ML17059C125

Insp Repts 50-220/98-05 & 50-410/98-05 on 980412-0523. Violations Noted.Major Areas Inspected:Reviews of Licensee Activities in Functional Areas of Operations,Engineering, Maint & Plant Support

ML17059C125
Person / Time
Site:	Nine Mile Point
Issue date:	07/07/1998
From:	Doerflein L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17059C123	List: ML17059C122 ML17059C124 ML17059C125
References
50-220-98-05, 50-220-98-5, 50-410-98-05, 50-410-98-5, NUDOCS 9807140059
Download: ML17059C125 (80)

See also: IR 05000220/1998005

Text

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket/Report Nos.:

50-220/98-05

50-41 0/98-05

License Nos.:

DPR-63

NPF-69

Licensee:

Niagara'ohawk Power Corporation

P. O. Box 63

Lycomnin, NY 13093

Facility:

Nine Mile Point, Units

1 and 2

Location:

Scriba, New York

Dates:

April 12 - May 23, 1998

Inspectors:

B. S.

T. A.

J. G.

L. A.

R. C.

R. A.

Norris, Senior Resident Inspector

Beltz, Resident Inspector

England, Reactor Engineer

Peluso, Radiation Specialist

'Ragland, Radiation Specialist

Skokowski, Resident Inspector

'I

Approved by:

Lawrence T. Doerflein, Chief

Projects Branch

1

Division of Reactor Projects

9807i4005980707

PDR

• DOCK 05000220

G

PDR

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TABLE OF CONTENTS

page

TABLE OF CONTENTS

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EXECUTIVE SUMMARY

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SUMMARYOF ACTIVITIES...: .

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Nuclear Regulatory Commission (NRC) Staff Activities ..

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

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General Comments................ ~........ ~...

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M8

Miscellaneous Maintenance Issues..........................

M8.1

(Closed) URI 50-220/96-01-03:

Inadequate Testing of Unit 1

Control Room Annunciators.... ~..........

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M8.2

(Closed) URI 50-410/96-10-01:

Post-Maintenance

Testing of

the Unit 2 Main Steam Line Radiation Monitors ... ~........

M8.3

(Closed) VIO 50-410/96-10-03:

Procedure Changes Not in

Accordance with TS Requirements

M8.4

(Closed) VIO 50-410/97-02-02:

Missed Unit 2 HPCS Actuation

, Instrumentation TS Surveillance Test ...................

M8.5

(Closed) LER 50-410/98-09:

Missed Battery TSSR Due to

Inappropriate Interpretation.

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I. OPERATIONS

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Conduct of Operations ......,,......................;......

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01.1

General Comments...................................

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01.2

Unit 1 - Failure to Follow Surveillance Test Procedure ~..........

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01.3

Conduct of Unit 2 Core Off-load Operations (60710) ...........

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01.4

Dropped Double Blade Guide During Unit 2 Off-load Operation .... 3

08

Miscellaneous Operations Issues .............."...........

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08.1

(Closed) LER 50-410/98-05:

Reactor Water Cleanup Isolation

on High Differential Flow Caused by Relief Valve Lifting......

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08.2

(Closed) LER 50-410/98-06:

Engineered Safety Feature Actuations

Due to Partial Loss of Offsite Power

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08.3

(Closed) LER 50-410/98-08;

HPCS Out of Service with One

Division of RHS in Suppression

Pool Cooling .......

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III. ENGINEERING..........

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Conduct of Engineering......... ~.......,

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General Comments..... '........

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E1.2

Unit 2 Emergency Diesel Generator Fuel Line

E3

Engineering Procedures

and Documentation

E3.1

Unit 2 ECCS Suction Strainer Modification

Leak ..........

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Table of Contents (cont'd)

E8

Miscellaneous Engineering Issues................. ~............ ~... 14

E8.1

(Closed) VIO 50-220 5, 50-410/96-01-05:

Failure to Complete

Safety Evaluation Prior to Installation of Temporary Modification .. 14

E8.2

(Closed) URI 50-220 5 50-410/96-14-02:

Potential Over-

pressurization Concerns Relative to NRC Generic Letter 96-06 ..

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E8.3

(Closed) URI 50-220/97-12-08:

Impact of Drywell to Wetwell

Bypass on Containment Pressure...... ~..................

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E8.4

(Closed) LER 50-220/98-05:

Unrecognized Violation of TS

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E8.5

(Closed) LER 50-220/98-06:

Design Deficiency Associated with

CREVS Radiation Monitors ............ ~...............

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(Closed) LER 50-410/98-07:

TS 3.0.3 Entry Due to Missed Logic

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System Functional Testing of Loss of Voltage and Degraded

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E8.7

(Closed) Part 21 50-220/98-01:

Defective GE SBM-Type

Switches at Unit

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IV. PLANT SUPPORT

R1

Radiological Protection and Chemistry (RPSC) Controls

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Radiological Environmental Monitoring Program Implementation

R1.2

Meteorological Monitoring Program Implementation ........

R1.3

Unit 1 and Unit 2 Tours

R1.4

Unit 2 Refueling Outage .. ~.........

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ALARAGoals and Initiatives ........ ~......

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Quality Assurance

in RPSC Activities.......................

R7.1

Quality Assurance Audit Program

R7.2

Quality Assurance of Analytical Measurements .. ~...... ~;

R7.3

Deviation Event Reports and Self Assessments........

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Status of Security Facilities and Equipment ~..... ~............

S2.1

Tour of the Protected Area Perimeter..... ~....

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S8, Miscellaneous Security and Safeguards

Issues

S8.1

(Closed) URI 50-220 5 50-410/96-06-06:

Fitness-for-Duty

Random Selection Process Software Altered.............

S8.2

Administrative Closure of Escalated Enforcement Items .....

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V. MANAGEMENTMEETINGS........................,................

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Exit Meeting Summary ....................................

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ATTACHMENT

ATTACHMENT1

Partial List of Persons Contacted

Inspection Procedures

Used

Items Opened, Closed, and Updated

List of Acronyms Used

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EXECUTIVESUMMARY

Nine Mile Point Units 1 and 2

50-220/98-05 & 50-41 0/98-05

April 12- May 23, 1998

This NRC inspection report includes reviews of licensee activities in the functional areas of

operations, engineering, maintenance,

and plant support.

The report covers a six-week

period of inspections and reviews by the resident staff and regional specialists in the areas

of environmental monitoring and outage radiation protection.

PLANT OPERATIONS

During performance of a Unit 1 surveillance test, the containment spray raw water inter-tie

check valve did not open with the required torque and the station shift supervisor (SSS)

failed to enter the core spray system TS 3.1.4.d action statement,

as required by the

surveillance test.

The relieving SSS identified the procedural non-compliance and took

prompt and appropriate action to comply with the surveillance procedure.

The failure to

properly implement the surveillance test is a violation of TS 6.8.1.

(VIO 50-220/98-05-01)

While transferring a double blade guide (DBG) from the spent fuel pool to the reactor

vessel, the DBG became disengaged

from the grapple and came to rest in the fuel transfer

canal.

NMPC determined that the root cause was the refueling crew did not properly verify

engagement

of the grapple.

NMPC's root cause investigation was methodical and

thorough, the root cause determination was technically sound, and the corrective actions

adequately addressed

the cause.

MAINTENANCE/SURVEILLANCE

During this inspection period, the NMPC staff self-identified that the TS required service

test of the Unit 2 Division I battery was not completed during the previous two refueling

outages.

NMPC had improperly credited the battery cyclic performance test for satisfying

the requirements of the service test.

NMPC requested

and was granted

a Notice of

Enforcement Discretion (NOED) to avoid the consequential

TS required shutdown.

The

NOED was exited on May 2; 1998 upon the unit achieving Cold Shutdown conditions and

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the service test was completed satisfactorily on May 7, 1998. Notwithstanding, the failure

to have properly service tested the Division I battery, since April 1995, is a violation of TS 4.8.2.1.d.

(VIO. 50-410/98-05-02)

ENGINEERING

During surveillance testing of the Unit 2 Division II EDG, a fuel leak developed between the

fuel filter and the fuel injectors.

NMPC determined that the leak was caused by vibration

of the fuel supply piping, which caused fretting of the pipe at a pipe support.

Subsequent

licensee investigation identified notable, but less severe, fretting on the Division I EDG fuel

supply piping. The fuel line supports were installed in 1993, but the specific design

change to install a protective grommet.was not adequately incorporated into the final

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Executive Summary (cont'd)

design package.

This is a violation of 10CFR50, Appendix B, Criterion III, "Design

Control." (VIO 50-410/98-05-03)

The design and installation of the new ECCS pump suction strainers appeared

adequate to

ensure sufficient net positive suction head for the pumps in the event of a loss of coolant

accident (LOCA).

During a review of Unit 1 operating procedures,

NMPC identified that the normally open

vent valves on the containment spray raw water heat exchangers

violated secondary

containment integrity, in that it provided a potential release path from the reactor building

to the environment.

This licensee identified and corrected violation of secondary

containment integrity requirements was not cited.

During a review of the control room emergency ventilation system initiation.logic, NMPC

determined that the system would not automatically initiate, as required.

Specifically, the

system would not automatically start as a result of a main steam line break or a loss of

coolant accident.

This licensee identified and corrected violation of 10CFR50, Appendix B,

Criterion XI, "Test Control," was not cited.

The inspectors observed that NMPC's follow-up of the Part 21 report concerning GE SBM-

type control switches and their identification of the susceptible switches at Unit 1 was

thorough and an example of an improving questioning attitude by the engineering staff.

PLANT SUPPORT

The licensee effectively maintained and implemented the Radiological Environmental

Monitoring Program in accordance with regulatory requirements.

The licensee performed,a

comprehensive review of an anomalous. indication of Iodine 131 in an environmental milk

sample.

Overall, the licensee effectively maintained meteorological monitoring system operability,

and satisfactorily performed channel calibrations and channel functional tests for the

meteorological instrumentation, with the exception of the wind speed channel.

The failure

to perform the channel calibration of the wind speed channel according to the channel

calibration definition in TS 1.4, in that the accuracy of the entire wind speed chan'nel was

not measured from the sensor to the channel output, constitutes

a violation of Unit 2 TS

3/4.3.7.3.

(VIO 50-410/98-05-06)

Housekeeping was adequate

in that aisles and walkways were clear and free of debris,

radiological boundaries and postings were clear, and access controls to radiologically

controlled areas were effective.

