ML17261B023

Responds to NRC 890222 Ltr Re Violations Noted in Insp Rept 50-244/89-17.Corrective Actions:Personnel Verified Safety Injection Block/Unblock Switch in Proper Position & Operator Procedure 0-1.1 Changed as Indicated

ML17261B023
Person / Time
Site:	Ginna
Issue date:	03/26/1990
From:	MECREDY R C ROCHESTER GAS & ELECTRIC CORP.
To:	RUSSELL W T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
NUDOCS 9004040007
Download: ML17261B023 (16)

See also: IR 05000244/1989017

Text

ACCELERATED

DISTRIBUTION

DEMONST$&TION SYSTEM REGULATORY

INFORMATION

DISTRXBUTION

SYSTEM (RIDS)ESSION NBR:9004040007

DOC~DATE: 90/03/26 NOTARIZED:

NO FACIL:50-244

Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION

MECREDY,R.C.

Rochester Gas&Electric Corp.RECIP.NAME

RECIPIENT, AFFILIATION

RUSSELL,W.T;

Region 1, Ofc of the Director SUBJECT: Responds to NRC 890222 ltr re violations

noted in Insp Rept 50-244/89-17.

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I~

ROCHESTER GAS f f A'f f~ff ff RTC If f,i i'TAN I AND ELECTRIC CORPORATION

~89 EAST AVENUE, ROCHESTER, N.Y.14849-pppg

March 26, 1990 TCKCRHONC ARCA COOK 71K 546 2700 Mr.William T.Russell Regional Administrator

U.S.Nuclear Regulatory

Commission

Region I 475 Allendale Road King of Prussia, Pennsylvania

19406 Subject: Response to Notices of Violation Inspection

Report No.50-244/89-17

R.E.Ginna Nuclear Power Plant Docket No.50-244 Dear Mr.Russell: This letter is in response to the February 22, 1989 letter from Jon R.Johnson, Chief, Projects Branch No.3 to Robert E.Smith, Senior Vice President, RG&E, which transmitted

Inspection

Report No.50-244/89-17.

In that report, two violations

were identified.

The following provides a reply to the violations

pursuant to 10 CFR 2.201.RESTATEMENT

OF VIOLATIONS

During inspection

at the R.E.Ginna Nuclear Power Plant from December 12, 1989 through January 8, 1990, the following violations

were identified

and evaluated in accordance

with the NRC Enforcement

Policy (10 CFR 2, Appendix C): Contrary to the above, a safety injection system design deficiency

was not promptly identified

and corrected when corporate engineering

was notified on or before October 20,'989 that failure of the safety injection block/unblock

switch could block automatic safety injection actuation on low pressurizer

pressure or low steam line pressure.Corporate engineering

did.not conclude that this problem existed at Ginna until about November 17, 1989, and site technical personnel were not informed about the deficiency

until December 19, 1989.This is a Severity Level IV violation (Supplement

I).~Qo~~l"/0040">0V07 200 c'OR ADOCI=000:..44

FDC A.10 CFR 50, Appendix B, Criterion XVI, and the Ginna Quality Assurance Manual, Section 16, require prompt identification

and correction

of conditions

adverse to quality including failures, malfunctions, deficiencies, defective material and equipment, and nonconformances.

4

B.10 CFR 50, Appendix B, Criterion V, and the Ginna Quality Assurance Manual, Section 5, require activities

affecting quality-to be accomplished

in accordance

with instructions, procedures, or drawings which include appropriate

quantitative

or qualitative

acceptance

criteria for determining

that important activities

have been satisfactorily

accomplished.

Contrary to the above, on December 15, 1989, maintenance

was performed on a safety-related

motor-operated

valve in the safety injection system in accordance

with a procedure which included an inappropriate

torque specification.

This is a Severity Level V violation (Supplement

I).RESPONSE TO VIOLATION A RG&E Position on Existence of Violation Rochester Gas and Electric Corporation (RG&E)concurs that a violation of Appendix B, Criterion XVI occurred.RG&E recognizes

that communication

between corporate engineering

and site personnel on issues of potential safety significance

should be formalized.

Our efforts to address this concern are provided in Section 4,"Long Term Enhancements".

As explained below, RG&E also believes that with respect to the issue identified

on October 20, 1989, we acted in a manner consistent

with the safety.significance

of the matter.2.Reason for Violation As Inspection

Report No.50-244/89-17 (p.7)indicates, RG&E received notice on October 20, 1989, from Westinghouse

Electric Corporation (Westinghouse)

of an apparent generic design deficiency

related to the type of safety injection (SI)block/unblock

switch used at various Westinghouse

reactors.The Westinghouse

letter, dated October 12, 1989, concluded that a"single failure of the switch (Westinghouse

OT2)could block either the automatic low pressurizer

pressure or the low steamline pressure SI signal in both trains"[emphasis supplied].

