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{{#Wiki_filter:UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 March 28, 2013 Mr. Ray Leib Site Vice President FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Mail Stop A-DB-3080 5501 North State Route 2 Oak Harbor, OH 43449-9760 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 -CLOSURE EVALUATION FOR REPORT PURSUANT TO TITLE 10 OF THE CODE OF FEDERAL REGULA TlONS, PART 50, SECTION 50.46(a)(3)
{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 28, 2013 Mr. Ray Leib Site Vice President FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Mail Stop A-DB-3080 5501 North State Route 2 Oak Harbor, OH 43449-9760
CONCERNING SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERRORS/CHANGES RELATED TO EMERGENCY CORE COOLING SYSTEM BYPASS AND UPPER PLENUM COLUMN WELDMENTS (TAC NO. ME8411)  
 
==SUBJECT:==
DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 - CLOSURE EVALUATION FOR REPORT PURSUANT TO TITLE 10 OF THE CODE OF FEDERAL REGULA TlONS, PART 50, SECTION 50.46(a)(3) CONCERNING SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERRORS/CHANGES RELATED TO EMERGENCY CORE COOLING SYSTEM BYPASS AND UPPER PLENUM COLUMN WELDMENTS (TAC NO. ME8411)


==Dear Mr. Leib:==
==Dear Mr. Leib:==
Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.46(a)(3), First Energy Nuclear Operating Company (the licensee), submitted a report describing two significant errors/changes identified in the emergency core cooling system (ECCS) evaluation model, and an estimate of the effects of the errors/changes on the predicted peak cladding temperature (PCT) for the Davis-Besse Nuclear Power Station, Unit No.1 (DBNPS). The report was submitted to the U.S. Nuclear Regulatory Commission (NRC or Commission) by letter dated March 16, 2012 (Agencywide Document Access and Management System (ADAMS) Accession No. ML12076A237), and supplemented by letter dated December 18, 2012 (ADAMS Accession No. ML12353A601).
The NRC staff finds that the report submitted pursuant to 10 CFR 50.46(a)(3), concerning ECCS evaluation model errors/changes pertaining to end of ECCS bypass and column weldments, satisfies 10 CFR 50.46 reporting requirements.
The report and supplemental information provided by AREVA NP Inc. enabled the NRC staff to (1) determine that it agrees with the licensee's assessment of the significance of the error. (2) confirm that the evaluation model remains adequate, and (3) verify that the licensee continues to meet the PCT acceptance criterion promulgated by 10 CFR 50.46(b).
The NRC staff does not agree that DBNPS's action is in compliance with the reanalysis requirement set forth in 10 CFR 50.46(a)(3)(ii), which states that the licensee "shall include with the report a proposed schedule for providing a reanalysis or talking other action as may be needed to show compliance with [10 CFR] 50.46 requirements." The NRC review of the report is complete and TAC No. ME8411 will be closed.
R. Leib -Please contact me at 301-415-3867.
if you have any Docket No.
Closure cc: Listserv sin7J, / ' L //!/i
/ I 1._ " M' hael Projec Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 CLOSURE EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION DAVIS-BESSE NUCLEAR POWER STATION, UNIT 1 REPORT DESCRIBING THE NATURE OF AND ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE OF A SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERRORS/CHANGES


==1.0 INTRODUCTION==
Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.46(a)(3),
First Energy Nuclear Operating Company (the licensee), submitted a report describing two significant errors/changes identified in the emergency core cooling system (ECCS) evaluation model, and an estimate of the effects of the errors/changes on the predicted peak cladding temperature (PCT) for the Davis-Besse Nuclear Power Station, Unit No.1 (DBNPS).
The report was submitted to the U.S. Nuclear Regulatory Commission (NRC or Commission) by letter dated March 16, 2012 (Agencywide Document Access and Management System (ADAMS)
Accession No. ML12076A237), and supplemented by letter dated December 18, 2012 (ADAMS Accession No. ML12353A601).
The NRC staff finds that the report submitted pursuant to 10 CFR 50.46(a)(3), concerning ECCS evaluation model errors/changes pertaining to end of ECCS bypass and column weldments, satisfies 10 CFR 50.46 reporting requirements. The report and supplemental information provided by AREVA NP Inc. enabled the NRC staff to (1) determine that it agrees with the licensee's assessment of the significance of the error. (2) confirm that the evaluation model remains adequate, and (3) verify that the licensee continues to meet the PCT acceptance criterion promulgated by 10 CFR 50.46(b).
The NRC staff does not agree that DBNPS's action is in compliance with the reanalysis requirement set forth in 10 CFR 50.46(a)(3)(ii), which states that the licensee "shall include with the report a proposed schedule for providing a reanalysis or talking other action as may be needed to show compliance with [10 CFR] 50.46 requirements."
The NRC review of the report is complete and TAC No. ME8411 will be closed.


Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.46(a)(3), First Energy Nuclear Operating Company (the licensee), submitted a report describing two significant errors/changes identified in the emergency core cooling system (ECCS) evaluation model, and an estimate of the effects of the errors/changes on the predicted peak cladding temperature (PCT) for the Davis-Besse Nuclear Power Station, Unit No.1 (DBNPS). The report was submitted to the U.S. Nuclear Regulatory Commission (NRC, Commission) by letter dated March 16, 2012 (Agencywide Document Access and Management System (ADAMS) Accession No. ML12076A237), and supplemented by letter dated December 18, 2012 (ADAMS Accession No. ML12353A601).
R. Leib                                    - 2 Please contact me at 301-415-3867. if you have any questions.
The NRC staff has evaluated the report, along with its supplemental information, and determined that it satisfies the requirements of 10 CFR 50.46(a)(3), but does not satisfy the intent of the reporting requirements, as discussed in the Statements of Considerations (SC) published in the Federal Register (FR), (53 FR 35996; September 16, 1988) for the realistic ECCS evaluations revision of 10 CFR 50.46. The NRC staff review is discussed in the following sections of this closure evaluation.
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a~~r Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346


==2.0 REGULATORY EVALUATION==
==Enclosure:==


2.1 Requirements Contained in 10 CFR 50.46 Acceptance criteria for ECCS for light-water nuclear power reactors are promulgated at 10 CFR 50.46. In particular, 10 CFR 50.46(a)(3)(i) requires licensees to estimate the effect of any change, or error in an acceptable evaluation model, or in the application of such a model, to determine if the change or error is significant.
Closure Evaluation cc: Listserv
For the purpose of 10 CFR 50.46, a significant change or error is one which results in a calculated peak fuel cladding temperature difference by more than 50 degrees Fahrenheit
 
("F) from the temperature calculated for the limiting transient using the last acceptable model, or is an accumulation of changes and errors such that the sum Enclosure
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 CLOSURE EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION DAVIS-BESSE NUCLEAR POWER STATION, UNIT 1 REPORT DESCRIBING THE NATURE OF AND ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE OF A SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERRORS/CHANGES
-of the absolute magnitudes of the respective temperature changes is greater than 50 of. For each change to or error discovered in an acceptable evaluation model, or in the application of such a model, Paragraph (a)(3)(ii) to 10 CFR 50.46 requires the affected licensee to report the change or error and its estimated effect on the limiting ECCS analysis to the Commis! least annually.
 
If the change or error is significant, the licensee is required to provide this within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 10 CFR 50.46 requirements.
==1.0      INTRODUCTION==
2.2 Additional Guidance Additional clarification concerning the intent of the reporting requirements is discussed in the Federal Register (53 FR 35996): [Paragraph (a)(3) of section 50.46] requires that all changes or errors in approved evaluation models be reported at least annually and does not require any further action by the licensee until the error is reported.
 
