ML053320361

E-mail, Ennis, NRR, to Daflucas, Entergy, VY EPU Supplements 38 and 39

ML053320361
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	11/22/2005
From:	Richard Ennis Office of Nuclear Reactor Regulation
To:	Daflucas R Entergy Nuclear Operations
References
%dam200601
Download: ML053320361 (6)

Text

I Rick -E-n-n-i-s----V--Y--E-P-U ---D--ra-ft---R-A-I-s-fo--r--S-u-r)-r) Ie--ments-38--and- 3-9--

I Rick Ennis VY EPU Draft RAIs for SuoDlements 38 and 39 Paae 1 Pagie II From: Rick Ennis To: Ronda Daflucas Date: 11/22/05 10:05AM

Subject:

VY EPU - Draft RAls for Supplements 38 and 39 Ronda, Attached are draft RAls from Marty Stutzke concerning the information provided in Supplements 38 and 39 (risk aspects of crediting containment overpressure).

Let me know if you want a conference call to discuss the questions. Also, please let me know ASAP when Entergy can provide a response. I will issue these as formal RAls after I have your response timeframe.

thanks, Rick 301-415-1420 CC: Brian Hobbs; Craig Nichols; Darrell Roberts; Jim DeVincentis; John McCann; Len Gucwa; Martin Stutzke; Richard Lobel

I cAtemP\GWI00002.TMP Ic:temp\GW1OOOO2.TMP PaQe 11l Page I Mail Envelope Properties (438333A2.86B: 11: 372)

Subject:

VY EPU - Draft RAIs for Supplements 38 and 39 Creation Date: 11/22/05 10:05AM From: Rick Ennis Created By: RXE@nrc.gov Recipients Action Date & Time TWGWPOO2.HQGWDOO1 MAS7 CC (Martin Stutzke) owvf2_o.OWFNDO Delivered 11/22/05 10:05 AM RML CC (Richard Lobel) ouTf4_o.OWFNDO Delivered 11/22/05 10:05 AM DJR CC (Darrell Roberts)

RXE BC (Rick Ennis) entergy.com bhobbs CC (Brian Hobbs) cnichol CC (Craig Nichols) jdevinc CC (Jim DeVincentis) jmccanl CC (John McCann)

LGUCW90 CC (Len Gucwa) rdafluc (Ronda Daflucas)

Post Office Delivered Route TWGWPOO2.HQGWDOOI Pending owvf2po.OWFNDO 11/22/05 10:05 AM ovvf4_po.OWFNDO 11/22/05 10:05AM Pending entergy.com Files Size Date & Time MESSAGE 1030 11/22/05 10:05AM draM raiO9 mcO761.wpd 32626 11/22/05 09:56AM Options Auto Delete: No Expiration Date: None Notify Recipients: Yes Priority: Standard Reply Requested: No Return Notification: None

1'-\emP\G1002.TMP -Page 2i I

I Concealed

Subject:

No Security: Standard To Be Delivered: Immediate Status Tracking: Delivered & Opened

REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED LICENSE AMENDMENT EXTENDED POWER UPRATE VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271 By letter dated September 10, 2003, as supplemented by letters dated October 1, and October 28 (2 letters), 2003, January 31 (2 letters), March 4, May 19, July 2, July 27, July 30, August 12, August 25, September 14, September 15, September 23, September 30 (2 letters),

October 5, October 7 (2 letters), December 8, and December 9, 2004, and February 24, March 10, March 24, March 31, April 5, April 22, June 2, August 1, August 4, September 10, September 14, September 18, September 28, October 17, October 21 (2 letters), October 26, October 29, and November 2, 2005, Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. submitted a proposed license amendment to the Nuclear Regulatory Commission (NRC) for the Vermont Yankee Nuclear Power Station (VYNPS). The proposed amendment, 'Technical Specification Proposed Change No. 263, Extended Power Uprate" would allow an increase in the maximum authorized power level for VYNPS from 1593 megawatts thermal (MWt) to 1912 MWt.

The NRC staff is reviewing your extended power uprate (EPU) amendment request and has determined that additional information is required to complete the review. The specific information requested is addressed below.

Probabilistic Risk Assessment (PRA) Licensing Branch A (APLA)

Reviewer: Marty Stutzke

1. Supplement 38, Attachment 1, page 9: Provide the engineering assessment that shows that the residual heat removal (RHR) and core spray (CS) pumps can operate at significantly reduced net positive suction head (NPSH) compared to the design NPSH, which is based on the results of tests conducted at Browns Ferry as described in NUREG/CR-2973. Have the conclusions of this engineering assessment been discussed with the pump manufacturer (Sulzer Bingham)? If so, does the pump manufacturer concur with the conclusions?

