ML20045B929

Responds to Concerns Raised by Constituent Hh Chase Re Conditions at Facility

ML20045B929
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	06/15/1993
From:	Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	Swett D HOUSE OF REP.
References
NUDOCS 9306210278
Download: ML20045B929 (6)

Text

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The lionorable Dick Swett June 15, 1993' United States House of Representatives 5

~a Washington, DC 20515

Dear Congressman Swett:

In a letter dated April 13, 1993, I provided you NRC's response to concerns raised by your constituent, Mr. H. Hamilton Chase, regarding conditions at the Vermont Yankee Nuclear Power Station. Mr. Chase's letter i

raised concerns that had also been raised by the New England Coalition on Nuclear Pollution in a letter dated February 9,1993, which was awaiting a response from the NRC.

In our letter of April 13, 1993, we-stated that we would furnish you a copy of the NRC response to the Coalition when it was available.

For your information and use, Enclosure 1 is Chairman Selin's response to the Coalition, and Enclosure 2 is a copy of my previous letter to you.

I trust that this information is responsive to your concerns.

Sincerely, 9306210278 930615 OfI$IMIIIOYI0dbY fDR ADOCK 05000272 James M. Tay1or James H.Snlezek PDR Executive Director for Operations

Enclosures:

j 1.

Letter to M. Daley dated May 26, 1993 i

2.

Letter to D. Swett e7 dated April 13, 1993

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February 5, 1993 Congressman Dick Swett 123 Cannon Dffice Building Tashington, D.C.

20515

Dear Congressman Swett,

I write as a Keene, New Hampshire citizen, concerned for the gen ;

eral saf ety of our livina area, regarding nuclear danger. This concern is directed at Verme-nkee.

'I Once again, we hav-the plant operator and the NRC working in a closed huddle, deciding which health safety risks should be made or known Io the general public.

The New England Coalition on Nuclear Pollution has for over two decades acted' responsibly, yet persistently, and without scare tactics or wild accusations. Yet the complaints entered and under discussion i

over the past 6 months, were set aside by agreement between the' plant 1

i operator and the NRC. The NRC is chartered to safeguard the public in_t-

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erest and safety,. not to help plants _in their financial stress. Why.

j Men are the documented plant safety concerns simply rejected. and the

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NRC made none of its criteria public, presented no facts-to back-up

-j opinion, and failed to address the full scope of the concerns put forth:

by the Coalition?

The history of both the plant and=the Coalition. deserves that Chairman Ivan Selin of NRC carry out a public inves t iga t ion:of condi-tions at Vermont Yankee and of oversight practices in NRC Region I. and

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make formal answers for public awareness.

These days, it takes public or citizen action'to get government'to reccgnize concerns and welfare of citizens, from pollution, to-S&L rob-bery. to environmental degradation,to economic woes..I hope a lot of citizens take up this cry. and write their Congress people, and Chair:-

4 man Selin, so the NRC will pay close attention.

Sincerely, g

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February 5, 1993 Congressman Dick Swett 128 Cannon Office Building Washington, D.C. 20515 h

Dear Congressman Swett,

I write as a Keene, New Hampshire citizen, concerned for the gen-eral saf ety of our living area, regarding nuclear danger. This concern F

is directed at Vermont Yankee.

Once again, we have the plant operator and the NRC working in a closed huddle, deciding which health or safety risks should be made known to the general public.

The New England Coalition on Nuclear Pol'lution~has for over two decades acted responsibly, yet persistently, and without scare tactics or wild accusations. Yet the complaints ente, red and under discussion over the past 6 months, were set aside by agreement between the plant operator and the NRC. The NRC is chartered to-safeguard the public int-erest and safety,, not to help plants in their financial stress. Why then are the documented plant safety concerns simply rejected, and the NRC made none of its criteria public, presented no facts to back up opinion, and failed to address the full scope of the concerns put forth by the Coalition?

The history of both the plant and the Coalition deserves that-Chairman Ivan Selin of NRC carry out.a public inves tigation df condi-tions at Vermont Yankee and of oversight practices in NRC Region 1,'and make formal answers for public awareness.

These days, it takes public or citizen ac'tlon'to get

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government to recognize concerns and welfare of citizens, from pollution, to S&L rob-bery, to environmental degradation,to economic woes. I hope a lot of citizens take up this cry, and write their Congres's people, and. Chair-man Selin, so the NRC will pay close attention.

Sincerely,_

CS MeleN*.KeneA S

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wAss m arou,o.c. zones May 26, 1993 CHAIRMAN Mr. Michael J. Daley New England Coalition on Nuclear Pollution, Inc.

Box 545 Brattleboro, Vermont 05301

Dear Mr. Daley:

On behalf of the Commission, I am responding to your letter of February 9, 1993, in which you expressed disappointment in the NRC Region I Administrator's answer to a letter you sent him on December 16, 1992.

I am also acknowledging your letters of April 8 and 11,1993,- in which you-expressed concern about the emergency diesel generators at the Vermont Yankee Nuclear Power Station. These two letters are being handled separately as a petition submitted under 10 CFR 2.206 and have been referred to the Director of the Office of Nuclear Reactor Regulation. The Commission and the NRC.

staff, including our Region I office, recognize the seriousness of the concerns you have raised, respect your views on equipment performance and other reactor safety issues at the Vermont Yankee Nuclear Power Station, and share your interest in the continued safe operation of the plant.

The Region I Administrator's letter of January 25, 1993, did.not deal lightly with the issues you raised. The staff evaluated the information in your letter, but believed these issues had been adequacely treated in the past. A summary of the approximately 40 events that you cited in your December letter, including the NRC's independent evaluation of both root causes and corrective actions, is provided in Enclosure 1.

The Commission too is concerned about patterns of poor performance and declining performance trends.

For this reason, the NRC inspection and assessment processes are performance oriented. The adequacy of a licensee's support of a plant (material, resource, and otherwise) is assessed on a continuing basis through independent inspections, the semi-annual senior NRC

- management review process, and our systematic assessment of licensee performance-(SALP) program. The type, degree, and schedule of inspectionsare reviewed at least semi-annually for each facility. Adjustments are made consistent with the licensee's performance as observed during inspections.

The most recently completed SALP evaluation at Vermont Yankee covered the period from March 1991 to August'1992 and reached conclusions about functional areas in which performance was rated as superior, where it was observed to be good and improving, and areas that showed declining performance.

The SALP Board concluded that Vermont Yankee conducted its activities in a safe manner.

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i The NRC's SALP evaluations, including the most recently issued report for Vermont Yankee, are an integrated assessment of inspection findings, enforcement actions, plant performance, and reportable events, such as those cited in your December letter. The SALP Board looks for common root causes indicative of potential problems, either by functional area (e.g., maintenance and surveillance) or cause code. Such causal analyses of reportable events are integral to the SALP process. A summary breakdown of all Vermont Yankee Licensee Event Reports (LERs) spanning the 6-year period of 1987 through 1992 is listed in Enclosure 2.

This analysis encompassed the last four SALP reports issued by the NRC. In fact, LERs were typically evaluated individually during NRC inspections as well as collectively in the SALP process. During the 6-year period of 1987 through 1992, there were 145 inspection reports issued involving more than 18,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of on-site-independent inspection. We believe the results summarized, both qualitatively in our SALPs and quantitatively in Enclosure 2,- support the conclusions stated in our earlier letter to you relative to equipment degradation.

The NRC will continue to monitor Vermont Yankee's performance.

In the near term, inspection initiatives have been undertaken specifically to evaluate the Vermont Yankee Nuclear Power Corporation's resolution of the fire barrier issues, emergency diesel generator reliability problems, and control rod drive surveillance test anomalies.

Further, the staff has scheduled a Service Water System Operational Performance Inspection for later this year to confirm system performance and plans to conduct an Operational Safety Team Inspection in June 1993 to further evaluate performance across several functional areas in an integrated fashion.

We want to assure you that any new information regarding Vermont Yankee's performance developed through our inspection process or from any other source will be evaluated promptly and that effective corrective actions will be taken as warranted by circumstances.

Sincerely, I

Ivan Selin r

Enclosures:

l 1.

Summary of Events 2.

Vermont Yankee LERs, 1987-1992 1

m.

ENCLOSURE 1

SUMMARY

OF EVENTS All licensee event reports (LERs) and other issues cited in the December 16, 1992, letter from the New England Coalition on Nuclear Pollution were (NECNP) previously reviewed and inspected by the NRC.

No new issues are evident and the events encompass a wide range of areas.

Generally, the licensee's corrective actions were independently found by the NRC to be adequate.

Issues Raised in 12/16/92 NECNP Letter

SUMMARY

REFERENCES ROOT CAUSE AND RESOLUTION Inadvertent emergency NRC Inspection Root cause attributed to an core cooling system Report (IR) 89-02, informal restoration procedure.

(ECCS) actuation while LER-89-15 The NRC determined that the repowering the "B" licensee's corrective actions were ECCS logic system.

adequate (personnel training, procedure development).

Inadvertent core spray IR 89-07, LER Root cause attributed to an (CS) and residual heat 89 16 inadequate test procedure. The removal (RHR) pump NRC determined that the licensee's start due to inadequate corrective actions were adequate procedure.

(design review, procedure modification ).

High pressure coolant IR 91-11, IR Root cause attributed to a injection (HPCI) 91-29, LER 91-07 component failure. The NRC inoperable (INOP) due determined that the licensee's to flow controller corrective actions were adequate setpoint drift.

(flow controller replaced, additional monitoring of controller setpoints).

