ML17250B199

Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983

ML17250B199
Person / Time
Site:	Ginna
Issue date:	06/08/1990
From:	MECREDY R C ROCHESTER GAS & ELECTRIC CORP.
To:	MARTIN T T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
NUDOCS 9006200487
Download: ML17250B199 (37)

See also: IR 05000244/1989081

Text

x REGULATOY INFORMATION

DISTRIBUTZOYSTEM (RIDE)x~x ACCESSION NBR:9006200487

DOC.DATE: 90/06/08 NOTARIZED:

NO;C FACIL:50-244

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

CREDY,R,C.

Rochester Gas&Electric Corp.RECIP.NAME

RECIPIENT AFFILIATION

MARTINET.T.

Region 1, Ofc of the Director SUBJECT: Responds to NRC 900509 ltr re violations

noted in Insp Rept 50-244/89-81.

DISTRIBUTION

CODE: IEOZD COPIES RECEIVED:LTR

1 ENCL 3 SIZE: TITLE: General (50 Dkt)-Insp Rept/Notice

of Violation Response NOTES:License

Exp date in accordance

with 10CFR2,2.109(9/19/72).

DOCKET 05000244 05000244 c RECIPIENT ID CODE/NAME PD1-3 PD RGN1 ERNAL: LPDR NSIC INTERNAL: AEOD AEOD/TPAD NRR MORISSEAU,D

NRR/DLPQ/LPEB10

NRR/DREP/PEPB9D

NRR/DST/DIR

8E2 NUDOCS-ABSTRACT

OGC/HDS2 FILE 01 COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME JOHNSONFA AEOD/DEIIB

DEDRO NRR SHANKMAN,S

NRR/DOEA DIR 11 NRR/DRIS/DIR

NRR/PMAS/ILRB12

OE~I REG~02 NRC PDR COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1<1 1~~+NG: P3 fo)579+/NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASCONTACT THE DOCUMENT CONTROL DESK ROOM Pl-37 (EXT.20079)TO ELIMINATE YOUR NAME FROM DISTRIBUTION

LISTS FOR DOCUMENTS YOU DON'T NEED!A D D, TOTAL NUMBER OF COPIES REQUIRED: LTTR 22 ENCL 22

AE;f, Nf e~~~5OAA 55455 ROCHESTER GAS AND ELECTRIC CORPORATION

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

June 8, 1990 TEEER<04C AREA CODE 7555 546 2700 Mr.Thomas T.Martin Regional Administrator

Region I U.S.Nuclear Regulatory

Commission

475 Allendale Road King of Prussia, PA 19406 Subject: Response to Inspection

Report 50-244/89-81

Safety System Functional

Inspect'ion

--RHR System R.E.Ginna Nuclear Power Plant NRC Docket 50-244 Dear Mr.Martin: This letter provides the initial 30-day response to the Safety System Functional

Inspection (SSFI)of the Residual Heat Removal (RHR)System at the R.E.Ginna Nuclear Power Plant, conducted, between November 6 and December 8, 1989.The NRC letter of May 09, 1990 from Marvin W.Hodges (NRC)to Robert C.Mecredy (RG&E)transmitted

the report for that inspection.

This letter provides the RG&E responses, pursuant to 10 CFR 2.201, to the two violations

issued in conjunction

with the SSFI report.In addition, we are providing schedule information

concerning

the unresolved

issues, including the postulated

flooding of the RHR room, identified

in the inspection

report.Additional

information

will be provided in the 120-day response to the SSFI report.The nuclear industry is going through major upgrade efforts involving configuration

management

and design basis documents.

RG&E is not alone in recognizing

the benefits of these improvements

and has been proceeding

with these efforts.On March 6, 1990 RG&E made a formal presentation

to NRC Region I staff and on March 27, 1990 made a presentation

to NRR regarding our configuration

management

program.We have completed three pilot system design basis documents and are reviewing them to determine the optimal specification

for the overall design basis document program for the remaining plant systems.In addition, RG&E has developed a separate program to provide further assurance that all design basis information

and, commitments

which may have been relied upon by the NRC are captured.The objective of the NRC SSFI of the RHR systems was to assess the capability

of that system to perform its design basis functions.

