ML20248L847
| ML20248L847 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 03/29/1996 |
| From: | Conte R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20248L142 | List:
|
| References | |
| FOIA-97-365 NUDOCS 9806110413 | |
| Download: ML20248L847 (21) | |
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From: Richard 'Contel ,
To: NCD2.KP2.WAC1, HBE, WND2.WNP3.DHD, DJC1, DJV, F M,...
Date: 3/29/96 3:05pm
Subject:
ALG 95-0222, VY AOG issue Bill Cook reported and my followup conversation with VY confirmed the following:
The VY project engineer in this case today called Don Reid, who reported that the PE appeared to be "under the influence.." and threatened to g. to the NRC if VY did not settle his early retirment package.
Mr. Reid reported that they are negotiating with him and his current demands are unreasonable. Mr. Reid also reported not feeling particularly threatened by the above and VY has nothing to hide.
CC: RWC, RMG, WHR, WFK, NCD2.KP2.TTM N
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DOCUMENT NAME: G:\ BRANCH 5\VYPE.AOG To 6 a copy of this alocument, indicate h the box: 'C' = Copy without attachment / enclosure *E~ m Copy wth attachment / enclosure
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NAME W. Ruland R. Conte DVito
- DATE OFFICIAL RECORD COPY bGYw b bw ' '
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Dear Mr. Gibson,
Enclosed is a copy of NRC Inspection Report 50-271/96-03 concerning the advanced off-gas system at the Vermont Yankee facility. The
. report was suppose to have been sent to you on May 3, 1996, but it l was not sent. I apologize for the delay.
l After Mr. Massey'has had an opportunity to review the report,
, please contact me so that we arrange for an interview of Mr. Massey l at the earliest convenient date.
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( ;, . l VERMONT YANKEE AOG SCENARIO February 28, 1996: Reviewed VY internal investigation and identified Engineer Massey as potential H&I victim referred to in news article. March 6, 1996: Paneled information-OI to call Massey RE:
- j. interview.
I March 7, 1996: Called Massey at work (not there) and left i message for him to call me (voice mail). l March 8, 1996: (Friday a.m.) Called Massey and Secy answered-Massey out, will be in Monday (I did not leave name or message). (1:45 p.m.) Contacted Massey at his house, explained why I was calling him. His response, "Do I-have to talk to you, I love ! Vermont Yankee. . . " Discussed interview and he requested to think about it and that I should call him back at work on Monday after he had an opportunity to talk to his lawyer. He said he was sick, and he sounded sick. March 11, 1996: (Monday) called Massey, he said call his lawyer (David Gibson). (1030 a.m.) Called Gibson- not in- left
-message for him to call me RE: Massey.
March 12, 1996: (1040 a.m.) Called Gibson (same as above). (1215 p.m.) Gibson called me, I was at lunch. (1230 p.m.) Called Gibson- on the phone- will call me back. 1 (1235 p.m.) Gibson called me- I explained my j intentions- he will encourage Massey to talk , to me and he will get back to me by the end of l the week. I also explained to him that the staff needed to talk to him regarding - the technical issues only, and asked if he would encourage Massey to cooperate regardless of their decision to talk to me. March 15, 1996: (Friday) Called Gibson- not in office- left message for him to call me. March 18, 1996: (Monday, 1:15 p.m.) Called Gibson (same as ; above and he has not contacted me as of this date, March 21, 1996. l
- _ _ _ _ _ _ ___ ~' ._ - - - - _ - - - _ _ _ _ - _ _ _ _ _ _ _ ~~~-,--m ___ _ ''~'----,_ ____ -~----_________
t-On March 20, 1996, I learned that John Calvert spoke with Gibson on March 19th and informed him that Massey was not willing to talk to him. l l l I 1 j 4
9 1 Y' % 6 c
, ?# UNITED STATES 6 NUCLEAR REGULATORY COMMISSION
$ g'# ' [*j ),
, REGloN t o a 475 ALLENDALE ROAD
,d KING OF PPUSSIA, PENNSYLVANIA 19406-1415 l
May 3, 1996 Mr. Donald Reid Vice President, Operations Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Ferry Road Brattleboro, Vermont 05301
SUBJECT:
NRC INSI ' TION REPORT 50-271/96-03
Dear Mr. Reid:
On March 30, 1996, the NRC completed an inspection at your Vermont Yankee reactor facility. The enclosed report presents the results of that inspection. During the 6-week period covered by this inspection period, the conduct of activities at the Vermont Yankee facility reflected safety conscious operations, sound engineering and maintenance practices, and proper implementation of radiological work controls. Collectively, the number of plant issues identified by the VY staff during this inspection period (for example, the torus water temperature and battery room block wall issues) reflected positively on your staff's renewed dedication to identifying and resolving problems. In view of the types and significance of issues being identified, we encourage the continued diligence of your staff in this area. Also during this inspection period, we reviewed the current technical status of the advanced off-gas (A0G) system, as part of the verification process for your letter (No. BVY 96-17), dated February 26, 1996. The inspector found no engineering, operation, or maintenance indications in the last five years that the A0G system functionality was impaired in such a manner that led to degraded conditions that exceeded the Technical Specification requirements either for minimum channel availability, A0G-related instrumentation, or i system operability, No reply to this report is necessary and your cooperation with us is appreciated. Sincerely,
<- . i l
. Richard J. Conte, Chief i Reactor Projects Branch 5 Division of Reactor Projects Docket No. 50-271 l 8 f l 4@'f!Too68 Sjp ;
L f 2 L Enclosure I: Executive Summary Enclosure 2: NRC Resident Inspection Report Enclosure 3: NRC Specialist Inspector A0G Review Report i cc w/ encl:
'R. Wanczyk, Plant Manager J. Thayer, Vice President, Vermont Yankee Nuclear Power Corporai J. Duffy, Licensing Engineer, Vermont Yankee Nuclear Power Corpt J. Gilroy, Director, Vermont Public Interest Research Group, Inc D. Tefft, Administrator, Bureau of Radiological Health, State ofpshire Chief, Safety Unit, Office of the Attorney General, Commonwealth Massachusetts R. Gad,' Esquire
.G. Bisbee, Esquire R. Sedano, State of Vermont, SLO-Oesignee T. Rapone, Massachusetts Executive Office of Public Safety State of New Hampshire, SLO Designee Commonwealth of Massachusetts, SLO Designee
3 Distribution w/ enc 1: Region I Docket Room-(with concurrences) PUBLIC Nuclear Safety Information Center (NSIC)
.D. Screnci, PA0 NRC Resident Inspector R.. Conte, DRP H. Eichenholz, DRP C. O'Daniell, DRP
- J. Calvert, DRS (Enclosure 3)
W. Ruland, DRS (Enclosure 3) Distribution w/ encl (VIA E-MAIL): W.. Dean, OEDO D. Dorman, NRR P. McKee, NRR Inspection Program Branch, NRR (IPAS) e
ENCLOSURE 1 EXECUTIVE
SUMMARY
Vermont Yankee Nuclear Power Station NRC Inspection Report 50-271/96-03 This integrated inspection included espects of licensee cgerations, engineering, maintenance, and plant support. The report covers a 6-week period of resident inspection. Operations Overall, the conduct of plant operations was professional and safety conscious this inspection period. The Plant Operations Review committee (PORC) exhibited.a clear and well defined safety focus during their examination of recent Inservice Testing and Appendix J Program discrepancies and during their review of a proposal for alternate Appendix R compensatory measures. Prompt and effective actions were taken by the VY staff to resolve the operability concern involving the loose valve operators on the manual isolation valves to both residual heat removal heat exchangers. r - Maintenance 3 A number of maintenance and testing activities were observed and found to be well coordinated, with good pre-evolutionary briefings and good _ communications. Plant staff response to the March 26 recirculation pump trip' was good, however, the apparent cause of the trip was identified to have been personnel error. . An inspection follow-up item (IFI 96-03-01) was assigned to a review VY's root cause evaluation and corrective actions. The VY staff's approach to monitoring and understanding the scram solenoid
- pilot valve VIT0N diaphragm degradation issue has been and continues to be aggressive. However, VY's increased frequency of individual rod scram time testing may potentially conflict with Technical Specifications 4.3.C.2 if appropriate administrative controls are not instituted. Pending further VY staff.and inspector. review, this-issue is unresolved (URI 96-03-02).
~The VY staff's decision to postpone the reactor core isolation cooling system
.and the."B"' emergency diesel generator (EDG) limiting condition for operation
-(LCO) maintenance outages, during this inspection period, demonstrated prudent decision making with safety benefits.
Encineerino Identification of the battery room masonry wall seismic qualification calculation' errors demonstrated an excellent questioning attitude on the part , of the individual engineer. The engineering and plant staff handling of this
' design non-conformance, with respect to promptly dispositioning the station batteries operability impact, was not timely. PORC's review of the station batteries operability determination was completed and, as referenced above, j the PORC's-decision to postpone the "B" EDG LCO maintenance outage was. i i
i i l L - l 96c5!SC%$ 2lpj2 -
prudent. The NRC staff review of this potentially degraded condition using l the guidance of Generic. Letter 91 18 was ongoing at the conclusion of the inspection period and was unresolved (URI 96-03-03). VY engineering and operating staffs' have appropriately dealt with the torus water temperature limit concern, to date, by pursuing further. design basis analyses and, in the interim, administrative 1y restricting torus water temperature to 90 degrees F. Pending completion of formal analysis of this potential design basis conflict and NRC staff review, this issue is unresolved (URI 96-03-04). Licensee identified and corrected discrepancies in the Inservice Testing and
-Appendix J Programs (reference LERs 96-001 and 90-004, respectively) were dispositioned as non-cited violations. These discrepancies were identified by the VY staff as a result of thorough corrective action for organizational problems identified via the Fire Protection and Appendix R Programs.
The inspector reviewed the current technical status of the advanced off-gas (A0G) system as part of the verification process for the licensee's letter (No. BVY 96-17), dated February 26, 1996. In particular, the inspector reviewed issues dealing with A0G system performance and with a system modification cancellation. The engineering staff's coordination with the plant staff, the quality of the consolidated as-built panel 9-50 electrical drawings, and the delineation and resolution of design issues for the planned A0G modification were generally very good. The inspector found no indication that the cancellation of the modification was driven by cost considerations other than the inherent cost risk associated with implementing a modification with possible incomplete documentation, such as installation and test instructions. The inspector found no engineering or maintenance indications in the last 5 years that A0G system functionality was impaired in such a manner that led to degraded conditions that exceeded the Technical Specification requirements. Recent initiatives including system-analyzed maintenance developed by the I&C engineering' staff and reliability-based maintenance. developed by the maintenance engineering staff were considered good. Plant' Support VY's ongoing systematic re-examination of the entire Fire Protection and Appendix R Programs identified a number of improperly installed fire dampers and incomplete test data for the switchgear rooms carbon dioxide suppression systems. .The compensatory measures for these discrepancies were promptly implemented and the proposed corrective actions deemed appropriate. i Conclusive system test results to support a system operability determination are still pending and this issue remains unresolved (URI 96-03-06). VY staff review of plant refueling practices identified that preceding the 1990 and 1992 refuel outages all three layers of reactor vessel shield blocks were removed while at power. This condition was determined to have been in conflict with the plant design basis. The apparent root cause of this problem was inadequate procedural guidance, but further evaluation was ongoing. Pending VY completion and inspector review of the final root cause evaluation, this issue is unresolved (URI 96-03-05). 1 ii E__________________________._____._________________.__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ __ g
ENCLOSURE 2 I U.S. NUCLEAR REGULATORY COMMISSION
,, REGION I 1
Docket No. 50-271 - Licensee Nu. DPR-28 Report No. 96-03 Licensee: Vermont Yankee Nuclear Power Corporation Facility: Vermont Yankee Nuclear Power Station Location: Vernon, Vermont Dates: February 2 - Narch 30, 1996 Inspectors: William A. Cook, Senior Resident Inspector Approved by: RichardJ. Conte,p61ef,ProjectsBranch5 Division of Reactor Projects l l l l l l l iii
TABLE OF C0KfENfS . EXECUTIVE
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . . . . . . . i TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
SUMMARY
OF PLANT STATUS . . . . . . . . . . . . . . . . . . . . . . . . . I 1.0 OPERATIONS ............................ I 1.1 Conduct of Operations . . . . . . . . . . . . . . . . . . . . I 1.2 Operational Status of Facilities and Equipment ....... 1 1.2.1 Safety Parameter Display System (SPDS) Out of Service . I 1.2.2 Ba' tery Room Block Wall 10 CFR 50.72 Notification . . . 2 1.2.3 To is Water Temperature Design Limit Concern ..... 2 1.3 Operatic Procedures and Documentation . . . . ....... 2 1.3.1 Mo :hly Statistical Reports . . . . . . . . . . . . . . 2 1.3.2 Fut1 Failure Status and Parameter Trends Report . . . . 2 1.4 Quality Assurance in Operations . . . . . . . . . . . . . . . 3 On Site Review Comittee Activities . . . . . . . . . . . . . 3 1.5 Miscellaneous Operations Issues . . . . . . . . . . . . . . . 3 (Closed) LER 50-271/96-06 . . . . . . . . . . . . . . . . . . 3 2.0 MAINTENANCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1 Conduct of Maintenance ................... 2.1 - 4 2.1.1 Review of Maintenance and Surveillance Testing .... 4 2.1.2 Recirculation Pump Trip Due to Maintenance Personnel Error . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.2 Maintenance Procedures and Documentation .......... 6 Single Rod Scram Time Testing Update ............ 6 2.3 Quality Assurance in Maintenance Activities . . . . . . . . . 7 LCO Maintenance Postponed . . . . . . . . . . . . . . . . . . 7 3.0 ENGINEERING . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 I 3.1 Conduct of Engineering ................... 8 I Battery Room Block Wall Seismic Qualification . . . . . . . . 8 3.2 Quality Assurance in Engineering Activities . . . . . . . . . 10 Torus Water Temperature Administrative Limit ........ 10 3.3 Miscellaneous Engineering Issues .............. 11 3.3.1 (Closed) LER 50-271/96-01 . . . . . . . . . . . . . . . 11 3.3.2 (Closed) LER 50-271/96-04 . . . . . . . . . . . . . . . 11 4.0 PLANT SUPPORT . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 4.1 Radiological Protection Controls .............. 12 Reactor Vessel Shield Block Rsuoval Outside Design Basis .. 12 4.2 Status of Fire Protecticn Facilities and Equipment ..... 13 4.2.1 Fire Dampers Not Properly Installed . . . . . . . . . . 13 4.2.2 Switchgear Rooms Carbon Dioxide Suppression System Decl ared Inoperable . . . . . . . . . . . . . . . . . . 14 5.0 REVIEW OF UFSAR COMMITMENTS . . . . . . . . . . . . . . . . . . . . 14 iv
6.0 MANAGEMENT MEETINGS . . . . . . . . . . . . . . . . . . . . . . . . 15 6.1 Exit Meeting Sumary .................... 15 INSPECTION PROCEDURES USED ....................... 15 ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . 15 i V
l I l DETAILS SUMARY OF PLANT STATUS , i l Vermont Yankee (VY) operated at 100 percent reactor power throughout this ! inspection period with the exception of power reductions to conduct planned rod pattern exchanges and surveillance testing. i On March 5, VY representatives met with the NRC staff in Headquarters at 1 Rockville, Maryland to discuss issues related to the Fire Protection and 10 CFR 50 Appendix R Programs at the station. The VY staff provided status of their progress and plans for establishing compliance with Appendix R prior to the completion of the 1996 refueling outage. On March 13, VY representatives met with the NRC staff in Headquarters to , discuss the reactor vessel shroud repair and reactor vessel internals ; inspection plans. Details of the 1995 outage shroud visual inspection results and planned 1996 outage shroud repair plans were shared with the NRC reviewers and inspection staff. The formal repair plan and plant modification approval request has not been submitted for NRC staff review; to date. On March 28, VY representatives met with the NRC staff in Headquarters to discuss the seismic qualification concerns involving the station 125 VDC battery room masonry block walls. The VY representatives provided additional information pertaining to the specific seismic qualification methodology being used to support their masonry block whil operability determination (reference Section 3.1.1 of this report). 1.0 OPERATIONS 1.1 Conduct of Operations Using Inspection Procedure 71707, the inspector conducted frequent reviews of on-going plant operations. The conduct of operations, as observed in the l control room and in the plant, was professional and safety conscious. Specific operating events and noteworthy inspector observations are detailed in the sections below. l 1.2 Operational Status of Facilities and Equipment l l 1.2.1 Safety Parameter Display System (SPDS) Out of Service (71707, 93702) ; I l On March I at 3:40 p.m., VY notified the headquarters duty officer in l accordance with 10 CFR 50.72 (Event No. 30053) that the SPDS was out of I service for greater than eight hours. The inspector determined that the Emergency Response Facility.Information System (EFIS), which includes the SPDS, had been removed from service at 7:40 a.m. for preventive maintenance. An unanticipated ' problem with the data acquisition system delayed restoration l of the system within the planned eight-hour work window. The ERFIS/SPDS t:as I returned to service at 8:35 p.m. on March 1. i l l The inspector verified that compensatory measures to manually conduct thermal l ! heat balance calculations were in-place, as well as, an alternate means to ! l verify reactor core thermal limits, since the ERFIS provides these functions. 1
2 Neither back-up method was warranted because the ERFIS was restored in less than 24 hours. The inspector concluded that VY made the appropriate 10 CFR 50.72 notification for this temporary loss of emergency assessment capability l and that appropriate compensatory measures were available. Inspector follow-up of a related 10 CFR 50.72 notification (Event No. 29846) on January 17, identified that similarly appropriate compensatory measures were available, but not needed. The January 17 SPDS outage lasted eight hours and 12 minutes and was the result of a planned system outage to replace the plant process computer system. The inspector determined that unforeseeable l minor software changes delayed restoration of the SPDS within eight hours on January 17. 1.2.2 Battery Room Block Wall 10 CFR 50.72 Notification (71707) On March 18, VY made a 10 CFR 50.72 notification identifying that the battery room masonry block wall did not meet current design basis criteria for seismic qualification. Additional details and inspector observations are in Section 3.1.1. I 1.2.3 Torus Water Temperature Design Linit Concern (71707) On March 26, VY made a 10 CFR 50.72 notification identifying that the current Technical Specification (TS) torus water maximum temperature limit may be non-conservative. This determination was based upon a new primary containment loss of coolant accident (LOCA) response analysis. Additional details and inspector observations are in Section 3.2.1. 1.3 Operations Procedures and Documentation (71707) 1.3.1 Monthly Statistical Reports l The inspector reviewed the Monthly Statistical Reports for January and February 1996, dated February 10 and March 10, 1996, respectively. The inspector verified that these reports were submitted in accordance with TS 6.7.A.3 and properly reflected the operating status of the facility during the months of January and February 1996. 1.3.2 Fuel Failure Status and Parameter Trends Report , The inspector reviewed the internal monthly fuel failure status and trend l reports for the months of December 1995, January 1996, and February 1996. These reports are generated by the site reactor engineering group to monitor and trend fuel performance by examining offgas radiation levels and reactor i coolant activity analyses. Between December 1995 and February 1996 no fuel I failures occurred and no abnormal trends were observed. The inspector found this plant performance monitoring appropriate and the summary reports well written and concise. l I l
3 1.4 Quality Assurance in Operations On Site Review Committee Activities
- a. Inspection Scope On February 28, the inspector observed routine Plant Operations Review Committee (PORC) meeting No. 96-019. The inspector observed the PORC review of two Licensee Event Reports (LERs 96-01 and 96-04) and Basis for Maintaining Operability (BMO) No. 95-07, Revision 1. These types of items obtain a PORC review prior to approval and issuance by the Plant Manager. At PORC meeting 96-019, the inspector also' observed a discussion pertaining to proposed alternate compensatory measures for Appendix R issues.
- b. Observations and Findings The inspector observed the PORC members' review and discussions of LERs 96-01 and 96-04'and BM0 95-07, Rev. I', to have been thorough and insightful.' The committee ensured that questions raised, pertaining to BM0 95-07 by the members from the previous PORC meeting quorum, were appropriately addressed and resolved in Revision 1 to BM0 95-07. The inspector also noted that the PORC members asked some broader safety questions during their discussion of LER 96-04 which resulted in appropriate PORC followup items. Additional NRC observations of PORC are discussed in Sections.2.3.1 and 3.1.
The observed discussion of the proposed alternative Appendix R compensatory measures, presented.by the Fire Protection Improvet.ent Plan (FPIP) task force, i demonstrated a clear understanding by the PORC u n bers of their chartered responsibilities. Briefly, the PORC rejected the FPIP task force's request to review and comment on their alternative compensatory measures proposal. The- ,
.PORC rejected-the review of this proposal because it had not already been !
scrutinized through appropriate. administrative procedural review and approval (including a 50.59 safety evaluation,'if needed) processes.
' The inspector considered PORC's refusal to engage in a discussion of the l
r.erits of the alternative Appendix R compensatory measures proposal without the conduct of prerequisite reviews as entirely appropriate.
- c. Conclusion The. inspector concluded that the PORC appropriately implementd its charter !
and exhibited a clear and well defined safety focus during the r review of the f above stated agenda topics. The PORC's rejection'of the FPIP task force ,
. proposal exemplified their high standards of procedural review and defense-in-
-depth through required administrative processes.
1.5 Miscellaneous Operations Issues (92700) I (Closed) LER 50-271/96-06: Potentially Inoperable Residual Heat Removal (RHR) , Service Water Valves Due to the Bolts Holding the Valve operators Being ; L Insufficiently Tight, hted March 14, 1996 L i
4 On February 14, VY made a 10 CFR 50.72 notification (Event No. 29975 identifying the entry into a TS required shutdown due to both RHR loops' heat exchanger manual outlet isolation valves (RHR-192A & B) being declared inoperative. Both RHR-192A & B being inoperable resulted in the containment cooling, RHR service water, and alternate cooling systems being declared inoperable at 10:15 a.m. VY declared the valves inoperable because the four blind bolts affixing the manual valve operators to the valve bonnets had become loose and, if sufficiently un-threaded, could have potentially prevented valve operation. The valve design is such that, if the operator became detached, the direction of flow through the valve could cause it to shut and stop cooling water flow to the heat exchanger. The inspector observed that corrective maintenance was promptly initiated and each valve was removed from service, repaired, tested and returned to service, one after the other. The sequencing of valve repairs ensured one train was available while repairs were effected on the other. Both trains were restored by 3:00 p.m..on February 14 and the TS shutdown terminated. During valve repair, the blind bolts were found to be slightly loosened, but not sufficiently to prevent operation of the manual valve operator. The inspector concluded that VY took prompt and effective action to resolve this system operability issue upon identification of its potential safety impact. The corrective maintenance was well planned, executed, and controlled. The LER was well written and concise. VY committed to submit a supplement to LER 96-06 upon completion of the root cause evaluation. 2.0 MAINTENANCE 2.1 Conduct of Maintenance 2.1.1 Rev'.ew of Maintenance and Surveillance Testing
- a. Inspection Scope Using Inspection Procedures 62703 and 61726, the inspector reviewed all or portions of the following maintenance and surveillance testing activities, e Monthly surveillance testing of the A and B emergency diesel generators conducted on February 20.
- Post-maintenance testing of the B control rod drive pump conducted on February 23 and March 1.
e- Control rod pattern exchange and single rod scram time testing conducted
.on February 27. . High pressure coolant injection systems full flow testing conducted in accordance with procedure OP-4120 on March 7.
- Post-modification testing of recirculation loop sample valves V2-39 and V2-40 conducted on March 3. 1
- b. Observations Findings and Conclusions The inspector monitored the pre-evolutionary briefings conducted in the control room and found the briefings to be well structured and comprehensive.
i
i 5 Testing personnel clearly understood the test acceptance criteria and tne ! step-by-step testing sequence. The inspector observed good communications and
.. coordination of testing personnel activities with routine plant evolutions.
2.1.2 Recirculation Pump Trip Due to Maintenance personnel Error (93702)
-a. Background On March 26 at 11:41 a.m., while operating at 100 percent reactor power, the "B" recirculation pump tripped. Operators appropriately responded to the i recirculation pump trip and stabilized the unit at approximately 50 percent power. The "A" recirculation punsp.was manually ramped back and control rods were manually insertei to exit the' Buffer Region of the TS Section 6.7.A.4 established Core Ops iting Limits Report, Figure 2.4-1 (power-to-flow map). -No core flow instabi - ties were observed and no reactor protection system challenges occurred.
I The VY staff-determined that the cause of the recirculation pump trip was l personnel error. The error involved a contractor electrician conducting ! routine scheduled meter calibrations. While performing the calibration of the ' T-7-1A transformer local ammeter (B phase) usina a generic work procedure, .the electrician mistakenly left the associated ameter circuit overcurrent trip . relays (518 and 50B) un-bypassed. Upon inserting the calibration test current to the ammeter, the overcurrent relays tripped causing the 4160V to 480V station service transformer T-7 supply breaker to trip which de-energized Bus
- 7. The de-energizing of Bus 7 removed the "B" recirculation pump motor-generator set lubricating oil pump from service, causing the motor-generator set to trip on low lubricating oil pressure, and causing the "B" recirculation pump trip. .