Thereafter, although reanalysis is not required solely because of such minor error, any subsequent calculated evaluation of ECCS performance requires use of a model with such error, and any prior errors, corrected.
Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.46(a)(3),
The NRC needs to be apprised of even minor errors or changes in order to ensure that they agree with the applicant's or licensee's assessment of the significance of the error or change and to maintain cognizance of modifications made subsequent to NRC review of the evaluation modeL .. Significant errors require more timely attention since they may be important to the safe operation of the plant and raise questions as to the adequacy of the overall evaluation model ... More timely reporting is required for significant errors or changes ... the final rule revision also allows the NRC to determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements.
First Energy Nuclear Operating Company (the licensee), submitted a report describing two significant errors/changes identified in the emergency core cooling system (ECCS) evaluation model, and an estimate of the effects of the errors/changes on the predicted peak cladding temperature (PCT) for the Davis-Besse Nuclear Power Station, Unit No.1 (DBNPS).
The report was submitted to the U.S. Nuclear Regulatory Commission (NRC, Commission) by letter dated March 16, 2012 (Agencywide Document Access and Management System (ADAMS)
Accession No. ML12076A237), and supplemented by letter dated December 18, 2012 (ADAMS Accession No. ML12353A601).
The NRC staff has evaluated the report, along with its supplemental information, and determined that it satisfies the requirements of 10 CFR 50.46(a)(3), but does not satisfy the intent of the reporting requirements, as discussed in the Statements of Considerations (SC) published in the Federal Register (FR), (53 FR 35996; September 16, 1988) for the realistic ECCS evaluations revision of 10 CFR 50.46. The NRC staff review is discussed in the following sections of this closure evaluation.
 
==2.0      REGULATORY EVALUATION==
 
2.1     Requirements Contained in 10 CFR 50.46 Acceptance criteria for ECCS for light-water nuclear power reactors are promulgated at 10 CFR 50.46. In particular, 10 CFR 50.46(a)(3)(i) requires licensees to estimate the effect of any change, or error in an acceptable evaluation model, or in the application of such a model, to determine if the change or error is significant. For the purpose of 10 CFR 50.46, a significant change or error is one which results in a calculated peak fuel cladding temperature difference by more than 50 degrees Fahrenheit ("F) from the temperature calculated for the limiting transient using the last acceptable model, or is an accumulation of changes and errors such that the sum Enclosure
 
                                                  - 2 of the absolute magnitudes of the respective temperature changes is greater than 50 of.
For each change to or error discovered in an acceptable evaluation model, or in the application of such a model, Paragraph (a)(3)(ii) to 10 CFR 50.46 requires the affected licensee to report the change or error and its estimated effect on the limiting ECCS analysis to the Commis!
least annually. If the change or error is significant, the licensee is required to provide this within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 10 CFR 50.46 requirements.
2.2     Additional Guidance Additional clarification concerning the intent of the reporting requirements is discussed in the Federal Register (53 FR 35996):
[Paragraph (a)(3) of section 50.46] requires that all changes or errors in approved evaluation models be reported at least annually and does not require any further action by the licensee until the error is reported. Thereafter, although reanalysis is not required solely because of such minor error, any subsequent calculated evaluation of ECCS performance requires use of a model with such error, and any prior errors, corrected.
The NRC needs to be apprised of even minor errors or changes in order to ensure that they agree with the applicant's or licensee's assessment of the significance of the error or change and to maintain cognizance of modifications made subsequent to NRC review of the evaluation modeL ..
Significant errors require more timely attention since they may be important to the safe operation of the plant and raise questions as to the adequacy of the overall evaluation model ... More timely reporting is required for significant errors or changes ... the final rule revision also allows the NRC to determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements.
The NRC staff considered the discussion in the Federal Register in its evaluation of the error report submitted by the licensee.
The NRC staff considered the discussion in the Federal Register in its evaluation of the error report submitted by the licensee.
3.0 TECHNICAL EVALUATION The report submitted by the licensee described the effects of two errors/changes which affect the large-break loss-of-coolant accident (LBLOCA) analysis of record. The first item is an error in the determination of the end of ECCS bypass. The second item is a change in the evaluation model (EM) to include the effects of the upper plenum column weldments.
 
==3.0     TECHNICAL EVALUATION==
 
The report submitted by the licensee described the effects of two errors/changes which affect the large-break loss-of-coolant accident (LBLOCA) analysis of record. The first item is an error in the determination of the end of ECCS bypass. The second item is a change in the evaluation model (EM) to include the effects of the upper plenum column weldments.
Based on the nature of the reported errors/changes, and on the magnitude of the effects on the PCT calculation, the NRC staff determined that a detailed technical review was necessary.
Based on the nature of the reported errors/changes, and on the magnitude of the effects on the PCT calculation, the NRC staff determined that a detailed technical review was necessary.
Based on the regulatory evaluation discussed above, the NRC staff's review was performed to ensure that the NRC staff agreed with the licensee's assessment of the significance of the errors/changes, and to enable the staff to verify that the evaluation model, as a whole, remains adequate.
Based on the regulatory evaluation discussed above, the NRC staff's review was performed to ensure that the NRC staff agreed with the licensee's assessment of the significance of the errors/changes, and to enable the staff to verify that the evaluation model, as a whole, remains adequate. Finally, the NRC staff's review also establishes that the licensee's proposed schedule for re-analysis is acceptable in light of the safety significance of the reported errors/changes.
Finally, the NRC staff's review also establishes that the licensee's proposed schedule for re-analysis is acceptable in light of the safety significance of the reported errors/changes.
 