Radiological controls for outage work were well planned and health physics personnel

maintained close oversight of work.

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Executive Summary (cont'd)

Procedure S-RPIP-5.4, "Dose Tracking and Timekeeping," lacked clarity with regard to the

method for determining the available administrative extremity exposure,

and several

examples of inaccurate determinations of available administrative extremity exposure were

identified.

ALARAgoals were effectively used as a tool to aid radiological planning to minimize

radiation exposure.

Numerous ALARAinitiatives including publication of a pre-outage

report, use of cameras,

use of temporary shielding, planned reactor vessel nozzle hydro

washes,

and an attempt to chemically decontaminate the reactor recirculation system

demonstrated

management

support and a commitment to maintaining radiation exposures

ALARA.

The contractor laboratory continued to implement effective QA/QC programs for the REMP,

and continued to provide effective validation of analytical results.

The laboratory

demonstrated

the ability to accommodate

and incorporate difficultmedia and geometries

into the program.

The programs are capable of ensuring independent checks on the

precision and accuracy of the measurements

of radioactive material in environmental

media.

The DER system and the self-assessment

program were effective in their use to identify,

evaluate, and resolve radiological program deficiencies.

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REPORT DETAILS

~ Nine Mile Point Units 1 and 2

50-220/98-05 5 50-410/98-05

April 12- May 23, 1998

SUMMARYOF ACTIVITIES

Niagara Mohawk Power Corporation (NlVIPC) Activities

Unit 1

Nine Mile Point Unit 1 (Unit 1) started the inspection period at full (100%) power.

On April

21, 1998, Unit 1 entered

a 7-day Technical Specification (TS) Limiting Condition of

Operation (LCO) due to the determination that the control room emergency ventilation

system (CREVS) would not automatically initiate as designed

(see Section E8.5 of this

inspection report).

Because the repairs could not be completed before the expiration of the

LCO, Unit 1 was shutdown on April 27.

Following CREVS modifications, the unit was

restarted shortly after the end of the inspection period, and obtained full power on June 1,

1 998.

Unit 2

Nine Mile Point Unit 2 (Unit 2) started the inspection period at 91% power, in a coast-

down condition as they neared the next refueling outage.

On May 2, the unit was

shutdown to start the sixth refueling outage (RFO6). The unit remained shutdown through

the end of the inspection period.

Nuclear Regulatory Commission (NRC) Staff Activities

Ins ection Activities

The NRC resident inspectors conducted inspection activities during normal, ba'ckshift, and

deep backshift hours.

In addition, specialist from Region

I conducted inspections in the

area. of environmental monitoring and outage radiation protection.

The results of the

inspection activities are contained in the applicable sections of this report.

U dated Final Safet

Anal sis Re ort Reviews

While performing the inspections discussed

in this report, the inspectors reviewed the

applicable portions of the Updated Final Safety Analysis Report (UFSAR) related to the

areas inspected.

The inspectors verified that the UFSAR wording was consistent with the

observed plant practices, procedures

and/or parameters.

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01

Conduct of Operations

01.1

General Comments

71707 '.

OPERATIONS

The resident inspectors conducted frequent reviews of ongoing plant operations to

determine if the units were operated safely and in accordance with licensee

procedures

and regulatory requirements.

The reviews included tours of accessible

and normally inaccessible

areas of both units, verification of engineeied safeguards

features

(ESF) system operability, verification of adequate

control room and shift

staffing, verification that the units were operated

in conformance with technical

specifications, and verification that logs and records accurately identified equipment

status or deficiencies. 'In general, the conduct of operations was professional and

safety-conscious;

specific events and noteworthy observations

are detailed in the

sections below.

01.2

Unit 1 - Failure to Follow Surveillance Test Procedure

a.

Ins ection Sco

e 71707

The inspectors reviewed the circumstances

surrounding the failure of Unit 1 to

perform actions specified by a surveillance test procedure due to unsatisfactory

surveillance test data.

b.

Observations

and Findin s

On April 22, 1998, during performance of NMPC surveillance test Procedure

N1-ST-

Q28, "Containment Spray Raw Water Inter Tie Check Valve'Quarterly Operability

Test," check valve 93-64 (containment spray raw water sub-loop 122 to core spray

loop 12 testable check valve) failed to open with the required torque.

The.day-shift

station shift supervisor (SSS) was notified of the valve failure at 12:07 p.m.; he

entered Unit 1 TS 3.3.7.b, with a 15-day LCO for an inoperable containment spray

loop and DER 1-98-0960 was initiated to address the concern.

At 8:43 p.m., the night-shift SSS noted that actions contained in N1-ST-Q28 had

not been completed.

Note

1 after Step 10.1.1 (Operations Review of the

Acceptance Criteria) stated that if check valve 93-64 failed, then loop 12 was to be

considered inoperable; subsequently,

TS LCO 3.1.4.d needed to be entered, which

required a shutdown be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in a cold shutdown condition

within the next 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

Note 2 stated that the LCO could be exited if one of the

blocking valves was closed.

The SSS directed both blocking valves shut, and

1 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.

The NRC inspection manual procedure or temporary instruction

(Tl) that was used as inspection guidance is listed for each applicable report section.

4

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exited LCO 3.1.4.d at 9:20 p.m. The failure to adhere to procedure N1-ST-Q28 is a

violation of the Unit 1 TS 6.8.1.

(VIO 50-220/98-05-01)

The inspectors discussed the check valve failure with the system engineer, who

stated that check valve 93-64 appeared to have been hydraulically locked.

The

system was vented and the valve retested satisfactorily.

The system engineer

stated that the mechanism for the hydraulic locking would be investigated further.

C.

Conclusion

During performance of a Unit 1 surveillance test, the containment spray raw water

inter-tie check valve did not open with the required torque and the station shift

supervisor (SSS) failed to enter the core spray system TS 3.1.4.d action statement,

as required by the surveillance test.

The relieving SSS identified the procedural non-

compliance and took prompt and appropriate action to comply with the surveillance

procedure.

The failure to properly implement the surveillance test is a violation of

TS 6.8.1.

(VIO 50-220/98-05-01)

01.3

Conduct of Unit 2 Core Off-load 0 erations

60710

The inspectors observed licensee and contractor, General Electric (GE), conduct of

operations during Unit 2 core off-load. The inspectors observed the evolution from

the control room, and the refuel floor, during both normal and back shift hours.

The

inspectors also reviewed applicable procedures

and TS to verify licensee

compliance.

The inspectors observed operations staff and GE personnel perform fuel movement

from the refuel bridge and considered the evolutions well controlled.

NMPC and GE

personnel exhibited good formal communication, and completed the evolution in

accordance with procedures.

'Before RFO6, NMPC replaced the triangular refueling mast with a heavier round

mast that included an installed camera and monitoring system.

The modification

required a TS amendment, which was approved by the NRC, via letter dated April

16, 1998.

During the off-load evolutions, no concerns were noted with'the

operation of the new refueling mast.

Moreover, the inspectors noted that the

installed camera system greatly enhanced the operators'bility to verify fuel bundle

serial numbers before grapple engagement.

01.4

Dro

ed Double Blade Guide Durin

Unit 2 Off-load 0 eration

Ins ection Sco

e 60710 93702

The inspectors reviewed the circumstances

surrounding the dropped double blade

guide (DBG) during the Unit 2 core off-load. A DBG is used to provide lateral

support for fully inserted control rods during off-load conditions; wHen fuel is in the

reactor vessel, control rods are supported laterally by the surrounding fuel bundles.

The inspectors assessed

the licensee's

response to the event, including the

4

T

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immediate actions, root cause determination, and corrective actions.

In addition,

the inspectors visually observed the location of the DBG as it rested in the transfer

canal, monitored a Station Operating Review Committee (SORC) meeting associated

with the event, reviewed the applicable procedure and DER, and discussed

related

issues with the Senior Reactor Operator (SRO) on the refuel bridge at the time of

the event, members of the root cause analysis team, and the Unit 2 Plant Manager.

Observations

and Findin s

On May 19, 1998, while transferring a DBG from the spent fuel pool (SFP) to the

reactor vessel, the DBG became disengaged

from the grapple and fell onto the fuel

transfer canal.

The refueling bridge SRO immediately stopped all fuel handling

activities and informed the control room of the situation.

Management and

technical support staff assisted

in the evaluation of the situation.

An underwater

video camera was used to view the grapple and the DBG bail (handle)

~ The grapple

was closed and the,"engaged" light on the refuel platform control panel was

illuminated. The DBG bail handle appeared to be intact with no indication of failure.

The licensee issued

a DER to record the event and a root cause analysis team was

formed.

Upon being informed of the event, the inspectors performed a visual

inspection of the refuel floor, the SFP and reactor cavity, and considered the DBG to

be in a stable condition.

NMPC developed

a plan and retrieved the DBG. They

inspected it for damage,

and returned it to its previous location in the SFP.

During NMPC's root cause investigation, the grapple and the DBG handle were

measured,

inspected

and tested to determine how the DBG may have dropped from

the grapple.

GE, the manufacturer of the grapple, was contacted for assistance.

By evaluating the dimensions of the grapple and attempting to duplicate the event,

NMPC determined the following as the most probable scenario:

When the grapple

was brought into position over the DBG, it was slightly mis-oriented such that the

DBG bail handle wedged diagonally between the bail handle channel on one side of

the grapple shroud and the corner of the shroud on the opposite side.

In this

position, the grapple was ready to engage the bail handle; however, the mast was

approximately one-inch higher then normal for DBG engagement.

When the signal

was given for the grapple to engage the bail, the middle hook of the grapple

traveled to the normal engaged position, but the outside hook came to rest against

the bail handle.

In this condition, the "engaged" light would not have been

illuminated since the necessary

contacts within the circuitry would not be closed.

The DBG was lifted to the "full-up"position and moved to the fuel transfer canal.

To transverse through the fuel transfer canal, the bridge operator rotated the mast

90'o align the DBG with the transfer canal.

As the bridge accelerated,

the DBG

slipped from the single grapple hook; at this time; the "grapple engaged" light

illuminated since the grapple hooks would have closed after the DBG fell.

The licensee tested this scenario several times with confirmatory results.

All details

noted during the tests corresponded to those described by the refueling operators

following the event with one exception.

During the post event interviews, the fuel

handler and spotter stated that the "engaged" light was lit before the DBG was

lifted. During the tests, the "engaged" light was never received before the blade

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guide lift. Based on the above, the licensee determined that the root cause

was'hat

the refueling crew did not properly verify grapple alignment, resulting in a

partial engagement of the DBG.

