A method for applying ROP setting value in advance for the transition core of an improved fuel heavy water reactor

Abstract

The present invention belongs to the field of reactor physics technology for heavy water reactor nuclear power plants, and specifically relates to a method for preemptively applying the ROP set value for a heavy water reactor transition core with improved fuel. The method comprises the following steps: Step 1: Calculating the X% increase in the ROP set value for the full core with both the original design fuel and the improved fuel; Step 2: Calculating the D EFPDs required to fully fill the high power region of the core with the improved fuel and the corresponding EFPD at the time point; Step 3: Calculating the conservative increase in the ROP set value by Y% compared to the original design fuel core after the fuel channels in the high power region are fully filled with the improved fuel; Step 4: ROP set value evaluation and calculation of the minimum increase by Y% at full power. The present invention has the beneficial effect of increasing the ROP margin earlier, thereby generating more power generation revenue for the nuclear power plant, and reducing the power reduction during refueling earlier, thereby enhancing the operational safety of the nuclear power plant.

Description

Technical Field

[0001] The invention belongs to the technical field of reactor physics of heavy water reactor nuclear power plants, and particularly relates to a method for applying a ROP set value in advance to an improved fuel transition core of a heavy water reactor. Background Art

[0002] ROP, or Regional Overpower Protection, is one of the trip parameters for the heavy water reactor (HWR) shutdown system. After long-term operation, HWR units may be forced to reduce their total power due to: 1) expansion of the pressure tube diameter after long-term neutron irradiation; 2) reduced heat transfer efficiency caused by scaling of steam generator heat transfer tubes; and 3) FAC (Fast Action Chain) phenomena such as those in the main loop piping. This reduces the total flow rate, the critical channel power (CCP) of the primary heat system, and the operating margin for Regional Overpower Protection (ROP).

[0003] To address the aforementioned aging phenomenon, one approach to increasing the ROP margin is to replace the entire core with improved fuel, improve the thermal-hydraulic conditions, and increase the critical channel power (CCP), thereby raising the ROP shutdown setting. However, due to the numerous channels in the core, replacing the original fuel with the improved fuel takes a relatively long time according to the normal refueling cycle (for example, replacing 37R fuel with 37M fuel in a CANDU-6 heavy water reactor takes approximately 700 days), thus delaying the application of the increased ROP margin for the improved fuel core. The current practice is to only apply the increased ROP margin for the improved fuel after the entire core has been replaced with the improved fuel, resulting in significant economic losses for the power plant (loss value = electricity price * delayed application time * increased ROP margin * unit rated power). Summary of the Invention

[0004] The purpose of the present invention is to provide a method for applying the ROP set value in advance in the fuel transition core of an improved heavy water reactor. This method is applicable to multi-channel heavy water reactors. Compared with the original method in which the new ROP set value can be applied only after the entire core is replaced, this method can improve the ROP margin earlier, thereby bringing more power generation revenue to the nuclear power plant, and can also reduce the power reduction during the fuel change earlier, thereby enhancing the operational safety of the nuclear power plant.

[0005] The technical solution of the present invention is as follows: a method for applying the ROP set value in advance in a heavy water reactor improved fuel transition core, comprising the following steps:

[0006] Step 1: Calculate the increase in ROP setpoint by X% for the full core of the original design fuel and the improved fuel;

[0007] Step 2: Calculate the time D EFPD required to fill the high power area of ​​the core with improved fuel and the corresponding EFPD at the time point;

[0008] Step 3: Calculate the conservative increase in the ROP setting value by Y% compared to the original design fuel core after the fuel channels in the high power area are filled with improved fuel.

[0009] In step 1, a complete and systematic ROP setting value analysis is performed for both cases where the core is fully loaded with the original design fuel and fully loaded with the improved fuel, confirming that the ROP setting value of the core with the improved fuel is X% higher than that of the core with the original design fuel. The X% is used in a subsequent assessment to determine whether it is necessary to use a transition core to apply the ROP setting value in advance.

[0010] In step 2, based on the normal refueling cycle of the heavy water reactor core, the time D EFPD required for the core high power area to be filled with the improved fuel from the original design fuel and the corresponding time point EFPD are calculated. The time point is (initial day of core replacement + D) EFPD.

