Space dose rate radiation protection method based on radiation supervision sample source item calculation

Abstract

The invention relates to the technical field of nuclear power plant radiation protection, and discloses a space dose rate radiation protection method based on irradiation supervision sample source item calculation, which comprises the following steps: step 1, carrying out source item analysis on a taken supervision sample and debris to obtain different nuclide types and quantities generated by neutron activation in a target material per unit volume at the irradiation stop moment, the method comprises the steps of 1, establishing a Monte Carlo calculation model, 2, calculating the activity of different radionuclides generated by activation in target materials of unit volume and unit mass when irradiation is stopped at different irradiation time, 3, establishing the Monte Carlo calculation model, and developing the space dose rate caused by unit mass supervision samples in different scenes, 4, calculating the dose rate distribution of the supervision samples, and 5, calculating the activity of different radionuclides generated by activation in the target materials of unit volume and unit mass when irradiation is stopped at different irradiation time. And 5, according to the dose rate distribution of the supervision sample calculated in the step 4, arranging the types and thicknesses of radiation protection materials in different scenes and positions, and formulating reasonable radiation protection measures in combination with site dose rate requirements.

Description

Technical Field

[0001] The invention relates to the technical field of nuclear power plant radiation protection, and in particular to a space dose rate radiation protection method based on irradiation supervision sample source term calculation. Background Art

[0002] In some pressurized water reactors of nuclear power plants, the upper cylindrical shell section of the reactor pressure vessel corresponds to the core position, and its inner surface is the area of ​​maximum neutron injection of the pressure vessel. Therefore, 12 support seats are arranged on the inner surface of the upper cylindrical shell section, with 2 in a group and each group evenly arranged. The support seats are used to place the steel "irradiation" supervision sample box for the reactor pressure vessel to monitor the changes in the material properties of the reactor pressure vessel under neutron irradiation.

[0003] At present, some nuclear power plant pressurized water reactor pressure vessels have irradiation monitoring samples that are welded and fixed on the inner surface of the pressure vessel. When removing them, special tools are needed to melt and cut the welding points before removing them. For other commercial light water pressurized water reactors, the monitoring samples are screwed in a special channel on the barrel inside the core of the hanging basket. When removing them, special tools are used to unscrew and remove them. The two methods are quite different. Since the pressurized water reactor where the irradiation monitoring sample is welded and fixed is located closer to the pressure vessel, it can better reflect and represent the changes in the properties of the pressure vessel under irradiation and high temperature. However, cutting is required for the removal process, which introduces the problem of increased cutting products and related radiation dosage. Therefore, in the process of removing the monitoring sample of the pressure vessel of the pressurized water reactor, it is necessary to consider not only the radiation protection of the monitoring sample operation process, but also the radiation protection of the monitoring sample cutting products.

[0004] Without conducting relevant source analysis, it is impossible to formulate reasonable radiation protection measures, which often results in insufficient or excessive protection measures. Summary of the invention

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

[0006] The present disclosure provides a space dose rate radiation protection method based on irradiation supervision sample source term calculation, comprising:

[0007] Step 1: Perform source term analysis on the taken surveillance samples and debris, and combine the surveillance sample material composition and natural abundance to obtain the types and quantities of different nuclides produced by neutron activation in the unit volume of target material at the time of irradiation cessation;

[0008] Step 2: Establish a neutron irradiation activation differential equation, and combine the above source term analysis results to calculate the activity of different radionuclides produced by activation per unit volume and per unit mass of the target material when irradiation is stopped at different irradiation times;

[0009] Step 3: Based on the ratio of target material nuclides and the source equivalent to 1.86 MeV gamma rays, a Monte Carlo calculation model is established to conduct spatial dose rate caused by unit mass monitoring samples under different scenarios;

[0010] Step 4: Calculate the dose rate distribution of the surveillance sample based on the calculated activity of the radionuclide and the spatial dose rate per unit mass of the surveillance sample;

[0011] Step 5: Based on the dose rate distribution of the surveillance samples calculated in step 4, arrange the types and thicknesses of radiation protection materials in different scenes and locations. Combined with the dose rate requirements of the site, 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 levels with shielding of different water layer thicknesses, and personnel irradiation dose rate levels with iron shielding above the water layer.

[0013] Optionally, step 1 includes:

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

[0015]

[0016] Where:

[0017] N m is the number of non-radioactive target nuclei of interest per unit volume of neutron-irradiated target material, in units of nuclei / cm 3 ;

[0018] N p The number of radioactive nuclei generated by the activation of target atomic nuclei per unit volume of target material irradiated by neutrons, in units of pieces / cm 3 ;

[0019] σ m is the microscopic activation cross section of the target nucleus, in units of 10 -24 cm 2 ;

[0020] σ p The microscopic "consumption" cross section of the radioactive nucleus generated by activation, in units of 10 -24 cm 2 ;

[0021] φ is the irradiation neutron flux density, in units of neutrons / cm 2 ·s;

[0022] φ b is the neutron flux density used to "consume" the radioactive nuclei produced by activation, in units of 10 -24 cm 2 ;

[0023] t——time;

[0024] λ p ——the probability that a radioactive nucleus will decay per unit time;

[0025] Assume that φ and φ b are constants, N m It is also considered not to change with irradiation time. At "t = 0, N p Under the initial condition of (t) = 0", formula (1) yields:

[0026]

[0027] Here, N p (t) is the number of radioactive nuclei generated by activation in a unit volume of target material when the irradiation is stopped for a continuous irradiation time of t, in units of nuclei / cm 3 .

