ML20149K694

Response to Request for Additional Information, to Proposed License Amendment Requests Addition of Analytical Methodology to the Core Operating Limits Report for a Small Break Loss of Coolant Accident

ML20149K694
Person / Time
Site:	Surry, North Anna
Issue date:	05/28/2020
From:	Gerald Bichof Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML20149K693	List: ML20149K694
References
20-149
Download: ML20149K694 (69)

Text

PROPRIETARY INFORMATION-WITHHOLD UNDER 10 CFR 2.390 VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 May 28, 2020 Serial No.:

NRA/DEA:

10 CFR 50.90 20-149 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001 Docket Nos.:

R2 50-338/339 50-280/281 License Nos.: NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA AND SURRY POWER STATIONS UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUESTS DPR-32/37 ADDITION OF ANALYTICAL METHODOLOGY TO THE CORE OPERA TING LIMITS REPORT FOR A SMALL BREAK LOSS OF COOLANT ACCIDENT {SBLOCA)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION AND ANALYSIS ERROR CORRECTION By letters dated July 12, 2018 and July 31, 2018 [Agency wide Document Access and Management System (ADAMS) Accession Nos. ML18198A118 and ML18218A170, respectively], Virginia Electric and Power Company (Dominion Energy Virginia) submitted license amendment requests (LARs) to revise the Technical Specifications (TS) for North Anna and Surry Power Stations (NAPS and SPS) Units 1 and 2, respectively, to allow each station to implement a fuel vendor-independent evaluation model for analyzing hypothetical small break loss-of-coolant accidents.

As part of its review of the LARs, the U. S. Nuclear Regulatory Commission (NRC) staff conducted an audit at the Dominion Energy Virginia corporate offices in Glen Allen, Virginia, from October 1-4, 2018. During the course of the audit, the NRC staff presented Dominion Energy Virginia staff with a detailed list of issues requiring further information.

An audit summary report was issued on October 25, 2018.

The NRC staff completed the initial review of the LARs and of information provided during the audit and determined that additional information was needed to complete their evaluation. An NRC request for additional information (RAI) was provided in a letter dated February 8, 2019 (ADAMS Accession No. ML19032A055) and Dominion Energy Virginia's response to the RAI was provided in a letter dated July 9, 2019 (S/N 19-083)

(ADAMS Accession No. ML19196A109). contains information that is being withheld from public disclosure under 10 CFR 2.390. Upon separation from Attachment 1, this page is decontrolled.

Serial No.20-149 Docket Nos.: 50-338/339/280/281 NAPS-SPS SBLOCA RAI Response Set 2 Page2 of4 As part of its review of the LARs, staff from the NRC conducted a supplemental audit at the Dominion Energy Virginia corporate offices in Glen Allen, Virginia, from January 22-24, 2020. As a result of its review and the interactions at the audit, the NRC staff has determined that additional information is needed to complete their evaluation. In a letter dated April 1, 2020 the NRC provided specific RAls. Attachments 1 and 2 provide Dominion Energy Virginia's response to the RAls. contains information proprietary to Framatome and is therefore supported by an affidavit signed by the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390.

Accordingly, it is respectfully requested that the proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390. A redacted, non-proprietary version of the information is provided in Attachment 2. The affidavit is provided in Attachment 3.

The information provided in this letter does not affect the conclusions of the significant hazards considerations or the environmental assessments included in the July 12, 2018 and July 31, 2018 LARs.

If there are any questions or if additional information is needed, please contact Mrs. Diane E. Aitken at (804) 273-2694.

Sincerely, Gerald T. Bischof 6Jd1~12 Senior Vice President -

uclear Operations & Fleet Performance COMMONWEAL TH OF VIRGINIA

)

)

COUNTY OF HENRICO

)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T. Bischof, who is Senior Vice President - Nuclear Operations & Fleet Performance of Virginia Electric and Power Company.

He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

~

AcknOWledged befo<e me u,;, ZB day of~*

2020.

MyCommlsslonExp;<es:'JIYpvVl 5/L :2,c!Z

~'z:c~

Notary Public Commitments made in this letter: None 1

DIANE E. AITKEN NOTARY PUBLIC REG. ffl63114 COMMONWEALTHOF\\IIRGNA

~

WCOMM1SSION EXPIRES MARCH 31, 2022.*

Attachments:

Serial No.20-149 Docket Nos.: 50-338/339/280/281 NAPS-SPS SBLOCA RAI Response Set 2 Page 3 of 4

1. RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION SUPPLEMENT 1 REGARDING FVI SBLOCA (PROPRIETARY)

2. RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION SUPPLEMENT 1 REGARDING FVI SBLOCA (NON-PROPRIETARY)

3. FRAMATOME AFFIDAVIT FOR WITHHOLDING PROPRIETARY INFORMATION

Serial No.20-149 Docket Nos.: 50-338/339/280/281 NAPS-SPS SBLOCA RAI Response Set 2 Page 4 of 4 cc:

U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector North Anna Power Station NRC Senior Resident Inspector Surry Power Station Ms. Karen R. Cotton Gross NRC Project Manager U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. G. Edward Miller NRC Senior Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. Marcus Harris Old Dominion Electric Cooperative Innsbrook Corporate Center, Suite 300 4201 Dominion Blvd.

Glen Allen, Virginia 23060 State Health Commissioner Virginia Department of Health James Madison Building - 7th Floor 109 Governor Street Room 730 Richmond, Virginia 23219 Serial No.20-149 Docket Nos. 50-338/339/280/281 RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION SUPPLEMENT 1 REGARDING FVI SBLOCA (NON-PROPRIETARY)

Virginia Electric and Power Company (Dominion Energy Virginia)

North Anna and Surry Power Stations Units 1 and 2

RAI 2 S1 Request:

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 2 of 61 The response to request for additional information (RAJ) 2 refers to calculated results from the FVJ-SBLOCA [fuel-vendor independent small-break Joss-of-coolant accident] and ASTRUM [Automated Statistical Treatment of Uncertainty Method] large-break loss-of-coolant accident methods as demonstrating that breaks in a size range between 10% of the cold leg crosssectional area and 1. 0 ft2 are adequately addressed. However, the response appears to be based on extrapolation of FVJ-SBLOCA and ASTRUM results into a range of the break spectrum (i.e., from approximately 0.4125 - 1.0 ft2) where no calculations for North Anna or Surry have been reported with either evaluation model.

Furthermore, the U.S. Nuclear Regulatory Commission (NRG) staff observed in Section 15.3.1.5.1 of the North Anna Updated Final Safety Analysis Report (UFSAR) that "The NO TRUMP computer code is used for Joss-of-coolant accidents due to small breaks less than one square foot." A similar description exists in Section 14.5.2.2 of the Surry UFSAR.