HPCI INOP due to IR 92-06, IR Root cause attributed to component degraded battery bus 92 04, LER 92-04 failure. The NRC determined that voltage.

the licensee's corrective actions were adequate (alarm response sheet revision, component replacement, engineering review, and personnel training). The NRC issued a Notice of Violation to the licensee for falling to make the required 10 CFR 50.72 report within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2 issues Raised in 12/16/92 NECNP Letter

SUMMARY

REFERENCES ROOT CAUSE AND RESOLUTION Reactor core isolation IR 89-09, IR Root cause attributed to component cooling (RCIC) INOP 89-12, IR 89-22, failure due to excessive cycling over due to a MOV failure.

and LER 89-14 a short period of time. The NRC determined that the licensee's corrective actions were adequate (component replacement and failure analysis, procedural enhancements).

RCIC INOP due to flow IR 92-12 and LER Root cause attributed to component controller drift.

92-15 failure. The NRC determined that the licensee's corrective actions were adequate (controller calibration and increased setpoint monitoring).

Inadvertent reactor IR 90 03, IR 9122 Root cause attributed to personnel scram due to a short and LER 90-09 error. The NRC determined that circuit on the vital AC the licensee's corrective actions bus which was caused were adequate (personnel training).

by personnel error.

Overloaded power IR 88-14, IR 88-19 Root cause attributed to a design supply in fire protection and LER 8812 deficiency. The NRC determined control panels.

that the corrective actions were adequate. The NRC it. sued a Notice of Violation to the licensee for inadequate design control.

Failure to meet IR 90-03 and LER Root cause attributed to design separation criteria for 90-08 deficiency. The NRC determined j

power cables to that the licensee's corrective actions Regulatory Guide 1.97 were acceptable (modification of the instrumentation instrument cabling).

Lack of redundancy in IR 89-09, IR Root cause attributed to an RHR service water 89-12, IR 91-03 inadequate modification design systems.

and LER 89-09 review. The NRC determined that the licensee's corrective actions were adequate (modification and engineering review).

3 issues Raised in 12/16/92 NECNP Letter

SUMMARY

REFERENCES ROOT CAUSE AND RESOLUTION Failure to meet IR 90-10, IR Multiple root causes were technical specifications 9015, LER 90-10 identified, including engineering for the emergency diesel interface with plant operations, generator (EDG) timeliness y incorporating changes operational readiness into the Final Safety Analysis test.

Report (FSAR), adequacy of technical review and the accuracy of vendor supplied Information.

The NRC determined that the licensee's corrective actions were adequate.

Reduced cooling water IR 91-13, IR Root cause attributed to an flow to the EDGs and 91-21, IR 91-19, inadequate safety evaluation. The station sen' ice air LER 91 12 SRC issued an SL III Violation for compressors due to high the licensee's failure to perform a service water system written safety evaluation as backpressure.

required by 10 CFR 50.59. The licensee developed a program to strengthen its safety review process.

The NRC is reviewing the effectiveness of this program.

Removal of a technical IR 89-17, LER Root cause attributed to inadequate specification 89 20 review of a procedural revision.

surveillance requirement The NRC determined that the from a procedure due to licensee's corrective actions were an inadequate review.

acceptable (detailed review of related procedures, procedural revision, and personnel instruction on review of technical specification i

procedures).

l

)

4 issues Raised in 12/16/92 NECNP Letter

SUMMARY

REFERENCES ROOT CAUSE AND RESOLUTION Reactor scram resu! ting IR 91-07, LER Root cause attributed to component from a loss of o!T site 91-05 failure (falled insulator in the 345-power (LOOP) that was kV switchyard). The NRC caused by the determined that the licensee's mechanical failure of a corrective actions were adequate 345-kV switchyard bus (component repair, switchyard which was attributed to inspection).

a broken high voltage insulator stack.

Reactor scram caused IR 91-12, IR Root cause attributed to an by LOOP caused by an 91-13, IR 91-19, inadequate procedum (maintenance inadequate procedure IR 91-22, LER guideline). The NRC determined guideline.

91-09.

that the licensee's corrective actions were adequate to support plant restart.

Reactor scram due to IR 91-14,91-22, Root cause attributed to a lightning loss of a 345-kV and LER strike. Some equipment problems switchyard.

91-14.

were noted during the plant trip which were corrected prior to plant startup.

Loss of all normal IR 87-16, LER Root cause attributed to a fault on power during shutdown 87-08 an off site transmission line and due to routing all o!T.

the routing of all off site power site power sources sources through one breaker. The through one breaker.

NRC determined that the licensee's planned corrective actions (procedural revisions) were appropriate.

Degraded grid IR 92-09, IR Root cause attributed to setpoint undervoltage (UV) 93-04, LER drift. The NRC determined that relays found below 92-12 the licensee's planned corrective technical specification actions (reset relays and limits.

engineering evaluation) wem appropriate.

i

a 5

issues Raised in 12/16/92 NECNP Letter

SUMMARY

REFERENCES ROOT CAUSE AND RESOLUTION Failed relay coil results IR 91-12, LER Root cause attributed to component in a primary 91-10 failure (lack of an established containment isolation service Ilfe for normally energized system actuation.

GE CR 120 relays). The NRC determined that the licensee's corrective actions (established a service life for this component) were appropriate.

Ioss of 'B' loop IR 91-07, IR Root cause attributed to an shutdown cooling due to 91-19, IR 91 11, inadequate procedure for the pressure switch IR 92-01, LER system configuration. Final NRC actuation.

91-06 review of the licensee's corrective actions were adequate (procedurt revision, and engineering analysis is pending).

Failure to perform daily IR 8917, IR Root cause attributed to inadequate instrument checks on 91-05, LER 89 23 procedural review. The NRC the low-pressure coolant determined that the licensee's injection (LPCI) system corrective actions were good crosstle monitor.

(procedural revision and review).

Containment isolation IR 91 14, IR Root cause attributed to component valve failure to close 9124, LER 9115 failure (erosion of the screw-in seat due to erosion / corrosion threads). The NRC determined and displacement of the that the licensee's corrective actions screw-in seat, were adequate (replaced valve seat and inspection of similar valves).

Turbine trip and reactor IR 88-08, IR Root cause attributed to component scram due to feedwater 8810, LER 88-07 failure. The NRC determined that controller malfunction.

the licensee's corrective actions were adequate.

6 issues Raised in 12/16/92 NECNP Letter

SUMMARY

REFERENCES ROOT CAUSE AND RESOLUTION Appendix J Type B and IR 84 26, IR Root cause attributed to component C failure due to check 85-08, IR 8810, failure. The NRC determined valve seat leakage.

IR 89-02, IR (NRC Inspection Report 50-271/89-9015, IR

02) that the licensee's corrective 91-19, IR 92-09, actions prior the 1990 refueling IR 92-16, IR outage were ineffective. During the 92-18, LER 8411, 1990 refueling outage, the licensee LER 89-07, replaced the valve's elastometric LER 9012, LER seating material with stellite. This 92 10 valve (FDW-96A) passed its leak test during the 1992 refueling outage. The NRC determined that the licensee's frequent repair of this valve, indicated earlier ineffective corrective action, but was not Indicative of a tolerance for a degraded condition.

Service water check IR 89 07, IR Root cause attributed to valves inoperable due to 91-19, LER 89-17 microbe induced corrosion (MIC).

corrosion of internal The NRC determined that the parts.

licensee's corrective actions were adequate (increased inspection l

frequency, and system piping upgrades).

Inadvertent scram and IR 92 06, IR Root cause attributed to personnel ECCS Initiation while 92-09, LER 92-014 error (poor coordination when restoring level restoring the vessel level reference transmitters to service, legs to service). The NRC determined that the licensee's corrective' actions were adequate (system restoration and personnel training).

7 issues Raised in 12/16/92 NECNP Letter

SUMMARY

REFERENCES ROOT CAUSE AND RESOLUTION Missed diesel fire pump 1R 91-07, IR Root cause attributed to an fuel oil suncillance due 91 11, IR 91-14, inadequate procedure. The NRC to inadequate '

91-19 LER 91-03 issued a non-cited violation to the procedure.

licensee, and concluded that the licensee's corrective actions wem adequate (oil sampled, procedure revised, review of all chemistry department procedures, and training enhancements).

Reactor vessel inventory IR 89 02, LER Root cause attributed to personnel decrease due to 89-13 error. The NRC determined that personnel error.

the licensee's corrective actions were adequate (procedural revisions and administrative controls).

Plant service water IR 88-03, IR Root cause attributed to an effluent stream not 88-06, IR 88-14, inadequate procedure. The NRC monitored due to an LER 88-01 determined that the licensee's inadequate procedure.

corrective actions were not properly implemented and issued a Notice of Violation (Severity Level IV) for this event.

Missed elliuent sample IR 88-14, IR 88 20 Root cause attributed to inadequate due to inadequate LER 8814 implementation of corrective corrective action in LER actions for LER 88-01. The NRC 88-01.

Issued a Notice of Violation to the licensee for failing to meet this technical specification requirement.

The licensee took additional corrective actions (procedural and log sheet revisions, and personnel training) which NRC determined were adequate.

8 Issues Raised in 12/16/92 NECNP Letter

SUMMARY

REFERENCES ROOT CAUSE AND RESOLUTION Missed RHR valve IR 89-22, LER Root cause attributed to an leakage surveillance due 89 24 Inadequate procedural review. The to incomplete procedure licensee's investigation determined review.

that this valve was not required to be in the IST Program and was subsequently removed from the program. The NRC determined that the licensee's corrective actions were appropriate.

Priman containment IR 90-15, IR Root cause attributed to a weakness isolation system (PCIS) 9018, LER 90-18 in the testing procedure. The hTC spurious actuation due determined that the licensee's to an inadequate corrective actions were appropriate procedure.

(procedural revisions).

Improper inservice flow IR 9216, LER Root cause attributed to personnel testing of the control 92 16 error. The NRC determined that room chilled water the licensee's corrective actions pump due to ASME were adequate (procedure revisions, Code misinterpretation IST Program review, personnel and subsequent missed training).

quarterly test due to incorrectly following the test procedure.