Following verification of the cause of Bus 7 being de-energized, the bus was 'l re-energized at 11:55 a.m. and electrically loaded. The calibration of the T- l 7 transfctmer local ammeter was aborted and all subsequent scheduled meter calibrations suspended following a detailed root cause evaluation and the implementation of corrective actions. The "B" recirculation pump was returned to service at 1:56 p.m. and reactor power subsequently restored to 100 percent. j b.- Observations and Findings-The inspector verified that reactor systems and balance of plant systems ! responded, as designed. One plant computer (ERFIS) software problem was identified involving the failure of the flow instability monitoring program , (SOLOMON) to properly initiate upon B recirculation pump trip. This software ' programming. anomaly was promptly and appropriately addressed prior to the conclusion of the inspection period. The inspector observed control room operators actions to exit the Buffer l Region of the power-to-flow map. This potential flow instability region was exited by 12:17 p.m. and the inspector noted prompt and appropriate actions by the control room operators to exit this operating region. The shift supervisor demonstrated effective control and coordination of the shift crew
6
- and the supporting plant staff who responded to this event. The shift crew l
communications were clear and concise with an excellent team approach to addressing this operating challenge. l Operators did experience some minor difficulty in selecting control rods per the insertion sequence procedure, but were successful in sddressing this rod
- select problem. The problem encountered was subsequently traced by the-i instrumentation and controls (IkC) staff to high resistance switch contacts in
! the reactor manual control system rod select matrix. The effected pushtutton l selector switch was replaced and successfully post-maintenance tested. The inspector reviewed the I&C staff actions to address this operating concern and considered them prompt and appropriate. c.. Conclusions
- Overall plant staff response to this event was good. Coordination and communications were effective in promptly diagnosing the cause of the "B" L recirculation pump trip and.in stabilizing the reactor plant. The root cause evaluation for this event was not completed by the close of this inspection period and will be reviewed _by the inspector for thoroughness and adequacy of l corrective actions. This is an inspection follow-up item (IFI 96-03-01).
! 2.2 , Maintenance Procedures and Documentation l . l Single Rod Scram Time Testing Update (92901) J
- a. Background As previously discussed in inspection report 50-271/95-25, the VY staff I observed an increase in the individual control rod notch 46 drop-out times.
This increase in scram time was attributed to an apparent degradation of the i scram solenoid pilot valve (SSPV) VITON elastomer diaphragms and the SSPV endcap design. Since the VY staff discovered their SSPV diaphragm concerns in early November 1995, other licentees have experienced similar problems and the ! Boiling Water Reactor Owners' Group (BWROG) Regulatory Response Group (RRG) has developed interim recommendations regarding the VITON diaphragm issue. By letter dated February 16, 1996, the BWROG RRG promulgated their interim recommendations to all affected boiling water reactor plants with dual-type SSPVs containing VITON diaphragms. The VY staff documented their endorsement i and proposed implementation plans for these recommendations by letter to the NRC, dated March.25, 1996. This letter stated that VY intends to meet or exceed the RRG's recommendations for testing both the SSPVs and the alternate rod' insertion (ARI) system. As stated in their March 25 letter to the NRC, VY will test 15 control rods during each rod pattern exchange. Prior to the issuance of this letter, VY . had single rod scramed 15 control rods on January 9 and 15 rods on February i 27 during scheduled rod pattern exchanges. An average of 0.005 seconds increase in notch 46 drop-out time was observed on February 27 for the 15 ; j control rods tested. The inspector noted that this increase was slightly less ; than anticipated by the reactor engineering staff (responsible for monitoring L_________.________.__________._____________________
7 f' control rod scram time results). Based upon the RRG recommendation, VY has scheduled rod pattern exchanges and individual rod scram time testing for April 23 (56 day interval), June 11 (49 day interval), and August 23 (73 day interval). In addition, VY stated that the recommendations for ARI system testing were verified to have already been instituted via the existing
. surveillance testing procedures.
b.- Observations and Findings The inspector verified that the number of control rods selected for scram time I testing and the interval between testing was consistent with the RRG recommendations. As previously discussed in inspection report 95-25, the VY . I staff has demonstrated and continues to demonstrate an aggressive approach to' l understanding and monitoring the recent performance of the control rod SSPVs. l' -This aggressive approach is again reflected in their commitment to the BWROG l RRG recommendations. ~Notwithstanding, VY TS 4.3.C.2 states that scram time
-testing of 50 percent of the control rod drives in each quadrant be conducted
- "not more frequently than 16 weeks nor less frequently than 32 weeks-1 intervals." The inspector notes that the lower bound (16-week interval) to l this scram time testing requirement limits the frequency.of testing of certain i control rods. ' The control rod scram time testing schedule outlined in the l March 25, 1996 letter potentially conflicts with TS 4.3.C.2, if. appropriate controls are not in place to ensure the proper rod selection for this testing.
-Pending.VY staff review of this observation and inspector verification of the procedural controls in place to ensure proper control rod selection and TS compliance, this issue is unresolved (URI 96-03-02).
- c. Conclusion l The VY staff's approach to monitoring and understanding the scram solenoid pilot valve VITON diaphragm degradation issue has been and continues to be i aggressive. However, VY's increased frequency of individual rod scram time !
testing may potentially conflict with Technt:a1 Specifications 4.3.C.2 if ! appropriate administrative ~ controls are not instituted. As discussed above, this issue is unresolved. 2.3 . Quality Assurance in Maintenance Activities LCO Maintenance Postponed (62703, 40500)
- a. . .0 observations and Findings l
L .During this inspection period, VY station management postponed scheduled j safety system limiting condition for operation (LCO) maintenance outages on
~
J two separate occasions. The first instance involved a planned maintenance i outage for the reactor core isolation cooling (RCIC) system scheduled to i comence the week of March 3. The inspector observed frequent discussions !
'during the preceding weeks' morning meetings concerning the readiness of the LC0 maintenance plan and related engineering concerns involving seismic ,
. qualification and containment integrity. On Friday March 1, the Plant Manager ;
postponed this maintenance activity based upon insufficient documented
- l. !
l
8 resolution of a number of these concerns. The second instance involved the postponement of the planned "B" emergency diesel generator (EDG) LC0 maintenance outage scheduled to commence the week of March 17. A recommendation to the Plant Manager to postpone this EDG , maintenance outage came from the PORC. On March 15, the PORC was reviewing i the station battery room issue (reference Section 3.1.1) and concluded that it i was not prudent to proceed with the "B" EDG maintenance outage due to the !
' battery room block wall seismic qualification concerns. The PORC concluded that the potential increased risk to safety system emergency power supplies was proper justification to postpone this elective maintenance activity.
- b. Conclusion The inspector concluded that the VY staff proceeded cautiously and i u thoughtfully in assessing the prudence of proceeding with the above planned j L
- LC0 maintenance outages. In both instances, VY chose to conservatively '
' postpone the elective maintenance until a more appropriate system work window I was available.
]
3.0- ENGINEERING l 3.1 Conduct of Engineering L Battery Room Block Wall Seismic Qualification (93702, 37551)- I
- a.
Background:
I [. On March 12 the design engineering staff initiated Event Report (ER) No. 96-1066 to identify two errors made in 1982 in the seismic qualification calculations for the battery room masonry block wall separating the two safety- l l related station batteries. The two errors involved-the incorrect assumption 1 that the wall was ~ constructed entirely of solid concrete blocks (only the L upper one-third of the wall is solid block and the lower two-thirds is hollow block), and that this common block wall _ was subjected to the static and dynamic-loading of two, not one, battery racks (the battery racks on either !
-' side of the wall are seismically braced by the block wall via through-wall t
threaded bolts and metal brackets). ! - The corrected seismic qualification calculation, which is derived from a linear / elastic methodology, identified the battery room block wall to have L exceeded the acceptance criteria (maximum allowable tensile stress) by l . approximately a. factor of six. - Consequently, the block. wall was. determined to l not be in accordance with the plant design basis and un operability assessment of this nonconforming condition was initiated by the engineering staff. Using a seismic qualification methodology referred to as " arching-action" (not a
- methodology reviewed and approved by the NRC staff for VY applications), the E
, -VY engineering staff was able to analytically demonstrate that the existing block wall and attached battery racks would survive a design basis earthquake.
Based upon this new engineering analysis, VY concluded the block wall could sustain a design basis seismic event and thus not adversely impact the
- operability of the station batteries.
l
9 . The inspector observed that VY's evaluation of this nonconforming casonry block wall configuration and its potential impact on station battery operability were generally consistent with the guidance of Generic Letter 91-
- 18. VY made an Emergency Notification System (ENS) call on Mat ch 18 in accordance with 10 CFR 50.72 (Event No. 30127) informing the NRC staff that the nonconforming condition placed the plant outside of its design basis. On March 28, VY representatives met with the NRC staff tr; discuss the details of the " arching-action" methodology and its specific application to the battery room block wall. Pending the NRC staff's final evaluation of the application of this unapproved methodology (for VY), this issue is unresolved (URI 96 03).
- b. Observations and Findings In monitoring the VY staff's handling of this design basis issue, the inspector made the following observations:
e Initiation of ER 96-1066 on March 12 (and subsequent March 18 ENS call) was not timely based upon the determination on March 6 that the seismic qualification calculation of record was both in error and the results exceeded the acceptance criteria.
. PORC review of the station batteries' operability determination on March 15 was completed and consistent with Generic Letter 91-18 guidance. No written explanation or summary was provided in advance or during the meeting to the PORC members, but the PORC member discussion was thorough and with clear safety focus.
. The PORC decision to. postpone (due to the station battery operability issues) the planned "B" EDG LC0 maintenance outage (scheduled to comence March 17) was prudent and indicative of a good safety perspective.
. The station was slow to formalize the Basis for Maintaining Operability (BMO) document which identifies the bases for the interim. acceptability of the nonconforming masonry block wall and the station batteries' operability assessment, and defines the corrective action plan to resolve the current design basis conflict.
- c. Conclusion The identification of this nonconforming design issue was excellent and demonstrated a good questioning attitude on the part of the individual engineer who identified the problem. The engineering and operating staffs' handling of this design discrepancy with respect to promptly dispositioning the station batteries operability impact was not timely. PORC's review of the operability determination was comleted and their decision to postpone the pending EDG LCO maintenance outage was prudent and indicative of an excellent safety focus.
l 1 10 3.2 ~ Quality Assurance in Engineering Activitie',
' Torus Water Temperature Administrative' Limit (37551)
- a. . Background'
- In November 1995, during a review of the Final Safety Analysis Repe-t (FSAR) for periodic update, the VY staff identified that elements of TS Amendment 88, approved on June 6,1985, were not completely incorporated in all appropriate FSAR sections. Amendment 88 involved an increase in initial (maximum allowable operating) torus water temperature from 90 to 100 degrees Fahrenheit
.(F). ' Closer examination of Amendment 88 and supporting documentation revealed that other safety analyses and emergency core cooling systems (ECCS) were potentially adversely impacted by this torus water temperature limit change.
Three specific areas impacted were: the loss of coolant accident (LOCA) containment analysis; maximum (post-accident) projected torus water ! temperature; and available net positive suction head to the core spray and low l
- pressure coolant injection pumps. A preliminary re-analysis by the VY )
engineering staffs of the LOCA containment response concluded that the plant would remain within its design basis assuming initial pool temperature-is at or below 90 degrees F. Consequently, by Standing order No.19, dated December 1, 1995, VY administrative 1y imposed a more restrictive torus water
. temperature operating limit of 90 degrees F.
b.- Observations and' Findings The inspector noted that continued engineering review and analysis of the ; above concorns lead VY to initiate the 10 CFR 50.72 notification (Event No. i 30175) on March 26.. A summary of the engineering staff actions were documented in a March 25, 1996 memorandum responding to Potential Adverse -l Condition Report No. 96-02. The inspector found the March 25 memorandum well I written and concise. The scope..of the torus _ water temperature issue was well -l defined and. recommendations'to resolve the outstanding potential safety issues !
'were appropriate. The inspector summarized VY's recommendations below.
.- Finalize, .via formal calculations, the preliminary assessment l restricting torus water temperature to 90 degrees F (target date - April 30,1996),
e Conduct add'itional reviews to confirm that the use of containment over- ! pressure is allowable for calculating NPSH to ECCS pumps.- ! e conduct formal analysis for maximum torus water temperature with a , ' spectrum'of LOCA break sizes, including a-stuck open safety relief l valve. Utilize the new containment model for this analysis, when available, (target date - December,1996). e Util.ize engineering design report.(EDR) 94-05 to track FSAR updates of , I the various elements of TS Amendment 88.
11
- c. Conclusion The inspector concluded that the VY engineering and operating staffs have been timely and thorough in bringing this complex design basis issue to appropriate
. resolution, to date. The administrative restriction of maximum operating torus temperature to 90 degrees F, pending the final analysis and a potential TS amendment, was a conservative safety decision. Pending completion of the VY staff's formal analysis of this potential design basis conflict and NRC staff review, the resolution of this issue remains unresolved (URI 96-03-04).
3.3 Miscellaneous Engineering Issues (92903, 92700). 3.3.1 (Closed) LER 5' 271/96-01: Technical Specification 4.6.E Not Met Due to Components Not :ncluded in the In-Service Test (IST) Program Scope,
. dated March.1, .996 The valve testing discrepancies documented in LER 96-01 were identified as a result of VY's in-depth IST review initiated as a result of corrective actions for a Notice of Violation- (reference inspection report 95-22 and LER 95-17).
The inspector, reviewed the IST deficiencies documented in LER 96-01, the BM0
- written to document the associated operability determinations and corrective action plans, and discussed these items with the responsible VY engineers. i The inspector concluded that VY's operabilit., determinations were adequately founded and that the interim and long term corrective actions to resolve the
-testing inadequacies were appropriate. The inspector did note that LER 96-01 did not discuss all of the IST valves addressed in BM0 No. 95-07, Rev.1, dated February 14,1996. .The valves not adequately tested per the IST program, as identified in BM0 95-07, and not documented in LER 96-01 were:
e High pressure' coolant injection (HPCI) pump suction check valve (V23-32) e HPCI discharge check valve (V23-18) e Both ' standby liquid control (SLC) pump discharge check valves (V11-43A & V11-438). The VY staff reviewed the inspector's observation and likewise concluded that the above valves should have been reported. At the conclusion of the i.. inspection, a supplement to LER 96-01 was being prepared to address these
. additional IST discrepancies.
The above. stated licensee identified and corrected Inservice Testing Program i violations are being treated as a non-cited violation, consistent with Section
- VII.B.1 of the "NRC Enforcement Policy." The inspector identified oversight in' reporting all of the recently discovered deficiencies (in LER 96-01) was considered an isolated case. .
3.3.2 (Closed) LER 50-271/96-04: Discrepancies. Identified in the Appendix J Leak Rate Testing Program, dated March 1, 1996 The inspector-determined that a self-assessment of the Appendix J testing program discovered these discrepancies. The self-assessment w'as initiated as i
12 a corrective action to the broad engineering management issues identified as contributors to the Appendix R/ Fire Protection Program problems identified in 1995 (reference inspection report 95-26 and Enforcement Action 95-268). For each of the Appendix J testing discrepancies, VY performed an operability determination and developed corrective actions to resolve the issue. This information was documented in BM0 No. 96-02. The inspector reviewed BM0 96-02 and LER 96-04 to assess the adequacy and thoroughness of VY's operability assessment and corrective action plans. The inspector concluded that the operability deterininations were adequate and the corrective actions appropriate, as discussed in the following paragraphs. The specific testing discrepancy involving the reactor building closed cooling water (RBCCW) system was dispositioned via structural engineers assessing the seismic qualification of the non-seismic Class 1 portions of the RBCCW system. Based upon followup discussions, the inspector considered this engineering assessment appropriate, but not well documented with respect to specific review criteria and methodology used, as well as, detailed RBCCW system walkdown observations. The testing discrepancy involving the core spray system valve was mitigated by the explication of an alternate testing method. The inspector found this alternate testing method appropriate for an interim operability determination, but as identified by the VY staff, this alternate testing methodology requires an NRC approved exemption to the ASME Code for long term use. A Code exemption was being pursued per the corrective action plan. Similarly, the specific Appendix J testing discrepancy involving the main steam isolation valves (MSIVs) was adequately being comper. sated for by an alternate leak rate testing methodology for an interim operability standpoint. However, VY plans to satisfy the requirements of their NRC approved exemption involving the MSIVs prior to the conclusion of the 1996 refuel outage. These licensee identified and corrected Appendix J testing violations are being treated as one non-cited violation, consistent with Section VII.B.1 of the "NRC Enforcement Policy." 4.0 PLANT SUPPORT 4.1 Radiological Protection Controls Reactor Vessel Shield Block Removal Outside Design Basis (92904)
- a. Background On February 21, VY made a 50.72 notification (Event No. 30005) identifying that immediately prior to entering the refueling outages in 1990 and 1992 all three layers of concrete shield blocks were removed from above the reactor vessel while the reactor was operating at greater than zero percent power.
These events were outside the plant design basis with respect to the 30-day radiation exposure to personnel in the Technical Support Center (TSC) following a design basis accident. Based upon conservative estimates, VY
13 concluded that the total dose to personnel in the TSC could potentially exceed five REM, which conflicts with the requirements of NUREG 0737, item II.B.2, and the VY design basis (reference LER 96-03).
- b. Observations and Findings The inspector determined that a procedural revision implemented prior to the 1995 refuel outage' prevented more than one layer of shield blocks from being removed prior to that reactor shutdown and subsequent refueling. Follow-up by
- the inspector also identified that, in addition to radiation shielding, the reactor vessel shield blocks provide significant mechanical barrier protection
! of the primary containment against tornado or high wind generated missiles. l similar to the radiation shielding minimum thickness calculations developed by the VY staff, the minimum missile protection requirements (thickness of concrete) provided by the reactor vessel shield blocks is 18 inches of concrete (reference FSAR Section 5.2). Accordingly, one layer of shield ' blocks (24 inches in depth) provides sufficient missile protection under the postulated design basis event.
- c. Conclusi.on At the close of the inspection period, VY had not completed their root cause l evaluation of this issue and had preliminarily attributed the cause to a lack i
of formal procedural guidance for the removal of vessel shield blocks. Pending completion of the licensee's root cause evaluation and NRC staff review, this issue is unresolved (URI 96-03-05). 4.2 Status of Fire Protection Facilities and Equipment 4.2.1 Fire Dampers Not Properly Installed (92904)
- a. Background on March 13, VY made a 50.72 notification (Event No. 30104) identifying that l 11 fire dampers installed in the ventilation duct work of the control room, cable vault, switchgear rooms, turbine lube oil room, and high pressure '
l coolant injection room were found to be installed contrary to the vendor's installation instructions. These dampers were all installed subsequent to l original construction and in response to fire protection program enhancements i L and 10 CFR 50, Appendix R upgrades made in the late Ig70's and early 1980's. The inspector determined that all of the specific installation deficiencies potentially impact the damper closure capability due to thermal expansion during a fire. The installation deficiencies include: improper caulking material; damper sleeves improperly installed or not installed around the damper; and, dampers installed in metal ventilation duct work which is of ; L insufficient gage to resist collapse during a fire. Two fire dampers wera
' identified which may not close under design conditions, in that, they are gravity drop vice spring assist and the ventilation flow rate may potentially inhibit damper closure.
l ___w--------------------- - - - - - - - - - - - - - - - - - - - - --
I j i 14 L b '. . Conclusion The inspector verified that compensatory fire watches were properly implemented and that the'VY engineering staff. was actively pursuing resolution of the identified damper problems. The inspector noted that the discovery of
.these damper design / installation concerns was the result of a broad ! programmatic review by VY of each element of their Fire Protection and !
Appendix R. Programs. This programmatic review is part of an ongoing
. corrective action to the previously identified Appendix R Program violations
.(reference inspection report 95-26 and escalated Enforcement Action No. 95- ;
'268). j 4.2.2 Switchgear Rooms Carbon Dioxide Suppression System Declared Inoperable (92904) a.: . Background On Narch 20, VY made a 10 CFR 50.7?. notification (Event No. 3014() identifying !
that the East and West switchgear. rooms carbon dioxide (CO2) suppression j systems have been declared inoperable due to insufficient post-modification ! test data. The inspector determined that another aspect of VY's broad ! programmatic review of-their Fire Protection Program (see Section 4.2.1) was'a I detailed review of the: plant C02' suppression systems. This review identified 1 that no full discharge testing results could be found to support post t j
. modification testing of the East and West switchgear CO2 suppression systems. l These CO2 suppression systems were modified via' a 1982 plant design change l (PDCR No. 82-14). j The inspector verified that a compensatory hourly fire watch was established in'accordance with Technical Specification 3.13.D, and that the CO2 suppression' system remained functional. Follow-up by the inspector identified that the VY staff has tentative plans to conduct switchgear room differential ;
prusure and test gas functional testing to demonstrate the CO2 suppression i system's' capacity to achieve and maintain a 50% CO2 concentration in the upper ! portions of the rooms. The date of this testing had not been established by
.the conclusion of the inspection period. l
- b. Conclusion The inspector' concluded that VY's immediate and planned corrective actions ,
were appropriate. Based upon insufficient testing ~ data, to date, the' l operability of.the East and West Switchgear Room C02 suppression systems is, ; and has been indeterminate. This issue is unresolved pending completion of
. testing and NRC inspector review of the test results (URI 96-03-06). ;
l' i 5.0 REVIEN OF UFSAR C005tITRENfS l
' A recent discovery ef a licensee operating their facility in a manner contrary l to the~' Updated Final Safety Analysis' Report (UFSAR) description highlighted the need for a focused review that compares the plant practices, procedures, l
-and/or parameters to the UFSAR descriptions. While performing the inspections 4 discussed in this report, the inspector reviewed the applicable portions of
!- 15 i the UFSAR the related to the areas inspected. The inspectors verified that the UFSAR wording was consistent with the observed plant practices, procedures, and/or parameters. As discussed in Section 3.2.1 above, the VY staff identified discrepancies (specifically Figure 14.6-7) where the applicable UFSAR sections had not been i revised subsequent to the NRC staff's issuance of TS Amendment 88 in June ! 1985. Also, as discussed in Section 4.1, the VY staff had on two occasions operated the unit with all of the reactor vessel shield blocks removed and thereby degraded the available primary containment missile protection. Both of these UFSAR issues are being addressed via the assigned unresolved item. 6.0 MANAGEMENT MEETINGS 6.1 Exit Meeting Summary i The inspectors presented the inspection results to members of VY management periodically throughout the inspection period and at the conclusion of the inspection on April 12, 1996. The licensee acknowledged the findings presented. , The inspectors asked the licensee whether any materials examined during the { inspection should be considered proprietary. No proprietary information was l identified. INSPECTION PROCEDURES USED , l IP 40500: Effectiveness of Licensee controls in Identifying, Resolving, and Preventing Problems ! IP 62703: Maintenance Observations IP 61726: Surveillance Observations l IP 71707: Plant Operations IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 93702: Prompt Onsite Response to Events at Operating Power Reactors IP 37551: Onsite Engineering IP 71750: Plant Support Activities IP 92901: Followup - Operations IP.92902: Followup - Maintenance IP 92903: Followup - Engineering IP 92904: Followup - Plant Support IP 90713: Review of Periodic & Special Reports ITEMS OPENED, CLOSED, AND DISCUSSED [ 9Pfll t 96003-01 IFI Recirculation pump trip due to maintenance personnel error. ( t 96003-02 URI Adequacy of administrative controls for control rod scram time testing per 4.3.C.2. ( 96003-03 URI Battery room block wall seismic qualification methodology acceptability and operability determination impact. l l
16 96003-04 URI Torus water temperature administrative limit reduced to 90 degrees from TS limit of 100 degrees. 96003-05 URI Reactor vessel shield block removal outside design bacis. 96003-06*- URI Operability verification testing of the East and West Switchgear Room CO2 Suppression System not performed. 95003-07 URI Resolve inconsistencies in license change request documentation involving recombiner catalyst service lifetime and any potential impact on catalyst maintenance. Closed 96001 LER TS 4.6.E not met due to components not included in IST program scope 96004 LER discrepancies identified in Appendix J 1eak rate testing program 96006 LER potentially inoperable RHR SW valves i l l i l I 1
l EHCLOSURE 3 U.S. NUCLEAR REGULATORY COMMISSION . l REGION I l-DOCKET / REPORT NO. 50-271/95-03 l" LICENSEE: Vermont Yankee Nuclear Power Corporation FACILITY: Vermont Yankee DATES: March 3 - 8, 1996 thru April 10 - 11, 1996 LOCATION: Vernon, VT INSPECTOR: # [E4 ?<f fh' n Calvdrt, Reactor Engineer Date L ,
/lectricalEngineeringBranch Division of Rea(. tor Safety . _ -
APPROVED: / illiam Rulan'd, Chief 4 I 4 Date / ! Electrical Engineering Branch i Division'of Reactor Safety , l-l .. l l l l I
REPORT DETAILS FOR VERM0KT YA KEE INSPECTION REPORT NO. 96-03 1.0 ADVANCED OFFGAS SYSTDI MODIFICATION REVIEW This inspection was conducted on March 3 - 8, 1996, at the Vermont Yankee site, and April 10 - 11, 1996, by telecon from the NRC Region I offices. 1.1 Scope and Background (IP37700)
- a. Scope The inspector reviewed the current technical status of the advanced off-gas (A0G) system, as part of the verification process for the licensee's letter to the NRC (No. BVY 96-17), dated February 26, 1996. The inspector reviewed those issues dealing with system performance deterioration and a modification cancellation. The radiation aspects of the A0G system are covered, in part, in reports 50-271/95-25, Section 5.1, and 50-271/95-24. The hydrogen analyzer aspects of the A0G system are covered in inspection reports 50-271/96-02, Section 3.2 and 50-271/95-25, Section 3.2.4.
