RS-15-266, Response to Request for Additional Information Regarding Request for a License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink.

Response to Request for Additional Information Regarding Request for a License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink.

ML15282A345
Person / Time
Site:	Braidwood
Issue date:	10/09/2015
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-15-266, TAC MF4671, TAC MF4672
Download: ML15282A345 (9)

Text

40() Wit it ('1(1 Romi Wir riivill, Awomw ExeLon G (/ ?OO() Off RS-1 5-266 10 CFR 50.90 October 9, 2015 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units I and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457

Subject:

Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units I and 2, Technical Specification 3.7.9, "Ultimate Heat Sink"

References:

1) Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Request for a License Amendment to Braidwood Station, Units I and 2, Technical Specification 3.7.9, 'Ultimate Heat Sink,"

dated August 19, 2014 (ML14231A902)

2) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC)

Preliminary RAIs for LAR Regarding Braidwood Station Technical Specification, "Ultimate Heat Sink," dated February 5, 2015 (ML15036A431)

3) Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units I and 2, Technical Specification 3.7.9, "Ultimate Heat Sink," dated March 31, 2015 (ML15090A604)

4) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC)

Preliminary Containment and Ventilation Branch RAI5 for Braidwood UHS LAR, dated March 9, 2015 (ML15069A004)

5) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC)

Revised Preliminary RAIs for LAR Regarding Braidwood Station Technical Specification, "Ultimate Heat Sink," dated March 24, 2015 (MI-1 5084A01 8)

6) Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units I and 2, Technical Specification 3.7.9, "Ultimate Heat Sink," dated April 30, 2015 (ML15120A396)

October 9, 2015 U.S. Nuclear Regulatory Commission Page 2

7) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC)

Additional RAI Regarding Containment Analysis for Braidwood UHS LAR (MF4671 and M174672), dated July 22, 2015

8) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC)

Need Clarification Conference Call Regarding Your April 30, 2015 Response to SCVB-RAI-1(a), dated August 12, 2015 In Reference 1, Exelon Generation Company, LLC, (EGC) requested an amendment to the Technical Specifications (TS) of Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units I and 2. The proposed amendment would modify TS 3.7.9, 'Ultimate Heat Sink (UHS)," by changing the maximum allowable temperature of the UHS from 100 OF to a maximum UHS temperature of 102°F. In Reference 2, the U. S. Nuclear Regulatory Commission (NRC) requested additional information related to its review of Reference 1. In Reference 3, EGC provided the requested information. In Reference 4, the NRC provided additional preliminary RAIs from the Containment and Ventilation Branch. In Reference 5, the NRC clarified the questions provided in Reference 4 and indicated that certain questions originally transmitted in Reference 4 had been eliminated and no longer required a response (i.e., RAIs 2, 3 and 4). Reference 6 provided a response to the Reference 4 and Reference 5 RAI5.

Reference 7 and Reference 8 requested additional information related to the Reference I submittal and Reference 6 follow-up response. Attachment I provides the response to the Reference 7 and Reference 8 requested information.

EGC has reviewed the information supporting a finding of no significant hazards consideration that was previously provided to the NRC in Attachment I of Reference 1. The information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), a copy of this letter and its attachment is being provided to the designated State of Illinois official.

Should you have any questions concerning this letter, please contact Ms. Jessica Krejcie at (630) 657-2816.

October 9, 2015 U.S. Nuclear Regulatory Commission Page 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 9th day of October 2015.

Respectfully, David M. Gullott Manager Licensing Exelon Generation Company, LLC

Attachment:

Response to Request for Additional Information cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, Braidwood Station Illinois Emergency Management Agency Division of Nuclear Safety

Response to Request for Additional Information ATTACHMENT Response to Request for Additional Information In Reference 1, Exelon Generation Company, LLC, (EGC) requested an amendment to the Technical Specifications (TS) of Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units I and 2. The proposed amendment would modify TS 3.7.9, "Ultimate Heat Sink (UHS)," by changing the maximum allowable temperature of the UHS from 100 °F to a maximum UHS temperature of 102°F. In Reference 2, the U. S. Nuclear Regulatory Commission (NRC) requested additional information related to its review of Reference 1. In Reference 3, EGC provided the requested information. In Reference 4, the NRC provided additional preliminary RAIs from the Containment and Ventilation Branch. In Reference 5, the NRC clarified the questions provided in Reference 4 and indicated that certain questions originally transmitted in Reference 4 had been eliminated and no longer required a response (i.e., RAls 2, 3 and 4). Reference 6 provided a response to the Reference 4 and Reference 5 RAIs.

Reference 7 and Reference 8 requested additional information related to the Reference I submittal and Reference 6 follow-up response. This attachment provides the response to the Reference 7 and Reference 8 requested information.