The UFSAR descriptions reviewed by the staff do not appear to define any portion of the postulated LOCA break size range as inherently non-limiting.

Paragraph 10 CFR 50.46(a)(1)(i) requires that "EGGS [Emergency Core Cooling System]

cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated Joss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated Joss-of-coolant accidents are calculated." As such, please address the following RA/s:

a) Compare the proposed range of small breaks for the FVI-SBLOCA methodology to the analyzed range of breaks for the current SBLOCA evaluation model. Please provide justification if adoption of the FVI SBLOCA methodology would result in a reduction to the analyzed break spectrum as compared to the current evaluation model.

b) Considering that the predicted limiting break size may in general be a function of, among other things, the evaluation model being used, please provide any evidence, such as calculated results using the FVI-SBLOCA and ASTRUM methods, demonstrating that these evaluation models will not predict limiting results for the LOCA event in the range of break sizes between 0.4125-1.0 ft2.

Response

2 S1.a Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 3 of 61 The FVI-SBLOCA methodology would result in an increase in the explicitly analyzed small-break spectrum as compared to the current SBLOCA evaluation model (NOTRUMP). The following table provides the range of break sizes analyzed with NOTRUMP compared to that analyzed with FVI-SBLOCA for both North Anna and Surry.

The North Anna analyzed spectrum was taken from Reference [2S1.a-1] for NOTRUMP and Reference [2S1.a-2] for FVI-SBLOCA. The Surry analyzed spectrum was taken from Reference [2S1.a-3] for NOTRUMP and Reference [2S1.a-4] for FVI-SBLOCA.

Analyzed Small-Break Spectrum North Anna Surry NOTRUMP 1.5" to 5.189" diameter 1.5" to 5.50" diameter FVI-SBLOCA 1.00" to 8.70" diameter 1.00" to 8.70" diameter

References:

[2S1.a-1]

[2S1.a-2]

[2S1.a-3]

[2S1.a-4]

North Anna UFSAR, Revision 55, Section 15.3.1, "Loss of Reactor Coolant From Small Ruptured Pipes or From Cracks in Large Pipes That Actuates the Emergency Core Cooling System (Small Break. Loss-of-Coolant Accident)."

ANP-3467P, Revision 0, "North Anna Fuel-vendor Independent Small Break LOCA Analysis," May 2018.

Surry UFSAR, Revision 51, Section 14.5.2, "Loss of Reactor Coolant From Small Ruptured Pipes or From Cracks in Large Pipes, Which Actuates Emergency Core Cooling System (Small Break Loss-of-Coolant Accident Analysis)."

ANP-3676P, Revision 0, "Surry Fuel-vendor Independent Small Break LOCA Analysis," July 2018.

2.S1.b Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 4 of 61 The evaluation model is a tool to support the demonstration of acceptable ECCS performance. While the specific break size which is identified as limiting is somewhat dependent on the evaluation model, the metric from the analysis application is solely the value used to compare to the criteria (e.g. maximum PCT). Those calculated values assure the performance of the plant is bounded across the spectrum of possible LOCAs.

The physics of a LOCA and actual plant performance are independent of the evaluation model.

The LOCA break spectrum can be divided into three general regions based on the expected physical phenomena: Small Break LOCA (SBLOCA) when total break area is approximately :5 0.4 fl2 ( ~10% of the cold leg pipe area); Large Break LOCAs (LBLOCA) when the total break area is, depending on plant and ECCS design, greater than approximately 1.6 fl2 to 2.4 fl2 (~40 to 60% of the cold leg pipe area); and Intermediate Break (IBLOCA) when the break area is in between SBLOCA and LBLOCA. Evaluation models also break the spectrum into sub-regions according to the phenomena and the models necessary to accurately capture the specific phenomena. ((

)) As was discussed above, the purpose of these methodologies is not to predict a PCT for every break size or location possible at the plant, but rather to demonstrate acceptable performance of a plant's ECCS design.

Accordingly, the methodologies limit the analysis to those types of breaks that present the most challenge to the ECCS. As a result, portions of the break spectrum and break locations, such as the hot leg, are commonly not explicitly analyzed.

With the evolution to more realistic analyses, which utilize statistics to determine the limiting values based on desired confidences, and with increased computing capabilities, the explicitly analyzed break spectrum and number of cases has increased. However, the physics of the event remain constant and conclusions can be made independent of the EM used. The phenomena and evolution of breaks within the intermediate range make them fundamentally less limiting than a SBLOCA or a LBLOCA. There will be a transition which starts at the boundaries of the ranges, but in general, the important thermal-hydraulic phenomena for the three regions can be outlined as follows:

LBLOCA o The core flow becomes negative upon break initiation and it becomes almost completely voided within a few seconds

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 5 of 61 o The rapid system depressurization allows for the full complement of the ECCS to be available shortly after the start of the transient o The temperature of the majority of the fuel rod will be substantially higher than the saturation temperature almost immediately o There is two-phase flow upwards out of the downcomer to the break o The reverse flow from the reactor vessel results in ECCS bypass such that a major portion of the accumulator water entering the DC from the intact legs is swept back out the break o The core remains essentially dry throughout the blowdown and refill phases of the LOCA o

Depending on the plant, the PCT can occur during blowdown, refill, or reflood SBLOCA o

Break opening does not empty the core; the core flow remains positive throughout the transient and core mass reduction is due to boil off o The system depressurizes more slowly and stabilizes at saturation pressure. Following loop seal clearing, the break flow becomes two phase and the system continues to depressurize o

The significant core heat-up occurs only after loop seal clearing o

The water level in the reactor vessel is several feet above the bottom of the core barrel, creating a water seal at the bottom of the downcomer o

Only a limited amount of steam can go back through the downcomer towards the break allowing almost all of the ECG injected in the intact cold legs to proceed to enter the downcomer and provide core cooling The intermediate breaks demonstrate characteristics of both regimes. Consistent with LBLOCA, the primary system depressurizes relatively fast and the core heat up begins early in the blowdown phase. Consistent with SBLOCA, the core flow remains positive and a water seal at the bottom of the downcomer remains such that all of the ECCS is available for cooling. The PCT trend as a function of break size depends on which phenomena dominate.

For the smaller IBLOCAs, the level remains in the core. The liquid flashes due to the depressurization and the voiding causes a substantial level swell. As a result, substantial cooling occurs in the upper regions of the core thereby preventing any substantial heat up. The heat up is terminated by the accumulator injection.

As the break size within the IBLOCA range increases, the amount of water in the reactor vessel following loop seal clearing decreases. The amount of water in the core therefore decreases, but the flashing and level swell increase due to the faster depressurization.

