ML16342C093
| ML16342C093 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 02/09/1999 |
| From: | Steven Bloom NRC (Affiliation Not Assigned) |
| To: | Rueger G PACIFIC GAS & ELECTRIC CO. |
| References | |
| TAC-MA1404, TAC-MA1405, NUDOCS 9902160242 | |
| Download: ML16342C093 (30) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 February 9,
1999 Mr. Gregory M. Rueger Senior Vice President and General Manager Pacific Gas and Electric Company Diablo Canyon Nuclear Power Plant P.O. Box 3 Avila Beach, California 93424
SUBJECT:
CLOSURE OF TAC REGARDING GENERIC IMPLICATIONOF THE PART-LENGTH CONTROL ROD DRIVE MECHANISMHOUSING LEAK AT DIABLOCANYON, UNITS 1 AND 2 (TAC NOS. MA1404 AND MA1405)
Dear Mr. Rueger:
By the enclosed letters dated August 11 and December 23, 1998, the NRC has responded to the Westinghouse Owners Group (WOG) positions regarding corrective actions to address generic aspects of the part-length control rod drive mechanism housing issue that originated as a result of the leak at Prairie Island Unit 2 on January 23, 1998. The WOG program is a voluntary industry initiative to address this issue.
By letters dated March 13, 1998, May 15, 1998, and January 15, 1999, Pacific Gas and Electric Company discussed its participation in the WOG initiative and the activities at Diablo Canyon Units 1 and 2 resulting from that initiative. The NRC staff's review of these letters and results was performed under TAC Nos MA1404 and MA1405.
As discussed in the December 23, 1998, letter, the NRC staff has concluded that, given the marginal increase in risk and the small number of welds with potentially reduced safety margins, the actions taken under the industry initiative are acceptable for protecting public health and safety. Accordingly, our review under TAC Nos. MA1404 and MA1405 are considered complete.
9902i60242 990209 PDR ADCICK 05000275 P
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Mr. Gregory M. Rueger February 9,
1999 Ifyou have questions regarding this lette'r, please contact me by phone at (301) 415-1313 or by electronic mail at sdb1@nrc.gov.
Sincerely, Original Signed By Docket Nos. 50-275 and 50-323
Enclosures:
- 1. NRC Letter dated 8/11/98
- 2. NRC Letter dated 12/23/98 cc w/encls:
See next page Steven D. Bloom, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation DISTRIBUTION:
Docket PUBLIC PDIV-2 Reading EAdensam WBateman S Bloom EPeyton OGC ACRS KBrockman, Region IV LSmith, RegionlV DHood Document Name:
DC1401.w d OFC DATE PDIV-2 SBloo 2/
I /99 PDIV-2 EPeeoV 2/ 8 /99 OFFICIAL RECORD COPY
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Mr. Gregory M. Rueger February 9,
1999 cc w/encls:
NRC Resident Inspector Diablo Canyon Nuclear Power Plant c/o U.S. Nuclear Regulatory Commission P. O. Box 369 Avila Beach, California 93424 Dr. Richard Ferguson, Energy Chair
'ierra Club California 1100 11th Street, Suite 311 Sacramento, California 95814 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower 8 Pavillion 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 Christopher J. Warner, Esq; Pacific Gas 8 Electric Company Post Office Box 7442 San Francisco, California 94120 Ms. Nancy Culver San Luis Obispo Mothers for Peace P. O. Box 164 Pismo Beach, California 93448 Chairman San Luis Obispo County Board of Supefvlsof's Room 370 County Government Center San Luis Obispo, California 93408 Mr. Truman Burris Mr. Robert Kinosian California Public Utilities Commission 505 Van Ness, Room 4102 San Francisco, California 94102 Mr. David H. Oatley, Vice President Diablo Canyon Operations and Plant Manager Diablo Canyon Nuclear Power Plant P.O. Box 3 Avila Beach, California 93424 Telegram-Tribune ATTN: Managing Editor 1321 Johnson Avenue P.O. Box 112 San Luis Obispo, California 93406 Mr. Steve Hsu Radiologic Health Branch State Department of Health Services Post Office Box 942732 Sacramento, California 94232 Diablo Canyon Independent Safety Committee ATtN: Robert R. Wellington, Esq.
