53840 Federal Register / Vol. 62, No. 200 / Thursday, October 16, 1997 / Notices

EFFECTIVE DATE: October 1, 1997.

ADDRESSES: Interested persons may

obtain a single copy of Management

Directive 5.6 by contacting Nancy

Belmore, Office of State Programs, U.S.

Nuclear Regulatory Commission,

Document Control Desk, P1–37,

Washington, DC 20555, telephone

(301)–415–2326.

FOR FURTHER INFORMATION CONTACT:

Kathleen N. Schneider, Office of State

Programs, U.S. Nuclear Regulatory

Commission, Document Control Desk,

P1–37, Washington, DC 20555,

telephone (301)–415–2320.

SUPPLEMENTARY INFORMATION: In 1995,

NRC implemented, on an interim basis,

a process to evaluate NRC regional

licensing and inspection programs and

Agreement State radiation control

programs that regulate the use of

radioactive materials in an integrated

manner using common performance

indicators (see 60 FR 54734; October 25,

1995). The NRC staff conducted the

interim program using Management

Directive 5.6, ‘‘Integrated Materials

Performance Evaluation Program’’ dated

September 12, 1995. On June 30, 1997,

the Commission approved SECY–97–

054, Final Recommendations on Policy

Statements and Implementing

Procedures for: ‘‘Statement of Principles

and Policy for the Agreement State

Program’’ and ‘‘Policy Statement on the

Adequacy and Compatibility of

Agreement State Programs.’’ NRC is

implementing IMPEP with the

corresponding revisions as a result of

the final policy statements. The revised

Management Directive is currently being

prepared in final form to incorporate the

final policy statements and comments

received during interim implementation

of IMPEP from the Regions and the

Agreement States.

NRC is rescinding the May 28, 1992,

‘‘NRC Review of Agreement State

Radiation Control Programs: Final

General Statement of Policy,’’ on

October 1, 1997. This policy is

superseded by IMPEP, which is no

longer considered an interim program.

SMALL BUSINESS REGULATORY

ENFORCEMENT FAIRNESS ACT: In

accordance with the Small Business

Regulatory Enforcement Fairness Act of

1996, the NRC has determined that

Management Directive 5.6 is not a major

rule and has verified this determination

with the Office of Information and

Regulatory Affairs of the Office of

Management and Budget.

Dated at Rockville Maryland this 7th day

of October, 1997.

For the Nuclear Regulatory Commission.

John C. Hoyle,

Secretary of the Commission.

[FR Doc. 97–27424 Filed 10–15–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–271]

Vermont Yankee Nuclear Power

Corporation, Vermont Yankee Nuclear

Power Station; Issuance of Partial

Director’s Decision Under 10 CFR

2.206

Notice is hereby given that the

Director, Office of Nuclear Reactor

Regulation, has taken action with regard

to a Petition dated December 6, 1996,

submitted by Mr. Jonathan M. Block, on

behalf of the Citizens Awareness

Network, Inc. (CAN). The Petition

requested evaluation of certain

Memoranda enclosed with the Petition

relating to the Vermont Yankee Nuclear

Power Station operated by the Vermont

Yankee Nuclear Power Corporation

(Licensee) to see if enforcement action

is warranted.

The first document enclosed with the

Petition is a CAN Memorandum dated

December 5, 1996, that reviews

information presented by the Licensee

at an enforcement conference held on

July 23, 1996, involving the minimum

flow valves in the Vermont Yankee

residual heat removal system. CAN

raises a concern that the corrective

action taken by the Licensee in opening

these valves may have introduced an

unreviewed safety question with regard

to containment isolation.

The second document enclosed with

the Petition is a CAN Memorandum

dated December 6, 1996, that contains a

review of certain Licensee Event Reports

(LERs) submitted by the Licensee in the

latter part of 1996. Various issues are

presented, such as fire protection,

tornado protection, thermal protection

for piping lines, equipment operability,

and equipment testing. On the basis of

its analysis of the LERs, CAN reaches

certain conclusions regarding the

performance of the Licensee and actions

that should be taken.

On the basis of these documents, CAN

requests that the NRC determine

whether enforcement action is

warranted pursuant to 10 CFR 2.206.

The Director of the Office of Nuclear

Reactor Regulation has granted the

Petition in that the NRC staff has

evaluated the majority of issues and

LERs raised in these Memoranda to see

if enforcement action is warranted based

upon the information contained therein.

