Design method of long-life in-pile neutron probe with gamma compensation

Abstract

A design method of gamma-compensated long-life in-core neutron probe is proposed to solve the problem of gamma noise interference in the mixed radiation field of nuclear reactors. The gamma current of the in-core neutron probe is calculated by solving the radiation field information of the in-core instrument tube and establishing the sensitivity separation model of the probe. Based on the decay constant and microscopic absorption cross section of the sensitive region of the probe, the expected service life of the probe is calculated. By adjusting the material and structure parameters, the optimal scheme of the probe that meets the design target is obtained. The in-core neutron probe has high signal-to-noise ratio and solves the problem of current interference in the gamma field from the principle design, which is conducive to improving the sustainable online monitoring capability of the core state parameters. The proposed design method is based on the mechanism of radiation-induced current of the probe and does not depend on the pre-experimental basis and manufacturing process upgrade, which is simple, efficient and feasible in engineering.

Description

Technical Field

[0001] This invention relates to the field of online radiation field monitoring and reactor core neutron detector technology, specifically to a design method for a gamma-compensated long-life reactor core neutron detector. Background Technology

[0002] Neutron radiation fields are always accompanied by the generation of gamma rays, and the differentiation between neutrons and gamma rays is a crucial technical problem that must be solved in neutron detection. The reactor core is a typical neutron-gamma mixed field, and its in-core neutron detectors mainly consist of fission ionization chambers based on nuclear fission and self-powered detectors based on neutron activation. The design of gamma-ray compensation electrodes significantly enhances the gamma-ray radiation resistance of ionization chamber detectors, but due to the high reaction energy and rapid burnup, fission ionization chambers cannot operate continuously in the in-core radiation field for extended periods, failing to meet the current requirements for online, long-lifetime neutron detection. Meanwhile, self-powered detectors based on activation methods are more susceptible to gamma-ray noise interference.

[0003] To address the extremely poor gamma-ray interference resistance of self-powered detectors, solutions such as background compensation core wires and axial current signal compensation have been applied in engineering practice. However, these measures only have a certain effect on the signal shielding of armored cables and cannot improve the gamma-ray resistance of the detector probe. For the design of self-powered detectors with n / γ discrimination, internationally, experimentally based low-Z material background current compensation methods and circuit signal adjustment-based external photon compensation methods have been proposed. Both approaches directly consider background current compensation through detector parameter adjustment; however, the first method suffers from a lack of theoretical guidance and significant design costs, while the second method has not yet solved the problem of measuring weak charges on insulators in practice. In recent years, with the development of neutron detector mechanism research and high-precision response calculation models, it has become possible to fundamentally solve gamma-ray noise interference in detector design. Furthermore, my country's nuclear reactors currently rely heavily on imports for in-core neutron detectors; therefore, the design of detector schemes with high signal-to-noise ratios and long lifespans plays a positive role in promoting the independent development of domestically produced detectors.

[0004] Against this backdrop, how to provide an effective design solution based on the response mechanism of in-core neutron detectors and target radiation fields to solve the problem of strong gamma noise in existing neutron detectors, thereby ensuring the accuracy of core measurements and online monitoring of state parameters, has become one of the real challenges restricting in-core neutron detection technology. Summary of the Invention

[0005] To overcome the problems existing in the prior art, the present invention aims to provide a design method for a gamma-compensated long-life in-reactor neutron detector to solve the problem of gamma-ray background noise current interference during neutron detection. Based on gamma-ray branch current calculation, the basic parameters of the detector are adjusted to compensate for the total gamma response current, and the expected operating lifetime of the detector is quantitatively calculated based on the microscopic absorption reaction rate in the detector's sensitive region, ensuring sustainable operation under target radiation field conditions.

