ML19345H057
| ML19345H057 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 04/15/1981 |
| From: | Paulson W Office of Nuclear Reactor Regulation |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-05.A, TASK-03-05.B, TASK-3-5.A, TASK-3-5.B, TASK-RR NUDOCS 8104300203 | |
| Download: ML19345H057 (28) | |
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UNITED STATES 8Y
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 h
April 15,1981 d
Docket No. 50-219
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MEMORANDUM FOR: Dennis M. Crutchfield, Chief Operating Reactors Branch #5, DL FROM:
Walter A. Paulson, Project Manager
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Operating Reactors Branch #5, D,L
SUBJECT:
SUMMARY
OF MEETING WITH JERSEY CENTRAL POWER & LIGHT COMPANY (JCP&LCo) ON MARCH 25, 1981 TO DISCUSS SYSTEMATIC EVALUATION PROGRAM (SEP) TOPICS III.5.A AND III.S.B HIGH ENERGY LINE BREAKS INSIDE AND OUT-SIDE CONTAINMENT, FOR THE OYSTER CREEK NUCLEAR GENERATING STATION On March 25, 1981, the staff met with representatives of Jersey Central Power & Light Company in Bethesda, Maryland. The purpose of the meeting was to discuss the status of the review of SEP Topics III.5.A and III.S.B
- for the Oyster Creek Nuclear Generating Station. A list of attendees is enclosed (Enclosure 1).
A summary of evaluations submitted by the licensee, responses to requests for additional information, and a chronology of meetings with the NRC was presented by MPR Associates, consultants to JCP&L (Enclosure 2). The licensee has concluded the following with regard to postulated high energy line breaks inside containment (Topic III.S.A):
1.
High energy pipe breaks inside containment have been evaluated in accordance with pertinent NRC criteria.
2.
The evaluations indicate that no single pipe break would cause demage to systems, components, or structures required to nitigate the consequences of the break or to achieve safe shutdown.
3.
Topic III-5.A is considered resolved for Oyster Creek.
We stated that we would prepare a Safety Evaluation for this topic.
The discussion on high energy line breaks outside containment (SEP TOPIC 11I.5.B) centered on our concerns regarding the s team supply line to the emergency condenser. This line has two isolation valves outside containment (none inside containment). The concern is that the effects of a line break outside containment could result in the loss of capability to close the isolation valves. Our letter dated
, July 10,1980 transmitted our evaluation of this topic and requested that JCP&L provide a schedule for modifications to this system. JCP&L's letter dated October 6,1980, proposed that a leak detection system be installed by the end of the next scheduled refueling outage (scheduled to begin about November 30, 1981). The proposed system would detect leakage from a through-wall crack. This system is predicated on the,
" leak before break" principle.
The NRC staff stated that our review of the licensee's October 6,1980 submittal indicated:
1.
The submittal did not demonstrate that barriers, separation, etc. are not practical.
2.
The submittal did not demonstrate that if this li'ne does fail, "le6k before break" is the expected mode of fa,ilure.
l The staff stated that we would provide a response to JCP&L's October 6, 1980 letter.
OdxA. as~
Walter A. Paulson, Project Manager Operating Reactors Branch #5 Divisio,n of Licensing -
Enclosures:
As stated cc w/ enclosures:
See next page l
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Mr. David P. Hoffman April 15',1981 cc Mr. Paul A. Perry, Secretary U. S. Environmental Protection Consumers Power Conpany Agency 212 West Michigan Avenue Federal Activities Branch Jackson, Michigan 49201 Region V Office ATTN: EIS C0ORDINATOR Judd L. Bacon, Esquire 230 South Dearborn Street Consumers Power Conpany Chicago, Illinois 60604 212 West Michigan Avenue l
' Jackson, Michigan 49201 Herbert Grossman, Esq., Chairman Atomic Safety and Licensing Board t
Joseph Gallo, Esquire U. S. Nuclear Regulatory Comission Isham, Lincoln & Beale Washington, D. C.
20555 1120 Connecticut Avenue Room 325 Dr. Oscar H. Paris
. Washington, D. C.
20036 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission Peter W. Steketee, Esquire Washington, D. C.
20555 505 Peoples Building Grand Rapids,. Michigan 49503 Mr. Frederick J. Shon l
Atomic Safety and Licensing Board Alan S. Rosenthal, Esq., Chairman U. S. Nuclear Regulatory Comission Atomic Safety & Licensing Appeal Board Washington, D. C.