The conclusion of the evaluation is that

no further enforcement action is

warranted for those issues and LERs that

are closed. LERs which remain open

will be resolved through the normal

inspection and enforcement process and

will be addressed in a Final Director’s

Decision after the NRC staff has

completed its evaluations. The reasons

for the staff’s conclusions are provided

in the ‘‘Partial Director’s Decision

Pursuant to 10 CFR 2.206’’ (DD–97–25),

the complete text of which follows this

notice and is available for public

inspection at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC

20037, and at the local public document

room located at Brooks Memorial

Library, 224 Main Street, Brattleboro,

VT 05301. A copy of the Decision will

be filed with the Secretary of the

Commission for the Commission’s

review in accordance with 10 CFR

2.206(c) of the Commission’s

regulations. As provided for by this

regulation, the Decision will become the

final action of the Commission 25 days

after the date of issuance, unless the

Commission, on its own motion,

institutes a review of the decision in

that time.

Dated at Rockville, Maryland, this 8th day

of October 1997.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

Partial Director’s Decision Pursuant to

10 CFR 2.206

I. Introduction

On December 6, 1996, Mr. Jonathan

M. Block, submitted a Petition to the

Office of the Secretary of the U.S.

Nuclear Regulatory Commission (NRC)

pursuant to Section 2.206 of Title 10 of

the Code of Federal Regulations (10 CFR

2.206). The Petition was submitted on

behalf of the Citizen’s Awareness

Network, Inc. (CAN or Petitioner), and

contained two Memoranda from CAN.

The first Memorandum enclosed with

the Petition is dated December 5, 1996.

It reviews information presented by the

Vermont Yankee Nuclear Power

Corporation (Licensee) at a

predecisional enforcement conference

held on July 23, 1996, involving the

minimum flow valves in the residual

heat removal (RHR) system at the

Vermont Yankee Nuclear Power Station

(Vermont Yankee facility). CAN raises a

concern that the corrective action taken

by the Licensee in opening these valves

may have introduced an unreviewed

safety question with regard to

containment isolation.
Federal Register / Vol. 62, No. 200 / Thursday, October 16, 1997 / Notices 53841

1Several statements in the December 5, 1996,

Memorandum are either unclear or incorrect. A

single power supply failure does not prevent RHR

minimum flow valves in both loops from operating,

contrary to the statement on page 2 of the

Memorandum. Minimum flow valves in both loops

will not remain open if a single power supply

failure occurs, contrary to the statement on page 3

of the Memorandum. Also, on page 4 of the

December 5, 1996, Memorandum, CAN questions

the remote manual closure capability of the

minimum flow valves. The minimum flow valves

have remote manual closure and opening

capability, but the pump protection logic will

override any remote manual closure or opening

signal.

2The NRC staff assumes Petitioner’s reference to

an ‘‘unreviewed safety question’’ is in the context

of the NRC’s regulation 10 CFR 50.59, ‘‘Changes,

Tests, and Experiments’’.

The second Memorandum enclosed

with the Petition is dated December 6,

1996, and contains a review of certain

licensee event reports (LERs) submitted

by the Licensee in the latter part of

1996. Various issues are presented, such

as fire protection, tornado protection,

thermal protection for piping lines,

equipment operability, and equipment

testing. On the basis of its analysis of

the LERs, CAN reaches certain

conclusions regarding Licensee

performance and actions that should be

taken. In the Petition, the Petitioner

requested that the NRC evaluate these

documents, pursuant to 10 CFR 2.206,

to see if enforcement action is warranted

based upon the information contained

therein.

On February 12, 1997, the NRC

informed the Petitioner in an

acknowledgement letter that the Petition

had been referred to the Office of

Nuclear Reactor Regulation for the

preparation of a Director’s Decision and

that action would be taken within a

reasonable time regarding the specific

concerns raised in the Petition.

II. Discussion

The NRC staff evaluation of these

documents follows.

A. The Residual Heat Removal System

The first document enclosed with the

Petition is a CAN Memorandum dated

December 5, 1996, that reviews

information presented by the Licensee

at a predecisional enforcement

conference held on July 23, 1996,

involving the minimum flow valves in

the Vermont Yankee RHR system.1 The

Vermont Yankee RHR system consists of

two loops. Each loop has two pumps

that take suction from the suppression

chamber. Each pump has a minimum

flow line equipped with a minimum

flow valve that returns flow to the

suppression chamber. The RHR pumps

start automatically to cool the reactor in

case of a loss-of-coolant accident

(LOCA). The minimum flow valves

close to prevent flow from being

diverted from the reactor core to the

suppression pool when flow is being

supplied from the RHR pumps to the

reactor core, and open automatically on

high pump discharge pressure to protect

the RHR pumps if other valves between

the RHR pumps discharge and the

reactor core are not yet open.

The Licensee discovered a

vulnerability to single failure which

could prevent the minimum flow valves

from opening to protect the RHR pumps

during a LOCA. To resolve this concern,

the Licensee changed the normal and

failed positions of these valves from

CLOSED to OPEN. The Petitioner is

concerned that the corrective action

taken by the Licensee in opening these

valves may have introduced an

unreviewed safety question with regard

to containment isolation.2 A pipe break

outside containment would breach

primary containment with an OPEN

minimum flow valve.

