ML18106A989

Provides Response to RAI Re GL 97-01, Degradation of Crdm/ Cedm Nozzle & Other Vessel Closure Head Penetrations, Dtd 980903

ML18106A989
Person / Time
Site:	Salem
Issue date:	12/15/1998
From:	SIMPSON E C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-97-01, GL-97-1, LR-N980550, TAC-M98591, TAC-M98592, NUDOCS 9812230226
Download: ML18106A989 (10)

Text

• Public Seivice Electric and Gas Company

• E. C. Simpson Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1700 Senior Vice President

-Nuclear Engineering United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

LR-N980550 DEC 1'51998 REQUEST FOR ADDITIONAL INFORMATION RELATED TO GENERIC LETTER 97-01 RESPONSE FOR SALEM GENERATING STATION UNITS NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 (TAC NOS M98591 AND M98592) REF: NRC Request for Additional Information (RAI) Related to Generic Letter 97-01 Response for Salem Generating Station Units Nos. 1 and 2 Dated September 3, 1998. The Attachment to this letter provides the information requested by the NRC request for additional information (RAI) related to Generic Letter 97-01 "Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations," (Reference).

The Attachment restates the NRC individual information requests and follows each request with the associated PSE&G response.

PSE&G's responses are based on generic responses provided by NEI as developed by the Alloy 600 Issue Task Group of the PWR Materials Reliability Project with input from the PWR Owners Groups and EPRI. If there are any additional questions, please do not hesitate to contact Richard Labatt (Principal Engineer) at (609) 339-1094.

Attachments 9a12i5 __ _ PDR ADOCK 05000272 p PDR a:\ Printed on Recycled Paper '}1

• Documeht Control Desk LR-N980550 C Mr. H. J. Miller, Administrator

-Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. P Milano, Licensing Project Manager -Salem U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. Scott Morris (X24) USNRC Senior Resident Inspector

-Salem Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625 DEC 15 1998 95-4933 Document Control Desk LR-N980550 PJD/ BC Senior Vice President

-Nuclear Operations (X04) General Manager -Salem Operations (SOS) Director -QA/Nuclear Training/Emergency Planning (114) Director -Licensing/Regulation

& Fuels Director -Design Engineering Manager -Business Planning & Co-Owners Affairs (N18) Manager -Salem Operations (S01) Manager -Mechanical Design (N29) Manager -System Engineering

-Salem (802) Project Manager -NRB (N38) J. J. Keenan, Esq. (N21) Records Management (N21) Microfilm Copy File: 3.5 Generic Letter 97-01 DEC 151999 NRC REQUEST

• ATTACHMENT 1 TO LR-N98055, Response to RAI Related to Generic Letter 97 -01 DEC 151998 1. In WCAP-14901 WEC did not provide any conclusions as to what the probabilistic failure model would lead the WOG to conclude with respect to the assessment of PWSCC in WEC-designed vessel head penetrations.

With respect to the probabilistic susceptibility model (e.g., probabilistic failure model) provided in WCAP-14901:

a. Provide the susceptibility rankings compiled for the WOG member plants for which WCAP-14901 is applicable.

In regard to other WOG member plants to which WCAP-14901 is applicable, include the basis for establishing the ranking of your plant(s) relative to the others. b. Describe how the probabilistic failure model in WCAP-14901 for assessing postulated flaws in vessel head penetration nozzles was bench-marked, and provided a list and discussion of the standards the model was marked against. c. Provide additional information regarding how the probabilistic failure models in WCAP-14901 will be refined to allow the input of plant-specific inspection data into the model's analysis methodology.

d. Describe how the variability in product forms, material specifi.cations, and heat treatments used to fabricate each CROM penetration nozzle at the WOG member utilities are addressed in the probabilistic crack initiation and growth models described or referenced in Topical Report No. WCAP-14901. PSE&G RESPONSE 1.a For industry planning purposes, plants have been grouped into three categories based on the predicted time to reach the allowable flaw depth limit. These results are provided in the industry histogram and the table provided as Attachments 2 and 3. PWR RPV head penetrations were analyzed using models developed by the Electric Power Research Institute (EPRI) and by the Westinghouse Electric Company. Both of these are probabilistic models which use the Monte Carlo method to handle uncertainties.

The histogram in Attachment 2 is based on the cumulative probability of at least one penetration in the head of each plant having a crack at the allowable depth, typically 75% through-wall.

The analysis results are reported as the time (in EFPYs of operation from January 1, 1997) for each subject plant to reach a reference probability level. This reference probability was established using the 1

• ATTACHMENT 1 TO LR-N98055' Response to RAI Related to Generic Letter 97-01 DEC 15 1998 results of the DC Cook Unit 2 inspection in 1994 and calculations made for DC Cook Unit 2 using the same methodology as for the subject plant. It is the probability that a 75% through-wall crack existed at DC Cook 2 at the time of its inspection.

