ML20198K926

Fourth Partial Response to FOIA Request for Documents. Documents Listed in App D Being Released in Entirety

ML20198K926
Person / Time
Issue date:	01/13/1998
From:	Racquel Powell NRC OFFICE OF ADMINISTRATION (ADM)
To:	Jim Hickey SHAW, PITTMAN, POTTS & TROWBRIDGE
References
FOIA-97-365 NUDOCS 9801150130
Download: ML20198K926 (2)

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6 RESPONSE Ta FREEDOM OF piNat y j eutiat (4 ui) k\\

,0NFORMATION ACT (FOIA) REQUEST (C p3, 36hg g att-UUCE,l t kuMBL k %l (lisophtable)

RtOut STE R J. Patrick Hickey PART 1,-AGENCY RECORDS RELE ASED OR NOT LOCATED (See checAraboats/

No erney records subject to the request have been located.

No additional syncy records subject to the request have been located, Requested records are available through another pubhc distribution program. See Comments section.

Arncy recorde subject to the request that are identified in Append;a(es) are already available for pubhc inspection and copying et the NRC Public Document Room,2120 L Street, N.W., Washmgton, DC, X

Apocy records subject to the request that are identified in Appendentes!

O are being made aetable for pubhc inspection and copying at the NRC Publ,c Document Room,2120 L Street, N.W., Washington, DC,in a folder under this FOI A number.

The nonproprietary ve sion of the proposal (s) that you agreed to accept in a telephone conversation with a rnember of my staf f is now being made available for public inspection and copying at the NRC Pubhc Document Room,2120 L Street, N W., Washington, DC, in a folder under this FOI A number, Apncy records subject to the request that are identified in Appendixtes) may be inspected and copied at the N RC Local Public Document Aoom identified in the Comments sectlon.

Lntitned is mformation on how you may obtain access to and the chargrs for copying records located at the NRC Public Docur-ent Room,2120 L Street, N!% Washington, DC.

y Apocy records subject to the request are enclosed.

• Records subject to the request have been referred to another Federal agency (ies) for review and tra:t resronse to you.

Fees You will be billed by tha NRC for fees totakng $

You will receive a refund from the NRC in the smount of 5 in view of N RC's response to this request, no further action is being taken on appeal letter dated

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PART 11. A-lNFORMATION WtTHHELD FROM PUBLIC DISCLOSURE Certrin information in the sequested records is being withheld from public disclosure pursuant to the exemptions described in and for the reasons stated in ftrt 11, B, C, and D. Any reieased portions of the documents for which only part of the record is being wit %eid are being made avlsilable for pablic inspection and copying in the NRC Public Documt nt Room,2120 L Street, N.W, Washington, DC in a folder mer this FOIA number, COMMENTS

• Copics of the records identified on Appendix D are enclosed.

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- APPENDlX D RECORDS BEING RELEASED IN THEIR ENTIRETY J

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MQ2 - DAIE-DESCRIPTION /(PAGE COUNT) 1, 5/3/96 Letter to D Reid, Vermont Yankee Nuclear Power Corp. from R. - Conte (Exhibit 16)(14 pages) 2.

4/9/97 Letter to D Reid, Vermont Yankee Nuclear Power Corp froni-R Conte (Exhibit 17) (7 pages) a 4

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September 19,1997-

. Mr. Russell A. Powell, Chief Freedom of informatior Act/ Local Public Document Room Branch U.S. Nuclear Regulatt / Commission

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Washington, D.C. 20555-0001 ~

Re:- Freedom of information Act (FOIA) Request

Dear Mr. Powell,

Pursuant to 5 U.S.C. 9 552 (a)(3) and 10 C.F.R. 9.23, I am requesting all documents relating to the investigation by the NRC Region i Office of Investigations of Case No. 1-96-005 (allegations related to Vermont Yankee).

-l agree in advance to pay any reasonable fees associated with this request.

. Thank y$u for your prompt attention to this request. I look forward to your response within ten (10) days, as required by NRC and FOIA regulations.

Sincerely, J.' Patrick Hickey G 1-AS?!? ?

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9 Case No. 1 96 005 Exhibit 16

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.LNUCLEAR REGULATORY COMMISSION.

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KING OF PRUSSIA, PENNSYLVANIA 194001415

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i i"E May 3.-1996L p

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Mr.i Donald R41d?

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-Vice President, Operations-Vement Yankee Nuclear Power Corporation _

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RD 5 Box 169 Ferry Road-3 Brattleboro,. Vemont 05301-f

SUBJECT:

--NRC INSP : TION REPORT 50-271/96-03 x,

Dear Mr. Reid:

I i0n March 30,lity. The enclosed report presents the results of that1996, the NRC reactor. faci

~1nspection.

