ML20137Q239

Insp Rept 50-302/85-41 on 850926-1023.Violations Noted: Failure to Post High Radiation Area,Failure to Verify Accuracy of Estimated Critical Positions Prior to Startup & Failure to Adhere to Protective Clothing Stds

ML20137Q239
Person / Time
Site:	Crystal River
Issue date:	11/18/1985
From:	Panciera V, Stecko T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20137Q175	List: IR 05000926/2010023 ML20137Q171 ML20137Q239
References
50-302-85-41, NUDOCS 8512050197
Download: ML20137Q239 (19)

See also: IR 05000302/1985041

Text

p ltro UNITED STATES

'o NUCLEAR REGULATORY COMMISSION

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[ '

n H EGION 11

g j 101 MARIETTA STREET, N.W.

• I, 's ATLANTA GEORGIA 30323

%,...../

Report No.: 50-302/85-41

Licensee: Florida Power Corp 3 ration ,

3201 34th Street, South

St. Petersburg, FL 33733

Docket No.: 50-302 License No.: DPR-72

Facility Name: Crystal River 3

Inspection Conducted: September 26 - October 23, 1985

Inspector: M ' ' _ _ h/

T. 7. Ste aa, aenior ResWent' MKpector Date Signed

Accompanying rsonne . J. E. Tedrow, Resident Inspector

Approvedby:\ . /6 o u m e // /

V. W. Par tNera,~ 5 Miori Chief D5th Signed

Pruject Section 28

Division of Reactor Projects

SUMMARY

Scope: This routine inspection involved 169 inspector-hours on site by two

resident inspectors in the areas of plant operations, security, radiological

controls, licensee event reports, nonconforming operations reports, THI task

action plan (NUREG 0737) followup, and ifcensee action on previous inspection

items. Numerous facility tours were conducted and facility operations observed.

Some of these tours and observations were conducted on backshifts.

Results: Three violations were identified: (Failure to post a high radiation

area, paragraph 5.b.(5)(a); Failure to verify the accuracy of two independent

estimated critical positions prior to reactor startup, paragraph 5.a; and Failure

to adhere to the requirements specified on a RWP for protective clothing,

paragraph 5.b.(5)(b)).

0512050197 051121

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! REPORT DETAILS

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1. Persons Contacted

Licensee Employees

W. Bandhauer, Nuclear Safety Supervisor

• G. Boldt, Nuclear Plant Operations Manager

• P. Breedlove, Nuclear Records Management Supervisor

R. Brown, Nuclear Electrical /I&C Supervisor

J. Buckner, Nuclear Security and Special Projects Superintendent

• J. Bufe, Nuclear Compliance Specialist

• M. Craven, Nuclear Security Officer

• 0. Green, Nuclear Licensing Specialist

E. Howard, Director, Site Nuclear Operations

• W Johnson, Nuclear Plant Engineering Superintendent

• K. Lancaster, Manager, Site Nuclear Quality Assurance

• P. McKee, Nuclear Plant Manager

• C, McLane, Building Serviceman

l *R. Pinney, Senior Nuclear Engineer

i V. Roppel, Nuclear Plant Engineering and Technical Service Manager

l E. Simpson, Director, Nuclear Operations Eng, and Licensing

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• P. Skramstad, Nuclear Chem / Rad Protection Superintendent

l *D. Smith, Nuclear Maintenance Superintendent

>

• E. Standard, Nuclear Mechanic

K. Vogel, Nuclear Senior Electrical /I&C Engineer

l *K. Wilson, Supervisor, Site Nuclear Licensing

l *R. Wittman, Nuclear Operations Superintendent

G. Moore, Chairman, Nuclear General Review Committee

Other personnel contacted included office, operations, engineering,

maintenance, chem / rad and corporate personnel.

• Attended exit interview

l 2. Exit Interview

The inspector met with licensee representatives (denoted in paragraph 1) at

the conclusion of the inspection on October 23, 1985. During this meeting,

the inspector summarized the scope and findings of the inspection as they

are detailed in this report with particular emphasis on the violations,

unresolved item, and inspector followup items. The licensee representatives

acknowledged the inspector's comments and did not identify as proprietary

any'of the materials provided to or reviewed by the inspectors during this

inspection.

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3. Licensee Action on Previous Inspection Items

(Closed) Violation (302/85-29-02): The inspector verified that a Short Term

Instruction (STI 85-36) was issued to ensure that licensed personnel are

familiar with the requirements of the Offsite Dose Calculation Manual

(ODCM). In addition, the inspector verified that the placards were

installed near the Reactor Building purge exhaust fan cont"ol switches to

provide information to operators prior to securing the purge fans.

Discussions with licensee representatives also indicate that controlled l

copies of the ODCM are now being maintained in the control room.

(Closed) Violation (302/85-29-04): The procedure SP-220 that was completed

on July 5, 1985. was reviewed to assure satisfactory performance. A

memorandum, dated July 8, 1985, and sent to shift supervisors to remind them

of the requirements for making a voluntary entry into an action statement,

was also reviewed. Based upon these reviews, this item is considered to be

closed.

