A method for estimating the release source term under emergency conditions in a sodium-cooled fast reactor based on core damage.

Abstract

This invention relates to a method for estimating the release term of a sodium-cooled fast reactor under emergency conditions based on core damage. The method calculates the core stockpile of nuclides based on their cumulative yield, and calculates the source term released from the core to the primary circuit sodium pool by combining the core damage share and the release share of each nuclide. It also calculates the source term of radionuclides released from the core to the covering gas cavity based on the release share from the core to the covering gas cavity. If the accident type is determined to be plug seal loss, the rate of change of the source term released to the containment is calculated; if the accident type is determined to be double-walled vessel meltdown or leakage, the rate of change of the gas-carried source term in the crater is calculated; and the source term released to the environment is further calculated. Using the estimation method disclosed in this invention, it is possible to calculate the dynamic, phased release term to the primary circuit, containment, and environment under different release pathways based on facility operation data and core stockpile during a sodium-cooled fast reactor accident, providing a basis for decision-making regarding appropriate protective measures in emergency situations.

Description

Technical Field

[0001] This invention belongs to the field of nuclear accident emergency assessment, specifically involving a method for estimating the release source term of a sodium-cooled fast reactor under emergency operating conditions based on core damage. Background Technology

[0002] The Fukushima accident has once again sounded the alarm for humanity, demonstrating that although the probability of a serious reactor accident is extremely low, it can still occur. Section 3.1.2 of the National Nuclear Emergency Response Plan requires that "accident condition diagnosis and release term analysis be carried out" after a nuclear accident. Section 7.6 of the Nuclear Safety Guideline HAD002 / 01-2019, "Emergency Preparedness and Response of Nuclear Power Plant Operators," requires operating units to "collect and understand the necessary data for assessment, including the evolution of the accident, source terms, and meteorological parameters of the nuclear power plant location and surrounding areas." Appendix B requires that the emergency plan "explain the methods and arrangements for obtaining parameters (estimated source terms, radiation measurement results of the containment and effluents, and meteorological parameters)." It also points out that "although the possibility of a nuclear power plant entering a nuclear emergency state due to error or accident after taking various preventive measures is very small, it cannot be completely ruled out. In order to strengthen emergency response capabilities so that an accident can be quickly and effectively controlled and its consequences mitigated, nuclear power plants should have comprehensive emergency plans and sufficient emergency preparedness." The estimation of the release source terms of nuclear power plants is a major component of emergency assessment.

[0003] Therefore, it is necessary to estimate the release source terms of the sodium-cooled fast reactor under emergency operating conditions, and then assess the impact of the release source terms on the surrounding environment and the measures that should be taken. Summary of the Invention

[0004] To address the shortcomings of existing technologies, the present invention aims to provide a method for estimating the release source terms of a sodium-cooled fast reactor under emergency operating conditions based on core damage. This method can calculate the dynamic, phased release source terms to the primary circuit, containment, and environment under different release pathways based on facility operation data, core accumulation, and core damage share during a sodium-cooled fast reactor accident, providing a basis for decision-making in taking appropriate protective measures in emergency situations.

[0005] To achieve the above objectives, the technical solution adopted by the present invention is: a method for estimating the release source term under emergency operating conditions of a sodium-cooled fast reactor based on core damage, the method comprising the following steps:

[0006] S11. Calculate the core stockpile of nuclides based on the cumulative yield of nuclides;

[0007] S12. Calculate the source term of the core released to the primary loop sodium pool based on the core accumulation of nuclides, the core damage share, and the release share of each nuclide.

[0008] S13. Calculate the source term of radionuclides released from the core to the covered gas cavity based on the core accumulation of nuclides, the core damage share, and the release share from the core to the covered gas cavity.

[0009] S14. If the accident type is determined to be loss of plug seal, the airborne radionuclides will be released from the covered air cavity into the containment. Calculate the rate of change of the source term released into the containment.

