Neutron-induced radiation detectors and radiation dosimetry systems containing the same

Abstract

Neutron-induced radiation detectors (NiRDs) and radiation dosimetry systems containing the same are described. Examples of NiRDs can include numerous components based on location, function, dimension, and / or constituent materials. A NiRD can be configured to directly measure an absorbed dose (or rate) of particular ionizing reaction products induced by neutrons. The NiRD can utilize various detection schemes for measuring the deposited energy of the reaction products, such as gaseous ionization, semiconducting, scintillating, or homogenous versions of these detection schemes. The NiRD can measure the deposited energy of the reaction products in forward emission, backward emission, or both. Examples of radiation dosimetry systems can include one or more NiRDs, a gamma ray detector, and readout electronics coupled to each of the radiation detectors. Example implementations of the radiation dosimetry systems are described for determining an equivalent dose (or rate) of a neutron beam generated for boron neutron capture therapy (BNCT).

Description

[0001]Attorney Docket No.069415 / 632275 / TLIFE.0006.WO NEUTRON-INDUCED RADIATION DETECTORS AND RADIATION DOSIMETRY SYSTEMS CONTAINING THE SAME CROSS-REFERENCE TO RELATED APPLICATIONS This application claims priority to U.S. Provisional Application Serial No.63 / 656,715, filed June 6, 2024, which is incorporated herein by reference in its entirety for all purposes. FIELD This specification relates generally to neutron-induced radiation detectors and radiation dosimetry systems containing the same. BACKGROUND Boron neutron capture therapy (BNCT) is a modality of treatment for a variety of types of cancer, including some of the most difficult types. BNCT is a technique that selectively aims to treat tumor cells while sparing the normal, healthy cells using a boron compound. The boron compound allows for efficient uptake by multiple cell types and selective drug accumulation at target sites, such as tumor cells. Boron loaded cells can be irradiated with neutrons, e.g., in the form of a neutron beam. The neutrons react with the boron to eradicate the tumor cells. Neutron beams for BNCT can be generated through various techniques. One such technique involves irradiation of a suitable neutron-generating target with a charged particle beam, such as a proton or deuteron beam. The charged particles react with nuclei in the target to emit a beam of neutrons having a characteristic energy spectrum (e.g., an epithermal energy spectrum) that can be delivered to the tumor cells. The total neutron dose absorbed by the tumor cells, as well as the healthy surrounding tissue, is a complex function of particle interactions at the cellular level. In addition to boron, these interactions result from neutron reactions with other chemical elements of the human body, such as nitrogen, hydrogen, carbon, and oxygen, which cumulatively make up about 99% of the mass percentage of soft tissue. For a BNCT treatment, these different neutron interactions have distinctive biological effects and are accounted for separately in the prescribed (or “equivalent”) neutron dose, which differs from the total absorbed dose that can be measured. 1 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO Traditionally, BNCT treatment planning systems are calibrated via neutron activation analysis. The prescribed neutron dose is calculated from the thermal neutron flux, which is inferred from the radioactivity of one or more metallic foils that have been activated by the source of neutrons. The fundamental assumption built into these neutron activation methods is that the neutron-induced dose components, e.g., the boron, nitrogen, and hydrogen dose components, can be accurately predicted based on the ability to predict the thermal neutron flux. However, these conventional treatment planning methods are limited and only capable of a coarse resolution over the approximately nine orders of magnitude in the neutron energy spectrum relevant to BNCT. Inaccuracies in the prescribed neutron dose calculated with these techniques can compromise BNCT treatment, such as reducing the effectiveness of tumor cell irradiation, increasing stochastic or deterministic health effects to the healthy tissue from neutron exposure, or all of the above. Hence, there is a need for improved systems, devices, and methods for accurately measuring, evaluating, and comparing the calculated and observed dose quantities for each individual neutron-induced dose component. SUMMARY This specification describes neutron-induced radiation detectors (NiRDs) and radiation dosimetry systems containing the same. Examples of radiation dosimetry systems including one or more NiRDs are described in the context of a boron neutron capture therapy (BNCT) system, where a radiation dosimetry system is configured to: (i) receive a neutron beam generated via irradiation of a neutron-generating target with a charged particle beam, and (ii) measure one or more neutron-induced dose components of the neutron beam. The radiation dosimetry system can also be configured to measure a gamma ray dose resulting from residual gamma rays in the neutron beam. The radiation dosimetry system can be used to precisely calibrate a treatment planning system that utilizes the neutron beam for BNCT. Although examples of BNCT applications are provided, the technologies described herein can also be deployed in other environments. Examples of which are described throughout this specification. In general, for a BNCT treatment, the total neutron dose absorbed by an organ or tissue is divided into four dose components that each have distinct biological effects and, therefore, are accounted for separately in the equivalent (or “prescribed”) dose. The four components include: (i) the gamma ray dose, (ii) the neutron-induced boron dose, (iii) the neutron-induced nitrogen 2 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO dose, and (iv) the neutron-induced hydrogen dose. These quantities are calculated individually within a treatment planning system and then summed with appropriate biological weighting factors to obtain the total equivalent dose. The treatment plan is evaluated to determine the limiting equivalent dose to the specific organ or tissue at risk, e.g., 8.5 sieverts (Sv) to the skin, and then prescribed in terms of a desired number of monitor units, which is a normalized unit of measure. The number of prescribed monitor units are delivered to the patient over an appropriate treatment time and can be observed in real-time during the BNCT treatment using a monitor (dosimetry) chamber calibrated to the monitor unit. This calibration ensures that the treatment planning system and the observed quantity from the monitor chamber produce the same result within an acceptable uncertainty bound. What is unique to BNCT, is that the prescribed neutron dose is an inferred quantity that differs from the (observable) absorbed dose that can be measured, which is generally not true for other modalities of treatment. Traditionally, during the commissioning phase, the treatment planning system is calibrated using a single scalar quantity – the thermal neutron flux. This is measured using metallic (e.g., gold) foil activation at a fixed depth in a tissue equivalent (or tissue approximate) material. The thermal neutron flux inferred from the metallic foil activation is compared to the calculated value in the treatment planning software and the observed output from the monitor chamber, thus correlating the three quantities. The fundamental assumption built into this method is that the neutron-induced dose components, e.g., the boron, nitrogen, and hydrogen components, can be accurately predicted based on the ability to predict the thermal neutron flux. This method also assumes that the sum of the calculated dose components and the absorbed dose observed in the monitor chamber are proportional. As a secondary check, the neutron energy spectrum can be unfolded with the metallic foil activation technique using many such foils. However, these foil activation methods are limited and only capable of a coarse resolution over the approximately nine orders of magnitude in the energy distribution relevant to BNCT, which can significantly increase the error and uncertainty in the treatment planning system. Inaccuracies in the treatment planning system can reduce the effectiveness of tumor irradiation, as well as increase stochastic or deterministic health effects to the healthy tissue. Among other aspects, this disclosure is aimed at improving the accuracy and precision of a BNCT treatment planning system, e.g., NeuCure, SERA, BDTPS, PHITS models, and other deterministic or Monte Carlo based treatment planning systems. The NiRDs and radiation 3 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO dosimetry systems disclosed herein can directly measure an absorbed dose (or an absorbed dose rate) of each neutron-induced dose component. Hence, the NiRDs and radiation dosimetry systems disclosed herein can be used to accurately calculate the equivalent dose (or the equivalent dose rate) of a neutron source to a specific organ or tissue in the human body. The predominant neutron-induced dose components depend, at least in part, on the energy spectrum of the neutron flux, which for BNCT applications is between about one (1) millielectronvolts (meV) and one (1) megaelectronvolt (MeV). Nevertheless, the NiRDs and radiation dosimetry systems can be configured for neutron energies above one (1) MeV, which can be relevant for other neutron sources and neutron-generating systems, e.g., fast neutron reactors (FNRs) and fusion reactors, that generate ultra-fast (e.g., relativistic) neutrons either intentionally or as byproducts. In these cases, the neutron-induced carbon and oxygen doses can play a significant (or dominate) role in the equivalent dose, e.g., due to inelastic neutron scattering and / or fast neutron capture. Moreover, neutron reactions with calcium and phosphorus can be non-negligible for hard tissue, e.g., bone matter and tooth enamel. For completeness, neutronics properties, other physical properties, and example materials for all non-trace elements of the human body are provided herein. The NiRDs and radiation dosimetry systems can be configured to directly measure the neutron-induced dose(s) resulting from neutron reactions with isotopes of any of these elements. As such, a NiRD can include various components based on location, function, dimension, and / or constituent materials. At a high-level, the NiRD generally includes an irradiator material and an active volume coupled to the irradiator material. The NiRD may also include a gamma ray shield enclosing the irradiator material and active volume. The irradiator material includes one or more isotopes of one or more elements of interest, where each isotope corresponds to one or more neutron-induced dose components being characterized by the NiRD. In general, the elements of interest include boron, gadolinium, and the non-trace elements of the human body, corresponding to the primary nuclear reaction channels in neutron capture therapy, e.g., BNCT, and other sources of neutron exposure. Each isotope reacts with neutrons within a particular energy spectrum via neutron capture and / or neutron scattering to produce ionizing reaction products. The reaction products can include gamma rays and / or one or more species of charged particles, where the precise species of ionizing particle(s) depends on the particular neutron reaction and isotope. For example, the 4 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO reaction products include alpha particles and lithium-7 ions for the10B(n,α)7Li reaction, protons for the14N(n,p)14C and1H(n,n)1H reactions, and gamma rays for the1H(n,ɣ)2H reaction. The irradiator material is to emit the reaction products into the active volume, depositing their energy The active volume is defined by an enclosure that is filled, at least partially, with an absorbing medium that is sensitive to the ionizing reaction products. The enclosure typically includes an entrance window that is transparent to the ionizing reaction products. For example, the thickness and material composition of the entrance window can be chosen based on the hardness (or penetration ability) of the reaction products. A layer of the irradiator material can be positioned on the entrance window to emit the reaction products therethrough, injecting the reaction products into the absorbing medium. In some implementations, the layer of the irradiator material can be positioned directly on the absorbing medium, or directly on an electrode interfacing with the absorbing medium. The absorbing medium is configured to absorb kinetic energy of the reaction products and produce secondary particles in response, e.g., via distinct mechanisms of linear energy transfer (LET). The secondary particles are used as the means for detection and measurement of the energy deposited within the active volume. For example, the secondary particles can either be directly measured as electric current pulses, or converted to electric current pulses (e.g., via the photoelectric effect), which are output by the active volume in response to ionizing events. In some implementations, the active volume is configured in pulse mode such that the total charged collected in an electric current pulse is proportional to the energy deposited by a single quantum of ionizing radiation, providing linear (or proportional) detection of the absorbed dose and dose rate. Pulse mode is particularly well- suited for neutron-induced radiation dosimetry and spectroscopy, allowing precise measurement of the species, counts, count rates, deposited energies, absorbed doses, absorbed dose rates, and reaction rates of the neutron-induced reaction products under a suitable calibration of the radiation dosimetry system. For example, the radiation dosimetry system can be calibrated with one or more radioactive, monoenergetic isotopes of known intensity and energy such as americium-241 (Am-241) and / or neptunium-237 (Np-237). The gamma ray shield is configured to absorb, scatter, or otherwise mitigate background gamma rays and / or residual gamma rays produced by a neutron source (e.g., a neutron- generating target). The generation of free neutrons is typically accompanied by a non-negligible 5 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO dose of gamma rays which can be independently measured by the radiation dosimetry system using a gamma ray detector. However, these residual gamma rays can ionize the irradiator material and / or the absorbing medium, contaminating the neutron-induced dose measurements. The gamma ray shield can increase the sensitivity of the NiRD to incident neutrons without significantly altering the neutron energy spectrum. For example, the gamma ray shield can reduce dark counts, deadtimes, and / or detector pileup resulting from residual gamma ray absorption within the active volume. In general, the configuration of the irradiator material, the active volume, and the absorbing medium residing therein depends on the detection scheme employed by the NiRD, and whether the reaction products are detected in forward emission, backward emission, or both. For example, in some implementations, the NiRD includes two separate active volumes with the irradiator material positioned therebetween, e.g., in a “sandwich” configuration, such that the energy of the reaction products is fully absorbed within the NiRD. In these cases, the two active volumes may be electrically connected to each other in parallel, summing the electric current pulses output from each. Thus, for example, the NiRD can measure the forward and backward emitting reaction products of a neutron capture reaction simultaneously, e.g., a forward (or backward) emitting alpha particle and a backward (or forward) emitting lithium-7 ion for the10B(n,α)7Li reaction. The NiRD can utilize different detection schemes for measuring the energy deposited by the reaction products, such as gaseous ionization detection schemes, semiconducting detection schemes, scintillating detection schemes, or homogenous versions of these detection schemes. For gaseous ionization NiRDs (or “gas-NiRDs”), the active volume is defined by a chamber (e.g., a sealed high- or low-pressure chamber), the absorbing medium is a fill gas, and the secondary particles are ion-pairs. Examples of gas-NiRDs utilizing cylindrical, spherical, and planar geometries for detecting neutron-induced reaction products in forward, backward, and dual (or isotropic) emission are provided herein. For semiconductor NiRDs (or “semi-NiRDs”), the active volume is defined by an insulating shell (e.g., a dielectric cladding), the absorbing medium is a semiconductor material, and the secondary particles are electron-hole pairs. Examples of semi-NiRDs utilizing different impurity doping and electrical contact designs for detecting neutron-induced reaction products in forward, backward, and dual emission are provided herein. 6 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO For scintillation NiRDs (or “scin-NiRDs”), the active volume is defined by a reflecting cavity (e.g., a silvered film or silvered glass reflector), the absorbing medium is a scintillator, and the secondary particles are scintillation photons. Examples of scin-NiRDs utilizing different photodetectors for converting scintillation photons and detecting neutron-induced reaction products in forward, backward, and dual emission are provided herein. For homogenous NiRDs (or “homo-NiRDs”), the irradiator material and absorbing medium are the same material, allowing simultaneous generation and detection of neutron- induced reaction products in the same bulk medium. Homo-NiRDs involve fill gases, semiconductor materials, and scintillators that include one or more isotopes of one or more of the elements of interest, while also being sensitive to the reaction products of interest. Examples of homo-NiRDs utilizing gaseous ionization, semiconducting, and scintillating detection schemes for homogenous, bulk detection of neutron-induced reaction products are provided herein. Examples of other types of NiRDs for bulk detection such as microstructured semi-NiRDs and microstructured scin-NiRDs are also provided herein. In the described examples, a radiation dosimetry system can include one or more NiRDs, a gamma ray detector, and readout electronics coupled to each of the radiation detectors for performing various radiation dosimetry and spectrometry measurements, such as absorbed dose and absorbed dose rate measurements. Each NiRD can include an irradiator material composed of a different material for detecting one or more neutron-induced dose components of a neutron source. Each NiRD can also employ a different detection scheme this is calibrated and optimized for its particular dose component(s). The gamma ray detector can detect a gamma ray dose from residual gamma rays produced by the neutron source. By combining the measurements from each radiation detector, the readout electronics can calculate the equivalent dose (or equivalent dose rate) to a particular organ or tissue with considerable precision, e.g., for use in calibrating a BNCT treatment planning system. The readout electronics can be electrically coupled to each of the radiation detectors via a switchboard or other appropriate circuitry. The readout electronics can include various electronic components such as a bias supply, a preamplifier, an amplifier, and a multichannel analyzer (MCA) for performing pulse-height (energy) spectroscopy. The readout electronics can also be configured for other types of measurements such as counting and timing measurements. The bias supply can be configured to apply a suitable bias voltage to a radiation detector during active 7 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO irradiation of the radiation detector, e.g., to configure the radiation detector for pulse mode. The preamplifier can be configured to receive an electric current pulse output from the radiation detector and convert it into a voltage pulse, e.g., with a characteristic pulse height and shape. The amplifier can be configured to amplify (e.g., linearly amplify) the voltage pulse, which can be subsequently processed and analyzed by the MCA. The MCA can apply signal processing techniques, such as pulse-height and pulse-shape discrimination, to a series of such output voltage pulses to perform various functions, e.g., determining the species of ionizing radiation generating the pulses, quantifying the energy deposited by the ionizing radiation, determining the accumulated or current dose rate of the ionizing radiation, determining the reaction rates, eliminating dark counts, dark (or leakage) currents, and / or amplifier noise, among other functions. Numerous examples of NiRDs and radiation dosimetry systems are disclosed that perform some, or all, of these abovementioned functions. Numerous exemplary materials are also disclosed with capabilities to perform one or more of these functions, or have capabilities that enable the NiRDs and radiation dosimetry systems to perform such functions. Particular embodiments of the subject matter described in this specification can be implemented so as to realize one or more of the following advantages. The NiRDs and radiation dosimetry systems disclosed herein allow for direct measurement, evaluation, and comparison of each of the neutron-induced dose components of a neutron source (e.g., a neutron-generating target) having a generic energy spectrum. For BNCT applications, e.g., that use a beam of epithermal neutrons for treatment, this can vastly improve the accuracy and uncertainty of a treatment planning system and the cross-calibration of a monitor (dosimetry) chamber. The disclosed NiRDs and radiation dosimetry systems provide active, real-time measurements and, therefore, can substantially reduce the procedural time involved for BNCT commissioning exercises compared to conventional methods, e.g., metallic foil activation methods. The disclosed NiRDs and radiation dosimetry systems can be used for neutron beam commissioning during the initial phase, as well as during BNCT treatment to actively monitor the absorbed neutron dose and / or dose rate, thereby eliminating the need to cross-calibrate a monitor chamber. The disclosed NiRDs and radiation dosimetry systems can also characterize neutron-induced doses in energy ranges beyond what is typical for BNCT, e.g., neutron-generating systems involving ultra-fast neutrons such as FNRs and fusion reactors. 8 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO Implementations of the NiRDs and radiation dosimetry systems disclosed herein can also be employed in safety protocols and neutron monitoring procedures, e.g., when personnel are unavoidably (or unintentionally) exposed to non-negligible doses of neutrons. For example, a radiation dosimetry system including an array of NiRDs can be implemented as a personalized multi-component dosimeter, e.g., a handheld or pocket dosimeter, that allows a user (or wearer) to accurately measure the equivalent neutron dose (or dose rate) resulting from neutron exposure. The radiation dosimetry system can track the accumulated dose over seconds, minutes, hours, weeks, months, years, etc. to mitigate stochastic health effects from such exposure, e.g., development of radiation-induced cancer. Alternatively, or in addition, the radiation dosimetry system can measure the current dose rate at the point of exposure to mitigate deterministic health effects, e.g., development of acute radiation syndrome. These applications can be useful for clinicians that administer BNCT (and other forms of nuclear medicine), as well as nuclear chemists, nuclear reactor operators, and other occupations or situations involving significant neutron exposure. The details of one or more embodiments of the subject matter of this specification are set forth in the accompanying drawings and the description below. Other features, aspects, and advantages of the subject matter will become apparent from the description, the drawings, and the claims. BRIEF DESCRIPTION OF DRAWINGS The details of the subject matter set forth herein, both as to its structure and operation, may be apparent by study of the accompanying figures, in which like reference numerals refer to like parts. The components in the figures are not necessarily to scale, emphasis instead being placed upon illustrating the principles of the subject matter. Moreover, all illustrations are intended to convey concepts, where relative sizes, shapes and other detailed attributes may be illustrated schematically rather than literally or precisely. FIG.1A is a schematic view depicting an example of a neutron beam system configured for use in boron neutron capture therapy (BNCT). FIG.1B is a schematic view depicting a more detailed illustration of the beamline integrated into the neutron beam system of FIG.1A. FIG.2A is a perspective view depicting an example of a neutron-generating target. 9 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIG.2B is a side view depicting an example of a target assembly for housing a neutron- generating target. FIG.2C is a cross-sectional side view depicting an example of a target assembly for housing a neutron-generating target. FIG.3A is a schematic diagram depicting an example of a radiation dosimetry system including a neutron-induced radiation detector (NiRD) and readout electronics electrically coupled to the NiRD. FIG.3B is a schematic diagram depicting an example configuration of the readout electronics in FIG.3A. FIG.3C is a circuit diagram of an example circuit topology for the radiation dosimetry system in FIG.3A. FIG.3D is an equivalent circuit diagram of an analog component of the circuit topology in FIG.3C. FIG.3E are plots of an example process for pulse-shape discrimination performed by a multichannel analyzer (MCA). FIG.3F are plots of an example process for pulse-height discrimination performed by an MCA. FIG.3G is an example histogram of counts versus channel (pulse height) for an irradiator material composed of boron nitride. FIG.3H is an example histogram of counts versus channel (pulse height) for an irradiator material composed of polyethylene. FIG.4A is a schematic diagram depicting an example of a multi-component radiation dosimetry system including multiple NiRDs, a gamma ray detector, and readout electronics electrically coupled to each of the radiation detectors. FIG.4B is a circuit diagram of an example switchboard for the multi-component radiation dosimetry system in FIG.4A. FIG.5A is a flow diagram of an example process for performing neutron-induced radiation dosimetry using the radiation dosimetry system in FIGs.3A-3D. FIG.5B is a flow diagram of an example process for performing multi-component radiation dosimetry using the radiation dosimetry system in FIGs.4A-4B. 10 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIG.5C is a flow diagram of an example process for performing neutron-induced radiation dosimetry. FIG.6A is a cross-sectional side view of an example irradiator material depicting a neutron capture event. FIG.6B is a cross-sectional side view of an example irradiator material depicting a neutron scattering event. FIGs.7A-7B are plots of neutron cross-section and reaction product energies for the10B(n,α)7Li reaction. FIGs.8A-8B are plots of neutron cross-section and reaction product energies for the14N(n,p)14C reaction. FIGs.9A-9B are plots of neutron cross-section and reaction product energies for the1H(n,n)1H and1H(n,ɣ)2H reactions. of neutron cross-section for12C(n,n)12C reactions and ultra-fast neutron for carbon-12. FIG.10B is a plot of neutron cross-section for16O(n,n)16O reactions and ultra-fast neutron charged reactions for oxygen-16. FIG.10C is a plot of stopping power versus energy for neutron-induced reaction products in boron nitride and polyethylene. FIGs.11A-11B are various views depicting an example of a gaseous ionization NiRD (“gas-NiRD”). FIG.12A is a cross-sectional side view depicting an example of a gas-NiRD configured for detection in backward emission. FIG.12B is a cross-sectional side view depicting an example of a gas-NiRD configured for detection in both forward and backward emission. FIGs.13A-13B are various views depicting an example of a gas-NiRD having a spherical configuration. FIGs.14A-14C are cross-sectional side views depicting additional examples of gas- NiRDs having planar configurations. FIGs.15A-15B are various views depicting an example of a semiconductor NiRD (“semi-NiRD”). 11 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIG.16A is a cross-sectional side view depicting an example of a semi-NiRD configured for detection in backward emission. FIG.16B is a cross-sectional side view depicting an example of a semi-NiRD configured for detection in both forward and backward emission. FIGs.16C-16D are cross-sectional side views depicting additional examples of semi- NiRDs. FIGs.17A-17B are various views depicting an example of a semi-NiRD having a microstructured configuration. FIG.18A is a cross-sectional side view depicting an example of an active volume for a semi-NiRD with two Schottky contacts. FIG.18B is an example current-voltage curve of an active volume with two Schottky contacts. FIG.19A is a cross-sectional side view depicting an example of an active volume for a semi-NiRD with two ohmic contacts. FIG.19B is an example current-voltage curve of an active volume with two ohmic contacts. FIG.20A is a cross-sectional side view depicting an example of an active volume for a semi-NiRD with a Schottky contact and an ohmic contact. FIG.20B is an example current-voltage curve of an active volume with a Schottky contact and an ohmic contact. FIGs.21A-21F are cross-sectional side views depicting additional examples of active volumes for a semi-NiRD. FIGs.22A-22B are various views depicting an example of a scintillation NiRD (“scin- NiRD”). FIG.23A is a cross-sectional side view depicting an example of a scin-NiRD configured for detection in backward emission. FIG.23B is a cross-sectional side view depicting an example of a scin-NiRD configured for detection in both forward and backward emission. FIGs.23C-23D are cross-sectional side views depicting additional examples of scni- NiRDs. 