ML20245D548

Insp Repts 50-338/89-11 & 50-339/89-11 on 890403-07. Violations,Deviations & Unresolved Items Noted.Major Areas Inspected:Licensee Conformance to Reg Guide 1.97, Instrumentation for Light Water Cooled Nuclear Power..

ML20245D548
Person / Time
Site:	North Anna
Issue date:	06/07/1989
From:	Conlon T, Fillion P, Mark Miller NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20245D519	List: ML20245D515 ML20245D527 ML20245D539 ML20245D548
References
RTR-REGGD-01.097, RTR-REGGD-1.097 50-338-89-11, 50-339-89-11, NUDOCS 8906270152
Download: ML20245D548 (16)

See also: IR 05000338/1989011

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Sa Rfc Uh3tTED STATES

. og%,

NUCLEAR REGULATORY COMMISSION -

1. #

REGION 11

I o 101 MARIETTA STREET,N.W.

E: I

ATLANTT. GEORGIA 30323

\*****p'f

Report Nos.: 50-338/89-11 and 50-339/89-11

Licensee: Virginia Electric and Power Company

Glen Allen, VA 23060

Docket Nos.: 50-338 and 50-339 License Nos.: NPF-4 and NPF-7

Facility Name: North Anna 1 and 2

Inspection Conducted: April 3-7,1989

Inspectors: P ppf ~

e C 6 ~7 [

P. Fillion '

Date Signed

C c P ir h / W f' #7 SV

M. Miller Dats Sitjned

Approved [ M# ._

M7 [

T. Cdnlon, Section Chief 'Dat'e Signed

Plant Systems Section

Engineering Branch

Division of Reactor Safety

SUMMARY

Scope:

This routine, announced inspection was in the area of the licensee's

conformance to Regulatory Guide (RG) 1.97, Instrumentation for Light-Water-

Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and

Following an Accident.

Results:

With the exception of two URI's and one deviation the plant is in compliance

with RG 1.97. There were no major weakness or strengths identified in the

areas inspected.

- UP.I C9-11-01, Possible unreviewed deviation from RG 1.97 in the

area of process computer isolation (paragraph 2.C).

- URI 89-11-02, Specific identification of RG 1.97 Indicators

(paragraph 2.C)

- Deviation 89-11-03, Deviation from RG 1.97 (neutron flux)

paragraph 2.c). i

- Violation 89-11-04, Failure to identify deficient conditions (not

related to RG 1.97) (paragraph 2. c)

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8906270152 8906?O

PDR ADOCK 05000338

Q PDC

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REPORT DETAILS j

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1. Persons Contacted

Licensee Employees 1

• D. Blankenship, I&C Engineer  ;

• G. Kane, Station Manager j'

• P. Kemn Supervisor of Licensing

P. Knutsen, Supervisor of Nuclear Engineering l

• H. V. Le, Systems Engineer  ;

M. Marino, Systems Engineer i

G. Mocarski, Loss Prevention Coordinator j

• R. Woodall, III, Systems Engineer

Other licensee employees contacted during this inspection included

engineers, operators, security force members, technicians, and'  ;

administrative personnel. J

• Attended exit interview

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

Instrumentation for Nuclear Power Plants to Assess Plant and Environs 1

Conditions During and Following an Accident (Regulatory Guide 1.97)

(25587).

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

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

variables and systems over their anticipated ranges for accident

conditions as appro Regulatory

Guide 1.97 (RG 1.97)priate to ensure

describes adequate

a method safety.

acceptable to the NRC staff for

complying with the Commission's regulations to provide instrumentation to

monitor plant variables and systems during and following an accident.