[0011] In step 2, the reactor is composed of multiple fuel channels. The fuel channels in the core are refueled according to a specific refueling method and refueling cycle. Refueling the entire core from the original design fuel to the improved fuel requires a longer period of time. Because the central core region has higher power and a shorter refueling cycle, the high-power region of the core will be replaced with the improved fuel sooner than the entire core.

[0012] The step 3 includes:

[0013] Step 31: Core physical analysis;

[0014] Step 32: Thermal hydraulic analysis;

[0015] Step 33: Uncertainty assessment and setting;

[0016] Step 34: ROP setting value evaluation and minimum boost Y% full power calculation.

[0017] Step 31 includes performing a simulated refueling during the transition from the original design fuel core to the improved fuel core, obtaining the time required for the high power area to be fully loaded with the improved fuel and the fuel mixing in the core from the results of the simulated refueling calculation, and evaluating the flux shape adopted. Usually, at this time point, the overall percentage of improved fuel bundles in the core is 95%.

[0018] The step 32 includes the main system thermal hydraulic boundary conditions, thermal modeling and critical channel power CCP calculation, including:

[0019] A. Consider establishing N conservative thermal-hydraulic boundary conditions

[0020] B. Based on the N assumed thermal and hydraulic boundary conditions, establish M CCP thermal models and calculate the CCP under N×M conditions;

[0021] The step 33 includes evaluating the uncertainty difference involved in the ROP calculation when the heavy water reactor core is loaded with the original design fuel and the improved fuel, and increasing the uncertainty when setting the core uncertainty in the "high power region" if the difference is large.

[0022] Step 34 includes calculating and comparing the ROP set values ​​of the complete calculation method for the original design fuel and the improved fuel core at the EFPD time point (the initial day of core replacement + D); calculating the ROP set values ​​for the "high power region" core using a combination of thermal-hydraulic conditions and uncertainties at the EFPD time point (the initial day of core replacement + D), and arranging them into a list; and using the minimum increase in the list, Y% full power, as the increase in the set value under the complete ROP calculation for the "high power region" core compared to the original design fuel core.

[0023] The beneficial effects of the present invention are as follows: during the transition period of the core of a heavy water reactor in which improved fuel is replaced, compared with the original method of applying the new ROP setting value only after the entire core is replaced with the improved fuel, the method of the present invention of applying the ROP setting value in advance during the transition period from the original design fuel to the improved fuel core can improve the ROP margin earlier, thereby bringing more power generation benefits to the nuclear power plant, and reducing the power reduction during the fuel replacement earlier, thereby enhancing the operational safety of the nuclear power plant.

[0024] For example, using the originally designed ROP setting analysis procedures, methods, and processes, during the transition period of a power plant from 37R fuel to 37M fuel core, a conservative assumption was used to calculate that after the fuel channels in the CPPF area were filled with 37M fuel bundles, the ROP setting value could be increased by at least 2.66% of full power based on the ROP setting value calculation results for the 37R core. BRIEF DESCRIPTION OF THE DRAWINGS

[0025] Figure 1 Apply the ROP setpoint assessment process in advance for transition cores;

[0026] Figure 2 Schematic diagram of the CPPF area. DETAILED DESCRIPTION

[0027] The present invention will be further described in detail below with reference to the accompanying drawings and specific embodiments.

[0028] The core of a heavy water reactor is composed of numerous fuel channels, which are refueled according to a specific refueling pattern and cycle. Refueling the entire core from the original design fuel to the improved fuel takes a long time. Because the central core has higher power and a shorter refueling cycle, the high-power region of the core is replaced with the improved fuel sooner than the entire core.

[0029] The main steps in calculating the ROP setpoint for a heavy water reactor include core physics calculations, thermal calculations, uncertainty analysis, and setpoint calculations. During the transition from the original fuel design to the improved fuel design, the ROP setpoint will gradually transition from the ROP setpoint for the original fuel design to the ROP setpoint for the improved fuel design. Since nearly all ROP detectors are located in the core's high-power region, and the channels that burn out first under various operating conditions are almost always located in this region, a specific core high-power region is selected as the analysis target for the transition core. This core state is hereafter referred to as the "improved fuel-high-power region" core.