This probability is somewhat lower than that for the actual observed crack depth of 6.8 millimeters (43% through-wall).

By grouping the plants based on their relative probability of experiencing a flaw at the allowable depth, the results of both the EPRI and Westinghouse models were normalized and presented in the same histogram.

The results plotted in Attachment 2 show: Seven plants calculated to have the same probability of having a crack at the allowable depth as DC Cook 2 in less than 5 EFPYs after January 1, 1997. Sixteen plants calculated to have the same probability of having a crack at the allowable depth as DC Cook 2 in 5 to 15 EFPYs after January 1, 1997. Salem Unit 1 is in this category Forty-six plants calculated to have the same probability of having a crack at the allowable depth as DC Cook 2 in more than 15 EFPYs after January 1, 1997. Salem Unit 2 is in this category.

1.b. The Westinghouse models and software used for the probabilistic analysis of reactor vessel head penetration nozzles were developed using the structural reliability and risk assessment (SRRA) methodology.

The application of this SRRA methodology to piping risk-informed ISi was extensively benchmarked against hand calculations, available failure data and alternative calculations as described in WCAP-14572, Revision 1, Supplement I (October 1997). NRC is currently planning to issue a SER accepting this application of SRRA by the end of 1998. As described in Table 4-2 of WCAP-14901 (July 1997), the SRRA probabilities for Alloy 600 PWSCC compare very well with inspection observations at four plants, where sufficient information existed to perform calculations for the worst head penetration nozzle at the time they were first inspected.

While two of the plants (D. C. Cook 2 and Ringhals 2) with relatively high calculated probabilities had observed flaw indications, two other plants with lower calculated probabilities (Almaraz 1 and North Anna 1) did not. The initial WOG probabilistic model was revised as a result of the North Anna 1 inspection observations and an independent peer review by Alloy 600 PWSCC specialists (Jim Begley and Brian Woodman) at APTECH Engineering in the spring of 1997. 1.c. There are two kinds of variations that are considered in the Westinghouse probabilistic analysis:

random and systematic.

The random variation is that due 2

• ATTACHMENT 1 TO LR-N98055, Response to RAI Related to Generic Letter 97-01 DEC 15 1998 to localized material variability and other effects with insufficient information available to completely characterize them. This could include the effect of the variation in surface roughness on crack initiation and the variation in the actual weld size on the local stress. For these types of uncertainties, a Baysean updating process has been developed by Westinghouse that could be used to combine the prior distribution on time to failure, which gives the initial calculated probability of failure with time, with the observations from the inspection.

The updated posterior distribution that is generated in this manner can then be used to generate an updated estimate of the probability of failure with time for each penetration that was inspected.

The systematic or mechanistic type variations, such as the time to crack initiation being inversely proportional to the stress to the 4th power, are included directly in the Westinghouse probabilistic model. If the observations from an inspection would differ significantly from what was calculated, then the basic model would need to be revised. This in fact has already occurred based upon the observations from the North Anna Unit I inspections.

The revised model now provides calculated probabilities that are consistent with the current inspection observations (see response to 1.d.). 1.d. Since the Westinghouse probabilistic analysis models are mechanistically based, uncertainties are provided to directly account for the variability in such fabrication related input parameters as nozzle wall thickness, material grain boundary carbide coverage and monotonic yield strength.

The Westinghouse mechanistic model also accounts for the variability in indirect fabrication related effects, such as the variation in surface roughness on crack initiation and the variation in the actual weld size on the local stress, where there is insufficient information to describe the causes and effects in a statistically significant manner. Specifically, the model input also includes the observed uncertainties on the coefficients used to calculate residual stress, initiation time and crack growth rate. NRC REQUEST 2. Table 1-2 in WCAP-14901 provides a summary of the key tasks in WEC's vessel head penetration nozzle assessment program. The table indicates that the Tasks for (1) Evaluation of PWSCC Mitigation Methods, (2) Crack Growth Data and Testing, and (3) Crack Initiation Characterization Studies have not been completed and are still in progress.

In light of the fact that the probabilistic susceptibility models appear to be dependent in part on PWSCC crack initiation and growth estimates, provide your best estimate when these tasks will be completed by WEC, and describe how these activities relate to and will be used to update the probabilistic susceptibility assessment of VHP nozzles at your plants. 3

• e ATTACHMENT 1 TO LR-N980550 DEC 151998 Response to RAI Related to Generic Letter 97 -01 PSE&G RESPONSE 2. The programs on crack growth testing and crack initiation have been essentially completed, and the program on mitigation is now underway and targeted for completion in mid-2000.

