ML20199L532

Insp Rept 50-354/97-10 on 971116-980103.Violations Noted. Major Areas Inspected:Licensee Operations,Engineering,Maint & Plant Support

ML20199L532
Person / Time
Site:	Hope Creek
Issue date:	02/03/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20199L447	List: ML20199L443 ML20199L479 ML20199L532
References
50-354-97-10, NUDOCS 9802090222
Download: ML20199L532 (29)

See also: IR 05000354/1997010

Text

,.

.

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No: 50 354-

License Nos:- NPF 57

Report No. 50 354/97-10

Licensee: Public Service Electric and Gas Company

Facility: - Hope Creek Nuclear Generating Station

l-

'

Location: P.O. Box 236

Hancocks Bridge, New Jersey 08038

Dates: November 16,1997 - January 3,1998

Inspecters: S. A. Morris, Senior Resident inspector -

J. D. Orr, Resident inspector

Approved by: James C. Linville, Chief, Projects Branch 3

Division of Reactor Projects

I

l

9002090222 980203

gDR ADOCK 05000354

PDR

.

.

.-

.-

EXECUTIVE SUMMARY-'

Hope Creek Generating Station

- NRC Inspection Report 50 354/9710

This integrated inspection report includes aspects of licensee operations, engineering,

maintenance, and plant support. The report covers a sever week period of resident

. Inspection.

Operations '

= Plant operators conducted an essentially error free startup, in spite of numerous emergent-.

plant equipment deficiencies. Operators exhibited excellent control of planned evolutk'ns

and responded ' adequately to unanticipated equipment malfunctions. Reactivity

management during the evolution was good. (Section 01.1)

Operators generally exhibited good performance with respect to procedure compliance,

peer checking, and communication during infrequently performed evolutions while

preparing for plant start up. . Weaknesses were evident with respect to technical .

specification required equipment status control.' (Section 04.1)

,

H ' Operating crew performance during a requalification exam simulator scenario was .

adequate. Though crew communications and event recognition were good, consistent

with observed performance in the actual conuoi room, operators failed to perform or

L evaluate certain expected actions during event response. Fidelity between the Hope Creek

simulator and the actual control room was adequate. Examination security measures were-

good.-(Section 05.1)

The inspectors continued to observe good overall performance with respect to quality-

improvement measures and oversight in operations. The inspectors noted that the recent -  !

n..snagement focus on operability determination backlog reduction was effective. -(Section

07)

Maintenance .

Maintenance department performance, particularly with respect to procedure development

and usage, was weak._ Repeat examples of inadequate procedure development and

implementation were identified.' in at least two cases, associated with maintenance on the 4

reactor core isolation cooling systnm and a control room ventilation system, the

' performance issues resulted in_ delays in returning these safety systems to an operable

status.' (Section M1.1)-

LA recently-developed surveillance procedure which implemented technical specification

acceptance criteria for electrical protection assembly under-frequency testing was

-inadequate. (Section E2.1)

ii

i

_ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ -- .. .. .

.

.

PSE&G personnel completed inservice testing of the "A" and "C" core spray pumps safely

and in accordance with the established procedure. (Section M2.') .

Enoineerina

Engineering support of safe plant operation was acceptable. Several modifications were

developed and installed during the refueling outage which improved the overall material

condition of the plant. However, the backlog of engineering work activities, particularly

with respect to completion of corrective action program related activities, had not been

significantly reduced. Specifically, the timeliness with which condition report evaluations

were completed and corrective actions established exceeded program expectations.

(Section E2.1)

PSE&G efforts at reducing the number of installed temporary modifications was effective.

Remaining modifications were properly evaluated, with proper revisions made in associated

drawings and procedures. However, several of the remaining modifications had been in

place well beyond that which was intended by PSE&G's administrative control program.

(Section E2.2)

PSE&G failed to establish adequate testing and procedure controls to ensure that the

operability of the reactor building to-torus vacuum breaker assemblies was maintained.

This issue was self-identified, properly reported, and promptly corrected. (Section E2.3)

The material condition of the reactor core isolation cooling system and high pressure

coolant injection system was poor as indicated by failed technical specification operability

testing during the reactor plant start up. Inadequate engineering practices contributed to

the failure of the reactor core isolation cooling system surveillance testing. (Section M2.1)

Resolution of a potentially generic deficiency in safety-related swing check valves was

timely and effective. (Section E8.1)

PSE&G implemented timely corrective actions following self-identified and reported

discrepancies in the safety auxiliaries cooling system. These discrepancies, which rangoo

from insufficiently-Justified system operating procedures to errors in design basis

assumptions, were identified during a comprehensive design and licensing basis validation.

(Section E8.6)

. Plant Suocort

The emergency operations facility was well maintained and secured. (Section P2)

PSE&G security personnel performed a thorough evaluation of an event involving an

unauthorized protteted area access. Corrective actions were good. (Section S4) l

ill

.. -- __ _ _ _ -_

,

. ,.

.

TABLE OF CONTENTS -

,

EXECUTIVE SUMM ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii

' TABLE O F C C NTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv -

' l . ' O pe r a t ion s . . . _ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -

.