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 6 of 61 Similarly, the accumulator injection occurs earlier. The interplay results in the potential for decreases or increases in PCT with increasing break size.

At a certain break size, the water level falls below the bottom of the active core. When this occurs, there is a period of complete core dryout prior to accumulator injection. The duration of the core dryout period increases with increasing break size and accordingly the PCT increases. The transition to the classical LBLOCA occurs depending on plant and ECCS design, at a total break area between 40% and 60% of the cold leg pipe area, when the level falls below the bottom of the core barrel allowing reverse flow of liquid and steam back through the downcomer to the break. ((

11 In conclusion, the phenomena associated with the IBLOCA event are such that it is less challenging to the LOCA criteria than SBLOCA or LBLOCA and therefore a plant does not need an explicit analysis in this regime in order to conclude that the ECCS will perform acceptably.

References:

{2S1.b-1]

))

EMF-2103P-A-003, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," June 2016

RAI 6 S1 Request:

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 7 of 61 The licensee's response to RA/ 6 provides a qualitative, historical review of background material concerning the analysis of reactor coolant pump trip timing for the SBLOCA event. Much of the historical analyses and derivative insights discussed therein originated in response to the 1979 accident at Three Mile Island (TM/), Unit 2. As discussed during the regulatory audit in October 2018, the post-TM/ analysis relied upon computer codes developed during the 1960s and 70s with significantly simpler modeling practices than modem codes (e.g., 10-20 fluid nodes, simplified field equations). The post-TM/ analysis also focused upon smaller break sizes (e.g., 2-4 inches), as opposed to the larger range of small breaks discussed in RA/ 6 (i.e., 5 inches and larger) that contemporary analyses show have the potential to be limiting for many Pressurized Water Reactors (PWRs) ll H The licensee's response to RA/ 6 did not describe sensitivity calculations applicable to North Anna or Surry in the range of reduced reactor coolant pump trip delay times and break sizes of interest to RA/ 6.

Based upon its review of the licensee's response to RA/ 6, the NRG staff concluded that the concerns expressed in RA/ 6 that trip times less than 5 minutes could be both (1) more limiting than the cases analyzed by the licensee for break sizes 5 inches and larger and (2) more likely than the cases analyzed by the licensee had not been adequately addressed. To probe the significance of the issue, the NRG staff performed preliminary sensitivity studies using the TRACE thermal-hydraulic code that considered ((

)).

During a regulatory audit held on January 22-24, 2020, the NRG staff audited a calculation report containing sensitivity analyses for reduced reactor coolant pump trip times using the EMF-2328 methodology that appeared to show similar results to the staff's calculations using TRACE. The EMF-2328 sensitivity results, which assumed ((

11-Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 8 of 61 Paragraph 10 CFR 50.46(a)(1)(0 requires that "EGGS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated." As described above, the sensitivity studies illustrate that such assurance has not been provided, because the presently limiting break size indicates the potential to return more severe results when the reactor coolant pump (RCP) trip timing sensitivity is also considered. As such, please provide the following additional information:

Adequately address the potential for ((

J] for North Anna and Surry that has the potential both to produce more severe consequences and to be more likely than the cases analyzed by the licensee ((

11-As applicable, ((

11-ldentify the value(s) of Reactor Coolant System (RCS) subcooling margin at which ((

11 for North Anna and Surry.

Response

The licensing basis SBLOCA analysis of record for North Anna and Surry is performed in accordance with the NRG Safety Evaluation for the EMF-2328, Supplement 1 methodology. The SE confirms the intended nature and criteria for the RCP trip studies and, separately, the treatment of RCP trip for the break spectrum:

"To prevent SBLOCAs from exceeding the criteria limits, the timing for tripping the RCPs during the event must also be identified." (Section 4.4)

"AREVA has agreed to evaluate a spectrum of hot and cold leg breaks to support the RCP trip procedure and determine/verify the trip timing consistent with the Emergency Operating Procedures (EOPs)." (Section 4.4)

. ((

))

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 9 of 61 These SE statements and the sensitivity studies provided as part of this RAI response establish the licensing basis analyses as those that assume RCP trip on reactor trip.

The following discussion provides relevant historical context along with the results of sensitivity studies conducted with a best estimate operator response RCP trip time in support of the conclusion regarding the assumptions of the licensing basis analysis case and to address RCP behavior with the potential to produce more severe consequences and to be more likely than cases analyzed with RCP trip coincident with reactor trip.

The manual RCP trip strategy following a SBLOCA at North Anna and Surry complies with Generic Letter 85-12 [Reference [6S1-1)) for Westinghouse NSSSs. The enclosure to GL 85-12 provides the NRG safety evaluation associated with the Westinghouse Owners Group (WOG) work justifying credit for manual operator action to trip the RCPs.

This safety evaluation documents the NRG acceptance of the WOG submittals

[References [6S1-2], [6S1-3], and [6S1-4)) in support of the RCP Trip strategy developed for Westinghouse NSSSs. The staff reviewed the plant-specific information for North Anna and Surry provided in References [6S1-5] and [6S1-6], respectively, and concluded in Reference [6S1-7] that the issue of TMI Action Plan Item 11.K.3.5 has been satisfactorily resolved for North Anna and Surry.

The WOG work concluded that automatic RCP trip is not required because adequate time for manually tripping the RCPs was demonstrated using 10 CFR Part 50, Appendix K assumptions as well as most probable best estimate analysis results. It was also concluded that the most probable best estimate analysis results demonstrate that the RCPs can be tripped at any time during the LOCA (if the operator should fail to trip the pumps when the trip criterion is reached) without exceeding the 10 CFR 50.46 acceptance criteria. As discussed in response to RAI 6, the studies performed in WCAP-9584, OG-110 and OG-117 [References [6S1-2], [6S1-3], and [6S1-4], respectively]

identified 5 minutes as the analytical limit for the RCP trip criteria for assurance that the SBLOCA transient results remain below 10 CFR 50.46 acceptance criteria.

The current NRG staff's concern with the existing NRG approved basis for RCP trip time stems from the dated nature of the post-TMI analyses that relied upon older computer codes and significantly simpler modeling practices than current day, as well as the focus of those analyses on the lower end of small break sizes, as opposed to the larger range of small breaks (identified in RAI 6 S1 as;:: 5 inches). To address the NRG staff's concerns surrounding the applicability of the historic analysis conclusions to larger small breaks, two sensitivity studies were performed to demonstrate the adequacy of the RCP trip criterion with respect to the 10 CFR 50.46 acceptance criteria. The sensitivity studies were performed using the North Anna model and an RCP trip time of 1 minute following the loss of RCS subcooling. The RCP trip time of 1 minute represents a best estimate

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 10 of 61 operator response time based on a review of the time critical operator verification program at North Anna.