[0028] Optionally, the activity A of different radionuclides generated by activation in a unit volume of the target material when the irradiation is stopped at different irradiation times in step 2 p (t) is calculated by the following formula:

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

[0030]

[0031] Here, t is the continuous irradiation time, A p The unit of (t) is Bq / cm 3 .

[0032] Optionally, the activity A of different radionuclides produced by activation in the target material per unit mass when the irradiation is stopped at different irradiation times in step 2 pm (t) is calculated by the following formula:

[0033] 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) are simplified to:

[0034]

[0035] The following equations (5) and (6) are derived from equations (3) and (4):

[0036]

[0037] or:

[0038]

[0039] Where:

[0040] A pm The unit of (t) is Bq / g;

[0041] n A is Avogadro's constant, which is 6.023×10 23 ;

[0042] A m is the mass number of the non-radioactive target nucleus of interest in the target material;

[0043] P m is the share of the total mass of the non-radioactive target nuclei of interest in the target material in the mass of the target material;

[0044] is the saturation factor for the activated radioactivity.

[0045] Optionally, the irradiation neutron flux density φ is taken as follows:

[0046] The monitoring sample is located between the core boundary and the outer boundary of the pressure vessel. In a pressurized water reactor, the radial distribution of the neutron flux from the core boundary to the outer boundary of the pressure vessel can be measured by the neutron probes of the in-core measurement and the out-of-core measurement.

[0047] The radial thickness of the radial components of the reactor, including the core, core barrel, water separation, core basket, pressure vessel cladding layer, pressure vessel, air space, insulation layer, concrete lining, the concrete lining is obtained from the equipment manual;

[0048] The above data can be used to find out the thermal neutron flux density and fast neutron flux density at the location of the surveillance sample. The irradiation neutron flux density φ at the location of the surveillance sample can be obtained by adding the thermal neutron flux density and the fast neutron flux density.

[0049] 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 the density of the material used in the surveillance sample.

[0050] Optionally, based on the calculated mass specific activity, combined with the mass of the surveillance sample, 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 levels in different scenarios calculated by Monte Carlo simulation, the absorption dose rate and equivalent dose rate at different distances can be calculated, that is, the external irradiation dose rate level around the removed surveillance sample and the cutting product.

[0051] Optionally, the size, shape and thickness of the special tools and containers are determined according to the external radiation dose rate level around the surveillance sample and cutting products.

[0052] Optionally, when shielding cannot be used, the results of the external radiation dose rate distribution can also be used to estimate the minimum distance that personnel must maintain.

[0053] Compared with the prior art, the technical solution provided by the embodiments of the present disclosure has the following advantages:

[0054] The embodiment of the present disclosure provides a method for radiation protection of space dose rate based on source term calculation of irradiation monitoring sample. The method first obtains the types and quantities of different nuclides produced by neutron activation in a unit volume of target material at the moment of irradiation stopping, and then establishes a differential equation for neutron irradiation activation. In combination with the above-mentioned source term analysis results, the activities of different radioactive nuclides produced by activation in a unit volume and unit mass of target material at different irradiation times are calculated. Based on the nuclide ratio of the target material and the source equivalent to 1.86MeV of gamma rays, a Monte Carlo calculation model is established to carry out space dose rates caused by unit mass monitoring samples in different scenarios. Based on the calculated activities of the radioactive nuclides and The spatial dose rate caused by the unit mass surveillance sample is calculated, and the surveillance sample dose rate distribution is calculated. Finally, according to the calculated surveillance sample dose rate distribution, the types and thicknesses of radiation protection materials in different scenes and positions are arranged. Combined with the site dose rate requirements, reasonable radiation protection measures are formulated to reduce the radiation dose rate of personnel. Therefore, based on the existing information and data of the nuclear power plant, the external exposure dose rate level around each radiation source of the surveillance sample and the cutting product can be calculated before starting work. This part is calculated through the Monte Carlo calculation model to avoid unnecessary radioactive exposure caused by radiation protection personnel in the process of detecting radiation dose levels, and reduce the exposure dose of personnel during the overhaul of the power plant. BRIEF DESCRIPTION OF THE DRAWINGS

[0055] Figure 1 It is a schematic diagram of the structure of the reactor pressure vessel in the current pressurized water reactor of a nuclear power plant;

[0056] Figure 2 is a radial distribution diagram of neutron flux from the core boundary to the outer boundary of the pressure vessel in an embodiment of the present invention;

[0057] Figure 3 A schematic diagram of a Monte Carlo calculation model in an embodiment of the present invention;

[0058] Figure 4 is the space dose rate level in the air when the monitoring sample is not shielded in the embodiment of the present invention;

[0059] Figure 5 Monitor sample space dose rate levels for different lead shielding in the embodiment of the present invention;

[0060] Figure 6 The dose rate level of the monitoring sample space at different water layer thicknesses in the embodiment of the present invention;

[0061] Figure 7 This is the dose rate change in a place with a water layer thickness of 200 cm + 2 cm iron shielding in the embodiment of the present invention.

