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Insp Rept 50-302/85-33 on 850727-0816.Violations Noted: Failure to Take & Analyze Grab Samples Each 24 H When RM-A12 Out of Svc,Failure to Have at Least One Operable Emergency Diesel Generator & Failure to Have Valve Control Procedures
	ML20135G016
Person / Time
Site:	Crystal River
Issue date:	08/29/1985
From:	Panciera V, Stetka T; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20135G002	List: IR 05000727/2008016; ML20135G000; ML20135G016;
References
	50-302-85-33, IEIN-85-058, IEIN-85-58, NUDOCS 8509180061
	Download: ML20135G016 (13)
	v • d • e

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Q RfC UNITED STATES

f NUCLEAR REGULATORY COMMISSION

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%' \ REGION ll

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[ 101 MARIETTA STREET.N.W.

, i, " c ATL ANTA, GEORGI A 30323

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Report No.: 50-302/85-33

Licensee: Florida Power Corporation

3201 34th Street, South

St. Petersburg, FL 33733

Docket No.: 50-302

License No. DPR-72

Facility Name: Crystal River 3

Inspection Conducted: July 27 - August 16, 1985

Inspector: '

T. F. Stetka," SenioFT<esident Insp . wor

~ ~, $ ateAY0$ Signed

Accompanying Personne E. Tedrow, Resident Inspector

. Sasser, Resident Inspector (0conee) (August 9-16)

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Approved by: . ], n , , , j fA ,, 5)/2')/ f'S

V'. W . Pa nbiWa','th fe f , P roj e c t S ec t i o n 2 B " l)a tE' Si gned

Division of Reactor Projects

SUMMARY

Scope: This routine inspection involved 112 inspector-hours on site by two

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

controls, Licensee Event Reports and Nonconforming Operations Reports, plant

startup preparations, and licensee 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 take and analyze grab

samples each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as required by Technical Specification (TS) 3.3.3.1 when

RM-A12 is out of service, paragraph 5.b.(1); failure to have at least one

operable EDG as required by TS 3.8.1.2.b, paragraph 6.b.(3); and failure to have

an adequate equipment control procedure for the performance of valve lineups as

required by TS 6.8.1.a, paragraph 7.)

8509180061 850829 2

PDR ADOCK 0500

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REPORT DETAILS

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1. Persons Contacted

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Licensee Employees .

! *S. Baggett, Nuclear Principal Mechanical / Structural Engineer t

G. Boldt, Nuclear Plant Operations Manager

2 *P. Bleedlove, Nuclear Records Management Supervisor

.( *C. Brown, Assistant Nuclear Outage and Modifications Manager

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J. Bufe, Nuclear Compliance Specialist

D. Fields, Nuclear Reliability Supervisor ;

i E. Howard, Director, Site Nuclear Operations '

W. Johnson, Nuclear Plant Engineering Superintendent

K. Lancaster, Manager, Site Nuclear Quality Assurance

i *J. Lander, Nuclear Outage and Modifications Manager

C. Long, Senior Quality Auditor

j *J. May, Welding Engineer

P. McKee, Nuclear Plant Manager

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T. Miller, Nuclear Shift Supervisor

W. Neuman, Senior Nuclear Inservice Inspection Specialist

P. Patel, Nuclear Quality Engineering Supervisor

, *W. Rossfeld, Nuclear Compliance Manager

, *P. Skramstad, Nuclear Chemistry and Radiation Protection Superintendent

l F. Sullivan, Nuclear Electrical /I&C Engineer

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

4 *G. Westafer, Manager, Nuclear Operations Licensing and Fuel Management

K. Wilson, Supervisor, Site Nuclear Licensing

R. Whittman, Nuclear Operations Superintendent

J. Wright, Nuclear Support Specialist, Chem / Rad Protection Services

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Other personnel contacted included office, operations, engineering,

maintenance, chem / rad and corporate personnel.