SCVB-RAI-1I SCVB-RAI-1 I The NRC staff is aware of Westinghouse's InfoGram lG-14-1, dated November 5, 2014, which states that the loss-of-coolant accident (LOCA) containment mass and energy (M&E) release analysis methodology was found to use the reactor coolant system (RCS) stainless steel volumetric heat capacity value lower than the ASME values. The staff has received information which indicates that the impact on the LOCA peak containment pressure in the current licensing basis would be significant if the analysis would use ASME values for the RCS metal volumetric heat capacity. Provide an update to your application dated August 19, 2014 (ADAMS Accession Number ML14231A902), that contains the results of the containment pressure, containment temperature, sump temperature responses and net positive suction head (NPSH) analysis using the ASME published volumetric heat capacity for the RCS metal. In your supplement dated April 30, 2015 (ADAMS Accession No. ML15128A186) you stated that, "All issues related to the subject Nuclear Safety Advisory Letters (NSALs) (NSAL-06-6, NSAL-1 1-5, and NSAL-14-2) were explicitly addressed (and subsequent corrections in the loss-of-coolant (LOCA) mass and energy (M&E) release analysis were made) in support of the Ultimate Heat Sink (UHS) analysis." Confirm, in your update, that these NSALs continue to be addressed in the requested updated analysis.

SCVB-RAI-I I Response:

Exelon has reviewed the above RAI and will be providing a response in a future transmittal.

SCVB-RAI-I 2:

Exelon letter dated April 30, 2015, response to SCVB-RAI-1 (a), third paragraph states:

"Braidwood Unit I specific analysis of this event models the hydraulics of the affected systems using RELAP5 up to the point of void collapse and determines the resulting flow and pressures, Page 1 of 5

ATTACHMENT Response to Request for Additional Information and then calculates water hammer forces using either the EPRI methodology or a method of characteristics-based program, such as HYTRAN."

Provide a description of the EPRI methodology and the program HYTRAN that was used for the analysis and, if possible, a reference to NRC approval of the methodology.

SCVB-RAI-12 ResDonse:

As discussed in Reference 6, the Braidwood Station Unit I and Unit 2 analysis utilized EPRI methodology and the program HYTRAN to calculate water hammer forces for evaluations performed in support of Generic Letter (GL) 96-06. The Braidwood Station application of EPRI methodology was described in various submittals, as outlined in "Byron Station, Units I and 2, and Braidwood Station, Units I and 2, Generic Letter 96-06, 'Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," dated February 24, 2004 (Reference 9). As documented in Reference 9, the NRC reviewed the various Braidwood Station submittals and indicated satisfactory resolution of GL 96-06 requested actions, including water hammer and two-phase flow conditions and thermally-induced overpresurrization. The Reference 9 letter includes reference to the NRC approval of EPRI methodology (ML020940132).

The GL 96-06 calculations were revised for Braidwood Station in support of the Measurement Uncertainties Recapture (MUR) project. In the revised analyses, RELAP continues to be used to determine the locations where the vapor voids would collapse, along with the corresponding fluid velocities of the two water slugs. As described in Reference 6, the Braidwood methodology includes calculating water hammer loads using either the EPRI methodology or a Method of Characteristics (MOC) based program, HYTRAN. The EPRI methodology has limitation of applicability based on voids closure velocity. The RELAP results for Braidwood show closure velocities above the EPRI applicability criteria, therefore, HYTRAN was used to calculate the forces from the void collapses.

HYTRAN is a PC based program that implements MOC to perform the waterhammer analysis.

It computes time-dependent forces on, and pressures and flow velocities in, the legs of a liquid filled piping system. Output gives pressures and flow velocities at the end nodes of each leg, and the internal nodes if desired, and also gives the unbalanced force on the leg.

The HYTRAN model is used to match pre-collapse steady state flows in the RELAP analysis, match flows just prior to the steam bubble collapse, collapse the steam bubble in the predicted time, and return to steady state flow after collapse. Loads from the collapsing voids are determined along each piping leg. The HYTRAN results are then post-processed to generate time-history files containing flow velocities, pressures and leg forces. The flows before and after the void collapse were matched to the RELAP flows for the HYTRAN results to be valid.

Furthermore, the HYTRAN results were ensured to be conservative by verifying that the rate of change of the velocity (i.e., slope of the velocity trace) in the HYTRAN analysis is greater than or equal to the rate of change of the velocity in the RELAP analysis.

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ATTACHMENT Response to Request for Additional Information The HYTRAN computer program is part of the original licensing basis for Braidwood Station and is described in UFSAR, Appendix D, Section D.8.6 and referenced in UFSAR section 3.9.1.2, "Computer Programs Used in Analyses":

HYTRAN (Hydraulic Transient Analysis) calculates pressures, velocities and force transients in a liquid filled piping network due to transients that are initiated by valve closure, pump trip, or by pressure changes at a piping terminal. HYTRAN was validated by comparison of sample problem results to results given in published documents.

The use of HYTRAN for this application has been validated by other utilities in support of GL 96-06 related analyses. A submittal (Reference 10) for Palisades Nuclear Power Plant compared HYTRAN results to results from the EPRI Rigid Body method (RBM). The report concluded that the RBM and HYTRAN/MOC loads correlate well, and the HYTRAN/MOC loads are almost always higher than the RBM method loads. The conservatism with using the HYTRAN program when compared to the EPRI RBM method was acknowledged by the NRC in the SER for the Palisade Plant GL 96-06 submittals (Reference 11).