[0062] Description of reference numerals:

[0063] 1-main body, 2-cover body, 3-core, 4-irradiation monitoring sample box, 5-monitoring sample, 6-lower end cover, 7-upper end cover, 8-multi-layer lead shielding layer, 9-internal cavity, 10-outer stainless steel wall. DETAILED DESCRIPTION

[0064] A specific implementation of the present invention is described in detail below in conjunction with the accompanying drawings, but it should be understood that the protection scope of the present invention is not limited by the specific implementation.

[0065] At present, some nuclear power plant pressurized water reactor pressure vessels have irradiation monitoring samples that are welded and fixed on the inner surface of the pressure vessel. When removing them, special tools are needed to melt and cut the welding points before removing them. For other commercial light water pressurized water reactors, the monitoring samples are screwed in a special channel on the barrel inside the core of the hanging basket. When removing them, special tools are used to unscrew and remove them. The two methods are quite different. Since the pressurized water reactor where the irradiation monitoring sample is welded and fixed is located closer to the pressure vessel, it can better reflect and represent the changes in the properties of the pressure vessel under irradiation and high temperature. However, cutting is required for the removal process, which introduces the problem of increased cutting products and related radiation dosage. Therefore, in the process of removing the monitoring sample of the pressure vessel of the pressurized water reactor, it is necessary to consider not only the radiation protection of the monitoring sample operation process, but also the radiation protection of the monitoring sample cutting products.

[0066] Without relevant source analysis, it is impossible to formulate reasonable radiation protection measures, which often results in insufficient or excessive protection measures. If the radiation parameters of the activated irradiated monitoring sample itself and the radiation parameters of the cutting product (cutting slag) can be analyzed, relevant personnel can be effectively guided to formulate reasonable radiation protection measures, saving manpower and material costs.

[0067] To this end, an embodiment of the present disclosure provides a space dose rate radiation protection method based on the calculation of the irradiation supervision sample source term. The present disclosure belongs to the field of radiation protection of nuclear power plants and provides a way of thinking for removing the supervision sample after activation and calculating its own radiation parameters and the radiation parameters of the cutting product (cutting slag), thereby effectively guiding relevant personnel to formulate reasonable radiation protection measures.

[0068] In order to more clearly explain the space dose rate radiation protection method based on irradiation supervision sample source term calculation in the present disclosure, the pressure vessel structure of the nuclear power plant pressurized water reactor is first explained, such as Figure 1 As shown, the body 1 and the cover 2 of the pressure vessel are connected by a flange 1, and the cover 2 corresponds to the core 3, so its inner surface is the maximum neutron injection area of ​​the pressure vessel. Therefore, the monitoring sample box 4 is fixed on the cover 2, and the irradiation monitoring sample box 4 is in the highest neutron flux area of ​​the pressure vessel to monitor the changes in material properties of the reactor pressure vessel under neutron irradiation.

[0069] Compared with the irradiation monitoring sample fixed in the dedicated channel of the core of the hanging basket by spiral method, the process of removing the irradiation monitoring sample fixed by welding on the inner surface of the pressure vessel is more complicated. Before calculating the radiation source item of the irradiation monitoring sample, the irradiation monitoring sample fixed by welding on the inner surface of the pressure vessel is cut, and the cutting debris is collected, and the monitoring sample and the debris are taken out. The operation process of removing the monitoring sample is divided into three stages:

[0070] The first stage is the cutting and removal stage;

[0071] The second stage is the grinding operation stage;

[0072] The third stage is the debris collection operation stage.

[0073] At least one embodiment of the present disclosure provides a method for space dose rate radiation protection based on radiation monitoring sample source item calculation, the calculation method is used to calculate the radiation parameters of various radiation sources of the radiation monitoring sample and the cutting product welded and fixed on the inner surface of the pressure vessel, and the calculation method of the radiation parameters of each radiation source includes:

[0074] Step 1: Perform source term analysis on the taken surveillance samples and debris, and combine the surveillance sample material composition and natural abundance to obtain the types and quantities of different nuclides produced by neutron activation in the unit volume of target material at the time of irradiation cessation;

[0075] Step 2: Establish a neutron irradiation activation differential equation, and combine the above source term analysis results to calculate the activity of different radionuclides produced by activation per unit volume and per unit mass of the target material when irradiation is stopped at different irradiation times;

[0076] Step 3: Based on the ratio of target material nuclides and the source equivalent to 1.86 MeV gamma rays, a Monte Carlo calculation model is established to conduct spatial dose rate caused by unit mass monitoring samples under different scenarios;

[0077] Step 4: Calculate the dose rate distribution of the surveillance sample based on the calculated activity of the radionuclide and the spatial dose rate per unit mass of the surveillance sample;

[0078] Step 5: Based on the dose rate distribution of the surveillance samples calculated in step 4, arrange the types and thicknesses of radiation protection materials in different scenes and locations. Combined with the site dose rate requirements (not higher than 0.5mSv / h), formulate reasonable radiation protection measures to reduce the radiation dose rate of personnel.

