ML15238A676
| ML15238A676 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/14/1982 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Tucker H DUKE POWER CO. |
| Shared Package | |
| ML15238A677 | List: |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8211060494 | |
| Download: ML15238A676 (9) | |
Text
October 14, 1982 Dot 1 N:
DEisenhut AEOD
- oktPWagner IE-Taylor & Jordan IqRCPDR RIngram ACRS-10 L PDR ETourigny Gray File ORB#4 Rdg OELD LSchneider Dockets Nos. 50-269, 50-270 and ASLAB 50-287 DBrinkman TBarnhart-1 2 RDiggs CMiles Mr. H. B. Tucker, Vice President Nuclear Production Department Duke Power Company P. 0. Box 33189 422 South Church Street Charlotte, North Carolina 28242&
Dear Mer. Tucker:
The staff and our consultant, Franklin Research Center, have completed our review of your analysis of a main steam line break with continued feedwater addition for the Oconee Nuclear Station. The results of our review are contained in the enclosed Safety and Technical Evaluation Reports.
Your submittals dated May 7, 1980 and July 23, 1982 on this subject were in response to IE Bulletin 80-04 issued February 8, 1980.
Based on our review, we conclu de that your analysis is acceptable and no further action is required of you regarding this subject.
Sincerely,
'IYRIGINAL SmiwzyB John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing
Enclosures:
- 1. Safety Evaluation Report
- 2. Technical Evaluation Report cc w/enclosures:
See next page pDR DOCK0500269
- ORB#4:D 4b C -AlRB 4: DL ORB#3:DLT OFFICE j?,lid_ _
SURNAME 'Joh F.
Stolz,/c Chief29V DATE.....................................................
NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960
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 Honorable James M. Phinney County Supervisor of Oconee County Walhalla, South Carolina 29621 Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission, Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Regional Radiation Representative EPA Region IV 345 Courtland Street, N.E.
Atlanta, Georgia 30308 William T. Orders Senior Resident Inspector U.S.
Nuclear Regulatory Commission Route 2, Box 610 Seneca, South Carolina 29678 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 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
UNITED STATES S"
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 October 14, 1982 Dockets Nos. 50- 29, 50-270 and 50-287 Mr. H. B. Tucker, Vi e President Nuclear Production '5 partment Duke Power Company P. 0. Box 33189 422 South Church Street Charlotte, North Carolina 8242
Dear Mr. Tucker:
The staff and our consultant, Fra 'In Research Center, have completed our review of your analysis of in steam line break with continued feedwater addition for the 0 nee N lear Station. The results of our review are contained '
the enc1 ed Safety and Technical Evaluation Reports. Your submitt dated May 7, 980 and July 23, 1982 on this subject were in resp e to IE Bulletin 0-04 issued February 8, 1980.
Based on our revi w, we conclude that your nalysis is acceptable and no further acti is required of you regardi g this subject.
i
- ncerely, ohn F. Stolz, Chie pe ating Reactors B nch #
ision of Licensin
Enclosures:
- 1. Safety Evaluation Report
- 2. Technical Evaluation Report cc w/enclosures:
See next page
SAFETY EVALUATION REPORT CONTAINMENT SYSTEMS BRANCH PAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION OCONEE NUCLEAR PLANT UNIT 1, 2 AND 3 Docket No.: 50-269, -270, -287
1.0 INTRODUCTION
In the summer of 1979, a pressurized water reactor (PUR) Licensee submitted a report to the NRC that identified a.deficiency in its original analysis of containment pressurization resulting from a postulated main steam Line break (MSLB). A reanalysis of the containment pressure response foLLowing a MSLB was performed, and it was determined that, if the auxiLiary feedwater (AFW) system continued to supply feedwater at runout conditions to the steam generator that had experienced the steam Line break, the containment design pressure would be exceeded in approximateLy 10 minutes. In other words, the Long-term blowdown of the water supplied by the AFW system had not been considered in the earLier analysis.
On October 1, 1979, the foregoing information was provided to aLL holders of operating Licenses and construction permits in IE Information Notice 79-24 E23.
Another Licensee performed an accident analysis review pursuant to the information furnished in the above cited notice and discovered that, with offsite electrical power availabLe, the condensate pumps would feed the affected steam generator at an excessive rate. This excessive feed had not been considered in the anaLysis of the postuLated MSLB accident.
A third Licensee informed the NRC of an error in the MSLB analysis for their plant.
For a zero or Low power condition at the end of core Life, the Licensee identified an incorrect postulation that the startup feedwater controL valves would remain positioned "as is" during the transient. In reality, the startup feedwater controL valves wiLL ramp to 80% fuLL open due to an override signal resuLting from the Low steam generator pressure reactor trip signal.
Reanalysis of the events showed that the rate of feedwater addition to the affected steam gene rator associated with the opening of the startup valve wouLd cause a rapid reactor cooLdown and resultant reactor-return-to-power response, a condition which is beyond the plant's design basis.
