ML13338A345

Closure Evaluation for Report Concerning Significant Emergency Core Cooling System Evaluation Model Errors/Changes

ML13338A345
Person / Time
Site:	Oconee
Issue date:	12/05/2013
From:	Richard Guzman Plant Licensing Branch II
To:	Batson S Duke Energy Carolinas
Guzman R
References
TAC ME9119, TAC ME9120, TAC ME9121
Download: ML13338A345 (11)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 5, 2013 Mr. Scott Batson Site Vice President Oconee Nuclear Station Duke Energy Carolinas, LLC 7800 Rochester Highway Seneca, SC 29672-0752

SUBJECT:

OCONEE NUCLEAR STATION, UNITS 1,2, AND 3 - CLOSURE EVALUATION FOR REPORT CONCERNING SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERRORS/CHANGES RELATED TO EMERGENCY CORE COOLING SYSTEM BYPASS AND UPPER PLENUM COLUMN WELDMENTS (TAC NOS. ME9119, ME9120, AND ME9121)

Dear Mr. Batson:

Pursuant to 10 CFR 50.46(a)(3), Duke Energy Carolinas, LLC, the licensee for Oconee Nuclear Station, Units 1,2, and 3 (ONS), 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. The report was submitted by letter dated March 9, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12073A354), as supplemented by letters dated December 7,2012 (ADAMS Accession No. ML12348A055), April 4, 2013 (ADAMS Accession No. ML13102A033) , and October 10, 2013 (ADAMS Accession No. ML13308B902).

The Nuclear Regulatory Commission (NRC) staff finds that the report submitted 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 peak cladding temperature acceptance criterion promulgated by 10 CFR 50.46(b).

S. Batson -2 A copy of the NRC staff evaluation is enclosed.

If you have any questions regarding this letter, please contact me at 301-415-1030 or via e-mail at Richard.Guzman@nrc.gov.

Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287

Enclosure:

NRC Staff Evaluation

cc w/encl: Distribution via Listserv

\.f.r>1'I REGu~ UNITED STATES

~~c, .l>o?, NUCLEAR REGULATORY COMMISSION

! ~ WASHINGTON, D.C. 20555*0001

~

~ ~

°

'v. ~

1'" ¥>o

• • • ic~ CLOSURE EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 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 10 CFR 50.46(a)(3), Duke Energy Carolinas, LLC, the licensee for Oconee Nuclear Station, Units 1, 2, and 3 (Oconee), submitted a report describing two significant errors/changes identified in the emergency core cooling system evaluation model, and an estimate of the effects of the errors/changes on the predicted peak cladding temperature (PCT). The report was submitted by letter dated March 9, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12073A354), as supplemented by letters dated December 7,2012 (ADAMS Accession No. ML12348A055), April 4, 2013 (ADAMS Accession No. ML13102A033), and October 10, 2013 (ADAMS Accession No. ML13308B902).

The licensee's RAI response letter dated April 4, 2013, also referred to an AREVA letter submitted to the NRC on March 28, 2013 (ADAMS Accession No. ML13091A075).

The U.S. Nuclear Regulatory Commission (NRC) staff has evaluated the report, along with its supplemental information, and determined that it satisfies the reporting requirements of 10 CFR 50.46(a)(3), and also the intent of the reporting requirements, as discussed in the statement of considerations published on September 16, 1988, in the Federal Register (FR), for the realistic Emergency Core Cooling System (ECCS) evaluations revision of 10 CFR 50.46 (53 FR 35996).

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 ECCSs for light water nuclear power reactors are found in 10 CFR 50.46. In particular, 10 CFR 50.46(a)(3)(i) requires licensees to estimate the effect of any change to 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 (OF) from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 OF.

- 2 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 nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually. If the change or error is significant, the licensee is required to provide this report 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 Section 50.46 requirements.

2.2 Additional Guidance Additional clarification concerning the intent of the reporting requirements is discussed in the Federal Register (53 Fed. Reg. 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 that 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 (CWs).