The purpose of this inspection was to verify that the licen,ee has an

instrumentation system for assessing variables and systems during and

following an accident, as discussed in RG 1.97. Under accident conditions

it is necessary that the operating personnel have (1) information that

permits the operator to take preplanned actions to accomplish a safe plant'

shutdown; (2) determine whe;her the reactor trip, Engineered

Safety-Feature System (ESFS), and that other manually initiated safety '

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systems important to safety are performing their intended functions; and

(3) provide information to the operators that will enable them to

determine the potential for causing a gross breach of the barriers to

radioactivity release and to determine if a gross breach of barrier has

occurred. For this reason multiple instruments with overlapping ranges

may be necessary. The required instrumentation must be capable of

i surviving the accident environment for the length of time its operability -

It is desirable that components continue to function

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is required.

following seismic events.

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As a result, five types of variables have been specified that serve as j

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 i

information to indicate whether plant safety functions are being  !

accomplished; Type C - Those variables that provide information to  !

indicate the potential for barriers being breached or the actual breach of I

barriers to fission product release; Type D - Those variables that provide

information to indicate operation of individual safety systems and oth(.r

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

detennining the magnitude of the release of radioactive materials and for >

continuously assessing such release. )

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The design and qualification criteria are separated into three 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 require:aen ts and is intended for key

n riables . Category 2 provides less stringent requirements and generally

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 '

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backup and diagnostic ins trumentation. A key variable is that 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 j

designer chooses for planned manual actions. Inspection of Categories 1 1

and 2 equipment was performed as described below,

a. Category 1 Instrumentation

The instrumentation listed in the Category 1 Table, of this section, I

was examined to verify that the design and qualification criteria of '

pG 1.97 had been satisfied. The instrumentation was inspected by

reviewing drawings, procedures, data sheets, other documentation, and I

performing walkdowns for visual observation of the installed i'

equipment. The following areas were inspected:

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

Q-List were reviewed for confirmation that the licensee had

addressed environmental qualification requirements for class 1E

equipment. 1

(2) Redundancy - Walkdowns were performed to verify by visual

observations the specified instruments were installed and

separation requirements were met. In addition, drawings were

reviewed to verify redundancy and channel separation.

(3) power Sources - Drawings were reviewed to verify the

instrumentation 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 I.97 or the SER.

Review of calibration procedures verified sensitivity and

overlapping requirements of RG 1.97 for instruments measuring

the same variable.

(6) Interfaces - The drawings and Q. List were reviewed to verify

that safety-related isolation devices were used when required to

isolate the circuits from non-safety systems.

(7) Direct Measurement - Drawings were reviewed to verify that the

parameters are directly measured by the sensors.

(8) Service, Testing, and Calibration - The maintenance program for

performing calibrations and surveillance was reviewed and

discussed with the licensee. Calibration and surveillance

procedures and the latest data sheets for each instrument were

reviewed to verify the instruments have a valid calibration.

CATEGORY 1 TABLE

Variable Instrument No. (Loop) Drawings

Steam Generator LT-1474 6007D07

Narrow Range L1-1474 L-1474

Level MUX

Computer

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LT-1475 6007D23

LI-1475 L-1475

MUX

Computer

LT-1476 6007D41

LI-1476 L-1476- "

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FR-1478

MUX

Computer

l LT-1484 6007D07

l LI-1484 L-1484

l MUX

Computer

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LT-1485 6007D23 i

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LI-1485 L-1485

MUX

Computer

LT-1486 6007D42  ;

LI-1486 L-1486

FR-1488 ,

MUX

Computer

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LT-1494 6007008

LI-1494 L-1494 4

MUX

Computer j

Steam Generator LT-1495 6007D42

Narrow Range LI-1495 L-1495

Level MUX

Computer i

LT-1496 6007D43 l

LI-1496 L-1496 {

FR-1498 -l

MUX q

Computer j

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RCS Cold Leg TE-1410 6007D13

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Temperature TR-1410 T-1410

MUX f

Computer ]

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T2-1420 6007D20

TR-1420 T-1420

MUX

Computer

TE-1430 6007D46 i

TR-1430 T-1430 j

MUX

Computer

Refuelin9 Water LT-QS-100A 7383D29-

Storage Tank LI-QS-100A LQS-100A

Level MUX 3

LT-QS-100B 7383D33

LI-QS-100B LQS-1008

MUX

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Variable InstrumentNo.(Loopl Drawings