This issue must be addressed in terms

of the Vermont Yankee facility licensing

basis. The basic design for early boilingwater reactors, including the Vermont

Yankee facility which was reviewed and

accepted by the NRC, considered the

piping of the RHR and Core Spray (CS)

Systems to be a closed extension of

primary containment. Failure of the

passive pressure boundary (piping) of

these systems during either the shortterm (injection phase) or long-term

(recirculation phase) course of a designbasis accident (DBA) was not a design

basis assumption. As a result, the RHR

and CS suction and minimum flow lines

were not provided with containment

isolation valves, or if valves were

provided in these lines, they were not

provided for the purpose of meeting

containment isolation requirements and

thus were not classified as containment

isolation valves. In most if not all cases,

the penetrations of concern in the older

plants were originally provided with at

least one valve capable of performing

the containment isolation function, and

these valves are periodically tested

under inservice testing (IST) program

requirements. The Vermont Yankee

minimum flow valves can be remotely

closed and are periodically tested under

the IST program.

For more recent facilities, emergency

core cooling system (ECCS) closed

systems outside containment are

required to have at least one recognized

isolation valve at each penetration. This

is not the case for the Vermont Yankee

facility.

In view of the licensing criteria

applicable to the Vermont Yankee

facility, maintaining the minimum flow

valves of the RHR system in the OPEN

position is permitted and acceptable.

The Vermont Yankee final safety

analysis report (FSAR) does not describe

the minimum flow valves as being in

the CLOSED position, and placing these

valves in the OPEN position is not a

change to the facility under the meaning

of 10 CFR 50.59 and no unreviewed

safety question is presented. For the

above reasons, no enforcement action is

warranted with regards to this issue.

B. Licensee Event Reports

The second document enclosed with

the Petition is a CAN Memorandum

dated December 6, 1996, that contains a

review of several LERs submitted by the

Licensee in the latter part of 1996.

Various issues are presented, such as

fire protection, tornado protection,

thermal protection for piping lines,

equipment operability, and equipment

testing. On the basis of its analysis of

the LERs, CAN reaches certain

conclusions regarding Licensee

performance and actions that it believes

should be taken. First, CAN requests

that the NRC and the Licensee review

all safety analyses conducted since

initial startup of the Vermont Yankee

facility with particular attention to their

role in providing a complete and up-todate FSAR. Second, the Licensee needs

to correct serious deficiencies in its

design change control process and

should undertake a historical review of

its design control documentation to

verify its accuracy. Third, the Licensee

should perform a global evaluation to

determine how many modifications

have been inadequately tested since

startup. Fourth, the Licensee needs to

initiate a thorough retraining program to

review and emphasize the underlying

safety purposes of Technical

Specifications, the FSAR, design bases

and NRC regulations in relation to

routine operation of the Vermont

Yankee facility, emergency

preparedness, and practical

implementation of the NRC’s ‘‘defense

in depth’’ philosophy. Finally, CAN

strongly recommends that the Licensee’s

Vermont Yankee staff receive training

on the proper use of the ‘‘Single

Failure’’ criterion.

The LERs identified in the CAN

Memorandum are briefly discussed

below.

(1) LER 96–13: ‘‘Two fire suppression

systems do not meet design

requirements due to personnel error

on the part of [the] vendor who

designed and installed the systems’’

CAN asserts that the LER did not

address the cause and consequences of
53842 Federal Register / Vol. 62, No. 200 / Thursday, October 16, 1997 / Notices

3The NRC’s policy and procedures for

determining the enforcement action that may be

warranted for a violation are discussed in NUREG–

1600, ‘‘General Statement of Policy and Procedures

for NRC Enforcement Actions’’ (Enforcement

Policy). Because regulatory requirements have

varying degrees of safety, safeguards, or

environmental significance, the first step in the

enforcement process is to evaluate the significance

of the violation and then assign a severity level to

the violation. A violation is assigned one of four

severity levels. As described in Section IV of the

Enforcement Policy, Severity Level I is assigned to

violations that are the most safety significant and

Severity Level IV is assigned to violations that are

the least safety significant. Consistent with the

recognition that violations have different degrees of

safety significance, the Enforcement Policy

recognizes that there are other violations of minor

safety or environmental concern that are below the

level of significance of Severity Level IV violations.

These minor violations are not normally the subject

of formal enforcement action and are not usually

described in inspection reports. To the extent that

such violations are described, they are usually

described as ‘‘Non-Cited Violations.’’

the foam suppression system deficiency,

which is one of the two fire suppression

systems addressed in this LER. CAN is

correct in that the Licensee did not

determine a precise root cause because

such a long time had elapsed since the

occurrence (1978). It is not unreasonable

for a licensee to be unable to ascertain

the exact root cause of a personnel error

that took place many years before (18

years in this case). Key points that are

considered in reviewing an LER are (1)

whether the specific problem is being

appropriately addressed, (2) whether the

potential for a broader problem exists

and, (3) if a broader problem exists,

whether it is properly addressed. In this

case, the Licensee reviewed its current

design process and procedures and

determined that a similar occurrence

would not be expected to occur now,

and the Licensee had two teams that

were actively reviewing the fire

protection design bases and searching

for the types of problems reported in the

LER. CAN is incorrect in stating that the

consequences of the foam system

deficiency were not discussed in the

LER. The Licensee stated that any fire in

the area would be contained and

suppressed, preventing its spread to

safety-related equipment.