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

01.1 u Reactor Startup and Power Ascension . . . . . -, . . . . . . . . . . . . . . 1

04 Operator Knowlooge and Performance . . . . . . . . . . . . . . . . . . . . . . s . . . 2

04.1 - General Operator Performance Observations . . . . . . . . . . . . . . . . 2

-05' Operator Training and Qualification . . . . . . . . . .:. . . . . . . . . . . . . . . . 3

05.1' Observation of Licensed Operator Requalification Exam . . . . . . . . 3

07_ Quality Assurance in' Operations . . . . . . . . . . . . . . . .-.. . . . . . . . . . . . . 4 ;

08 Miscellaneous Operations issue . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

08.1 -(Closed) VIO 50-354/97-07-01: failure to promptly identify inoperable

electric fire water pump . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

08.2 (Closed) LER 50 354/97 32: technical specification required shut

down due to concurrent inoperability of the high pressure coolant

injection and reactor core isolation cooling systems . . . . . . . . . . . . 5

li . - M aint e n a nc e . . . . . . . . . . . . .- . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

.

M1 - Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . .' . . . . . . . . . 5

L M1.1 General Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -

M2' Maintenance and Material Condition of Facilities and Equipment . . = . . . . . 7 .

M2.1 Maintenance and Surveillance of High Precsure Safety Systems . . .:7 -

M2.2. Core Spray System inservice Testing Observation - . . . , . . . . . . . .--8

M8 . Miscellaneous Maintenance issuosi . . . . . . . . . . . . . . . . . . . . . . . . . . -. 9 -

' M8.1 (Closed) VIO 50-354/E96125-01013: Failure to plan appropriate ;

post-maintenance technical specification (TS) surveillance testing '. 9l

M8.2 (Closed) VIO 50 354/E96-125-01023: Failure to conduct appropriate -

post maintenance technical specification surveillance testing . . . . 9. .

M8.3 -(Closed) VIO 50-354/97-04-02:high pressure coolant injection (HPCI)

system feed water injection line inoperable . . . . . . . . . . . . . . . . 10 - '

M8.4 (Closed) LER 50 354/97 27:operaticn in a technical specification (TS)

prohibited condition due to improperly performed surveillances for -

determining specific gravity of the 125 Vdc and 250 Vdc batteries

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

M8.5' l Closed) LER 50-354/97-28: technical specification (TS) prohibited

-

condition - f ailure to perform secondary containment isolation -

actuation system surveillances . . . . . . . . . . . . . . . . . . . . . . . . 10

M8.6 (Closed) LER 50 354/97-29: operation in a technical specification (TS)

- prohibited condition due to removal of a residual heat removal (RHR) -

shutdown coeling suction line snubber . . . . . . . . . . . . . . . . . . . 11

- M8.7 (Closed) LER 50-354/97-30: failed inservice test surveillance for

reactor building-to- torus vacuum breakers . . . . . . . . . . . . . . . . 11

1

iv

- - _

[

____J

- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

.

.

. .

.

-

...

1. . -

.

'

s 111. Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1

E2- Engineering Support of Facilities and Equipment ................. 11

E2.1 General Observations -. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

,

E2.2 Review of Temporary Modifications . . . , , . . . . . . . . . . . . . . . . .~ 13 -

L. E2.3 Inadequate Testing of Reactor Building toLTorus Vacuum Breakers-

..............................................15

E8 Miscellaneous Engineering issues ". . . . . . . . . . . . . . . . . . . . . . . . . . . 17

.

E8.1 .(Closed) P2150 354/9511: deficiency in Anchor Darling swing -

<< ' chec k valves . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . 17

- E8.2 - (Closed) VIO 50-354/96-07-01:two examples of a failure 'o evaluate-

_. changes made to the facility in accordance with 10 CFR 50.59 . .17

^ E8.3- (Closed) VIO 50-354/E96125-02013: failure to promptly identify and .

.. correct a condition adverse to quality . . . . . . . . . . . . . . . . . . . . 18

T E8.4 (Closed).VIO 50 354/E96-125-02023:f ailure to promptly identify and

correct a conditlan adverse to quality . . . . . . . . . . . _ .-. . . . . . . 18

E8.5- -(Closed) VIO 50 354/E96-125-03013: failure to obtain NRC approval

before making a change to the facility 1. . . . . .. . . . . . . . . . . . . . . 18

E8.6 (Closed) LER 50 354/96 22: condition alone that could have

iprevented removal of residual heat - safety auxiliaries cooling system

(SACS) deficiencies . . .= . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

1 1V. Plant Support - . . . . . . . . . . . . . - , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - . . . 2 0 :

P2_- Status of Emergency Preparedness Facilities, Equipment, and Resources'

...........-........>...............................20

S4 Security and Safeguards Staff Knowledge and Performance . . . . . . . . . . 20

- V. Management Meetings '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . : 21.

. _

\

-

_ _ _ __ __

_

-- i

,.

v

h'

_

-

. . . , . . . . . _ _ . . . . . - . . . _ , _ _ - - - - - - _ _ - _ _ _ _ _

_ _ - - _ .

. _ - _ _ _ - _ _ _ _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ - -

.

.

Egnort Details

1. Operations

01 Conduct of Operations

Qld Reactor Startuo and Power Ascension

a. Insnection Scoce (71707)

The inspectors observed Hope Creek restart activities from November 30,1997 to

December 8,1997. Specifically, the inspectors monitored the reactor start up to

criticality, extended reactor plant operation in Operational Condition 2, low pressure

testing of the high pressure coolant injection (HPCI) system, a power and pressure

reduction from 900 psig to 100 psig to resolve discrepancies with a jet pumo

, operability surveillance, and safety relief valve lift testing.

