ML20149K760
| ML20149K760 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/25/1997 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-01, GL-97-1, NUDOCS 9707300135 | |
| Download: ML20149K760 (10) | |
Text
N Dave Morey 3:uthern Nuclear
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Vice President Op: rating Comp:ny s
Farley Project P.O. Box 1295 Bkmingham. Alabama 35201 Tel 205.992.5131 SOUTHERN COMPANY July 25, 1997 Energy:o Serve YourWorld" Docket No.:
50-348 10 CFR 50.55a 50-364 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Joseph M. Farley Nuclear Plant Response to Generic Letter 97-01 Ladies and Gentlemen:
Generic Letter (GL) 97-01, Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations, was issued to request licensees to describe their program for insuring the timely inspection of PWR control rod drive mechanism (CRDM) and other closure head penetratiors. Enclo_ sed is the Southern Nuclear Operating Company (SNC) response for Farley Nuclear Pbnt (FNP) relative to the information requested by GL 97-01.
Prior to issuance of GL 97-01, SNC worked with the Westinghouse Owners Group (WOG), the Electric Power Research Institute (EPRI), and the Nuclear Energy Institute (NEI) to understand the operational experience, identify technical issues, cause factore, relative importance, and solutions to the Alloy 600 issues identified by the NRC. One of these tasks was the development of safety evaluations by each of the PWR Owners Groups (PWROGs) that characterized the initiation of damage, propagation, and consequences associated with Alloy 600 head penetration cracking. The WOG safety evaluation is contained in WCAP-13565, Alloy 600 Reactor Vessel Head Adaptor Tube Cracking Safety Evaluation, and is applicable to FNP. The NRC reviewed the safety evaluations prepared by the PWROGs and issued a safety evaluation report (SER) to the Nuclear Management and Resources Council (NUMARC) on November 19,1993. The WOG safety evaluation and the resulting NRC SER establish the basis for the continued operation of FNP.
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9707300135 970725 Y
PDR ADOCK 05000348 ll! ! !
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U. S. Nuclear Regulatory Commission Page 2
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Mr. D. N. Morey states he is vice president of SNC, is authorized to execute this oath on behalf of SNC, and to the best of his knowledge and belief, the facts set forth in this letter are true.
Respectfully submitted, SOUTIIERN NUCLEAR OPERATING COMPANY J
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Dave Morey Sworn to and subscribed before me thishay of I997 7Wac& Aad @rv l
Notary Publi8 My Commission Expires:
/Zd-I /,/k97 REM / cit:GL9701 Enclosure ec:
Mr. L. A. Reyes, Region II Administrator Mr. J. I. Zimmerman, NRR Project Manager Mr. T. M. Ross, Plant Sr. Resident Inspector Mr. T. A. Reed, NRR - Materials and Chemical Engineering Branch Dr. D. E. Williamson, State Department of Public IIealth
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ENCLOSURE JOSEPH M. FARLEY NUCLEAR PLANT RESPONSE TO GENERIC LETTER 97-01
i Response to Generic Letter 97-01 GL 97-01 Requested Information Item 1.1 A description ofallinspections ofCRDAf no::le and other 171Ps (vessel head penetrations) performed to the date ofthis generic letter, including the results ofthese inspections.
SNC Response to Requested Information item 1.1 Consistent with WCAP-13565, Alloy 600 Reactor Vessel llead Adaptor Tube Cracking Safety j
Evaluation, and the resulting NRC SER dated November 19,1993, Southern Nuclear Operating Company (SNC) performs GL 88-05 walkdowns following shutdown for refueling of either of the Farley Nuclear Plant (FNP) units to identify boric acid deposits that could be indicative of reactor coolant system (RCS) leakage. These walkdowns include a visual inspection of the reactor vessel head and to date, no significant boric acid deposits have been identified that indicates leakage from the subject reactor vessel head penetration nozzles. In addition to the GL 88-05 walkdowns, SNC
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inspections of 29 out of 69 head penetrations on Unit I and 32 out of 69 head penetrations on Unit 2
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between the reactor vessel head insulation and the top of the reactor head using a remote camera did not identify any evidence ofleakage from the penetrations. These inspections were performed in 1995 and included penetrations for CRDMs, instrument columns, head adapter plugs, and latch
- housings. Also in 1995, field replication of one of the material heats used in Unit I and two of the material heats used in Unit 2 were performed to determine the material microstructure, a factor in the i
models used to assess the susceptibility of the reactor vessel head penetrations to primary water stress corrosion cracking (PWSCC). The penetrations selected for replication were chosen based on their location in the periphery of the closure head in the area of the highest residual stress.
GL 97-01 Requested Information Item 1.2
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1 Ifaplan has been developed to periodically inspect the CRDAf no::le and other F71Ps:
A. Provide the scheduleforfirst, and subsequent, inspections ofthe CRDAf no::le and other 17IPs, including the technical basisfor this schedule.
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B. Provide the scopefor the CRDAf no::le and other VHP inspections, including the total number ofpenetrations (and how many will be inspected), which penetrations have thermal sleeves, which are spares, and which are instrument or otherpenetrations.