[0006] To achieve the above objectives, the present invention adopts the following technical solution:

[0007] A design method for a gamma-compensated long-life in-reactor neutron detector includes the following steps:

[0008] Step 1: Read the geometry and material information of the fuel assembly containing the in-core instrument tubes, and solve the steady-state neutron transport equations to obtain the neutron and photon energy spectra at the in-core instrument tubes;

[0009] Step 2: Read the initial material and structural parameters of the in-pile neutron detector, establish the Monte Carlo model of the in-pile neutron detector based on the neutron energy spectrum and photon energy spectrum obtained in Step 1, calculate the neutron sensitivity and photon sensitivity of the in-pile neutron detector using formula (1) and formula (2) respectively, and calculate the instantaneous sensitivity coefficient of the in-pile neutron detector using formula (3).

[0010]

[0011]

[0012] In the formula:

[0013] S n —The neutron sensitivity of the in-pile neutron detector;

[0014] S γ —The photon sensitivity of the in-pile neutron detector;

[0015] Q n (t)——The effective charge generated by the in-pile neutron detector and neutron reaction at time t;

[0016] Q γ —The effective charge generated by the reaction of photons with the in-pile neutron detector;

[0017] φ n — Neutron flux density in the sensitive region of the in-pile neutron detector;

[0018] φ γ —Photon flux density in the sensitive region of the in-pile neutron detector;

[0019] L – Length of the in-pile neutron detector;

[0020] t — time;

[0021]

[0022] In the formula:

[0023] K P —Instantaneous sensitivity coefficient of the in-pile neutron detector;

[0024] Q n (0) — The amount of effective charge generated by the in-pile neutron detector and neutron reaction at the initial moment;

[0025] Step 3: Based on the neutron and photon energy spectra at the in-pile instrument tubes obtained in Step 1, and the neutron sensitivity, photon sensitivity, and transient sensitivity coefficients of the in-pile neutron detector obtained in Step 2, calculate the gamma current of the in-pile neutron detector using formula (4).

[0026]

[0027] In the formula:

[0028] I P —Gamma current of the in-pile neutron detector;

[0029] ΔE n —The energy range of the neutron energy spectrum at the in-core instrumentation tubes;

[0030] φ n (E) — Neutron energy spectrum at the instrumentation tubes inside the reactor;

[0031] S n (E)——Neutron sensitivity of the in-reactor neutron detector at energy E;

[0032] ΔE γ —The energy range of the photon energy spectrum at the in-core instrument tubes;

[0033] φ γ (E)——Photon energy spectrum at the instrumentation tubes inside the reactor;

[0034] S γ (E) — Photon sensitivity of the in-reactor neutron detector at energy E;

[0035] Step 4: Based on the initialization material and structural parameters of the in-pile neutron detector read in Step 2, obtain the decay constant and microscopic absorption cross section of the sensitive region of the in-pile neutron detector;

[0036] Step 5: Based on the decay constant and microscopic absorption cross section of the sensitive region of the in-pile neutron detector obtained in Step 4, and the neutron energy spectrum at the in-pile instrument tube obtained in Step 1, calculate the expected operating lifetime of the in-pile neutron detector using formula (5).

[0037]

[0038] In the formula:

[0039] L – Expected operational lifetime of the in-pile neutron detector;

[0040] λ — decay constant of the sensitive region of the neutron detector in the reactor;

[0041] σ a (E)——Microscopic absorption cross section of the sensitive region of the in-pile neutron detector at energy E;

[0042] —The microscopic absorption reaction rate in the sensitive region of the in-pile neutron detector;

[0043] Step 6: Adjust the material and structural parameters of the in-pile neutron detector in Step 2, and repeat Step 2 to Step 5; based on the gamma current of the in-pile neutron detector obtained in Step 3 and the expected operating life of the in-pile neutron detector obtained in Step 5, make both meet the design target shown in Formula (6), and the corresponding material and structural parameters of the in-pile neutron detector are used as the optimized design of the gamma-compensated long-life in-pile neutron detector.