[0030] For example, the CANDU6 reactor has 380 fuel channels, each with 8 fuel bundles. Under normal circumstances, the 8-bundle refueling method is adopted. Each fuel channel needs to undergo 8-bundle refueling twice, and all 12 original fuel bundles in the channel are replaced with 37M fuel. Figure 2 The numbers in the channels in the figure are the number of times each channel uses 37M fuel bundles for refueling.

[0031] The main steps in calculating the ROP setpoint for the CANDU6 reactor include core physics calculations, thermal calculations, uncertainty analysis, and setpoint calculations. During the transition from the original design reference fuel (37R fuel) core to the 37M fuel core, the ROP setpoint will gradually transition from the 37R ROP setpoint to the 37M ROP setpoint. Since nearly all ROP detectors are located in the CPPF region, and the channels that burn dry first under various operating conditions are almost always located in the high-power region of the core, the CPPF region was selected as the analysis target for the transition core. This core state is hereafter referred to as the "37M-CPPF" core.

[0032] A method for applying a ROP setting value in advance to an improved fuel transition core of a heavy water reactor comprises the following steps:

[0033] Step 1: Calculate the increase in ROP set point value by X% for the original design fuel and the improved fuel in the full core.

[0034] A comprehensive ROP setting analysis was conducted for both cores loaded entirely with the original design fuel and cores loaded entirely with the improved fuel. It was confirmed that the ROP setting for the improved fuel core was approximately X% higher than that for the original design fuel core. This X% was used in subsequent assessments to determine whether it was necessary to preemptively apply the ROP setting using a transition core.

[0035] Step 2: Calculate the time D EFPD required to fill the high power area of ​​the core with improved fuel and the corresponding time point (initial day of core replacement + D) EFPD.

[0036] Based on the normal refueling cycle of the heavy water reactor core, the time D EFPD required for the high-power area of ​​the core to be filled with the improved fuel from the original design fuel is calculated, as well as the corresponding time point (the initial day of core replacement + D) EFPD.

[0037] For example, a CANDU6 reactor has 380 fuel channels, each containing eight fuel bundles. Using the normal eight-bundle refueling method, each channel requires two eight-bundle refuelings, until all 12 fuel bundles within the channel are replaced with 37M fuel. The refueling cycle for each fuel channel ranges from six months to one year. Filling all 380 fuel channels with 37M fuel requires nearly 700 EFPDs, or approximately 23 months. The high-power region (CPPF) of the core has a shorter refueling cycle, requiring approximately (D EFPDs) a year to fill the CPPF with 37M fuel. Calculations show that a full CPPF load of 37M fuel requires approximately (the initial day of core replacement + D) EFPDs.

[0038] Step 3: Using a conservative analysis method, calculate the conservative increase in the ROP setting value by Y% compared to the original design fuel core after the fuel channels in the high-power area are filled with improved fuel.

[0039] A conservative analysis method is used to calculate the conservative increase in the ROP setting by Y% compared to the original design fuel core when the fuel channels in the high power region of the heavy water reactor are filled with improved fuel. The main steps include the following:

[0040] Step 31: Core Physics Analysis

[0041] A simulated refueling is performed during the transition from the original design fuel core to the improved fuel core. The results of the simulated refueling calculations reveal the time required to fully load the high-power region with the improved fuel, the fuel mix in the core, and the appropriate flux shape to be used. Typically, at this point in time, the percentage of improved fuel bundles in the core is around 95%. Because the core physics of the original design and improved fuels are similar, the same flux shapes can be used.

[0042] For example, the CANDU-6 heavy water reactor simulated refueling during the transition from a 37R core to a 37M core. The results of the simulated refueling calculations revealed the time required to fully load the CPPF area with 37M fuel and the fuel mixing in the core, and the flux shape used was evaluated.

[0043] Simulations show that after approximately D FPDs (approximately one year), all fuel channels in the CPPF region have undergone two refueling cycles, with each channel containing all 12 fuel bundles filled with 37M fuel. Transition core simulations indicate that after approximately one year, half of the channels in the CPPF region (approximately 60% of all channels) and half of the channels in the non-CPPF region (40% of all channels) are filled with 37M fuel, meaning that 80% of the channels in the core are fully loaded with 37M fuel bundles. The remaining 20% ​​of the channels are mixed fuel, consisting of four 37R and eight 37M fuel bundles, all located in the non-CPPF region. The overall percentage of 37M fuel bundles in the core is close to 95%. Due to the high physical similarity between the 37R and 37M cores, the same flux shapes are used.