The first sensitivity study retains all conservatisms prescribed in EMF-2328 and Supplement 1 for the break spectrum analysis (i.e., consistent with Appendix K), but assumes a best estimate operator action time. The results of this study for cold leg and hot leg breaks are presented in Tables 6.S1-1 and 6.S1-2, respectively.

Table 6.51-1: North Anna Cold Leg Break

((

))

((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 11 of 61 Table 6.51-2: North Anna Hot Leg Break

))

These results demonstrate that the analyses using Appendix K assumptions with a best estimate operator trip time continue to meet the 10 CFR 50.46 acceptance criteria. The limiting break with RCP trip 1 minute after the loss of subcooling margin has a higher calculated PCT than the AOR limiting break with RCP trip at the time of reactor trip

((

)). The hot leg break results show no sign of any appreciable heat up.

The second sensitivity study is performed with relaxations to the EM assumptions. Best estimate assumptions are more appropriate for evaluating operator action times. When performing analytical work to support time critical operator actions, the use of overly conservative assumptions is not desirable as it may put undue time pressures on the operators. It may also lead to actions which may place the plant in a less favorable condition. This second sensitivity study demonstrates the conservatism in the first sensitivity by applying two best estimate assumptions, all other assumptions are kept consistent with Appendix K. As was done for Millstone Unit 2 [References [6S1-8] and

[6S1-9)), the decay heat multiplier is reduced from 1.2 to 1.0 and the critical break flow model is changed from Moody to Homogeneous Equilibrium. Relaxing these conservative Appendix K assumptions provides an analysis which is more representative of a realistic response. The results of the study are shown in Table 6.S1-3.

((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 12 of 61 Table 6.51-3: North Anna Cold Leg Break

))

The sensitivity studies submitted in this response confirm that PCT results are less than 2200 °F with the use of a best estimate operator response time for RCP trip. In addition, the sensitivity studies demonstrate that use of a best-estimate operator response time with relaxed Appendix K assumptions results in PCTs well below the results of the licensing basis analyses. Use of best-estimate operator response times is most appropriately used in combination with best-estimate assumptions. These results are consistent with the relevant historical context regarding the basis for setting a 5 minute time for RCP trip as a time critical action.

The base analysis, 5 minute RCP trip study, and the sensitivity studies presented in this RAI response quantitatively demonstrate that the 10 CFR 50.46 acceptance criteria continue to be met under the FVI-SBLOCA application of the EMF-2328, Supplement 1 methodology. Thus, more likely RCP trip behavior is shown not to produce more severe consequences and the licensing basis analyses are those that assume RCP trip coincident with reactor trip.

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 13 of 61 Subcooling Margin RAI 6 S1 also requested additional details about the subcooling margin ((

)). Further clarification from the NRC confirmed this request was specific to cases with a 5 minute RCP trip delay.

Subcooling margin is used to start the delay to the pump trip.

((

References:

[6S1-1]

))

Letter from the United States Nuclear Regulatory Commission to All Applicants and Licensees with Westinghouse (W) Designed Nuclear Steam Supply Systems (NSSSs), "Implementation of TMI Action Item 11.K.3.5, "Automatic Trip of Reactor Coolant Pumps" (Generic Letter No. 85-12),"

dated June 28, 1985.

[6S1-2]

[6S1-3]

[6S1-4]

[6S1-5]

[6S1-6]

[6S1-7]

[6S1-8]

[6S1-9]

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 14 of 61 WCAP-9584, "Analysis of Delayed Reactor Coolant Pump Trip During Small Loss of Coolant Accidents for Westinghouse Nuclear Steam Supply Systems," August 1979.

Westinghouse Owner's Group Letter, OG-110, "Evaluation of Alternate RCP Trip Criteria," October 6, 1983.

Westinghouse Owner's Group Letter, OG-117, "Justification of Manual RCP Trip for Small Break LOCA Events," March 9, 1984.

Letter from W. L. Stewart (VEPCO) to H. R. Denton (USNRC), "Virginia Electric and Power Company, North Anna Power Station Unit Nos. 1 and 2, Response to Generic Letter 85-12, Automatic Trip of Reactor Coolant Pumps," dated February 14, 1986.

Letter from W. L. Stewart (VEPCO)to H. R. Denton (USNRC), "Virginia Electric and Power Company, Surry Power Station Unit Nos. 1 and 2, Response to Generic Letter 85-12, Automatic Trip of Reactor Coolant Pumps," dated December 6, 1985.

Letter from L. B. Engle (USN RC) to W. R. Cartwright (VEPCO), "North Anna Power Station, Units Nos. 1 and 2, and Surry Power Station, Units 1 and 2 /

NUREG-0737, Action Plan Item 11.K.3.5, RCP Trip Issue (TAC Nos. 49665, 49666, 49681 and 49682)," dated March 20, 1989.

Letter from M. D. Sartain (Dominion) to USNRC, "Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 2, Proposed License Amendment Request, Small Break Loss of Coolant Accident Reanalysis,"

Dominion Serial No.15-411, ADAMS Accession No. ML15253A205, dated September 1, 2015.

Letter from R. V. Guzman (USN RC) to D. A. Heacock (Dominion), "Millstone Power Station, Unit No. 2 - Issuance of Amendment Re: Small Break Loss of Coolant Accident Reanalysis (CAC No. MF6700)," Dominion Serial No.16-397, ADAMS Accession No. ML16249A001, dated September 30, 2016.

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 15 of 61 RAI 7 S1 Request:

The response to RA! 7 indicates the results of the refueling water storage tank (RWST) drain down sensitivity study support that the analyses presented in ANP-3467P and ANP-3676P remain bounding. The response to RA! 7 also indicates that the ((

11, that this value was assessed as bounding based on

((

lJ. The response did not provide sufficient detail to conclude that the supporting analyses establish ((

lJ.

Paragraph 10 GFR 50.46(a)(1)(i) requires that EGGS cooling performance must be "calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated." As such, please address the following RA!s for both North Anna and Surry:

a) Identify the break sizes considered in the RWST drain down sensitivity analyses and provide justification for the adequacy of the assessed break sizes.

b) ((

lJ.

c) Supplement Tables 4-2 of both ANP-3467P and ANP-3676P by indicating ((

lJ.

d) Provide adequate technical basis for the conclusion that a post-RWST drain down

((

11 is an appropriately conservative ((

Response

7 S1.a lJ.

All the break sizes analyzed as part of the AOR were assessed in the RWST drain down study. The method used to assess the spectrum with respect to the RWST drain down sensitivity is further discussed in Part b.

7 S1.b Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 16 of 61 The range of the break sizes included in the RWST drain down sensitivity analyses is adequate on the basis that all break sizes analyzed as part of the AOR were assessed.