The inspector' interviewed personnel associated with the A0G system and/or modification such as the design engineer, electrical engineer, I&C engineer, maintenance engineer, project engineering manager, and operations support engineer. The former project manager for the A0G modification declined to be interviewed by the inspector for this inspection. The inspector reviewed pertinent licensee documents associated with the A0G system and modification such as the UFSAR, LERs, licensee safety evaluation, I&C work orders list, I&C procedures list, I&C system-analyzed maintenance (SAM), operator rounds procedures, reliability-based maintenance (RBM), system specifications / drawings, modification drawings, drawing modification job file, design engineering memorandums, and project engineering memorandums. The inspector performed walkdowns of the control room A0G panel and accessible equipment in the A0G building.
- b. Background The A0G system was added to the Vermont Yankee plant in 1973. The purpose of the system is to process noncondensible gases removed from the main condenser to limit radioactive gaseous release to as low as reasonably achievable. The guard bed, adsorbers and associated system components are safety Class 3 (processes or houses radioactive waste) in the licensee's classification criteria. All other parts of the system are classified as non-nuclear safety.
The system performs six processes on the main condenser gases before release to the plant stack: hydrogen dilution; hydrogen recombination; preliminary delay for decay of radioactive gases; moisture removal; charcoal adsorption; i and final delay for decay of radioactive gases. Except for the passive delay pipes and charcoal adsorbers, the A0G consists of two trains of equipment with cross connection capabilities for certain equipment.
L 2 A bypass line-is installed downstream of the hydrogen dilution /recombiner trains and the preliminary delay pipe. The bypass does not include bypass of the hydrogen dilution /recombiner trains. The bypass permits continued reactor operation if portions of both trains of the moisture removal and the charcoal adsorbers were to become inoperable during normal operations. The bypass line is joined to the input of the final delay pipe for transport to the plant stack. The radiation is redundantly monitored after the bypass line, just before the final delay pipe. If the radiation monitoring levels of the monitors exceed a preset level, automatic action occurs to shut off flow to the stack from the A0G system. The plant stack has additional radiation monitoring, but no automatic control function. l 2.0 OBSERVATIONS A W FININGS 2.1 Review of 10 CFR 50.59 Safety Evaluation for the Planned Modification The inspector reviewed the document, "EDCR 94-02 Enclosure (A) Safety Evaluation." ' The equipment involved in the modification was classified non-nuclear safety. The evaluation arrived at the appropriate conclusions and showed that neither the functions of any safety-related system would be
. degraded, nor would the margin of safety be degraded as defined in the Technical Specification for the A0G system.
2.2 Review of Modification Planning The modification, EDCR 94-402, was designed by the licensee to improve the reliability operations and maintainability of the system. The modification . was originally scheduled for the 1995 and 1996 refueling outages, but has since been planned to be separated into a set of smaller tasks that could be l performed as minor modifications for the'1995 and 1996 outages. ! An initial licensee engineering evaluation was made at the beginning of the u _ modification design in 1992 to identify problems in the areas of reliability, operation, and performance of maintenance activities. The major areas listed below were identified as needing improvement, i L (1) Provide a verified, unified set of drawings for operations, maintenance, and I&C activities. Perform field verification of the as-built wiring.
- Revise the A0G control panel drawings, especially the electrical l
independence and wiring areas. The original 1973 contractor drawings were difficult to read because of layout, lettering,' lack of detail, and reproduction quality. The
. drawings and the actual electrical installation did not incorporate VY iL engineering standards, for example, redundancy and instrument fusing. !
The. revision control of the drawings was poor. In the case of the ) piping and instrumentation diagrams (P&ID), there were two identical i sets of drawings-done by two different contractors. j i
3 An example of the wiring problems was that some of the neutral wires were connected from different power supplies than the hot wires. This was viewed by the licensee as a condition that could cause erratic operation of the instruments. This is further discussed in the Section on " Status of the Modification." The' major weakness the licensee engineering identified was that, if a postulated failure of a single link in the instrument ac bus occurred, it could cause extensive failure of the A06 instrumentation. This is further discussed in Section on " Status of the Modification." The licensee stated that the functionality of the system was not hampered by these discrepancies. Maintenance could be performed, but
- not efficiently. Over the years, as minor changes were made, the confidence in the accuracy and completeness of the electrical drawings
.was questioned by operations and maintenance personnel.
.(2) Change the level control and pumping system for the A0G condensate drain tank, TK-104-1. This is an interfacing system to the A0G that is used during normal operations.
The tank receives extracted moisture from the A0G process lines. The water is then pumped to the main condenser. The tank' level provides a water barrier between the condenser vacuum and the A0G process lines. During start-ups from extended outages or in cold weather, a large volume of condensate is produced and A0G system startup times are increased. During normal plant operation, a low level of condensate is produced. There were instances when the tank was pumped dry, which the licensee found the root cause'to be regulating valve controller failure. The licensee's focus of this item was availability, rather than functionality, of the system. The modification was to replace the level control system with a type that integrated pump control and protection.
'(3) Upgrade the pressure rating for the steam jet air ejector (SJAE) inter-condenser.
This system operating pressure upgrade would increase the design margin to absorb pressure transients associated with the isolation of the A0G recombiner inlet valves. An installed rupture disc downstream of the inter-condenser currently prevents A06 system damage due to inadvertent system isolation and/or hydrogen detonations. Over the past several , years, this rupture disc has actuated due to over-pressure conditions L approximately eight times, necessitating a unit shutdown to replace the disc. With the increased inter-condenser pressure rating, the setpoint of the rupture disc could likewise be increased. The inspector < determined that system availability, rather than system functionality, was the focus of this item. The inspector notes that the potential radiological release consequences . of a rupture of the f.9G rupture disc has been the subject of previous
4 , NRC staff review and follow-up (reference inspection reports 92-01, 92-15, 93-25, and 94-27). As documented in these reports, the licensee modified the turbine building ventilation system to exhaust to the main stack to ensure proper filtering and monitoring of gaseous radiological releases. Until this modification was completed in the Fall of 1993, the licensee took appropriate. interim measures to monitor potential releases via-the turbine building ventilation system pathway. The inspector also reviewed the applicable off-normal operating procedure (No. 0N-3151,. 0ff Gas Explosion / Rupture Disc Failure) and verified that appropriate procedural guidance was in place to ensure prompt actions are taken should this event occur. In addition to the above items, eight other areas addressed in the . modification package similarly focused on system availability, maintainability and routine operations. Examples were the replacement of analog recorders with digital chartless type recorders, replacement of analog controllers with digital controllers, circuit changes to eliminate spurious A0G annunciator alarms, and re-configuration of the instrument air supply to valves to prevent unnecessary loss of recombiner heat exchangers due to loss of instrument air. 2.3 Review of Modification Design The design engineering coordination with the operations, I&C, and maintenance functions to ident.ify concerns, analyza the concerns, evaluate the alternatives, and document the design bases was performed well. This was indicated by the depth of the design analyses and the active engineering participation in site walkdowns. The consolidation and updating of the as-built drawings of the A0G control room panel 9-50 was performed well. The licensee performed a detailed point-to-point walkdown and verification of the 9-50 panel as-built drawings and ! found only one minor difference, which was corrected on the drawings. The licensee review at the site identified wiring discrepancies with the planned modification drawings that could have potentially complicated the ! installation process. This indicated good peri'ormance of the overall drawing review process, but indicated weakness in the engineering drawing check procedures for the modification. The licensee checked the electrical wiring in the A0G building to confirm conformance with the original system installation drawings. Several
. discrepancies were found and corrected. None of the discrepancies affected i
. proper functioning of the system, according to the licensee.
2.4 Status of the Modification ! The licensee management made a decision to not include the modification in the !
; April 1995 refueling outage. -The inspector reviewed two internal VY l documents, "A0G Mods Design Change," December 22, 1994, (VY Vice President, l Engineering to VY Department Manager and Project Manager), and "A0G Design Changes," January 24, 1995, (VY Vice President, Engineering to VY Department ; Manager) for reasons why the decision was made. The inspector found from the' I l u
I E__._____________________________________________________. _ . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . . . _ . .
5 review, supplemented by personnel interviews, information to suggest that the main reason for the management decision was because all the documentation l t necessary for the modification package had slipped schedule milestones. ! The modification design drawings were completed and put under drawing control, but the modification was cancelled and not implemented as part of the 1995 refueling outage. The inspector noted that the 9-50 panel wiring, where the neutral wires were from different power supplies than the hot wires, was changed as a task during the 1995 refueling outage. The change of the postulated failure of a single link in the instrument AC electrical bus and the effect on the instrumentation was not implemented as a separate task during the 1995 outage. The licensee has this task under review for future implementation. The inspector noted that if this postulated failure of a single link were to occur and if all A0G instrumentation would consequently fail, it would be covered by the Limiting Condition for Operation (LCO) in the requirements for minimum number of channels operable of Technical Specification (TS) Table 3.9.2 (2a,b,c), Gaseous Effluent Monitoring Instrumentation, which covers radiation monitoring, flow rate, and hydrogen monitors. This TS would permit continued plant operation and release of off-gas effluents via this pathway for up to seven days provided one stack radiation monitoring system was operable and off-gas temperatures and pressures were continuously monitored. 2.5 Review of Maintenance Engineering
- a. Instrument and Controls (I&C)
The I&C engineering group started a system-analyzed maintenance (SAM) project for the A0G system in 1992. The inspector reviewed the document "VY I&C Preventative Maintenance, System Analyzed Maintenance Project (SAM)," Revision 2, dated February 6, 1996, that described the purpose, objectives, and methodology of the project. The document described the method for classifying the instruments of a system according to designed function, safety class, importance to function, vendor recommendations, maintenance history, operating experience. An appropriate maintenance task with an associated interval is then assigned. The A0G system has 370 instruments, 318 of which have been classified to date and are planned for final engineering review. Approximately 85% of the instruments were classified as functionally important. The inspector reviewed a sample of ten completed worksheets for the functionally important equipment and noted that the function was clearly described, and the general requirements for the maintenance task was delineated.
- b. Mechanical The maintenance engineering group had a project similar to the IAC SAM project called reliability-based maintenance (RBM). The A0G mechanical components had all been classified and had been reviewed. The inspector selected a valve in the recombiner drain (HO-0G-587) and reviewed the basis of classification, and
i l (' 6 L l Enoted that the functional importance basis and maintenance actions were appropriate for the valve service. l-l
.2.6 Review of Maintenance Status
'The inspector reviewed a listing of work orders on the A0G system from 1991 to the present. The listing showed that the operators were alert to the identification of off-normal conditions, such as motor bearings making noise, pump cycling, air fitting leaking, and valve packing leaking that'could lead to degraded conditions. The inspector determined that the listing showed no l indication that functional problems occurred that were not addressed by corrective actions.
The inspector deterr ied that there were I&C instrument calibration procedures ; l- for the system insti tentation. Additionally, there were functional /calibrati i procedures for the A0G hydrogen moni'; ors, radiation monitors, trip systeL, moisture detectors. The work order status showed that 4 instrument calibration and functional checks had been performed. The inspector reviewed the document " Advanced Off-Gas Hydrogen Analyzer and Recombiner Catalyst," memo number VYI-2/96, dated January 5, 1996. The document stated that the licensee performed an analysis of temperature indications across a.recombiner from actual plant data covering the period l- from 1980 to 1995. The trend of temperature showed that the recombiner was operable, but indicated a loss of catalyst efficiency. The outlet temperature. data was lower than expected, which could indicate higher output hydrogen concentration. The licensee performed a grab sample check of the output that showed that the hydrogen concentration was well within the normal plant operating band. Engineering recommended that the inside and outside surfaces of the' thermocouple wells be cleaned and the thermocouple be checked as the corrective action for the low outlet temperatures. Engineering also found that the design specification for the catalyst life was ' for 18 months to 5 years, so they conservatively recommended replacement of
- the catalysts at the next refueling outage, even though the data showed that l the recombiners are properly performing their' function. This recommendation
- is being reviewed by licensee management. Further discussion of the catalyst L life is found-in Section 3.0.
2.7 System Walkdown l The A0G building was very well laid out and the inspector considered redundancy, physical separation, shielding, and maintainability. The material condition was generally very good. The A06 control room panel CRP 9-50 arrangement provided the necessary ; readouts, recorders, system mimics, and controls to operate.and determine system status. The inspector verified that the following indications were within normal operating ranges: radiation level at the SJAE; the recombiner inlet / outlet temperatures; recombiner outlet flow; hydrogen analyzer percent of combustible limit; guard bed inlet radiation; first section adsorber outlet radiation; and system outlet flow. The panel had an utensive temperature i Y________________-____.__________________---___-=-_-_---___--_-___--_____-___-_--_-----_------ - - - - - - - - - - -
7 monitoring panel for indication of system temperatures at heat exchangers, for
' example. Additionally, the panel had pressure indications for key points in the system.
3.0 REVIEW 0F UFSAR Als ColglITNENTS A recent discovery of a licensee operating their facility in a manner contrary 1
.to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR descriptions.
l While performing the inspections discussed in this report, the inspectors ; reviewed Section 9.4, " Gaseous Radwaste System," of the UFSAR that related to ; the areas inspected. The inspectors verified that the UFSAR wording for the ! A0G sy 6 i was consistent with the observed plant practices, procedures and/or ; paramet cs. The modifications of EDCR 94-402 were not checked versus the l UFSAR because the EDCR was cancelled and planned to be implemented at a later l date. I As a result of their follow-up of the' A0G system hydrogen monitor issues, the licensee noted that in their-licensing change request to the NRC for the A0G system of June 16, 1972, Attachment A, Appendix A. " Description of Offgas Processing System Components," page A-3, that the recombiner catalyst lifetime was estimated to be equivalent to plant lifetime when operating in the steam /offgas environment. They also noted inconsistent lifetime catalyst information in the proprietary Appendix C, page C-3 of Attachment A to the
' licensing change request, which described the catalyst expected life as less j than plant' lifetime. The licensee's resolution of the inconsistent recombiner !
catalyst lifetime in the A0G license change request documents and any impact j on catalyst maintenance is an unresolved item (URI 50-271/96-03-07). 4.0 MANAGEMENT OVERSIGHT Project engineering management initiated an independent engineering review of the design implementation process. Inspection information suggested that plant and engineering management made the decision to not include the modification in the 1995 shutdown because all the documentation necessary for '
- the modification package had slipped schedule milestones. The inspector l inferred from these actions that management was actively involved in the oversight of the modification.
-The inspector found no indication that the cancellation of the modification was driven by cost considerations other than the inherent' cost risk associated with implementing a modification with possible incomplete documentation.
1
5.0 CONCLUSION
S The licensee's resolution of the inconsistent recombiner catalyst lifetime in ' l the A0G license change request documents and any impact on catalyst I maintenance is an unresolved item (URI 50-271/96-03-07). l
8 . The inspector found no engineering or maintenance indications in the last 5 l' years.that A0G system functionality was impaired in such a manner that led to degraded conditions that exceeded the Technical Specification requirements l either for minimum channel availability, or A0G system instrumentation, or A0G l system operability. The inspector found no indication that the cancellation of the modification l was driven by cost considerations other than the inherent cost risk associated with implementing a modification with possible incomplete documentation, such as installation and test instructions. The engineering coordination with the piant staff, the quality of the consolidated as-built panel 9-50 electr; cal drawings, and the delineation and resolution of design issues for the plaaned A0G modification were generally l very good. The _IAC and maintenance engineering initiatives regarding A0G system analyzed I&C maintenance and reliability-based mechanical maintenance for components were good. 6.0 EXIT MEETING The findings of the inspection were presented and discussed with Mr. D. Reid, Vice President of Operations and members of the licensee's staff on March 8, 1996, as listed in'Section 7.0. The licensee acknowledged the
-findings' presented.
The inspector telephoned the licensee on April 10 and 11, 1996, for additional information on the maintenance for the hydrogen recombiner catalyst. , l The inspector received and reviewed proprietary material during the inspection and used the material for technical reference. No proprietary information was knowingly included in the report. 7.0 LIST OF PEOPLE CONTACTED Vermont Yankee Nuclear Power Corporation J. Bolvin Manager, Technical Support E. Bowman Operations Engineer B. Buteau Manager, Engineering Reorganization Coordination
*D. Calsyn Supervisor, Quality Assurance L..Casey** Design Engineer, YNSD
*R. Clark Executive Director, Quality Assurance
*P. Corbett Manager, Project Engineering
*J. DeVincento Manager,' Performance Engineering F. Helin Manager, Reactor Engineering
*S. Jefferson' Assistant to the Plant Manager G. Maret; Superintendent, Operations
*D. McElwee** State Liaison Engineer
*S. Miller ** Manager, Design Engineering, YNSD
'*J. Pelletier Executive Director. YNS
*D. Reid Vice President, Operations
9 R. Routhier Electrical Engineer
*J. Thayer Vice President, Engineering J. Todd Maintenance engineer R. Wanczyk ' Plant Manager ,
M. Watson ** Manager, I&C Vermont Department of Public Service
*W. Sherman State Nuclear Engineer U.S.N.R.C.
*J. Calvert Reactor Engineer, ElectricJ1 Engineering Branch
*W. Cook Senior Resident Inspector
- Present at exit meeting on March 8, 1996
** Contacted by telephone April 10 - 11, 1996 1
i l , 4 l l 1 f L___________ ______.__ _ __._ -____ ______ _ _ _ _ _ . . _ _ _ _ _ _ . _ _ _ _ . _ _ _ . . _ _ _ _ _ _ _ _ ___._______._._...__._______._._._________J
E I L l-d' M '# (Il J fe[ (,MJA - v/d May 3,1996 7 Mr. Donald Reid - l Vice President, Operations Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Ferry Road Brattleboro, Vermont 05301
SUBJECT:
NRC INSPECTION REPORT 50-271/96-03
Dear Mr. Reid:
On March.30,1996, the NRC completed an inspection at your Vermont Yankee reactor __ facility. The enclosed report presents the results of that inspection. During the 6-week period covered by this inspection period, the conduct of activities at the _. l Vermont Yankee facility reflected safety conscious operations, sound engineering and ! . maintenance practices, and proper implementation of radiological work controls.
- . Collectively, the number of plar.t issues identified by the VY staff during this inspection period (for example, the torus water temperature and battery room block wallissues) reflected positively on your staff's renewed dedication to identifying and resolving problems. In view of the types and significance of issues being identified, we encourage l the continued diligence of your staff in this area.
l' Also during this inspection period, we reviewed the current technical status of the i advanced off-gas (AOG) system, as part of the verification process for.your letter (No. BVY 96-17), dated February 26,~ 1996. The inspector found no engineering, operation, or maintenance indications in the last five-years that the AOG system functionality was impaired in such a manner that led to degraded conditions that exceeded the Technical ! . Specification requirements either for minimum channel availability, AOG-related instrumentation, or system operability. ! No reply to this report is necessary and your cooperation with us is appreciated. i- l l Sincerely, . l l l I l- ORIGINAL SIGNED BY: Richard J. Conte, Chief Reactor Projects Branch 5 l Division of Reactor Projects o ,
,l 3pGl5&v58 pp- .
Docket No. 50-271 Enclosure 1: . Executive Summary Enclosure 2: NRC Resident inspection Report Enclosure 3:. NRC Specialist inspector AOG Review Report cc w/ encl: R. Wanczyk, Plant Manager
. J. Thayer, Vice President, Vermont Yankee Nuclear Power Corporation
. J. Duffy, Licensing Engineer, Vermont Yankee Nuclear Power Corporation .
. J. Gilroy, Director, Vermont Public Interest Rese6rch Group, Inc.
D. Tefft, Administrator, Bureau of Radiological Health, State of New Hampshire Chief, Safety Unit, Office of the Attorney General, Commonwealth of Massachusetts R. Gad, Esquire G. Bisbee, Esquire R. Sedano, State of Vermont, SLO Designee T. Rapone, Massachusetts Executive Office of Public Safety State of New Hampshire, SLO Designee Commonwealth of Massachusetts, SLO Designee
Distribution w/ encl: Region i Docket Room (with concurrences) PUBLIC Nuclear Safety Information Center (NSIC) D. Screnci, PAO NRC Resident inspector R. Conte, DRP H. Eichenholz, DRP C. O'Daniell, DRP J. Calvert, DRE., (Enclosure 3) W. Ruland, DRS (Enclosure 3) Distribution w/enci (VIA E-MAIL): W. Dean, OEDO D. Derman, NRR j P. McKee, NRR Inspection Program Branch, NRR (IPAS) , i l 1 l l 1 i l l l l l l DOCUMENT NAME: G:\ BRANCH 5W9603Rl.CPC ! Te esselve e sepy of this document. Inascate in the ben: 'C' = Copy without ettschment/ enclosure *E' = Copy wkh ettechmentlenclosure *N* = No copy OFFICE Rl/DRP Rl/DRP / l / NAME WCook RConte l DATE 05//96 05/196 OFFICIAL RECORD COPY r I. L-___-_____-___-__
ENCLOSURE 1 l EXECUTIVE
SUMMARY
Vermont Yankee Nuclear Power Station l NRC Inspection Report 50-271/96-03 j l This integrated inspection included aspects of licensee operations. engineering, i maintenance, and plant support. The report covers a 6-week period of resident inspection. Operations
'Overall, the conduct of plant operations was professional and safety conscious this inspection period. The Plant Operations Review committee (PORC) exhibited a clear and well defined safety focus during their examination of recent inservice Testing and Appendix J Program discrepancies and during their review of a proposal for alternate Appendix R compensatory measures. Prompt and effective actions were taken by the VY staff to resolve the operability concern involving the loose valve operators on the manual isolation valves to both residual heat removal heat exchangers.
Maintenance A number of maintenance and testing activities were observed and found to be well coordinated, with good pre-evolutionary briefings and good communications. Plant staff response to the March 26 recirculation pump trip was good, however, the apparent cause i of the trip was identified to have been personnel error. An inspection follow-up item (IFl j 96-03-01) was assigned to review VY's root cause evaluation and corrective actions. i The VY staff's approach to monitoring and understanding the scram solenoid pilot valve
' VITON diaphragm degradation issue has been and continues to be aggressive. However, VY's increased frequency of individual rod scram time testing may potentially conflict with -
Technical Specifications 4.3.C.2 if appropriate administrative controls are not instituted. Pending further VY staff and inspector review, this issue is unresolved (URI 96-03-02). 1 The VY staff's decision to postpone the reactor core isolation cooling system and the "B" ] emergency diesel generator (EDG) limiting condition for operation (LCO) maintenance j outages, during this inspection period, demonstrated prudent decision making with safety benefits. - Enaineerina ; identification of the battery room masonry wall seismic qualification calculation errors demonstrated an excellent questioning attitude on the part of the individual engineer. The engineering and plant staff handling of this design non-conformance, with respect to , promptly dispositioning the station batteries operability impact, was not timely. PORC's l review of the station batteries operability determination was completed and, as referenced
~a bove, the PORC's decision to postpone the "B" EDG LCO maintenance outage was prudent. The NRC staff review of this potentially degraded condition using the guidance of i Generic Letter 91-18 was ongoing at the conclusion of the inspection period and was l
I
%CQGd9 & Nf f
unresolved (URI 96-03-03). !
' VY engineering and operating staffs' have appropriately dealt with the torus water !
! temperature limit concern, to date, by pursuing further design basis analyses and, in the l Interim, administratively restricting torus water temperature to 90 degraes F. Pending completion of formal analysis of this potential design basis conflict and NRC staff review, j this issue is unresolved (URI 96-03-04). l l L8mnsee identified and corrected discrepancies in the Inservice Testing and Appendix J !' . Programs (reference LERs 96-001 and 90-004, respectively) were dispositioned as non-cited violations. These discrepancies were identified by the VY staff as a result of l l thorough corrective action for organizational problems ider.tified via the Fire Protection and !