[0010] S15. If the accident type is determined to be a double-walled container meltdown or leakage, radioactive material enters the crater. Calculate the rate of change of the gaseous source term in the crater.

[0011] S16. Calculate the source terms released to the environment based on the rate of change of the source terms released to the containment and the rate of change of the airborne source terms in the crater.

[0012] Furthermore, the cumulative yield of nuclides in step S11 is the cumulative yield of nuclides under fast neutron conditions, which is obtained by consulting publicly available data from foreign literature.

[0013] Furthermore, the release proportions of each nuclide in step S12 were obtained by consulting publicly available data from foreign literature.

[0014] Furthermore, in step S13, the proportion of elements released from the reactor core to the cover gas at different temperatures is obtained based on the chemical equilibrium method disclosed in foreign literature.

[0015] Furthermore, step S14 includes the following sub-steps:

[0016] S141. Calculate the flow velocity of sodium vapor towards the containment vessel;

[0017] S142. Calculate the separation coefficient of radionuclides as sodium coolant evaporates;

[0018] S143. Calculate the aerosol deposition rate;

[0019] The rate of change of the source term released into the containment structure is calculated based on the parameters mentioned above.

[0020] Furthermore, step S15 includes the following sub-steps:

[0021] S151. Calculate the combustion rate of sodium;

[0022] S152. Calculate the combustion separation coefficients of different nuclides;

[0023] S143. Calculate the aerosol deposition rate; calculate the rate of change of the gaseous source term in the crater based on the above parameters.

[0024] Furthermore, when some input parameters were unavailable, the combustion rate of sodium was determined using experimental results from the Humphrey Laboratory in the United States.

[0025] Furthermore, the combustion separation coefficients of different nuclides were obtained by consulting publicly available experimental data from abroad.

[0026] The advantages of this invention are as follows: By employing the method disclosed in this invention for estimating the release source terms of a sodium-cooled fast reactor under emergency operating conditions, it is possible to calculate the dynamic, phased release source terms to the primary circuit, containment, and environment under different release pathways based on facility operation data, core accumulation, and core damage share when a sodium-cooled fast reactor accident occurs. This provides a reliable basis for decision-making regarding the assessment of the impact of the release source terms on the surrounding environment and the protective measures to be taken in emergency situations. Attached Figure Description

[0027] Figure 1 This is a flowchart illustrating a method for estimating the release source term under emergency conditions in a sodium-cooled fast reactor based on core damage, as described in an embodiment of the present invention. Detailed Implementation

[0028] The present invention will now be further described with reference to the accompanying drawings and specific embodiments.

[0029] Example 1

[0030] like Figure 1 As shown, this embodiment of the invention provides a method for estimating the release source term under emergency conditions of a sodium-cooled fast reactor based on core damage. The method includes the following steps:

[0031] S11. Calculate the core stockpile of nuclides based on the cumulative yield of nuclides.

[0032] According to foreign literature, the core stock is related to the reactor's power operation history, cumulative yield, and the decay of various nuclides. The core stock is calculated using the following formula:

[0033]

[0034] In the formula: A core,i (t) represents the core accumulation of radionuclide i at time t, Bq; p represents the reactor thermal power, W; Y i λ represents the cumulative yield of nuclide i, dimensionless; i Let be the decay constant of nuclide i, s -1 t is the equivalent full-power operating time of the reactor, in seconds.

[0035] The cumulative yield is the cumulative yield under fast neutron conditions. The cumulative yield data for several fission products in Table 1 can be obtained by consulting publicly available data in foreign literature.

[0036] Table 1 Cumulative Production of Fission Products

[0037]

[0038] S12. Calculate the source term of core release to the primary loop sodium pool based on the core accumulation of nuclides, core damage share, and release share of each nuclide.