'A' emergency diesel IR 91-19 Root cause attributed to component generator fuel oil failure. The NRC determined that transfer pump the licensee's post maintenance operability, testing was adequate to ensure operability.

'B' emergency diesel IR 91-19 Root cause attributed to component generator failure to failure (wear of the fuel oil safety start.

valve). The NRC determined that the licensee's investigation and repair efforts were thorough.

l

L 9

Issues Raised in 12/16/92 NECNP Letter

SUMMARY

REFERENCES ROOT CA'USE AND RESOLUTION "B" EDG failure to start IR 92-06 Root cause(s) attributed to during ECCS testlog.

component malfunction (s). The first failure was due to incomplete resetting of diesel governor shutdown plunger which the licensee attributed at first to the advanced age of the governor. A second failure occurred when the diesel generator output breaker failed to shut due to binding on an auto start relay. The NRC -

determined that the licensee's

- repairs and investigations were adequate.

i

ENCLOSURE 2 VERMONT YANKEE LERs 1987 THROUGH 1992 Of the 115 LERs issued in the 6-year period from 1987 through 1992, a majority (36%)

involved personnel errors - not an uncommon cause given the period of time or complexity of operations and testing. Deficient procedures account for more than 20% of the events involved, and this is an i' sue for human performance and administrative controls (and not s

equipment degradation). Smaller fractions for design deficiencies, external, other and unknown causes account for less than 20% of the events.

CAUSE CODES SUBTOTAIS FUNCTIONAL AREA A

B C

D E

X (BY AREA)

Operations 17 1

1 6

17 7

49 Radiological Controls 7

0 0

1 0

0 8

Maintenance and Surveillance 16 a

1 11 11 4

43 Emergency Preparedness 0

Security 0

Engineering and Technical Support 1

8 0

5 1

0 15 Safety Assessment /

Ouality verification 0

Subtotals 41 9

2 23 29 11 115 Notes:

1. LERs issued from 1987 through 1992

2. Cause codes identified are based upon NRC staff evaluations and inspections of the events, and may in certain instances differ from those specified in the LER.

3. Cause codes are:

A.

Personnel Error B.

Design C.

External or UnknowTi D.

Procedure Inadequacy E.

Component Failure X.

Other

4. LERs in the functional area of security do not include safeguards reports under 10 CFR Part 73.

2 The balance (29 events or approximately 25%) involve component failures. In evaluating the potential degradation of safety equipment due to deficiencies in test or maintenance practices, reportable events caused by component failures are of interest. A breakdown of these component failures over the six year period shows that relatively few of the LERs caused by component failure were attributed to deficiencies in maintenance and surveillance test programs. Such deficiencies would be othenvise indicative of weakness in predictive and preventive methods. A chronological breakdown by SALP period is provided below and indicates that there is no increasing trend over the period in question.

COMPONENT FAILURES SALP REPORT DUE TO MAINTENANCE PERIOD NUMBER TOTAL OR TEST 1/87 - 6/88 87-99 9

3 7/88 - 9/89 88-99 7

3 10/89 - 3/91 89-99 5

0 3/91 - 8/92 91-99 7

4 8/92 - 12/92 Current 1

1

April lj, 1993 J

The Honorable Dick Swett United States House of Representatives l

Washington, DC 20515

Dear Congressman Swett:

I am responding to a letter of February 5,1993, from your constituent, Mr. H. Hamilton Chase, regarding his concern "for the general safety of our living area" with respect to conditions at the Vermont Yankee Nuclear Power Station (Vermont Yankee).

The Coalition has addressed its concerns regarding safety conditions at Vermont Yankee in several letters to the NRC during the last 6 months. The NRC has responded to the Coalition's concerns in letters dated November 2, 1992, and January 25,.1993. Copies of these letters are enclosed for your information. The NRC staff is currently preparing a response to the Coalition's most recent letter of February 9,1993. That letter shared Mr. Chase's concerns regarding publication of NRC criteria, the alleged failure of the NRC staff to address the full scope of the Coalition's -

concerns, and the request for public investigation of conditions at Vermont Yankee. We will send you a copy of our response to the Coalition when it is issued.

On May 20, 1993, Dr. Thomas Murley, Director of the Office of Nuclear Reactor Regulation, and Mr. Tim Martin, Regional Administrator, NRC Region I, will meet with the Vermont State Nuclear Advisory Panel. They will be prepared to discuss issues dealing with Vermont Yankee's performance.

Mr. Chase's statements regarding the relationship between the plant operator and the NRC, and his request for public investigation of the oversight practices in NRC Region I, have been provided to the NRC's Office of the Inspector General.

A copy of Mr. Chase's letter is enclosed for your convenience.

I trust that this information is responsive to your concerns.

Sincerely, a 2.,ned by Jaidii6Ii5y1or Executive Director for Operations

Enclosures:

1.

Letter dated November 2, 1992 1

2.

Letter dated January 25, 1993 3.

Letter from H. Chase dated February 5, 1993 q$nA i

Enclosure.1-p** **cy 0,

UNITED STATES

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See Attached Page November 2, 1992 o

CHAIRMAN Mr. Michael J. Daley New England Coalition on Nuclear Pollution, Inc.

Box 545 Brattleboro, Vermont 05302

Dear Mr. Daley:

On behalf of the Commission, I am responding to your letter of September 15, 1992, in which you stated that various operational practices and plant design features at the Vermont Yankee Nuclear Power Station require our immediate attention and action.

Your concerns relate to an event which occurred on January 13, 1992, in which the rupture disk on the inlet to the advanced off-gas (A0G) system ruptured, and to the operation and design of the turbine building roof exhaust fans. A brief background of the' event and the subsequent NRC review is provided in Enclosure 1.

Twelve separate concerns' raised in your letter are addressed in Enclosure 2.

The two-day " delay" in assessing the efficacy of an on-line repair to the A0G' rupture disc had no safety. impact on continued plant operation and'did not-violate operating procedures. Vermont-Yankee management imposed thettwo day period to assess whether on-line repair was a viable approach to the problem while the-interim repair remained in place. At the end of the two days, the i

Plant Operations Review Committee, weighing factors including as low as

~

reasonably achievable (Al. ARA), worker safety, and continued A0G' system reliability, concluded that on-line repair was~not viable.

Plant management then directed the plant to shut down.

The NRC did not find any immediate safety or regulatory issues that required the' plant to shut down.

The design purpose of the non-safety related rupture disc is to protect the downstream A0G system from overpressure transients, such as a hydrogen explosion. ~The 1

A0G system remained capable of fulfilling its function to minimize gaseous radioactive releases from the plant with the interim repair in place.

The NRC concluded that the event had minimal effect on plant workers and -the i

public. The main release from the January 13, 1992 event was approximately 1

six minutes in duration.

In January 1992, as part of the.NRC's followup to the event, an NRC radiation specialist reviewed Vermont' Yankee's Stack Release Event Report, and-the weekly turbine building roof vent effluent results.

The' stack release during the event indicated an estimated dose. rate of approxi-mately one-tenth the allowable technical. specification release rate limit.

i The' estimated dose from this event was approximately one-hundredth of one-percent of the. technical specification annual limit.

Although the NRC did not quantify the contribution of this particular event to the turbine building roof exhaust, the NRC reviewed the total turbine building roof effluent dose assessment calculations for -the weeks of. January 7, January 14 and

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January 21, 1992.

The.results for these weeks were 0.57 millirem, 0.59 millirem and 0.62 millirem, respectively. None of these projected dose' calculations exceed any regulatory limit. No significant; change in the total release between the week of the event (January 14) and the previous and subsequent weeks (January 7 and 21) was observed.

Further, independent inspections by the NRC and evaluation of Vermont Yankee's Semi-Annual Effluent reports for the first two quarters of'1992 indicate no substantive differences in reported releases between the months of January and February 1992.

Therefore, we have concluded that a public investigation of the radiological consequences of the January 13, 1992 event is not warranted.

Regarding the issue of the turbine building roof exhaust pathway, the NRC has conducted several inspections over the past eight years that address Vermont 1

Yankee's design and programs associated with radioactive effluent control. A continuous charcoal cartridge.and particulate filter radiation monitoring system was added to the turbine building roof vent pathway in October 1991.

Most recently, the NRC concluded that Vermont Yankee's November 1991 commitment to reroute the turbine building roof exhaust to the plant stack by i

Fall 1993 was acceptable.

The Commission appreciates your interest and efforts to ensure that we know of important safety matters and issues. We. share your. concern for potentially unmonitored release paths and unnecessary releases to the environment,'however small. We believe that our regular inspections of effluent ' controls and environmental monitoring programs, in conjunction with our other inspection activities, provide appropriate oversight at Vermont Yankee to ensure adequate protection of the public health and safety.

Sincerely, o

Ivan Selin

Enclosures:

===1.

Background===

2.

Response to Twelve Concerns r

i f

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e ENCLOSURE 1 BACKGROUND On January 13, 1992, a pressure transient within the advanced off-gas (A0G) system resulted in the rupturing of the A0G system rupture disk and the release of radioactive gases and particulates into the steam jet air ejector (SJAE) room of the turbine building (NRC Inspection Report (IR) 50-271/92-01 and LER 92-03).

The pressure transient was caused by a maintenance activity in which cooling water to the A0G system was inadvertently isolated.

As documented in IR 50-271/92-01, the NRC noted that approximately 4 minutes after the A0G rupture disk ruptured, licensed operators restored the A0G system to service.

Once the A0G system was restored to service, the system was operated at a negative pressure, and no additional release of gas to the SJAE room occurred.

Based on the radiation monitors in the stack, tht majority of the release lasted for approximately six minutes.

Vermont Yankee implemented a temporary repair of the A0G system rupture disk by placing a metal bucket over the ruptured disk.

This action minimized air leakage into the A0G system to help maintain a negative pressure in the A0G system.