Prior to continuing core off-load, NMPC implemented several corrective and

preventive actions, as documented

in DER 2-98-1415.

The actions included:

(1)

adding a column to the fuel move sheets to document the receipt of the "grapple

engaged" light prior to lifting fuel assemblies

or blade guides, and (2) training of the

refueling crews on the issues associated with the event.

Based on an analysis from

GE, NMPC further concluded that physical design differences between the blade

guide bail handle and the fuel assembly bail handle would prevent a similar

occurrence with a fuel assemble.

NMPC completed core off-load on May 23,

without further incident.

Based on the inspectors review, NMPC's root cause investigation appeared

methodical and thorough, the root cause determination was technically sound, and

the corrective actions adequately addressed

the cause.

C.

Conclusion

While transferring a double blade guide (DBG) from the spent fuel pool to the

reactor vessel, the DBG became disengaged from the grapple and came to rest in

the.fuel transfer canal.

NMPC determined that the root cause was the refueling

crew did not properly verify engagement

of the grapple.

NMPC's root cause

investigation was methodical and thorough, the root cause determination was

technically sound, and the corrective actions adequately addressed

the cause.

08

Miscellaneous Operations Issues

08.1

Closed

LER 50-410 98-05: Reactor Water Cleanu

Isolation on Hi h Differential

Flow Caused

b

Relief Valve Liftin

92700

On March 17, 1998, while placing a reactor water cleanup (RWCU) system filter

demineralizer in service, Unit 2 experienced

an automatic isolation of the 'system

due to high differential flow. The operators verified proper system response,

implemented the TS action statements,

and wrote a DER to investigate the cause of

the system isolation.

Subsequently,

NMPC identified that the filter demineralizer

relief valve lifted, which caused the high differential flow and resulted in the system

isolation.

The relief valve was disassembled

and NMPC determined that the valve

disk seating surface was degraded.

A new valve was installed.

The day of the

event, the inspectors discussed the RWCU system isolation with the Unit 2 licensed

operators,

and verified the appropriate implementation of TS.

During this inspection period, the inspectors discussed

the root cause determination

with the system engineer and observed the SORC's review of the associated

DER

and Licensee Event Report (LER) 98-05.

No concerns were identified. The

inspectors verified that the LER was completed in accordance with the requirements

of 10CFR50.73.

This LER is"closed.

4t

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Closed

LER 50-410 98-06:

En ineered Safet

Feature Actuations Due to Partial

Loss of Offsite Power 92700

The technical issues associated with this LER were reviewed and documented

in

Section 01.2 of inspection report (IR) 50-410/98-02.

However, the LER provided

additional information regarding two equipment performance abnormalities that

occurred during the loss of Line 5, one of the two 115KV sources of offsite power.

Specifically, the Division II hydrogen/oxygen (H,O,) sample pump and the Division II

cable spreading area unit cooler both tripped during the transient and had to be

manually restarted.

These components were not expected to trip since Division II

was being powered by Line 6, the other 115KV offsite power source.

NMPC

evaluated the circuits associated with these components

and verified that there was

no inter-tie with Division I or III power.

However, they did identify a loose

connection within the control circuitry for each component.

Also, the loss of Line 5

caused

a transfer of load to Line 6, resulting in a momentary drop in Line 6 voltage.

NMPC determined that the loose connections, combined with the voltage drop on

Line 6, was sufficient to cause these components to trip. Although loose

connections existed, NMPC concluded that the equipment was able to perform the

intended design function. The inspectors considered NMPC's conclusion to be

technically sound.

The inspectors verified that the LER was completed in accordance with the

requirements of 10CFR50.73.

This LER is closed.

r

Closed

LER 50-410 98-08: HPCS Out of Service with One Division of RHS in

Su

ression Pool Coolin

Ins ection Sco

e 92700

The inspectors reviewed the LER, the UFSAR, related DERs, and the NMPC design

specification related to the emergency core cooling system (ECCS).

In addition, the

inspectors discussed the issue with the system engineer and a licensing engineer,

and observed

a Unit 2 SORC meeting regarding the issue.

The inspectors verified

the completion of the LER in accordance with 10CFR50.73.

Observations

Findin s and Conclusions

In March 1998, as a result of operating experience reviews, NMPC initiated a DER

to investigate

a concern regarding the operability of the residual heat removal

system (RHS) while operating in the suppression

pool cooling mode.

Specifically, as

originally designed, the closing time (120 seconds) of the suppression

pool cooling

valve was slower than the opening time (65 seconds) of the low pressure coolant

injection (LPCI) mode injection valve. Therefore, during suppression

pool cooling,

full LPCI flow. would not be available until the suppression

pool cooling valve was

completely closed.

As a result, during suppression

pool cooling, the RHS division

would be inoperable for the LPCI mode.

NMPC reviewed operational history and

determined that, based on this scenario, TS requirements had not been exceeded.

Afterwards, NMPC determined that during RHS full flow test, the system would also

J

4

be inoperable, because the test configuration uses the suppression

pool cooling

valve.

Licensee review identified that, on January 25, 1996, Unit 2 had operated for

twelve minutes with the high pressure

core spray (HPCS) system out-of-service at

the same time that Division II RHS was in the full flow test configuration.

In this

configuration, TS 3.0.3 was applicable since no specific TS requirement existed to

cover this condition.

Based on the information provided in the LER, and a review of

the applicable TS, the inspectors concluded that no TS violation had occurred,

because the twelve minutes that HPCS and Division II RHS were concurrently

inoperable was within the one hour LCO allowed by TS 3.0.3.

However, the failure

to have adequately reviewed the suitability of the suppression

pool cooling/full flow

test valve closing time with respect to the LPCI functions of RHS constitutes

a

.

violation of minor significance and is not subject to formal enforcement action.

The inspectors verified that the LER was completed in accordance with the

requirements of 10CFR50.73.

This LER is closed.

II. MAINTENANCE

M1

Conduct of Maintenance

M1 ~ 1

General Comments

61726 62707

The resident inspectors periodically observed plant maintenance

activities and the

performance of various surveillance tests.

As part of the observations, the

inspectors evaluated the activities with respect to the requirements of the

Maintenance

Rule, as detailed in 10CFR50.65.'n general, maintenance

and

surveillance activities were conducted professionally, with the work orders (WOs)

and necessary

procedures

in use at the work site, and with the appropriate focus on

safety.

Specific activities and noteworthy observations

are detailed in the

inspection report.

The inspectors reviewed procedures and,observed

all or portions

of the following maintenance/surveillance

activities:

N1-ESP-RPS-331

N1-IPM-036-010

N1-IPM-036-040

N1-ISP-036-1 03

N1-ISP-036-1 04,

Reactor Protection System Motor-Generator Set

Instrument Channel Test Excluding Output Contractors

Anticipated Transient Without Scram /Alternate Rod

Injection Instrument Calibration

Yarway Reactor Level Local Indicator Calibration

Hi/Lo Rx Water Instrument Trip Channel Calibration

Low-Low Rx Water Level Instrument Trip Channel

Calibration.

P

Surveillance activities are included under "Maintenance.

For example,

a section involving surveillance observations might

be included as a separate sub-topic under M1, "Conduct of Maintenance."

f

e

~

N1-ST-V3

N2-ESP-BYS-R681

N2-FHP-1 3.1

N2-MSP-EGS-R02

N2-OSP-EGS-R04

N2-PM-S001

WO 97-12419-00

WO 97-16778-01

WO 97-16778-03

WO 97-16778-12

'WO 98-1192-00

WO 98-21 59/60

WO 98-2243-02

WO 98-2211-00

WO 98-2211-02

WO 98-2211-04

WO 98-2327-04

Rod Worth Minimizer Operability Test APRM/IRM

Overlap Verification

Div I/II/IIIBattery Service Test

Complete Core Off-load

Diesel Generator Inspection Division 3 (EGS "EG2)

Operating Cycling Diesel Generator Simulated Loss of

Offsite Power with ECCS Division I 5, II

Refueling Platform and Grapple Inspection

N2-MSP-EGS-R002- Diesel Generator Inspection - Div.

III (18 month, 6 5 12 year Requirements)

Pre-staging Activities to Support ECCS Suction Strainer

Modification N2-97-067

Remove Existing ECCS Strainer 2CLS "STR1 and

Replace with New Strainer

Transport of New ECCS Suction Strainers into the

Suppression

Pool and Removal of Old ECCS Suction

Strain ers

Work Order for N1-ESP-RPS-331

Input New Set Points for Control Room Ventilation

Radiation Monitor ¹11/12

Relocate Radiation Detector per DDC 1M00571

Fabricate

5, Install Plate for Controls per DDC 1F00460

Make Wire Changes to Provide Manual Control to TCV-

210.1-56 per 1F0460 Latest Revision

Support Ears to Mount Valve Position Indicator, Plate

Broken Off

Perform MFT for CRAG Mod.(N1-MFT-053)

.

M8

Miscellaneous Maintenance Issues

M8.1

Closed

URI 50-220 96-01-03: Inade

uate Testin

of Unit 1 Control Room

=Annunciators

a.

Ins ection Sco

e 92902

The inspectors reviewed NMPC's response to a lack of preventive maintenance

and

calibration of annunciator inputs at Unit 1. The inspectors discussed the disposition

of the associated

DER with Unit 1 management.

b,

Observations

Findin s and Conclusion

.

The inspectors reviewed the disposition of DER 1-96-0148regarding

the

January 20, 1996 trip of the Unit 1 ¹11 reactor feed pump due to a relay failure.

The DER addressed

the failure of the Agastat relay and associated

corrective

actions.

However, the unresolved item was initiated due to the inspectors'oncern

that the associated

control room annunciator did not alarm, as expected,

and to

assess

the adequacy of periodic control room annunciator testing.

t

i

0

Discussions with the Unit 1 Technical Support Manager and the General Supervisor

of Operations revealed that all annunciator inputs are either tested or determined to

be not necessary for safe operation.

After a licensee review of this concern, the

reactor feed pump trip annunciator and a few additional annunciators were found to

have been improperly categorized

as not necessary for safe operation.

The

inspectors determined that this oversight of annunciator testing constituted

a

violation of minor safety significance and is not subject to formal enforcement

action.

