ML20199E471

Insp Rept 50-333/97-08 on 971027-1221.Violations Noted. Major Areas Inspected:Operations,Maint & Plant Support

ML20199E471
Person / Time
Site:	FitzPatrick
Issue date:	01/20/1998
From:	Rogge J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20199E413	List: IR 07100027/2012021 ML20199E408 ML20199E471
References
50-333-97-08, 50-333-97-8, NUDOCS 9802020164
Download: ML20199E471 (31)

See also: IR 05000333/1997008

Text

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U.S. NUCLEAR REGULATORY COMMISSION

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Region I

License No.:

DPR-59

Report No.:

97-08

Docket No.:

50-333

Licensee:

New York Power Authority

Post Office Box 41

Scriba, New York 13093

Facility Name:

James A. FitzPatrick Nuclear Power Plant

Dates:

October 27,1997 through December 21,1997

-' Inspectors:

G. Hunegs, Senior Resident inspector -

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R. Fernandes, Resident inspector

J. McFadden, Radiation Specialist

Approved by:

John F. Rogge, Chief, Projects Branch 2

Division of Reactor Projects.

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9802020164 980120

PDR

ADOCK 05000333

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EXECUTIVE SUMMARY

James A. FitzPatrick Nuclear Power Plant

NRC Inspection Report 50-333/97 08 --

Operations

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The shutdown for the forced outage conducted on December 7 was safe and well

controlled. Good command and control, communication and procedure adherence

-were noted. Operator observations, involving a degraded residual heat removal

system pipe support and mislabeled containment isolation valve, demonstrated

good operation practices.- The reactor startup following the outage was performed

in a safe and prudent manner,

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An operator error was made while performing an electrical ground isolation

abnormal operating procedure. Specifically, breakers were operated out of

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- sequence, resulting in the inadvertent automatic operation of high pressure coolant

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injection (HPCl) system valves. Although the valve operation had minor safety

consequences as the HPCI system was out of service for maintenance, the

improper performance of an abnormal operating procedure was determined to be a

--violation. Additionally,,the pre-evolutior, brief for the operations staff was weak in

that the assignment of personnel to conduct breaker manipulations was not made,

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The inspector observed portions of the Safety Review Committee meeting

conducted on November 20-21,1997 and noted that the meeting demonstrated

good safety oversight of station activities.

Maintenance

- During emergency diesel generator maintenance activities, extensive supervisor

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involvement was noted. Additionally, pre-evolution briefs wers conducted for

activities where warranted and procedures were in use. Emergent issues including

a lost lube oil valve disc retaining nut and damaged piston assembly resulted in the -

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work activity taking longer than originally scheduled. These emergent issues were

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effectively addressed through good coordination between operations,

maintenance, quality assurance, technical services and supervisor oversight,

The process to control work activities associated with troubleshooting to locate a

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direct current ground was unsatisfactory and resulted in an invalid engineered

safeguards feature (ESF) actuation signal for the high pressure coolant injection

(HPCl) steam supply valves. The HPCI system was out of service for scheduled

maintenance. Operators did not recognize that the troubleshooting activities made

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the primary containment isolation system (PCIS) function inoperable and therefore

did not enter the appropriate Technical Specification Limiting Condition for

Operation (LCO) action statement. The licensee's immediate corrective actions

were appropriate and the root cause analysis was critical of the oaeration staff's

handling of the troubleshooting activities, but lacked in-depth review of the work

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Executive Summary (cont'd)

control process for the activity. Additionally, the licensee's use of junction boxes

for temporary storage of parts was considered to be a poor work practice. The

failure to enter the TS LCO was a violation.

The work package to prepare for replacement of the low pressure coolant injection

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(LPCI) battery was weak in that the impact of removing a portion of the battery

enclosure ,on LPCI battery operability was not considered prior to beginning the

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work. Although the work was stopped, the licensee subsequently determined that

- the work would not impact battery operability. Additionally, plant drawings for the'

structure were not reviewed prior to the work being performed which contributed

to confusion in performing the task.

Enaineerina

Environmental qualification (EO) components for the high pressure coolant

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injection (HPCI) system were erroneuasly removed from the EQ program in 1993,

and in f act, may not have originally met EQ criteria because of installed

unrecognized test jacks which affected the EQ of the system. The licensee

prepared a justification for continued operation (JCO) which provided reasonable

assurance that the equipment would perform its safety function. The licensee was

slow to pursue the JCO because the impact of this non-EQ component on HPCI

system operability was not initially recognized. Once the problem was recognized,-

.the licensee was aggressive in resolving the issue. The EQ issue was

appropriately resolved through removing the component connection to the test

-Jacks and inserting the previously removed components back into the scope of the

EQ prooram. The licensee's erroneous removal of HPCI components from the EQ -

program was a violation of 10 CFR 50.49.

The licensee's program to monitor safety relief valve (SRV) leakage was effective.

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Licensee management exercised good judgement in electing to shutdown the plant

to effect repairs to leaking SRVs.

Plant Sunoort

Overall, the solid radioactive waste program and ectivities and the program for the

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transportation of radioactive naterials and its related activities were well managed

and effective. The quality assurance audits and surveillance reports were

thorough, programmatic and well documented.

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Training for personnel involved with solid radioactive waste activities was

appropriate in scope and depth. However, the training program was not well

organized and documented and therefore the administration of the training program

was a weakness.

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Executive Summary (cont'd)-

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Radiological controls in administrative areas relative to the elevated radiation levels-

-due to hydrogen injection were proper and adequate.

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- On December 11,1997, an emergency plan joint drill was conducted with the

licensee and Nine Mile Point participating. The emergency preparedness (EP) drill

demonstrated solid performance of the EP staff and licensee organization.

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TABLE OF CONTENTS

EXECUTIVE SUMM ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . II

TA BLE O F CO NT E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

1. O PE R ATI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - . . . . . . . . . . . 1

- 01

Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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'01.1

Operational Safety Verification . . . . . . . . . . . . . . . . . .-. . . . -. . . 1

01.2 Plant Shutdown due to Safety Relief Valve Leakage

.......--...2

04

Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 3

04.1 Battery Ground Isolation Procedure Error (Violation 50-333/97007-01)

...............................................3

07

Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

07.1 Licensee Self Assessment Activities

.....................5

. ll . M AI NT E N A N C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

-M1

Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

M1.1 General Comments on Maintenance and Surveillance Activities . . . 5 -

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M1.2

"A"_ Emergency Diesel Generator Scheduled Maintenance . . . . . . 6

M4 Maintenance Staff Knowladge and Performance . . . . . . . . . . . . . . . . . .

7_ :

M4.1 Invalid Engineered Safeguards Feature (ESF) Actuation and Failure to.

Perform Technical Specification Required Actions While Performing

Troubleshooting (Violation 50-333/97008-02) . . . . . . . . . . . .. . . . .L7

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M4.2 Low Pressure Coolant injection (LPCI) Battery Replacement . . . . . 9

lil . E NGI N EERI NG . . . . . . . . . . . . . . - . . . . . - . . . . . . . . . . . - . . . . . . . . . . . . . . . . . . . . 10

E1- Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10

E1.1

Environmental Qualification of components in the High Pressure

Coolant Injection System (Violation 50-333/97008-03)

...............................................10

E2

Engineering Support of Facilities and Equipment

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E2.1 ' Licensee Monitoring of Leaking Safety Relief Valves (SRVs) . . . . 12

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E8

Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . .- 12

E8.1 = (Closed) Inspector Follow up Item (IFI) 50-333/96007-02......