LT-QS-100C 7383D39

LI-QS-100C LQS-100C

MUX

RCS Wide Range PT-1402 6007D89

Pressure PI-1402A & B P-1402

MUX

Computer

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PT-1403 6007070

PI-1403A & B 6008D39  ;

MUX P-1403  ;

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Computer

Containment PT-LM-100A 7382D10

Intermediate PI-LM-100A ,

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Range Pressure MUX

Computer

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PT-LM-1008 7382021

PI-LM-100B

MUX 1

Computer j

PT-LM-100C 7382026

PI-LM-1000

MUX

Computer

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PT-LM-1000 7382031 l

PI-LM-1000 l

MUX l

Computer

Neutron NFD-190 Channel 1 NF-NM190

Flux NFI-190A NF-NM290

NFI-190B

MUX

NFD-1270 Channel 2 NF-NH1270

NFI-1270A NF-NH2270

NFI-12708

MUX

Core Exit TE-1E TESK-RC022

, Temperature TE-51E TESK-RC023

l MUX A TESK-RC024

MUX B

ICCM A '

ICCM B

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Variable Instrument No. (Loopl Drawings i

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Pressurizer ZS-PCV-1455C 11715-ESK-60N

PORV A1, A2, B1, B2 11715-ESK-6NR. I

Position Indicator ZS-PCV-1456 PVC-1456

A1, A2, B1, B2 PVC-1455

Containment H A-HC-101-1 A-HC101-1

Hydrogen H I-HC-101-1 & 2

Concentration H R-HC-101-1

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H A-HC-201-1 A-HC201-1 1

H I-HC-201-1 & c l

H R-HC-201-1 l

Notes: (1) MUX is multiplexer input for SPDS and computers other l

than the plant process computer. I

(2) Computer is plant prncess computer. j

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(3) The instruments listed are for Unit 1, the same Unit 2 1

instruments were reviewed. .)1

b. Category 2 Instrumentation

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The instrumentation listed in the Category 2 Table, of this section, i

was examined to verify that the design and qualification criteria of q

RG 1.97 had been satisfied. The instrumentation was inspected by  !

reviewing drawings, procedures, data sheets, other documentation, and ,

performing walkdowns for visual observation of the installed

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equipment. The following areas were inspected:

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

Q-List were reviewed for confirmation that the licensee had

addressed environmental qualification requirements for Class 1E

equipment.

(2) Power Sources - Drawings were reviewed to verify the

instrumentation is energized from a high quality or safety-

related power source.

(3) Display and Recording - Walkdowns were performed to verify by

visual observation that the specific display and recording

instruments were installed. Drawings were reviewed - to verify

there was at least one recorder, where required by RG 1.97, in a

redundant channel and two indicators, one per division (channel)

for each measured variable.

(4) Range - Walkdowns were performed to verify the actual range of

the indicators / recorders was as specified in RG 1.97 er the SER.

Also calibration procedures were reviewed to verify sensitivity

and overlapping requirements of RG 1.97 for instruments

measuring the'same variable.

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(5) Interfaces - The drawings and Q-list were reviewed to verify

that safety-related isolation devices are used when required to

isolate the circuits from computer systems (not safety-related).

(6) Direct Measurements - Drawings were reviewed to verify that the i

parameters are directly measured by the sensors.

(7) Service, Testing, and Calibration - The maintenance program for '

performing calibrations and surveillance was reviewed and

discussed wit!. the licensee. Calibration and surveillance

procedures and the latest data sheets for each instrument were i

reviewed to verify the instruments have a valid calibration.  ;

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CATEGORY 2 TABLE

Variable Instrument No. (Loop) Drawings .4

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HPSI FT-1940 7392008  !