M8.2

Closed

URI 50-410 96-10-01: Post-Maintenance

Testin

of the Unit 2 Main

Steam Line Radiation Monitors

92902

During a review of the post-maintenance

testing (PMT) of the Unit 2 main steam

line (MSL) radiation monitor, the inspectors noted that the TS required trip signals

were cleared to support the PMT. The inspectors questioned the removal of the trip

signals while the channel was still inoperable and the LCO action statement was still

effective.

The licensee stated that the trip signal needed to be cleared to perform

the PMT surveillance before the channel could be declared operable.

Notwithstanding good justification, the removal of the LCO required trips prior to

declaring the MSL radiation monitor operable appeared to be in conflict with Unit 2

TS 3.3.1.

NMPC challenged this interpretation of TS 3.3.1 and submitted a letter.to

the NRC, dated October 21, 1996, requesting

a clarification of "... a longstanding

[industry] practice that permitted the conduct of TS surveillance testing needed to

demonstrate that previously inoperable equipment had been restored to an operable

condition."

On November 21, 1996, the NRC responded with the following: "... It is not the

intent of TS 3.0.2 to preclude the return to service of a component that has been

replaced or repaired when it can reasonably

be considered operable except for the

'ompletion of.surveillance testing to confirm its operability.

The NRC staff has

addressed

this existing ambiguity in TS 3.0.2 by adding TS 3.0.5 to the Standard

Technical Specifications for BWR/4, Revision 1. TS 3.0.5 states "... equipment

removed fiom service or declared inoperable to comply with ACTIONS may be

returned to service under administrative control solely to perform testing required to

demonstrate

its OPERABILITYor the OPERABILITYof other equipment

. ~."

Based upon the above, NMPC's post-maintenance

testing practices are acceptable

and there is no violation of regulatory requirements.

This unresolved item is closed.

M8.3

Closed

VIO 50-410 96-10-03: Procedure

Chan

es Not in Accordance with TS

Re uirements

92902

In August 1996, during repairs to the Unit 2 control building chillers, a procedure

change evaluation (PCE) was processed to change the service water low flow set

point. However, the PCE was processed

as an editorial change and did not receive

the approval of a senior reactor operator, as required by the Unit 2 TS 6.8.3. This

was documented

on DER 2-96-3284.

P

0

10

The inspectors reviewed the DER disposition and NMPC's response to the Notice of

Violation. Corrective actions included revising the common site procedure which

controls PCEs (NIP-PRO-04) to clarify and limitwhat may be considered

an editorial

change to a procedure.

The inspectors have identified no further examples of

improperly processed

PCEs.

This violation is closed.

M8.4

Closed

VIO 50-410 97-02-02: Missed Unit 2 HPCS Actuation Instrumentation TS

Surveillance Test 92702

In March 1997, NMPC identified that their test procedures failed to satisfy the TS

surveillance requirements

(TSSR) for response time testing of the HPCS actuation

instrumentation.

NMPC's letter, dated June 16, 1997, provided the root cause and

corrective actions for this violation; much of the same information was also

contained in LER 50-410/97-01.

The inspectors'eview of the immediate corrective

actions was detailed in IR 50-410/97-02 and the inspectors'eview of the

associated

LER was documented

in IR 50-410/97-04.

The inspector determined

that the licensee reviewed other ECCS response time tests with no additional

deficiencies identified. With respect to actions to prevent recurrence,

NMPC

credited improvements to their procedure review process

as barriers to prevent

recurrence.

The inspectors re'viewed the enhanced

administrative control

procedures

and identified no concerns.

This violation is closed.

M8.5

Closed

LER 50-410 98-09: Missed Batter

TSSR Due to Ina

ro riate

aO

Ins ection Sco

e 61726 92700

The inspectors reviewed the LER, DER, and applicable TSs, surveillance test

procedures

and test results, and discussed this issue with responsible individuals.

In addition, discussions were held with NRC management

and technical staff

members from the Region

I Office and the Office of Nuclear Reactor Regulations

(NRR) with regard to enforcement discretion.

Also, the inspectors verified the

completion of the LER in accordance with 10CFR50.73.

b.

Observations

and Findin s

On April 17, 1998, NMPC determined that TSSR 4.8.2.1.d for the Division I 125

volt battery had not been met for Unit 2 from April 5, 1995, to the present.

Specifically, during RFO4 and RFO5, credit was inappropriately taken for the battery

performance test in lieu of the battery service test.

Upon identification of the

missed surveillance, the Unit 2 SSS declared the Division I battery inoperable.

Before initiating the TS required plant shutdown, NMPC notified the NRC and

requested

enforcement discretion to delay the testing of the battery until RFO6,

scheduled to begin May 2, 1998.

The NRC verbally granted enforcement discretion

from the TS requirements until the next Unit 2 entry into Cold Shutdown, but not

later than May 3, 1998. This discretion allowed NMPC to avoid an unnecessary

'lant shutdown.

The enforcement

discretion was granted provided that the

Division II and III batteries remained'operable

and the Division I battery cell-to-cell

11

resistance

check were performed weekly.

On April 21, 1998, the written Notice of

Enforcement Discretion (NOED) was docketed.

Notwithstanding, the failure to

complete the required Division I battery service test, since April 1995, is a violation

of TS 4.8.2.1.d.

(VIO 50-410/98-05-02)

On May 2, 1998, Unit 2 shutdown for RFO6 and the NOED was exited upon the

unit achieving the Cold Shutdown condition.

On May 7, NMPC successfully

completed the service test on the Division I battery.

The inspectors reviewed the

test procedure and results, and identified no concerns.

The inspectors verified that the LER was completed in accordance with the

requirements of 10CFR50.73.

This LER is closed.

c;

Conclusion

During this inspection period, the NMPC staff self-identified that the TS required

service test of the Unit 2 Division I battery was not completed during the previous

two refueling outages.

NMPC had,improperly credited the battery cyclic

performance test for satisfying the requirements of the service test.

NMPC

requested

and was granted a Notice of Enforcement Discretion (NOED) to avoid the

consequential

TS required shutdown.

The NOED was exited on May 2, 1998 upon

the unit achieving Cold Shutdown conditions and the service test was completed

satisfactorily on May 7, 1998.

Notwithstanding, the failure to have properly service

tested the Division I battery, since April 1995, is a violation of TS 4.8.2.1.d.

(VIO 50-41 0/98-05-02)

III. ENGINEERING

E1

Conduct of Engineering

E1.1

General Comments

37551

The resident inspectors frequently reviewed design and system engineering

activities, including justifications for operability determinations,

and the support by

the engineering organizations to plant activities.

E1.2

Unit 2 Emer enc

Diesel Generator Fuel Line Leak

a.

Ins ection Sco

e 37551

During a Unit 2 surveillance test of the Division II emergency diesel generator (EDG),

a fuel leak developed in the pipe between the fuel filters and the fuel injectors.

The

inspectors assessed

NMPC's actions to address

and evaluate the leak. The

assessment

included a visual inspection of the damaged fuel line and the

susceptible location on the other Unit 2 EDGs, and a review of associated

DERs,

SSS's logs, plant modifications, UFSAR and TS sections,

and immediate and long-

k

12

term corrective actions.

The inspectors discussed

issues related to the event with

the Unit 2 Plant Manager and members of the Unit 2 system engineering staff.

Observations

and Findin s

On April 14, 1998, during a surveillance test of the Division II EDG, a fuel leak

dev'eloped in the fuel line pipe between the fuel filters and the fuel injectors.

The

operators immediately stopped and shutdown the EDG. The SSS declared the EDG

inoperable, the appropriate TS LCO was entered; and a DER was initiated to record

the event.

Upon investigation, NMPC concluded that vibration of the fuel line pipe

caused fretting of the piping at the location of a pipe support.

NMPC generated

a

work order to replace the pipe and, after repairs were completed, declared the EDG

operable on April 15.

As required by TS, NMPC evaluated the other EDGs for susceptibility to the same

failure mechanism.

The Division III EDG is of a different design and is not

susceptible.

However, NMPC identified notable degradation of the fuel line at the

same location on the Division I EDG. After close examination, NMPC concluded

that the degradation was not significant enough to impact operability.

The

inspectors discussed the basis for the operability determination with the SSS and

the system engineer, and considered it adequate.

I

On April 18 NMPC replaced the Division I EDG degraded fuel line pipe.

Besides

replacing the pipe, NMPC evaluated and incorporated

a design change to install a

rubber grommet between the pipe and the support to prevent recurrence.

Subsequently,

a similar design change was completed for the Division II EDG pipe

support.

Following the event, NMPC evaluated the consequence

of the Division II EDG fuel

line leak and concluded that based on the size of the leak and the available fuel

within the storage tank, the EDG was capable of operating at rated load for seven

days, as designed.

The resident inspectors and region-based

specialists reviewed

the evaluation and determined that the licensee's conclusion was acceptable.

NMPC reviewed the history of these fuel lines and revealed that a 1993 design

change added the pipe supports to correct previous fuel line leaks.

Further

investigation revealed that the design package included a rubber grommet at the

pipe support, to compensate for system vibration. However, the requirement to

install the protective grommet was not adequately incorporated into the final design

package.

The failure to translate

a specific design change to correct an identified

design deficiency into the design package

is contrary to 10CFR50, Appendix B,

Criterion III, "Design Control," and is a violation.

(VIO 50-410/98-05-03)

Conclusion

'

During surveillance testing of the Unit 2 Division II EDG, a fuel leak developed

between the fuel filter and the fuel injectors.

NMPC determined'that the leak was

caused by vibration of the fuel supply piping, which caused fretting of the pipe at a

r

13

pipe support.

Subsequent

licensee investigation identified notable, but less severe,

fretting on the Division I EDG fuel supply piping. The fuel line supports were

installed in 1993, but the specific design change to install a protective grommet

was not adequately incorporated into the final design package.

This is a violation of

10CFR50, Appendix B, Criterion III, "Design Control." (VIO 50-410/98-05-03)

E3

Engineering Procedures

and Documentation

E3.1

Unit 2 ECCS Suction Strainer Modification

a.

Ins ection Sco

e 37551

In response to NRC Bulletin 96-03, NMPC installed new ECCS suction strainers in

the Unit 2 suppression

pool. The inspectors reviewed the engineering design

documents, the associated

work orders, and observed the installation of the new

strainers.

b.

Observations

Findin s

and Conclusion

In May 1996, the NRC issued Bulletin 96-03 ("Potential Plugging of Emergency Core

Cooling Suction Strainers by Debris in Boiling Water Reactors" ) which addressed

concerns that the strainers would become plugged by debris during a loss of coolant

accident.