12

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E8.2 (Closed) Unresolved item (URI) 50 333/95006-03 ........... 13

IV. Plant Support

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R1

Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . . - 14

R1.1 Implementation of the Solid Radioactive Waste Program

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R1.2 Compliance with NRC and Department of Transportation (DOT)

Regulations for Shipping of Low Level Radioactive Waste (LLRW) for

Disposal and Transportation of Other Radioactive Materials

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Table of Contents (cont'd)

R1.3 Elevated Radiation Exposure Levels Due Hydrogen injection . . . .

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Staff Training and Qualification in RP&C (Inspector Follow-up Item (IFI) 50-

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333/97-008-04)

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R7

Quality Assurance in RP&C Activities . .

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R8

Miscellaneous RP&C lssues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

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R8.1

(Closed) Violation 50 333/96007-08 . . . . . . . . . . . . . . . . . . . .

19

R8.2 (Closed) Violation 50-333/96007 09 . . . . . . . , . . . . . . . . . 19

P1

Conduct of EP Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

P1.1

Emergency Plan Drill . . . . . . . . . . . . . . . . . .

........... 19

P8

Miscellaneous EP lssues (EA 98-008)(NCV 97-008-G

,...........20

V. M AN AG EM ENT M EETING S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

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Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

ATTACHMENT

Attachment 1 - Partial List of Persons Contacts

- Inspection Procedures Used

- Items Opened, Closed, and Discussed

- List of Acronyms Used

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Report Details

lummarv of Plant Status

' The unit began this inspection period at 100 percent power. On December 7, the plant

was taken to cold shutdown to repair leaking safety relief valves (SRVs). The plant was

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taken critical on December 13 and returned to 100 percent power on December 17. The

- plant continued operation at 100 percent through the end of the inspection period,

l. OPERATIONS

01

Conduct of Operations'

01.1

- Operational Safety Verification

a.

Insoection Scoce

The inspectors observed plant operation and verified that the facility was operated

safely and in accordance with procedures and regulatory requirements. Regular

tours were conducted of the plant with focus on safety related structures and

systems, operations, radiological controls and security. Additionally, the

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operability of engineered safety features, other safety related systems and on-site

and off site power sources was verified. -The inspectors performed walk downs of-

accessible portions of several safety related systems.

The inspection activities 'during this report period included inspection during

normal, back shift and weekend hours. Regular tours were conducted of the -

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following plant areas:

control room

secondary containment building

radiological control point

electrical switchgear rooms

emergency core cooling system pump rooms

security access point

protected area fence

intake structure

diesel generator rooms

Control room instruments and plant computer indications were observed for

correlation between channels and for conformance with technical specification

(TS) requirements. The inspectors observed various alarm conditions and

confirmed that operator response was in accordance with plant operating

procedures. Compliance with TSs and implementation of appropriate action

statements for equipment out of service was inspected. Plant radiation monitoring

' Topical headings such as 01, M8, etc., are used in accordance with the NRC

standardized reactor inspection report outline. Individual reports are not expected to

address all outline topics.

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system indications reviewed for unexpected changes. Logs and records were

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reviewed to determine if entries were accurate and identified equipment status or -

deficiencies. These records included operating logs, turnover sheets, system

safety tags, and temporary modifications. Control room and shift manning were

compared to regulatory requirements and portions of shift turnovers were

observed. Daily supervisor meetings were attended to assess personnel focus on

risk significant items and plant priorities,

b.

Observations and Findinos

Overall, the licensee operated the plant safely. Plant activities were performed in

accordance with procedures and effective controls were implemented for safe

plant operation. Overall, equipment operability, material condition and

housekeeping conditions were good,

c.

Conclusions

Overall, the licensee operated the plant safely and activities were performed in

conformance with requirements. Effective controls were implemented to achieve -

safe operation of the plant.

01.2

Plant Shutdown due to Safety Relief Valve Leakage

a.

inspection Scoce

During the inspection period the licensee noted an increasing trend in safety relief

valve (SRV) leakage to the torus. On December 1, the licensee elected to enter a

forced outage in order to replace the leaking SRVs. The inspectors witnessed

various portions of the shutdown preparations, power reduction, and reactor

cooldown and depressurization activities on December 7. The inspectors'

objective was to determine the effectiveness of management controls in ensuring

a safe transition to shutdown, in addition, the inspectors observsd portions of the

reactor startup conducted on December 13,1997. Inspector attemion was

focused on reactivity control, operator procedure use and communications,

b.

Observations and Findinas

The unit was shutdown per operating procedure (OP)-65, Start-up and Shutdown

Procedure. Power reduction was performed in accordance with reactor analyst

procedure (RAP)-7.3.16,I'lant Power Changes, and the main generator was

removed from service on December 7,in accordance with applicable ope:ating

procedures. The unit was in cold shutdown at 3:48 a.m. and the reactor mode

switch was taken to the refuel position at 4:47 a.m. on December 8.

The inspectors noted good command and control of unit shutdown activities.

Communications were professional and precise with three-point communications

used. Coordination of various shutdown activities by licensed operators was very

good. Appropriate oversight of personnel during manipulation of the reactor

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controls was noted. For example, a second checker for control rod motion and

selection was stationed. in addition, senior licensee management personnel were

assigned for shift coverage.

Prior to the startup of tha plant, the operations staff noted two plant deficiencies. -

One deficiency involved dislodged grouting material from behind a pipe support in

the residual heat removal system, and the second issue involved a mislabeled

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containment isolation valve. Both issues were adequately addressed by the

licensee prior to startup, and reviewed by the inspectors. The latter event will be

further reviewed following the issuance of a licensee event report (LER).

The startup was characterized by clear operator communications and procedure

use, attentive management oversight, and effective control by shift supervision.

Shift tumover meetings were performed in a controlled manner and crew briefings

were good. Senior operations management personnel were designated to provide

continuous oversight,

c.

Conclusions

The shutdown for the forced outage was safe and well controlled. Good

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command and control, communication and procedure adherence were noted. The

observations by the operators demonstrated good operational practices. The

reactor startup following the outage was performed in a safe and prudent manner.

04

Operator Knowledge and Performance

04.1

- Battery Ground Isolation Procedure Error (Violation 50-333/97007 01)

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Insoection Scone

On October 23, the operators entered abnormal operatino procedure (AOP)-23,

Direct Current (DC) Power System Ground Isolation, in response to indications of a

ground on the B" battery. Testing involving the high pressure coolant injection

(HPCI) logic system had just been completed prior to the ground appearing on the

control room instrumentation, so the control room staff elected to proceed to the

portion of the AOP which isolates the HPCI logic circuitry. During performance of

the procedure, the operators failed to open the power supply breakers for 23MOV-

57 and 23MOV-58, the HPCI liooster pump suction from the suppression pool

downstream and upstream isoldion valves respectively. This resulted in the

valves automatically opening when the correct circuit breaker,71DCB2 Breaker 6,

HPCI Logic Power Supply, was opened in the improper sequence. The event

occurred during a HPCI maintenance limiting condition for operation (LCO) and

therefore the system was already considered inoperable. The inspector reviewed

procedures, plant logs and conducted interviews with station personnel involved in

the performance of the ground isolation procedure.