Flow FI-1940 SI-036 l

7392D18 1

FT-1940-1

FI-1940-1 SI-053 -}

FT-1943 7392D18

FI-1943 SI-037

FT-1943-1 7392008 i

FI-1943-1 SI-054

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LPSI FT-1945 6008D08  ;

F1ow FI-1945 F-1945

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MUX 4

FT-1946 6008D37

FI-1946 F-1946 ,

MUX

Boric Acid FT-1110 6007D90

Charging FI-1110 F-1110- ,

Flow MUX l

Component TE-SW-111 7382053 l

Cooling Water TI-SW-111 i

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Temperature MUX

Containment TE-RS-150A 7382D11

Sump Water TI-RS-150A T-RS150A  ;

Temperature MUX

TE-RS-150B T-RS150B .

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Variable Instrument No. (Loop) Drawings

Accumulator PT-1921 6008D06

Tank Pressure PI-1921

PT-1923 5008D41

PI-1923

PT-1925 6008008

PI-1925

Ti- 927

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6008D41

PI-1927

MUX  !

' Accumulator PT-1929 6008007

Tank Pressure PI-3929

PT-1931 6008D42

PI-1931

MUX

4160 VAC 0-5KV Voltmeter 11715-FE-1D

BUS IH and IJ

Voltage

Notes: (1) MUX is multiplexer input for SPDS and computers other

than the plant process computer.

(2) Computer is the plant process computer.

(3) The instruments listed are for Unit 1, the 'same Unit 2 -i

instruments were reviewed. l

c. Discussion

In RG 1.97, Rev. 3, the design and qualification criteria for H

instrumentation include the following three requirements: j

(1) Redundant channels should be electronically independent and

physically separated from each other and from equipment not

classified important to safety in accordance with RG 1.75, l

" Physical Independence of Electric Systems," up to and including ,

any isolation device.

(2) No single failure within the accident-monitoring I

instrumentation should prevent the operators frora . being

presented the information necessary for them to determine the

safety status of the plant etc.

(3) The transmission of signals. for other use should be' through

isolation devices that are. designated as part of the monitoring

instrumentation and that meet the provisions of this document.

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In the North Anna design, qualified isolation devices located at the

Reactor Protection System process racks isolate the protection

portion of the circuitry from the indication portion. The indication

portion of. the circuitry includes the transmission of signals to  ;)

instruments on the main control board and the plant process computer. 3

Since the computer is not electrically isolated from the: instrument

signal, the computer is, in effect, part of the accident-monitoring

l instrumentation. Wiring from each of the process cabinets, for the

four redundant channels, to instruments on,the main control board was-

' designated as non-safety-related in the original plant design.

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Therefore, physical separation was not required, and does not exist

between redundant channels in the indication portion or the

circuitry. The NRC does not require back-fitting" of- accepted -

original plant designs to meet the requirements of RG'1.75' for

. physical separation.

Criteria (2) 'and (3) stated above are. violated by virtue of the fact

that the plant trocess computer is connected to the accident-monitoring :

instrumentation. The computer . represents a potential . common mode

failure. All conputer input cards are powered from a common power

stfply which means that a _ voltage surge appearing on the computer

power supply could. propagate to all redundant channels. In Tuch a

scenario, the surge could damage the isolator at the protection

system / indication interface' for multiple redundant channels. Damage

to the isolators would result in loss of indication. The issue

reduces to whether or not isolation devices are required at the

computer input points.

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' Neither the licensee's RG 1.97 submittal nor the NRC's. Safety

Evaluation Report ' address the. apparent deviation from. RG 1.97

described . above, i.e., plant process computer not isolated ,from

accident-moni toring instrumentation. Therefore, the inspector- 4

concluded that the particular design at North Anna may not have been .!

reviewed in relation to RG 1.97 at 'the time the SER was written.  !

