A space dose rate radiation protection method based on irradiation supervision sample source term calculation

Abstract

The present application relates to the technical field of nuclear power plant radiation protection, and discloses a kind of space dose rate radiation protection method based on irradiation supervision sample item calculation, comprising: step 1: the source item analysis is carried out to the supervision sample and scrap taken out, obtain the different nuclide species, quantity produced in unit volume of target material due to neutron activation at irradiation stop moment, step 2: the activity of different radioisotopes produced in unit volume and unit mass of target material due to activation at different irradiation time stop irradiation is calculated, step 3: a Monte Carlo calculation model is established, and the space dose rate caused by unit mass of supervision sample under different scenarios is carried out, step 4: the dose rate distribution of supervision sample is calculated, step 5: according to the dose rate distribution of supervision sample calculated in step 4, the kind and thickness of radiation protection material in different scenarios and positions are arranged, combined with the dose rate requirement of place, and reasonable radiation protection measures are formulated, and the present application can assist in reducing the dose of personnel during the overhaul of power station.

Description

TECHNICAL FIELD

[0001] The present application relates to the technical field of nuclear power plant radiation protection, in particular to a spatial dose rate radiation protection method based on irradiation supervision sample source term calculation. BACKGROUND

[0002] In some nuclear power plant pressurized water reactors, the upper cylindrical shell section of the reactor pressure vessel corresponds to the core part, and the inner surface thereof is the maximum neutron fluence area of the pressure vessel. Therefore, 12 support seats are arranged on the inner surface of the upper cylindrical shell section, and each two support seats form a group and are uniformly arranged between the groups. The support seats are used to place the steel "irradiation" supervision sample box of the reactor pressure vessel to monitor the change of the material properties of the reactor pressure vessel under the action of neutron irradiation.

[0003] Currently, the irradiation supervision sample of the pressurized water reactor pressure vessel of some nuclear power plants is welded to the inner surface of the pressure vessel, and when it is taken out, a special tool is needed to melt and cut the welding point and then take it out. For other commercial light water pressurized water reactors, the supervision sample is screwed into a special hole on the surrounding barrel in the hanging basket core, and when it is taken out, a special tool is used to unscrew and take it out. The two methods are quite different. Since the irradiation supervision sample of the pressurized water reactor is welded to the inner surface of the pressure vessel, it can better reflect and represent the change of the properties of the pressure vessel under irradiation and high temperature. However, for the taking-out process, cutting is required, which introduces the problem of cutting products and related radiation measurement increase, so that the taking-out process of the supervision sample of the pressurized water reactor pressure vessel not only needs to consider the radiation protection of the operation process of the supervision sample, but also needs to consider the radiation protection of the cutting products of the supervision sample.

[0004] Without related source term analysis, it is impossible to develop reasonable radiation protection measures, which often leads to insufficient protection or excessive protection. SUMMARY

[0005] In order to solve the above technical problems, the present application provides a spatial dose rate radiation protection method based on irradiation supervision sample source term calculation, which can calculate the external exposure dose rate level around the supervision sample and cutting products before the work is carried out, avoid unnecessary radioactive exposure caused by the radiation dose level detection of the radiation protection personnel, and reduce the personnel dose during the overhaul of the power plant.

[0006] The present application provides a spatial dose rate radiation protection method based on irradiation supervision sample source term calculation, which comprises:

[0007] Step 1: source term analysis is performed on the taken-out supervision sample and debris, the material components and natural abundance of the supervision sample are combined, and the types and quantities of different nuclides produced in the unit volume of target material due to neutron activation at the irradiation stop moment are obtained;

[0008] Step 2: Establish neutron irradiation activation differential equation, combined with the above source item analysis results, calculate the activity of different radionuclides generated by activation in unit volume and unit mass of target material at different irradiation times and stopping irradiation;

[0009] Step 3: Based on the proportion of target material nuclides, the source is equivalent to 1.86 MeV gamma rays, a Monte Carlo calculation model is established, and the spatial dose rate caused by unit mass of supervision sample in different scenarios is carried out;

[0010] Step 4: Based on the calculated activity of radionuclides and the spatial dose rate caused by unit mass of supervision sample, the dose rate distribution of supervision sample is calculated;

[0011] Step 5: According to the dose rate distribution of supervision sample calculated in step 4, arrange different scenarios, positions, types and thicknesses of radiation protection materials, and combined with the dose rate requirement of the place, formulate reasonable radiation protection measures to reduce the radiation dose rate of personnel.

[0012] Optionally, the scenarios in step 3 include spatial dose rate distribution without shielding, spatial dose rate distribution with different lead shielding, personnel dose rate level with different water layer thickness shielding, and personnel radiation dose rate level with iron shielding above the water layer.

