IR 05000259/1987022

Insp Repts 50-259/87-22,50-260/87-22 & 50-296/87-22 on 870526-29.Violations Noted.Major Areas Inspected:Previously Identified Enforcement Matters & Program for Inservice Testing Pumps & Valves

ML20237K993
Person / Time
Site:	Browns Ferry
Issue date:	08/21/1987
From:	Blake J, Girard E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20237K972	List: IR 05000259/1987022 ML20237K970
References
50-259-87-22, 50-260-87-22, 50-296-87-22, GL-87-06, GL-87-6, NUDOCS 8708270374
Download: ML20237K993 (13)

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[ 'o UNITED STATES I" ~ ~ ,% NUCLEAR REGULATORY COMMISSION

$  : E FIEGION 11  ;

'# 101 MARIETTA ST., N.W., SUITE 3100 o,

ATLANTA, GEORGIA 30303

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Report Nos.: 50-259/87-22, 50-260/87-22, and 50-296/87-22 Licensee: Tennessee Valley Authority 6N 3.8A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Docket Nos.: 50-259, 50-260 and 50-296 License Nos.: DPR-33, DPR-52, and DPR-68 i Facility Name: Browns Ferry 1, 2, and 3 Inspection CondU d: 26-29, 1987 Inspecto . - -

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9/S/ 97 . lake, Section Chief Date' Signed M e ials and Processes Section y sion of Reactor Safety SUMMARY  !

Scope: This routine, unannounced inspection was conducted in the areas of previously identified enforcement matters and program for inservice testing pumps and valve Results: Two potential violations were identified - Inadequate Corrective Action for Knowing Violation of Requirements, paragraph 3.a.; and Inadequate Response to Employee Concern Program Recommendation, paragraph PDR ADOCK 05000259 G PDR

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I REPORT DETAILS i Persons Contacted

• R. Lewis, Plant Manager

• C. Turnbow, Department Director, Nuclear Construction

• May, Manager of Licensing

• C. McFall, Compliance Engineer

• L. Clardy, Quality Assurance Surveillance Supervisor

• J. Martin, Assistant Plant Manager

• R. Young, Modifications Manager

• J. Looney, Assistant Site Representative, Employee Concerns Program (ECP)

J. Pleva, Assistant Site Representative, ECP C. Elledge, Site Representative, ECP P. Mann, Site Representative, ECP H. Hodges, Engineer, Technical Support Mechanical Test Section J. Savage, Compliance Engineer NRC Resident Inspectors

• Paulk, Senior Resident Inspector

• Brooks, Resident Inspector

• C. Patterson, Resident Inspector

• Attended exit interview Exit Interview The inspection scope .and findings were summarized on May 29, 1987, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection finding No dissenting comments were received from the licensee. The inspector identified two rew potential enforcement items stemming from this inspec-tion and informed the licensee that they would be notified of the status of these items after their review with Region II managemen In a telephone call on July 2,1987, Region II informed the licensee that the new items would be identified as follows:

Violation 259, 260, 296/87-22-01, Inadequate Corrective Action for Knowing Violation of Requirements, paragraph Violation 259, 260, 296/87-22-02, Inadequate Response to Employee ,

Concern Program Recommendation, paragraph On Auoust 20, 1987, the licensee was informed by the Office of Special Projects, TVA Projects Division Section Chief, that the above items are being reclassified as unresolved items pending further evaluation by NRC management for enforcement actio __

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The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspectio . Licensee Action on Previous Enforcement Matters (Closed) Unresolved Item 259, 260, 296/85-07-01): Adequacy of Actions Taken With Regard to Allegations Concerning Category 1 Support (1) Introduction References: (a) NRC Inspection Report 259, 260, 296/85-07, dated February 28, 1985 (b) Tennessee Valley Authority, ECP Investiga-tion Report, Concern No. ECP 86-BF-566-001, approved January 22, 1987 (c) NRC Inspection Report 260/80-34, dated January 22, 1981 (d) Nuclear Safety Review Staff (NSRS) Inves-tigation Report No. I-84-30-BFN, Investiga-tion of Problems Concerning Category 1 Support at Browns Ferry Nuclear Plant, issued December 7, 1984 This item identified the concern that a general foreman and a project manager had sought to meet schedule at the expense of quality by:

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encouraging submittal of work to inspection without having taken proper actions to assure that it conformed to requirements

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knowingly performing work before it was properly authorized, possibly resulting in safety systems being inoperable when neede NOTE: A specific incident involving Residual Heat Removal (RHR) system piping was described as an example where performance of work before authorization may have made the system inoperable while it was still needed for maintaining safe shutdown condition When this concern was identified by the NRC inspector during NRC '

Inspection 85-07 (Ref. (a)), the plant manager indicated that he would have the matter investigate This investigation was subsequently performed by the licensee's Employee Concerns Program (ECP) personnel and the investigation report (Ref. (b))

and supporting data were reviewed by the NRC inspector during the current inspectio In addition, the inspector discussed the investigation with involved TVA personnel. The inspector's findings and his conclusions regarding the licensee's actions in the matter are described and discussed belo i

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(2) Findings (a) The ECP investigation concluded that unauthorized work had been performed by the general foreman but that for the specific incident named, involving RHR piping and supports, evidence indicated that the piping was correctly logged out of service for modification wor The " unauthorized work" referred to in the concern and the investigation was determined to involve modifications of Units 1 and 3 pipe supports during 1983 and early 198 The unauthorized work consisted of performing work in accordance with unapproved work plans, a violation of Browns Ferry Standard Practice BF The ECP investigation found that the " specific incident" referred to occurred in Unit 1, in April or May 1983, and that, in the (specific) incident, the general foreman exposed personnel to an increased safety hazard and increased the likelihood of damage to associated hardwar The NRC inspector verified that the data obtained in the ECP investigation supported the above conclusion (b) The extent to which unauthorized work was performed could not be readily determined from record The inspector found, however, that ECP investigation testimony and such records as could be determined to indicate unauthorized work showed that the occurrence was not an isolated inciden (c) Testimony of individuals contacted in the ECP investigation indicates that the general foreman was fully aware that procedures were being violated by his having unauthorized work performed. General foreman is a second line supervi-sory positio (d) The ECP investigation did not specifically address the project (Modifications engineering) manager's involvement in the performance of unauthorized work. However, the NRC inspector found that the testimony obtained indicates that the project manager e.nd other Modifications engineering personnel were aware of the unauthorized work, apparently did not encourage it, but took no significant action to stop i One Modification engineer did state that he reported such unauthorized work to his supervisor, but it continued and the supervisor never informed him of any action taken or inquired as to whether the situation was correcte __ - ___ _ ___ _ _-___ - __ -

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general foreman's performance may have been incorrect, l his actions were " accepted" by Modifications management i

personnel (e.g.., the project manager) and it was difficult to determine the degree of fault that should be placed on him. The investigation also concluded, however, that the general foreman's credibility is now tarnished and that I worker confidence in him to be responsible and accountable is questionable from many sources. The NRC inspector found that the testimony obtained in the ECP investigation and associated data supported these conclusion (f) As noted in (a) above, investigation of a named incident involving RHR piping and supports found that personnel had been exposed to increased safety hazards and hardware to increased likelihood of damage. The hazard to personnel and hardware determined in the investigation of this incident involved removal of two pipe hangers that supported about 40 feet of 24-inch diameter RHR piping (containing water). This removal left the length of line apparently inadequately supported and subject to excessive stresse The investigation recommended that this line should be examined for evidence of movement and/or for configuration differences caused by the lack of adequate support. The ECP investigation found that the incident involved Unit 1 RHR pipe hangers H-11 and -12. At the time of the incident the hangers were scheduled to be removed per Work Plan 10270. The plan had not been approved when hanger removal was performed. The NRC inspector found this in agreement with the data and testimon (g) Based on his review of ECP investigation testimony, the NRC inspector found that the licensee may not have prepared and implemented adequate instructions for temporary support of piping while permanent hangers were being removed in the 1983 Unit I hanger replacement work. Testimony indicates that temporary support instructions were added to the work planning for subsequent 1983 Unit 3 hanger replacement wor (h) The NRC inspector was not able to clearly define a period of time during which the unauthorized work occurre However, all of the incidents referred to in the investiga-tion testimony occurred during 1983 and the first quarter of 198 (i) The inspector found that the general foreman referred to in the concern had a previous history of violating require-ments to expedite work performanc In an incident in ,

October 1980, as a foreman (prior to his promotion to general foreman), he had performed welding for which he was l