[0044] I P ≤I s L>L s (6)

[0045] In the formula:

[0046] I s —Design limits for gamma current in in-pile neutron detectors;

[0047] L s —Design limits for the expected operational lifetime of in-pile neutron detectors.

[0048] Preferably, the Monte Carlo model of the in-pile neutron detector described in step 2 uses the neutron energy spectrum and photon energy spectrum as radiation field input parameters, and the detector material and structure as calculation unit input parameters.

[0049] Preferably, the decay constant and microscopic absorption cross section of the sensitive region of the in-pile neutron detector described in step 4 are obtained from a material nuclear database.

[0050] The designed in-reactor neutron detector is a self-powered detector; the material of the sensitive region of the in-reactor neutron detector is... 51V, the detector structure is a coaxial cylindrical shape, and the gamma noise ratio of the in-pile neutron detector is less than 0.5%.

[0051] Compared with the prior art, the present invention has the following outstanding advantages:

[0052] 1. In this invention, the gamma compensation design is based on the radiation-induced current mechanism of the in-pile neutron detector, ensuring that the branch gamma currents from radiation trapping and fission transients compensate each other, thus solving the interference of gamma noise from the design principle.

[0053] 2. In this invention, the design method is based on the Monte Carlo model of the in-pile neutron detector and does not rely on costly experimental exploration; the design goal is achieved by adjusting the material and structural parameters of the in-pile neutron detector, which is highly adaptable to the current manufacturing technology and the solution is engineering feasible.

[0054] 3. The gamma-compensated long-life in-core neutron detector provided by the design method of this invention can not only significantly improve the signal-to-noise ratio of the in-core neutron detector itself, but also enhance the core monitoring capability based on the in-core neutron detector signal, making it usable for control and protection under transient conditions. Attached Figure Description

[0055] Figure 1 This is a flowchart of the method of the present invention.

[0056] Figure 2 This invention relates to the neutron and photon energy spectra obtained for a specific nuclear reactor type.

[0057] Figure 3 This is a schematic diagram illustrating the dynamic response trend of an in-core neutron detector obtained for a certain nuclear reactor type during a load shedding test. Detailed Implementation

[0058] The present invention will be further described in detail below with reference to the accompanying drawings and specific embodiments:

[0059] Specific steps are as follows: Figure 1 As shown, this invention is a design method for a gamma-compensated long-life in-reactor neutron detector. The main steps include establishing a Monte Carlo model of the detector, calculating the gamma current and expected operating lifetime of the in-reactor neutron detector, and optimizing the detector parameters. The specific steps are as follows:

[0060] Step 1: Read the geometry and material information of the fuel assembly containing the in-core instrumentation tubes, solve the steady-state neutron transport equations to obtain the neutron and photon energy spectra at the in-core instrumentation tubes, see [link to relevant documentation]. Figure 2 ,Depend on Figure 2 A significant gamma-ray radiation field can be observed at the instrument tubes inside the nuclear reactor.

[0061] Step 2: Read the initial material and structural parameters of the in-pile neutron detector. Based on the neutron energy spectrum and photon energy spectrum obtained in Step 1, establish a Monte Carlo model of the in-pile neutron detector in Geant4. Calculate the neutron sensitivity and photon sensitivity of the in-pile neutron detector using formulas (1) and (2) respectively, and calculate the transient sensitivity coefficient of the in-pile neutron detector using formula (3).