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b.

Findinos and Ohscrvatiores

AOP-23, DC power System 0 Ground Isolation, provides steps which attempt to

locate the source of a ground in the DC power system. The procedure contains

general steps in the main body of the text and lists the specific breakers to be

utilized in the isolation of the grounds h on attachment to the procedure.- In b.lef,-

the procedure directs the operators to establish communications between the

control room and the operator at the specified breaker, perform any actions

required by the breaker attachment sheet, enter the any applicable LCOs, and open

the isolation breaker. The ground detector in the control room is then monitored

to see the effect, if any, of opening the isolation breaker. The process is repeated

until the ground is isolated. More specifically, in the attachment to the procedure,

tables identify the isolation breaker to be opened, its corresponding circuit or

component, and the actions required prior to opening the isolation breaker. In this

particular event, the operator be.:ame focused on selecting the proper isolation

breaker and omitted the requirements of the proceduro to open the power supply

breakers for 10MOV 57 and 10MOV 58.

In discussion with the plant staff, the inspector learned that the pre-evolution bdef

for the operations staff was not specific. The operators had been monitoring the

ground circuit prior to the alarming condition being reached, taken out the AOP,

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and discussed the most probable circuit to check based on recent HPCI system

testing. The inspector noted that all the control room staff had been included in

the discussions of current plant conditions, including the selection of an additional

operator to perform a peer check of the isolation breaker operation. However, the .

assignment or discussion of who was going to open the breakers to 10MOV-57

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and 10MOV 58 was not discussed as part of the brief.

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The impact of the procedure error was to cause the HPCI booster pump suction to

shift from the condensate storage tank (CST) to the torus. The torus suction

valves are designed to go open on low CST water level or high suppression pool

level to ensure that HPCI has a makeup water source. This action occurred

because the HPCI logic circuitry, following a power loss to the CST level

instrumentation when breaker six was opened, caused the suction valves to

automatically go open. As previously stated, the system was undergoing

maintenance and thus was already considered inoperable. The impact was limited

to unnecessarily challenging the HPCI logic circuitry and cycling valves.

Immediate correct!- n actions were to restore the power to the logic circuitry,

reposition the valves, and re-perform the procedure correctly. The electrical

ground was subsequently located and fixed.

c.

Conclusions

The operator error in performing the actions of the AOP had minor safety

consequences, however, the proper performance of abnormal operating procedures

is of high importance and was determined to be a violation. (50 333/97008-01)

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Additionally, the pre evolution brief for the operations staff was weak in that the

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assignment of personnel to conduct breaker manipulations was not made..

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Quality Assurance in Operations

07.1

Licensee Self-Assessment Activities

a.

insoection Scooe

- During tlie inspection period, the inspectors reviewed multiple licensee self-

assessment activities, including portions of the Safety Review Committee (SRC)

meeting conducted on November 20 - 21,1997. Observations of the SRC

meeting are noted below,

b.

Observations and Findinas

Recent plant history and issues and performance indicators, operational review and

human performance trends were discussed. Specific issues that were discussed in

depth included nuclear personnel turnover and engineering lack of rigor. SRC

members demonstrated a good questioning attitude and good interaction with the' -

Indian Point 3 representative were noted. Follow up items were developed where

appropriate,

c.

Conclusions

. The SRC meeting demonstrated good safety overright of station activities.

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11. MAINTENANCE

M1

Conduct of Maintenance

M1.1

General Comments on Maintenance and Surveillance Activities

a.

Insoection Scooe

The inspectors observed selected maintenance activities to verify that activities

were conducted in a manner sufficient to ensure reliable, safe operation of the

plant. The inspectors observed selected surveillance tests to determine whether

the tests were conducted in accordance with technical specification and other

requirements.

The inspectors observed all or portions of the following work activities:

WR 97-06988-06 Replace "B" control rod drive (CRD) pump and restore

temporary modification 97-095,

WR 97-08389-01 Investigate and repair source of water leakage in main stack

room.

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WR 97-08063-02 Troubleshoot / repair valve positioner, feedwater heater 33E 5A

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drain valve operator.

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WR 97 06476-00 Troubleshoot / repair pre-cooler drain lines.

The inspectors observed portions of the following surveillance activities:

ST-3P

Core spray flow rate and valve inservice test.

ST SD

Average Power Range Monitor Calibration.

ISP 94

Reactor Protection System Electrical Protection Assembly

Functional Test / Calibration.

b.

Observations and Findinas

The inspectors found the work performed under these activities to be professional

and thorough. Technicians were experienced and knowledgeable of their assigned

task. Activities were conducted appropriately and in eccordance with procedural

and administrative requirements. Good coordination and communication were

observed during performance of the surveillance activities.

c.

Conclusions

Overall, the above maintenance and surveillance activities were well conducted,

with good adherence to both administrative requirements and maintenance and

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surveillance procedures.

M1.2

"A" Emergency Diesel Generator Scheduled Maintenance

a.

Insoection Scooe

The "A" emergency diesel generator (EDG) was scheduled for planned

maintenance from November 3 to 5,1997. Activities to be completed incit.ded

routine preventive and corrective maintenance, fuel oil replacement, and power

pack assembly (cylinder liner, pistons and associated components) replacement.

The inspector observed selected activities including procedure use, quality

assurance, and supervisor oversight. During the planned mainunance, emergent

work including a replaced power pack failure and the identification of a missing nut

on a lube oil check valve were also reviewed.

b.

Observations and Findinos

During performance of maintenance procedure (MP) 93.11, the lube oil gallery

supply check valve was inspected due to industry information which documented

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a history of problems with the valve Mechanics identified that the valve disc

retaining nut was missing, and the disc was lying on the bottom of the valve. The

valve is a % inch swing check valve. The licensee initiated deficiency and event

report (DER) 97-1545 to investigate the problem and to analyze the impact of the

missing nut. The check valve is located between the lubo oil cooler and main lube

oil pump discharge in a line used for lube oil warm up when the engine is

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shutdown. On November 5, a licensee quality assurance (OA) inspector found a -

brass nut of similar proportions located outside the screenwell and the licensee

determined that the nut was the missing nut.

The licensee's basis for this determination was that the lube oil cooler had been

disassembled prior to the valve inspection. The fact that the nut was missing was '

not known at the time of the lube oil cooler inspection. Detection of tne nut

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during tube oiler cooler maintenance would be difficult, due to the size of the nut

and the amount of oil present. Since the nut was not found in the locations where ~

lt would be expected to be based on lube oil flow paths, the licensee concluded

that the nut was removed from the cooler without detection during cooler

maintenance. The inspector concluded that the licensee's analysis was

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reasonable.

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Another major activity completed was that the power pack assemblies for all

cylinders were replaced. During EDG post work testing, a high crankcase pressure

alarm was observed after about 15 minutes of EDG operation and the engine was

shutdown. It was determined that one piston and liner was damaged.

Specifically, the bottom end of the piston skirt was broken and other internal parts '

were damaged. The assembly was removed and replaced with a rebuilt

assembly, broken parts were retrieved from the lube oil sump and the lube oil

strainer and filter were changed. -The licensee is awaiting the results of an

equipment failure evaluation for the damaged power pack assembly.-

The EDG limiting condition for operation (LCO) was exited on November 8. The

- delay in completing the work activities was a result of emergent work.

c.