[0039] After core damage, fuel and fission products may be released into the primary loop sodium pool. The types and amounts of nuclides released depend on the extent of core damage. The source term for the primary loop sodium pool is calculated using the following formula:

[0040] A pool,i (t)=A core,i (t)*N damage,i

[0041] In the formula: A pool,i (t) represents the activity of radionuclide i in the sodium pool at time t, Bq; A core,i (t) represents the core accumulation of radionuclide i at time t, Bq; N damage,i The fraction of radionuclide i released from the reactor core into the primary circuit due to core damage is dimensionless.

[0042] Core damage results in the release fraction N of radionuclide i from the core into the primary circuit. damage,i It is the product of the core damage share and the share of each nuclide released from the core into the sodium pool.

[0043] After the cladding breaks down, all the inert gas is released into the sodium pool of the primary coolant circuit. Simultaneously, some radioactive materials in the fuel are also released. The release percentages of radionuclides from the reactor core to the primary coolant circuit are shown in Table 2. The nuclide release percentage data in Table 2 were obtained by reviewing publicly available foreign literature.

[0044] Table 2. Proportion of each nuclide released from the reactor core to the sodium pool with the inert gas.

[0045] Nuclide Release share Non-volatile fission products, such as La, Zr, Nb, Ba, and Sr. 1％ Pu、U 0.005％ Alkali metals, halogens 25％ Te,Se 2.5％

[0046] S13. Calculate the source term of radionuclides released from the core to the covering gas cavity based on the core accumulation of nuclides, the core damage share, and the release share of nuclides from the core to the covering gas cavity.

[0047] The source term covering the air cavity is calculated using the following formula:

[0048] A cover,i (t)=A core,i (t)*N pool,i (t)

[0049] In the formula: A cover,i (t) represents the activity of radionuclide i in the covered air cavity at time t, Bq; N pool,i (t) represents the fraction of radionuclide i released from the core into the covering gas cavity due to core damage, and is dimensionless.

[0050] Core damage results in the release of radionuclide i from the core into the portion N of the covering gas cavity. pool,i (t) is the product of the core damage share and the share of nuclides released from the core into the covering gas cavity.

[0051] Table 3 shows the proportions of elements released from the reactor core to the covering gas cavity at different temperatures, obtained using chemical equilibrium methods from foreign literature.

[0052] Table 3. Proportion of nuclides released from the reactor core into the covered gas cavity at different temperatures.

[0053]

[0054]

[0055] S14. If the accident type is determined to be loss of plug seal, the airborne radionuclides will be released from the covered gas chamber into the containment. Calculate the rate of change of the source term released into the containment.

[0056] If a transient supercritical event occurs, leading to a complete loss of plug seal, or if the main vessel overpressure venting fails, resulting in a loss of plug airtightness, gaseous radionuclides will be released from the covering gas chamber to the reactor top shield, and subsequently to the containment. The rate of change of the source term released to the containment in this scenario is calculated using the following formula:

[0057]

[0058] In the formula: A contain,i (t) represents the activity of radionuclide i in the containment at time t, Bq; V topshield For the volume of the top protective cover space, m 3 ;v Na S represents the flow velocity of sodium vapor towards the containment, in m / s; topshield The area of ​​the leakage path from the top of the reactor shield to the containment is given in m. 2 ;K D,i is the separation coefficient of radionuclide i as it evaporates with sodium coolant, dimensionless; f0 is the sum of containment ventilation flow rate and natural leakage flow rate, m 3 / s;V contain The volume of the containment hall's atmospheric space is m. 3 ;λ dep For sodium-cooled fast reactors, the deposition rate is mainly determined by two mechanisms: gravity and Brownian diffusion. -1 .

[0059] Step S14 includes the following sub-steps:

[0060] S141. Calculate the flow velocity v of sodium vapor towards the containment. Na ;

[0061] According to experimental results from the Karlsruhe Nuclear Research Center in Germany, if instantaneous supercriticality occurs, the plug seal is completely lost, and sodium vapor is continuously released into the containment. At the typical temperature of the primary loop sodium pool, the flow velocity of sodium vapor towards the containment is 17 m / s. If the overpressure protection of the main vessel fails, causing the plug to lose only its airtightness, the sodium vapor in the covered gas chamber enters the reactor top shield, and the flow velocity of sodium vapor towards the containment is 1.3 m / s. Fission products enter the containment along with the sodium vapor.