FoLLowing the identification of these deficiencies in the original MSLB accident analysis, the NRC issued IE BuLLetin 80-04 on February 8, 1980. This buLLetin required aLL Licensees of PWRs and near-term PWR operating License applicants to do the foLLowing:
- 1.
Review the containment pressure response anaLysis to determine if the potential for containment overpressure in the event of a MSLB inside containment included the impact of runout flow from the auxiLiary feedwater system and the impact of other energy sources such as continuation of feedwater or condensate flow.
In your review, consider the ability to detect and isolate the damage steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout fLow.
- 2.
Review your analysis of the reactivity increase which resuLts from a MSLB inside or outside containment. This review should consider the reactor cooLdown rate and the potentiaL for the reactor to return to power with the most reactive controL rod in the fuLLy withdrawn position. If your previous anaLysis did not consider aLL potential water sources (such as those Listed in 1 above) and if the reactivity increase is greater than previous analysis indicated, the report of this review should include:
- a. The boundary conditions for the anaLysis, e.g., the end of Life shutdown margin, the moderator temperature coefficient, power Level and the net effect of the associated steam generator water inventory on the reactor system cooLing, etc;
- b. The most restrictive single active faiLure in the safety injection system and the effect of that faiLure on deLaying the delivery of high concentration boric acid soLution to the reactor cooLant system;
- c. The effect of extended water supply to the affected steam generator on the core criticality and return to power; and
- d. The hot channeL factors corresponding to the most reactive rod in the fuLLy withdrawn positions at the end of Life, and the Minimum Departure from Nucleate BoiLing Ratio (MDNBR) values for the analyzed transient.
- 3.
If the potential for containment overpressurization exists or the reactor return-to-power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action. If the unit is operating, provide a description of any interim action that wiLL be taken untiL the proposed corrective action is completed."
FoLLowing the Licensee's initial response to IE BuLLetin 80-04, a request for additional information was developed to obtain aLL the information necessary to evaluate the Licensee's analysis.
The results of our evaluation for Oconee NucLear PLant, Units 1, 2 and 3 (Oconee 1, 2 and 3) are provided beLow.
2.0 EvaLuation Our consultant, the Franklin Research Center (FRC), has reviewed the submittaLs made by the Licensee in response to IE BuLLetin 80-04, and prepared the attached Technical Evaluation Report. We have reviewed this evaluation and concur in its bases and findings.
3.0 Conclusion Based on our review of the enclosed Technical Evaluation Report, the foLLowing conclusions are made regarding the postulated MSLB with continued feedwater addition for Oconee 1, 2 and 3:
- 1. There is no potential for containment overpressurization resuLting from a MSLB with continued feedwater addition because the main feedwater system is isolated and auxiliary feedwater flow to the affected steam generator is restricted.
- 2. The emergency feedwater (EFW) pumps are adequately protected against a runout flow condition and therefore can be expected to carry out their intended function without incurring damage in the event of a MSLB.
- 3. ALL potentia water sources were identified and, although a reactor return-to-power is predicted, the specified acceptable fueL design Limits are not exceeded; therefore, the FSAR reactivity increase anaLysis remains vaLid.
- 4. No further action regarding IE BuLLetin 80-04 is required.
4.0 References
- 1.
"AnaLysis of a PWR Main Steam Line Break with Continued Feedwater Addition," NRC Office of Inspection and Enforcement, February 8, 1980, IE BuLLetin 80-04
- 2.
"Overpressurization of the Containment of a PWR PLant After a Main Steam Line Break,"
NRC Office of Inspection and Enforcement, October 1, 1979, IE Information Notice 79-24
- 3.
W. 0. Parker, Jr.
(DPC)
Letter to J. P. O'ReiLLy (NRC, Region II)
Subject:
IE BuLLetin 80-04 May 7, 1980
- 4.
U. 0.
Parker, Jr.
(DPC)
Letter to H. R. Denton (NRR)
Subject:
Response to Request for Additional Information, PWR, Main Steam Line Break with Continued Feedwater Addition Oconee NucLear Station Units 1, 2, and 3 July 23, 1982
- 5.
Oconee NucLear Station Final Safety Analysis Report Duke Power Company, 1982
- 6.
Technical Evaluation Report "PWR Main Steam Line Break with Continued Feedwater Addition -
Review of Acceptance Criteria" FrankLin Research Center, November 17, 1981 TER-C5506-119
- 7.
"Criteria for Protection Systems for NucLear Power Generating Stations" Institute of ELectrical and Electronic Engineers, New York, NY, 1971 IEEE Std 279-1971
- 8.
Standard Review PLan, Section 4.2 "FueL
System Design
NRC, JuLy 1981 NUREG-0800
- 9.
Standard Review PLan, Section 15.1.5 "Steam System Piping Failures Inside and Outside of Containment (PUR)"
NRC, July 1981 NUREG-0800