~ Appendix R Programs. I i
i The inspector reviewed the current technical status of the advanced off-gas (AOG) system as part of the verification process for the licensee's letter (No. BVY 96-17), dated February ] 26,1996. In particular, the inspector reviewed issues dealing with AOG system j l performance and with a system modification cancellation. The engineering staff's coordination with the plant staff, the quality of the consolidated as-built panel 9-50 electrical drawings, and the delineation and resolution of design issues for the planned i AOG modification were generally very good. The inspector found no indication that the ; cancellation of the modification was driven by cost considerations other than the inherent l cost risk associated with implementing a modification with possible incomplete l documentation, such as installation and test instructions. The inspector found no L engineering or rnaintenance indications in the last 5 years that AOG system functionality l was impaired in such a manner that led to degraded conditions that exceeded the l Technical Specification requirements. Recent initiatives including system-analyzed maintenance developed by the l&C engineering staff and reliability-based maintenance developed by the maintenance engineering staff were considered good. Plant Succort VY's ongoing systematic re-examination of the entire Fire Protection and Appendix R l- Programs identified a number of improperly installed fire dampers and incomplete test data l- for the switchgear rooms carbon dioxide suppression systems. The compensatory measures for these discrepancies were promptly implemented and the proposed corrective actions deemed appropriate. Conclusive system test results to support a system operability determination are still pending and this issue remains unresolved (URI 96-03-06).
. VY. staff review of plant refueling practices identified that preceding the 1990 and 1992
]
refuel outages all three layers of reactor vessel shield blocks were removed while at power. This condition was determined to have been in conflict with the plant design basis. The apparent root cause of this problem was inadequate procedural guidance, but further evaluation was ongoing. Pending VY completion and inspector review of the final root cause evaluation, this issue is unresolved (URI 96-03-05). 1 L---________-_.____._-___--__-_----_---._-___-__-_--. _ _ . _ . l
I ENCLOSURE 2 l l U.S. NUCLEAR REGULATORY COMMISSION REGION I i t F Docket No. 50-271 l Licensee No. DPR 28 Report No. 96-03 l l Licensee: Vermont Yankee Nuclear Power Corporation Facility: Vermont Yankee Nuclear Power Station Location: Vernon, Vermont Dates: February 2 - March 30,1996 Inspectors: William A. Cook, Senior Resident inspector l l l Approved by: Richard J. Conte, Chief. Projects Branch 5 Division of Reactor Projects i
I f TABLE OF CONTENTS EXECUTIVE
SUMMARY
i ! TABLE OF CONTENTS iv i .
SUMMARY
OF PLANT STATUS 1 l' l 1.0 OPERATIONS 1 1.1 Conduct of Operations 1 1.2 Operational Status of Facilities and Equipment 1 1.2.1 Safety Parameter Dispir.y System (SPDS) Out of Service 1 l 1.2.2 Battery Room Block Wall 10 CFR 50.72 Notification 2 1.2.3 Torus Water Temperature Design Limit Concern 2 1.3 Operations Procedures and Documentation 2 1.3.1 Monthly Statistical Reports. 2 1.3.2 Fuel Failure Status and Parameter Trends Report 2 l 1.4 - Quality Assurance in Operations 3 On Site Review Committee Activities 3 l 4 1.5 Miscellaneous Operations issues 3 l (Closed) LER 50-271/96-06 3 I
'2.0 MAINTENANCE 4 2.1 Conduct of Maintenance 4 2.1.1 Review of Maintenance and Surveillance Testing 4 2.1.2 Recirculation Pump Trip Due to Maintenance Personnel Error 5 ,
2.2 Maintenance Procedures and Documentation 6 l Single Rod Scram Time Testing Update' 6 l 2.3 Quality Assurance in Nialntenance Activities 7 j LCO Maintenance Postponed 7 L 3.0 ENGINEERING 8 ! l 3.1 - Conduct of Engineering 8 , Battery Room Block Wall Seismic Qualification 8 3.2 Quality Assurance in Engineering Activities 10 Torus Water Temperature Administrative Limit 10 ,
.3.3 Miscellaneous Engineering issues 11 ,
- 3.3.1 (Closed) LER 50-271/96-01 11 !
3.3.2 (Closed) LER 50-271/96-04 11
' 4.0 PLANT SUPPORT 12 4.1 Radiological Protection Controls 12 Reactor Vessel Shield Block Removal Outside Design Basis 12 4.2 Status of Fire Protection Facilities and Equipment 13 4.2.1 Fire Dampers Not Property Installed 13 4.2.2 Switchgear Rooms Carbon Dioxide Suppression System Declared inoperable 14 '
w
l f i 5.0 REVIEW OF UFSAR COMMITMENTS 14 6.0 MANAGEMENT MEETINGS 15 6.1 Exit Meeting Summary 15 t INSPECTION PROCEDURES USED 15 L ITEMS OPENED, CLOSED, AND DISCUSSED 15 l i I l l l [ t l
- l. ,
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DETAILS
SUMMARY
OF PLANT STATUS Vermont Yankee (VY) operated at 100 percent reactor power throughout this inspection period with the exception of power reductions to conduct planned rod pattern exchanges and surveillance testing. i On March 5, VY representatives met with the NRC staff in Headquarters at Rockville, Maryland to discuss issues related to the Fire Protection and 10 CFR 50 Appendix R
' Programs at the station. The VY staff provided status of their progress and plans for establishing compliance with Appendix R prior to the completion of the 1996 refueling outage.
On March 13, VY representatives met with the NRC staff in Headquarters to discuss the
. reactor vess31 shroud repair and reactor vesselinternals inspection plans. Details of the 1995 outage shroud visual inspection results and planned 1996 outage shroud repair plans were shared with the NRC reviewers and inspection staff. The formal repair plan and plant modification approval request has not been submitted for NRC staff review, to date.
On March 28, VY. representatives met with the NRC staff in Headquarters to discuss the seismic qualification concerns involving the station 125 VDC battery room masonry block - walls. The VY representatives provided additionalinformation pertaining to the specific seismic qualification methodology being used to support their masonry block wall operability determination (reference Section 3.1.1 of this report). 1.0 OPERATIONS.
.1.1 Conduct of Operations
- Using inspection Procedure 71707, the inspector conducted frequent reviews of on-going plant operations. The conduct of operations, as observed in the control room and in the plant, was professional and safety conscious. Specific operating events and noteworthy inspector observations are detailed in the sections below.
1.2 Operational Status of Facilities and Equipment - l 1.2.1 Safety Parameter Display System (SPDS) Out of Service (71707,93702) On March 1 at 3:40 p.m., VY notified the headquarters duty officer is: accordance with 10 CFR 50.72 (Event No.- 30053) that the SPDS was out of service for greater than eight
* ~ hours. The inspector determined that the Emergency Response Facility information I System (EFIS), which includes the SPDS, had been removed from service at 7:40 a.m. for :
. preventive maintenance. An unanticipated problem with the data acquisition system I delayed restoration of the system within the planned eight-hour work window. The ERFIS/SPDS was returned to service at 8:35 p.m. on March 1. I
' The inspector verified that compensatory measures to manually conduct thermal heat .
balance calculations were in-place, as well as, an alternate means to verify reactor core l l 1
thermal Umits, since the ERFIS provides these functions. Neither back-up method was warranted because the ERFIS was restored in less than 24 hours. The inspector concluded that VY made the appropriate 10 CFR 50.72 notification for this temporary loss of emergency assessment capability and that appropriate compensatory measures were available. Inspector follow-up of a related 10 CFR 50.72 notification (Event No. 29846) on January 17, identified that similarly appropriate compensatory measures were available, but not needed. The January 17 SPDS outage lasted eight hours and 12 minutes and was the result of a planned system outage to replace the plant process computer system. The inspector determined that unforeseeable minor software changes delayed restoration of the SPDS within eight hours on January 17. 1.2.2 Battery Room Block Wall 10 CFR 50.72 Notification (71707) On March 18, VY made a 10 CFR 50.72 notification identifying that the battery room masonry block wall did not meet current design basis criteria for seismic qualification. Additional details and inspector observations are in Section 3.1.1. P 1.2.3 Torus Water Temperature Design Limit Concern (71707) On March 26, VY made a 10 CFR 50.72 notification identifying that the current Technical Specification (TS) torus water maximum temperature limit may be non-conservative. This determination was based upon a new primary containment loss of coolant accident (LOCA) response analysis. Additional details and inspector observations are in Section 3.2.1. l 1.3 Operations Procedures and Documentation (71707) j i 1.3.1 Monthly Statistical Reports J The inspector reviewed the Monthly Statistical Reports for January and February 1996, dated February 10 and March 10,1996, respectively. The inspector verified that these reports were submitted in accordance with TS 6.7.A.3 and properly reflected the operating status of the facility during the months of January and February 1996. 1.3.2 Fuel Failure Status and Parameter Trends Report The inspector reviewed the internal monthly fuel failure status and trend reports for the months of December 1995, January 1996, and February 1996. These reports are , generated by the site reactor engineering group to monitor and trend fuel ! performance by examining offgas radiation levels and reactor coolant activity analyses. Between December 1995 and February 1996 no fuel failures occurred , and no abnormal trends were observed. The inspector found this plant performance monitoring appropriate and the summary reports well written and concise. _ _ _ _ _ _ _ - _ _ _ - _ = _ _ - _ _ _ _ - _ _ - _ _ _ _ _
l 1.4 Quality Assurance in Operations On Site Review Committee Activities
- a. Inspection Scope
~
l ' On February 28, the inspector observed routine Plant Operations Review Committee (PORC) meeting No. 96-019. The inspector observed the PORC review of two Licensee - i Event Reports (LERs 96 01 and 96-04) and Basis for Maintaining Operability (BMO) No. 95-07, Revision 1. These types of items obtain a PORC review prior to approval and issuance by the Plant Ma' nager. At PORC meeting 96-019, the inspector also observed a ! . discussion pertaining to proposed alternate compensatory mensures for Appendix R issues.
- b. Observations and Findings The inspector observed the PORC members' review and discussions of LERs 96-01 and i 96-04 and BMO 95-07, Rev.1, to have been thorough and insightful. The committee ,
ensured that questions raised, pertaining to BMO 95-07 by the members from the previous ) PORC meeting quorum, were appropriately addressed and resolved in Revision 1 to BMO l l 95-07. The inspector also noted that the PORC members asked some broader safety l l' questions during their discussion of LER 96-04 which resulted in appropriate PORC l followup items. Additional NRC observations of PORC are discussed in Sections 2.3.1 and 3.1. The observed discussion of the proposed alternative Appendix R compensatory measures, presented by the Fire Protection improvement Plan (FPlP) task force, demonstrated a clear understanding by the PORC members of their chartered responsibilities. Briefly, the PORC rejected the FPlP task force's request to review and comment on their alternative compensatory measure proposal. The PORC rejected the review of this proposal because it had not already been scrutinized through appropriate administrative procedural review and approval (including a 50.59 safety evaluation, if needed) processes. The inspector considered PORC's refusal to engage in a discussion of the merits of the ' alternative Appendix R compensatory measures proposal without the conduct of
. prerequisite reviews as entirely appropriate.
- c. Conclusion The inspector concluded that the PORC appropriately implemented its charter and exhibited a clear and well defined safety focus during their review of the above stated l agenda topics. The PORC's rejection of the FPlP task force proposal exemplified their high standards of procedural review and defense-in-depth through required administrative 3 processes.
- 1.5 - Miscellaneous Operations issues (92700) 1(Closed) LER 50 271/96-06: Potentially inoperable Residual Heat Removal (RHR) Service Water Valves Due to the Bolts Holding the Valve Operators Being insufficiently Tight,
' dated March 14,1996 l.
w_____-__-______________
1 On February 14, VY made a 10 CFR 50.72 notification (Event No. 29975 identifying the entry into a TS requirad shutdown due to both RHR loops' heat exchanger manual outlet isolation valves (RHR-192A & B) being declared inoperable. Both RHR-192A & B being
-inoperable,resulted in the containment cooling, RHR service water, and alternate cooling -systems being declared inoperable at 10:15 a.m. VY declared the valves inoperable because the four blind bolts affixing the manual valve operators to the valve bonnets had l become loose and,if sufficiently un-threaded, could have potentially prevented valve operation. The valve design is such that,if the operator became detached, the direction of flow through the valve could cause it to shut and stop cooling water flow to the heat exchanger.
The inspector observed that corrective maintenance was promptly initiated and each valve was removed from service, repaired, tested, and returned to service, one after the other. The sequencing of valve repairs ensured one train was available while repairs were effected on the other. Both trains were restored by 3:00 p.m. on February 14 and the TS shutdown terminated. During valve repair, the blind bolts were found to be slightly loosened, but not sufficiently l to prevent operation of the manual valve operator. The inspector concluded that VY took l prompt and effective action to resolve this system operability issue upon identification of !. Its poten** safety impact. The corrective maintenance was well planned, executed, and l controlled. The LER was well written and concise. VY committed to submit a supplement to LER 96-06 upon completion of the root cause evaluation. 2.0 AllAINTENANCE I 2.1 Conduct of Maintenance 2.1.1 Review of Maintenance and Surveillance Testing
- a. Inspection Scope Using inspection Procedures 62703 and 61726, the inspector reviewed all or portions of the following maintenance and surveillance testing activities.
Monthly surveillance testing of the A and B emergency diesel generators conducted on February 20. Post-maintenance testing of the B control rod drive pump conducted on February 23 l and March 1. Control rod pattern exchange and single rod scram time testing conducted on February 27. High pressure coolant injection systems full flow testing conducted in accordance , with procedure OP-4120 on March 7. I Post-modification testing of recirculation loop sample valves V2-39 and V2-40 conducted on March 8. I
- b. Observations Findings and Conclusions l
The inspector monitored the pre-evolutionary briefings conducted in the control room and found the briefings to be well structured and comprehensive. Testing personnel clearly understood the test acceptance criterla and the step-by-step testing sequence. The inspector observed good communications and coordination of testing personnel activities with routine plant evolutions. 2.1.2 Recirculation Pump Trip Due to Maintenance Personnel Error (93702)
- a. Background On March 26 at 11:41 a.m., while operating at 100 percent reactor power, the "B" recirculation purra tripped. Operators appropriately responded to the recirculation pump trip and stabilized the unit at approximately 50 percent power. The "A" re, circulation pump was manually ramped back and control rods were manually inserted to exit the Buffer Region of the TS Section 6.7.A.4 established Coze Operating Limits Report, Figure 2.4-1 (power-to-flow map). No core flow instabilities were observed and no reactor protection system challenges occurred.
The VY staff determined that the cause of the recirculation pump trip was personnel error. The error involved a contractor electrician conducting routine scheduled meter calibrations. While performing the calibration of the T-7-1 A transformer local ammeter (B phase) using a generic work procedure, the electrician mistakenly left the associated ammeter circuit overcurrent trip relays (518 and 50B) un-bypassed. Upon inserting the calibration test current to the ammeter, the overcurrent relays tripped causing the 4160V to 480V station service transformer T-7 supply breaker to trip which de-energized Bus 7. The de-energizing of Bus 7 removed the "B" recirculation pump motor-generator set lubricating oil pump from service, causing the motor-generator set to trip on low lubricating oil pressure, and causing the "B" recirculation pump trip. Following verification of the cause of Bus 7 being de-energized, the bus was re-energized .
. at 11:55 a.m. and electrically loaded. The calibration of the T-7 transformer local ammeter was aborted end all subsequent scheduled meter calibrations suspended following a detailed root cause evaluation and the implementation of corrective actions.
The "B" recirculation pump was returned to service at 1:56 p.m. and reactor power subsequently restored to 100 percent.
- b. Observations and Findings The inspector verified that reactor systems and balance of plant systems responded, as
. designed. One plant computer (ERFIS) software problem was identified involving the failure of the flow instability monitoring program (SOLOMON) to properly initiate upon B recirculation pump trip. This software programming anomaly was promptly and appropriately addressed prior to the conclusion of the inspection period.
The inspector observed control room operators actions to exit the Buffer Region of the power-to-flow map. This potential flow instability region was exited by 12:17 p.m. and the inspector noted prompt and appropriate actions by the control room operators to exit this operating region. The shift supervisor demonstrated effective control and coordination of the shift crew and the supporting plant staff who responded to this event. The shift
crew communications were clear and concise with an excellent team approach to addressing this operating challenge. Operators did experience some minor difficulty in selecting control rods per the insertion sequence procedure, but were successfulin addressing this rod select problem. The
. problem encountered was subsequently traced by the instrumentation and c,ntrols (l&C) staff to high resistance switch contacts in the reactor manual control system rod select j matrix. The effected pushbutton selector switch was replaced and successfully post-maintenance tested. The inspector reviewed the I&C staff actions to address this operating concern and considered them prompt and appropriate.
- c. Conclusions Overall plant staff response to this event was good. Coordination and communications were effective in promptly diagnosing the cause of the "B" recirculation pump trip and in stabilizing the reactor plant. The root cause evaluation for this event was not completed
. by the close of this inspection period and will be reviewed by the inspector for thoroughness and adequacy of corrective actions. This is an inspection follow-up item (IFl 96-03-01). '
2.2 Maintenance Procedures and Documentation Single Rod Scram Time Testing Update (92901)
- a. Background i
As previously discussed in inspection report 50-271/95-25, the VY staff observed an 'i increase in the individual control rod notch 46 drop-out times. This increase in scram time I was attributed to an apparent degradation of the scram solenoid pilot valve (SSPV) VITON i elastomer diaphragms and the SSPV endcap design. Since the VY staff discovered their I
, SSPV diaphragm concerns in early November 1995, other licensees have experienced similar problems and the Boiling Water Reactor Owners' Group (BWROG) Regulatory Response Group (RRG) has developed interim recommendations regarding the VITON diaphragm issue.
4 By letter dated February 16,1996, the BWROG RRG promulgated their interim recommendations to all affected boiling water reactor plants with dual-type SSPVs containing VITON diaphragms. The VY staff documented their endorsement and proposed implementation plans for these recommendations by letter to the NRC, dated March 25, 1996. This letter stated that VY intends to meet or exceed the RRG's recommendations for testin. 'soth the SSPVs and the attemate rod insertion (ARI) system. As stated in their March 25 letter to the NRC, VY will test 15 control rods during each rod . pattern exchange. Prior to the issuance of this letter, VY had single rod scraramed 15 control rods on January 9 and 15 rods on February 27 during scheduled rod pattern exchanges. An average of 0.005 seconds increase in notch 46 drop-out time was observed on February 27 for the 15 control rods tested. The inspector noted that this increase was slightly less than an:Icipated by the reactor engineering staff (responsible for , monitoring control rod scram time results). Based upon the RRG recommendation, VY has
scheduled rod pattern exchanges and individual rod scram time testing for April 23 (56 day interval), June 11 (49 day interval), and August 23 (73 day interval). In addition, VY l stated that the recommendations for ARI system testia.g were verified to have already been instituted via the existing surveillance testing p ocedures.
- b. Observations and Findings l
The inspector verified that the number of control rods selected for scram time testing and l the interval between testing was consistent with the RRG recommendations. As l previously discussed in inspection report 95-25, the VY staff has demonstrated and continues to demonstrate an aggressive approach to understanding and monitoring the recent performance of the control rod SSPVs. This aggressive approach is again reflected in their commitment to the BWROG RRG recommendations. Notwithstanding, VY TS l 4.3.C.2 states that scram time testing of 50 percent of the control rod drives in each quadrant be conducted "not more frequently than 16 weeks nor less frequently than 32
- weeks intervals." The inspector notes that the lower bound (16-week interval) to this scram time testing requirement limits the frequency of testing of certain control rods. The control rod scram time testing schedule outlined in the March 25,1996 letter potentially conflicts with TS 4.3.C.2,if appropriate controls are not in place to ensure the proper rod i
selection for this testing. Pending VY staff review of this observation and inspector verification of the procedural controls in place to ensure proper control rod selection and TS compliance, this issue is unresolved (URI 96-03-02).
- c. Conclusion l The VY staff's approach to monitoring and understanding the s: ram solenoid pilot valve ,
VITON diaphragm degradation issue has been and continues to be aggressive. However, ! VY's increased frequency of individual rod scram time testing may potentially conflict with
- l. Technical Specifications 4.3.C.2 If sippropriate administrative controls are not instituted.
l As discussed above, this issue is uriresolved, i i l 1 I r . 2.3 Quality Assurance in Maintenance Activities i LCO Maintenance Postponed (62703,40500) :
- a. Observations and Findings
. During this nspection period, VY station management postponed scheduled safety system limiting condition for operation (LCO) maintenance outages on two separate occasions.
The first instance involved a planned maintenance outage for the reactor core isolation cooling (RCIC) system scheduled to commence the week of March 3. The inspector observed frequent discussions during the preceding weeks' morning meetings concerning the readiness of_the LCO maintenance plan and related engineering concerns involving seismic qualification and containment integrity. On Friday March 1, the Plant Manager
. postpcned this mainter.ance activity based upon insufficient documented resolution of a number of these concerns.
The second instance it.volved the postponement of the planned "B" emergency diesel
Generator (EDG) LCO maintenance outage scheduled to commence the week of March 17. A recommendation to the Plant Manager to postpone this EDG maintenance outage came from the PORC. On March 15, the PORC was reviewing the station battery room issue (reference Section 3.1.1) and concluded that it was not prudent to proceed with the "B" EDG maintenance outage due to the battery room block wall seismic qualification concerns. The PORC concluded that the potentialincreased risk to safety system emergency power supplies was proper justification to postpone this elective maintenance activity.
- b. Conclusion The inspector concluded that the VY staff proceeded cautiously and thoughtfully in essessing the prudence of proceeding with the above planned LCO maintenance outages.
In both instances, VY chose to conservatively postpone the elective maintenance until a more appropriate system work window was available. 3.0 ENGINEERING
' 3.1 Conduct of Engineering Battery Room Block Wall Seismic Qualification (93702,37551)
- a. Background On March 12 the design engineering staff initiated Event Report (ER) No. 96-1066 to identify two errors made in 1982 in the seismic qualification calculations for the battery room masonry block wall separating the two safety related station batteries. The two errors involved the incorrect assumption that the wall was constructed entiraly of solid concrete blocks (only the upper one-third of the wall is solid block and the lower two-thirds
- is hollow block), and that this common block wall was subjected to the static and dynamic l' loading of two, not one, battery racks (the battery racks on either side of the wall are seismically braced by the block wall via through-wall threaded bolts and metal brackets). !
l I g The corrected seismic qualification calculation, which is derived from a linear / elastic methodology, identified the battery room block wall to have exceeded the acceptance criteria (maximum allowable tensile stress) by approximately a factor of six. Consequently, I i the block wall was determined to not be in accordance with the plant design basis and an l operability assessment of this nonconforming condition was initiated by the engineering staff. Using a seismic qualification methodology referred to as " arching-action" (not a methodology reviewed and approved by the NRC staff for VY applications), the VY ' engineering staff was able to analytically demonstrate that the existing block wall and attached battery racks would survive a design basis earthquake. Based upon this new engineering analysis, VY concluded the block wall could sustain a design basis seismic j event and thus not adversely impact the operability of the station batteries. j i i
The inspector observed that VY's evaluation of this nonconforming masonry block wall configuration and its potential impact on station battery operability were generally consistent with the guidance of Generic Letter 91-18. VY made an Emergency Notification System (ENS) call on March 18 in accordance with 10 CFR 50.72 (Event No. l. 30127) Informing the NRC staff that the nonconforming condition placed the plant outside of its design basis. On March 28, VY representatives met with the blRC staff to discuss the detniis of the " arching-action" methodology and its specific application to the battery
. room block wall. Pending the NRC staff's final evaluation of the application of this j unapproved methodology (for VY), this issue is unresolved (URI 96-03-03).