[0062] S142. Calculate the separation coefficient K of radionuclides as they evaporate with sodium coolant. D,i ;

[0063] The separation coefficients of different nuclides as they evaporate with sodium coolant are calculated as follows: Separation coefficient K at equilibrium. D The gas-liquid equilibrium coefficients of Cs, I, and Te, as measured experimentally in Japan, are as follows.

[0064] log K D [Cs] = 1940 / T(K) - 0.738

[0065] Where: K D [Cs] is the gas-liquid separation coefficient of Cs, dimensionless; T is the sodium pool temperature in K, dimensionless.

[0066] log K D [Te] = -8832 / T(K) + 5.670

[0067] Where: K D [Te] is the gas-liquid separation coefficient of Te, which is dimensionless.

[0068] log K D [I] = -215 / T(K) - 0.271

[0069] Where: K D [I] is the gas-liquid separation coefficient of I, which is dimensionless.

[0070] Non-equilibrium separation coefficient K' D and K D The relationship is shown in the following formula.

[0071] K' D / K D =D v,i / D v,a

[0072] D v,i Let m be the diffusion coefficient of sodium vapor. 2 / s;D v,a Let m be the diffusion coefficient of fission product i. 2 / s.

[0073] For a mixture of two gases, the diffusion coefficient is calculated using the following formula.

[0074]

[0075] B = [10.85 - 2.50(1 / M1 + 1 / M2)] 1 / 2 ]×10 -4

[0076] I D =κT / ε 12

[0077] In the formula: D is the diffusion coefficient of the mixture of two gases, in cm. 2 / s; T is the gas temperature, K; M1 and M2 are the molar masses of the two components, g / mol; P is the absolute pressure, atm; r 12 For the collision diameter, 10 -10 m; κ is the Boltzmann constant, 1.38 × 10⁻⁶. -23 J / K; ε 12 The value is the intermolecular interaction energy, J.

[0078] S143, Calculate the deposition rate λ dep

[0079] Thermophoresis is caused by temperature differences in the gas, and the deposition rate is proportional to the temperature gradient on the walls. Under emergency conditions, the temperature gradient in the gas space of the demonstration fast reactor containment hall is unavailable, making it difficult to accurately calculate the aerosol removal effect of thermophoresis. The diffusion-trapping phenomenon of aerosols is caused by the condensation of vapors in their carrier gas on the wall surfaces. Since there is no vapor in the demonstration fast reactor containment, diffusion-trapping does not need to be considered.

[0080] Therefore, for sodium-cooled fast reactors, there are mainly two mechanisms: gravity and Brownian diffusion, hence the deposition rate λ dep

[0081] λ dep =λ grav +λ diff

[0082] In the formula: λ grav For gravity deposition rate, s -1 ;λ diff For Brownian diffusion deposition rate, s -1 .

[0083] (1) Gravity deposition rate λ grav Calculate using the following formula.

[0084]

[0085] In the formula: Awall The area of ​​the containment vessel available for gravity settlement, in meters. 2 ;v grav V is the gravitational settling velocity of the aerosol, m / s; V is the containment volume, m³. 3 .

[0086] Gravitational settling velocity v of aerosols grav Calculate using the following formula.

[0087]

[0088] In the formula: d p ρ represents the aerosol particle size in meters (m); p Density of aerosol, kg / m³ 3 g is the acceleration due to gravity, m / s² 2 μ is the dynamic viscosity of air at 298 K, 1.8 × 10⁻⁶. -5 N·s / m 2 ; χ is the dynamic shape factor; C m The Cunningham slip correction factor is calculated using the following formula.