(Closed) Violation (302/85-29-05): The inspector verified that the licensee

completed the corrective actions as follows:

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A letter written on July 13, 1985, provided guidance to maintenance

personnel clarifying the intent and scope of procedure CP-113;

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An investigation to determine if any jumpers were improperly installed

was conducted and completed on October 18, 1985;

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Procedure CP-113 was revised as revision 44 on August 26, 1985, to

clarify the intent of section 5.4 of the procedure tha*. addresses

jumper control; and,

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Procedure MP-1088 was revised to delete the 2"-6" requirement. The

procedure now requires that the water level be below the reactor vessel

flange.

Based upon this review, this item is considered to be closed.

(0 pen) Inspector Followup Item (302/84-30-04): During the last outage

(Refuel V), the licensee replaced a substantial portion of the nuclear

services seawater (RW) system piping with piping that has been relined with

a superior coating substance in accordance with modification (MAR)

83-06-27-01. This work was observed by the inspectors as documented in NRC

Inspection Report 50-302/85-26. The licensee intends to add a refueling

outage visual inspection to the Preventative Maintenance (PM) program to

ensure that pipe integrity is monitored. This item remains open pending the

addition of this visual inspection program to the PM program.

(Closed) Inspector Followup Item (302/85-37-01): The licensee has clarified

procedure OP-103 and relabeled the level column in figure 7.10 to read

" Dipstick",

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(Closed) Inspector Followup Item (302/85-37-04): The inspector has reviewed

the licensee's post trip evaluation for the reactor trip which occurred at

. 6:44 p.m. , on August 20, 1985. The inspector reviewed the plant's para-

meters and condition and has no further questions on this item.

(Closed) Unresolved Item (302/85-26-08): The licensee has revised the

Station Battery Service Test, SP-523, to reflect the higher discharge rates.

(Closed) Inspector Followup Item (302/85-33-07): The licensee has

incorporated the guidelines discussed in Information Notice (IEN) 85-58 in

the reactor trip breaker maintenance procedure PM-118. A receipt inspection

is done on all breakers received from the General Electric Service Shop in

Atlanta, Georgia. This receipt inspection includes the performance of

PM-118 before the breaker is accepted.

(Closed) Inspector Followup Item (302/85-19-03): The inspector has reviewed

the work package to install high temperature and radiation resistant control

cables for valves RCV-11 and RCV-13. These cables were installed in

accordance with modification (MAR) 81-11-23. The licensee has also

performed an equipment qualification modification on all safety related

valves in hostile environraents. The inspector considers this action

sufficient and has no further questions on this item.

(Closed) Unresolved Item (302/85-37-02): Evaluation of this item by NRC

Region II has determined that the licensee's compensatory measures were

adequate. The licensee will install an alarm system for the door in

question and in the interim will maintain a security guard post for this

door.

4. Unresolved Items

Unresolved items are matters about which more information is required to

determine whether they are acceptable or may involve violations or devia-

tions. A new unresolved item is identified in paragraph 8.c of this report.

5. Review of Plant Operations

The plant started this inspection period in power operation (Mode 1). While

reducing power for turbine generator repairs on September 27, 1985, a

control rod group dropped into the reactor core causing the plant to be

shutdown to the hot standby mode (Mode 3) (see paragraph 8.a for details of

this event). Following repairs to the control rod auxiliary power supply on

October 1, the reactor was restarted at 5:13 a.m. and the plant reentered

Mode 1 at 6:38 a.m. On October 2, the plant was shut down to Mode III for

repairs to a control rod drive position indicator. By 1:50 p.m. , on

October 2, this repair was completed and the reactor was again taken

critical. Power operation was resumed at 2:45 p.m., on October 2.

On October 9, at 9:34 a.m. , the reactor was manually tripped from approxi-

mately 96% of full power due to inadvertent closure of two main steam ,

isolation valves (see paragraph 8.b for details of this event). Following  !

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repairs to the main steam isolation portion of the Emergency Feedwater

Initiation and Control (EFIC) system, the reactor was restarted at 6:47 p.m.

and the plant entered Mode 1 at 7:51 p.m. , on October 9, where it remained

for the duration of this inspection period. .~ -

a.. Shift Logs and Facility Records-

The inspector reviewed records and discussed various entries with -

operations personnel to verify compliance with the Technical Specifica- ,

tions (TSs) and the licensee's administrative procedures. -

The following records were reviewed: A

Shift Supervisor's Log; Reactor Operator's Log; Equipment Out-0f-

Service Log; Shift Relief Checklist; Auxiliary Building Operator's Log; -

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Active Clearance Log; Short Term Instructions (STIs); Selected _

Chemistry / Radiation Protection Logs; and Completed Operation Procedures '

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(ops).

In addition to these record reviews, the inspector independently

verified clearance order tagouts.

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On October 10, during a review of the data completed on October 9 for

procedure OP-210, Reactor Startup, the inspector noted that the two ,

Estimated Critical Position (ECP) calculations required by

section 6.2.1 resulted in final reactivity values of -0.55% Ak/k and

-0.24% Ak/k. This provided a difference between these values of 0.31%

Ak/k. In addition to requiring these two separate and independent

calculations, step 6.2.1 also specifies that the two ECP calculations

be in agreement within 10.1% Ak/k prior to proceeding with the reactor

startup. No other ECP calculations were performed and a reactor

startup was made based upon these results.