[0044] Step 32: Thermal-Hydraulic Analysis

[0045] The thermal-hydraulic analysis of a heavy water reactor (HWR) includes the thermal-hydraulic boundary conditions of the main system, thermal modeling, and critical channel power (CCP) calculation. The main steps include: A. Establishing N conservative thermal-hydraulic boundary conditions; B. Establishing M CCP thermal models based on these N conservative thermal-hydraulic boundary conditions and calculating the CCP under N×M conditions.

[0046] For example, for the CANDU-6 heavy water reactor, thermal-hydraulic analysis includes the thermal-hydraulic boundary conditions of the main system, thermal modeling, and critical channel power (CCP) calculation.

[0047] Thermal-hydraulic boundary conditions of the main system

[0048] For the "37M-CPPF" reactor core, the following three conservative NUCIRC thermal-hydraulic models are established:

[0049] Option 1:

[0050] Both the CPPF and non-CPPF areas of the core are loaded with 37M fuel. Because channels loaded with 37M fuel have slightly lower pressure drops and slightly higher flow rates than channels loaded with 37R fuel, it is conservative to assume that 37M fuel is loaded in the non-CPPF areas, allowing for a lower flow rate in the CPPF fuel channels.

[0051] The inter-header pressure difference is based on the data of the 37M core at 80% FP power. The inter-header pressure difference of the 37M core is small, and the inter-header pressure difference used in the 37M core model is conservative;

[0052] Normalize the total coolant flow to the total flow of the 37R core. The total flow of the 37M core is greater than that of the 37R core, so this is conservative.

[0053] Option 2:

[0054] The non-CPPF area of ​​the core is loaded with 37R fuel; the rest is the same as option 1.

[0055] Option 3:

[0056] The pressure drop between headers is based on the 37R core data; the rest is the same as Option 1.

[0057] Critical Channel Power CCP Calculation

[0058] Conservatively assuming that all channels in the non-CPPF area are loaded with 37R fuel, the NUCIRC program ITYPE2 is used to establish the CCP thermal model for "all CPPF areas are loaded with 37M fuel, and all non-CPPF areas are loaded with 37R fuel" for the three thermal-hydraulic boundary conditions mentioned above (Option 1, Option 2, Option 3). The critical channel power (CPP) corresponding to the three main system boundary conditions (Option 1, Option 2, Option 3) is calculated as the input for the ROVER-F program setting value calculation.

[0059] Step 33: Uncertainty Assessment and Setting

[0060] The uncertainty differences involved in ROP calculations for heavy water reactor cores loaded with original design fuel and improved fuel are evaluated. For those with large differences, the uncertainty is conservatively increased when setting the core uncertainty in the "high power region", while for those with small deviations, the uncertainty is kept unchanged.

[0061] For example, for the "37M-CPPF" core state of the CANDU-6 reactor type, due to differences in channel flow distribution, pressure tube creep rate, and core fuel mixture fuel string configuration, for the sake of conservatism, on the basis of the uncertainty of the complete 37M core, the uncertainty of the inter-manifold pressure drop, integrated two-phase flow, and CHF is increased by 50%, and the uncertainty of the pressure tube creep is increased by 100%.

[0062] The inlet header temperature is related to steam generator performance. Steam generator performance does not differ significantly between cores loaded with 37M and 37R fuel, so the uncertainty in the inlet header temperature remains unchanged. The outlet pressure is controllable, so its uncertainty also remains unchanged. The uncertainty in the core flow rate is based on ultrasonic flow measurement. For models with a 37M-CPPF core and a full core of 37M or 37R fuel, the uncertainty in the total core flow rate remains unchanged.

[0063] Step 34: ROP setting value evaluation and minimum boost Y% full power calculation

[0064] For the heavy water reactor core, the ROP setting values ​​of the complete calculation method for the original design fuel and the improved fuel core at the EFPD time point (the initial day of core replacement + D) are calculated and compared; the ROP setting values ​​for the "high power region" core at the EFPD time point (the initial day of core replacement + D) are calculated using the above (2) conservatively assumed thermal-hydraulic conditions and (3) uncertainties, and compiled into a table; the minimum increase in the list, Y% full power, is conservatively used as the increase in the setting value under the complete ROP calculation for the "high power region" core compared to the original design fuel core.