Because the design deficiencies

addressed in this LER were licenseeidentified and corrected, they were

treated as Non-Cited Violations in

Inspection Report 50–271/96–05 in

accordance with Section VII.B.1 of the

NRC’s Enforcement Policy,3 and the

LER was closed in Inspection Report

50–271/96–06. Further enforcement

action is not warranted.

(2) LER 96–14: ‘‘Failure to provide

tornado protection for diesel

generator rooms as specified in the

Final Safety Analysis Report due to

unknown cause’’

The FSAR states that large venting

areas are provided to vent the diesel

generator room in the event of a tornado

to provide pressure equalization. The

LER notes that the facility as

constructed did not include venting.

CAN asserts that ‘‘flaws in the FSAR

cause serious, rippling effects

throughout VY’s [Vermont Yankee

facility’s] safety systems’’ and that the

Licensee ‘‘must include assessments of

the impact of the deficient conditions

upon all affected programs.’’

The Licensee took immediate action

to insure emergency diesel generator

(EDG) operability in the absence of the

pressure relief panels. The Licensee

took immediate compensatory measures

which included blocking open the EDG

room doors and posting fire and security

watches. The Licensee took additional

compensatory actions for the restoration

of operability of the diesel and day tank

enclosures during cold weather months

when the EDG doors had to be shut. An

NRC inspector verified that the

recommended compensatory measures

were properly implemented.

The discrepancy between the actual

plant design and the FSAR is a de facto

change to the facility as described in the

safety analysis report, and thus required

an evaluation to meet the requirements

of 10 CFR 50.59. The failure to perform

such a 10 CFR 50.59 evaluation was

categorized as a Severity Level IV

violation, and was dispositioned in

Inspection Report 96–11 as a Non-Cited

Violation in accordance with Section

VII.B.1 of the Enforcement Policy.

Other plants have been found to have

FSARs which do not properly describe

the facilities. Consequently, for this

reason and as a result of lessons learned

from events at Millstone Nuclear Power

Station and Maine Yankee Atomic

Power Station, on October 9, 1996, the

NRC requested information from all

power reactor licensees, to verify,

among other things, that the plant

FSARs properly describe the facilities,

and that the systems, structures, and

components are consistent with the

design basis. In conjunction with this

request for information, and in order to

encourage licensees to identify

discrepancies, the Commission

approved a modification to the NRC

Enforcement Policy that allows the NRC

staff to exercise enforcement discretion

for a period of 2 years for violations

related to FSAR discrepancies identified

by licensees. The policy revision was

published in the Federal Register on

October 18, 1996 (61 FR 54461).

In the Licensee’s response to this

request for information dated February

14, 1997, the Licensee committed to

complete its FSAR verification program

in 1998.

CAN raises a concern about a

potential error in the Licensee’s

statement in this LER of no prior

occurrences, based on a James A.

Fitzpatrick Nuclear Power Plant report

of a similar problem. Licensees are only

required to report prior similar

occurrences at their facility, and not at

any other facility. Therefore, the

Licensee was accurately reporting that a

similar event had not previously

occurred at Vermont Yankee Nuclear

Power Station. This LER is closed.

Further enforcement action is not

warranted. The Licensee has issued a

supplement to this LER to document the

long term corrective actions to vent the

EDG room in the event of a tornado to

provide pressure equalization. This LER

supplement remains open pending NRC

inspection of the Licensee’s

modifications to the EDG room to

provide the required pressure

equalization.

(3) LER 96–15: ‘‘Original B31.1 ANSI

Code Section that Required

Overpressurization Relief for

Isolated Piping Sections was not

Considered during [the] Original

Design’’

Certain piping sections, which would

be isolated after a LOCA, were found to

lack overpressure protection contrary to

code requirements. The water in this

piping could expand because of the high

temperatures accompanying a LOCA

and exceed the design pressure rating of

the piping. CAN asserts that the

Licensee failed to take advantage of

earlier opportunities to identify this

design error when making modifications

to the six systems discussed in the LER.