Legal Counsel 857 Cass Street; Suite D Monterey, California 93940
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UNITED STATES NVCLEAR REGULATORY COMMlSSION WASH IN QTON, O.c. OR~
August ll. 1998 Mr. Lou LIberatori, Chairman WOG Steering Committee Indian Point Unit 2 Broadway 8 Bleakley Ave.
Buchanan, NY 10511
SUBJECT:
PART-LENGTH CONTROL ROD DRIVE MECHANISMHOUSING ISSUE
Dear Mr. Liberatori:
This letter contains the NRC staff's evaluation of the Westinghouse Owner's Group, (WOG) proposed resolution of the part-length control rod drive mechanism (CRDM) housing issue that onginated as a result of the leak that occurred at Prairie island Unit 2 on January 23, 1998.
Following the staff's review of the initial information on this event, the NRC formally requested WOG to activate the Regulatory Response Group on February 20, 1998. The staff met with the RRG on February 27, 1998 to discuss this issue, On March 6, 1998, RRG issued a letter that requested the affected owners to docket their plans for addressing the issue within 30 days'and initiate compensatory measures for RCPB leakage.
The options identified by RRG for the plan were:
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Remove the housings and cap the reactor head penetrations
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Perform nondestructive examinations to confirm the absence of any cracldng
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. Perform additional records search to better identify applicability and obtain other data to confirm the absence of any cracking
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Address the capability of using additional RCS leakage monitoring awareness while the Issue is being resolved The NRC found these recommendations an appropriate and acceptable response to the identification of the QA breakdown at the vendor's shop and as suitable corrective actions for the potential very large defects that jeopardize RCS integrity.
The staff met with WOG representatives In a number of public meetings, the most recent of which was held on June 11, 1998. During this meeting WOG provided its conclusions based on the weld inspections, fabrication records review, safety assessment, and statistical evaluation of the inspections planned and performed (assuming that no additional flaws are Identified in the planned inspections).
Its conclusions are that (1 ) the Prairie island flaw was an Isolated event, (2) there ls 95 percent confidence that about 95 percent of the remaining welds do not have Saws, and (3) continued operation of plants willnot result in a significant increase in risk, WOG plans to dose this part-length CRDM housing issue, Ifno further unacceptable flaw Is Identified In the currently planned weld Inspections.
ENCLOSURE 1
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~ 0 August 11, 1998 After completing its evaluation, the staff disagreed with WOG's conclusions. The staff conveyed this determination to WOG by telecon on June 25, 1998. Specifically, the staff determined that the inspections performed to date and inspections currently planned are not adequate to assure that a similar CRDM housing weld flaw found at Prairie island would not be present at another facility. The staff disagreed with aspects of the mechanistic, statistical, and risk evaluations presented by WOG. In light of the break down in the quality assurance program at the vendor's shop and the need to maintain the pressure boundary integrity, the staff disagreed with WOG's approach of using the 95/95 criterion of 95 percent confidence that 95 percent ofthe welds would not have fiaws ofinterest to justify the sampling size of the weld inspection. This approach does not provide high assurance that the Type 309 weld buttered 403 components manufactured at Royal industries satisf'y the applicable regulation, IncIuding the required specified margins for structural integrity. In its evaluation, the staff determined that use of the acceptance criterion suggested by WOG would be inadequate to catch (with 954k confidence) as many as six defective welds in the population of 182 uninspected welds even ifno additional flaws are found in the proposed WOG sample.
The detailed staff evaluation is enclosed.
