ML18051A472
| ML18051A472 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/22/1983 |
| From: | Wambach T Office of Nuclear Reactor Regulation |
| To: | Vandewalle D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| LSO5-83-06-049, LSO5-83-6-49, NUDOCS 8306280439 | |
| Download: ML18051A472 (8) | |
Text
Docket No. 50-255 LS05-83-06-049 Mr. David J. Vande\\tJalle Nuclear Licensing Administrator Consumers Power Company 1945 West Parna 11 Road Jackson, Michigan 49201
Dear Mr. VandeWalle:
June 22,
- 1983
SUBJECT:
PALISADES NUCLEAR PO~JER PLANT - IPSAR SECTIONS 4. 9 AND 4.15, PIPE BREAK INSIDE CONTAINMENT ANO REACTOR COOLANT PRESSURE BOUNDARY ( RCPB} LEAKAGE DETECTION REQUIREMENTS In the Integrated Plant Safety Assessment Report (IPSAR} for the Palisades Nuclear Power Plant, NUREG-0820, the subject items were identified as requiring "refined engineering analysis or continuation of ongoing evalua-tion. II Based upon review of your December 9, 1982 submittal, the staff concludes that IPSAR Section 4.9 is complete. It is the staff's position that you should propose corrective measures for the two break locations that affect multiple instrument lines and, in accordance with IPSAR Section 4.15.2, that you should propose Technical Specification changes concerning operability of leakage detection systems.
You are requested to P.ro.vide your proposed resolution for these issues within 90 days of receipt of this letter.
Enclosure:
Evaluation on IPSAR Sections 4.9 and 4.15 cc w/enclosure:
See next page Sincerely, Original signed by/
Thomas V. Wambach, Project Manager Operating Reactors Branch #5 Division of Licensing
~
~'<)' OJ
~4-'
vs
\\)
c}tt~
~ 1"~
. ~.
~-o~*
OFFICE *** ~~P.~.:.P.hf.~(~... ~.~P..~.:R.......... ~.~-~-~.:.Q.~...... ~.~.~-~.:.Q.~.fi.M..... ~.~... ~.:.Q.~......... 9.~~~?.:.P.~........ Q~....... :... 9.....
suRNAME ~.. f:Mc K~nna.; dk... C.G.r..i me.$.......... M.8.o.Y.l.e.............TM.1.c.b.ae.1.s......... ~Rv.s.s.e..J.1....... IW9.UJD.9.~h *..
1
- Cr.lJ;ts;.b.f.i.e.l.d DATE *.* 6/.'5./.83.......... E./...-:f)/.23........... 6;..,'1..;.23........... 6./..!.Vfil............ 6./.~./E.3........ 6/.~.~83.......... 6;.Y.?1.23.......
NRG FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960
Mr. David J. VandeWalle cc M.
I._~iller, Esquire Ilham, Lincoln & Beale Suite.4200 One First National Plaza Chicago, Illinois 60670 Mr. Paul A. Perry, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 James G. Keppler, Regional Administrator Nuclear Regulatory Corranission, Region III 799 ~oosevelt Road Glen Ellyn, Illinois 60137 Township Supervisor Covert Townshi Route l, Box 10 Van Buren County, Michigan 49043 Office of the Governor (2)
Room 1 - Capitol Building Lansing, Michigan 48913 Palisades Plant ATTN:
Mr. Robert Montross Plant Manager Covert, Michigan 49043 U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN:
Regional Radiation Representative 230 South Dearborn Street Chicago, Illinois 60604 Resident Inspector c/o U. S. NRC Palisades Plant Route 2, P. O. Box 155 Covert, Michigan 49043 Mr. Ronald Callen, Supervisor Advance Planning and Review Section Michigan Public Service Commission 6545 Mercantile Way Post Office Box 30221 Lansing, Michigan 48909
I.
INTRODUCTION PALISADES NUCLEAR POWER PLANT IPSAR SECTIONS 4.9 AND 4. 15 HIGH ENERGY LINE BREAKS INSIDE CONTAINMENT AND RCPB LEAKAGE DETECTION REQUIREMENTS The final Integrated Plant Safety Analysis Report lIPSAR) for the Palisades Nuclear Plant (NUREG-0820) (Reference 1), issued in October 1982, documented in Section 4.9 the licensee's commitment to complete review of high energy pipe breaks inside containment using staff guidance transmitted with the staff safety evaluation of Topic III-5.A. This guidance included criteria for acceptable interactions as well as a method for resolution of break locations where remedial measures are impractical which consists of a fracture mechanics analysis that demonstrates that a given size flaw remains stable under postulated loads and that smaller flaws are detectable.
Similarly, the evaluation of the leakage detection systems (IPSAR Section 4.15) concluded that the need for improved leak rate sensitivitv should be determined concurrent with the resolution of pipe break effects. The staff safety evaluation on Topic III-5.A, Effects of High Energy Line Breaks Inside Containment was issued by letter dated December 4, 1981 (Reference 2).