This matter is called .an Unresolved Item pending' further review by )

the NRC. Unresolved Item 89-11-01, Possible Unreviewed Deviation "

from RG 1.97 (process computer isolation). ^1

RG 1.97 Revision 3 states that the instruments designated as Types.A,

B, and C'and Categories 1 and 2.should be specifically. identified on i

the control panels so that the operator- can eas fly discern that they

are intended for use . under accident conditions. Examples 'of

acceptable methods-for accomplishing this requirement are ident4 fica-

tion labels having a different background color than other labels or

instrument bezels color coded to . indicate RG 1.97 instruments. The 1

licensee's RG 1.,97 . correspondence does not address s'pecific

identification of instruments o'n the control panels, nor is specific

identification of RG 1.97 instruments incorporated into the ' control

panels. The licensee stated during: the inspection thatr "The

Virginia Power approach for compliance with EQ requirements was ato {

provide qualified signals to . the: instrumentation racks ' for all.

channels associated with each variable. Accordingly, all instrument

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channels for each variable met the same criteria, and special l

demarcation for RG 1.9'I compliance was not_ necessary. The Control {

Room Design Review performed in response to NLfREG-0737 Supplement l' j

evaluated control room labels for human factors considerations. A i

[ internal] memo requested that CRDR evaluate the specific criteria of l

RG 1.97. The response to that memo was that this would be further I

evaluated and addressed by Corrective Action CA29E identified as an I

commitment in letter SN 85-268C dated June _30,1986." This matter is

identified as an unresolved item pending the completion of CA29E and 4

subsequent NRC review. Unresolved Item 338, 339/89-11-02, Specific l

Identification of RG 1.97 Indicators.

The design and qualification criteria for Category 1 variables

requires that recording of instrumentation readout information should '

be provided. Where direct and immediate trend or transient

information is essential for operator information or action, the i

recording should be continuously available on dedicated recorders. ,

Otherwise, it [ recording information] may be continuously updated,  :

stored in computer memory, and displayed on demand. Intermittent

displays such as scanning recorders may be used in some cases. The

RG defines neutron flux as a Type B, Category 1 variable and, j

therefore, at least one channel should be recorded. The licensee's i

submittal dated January 31, 1984, also defines neutron flux as a

Type B, Category 1 variable, and states that the control room display

requirements are met. The submittal also states that a new excore

flux monitoring system will be installed to meet the environmental

qualification criteria. This new system was installed and, since it

does not include a dedicated (strip chart) recorder, apparently the

intention was to rely on the safety parameter display system to

provide the recording function. The NRC inspector asked the Software

Analyst Engineer to demonstrate the recording function for the

neutron flux qualified channels, but he was not able to do so. The

neutron flux qualified channel signals were transmitted to the SPDS;

however, the SPDS was not programmed to record this input data for

display on demand.

The conclusion reached during the inspection was that trend

information for the neutron flux channel was not available on

dedicated recorders nor stored in computer memory for display on

demand. Furthermore, it was not detennined whether or not this

variable is an isolated case of that problem.

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The failure to provide recording for neutron flux constitutes a

deviation from RG 1.97 and, therefore, a deviation from the

licensee's commitment to comply with the regulatory guide. Deviation

338, 339/89-11-03, Deviation from RG 1.97 (neutron flux).

RG 1.97 defines component cooling water temperature to ESF system as

a Type D, Category 2 variable. The licensee, in his RG 1.97

submittal dated January 31, 1984, stated that the instrumentation for

compnnent cooling water temperature to ESF systems did not meet

RG 1.97 with respect to range, environment 7 qualification, power

source nor control room display. The submittal also indicated that

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they would install a temperature channel to monitor operation at the

charging pump cooling system. In a supplementary submittal dated 3

May 10, 1985, the licensee stated the following: {

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"The Service Water System provides cooling water to Engineered ,

Safety Feature (ESF) systems. Therefore, service water j

temperature is the North Anna equivalent variable for the j

' component cooling water temperature to ESF system' variable )

specified in Regulatory Guide l.97. l

" Service water temperature instrument channels will be modified )

to meet the requirements of Regulatory Guide 1.97.- These j

modifistions are now in the design stage and will provide i

Service Water Temperature information to th- TSC and E0F in

addition to the existing temperature indicating meter on the

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main control board. Replacement of the existing RTD's will not' 1

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be necessary because they are located in the Turbine Building

and the Service Water Pump House and were originally specified

to function in the areas in which they are located." j

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The Safety Evaluation Report found this commitment acceptable. In  !

providing indication for this variable, qualified temperature I

transmitters were installed to monitor temperature of the auxiliary j

service water pump discharge.  !