20555 U. S. Nuclear Regulatory Comission Washington, D. C.
20555 Big Rock Point Nuclear Power Plant ATTN: Mr. C. J. Hartman Mr. John O'Neill, II Plant Superintendent Route 2, Box 44 Charlevoix, Michigan 49720 Maple City, Michigan 49664 Christa-Maria Charlevoix Public Library Route 2, Box 108C 107 Clinton Street Charlevoix, Michigan 49720 l
Charlevoix, Michigan William J. Scanlon, Esquire
' Chairman 2034 Pauline Boulevard County Board of Supervisors Ann Arbor, Michigan 48103 Charlevoix County 1
Charlevoix, Michigan 49720 Resident Inspector -
Big Rock Point Plant Office of the Governor- (2) c/o U.S. NRC Room 1 - Capitol Building RR #3, Box 600 Lansing, Michigan 48913 Charlevoix, Michigan 49720 Director, Criteria and Standards Mr. Jim E. Mills Division Route 2, Box 108C Office of Radiation Programs Charlevoix, Michigan 49720 (ANR-460)
U. S. Environmental Protection Thomas S. Moore l
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Agency Atomic Safety & Licensing Appeal Board Washington, D. C.
20460 U. S. Nuclear Regulatory Comission Washington, D. C.
20555
Mr. David P. Hoffman April 15,1981 Cc Dr. John H. Buck Atomic Safety and Licensing Appeal Board' U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Ms. JoAnn Bier 204 C1.inton Street
'Charlevoix, Michigan 49720 e
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ATTENDANCE LIST NAME AFFILIATION W. Paulson NRC W. Russell NRC H. Brammer NRC/MEB Y. C. Li NRC/MEB E. MeKenna NRC JJohnson MPR Assoc.
J. Knubel JCP&L/GPU P. Y. Chen NRC R. Hermann NRC R. Klecker NRC/MTEB H. Walker NRC/MTEB W. Schmidt MPR Assoc.
D. Strawson MPR Assoc.
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CHRONOLOGY OF KEY EVENTS HIGH ENERGY LINE BREAKS OUTSIDE CONTAINMENT 1.
DECEMBER 1972 - GIAMBUSSO LETTER ASKED LICENSEES TO SUBMIT INFORMATION TO AEC ON THE EFFECTS OF PIPE BREAKS OUTSIDE CONTAINMENT.
2.
JULY 1974 - NIENDMENT'NO. 75 TO OYSTER CREEK FSAR JCP&L RESPONSE TO GIAMBUSSO LETTER.
WENT THROUGH 5 REVISIONS BEFORE ACCEPTABLE TO AEC BY DECEMBER 1976.
3.
FEBRUARY 1980 - JCP&L LETTER TO NRC PRESENTED RESULTS OF ADDITIONAL EVALUATIONS WHICH INDIdATED THAT BREAKS AT CERTAIN LOCATIONS IN THE EMERGENCY CONDENSER PIPING ON THE 75' ELEVATION COULD RESULT IN DAMAGE TO BOTH l
ISOLATION VALVES IN THE BROKEN LOOP.
STATED THAT THREE GENERAL METHCDS OF PREVENTING UNACCEPTABLE INTERACTION WITH THESE VALVES WERE UNDER REVIEW.
A.
PIPE WHIP RESTRAINTS AND JET IMPINGEMENT BARRIERS.
B.
ADDITION OF NEW ISOLATION VALVES INSIDE CONTAINMENT.
C.
CLOSE ONE ISOLATION VALVE OUTSIDE C;ONTAINMENT DURING NORMAL OPERATION.
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JULY 1980 - NRC LETTER TO JCP&L PRESENTED NRC EVALUATION OF PIPE BREAKS OUTSIDE CONTAINMENT UNDER SEP TOPIC III-5.B.
A.
ENCLOSURE 1 - NRC DETAIL EVALUATION.
B.
ENCLOSURE 2 - REQUEST FOR ADDITIONAL INFORMATION (4 ITEMS).
C.
ENCLOSURE'3 - STAFF POSITIONS.
REQUESTED SCHEDULE POR PLANT MODIFICATIONS FOR 2 AREAS WHERE NRC EVALUATION CONCLUDED INADEQUATE PROTECTION EXISTED.