This assessment begins with the review of the quench time and a conservatively calculated switchover time for each break size in the spectrum (See part c). The quench time is the time after the heatup when the PCT reaches and remains at the saturation temperature. The quench time can be confirmed by looking at PCT plots. The drain down timing during the transient is calculated for each break size in the break spectrum.

((

)) The fluid in the containment sump has mostly re-condensed in the containment after removing heat from the core, meaning that it is at a higher temperature compared to the fluid in the RWST. The high-temperature SI fluid would remove heat less effectively and potentially prolong quenching. However, if the core is already quenched by the time of RWST draining, the hotter SI injection fluid would not cause a significant temperature excursion. Thus, only the cases in which the RWST drains before the core is quenched were evaluated to ensure that the results of the AOR break spectrum limiting case are not challenged. These cases were evaluated with the increased pumped SI temperature to simulate the pumped SI switching suction from the RWST to the containment sump.

Specifically, for the Surry FVI-SBLOCA analysis, ((

)) As an example, Figure 7.S1-1 presents a comparison of PCT from the AOR break spectrum to the RWST drain down study for the North Anna ((

)).

Figure7.S1-1: ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 17 of 61 11 In all re-analyzed cases, the time of quench was extended relative to the base cases.

However, the results of the re-analyzed cases did not challenge those of the limiting case from the AOR break spectrum.

7 S1.c Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 18 of 61 Table 7.S1-1 and Table 7.S1-2 provide the comparisons of calculated switchover time and the AOR quench time for break sizes that are ((

)). Break sizes

((

)) are not included because the core quench for those cases occurs significantly earlier than the time of switchover; therefore, the PCT is assessed to not be impacted by the RWST drain down sensitivity study.

Table 7.51-1: North Anna Break Spectrum Quench and Switchover Times

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 19 of 61 Table 7.51-2: Surry Spectrum Quench and Switchover Times

7 S1.d Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 20 of 61 A post-RWST drain down SI water temperature of ((

11 is an appropriately conservative SI temperature under SBLOCA conditions. As stated in the prior RAI 7 response, the ((

11 value is based on the containment response analyses to the Large Break LOCA (LBLOCA) conditions presented in the North Anna and Surry UFSARs. During a LBLOCA, the rate of mass and energy release to the containment sump is greater than that for a SBLOCA. Because the mass and energy release to the sump in this case is greater than that of the SBLOCA, the sump temperature of

((

11 is also bounding of SBLOCA conditions when the pumped SI suction is switched from the RWST to the containment sump.

Containment sump temperature rises rapidly after both a LB and SBLOCA as the containment spray system condenses steam from the containment atmosphere and break flow from the RCS begins to fill the sump. This temperature increase continues until the recirculation spray (RS) pumps start. Once the RS pumps are operating and passing the sump heat to the recirculation spray heat exchanger (RSHX), containment vapor and liquid temperature drop rapidly. The containment response analysis that is the basis for

((

11 assumed one train of minimum RS pump flow rates and SW flow rates, and modeled degraded RSHX performance for the applicable range of SW temperatures and containment air pressures. On this basis, ((

11 was chosen as a bounding conservative maximum safety injection temperature after switchover.

The determination of the RWST Switchover Time for each break was based ((

)). The use of the minimum usable RWST water volume and maximum spray flow rate leads to an earlier calculated time that sump switchover would be expected to occur. The calculated times for the sump switchover are well after the time the containment sump liquid temperature drops below ((

11 based on the UFSAR containment response analyses.

RAI 8 S1 Request:

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 21 of 61 The NRG staff requested in part b. that the licensee estimate the observed change in peak cladding temperature associated with an S-RELAP5 code modification autonomously implemented by Framatome following the NRG staff's review and approval of the SBLOCA evaluation model described in EMF-2328(P)(A).

The requirements in Appendix K to 10 CFR 50 reflect the importance of performing comparisons of evaluation model predictions against relevant test data. The assessment of the EMF-2328 evaluation model against test data, which constitutes part of the NRG staff's basis for finding the evaluation model acceptable, is specifically discussed in Section 4.5 of the NRG staff's safety evaluation of Revision O and Section 5.3 of the NRG staff's safety evaluation of Supplement 1.

Confirmation of the impact of the autonomously implemented code modification on the calculated peak cladding temperature and other relevant figure of merit specified in 10 CFR 50.46(b) is necessary to confirm whether (1) the existing evaluation model assessment remains valid or (2) a new assessment is necessary with the modified evaluation model the licensee proposes to apply to North Anna and Surry.

Therefore, please provide the following information:

a) Please provide a valid estimate of the magnitude of the impact on peak cladding temperature and other relevant figures of merit specified in 10 CFR 50.46(b) that is associated with the S-RELAP5 modification Framatome autonomously implemented following the NRG staff's review and approval of the EMF-2328(P)(A) evaluation model.

b) Please justify that the autonomously implemented code modification does not adversely affect previously reviewed assessments of the EMF-2328 evaluation model in both Revision O and Supplement 1. Please include available supporting evidence, such as calculated results from a vendor continuity of assessment evaluation, which demonstrates the impact of reanalyzing assessment cases with the modified code version.

c) Please clarify ((

]}.

Response

8 S1.a & 8 S1.b

((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 22 of 61

11 Table 8.51-1: ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 23 of 61 11

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 24 of 61

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 25 of 61 Figure 8.S1-1: ((

References:

[8S1-1]

[8S1-2]

))

EMF-2328(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based.

EMF-2328(P)(A), Revision 0, Supplement 1 (P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based.

8 S1.c

((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 26 of 61

))

RAl1151 Request:

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 27 of 61 The response to RA/ 11 appropriately identified ((

11 Calculation of ((

11 is an essential element of adequately predicting the termination of the cladding heatup, and hence the peak cladding temperature and maximum local oxidation; therefore, ((

11 must be taken into account when demonstrating that the acceptance criteria contained in 10 CFR 50.46(b)(1) and (2) are satisfied. ((

lJ. Confirm that the response to RA/ 11 insofar as it pertains to ((

lJ. If not, please provide a detailed description and justification for the ((

11 in accordance with Section II, "Required Documentation," of Appendix K to 10 CFR 50.

Response

The response to RAI 11 provided two distinct categorizations - on one hand it provided in Table 11-1 a ((

)), and, on the other, it provided in Table 11-2 the ((

)). Table 11-1 provided

((

)) are not included in Table 11-1. Table 11-2 provided

((

)). Unfortunately, the answer did not provide a clear dissociation between the two categorizations. It is evident that these two categorizations do not necessarily overlap. In other words, there are ((

)).