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 agrees with the licensee's assessment of the significance of the errors/changes, and to enable the NRC staff to verify that the evaluation model, as a whole, remains adequate.

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

The first item is an error in the determination of the end of EGGS bypass resulting in a decrease in PGT of 80 oF. The second item is a change to include the effects of the upper plenum GWs resulting in a PGT increase of 80 of. The nature of the errors, and the methods used to estimate their effects on the calculated PGT is discussed in greater detail in the AREVA letter submitted to the NRG on March 28, 2013.

ECCS Bypass Error/Change A mathematical error was identified with the control variables in the energy balance calculations used to determine the complete end of EGGS bypass in LBLOGA applications. As defined in the AREVA letter submitted on March 28, 2013, the complete end of bypass is achieved when all of the injected EGGS liquid reaches the lower plenum before core reflood analysis begins.

When the EGGS flows can condense all of the steam flowing into the upper downcomer region, the end of blowdown has occurred. AREVA further defines the complete end of bypass time as the earliest end of blowdown time or the time at which the EGGS 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 Oconee has 177 fuel assemblies and uses a Lowered Loop (LL) design Nuclear Steam Supply System (NSSS). The effect of the EGGS bypass error was identified from analyses of a 205 fuel assembly plant. The control variables in the evaluation model are common to both 205 fuel assembly and 177 fuel assembly plants; therefore, the licensee determined that the error applies also to the Oconee EGGS evaluation. A blowdown reanalYSis was performed using RELAP5 (Reference 1) with corrected bypass variables for a 177 fuel assembly LL plant. The case analyzed was for a 2.506-ft peak power location at beginning of life.

With the error corrected, the analysis showed a reduction in the end of EGGS bypass time by roughly 2 seconds. When the end of bypass time occurred earlier, the amount of ECCS fluid that was not bypassed increased. The previously bypassed EGGS liquid added to the lower plenum and caused the bottom of core recovery to occur earlier. This caused the core refill period to become shorter, and reflood began sooner. Since core reflood occurred sooner, core cooling began earlier, which caused the predicted PGT to decrease.

The licensee reported that the limiting PGT decreased by 90.3 OF for the ruptured node and 45.4 OF for the unruptured node in a 177 fuel assembly LL plant. The licensee conservatively reported a PGT decrease of 80 OF for ruptured segments and 40 OF for unruptured segments as a result of the end of bypass time occurring earlier in the LBLOGA analysis for 177 fuel assembly LL plant.

The results from small-break loss-of-coolant accident (SBLOGA) analyses were not impacted since the same EGGS bypass modeling is not used for SBLOGA.

-4 Column Weldment Model Error/Change The change in the EM was caused by the inability of the RELAPS model to account for the effects of CWs over the hot bundle. Sensitivity studies were performed with upper plenum CW added to the EM. Column weldments are also known as control rod guide tube housings. They support the control rod and 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.

The change estimate of the column weldment issue was developed using an iterative process that adapted a 20S-FA plant's model to the 177-FA plant design. A LBLOCA RELAPS model was initially developed for a 20S fuel assembly plant. Then, the CW design from the 20S-FA plant was modeled on top of the hot bundle of a 177 fuel assembly plant with a LL design. This simplified approach was used because the CW design details for a 177 fuel assembly plant were not readily available when the CW modeling issue was first assessed. When the effect of CWs was being analyzed for a Raised Loop (RL) plant, additional details of the CWs for a 177 fuel assembly plant had been developed.

Estimation of the Effect of Column Weldments in the PCT Calculation The licensee reported that when CWs modeled for a 20S fuel assembly plant were incorporated over the hot channel in a 177 fuel assembly LL plant, the RELAPS blowdown analysis resulted in an increase in end-of-blowdown fuel temperature of 3S.6 of for the peak unruptured fuel cladding segment. The licensee estimated this temperature increase to be 40 of. The fuel temperature increase observed at end-of-blowdown is generally maintained throughout the refill period because the fuel heats up adiabatically during this phase. The actual temperature increase at beginning-of-core recovery was analyzed to be 30.2 of, indicating that estimation of 40 of was conservative.

