ML20133B542
| ML20133B542 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/12/1985 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML20133B528 | List: |
| References | |
| AEP:NRC:0906E, AEP:NRC:906E, NUDOCS 8508060203 | |
| Download: ML20133B542 (3) | |
Text
{{#Wiki_filter:7 \\ \\ INDIANA & MICHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 April 12, 1985 AEP:NRC:0906E Donald C. Cook Nuclear Plant Docket Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 ADDITIONAL INFORMATION CONCERNING NRC INSPECTION REPORT 50-315/84-13 (DRS) ; 50-316/84-15 (DRS) Mr. James G. Keppler U.C. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Keppler:
This letter provides additional information concerning two issues addressed in the subject report. The first issue involves the implementation of activities related to a 1979 Interpretation (XI-01-79-18) of ASME Code, Section XI-1977. The second issue involves an assessment of our valve program to determine program modifications, which would be needed to assure adequate tracking of valve failure rates and to assure an adequate valve trending program is in place. Our letter AEP:NRC:0906B, dated January 30, 1985 indicated that a written description of the results of the assessment would be provided by April 1,1985. Based on a discussion with Mr. P. Wohld of your staff on March 26, 1985, we agreed to provide by mid-April, 1985 a description of the results of the assessment and a description of how we will implement Interpretation XI-01-79-18. Issue 1 Based on discussions with your staff, Interpretation XI-01-79-18, states that surveillance, of accessible valves with remote indication, are to be performed with an observer near the valves. The observer, along with another cbserver near the remote indicators, would assure agreement between valve movement and the remote indicators. Per our agreement with your staff on March 27, 1985, the following describes the activities we will implement with regard to Interpretation XI-01-79-18 and the ISI valve program as discussed with the Office of Nuclear Reactor Regulation on March 5-7, 1985. o Units 1 and 2 procedures will be revised by the end of the Unit 2 refueling outage (approximately February,1986) to address the above. This will be done by revising our procedures to comply with Paragraph IWV-3300 of the ASME B&PV Code, Section XI-1980 Edition, plus Addendum through Winter 1981. 8508060203 850731 {" ^N 0*$$5 APR FI g
Mr. Jame3 G. K:pplcr AEP:NRC 0906E g h o Unit 2 accessible valves equipped with remote indication will be surveilled per the revised procedures by the end of the Unit 2 refueling outage, Those Unit 1 valves, which have no other check (such as a partial o or full flow test) to indicate they will perform their safety function when needed, will be identified. The remote indications for the Unit 1 valves in this category will be verified per the 19^0 Code by the end of the current Unit 1 refueling outage (approximately late-July, 1985) or the valves will be tested. o Unit 1 accessible valves equipped with remote indication will be surveilled per the revised procedures, during the next scheduled surveillance following the issuance of the revised procedure. These valves will be surveilled by the end of the next Unit 1 refueling outage (approximately October,1986), o should a problem be discovered during the surveillance of an accessible Unit 2 valve, the commensurate Unit 1 valve will be evaluated when available to determine whether remote indication should be verified. Issue 2 The following is a description of the valve program modifications that have or will be implemented as a result of the aforementioned assessment, concerning valve failure rates and valve trending. Also provided are the target dates for completing the modifications: o Procedure No. 12-QHP-5070ISI.014 dated April 1, 1985 delineates (1) when valves are to be placed on increased frequency, (2) the criteria for determining a valve failure, and (3) provides criteria for valve trending. o In support of a trending program which is responsive to valve degradation, two aids will be developed. The first aid consists of a set of curves (graphs), which will be used to assure that valve degradation does not continue beyond the established limit. The second aid is a matrix, which identifies like valves that could be susceptible to a common failure. These aids are targeted to be completed by December 31, 1985. o To ensure that unacceptable valve degradation or valve failures are properly considered for trending and tracking, the appropriate plant department (ISI section) has been added to the distribution list for Condition Reports that document valve failures and unacceptable degradation, o Procedures are in place to assure that the respective Condition Reports are reviewed by cognizant personnel to identify the cause of the valve failures or unacceptable valve degradation, and the appropriate corrective actions. b
IR ~ 3 Mr. Jame2 G. Kepplsr AEP:NRC 0906E \\ o= Data accumulated as part of the ISI valve trending program will be used as a tool.in scheduling timely maintenance and replacemerit. This'dbcumenthasbeenpreparedfollowingCorporateprocedureswhich . incorporate a reasonable set of controls to insure its accuracy and completeness prior to' signature by the undersigned. Very truly yours, b{ g$ / / M. . Ale. ch p Vice Pres dent 4kl7 cm cct -John E. Dolan W. G. Smith, Jr. - Bridgman R. C. Callen G. Bruchmann G. Charnoff NRC Resident Inspector - Bridgman E.