SNC Response to Requested Information Item 1.2 j
SNC had originally scheduled volumetric examinations for FNP 1 and 2 in the year 2001. WCAP-13565, Alloy 600 Reactor Vessel Head Adaptor Tube Cracking Safety Evaluation, and the resulting NRC SER provided the basis for the original SNC plan. Specifically, WCAP-13565 and the NRC SER concluded:
(1) axial flaws would be more likely than circumferential flaws; (2) catastrophic failure of a penetration is extremely unlikely because a flaw would be detected during GL 88-05 boric acid leakage surveillance walkdowns before it reached the critical flaw size;
. Response to Generic Letter 97-01 (3) the Westinghouse Owners Group (WOG) susceptibility model considers the appropriate parameters affecting IGSCC and should provide a reasonable ranking of plant susceptibilities; (4) based on existing leakage monitoring requirements, there is reasonable assurance that leakage in excess of the 1.0 gpm technical specification limit would be detected prior to any unstable extension of the flaw; and, (5) CRDM cracking at the reactor vessel heads is not a significant safety issue as long as the surveillance walkdowns in accordance with GL 88-05 continue.
In response to GL 97-01, SNC has elected to participate in the Nuclear Energy Institute (NEI)/WOG RPV head penetration integrated inspection program described in the SNC Response to Requested Information Item 1.4 below. As a participant in the NEI/WOG integrated inspection program, inspections of the FNP reactor vessel head penetration nozzles will be based on the NEl/WOG j
integrated inspection program.
GL 97-01 Requested Information item 1.3 Ifa plan has.n_pt been developed to periodically inspect the CRDM no::le and other VHPs, provide the analysis that supports why no augmentedinspection is necessary.
SNC Response to Requested Information Item 1.3 j
Based on SNC participation in the NEI/WOG integrated inspection program, Requested Information Item 1.3 is not applicable to FNP.
GL 97-01 Requested Information Item 1.4 In light ofthe degradation ofCRDM no::le and other VHPs described above, provide the analysis that supports the selected course ofaction as listed in either 1.2 or 1.3, above. In particidar, provide a description ofall relevant data and/or tests used to develop crack initiation and crack growth models, the methods and data used to validate these models, theplant-specific inputs to these models, and how these models substantiate the susceptibility evaluation. Also, ifan integrated industry inspection program is being relied on, provide a detailed description ofthis program.
SNC Response to Requested Information Item 1.4 The data, tests, and methods used in developing the crack initiation and crack growth models on which the SNC management strategy for addressing the FNP vessel head penetration cracking issue is based are provided in Sections 2 and 3 of WCAP-14901, Background and Methodology for Evaluation of Reactor Vessel Closure IIead Penetration Integrity for the Westinghouse Owners Group.
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.m Response to Generic Letter 97-01 SNC is a participant in the WOG analysis program. A plant specific probability an:.iysis using the methodology described in Section 4 of WCAP-14901 has been performed for FNP. The plant specific input parameters to the analysis are provided in Table I and Table 2 of this response for Unit I and Unit 2, respectively. The analysis results will be incorporated into the NEl/WOG integrated inspection program, for use in detennining the need for a plant specific inspection. This integrated inspection program includes all three PWR owners groups, the Electric Power Research Institute (EPRI), and NEl who are cooperatively working to compile information on the estimated operating time from January 1, 1997, needed to initiate and propagate a crack 75% through wall in a vessel penetration for all operating PWRs in the United States. This information will be evaluated to determine if an adequate number of plants have or are planning to inspect in the near future. This evaluation is scheduled to be completed j
and detailed inspection plans for the industry to be provided to the NRC by the end of 1997. A copy of WCAP-14901 has been provided to the NRC by the WOG.
TABLE 1 FARLEY UNIT 1 INPUT VALUES USED FOR PROBABILtSTic ANALYSIS Grain f*ng"le d
hg"" rah P
Case Penetration No.
Temperature Y.S. (ksi)
()
(%)
1 62-69 596.5 F 42.6 32.7 50.2til 2
58-61 40.0 32.7 50.211 3
50-57 38.6 32.7 50.211 4
46-49 37.3 32.7 50.201 5
38-45
.33.1 32.7 50.203 6
30-37 28.6 32.7 50.233 7
26-29 27.0 32.7 50.2DI 8
22-25 25.4 32.7 50.2UI __
9 18-21 19.8 39.0 44.7 10 14-17 19.8 32.7 50.201 11 10-13 17.6 32.7 50.201 12 6-9 12.4 32.7 50.2Ul 13[21 2-5 8.7.
32.7 50.2Di VI Values from field replication 121 This case also used to bound penetration number 1 Page 3 of 7
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Kesponse to Generic Letter 97-01 4
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l TABLE 2 FARLEY UNIT 2 INPUT VALUES USED FOR PROBABILISTIC ANALYSIS 3
Grain
""d
. Case Penetration No.