-3 3.1 Summary of Technical Information in the Report The licensee's report indicated two errors/changes which affect the PCT for LBLOCA analysis.
                                                  -3 3.1     Summary of Technical Information in the Report The licensee's report indicated two errors/changes which affect the PCT for LBLOCA analysis.
The first item is an error in the determination of the end of ECCS bypass resulting in a decrease in PCT of 80 OF. The second item is a change to include the effects of the upper plenum column weldments resulting in a PCT increase of 80 OF. The nature of the errors, and the methods used to estimate the effects on the calculated PCT, is discussed in greater detail in the AREVA generic request for additional information (RAI) Response letter "Generic RAI Response to a day 10 CFR 50.46 Report of Significant PCT Change," December 6,2012 (ADAMS Accession No. ML12342A381).
The first item is an error in the determination of the end of ECCS bypass resulting in a decrease in PCT of 80 OF. The second item is a change to include the effects of the upper plenum column weldments resulting in a PCT increase of 80 OF. The nature of the errors, and the methods used to estimate the effects on the calculated PCT, is discussed in greater detail in the AREVA generic request for additional information (RAI) Response letter "Generic RAI Response to a 30 day 10 CFR 50.46 Report of Significant PCT Change," December 6,2012 (ADAMS Accession No. ML12342A381).
ECCS Bypass Error A mathematical error was identified with the control variables in the energy balance calculations used to determine the complete end of ECCS bypass in LBLOCA applications.
ECCS Bypass Error A mathematical error was identified with the control variables in the energy balance calculations used to determine the complete end of ECCS bypass in LBLOCA applications. As defined in the generic RAI responses from AREVA, the complete end of bypass is achieved when all of the injected ECCS liquid reaches the lower plenum before core reflood analysis begins. When the ECCS flows can condense all of the steam flowing into the upper downcomer region, the end of blowdown has occurred. The complete end of bypass time is the earliest end of blowdown time or the time at which the ECCS flows can condense all the steam flowing into the upper downcomer region. The error in the control variables incorrectly calculated the complete end of bypass time and liquid mass that should have remained in the vessel at the end of blowdown.
As defined in the generic RAI responses from AREVA, the complete end of bypass is achieved when all of the injected ECCS liquid reaches the lower plenum before core reflood analysis begins. When the ECCS flows can condense all of the steam flowing into the upper downcomer region, the end of blowdown has occurred.
Estimation of the Effect of ECCS Bypass in the PCT Calculation The DBNPS has 177 fuel assemblies and uses a raised loop (RL) design nuclear steam supply system (NSSS). The effect of the ECCS bypass error was analyzed explicitly for a 205-fuel assembly plant and applied to a 177-fuel assembly plant. The control variables are common to both 205-fuel assembly and 177 -fuel assembly plants; therefore, the correction is applicable to DBNPS. A blowdown reanalysis was performed using the computer code RELAP5 with corrected bypass variables for a 177-fuel assembly RL plant. The case analyzed was for a 9.536-foot (ft) peak power location at beginning of life. The NRC staff determined that a power shape that uses a high peaking elevation in the core is appropriate because the higher elevation will remain uncovered the longest. This tendency occurs because the reflood begins in the bottom of the core and proceeds to the top, allowing for the longest adiabatic heatup to occur at a higher elevation.
The complete end of bypass time is the earliest end of blowdown time or the time at which the ECCS flows can condense all the steam flowing into the upper downcomer region. The error in the control variables incorrectly calculated the complete end of bypass time and liquid mass that should have remained in the vessel at the end of blowdown.
With the correction accounted for, the analysis showed a reduction in the end of ECCS bypass time by roughly two seconds. When the end of bypass time occurred earlier, the amount of ECCS fluid that was not bypassed increased. The ECCS bypass liquid added to the lower plenum caused the bottom of core recovery to occur earlier. This caused the core refill period to become shorter and reflood began earlier. Since core reflood occurred sooner, core cooling began earlier which caused the predicted PCT to decrease.
Estimation of the Effect of ECCS Bypass in the PCT Calculation The DBNPS has 177 fuel assemblies and uses a raised loop (RL) design nuclear steam supply system (NSSS). The effect of the ECCS bypass error was analyzed explicitly for a 205-fuel assembly plant and applied to a 177 -fuel assembly plant. The control variables are common to both 205-fuel assembly and 177 -fuel assembly plants; therefore, the correction is applicable to DBNPS. A blowdown reanalysis was performed using the computer code RELAP5 with corrected bypass variables for a 177 -fuel assembly RL plant. The case analyzed was for a 9.536-foot (ft) peak power location at beginning of life. The NRC staff determined that a power shape that uses a high peaking elevation in the core is appropriate because the higher elevation will remain uncovered the longest. This tendency occurs because the reflood begins in the bottom of the core and proceeds to the top, allowing for the longest adiabatic heatup to occur at a higher elevation.
The licensee reported that the limiting PCT decreased by 96.7 of for the ruptured node and 51.7 of for the unruptured node in a 177-fuel assembly RL plant. The licensee observed that there is a typical two-to-one difference in PCT between ruptured and unruptured fuel cladding segments resulting from the metal to water reaction. The licensee conservatively reported a PCT decrease of 80 of for ruptured segments and 40 of for unruptured segments due to the end of bypass time occurring earlier in the LBLOCA analysis for 177-fuel assembly
With the correction accounted for, the analysis showed a reduction in the end of ECCS bypass time by roughly two seconds. When the end of bypass time occurred earlier, the amount of ECCS fluid that was not bypassed increased.
 
The ECCS bypass liquid added to the lower plenum caused the bottom of core recovery to occur earlier. This caused the core refill period to become shorter and reflood began earlier. Since core reflood occurred sooner, core cooling began earlier which caused the predicted PCT to decrease.
                                                -4 RL plant.
The licensee reported that the limiting PCT decreased by 96.7 of for the ruptured node and 51.7 of for the unruptured node in a 177 -fuel assembly RL plant. The licensee observed that there is a typical two-to-one difference in PCT between ruptured and unruptured fuel cladding segments resulting from the metal to water reaction.
The results from a small-break-Ioss-of-coolant accident (SBLOCA) analyses were not impacted since the ECCS bypass is not used for SBLOCA.
The licensee conservatively reported a PCT decrease of 80 of for ruptured segments and 40 of for unruptured segments due to the end of bypass time occurring earlier in the LBLOCA analysis for 177 -fuel assembly
Column Weldment Model Change The change in the EM was caused by the inability of the RELAPS model to account for the effects of column weldments over the hot bundle. Sensitivity studies were performed with upper plenum column weldments added to the EM. Column weldments allow a portion of the flow exiting from the core channels underneath them to reach the upper head. This makes them important in determining the temperature of the fluid reaching the upper head.
-4 RL plant. The results from a small-break-Ioss-of-coolant accident (SBLOCA) analyses were not impacted since the ECCS bypass is not used for SBLOCA. Column Weldment Model Change The change in the EM was caused by the inability of the RELAPS model to account for the effects of column weldments over the hot bundle. Sensitivity studies were performed with upper plenum column weldments added to the EM. Column weldments allow a portion of the flow exiting from the core channels underneath them to reach the upper head. This makes them important in determining the temperature of the fluid reaching the upper head. A LBLOCA RELAPS model was initially developed for a 20S-fuel assembly plant incorporating column weldments (CW) on top of the hot bundle of a 177 -fuel assembly plant with a lowered loop (LL) design nuclear steam supply system. Since the column weldment design details for a 177 -fuel assembly plant were not readily available when the column weldments were first assessed, a simplified approach used the column weldment model that was developed for the 20S-fuel assembly plant and incorporated it over the 177 -fuel assembly LL plant. The DBNPS has 177 fuel assemblies and uses a RL design. When the effect of column weldments was being analyzed for a RL plant, additional details of the column weldments for a 177-fuel assembly plant were developed.
A LBLOCA RELAPS model was initially developed for a 20S-fuel assembly plant incorporating column weldments (CW) on top of the hot bundle of a 177-fuel assembly plant with a lowered loop (LL) design nuclear steam supply system. Since the column weldment design details for a 177-fuel assembly plant were not readily available when the column weldments were first assessed, a simplified approach used the column weldment model that was developed for the 20S-fuel assembly plant and incorporated it over the 177-fuel assembly LL plant.
Estimation of the Effect of Column Weldments in the PCT Calculation The licensee reported that when column weldments modeled for a 20S-fuel assembly LL plant were incorporated over the hot channel in a 177 -fuel assembly LL plant, the RELAPS blowdown analysis resulted in an increase in PCT of 3S.6 OF for the peak unruptured fuel cladding segment. Due to the metal to water reaction inside the fuel, there is a two-to-one variation in PCT change between ruptured and unruptured segments.
The DBNPS has 177 fuel assemblies and uses a RL design. When the effect of column weldments was being analyzed for a RL plant, additional details of the column weldments for a 177-fuel assembly plant were developed.
The temperature change of 3S.6 OF was doubled by the licensee to account for both sides of the ruptured node undergoing the exothermic reaction.
Estimation of the Effect of Column Weldments in the PCT Calculation The licensee reported that when column weldments modeled for a 20S-fuel assembly LL plant were incorporated over the hot channel in a 177-fuel assembly LL plant, the RELAPS blowdown analysis resulted in an increase in PCT of 3S.6 OF for the peak unruptured fuel cladding segment. Due to the metal to water reaction inside the fuel, there is a two-to-one variation in PCT change between ruptured and unruptured segments. The temperature change of 3S.6 OF was doubled by the licensee to account for both sides of the ruptured node undergoing the exothermic reaction. This resulted in an estimated increase in fuel temperature of 71.2°F for the rupture-limited cladding segments in a 177-fuel assembly LL to account for upper plenum column weldments modeled from a 20S-fuel assembly LL plant.
This resulted in an estimated increase in fuel temperature of 71.2°F for the rupture-limited cladding segments in a 177 -fuel assembly LL to account for upper plenum column weldments modeled from a 20S-fuel assembly LL plant. Typically, the PCT will increase in proportion to the end of blowdown fuel temperature.
Typically, the PCT will increase in proportion to the end of blowdown fuel temperature. For this case, there was a favorable shift of rupture time into the blowdown phase causing the rupture segment PCT to increase by 26.2 OF from the computer code BEACH calculation. This is compared to an increase in fuel temperature of 71.2°F from the end of blowdown calculation using RELAPS. BEACH calculated an increase in PCT of 11 .S OF for the unruptured fuel segment. This can be seen in Table 1, below.
For this case, there was a favorable shift of rupture time into the blowdown phase causing the segment PCT to increase by 26.2 OF from the computer code BEACH calculation.
When a 177-fuel assembly plant with a RL design (like DBNPS) was being analyzed, details of column weldments of a 177-fuel assembly plant had been developed. These details were used to compare results of the column weldment models between a 177-fuel assembly plant and a 20S-fuel assembly plant. The licensee reported that when the cases were analyzed with column weldments from the 177-fuel assembly plant, the PCT increased by 3 OF for an unruptured fuel segment. This result was doubled for the ruptured fuel segment. This results in a PCT increase of 14.S OF for the unruptured fuel segment and 32.2 OF for the ruptured fuel segment in the LL
This is compared to an increase in fuel temperature of 71.2°F from the end of blowdown calculation using RELAPS. BEACH calculated an increase in PCT of 11 .S OF for the unruptured fuel segment. This can be seen in Table 1, below. When a 177-fuel assembly plant with a RL design (like DBNPS) was being analyzed, details of column weldments of a 177-fuel assembly plant had been developed.
 