CAN is correct in that the LER

represented the first discovery of this

problem, although modifications had

been made to the affected systems

earlier. This potential

overpressurization problem has been

identified at other plants, as evidenced

by the issuance of NRC Information

Notice (IN) 96–49 on August 20, 1996,

and NRC Generic Letter (GL) 96–06 on

September 30, 1996. The Licensee did

maintain an awareness of events in this

area and identified this issue at its site

before the generic communications

referred to above were issued. The NRC

staff encourages licensee initiatives to

identify and correct safety problems that

may be generic to the industry in

advance of generic NRC staff

communications to the industry. The

Licensee’s corrective actions included a

design change which provided the

required overpressure protection for the
Federal Register / Vol. 62, No. 200 / Thursday, October 16, 1997 / Notices 53843

affected lines. The change was

completed in the 1996 refueling outage.

This LER remains open. Responses

from power reactor licensees to GL 96–

06 were received by the NRC staff in

February 1997 and are undergoing

review to assure that the overpressure

protection issue is being adequately

addressed and resolved. Following this

generic review, a determination will be

made of whether enforcement action is

warranted for specific plants.

Information regarding the completion of

this activity and any enforcement action

taken will be publicly available in the

plant specific Inspection Reports. This

LER will be further discussed in a Final

Director’s Decision when the LER is

closed.

(4) LER 96–18: ‘‘Inadequate Installation

and Inspection of Fire Protection

Wrap Results in Plant Operation

Outside of Its Design Basis, A

Single Fire Would Impact Multiple

Trains of Safety-Related

Equipment’’

CAN asserts that this deficiency had

significant adverse safety implications.

The reported deficiency consisted of a

small gap in the fire barrier installed on

a cable tray support. The cable tray

contained wiring to support operation of

the ECCS. The NRC staff does not

consider CAN’s claim, that a fire could

have rendered both divisions of the

ECCS inoperable, credible. The

Licensee’s evaluation found that

existing fire protection analyses were

very conservative, and that, with the

combustible loading and fire detection

and suppression equipment in the area,

no credible fire threat could challenge

the functionality of the ‘‘as found,’’

wrapped cable. The Licensee has acted

appropriately to correct the fire barrier

deficiency and to prevent similar

problems in the future. With the

combustible loading, fire detection, and

suppression equipment in the area, the

NRC staff conceptually agrees with the

Licensee’s conclusion that no credible

fire threat could challenge the

functionality of the ‘‘as found’’ wrapped

cable. Inspection activities were

performed the week of August 18, 1997

to verify the Licensee’s conclusion.

This LER remains open. Results of the

inspection and any enforcement action

as a result of this inspection activity

will be made publicly available through

plant specific Inspection Reports. This

LER will be further discussed in a Final

Director’s Decision when the LER is

closed.

(5) LER 96–19: ‘‘Half scram and group

III containment isolation caused by

loose Reactor Protection System

breaker termination’’

The NRC staff agrees with CAN that

this event presented no significant risk

to public health and safety. This LER is

closed. No violation was involved,

therefore the NRC staff concludes that

enforcement action is not warranted.

(6) LER 96–20: ‘‘Inadequate vender [sic]

design activity and Licensee design

verification result in inability to

demonstrate Fire Suppression

System Operability’’

This LER involved the inability of the

carbon dioxide fire suppression system

to fully extinguish a deep-seated fire, as

required. The Licensee stated in the LER

that this event had no safety

significance. The NRC staff considered

this LER to have little apparent actual or

potential safety significance. This

conclusion was based on the Licensee’s

analysis that although the carbon

dioxide suppression systems might not

fully extinguish a deep-seated fire, the

suppression and detection systems

would function. Fire detection would

alert the fire brigade, and because the

carbon dioxide fire suppression system

had reduced the fire, the fire brigade

could extinguish the fire more easily.

The NRC staff closed this LER in

Inspection Report 96–11. Pending

inspector review of the Licensee’s

corrective actions, the unresolved item

initiated for this issue in Inspection

Report 96–08 (URI 96–08–01) was left

open. As documented in Inspection

Report 97–05, unresolved item 96–08–

01 was closed and a Non-Cited

Violation was issued, consistent with

Section VII.B.1 of the NRC Enforcement

Policy. Further enforcement action is

not warranted.

CAN asserts that this LER reveals a

serious deficiency in the Licensee’s

design change control process, and that

the Licensee should determine how

many other modifications have been

inadequately tested since startup. The

NRC staff agrees that this event

demonstrated a weakness in the

Licensee’s modification and testing

programs associated with fire

protection. As noted under the

discussion regarding LER 96–13, the

Licensee has initiated reviews of the fire

protection design bases to search for

these types of problems, and believes

that the current design process and

procedures are adequate to prevent

similar problems. As discussed earlier,

by letter dated October 9, 1996, the NRC

staff requested information from all

licensees, to verify, among other things,

the adequacy of the design change

control process and to determine the

rationale for concluding that designbasis requirements are properly

translated into operating, maintenance,

and testing procedures. The Licensee

responded by letter dated February 14,

1997.

