ML17262A133

Responds to Violations & Several Unresolved Items Noted in SSFI Rept 50-244/89-81.Update of Appropriate Unresolved Items Encl.Specific Actions Re All NRC Unresolved Items Being Tracked to Completion

ML17262A133
Person / Time
Site:	Ginna
Issue date:	09/11/1990
From:	MECREDY R C ROCHESTER GAS & ELECTRIC CORP.
To:	MARTIN T T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
NUDOCS 9009200182
Download: ML17262A133 (2)

See also: IR 05000244/1989081

Text

k'L ROCHESTER GAS AND ELECTRIC CORPORATION

ROBERT C htECREDY Vi<e hetident Cinne t4uclee<Ptoduction

/'.tone>t*te~89 EAST AVENUE, ROCHESTER N.Y.14649-0001

TELEPHONE AREA CODE 71B 546 2700 September 11, 1990 Mr.Thomas T.Martin Regional Administrator

U.S.Nuclear Regulatory

Commission

Region I 475 Allendale Road King of Prussia, PA 19406'ubject:

120-day Response to Inspection

Report 50-244/89-81

Safety System Functional

Inspection

on the RHR System R.E.Ginna Nuclear Power Plant Docket No.50-244 Reference: (a)NRC Inspection

Report 50-244/89-81, dated May 9, 1990(b)RG&E letter from R.C.Mecredy-to NRC, T.T.Martin, dated June 8, 1990 Dear Mr.Martin: Reference (a)requested a response to two Notices of Violation and several unresolved

items within 30 days and a written evaluation

of the deficiencies

identified

in Section 2.1 of the Inspection

Report within 120 days.In our 30-day response to two Notices of Violation, Ref.(b), we summarized

our proposed resolution

or schedule for resolution

for the unresolved

items 89-81-01 through 10.The NRC unresolved

item description

and our proposed resolution

of each of these was discussed in enclosures

C and E of Ref.(b).An update of the appropriate

unresolved

items is provided as Attachment

A to this response.Specific actions regarding all NRC unresolved

items are being tracked to completion

by RG&E.~U~cIQ et\A oo 1'Q~C3 oP o+The identified

weaknesses

in Section 2.1 of the Inspection

Report collectively

raised an NRC concern as to the effectiveness

of RG&E's current practices to establish engineering

assurance.

This was identified

as unresolved

item 89-81-11 and is the focus of this response.RG&E recognizes

that unresolved

item 89-81-11, engineering

assurance, is characterized

by broad programmatic

issues.RG&E has completed an evaluation

of the deficiencies

and concerns raised in the Inspection

Report by performing

an internal assessment

of the underlying

issues identified

in Section 2.1 and the examples discussed in Section 2.2.Our assessment

has been documented

in an g Q~o//(]

Cl

internal report entitled"Systematic

Assessments

of Engineering

Assurance Issues'and RHR SSFI Concerns" dated 9/11/90.RG&E believes that the underlying

concerns necessitate

both interim.and

long term activities

to resolve.Our approach in performing

the-internal assessment

and a summary of the high priority actions are presented in Attachment

B.This attachment

is a summary of the considerable

efforts of an RG&E SSFI Assessment

Team composed of a group of experienced

RG&E staff and management

personnel.

The primary task of'the RG&E SSFI Assessment

Team was to prepare a report to RG&E's management

which recommended

the most effective interim actions needed to begin the process of strengthening

the engineering

processes and.controls.The RG&E SSFI Assessment

Team was composed of nine senior engineers and staff.The team met to re-examine

the inspection

report and categorize

the deficiencies

by topical areas.The team also evaluated a report prepared by an RG&E consultant

who independently

identified

the programmatic

concerns.The assessment

consisted of individual

evaluations

by team members as well as working sessions as a group.The RG&E Assessment

Team grouped the NRC identified

deficiencies

into the following topical areas:~Improved~Improved~Improved~Improved~Improved~Improved Method of Identifying

and.Assessing Safety Concerns Design Control and Reviews Design Interface Control Documentation

Associated

with Design Bases Documentation

Associated

with Modifications

Engineering/Plant

Communications

The team then established

interim actions and long term corrective

actions for each topical area.The team prioritized

the interim actions and established

a proposed schedule.The interim actions were recommended

based upon achieving a fundamental

improvement

on the engineering

process.Interim actions are those actions which can be implemented

immediately

or within a period.of up to a year.Long-term corrective

actions were also recommended,.