5.
OCTOBER'1980 - JCP&L LETTER TO NRC JCP&L RESPONSE TO NRC LETTER, DATE6 JULY 19'80.
A.
ATTACHMENT 1 - RESPONSE TO ENCLOSURE 3 OF NRC LETTER.
1)
EMERGENCY CONDENSER - PLAN TO INSTALL LEAK DETECTION SYSTEM.
EITHER ACOUSTIC (WESTINGHOUSE)
OR MOISTURE STRIP (TECHMARK).
2)
MAIN STEAM /FEEDWATER - NRC EVALUATION NOT CORRECT.
NO MODIFICATIONS ARE REQUIRED.
B.*
NTTACHMENT 2 - RESPONSE TO ENCLOSURE 2 OF NRC LETTER.
1)
COMPARISON OF OYSTER CREEK PIPING WITH SECTION B.l.b OF BTP MEB 3-1.
B.
ATTACHMENT 2 (CONT'D) 2)
COMPARISON OF OYSTER CREEK PIPING WITH SECTION B.2.c OF BTP ASB 3-1.
3)
EVALUATION OF EFFECTS OF PIPE BREAKS IN REACTOR WATER CLEANUP SYSTEM ON CABLE TRAY 13A.
4)
EVALUATION OF POTENTIAL FOR FLOODING OF CABLE SPREADING ROOM.
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OYSTER CREEK NUCLEAR GENERATING STATION UNIT 1 SEP TOPIC III-5.A HIGH ENERGY PIPE BREAKS INSIDE CONTAINME!!T
SUMMARY
1.
Pertinent NRC Criteria a.
Regulatory Guide 1.46, " Protection Against Pipe Whip Inside Containment" b.
Standard Review Plan 3.6.2, " Determination of Break Locations and Dynamic Effects Associated with Postulated Rupture of Piping" 2.
SEP Meeting Summary 3.
Evaluations of High Energy Pipe Breaks Inside Containment a.
Prior to SEP b.
As part of SEP i
4 Conclusions P00R ORIGINAL
SEP MEETING
SUMMARY
1.
September 7, 1978 JCP&L summarized all prior work pertaining to Topic III-5.A.
A document containing all pertinent docketed information was submitted to NRC.
2.
February 6, 1979 JCP&L submitted a summary of all stress values used to define postulated break locations in previous evaluations, plus revisions to the seismic stress for a few systems.
NRC identified further information required to resolve Topic III-5.A.
3.
August 16, 1979 The JCP&L report submitted on July 30, 1979, in response to the NRC request for further information was discussed.
No further actions were requested of JCP&L.
4.
May 1980 NRC indicated new criteria are being considered.
These may include defining postulated break i
locations based on an " effects oriented" approach.
These may also recognize the," leak-before-break" concept, and permit leak detection systems where it is impractical to (1) re-route piping, (2) move components, or (3) provide jet deflectors.
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P00R ORIGINAL.
1
.'.a EVALUATIONS PRIOR TO SEP (See Primarily Attachment 7C in Document dated Septembsr 7, 1978).
1.
Method of defining postulated break locations.
In full compliance with Regulatory Guide 1.46.
2.
Criteria for defining " unacceptable" postulated breaks, a.
Due to pipe whip:
Contact with another high energy system or safety system having smaller nominal diameter or righter wall thickness.
Contact with the steel containment vessel.
Contact with electric conduit, cable trays or notor operators required for a safety function.
No consideration given to whether systems are redundant or required for safe shutdown.
b.
Due to steam jets:
Impingement on any electric cable tray or motor operator required for a safety function, again with no consideration of redundancy or effect on safe shutdown capability.
3.
Overall results:
17 different high energy piping systems located inside containment.
150 different postulated break locations.
5 4
P00R BRIGINAL t
EVALUATIONS PRIOR TO SEP (cont'd) 4.
Break categories which vere identified:
Category 1; 89 postulated breaks with no
" unacceptable" consequences.
Category 2; 48 postulated breaks with
" unacceptable" consequences, but at low stress locations (less than 50 percent of the stre'ss criterion in Regulatory Guide 1.46).
Category 3, 13 postulated breaks with
" unacceptable" consequences at high stress locations (greater than 50 percent of the stress criterion in Regulatory Guide 1.46).