As part of that inspection, the SSFI team assessed the overall design control program and other work processes used by RG&E.The review of these programmatic

aspects was far broader than the RHR system.Special emphasis was placed upon the engineering

processes and their interfaces

with other activities.

900b200487

900b08 PDR ADOCK 05000244 8 PNU ggIJ t~fP(

2The primary result of the SSFI was that no situations

were identified

that would prohibit the RHR system from performing

its intended functions under normal and design basis accident conditions.

As would be expected from an SSFI of any nuclear power plant, and in particular

one of the early SEP plants, the SSFI identified.

areas where improve-ment is warranted.

Two Severity Level IV violations

were cited., and ten specific unresolved

items were documented.

The NRC letter of May 09, 1990 requires that the violations

be addressed, pursuant to 10 CFR 2.201, within 30 days.The letter also requests that RG&E provide its evaluation

of the specific unresolved

items and planned actions, within 120 days.In addition, the NRC letter requests that RG&E also provide schedule information

regarding the actions to address the unresolved

items, within 30 days.The schedules requested are exclusive of unresolved

item 89-81-11, Engineering

Assurance, for which a response was requested in 120 days.k Responses to two violations

identified

ar'e provided as Enclosures

A&B to this letter.The first violation involved.not maintaining

an up-to-date load profile for the batteries.

The actual capability

of the batteries was not an issue, only the adequacy of the testing.RG&E had already reached a state of full compliance

on this matter when the SSFI report was received.>>The second violation cited had two parts.The first part involves having not already developed a periodic testing program for the molded case circuit breakers.The second part involves not having an explicit acceptance

criterion in the test procedure for the setpoints of the dc undervoltage

alarm relays.Although a generally accepted periodic test method for molded case circuit breakers is not available in the industry today, we choose not to take issue with this violation.

The industry is currently examining the need for and/or requirements

for molded case circuit breaker testing.RG&E will implement, when.available, those testing methods and requirements

endorsed by the industry.With regard to the acceptance

criterion for the undervoltage

relay setpoints, we had already reached a state of full compliance

on this matter when the SSFI report was received.In addition, on our own initiative, we have expanded this concern to include the ac undervoltage

relays for the safety buses.In addition to these violations, NRC also identified

ten unresolved

items.The identification

of these items is contained in Enclosure C.Several of these unresolved

items have already been completed and several more are in process.During the RG&E review of the SSFI report, management

recognized

that many of the unresolved

items were examples of broader, underlying, programmatic

concerns.Many of these concerns focused on engineering

functions and, controls'.

Because RG&E understands

the importance

of resolving the programmatic

and management

issues as well as the specific items cited by the NRC, we are developing

a systematic

approach to address both types of concerns.This approach is a two-part, parallel effort.The first part focuses on the management

processes in a disciplined

manner, while the second.part focuses on the resolution

of the specific unresolved

items.

P

To begin the review of the broader concerns, we have re-reviewed

the SSFI report and the cited issues, and have categorized

them into general topical areas.For example, unresolved

item 89-81-05 involves not having a mechanism to assure that design calculations

are main-tained up-to-date.

We see this specific item as being part of a more general area called design control.Enclosure D is a preliminary

categorization

of the unresolved

items into the general topical areas.In addition, RG&E is initiating

a more detailed review of the work processes and their controls for each of the general areas which contain significant

identified, weaknesses.

This review will encompass identifying

the cause of the violations, as well as the unresolved

issues, identified

by the SSFI report.Enclosure E contains the schedular information

as requested by the staff.We have separated this schedule information

into two catego-ries: resolution

completed and scheduled for resolution.

RG&E has resolved items 89-81-04, 06, 07A, and 10 as identified

in Enclosure C.In particular, RG&E has promptly resolved the issue regarding flooding of the RHR pump room.The UFSAR has been updated, and the EOPs and training documents have been revised.A detailed account of those actions taken to resolve the items identified

above are con-tained in Enclosure E.RG&E believes that the approach outlined in this letter assures proper and complete resolution

of the specific issues identified

as well as the more programmatic

issues discussed.