Many of the longer term recommendations

are already embodied in two major programs: Configuration

Management (CM)and the Engineering

Procedures

Upgrade Program.A report on the Configuration

Management

program descriptions

and schedule was presented to members of Region I and NRR on March 6 and March 27, 1990, respectively.(Individual

projects within the CM program may be examined within the enclosure to Inspection

Report 90-03, dated April 18, 1990).The Engineering

Procedures

Upgrade program has been initiated and will include an external assessment

of our current procedures

by an independent

consultant.

It is expected that this assessment

will be complete by year-end.The results of RG&E's internal SSFI Assessment

will become an input to the Engineering

Procedures

Upgrade and the Configuration

Management

Programs.Attachment

B is a summary report of the Assessment

Team activities

and recommendations.

RG&E believes that many of the deficiencies

noted under the engineering

assurance unresolved

item 89-81-11 had been recognized

prior to the RHR SSFI and have been enveloped under the various Configuration

Management

Projects, such as the Setpoint Verification

Program and Design Basis Documentation

Projects.We have recognized

that interim actions are necessary to sufficiently

strengthen

the engineering

processes, procedures

and documentation

of information

to bridge the gap to these longer term programs.We have begun the process to implement these actions and plan to examine their effectiveness, during 1991.We believe that the interim actions planned are the most effective measures for RG&E to provide adequate resolution

of the identified

deficiencies

while we are implementing

the long term Configuration

Management

Programs.The NRC Inspection

Report identified

the Engineering

A'ssurance

deficiencies

in sections 2.2.1.1, 2.2.1.2, 2.2.3.2, and 2.2.3.3.The RG&E assessment

concentrated

on the root causes of these deficiencies

and not just the examples themselves.

Nevertheless, the specific analyses, reports, and drawings that require revisions or corrections

as described in these sections are being revised or updated as necessary, including for example, the calculations

discussed in section 2.2.1.2 and the drawings identified

in section 2.2.3.3(B).

, Our specific evaluations

relative to the examples found in these four sections are contained in the Assessment

Team report.We believe that the systematic

assessment

discussed in Attachment

B is a thorough and appropriate

response to unresolved

item 89-81-11.Please notify us if you believe we have not interpreted

the NRC report correctly.

Very truly yours, GAHK118 Atta'chment

Robert C.Mecredy xc: U.S.Nuclear Regulatory

Commission (original)

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

I-3 Washington, D.C.20555 Ginna Senior Resident Inspector

Attachment

A UPDATE OF UNRESOLVED

ITEMS FROM RG&E's ENCLOSURE E OF, JUNE 8, 1990 URI 89-81-02 (Section 2.2.1.4)Resolution

of Safety Concerns June 8, 1990: An interim process for handling safety concerns is under development

and will be discussed in our 120 day response.Update: RG&E has developed a formal process in Procedure QE-1603, Documenting

and Reporting of Conditions

Adverse to Quality, for handling potential safety concerns identified

by Nuclear Engineering

Services personnel.

These among the conditions-which would be reported, are: nonconformances, deviations, deficiencies, failures, malfunctions, defective material and equipment, vendor technical reports, design basis documentation, and the material condition of the plant structures, systems or components.

The procedure requires that potential safety concerns be documented

and tracked by Nuclear Engineering

Services personnel.

The process provides for: ensuring that a preliminary

safety evaluation

is performed by Nuclear Safety and Licensing;

transmitting

information

concerning

conditions

that involve a safety concern to the Technical Manager, Ginna Station;disposition

of safety concerns through the appropriate

process such as a nonconformance

report (NCR), and.identified

deficiency

report (IDR);reviewing the condition and preliminary

safety evaluation

by Ginna Station Technical Section against the criteria for reporting events (A-25.1);dispositioning

the condition through the appropriate

process such as corrective

action reports (CAR), procedure change notice (PCN)and work request/trouble

request (WR/TR);providing the initiator with the feedback on the disposition.