Vermont Yankee management established a two-day time frame to asseu the option of performing an on-line repair of the A0G disk.

This asses m.nt was not completed by the self-imposed deadline of January 15, 1992. As a result, Vermont Yankee management directed the operators to shut down the plant to perform an off-line repair. The NRC determined that the licensee's actions were acceptable.

The two day " delay" in making repairs did not violate operating procedures in effect at the time.

The NRC concluded that the event had minimal effect on plant workers and the public based on NRC reviews of effluent releases from the plant during periods of time which bound the January 13th event.

In January 1992, as part of the NRC's followup to the event, an NRC radiation specialist reviewed Vermont Yankee's Stack Release Event Report, and the weekly turbine building roof vent effluent results.

The stack release during the event indicated an estimated dose rate of approximately one-tenth the allowable technical speci#.tcation release rate limit.

The estimated dose from this event was approximately one-hundredth of one percent of the technical specification annual limit.

Although the NRC staff did not quantify the contribution of this event to the turoine building roof exhaust, it reviewed the cumulative turbine building roof effluent dose calculations for the weeks of January 7 January 14 and January 21, 1992.

The results for these weeks were 0.57 millirem, 0.59 millirem and 0.62 millirem, respectively.

No significant change in the total release between the week of the event (January 14) and the previous and subsequent weeks (January 7 and 21) was observed.

i

- In July 1992, as part of a routine radiological effluent inspection (IR 50-271/92-15), the NRC reviewed the radiological monitoring results and projected dose assessments for the turbine building roof vent pathway considered (the ground level release point) from January 1992 to June 1992 and the available 1992 data for the semiannual effluent report.

Estimated dose for the turbine building roof exhaust pathway for the month of January, which includes any effects from the January 13th event, did not differ significantly from the following month's estimated dose.

The Vermont Yankee Effluent and Waste Disposal Semiannual Report for the first and second quarters of 1992 indicates that the total curies (C1) released from the elevated and ground level release points were within allowable limits.

i I

ENCLOSURE 2 RESPONSE TO 12 CONCERNS RAISED IN 9/15/92 LETTER FROM NEW ENGLAND COALITION ON NUCLEAR POLLUTION, INC.

1.

Vermont Yankee routinely operates the turbine building roof fans in contradiction of its Final Safety Analysis Report (FSAR).

The fans are operated year round contrary to the FSAR which indicates summer operation only. The fans are operated to reduce elevated levels of radioactive gases in the turbine building because of the leaky fuel.

In April 1991, while finding no safety concern, the NRC staff identified an inconsistency between Vermont Yankee's actual operating practice and its FSAR description regarding operation of the turbine building roof fans throughout the year (NRC IR 50-271/91-09).

In November 1991, the licensee issued Revision 9 to Vermont Yankee's FSAR in which it deleted reference to operation of the turbine building roof exhaust fans cnly during the summer.

The purpose of the fans is to remove heat from the turbine building; however, any coincidental removal of radioactive gases is monitored. Current operation of the system provides for operating the fans in an automatic mode. in which they cycle on at a temperature of 80 degrees Fahrenheit.

This does result in year-round operation when the plant is at power because of their location over the turbine.

However, the purpose of the fans is not to reduce airborne contamination due to leaking fuel.

If elevated radiation levels inside the turbine building are detected, the off-normal operating procedures require operators to manually secure the fans.

2.

Significant amounts of radioactive materials, released from the steam jet air ejector rupture disk, were vented directly to the environment through the unfiltered turbine building roof exhaust fans. The turbine building exhaust fans are a known pathway for uncontrolled radioactive releases.

During the January 1992 A0G disc rupture, the release pathway from the turbine building exhaust fans was monitored.

In October 1991, Vermont Yankee began operating a continuous charcoal cartridge and particulate filter system to quantify the total amount of radioactive materials (iodines, particulates, and tritium) released through this pathway.

In September 1991, Vermont Yankee revised its Offsite Dose Calculation Manual to incorporate a radioactive gaseous effluent monitoring program for this pathway.

In September 1991, the NRC used its mobile laboratory as part of its independent measurements program to verify Vermont Yankee's capability for analyzing radioactive effluents (IR 50-271/91-15).

The NRC confirmed that the licensee can accurately quantify radioactivity on charcoal cartridges and particulate filters from the turbine building exhaust sampling system.

In November 1991, Vermont Yankee committed to reroute the turbine building roof exhaust fans to discharge through the plant stack by the end of the refueling outage in 1993.

In July 1992, the NRC determined that the current effluent monitoring system for the turbine building exhaust fans was acceptable (IR 50-271/92-15) with regard to sampling the type of radioactive material that could be released pending the reroute of the exhaust in Fall 1993.

i

. 3.

The radioactive gas level in the reactor building was higher than that in the turbine building, this indicates that the release through the turbine roof exhaust fans contaminated the reactor building intake air supply. The release from the ruptured disk was of poor design because the gases migrated throughout the turbine building and eventually into the reactor building before apparently entering the plant exhaust.

Although the reactor building radioactive gas levels were increased to 0.19 maximum permissible concentration (MPC) after the A0G rupture disc ruptured, the release pathway was not through the turbine roof exhaust fans to the reactor building intake air supply.

The reactor building is maintained at a slightly negative pressure (vacuum) by design.

Therefore, air from the turbine building flows to the reactor building past the airlock door seals, which are not designed to be fully leak-tight.

The airlock allows air to flow into the reactor building when the airlock is operated to allow personnel to pass between the reactor and turbine buildings.

Following the January 13, 1992 event, the particulate concentration level in the reactor building increased primarily because air was transferred through the above described paths.

The transfer of particulates from the turbine building roof exhaust to the reactor building air intake was not considered as a source because particulates would be diluted when the turbine building roof exhaust is mixed with the atmosphere. Additionally, the turbine building roof exhaust radiation measurements did not indicate increased activity during this event, as documented in NRC IR 50-271/92-15, in which the NRC staff noted no significant changes in total emissions in January 1992 in relationship to following month of February.

4 Because the plant was operating with leaky fuel, workers were needlessly exposed to radiation levels as high as 38% of MPC (Maximum Permissible Concentration).

The radiation dose to workers from this event was negligible.

At the time of the release, no personnel were in the steam jet air ejector room where the release took place.

Personnel in the A0G building were immediately evacuated, and the whole body count of these individuals indicated no internal contamination.

Personnel exposures were minimized for activities that were necessary for assessing and implementing interim corrective actions.

It should be noted that total dose is determined with a calculation that includes both duration of exposure and exposure rate. For example, a person exposed to 38 percent of MPC for only 6 minutes would incur an exposure less than one millirem, which is significantly below Federal limits.

5.

Counts per minute (CPM) at the stack increased from 250 to 130,000 CPM indicating considerable exhaust activity that bypassed the charcoal filter.

Although the charcoal filter was bypassed when the A0G system was isolated and stack activity increased to 130,000 CPM, the duration was short.

The release rate during the event was less than one-tenth of the corresponding rate allowed by NRC regulations.

The duration of the release at the Vermont Yankee site on January 13, 1992, was not continuous but an instantaneous release (about 6 minutes), during which the peak release rate was 130,000 CPM.

The duration of the release wa one of the critical factors used to quantify the total release. The dose calculation to the public from the stack during this event was approximately one-thousandth of one millirem.

5.

Vermont Yankee's claim that no limits were exceeded was questionable since the release through the roof exhaust may not have been accounted for and that the flow path may not have adequately been understood.

The roof exhaust was accounted for, and the effluent pathways were understood.

In September 1991, Vermont Yankee revised its Offsite Dose Calculation Manual to incorporate a radioactive gaseous effluent monitoring program for the turbine roof exhaust pathway.

In July 1992, the NRC inspected 1he facility to review the adequacy of the licensee's, radioactive liquid and gaseous effluent control programs (NRC IR 50-271/92-15).

The NRC reviewed the measurement results of the weekly noble gas and particulate samples for the turbine building vent and main stack both before and after the incident; the measurements indicated no elevated radioactivity levels through the turbine building pathway because of this event.

Tables 18 and IC in the Vermont Yankee " Effluent and Waste Disposal Semiannual Report for the First and Second Quarters, 1992," document the total curies from the elevated and ground level releases.

7.

Since particulates were involved, has there been an adequate assessment of leaching of long term radioactivity from the roof and its long term effect on the environment?

Prior to this event, the licensee assessed the leaching (migrating) of radionuclides from the turbine building roof.

During an inspection conducted in March 1991, the NRC reviewed this assessment including soil and water sampling and measurement techniques (IR 50-271/91-09). The NRC found no impact on the public health and the environment. Based on the NRC staff's reviews of the weekly turbine building exhaust effluents during the January 13th event, the assessment of leaching of radionuclides remains adequate.

. 8.

Contrary to past practice, Vermont Yankee continued to operate for two days with leakage from the rupture disk past a metal bucket which was placed over the ruptured disc.

Previous A0G events were caused by a hydrogen explosion in the A0G system or an initiating event that also resulted in a plant trip.

This recent situation, in which a rupture disk ruptured because the A0G system was isolated, was unique.

The NRC found no immediate safety or regulatory issues that required the plant to be shut down.

The metal bucket was an interim repair to reduce air leakage into the A0G system.

Since the A0G system is operated at a vacuum, air flow by the bucket would otherwise be into the A0G system.

Therefore, the bucket was not intended to prevent gas releases out from the A0G system.

After the self-imposed two day period, Vermont Yankee management concluded that an on-line repair was not viable and shut the plant down.

9.

In private consultation the NRC forced management to place the plant off line and make repairs.

There was no private consultation between the NRC and Vermont Yankee.

The NRC did not force management to take the plant off line and make repairs. Vermont Yankee corporate management decided to take the plant off line to replace the A0G rupture disk, since an on-line repair was not acceptable due to ALARA and worker safety concerns, and in consideration of the continued reliability of the A0G system.