[0013] Optionally, step 1 includes:

[0014] The number of activated radioactive nuclei generated by the non-radioactive target atomic nucleus of interest in the target material irradiated by neutrons satisfies the following differential equation:

[0015] (1)

[0016] In the formula:

[0017] N0 is the number of non-radioactive target atomic nuclei of interest in unit volume of target material irradiated by neutrons, with the unit of cm 3 ;

[0018] N is the number of radioactive nuclei generated by the activation of target atomic nuclei in unit volume of target material irradiated by neutrons, with the unit of cm 3 ;

[0019] is the microscopic activation cross section of the target atomic nucleus, with the unit of 10 -24 cm 2 ;

[0020] is the microscopic "consumption" cross section of the radioactive nucleus generated by activation, with the unit of 10 -24 cm 2 ;

[0021] is the irradiation neutron flux density, unit: cm 2 ·s;

[0022] is the neutron flux density for "consumption" of the generated radioactive nuclei, unit: 10 -24 cm 2 ;

[0023] time;

[0024] is the probability of decay of one radioactive nucleus in unit time;

[0025] Assuming and are constants, also considered as not changing with irradiation time, under the initial condition of =0, =0, from equation (1) we get:

[0026] (2)

[0027] Here, is the continuous irradiation time, is the number of radioactive nuclei generated in unit volume of target material at the moment of stopping irradiation due to activation, unit: cm 3 .

[0028] Optionally, the activity of different radioactive nuclides generated in unit volume of target material at the moment of stopping irradiation due to activation in step 2 with different irradiation times is calculated by the following formula:

[0029] Equation (2) can also be expressed in another form as:

[0030] (3)

[0031] Here, is the continuous irradiation time, unit: Bq / cm 3 .

[0032] Optionally, the activity of different radioactive nuclides generated in unit mass of target material at the moment of stopping irradiation due to activation in step 2 with different irradiation times is calculated by the following formula:

[0033] If the third term on the right side of equation (1) is not considered, or the third term is negligible compared with the second term, then equation (2) and equation (3) are simplified as:

[0034]

[0035] (4)

[0036] From formula (3) and formula (4), the following formula (5) and formula (6) are derived:

[0037] (5)

[0038] Or:

[0039] (6)

[0040] In the formula:

[0041] The unit of is Bq / g;

[0042] is Avogadro constant, taking the value of 6.023 x 10 23 ;

[0043] is the mass number of the non-radioactive target nucleus of interest in the target material;

[0044] is the share of the total mass of the non-radioactive target nucleus of interest in the target material in the mass of the target material;

[0045] is the saturation factor of the activation radioactivity.

[0046] Optionally, the irradiation neutron flux density takes the following values:

[0047] The neutron flux radial distribution from the core boundary to the outer boundary of the pressure vessel in a pressurized water reactor can be measured by the in-core and out-of-core neutron probes at the position where the surveillance sample is located;

[0048] The radial thickness of each part of the reactor in the radial direction, including the core, the core barrel, the water gap, the core basket, the pressure vessel cladding layer, the pressure vessel, the air gap, the thermal insulation layer, the concrete lining, and the concrete lining can be found in the equipment manual;

[0049] The thermal neutron flux density and the fast neutron flux density at the position of the surveillance sample can be found from the above data, and the sum of the thermal neutron flux density and the fast neutron flux density is the irradiation neutron flux density at the position of the surveillance sample.

[0050] Optionally, after obtaining the saturation mass specific activity of the surveillance sample by formula 6, the volume specific activity is obtained by using the density of the surveillance sample material.

[0051] Optionally, according to the calculated specific activity, in combination with the mass of the supervision sample, the size of the cutting product and the total mass, and in reference to the absorption dose rate constant of the corresponding nuclide and the dose rate level in different scenarios calculated by the Monte Carlo simulation, the absorption dose rate and the equivalent dose rate at different distances, i.e. the external exposure dose rate level around the supervision sample and the cutting product, can be calculated.

[0052] Optionally, the size, shape and thickness of the special tool and container for the supervision sample and the cutting product can be determined according to the external exposure dose rate level around the supervision sample and the cutting product.

[0053] Optionally, when shielding cannot be used, the minimum distance that the personnel must maintain can also be estimated according to the results of the external exposure dose rate distribution.

[0054] Compared with the prior art, the technical scheme provided by the embodiments of the present disclosure has the following advantages:

[0055] The method for calculating the spatial dose rate radiation protection based on the irradiation supervision sample source term provided by the embodiments of the present disclosure first obtains different nuclide types and quantities generated in the target material per unit volume due to neutron activation at the irradiation stop time, then establishes a neutron irradiation activation differential equation, calculates the activity of different radioactive nuclides generated in the target material per unit volume and per unit mass due to activation at different irradiation time stop irradiation in combination with the above source term analysis results, establishes a Monte Carlo calculation model based on the target material nuclide proportion and the source equivalent 1.86 MeV gamma rays, carries out the spatial dose rate caused by the unit mass supervision sample in different scenarios, calculates the supervision sample dose rate distribution based on the calculated activity of the radioactive nuclides and the spatial dose rate caused by the unit mass supervision sample, and finally, according to the calculated supervision sample dose rate distribution, arranges the types and thicknesses of the radiation protection materials in different scenarios and positions, combines the site dose rate requirement, and formulates reasonable radiation protection measures to reduce the personnel irradiation dose rate, so that the external exposure dose rate level around each radiation source of the supervision sample and the cutting product can be calculated before the work is carried out according to the existing data and information of the nuclear power plant, and this part is calculated by the Monte Carlo calculation model to avoid unnecessary radioactive irradiation caused by the radiation protection personnel in the process of detecting the radiation dose level and to reduce the personnel exposure dose in the process of the power plant overhaul. BRIEF DESCRIPTION OF DRAWINGS