[0079] The present disclosure provides a method for radiation protection of space dose rate based on source term calculation of irradiation supervision sample. The method first obtains the types and quantities of different nuclides produced by neutron activation in a unit volume of target material at the moment of irradiation stopping, and then establishes a differential equation for neutron irradiation activation. Combined with the above-mentioned source term analysis results, the activities of different radioactive nuclides produced by activation in a unit volume and unit mass of target material at different irradiation times are calculated. Based on the nuclide ratio of the target material and the source equivalent to 1.86MeV of gamma rays, a Monte Carlo calculation model is established to carry out unit mass calculations under different scenarios. The space dose rate caused by the surveillance sample is calculated based on the calculated activity of the radioactive nuclides and the space dose rate caused by the surveillance sample per unit mass, and the distribution of the surveillance sample dose rate is calculated. Finally, according to the calculated distribution of the surveillance sample dose rate, the types and thicknesses of radiation protection materials in different scenes and positions are arranged. Combined with the site dose rate requirements, reasonable radiation protection measures are formulated to reduce the radiation dose rate of personnel. Therefore, based on the existing information and data of the nuclear power plant, the external exposure dose rate level around each radiation source of the surveillance sample and the cutting product can be calculated before the work is carried out. This part is calculated by the Monte Carlo calculation model, such as Figure 3 As shown, the Monte Carlo calculation model includes a monitoring sample 5, which is placed inside a shielding barrel. The shielding barrel consists of a lower end cover 6, an upper end cover 7, a multi-layer lead shielding layer 8, an internal cavity 9 and an outer stainless steel wall 10, so as to avoid unnecessary radioactive exposure caused by radiation protection personnel in the process of detecting radiation dose levels and reduce the exposure dose of personnel during the overhaul of the power plant.

[0080] 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 shielding of different water layer thicknesses, and the personnel irradiation dose rate level with iron shielding above the water layer.

[0081] Specifically, step 1 includes:

[0082] The number of activated specific radioactive nuclei generated by the non-radioactive target nuclei of interest in the target material irradiated by neutrons satisfies the following differential equation:

[0083]

[0084] Where:

[0085] N m is the number of non-radioactive target nuclei of interest per unit volume of neutron-irradiated target material, in units of nuclei / cm 3 ;

[0086] N p The number of radioactive nuclei generated by the activation of target atomic nuclei per unit volume of target material irradiated by neutrons, in units of pieces / cm 3 ;

[0087] σ m is the microscopic activation cross section of the target nucleus, in units of 10 -24 cm 2 ;

[0088] σ p The microscopic "consumption" cross section of the radioactive nucleus generated by activation, in units of 10 -24 cm 2 ;

[0089] φ is the irradiation neutron flux density, in units of neutrons / cm 2 ·s;

[0090] φ b is the neutron flux density used to "consume" the radioactive nuclei produced by activation, in units of 10 -24 cm 2 ;

[0091] t——time;

[0092] λ p ——the probability that a radioactive nucleus will decay per unit time;

[0093] Assume that φ and φ b are constants, N m It is also considered not to change with irradiation time. At "t = 0, N p Under the initial condition of (t) = 0", formula (1) yields:

[0094]

[0095] Here, N p (t) is the number of radioactive nuclei generated by activation in a unit volume of target material when the irradiation is stopped for a continuous irradiation time of t, in units of nuclei / cm 3 .

[0096] Specifically, the activity A of different radionuclides generated by activation in a unit volume of target material when irradiation is stopped at different irradiation times in step 2 is p (t) is calculated by the following formula:

[0097] Formula (2) can also be expressed in another form:

[0098]

[0099] Here, t is the continuous irradiation time, A p The unit of (t) is Bq / cm 3 .

[0100] Furthermore, the activity A of different radionuclides produced by activation in the target material per unit mass when the irradiation is stopped at different irradiation times in step 2 is pm (t) is calculated by the following formula:

[0101] 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) are simplified to:

[0102]

[0103] The following equations (5) and (6) are derived from equations (3) and (4):

[0104]

[0105] or:

[0106]

[0107] Where:

[0108] A pm The unit of (t) is Bq / g;

[0109] n A is Avogadro's constant, which is 6.023×10 23 ;

[0110] A m is the mass number of the non-radioactive target nucleus of interest in the target material;

[0111] P m is the share of the total mass of the non-radioactive target nuclei of interest in the target material in the mass of the target material;

[0112] is the saturation factor for the activated radioactivity.

