Text

Responds to from Constituent Hh Chase Re Concern for Safety of Living Area W/Respect to Conditions at Plant.Informs That T Murley & T Martin Will Meet W/Vermont State Nuclear Advisory Panel on 930520 to Discuss Issues
	ML20035F370
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	04/13/1993
From:	Taylor J; NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	Swett D; HOUSE OF REP.
References
	CCS, NUDOCS 9304210205
	Download: ML20035F370 (2)
	v • d • e

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April 13, 1993 The Honorable Dick Swett i

United States House of Representatives l

l Washington, DC 20515

Dear Congressman Swett:

I am responding to a letter of February 5,1993, from your constituent, l

Mr. H. Hamilton Chase, regarding his concern "for the general safety of our l

living area" with respect to conditions at the Vermont Yankee Nuclear Power Station (Vermont Yankee).

l The Coalition has addressed its concerns regarding safety conditions at i

Vermont Yankee in several letters to the NRC during the last 6 months. The i

NRC has responded to the Coalition's concerns in letters dated November 2, j

1992, and January 25, 1993. Copies of these letters are enclosed N r your information. The NRC staff is currently preparing a response to the l

Coalition's most recent letter of February 9,1993.

That letter shared Mr. Chase's concerns regarding publication of NRC criteria, the alleged l

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 tc the Coalition when it is l

1ssued.

t 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 i

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 I

this information is responsive to your concerns.

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Sincerely, i

l Winal signed by Niined.Nihylor Executive Director l

for Operations

Enclosures:

l 1.

Letter dated November 2, 1992 DISTRIBUTION:

See page 2.

2.

Letter dated January 25, 1993 3.

Letter from H. Chase dated February 5,1993

see previous pages for concurrence l

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H. Thompson R. Thompson J. Blaha C. Norsworthy T. Murley/F. Miraglia B. Clayton B. Russell D. Dorman w/ incoming T. Gody T. Clark J. Part. low J. Linville, RI J. Knubel T. Martin, RI PDI-3 Reading w/ incoming J. Linville, RI S. Varga D. Williams, OIG l

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November 2,1992 CHAIRMAN t

t 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-i gas (A0G) system ruptured, and to the operation ano design of tne turbine building roof exhaust fans. A brief background of the event and the suosequent NRC review is provided in Enclosure 1.. Twelve separate concerns i

raised in your letter are addressed in Enclosure 2.

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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 the two day l

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 (ALARA), worker safety, and continued A0G system l

reliability, concluded that on-line repair was not viable.

Plant management I

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 l

A0G system from overpressure transients, such as a hydrogen explosion. The l

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

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.

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 Originated by: TMartin, RI M !! ! LGACLil.

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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 j

subsequent weeks (January 7 and 21) was observed.

Further, independent inspections by the NRC and evaluation of Vermont Yankee's Semi-Annual Effluent i

I 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 corducted several inspections over the past eight years that address Vermont Yankee's design and programs associated with radioactive effluent control.

A continuous charcoal cartridge and particulate filter radiation monitoring system was added to the turoine building roof vent pathway in Ocicber 1991.

Most recently, the NRC concluded that Vermont Yankee's November 1991 l

commitment to reroute the turbine building roof exhaust to the plant stack by 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,

^

Ivan Selin

Enclosures:

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===1.

Background===

2.

Response to Twelve Concerns l

O ENCLOSURE I 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.

l As documented in IR 50-271/92-01, the NRC noted that approximately 4 minutes l

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, the 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 assess the option of performing an on-line repair of the A0G disk.

This assessment l

was not completed by the self-imposed deadline of January 15, 1992. As a I

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 wnich bound the January 13th event.

In January 1992, as part of the NRC's followup to the event, an NRC radiation specialist reviewed Vermont i

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 spec".ication release rate limit.

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

Althougn 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 suosequent weeks (January 7 and 21) was observed.

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.' 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 (Ci) released from the elevated and ground level release points were within allowable limits.

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

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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 l

radioactive gases in the turbine building because of the leaky fuel.

l 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 l

reference to operation of the turbine building roof exhaust fans only 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 ::utomatic 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.

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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 j

filter system to quantify the total amount of radioactive materials (iodines, particulates, and tritium) released through this pathway.

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September 1991, Vermont Yankee revised its Offsite Dose Calculation l

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 I

radioactivity on charcoal cartridges and particulate filters from the I

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.

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3.

