ML043570342

Email Ennis, NRR, to Daflucas, Entergy, VY EPU Draft Rais, Round 5

ML043570342
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	12/07/2004
From:	Richard Ennis Office of Nuclear Reactor Regulation
To:	Daflucas R Entergy Nuclear Northeast
References
Download: ML043570342 (12)

Text

Rick Ennis - VY EPU Draft RAIs Page I Rick Ennis - VY EPU Draft RAts Page 1 

From:

Rick Ennis To:

Ronda Daflucas Date:

Tue, Dec 7, 2004 2:35 PM

Subject:

VY EPU Draft RAts

Ronda, Attached are the draft round 5 RAts for the EPU amendment request. Please let me know your response timneframe ASAP so I can finalize the RAI cover letter.

Also, let me know if you would like to schedule a conference call with the reviewers.

thanks, Rick 301-415-1420 CC:

Allen Howe; Devender Reddy; Donna Skay; John Boska; Richard Lobel; Zena Abduttahi

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VY EPU Draft RAIs Tue. Dec 7, 2004 2:35 PM Rick Ennis RXE@nrc.gov Recipients owf2_po.OWF'NDO ZXA CC (Zena Abdullahi) owf4_po.OWFN_DO RXE BC (Rick Ennis)

Action Delivered Opened Delivered Opened Date & Time 12/07 2:35 PM 12/07 3:58 PM 12/07 2:35 PM 12/07 2:32 PM entergy.com rdafluc (Ronda Daflucas)

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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 on 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, and October 7, 2004 (2 letters), (References 1 through 21), Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (Entergy or the licensee), submitted a proposed license amendment to the U.S. Nuclear Regulatory Commission (NRC) for the Vermont Yankee Nuclear Power Station (VYNPS). The proposed amendment, "Technical Specification Proposed Change No. 263, Extended Power Upratew 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 in the following request for additional information (RAI).

Enclosure

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Plant Systems Branch (SPLB)

Balance of Plant Section (SPLB-A)

Reviewer: Devender Reddy

10.

EPU Transient Testing As discussed in the NRC's "Review Standard for Extended Power Uprates," RS-001, Safety Evaluation template Section 2.12, "Power Ascension and Testing Plan," the purpose of the EPU test program is to demonstrate that structures, systems, and components (SSCs) will perform satisfactorily in service at the proposed EPU power level. The test program also provides additional assurance that the plant will continue to operate in accordance with design criteria at EPU conditions. The NRC's acceptance criteria for the proposed EPU test program are based on 10 CFR Part 50, Appendix B, Criterion Xl, which requires establishment of a test program to demonstrate that SSCs will perform satisfactorily in service. Specific review criteria are contained in NUREG-0800, Standard Review Plan (SRP) Section 14.2.1, "Generic Guidelines for Extended Power Uprate Testing," Draft Revision 0, dated December 2002.

SRP Section 14.2.1 directs the NRC staff to assess the adequacy of the licensee's evaluation of the aggregate impact of EPU plant modifications, setpoint adjustments, and parameter changes that could adversely impact the dynamic response of the plant to anticipated operational occurrences. The staff's review is intended to ensure that the performance of plant equipment important to safety that could be affected by integrated plant operation or transient conditions is adequately demonstrated prior to extended operation at the requested EPU power level. Licensees may propose a test program that does not include all of the power-ascension testing that would normally be included in accordance with the guidance provided in SRP 14.2.1, provided each proposed test exception is adequately justified. If a licensee proposes to omit a specified transient test from the EPU testing program based on favorable operating experience, the applicability of the operating experience to the specific plant must be demonstrated. Further, the licensee shall address the potential for any new thermal-hydraulic phenomena or system interactions that may be introduced as a result of the EPU or planned modifications.

Also, if the basis for elimination of a transient test relies on the use of analytical methods, the licensee should address the conformance to limitations associated with the analytical methods. Plant design details (such as configuration, modifications, and relative changes in setpoints and parameters), equipment specifications, operating power level, test specifications and methods, operating and emergency operating procedures; and adverse operating experience from previous power uprates must be considered and addressed.