Based on the staff evaluation results described above, the inspection program for Type 309 weids proposed by the WOG appears to leave a potentially significant number of severely degraded components in service. An inspection program that results in high assurance that no degraded components are left in service is the appropriate goal. To accomplish this, it thus would appear necessary to either inspect essentially all the components with a qualified ultrasonic examination, or remove the components.
The staff finds the statistically based inspection program proposed by WOG for part-length CRDM that used Inconel weld filler(Alloy82) is acceptable.
The staffs basis for this is that no failures have been identified with these components and they are considered to be less susceptible to the mechanism that generated the flaw in the Type 309 weld; therefore, an inspection program based on the 95/95 criterion is acceptable for sampling the population.
f would appreciate ifyou would address the concerns described above. Specifically, the WOG should address (1) whether they agree with the staff statistical analysis regarding the potential number of cfefectivc welds that could be left in service, (2) ifyou agree with the staff analysis, why you believe leaving up to six defective welds in service is acceptable and (3) what modifications you would propose to your inspection program to address the staff concerns.
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3 August ll. 1998 Please provide your response within 14 days of receipt of this letter so that the staff can resolve this issue in the near term and take any regulatory action deemed necessary.
Project No, 694
Enclosure:
As stated cc w/encl: See next page Brian W. Sheron, Acting Associate Director forTechnical Review Office of Nuclear Reactor Regulation
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Staff Evaluation on WOG's Pro osed Ins ection Pr ram or Part-Len CRDM Housln Issue
1.0 BACKGROUND
'n January 23, 1998, a non-isolable reactor coolant pressure boundaty leak of 0.26 g.p.m. was discovered in a part-length CRDM housing at the G-9 core location of the Prairie-Island Unit 2 reactor while it was operating.
Metallurgical evaluation of the failed housing confirmed ultrasonic (UT) examination results that a very deep 360'ong, partial through-wall crack was present.
The metallurgical evaluation results showed the flaw had been undersized by UT examination results. The failure mechanism was identified as hot tearing associated with the fabrication of the Type 309 austenltic stainless steel (309) weld buttering at the Type 403 martensitic stainless steel (403) forging. Chemical analysis results identified the following contaminants on the fracture face of the failed component: sulfur, copper, boron, and zinc.
Following staff's review of the Information provided by the licensee, the NRC formally requested WOG to activate the RRG on February 20, 1998. The staff met with RRG on February 27, 1998, to discuss this issue. On March 6, 1998, RRG issued a letter that requested the affected owners to docket their plans for addressing the issue within 30 days and initiate compensatory measures for RCPB leakage.
The options identified by RRG for the plan were:
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Remove the housings and cap the reactor head penetrations
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Perform non-destructive examinations to confirm the absence of any cracking
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Perform additional records search to better identify applicability and obtain other data to confirm the absence of any cracking
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Address the capability of using additional RCS leakage monitoring awareness while the Issue is being resolved The NRC found these recommendations an appropriate and acceptable response to the Identification of the QA breakdown at the vendor's shop and as suitable corrective actions for the potential vela large defects that jeopardize RCS integrity.
To date, affected WOG member utilities have inspected or repaired or committed to inspect or repair 10" CRDM 308/309/403 weldments on 51 assemblies at nine operating plants. There is a total population of 284 welds of the type of interest (l,e.Type 309) ln 137 installed assemblies and five spare assemblies at 21 operating plants.
On June 11, 1998, representatives of WOG summarized this Issue at a public meeting with the staff, Based on the weld inspections, fabrication records review, safety assessment, and statisticai evaluation of the inspections planned and performed (assuming no flaws are Identified In the planned inspections), WOG concluded that (1) the Prairie Island flaw-was an Isolated event, (2) there is 95 percent confidence that about 95 percent of the remaining welds do not have flaws, and*(3) continued operation of plants willnot result in a significant increase In risk.
WOG Indicated It plans to close this part-length CRDM housing issue, lfno further unacceptable fiaw is identified in the currently planned weld inspections.
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2.0 EVALUATION Material En ineerin The staff agrees with the industry finding that the significant cracking at Prairie Island Unit 2 was fabrication-related.