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 28 of 61 During the ((

)) following loop seal clearing, pressure imbalances are relieved following the violent expulsion of the liquid plugs in the loop seals with subsequent random redistribution of the liquid mass to other parts of the primary such as cold legs and the downcomer, and including expulsion of liquid through the break for the broken loop. Following loop seal clearing the core liquid level recovers rapidly to about the level of the cold leg, ((

)).

BAI 12,a S1 Request:

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 29 of 61 The response to RAJ 12.a identified ((

11 parameters that would be considered in determining whether a given analysis pet1ormed with the FVI SBLOCA methodology remains valid for an upcoming fuel cycle, or whether a new analysis must be pet1ormed:

((

11 According to the response to RAJ 12.a, other fuel-related parameters associated with

((

11 would not be considered in the applicability determination.

Due to the non-linear impacts of fuel behavior on the LOCA event (e.g., with respect to cladding deformation and rupture), development of a set of acceptance criteria for continued analysis applicability that is both simplified and universal may be challenging.

For instance, although the ((

11 presented in response to RAJ 19 may apply to existing plant conditions, it does not necessarily apply to other conditions (e.g.,

more severe cladding temperature transients).

Moreover, the responses to RAJ 12 and other questions (e.g., RA/s 20, 21, and 22) did not offer adequate validation that fuel parameters other than those considered by the licensee would not affect the continued applicability of an analysis pet1ormed using the FVI SBLOCA methodology to future fuel cycles.

Based upon the information in the RAJ responses and previously docketed submittals, the NRG staff's review needs additional information to confirm that the ((

11 criteria proposed by the licensee would be sufficient, in general, to assure an acceptable determination of the peak cladding temperature and other figures of merit specified in 10 CFR 50.46(b) on a cycle-specific basis.

Therefore, to assure that the proposed FVI SBLOCA methodology is capable of acceptably calculating figures of merit for comparison with the acceptance criteria specified in 10 CFR 50.46(b), please (1) provide adequate evidence demonstrating that the proposed criteria for determining the continued applicability of an analysis on a cycle-specific basis are valid or (2) propose a modified set of criteria or methodology for making the determination that is supported by adequate justification.

Response

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 30 of 61 RAI 12.a S1 requests additional evidence to demonstrate that the proposed reload checks are adequate for determining the continued applicability of a FVI-SBLOCA analysis on a cycle-specific basis. Based on discussions held with staff during the January 2020 Audit, additional background is warranted regarding the Dominion Energy reload evaluation process. Dominion Energy validates reload design patterns using a process governed by an internal Reload Safety Analysis Checklist. This process is compliant with NRG-approved Topical Report VEP-FRD-42-A [Reference 12.aS1-1] and involves an assessment of cycle specific variations between the licensing basis analyses and a given reload core pattern.

Cycle specific variations are defined as those parameters that can change during a typical core reload. These parameters may include, but are not limited to, uranium enrichment, loading pattern, core peaking factors, burnup, and power history. Minor fuel component changes (e.g., bottom nozzle skirt hole pattern change) may be made by the fuel vendor in accordance with their NRG approved fuel change processes. These changes are also assessed in a manner similar to cycle specific variations due to the core pattern change.

In the consideration of whether a parameter change is a cycle specific variation, parameters such as the following remain the same from cycle-to-cycle:

Fuel product beyond minor fuel component changes Peaking factor limit Cycle length (within normal cycle-to-cycle variations; e.g., a nominal 18 month cycle)

Burnable absorber (integral or discrete)

Given that the typical cycle-specific variations are limited for a normal reload design, fuel products and core operational plans do not change in a manner that impacts the approved LOCA analyses. Assessment of the North Anna and Surry FVI-SBLOCA analyses shall be performed on a reload basis as part of the Dominion Energy process. In addition to existing reload pattern checks, the neutronics parameters determined to have the potential to influence highly ranked PIRT phenomena, as discussed in the prior response to RAI 12.a, shall be incorporated in the reload checks associated with the FVI-SBLOCA analyses due to the use of a representative Framatome fuel model.

Changes to fuel products and core operational plans in excess of typical cycle-specific variations, such as those listed above, require additional engineering review to determine impacts to the station design and licensing bases in accordance with the recommendations of INPO SOER 96-02 [Reference 12.aS1-2] and SOER-03-2

[Reference 12.aS1-3]. These INPO SOERs provide recommendations for evaluating new

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 31 of 61 fuel assembly and fuel/core component design features, changes to fuel management strategy, and new or unique core design and operating strategies to determine if the change is a significant change. Should a change be identified as significant (i.e., in excess of typical cycle-specific variations), additional design work, up to and including the potential to request NRG approval, is required in order to demonstrate compliance with existing regulations (e.g., 10 CFR 50.59, 10 CFR 50.46).

Dominion Energy's compliance with the licensed reload assessment Topical Report

[Reference 12.aS1-1], fuel vendor compliance with licensed fuel change processes, and the additional reload checks discussed in the prior response to RAI 12.a ensure that cycle specific variations are appropriately assessed against the inputs assumed in the FVI-SBLOCA analyses for both North Anna and Surry.

References:

[12.aS1-1]

VEP-FRD-42-A, Revision 2, Minor Revision 2, "Reload Nuclear Design Methodology," October 2017.

[12.aS1-2]

INPO Significant Operating Experience Report (SOER) 96-02, "Design and Operating Conditions for Reactor Cores," November 19, 1996.

[12.aS1-3]

INPO Significant Operating Experience Report (SOER) 03-02, "Managing Core Design Changes," July 22, 2003.

BAI 12,b 51 Request:

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 32 of 61 The response to RA/ 12.b appears to identify ((

11 criteria for determining whether the FVI SBLOGA methodology may be applied to a given fuel design:

((

11 The response to RA/ 17 further appears to add an additional criterion that ((

11:

Fuel has the same lattice structure as the current resident fuel The NRG staff's review of these criteria identified the potential for applications of the FVI SBLOGA methodology to fuel designs with properties and features that extend beyond the set of characteristics explicitly considered in the NRG staff's review of the Framatome codes supporting the methodology. For instance, EMF-92-116(P)(A) indicates that RODEX2 may be used to model fuel for U.S. PWRs using uranium dioxide and urania-gadolinia pellets, and Zircaloy-4 or M5 cladding. The use of the RODEX2 code for fuel incorporating other types of cladding materials, burnable poisons, and other design features not currently used in Framatome fuels, but which may be incorporated into other vendors' fuel designs, has not been previously reviewed. Existing reviews of the RODEX2 code also did not previously review the impacts of many novel design features being considered for future fuel rods, such as advanced cladding alloys and coatings; different fuel materials, enrichments, geometries, dopants, and coatings; and advanced fabrication techniques. The NRG staff's safety evaluation on EMF 116(P)(A) further assessed the RODEX2 code only up to Framatome's currently licensed maximum burnup of 62 GWdlMTU. Although the criteria proposed by the licensee in response to RA! 12.b would apparently allow the previously reviewed application scope for the RODEX2 code to be exceeded, adequate justification for application of RODEX2 to such expanded applications has not been provided. Furthermore, note that this discussion applies not only to ((

lJ.