Entergy's test program primarily includes steady state testing with some minor load changes, and no large-scale transient testing is proposed. Sufficient information has not been provided to demonstrate that in the absence of large-scale transient testing, the integrated plant response during transient conditions will be as expected. Entergy is therefore requested to either: a) provide additional information in accordance with the guidance provided in SRP Section 14.2.1 that explains in detail how the proposed EPU startup and power ascension test program, in conjunction with the original test results

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and applicable industry experience, assures the plant will respond as expected during postulated transient conditions following implementation of the proposed EPU given the revised operating conditions that will exist and plant changes that are being made; or b) describe transient testing that will be included in the power ascension test program in order to provide this assurance, and explain in detail how the proposed transient testing will demonstrate that SSCs will perform satisfactorily in service at the proposed EPU power level.

11.

Followup on Response to RAI SPLB-A-7, Item b (Spent Fuel Pool Cooling - Heat Removal Capability and Limiting Case for Core Off load)

The licensee was requested to address the limiting cases for normal batch off load and full core offload in accordance with the Updated Final Safety Analysis Report (UFSAR),

Section 10.5.5, which states: "Considering one train (one heat exchanger and one pump), this heat removal capability encompasses the normal maximum heat load from completely filling the pool with 3,353 spent fuel assemblies from the last normal discharge..."

In its response the licensee stated that, "...., the configurations presented in the VYNPS UFSAR, Section 10.5, Page 10.5-9, present scenarios more conservative than SRP Section 9.1.3 in that the batch offload configuration assumes more than a single failure (failure of both the NFPCS [normal fuel pool cooling system ] trains and the failure of one SFPCS)...'

The licensee has not adequately considered and addressed the plant licensing basis as reflected in the UFSAR. Because the NFPCS is not safety-related, it is not credited in the limiting case. This is consistent with the guidance provided in SRP 9.1.3.

Therefore, the licensee is requested to address the question as originally posed by the staff in RAI SPLB-A-7, Item b.

12.

Followup on Response to RAI SPLB-A-3 (Turbine Overspeed)

The increase in main steam flow rate and rotor inertia considerations increase the likelihood that the main turbine speed will overshoot and exceed design specifications during postulated events. Identify the worst-case scenario that could lead to main turbine overspeed and discuss in detail what measures will been taken to assure that this condition will not occur, including testing that will be completed to confirm that the combination of increased main steam flow and inertial effects will not cause the turbine design specifications to be exceeded.

13.

High Energy Line Breaks (HELBs)

Referring to Section 10.1.2 of NEDC-33090P, additional discussion is needed to explain why safety-related SSCs will not be affected due to postulated HELBs at the proposed EPU conditions. Specifically, the section titled "Liquid Line Breaks" should state a conclusion regarding the ability of safety-related SSCs to perform as intended at the proposed EPU conditions.

Probabilistic Safety Assessment Branch (SPSB)

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Containment and Accident Dose Assessment Section (SPSB-C)

Reviewer: Richard Lobel

34.

What is the temperature criterion for piping attached to the torus? Is this criterion satisfied under power uprate conditions?

35.

The response to RAI SPSB-C-1 provided in Attachment 2 to Supplement 8 indicates that pumps taking suction from the suppression pool have adequate net positive suction head (NPSH) without requiring credit for containment accident pressure when best-estimate assumptions are used. Please submit the thermal-hydraulic analyses (both the containment response analysis and the NPSH calculation) that support this statement.

36.

(a)

The licensee's December 29,1999, letter to the NRC, concerning installation of larger emergency core cooling system suction strainers in accordance with NRC Bulletin 96-03, stated that the head loss correlation of NUREG/CR 6224 was used. Verify that this correlation was used within its range of applicability considering debris materials present in the VYNPS containment, bed thickness, suppression pool temperature, and approach velocity.

(b)

The head loss due to loss-of-coolant accident (LOCA)-generated debris appears to be low. Please provide the calculation of head loss, VYC-1 924, Revision 0 (Reference 2 to calculation VYC-0808, Revision 8).

37.

What is the coping period for station blackout (SBO)? Is the NPSH analysis for SBO consistent with this?

38.

Please provide the results of the EPU Appendix R and SBO containment analyses.

These analyses were included in Revision 6 of calculation VYC-0808 but are not included in Revision 8 of this calculation.

39.

Please provide for NRC review calculation GE-VYNPS-AEP-346, Revision 2 (Reference 10 to calculation VYC-0808, Revision 8).

40.