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 W(

plant; issues identified by the VY staff during this inspection period (for LEy 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. SVY 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-I degraded conditions that exceeded the Technical Specification requirements Leither for-minimum channel availability, A0G-related instrumentation,. or system operability.

No reply to this report is necessary and your cooperation with us is I

. appreciated.

Sincerely, y

. Richard J. Conte, Chief Reactor. Projects Branch 5 Division of Reactor Projects Docket No. 50-271 e

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96.005:

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. Enclosure 1:

Executive Summary : NRC Resident Inspection Report Enclosure'3: NRC Specialist Inspector A0G_ Review Report ccw/ enc 1:-

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,- Vennont Public Interest Research Group. Inc.

D. Tofft, 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 t

commonwealth of Massachusetts, SLO Designee l.

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EXECUTIVE

SUMMARY

Vermont-Yankee' Nuclear power Station I

-NRC Inspection Report 50-271/gf-03 This integrated inspection included aspects of licensee' operations >

engineering, maintenance, and plant support. The report covers a 6-week

. period of resident inspection.

Oserations l

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-Overall, the conduct of plant operations was professional and safety conscious

- this inspection period. The plant Operations Review committee (p0RC) ion of exhibited a clear and well defined safety focus during their examinat recent & service Testing and Appendix J Program eiscrepancies and during their review cf 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 omrators on the manual isolation valves to both residual heat removal heat excsangers.

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

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was good, however, the apparent cause of the trip was identified to have been A

personnel error.

An inspection follow-up item (IFI g6-03-01) was assigned to review VY's root cause evaluation and corrective actions.

The VY staff's approach to monitoring and understanding the scram salenoid l

pilot valve VIT0N diaphragm degradat' on issue has been and continues to be I

aggressive. However, VY's increased frequency of individual rod scram time l

testing may potentially conflict with Technical Specifications 4.3.C.2 if appropriate administrative controls are not instituted.

pending further VY j

staff and inspector review, this issue is unresolved (URI g6-03-02).

The VY staff's decision to stpone the reactor core isolation cooling system l

and the '8' emergency diese generator (EDG) limiting condition for operation maintenance outages, during this inspection period, demonstrated prudent ion making with safety benefits.

l Eneineering L

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 romptly dispositioning the station L

-batteries operability impact, was not ti ly. p0RC's review of the station l

batteries operability determination was completed and. as referenced above, l-the PORC's decision to postpone the '8' EDG LC0 maintenance outage was i

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prudent. - The NRC staff review of this potentialky degraded condition using the guidance of Generic Letter 91-18 was ongoing at the conclusion of the C

inspect 16n period and was unresolved (URI 96-03-03).

W engineering and operating staffs' have appropriately dealt with-the torus-

-water temperature-limit concern, to date,ively restricting torus waterby pursuine.further analyses and,-in the interim, adeintstrat temperature to 90 degrees F.

pending completion of format analysis = of this

'l potential design bask conflict and NRC staff review, this issue is unresolved i

(URI96-03-04).

Licensee identified and corrected discrepancies in the Inservice Testing and Appendix J Programs (reference LERs96-001 and 90-004, respectively) were dispositioned as non-cited violstions. These discrepancies'were identified by the W staff as a result of thorough corrective action for organizational problems identified via the Fire protection and Appendix R Programs.

J Tne inspector reviewed the current technical status of the advanced off-gas i

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(A0G) systou as part of the verification process for the licensee's letter (No. BW 96-17), dated February 26, 1996.- In particular, the inspector reviewed issues dealing with A0G system performance and with a system modification cancellation. Tne engineering staff's coordination with the plant staff, the quality of the consolidated as-built panel 9-50 electrical 1

drawings, and the delis.eation and resolution of design issues for the planned A06 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

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with possible incomplete documentation, such as installation and test

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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 1

L Specification requirements.

Recent initiativ6s including system-analyzed maintenance developed by the I&C engineering staff and reliability-based maintenance developed by the maintenance engineering staff were considered good.

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flan 13unnart W's ongoing systematic re-examination of the entire Fire Prctection and Appendix R Programs-identified a number of improperly installed fire dampers and incomplete test data for the switchgear. rooms carbon dioxide suppression l

systems. The-com>ensatory measures for these discrepancies were promptly implemented and tse proposed corrective actions deemed appropritte.'

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 s sparent root cause of this problem was inadequate procedural gu dance, but furtier evaluation was ongoing.

pending W completion and inspector review of the final root cause evaluation, l

this issue is unresolved (URI g6-03-05).