Additionally, the licensee's response to RA/ 13 did not provide assurance that a

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 33 of 61 sufficiently robust administratively controlled process exists to incorporate potentially incomplete data for other fuel cladding types and other design features into analyses performed using the FVI SBLOCA methodology.

Finally, regarding application of the FVI SBLOCA methodology to all fuel designs clad with a zirconium-based alloy, the NRG staff noted that a comparison of the response to RA/ 21 and results presented in ANP-331 SP indicates ((

)). Additional discussion of the potential for fuels clad with different zirconium-based alloys to behave differently during an SBLOCA event is presented below in RA! 18 S1.

To assure that the proposed FVI SBLOCA methodology is capable of acceptably calculating figures of merit for comparison with the acceptance criteria specified in 10 CFR 50.46(b), please list the criteria that would be used to determine applicability of the FVI SBLOCA methodology to a specific fuel design. Furthermore, please (1) provide adequate justification that the proposed criteria for determining the applicability of the FVI SBLOCA methodology to a specific fuel design are valid or (2) propose a modified set of criteria or methodology for making the determination that is supported by adequate justification.

Response

Following the NRG audit in January of 2020, Dominion Energy is altering the request for approval of the FVI-SBLOGA analysis application to only the currently loaded and operating fuel types at North Anna and Surry, with minor fuel component changes made by the fuel vendor in accordance with their NRG approved fuel change processes, and Framatome fuel products. This alteration significantly limits the range of potential fuel design changes that can be assessed using the submitted FVI-SBLOGA analysis.

((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 34 of 61

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 35 of 61

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 36 of 61 11 Table 12.51-1: Fuel 8 Results

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 37 of 61 Table 12.51-2: FVI Results Recap for Same Break Sizes

Figure 12.S1-1: ((

Figure 12.S1-2: ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 38 of 61

))

))

Figure 12.51-3: ((

Figure 12.51-4: ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 39 of 61

))

))

Figure 12.51-5: ((

Figure 12.51-6: ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 40 of 61

))

))

Figure 12.51-7: ((

Figure 12.51-8: ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 41 of 61 11 11

Figure 12.51-9: ((

Figure 12.51-10: ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 42 of 61 11 11

Figure 12.51-11: ((

Figure 12.51-12: ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 43 of 61 11 11

BAI 13 S1 Request:

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 44 of 61 The response to RA/ 13 discusses an "established process" for determining whether the characteristics of an alternative cladding material may be represented by the modeling practices used in an analysis performed with the FVI SBLOCA methodology. In the event limited data is available for performing an assessment, the licensee's response states that compensatory actions may be taken, including defining a conservative peak cladding temperature penalty that would be intended to account for data limitations.

In accordance with Criterion Ill of Appendix B to 10 CFR 50, the establishment of design control measures is required to assure that applicable regulatory requirements and design basis information is correctly translated into specifications, drawings, procedures, and instructions. Criterion Ill identifies that such design control measures shall be applied to, among other things, accident analyses. Criterion Ill further requires that measures shall also be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components.

To assure that analytical predictions made by the proposed FVI SBLOCA methodology would representatively model the actual plant design, in accordance with the above regulatory requirements, the following information is needed to support the licensee's proposed fuel modeling practices:

a) Please identify the document that contains the established process for evaluating alternative cladding materials and summarize the key elements of this process.

b) Please clarify whether the established process applies solely to the evaluation of alternative cladding materials or whether it applies more generally to any fuel properties which may diverge from those that have been explicitly evaluated using the FVI SBLOCA methodology. If the established process applies only to alternative fuel cladding materials, then please further discuss how impacts of differences in other fuel properties would be assessed.

c) Please adequately describe and justify in particular the process for defining a conservative penalty on peak cladding temperature under conditions where available data is limited, considering that data limitations may challenge the capability to define a conservative penalty.

Response

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 45 of 61 As discussed in the NRG audit held at Dominion Energy in January of 2020, and in the response to RAI 12.b S1, the request for approval for application of the North Anna and Surry FVI-SBLOGA analyses is limited to the resident fuel products with minor component changes by the fuel vendor in accordance with their NRG approved fuel change processes and the Framatome fuel products that meet the following criteria:

The same 15x15 or 17x17 array as the resident fuel product, An NRG-approved Zirconium-based cladding material, Natural or Slightly Enriched Uranium Dioxide fuel, with allowances for currently available integral poisons, and Within the existing method applicability and limitations of the supporting codes and methods Failure to meet those criteria would require an NRG submittal to address the SBLOGA analyses.

Because of the stringent criteria for application of the FVI-SBLOGA analysis, the potential concerns described in RAI 13 S1 are eliminated. Specifically, a) The added constraints with respect to both the cycle-specific assessment in RAI 12.a S1 and those above ensure that sufficient information exists to perform an informed assessment of alternative cladding materials.

The alternate clad materials are restricted to NRG-approved materials.

ZIRLO and Optimized ZIRLO, have been assessed with this submittal and the RAI responses; Framatome methods have been updated to support current Framatome clad products Similarly, when considering Framatome designs other than those analyzed, adequate information is available to determine if the design changes are within the applicability b) As demonstrated by the sensitivity study discussed in RAI 12.b S1, the constraints placed on the FVI-SBLOGA analysis application prevent variation in the fuel design parameters from having a substantial impact on the transient results.

c) The scenario in which available data for the fuel product is limited is precluded by the limitations on the FVI-SBLOGA analysis application described above. In the event that the data for the fuel product in consideration is limited, an NRG submittal would be required to address the SBLOGA analyses.

BAI 16 S1 Request:

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 46 of 61 The response to RAJ 16 provided evidence generated during the review of a design certification for an evolutionary pressurized water reactor design, suggesting that ((

lJ. Considering the following:

Whereas the limiting SBLOCA event for the design certification application exhibited ((

11; The RAJ response associated with the design certification review (ML090970699) indicated that the time required for ((

For some of the limiting ((

ANP-3676P, the ((

For Surry in particular, the ((

11 during the transient, 11; 11 described in ANP-3467P and 11; and The NRG staff is not able to conclude that the effects of TCD ((

l]. Paragraph 10 CFR 50.46(a)(1)(i) requires the consideration of postulated LOCAs of sufficient sizes, locations, and other properties sufficient to provide assurance that the most severe hypothetical LOCAs are calculated. In addition, paragraph 1, "The initial stored energy in the fuel," of Section A, "Sources of heat during the LOCA," of Part I, "Required and Acceptable Features of the Evaluation Models," of Appendix K, "EGGS Evaluation Models," to 10 CFR 50 requires that the U02 thermal conductivity be evaluated as a function of burnup and temperature. ((

lJ. Provide sufficient information that is specifically applicable to the requesting plants to demonstrate that the existing approach ((

11, consistent with the required and acceptable features of EGGS evaluation models set forth in Appendix K to 10 CFR 50. If the existing approach ((

lJ.