Potential conditions

adverse to quality that are discovered

by Ginna Station personnel are dispositioned

by one of the current processes under the Maintenance

Work Request and Trouble Report (A-1603), Corrective

Action Report (A-1601), and Reporting of Unusual Plant Conditions (A-25).Interim and long term corrective

actions recommended

as part of the RG&E SSFI Assessment

are described in Attachment

B under the general topical area Improved Process for Reporting and Assessing Safety Concerns.A-1

URI, 89-81-05 (Section 2.2.2.2)Electrical

Load Growth Program June 8, 1990: 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.Update: RG&E issued a change to engineering

procedure QE-301 Rev.11 with issuance of PDR 0609 dated 7/9/90.This change requires that our design process ensure that the effects of all load changes on the station batteries or diesel-generators

shall be addressed, including the requirement

that these be evaluated and shown to be within the margin allowed by the current loading analysis.During the RG&E SSFI Review Team Assessment

it was noted that other examples were identified

which could be placed within the issue of establishing

a mechanism to evaluate the cumulative

effects of modifications.

Interim and long term corrective

actions are described within Attachment

B under the general topical area Improved Documentation

Associated

with Modifications.

URI 89-81-07 (Section 2.2.4.4.a., b., and d.)Control Room P&ID's June 8, 1990: 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.Update: This concern was manifested

in two areas, drawing change requests (DCRs)and training material.DCR Process: The timeliness

of processing

drawing changes has been enhanced through implementation

of Revision 3 of A-606, Drawing Change Requests procedure.

This upgrade directs the timely upgrading of drawings used in the Control Room and Technical Support Center.Posting of drawing changes is required within 2 working days from their receipt by Central Records.(General practice has been same day posting).The approval and tracking of the DCR process is currently assigned to the Technical Section at Ginna Station.Plans are being made to transfer data entry, control of the database and distribution

to the Document Control department.

Another enhancement

includes a monthly DCR status report that is distributed

to management.

The DCR process has been given increased emphasis through procedural

contr'ols.

With its present method, RG&E believes that timely posting and effective tracking and trending of the DCR system will be achieved..

A-2

4

Trainin Material: RG&E is committed to ensuring that all training material available to Licensed operators is as correct and current as possible.During the inspection

we agreed that the Lesson Text RG&E-25 contained an invalid value for the time available to isolate a 50 gpm seal leak to prevent RHR pump motor flooding.Other information

relative to plant modifications

on valve:numbers (EWR 4761)and piping modification (EWR 4675)also had not yet been incorporated

into the Lesson Text because the Training Change Request had not yet been implemented.

After the discrepancy

on the isolation time was identified

to RG&E, the Lesson Text was immediately

corrected.

For clarity, the Lesson Texts have been renamed TRAINING SYSTEMS DESCRIPTIONS.

However, these are not defined as Controlled

Configuration

material.These documents are not meant to take the place of approved plant procedures, engineering

design documents, plant drawings or vendor technical manuals.They are used as reference material, arranged by system, that contain a conceptual

overview of.that system designed to be used.as a job and.training aid.We understand

the NRC's concern over the fact that this material is available for use in the control room, may be frequently

used, and is not designated

as being potentially

out of date.We believe the primary concern is the timeliness

of updating this material to reflect the plant configuration, not over the control over these documents.

Therefore, strict control has been placed over these documents.

There is a master copy controlled

b'y the Training Department

and all changes are controlled

by procedure TR 5.9 (Training Change Request/Notice).

Records are kept of controlled

copy holders of controlled

copies.Placement of this material in the control room is controlled

by the Training Department.

This is the norm, not the exception, in the utility industry.We understand

and recognize the underlying

concern identified

over the timeliness

of providing current and controlled

material to those who may use it.Procedures

TR 5.5.1 (Tracking Plant Changes)is currently the process that has been developed and implemented

to identify and track plant changes and include those changes in training material where appropriate.

We have taken additional

steps in order to provide training material that is as current as possible and to better define the purpose of this material.1.Place a copy of the"Information

Letter" in all Training System Descriptions

that are affected by a plant modification.

This action has been implemented

and will be controlled

by Configuration

A-3

2.Management

Training Guidelines, CMTG-3.0, Preparation

and Use of Information

Letters.The information

letter provides current information

on a modification.

The letter will remain part of the system description

until the system description

has been revised.to incorporate

the new modification.

I The first page of each Training System Description

will be stamped TRAINING INFORMATION

ONLY.This action has been initiated..

3.Include the date of revision for each page of the Training System Description.

This action has been implemented.