These details were used to compare results of the column weldment models between a 177 -fuel assembly plant and a 20S-fuel assembly plant. The licensee reported that when the cases were analyzed with column weldments from the 177-fuel assembly plant, the PCT increased by 3 OF for an unruptured fuel segment. This result was doubled for the ruptured fuel segment. This results in a PCT increase of 14.S OF for the unruptured fuel segment and 32.2 OF for the ruptured fuel segment in the LL
                                                  - 5 plant as seen in Table 1.
-plant as seen in Table 1. When column weldments were applied to the RL design, the PCT increased by 8.9 of for the unruptured segment. The ruptured segment yielded a decrease in PCT; therefore, it is not the limiting case for a raised loop analysis.
When column weldments were applied to the RL design, the PCT increased by 8.9 of for the unruptured segment. The ruptured segment yielded a decrease in PCT; therefore, it is not the limiting case for a raised loop analysis. These results can be seen in Table 1.
These results can be seen in Table 1. Table 1: Effect of Column Weldments on PCT No. of Fuel Assemblies Loop Design Ruptured!
Table 1: Effect of Column Weldments on PCT Ruptured!            No. of Fuel No. of Fuel                       Unruptured Fuel      Assemblies CW Assemblies      Loop Design            Segment            modeled for       Effect on PCT 177       Lowered loop         Unruptured                205          Increase 11.5° 177        Lowered loop           Ruptured                205          Increase 26.5° 177        Lowered loop         Unruptured               177
Unruptured Fuel Segment No. of Fuel Assemblies CW modeled for Effect on PCT 177 177 177 177 177 177 Lowered loop Lowered loop Lowered loop Lowered loop Raised loop Raised loop Unruptured Ruptured Unruptured Ruptured Unruptured Ruptured 205 205 177 177 177 177
* Increase 14.5° !
* I Increase 11.5° Increase 26.5° Increase 14.5° Increase 32.2° Increase 8.9 0 Decrease ! The licensee used a generic increase of 40 of for unruptured fuel segments and 80 of for ruptured fuel segments for the effect of column weldments.
177        Lowered loop          Ruptured                177        I Increase 32.2° 177        Raised loop          Unruptured                177            Increase 8.9 0 177        Raised loop          Ruptured                177              Decrease The licensee used a generic increase of 40 of for unruptured fuel segments and 80 of for ruptured fuel segments for the effect of column weldments. These estimates are bounding and conservative for the analyses completed for DBNPS.
These estimates are bounding and conservative for the analyses completed for DBNPS. An evaluation of the impact of CW on SBLOCA analyses was performed.
An evaluation of the impact of CW on SBLOCA analyses was performed. It was concluded that SBLOCA analyses are unaffected by the CW modeling because the net flow remains upward during these slower evolving transients.
It was concluded that SBLOCA analyses are unaffected by the CW modeling because the net flow remains upward during these slower evolving transients.
Reported Results Following the correction for ECCS bypass and the column weldment model change, the current predicted PCT for LBLOCA at DBNPS is 2119 of.
Reported Results Following the correction for ECCS bypass and the column weldment model change, the current predicted PCT for LBLOCA at DBNPS is 2119 of. 3.2 Summary of Staff Evaluation I n its evaluation, the NRC staff reviewed (1) the approach used to estimate the effects of the ECCS bypass error and the effects of upper plenum column weldments, (2) the estimated effect of both errors/changes, and (3) the licensee's proposal to not perform a reanalysis in consideration of the approach used to estimate the effects of the errors/changes.
3.2     Summary of Staff Evaluation In its evaluation, the NRC staff reviewed (1) the approach used to estimate the effects of the ECCS bypass error and the effects of upper plenum column weldments, (2) the estimated effect of both errors/changes, and (3) the licensee's proposal to not perform a reanalysis in consideration of the approach used to estimate the effects of the errors/changes. As discussed in the following paragraphs, the NRC staff determined that the licensee's estimates are acceptable; however, the NRC staff does not agree with the proposal to not perform a reanalysis.
As discussed in the following paragraphs, the NRC staff determined that the licensee's estimates are acceptable; however, the NRC staff does not agree with the proposal to not perform a reanalysis.
To estimate the effects of the ECCS bypass error, the licensee analyzed the effect of correcting the control variables in the energy balance equation used to determine the complete end of bypass time in LBLOCA applications. The effect was analyzed explicitly for a 205-fuel assembly plant and applied to 177-fuel assembly plants. The control variables were common to both 205 fuel assembly plants and 177-fuel assembly plants; therefore, the correction is applicable to
To estimate the effects of the ECCS bypass error, the licensee analyzed the effect of correcting the control variables in the energy balance equation used to determine the complete end of bypass time in LBLOCA applications.
 