SCVB-RAI-1 3 Exelon letter dated April 30, 2015, response to SCVB-RAI-1(a), last paragraph states:

"The RCFC Unit 2 piping analyses are based on the Unit I RCFC piping analytical models, with unit specific differences reconciled during original construction. Unit specific RCFC piping differences that are applicable to the 96-06 analysis will be reconciled as part of Regulatory Commitment #1 of Reference 1, prior to the implementation of the UHS LAR."

Explain what is implied by reconciliation of unit specific differences (if possible, provide a reference of NRC approval during initial licensing) and how the analysis for Braidwood Unit 2 will be performed using the analysis for Unit I SCVB-RAI-1 3 Resronse:

As described in Reference 6, for Braidwood Unit 2, the original analyses for the piping subsystems that are evaluated for Generic Letter (GL) 96-06 are based on the analyses of the corresponding Unit I piping subsystems.

During original design and construction, the Braidwood Unit 2 analyses documented the review of the as-built piping configurations (piping geometry and support configuration) with respect to the Braidwood Unit I analyzed configuration. Variances were evaluated and reconciled. The analyses document that all piping stresses for the reconciled Braidwood Unit 2 geometry are within code allowable. The results of the Braidwood Unit 2 piping subsystem analyses confirmed compliance with applicable Codes and were reviewed as part of the original Unit 2 licensing. Use of the piping analytical models of the lead Unit (i.e., Unit 1) to the design of the replicate unit (i.e., Unit 2) is common industry practice.

Page 3 of 5

ATTACHMENT Response to Request for Additional Information The GL 96-06 RELAP model includes the entire flow path from the Ultimate Heat Sink through the Reactor Containment Fan Coolers (RCFCs) and back to the UHS. A simplified model of the flow paths parallel to the RCFCs is included in the analysis. The parallel flowpaths are not modeled in detail as void formation occurs outside these parallel flow paths. Based on similarities between Unit I and Unit 2 Essential Service Water (SX) piping configurations (as documented in the reviews of the as-built piping subsystems), the results of the RELAP model using the Braidwood Unit I configuration are considered representative for both Braidwood Units.

The forces on the piping that were calculated with HYTRAN were reviewed against the structural capacity of the pipe supports. The support loads and configuration were obtained from the stress analyses for the affected Braidwood Unit I piping subsystems. The regulatory commitment is to validate that the structural capacity of the Unit 2 pipe support configuration is equivalent to the capacity of the Unit I pipe support configuration. This commitment will support the conclusion that the Unit 2 pipe supports can accommodate forces developed from void collapses.

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ATTACHMENT Response to Request for Additional Information

REFERENCES:

1) Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Request for a License Amendment to Braidwood Station, Units I and 2, Technical Specification 3.7.9, 'Ultimate Heat Sink," dated August 19, 2014 (ML14231A902)

2) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC) Preliminary RAls for LAR Regarding Braidwood Station Technical Specification, "Ultimate Heat Sink," dated February 5, 2015 (ML15036A431)

3) Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units I and 2, Technical Specification 3.7.9, "Ultimate Heat Sink," dated March 31, 2015 (MLI 5090A604)

4) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC) Preliminary Containment and Ventilation Branch RAls for Braidwood UHS LAR, dated March 9, 2015 (MLI 5069A004)

5) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC) Revised Preliminary RAls for LAR Regarding Braidwood Station Technical Specification, "Ultimate Heat Sink," dated March 24, 2015 (ML15084A018)

6) Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units I and 2, Technical Specification 3.7.9, "Ultimate Heat Sink," dated April 30, 2015 (ML15120A396)

7) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC) Additional RAI Regarding Containment Analysis for Braidwood UHS LAR (MF4671 and MF4672),

dated July 22, 2015

8) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC) Need Clarification Conference Call Regarding Your April 30, 2015 Response to SCVB-RAI-1(a), dated August 12, 2015

9) Letter from Mahesh Chawla (U.S. NRC Office of Nuclear Reactor Regulation) to Mr.

Christopher M. Crane, (Exelon Generation Company, LLC), "Byron Station, Units I and 2, and Braidwood Station, Units I and 2, Generic Letter 96-06, 'Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions' (TAC NOS. M96789, M96790, M96782, AND M96783)," dated February 24, 2004 (ML040220400).

10)Letter from Daniel J Malone (Nuclear Management Company, LLC) to U.S. Nuclear Regulatory Commission, "Resolution of Generic Letter (GL) 96-06 Watterhammer Issues revised Response Requested Actions to Address GL-96-06," dated August 18, 2004 (ML042370019) 11)Letter from David H. Jaffe (U.S. NRC Office of Nuclear Reactor Regulation) to Mr. Daniel J. Malone (Nuclear Management Company), "Palisades Plant Generic Letter 96-06 Regarding Piping Thermal Overpressurization, Waterhammer, and Two-Phase Flow Issue (TaG No. M96844), dated January 11, 2005 (ML040560007)

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