Failure to adhere to the requirements of procedure OP-210 is contrary

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to the procedure adherence requirements of TS 6.8.1.a and is considered

to be a violation.

Violation (302/85-41-01): Failure to adhere to the requirements of

procedure OP-210 regarding ECP calculation agreement.

b. Facility Tours and Observations

Throughout the inspection period, facility tours were conducted to

observe operations and maintenance activities in progress. Some

operations and maintenance activity observations were conducted during

backshifts. Also, during this inspection period, licensee meetings

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.were attended by the inspector to observe planning and management

activities. The meeting attendance included the Nuclear Generator

Review Committee (NGRC) meeting (the Offsite Review Committee) heid on

October 8, 1985.

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The facility tours and observations encompassed the following areas:

~,, security perimeter fence; control room; emergency diesel generator

i room; auxiliary building; intermediate building; battery rooms; and

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- electrical switchgear rooms.

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During these tours, the following observations were made:

E ~ '(1) Monitoring Instrumentation - The following instrumentation was

1. 1 observed to verify that indicated parameters were in accordance

> with TSs for the current operational mode:

s. Equipaint operating status; area, atmospheric, and liquid radia-

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N- tion monitors; electrical system lineup; reactor operating para-

. meters; and auxiliary equipment operating parameters.

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- A No violations or deviations were identified.

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-(2)- Safety Systems Walkdown - The inspector conducted a walkdown of

the nuclear services and decay heat seawater, containment hydrogen

, nonitoring (WS) system, and accessible areas of the core flood

> < system to . verify - that the system's drawings and procedures

correctly reflect "as-built" plant conditions.

As a result of the walkdown on the WS system, there appears to be

a discrepancy between the valve labeling and the flow diagram.

See paragraph 7.1 of this report for details.

-No violations or deviations were identified.

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(3)' Shift Staffing - The ' inspector verified that operating shif t

/ staffing was in accordance with TS requirements and that control

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' room operations were being conducted in .an orderly and

professional manner. In addition, the. inspector observed shift

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turnovers on various occasions to verify the continuity of plant

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' .7 status, operational problems, and other pertinent plant informa-

7- w.., tion during these turnovers- .

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5 No violations or deviations were identified.

.f. s (4)., - Plant Housekeeping Conditions - Storage of material 'and '

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.t - Q -. components, and cleanliness conditions of various areas throughout

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+ the facility were observed to determine whether safety and/or fire

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. hazards existed.

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~'4- - -No; violations or deviations were identified.

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(5) Radiation Areas - Radiation Control Areas .(RCAs) were observed to

• Cerify proper identification and implementation. These observa-

tions included selected, licensee conducted surveys, review of

step-off' pad = conditions, disposal of contaminated clothing, and

area postings. Area postings were independently. verified for

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accuracy through the use of the inspector's own radiation

monitoring instrument. The inspector also reviewed selected

s' radiation work permits and observed personnel use of protective

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clothing, respirators, and personnel monitoring devices to assure

that the' licensee's radiation monitoring policies were being

followed.

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As a result'of these reviews, the following items were identified:

(a) On October 16,. 1985, while checking area postings, the

inspector identified two solid radwaste storage drums (55

"

gallon drums) within the radwaste storage area that had

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indicated dose rates of 300 millirem. per hour (mrem /hr) to

500 mrem /hr on contact, and approximately 100 to 150 mrem /hr

at 18 inches from the drums. These drums were not posted as

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a high radiation area.

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When licensee personnel wede notified of this finding, they

responded with their own ins'trumentation. Upon verifying the

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inspector's findings, the area was properly barricaded and

posted as a high radiation-area.

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Technical Specification (TS) 6.12.1.a requires all radiation

areas that have dose rates greater than 100 mrem /hr and less

than 1000 mrem /hr to be barricaded and conspicuously posted

as a high radiation area. Failure to adhere to the require-

ments of TS 6.12.1.a is considered to be a violation.

Violation (302/85-41-02)': Failure to barricade and post a

high' radiation area as required by TS 6.12.1.a.

(b) ~0n October 7, while observing troubleshooting on the "A" High

- Pressure Injection Pump (MVP-1A), the inspector noticed an

electrician working inside the contaminated area around the

A pump and 'not wearing anti-contamination protective clothing

on his head or rubber shoe covers on his feet. The inspector

contacted the health physics office to. determine the correct

protective clothing to be worn for this work and reviewed the

radiation-' work permit- established for- this type = of work

-(S85-0399). - As a result of this review, it was determined

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that a hood and rubber shoe covers were required to be worn

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inside the contaminated area. When health physics

technicians were made aware of this finding by the-inspector,

a- the technicians ~ instructed the electrician to wear the

k required protective clothing.

- Chemistry and Radi.ation Protection Procedure RSP-101, Basic -

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M Radiologicalc Safeel Information and Instructions for

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" Radiation Workers", step 3.1.3.4 requires that the require-

t; }. ~ ments established on RWPs be observed and adhered to.

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Failure to adhere to the requirements of procedure RSP-101 is

contrary to the requirements of TS 6.8.1.a and is considered

to be a violation

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Violation (302/85-41-03): Failure to adhere to the require-

ments of procedure RSP-101 to wear protective clothing

established on a RWP.