[0056] Figure 1 FIG. 1 is a structural schematic diagram of the reactor pressure vessel in the current nuclear power plant pressurized water reactor;

[0057] Figure 2 FIG. 3 is a neutron flux radial distribution diagram from the core boundary to the outer boundary of the pressure vessel in the embodiments of the present disclosure;

[0058] Figure 3A schematic diagram of a Monte Carlo calculation model in an embodiment of the present application;

[0059] Figure 4 A spatial dose rate level of a supervision sample in the air without shielding in an embodiment of the present application;

[0060] Figure 5 A spatial dose rate level of a supervision sample in different lead shielding in an embodiment of the present application;

[0061] Figure 6 A spatial dose rate level of a supervision sample in different water layer thickness in an embodiment of the present application;

[0062] Figure 7 A dose rate change of a place in a 200 cm water layer thickness + 2 cm iron shielding in an embodiment of the present application.

[0063] BRIEF DESCRIPTION OF DRAWINGS

[0064] 1 - body, 2 - cover, 3 - core, 4 - irradiation supervision sample box, 5 - supervision sample, 6 - lower end cover, 7 - upper end cover, 8 - multi-layer lead shielding layer, 9 - internal cavity, 10 - outer layer stainless steel wall surface. DETAILED DESCRIPTION

[0065] One specific embodiment of the present application will be described in detail below with reference to the accompanying drawings, but it should be understood that the scope of protection of the present application is not limited by the specific embodiment.

[0066] At present, the irradiation supervision sample of the pressurized water reactor pressure vessel of some nuclear power plants is welded to the inner surface of the pressure vessel, and when it is taken out, special tools are needed to melt and cut the welding point and then take it down. For other commercial light water pressurized water reactors, the supervision sample is screwed into the special hole on the surrounding barrel in the core of the basket, and when it is taken out, special tools are used to unscrew and take it out. The two methods are quite different. Since the irradiation supervision sample of the pressurized water reactor is welded to the inner surface of the pressure vessel, it can better reflect and represent the changes in the properties of the pressure vessel under irradiation and high temperature, but for the taking-out process, cutting must be performed, thereby introducing cutting products and increasing the problem of related radiation metrology, so that the supervision sample taking-out process of the pressurized water reactor pressure vessel not only needs to consider the radiation protection of the supervision sample operation process, but also needs to consider the radiation protection of the cutting products.

[0067] Without related source item analysis, it is impossible to develop reasonable radiation protection measures, often resulting in insufficient protection measures or over-protection. If the radiation parameters of the activated irradiation supervision sample itself and the radiation parameters of the cutting products (cutting slag) can be analyzed, it can effectively guide relevant personnel to develop reasonable radiation protection measures and save manpower and material resources.

[0068] To this end, the embodiment of the present disclosure provides a spatial dose rate radiation protection method based on irradiation supervision sample source item calculation, which belongs to the field of nuclear power plant radiation protection, and provides a thought for taking out the activated supervision sample and calculating the radiation parameters of the supervision sample and the cutting product (cutting slag), thereby effectively guiding relevant personnel to develop reasonable radiation protection measures.

[0069] In order to more clearly illustrate the spatial dose rate radiation protection method based on irradiation supervision sample source item calculation in the present disclosure, the structure of the pressurized water reactor pressure vessel of the nuclear power plant is first described, as shown in the figure. Figure 1 As shown in the figure, the body 1 and the cover 2 of the pressure vessel are connected through the flange 1, and the cover 2 corresponds to the core 3 part, so the inner surface thereof is the maximum neutron flux area of the pressure vessel. Therefore, the supervision sample box 4 is fixed on the cover 2, and the irradiation supervision sample box 4 is located in the highest neutron flux area of the pressure vessel to monitor the change of the material characteristics of the reactor pressure vessel under the action of neutron irradiation.

[0070] Compared with the irradiation supervision sample fixed in the basket core special hole in a spiral manner, the irradiation supervision sample fixed on the inner surface of the pressure vessel by welding is more complex to take out. Before calculating the radiation source item of the irradiation supervision sample, the irradiation supervision sample fixed on the inner surface of the pressure vessel by welding is cut, and the cutting debris is collected. The taking out operation process of the supervision sample is divided into three stages:

[0071] The first stage is the cutting and taking out stage.

[0072] The second stage is the polishing operation stage.

[0073] The third stage is the debris collection operation stage.

[0074] The embodiment of the present disclosure provides a spatial dose rate radiation protection method based on irradiation supervision sample source item calculation, and the calculation method is used for calculating the radiation parameters of each radiation source of the irradiation supervision sample fixed on the inner surface of the pressure vessel by welding and the cutting product. The calculation method of the radiation parameters of each radiation source includes:

[0075] Step 1: source item analysis is performed on the taken-out supervision sample and debris, the supervision sample material components and natural abundance are combined to obtain different nuclide types and quantities generated in the target material per unit volume at the irradiation stop moment due to neutron activation;

[0076] Step 2: a neutron irradiation activation differential equation is established, and the activity of different radioactive nuclides generated in the target material per unit volume and per unit mass at different irradiation time stop irradiation is calculated in combination with the above source item analysis result.