Very truly yours, Robert C.Mec e y Division Manager Nuclear Production

GAHN108 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 Ginna NRC Senior Resident Inspector

0

ENCLOSURE A Response to Notice of Violation 50-244/89-81

Violation 1

.Ins ection Re ort 44/88-81 VIOLATION 1:STATEMENT OF VIOLATION-

10 CFR 50, Appendix B, Criterion III, requires in part that measures be established

to ensure that applicable

regulatory

requirements

and design bases are translated

into specifications

and procedures.

These measures shall provide for verifying the adequacy of design by performance

of design reviews.Ginna Station Quality Assurance Manual, Section No.11,"Test Con-trol," requires that engineering

establish design test requirements

and that testing be performed in accordance

with approved procedures

which incorporate

the requirements

and acceptance

criteria contained in applicable

Technical Specifications

and regulatory

requirements.

Contrary to the above, on November 15, 1989, the design reviews for Engineering

Work Request (EWR)3891 were inadequate

in that the EWR did not establish the battery load, requirements

thereby resulting in a battery load.profile used during the service test.not reflecting

the design basis load requirements.

This is a Severity Level IV Violation (Supplement

1).ACCEPTANCE

OF VIOLATION:

RG&E agrees that it did not update the battery load profile as part of EWR 3891.DISCUSSION:

The purpose of EWR 3891 was to replace the batteries because they were nearing the end of their service life and, while replacing them, to increase the capacity margin.EWR 3891 did not include an updat-ing of the battery test profile because it had been determined

that no large loads had been added.to the battery since the original load profile had been developed.

The battery load profile was based upon the original Westinghouse

design data.That information

was consistent

with industry practice at the time it was developed.

Analytical

techniques

were not as sophisticated

as those in use today.Rather than explicitly

quanti-fying such factors as momentary loads and.the load starting currents, it was general practice to provide additional

battery sizing based upon experience

and engineering

judgement.

Today's standards (such as IEEE standard 485)suggest a more refined, more precisely quanti-fied analysis.The actual battery capacity was sufficient

to provide its safety functions.

The battery has been shown to have adequate capacity as confirmed by a physical test.Although there is no requirement

for the Ginna Nuclear Power Plant to incorporate

all newly-developed

industry standards, we believe it prudent to use the current industry standards for developing

revised battery load profiles, and have done so.A-1

CORRECTIVE

STEPS T A preliminary

analysis, performed during the inspection, demonstrated

that the battery size is adequate.The revised battery size calculation

had, been finalized subsequent

to the NRC inspection

and prior to the receipt of the inspection

report, which confirms that the battery size is not a concern.An improved battery load profile has been developed which incorpo-rates calculational

improvements

contained in current industry standard IEEE 485-1983.The upgraded battery load profile (Design Analysis EWR 3341"Sizing of Vital Batteries", dated March 12, 1990)has been transmitted

by Engineering

to the plant staff, and the battery testing procedures'PT-10.2

and PT-10.3, Battery Service Tests)have been revised.The batteries were tested during the recent outage using the revised procedures.

The results demonstrated

the adequacy of the battery capacity.CORRECTIVE

STEPS TO BE TAKEN TO PREVEBVP REClJRRENCE:

The applicability

of this violation has been broadened by RG&E to assure that not only the important dc electrical

loads are analyzed and tested, but also that the important ac electrical

loads which may impact the operation of the plant emergency diesel generators

are identified

and tracked.We have implemented

an electrical

load'rowth

program as described under unresolved

item 89-81-05.DATE WHEN FULL COMPLIANCE

WILL BE ACHIEVED: Engineering

established

updated battery load requirements.

The battery test procedures

have been revised and the batteries have been tested using the new procedure.