12 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIGs.24A-24B are various views depicting an example of a scin-NiRD having a microstructured configuration. FIGs.25A-25C are cross-sectional side views depicting examples of photodetectors for a scin-NiRD. FIGs.26A-26C are cross-sectional side views depicting examples of homogenous NiRDs (“homo-NiRDs”). DETAILED DESCRIPTION Before the present subject matter is described in detail, it is to be understood that this disclosure is not limited to the particular embodiments described, as such may, of course, vary. The terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting, since the scope of the present disclosure will be limited only by the appended claims. The term “particle” is used broadly herein and, unless otherwise limited, can be used to refer to an electron, a proton (or H+ ion), or a neutron, as well as a species having more than one electron, proton, and / or neutron, e.g., other ions, atoms, and molecules. Examples of neutron-induced radiation detectors (NiRDs) and radiation dosimetry systems containing the same are described herein, which can be used to characterize one or more neutron-induced dose components of neutron radiation generated by a neutron source, as well as any residual gamma radiation produced by the neutron source. The example NiRDs and radiation dosimetry systems described herein can be used for any type of neutron source in which neutron- induced radiation dosimetry and / or spectrometry is desired. One example is a neutron beam system, e.g., including a reactor or a particle accelerator, for generation of a neutron beam for use in boron neutron capture therapy (BNCT). BNCT utilizes a beam of epithermal neutrons, e.g., with an epithermal energy spectrum between one (1) electronvolt (eV) and 30 kiloelectronvolts (keV), for cancer treatment. In BNCT, the neutrons can be generated from nuclear reactions of charged particles, such as protons or deuterons, colliding with either a beryllium or lithium target. The neutrons can then be administered to a patient, irradiating boron loaded tumor cells. Before, during, and / or after treatment, the NiRDs and radiation dosimetry systems can be used to quantify an equivalent dose absorbed by the tumor cells, as well as the dose absorbed by the surrounding healthy tissue, for devising an appropriate treatment plan that 13 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO maximizes the efficacy of BNCT while minimizing stochastic and / or deterministic health effects to the healthy tissue. For ease of description, this specification is organized as follows. Section I describes examples of neutron beam systems that can be configured for BNCT. Section II describes examples of radiation dosimetry systems including one or more NiRDs for performing neutron- induced radiation dosimetry and spectroscopy. Section III describes examples of gaseous ionization NiRDs (or “gas-NiRDs”) utilizing gaseous ionization detection schemes. Section IV describes examples of semiconductor NiRDs (or “semi-NiRDs”) utilizing semiconducting detection schemes. Section V describes examples of scintillation NiRDs (or “scin-NiRDs”) utilizing scintillating detection schemes. Section VI describes examples of homogeneous NiRDs (or “homo-NiRDs”) utilizing homogenous versions of these abovementioned detection schemes. Examples of the NiRDs and radiation dosimetry systems described herein are not intended to be viewed in isolation from each other. All features, elements, components, and functions described with respect to any example NiRD or radiation dosimetry system provided herein are freely combinable and substitutable with those from any other example NiRD and radiation dosimetry system unless explicitly stated otherwise. Section I: Examples of Neutron Beam Systems For ease of description, examples of the NiRDs and radiation dosimetry systems described herein will be done so in the context of a neutron beam system that converts a charged particle beam into a neutron beam for use in BNCT. A radiation dosimetry system including one or more NiRDs can be configured to accurately measure one or more neutron-induced dose components (e.g., including a neutron-induced boron, nitrogen, and / or hydrogen dose) of the neutron beam upon irradiation of a patient. The radiation dosimetry system can, therefore, accurately predict the equivalent (or “prescribed”) dose to the patient that accounts for the distinct biological effects of each neutron-induced dose component. However, the NiRDs and radiation dosimetry systems are not limited to BNCT applications. The NiRDs and radiation dosimetry systems described herein can be used in other neutron-generating systems, even those outside of BNCT applications utilizing different energy ranges, e.g., particle colliders, fission or fusion reactors, fast neutron reactor (FNRs), irradiation tests, isotope production, transmutation, crystallography, neutron imaging, among others. For example, the NiRDs and radiation dosimetry systems described herein can be configured for experiments in particle physics, 14 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO nuclear energy production, materials science, life science, chemistry, and molecular biology, among other applications. FIG.1A illustrates a schematic view of an example beam system 100 configured as a neutron beam system for use in BNCT. At a high-level, the beam system 100 is configured to create a charged particle beam 61 and propagate it toward a neutron-generating target 60 to generate a neutron beam 70. The neutron beam 70 is then directed towards a radiation dosimetry system 150 to be characterized for BNCT treatment of a patient, e.g., to determine an equivalent (or “prescribed”) neutron dose for the patient. The beam system 100 includes a charged particle source 20, a low-energy beamline (LEBL) 30, a particle accelerator 40, and a high-energy beamline (HEBL) 50. The charged particle source 20 is configured to generate the charged particle beam 61, which is output to the LEBL 30. The LEBL 30 is configured to transport the charged beam 61 from the charged particle source 20 to the accelerator 40. The accelerator 40 is configured to accelerate the charged particle beam 61 to a higher energy. The HEBL 50 extends from the accelerator 40 to the target 60 housed within a target assembly 65 portion of the HEBL 50, see FIGs.2A-2C for example. The HEBL 50 transfers the charged particle beam 61 from an output of the accelerator 40 to the target 60, where it irradiates the target 60 and is converted to the neutron beam 70. FIG.1B is a schematic view depicting a detailed illustration of the beamline integrated into the neutron beam system 100 of FIG.1A. Here, the beam system 100 includes a pre- accelerator system 26 forming at least a portion of the LEBL 30, where the pre-accelerator system 26 serves as a charged particle beam injector. The beam system 100 includes a high- voltage tandem accelerator 40 coupled to the LEBL 30, and the HEBL 50 extending from the tandem accelerator 40 to a neutron-generating target 60, as described with reference to FIG.1A. The LEBL 30 transfers a negative ion beam (e.g., hydrogen ions (H-)) from an ion source 20, through the pre-accelerator 26 which boosts the energy level of the ion beam and converges the ion beam, to an input (e.g., an input aperture) of the accelerator 40. The accelerator 40 is powered by a high-voltage power supply 42 coupled thereto. The accelerator 40 includes a vacuum tank, a charge-exchange tube, accelerating electrodes, and a high-voltage feedthrough. The accelerator 40 can, in some implementations, accelerate a hydrogen ion beam to produce a proton beam with an energy generally equal to twice the voltage applied to the accelerating electrodes positioned within the accelerator 40. The energy level of the proton beam can be 15 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO achieved by accelerating the beam of negatively charged hydrogen ions from the input of the accelerator 40 to the innermost high-potential electrode, stripping two electrons from each hydrogen ion, and then accelerating the resulting protons downstream by the same voltages encountered in reverse order. The HEBL 50 transfers the proton beam from the output of the accelerator 40 to the neutron-generating target 60 positioned at the end of a branch 71 of the beamline extending into a patient treatment room. The beam system 100 is configured to direct the proton beam to one or more targets 60 and associated target areas. In some implementations, the HEBL 50 includes multiple (e.g., three) branches 71, 81, and 91 extending into multiple different patient treatment rooms, with each branch terminating at a respective neutron-generating target 60. The HEBL 50 includes a pumping chamber 51, quadrupole magnets 52 and 72 to prevent de-focusing of the proton beam, dipole (or bending) magnets 56 and 58 to steer the proton beam towards one or more targets 60, beam correctors 53, diagnostics such as current monitors 54 and 76, a fast beam position monitor 55 section, and a scanning magnet 74 for branch 71. Branches 81 and 91 can contain components similar to branch 71. The design of the HEBL 50 depends on the configuration of the treatment facility (e.g., a single-story treatment facility, a two-story treatment facility, and the like). The proton beam can be delivered to the target 60, e.g., positioned in (or proximate to) a treatment room containing a radiation dosimetry system 150, with the use of the dipole magnet 56. The quadrupole magnets 72 can be configured to focus the proton beam to a certain size at the target 60. The proton beam can pass one or more scanning magnets 74, which provide lateral movement of the proton beam onto the target 60’s surface in a desired pattern (e.g., spiral, curved, stepped in rows and columns, combinations thereof, and others). The beam lateral movement can enable generation of smooth and even time-averaged distribution of the proton beam on the target 60, preventing overheating of the target 60 and making the neutron generation as uniform as possible within the target 60, e.g., in a neutron-generating layer 121 of the target 60 as shown in FIG.2A. The scanning magnets 74 can be configured to direct the proton beam to a current monitor 76, which measures beam current. The beam current value can be used to operate a safety interlock. The target assembly 65 containing the target 60 can be physically separated from the high-energy beamline volume with a valve 77. A function of the valve 77 is to separate the vacuum volume of the beamline from the target 60 during removal of a used target and 16 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO loading of a new target. In some implementations, instead of being bent by 90 degrees by a dipole magnet 56, the proton beam can be directed straight to one or more quadrupole magnets 52 located in the horizontal beamline. The proton beam could be bent by another bending magnet 58 to a preset angle, depending on a setting requirement (e.g., location of a patient or a treatment room configuration). In some implementations, the bending magnet 58 can be arranged at a branch point in the beamline and can be configured to direct the proton beam in one of two directions for two different treatment rooms located on the same floor of a medical facility. The beam system 100 as described with respect to FIGs.1A-1B is an example configuration that can be used to generate charged particle and neutron beams. Different configurations of the beam system 100 can utilize accelerators other than electrostatic tandem accelerators, as well as targets 60 that are either fixed or rotating. Examples of the radiation dosimetry systems 150 described herein are not limited to use with any one type of beam system, neutron beam generating system, or neutron-generating system. The radiation dosimetry system 150 can be configured to accurately measure one or more neutron-induced dose components of any neutron source such as spontaneous fission sources, radioactive isotopes, plasma focus and pinch devices, inertial electrostatic confinement devices, high-energy bremsstrahlung photoneutron and photofission systems, high-energy particle accelerators, nuclear fission reactors, nuclear fusion reactors, among others. For example, the radiation dosimetry system 150 can be used to characterize dosages from neutron sources for applications in physics, engineering, medicine, petroleum exploration, biology, chemistry, and nuclear power. FIG.2A is a perspective view of an example neutron-generating target 60. In this example, the target 60 has a neutron-generating layer 121 with a charged particle receiving face surface 122. The neutron-generating layer 121 is positioned on (or in proximity to) a substrate 123. In some cases, the neutron-generating layer 121 is covered with one or more additional layers for protection. The neutron-generating layer 121 can also have one or more underlying layers between the layer 121 and the substrate 123, e.g., to resist blister formation. A charged particle beam, such as a proton or deuteron beam, incident upon the receiving face surface 122, passes into the target 60 and causes the neutron-generating layer 121 to undergo a reaction that generates neutrons. This is the7Li(p,n)7Be nuclear reaction in the case where the neutron- generating layer 121 is composed of lithium-7 and the charged particle beam is a proton beam. The neutron-generating layer 121 can alternatively be beryllium-9, and neutrons can be 17 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO generated with a proton beam (9Be(p,n)9B) or deuteron beam (9Be(d,n)10B) at different energies. The substrate 123 can be a material with excellent thermal conductivity, such as copper or aluminum, to assist in removal of the heat generated by the reactions. FIG.2B is a side view of an example target assembly 65 that can form a terminal portion of the HEBL 50. A target 60 (not shown) can be contained within the assembly 65 at or near an end 67 of the assembly 65. The charged particle beam enters the assembly 65 at an end 66 and travels to the opposite end 67 where it impacts the target 60. Various cooling channels 68 are routed to and from the end 67 for the insertion and removal of coolant used to regulate the temperature of the target 60 during use. Numerous sensors can be included in the assembly 65 to monitor the temperature and radioactivity of (and around) the assembly 65. Also shown is the valve 77 in the example form of a gate valve. FIG.2C is a cross-sectional view of an example target assembly 65 omitting components such as the valve 77, coolant channels, and sensor connections for clarity. A sidewall 62 has a tubular shape and contains an interior space 64 at a vacuum or near vacuum level. A target 60 is positioned at the end 67 and held in place by an end cap 63. Variations of this construction are possible, such as with the target 60 surrounded by a sidewall 62 of the assembly 65. A charged particle beam 61 is directed through an interior space 64 of the assembly 65 and scanned across the target 60 by the scanning magnet 74 located upstream on the HEBL 50 (not shown). Neutrons produced by the target 60 will be emitted at some level in virtually all directions from the target 60, but the majority of the neutrons will be emitted in a disperse but generally forward directed path. This is depicted here as a neutron beam 70 (e.g., in raw form). Section II: Examples of Radiation Dosimetry Systems In the fields of health physics and radiation protection, radiation dosimetry is the measurement, calculation, and assessment of the ionizing radiation dose absorbed by an object, typically the human body. Ionizing radiation can cause stochastic (or probabilistic) health effects when tissue is exposed to absorbed doses above certain thresholds. Examples of stochastic health effects include damage to cellular DNA, onset cell death, and radiation-induced cancer. Absorbed doses may be accumulated incrementally over an extended period time, e.g., on the order of months or years, where each incident of an absorbed dose increases the probability of a stochastic health effect. Ionizing radiation can also cause deterministic health effects when tissue is exposed to absorbed dose rates above certain thresholds. Examples of deterministic health 18 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO effects include radiation burns, immediate cellular damage, and acute radiation syndrome. Deterministic health effects generally occur immediately, or almost immediately, e.g., within seconds, minutes, or hours of exposure. In radiation dosimetry, absorbed doses and dose rates can be quantified taking into account the type of ionizing radiation and tissue being exposed, as different species of ionizing particles can be more or less damaging to different types of tissue. For example, dose-for-dose, neutrons are generally more biologically hazardous than photons. How these dosages are quantified will be explained more thoroughly below. In general, ionizing radiation can be classified as directly or indirectly ionizing. Directly ionizing radiation includes charged particles such as protons (hydrogen ions), alpha particles (helium-4 ions), beta particles (high-energy electrons and positrons), and other charged nuclei that can directly ionize atoms in a medium through the Coulomb force, if they have sufficient kinetic energy. Indirectly ionizing radiation includes neutral particles such as photons and neutrons that ionize atoms through so-called “secondary ionization”. Secondary ionization refers to the liberation of high-energy charged particles in a medium via interaction with the neutral particles – the charged particles subsequently ionize atoms in the medium through the Coulomb force. High-energy photons, such as x-rays and gamma rays, indirectly ionize atoms through the electromagnetic interaction, which generally involves absorption or scattering of a photon to produce one or more beta particles that subsequently ionize the medium. For example, for incident energies between about 100 eV and one (1) MeV, photons mainly ionize atoms indirectly through the photoelectric or Compton effect, both of which cause the ejection of a beta particle. For incident energies above about one (1) MeV, photons can also ionize atoms indirectly through pair-production, the generation of an electron-positron pair. As used herein, photons are referred to as “gamma rays” if they are produced by a nuclear reaction, subatomic particle decay, or radioactive decay within a nucleus. Photons are referred to as “x-rays” if they are produced outside the nucleus, e.g., in an x-ray vacuum tube. The generic term “photon” is used to describe both. Conversely, neutrons indirectly ionize atoms through the nuclear (strong) interaction, which generally involves various neutron capture and neutron scattering reactions. Fissile materials that undergo nuclear fission via neutron reactions are not described in detail herein. Unlike photons, neutrons can indirectly ionize matter even at low, thermal energies (e.g., below 19 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO one (1) eV) due to neutron capture processes. In a neutron capture event (e.g., FIG.6A), an isotope absorbs an incident neutron, transmuting and exciting the isotope. The transmuted isotope then decays with the emission of reaction products that can include gamma rays and / or one or more species of charged particles. The reaction products have kinetic energies corresponding to the sum total of the incident neutron’s kinetic energy and the released rest energy (Q-value) of the reaction. The gamma rays and charged particles then indirectly and directly ionize atoms as described above. In a neutron scattering event (e.g., FIG.6B), an isotope scatters an incident neutron, either elastically or inelastically, and recoils in response due to the transfer of kinetic energy. If the isotope is charged upon scattering, e.g., via loss of one or more electrons, the recoiling isotope can directly ionize atoms as described above. In the case of inelastic scattering, the recoiling isotope is also excited. The excited isotope then decays with the emission of ionizing reaction products, e.g., including gamma rays and / or one or more species of charged particles. In radiation dosimetry, the absorbed dose (^^^ is a measure of the energy deposited in matter by ionizing radiation, per unit mass. The absorbed dose has units of gray (Gy), defined as one Joule of energy absorbed per kilogram of matter. In the limit of infinitesimal spatial extent, the absorbed dose is defined as the ratio of the mean energy ^^ deposited by ionizing radiation(from all sources) in an item ^^^^ having a mass distribution ^^ within a volume ^^ of the item:d^^ 1 d^^ ^^ ൌ(1.1) ൌ . ^^ ^^ ൌ d^^ / d^^ is the energy deposited in the item per unit volume. At the molecular level, the absorbed dose ^^ ൌ^^^^^,^^, ^^^ is generally a complex, inhomogeneous quantity as it depends on several factors suchas the energy and spatial distribution of the incident radiation, the kinetics of the charged particles liberated, the local mass and atomic distribution of the item, among other properties. As this disclosure is concerned with gamma and neutron radiation, the absorbed doses described herein are due to indirectly ionizing radiation. For indirectly ionizing radiation in charged particle equilibrium (CPE), the absorbed dose is equivalent to the kerma (“Kinetic Energy Released per unit Mass”). CPE holds for indirectly ionizing radiation given the following conditions are satisfied: a. The atomic composition of the item is homogeneous. 20 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO b. The mass density of the item is homogeneous. c. There exists a uniform field of indirectly ionizing radiation through the item. d. No inhomogeneous electric or magnetic fields are present within the item. For ease of description, it is assumed that the item being irradiated is in CPE which simplifies the analysis, e.g., facilitating analytic (or empirical) expressions of the absorbed dose in the item. CPE is generally satisfied, at least approximately, for a relatively homogenous item (e.g., a specific organ or tissue) irradiated with a neutron source having a relatively uniform flux over the item (e.g., a neutron beam 70). Note, it is possible for CPE to exist in an item without satisfying all the above conditions under certain geometrical constraints. Given the constraints of CPE, it is often more practical to work with the average dose(^^்^ absorbed by the item which is a reliable metric ^^ ^ ^^் under such conditions. The averagedose is calculated by mass-averaging over the total volume ^^ of the item:^^^^^d^^ ^^ 1 ^^ (1.2) ^^் ்்ൌ ൌ ൌ . ^^ Here, ^^்is the mean of the item, and ^^் ൌ ^^் / ^^ is the (homogenous) mass density of the item. The average dose mayalso be referred to as the mean specific energy deposited in the item. The average dose is agnostic to the type of indirectly ionizing radiation absorbed by the item as it a measure of the total energy deposited. Nevertheless, since the mean deposited energy from photons ^^்,ఊandneutrons ^^்,^ sums linearly ^^் ൌ ^^்,ఊ ^ ^^்,^, the average dose can be separated into an averagephoton dose ^^்,ఊand an average neutron dose ^^்,^absorbed by the item as: ^^் ൌ ^^்,ఊ ^ ^^்,^. (1.3)Typically, the absorbed doses are calculated by integrating the absorbed dose rates over an exposure time. For example, a deterministic or Monte Carlo solver can be used to generate an estimate of the energy-dependent photon and neutron fluxes. This quantity can then be multiplied with an energy-dependent response function (e.g., an energy-dependent absorption coefficient or neutron cross-section) to determine the absorbed dose rate for the item of interest. This product is then integrated or summed over the photon or neutron energy (and the exposure time) to determine the absorbed dose. In this case, the absorbed photon dose rate (^^^ ்,ఊ) after integrating over the incidentphoton energy can be expressed as: 21 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO d^^ ^^ ൫^^ ൯ (1.4) ^^^ ൌ ்,ఊ ൌ ^ ்,^୬ ఊ்,ఊ Φఊ൫^^ ൯^^ d^^ . d^^ఊ^^ఊ ఊ்^^ is the photon energy-absorption Conversely, the average neutron dose is the sum of each neutron-induced dose component (^^்,^^) resulting from a particular type of neutron reaction (^^) with a particular isotope (^^) in the item. Energy deposition by neutrons is generally more complicated than photons due to the number of nuclear reaction channels present. For example, hydrogen-1 has a (relatively) high probability of both neutron capture and scattering reactions which produce different reaction products and, therefore, different means of energy deposition. In terms of theabsorbed neutron dose rate (^^^ ்,^), after integrating over the incident neutron energy, this can beexpressed as: d^^ ^^ ^^^(1 ்,^ ൌ ்,^ ൌ ^ ^^^ ்,^^ ൌ ^ ்,^.5) ^Φ^^^^^^^^^^^^^^^^^^^^^^^^d^^^. ^^ (homogeneous) atom density of the isotope in the item, ^^^^is the neutron cross-section for the isotope and reaction, ^^^^is the average energy deposited in the item due to the reaction with theisotope, and ^^^ is the incident neutron energy. Note, the ratio ^^்,^ / ^^் ൌ ^^்,^^^^ / ^^^ can also beexpressed in terms of the mass fraction ^^்,^of the isotope in the item, the molar mass ^^^of the isotope, and Avogadro’s number ^^^. The average energy deposited ^^^^depends on whether the reaction is a neutron capture or neutron scattering reaction. For neuron capture reactions, the average energy deposited in theitem is ^^^^^^^^^ ൌ ^^^ ^ ^^^^, where ^^^^ is the Q-value of the reaction with the isotopecorresponding to released ^^^^ ^ 0 or absorbed ^^^^ ^ 0 rest energy. For elastic neutron scatteringreactions, the average energy deposited in the item is equal to the average recoiling energy of theisotope ^^^^^^^^^ ൌ ^^^ െ ^^ᇱ^ ^ 2^^^^^^^^^ / ^^^^ ^ ^^^^ଶ, where ^^ᇱ^ is the energy of the scatteredneutron, ^^^the mass of the isotope, and ^^^is the mass of the neutron. Inelastic neutron scattering occurs for neutron energies above the inelastic threshold energy of the isotope ^^^୦,^ൌ^^^,^^^^^ ^ ^^^^ / ^^^, where ^^^,^ is the energy level of the isotope’s first excited state. In this case,22 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO the average energy deposited in the item also includes the Q-value of the reaction with theisotope ^^^^^^^^^ ൌ ^^ ᇱ^ െ ^^^ ^ ^^^^.The neutron cross-section (^^^^) quantifies the probability of the reaction occurring with the energy ^^^. The total neutron cross-section ^^௧,^of an isotope is the sum over each reaction type: ^^௧,^^^^^^ ൌ ^ ^^^^ ^^^^^ ൌ ^^^,^^^^^^ ^ ^^^,^^^^^^.(1.6) ^ The total with the isotope at a total neutron scattering cross-section (^^^,^) and the total neutron capture cross-section (^^^,^). A related quantity, the neutron reaction rate ^^்,^of the item, is given as the of the reaction rates ^^்,^^of each isotope and reaction type: ^^ ൌ ^^^ൌ ^ ^^ ^Φ ^^^ ^^^ ^^^ ^d(1.7) ்,^ ்,^^ ்,^ ^ ^ ^^ ^ ^^^where Σ் total number of neutron reactions that occur in the item per unit time and volume. Tbl.1 below lists neutronics and other physical properties for select elements of interest in the periodic table. Specifically, Tbl.1 tabulates the eleven non-trace elements of the human body. Each non-trace element composes at least 0.1% mass percentage of the human body, pertaining to the most relevant neutron reaction channels. The remaining trace elements of the human body together compose <0.1% of the human body’s mass percentage. Isotopes of elements that may be used for neutron capture therapy, such as non-radioactive isotopes boron- 10 and gadodolinum-157, are also tabulated in Tbl.1. Other elements (and isotopes thereof) that may be of interest include those introduced from past medical procedures of a patient. These correspond to items or objects that may introduce additional nuclear reaction channels when irradiated with neutrons, such as dental implants and fillings (e.g., composed of silver, titanium, or ceramic), sensory and neurological implants such as cochlear implants, cardiovascular 23 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO implants such as heart valves, orthopedic implants such as pins, rods, screws, and plates (e.g., composed of titanium, silicone, and / or apatite), other types of medical implants, and the like. In Tbl.1, total neutron scattering and capture cross-sections in the epithermal energy range are reported for each isotope. Trailing digits in parentheses provide the standard errors. For most isotopes, neutron scattering cross-sections are independent of the incident neutron energy in the epithermal energy range, exhibiting a plateauing behavior as shown in FIGs.9A, 10A, and 10B. Conversely, neutron capture cross-sections are inversely proportional to the neutron energy in the epithermal energy range, corresponding to the so-called “1 / v law” as shown in FIGs.7A, 8A, and 9A. The total neutron capture cross-sections are reported in Tbl.1 for a neutron energy of 25.30 meV, as is convention. This neutron energy corresponds to the most probable speed (~2200 meters per second (m / s)) of free neutrons under a Maxwell-Boltzmann distribution at a temperature of 20 degrees Celsius (˚C). Referring to Tbl.1, a total neutron scattering cross-section (^^^,^) of about 5 barns or more corresponds to an isotope with a relatively high probability of scattering free neutrons, with some isotopes, e.g., hydrogen-1, having scattering cross-sections on the order of hundreds of barns. For example, the total neutron scattering cross-section can be about 6 barns or more, 7 barns or more, 8 barns or more, 9 barns or more, 10 barns or more, 15 bn or more, 20 barns or more, 25 barns or more, 50 barns or more, 75 barns or more, 100 barns or more, depending on the isotope. A total neutron capture cross-section (^^^,^) of about 1 barn or more corresponds to an isotope with a relatively high probability of absorbing free neutrons, with some isotopes, e.g., boron-10 and gadolinium-157, having capture cross-sections on the order of thousands to hundreds of thousands of barns. For example, the total neutron capture cross-section can be about 2 barns or more, 3 barns or more, 4 barns or more, 5 barns or more, 10 barns or more, 15 bn or more, 20 barns or more, 25 barns or more, 50 barns or more, 75 barns or more, 100 barns or more, 200 barns or more, 500 barns or more, 1000 barns or more, 2000 barns or more, 5000 barns or more, 10000 barns or more, 100000 barns or more, depending on the isotope. For reference, a barn is a metric unit of area equal to 10−28m2. Table 1: Physical properties for select elements of interest in the periodic table. Rows 1-8 list, for each element, the element’s name, the element’s chemical symbol, the element’s precent mass (PM) of the human body, the element’s percent atoms (PA) of the human body, the element’s 24 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO naturally occurring isotopes, each isotope’s natural abundance (NA), and each isotope’s total neutron cross-section in the epithermal energy range. Total neutron cross-sections are tabulated in terms of the total neutron scattering (^^^,^) and capture (^^^,^) cross-sections for the isotope (see Eq. (1.6)). Total neutron capture cross-sections are reported at a neutron energy of 25.30 meV. PM PA NA Neutron Cross-Section [barn] Element Symbol Isotope [%] [%] [%] Scattering Capture ) ) ) 25 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 40Ca 96.941 2.90(2) 0.41(2) 42Ca 0.647 1.42(8) 0.68(7) ) .) t oug t e absorbed dose்n q. ( .3) accounts or t e tota energy depos ted by (indirectly) ionizing radiation, when the item in question is an organ or tissue, it does not account for the differing biological effects of each dose component. This is quantified by the so-called “equivalent” dose ^^். The equivalent dose ^^்has units of sieverts (Sv), defined as the gray- equivalent (Gy-eq), and is calculated as a weighted sum over the absorbed dose for each dose component: ^^் ൌ ^^்,ఊ^^்,ఊ ^ ^ ^^்,^^^^்,^^ .(2.1)Similarly, the the absorbed dose rate for each are a biological) weighting factors for the particular organ or tissue (^^), which are usually specified by a regulatory body such as the International Commission on Radiological Protection (ICRP) or the International Commission on Radiation Units and Measurements (ICRU). These weighting factors are inferred using other experiments and techniques which will not be discussed further herein. In general, an equivalent dose of radiation is estimated to have the same biological effectas an equal dose of gamma rays, which is given a weighting factor of ^^்,ఊ ൌ 1. For example, anabsorbed dose of one (1) Gy by alpha particles can lead to an equivalent dose of about 20 Sv. 26 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO Notice then, by definition, the equivalent dose ^^்is a calculated (inferred) quantity that is neither equal, nor proportional, to the (measurable) absorbed dose ^^். For BNCT applications, the equivalent (“prescribed”) dose delivered to an organ or tissue is typically divided into four components: (i) the gamma ray dose, (ii) the neutron-induced boron dose, (iii) the neutron-induced nitrogen dose, and (iv) the neutron-induced hydrogen dose. These are summarized concisely in Eq. (2.2) below: ^^் ൌ ^^்,ఊ ^ CBE^ ൈ ^^்,^ ^ RBE^ ൈ ^^்,^ ^ RBEୌ ൈ ^^்,ୌ. (2.2)The equivalent dose from each component is the product of the biological weighting factor and the absorbed dose, where the gamma, boron, nitrogen, and hydrogen dose components are identified by the subscripts γ, B, N, and H, respectively. The relative biological effectiveness (RBE) and the combined biological effectiveness (CBE) (which is used for boron for combining its RBE and cellular distribution), are correction factors that account for how damaging the radiation is at the cellular level relative to an equivalent absorbed gamma ray dose. RBE and CBE values are usually assumed constant for a given particle species, particle energy distribution, and tissue type (e.g., skin, tumor, or soft tissue). The gamma ray dose (^^்,ఊ) is attributed to residual gamma rays in the BNCT neutron source, which ionizes the (mostly) Compton-scattered electrons. The scattered electrons can have kinetic energies in the range of about one (1) MeV to 10 MeV and can travel about one (1) millimeter (mm) to 10 mm in the organ or tissue, creating an ionization cloud along their trajectories. The neutron-induced dose components (boron, nitrogen, and hydrogen) are attributed to reactions with thermal, epithermal, and fast neutrons. This spans a neutron energy range from about one (1) meV to one (1) MeV for a typical BNCT neutron source. The reaction products from these neutron-induced reactions include positively charged particles that deposit their energy locally, within about one (1) micron (µm) to 10 µm from where they are produced. As used herein, “thermal neutrons” will refer to neutrons having kinetic energies of one (1) eV or less, “epithermal neutrons” will refer to neutrons having kinetic energies between one (1) eV and 30 keV, and “fast neutrons” will refer to neutrons having kinetic energies of 30 keV or more. Although other definitions can be used, this demarcation is the standard convention for BNCT applications. “Ultra-fast neutrons” (e.g., relativistic neutrons) will refer to neutrons having kinetic energies of one (1) MeV or more. 27 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIGs.7A-7B are plots of neutron cross-section and reaction product energies for boron- 10. The boron dose (^^்,^) is attributed to the10B(n,α)7Li reaction resulting from capture of thermal and epithermal neutrons by boron-10 loaded in the tissue, which produces an alpha particle (of ~1.47 MeV) and a lithium-7 ion (of ~840 keV). The probability of this reaction occurring at a particular neutron energy is related to its neutron cross-section as depicted in FIG. 7A. The energetics of this reaction are depicted in FIG.7B, which shows the reaction product energies as a function of angle with respect to an incident neutron at normal incidence. As the center-of-mass energy is increased by the incident neutron energy, the reaction products become more energetic in the forward direction. FIGs.8A-8B are plots of neutron cross-section and reaction product energies for nitrogen-14. The nitrogen dose (^^்,^) is attributed to the14N(n,p)14C reaction resulting from capture of thermal and epithermal neutrons by nitrogen-14 in the tissue, which produces a proton (of ~590 keV) and a residual carbon-14 nucleus (of ~40 keV). The probability of this reaction occurring at a particular neutron energy is related to its neutron cross-section as depicted in FIG. 8A. The energetics of this reaction are depicted in FIG.8B, which shows the reaction product energies as a function of angle with respect to an incident neutron at normal incidence. Similar as above, as the center-of-mass energy is increased by the incident neutron energy, the reaction products become more energetic in the forward direction. FIGs.9A-9B are plots of neutron cross-section and reaction product energies for hydrogen-1. The hydrogen dose (^^்,ୌ) is attributed to the1H(n,n)1H reaction resulting from (mostly) elastic scattering of epithermal and fast neutrons by hydrogen-1 in the tissue, which produces a recoiling proton. A second hydrogen dose may also be considered that is attributed to the1H(n,ɣ)2H reaction, which is typically captured in the gamma ray dose (^^்,ఊ) together with the other sources of gamma rays. This reaction can be a significant component of the gamma ray dose since the point of origin is often within the patient. The1H(n,ɣ)2H reaction results from capture of thermal and epithermal neutrons by hydrogen-1 in the tissue, which produces a gamma ray (of ~2.2 MeV) and a residual hydrogen-2 (deuterium) nucleus. The probability of the1H(n,n)1H and1H(n,ɣ)2H reactions occurring at a particular neutron energy are related to their neutron cross-sections as depicted in FIG.9A. The energetics of the1H(n,n)1H reaction are depicted in FIG.9B, which shows the energy of the recoiling proton as a function of angle with respect to an incident neutron at normal incidence. Note, the proton is always emitted in the 28 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO forward direction due to conservation of momentum and kinetic energy in an elastic scattering event. FIGs.10A and 10B are plots of neutron cross-section for carbon-12 and oxygen-16, respectively. The neutron-induced carbon and oxygen doses are usually trivial for the energy ranges relevant to BNCT, at least in cases when a neutron-generating target 60 is composed of lithium. When the target 60 is composed of beryllium, and the acceleration potential of an accelerator 40 is increased for neutron generation (e.g., increased relative to a lithium-based target 60), the neutron energy spectrum also increases. This can lead to small but non-negligible carbon and oxygen doses. Nevertheless, the dose from elastic scattering reactions is still generally dominated by hydrogen. Particularly, for neutron energies below one (1) MeV, the carbon and oxygen doses are attributed to the12C(n,n)12C and16O(n,n)16O scattering reactions, which produce recoiling carbon-12 and oxygen-16 nuclei, respectively. As these reaction products are uncharged, they do not ionize tissue, instead causing superficial displacement damage via non-ionizing energy loss (NIEL). However, the carbon and oxygen doses can be significant, even the dominate form of ionization in tissue, for neutron energies above one (1) MeV. These energy ranges can be applicable to neutron sources that generate ultra-fast neutrons, such as fast neutron reactors (FNRs) and nuclear fusion reactors. Numerous nuclear reaction channels open up in the ultra-fast neutron energy spectrum for carbon-12 and oxygen-16. For example, above the inelastic threshold energy of 4.8 MeV, carbon-12 exhibits the12C(n,n’)12C* inelastic scattering reaction which results in emission of a gamma ray by the excited