!

b. Observations and Findinos

f

'

In general, operator activities conducted during the plant start up were safe and

deliberate, procedure adherence was very good, command and control was evident,

and the control room atmosphere was quiet and unobstructed. The operators

anticipated upcoming evolutions and questioned unexpected results. The inspectors

noted that there was continuoc eactor engineering personnel coverage during

reactivity manipulations. Apprwate decisions were made after equipment

deficiencies were identified. As an example, the control room operators,

anticipating a change to Operational Condition 1 from Operational Condition 2,

performed the " Recirculation Jet Pump Operability - Daily (HC.OP-ST.BB OOO1(Q))"

surveillance to verify that the average power range monitor (APRM) recirculation

flow units were reading accurately. Rasults of this surveillance did not meet the

acceptance criteria. Had the operators not performed this surveillance early, they

would have been challenged with a potential inoperability of the APRMs in

Operational Condition 1. This anomalous condition was promptly resolved before

startup activities proceeded.

However, the operators were challenged with several equipment issues that delayed

the plant start up and required abnormal plant start up operations. For instance, the

initial reactor start up wa. delayed to resolve a higher than normal reactor water

cleanup (RWCU) system differential flow. The operators questioned the abnormal

indication when the RWCU system was restored to service after outage

maintenance to repair a piping flange leak. A portion of the RWCU was taken out

of service and drained for a boroscope inspection of a flow measuring orifice.

PSE&G engineers suspected, based on system parameters, that the flow orifice may

have been installed backwards during the refuel outage maintenance activities. The

boroscope inspection confirmed that the orifice was reversed. (See Section M1.1

for details)

On December 5,1997, five days into the plant start up, while performing reactor

core isolation cooling (RCIC) system testing at rated plant pressure, control room

operators discovered that the turbine speed controller had no effect on steam flow.

___n

_ ___ ___-_-____ - ___ - _ - - -_

.

.

2

The RCIC system was declared inoperable. Control room operators performed the

high pressure coolant injection (HPCI) system testing at rated plant pressure later

that same day. After 24 minutes of operation, the operators tripped the HPCI

turbine due to high vibrations. With both the RCIC and HPCI systems inoperable,

operatore entered technical speeltication 3.0.3 and commenced a plant shutdown to

iess than 150 psig as required. PSE&G personnellater determined that both the

RCIC and HPCI malfunctions were related to earlier outage sctivities on those

i systems. (See Section M2.1 for details)

f

)

l c. Conclusions

Plant operators conducted an error free startup, in spite of numerous emergent plant

equipment deficiencies. Operators exhibited excellent control of planned evolutions

and they responded adequately to unanticipated equipment malfunctions. Reactivity

management during the evolution was good.

04 Operator Knowledge and Performance

.QL1 General Ooerator Performance Observations

a. inspection Scone (71707)

Throughout the report period, the inspectors observed numerous operations

department activities both in the control room and in the field. Several tours with

equipment operators were conducted to evaluate non-licensed operator

performance,

b. Observations and Findinas

Station operators performed several infrequent evolutions during the report period

while preparing the plant for restart from the refueling outage. The inspectors

observed generally good performance with respect to procedure compliance, peer

checking, coordination and communication with other departments, and use of

shutdown risk assessment information. As an example, when it became evident

that the impending plant start up would be performed with only two of the three

reactor feed pumps available, operators evaluated the viability of plant operation

with only two feed pumps during a " test" start up in the Hope Creek simulator.

Good command and control of a post-refueling reactor vesselinservice leak rate test

was evident. Control rod scram time testing and control rod drive mechanism

" friction" testing were completed in accordance with applicable procedures and

without incident.

The inspectors questioned operators regarding specific instances of inadequate

safety-related equipment status control. For example, on December 2,1997, the

inspectors determined that the technical specification (TS) action statement tracking

log, maintained by the shift supervisor as a ready reference of the status of all TS-

related equipment, did not accurately account for the "B" primary containment

hydrogen / oxygen analyzer being out of service for maintenance. Other similar

i

l

_ _ _ __ _. _ _ _._.__._. _ _._ _ . _ _ _ _ -

. _ _ _ _ _

. .

.

l

3  :

'

, examples of.this weaknese had been noted previously by NRC inspectors with

. respect to the operability of the standby liquid control system. Despite the

} weaknesses in maintaining the status log, the inspectors determined from narrative

log reviews, operating shift turnover sheets, and individual Interviews that operators ,

were aware of the actualinoperable system conditions. This minor violation of TS  !

6.8.1.a, which requires administrative procedures be implemented for equipment i

status control, is being treated as a Non Cited Viol? tion consistent with Section IV j

of the NRC Enforcement Policy. (NCV 50 354/97<10 01)

-

l

i

On November 19,1997, the inspectors noted that the reactor pknt mode switch in  !

the control room was in " Refuel," permitting single control rod witndrawals, in spite l

?- of the fact that no rod withdrawals had been conducted for nearly twelve hours. l

Operators had temporarily interrupted scheduled contro! rod testing to address other '

emergent concerns, but elected to leave the modo switch in " Refuel" in part

- because the duration of the rod testing delay was indeterminate, and also because

returning the modo sw tch to " Shutdown" would gsnerate an undesired scram

i, signal. While no consequence resulted from this issue, the inspectors judged that -l

the operators' decision to remain in the " Refuel" mode for an extended period was _ 1

'

not consistent with PSE&G management expectations for conservative operating .

practices.

l

.

c.. Conclusions [

C: erstors exhibited good performance with respect to procedure compliance, peer 3

checking, and communication during infrequently performed evolutions while  :

prepaling for plant start up.' Weaknenes were evident with respect to techical-

specification required equipment status control. .

05' Operator Training and Qualification

QLj, Observation of Licensed Onerator Reaualification Exam

s.. Inanaction Scone (71707) l

The i,.;pectors observed ons operating crew during a routine licensed operator

requalification examination in the Hope Creek control room simulator. Crew:

communications, procedure usa;,e, and control board operation were evaluated.