[0113] Optionally, the irradiation neutron flux density φ is taken as follows:

[0114] The monitoring sample is located between the core boundary and the outer boundary of the pressure vessel. In a pressurized water reactor, the radial distribution of the neutron flux from the core boundary to the outer boundary of the pressure vessel can be measured by the neutron probes of the in-core measurement and the out-of-core measurement, such as Figure 2 As shown;

[0115] The radial thickness of the radial components of the reactor, including the core, core barrel, water separation, core basket, pressure vessel cladding layer, pressure vessel, air space, insulation layer, concrete lining, the concrete lining is obtained from the equipment manual;

[0116] The above data can be used to find out the thermal neutron flux density and fast neutron flux density at the location of the surveillance sample. The irradiation neutron flux density φ at the location of the surveillance sample can be obtained by adding the thermal neutron flux density and the fast neutron flux density.

[0117] Optionally, the probability that a radioactive nucleus decays per unit time λ p Refer to the kernel parameter values ​​given in the kernel database.

[0118] 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 the density of the material used in the surveillance sample.

[0119] Optionally, based on the calculated mass specific activity, combined with the mass of the surveillance sample, 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 levels in different scenarios calculated by Monte Carlo simulation, the absorption dose rate and equivalent dose rate at different distances can be calculated, that is, the external irradiation dose rate level around the removed surveillance sample and the cutting product.

[0120] Optionally, the size, shape and thickness of the special tools and containers are determined according to the external radiation dose rate level around the surveillance sample and cutting products.

[0121] Optionally, when shielding cannot be used, the results of the external radiation dose rate distribution can also be used to estimate the minimum distance that personnel must maintain.

[0122] The overall process is as follows

[0123] Activation specific radioactivity calculation formula

[0124] The number of activated specific radioactive nuclei generated by the non-radioactive target nuclei of interest in the target material irradiated by neutrons satisfies the following differential equation:

[0125]

[0126] Where:

[0127] Nm is the number of non-radioactive target nuclei of interest per unit volume of neutron-irradiated target material, in units of nuclei / cm 3 ;

[0128] N p The number of radioactive nuclei generated by the activation of target atomic nuclei per unit volume of target material irradiated by neutrons, in units of pieces / cm 3 ;

[0129] σ m is the microscopic activation cross section of the target nucleus, in units of 10 -24 cm 2 ;

[0130] σ p The microscopic "consumption" cross section of the radioactive nucleus generated by activation, in units of 10 -24 cm 2 ;

[0131] φ is the irradiation neutron flux density, in units of neutrons / cm 2 ·s;

[0132] φ b is the neutron flux density used to "consume" the radioactive nuclei produced by activation, in units of 10 -24 cm 2 ;

[0133] t——time;

[0134] λ p ——the probability that a radioactive nucleus will decay per unit time;

[0135] Assume that φ and φ b are constants, N m It is also considered not to change with irradiation time. At "t = 0, N p Under the initial condition of (t) = 0", formula (1) yields:

[0136]

[0137] Here, N p (t) is the number of radioactive nuclei generated by activation in the target material per unit volume at the time of stopping irradiation for the continuous irradiation time t (pieces / cm 3 ).

[0138] Formula (2) can also be expressed in another form:

[0139]

[0140] Here, A p(t) is the activity of radionuclides generated by activation in the target material per unit volume at the time of cessation of irradiation when the irradiation lasts for t, Bq / cm 3 .

[0141] 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) are simplified to:

[0142]

[0143] From (3) and (4), we can deduce (5) and (6). pm (t) is the activity of radionuclides generated by activation per unit mass of target material when the irradiation lasts for t (Bq / g):

[0144]

[0145] or:

[0146]

[0147] Where:

[0148] n A is Avogadro's constant, which is 6.023×10 23 ;

[0149] A m is the mass number of the non-radioactive target nucleus of interest in the target material;

[0150] P m It is the proportion of the total mass of the non-radioactive target nuclei of interest in the target material to the mass of the target material.

[0151] In most cases, the activity of activated radionuclides can be calculated using equations (4) and (6). By analyzing equations (4) and (6), the following very useful information can be obtained:

[0152] In the formula It is the saturation factor of activated radioactivity. When the irradiation time is 1 half-life, it reaches 50% of the saturation value; when the irradiation time is 2 half-lives, it reaches 75% of the saturation value; longer irradiation time does not contribute much to the increase of activated radioactivity. It can be roughly considered that when the irradiation time reaches more than 5 half-lives, the activated radioactivity reaches the saturation value;

[0153] When the irradiation time is less than one half-life, the activated radioactivity is approximately linearly related to the irradiation time.

[0154] After the irradiation stops, the remaining radioactivity after a period of cooling can be calculated using the commonly used decay formula.

[0155] Related input parameters

[0156] 2.1 Neutron flux density

[0157] The monitoring sample is located between the core boundary and the outer boundary of the pressure vessel. In the pressurized water reactor, the radial distribution of the neutron flux from the core boundary to the outer boundary of the pressure vessel can be measured by the neutron probes of the in-core measurement and the out-of-core measurement.