(7) LER 96–21: ‘‘Inadequate procedural

controls of MOV Limit Switch

Settings result in a potential

common cause failure mode with

the capacity to affect multiple safety

significant components’’

This LER involved two limit switches

on shutdown cooling suction motoroperated valve (MOV) to the ‘‘D’’ RHR

pump. The switches measure valve

travel towards the open position. One

open limit switch permits the pump

motor to start after the valve position is

sufficiently open, and the other limit

switch stops valve travel so that the

motor doesn’t drive the valve too far and

damage the valve. The Licensee

identified that a modification to the

valve’s motor operator resulted in the

improper setting of these two limit

switches.

Inspector follow-up, as documented

in Inspection Report 97–05, led to the

conclusion that this error was of low

safety significance. The failed start of

the ‘‘D’’ RHR pump because of this limit

switch error on the shutdown cooling

suction valve affected only the

shutdown cooling mode of operation of

the RHR system. The failure did not

impact the other modes of RHR system

operation and the safety design bases

functions of the RHR system. Further,

prompt Licensee action was taken to

check the other recently modified

MOVs. Their limit switches were found

to be properly set and therefore their

safety functions were unaffected. This

licensee-identified and corrected

violation resulted in the issuance of a

Non-Cited Violation, consistent with

Section VII.B.1 of the NRC Enforcement

Policy. This LER is closed. Further

enforcement action is not warranted.

(8) LER 96–22: ‘‘Combination of poor

man-machine interface, an

inadequate procedure, inadequate

Operating Experience Review

results in a common cause failure

mechanism, and an Emergency

Diesel Generator to exceed Tech

Spec [sic] outage time’’

The output breaker for one emergency

diesel generator (EDG) was found to be

incapable of closing because of a

missing cotter pin which was necessary

for a mechanical linkage. As a result of

the absence of this cotter pin, the

breaker closing springs failed to

recharge, rendering the breaker

incapable of being closed from the

control room. The only indication that

the closing springs had failed to

recharge was a mechanical flag indicator

located behind the breaker cubicle door.
53844 Federal Register / Vol. 62, No. 200 / Thursday, October 16, 1997 / Notices

4CAN asserts that the Licensee misconstrues the

purposes of TS Limiting Conditions for Operation

(LCOs) as part of a ‘‘chronic pattern of

misunderstanding’’ of TS, FSAR design bases, and

NRC regulations. For the reasons described herein,

LER 96–22 does not provide a basis for this

assertion.

No licensee procedures required

verification of the closing spring status.

The closing springs were apparently in

an uncharged condition for over three

weeks without discovery. Because the

periodic surveillance interval for the

breaker is greater than the EDG limiting

condition for operation (LCO), the

Licensee unknowingly operated in

violation of its Technical Specification

(TS) governing diesel generator

operability. After reviewing the

Licensee’s root cause analysis of this

event, the NRC staff determined that the

missing cotter pin would not reasonably

have been expected to be detected by

the Licensee’s existing quality assurance

program or through other related control

measures. 4 The Licensee identified the

EDG inoperability, investigated to

determine when the problem arose, and

reported that the LCO time was

exceeded. The Licensee responded to

the inoperable equipment when the

inoperability was discovered. The

Licensee did not intentionally exceed an

LCO. Rather, the Licensee discovered an

equipment problem caused by a

malfunction beyond its control which

meant that, in hindsight, an LCO had

been exceeded. The Licensee is

designing a modification for this and

other circuit breakers of similar design

to allow monitoring of the charging

status of the closing springs without

having to open the circuit breaker

cubicle door.

Because the EDG inoperability was

not avoidable by reasonable Licensee

quality assurance measures or

management controls, the NRC did not

issue a Notice of Violation for this issue.

This is consistent with Section VI.A of

the Enforcement Policy. This LER is

closed. The NRC staff concludes that

further enforcement action is not

warranted.

(9) LER 96–23: ‘‘Inadequate Surveillance

Procedure results in failure to meet

Technical Specification

requirements for Radiation Monitor

Functional Testing’’

The reactor building and refueling

floor radiation monitor test procedure

did not verify the high alarm contact

actuation as required by TS. The NRC

staff agrees with CAN that this event

presented no significant risk to public

health and safety. Considering that the

monitors were verified to be fully

functional, and were in the condition

required by Plant Technical

Specifications, this specific event

appears to have been limited to an

inadequate testing methodology. The

Licensee’s corrective actions included

revising the deficient surveillance test

procedure to properly test the high

alarm output contacts.

However, the LER remains open as

the NRC staff has not completed its

inspection activities related to this LER.