Conclusions

Extensive supervisor involvement was noted. Additionally, pre-evolution briefs

were conducted for activities where warranted and procedures were in use.

Emergent issues including the lost lube oil valve disc retaining nut and the

damaged piston assembly resulted in the work activity taking longer than originally

scheduled. These emergent issues were effectively addressed through good

coordination Letween operations, maintenance, quality assurance, technical

services and supervisor oversight.

M4

Maintenance Staff Knewledge and Performance

M4.1

invalid Engineered Safeguards Feature (ESF) Actuation and Failure to Perform

Technical Specification Reauired Actions While Performing Troubleshooting

(Violation 50-333/97008-02)

a.

Insoection Scooe

On October 24, while performing troubleshooting to locate a ground on the "B"

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DC power system, an inadvertent short across a pair of test jacks in an electrical

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panel caused a partial isolation signal for *.he high pressure coolant injection (HPCI)

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system. The HPCI system was not operating at the time and was in a scheduled

limiting condition for operation (LCO) for maintenance. The expected isolation of

the appropriate HPCI steam supply valves did not occur. The operations staff

subsequently determined that fuses removed from the HPCI logic circuitry during

troubleshooting, prevented the associated isolation valves from automatically

closing. The maintenance activity disabled the primary containment isolation

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function of the HPCI valves without entering the appropriate LCO. Following the

discovery of this, the licensee completed the actions required by Technical

Specifications (TS) by isolating the outboard HPCI steam isolation valve, 23MOV-

60. The inspector reviewed the licensee's root cause analysis for the event,

conducted interviews, attended management meetings on the event, and reviewed

station procedures to assess the event regarding safety significance and work

control processes.

b.

Observations and Findinas

On October 24,1C37, maintenance activities to repair a ground problem were

conducted which rendered the primary containment isolation system (PCIS)

function of the outboard HPCI steam isolation valves inoperable, however, the

applicable LCO action statement was not entered. If one or more of the

containment isolation valves are inoperable. Technical Specifications require, in

part, that the affected penetration be isolated within four hours by use of at least

one deactivated automatic valve secured in the closed position. Operators did not

recognize that PCIS was disabled until after a maintenance error caused a short of

the logic circuitry which caused an invalid engineered safeguards feature (ESF)

actuation signal sixteen hours after disabling the logic.

The root cause analysis identified severai nppropriate actions. These included

the failure to recognize the impact on the PCJ function of the HPCI isolation

valves when removing logic fuses during surveillance test ST 2M, ECCS Trip

Systems Bus Power Monitors Functional Test, disabling the same PCIS function

during trouble shooting without entering the applicable LCO, and failing to enter

the correct LCO when the condition was recognized. Several causes were

identified by the licensee for the inappropriate actions identified above.

Surveillance test ST-2M was inadequate, in that it did not recognize disabling the

PCIS function of the HPCI valves, a less than adequate review of the short form

temporary operating procedure and protective tag out, and inadequate training on

a previous technical specification change which resulted in operators using the

incorrect section of the TS. The licensee developed twelve recommended

corrective actions, including revising procedures to capture the lessons learned,

training and review of the event with operators, and review of all surveillance test

and operating procedures to identify the impact of fuse removal on TS.

The inspector also reviewed the troubleshooting work request which led to the

event and determined that the impact on PCIS was also missed during the work

control process. The inspector noted that this issue was not addressed in the root

csuse analysis. The licensee subsequently reviewed the work control process and

determined that the troubleshooting process for this emergent work item relied on

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the workers to determine the affects of their actions in the field during work

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execution. Had a detailed review of the logic by the workers and the operations

staff been conducted, the potentialimpact of the work could have been identified.

The mind set of the plant staff was that the system was in an existing LCO and

tagged out, therefore work would not impact the plant. This was an incorrect

assumption as identified when a technician inadvertently shorted two terminals in

a junction bux. The issues surrounding the performance of the troubleshooting

were discussed in Licensee Event Report 97-011 and will be addressed when the

LER is reviewed.

c.

Conclusions

The process to control work activities associated with troubleshooting to locate a

DC ground was unsatisfactory and resulted in an invalid ESF actuation. Operators

did not recognize that the troubleshooting activities made the PCIS function

inoperable and therefore did not enter the appropriate LCO action statement. The

licensee's immediate corrective actions were appropriate and the root cause

analysis was critical of the operations staff's handling of the trouble shooting

activities but lacked in-depth review of the work control process for the activity.

The failure to enter the LCO was determined to be a violation (50-333/97008 02).

M4.2

Low Pressure Coolant Injection (LPCI) Battery Replacement

a.

insoection Scone

The inspector observed preparations for the "B" LPCI battery replacement in the

reactor building. The mechanics were trying to remove several sections of metal

panels that make up one of the walls to "B" LPCI battery enclosure, in discussion -

with the maintenance personnel the inspector learned that the wall was much

more intricate than the maintenance crew had expected. The responsible engineer

was notified and after further discussion the licensee determined that the work

should not be continued. A horizontal top corner piece of the structure had been

removed to allow access to the vertical wall sections, but no other pieces were

removed. The inspector reviewed the licensee's work planning and discussed the

activity with the licensee personnel,

b.

Observations and Findinas

Work request (WR) 96-05333-07,was written to remove panels from the west

wall of the "B" LPCI enclosure, to facilitate the installation of a temporary load

handling monorail. The monorail was to be used to replace the existing LPCI

battery cells with new cells during the upcoming scheduled LCO maintene -

period. In follow up interviews with the plant staff the inspector learned that the

original maintenance package did not consider potential fire protection and seismic

issues associated with the removal of various battery enclosure panels, in

discussion with the planning staff the inspector discovered that the enclosure

drawings were not reviewed as part of the work package planning which

contributed to a leck of detailin the work package. The licensee initiated a

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deficiency event report to investigate the issue and utilized another method to

exchange the battery nlis. The licensee's investigation concluded that the battery

enclosure was not necessary for LPCI battery operability. The battery replacement

work package was weak in that the impact of the work activity was not assessed.

Additionally, the work control procedure did not include a requirement to include

all structures, systems and components (SSCs) when reviewing work for seismic

concerns. The corrective actions were appropriate for the above findings,

c.

Conclusions

The inspector concluded that the work package to prepare to replace the LPCI

battery was weak in that the impact of the work on the LPCI battery operability

was not considered prior to beginning the work. Although the work was stopped,

the licensee subsequently determined that the work would not impact battery

operability. Additionally, plant drawings for the structure were not reviewed prior

to the work being performed which contributed to confusion in performing the

tank.

Ill. ENGINEERING

E1

Conduct of Engineering

E1.1

. Environmental Qualification of components in the High Pressure Coolant Injection -

System (Violation 50-333/97008-03)

a.

Insoection Scoce

On October 24,'1997, while performing troubleshooting for a DC ground, a nut

was dropped across test Jacks located in a junction box. The resulting short

caused a HPCl isolation signal. The identification of electrical test jacks on

>

junction boxes for HPCI and RCIC isolation circuits raised questions concerning the

operability and environmental qualification (EO) of the associated components.