[0077] Step 3: Based on the proportion of target material nuclides, the source is equivalent to 1.86 MeV gamma rays, a Monte Carlo calculation model is established, and the spatial dose rate caused by unit mass of supervision sample in different scenarios is carried out;

[0078] Step 4: Based on the calculated activity of radionuclide and the spatial dose rate caused by unit mass of supervision sample, the supervision sample dose rate distribution is calculated;

[0079] Step 5: According to the supervision sample dose rate distribution calculated in step 4, the types and thicknesses of radiation protection materials in different scenarios and positions are arranged, and combined with the site dose rate requirement (not higher than 0.5 mSv / h), reasonable radiation protection measures are formulated to reduce the personnel irradiation dose rate.

[0080] The spatial dose rate radiation protection method based on the source term calculation of the irradiation supervision sample provided by the embodiment of the disclosure first obtains the different nuclide types and quantities generated in the unit volume of target material due to neutron activation at the irradiation stop moment, and then establishes a neutron irradiation activation differential equation, combines the above source term analysis results, calculates the activity of different radionuclides generated in the unit volume and unit mass of target material due to activation at different irradiation time stop irradiation, based on the proportion of target material nuclides, the source is equivalent to 1.86 MeV gamma rays, a Monte Carlo calculation model is established, and the spatial dose rate caused by unit mass of supervision sample in different scenarios is carried out, based on the calculated activity of radionuclide and the spatial dose rate caused by unit mass of supervision sample, the supervision sample dose rate distribution is calculated, finally, according to the calculated supervision sample dose rate distribution, the types and thicknesses of radiation protection materials in different scenarios and positions are arranged, combined with the site dose rate requirement, reasonable radiation protection measures are formulated to reduce the personnel irradiation dose rate, so that the existing data and data of the nuclear power plant can be calculated to obtain the surrounding external exposure dose rate level of each radiation source of the supervision sample and the cutting product before the work is carried out. This part is calculated by a Monte Carlo calculation model, as shown in Figure 3 The Monte Carlo calculation model includes a supervision sample 5 placed inside a shielding barrel, which is composed of a lower end cover 6, an upper end cover 7, a plurality of lead shielding layers 8, an internal cavity 9 and an outer stainless steel wall 10, to avoid unnecessary radioactive irradiation of the radiation protection personnel during the detection of the radiation dose level and reduce the personnel exposure dose during the overhaul of the power plant.

[0081] Specifically, the scenarios in step 3 include the spatial dose rate distribution without shielding, the spatial dose rate distribution with different lead shielding, the personnel dose rate level with different water layer thickness shielding, and the personnel irradiation dose rate level with iron shielding above the water layer.

[0082] Specifically, step 1 includes:

[0083] The number of radioactivity nuclei produced by activation of the non-radioactive target nuclei of interest in the target material irradiated by neutrons satisfies the following differential equation:

[0084] (1)

[0085] where:

[0086] N is the number of non-radioactive target nuclei of interest per unit volume in the target material irradiated by neutrons, in units of # / cm 3 ;

[0087] N is the number of radioactivity nuclei produced by activation of the target nuclei per unit volume in the target material irradiated by neutrons, in units of # / cm 3 ;

[0088] is the microscopic activation cross section of the target nuclei, in units of 10 -24 cm 2 ;

[0089] is the microscopic "depletion" cross section of the radioactivity nuclei produced by activation, in units of 10 -24 cm 2 ;

[0090] is the neutron flux density, in units of # / cm 2 ·s;

[0091] is the neutron flux density used to "deplete" the radioactivity nuclei produced by activation, in units of 10 -24 cm 2 ;

[0092] is time;

[0093] is the probability that one radioactivity nucleus will decay in a unit of time;

[0094] Assuming and are constants, is also considered to be constant and does not change with irradiation time, under the initial conditions of = 0, = 0, equation (1) becomes:

[0095] (2)

[0096] Here, is the duration of continuous irradiation The number of radioactive nuclei generated by activation per unit volume of target material at the moment of stopping irradiation, in units of pieces / cm 3 .

[0097] Specifically, the activity of different radionuclides generated by activation per unit volume of target material at the moment of stopping irradiation in step 2 for different irradiation times The following formula is used for calculation:

[0098] Formula (2) can also be expressed in another form as:

[0099] (3)

[0100] Here, t is the duration of irradiation, The unit of A is Bq / cm 3 .

[0101] Further, the activity of different radionuclides generated by activation per unit mass of target material at the moment of stopping irradiation in step 2 for different irradiation times The following formula is used for calculation:

[0102] If the third term on the right side of formula (1) is not considered, or the third term is negligible compared with the second term, then formula (2) and formula (3) are simplified as:

[0103]

[0104] (4)

[0105] From formula (3) and formula (4), the following formula (5) and formula (6) are derived:

[0106] (5)

[0107] Or:

[0108] (6)

[0109] In the formula:

[0110] The unit of A is Bq / g;

[0111] Ave is Avogadro's constant, with a value of 6.023 x 10 23 ;

[0112] A is the mass number of the non-radioactive target nucleus of interest in the target material;

[0113] A is the proportion of the total mass of the non-radioactive target nucleus of interest in the target material in the mass of the target material;

[0114] The saturation factor for the activation radioactivity.