The effect was analyzed explicitly for a 205-fuel assembly plant and applied to 177 -fuel assembly plants. The control variables were common to both fuel assembly plants and 177 -fuel assembly plants; therefore, the correction is applicable to DBNPS as a 177 -fuel assembly plant. To estimate the effects of upper plenum column weldments, the licensee included column weldments over the hot channel in a model of a 205-fuel assembly plant. This model was incorporated over the hotel channel in a 177 -fuel assembly lowered loop plant and a RELAP5 blowdown analysis was completed.
                                                -6 DBNPS as a 177-fuel assembly plant.
Sensitivity studies were performed using column weldments modeled for a 177 -fuel assembly raised loop plant. The effects of the studies showed the generic estimate of the effect of column weldments was conservative for RL plants. The licensee estimated the effect of the ECCS bypass error to be a decrease in PCT of 80 oF. The licensee estimated the effect of upper plenum column weldments to be an increase in PCT of 80 of. The current predicted PCT for LBLOCA at DBNPS is 2119 of. As stated in 10 CFR 50.46(a)(3)(ii), the licensee "shall include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with [10 CFR] 50.46 requirements." As described in Section 2.0, above, the SOC explain further that "the final rule revision also allows the NRC to determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements." In the RAI issued by the NRC on October 15, 2012 (ADAMS Accession No. NIL 12279A037), the NRC staff requested that the licensee "justify not providing a schedule for reanalysis or taking other action to show compliance with 10 CFR 50.46." The RAI response was provided by the licensee on December 18, 2012 (ADAMS Accession No. ML12353A601).
To estimate the effects of upper plenum column weldments, the licensee included column weldments over the hot channel in a model of a 205-fuel assembly plant. This model was incorporated over the hotel channel in a 177-fuel assembly lowered loop plant and a RELAP5 blowdown analysis was completed. Sensitivity studies were performed using column weldments modeled for a 177-fuel assembly raised loop plant. The effects of the studies showed the generic estimate of the effect of column weldments was conservative for RL plants.
The licensee estimated the effect of the ECCS bypass error to be a decrease in PCT of 80 oF.
The licensee estimated the effect of upper plenum column weldments to be an increase in PCT of 80 of. The current predicted PCT for LBLOCA at DBNPS is 2119 of.
As stated in 10 CFR 50.46(a)(3)(ii), the licensee "shall include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with [10 CFR] 50.46 requirements." As described in Section 2.0, above, the SOC explain further that "the final rule revision also allows the NRC to determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements."
In the RAI issued by the NRC on October 15, 2012 (ADAMS Accession No. NIL12279A037), the NRC staff requested that the licensee "justify not providing a schedule for reanalysis or taking other action to show compliance with 10 CFR 50.46." The RAI response was provided by the licensee on December 18, 2012 (ADAMS Accession No. ML12353A601).
The licensee indicated in the response that both the error and change to the ECCS evaluation model that were presented in the 10 CFR 50.46 30-day report were analyzed in detail and that the impact of both items does not result in challenge to the 10 CFR 50.46(b) acceptance criteria.
The licensee indicated in the response that both the error and change to the ECCS evaluation model that were presented in the 10 CFR 50.46 30-day report were analyzed in detail and that the impact of both items does not result in challenge to the 10 CFR 50.46(b) acceptance criteria.
The licensee also concluded that the evaluation model is considered acceptable since the error and change have been analyzed and there are no other known errors or changes in the model at this time. The NRC staff determined that the PCT error evaluations are supported by explicit analysis using the Babcock &Wilcox plant ECCS evaluation model, and the error-adjusted LBLOCA PCTs for DBNPS remain below the 10 CFR 50.46(b) regulatory acceptance criteria.
The licensee also concluded that the evaluation model is considered acceptable since the error and change have been analyzed and there are no other known errors or changes in the model at this time.
NRC Information Notice 97-15. Supplement 1, "Reporting of Errors and Changes in Break/Small-Break Loss-of-Coolant Evaluation models of Fuel Vendors and Compliance with 10 CFR 50.46(a)(3)" describes, among other things, a prior issue at another plant. in which a proposed schedule for reanalysis was not included with a report submitted pursuant to 1 0 CFR 50.46(a)(3).
The NRC staff determined that the PCT error evaluations are supported by explicit analysis using the Babcock & Wilcox plant ECCS evaluation model, and the error-adjusted LBLOCA PCTs for DBNPS remain below the 10 CFR 50.46(b) regulatory acceptance criteria.
To address the NRC staff review, the licensee indicated that they would provide such a reanalysis.
NRC Information Notice 97-15. Supplement 1, "Reporting of Errors and Changes in Large Break/Small-Break Loss-of-Coolant Evaluation models of Fuel Vendors and Compliance with 10 CFR 50.46(a)(3)" describes, among other things, a prior issue at another plant. in which a proposed schedule for reanalysis was not included with a report submitted pursuant to 10 CFR 50.46(a)(3). To address the NRC staff review, the licensee indicated that they would provide such a reanalysis. The NRC staff is unable to determine that either DBNPS's report or RAI response addresses the reanalysis requirement.
The NRC staff is unable to determine that either DBNPS's report or RAI response addresses the reanalysis requirement.
In summary, the NRC staff reviewed the licensees report estimating the effect of the ECCS bypass error and column weldments on the LBLOCA analyses for DBNPS. Since the evaluation included explicit analyses of the ECCS bypass error and column weldments in the evaluation model, the NRC staff concluded that the error estimates were acceptable. The NRC staff does not agree that DBNPS action is in compliance with the reanalysis requirement set forth in 10 CFR 50.46(a)(3)(ii). "Other action" should not be interpreted as no action.
In summary, the NRC staff reviewed the licensees report estimating the effect of the ECCS bypass error and column weldments on the LBLOCA analyses for DBNPS. Since the evaluation included explicit analyses of the ECCS bypass error and column weldments in the evaluation model, the NRC staff concluded that the error estimates were acceptable.
 
The NRC staff does not agree that DBNPS action is in compliance with the reanalysis requirement set forth in 10 CFR 50.46(a)(3)(ii). "Other action" should not be interpreted as no action.
                                                -7
-7  


==4.0 CONCLUSION==
==4.0     CONCLUSION==


Based on the considerations discussed above, the NRC staff finds that the report submitted pursuant to 10 CFR 50.46(a)(3), concerning ECCS evaluation model errors/changes pertaining to end of ECCS bypass and column weldments satisfies 10 CFR 50.46 reporting requirements.
Based on the considerations discussed above, the NRC staff finds that the report submitted pursuant to 10 CFR 50.46(a)(3), concerning ECCS evaluation model errors/changes pertaining to end of ECCS bypass and column weldments satisfies 10 CFR 50.46 reporting requirements.
The report and supplemental information provided by AREVA NP Inc. enabled the NRC staff to (1) determine that it agrees with the licensee's assessment of the significance of the error, (2) con'firm that the evaluation model remains adequate, and (3) verify that the licensee continues to meet the PCT acceptance criterion promulgated by 10 CFR 50.46(b).
The report and supplemental information provided by AREVA NP Inc. enabled the NRC staff to (1) determine that it agrees with the licensee's assessment of the significance of the error, (2) con'firm that the evaluation model remains adequate, and (3) verify that the licensee continues to meet the PCT acceptance criterion promulgated by 10 CFR 50.46(b).
However, the NRC staff does not agree that DSNPS's action is in compliance with the reanalysis requirement set forth in 10 CFR 50.46(a)(3)(ii), which states that the licensee "shall include with the report a proposed schedule for providing a reanalysis or talking other action as may be needed to show compliance with [10 CFR] 50.46 requirements." Principal Contributors:
However, the NRC staff does not agree that DSNPS's action is in compliance with the reanalysis requirement set forth in 10 CFR 50.46(a)(3)(ii), which states that the licensee "shall include with the report a proposed schedule for providing a reanalysis or talking other action as may be needed to show compliance with [10 CFR] 50.46 requirements."
A. Guzzetta, NRR B. Parks, NRR Date of issuance:
Principal Contributors: A. Guzzetta, NRR B. Parks, NRR Date of issuance: March 28, 2013
March 28, 2013 R. Leib -2 Please contact me at 301-415-3867, if you have any Docket No.
 