The Bulletin proposed several options, NMPC chose the installation of

large capacity passive strainers at Unit 2; in that the existing strainers could not

accommodate the projected debris loading.

NMPC designed and installed new

suction strainers for the residual heat removal, low pressure

core spray, and high

pressure core spray systems to satisfy the system pump net positive suction head

requirements.

The new strainers are of a stacked-disk design and constructed of

type 304 stainless steel.

All work was accomplished

in accordance with the

respective TSs for the associated

safety systems.

In addition to the work orders (listed in Section M1.1) for the removal and

installation of the strainers, the inspectors reviewed the below listed documents

related to the strainer modification. The inspectors identified no concerns or

unreviewed safety questions,

and the proposed changes to the UFSAR appeared

appropriate.

NMP2-41 5M

DDC 2M11330

DCN N2-97-067

DDC 2S11055A

DDC 2S11067

Engineering Specification for Bidding Purposes for

Replacement of ECCS Suction Strainers

Design Document Change to NMP2-415M after Award

Design Change Notification for the'Modifications

Required to Provide Access for the ECCS Strainer

Replacement - Structural, Mechanical, Electrical, and

ALARA

Removal of South Suppression

Pool Hatch Wall &

Mezzanine for ECCS Strainer Replacement

Field Tolerances for Installation of ECCS Strainers

t

,14

DDC 2M11294

SE 98-033

LDCR 2-97-150

Technical Justification and Installation of ECCS

Strainers

Safety Evaluation for ECCS Suction Strainer

Replacement

Licensing Document Change Request for UFSAR

Changes due to ECCS Suction Strainer Replacement

The design and installation of the new ECCS pump suction strainers appeared

adequate to ensure sufficient net positive suction head for the pumps in the event

of a loss of coolant accident (LOCA).

E8

.

Miscellaneous Engineering Issues

E8.1

Closed

VIO 50-220 5 50-410 96-01-05: Failure to Com lete Safet

Evaluation

Prior to Installation of Tem orar

Modification 92903

E8.2

On January 31, 1996, the inspectors identified the installation of an emergency

temporary modification on the Unit 2 circulating water system prior to the

completion of the required 10CFR50.59 safety evaluation.

Furthermore, NMPC

Procedure GAP-DES-03, "Control of Temporary Modifications," Revision 4,

permitted the installation of emergency temporary modifications.

NMPC letter,

dated May 22, 1996, documented the root. cause and corrective actions for this

violation. The inspectors'eview of the immediate corrective actions was

documented

IR 50-220 &. 50-410/96-01. With respect to actions to prevent

recurrence,

NMPC removed the provision for emergency temporary modification

from the temporary modifications procedure.

Based on the inspectors'eview of

the licensee's corrective and preventive actions, and observation that there have

been no subsequent

installations of a temporary modification prior to the completion

of the associated

safety evaluation, this violation is closed.

Closed

URI 50-220 5 50-410 96-14-02: Potential Over- ressurization Concerns

Relative to NRC Generic Letter 96-06 92903

During the evaluation of NRC Generic Letter (GL) 96-06, "Assurance of Equipment

Operability and Containment Integrity during Design-Basis Accident Conditions,"

NMPC identified some system piping penetrating the drywell that could potentially

be over-pressurized

during a design basis LOCA. Subsequent

NMPC examination of

each penetration,

in accordance with the guidance provided in GL 96-06, concluded

that the systems remained operable.

An unresolved item was assigned to track the

licensee's resolution of this generic issue and to assess

whether this condition was

potentially outside the design bases of the plants.

Subsequent

discussions between the NRC Region

I Office and NRR concluded that

thermal over-pressurization

is not necessarily

a condition outside the design bases

of the plant.

NRC-staff follow-up of each licensee's actions to address

GL 96-06

issues will be assessed

via a future inspection activity. This unresolved item is

closed.

I

15

j

E8.3

Closed

URI 50-220 97-12-08: Im act of Dr well to Wetwell B

ass on

Containment Pressure

90712 92903

On October 12, 1997, GE issued

a 10CFR21 (Part 21) notification regarding the

'ossible reduction in the pressure

suppression capability of the torus due to bypass

leakage between the drywell and the torus.

During review of the Part 21

notification for Unit 1, NMPC determined that, although the specific issue described

in the Part 21 was not a concern at Unit 1, other conditions may challenge the

pressure suppression

capability of the torus.

Particularly, during drywell and torus

inerting, deinerting, and primary containment pressure maintenance

evolutions, the

drywell and torus vent valves were usually open concurrently, establishing

a

drywell-to-torus bypass pathway.

Upon identification of this vulnerability, NMPC

issued

a procedure change to prohibit concurrent opening of both the'rywell and

torus vent and purge valves during primary containment venting, filling, and make-

up evolutions.

NMPC issued

LER 50-220/97-15to document this condition.

However, at that

time, NMPC had yet to complete their analysis to determine whether the

containment design pressure would be exceeded

should a LOCA have occurred

while the drywell-to-torus valves were open.

The inspectors'eview of NMPC's

evaluation of the Part 21, and the LER were provided in NRC IR 50-220/97-12.

As

documented

in the IR, the issue was unresolved pending the completion of NMPC's

analysis.

Subsequently,

NMPC completed their analysis and determined that the containment

post-accident pressure, with the bypass pathway, would not have exceeded

maximum design pressure.

Based on this analysis, NMPC retracted

LER 50-220/97-

15 via letter, dated May 1, 1998.

This unresolved item is closed.

E8.4

Closed

LER 50-220 98-05: Unreco

nized Violation of TS Secondar

Containment

a 0

Ins ection Sco

e

92700

The inspectors reviewed NMPC's analysis and corrective actions associated with

the discovery of a breach of Unit 1 secondary containment integrity due to normally

open vents on the containment spray raw water heat exchanger.

b.

Observations

and Findin s

In April 1998, with Unit 1 at full power, NMPC discovered

a breach of secondary

containment.

Specifically, the containment spray raw water (CSRW) heat

exchanger vents were normally open, in accordance with the operating procedure

(N1-OP-14, "Containment Spray" ). The open vents provided a potential

unmonitored release path from the secondary containment (reactor building)

atmosphere,

via the reactor building drain system, through the open CSRW vents,

to the service water system, and to the environment via the service water discharge

to Lake Ontario.

Unit 1 TS, Section 3.4.0, requires reactor building integrity be

maintained during power operation; since the definition of reactor building integrity

16

was not satisfied,

a reactor'hutdown was initiated in accordance with TS 3.0.1.

The procedure was revised, the vent valves were closed, and the shutdown was

terminated at 94% power.

DER 1-98-0903 was written, and the appropriate NRC

notifications were made as required by 10CFR50.72.

The inspectors determined that the original design for the containment spray system

was to operate with the vents open to maximize heat exchanger performance.

In

1986, the vents were closed following the replacement of the heat exchangers.

In

1991, DER 1-91-Q-1417 was written to address concerns associated with the

vents being closed and the effect of non-condensible

gas build-up in the heat

exchangers.

The DER disposition stated that testing 'of the containment spray heat

exchangers

showed that the system would perform its design basis function with

the vents closed.

In addition, the DER noted that the Tube Exchanger

Manufacturers Association (TEMA) recommends that the heat exchanger

be

operated with the vents open.

Therefore, in 1992, a plant change request was

processed to operate with the vents valves open, which would maximize system

performance.

In response to the containment concern and in support of the

decision to close the vent valves, NMPC performed an Operability Determination

and a 10CFR50.59 Applicability Review.

The inspectors discussed this issue with the responsible staff. members and unit-

management,

reviewed the associated

documentation,

and considered the actions

taken and decisions made by'NMPC to have been appropriate.

The LER identified

the cause of the event as inadequate

design analysis, in that, the personnel involved

did not consider the interaction between the open vents and secondary containment

integrity. The analysis of the event revealed that surveillance tests conducted in

1996 and 1997 showed that the reactor building emergency ventilation system

(RBEVS) was able to maintain a negative pressure relative to the environment.

NMPC concluded that operation with the vent valves open would not affect the

ability of the RBEVS to maintain negative pressure

in the event of a design basis

accident.

Notwithstanding', plant operation with secondary containment integrity not properly

established

is a violation of the Unit 1 TS, Section 3.4.0.

However, this non-

repetitive, licensee identified and corrected violation is being treated as a Non-Cited

Violation, consistent with Section VII.B.1 of the NRC Enforcement Policy.

(NCV 50-220/98-05-04)

The inspectors verified that the LER was completed in accordance with the

requirements of 10CFR50.73.

This LER is closed.

Conclusion

During a review of Unit 1 operating procedures,

NMPC identified that the normally

open vent valves on the containment spray raw water heat exchangers

violated

secondary containment integrity, in that it provided a potential release path from the

reactor building to the environment.

This licensee identified and corrected violation

of secondary containment integrity requirements was not cited.

ass

17

Closed

LER 50-220 98-06: Desi

n Deficienc

Associated with CREVS Radiation

Monitors

Ins ection Sco

e 37551

90712

NMPC identified that the radiation monitors for the Unit 1 control room emergency

ventilation system (CREVS) would not have automatically initiated the system in the

event of a main steam line break (MSLB). The unit was shutdown because

repairs

were not able to be completed within the time allowed by the TSs.

The inspectors discussed the issue with various engineering personnel,

and Unit 1

management.

In addition, they monitored portions of the reactor shutdown and the

system modifications.

The inspectors reviewed the DERs, the associated

Safety

Evaluation, implementation of the TS amendment,

and the modification functional

test; the inspectors also performed an in-office review of the LER.

b.

Observations

and Findin s

On April 21, 1998, NMPC identified that the Unit 1 CREVS would not have

automatically initiated, as designed, following a MSLB. The SSS appropriately

entered TS LCO 3.4.5.e, which allowed 7 days to restore the system to an operable

status or required the reactor be shutdown.

NMPC identified that the trip settings

for the radiation monitors installed on the intake of the CREVS were set too high.

Unable to adjust the radiation monitors to properly initiate CREVS for a loss of

coolant accident (LOCA), NMPC initiated a plant shutdown on April 28, which was

completed on April 29.

The inspectors determined that the radiation set point was less than or equal to 800

counts per minute (~800 cpm); this was below the Unit 1 TS required set point of

1000 cpm.