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Attended exit interview

2. Exit Interview

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

j the conclusion of the inspection on August 26, 1985. During this meeting,

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

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

j unresolved items, and inspector followup items.

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! Also during this meeting, the inspectors discussed the observations from

walkdowns conducted on the High Pressure Injection (HPI) and Emergency

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Feedwater (EFW) systems.

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The following items were identified:

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The following valves in the HPI system were missing identification

tags: MUV-413, MUV-414, MUV-415, and MUV-335;

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Modifications to the EFW system have removed valves EFW-73 and EFW-74

but the system drawing, FD-302-082, has not yet been revised; and

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Modifications to the EFW system have also removed valve COV-104 and

taken the internals from valve CDV-103 but both these valves are still

shown on drawing FD-302-101 and listed in procedure SP-381, Locked

Valve List.

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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.

3. Licensee Action on Previous Inspection Items

(Closed) Inspector Followup Item (302/85-02-01): The licensee has completed

a detailed review of their dose assessment which includes the RADDOSE II

computer program. As a result of this review, it was determined that the

errors made during the emergency plan exercise were caused by personnel

error and not the computer program. The licensee is taking corrective

actions to minimize a repetition of these occurrences.

(0 pen) Inspector Followup Item (302/85-29-03): The licensee has received

verbal notification from their architect-engineer, Gilbert Associates (GAI),

that at least one of the AHF-22 fans per diesel must be running during

diesel operation. It appears that these fans are necessary to assure proper ,

generator cooling. The licensee has written a Short Term Instruction (STI)

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to assure operators are aware of the importance of these fans. The

licensee's diesel operating procedure, SP-354, already contains the

requirement that these fans must be running for diesel operation. This item

remains open pending receipt of GAI's written evaluation and further review

of this item by the inspectors.

(Closed) Inspector Followup Item (302/85-08-05): The licensee has replaced

valves CHV-68 and CHV-69 with the new Fisher valves.

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 5.b.(2) of this

report.

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5. Review of Plant Operations

At the beginning of this inspection period, the plant was in the cold

shutdown mode (Mode 5). Plant heatup was commenced and the plant reached

the hot shutdown mode (Mode 4) at 0543 on August 7,1985, followed by the

hot standby mode (Mode 3) at 2359 on August 8,1985, 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 (TS) and the licensee's administrative procedures.

The following records were reviewed:

Shift Supervisor's Log; Reactor Operator's Log; Shift Relief Checklist;

Auxiliary Building Operator's Log; Active' Clearance Log; Daily Opera-

ting Surveillance Log; Short Term Instructions (STIs); selected

Chemistry / Radiation Protection Logs; and Outage Jhift Manager's (OSM)

Log.

In addition to these record reviews, the inspector independently

verified clearance order tagouts.

No violations or deviations were identified.

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

were attended by the inspector to observe planning and management

activities.

The facility tours and observations encompassed the following areas:

Security Perimeter Fence; Control Room; Emergency Diesel Generator

Room; Auxiliary Building; Intermediate Building; Battery Rooms;

Electrical Switchgear Rooms; and Reactor Building.

During these tours, the following observations were made:

(1) Monitoring Instrumentation - The following instrumentation was

observed to verify that indicated parameters were in accordance

with the TS for the current operational mode:

Equipment operating status; area, atmospheric, and liquid

radiation monitors; electrical system lineup; reactor

operating parameters; and auxiliary equipment operating

parameters.

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On August -9, while examining the area radiation monitors, the

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inspector noted that monitor RM-A12, which monitors the main

condenser vacuum pump, was removed from the panel and therefore

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out of service. TS 3.3.3.1, item 2.C in table 3.3-6 requires that

grab samples be taken and analyzed for gross activity at least

once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months whenever this monitor is out of service.

The inspector discussed the missing instrument with -licensee

personnel and questioned whether grab samples were being obtained.

The inspector 'also reviewed work requests to determine when the

instrument was removed from service.