From the contamlnants found on the failed component's fracture faces, It appears that the failed component was probably inadequately cleaned prior to weld buttering.
Some of the elements found on the fracture faces are usually contained in commercial cutting lubricants. The staff agrees that the hot tearing most likelyoccurred during solidification of the weld butter and that subsequent post-weld heat treatment (PWHT) may have extended the cracking. Further, the heavy oxide scale found on fracture faces indicates that the open crack was subjected to the high temperatures of the PWHT.that are well above plant operating temperatures.
Because of the large thermal expansion mismatch between the 309 and 403 materials, care must be taken to minimize solidification'cracking. The presence of contaminants, from perhaps residual cutting lubricant, would increase the chances for solidification nacking.
The CRDM housing is a safety-related Code component that was manufactured under a quality assurance program that was intended to satisfy 10 CFR 50, Appendix B. The cleaning prior to weld buttering was specified In the controlling procedure.
The surface and volumetric examinations performed failed to assure quality in that a component with severe cracking was not rejected. Therefore, it is clear that the quality assurance program broke down for the failed component, in particular, with respect to Appendix B, Criterion IX-Control of Special Processes, in that cleaning prior to welding appears to be not as specified and Criterion X-inspection, in that the examinations performed for the work operation did not identify the unacceptable defect.
The severity of the cracking found in the Prairie island Unit 2 part-length CRDM was among the worst identified in a safety-related component at an operating nuclear power plant. The cracking found was well in excess of the depth that could be accepted by analysis pursuant to Section XI, IWB 3600, One portion of the 360'ircumferential crack was through-wali and other portions of the crack were in excess of the approximately 75% Code maximum flaw depth limitation.
Nonetheless, limit load fracture analysis was performed by WOG to determine the margins that existed in the flawed component. WOG stated that the average remaining ligament in the cracked component from metallography was about 25%, The failure pressure was calculated to be 2900 psi for the 309 weld, Based on WOG analysis, there was a margin of about 1.3 (2900/2260) to failure for normal and upset conditions; the ASME code-required margin of safety Is 2.77. However, the staff was unable to confirm that the average uncracked ligament was 25%. It Is not clear, from review of the metallography presented (WCAP-15054), Ifthe 25%
average ligament includes a "mixed zone" of small ligaments across the fracture face. WOG stated that additional capacity existed because the actual strength forthe 309 weld Is about 10%
higher than used In the analysis.
Arguments regarding the margins available for component integrity for OBE and SSE loadings are based on a calculation using material alfowables higher than the design aiiowable. Further, WOG argues that since the flawed component passed a hydro test at 3450 psl and crack growth in service has not been identified, a margin of 1.5 is Inferred, The staff does not agree that a margin based on pressure alone Is Indicative of component integrity structural margins. The staffs view is that the actual margin to failure fs smaller than ciaimed, but the margin Is essentially indeterminate.'n part, this comes from the staff's review of the metallography and from a mview of operating history. Prior to the previous refueling no leakage had been reported for the component, A leak in this reactor coolant pressure boundary (RCPB) component was discovered while the plant was operating, and the plant was taken out of service as required by the technical specifications (TS). To the staff s
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4 knowfedge no unusual loadings from transients or other events had occurred prior to the discovery of the leak. The staff understands that some work was done on the reactor head during the last refueling outage and that the head was removed from and replaced on the reactor vessel during the refueling, It is possible that a load from either bumping the head during movement or when work was being performed was of sufficient magnitude to cause the crack to open and leak during the subsequent cycle of operation.
The regulations applicable to this issue are as follows:
1.
10 CFR 50.55a(g)(4) requires that throughout the service life of a boiling or pressurized waterwooied nuclear power facility, components that are ASME Code Class 1, 2, or 3 must meet the requirements set forth in the applicable edition and addenda of Section XI for the facility.
2.
