ML15238B443
| ML15238B443 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/17/1981 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| TASK-2.K.2.20, TASK-TM NUDOCS 8109020047 | |
| Download: ML15238B443 (7) | |
Text
AUGUST 1 798, Zk 3 E)
Dockets Nos. 50-269, 270 and 287 Mr. William 0. Parker, Jr.
Vice President - Steam Production Duke Power Company P. 0. Box 33189 Charlotte, North Carolina 28242
Dear Mr. Parker:
As-part of the NUREG-0737 Implementation Plan for Operating Reactors, the NRC has completed the review of Item II.K.2.20, System Response to Small Break LOCA, for the Oconee Nuclear Station.
Your submittal, which included the B&W letter, Reference 1, of the enclosed Safety Evaluation Report, provided sufficient information for us to conclude that small primary system breaks which result in a stuck open PORV, will not result in unanalyzed consequences even when assuming a concurrent sinale failure. We conclude Item II.K.2.20 is completed for your plant.
Sincerely, QORIGERL ZIGZDBY JO F. STOILTP John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing
Enclosure:
SER DLTRBUTI0N cc w/enclosure:
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.-C. 20555 August 17, 1981 Dockets Nos. 50-269, 270 and 287 Mr. William 0. Parker, Jr.
Vice President - Steam Production Duke Power Company P. 0. Box 33189 Charlotte, North Carolina 28242
Dear Mr. Parker:
As part of the NUREG-0737 Implementation Plan for Operating Reactors, the NRC has completed the review of Item II.K.2.20, System Response to Small Break LOCA, for the Oconee Nuclear Station. Your submittal, which included the B&W letter, Reference 1, of the encTosed Safety Evaluation Report, provided sufficient information for us to conclude that small primary system breaks which result in a stuck open PORV, will not result in unantlyzed consequences even when assuming a concurrent single failure. We conclude Item II.K.2.20 is completed for your plant.
Sincerely, John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing
Enclosure:
SER cc w/enclosure:
See next page
Duke Power Company cc w/enclosure(s):
Mr. William L. Porter Duke Power Company P. 0. Box 33189 422 South Church Street Office of Intergovernmental Relations Charlotte, North Carolina 28242 116 West Jones Street Raleigh, North Carolina 27603 Oconee County Library 501 West Southbroad Street Walhalla, South Carolina 29691 Honorable James M. Phinney County Supervisor of Oconee County Walhalla, South Carolina 29621 Reqional Radiation Representative EPA Region IV 345 Courtland Street, N.E.
Atlanta, Georgia 30308 Mr. Francis Jape U.S. Nuclear Regulatory Commission Route 2, Box 610 Seneca, South Carolina 29678 Mr. Robert B. Borsum Babcock & Wilcox NucTear Power Generation Division Suite 420, 7735 Old Georgetown Road Bethesda, Maryland 20014 Manager,, LIS NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 J. Michael McGarry, III, Esq.
DeBevoise & Liberman 1200 17th Street, N.W.
Washington, D. C. 20036
TASK ACTION ITEM II.K.2.20 SAFETY EVALUATION REPORT In a letter from 0. F. Ross, dated August 21, 1979, the NRC informed all licensees.
with Babcock and Wilcox reactors of the ACRS ECCS Subcommittee concern that B&W plants have not been analyzed to withstand postulated small breaks which result in system repressurization to the PORV set point.
These analyses would assume that the PORV remained stuck open for the remainder of the transient.
System repressurization may occur during a small break in plants with B&W NSSSs via the following means:
- 1) Loss of heat sink (i.e., loss of auxiliary feedwater),
- 2) HPI flow exceeds break flow, and
- 3) Loss of natural circulation.
2 In response to the NRC concerns, B&W analyzed a 0.01 ft cold leg break concurrent with loss of auxiliary feedwater. Previous analyses (Reference-1) have demonstrated that break areas greater than 0.01 square feet (concurrent with loss-of-auxiliary 2
feedwater) will not pressurize the system to the PORV set point. The 0.01 ft break is considered limiting in that it maximizes inventory depletion.
The analyses assumed that once actuated, the PORV stuck in the open position. The PORV in Jhe stuck open position was analyzed to have su;ficient capacity Lo depressu. ize the system. For this bounding analysis, the mixture level within the reactor vessel dropped to a minimum of three feet above the top of the core.
Core uncovery was prevented by the two high pressure injection pumps.
Less limiting events which could pressurize the primary system to the PORV.
setpoint include breaks which do not exceed the HPI capacity. These breaks 2
are typically much less than 0.01 ft area, and are less severe than the one discussed above, due to less inventory loss.
Breaks less than 0.01 ft2 (without assuming a stuck-open PORV) have been analyzed and documented in Reference 1.
-2 In addition to the initiating events described above, repressurization could also occur as a result of interrupting natural circulation.
Temporary interruption of natural circulation can occur when the apex of the hot leg "candy-cane" collects a sufficient volume of steam to interrupt flow to the steam generator.
The system is predicted to pressurize until the break drains the steam sufficiently to uncover a condensing surface in the steam generators. Once steam can be condensed, the system will depressurize.
Based on previous analyses (Reference 1) it is concluded that no break size in the primary system will result in opening of the PORV as long as the auxiliary feedwater is available.
Based on our review of the B&W licensee's submittals, we conclude that small primary system breaks which result in the opening of a PORV (assumed to stick open) will not provide adverse consequences, when assuming a single failure.
As such, we consider Item II.K.2.20 completed by issuance of this SER. Moreover, we do not believe it necessary for Item II.K.2.20 to be addressed by present and future applicants as a licensing condition.
Reference:
Reference 1:
Letter J.H. Taylor (B&W) to S. A. Varga (NRC), "Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177-Fuel Assembly Plant," May 7, 1979._
ATTACHMENT 4 Item III.A.1.2 -
Staffing Levels for Emergency Situations Carolina Power & Light Company is currently preparing a revision to the H. B.
Robinson Unit 2 Emergency Plan to incorporate recommendations and correct deficiencies identified by the NRC during a recent Emergency Preparedness Appraisal. The response to Item II.A.1.2 will be included in this revision to the Robinson Emergency Plan and will be submitted to the NRC by July 14, 1982.
~.
0 ATTACHMENT 5 Item III.A.1.2 - Upgrade Emergency Support Facilities CP&L has been involved in discussions with the NRC Staff regarding the requirements for emergency response capability. Based on the NUREG-0696 requirements and SECY-82-111, CP&L submitted a letter dated April 9, 1982 requesting NRC concurrence with the location of our proposed emergency response facilities and the security provisions for including the Technical Support Center within the protected area of the plant. Based on Staff concurrence received in a letter dated May 19, 1982, CP&L is proceeding with preliminary engineering for the facilities. At present, we anticipate submitting a radiological habitability analysis for the facilities in July 1982 for Staff review and concurrence. Upon receiving Staff approval of the analysis, we will finalize the engineering design and begin construction of the facilities.