As a result of this review, it appears that RM-A12 was removed

from service on the evening of August 7. A grab sample was taken

and analyzed at 6: 27 p.m. on August 7, but only because this

sample - is required to be taken weekly and this time period

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happened to coincide with the due date. Additional grab samples

were not taken until approximately 10:30 a.m., on August 9, when

the inspector asked about the missing instrument. Therefore, the-

time period between samples was approximately 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months .

l Failure to take and analyze these grab samples at least once every.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is contrary to the requirements of TS 3.3.3.1 and is

considered to be a violation.

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Violation (302/85-33-01): Failure to take and analyze grab

samples each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as required by TS 3.3.3.1 when RM-A12 is

out-of-service.

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

the high pressure injection (HPI), emergency feedwater (EFW), and

control rod drive (CRD) systems to verify that the linaups were in

accordance with license requirements for system oper oility and

i that the system drawings and procedures correctly reflect

"as-built" plant conditions.

, On August 2, 1985, the inspectors completed a walkdown of the EFW

system. During this walkdown, which utilized feedwater system

drawing FD-302-081, the inspectors were unable to determine the .'

location of valve FWV-159, a vent valve on the "B" Once Through

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Steam Generator (OTSG) EFW line. The inspectors requested help

from the licensee but, even with the aid of an Auxiliary Nuclear

Operator (AN0), were unable to find the valve. The EFW system has

recently undergone major modifications to install a new Emergency

.Feedwater Initiation and Control (EFIC) system and these modifica-

tions have replaced the section of pipe in which this valve is

supposed to be located. 'After discussions with licensee

personnel, it appears that the EFIC modifications package still

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requires the installation of vent valve FWV-159. The licensee

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i will provide information to the inspector to demonstrate that

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either valve FW-159 actually does exist in the field, or that the

modification package 1for the EFW system has deleted this valve

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because it is no longer required.

Unresolved Item (302/85-33-02): Review the licensee's determina-

tion that valve FWV-159 actually exists or that EFIC modifications

have deleted this valve.

(3) Shift Staffing - The inspector verified that operating shift

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staffing was in accordance with TS requirements and that control

room operations were being conducted in an orderly and profes-

sional manner. In addition, the inspector observed shift turn-

overs on various occasiens to verify the continuity of plant

status, operational problems, and other pertinent plant informa-

tion during these turnovers.

No violations or deviations were . identified.

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(4) Plant Housekeeping Conditions - Storage of material.and components

and cleanliness conditions of various areas throughout the

facility were observed _to determine whether safety and/or fire

hazards existed.

1 No violations or deviations were identified.

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l (5) Radiation Areas - Radiation Control Areas (RCAs) were observed to ,

verify proper identification and implementation. These observa- '

tions included selected licensee conducted surveys, review of

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

, area posting. Area postings were independently verified for

accuracy through the use of the inspector's own radiation moni-

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toring instrument. The inspector also reviewed selected radiation

work permits and observed personnel use' of protecthe clothing,

respirators, and - personnel monitoring devices to assure that the

! licensee's radiation monitoring policies were being followed. .

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No violations or deviations were identified.

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

security barriers are intact, guard forces are on duty, and access

, to Protected Area (PA) is controlled in accordance with the

facility security plan. -Personnel within the PA were observed to

, verify proper display of badges and that personnel requiring ;

i escort were properly escorted. ' Personnel within vital areas were

i 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 equipment,

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, was utilized; and

TS requirements were followed.