[0158] The radial thickness of the radial components of the reactor (including the core, core cask, water isolation distance, core basket, pressure vessel cladding layer, pressure vessel, air space, insulation layer, concrete lining, concrete lining, etc.) can be found in the equipment manual.

[0159] The above data can be used to obtain the thermal neutron flux and fast neutron flux at the location of the monitored sample.

[0160] 2.2 Half-life / decay constant

[0161] The nuclear cross section and the half-life / decay constant of the nuclides generated by activation can refer to the nuclear parameter values ​​given by the China Nuclear Database, as shown in Table 1:

[0162] Table 1 Nuclear cross sections of relevant nuclear reactions and half-lives / decay constants of generated nuclides

[0163]

[0164] The (n, p) reaction cross section for the activation reaction of two fast neutrons is the average (n, p) reaction cross section of the entire fission neutron spectrum.

[0165] The fast neutron flux distribution given by FSAR is not the full fission neutron spectrum. Considering the large difference in cross-section conditions, the activation radioactivity produced by these two fast neutron (n, p) reactions was not verified.

[0166] When estimating the radiation field, the main consideration is the emission of higher energy gamma rays. 60 Co and 59 Contribution of Fe, 58 Co and 54 Mn even 51 The contribution of Cr can be relatively neglected.

[0167] 2.3 Composition of target nuclides in activated metal materials

[0168] The activated metal material is selected according to the actual material used in the power station, 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:

[0169] Table 2 Element composition of stainless steel, natural abundance and mass fraction of target nuclides

[0170]

[0171]

[0172] Estimation results

[0173] By using formula (6) and substituting the above parameters into the equation, 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 7.8 g / cm3 for stainless steel).

[0174] The time for taking out the surveillance samples of this unit was initially set at 4, 8, 12, 20, 28 and 36 years. For this purpose, the corresponding time was used as the moment to stop irradiation, and the corresponding mass specific activity was estimated, that is, the mass specific activity of the taken out surveillance samples and cutting products.

[0175] 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 trend was 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.

[0176] Radiation protection applications

[0177] 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 absorption dose rate and equivalent dose rate at different distances can be easily calculated, that is, the external radiation dose rate level around the removed surveillance sample and the cutting products.

[0178] 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.

[0179] When shielding cannot be used, the results of the external radiation dose rate distribution can also be used to estimate the minimum distance that personnel must maintain; the staff can be guided to determine the size, shape and thickness of their special tools and containers based on the actual different external radiation dose rate levels calculated, so that the shielding effect reaches a reasonably feasible and as low result as possible, which ensures safety and saves costs; when shielding cannot be used, the staff can be guided to understand the minimum distance that must be maintained from the radiation source; the staff can be guided to fully collect the cutting products (debris - formed by melting in water, close to a sphere) to prevent the debris from being transferred to the auxiliary system after the core is activated and then forming "hot particles" on the pipeline; the design of the present invention can guide the staff to prepare corresponding protective measures in advance and reasonably arrange the work plan in advance, which can effectively shorten the actual construction period of the corresponding work.

[0180] The present disclosure is described below by a specific embodiment. In order to keep the following description of the embodiment of the present disclosure clear and concise, detailed descriptions of known functions and known components may be omitted.

[0181] Example 1

[0182] The thermal neutron activation saturated mass specific activity of the surveillance sample is calculated by formula (7). The thermal neutron flux, nuclear cross section, half-life / decay constant of the nuclides generated by activation, and mass fraction of each element in the target nuclide at the surveillance sample location can be calculated / queried in the following way.

[0183] 1) Obtaining neutron flux

[0184] The monitoring sample is located between the core boundary and the outer boundary of the pressure vessel. Therefore, the neutron flux can be measured by the radial distribution of the neutron flux from the core boundary to the outer boundary of the pressure vessel in the pressurized water reactor, which can be measured by the neutron probes of the in-core measurement and the out-of-core measurement, such as Figure 2 As shown;

[0185] The radial thickness of the radial components of the reactor (including the core, core cask, water isolation distance, core basket, pressure vessel cladding layer, pressure vessel, air space, insulation layer, concrete lining, concrete lining, etc.) can be checked in the equipment manual.

[0186] Based on the above data, the thermal neutron flux and fast neutron flux at the location of the monitoring sample are obtained from the manual. Some figures in the manual are shown in Figure 2 .

[0187] 2) Target nuclide half-life / decay constant

[0188] The nuclear cross section and the half-life / decay constant of the nuclides generated by activation are the nuclear parameter values ​​given by the China Nuclear Database. The half-life / decay constant of the nuclides involved are shown in Table 1:

[0189] The two fast neutron (n, p) activation reactions are based on the average (n, p) reaction cross section of the entire fission neutron spectrum. When estimating the radiation field, 58 Co and 54 Mn even 51 The contribution of Cr is relatively small and can be ignored. Therefore, this example mainly considers the emission of higher energy gamma rays. 60 Co and 59 The contribution of Fe is equivalent to a γ-ray energy of 1.86 MeV.