NMPC determined that the set point for MSLB was 6210 cpm, but the

existing monitors'could not be set low enough to detect LOCA conditions.

To

'ensure that the CREVS would initiate for both a MSLB and a LOCA, NMPC proposed

a TS amendment to change the CREVS automatic initiation signal from high

radiation to signals from the reactor protection system for MSLB (main steam line

high flow or main steam line tunnel high temperature)

and LOCA (high drywell

pressure

or low-low reactor pressure vessel water level).

In addition, the set point

for the radiation monitors was adjusted to ~193 cpm.

The NRC approved this TS

amendment on May 23, 1998.

NMPC determined the cause of the event to be an inadequate

engineering

evaluation in 1984. A contributing factor was inadequate

design control.

Corrective actions included incorporation of the lessons learned from this event into

the engineering department continuing training program, and a review of other

radiation monitor set point calculations.

The inspectors discussed the modifications

with the CREVS system engineer and control room personnel,

and reviewed the

completed DERs, safety evaluation and the post-modification test, and had no

concerns.

This event was of low safety significance, in that, the emergency

4

18

procedures

require the operators to verify the CREVS is in operation, or to start the

system manually, in the event of a MSLB or LOCA. Notwithstanding, the failure

to properly evaluate the initiation logic for the CREVS is a violation of 10CFR50,

Appendix B, Criterion XI, "Test Control." This non-repetitive, licensee identified

and corrected violation is being treated as a Non-Cited Violation, consistent with

Section VII.B.1 of the NRC Enforcement Policy.

(NCV 50-220/98-05-05)

The inspectors verified that the LER was completed in accordance with the

requirements of 10CFR50.73.

This LER is closed.

Conclusions

During a review of the control room emergency ventilation system initiation logic,

NMPC determined that the system would not automatically initiate, as required.

Specifically, the system would not automatically start as a result of a main steam

line break or a loss of coolant accident.

This licensee identified and corrected

violation of 10CFR50, Appendix B, Criterion XI, "Test Control," was not cited.

Closed

LER 50-410 98-07: TS 3.0.3 Entr

Due to Missed Lo ic S stem

Functional Testin

of Loss of Volta e and De raded Volta e Channels

90712

The technical issues associated with this LER were described in Section E3.1 of IR

50-410/98-02.

The inspectors performed an in-office review and verified that the

LER was completed in accordance with 10CFR50.73.

This LER is closed.

Closed

Part 21 50-220 98-01: Defective GE SBM-T

e Switches at Unit 1

Ins ection Sco

e 36100

In January 1998, NMPC initiated a DER as a result of a GE Nuclear Energy (GENE)

.

Part 21 notification of an adverse condition related to the spring-return function of

some GE provided control switches that could damage the associated

control

circuits.

In March 1998, GENE issued

a revision to the notification explaining the

failure mode and root cause.

The inspectors reviewed the revised Part 21 and the

DER disposition.

Observations

Findin s and Conclusion

On March 19, 1998, GENE revised the Part 21 notification regarding

a failure of

some GE SBM-type control switches having a spring-return feature.

GE

manufactures these switches as commercial grade.

The switches were purchased

as commercial grade and then dedicated for safety related applications by NMPC.

During the review of the original Part 21 notification, NMPC identified seven

potentially affected switches in safety-related functions for Unit 1 and NMPC

engineering initiated DER 1-98-0202.

The inspectors reviewed the revised Part 21 notification, the DER disposition, and

discussed the issue with the responsible maintenance

engineer.

GENE determined

19

the failure mechanism to be binding between the rear bearing and the casing

support caused

by shrinking of the casing support due to "post-mold cure."

The

mold for the casings had worn such that the bearing support hole was at the

minimum allowable value, and post-mold cure caused the hole to shrink. When

operated manually, the switch contacts operated properly.

GENE also determined

that switches in service for over two-years were not subject to the failure mode.

The licensee identified that all of the questionable switches at Unit 1 were greater

than two-years old, except one.

The inspector determined that the remote manual

control switch for the emergency cooling condenser vent to torus blocking valve

(the one exception) will be monitored until the two-year period is over.

The

inspector noted that it is NMPC's standard operating practice to verify that control

switches spring return to the normal position and that this verification may be

subject to peer (dual) verification.

The inspectors observed that NMPC's follow-up of the Part 21 report concerning GE

SBM-type control switches and their identification of the susceptible switches at

Unit 1 was thorough and an example of an improving questioning attitude by the

engineering staff. This Part 21 report is closed.

IV. PLANT SUPPORT

Using NRC IP 71750, the resident inspectors routinely monitored the performance

of activities related to the areas of radiological controls, chemistry, emergency

preparedness,

security, and fire protection.

Minor deficiencies were discussed with

the responsible management,

and significant observations

are detailed below.

R1

Radiological Protection and Chemistry (RP8cC) Controls

R1.1

Radiolo ical Environmental Monitorin

Pro ram Im lementation

a.

Ins ection Sco

e 84750

The following areas of the Radiological Environmental Monitoring Program (REMP)

were assessed

and reviewed:

selected sampling and analysis procedures,

analytical data from 1998,

selected sampling techniques,

operability and calibration of air samplers,

1996 and 1997 Land Use Census results,

1996 and 1997 Annual Radiological Environmental Operating Reports, and

licensee's investigation after identifying iodine-131 (I>>,) in milk in April 1997

C

1

20

Observations

and Findin s

The sampling and analysis procedures

provided appropriate guidance to perform

REMP tasks.

Sampling techniques were appropriate to collect environmental sample

media.

The air sampling equipment and water compositors were operable during

1997 to present,

as evidenced

in the sample logs and sample analysis results.

The

air sampling equipment calibration results were within the established tolerances,

and calibrations were performed within the frequency specified in the procedure.

A

Land Use Census was performed 1996 and 1997 during the growing season,

as

required by the TS.

The 1996 and 1997 Annual Radiological Environmental Operating Reports included

results of the environmental monitoring program, program changes,

land use

census,

and inter-laboratory comparison program, as required by TS. The reports

provided a comprehensive

summary of the results of the REMP around the site and

met the TS reporting requirements.

I)3$ was detected

in a routine indicator milk sample during the week of April 2 1

1997 at a concentration of 0.5 pCi/L. The licensee immediately conducted

an

investigation and discussed this issue with the NRC in April 1997. The primary

analytical contract laboratory immediately investigated the analysis results by re-

analyzing the sample and confirming the results with another laboratory.

The

investigation was detailed and exhaustive.

The licensee concluded that:

(1) the

source of the iodine could not be determined;

(2) that it was unlikely the source of

the I>>, was from either Nine Mile Point or from the J. A. FitzPatrick plant; and, (3)

the dose was insignificant compared to the doses received from natural sources.

The details and conclusions. of the investigation as a result of the I>>, was

documented

in the 1997 Annual Radiological Environmental Operating Report, as

.required by TS.

Conclusions

The licensee effectively maintained and implemented the Radiological Environmental

Monitoring Program in accordance with regulatory requirements.

The licensee

performed a comprehensive

review of an anomalous indication of I>>, in an

environmental milk sample.

Meteorolo ical Monitorin

Pro ram lm lementation

Ins ection Sco

e 84750

The inspectors reviewed the implementation of the meteorological monitoring

program (MMP); specifically, the status of the meteorological instrumentation

including, system operability, and the associated

channel calibration and channel

functional test procedures

and results were reviewed for the period of July 1996 to

May 1998.

I

0

21

Observations

and Findin s

Channel calibrations, channel checks, and channel functional tests were performed

within the frequency recommended

in Table 4.3.7.3-1 of TS 3/4.3.7.3 and

Regulatory Guide 1.23, Revision 1. The wind speed, wind direction, and

temperature

sensors

on the towers were operable; and applicable data was

available.

The associated

procedures

provided appropriate guidance to perform

channel functional tests and channel calibrations for all the channels, except for the

wind speed channels.

The required meteorological monitoring instrumentation channels (wind speed, wind

direction, and delta temperature)

shall be demonstrated

operable by the semi-annual

performance of the channel check and channel calibration operations,

as required by

TS 3/4.3.7.3.

Prior to May 22, 1998, NMPC had not performed a channel

calibration of the wind speed channel.

According to Unit 2 TS, Section 1.4, a

channel'calibration "... shall be the adjustment,

as necessary,

of the channel output

so that it responds with the necessary

range and accuracy to known values of the

parameter which the channel monitors.

The channel calibration shall encompass

the

entire channel including the sensor andalarm and/or trip functions, and shall include

the channel functional test."

Relative to wind speed, the licensee's calibration did

not include the wind speed sensor,

as required by the TS. Therefore, the accuracy

of the wind speed channel was not measured during channel calibrations.

Failure to

perform the channel calibration of the wind'speed channel, in accordance with the

Unit 2 TS, Section

1 4, constitutes

a violation of Unit 2 TS 3/4.3.7.3.

(VIO 50-410/98-05-06)

Conclusion

Overall, the licensee effectively maintained meteorological monitoring system

operability, and satisfactorily performed channel calibrations and channel functional

tests for the meteorological instrumentation, with the exception of the wind speed

channel.

The failure to'erform the channel calibration of the wind speed channel

according to the channel calibration definition of TS 1.4, in that, the accuracy of the

entire wind speed channel was not measured from the. sensor to the channel output,

constitutes

a violation of Unit 2 TS 3/4.3.7.3.

(VIO 50-410/98-05-06)

Unit 1 and Unit 2 Tours

Ins ection Sco

e 83750

A review was performed of housekeeping,

radiological boundaries,

and access

controls.

Information was gathered through tours of Unit 1 and Unit 2 reactor

buildings and drywells, and through discussions with cognizant personnel.

Observations

and Findin s

Housekeeping was adequate

in that walkways and aisles were clear and free of

debris, and major plant work areas were generally well illuminated. Some examples

22

of poor lighting were observed

in the Unit 2 drywell; reportedly due to a trip of a

temporary power breaker switch.

This was corrected the next day by re-

distributing several temporary lighting strings to another temporary power box.

High radiation areas and contaminated

areas were well delineated and clearly

posted.

A selective examination of the access to areas with dose rates greater than

1000 mrem per hour at 30 centimeters revealed appropriate controls, such as

locked doors or flashing lights. Access to radiologically controlled areas. was well

controlled with radiation work permits, health physics briefings, use of electronic

dosimetry, and radiological postings.

C.

Conclusion

Housekeeping

was adequate

in that aisles and walkways were clear and free of

debris, radiological boundaries and postings were clear, and access controls to

radiologically controlled areas were effective.