In this evaluati.on, the staff concluded that the criteria and methodology were generally acceptable; however, additional information on the criteria used to determine acceptability of target piping interactions (functionality) was requested.. Also, over 200 break locations were unresolved.
As discussed above, the licensee was to further evaluate these break locations.
II.
EVALUATION By letters dated August 16, 1982 {Reference 3) and December 9, 1982 (Reference 4), CPCo submitted additional analyses of high energy line breaks inside containment to resolve the above issues.
A.
Target Piping Acceptability Criteria In the topic safety evaluation (Reference 2), the staff requested addi-tional justification for the acceptance criterion for target piping, since the "functional capability" of the impinged pipe had not been directly addressed by the licensee. The staff's specific concern is
- that some piping systems are required to deliver certain ra~ed. flow.....
- and should be designed to retain dimensional stability when stressed to the allowable limits associated with the emergency and faulted condi-tions; i.e., the functional capability of the piping is required to be demonstrated.
In response to the staff's concern, the licensee submitted Reference 4.
The licensee stated that a three dimensional computer model of the pipe as a long cylindrical thin shell was used to conduct a non-linear elastic-plastic analysis.
The analysis showed that the maximum flow area reduction experienced by the target pipe in the Palisades pipe break study was only
~. 9%.
Therefore, the functional capability of the target pipe is not significantly affected.* Based on a review of the information in Reference 4, we have determined that the licensee's target pipe evaluation is acceptable.
B.
Disposition of Unresolved Breaks The break locations that were unresolved after the preliminary review (in 1981) with conservative screening criteria were reevaluated in more detail.
The number of unresolved break locations were reduced to 139 after locations were moved to welds and structural discontinuities only instead of at any point in the system.
The mechanistic approach was used on some lines to eliminate 42 locations. Detailed structural evaluation of interactions with the reactor cavity wall, containment liner,. penetrations, and concrete embedment eliminated 18 locations (24 inter~ctions).
The assessment of effects on safe shutdown capability and consideration of the jet zone of influence Lfrom break locatio_rrs _th~t w~re !l}f?V~q)_
reduced the unresolved locatipns to 14.
Twelve* of the remaining locations were then considered with the staff. gu1diin-ce--f6-r reso*1 ution of break locati-ons where re~dial measures are impractical (transmitted with the staff topic evaluatton in Reference 2).
The other two locations are still unresolved (see Section C.3).
C.
- Results
- 1.
Break Locations The licensee used both the mechanistic and the simplified mechanistic approaches (see definitions in Reference 5) *to reduce the number of break locations to consider.
Based on review of the licensee*~
submittal (Reference 4), the staff concludes that the licensee has adequately identified the most likely break locations inside contain-ment and that the methodology for selection is acceptable.
- 2.
Structural Evaluations The licensee's structural analyses were not reviewed by the staff in any deta i 1 ; however, th.ey wi 11 be considered in conjunct ion with review of code, load and load combination changes (see IPSAR Section 4.12).
- 3.
Effects of Safe Shutdown Capability The staff examined the target reevaluations presented in Appendix C of Reference 4, concerning the plant capability to attain safe shutdown following a postulated pipe break.
The approach used by the licensee was to examine each unresolved interaction between a postulated break and a safety-related target to determine whether a safe shutdown could still be performed.
The single failure criterion was used*.
For example, for a non-LOCA break, some degradation of.safety. injection capability would be acceptable.
For some secondary system break locations, the contain-ment wall may be impacted.
However, the possible.radiological consequences of such a scenario would be a small fraction of 10 CFR Part 100 guidelines.
Some small line breaks. could fail other small lines;.. these cases were considered* acceptable if the break area was enveloped by existing safety analysis, the ensuring. break.
is _fo_-'t_he same loop of~tl:i~_RCS ari9_damag*e* to in-core.instr-~mentatian_
lines is prevented.
Two breaks (one in* the ch~ging line,* one in the letdown line) were identified which could damage instrument lines for steam generator pressure and level for both steam generators as well as the pressurizer pressure.and level instrument lines.
The* staff finds __
this lass of monitoring instruments *~nacceptabl e and coricl udes _tha.t __ _
corrective measures are required.
Pipe whip from one break location in the three-inch pressurizer spray line could strike a cable tray containing power cables to one of the two hydrogen recombiners.
A postulated single failure could result in no recombiners being available.
The licensee performed a preliminary leak before break fracture mechanics analysis (see below) which showed a stable 90° crack with a leak rate of approximately O. l gpm.
A leak rate of this magnitude is probably too small to be readily detectable by present in-plant leak detection equipment.
However, the leak would be detected in the daily calculation of the primary coolant inventory.
Based upon the preliminary deterministic fracture mechanics analysis, the licensee believes that the pas*~1 b1Jf_tY. of the.3-inch pres-sur1_zer __
spray line break resulting in a pipe whip is unlikely and, therefore, that the need to modify the plant to add whip restraints or a barrier to protect the target cable tray is unwarranted.
