3F1212-03, Response to Request for Additional Information Regarding 10 CFR 50.46 Notification of Change in Peak Cladding Temperature for Large Break Loss of Coolant Accident Analysis

Response to Request for Additional Information Regarding 10 CFR 50.46 Notification of Change in Peak Cladding Temperature for Large Break Loss of Coolant Accident Analysis

ML12352A283
Person / Time
Site:	Crystal River
Issue date:	12/10/2012
From:	Wunderly B Duke Energy Corp, Florida Power Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F1212-03, TAC ME8408
Download: ML12352A283 (5)

Text

PDuke V Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10CFR50.46 December 10, 2012 3F1212-03 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Request for Additional Information Regarding 10 CFR 50.46 Notification of Change in Peak Cladding Temperature for Large Break Loss of Coolant Accident Analysis (TAC No.ME 8408)

References:

1. CR-3 to NRC letter, 3F0312-04, dated March 19, 2012, "Crystal River Unit 3 - 10 CFR 50.46 Notification of Change in Peak Cladding Temperature for Large Break Loss of Coolant Accident Analysis" (Accession No.

MU12081A278)

2. Email from F. Saba (NRC) to Dan Westcott (CR-3) dated October 26, 2012, "RAIs regarding Cr stal River 30-day report for ECCS model changes pursuant to 10-CFR 50.46 requirements (ME8408)" (Accession No. ML12304A068)

3. AREVA to NRC letter, AREVA - ANP-3180, dated December 6, 2012, "177 Fuel-Assembly Plant RAI Responses to a 30-Day 50.46 Report of Significant PCT Change"

Dear Sir:

By letter dated March 19, 2012, Florida Power Corporation (FPC) submitted a 30-day report, pursuant to 10 CFR 50.46(a)(3)(ii), regarding the impact on Peak Cladding Temperature from two errors in the Emergency Core Cooling System (ECCS) evaluation model used to assess a postulated Large Break Loss of Coolant Accident for Crystal River Unit 3 (CR-3) (Reference 1).

This information is specific to the application of the AREVA ECCS evaluation model for Babcock & Wilcox plants as applied to CR-3.

On October 26, 2012, via electronic mail, the NRC provided a request for additional information (RAI) regarding the March 19, 2012 submittal (Reference 2). This letter provides the response to the RAI.

In response to the first RAI question, AREVA has submitted a generic response to the NRC (Reference 3). FPC is providing a response to the second RAI question in the enclosure to this letter.

This correspondence contains no new regulatory commitments.

If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Superintendent, Licensing and Regulatory Programs at (352)563-4796.

Sincerely, lair derly Director-Engineering-Nuclear Crystal River Nuclear Plant BPW/sam.& a Crystal River Nuclear Plant 15760 W. Power Line Street Crystal River, FL 34428

U. S. Nuclear Regulatory Commission Page 2 of 2 3F1212-03

Enclosure:

Response to Request for Additional Information - 10 CFR 50.46 Notification of Change in Peak Cladding Temperature for Large Break Loss of Coolant Accident Analysis xc: NRR Project Manager Regional Administrator, Region II Senior Resident Inspector

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 ENCLOSURE RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

- 10 CFR 50.46 NOTIFICATION OF CHANGE IN PEAK CLADDING TEMPERATURE FOR LARGE BREAK LOSS OF COOLANT ACCIDENT ANALYSIS

U. S. Nuclear Regulatory Commission Enclosure 3F1212-03 Page 1 of 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - 10 CFR 50.46 NOTIFICATION OF CHANGE IN PEAK CLADDING TEMPERATURE FOR LARGE BREAK LOSS OF COOLANT ACCIDENT ANALYSIS By letter dated March 19, 2012 (Reference 1), Florida Power Corporation, the licensee for Crystal River Unit 3 (CR-3) Nuclear Generating Plant submitted a notice reporting a change or error discovered in an evaluation model or in the application of such a model that affects the peak cladding temperature (PCT) calculation. This report was submitted pursuant to the requirements of 10 CFR 50.46, which requires, in part, that licensees report a change in the evaluation model used resulting in a significant change in PCT (greater than 50'F). The intent of this requirement is to enable the staff to establish the safety significance of this change (See FR Volume 53, No. 0180, pp. 35996-36005).

The NRC staff has reviewed the licensee's 30-day report and requests the licensee respond to the following questions:

REOUEST FOR ADDITIONAL INFORMATION (RAI) 1.

There are two changes to PCT for Large Break Loss of Coolant Accident (LBLOCA) analysis discussed in the report submitted by the licensee. The first change is an Evaluation Model (EM) application error in the determination of the end of Emergency Core Cooling System (ECCS) bypass which resulted in an 80'F decrease in PCT. The second change is an EM modeling change to include the effects of the upper plenum column weldments which resulted in an 80'F increase in PCT.

Provide the analysis that lead to each change having an 80 degree change in PCT.

Response

In response to the first RAI question, AREVA has submitted a generic response to the NRC (Reference 2).

RAI 2.

10 CFR 50.46(a)(3)(ii) states: ... If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 50.46 requirements ... "

The PCT for LBLOCA for Oconee Units 1, 2, and 3 has changed by an absolute value of 160'F since the analysis was performed. Simply reporting the changes and errors in the methodology does not satisfy the intent of the regulation.

Justify not providing a schedule for reanalysis or taking other action to show compliance with Section 50.46.

U. S. Nuclear Regulatory Commission Enclosure 3F1212-03 Page 2 of 2

Response

The response to RAI 1 provides additional detail regarding the analytical bases for the two PCT error estimates, which were based upon explicit RELAP5/Mod2-B&W code runs for the nuclear steam supply systems designed by Babcock & Wilcox (B&W). One error that was corrected in the evaluation models was specific to the determination of the end of the ECCS bypass time. A separate error correction to the ECCS evaluation model was made based upon the effects of the upper plenum column weldments.

As evidenced by the information provided in the response to RAI 1, both of these error corrections have been analyzed in detail. Furthermore, the error corrections in the ECCS evaluation model do not result in any challenge to the 10 CFR 50.46(b) acceptance criteria. As the individual error corrections have been identified for the applicable evaluation model, and there are no other known changes identified at this time, the overall evaluation model is considered bounding and complete.

The corrected ECCS evaluation model, as discussed in RAI 1, will be used for any future analyses.

In summary, the response to RAI 1 establishes the following:

" The error-adjusted LBLOCA PCTs for CR-3 remain below the 10 CFR 50.46(b) acceptance criteria. The two changes of 80'F offset each other with no net change to the current PCT. The resulting maximum PCT for the LBLOCA analysis is 1994°F (Reference 1).

• The Small Break Loss of Coolant Accident (SBLOCA) analysis is not affected by the ECCS evaluation model errors.

" The response provides additional information regarding the nature of the PCT error evaluations which are supported by explicit analyses using the B&W plant ECCS evaluation model.

• The analysis, with the identified corrections, is considered adequate to demonstrate compliance with the requirements of 10 CFR 50.46.

Based on the above, there are no adverse impacts to safety. Therefore, further LBLOCA reanalysis for CR-3 is not warranted.

REFERENCES

1. CR-3 to NRC letter, 3F0312-04, dated March 19, 2012, "Crystal River Unit 3 - 10 CFR 50.46 Notification of Change in Peak Cladding Temperature for Large Break Loss of Coolant Accident Analysis" (Accession No. ML12081A278)

2. AREVA to NRC letter, AREVA - ANP-3180, dated December 6, 2012, "177 Fuel-Assembly Plant RAI Responses to a 30-Day 50.46 Report of Significant PCT Change"