ML20115H177
| ML20115H177 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 07/16/1996 |
| From: | Duffy J VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BVY-96-89, GL-88-20, NUDOCS 9607220252 | |
| Download: ML20115H177 (4) | |
Text
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VERMONT YANKEE l'
NUCLEAR POWER CORPORATION I
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Ferry Road, Brattleboro. VT 05301-7002 REPLV TO ENGINEERING OFFICE 580 MAIN STREET j
BoLTON, MA 01740 i
(508) 779-6711 l
July 16,1996 BVY 96-89 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
References:
(a) License No. DPR-28 (Docket No. 50-271)
(b) Letter, VYNPC to USNRC, BVY 96-61, dated May 9,1996 (c) Letter, VYNPC to USNRC, BVY 96-76, dated May 10,1996 (d) Letter, VYNPC to USNRC, BVY 96-77, dated May 11,1996 (e) Letter, USNRC to VYNPC, NVY 96-106, dated June 20,1996 (f) USNRC Generic Letter 88-20, NVY 88-259, dated November 23,1988
Subject:
Response to Request for Additional Information Regarding Use of RELAP5YA-BWR j
Code for Vermont Yankee On May 30 and 31,1996, the NRC performed an audit of RELAP5YA-BWR code and the limiting case l
design basis accident analysis as part of the review of Reference (b), LER 96-10. In References (c) and I
(d) Vermont Yankee provided additional information regarding issues discussed during the audit. In Reference (e) the NRC requested additional information in order to complete its review. The requested information is attached.
We trust that the information provided is acceptable; however, should you have any questions, please contact this office.
l Sincerely, VERMONT YANKEE NUCLEAR POWER CORPORATION NfAd
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ames J. Du y l
9607220252 960716 l
PDR ADOCK 05000271 Licensing Engineer P
PDR c: USNRC Region I Administrator USNRC Resident Inspector-VYNPS
//[d/
f USNRC Project Manager-VYNPS
s VERMONT YANKEE NUCLEAR POWER CORPORATION United States Nuclear Regulatory Commission July 16,1996
. Attachment Page 1 of 3 Request for Additional Information Vermont Yankee Nuclear Power Station
. NRC Reauest for Additional Information #1 The Licensee is requested tojustify for the Cycle 16 emergency core cooling system evaluation performed with the methodology of General Electric's SAFE /REFLOOD model, the worst single failure scenario identified in LER %-010, including break size, break location, and assumptions, would not violate the peak cladding temperature limits of 10 CFR 50.46.
Vermont Yankee Resnonse to Reauest for Addi ionni Information #1 t
In Reference (c) Vermont Yankee provided the NRC with information regarding the impact of this recently identified single failure scenario on previous Vermont Yankee LOCA analyses. As described in Reference (c) the maximum peak cladding temperature (PCT) cases for the design basis LOCA analyses prior to Cycle 17 (intermediate-to-large sized breaks) would be unaffected by the availability of the RHR pumps due to the timing of the event and because the analyses assumed the RHR pumps were not available. For other scenarios, such as the one identified in LER 96-10, the SAFE /REFLOOD model is overly conservative. Using more realistic analysis techniques,(e.g. SAFER /GESTR), one core spray pump plus the automatic depressurization system (ADS) will most likely maintain adequate cooling.
Further, best estimate calculations using previous models (i.e. SAFE) showed acceptable results (i.e.
PCT < 2200 F) with one core spray pump and ADS for any break size and location.
N]LCleauest for Additional Information #2 The licensee is requested to address the root cause for not identifying the discrepancy between the plant operating configuration and the limiting case single failure assumptions for a period of twenty-two years.
Vermont Yankaa Re=nnnae to Reauest for Additional Information #2 The causes for not identifying the discrepancy between the plant operating configuration and the limiting case single failure assumptions for a period of twenty-two years are as follows:
- 1. The Vermont Yankee design change process is very specific and robust in its guidance regarding design input considerations. However, this robust proce:,s has only recently been extended to engineering evaluations ofindustry events.