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-EMCLOSURE 3

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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 4

't INSPECTOR:

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.n Calvirt, Reactor Engineer Date ectrical Engineering Branch Division of Reactor Safety s

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APPROVED:

11111am Rulan'd, Chief Cate /

Electrical Engineering Branch Division of Reactor Safety 4

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REPORT DETAILS FOR VUNIONT YAKEE

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IN$pECTION REPORT NO. 96-43

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1.0; ' ADVANCED OFFSAS SYSTM NODIFICATION REVIDf L

This inspection was conducted on March 3 - 8. 1996,- at the Vermont Yankeer site,;and Apr11:10 - 11,:1996, by telecon from the NRC Region I off1ces.

i' 1.1 Scope and Background (Ip37700)-

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Scope The inspector reviewed the current technical status of the advanced off-gas i

_ (A06): system, as part of the verificat16n process for th6 licensee's' letter to-the IRC (No. SVY 96-17), dated February 26, 1996. The inspector reviewed those issues dealing w9th system perfomance deterioration and a modification I

cancellation. The radiation aspects of the A06 system are covered, in part, l

-in reports'50-271/95-25, Section 5.1, and 50-271/95-24. The hydrogen analyzer aspects of the A06 system are covered in inspection reports l

Section 3.2 and 50-271/95-25, Section 3.2.4.

50-271/96-02, I

The inspector interviewed personnel associated with the A06 system and/or modification such as the design engineer, electrical engineer, I&C engineer,_

maintenance engineer, project engineering manager, and operations support l

engineer. The former project manager for the A06 modification declined to be interviewed by the inspector for this inspection.

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The inspector reviewed pertinent licensee documents associated with the A0G system and modification such as the UFSAR, LERs, licensee safety evaluation,

!&C work orders Itst, I&C procedures list, I&C system-analyzed maintenance L

($AM), operator rounds procedures, reliability-based maintenance-[RBM), systez specifications / drawings,- modification drawings, drawing modificatSon job file, design engineering memorandums, and project engineering memorandums.

The inspector performed walkdowns of the control room A06 panel and accessible equipment in the A06 building.

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Background

The A06 system iras added to the Vermont Yankee plant in 1973. Tha 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 Itack: hydrogen dilution; hydrogen recombination; preliminary

delay for decay.of radioactive gases; soisture removal; charcoal adsorption; and final delay for decay of radioactive gases. Except for the passive delay pipes and charcoal adsorbers, the A06 consists of two trains of squipment with cross connet. tion capabilities.for certain equipment.

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J A bypass lino is installed downstream of the hydrogen dilution /recombiner

trains and the preliminary delay pipe. h bypass does not include bypass of the hydrogen dilution /recombiner t?ains~. W bypass permits continued reactor

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! operation if portions of both trains of the moisture removal and the charcoal-3 adsorbers were to become~ inoperable during normal operations. The bypass line.

. is joieed 3o the input 1T the final delay pipe for transport to the plant stack.

4 h radiatten it reduadattly monitored aftsr the bypass line, just before the final delay ph;4.- If the radiation monitoring levels of the monitors extwed a

- preset 1er. automatic action occurs to shut off flow to the stack from the A06 system. The plant stock has additional radiation monitoring, but no l:

automatic control function.

2.0 GRSERVATIONS AIS FIIBINGS t.1 Review of 10 CFR 50.59 Safety Evaluation for the planned Nodification 1he inspector reviewed the document. *EDCR 34-02 Enclosure (A)lassified non-Safety Evaluation." h equipment involved in the modification was c nuclear safety. The evaluation arrived at the appropriate conclusions and showed that neither the functions of any safety-related system would be 4

degraded, nor would the margin of safety be degraded as defined in the Technical Specification for the A06 system.

2.2 Review of Nodification planning -

The modification, EDCR 94-402, was designed by the licensee to improve the reliability operations and maintainabiltty 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 j

performed as minor modifications for the 1995 and 1996 outages.

An initial licensee engineering evaluation was made at the beginning of the j

' modification design in 1992 to identify problems in the areas of reliability, i

L operation, and performance of maintenance activities. W major areas listed L

below were identified as needing improvement.

L (1) provide a verified, unified set of drawings for operations, maintenance, l

and It<C activities. perform field varification of the as-built wiring.

L Revise the A06 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 engineering standards, for example, redundancy and instrument fusing.

The revision centrol of the drawings was poor..In the case of the piping and instrumentation diagrams (P&ID), there were two identical

-sets of drawings done by two different contractors.

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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 n-postulated failure of a sinnie link in the instroent ac bus occurred, it could cause extensive fa'lurs of the A0G 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. Mainterance could be performed, but i

not efficiently. Over tw years, as minar changes were made, the confidence in the accuracy and completeness of the electrical drawings was questioned by operations and maintenance personnel.

f (2)

Change the level control and pumping system for the A0G condensate drain tank, T b104-1. This is an interfacing system to the A06 that is used during normal operations.