The following tests were observed and/or data reviewed:

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SP-102, Control Rod Drop Time Tests;

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SP-204, Class 1 System Leakage Test for Inservice Inspection;

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SP-210, ASME Class 3 Hydrostatic Testing;

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SP-295, Surveillance Capsule Insertion / Removal (and procedure

FP-502 for the supporting data);

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SP-323, Evacuation and Fire Alarm Demonstration;

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SP-332, Monthly Feedwater Isolation Functional Tests;

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SP-340, ECCS Pump Operability;

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SP-349, Emergency Feedwater System Operability Demonstration;

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SP-381, Locked Valve List (Position Verification of Locked

Valves);

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SP-414, High Pressure Injection Flow Verification Test;

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SP-417, Refueling Interval Integrated Plant Response to

Engineered Safeguards Actuation;

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SP-422. RC System Heatup and Cooldown Surveillance;

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SP-440, Unit Startup Surveillance Plan; and

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SP-457, Refueling Interval ECCS Response to a Safety Injec-

tion Test Signal.

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No violations or deviations were identified.

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(9) Maintenance Activities - The inspector observed maintenance l

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activities to verify that correct equipment clearances were in

I effect; Work Requests and Fire Prevention Work Permits, as

required, were issued and being followed; Quality Control

i personnel were available for inspection activities as required; ,

j and TS requirements were being followed. i

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Maintenance was observed and work packages were reviewed for the

following maintenance activities:

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Replacement of engineered safeguards actuation relay for

MUV-41, DWV-160, and CFV-26; ,

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Replacement . of a torque switch for valve EFW-1, emergency

feedwater pump #2 (EFP-2) main condenser hotwell suction

isolation valve; )

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Troubleshooting of the " A'.' Emergency Diesel Generator

(EDG-3A) for relay problems and governor oil change;

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Control Rod Drive Mechanism (CRDM) breaker replacement for

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rod #7 in rod group #7;

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Replacement of EDG-3A in accordance with procedure MP-117;

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Verification of CRDM power patch panels, group patch panels,

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and position indicator panels to. incorporate cycle 6 rod

positions in accordance with maintenance procedure MP-108-A.

! No violations or deviations were identified.

(10) Radioactive Waste Controls - Selected liquid releases and solid

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waste compacting activities were observed to verify that approved

I procedures were utilized, that appropriate release approvals were

obtained, and that required surveys were taken.

No violations or deviations were identified.

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

seismic restraints (snubbers) on safety-related systems were

observed to ensure that-fluid levels were adequate and no leakage

was evident, that restraint settings were appropriate, and that

anchoring points were not binding.

No violations or deviations were identified.

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6. Review of Licensee 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 corrective actions

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

reviewed as they occurred to determine if the TS were satisfied.

LER 84-15 reported that inadvertent Emergency Diesel Generator (EDG)

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starting was caused by failure of an air filter petcock on the air

start system and that replacement of the filter was being evaluated.

This evaluation resulted in a modification (MAR 85-02-01) that replaced

1 the air filters with a more durable filter.

This modification has been installed and tested. The inspectors have

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reviewed the modification package and examined the installation and

note that this installation should prevent recurrence of these events.

This LER is considered to be closed.

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) b. The inspector reviewed Nonconforming Operations Reports (NCORs) to

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

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identified in the reports or during subsequent reviews have been

j accomplished or are being pursued for completion, generic items are ;

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

j reported as required by TSs.

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I All NCORs were reviewed in accordance with the current NRC enforcement

i policy.

As a result of this review, the following items were identified:

I (1) NCOR 85-142 reported that several anchor supports for three cable

trays (trays 134, 135, and 212) are pulling away from ~ the walls

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inside the reactor building. The licensee has investigated the

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condition of the cable trays and a determination as to whether the

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trays are safety-related and/or seismically qualified has been

. made, but this determination was not available for the inspector's

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review.

) Inspector Followup Item (305/85-33-03): Review the licensee's

investigation into the condition of cable trays 134,135, and 212

l and the determination as to whether the trays are safety related ,

and/or seismically qualified.

(2) NCOR 85-118 reported a non-terminated circuit in the Non-Nuclear

i Instrument (NNI) cabinet for the third mechanical seal cavity

! bleedoff pressure transmitter (RC-19A-PT2) for Reactor Coolant

{ Pump (RCP) 3A-2. The licensee is presently evaluating the cause

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and safety significance of this item.