(6) Security Control - Security controls were observed to verify

that security barriers are intact, guard forces are on duty,

and access to the protected area (PA) is controlled in

accordance with the facility security plan. Personnel within

.the PA were observed to ensure proper display of badges and

that personnel. requiring escort were properly escorted.

Personnel within vital areas were observed to ensure proper

authorization for the area.

No violations or deviations were identified.

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(7) Fire Protection - Fire protection activities, staffing and

equipment were observed to verify that fire brigade staffing

was appropriate and that fire alarms, extinguishing equip-

ment, actuating controls, fire fighting equipment, emergency

equipment, and fire barriers were operable.

No violations or deviations were identified.

(8) Surveillance - Surveillance tests were observed to verify

that approved procedures were being used; qualified personnel

were conducting the tests; tests were adequate to verify

equipment operability; calibrated equipment, as required,

were utilized; and TS requirements were followed.

The following tests were observed and/or data reviewed:

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- SP-130, Engineered Safeguards Monthly Functional Tests;

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SP-146, EFIC Monthly Functional Test;

.SP-163A,' Waste Gas-H2 2 /0 Analyzer Channel Calibration;

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SP-317, RC System Water. Inventory Balance;

-- SP-333, Control Rod Exercises;

- SP-335,' Radiation Monitoring Instrumentation Functional

Test;

- SP-421, Reactivity Balance Calculations; and,

- SP-422,'RC System Heatup and Cooldown Surveillance.

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No violations or deviations were identified.

(9) Maintenance Activities - The inspector observed maintenance

- activities to verify that correct equipment clearances were

in effect; work requests and fire prevention work permits, as

required, were issued and being followed; quality control

personnel were available for inspection activities as

required; and TS requirements were being followed.

Maintenance was observed and work packages were reviewed for

the following maintenance activities:

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- Troubleshooting the control rod drive system;

- Control rod position indication relay replacement for

control rod number 4 in rod group number 3;

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- Electrical checks of the control rod power train in

accordance with procedure PM-126;

- Troubleshooting of oil leaks on the "A" High Pressure

Indication Pump (MVP-1A) in accordance with procedure

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HP-531;

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- Reactor trip. breaker inspection and testing in accordance

with procedure'PM-118;

- Troableshooting.the presence of a half trip signal on the

Emergency Feedwater Initiation and' Control (EFIC) system's

main steam line isolation channel in accordance with

procedure MP-531;

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- Replacement of the "B"LDecay Heat Closed Cycle Cooling

Heat Exchanger (DCHE-18) channel heads in accordance with

procedures MP-122, PM-112 and modification MAR 83-05-10-01;

- Troubleshooting of reactor coolant' system flow indication

oscillations in the "B" channel of the_ reactor protection

system in accordance with_ procedure MP-531;

- Replacement of_the sightglass on the "A" Emergency Diesel

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Generator-(EDG-3A) governor; and- ,

- Troubleshooting the EDG-3A drop' target annunciator panel.

While observing ' reactor trip breaker maintenance and testing

which is-performed in accordance with procedure PM-118, the

inspector noticed that . step 7.4.3.2 of the procedure, which 'j

recorded the "as left" pickup and_ dropout voltage settings '

applied to the undervoltage _ trip device, did not require

verification by Quality Control (QC) personnel although a QC .l

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inspector did verify the "as found" pickup and dropout

voltages. The inspector ' discussed this omission with

i- licensee personnel who =then agreed that the "as lef t"

voltages should be verified by a QC inspector. Step 7.4.3.2

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. of- the procedure was performeo and verified by a QC inspector

and the licensee plans to revise procedure PM-118 to require

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this verification. This revision to PM-118 will be reviewed

during future _ inspections. This item will be tracked along

with previously' identified concerns on procedure PM-118.

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identified in' inspector followup item (302/85-07-04). '

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As a result of ~ the work package for troubleshooting the

EDG-3A drop target annunciator panel, the inspector noticed

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. that this work was not classified as safety related even

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though electrical . circuit drawings appear to classify this

system- as . safety related. See paragraph'8.c of this report

for details.

(10) Radioactive Waste Controls -Solid waste compacting and

selected liquid and. gaseous waste . releases were observed to

verify .that approved procedures were utilized, _ that appro-

priate . release approvals were obtained, and .that required

surveys were taken.

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No violations or deviations were identified.

j.. -(11)_ Pipe Hangers and Seismic Restraints. - Several pipe hangers

L and seismic restraints (snubbers) _ on safety related systems

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were' observed to ensure that fluid levels were. adequate and

no leakage was _ evident, that restraint settings were appro-

priate,'and that anchoring points were not binding.

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l No violations or deviations were identified.

6. ^ Review of Lic'ensee Event Reports and Nonconforming Operations Reports-

' a .- Licensee Event Reports (LERs). were: reviewed for potential; generic

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impact, to detect trends, < and to determine 'whether corrected actions

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1 appeared appropriate. Events, which were reported immediately, were

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reviewed:as'they-occurred to_ determine.if~the TSs were satisfied,

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LERs' 85-08,' 85-14, 85-16, and 85-17 were . reviewed in accordance with

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' current NRC enforcement policy. . LERs~ 85-08, 85-16, Land - 85-17 ~ are

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closed. cLER_85-14 will remain.open for the following reason:

LER.85-14 reported two succes'sive actuations of_the-Emergency Feedwater' -

Initiation'and: Control System-(EFIC). As part of the corrective action;

to ' preventXrecurrence of this_ type of event, the licensee plans to

lprovideLa' detailed review and discussion Lof the entire event to all ,

operatorsivia the Operator's Study Book. This item will. remain open

pending completion.of this corrective action.