10 CFR 50, Appendix B, Criterion XVI,states that measures willbe established to assure that conditions adverse to safety, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly detected and corrected.
The leaking component at Prairie Island did not satisfy the above requirements in that Section XI margins to failure were not maintained and there was through-waIl leakage in the RCPB. This resulted in a reduction in defense in depth since the reactor coolant pressure barrier was breached.
As explained in the following section on the staff's statistical evaluation, the Inspection program proposed by WOG is inadequate.
Statistical Evaluation At the June 11, 1998, WOG/NRC meeting, WOG reported that 35 weld inspections were performed, and found no defective welds. Based on this information, the staff performed its independent statistical evaluation, and found that this inspection would not catch (with 95%
confidence) as many as 21 defective welds in the remaining population of 248 uninspected welds.
The inspection program proposed by WOG is inadequate from a statistical point of view. It cloes not provide adequate confidence that appropriate corrective actions are taken to ensure the Type 309 weld buttered 403 components manufactured at Royal Industries satisfy the'applicable regulation, specifically the required margins for structural integrity. In order to attain this goal, it would be necessary to demonstrate with 95% confidence that there are no flaws remaining in the uninspected welds. Even lfno defective welds are. found fn the 66 additional welds which are to be inspected, and accounting for the 35 welds already Inspected, what can be demonstrated with 95% confidence, is only that there are less than seven defective welds in the remaining population of 182 uninspected welds.
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The staff has the following additional comments on WOG's statisticai evaluation presented at the 5/6/98 meeting (Statistical Evaluation vfewgraphs):
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The inspection results to date are not '1 flaw fn 36 wefds inspected'ut rather zero flaws in 35 welds inspected.
The one ffaw found was not the result of a random inspection and should therefore not be counted.
- 1. From the '82'ub-lot column, it appears that p s.0271 with at least 95%
assurance.
However, there are two problems with this conclusion, First, from page 6, the value.0271 ls the posterior mean.
However, the Perdue-Abramson method does not use mean values.
It uses the posterior distribution which is
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given on page 6. From this distribution, the probabilfty of p s.0278 is 235+.637
~.872, Thus, p s.0278 with only 87 percent assurance (not with 95 pe;cent as claimed by WOG). The only statement that can be made with at feast 95%
assurance (actually, 100%) is that p s.0729.
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2, From the 309'ub-lot column, it appears that p s.0278 with at least 95%
assurance.
Because this population is described by the prior distribution, from page 6 the probability that p s.0278 is.185+.630 ~.815. Thus, p s.0278 with only 81 percent assurance (not with 95 percent as claimed by WOG).
Risk Assessment WOG risk analysis is the product of three numbers:
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the probability. that a reactor willhave at least one flaw,'
the frequency of operational events that might cause the flaw to fail catastrophically enough to create a LOCA, and
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the probability of failing to mitigate the LOCA.
The staff disagrees with the Westinghouse analysis on the first two values.
The Westinghouse analysis uses a probability of '-0.05 for a flaw to exist,'resumably In a single plant. However, that is apparently taken from its statement that there Is 95% confidence that the whole population of wefds is less than 5% flawed. For a PRA, the appropriate value is the probability that one or more of the 8 to 16 welds in the plant Is flawed. To determine that probability properfy, the mean or 'best estimate'alue of the flaw rate should be used, not the 95th percentile value. Assuming that the inspection of the sample of 101 welds Is completed without discovery of another severe flaw, there is 50% confidence that the rate of flaws In the remaining population is fess than 0.7/1 01&,0069; so there is 50% confidence that there are no more than 1.26 flaws in the remaining uninspected population of 182 wefds. Together with the known flaw, that makes a total flaw occurrence rate of 2,26/284~ 0,008 for the whole population.
At this rate, the probability that one of the welds in a plant willbe flawed Is between 0.062 for 8 welds and 0.12 for 16 welds.