[0115] The value of the irradiation neutron flux density is as follows:

[0116] The neutron flux radial distribution of the position where the surveillance sample is located between the core boundary and the pressure vessel outer boundary in the PWR can be measured by the in-core and out-core neutron probes, as shown in Figure 2 ;

[0117] The radial thickness of each part of the reactor in the radial direction, including the core, the core barrel, the water gap, the core basket, the pressure vessel cladding layer, the pressure vessel, the air gap, the thermal insulation layer, and the concrete lining, is obtained from the equipment manual.

[0118] The thermal neutron flux density and the fast neutron flux density at the position of the surveillance sample can be obtained from the above data, and the irradiation neutron flux density at the position of the surveillance sample is obtained by adding the thermal neutron flux density and the fast neutron flux density. .

[0119] Optionally, the probability of decay of a radioactive nucleus per unit time The nuclear parameter value is obtained from the reference nuclear database.

[0120] Optionally, after obtaining the thermal neutron activation saturation mass specific activity of the surveillance sample by formula 6, the volume specific activity is obtained by using the density of the surveillance sample.

[0121] Optionally, according to the calculated mass specific activity, combined with the mass of the surveillance sample, the size and total mass of the cut product, and referring to the corresponding nuclear absorption dose rate constant and the dose rate level under different scenarios calculated by Monte Carlo simulation, the absorption dose rate and equivalent dose rate at different distances, i.e., the external exposure dose rate level around the removed surveillance sample and cut product, can be calculated.

[0122] Optionally, according to the external exposure dose rate level around the surveillance sample and the cut product, the size, shape, and thickness of the dedicated tool and container are determined.

[0123] Optionally, when shielding cannot be used, the minimum distance that personnel must maintain can also be estimated from the results of the external exposure dose rate distribution.

[0124] The overall process is as follows

[0125] Activation ratio radioactivity calculation formula

[0126] The number of activated nuclei of the non-radioactive target nuclei of interest in the target material irradiated by neutrons, N, is given by the following differential equation:

[0127] (1)

[0128] where:

[0129] N is the number of non-radioactive target nuclei of interest per unit volume in the target material irradiated by neutrons, in units of cm 3 ;

[0130] N is the number of radioactive nuclei generated by activation of the target nuclei per unit volume in the target material irradiated by neutrons, in units of cm 3 ;

[0131] σ is the microscopic activation cross-section of the target nuclei, in units of 10 -24 cm 2 ;

[0132] σ is the microscopic "depletion" cross-section of the activated radioactive nuclei, in units of 10 -24 cm 2 ;

[0133] Φ is the neutron flux density, in units of cm 2 ·s;

[0134] Φ is the neutron flux density for "depleting" the activated radioactive nuclei, in units of 10 -24 cm 2 ;

[0135] t is time;

[0136] λ is the probability that one radioactive nucleus will decay in a unit of time;

[0137] Assuming and are constants, and are also considered to be independent of irradiation time, under the initial conditions of = 0, = 0, equation (1) gives:

[0138] (2)

[0139] Here, is the duration of continuous irradiation The number of radioactive nuclei generated by activation in the target material per unit volume at the time of stopping irradiation (number / cm 3 ).

[0140] Formula (2) can also be expressed in another form:

[0141] (3)

[0142] here, The continuous irradiation time is , the activity of radionuclides generated by activation in unit volume of target material at the moment of cessation of irradiation, Bq / cm 3 .

[0143] If the third term on the right side of equation (1) is not necessary, or the third term is negligible compared to the second term, equations (2) and (3) can be simplified to:

[0144]

[0145] (4)

[0146] From (3) and (4), we can deduce (5) and (6). The activity of radionuclides generated by activation per unit mass of target material at the moment of cessation of irradiation for a continuous irradiation time of t (Bq / g):

[0147] (5)

[0148] Or:

[0149] (6)

[0150] Where:

[0151] is Avogadro's constant, which is 6.023×10 23 ;

[0152] is the mass number of the non-radioactive target nucleus of interest in the target material;

[0153] It is the proportion of the total mass of the non-radioactive target nuclei in the target material to the mass of the target material.

[0154] The activity of activated radionuclides can be calculated using equations (4) and (6) in most cases. By analyzing equations (4) and (6), the following very useful information can be obtained:

[0155] In the formula The saturation factor for activation radioactivity. When the irradiation time is 1 half-life, 50% of the saturation value is reached; when the irradiation time is 2 half-lives, 75% of the saturation value is reached; longer irradiation time has little contribution to the increase of activation radioactivity, and it can be approximately considered that when the irradiation time is more than 5 half-lives, the activation radioactivity reaches the saturation value;

[0156] When the irradiation time is less than 1 half-life, the activation radioactivity activity is approximately linearly related to the irradiation time.

[0157] After the irradiation is stopped, the remaining radioactivity after a period of cooling can be calculated by using the common decay formula.

[0158] Related input parameters

[0159] 2.1 Neutron flux density

[0160] The position where the surveillance sample is located is between the core boundary and the outer boundary of the pressure vessel, and in this PWR, the radial distribution of neutron flux from the core boundary to the outer boundary of the pressure vessel can be measured by the in-core and out-of-core neutron probes.

[0161] The radial thickness of each part of the reactor (including the core, the core barrel, the water gap, the core basket, the pressure vessel cladding layer, the pressure vessel, the air gap, the thermal insulation layer, the concrete lining, the concrete lining, etc.) can be found in the equipment manual.

[0162] From the above data, the thermal neutron flux and fast neutron flux at the position of the surveillance sample can be found.