Additionally, simulator f_idelity to the actual control room configuration was ~

'

assessed. Lastly, the inspectors witnessed PSE&G instructor performance during -  ;

the conduct of the exam scenario, ,

.-  !

v

..

_ _ _ _ _ _ _ _ _ _ . ..

.

.

.. . . . ._. .. .

,

4

4

,

b. ' Observations and Findinas .

On December 19,1997, the inspectors observed the conduct of a routine licensed

operator requalification exam simulator scenario. Security measures to ensure that

. examination materials were not comprised prior to or during the examination were

. good. The inspectors did not detect any incidences of PSE&G evaluators "cuelng"

.

the examinees. Simulator fidelity with the actual Hope Creek control room was

adequate.- Recently implemented control room modifications were noted to be

.

. duplicated in the simulator, however emergency operating proccdures were " pre-

staged" at the shift supervisor desk prior to exam commencement, and well before

conditions developed which would have required their use.

During the exam scenario, the inspectors observed that operators failed to insert a

l " half scram" signal in the reactor protection system following an event which

'

should have automatically cour,ed .s half scram signal but did not. The inspectors

learned that the evaluators also detected this failure, which led to their conclusion -

that the operating crew was * weak" overall in this scenario. Additionally, the

inspectors identified inconsistencies in plant operating procedures with respect to  ;

recirculation system operation. Specifically, the operating crew intentionally '

reduced reactor recirculation flow to a minimum prior to inserting a manual scram,

tyhich resulted in a brief period of plant operation in the power / flow *instsility

region." While the practice of minimizing recirculation flow prior to a planned scram

was in accordance with the procedure employed during the exam scenario, other

operational guidance specifies that recirculation flow never be intentionally reduced

. into the instability region.- Crew communications were good during the exam

l scenario, consistent with observed performance in the actual control room during

plant events.-

- The inspectors discussed their findings with the Hope Creek' operations manager

following completion of the exam. This individual expressed agreement with the

inoted observations, and planned to address them appropriately.

_c. Conclusions -

'

Operating crew performance during a requalification exam simulator scenario was

-adequate. Though communications and evtnt recognition were good, consistent

with observed performance in tM actual control room, operators f ailed to perform or

evaluate certain expected actions during event response. Fidelity between the Hope

. Creek simulator and the actual control room was adequate. Examination security

. measures were good.

07 Quality Assurance in Operations

[Ehe inspectors continued to observe gbbd overall hrformance with respect to

-quality improvement measures and oversight in operations. The inspectors

reviewed quality assurance department reports, interviewed auditors, and attended

an offsite Nuclear Review Board meeting. These activities provided effective

performance assessment feedback to operations management, and in a timely and

useful manner.-- Performance trending in various categories was also evident.

_

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _.

.

.

5

The Inspectors noted that a recent management focus on operability determination

backwg reduction was effective. At the conclusion of the refuel outage, only eight

operability determinations remained from the initial pre outage population of thirty-

five. = Appropriate safety evaluations were completed for the items which remained

to ensure that continued operation with degraded componentn did not pose a threat

to plant safety.

08 Miscellaneous Operations issue

.QHJ IClosed) VL a354/97 07-01:f ailure to promptly identify inoperable electric fire

water pump. ihe inspectors reviewed PSE&G's letter dated December 12,1997

which provided a response to this violation. PSE&G attributed the cause of this

event to personnel error, in that operators failed to follow a procedure precaution

during an electrical bus swap. Further errors were committed by both operations

and fire protection personnel in that these individuals failed to detect the condition

during localinspection tours of the equipment. The inspectors verified that PS2&G

completed the corrective actions stated in the noted letter, which ;ncluded individual

disciplinary action, additional training, and procedure enhancements. The fira water

pump was promptly returned to an operable status upon discovery of this issue.

W (Closed) LER 50 354/97-32: technical specification required shut down due to

- concurrent innoerability of the high pressure coolant injection and reactor core

isolation cooling systems. This issue is described in detail in Section M2.1 of this

report. No new information was provided in this LER.

11. Maintenance

1

!  ; M1. Conduct of Molntenance

ML1 Deneral Observations

a. Inspection Scope (62707)

The inspectors observed or reviewed numerous maintenance department activities

during the report period, which included outage work practices and plant startup

testing evolutions.

b. Observation and Findinas

At the conclusion of the refueling outage and during the reactor start up, the

inspectors noted several maintenance department performanco deficiencies which

-impacted plant operation. For example, the inspectors conducted a thorough

cor.tainment close-out inspection tour prior to plant start up, and identified several

items left in the drywell by maintenance personnel, including tools, loose insulation,

and welding rods. This indicated a lack of sensitivity on the part of maintenance

technicians and supervisors regarding the need to minimize foreign materialin this

area, particularly in light of the fact that the emergency core cooling suction strainer

modification was not fully implemented during the outage.

(

l

_ _ _ _ ___ __.

.

.

6

On November 24,1997, operators experienced two automatic trips of the BK400

safety related control room venthtlon system chiller unit due to high condenser

backpressure. Subsequent PSE&G investigation determined that a recent

maintenance activity to calibrate a system pneumatic pressure controller was not

properly completed. Specifically, a gain adjustment in the control circuit for the

condenser cooling water pressure control valve was laft in the wrong position '

following the calibration, resulting in insufficient coolir.g to the unit during operation.