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j Inspector Followup Item (302/85-33-04): Review the licensee's !

evaluation of the non-terminated circuit in the NNI cabinet for

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RC-19A-PT2 for cause and safety significance.

(3) NCOR 85-137 reported that a portion of surveillance procedure

! SP-135, Engineered Safeguards Actuation System Response Time Test,

j which tests the B emergency diesel generator (B-EDG) automatic l

l load sequence timers as required by TS 4.81.2, was not completed

until July 25 at which time the automatic load sequence timers

were found to be operable. This surveillance is necessary to ,

j assure that the B-EDG is operational. The licensee's assessment

1 of this event as documented in this report was only that "No core

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alterations or positive reactivity additions in progress."

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I' The B-EDG was declared inoperable on June 27 at which time the I

A-EDG was made inoperable for planned maintenance. As of July 25,

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the A-EDG was still inoperable. A review of logs by the inspector s

indicates that a core alteration (the reactor vessel . head was

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j installed) occurred on July 10 and that the plant changed modes

! from the refueling mode (Mode 6) to the cold shutdown mode

j (Mode 5) on July 18.

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j Failure to have at least one operable EDG is contrary to the

requirements of TS 3.8.1.2.b. Also, since an operational mode

i change was made without an operable EDG, this event is contrary to

j the requirements of TS 3.0.4 which requires that the Limiting

l Conditions for Operation (LCO) be met prior to changing opera-

j tional modes,

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Violation (302/85-33-05): Failure to have an operable EDG as !

j required by TS 3.8.1.2.b and to meet the LCO requirement prior to

1 changing operational modes as required by TS 3.0.4.

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7. Design, Design Changes and Modifications !

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

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

j changes were performed in accordance with technically adequate and approved.

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procedures, that subsequent testing and test results met acceptance criteria

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

i drawings' and facility procedures were revised as necessary. This review l

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

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The following modification approval records (MARS) were reviewed and/or

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associated testing observed: '

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Replacement of High Pressure Injection Valves MUV-23, MUV-24, MUV-25, i

and MUV-26 in accordance with MAR 83-09-22-01; and functional stroke '

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test in accordance with procedure TP-1-

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Replacement of control complex chiller cooling water flow control

valves CHV-68 and CHV-69 in accordance with MAR 81-03-09-01;

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Integrated Test of the Remote Shutdown Panel in accordance with

procedure TP-18 of MAR 77-07-01;

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Functional testing of the Emergency Feedwater Initiation and Control

(EFIC) system in accordance with the following test procedures of i

i MAR 80-10-66:

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j Tp-6, EFIC System Test

TP-6L, EFIC Fill Rate Test

j TP-9, EFIC Final Test ;

TP-14, Operability Test of EFV-55,56,57,58

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Installation of a fuel oil relief valve on the "A" Emergency Diesel

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j Generator (EDG-3A) in accordance with MAR 85-03-05-01; and

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Replacement of diesel generator air filters in accordance with ;

i MAR 85-02-02-01. ;

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] On August 8, while observing the performance of TP-14 on the EFIC system, '

4 the inspector noted that TP-14 required the completion of the valve check

! list in procedure OP-605, Feedwater System. This valve check list was

completed and signed off within about one hour of the commencement of TP-14.

When test TP-14 was run, the test did not work because there was no flow

j from the emergency feedwater pump (EFP-1). Investigation by the licensee

j determined that valve EFV-7, the discharge stop check valve for EFP-1 was ,

j shut. The valve check list for OP-605 requires this valve to be opened and

] it was signed off on this checklist as open.

l The licensee's Operations Section Implementation Manual (OSIM) requires

valve position indication to be verified by either hands-on operation of the

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l valve, visual observance of valve stem position, or by the fact that the

l system is operating properly (e.g., the system is pressurized and fluid is

! not coming out of vent and/or drain valves). In this-case, the operator

apparently verified the valve position by stem observation only.

