ML20211C737

Discusses Closure of Staff Review Re Generic Implication of Part Length Control Rod Drive Mechanism Housing Leak on 980123.Enclosed NRC 980811 & 1223 Ltrs Responded to WOG Positions Re Corrective Actions

ML20211C737
Person / Time
Site:	Prairie Island
Issue date:	08/17/1999
From:	Kim T NRC (Affiliation Not Assigned)
To:	Richard Anderson NORTHERN STATES POWER CO.
References
TAC-MA0751, TAC-MA6200, NUDOCS 9908260085
Download: ML20211C737 (4)

Text

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August 17, 1999 Mr. Roger O. Anderson, Director Nuclear Energy Engineering Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT - CLOSURE OF STAFF REVIEW REGARDING GENERIC IMPLICATION OF THE PART-LENGTH CONTROL ROD DRIVE MECHANISM HOUSING LEAK (TAC NOS. MA0751 AND MA6200)

Dear Mr. Anderson:

By the enclosed letters dated August 11 and December 23,1998, the NRC has responded to Westinghouse Owners Group's (WOG's) positions regarding corrective actions to address i

generic aspects of the part-length control rod drive mechanism (CRDM) 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 i.7dustry initiative to address this issue.

Separate from the WOG program, Northern States Power Company stated in Licensee Event Report (LER) 98-002-02, dated May 24,1999, that it had chosen to remove and cap all part-length CRDMs in both units. All four part-length CRDMs for Unit 2 were removed and capped in February 1998, and all four part-length CRDMs for Unit 1 were removed and capped during the refueling outage in April-May 1999. Accordingly, our review of this issue under TAC Nos. MA0751 and MAG 200 is considered complete.

If you have questions regarding this letter, contact me by phone at (301) 415-1392 or hv electronic mail at tjk3@nrc. gov.

i Sincerely, TNN, ben 8tNojNf Manager, Section 1 f

9908260085 990817 PDR ADOCK 05000282 Proj,ect Directorate lli P

PDR Division of Licensing Project Management i

Office of Nuclear Reactor Regulation i

Docket Nos. 50-282 81d 50-306 g p ;=

Enclosures:

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1. Letter from B. Sheron to L. Liberatori dated December 23,1998 (ACN 9812310167) h

2. Letter from B. Sheron to L. Liberatori dated August 11 N

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DOCUMENT NAME: G:\\PDill-1\\ PRAIRIE \\LTRCRDM.'NPD *See previous concurrences To recew a copy of this document, indicate in the bor "C" e C@y wOcut attachment /encbsure "E". Copy with anachment/enctnsure "N" No copy OFFICE PM:PD31_

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e August 17,1999 Mr. Roger O. Anderson, Director Nuclear Energy Engineering Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT - CLOSURE OF STAFF REVIEW REGARDING GENERIC IMPLICATION OF THE PART-LENGTH CONTROL ROD DRIVE MECHANISM HOUSING LEAK (TAC NOS. MA0751 AND MA6200)

Dear Mr. Anderson:

By the enclosed letters dated August 11 and December 23,1998, the NRC has responded to Westinghouse Owners Group's (WOG's) positions regarding corrective actions to address generic aspects of the part-length control rod drive mechanism (CRDM) 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.

Separate from the WOG program, Northern States Power Company stated in Licensee Event Report (LER) 98-002-02, dated May 24,1999, that it had chosen to remove and cap all part-length CRDMs in both units. All four part-length CRDMs for Unit 2 were removed and capped in February 1998, and all four part-length CRDMs for Unit 1 were removed and capped during the refueling outage in April-May 1999. Accordingly, our review of this issue under TAC Nos. MA0751 and MA6200 is considered complete.

If you have questions regarding this letter, contact me by phone at (301) 415-1392 or by electronic mail at tjk3 @nrc. gov.