Staff calculations indicate that it is not necessary to credit the reduced values of required NPSH given in Curves E12.5.522-1 B and E12.5.522-2B in Attachment 5 of calculation VYC-080,8 Revision 8. Using the long-term required NPSH values given in Table SPSB-C-12-1 (from the July 2, 2004 response to staff RAls), the containment accident pressure that must be credited is slightly higher than the values calculated in Tables 4.1, 4.4 and 4.5, but still sufficiently below the conservatively calculated pressure shown to be available. Please justify use of the reduced values of required NPSH and assess the use of the values given in Table SPSB-C-12-1 of your July 2, 2004, letter instead.

41.

Describe the worst-case single failure assumed for the NPSH calculations and the basis for this assumption.

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42.

Do the reactor pressure vessel safety valves that discharge into the drywell atmosphere open during an anticipated transient without scram (ATWS)? If so, please explain why no debris is assumed to be generated during an ATWS.

43.

Explain why the emergency operating procedure (EOP) NPSH curves are still valid without change for the power uprate conditions. Are accident-generated debris included in the calculations? Is credit taken for the minimum available NPSH shown in curves on pages 18 and 19 of 19 of Attachment 5 to calculation VYC-0808, Revision 8? Were the EOP curves calculated with the same computer program used to calculate the temperatures and pressures used in VYC-0808, Revision 8?

44.

The response to SPSB-C-1 0, dated July 2, 2004, contains a calculation which shows that with two heat exchangers operating but all other conservative assumptions of the licensing basis calculation unchanged, the suppression pool temperature is reduced from 194 F to 169 F. Is the flow through each heat exchanger due to just one residual heat removal (RHR) pump and one service water pump? Under what conditions would the operator actually use both trains of RHR to cool the suppression pool as opposed to using one train to cool the suppression pool and one train to inject water into the reactor vessel? The RAI response states that the calculation was not performed to QA program requirements. The staff requests that this calculation be verified according to the VYNPS Appendix B program.

45.

The minimum required NPSH values recommended by the pump vendor were based on operating conditions supplied by the licensee (Page 6 of 19 of Attachment 5 and Page 8 of 58 of calculation VYC-0808, Revision 8). If these suppression pool temperature values were based on pre-power uprate temperatures, why are the recommended times at minimum required NPSH still valid?

46.

Calculation VYC-0808, Revision 8, Attachment 5, Page 7 of 19, states that the RHR pumps were run for only a few minutes at reduced NPSH. Please explain why this is sufficient time to observe pump behavior at reduced NPSHA, as stated in the Attachment.

47.

Regarding calculation VYC-0808 Revision 8, Attachment 5 Page 8 of 19, and Page 6 of 58, Section 2.1, what is the "minimum operational NPSH"?

48.

Regarding calculation VYC-0808 Revision 8, Page 8 of 58, what is the basis for the limit of 8000 hours0.0926 days <br />2.222 hours <br />0.0132 weeks <br />0.00304 months <br /> on impeller life? What is the licensing basis time the pumps must operate after a postulated design-basis LOCA? To what measured percentage reduction in pump discharge head does the value of minimum available NPSH after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> correspond?

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49.

Calculation VYC-0808, Revision 8 indicates that for the short-term (within 10 minutes)

LOCA analysis, it was assumed there was no operator action to initiate suppression pool cooling or to secure or throttle ECCS pumps. For the long-term (after 10 minutes)

LOCA analysis, it was assumed that the operators initiate torus cooling, all but one Core Spray pump and one RHR pump are secured, and that the Core Spray pump is throttled to 3,500 gpm. What is the impact on available NPSH if the operator does not take action as assumed or if he/she delays in taking action? How long could the operators delay and still be successful? Please discuss both the conservative design-basis analysis and the best-estimate analysis used to support the PRA.

50.

Regarding calculation VYC-0808, Revision 8, Table 4.4, please verify that the interpolated NPSHr value of 24.8 ft for Q= 3500 gpm is correct.

51.

Regarding the response to RAI SPSB-C-27, why is it considered acceptable to credit non-safety related systems such as main feedwater and turbine bypass valves in analyzing the stuck open reactor vessel relief valve? What would be the suppression pool temperature if only safety related equipment were credited to depressurize and shut down the reactor?

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Reactor System Branch (SRXB)

Boiling Water Reactors and Nuclear Performance Section (SRXB-A)

Reviewer: Zena Abdullahi Note: The RAI dated May 28, 2004, included 3 questions from SRXB-A designated as questions 1, 2, and 3. The RAI dated September 1, 2004, included 2 questions from SRXB-A designated as questions 1 and 2. The September 1, 2004, questions should have been designated as questions 4 and 5. Therefore, this RAI is designated as SRXB-A question 6.