~

Contrary to PSE&G requirements, "as found" data was not recordd, and procedure

steps were not initialed as being satisfactorily completed. Additionally, the

technician involved in the worL activity was neither trained not qualified to work on

pneumatic controllers. Lastly, supervisory review of the completed work package

failed to detect any errors. This lack of attention to detail and failure to follow

established procedures while performing maintenance on a safety related

component was a violation of technical specification '3.8.1.a. (VIO 5C 354/9710-

02)

On November 29,1997, based on anomalous control room indications, PSE&G

l

engineers determined that a orifice plate flow element in the reactor water cleanup

(RWCU) system had been installed backwards following system piping flange leak

repairs. Rather than reverse the orifice to its proper orientation, engineers

completed a design modification and safety evaluation which re calibrated the

associated flow indication circt.it to account for the new configuration. During

f subsequent reviews into the causal f actors of this issue, the inspectors determined

that the work order guidance provided to the technicians performing the RWCU

piping flange repairs failed to specify the correct orifice plate installation.

Additionally, the technicians did not demonstrate quality workmanship during the

activity in that the "6s found" orifice orientation was not recorded as an aid for

reassembly. This failure to establish appropriate instruction for maintenance on

components which can affect to performance of safety-related equipment,in this

case RWCU isolation controls, was a violation of 10 CFR 50 Appendix B, Criterion

V. (VIO 50 354/9710 03).

On December 30,1997, while attempting to restore from on line reactor core

isolation cooling (RCIC) system maintenance, technicians were unable to open the

turbine trip throttle valve. Subsequent investigation, which include discussions with

appropriate vendor representatives, determined that the mechanical overspeed trip

device was improperly reset which in turn resulted in the inability to reset the

turbine trip throttle valve. Neither the RCIC system operating or maintenance

procedure adequately described the proper method to ensure this interaction was

minimized or precluded oilor to returning the system to standby service. The

inspectors noteti that this problem was not previously understood in spite of

previous industry experience, including a similar 1993 issue with auxillary feedwater

pump turbines at the Salem station. The f ailure to establish appropriate procedures

for the RCIC system as a violation of technical specification 6.8.1.a. (VIO 50-

354/07 10-04)

Lastly, the inspectors identified an improperly developed TS surveillance test

3 procedure for electrical protection assemblies. This issue is described in detai! in

section E2.1 of this report.

i

.

f

7

c. Conclusions

Maintenance department performance during the report period, particularly with

respect to procedure development and usage, was weak. Three examples of

inadequate work controls were identified, which resulted in delays in returning

safety systems to an operable status.

M2- Maintenance and Material Condition of Facilities and Equiprnent

M2d Maintenance and Surveillance of Hioh Pressure Safety Systems

a. inspection Scooe (61726,62707)

During the plant start up following RFO7, the inspectors observed surveillance

testing and maintenance associated with the high pressure coolant injection (HPCI) -

system, the reactor core isolation cooling (RCIC) system, and the automatic

depressurization system safety relief valves (SRV).

b. Observations and Findinog

Station operators experienced several equipment failures with high pressure safety

systems during technical specification-required surveillance testing. For example, in

spite of a high quality pre evolution briefing prior to 800 psig SRV " lift" tests in

Operational Condition 2, difficulties were encountered with the set up of measuring

and test equipment which required operators to re perform the test. Additionally, a

faulty acoustic monitor was identified on the "C" SRV which required replacement

and ratesting. The tests were conducted with good management oversight of the

activitica and effective monitoring of critical plant parameters.

On December 5,1997, the RCIC system was declared inoperable when control

room operators discovered that turbine controls were ineffective at changing turbine

speed. PSE&G later determined that the governor valve stem was binding in the

valve body packing gland. A detailed root cause investigation into this event, wnich

was effect vely coordinated between operations, maintenance and engineering

departmems, determined that the valve stem b!nding was primarily the result of

inadequate engineering review of an earlier March 1997 design change which

replaced the RCIC governor valve stem with a stem of a different material. This

modification was implemented as a corrective action to earlier governor valve stem

f ailures in December 1996, when the RCIC turbine suffered an overspeed trip

because of corrosion on the governor valve stem. PSE&G engineers, in concert

with contracted turbine vendor personnel, determined that the new valve stem

material exhibited c thermal expansic,n rate three times that of the original stem,

which resulted in stem binding when subjected to rated reactor temperature and -

pressure conditions. Additionally, manufacturing tolerances of the replacement

stem were not equivalent to that of the original, exacerbating the issue. Corrective

actions to restore the system focused on widening the tolerances of the governor

valve packing gland spacers to accommodate in the increased thermal growth rate.

The system was properly tested and restored to operable status on December 12,

. _ - .

- _ _ _ _ - _ _ _ - _ .

.

-

.

8

1997. However, inadequate corrective actions were implemented folic, wing the

previous RCIC governor valve stem failure, in violation of 10 CFR 50 Appendix B,

Criterion XVI " Corrective Action." (VIO 50 354/97 10 05)

in addition to the December 5,1997 RCIC testing described above, operators

condected HPCI system surveillance testing later on that same day. This test also

f ailed due to high vibrations detected approximately 20 minutes into the test run.