carbon-12 nucleus. For energies above 8.3 MeV and 8.8 MeV, the12C(n,n’)12C*→8Be+α and12C(n,α)9Be reactions can also occur, respectively. As another example, oxygen-16 exhibits fast neutron activation due to an (n,p) reaction, corresponding to the16O(n,p)16N→16O+ ɣ+β- radioactive decay process. This results in the emission of a proton, followed by a gamma ray and a beta particle from the unstable nitrogen-16 nucleus. Such dose components from carbon and / or oxygen can also be included in the equivalent dose (Eq. (2.1)) to quantify the effects due to ultra- fast neutrons. Nitrogen, hydrogen, carbon, and oxygen make up about 99% of the mass percentage of soft tissue and, therefore, are the primary neutron-induced dose mechanisms from external neutron radiation. The dose from injected (or ingested) neutron capture compounds, e.g., including boron and / or gadolinium, is specific to neutron capture therapies such as BNCT. Still, 29 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO the dose components from the other non-trace elements of the human body, e.g., calcium, phosphorus, potassium, sulfur, sodium, chlorine, and / or magnesium, can also be included in the equivalent dose depending on the situation, such as the neutron energy spectrum, the level of error in the equivalent dose, and / or the type of tissue under consideration. For example, calcium- 40 and phosphorus-31 exhibit thermal neutron capture via the40Ca(n,α)37Ar and31P(n,ɣ)32P reactions, respectively. The neutron-induced calcium and phosphorus doses are particularly relevant for hard tissue, e.g., bone matter and tooth enamel, as calcium and phosphorus can make up about 45% to 60% of the mass percentage of tissue in these cases. Nevertheless, these neutron-induced doses are difficult to measure in practice. Currently, no detectors exist that measure these individual dose quantities directly. Thus, the calibration of a BNCT treatment planning system, for example, relies on secondary parameters, causing a large uncertainty in its predictive capability. The one exception is the gamma ray dose. Gamma ray detectors, such as ionization chambers, photodiodes, gamma ray spectrometers, and thermoluminescent detectors, measure the gamma ray dose absorbed within a specific mass of matter. These gamma ray detectors can be calibrated readily using a gamma ray source of known intensity, energy, and dose rate, such as cobalt-60 (Co-60) or cesium-137 (Cs-137). For most gamma ray detectors, the response to photon dose is uniform over a broad range of energies, which allows calibrations to be transferred between systems straightforwardly. Such systems are well-known and will not be discussed further herein. That said, calibration of neutron detectors, particularly for BNCT applications, is significantly more difficult to achieve than gamma ray detectors. As shown in Eq. (1.5), the absorbed neutron dose (^^்,^) is strongly dependent on the neutron energy spectrum, the item with which the neutrons interact, the reaction products, and the energetics of the reaction products. Available neutron sources for calibration purposes are characterized by the energy spectrum they produce. For example, a radioactive source that emits neutrons through radioactive decay (e.g., californium-252 (Cf-252)) or through nuclear reaction channels (e.g., americium-beryllium (AmBe)) produce neutrons at a known intensity, but over a broad range of energies. These ranges of neutron energies are not directly applicable to a BNCT treatment. Thus, the response from a monitor (dosimetry) chamber calibrated to a Cf-252 or AmBe source will produce erroneous results when transferred to a BNCT neutron source without correlating the neutron energy spectrum. This happens to be the unknown variable for BNCT, as well as 30 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO other applications that utilize particle accelerators to generate neutrons. Similar arguments can be made for neutrons produced from a nuclear reactor, which are thermalized (< 1 eV), or a nuclear fusion source, e.g., deuterium-tritium fusion (~14.1 MeV neutrons) and deuterium-deuterium fusion (~2.45 MeV neutrons), both of which are too energetic for accurate comparison to BNCT. In practice, the neutron-induced dose components, as described above, are currently unable to be measured directly. Common practice is to assume the functions that describe the energy deposition (e.g., Eq. (1.5)), reaction probability (e.g., Eq. (1.6)), and reaction rate (e.g., Eq. (1.7)) are faithful, and therefore, the only unknown quantities are the neutron flux intensity and its energy distribution. This is a relatively dangerous assumption since the calibration of a BNCT treatment is traditionally based on computed dose components that cannot be directly compared against experiment. The traditional method for calibrating a BNCT treatment planning system is neutron activation analysis. This technique involves neutron activation of a known metallic foil to infer the neutron flux intensity (a single foil) or the energy distribution of the neutron flux (multiple foils). Measurement of energy distributions involves many foils that have distinctive resonances in the energy-dependent cross-section or energy threshold reactions. In general, the characteristic neutron flux ^^^^^^^^ can be inferred if the saturation activity ^^^, the energy-dependent cross-section ^^^^^^^, the mass ^^, and the molar mass ^^ of the (e.g., pure) sample are known: ^^^ ൌ ^^^ ^^^ ^^^^^^^^^^^^^^^^^^.(3.1) The saturation activity gamma ray spectroscopy techniques. This process can be described mathematically as: ^^^^^ െ ^^^(3.2) ^^^A metallic foil of amount of time ^^^. The foil is then removed from the neutron field and placed on a gamma ray spectrometer, typically a high-purity germanium (HP-Ge) detector, to count the net number ofphotons emitted within some initial count time ^^^, and stop time ^^ଶ. Here, ^^^ െ ^^^ is the netcount rate (real events ^^ minus background events ^^) of the characteristic photon that is emitted from the isotope with a characteristic decay rate ^^, and a pre-determined detection efficiency and branching ratio ^^. The neutron flux for monoenergetic emission is the ratio of the saturation activity and the cross-section. For multiple foil samples, Eq. (3.1) takes a matrix form, where the 31 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WOenergy dependent neutron flux can be inferred by multiplying the results vector ℛ ൌ^^^ି^^^ಮ,భெభ^భ,^ಮ,మெమ^మ , … ^ by the inverse of the energy-dependent neutron cross-section:^^^^^^^^ ൌ ℛ^^ି^^^^^^. (3.3) can be accurate for a limited set of situations, it is time consuming as it involves exposing and counting many individual foils which can take from several hours to several days depending on the complexity. For BNCT, this method offers limited resolution of the neutron energy spectrum which, as described above, can span more than nine orders of magnitude. In most cases, there are six to eight discrete energy points that are defined, but it is common to unfold hundreds of energy bins. This is, therefore, an ill-posed method as the number of outputs far exceeds the number of inputs. To overcome some, or all, of these abovementioned issues, this specification introduces radiation dosimetry systems 150 including one or more NiRDs 200 that can directly measure an absorbed dose and / or an absorbed dose rate of each neutron-induced dose component. For example, the described NiRDs 200 can utilize thin film and active detector technologies to precisely quantify the absorbed dose or rate of a neutron-induced reaction. Various NiRD 200 configurations and detection schemes are described that can measure one or more neutron- induced dose components in forward emission, backward emission, or both. These and other features related to the NiRDs 200 and radiation dosimetry systems 150 are described in more detailed below. FIG.3A shows a schematic diagram of an example radiation dosimetry system 150 configured as an active radiation detector system. The radiation dosimetry system 150 includes a neutron-induced radiation detector (or “NiRD”) 200 and readout electronics 300 coupled to the NiRD 200 for performing active, real-time readout. Here, “active” detection and readout refers to the voltage biasing and / or external energy supplied to the NiRD 200 by the readout electronics 300. For example, with appropriate voltage biasing from the readout electronics 300, the NiRD 200 can produce an immediate, observable signal for an individual quantum of ionizing radiation. This may facilitate faster and / or more sensitive detection of ionizing radiation compared to “passive” detectors which typically operate in a cumulative mode – exposure to ionizing radiation causes the accumulation of a signal that is stored in the passive detector. At a high-level, the NiRD 200 includes an irradiator material 210 and an active volume 220 coupled to the irradiator material 210. As shown in FIG.3A, a layer of the irradiator material 32 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 210 is positioned on an entrance window 224 of the active volume 220. In this example, the NiRD 200 further includes a gamma ray shield 202 enclosing the irradiator material 210 and active volume 220. For example, the gamma ray shield 202 can be a hollow cap (or shell) that can be placed over the irradiator material 210 and active volume 220. As further shown in FIG. 3A, ports IO1 and IO2 of the readout electronics 300 are connected to respective ports IO1 and IO2 of the active volume 220 via bias supply lines 252, with bias supply line 252-2 connected to a ground 315, e.g., a signal ground, a chassis ground, Earth ground, or other known reference voltage. A readout line 250 connects port IO3 of the readout electronics 300 to bias supply line 252-1 and is used for active readout of the NiRD 200. In the example of FIG.3A, as well as other examples provided herein, the NiRD 200 has an overall geometry that is rotationally symmetric about a central axis 170. The irradiator material 210 has an incident surface 211 orthogonal to the central axis 170 to accept neutrons propagating in this direction. For demonstrative purposes, the radiation dosimetry system 150 is configured to measure one or more neutron-induced dose ^^^^^^ components of a neuron beam 70 generated by the beam system 100, e.g., to calibrate a BNCT treatment planning system. The NiRD 200 is arranged such that the beam axis of the neutron beam 70 is coincident with the central axis 170. At least a portion of the neutron beam 70 irradiates the incident surface 211 of the irradiator material 210. For example, the NiRD 200 can be positioned within a beam diameter, e.g., a full-width-half-max (FWHM), of the neutron beam 70 to receive a uniform neutron flux across the incident surface 211 of the irradiator material 210. In some implementations, the neutron beam 70 may first be collimated with a neutron collimator, neutron aperture, or neutron lens before irradiating the irradiator material 210. The irradiator material 210 is configured to react with neutrons in the neutron beam 70 and emit products 520 of the reactions into the active volume 220, which are detected therein. The active volume 220 is defined by an enclosure 226 that is filled, at least partially (e.g., mostly), with an absorbing medium 600 that is sensitive to ionizing radiation. The absorbing medium 600 may also be referred to as an “absorber” for shorthand. In the described examples, the enclosure 226 can be a gas-filled chamber 226A, an insulating shell 226B, or a reflecting cavity 226C, and the absorbing medium 600 can be a fill gas 600A (e.g., Sec. III), a semiconductor material 600B (e.g., Sec. IV), or a scintillator 600C (e.g., Sec. V), respectively, that resides therein. The reaction products 520 are transmitted through the entrance window 224 33 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO and injected into the absorbing medium 600. The absorbing medium 600 is configured to absorb kinetic energy of the reaction products 520 and generate secondary particles 530 in response, which are ion-pairs 530-e / i for a gas 600A, electron-hole pairs 530-e / h for a semiconductor 600B, and scintillation photons 530-ɣ for a scintillator 600C. In general, the active volume 220 also includes one or more components that are actively voltage biased by the readout electronics 300, e.g., a pair of electrodes 222-1 and 222-2 or a photodetector 204, that are used for collecting the secondary particles 530. The sensitivity of the absorbing medium 600 to ionizing radiation, particularly gamma rays and / or one or more species of charged particles, can be defined in terms of its conversion efficiency (^^). The conversion efficiency refers to the number of secondary particles 530 produced in the absorbed medium 600 per unit of energy deposited by the ionizing radiation. Theconversion efficiency can also be related to the average energy deposited (^^ ൌ ^^ି^) by ionizingradiation to create a single second particle 530. In general, the conversion efficiency is dependent on the material composition of the absorbing medium 600. In some cases, the conversion efficiency may also depend (e.g., modestly) on the species of the ionizing radiation. As used herein, the absorbing medium 600 can be considered “sensitive” to ionizing radiation if it has a conversion efficiency of at least one (1) secondary particle per keV of deposited energy by theionizing radiation, that is, if ^^ ^ 1 / keV. For example, the absorbing medium 600 can have aconversion efficiency of 1 / keV or more, 2 / keV or more, 3 / keV or more, 4 / keV or more, 5 / keV or more, 6 / keV or more, 7 / keV or more, 8 / keV or more, 9 / keV or more, 10 / keV or more, 15 / keV or more, 20 / keV or more, 25 / keV, 50 / keV or more, 75 / keV or more, 100 / keV or more, 200 / keV or more, 300 / keV or more, 400 / keV or more, 500 / keV or more, or 1000 / keV or more. For gases 600A, the conversion efficiency refers to the number of ion-pairs 530-e / i produced per unit of energy deposited by ionizing radiation. For semiconductor materials 600B, the conversion efficiency refers to the number of electron-hole pairs 530-e / h produced per unit of energy deposited by ionizing radiation. For scintillators 600C, the conversion efficiency refers to the number of scintillation photons 530-ɣ produced per unit of energy deposed by ionizing radiation. Note, the conversion efficiency characterizes the number of secondary particles 530 generated at the point of interaction of the reaction products 520 with the absorbing medium 600. Some of these secondary particles 530 may be lost before being collected by the active volume 34 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 220, e.g., due to recombination and diffusion losses for ion-pairs 530-e / i and electron-hole pairs 530-e / h, or light leakage and reabsorption losses for scintillation photons 530-ɣ. Generally, the entrance window 224 forms a planar piece of the enclosure 226 that is optically thin (e.g., transparent) to the reaction products 520 emitted by the irradiator material 210. For example, the entrance window 224 may be composed of a different material than the enclosure 226 and / or be the thinnest portion of the enclosure 226. The thickness and material composition of the entrance window 224 can be chosen based on the “hardness” (or penetration ability) of the reaction products 520. “Soft radiation” such as alpha particles, protons, and other heavy charged particles have lower penetration ability than “hard radiation” such as gamma rays and beta particles. An entrance window 224 composed of one or more low-Z materials, such as beryllium, is generally appropriate for both hard and soft radiation. An entrance window 224 composed of aluminum can be appropriate for hard radiation. Other example materials that can be used for the entrance window 224 are provided elsewhere herein. In some implementations, the entrance window 224 can be omitted and the irradiator material 210 itself may form this piece of the enclosure 226. For example, the irradiator material 210 can be directly deposited on an electrode 222 interfacing a semiconductor material 600B. As another example, the irradiator material 210 can be directly deposited on a scintillator 600C. Such implementations of windowed or windowless NiRDs 200 may depend on the detection scheme utilized by the NiRD 200, which will be discussed in more detail in the following. The transparency of the entrance window 224 to ionizing radiation, particularly gamma rays and / or one or more species of charged particles, can be defined in terms of its thickness (^^^^) relative to the mean range (Δ^^) of the ionizing radiation in the entrance window 224. The mean range of an ionizing particle in the entrance window 224 is the average distance the ionizing particle travels before being fully stopped. In general, the mean range is dependent on the species of the ionizing particle, the incident energy of the ionizing particle, and the material composition of the entrance window 224. For gamma rays, the mean range is also referred to as the mean free path, attenuation length, or absorption length. The fraction of gamma rays that are not stopped by the entrancewindow 224 is the transmission ^^ ൌ exp^െ^^^^ / Δ^^^ of the entrance window 224. For example,the entrance window 224 can have a transmission of 90% or more, 95% or more, 99% or more, 99.9% or more, 99.99% or more, or 99.999% or more for gamma rays having kinetic energies of 35 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 1 keV or more, 10 keV or more, 100 keV or more, 1 MeV or more, 10 MeV or more, or 100 MeV or more. For charged particles, the mean range is related to the stopping power Δ^^ ൌ^ఌబ^ ^^ି^^^^^d^^ , where ^^^ is the incident energy of the charged particle and ^^^^^^ ൌ െd^^ / d^^ isof the charged particle in the entrance window 224. If the thickness ^^^^^ window 224 is less than the mean range of a charged particle with incident energy ^^^, then the charged particle is generally transmitted through the entrance window 224 but with some energy loss. The energy lost Δ^^ ൌ ^^^ െ ^^ௗ through the entrance window 224 canbe found by solving ^^^^ ൌ ^ఌబఌ^ ^^ି^^^^^d^^, where ^^ௗis the energy of the charged particle after traversing the entrance window 224. For a thin entrance window 224, the energy loss is approximately Δ^^ ^ ^^^^^^^^^^^.As used herein, the entrance window 224 can be considered “transparent” to ionizing radiation if it has a thickness of 10% or less than the mean range of the ionizing radiation having kinetic energies of 1 keV or more, that is, if ^^^^ / Δ^^ ^ 0.1 for ^^^ ^ 1 keV. For example, theentrance window 224 can have a thickness of 10% or less, 5% 1% or less, 0.1% or less, 0.01% or less, 0.001% or less, or 0.0001% or less than the mean range of ionization energy having kinetic energies of 1 keV or more, 10 keV or more, 100 keV or more, 1 MeV or more, 10 MeV or more, or 100 MeV or more. The radiation dosimetry system 150 uses the secondary particles 530 produced by the absorbing medium 600 as the means for detection and measurement of the energy deposited within the active volume 220. Particularly, for each count (or ionizing event) from a single ionizing particle (or quantum of radiation), at least a portion of the secondary particles 530 resulting from the ionizing event are collected by the active volume 220. The active volume 220 then produces a corresponding electrical signal as output. In the described examples, the produced electrical signal is an electric current pulse (^^ௗ) that flows through the readout line 250 to the readout electronics 300 for signal processing. However, in other examples, the produced electrical signal may be in the form of a voltage pulse or an oscillation in an otherwise non- pulsed current or voltage signal. In general, the shape and amplitude of an electric current pulse characterizes the species and amount of energy deposited by the ionizing particle. The readout electronics 300 can process 36 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO a series of such electric current pulses to determine an absorbed dose (or rate) of the reaction products 520 induced by the neutron beam 70. During active readout, the readout electronics 300 can apply a bias voltage (^^) to the active volume 220, which places the active volume 220 in a suitable electrical configuration for collecting, multiplying, and / or converting the secondary particles 530 into the current pulse. Generally, the bias voltage configures the active volume 220 for pulse mode, enabling pulse-height (energy) spectroscopy. For example, the bias voltage can induce gas multiplication effects in a fill gas 600A, induce charge carrier transport in a semiconductor material 600B, reverse bias a photodetector 204 coupled to a scintillator 600C, among other effects. Such processes are described in more detail in Secs. III-V. In this section, the overall functions performed by the irradiator material 210, active volume 220, absorbing medium 600, and readout electronics 300 are described. The gamma ray shield 202, if present, is configured to absorb, scatter, or otherwise mitigate residual gamma rays in the neutron beam 70. Generation of the neutron beam 70 is typically accompanied by a non-negligible dose of gamma rays, e.g., as byproducts from nuclear reactions in the target 60. This gamma ray dose (^^ఊ) can be independently measured by the radiation dosimetry system 150 using a detector 201, see FIGs.4A-4B for example. However, these residual gamma rays can ionize the absorbing medium 600, contaminating the neutron-induced dose measurements. The gamma ray shield 202 generally inhibits these residual gamma rays from interacting with the irradiator material 210 and active volume 220, thereby increasing the sensitivity of the NiRD 200 to incident neutrons without significantly altering the neutron energy spectrum of the neutron beam 70. For example, the gamma ray shield 202 can reduce dark counts, deadtimes, and / or detector pileup resulting from residual gamma ray absorption in the active volume 220. In some implementations, the gamma ray shield 202 is composed of lead. The gamma ray shield 202 can have a thickness in a range from about one (1) millimeter (mm) to 10 centimeters (cm). The chemical composition of the irradiator material 210 is chosen to isolate the neutron- induced dose ^^^^^^ component(s) of the neutron beam 70. That is, the irradiator material 210 includes one or more isotopes of one or more of the elements of interest, where each isotope ^^^^ is associated with one or more types of neutron reactions ^^^^ being characterized by the NiRD 200. In general, the elements of interest include boron, gadolinium, and the non-trace elements of the human body, as tabulated in Tbl.1. To wit, boron-10 is a suitable non-radioactive isotope for 37 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO neutron capture therapy (i.e., BNCT), with galodinium-157 being a promising candidate. The non-trace elements compose (nearly) 100% of the mass percentage of the human body and, therefore, pertain to most (or all) of the relevant nuclear reaction channels in the human body. To reiterate, isotopes are distinct nuclear species (or nuclides) of the same chemical element. That is, they have the same number of protons in their nuclei and belong to the same element in the periodic table. However, isotopes differ in their nucleon numbers due to having a different number of neutrons in their nuclei. While isotopes of a given element have (almost) the same chemical properties, they have different atomic masses and physical (e.g., nuclear) properties, as shown in Tbl.1. Tbl.2 below lists examples of irradiator materials 210 that include one or more of the elements of interest, e.g., as a pure element or in a compound. In BNCT applications for example, one choice for the irradiator material 210 for characterizing the neutron-induced boron (^^^) and nitrogen (^^^) dose components is boron nitride, which contains equal parts by atom of boron and nitrogen. Boron nitride is a semiconductor material and can be deposited in layers ranging from a few atoms thick to a few microns. To characterize the neutron-induced hydrogen dose (^^ୌ) component, another choice for the irradiator material 210 is polyethylene, which can be deposited in layers as thin as two to five microns. Polyethene is also capable of characterizing the neutron-induced carbon dose, e.g., resulting from ultra-fast neutrons. Alternatively, the carbon dose can be characterized by an irradiator material 210 composed of diamond (or graphite). Synthetic diamond can be deposited as monocrystalline, e.g., via Chemical Vapor Deposition (CVD), in layers on the order of microns or less. More generally, the irradiator material 210 can be an element of interest in pure form, a compound composed of two or more of the elements of interest, a compound composed of one or more of the elements of interest and one or more other, different elements, or a material composed of one or more other, different elements that is doped with one or more of the elements of interest. The irradiator material 210 can be in a solid, liquid, or gas state depending on the implementation. Examples of irradiator materials 210 that can characterize the neutron- induced doses for each of the elements of interest are tabulated in Tbl.2 and discussed further below. Many of these example irradiator materials 210 include two or more of the elements of interest in a compound and, therefore, can be used to characterize multiple neutron-induced dose components. 38 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO To characterize the neutron-induced hydrogen dose(s), the irradiator material 210 can be composed of hydrogen (H2) or a hydrogen compound. Examples of hydrogen compounds include hydrocarbons, hydrides, plastics, polymers, rubbers, oils, paraffins, acrylics, nylons, polyesters, and the like. Select examples include water (H2O), methane (CH4), polyethylene (CH2), poly(methyl methacrylate) (C5H8O2), polyethylene terephthalate (C10H8O4), hydrogen peroxide (H2O2), hydrogen chloride (HCl), carbonaceous sulfur hydride (CH8S), hydrogen sulfide (HS), hydrogen disulfide (H2S2), and other compounds provided herein. To characterize the neutron-induced boron dose(s), the irradiator material 210 can be composed of boron (B) or a boron compound. Examples of boron compounds include boranes, carboranes, boron nitrides, metal borides, borosilicate glasses, and the like. Select examples include boron nitride (BN), boron monoxide (BO), boron trioxide (B2O3), boron carbide (B4C), borax (Na2H20B4O17), ortho-carborane (C2B10H12), titanium diboride (TiB2), and other compounds provided herein. To characterize the neutron-induced carbon dose(s), the irradiator material 210 can be composed of carbon (C) or a carbon compound. Examples of carbon compounds include carbonates, bicarbonates, carbonyls, oxalates, oxocarbons, carbon nitrides, carbon sulfides, and the like. Select examples include carbon dioxide (CO2), carbon monoxide (CO), carbon disulfide (CS2), carbonyl sulfide (COS), ammonium carbonate ((NH4)2CO3), ammonium bicarbonate (NH4HCO3), cyanuric chloride (C3Cl3N3),cyanamide (CN2H2), and other compounds provided herein. To characterize the neutron-induced nitrogen dose(s), the irradiator material 210 can be composed of nitrogen (N2) or a nitrogen compound. Examples of nitrogen compounds include nitrides, nitrates, azides, nitrido complexes, nitrogen halides, nitrogen oxides, amines, amides, and the like. Select examples include nitrous oxide (N2O), dinitrogen trioxide (N2O3), silicon nitride (Si3N4), calcium nitride (Ca3N2), triphosphorus pentanitride (P3N5), tetrasulfur tetranitride (S4N4), and other compounds provided herein. To characterize the neutron-induced oxygen dose(s), the irradiator material 210 can be composed of oxygen (O2) or an oxygen compound. Examples of oxygen compounds include oxides (e.g., metal oxides such as the noble metals), carbonates, chlorates, nitrates, silicates, and many of the other compounds provided herein. 39 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO To characterize the neutron-induced sodium dose(s), the irradiator material 210 can be composed of sodium (Na) or a sodium compound. Examples of sodium compounds include sodium salts, sodium oxides, sodium electrides, sodides, sodium phosphates, organosodium compounds, intermetallic compounds, and the like. Select examples include sodium chloride (NaCl), sodium hydroxide (NaOH), sodium nitrate (NaNO3), sodium carbonate (Na2CO3), sodium bicarbonate (NaHCO3), sodium oxalate (Na2C2O4), sodium sulfide (Na2S), sodium thiosulfate (Na2S2O3), and other compounds provided herein. To characterize the neutron-induced magnesium dose(s), the irradiator material 210 can be composed of magnesium (Mg) or a magnesium compound. Examples of magnesium compounds include magnesium hydrides, magnesium halides, magnesium oxohalides, magnesium oxides, chalcogenides, and the like. Select examples include magnesium oxide (MgO), magnesium diboride (MgB2), magnesium carbonate (MgCO3), magnesium chloride (MgCl2), magnesium citrate (C6H6MgO7), magnesium hydroxide (Mg(OH)2), epsomite (MgSO4^7H2O), and other compounds provided herein. To characterize the neutron-induced phosphorus dose(s), the irradiator material 210 can be composed of phosphorus (P) or a phosphorus compound. Examples of phosphorus compounds include phosphorus salts, phosphides, phosphines, phosphorus nitrides, phosphorus sulfides, and the like. Select examples include phosphorus sesquisulfide (P4S3), phosphoric anhydride (P4O10), superphosphate Ca(H2PO4)2, and other compounds provided herein. To characterize the neutron induced sulfur dose(s), the irradiator material 210 can be composed of sulfur (S) or a sulfur compound. Examples of sulfur compounds include metal sulfides, sulfur halides, sulfur oxides, sulfur salts, and many of the other compounds provided herein. To characterize the neutron induced chlorine dose(s), the irradiator material 210 can be composed of chlorine (Cl) or a chlorine compound. Examples of chlorine compounds include chloride salts, chlorites, chlorides, chlorine oxides, organochlorine compounds, and the like. Select examples include chlorine dioxide (ClO2), polyvinyl chloride (C2H3Cl), sodium chlorate (NaClO3), and other compounds provided herein. To characterize the neutron induced potassium dose(s), the irradiator material 210 can be composed of potassium (K) or a potassium compound. Examples of potassium compounds include potash, potassium salts, potassium oxides, potassium nitrates, potassium sulfates, and the 40 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO like. Select examples include potassium chloride (KCl), potassium oxide (K2O), potassium peroxide (K2O2), potassium citrate (K3C6H5O7), potassium nitrate (KNO3), potassium chlorate (KClO3), potassium carbonate (K2CO3), potassium bicarbonate (KHCO3), potassium oxalate (K2C2O4), potassium sulfate (K2SO4), and other compounds provided herein. To characterize the neutron induced calcium dose(s), the irradiator material 210 can be composed of calcium (Ca) or a calcium compound. Examples of calcium compounds include calcium phosphates (e.g., calcium phosphate ceramics (CPCs)), calcium salts, calcium oxides, calcium sulfates, and the like. Select examples include calcium carbide (CaC2), calcium borate Ca3(BO3)2, calcium carbonate (CaCO3), calcium bicarbonate (Ca(HCO3)2), calcium oxalate (CaC2O4), dicalcium phosphate (CaHPO4), tricalcium phosphate (Ca₃(PO₄)₂), calcium hydroxyapatite (Ca10(PO4)6(OH)2), calcium bisulfate (Ca(HSO3)2), and other compounds provided herein. For reference, CPCs are a class of biomaterials that are commonly used to simulate and / or graft upon bone matter and tooth enamel, making them particularly useful for characterizing the calcium and phosphorus doses in hard tissues. To characterize the neutron induced gadolinium dose(s), the irradiator material 210 can be composed of gadolinium (Gd) or a gadolinium compound. Examples of gadolinium compounds include gadolinium oxides, gadolinium salts, gadolinium halides, and the like. Select examples include digadolinium trioxide (Gd2O3), gadolinium nitrate (Gd(NO3)3), gadolinium trichloride (GdCl3), and gadolinium oxysulfide (Gd2O2S). 