- b. ' Observations and Findings in monitoring the VY staff's handling of this design basis issue, the inspector made the l following observations:
L Initiation of ER 96-1066 on March 12 (and subsequent March 18 ENS call) was not !. timely based upon the determination on March 6 that the seismic qualification I calculation of record was both in error and the results exceeded the acceptance
- - criteria. -
PORC review of the station batteries' operability determinatica on March 15 was completed and consistent with Generic Letter 91-18 guidance. No written explanation or summary was provided in advance or during the meeting to the ! l PORC members, but the PORC member discussion was thorough and with clear ! L safety focus. The PORC decision to postpone (due to the station battery operability issues) the planned "B" EDG LCO maintenance outage (scheduled to commence March 17) was prudent and indicative of a good safety perspective. The station was slow to formalize the Basis for Maintaining Operability.(BMO) document which identifies the bases for the interim acceptability of the nonconforming masonry block wall ar.d the station batteries' operability assessment, and defines the corrective action plan to resolve the current design basis conflict. I c. Conclusion I 1 1 The identification of this nonconforming design issue was exce!!ent and demonstrated a ; good questioning attitude on the part of the individual rngineer who identified the problem. The engineering and operating staffs' handling of this design discrepancy with respect to promptly dispositioning the station batteries operability impact was not timely. PORC's review of the operability determination was comleted and their i decision to postpone the pending EDG LCO maintenance outage was prudent and indicative of an excellent safety focus. ' 4
3.2 Quality Assurance in Engineering Activities Torus Water Temperature Administrative Limit (37551)
- a. Background in November 1995, during a review of the Final Safety Analysis Report (FSAR) for periodic update, the VY staff identified that elements of TS Amendment 88, approved on June 6, 1985, were not completely incorporated in all appropriate FSAR sections. Amendment 88 involved an increase in initial (maximum allowable operating) torus water temperature from 90 to 100 degrees Fahrenheit (F). Closer examination of Amendment 88 and supporting documentation revealed that other safety analyses and emergency core cooling systems (ECCS) were potentially adversely impacted by this torus water temperature limit change.
Three specific areas impacted were: the loss of coolant accident (LOCA) containment analysis; maximum (post-accident) projected torus water temperature; and available net positive suction head to the core spray and low pressure coolant injection pumps. A preliminary re-analysis by the VY engineering staffs of the LOCA containment response concluded that the plant would remain within its design basis assuming initial pool temperature is at or below 90 degrees F. Consequently, by Standing order No.19, dated December 1,1995, VY administratively imposed a more restrictive torus water temperature operating limit of 90 degrees F.
- b. Observations and Findings The inspector noted that continued engineering review and analysis of the above concerns lead VY to initiate the 10 CFR 50.72 notification (Event No. 30175) on March 26. A summary of the engineering staff actions were documented in a March 25,1996 memorandum responding to Potential Adverse Condition Report No. 96-02. The inspector found the March 25 memorandum well written and concise. The scope of the torus water temperature issue was well defined and recommendations to resolve the outstanding potential safety issues were appropriate. The inspector summarized VY's recommendations below.
Finalize, via formal calculations, the preliminary assessment restricting torus water temperature to 90 degrees F (target date - April 30,1996). Conduct additional reviews to confirm that the use of containment over-pressure is allowable for calculating NPSH to ECCS pumps. Conduct formal analysis for maximum torus water temperature witn a spectrum of LOCA break sizes, including a stuck open safety relief valve. Utilize the new containment model for this analysis, when available, (target date - December, 1996). Utilize engineering design report (EDR) 94-05 to track FSAR updates of the various elements of TS Amendment 88.
i r.
- c. Conclusion The inspector concluded that the VY engineering and operating staffs have been timely l and thorough in bringing this complex design basis issue to appropriate resolution, to date.
The administrative restriction of maximum operating torus temperature to 90 degrees F, pending the final analysis and a potential TS amendment, was a conservative safety decision. Pending completion of the VY staff's formal analysis of trds potential design basis conflict and NRC staff review, the resolution of this issue remains unresolved (URI 96-03-04). 3.3 Miscellaneous Engineering Issues (92903,92700) l [ 3.3.1 (Closed) LER 50-271/96-01: . Technical Specification 4.6.E Not Met Due to Components Not included in the in-Service Test (IST) Program Scope, dated March 1,1996 The valve testing discrepancies documented in LER 96-01 were identified as a result of VY's in-depth IST review initiated as a result of corrective actions for a Notice of Violation (reference inspection report 95-22 and LER 95-17). The inspector reviewed the IST l deficiencies documented in LER 96-01, the BMO written to document the associated ! operability determinations and corrective action plans, and discussed these items with the responsible VY engineers. The inspector concluded that VY's operability determinations were adequately founded and that the interim and long term corrective actions to resolve the testing inadequacies were appropriate. The inspector did note that LER 96-01 did not discuss all of the IST valves addressed in BMO No. 95-07, Rev.1, dated February 14, l 1996. The valves not adequately tested per the IST program, as identified in BMO 95-07, i and not documented in LER 96-01 were: High pressure coolant injection (HPCl) pump suction check valve (V23-32) HPCI discharge check valve (V23-18) Both standby liquid control (SLC) pump discharge check valves (V11-43A &
. V11-43B) - l l
. The VY staff reviewed the inspector's observation and likewise concluded that the above I H - valves should have been reported.' At the conclusion of the inspection, a supplement to LER 96-01 was being prepared to address these additional IST discrepancies. 4 l
j - The above stated licensee identified and corrected Inservice Testing Program violations are ' being treated as a ncn-cited violation, consistent with Section Vll.B.1 of the "NRC
. Enforcement Policy." The inspector identified oversight in reporting all of the recently l discovered deficiencies (in LER 96-01) was considered an isolated case.
3.3.2 (Closed) LER 50-271/96-04: Discrepancies identified in the Appendix J Leak Rate Testing Program, dated March 1,1996 [ . The lnspector determined that a self-assessment of the Appendix J testing program discovered these discrepancies. The self assessment was initiated as L , 1 O__.________--___ _ _ _ _ _ _ - _ .
l l a corrective action to the broad engineering management issues identified as contributors to the Appendix R/ Fire Protection Program problems identified in 1995 (reference
- inspection report 95-26 and Enforcement Action 95 268).
For each of the Appendix J testing discrepancies, VY performed an operability determination and developed corrective actions to resolve the issue. This information was l' documented in BMO No. 96-02. The inspector reviewed BMO 96-02 and LER 96-04 to l assess the adequacy and thoroughness of VY's operability assessment and corrective ( action plans. The inspector concluded that the operability determinations were adequate L . and the corrective actions appropriate, as discussed in the following paragraphs. The specific testing discrepancy involving the reactor building closed cooling water (R8CCW) system was dispositioned via structural engineers assessing the seismic qualification of the non-seismic Class 1 portions of the RBCCW system. Based upon l followup discussions, the inspector considered this engineering assessment appropriate, but not well documented with respect to specific review criteria and methodology used, as l well as, detailed RBCCW system walkdown observations. The testing discrepancy involving the core spray system valve was mitigated by the application of an alternate testing method. The inspector found this alternate testing method appropriate for an interim operability determination, but as identified by the VY staff, this alternate testing methodology requires an NRC approved exemption to the ASME Coda for long term use. A Code exemption was being pursued per the corrective action plan. Similarly, the specific Appendix J testing discrepancy involving the main steam isolation valves (MSIVs) was adequately being compensated for by an alternate leak rate testing methodology for an interim operability standpoint. However, VY plans to satisfy the requirements of their NRC approved exemption involving the MSIVs prior to the conclusion of the 1996 refuel outage. These ~censee identified and corrected Appendix J testing violations are being treated as one non-cited violation, consistent with Section Vll.3.1 of the "NRC Enforcement Policy." l 4.0 PLANT SUPPORT 4.1 Radiological Protection Controls Reactor Vessel Shield Block Removal Outside Design Basis (92904)
- a. Background !
l On February 21, VY made a 50.72 notification (Event No. 30005) identifying that i immediately prior to entering the refueling outages in 1990 and 1992 all three layers of concrete shield blocks were removed from above the reactor vessel while the reactor was ( operating at greater than zero percent power. These events were outside the plant design I basis with respect to the 30-day radiation exposure to personnelin the Technical Support Center (TSC) following a design basis accident. Based upon conservative estimates, VY l l l
concluded that the total dose to personnel in the TSC could potentially exceed five REM, which conflicts with the requirements of NUFiEG 0737, item II.B.2, and the VY design basis (reference I.ER 96-03).
- b. Observations and Findings The inspector determined that a procedural revision implemented prior to the 1995 refuel outage prevented more than one layer of shield blocks from being removed prior to that reactor shutdown and subsequent refueling. Follow-up by the inspector also identified that, in addition to radiation shielding, the reactor vessel shield blocks provide significant mechanical barrier protection of the primary containment against tomado or high wind generated missiles. Similar to the radiation shielding minimum thickness calculations developed by the VY staff, the minimum missile protection requirements (thickness of concrete) provided by the reactor vessel shield blocks is 18 inches of concrete (reference
. FSNt Section 5.2). Accordingly, one layer of shield blocks (24 inches in depth) provides sufficient missile protection under the postulated design basis event.
- c. Conclusion At the close of the inspection period, VY had not completed their root cause evaluation of this issue and had preliminarily attributed the cause to a lack of formal procedural guidance for the removal of vessel shield blocks. Pending completion of the licensee's root cause evaluation and NRC staff review, this issue is unresolved (URI 96-03-05).
4.2 Status of Fire Protection Facilities and Equipment 1 4.2.1 Fire Dampers Not Properly Installed (92904) )
)
- a. Background
- On March 13, VY made a 50.72 notification (Event No. 30104) Identifying that 11 fire dampers installed in the ventilation duct work of the control room, cable vault, switchgear rooms, turbine lube oil room, and high pressure coolant injection room were found to be installed contrary to the vendor's installation instructions. These dampers were all installed subsequent to original construction and in response to fire protection program enhancements and 10 CFR 50, Appendix R upgrades made in the late 1970's and early 1980's. The inspector determined that all of the specific installation deficiencies potentially impact the damper closure capability due
- to thermal expansion during a fire. The installation deficiencies include: improper caulking material; damper sleeves improperly installed or not installed around the damper: and, dampers installed in metal ventilation duct work which is of I~
insufficient gage to resist collapse during a fire. Two fire dampers were identified which may not close under design conditions,in that, they are gravity drop vice spring assist and the ventilation flow rate may potentially inhibit damper closure.
- b. Conclusion The inspector verified that compensatory fire watches were properly implemented and that the VY engineering staff was actively pursuing resolution of the identified damper problems. The inspector noted that the discovery of these damper design / installation concerns was the result of a broad programmatic review by VY of each element of their Fire Protection and Appendix R Programs. This programmatic review is part of an ongoing corrective action to the previously identified Appendix R Program violations (reference inspection report 95-26 and escalated Enforcement Action No. 95-268).
4.2.2 _ Switchgcar Rooms Carbon Dioxide Suppression System Declared inoperable I (92904) I La. Background On March 20, VY made a 10 CFR 50.72 notification (Event No. 30146) identifying that the East and West switchgear rooms carbon dioxide (CO2) suppression systems have been declared inoperable due to insufficient post-modification test data. The inspector determined that another aspect of VY's broad programmatic review of their Fire Protection I Program (see Section 4.2.1) was a detailed review of the plant CO2 suppression systems. l This review identified that no full discharge testing results could be found to support post-modification testing of the East and West switchgear CO2 suppression systems. These CO2 suppression systems were modified via a 1982 plant design change (PDCR No. 82-14). The inspector verified that a compensatory hourly fire watch was established in accordance with Technical Specification 3.13.D, and that the CO2 suppression system remained functional. Follow-up by the inspector identified that the VY staff has tentative
- plans to conduct switchgear room differential pressure and test gas functional testing to demonstrate the CO2 suppression system's capacity to achieve and maintain a 50% CO2 concentration in the upper portions of the rooms. The date of this testing had not been established by the conclusion of the inspection period.
b.: Conclusion The inspector concluded that VY's immediate and planned corrective actions were appropriate. Based upon insufficient testing data, to date, the operability of the East and LWest Switchgear Room CO2 suppression systems is, and has been indeterminate. This issue is unresolved pending completion of testing and NRC inspector review of the test results (URI 96-03-06). 5.0 REVIEW OF UFSAR COMMITMENTS A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a focused review that compares the plant practices, procedures, and/or parameters to the UFSAR descriptions. While performing the inspections discussed in this report, the inspector reviewed the applicable portions of the UFSAR the related to the areas
' inspected. The inspectors verified that the UFSAR woroing was consistent with the i
!- observed plant practices, procedures, and/or parameters. As discussed in Section 3.2.1 above, the VY staff identified discrepancies (specifically i Figure 14.6-7) where the applicable UFSAR sections had not been revised subsequent to l
. the NRC staff's issuance of TS Amendment 88 in June 1985. Also, as discussed in Section 4.1, the VY staff had on two occasions operated the unit with all of the reactor vessel shield blocks removed and thereby degraded the available primary containment i missile protection. Both of these UFSAR issues are being addressed via the assigned unresolved item.
6.0 MANAGEMENT MEETINGS l 6.1 ' Exit Meeting Summary The inspectors presented the inspection results to members of VY management 1 periodically throughout the inspection period and at the conclusion of the inspection on April 12,1996. The licensee acknowledged the findings presented.
]
l The inspectors asked the licensee whether any materials examined during the inspection
- should be considered proprietary. No proprietary information was identified.
. INSPECTION PROCEDURES USED L
[ IP 40500: Effectiveness of Licensee Controls in identifying, Resolving, and Preventing Problems IP 62703: Maintenance Observations ! IP 61726: Surveillance' Observations IP 71707: Plant Operations IP 92700: 'Onsite Followup of Written Reports or Nontoutine Events at Power Reactor Facuities IP 93702: Prompt Onsite Response to Events at Operating Power Reactors-l IP 37551: Onsite Engineering I- IP 71750: Plant Support Activities l IP 92901: Followup - Operations j Followup - Maintenance
. IP. 92902:
IP 92903: Followup - Engineering IP 92904: Followup - Plant Support ; ! IP 90713: Review of Periodic & Special Repurts i l . ITEMS OPENED, CLOSED, AND DISCUSSED 1_ Onen I I
-_96003-01 IFl Recirculation pump trip due to maintenance personnel error.
96003-02 URI Adequacy of administrative controls for control rod scram time i testing per 4.3.C.2. i 9600343 . URJ Battery room block wall seismic qualification methodology -
' acceptability and operability determination impact. )
96003 URI Torus water temperature administrative limit reduced to 90 degrees 1 i
from TS limit of 100 degrees. 96003-05 URI Reactor vessel shield block removal outside design basis. 96003-06 URI Operability verification testing of the East and West Switchgear Room CO2 Suppression System not performed. 96003-07 URI Resolve inconsistencies in license change request documentation involving recombiner catalyst service lifetime and any potential impact on catalyst maintenance. Closed 96001 LER TS 4.6.E not met due to components not included in IST program scope 96004 LER discrepancies identified in Appendix J leak rate testing program 96006 LER potentially inoperable RHR SW valves l l l l
U.S. NUCLEAR REGULATORY COMMISSION REGION I . DOCKET / REPORT NO. 50-271/96-03 LICENSEE: -Vermont Yankee Nuclear Power Corporation FACILITY: Vermont Yankee DATES: March 3 - 8,1996 thru April 10 -11,1996 LOCATION: Vernon, VT ORIGINAL SIGNED BY: 4/26/96 INSPECTOR: 4 John Calvert, Rcactor Engineer Date . . . Electrical Engineering Branch Division of Reactor. Safety ORIGINAL SIGNED BY: 4/26/96
- APPROVED:
William Ruland, Chief Date Electrical Engineering Branch Division of Reactor Safety AM!TM(rf Up
I
)
{ REPORT DETAILS FOR VERMONT YANKEE INSPECTION REPORT NO. 96-03
' 1.0 ' ADVANCED OFFGAS SYSTEM MODIFICATION REVIEW This inspection was conducted on March 3 - 8,1996, at the Vermont Yankee site, and April 10 -.11,1996, by telecon from the NRC Region I offices.
)
1.1 Scope and Background (IP37700) 1
)
- a. Scope The inspector reviewed the current technical status of the advanced off-gas (AOG) system, as part of the verification process for the licensee's letter to the NRC (No. BVY 96-17)/ dated February 26,1996. The inspector reviewed those issues dealing with system performance deterioration and a modification cancellation. The radiation aspects of the AOG system are covered,in part,in reports 50-271/95 25, Section 5.1, and 50-271/95-24. The hydrogen analyzer aspects of the AOG system are covered in inspection reports 50 271/96-02, Section 3.2 and 50-271/95-25, Section 3.2.4.
The inspector interviewed personnel associated with the AOG systern and/or modification I such as the design engineer, electrical engineer, I&C engineer, maintenance engineer, { project engineering manager, and operations support engineer. The former project manager for the AOG modification declined to be interviewed by the inspector for this inspection. The inspector reviewed pertinent licensee documents associated with the AOG system and modification such as the UFSAR, LERO, licensee safety evaluation, I&C work orders list, l&C procedures list, l&C system-analyzed maintenance (SAM), operator rounds procedures, reliability-based maintenance (RBM), system specifications / drawings, modification drawings, drawing modification job file, design engineering memorandums, and project engineering memorandums. The inspector performed walkdowns of the control room AOG panel and accessible equipment in the AOG building,
- b. Background l
The AOG system was added to the Vermont Yankee plant in 1973. The purpose of the system is to process noncondensible gases removed from the main condenser to limit
~
radioactive gaseous release to as low as reasonably achievable. The guard bed, adsorbers ) and associated system components are safety Class 3 (processes or houses radioactive l waste)in the licensee's classification criteria. All other parts of the system are classified as non-nuclear safety. < The system performs six processes on the main condenser gases before release to the ' plant stack: hydrogen dilution: hydrogen recombination; preliminary delay for decay of radioactive gases; moisture removal; charcoal adsorption; and final delay for decay of radioactive gases. Except for the passive delay pipes and charcoal adsorbers, the AOG consists of two trains of equipment with cross connection capabilities for certain
l equipment. . 1 l l l l l 1 i I t l l 1 i l l l I i l I I i l h l I t'
n
' A bypass line is installed downstream of the hydrogen dilution /recombiner trains and the preliminary delay pipe. The bypass does not include bypass of the hydrogen dilution /recombiner trains. The bypass permits continued reactor operation if portions of both trains of the moisture removal and the charcoal adsorbe.~. were to become inoperable during normal operations. The bypass line is joined to the input of the final delay pipe for transport to the plant stack.
The radiation is redundantly monitored after the bypass line, just before the final delay pipe. If the radiation monitoring levels of the monitors exceed a preset level, automatic action occurs to shut off flow to the stack from the AOG system. The plant stack has additional radiation monitoring, but no automatic control function. 2.0 OBSERVATIONS AND FINDINGS 2.1 Review of 10 CFR 50.59 Safety Evaluation for the Planned Modification
. The inspector reviewed the document, "EDCR 94-02 Enclosure (A) Safety Evaluation."
The equipment involved in the modification was classified non-nuclear safety. The I evaluation arrived at the appropriate conclusions and showed that neither the functions of any safety-related system would be degraded, nor would the margin of safety be degraded as defined in the Technical Specification for the AOG system. 2.2 Review of Modification Planning l The modification, EDCR 94-402, was designed by the licensee to improve the reliability operations and maintainability of the system. The modification was originally scheduled for the 1995 and 1996 refueling outages, but has since been planned to be separated into l ! a set of smaller tasks that could be performed as minor modifications for the 1995 and 1996 outages. An initiallicensee engineering evaluation was made at the beginning of the modification ) design in 1992 to identify problems in the areas of reliability, operation, and performance of maintenance activities. The major areas listed below were identified as needing improvement. i (1) Provide a verified, unified set of drawings for operations, maintenance, and I&C ! activities. Perform field verification of the as-built wiring. Revise the AOG control ! panel drawings, especially the electrical independence and wiring areas. The original 1973 contractor drawings were difficult to read because of layout, j lettering, lack of detail, and reproduction quality. The drawings and the actual electrical installation did not incorporate VY engineering standards, for example, i redundancy and instrument fusing. The revision control of the drawings was poor. In the case of the piping and instrumentation diagrams (P&lD), there were two identical sets of drawings done by two different contractors.
i An example of the wiring problems was that some of the neutral wires were connected from different power supplies than the hot wires. This was viewed by i the licensee as a condition that could cause erratic operation of the instruments. This is further discussed in the Section on " Status of the Modification." The major weakness the licensee engineering identified was that, if a postulated i failure of a single link in the instrument ac bus occurred, it could cause extensive failure of the AOG instrumentation. This is further discussed in Section on " Status of the Modificat!on." The licensee stated that the functional:ty of the system was not hampered by these discrepancies. Maintenance could be performed, but not efficiently. Over the years, as minor changes were made, the confidence in the accuracy and completeness of the electrical drawings was questioned by operations and l maintenance personnel. (2) Change the level control and pumping system for the AOG condensate drain tank, TK-104-1. This is an interfacing system to the AOG that is used during normal cperations. The tank receives extracted moisture from the AOG process lines. The water is then pumped to the main condenser. The tank level provides a water barrier between the condenser vacuum and the AOG process lines. During start-ups from extended outages or in cold weather, a large volume of condensate is produced and AOG system startup times are increased. During normal plant operation, a low level of condensate is produced. There were instances when the tank was pumped dry, which the licensee found the root cause i to be regulating valve controller failure. ! The licensee's focus of this item was availability, rather than functionality, of the system. The modification was to replace the level control system with a type that integrated pump control and protection. (3) Upgrade the pressure rating for the steam jet air ejector (SJAE) inter-condenser. This system operating cressure upgrade would increase the design margin to absorb i pressure transients as.eced with the isolation of the AOG recombiner inlet valves. An installed rupture disc downstream of the inter-condenser currently l. l prevents AOG system damage due to inadvertent system isolation and/or hydrogen , L detonations. Over the past seveial years, this rupture disc has actuated due to L over-pressure conditions approximately eight times, necessitating a unit shutdown l_ to replace the disc. With ae increased inter-condenser pressure rating, the setpoint of the rupture disc could hewise be increased. The inspector determined that system availability, rather than system functionality, was the focus of this item. The inspector notes that the potential radiological release consequences of a rupture of the AOG rupture disc has been the subject of previous NRC staff review , and follow-up (reference inspection reports 92-01,92-15,93-25, and 94-27). As documented in these reports, the licensee modified the turbine building ventilation
system to exhaust to the main stack to ensure proper filtering and monitoring of gaseous radiological releases. Until this modification was completed in the Fall of 1993, the licensee took appropriate interim measures to monitor potential releases via the turbine building ventilation system pathway. The inspector also reviewed the applicable off-normal operating procedure (No. ON-3151, Off Gas Explosion / Rupture Disc Failure) and verified that appropriate procedural guidance was in place to ensure prompt actions are taken should this event occur. In addition to the above items, eight other areas addressed in the modification package similarly focused on system availability, maintainability and routine operations. Examples were the replacement of analog recorders with digital chartless type recorders, replacement of analog controllers with digital controllers, circuit changes to eliminate spurious AOG annunciator alarms, and re-configuration of the instrument air supply to valves to prevent unnecessary loss of recombiner heat exchangers due to loss of
. Instrument air.
2.3 Review of Modification Design The design engineering coordination with the operations, l&C, and maintenance functions to identify concerns, analyze the concerns, evaluate the alternatives, and document the design bases was performed well. This was indicated by the depth of the design analyses and the active engineering participation in site walkdowns. The consolidation and updating of the as-built drawings of the AOG control room panel 9 50 was performed well. The licensee performed a detailed point-to-point walkdown and verification of the 9-50 panel as-built drawings and found only one minor difference, which was corrected on the drawings. The licensee review at the site identifie/ .viring discrepancies with the planned modification drawings that could have potentially complicated the installation process. This indicated good performance of the overall drawing review process, but indicated weakness in the engineering drawing check procedures for the modification. The licensee checked the electrical wiring in the AOG building to confirm conformance with the original system installation drawings. Several discrepancies were found and corrected. None of the discrepancies affected proper functioning of the system, according to the licensee. 2.4 Status of the Modification The licensee management made a decision to not include the modification in the April 1995 refueling outage. The inspector reviewed two internal VY documents, "AOG Mods Design Change," December 22,1994, (VY Vice President, Engineering to VY Department Manager and Project Manager), and "AOG Design Changes," January 24,1995, (VY Vice President, Engineering to VY Department Manager) for reasons why the decision was made. The inspector found from the review, supplemented by personnel interviews, information to suggest that the main reason for the management decision was because all ) the documentation necessary for the modification package had slipped schedule milestones. __ _ _ -. _ -- _ _ - _ _ - . - _ - _ _ _ l
l l The modification design drawings were completed and put under drawing control, but the j modification was cancelled and not implemented as part of the 1995 refueling outage.
