IR 05000324/1990005

Insp Repts 50-324/90-05 & 50-325/90-05 on 900129-0202. Violations Noted.Major Areas Inspected:Licensee Conformance to Reg Guide 1.97, Instrumentation for Light Water Cooled Nuclear Power Plant to Assess Plant & Environs..

ML20012C261
Person / Time
Site:	Brunswick
Issue date:	02/26/1990
From:	Conlon T, Hunt M, Merriweather N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20012C258	List: IR 05000324/1990005 ML20012C257 ML20012C260
References
RTR-REGGD-01.097, RTR-REGGD-1.097 50-324-90-05, 50-324-90-5, 50-325-90-05, 50-325-90-5, NUDOCS 9003200412
Download: ML20012C261 (10)

Text

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• • Me UNITED ST ATES ug'g NUCLEAR REGULATORY COMMIS$lON

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ATLANTA, GEORGI A 30323 3..

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Report Nos.: 50-325/90-05 and 50-324/90-05 Licensee:

Carolina Power and Light Company F

P. O. Box 1551 L

Raleigh, NC 27602 i

Docket Nos.:

50-325 and 50-324 License Nos.: DPR-71 and DPR-62

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Facility Name:

Brunswick 1 and 2 Inspection Conducted: January 29 - February 2,1990 i

Inspectors: g.

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,M. D. Hunt F

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.N' Merriweather Date Signed Approved M Mt '

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i. E. Conlon, Chief ate Signed Plant Systems Section Engineering Branch Division of Reactor Safety SUMMARY

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Scope:

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This. routine announced inspection was conducted in the areas of the licensee's conformance to Regulatory Guide (RG) 1.97, Instrumentation for Light - Water Cooled Nuclear Power Plant to Assess Plant and Environs Conditions During and Following an Accident.

Results:

In the. areas inspected, two violations were identified.

Although some deficiencies still exist regarding certain Category 1 instruments, the licensee has performed -the installation and modification of instruments to comply with

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Regulatory Guide 1.97, Revision 2.

In the areas inspected, one. licensee-identified violation and one NRC violation were identified regarding a failure to perform an adequate engineering evaluation of a condition adverse to quality and a failure to follow procedures i

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for implementation of modifications. With the exception of the above items, the inspectors considered the licensee in compliance with their response to Regulatory Guide 1.97, Revision 2, and the Safety Evaluation Report.

The as-built plant installation and drawings do not reflect the information which was submitted for NRR approval in several instances but no ma,ior discrepancies exist.

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REPORT DETAILS-F (

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-3, Persons Contacted b.f Licensee Employees

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• R. Allen, Senior Specialist, Nuclear Engineering Design (NED)

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• C, F. Blackmon, Manager, Operations

• S. H. Callis, Licensing On-site Representative.

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• A. G. Cheatham, Manager. Environmental / Radiation Control

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W.>J. Dorman, Manager, QA/QC

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'J.'L. Harness, General Manager l

• A. Harris, Regulatory Compliance.

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'R._E..Helme, Manager, Techni. cal' Support

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J. R. Holder, Manager, Outage Management and Modifications

• J. A. McKee, Manager, QA.

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• D.'E. Moore, Unit Manager, Engineering (NED)

• J. W. Moyer,-' Technical Assistant to Plant _ General Manager

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F R. Phillips;.EQ Engineer

• R. M. Paulk,. Supervisor, Regulatory Compliance M. Sawtschenko, Senior Specialist, Operations

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• R. B. Starkey, Manager, Brunswick Nuclear Plant.

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• S. L. Smith, Manager Unit-1 I&C Maintenance

' R; L. Warden, Manager; Maintenance I

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NRC Resident Inspector L

• W. Ruland, Senior Resident Inspector I

' Attended exit interview l

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Acronyms and initialims used through'out this report are listed in the last

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paragraph.