The letter also stated that the probability

of switch failure was"10'10'/yr":and that, while a design change was recommended, the situation was"not an immediate safety concern." In addition, the Westinghouse

letter referred to a Licensee Event Report (LER), No.88-007-00, submitted by Wisconsin, Electric Power Company (Wisconsin

Electric)on September 16, 1988, concerning

the same issue at the Point Beach Nuclear Plant (Point Beach).The Wisconsin Electric LER concluded that"this condition will not have a significant

impact on the health and safety of the general public or the employees of the Point Beach Nuclear Plant."

'

The LER noted that the Point Beach facility was operating at 100%capacity when the concern was identified

and that design change would not'e made until the next scheduled outage.Upon receipt of the Westinghouse

notification

on October 20, 1989, RG&E (corporate)

initiated a timely review for applicability

to Ginna Station.Based on the Wisconsin Electric LER and on Westinghouse's

calculation

of the low probability

of switch failure, it was apparent that the matter did not constitute

an immediate safety concern.When it was identified

that the switch configuration

was applicable

to Ginna Station, an internal engineering

recommendation

was made consistent

with the guidance of the Westinghouse

letter and attached LER, that an EWR be initiated.

This was completed on November 17, 1989.This recommendation

was then evaluated within Nuclear Safety and Licensing, resulting in a discussion

with site technical support personnel relative to this situation on December 19, 1989.On December 20, site personnel initiated a Ginna Station Event Report per Procedure A-25.1 (Event No.89-168).The event report indicated that the site Plant Operations

Review Committee (PORC)had, on December 20, 1989, concluded that plant operation could continue for the following reasons: 1.Westinghouse

stated that the.probability

of failure was very low (i.e., 10'o 10'/yr);2.Emergency Operating Procedures

directed Operators to use manual SI initiation

where indicators

show automatic initiation

has failed;3.A separate automatic SI initiating

mechanism would activate when containment

pressure reached 4 psig;4.During depressurization, a bistable light will'lert operators of a blocked SI signal;and 5.Visual verification

of the SI switch plunger position indicates that the contacts are in the proper position.The violation states that the time between October 20, 1989, when RG&E (corporate)

was notified by Westinghouse, and the communication

of this information

to the site technical staff on December 19, 1989, shows that the SI design deficiency

was not promptly identified

and corrected, and indicates problems in communication

between corporate engineering

and site personnel.

While RG&E does not deny this violation, we believe that the actions taken by RG&E were appropriate

in view of RG&E's preliminary

conclusion

that the issue did not constitute

an immediate safety concern.

RG&E believes that Appendix B, Criterion XVI does not establish a precise time limit for resolution

of safety issues.Rather, issues such as"promptness" or"timeliness" are subjective

matters that inherently

depend upon the safety significance

of the situation.

Given that RGGE had a documented

recommendation

from Westinghouse

that no immediate safety concern existed (as corroborated

by the Point Beach LER), its actions toward resolution

of the issue were prompt and timely.Any other interpretations

of Criterion XVI would be counter to public health and safety because it would require licensees to treat all deficiencies

or non-conforming

items the same (i.e., regardless

of safety significance).

This same basic philosophy

was affirmed in an analogous context'in recent guidance issued by NRC's Office of Nuclear Reactor Regulation

'(NRR).Specifically, on July 19, 1989, Dr.T.E.Murley, Director, NRC/NRR, sent a memorandum

to all of the regional administrators

entitled"Guidance on Action To Be Taken Following Discovery of Potentially

Nonconforming

Equipment." In his memorandum, Dr.Murley stated that"[t]here is no generally appropriate

timeframe in which operability

determinations

should be made." For equipment which is"clearly inoperable," an immediate declaration

of inoperability

should be made and the appropriate

technical specifications

followed.However, Dr.Murley's memorandum

contrasts this situation with those where equipment nonconformances

simply raise the issue of operability.

In such situations

Dr.Murley states that: operability

determinations

should be made by licensees as soon as racticable, and in a timeframe commensurate

with the a licable e ui ment's im ortance to safet usin the best information

available,(e.g., analyses, a test or partial test, experience

with operating events, engineering

judgement or a combination

of the factors)(emphasis supplied).

Although this guidance relates to timing of operability

determinations, it is equally appropriate

with respect to resolution

of open items under Criterion XVI.Consistent

with this philosophy

and based on the best information

available, future cases of this type will be resolved"as soon as practicable" and in a time commensurate

with the safety significance

of the matter.Communication

between corporate and site personnel will be initiated promptly once applicability

to Ginna Station is determined.