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i not officially qualified, used uncontrolled welding 'l material, failed to have a fire watch or burning permit for j welding and the welds he was responsible for were under-sized. These violations, in part, were previously cited in NRC Inspection Report 260/84-34 (Ref. (c)). According to ECP personnel, these previous violations by the general foreman were not described in his current personnel history record - they may have been recorded there at one time; but, if so, they had been purged after a period of tim (j) The NRC inspector found that TVA Corrective Action Reports (CARS)84-038 (5/4/84) and 84-058 (8/29/84), documented evidence of previous (1983 and 1984) work performed without approved work plans. The cause of the instance documented on CAR 84-038 was identified as an oversight by the general foreman and foreman. The general foreman is the one !

referred to in the concerns investigated. For CAR 84-058, the evidence itself (weld rod issuances assigned to the work plan prior to its approval) was discounted as being errte These cause determinations are contrary to ECP investigation testimony which indicated widespread knowledge that the named general foreman had been respon-sible for performance of unauthorized wor (k) ndditional charges of knowing violation of procedures by performance of unauthorized work (by the subject general foreman) were examined in a 1984 NSRS investigation (Re (d)). The NSRS investigation concluded that work had been performed using unapproved work plans. The cause of the performance of unauthorized work was stated to be that

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"the work control procedure (BF 8.3) did not contain the necessary flexibility to perform some activities without violating the procedural requirements." The investigation also concluded that Browns Ferry QC was aware of this adverse condition and had taken appropriate action to identify the condition and adequately monitor it." The inspector concluded that the important deficiency that was rot clearly identified by NSRS was that craft management personnel had knowingly and willfully violated procedures and that Modifications engineering management was aware of this and failed to take action to have it correcte The NRC inspector noted that a violation of requirements by i a QC inspector that was examined in the NSRS investigation

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had resulted in termination of the QC inspector but that no l apparent personnel action was taken against the general foreman or other management personnel who had apparently

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. 6 (1)l The inspector. -found that .the ECP in'vestigation; testimony indicated :the .following as apparent _ causes of the perform-ance of unauthorized work:

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The process for . obtaining _ . approval of : work plans needed for performance of required modifications.was inefficient and delaying work for proper.. approvals:

had adversely. affected the schedule cfor completing -

modifications and getting the plant back into servic Ability to meet schedule was the primary factor in evaluating job ' performance for promotion and pa raise Previous violations of requirements by. the general foreman had resulted .in no significant adverse

' personnel actions against hi (m) Because of the delays in obtaining approved work plans referred to in'(1) above, the plant manager had determined .

that prefab (prefabrication) work on pipe: supports could'be performed before the work plans were fully approved in order to expedite the work. This was on a risk basis - the risk being that, if requirements changed, items might have to be scrapped. Authorization to perform prefabrication (prefab) work required the plant manager's signature on control form BF-62. According to testimony reviewed by the NRC inspector, there had been no written definition of the limits on " prefab". The NRC inspector's review found, however, -that " unauthorized rork" stated to have been performed by .the general foreman would not have been considered . acceptable under the most liberal interpreta-tions of prefab authorization described by engineers and craft personnel interviewed in the ECP investigation. All of the individuals agreed, for example, that removal of i hangers from piping without full work approval would be !

unacceptabl (n) The NRC inspector's review found no specific instance where performance of unauthorized work had been confirmed to have adversely affected plant operational safety or hardware. The possible deficiency in RHR hardware referred to in (a) and (f) above has not been sufficiently investi-gated to determine whether it represents inadequacies in ,

installation or hardware. Further, if the RHR example is deficient, it may be due more to inadequate procedures 4 rather than to performance of unauthorized wor ]

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(o) The inspector found no indication ' that management-. above

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the ' general foreman ' and . project manager .(Modifications engineering supervisor) level knew of the performance of unauthorized work during the 1983.through early 1984 period in questio Upper management (plant manager) was subse-

'quently aware of charges of the' unauthorized work in mid-to late 1934 through CAR 84-038, CAR 84-058 and the NSRS investigation (Ref. (d)). When the NSRS investigation concluded .that the unauthorized work had been performed, it -also noted procedural and training corrective actions that were presumed to preclude recurrenc As ~ stated previously, the need for any specific action against the general foreman or other management personnel who' knew of