These actions were completed prior to the receipt of the NRC inspection

report.RG&E is in full compli-ance.A-2

ENCLOSURE B Response to Notice of Violation 50-244/89-81

Violation 2

I P

~RG&E/Ginna

Ins ecti Re ort 50-244/89-81

VIO N 2: STATEMENT OF VIOLATION:

R.E.Ginna Technical Specifications Section 6.8.1 requires that written procedures

be established

and.implemented.

for activities

such as surveillance

and testing activities

of safety-related

equipment.

Ginna Station Quality Assurance Manual,Section II,"Test Control," establishes

the requirements

for establishing

and implementing

test programs to demonstrate

that safety-related

systems and components

will perform satisfactorily.

Furthermore, this section requires that testing shall be performed in accordance

with written procedures

which incorporate

acceptance

criteria.Contrary to the above, on December 9, 1989, Class 1E 480V ac molded case circuit breakers have not been subjected to scheduled periodic testing.Furthermore, there is no established

acceptance

criteria for testing the dc undervoltage

relay al'arms in Procedure PT-11,"60-Cell Battery Banks'AB'his is a Severity Level IV Violation (Supplement

1).ACCEPTANCE

OF VIOLATION:

RG&E agrees that the periodic testing program of safety-related

equipment at the Ginna.Nuclear

Power Plant does not currently include molded case circuit breakers.RG&E also agrees that the Ginna periodic test procedure PT-11"60-Cell Battery Banks'AB'" did not specify an acceptance

criterion for the setpoint of the dc undervoltage

relay alarms.This violation has two parts which are addressed separately

below: Part 1: Molded Case Circuit Breaker Testing DISCUSSION:

Molded case circuit breakers are designed for nuclear and non-nuclear

applications.

This type circuit breaker is sealed.and does not include design features to test all the capabilities

of the breaker beyond functional

tests.RG&E realizes the importance

of assuring proper operation of these breakers.RG&E has not been lax in its attention to the importance

of testing molded case circuit breakers.This problem was self-identified

by RG&E and was incorporated

into the RCM program.On our own initiative, we developed and implemented

receipt-inspection

testing for all new molded case circuit breakers at Ginna.We have also performed testing on molded, case circuit breakers in an effort to determine their characteristics.

Three years ago, RG&E performed special testing of all of its exist-ing magnetic only, molded case circuit breakers at Ginna Station on a special one-time basis.Successful

operation has indicated no known degradation.

B-1

While the functioniO~

of molded case circuit hkers is important to safety and while there is an NRC requirement

for a test program to assure that safety-related

structures, systems and.components

will perform satisfactorily, there is no specific requirement

to test periodically

every piece of equipment.

As stated.in Appendix B, Criterion XI,"The test program shall include, as appropriate, operational

tests...of structures, systems and components." The term"as appropriate" is applicable

and includes the availability

of appropriate

test methods.Molded case circuit breakers are not designed for in situ testing and would require determination

and retermination

to perform the testing.The vendors of this equipment have also not made recommendations

for periodic testing.Because of generic applicability, periodic testing for molded case circuit breakers has been an industry-wide

issue and no generally accepted test method has been developed at this time.The nuclear industry has responded to the NRC through NUMARC concern-ing molded case circuit breaker testing and RG&E is pursuing this in conjunction

with this effort.CORRECTIVE

STEPS TAKEN: RG&E is continuing

to work toward developing

appropriate

test methods for molded case circuit breakers, as part of the Reliability

Centered Maintenance (RCM)program.The Ginna Nuclear Power Plant is one of the two"pilot plants" in the nation for the EPRI sponsored RCM program.CORRECTIVE

STEPS TO BE TAKEN TO PR1DGQFZ RECURRENCE-

The industry is currently examining the need for, and benefits of, molded case circuit breaker testing.RG&E will continue to work closely with the industry and EPRI to determine appropriate

test methods and.requirements.

DATE WHEN FULL COMPLIANCE

WILL BE ACHIEVED: Although RG&E does not consider this a compliance

matter, RG&E will implement, when available, those testing methods and requirements

endorsed by the industry.Part 2: Undervoltage

Relay Alarm Acceptance

Criteria CORRECTIVE

STEPS TAKEN: The periodic test procedure PT-11"60-Cell Battery Banks'AB'" has been revised to explicitly

define the acceptance

band/criterion

for the dc undervoltage

alarm relays.The dc relays have subsequently

been calibrated

and tested.The relays have been verified to perform within the specified acceptance

criterion.