There were no additional

interim remedial actions recommended

as part of the RG&E SSFI Review Team Assessment

other than those above.Long term corrective

actions are described in Attachment

B under general topical area Plant Design Information/Design

Bases.URI 89-81-06 (Section 2.2.2.3)Molded Case Circuit Breaker June 8, 1990: The industry is currently examining the need for, and benefits of, molded case circuit breakers testing.RG&E will continue to work closely with the industry and EPRI to determine the appropriate

test methods and requirements.

Update: During August RG&E personnel from Engineering

and Plant Maintenance

visited the Diablo Canyon Power Plant to inspect the equipment and procedures

for periodic testing of molded case circuit breakers used by PG&E.Results of the first cycle of testing by the Diablo Canyon staff were also discussed.

A similar program for Ginna Station appears to be technically

feasible, subject to additional

evaluation

and procurement

of test equipment, development

of procedures, and performance

of a trial test program.It is anticipated

that a trial program can be initiated within the next year.It is estimated that the first cycle of a test program would require four or more years to complete following the successful

completion

of a trial test.A-4

Attachment

B SYSTEMATIC

ASSESSMENT

OF ENGINEERING

ASSURANCE ISSUES AND RHR SSFI CONCERNS This attachment

is a summary of the SSFI Assessment

Team approach and recommended

actions extracted.

from the RG&E report with the same name.

Systematic

Assessment

of Engineering

Assurance Issues and RHR SSFI Concerns TABLE OF CONTENTS 1.0 INTRODUCTION

1.1~pur ose 1.3 Sco e of Review and Recommendations

1.4 Review Team 2.0 SYSTEMATIC

ASSESSMENT

'I 3.0 ISSUES AND CONCERNS ADDRESSED 4.0 SUMMARY OF RECOMMENDATIONS

APPENDICES:

Appendix A, Programmatic

Concerns Listing Page i

0

Systematic

Assessment

of Engineering

Assurance Issues and RHR SSFI Concerns 1.0 INTRODUCTION

1e2~Pur ose This document provides the results of a systematic

assessment

of issues and concerns raised by the NRC',s Safety System Functional

Inspection (SSFI)conducted during November and December, 1989, which focused on the Ginna Station Residual Heat Removal (RHR)System.This summary report also presents interim remedial actions and long-term corrective

actions.Other steps toward improvements, already in progress, are also listed..Back round A safety System Functional

Inspection (SSFI)was performed by an NRC team from November 6 to December 8, 1989, at RG&E facilities (Ginna Station and the corporate offices), and is documented

in a letter from the NRC dated May 9, 1990.The objective of the SSFI was to assess the capability

of the Ginna Residual Heat Removal (RHR)system to perform its design basis safety functions.

The NRC inspection

team evaluated the adequacy of operational

procedures, test practices, ,and maintenance

policies as they contribute

to RHR system reliability.

The NRC team also addressed the quality of engineering

support activities.

The NRC team did not identify any conditions

that would prohibit the RHR system from performing

its intended functions under normal and design basis accident conditions.

However, it was stated by the NRC that complete verification

of system reliability

was not possible since the design basis calculations

for the RHR system were not readily available.

The NRC SSFI team did have one immediate concern and, as a result, RG&E was requested to promptly resolve a discrepancy

regarding the potential flooding of the RHR pump room.Our actions taken to resolve this were documented

in Enclosure E to our June 8, 1990 response (URI 89-81-10).

In addition it appeared to the NRC inspection

team that two activities

were not conducted in full compliance

with NRC requirements, as described in the Notice of Violation (NOV)enclosed as Appendix A to the NRC SSFI Inspection

Report.B-1

The RG&E response to the NOV included a schedule for resolving the unresolved

items (exclusive

of 89-81-11 discussed above)identified

in the NRC SSFI report.The RHR SSFI Inspection

Report also cited, a number of concerns which could be associated

with broader programmatic

issues.The NRC inspection

team concluded that weaknesses

exist in engineering

support and plant modification

activities.