2. Supplement 38, Attachment 1, page 10: The NPSH testing conducted at Browns Ferry indicates that the RHR pumps can operate without significant vibration at roughly 60%

to 70% of the vendor recommended required NPSH for flow rates near the pump design value. Further, the RHR pump head would be reduced about 12%, but the pump would still be operating above the "knee" of the pump curve. Based on the Individual Plant Examination (IPE) submittal for VYNPS, the NRC staff understands that the success criteria for low pressure coolant injection (LPCI) is one RHR pump. When this success criteria was determined, how was the reduction in RHR pump head that occurs when the pump is cavitating considered?

3. Supplement 38, Attachment 1, page 19: It is stated that the maximum leak size was determined to be 27 x La, using the conservative Title 10 of the Code of Federal Regulations Part 50, Appendix K containment analysis. Using more realistic assumptions typical of PRA, what is the leak size required to cause inadequate containment overpressure? In terms of La, what is the definition of 'large" used in the determination of the large early release frequency (LERF)? Note that Supplement 39, Attachment 1, page 20, Section 3.4.1, first paragraph indicates that the maximum leak size is approximately 60 x La.

4. Supplement 38, Attachment 1, page 19 and Supplement 39, Attachment 1, pages 12 and 13: It is stated that EPRI TR-1009325 was used to determine the probabilities of containment pre-existing leakage. The NRC staff has not yet accepted this reference as a technical basis for granting permanent 15-year integrated leak rate test (ILRT) intervals. In fact, the Nuclear Energy Institute (NEI) submitted an updated version of this document for further staff review on October 26, 2005. The staff notes that the technical basis for containment leakage probabilities used to justify the one-time 15-year ILRT interval that was granted in VYNPS Amendment No. 227, dated August 31, 2005, was EPRI TR-1 04285, and that the containment leakage probabilities in this report are notably higher than those provided in EPRI TR-1 009325. Either justify the use of EPRI TR-1009325 as an acceptable source of containment leakage probabilities, or reassess the change in core-damage frequency (CDF) caused by crediting containment accident pressure using containment leakage probabilities that are consistent with the recently granted one-time 15-year ILRT interval.

5. Supplement 38, Attachment 1, page 20 and Supplement 39, Attachment 1, page 22:

Provide the high confidence of low probability of failure (HCLPF) values used in the Seismic Margins Analysis (SMA) of VYNPS for the following: reactor coolant system piping, reactor vessel supports, safety relief valves (SRVs), and the containment.

6. Supplement 38, Attachment 1, page 20 and Supplement 39, Attachment 1, page 22:

Could a fire simultaneously cause a stuck-open relief valve and a failure of the containment isolation (Cl) system?

7. Supplement 39, Attachment 1, General: Is the overall intent of the risk evaluation of the proposed containment overpressure credit is to provide a sensitivity analysis that investigates modeling uncertainty in the baseline post-EPU PRA? The NRC staff notes that Supplement 38 indicates no overpressure credit is required using realistic assumptions. Hence, there should be no changes between the pre-EPU and post-EPU PRA models with respect to their treatment of the proposed overpressure credit.

8. Supplement 39, Attachment 1, General: Does the change in CDF only consider the impact of the proposed overpressure credit, or does if also include the impact of other changes resulting from the proposed EPU (e.g., shorter operator times due to higher decay heat)?

9. Supplement 39, Attachment 1, page 13: It is stated that containment integrity (Event IP) is considered when the hardened torus vent is being used (Event VT) to prevent over-pressurization failure of the containment following a loss of torus cooling (Event TC). It is difficult to interpret the event tree logic (e.g., the large LOCA event tree) in the context of this statement since Event IP appears before Event VT. To help clarify the NRC

staff's understanding of the modeling approach taken, provide a narrative explanation of each core-damage sequence in the large LOCA event tree.

10. Supplement 39, Attachment 1, page 14: If the containment is not intact (Event IP occurs), why is it possible to credit alternative injection and containment overpressure (COP) control (Event Al)?

11. Supplement 39, Attachment 1, page 18 and Tables 3.2A and 3.3: On page 18, it is stated that CONFIG#1 represents the risk when the COP is not available and CONFIG#2 represents the risk when the COP is available. However in Tables 3.2A and 3.3, the CDF associated with CONFIG#1 is lower than for CONFIG#2. Please clarify.

Also, note that in Table 3.3, the total CDF for CONFIG#1 is incorrect (typographical error).

12. Supplement 38, Attachment 1, page 18: It is stated that the only difference between the model cases lies in Endstate Bin IIV. However, Table 3.2A indicates that Endstate Bins ID, IIIC, IVA, and IC also change. Please clarify.

13. Supplement 39, Attachment 1, page 18: What is the impact of the proposed containment overpressure credit on the conditional containment failure probability?