[0065] For example: In the ROP calculation of the transition core of the CANDU-6 reactor type in the 37M fuel improvement change project of the Qin-3 Power Plant, the ROP set values ​​of the intact 37R and 37M cores at the time point of 6637EFPD (the initial day of core replacement + D) are calculated and compared; the ROP set values ​​of the "37M-CPPF" core at the time point of 6637EFPD (the initial day of core replacement + D) are calculated under some conservatively assumed thermal-hydraulic conditions and uncertainties, and compiled into a list; the minimum increase of 2.66% in the list is conservatively adopted as the increase in the set value of the "37M-CPPF" core compared to the intact 37R core at full power.

[0066] First, the set values ​​for the complete 37R and 37M cores of the 6637EFPD were calculated, and the results are shown in Table A.

[0067] Table A 6637EFPD Complete 37R Core and 37M Core ROP Setting Values

[0068]

[0069]

[0070] For three conservatively assumed thermal-hydraulic boundary conditions (Option 1, Option 2, Option 3), the nominal uncertainty and the uncertainty of "inter-manifold pressure drop, integrated two-phase flow and CHF, and pressure tube creep" were added, for a total of six combinations. The ROP trip setting values ​​of the 6637EFPD "37M-CPPF" core were calculated using the ROVER-F program, see Table B.

[0071] Table B 6637EFPD "37M-CPPF" core ROP setting values ​​(three boundary conditions, nominal uncertainty and increased uncertainty)

[0072]

[0073] The increase in the set values ​​of the 6637EFPD "37M-CPPF" core compared to the complete 37R core is shown in Table C. The minimum increase is 2.66%, corresponding to the selection of 1 for the main system thermal-hydraulic boundary condition, which increases uncertainty.

[0074] Table C 6637EFPD "37M-CPPF" core compared to 37R core setting value improvement

[0075]

[0076] The main approach to be protected is the conservative analysis and early application of ROP setting values ​​described above, which calculates the conservative increase in ROP setting value by Y% compared to the original design fuel core after the fuel channels in the high-power area are filled with improved fuel.

[0077] Example 1: A 37M fuel improvement project for a CANDU-6 reactor of a power plant was carried out to apply a 2.66% full power ROP set point increase in advance after the transition core was fully loaded with 37M fuel in the CPPF high power region.

[0078] The CANDU6 reactor has a total of 380 fuel channels, each containing 8 fuel bundles. Normally, the 8-bundle refueling method is used. Each fuel channel needs to undergo two 8-bundle refuelings, and all 12 original fuel bundles in the channel are replaced with 37M fuel. Figure 2 The numbers in the channels are the number of times each channel uses 37M fuel bundles for refueling.

[0079] The main steps in calculating the ROP setpoint for the CANDU6 reactor include core physics calculations, thermal calculations, uncertainty analysis, and setpoint calculation. During the transition from a 37R fuel core to a 37M fuel core, the ROP setpoint will gradually transition from the 37R ROP setpoint to the 37M ROP setpoint. Since nearly all ROP detectors are located in the CPPF region, and the channels that burn out first under various operating conditions are almost always located in the core's high-power region, the CPPF region was selected for analysis during the transition core. This core state is hereafter referred to as the "37M-CPPF" core.

[0080] 1. A comprehensive analysis of the ROP setting for both reactor cores loaded entirely with the original design fuel and those loaded entirely with the improved fuel confirmed that the ROP setting for the improved fuel core is approximately 5% FP higher than that for the original design fuel core. This 5% FP will be used to assess whether it is necessary to preemptively apply the ROP setting using a transition core.