41 LEGAL02 / 46174409v1 OsWcfo.int60o0rtne.tnd umeG- - - - - - - - - - - - - - -0.EeFnelIr emeLof hlce aT / 1ae5.7l ts2bil2T36e 3 / e 15S -14.t 2s9 s6 er0.ewtoo nRi.Nf0to1e 2ksctslonDeay miree el tanr eemotthtrAfootaeirdoarrmireohetnsotseil du1lcwnoit Ra.thtse0r1et2nsil faioresttanemrmeolteai edhat crirfiof smo eehsiter cslepp’otene)e et)2)dir)2mraxp tl ma alOldS eatcel ley aci)rchtH2Cyh) 2SH)) h(H( r H(O3 ) )71esly eh2aeht HhSpf tpeedi eudfl ( e )ediNB(O2C e B4O4: o)4C( e ret xiorurossdiffluB( di(B e(eB02 12desd)Hemnletyen elphculoussidedxo dix diHv290lalmrn O2C(elht)el)4 n nec n nirtinooi bra a 44baeretuoH( en yheO2 yhO8egegan )egegn m rtcN(716Tre nh i prt teah tem(8 teHororoS8orornsmonononoxa4 / otate ylylH5 yl 01dydy braHdydyrororor r 20oderC W MoPoPC( oPC(H H CC(H H B B BoBoL A netB G aniE L O W.60d00G -.EFIL aT - / C57223K -6 / 5149lC -60.oNteS -kcoDyP -enrottAgM -aN - O - N - CseY BseY s HeY )2) ) 341N )3 eN4 )H0)2B12Bi ) ) )Set l 3C O2)2dirt Si(e )H lOT)2OaS 2OetaanC3))N()4N3 nadi CaN C((OC(C(C(nob(H2OeNi 3 aCtnrtiN((eeendaiCr( edi ee obraed 2N2 dixS(eo ep nd x di di raciirN( oa e dii(e eds rtdi x 1drnoubbri i ofl fl c)bd xonious usm3m )3olCh(e rce dit dirirtuiroetrroolrd v904o apc- mudmid luiOuiOcdi xn n nynnCone tin hufhcyh 4)2nCirmsgo n mpslu716moio ht naobo o oo 4 oH rbrbrbm H m4unau rana ortinocuiio) chpN5 sam rum iui4 / 2CrtOiTaCaCarCam N m H C A((AN( yCytCinNil lDiSaiCr 3tTP( ed d0ToSoS L A GE L O W.60d0G -0.EFIL aT / C -57223K -6 / 5149lC -60.oNteS -kcoDyP -enrottAgM -asNeY s OeY NseY C - B - H - )3)3)7)3)3O )3) C ) O2O) 2OgS4 )01 22B Cg l)CMP(O )4 l) ) OH4S) g g 6 ) e 4O C3 3)lO3C a2 aO N2 ) a 2OM gM ( MH6NaN(C (e2S2N( (eO2dif P(P2)2H2OlC Meet (eC(di H7 luedH( OlC(CaK(N( ettaaNa et ( diandieetan(N(af edrioobro t xaro^ r4 siiraC( eNdOuqdyCedi ( eedirdanobetelrobraaluxbiralhti ySnt adif so o d c c c hg seh etMsna adhirxota oolhrolh 1cv9uironaccibxlousihtmum m m m m)i2 ( sucips idcp ms uiuiuiuiui) etrorool lh04he yc mu47mum m m m m iuiuiuiuiuiesnesnesnesnesneH n Oi(mh h popspsr neir nimiodovs16yuisa4 / 2CodSodSod d d gagagagagaga gsSoSoSoS M M M M M MM( po o p lEhPhPuShl d t0CoPoSoP L A GE L O W.60d00G -.EFIL aT - / C5722s3Ke6Y / 5149lC -60.oNteS -kcoDyP -enrottAgM -aN - s OeY N - C - B - H - )3) )4)2) O7) )3O 3 C ) O ) 4 )3P3O2 5) O 2 H6 3OlC H 2 K(O2) O4) )23O ) C O4HaO e S ) O C K OKC3NK ( etC2S22CO Ba 2C CC ( e tiH t(aK2 ((edK K((eetanK(K(aC( 3( et aetataap)C(ei(dxioete txreatrt aaarrnobeteti oolbrraacl t(aa CetaC iax feldieanno (h hbetpspasybtaobr)alohox ) 2ehorHtaOfl1vdno pic nhcac b ous raroraci)2axp p ) d(6 us90uomum m m m m m m m c bac bpis uis uis uis uis uis uis uis uismum m mO3o mm Cmui u₂i)y₄h)4 ib44OmOPm716msoat sasasasasasasasaic ulic uic uicH(uiccl calacP(₃uic(01 uic4 / 2CotPotPot t t t t tPoPoPoPoPoPoPalCalCalCa alCC( acCiiDr alTC( a alCC( a 0C L A GE L O W.60ds0 e0G.Y EFIL aT / C -57223K - 6 / 5149lC- -seY -60.oNteS- - -sekYcoDyP- - - -enrottAgM- - - -aN- - - -sesesO Y Y- eY 64N -seY- -C- - - -B- - - -H- - - -)) )) 3S2O3)33 lO2d 2OCdG(N( GdGed (e (ediGdidix(oiertrtaofrllthusiciymn rtxo1vdnuuionm m m90il uiu u44poni iinini 716mdalol l 4og dodod / 2CiDaGaGa 0G L A GE L Attorney Docket No.069415 / 632275 / TLIFE.0006.WO Note, as demonstrated in Tbl.1 and mentioned above, different isotopes of the same chemical element react differently with neutrons so the atomic concentration of each isotope in the irradiator material 210 is a point of consideration. Most materials of a particular chemical formula are made up of the natural abundance (NA) of isotopes of those chemical elements. Hence, a material can be chemically pure while still being isotopically impure. For example, carbon-12 and carbon-13 have a natural abundance of 98.89% and 1.11%, respectively, so naturally occurring diamond (or graphite) will generally be composed of carbon-12 and carbon- 13 isotopes in these proportions. However, some materials, such as synthetic diamond, can be enriched or made isotopically pure using advanced manufacturing techniques. With that in mind, in some implementations, the irradiator material 210 is composed of one or more of the elements of interest, e.g., to eliminate extraneous nuclear interactions with other elements that may introduce dark counts in the active volume 220. For example, the irradiator material 210 can be composed of one or more of the elements of interest with a chemical purity of about 90% or more, 95% or more, 99% or more, 99.9% or more, 99.99% or more, or 99.999% or more. In some further implementations, the irradiator material 210 is composed of a single isotope of each of the selected elements, e.g., to eliminate extraneous nuclear interactions with other isotopes that may also introduce dark counts in the active volume 220. For example, the irradiator material 210 can be composed of a single isotope of each of the selected elements with an isotopic purity of about 90% or more, 95% or more, 99% or more, 99.9% or more, 99.99% or more, or 99.999% or more. In some implementations, the irradiator material 210 is homogenous, which is generally preferable for dose measurements as the irradiator material 210 emits the reaction products 520 (mostly) isotropically. That is, the atomic composition of the irradiator material 210 is homogeneous, and the mass density of the irradiator material 210 is homogeneous. The irradiator material 210 can be polycrystalline or monocrystalline, with monocrystalline generally being preferred to eliminate impurities, dislocations, grain boundaries, traps, or other defects that may cause unwanted scattering or trapping of the reaction products 520. It should be mentioned, however, that irradiation of the irradiator material 210 with neutrons may eventually degrade its homogeneity, chemical purity, and / or isotopic purity. For example, neutron scattering events can cause dislocation damage in the irradiator material 210, neutron capture events can transmute isotopes in the irradiator material 210, among other effects. 47 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO Hence, in some implementations, the irradiator material 210 is replaceable, which may be accomplished with an entrance window 224 that is removable from the enclosure 226. A new entrance window 224, with a new irradiator material 210 deposited thereon, may then be attached to the enclosure 226. Alternatively, or in addition, the enclosure 226 itself may be removable, e.g., configured as a hollow cap or multiple separatable pieces, that can be removed from the active volume 220 and a new enclosure 226 placed thereon. FIGs.6A and 6B are cross-sectional side views of the irradiator material 210 depicting examples of a neutron capture event 500-C and a neutron scattering event 500-S, respectively. Here, the irradiator material 210 is formed into a layer having a thickness of ^^୍. For example, in some implementations, the irradiator material 210 can have a thickness in a range from about 10 nanometers (nm) to 10 microns (µm). In some implementations, the irradiator material 210 can have a total surface area in a range from about one (1) millimeter-squared (mm2) to 500 centimeter-squared (cm2). For most of the example NiRDs 200 described herein, the irradiator material 210 is arranged as a uniform layer on the entrance window 224 (or on another planar surface of the active volume 220), providing a uniform incident surface 211 for the neutron beam 70. A few example NiRDs 200 implement more sophisticated configurations, such as spherical (e.g., FIGs.13A-13B) and microstructured (e.g., FIGs.17A-17B and 24A-24B) geometries, that can be suitable for dose detection in isotropic emission, diffuse neutron sources, situations involving multiple neutron sources, among others. Referring to the example neutron capture event 500-C shown in FIG.6A. An incident neutron 510-I impinges on the incident surface 211 of the irradiator material 210 and is absorbed by an isotope therein. The isotope then emits particles 520-X and 520-Y as reaction products, with particle 520-X propagating in the forward direction 522-X and particle 520-Y propagating in the backward direction 524-Y. The particular direction (forwards or backwards) that the particles 520-X and 520-Y are emitted is generally a probabilistic event, with the stipulation that momentum and total energy (including the incident neutron energy and Q-value) is conserved during the reaction. On average, each of particles 520-X and 520-Y will be emitted in the forward and backward directions ~50% of the time (similarly for neutron capture events 500-C involving more than two reaction products 520). The active volume 220 can register one or both of particles 520-X and 520-Y as ionizing events, generating an electric current pulse (^^ௗ) in response. Particularly, if the NiRD 200 is configured for forward emission, the active volume 48 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 220 registers particle 520-X as an ionizing event. If the NiRD 200 is configured for backward emission, the active volume 220 registers particle 520-Y as an ionizing event. If the NiRD 200 is configured for both forward and backward emission, e.g., dual, or isotropic emission, the active volume 220 registers both particles 520-X and 520-Y as a coincident ionizing event. Consider an irradiator material 210 composed of boron nitride as an example. In this case, the irradiator material 210 can capture neutrons via the10B(n,α)7Li and14N(n,p)14C reactions, which are characterized in FIGs.7A-8B. In a10B(n,α)7Li capture event, a boron-10 isotope absorbs an incident neutron 510-I which subsequently breaks up into an alpha particle (e.g., particle 520-X) and a lithium-7 ion (e.g., particle 520-Y). The active volume 220 can detect one or both of the alpha particle and lithium-7 ion as both particles are ionizing. In a14N(n,p)14C capture event, a nitrogen-14 isotope absorbs an incident neutron 510-I which subsequently breaks up into a proton (e.g., particle 520-X) and a carbon-14 nucleus (e.g., particle 520-Y). The active volume 220 can detect the proton as this particle is ionizing. Consider an irradiator material 210 composed of polyethylene as another example. In this case, the irradiator material 210 can capture neutrons via the1H(n,ɣ)2H reaction, which is characterized in FIG.9A. In a1H(n,ɣ)2H capture event, a hydrogen-1 isotope absorbs an incident neutron 510-I which subsequently emits a gamma ray (e.g., particle 520-X) and decays to a hydrogen-2 nucleus (e.g., particle 520-Y). The active volume 220 can detect the gamma ray as this particle is ionizing. Referring to the example neutron scattering event 500-S shown in FIG.6B. An incident neutron 510-I impinges on the incident surface 211 of the irradiator material 210 and is scattered 510-S by an isotope therein (either elastically or inelastically). The isotope then emits particle 520-Z as the reaction product, which propagates in the forward direction. If the particle 520-Z is charged upon scattering, the active volume 220 can register the particle 520-Z as an ionizing event, generating an electric current pulse (^^ௗ) in response. Alternatively, or in addition, if the particle 520-Z is excited upon inelastic scattering, the active volume 220 can register any ionizing particle(s), e.g., gamma rays and / or other charged particles, emitted from the excited particle 520-Z when it subsequently decays. Consider the example above when the irradiator material 210 is composed of polyethylene. In this case, the irradiator material 210 can scattering neutrons via the1H(n,n)1H reaction, which is characterized in FIGs.9A-9B. In a1H(n,n)1H scattering event, a hydrogen-1 49 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO isotope scatters an incident neutron 510-I, generating a recoiling proton (e.g., particle 520-Z). The active volume 220 can detect the recoiling proton as this particle is ionizing. There are several advantages for developing this kind of measurement. Firstly, the radiation dosimetry system 150 allows for an active, real-time measurement of an observed reaction rate, as described in Eq. (1.7). Secondly, the energy distribution of the reaction products 520 can be detected by the NiRD 200 in either forward emission, backward emission, or both. In BNCT applications for example, independently measured forward and backward quantities of neutron capture reactions 500-C can be compared to results obtained from a treatment planning system either separately or combined. The combination of these two observables provides a direct measurement of the absorbed dose triple product: (i) neutron flux, (ii) neutron cross- section, and (iii) energy deposition, as described in Eq. (1.5). Alternatively, when the NiRD 200 is configured for dual (or isotropic) emission, the output of the NiRD 200 provides a direct measurement of the absorbed dose triple product. Thus, the NiRD(s) 200 described herein can provide a high precision correction to a BNCT treatment planning system. Note, the thickness of the irradiator material 210 is a particular design feature that can be optimized for neutron-induced dose measurements. For example, for an infinitely thin irradiator material 210, the particle distributions for neutron capture events 500-C will have an average energy corresponding to isotropic emission, and a width corresponding to a Doppler shift induced by the incident neutron energy. The particle distributions for neutron scattering events 500-S will have an average energy corresponding to conical forward-directed emission, resulting from the kinetic energy imparted to the recoiling particle. However, the observed energies will generally have losses due to the thickness of irradiator material 210 and entrance window 224 traveled. This process can be described according to linear energy transfer (LET) and is well- known for a given charged particle species and material. These relationships are quantified for the abovementioned neutron reactions and irradiator materials 210 in FIG.10C, which shows aplot of stopping power (^^ ൌ െd^^ / d^^) versus energy (^^) for neutron-induced reaction products520 in boron nitride and polyethylene. These LET effects can be accounted for during calibration of the radiation dosimetry system 150, which involves exposing the NiRD 200 to one or more monoenergetic sources of charged particles of known intensity and energy. These and other calibration steps are described in more detail below. 50 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIG.3B is a schematic diagram depicting an example configuration of the readout electronics 300. As shown in FIG.3B, the readout electronics 300 can include a bias supply 310, a preamplifier 320, an amplifier 330, and a multichannel analyzer (MCA) 340. Here, the readout electronics 300 are configured for pulse-height (energy) spectroscopy, facilitating measurement of absorbed doses and absorbed dose rates. The general functions of each electronic component of the readout electronics 300 is summarized below. A more detailed circuit implementation of the readout electronics 300 is provided with reference to FIG.3C. The bias supply 310 is configured to generate a bias voltage over ports IO1 and IO2. The bias supply 310 can be any suitable power supply, such as a high-voltage power supply. The preamplifier 320 is configured to receive a series of electric current pulses over port IO3, e.g., from the active volume 220, and convert each electric current pulse into a respective voltage pulse. The preamplifier 320 can be any suitable low-noise charge-sensitive preamplifier, such as a resistive feedback charge-sensitive preamplifier. The amplifier 330 is configured to receive the voltage pulses from the preamplifier 320 and amplify each voltage pulse. In some implementations, the amplifier 330 may also perform additional pulse-shaping on the voltage pulses, e.g., to reduce the length of the voltage pulse tails and / or provide higher resolution in the pulse shape before an analog-to-digital conversion stage. The amplifier 330 can be any suitable linear amplifier or linear pulse-shaping amplifier, such as an electrometer. The MCA 340 is configured to receive the amplified (output) voltage pulses from the amplifier 330 and analyze each amplified voltage pulse to determine a histogram “or spectrum” of counts versus pulse-height (energy). The MCA 340 can then determine an absorbed dose or rate from the histogram (and present it to a user, e.g., on a user interface or display). The MCA 340 can be any suitable analog or digital MCA, such as a digital workstation MCA or a digital portable MCA. FIG.3C is a circuit diagram of an example circuit topology 150-CD for the radiation dosimetry system 150 in FIG.3A. Particularly, the circuit topology 150-CD provides an example implementation of the readout electronics 300 configuration in FIG.3B for performing pulse- height (energy) spectroscopy. The bias supply 310 is a DC power supply in this implementation, such as a battery that generates a DC voltage. Alternatively, the bias supply 310 can be an AC-to-DC converter that 51 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO converts an AC voltage source, e.g., an electrical outlet, to DC voltage. The bias supply 310 includes (or can be represented by) a voltage source (^^^^ connected in series with a (e.g., large) source resistor (^^^^ between ports IO1 and IO2 of the readout electronics 300. This applies the bias voltage (^^^ across ports IO1 and IO2 of the active volume 220, e.g., while injecting little or no residual current into the electrical current signal output by the active volume 220. The preamplifier 320 is a resistive feedback charge-sensitive preamplifier in this implementation. The preamplifier 320 includes an operational amplifier (opamp) 322 that has its output connected to its inverting input via a feedback resistor (^^^) and a feedback capacitor (^^^) that are in parallel with each other. A (e.g., large) coupling capacitor (^^^) is connected between the inverting input (-) of the opamp 322 and port IO3. The non-inverting input (+) of the opamp 322 is connected to ground 315. Note that connection to the opamp 322’s dual- or single-supply voltages are omitted in FIG.3C for clarity. The preamplifier 320 further includes a combinational clipping network 324 connected to the output of the opamp 322. The clipping network 324 includes an input resistor (^^^^), a first diode (^^^), a first voltage source (^^^), a second diode (^^ଶ), a second voltage source (^^ଶ), and an output resistor (^^^ଶ). The first diode and voltage source form a positively biased clipper. The second diode voltage source form a negatively biased clipper in parallel with the positively biased clipper. The clipping network 324 can be placed before the amplifier 330 gain stage to reduce the decay-time constant of the voltage pulses output by the preamplifier 320. The amplifier 330 is a linear amplifier in this implementation, such as an electrometer. The amplifier 330 includes an opamp 332 that has its inverting input (-) connected to the output of the clipping network 324 by an input resistor (^^^^). The non-inverting input (+) of the opamp 332 is connected to ground 315. An output resistor (^^^ଶ) connects the output of the opamp 332 to its inverting input (-) for gain feedback. Note that connection to the opamp 332’s dual- or single-supply voltages are omitted in FIG.3C for clarity. The combined effect of the preamplifier 320 and amplifier 330 is to convert, shape, and amplify the electric current pulse (^^ௗ) into an amplified (output) voltage pulse (^^^) that is received as input by the MCA 340. Here, the output voltage pulses are referenced against ground 315. The MCA 340 is a digital MCA in this implementation, such as a digital workstation MCA or a digital portable MCA. The MCA 340 includes an analog-to-digital converter (ADC) 342, a field-programmable gate array (FPGA) 344, and a communications interface 346 (e.g., 52 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO including a user interface or display). In some implementations, the ADC 342 is a fast and high- resolution ADC. For example, the ADC 342 can have a bit resolution of 14 bits or more and a sampling rate of 60 million samples per second (MSPS) or more. The ADC 342 receives the output voltage pulse from the amplifier 330 and converts it into a digital signal that can be efficiently processed by the FPGA 344. The FPGA 344 can perform processes such as pulse- height and pulse-shape discrimination to generate the histogram of counts for a series of such digitized voltage pulses, see FIGs.3E-3H for example. The results of which, e.g., including the computed neutron-induced dose(s) and / or rate(s), can then be displayed on the communications interface 346, e.g., for presentation to a user. To perform such functions, the FPGA 344 can include various digital signal processing modules including, but not limited to, a digital filter (e.g., Sallen-Key filter), an energy reconstruction filter (e.g., a trapezoidal filter), a baseline restorer, a timing trigger, a pulse pile-up rejector, among other modules. In the implementation of FIG.3A, discrimination, filtering, and other signal processing operations is performed digitally by the FPGA 344 of the MCA 340. However, in other implementations, some of the signal processing performed by the FPGA 344 can be performed by separate circuity, e.g., via analog or digital means. For example, a discriminator circuit (e.g., a Foster-Seeley discriminator circuit) can be integrated in the MCA 340, either preceding or proceeding the FPGA 344, to discriminate on the pulse height, shape, and / or temporal characteristic of the voltage pulses. FIG.3D is an equivalent circuit diagram of an analog component 150-EC of the radiation dosimetry system 150-CD circuit topology in FIG.3C, that is, excluding the MCA 340. Assuming a relatively large coupling capacitance (^^^), the equivalent circuit 150-EC can be represented as a parallel RC circuit, including an effective resistance (^^) and an effectivecapacitance (^^), plus an output voltage gain (^^ ^ 1). This corresponds to a time constant of ^^ ൌ^^^^ for the equivalent circuit 150-EC. For the example NiRDs 200 described herein, the charge collected ^^ௗ^^^^ by the readout electronics 300 is a direct measure of the energy deposition by an ionizing particle in the active volume 220. This is referred to as “pulse mode” as the active volume 220 produces an electric current pulse that is proportional to the energy absorbed from an ionizing event. As mentioned above, the readout electronics 300 can appropriately voltage bias the NiRD 200 to operate in pulse mode. That said, the NiRDs 200 and / or readout electronics 300 can also be configured in 53 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO other modes of operation such as “current mode” or “mean squared volage mode” (e.g., “Campbelling mode”) via suitable biasing voltages to the active volume 220 and circuit configurations of the readout electronics 300. These modes can be suitable for counting or timing experiments, e.g., to determine reaction rates, decay curves, and / or statistical properties of the neutron-induced reactions. For example, the radiation dosimetry system 150 can characterize a neutron reaction involving emission of a prompt gamma ray by measuring counts versus time, which produces a decay curve with a characteristic half-life that can be used for isotope identification of the irradiator material 210. When considering pulse mode operation, an ionizing particle deposits its kinetic energy within the absorbing medium 600 over an interaction (or stopping) time, which can be on the order of nanoseconds for gases 600A or picoseconds for semiconductors 600B and scintillators600C. The number of secondary particles 530 (^^^^ୡ ൌ ^^^^ௗ) produced in the absorbing medium600 is equal to the total energy deposited (^^ௗ) by the ionizing particle multiplied by the conversion efficiency (^^) of the absorbing medium 600. Due to active voltage biasing of the active volume 220, at least a portion of the secondary particles 530 are then collected, converted, and / or multiplied for a total duration of ^^^, where ^^^represents the charge collection time (orpulse length) of the resultant electric current pulse ^^ௗ^^^^ ൌ ^^^ௗ^^^^ output by the active volume220. The charge collection time may vary significantly between different NiRD 200 implementations. For example, for gases 600A, the charge collection time can be on the order of milliseconds or microseconds, whereas for semiconductors 600B and scintillators 600C, the charge collection time can be on the order of nanoseconds. These times reflect both the mobility of charge carriers within the active volume 220 and the average distance that is traveled before collection. The shape of the current pulse is indicative of the species of the ionizing particle, e.g., via distinct mechanisms of LET in the absorbing medium 600. Moreover, in pulse mode, thetotal charge collected ^^ௗ ൌ ^^ௗ^^^^^ ൌ ^^^^^^^^ୡ ൌ ^^^^^^^^ௗ by the readout electronics 300 isproportional to the number of secondary particles 530 generated in the absorbing medium 600 and, therefore, the total energy deposited by an ionizing event. Here, ^^ is the elementary chargeand ^^ is the total gain ^^ ^ 1 or loss ^^ ^ 1 of the signal after output from the active volume 220.Note, the actual shape of the pulse produced from the flow of charge carriers within the active volume 220 can be addressed using the Shockley-Ramo theorem. 54 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO Referring to the equivalent circuit 150-EC, for short current pulses compared to the timeconstant ^^^ / ^^ ≪ 1, the equivalent circuit 150-EC behaves like an integrator during the chargingphase 0 ^ ^^ ^ ^^^ of the effective capacitor (^^):௧ ି (4.1) ^^^^^^^ ^ ^^^^ ^ ^ ^^ௗ^^^ᇱ^^^^^ᇱ ൌ ^^^^ି^^^ௗ^^^^.^ Hence, the output accumulated chargethe readout electronics 300. Particularly, the leading edge of the voltage pulse for 0 ^ ^^ ^ ^^^ isgoverned, at least in part, by the energy deposition mechanism of the particular ionizing particle within the active volume 220. The MCA 340 can operate in pulse-shape mode (PSM), e.g., using pulse-shape discrimination (PSD), to determine the particular species of reaction product(s) 520 generating the voltage pulse. This is depicted more clearly in FIG.3E which shows plots of an example process for PSD performed by the MCA 340. The top plot of FIG.3E shows several current pulses 350 each having a different pulse shape. These include a Dirac pulse 350A, a square pulse 350B, a triangle pulse 350C, and a raised-cosine pulse 350D that may be generated by different species of ionizing particles. For example, the Dirac pulse 350A may correspond to a gamma ray, the square pulse 350B may correspond to an alpha particle, the triangle pulse 350C may correspond to a lithium-7 ion, and the raised-cosine pulse 350D may correspond to a proton. The bottom plot of FIG.3E shows corresponding output voltage pulses 352A-352D for each of the current pulses 350A-350D. The MCA 340 can use the pulse shape of the output voltage pulse’s leading edge to determine the species of ionizing particle. For example, the MCA 340 can be pre-programmed with the pulse shapes of the ionizing particles. Alternatively, or in addition, the MCA 340 can be calibrated to the pulse shapes by exposing the NiRD 200 to a (e.g., monoenergetic) source of each of the ionizing particles. Such implementations of the MCA 340 can be useful in situations involving mixed reaction products 520 when, for example, two dissimilar ionizing particles have similar, overlapping energies but deposit their energies via different mechanisms of LET. Referring again to the equivalent circuit 150-EC, the output voltage pulse during thedischarging phase ^^^ ^ ^^ through the effective resistor (^^) is then:^^^^^^^ ^ ఛ.(4.2)LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO The pulse height ℎ ൌ ^^^^ି^^^ ൌ ^ ି^ௗ ^^^ ^^^^^^^^ௗ of the output voltage pulse is thereforedirectly proportional to the total deposited energy by an ionizing particle. The trailing edge of the voltage pulse is governed (mostly) by the circuit properties of the readout electronics 300 which can be optimized, e.g., to reduce the decay-time constant. The MCA 340 can operate in pulse- height mode (PHM), e.g., using pulse-height discrimination (PHD), to measure the number of counts per channel (pulse height) for a series of voltage pulses. The MCA 340 can also use PHD to filter noise from the voltage pulses and reject dark counts. This is depicted more clearly in FIG.3F which shows plots of an example process for PHD performed by the MCA 340. The top plot of FIG.3F shows several voltage pulses 352-1 through 352-7 having different pulse heights. These pulses 352 are situated on a background noise signal 354, e.g., including amplifier and / or thermal noise. The MCA 340 can set an acceptance window 360 defining the range in pulse heights that are accepted. The MCA 340 can also set an acceptance level 362 defining the average pulse height that is accepted. Voltage pulses 352 that are outside the acceptance window 360 are rejected by the MCA 340, e.g., pulses 352-1, 352-2, 352-4, and 352-7. For example, pulses 352-1 and 352-4 may correspond to dark counts from extraneous charged particles and pulses 352-4 and 352-7 may correspond to dark counts from residual gamma rays. The bottom plot of FIG.3F shows the output voltage pulses 352-3, 352-5, and 352- 6 that are accepted by the MCA 340, which have also been filtered of the background noise signal 354. During or after such signal processing, the MCA 340 can then determine the histogram of counts over a particular exposure time of the neutron beam 70, and thereby the neutron-induced dose components (^^^^^ absorbed by the NiRD 200. For example, the radiation dosimetry system 150 can first be calibrated by exposing the NiRD 200 to a monoenergetic source of known intensity and energy (^^^), which produces individual ionizing events of a known, reference dose^^^ ൌ ^^^ / ^^ୟୠ^, where ^^ୟୠ^ is the mass of the absorbing medium 600. The reference dose ^^^ isproportional to the pulse height ^^^ ∝ ℎ^ of the output voltage pulse produced by an ionizingevent from these monoenergetic emissions. One such monoenergetic, reference source that can be used for calibration (e.g., point and / or linear calibration) of the NiRD 200 is americium-241 (Am-241) which decays predominately via alpha decay, producing an alpha particle and a weak gamma ray (with energy of 59.541 keV) as a byproduct. The alpha particle is produced with an energy of 5.486 MeV for 56 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 85% of the time. This alpha emission can be used for calibration of the NiRD 200 and is widely accepted for standard α-energy. For reference, the alpha particle is produced with an energy of 5.433 MeV for 13% of the time, and an energy of 5.388 MeV for the remaining 2% of the time. These alpha emissions may also be used for calibration of the NiRD 200, with the 5.486 MeV alpha emission generally being preferred. Using the reference source for calibration, the MCA 340 can calculate the neutron- induced doses from the number of counts ^^^ℎ^ of each channel (ℎ) as: 1 ℎ 1 ^^^̅(5.1) ^^ ൌ^^^ ^ ^^^ℎ^^^^ൌ^ ^^^ℎ^^^ ^ℎ^ൌ . ^^ ℎ ^^ௗ^^^ௗ^ℎ^ ൌ particular energy source. mean energy deposited in the volume of the absorbing medium 600 from a isotope (^^) and reaction type (^^). ℎ ∈ ^^, ^^ refers to the sum over channels corresponding to the reaction product(s) 520generated from the particular isotope and reaction type, which can be determined by the MCA 340 using PSD and PHD as described above. The MCA 340 can determine the neutron-induceddose rates (^^^ ^^) absorbed in a similar manner:d^^ ^^^ ^^1 (5.2) ^^ ൌൌ ^^^ ^^^ ^ℎ^^^ௗ^ℎ^where ^^^ ^^ூis the (homogenous) mass density of the irradiator material 210, and ^^ூ,^is the (homogeneous) atom density of the isotope in the irradiator material 210. Eqs. (5.1) and (5.2) account for the fact that the irradiator material 210 generates the reaction products 520 from the incident neutrons 510, but the absorbing medium 600 absorbs the dose from the reaction products 520. The MCA 340 can also determine the reaction rates (^^ூ,^^) of the irradiator material 210 as: 1 ^^^^ ^ℎ^ ^Φ ^^^ ^^^ ^^^ ^(5.3) ^^ ^^ ^ d^^where ^^ூis the appropriately scale the counts and count rates per channel to obtain the neutron-induced doses, dose rates, and reaction rates in Eqs. (5.1)-(5.3) if the reaction products 520 are not fully 57 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO absorbed in the NiRD 200, e.g., if only the forward 522 or backward 524 emitting reaction products 520 are detected for a neutron capture reaction 500-C. The neutron-induced doses and rates in Eqs. (5.1) and (5.2) correspond to a single point calibration of the dose measured by the NiRD 200. However, higher-order calibrations may also be used by the MCA 340 to calibrate the NiRD 200, such as a linear calibration that incorporates one or more additional reference sources. For example, neptunium-237 (Np-237) decays almost exclusively via alpha decay, producing an alpha particle with an energy of 4.958 MeV for ~100% of the time. Np-237 can be used in addition to Am-241 to expand the point calibration of the measured dose to a linear calibration. Moreover, the MCA 340 can use the pulse-height spectrum of the monoenergetic, reference source(s) to compensate for errors in the calculated dose components. One example of error compensation using the reference sources involves correcting for energy loses (∆^^) of the reaction products 520 in the entrance window 224 of the NiRD 200. Alternatively, or in addition, energy losses due to one or more dead layers in the NiRD 200 may also be compensated for using this approach. For reference, a dead layer is an inactive region that typically forms on the surface of a semiconductor material 600B, e.g., at the interfaces between the semiconductor 600B and an electrode 222, which the reaction products 520 may need to pass through before interacting with the active region of the semiconductor 600B. A dead layer can form due to natural and / or induced passivation of a semiconductor material 600B, or during the formation of an electrode 222 on the semiconductor 600B. In general, the energy deposited in the entrance window 224 by a reaction product 520 is lost, thus the (raw) measured signal will have some offset due to the lost energy. The reference sources can be used to determine and correct for this offset. Particularly, emissions from a reference source, e.g., alpha emissions from an Am-241 or Np-237 source, pass through the entrance window 224 and lose some energy due to LET, but deposit enough energy in theabsorbing medium 600 to generate a characteristic peak d^^ / dℎ ൌ 0, where ^^ is the number ofcounts and ℎ is the pulse height, as described above. The MCA 340, or a user of the MCA 340, can then correlate the known emission energy to the peak location in the pulse-height spectrum. This process can be performed for both low energy emissions and high energy emissions thus creating two points (ℎ^,^^^) and (ℎଶ,^^ଶ) that establish a linear calibration curve (^^^) relating particle intensity to a measured pulse height: 58 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO ^^ െ ^^^^^^ℎ^ ൌ ^^(5 ^^ ^ℎ െ ℎ^^ ଶ ^.4) ℎെ ℎ. ଶ^When the calibrationnet energy loss of the entrance out the LET effects of the entrance window 224. Alternatively, or in addition, the extrapolated calibration curve provides the minimum threshold energy (ℎ^) needed to produce a measurable signal: ^^ ℎ^ ൌ ℎ^ െ ^ℎଶ െ ℎ^^ ^(5.5) െ^^. ^ Different species of improved accuracy, this can characterized by the NiRD 200 using one or more respective reference sources. In other words, the MCA 340 can maintain a respective calibration curve for each species of reaction product 520. In other cases, a reference source can be used to define the solid angle incident upon the active volume 220. This scenario, most common under vacuum, can be useful if there are collimation features incorporated between the irradiator material 210 and the active volume 220. For integrated units of the irradiator material 210 and active volume 220, the NiRD 200 will generally have a 2π efficiency, which can also be verified with a reference source, e.g., an Am- 241 source. Another example of error compensation using a reference source involves correcting for the energy resolution of the NiRD 200. The pulse-height spectrum of a monoenergetic reference source is a direct measure of the response function ^^^ℎ^ of the NiRD 200, which characterizes broadening effects in the spectrum. For example, in some implementations, the response function of the NiRD 200 may be approximated as a Gaussian function: ଶ ^^1 ℎ (5.6) ^^ ℎwhere ^^ is the energy broadening due the response function using a few different methods. As one example, the MCA 340 can be programmed with software that includes a compensating response function ^^^ℎ) as a convolution to the raw output of the active volume 220. This effectively removes the broadeningeffects from the pulse-height spectrum ^^^ℎ^ ∗ ^^^ℎ^ ^ 1, where ∗ represents the convolution59 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO operator. In other implementations, for instance, if the signal-to-noise ratio of the NiRD 200 is relatively large, the MCA 340 can deconvolve the response function of the NiRD 200 from the raw output of the active volume 220, e.g., by correcting the number of counts per channel ^^^ℎ^ → ^^ି^^ℎ^ ∗ ^^^ℎ^ and the count rate per channel ^^^ ^ℎ^ → ^^ି^^ℎ^ ∗ ^^^ ^ℎ^.In the radiation dosimetry system 150 can be implemented for absolute, neutron-induced radiation dosimetry. Relative dosimetry is the typical method for measuring absorbed doses in clinical applications, such as BNCT applications. Here, once the NiRD 200 is calibrated against one or more reference sources, the measured neutron- induced doses, dose rates, and reaction rates in Eqs. (5.1)-(5.3) provide a direct measure of the total deposited energy and reaction probability for each isotope and reaction type, allowing calibration of a treatment planning system to a particular organ or tissue. In some implementations, such as those involving absolute or reference dosimetry, the MCA 340 may perform a dose conversion operation on each measured neutron-induced dose component. A dose conversion operation is typically involved when calibrating a treatment planning system for multiple different types of tissue, e.g., due to tissues having different atomic compositions, different radiation spectra, and / or different biological weighting factors. For example, the MCA 340 may compensate for one or more of these properties by scaling the neutron-induced dose components in Eq. (5.1) as: ^^்,^^ ൌ ^^்,^^^^^^ , (5.7)where ^^்,^^is a dose conversion factor for the organ or tissue. In some cases, the dose conversion factor may be approximated as ^^்,^^ ൌ ^^^ୟୠ^ / ^^ூ^൫^^்,^ / ^^ூ,^൯, where ^^ூ,^ is themass fraction of the isotope in the irradiator of the isotope in the organ or tissue. The MCA 340 can determine the absorbed dose rates and the reaction rates to the organ or tissue in a similar manner ^^^ ்,^^ ൌ ^^்,^^^^^ ^^ and ^^்,^^ ൌ൫^^்,^ / ^^ூ,^൯^^ூ,^^. A dose conversion operation may also be involved when the radiation dosimetry system 150 is implemented as a personalized dosimeter, e.g., a handheld or pocket dosimeter, that performs absolute or reference dosimetry for a particular user of the radiation dosimetry system 150. In such implementations, one or more NiRDs 200 of the radiation dosimetry system 150 may be placed on the surface of the user’s skin. The MCA 340 can then use a dose conversion factor specific to the user’s skin tissue to determine the neutron-induced dose. 60 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO Thus, the radiation dosimetry system 150 provides a highly accurate platform for measuring individual neutron-induced dose(s). In a radiation dosimetry system 150-MC including multiple NiRDs 200 and a gamma ray detector 201, the radiation dosimetry system 150-MC can determine multiple dose components, e.g., pertaining to a BNCT treatment, and thecorresponding equivalent neutron dose (^^்) or dose rate (^^^ ்) to the organ or tissue as in Eq.