Telephone conversation with the licensee on June I and 2,1989, 1

disclosed that qualified temperature and flow elements are installed

in the Servie Water System that meet the intent of F31.97. These

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instruments are as follows:

l - TE-SW-111 and -211 on discharge of Aux Service Water Pumps 1

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- FE-SW-100A, -1008, -100C, and -100D on discharge flow of

RSHX. The power supply meets requirements for type D

variable.

- TE-SW-108 and 109 on discharge from main Service Water Pumps.

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1 - FT-SW-109A and 109B from charging pump gear box and seal

I cooler

- FT-SW-108A and 108B from charging pump lube oil cooler.

In the course of performing plant walkdowns related to RG 1.97

equipment, the following deficient conditions were identified:

(1) In the main control room, the control board on which most of the

RG 1.97 indicators were mounted is an open-back type of panel.

The panel is about 20 feet long and seven feet high for each

Unit. Because of previous events that involved workers-

inadvertently short-circuiting wiring terminal points, a

decision was made by the responsible manager to not allow

cleaning and vacuuming work to take place in the area behind the

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main control board. Consequently, the interior of .the' main 4

control panel, including the terminal blocks, had not been .l

cleaned for several years. The amount of. dust buildup had {

reached the point where it could affect the integrity of the ,

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wiring.

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(2) The Nuclear Instrumentation System Cabinets which contain the i

excore neutron flux instruments were opened for inspection after I

finding condition (1). Two of these NIS cabinets per Unit j

contained an un-terminated, un-taped multiconductor cable. One i

of these cables was tagged INMS2WX110. It was later determined j

that the cable was actually a spare cable. To have an j

un-terminated, un-taped, improperly tagged cable in the NIS i

cabinet represents a deficient condition, and a departure from {

plant procedures.

(3) The control room floor is a false floor er computer floor type

construction which provides a cabla routing space. Behind the

main control panel at a point where the Units 1 and 2 panels

meet, one of the removable floor boards was loose or not. q

properly supported. This constituted a deficient condition j

because a person could fall on the loose board and injure j

himself or fall into the open-back control board. ]

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(4) Panels located at the opposite side of the control room from the

main control board were also inspected. The NRC inspector i

noticed that the bottom of the panel was not properly sealed. .l

The panels themselves had large openings in the bottom to allow I

cable entry from the raised floor cable routing space (as  !

defined by NFPA Std 75). According to the original design, fire

retardent boards had been installed at the bottom of these

panels, after the cable work was complete. The purpose of the  ;

boards may be to help maintain halon gas concentrations  !

discharged from the halon fire suppession system installed in

the raised floor cable routing space. In nine of these panels,

the bottom sealing bocrds were badly broken or missing.

The four deficient conditions described above. '.e., the heavy dust

in the control panel, the improperly spared ca .es, the loose floor

boards and the broken sealing boards, had sufficient safety

significance to constitute a violation of NRC requirements in the

area of identifying and correcting deficient conditions. It is

identified as Violation 338, 339/89-11-04, Failure to Identify

Deficient Conditions.

The licensee made several plant modifications or equipment upgrades

in order to comply with RG 1.97. Part of the work of the NRC

inspection was to determine the scope of these plant modifications

for the audit sample. The relevant modification packages were

reviewed to the extent of verifying the scope of work and status.