10.

The practice of leaving the turbine building door open in all seasons is questionable.

The staff reviewed the effect of opening the turbine building truck bay door (IR 50-271/90-13).

The results indicated that no technical specification requirements were violated. The licensee committed to obtain quantitative measurements to verify that the air flow,was into the turbine building.

The licensee completed this review.

A preliminary review of the licensee's assessment by the NRC's resident inspector indicated that normal turbine building ventilation system operation, with the turbine building truck bay door opened, ensures an appropriate inward air flow.

11.

The exhaust fans have no automatic emergency stop system to ensure that this dirset pathway is sealed during releases.

Radiation levels inside the turbine building are continuously monitored, and an abnormal level will actuate an alarm in the control room.

if an alarm occurs, operators following an approved off-normal operating procedure are required to secure the turbine building roof exhaust fans using the manual emergency stop switch located on the heating and ventilation control room panel and close the turbine building door.

. 12.

Operation of the plant with leaky fuel is in conflict with ALARA.

Operations of the plant with leaking fuel led to slightly elevated exposure levels.

The As low As Reasonably Achievable (ALARA) program is based on minimizing collective exposure to the extent reasonable; Vermont Yankee's operation with leaking fuel was not in conflict with this program. Other measures also contribute to an effective ALARA program, including minimizing steam leaks, core management, and enhanced monitoring.

These measures were incorporated in the Vermont Yankee Failed Fuel Action Plan (FFAP), which was reviewed by the staff and determined acceptable (IR 50-271/91-29).

The FFAP minimized the impact of failed fuel on maintenance.

Fuel performance since the March 1992 refueling indicates no failed fuel in the operating core.

.c New England Coalition on Nuclear Pollution,Inc.

Box 543. Brattleboro. Vermont 03302 Phone 602)23 0330 September 15, 1992 Ivan Selin, Chairman U.S.

Nuclear Regulatory Commission Washington, DC

Dear Chairman Selin:

We wish to alert you to a situation at the Vermont Yankee Nuclear Power Station that requires your immediate attention and corrective action.

Vermont Yankee routinely operates turbine building roof ex-haust fans in contradiction of its Final Safety Analysis Report (FSAR).

This practice may have resulted in an in-adequately monitored release of radioactive materials (par-ticulate and gaseous) to the environment during an incident on January 13, 1992 (see the Feb. 21,'92 NRC inspection report No.50-271/92-01, hereafter NRC report,and Licensee Event Report LER 92-003: A0G RUPTURE DISC-TEMPORARY REPAIR NOT WITHIN SYSTEM DESIGN BASES Feb. 13,'92).

On Monday, January 13, 1992 at approximately 1:30 PM, with the reactor at 100% power, maintenance personnel in-advertently shut off the Advanced Off-Gas System (A0G) which caused a rapid build-up of pressure bursting the steam jet air ejector rupture disc and releasing radioactive gases and particles into the turbine building.

Because of poor design, the release from the rupture disc was not immediately contained.

Instead, the gases and par-ticles migrated throughout the turbine building and eventually into the reactor building, before apparently en-tering the plant exhaust.

Because the plant was operating with leaky fuel, workers were needlessly exposed to radiation levels as high as 38%

of Maximum Permissible Concentrations during this incident.

This is the second fuel cycle in a row that management has decided to operate the plant with leaky fuel, a practice in j

conflict with keeping radiation doses as low as reasonably achievable.

In a July 1990 letter, plant workers warned your Staff about this potentially dangerous condition'at the plant.

Concentrationsaof radioactive gases and particles in this uncontrolled release were at least ten times higher j

than they would have been under normal operating conditions.

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NECNP-2 counts per minutes at the stack went from a pre-event level l of 250 cpm to 130,000 cpm (NRC report), indicating consider-able exhaust activity that by-passed the charcoal filter i system and catalytic converter. These systems are designed to reduce radiation levels of reactor off-gas by a factor of 10,000 before release to the environment. In addition to this monitored and recognized release path-way, we suspect that significant amounts of the radioactive materials released from the steam jet air ejector rupture disc were vented directly to the environment through the un-filtered turbine building roof exhaust fans. These exhaust fans have no automatic emergency stop system to assure that this direct pathway is sealed during releases of this type (FSAR 10.12-4 and Fig 10.12). Considering the close proximity of the plant to the Vernon Elementary school (within 540 yards), and the presence of particulates in the release, it is surprising to find no discussion of this pathway in either NRC or Vermont Yankee reports. Yet there is every reason to believe these fans were in op-eration during this incident. Year-round operation has be-come common practice in recent years to reduce elevated levels of radioactive gases in the turbine building because of the leaky fuel. This practice contradicts the plant FSAR which indicates summer operation only. Complaints from con-cerned workers had alerted Vermont Yankee management and your Staff to this questionable practice, as well as to the practice of leaving turbine building doors open in all sea-sons. The turbine building exhaust fans are a known pathway for uncontrolled radioactive releases. This pathway was the subject of considerable discussion 9 years ago in associa-tion with contamination of Connecticut River sediment with Co-60 (see Memo for Darrell Eisenhut from Richard W. Starostecki re: LOW LEVEL RADIOACTIVE EMISSIONS FROM BVR TURBINE ROOF VENTS 11/7/83). Further, the NRC report notes particulate levels in the reactor building higher than levels measured in the turbine building where the release took place. The buildings are divided by airlocks and have separate ventilation systems. Neither NRC nor Vermont Yankee reports explain how the release crossed the boundary between the buildings. How-ever, in a past incident freon released on the roof entered the control room air supply (LER BS-012-00: CONTROL ROOM HABITABILITY SYSTEM ACTIVATION). Did releases via the tur-bine building roof exhaust fans contaminate the reactor building intake air supply? Vermont Yankee claims that the releases associated with this

.o ~ NECNP-3 incident didn't exceed any limits (LER 92-003, NRC report). Ve seriously question the reliability of this claim, given the nature of the release and the existence of the roof ex-haust pathway. Specifically, were Radiation Protection or Health Physics personnel aware of the beyond-FSAR practice of operating the turbine building roof exhaust fans? Did they adequately understand the probable flow paths of the release from the rupture disc? Since particulates were in-volved, has Vermont Yankee adequately assessed the long-term leaching of radioactivity off the roof and its effect on the environment and the public. The NRC report notes only that "The surveys taken appear to be properly documented and of adequate detail to assess conditions within the plant." (em-phasis added). Within six minutes of the discovery of the rupture, the A0G vas restored to service and leakage out the ruptured disc was minimized by placing a metal bucket over it.

However, the plant then continued operation for two days instead of immediately shutting down to repair the ruptured disc.

The steam jet air ejector rupture disc has burst many times in the past twenty years. Until this incident, the plant has never continued to operate or attempted an on-line repair. Records from 1973 demonstrate that the plant moved to shut down within 5-7 hours of discovery of a steam jet air ejector rupture disc rupture (Abnormal Occurrence Nos. A0-73-27, A0-73-26, A0-73-25). Yet, plant management deliberated for two days about whether an on-line repair could be done, or if they could continue operations in a degraded mode (ie. with a bucket over the leak). Meanwhile, control room operators were required to perform constant surveillance of the degraded A0G to prevent further uncontrolled releases (LER 92-003: A0G RUPTURE DISC TEMPORARY REPAIR NOT WITHIN SYSTEM DESIGN BASES Feb. 13,'92). We understand that management had decided it was possible to continue operations in the degraded mode until your Staff, in private consultation at Corporate Offices, forced manage-ment to take the plant off-line and make repairs. Given these facts and uncertainties, we feel that Vermont Yankee operated in a manner that recklessly endangered workers, and represented a significant increased risk to the public. We require you to:

1. Determine why there was a two day delay in making repairs and whether this was a violation of operating procedures in force at the time.

If so, to identify those responsible and carry out disciplinary action.

s 4 NECNP-4 2. Conduct a public investigation of the release to determine the full extent of worker and public exposure, and to make available (with supporting documents): total curies released, radiation doses to workers and public including pathway analysis, and stack monitor data for the period just before and twelve hours after the event.

3. Require Vermont Yankee to accelerate its plans to isolate all turbine building exhaust points and vent them into a filtered and monitored system.

4 Conduct a plant-wide design study to identify any other inadequately filtered and monitored pathways to the environ-ment and require Vermont Yankee to correct these flaws. We look forward to your prompt attention to this situation. 1 Sincer,ely, g-l . /. h' n i ,.. ~ Michael J. Daley for Board 1 of the NeJ' England CDalition on Nuclear Pollution cc. Vermont Yankee Governor Howard Dean media l 4

[ p ena, u o, tJNITED STATES y' g NUCLEAR REGULATORY COMMISSION 24 i" g REGION I ^ tH 475 ALLENDALE ROAD q.s,,/ xiNo or PRusslA. PENNSYLVANIA 19406 1415 x Docket No. 50-271 Mr. Michael Daley New England Coalition on Nuclear Pollution, Inc. Box 545 Brattleboro, Vermont 05302

Dear Mr. Daley:

I am responding to your letter to Chairman Selin dated December 16, 1992, in which you expressed concern that " budgetary pressures" are causing " systematic degradation of safety related components at the Vermont Yankee plant." Additionally, you expressed concern that NRC oversight has neither identi6ed nor addressed this issue, in part, due to limited NRC resources. The NRC is concerned with events, such as those cited in your letter. Such events, individually and collectively, have been and continue to be critically reviewed. However, the NRC's independent review of events at the Vermont Yankee Nuclear Power Station (VYNPS) does not support your assertion that a " systematic degradation of safety-related components" is occurring. The NRC previously reviewed each of your cited examples; however, no apparent connection between the examples cited and any degradation of equipment caused by neglect from budgetary pressures was found. These events were attributed to other causes, including procedural deficiencies, personnel errors, manufacturing defects, design issues and ineffective corrective actions. Further, these events as an aggregate are not " extraordinary" or " pervasive," considering frequency and safety signi6cance. Your characterization of certain events warrants elucidation; specifically a feedwater check valve (FDW-96A), two intermediate range monitors (IRMs), and two average power range monitors (APRMs) are cited as examples where equipment remained deficient for extended periods without appropriate resources applied to correct the deficiency. In your letter, you stated, "This (cost cutting), in turn, leads to non-conservative judgements about running the plant with equipment in a degraded mode like the tolerance, for over six years. of leakage in feedwater check valve 96A." The licensee is required to test this valve every refueling outage to ensure that leakage is within limits before plant startup. As documented in NRC Inspection Repon 50-271/89-02 (IR 89-02), FDW-96A failed leakage tests during the 1983,1984.1985 and 1989 refueling outages. Although the licensee repaired this vahe after every test failure, and before the subsequent plant startups, the NRC concluded in IR 89-02 that the licensee's corrective actions to prevent recurrence were in fact ineffective. During the 1990 refueling outage, after another failure, the licensee reached a root cause determination and replaced the valve's clastomeric seating material with stellite. FDW-96A f

Mr. Michael Daley 2 passed its leak test during the subsequent refueling outage in 1992. The licensee's frequent repair of this valve, although indicative of earlier ineffective corrective action, was not indicative of tolerance of a degraded condition. Another example, which you cited, was that the plant operated for six months with the "E" and "F" iRM channels inoperable. Although the "E" and "F" IRM channels became inoperable during the April 1992 startup, the remaining four IRMs were operable and met the minimum number required by technical specification. Each of the redundant reactor protection system trains has one more IRM channel than necessary. This allows one bypassed channel per train. Further, when the plant reaches about 40 percent power, the IRMs are withdrawn from the core and no longer provide protective signals. Therefore, it is reasonable for the licensee to wait until the next refueling outage to repair the IRMs and avoid unnecessary personnel exposure. This current status of the IRMs does not present a safety problem, and a reasonable long-term solution is bemg pursued. You expressed another concern with two unshared APRM channels bypassed. The licensee operated with two shared, not unshared, APRM channels bypassed. Similar to the IRM channels. VYNPS Technical Specifications allow these channels to be bypassed. The licensee placed these APRM channels in bypass to minimize spurious actuation caused by local power range monitors (LPRMs), which provide an input signal to the APRM channels. These spunous actuations. which do not affect the APRM operability, occurred after the licensee began upgrading the LPRMs and have also been observed at other plants with similar upgrades. The hcensee implemented a vendor-developed solution to these spurious actions, and currently has the two channels still in bypass, while they monitor their effectiveness. The licensee's bypass of APRM channels does not present a safety problem, and a reasonable long-term solution is being pursued. Regarding your concern about NRC oversight, the NRC has sufficient resources to address safety issues at nuclear power plants. Problems with licensee performance are identified and addressed through our inspection program and our periodic assessments of plant performance. NRC inspections, primarily performance-based, frequently focus on the licensee's effectiveness in identifying deficiencies, determining underlying problems, analyzing root causes, and correcting deficiencies. Basic inspection activities, " core" inspections, confirm with reasonable assurance that the health and safety of the public are maintained. Additionally, discretionary mspection activities, " initiative" inspections, primarily focus upon suspect programmatic weaknesses. NRC managers and inspectors, periodically and collectively, conduct assessments of plant performance through the Systematic Assessment of Licensee Performance (SALP) process, and plant performance reviews. During the SALP process, the NRC reviews licensee activities. Licensee Event Repons and inspection findings to ascertain the licensee's performance and to detect underlying problems or trends. In the recent SALP, we noted a decline in performance in three areas: however, overall performance at VYNPS was good. More frequently, plant performance reviews and briefings are conducted by NRC management, as well as inspectors.

Mr. Michael Daley 3 As a regulatory agency, the NRC's oversight of plants and imposition of requirements must not be arbitrary. To operate in this manner would be detrimental to our objective of public health and safety in part, because it may divert the focus from safety significant issues to those which are not. Consequently, the NRC's revisions of requirements and use of enforcement discretion enhance our effectiveness in assuring public health and safety, without diverting resources toward activities that provide minimal or no safety benefit. The NRC has carefully considered all of the events that you have cited in your letter. None of the events or problems cited in your letter, which span essentially the last four years, are new to the NRC. Moreover, none of the ir. formation presented in your letter, either individually or collectively, would cause the NRC to alter the assessments reached in our previous SALP reports. With the absence of any new information or insights into plant performance, a public insestigation, such as the one that you suggested, would not be warranted. I thank you for your interest, and trust that your main concerns were addressed. If you have additional questions regarding events or NRC inspections at VYNPS, you may wish to contact Mr. Gene Kelly of my staff at 215-337-5183. Sincerely, Y gQ ~ Thomas T. Martin Regional Administrator cc: Public Document Room (PDR) Local Public Document Room (LPDR)' Nuclear Safety information Center (NSIC) State of Vermont Commonwealth of Massachusetts (2)

r (,

• New England Coalition on Nuclear Pollution,Inc.

i Box 545, Brattleboro, Vermont 05302 Phone (802) 257 0336 December 16, 1992 Ivan Selin, Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555 i

Dear Chairman Selin:

We have received your response to our letter of September 15, 1992 raising concerns about the rupture of the steam jet air ejector rupture disc and subsequent release of radiation. We appreciate your. detailed reply to those concerns and are preparing our response. The purpose of this letter, however, is to alert you to an alarming and potentially disastrous situation:. systematic degradation of safety components at the Vermont Yankee plant. Because of an ongoing pattern of industry mismanagement and regulatory neglect, Michael Mulligan resigned his position as a control room operator for the Vermont Yankee nuclear plant, and joined our organization. In July, 1990, after many months of repeated efforts to i bring important safety problems to the attentionuof plant management and NRC officials failed to produce substantive change, Mr. Mulligan wrote the Coalition an anonymous letter [ regar{ing four major areas of concern at the plant: gashed fuel

pins, spent fuel pool cooling, shift staffing, and nocturnal burning of waste oil.

He sent similar letters to the NRC and to the State of Vermont Nuclear Engineer, who eventually made the letter public at the request of then Governor Madeline Kunin. Shortly thereafter, Mr. Mulligan contacted us by telephone still anonymously -- and made it clear that his primary I concern was much larger. The plant, he told us, was headed for catastrophe, because the management was shortsightedly focusing on the bottom line, and because the NRC has failed to recognize the consequences of that shortsightedness. The only hope, he told us, was to open Vermont Yankee's operations to the light of day. Only public demands for safer operation would force management to allocate more resources on safety concerrs. The issues Mr. Mulligan has brought to our attention have 1. Gashed pins increase off gas levels and radiation doses to the public and especially to plant workers. Mike and other Vermont Yankee workers were especially concerned about their in-creased exposure. ( jh( I2/ k k 1 Educating Ihe Pu b lic in Clean Alternatives to Nuclear Po w e r

been profoundly shocking to us. We believe they go straight to the core of the problems confronting this industry. The Coalition has always known Vermont Yankee's design was vulnerable to severe accidents, and that the plant participated in the broadly unacceptable risks inherent in the use of nuclear technology, but we must admit to having taken a margin of comfort in the notion that Vermont Yankee was one of the better run utilities. However, we no longer enjoy the comfort of that illusion. By themselves, the circumstances surrounding Mr. Mulligan's resignation are an indication of an extraordinary situation at the plant. Combined with an investigation of the issues he has brought to our attention, the situation proves to be nothing short of alarming. We have checked and verified each of Mr. Mulligan's reports in NRC documents, not because we doubted his story, but because we knew others would. There is simply no question that a variety of major safety systems at Vermont Yankee have had substantial difficulties during the past four years. We list them here, with accompanying footnotes documenting the equipment failures. Each document is available in the NRC public document room; most can also be found on the NUDOCS computerized document system.

1) emergency core cooling system (ECCS),I, which includes:

the gore spray system,' HPCI ang the RCIC system; 1. Licensee Event Report LER 89-015-00 (hereafter, simply LER): " Spurious Relay Actuation Caused ECCS Initiation Signal Due to Lack of Procedure for Reenergizing Local Instrument Cabinet." Event Date: 3/10/89.

2. LER 89-015--00 and LER 89-015-01: " Primary Containment Leak Rate Test Caused Inadvertent Core Spray and RHR Pump Start Due to Inadequate Procedure." Event Date: 3/30/89.

3. LER 91-007-00: "HPCI Declared Inoperable Due to Flow Con-troller Set Point Drift," Event Date: 3/13/91. See

also, LER 92-004-00: "High Pressure Coolant Injection System Inoperable Due to Degradation of Station Battery Bus Voltage Caused by Failed Battery Charger Component." Event Date: 2/20/92.

4. LER 89-014-00: " Reactor Core Isolation Cooling System Inoperable Due to Motor Burn Out on RCIC-21 Valve," Event Date: 7/18/89; and LER 92-015-00: " Reactor Core Isolation Cooling System Inoperable Due to Flow Controller Setpoint Drift," Event Date: 4/24/92. 2

I l l I

2) ~AC and DC power systems which include:

~ !! the emergency diesel generators,2 1 1. LER 90-009-00: " Inadvertent Reactor Scram Due to a Short Circuit on the Vital AC Bus as a Result of Personnel Error," Event Date: 6/1/90. Also, LER 88-012-00: " Overloaded Power Supply in Vital Fire Protection Control Panels," Event Date: 9/28/88; LER 90-008-00: " Failure to Meet Separation Criteria for Power Cables to Regulatory Guide 1.97 Instrumentation Loops," Event Date: 5/29/9C. See also, LER 89-009-00: " Lack of Redundan-cy in Residual Heat Removal Service Water Systems," Event Date: 5/28/89. 2. LER 90-010-00: " Failure to Meet Technical Specifications for Diesel Generator Operation Readiness Test," Event Date: 8/16/90; LER 90-010-01 and LER 90-010-02: " Failure to Meet Tech-nical Specifications for Diesel Testing Generator," Event Date: 8/15/90; and LER 91-012-00 and LER 91-012-01: Reduced Cooling Water Flow to Diesel Generator Heat Exchangers and Station Serv-ice Air Compressors Due to High Service Water System Backpressure Caused by Weak Design." Event Date: 4/23/91. Also, Harold Eichen-

holz, Thomas G. Hiltz, and Richard S.