[0062]

[0063]

[0064] In the formula:

[0065] S n —The neutron sensitivity of the in-pile neutron detector;

[0066] S γ —The photon sensitivity of the in-pile neutron detector;

[0067] Q n (t)——The effective charge generated by the in-pile neutron detector and neutron reaction at time t;

[0068] Q γ —The effective charge generated by the reaction of photons with the in-pile neutron detector;

[0069] φ n — Neutron flux density in the sensitive region of the in-pile neutron detector;

[0070] φ γ —Photon flux density in the sensitive region of the in-pile neutron detector;

[0071] L – Length of the in-pile neutron detector;

[0072] t — time;

[0073]

[0074] In the formula:

[0075] K P —Instantaneous sensitivity coefficient of the in-pile neutron detector;

[0076] Q n (0) — The amount of effective charge generated by the in-pile neutron detector and neutron reaction at the initial moment;

[0077] Step 3: Based on the neutron and photon energy spectra at the in-pile instrument tubes obtained in Step 1, and the neutron sensitivity, photon sensitivity and transient sensitivity coefficients of the in-pile neutron detector obtained in Step 2, calculate the gamma current of the in-pile neutron detector using formula (4), where the first term on the right side of the equation is the γ-ray branch current from radiation capture, and the second term is the γ-ray branch current from fission transients.

[0078]

[0079] In the formula:

[0080] I P —Gamma current of the in-pile neutron detector;

[0081] ΔE n —The energy range of the neutron energy spectrum at the in-core instrumentation tubes;

[0082] φ n (E) — Neutron energy spectrum at the instrumentation tubes inside the reactor;

[0083] S n (E)——Neutron sensitivity of the in-reactor neutron detector at energy E;

[0084] ΔE γ —The energy range of the photon energy spectrum at the in-core instrument tubes;

[0085] φ γ (E)——Photon energy spectrum at the instrumentation tubes inside the reactor;

[0086] S γ (E) — Photon sensitivity of the in-reactor neutron detector at energy E;

[0087] Step 4: Based on the initialization material and structural parameters of the in-pile neutron detector read in Step 2, obtain the decay constant and microscopic absorption cross section of the sensitive region of the in-pile neutron detector;

[0088] Step 5: Based on the decay constant and microscopic absorption cross section of the sensitive region of the in-pile neutron detector obtained in Step 4, and the neutron energy spectrum at the in-pile instrument tube obtained in Step 1, calculate the expected operating lifetime of the in-pile neutron detector using formula (5).

[0089]

[0090] In the formula:

[0091] L – Expected operational lifetime of the in-pile neutron detector;

[0092] λ — decay constant of the sensitive region of the neutron detector in the reactor;

[0093] σ a(E)——Microscopic absorption cross section of the sensitive region of the in-pile neutron detector at energy E;

[0094] —The microscopic absorption reaction rate in the sensitive region of the in-pile neutron detector;

[0095] Step 6: Adjust the material and structural parameters of the in-core neutron detector in Step 2. Specifically, with the basic size of the detector as preferred, repeat Steps 2 to 5. Based on the gamma current of the in-core neutron detector obtained in Step 3 and the expected operating life of the in-core neutron detector obtained in Step 5, ensure that both meet the design target shown in Formula (6). The corresponding material and structural parameters of the in-core neutron detector are then used as the optimized gamma-compensated long-life in-core neutron detector. The dynamic response trend of this optimized gamma-compensated long-life in-core neutron detector during the load shedding test is shown in [reference needed]. Figure 3 ;

[0096] I P ≤I s L>L s (6)

[0097] In the formula:

[0098] I s —Design limit for gamma current of in-pile neutron detector, I s One percent of the total detector current can be taken;

[0099] L s —Design limit for the expected operational lifetime of the in-reactor neutron detector, L s A period of 6 years is acceptable.

[0100] Accordingly, a gamma-compensated long-life in-reactor neutron detector designed using the method of this invention includes: a self-powered detector; and a detector sensitive region material of... 51 V, the detector structure is a coaxial cylindrical shape.

[0101] In this invention, the characteristic information of the fuel assembly containing the in-core instrument tube in step 1 and the initialization information of the in-core neutron detector in step 2 are arbitrary. The method of this invention does not have specific selectivity for the design and initialization of fuel assemblies and in-core neutron detectors. Its design process based on the principle of γ branch current compensation is universal and is not limited by human factors and practical experience.