(2.1). An example of such a multi-component radiation dosimetry system 150-MC is described below with reference to FIGs.4A-4B. For completeness, a few examples of histograms (or pulse-height spectrums) that may be generated by the MCA 340 are shown in FIGs.3G-3H. FIG.3G is an example histogram of counts ^^^ℎ^ versus channel (ℎ) for an irradiator material 210 composed of boron nitride. As shown in FIG.3G and described above, the MCA 340 can individually determine the neutron- induced boron (^^^) and nitrogen (^^^) doses resulting from10B(n,α)7Li and14N(n,p)14C reactions with the irradiator material 210. FIG.3H is an example histogram of counts ^^^ℎ^ versus channel (ℎ) for an irradiator material 210 composed of polyethylene. As shown in FIG.3H and described above, the MCA 340 can individually determine the neutron-induced hydrogen dose(s) (^^ୌ) resulting from1H(n,n)1H and1H(n,ɣ)2H reactions with the irradiator material 210. FIG.5A is a flow diagram of an example process 700 for performing neutron-induced radiation dosimetry to determine one or more neutron-induced dose components of a neutron source, e.g., a neutron-generating target 60 that generates a neutron beam 70 for BNCT. The process 700 is an example of a dosimetry protocol that can be performed by the radiation dosimetry system 150 depicted in FIGs.3A-3D. In some implementations, the process 700 may briefly begin with connecting the NiRD 200 to the readout electronics 300. The process 700 proceeds by applying a bias voltage (^^) to the active volume 220, using the bias supply 310, to configure the active volume 220 in pulse mode (710). As explained above, pulse mode enables pulse-height (energy) spectroscopy as the active volume 220 produces an electric current pulse that is proportional to the energy absorbed from an ionizing event. The process 700 includes irradiating the irradiator material 210 with neutrons 510, generated by the neutron source, to emit reaction products 520 into the active volume 220 (720). 61 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO As explained above, the irradiator material 210 includes one or more isotopes of one or more of the elements of interest. The neutrons 510 react with the isotope(s) to generate the reaction products 520. The kinetic energy of the reaction products 520 is absorbed by the absorbing medium 600, producing secondary particles 530 in response. Due to the applied voltage bias, the active volume 220 collects, multiplies, and / or coverts at least a portion of the secondary particles 530 into a series of electric current pulses (^^ௗ^, where each electric current pulse characterizes the energy absorbed due to an ionizing event from a reaction product 520. The process 700 includes converting each electric current pulse produced by the active volume 220, using the preamplifier 320, into a respective voltage pulse (730). The process 700 includes amplifying each voltage pulse, using the amplifier 330, to generate a respective amplified (output) voltage pulse (^^^) (740). The process 700 includes analyzing each amplified (output) voltage pulse, using the MCA 340, to determine a respective neutron-induced dose (^^^^) for each isotope (^^) and each type of neutron reaction (^^) with the isotope (750). As above, the MCA 340 can use PHD, PSD, and various other signal processing techniques to determine the neutron-induced dose(s). FIG.4A is a schematic diagram depicting an example of a multi-component radiation dosimetry system 150-MC including an array of NiRDs 200-1 through 200-N, a gamma ray detector 201, readout electronics 300, and a switchboard 302 electrically coupling each of the radiation detectors 200 and 201 to the readout electronics 300. FIG.4B is a circuit diagram of an example configuration of the switchboard 302. In this example, the switchboard 302 includes multiple on / off switches 304-1 through 304-N for the NiRDs 200-1 through 200-N and an on / off switch 304 for the gamma ray detector 201, where the switches 304 interconnect the various ports of the switchboard 302. Ports IO1-IO2 of the readout electronics 300 are connected to respective ports SO1-SO2 of the switchboard 302 via the bias supply lines 252. Port IO3 of the readout electronics 300 is connected to port SO3 of the switchboard 302 via the readout line 250. Each radiation detector 200 and 201 has respective ports IO1-IO2 connected to corresponding ports IO1-IO2 of the switchboard 302. The radiation dosimetry system 150-MC is configured to measure: (i) multiple neutron- induced dose components (^^^^) of the neutron beam 70 using the NiRDs 200, and (ii) a gamma ray dose (^^ఊ) of the neutron beam 70 using the gamma ray detector 201. In some 62 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO implementations, the gamma ray detector 201 is a gamma ray spectrometer, e.g., a high-purity Germanium (HP-Ge) radiation detector. The radiation dosimetry system 150-MC is further configured to determine an equivalent dose (^^^ from each of the neutron-induced doses and the gamma ray dose, e.g., according to Eq. (2.1), for a particular organ or tissue. As mentionedabove, the radiation dosimetry system 150-MC can also determine the equivalent dose rate (^^^ )from the neutron-induced dose rates and the gamma ray dose rate in a similar fashion. Each NiRD 200-1 through 200-N includes an irradiator material 210-1 through 210-N composed of a different material. Each irradiator material 210-1 through 210-N is coupled to a respective active volume 220-1 through 220-N for measuring one or more of the neutron-induced dose components. In BNCT applications for example, the radiation dosimetry system 150-MC can include two NiRDs 200-1 and 200-2, with one NiRD 200-1 having an irradiator material 210-1 composed of boron nitride for measuring the neutron-induced boron and nitrogen doses, and the other NiRD 200-2 having an irradiator material 210-2 composed of polyethylene for measuring the neutron-induced hydrogen dose(s). Each active volume 220-1 through 220-N may utilize a different detection scheme that is optimized for measuring its particular neutron-induced dose component(s). In this implementation, each NiRD 200-1 through 200-N also includes a respective gamma ray shield 202-1 through 202-N enclosing the respective irradiator material 210-1 through 210-N and active volume 220-1 through 220-N. In other implementations, the irradiator materials 210 and active volumes 220 of the NiRDs 200 may be enclosed by a single gamma ray shield 202. Using the switchboard 302, the radiation dosimetry system 150-MC can selectively connect each of the radiation detectors 200 and 201 in parallel with the readout electronics 300 or sequentially activate / deactivate them (via the switches 304). This allows the radiation dosimetry system 150-MC to perform parallel or sequential measurement of the dose components. For example, each NiRD 200 and the gamma ray detector 201 may be irradiated, respectively, with neutrons and gamma rays in the neutron beam 70 and output electric current pulses in response. The readout electronics 300 can then processes the electric current pulses in parallel, or the respective current pulses generated by each detector 200 and 201 separately, to determine the dose components. The readout electronics 300 can then determine the equivalent dose after combining each of the measured dose components. 63 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some implementations, the radiation dosimetry system 150-MC may be configured as a multi-component dosimeter, e.g., a handheld or pocket dosimeter, that allows a user to actively measure the equivalent neutron dose (or rate) resulting from neutron exposure. The radiation dosimetry system 150-MC can track the accumulated dose over seconds, minutes, hours, weeks, months, years, etc. to mitigate stochastic health effects from such neutron exposure, e.g., development of radiation-induced cancer. The radiation dosimetry system 150-MC can also measure the current dose rate at the point of exposure to mitigate deterministic health effects, e.g., development of acute radiation syndrome. FIG.5B is a flow diagram of an example process 701 for performing multi-component radiation dosimetry to determine an equivalent dose induced by a neutron source, e.g., a neutron- generating target 60 that generates a neutron beam 70 for BNCT. The process 701 is an example of a dosimetry protocol that can be performed by the radiation dosimetry system 150-MC depicted in FIGs.4A-4B. In some implementations, the process 701 may briefly begin with connecting each NiRD 200-1 through 200-N, the gamma ray detector 201, and the readout electronics 300 to the switchboard 302. In general, the respective irradiator material 210 of each NiRD 200 includes one or more respective isotopes of one or more of the elements of interest. In other words, each NiRD 200 is configured to characterize one or more respective neutron-induce dose components of the equivalent dose. For each NiRD 200-1 through 200-N, the process 701 includes performing operations (760) and (700) for the NiRD 200. The process 701 includes irradiating the irradiator material 210 of the NiRD 200 with neutrons 510 generated by the neutron source (760). Note, this process 901 may occur sequentially or simultaneously for each of the various NiRDs 200-1 through 200-N. For example, a user or an automated means (e.g., actuators) may manually place and remove each NiRD 200 from a neutron beam 70 generated by the neutron source. Alternatively, each NiRD 200 may be simultaneously placed in the neutron beam 70. As another example, if the neutron source generates neutrons isotropically, then each NiRD 200-1 through 200-N can receive neutrons from the neutron source simultaneously as there is generally no directional preference. 64 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO The process 701 includes determining, using the readout electronics 300, a respective neutron-induced dose for each isotope in the irradiator material 210 and each type of neutron reaction with the isotope (700). This process 700 is described above with reference to FIG.5A. The process 701 includes irradiating the gamma ray detector 201 with gamma rays generated by the neutron source (770). The process 701 includes determining, using the readout electronics 300, a gamma ray dose for the gamma rays (780). The process 701 includes determining an equivalent dose of the neutron source, using the readout electronics 300, from the gamma ray dose and each of the neutron-induced doses (990). FIG.5C is a flow diagram of an example process 800 for performing radiation dosimetry of a neutron source. The process 800 includes irradiating an irradiator material with neutrons generated by the neutron source (810), where the irradiator material includes one or more isotopes of one or more of the elements of interest. The process 800 includes receiving, using readout electronics, a series of electrical signals generated by an active volume coupled to the irradiator material (820). The process 800 includes processing, using the readout electronics, each of the electrical signals to determine a neutron-induced dose for an isotope of an element of the irradiator material corresponding to a neutron reaction with the isotope (830). In implementations involving one or more additional, different irradiator materials each coupled to a respective active volume, the abovementioned operations of the process 800 can be repeated for each additional irradiator material and respective active volume. In implementations involving a gamma ray detector, the process 800 can further include the following operations. The process 800 can include irradiating the gamma ray detector with gamma rays generated by the neutron source. The process 800 can include receiving, using the readout electronics, a series of electrical signals generated by the gamma ray detector. The process 800 can include processing, using the readout electronics, each of the electrical signals generated by the gamma ray detector to determine a gamma ray dose of the neutron source. 65 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO The process 800 can further include calculating, using the readout electronics, an equivalent dose of the neutron source from the gamma ray dose and each of the neutron-induced doses. The equivalent dose can be a weighted sum of the gamma ray dose and each of the neutron-induced doses. In some implementations of the process 800, each weight of the weighted sum is a relative biological effectiveness (RBE) value or a combined biological effectiveness (CBE) value. In some further implementations of the process 800, e.g., for BNCT applications, the equivalent dose includes: a gamma ray dose, a neutron-induced boron dose, a neutron-induced nitrogen dose, and a neutron-induced hydrogen dose. Section III: Examples of Gaseous Ionization NiRDs (gas-NiRDs) FIGs.11A-11B are various views depicting an example of a gaseous ionization NiRD (“gas-NiRD”) 200A. FIG.11A is an isometric view of the gas-NiRD 200A. FIG.11B is a cross- sectional side view of the gas-NiRD 200A. For the gas-NiRD 200A, the active volume 220A is defined by a chamber 226A enclosing a first electrode 222-1, a second electrode 222-2, and a fill gas 600A dispersed between the first 222-1 and second 222-2 electrodes. For example, the chamber 226A can be an atmospheric pressure chamber, a high-pressure chamber (e.g., of about 5 atmospheres (atm) or more), or a low-pressure chamber (e.g., of about 1 atm or less). The chamber 226A may be composed of fiberglass, glass, aluminum, or other suitable material for maintaining a sealed environment. The electrodes 222-1 and 222-2 may be composed of graphite, aluminum, tin, stainless steel, or other suitable conducting material. The pressure (^^) within the chamber 226A can be chosen to establish a desired concentration of the gas 600A, e.g., to reduce the stopping distance of ionizing particles and / or enhance gas multiplication effects. In some implementations, the gas 600A is air or an inert gas such as neon, argon, krypton, or xenon. In other implementations, the gas 600A is a gas mixture including an inert gas and a quench gas. For example, the quench gas can be methane or a halogen such as chlorine or bromide. A common gas mixture that may be used for the fill gas 600A is P-10 which includes 90% argon and 10% methane. 66 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO As shown, the gas-NiRD 200A is rotationally symmetric about the central axis 170 such that the electrodes 222-1 and 222-2 are cylindrical and coaxial with each other, with the second electrode 222-2 positioned about the first electrode 222-1. The first electrode 222-1 is a solid cylinder positioned on the central axis 170 and has an outer radius of ^^^^. For example, the first electrode 222-1 can have an outer diameter in a range from about 10 µm to 1 cm. The second electrode 222-2 is a cylindrical shell having an inner radius of ^^ୋand an outer radius of ^^^ଶ,corresponding to a radial thickness of ^^^ଶ െ ^^ୋ. For example, the second electrode 222-2 canhave an inner diameter in a range from about 10 mm to 10 cm and a radial thickness in a range from about 10 µm to 1 cm. Each electrode 222-1 and 222-2 has a length of ^^^. For example, the first 222-1 and second 222-2 electrodes can each have a length in a range from about 10 mm to 25 cm. The chamber 226A is positioned about and laterally bounds the second electrode 222-2 establishing the active volume 220A with a cylindrical shape. The chamber 226A has a radius of^^େ୦, a thickness of ^^େ୦, and a total length of ^^େ୦ ^ ^^^ ^ ^^^^, where ^^^^ is the thickness of theentrance window 224. The entrance window 224 forms a top, planar piece of the chamber 226A that is orthogonal to the central axis 170. The entrance window 224 longitudinally bounds the electrodes 222 and the gas 600A dispersed therebetween. The irradiator material 210 is a uniform layer positioned on the entrance window 224, which in this example is a circular disc having a thickness of ^^୍and a radius of ^^େ୦. For example, the entrance window 224 can have a thickness in a range from about 10 µm to one (1) mm, and the irradiator material 210 can have a thickness in a range from about 10 nm to 10 µm. The gas-NiRD 200A is configured for detection in forward emission. Particularly, the irradiator material 210 is arranged to receive incident neutrons 510-I upon an outward-facing incident surface 211. Hence, forward emitting reaction products 520 from neutron capture 500-C and scattering 500-S events are registered as counts in the active volume 220A. The entrance window 224 is optically thin (e.g., transparent) to the reaction products 520, which are injected into the gas 600A. For example, the entrance window 224 may be composed of aluminum, beryllium, glass, quartz, or a silicate such as a mica. As shown, a reaction product 520-Z entering the gas 600A collides with atoms, leaving an ionizing track 526 of ion-pairs 530-e / i along its trajectory. The number of ion-pairs 530-e / i produced is proportional to the energy of the particle520-Z if it is fully stopped within the gas 600A. In general, the energy (^^ ൌ ^^ି^) involved to67 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO produce a single ion-pair 530-e / i is proportional to the ionization potential of the gas 600A, but is generally larger than the ionization potential by about a factor of about 1.5-3. For example, argon has an ionization potential of 15.7 eV, and the energy to produce an ion-pair 530-e / i is about ^^~26.4 eV. Then, a particle 520-Z having an energy of one (1) MeV that is fully stopped by argon may create about 40,000 ion-pairs 530-e / i. Here, the first electrode 222-1 is configured as an anode connected to port IO1, and the second electrode 222-2 is configured as a cathode connected to port IO2, which is grounded 315. The bias voltage ^^^^ is applied across the electrodes 222-1 and 222-2 to generate an electric field ^^^^ within the gas 600A, between the equipotential surfaces of the electrode 222-1 and 222-2.Assuming relatively thin electrodes 222 and ^^^^ / ^^^ଶ ≪ 1, the electric field can be represented, atleast approximately, as electric field lines radiating normally between the equipotential surfaces of the electrodes 222-1 and 222-2: ^^^^^^ ൌ ^^^^^^^̂^ ൌെ1 ^^ (6.1) ^̂^, log ^^^^ଶ / ^^^^^ ^^ where ^^^^ ^ ^^ ^ ^^^^ is the radial coordinate from the central axis 170. In general, thebias voltage is such that the active volume 220A is configured for pulse mode which may utilize gas multiplication effects. In this case, the active volume 220A can be radially sectioned into an“ion drift” region for ^^^ଶ ^ ^^ ^ ^^^, and an “avalanche” region for ^^^ ^ ^^ ^ ^^^^ in the localvicinity of the first electrode 222-1. ^^^is the critical radius and denotes the radial boundary between the ion drift and avalanche regions where the electric field equals the minimum threshold field to induce Townsend avalanches. Due to the geometry of the cylindrical gas-NiRD 200A, the electric field is strong near the first electrode 222-1 owing to the ^^ି^radial dependance. To illustrate, suppose a bias voltage of 2000 Volts is applied to the active volume220A, with radii for the electrodes 222 of ^^^^ ൌ 0.1 mm and ^^^ଶ ൌ 1 cm. The electric field at thefirst electrode 222-1’s surface is then ~5 ൈ 10^ V / m.Generally, the electric field is weak enough in the ion drift region to not cause gas multiplication effects, but strong enough to prevent recombination of the ion-pairs 530-e / i. This causes electrons 530-e to drift towards the first electrode 222-1, and positive ions 530-i drift towards the second electrode 222-2. In the immediate vicinity of the second electrode 222-2, the field strength is large enough to produce Townsend avalanches for each electron 530-e, substantially multiplying the charge (^^ௗ) collected at the electrodes 222-1 and 222-2. Hence, in 68 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO theses case, the electric current pulse (^^ௗ) output from the active volume 220A is generally dominated from collection of the electrons 530-e, e.g., corresponding to the “proportional counter” gaseous ionization region. The gas multiplication factor (^^), where ^^ represents the relative gain of the electric current pulse, can be modeled, at least approximately, as: ா^^^^n^^ ൌ ^ ^^^^^^^ (6.2) l ^^^ ^^^^, ^^^^ where ^^ is the first of electrons per unit path length. ^^ln 2^^ ln ^^ ൌ ^ln ൬ ^ െ(6.3) log ^^^ / ^^ ^ln ^^൨,^ଶ ^^∆^^ ^^^^^^ln ^^^ / ^^^^^ where ^^ is a property of the an avalanche for (^^ / ^^), and ∆^^ఒis the gas multiplication effects can improve the signal-to-noise ratio of the gas-NiRD 200A, as well reduce the amount of electronic amplification implemented by the readout electronics 300. FIG.12A is a cross-sectional side view depicting an example of a gas-NiRD 200A-BE configured for detection in backward emission. The gas-NiRD 200A-BE is configured similarly as the gas-NiRD 200A of FIGs.11A-12B except it is arranged to receive incident neutrons 510-I upon an inward-facing incident surface 211 of the irradiator material 210. Hence, backward emitting 524-Y reaction products 520-Y from neutron capture events 500-C are registered as counts in the active volume 220A. FIG.12B is a cross-sectional side view depicting an example of a gas-NiRD 200A-DE configured for detection in both forward and backward emission (aka dual emission). Here, the gas-NiRD 200A-DE may be referred to as being in a “sandwich” configuration. As shown, the gas-NiRD 200A-DE includes a first active volume 220A-1 for detection in forward emission and a second active volume 220A-2 for detection in backward emission. This allows the energy produced by a neutron capture event 500-C to be fully absorbed in the gas-NiRD 200A-DE as both forward emitting 522-X reaction products 520-X and backward emitting 524-Y reaction products 520-Y from neutron capture events 500-C are registered as counts. In this example, the active volumes 220A-1 and 220A-2 are connected in parallel to combine the electric currentpulses generated by each ^^ௗ ൌ ^^ௗ,^ ^ ^^ௗ,ଶ, where ^^ௗ,^ and ^^ௗ,ଶ are the electric current pulses from69 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO ionizing events in the first 220A-1 and second 220A-2 active volumes, respectively. Hence, coincident ionizing events from a single neutron capture reaction 500-C produce a single, higher- energy peak, which generally leads to more robust detection. As shown, a common chamber 226A, e.g., having two sections that can be joined together, laterally bounds the cathodes 222-1-2 and 222-2-2 to establish the two cylindrically shaped active volumes 220A-1 and 220A-2. The irradiator material 210 is arranged in a bisecting, central plane of the chamber 226A, sandwiched between the two active volumes 220A-1 and 220A-2. The gas-NiRD 200A-DE has mirror (or reflection) symmetry relative the central plane such that the second active volume 220A-2 is the mirror image of the first active volume 220A-1. Respective entrance windows 224-1 and 224-2 positioned on the irradiator material 210 separate the irradiator material 210 from the respective gases 600A-1 and 600-2 in each of the active volumes 220A-1 and 220A-2. The anode 222-1-1 of the first active volume 220A-1 is electrically connected to the anode 222-2-1 of the second active volume 220A-2. Likewise, the cathode 222-1-2 of the first active volume 220A is electrically connected to the cathode 222-2-2 of the second active volume 220A-2. This sets the two active volumes 220A-1 and 220A-2 in parallel with each other such that they are biased by the same bias voltage and their electrical current pulses sum. FIGs.13A-13B are various views depicting an example of a gas-NiRD 200A-S having a spherical configuration. FIG.13A is an isometric view of the gas-NiRD 200A-S. FIG.13B is a cross-sectional view of the gas-NiRD 200A-S. In this example, the gas-NiRD 200A-S is rotationally symmetric about a central point 171 and configured for detection in isotropic emission. Particularly, the irradiator material 210 can receive incident neutrons 510-I upon its incident surface 211 about a full solid angle and emit reaction products 520 in all directions within the solid angle. The spherical gas-NiRD 200A-S can be advantageous for detecting doses due to neutron capture reactions 500-C as no direction of emission is preferred. As shown, the irradiator material 210 is a solid sphere positioned on thecentral point 171. The irradiator material 210 has a radius of ^^ and is positioned at the center ofthe gas-NiRD 200A-S. For example, the irradiator material 210 can have an outer diameter in a range from about 10 µm to 1 cm. The irradiator material 210 is positioned within multiple spherical shells that establish the active volume 220A. The entrance window 224 of the chamber 226A is positioned radially on 70 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO the irradiator material 210 to radially bound the first electrode 222-1, separating the irradiator material 210 from the active volume 220A. The entrance window 224 has an outer radius of ^^^^corresponding to a radial thickness of ^^^^ െ ^^^^. For example, the entrance window 224 canhave a radial thickness in a range from about 10 µm to 1 mm. The first electrode 222-1 is a spherical shell positioned radially about the entrance window 224. The first electrode 222-1 hasan outer radius of ^^^^ corresponding to a radial thickness of ^^^^ െ ^^^^. The irradiator material210, entrance window 224, and first electrode 222-1 are suspended in place by a stem 227 that extrudes radially inward from the chamber 226A’s outer wall. The second electrode 222-2 is a spherical shell positioned radially about the first electrode 222-1 and the gas 600A is dispersed therebetween. The second electrode 222-1 has an outer radius of ^^^ଶcorresponding to a radialthickness of ^^^^ െ ^^ୋ. The gas 600A occupies a volume within the chamber 226A correspondingto a radial thickness of ^^ୋ െ ^^^^. The outer wall of the chamber 226A is positioned about andradially bounds the second electrode 222-2. The chamber 226A has an outer radius of ^^େ୦corresponding to a radial thickness of ^^େ୦ െ ^^^ଶ.The electric field generated within the gas 600A for the spherical gas-NiRD 200A-S can be represented, at least approximately, as electric field lines radiating normally to the equipotential surfaces of the electrodes 222-1 and 222-2: ^^^െ^^ ^^^ ൌ ^^^^^^^̂^ ൌ ^^^^^ଶ^^ (6.4) ^^^ ^ଶ^̂^, ^ଶ െ ^^^^^^ where ^^^^ ^ ^^ ^ ^^^^ is the radial coordinate from the center of the irradiator material210. As for the cylindrical gas-NiRDs 200A described above, the spherical gas-NiRD 200A-S can be appropriately biased for gas multiplication effects such that active volume 220A isradially sectioned into an ion drift region for ^^^ଶ ^ ^^ ^ ^^^, and an avalanche region for ^^^ ^ ^^ ^^^^^in the immediate vicinity of the first electrode 222-1. Due to the geometry of the spherical gas-NiRD 200A-S, the electric field is strong near the first electrode 222-1 owing to the ^^ିଶdependance. To illustrate, suppose a bias voltage of 2000 Volts is applied to the active volume220A, with radii for the electrodes 222 of ^^^^ ൌ 1 mm and ^^^ଶ ൌ 1 cm. The electric field at thefirst electrode 222-1’s surface is then ~2 ൈ 10^ V / m. Using the formula in Eq. (6.2), anexpression for the resulting gas multiplication factor (^^) can also be determined for the spherical gas-NiRD 200A-S. 71 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIGs.14A-14C are cross-sectional sides views depicting additional examples of gas- NiRDs 200A-P having planar configurations. Referring to FIG.14A, the gas-NiRD 200A-P is rotationally symmetric about the central axis 170. The electrodes 222-1 and 222-2 are arranged in respective planes that are orthogonal to the central axis 170. The electrodes 222-1 and 222-2 are planar and parallel to each other. The first electrode 222-1 is a parallel plate having a thickness of ^^^^. The second electrode 222-2 is also a parallel plate having a thickness of ^^^ଶ. For example, the first 222-1 and second 222-2 electrodes can each have a thickness in a range from about 10 µm to 1 cm. Here, each electrode222-1 and 222-2 has a circular shape in the ^^ െ ^^ plane with a diameter of ^^ୋ. For example, thefirst 222-1 and second 222-2 electrodes can each have a maximum lateral dimension in range from about 5 mm to 10 cm. In other implementations, the electrodes 222-1 and 222-2 may have square, rectangular, polygonal, or other appropriate shapes. The electrodes 222-1 and 222-2 are positioned from each other by a distance of ^^ୋalong the central axis 170, with the gas 600A dispersed therebetween. For example, the first 222-1 and second 222-2 electrodes can be positioned about 1 mm to 20 cm apart. The chamber 226A laterally bounds the electrodes 222 and gas 600A establishing the active volume 220A with a cylindrical shape. The chamber 226A has a thickness of ^^େ୦, and atotal height of ^^େ୦ ^ ^^^ଶ ^ ^^ୋ ^ ^^^^ ^ ^^^^ along the central axis 170, where ^^^^ is thethickness of the entrance window 224. The entrance window 224 forms a top, planar piece of the chamber 226A that is orthogonal to the central axis 170. In this example, the entrance window 224 is positioned on and longitudinally bounds the first electrode 222-1. The entrance window 224 has respective surfaces parallel to the first electrode 222-1. The irradiator material 210 is a uniform layer, which in this example is a circular disc having a thickness of ^^୍and a diameter of^^ୋ ^ 2^^େ୦. The irradiator material 210 is positioned on the entrance window 224 and has anincident surface 211 parallel thereto. The gas-NiRD 200A-P is configured for detection in forward emission. Particularly, the irradiator material 210 is arranged to receive incident neutrons 510-I upon the outward-facing incident surface 211. Hence, forward emitting reaction products 520 from neutron capture 500-C and scattering 500-S events are registered as counts in the active volume 220A. 72 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO The electric field generated within the gas 600A for the planar gas-NiRD 200A-P can be represented, at least approximately, as a uniform electric field with electric field lines parallel to the equipotential surfaces of the electrodes 222-1 and 222-2: ^^ ^^ ൌ ^^^̂^ ൌ െ(6.5) ^̂^. ^^ୋIn general, the bias voltage is such that the active volume 220A is configured for pulse mode. The electric field is weak enough to not cause gas multiplication effects, but strong enough to prevent recombination of the ion-pairs 530-e / i. This causes electrons 530-e to drift towards the first electrode 222-1, and positive ions 530-i to drift towards the second electrode 222-2. Hence, in these cases, the electric current pulse (^^ௗ) output from the active volume 220A is a combination of charge collected from both the electrons 530-e and the ions 530-i, e.g., corresponding the “ionization chamber” gaseous ionization region. Generally, a biasing voltage large enough to induce avalanches for a uniform electric field is impractical. For example, withan electrode spacing of ^^ୋ ൌ 1 cm, an applied bias voltage of ~ 50,000 Volts would be involvedto achieve an electric field of ~5 ൈ 10^ V / m. Moreover, if the electric field is strong enough tocause Townsend avalanches within the entire volume of the gas 600A, self-multiplication is likely to occur, entering an operating region of limited or no proportionality, e.g., corresponding to the “Geiger-Mueller” gaseous ionization region. This region is avoided for dosimetry and spectroscopy measurements but can be utilized for counting and timing measurements. Note, each of the example gas-NiRDs 200A can be biased in any of the gaseous ionization regions, if desired, to perform such measurements. FIG.14B is a cross-sectional side view depicting an example of a planar gas-NiRD 200A-P-BE configured for detection in backward emission. The gas-NiRD 200A-P-BE is configured similarly as the gas-NiRD 200A-P of FIG.14A except it is arranged to receive incident neutrons 510-I upon an inward-facing incident surface 211 of the irradiator material 210. FIG.14C is a cross-sectional side view depicting an example of a planar gas-NiRD 200A-P-DE in a sandwich configuration for detection in dual emission. Like the cylindrical gas- NiRD 200A-DE of FIG.12B, the planar gas-NiRD 200A-P-DE utilizes a second active volume 220A-2 that is the mirror image of the first active volume 220A-1. As shown, the gas-NiRD 200A-P-DE uses the first active volume 220A-1 for detection in forward emission and the 73 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO second active volume 220A-2 for detection in backward emission. A common chamber 226A establishes the two active volumes 220A-1 and 220A-2 between respective entrance windows 224-1 and 224-2. The irradiator material 210 is arranged in a bisecting, central plane of the chamber 226A, sandwiched between the two active volumes 220A-1 and 220A-2. The entrance windows 224-1 and 224-2 are positioned on respective anodes 222-1-1 and 222-2-1 of the active volumes 220A-1 and 220A-2, with the irradiator material 210 positioned therebetween. Section IV: Examples of Semiconductor NiRDs (semi-NiRDs) FIGs.15A-15B are various views depicting an example