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b. The inspector reviewed non-comforming operations reports (NCORs) to

verify the following: compliance with the TS, corrective actions as

identified in the reports or during subsequent reviews have been

accomplished or are being pursued for completion, generic items are

, identified and reported as required by 10 CFR Part 21, and items are

reported as required by TS.

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All NCORs were reviewed in accordance with the current NRC Enforcement

Policy.

No kiolations or deviations were identified.

7. NUREG 0737 - TMe Task Action Plan Item Review

The following TMI Task Action Plan items were reviewed and applicable

installations verified to determine the status of completion:

a. I.D.2. , Plant Safety Parameter Display System (SPDS): The licensee has

completed installation of the SPDS in accordance with their commitment

stated in a letter to the NRC dated June 19, 1985.

To verify implementation of this item, the inspector reviewed the

completed modification package (MAR 81-06-38) which included system

test procedures T/P-1, T/P-2, and T/P-3 and examined system installa-

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tion. Due to problems with eight computer points (Nos. 227, 228, 232,

235, 236, 237, 240, and 247), test procedure T/P-3 has not been

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completed. The -licensee is actively pursuing resolution to these

problems. The licensee's status on this item is considered to be

complete. Further progress on resolution of the computer point

problems will be tracked as an Inspector Followup Item.

Inspector Followup Item (302/85-41-04): Review the license's progress

to resolve the SPDS computer point problems. (Item I.D.2.)

b. 11.B.1., Reactor Coolant System (RCS) Vents: The inspectors verified

installation of the upper "J" leg and pressurizer vents in accordance

with ' modification (MAR) 80-04-73 as discussed in NRC inspection report

50-302/83-18. During this reporting period, the inspector verified

that the system was periodically tested and that operation of the vents

was covered by procedures. The system is being tested in accordance

with procedure SP-171, Reactor Coolant High Point Vents Functional '

Test, and is covered operationally by procedure EP-290, Inadequate Core

Cooling.

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By letter dated July 10, 1985, from the NRC to FPC, the licensee was

granted a ' permanent exemption for the installation of the reactor

vessel head vent. Based upon this letter and NRC reviews of other

aspects of this item, the licensee's commitments for item II.B.1. are

considered to be complete and this item is closed.

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c. II.B.2. , Plant Shielding (Equipment Qualification): All shielding and

valve modifications were reviewed and examined by the NRC as documented l

in NRC Report 50-302/83-24. i

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In a letter dated May 5, 1980, the subject of which included the

staff's evaluation of the implementation of " Category A Lessons .

Learned" requirements, the NRC stated that "The licensee will complete  !

l this review (for equipment qualification) for electrical equipment as [

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part of. its . response to IE Bulletin 79-01B". This position was

confirmed in a June 1, 1980 letter from FPC to NRC. This letter  :

referred to NUREG 0578, item 2.1.6.B, which was involved with the [

qualification of electrical equipment. The licensee responded to IEB

79-01B on October 31, 1980.

During the time period from 1980 to the present, the requirements of IE  !

Bulletin 79-01B were superseded by a new ruling designated as 10 CFR '

50.49. During the period of March 4-8, 1985, an NRC inspection was ,

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conducted (NRC -Inspection Report 50-302/85-09) to review the pre-

implementation of the 10 CFR 50.49 rule. This inspection identified

eight open items that were to be completed by the end of Refuel V

(which ended in August 1985). Further activities concerning equipment

qualification will be tracked in accordance with the open items of NRC  !

Report 50-302/85-09 and action on item II.B.2 is considered to be  !

complete,

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d. II.E.1.1., Auxiliary Feedwater System Evaluation: Modifications to the

emergency feedwater (EFW) system have been completed except for the

installation of a new EFW tank which FPC has committed to complete, in

a letter dated June 26, 1984, by the end of their next refueling outage

(Refuel VI). These modifications 'were reviewed and verified complete

by system walkdowns as follows: ,

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Replacement of the turbine driven EFW pump (EFP-2) steam admission  ;

. valve ASV-5 and addition of a new in parallel" steam admission '

valve ASV-204 in accordance with modifications (MARS) 85-04-02-01

and 80-11-48-01 respectively. These MARS and their completed

installations were verified complete as . documented in NRC

Inspection Report 50-302/85-29;

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Removal of EFW pump recirculation return line' valve COV-104 and

-removal of the internals of EFW system suction valve (CDV-103)

from the condensate storage tank (CST) in accordance with MAR

83-04-31-01; and,

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Electrical disconnection of valves EFV-3,4,7, and 8 in accordance

with temporary MAR T82-07-34-01. This temporary MAR was made a

permanent modification by-MAR 77-07-01-03A. (Note: This modifi-

cation was required because valves EFV-3,4,7, and 8 were

electrically supplied by non-safety-related motor control

centers.-) The inspector verified the electrical disconnection of '

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valves EFV-4 and 8 at both the motor control center and the valve

operators.