[0081] 2. Calculate the time required to fully load the core's high-power region with improved fuel, D EFPDs, and the corresponding time point (the initial day of core replacement + D) EFPDs. The CANDU6 reactor has 380 fuel channels, each containing eight fuel bundles. Using the normal 8-bundle refueling method, each channel requires two 8-bundle refuelings, until all 12 fuel bundles in the channel are replaced with 37M fuel. The refueling cycle for each fuel channel ranges from six months to one year. Filling all 380 fuel channels with 37M fuel requires nearly 700 EFPDs, or approximately 23 months. The core's high-power region (CPPF) has a shorter refueling cycle, requiring approximately (D EFPDs) one year to fully load the CPPF with 37M fuel. The calculated time for a full CPPF with 37M fuel is approximately (the initial day of core replacement + D) EFPDs, or 6637 EFPDs in this example.

[0082] 3. Using a conservative analysis method, calculate the conservative increase in ROP setpoint by Y% compared to the original fuel core design when the fuel channels in the high-power region are filled with the improved fuel. The main steps include the following:

[0083] 3.1 Core physics analysis

[0084] A simulated refueling was carried out during the transition from the original design fuel core to the improved fuel core. The time required to fully load the high power area with the improved fuel and the fuel mixing in the core were obtained from the results of the simulated refueling calculation, and the flux shape adopted was evaluated.

[0085] A simulated refueling of the CANDU-6 heavy water reactor during the transition from a 37R core to a 37M core was performed. The time required to fully load the CPPF region with 37M fuel and the fuel mixing in the core were calculated from the simulated refueling calculation results, and the flux shape adopted was evaluated.

[0086] Simulations show that after approximately D FPDs (approximately one year), all fuel channels in the CPPF region have undergone two refueling cycles, with each channel containing all 12 fuel bundles filled with 37M fuel. Transition core simulations indicate that after approximately one year, half of the channels in the CPPF region (approximately 60% of all channels) and half of the channels in the non-CPPF region (40% of all channels) are filled with 37M fuel, meaning that 80% of the channels in the core are fully loaded with 37M fuel bundles. The remaining 20% ​​of the channels are mixed fuel, consisting of four 37R and eight 37M fuel bundles, all located in the non-CPPF region. The overall percentage of 37M fuel bundles in the core is close to 95%. Due to the high physical similarity between the 37R and 37M cores, the same flux shapes are used.

[0087] 3.2 Thermal-hydraulic analysis

[0088] The thermal-hydraulic analysis of a heavy water reactor (HWR) includes the thermal-hydraulic boundary conditions of the main system, thermal modeling, and critical channel power (CCP) calculation. The main steps include: A. Establishing N conservative thermal-hydraulic boundary conditions; B. Establishing M CCP thermal models based on these N conservative thermal-hydraulic boundary conditions and calculating the CCP under N×M conditions.

[0089] For example, for the CANDU-6 heavy water reactor, thermal-hydraulic analysis includes the thermal-hydraulic boundary conditions of the main system, thermal modeling, and critical channel power (CCP) calculation.

[0090] A. Thermal-hydraulic boundary conditions of the main system

[0091] For the "37M-CPPF" reactor core, the following three conservative NUCIRC thermal-hydraulic models are established:

[0092] Option 1:

[0093] Both the CPPF and non-CPPF areas of the core are loaded with 37M fuel. Because channels loaded with 37M fuel have slightly lower pressure drops and slightly higher flow rates than channels loaded with 37R fuel, it is conservative to assume that 37M fuel is loaded in the non-CPPF areas, allowing for a lower flow rate in the CPPF fuel channels.

[0094] The inter-header pressure difference is based on the data of the 37M core at 80% FP power. The inter-header pressure difference of the 37M core is small, and the inter-header pressure difference used in the 37M core model is conservative;

[0095] Normalize the total coolant flow to the total flow of the 37R core. The total flow of the 37M core is greater than that of the 37R core, so this is conservative.

[0096] Option 2:

[0097] The non-CPPF area of ​​the core is loaded with 37R fuel; the rest is the same as option 1.

[0098] Option 3:

[0099] The pressure drop between headers is based on the 37R core data; the rest is the same as Option 1.

[0100] B. Critical Channel Power CCP Calculation

[0101] Conservatively assuming that all channels in the non-CPPF area are loaded with 37R fuel, the NUCIRC program ITYPE2 is used to establish the CCP thermal model for "all CPPF areas are loaded with 37M fuel, and all non-CPPF areas are loaded with 37R fuel" for the three thermal-hydraulic boundary conditions mentioned above (Option 1, Option 2, Option 3). The critical channel power (CPP) corresponding to the three main system boundary conditions (Option 1, Option 2, Option 3) is calculated as the input for the ROVER-F program setting value calculation.