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J Based upon these findings, the OSIM valve verification procedure is

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considered to be inadequate to provide positive equipment control. An

! inadequate equipment control procedure is contrary to the requirements of

TS 6.8.1.a. as listed in Regulatory Guide 1.33 and is considered to be a

violation.

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i Violation (302/85-33-06): Failure to have an adequate equipment control

j procedure for the performance of valve lineups as required by TS 6.8.1.a.

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8. Review of IE Notices (IENs)

The inspector discussed IEN 85-58, Failure of General Electric Type AK-2-25

Reactor Trip Breakers, with licensee representatives and has been informed

that appropriate checks of the critical parameters of the breakers has been

accomplished in accordance with procedure PM-118. The inspectors will

verify implementation of the guidelines listed in IEN 85-58.

Inspector Followup Item (302/85-33-07): Verify the implementatior, of the

guidelines discussed in IEN 85-58 in the reactor trip breaker maintenance

procedure PM-118.

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1 9. Loss of Non-Nuclear Instrumentation (NNI) Emergency Procedure Review

The inspector reviewed emergency procedures (EPs), abnormal procedures

l (APs), and verification procedures (VPs) to verify that the loss of NNI

j event was properly covered.

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In 1983, the licensee revised their EPs and APs from " event" oriented

j procedures to " symptom" oriented procedures as recommended in NUREG 0737.

j These revisions also established the verification' procedures primarily for

, use of the shift technical advisors (STAS) to assist operators in the

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analysis of an accident.

! Following their event of February 26, 1980, the licensee committed to

i develop and implement a procedure that addressed a loss of NNI. This

] commitment was confirmed by an NRC Confirmatory Order. The transition from

j event oriented to symptom oriented procedures made coverage of a loss of NNI

event doubtful.

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The inspector compared the old loss of NNI procedure (EP-114) with proce-

dures AP-580, Reactor Protection System (RPS) Actuation, and VP-580, Plant

Safety Verification Procedure.

l The major result of a loss of NNI is a reactor and turbine trip. The old

EP-114 directed the operators to perform actions that would restore the

i plant to a stable condition following these trips. The only apparent t

i difference between EP-114 and the reactor / turbine trip procedures was an

i additional statement to " Assess plant condition using redundant or alterna-

tive instrumentation."

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The new AP-580 directs the actions to be taken to recover from a reactor /

turbine trip. This procedure provides more direction for response to the

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reactor / turbine trip. The VP-580 provides an additional check to verify

that conditions are as expected.

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l Following review of these procedures, the inspector stated that it may be

! appropriate to add a caution or note to the AP or VP to remind operators to

! assure their instrumentation operable. The licensee acknowledged the

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inspector's comment and stated that operator training has taught operators

to verify instrument status by observation of other plant parameters. The

licensee representative also stated that the inspector's comments were

appropriate and that the procedures would be evaluated to determine if such

a note would enhance the procedure.

Inspector Followup Item (302/85-33-08): Review the licensee's evaluation to

add loss of instrumentation caution notes to procedures AP-580 and/or

VP-580.

10. Licensed Operator Training Observations

The inspector observed licensee conducted licensed operator training

activities to evaluate the quality of the training. The following training

activities were observed:

-

Oral boards and system walkthroughs for three candidates;

-

Operation of the newly installed remote shutdown panel (RSP); and

-

Operation of the turbine trip valve on the turbine driven emergency

feedwater pump (EFP-2).

The inspector noted that the training on the RSP was being conducted due to

a commitment made to the Nuclear General Review Committee (NGRC), the

licensee's offsite safety review committee. This training was judged

effective since it provided each operating shift with hands-on training on

the new panel.

Due to the problems at the Davis-Besse plant, as reported in NUREG-1154, the

licensee decided to conduct training on the EFP-2 turbine trip valve to

assure operators (both licensed and non-licensed) were knowledgeable in the

trip valve operation.

No violations or deviations were identified.

ML20135G016

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