Typically, the PCT will increase in proportion to the end-of-blowdown fuel temperature. The licensee notes that 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. Therefore, the licensee estimated that the ruptured cladding segment would experience an 80 of increase in PCT, estimated based on a 40 of increase in end-of-blowdown temperature.

When the licensee analyzed this case using BEACH (Reference 2), there was a favorable shift of rupture time into the blowdown phase causing the ruptured segment PCT to increase by 26.2 of from the BEACH calculation. BEACH calculated an increase in PCT of 11.S OF for the unruptured fuel segment. (This is compared to an increase in fuel temperature of 3S.6°F from the end of blowdown calculation using RELAPS.)

Since analysis using a LL design with a 177 fuel assembly plant was not completed, the geometric properties between a 20S fuel assembly plant and 177 fuel assembly plant were compared. When CWs are explicitly modeled in the upper plenum model of the LBLOCA analyses, the structures impede coolant flow entering the top of the core after the core How reverses direction. The crossflow into the lower CW holes and slots from the upper plenum is restricted in the reverse-flow direction due to the form losses. The hole and slot sizes in the lower CWs are identical between the 20S fuel assembly and 177 fuel assembly plant designs;

~ 5 therefore, the controlling resistances and flow areas for crossflow into or out of the lower CW are the same.

Even though the hole and slot sizes in the lower CWs are identical, the flow areas inside the CWs for the 177 fuel assembly plants are 7 percent smaller than those in a 205 fuel assembly plant. This difference is related to the 15x15 versus 17x17 fuel bundle arrangements and the 177 fuel assembly plant having a different number of control rods than the 205 fuel assembly plant.

Analyses for a 177 fuel assembly CW plant with a LL design were not explicitly completed. A sensitivity study was done using a 177 fuel assembly CW in a RL design. The results of this study show an increase in PCT of 3 of when compared to a 205 fuel assembly CW with a LL design. The PCT change was due to the difference in flow area inside the CW between the 177 and 205 fuel assembly plants.

Compared to a case with only the corrected ECCS bypass calculation, it was observed that modeling the 177 fuel assembly CW in a RL plant increased PCT by 8.9 OF for an unruptured segment. For the LL plant, the PCT increased 11 .5 of for an un ruptured segment, after modeling a simplified version of the 205 fuel assembly CW, compared to a case with only the corrected ECCS bypass calculation.

The licensee used a generic increase of 40 of for unruptured fuel segments and 80 OF for ruptured fuel segments for the effect of CWs. The largest increase in PCT for the LL design with ruptured fuel segments was explicitly analyzed to be 26.2 of when CW were included in the analyses. The generic increase in PCT of 80 of that the licensee applied to rupture limited LL plants is considered bounding and conservative because there is significant margin over the calculated PCT increase of 26.2 of. The 3 of increase in PCT from analyzing CW in a 177 fuel assembly plant is also well within the margin.

An evaluation of the impact of CW on SBLOCA analyses was performed by the licensee. 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 CW model change, the current predicted PCT for LBLOCA at Oconee is as follows:

Unit I Fuel TYlle PCT eF) l r

1 Mk~B11 2035 l=_1 Mk-B-HTP 2020 I 2 Mk-B-HTP 1913 3 Mk-B11 2035 t_3__ ' Mk-B-HTP I 2020

-6 3.2 Summary of NRC 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 CWs, (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 the licensee's estimates of the error and change are acceptable.

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 identified from analyses of a 205 fuel assembly plant. The control variables are common to both 205 fuel assembly plants and 177 fuel assembly plants; therefore, the correction is applicable to Oconee. The NRC staff determined that this estimate was acceptable because explicit analysis was completed to evaluate the effect of correcting the control variables.

To estimate the effects of upper plenum CWs, the licensee included CWs over the hot channel in a model of a 205 fuel assembly plant. This model was incorporated over the hot channel in a 177 fuel assembly LL plant and a RELAP5 blowdown analysis was completed. Sensitivity studies were performed using CW modeled for a 177 fuel assembly RL plant. The effects of the studies showed the generic estimate of the effect of CW was conservative for LL plants. The NRC staff determined that this estimate was acceptable because the effect of including upper plenum column weldments was explicitly analyzed.