Since both HPCI and RCIC were declared inoperable while in Operational Condition

2, operators entered technical specification (TS) 3.0.3 and reduced reactor plant

power and pressure to the point at which neither sistem was required to be

operable (l.a.< 150 psig). Operators stabilzod the plant in the appropriate condition

safely and well within TS prescribed criteria. Again, a comprehensive team

comprised of individualc from various departments and vendor representatives was

assembled to develop a troubleshooting plan and determine the root cause of the

problem. After a thorough review, which included intrusive system inspections, the

team concluded that excessive moisture collection in the turbine glands corroded

the internal carbon steam seals and caused the noted turbine vibrations. The

inspectors reviewed the team's efforts and found that the team's conclusions were

reasonable. Corrective actions included the replacement of the gland seal

components, and the initiation of a system modification to provide improved gland

l

'

seal cavity draining. The system was successfully tested and restored to operability

on December 12,1997,

c. Conclusions

The material condition of the reactor core isolation cooling system and the high

pressure coolant injection sys'3m was poor as indicated by failed technical

specification operability testing during the reactor plant start up, inadequate

engineering practices contributed to the failure of reactor core isolation cooling

system surveillance testing.

M2.2 Core Sorav system inservice Testina observation

a. Insoection Scope (61726)

The inspectors observed the performan:e of the "A" and "C" core spray pump in-

service tests (IST). Other than the pre-evolution brief conducted in 'he control

room, the inspectors focused on the actions of the operators and technicians in the

local core spray pump rooms,

b. Observations and Findinas

The pre-job brief for the quarterly inservice test was thorough and was conducted in

accordance with the operations department standard checklist. A.ll participants in

the evolution were present and actively contributed lessons learned and contingency

plans for potential unexpected occurrences,

i

I

l

1

._. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ - _ _ - _ -

.

9

The inspectors interviewed the cognizant equipment operator and instrument '

technician regarding the installation and use of their respective measuring and test

equipment (M&TE). These individuals were knowledgeable in the use of the M&TE.

The inspectors observed that the MGTE was properly installed and within

calibration.

The inspectors observed the instrument technician spill some potentially

contaminated water from a system test connection. The technician reported the

potential problem to the radiological controls technician when he arrived on scene to

support the surveillance. The inspectors determined that their actions were prompt

and reasonable to preclude any potential radiological problems,

c. Conclusions

PSE&G technicians completed inservice testing of the "A" and "C" core spray

pumps safely and in accordance with the established procedura and technical

specification 4.0.5,

M8 Miscellaneous Maintenance issues

i

l MB.1 (Closed) VIO 50 354/E96-125-01013: Failure to plan appropriate post maintenance

techriical specification (TS) Surveillance testing. NRC inspectors identified that

PSE&G had not planned to perform TS 4.1.3.2.b required control rod scram time

testing before entering Operational Condition 1. This testing was required to

demonstrate the operability of over sixty control rods which had undergone

maintenance during RFO6. Similarly, control rod scram time testing had not been

adequately performed prior to restart from earlier refueling outages, specifically

RF03 and RF05 (see M8.2 below). The inspectors reviewed the corrective actions

implemented as described in PSE&G's November 22,1996 violation response letter.

PSE&G's actions immediately following identification of this issue were prompt and

comprehensive. These actions included conduct of the proper TS surveillance

testing, a review of other " conditional" TS requirements, and independent

assessments of the station's readiness for a change in operational condition.

Subsequently, in-depth TS training was provided to operations personnel to

reinforce the nc ' to operate and maintain the plant in accordance with the

licensing basis, Lastly, the inspectors reviewed PSE&G control rod testing plans

following RFO7 maintenance. The inspectors verified that all required scram time

tests were completed in the proper operational condition and in compliance with TS 4.1.3.2.b.

MM IChsed) VIO 50-354/E96-125 01023: Failure to conduct appropriate post-

maintenance technical specification surveillance testing. This issue involved the

discovery that following maintenance in RF03 and RF05 which had the potential to

affect control rod scram insertion times, no TS 4.1.3.2.b surveillance testing was

performed as required. The corrective actions for this issue are identical to those

described for the event described in section M8.1 above,

w

. - -_ _ _ - . _- - -. - -- . - - - . . -_ -

.

.

10

MHJ (Closed) VIO 50 354/97 04 02:high pressure coolant injection (HPCI) system feed

water injection line inoperable. This event, along with PSE&G's root cause analysis

and proposed corrective actions to prevent recurrence, were described in detail in

LER 50 354/9713 00and 01. The inspectors reviewed these documents and the

stated corrective actions, and documented their findings in NRC Inspection Report

50 354/97 09. Additionally, PSE&G responded directly to the violation in a

September 9,1997 letter to the NRC. This letter did not provide any significant

additionalinforrration beyond what was described in the LER and its supplement,

with the exception that PSE&G disagreed with the inspectors' assessment that the .

" initial decision to maintain the operability of the HPCI system was non-

conservative." While the inspectors acknowledged PSE&G's argument that

operators made the decision in accordance with the " normal plant (operability

_

determination) process," the inspectors maintained that the condition of the HPCI

' system feedwater injection valve was not certain at the time the decision was

made, and that the organization did not effectively validate, verify, and interpret

information gathered through troubleshooting.

MHd (Closed) LER 50 354/97 27: operation in a technical specification (TS) prohibited

condition due to improperly performed surveillances for determining specific gravity

of the 125 Vdc and 250 Vdc batteries. TS 4.8.2.1.b.1 requires that specific gravity

of the safety related batteries be determined every 92 days and be within the -

specifications provided in Table 4.8.2.1 1. TS Table 4.8.2.1 1 requires that each

cell's specific gravity be corrected for temperature for every individual cell. Hope

Creek previously used, since initial operation, every sixth cell to correct for

temperature. The inspectors noted in the LER that this change in methodology

produced very minor changes in specific gravity measurements. The inspectors

verified by on site inspection that PSE&G completed procedure changes for the

affacted surveillances. This licensee identified and corrected violation of TS is being

treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC

Enforcement Policy. (NCV 50 354/9710 06)

l M{Lii (Closed) LER 501354/97 28: technical specification (TS) prohibited condition -

failure to perform secor.dary containment isolation actuation system surveillances.