The NRC staff will look historically to

see if this is an isolated case as part of

the enforcement consideration. On

January 10, 1996 the NRC issued

Generic Letter (GL) 96–01 , ‘‘Testing of

Safety-Related Logic Circuits,’’ that

requested, among other things, that all

power reactor licensees review their

surveillance test procedures to ensure

that all portions of the logic circuitry are

being tested. The Licensee’s response to

GL 96–01, due to be sent to the NRC in

September 1997, will be evaluated with

respect to the Licensee’s long-term

corrective action for logic testing

procedures, because any associated

corrective action could be considered in

determining whether enforcement

action is warranted. Information

regarding any enforcement action taken

will be available publicly in plantspecific Inspection Reports. This LER

will be further discussed in a Final

Director’s Decision when the LER is

closed.

(10) LER 96–25: ‘‘Inadequate testing

leads to misadjustment of isolation

valve mechanical stop and failure to

meet Technical Specification leak

rate limits for containment purge

isolation valve’’

This LER involved a containment

isolation valve which leaked in excess

of TS requirements. The amount of

valve leakage was influenced by the

direction in which the valve was leak

tested and the adjustment of a

mechanical stop. CAN’s concern

appears to be that the Licensee failed to

apply the single-failure criterion in

assessing the significance of the failure

in its LER. section 50.73(b)(3) requires

that an LER contain an assessment of

the safety consequences and

implications of the event, including the

availability of other systems or

components that would have performed

the safety function of the failed system

or component. In this case, the

requirement is that the assessment

include the availability of a redundant

component (valve) that would have

performed the safety function (torus

isolation). Petitioner’s issue is thus

whether the LER should have, in

addition, assessed the potential

radiological consequences had a designbasis accident (DBA) occurred with

failure of the redundant isolation valve.

Compliance with section 50.73(b)(3)

does not require that the assessment

consider an additional single failure

beyond the failure which forms the

basis for the assessment. On the basis of

required reporting, LER 96–25 was not

deficient in omitting discussion of the

potential consequences of failure of the

redundant valve. Inspection Report 50–

271/96–11 dispositioned this Severity

Level IV TS violation as a Non-Cited

Violation in accordance with the criteria

for enforcement discretion in Section

VII.B.1 of the Enforcement Policy.

Although the event was considered to

be of more than minor safety

significance, the outboard valves had

successfully passed all previous tests,

and thus the demonstrated containment

integrity was always maintained for the

two affected penetrations. This LER is

closed. No further enforcement action is

warranted.

C. Summary

In summary, with respect to CAN’s

concern that an unreviewed safety

question with respect to containment

isolation may have been introduced by

Licensee actions in opening the RHR

minimum flow lines, the NRC staff

determined that no unreviewed safety

question was introduced and, therefore,

no enforcement action is warranted.

With respect to CAN’s concerns related

to the LERs, the NRC staff finds that the

Enforcement Policy has been applied

consistently for the LERs that have been

closed and further enforcement action is

not warranted.

For those LERs which remain open

the Inspection/Enforcement process will

continue until the staff has completed

its investigation and consideration of

the issues involved. LER closure and

enforcement action, as appropriate, will

be documented publicly as is NRC staff

practice, and will documented in a

Final Director’s Decision.

With regard to CAN’s overall

conclusions based on its analysis of the

above LERs, the NRC staff has reached

the following conclusions:

With respect to CAN’s conclusion that

the NRC and the Licensee should review

all safety analyses conducted since

startup of the Vermont Yankee facility

with particular attention to their role in

providing a complete and up-to-date

FSAR, the NRC staff has taken actions

as noted in the discussion above related

to LER 96–14 with respect to identifying

and correcting FSAR inaccuracies. This

action was taken in a request on October

9, 1996, to all licensees, including

Vermont Yankee, to provide the

requested information. In addition, the

NRC staff has implemented a series of
Federal Register / Vol. 62, No. 200 / Thursday, October 16, 1997 / Notices 53845

engineering design inspections,

including an inspection to verify

portions of the Licensee’s design control

process and maintenance of the

Licensees’s FSAR commitments. The

results of the NRC design inspection

conducted at Vermont Yankee were

reported in Inspection Report 97–201

dated August 27, 1997.

With respect to CAN’s conclusion that

the Licensee needs to correct serious

deficiencies in its design change control

process and should undertake a

historical review of its design control

documentation to verify its accuracy,

the NRC staff has taken action as noted

in the discussion related to LER 96–20

with respect to identifying and

correcting design change control process

deficiencies. In the October 9, 1996

letter to all licensees, including

Vermont Yankee, the NRC staff

requested information to verify, among

other things, the adequacy of the design

change control process and to determine

the rationale for concluding that designbasis requirements are properly

translated into operating, maintenance,

and testing procedures. As also noted in

the discussion related to LER 96–20, the

Licensee has undertaken a review of the

fire protection design bases to search for

the type of problems involved in LER

96–20, and believes that the current

modification programs are adequate to

prevent similar problems.

With respect to CAN’s conclusion that

the Licensee should perform a global

evaluation to determine how many

modifications have been inadequately

tested since startup, as noted in the

discussion related to LER 96–20, the

Licensee has been required to provide

verification of the design change control

process, including among other things

the rationale for concluding that design

basis requirements are translated into

testing procedures.