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2.. (Inspection of. Licensee's Implementation of Multiplant Action A-17:

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' Instrumentation for Nuclear Power Plants to Assess Plant and: Environs

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C Condition's During and Following an Accident (Regulatory Guide.l.97)

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P Criterion 13, " Instrumentation: and Control," of Appendix A to 10 CFR-

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'Part 50 includes a requirement that instrumentation be provided to monitor--

-variables and systems over1'their anticipated ranges for accident-conditions as appropriate to ensure adequate safety.

Regulatory i-Guide 1.97 (RG;1.97) describes a method acceptable to the NRC staff for.

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complying with the Commission's regulations to provide instrumentation to i

monitor plant variables and systems during and following an accident.

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s The purpose of this. inspection was t'o verify that the licensee has an instrumentation system for assessing variables and systems during and following an accident, as discussed in Regulatory Guide (RG) 1.97.- Under

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accident conditions, it is-necessary that the operating personnel have

(1) information that permits the operator to take preplanned actions to L

accomplish a safe plant shutdown; (2) determine whether the reactor-I tripped, Engineered Safety-Feature Systems (ESFS) actuated, and that

'other manually initiated safety systems important to safety are performing b

their intended functions; and (3)_ provide information to operators that i

will enable them'to determine the potential for causing a gross breach of b

. the barriers to radiation release and to determine if a gross breach of I'

barrier has - occurred.

For this reason, _ multiple instruments with overlapping ranges may be necessary. The required instrumentation must be capable of survivinn the accident environment for the. length of time its

E operability is required.

It is desirable that components continue to b

function following seismic events.

As a result, five types of variables have been specified that serve as guides in defining criteria and the selection of accident-monitoring instrumentation. The types are:

Type A - Those variables that provide information needed to permit the control room operating personnel to take specified manual actions for which no automatic control is provided and that are required for safety systems to accomplish their _ functions for design basis accident events; Type B - Those. variables that provide information to indicate whether plant safety functions are 'being accom-t plished; Type C - Those variables that provide information to indicate the potential.for barriers being breached or the actual breach of barriers to fission product-release; Type D - Those variables that provide-information L

to' indicate operation of individual safety systems and other systems-

.important to safety; Type E - Those variables to be monitored in-

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determining.the magnitude of the release of radioactive materials and for h

continuously assessing such release.

.The design ' and qualification criteria - are separated into. the separate categories that provide a graded approach to requirements depending on the importance to safety of the measurement of a specific variable, Category 1-

provides the most stringent requirements and is intended for key

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variables.

Category 2 provides less stringent requirements and generally F

applies to -instrumentation designated for indicating systems operating status. Category 3 is intended to provide requirements that will ensure that high quality off-the-shelf instrumentation is obtained'and applies to backup and diagnostic instrumentation.

A key variable is the single accomplishment of a safety function (Types B and C),-or the operation of a

. safety system (Type D), or radioactive material release (Type E). Type A

variables are plant specific and -depends on the operations that -the designer chooses for planned manual actions.

Inspection of Categories 1 and 2 equipment was performed as described below.

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Category I and 2 Instrument for Units 1 and 2

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L The instrumentation listed in the Table was examined to verify that j

.the design and qualification criteria of RG 1.97 had been satisfied.

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The instrumentation was inspected ~by reviewing drawings, procedures,

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data sheets, other documentation and performing walkdowns for visual observation of selected installed. equipment including CR indicators-l E

and recorders..The following areas were inspected:

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(1) Equipment Qualification - The EQ Master Equipment List and the L

- Q-List were reviewed for confirmation that-the licensee had

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addressed environmental qualification requirements for Class IE

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(2) Redundancy - Walkdowns were performed to verify by - visual observation that selected instruments were installed as specified and that separation -requirements were met.

In addition, drawings for all listed Category 1 instrumentation were reviewed to verify redundancy and channel separation, i

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(3) Power Sources - Drawings were reviewed to verify the instrumenta '

tion is energized from a safety-related power source.

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(4) Display and Recording - Walkdowns were' performed to verify by-visual observation that the specified display and recording instruments were installed.

Drawings were reviewed to' verify there was at least one recorder in a redundant channel and two

~ indicators, one per division (channel) for each -measured variable.