This event involved :.he self identified discovery that outage planning personnel

failed to include channel functional tests required by TS in the outage plan. As a

L result, surveillance activities required by TS 4.3.2.1 were not performed to

demonstrate operability of the secondary containment isolation achation logic

channels during core alterations or during operations with a pot .al to drain the

reactor vessel. PSE&G attributed this failure to personnel error in that outage

planners inappropriately used an uncontrolled TS surveillance matrix while

developing the refueling outage testing schedule. This matrix did not accurately

reflect all TS surveillance testing requirements. PSE&G procedures require that the

L contro!!ed surveillance testing matrix, which was recently validated by the technical

'

specification surveillance improvement project, be used for planning and scheduling

TS required testing activities. Corrective actions to preclude recucence of this issue

involved Individual disciplinary measures and re emphasis on mapp;nent

expectations regarding the use of only controlled documents. Based on an in-office

review the inspectors concluded the corrective actions were adequate.

l

l

. . . . . _ . _.

- _ _ - _ - _ _ _ _ _ _ _ -

.

11

This licensee identified and corrected violation of TS 4.3.2.1 is being treated as a

Non Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.

(NCV 50-354/9710 07)

MlLQ (Closed) LER EO 354/97 29: operation in a technical specification (TS) prohibited

condition due to removal of a residual heat removal (RHR) shutdown cooling suction

line snubber. On November 18,1997, during implementation of a design change

package (DCP) to replace snubber 1 P BC 144 H002on the RHR shutdown cooling

suction line, personnel identified that the snubber had been mistakenly removed in

1992 during the snubber reduction program. PSE&G's investigation determined that

individual work orders to remove snubbers were not generated from the approved

snubber reduction DCP in 1992. PSE&G performed a comprehensive walkdown of

the shutdown cooling suction line and did not identify any additional missing

snubbers. PSE&G also performed a review of outage work orders and did not

identify any activities that would have unintentionally removed snubbers.

The inspectors interviewed licensee personnel in PSE&G's inservice inspection and

work planning departments to determine the effectiveness of the corrective actions.

All required system snubbers were replaced with a new model during RF07. The

inspectors determined that appropriate licensee personnel v'ere knowledgeable of

this event and that they had understood the lessons learned. This licensee

identified and corrected violation of TS 3.7.5 is being treated as a Non-Cited

Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-

354/97 10 08)

ML2 (Closodi LER 50-354/97-30: failed inservice test surveillance for reactor building to-

,

!

torus vacuum breakers. This issue was discussed in detailin section M2.3 of this

report. This LER accurately described the circumstances involved in the discovely

and resolution of the issue.

Ill. Enainee.ririg

E2 Engineering Support of Facilities and Equipment

E2d General Observations

a. Insocction Scoce (37551)

The inspectors reviewed several engineering department work products, including a

quarterly maintenance rule performance report, se,veral condition report root cause

evaluations, and design change package safety evaluations. Intervicws were

conducted with engineering personnel as appropriate.

b. Observations and Findinas

During the report period, the PSE&G engineering department completed a

reorganization which resulted in a staff reduction of nearly eighty employees.

_______ _

.

12

Additionally, the department completed a comprehensive " engineering backlog"

review which was a systematic assessment of all outstanding PSE&G engineering

work items in ordar to prioritize necessary work activities and eliminate non-

productive efforts. The inspectors concluded that this latter activity was a positive

initiative to fully understand the volume of needed enginecting work following the

reorganization. Further, based on personnelinterviews with operations and

maintenance department staff, the inspectors found that these personnel were

sometimes unaware of the appropriate engineering points-of contact.

The inspectors noted that the backlog of corrective action program activities

assigned to the engineering department continued to exceed programmatic

expectations. Additionally, two condition report evaluations reviewed by the

inspectors were not completed until well after programmatic time constraints.

These two condition reports involved: (1) a question regarding the design adequacy

of an unmonitored turbine building battery ventilation exhaust path, and (2) a f ailure

of a solenold-operated valve in the MSIV sealing system.

The inspectors reviewed the effectiveness of some of the design change packages

(DCP) impleme'ited during the refueling outage. For example, the reactor

recirculation pump seal modification, in conjunction with modifications which

improved Leal drain lines, appeared to be effective in that initial seal cavity pressure

trends during the first month of operation after the outage were much improved

over the old design. Additionally, the unidentified leakage collected in the drywell

floor drain sumps exhibited significant improvement,largely due to this modification.

The drywell equipment drain sump identified leak rate also improved. Plant

operators benefitted from other engineering department initiated modifications,

including elimination of the rod sequence control system, removal of recirculation

pump motor generator set ventilation runback signals, and installation of 4160 VAC

switchgear panel test points. Virtually every mechanteal mubber in the facility was

replaced with a newly designed and engineered hydraulic design.

The inspectors performed a comprehensive review of DCP 4HE-0382 which

replaced logic cards in the reactor protection system (RPS) end power range neutron

monitoring system (PRNMS) electrical protection assemblies (EPA). This

modification was developed based on vendor recommendations. The 10 CFR 50.59

applicability review associated with the DCP was sufficiently detailed and properly

referenced the necessary licensing basis documents to permit the inspectors to

independently reach the same conclusion es PSE&G, specifically that the

modification did not require a safety evaluation.