With respect to CAN’s conclusion that

the Licensee needs to initiate a thorough

retraining program to review and

emphasize the underlying safety

purposes of TSs, the FSAR, design bases

and NRC regulations in relation to

routine operation of the Vermont

Yankee facility, emergency

preparedness, and practical

implementation of the NRC’s ‘‘defense

in depth’’ philosophy, the NRC staff

disagrees. In the discussion related to

LER 96–22, the NRC staff addresses

CAN’s assertion that the Licensee

misconstrues the purposes of TS LCO as

part of a ‘‘chronic pattern of

misunderstanding’’ of TS, FSAR design

bases and NRC regulations. The NRC

staff finds no basis to require such a

retraining program.

Finally, CAN strongly recommends

that the Licensee’s Vermont Yankee staff

receive training on the proper use of the

‘‘Single Failure Criterion.’’ In the

discussion related to LER 96–25, the

NRC staff addresses what seems to be

the basis for CAN’s recommendation:

i.e. the perception that the Licensee

failed to properly apply the Single

Failure Criterion in assessing the

significance of a leaking isolation valve

in LER 96–25. Compliance with Section

50.73 does not require that the

assessment consider an additional

single failure. The enforcement

conference related to the minimum flow

valves concerned a problem in

implementation of the Single Failure

Criterion; not a misunderstanding of the

requirements of the Single Failure

Criterion. Because the Licensee did not

err in the instance described in LER 96–

25 and the Petition provides no other

instances in which problems were

caused by a misunderstanding of the

Single Failure Criterion, the NRC staff

finds no basis for requiring additional

training.

III. Conclusion

The NRC staff has reviewed the

information submitted by the Petitioner.

The Petitioner’s request is granted in

that the NRC staff has evaluated the

majority of issues and LERs raised in the

Memoranda provided by the Petitioner

to see if enforcement action is warranted

based on the information contained

therein. The NRC staff has discussed

each Memorandum above and described

any related enforcement action taken for

those issues and LERs which are closed.

The NRC will continue the same process

and consideration for the LERs that

remain open and documentation of any

inspection and/or enforcement action

will be consistent with agency practices

and will also be the subject of a Final

Director’s Decision.

As provided in 10 CFR 2.206(c), a

copy of this Decision will be filed with

the Secretary of the Commission for the

Commission’s review. This Decision

will become the final action of the

Commission 25 days after issuance,

unless the Commission, on its own

motion, institutes review of the Decision

in that time.

Dated at Rockville, Maryland, this 8th day

of October 1997.

For the Nuclear Regulatory Commission.

Original signed by

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 97–27417 Filed 10–15–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–213]

Connecticut Yankee Atomic Power

Company, Haddam Neck Plant; Notice

of Meeting Regarding Post-Shutdown

Decommission Activities Report

The Nuclear Regulatory Commission

(NRC) previously published a Notice of

Receipt (62 FR 46783, dated September

4, 1997) of the Post-Shutdown

Decommissioning Activities Report

(PSDAR), dated August 22, 1997, for the

Haddam Neck Plant (HNP) located in

Middlesex County, Connecticut, Town

of Haddam.

In that previous Federal Register

Notice, the NRC stated that it would

hold a public meeting in the vicinity of

the Haddam Neck Plant within 60 days

of that notice. The purpose of this

informational meeting is to (1) describe

the licensee’s planned activities, (2)

describe the regulatory process for

decommissioning, (3) hear public

comments regarding health and safety,

and protection of the environment

during decommissioning, and (4)

provide an opportunity for State and

local representatives to participate. The

licensee, Connecticut Yankee Atomic

Power Co., will discuss their plans to

decommission the Haddam Neck Plant.

The NRC will discuss the PSDAR and

the license termination process, and

describe the program for future plant

oversight. The public will have an

opportunity to comment on the PSDAR

and the meeting will be transcribed by

a court reporter.

The meeting is scheduled for Monday,

October 27, 1997 from 6:30 pm to 10:00

pm in the auditorium of the Haddam

Killingsworth High School on Little City

Road in Higganum, Connecticut.

The PSDAR is available for public

inspection at the local public document

room located in the Russell Library, 123

Broad Street, Middletown, CT 06457,

and at the Commission Public

Document Room, 2120 L Street, NW.,

Washington, DC 20037.

Dated at Rockville, Maryland, this 9th day

of October 1997.

For the Nuclear Regulatory Commission.

Seymour H. Weiss,

Director, Non-Power Reactors and

Decommissioning Project Directorate,

Division of Reactor Program Management,

Office of Nuclear Reactor Regulation.

[FR Doc. 97–27416 Filed 10–15–97; 8:45 am]

BILLING CODE 7950–01–P