B-2

~1 0

.CORRECTIVE

STEPS TAKEN TO PREVIXT RE E: The applicability

of this violation has been broadened by RG&E to assure that not.only the test procedures

for dc undervoltage

alarm relays have explicit acceptance

criteria, but also that the test procedures

for the ac undervoltage

relays for the safeguards

buses have explicit acceptance

criteria.The test procedure, PT-11 for the dc undervoltage

alarm relays has been revised and PT-9.1 for the 480V ac safeguards

buses is being revised to provide explicit acceptance

criterion.

DATE WHEN FULL COMPLIANCE

WILL BE ACHIEVED: The test procedure for the dc undervoltage

alarm relays has been revised to provide an explicit acceptance

band/criterion.

This action was completed prior to the receipt of the NRC inspection

report.RG&E is in full compliance.

B-3

0 t'

ENCLOSURE C Identification

of Specific Unresolved

Items Note: The statements

of issues have been directly extracted from the SSFI report.In a few instances the issues have been condensed and paraphrased.

r 0 0 e

.89-81-01 Service r Single Failure Suscept lity Potential loss of cooling water[flow]to both emergency diesel generators

during or following a seismic event.The cooling water for the water jacket heat exchanger and lube oil heat exchanger discharges

through a common non-safety

non-seismic

10-inch discharge pipe.The cooling water discharge pipe would have to fail[or has been postulated

by the NRC SSFI team to fail]so as to prevent[block/pinch

off]the flow of the service water.89-81-02 Resolution

of Safety Concerns The licensee was unable to provide the team with a documented

or verifiable

process available at RG&E that addresses how safety concerns raised outside the normal engineering

process are brought to the attention of the Nuclear Safety and Licensing staff and resolved.89-81-03 RHR Pump NPSH A consultant

independently

evaluated the available NPSH during post-accident recirculation

mode from the containment

sump and, a prelimi-nary result indicates that there may be some modes of operation of the RHR pumps under which adequate NPSH is not available.

Licensee is evaluating

the validity of these modes and the probability

of occurrence.

Licensee is also evaluating

the possibility

that the consultant's

analytical

model was too conservative.

89-81-04 Class 1E Battery Testing Failure to test the batteries with a load profile which truly repre-sented the load demand on the battery is considered

a violation of 10 CFR 50, Appendix B, Criterion III.89-81-05 Electrical

Load Growth Control Program RG&E does not have a mechanism to assure that plant calculations

affected by modifications

are updated to ensure that they are main-tained up-to-date

and accurate.The design process provides guidance to engineers to review the system capacity and other attributes, but the guidance addresses only specific modifications

as they are performed.

There is no formal load tracking system to ensure that system capacity is reviewed for the integrated

effect of several modifications

instead of just one.The licensee stated that an on-line program to capture electrical

load growth and update affected calculations

would be developed.

89-81-06 Molded Case Circuit Breakers and, Undervoltage

Relay Alarms Failure to periodically

test the molded case circuit breakers and not establishing

an acceptance

criteria for the undervoltage

relay alarms are a violation of facility Technical Specifications 6.8.1, which requires testing of safety-related.

components

in accordance

with established

procedures.

.89-81-07A

Calibrate of Control Room Instrume The control room dc voltmeters

are not calibrated

on a periodic basis to ensure reliable system voltage indication

to operators.

89-81-07B Control Room P&IDs Piping and Instrument

Diagram (P&ID)updates and Design Change Requests (DCRs)posted in the control room were reviewed by the team.It was noted that the RHR system P&ID (33013-1247)

did not reflect the current valve position configuration

for the RHR system.Also, the existing DCRs outstanding

against this drawing could not be used.to derive the correct valve positions in that DCRs 1247-4, and 1247-5 had not been approved by RG&E Engineering

and did not reflect the current position of valve 822B.Processing

of DCRs does not always occur in a timely manner such that the control room P&IDs can be immediately

updated.Plant operations

organization

makes permanent changes to system valve positions, there is not an immediate markup or annotation

made on the effected draw-ings.The team noted that permanent changes to valve positions in system operating procedures

are occurring without the prior concurrence

of RG&E engineering.