[0089]

[0090] Where: λ is the mean free path of air at 298K, 0.069×10 -6 m;F slip The slip coefficient is set to 1.257.

[0091] (2) Brownian diffusion deposition rate λ diff Based on the velocity v of aerosols due to Brownian diffusion diff The velocity of aerosols due to Brownian diffusion is calculated using the following formula.

[0092]

[0093] In the formula: v diff σ is the velocity of the aerosol due to Brownian diffusion, in m / s; σ is the Boltzmann constant, 1.38 × 10⁻²³ J / (s·m²). 2 ·K 4 T represents the temperature of the containment hall, in K; C represents the temperature of the containment hall. m U is the Cunningham slip correction factor; μ is the dynamic viscosity of air, N·s / m 2 ; χ is the dynamic shape factor; d p Δ is the aerosol particle size, in meters; Δ is the diffusion boundary layer thickness, with a default value of 10. -5 m.

[0094] S15. If the accident type is determined to be a double-walled container meltdown or leakage, radioactive material enters the crater. Calculate the changes in airborne radionuclides in the crater over time.

[0095] If the double-walled container melts through or leaks, sodium coolant carrying fission products will enter the crater. Since air exists within the crater, sodium fire carrying fission products will form radioactive aerosols and release them into the crater's gas space. The following formula can be used to calculate the change in airborne radioactivity in the crater over time:

[0096]

[0097] In the formula: A cavity,i (t) represents the activity of radionuclide i in the crater at time t, in Bq; f1 is the combustion rate of sodium, kg / s; f2 is the pile ventilation flow rate, m³ / s. 3 / s;V cavity Let V be the volume of the free space of the heap.

[0098] (1) Burning rate of sodium The calculation is performed using the following method:

[0099] According to the results of the large-scale (20.4m high, 7.62m diameter) pool-type sodium fire experiment CSTF at the Humphrey Laboratory in the United States, 25.5% of sodium forms an aerosol. The combustion rate was calculated using the model of the French sodium-cooled fast reactor severe accident analysis program ASTEC-Na:

[0100]

[0101] In the formula:

[0102] The combustion rate is expressed in kg / s; H2 g ρ is the gas mass transfer coefficient, m / s; g The density of the gas is kg / m³. 3 C O2 Oxygen concentration; S pool Let m be the area of ​​the sodium pool, and m be the area of ​​the crater. 2 .

[0103] In emergency situations, if parameters such as oxygen concentration cannot be obtained in real time, based on experimental research results from the Humphrey Laboratory in the United States, assuming that the oxygen concentration and flame temperature remain constant, the average combustion rate of sodium during a pool fire is 38 kg / (h·m). 2 ).

[0104] (2) Calculate the combustion separation coefficient K C

[0105] Define the combustion separation coefficient K C as follows:

[0106]

[0107] In the formula: The release rate of nuclide i is expressed in kg / s; m i,pool The remaining mass of nuclide i in the sodium pool is in kg. The combustion rate of sodium is expressed in kg / s; m Na,pool The mass of sodium in the sodium pool, including metallic sodium and sodium oxide, is expressed in kg.

[0108] According to the FANAL experimental results published by the French IRSN, the combustion separation coefficient K of different nuclides or elements C See Table 4. Due to the combustion separation coefficient K C The combustion-entrained nuclides can be calculated based on the combustion separation coefficient, as obtained from Table 4.

[0109] Table 4 Combustion Separation Coefficient K C

[0110] Serial Number nuclide or element name Combustion Separation Coefficient 1 Na-22 1.02±0.2 2 Cs 5.2±0.5 3 I (NaI form) 1.6±0.1 4 Ag-110m 1.30±0.4 5 Sr <![CDATA[3.4×10 -3 ]]> 6 Zr <![CDATA[<2×10 -3 ]]> 7 Ru <![CDATA[5.1×10 -2 ]]> 8 Sn <![CDATA[<6×10 -3 ]]> 9 Sb 0.17 10 Te <![CDATA[1.2×10 -2 ]]> 11 Ce <![CDATA[1.3×10 -3 ]]>

[0111] S16. Calculate the source terms released to the environment based on the rate of change of the source terms released to the containment and the rate of change of the airborne source terms in the crater.