The tank receives extracted moisture fron,the A06 process lines. The water is then pumped to the main condenser. The tank level provides a water barrier Jetween the condenser vacuum and the A06 process lines.

ir During start-ups from extended outages or in cold weather, a large

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volume of condensate is produced and A06 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.

l 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 staan jet air ejector (SJAE) inter-condenser.

l 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 approximately eight times, necessitating a unit shutdown to replace the disc. With the uncreased inter-condenser pressure rating, the setpoint of the rupture disc could likewise be increased. The inspector determined that system availability, rather than system functionality, j

was the focus of this ites.

The inspector notes that the potential radiological release cotisequences

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of a rupture of the A0G rupture dise has been the subject of previous exy &

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NRC staff review and follow-up (reference inspection reports 92-41, 92-r 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 modificat' on was completed in the Fall of 1993, the licensee took appropriate interin measures to monitor potential releases via the turbine building ventilation system pathway. The inspector also reviewed the app 1' ctble 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 operat'ons. Examples were the replacement ~of analog recorders with digital chartless type recorders, replacement of analog controllers with digital controllers, circuit changes to eliminate spurious Aos annunciator alarms, and re-configuration of the instrument air supply to valves to prevent c

unnecessary loss of recombiner heat exchangers due to loss of instrument air.

1.3 Review of Nodification Design The design engineering coordination with the operations, 1&C, and maintenance j

functions to ' dentify concerns, analyze the concerns, evaluate the alternatives, and document the design bases was performed well. This was i{r-participation in site walkdowns.

indicated by the depth of the design analyses and the active engineering l

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-l to-point walkdown and verification of the g-50 panel as-built drawings and found only one minor difference, whicn 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 insta)1ation process. This-indicated good wrformance of the overall drawing review process,-but indicated weakness in tae engineerir.g drawing check procedures for the modification.

The licensee checked the electrical wiring in the A0G building to confirm conformance with the original systen installation drawings. $everal discrepancies were found and corrected. None of the discrepancies affected proper functioning of the system, according to the licensee.

L 1.4 Status of the Rodification The licensee managment made a decision to not include the modification in the April 1995 refueling outage. The inspector reviewed two internal VY documents, "A06 Mods Design Change,' December 22, 1994, (VY Vice President, Engineering to VY Department Manager and project Manager), and "A0G Design

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Changes,' January 24, 1995, (VY Vice President, Engineering to VY Department Manager) for reasons why the decision was made. The inspector found from the EX iT

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. review, supplemented by personnel interviews, information to suggest that the main. reason for the management decision was because all the documentation

. necessar,y 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 g-50 panel wiring, where the neutral wires were from different power supplies than the hot wires, was j

changed as a task during the 1995 refueling outage.,

4 The change of the postulated failure of a single link in-the instrument AC electrical bus-and the effect on the instrumentstion was not implemented as a i

separate task-during the 1995 outage..The licensee has this. task under review i

for future implementation. The inspector noted that if this postulated' failure of a single link were to occur and if all AD instrumentation would i

consequently fall it would be covered by the Limiting Condition for 0peration (LC0) in the requirements for minimum number of channels operable of Technical Specification (TS) Table 3.g.2 (ta,b.c), Gaseous Effluent Monitoring i

i Instrumentation, which covers radiation monitoring, flow rate, and hydrogen l

i 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 j

. pressures were continuously monitored.

2.5 Review of Raintenance Engineering a.

Instrument and Controls (I&C) l The I&C engineering group started a system-analyzed maintenance (SAM) project ll for the A04 system in 1992. The inspector reviewed the document 'VY I&C Preventative Raintenance, System Analyzed Maintenance Project (SAM),"

Revision 2, datad February 6, 1996, that described the purpose, objectives, and methodology af 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 e'xperience. An appropriate i

l maintenance task with an associated interval is then assigned.

g The A06 system has 370 instruments, 318 of which have been classified to date and are planned for final engineering review. Approximately 85% of the L

instruments were classified as functionally important. The inspector reviewed l

L a sample of ten completed worksheets for the functionally important equipment i

and noted that the function was clearly described, and the general requirements for the maintenance task was delineated.

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Rochenical The maintenance engineering group had a project similar to the IAC SAM project.

-called reliability-based maintenance (RBM). The A06 mechanical components had all keen classified and had.been reviewed. The inspector selected a

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the recombiner drain (H0-06-587) and reviewed the basis of classification, and b

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noted that the functional importance basis and maintenance actions were appropriate for the valve service.

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2.5 Review of Raintenance Status The inspector reviewed a listing of work orders on the A06 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 prohloss occurred that were not addressed by corrective actions.

The inspector detere led that there were I&C instrument calibration procedures for the system insti tentation. Additionally, there were functional /calibrati i procedures for the A0G hydrogen monitors, radiation menitors, trip systei, 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 C talyst ' memo number VYI-2/g6, 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 than expected, which could indicate higher output hydrogen L

concentration. The licensee performed a grab sample check of the output that showed that the hydrogen concentration was well within the normal pitnt operating band.