The inspector reviewed the licensee's EQ program calculations, justification for

continued operation (JCO) and conducted a physical walkdown of the affected

areas,

b.

Observations and Findinas

On October 24,1997, during troubleshooting on a pressure switch for the source

of a DC ground, a nut was dropped across two hot test points in a junction box,

located in the west crescent area, which initiated a HPCI isolation trip signal.

Fuses pulled for the troubleshocting prevented the actual system isolation. The

inspector noted that the junction box was marked as EQ, however, the pressure

switches locatrd in the junction box had been removed from the EQ program.

Test jacks were also located in the bottom of three additional junction boxes and

were not identified on plant drawings. The concern was that the test Jacks may

not maintain electrical integrity in a high energy line break (HELB) and therefore the

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potential existed to impact the HPCI steam line isolation function. Similar test

Jacks were located in junction boxes associated with RCIC,

At the time of the event, the licensee did not initially recognize the need to

determine the operability of the affected components. Subsequently, an

operability review for HPCI and RCIC was completed on November 4,1997, and

c licensee prepared a JCO, JAF EQ JCO-97-002, Plant Operation with Test

a

Jacks installed in Junction Boxes JB-R2550D snd JB-R2550E for 23 PS-86A,B,C

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ano D to justify continued operation with the test Jacks installed. The licensee's

operability review determined that the test jacks did not affect the operability of

the circuits.

The licensee reviewed the EQ status of the associated junction boxes. It was

determined that on March 3,1993, the licensee deleted approximately

15 c 'ponents from the EQ program for harsh environment plant electrical

equi w snt. The analysis was documented in JAF CALC-HPCI-00820 and was

t

prepared to show that HPCI electrical components would not be subject to a harsh

environment during a HELB. The licensee determined that a nonconservative

asrumption was made in thu calculation which resulted in removing the HPCI

components from the EQ program.

The licensee's corrective actions included walkdowns to identify any other similar,

test Jacks that posed EQ issues. The results of the wa!kdown determined the

extent of the condition was limited to HPCI and RCIC. Additionally, the licensee

removed the electrical connections to the HPCI and RCIC test lugs under a plant

modification. A longer term action review other components removed from the EQ

program was in progress.

.

The inspector noted a station work practice where technicians occasionally used

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junction boxes to temporarily store various objects while working on components.

Typically, this practice was used when technicians ware working on grates where

there was not a readily available place to temporarily store small tools or

components, The licensee reviewed their practices in this area and determined

that this practice would no longer be used,

c.

Conclusions

Following the initial event, the licensee was slow to pulsue the JCO because the

EQ aspects were not readily recognized. The EQ components were erroneously

removed from the program in 1993, and in fact, did not orienally meet EQ criteria

because of the unrecognized installed test jacks. A JCO was prepared which

provided reasonable assurance that the equipment would perform its safety

function. The EQ issue was appropriately resolved through removing the

connection to the test jacks and inserting the previously removed components into

the scope of the EQ program.10 CFR 50.49, Environmental Qualification of

Electric Equipment important to Safety for Nuclear Power Plants, describes EQ

program requirements. Contrary to these requirements, the licensee erroneously

removed HPCI components from the EQ program (VIO 50-333/97008-03).

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E2

Engineering Support of Facilities and Equipment

E2.1

Licensee Monitoring of Leaking Safety Relief Valves (SRVs)

a.

Insoection Scoce

Because of a history of safety relief valve leakage and an industry event where an

SRV inadvertently opened, the licensee monitors SRV leakage. The inspectors

reviewed the licensee's SRV monitoring program and discussed the issue with

licensee personnel. Total SRV leakage recently increased to a point where the

licensee elected to shutdown and repair the leaking SRVs as described in Section

01.2.

b.

Observations and Pndinos

The licensee use: 11 Target Rock 2 wage pilot actuated safety relief valves for

pressure relief of the reactor vessel. The licensee monitors SRV tailpipe

temperature and calculates leakage based on torus heat up rate. The licensee had

previously developed an action plan to schedule a plant shutdown at a torus

.

heatup rate corresponding to an SRV leakage rate of 400 lbs/hr and to shutdown

the plant at a torus heatup rate corresponding to 600 lbs/hr. As of November 21, ,

the licensee was operating with indication of 3 leaking SRVs and a leak rate of

450 lbm/hr with most leakage attributed to "C" SRV main seat leakage.

The inspectors monitored the licensee's performance related to SRV leakage. The

licensee closely tracked SRV performance through daily torus heat up rate

4

calculations and observations of SRV tailpipe temperature. In addition, the

inspectors noted that SRV performance is routinely scheduled for discussion at the

department manager's meetings.

c.

Conclusiqng

The licensee's program to monitor SRV leakage was effective. Licensee

management exercised good judgement in electing to shutdowr the plant on

December 7th to effect repairs to leaking SRVs.

E8

Miscellaneous Engineering issues

E8.1

(Closed) inspector Follow up item (IFI) 50-333/96007-02: Affect of reactor water

cleanup and contro! rod drive flow on alternt

decay heat removal (ADHR) pre-

operational testing. During refueling outage 12, the inspectors noted that the

control rod drive (CRD) and reactor water clean-up (RWCU) sys.tems were in

service providing approximately 240 gallons per minute flow to the reactor vessel

and providing additional refueling cavity mixing during the pre-operational testing

of the ADHR system. This was of concern to the inspectors because the intent of

the pre-operational testing was to ensure that the alternate deca / heat removal

system was capable of removing the heat generated by the spent fuelin both the

reactor vessel and in the spent fuel pool. The test was to demonstrate that the

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ADHR system would remove heat from the reactor cavity and the spent fuel pool

using natural circulation. The inspector's concern was that the added circulation

and cooling water may result in non conservative results with respect to the

capabilities of the system. Additionally, the original calculations for the heat

removal capacity did not account for the additional circulation provided by the

CRD and RWCU systems, which were in service during the test. The licensee

subsequently revised the computer model used to independently verify the General

Electric design calculations on the ADHR system, JAF cal.C DHR 02380, ADHR

System Thermal Hydraulic Analysis, and incorporated the CRD and RWCU system

flows. The revised calculation demonctrated that there was no significant change

in the natural circulation flow characteristics of the reactor vessel and the surf ace

water temperature of the SFP differed from the original calculation by one degree

Fahrenheit, which was considered to not have a materialImpact.

E8.0

(Closed) Unresolved item (URI) 50 333/05006 03: Components Missing from

Control Room Ventilation Drawings. As part of a system walkdown, the inspector

identified that two motor operated dampers (MODS), MOD 113 and MOD 114, and

flow element (FE) 102 were omitted from as built drawing FB 35C, Rev.12,

Equipment Room Heating, Vent and Air Conditioning. Hcwever, the inspector

noted that the components were identified on the cont:ol rocin flow diagram FB-

45A and identified in the Final Safety Analysis Report (FSAR), The inspector was

concerned that despite a temporary modificaHon ar 1 a minor modification being

processed for two safety relat6d components, this deficiency in the as built

drawing was not identified by the licensee. The licensee subsequently determined

,

that the error occurred during original construction and that the drawing was not

required to be updated because of its classification. The System Engineering

Standing Order, (SESOF2, classified the drawing as a type "C" drawing, and per

Design Change Manual, (DCM) 22, drawing changes are not required until five

changes have been posted to the drawing or with the department man %er's

approval. The classification of the drawing signifies that it is used to facilitate

design and maintenance that has a low freauency use. The current practice is for

users to verify the current status of changes to drawings in the drawing status log

prior to use. In the review of the event, however, the nicensee determined that the

modification process should have listed the drawing as an "affected drawing" so

that the appropriate revision process would have been implemented. The

modification removed the motors from the dampers and put them in a f ail safo

position and they currently do not provide any safety function. The inspector

reviewed the corrective actions including revision of the appropriate drawings,

updating the plant equipment database and training on the issue. The inspector

determined the corrective actions were appropriate and that a violation did not

exist because the licensee practices were in accordance with their procedural

requirements for drawing update and were appropriate to the circumstances.