[0102] In step 2, the neutron and photon flux density and the effective charge generated by the reaction with neutrons and photons in the sensitive region of the in-pile neutron detector in formulas (1) and (2) can be obtained by the track length counting method; in step 3, the neutron energy spectrum and photon energy spectrum are preferably fine group structures; in step 4, the decay constant and microscopic absorption cross section of the sensitive region of the in-pile neutron detector can be obtained from the decay nucleus database and the evaluation nucleus database, respectively.

[0103] To verify the effectiveness of the present invention, Figure 3 The dynamic response curve of an in-pile neutron detector designed using this invention is shown during a load shedding test. Figure 3 It can be seen that the designed in-reactor neutron detector has an effective online tracking capability for the neutron-gamma mixed radiation field of the nuclear reactor, and the effective current source under transient operating conditions is the neutron signal. The unit neutron sensitivity of this in-reactor neutron detector is 2.55 × 10⁻⁶. -23 A·cm·s -1 The unit photon sensitivity is -4.02 × 10⁻⁶. -24 A·cm·s -1 During steady-state operation and transient load shedding tests, its gamma noise ratio was less than 0.5%, achieving a high signal-to-noise ratio for the detector. Furthermore, due to the adoption of… 51 V, as a sensitive material, has its size parameters designed to compensate for the self-shielding effect of the in-reactor neutron detector, and its design life is higher than 12 years, which can meet the design requirements for long-term continuous neutron detection.

[0104] This invention presents a design method for a gamma-compensated long-life in-reactor neutron detector. Based on the principle of gamma branch current compensation, specifically the mutual compensation of two gamma-ray branch currents originating from radiation trapping and fission instantaneous events (expressed in formula (4), this method addresses the gamma noise interference problem of the neutron detector in the mixed radiation field of a nuclear reactor. The total gamma current of the in-reactor neutron detector is calculated by solving for the radiation field information of the in-reactor instrumentation tubes and establishing a detector sensitivity separation model. Based on the microscopic absorption cross-section of the detector's sensitive region and the neutron energy spectrum of the in-reactor instrumentation tubes, the microscopic absorption reaction rate of the detector's sensitive region is calculated, thereby quantitatively calculating the detector's expected operating lifetime. By adjusting material and structural parameters, an optimized detector scheme that meets the design objectives is obtained. This in-reactor neutron detector has a high signal-to-noise ratio and solves the current interference in the gamma field from a design perspective, which is beneficial for improving the sustainable online monitoring capability of the reactor core state parameters. The proposed design method is based on the reaction principle between the detector and the mixed radiation field of the nuclear reactor, namely the radiation-induced current mechanism. It does not rely on prior experimental foundations or manufacturing process upgrades, is simple and efficient, and possesses engineering feasibility.

[0105] The design method and corresponding scheme of the gamma-compensated long-life in-reactor neutron detector of this invention can provide a design reference for the research and development of in-reactor neutron detection systems in nuclear reactors. This invention can effectively solve the gamma noise problem in online monitoring of reactor operation and improve the safety and control capabilities of nuclear reactors based on in-reactor measurement signals.