Sincerely, TheYm,* den 8t"lYojNi Manager, Section 1 Project Directorate ill Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306

Enclosures:

1. Letter from B. Sheron to L. Liberatori dated December 23,1998 (ACN 9812310167)

2. Letter from B. Sheron to L. Liberatori dated August 11,1998 (ACN 9808120260) cc w/encls: See next page DISTRIBUTION w/o encis*:

Docket File PD31 Rdg ACRS*

RLanksbury, Rlli PUBLIC OGC*

DHood' KKarwoski*

DOCUMENT NAME: G:\\PDill-1\\ PRAIRIE \\LTRCRDM.WPD *See previous concurrences

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August 17, 1999 Mr. Roger.O. Anderson, Director

Nuclear Energy Engineering Northem States Power Company

. 414 Nicollet Mall Mmneapolis, MN 55401

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT - CLOSURE OF STAFF REVIEW REGARDING GENERIC IMPLICATION OF THE PART-LENGTH CONTROL ROD DRIVE MECHANISM HOUSING LEAK (TAC NOS. MA0751 ~

AND MA6200) o -.., ~,

Dear Mr. Anderson:

By the enclosed letters dated August 11 and December 23,1998, the NRC has responded to

' Westinghouse Owners Group's (WOG's) positions regarding corrective actions to address

generic aspects of the part-length control rod drive mechanism (CRDM) 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.

Separate from the WOG program, Northern States Power Company stated in Licensee Event Report (LER) 98-002-02, dated May 24,1999, that it had chosen to remove and cap all part-length CRDMs in both units. ' All four part length CRDMs for Unit 2 were removed and capped in February 1998, and all four part-length CRDMs for Unit 1 were removed and capped during the refueling outage in April May 1999. Accordingly, our review of this issue under

.. TAC Nos. MA0751 and MA6200 is considered complete.

If you have questions regarding this letter, contact me by phone at (301) 415-1392 or by electronic mail at tjk3Onrc. gov.

Sincerely, y

Tae Kim, Senior Project Manager, Section 1.

' Project Directorate ill Division of Licensing Project Management

- Office of Nuclear Reactor Regulation

~

Docket Nos. 50-282 and 50-306 i

Enclosures:

1. Letter from B. Sheron to L. Liberatori dated December 23,1998 (ACN 9812310167)

2. Letter from B. Sheron to L. Liberatori dated August 11,1998 (ACN 9808120260) cc w/encIs: See next page u-

~

Mr. Roger O. Anderson, Director Prairie Island Nuclear Generating Northern States Power Company Piant cc:

J. E. Silberg, Esquire Shaw, Pittman, Potts and Trowbridge Site Licensing 2300 N Street, N. W.

Prairie Island Nuclear Generating Washington DC 20037 Plant Northern States Power Company Plant Manager 1717 Wakonade Drive East Prairie Island Nuclear Generating Welch, Minnesota 55089 Plant swiuiem 6tates Power Company Tribal Council 1717 Wakonade Drive East Prairie Island Indian Community Welch, Minnesota 55089 ATTN: Environmental Department 5636 Sturgeon Lake Road Adonis A. Nebiett Welch, Minnesota 55089 Assistant Attorney General Office of the Attorney General Site General Manager 455 Minnesota Street Prairie Island Nuclear Generating Suite 900 Plant St. Paul. Minnesota 55101-2127 Northern States Power Company 1717 Wakonade Drive East U.S. Nuclear Regulatory Commission Welch, Minnesota 55089 Resident inspector's Office 1719 Wakonade Drive East Welch, Minnesota 55089-9642 Regional Administrator, Region ill U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Mr. Stephen Bloom, Administrator Goodhue County Courthouse Box 408 Red Wing, Minnesota 55066-0408 Commissioner Department of Public Service 121 Seventh Place East Suite 200 St. Paul, Minnesota 55101-2145 ApW99

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nf, wasumorow, p.c. mess.ean December 23, 1998 Mr. Lou Liberatori, Chairman WOG Steering Committee Indian Point Unit 2-1 Broadway & Blealdey Ave.

Buchanan, NY10511

SUBJECT:

PART-L NGTH CONTROL ROD DRIVE MECHANISM HOUSING ISSUE Dear Mr.