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his actions was not indicated in the'NSRS repor (p) Although -licensee management agreed to investigate the concern described in NRC Unresolved Item 85-07-01 during a February 1985 NRC inspection, the investigation was not initiated until June 1986, not completed until November 1986, and corrective actions were not agreed to-until' April 198 (q) The licensee had three general foremen responsible for performing the modification work referred to in this-matter. The inspector found that the testimony and related data only indicated significant violations of requirements by the one subject general foreman named in the original concer (r) The NRC inspector found that the licensee had responded to the ECP investigation findings stating that the following corrective actions had been or would be performe The subject general foreman is now directly respon-sible to a craft supervisor who is responsible to a Modifications Section Superviso This change provides more consistent and higher level supervision of the general foreman. Also, there has been a change in philosophy that places procedural compliance foremost in the managers reinds. Additionally, the general foreman was counseled regarding the errors in performing unauthorized work and documentation of the counseling was placed in his personnel history recor CAR 87-0061 has been initiated to address the remaining RHR piping concern (uncorrected line movement or a configuration difference resulting from the removal of supports). Also, item 10 of Corrective Action Tracking Document 30704-BFN-02

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, 8 addresses the situation and indicates that a walkdown of the RHR system will be conducted prior to Unit 3 startu Any evidence indicating the need for reanalysis will be submitted for design engineering dispositio The inspector reviewed CAR 87-0061 and found that the actions stated did not correctly address the RHR system concern described in the ECP investigation. It identified the concern to be addressed as temporary supports left on the Unit 3 RHR system piping. The concern should have been for Unit 1 RHP. piping location or configuration differing from installation requirements that applied when the modifications were performed ar.d inspecte NOTE: The NRC inspector found that the report of the ECP investigation did not clearly state that the RHR piping of concern was in Unit This could be ascertained by reading the report very closely, howeve The concern for piping location and configuration was clearly described in the repor Licensee personnel informed the inspector that Corrective Action Tracking Document 30'/04-BFN-2 also did not correctly identify the concern and its locatio (s) The inspector found that although some of the ECP investi-gation testimony indicated that records might have been falsified, no specific charges were made and no evidence of intentional falsification had been confirme (t) Concern that the subject general foreman had encouraged submittal of work to inspection without having taken proper actions to assure that it conformed to requirements was not '

I specifically investigated by the licensee. The inspector's review of ECP investigation data did not indicate any individuals had concerns that any specific hardware should be considered suspect on the basis of such charge I Reinspection of hardware described in the NSRS investiga- ]

tion (Ref. (d)) revealed only one incident where work i appeared to have been submitted for QC inspection before it had been checked by craft personnel. This does not disprove the concern, as QC may have detected and obtained corrections for nonconforming item I i

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(3) Conclusions-(a) 'A general foreman knowingly and willfully violated TVA -

procedural requirements given in Browns Ferry Standard Practice BF 8.3, by having work peformed before associated work plans were approved. It did not appear that Modifica-tions engineering personnel (including a supervisor involved in the~ work) encouraged this violation of require-ments. However, they were aware of it and took no apparent action to stop it's occurrence. This occurred in 1983 and

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early 198 Prior to the NRC reporting concern regarding the general foreman's actions, TVA failed to properly identify the general foreman's knowing ' and willful violation 'of the requirements or to take any ' action to assure that the general foreman, who had a previous history of violating requirements, would not violate requirements-in other wor Furthermore, even after the NRC had informed TVA of a concern regarding the general foreman's actions, TVA was not prompt in verifying the extent of violation or in determining corrective actio Although the concern was reported to TVA in February 1985, along with_a more serious associated charge that his actions might have endangered the capability of a safety system while it was needed. for safe shutdown, TVA did not begin to investigate the matter until June 1986. TVA's investigation report on the matter was not issued until January 1987, and proposed corrections were not accepted until April 1987, over two years after the NRC report of the concer Additionally, other superv.isory personnel (through second line management) knew of the general foreman's actions when they occurred and failed to see that the violations-were properly identified and corrected. TVA's actions in regard to the - matter described appear to violate 10 CFR 50, Appendix B, Criterion XVI requirements in that the licensee's measures to assure prompt identification and correction of conditions adverse to quality were ineffective. This is identified as an Unresolved Item 259,260,296/87-22-01, Untimely Corrective Action for Known Violation of Requirement This matter is undergoing further evaluation by NRC management for enforcement actio (b) A concern that a specific incidence of performance of unapproved work in violation of procedural requirements had endangered the capability of a safety system while it was needed for safe shutdown was determined to be incorrec Further, there was insufficient evidence to conclude that performance of unapproved work had directly resulted in any significant hardware deficiencie I