Claims

1. A design method for a gamma-compensated long-life in-reactor neutron detector, characterized in that, Includes the following steps: Step 1: Read the geometry and material information of the fuel assembly containing the in-core instrument tubes, and solve the steady-state neutron transport equations to obtain the neutron and photon energy spectra at the in-core instrument tubes; Step 2: Read the initial material and structural parameters of the in-pile neutron detector, establish the Monte Carlo model of the in-pile neutron detector based on the neutron energy spectrum and photon energy spectrum obtained in Step 1, calculate the neutron sensitivity and photon sensitivity of the in-pile neutron detector using formula (1) and formula (2) respectively, and calculate the instantaneous sensitivity coefficient of the in-pile neutron detector using formula (3). In the formula: S n —The neutron sensitivity of the in-pile neutron detector; S γ —The photon sensitivity of the in-pile neutron detector; Q n (t)——The effective charge generated by the in-pile neutron detector and neutron reaction at time t; Q γ —The effective charge generated by the reaction of photons with the in-pile neutron detector; φ n — Neutron flux density in the sensitive region of the in-pile neutron detector; φ γ —Photon flux density in the sensitive region of the in-pile neutron detector; L – Length of the in-pile neutron detector; t — time; In the formula: K P —Instantaneous sensitivity coefficient of the in-pile neutron detector; Q n (0) — The amount of effective charge generated by the in-pile neutron detector and neutron reaction at the initial moment; Step 3: Based on the neutron and photon energy spectra at the in-pile instrument tubes obtained in Step 1, and the neutron sensitivity, photon sensitivity, and transient sensitivity coefficients of the in-pile neutron detector obtained in Step 2, calculate the gamma current of the in-pile neutron detector using formula (4). In the formula: I P —Gamma current of the in-pile neutron detector; ΔE n —The energy range of the neutron energy spectrum at the in-core instrumentation tubes; φ n (E) — Neutron energy spectrum at the instrumentation tubes inside the reactor; S n (E)——Neutron sensitivity of the in-reactor neutron detector at energy E; ΔE γ —The energy range of the photon energy spectrum at the in-core instrument tubes; φ γ (E)——Photon energy spectrum at the instrumentation tubes inside the reactor; S γ (E) — Photon sensitivity of the in-reactor neutron detector at energy E; Step 4: Based on the initialization material and structural parameters of the in-pile neutron detector read in Step 2, obtain the decay constant and microscopic absorption cross section of the sensitive region of the in-pile neutron detector; Step 5: Based on the decay constant and microscopic absorption cross section of the sensitive region of the in-pile neutron detector obtained in Step 4, and the neutron energy spectrum at the in-pile instrument tube obtained in Step 1, calculate the expected operating lifetime of the in-pile neutron detector using formula (5). In the formula: L – Expected operational lifetime of the in-pile neutron detector; λ — decay constant of the sensitive region of the neutron detector in the reactor; σ a (E)——Microscopic absorption cross section of the sensitive region of the in-pile neutron detector at energy E; —The microscopic absorption reaction rate in the sensitive region of the in-pile neutron detector; Step 6: Adjust the material and structural parameters of the in-pile neutron detector in Step 2, and repeat Step 2 to Step 5; based on the gamma current of the in-pile neutron detector obtained in Step 3 and the expected operating life of the in-pile neutron detector obtained in Step 5, make both meet the design target shown in Formula (6), and the corresponding material and structural parameters of the in-pile neutron detector are used as the optimized design of the gamma-compensated long-life in-pile neutron detector. I P ≤I s ,L>L s (6) In the formula: I s —Design limits for gamma current in in-pile neutron detectors; L s —Design limits for the expected operational lifetime of in-pile neutron detectors.

2. The design method of a gamma-compensated long-life in-reactor neutron detector according to claim 1, characterized in that: The Monte Carlo model of the in-pile neutron detector described in step 2 uses the neutron energy spectrum and photon energy spectrum as radiation field input parameters, and the detector material and structure as calculation unit input parameters.

3. The design method of a gamma-compensated long-life in-reactor neutron detector according to claim 1, characterized in that: The decay constant and microscopic absorption cross section of the sensitive region of the in-pile neutron detector mentioned in step 4 were obtained from a material nuclear database.

4. The design method of a gamma-compensated long-life in-reactor neutron detector according to claim 1, characterized in that, The designed in-reactor neutron detector is a self-powered detector; the material of the sensitive region of the in-reactor neutron detector is... 51 V, the detector structure is a coaxial cylindrical shape, and the gamma noise ratio of the in-pile neutron detector is less than 0.5%.

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