- The Inspector noted that the 9-50 panel wiring, where the neutral wires were from i different power supplies than the hot wires, was changed as a task during the 1995 l refueling outage.
l l The change of the postulated failure of a single link in the instrument AC electrical bus and the effect on the instrumentation was not implemented as a separate task during the 1995 l outage. The licensee has this task under review for future implementation. The inspector l noted that if this postulated failure of a single link were to occur and if all AOG l instrumentation would consequently fail,it would be covered by the Limiting Condition for l Operation (LCO)in the requirements for minimum number of channels operable of Technical Specification 'TS) Table 3.9.2 (2a,b,c), Gaseous Effluent Monitor!ng Instrumentation, which covers radiation monitoring, flow rate, and hydrogen monitors. This TS would permit continued plant operation and release of off-gas effluents via tNs pathway for up to seven days provided one atack radiation monitoring system was operable and off-gas temperatures and pressures were continuously monitored. 2.5 Review of Maintenance Engineering
- a. Instrument and Controls (l&C)
The 1&C engineering group started a system-analyzed maintenance (SAM) project for the AOG system in 1992. The inspector reviewed the document "VY l&C Preventative Maintenance, System Analyzed Mainterience Project (SAM)," Revision 2, dated February 6,1996, that described the purpose, objectives, and methodology of the project. The document described the method for classifying the instruments of a system according to designed function, safety class, importance to function, vendor recommendations, maintenance history, operating experience. An appropriate maintenance task with an associated interval is then assigned. ! The AOG system has 370 instruments,318 of which have been classified to date and are
~
I planned for final engineering review. Approximately 85% of the instruments were classified as functionally important. The inspector reviewed a sample of ten completed worksheets for the functionally important equipment and noted that the function was l clearly described, and the general requirements for the maintenance task was delineated. z b. Mechanical The maintenance engineering group had a project similar to the I&C SAM project' called reliability-based maintenance (RBM). The AOG mechanical components had all been classified and had been reviewed. The inspector selected a valve in the recombiner drain (HO-OG-587) and reviewed the basis'of classification, and noted that the functional importance basis and maintenance actions were appropriate for the valve service. 2.6 Review of Maintenance Status a-_________-_-____-______ _
I
' The inspector reviewed a listing of work orders on the AOG system from 1991 to the present. The listing showed that the operators were alert to the identification of off-normal conditions, such as motor bearings making noise, pump cycling, air fitting leaking, and valve packing leaking that could lead to degraded conditions. The inspector determined that the listing showed no indication that functional problems occurred that were not addressed by corrective actions.
The inspector determined that there were I&C instrument calibration procedures for the
. system instrumentation. Additionally, there were functional / calibration procedures for the l AOG hydrogen monitors, radiation monitors, trip system, moisture detectors. The work order status showed that instrument calibration and functional checks had been performed.
The inspector reviewed the document " Advanced Off-Gas Hydrogen Analyzer and Recombiner Catalyst," memo number VYl-2/96, dated January 5,1996 The document stated that the licensee performed an analysis of temperature indications across a recombiner from actual plant data covering the period from 1980 to 1995. The trend of temperature showed that the recombiner was operable, but indicated a loss of catalyst efficiency. The outlet temperature data was lower then expected, which could indicate higher output hydrogen concentration. The licensee performed a grab sample check of the output that showed that the hydrogen concentration was well within the normal plant operating band. Engineering recommended that the inside and outside surfaces of the thermocouple wells be cleaned and the thermocouple be checked as the corrective action for the low outlet temperatures. Engineering also found that the design specification for the catalyst life was for 18 months to 5 years, so they conservatively recommended replacement of the catalysts at the next refueling outage, even though the data showed that the recombiners are properly performing their function. This recommendation is being reviewed by licensee management. Further discussion of the catalyst life is found in Section 3.0. 2.7 System Walkdown The AOG building was very welllaid out and the inspector consifered redun ancy, physical separation, shielding, and maintainability. The material condition was generally very good. The AOG control room panel CRP 9-50 arrangement provided the necessary readouts, recorders, system mimics, and controls to operate and determine system status. The inspector verified that the following indications were within normal operating ranges: rediation level at the SJAE; the recombiner inlet / outlet temperatures; recombiner outlet flow; hydrogen analyzer percent of combustible limit; guard bed inlet radiation; first section adsorber outlet radiation; and system outlet flow. The panel had an extensive temperature monitoring panel for indication of system temperatures at heat exchangers, for example. Additionally, the panel had pressure indications for key points in the system. j 3.0 ' REVIEW OF UFSAR AND COMMITMENTS A recent discovery of a licensee operating their facility in a manner contrary to the i -___ __ _ \
L 1 l l Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the l UFSAR descriptions. While performing the inspections discussed in this report, the inspectors reviewed Section l 9.4, " Gaseous Radwaste System," of the UFSAR that related to the areas inspected. The inspectors verified that the UFSAR wording for the AOG system was consistent with the observed plant practices, procedures and/or parameters. The modifications of EDCR 94-402 were not checked versus the UFSAR because the EDCR was cancelled and planned to be implemented at a later date. As a result of their follow-up of the AOG system hydrogen monitor issues, the licensee noted that in their licensing change request to the NRC for the AOG system of June 16, 1972, Attachment A, Appendix A, " Description of Offgas Processing System Components," page A-3, that the recombiner catalyst lifetime was estimated to be equivalent to plant lifetime when operating in the steam /offgas environment. They also noted inconsistent lifetime catalyst information in the proprietary Appendix C, page C-3 of Attachment A to the licensing change request, which aescribed the catalyst expected life j as less than plant lifetime. The licensee's resolution of the inconsistent recombiner catalyst lifetime in the AOG license change request documents and any impact on catalyst maintenance is an unresolved item (URI 50-271/96-03-07). 4.0 MANAGEMENT OVERSIGHT l Project engineering management initiated an independent engineering review of the design implementation process. Inspection information suggested that plant and engineering management made the decision to not include the modification in the 1995 shutdown because all the documentation necessary for the modification package had slipped schedule milestones. The inspector inferred from these actions that management was actively involved in the oversight of the modification. The inspector found no indication that the cancellation of the modification was driven by cost considerations other than the inherent cost risk associated with implementing a
- modification with possible incomplete documentation.
5.0 CONCLUSION
S The licensee's resolution of the inconsistent recombiner catalyst lifetime in the AOG license change request documents and any impact on catalyst maintenance is an unresolved item (URI 50 271/96-03-07). The inspector found no engineering or maintenance indications in the last 5 years that t AOG system functionality was impaired in such a manner that led to degraded conditions ; that exceeded the Technical Specification requirements either for minimum channel availability, or AOG system instrumentation, or AOG system operabiGty. The inspector found no indication that the cancellation of the modification was driven by cost considerations other than the inherent cost risk associated with implementing a l l
p l r modification with possible incomplete documentation, such as installation and test instructions. The engineering coordination with the plant staff, the quality of the consolidated as-built panel 9-50 electrical drawings, and the delineation and resolution of design issues for the planned AOG modification were generally very good. , The 1&C and maintenance engineering initiatives regarding AOG system analyzed l&C maintenance and reliability-based mechanical maintenance for components were good.
- . 6.0 EXIT MEETING The findings of the inspection were presented and discussed with Mr. D. Reid, Vice i l President of Operations and members of the licensee's staff on March 8,1996, as listed in Section 7.0. The licensee acknowledged the findings presented.
l The inspector telephoned the licensee on April 10 and 11,1996, for additional information j i on the maintenance for the hydrogen recombiner catalyst. l The inspector received and reviewed proprietary material during the inspection and used the material for technical reference. No proprietary Information was knowingly included in l l the report. ! l . 7,0 LIST OF PEOPLE CONTACTED l l t Vermont Yankee Nuclear Power Corporation l J. Bolvin Manager, Technical Support l . E. Bowman Operations Engineer B. Buteau Manager, Engineering Reorganization Coordination i l
- D. Calsvn Supervisor, Quality Assurance L. Casey*
- Design Engineer, YNSD
*R.' Clark Executive Director, Quality Assurance )
- P. Corbett Manager, Project Engineering
'J. DeVincento Manager, Performance Engineering F. Helin Manager, Reactor Engineering l
'S, Jefferson Assistant to the Plant Manager !
G. Maret Superintendent, Operations I L 'D. McElwee*
- State Liaison Engineer L
'S. Miller *
- Manager, Design Engineering, YNSD l
'J. Pelletier Executive Director, YNS l
- D. Reid Vice President, Operations '
j R. Routhier Electrical Engineer t
'J. Thayer Vice President, Engineering J.Todd Maintenance enginee R. Wancryk ' Plant Manager M. Watson *
- Manager, l&C Vermont Department of Public Service
'W. Sherman State Nuclear Engineer l
l , i
. I i ! l U.S.N.R.C.
'J. Calvert Reactor Engineer, Electrical Engine'ering Branch
*W. Cook Senior Resident inspector
- Present at exit meeting on March 8,1996
** Contacted by telephone April 10 -11,1996
)
l l 1 i r. l 1 I l i l l I
Allegation Rscsipt Rzport Page 1 of
, (Use also for staff suspected wror.gdoing)
Date/ Time , Received:
- o 13 96 08764 %. Allegation No.
(leave blank) 3 Employee Receiving Allegation or suspectiny wrongdoing %) f-d F A' 6U h (first two initials and last name): *R u Ca t. leger Name:
- Home Address:
- O Home Phone
- City / State / Zip:
- Jlleger's Employer:
- Alleger's Position /
Title:
- N j Facility: 1 I Pk aw 0 h k Docket or Mtls. License No. : b)l Was alleger informed of NRC identity protection policy? Yes No O If a licensee employee or contractor, I did they raise the issue to their management? Yes No
- Does the alleger object to referral of issues to the licensee? Yes No l Was confidentiality requested? Yes No f
l Was confidentiality initially granted? Yes No L Individual Granting Confidentiality: Criteria for determining whether the issue is an allegation: l Is it a declaration, statement, or assertion of impropriety or inadequacy? / No l Is the impropriety or inadequacy associated with NRC regulated activities? / No ) l Is the validity of the issue unknown? / No f If No to any of the above questions, the issue is not an allegation and should be, handled by other appropriate methods (e.g. as a request for information or an OSHA referral). Allegation Summary or staff suspected wrongdoing: (Summarize each concern here - provide additional detail on reverse side of form, if necessary) bSCQ Sf'c% --W ev nr. Js% [ Umb & v " l la h J1*}$'*[ Y OcT & ) _ (1),/% Ce+ LZD w%l ci;w2 3r & cf 40cyc [ t k. 41 cn dj r w iti , d clu,a+&J " n c d, Lt an-a rs u"A
- $ SPPc/N /W In C& ph& [sypb.J uA i'h hv, n na. ) i: /Str -
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W ad- M .74 A n~s Co@ em$ ' Number of Concerns:
. Type of Regulated Activity (a) _ Reactor (d) _,, Safeguards (b) Vendor (e) Other:
(c) _ Materials (Specify) Functional Area (s): (a) Operations (e) Emergency Preparedness (b) Construction (f) Onsite Health and Safety \ (c) Safeguards (g) Offsite Health and Safety (d) Transportation (h) Other:
- Do D91. complete these sections for issues of staff suspected wrongdoing.
R Q' g hcm&h $L Ch I] La N & $ r. newrect-_- __ -__
Page 1 of i Detailed Description of Allegation or staff suspected wrongdoing: (Do not state the alleger's name in this section - simply refer to the individual as the alleger) ! The attached redacted version of a letter to the licensee from the lawyers for a licensee project engineer (LPE) (who is to interviewed by OI and DRS in another case file) was received by facsimile in Region I on 5/21/96. The letter indicated generally that uncorrected safety violations exist at the licensee's facility dealing with the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems. The plant manager, VP operations and VP Engineering discussed the matter in two telecons on the same date with DRP personnel. They all indicated that the unredacted portion deals with a , retirement settlement proposal for the LPE and that information is not nuclear I safety related. The licensee indicated that the entire letter is available for NRC staff review. 1 The major issues of the letter are as follows:
- 1. In late October 1995, another engineer / manager allowed circuits for the HPCI/RCIC system to be disconnected resulting in degraded mode due to errors made.
- 2. The errors were incorrect procedure (apparent white tag process) implemented in 1988 to keep circuits deenergized contrary to GE recommendations - safety violatic,n (potential safety hazard but not immediate threat to nuclear safety).
- 3. No action has been taken by licensee to date; the other engineer has not '
acknowledged receipt of LPE report. The telecons noted above established the following additional information:
- 1. The plant manager confirmed that as far back as 1988, white tag (s)
(equivalent to standard RED - DANGER DO NOT OPERATE) were installed on circuit breakers for heaters used with for certain HPCI/RCIC pump motors for environmental considerations (to maintain low humidity to avoid excessive moisture build up especially when components have been in a long term layup). The circuits were deenergized because of abnormal deterioration on related electrical insulation.
- 2. The affected components were heaters in the motors for the following pumps: RCIC Vacuum Tank Condensate Pump, RCIC Vacuum Tank Vacuum Pump, HPCI Gland Seal Condensate Hotwell Pump, HPCI Gland Seal Exhauster Pump, WPCI Auxiliary 011 Pump.
- 3. Last Year (1995) .a licensee white tag audit revealed a potential problem with configuration control with the long term use of the white tags to keep these heaters deenergized. The LPE was assigned the action to resolve a Plant Operation Review Committee (PORC) followup item.
P 3 ' "'} ---N
' 4. The LPE and the another engineer had differing views on the proposed GE resolution to reenergized the circuits. The LPE wanted to implement the GE recommendation and the other engineer did not.apparently because of the
. original problem on the insulation.
Currently:
- 1. In light of the recent letter from the lawyers for the LPE, the licensee l
has assigned a third engineer to do an independent review, document'an operability determination, and report back to the PORC by 5/22 in the p.m. 2.. The LPE is still employed by the licensee and is working offsite but still appears to have site access. 1 l l' l l
. . . I L._________--__-_________-_____-____-.______._-___
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{W q. g ses- L es J *" * ** r war g gO May 16, 199G i a Peter Robb, Esquire P.O. newRachlin Downs, s Martin p Brattleboro, Vermont 05302 Res Wassey - Verssont Yankee Dear Peter f y , v1 S 3:T4& - A O-7 bdc 5 -,
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In addition I am informed by Mr formed a review ,on the high pressure c. oolant Massey in that he hac per-reactor core isolation coolant systema circuits.jection and According the to the investigation and report performed by Mr. Massey, it appears that arrors were made with respect to the systems and that the systems were operating in a " degraded mode *, Apparently one of g the circuits Mr. cornett.had been disconnected with the express permission of -
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,. fB 86110N 1312 VT Y W it iUCLEPR POWEH I M 10. BU225(4bWI r. ue 05/20/S6 MON 09*29 FAI 802 258 2288 DRY /2 S 002 ?
- f. 5 Peter Robb, F. squire May 2 6, 1996 Wb(
Page 2 Upon investigation by Mr. Masser per a recent assignment given to him, the disconnection of the circuit appears to have been incorrect procedure. The disconnection had taken place as long ago as 1988. Mr. Masaoy contacted General Electric Nuclear Ener" che heater circuit breaker and was informed brY to inquirnTasCENE to the correctness that the of havi.ng disco cirevit breakers should be reconnected and turned on in order to prev +nt possible moisture damage to the insulation of the wiree>, which in turn could cause problemo with operation of the motors that are involved with the systems. I am not sufficiently famil-Lac with technical aspects of the systems to state wa.th any certninty that the above sum ==cy is accurato. However, I do beliswa that it captures the essence of the findings of Mr, Masseiy, which appear to have been borne out as being accurate. Mr. Massey prepared a written report on this matter which was turned over to Mr. corbett in late October of.1995, To his knowledge, Mr. Corbett has not acknowledged receipt of the re-Port;indings the f of Mr. Mussey.nor has any action been taken by vermont Yankee concerning According to Mr. Massey, the condition of the disconnected ciret.it breaker represents a safety violation and is a potential hazard, although not an immndiate threat to the safety of the nuclear plant. W
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f1AY-21-19SS 08:36 VERM PC P.01 fG/ r 'b l
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VERMONT YANKEE NUCLEAR POWER STATION RESIDENT INSPECTOR OFFICE VERMONT YANKEE FAX NO.: 802 257-7791 OFFICE NO.: 802 257-4319/4310
, ON CmEF, DRP,3A TO:
FROM: VElb10NT YANKEE [, fg
// Bill Cook, Sr. Reylde pector
,,..~
// PaulI , Resident Inspector Curtis, Resident Office Assistant DATE SENT: NA! T mIE: Y 'd b NUMBER OF PAGES INCLUDING THIS FORM:
6___..._. . _ . _ _ _ . ._ _ . _ . _ _ _ _ _ ___
ALLEGATION DISPOSITION ..EC%J ev. Allsgation No.: 11/01/95 RI- -A- W Branch Chief (AOC) : I-
's Site: ,,,, V 4 % o d M Panel Date: 5 ell % 00A Acknowledged: Yes confidentiality crantedi No Qves !
No . ALLEGATION PANEL DECISIO * (Pre ous Allegation Panels on issue: Yes No ) Attendees: Chair - - Branch Chief (AOC) - SAC - OI Rep. - RI Counsel - Others - DISPOSITION ACTIONS: concurrences), responsible (State actio person,ns required for closureECD and expected closure (including special documentation)
- 1) Acknowledgement letter (if known alleaer) d -
b) U _9 O Dd r Responsible Person: ECD Closure Documentation: Co leted:
- 2) 08O - OM d iX QUJJ (A/ iN -) ClM (7 f)l/b c., "a ) af a#
L A v- C W r , N ~ % heJ tw eA b6 U L t2 % a 6- %L U h.Ja 3-h4 en A Responsible Person: b O/b ECD: hY $b Closure Documentation: C
- 3) ( o n a $ dA 0 0 Uh n b & &p(npleted:
Chdl b4 h Y'IOE CL 6 k P5LvY w S M 4A v t[) v )E'I i ~J Ik Ud) C Vhn ) k / 6T M ex v n Ss Ul ' 0 Responsible Persons b L, N4 )bb ECD: bf#7 / 82) # . Closure o7umentatyn: Cqmpleted:
- 4) DS 1 ! 84 1 ) OR9 N v'c ()c/ M ) IW) ed ff., b 1 c-rh iSS'W. b 5 o (A cN d N8A R cd d vu sl i c } eb n CA_r"h'e Responsible Persons (M vd R3 f 0N I ECD:
Closure Documentation: Completed:
- 5) 7- I w P I o ~ k d W T* $r V1N/ NY v & <GA 'dVn o r <<m% A ; 2 u '
O Responsible Person: 6 ECD: Closure Documentation: Completed: Safety Significance Assessment: 4 NOTES: Factors to Consider Prior to Peferral to a ideens(' If an allegation is to be ref erred to a licenses, following factors are to be consideredi . m.imm a.tm, nesagenuaarn=nemun.:e> - "- >" @'
- b. The possibility that the release of information could bring bann to the alleger.
- c.
g e g t g g eg g o g y d g g g icensee, applicant, or vendor management or those parties who would d. g g the alleger has voiced objectis a over the release of the allegation to the licensee, applicant. or
- e. gegrg gtg og gn}n{oggreceived f rom another governmental agency, which does not approve
- f. Whether the alleger has already taken this concern to the licensee with unsatisf actory results.
b* En t @ a Es*I SIkga$fe crdNka$1steknNtI 8 above$n$ncate that i $ questIonab$e whN ber a referrali $s e ap$ks Y a $[ckna osNaNs propriate. e$ h!ethe Distribution Penal Attandaac. Daninnal rmneal AT Dese ~ skia %--- #-rr_n' - . . - -
Qh;. ions 'for Resolution; e - Lic n000.R;fcrr:1 (Div. Dir C ' ' Referral)../ oncurrence Document NRCRe@ ired (First Review Consider of Response Factors
- Resp. - AOC)Prior to R forral to Another Agency (OSHA, etc. - Resp. - SAC)
R3ferral to an Agreement State (MD, ME, NH, NY, RI - Resp. - SAC) R forral to Another NRC' Office (OIG, NRR, Other Regions - Resp. - SAC) R: quest for Additional Info.(Frcm alleger, licensec, others - Resp. - AOC)
' Closeout Letter / Memo (If no further action planned - Resp. - AOC)
Inspection (Resident / Specialist routine or reactive) IF H&ID INVOLVED:
- 1) \
process and the need to fi rmed of th DOL hastheindividualbeeninf$eacompla$nt Yes I (has DOL information package been provided?)within 180 days No l i
- 2) has the individual filed a complaint with DOL Yes No i 3) if the complainant filed directly with DOL have they been Yes No contacted to obtain their technical concer,ns (Resp. - SAC)
- 4) is DOL a chilling effect letter warranted: Yes No conc nding iationin/w avor of al eger)icensee pr or to DOL decision)
ADDITIONAL NOTES: ' 4 i l m l i
[ U. 93 ' IC l:
.[
p, i8-
- NUCLEAR REGULATORY COMMISSION REGloN l '
s, a 475 ALLENoALE RoAo
,g" . KING oF PMUsSIA, PENNSYLVANIA teso61415 August 8, 1996
. Mr. Donald Reid .
. Vice President, Operations
- Vermont Yankee Nuclear. Power Corporation RD 5, Box-169 Ferry Road.
Brattleboro, Vermont' 05301
SUBJECT:
NRC INTEGRATED INSPECTION REPORT 50-271/96-06 i D' ear Mr. Reid:'
- 1. .
On July 13,1996, the NRC completed an inspection at your Vermont Yankee reactor i facility.:The enclosed report presents the results of that inspection.
. During'the 9-week period covered by this inspection period,' your conduct of activities at the Vermont-Yankee facility was generally characterized by safety-conscious operations, sound engineering'and maintenance practices, and careful radiological work controls. Your.
staff continues to demonstrate a questioning attitude in identifying program discrepancies
- .and/or old desig6 problems (for example, the Appendix J valve testing oversights and the l'
- emergency. diesel generator room differential protection inadequacies discovered during this inspection period). We encourage continued emphasis by the VY staff in these types of endeavors.
1
- 1. . .
Within the area of plant support, we found that you continued to implement overall { E effective Radiological Environmental Monitoring and Meteorological Monitoring Programs. E . No reply to this report is necessary and your cooperation with us is appreciated. Sincerely, 1 Richard J. Conte, chief n Reactor Projects Branch 5 Division of Reactor Projects
Enclosure:
NRC Inspection Report 50-271/96-06 g j D ' Docket No.- 50 271 [ ll
> i-
- - _ - _ _ _ _ _ _%(( 6/76~'_8[
Mr. Donald A. Reid 2
- cc w/enci:ce w/ encl:
R. Wanczyk, Plant Manager ! J. Thayer, Vice President, Vermont Yankee Nuclear Power Corp. J. Duffy, Ucensing Engineer, Vermont Yankee Nuclear Power Corporation ! J. Gilroy, Director, Vermont Public Interest Research Group, Inc. f D. Tefft, Administrator, Bureau of Radiological Health, State of New Hampshire ! Chief, Safety Unit, Office of the Attorney General, Commonwezith of l
' Massachusetts R. Gad, Esquire
! G. Bisbee, Esquire T. Rapone, Massachusetts Executive Office of Public Safety State of New Hampshire, SLO Designee R. Sedano, State of Vermorit, SLO Designee Commonwealth of Massachusetts, SLO Designee
)
l l l f k--------------
l l~ U.S. NUCLEAR REGULATORY COMMISSION REGION I l Docket No. 50-271 Licensee No. DPR-28 Report No. 96-06 Licensee: Vermont Yankee Nuclear Power Corporation Facility: Vermont Yankee Nuclear Power Station Location: Vernon, Vermont Dates: May 12 - July 13,1996 Inspectors: William A. Cook, Senior Resident inspector Harold Eichenholz, Project Engineer, DRP Laurie A. Peluso, Radiation Physicist, DRS Michael K. Webb, NRR Russell Gibbs, Reactor Engineer, DRP Approved by: Richard J. Conte, Chief, Projects Branch 5 Division of Reactor Projects l l l 4 j l I i i
.-qbC2!'iv484 .29pf- l
E)'ECUTIVE
SUMMARY
Vermont Yankee Nuclear Power Station NRC Inspection Report 50-271/96-06 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 9-week period of resident inspection; in addition, it includes the results of announced inspections by a regional specialist and a regional projects inspector. l Ooerations The inadvertent individual rod scram of rod 26-07 vice rod 26-11 on May 20,1996 was i~ the result of personnel error lack of self checking. The VY staff critique was prompt and j- thorough and the corrective actions implemented to pievent recurrence were appropriate. l
-)
Based upon the observation of a number of Plant Operations Review Committee (PORC) I meetings, the inspectors concluded that the PORC implemented its charter appropriately l and demonstrated a well defined safety focus during their review of the agenda topics. I Maintenance VY staff performance during a number of maintenance and surveillance activities was observed to have been good. l L The VY staff properly ensured the operability of the HPCI' system and its components after ; the completion of the LCO maintenance outage and prior to formally returning the system to service. Very good supervisory oversight was observed. Enoineerino
- The inspector concluded that the VY staff's handling of the HPCl/RCIC motor heater issue differing professional opinion was appropriate and timely. The prompt dispositioning of the .
- . ' potential motor heater operability impact was consistent with Generic Letter 91-18 l guidance and the technical basis for leaving the heater circuits de-energized was well founded. The recommendations for increased frequency of motor insulation meggaring and periodic motor inspections were appropriate preventive maintenance measures.
i The VY staff's discovery of Appendix J program discrepancies this inspection period reflect ; positively on the thoroughness of their review of the program. Pending NRC staff review ! ' of the exemption request for the specific valves of interest, this issue is unresolved (URI 96-06-01). 1 The training performed for the refueling bridge design changes was appropriately performed prior to declaring the modifications operable and was thorough. The installation of the hoist loaded interlock and the human factors changes implemented were favorable for fuel j b handling operations. H l
I l L l The implementation of the Fuel Handling Training Program was effective. The requirement l I for classroom and hands-on training, as part of requalification prior to refuel operations, l appropriately refreshed operator awareness of this infrequently performed evolution. ) f The identification of the emergency diesel generator room non-conforming design issue 1 demonstrated a good questioning attitude on the part of the individual e'igineer wrio !