[0112] The source items released to the environment are calculated using the following formula:

[0113]

[0114] Where: N filter-con,i The fraction of nuclide i passing through the containment ventilation system filter, dimensionless; N filter-cav,i The fraction of nuclide i passing through the reactor pit ventilation system filter is dimensionless.

[0115] As can be seen from the above embodiments, the method for estimating the release source term in emergency operating conditions of a sodium-cooled fast reactor based on core damage disclosed in this invention can calculate the dynamic and phased release source term to the primary circuit, containment and environment under different release pathways based on facility operation data, core accumulation and core damage share when a sodium-cooled fast reactor accident occurs. This provides a reliable basis for decision-making to assess the impact of the release source term on the surrounding environment and what protective measures should be taken in emergency situations.

[0116] The method described in this invention is not limited to the embodiments described in the specific implementation. Other implementation methods derived by those skilled in the art based on the technical solution of this invention also fall within the scope of technical innovation of this invention.

Claims

1. A method for estimating the release source term under emergency conditions in a sodium-cooled fast reactor based on core damage, the method comprising the following steps: S11. Calculate the core stockpile of nuclides based on the cumulative yield of nuclides; S12. Calculate the source term of the core released into the primary loop sodium pool based on the core accumulation of nuclides, the core damage share, and the release share of each nuclide. S13. Calculate the source term of radionuclide release from the core to the covering gas cavity based on the core accumulation of nuclides, the core damage share, and the release share of nuclides from the core to the covering gas cavity. S14. If the accident type is determined to be loss of plug seal, the airborne radionuclides will be released from the covered air chamber into the containment. Calculate the rate of change of the source term released into the containment. S15. If the accident type is determined to be a double-walled container meltdown or leakage, radioactive material enters the crater. Calculate the rate of change of the gaseous source term in the crater. S16. Calculate the source terms released to the environment based on the rate of change of the source terms released to the containment and the rate of change of the airborne source terms in the crater; Step S14 includes the following sub-steps: S141. Calculate the flow velocity of sodium vapor towards the containment. S142. Calculate the separation coefficient of radionuclides as sodium coolant evaporates; S143. Calculate the aerosol deposition rate; Calculate the rate of change of the source term released into the containment structure based on the above parameters; Step S15 includes the following sub-steps: S151. Calculate the combustion rate of sodium; S152. Calculate the combustion separation coefficients of different nuclides; S143. Calculate the aerosol deposition rate; The rate of change of the gaseous source term in the crater is calculated based on the above parameters.

2. The method for estimating the emergency release source term of a sodium-cooled fast reactor based on core damage as described in claim 1, characterized in that: The cumulative yield of nuclides in step S11 is the cumulative yield of nuclides under fast neutrons, which is obtained by consulting publicly available data from foreign literature.

3. The method for estimating the emergency release source term of a sodium-cooled fast reactor based on core damage as described in claim 1, characterized in that: The release proportions of each nuclide in step S12 were obtained by consulting publicly available data from foreign literature.

4. The method for estimating the emergency release source term of a sodium-cooled fast reactor based on core damage as described in claim 1, characterized in that: In step S13, the proportion of elements released from the reactor core to the cover gas at different temperatures is obtained based on the chemical equilibrium method disclosed in foreign literature.

5. The method for estimating the emergency release source term of a sodium-cooled fast reactor based on core damage as described in claim 1, characterized in that: When some input parameters were unavailable, the combustion rate of sodium was determined using experimental results from the Humphrey Laboratory in the United States.

6. The method for estimating the emergency release source term of a sodium-cooled fast reactor based on core damage as described in claim 1, characterized in that: The combustion separation coefficients of different nuclides were obtained by consulting publicly available experimental data from abroad.

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