These weaknesses

were listed in Section 2.1, and were discussed in Section 2.2, of the NRC SSFI Inspection

Report, and have been assigned unresolved

item number 89-81-11.The SSFI Inspection

Report required'G&E

to"provide their evaluation

of those weaknesses

within 120 days".The identified

weaknesses

were placed under the broad category of"engineering

assurance" by the NRC.RG&E committed'n

the June 8, 1990 30-day response to conduct a review of its engineering

process using a systematic

approach.RG&E elected to perform its evaluation

of the"engineering

assurance" issues by utilizing a review team approach.The results of the review team approach was intended to provide the basis toward.resolution

of NRC SSFI Inspection

unresolved

item 89-81-11.1.3 Sco e.of the Review and'Recommendations

The scope of the SSFI review was established

by RG&E Management

prior to the initiation

of the review team effort.RG&E Management

provided general guidance for conduct of the review'team effort as well as specific guidance on the scope of potential recommendations.

To ensure that a thorough evaluation

was conducted, the review team examined the material found in the following documents:

a~NRC SSFI Report no.50-244/89-81, dated, May 9, 1990 b.RG&E's 30-day response letter to the NRC dated June 8, 1990 c~Commitment

and Action Tracking System (CATS)commitments

established

by the NRC inspection

report and RG&E's June 8, 1990 letter.d.e.INPO Good Practices,"Guidance for the Conduct of Design Engineering" (INPO 88-016)December, 1988 Grove Engineering

Review Report dated July 10, 1990 Applicable

sections of the RG&E Configuration

Management (CM)Plan.B-2

0

g, NRC Safety System Functional

Inspection

Guidelines, Appendix C (issued 11/12/86).

h.EPRI, Nuclear Safety Analysis Center Document, NSAC/121,"Guidelines

for Performing

Safety System Functional

Inspections (November 1988).NQA-1"Quality Assurance Program Requirements

for Nuclear Power Plants" (1979)j.ANSI N45.2.11,"Quality Assurance Requirements

for the Design of Nuclear Power Plants" (1974)k.NRC'egulatory

Guide 1.64,"Quality Assurance Requirements

for the Design of Nuclear Power Plants" l.QE-series Engineering

Procedures

m.A-series Ginna Station Administrative

Procedures

With this reference material as background

information

the review team members proceeded to evaluate the NRC concerns and make recommendations

for corrective

action to RG&E Management

through the Department

Manager, Nuclear Engineering

Services.The following is'summary of the guidance provided by RG&E Management:

a.Issues addressed were to focus on, but not, limited to, those contained in the NRC SSFI Report (IR 89-81).b.c~Concerns cited by the NRC were to be accepted as valid.No effort was to be expended on questioning

either the cited concerns or the examples used in the NRC SSFI Report.Recommendations

were to take the form of interim actions and long-term corrective

actions.d.Recommendations

for interim actions were to be limited to the following items: i.Changes to~existin Engineering

and Ginna Station Procedures.

ii.Creation of a limited number of new Engineering

QE or Administrative

Procedures

and/or Ginna Administrative

Procedures.

iii.Issuance of policy statements (at discipline, department

or corporate level.)B-3

e.iv.Development

of discipline-specific

implementing

documents (such as design guides, standards, etc.).v.Reassignment

of duties to personnel within specific disciplines.

Recommendations

were'o be achievable

utilizing staff levels that are currently authorized.

1.4 Review Team RG&E Management

selected a review team to act on their behalf consisting

of a group of nine experienced

personnel representing

the following areas: Mechanical

Engineering, Electrical

Engineering, Structural

and Construction

Engineering, Nuclear Safety and Licensing, Configuration

Management, Document Control/Records

Man'agement, Ginna Technical Section, and Nuclear Engineering

Services Department

staff.2.0 SYSTEMATIC

ASSESSMENT

The multi-discipline

RG&E SSPI Review Team performed an assessment

of the issues and concerns generated by the NRC RHR SSFI.The team began by establishing

the following definition

of"Engineering

Assurance":

En ineerin Assurance:

The planned and systematic

actions necessary to provide adequate confidence

that engineering

activities

are performed in a consistent

manner with adherence to plant licensing basis, applicable

procedures, regulations

and accepted industry standards.

The review team members formed."focus groups" which were assigned individual

detail items from the programmatic

concerns listing established

in the initial breakdown of issues (appendix A).Individual

assessments

were made and the issues grouped and documented

as part of RG&Es"Systematic

Assessment

of Engineering

Assurance and RHR SSPI Concern Report." The review was based on the review team's own assessment

as well as on detailed information

obtained through discussions

with other cognizant engineering

and, plant personnel.