of a semiconductor NiRD (“semi-NiRD”) 200B. FIG.15A is an isometric view of the semi-NiRD 200B. FIG.15B is a cross-sectional side view of the semi-NiRD 200B. For the semi-NiRD 200B, the active volume 220B is defined by an insulating shell 226B enclosing a first electrode 222-1, a second electrode 222-2, and a semiconductor material 600B. The semiconductor 600B is positioned between the electrodes 222-1 and 222-2 and forms respective electrical contacts (or junctions) thereto. For example, the insulating shell 226B (or dielectric cladding) can be composed of silicon dioxide, aluminum (III) oxide, titanium dioxide, hafnium (IV) oxide, zirconium dioxide, strontium peroxide, niobium pentoxide, gadolinium (III) oxide, calcium dioxide, combinations thereof, or any appropriate insulating (or dielectric) material. In these cases, the entrance window 224 may be composed of the same material as the insulating shell 226B or a different insulating (or dielectric) material. The semiconductor 600B is preferably monocrystalline, e.g., to eliminate traps and defects that degrade charge carrier transport. The semiconductor 600B can be composed of silicon, silicon carbide, thallium (I) bromide, germanium, gallium arsenide, gallium nitride, indium phosphide, indium gallium nitride, aluminum arsenide, aluminum gallium nitride, aluminum gallium arsenide, indium gallium arsenide, diamond, selenium, cadmium telluride, cadmium zinc telluride, boron nitride, mercury (I) iodide, combinations thereof, or any appropriate semiconducting material. The semiconductor material 600B can be intrinsic (undoped), n-doped, p-doped, or combinations thereof (see FIGs.18A-21F for example). Different combinations of semiconductor materials 600B, electrical contacts, and doping schemes can be used to optimize various different properties for a particular species of reaction product(s) 520, e.g., the energy deposition (e.g., pulse shape) of the reaction product(s) 520, charge collection time of the charge carriers, amount of charged collected by the charge carriers, improve charge carrier transport and reduce 74 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO recombination effects, improve current-voltage characteristics, among other properties. For example, the semi-NiRD 200B can be configured as a diffused junction detector, a surface barrier detector, a fully-depleted detector, a semiconductor diode detector, a detector with ion- implanted electrodes, among others. The semi-NiRD 200B has a similar configuration as the planar gas-NiRD 200A-P but employing solid-state materials. Particularly, the semi-NiRD 200B is rotationally symmetric about the central axis 170 with the electrodes 222-1 and 222-2 arranged in respective planes that are orthogonal to the central axis 170. The electrodes 222-1 and 222-2 are planar and parallel to each other. The first electrode 222-1 is a uniform layer formed on the semiconductor 600B having a thickness of ^^^^. The second electrode 222-2 is also uniform layer formed on the semiconductor 600B having a thickness of ^^^ଶ. For example, the first 222-1 and second 222-2 electrodes can each have a thickness in a range from about 10 nm to 1 µm. Here, each electrode222-1 and 222-2 has a circular shape in the ^^ െ ^^ plane with a diameter of ^^ୋ. For example, thefirst electrode 222-1, second electrode 222-2, entrance window 224, irradiator material 210, and semiconductor material 600B can each have a maximum lateral dimension in a range from about 5 mm to 10 cm. In other implementations, the electrodes 222-1 and 222-2 may have square, rectangular, polygonal, or other appropriate shapes. The electrodes 222-1 and 222-2 are positioned on opposite sides of the semiconductor material 600B which has a thickness of ^^ୗalong the central axis 170. For example, the semiconductor material 600B can have a thickness in a range from about 1 µm to 500 µm. The insulating shell 226B laterally bounds the electrodes 222 and semiconductor material 600B, establishing the active volume 220B with a cylindricalshape. The insulating shell 226B has a thickness of ^^େ୪, and a total height of ^^େ୪ ^ ^^^ଶ ^ ^^ୗ ^^^^^ ^ ^^^^ along the central axis 170, where ^^^^ is the thickness of the entrance window 224.The entrance window 224 forms a top, planar piece of the insulating shell 226B that is orthogonal to the central axis 170. The entrance window 224 is positioned on and longitudinally bounds the first electrode 222-1 and has surfaces parallel thereto. The irradiator material 210 is uniform layer in the shape of a circular disc having a thickness of ^^୍and a diameter of ^^ୋ^ 2^^େ୪. The irradiator material 210 is positioned on the entrance window 224 and has an incident surface 211 parallel thereto. For example, the entrance window 224 can have a thickness in a range from about 1 nm to 100 nm, and the irradiator material 210 can have a thickness in a range from about 10 nm to 100 nm. 75 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO The semi-NiRD 200B is configured for detection in forward emission. Particularly, the irradiator material 210 is arranged to receive incident neutrons 510-I upon the outward-facing incident surface 211. Hence, forward emitting reaction products 520 from neutron capture 500-C and scattering 500-S events are registered as counts in the active volume 220B. As shown, a reaction product 520-Z entering the semiconductor material 600B collides with atoms, leaving an ionizing track 526 of electron-hole pairs 530-e / h along its trajectory. The number of electron- hole pairs 530-e / h produced is proportional to the energy of the particle 520-Z if it is fullystopped within the semiconductor 600B. In general, the energy (^^ ൌ ^^ି^) involved to produce asingle electron-hole pair 530-e / h from ionizing radiation is proportional to the bandgap of the semiconductor 600B, but is generally larger than the bandgap by about a factor of about 2-3. For example, intrinsic silicon has a bandgap of 1.12 eV at room temperature, and the energy to produce an electron-hole pair 530-e / h is about ^^~3.6 eV. Then, a particle 520-Z having an energy of one (1) MeV that is fully stopped by silicon may create about 300,000 electron-hole pairs 530-e / h. Here, the first electrode 222-1 is configured as an anode connected to port IO1, and the second electrode 222-2 is configured as a cathode connected to port IO2, which is grounded 315. The bias voltage ^^^^ is applied across the electrodes 222-1 and 222-2 to generate an electric field ^^^^ through the semiconductor 600B, between the equipotential surfaces of the electrode 222-1 and 222-2. In general, the bias voltage is such that the active volume 220B is configured for pulse mode, which generally depends on the doping scheme implemented for the semiconductor material 600B. For example, a semiconductor 600B including a PN junction (e.g., FIGs.21A- 21B) or a PIN junction (e.g., FIGs.21C-21D) is generally reverse biased. Consequently, the totalelectric field profile (^^ ^ ^^ୠ୧) through the semiconductor 600B can include a built-in, spatiallyvarying electric field (^^ୠ୧ ൌ ^^ୠ୧^^^^) due to various semiconducting junctions within thesemiconductor 600B (and the electrical junctions at the electrodes 222). Consider an intrinsic or uniformly doped semiconductor material 600B as an example. In these cases, the electric field can be represented, at least approximately, as a uniform electric field within the bulk of the semiconductor 600B: ^^ ^^^̂^^^ (7.1) ^̂^. LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO The electric field is strong enough to induce charge carrier transport and reduce (or prevent) recombination of the electron-hole pairs 530-e / h. This causes electrons 530-e to drift towards the first electrode 222-1, and holes 530-h to drift towards the second electrode 222-2. Thus, the electric current pulse (^^ௗ) output from the active volume 220B is induced by charge carrier drift of the electrons 530-e and the holes 530-h. Note, application of the bias voltage tothe active volume 220B may be accompanied by a static, leakage (or dark) current ^^ௗ → ^^ௗ ^ ^^^due to thermal generation of free electron-hole pairs 530-e / h. For wide-bandgap semiconductors600B like diamond and silicon carbide, this is typically negligible ^^^ ^ 0 even at roomtemperature and can be corrected for by the readout electronics 300 if present, e.g., using PHD orfiltering. As mentioned above, the current-voltage characteristics of the current pulse ^^ௗ ൌ ^^ௗ^^^^from an ionization event and the leakage current ^^^ ൌ ^^^^^^^ depend on the type of electricaland / or semiconducting junctions implemented in the active volume 220B. Examples of these are discussed below with reference to FIGs.18A-21F. FIG.16A is a cross-sectional side view depicting an example of a semi-NiRD 200B-BE configured for detection in backward emission. The semi-NiRD 200B-BE is configured similarly as the semi-NiRD 200B of FIGs.15A-15B except it is arranged to receive incident neutrons 510- I upon an inward-facing incident surface 211 of the irradiator material 210. FIG.16B is a cross-sectional side view depicting an example of a semi-NiRD 200B-DE in a sandwich configuration for detection in dual emission. Like the planar gas-NiRD 200A-P- DE of FIG.14C, the semi-NiRD 200B-DE utilizes a second active volume 220B-2 that is the mirror image of the first active volume 220B-1. As shown, the semi-NiRD 200B-DE uses the first active volume 220B-1 for detection in forward emission and the second active volume 220B-2 for detection in backward emission. A common insulating shell 226B establishes the two active volumes 220B-1 and 220B-2 between respective entrance windows 224-1 and 224-2. The irradiator material 210 is arranged in a bisecting, central plane of the semi-NiRD 200B-DE, sandwiched between the two active volumes 220B-1 and 220B-2. The entrance windows 224-1 and 224-2 are positioned on respective anodes 222-1-1 and 222-2-1 of the active volumes 220B- 1 and 220B-2, with the irradiator material 210 positioned therebetween. FIGs.16C-16D are cross-sectional side views depicting additional examples of semi- NiRDs 200B. The semi-NiRD 200B’ of FIG.16C is configured similarly as the semi-NiRD 200B of FIGs.15A-15B except the entrance window 224 is absent. Here, the irradiator material 77 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 210 is positioned directly on the anode 222-1 of the active volume 220B. The semi-NiRD 200B’- DE of FIG.16D is configured similarly as the semi-NiRD 200B-DE of FIG.16B except the entrance windows 224-1 and 224-2 are absent. Here, the irradiator material 210 is positioned directly on the anodes 222-1-1 and 222-2-1 of the active volumes 220B-1 and 220B-2. FIGs.17A-17B are various views depicting an example of a semi-NiRD 200B-MS having a microstructured configuration. FIG.17A is an isometric view of the semi-NiRD 200B- MS. FIG.17B is a cross-sectional side view of the semi-NiRD 200B-MS. In this example, the irradiator material 210 is embedded within the semiconductor material 600B. The irradiator material 210 is segmented into multiple pieces 210-1 through 210- N that are arranged into a two-dimensional array within the semiconductor material 600B. Each piece of the irradiator material 210 is a solid cylinder of the same size and shape, having a diameter ^^୍and total length ^^ୈcorresponding to the height of the semiconductor 600B. The cylindrical pieces are parallel to one another along the ^^-direction and extend through the semiconductor 600B to each of the electrodes 222-1 and 222-2. Adjacent pieces of the irradiator material 210 are spaced a distance of ^^୍in the ^^-direction and a distance of ^^୍in the ^^-direction.In some implementations, the pieces may be spaced equidistance from each other ^^୍ ൌ ^^୍ to forma square, symmetric array. As shown, each electrode 222 and the semiconductor 600B has a rectangular geometry with a length of ^^ୈin the ^^-direction and a width of ^^ୈin the ^^-direction. No entrance window 224 for the reaction products 520 is present as these are directly injected into the semiconductor 600B to generate electron-hole pairs 530-e / i. The microstructured semi- NiRD 200B-MS can be suitable for detection in dual (or isotropic) emission as both forward emitting 522-X particles 520-X and backward emitting 524-Y particles 520-X from neutron capture events 500-C can be detected in the active volume 220B. FIGs.18A-20B show examples of active volumes 220B for the semi-NiRD 200B that involve different electrical contacts (or junctions) between the electrodes 222 and the semiconductor material 600B. Note, the insulating shell 226B is omitted in FIGs.18A-20B for clarity. FIG.18A is a cross-sectional side view depicting an example of an active volume 200B- SS for the semi-NiRD 200B with two Schottky contacts, e.g., two rectifying contacts, formed between the electrodes 222 and the semiconductor material 600B. Here, both the first 222-1 and second 222-2 electrodes are metal layers 622-1 and 622-2 formed on the semiconductor 600B, 78 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO e.g., via CVD. For example, the electrodes 222-1 and 222-2 can each be composed of aluminum, chromium, titanium, nickel, gold, silver, copper, platinum, tungsten, an alloy thereof, or any suitable conducting material. In some implementations, one or both of the electrodes 222-1 and 222-2 may include multiple (e.g., two or three) metal layers 622 in a multilayer stack configuration. Examples of multilayer electrode 222 stacks include titanium / platinum / gold stacks, chromium / gold stacks, among others. In this case, the semiconductor 600B may be intrinsic (“i"), n-doped (“n”), or p-doped (“p”). FIG.18B is an example current-voltage curve ^^^^^) of an active volume 200B-SS with two Schottky contacts showing the double-diode current shape versus bias voltage. FIG.19A is a cross-sectional side view depicting an example of an active volume 220B- OO for the semi-NiRD 200B with two ohmic contacts, e.g., two non-rectifying contacts, formed between the electrodes 222 and the semiconductor material 600B. Here, both the first 222-1 and second 222-2 electrodes are implantation layers 621-1 and 621-2 formed on (or in) the semiconductor 600B, e.g., via ion implantation. In this example, the first electrode 222-1 is a heavily n-doped (“n+”) implantation layer 621-1 and the second electrode 222-2 is a heavily p- doped (“p+”) implantation layer 621-2. This produces a substantial number of free electrons and holes in each of the electrodes 222-1 and 222-2, respectively, to exhibit non-rectifying behavior. FIG.19B is an example current-voltage curve ^^^^^) of the active volume 220B-OO with two ohmic contacts showing the ohmic (linear) current shape versus bias voltage. FIG.20A is a cross-sectional side view depicting an example of an active volume 220B- SO for a semi-NiRD 200-B with a Schottky contact and an ohmic contact, e.g., one rectifying and one non-rectifying contact, formed between the electrodes 222 and the semiconductor material 600B. Here, the first electrode 222-1 is a heavily n- or p-doped implantation layer 621 and the second electrode 222-2 is a metal layer 622. FIG.20B is an example current-voltage curve ^^^^^) of the active volume 220B-SO with a Schottky and ohmic contact showing the single- diode current shape versus bias voltage. FIGs.21A-21F are cross-sectional side views depicting additional examples of active volumes 220B for the semi-NiRD 200B that involve semiconducting junctions in the bulk of the semiconductor material 600B (as well as other electrode configurations). Note, the insulating shell 226B is omitted in FIGs.21A-21F for clarity. 79 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIG.21A shows an example active volume 220B-PN1 with a PN junction formed in the semiconductor material 600B. Here, both the first 222-1 and second 222-2 electrodes are metal layers 622-1 and 622-2 formed on the semiconductor 600B. The semiconductor 600B includes a p-doped region in electrical contact with the first electrode 222-1 and an n-doped region in electrical contact with the second electrode 222-2. FIG.21B shows another example active volume 220B-PN2 with a PN junction formed in the semiconductor material 600B. Here, the first electrode 222-1 is a heavily n- or p-doped implantation layer 621 and the second electrode 222-2 is a metal layer 622. Similar as above, the semiconductor 600B includes a p-doped (or n-doped) region in electrical contact with the first electrode 222-1 and an n-doped (or p-doped) region in electrical contact with the second electrode 222-2. FIG.21C shows an example active volume 220B-PIN1 with a PIN junction formed in the semiconductor material 600B. Here, both the first 222-1 and second 222-2 electrodes are metal layers 622-1 and 622-2 formed on the semiconductor 600B. The semiconductor 600B includes a p-doped region in electrical contact with the first electrode 222-1, an n-doped region in electrical contact with the second electrode 222-2, and an intrinsic (undoped) region between the p- and n- doped regions. FIG.21D shows another example active volume 220B-PIN2 with a PIN junction formed in the semiconductor material 600B. Here, the first electrode 222-1 is a heavily n- or p-doped implantation layer 621 and the second electrode 222-2 is a metal layer 622. Similar as above, the semiconductor 600B includes a p-doped (or n-doped) region in electrical contact with the first electrode 222-1, an n-doped (or p-doped) region in electrical contact with the second electrode 222-2, and an intrinsic region between the p- and n-doped regions. FIG.21E shows an example active volume 220B-MOS1 with a metal-oxide- semiconductor (MOS) structure formed between the electrodes 222 and semiconductor material 600B. Here, the first electrode 222-1 is a gate electrode that includes an insulator layer 624 formed on the semiconductor 600B and a metal layer 622-1 formed on the insulator layer 624. The second electrode 222-2 is a body electrode that includes a metal layer 622-2 formed on the semiconductor 600B. In this case, the semiconductor 600B may be intrinsic, n-doped, or p- doped. 80 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIG.21F shows another example active volume 220B-MOS2 with an MOS structure formed between the electrodes 222 and semiconductor material 600B. Here, the first electrode 222-1 is a gate electrode that includes an insulator layer 624 formed on the semiconductor 600B and a heavily n- or p-doped semiconductor layer 623 formed on the insulator layer 624. The second electrode 222-2 is a body electrode that includes a metal layer 622 formed on the semiconductor 600B. Similar as above, the semiconductor 600B may be intrinsic, n-doped, or p- doped. Section V: Examples of Scintillation NiRDs (scin-NiRDs) FIGs.22A-22B are various views depicting an example of a scintillation NiRD (“scin- NiRD”) 200C. FIG.22A is an isometric view of the scin-NiRD 200C. FIG.17B is a cross- sectional side view of the scin-NiRD 200C. For the scin-NiRD 200C, the active volume 220C is defined by a reflecting cavity 226C enclosing a scintillator 600C. The reflecting cavity 226C may also partially or fully enclose a photodetector 204 coupled to the scintillator 600C. The photodetector 204 can be any suitable photodetector that is sensitive scintillation photons 530-ɣ such as a photodiode (e.g., FIGs.25A- 25B) or a photomultiplier (e.g., FIG.25C). The reflecting cavity 226C can be a mirror-backed reflector (e.g., a silvered film or glass reflector) or composed of a suitable highly reflective material such aluminum, gold, silver, or an alloy thereof. The reflecting cavity 226C can also be a thin film of reflecting paint (e.g., white reflecting paint) or dry reflecting powder (e.g., magnesium oxide powder) applied to the outer surfaces of the scintillator 600C. In any of these cases, the entrance window 224 may be composed of the same material as the reflecting cavity 226C or a different highly reflective material. The scintillator 600C can be any suitable organic or inorganic scintillator, which may be doped with one or more elements that activate the scintillator 600C. In some examples, the scintillator 600C is a liquid scintillator Examples of organic scintillators include anthracene, stilbene, and naphthalene. Examples of inorganic scintillators, e.g., intrinsic and doped scintillators, include barium fluoride, bismuth germanate, cadmium tungstate, calcium fluoride doped with europium, calcium iodide doped with thallium, calcium tungstate, cesium iodide, cesium iodide doped with sodium, cesium iodide doped with thallium, gallium arsenide doped with silicon and boron impurities, gadolinium oxysulfide, lanthanum bromide doped with cerium, lanthanum chloride doped with cerium, lead tungstate, lutetium iodide, LYSO 81 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO (Lu1.8Y0.2SiO5(Ce)), sodium iodide doped with thallium, yttrium aluminum garnet doped with cerium, zinc sulfide doped with silver, and zinc tungstate. Like the semiconductor material 600B, the scintillator 600C can be chosen to optimize various different properties for a particular species of reaction product(s) 520, e.g., the energy deposition (e.g., pulse shape) of the reaction product(s) 520, stopping distance of the reaction product(s) 520, number of scintillation photons produced, among other properties. As shown, the scin-NiRD 200C is rotationally symmetric about the central axis 170. The reflecting cavity 226C is positioned about and laterally bounds the scintillator 600C (and at least a portion of the photodetector 204), establishing the active volume 220C with a cylindrical shape. The scintillator 600C is a solid cylinder (or disk) having a height of ^^ୗୡand a diameter of ^^ୈ. The scintillator 600C is positioned on the photodetector 204 and forms an optical contact at an optical coupling surface 205, e.g., a cleaned, polished, applied with an optically clear adhesive (OCA), and / or refractive index matched interface. The photodetector 204 is shown here schematically (not to scale) and has a total length of ^^^୦. The reflecting cavity 226C has athickness of ^^ୖ, and a total height of ^^^^ ^ ^^ୗୡ ^ ^^^୦ along the central axis 170, where ^^^^ isthe thickness of the entrance window 224. Note, in cases when the photodetector 204 is a photomultiplier 204-PM, the reflecting cavity 226C may only bound the scintillator 600C. The entrance window 224 forms a top, planar piece of the reflecting cavity 226C that is orthogonal to the central axis 170. The entrance window 224 is positioned on and longitudinally bounds the scintillator 600C. The irradiator material 210 is a uniform layer in the shape of a circular dischaving a thickness of ^^୍ and a diameter of ^^ୈ ^ 2^^ୖ. The irradiator material 210 is positioned onthe entrance window 224 and has an incident surface 211 parallel thereto. The scin-NiRD 200C is configured for detection in forward emission. Particularly, the irradiator material 210 is arranged to receive incident neutrons 510-I upon its outward-facing incident surface 211. Hence, forward emitting reaction products 520 from neutron capture 500-C and scattering 500-S events are registered as counts in the active volume 220C. As shown, a reaction product 520-Z entering the scintillator 600C collides with atoms, leaving an ionizing track 526 along its trajectory. The atoms are excited by the particle 520-Z and then quickly decay (e.g., on the order of nanoseconds or less) from their excited states to generate scintillation photons 530-ɣ. The number (^^^^ୡ) of scintillation photons 530-ɣ produced is proportional to the energy of the particle 520-Z if it is fully stopped within the scintillator 600C. In general, the 82 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WOenergy (^^ ൌ ^^ି^) involved to produce a single scintillation 530-ɣ is proportional to the band gapof the scintillator 600C, but is generally larger than the band gap by a factor of about 2-15, with the theoretical and experimental limit being about 2-3. This variation in the conversion efficiency is because the generation of scintillation photons 530-ɣ is a relatively complicated (and potentially inefficient) process involving conversion of ionizing radiation to electron-hole pairs, transport and energy transfer of the electron-hole pairs to the luminescence centers, and luminescence at the luminescence centers where the electron-hole pairs recombine radiatively. For example, sodium iodide has a band gap of 5.89 eV at room temperature, and the average energy to produce a scintillation photon 530-ɣ is about ^^~26.3 eV when activated with thallium. Then, a particle 520-Z having an energy of one (1) MeV that is fully stopped by sodium iodide may create about 38,000 scintillation photons 530-ɣ. The time evolution of the number of emitted scintillation photons 530-ɣ in a single ionization event can often be described by a linear superposition of one or more exponential decays. For two decays, this may be expressed, at least approximately, as: ^^ୡ^ ^^^ ^^ (8.1) ^^ ^^ ൌ ^^ exp^െ ^ ^ ^^ exp ൬െ^, ^^ where ^^ and ^^^ respectively, and ^^^^ୡ^0^ ൌ ^^ ^ ^^ providing the total number of scintillon photons 530-ɣproduced. While the fast component usually dominates, the relative amplitude, ^^ and ^^, of the two components depend on the scintillator 600C. Both of these components can be functions of the LET (stopping power) such that the overall decay-time constant varies with the type of ionizing reaction product 520. Such scintillators 600C enable PSD as described above. In general, the scintillation photons 530-ɣ radiate isotopically from the atoms as they decay, which are confined by the reflecting cavity 226C. Scintillation photons 530-ɣ that are incident upon the optical coupling surface 205 are absorbed by the photodetector 204 and contribute to the electric current pulse (^^ௗ). Here, the bias voltage ^^^^ is applied across the ports IO1 and IO2 of the photodetector 204. As described above, the bias voltage generally configures the active volume 220C for pulse mode, which depends on the photodetector 204 implemented. For example, a photodiode 204-PD can be reverse biased while a photomultiplier 204-MP can be biased with a large negative voltage to accelerate and multiply photoelectrons. Examples of these are discussed below with reference to FIGs.25A-25C. 83 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIG.23A is a cross-sectional side view depicting an example of a scin-NiRD 200C-BE configured for detection in backward emission. The scin-NiRD 200C-BE is configured similarly as the scin-NiRD 200C of FIGs.22A-22B except it is arranged to receive incident neutrons 510-I upon an inward-facing incident surface 211 of the irradiator material 210. FIG.23B is a cross-sectional side view depicting an example of a scin-NiRD 200C-DE in a sandwich configuration for detection in dual emission. Like the semi-NiRD 200B-DE of FIG. 16B, the scin-NiRD 200C-DE utilizes a second active volume 220C-2 that is the mirror image of the first active volume 220C-1. As shown, the scin-NiRD 200C-DE uses the first active volume 220C-1 for detection in forward emission and the second active volume 220C-2 for detection in backward emission. The reflecting cavity 226C is divided into two separate cavities by respective entrance windows 224-1 and 224-2 which establishes the two active volumes 220C-1 and 220C-2. The irradiator material 210 is arranged in a bisecting, central plane of the dual- cavity 226C, sandwiched between the two active volumes 220C-1 and 220C-2. The entrance windows 224-1 and 224-2 are positioned on respective scintillators 600C-1 and 600C-2 of the active volumes 220B-1 and 220B-2, with the irradiator material 210 positioned therebetween. Each scintillator 600C-1 and 600C-2 is coupled to a respective photodetector 204-1 and 204-2 at a respective coupling surface 205-1 and 205-2 for absorbing the scintillation photons 530-ɣ produced from an ionizing event. In this example, the photodetectors 204-1 and 204-2 areconnected in parallel to combine the electric current pulses generated by each ^^ௗ ൌ ^^ௗ,^ ^ ^^ௗ,ଶ,where ^^ௗ,^and ^^ௗ,ଶare the electric current pulses from ionizing events in the first 220C-1 and second 220C-2 active volumes, respectively. The scin-NiRD 200C-DE can be well-suited for a fully solid-state implementation that uses two photodiodes 204-PD as photodetectors 204, as the structure may be fabricated using semiconductor device fabrication techniques. FIGs.23C-23D are cross-sectional side views depicting additional examples of scni- NiRDs 200C. The scin-NiRD 200C’ of FIG.23C is configured similarly as the scin-NiRD 200C of FIGs.22A-22B except the entrance window 224 is absent. Here, the irradiator material 210 is positioned directly on the scintillator 600C of the active volume 220C. The semi-NiRD 200C’- DE of FIG.23D is configured similarly as the scin-NiRD 200C-DE of FIG.23B except the entrance windows 224-1 and 224-2 are absent. Here, the irradiator material 210 is positioned directly on the scintillators 600C-1 and 600C-2 of the active volumes 220C-1 and 220C-2. 84 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO FIGs.24A-24B are various views depicting an example of a scin-NiRD 200C-MS having a microstructured configuration. FIG.24A is an isometric view of the scin-NiRD 200C-MS. FIG. 24B is a cross-sectional side view of the scin-NiRD 200C-MS. The scin-NiRD 200C-MS is configured similarly to the semi-NiRD 200B-MS in FIGs. 17A-17B. Particularly, the irradiator material 210 is embedded within the scintillator 600C. The irradiator material 210 is segmented into multiple pieces 210-1 through 210-N that are arranged into a two-dimensional array within the scintillator 600C. Each piece of the irradiator material 210 is a solid cylinder of the same size and shape, having a diameter ^^୍and total length ^^ୈcorresponding to the height of the scintillator 600C. The cylindrical pieces are parallel to one another along the ^^-direction and extend through the scintillator 600C, between a top wall of the reflecting cavity 226C and the photodetector 204. Adjacent pieces of the irradiator material 210 are spaced a distance of ^^୍in the ^^-direction and a distance of ^^୍in the ^^-direction. In someimplementations, the pieces may be spaced equidistance from each other ^^୍ ൌ ^^୍ to form asquare, symmetric array. As shown, the scintillator 600C has a rectangular geometry with a length of ^^ୈin the ^^-direction and a width of ^^ୈin the ^^-direction. No entrance window 224 for the reaction products 520 is present as these are directly injected into the scintillator 600C to generate scintillation photons 530-ɣ. The microstructured scin-NiRD 200C-MS can be suitable for detection in dual (or isotropic) emission as both forward emitting 522-X particles 520-X and backward emitting 524-Y particles 520-X from neutron capture events 500-C can be detected in the active volume 220C. FIGs.25A-25C are cross-sectional side views depicting examples of photodetectors 204 for a scin-NiRD 200C. FIG.25A-25B show examples of photodiodes 204-PD1 and 204-PD2, and FIG.26C shows an example of a photomultiplier 204-PM. FIG.25A shows an example photodiode 204-PD1 configured as PIN photodiode. The photodiode 204-PD1 is configured similarly as the active volume 220B-PIN1 of FIG.21C but for detecting low-energy scintillation photons 530-ɣ which generate electron-hole pairs via the photoelectric effect. Here, the photodiode 204-PD1 includes a first electrode 222-1 configured as an anode connected to port IO1, and a second electrode 222-2 configured as a cathode connected to port IO2, which is grounded 315. The photodiode 204-PD1 is reverse biased such that the electric current pulse (^^ௗ) resulting via collection of the electron-hole pairs at the electrodes 222- 1 and 222-2 includes little leakage (or dark) current. 85 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO As shown, the first 222-1 and second 222-2 electrodes are metal layers formed on a semiconductor material 620. For example, the semiconductor 620 can include any of those listed above for the semiconductor material 600B of the semi-NiRD 200B. The semiconductor 620 is positioned between the first 222-1 and second 222-2 electrodes and forms respective electrical contacts thereto. The first electrode 222-1 is a uniform layer. The second electrode 222-2 is alayer having a cross-section in the ^^ െ ^^ plane in the shape of an annulus. This forms anoptically transmissive gap for an optical entrance window 206 that is transparent to the scintillation photons 530-ɣ. In this case, the optical window 206 is an air gap but a suitable dielectric material such as glass, e.g., for refractive index matching, may also be used. The second electrode 222-2 laterally bounds an antireflective coating 229 and the optical window 206. The antireflective coating 229 is formed on the semiconductor 620 to reduce reflections at the semiconductor 620 interface. For example, the antireflective coating 229 can be composed of magnesium fluoride, silicon nitride, silicon dioxide, titanium dioxide, or aluminum oxide. The optical window 206 is formed on the antireflective coating 229. The semiconductor 620 includes a p-doped region in electrical contact with the second electrode 222-2, an n-doped region in electrical contact with the first electrode 222-1, and an intrinsic region between the n- and p- doped regions. FIG.25B shows an example photodiode 204-PD2 configured as an avalanche photodiode. The photodiode 204-PD2 is configured similarly to the photodiode 204-PD1 of FIG. 25A but using a different doping scheme. Particularly, from top to bottom, the semiconductor material 620 includes a heavily p-doped region in electrical contact with the second electrode 222-2, an intrinsic region, a p-doped region, and a heavily n-doped region in electrical contact with the first electrode 222-1. Generally, the photodiode 204-PD2 is designed to experience avalanche breakdown at a specified reverse bias voltage. FIG.25C shows an example photomultiplier 204-PM configured with a photomultiplier tube (PMT) 207. In solid-state implementations, a silicon photomultiplier (SiPM) may also be used as the photomultiplier 204-PM. For example, typical SiPMs include a large array of avalanche photodiodes with a density of up to 10,000 avalanche diodes per square millimeter. As shown, the PMT 207 is a glass container evacuated of ambient air to form a vacuum 650 therein, e.g., using a tight and durable glass-to-metal seal like other types of vacuum tubes. The PMT 207 encloses a first electrode 222-1 configured as a photocathode connected to port 86 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO IO1, and a second electrode 222-2 configured as an anode. The PMT 207 further encloses a focusing electrode 223 and a series of dynodes 225-1 through 225-N that form an electron multiplier. In this case, four dynodes 225-1, 225-2, 225-3, and 225-4 are shown but many more dynodes 225 may be used in practice, e.g., ten or more, fifteen more, twenty or more, etc. A voltage distribution across the dynodes 225 is produced by a voltage divider circuit that includes a series of resistors (^^). In the example, the photocathode 222-1 is held at a high negative voltage (e.g., ~ 1000 Volts), while the anode 222-2 is held near ground 315 by an output resistor (^^^). The voltage distribution on the dynodes 225 generates an electric field within the PMT 207 that can accelerate electrons therein. In some implementations, the voltage divider may also include capacitors connected in parallel with the respective resistors of the final dynodes 225, e.g., the final one, two, three, four, or five dynodes 225. These capacitors act as local reservoirs of charge to maintain the voltages on the dynodes 225 while secondary electrons 531-e propagate through the PMT 207. Note, many variations of this design may be used in practice, the configuration shown in FIG.25C is merely illustrative. In this example, the photomultiplier 204-PM is in a head-on or end-on (transmission mode) configuration where scintillation photons 530-ɣ enter the flat, circular optical window 206 of the PMT 207 and interact with the photocathode 222-1 positioned thereon. For example, the photocathode 222-1 can be a thin conducting layer deposited on the inner surface of the optical window 206. A scintillation photon 530-ɣ transmitted through the optical window 206 strikes the photocathode 222-1 ejecting one or more photoelectrons 530-e due to the photoelectric effect. The electron(s) 530-e pass through the focusing electrode 223 toward the dynodes 225, where the electron(s) 530-e are accelerated by the electric field and multiplied by each dynode 225. This generates an exponentially increasing number of secondary electrons 531-e via secondary emission. For example, each dynode 225 may produce an average of five secondary electrons for each incoming electron. Then, a series of twelve dynodes 225 will produce an average of 5^ଶ~10଼secondary electrons 531-e for each photoelectron 530-e. This results in a sharp electric current pulse (^^ௗ) across the output resistor. Such electron multiplication effects can improve the signal-to-noise ratio of the scin-NiRD 200C, as well reduce the amount of electronic amplification implemented by the readout electronics 300. 