6.

Table 1-1 in Attachment 6 of the application dated September 10, 2004, lists all the nuclear steam system codes used for the EPU request. Section 1.2.2 of Attachment 6, "Computer Codes," indicates that the VYNPS application of these codes complies with the limitations, restrictions, and conditions specified in the applicable NRC safety evaluation report (SER) that approved each code, with exceptions as noted in Table 1-1.

Similarly, review the fuel vendor's analytical methods and code systems used to perform the safety analyses supporting the VYNPS EPU application and provide the following information:

(a)

Confirm that the steady state and transient neutronic and thermal-hydraulic analytical methods and code systems used to perform the safety analyses supporting the EPU conditions are being applied within the NRC-approved applicability ranges.

(b)

Confirm that for the EPU conditions, the calculational and measurement uncertainties applied to the thermal limits analyses are valid for the predicted neutronic and thermal-hydraulic core and fuel conditions.

(c)

Confirm that the assessment database and the assessed uncertainty of models used in all licensing codes that interface with and/or are used to simulate the response of VYNPS during steady state, transient or accident conditions remain valid and applicable for the EPU conditions.

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8 REFERENCES

1)

Entergy letter (BVY 03-80) to NRC dated September 10, 2003, nVermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Extended Power Uprate"

2)

Entergy letter (BVY 03-90) to NRC dated October 1, 2003, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 1, Extended Power Uprate -Technical Review Guidance"

3)

Entergy letter (BVY 03-95) to NRC dated October 28, 2003, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 2, Extended Power Uprate - Grid Impact Study"

4)

Entergy letter (BVY 03-98) to NRC dated October 28, 2003, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 3, Extended Power Uprate - Updated Information"

5)

Entergy letter (BVY 04-009) to NRC dated January 31, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 4, Extended Power Uprate - NRC Acceptance Review"

6)

Entergy letter (BVY 04-008) to NRC dated January 31, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 5, Extended Power Uprate - Response to Request for Additional Information"

7)

Entergy letter (BVY 04-025) to NRC dated March 4, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 6, Extended Power Uprate - Withholding Proprietary Information"

8)

Entergy letter (BVY 04-050) to NRC dated May 19, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 7, Extended Power Uprate - Confirmatory Results"

9)

Entergy letter (BVY 04-058) to NRC dated July 2, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 8, Extended Power Uprate - Response to Request for Additional Information"

10)

Entergy letter (BVY 04-071) to NRC dated July 27, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 9, Extended Power Uprate - Revised Containment Overpressure Envelope"

11)

Entergy letter (BVY 04-074) to NRC dated July 30, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 10, Extended Power Uprate - Response to Request for Additional Information"

12)

Entergy letter (BVY 04-081) to NRC dated August 12, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 11, Extended Power Uprate - Response to Request for Additional Information"

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13)

Entergy letter (BVY 04-086) to NRC dated August 25, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 12, Extended Power Uprate - Revised Grid Impact Study"

14)

Entergy letter (BVY 04-097) to NRC dated September 14, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 13, Extended Power Uprate - Response to Steam Dryer Action Items"

15)

Entergy letter (BVY 04-098) to NRC dated September 15, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 14, Extended Power Uprate - Response to Request for Additional Information"

16)

Entergy letter (BVY 04-100) to NRC dated September 23, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 15, Extended Power Uprate - Response to Steam Dryer Action Item No. 2"

17)

Entergy letter (BVY 04-101) to NRC dated September 30, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 16, Extended Power Uprate - Additional Information Related to Request for Additional Information EMEB-B-5"

18)

Entergy letter (BVY 04-107) to NRC dated September 30, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 17, Extended Power Uprate - Response to Request for Additional Information related to 10 CFR 50 Appendix R Timeline"

19)

Entergy letter (BVY 04-106) to NRC dated October 5, 2004, Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 18, Extended Power Uprate - ECCS Pump Net Positive Suction Head Margin"

20)

Entergy letter (BVY 04-109) to NRC dated October 7, 2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 19, Extended Power Uprate - Initial Plant Test Program"

21)

Entergy letter (BVY 04-113) to NRC dated October 7,2004, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 20, Extended Power Uprate - Meeting on Steam Dryer Analysis"