Barkley: " Inspection Report 50-271/91-19," Section 4.2.1: "'A' Emergency Diesel Generator Fuel 011 Transfer Pump Operability" and Section 4.2.2: "'B' Emergency Diesel Generator Failure to Start." These events took place on July 25 and July 25, 1991 respectively.Also. H. Eichenholz and P.

Harris,

" Inspection Report 50-271/92-06," Section 4.2.2: "'B' EDG Maintenance Associated with the ECCS Tests:" "The April 5 test was not successful because the "B" EDG failed to start, due to incom-plete resetting of the diesel governor shutdown plunger following the last operation of the diesel on April 3.... Vermont Yankee preliminar11y determined that the root cause for the first failure the advanced age of the "B" diesel generator... However, based was on satisfactory performance during surveillance testing, and in

part, due to unavailability of parts, VY was reasonably assured that the "B" EDG governor would continue to perform its safety function until its scheduled replacement in May, 1992."

On June 3, 1992, Vermont Yankee submitted to NRC a " Request for Temporary Waiver of Compliance from Technical Specification LCO Requirements Pertaining to Emergency Diesel Generator " (BV 92-058). Technical Specification 3.5.H.1 requires that "During any period when one of the standby diesel generators is inoperable, continued reactor operation is permissible only during the succeed-ing seven days " On June 4, 1992, Charles W. Hehl, Director of the Division of Reactor Projects at the NRC granted a " Temporary Waiver of Compliance" allowing Vermont Yankee to run an additional 24 hours without a safety backup diesel generator. 3

-- emergency battery systems I batt'5' 'h*'8' "'**** ' -- emergency -- switchyard bus -- and relays

3) residual heat removal (RHR) systems, whichinegudes:

-- RHR service water systems and pumps 1. LER 89-020-00: " Removal of a Technical Specification Sur-reillance Requirement from a Procedure Due to an Inadequate Technical Specification Review," Event-Date: 8/11/89: "On 8/11/99, with the plant at 100% power, Vermont Yankee discovered the procedure controlling battery maintenance and testing was not consistent with Technical Specification requirements." 2. LER 92-004-00: "High Pressure Coolant Inj ec tion System Inoperable Due to Degradation of Station Battery Bus Voltage Caused by Failed Battery Charger Component." Event Date: 2/20/92.

3. LER 91-005-00: " Reactor Scram Due to Mechanical Failure of 345 kV Switchyard Bus Caused by Broken High Voltage Insulator Stack," Event Date: 3/13/91; LER 91-009-00: " Reactor Scram Due to Loss of Normal Off-Site Power (LNP) Caused by Inadequate Proce-dure Guideline," Event Date: 4/23/91; and LER 91-014-00: " Reactor Scram Due to Loss of 345 kV Switchyard Caused by Defective Off-site Carrier Equipment," Event Date: 6/15/91.

See also. NRC Information Notice 91-81: " Switchyard Problems that Contribute to Loss of Offsite Power," December 16, 1991. See

also, LER 87-003-00 and LER 87-008-01: " Loss of Normal Power During Shut-down Due to Routing All Off site Power Sources Through One Break-er," Event Date: S/17/87.

4. LER 92-012-00: " Degraded Grid Undervoltage Relays Found Below Technical Specifications Limits," Event Date: 3/31/92.

Also, LER 91-010-00: " Failed Relay Coil Results in Primary Con-tainment Isolation Sy: tem Actuation," Event Date: 4/12/91.

5. LER 89-009-00: " Lack of Redundancy in Residual Heat Remov-al Service Water Systems," Event Date: 5/2S/89; LER 91-005-00, LER 91-005-01, and LER 91-005-02: " Loss of [RHR) 'B' Loop Shut-down Cooling Due to Pressure Switch Activation," Event Date: 3/14/91: "On 3/14/91 at 0450 hours, with reactor vessel cooldown in progress following a reactor scram on 3/13/91... and with the "B" loop Residual Heat Removal (RHR) (BO*) System flushed and lined up for Shutdown cooling, a Group 4 Primary Containment Isolation Signal (PCIS) (JM)* was received during two attempted starts of the "B" RHR pump and closure of Shutdown Cooling Suc-tion Isolation valves. Also, LER 89-023-00: Failure to Perform Daily Instrument Checks on the Low Pressure Coolant Injection System Crosstie Monitor Due to Interpretation of Tech. Spec. Requirements", Event Date: 9/11/89: " Vermont Yankee Technical l Specification 4.2.A. Table 4.2.1, requires an instrument check of l 4

4) feedwater systeml and. heck valves 2 ; and

5) service water system check valves 3,

... Continued... the indication for the residual heat removal (RHR) system crosstie valve, RHR-20, be completed'once per day. Contrary to this requirement, it was discovered, on 9/11/89, that the indica-tion to the valve had not been available from 3/20/89, when the power supply breaker to the indication was removed...."

Also, LER 91-015-00: " Containment Isolation Valve Failure to Close Due to Erosion / Corrosion and Displacement of Screw-in Seat,"

Event Date: 5/14/91: "On June 14, 1991 Residual Heat Removal Valve V10-34A Failed to Close." 1. LER 88-007-00: Main Turbine Trip and Reactor Scram from Feedwater Flow Controller Malfunction Due to Failed Feedwater Flow Integrator," Event Date: 5/18/88. 2. LER 92-010-00: "1992 Appendir J Type B and C Failure Due to Seat Leakage," Event Date: 3/8/92. "On 3/8/92, 3/12/92, and 3/17/92 Liquid Radwaste Valve LRW -83 (EIIS=WD), Feedwater Check Valve FDW-28B (EIIS=SJ) and Control Rod Valves CRD-413A and 413-B (EIIS=AA) were found to have seat leakage above that per-mitted by Technical Specification 3.7.A.4." LER 90-012-01: "1990 Appendir J_ Type B and C Failure Due to Seat Leakage," Event Date: 9/3/90: "On 9/3/90 and 9/5/90 Feedwater Check valve FDW-95A (EIIS=SJ) and Primary Containment Atmospheric Control valve PCAC-5B (EIIS=BB) were found to have seat leakage above that permitted by Technical Specification 3.7.A.4." r LER 89-007-00: "1989 Appendir J Type B and C Failure Due to Seat Leakage," Event Date: 2/15/89: "On 2/15/89, 2/17/89, 3/5/89 and 3/7/89 ... Liquid Radwaste Valves LRW-83, LRW-94, LRW-95 (EIIS=WK), Primary Containment Atmospheric Control valve. PCAC-8,9,10,23 and PCAC-5,7,6A,7A 7B (EIIS=BB),' Containment Air Com-pressor Discharge Check Valve CA-89C (EIIS = LD) and Feedwater Check valve FDW-95A (EIIS=SJ) were found to have seat. leakage above that permitted by Technical Specification 3.7.A.4." LER 84-011-01 and 84-011-02: " Update on Leaking Containment Isolation Valves," Event Date: 5/15/84: FDW-95A and CA-89C ... were found to have seat leakage above that permitted by Technical Specification 3.7.A.4." j IT SHOULD BE NOTED THAT THE SAME FEEDWATER CHECK VALVE FDW 95A -- WAS REPORTED LEAKING FOR AT LEAST $ YEARS. FROM 1984 THROUGH 1990. 3. LER 89-017-00: " Service Water Check Valves Inoperable 'Due to Corrosion of Internal Parts " Event Date: 3/30/89. 5

In addition, there have been equipment problems in other key areas during the levelindicators,{astand the diesel fire pump,g example, four years as well: fo core water among others. As though all of this weren't enough, maj or questions have been raised during this same geriod about personnel training programs and plant procedures, about the plant's emergency 1. LER 92-014-00: " Inadvertent Scram and ECCS Initiation While Shutdown When Restoring Four Level Transmitters to Service," Event Date: 4/12/92. Also, letter from Ernest C. Hadley, attorney for We the People, Inc. to Ivan Selin, July 21,

1992, concerning generic problems with water level instrumenta-tion at U.S. nuclear reactors.

2. LER 91-003-00: " Missed Diesel Fire Pump Fuel Oil Surveil-lance Due to Inadequate Procedure," Event Date: 2/27/91. 3.

Williams, J.H.; Conte, R.J.