Closure cc: Listserv DISTRIBUTION:
R. Leib                                   -2 Please contact me at 301-415-3867, if you have any questions.
PUBLIC LPL3-2 R/F RidsAcrsAcnw_MailCTR Resource RidsNrrSrxb Resource RidsRgn1 MailCenter Resource RidsNrrDorlDpr Resource Sincerely, 1 RAJ Michael Mahoney, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation AGuzzetta, NRR RidsNrrDorlLpl3-2 Resouce RidsNrrPMDavis-Besse Resource RidsNrrLASRohrerResource BParks, NRR ADAMS A ccesslon N *8 Iy M aed M h 12 , 2013 0.: ML13086A165 erno D t arc OFFICE LPL3-2/PM LPL3-2/LA DSS/SRXB/BC LPL3-2/BC NAME IVIMahoney SRohrer CJackson JBowen MMahon DATE 3/28/13 3/27/13 3/12/13* 3/28/13 3/28/13 OFFICIAL RECORD}}
Sincerely, 1 RAJ Michael Mahoney, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346
 
==Enclosure:==
 
Closure Evaluation cc: Listserv DISTRIBUTION:
PUBLIC                                    AGuzzetta, NRR LPL3-2 R/F                                RidsNrrDorlLpl3-2 Resouce RidsAcrsAcnw_MailCTR Resource              RidsNrrPMDavis-Besse Resource RidsNrrSrxb Resource                      RidsNrrLASRohrerResource RidsRgn1 MailCenter Resource              BParks, NRR RidsNrrDorlDpr Resource ADAMS A ccesslon N0.: ML13086A165                        *8Iy Merno Daet d Marc h 12, 2013 OFFICE LPL3-2/PM NAME      IVIMahoney LPL3-2/LA SRohrer DSS/SRXB/BC CJackson LPL3-2/BC JBowen LPL3-21~1 MMahon DATE       3/28/13         3/27/13           3/12/13*         3/28/13         3/28/13 OFFICIAL RECORD COPY}}

Revision as of 21:12, 4 November 2019

Closure Evaluation for Report Pursuant to Title 10 of the Code of Federal Regulations, Part 50, Section 50.469(a)(3) Concerning Significant Emergency Core Cooling System Evaluation Model Errors/Changes
ML13086A165
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 03/28/2013
From: Michael Mahoney
Division of Operating Reactor Licensing
To: Lieb R
FirstEnergy Nuclear Operating Co
Michael Mahoney, NRR/DORL 415-3867
References
TAC ME8411
Download: ML13086A165 (10)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 28, 2013 Mr. Ray Leib Site Vice President FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Mail Stop A-DB-3080 5501 North State Route 2 Oak Harbor, OH 43449-9760

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 - CLOSURE EVALUATION FOR REPORT PURSUANT TO TITLE 10 OF THE CODE OF FEDERAL REGULA TlONS, PART 50, SECTION 50.46(a)(3) CONCERNING SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERRORS/CHANGES RELATED TO EMERGENCY CORE COOLING SYSTEM BYPASS AND UPPER PLENUM COLUMN WELDMENTS (TAC NO. ME8411)

Dear Mr. Leib:

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.46(a)(3),

First Energy Nuclear Operating Company (the licensee), submitted a report describing two significant errors/changes identified in the emergency core cooling system (ECCS) evaluation model, and an estimate of the effects of the errors/changes on the predicted peak cladding temperature (PCT) for the Davis-Besse Nuclear Power Station, Unit No.1 (DBNPS).

The report was submitted to the U.S. Nuclear Regulatory Commission (NRC or Commission) by letter dated March 16, 2012 (Agencywide Document Access and Management System (ADAMS)

Accession No. ML12076A237), and supplemented by letter dated December 18, 2012 (ADAMS Accession No. ML12353A601).

The NRC staff finds that the report submitted pursuant to 10 CFR 50.46(a)(3), concerning ECCS evaluation model errors/changes pertaining to end of ECCS bypass and column weldments, satisfies 10 CFR 50.46 reporting requirements. The report and supplemental information provided by AREVA NP Inc. enabled the NRC staff to (1) determine that it agrees with the licensee's assessment of the significance of the error. (2) confirm that the evaluation model remains adequate, and (3) verify that the licensee continues to meet the PCT acceptance criterion promulgated by 10 CFR 50.46(b).

The NRC staff does not agree that DBNPS's action is in compliance with the reanalysis requirement set forth in 10 CFR 50.46(a)(3)(ii), which states that the licensee "shall include with the report a proposed schedule for providing a reanalysis or talking other action as may be needed to show compliance with [10 CFR] 50.46 requirements."

The NRC review of the report is complete and TAC No. ME8411 will be closed.

R. Leib - 2 Please contact me at 301-415-3867. if you have any questions.

sin7J, / '

//!/i i~

L Mah~ney.

M' hael

/~-' / I Projec 1._

Plant Licensing Branch 111-2

"

a~~r Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346

Enclosure:

Closure Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 CLOSURE EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION DAVIS-BESSE NUCLEAR POWER STATION, UNIT 1 REPORT DESCRIBING THE NATURE OF AND ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE OF A SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERRORS/CHANGES

1.0 INTRODUCTION

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.46(a)(3),

First Energy Nuclear Operating Company (the licensee), submitted a report describing two significant errors/changes identified in the emergency core cooling system (ECCS) evaluation model, and an estimate of the effects of the errors/changes on the predicted peak cladding temperature (PCT) for the Davis-Besse Nuclear Power Station, Unit No.1 (DBNPS).

The report was submitted to the U.S. Nuclear Regulatory Commission (NRC, Commission) by letter dated March 16, 2012 (Agencywide Document Access and Management System (ADAMS)

Accession No. ML12076A237), and supplemented by letter dated December 18, 2012 (ADAMS Accession No. ML12353A601).

The NRC staff has evaluated the report, along with its supplemental information, and determined that it satisfies the requirements of 10 CFR 50.46(a)(3), but does not satisfy the intent of the reporting requirements, as discussed in the Statements of Considerations (SC) published in the Federal Register (FR), (53 FR 35996; September 16, 1988) for the realistic ECCS evaluations revision of 10 CFR 50.46. The NRC staff review is discussed in the following sections of this closure evaluation.

2.0 REGULATORY EVALUATION

2.1 Requirements Contained in 10 CFR 50.46 Acceptance criteria for ECCS for light-water nuclear power reactors are promulgated at 10 CFR 50.46. In particular, 10 CFR 50.46(a)(3)(i) requires licensees to estimate the effect of any change, or error in an acceptable evaluation model, or in the application of such a model, to determine if the change or error is significant. For the purpose of 10 CFR 50.46, a significant change or error is one which results in a calculated peak fuel cladding temperature difference by more than 50 degrees Fahrenheit ("F) from the temperature calculated for the limiting transient using the last acceptable model, or is an accumulation of changes and errors such that the sum Enclosure

- 2 of the absolute magnitudes of the respective temperature changes is greater than 50 of.

For each change to or error discovered in an acceptable evaluation model, or in the application of such a model, Paragraph (a)(3)(ii) to 10 CFR 50.46 requires the affected licensee to report the change or error and its estimated effect on the limiting ECCS analysis to the Commis!

least annually. If the change or error is significant, the licensee is required to provide this within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 10 CFR 50.46 requirements.