Engineering recommended that the inside and outside t,urfaces of the thermocouple wells be cleaned and the thermocouples be checked as the corrective action for the 1cw outlet temperatures.

Engineering also found that the design specification for the catalyst life was for 18 months to 5 years, so they conservatively recomended replacement of the catalysts at the next refueling outage, even though the data showed that the recombiners are properly performing their function. This recomendation 1s being reviewed by licensee management.

Further discussion of the catalyst life is found in section 3.0.

2.7 system Walkdown The A06 building was very well laid out and the inspector considered redundancy, physical separation, shleiding, and maintainability. The material condition was generally very good.

The A06 control room panel CRP g-50 arrangement provided the necessary 4 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

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. radiation; and system outlet flow. The panel had an extensive temperature EXHlBIT /M?

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1 monitoring panel-for indication of system temperatures at heat exchanners, for example. Additionally the panel had pressure indications for key po'nts in i

the system.

3.0 ; REVIEN OF WSAR Ale ColglITNENTS -

A recent discovery of a licensee operating their facility in a manner contrary

'to the Updated Final Safety Analysis Re art (UFSAR) description highlighted the need for a special= focused review tiat compares plant practices.

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-procedures and/cr parameters to the UFSAR descriptions.

While performing the inspections-discussed in this report, the inspectors reviewed Section g.4, "Saseous Radwaste System,' of the UFSAR that related to the areas inspected. -The inspectors verified that the UFSAR wording for the A06 system was consistent witi the observed plant practices, procedures and/or parameters.- The modifications of EDCR g4-402 were not checked versus the UFSAR because the EDCR was cancelled and planned to be implemented at a later date.

3

- As a result of their follow-up of the A0G system hydrmen monitor issues, the i

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 Off9as-processing System components," page A-3, that the recombiner catalyst 'ifetime was estimated to be equivalent to plant lifetime when operating in the steam /offgas environment. They also sooted inconsistent lifetime catalyst

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information in the proprietary Appendix C, page C-3 of Attachment A to the A

licensing change request, which described the catalyst expected life as less than plant lifetime.- The licensee's resolution of the inconsistent-recombiner catalyst lifetime in the A06 license change request documents and any impact on catalyst maintenance is an unresolved ' tea (URI 50-271/96-43-07).

4.0 RANAGEMENT OVER$IGHT project engineering management initiated an independent engineering review of the design implementation process.

Ins section information suggested that plant and engineering management made tie decision to not include the modification in the 1995 shutdown because all the documentation necessary for t

the modification package had slipped schedule allestones. The inspector inferred from these actions that management was actively involved in the oversight of the modification.

Tha 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 doctmentation.

6.0 CONCLUSION

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.1he licensee's resolution of the inconsistent recombiner catalyst lifetime in the A0G license change request documents and any impact on catalyst maintenance is an unresolved ites (URI 50-271/96-03-07).

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1 The insMor found no engineering or maintenance indications-in the last 5 years' t5at A06 system functionality was impaired in such a manner that led to degraded conditions that. exceeded the Technical Specification requirements seither for minimum channel: availability, or A06 system instrumentation, or A0G system operability. -

'The insputor 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 thstallation and test instructions.

The engineering coordination with the plant staff, the quality of the

-and the delineation and consolidated as-built panel g-50 electrical drawings,ication were generally o

resolution of design issues for the planned A06 modif

-very good.

I The I&C and maintenance engineering initiatives re2arding A0G system analyzed 1&C maintenance and reliab lity-based mechanical maintenance for components 1

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 B, 1996, as listed in Section 7.0.

The licensee acknowledged the

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l findings presented.

l The inspector telephoned the licensee on April 10 and 11,1996, for additional information on the maintenance for the hydrogen recombiner catalyst.

The inspector received and reviewed proprietary material during the inspection i

and used the material for te,:hnical reference. No proprietary information was i

l knowingly included = in the report.

7.0 - LIST OF PEOPLE CONTACTED 1'

Vermont Yankee Nuclear Power Cornorat 4D J. Rolvin Manager, Technical Support

.E. Bowman Operations Engineer B.'Buteau Manager, Engineering Reorganization Coordination

• D..Calsyn Supervisor, Quality Assurance L. Casey**

Dasign Engineer, VMSD:

• 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 i

. G. Maret Superintendent. Operations

• D. ' McElwee** -

State Liaison Engineer L

• S.- Miller **

Manager, Design Engineering, YNSD

• J. Pelletier Executive Director, YNS f-y

• D. Reid.