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The Westinghouse analysis uses '-1E43-1E45/year abnormal event frequency's the frequency of occurrence of events that might cause catastrophic failure of a flawed weld fn these CRDM housings. That is apparently based on the facts that the flawed housing suNfved a 3450 psi hydrostatic test after fabrication and was analyzed by Westinghouse to be capable withstanding an operating basis earthquake.
However, for reasons addressed elsewhere in this review, the staff is not confident fn that part of the Westinghouse anafysfs.
It is known that the flawed weld survived 23 years of service at Prairie fsfand Unit 2, so one estimate is that the freqvency of occurrence of events that would cause weld failure is probably fess than 1/23 years ~
4.3 x 10'/reactor-year.
Since none of the operating PWRs have experienced pressure transients exceeding 3450 psi or earthquakes exceeding the magnitude of operating basis earthquakes in approximately 1500 combined years of operation, one could estimate the occurrence rate of events that could fail this flawed weld as fess than 1/1500 years c 6.7 x 10"/reactor-year.
This value Is Just fnside the upper range suggested by Westinghovse.
However, It fs not cfear that the flawed weld at Prairie Island actually would have survived all of the operational occurrences experienced at the other PWRs to date, Although corrosive degradation of the weld during fts service life was not evident, It was observed to begin leaking noticeably during the current cycle operation.
Some sort of stress imposed during the outage is suspected of producing the teak that occurred later, although no actual stress inducing occurrence was noted, However, other plants have experienced such events as moderate earthqvakes, cable snags and impact loading while moving the upper heads and other loads, lt is not clear how these occurrences at the other plants would have affected the flawed CRDM housing.
Degradation during an outage may potentially make the flawed weld more susceptible to failure during operational events.
The Westinghouse analysis used a 'LOCA CCDP - 1E42 -1 E44.'he staff agrees that the probability is in this range for failing to prevent core damage, given a LOCA of this size. The staff's analysis uses a value of 1 x 10'or the conditional core damage probability due to small to medium LOCAs, This is consistent with the results of a variety of NRC and industry PRAs.
Combining the staff's factors provides a range from 5 x 10~ to 4 x 10~/year over which the staff's confidence varies from good to poor that the core damage frequency due to failure of a flawed CRDM housing has been bounded, The staff's range generally overlaps and slightly exceeds the upper part of the range suggested by Westinghouse, which is '1E-06/yr to1 E-10/yr.'.0 CONCLUSION
'The staff has concluded the following:
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The feaking component at Prairie Island did not meet the cvrrent regulation, In order to assure that the remainder of the population of the 309 weld buttered 403 components manufactured at Royal Industries have the required specified margins for structural Integrity, and to satisfy applicable quality assurance requirements, corrective actions are necessary to provide a high confidence that the deficiencies revealed by the discovery of the weld flaw at Prairie Island Unit 2 did not result ln a similar CRDM housing weld flaw at another facility.
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The inspection program proposedby WOG for the Type 309 welds is inadequate from a statistical point of view. As stated ln the staffs statistical evaluation, WOG's current inspection plan is inadequate to catch (with 95% confidence) as many as six defective
. welds in the remaining population of $ 82 uninspected welds, even lfno additional defective welds are found in the sample. There would be only 36% confidence that no defective weld remains in the uninspected population. Based on the staf's evaluation a combination of inspection or repair of 100% of the 309/403 partial length CRDMs is appropriate.
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The level of risk associated with this issue at plants which have not yet inspected similar welds may be small (i,e., the CDF increment is ln or below the mid-10~/reactor-year range).
Considering this level of risk, the staff has concluded that lt is not appropriate to require immediate action that would subject plants to additional startup and shutdown activities. The staff considers it more prudent to implement the necessary inspection or repair during the next refueling outage.
This should allow planning and qualification of inspection and repair methods that willminimize personnel exposure'and best integrate with other refueling activities.