The inspector was satisfied that the necessary work had been

accomplished. The plant modifications for the audit sample are

summarized below:

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SUMMARY OF MODIFICATIONS

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Variable Package No. Scope Completed

Accumulator 86-11 & 12 Upgrade to EQ 10/2/87

Tk Pres

HPSI Flow 84-55 & 56 Upgrade to EQ and relocate 12/20/87

82-14

Component 84-49 & 50 Replace equipment 12/5/88

Cooling Water

Core Exit 85-07 & 08 Entire System upgrade to 6/14/87

Temp meet range, EQ, seismic,

redundancy

RCS Cold 81-54 Replace RTD to make EQ 4/10/87

Leg Temp

RCS Wide 82-14 Upgrade to EQ 5/27/87

Range Pres.

RWST Level 84-47 & 48 Upgrade transmitters to 12/19/85

seismic i

l

PORV Position 84-17 & 18 Replace limit switches to 6/14/87 I

meet EQ, seismic, redundancy I

l

Cont Inter- 82-14 Upgrade to EQ 5/27/87 l

mediate Pres.  !

\

Neutron 83-30 & 31 New system installed 6/24/S8

Monitoring  !

LHSI Flow 82-14 Upgrade to EQ 5/27/87

• Date that Station Manager signed package for Unit I as being

essentially complete.

3. Corrective Action Program

With respect to URI 338, 339/89-11-01, Electrical Independence of RG 1,97

l Instrumentation, the licensee's official position is that their design

l meets the intent of RG 1.97 and, therefore, any internal audits would not I

have identified problems in this area. URI 338, 339/89-11-02, Specific '

Identification of RG 1.97. Indicators, is being addressed by .the licensee

under their control room design review effort. The licensee deviated from

I

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the RG for one variable (Deviation '338, 339/89-11-03). This is attributed "

to personnel error and inattention to detail. The problems described in

Violation 338, 339/89-11-04, Failure to Identify Deficient Conditions,

should have been identified by the. licensee in the course of performing

plant walkdowns. Three of the conditions had probably existed for at

least a year. The improperly spared cables should also have been

identifled at the time the modification work was performed.

4. Exit Interview

The inspection scope and results were sunnarized on April 7,1989,.with

those persons indicated in paragraph 1. The inspectors described; the

areas inspected and discussed in detail the inspection results listed

below. Proprietary information is not contained in this report.

Dissenting connents were not received from the licensee.

338, 339/89-11-01 Unresolved Item - Possible unreviewed deviation from RG

1.97 (Process Computer Isolatio.1)

338, 339/89-11-02 Unresolved Item - Specific Identification of RG 1.97

Indicators

!

338,339/89-11-03 Deviation - Deviation from RG 1.97

338,339/89-11-04 Violation - Failure to Identify Deficient Conditions i

The Station Manager commented on the violation. He stated that failure to

remove dust from the control panel did not represent a shortcoming of

their program but, rather, a failure to implement Procedure ADM-20.48, 4

Station Material Condition and Housekeeping (dated November 23, 1988). "

The Station Manager also stated that corrective actions for each of the

deficiencies had already been initiated. "The control panel was cleaned

last evening, but not to .ny satisfaction. It will be vacuumed again," he -

said. It was also stated that they determined the halon system would have j

performed its function even though the sealing boards were broken or j

missing.  !

i

5. Acronyms and Initialisms ]

l

EQ -

Environmental Qualification

FI -

Flow Indicator

FT -

Flow Transmitter

HPSI - High Pressure Safety Injection

HA - Hydrogen Analyzer

HI - Hydrogen Analyzer Indicator

HR -

Hydrogen Analyzer Recorder

I CM - Inadequate Core Cooling Monitor

LI -

Level Indicator

LPSI - Low Pressure Safety Injection l

LT -

Level Transmitter

MUX - Multiplexer

NFD -

Nuclear Flux Detector

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.

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1

NFI -

Nuclear Flux Indicator l

PCV -

Pressure Control Valve j

PI -

Pressure Indicator  !

I

PORV - Power Operated _ Relief Valve

PT -

Pressure Transmitter

RCS - Reactor Cooling System

RTD - Resistance Type Temperature Detector  !

SER - Safety. Evaluation Report i

TE - Temperature Element  !

TI - Temperature Indicator .

TR - Temperature Recorder  !

PS -

Position Switch

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