UFSAR, sections 5.4.5.3.5 and 5.4.5.2, refers to two remotely operat-ed valves which can be utilized to isolate an RHR loop from outside the pump room.The system walkdown and the upgraded P&IDs indicate that there is no longer any method available to isolate an RHR loop remotely (i.e., via reach rods).Although this information

has been removed from the RHR P&ID, there is no identified

punchlist item to delete this information

from the UFSAR.The team noted that uncontrolled

training material (Lesson Texts)have not been updated to reflect system changes accomplished

during the last outage.There is no station requirement

to maintain this training material current.The inspection

team considers that making this type of information

available to control room operators in such an uncontrolled

manner represents

a notable program weakness.The lack of timely operating information

updates for control room use is considered

an unresolved.

item.89-81-08 Equipment Environmental

Qualification

Evaluation

The NRC questioned.

the basis for the assumption

that RHR pump seal failure will occur after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.The NRC requests RG&E to sub-stantiate the method of detecting any leak in the RHR pump room if the pump seal were to fail before the stated 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.C-2

The safety relief valve test procedures

contain general and minimal instructions

for performing

the relief setpoint test.Standard test practices are not always performed.

or documented.

As written, the test procedure requires only one successful

setpoint test.Data from relief valve testing has been recorded inaccurately

and.inconsistent-

ly in some cases.The NRC concluded that RG&E should formalize test procedures

instructions

and data recording requirements.

During the on-going procedure upgrade effort, RG&E should assure that valve test procedures

incorporate

all new (1986)ASME Code Section XI, IWV-3512, and ANSI/ASME OM-1-1981 requirements

for safety relief valves.In particular, more than one successful"pop test" at the designated

lift pressure should be performed and the results comp-letely and accurately

documented.

Valve setpoint and leak testing should also be performed with the allowable specification

listed in the procedure.

Valve test results and data should accurately

reflect'he results of all test activities.

RG&E should also consider the benefits of adding other periodic valve tests such as the as-found relief lift setpoint, valve accumulation, and.valve capacity.89-81-10 Translation

of FSAR Requirements

into Operating Procedures

The Ginna UFSAR contains"operational" information

and data which the inspectors

determined

to be invalid and, without a supporting

design basis., Specifically, Section 5.4.5.3.5 states that in the event of a 50 gpm RHR pump seal leak and loss of both pump room sump pumps, operators have 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to isolate the leak before the RHR pump motors become flooded.The team determined

that a 50 gpm leak into the pump room, with two failed sump motors;cannot be sustained in the RHR pump room for four hours before flooding the pump motors.It was suggested.that the four hour allowance was originally

intended just to indicate a rough system margin for coping with gross leakage in the pump pit.The team was unable to find any consideration

of this in any of the available design documents associated, with the RHR system.It also could not be found in any of the system operating or emergency procedures.

The alarm response procedure for the high sump level alarm requires control room operators to dispatch an auxiliary operator to investigate

possible pump room flooding, however there is no reference to maximum time limit to isolate a leaking RHR train if necessary.

The team reviewed the instrumentation

devices available to control, room operators which would indicate RHR leakage in the pump room.The only known indication

would be from a high level sump alarm.However, the sump alarm instrument

is not qualified for service in a harsh environment.