R1s4

Unit 2 Refuelin

Outa

e

a

~

Ins ection Sco

e 83750

A review was performed of radiological controls implemented for the Unit 2

refueling outage work. Specific areas evaluated included the refuel floor, drywell,

and suppression

pool. Information was gathered through tours of the facility,

interviews with cognizant personnel, attendance

at several drywell work scheduling

meetings, and selected examinations of reviews to maintain radiation exposures

as-

low-as-is-reasonably-achievable

(ALARA).

b.

Observations

and Findin s

Refuel'Floor:

ALARAReview No. 98-2-23, "Refuel Floor Activities," was used as

the major radiological control plan for work on the refuel floor. It included basic

ALARArequirements for refuel floor activities, and specific requirements for reactor

pressure vessel disassembly/reassembly,

underwater activities, and cavity/storage

pit decontamination.

The review was based on lessons learned during previous

outages and pertinent industry events.

The review included specific requirements

for work coordination, pre-job briefings, dose minimization, and contamination

controls.

The chief radiation protection (RP) technician maintained close oversight

of personnel access,

was thoroughly knowledgeable of ongoing work and

radiological controls, communicated well with plant work groups, and ensured that

personnel were instructed on radiological conditions and requirements prior to work.

~Dr well Major radiological controls for the drywall included close health physics

oversight of drywell access,

radiation work permits and ALARAreviews, extensive

use of temporary shielding, flushing of drain lines and reactor vessel nozzles, and

specific work planning and control. A chemical decontamination of the recirculation

system had been planned to minimize dose for recirculation valve work, but was

canceled when the system could not be fully isolated during the scheduled work

window. To compensate for this cancellation, several reactive drywell planning

23

meetings were conducted to establish

a revised work plan that accomplished

required work while minimizing radiation exposure.

Major station work groups

attended the meeting including outage planning, operations, maintenance,

engineering, and radiation protection.

Su

ression Pool:

Major work performed in the suppression

pool involved a

modification to replace emergency core cooling system suction strainers.

The

suction strainers were located below suppression

pool water level, requiring work to

be performed by divers.

Extremity dosimetry was issued to the divers in

accordance with procedural guidance,

and licensee actions including close oversight

by health physics personnel and use of administrative limits ensured that extremity

exposures were within regulatory and administrative limits. However, several

examples were identified in which the available administrative extremity exposure

limits were incorrectly determined and documented

on Procedure S-RPIP-5.4, "Dose

.Tracking and Timekeeping," Attachment 1: "Dose Tracking and Timekeeping

Worksheet."

Procedural guidance did not specify the exact method for determining

"available exposure," and several examples were identified in which the

administrative available exposure for the "extremity" was calculated by subtracting

the accrued whole body dose (rather than the accrued extremity dose) from the

administrative extremity dose limit.

The radiation protection manager acknowledged that the observed method for

determining the available administrative exposure for the extremity was incorrect,

and stated that instructions for determining the available administrative exposure for

the extremity would be clarified by a revision to procedure S-RPIP-5.4, "Dose

Tracking and,Timekeeping."

This failure constitutes

a violation of minor safety

significance and is not subject to formal enforcement action.

C.

Conclusions

Radiological controls for outage work were well planned and health physics

personnel maintained close oversight of work.

Procedure

S-RPIP-5 4, "Dose Tracking and Timekeeping," lacked clarity with regard

to the method for determining the available administrative extremity exposure,

and

several examples of inaccurate determinations of available administrative extremity

exposure were identified. The radiation protection manager stated that instructions

for determining the available administrative extremity exposure would be clarified by

a procedure revision,

R1.5

ALARAGoals and Initiatives

a 0

Ins ection Sco

e 83750

A review was performed of the use of goals to maintain radiation exposures

ALARA,and of ALARAinitiatives implemented for the Unit 2 refueling outage.

Information was gathered through reviews of ALARAgoals, tours of the plant,

J

P

24

discussions with cognizant personnel,

and a review of Safety Evaluation

No.98-040, "Chemical Decon of RCS."

b.

Observations

and Findin s

Outage exposure estimates were detailed, appeared

reasonable,

and were frequently

used to evaluate performance with regard to radiation exposure.

Exposure

estimates were established for work groups, major jobs, and the entire Unit 2

refueling outage.

To ensure the usefulness of the exposure goals, prompt changes

were made to reflect major changes

in work scope and cancellation of an attempted

chemical decontamination of the recirculation system.

The following examples of ALARAinitiatives were noted:

A camera monitor was set-up at the entrance to the radiologically controlled

area that allowed for viewing of multiple in-plant job locations including

specific drywell valves, under vessel areas,

and various locations of the

refueling floor.

A pre-outage ALARAr'eport was published that summarized radiological

control outage planning efforts. The document demonstrated

thorough

planning and preparation for outage work.

A chemical decontamination of the reactor recirculation system was planned

and set-up to reduce dose rates for major recirculation valve work. The

planned decontamination involved the low oxidation-state metal ion (LOMI)-

alkaline permanganate

(AP)-LOMI process (LOMI-AP-LOMI). However, the,

chemical decoritamination effort was canceled after water leakage through

the jet pump "ram" heads and through the pump discharge valve could not

be stopped during the scheduled work window.

Plans were in place to hydro wash eleven vessel nozzles including five

reactor

recirculation discharge nozzles, one jet pump instrumentation nozzle,

one HPCS nozzle, three feedwater nozzles, and one LPCI nozzle.

Numerous temporary shielding applications for job specific and general area

'ose

reduction were observed.

Examples included ten recirculation

discharge nozzles, the north scram dump column, and water shields for the

chemical decontamination resin columns.

c.

Conclusions

ALARAgoals were effectively used as a tool to aid radiological planning to minimize

radiation exposure,

Numerous ALARAinitiatives including publication of a pre-

outage report, use of cameras,

use of temporary shielding, planned reactor vessel

nozzle hydro washes,

and an attempt to chemically decontaminate the reactor

recirculation system demonstrated

management support and a commitment to

maintaining radiation exposures ALARA.

A

25

0

R7

Quality Assurance

in RP5C Activities

R7.1

Qualit

Assurance Audit Pro ram

a.

Ins ection Sco

e 84750

The inspector reviewed the following Quality Assurance

(QA) audit reports:

~

97015

~

96022

Environmental Protection, Radioactive Effluents, Radiological

Material Processing,

Transport and Disposal

Radiological Effluents, REMP, Offsite Dose Calculation Manual,

Radioactive Material Processing

b;

Observations

and Findin s

The objectives of the 97015 audit covered specific areas of the REMP and the

objectives of the 96022 audit covered specific areas of the REMP and MMP. Both

audits 'were conducted similarly. Previous DERs were reviewed and followed up for

completeness

and effectiveness of corrective actions.

The auditors reviewed

personnel performance, program implementation, and records.

No significant issues

were identified.

c.

Conclusion

The licensee met the QA audit requirements.

The audits were thorough and of

sufficient depth to assess

the REMP and MMP.

R7.2

Qualit

Assurance of Anal tical Measurements

a.

Ins ection Sco

e 84750

The following aspects of the Quality Assurance/Quality Control (QA/QC) program of

the primary contractor laboratory for the period of July 1996 to May 1998 were

reviewed:

~

the results of the internal QC program, including efficiency and resolution

checks, daily instrument energy checks, control charts of instrument

performance,

and routine calibrations; and

the results of the QA program, including the Inter-laboratory Comparison

(cross-check)

Program.

b.

Observations

and Findin

s

The QA/QC program for analyses of REMP samples

is conducted by the primary

analytical contract laboratory, J. A. FitzPatrick Environmental Laboratory.

The

laboratory implemented intra-laboratory (QC) and inter-laboratory (QA) programs.

The intra-laboratory (QC).program included efficiency and resolution, checks, daily

k

26

instrument energy checks, control charts of instrument performance,

and routine

calibrations.'he results for 1996 and 1997 were compiled and documented

in the

Environmental Laboratory QA/QC Report.

The results from 1996 through 1998

were within the acceptance

criteria. The laboratory continued to participate in an

Inter-laboratory Comparison Program provided by a vendor (Analytics, Inc.). The

laboratory's participation in this program was effective.

In addition to the above required comparison programs, the laboratory participated

in a cross check program with the Environmental Measurements

Laboratory (EML),

Department of Energy.

The analysis results of this program were generally in

agreement, with occasional disagreements

in certain samples.

The laboratory had

conducted

an investigation and determined the cause of the disagreements.

EML

provided sample media and geometries different from the usual sample media and

geometries provided by Analytics, Inc. and the licensee.

The laboratory

accommodated

and incorporated different and difficultmedia and geometries into

the program.

The licensee issued the 1997 Annual Radiological Environmental

Operating Report, as required by TS.

Conclusion

The contractor laboratory continued to implement effective QA/QC programs for the

REMP, and continued to provide effective validation of analytical results.

The

laboratory demonstrated

the ability to accommodate

and incorporate difficultmedia

and geometries into the program.

The programs were capable of ensuring

independent checks on the precision and accuracy of the measurements

of

radioactive material in environmental media.

Deviation Event Re orts and Self Assessments

Ins ection Sco

e 83750

A review was performed to evaluate methods used to identify, evaluate, and resolve

radiological control program deficiencies.

Information was gathered by a selected

review of radiological control issues documented

in DERs and a review of the self-

assessment

procedure.

Observations

and Findin s

The DER system had a high volume, low threshold, and the staff readily used the

system to address program deficiencies.

Ten DERs were selected to evaluate the

effectiveness of the system for resolving problems.

Problem evaluations including

identification of cause and corrective actions taken were reasonable

and

commensurate

with the significance of identified issues.

Self-assessment

Procedure NIP-ECA-05, "Posting and Surveys," was thorough in

that it included a review of radiation work permits, interviews with cognizant

personnel, extensive walkdowns of Unit 1 and Unit 2, and a compliance review with

respect to procedures

and the updated safety analysis report.

Assessment

team

L

27

members were well qualified and included four specialists, three supervisors, two

chief technicians,

and one peer evaluator from Diablo Canyon.

Numerous strengths

and opportunities for improvement were identified. Significant issues were placed

into the DER system, personnel were assigned to resolve identified issues,

and due

dates for completion were established.

A selected review of issues placed into the

DER system and interviews with responsible personnel indicated that adequate

progress was being made toward resolution of identified issues.

C.