The radioactive gas level in the reactor building was higher than that in tht 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 l

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 j

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-l 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).

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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.

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5.

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

filter.

Although the charcoal filter was bypassed when the A0G system was isolated and stacs 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 was 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.

J 6.

Vermont Yankee's claim that no limits were exceeded was questionable l

since the release through the roof exhaust may not have been accounted for and that the flow path may not have adequately been understood.

i 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 the facility to review the adequacy of the licensee's radioactive liquid and gaseous effluent control programs (NRC 1R 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 l

stack both before and after the incident; the measurements indicated no I

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.

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l 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 i

off line to replace the A0G rupture disk, since an on-line repair was i

not acceptable due to ALARA and worker safety concerns, and in l

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 comr tted to obtain quantitative measurements to verify that the air floi'aas 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 direct 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.

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12.

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

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

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 i

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.

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New England Coalition on Nuclear Pollution.Inc.

Box 5-15. Brattleboro. Vermont 05502 Phone <50m 25 -0330 t

i September 15, 1992 I

Ivan Selin, Chairman l

U.S.

Nuclear Regulatory Commission l

Washington. DC l

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-I 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 l

caused a rapid build-up of pressure bursting the steam jet l

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.

I This is the second fuel cycle in a row that management has l

decided to operate the plant with leaky fuel, a practice in 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.

Concentrations of radioactive gases and particles in this uncontrolled release were at least ten times higher l

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 of 250 cpm to 130,000 cpm (NRC report), indicating consider-able exhaust activity that by-passed the charcoal filter 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.

l In addition to this monitored and recognized release path-vay, 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 l

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).

l 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 i

reports.

Yet there is every reason to believe these fans were in op-eration during this incident.

Year-round operation has be-co=e common practice in recent years to reduce elevated l

levels of radioactive gases in the turbine building because i

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 l

your Staff to this questionable practice, as well as to the l

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-l tion with contamination of Connecticut River sediment with co-60 (see Memo for Darrell Eisenhut from Richard V.

Starostecki re: LOW LEVEL RADIOACTIVE EMISSIONS FROM BWR 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 85-012-00: CONTROL ROOM HABITABILITY SYSTEM ACTIVATION).

Did releases via the tur-bine building roof exhaust fans contaminate the reactor building intake air supply?

Ver=ont Yankee claims that the releases associated with this l

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incident didn't exceed any limits (LER 92-003, NRC report).

i 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 j

adequate detail to assess conditions within the plant." (em-phasis added).

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Within six minutes of the discovery of the rupture, the AOG vas restored to service and leakage out the ruptured disc i

was minimized by placing a metal bucket over it.

However, i

the plant then continued operation for two days instead of i= mediately 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 i

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.

AO-73-27, AO-73-26, AO-73-25).

Yet, plant =anagement 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 I

leak).

Meanwhile, control room operators were required to l

perform constant surveillance of the degraded AOG to prevent further uncontrolled releases (LER 92-003: AOG RUPTURE DISC l

TEMPORARY REPAIR NOT WITHIN SYSTEM DESIGN BASES Feb.

13,*921.

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.

I Given these facts and uncertainties, we feel that Vermont Yankee operated in a manner that recklessly endangered l

vorkers, 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.

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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 i

i pathway analysis, and stack monitor data for the period just before and twelve hours after the event.

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3. Require Vermont Yankee to accelerate its plans to isolate j

all turbine building exhaust points and vent them into a

"'tered and monitored system.

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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.

Sin ce rely,

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Michael J. Daley for 2t' Board of the NeJ' England Cdalition on Nuclear Pollution ec. Vermont Yankee Governor Howard Dean media I

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' Enclosure 2 I

a nac o,

UNITED STATES j

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.uso oF MUSSIA. PENNSYLVANIA 19406 141s JA!l Docket No. 50-271 i

Mr. Michael Dalev New England Coalition'on l

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 j

expressed concern that " budgetary pressures" are causing " systematic degradation of safety j

related components at the Vermont Yankee plant." Additionally, you expressed concern that NRC oversight has neither identified nor addressed this issue, in part, due to limited NRC i

resources.

i The NRC is concerned with events, such as those cited in your letter. Such events, individually j

l and collectively, have been and continue to be critically reviewed. However, the NRC's j

independent review of events at the Vermont Yankee Nuclear Power Station (VYNPS) does not l

suppon your assertion that a " systematic degradation of safety-related components" is occurring.

J l

The NRC previously reviewed each of your cited examples; however, no apparent connection l

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.