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IV. Plant Suonort

R1

Radiological Protection and Chemistry (RP&C) Controls

R1.1

Implementation of the Solid Radioactive Waste Program

a.

Insoection Scone

The inspector reviewed the licensee's solid radioactive waste program.

-Information was gathered through observation of activities, tours of the

radiologically controlled areas including the radioactive waste building, discussions

with cognizant personnel, and review and evaluation of procedures and

documents,

b.

Observations and Findinas

'In reviewing the implementation of the solid radioactive waste program, the liquid

wasta processing and solidification methods were inspected, the dry active waste

operation was evaluated, and storage locations were inspected. Liquid radioactive

waste was processed through a vendor filter /demineralizer skid. There was a

" green is clean" program, and potentially contaminated materials were sorted and '

frisked to minimize the generation of radioactive waste. Offsite contracted

services were available for equipment / parts decontamination and for

supercompaction or incineration of dry active waste. The interim waste storage

building provided five high bay areas for storage of low level waste and a separate

area with shielded concrete cells for storage of higher level waste such as high

integrity containers (HICs) filled with spent powder or resin. A large portion of this'

stcrage capacity was still avsilable.

.The volume of solid radioactive waste and especially of dry active waste had

steadily and significantly decreased over the last several years. Numerous

initiatives to reduce waste were evident. This included the establishment of a low-

level radioactive waste reduction team. Reusable wrist and ankle straps in place

of masking tape for protective clothing purposes, reusable bags, the wearing of

i

cotton glove liners into the whole body contamination monitors, and the

evaluation / implementation of good radioactive waste reduction practices from

other licensees and from a utility research group have contributed to the decrease

in radioactive waste volume.

Housekeeping was good; aisle ways were clear and clean; storage areas were

clean and orderly; contaminated areas were minimized; radioactive material was

clearly and properly labeled and stored in an orderly fashion.

c.

Cgnplusions

The implementation of the solid radioactive waste program was well managed and

effective.

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R1.2

Compliance with NRC and Deptatment of Transportation (DOT) Regulations for

Shipping of Low Level Radioact!ve Waste (LLRW) for Disposal and Transportation

of Other Radioactive Materials

a.

Insoection Scona

The inspector reviewed the licensee's transportation of radioactive materials.

Information was gathered through observation of activities, discussions with

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cognizant personnel and review and evaluation of procedures and documents.

Temporary Instruction (TI) 2515/133,lmplementation of Revised 49 CFR Parts

100179 and 10 CFR 71 was completed during this review,

b.

Observations and Findings

The shipping records for several past radioactive waste shipments were reviewed.

There records were found to be appropriate and complete The Inspector

observed the transfer of a HIC containing spent resln into a shipping cask. The

waste classification and Department of Transportation (DOT) shipment type

_

determination for this shipment were evaluated and met regulatory requirements.

After this waste shipment was on the public highway, the inspector tested the

emergency response information telephone process (10 CFR 49, Subpart G,

Emergency Response Information) by calling, in the evening, the emergency

response telephone number which was on the radioactive waste manifest. The

inspector's call was answered, and the emergency response information described

in 10 CFR 49.602 was made available by the recipient of the callin a timely

manner,

c.

Conclusions

Good perfolmance was demonstrated in the area of packaging and transportation

of solid radioactive waste.

_

RI .3

Elevated Radiation Exposure Lavels Due Hydrogen injection

a.

inspection Scone

The inspector reviewed the licensee's radiological controls in administrative areas

relative to the elevated radiation levels due to hydrogen injection into the reactor

coolant system, ha.foimation was gathered through observation of a radiation

survey, conduct of a radiation survey, tours of the affected locations, discussions

with cognizant,pers.onnel, and review and evaluation of procedures and

documents,

b.

Observations and Findinga

Elevated dose rates outside the radiologically controlled area (RCA) due to

hydrogen injection, and the radiological controls in those areas were reviewed.

This review focused on the second floor of the old administration building and on

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the second floor of warehouse No.1 since these areas were representative of the

,

affected areas which were outside the RCA, within the protected area, and which

were occupied by mainly administrative personnel. These locations were included

in the licenses's routine radiation survey program.

Dose rates on elevation 290 of warehouse No.1.'arled from 10 to 70 microrem

per hour with reactor power level at 100% based on the licensee's routine

radiation survey results. Dose rates on elevation 286 of the old administrative

building varied from 10 to 150 microrem per hour with reactor power level at

100% based on the licensee's routine radiation survey results and on an

independent survey by the inspector using a calibrated licensee radiation survey

meter. At the time of these surveys, the hydrogen injection rate was

approximately 18.5 cubic feet per minute. The highest dose rates in the latter

area were along the outside window areas, and the dose rates gradually decreased

with distance away from these window areas.

,

Current occupancy factors were observed for each of the two areas and appeared

to approximate 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per wesk. Assuming a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> work week, continuous

occupancy during the work week, and 50 work v.'eeks per year,150 and

10 microrem per hour would equate to 300 and 20 millirem per year, respectively.

For perspective, the average background radiation level in the United States is

about 10 microrem per hour or approximately 100 millirem per year (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per

day and 365 days per year).

10 CFR 19.12, " Instruction to Workers," requires that all individuals who in the

course of their employment are likely to receive in a year an occupational dose in

excess of 100 millirem shall receive radiation protection instruction commensurate

with potential radiological health protection problems present in the workplace.

,

Radiation safety training records for several white badged individuals (non-

radiation workers)in each of these areas were inspected. The individuals were

confirmed to have received general employee training, which includes basic

radiation protection training on an annual basis. The inspector confirmed that the

training included discussion of the guidance in Regulatory Guide 8.13, " Instruction

Concerning Prenatal Radiation Exposure."

10 CFR 20.1502 requires the use of individual radiation monitoring devices for

adults likely to receive a dose in excess of a total effective dose equivalent of

,

500 millirem per year; and for the embryo / fetus of a declared pregnant women for

whom the embryo / fetus is likely to receive a dose in excess of 50 millirem during

the entire pregnancy. Inspector observations noted that a number of the

administrative (non-radiation worker) personnel had been provided individual

radiation monitoring devices even though this was not required by 10 CFR 20.1502. In some of these administrative areas, one would be likely to receive a -

dose in excess of 50 millirem within a nine month period, and, in such a case,

10 CFR 1502 would require the use of a individual radiation monitoring device for

the embryo / fetus of a declared pregnant woman. The inspector confirmed that a

declared pregnant woman had been provided such dosimetry.