~ identified the problem and an appropriate review of industry operating experience by the j VY staff. The engineering and operations staffs' handling of the (temporary) resolution of this design discrepancy and the PORC review of the operability assessment were prompt l and thorough. l l
The Quality Assurance staff audit of the VY Environmental Qualification pro.pr.) was thorough and well documented. Station management was receptive and responsive to the ' audit findings and assessments. ) I Plant Succort ) VY continued to implement an overall effective Radiological Environmental Monitoring Program (REMF) and Meteorological Monitoring Program (MMP) with respect to the management controls, quality assurance audits, quality assurance for analytical measurements, and implementation of the Offsite Dose Calculation Manual (ODCM). The licensee's upgrade to instrumentation for meteorological monitoring was noteworthy. This failure to adhere to the required personnel frisking requirements for existing the radiologically control area to the AOG building roof constitutes a violation of minor significance and is being treated as a non-cited violation, consistent with Section IV of the NRC Enforcement Policy. During the annual Emergency Preparedness Exercise, the licensee's overall performance was very good. The emergency facilities were generally staffed and activated in a prompt I manner. Good management direction and control was observed at all of the facilities. The classifications of the simulated events were timely and accurate, and off site notifications were property completed within 10 minutes. Communiccions among the teams was very good, resulting in a smooth translation of information, and good turnover from the simulator control room and Technical Support Center to the Emergency Off-site Facility, i I ... ' til
i l 1 TABLE OF CONTENTS EX EC UTIVE S U M MARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii TA B LE O F C O NTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
- 1. O pe r a tio n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 Retraction of 10 CFR 50.72 Notifications ................. 1 01.2 Nuclear Alert System (NAS) Temporarily Out-of. Service . . . . . . . 2 01.3 10 CFR 50.72 Notification, June 5,1996 . . . . . . . . . . . . . . . . . 2 01.4 10 CFR 50.72 Notification, June 24,1996 ................ 2 i 04 Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 3 l 04.1 Inadvertent Rod Scram . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 )
L 07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 07.1 Licensee Self Assessment Activities . . . . . . . . . . . . . . . . . . . . . 4
- 11. M a i nt e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A M1 Conduct of Maintenance .................................. 4 "
M1.1 General Comments ................................. 4 M1.2 LCO Maintenance Plan for the HPCI System . . . . . . . . . . . . . . . . 5 111. Engin e e rin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 E2 Engineering Support of Facilities and Equioment . . . . . . . . . . . . . . . . . . 6
.................................................... 6 E3 Engineering Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . 7 j E3.1 Residual Heat Removal (RHR) Appendix J issues ............ 7 E3.2 Refuel Bridge Modifications ........................... 9 ;
E7 Quality Assurance in Engineering ' Activities . . . . . . . . . . . . . . . . . . . . 10 l E7.1 Effect of Design Basis Tornado Pressure Load on Diesel Room Enclosures ...................................... 10 E7.2 Review of QA Audit No. VY 96-25 . . . . . . . . . . . . . . . . . . . . . 11 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 ES.1 Review of Licensee Event Reports (LERs) . . . . . . . . . . . . . . . . . 12 ; IV. Plant Support .............................................. 13 R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 13 R1.1 Implementation of the Radiological Environmental Monitoring Program........................................ 13 R1.2 Meteorological Monitoring Program (MMP) . . . . . . . . . . . . . . . . 14 R6 RP&C Organization and Administration ....................... 15 l- R6.1 Management Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 l R7 Quality Assurance in RP&C Activities ........................ 15 R7.1. Quality Assurance Audit Reports . . . . . . . . . . . . . . . . . . . . . . . 15 R7.2. Quality Assurance of Analytical Measurements ............ 16 R8 Miscellaneous RP&C issues -Inappropriate Personnel Frisking Event .. 17 P4 Staff Knowledge and Performance in EP . . . . . . . . . . . . . . . . . . . . . . 18 iv i i
j l P8 MiscillanIous EP IssuIs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 P.8.1 (Closed) IFl 9 5-20 01 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 P.8.2 (Closed) IFl 9 5 20-0 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 F5 Fire Protection and Staff Training and Qualification . . . . . . . . . . . . . . . 20 21 V. Ma nagement Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 X1 Updated Final Safety Analysis Report (UFSAR) Review . . . . . . . . . . . . 21 X2 22 INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 i 23 4 PARTIAL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 LI ST O F ACR O NYMS USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . o 6 e
- . V L__________________
DETAILS Summarv of Plant Status i During this inspection period, the Vermont Yankee (VY) staff. operated the unit at full power, with a few exceptions involving planned down powers to conduct maintenance and I surveillance testing. On June 20 the licensee announced the decision to pruceed with a
' shroud repair to address the industry-wide shroud weld cracking problem. The repairs are planned to be made during the scheduled Fall 1996 refueling outage. The refueling outage start date was postponed from August 24 to September 8 to support projected New England electrical power demands. On May 23 the !icensee conducted their annual NRC observed Emergency Preparedness Exercise (reference section P).
'A specialist inspector conducted a review of the environmemal monitoring program during
. this inspection period (reference section R).
Region based inspectors and headquarters' inspectors provided support to the resident office during this inspection period,' until the permanent assignment of a new resident inspector to the office on June 30.
- l. Operations
~ 01 - Conduct of Operations' 01.1' Retraction of 10 CFR 50.72 Notifications L ' Insoection Scoce-
- On May 29, the VY staff notified the Headquarters Duty Officer of two 10 CFR 50.72 notifications (Event Nos. 30511' and 30523, dated May 21 and May 22,1996,' !
respectively) that were being retracted.' After further review the VY staff concluded that I neither event was reportable. The inspector reviewed these events and the licensee's 10 CFR 50. 72 notifications. j i Observations and Findinas Event No. 30511 involved the potential inoperability of the high pressure coolant injection - l (HPCI) system. At approximately 12:15 p.m. on May 21, a number of control room annunciators were received and the HPCI system suction valves swapped from the CST to ; the torus. Initial assessment by the control room operators resulted in the HPCI system , being ' declared inoperable and I&C technicians were called to investigate the as-found i condition. The I&C technicians could not determine the precise cause of the annunciator ! L ' alarms and valve actuation, but identified a common power supply (GEMAC 5000 series)_ l whi:h was the most likely cause of the problem and subsequently replaced it. Further VY ; staff review concluded that the HPCI system was unaffected by the power supply fault L' L and that the HPCI system remained operable and could have performed its intended design i function. ' Therefore, the May 21 10 CFR 50.72 notification was retracted. . Event No. 30523 involved the apparent failure to open of a primary containment isolation valve in the containment atmosphere sampling' system. Valve V6-109-76A was being I c __ _ __ _ - _ - _ - _ _ - _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - _ - _ - _ _ _ _ _ - _-___ _ -_ _ __-_-__
2 opened following planned maintenance which replaced the reactor building ventilation radiation monitor. By valve position indication, valve V6-109-76A appeared to not open. Control room operators promptly closed V6-109-76B to isolate the effected containment penetration. Subsequent investigation identified that isolation valve V6-109-76A was functional and that the problem was valve position indication only. The valve position open limit switch was adjusted by the I&C staff and the valve stroked-tested satisfactorily. Consequently, the containment atmosphere sampling system remained operational and this condition was not reportable and the May 22 notification was retracted. L Conclusions The inspector concluded that the VY staff had taken appropriate action in accordance with Technical Specifications upon discovery and initial assessment of these problems. The 10 CFR 50.72 notifications were made in accordance with regulatory requirements and reflected a timely and conservative reporting threshold. Likewise, the retraction of bnth notifications was appropriate and well founded. 01.2 Nuclear Alert System (NAS) Temporarily Out-of-Service On June 18, at 3:10 p.m. the VY control room operators notified the NRC staff (Event No. 30640)in accordance with 10 CFR 50.72, that a portion of the Nuclear Alert System (dedicated telephone lines between the Emergency Off-site Facility (EOF) and state Emergency Operations Centers (EOCs) was found out-of-service in the EOF. Alternate means of communication were available to compensate for this deficiency. The NAS telephone service was restored between the EOF and EOCs by 4:35 p.m. on June 19. The cause of the equipment problem was determined to be line noise introduced by local telephone company equipment. Actions by the local telephone company corrected the conditic,n. The inspector concluded that the VY staff appropriately addressed this communications problem and initiated timely action to resolve it. 01.3 10 CFR 50.72 Notification, June 5,1996 At 2:47 p.m. on June 5, the VY staff made a 10 CFR 50.72 notification (Event No. 30584) identifying to the NRC staff that both emergency diesel generator rooms may not be able to withstand the differential pressures associated with the design basis tornado event. Inspector followup of this event is documented in section E7.1 of this report. 01.4 10 CFR 50.72 Notification, June 24,1996 At 2:00 p.m. on June 24, the VY staff made a 10 CFR 50.72 notification (Event No. 30675) informing the NRC staff that an event requiring Vermont state agency notification had occurred. The event involved the inadvertent release of 400 500 gallons of sodium hypochlorite to the Connecticut River via a break in the storage tank's outlet line. The cause of the oulet line (PVC piping) break was due to the impact caused by a portable eye-wash station that was inadvertently knocked over by a technician in the area. The inspector verified no adverse impact or challenge to safety related systems resulted from this event. The control room staff's notification of the NRC staff was timely and in accordance with 10 CF3 50.72 requirements. l L___.___________._
3 04 Operator Knowledge and Performance 04.1 Inadvertent Rod Scram a inspection Scoce On May 30 during the post-maintenance testing of control rods fol!owing scram solenoid pilot valve (SSPN) diaphragm replacement, a control room operator inadvertently scrammed control rod 26-07 instead of 26-11. The inspector conducted a followup of this event to determine the cause of the error and to assess the corrective actions taken by the VY staff. L Observations and Findinos A critique of the event performed immediately following the inadvertent rod scram involving the control operators, shift supervisor, and Operations Department management identified that the control room operator responsible for scramming the control rods from the test panel failed to adequately self-check his actions. The test procedure had been followed conectly and successfully for a number of rods prior to this event. However, following installation of the scram time recorder test plug into the rod 26-11 test jack, the operator reached under the jack vice above the jack and actuated the individual rod scram switch for rod 26-07. Control room operators immediately identified the error and properly entered Operations Procedure (OP)-2111, inadvertent Control Rod Mispositioning, section Q, and implemented the proper notification and recovery measures. Rod 26 07 was returned to position 48 (fully withdrawn) and rod 26-11 was subsequently successfully scram time-tested. As corrective action, all personnel involved with SSPV diaphragm replacements and individual scram-time-testing were briefed prior to proceeding with additional rods. The briefings stressed the importance of attention to detail and the use of STAR (Stop, Think, Act, and Review) techniques. In addition, subsequent individual rod scram tests utilized a second checker at the test panel to minimize the chance of another error. No further personnel errors occurred during this work activity. L Conclusion The inadvertent individual rod scram of rod 26-07 vice rod 26-11 on May 20,1996 was the result of personnel error -lack of self checking. The VY staff critique was prompt and thorough and the corrective actions implemented to prevent recurrence were appropriate.
4 07 Quality Assurance in Operations , 07.1 Licensee Self-Assessment Activities A Insoection Scoce On June 12, the inspector observed routine Plant Operations Review Committee (PORC) meeting No. 96-050. Additionally, on June 13 and 14, the inspector observed PORC meetings No. 96-051 and 96-052. At PORC meeting No. 96-050, the inspector observed the PORC review of Basis for Maintaining Operation (BMO) No. 96-08, an Appendix R exemption request regarding emergency lighting, and Event Report (ER) No. 96-0030. At PORC meetings No. 96-051 and 96-052, the inspector observed PORC review of two Temporary Modifications (TMs) and a proposed Technical Specification (TS) request regarding core shutdown margin. L Observations and Findinos The inspector observed the PORC members' review and discussion of BMO 96-08 to be extensive and noted that the participants appeared thoroughly prepared to discuss the BMO. The inspector also noted that the members looked beyond the immediate situation and asked questions regarding long-term or permanent resolution of the circumstances that had created the need for the BMO. A further discussion of BMO 96-08 is included in Section E7.1. The observed discussion of the core shutdown margin TS change request demonstrated an ability by the PORC members to ask probing questions. The members' concerns resulted in a decision that a more in-depth review of the shutdown margin determination should be conducted to provide a greater level of detailin the proposed TS change request. In addition, the discussions contributed to more consistent terminology in the proposed change request. L Conclusions The inspector concluded that the PORC implemented its charter appropriately and demonstrated a well defined safety focus during their review of the agenda topics discussed above. II. Maintenance M1 . Conduct of Maintenance
-M1.1 General Comments 1 &- Insoection Scoce ]
l Using inspection procedures 62703 and 61726 as guidance, the inspectors observed all or portions of the following work activities: i [ L l l , I w - - - - - - - - - - -
l ( 5 l l e op 4323, Main Steam Line High Flow Functional / Calibration
.e op 4126, Monthly Diesel Test j e OP 5378, SCRAM Solenoid Pilot Valve (SSPV) Online Maintenance lL e OP 4424, Control Rod SCRAM Testing and Data Reduction
[ h Observations and Findinos The inspectors found the licensee's performance during these maintenance and surveillance activities was professional and thorough. Prior to performing wcrk, the licensee conducted good pre-job briefings appropriate for the maintenance or testing activity. The briefings ranged from informal discussions among the assigned technicians to ensure that each knew and understood the job to minimize errors during routine surveillance, such as the Main Steam Line High Flow instruments and the monthly Diesel Test, to more formal detailed briefings to coordinate significant maintenance, such as the SCRAM SSPVs and
. the LCO maintenance outage for the HPCI System. The licensee utilized approved procedures which included guidance for quality control such as independent verification at
; specific hold points to ensure critical work steps were performed properly. The level of supervision was appropriate for the jobs worked and the licensee used appropriate measuring and test equipment (M/TE) to perform functional tests and calibrations. The
. licensee ensured the operability of plant systems and components after the completion of the above activities via appropriate surveillance testing.
i L Conclusion ! VY staff performance observed for the above mentioned work activities was good. 1 M1.2 LCO Maintenance Plan for the HPCI System i L' 'Insoection Scoce i Using the guidance of inspection procedure 62703, the inspectors observed portions of the , following work activities involving the LCO Maintenance Plan for the High Pressure Cooling Injection (HPCI) system: e Installation and Test Procedure for EDCR 95-407, i ECN-3, "HPCI Pump Suction from Suppression Chamber Valve V23-57"
- OP 5210, Rev 6, MCC Inspections-e' OP 5219, Rev 8, Diagnostic Testing of Motor Operated Valves l'
e OP 5220, Rev 19, Limitorque Operator j
. OP 2120, Rev 25, High Pressure Coolant injection System !
1
6 E Observations and Findinos The inspector reviewed the inspection plan and found it to be comprehensive. it included: work activities to be performed; a Maintenance and I&C schedule; support requirements; a contingency plan; a compensatory action plan; an approach for alternate testing, if required; and justification for performing the maintenance on the HPCI system. Several plant departments and contractors were used to complete the plan within the 7-day LCO work window. The licensee demonstrated very good communication between the groups involved and completed the plan within the targeted time period. Work observed was thoroughly and comprehensively completed. The licensee adhered closely to the plan and to the procedures used to perform the specific maintenance and other associated work. L- : Conclusions 1 Tht licensee properly ensured the operability of the HPCI system and its components after the completion of the LCO maintenance outage and prior to formally retuming the system to service. Very good supervisory oversight was observed. 111. Engineering E2 Engineering Support of Facilities and Equipment E2.1 Review of Event Report No. 96 0325 i a. Inspection Scope On May 21, during the station engineering staff's review of a Plant Operation Review Committee (PORC) follow item (PFI 95-038-01), an employee raised a differing professional opinion in the handling of a potential safety system operability issue. The issue involved a VY engineering decision made in July 1988 to de-energize lighting panel No.1M, circuit
#54 which feeds the reactor core isolation cooling (RCIC) motor heaters for the gland seal-condenser pump and the gland seal vacuum pump. In addition, power panel No. 6A, l circuit #7, which feeds the high pressure coolant injection (HPCI) motor heaters for the i auxiliary oil pump P85-1 A, gland seal condenser blower, and gland seal condenser l condensate pump was de-energized. The reason for de-energizing these circuits was to )
L protect the motor wiring insulation from excessive heat-related degradation. Motor I winding and power lead insulation deterioration was noted during motor inspections . conducted during the 1988 outage. Maintenance Requests (MRs) and associated safety evaluations were performed in 1988 to properly control the de-energization of these , circuits. l j: The differing professional opinion raised a concern that the de-energization of the motor heaters potentially adversely impacted the operability of the HPCI and RCIC systems l i because of moisture intrusion into the commutator and. brush rigging. As this differing view conflicted with an engineering determination made in 1988, the engineering staff ; initiated an Event Report (No. 96-0325) to ensure a prompt re-examination of system l operability and the timely resolution of this issue. The inspector reviewed the licensee's corrective' action plan and technical resolution of this issue. l l t i_'_______-__-__-_a_:_-_____________-___--_-______._
7 The inspector reviewed a variety of documents associated with Event Report 96-0325 including: PFI 95-038-01 response dated 10/25/95; Assessment of motor heater de-energization on HPCI and RCIC operability, revision 1, dated 5/22/96; and Performance Engineering and Plant Support Department Memorandum, dated 7/19/96 and associated references. The inspector also examined the action plan prepared to address this issue anc the timekne developed from the documentation review conducted by the VY staff.
); Observations and Andinos As documented in the 7/19/96 memorandum and supported by a number of engineering evaluations and assessments, the motor heaters for the affected HPCI and RCIC motors are not required to be energized to support their associated systems' safety function or to ensure protection against log-term motor moisture intrusion. The basis for the licensee's conclusion was that the normal ambient temperatures in the RCIC and HPCI pump rooms result in relative humidity ranges of approximately 20 to 60 percent, dependent upon the outside relative humidity. Thus, the motors are not routinely subjected to a high humidity environment. Secondly, based upon a broad industry review, continuous motor heater energization has shown to have a negative impact on motors' 40-year qualified life. In addition, review of the original General Electric / Terry Turbine environmental qualification (EO) testing of the HPCI and RCIC systems determined that both systems functioned properly under 100 percent humidity conditions (established prior to system initiation).
Lastly, a review of the postulated EQ events identified that neither HPCI nor RCIC systems are credited with availability when the steam environment is in the vicinity of these systems. Notwithstanding, recommendations developed from the VY engineering staff review included an increase in the frequency of motor insulation meggaring from every six years to every three years, and an increased inspection frequency of the motor brushes and brush rigging. L Conclusions The inspector concluded that the VY staff's handling of the HPCl/RCIC motor heater issue differing professional opinion was appropriate and timely. The prompt dispositioning of the potential motor heater operability impact was consistent with Generic Letter 91-18 guidance and the technical basis for leaving the heater circuits de-energized was well founded. The recommendations for increased frequency of motor insulation maggaring and periodic motor inspections were appropriate preventive maintenance measures. E3 Engineering Procedures and Documentation E3.1 Residual Heat Removal (RHR) Appendix J lssues 32 Insoection Scoce Event Report (ER) 96-0420 was prepared by the licensee documenting issues regarding the Appendix J Program. The issues were self-identified during a program review of the technical basis for exempting RHR system valves from Appendix J Type C testing requirements. The significance of the issues is that the technical basis for the exemptions is not fully supported by system operating procedures (EOPs) and the various system
8 configurations given the required range of single failures. , I The valves addressed in the report include the LPCI injection valves, drywell/ torus spray isolatio'n valves, and the shutdown cooling isolation valves. The exemption basis for the LPCIinjection and spray valves assumes that the RHR system will continually operate at
>Pa (44 psig) after a design basis LOCA for 30 days to provide the required water seal, but plant EOPs do not specify such operation.
A licensee self-imposed exemption of type C testing had been implemented at VY for the shutdown cooling valves without formal exemption request approval by the NRC staff. The inspector notes that based upon the licensee's records there was previous correspondence between NRC and VY on this issue, but it was never finalized as a formal approved exemption by the NRC. The technical basis for this relief was to be the " design of the RHR system at'the associated penetration and ALARA concerns for piping modifications."
- h. - Observations and Findinos Subsequent to the ER initiation, a Plant Operations Review Committee (PORC) meeting was held on July 3,1996 tc discuss the issue and develop a course of action for disposition. A thorough discussion took place amongst the PORC members of the issues which included a re-verification of the operability evaluation provided with the ER and the needed actions to verify that VY remains within its design basis.
The conclusion of the PORC was that an immediate safety concern does not exist because the technical basis for the system's defined Technical Specification operability, in their view, has been maintained. This basis is thht during a design basis event the RHR system would be filled with water and expected to operate post-accident in various modes. Further, the system was designed as a closed seismic loop within the reactor building and is essentially considered an extension of the primary containment. In support of the PORC's conclusion, the following items were examined and will be use in l support of an exemption request for these Appendix J program oversights:
- A detailed review of the FSAR and Technical Specifications.
-- Visual examination of the affected valves for gross valve packing leakage.
- Review drawings and open/ closed work orders.
- . Formal documentation of the operability evaluation via a Basis for Maintaining Operability (BMO).
L Conclusions The VY staff's discovery of Appendix J program discrepancies this inspection period reflect v sitively on the thoroughness of their review of the program. In light of the obvious i oversight to finalize the Appendix J exemption request and pending NRC staff review of ! i- ' the exemption request for the specific valves of interest, this area is unresolved (URI 96- , 1 06-01). L
L 9 L E3.2 Refuel Bridge Modifications - t
- 3,, - Insoection Scoce This inspection was limited in scope and focused on the verification that operator training l
was properly performed and approved prior to Plant Design Change Requests (PDCR) 92-L 11'and 94-003 being declared operable. These design changes were incorporated in i Phases 1 and 2 of the Refuel Bridge Upgrade Project. Phases 1 and 2 of the design upgrade made multiple changes to the refuel bridge to increase its reliability, but this inspection focused only on Temporary Modification (TM) 93 . 053, and the hoist joy stick directional control switch as part of PDCR 92-11. TM 93-053 -
- installed an interlock in the fuel grapple hoist lift circuitry such tnat upward movement of
< the hoist is blocked when the grapple is loaded to greater than 485 pounds, unless both grapple hooks are fully closed. This interlock was installed in response to a dropped fuel assembly event in' September 1993. The modification to the hoist joy stick control switch reversed the directional control in response to a " bumped" fuel assembly event that also
. occurred in September 1993.
L The aspects of PDCR 94-003 inspected included the installation of a human factored (: control console for the bridge, trolley, and main hoist. b Observ'stions and Findinos L .The' inspector interviewed the cognizant modification engineer to determine the proccss l whereby the Training _ Department was informed of plant design changes. The engineer l referenced AP 6000, Plant Design Change Requests, &nd pointed out the procedure p requirement to forward a " text only" copy of the modification package to the Training l Department. The inspector discussed the modifications with the Training Department's instructor who is L currently in charge of refueling operations training. The instructor demonstrated that training was satisf actorily performed, as required, by review of training material and - L
- - student attendance records. The instructor also demonstrated to the inspector the current requirement for an engineering review of modification training packages. The instructor stated that the purpose of the review was to ensure that the Training Department efforts l
L reflect the actual design changes. The inspector reviewed TTD 4.2, instructor Guides, to ) ! confirm this requirement. _ The instructor also explained the Fuel Handling Training i L Program. This program requires formalinitial and requalification training for individuals involved in fuel handling. In order to evaluate the effectiveness of the refuel bridge modification training, the l i I inspector observed operation of the refuel bridge by a qualified operator. The operator referenced plant procedures OP-1100, Refuel Platform Operation and OP 1101, Management of Refuelitig Activities & Fuel Assembly Movement. The inspector observed , movement of the bridge in automatic and manual to various spent fuel pool coordinates l and obsen,3d main hoist operation with a " dummy _" fuel assembly. Phase 2 modifications !
, to the refuel bridge that were observed in operation included the underwater camera l l
L w_- _ _ - _ - _ _ _
10 (which is used to verify grapple position), grapple open/ closed indications, and the computerized display of refuel bridge location and status of hoist loaded conditions. The inspector evaluated the operator's familiarity with the reference procedures throughout the refuel bridge operation and found his knowledge and execution to be good. Additionally, the inspector questioned the operator's opinion of the control console regarding human factors considerations and the operator's response was favorable. The operator was particularly pleased with the addition of the underwater camera and automatic bridge operation features.