Temperature fet"p Y.S. (ksi) 9gle Co rag
()
(%)
1 62-69 596.9 F 42.6 48.5 23.0 a
2 58-61 40.0 48.5 23.0 l
3 50-57 38.6 35.0 27.7 4
46-49 37.3 48.5-23.0 5
38-45 33.1 48.5 23.0 6
30-37 28.6
'48.5 23.0 7
26-29 27.0 48.5 23.0 8
22-25 25.4 48.5 23.0 l
9 14-21 19.8 48.5-23.0 l
10 10-13 17.6 48.5 23.0 11 6-9 12.4 48.5 23.0 i
12 2-5 8.7 48.5 23.0 l
13 1
0 48.5 23.0 f
IU Values from field replication-
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Response to Generic Letter 97-01 GL 97-01 Requested Information Item 2 Provide a description ofany resin head intrusions, as described in IN 96-11, that have exceeded the current EPR1 PlVR Primary IVater Chemistry Guidelines recommendationsforprimary water sulfate levels, including thefallowing information: (See Requested Information Items 2.1 through 2.6)
SNC Response to Requested Information Item 2 SNC has reviewed the plant historical records and determined that a resin ingress similar to those which occurred in 1980 and 1981 at the Jose Cabrera (Zorita) plant has not occurred at FNP. This data search was structured to identify any resin intrusion event into the primary coolant system that exceeded a magnitude of I cubic foot (approximately 30 liters). The threshold of I cubic foot was l
chosen as a conservative lower bound since it represents less than 15% of the estimated volume of resin released into the reactor coolant system during the two events at Jose Cabrera.
1 Routine analysis for sulfate in reactor coolant was not performed prior to June 1994 for FNP. For the period of plant operation following June 1994, a sulfate concentration in the range of 15 to 17 ppm was used as an indicator of action resin ingress equivalent to a volume of I cubic foot.
For the period of plant operation prior to the routine analysis for sulfate in reactor coolant, the data search was based on a review of the plant's reactor coolant chemistry records relative to conductivity of the reactor coolant. An elevation of 28 S/cm in conductivity was the value used as an indicator of action resin ingress equivalent to a volume of I cubic foot.
One occurrence was identified where the RCS conductivity increase exceeded the 28 S/cm criterion for approximately a two week period in February 1988. A review of additional RCS chemistry parameters during this period determined that the elevated conductivity was not the result of a resin bead intrusion. Specilically, the elevated conductivity was not accompanied by a corresponding depression of pli, elevation in lithium, or elevation in suspended solids.
Requested Information Item 2 implied that any increase above the limits specified in the EPRI PWR Primary Water Chemistry Guidelines were indicative of a potential resin bead intrusion. It is true that a resin bead intrusion would result in sulfate values that exceed the EPRI Primary Water Chemistry Guidelines, but it is not true that all sulfate measurements in excess of the EPRI guidelines are indicative of a resin intrusion. Therefore, SNC used the above criteria to determine that a resin bead intrusion has not occurred at FNP.
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Response to Generic Letter 97-01 i
l GL 97-01 Requested Information Item 2.1 l
l Were the intrusions action, anion, or mixed bed?
SNC Response to Requested Information Item 2.1 Based on the determination that no resin bead intrusions have occurred at FNP, Requested Information Item 2,1 is not applicable to FNP.
GL 97-01 Requested Information Item 2.2 What were the duration 's ofthese intrusions?
l SNC Response to Requested Information Item 2.2 Based on the determination that no resin bead intrusions have occurred at FNP, Requested Information Item 2.2 is not applicable to FNP.
GL 97-01 Requested Information Item 2.3 1
Does the plant's RCS water chemistry Technical Specificationsfollow the EPRIguidelines?
SNC Response to Requested Information Item 2.3 The FNP Technical Specifications provide limits on RCS dissolved oxygen, chlorides, and fluorides, l
and these limits are consistent with Action Levels II and III of the EPRI Primary Water Chemistry Guidelines.
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GL 97-01 Requested Information Item 2.4 1
Identify any RCS chemistry excursions that exceed the plant administrative limitsfor thefollowing species: sulfates, chlorides orfluorides, o.tygen. boron, andlithium.
SNC Response to Requested Information Item 2.4 5
Based on the determination that no resin bead intrusions have occurred at FNP, Requested Information Item 2.4 is not applicable to FNP.
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Response to Generic Letter 97-01 l
l GL 97-01 Requested Information item 2.5 Identify any conductivity excursions which may be indicative ofresin intrusions. Provide a technical assessment ofeach excursion and anyfollow-up actions.
SNC Response to Requested Information Item 2.5 See SNC response to Requested Information item 2 regarding review of RCS conductivity data.
l GL 97-01 Requested Information item 2.6 h ovide an assessment ofthepotentialfor any ofthese intrusions to result in a sigmficant increase in trae probabilityfor IAA of17iPs and any associatedplan for inspections.
SNC Response to Requested Information Item 2.6 Based on the determination that no resin bead intrusions have occurred at FNP, Requested Information item 2.6 is not applicable to FNP.
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