r ae + UNITED STATES NUCLEAR REGULATORY COMMISSION -E WASHING TON, D. C. 20555 / April 11, 1985 h Docket Nos. 50-315 and 50-316 Mr. John Dolan, Vice President Indiana and Michigan Electric Company c/o American Electric Power Service Corporation 1 Riverside Plaza Columbus, Ohio 43216
Dear Mr. Dolan:
In response to the Indiana and Michigan Electric Company letter dated November 28, 1984, we hav.e completed our review of your method of performing the monthly pump test required by Technical Specification 4.7.1.2.a.2.b for the turbine driven pump for. D. C. Cook, Units 1 and 2. The subject specification 'r'eq'ufres that the turttine driven pump be demonstrated operable every 31 days by verifying that "the steem turbine driven pump develops a discharge pressure of >1285 psig at a flow of> 700 gpm when the secondary steam pressure is greater han 310 psig." The plaYt procedures for performing this test provide for adjusting the turbine speed to 4350 rpm and establishing a flow of 700 gpm. The pump discharge pressure is then recorded and compared to the 1285 psig (corrected for temperature) minimum requirement of the specification. k Region 3 (refer to Inspection Reports 50-315/84-13 and 50-316/84-15) expressed a concern that the pressure, when corrected for temperature, allows the test to be found acceptable when the pressure is less than that stated in the surveillance requirements (1285 psig). While we agree with the licensee's philosophy for correcting the discharge pressure for temperature and with the correction factors identified in the procedures, we question the methods identified in the procedure for measuring the pumped fluid temperatures. The IMEC procedures for determining condensate temperature call for placing a handheld pyrometer against the suction piping. We do not believe this is an accurate nettod of determining condensate temperature. Further, since the corrNt,)n factors involved with the range of temperatures expected in the c% un i tank (40*F/+ 1 psig - 100*F/-8 psig) are small, and may be no g e. 37-. Un inherent instrument errors and reading inaccuracies, the correction toe.cr obtained by this method appears unwarranted and unnecessary. However, we do agree that a temperature correction factor for determining the required discharge pressure during the surveillance testing procedure does not invalidate the test results on a technical basis. Based on our review, we conclude that IMEC should take one of two actions: 1. Revise the specification to identify that the 1285 psig acceptance ~ pressure is based on a fluid temperature of 60'F and develop a more t accurate method of determining the temperature of the pump fluid, or ( seh U 1DW OO o-gin -{ yp
Mr. Dolan April 11, 1985 2. Delete the correction factor from the surveillance procedure. If you have any further questions on this matter, please let us know. Si cerely, .even
- rhW, Operating Reactors I h #1 Division of Licensin cc:
3 W. Shafer, RIII B. Jurgensen, RI e e 9 J t
F p mg[o g 0; g UNITED STATES t a o NUCLEAR REGULATORY COMMISSION h
- f WASHINGTON, D. C. 20555
@tiYROUTIQ-Ps \\,*****/ APR 1 i 1985 %~w Jt- [n-i{- = v Docket No.: 50-341 FIL M D MEMORANDUM FOR: Richard L. Spessaro, Director Division of Reactor Safety Region III FROM: Hugh L. Thompson, Jr., Director Division of Licensing Office of Nuclear Reactor Regulation
SUBJECT:
RESPONSE TO REQUEST FOR TECHNICAL ASSISTANCE REGARDING MAXIMUM STR0KE TIME TESTING FOR IST OF VALVES We have reviewed the infonnation submitted in your request for technical assistance dated November 14, 1984 regarding testing of the maximum stroke time as part of the in-service testing (IST) program at the Fermi-2 facility. Our basic position on this request is that the applicant has committed to comply with the requirements of the ASME Code and has not requested specific relief from the applicable portion of the ASME Code. Our response is directed towards the third concern outlined in your letter (i.e., the acceptability of baseline data established for valve testing in accordance with Section XI of the ASME Code) since the first two concerns were previously resolved. Acceptability of Baseline Data Established for Valve Testing per Section XI With respect to the applicant's procedures for measuring valve stroke times, as described in your letter dated November 14, 1984, the staff agrees that these procedures are not in accordance with the requirements of Section XI. Subsection IWV-3a17 of the ASME Code (the Code). The use of such procedures would require prior written relief by the staff from the specific requirements of the Code. The specific applicable Code requirements are: IWV-3417 Corrective Action (a) If, for power operated valves, an increase in stroke time of 25% or more from the previous test for valves with full-stroke times greater than 10 sec or 50% or more for valves with full-stroke times less than or equal to 10 sec is observed, test frequency shall be increased to once each month until corrective action is taken, at which time the original test frequency shall be resumed. In any case, any abnormality or erratic action shall be reported. (Emphasis added).