5.
Corrective action:
JCP&L recommended increased surveillance for pipe welds in Category 3.
NRC letter of February 12, 1975 (see Attachment 8 to report of September 7, 1978) requested increased inspection frequency.
JCP&L modified the original surveillance program to be consistent with NRC's request (see Attachment 10 to report of September 7, 1978).
The surveillance program has been implemented.
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P00R BRIGINAI.
EVALUATIONS FOR SEP
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STRESS SUMMARIES (See Report dated February 6, 1979) is All stress values used previously to define postulated break locations were submitted to the NRC.
is Concervative seismic stresses were originally employed Where seismic stresses were not located or available.
These were modified by:
Performing dynamic response spectra analyses, cLr_
6 6
Locating the original seismic evaluations, h5 The effect was to reduce the number of breaks firem 150 tc 141.
Revised break categories are:
6 Category 1:
83 postulated breaks 6
Category 2:
53 postulated breaks 6
Category 3:
5 postulated breaks l
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EVALUATIONS FOR SEP RESPONSE TO NRC REQUEST FOR INFORMATION (See Document Submitted to NRC on July 30, 1979) 1.
NRC requested further information for Category 2 and 3 postulated breaks:
Scoping calculations to determine if the containment wall can be penetrated.
Evaluations to determine if the breaks would prevent safe shutdown, considering:
(1)
Primary effects of pipe. whip.
(2)
Jet impingement on electrical equipment.
(3)
Mechanical effects of jets on piping and structures.
(4)
Secondary (cascading) effects of pipe whip.
Inherent capability of piping to withstand circu:sf erential breaks.
2.
Containment wall penetration:
No postulated break results in penetration becausei A plastic hinge does not form, oji The whipping pipe does not attain sufficient energy, or The break is in piping of 14-inch diameter or greater, so wall deformation is terminated by the concrete shield wall before' failure.
l 3.
Primary effects of pipe whip:
No postulated breaks l
prevent a safe shutdown because:
l Physical separation prevents contact with more than l
.one' redundant ~ train of a safety system, and Single active component failure of the intact train of safe shutdown systems does not prevent safe shutdown.
l Further, No postulated break results in contact with safe shutdown equipment.
P00R BRIGINAl.
RESPONSE TO NRC REQUEST FOR INFORMATION (cont'd) 4.
Jet impingement on electrical equipment:
A single cable tray contains power supply cable to the ADS electromatic relief valves.
However, no postulated break will prevent safe shutdown because:
The break is in'large piping (8-inch or greater),
which does not require ADS operation, or For the specific postulated breaks in small piping (less than 8-inch), at least two relief valves and one emergency condenser remain operable, and can prevent excessive clad temperature.
5.
Mechanical effects of jets:
Jets impinging on adjacent
' piping were assumed without specific analyses to affect piping integrity.
Evaluations then show that no postulated break will prevent safe shutdown because:
Physical separation prevents contact with more than one redundant train of a safety system, and The unaffected train can withstand a single active failure.
6.
Secondary effects of pipe whip:
In each case, the damaged " target" pipe occurs in a system which the above analyses show will not prevent safe shutdown.
Accordingly, there should be no adverse secondary effects.
7.
Inherent capability of piping:
Partial through-wall cracks were considered, including rapidly developing cracks (dynamic load f actor of two) and slowly l
developing cracks.
The range of. tolerable crack sizes, in percent of total circumference, are:
82% to 95% for rapidly developing cracks.
91% to 97.5% for slowly developing cracks.
4 l
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POOR ORIGINAL l
CONCLUSIONS High energy pipe breaks inside containment have bean
. evaluated in accordance with pertinent NRC criteria.
The evaluations indicate that no single pipe break would cause dar. age to systems, components, or structures required co mitigate the consequences of the break or to achieve safe shutdown.
Topic III-5.A is considered resolved for. Oyster Creek.
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MEETING
SUMMARY
DISTRIBUTION Docket NRC PDR Local PDR ORB Reading J. Olshinski S. Varga T. Ippolito R. Clark JStolz D. Crutchfield OELD OI&E (3)
HSmith WPaul son ACRS (10)
NSIC TERA W. Russell R. Hennann E. Mckenna H. brammer Y. C. Li P. Y. Chen R. Klecker H. Walker l
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P00R ORlGlNAL
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