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(c)' The ECP L investigation of . the- specific incidence of' '

l unapproved. work referred to in (b) above, identified a (;

concern that .certain Unit' 1 PHR piping might' have been 4

. inadequately supported during support replacement work- and recommended that the - piping be examined ' to assure that there had been no configurational changes or line movement that might ' require reanalysi TVA's identification of -!

this concern in Corrective Action Report (CAR) 87-0061 and- ']

Corrective Action Tracking Document.'30704-BFN-02,; Item 10 was; deficient. . The concern was identified as the possible -

existence of temporary supports'. improperly remaining o Unit 3 piping, whereas, the. concern should ' have - been identified as excessive. lire movement or configuration .

deficiencies for Unit.1 piping.. TVA's failu,e to properly

' identify the concern. for examination and evaluation is being further evaluated by NRC management. This is identi-fied .as an unresolved item 259, 260, 296/87-22-02, Inadequate Response to Employee Concern Program Recommen-dation. The matter of. the adequacy of the Unit.1 piping and of the procedures used in providing temporary support of such piping wil? be examined by Region II. in their evaluation of = the corrective actions for the violatio The matter of why the licensee failed to properly describe the piping concern in their CAR and Corrective Action Tracking Document will be examined by Region II in evaluat-ing the~ licensee's determination of caus Unresolved Item 85-07-01 is considered . close Matters formerly covered by the item which require further evaluation will be addressed through the two violations identifiec conve. and through routine NRC inspection (Closed) Violation (259,260,296/84-40-01): Improper Acceptance Criteria for Pressure Isolation Valves This violation documented deficiencies in the licensee's leakage acceptance criteria for pressure isolation valves (PIVs). The permissible leak rate values specified by the licensee were ,

considered excessive as leakage could be permitted up to the limits of relief valves in the low pressure line In a response letter, dated December 18, 1984, the licensee stated their corrective actions which provided for lower permissible leakage rate Region II was concerned that even these lower rates might be excessive and inte'nded that the matter be resolved in the NRC evaluation of the licensee's inservice testing program for pumps and valve Subsequently, the 1

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NRC has ' determined that permissible . leakage rates for PIVs will be

'. treated as . a generic issue. The first action in . addressing th matter as a generic issue was initiated with the issuance' of NRC-Generic Letter 87-06. As 'this alternate NRC action will address Region II's concerns, the violation item will be considered resolve . Unresolved Items Unresolved items were identified during this inspection. Details for two unresolved items are provided in paragraph 3.a.(3) of this repor . Program for Inservice Testing. of Pumps and Valves (73756)

Region. II is currently performing a ret iew of the licensee's program for inservice testing (I3T) of pumps and -valves. This program was submitted for NRC revier in a letter from R. Gridley (TVA) to D.' Muller (NRC), dated-December 23, 198 During the current inspection, the NRC inspector and cognizant licensee engineers discussed the. IST program and several of its contained requests for relief from ASME Section XI (80W80). IST requirements. The inspector reviewed licensee documentation in support of a request to extend the testing interval for certain testable check valves (Licensee Relief Request PV-27). The inspector found that the documentation supported contentions made by the licensee as the basis for their relues The documentation reviewed was as followed:

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Failure evaluation for Nonconformance Report (NCR) BFNMEB8502 R0, approved July 3, 198 Memorandum from .N. Beasley (Browns Ferry Engineering Project Project Manager) to G. Hall (Nuclear Power Design Services Manager) dated April 4, 1985, describing vendor information indicating possible design problems in testable check valves and indicating how these problems would be addresse Browns Ferry Technical Specifications 4.5A and B

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Memorandum from N. Beasley to G. Hall dated July 10, 1985, providing a safety evaluation on the spurious actuation of testable check valve actuator B! owns Ferry Final Safety Analysis Report, Section 6.4, page 6.4-10; Section 6.5, page 6.5-21, and Section 7.3, pages 7.3-6 and 11  !

, (information on testable check valves).