Funher these events as an aggregate are not " extraordinary" or. " pervasive,"

considering frequency and safety significance.

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 le:ter, 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 leaktge in feedwater check valve 96A." The licenser is required to test this valve every refueling cutage to ensure that leakage is within limits before plant stanup.

As documented in NRC Inspection Report 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 esery 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 elastomeric seating material with stellite. FDW-96A QMftTOb 2 I $

l

,-~.

i

1 l

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 l

of tolerance of a degraded condition.

Another example, which you cited, was that the plant operated for six months with the "E" ano "F" IRM channels inoperable. Although the "E" and "F" IRM channels became inoperable i

during the April 1992 startup, the remaining four IRMs were operable and met the minimum l

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 l

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 being 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 Techrical 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 spurious actuations, which do not affect the APRM operability, occurred after the licensee began l

upgrading the LPRMs and have also been observed at other plants with similar upgrades. The I

bcensee 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 phnts. Problems with licensee performance are identified and addressed though our inspection program and our periodic assessments of plant performance.

NRC inspections, primarily performance-based, frequently focus on the licensee's effectiveness i

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 l

mspection activities, " initiative" inspections, primarily focus upon suspect programmatic weaknesses.

NRC managers and inspectors, periodically and collectively, conouct 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 Reports 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 j

m three areas: nowever, overall performance at VYNPS was good. More frequently, plant performance reviews and briefings are conducted by NRC management, as well as inspectors.

t Mr. Michael Daley 3

j As a regulatory agency, the NRC's oversight of plants and imposition of requirements must not i

be arbitrary. To operate in this manner would be detrimental to our objective of public health and safety, in part, because it may diven 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 auuring 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 information 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 investigation, 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,

[

~

Thomas T. Martin Regional Administrator ec:

Public Document Room (PDR)

Local Public Document Room (LPDR)-

Nuclear Safety information Center (NSIC)

State of Vermont Commonwealth of Massachusetts (2)

)

l l

'1 New England Coalition on Nuclear Pollution,Inc.

Box 545, Brattleboro, Vermont 05302 Phone (802) 257-0336 December 16, 1992 l

Ivan Selin, Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555 i

Dear Chairman Selin:

l

)

We have received your response to our letter of September l

15, 1992 raising concerns about the rupture of the steam jet air ej ector rupture disc and subsequent release of radiation.

We appreciate your detailed reply to those concerns and are preparing our response.

l The purpose of this letter, however, is to alert you to an l

alarming and potentially disastrous situation: systematic l

degradation of safety components at the Vermont Yankee plant.

l 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 bring important safety problems to the attention of plant management and NRC officials failed to produce substantive change, Mr. Mulligan wrote the Coalition an anonymous letter pins,ging regar four major areas of concern at the plant: gashed fuel spent fuel pool cooling, shift staffing, and nocturnal burning of waste oil.

He sent similar letters to the NRC and to l

the State of Vermont Nuclear Engineer, who eventually made the letter public at the request of then Governor Madeline Kunin.

l Shortly thereafter, Mr. Mulligan contacted us by telephone still anonymously -- and made it clear that his primary 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 vould force management to allocate more resources on safety concerns.

l 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.

l l 5 GL)4GOTW y

Edu ca tin g Ihe Pu blic in Cle a n Alternatives to Nuclear Po wer

l!.-

1 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 l

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.

l We have checked and verified each of Mr. Mulligan's reports l

in NRC documents, not because we doubted his story, but because l

we knew others would.

There is simply no question that a variety l

of major safety systems at Vermont Yankee have had substantial difficulties during the past four years.

We list them here, with l

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.

(ECCS) I, which includes:

1) emergence core cooling system

-- the gore spray system,'

-- HPCI anj l

-- 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.

I

2. LER 89-015--00 and LER 89-015-01: " Primary Containment Leak l

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 Inj ection System Inoperable Due j

to Degradation of Station Battery Bus Voltage Caused by Failed i

Battery Charger Component." Event Date: 2/20/92.

1 4.

LER 89-014-00:

" Reactor Core Isolation Cooling System j

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

I I

2) AC and DC power systems which include:

-- the emergency diesel generators,2 l

l 1.

LER 90-009-00: " Inadvertent Reactor Scram Due to a

Short l

Circuit on the Vital AC Bus as a Result of Personnel Error,"

Event Date:

5/1/90.

Also, LER 88-012-00:

" Overloaded Power l

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: 5/29/90.