Conclusions

C

The DER system and the self-assessment

program were effective in their use to

identify, evaluate, and resolve radiological program deficiencies.

S2

Status of Security Facilities and Equipment

S2.1

Tour of the Protected'Area

Perimeter

71750

The inspectors toured the Nine Mile Point Nuclear Station protected area perimeter

and found the fence and perimeter detection systems intact.

In addition, since the

tour was completed at approximately 1:30 a.m., the inspectors visually assessed

protected area lighting and found it to be acceptable.

S8

Miscellaneous Security and Safeguards

Issues

Process Software Altered 92904

S8.1

Closed

URI 50-220 5 50-410 96-06-06: Fitness-for-Dut

Random Selection

In May 1996, NMPC discovered that two contractors had intentionally altered the

fitness-for-duty (FFD) computer software code.

Specifically, the alteration excluded

the two individuals from the random selection process for FFD testing.

An

unresolved item was opened, pending completion of NMPC's internal investigation,

and subsequent

NRC review of the results.

S

On April 28, 1998, the NRC issued an enforcement action letter (EA 97-185) to

NMPC, stating that the failure to ensure that individuals were tested in a statistically

random and unpredictable manner was a violation of 10CFR26.24, and constituted

a Severity Level III violation. However, based on the NRC Enforcement Policy,

Section VII.B.6, the NRC decided to exercise discretion and not issue

a Notice of

Violation. (NCV 50-220 5 50-410/98-05-08)

Unresolved item 50-220 5 50-

410/96-06-06 is closed.

S8.2

Administrative Closure of Escalated Enforcement Items 92904

The below escalated

enforcement items (EEls) are being administratively closed,

based on the issuance of the enforcement action letter (EA 98-234), dated May 20,

1998, and the associated

determination:

28

EEI 50-220 & 50-410/98-01-01 was reclassified as a Level IV violation-

VIO 50-220 & 50-410/98-01-01

EEI 50-220 & 50-410/98-01-02was withdrawn

V. MANAGEMENTMEETINGS

X1

Exit Meeting Summary

At periodic intervals, and at the conclusion of the inspection period, meetings were

held with senior station management to discuss the scope and findings of this

inspection.

The exit meetings for specialist inspections were conducted upon

completion of their onsite inspection:

~

Environmental Monitoring

May 23, 1998

~

Outage Radiation Protection

May 23, 1998

The final exit meeting occurred on June 12, 1998.

During this meeting, the

resident inspector findings were presented.

NMPC did not dispute any of the

inspectors findings or conclusions.

Based on the NRC Region

I review of this

report, and discussions with NMPC representatives,

it was determined that this

report does not contain safeguards

or proprietary information.

~a

ATTACHMENT1

PARTIALLIST OF NMPC PERSONS CONTACTED

Nia ara Mohawk Power Cor oration

R. Abbott

D. Barcomb

D. Bosnic

J. Burton

H. Christensen

J. Conway

G. Correll

R. Dean

A. DeGracia

S. Doty

K. Dahlberg

G. Helker

A. Julka

P. Mezzafero

B. Murtha

L. Pisano

N. Rademacher

R. Randall

V. Schuman

R. Smith

C. Terry

C. Merritt

K. Ward

D. Wolniak

Vice President,

Nuclear Engineering

Manager, Unit 2 Radiation Protection

Manager, Unit 2 Operations

Manager, Training

Manager, Security

Vice President,

Nuclear Generation

Manager, Unit

1 Chemistry

Manager, Unit 2 Engineering

Manager, Unit 1 Work Control

Manager, Unit 1 Maintenance

Plant Manager, Unit 2 tActing)

Manager, Unit 2 Work Control

Director, ISEG

Manager, Unit 1 Technical Support

Manager, Unit 1 Operations tActing)

Manager, Unit 2 Maintenance

Manager, Quality Assurance

Manager, Unit I Engineering

Manager, Unit 1 Radiation Protection

Plant Manager, Unit

1

Vice President,

Nuclear Safety Assessment

&. Support

Manager, Unit 2 Chemistry

Manager, Unit 2 Technical Support

Manager, Licensing

New York Power Authorit

B. Gorman

D. Kiepper

A. McKeen

Environmental Supervisor, J. A. FitzPatrick Environmental Laboratory

I&C Manager

Radiological and Environmental Services Manager

Attachment

1 (cont.)

INSPECTION PROCEDURES USED

IP 36100

IP 37551

IP 60710

IP 61726

IP 62707

IP 71001

IP 71707

IP 71750

IP 83750

IP 84750

IP 90712

IP 92700

10 CFR Part 21 Inspections at Nuclear Power Plants

On-Site Engineering

Refueling Activities

Surveillance Observations

Maintenance Observations

Licensed Operator Re-qualification Program Evaluation

Plant Operations

Plant Support

Occupational Radiation Exposure

Radioactive Waste Treatment, and Effluent and Environmental Monitoring

In-Office Review of Written Reports of Non-Routine Events at Power

Reactor Facilities

Onsite Follow-up of Written Reports of Non-Routine Events at Power

Reactor Facilities

IP 92902

IP 92903

IP 92904

IP 93702

Follow-up - Maintenance

Follow-up - Engineering

Follow'-up - Plant Support

Prompt Onsite Response to Events at Operating Power Reactors

A-2

t

r<

Attachment

1 (cont.)

ITEMS OPENED

CLOSED AND UPDATED

OPENED

50-220/98-05-01

50-41 0/98-05-02

VIO

Failure to follow procedure, resulting in a missed plant

shutdown

VIO

Failure to conduct surveillance test on batteries

50-410/98-05-03

VIO

Failure to perform adequate

design for EDG modification

on fuel line

50-220/98-05-04

50-220/98-05-05

50-41 0/98-05-06

NCV

Failure to maintain secondary containment integrity

NCV

Failure to properly evaluate control room emergency

ventilation s stem initiation lo ic

y

g

VIO

Failure to perform calibration of wind speed channel

50-220 5

50-410/98-05-08

NCV

Failure to ensure individuals were randomly tested for

fitness-for-duty

CLOSED

50-220/98-05-04

50-220/98-05-05

50-220 5

50-41 0/98-05-08

50-220/96-01-03

50-220 8L

50-41 0/96-01-05

50-220 5.

50-41 0/96-06-06

NCV

NCV

NCV

URI

VIO

URI

Failure to maintain secondary containment integrity

Failure to properly evaluate control room emergency

ventilation system initiation logic

Failure to ensure individuals were randomly tested for

fitness-for-duty

Lack of Testing of control room annunciators

Failure to perform safety evaluation prior to installation of

temporary modification

Apparent Tampering of fitness-for-duty computer

50-410/96-1 0-03

VI0

Procedure changes not in accordance with TS

requirements

50-220 5

50-41 0/96-1 4-02

50-41 0/97-02-02

50-220/97-1 2-08

50-220/98-05

50-220/98-06

50-41 0/98-05

URI

VIO

URI

LER

LER

LER

Over-pressurization

concerns relative to GL 96-06

Missed TS surveillance on HPCS actuation

instrumentation

Impact of drywell-to-wetwell bypass on containment

pressure

Unrecognized Violation of TS Secondary Containment

Design Deficiency Associated with Control Room

Emergency Ventilation System Radiation Monitors

Reactor Water Cleanup Isolation on High Differential Flow

Caused by Relief Valve Lifting

A-3

la

Attachment

1 (cont.)

CLOSED

50-41 0/98-06

50-41 0/98-07

50-41 0/98-08

50-41 0/98-09

50-41 0/96-1 0-01

50-220/98-01

UPDATED

LER

LER

LER

LER

URI

Part 21

Engineered Safety Feature Actuations Due to Partial Loss

of Offsite Power

TS 3.0.3 Entry Due to Missed Logic System Functional

Testing of Loss of Voltage and Degraded Voltage

Channels

HPCS Out of Service with One Division RHS in

Suppression

Pool Cooling

Missed Battery Technical Specification Surveillance

Requirements

Due to Inappropriate Interpretation

Post-maintenance

testing of Unit 2 MSLRM

Defective GE SBM-Type Switches at Unit 1

50-220 5.

50-41 0/98-08-01

EEI

Failure to properly control, store, and classify safeguards

information - changed to VIO 50-220 5 50-410/98-02-01

WITHDRAWN

50-220 8L

50-41 0/98-08-02

EEI

Failure to report an event in accordance with

10CFR73.71

A-4

Attachment

1 (cont.)

LIST OF ACRONYMS USED

ALARA

cpm

CFR

CREVS

DBG

DER

EA

ECCS

EDG

EEI

EML

ESF

FFD

GE

GENE

GL

HPCS

~is~

IR

LCO

LER

LOCA

LPCI

MMP

MSLB

NCV

NMPC

NOED

NRC

QA

QC

Part 21

PCE

RBEVS

REMP

RFO

RHS

RP5C

SFP

SORC

SRO

SSS

TEMA

As Low As Is Reasonably Achievable

counts per minute

Code of Federal Regulations

Control Room Emergency Ventilation System

Double Blade Guide

Deviation/Event Report

Enforcement Action

Emergency Core Cooling System

Emergency Diesel Generator

Escalated Enforcement Item

Environmental Monitoring Laboratories

Engineered Safeguards

Feature

Fitness for Duty

General Electric

General Electric Nuclear Energy

Generic Letter

High Pressure

Core Spray

Iodine 131

Inspection Report

Limiting Condition for Operation

Licensee Event Report

Loss of Coolant Accident

Low Pressure

Coolant Injection

Meteorological Monitoring Program

Main Steam Line Break

Non-Cited Violation

Niagara Mohawk Power Corporation

Notice of Enforcement Discretion

Nuclear Regulatory Commission

Quality Assurance

Quality Control

10 CFR 21

Procedure Change Evaluation

Reactor Building Emergency Ventilation System

Radiological Effluents Monitoring Program

Refueling Outage

Residual Heat Removal System

Radiological Protection 5. Chemistry

Spent Fuel Pool

Station Operating Review Committee

Senior Reactor Operator

Station Shift Supervisor

Tube Exchanger Manufacturer's Association

A-5

Attachment

1 (cont.)

TS

TSSR

UFSAR

Unit 1

Unit 2

URI

VIO

WO

Technical Specification

Technical Specification Surveillance Requirement

Updated Final Safety Analysis Report

Nine Mile Point Unit 1

Nine Mile Point Unit 2

Unresolved Item

Violation

Work Order

A-6

f'