Operating procedures, emergency procedures, and operator.training material do not reflect the limiting design basis of the system.The apparently

unsupported

4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> flooding limit is considered

an un-resolved item pending verification

of the value by the licensee or correction

of the UFSAR.C-3

O 89-81-11 Engineer'ssurance

The design control measures as implemented/practiced

by the licensee's

engineering

department

were weak, and did not favorably compare to good engineering

assurance practices generally accepted in the industry.There was lack of consistency

in the implementation

of approved engineering

procedures

among the various departments

and engineering

management

did not appear to be cognizant of this incon-sistency.There was a lack of formal design interface control, lack of control over external communication

with design consultants, and a lack of control over design documents/modification

packages during the development

and implementation

phase.C-4

ENCLOSURE.D Preliminary

Categorization

of Issues Note: The categories

contained in Enclosure D were selected topics in 10 CFR 50 Appendix B and other sources.To begin the review of the broader concerns, we have reviewed the SSFI report and the cited issues, and have categorized

them into general topical areas.For example, unresolved

item 89-81-05 involves not having a mechanism to assure that design calculations

are maintained

up-to-date.

We see this specific item as being part of a more general area called design control.Enclosure D is a preliminary

categorization

of the unre-solved items into the general topical areas..It is currently planned to categorize

all the concerns identified

in the inspection

report.

DESIGN CONTROL General Control of Design Inputs Control of Design Process SSFI URI 89-81-05: SSFI URI 89-81-08: Electrical

Load Growth Con-trol Program Equipment Environmental

Qual-ification Evaluation

Control of Design Outputs SSFI URI 89-81-07B:

Control Room P&IDs Control of Design Interfaces

and.Coordination

Control of Design Changes Design Reviews/Engineering

Assurance SSFI URI 89-81-05: Electrical

Load Growth Control Program SSFI URI 89-81-11: Engineering

Assurance Specific Design Concerns SSFI URI 89-81-01: Service Water Single Failure Susceptibility

SSFI URI 89-81-03: RHR Pump NPSH PROCEDURES

SSFI URI 89-81-09: Safety Relief Valve Testing DOCUMENT CONTROL SSFI URI 89-81-07B:

Control Room P&IDs

0

ORGANIZATIONAL

.ACES SSFI URI 89-81-02: Resolution

of Safety Concerns SSFI URI 89-81-07B:

Control Room P&IDs SSFI URI 89-81"10: Translation

of FSAR Require-ments into Operating Proce-dures HANDLING OF SAFETY CONCERNS SSFI URI 89-81-02: Resolution

of Safety Concerns SURVEILLANCE

TESTING MAINTENANCE

SSFI URI 89-81-07A:

Calibration

of Control Room Instruments

SSFI URI 89-81-09: Safety Relief Valve Testing D-2

ENCLOSURE E Resolution

of Specific Issues Note: We have separated.

the schedule information

contained in this enclo-sure into two categories:

resolution

completed, and schedule for resolution.

Listed first are those items for which RG&E has complet-ed resolution.

Those measures taken by RG&E are identified.

Some of the unresolved

items listed cannot be adequately

resolved, without addressing

the broader more programmatic

issues such as design control and engineering

assurance and require more time to resolve than the specific items.The schedules provided for some items may change as RG&E further identifies

the underlying

concerns.An updated schedule will be provided in the 120 day response.

0

~Resolution

Com lete~89-81-04 Class 1E Battery Testing This item was resolved prior to receipt of the SSFI report.see Enclosure A for actions taken for resolution.

Please 89-81-06 Undervoltage

Relay Alarms and Molded Case Circuit Breakers Please see Enclosure B for actions taken for resolution.

89-81-07A Calibration

of Contxol Room Instruments

This item was resolved prior to receipt of the SSFI report.The actions taken to resolve this issue include: 1)Calibration

of all control room dc bus voltmeters

during the recent refueling outage (the voltmeters

were found to be within the specified acceptance

criteria).

2)All dc bus voltmeters

are now calibrated

per Calibration

Proce-, dure CP-514 on an annual basis.3)All emergency diesel generator and various secondary system power meter calibrations

have been added to the CP-500 series procedures, and the meters were calibrated

during the 1990 refueling outage.89-81-10 Translation

of the FSAR Requirements

into Operational

Procedures

This item was resolved promptly.The actions taken to resolve this issue include: 1)Performance

of a reanalysis, during the SSFI inspection, which determined

that operators have two hours to respond.(Design Analysis, 10CFR50.59

Safety Evaluation, NSL-0000-015, Rev.0, dated December 8, 1989, Residual Heat Removal Leakage Provi-sions.)T 2)Update of UFSAR sections 5.4.5.3.5, 5.4.5.2 and 6.3.3.8, submit-ted as part of the UFSAR update on December 16, 1989.3)Revision of Training System Description

RGE-25 during the inspection.