Vice President, Operations EX BIT /b S

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R. Routhier Electrical Engineer

• J. Thayer Vice President,' Engineering J. Todd Maintenance engineer-R. Wancryk-Plant Manager N. Watson **

Manager, I&C Vermont Denartment of Public Service

. *W. Sherman-State Nuclear Engineer U.S.N.R.C.'

• J. Calvert Reactor Engineer, Electrical Engineering Branch

• W. Cook Senior Ressdent Inspector

• Present at exit meeting on March 8, 1996

• • Contacted by telephone April 10 - 11, 1996

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l EXHIBIT 17

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\\ C-DhL Case No. 1 96 005 Exhibit 17

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April 9,1997 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 50471/97-02 NOTICE OF VIOLATION

Dear Mr. Reid:

On March 8,1997, the NRC completed an inspection at your Vermont Yankee reactor facility. The enclosed report presents the results of that inspection.

As discussed in Section E1.2 of the enclosed report, a root cause evaluation of the mixing of electrical power supply neutral and ground wires in the advanced off gas (AOG) system had not been completed during the on site inspection by a representative of the Office of

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Nuclear Reactor Regulations. The neutral mixing problem and other wiring error problems appear to be limited to the AOG system, and they do not adversely impact AOG system operable. However, the full extent of this wiring control problem has yet to be determined and resolved. Without the completed root cause analysis, we are concerned that the identified non safety related neutrals and grounds wiring problem may potentially involve safety related systems. Upon completion of your staff's evaluatir.n of this issue, please provide us, in writing, *,'our findings, corrective actions, and basis for assuring that those problems do not adversely impact safety related or important to safety systems. Also, we request that you provide un with a status of resolution of all of the AOG system wiring errors and, if not sesolved, provide a date when these errors will be corrected.

Inspector review of the November 25,1996 event involving the on line de energization of the 480V electrical bus No. 6 for preventive maintenance identified that your staff failed to implement the administrative cut'ol requirements for removing power supplies from service. This failure to implement the procedural requirements of Administration Procedure (AP) 0125 and Technical Specif! cation 8.5 is a violation. Of particular concern to us was that this evolution was previously performed during unit outages and that more rigorous reviews were not conducted prior to conducting this planned activity at power. We viewed the shift supervisor's performance as the last line of defense which failed leading to this event and the maintenance and planning staff's failures as more programmatic.

This violation 4 cited in the enclosed Notice of Violation, and the circumnances surrounding the violation are described in detailin the enclosed report. Please note that you are required to respond to this letter and steuld follow the instructions specified in the

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enclosed Notice when preparing your response. The NRC will use your response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements.

EXHIBIT CMENO.

1-96-005' e

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C Mr. Donald Reid

' 2 in accordance with 10 CFR 2.790 of the NRC's " Rules of Practices," a copy of this letter

[

end the enclosed Notice will be placed in the NRC Public Document Room.

Sincerely, ORl0lNAL SIGNED BY:

Richard J. Conte, Chief Reactor Projects Branch 5 i

Division of Reactor Projects j

Enclosures:

1. Notice of Violation

2. NRC Inspection Report 50 271/97 02 Docket No. 50 271 cc w/ encl:

R. McCullough, Operating Experience Coordinator. Vermont Yankee R. Wancryk, Director, Safety and Regulatory Affairs

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G. Marot, Plant Manager J. Duffy, Licensing Engineer, Vermont Yankee Nuclear Power Corporation J. Gilroy, Director, Vermont Public Interest Research Group, Inc.

D. Tofft, 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 T. Rapone, Massachusetts Executive Office of Public Safety l

State of New Hampshire, SLO Designee State of Vermont, SLO Designee Commonwealth of Massachusetts, SLO Designee l

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EXHIBIT /

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e ENCLOSURE 2 U.S. NUCLEAR REGULATORY COMMISSION REGION I Docket No.

50 271 Licensee No.

DPR 28 Report No.-

97 02 Licensee: -

Vermont Yankee Nuclear Power Corporation Facility:

Vermont Yankee Nuclear Power Station Location:

Vernon, Vermont

' ' Dates:

January 19 March 8,1997 Inspectors:

William A. Cook, Senior Resident inspector Edward C. Knutson, Resident inspector Laurie A. Peluso, Radiation Physicist, Reglon i Thomas Koshy, Sr., Reactor Systems Engineer, NRR

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Approved by:

Richard J. Conte, Chief, Projects Branch 6 Division of Reactor Projects 4

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The Operations department root cause evaluation for the November 25,1996 electrical bus l'r No. 6 on line de energiration event was adequately self critical and identified the principal causal factors for the event. The inspector concluded that the root cause of the event was a failure to fo!!ow AP 0125, Appendix B and cited this procedural non compliance as a violanon of regulatory requirements (VIO 97 02-04). Also, the inspector concluded, as did the hconsee, that a weakness in the work planning process significantly contributed to this evolution not being reviewed mcre thoroughly, in advance.