Furthermore, a modification to add local leak detection to monitor one weld is also not warranted.
The weld associated with the break location is presently examined to ASME Section XI Class l requirements and is scheduled for inspection during the 1985 refueling outage.
- Following a small break LOCA of the 3-inch pressurizer spray line, the hydrogen recombiner is not needed until a 2% hydrogen volume concentration is reached in.containment.
For comparative purposes, calculations performed for the Palisades containment show a 2%
volume percent. of hydrogen volume concentration is not reached until 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> have elapsed following a design basis accident (DBA).
Plant operating procedure OMP-22 requires a hydrogen recombiner to be energized. by the time hydrogen concentration has reached 2%.
If at that time the recombiner is found to be inoperative, due to cable severance, there would be adequate time available to restore power to the redundant hydrogen recombiner before a combustible mtxture of 4% is reached at about 350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br />.
It is reasonable to assume that given this time frame of 350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br /> associated with a OBA, the diesel generator malfunction wil 1 be.
corrected or offsite power will be restored.
The staff finds this resolution acceptable.
Based on our review of the licensee's general guidelines and of the detailed results, the staff concludes that the licensee has adequately addressed the postulated break interactions except for the two cases involving i:D,strument lines discussed above~
- 4.
Resolution of 12 Locations (Where Remedial Measures are Impratt ital ),
As discussed in Reference 2, the basic approach is to show that conditions that could lead to a double-ended rupture do not exist.
The fracture mechanics.evaluation is done to show that a given size flaw remains stable under postulated loads and that smaller flaws are detectable by leakage detection systems.
The jet impingement effects from cracks corresponding to the detectable leakage must also be considered.
Ca)
Fracture Mechanics Appendix G of the licensee's submittal (Reference 4) was reviewed to determine if the fracture mechanics analysis of the twelve postulated break locations met the staff's crite.ria given in Reference 2.
The staff has determined that significant margins against pipe break, given seismic and operating loads, exist at these locations. The analyses were performed with conservative values for postulated crack size and material properties. This analysis demonstrated to the staff's satisfaction that a crack of a size which would result in a 10 gpm leak would be stable in the presence of seismic and operating loads for a considerable peri ad of time.
e
.. - (b)
Effects of Jets The 1icensee assessed the jet impingement 1oad from f1aw sizes corresponding to a 10 gpm 1eak. Targets in the vicinity of these break 1ocations were reviewed to ensure that no unacceptab1e damage wou1d occur from these jets.
The staff finds this acceptable.
- 5.
Leakage Detection As discussed in Sect.ion 4.15 of the Palisades IPSAR, the staff found that the leakage detection systems are able to detect a 1 gpm leak from the reactor coolant pressure boundary to the containment within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In the IPSAR, the staff concluded that the need for improved sensitivity should be.determined from the fracture mechanics evaluation. A staff requirement for-Technical Specification modifications to impose operability requirements for 1eakage detection systems was deferred until the pipe break eva1uation was comp1eted.
~
Given the* significant resi~tance of the* subject piping to pipe break and given the sensi:;.tivity of the existing leakage detection system15 the staff finds the fracture mechanics evaluation accept-*
able and that no leak detection system modifications due to the subject postu1ated break locations need* be made.
Therefore, in accordance with IPSAR Section 4.15.2, the 1icensee should submit a request for amendment to modify the Technical Specifications concerning operability of leak detection systems that monitor leakage to the containment.
III.
CONCLUSION The staff has reviewed the 1icensee's ana1yses of high energy line breaks inside containment.and concludes that the subject topic and thus IPSAR Section 4.9, is complete.
Staff review of structural considerations will be coordinated with review of IPSAR Section 4.12 on Code, Load and Load Combination changes.
It is the staff's position that the licensee should propose corrective measures for the two indicated break locations that affect multiple instrument lines.
No modifications to p1ant leakage detection systems are required as a result of this review; therefore, the 1icensee should propose Technical Specification.changes concerning operability of insta11ed leakage detection systems in accordance with IPSAR Section 4. 15.2~
IV.
REFERENCES
.. 1.
NUREG-0820, Integrated Plant Safety Assessment Report for the Palisades Nuclear Power Plant, dated October 1982.
- 2.
Letter from D. M. Crutchfield (NRC) to D.*
- P. Hoffman {CPCo), dated December 4, 1981.
- 3.
Letter from R. A. Vincent (CPCo) to D. M. Crutchfield (NRC), dated August 16,. 1982.
- 4.
Letter from K. A. Toner (CPCo) to D. M. Crutchfield (NRC)~
transmitti.ng EDS Report No. 02-0540-1108, November 1982,. dated December 9, 1982.
- 5.
Letter from D.. K. Davis (NRC) to J. Mc.Ewen ( KMC), SEP Owners Group, dated July 20, 1978.