See also, LER 89-009-00: " Lack of Redundan-cy in Residual Heat Removal Service Water Systems," Event Date:

5/28/39.

2.

LER 90-010-00: " Failure to Meet Technical Specifications for Diesel Generator Operation Readiness Test,"

Event Date:

8/15/90; LER 90-010-01 and LER 90-010-02: " Failure to Meet Tech-nical Specifications for Diesel Testing Generator," Event Date:

2 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 i

Caused by Weak Design." Event Date: 4/23/91. Also, Harold Eichen-

holz, Thomas G. Hiltz, and Richard S. Barkley: " Inspection Report 1

50-271/91-19," Section 4.2.1:

"'A' Emergency Diesel Generator Fuel Oil 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 50-271/92-05," 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 preliminarily determined that the root cause for the first failure was the advanced age of the "B" diesel generator However, based 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 6

" Request for Temporary Vaiver of Compliance from Technical Specification LCO Requirements Pertaining to Emergency Diesel Generator,"

(BV 92-l 058).

Technical Specification 3.5.H.1 requires that "During any

)

period when one of the standby diesel generators is inoperable.

l continued reactor operation is permissible only during the succeed-ing seven days...." On June 4, 1992, Charles V.

Hehl, Director of the Division of Reactor Proj ects at the NRC granted a

" Temporary Waiver of Compliance" allowing Vermont Yankee to run an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months without a safety backup diesel generator.

3 l

l

(

emergency battery systems,I

{

-- emergency batte charge system,2

-- switchyard bus,5y

-- and relays t

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

l

-- RHR service water systems and pumps l

1.

LER 89-020-00: " Removal of a Technical Specification Sur-

[

veillance Requirement from a Procedure Due to an Inadequate Technical Specification Review "

Event-Date:

S/11/89:

"On 8/11/S9, with the plant at 100% power, Vermont Yankee discovered l

the procedure controlling battery maintenance and testing was not consistent with Technical Specification requirements."

r 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 Da*e: 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 15, 1991.

See

also, LER 37-003-00 and LER 87-008-01: " Loss of Normal Power During Shut-down Due to Routing All Off site Power Sources Through One Break-l er," Event Date: S/17/S7.

l 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-l tainment Isolation System 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-l down Cooling Due to Pressure Switch Activation,"

Event Date:

3/14/91: "On 3/14/91 at 0450 hours0.00521 days 0.125 hours 7.440476e-4 weeks 1.71225e-4 months , with reactor vessel cooldown in progress following a reactor scram on 3/13/91... and with the "B"

loop Rosidual 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 of the "B" RHR pump and closure of Shutdown Cooling Suc-starts l

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 Specification 4.2.A, Table 4.2.1, requires an instrument check of l

4 l

4) feedwater systeml and check 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-i 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 Appendiz 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 Appendiz 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."

LER 89-007-00: "1989 Appendiz 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=WI),

Primary Containment Atmospheric Control valve. PCAC-8,9,10,23 and PCAC-5,7,5A,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."

IT SHOULD BE NOTED THAT THE SAME FEEDWATER CHECK VALVE FDW 95A -- WAS REPORTED LEAKING FOR AT LEAST 5 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 l

l

_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.

l As though all of this weren't enough, major questions have been raised during this same geriod about personnel training programs and plant procedures, about the plant's emergency l

l I

l 6

l i

l 1.

LER 92-014-00:

" Inadvertent Scram and ECCS Initiation l

While Shutdown When Restoring Four Level Transmitters to Service."

Event Date:

4/12/92.

Also, letter fros' 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 50-271/91-92 on 911021-25.

Deficien-l 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 event.

Another 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, Rey, 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' Testing 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 J

I operating procedures (EOPs)l, and about plant security 2, Problems with training and security were identified, in part, with a lack of adequate funding.

Ve 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 i

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.

i 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?

i Ve can see no other explanation for this than that utility j

decision-making is unduly driven by the bottom line and that j

your staff is in some way acquiescing in this state of affairs.

1 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 i

)

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 just 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

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

Weaknesses and deficiencies i

noted...."

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

October 13,1992.

7-

s.

Cost cutting and cost containment activities lead to subtle interactions that impact plant operations, such 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 95Al These activities are multiplying just as many of the plant's key components are feeling the effec"s of age-related degradation.

Thus, rather than improting plant safety through increased vigilance, management is moving the plant in the 1

opposite direction. For example, it has pushed hard to tsduce the j

time spent for planned outages.