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The inspector also reviewed licensee documents titled * Radiological Technical

Information Document (RTID) No. 93 011, Basis for Individual Monitoring

Requirements Within the Restricted Area of the JAFNPP, December 13,1993,"

"First Quarter 1997 Restricted Area Dose Evaluation, April 23,1997," "Second

Quarter 1997 Restricted Area Dose Evaluation, July 18,1997," " Third Quarter

1997 Restricted Area Dose Evaluation, November 3,1997," and "RTID 96-005,

Radiological Assessment of Doses / Dose Rates in Non RCA Occupied Areas,

June 19,1996."

Based on monitoring and survey data by the licensee and the NRC, observations in

several affected areas, and review of training and documents describing the

licensee's evaluation of this issue, the inspector concluded that the radiation

badging requirements in 10 CFR 20.1502 and the training requirements in 10 CFR 19.12 were being followed,

c.

Conclusions

Radiological controls in administrative areas relative to the elevated radiation levels

due to hydrogen injection were proper and adequate.

R5

Staff Trainin9 and Qualification in RP&C (Inspector Follow up ltem (IFI) 50 333/97-

008 04)

a.

Insoection ScRD.t

The inspector reviewed the qualifications and training of selected radioactive

waste personnel. Information was gathered through discussions with cognizant-

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personnel, and review and evaluation of documents,

b.

Observations and Findinas

Training department personnel stated that the training for radioactive waste-

processing, handling / transferring, packaging, and shipping was provided by

contractors and that the training courses were reviewed and approved by licensee

personnel before implementation. The inspector reviewed the course materials

used for the training of the radioactive waste handlers and shippers. The scope

and depth of the course materials was fully adequate. The inspector verified that

these individuals had been recently trained in the aforementioned topics and that

the two individuals who were responsible for classifying waste and determining

DOT chipment type had been retrained on the applicable computer program in mid-

1997. Additionally, it was confirmed that allindividuals authorized to sign

shipping paperwork had received recent training on the shipping reguistions.

However, a documented description of the required training for radioactive waste

processors, handlers /transferors, classifiers, and shippers was not available, and

the training record database was incomplete in that the latest training for the

computer program used for classifying and typing waste shipments had not been

entered. .These administrative deficiencies were considered a program weakness.

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The licensee stated that a matrix of required training and the frequency of same

for radioactive waste processors, handlers /transferors, classifiers, and shippers

would be developed and kept available for review and that the training record

database would be updated and kept updated. This issue will be reviewed during

a subsequent inspection (IFl 50 333/97008-04).

c.

Conclusiong

The training and retraining for the radioactive waste handlers /transferors and

shippers was appropriate in scope and depth, and records of this training were

adequately maintained. However, a documented description of required training

was not available, and the training record database was incomplete. Therefore,

the training program was not well organized and documented.

R7

Quality Assurance in RP&C Activitles

a.

Insoection Scoce

The inspector reviewed the licensee's quality assurance (OA) activities for solid

radioactive waste management and transportation of radioactive materials.

Information was gathered through discucslons with cognizant personnel and

review and evaluation of documents,

b.

Qharvations and Findinas

Audit A9617J, conducted in the Fall of 1996, covered the implementation of the

revised DOT and NRC radioactive material shipping requirements. This audit

resulted in two Deviation and Event Reports (DEPis) and two Recommendations.

-

The inspector's review of the audit checklist and audit report showed that the

audit was thorough and programmatic.

Audit A97 05J, conducted in February of 1997, covered the Process Control

Program (PCP) and Regulatory Guide 1.21. This audit resulted in the issuance of

two DERs and one Recommendationin the PCP area. The audit checklist and

audit report portions dealing with the PCP were reviewed and were found to be in-

depth efforts.

Six QA surveillance reports, performed from November 1996 to September 1997,'

were evaluated. These reports covered receipt inspections of radioactive waste

shipping casks and liners, shipment inspections of radioactive waste in casks,

review of documentation packages for several waste shipments, and the release of

material from the radiologically controlled area. There were no resultant DERs or

recommendations based on these reports. The surveillance reports showed that

the surveillance activities were detailed and well documented.

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c.

Conclusions

The cuality Assurance audits and surveillance reports were thorough,

programmatic, and well documented.

R8

Miscellaneous RP&C lasues

R8.1

(Closed) Violation 50 333/96007 08: Failure to follow plant Technical

Specification for locked high radiation area entry (i.e., contractor in the drywell

l

with his alarming dosimeter turned off). The inspector reviewed the corrective

actions described in the licensee's response letter dated February 21,1997. The

corrective actions were reasonable and comprehensive. No similar problems were

identified.

R8.2

(Closed) Violation 50 333/96007-09: Failure to follow a formal quality assurance

program (i.e., failure to promptly identify and correct a deviation involving the lack

'

of current certification of technicians for use of a computer program code used to

classify shipments). The inspector reviewed the corrective actions described in

the licensee's response letter dated February 21,1997. The corrective actions

were appropriate and complete. No similar problems were identified.

P1

Conduct of EP Activities

P1.1

Emergency Plan Drill

a.

Insoection Scoce

On December 11,1997, an emergency plan joint drill was conducted with the

licensee and Nine Mile Point participating. The purpose of the drill was to

demonstrate that various emergency preparedness functions could be performed

jointly from the emergency operations f acility (EOF). The drill was a partial scale

drill and had limited participation by Oswego County.

The inspector observed and evaluated the performance of licensee emergency

response personnelin the EOF including staffing and activation; facility

management and control; accident assessment and classification; offsite dose

assessment; protective action decision making and implementation; notifications

and communications; and interaction with the Oswego County personnel,

b.

Observations and Findinos

The emergency was properly classified. The reactor condition and emergency was

continuously reassessed._ Environmental sampling teams were appropriately

deployed. Offsite dose assessment and protective action recommendations were

appropriate. Communications within the Emergency Operations Facility were

frequent with proper notifications and interaction with county personnel noted. A

particular strength noted was the good coordination between the emergency

directors from the Nine Mile Point and FitzPatrick f acilities in setting priorities.

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c.

Conclusions

The emergency preparedness drill dernonstrated solid performance of the EP staff

and licensee organization.

P8

Miscellaneous EP lssues (EA 98-008)(NCV 97 008 05)

(Closed) IFl 50 333/97002 03(NCV 97 008 05): Adequacy of emergency

procedures governing evacuation from areas near the new fuel storage vault and

f ailure to meet requirements of 10 CFR 70.24 for new fuel criticality monitors.

This issue involved the f ailure to have in place either an adequate criticality

monitoring system for storage and handling of new (non irradiated) fuel or an NRC

approved exemption to this requirement contained in 10 CFR 70.24. The issue

was previously left as an inspector follow up item pending additionalinternal NRC

guidance regarding the adequacy of the existing monitoring system as well as the

emergency procedures governing evacuation from areas near the new fuel storage

.

vault.

10 CFR 70.24 requires that each licensee authorized to possess more than a small

amount of special nuclear material (SNM) maintain in each area in which such

materielis handled, used or stored a criticality monitorin0 system which will

energize clearly audible alarm signals of accidental criticality occurs. The purpose

of 10 CFR 70.24 is to ensure that, if a criticality were to occur during the handling

of SNM, personnel would be alerted to that f act and would take appropriate

action.