This letter provides the staffs response to your letter of October 15,1998, transmitting WCAP 15126,' Technical Assessment of the Part Length CRDM Housing Motor Tube Cracking in Westinghouse Owners Group Plants." Your October 15,1998, letter was in response to the August 11,1998, NRC letter on this subject and contains the Westinghouse Owners Group (WOG) position regarding corrective 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 January 23,1998. The staff considers the WOG program as a voluntary industry initiative in lieu of a regulatory action to address this issue. The staff notes that affected licensees heve provided commitments to follow the recommendations of the -

WOG in addressing this issue.

In our August 11,1998, letter we requested that the WOG adoress whether it agreed with the staff's statistical analysis regarding the potential number of defective welds that could be left in service if WOG agreed with the staff analysis, then we requested that the WOG address whyit believes leaving up to six defective welds in service is acceptable. Finally, we asked what modifications WOG would propose to the inspection program to address the staff concems.

WCAP 15126 contains conclusions similar to staff conclusions regrarding the potential number of defective welds that could be left in service.' However, to address the latter two questions, the WCAP refers to USNRC Regulatory Guide 1.174,"An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis" and contains an assessment of the probability of core melt as the basis for your conclusion that no further actions beyond the approximately 36% sample of welds inspected or replaced are necessary.

From the review of the information provided regarding fabrication history and metallurgical root cause analysis, it cannot be precluded that additional cracked housings remain in service.

Further, if cracks similar to those found at Prairie Island were in service, safety margins would be significantly less than specified by 10 CFR 50.55a through its implementation of Section XI of the ASME B&PV code for the CRDM housings. The sampling based inspection program for Type 309 welds performed by the WOG provides a 95% confidence that less than about 3% of

. the uninspected welds are likely to be defective. We agree this would limit the potentially -

significant number of severely degraded components in service to that assumed in the WOG risk -

assessment..

Inclosure 1 wu_

r' Lou Liberatori 1 i

We compared the WOG's resolution of this case, including its use of probabilistic risk assessment, with the guidance provided in Regulatory 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 for the range of defects that might be 4

p :::.9t is of the o'rer of 10 per reactor year. Given this marginalincrease in risk and the I

small number of welds with potentially reduced safety margins, we conclude that the actions i

taken are acceptable for protecting public health and safety.

]

Sincerely,

[ original signed by:)

I Brian W. Sheron, Acting Associate Director for Technical Review l

Office of Nuclear Reactor Regulation cc: N. Liparulo A. Drake J. Bastin H. Sepp e

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August 11. 1998 Mr. Lou Liberatori, Chairman WOG Steering Committee Indian Point Unit 2 I

Broadway & Bleakley Ave.

I Buchanan, NY 10511

SUBJECT:

PART-LENGTH CONTROL ROD DRIVE MECHANISM HOUSING ISSUE

Dear Mr. Liberatori:

This letter contains the NRC staffs evaluation of the Westinghouse Owners Group (WOG) proposed resolution 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 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:

Remove the housings and cap the reactor head penetrations

_ Perform non destructive examinations to confirm the absence of any cracking e

Perform additional records search to better identify applicability and obtain other data to confirm the absence of any cracking 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 vendofs 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 is 95 percent confidence that about 95 percent of the remaining welds do not have flaws, and (3) continued operation of plants will not result in a significant increase in risk. WOG plans to close this part length CRDM housing issue, if no further unacceptable flaw is identified in the currently planned weld inspections.

9-2 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 ancther 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 of the welds would not have flaws ofinterest to justify the sampling size of the weld inspection. This approach does not provide h gh assurance that the Type 309 weld buttered 403 components manufactured at Royal Industries satisfy the applicable regulation, including the required specified margins for structuralintegrity. In its evaluation, the staff determined that use of the acceptance criterion suggested by WOG would be inadequate to catch (with 95% confidence) as many as six defective welds in the population of 182 uninspected welds even if no 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 welds 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 (Alloy 82) is acceptable. The staff's 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.

I would appreciate if you would address the concems described above. Specifically, the WOG should address (1) whether they agree with the staff statistical analysis regarding the potential number of defective welds that could be left in service, (2)if you 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 concems.

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August 11, 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.