[0163] 2.2 Half-life / decay constant

[0164] The nuclear cross section and the half-life / decay constant of the activation generated nuclide can be referred to the nuclear parameter values given in the Chinese nuclear database as shown in Table 1 below:

[0165] Table 1 Nuclear cross section of related nuclear reactions and half-life / decay constant of generated nuclides

[0166]

[0167] Among them, the (n, p) reaction cross section averaged over the entire fission neutron spectrum for the two fast neutron (n, p) activation reactions.

[0168] However, the fast neutron flux distribution given in the FSAR is not the full fission neutron spectrum. Considering the large difference in cross section conditions, the activation radioactivity produced by the two fast neutron (n, p) reactions is not checked.

[0169] And in the estimation of the radiation field, the contribution of 60 Co and 59 Fe emitting high-energy γ rays is mainly considered,58 Co and 54 Mn even 51 The contribution of Cr can be relatively neglected.

[0170] 2.3 Composition of target nuclides in activated metal materials

[0171] The activated metal material is selected according to the actual material used in the power plant, which can be found in the equipment technical documents. It is generally stainless steel. The composition of the target nuclides for this unit can be found in Table 2:

[0172] Table 2 Element composition, target nuclides, natural abundance and mass fraction in stainless steel

[0173]

[0174] Estimation results

[0175] Using formula (6), after substituting the above parameters, the thermal neutron activation saturation mass specific activity of the surveillance sample can be estimated, and then the volume specific activity can be obtained by using the density of the material (such as 7.8g / cm3 for stainless steel).

[0176] The time for taking out the surveillance samples of this unit is initially set at 4 years, 8 years, 12 years, 20 years, 28 years and 36 years. Therefore, the corresponding time is used as the moment to stop irradiation, and the corresponding mass specific activity is estimated, that is, the mass specific activity of the taken out surveillance samples and cutting products.

[0177] Expression based on saturation factor It can be seen that the mass specific activity of 60Co in the surveillance sample and its cutting products increased rapidly in the first few years, and the increase rate tended to be very slow after 20 years. 51 Cr, 59 Fe, 58 Co、 54 Since nuclides such as Mn have short half-lives, their mass specific activities can be considered to have reached the corresponding saturation value from the time they are first taken out four years later.

[0178] Radiation protection applications

[0179] Based on the calculated mass specific activity, combined with the mass of the surveillance sample, the size and total mass of the cutting products (debris, including those produced by grinding), and referring to the absorption dose rate constant of the corresponding nuclide, the absorbed dose rate and equivalent dose rate at different distances can be easily calculated, that is, the external exposure dose rate level around the removed surveillance sample and cutting products.

[0180] The actual different external radiation dose rate levels determine the size, shape and thickness of the special tools and containers to make the shielding effect as low as reasonably practicable.

[0181] For the time when shielding can not be used, with the results of the external exposure dose rate distribution, the minimum distance that personnel must maintain can also be estimated; it can guide the staff to determine the size, shape and thickness of their special tools and containers according to the actual different external exposure dose rate levels calculated, so that the shielding effect reaches a reasonable and feasible minimum result, that is, to ensure safety and save costs; it can guide the staff to understand the minimum distance that must be maintained from the radiation source when shielding cannot be used; it can guide the staff to fully collect the cutting products (fragments - formed by melting in water, close to spherical) to prevent the activated fragments in the core from being transferred to the auxiliary system and then forming 'hot particles' on the pipeline; the design of the present application can guide the staff to prepare corresponding protective measures in advance and reasonably arrange the work plan, which can effectively shorten the actual work period of the corresponding work.

[0182] The present disclosure will be described below through a specific embodiment. In order to keep the following description of the embodiments of the present disclosure clear and concise, the detailed description of known functions and known components can be omitted.

[0183] Embodiment 1

[0184] The saturated mass specific activity of the monitor sample in thermal neutron activation is calculated by formula (7). The thermal neutron flux at the location of the monitor sample, the nuclear cross section, and the half-life / decay constant of the activated nuclide, and the mass fraction of each element in the target nuclide can be calculated / queried in the following way.

[0185] 1) Obtain the neutron flux

[0186] The monitor sample is located between the core boundary and the outer boundary of the pressure vessel, so the neutron flux can be obtained from the radial distribution of the neutron flux from the core boundary to the outer boundary of the pressure vessel in the PWR, which can be measured by the in-core and out-of-core neutron probes, as shown in Figure 2 .

[0187] The radial thickness of each component of the reactor (including the core, the core barrel, the water gap, the core basket, the pressure vessel, the air gap, the thermal insulation layer, the concrete lining, and the like) can be found in the equipment manual.

[0188] Based on the above data, the thermal neutron flux and the fast neutron flux at the location of the monitor sample can be obtained from the manual, and part of the manual is shown in Figure 2 .

[0189] 2) Half-life / decay constant of target nuclide

[0190] The nuclear cross section and the half-life / decay constant of the activated nuclide are given by the nuclear parameter values from the China Nuclear Database, and the half-life / decay constant of the nuclide is shown in Table 1:

[0191] Where two fast neutron (n, p) activation reactions are averaged by the (n, p) reaction cross section of the whole fission neutron spectrum. 58 Co and 54 Mn even 51 Cr's contribution is small and can be relatively ignored, so the calculation of the present example mainly considers the contribution of 60 Co and 59 Fe, and the equivalent gamma ray energy is 1.86 MeV.