[0102] 3.3 Uncertainty assessment and setting

[0103] The uncertainty differences involved in ROP calculations for heavy water reactor cores loaded with original design fuel and improved fuel are evaluated. For those with large differences, the uncertainty is conservatively increased when setting the core uncertainty in the "high power region", while for those with small deviations, the uncertainty is kept unchanged.

[0104] For the CANDU-6 core state of the "37M-CPPF" reactor, due to differences in channel flow distribution, pressure tube creep rate, and core fuel mixture and fuel string configuration, for the sake of conservatism, the uncertainties of the inter-manifold pressure drop, integrated two-phase flow, and CHF are increased by 50%, and the uncertainty of the pressure tube creep is increased by 100% on the basis of the uncertainties of the complete 37M core.

[0105] The inlet header temperature is related to steam generator performance. Steam generator performance does not differ significantly between cores loaded with 37M and 37R fuel, so the uncertainty in the inlet header temperature remains unchanged. The outlet pressure is controllable, so its uncertainty also remains unchanged. The uncertainty in the core flow rate is based on ultrasonic flow measurement. For models with a 37M-CPPF core and a full core of 37M or 37R fuel, the uncertainty in the total core flow rate remains unchanged.

[0106] 3.4 ROP setting value evaluation and calculation of minimum increase Y% full power

[0107] For the heavy water reactor core, the ROP setting values ​​of the complete calculation method for the original design fuel and the improved fuel core at the EFPD time point (the initial day of core replacement + D) are calculated and compared; the ROP setting values ​​for the "high power region" core at the EFPD time point (the initial day of core replacement + D) are calculated using the above (2) conservatively assumed thermal-hydraulic conditions and (3) uncertainties, and compiled into a table; the minimum increase in the list, Y% full power, is conservatively used as the increase in the setting value under the complete ROP calculation for the "high power region" core compared to the original design fuel core.

[0108] In the ROP calculation of the transition core of the CANDU-6 reactor type in the 37M fuel improvement change project of the Qin-3 Power Plant, the ROP set values ​​of the intact 37R and 37M cores at the time point of 6637EFPD (the initial day of core replacement + D) are calculated and compared; the ROP set values ​​of the "37M-CPPF" core at the time point of 6637EFPD (the initial day of core replacement + D) are calculated under some conservatively assumed thermal-hydraulic conditions and uncertainties, and compiled into a list; the minimum increase of 2.66% in the list is conservatively adopted as the increase in the set value of the "37M-CPPF" core compared to the intact 37R core at full power.

[0109] First, the set values ​​for the complete 37R and 37M cores of the 6637EFPD were calculated, and the results are shown in Table A.

[0110]

[0111] Table A 6637EFPD Complete 37R Core and 37M Core ROP Setting Values

[0112] For three conservatively assumed thermal-hydraulic boundary conditions (Option 1, Option 2, Option 3), the nominal uncertainty and the uncertainty of "inter-manifold pressure drop, integrated two-phase flow and CHF, and pressure tube creep" were added, for a total of six combinations. The ROP trip setting values ​​of the 6637EFPD "37M-CPPF" core were calculated using the ROVER-F program, see Table B.

[0113]

[0114] Table B 6637EFPD "37M-CPPF" core ROP setting values ​​(three boundary conditions, nominal uncertainty and increased uncertainty)

[0115] The increase in the set values ​​of the 6637EFPD "37M-CPPF" core compared to the complete 37R core is shown in Table C. The minimum increase is 2.66%, corresponding to the selection of 1 for the main system thermal-hydraulic boundary condition, which increases uncertainty.

[0116]

[0117] Table C 6637EFPD "37M-CPPF" core compared to 37R core setting value improvement

[0118] Therefore, using the previously designed ROP setting analysis procedures, methods, and processes, and using conservative assumptions for the transition period from 37R fuel to 37M fuel cores at Qin No. 3 Power Plant, we calculated that, after the CPPF fuel channels were fully loaded with 37M fuel bundles, the ROP setting could be increased by at least 2.66% of full power, based on the calculated ROP setting for the 37R core. This allowed the 2.66% ROP setting increase to be applied in advance, after the CPPF fuel channels were fully loaded with 37M fuel bundles.