- g. Conclusions The training performed for the refueling bridge design changes was performed prior to declaring the modifications operable and was thorough. The installation of the hoist loaded intenock and the human factors changes made were favorable for fuel handling operations.
.}
The implementation of the Fuel Handling Training Program was effective. The requirement for classroom and hands-on training, as part of requalification prior to refue! operations, appropriately refreshed operator awareness of this infrequently performed evolution. E7 Quality Assurance in Engineering Activities E7.1 . Effect of Design Basis Tornado Pressure Load on Diesel Room Enclosures 3, insoection Scoce On June 5, the Performance Engineering staff initiated ER No. 96-0343 to identify the discovery that pressure relief panels that were incorporated into the original design of the emergency diesel generator (EDG) and the EDG day tank room enclosures were not -l installed. The pressure relief panels control the maximum differential pressure effect for
' the design of the walls when subjected to the design basis tornado. The lack of required venting capability p! aces these rooms outside the design basis of the plant as described in Final Safety Analysis Report (FSAR), section 12.2.1. The licensee made a 10 CFR 50.72 notification (Event No. 30584) and took prompt measures to block open the doors for room over-pressure protection and posted compensatory fire and security watches at the open doors.
The Performance Engineering staff's identification of the deficiency was prompted by the ' review of NRC Information Notice 96-06, " Design and Testing Deficiencies of Tornado Dampers at Nuclear Power Plants." Subsequent investigation by the Design Engineering staff was ur:able to locate any information that would support a basis that the pressure relief devices are not necessarr or that existing features meet the venting requirement. j The immediate compensatory measures of blocking open the doors and stationing fire and i security watches were implemented until June 17, when Temporary Modifications (TMs) l l 96-010 and 96-011 were implemented for EDG-1 and EDG-2 room exhausts, respectively. i The TMs blocked open the EDG room exhaust dampels in the normal fully open position, allowing the room doors to be shut and the compensatory watches to be secured. This is an inte..im measure and the TMs must be removed and the dampers restored for cold i L weather operation by October 15. Although the EDG room doors were shut, the EDG day E-- _ _ - - _ - - - - - - - _ - - - - - - - - - - - - - - . - - - - - - . - - - - - - - - - - - - - - - - - - - - - - - - . -- -.- -- -- ----- - - - - - - - - - - - _ _ _ _ _ _ - - - - - _ - - - - - - . -
11
. tank rooms remained in the blocked open positions and compensatory fire and security watches were required to be continued because alternate relief paths were not immediately available for the day tank rooms. At the end of the inspection period, the Design Engineering staff was pursuing a permanent modification or justification for operation without a permanent modification through the upcoming winter season.
A Observations and Findiggg
'In observing the VY staff's response to this design basis issue, the inspector made the following observations:
e initiation of ER 96-0343 on June 5 (and subsequent 10 CFR 50.72 notification and compensatory. measures on the same day) was timely. e -PORC review of Basis for Maintaining Operation (BMO) No. 96-08 on June 12 was consistent with General Letter 91-18 guidance. A written pre-PORC description was thorough and enabled the PORC members to develop a clear safety focus. L Qnclusions The identification of this non-conforming design issue demonstrated a good questioning attitude on the part of the individual engineer who identified the problem and an appropriate review of industry operating experience by the VY staff. The engineering and operations staffs' handling of the (temporary) resolution of the discrepancy and the PORC review of the operability determination were prompt and thorough. E7.2 Review of QA Audit No. VY 96 25 L Insoection Scoce The inspector attended the exit meeting for QA Audit No. VY 95-25, Functional. Area Audit
- Environmental Qualification (EO), conducted on May 16 and reviewed the detailed audit report to assess the depth and scope of the audit, the significance of the audit findings, ' and the licensee's dispositioning of the findings.
A Observations and Findinos The inspector observed the formal presentation of the audit findings and assessments by the lead auditor to the Vice President of Engineering, Plant Manager, and responsible staff members. The presentation was clear and concise with appropriate discussion of audit finding details to ensure good comprehension of the items identified. VY management wa.' attentive and responsive to the auditor's finding. Good discussions evolved to ensure a thorough understanding of the concems and of the auditor's recommendations for improvement. j. The inspector found the audit to be well written and consisting of appropriate detail. Five Event Reports were written to capture the specific and program related audit findings. Per the audit report executive summary, no significant technical inadequacies were identified,
- l
i t 12 I I but increased management attention was recommended by the audit team with respect to , overall EQ Program documentation and administrative controls. The inspector noted that this audit was conducted in support of the broad corrective actions generated as a result of the programmatic deficiencies identified in the VY Fire Protection and Appendix R areas. L Conclusions l The Quality' Assurance staff audit of the VY Environmental Qualification program was thorough and well documented. Station management was receptive and responsive to the ( audit findings and assessments. E8 Miscellaneous Engineering issues E8.1 Review of Licensee Event Reports (LERs) L- Insoection Scoce The inspector reviewed the LERs listed below using the inspector guidance of Manual Chapter 92700 The events reported in these LERs may have been examined and
. documented in earlier inspections and have been referenced accordingly.
I LER No. 96-002, Containment spray header in torus not tested in accordance with ASME Section XI due to personnel error, dated 3/11/96. Reference inspection report 96-05, section M8.1. l l LER No. 96-003, Reactor shield blocks removed during power operations to facilitate outage scheduling due to lack of procedural guidance, dated 3/1/96. Reference inspection report 96-03, section 4.1. _ 1 i l, . LER No. 96 005, Failure to identify and take Technical Specifications required actions
- during design change installation activities due to human error, dated 3/28/96. Reference inspection report 96-05, section 07.1. ,
LER 96-007, Vital fire dampers not installed in accordance with manufacturers instructions, dated 4/3/96. Reference inspection report 96-05, section 02.1. l l, l LER 96-008, Error identified in block wall calculations due to personnel error which places f L plant outside of design basis, dated 4/4/96. reference inspection report 96-05, section l (' E8.1.
- i. '
' LER 96 010, Failure to previously identify dependency between the residual heat removal _j minimum flow valve and the cross-powered residual heat removal pumps, dated 5/9/96. !
' Reference inspection report 96-07.
LER 96-013,'Two fire suppression systems do not meet design requirements due to personnel error on the part of vendor who designed and installed the systems, dated l 5/23/96. Reference inspection report 96-05, section F7.1. l l- j
l 13 h Findinas and Conclusion
]
Based upon in-office review and follow-up of the above LERs and the detailed review of the events as previously documented in the referenced inspection reports, the inspector found the LERs were appropriately written and that the licensee satisfied the reporting requirements of 10 CFR 50.73. IV. Plant Support : i R1 Radiological Protection and Chemistry (RP&C) Controls i i R1.1- Implementation of the Radiological Environmental Monitoring Program a insoection Scoce (84750) The inspector observed and assessed the licensee's capability to implement the Radiological Environmental Monitoring Program (REMP). The inspector reviewed the REMP precedure manual, visited selected sampling locations to confirm that samples were being obtained from the locations specified in the Offsite Dose Calculation Manual (ODCM), witnessed the licensee and the contractor collect and prepare samples, examined the air samplers to determine operability and calibration status, and reviewed the results of the l Land Use Census. The above areas were inspected against specific requirements in the j Technical Specifications (TS), ODCM, and the Updated Final Safety Analysis Report j (UFSAR). A Observations and Findinas l The REMP procedure manual contained all the procedures necessary to collect and prepare environmental sample media. The procedure included air, milk, water sampling methods, dry gas meter calibration calculations for the air samplers, and a method for conducting the
' Land Use Census. The procedures were of good technical content, concise, and provided !
the required direction and guidance for implementing an effective REMP. i The sampling stations included air samplers for airborne iodines and particulate, water sampling stations, a milk station, and several thermoluminescent dosimeter (TLD) stations for measurement of direct ambient radiation. All observed air sampling equipment were operational and calibrated at the time of the inspection. The downstream water compositor was inoperable from April 29,1996 to May 29,1996 because the water level was unusually high and silt accumulated in the pump. The licensee's compensatory actions to collect daily water grab samples were appropriate. The licensee plans to document this compensatory action in the next Annual Radiological Environmental
- Surveillance Report. The licensee and the contractor demonstrated good sample practices while collecting water grab samples and followed the appropriate procedure.
i l 1
14
,cf Conclusions Based on the above review, direct observations, discussions with personnel, and -
examination of procedures and records for calibration of equipment, the inspector determined that the licensee r:ontinued to effectively implement the REMP in accordance i,
' with the TS, ODCM, and UFSAR commitments.
R1.2 Meteorological Monitoring Program (MMP) i
& Insoection Scoce (84750)
- The inspector observed and evaluated the licensee's MMP to determine whether the instruments and equipment were operable, calibrated, and maintained. The inspector also is reviewed the results of upgrading the meteorological monitoring system. The MMP was
' inspected against Section 2.3 of the UFSAR, Regulatory Guide 1.23, and Section 6.0 of
- the Emergency _ Plan.
A Observatioris and Findinos The Instrument and Controls (l&C) department had the responsibility to maintain and calibrate the meteorological monitoring instrumentation. The inspector noted that I&C personnel were very knowledgeable about the system. The calibration procedures emohasized the wind speed, wind direction, and temperature sensors, and other related components. Calibrations of'the meteorological instrumentation were performed every six months according to the Regulatory Guide 1.23 guidance. All reviewed calibration results were within the established acceptance criteria. , in August 1995, the licensee completed work to upgrade the meteorological monitoring system. All sensors, recording devices and electronics, with the exception underground
- wiring, were replaced. The digital strip charts were replaced with video-graphic recorders (CRT) which have disk and magnetic tape storage capability. The video-graphic recorder eliminates the need to exchange strip chart paper; instead the magnetic tape is replaced.
The initiative to upgrade the meteorological monitoring system was very good. L ' Conclusion Based on the above review, direct observations, discussions with personnel, and examination of procedures and records for calibration of equipment,'the inspector
. determined that the licensee continued to effectively implement the MMP in accordance with the UFSAR commitments, Regulatory Guide 1.23, and the Emergency Plan.
l l ___ - _ _ - - _ a
15 R6 RP&C Organization and Administration R6.1 - Management Controls L Insoection Scoce (84570) The inspector reviewed the organization changes and the annual environmental reports to verify the implementation of the TS requirements. 1 Observations and Rndings The inspector reviewed the organizations responsible for the REMP and MMP and discussed with the licensee any changes to the programs since the previous inspection conducted in November 1994. There had been no changes in the organizations since the previous inspection. I The inspector also reviewed the Annual Radiological Environmental Surveillance Reports for timely deportability and the results of the routine analysis of REMP samples and quality assurance results. The Annual Radiological Environmental Surveillance Reports for 1994 and 1995 provided a comprehensive summary of the analytical results of the REMP around the Vermont Yankee site and met TS reporting requirements. No obvious omissions, anomalous data, or adverse trends were identified. E Conclusion Based on the above review, the inspector determined that the licensee maintained effective I management controls for implementation of the REMP and MMP. ) R7 ' Quality Assurance in RP&C Activities - 1 R7.1 Quality Assurance Audit Reports ! i
& Insoection Scoce (84750)- .
The inspector reviewed Quality Assurance (QA) audit reports against criteria contained in TS requirement 6.2.B.5. A - Observations and Rndinos The inspector reviewed the following QA audit reports for REMP and MMP. ' l Audit Report No. VY95-02, dated March 2,1995 ' ( Audit Report No. VY96-02, dated March 1,1996. Both scope and technical depth of the audits were excellent to assess the REMP and MMP. Both audits thoroughly assessed the REMP for strengths and weaknesses. These audits were conducted by the Quality Assurance Audit Group and technical specialists. Few findings and several recommendations were identified as a result of the audits. The e
16 responsible departments incorporated these findings and recommendations in a timely manner to enhance the REMP and MMP. In an effort to provide a means to oversee quality assurance at the Yankee Atomic Environmental Laboratory (YAEL) and minimize the impact on the laboratory, the five sponsor companies (Vermont Yankee, Yankee Atomic, Maine Yankee, Pilgrim, 'nd Seabrook) formed the Laboratory Quality Control Audit Committee (LOCAC) to conduct annual quality assurance audits. The LOCAC audit scope covered several areas including procedures, analysis, quality control, and qur.lity assurance programs. The inspector reviewed the LOCAC audit, "1995 Audit of Yankee Atomic Environmental Laboratory", are noted that the audit was thorough and of sufficient technical depth to assess the strengths and weaknesses of the laboratory programs. Response to the findings and recommendations were timely. c Conclusions Based on the above review, the inspector determined that the licensee conducted audits of sufficient technical depth to assess the quality of the REMP and MMP. R7.2. Quality Assurance of Analytical Measurements a inscection Scoce (84750) The inspector reviewed the licensee's Quality Assurance (QA) Program for analytical measurements of radiological environmental samples including the Interlaboratory Comparison Program (EPA Cross-check Program), required by the TS and ODCM. h Observations and Findinos The QA program consisted of measurements of blind duplicate (split) samples. Samples were collected and submitted by the licensee to the contractor, Yankee Atomic Environmental Laboratory (YAEL), where they were prepared and analyzed. The results were sent to a Senior Environmental Engineer at Yankee Nuclear Services Division (YNSD) for review, assessment and reporting. The inspector reviewed the analytical results and noted that the results were in agreement, with few exceptions. Where discrepancies were found, reasons for the differences were investigated and resolved. The laboratory participated in the EPA Cross-check Program. The results were within the EPA's acceptance criteria. In 1996, the licensee started to use a vendor laboratory, Analytics, Inc., to continue the Interlaboratory Comparison Program since the EPA would no longer provide this service after December 1995. L Conclusion Based on the above reviews and discussions, the inspsetor determined that the licensee continued to implement a very good laboratory analysis quality assurance program in accordance with regulatory requirements.
17 R8 Miscellaneous RP&C lssues -Inappropriate Personnel Frisking Event a Insoection Scoce On July 2 while accompanying the auxiliary operator (AO) assigned to plant outside dutie;, the inspector observed the AO perform a hand and foot frisk versus a whole body frisk when exiting the Radiation Controlled Area (RCA). L Observations and Findinos The inspector and AO were in the Advanced Offgas (AOG) Building during routine rounds. When the AO prepared to leave the RCA to the roof area without performing a whole body frisk the inspector questioned the AO as to the posted requirement for a whole body frisk. The AO responded that the expectation for frisking to enter the root area was a hand and foot frisk only, because prior to exiting the AOG building a whole body frisk would be performed at the lower elevation exit / entrance of the building (the only allowed ingress and egress point). The inspector and AO then completed a hand and foot frisk and entered the roof area. Prior to exiting the AOG building, the inspector and AO performed a whole body frisk, as required. The following day (July 3), the inspector again questioned the AO to gain a better understanding of the basis for the expectation that a whole body frisk was not required to enter the roof area of the AOG building. Following this discussion, the AO discussed h;s actions and understanding of frisking expectations with the Radiation Protection Department. The AO's interpretation of frisking requirements for the AOG building roof access was determined at that time to be in conflict with the requirements delineated in Radiation Protection Procedure (RP)-4532, Personnel Mcnitoring When Exiting Restricted Areas, revision 22. ER No. 96-0422 was subsequently written by the AO. Following this event the licensee issued a Night Order on July 3,1996 to inform the Operations Department personnel of this event and required verbatim compliance of radiation protection signs. Additionally, the plant general manager issued a memorandum reminding plant personnel of the importance of strict adh';rence to plant postings and if any questions arise to get clarification before proceeding. L Conclusions The failure to adhere to the frisk posting is an example of failure to follow procedures and is contrary to TS 6.5. However, this failure to adhere to the required personnel frisking requirements to the AOG building roof constitutes a violation of minor significance and is being treated as a non-cited violation, consistent with Section IV of the NRC Enforcement Policy.
i-I 18 L
^P4 Staff Knowledge and Performance in EP
& ' garcise Evaluation Scoce The inspector observed and evaluated the licensee's off-year Emergency Preparedness Exercise on May 23,1996 to assess the adequacy of the licensee's emergency response
. program, implementation.of the Emergency Plan, the emergency plan implementing procedures, and the training program for emergency response. The inspector assessed classification of emergency conditions, notification of off-site agencies, overall communications, determination of protective action recommendations, and the post-exercise critique.
O Emeraency Resoonse Facility Observations and Criticue b.1 Simulator Control Room (SCR) l Communication between the Shift Supervisor (SS) and the SCR team was very good. Briefings were generally concise and detailed. The SCR team quickly determined abnormal conditions and correctly interpreted simulated events. The SS immediately understood the l
' direction of the scenario and declared the appropriate emergency classification levels (ECL) ll
! using emergency action levels (EALs) based on the plant conditions. Detection and classification of events were timely and correct. Notifications to the States and NRC were timely and accurate. The team was confident and knowledgeable, and demonstrated good communication and teamwork while implementing the actions of the emergency plan implementing procedures. f During the previous exercise, (see Section 6.0 of NRC Inspection Report 50-271/95-20, l l:
' dated November 6,1996 for details) the inspector noted untimely plant announcements L for the Unusual Event (UE), Alert, and Site Area Emergency declarations. In addition, the ~
SCR team did not implement the immediate actions of procedures OP 3500,3501 and e 3502 that direct prompt in-plant announcements. i' ..
. During this inspection, the inspector noted that all announcements were accurate and i.
timely. The announcements were made within 10 minutes of determining emergency L levels.~.The team followed procedures appropriate for the events of the scenario. L L b.2 . Technical Support Center (TSC) The inspector observed and evaluated the performance' of the licensee in the TSC including management and control, notifications and communications, assistance and support to the SCR, dispatch and coordination of monitoring teams. The TSC team properly managed site activities and provided technical support to reactor
. operations. A Senior Reactor Operator (SRO) was utilized to relay scenario information obtained from the SCR to the TSC team. The communicator between the TSC, the Operational Support Center (OSC) and other centers, was located in the TSC which resulted in good communications between the teams. Hourly briefings were accurate,
' informing the team about the status of the plant and the scenario. Recommendations were a__-__ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _-- -_. . _ _ _ . - _ _ _ _ - _ _ - _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ - -
m I 19 appropriate and provided technical support to the SCR team. Dispatch and coordination of monitoring teams was quick and well coordinated. Status boards and log books were effectively utilized, providing visual support and a record of the event. During the previous exercise, (see Section 7.0 of NRC Inspection Report 50-271/95-20, dated November 6,1996 for details), the inspector noted that no means to maintain an up-l to-date status of operating emergency core cooling (ECCS) equipment had been provided in the TSC. During this exercise, the inspector noted that the licensee had and utilized status boards of the ECCS, the Primary Containment Isolation System (PCIS), and the 125VDC electrical buses. These boards were instrumentalin determining operating systems status. b.3 Emergency Operating Facility (EOF) The inspector observed and evaluated the performance of the licensee in the EOF including staffing and activation, facility management and control, notifications and communications, assistance and support to the SCR, and interaction with offsite officials. Staffing and activation of the EOF was complete within 45 minutes after the announcement of the alert declaration. The Site Recovery Manager (SRM) promptly obtained information from the SCR and TSC, quickly became cognizant of the scenario, and fully activated the EOF in a timely manner. At the same time, the EOF staff quickly organized equipment to perform dose assessment, on-site and off site dose projection, and radiation protection support. ' Status boards were initiated quickly and contained correct and accurate data. The initial briefing to the State representatives was observed. The briefing, performed by an SRM staff member, was accurate and included a concise but detailed description of the scenario. Overall, facility management and control, communications, staffing and activation, and support to the SCR were good. b.4 Licensee Exercise Critique The inspector observed and evaluated the critique held in the TSC following the conclusion of the exercise and the formal critique presented by an exercise coordinator. During the preliminary critique, the licensee noted strengths, weaknesses, and elements of the TSC team that were an improvement compared to the previous exercise. Overall, the formal critique was balanced, citing strengths, and areas for improvement. The licensee candidly identified weaknesses and categorized them toward training considerations. The findings were reasonable and accurate. L Qvysit Exercise Conclusions The licensee's overall performance was very good. The facilities were generally staffed r and activated in a prompt manner. Good direction and control was observed at all of the i facilities. The four classifications of the simulated events w2re timely and accurate and , off-site notifications were completed within time 10 minutes. Communications among the j teams was very good, resulting in a smooth translation of information, and turnover from I SCR, TSC, to EOF. l
20 P8 Miscellaneous EP issues P.8.1 (Closed) IFl 95-20 01: Untimely plant announcements for the UE, Alert, and site
, Ares emergency declarations.
The Simulator Control Room (SCR) staff did not implement the immediate actions of procedures OP 3500,3501 and 3502 that direct prompt in-plant announcements. During ! the EP Exercise on May 23,1996, the inspector noted that licensee promptly announced I the emergency declarations, as expected. The announcements were made within 5 minutes after determining emergency level. l P.8.2 (Clo;:ed) IFl 95-20-02: No means provided to maintain up-to date status of safety related equipment !n the Technical Support Center (TSC). TSC personnel were inquiring as to the status of operating emergency core cooling (ECCS) equipment. No status boards existed to provide the operability status of ECCS and engineered safety features. During the EP Exercise on May 23,1996, the inspector noted that the licensee developed three status boards. One depicting the ECCS systems, one describing the PCIS, inct another referencing the 125V DC buses. These status boards were located in the TSC, EOF, and the SCR. The boards were effective in providing operability status of the affected systems, u F5- Fire Protection and Staff Training and Qualification L insoection Scoce The licensee conducted a fire drill on the afternoon of June 11. The inspector monitored ! the fire brigade actions at the scene and the post-drill critique. j h ' Observations and Findinos Licensee training personnel simulated a fire in the main transformer. The fire brigade l responded promptly to the control room announcement in accordance with the Fire Brigade and Fire Fighting operating procedure. Fire brigade members appeared efficient and knowledgeable of applicable site practices. In addition, the inspector observed that training : personnel provided effective training by providing useful, perceptive comments in the post-drill critique. E Conclusion
-Training personnel used a fire drill to provide effective fire brigade training.
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i l 21 V. Management Meetings X1 Exit Meeting Summary The inspectors presented the inspection results to VY station management periodically throughout the inspection period. . Subsequent to the conclusion of the inspection period, the inspection findings and conclusions were discussed with Mr. Robert Wanczyk and other members of his staff on July 23,1996. The licensee acknowledged the findings ' presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified. ! X2 Updated Final Safety Analysis Report (UFSAR) Review i A recent discovery of a licensee operating their facility in a manner contrary to the Updatec ; Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused l review that compares plant practices, procedures and/or parameters to the UFSAR l description. While performing the inspection discussed in this report, the inspector l reviewed the applicable portions of the UFSAR that related to the areas inspected. The inspector verified that the UFSAR wording was consistent with the observed plant practices and procedures and/or parameters. 1 l j e l 1
22 INSPECTION PROCEDURES USED 84750 Radioactive Waste Treatment, and Effluent and Environmental Monitoring 62703 Maintenance Observations 61726 Surveillance Observations 71707 Plant Operations 92700 Onsite Followup of Written Reports of Non-Routine Events 93702 Prompt Onsite Response to Events at Operating Power Reactors 37551 Onsite Engineering 71750 Plant Support Activities 40500 Effectiveness of Licensee Controls in Idenfitying, Resulting and Preventing Problems ITEMS OPENED, CLOSED, AND DISCUSSED CLOSED IR 95-20-01 Untimely plant announcements for the UE, Alert, and Site Area Emergency declarations. IFl 95-20-02 No means provided to maintain up-to-date status of safety related equipment in the Technical Support Center. OPEN URI 96-06-01 Appendix J testing of certain valves has not been conducted and specific exemption requests are pending licensee submittal and NRC staff review. I l 1 J l
l 1 23 PARTIAL LIST OF PERSONS CONTACTED ! Licensee l E. Undamood, Technical Services Superintendent G. Maret, Operations Superintendent E. Porter, Emergency Plan Coordinator D. Reid, Vice President Operations R. Sojka, Operations Support Manager R. Wanczyk, Plant Manager S. Skibniowsky, Chemistry Manager M. Desilets, Radiatior: Protection Manager
' The inspector also interviewed other licensee and contractor personnel.
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24 LIST OF ACRONYMS USED EAL Emergency Action Level ECCS Emergency Core Cooling System ECL Emergency Classification Level EOF Emergency Operations Facility NRC Nuclear Regulatory Commission OSC Operations Support Center PCIS Primary Containment isolation System SCR Simulator Control Room SRO Senior Reactor Operator SS Shift Supervisor TSC Technical Support Center UFSAR Updated Final Safety Analysis Report EPA Environmental Protection Agency MMP Meteorological Monitoring Program ODCM Offsite Dose Calculation Manual QA Quality Assurance REMP Radiological Environmental Monitoring Program TLD Thermoluminescent Dosimeter TS Technical Specifications YAEL Yankee Atomic Environmental Laboratory YNSD- Yankee Nuclear Services Division 1 i i l w__-__________}}