[0192] 3) Mass fraction of target nuclides

[0193] The activated metal material is selected according to the actual supervision sample material of the nuclear power plant, which can be found in the equipment technical document, and the supervision sample material of the present example is stainless steel, and the composition of the target nuclides involved in the unit is shown in Table 2:

[0194] 4) Calculate the mass specific activity / volume specific activity

[0195] Substitute the neutron flux, nuclear cross section, half-life / decay constant of the activated nuclide, mass fraction of each element in the target nuclide and other parameters into formula (7) to calculate the saturated mass specific activity of the supervision sample thermal neutron activation. Combined with the density of the supervision sample material (such as stainless steel 7.8 g / cm 3 ), the volume specific activity can be obtained, as shown in Table 3.

[0196] Table 3 Saturated mass specific activity of supervision sample thermal neutron activation

[0197]

[0198] 5) Mass specific activity of neutron activation for different years

[0199] The supervision sample of the unit involved in the present application is initially taken out at 4 years, 8 years, 12 years, 20 years, 28 years and 36 years, so the corresponding time is taken as the stopping irradiation time, and the corresponding mass specific radioactivity is estimated, that is, the mass specific radioactivity of the supervision sample and the cutting product, and the calculation results are shown in Table 4.

[0200] Table 4 Mass specific activity of supervision sample thermal neutron activation for different years

[0201]

[0202] According to the expression of the saturation factor It can be seen that the mass specific activity of 60 Co in the supervision sample and its cutting product increases rapidly in the first few years, and the increase rate tends to be very slow after 20 years, while 51 Cr, 59 Fe, 58 Co,54 Mn isotopes can be considered to reach the corresponding saturation value from the first removal at 4 years due to the short half-life.

[0203] 6) Dose rate levels at different locations

[0204] According to the calculated mass activity, combined with the mass of the surveillance sample, the size and total mass of the cutting product (debris, including that produced by grinding), and the involved isotope absorbed dose rate constant, the absorbed dose rate and equivalent dose rate at different distances, i.e. the external exposure dose rate level around the removed surveillance sample and the cutting product, can be calculated.

[0205] Figure 4 The spatial dose rate level in the air without shielding for the surveillance sample. At 1 m, the dose rate is about 370.07 mSv / h, and then attenuates to 14.33 mSv / h at 5 m, with an attenuation of about 96.1% from 1 m to 5 m. However, engineering requires that when the site dose rate is greater than 0.5 mSv / h, the staff should stop working, evacuate to a safe area, and report to the radiation protection personnel to confirm the radiation status. Therefore, the spatial dose rate is high when the surveillance sample is removed, and shielding measures need to be taken for the surveillance sample.

[0206] Figure 5 The spatial dose rate level of the surveillance sample with different lead shielding. For different lead shielding thicknesses, the dose rate of the surveillance sample changes, and the change trend shows a logarithmic function trend. The dose rate of the surveillance sample decreases rapidly when the lead shielding thickness is small, and then tends to a stable value. When the lead shielding thickness is 20 cm, the dose rate on the surface of the lead is 0.94 mSv / h, and when the lead shielding thickness is 23 cm, the dose rate on the surface of the lead is 0.18 mSv / h. Therefore, for 1.86 MeV gamma rays, 23 cm of lead material can reduce the dose rate generated by the surveillance sample to below 0.5 mSv / h, and the minimum thickness of 25 cm of the surveillance sample dedicated shielding barrel used in engineering can meet the shielding requirements.

[0207] Figure 6 The spatial dose rate level of the surveillance sample with different water layer thicknesses. When the water layer thickness is 160 cm, the water surface dose rate is 0.53 mSv / h, and when the water layer thickness is 180 cm, the water surface dose rate is 0.19 mSv / h. Therefore, when the water layer thickness is 180 cm, the site dose rate can meet the requirement of being below 0.5 mSv / h. When the surveillance sample is placed in the dedicated protective shielding container, the thickness of the water layer is reduced to 200 cm, at which time the water surface dose rate is 0.06 mSv / h. At this time, the dose rate of the surveillance sample placed in the shielding barrel meets the requirements. Figure 7For water layer thickness is 200cm, water layer above has 2cm iron shielding material, iron material above is air place dose rate variation, in water layer above additional 2cm iron material, can attenuate from 200cm water layer thickness dose rate to 0.02 mSv / h, therefore, water layer thickness 200cm+2cm, supervision sample to staff's radiation is smaller.

[0208] The above invention is only a few specific embodiments of the present application, but the embodiments of the present application are not limited to this, any changes that can be thought of by those skilled in the art should fall within the scope of the present application.