[0119] Finally, it should be noted that the above embodiments are only used to illustrate the technical solutions of the present invention, rather than to limit it. Although the present invention has been described in detail with reference to the aforementioned embodiments, those skilled in the art should understand that modifications to the technical solutions described in the aforementioned embodiments, or equivalent replacement of some or all of the technical features therein, do not deviate the essence of the corresponding technical solutions from the scope of the technical solutions of the embodiments of the present invention.

Claims

1. A method for applying ROP setting value in advance in a heavy water reactor improved fuel transition core, characterized in that: The steps include: Step 1: Calculate the increase in ROP set point value by X% for the full core with the original design fuel and the improved fuel; In step 1, an ROP setting value analysis is performed for both cases where the core is fully loaded with the original design fuel and fully loaded with the improved fuel, confirming that the ROP setting value of the core with the improved fuel is X% higher than that of the core with the original design fuel. The X% is used in a subsequent assessment to determine whether it is necessary to use a transition core to pre-apply the ROP setting value. Step 2: Calculate the time D EFPD required to fill the high power area of ​​the core with improved fuel and the corresponding EFPD at the time point; Step 2 calculates the time D EFPDs required for the core high power area to be fully loaded with the improved fuel from the original design fuel, and the corresponding time point EFPDs based on the normal refueling cycle of the heavy water reactor core. The time point is: the initial day of core replacement + D EFPDs. In step 2, the reactor is composed of a plurality of fuel channels. The fuel channels in the core are refueled according to a certain designed refueling method and refueling cycle. It takes a long time to replace the original design fuel of the entire core with the improved fuel. Since the power of the middle core region is higher and the refueling cycle is shorter, the high-power region of the core will be replaced with the improved fuel earlier than the entire core. Step 3: Calculate the conservative increase in the ROP setting value by Y% compared to the original design fuel core after the fuel channels in the high power area are filled with improved fuel.

2. The method for applying the ROP setting value in advance to the transition core of an improved fuel of a heavy water reactor according to claim 1, characterized in that: The step 3 includes: Step 31: Core physical analysis; Step 32: Thermal hydraulic analysis; Step 33: Uncertainty assessment and setting; Step 34: ROP setting value evaluation and minimum boost Y% full power calculation.

3. The method for applying the ROP setting value in advance for the transition core of an improved fuel of a heavy water reactor according to claim 2, characterized in that: Step 31 includes performing a simulated refueling during the transition from the original design fuel core to the improved fuel core. The time required for the high power area to be fully loaded with the improved fuel and the fuel mixing in the core are calculated from the results of the simulated refueling, and the flux shape adopted is evaluated. Usually, at this time point, the overall percentage of improved fuel bundles in the core is 95%.

4. The method for applying the ROP set value in advance to the transition core of an improved fuel of a heavy water reactor according to claim 2, characterized in that: The step 32 includes the main system thermal hydraulic boundary conditions, thermal modeling and critical channel power CCP calculation, including: A. Consider establishing N conservative thermal-hydraulic boundary conditions B. Based on the N assumed thermal and hydraulic boundary conditions, M CCP thermal models are established, and the CCPs under N×M conditions are calculated.

5. The method for applying the ROP setting value in advance for the transition core of an improved fuel of a heavy water reactor according to claim 2, characterized in that: Step 33 includes evaluating the uncertainty differences involved in ROP calculations when the heavy water reactor core is loaded with the original design fuel and the improved fuel, and increasing the uncertainty when setting the core uncertainty in the "high power region" if the difference is large.

6. The method for applying the ROP set value in advance in the transition core of an improved fuel of a heavy water reactor according to claim 2, characterized in that: Step 34 includes calculating and comparing ROP setpoints for the complete calculation method for the cores with the original design fuel and the improved fuel at the time point of the initial day of core replacement plus D EFPDs; calculating ROP setpoints for the core in the "high power region" using a combination of thermal-hydraulic conditions and uncertainties at the time point of the initial day of core replacement plus D EFPDs, and arranging them into a list; and determining the increase in the setpoint under the complete ROP calculation for the core in the "high power region" at full power, Y%, as compared to the core with the original design fuel, using the minimum increase in the list, Y%.

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