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Westinghouse Owners Group Project No. 694 cc:
Mr. Nicholas Uparulo, Manager Equipment Design and Regulatory Engineering Westinghouse Electric Corporation Mail Stop ECE 4-15 P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Andrew Drake, Project Manager Westinghouse Owners Group Westinghouse Electric Corporation Mail Stop ECE 5-16 P,Q, Box 355 Pittsburgh, PA 15230-0355 Mr. Jack Bastin, Director.
Regulatory Affairs Westinghouse Electric Corporation 11921 Rockville Pike Suite 107 Rockville, MD 20852 Mr, Hank Sepp, Manager Regulatory and Licensing Engineering Westinghouse Electric Corporation PO Box 355 Pittsburgh, PA 15230-0355
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASH IHOTON, Doc. RkR$400l December 23, 1998 Mr. Lou uberatorl, Chairman WOG Steering Committee Indian Point Unit2 Broadway 8 Bleakley Ave.
'Bvchanan, NY10511
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SUBJECT:
PART-LENGTH CONTROL ROD DRlVEMECHANlSMHOVSING ISSUE Dear Mr.
orL This letter provides the staffs response to your letter of October 15, 1998, transmitting WCAP-15126, "Technical Assessment of the Part Length CRDM Housing MotorTube Cracking fn Westinghouse Owners Grovp Plants.'our October 15, 1998, letter was ln response to the Augvst11, 1998, NRC letter on this subject and contains the Westinghouse Owners Group (WOG) position regarding cor'rective actions to address generic aspects of the part. length control rod drive mechanism (CRDM) housing issue that originated as a result of the leak that occurred at Prairie island, Unit 2, on January23,1998.
The staff considers the WOG program as a voluntary Industry initiative ln lieu of a regulatory action to address this issue. The staff notes that affected licensees have provided commitments to followthe recommendations of the WOG ln addressing this issue.
In our August 11, 1998, letter we requested that the WOG address whether It agreed with the staff's statisticai analysis regarding the potential number of defective welds that could be left ln service.
IfWOG agreed with the staff analysis, then we req'uested that the WOG address why It
. believes leaving up to six defective welds in service ls acceptable.
Finally, we asked what modifications WOG wovtd propose to the Inspection program to address the staff concerns.
WCAP-15126 contains conclusions similar to staff conclusions regarding the potential number of defective welds that covld be left ln service. However, to address the latter two qvestlons, the WCAP refers to USNRC Regulatory Guide 1.174, "AnApproach for Using Probabilistic Risk Assessment ln Risk-informed Decisions on Plant-Specific Changes to the Licensing Basis" and contains an assessment of the probability of core melt as the basis foryour conclusion that no further actions beyond the approximately 36% sample of welds fnspected or replaced'are necessary.
From the review of the Information provided regarding fabrication history and metallurgical toot cause analysis, lt cannot be precluded that additional cracked housings remain ln Service.
Further, Ifcracks similar to those fovnd at Prairie Island were ln service, safety margins would be slgnwcantly less than specified by10 CFR 50.55a through Its Implementation of Section XIof the ASME BB,PV code forthe CRDM housings. The sampling based inspection program for Type 309 welds performedby the WOG provides a 95% confidence that less than abovt 3% of the uninspected welds are likelyto be defective. We agree this would limitthe potentially significant nvmber of severely degraded components h service to that assumed In the WOG risk assessment.
ENCLOSURE 2'
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Lou Uberatori We compared the WOG's resolution of this case, including its use of probabllistlc <k assessment, with the guidance provided ln ReguIatoty Guide 1.174. As noted above, with 95%
confidence safety margins should be unaffected for all but as few as 3% of the uninspected welds..
We agree that the incremental core damage frequency forthe range of defects that might be present ts of the order of 10~ per reactor year. Given this marginal increase in risk and the small number of welds with potentially reduced safety margins, we conclude that the actions taken are acceptable for protecting public health and safety.
Sincerely,
'original signed by:)
Brian W. Sheron, Acting Associate Director for Technical Review Office of Nuclear Reactor Regulation cc: N. Uparulo A. Drake J. Bastin H. Sepp
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