-With regard to the installation of missile shields for EFP-2, the

licensee has determined through analysis that missile shields are not

required. The results of this analysis was transmitted to the NRC in a

letter dated August 8,.1984.

The licensee has not completed their Probability Risk Analysis (PRA).

In a letter to the NRC dated October 16, 1985, the licensee stated that

while their PRA was in its final stage, they intended to wait for the

issuance of the NRC generic letter concerning the reliability.of the

i auxiliary feedwater systems before they would provide their submittal.

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As a result of this review, it appears that two issues, building of the

EFW tank and issuance of the PRA, remain incomplete. These items will

be reviewed during-subsequent inspections and this item remains open.

e. II.E.1.2., Auxiliary Feedwater System Initiation and Flow: The

licensee was required to install a safety-related automatic initiation

and flow indication system for the emergency feedwater. (EFW) system.

This installation was to be followed by Technical Specification (TS)

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submittals to ensure that this system is included in the license

conditions.

The ~ licensee has installed and tested the Emergency' Feedwater

Initiation and Control (EFIC) system in accordance with modification

(MAR) 80-10-66. Installation and testing of the EFIC was reviewed and

observed by the inspectors as documented in NRC Inspection Report

50-302/85-33. The licensee also " submitted a TS change request and

' Amendment number 78 was issued to add the EFIC system to the TSs.

Based upon this review, the licensee's actions appear te be complete

and this item is considered to be' closed.

f. II.F.1.1. , Accident Monitoring - Noble Gas Monitor: The licensee has

completed installation of the modifications. required to satisfy this

item. These modifications were reviewed and verified complete- by.

system walkdowns as follows:

' Addition of Main Steam Line (MSL)- monitors in accordance with

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modification (MAR) 80-05-78. This MAR was verified complete as

documented in NRC Inspection Report 50-302/81-15;

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Addition of mid and high range monitors for. the reactor building

purge monitor '(RMA-1) and the - auxiliary building ventilation

monitor (RMA-2) in accordance with MAR 81-04-66-01. In addition

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, to.the. reviews conducted during this-inspection, operation of this

system was. observed during a post-implementation inspection as

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documented.in NRC Inspection Report 50-302/84-07.

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The licensee has been unable to calibrate the mid and high range

monitors due to the unavailability of calibration gases and procedures.

This 'was reported to the NRC in Special Report 85-03 dated August 21,

1985, as required by TS 3.3.3.9, action statement 29. This calibration

is expected to be completed by November 20, 1985. Activity on this

item will continue to be tracked by an existing Inspector Followup Item

(302/85-05-03).

During the system walkdowns, the inspector noted that two valves,

WSV-95 and ' WSV-96 appear to be labeled incorrectly and appear

incorrectly labeled on the system flow diagram (FD 302-694). This

finding was discussed with licensee personnel. The licensee will

examine the system and drawings and make applicable changes.

This is considered to be an Inspector Followup Item. This item will be

. reviewed in combination with the item identified in paragraph 7(i)

following.

Based upon this review, the licensee's activities appear to be complete

and this item is considered to be closed.

g. II.F.1.2, Accident Monitoring - Iodine / Particulate Sampling: The

licensee has completed modification (MAR) 81-04-66-03 that provided the

enhanced iodine and particulate sampling capability. This MAR has been

reviewed and the system walked down to verify completion. Operation of

this system was also observed during a post-implementation inspection

as documented in NRC Inspection Report 50-302/84-07.

Based upon this review, the licensee's actions appear to be complete

and this item is considered to be closed.

h. II.F.1.3, Accident Monitoring - Containment High Range Monitor: The

licensee has completed modification (MAR) 81-04-68 that increased the

range of the reactor building radiation monitor. This MAR has been

reviewed and a partial system walkdown conducted to verify completion.

Based on this . review, the licensee's actions appear to be complete and

this item is considered to be closed.

i. II.F.1.6, Accident Monitoring - Containment Hydrogen: The licensee has

completed modification (MAR) 79-11-70-02 that added the capability to

monitor post accident reactor. building hydrogen levels. This MAR and

is associated testing has been reviewed and the system walked down to

verify completion.

During a walkdown of this system, the inspector noted that there

appeared to be a discrepancy between the. system installation and the

flow diagram of the system (FD 302-693). This discrepancy appears to

be limited to labeling of the following valves.

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WSV-38 and 39;

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WSV-47 and 48; and,

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WSV-659 through 662.

This information was presented to licensee personnel for their review.

This finding will be tracked as an inspector followup item in combina-

tion with that identified in paragraph 7(f).

Inspector Followup Item (302/85-41-05): Review the licensee's progress

to resolve drawing discrepancies for flow diagrams FD 302-694 and

FD 302-693.

As the result of further correspondence between FPC and the NRC, the

licensee' has agreed to modify their monitoring system to allow more

frequent testing. This modification, MAR 85-09-09-01, has been issued

and work is expected to start in about three weeks. The licensee has

committed to have this MAR completed by December 31, 1985. The

inspector reviewed this MAR and will continue to track the installation

as an inspector followup item.

Inspector Followup Item (302/85-41-06): Review the licensee's progress

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to install the hydrogen monitoring testing MAR 85-09-09-01.