Response

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 47 of 61 Sensitivity studies were conducted to determine the effect of an increase in fuel internal energy representing TCD. A case from both the Surry and North Anna SBLOCA break spectrum (References [16S1-1] and [16S1-2]) that results in multiple loop seal clearings, i.e. one of the larger break cases, is chosen for continuing studies. ((

))

Table 16.51-1: North Anna and Surry

((

))

((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 48 of 61 Table 16.51-2: North Anna and Surry

))

With the models re-initialized, the chosen break cases were re-run. PCT comparisons between the AOR and the sensitivity studies are included in the following table.

Table 16.51-3: North Anna and Surry Sensitivity

((

))

A plot of the fuel temperature response from the Surry SBLOCA sensitivity study follows.

((

))

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 49 of 61 Figure 16.51-1: ((

References:

[16S1-1]

[16S1-2]

))

ANP-3676P RO, Surry Fuel-vendor Independent Small Break LOCA Analysis.

ANP-3467P RO, North Anna Fuel-vendor Independent Small Break LOCA Analysis.

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 50 of 61 RAI 18 S1 (ADEQUACY OF RUPTURE STRAIN VS. RUPTURE TEMPERATURE CORRELATION)

Regulatory Basis Appendix K to 10 CFR 50, Part I, "Required and Acceptable Features of the Evaluation Models," Section B, "Swelling and Rupture of the Cladding and Fuel Rod Thermal Parameters," requires, in part, that, "To be acceptable the swelling and rupture calculations shall be based on applicable data in such a way that the degree of swelling and incidence of rupture are not underestimated."

In addition to the Appendix K requirement, 10 CFR 50.46(a)(1)(i) requires the calculation of a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe hypothetical loss-of-coolant accidents are calculated. Paragraph 10 CFR 50.46(b)(1) requires that the peak cladding temperature remain below 2200 °F, and 10 CFR 50.46(b)(2) limits the maximum amount of local oxidation to 17% of the cladding thickness before oxidation, with oxidation calculated for both cladding surfaces if rupture is calculated to occur.

Description of Rupture Strain/Temperature Relationship The response to RAJ 18 ((

11

((

11 The NRG staff relied on this general description of the high-temperature plastic deformation behavior of zirconium alloys, in concert with ((

lJ.

Review of RAJ 18 Response

((

))

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 51 of 61 Assessment of RAJ 18 Response Against North Anna Rupture Results The response to RAJ 18 states, ((

11 The North Anna break spectrum analysis described in ANP-3467P included ((

11 Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 52 of 61 Consideration of an Alternative, Advanced Zirconium Alloy Cladding Consider the following discussion, as provided by Framatome Cogema Fuels (now Framatome) in Appendix C to BAW-10227P-A, "Evaluation of Advanced Cladding and Structural Material (MS) in PWR Reactor Fuel," in ((

11:

((

((

11 11 Additional consideration of MS data identifies ((

11

Request:

Given that the FVI SBLOCA analyses ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 53 of 61 11, additional justification is required ((

11 remains consistent with 10 CFR 50.46(a)(1)(i) and 10 CFR 50, Appendix K requirements identified above. Please provide an adequate demonstration that ((

11 does not underestimate the degree of swelling, consistent with the applicable Appendix K requirements identified above, and ensure that the demonstration is applicable within the range of peak fuel cladding temperatures permitted by the acceptance criteria contained in 10 CFR 50.46(b)(1), i.e.,

up to 2200 °F. Provide justification that the model adequately considers ((

11, regarding overall effect on essential figures of merit.

Response

((

11 Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 54 of 61 The North Anna Fuel Vendor Independent (FVI) analysis of record (AOR) uses a swelling and rupture model ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 55 of 61 11 Table 18.51-1: AOR Results and Rupture Data

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 56 of 61 Table 18.S1-2: Sensitivity Study Results and Rupture Data

Figure 18.51-1: ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 57 of 61 11 Figure 18.51-2: PCT Comparison

((

((

Figure 18.51-3: ((

11 11 Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 58 of 61 11 Comparison

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 59 of 61 Figure 18.S1-4: ((

11 Break Cladding Temperature Comparison Figure 18.S1-5: ((

11 Break Hot Assembly Collapsed Level Comparison

((

((

11 Figure 18.51-6: ((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 60 of 61 11 11 Break Cladding Temperature Comparison

Conclusion

((

Serial No.20-149 Docket Nos. 50-338/339/280/281 Page 61 of 61

))

Serial No.20-149 Docket Nos. 50-338/339/280/281 FRAMATOME AFFIDAVIT FOR WITHHOLDING PROPRIETARY INFORMATION Virginia Electric and Power Company (Dominion Energy Virginia)

North Anna and Surry Power Stations Units 1 and 2

AFFIDAVIT Serial No.20-149 Docket Nos. 50-338/339/280/281 2 of 4

1.

My name is Gayle Elliott. I am Deputy Director, Licensing & Regulatory Affairs for Framatome Inc. (Framatome) and as such I am authorized to execute this Affidavit.

2.

I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.

3.

I am familiar with the Framatome information contained in Attachment 1 to correspondence from Mr. Mark D. Sartain (Virginia Electric and Power Company) to Document Control Desk (U.S. Nuclear Regulatory Commission), entitled, "Virginia Electric and Power Company, North Anna and Surry Power Stations Units 1 and 2, Proposed Licensing Amendment Requests, Addition of Analytical Methodology to the Core Operating Limits Report for a Small Break Loss of Coolant Accident (SBLOCA) Response to Request for Additional Information and Analysis Error Correction," Docket Nos.: 50-338/339 and 50-280/281, dated May 2020 and referred to herein as "Document." Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.

4.

This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5.

This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is

Serial No.20-149 Docket Nos. 50-338/339/280/281 3 of 4 made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6.

The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:

(a)

The information reveals details of Framatome's research and development plans and programs or their results.

(b)

Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c)

The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.

(d)

The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability.

(e)

The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(d) and 6(e) above.

7.

In accordance with Framatome's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8.

Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

Serial No.20-149 Docket Nos. 50-338/339/280/281 4 of 4

9.

The foregoing statements are true and correct to the best of my knowledge, information, and belief.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on: May 20, 2020 Gayle Elliott