(5) Range -- Walkdowns were performed to verify the actual range of the indicator / recorders was as _ specified in RG 1.97 or the SER.

Review of calibration procedures verified sensitivity and overlapping requirements of RG.I.97 for instruments measuring

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the same variable.

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(6) Interf aces - The drawings and Q-List were viewed to verify that

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safety-related isolation devices were used when required _ to.

isolate _the circuits from nonsafety systems.

(7) Direct-Measurement - Drawings were reviewed to verify that the parameters _are directly measured by the sensors.

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(8) Service, Testing, and Calibration - The maintenance program for performing-calibrations and surveillances was reviewed and discussed with the licensee.

Calibration and surveillance procedures and the latest calibration completion date for each instrument were reviewed to verify the instruments have a valid

calibration.

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Category 1 Instruments

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hi Units 1 and 2 unless otherwise indicated

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Instrument Number Variable Channel or Train

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' RPV Pressure-C32-PT-N005 A&B

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C32-PI-R605 A&B

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' Alternate RPV Pressure B21-PT-N045 A&D-

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- Instrumentation (seismic only)

B21-PTM-N045 A&D

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RPV Water Level B21-LT-NO36 h

S21-LT-N037

821-LT-N026 A&B B21-LT-N027 A&B

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B21-LI-R610 B21-LR-R615 i

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B21-LI-R604 A&B-l

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B21-LI-R605 A&B-

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Suppression Pool Water Temperature CAC-TR-4426 1&2 j

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Suppression Pool. Water Level.

CAC-LT-2601

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Drywell Pressure CAC-PT-4175-F CAC-PT-4176~

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CAC-PT-1257-2A&2B Suppression Pool Pressure -

CAC-PI-1257-2A&2B C

Drywell/ Suppression Pool CAC-AT-4409

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. H /0 Analyzers CAC-AR-4409 2 2 CAC'-AT-4410-

'CAC-AR-4410

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Primary' Safety System B21-FT-4157-Relief Valve Position to

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B21-FT-4167 Logic light DSI-2A to

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Units 1 and 2 unless otherwise indicated l

is Instrument Number Variable Channel or Train

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(cont'd)

RCIC Flow Control E51-FT-N003

E51-FIC-R600 Primary Containment Radiation D22-RM-4195 (High Range)

D22-RM-4196

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D22-RM-4197 p

D22-RM-4198 022-RI-4195 r

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D22-RI-4197 h

D22-RI-4198

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I HPCI Flow E41-FT-N008

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E41-FIC-R600 j

Core Spray System E21-FT-N003 A&B'

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E21-FI-601 A&B h

RHR System Flow E11-FT-N015 A&B L

- Ell-FI-R603: A&B b.-

Discussion'and Conclusions

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.During the examination and verification of the variables as defined i'

by RG 1.97, the inspectors reviewed various licensee modifications E

which.were made to meet the category I variable requirements. ' The

following conditions were identified.

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~(1) The RPV pressure indicators C32-PI-R605A and R605B were not-

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properly identified on the simulator control: panels. While this s

L-is not a safety issue, the simulator:is a training tool and should reflect the correct, control board configuration..This Lv item was corrected during the inspection.

(2).The-RPV pressure indicator C32-PI-R605B had been installed on

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the-Unit 2 control board, calibrated and placed in service but

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the wiring drawing 2-FP-50014, sheet 4 had not been revised ' to-E show.-the ! installation of the additional indicator.- The L

indicator should have been added at revision level J but was not

and.the' drawing.is'now at revision level M.

This is an example of the failure to follow procedures.

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(3) 'A power supply module (CAC-PY-1257-2A/2B) had been designated as

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spare in' error.

Drawing 0-FP-07917 was revised to remove the.

[Je module. When the error was identified, the drawing.was revised l

to restore the module but the proper wiring connectionr details

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were not restored or identified on the drawing.

The instrument

. loop was never removed from service.