[0190] 3) Mass fraction of target nuclides

[0191] The activated metal material is selected according to the actual monitoring sample material of the nuclear power plant, which can be found in the equipment technical documents. The monitoring sample material of this example is stainless steel. The composition of the target nuclides of the unit involved is shown in Table 2:

[0192] 4) Calculate mass specific activity / volume specific activity

[0193] Substitute the parameters such as neutron flux, nuclear cross section, half-life / decay constant of the generated nuclides, and mass fraction of each element in the target nuclide into equation (7) to calculate the saturated mass specific activity of the thermal neutron activation of the surveillance sample. Combined with the density of the surveillance sample material (e.g. 7.8 g / cm for stainless steel), 3 ) can be used to obtain the volume specific activity, see Table 3.

[0194] Table 3 Thermal neutron activation saturation mass specific activity of surveillance samples

[0195]

[0196] 5) Neutron activation mass specific activity in different years

[0197] The removal time of the monitoring sample of the unit involved in the present invention 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 time to stop irradiation, and the corresponding mass specific radioactivity is estimated based on this, that is, the mass specific radioactivity of the removed monitoring sample and the cutting product. The calculation results are shown in Table 4.

[0198] Table 4 Mass specific activity of thermal neutron activation of surveillance samples in different years

[0199]

[0200] According to the expression of saturation factor It can be seen that the monitoring sample and its cutting products 60 The mass activity of Co increased rapidly in the first few years, but the increase tended to be very slow after 20 years. 51 Cr, 59 Fe, 58 Co. 54Since 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.

[0201] 6) Dose rate levels at different locations

[0202] 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 the radionuclide absorption dose rate constant involved, the absorbed dose rate and equivalent dose rate at different distances can be calculated, that is, the external exposure dose rate level around the removed surveillance sample and cutting products.

[0203] Figure 4 This is the level of space dose rate in the air when the monitoring sample is not shielded. The dose rate is about 370.07mSv / h at 1m, and then decays to 14.33mSv / h at 5m, which is about 96.1% decayed from 1m to 5m. However, engineering requirements require that when the site dose rate is greater than 0.5mSv / h, the staff should stop working, evacuate to a safe area, and report to the radiation protection personnel to confirm the radiation status. Therefore, when the monitoring sample is taken out, the space dose rate is high, and shielding measures need to be taken for the monitoring sample.

[0204] Figure 5 The spatial dose rate level of the surveillance sample under different lead shielding conditions. The dose rate change of the surveillance sample under different lead shielding thicknesses shows a logarithmic function trend. When the lead shielding thickness is small, the dose rate of the surveillance sample drops rapidly and then tends to a stable value. When the lead shielding thickness is 20cm, the lead surface dose rate is 0.94mSv / h, and when the lead shielding thickness is 23cm, the lead surface dose rate is 0.18mSv / h. Therefore, for 1.86MeV gamma rays, 23cm of lead material can reduce the dose rate of the surveillance sample to less than 0.5mSv / h. The minimum thickness of the special shielding barrel for the surveillance sample used in the project is 25cm, which can meet the shielding requirements.

[0205] Figure 6 The spatial dose rate level of the surveillance sample at different water layer thicknesses is 0.53mSv / h when the water layer thickness is 160cm, and 0.19mSv / h when the water layer thickness is 180cm. Therefore, when the water layer thickness is 180cm, the site dose rate can meet the requirement of less than 0.5mSv / h. Before the surveillance sample is placed in a special protective shielding container, the thickness of the water layer is reduced to 200cm, and the water surface dose rate is 0.06mSv / h. At this time, the dose rate of the surveillance sample placed in the shielding barrel meets the requirement. Figure 7This is the change in dose rate in a place where the water layer thickness is 200cm, there is 2cm iron shielding material above the water layer and there is air above the iron material. Adding 2cm of iron material above the water layer can attenuate the dose rate of the 200cm water layer thickness to 0.02mSv / h. Therefore, when the water layer thickness is 200cm+2cm, the radiation to the staff in the surveillance sample is smaller.

[0206] The above disclosures are only several specific embodiments of the present invention. However, the embodiments of the present invention are not limited thereto. Any changes that can be conceived by those skilled in the art should fall within the protection scope of the present invention.

Claims

1. A method for space dose rate radiation protection based on irradiation supervision sample source term calculation, characterized in that: include: Step 1: Perform source term analysis on the taken surveillance samples and debris, and combine the surveillance sample material composition and natural abundance to obtain the types and quantities of different nuclides produced by neutron activation in the unit volume of target material at the time of irradiation cessation; Step 2: Establish a neutron irradiation activation differential equation, and combine the above source term analysis results to calculate the activity of different radionuclides produced by activation per unit volume and per unit mass of the target material when irradiation is stopped at different irradiation times; Step 3: Based on the ratio of target material nuclides and the source equivalent to 1.86 MeV gamma rays, a Monte Carlo calculation model is established to conduct spatial dose rate caused by unit mass monitoring samples under different scenarios; Step 4: Calculate the dose rate distribution of the surveillance sample based on the calculated activity of the radionuclide and the spatial dose rate per unit mass of the surveillance sample; Step 5: Based on the dose rate distribution of the surveillance samples calculated in step 4, arrange the types and thicknesses of radiation protection materials in different scenes and locations, and formulate reasonable radiation protection measures to reduce the radiation dose rate of personnel in combination with the dose rate requirements of the site.