Si

rely, Brian W. Sheron, Acting Associate Director for Technical Review Office of Nuclear Reactor Regulation Project No. 694

Enclosure:

As stated cc w/ encl: See next page e

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Staff Evaluation on WOG's Proposed Insr=ction Proaram for Part Lenoth CRDM Housina lest=

1.0 BACKGROUND

On January 23,1998, a non-isolable reactor coolant pressure boundary 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' long, 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 Tvm '409 austenitic stainless steel (309) weld buttering at the Tyos 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:

Remove the housings and cap the reactor head penetrations Perform non-destructive examinations to confirm the absence of any cracking Perform additional records search to better identify applicability and obtain other data to confirm the absence of any cracking 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.

To date, affected WOG member utilities have inspected or repaired or committed to inspect or repair 102 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 (i.e. Type 309) in 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 statistical evaluation of the inspections planned and per.'ormed (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 will not result in a significant increase in risk.

WOG indicated it plans to close this part length CRDM housing issue, if no further unacceptable flaw is identified in the currently planned weld inspections.

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2.0 EVALUATION Materials Enoineerina The staff agrees with the industry finding that the significant cracking at Prairie Island Unit 2 was fabrication related. From the contaminants found on the failed component's fracture faces,it appears that the failed component was probably inadequately cleaned prior to weld buttedng.

Some of the elements found on the fra'cture faces are usually contained in commercial cutting lubricants. The staff agrees that the hot tearing most likely occurred 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

=':2!:, care must be taken to minimize solidification cracking. The presence of contaminants, from perhaps residual cutting lubricant, would increase the chances for solidification cracking.

The CRDM housing is a safety related Code component that was mabufactured 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 l

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 wellin excess of the depth that could be accepted by analysis pursuant to Section XI, lWB 3600. One portion of the 360' circumferential crack was through wall 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 cxisted 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 (2900/2250) 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),if the 25%

average ligament includes a " mixed zone" of smallligaments across the fracture face. WOG stated that additional capacity existed because the actual strength for the 309 weld is about 10%

higher than used in the analysis. Arguments regarding the margins available for component i

integrity for OBE and SSE loadings are based on a calculation using material allowables higher than the design allowable. Further, WOG argues that since the flawed component passed a hydro test at 3450 psi and crack growth in service has not been identified, a margin of 1.5 is i

' nferred. The staff does not agree that a margin based on pressure alone is indicative of i

component integrity structural margins. The staff's view is that the actual margin to failure is j

smaller than claimed, but the margin is essentially indeterminate. In part, this comes from the 4

staff's review of the metallography and from a review of operating history. Prior to the previous refueling no leakage had been repo 1ed 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 i

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3 knowledge no unusualloadings 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 -

water cooled 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 will be established to assure that conditions advarse 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. wore not maintained and there was through wall leakage in the RCPB. This resulted in a reduction in defense in depth since the reactor coolant pressure barrier was j

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 does 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 if no defective welds are found in 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 stsff has the following additional comments on WOG's statistical evaluation presented at the 5W98 meeting (Statistical Evaluation viewgraphs):

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The inspection results to date are not '1 flaw in 36 welds inspected' but rather l

zero flaws in 35 welds inspected. The one flaw found was not the result of a l

random inspection and should therefnre not be counted.

Etat.A:

1. From the '82' Sub-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 is the posterior mean. However, the Perdue Abramson method does not use mean values. It uses the posterior distribution which is given on page 6. From this distribution, the probability of p s.0278 is.235+.637

=.872. Thus, p s.0278 with only 87 percent assurance (not with 95 percent as claimed by WOG). The only statement that can be made with at least 95%

assurance (actually,100%) is that p s.0729.

2. From the *309' Sub lot column, it appears that p s.0278 with at least 95%

assurance. Because this population is described by the prior distribution, from l

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).

l Risk Assessment WOG risk analysis is the product of three numbers:

the probability that a reactor will have at least one flaw, i

l the frequency of operational events that might cause the flaw to fall catastrophically

. enough to create a LOCA, and the probability of failing to mitigate the LOCA.

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The staff disagrees with the Westinghouse analysis on the first two values.