87 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO Section VI: Examples of Homogeneous NiRDs (homo-NiRDs) FIGs.26A-26C are cross-sectional side views depicting examples of homogenous NiRDs (“homo-NiRDs”) 200*. Homo-NiRDs 200* allow simultaneous generation and detection of the reaction products 520 within the same bulk medium of the irradiator material 210. Homo-NiRDs 200* involve gases 600A, semiconductor materials 600B, and scintillators 600C that include one or more isotopes of one or more of the elements of interest, while also being sensitive to the reaction products 520 of interest. The homo-NiRDs 200* are generally configured for detection of the reaction products 520 in dual (or isotropic) emission within the active volume 220*, which is the full volume of the NiRD 200* in these cases. FIG.26A shows an example of a homo-NiRD 200A* utilizing a gaseous ionization detection scheme. The homo-NiRD 200A* is configured similarly to the planar gas-NiRD 200A- P of FIG.14A except the irradiator material 210 is also the fill gas 600A. As an example, the irradiator material 210 (600A) can include boron trifluoride (BF3), e.g., enriched to 96% boron- 10, to characterize the boron dose. As another example, the irradiator material 210 (600A) can include hydrogen gas (H2) or methane gas (CH4) to characterize the hydrogen dose. As yet another example, the irradiator material 210 (600A) can include nitrogen gas (N2) to characterize the nitrogen dose. Note, the homo-NiRD 200A* can also be configured similarly to the cylindrical gas-NiRD 200A and the spherical gas-NiRD 200A-S as the configuration in FIG. 26A is merely illustrative. FIG.26B shows an example of a homo-NiRD 200B* utilizing a semiconducting detection scheme. The homo-NiRD 200B* is configured similarly to the semi-NiRD 200B of FIGs.15A-15B except the irradiator material 210 is also the semiconductor material 600B. As an example, the irradiator material 210 (600B) can be composed of boron nitride (BN) to characterize the boron and nitrogen doses. As another example, the irradiator material 210 (600B) can be composed of boron carbide (BC4) to characterize the boron and / or carbon doses. As yet another example, the irradiator material 210 (600B) can be composed of diamond (C), e.g., synthetic diamond, to characterize the carbon dose in isolation. FIG.26C shows an example of a homo-NiRD 200C* utilizing a scintillation detection scheme. The homo-NiRD 200C* is configured similarly to the scin-NiRD 200C of FIGs.22A- 22B except the irradiator material 210 is also the scintillator 600C. As an example, to characterize the boron dose, the irradiator material 210 (600C) can be a liquid organic scintillator 88 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO loaded with boron-10 or a glass scintillator that is cerium-activated with boron silicates. As another example, the irradiator material 210 (600C) can be composed of sodium iodide doped with thallium to characterize the sodium dose. As yet another example, the irradiator material 210 (600C) can be composed of calcium iodide doped with thallium to characterize the calcium dose. Various aspects of the present subject matter are set forth below, in review of, and / or in supplementation to, the examples described thus far, with the emphasis here being on the interrelation and interchangeability of the following embodiments and the aspects thereof. In other words, an emphasis is on the fact that each aspect of the embodiments can be combined with each and every other aspect unless explicitly stated or taught otherwise. In a first group of embodiments, a neutron-induced radiation detector (NiRD) is provided, the NiRD including: an active volume defined by an enclosure including an entrance window transparent to ionizing radiation; an absorbing medium sensitive to the ionizing radiation, the absorbing medium at least partially filling the enclosure; and a layer of an irradiator material positioned on the entrance window, opposite the absorbing medium, where the irradiator material includes one or more elements selected from the group consisting of: boron, gadolinium, and non-trace elements of the human body. In some embodiments of the first group, the absorbing medium is sensitive to ionizing radiation comprised of gamma rays and / or one or more species of charged particles, and the entrance window is transparent to the ionizing radiation. In some embodiments of the first group, the non-trace elements of the human body consist of: hydrogen, carbon, nitrogen, oxygen, sodium, magnesium, phosphorus, sulfur, chlorine, potassium, and calcium. In some embodiments of the first group, the irradiator material is composed of the one or more elements. In some embodiments of the first group, the irradiator material is composed of boron nitride or polyethylene. In some embodiments of the first group, the irradiator material is monocrystalline. In some embodiments of the first group, the layer of the irradiator material has a thickness in a range from 10 nanometers (nm) to 10 microns (µm). 89 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the first group, the layer of the irradiator material has a total surface area in a range from 1 millimeter-squared (mm2) to 500 centimeter-squared (cm2). In some embodiments of the first group, the NiRD further includes a gamma ray shield enclosing the active volume and irradiator material. In some embodiments of the first group, the gamma ray shield is a hollow cap. In some embodiments of the first group, the gamma ray shield is composed of lead. In some embodiments of the first group, the gamma ray shield has a minimum thickness in a range from 1 millimeter (mm) to 10 centimeters (cm). In some embodiments of the first group, the NiRD has rotational symmetry about a central axis. In some embodiments of the first group, the enclosure is positioned laterally about the central axis; the entrance window is a planar piece of the enclosure that is orthogonal to the central axis; and the layer of the irradiator material has an incident surface orthogonal to the central axis. In some embodiments of the first group, the active volume includes: a first electrode; and a second electrode spaced apart from the first electrode, where the absorbing medium is located between the first and second electrodes. In some embodiments of the first group, the first electrode is an anode, and the second electrode is a cathode. In some embodiments of the first group, the cathode is grounded. In some embodiments of the first group, the enclosure is a chamber having the first electrode, second electrode, and absorbing medium arranged therein; and the absorbing medium is a gas dispersed between the first and second electrodes. In some embodiments of the first group, the chamber is composed of glass, fiberglass, or aluminum. In some embodiments of the first group, the entrance window is composed of aluminum or a silicate. In some embodiments of the first group, the silicate is a mica. In some embodiments of the first group, the entrance window has a thickness in a range from 1 µm to 1 mm. 90 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the first group, the first and second electrodes are composed of graphite, aluminum, tin, or stainless steel. In some embodiments of the first group, the gas includes air or an inert gas. In some embodiments of the first group, the inert gas is neon, argon, krypton, or xenon. In some embodiments of the first group, the gas includes a quench gas. In some embodiments of the first group, the quench gas is methane or a halogen. In some embodiments of the first group, the halogen is chlorine or bromide. In some embodiments of the first group, the gas has a pressure of one atmosphere (atm) or less. In some embodiments of the first group, the gas has a pressure of five atm or more. In some embodiments of the first group, the first electrode, second electrode, and chamber are cylindrical and concentric with the central axis. In some embodiments of the first group, the first electrode is a solid cylinder positioned on the central axis; the second electrode is a cylindrical shell positioned radially about the first electrode; and the chamber is positioned radially about the second electrode. In some embodiments of the first group, the chamber radially bounds the second electrode; and the entrance window longitudinally bounds the first electrode, second electrode, and gas therebetween. In some embodiments of the first group, the first electrode has an outer diameter in a range from 10 µm to 1 cm. In some embodiments of the first group, the second electrode has a radial thickness in a range from 10 µm to 1 cm. In some embodiments of the first group, the second electrode has an inner diameter in a range from 10 mm to 10 cm. In some embodiments of the first group, the first and second electrodes each have a length in a range from 10 mm to 25 cm. In some embodiments of the first group, the first and second electrodes are arranged in respective planes that are parallel to each other and orthogonal to the central axis. In some embodiments of the first group, each of the first and second electrodes is a respective parallel plate; and the chamber is positioned laterally about the first electrode, second electrode, and gas therebetween. 91 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the first group, the chamber laterally bounds about the first electrode, second electrode, and gas therebetween; and the entrance window longitudinally bounds the first electrode. In some embodiments of the first group, the first electrode, second electrode, entrance window, and layer of the irradiator material each have a circular or rectangular cross-section orthogonal to the central axis. In some embodiments of the first group, the first and second electrodes are positioned 1 mm to 20 cm apart. In some embodiments of the first group, the first and second electrodes each have a thickness in a range from 10 µm to 1 cm. In some embodiments of the first group, the first electrode, second electrode, entrance window, and layer of the irradiator material each have a maximum lateral dimension in range from 5 mm to 10 cm. In some embodiments of the first group, the enclosure is an insulating shell having the first electrode, second electrode, and absorbing medium arranged therein; and the absorbing medium is a semiconductor material positioned between the first and second electrodes and forming respective electrical contacts thereto. In some embodiments of the first group, the insulating shell and entrance window are composed of silicon dioxide, aluminum (III) oxide, titanium dioxide, hafnium (IV) oxide, zirconium dioxide, strontium peroxide, niobium pentoxide, gadolinium (III) oxide, or calcium dioxide. In some embodiments of the first group, the semiconductor material is composed of silicon, silicon carbide, thallium (I) bromide, germanium, gallium arsenide, gallium nitride, indium phosphide, indium gallium nitride, aluminum arsenide, aluminum gallium nitride, aluminum gallium arsenide, indium gallium arsenide, diamond, selenium, cadmium telluride, cadmium zinc telluride, boron nitride, or mercury (I) iodide. In some embodiments of the first group, the semiconductor material is monocrystalline. In some embodiments of the first group, the first and second electrodes are arranged in respective planes that are parallel to each other and orthogonal to the central axis. 92 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the first group, each of the first and second electrodes is a respective layer positioned on the semiconductor material; and the insulating shell is positioned laterally about the first electrode, second electrode, and semiconductor material therebetween. In some embodiments of the first group, the insulating shell laterally bounds the first electrode, second electrode, and semiconductor material therebetween; and the entrance window is a layer positioned on the first electrode. In some embodiments of the first group, the first electrode, second electrode, entrance window, irradiator material, and semiconductor material each have a circular or rectangular cross-section orthogonal to the central axis. In some embodiments of the first group, the first electrode is a metal layer. In some embodiments of the first group, the first electrode is composed of aluminum, chromium, titanium, nickel, gold, silver, copper, platinum, tungsten, or an alloy thereof. In some embodiments of the first group, the first electrode is an implantation layer of the semiconductor material. In some embodiments of the first group, the first electrode is a p+-doped implantation layer. In some embodiments of the first group, the second electrode is a metal layer. In some embodiments of the first group, the second electrode is composed of aluminum, chromium, titanium, nickel, gold, silver, copper, platinum, tungsten, or an alloy thereof. In some embodiments of the first group, the second electrode is an implantation layer of the semiconductor material. In some embodiments of the first group, the second electrode is an n+-doped implantation layer. In some embodiments of the first group, the semiconductor material is intrinsic (undoped), n-doped, or p-doped. In some embodiments of the first group, the semiconductor material includes: an n-doped region having the first electrode positioned thereon; and a p-doped region having the second electrode positioned thereon. In some embodiments of the first group, the n-doped and p-doped regions form a PN junction. 93 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the first group, the semiconductor material further includes an intrinsic (undoped) region between the n-doped and p-doped regions. In some embodiments of the first group, the intrinsic, n-doped, and p-doped regions form a PIN junction. In some embodiments of the first group, the first electrode, second electrode, semiconductor material, entrance window, and layer of the irradiator material each have a maximum lateral dimension in a range from 5 mm to 10 cm. In some embodiments of the first group, the semiconductor material has a thickness in a range from 1 µm to 500 µm. In some embodiments of the first group, the first and second electrodes each have a thickness in a range from 10 nm to 1 µm. In some embodiments of the first group, the entrance window has a thickness in a range from 1 nm to 100 nm. In some embodiments of the first group, the layer of the irradiator material has a thickness in a range from 10 nm to 100 nm. In some embodiments of the first group, the active volume includes a photodetector coupled to the absorbing medium. In some embodiments of the first group, the enclosure is a reflecting cavity having the photodetector and absorbing medium arranged therein; the photodetector includes an optical window transparent to scintillation photons; and the absorbing medium is a scintillator positioned on the optical window of the photodetector and forming an optical contact thereto. In some embodiments of the first group, the reflecting cavity and entrance window are composed of aluminum, gold, silver, or an alloy thereof. In some embodiments of the first group, the scintillator is an organic scintillator. In some embodiments of the first group, the organic scintillator is anthracene, stilbene, or naphthalene. In some embodiments of the first group, the scintillator is an inorganic scintillator. In some embodiments of the first group, the inorganic scintillator is thallium-doped sodium iodide or thallium-doped cesium iodide. In some embodiments of the first group, the photodetector is a photodiode. 94 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the first group, the photodiode is a PIN photodiode or an avalanche photodiode. In some embodiments of the first group, the photodetector is a photomultiplier. In some embodiments of the first group, the optical window is arranged in a plane parallel to the entrance window; the entrance window is a layer positioned on the scintillator; and the reflecting cavity laterally bounds the scintillator and at least a portion of the photodetector. In some embodiments of the first group, the photodiode includes: a first electrode arranged in the plane of the optical window; a second electrode separated from the first electrode, the second electrode arranged in a plane parallel to the plane of the optical window; and a semiconductor material positioned between the first and second electrodes and forming respective electrical contacts thereto. In some embodiments of the first group, each of the first and second electrodes is a respective layer positioned on the semiconductor material; and the first electrode has an annulus cross-section orthogonal to the central axis. In some embodiments of the first group, the photodiode further includes an antireflective coating arranged in the plane of the optical window, the antireflective coating positioned on the semiconductor material; the optical window is a layer positioned on the antireflective coating; and the first electrode laterally bounds the optical window and antireflective coating. In some embodiments of the first group, the optical window is composed of an air gap or a dielectric material. In some embodiments of the first group, the photomultiplier includes a photomultiplier tube having a series of electrodes arranged therein; and the optical window is a planar piece of the photomultiplier tube. In some embodiments of the first group, the series of electrodes includes: a photocathode, an anode, and a series of dynodes positioned between the photocathode and anode. In some embodiments of the first group, the photocathode is positioned on the optical window. In some embodiments of the first group, the series of electrodes further includes a focusing electrode positioned between the photocathode and a first of the dynodes. In some embodiments of the first group, the series of electrodes are electrically coupled by a voltage divider circuit. 95 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the first group, the active volume is a first active volume; the NiRD further includes a second active volume configured as the mirror image of the first active volume with respect to a central plane that is orthogonal to the central axis; and the layer of the irradiator material is arranged in the central plane and further positioned on the entrance window of the second active volume, opposite the absorbing medium of the second active volume. In some embodiments of the first group, the first electrodes of the first and second active volumes are electrically connected to each other; and the second electrodes of the first and second active volumes are electrically connected to each other. In a second group of embodiments, a NiRD having rotational symmetry about a central point is provided, the NiRD including: an active volume defined by a chamber enclosing: a first electrode; a second electrode spaced apart from the first electrode; and a gas dispersed between the first and second electrodes; and an irradiator material coupled to the active volume, where the irradiator material includes one or more elements selected from the group consisting of: boron, gadolinium, and non-trace elements of the human body. In some embodiments of the second group, the chamber, first electrode, second electrode, and irradiator material are spherical and concentric with the central point. In some embodiments of the second group, the irradiator material is a solid sphere positioned on the central point; the first electrode is a spherical shell positioned about the irradiator material; and the second electrode is a spherical shell positioned about the first electrode. In some embodiments of the second group, the irradiator material has an outer diameter in a range from 10 µm to 1 cm. In some embodiments of the second group, the first electrode has a radial thickness in a range from 10 µm to 1 cm. In some embodiments of the second group, the second electrode has a radial thickness in a range from 10 µm to 1 cm. In some embodiments of the second group, the second electrode has an inner diameter in a range from 10 mm to 10 cm. In some embodiments of the second group, the chamber includes an entrance window transparent to the ionizing radiation; and the entrance window is positioned radially on the irradiator material. 96 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the second group, the entrance window has a radial thickness in a range from 10 µm to 1 mm. In a third group of embodiments, a NiRD is provided, the NiRD including: an active volume defined by an enclosure filled, at least partially, with an absorbing medium; and an irradiator material embedded within the absorbing medium, where the irradiator material includes one or more elements selected from the group consisting of: boron, gadolinium, and non-trace elements of the human body. In some embodiments of the third group, the irradiator material is segmented into multiple pieces. In some embodiments of the third group, the pieces of the irradiator material are arranged into an array within the absorbing medium. In some embodiments of the third group, the pieces of the irradiator material are arranged equidistant from one another in the array. In some embodiments of the third group, each piece of the irradiator material has a same size and shape. In some embodiments of the third group, each piece of the irradiator material has a cylindrical shape. In some embodiments of the third group, the cylindrical pieces of the irradiator material are parallel to one another. In some embodiments of the third group, the enclosure is an insulating shell, and the absorbing medium is a semiconductor material arranged therein. In some embodiments of the third group, the active volume includes: a first electrode; and a second electrode spaced apart from the first electrode, where the semiconductor material is positioned between the first and second electrodes and forms respective electrical contacts thereto. In some embodiments of the third group, the enclosure is a reflecting cavity, and the absorbing medium is a scintillator arranged therein. In some embodiments of the third group, the active volume includes a photodetector coupled to the scintillator. 97 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the third group, the photodetector includes an optical window transparent to scintillation photons; and the scintillator is positioned on the optical window of the photodetector and forms an optical contact thereto. In a fourth group of embodiments, a NiRD is provided, the NiRD including: an active volume defined by an enclosure filled, at least partially, with an irradiator material, where the irradiator material includes one or more elements selected from the group consisting of: boron, gadolinium, and non-trace elements of the human body. In some embodiments of the fourth group, the active volume includes: a first electrode; and a second electrode spaced apart from the first electrode, where the irradiator material is located between the first and second electrodes. In some embodiments of the fourth group, the enclosure is a chamber having the first electrode, second electrode, and irradiator material arranged therein; and the irradiator material is a gas dispersed between the first and second electrodes. In some embodiments of the fourth group, the enclosure is an insulating shell having the first electrode, second electrode, and irradiator material arranged therein; and the irradiator material is a semiconductor material positioned between the first and second electrodes and forming respective electrical contacts thereto. In some embodiments of the fourth group, the active volume includes a photodetector coupled to the irradiator material. In some embodiments of the fourth group, the enclosure is a reflecting cavity having the photodetector and irradiator material arranged therein; the photodetector includes an optical window transparent to scintillation photons; and the irradiator material is a scintillator positioned on the optical window of the photodetector and forming an optical contact thereto. In a fifth group of embodiments, a radiation dosimetry system is provided, the radiation dosimetry system including: the NiRD of any of the embodiments of the first, second, third, or fourth groups; and readout electronics electrically coupled to the active volume, where the readout electronics are configured, during irradiation of the irradiator material with neutrons, to: receive a series of electrical signals generated by the active volume; and process the electrical signals to determine a respective neutron-induced dose for an isotope of each element corresponding to a respective neutron reaction with the isotope. 98 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the fifth group, the series of electric signals are a series of electric current pulses, and the readout electronics includes: a bias supply configured to apply a bias voltage to the active volume; a preamplifier configured to: receive each electric current pulse from the active volume; and convert the electric current pulse into a respective voltage pulse; an amplifier configured to: receive each voltage pulse from the preamplifier; and amplify the voltage pulse to generate a respective amplified voltage pulse; and a multichannel analyzer (MCA) configured to: receive each amplified voltage pulse from the amplifier; and analyze each of the amplified voltage pulses to determine each neutron-induced dose. In some embodiments of the fifth group, the bias supply is a high-voltage power supply. In some embodiments of the fifth group, the preamplifier is a resistive feedback charge- sensitive preamplifier. In some embodiments of the fifth group, the amplifier is an electrometer. In some embodiments of the fifth group, the MCA includes: an analog-to-digital converter (ADC), a field programmable gate array (FPGA), and a communications interface. In some embodiments of the fifth group, the ADC has a bit resolution of 14 bits or more, and a sampling rate of 60 million samples per second (MSPS) or more. In some embodiments of the fifth group, the NiRD is a first NiRD; the radiation dosimetry system further includes one or more additional NiRDs each configured according to any of the embodiments of the first, second, third, or fourth groups; and the readout electronics are electrically coupled to the active volume of each additional NiRD. In some embodiments of the fifth group, the irradiator material of each NiRD is composed of a different material. In some embodiments of the fifth group, the NiRDs include: a NiRD having an irradiator material composed of boron nitride; and a NiRD having an irradiator material composed of polyethylene. In some embodiments of the fifth group, the radiation dosimetry system further includes a gamma ray detector electrically coupled to the readout electronics. In some embodiments of the fifth group, the gamma ray detector is a gamma ray spectrometer. 99 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO In some embodiments of the fifth group, the radiation dosimetry system further includes a switchboard electrically coupled between the readout electronics, the gamma ray detector, and each NiRD. In a sixth group of embodiments, a beam system is provided, the beam system including: the radiation dosimetry system of any embodiment of the fifth group; a particle accelerator; a neutron-generating target; and a beamline extending from the accelerator to the neutron- generating target. In some embodiments of the sixth group, the particle accelerator is configured to output a charged particle beam, the beamline is configured to propagate the charged particle beam to the neutron-generating target, the neutron-generating target is configured to receive the charged particle beam and generate a neutron beam, and the radiation dosimetry system is configured to receive the neutron beam and measure one or more neutron-induced doses of the neutron beam. In some embodiments of the sixth group, the charged particle beam is a proton beam or a deuteron beam. In some embodiments of the sixth group, the neutron-generating target is composed of lithium or beryllium. In some embodiments of the sixth group, the beam system is configured for boron neutron capture therapy (BNCT). In a seventh group of embodiments, a method for neutron-induced radiation dosimetry of a neutron source is provided, the method including: irradiating an irradiator material with neutrons generated by the neutron source, where the irradiator material includes one or more elements selected from the group consisting of: boron, gadolinium, and non-trace elements of the human body; receiving, using readout electronics, a series of electrical signals generated by an active volume coupled to the irradiator material; and processing, using the readout electronics, each of the electrical signals to determine a neutron-induced dose for an isotope of an element of the irradiator material corresponding to a neutron reaction with the isotope. In some embodiments of the seventh group, the method further includes, for each of one or more additional, different irradiator materials coupled to a respective active volume: irradiating the irradiator material with neutrons generated by the neutron source, where the irradiator material includes one or more respective elements selected from the group consisting of: boron, gadolinium, and the non-trace elements of the human body; receiving, using the 100 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO readout electronics, a respective series of electrical signals generated by the respective active volume coupled to the irradiator material; and processing, using the readout electronics, each of the respective electrical signals to determine a respective neutron-induced dose for an isotope of a respective element of the irradiator material corresponding to a respective neutron reaction with the isotope. In some embodiments of the seventh group, the method further includes: irradiating a gamma ray detector with gamma rays generated by the neutron source; receiving, using the readout electronics, a series of electrical signals generated by the gamma ray detector; and processing, using the readout electronics, each of the electrical signals generated by the gamma ray detector to determine a gamma ray dose of the neutron source. In some embodiments of the seventh group, the method further includes: calculating, using the readout electronics, an equivalent dose of the neutron source from the gamma ray dose and each of the neutron-induced doses. In some embodiments of the seventh group, the equivalent dose is a weighted sum of the gamma ray dose and each of the neutron-induced doses. In some embodiments of the seventh group, each weight of the weighted sum is a relative biological effectiveness (RBE) value or a combined biological effectiveness (CBE) value. In some embodiments of the seventh group, the equivalent dose includes: a gamma ray dose, a neutron-induced boron dose, a neutron-induced nitrogen dose, and a neutron-induced hydrogen dose. While this specification contains many specific implementation details, these should not be construed as limitations on the scope of what is being claimed, which is defined by the claims themselves, but rather as descriptions of features that may be specific to particular embodiments of particular inventions. Certain features that are described in this specification in the context of separate embodiments can also be implemented in combination in a single embodiment. Conversely, various features that are described in the context of a single embodiment can also be implemented in multiple embodiments separately or in any suitable subcombination. Moreover, although features may be described above as acting in certain combinations and even initially be claimed as such, one or more features from a claimed combination can in some cases be excised from the combination, and the claim may be directed to a subcombination or variation of a subcombination. 101 LEGAL02 / 46174409v1 Attorney Docket No.069415 / 632275 / TLIFE.0006.WO Similarly, while operations are depicted in the drawings and recited in the claims in a particular order, this by itself should not be understood as requiring that such operations be performed in the particular order shown or in sequential order, or that all illustrated operations be performed, to achieve desirable results. In certain circumstances, multitasking and parallel processing may be advantageous. Moreover, the separation of various system modules and components in the embodiments described above should not be understood as requiring such separation in all embodiments, and it should be understood that the described program components and systems can generally be integrated together in a single software product or packaged into multiple software products. Particular embodiments of the subject matter have been described. Other embodiments are within the scope of the following claims. For example, the actions recited in the claims can be performed in a different order and still achieve desirable results. As one example, the processes depicted in the accompanying figures do not necessarily require the particular order shown, or sequential order, to achieve desirable results. 102 LEGAL02 / 46174409v1