Claims

1. A method for spatial dose rate radiation protection based on irradiation supervised sample term calculation, characterized in that, The method comprises the following steps: Step 1: source term analysis is performed on the removed surveillance sample and debris, and combined with the material composition of the surveillance sample and natural abundance, the types and quantities of different nuclides generated in unit volume of the target material at the moment of stopping irradiation due to neutron activation are obtained; Step 2: a neutron irradiation activation differential equation is established, and combined with the source term analysis result, the activity of different radioisotopes generated in unit volume and unit mass of the target material at the moment of stopping irradiation at different irradiation times is calculated; Step 3: based on the nuclide proportion of the target material and the source equivalent to 1.86 MeV gamma rays, a Monte Carlo calculation model is established, and the spatial dose rate caused by unit mass of the surveillance sample in different scenarios is carried out; Step 4: based on the calculated activity of the radioisotopes and the spatial dose rate caused by unit mass of the surveillance sample, the surveillance sample dose rate distribution is calculated; Step 5: according to the surveillance sample dose rate distribution calculated in the step 4, the types and thicknesses of the radiation protection materials in different scenarios and positions are arranged, and combined with the site dose rate requirement, reasonable radiation protection measures are formulated to reduce the personnel irradiation dose rate; The scenarios in the step 3 include the spatial dose rate distribution without shielding, the spatial dose rate distribution with different lead shielding, and the personnel dose rate level with different water layer thickness shielding and the personnel irradiation dose rate level with iron shielding above the water layer; The step 1 comprises: The number of activated radioisotopes generated by the non-radioactive target atomic nucleus concerned in the target material irradiated by neutrons satisfies the following differential equation: （1） In the formula: Nf = number of non-radioactive target nuclei of interest per unit volume in the target material irradiated by neutrons 3 ; The number of radioactive nuclei per unit volume in a target material irradiated by neutrons, generated by the activation of target atomic nuclei, in units of pieces / cm 3 ; Microscopic cross section in units of 10-34 cm2 for the target nucleus -24 cm 2 ; Microscopic "consumption" cross section for the activation of the generated radioactive nuclei, in units of 10 -24 cm 2 ; is the irradiation neutron flux density in units of cm"2s"1 2 · s; The neutron flux density for "burning" the generated radioactive nuclei of the activation, in units of 10 -24 cm 2 ; - time; - the probability that a radioactive nucleus will decay in a unit of time; Assume and are constants, are also considered to be constant with respect to irradiation time, = 0, = 0, the initial conditions, from equation (1) （2） Here, The continuous irradiation time is The number of radioactive nuclei generated by activation in the target material per unit volume at the irradiation stop time is 3 ; The activity of the different radionuclides produced by activation in the target material per volume at the end of irradiation for different irradiation times in step 2 This is calculated by the following formula: （3） Here, for the duration of irradiation, in Bq / cm 3 .

2. A method for spatial dose rate radiation protection based on calculation of source terms for irradiance monitoring as claimed in claim 1, wherein, The activity of the different radionuclides produced by activation in the target material per mass at the end of irradiation for different irradiation times in step 2 This is calculated by the following formula: Formula (2) and formula (3) are simplified as: （4） Formula (5) and formula (6) are derived from formula (3) and formula (4) as follows: （5） Or is: （6） In the formula: in units of Bq / g; wherein A is Avogadro's number, having a value of 6.023 x 1023 23 ; A is the mass number of the non-radioactive target nucleus of interest in the target material; the fraction of the total mass of the non-radioactive target nuclei of interest in the target material in the mass of the target material; Saturation factor for activation of radioactivity.

3. A method for spatial dose rate radiation protection based on the calculation of source terms for the irradiance monitor according to claim 1 or 2, characterized in that, The irradiation neutron flux density is as follows: The position of the surveillance sample is between the core boundary and the outer boundary of the pressure vessel, and the radial distribution of the neutron flux from the core boundary to the outer boundary of the pressure vessel in the pressurized water reactor can be measured by the in-core and out-core neutron probes; The radial thickness of each component of the reactor in the radial direction, including the core, the core barrel, the water distance, the core basket, the pressure vessel welding layer, the pressure vessel, the air spacing, the thermal insulation layer, the concrete lining and the concrete lining, is obtained from the equipment manual; The thermal neutron flux density and the fast neutron flux density at the position of the surveillance sample can be found by the neutron flux radial distribution and the radial thickness of each part of the reactor, and the irradiation neutron flux density at the position of the surveillance sample is obtained by adding the thermal neutron flux density and the fast neutron flux density .

4. A method for spatial dose rate radiation protection based on calculation of source terms from irradiance monitoring as claimed in claim 2, wherein, After obtaining the saturated mass specific activity of the surveillance sample by formula (6), the volume specific activity is obtained by the density of the surveillance sample.

5. A method for radiation protection based on the calculation of spatial dose rate from a source term of irradiance monitoring according to claim 2 or 4, characterized in that, According to the calculated mass specific activity, combined with the mass of the surveillance sample and the size and total mass of the cutting product, and referring to the absorption dose rate constant of the corresponding nuclide and the dose rate level in different scenarios calculated by the Monte Carlo simulation, the absorption dose rate and equivalent dose rate at different distances, i.e. the external exposure dose rate level around the removed surveillance sample and the cutting product, are calculated.

6. A method for spatial dose rate radiation protection based on calculation of source terms for irradiance monitoring as claimed in claim 5, wherein, The size, shape and thickness of the special tool and container for the surveillance sample and the cutting product are determined according to the external exposure dose rate level around the surveillance sample and the cutting product.

7. A method for spatial dose rate radiation protection based on calculation of source terms for irradiance monitoring as claimed in claim 6, wherein, When shielding cannot be used, the minimum distance that the personnel must maintain is estimated according to the external exposure dose rate distribution result.

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