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The Westinghouse analysis uses a probability of *-0.05 for a flaw to exist,' presumablyin a l

single plant. However, that is apparently taken from its statement that there is 95% confidence l

I that the whole population of welds 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 property, the mean or 'best estimate

• value of the flaw rate should be used, not the i

95th percentile value. Assuming that the inspection of the sample of 101 welds is completed I

l without discovery of another severe flaw, there is 50% confidence that the rate of flaws in the i

remaining population is less than 0.7/101=0.0069; so there is 50% confidence that there are no l

more than 1.26 flaws in the remaining uninspected population of 182 welds. Together with the l

known flaw, that makes a total flaw occurrence rate of 2.26/284= 0.008 for the whole population.

l At this rate, the probability that one of the welds in a plant will be flawed is between 0.062 for 8 welds and 0.12 for 16 welds.

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The Westinghouse anafysis uses '-1 E 1 E-05/ year abnormal event frequency" as the frequency of occurrence of events that might cause catastrophic failure of a flawed weld in these CRDM housings. That is apparently based on the facts that the flawed housing survived a 3450 psl hydrostatic test after fabrication and was analyzed by Westinghouse to be capable l

withstanding an operating basis earthquake. However, for reasons addressed elsewhere in this review, the staff is not confident in that part of the Westinghouse analysis. It is known that the flawed weld survived 23 years of service at Prairie Island Unit 2, so one estimate is that the 4.3 x 10'y/ reactor year. Since noni of the operating PWRs have expe exceeding 3450 psi or earthquaket exceeding the magnitude of operating basis earthquakes in approximately 1500 combined yeart of operation, one could estimate the occurrence rate of events that could fall this flawed weld as less than 1/1500 years = 6.7 x 10 / reactor-year. This d

i value is just inside the upper range suggested by Westinghouse. However, it is not clear 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 its 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 leak that I

occurred later, although no actual stress inducing occurrence was noted. However, other plants I

have experienced such events as moderate earthquakes, cable snags and impact loading while l

moving the upper heads and other loads. It 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 - 1E 02 -1E 04.* The staff agrees that the probability is in this range for failing to prevent core damage, given a LOCA of this size. The i

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staff's analysis uses a value of 1 x 10 for the conditional core damage probability due to small to mediam LOCAs. This is consistent with the results of a variety of NRC and industry PRAs.

I Combining the staff's factors provides a range from 5 x 10 to 4 x 10*/ year over which the staff's 4

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 to1E 10/yr.'

3.0 CONCLUSION

i The staff has concluded the following:

The leaking component at Prairie Island did not meet the current regulation. In order to assure that the remainder of the population of the 309 weld buttered 403 components manufactured at Royalindustries 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 in a similar CRDM housing weld flaw at another facility.

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I The inspection program proposed by WOG for the Type 309 welds is inadequate from a

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j statistical point of view. As stated in the staff's 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 182 uninspected welds, even if no additional i

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 staff's evaluation a combination of inspection or repair of 100% of the 309/403 partiallength CRDMs is appropriate.

• 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 in or below the mid 10 / reactor-year 4

range). Considering this level of risk, the staff has concluded that it 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 tne necessary inspection or repair during the next refueling outage. This should allow planning and qualification of inspection and repair methods that will minimize personnel exposure and best integrate with other refueling activities.

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[*

I, Westinghouse Owners Group Project No. 694 oc:

Mr. Nicholas Liparulo, Manager Equipment Design and Regulatory Eng!neering Westinghouse Electric Corporation

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Mail Stop ECE 4-15 1

P.O. Box 355 Pittsburgh, PA 15230-0355 i

Mr. Andrew Drake, Project Manager Westinghouse Owners Group

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"'::.;P.;ase Electric Corporation Mail Stop ECE 516 P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Jack Bastin, Director Regulatory Affairs Westinghouse Electric Corporation 11921 Rockville Pike Suite 107 Rockville, MD 20852 i

Mr. Hank Sepp, Manager

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Regulatory and Licensing Engineering l

Westinghouse Electric Corporation PO Box 355 Pittsburgh, PA 15230-0355 O

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