3.0 ISSUES AND CONCERNS ADDRESSED The review team regrouped all issues and programmatic

concerns into six topical areas, as listed below: s-4

3.1 To ical Area 1: "Improved Process for Reporting and Assessing Safety Concerns" a: "Process for Handling Safety Concerns Outside the EWR Process" 3.2 To ical Area 2:."Improved

Design Control and Reviews" a~b.c d.e.f."Engineering

Management""Engineering

Assurance""Timeliness

of DCR Processing""Design Reviews""PAID Upgrade Program""Design Control" 3.3 To ical Area 3: "Improved Design Interface Control" a."Procedural

Inconsistency" 3.4 To ical Area 4: "Improved Design Documentation/Design

Bases" a~b.c d.e.f.g,"Interdisciplinary

Review of Non-Mods""Calculations""Deletion of Information

from PGIDs""Valve Identification

Differences""Design Basis Information" Controlled

Instrument

List Training Material 3.5 To ical Area 5: Modifications" 1"Improved Documentation

Associated

with a."Invalid Information

in UFSAR" b."Assessment

of Cumulative

Effects of Modifications" c."Issuance of Design Outputs" d."Control of EWRs" 3.6 To ical Area 6: "Improved Engineering/Plant

Communications" a."Engineering/Plant

Interface" b."Acceptance

Criteria" c."Adequacy of SRV Test Acceptance

Criteria" 4.0 Summar of Recommendations

The following is a general summary of the most significant

interim actions and long term corrective

actions.4.1 RG&E's management

has ensured close control and quality engineering

services through their interaction

and review of design, but written procedures

do not make that control sufficiently-explicit.

B-5

As interim actions, a single procedure will be developed that outlines the entire scope of the design process.Discipline

design guides for generation

of design criteria, design analyses and design verification

documents will be initiated.

Also, the integrated.

assessment

process will be separately

" proceduralized.

Applicable

procedures

will be revised to establish the requirements

for review,'a'pproval, and issuance of vendor documents.

In, the longer term, we plan to complete the upgrade of engineering

procedures

and processes to reflect industry standards of good practice, efficiency

and rapid response.4.2 4.3 4.4 Engineering

procedures

contain a strong bias to modification

design.This has proved to be well suited toward major stand alone design projects but is not as effectively

used to aggressively

support all of the engineering

activities

associated

with a well-maintained.

operating plant.Close-out EWR documentation

can be protracted, because the scope of a modification

may be in'creased

over time, causing design documents to remain open.As interim actions, we plan to limit the practice of increasing

the scope of a design modification

during the interval between turnover of the modification

in the plant and records close-out.

We will begin to transmit EWR Design Packages to Document Control concurrent

with the issuance of the construction

package.In this way, the list of applicable

design documents will also be established

for the modification

to be installed..

The UFSAR change process will be proceduralized

and integrated

with the above turnover process.Interim activities

are needed to begin to capture, retain, provide access to, and organize design basis information

as part of the normal ongoing engineering

activities.

RG&E has incorporated

a design basis documentation

project under the Configuration

Management

Program.This program will be implemented

over the next several years focusing on the safety systems.Efforts will be made to identify the types of materi'al and documents that contain design basis information

and to begin to index and organize it in Document Control.In the longer term, the Design Basis Documentation

project will develop Design Basis Documents for the major plant systems and equipment.

Because of the major modification

bias used in the development

of QE procedures

and the major upgrade programs that have taken place over the years, the engineering

department

is not formatted on a system basis.B-6

li 1 i

The Configuration

Management

Program is being developed on a systems basis.The Q-List has defined system boundaries

that will be useful as we index design.documents

in Document Control and develop Design Basis Documents.

4.5 Many of the underlying

'concerns and long term corrective

actions are currently part of the existing or planned programs within Configuration'anagement

and Engineering

Procedures

Upgrade Programs.The Design Basis Documentation, Setpoint Verification, Q-List, Document Control Enhancement, and.Engineering

Controlled

Configuration

Drawing Upgrade Programs are individual

parts of this program.The specific actions recommended

by the SSFI Assessment

team will be reviewed by the RG&E personnel responsible

for the CM projects together with management

to make any needed revisions to the scope of these projects.4.6 Engineering

activities

are performed by Nuclear, Engineering

Services (NES)and station technical staff.The design process must ensure consistency

between these activities.