The licensee determinations of maximum stroke time requirements for valves was discussed. The NRC position on setting maximum stroke times is included as an attachment to this report.

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NRC Position On Setting Maximum Stroke 4

' Times for Inservice Testing of Valves, Memo to R. Spessard from H. L. Thompson, J dated 4/11/85 l

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^$ e- og[*g . UNITED STATES f' 3 NUCLEAR REGUL'ATORY COMMISSION I

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WASHINGTON. D. C. 20665 s%

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)4-APR 1 ' '9R5 Docket No.: 50-341

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MEMORANDUM FOR: Richard L. Spessard, 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 STROKE TIME TESTING -FOR IST OF VALVES 4 We have reviewed the infomation 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 Femi-2 facilit Our basic position on this request is that the applicant has committed to comply with the requirernents of the ASME Code and has not requested specific relief from the applicable portion of the ASME Code. Our response is directed ,

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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 resolve Acceptability of Baseline Data Established for Valve Testina 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, j Subsection IWV-3a17 of the ASME Code (the Code). The use of sect protecNres 4 would require prior written relief by the staff from the specific requirements of the Cod The specific applicable Code requirements are:

IWV-3417 Corrective Action

(a) If, for power operated valves, an increese in stroke time of

! 25% or more from the previous test for valves with full-stroke times

- greater than 10 set 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 taker., at )

which time the original test f requency shell be resumed. Ir any )

case, any abnormality or erratic action shall be reporte (Errphasis added).

Contact: fi. Lynch, 492 7050

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ATTACHMENT

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R. L..Spessard,' Director -2- APR y yg

(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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the valve shall be declared inoperative. When corrective action is requi' red as a result of tests made during cold shutdown, the condition shall be corrected before startup. A retest showing

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acceptable operation shall be run following any required corrective action before the valve is returned to servic 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 inoperabl It appears that the applicant's practice for establishing maximum limiting stroke times for valves is also inconsistent with the staff's interpretation i of the Cod 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 comoonent (valve) operability, not minimum system requirements. System (or component) response time limitations, as stated in

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'Mapplicant's FSAR or in the plant Technical Specifications , are also time

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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 tim 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 re:olve hi4 N E LfHugh L. Tho'mpson, r., Director Division of Licensing Office of Nuclear Reactor Regulation

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NUCLEAR REGULATORY COMMISSION WASMlWGTON. D. C. 20655 k,***** .

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APR 1 ' '985 Docket No.: 50-341 --

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MEMORANDUM FOR: Richard L.'Spessard, Director

. Division of Reactor Safety Region 111

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FROM:

Hugh L. Thompson, Jr. , Director Division of Licensing Office of Nuclear Reactor Regulation 3 j

SUBJECT: l RESPONSE TO REQUEST FOR TECHNICAL ASSISTANCE REGARDI-l MAXIMUM STROKE TIME TESTING FOR IST OF VALVES j

We have reviewed the information 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-? facilit 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 c

towards the * a d 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 resolve Acceptability of Baseline Data Established for Valve Testino 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 sta'f agrees that these procedures are not in accordance with the requirements of Section XI, Subsection 1WV-3417 of the ASME Code (the Code).The use of such procedures would of the Code.require prior written relief by the staff from the specific requirements The specific applicable Code requirements are: .

IWV-3417 Corrective Action (a) If, for power nperated valves, an increese in stroke time of 25% or more from the previous test for valves with full-stroke times greater than 10 set or 50t 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 taker., at which time the original test frequency shall be resumed. It any case, any abnormality or erratic action shall be reporte (Eephasis added).

Contact: M. Lynch, 492-7050 15peitg496sd

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ATTACHMENT R., L. Spessard, Director -2- Apg ,

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. y 1 1985 '

(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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the valve shall be declared inoperative. When corrective action is requi' red 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 4 action before the valve is returned to servic '

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 perfonned on the valve so that it will not become inoperabl 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 Cod 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 requirement System (or component) response time limitations, as stated in

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'TWiipplicant'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 tim Inasmuch as the IST program requirements become applicable when Detroit Edison declares th.at the Femi-2 facility has gone " commercial," you should bring this matter to its attention so that it can be properly resolve hM$ M i

,e Hugh L. Thompson, r., Director '

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Division of Licensing Office of Nuclear Reactor Regulation

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