Based upon this review, the licensee's actions appear to be complete

and this item is considered to be closed.

j. II.F.2., Instrumentation for Detection of Inadequate Core Cooling

(ICC):

The licensee has completed installation of the modifications (MARS)

required to satisfy this item. These modifications were reviewed and

verified complete by appropriate system walkdowns as follows:

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Removal of the center control rod drive mechanism (CRDM) in

accordance with MAR 83-03-04-12;

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Interconnections between the main control board and the ICC

instrument panel in accordance with MAR 80-10-66-17, field change

' notices (FCNs) 5 and 8; and,

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Testing of the completed system in accordance with MAR

83-03-04-11, test procedures (T/Ps) 1, 2, and 3.

The licensee could not obtain satisfactory completion of T/P-3 for the

void trending system due to the fact that incorrect initial calibration i

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- values had been' set into certain instrumentation modules. Therefore, .

while the ICC system is installed, it is not fully operable. l

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By letter dated September 6,1983, the NRC required the licensee to

install the ICC system but also directed that the system not be made

operational pending resolution by the NRC of instrument usage.

The licensee has met their commitment to install the instrumentation

and therefore action on this item is considered to be complete.

Operability of the instrumentation, i.e. , satisfactory completion of

T/P-3, will be tracked as an inspector followup item.

Inspector Followup Item (302/85-41-07): Review the licensee's progress

to complete T/P-3 of MAR 83-03-04-11 for the ICC system.

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k. II.K.3.5, Automatic Trip of the Reactor Coolant Pumps (RCPs): In

generic letter 83-10 dated February 8,1983, the NRC determined that

automatic tripping of RCPs was not necessary and that plant specific

procedures directing RCP trips was adequate. In response to that

letter, FPC in a letter to the NRC dated February 27, 1985, provided I

setpoints for use by the operators to trip the RCPs. These setpoints

were incorporated into procedure AP-380, Engineered Safeguards System

Actuation, as revision 4 dated July 30, 1985.

The inspector verified the inclusion of these setpoints into AP-380 and

considers licensee action on this item to be complete.

1. III. A.1.2. , Upgrade Emergency Support Facilities: The licensee has

completed construction and has implemented use of the Technical Support

Center (TSC) _ and the Emergency Offsite Facility (E0F). In addition,

the event recall system has been installed and is fully operable.

The inspectors have observed use of the TSC and the recall system.

Inspection of these facilities remains to be completed by the NRC

Region II Emergency Preparedness section. This item will remain open

pending completion of this inspection.

m. III. A.2.4&5, Emergency Preparedness - Installation. and Implementation

of the Offsite Dose Computer: The licensee has completed installation

of the new meteorological data acquisition system and has implemented

its usage. The licensee has had problems with this new system and is

continuing to resolve _ these problems. A discussion of the' problems

with the new system is. documented in NRC Inspection Report

50-302/84-18. Further activities to resolve !these problems will be

tracked in accordance with Inspector Followup Item (302/84-18-06).

The'-licensee has decided to not use the Emergency Dose Assessment

System (EDAS) they had previously committed to develop. In a letter to

the NRC dated December 21, 1984, the licensee informed the NRC of their

desire to utilize their existing systems to meet this requirement

because of developmental problems with the EDAS. The present system

will utilize existing computers and emergency plan procedures

EM-204(A), EM-204(B), and EM-204(C).

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The inspector has reviewed these procedures and determined that the

procedures provide the stated dose calculations. These procedures were

also previously reviewed by the NRC as documented in NRC Inspection

Reports 50-302/84-18, 50-302/84-28, and 50-302/85-02. As a result of

these reviews, one procedure, EM-204(C), has an , outstanding Inspector

Followup Item (302/84-18-04) pending. Further activity in this area

will be tracked by this followup item.

Based upon these reviews, the licensee's activities on this item

appears to be complete. Outstanding specific items will continue to be

. tracked by the identified Inspector Followup Items.

n. III.D.3.4, Control Room Habitability: The licensee has completed the

installation and has the control room toxic gas monitoring system

operable. Verification of this installation is documented in NRC

Inspection Report 50-302/84-09 and operation and testing of this system

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.is periodically observed by the inspectors.

In their submittal to the NRC dated January 30, 1981, the licensee

committed to provide, in addition to the toxic gas monitors, a system

that would allow pressurizing of the control room. This submittal was

approved by the NRC in a Safety -Evaluation Report (SER) dated

February 17, 1982. During further correspondence between the NRC and

FPC regarding this issue, the licensee was requested to provide imple-

mentation dates. In the licensee's response that provided these dates,

the control room pressurizing issue was overlooked.

During discussions with licensee personnel, it has been determined that

the licensee is not intending to pressurize the control room but will

instead go to a zone isolation concept. The licensee has let a

, contract to Gilbert Associates Incorporated (GAI) to re-analyze a study

of control room habitability during accident conditions. This

j re-analysis will be done utilizing the existing control room ventila-

, tion system. Based upon the results of this study, modifications to

the control room ventilation system, if required, will be made.

The licensee expects to complete this study by March,1986. Any

required modifications will be completed by the end of the next

refueling (Refuel VI).

This item will remain open pending completion- of the required control

. room pressurizing modifications.