2. A method for space dose rate radiation protection based on irradiation supervision sample source term calculation according to claim 1, characterized in that: 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 shielding of different water layer thicknesses, and the personnel irradiation dose rate level with iron shielding above the water layer.

3. A method for space dose rate radiation protection based on irradiation supervision sample source term calculation as claimed in claim 2, characterized in that: The step 1 comprises: The number of activated specific radioactive nuclei generated by the non-radioactive target nuclei of interest in the target material irradiated by neutrons satisfies the following differential equation: Where: N m is the number of non-radioactive target nuclei of interest per unit volume of neutron-irradiated target material, in units of nuclei / cm 3 ; N p The number of radioactive nuclei generated by the activation of target atomic nuclei per unit volume of target material irradiated by neutrons, in units of pieces / cm 3 ; σ m is the microscopic activation cross section of the target nucleus, in units of 10 -24 cm 2 ; σ p The microscopic "consumption" cross section of the radioactive nuclei generated by activation, in units of 10 -24 cm 2 ; φ is the irradiation neutron flux density, in units of neutrons / cm 2 ·s; φ b is the neutron flux density used to "consume" the radioactive nuclei produced by activation, in units of 10 -24 cm 2 ; t——time; λ p ——the probability that a radioactive nucleus will decay per unit time; Assume that φ and φ b are constants, N m It is also considered not to change with irradiation time. At "t=0, N p Under the initial condition of (t) = 0", formula (1) yields: Here, N p (t) is the number of radioactive nuclei generated by activation in a unit volume of target material when the irradiation is stopped for a continuous irradiation time of t, in units of nuclei / cm 3 .

4. A method for space dose rate radiation protection based on irradiation supervision sample source term calculation as claimed in claim 3, characterized in that: The activity A of different radionuclides generated by activation in the target material per unit volume when the irradiation is stopped at different irradiation times in step 2 p (t) is calculated by the following formula: Formula (2) can also be expressed in another form: Here, t is the continuous irradiation time, A p The unit of (t) is Bq / cm 3 .

5. A method for space dose rate radiation protection based on irradiation supervision sample source term calculation as claimed in claim 4, characterized in that: The activity A of different radionuclides produced by activation in the target material per unit mass when the irradiation is stopped at different irradiation times in step 2 pm (t) is calculated by the following formula: 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) are simplified to: A p (t)=N m s m φ[1-e -λpt ] (4) The following equations (5) and (6) are derived from equations (3) and (4): or: Where: A pm The unit of (t) is Bq / g; n A is Avogadro's constant, which is 6.023×10 23 ; A m is the mass number of the non-radioactive target nucleus of interest in the target material; P m is the share of the total mass of the non-radioactive target nuclei of interest in the target material in the mass of the target material; [1-e -λpt ] is the saturation factor of activated radioactivity.

6. A method for space dose rate radiation protection based on irradiation supervision sample source term calculation as claimed in claim 3 or 5, characterized in that: The value of the irradiation neutron flux density φ is as follows: The monitoring sample is located between the core boundary and the outer boundary of the pressure vessel. In a pressurized water reactor, the radial distribution of the neutron flux from the core boundary to the outer boundary of the pressure vessel can be measured by the neutron probes of the in-core measurement and the out-of-core measurement. The radial thickness of the radial components of the reactor, including the core, core barrel, water separation, core basket, pressure vessel cladding layer, pressure vessel, air space, insulation layer, concrete lining, the concrete lining is obtained from the equipment manual; The thermal neutron flux density and the fast neutron flux density at the location of the monitoring sample can be found through the above data. The irradiation neutron flux density φ at the location of the monitoring sample can be obtained by adding the thermal neutron flux density and the fast neutron flux density.

7. A method for space dose rate radiation protection based on irradiation supervision sample source term calculation as claimed in claim 5, characterized in that: After obtaining the thermal neutron activation saturation mass specific activity of the monitoring sample through formula 6, the volume specific activity is obtained through the density of the material used in the monitoring sample.

8. A method for space dose rate radiation protection based on irradiation supervision sample source term calculation as claimed in claim 5 or 7, characterized in that: According to the calculated mass specific activity, combined with the mass of the surveillance sample, 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 Monte Carlo simulation, the absorbed dose rate and equivalent dose rate at different distances can be calculated, that is, the external irradiation dose rate level around the removed surveillance sample and the cutting product.

9. A method for space dose rate radiation protection based on irradiation supervision sample source term calculation as claimed in claim 8, characterized in that: The size, shape and thickness of the special tools and containers are determined according to the external radiation dose rate level around the surveillance sample and the cutting product.

10. A method for space dose rate radiation protection based on irradiation supervision sample source term calculation as claimed in claim 9, characterized in that: When shielding cannot be used, the minimum distance that personnel must maintain can also be estimated based on the results of the external radiation dose rate distribution.

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