This is' an example of 1[,

failure to follow procedures,

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-(4) New. pressure transmitters, 1-CAC-PT-1257-2B and L

2-CAC-PT-1257-2B, were added to meet RG 1.97' commitments.

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However, the transmitters were installed and received initial

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calibration but were not added to the Q list and were not L

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entered into the preventative maintenance program.

These transmitters had not been calibrated in approximately two years.

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These pts are for RG 1.97 indication only and are not. a

f, technical specification component.

This is an example of

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U failure to follow procedures.

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-The above listed items (2, 3 and 4) are all examples of failure to ~

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follow procedures.

10 CFR 50, - Appendix B, Criterion V requires a

activities affecting safety to be performed in acco.rdance with l

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documented procedures.

Procedure 1 ENP-03, Plant Modification, t

requires that all drawings, 0-lists and EQ-lists be properly -revised

before the modification package is closed. The above listed Items 2,

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' 3 and-4 are examples of violation 50-325,324/90-05-01, Failure to

-Follow Procedures for Implementation of Modifications,

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Several discrepancies.were noted between the licensee's submittal

upon-which the SER was developed for the design requirements of j!

- category 1 variables and.what actually now exists in the units.

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b These had been identified by-the licensee in NCR No. A-87-017 dated June-19, 1987, which identified nonreduntant power supplies for reactor vessel pressure indication for train A-The' corrective

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action was to add indicator C32-PI-R605B for both units and the modification package was closed.- :Later NCR No, A-88-020 issued

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July 5, 1988, identified the fact that RVP indication loops are

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. powered f rom nonqualified. power supplies and contain non-Q-list

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F components (selector switch and recorder).

It also identified PCIS

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valves for which indication was powered from non-Q-list and nonclass

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1-E sources which would result in.a loss of position. indication upon..

.a loss of offsite power.

Although the nonseismic power supply - for RVP instrument loops was-

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identified later by the licensee (NCR A-88-020), it appears that a-

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more thorough investigation of the overall problem would have discovered that the power supplies did'not meet the commitments of e

RG 1.97.

This appears to be a failure to take adequate corrective action for conditions adverse to quality.

This is identified as a-

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noncited violation NCV-324,325/90-05-02, failure to perform an

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adequate engineering evaluation. This meets the intent of 10 CFR 50

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L Part 2, Appendix C, regarding discretionary enforcement.

The licensee has taken compensatory actions by listing alternative L

qualified instrumentation which is available. The licensee committed to-revise appropriate annunciation response procedures for seismic t

events when this alarm is received.

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Exit Interview I

The inspection scope and results were summarized on January 12, 1990, with-

those persons indicated in paragraph 1.

The inspector described the area F

inspected and discussed in detail the inspection results listed below.

Proprietary information is not contained in'this report, 4.

Acronyms and Initialisms

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f AT Analyzer Transmitter AR Analyzer Recorder CR Control Room EQ

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Environmental Qualification EQ DOC PAC Environmental Qualification Documentation Package ERDADS Emergency Response Data Acquisition Display System FSAR

. Final Safety Analysis-Report HRRM:

High Range Radiation Monitor-HT Heat IFI Inspector Follow-up Item FI Flow Indicator FIC'

Flow Indicating Controller FR Flow Recorder FT -

Flow Transmitter HPCI High Pressure Coolant Injection

'LI Level Indicator LR-Level Recorder

- LT Leve11 Transmitter

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'NCR Nonconformance. Report NCV Noncited Violation ND-Neutron Detector PAM-Post' Accident-Monitoring PI Pressure Indicator PR Pressure Recorder PT Pressure Transmitter PTM Pressure Trip Module RCIC Reactor Core Isolation Cooling RI Radiation Indicator RC Reactor Coolant-RCS Reactor Coolant System i

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8 RG Regulatory Guide RM Radiation Monitor RHR Residual Heat Removal RPV Reactor Pressure Vessel RVP Reactor Vessel Pressure RR~

Radiation Recorder SER Safety Evaluation Report SIS Safety Ihjection System TE Temperature Element TI Temperature Indicator.

TR Temperature Recorder

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