Contact:
M. Lynch, 492-7050 ApR 1 b M ,yA ll Ohm i yy t I' app. J
3 e
- C f.
e APRy1 885 R. L. Spessard, Director. (b) If a valve fails to exhibit the required change of valve stem or disk position or exceeds its specified limiting value of full-stroke time by this testing, then corrective action shall be initiated immediately. If the condition is not, or cannot be, corrected within 24 hours, the valve shall be declared inoperative. When corrective action is required as a result of tests made during cold shutdown, the condition shall be corrected before startup. A retest showing acceptable operation shall be run following any required corrective action before the valve is returned to service. As cited above, each in-service test valve stroke time is required to be com-pared to the previous in-service test valve stroke time and is not related in any way to the design or purchase specification for a valve. Additionally, the staff does not interpret a corrective action to be the acceptance of the new stroke time measured on the first monthly test. When a valve has exceeded this criterion on one in-service test, the monthly frequency must be maintained until maintenance is performed on the valve so that it will not become inoperable. It appears that the applicant's practice for establishing maximum limiting stroke times for valves is also inconsistent with the staff's interpretation of the Code. Subsection IWV is specifically a " component" test code and, therefore, requires that the owner specify the maximum limiting stroke times for each power operated valve (IWV-3413). It is the staff's position that these limiting values of full stroke time are required to be based on reason-able engineering judgement of component (valve) operability, not minimum system requirements. System (or component) response time limitations, as stated in the applicant's FSAR or in the plant Technical Specifications, are also time limitations placed on each subcomponent of that system (or component). How-ever, the staff's position is that these response time limitations should rarely take precedence over a component-oriented limiting valve stroke time. Inasmuch as the IST program requirements become applicable when Detroit Edison declares that the Fermi-2 facility has gone " commercial," you should bring this matter to its attention so that it can be properly resolved. 2242 2:, 44 L Hugh L. Th npson, Jr., Director Division of Licensing A Office of Nuclear Reactor Regulation i
y_ pesog ,y A. UNITED STATES + 1L NUCLEAR REGULATORY COMMISSION n. I WASHINGTON, D C. 20555 b j +AM October 19, 1984 MEMORANDUM FOR: R. L. Spessard, Director Division of Reactor Safety, Region III FROM: Darrell G. Eisenhut, Director Division of Licensing, NRR SUBJECT; RELIEF REQUESTS FROM LEAK TESTING REQUIREMENTS AS STATED IN SECTION XI, SUBSECTION IWV-3420 0F THE ASME CODE TIA 84-62
REFERENCE:
R. L. Spessard memorandum to D. G. Eisenhut dated July 19, 1984; Request for Technical Assistance - Relief Requests from Leak Testing Requirements as Stated in Section XI, Subsection IWV-3420 of the ASME Code (AITS F03043684) Your July' 19, 1984 memo noted that the Commission has granted relief from leak rate testing requirements of IWV-3420 for containment isolation valves and permitted 10 CFR Part 50, Appendix J, type C testing as an alternative. This -practice has led to two questions: 1. Does granting such relief exempt licensees from specifying discreet or weighted leak rates for Category A valves addressed by the relief request? 2. Does granting such relief exempt licensees from leak rate analysis and corrective action requirements as stated in'IWV-3426 and 3427, respectively as well as those requirements stated in IWV-3420 through IWV-3425? As requested, we have reviewed the questions and the implications of the granting of exemptions-from Section XI, IWV-3420 of the ASME Code. Section XI of the ASME Code requires individual testing for each component in-the IST program, including individual acceptanc.e criteria. Containment Isolation Valves (CIVs) are required to be individually included in the IST program because of their accident mitigation service requirements. However, since licensees are required to perform leak rate testing of CIVs in accordance with 10 CFR Part 50, Appendix J, NRR has routinely granted relief from the leak rate test requirements of the ASME Code for these components. For cases where this relief is granted the staff requires that the licensee still meet i the Analysis of Leak Rates and Corrective Action requirements of the Code, paragraphs IWV-3426 and IWV-3427 of the 1980 Edition, respectively. The staff believes that a " weighted" approach is the most appropriate method of assignino allowable leak rates. This method is based on the existence of s a linear relationship between valve sizes with respect to allowable leakage l (i.e., a 6" valve would be allowed twice the leakap of a 3" valve). Additionally, when the allowable leak rates are added up for all type C tested CIVs, the total should not exceed 0.6 L. This allows a certain amount of A ~ .-m 4,* r - $ @;,1 "6 OCT291994 pp.
/ l t 2 flexibility since the 0.6L value specified by Appendix J is the maximum allowedforthecombinedc0mulativeleakratesoftypeCtestedCIVsand containment penetrations as determined by type B testing. This completes NRR review pursuant to TIA 84-62. ) t-[bk' UC' Darrell . EikEnhu, Director Division @[ofLicensing
- 3. -
i l Office of Nuclear Reactor Regulation cc: R. Wessman, NRR C. E. Norelius, RIII T. T. Martin, RI J. A. Olshinski, RII R. Denise, RIV T. W. Bishop, RV J. M. Taylor, IE J. G. Partlow, IE R. J. Bosnak, NRR F. C. Cherny, NRR J. D. Page, NRR 1 ) 4 O e e l}}