This means that the pismo can make more money (since even scheduled plant shutdowns are expensive), but it also reduces the amount of time and resources available for firing 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.

1 Increasing the fuel cycle from 12 months to 13 months generates more profits, but it also increases the strain on the system.

Plant components work Sarder and longer, with less frequently scheduled maintena:

a.

To maintain short outage times, the utility has bedun tu shift various maintenance activities normally performed during an outage into periods when the plant is operating at full power.

This aractice has had unsettling consequences, resulting in a r er:

- scram that seriously challenged safety eqgipment and. release of radiation into the environment Proposed new NRC regulations would actually reduce the plant's accountability by extending reporting periods to match l

1. LER S4-011-01 and B4-011-02: " Update on Leaking Containment Isolation Valves " Event date: 5-15-84: "...FDW 95A and 39C...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

i the longer cyclesl, 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 l

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 atility's internal practice for handling MRs allows a screening l

of requests before they are actually logged onto the official

{

i computerized tracking system.

Given the atmosphere we have been describit;, it is unreasonable for your Staff to assume that this screenin. is performed with safety considerations as the prime criterion.

Your staff's inability, or unwillingness, to identify the pattern desec) bed 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 I

too little time to explore underlying causes.

Officials from the NRC Region I inspection branch told i

members of the Vermont State Nuclear Advisory Panel as much at a December 2, 1992 meeting.

Regional Supervisor E.

Kelley spoke of i

the difficult " art" of allocating limited personnel and resources 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 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 1.

" Reducing the Regulatory Burden on Nuclear Licensees,"

Proposed Rule RIN 3150-AE 30, Federal Register, June 18,

1992, pp.

27137-27191 and " Review of Reactor Licensee Reporting Re-quirements," Federal Register, June 19, 1992, pp. 27394-5.

9

fundamental issues in maj or program areas"I.

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 suf ficientig j eopardizes public safety the shut down of the plant This worries us.

to warrant 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 j udgments of what constitutes safe operation.

For example, for over 5 months the NRC has tolerated the 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 operation, 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 furetion, 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 i

j 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 l.

NRC presentation on the Vermont Yankee Inspection Program and Recent SALP Report. December 2, 1992, before the Vermont State Nuclear Advisory Panel.

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 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 Pollution has for over two decades acted responsibly, yet persistently, and without scare tactics or wild accusations. Yet the complaints entered 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-I erest and safety,, not to help plants in their financial stress. Thy l

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 t iga t ion:of condi-tions at Vermont Yankee and of oversight practices in NRC Region I,

and make formal answers for public awareness.

These days, it takes public or citizen action to get 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 Congress people, and Chair-man Selin, so the NRC will pay close attention.

Sincerely, g

gg 63 m aplelke,Keene,N E

OFFICE OF THE SECRETARY CORRESPONDENCE CONTROL TICKET PAPER NUMBER:

CRC-93-0230 LOGGING DATE: Mar 23 93 ACTION OFFICE:

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AUTHOR:

KURT JOHNSON

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ADDRESSEE:

LEGIS ASST TO CONGRESSMAN SWETT LETTER DATE:

Mar 10 93 FILE CODE: IDR-5 VY

SUBJECT:

SAFETY OF RESIDENTS AROUND THE VERMONT YANKEE PLANT ACTION:

Direct Reply DISTRIBUTION:

EDO SPECIAL HANDLING: NONE s

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NOTES:

DATE DUE:

Apr 6 93 SIGNATURE:

DATE SIGNED:

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NRC FOR SIGNATURE OF:

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Exocutive Director DESC:

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ENCLOSES LETTER FROM HAM CHASE, CLEAN WAY Taylor INDUSTRIES, INC. RE SAFETY OF RESIDENTS AROUND Sniezek THE VERMONT YANKEE PLANT Thompson Blaha DATE: 03/25/93 Knubel l

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ML20035F370

Contents

Text

Dear Congressman Swett:

Enclosures:

Dear Mr. Daley:

Enclosures:

Dear Chairman Selin:

Dear Mr. Daley:

Dear Chairman Selin:

Dear Congressman Swett,

SUBJECT:

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ML20035F370

Text

Dear Congressman Swett:

Enclosures:

Dear Mr. Daley:

Enclosures:

Dear Chairman Selin:

Dear Mr. Daley:

Dear Chairman Selin:

Dear Congressman Swett,

SUBJECT:

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