These programs have thus far served to confirm the assumptions used in the original safety evaluations and models. As additional information becomes available from the referenced testing, the models will be reviewed and updated as necessary.

No major changes are anticipated.

NRC REQUEST 3. In the NEI letters of January 29, 1998 (Ref. 1 ), and April 1, 1998 (Ref. 2), NEI indicated that inspection plans have been developed.

for the VHP nozzles at the Farley Unit 2 plant in the year 2002, and the Diablo Canyon Unit 2 plant in the year 2001, respectively.

The staff has noted that although you have endorsed the probabilistic susceptibility model described in WCAP-14901, Revision 0, other WOG member licensees have endorsed a probabilistic susceptibility model developed by an alternate vendor of choice. The WOG's proposal to inspect the VHP nozzles at the Farley Unit 2 and Diablo Canyon Unit 2 plants appears to be based on a composite assessment of the VHP nozzles at all WOG member plants. Verify that such a composite ranking assessment has been applied to the evaluation of VHP nozzles at your plants. If composite rankings of the VHP nozzles at WOG member plants have been obtained from the composite results of the two models, justify why application of the probabilistic susceptibility model described in WCAP-14901, Revision 0, would yield the same comparable relative rankings of the VHP nozzles for your plants as would application of the alternate probabilistic susceptibility model used by the WOG member plants not subscribing to WCAP-14901, Revision 0. Comment on the susceptibility rankings of the VHP nozzles at your plants relative to the susceptibility rankings of the VHP nozzles at the Farley Unit 2 and Diablo Canyon Unit 2 plants.

• PSE&G RESPONSE 3. Salem's inspection plans will be the result of the plant's economic situation, along with future operational plans. The Salem units probabilistic susceptibility as compared to other units is provided in the histogram included with this transmittal (Attachments 2 and 3). Salem's position in the histogram is one of the many considerations that must be evaluated in making inspection decisions.

Salem has only obtained composite rankings from WCAP-14901.

Concerning the susceptibility rankings of the VHP nozzles at Salem relative to the susceptibility rankings of the VHP nozzles at the Farley Unit 2 and Diablo Canyon Unit 2 plants, the relative rankings are shown in Attachments 2 and 3. As is shown in the Attachments, Salem Unit 1 falls in the 5-15 EFPY assessment group (the same group as Diablo Canyon 2) while Farley Unit 2 falls in the < 5 4

.._ * *

• ATTACHMENT 1 TO LR-N98055, Response to RAI Related to Generic Letter 97 -01 EFPY assessment group. Salem Unit 2 falls in the > 15 EFPY assessment group. 5 DEC 15 1998 50* .** *45.* .. 0 35 c '* t: m a: 30 .. Ch = 25 .*** 0 i 20; i 15* "

.*

• LR-N980550 ATTACHMENT 2 *aPtantsThat.Have

.. AlreactY.

Pertormed . . . *

Announeed

• Plans to Inspect

• mJSister Plants to lead LI.nits That Have Already Inspected . ii Plants $--.. cf---< 5EFPY s 5-15EFPvs

> 15 EFPVs Effective full power years (EFPYs) from 1/1/97 until probability of having a crack at the allowable depth matches DC .Cook 2 probability of one 75% through-wall crack at time of its 1994 inspection . Industry Histogram for IWV Head PWSCC DEC 15 1998

,,

• LR-N980550 ATTACHMENT 3 TabJe 1. Identification of Plants in Industj Histogram for RPV Head Nozzle PWSCC Status Plants That Have Already Performed Inspections Plants That Have Announced Plans to Inspect Sister Plants to Lead Units That Have Already Inspected Other Plants 00 North Anna 1 Oconee 2 rar1ey conee 1 Oconee 3 Surry l rys a 1ver Diablo Canyon 2 Ginna San Onofre 3 Davis Besse North Anna 2 Robinson 2 Salem 1 Surry 2 Turkey Point 4 Waterford 3 an norre AN02 Braidwood I Braidwood 2 Byron I Byron 2 Callaway s Calvert Cliffs l *Calvert Cliffs 2 Catawba 1 Catawba 2 Comanche Peak 1 1 Comanche Peak 2 Diablo Canyon 1 Farley 1 Fort Calhoun Indian Point 2 Indian Point 3 Kewaunee McGuire 1 McGuire 2 Millstone 3 Palo Verde I Palo Verde 2 Palo Verde 3 Point Beach 2 Prairie Island 1 Prairie Island 2 Salem 2 . Seabrook Sequoyah 1 Sequoyah 2 Shearon Harris South Texas 1 South Texas 2 St. Lucie I St. Lucie 2 Summer TMI 1 Turkey Point 3 Vogtle l Vogtle 2 Watts Bar 1 Wolf Creek DEC 1998