Corrective

Ste s Which Have Been Taken and the Results Achieved Corporate and site technical staff and the PORC have reviewed the circumstances

surrounding

the potentially

generic design deficiency

related to the control room SI block/unblock

switch.As stated in LER 89-016, the.following actions were taken:

Knowledgeable

personnel inspected the plunger position of the SI Block/Unblock

Switch and verified that theswitch contacts were in the proper position.~Operating Procedure 0-1.1 (Plant Heatup From Cold Shutdown to Hot Shutdown)was changed to add the following note and check-off to Step 5.11.6: NOTE: Prior to placing the SI Block/Unblock

Switch to the normal position, station an operator inside the MCB in direct observation

of the SI Block/Unblock

Switch to observe that both plunger tips are recessed inward after the switch is placed.to normal position.-

Block switch plunger t'ips position inward~An RG&E operator aid tag was.placed on the.MCB adjacent to the SI Block/Unblock

Switch denoting the note-from 0-1.1.~An RG&E operator aid tag was also placed inside the MCB adj acent to the rear of the SI Block/Unblock

Switch stating the following:

This is the switch we verify that the plunger's tips are recessed inward when the switch is placed to normal (labeled LAK).A spare switch of similar design has been placed in the Control Room for the purpose of training the operators to recognize the differences

in plunger position.These actions are considered

adequate to provide reasonable

assurance of SI system operability

until the situation can be permanently

dispositioned.

Finally, EWR 5025 was initiated to provide for the installation

of independent

SI block/unblock

switches for each SI train which is planned for the 1991 refueling outage.4.Corrective

Ste s Which Will Be Taken to Avoid Further Violation RG&E has recently taken steps to upgrade the overall corrective

action program for Ginna Station.The need for improvements

was noted during the course of the RHR System Safety System Functional

Inspection (SSFI), and is also considered

appropriate

due to RG&E's initiation

of a comprehensive

Configuration

Management/Design

Basis Program.We are working with the NUMARC Design Basis Issues Working Group to develop an improved problem identification

and resolution

program.The improved program will:~Improve the process of identifying, analyzing, and resolving problems;

~Improve the RG&E internal review process, including formalized

means of communication

between corporate engineering

and site personnel on issues of potential safety significance;

and Part of the implementation

of this effort will include specific procedural

upgrades, enhancement

of our corrective

action tracking system, and the issuance of a corporate policy which addresses problem identification

and reporting.

We believe that this broad effort, when fully implemented, will improve our capability

to consistently

identify and disposition

potential safety issues commensurate

with their significance.

5.Date When Full Com liance Will Be Achieved Long term and short term actions and schedules have been described above.Formal guidance concerning

communication

between corporate and site personnel on identified

problem issues is under development, and is targeted for completion

by July 1990.RESPONSE TO VIOLATION B Rochester Gas and Electric concurs with this violation as stated below.Reason for Violation Rochester Gas and Electric agrees that, Ginna Station does not have an established

written policy regarding consideration

of inherent inaccuracy

of calibrated

measuring and test, equipment (M&TE)when developing

acceptance

criteria.As-a common practice, torquing methods address only instru-ment"indication" and are not meant to include the instrument

accuracy.This practice is based on the fact that torque is only a general indicator of bolting pre-load because of the inaccuracies, e.g., lubrication, thread fit, thread condition, etc., inherent in the torque equation.When highly accurate bolt pre-loading

is required, means other than torque is used, i.e., stud elongation

to determine bolt pre-load.The Corrective

Ste s Which Have Been Taken and the Results Achieved Due to the successful

completion

of post maintenance

testing, no action regarding the valve packing adjustment

has been taken.A-1603.4,"Work Order Scheduling" was revised to require work and testing to be completed on individual

trains prior to starting maintenance

on a redundant train.

'The Corrective

Ste s Which Will Be Taken to Avoid Further Violation 1.Administrative

procedure A-1603.3,"Work Order Planning" will be revised to state a Ginna Station policy regarding consideration

of M&TE inherent inaccuracy

and provide direction for development'f

acceptance

criteria utilizi'ng

this equipment.

2.A new procedure for packing adjustment

is being developed to provide specific direction for adjustment

of valves repacked under the Valve Packing Improvement

Program and to provide a method of maintaining

and updating valve packing data.The Date When Full Com liance Will Be Achieved The anticipated

effective date of the above procedures

is May 1, 1990, for the maintenance

procedures

and June 30, 1990, for the administrative

procedure.

Very truly yours,Robert C.Me dy Division Manager Nuclear Production

GJWN093 Enclosures

xc: U.S.Nuclear Regulatory

Commission (original)

Document Control Desk Washington, D.C.20555 Allen R.Johnson.(Mail Stop 14D1)Project Directorate

I-3 Washington, D.C.20555 Nicholas S.Reynolds, Esq.Bishop, Cook, Purcell and Reynolds 1400 L.Street, N.W.Washington, D.C.20005-3502

Ginna NRC Senior Resident Inspector

~l 0