As interim actions, we plan to increase the controls over setpoint changes and reporting of safety concerns.A process will be developed to ensure that proposed setpoint changes are given the appropriate

review prior to their issuance.The PCAQ process (Potential

Condition Adverse to Quality)has been implemented

in QE-1603 for Nuclear Engineering

Services personnel to provide identification

and disposition

of potential safety concerns and provide a vehicle to improve the interface with technical personnel at the plant.We also plan to develop a streamlined

approval process for technical support projects not involving modifications.

In the long term we will examine establishing

a single process for all Nuclear Division personnel to report specific concerns which may have safety significance.

Processes will be developed to ensure commonality

of procedures

between NES and the Technical Section at Ginna.B-7

0 0 0

SYSTEMATIC

ASSESSMENT

OF ENGINEERING

ASSURANCE ISSUES AND RHR SSFI CONCERNS APPENDIX A

Cl0

Pro rammatic Concerns Areas Involvin Si ificant Identified

Weaknesses

Res onse Sco e Listin ENGINEERING

MANAGEMENT

Weakness in managerial

and administrative

controls Management

relies on engineer's

experience

instead of formal controls Engineering

management

has not provided clear guidance and procedural

controls over design change process Lack of Engineering

Assurance Practices Organizational

Interfaces

w~Control of documentation, engineering

design interfaces, and engineering

communications

with external organizations

is poor.Lack of criteria for.determining

when engineering

concurrence

is needed DCR's not processed in a timely manner Design output not properly distributed

UFSAR contains invalid information

No process for handling safety concerns identified

outside the engineering

process P&ID change did not result in an UFSAR change as appropriate

Engineering

Discipline

Interfaces

Each discipline

has its own interpretation

of engineering

procedure requirements.

Engineering

Management

has a different perception

than the engineering

staff P&ID changes occurred without an interdisciplinary

review CONFIGURATION

MANAGEMENT

Plant Baseline Configuration

Lack of complete and consistent

nomenclature

between P&IDs and procedures, UFSAR and QA Manual Deletion of information

from P&IDs Design Bases r~Design Basis Calculation

not available or do not exist~Lack of documented

design basis is a generic weakness~UFSAR contains invalid information

without a supporting

design basis~No calculation

list or formalized

overall listing

~Operating procedures, emergency procedures, and operator training.,material

do not reflect the limiting design basis of the system Design Modifications

RG&E does not have a mechanism for accounting

for synergistic

effects of modifications (electrical

calcs, pipe stress calcs)Numerous weaknesses

exist in engineering

support and plant modification

activities

Document Control ,~UFSAR contains invalid information

Lack of comprehensive

controlled

instrument

list Weakness in management

control system to assure complete and consistent

design output is issued and distributed

PSIDs issued have removed and revised information.

Team concerned how RG6E maintains traceability

of this information

Informational

inconsistencies

exist between documents DCR processing

is not timely EWRs remain in personal control of responsible

engineer EWRs lack index Completed EWRs are not processed into the document control system in a timely manner Uncontrolled

training material ENGINEERING

PROCEDURES

Design Reviews Review process lacks depth Review and verification

does not strictly follow ANSI N45.F 11 Inadequate

Review Independent

Verification

not done in accordance

with engineering

procedures

Calculations-

Generic weakness in review and approval of calculations

Calculational

control program (ANSI N45.2.11)is weak No list of calculations, no way to track past calculations

Setpoints~Instrument

loop setpoints may not account for loop inaccuracies

~Acceptance

criteria not established

in test procedure for setpointof

undervoltage

alarm relays Other Lack of formal control of engineering

and design documents RG&E design control measures do not compare favorably with accepted industry practices UFSAR contains invalid information

Lack of interface control with internal and external organizations

SRV testing procedures

contain general and minimal information

SAFETY CONCERNS Inability to properly identify safety concerns (battery load profile deficiencies

not discovered)

Inability to assess safety concerns (poor root cause analysis)No mechanism to disposition

safety concerns identified

outside of the normal engineering

process (PIC-629 EWR did not reflect any action taken on identified

concern TESTING DC undervoltage

test inadequate

SRV testing inadequate

MCCBs not tested periodically

SPECIFIC DESIGN CONCERNS SW Single Failure Inadequate

RHR NPSH Jumper cable exceeded minimum allowed bend radius Battery rack do not have a grounding cable RHR pump seal failure (Eg)RHR pump seal failure causing loss of both RHR pumps (single failure)