Per the original innpector concern raised in inspection report 9416, the licenseo completed an appropriate representative sample of harsh environment susceptible instrument loop error calculations and confirmed that, although some instrument TS allowable values may be exceeded, the "analyticallimits" did not compromise the bounding LOCA safety analyses.

Accordingly, URI 9416 02 is closed. However, the licensee's application of ISA S67.041987 and basis for concluding that the instrument loop derived " analytical limit" may exceed the TS allowable value remain unresolved. This unresolved item, along with the NRC staff's review of the Instrument Setpoint Program results, will be examined in a future inspection (URI 97 02 09).

Enoineerino The licensee has adequately addressed the issue regarding the absence of heaters for the motors of the HPCI and RCIC systems. The licensee's routine preventive maintenance program and the timeliness of any needed corrective maintenance provide reasonable assurance of continued operability of these components.

The NRR inspector concluded that there was no immediate safety concern involving the mixing of electrical neutrals and ground wires on this issue since only a narrow range of electrical f ault protection is potentially degreded. An Unresolved item (URI 97 02 05) will track future NRC reviews of VY staff actions to ensure resolution of any further electrical separation issues involving the AOG system, including an evaluation of the cause for this problem and on assessment of the adequacy of corrective actions. The performance history of the AOG system demonstrates the system's capability to comply with the applicable NRC and state effluent release regulations.

A number of plant design issues were identified during this inspection period requiring additional licensee action and inspector follow up/ verification of corrective actinn adequacy, inspection follow items have been assigned to these issues, as annotated in the report.

Plant Suongn Radiation worker and radiation protection staff performance associated with the free releasing of the Framatome bridge assembly was generally good, but some inattention with respect to the monitoring of the clean area boundary was noted.

On February 12, the licensee performod a practice emergency plan exercise for training purposes. The practice exercise involved participation by the local offsite ernergency response organizations. The inspector observed thc licensee's activities from the Technical C

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the rooms warmer whenever the plant is at power. (Reactor steam production is needed for these systems to operate). If condensation was to occur, it should be away from these rooms where the temperature is relatively lower.

i Additionally, the licensee's preventive maintenance program includes the measuring of the i

insulation resistance for these effected motors. These measurements should provide en l

early detection of any insulation degradation. The test results and motor performance in the last seven years, after the space heaters were disconnected, have been acceptable. The motors that were part of the original plant equipment continued to function, as dw..t,nstrated in the surveillance tests, in its required service.

. The primary safety function of the HPCI and MCic systems is to mitigate a small break LOCA while the reactor coolant system pressure remains above the capability of low l

' pressure safety injection systems. The HPCI and RCIC rooms should remain unchallenged l

during a design bases small break LOCA and therefore, these systems should be able to perform their prescribed safety function. in case of a large break LOCA, these systems are

. r.3t credited and their contribution is insignificant in relation to the large volume low i

pressure safety systems required for accident mitigation.

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In case of a HELB inside the reactor building, the credited systems are automatic i

depressurization, and "A" and "B" containment spray. More specifically, the HPCI and l

RCIC system motors are not relied on as the primary emergency core cooling systems when r

they are susceptible to degradation from the accident environment. Additionally, the l

demand for HPCI and RCIC is just after the accident and therefore the system should begin i

v 1

its function before any potential condensation could reach motor windings. This delay in environmental challenge to these motors would further reduce the possibility for i

condensation because of the motor heat up from its operation. This is in agreement with the licensee's classification of these motors for equipment qualification, i

The NRR inspector concurred with the prior regional assessment documented in inspection reports g6 Og and g6 06 that the motor heaters were not required for the HPCl and RCIC i

systems to perform their safety function, i

c.

Conclusions The licensee has adequately addressed the issue regarding the absence of hesters for the motors of the HPCI and RCIC systems. The licensee's routine preventive maintenance program and the timeliness of any needed corrective maintenance provide reasonable L

assurance of continued operability of these components.

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.E1.2 Mont and Electrieel Power System e.;

inanection Scone (37700)

The NRR inspector reviewed the advanced off gas (AOG) system and its associated j

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. electrical power, control, and instrumentation power supplies to verify that electrical protection and signal separation was in compliance with the Updated Final Safety Analysis M

Report (UFSAR) commitments.

EXHIBIT PAGEdOF 7 PAGE(S) f

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Observation and Findinas The inspector examined drawing revisions that addressed the failure to follow wiring practices that could de0rade the performance of circuit protective devices to promptly isolate electrical f aults.