- 2. ' For plants of Vermont Yankec's vintage, it is common that design basis documentation necessary for performing engineering evaluations exists in hard-copy format, indexed to specific systems. While this system has been used successfully for many years, the inability to perform key-word searches
. makes the process less efficient, potentially leading to missed opportunities in identifying problems in areas indirectly related to, or not specifically connected to, an issue being evaluated.
Design engineers have been instructed to use the guidance from the design change process when performing engineering evaluations ofindustry events. Vermont Yankee will prepare user-friendly design basis documentation packages for safety-significant systems as defined by the Individual Plant Examination (IPE). Failure modes and effects analyses will also be included for the appropriate systems
d VERMONT YANXEE NUCLEAR POWER CORPORATION
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United States Nuclear Regulatory Commission l
July 16,1996 l
Attachment l
Page 2 of 3 and programs. Clear management expectations have been established and reviewed with department l
personnel regarding the need to perform comprehensive, detailed, evaluations ofindustry events.
Additionally, the Vermont Yankee Event Report procedure requires a Potentially Similar Conditions l
Assessment for all Level 1,2, & 3 Event Reports. This assessment has found several similar conditions that have been addressed as part of our Root Cause Analysis process.
NRC Reauest for AdditionalInformation #3 The licensee is also requested to provide verification that all design basis accident scenarios identified through the Individual Plant Examination program have been accounted for in the licensing basis analyses.
Vermont Yankee Resoonse to Request for Additional Information #3 Design basis accident scenarios, including the design basis accident scenario identified in LER 96-10, were not identified in the IPE. Rather, the IPE identified and analyzed " severe" accident scenarios which are, by definition, beyond the design basis. His is consistent with the purposes of the IPE outlined in Generic Letter 88-20, Reference (f).
While the IPE was not intended to identify or analyze design basis accident scenarios, the IPE can provide insight into the risk associated with design basis accident scenarios. He insight comes from noting that the IPE's beyond design basis accident scenarios are design basis accident scenarios involving additional (beyond design basis) failures.
The design basis scenario ofinterest here is:
- 1. a small or intermediate LOCA, and
- 2. a single power supply failure that causes direct failure of one Core Spray, the HPCI system and two LPCI pumps, and the potential indirect failure of a third LPCI pump (via failure of the minimum flow valve powered by the failed power supply).
His scenario does not cause core damage in the IPE because one Core Spray pump (and possibly one LPCI pump, depending on the break size and location) remains available for injection. The IPE risk significance of this scenario depends on the frequency of the scenario and on the conditional probability (given the scenario) that core damage will occur as a result of additional (beyond design basis) failures. The IPE risk significance is summarized by the following points:
1.
The LPCI minimum flow valves and their power supply dependencies were modeled in the IPE. The IPE assumed that failure of a LPCI minimum flow valve to open (due to either valve mechanical failure or power supply failure) would cause failure of the associated two LPCI pumps (due to deadheaded operation) for small and intermediate LOCA initiating events.
- 2. The IPE calculated a total Core Damage Frequency of 4.3 E-06 per year. Of this total, only 1.3E-07 per year is due to LOCA events (including inadvertent / stuck open safety relief valve events) where core damage is caused by loss of ECCS injection. Of this 1.3E-07, about 1.0E-07 is attributed to small and
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VERMONT YANKEE NUCLEAR POWER CORPORATION United States Nuclear Regulatory Commission July 16,1996 Attachment Page 3 of 3 intermediate LOCAs. Thus, the total risk of core damage from LOCA events is small in both relative l
(about 2% of CDF) and absolute (about E-07 per year) terms.
- 3. Vermont Yankee examined the detailed event sequences (i.e. combinations of system failures) for all.
small and intermediate LOCA sequences with a frequency greater than 1E-09 per year. None of these l
j sequences involved the subject failure mode (i.e., power supply failure causing failure of a Core Spray pump, the HPCI system and three LPCI pumps, where failure of one or more LPCI pumps is caused by minimum flow valve failure) as a contributing cause of core damage.
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l Because the subject failure mode was not risk significant in the IPE, the IPE did not identify this failure mode (or associated accident sequences) as warranting further attention.
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