2.2 Additional Guidance Additional clarification concerning the intent of the reporting requirements is discussed in the Federal Register (53 FR 35996):

[Paragraph (a)(3) of section 50.46] requires that all changes or errors in approved evaluation models be reported at least annually and does not require any further action by the licensee until the error is reported. Thereafter, although reanalysis is not required solely because of such minor error, any subsequent calculated evaluation of ECCS performance requires use of a model with such error, and any prior errors, corrected.

The NRC needs to be apprised of even minor errors or changes in order to ensure that they agree with the applicant's or licensee's assessment of the significance of the error or change and to maintain cognizance of modifications made subsequent to NRC review of the evaluation modeL ..

Significant errors require more timely attention since they may be important to the safe operation of the plant and raise questions as to the adequacy of the overall evaluation model ... More timely reporting is required for significant errors or changes ... the final rule revision also allows the NRC to determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements.

The NRC staff considered the discussion in the Federal Register in its evaluation of the error report submitted by the licensee.

3.0 TECHNICAL EVALUATION

The report submitted by the licensee described the effects of two errors/changes which affect the large-break loss-of-coolant accident (LBLOCA) analysis of record. The first item is an error in the determination of the end of ECCS bypass. The second item is a change in the evaluation model (EM) to include the effects of the upper plenum column weldments.

Based on the nature of the reported errors/changes, and on the magnitude of the effects on the PCT calculation, the NRC staff determined that a detailed technical review was necessary.

Based on the regulatory evaluation discussed above, the NRC staff's review was performed to ensure that the NRC staff agreed with the licensee's assessment of the significance of the errors/changes, and to enable the staff to verify that the evaluation model, as a whole, remains adequate. Finally, the NRC staff's review also establishes that the licensee's proposed schedule for re-analysis is acceptable in light of the safety significance of the reported errors/changes.

-3 3.1 Summary of Technical Information in the Report The licensee's report indicated two errors/changes which affect the PCT for LBLOCA analysis.

The first item is an error in the determination of the end of ECCS bypass resulting in a decrease in PCT of 80 OF. The second item is a change to include the effects of the upper plenum column weldments resulting in a PCT increase of 80 OF. The nature of the errors, and the methods used to estimate the effects on the calculated PCT, is discussed in greater detail in the AREVA generic request for additional information (RAI) Response letter "Generic RAI Response to a 30 day 10 CFR 50.46 Report of Significant PCT Change," December 6,2012 (ADAMS Accession No. ML12342A381).

ECCS Bypass Error A mathematical error was identified with the control variables in the energy balance calculations used to determine the complete end of ECCS bypass in LBLOCA applications. As defined in the generic RAI responses from AREVA, the complete end of bypass is achieved when all of the injected ECCS liquid reaches the lower plenum before core reflood analysis begins. When the ECCS flows can condense all of the steam flowing into the upper downcomer region, the end of blowdown has occurred. The complete end of bypass time is the earliest end of blowdown time or the time at which the ECCS flows can condense all the steam flowing into the upper downcomer region. The error in the control variables incorrectly calculated the complete end of bypass time and liquid mass that should have remained in the vessel at the end of blowdown.

Estimation of the Effect of ECCS Bypass in the PCT Calculation The DBNPS has 177 fuel assemblies and uses a raised loop (RL) design nuclear steam supply system (NSSS). The effect of the ECCS bypass error was analyzed explicitly for a 205-fuel assembly plant and applied to a 177-fuel assembly plant. The control variables are common to both 205-fuel assembly and 177 -fuel assembly plants; therefore, the correction is applicable to DBNPS. A blowdown reanalysis was performed using the computer code RELAP5 with corrected bypass variables for a 177-fuel assembly RL plant. The case analyzed was for a 9.536-foot (ft) peak power location at beginning of life. The NRC staff determined that a power shape that uses a high peaking elevation in the core is appropriate because the higher elevation will remain uncovered the longest. This tendency occurs because the reflood begins in the bottom of the core and proceeds to the top, allowing for the longest adiabatic heatup to occur at a higher elevation.

With the correction accounted for, the analysis showed a reduction in the end of ECCS bypass time by roughly two seconds. When the end of bypass time occurred earlier, the amount of ECCS fluid that was not bypassed increased. The ECCS bypass liquid added to the lower plenum caused the bottom of core recovery to occur earlier. This caused the core refill period to become shorter and reflood began earlier. Since core reflood occurred sooner, core cooling began earlier which caused the predicted PCT to decrease.

The licensee reported that the limiting PCT decreased by 96.7 of for the ruptured node and 51.7 of for the unruptured node in a 177-fuel assembly RL plant. The licensee observed that there is a typical two-to-one difference in PCT between ruptured and unruptured fuel cladding segments resulting from the metal to water reaction. The licensee conservatively reported a PCT decrease of 80 of for ruptured segments and 40 of for unruptured segments due to the end of bypass time occurring earlier in the LBLOCA analysis for 177-fuel assembly

-4 RL plant.

The results from a small-break-Ioss-of-coolant accident (SBLOCA) analyses were not impacted since the ECCS bypass is not used for SBLOCA.

Column Weldment Model Change The change in the EM was caused by the inability of the RELAPS model to account for the effects of column weldments over the hot bundle. Sensitivity studies were performed with upper plenum column weldments added to the EM. Column weldments allow a portion of the flow exiting from the core channels underneath them to reach the upper head. This makes them important in determining the temperature of the fluid reaching the upper head.

A LBLOCA RELAPS model was initially developed for a 20S-fuel assembly plant incorporating column weldments (CW) on top of the hot bundle of a 177-fuel assembly plant with a lowered loop (LL) design nuclear steam supply system. Since the column weldment design details for a 177-fuel assembly plant were not readily available when the column weldments were first assessed, a simplified approach used the column weldment model that was developed for the 20S-fuel assembly plant and incorporated it over the 177-fuel assembly LL plant.

The DBNPS has 177 fuel assemblies and uses a RL design. When the effect of column weldments was being analyzed for a RL plant, additional details of the column weldments for a 177-fuel assembly plant were developed.

Estimation of the Effect of Column Weldments in the PCT Calculation The licensee reported that when column weldments modeled for a 20S-fuel assembly LL plant were incorporated over the hot channel in a 177-fuel assembly LL plant, the RELAPS blowdown analysis resulted in an increase in PCT of 3S.6 OF for the peak unruptured fuel cladding segment. Due to the metal to water reaction inside the fuel, there is a two-to-one variation in PCT change between ruptured and unruptured segments. The temperature change of 3S.6 OF was doubled by the licensee to account for both sides of the ruptured node undergoing the exothermic reaction. This resulted in an estimated increase in fuel temperature of 71.2°F for the rupture-limited cladding segments in a 177-fuel assembly LL to account for upper plenum column weldments modeled from a 20S-fuel assembly LL plant.

Typically, the PCT will increase in proportion to the end of blowdown fuel temperature. For this case, there was a favorable shift of rupture time into the blowdown phase causing the rupture segment PCT to increase by 26.2 OF from the computer code BEACH calculation. This is compared to an increase in fuel temperature of 71.2°F from the end of blowdown calculation using RELAPS. BEACH calculated an increase in PCT of 11 .S OF for the unruptured fuel segment. This can be seen in Table 1, below.