The VY AOG system design la unique to the nuclear industry. The system was designed through a contractor. During design improvement reviews performed between 1994 and 1995, the plant staff noticed that the panel wiring did not agree with the drawings. In a May 1995 memorandum, the licensee documented the completion of as built drawing toviews and the return of system wiring to sound electrical wiring practices. The licensee identified four wiring errors in control room pt,nel 9 50, where the neutral of the plant's instrument AC system was used as an electrical return path for the AOG system power

sources, in an October 8,1996 memorandum, VY requested a re review of the grounding and neutra' connections in response to additional questions raised in this area and an event at the Palo Verde Nuclear Power Station involving a short circuit causing smoke in two fire zones.

From this review and as documented by memorandum dated January 8,1997, one more wiring deficiency involving crossing the neutral connection was found, along with other minor drawing errors. This crossing error involved connecting a non safety control circuit neutral to a different non safety AOG motor control center located in the same room.

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The wiring errors identified were two kinds: (1) neutral side of power cables connected to the instrument AC system, and (2) neutral of the control wire connected to the neutral of an adjacent motor control center. The first kind of error could cause electrical pulses to travel onto an instrument bus and cause interference to sensitive instrumentation. The first and the second kind of errors degrade the fault isolation capability, if the neutral connection for the respective power t.ource is not in the direct wiring, but through a different MCC or instrument bus and then through the plant ground to the respective MCC, the impedance in the circuit is increased in proportion to the length of the wire. This added impedance would reduce the fault current and could prevent or delay the protective device from clearing the fault. The potential problem with the medlurn to high impedance faults, on the wires that run between field equipment and control room panels, is that they could remain uncleared.

The inspector determined that the low impedance type fault should not be a problem in the deficiencies identified.

The UFSAR, Section 8.4.6.6, " Intermixing of Cables," revision 12, states that " low level instrumentation cables are routed in separate trays from control cab;es." These low level cables are defined as anything carrying less than 50 volts. The electrical connection of neut:als between instrumentation and control cables invalidates the intent of the FSAR commitment in physical separation. The NRR inspector reviewed the plant trip history for the last 3 years, and did not reveal any questionable association with AOG system problems.

The licensee's focus on this electrical neutrals to ground mixing problem was limited to the AOG system. A root cause evaluation of this concern was not done, but the licensee initiated an evaluation prior to the completion of the inspector's onsite visit. The NRR EX IBIT PAGE OF '7 PAGE(S)

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Idapector's review concluded that the primary cause was that the' licensee's contractor

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i lacked knoC;;: in good wiring practices. Good wiring practices are essential to circuit

protection to prevent the spread of electrical faults. Based on the inapoctor's inquiry, the licensee agreed to look into any of this contractor's work and later confirmed that the l

contractor did not work on any other non safety related or safety related systems. The j

inspector _also determined that the problems associated with the AOG system were not i

tracked through the present problem resolution (Event Report) process. Accordingly, an l

Unresolved item (URI 974245) has been assigned to ensore NRC follow up revie, of the j

VY staff's action to conduct a root cause evaluation for these AOG system groundmg l

. issues and to. ensure mised grounding is not a broader concem to other systems in the plant.

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EgghaWan

-The NMR inspector concluded that there is no immediate safety concem on this issue since

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only a narrow range of electrical fault protection is potentially degraded. An Unresolved

. Item (URI 974248) will track future NRC reviews of VY staff actions to ensure resolution of any further electrical separation issues involving the AOG system, including an evaluation-l of the cause for this problem and an assessment of the adequacy of corrective actions. The i

performance history of the AOG system demonstrates the system's capability to comply l

j.

with the applicable NRC and state effluent release regulations.-

i

!c E7 Quality Assurance in Engineering Activities E7.1 Design Basis Documentation / Technical Specification improvement Projects issues The licensee's Design Basis Documentation (DBD) and ITS projects have the potential for identifying inconsistencies between the design, licensing, and operating bases of plant

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structures, systems, and components. Such inconsistencies will be documented in this section of the report and tracked to resolution as inspection follow items, p

During the inspection period, the licensee identified the following issues:-

Event Reoort 97177, use of the main station batteries standbv charoer " CAB".

VY has two trains of emergency electrical power, Division I and Division ll. Each division includes a 125VDC main station battery (A 1 and 81) and a dedicated battery charger

. (CA 1 and CB 1) which receives power from the AC portion of the division. A standby.

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' battery charger, CAB, can be connecteo to eit_her battery, in the event that the dedicated i

charger falls or requires maintenance. Technical Specification 3.10.A.2.b allows the atendby charger to be used indefinitely in place of either of the dedicated chargers, i

However, the standby charger can only be powered from one source, motor control center MCC 88, which is a Division I electrical power source.--if the standby charger was being

=

used in place of the Division il dedicated charger and a loss of Division i AC power occurred, loss;of both' division DC systems would eventually occur due to inability to

. recharge either of the batteries." in response to this finding, the licensee instituted administrative controls to restrict the use of the standby charger with the. Division ll battery,

? (IFl 974248)

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