4)Revision of EOPs prior to receipt of the inspection

report.(Procedure

E-1, Loss of Reactor or Secondary Coolant, Step 18 was added and ES-1.3, Transfer to Cold Leg Recirculation, a note before Step 9 was added.)

.Schedule for Resolu.n 89-81-01 Service Water Single Failure Susceptibility

As noted in the inspection

report, the failure of the 10 inch dis-charge line in a manner which would stop service water flow to the diesel generators

is a low probability

event.This event is also beyond the design and licensing basis of the plant.Nevertheless, RG&E plans to further evaluate the potential risk of this scenario during the PRA/IPE effort.Our IPE is currently scheduled to be submitted in the third quarter of 1991.89-81-02 Resolution

of Safety Concerns An interim process for handling safety concerns is under-development

and will be discussed in our 120 day response.89-81-03 RHR Pump NPSH Documentation

of the analysis findings is scheduled to be completed by December 31, 1990.In addition, RG&E plans to consider this matter in the PRA/IPE.89-81-05 Electrical

Load Growth Program RG&E has implemented

an interim process for all modifications

to perform the following actions: Current system loadings for the dc batteries have been estab-lished in Design Analysis, EWR 3341, Sizing of Vital Batteries, and for the diesel generator loads in Design Analysis, EWR 4136, Diesel Generator Loading.2)An Electrical

Engineering

Design Guide, Electrical

Interface Checklist EDG-15D, Rev.0, is being implemented

on all modifica-tions which requires identification

of load changes to the dc batteries and the diesel generator ac loads.3)A process controlled

by Electrical

Design Guide, Design Verifi-cation Model EDG-15B, Rev.0, has been established

within the Electrical

Engineering

Design Verification

Group which updates the loading data for the impacted power supply and determines

the remaining capacity margin for ac and dc loads.We are taking actions to integrate this process into the appropriate

Engineering (QE)procedures.

We anticipate

completion

of these actions by the date of our 120 day response.-

89-81-07B Control Room P&IDs RG&E has considered

the examples identified

by the staff which resulted in the staff's conclusion

that information

updates for control room use are not implemented

in a timely manner.RG&E has resolved several of the examples identified.

These include: E-2

C 4'C 0

.1)RG&E has impleilhted

improved controls inDrawing Change Request (DCR)process.RG&E has assigned Z" Station Engineer with responsibi

lity for tracking and processing

all DCRs.~~2)The UFSAR has been reviewed to assure that the appropriate

information

with regard to the isolation of the RHR pump seal is correct.3)RGGE has revised the lesson text to reflect the revised RHR'pump seal leakage time limitation

of two hours.An interim process for enhancing the update process for control room information

is currently under review and will be discussed, in the 120 day response.89-81-08 Equipment Environmental

Qualification

Evaluation

The passive failure'of a RHR pump seal is assumed to occur at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, consistent

with SRP 15.6.5.The consequences

of this assumed passive failure, concurrent

with the assumed design basis LOCA, was evaluated, by the NRC during the review of SEP Topic XV-19 and found to be acceptable.

Nevertheless, RGGE plans to further evaluate this scenario during the PRA/IPE effort with its attendant requirement

to perform an internal flooding analysis.Our IPE is currently sched-uled to be submitted in the third quarter of 1991.The results of this evaluation

will determine if the upgrade of the sump level switches to a safety-related

status is recommended.

89-81-09 Safety Relief Valve Testing and, Documentation

RGGE has commit/ed to incorporate

ASME Code Section ZI-IWV-3512

(1986)and implement ANSI/ASME OM-1-1987 as part of the IST Program Upgrade.Procedure changes to incorporate

these requirements

were completed.

prior to receipt of the SSFI report.RGfiE will have completed.

all testings under these new requirements

by December 31, 1994.E-3

e