Claims

Attorney Docket No.069415 / 632275 / TLIFE.0006.WO CLAIMS What is claimed is:

1. A neutron-induced radiation detector (NiRD), comprising: an active volume defined by an enclosure comprising an entrance window; an absorbing medium at least partially filling the enclosure; and a layer of an irradiator material positioned on the entrance window, opposite the absorbing medium, wherein the irradiator material comprises one or more elements selected from the group consisting of: boron, gadolinium, and non-trace elements of the human body.

2. The NiRD of claim 1, wherein the absorbing medium is sensitive to ionizing radiation comprised of gamma rays and / or one or more species of charged particles, and the entrance window is transparent to the ionizing radiation.

3. The NiRD of any preceding claim, wherein the non-trace elements of the human body consist of: hydrogen, carbon, nitrogen, oxygen, sodium, magnesium, phosphorus, sulfur, chlorine, potassium, and calcium.

4. The NiRD of claim 3, wherein the irradiator material is composed of the one or more elements.

5. The NiRD of claim 4, wherein the irradiator material is composed of boron nitride or polyethylene.

6. The NiRD of any preceding claim, wherein the irradiator material is monocrystalline.

7. The NiRD of any preceding claim, wherein the layer of the irradiator material has a thickness in a range from 10 nanometers (nm) to 10 microns (µm).

8. The NiRD of any preceding claim, wherein the layer of the irradiator material has a total surface area in a range from 1 millimeter-squared (mm2) to 500 centimeter-squared (cm2). 103 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 9. The NiRD of any preceding claim, further comprising a gamma ray shield enclosing the active volume and irradiator material.

10. The NiRD of claim 9, wherein the gamma ray shield is a hollow cap.

11. The NiRD of any of claims 9-10, wherein the gamma ray shield is composed of lead.

12. The NiRD of any of claims 9-11, wherein the gamma ray shield has a minimum thickness in a range from 1 millimeter (mm) to 10 centimeters (cm).

13. The NiRD of any preceding claim, having rotational symmetry about a central axis.

14. The NiRD of claim 13, wherein: the enclosure is positioned laterally about the central axis; the entrance window is a planar piece of the enclosure that is orthogonal to the central axis; and the layer of the irradiator material has an incident surface orthogonal to the central axis.

15. The NiRD of any preceding claim, wherein the active volume comprises: a first electrode; and a second electrode spaced apart from the first electrode, wherein the absorbing medium is located between the first and second electrodes.

16. The NiRD of claim 15, wherein the first electrode is an anode, and the second electrode is a cathode.

17. The NiRD of claim 16, wherein the cathode is grounded.

18. The NiRD of any of claims 15-17, wherein: 104 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO the enclosure is a chamber having the first electrode, second electrode, and absorbing medium arranged therein; and the absorbing medium is a gas dispersed between the first and second electrodes.

19. The NiRD of claim 18, wherein the chamber is composed of glass, fiberglass, or aluminum.

20. The NiRD of any of claims 18-19, wherein the entrance window is composed of aluminum or a silicate.

21. The NiRD of claim 20, wherein the silicate is a mica.

22. The NiRD of any of claims 18-21, wherein the entrance window has a thickness in a range from 1 µm to 1 mm.

23. The NiRD of any of claims 18-22, wherein the first and second electrodes are composed of graphite, aluminum, tin, or stainless steel.

24. The NiRD of any of claims 18-23, wherein the gas comprises air or an inert gas.

25. The NiRD of claim 24, wherein the inert gas is neon, argon, krypton, or xenon.

26. The NiRD of any of claims 18-25, wherein the gas comprises a quench gas.

27. The NiRD of claim 26, wherein the quench gas is methane or a halogen.

28. The NiRD of claim 27, wherein the halogen is chlorine or bromide.

29. The NiRD of any of claims 18-28, wherein the gas has a pressure of one atmosphere (atm) or less. 105 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 30. The NiRD of any of claims 18-28, wherein the gas has a pressure of five atm or more.

31. The NiRD of any of claims 18-30, wherein the first electrode, second electrode, and chamber are cylindrical and concentric with the central axis.

32. The NiRD of claim 31, wherein: the first electrode is a solid cylinder positioned on the central axis; the second electrode is a cylindrical shell positioned radially about the first electrode; and the chamber is positioned radially about the second electrode.

33. The NiRD of claim 32, wherein: the chamber radially bounds the second electrode; and the entrance window longitudinally bounds the first electrode, second electrode, and gas therebetween.

34. The NiRD of any of claims 32-33, wherein the first electrode has an outer diameter in a range from 10 µm to 1 cm.

35. The NiRD of any of claims 32-34, wherein the second electrode has a radial thickness in a range from 10 µm to 1 cm.

36. The NiRD of any of claims 32-35, wherein the second electrode has an inner diameter in a range from 10 mm to 10 cm.

37. The NiRD of any of claims 32-36, wherein the first and second electrodes each have a length in a range from 10 mm to 25 cm.

38. The NiRD of any of claims 18-30, wherein the first and second electrodes are arranged in respective planes that are parallel to each other and orthogonal to the central axis.

39. The NiRD of claim 38, wherein: 106 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO each of the first and second electrodes is a respective parallel plate; and the chamber is positioned laterally about the first electrode, second electrode, and gas therebetween.

40. The NiRD of claim 39, wherein: the chamber laterally bounds about the first electrode, second electrode, and gas therebetween; and the entrance window longitudinally bounds the first electrode.

41. The NiRD of any of claims 39-40, wherein the first electrode, second electrode, entrance window, and layer of the irradiator material each have a circular or rectangular cross-section orthogonal to the central axis.

42. The NiRD of any of claims 39-41, wherein the first and second electrodes are positioned 1 mm to 20 cm apart.

43. The NiRD of any of claims 39-42, wherein the first and second electrodes each have a thickness in a range from 10 µm to 1 cm.

44. The NiRD of any of claims 39-43, wherein the first electrode, second electrode, entrance window, and layer of the irradiator material each have a maximum lateral dimension in range from 5 mm to 10 cm.

45. The NiRD of any of claims 15-17, wherein: the enclosure is an insulating shell having the first electrode, second electrode, and absorbing medium arranged therein; and the absorbing medium is a semiconductor material positioned between the first and second electrodes and forming respective electrical contacts thereto. 107 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 46. The NiRD of claim 45, wherein the insulating shell and entrance window are composed of silicon dioxide, aluminum (III) oxide, titanium dioxide, hafnium (IV) oxide, zirconium dioxide, strontium peroxide, niobium pentoxide, gadolinium (III) oxide, or calcium dioxide.

47. The NiRD of any of claims 45-46, wherein the semiconductor material is composed of silicon, silicon carbide, thallium (I) bromide, germanium, gallium arsenide, gallium nitride, indium phosphide, indium gallium nitride, aluminum arsenide, aluminum gallium nitride, aluminum gallium arsenide, indium gallium arsenide, diamond, selenium, cadmium telluride, cadmium zinc telluride, boron nitride, or mercury (I) iodide.

48. The NiRD of any of claims 45-47, wherein the semiconductor material is monocrystalline.

49. The NiRD of any of claims 45-48, wherein the first and second electrodes are arranged in respective planes that are parallel to each other and orthogonal to the central axis.

50. The NiRD of claim 49, wherein: each of the first and second electrodes is a respective layer positioned on the semiconductor material; and the insulating shell is positioned laterally about the first electrode, second electrode, and semiconductor material therebetween.

51. The NiRD of claim 50, wherein: the insulating shell laterally bounds the first electrode, second electrode, and semiconductor material therebetween; and the entrance window is a layer positioned on the first electrode.

52. The NiRD of any of claims 50-51, wherein the first electrode, second electrode, entrance window, irradiator material, and semiconductor material each have a circular or rectangular cross-section orthogonal to the central axis.

53. The NiRD of any of claims 50-52, wherein the first electrode is a metal layer. 108 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 54. The NiRD of claim 53, wherein the first electrode is composed of aluminum, chromium, titanium, nickel, gold, silver, copper, platinum, tungsten, or an alloy thereof.

55. The NiRD of any of claims 50-52, wherein the first electrode is an implantation layer of the semiconductor material.

56. The NiRD of claim 55, wherein the first electrode is a p+-doped implantation layer.

57. The NiRD of any of claims 50-56, wherein the second electrode is a metal layer.

58. The NiRD of claim 57, wherein the second electrode is composed of aluminum, chromium, titanium, nickel, gold, silver, copper, platinum, tungsten, or an alloy thereof.

59. The NiRD of any of claims 50-56, wherein the second electrode is an implantation layer of the semiconductor material.

60. The NiRD of claim 59, wherein the second electrode is an n+-doped implantation layer.

61. The NiRD of any of claims 50-60, wherein the semiconductor material is intrinsic (undoped), n-doped, or p-doped.

62. The NiRD of any of claims 50-60, wherein the semiconductor material comprises: an n-doped region having the first electrode positioned thereon; and a p-doped region having the second electrode positioned thereon.

63. The NiRD of claim 62, wherein the n-doped and p-doped regions form a PN junction.

64. The NiRD of claim 62, wherein the semiconductor material further comprises an intrinsic (undoped) region between the n-doped and p-doped regions. 109 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 65. The NiRD of claim 63, wherein the intrinsic, n-doped, and p-doped regions form a PIN junction.

66. The NiRD of any of claims 50-65, wherein the first electrode, second electrode, semiconductor material, entrance window, and layer of the irradiator material each have a maximum lateral dimension in a range from 5 mm to 10 cm.

67. The NiRD of any of claims 50-66, wherein the semiconductor material has a thickness in a range from 1 µm to 500 µm.

68. The NiRD of any of claims 50-67, wherein the first and second electrodes each have a thickness in a range from 10 nm to 1 µm.

69. The NiRD of any of claims 50-68, wherein the entrance window has a thickness in a range from 1 nm to 100 nm.

70. The NiRD of any of claims 50-68, wherein the layer of the irradiator material has a thickness in a range from 10 nm to 100 nm.

71. The NiRD of any of claims 1-14, wherein the active volume comprises a photodetector coupled to the absorbing medium.

72. The NiRD of claim 71, wherein: the enclosure is a reflecting cavity having the photodetector and absorbing medium arranged therein; the photodetector comprises an optical window transparent to scintillation photons; and the absorbing medium is a scintillator positioned on the optical window of the photodetector and forming an optical contact thereto.

73. The NiRD of any of claim 72, wherein the reflecting cavity and entrance window are composed of aluminum, gold, silver, or an alloy thereof. 110 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 74. The NiRD of any of claims 71-73, wherein the scintillator is an organic scintillator.

75. The NiRD of claim 74, wherein the organic scintillator is anthracene, stilbene, or naphthalene.

76. The NiRD of any of claims 71-73, wherein the scintillator is an inorganic scintillator.

77. The NiRD of claim 76, wherein the inorganic scintillator is thallium-doped sodium iodide or thallium-doped cesium iodide.

78. The NiRD of any of claims 71-77, wherein the photodetector is a photodiode.

79. The NiRD of claim 78, wherein the photodiode is a PIN photodiode or an avalanche photodiode.

80. The NiRD of any of claims 71-77, wherein the photodetector is a photomultiplier.

81. The NiRD of any of claims 71-80, wherein: the optical window is arranged in a plane parallel to the entrance window; the entrance window is a layer positioned on the scintillator; and the reflecting cavity laterally bounds the scintillator and at least a portion of the photodetector.

82. The NiRD of claim 81 when also dependent on claim 78, wherein the photodiode comprises: a first electrode arranged in the plane of the optical window; a second electrode separated from the first electrode, the second electrode arranged in a plane parallel to the plane of the optical window; and a semiconductor material positioned between the first and second electrodes and forming respective electrical contacts thereto. 111 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 83. The NiRD of claim 82, wherein: each of the first and second electrodes is a respective layer positioned on the semiconductor material; and the first electrode has an annulus cross-section orthogonal to the central axis.

84. The NiRD of claim 83, wherein: the photodiode further comprises an antireflective coating arranged in the plane of the optical window, the antireflective coating positioned on the semiconductor material; the optical window is a layer positioned on the antireflective coating; and the first electrode laterally bounds the optical window and antireflective coating.

85. The NiRD of claim 84, wherein the optical window is composed of an air gap or a dielectric material.

86. The NiRD of claim 81 when also dependent on claim 80, wherein: the photomultiplier comprises a photomultiplier tube having a series of electrodes arranged therein; and the optical window is a planar piece of the photomultiplier tube.

87. The NiRD of claim 86, wherein the series of electrodes comprises: a photocathode, an anode, and a series of dynodes positioned between the photocathode and anode.

88. The NiRD of claim 87, wherein the photocathode is positioned on the optical window.

89. The NiRD of any of claims 87-88, wherein the series of electrodes further comprises a focusing electrode positioned between the photocathode and a first of the dynodes.

90. The NiRD of any of claims 86-89, wherein the series of electrodes are electrically coupled by a voltage divider circuit. 112 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 91. The NiRD of any of claims 13-90, wherein: the active volume is a first active volume; the NiRD further comprises a second active volume configured as the mirror image of the first active volume with respect to a central plane that is orthogonal to the central axis; and the layer of the irradiator material is arranged in the central plane and further positioned on the entrance window of the second active volume, opposite the absorbing medium of the second active volume.

92. The NiRD of claim 91 when also dependent on claim 15, wherein: the first electrodes of the first and second active volumes are electrically connected to each other; and the second electrodes of the first and second active volumes are electrically connected to each other.

93. A NiRD having rotational symmetry about a central point, the NiRD comprising: an active volume defined by a chamber enclosing: a first electrode; a second electrode spaced apart from the first electrode; and a gas dispersed between the first and second electrodes; and an irradiator material coupled to the active volume, wherein the irradiator material comprises one or more elements selected from the group consisting of: boron, gadolinium, and non-trace elements of the human body.

94. The NiRD of claim 93, wherein the chamber, first electrode, second electrode, and irradiator material are spherical and concentric with the central point.

95. The NiRD of claim 94, wherein: the irradiator material is a solid sphere positioned on the central point; the first electrode is a spherical shell positioned about the irradiator material; and the second electrode is a spherical shell positioned about the first electrode. 113 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 96. The NiRD of claim 95, wherein the irradiator material has an outer diameter in a range from 10 µm to 1 cm.

97. The NiRD of any of claims 95-96, wherein the first electrode has a radial thickness in a range from 10 µm to 1 cm.

98. The NiRD of any of claims 95-97, wherein the second electrode has a radial thickness in a range from 10 µm to 1 cm.

99. The NiRD of any of claims 95-98, wherein the second electrode has an inner diameter in a range from 10 mm to 10 cm.

100. The NiRD of any of claims 95-99, wherein: the chamber comprises an entrance window transparent to the ionizing radiation; and the entrance window is positioned radially on the irradiator material.

101. The NiRD of claim 100, wherein the entrance window has a radial thickness in a range from 10 µm to 1 mm.

102. A NiRD, comprising: an active volume defined by an enclosure filled, at least partially, with an absorbing medium; and an irradiator material embedded within the absorbing medium, wherein the irradiator material comprises one or more elements selected from the group consisting of: boron, gadolinium, and non-trace elements of the human body.

103. The NiRD of claim 102, wherein the irradiator material is segmented into a plurality of pieces.

104. The NiRD of claim 103, wherein the pieces of the irradiator material are arranged into an array within the absorbing medium. 114 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 105. The NiRD of claim 104, wherein the pieces of the irradiator material are arranged equidistant from one another in the array.

106. The NiRD of any of claims 103-105, wherein each piece of the irradiator material has a same size and shape.

107. The NiRD of claim 106, wherein each piece of the irradiator material has a cylindrical shape.

108. The NiRD of claim 107, wherein the cylindrical pieces of the irradiator material are parallel to one another.

109. The NiRD of any of claims 102-108, wherein the enclosure is an insulating shell, and the absorbing medium is a semiconductor material arranged therein.

110. The NiRD of claim 109, wherein the active volume comprises: a first electrode; and a second electrode spaced apart from the first electrode, wherein the semiconductor material is positioned between the first and second electrodes and forms respective electrical contacts thereto.

111. The NiRD of any of claims 102-108, wherein the enclosure is a reflecting cavity, and the absorbing medium is a scintillator arranged therein.

112. The NiRD of claim 111, wherein the active volume comprises a photodetector coupled to the scintillator.

113. The NiRD of claim 112, wherein: the photodetector comprises an optical window transparent to scintillation photons; and 115 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO the scintillator is positioned on the optical window of the photodetector and forms an optical contact thereto.

114. A NiRD, comprising: an active volume defined by an enclosure filled, at least partially, with an irradiator material, wherein the irradiator material comprises one or more elements selected from the group consisting of: boron, gadolinium, and non-trace elements of the human body.

115. The NiRD of claim 114, wherein the active volume comprises: a first electrode; and a second electrode spaced apart from the first electrode, wherein the irradiator material is located between the first and second electrodes.

116. The NiRD of claim 115, wherein: the enclosure is a chamber having the first electrode, second electrode, and irradiator material arranged therein; and the irradiator material is a gas dispersed between the first and second electrodes.

117. The NiRD of claim 115, wherein: the enclosure is an insulating shell having the first electrode, second electrode, and irradiator material arranged therein; and the irradiator material is a semiconductor material positioned between the first and second electrodes and forming respective electrical contacts thereto.

118. The NiRD of claim 117, wherein the active volume comprises a photodetector coupled to the irradiator material.

119. The NiRD of claim 118, wherein: the enclosure is a reflecting cavity having the photodetector and irradiator material arranged therein; the photodetector comprises an optical window transparent to scintillation photons; and 116 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO the irradiator material is a scintillator positioned on the optical window of the photodetector and forming an optical contact thereto.

120. A radiation dosimetry system, comprising: the NiRD of any preceding claim; and readout electronics electrically coupled to the active volume, wherein the readout electronics are configured, during irradiation of the irradiator material with neutrons, to: receive a series of electrical signals generated by the active volume; and process the electrical signals to determine a respective neutron-induced dose for an isotope of each element corresponding to a respective neutron reaction with the isotope.

121. The radiation dosimetry system of claim 120, wherein the series of electric signals are a series of electric current pulses, and the readout electronics comprises: a bias supply configured to apply a bias voltage to the active volume; a preamplifier configured to: receive each electric current pulse from the active volume; and convert the electric current pulse into a respective voltage pulse; an amplifier configured to: receive each voltage pulse from the preamplifier; and amplify the voltage pulse to generate a respective amplified voltage pulse; and a multichannel analyzer (MCA) configured to: receive each amplified voltage pulse from the amplifier; and analyze each of the amplified voltage pulses to determine each neutron-induced dose.

122. The radiation dosimetry system of claim 121, wherein the bias supply is a high-voltage power supply.

123. The radiation dosimetry system of any of claims 121-122, wherein the preamplifier is a resistive feedback charge-sensitive preamplifier. 117 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 124. The radiation dosimetry system of any of claims 121-123, wherein the amplifier is an electrometer.

125. The radiation dosimetry system of any of claims 121-124, wherein the MCA comprises: an analog-to-digital converter (ADC), a field programmable gate array (FPGA), and a communications interface.

126. The radiation dosimetry system of claim 125, wherein the ADC has a bit resolution of 14 bits or more, and a sampling rate of 60 million samples per second (MSPS) or more.

127. The radiation dosimetry system of any of claims 120-126, wherein: the NiRD is a first NiRD; the radiation dosimetry system further comprises one or more additional NiRDs each configured according to any of claims 1-113; and the readout electronics are electrically coupled to the active volume of each additional NiRD.

128. The radiation dosimetry system of claim 127, wherein the irradiator material of each NiRD is composed of a different material.

129. The radiation dosimetry system of claim 128, wherein the NiRDs include: a NiRD having an irradiator material composed of boron nitride; and a NiRD having an irradiator material composed of polyethylene.

130. The radiation dosimetry system of any of claims 127-129, further comprising a gamma ray detector electrically coupled to the readout electronics.

131. The radiation dosimetry system of claim 130, wherein the gamma ray detector is a gamma ray spectrometer.

132. The radiation dosimetry system of claim 131, further comprising a switchboard electrically coupled between the readout electronics, the gamma ray detector, and each NiRD. 118 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 133. A beam system, comprising: the radiation dosimetry system of any of claims 120-132; a particle accelerator; a neutron-generating target; and a beamline extending from the accelerator to the neutron-generating target.

134. The beam system of claim 133, wherein: the particle accelerator is configured to output a charged particle beam, the beamline is configured to propagate the charged particle beam to the neutron- generating target, the neutron-generating target is configured to receive the charged particle beam and generate a neutron beam, and the radiation dosimetry system is configured to receive the neutron beam and measure one or more neutron-induced doses of the neutron beam.

135. The beam system of claim 134, wherein the charged particle beam is a proton beam or a deuteron beam.

136. The beam system of any of claims 133-135, wherein the neutron-generating target is composed of lithium or beryllium.

137. The beam system of any of claims 133-136, configured for boron neutron capture therapy (BNCT).

138. A method for neutron-induced radiation dosimetry of a neutron source, the method comprising: irradiating an irradiator material with neutrons generated by the neutron source, wherein the irradiator material comprises one or more elements selected from the group consisting of: boron, gadolinium, and non-trace elements of the human body; receiving, using readout electronics, a series of electrical signals generated by an active volume coupled to the irradiator material; and 119 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO processing, using the readout electronics, each of the electrical signals to determine a neutron-induced dose for an isotope of an element of the irradiator material corresponding to a neutron reaction with the isotope.

139. The method of claim 138, further comprising, for each of one or more additional, different irradiator materials coupled to a respective active volume: irradiating the irradiator material with neutrons generated by the neutron source, wherein the irradiator material comprises one or more respective elements selected from the group consisting of: boron, gadolinium, and the non-trace elements of the human body; receiving, using the readout electronics, a respective series of electrical signals generated by the respective active volume coupled to the irradiator material; and processing, using the readout electronics, each of the respective electrical signals to determine a respective neutron-induced dose for an isotope of a respective element of the irradiator material corresponding to a respective neutron reaction with the isotope.

140. The method of claim 139, further comprising: irradiating a gamma ray detector with gamma rays generated by the neutron source; receiving, using the readout electronics, a series of electrical signals generated by the gamma ray detector; and processing, using the readout electronics, each of the electrical signals generated by the gamma ray detector to determine a gamma ray dose of the neutron source.

141. The method of claim 140, further comprising: calculating, using the readout electronics, an equivalent dose of the neutron source from the gamma ray dose and each of the neutron-induced doses.

142. The method of claim 141, wherein the equivalent dose is a weighted sum of the gamma ray dose and each of the neutron-induced doses. 120 LEGAL02 / 46174409v1Attorney Docket No.069415 / 632275 / TLIFE.0006.WO 143. The method of claim 142, wherein each weight of the weighted sum is a relative biological effectiveness (RBE) value or a combined biological effectiveness (CBE) value.

144. The method of claim 143, wherein the equivalent dose comprises: a gamma ray dose, a neutron-induced boron dose, a neutron-induced nitrogen dose, and a neutron-induced hydrogen dose. 121 LEGAL02 / 46174409v1

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