When a 177-fuel assembly plant with a RL design (like DBNPS) was being analyzed, details of column weldments of a 177-fuel assembly plant had been developed. These details were used to compare results of the column weldment models between a 177-fuel assembly plant and a 20S-fuel assembly plant. The licensee reported that when the cases were analyzed with column weldments from the 177-fuel assembly plant, the PCT increased by 3 OF for an unruptured fuel segment. This result was doubled for the ruptured fuel segment. This results in a PCT increase of 14.S OF for the unruptured fuel segment and 32.2 OF for the ruptured fuel segment in the LL

- 5 plant as seen in Table 1.

When column weldments were applied to the RL design, the PCT increased by 8.9 of for the unruptured segment. The ruptured segment yielded a decrease in PCT; therefore, it is not the limiting case for a raised loop analysis. These results can be seen in Table 1.

Table 1: Effect of Column Weldments on PCT Ruptured! No. of Fuel No. of Fuel Unruptured Fuel Assemblies CW Assemblies Loop Design Segment modeled for Effect on PCT 177 Lowered loop Unruptured 205 Increase 11.5° 177 Lowered loop Ruptured 205 Increase 26.5° 177 Lowered loop Unruptured 177

  • Increase 14.5°  !

177 Lowered loop Ruptured 177 I Increase 32.2° 177 Raised loop Unruptured 177 Increase 8.9 0 177 Raised loop Ruptured 177 Decrease The licensee used a generic increase of 40 of for unruptured fuel segments and 80 of for ruptured fuel segments for the effect of column weldments. These estimates are bounding and conservative for the analyses completed for DBNPS.

An evaluation of the impact of CW on SBLOCA analyses was performed. It was concluded that SBLOCA analyses are unaffected by the CW modeling because the net flow remains upward during these slower evolving transients.

Reported Results Following the correction for ECCS bypass and the column weldment model change, the current predicted PCT for LBLOCA at DBNPS is 2119 of.

3.2 Summary of Staff Evaluation In its evaluation, the NRC staff reviewed (1) the approach used to estimate the effects of the ECCS bypass error and the effects of upper plenum column weldments, (2) the estimated effect of both errors/changes, and (3) the licensee's proposal to not perform a reanalysis in consideration of the approach used to estimate the effects of the errors/changes. As discussed in the following paragraphs, the NRC staff determined that the licensee's estimates are acceptable; however, the NRC staff does not agree with the proposal to not perform a reanalysis.

To estimate the effects of the ECCS bypass error, the licensee analyzed the effect of correcting the control variables in the energy balance equation used to determine the complete end of bypass time in LBLOCA applications. The effect was analyzed explicitly for a 205-fuel assembly plant and applied to 177-fuel assembly plants. The control variables were common to both 205 fuel assembly plants and 177-fuel assembly plants; therefore, the correction is applicable to

-6 DBNPS as a 177-fuel assembly plant.

To estimate the effects of upper plenum column weldments, the licensee included column weldments over the hot channel in a model of a 205-fuel assembly plant. This model was incorporated over the hotel channel in a 177-fuel assembly lowered loop plant and a RELAP5 blowdown analysis was completed. Sensitivity studies were performed using column weldments modeled for a 177-fuel assembly raised loop plant. The effects of the studies showed the generic estimate of the effect of column weldments was conservative for RL plants.

The licensee estimated the effect of the ECCS bypass error to be a decrease in PCT of 80 oF.

The licensee estimated the effect of upper plenum column weldments to be an increase in PCT of 80 of. The current predicted PCT for LBLOCA at DBNPS is 2119 of.

As stated in 10 CFR 50.46(a)(3)(ii), the licensee "shall include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with [10 CFR] 50.46 requirements." As described in Section 2.0, above, the SOC explain further that "the final rule revision also allows the NRC to determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements."

In the RAI issued by the NRC on October 15, 2012 (ADAMS Accession No. NIL12279A037), the NRC staff requested that the licensee "justify not providing a schedule for reanalysis or taking other action to show compliance with 10 CFR 50.46." The RAI response was provided by the licensee on December 18, 2012 (ADAMS Accession No. ML12353A601).

The licensee indicated in the response that both the error and change to the ECCS evaluation model that were presented in the 10 CFR 50.46 30-day report were analyzed in detail and that the impact of both items does not result in challenge to the 10 CFR 50.46(b) acceptance criteria.

The licensee also concluded that the evaluation model is considered acceptable since the error and change have been analyzed and there are no other known errors or changes in the model at this time.

The NRC staff determined that the PCT error evaluations are supported by explicit analysis using the Babcock & Wilcox plant ECCS evaluation model, and the error-adjusted LBLOCA PCTs for DBNPS remain below the 10 CFR 50.46(b) regulatory acceptance criteria.

NRC Information Notice 97-15. Supplement 1, "Reporting of Errors and Changes in Large Break/Small-Break Loss-of-Coolant Evaluation models of Fuel Vendors and Compliance with 10 CFR 50.46(a)(3)" describes, among other things, a prior issue at another plant. in which a proposed schedule for reanalysis was not included with a report submitted pursuant to 10 CFR 50.46(a)(3). To address the NRC staff review, the licensee indicated that they would provide such a reanalysis. The NRC staff is unable to determine that either DBNPS's report or RAI response addresses the reanalysis requirement.

In summary, the NRC staff reviewed the licensees report estimating the effect of the ECCS bypass error and column weldments on the LBLOCA analyses for DBNPS. Since the evaluation included explicit analyses of the ECCS bypass error and column weldments in the evaluation model, the NRC staff concluded that the error estimates were acceptable. The NRC staff does not agree that DBNPS action is in compliance with the reanalysis requirement set forth in 10 CFR 50.46(a)(3)(ii). "Other action" should not be interpreted as no action.

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4.0 CONCLUSION

Based on the considerations discussed above, the NRC staff finds that the report submitted pursuant to 10 CFR 50.46(a)(3), concerning ECCS evaluation model errors/changes pertaining to end of ECCS bypass and column weldments satisfies 10 CFR 50.46 reporting requirements.

The report and supplemental information provided by AREVA NP Inc. enabled the NRC staff to (1) determine that it agrees with the licensee's assessment of the significance of the error, (2) con'firm that the evaluation model remains adequate, and (3) verify that the licensee continues to meet the PCT acceptance criterion promulgated by 10 CFR 50.46(b).

However, the NRC staff does not agree that DSNPS's action is in compliance with the reanalysis requirement set forth in 10 CFR 50.46(a)(3)(ii), which states that the licensee "shall include with the report a proposed schedule for providing a reanalysis or talking other action as may be needed to show compliance with [10 CFR] 50.46 requirements."

Principal Contributors: A. Guzzetta, NRR B. Parks, NRR Date of issuance: March 28, 2013

R. Leib -2 Please contact me at 301-415-3867, if you have any questions.

Sincerely, 1 RAJ Michael Mahoney, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346

Enclosure:

Closure Evaluation cc: Listserv DISTRIBUTION:

PUBLIC AGuzzetta, NRR LPL3-2 R/F RidsNrrDorlLpl3-2 Resouce RidsAcrsAcnw_MailCTR Resource RidsNrrPMDavis-Besse Resource RidsNrrSrxb Resource RidsNrrLASRohrerResource RidsRgn1 MailCenter Resource BParks, NRR RidsNrrDorlDpr Resource ADAMS A ccesslon N0.: ML13086A165 *8Iy Merno Daet d Marc h 12, 2013 OFFICE LPL3-2/PM NAME IVIMahoney LPL3-2/LA SRohrer DSS/SRXB/BC CJackson LPL3-2/BC JBowen LPL3-21~1 MMahon DATE 3/28/13 3/27/13 3/12/13* 3/28/13 3/28/13 OFFICIAL RECORD COPY

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