1 Bettehausen, L. " Training Program Inspection Report 50-271/91-82 on 911021-25. Deficien-cies noted...." Also LER 89-013-00 and LER 89-013-01: " Reactor Vessel Inventory Decrease Due to Personnel Error," Event Date: 3/10/S9. PNO-I-89-021, a notice of unusual event, covers the same 5aother set of events due to incorrect procedures is de-scribed in LER SS-001-00 and LER 88-001-01: " Plant Service Water Effluent Stream Not Monitored Due to Procedure Deficiency," Event Date: 2/11/88, Inspection Report 50-271/88-03 and Notice of Violation from [the same] Inspection Report," and LER 88-014-00: " Missed Effluent Sample Due to Inadequate Corrective Action in LER 88-01, Rev. 1," Event Date: 10/19/88. Other reports triggered by incorrect procedure include: LER 89-24-00 and 89-24-01: " Missed Residual Heat Removal Valve Leak-age Surveillance Due to Incomplete Procedure Review," Event Date: 9/13/39; LER 90-018-00: " Primary Containment Isolation System Spurious Actuation Due to an Inadequate Procedure " Event Date: 10/10/90; and LER 92-015-00: " Improper Inservice Flow Testina of the Control Room Chilled Water Pump Due to ASME Code Misinterpre-tation and Subsequent Missed Quarterly Test Due to Incorrectly Following the Surveillance Procedure," Event Date: 4/22/92. See also, LER 89-015-00. LER 89-015-00 LER 89-015-01, LER 89-020-00, LER 89-023-00, LER 91-003-00, and LER 92-014-00, all of which are cited above. 5

b e operating procedures (EOPs)l, and about plant security 2, Problems with training and security were identified, in part, with a lack of adequate funding. We do not intend to detail in this letter each of the problems we have just enumerated: they are already well documented. Instead, we want to point to the extraordinary and pervasive pattern of these shortcomings. It may be true that no one of these shortcomings, by itself, constitutes an adequate reason to challenge the ongoing operation of this plant. But when they are combined as they have been here, the possibility is raised that disastrous results could ensue. Each of these malfunctions and system degradations has already been brought to the attention both of management and of the NRC. The question then arises: why has the systematic degeneration of this plant been allowed to continue? Why has this not been corrected? We can see no other explanation for this than that utility decision-making is unduly driven by the bottom line and that your staff is in some way acquiescing in this state of affairs. In recent years, plant workers and mid-level management alike have been keenly aware of subtle and not-so-subtle messages from top management that maintaining or improving the plant's capacity factor - acknowledged to be one of the highest in the industry -- is far more important than resolving safety issues. In its day-to-day scramble to produce more electricity and therefore higher profits, plant management has created an atmosphere that causes employees to think twice before raising safety concerns that might j eopardize corporate financial goals. For instance, the plant j ust recently shut down because of problems with a recirculation pump controller. This pump controller has experienced chronic problems and its erratic behavior has been of considerable concern to on-duty operating personnel. Yet time after time the utility has attempted a quick fix. A look at the maintenance history of this piece of equipment would reveal a resistance to carrying out a thorough troubleshooting that might lead to unwelcome down time. 1. Bennett, F.P.; Conte, R.J. & Bettehausen, L. " Inspection Report

1. 50-271/92-80: Emergency Operating Procedures inspection 50-271/92-80, on 920224-28.

Weaknesses and deficiencies noted...."

2. Initial Systematic Assessment of Licensee Performance Report No. 50-271/91-99.

October 13,1992. 7-

Cost cutting and cost containment activities lead to subtle interactions that impact plant operations, auch as tight inventory control resulting in the unavailability of parts. As noted above, this occurred with the fuel pool motor, and the governor on the "B" emergency diesel generator. This, in turn, leads to non-conservative j udgments about running the plant with equipment in a degraded mode like the tolerance, for over six years, of leakage in feedwater check valve 95A I These activities are multiplying just as many of the plant's key components are feeling the effects of age related degradation. Thus, rather than improving plant safety through increased vigilance, management is moving the plant in the opposite direction. For example, it has pushed hard to reduce the time spent for planned outages. This means that the plant can make more money (since even scheduled plant shutdowns are expensive), but it also reduces the amount of time and resources available for fixing maj or saf ety systems. In large measure, the degradation of the switchyard equipment appears to stem from lack of time and resources during the shortened outages to perform necessary testing and maintenance. Increasing the fuel cycle from 12 months to 18 months generates more profits, but it also increases the strain on the system. Plant components work harder and longer, with less frequently scheduled maintenance. To maintain short outage times, the utility has begun to shift various maintenance activities normally performed during an outage into periods when the plant is operating at full power. This practice has had unsettling consequences, resulting in a reactor scram that seriously challenged safety eqgipment 2, and a release of radiation into the environment Proposed new NRC regulations would actually reduce the plant's accountability by extending reporting periods to match

1. LER 84-011-01 and 84-011-02: " Update on Leaking Containment Isolation Valves," Event date: 5-15-84:

"...FDW 95A and S9C...were found to have seat leakage above that permitted by Technical Specification 3.7.A.4." reports continuing into 1990. 2.LER 91-009-00: " Reactor Scram Due to Loss of Normal Off-Site Power (LNP) Caused by Inadequate Procedure Guideline," event date 4/23/91. 3.LER 92-003: "A0G Rupture Disc Temporary Repair Not Within System Design Basis" February 13, 1992 and NECNP letter'to Chairman Selin date September 15, 1992. 8

the longer cycles I, at a time when aging plants clearly require greater regulatory scrutiny. Informed of all of the system failures noted above, the NRC has imposed no fines and no shutdowns, and Staff regulatory practice seems focused on symptoms without any real understanding of the underlying pattern leading to the failures on such a wide scale. On-site NRC inspectors, informed of ongoing uncorrected conditions with potential safety implications respond to employees by calling for the utility to "self-correct" and for employees to submit more " maintenance requests" (MRs). Yet the utility's internal practice for handling MRs allows a screening of requests before they are actually logged onto the official computerized tracking system. Given the atmosphere we have been describing, it is unreasonable for your Staff to assume that this screening is performed with safety considerations as the prime criterion. Your staff'. inability, or unwillingness, to identify the pattern describeu here is part of a structural weakness in the oversight program. Because of limited resources, NRC must focus on individual problems and their resolution, leaving inspectors too little time to explore underlying causes. Officials from the NRC Region I inspection branch told members of the Vermont State Nuclear Advisory Panel as much at a December 2, 1992 meeting. Regional Supervisor E. Kelley spoke of the difficult " art" of allocating limited personnel and resources L to the twenty nuclear plants in the region. Senior Vermont Yankee Resident Inspector Harold Eichenholz and his partner, Paul Harris, mentioned a high reliance on the utility's ability to identify and correct its own problems because with only two inspectors on site, they must " choose and prioritize" the issues i they follow. The very nature of the problem we are describing would not be amenable to either self-identification or self-correction. In addition, the recent SALP Report identified deficiencies in Vermont Yankee's self-assessment and engineering evaluations, concluding that " Performance declines [three deratings out of seven SALP categories] attributed to the failure of self-assessment programs to effectively identify i 1. " Reducing the Regulatory Burden on Nuclear Licensees," Proposed Rule RIN 3150-AE 30, Federal Register, June 18,

1992, pp.

27187-27191 and " Review of Reactor Licensee Reporting Re-quirements," Federal Register, June 19, 1992, pp. 27394-5. 9

o f undamental issues in maj or program areas"I. l In addition to having limited resources in the field, your agency has no policy to determine when the type of systematic failure we are describing sufficientig jeopardizes public safety to warrant the shut down of the plant This worries us. Experience has taught us that simply bringing these matters.to the attention of your Staff will not lead to action to counteract these trends at Vermont Yankee. Many NRC decisions have, in fact, served to reinforce Vermont Yankee's misguided activities by relying too heavily on the utilities judgments of what constitutes safe operation. For example, for over 6 months the NRC has tolerated the I operation of Vermont Yankee with the E and F Intermediate Range Monitors (IRMs) inoperable and two unshared Average Power Range Monitors (APRMs) in bypass. This problem was discovered at the beginning of start-up after the March refueling outage. Yet the utility did not halt the start-up to repair the IRMs, even though plant technical specifications, the FSAR, and plant procedures require, as a minimum condition for operatien, two operable APRM downscale scram per channel. NRC is allowing the utility to avoid a shutdown to correct this deficiency in the reactor protection system, although neither Vermont Yankee (after 20 years running this reactor!), or your staff, can determine the importance of this function for protecting public safety. Since there is no way to predict or determine when the reactor might enter a power level requiring this protective function, the NRC decision to allow operation in this degraded mode represents an unacceptable trade-off of safety interests for the utility's interests. NRC allowed the increase of intervals between the inspection and overhaul of the emergency diesel generators when the utility shifted to 18 months between outages, despite the fact that these machines are over 20 years old and near or beyond the end*of their useful lives. The protracted and nearly intractable problems with the "A" EDG documented above, and the first ever failure of the "B" EDG to start (twice!) during an integrated ECCS test, casts doubt on the wisdom of allowing such reductions. Compounding the generator failures themselves. NRC has made 1. NRC presentation on the Vermont Yankee Inspection Program and Recent SALP Report, December 2, 1992, before the Vermont State Nuclear Advisory Panel. See also SALP Report No.50-271/91-99. 2.GAO report " NUCLEAR REGULATION -- Efforts to Ensure Nuclear Power Plant Safety can be Strenthened" GA0/RCED-87-141. 10 'l 1

.4 o questionable judgments about Vermont Yankee's reliance on the Vernon tie-line when granting Limited Condition of Operation requests. This has led to situations where only one back-up generator has been available for emergencies for as long as eight days at a time with the plant running at full power. The burden of owning and operating such a complex machine as a nuclear power plant demands an unwavering commitment to perfect housekeeping. Every safety system is needed, and its perfect operation must be assumed to be essential. This simply should not be a matter for negotiation between management and regulators. NRC must therefore ensure that the maintenance of essential safety systems is immune to budgetary pressures of any kind. The evidence we have presented here suggests that the NRC's current oversight activities at Vermont Yankee are failing to achieve this goal. k'e hope Vermont Yankee is only in the initial stages of degradation due to a neglect of preventive maintenance from the combined factors of cost-cutting, cost containment, and over-emphasis on capacity factor. But only a comprehensive analysis of Vermont Yankee's decisions in these areas can demonstrate this, and only immediate steps to halt these misguided decisions can curtail further deterioration. Since the situation we have described undermines public confidence in Vermont Yankee's dedication to a " safety first" philosophy, we call on you to conduct a public investigation of the issues we have raised, and allow the public opportunity to participate in any corrective action plan you develop. Sincerely, + 4/ Michael Dal , with Joh Greenberg and Michael Mulligan, for the Board of the New England Coalition on Nuclear Pollution 11}}