Most nuclear power plant licensees were granted exemptions from 10 CFR 70.24

during the ' construction of their plants as part of the Part 70 license issued to

permit the receipt of the initial core. Generally, these exemptions were not

explicitly renewed when the Part 50 operating license was issued, which

contained the combined Part 50 and Part 70 authority, in August 1981,the

Tennessee Valley Authority (TVA),in the course of reviewing the operating

licenses for its Browns Ferry facilities, noted that the exemption to 10CFR 70.24

that had been granted during the construction phase had not been explicitly

granted in the operating license. By letters dated August 11,1981,and

August 31,1987, TVA requested an exemption from 10 CFR 70.24. On May 11,

1988, NRC informed TVA that the previously issued exemptions are still in effect

even though the specific provisions of the Part 70 licenses were not incorporated

into the Part 50 license." Notwithstanding the correspondence with TVA, the

NRC has determined that, in cases where a licensee received the exception as part

of the Part 70 licenses issued during the construction phase, both the Part 70 and

Part 50 licenses would be examined to determine the status of the exemption.

The NRC view now is that unless a licensee's licensing b. sis specified otherwise,

an exemption expires with the expiration of the Part 70 license. The NRC intends

to amend 10 CFR 70.24 to provide for administrative contcols in lieu of criticality

monitors.

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21

The NRC has concluded that a violation of 10CFR 70.24 existed at FitzPatrick due

to the inadequacy of the existing new fuel vault radiation monitor as a

-

comprehensive criticality monitoring system for new fuel handling and storage.

The NRC has also determined that numerous other licensees have slmilar

circumstances that were caused by confusion regarding the continuation of an

exemption to 10 CFR 70.24 originally issued prior to issuance of the Part 50

license. After considering all the factors that resulted in these violations, the NRC

.

has concluded that while a violation did exist, it is appropriate to exercise

'

enforcement discretion of violations involving Special Circumstances in accordance

with Section Vil B.6 of the " General Statement of Polley and Procedures for NRC -

Entorcement Actions" (Enforcement Policy), NUREG 1600. Pending amendment to

10 CFR 70.24, further enforcement action will not be taken for f ailure to meet

'

l

10 CFR 70.24 provided an exemption to this regulation is obtained by NYPA

before the next receipt of fresh fuel or before the next planned movement of fresh

fuel at FitzPatrick. This item is tracked as non cited violation (NCV 97 008-05).

!

l

V. MANAGEMENT MEETINGS

X1

Exit Meeting Summary

'

The inspectors presented the inspections results to members of the licensee

management at the conclusion of the inspection on January 13,1998. The

licensee acknowledged the findings presented.

,

The inspectors asked the licensee whether any materials examined during the

inspection should be considered proprietary. No proprietary information was

identified.

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ATTACHMENT 1

,

PARTIAL LIST OF PERSONS CONTACTED

Licensee

G. Brownell, Licensing Engineer

M. Colomb, Site Executive Officer

D. Lindsey, General Manager, Operations

J. Maurer, General Manager, Support Services

A. McKeen, Radiologicel and Environmental Services Manager

T. Phelps, Radiological Supervisor

D. Ruddy, Director, Design Engineering

J. Solini, Sr. QA Engineer

D. Topley, General Manager, Maintenance

A. Zaremba, Licensing Manager

INSPECTION PROCEDURES USED

37551 Onsite Engineering

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62707 Maintenance Observations

61726 Surveillance Observations

.

71707 Plant Operations

1

71750 Plant Support

83724 External Occupational Exposure Control and Personal Dosimetry

,

86750 Solid Radioactive Waste Management and Transportation of Radioactive Materials

92702 Follow up on Corrective Actions for Violations and Deviations

i

Tl 2515/133 Implementation of Revised 49 CFR Parts 100179 and 10 CFR 71

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Attachment 1

2

ITEMS OPENED, CLOSED, AND DISCUSSED

Onened

50 333/97008-01

VIO

Improper performance of DC ground abnormal operating

procedure resulted in HPCIlogic actuation.

50 333/97008 02

VIO

Failure to enter Technical Specification Limiting Condition for

Operation while troubleshooting electrical grounds.

50 333/97008 03

VIO

Erroneously removal of HPCI components from the

10CFR50.49 environmental qualification program.

50 333/97008-04

IF!

Radioactive waste training pro 0 ram was not well organized and

documented

50 333/97008 05

NCV Failure to meet 10 CFR 70.24 requirements or to obtain a valid

exemption from this regulation,

i

G91td

50 333/95006-03

URI

Components missing from control room ventilation drawings

50 333/96007 02

IFl

Affect of RWCU and CRD flow on alternate decay heat

removal preoperational testing

50 333/97002-03

IFl

Adequacy of emergency procedures governing evacuation from

areas near the new fuel storage vault and f ailure to meet

requirements of 10 CFR 70.24 for new fuel criticality monitors

50-333/96007 08

VIO

Failure to follow plant Technical Specification for locked high

!

radiation area entry

50-333/97008 05

NCV Failure to meet 10 CFR 70.24 requirements or to obtain a valid

,

'

exemption from this regulation.

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EA 98-008

NCV Adequacy of emeroency procedures governing evacuation

from areas near the new fuel storage vault and f ailure to meet

i

requirements of 10 CFR 70.24 for new fuel criticality monitors.

l

50 333/96007-09 VIO

Failure to follow a formal quality assurance program

Discussed

None

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.

Attachment 1

3

LIST OF ACRONYMS USED

ADHR

Alternate Decay Heat Removal

AOP

Abnormal Operating Procedure

CFR

Code of Federal Regulations

CRD

Control Rod Drive

CST

Condensate Storage Tank

DC

Direct Current

DCM

Design Change Manual

DER

Deficiency & Event Report

DOT

Department of Transportation

EDG

Emergency Diesel Generator

EOF

Emergency Operations Facility

EQ

Environmental Qualification

ESF

Engineered Safety Feature

FE

Flow Element

FR

Federal Register

HIC

High Integrity Container

HPCI

High Pressure Coolant Injection

IFl

Inspection Follow up item

IR

inspection Report

LCO

Limiting Condition for Operation

LER

Licensee Event Report

LLRW

Low Level Radioactive Waste

LPCI

Low Pressure Coolant injection

MOD

Motor Operated Damper

MOV

Motor Operated Valve

MP

Maintenance Procedure

NCV

Non-Cited Violation

NRC

Nuclear Regulatory Commission

OP

Operating Procedure

PCIS

Primary Containtnent isolation System

PCP

Process Control Program

QA

Quality Assurance

QC

Quality Control

RAP

Reactor Analyst Procedure

RCA

Radiological Controlled Area

RP&C

Radiological Protection and Chemistry

RTID

Radiological Technical Information Document

RWCU

Reactor Water Clean-Up

SESO

System Engineer Standing Order

SNM

Special Nuclear Material

SRC

Safety Review Committee

SRV

Safety Relief Valve

SSC

Structures, Systems & Components

Tl

Temporary instruction

TS

Technical Specification

.

.

Attachment 1

4

-TVA

Tennessen Valley Authority

UFSAR

Updated Final Safety Analysis Report

VIO

Violation

WR

Work Request

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