8. Design, Design Changes and Modifications

Installation of new or modified systems were reviewed to verify that the

changes were reviewed and approved in accordance with 10 CFR 50.59, that the

changes were performed in accordance with technically adequate and approved

pruedures, that subsequent testing and test results met acceptance criteria

or deviations were resolved -in an acceptable manner, and that appropriate

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drawings and facility procedures were revised as necessary. This review

included selected observations of modifications and/or testing in progress.

Modifications reviewed during this inspection period are delineated in  !

paragraph 7 of this report.  ;

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9. Nonroutine Event Followup "

a. At 11:45 p.m.", on September 27, 1985, after a power reduction to '

approximately 13% of full power to conduct turbine generator repairs,

control rod group number 7 dropped into the reactor core. Control rod

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number ,1 in rod group number 7 (rod 1-7) was being powered from the

control- rod drive auxiliary power supply to align this control rod with i

the rest of the control rods in group number 7. The dropping of the ,

control rod group occurred while transferring rod 1-7 from the i

auxiliary power supply back to its normal group power supply. The '

addition of the negative reactivity resulting from the control rod

group insertion placed the reactor in a shutdown condition. The plant  ;

was placed in the hot standby (Mode 3) condition at 12:20 a.m. , on i

September 28, to effect repairs to the control rod drive system.

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The cause for the dropped control rod group is attributed to a faulty {

jogging motor in the auxiliary power supply's programmer control -

assembly. This faulty motor created a voltage transient which inter- i

rupted power to the entire group 7 control rods allowing the entire

group to lose power momentarily and drop into the core. Repairs to the  !

control rod drive auxiliary power supply were completed and the reactor

was taken critical at 5:13 a.m. , on October 1, followed by resumption

of power operation at 6:38 a.m. ,

The inspector observed the activities of troubleshooting the cause for

the dropped rod group and observed the plant startup following repairs. ,

No violations or deviations were identified.  !

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b. At 9:34 a.m. , on October 9,1985, the reactor was manually tripped due

to the inadvertent closure of two Main Steam Isolation Valves (MSIVs),

MSV-413 and MSV-414, for the "B" Once Through Steam Generator (OTSG).

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Troubleshooting on the Emergency Feedwater Initiation and Control  !

(EFIC) system was in progress at the time the MSIVs closed.

The EFIC system requires two out of four main steam isolation channels

to trip to cause the MSIVs to close. Troubleshooting on the EFIC i

system was in progress to determine the cause for a half trip signal

being present on the main steam isolation channel for the B OTSG. The

8 EFIC cabinet was deenergized to replace a main steam isolation trip

module believed to be the cause of the problem. Unknown to the

troubleshooting technicians, this module replacement did not correct

the cause of the problem and the half trip signal was still present.

When an EFIC cabinet.is-reenergized all trip modules assume trip status

until reset -by the technicians. Upon reenergizing the B EFIC cabinet,

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the necessary two trip signals were present (one from the uncorrected

half trip problem and one from the recently energized B EFIC channel)

and MSV-413 and 414 closed. The inspectors arrived in the control room

shortly after the reactor trip had oc' curred and verified the status of

safety systems, plant status, and plant parameters. The inspectors

also observed the trouble. shooting and maintenance on the EFIC system

and reviewed the licensee's post trip evaluation.

No violations or deviations were identified.

c. At 12:35 p.m. , on October 8,1985, the inspectors were informed of an

inadvertent start of the "A" Emergency Diesel Generator (EDG-3A). As

part of the review for this event, the inspector reviewed a work

package for troubleshooting the EDG-3A drop target annunciator panel

which was in progress at the time the diesel started. After

discussions with the electricians performing the work, the inspector

learned that when an attempt was made to deenergize the drop target

panel by opening a breaker on the Engineered Safeguards Diesel

Generator D. C. Panel 3A (DPDP-6A), the wrong breaker was inadvertently

opened which deenergized two emergency diesel generator air start

solenoid valves, EGV-36 and EGV-37. This allowed the admission of air

to the diesel and it started. The inspector noted that the trouble-

shooting being performed was not classified as safety related.

The inspector reviewed the Safety Listing to determine the safety

classification of the EDG-3A drop target annunciator panel. The Safety

Listing references Electrical Circuit Schedule Drawings (E-212 series)

to determine the safety classification of specific electrical circuits.

The inspector reviewed drawing E-212-027 and based on the information

from this drawing it appears that the drop target panel should be

classified as safety related. The inspector contacted licensee repre-

sentatives to obtain more precise information. The licensee is

presently researching this item to determine the safety classification

of this panel. This item will be considered unresolved pending the

inspector's review of this safety deterw11 nation.

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Unresolved Item (302/85-41-08): Review the licensee's determination of

the safety classification of the EDG-3A drop target annunciator panel.

While reading the Safety Listing, the inspector was unable to find

DPDP-6A listed. The inspector discussed this omission with licensee

representatives and it was determined that DPDP-6A should be classified

as safety related and included in the Safety Listing. It appears that

a previous revision to the Safety Listing had incorrectly omitted this

panel from the current listing. The licensee plans to revise the

Safety Listing to include .DPDP-6A. The inspectors have previously

identified their concerns over the adequacy of the Safety Listing to

properly classify systems in NRC inspection report 50-302/85-29

(Unresolved Item 302/85-29-01). This item appears to be another

example of. safety listing inadequacy.

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