The licensee estimated the effect of the ECCS bypass error to be a decrease in PCT of 80 OF for ruptured fuel segments and 40 OF for unruptured fuel segments. The licensee estimated the effect of upper plenum CW to be an increase in PCT of 80 OF for ruptured fuel segments and 40 OF for unruptured fuel segments. The current predicted PCT for LBLOCA at Oconee is as follows:

FU~it--

---~----

FueIT~~e PCT (OF)

Mk-B11 2035 r--1-~~ Mk-B-HTP 2020

~~-----~~~

2 I Mk-B-HTP 1913 3 I Mk-B11 2035 C~ __li1~j~ljlf_ ~~ 2020 The NRC reviewed the current estimated PCTs at the Oconee Units, and determined that the reported PCTs continue to remain below the 10 CFR 50.46(b)(1) acceptance criterion, and therefore, the reported PCTs remain acceptable.

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 the Regulatory Evaluation, the statements of consideration explain further that the final rule revision also allows the NRC to

-7 determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements."

In the Request for Information (RAI) issued on October 15, 2012 (ADAMS Accession No. ML12278A273), 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." In the RAI Response submitted on December 7,2012, the licensee indicates 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 concludes that the evaluation model is considered adequate 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 B&W plant ECCS evaluation model, and the error adjusted LBLOCA PCTs for ONS remain below the 10 CFR 50.46(b) regulatory acceptance criteria.

The NRC staff issued another RAI on September 12, 2013 (ADAMS Accession No. ML13309A446) asking the licensee to justify how the generic analysis for the B&W plant ECCS evaluation model satisfies the requirement, in 10 CFR 50.46(a)(1 )(i), to calculate ECCS cooling performance "in accordance with an acceptable evaluation model". The RAI states:

In light of the presently reported, significant, estimated effects of errors and changes, explain how the present ECCS cooling performance has been calculated in accordance with an acceptable evaluation model, such that any other action, as provided in 10 CFR 50.46(a)(3), has been taken to show compliance with 10 CFR 50.46 requirements, including those contained in 10 CFR 50.46(a)(1).

The licensee uses an evaluation model that conforms to 10 CFR 50, Appendix K, to evaluate ECCS performance. The use of an Appendix K-based evaluation model leads to a conservative estimate of PCT. Included in the licensee's RAI response submitted on October 10, 2013, is clarification that the current evaluation model (with the errors corrected) concluded the actual net PCT would decrease, but the licensee reported a net change of zero in PCT. Therefore, the existing results remain conservative compared to the use of an evaluation model that both conforms to Appendix K and explicitly corrects the reported changes and errors. The NRC staff concludes that the licensee's explicit analysis of both significant errors paired with the conservative estimate of PCT using an already conservative evaluation model is acceptable and satisfies all requirements of 10 CFR 50.46.

In summary, the NRC staff reviewed the licensees report and supplemental information estimating the effect of the ECCS bypass error and CWs on the LBLOCA analyses for Oconee.

Since the evaluation included explicit analyses of the ECCS bypass error and CWs in the evaluation model, the NRC staff concluded that the error estimates were acceptable.

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

- 8 to end of ECCS bypass and CW satisfies the intent of the 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 of 10 CFR 50.46(b).

5.0 REFERENCES

1. BAW-10164P-A, Revision 6, "RELAP5/MOD2-B&W - An Advanced Computer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis," June 2007.

2. BAW-10166P-A, Revision 5, "BEACH - A Computer Program for Reflood Heat Transfer during LOCA," November 2003.

ML13338A345

• via memo dated 11/25/13 OFFICE DORULPL2-1/PM DORULPL2-1/LA DSS/SRXB/BC DORULPL2-1/BC DORULPL2-1/PM NAME RGuzman SFigueroa CJackson* RPascarelli RGuzman DATE 12/5/13 12/5/13 11/25113 12/5/13 l12/5/13