In conjunction with this DCP evaluation, the inspectors reviewed the completed

work orders and installation and test procedures utilized to complete the logic card

modification. The work order documentation was sufficiently detailed, and all

necessary approvals, tag outs, and retests were specified. The post maintenance

test required successful completion of procedure HC.MD ST.SB 0000(Q) Revision 0,

~16 Month Electrical Protection Assembly Channel Calibration." Two deficiencies

vvere identified as a result of the inspector's review of this completed procedure.

Specifically, maintenance technicians failed to document "as found" voltage and

i

_ _. ..

__-____ - _ _

.

.

13

frequency readings to et least the same number of signibcint digits 3 stated in the

procedure test acceptance criteria, meaning that the recorded data could not be

adequately evaluated for acceptability. Secondly, step 5.9.8 of the procedure

specified the TS acceptance criteria for under frequency measurements as "2 57

hertz." The actual TS 4.8.4.6.b criterie for the PRNMS EPA's is "57 hertz 0%

+ 2%." Though "as lef t" setpoints were within the actual TS criteria, the f ailure to

establish surveillance test procedures consistent with TS requirements waa a

violation of TS 6.8.1.d. (VIO 50 354/9710 09)

Additionally the inspector noted that the reactor core isolation cooling (RCIC)

system surveillance test failure on December 5,1997, was primarily attributed to

inadequate development of corrective actions to previous RCIC governor valve stem

issues. (See Section M2.1 of this report for details)

c. Conclusions

Engineering support of safe plant operation was acceptable. Severa! modifications

were developed and mstalled during the refueling outage which improved the overall

material condition of the plant. However, the backlog of engineering work

activities, particularly with respect to completion of correctivt action program

related activities, was not significantly reduced. Engineering support with respect

to developing efffective corrective actions to previous reactor core isolatinn cooling

system f ailures was inadequate. A surveillance procedura which implemented

technical specification acceptance criteria for electrical protection assembly under-

frequency testing was inadequate.

E2,2 Review of Temocrarv Modifications

a. IDipaction Scone (37551)

The inspectors reviewed the status and tracking of temporary modifications to

verify that they met regulatory and PSE&G requirements for acceptability.

Additionally, the inspectors conducted a thorough evaluation of one temporary

modification, which included interviews with the applicable system manager and

reviews of affected drawings and procedures.

r

b. Observatione and Findinos

Hope (' reek had 24 active temporary modifications prior to the beginning of RFO7.

By the end of the outage, only eight modifications remained, indicating a good

management focus on temporary modificction reduction over the period. However,

of the eight, five were greater than one year old. While appropriate safety

evaluations had been completed Justifying the continued existence of each

remaining modification, the inspectors determined that the intent of PSE&G's

nuclear administrative procedure NO.NA-AP.ZZ-0013 (NAP-13), "Controlof

Temporary Modifications," hbd not been satisfied. This procedure 3tipulates that

such modifications should be "of short duration," and " generally not used for longer

than 90 days." The oldest temporary modification at the facility was greater than

.

_- _ _ _ _ _ _ _ _

___-__- - -

.

14

three years old. All eight modifications had planned removal dates before or during

the next refueling outage.

Additionally, the inspectors noted that on December 12,1997, a quality assurance

inspector identified numerous minor deficiencies in the temporary modification

tracking log in the Hope Creek work contro! center. While none of the

discrepancies affected the operability of r.ny safety-re' .ted systems, they

collectively indicated tnat PSE&G's programmatic requirements were not being

satisfactorily implemented. The condition report documenting this issue thoroughly

described the deficiencies.

The inspectors performed a detailed assessment of tempor . / modification #94-

026, titled " Drain KP (MSIV Sealing) System Water from KL (PCIG) system," which

was installed on September 10,1994. This modification was installed tc mitigate

the effects of leakage from the main steam system into the normally idle MSIV

sealing system. This leakage was negatively affecting the operational performance

of the primary containment instrument gas (PCIG) system, which supplies

pneumatic control for several safety-related applications. The modification involved

installation of a drain hose to an MSIV systern test connection, and the opening of

normally deenergized solenoid operated va: /es. /

The 10 CFR 50.59 safety evaluation associated with this modification was

adequate; however, in part because of when it was written, it did not meet current

PSE&G engineering department expectations for quality. As a result of inspector

questioning, the evaluation was re written (and re approved by the station

operations review committee) using the current engineering department standards.

MSIV Sealing and PClG system drawings we.e properly revised to reflect installation

of the modification. Affected operating procedures were also revised appropriately.

The inspecto s asked why this modification had been permitted to endure beyond

NAP 13 expectations, and learned that initial attempts to resolve the issue with  ;

corrective maintenance were unsuccessful. Later, a decision was made to remove '

the MSIV Sealing system entirely, but licensing actions were first required to effect

the change. Ultimately, this licensing action was deferred to the current operating

cycle due to PSE&G management placing greater priority on other needed licensing

actions.

c. Conclusions

PSE&G efforts at reducing the number of installed temporary modifications was

effective. Remaining modifications were properly evaluated, with proper revisions

made in associated drawings and procedures. However, several of the remaining

modifications had been in place well beyond that which was intended by PSE&G's

administrative control program.

l

_ _ _ _ - _ _

_ _ _ _ _ _ _ _______ - _-_ _ _ _-

.

15

E2J inadeauste Testina of Reactor Buildina to Torus Vacuum Breakers

a. Insocction Scoce (61726,37501)

The inspectors reviewed PSE&G's actions to address insufficient reactor building to-

'

torus vacuum breaker (RBTVB) testing which directly impacted their operability.

Specifically, excessive back-leakage tnraugh instrument air (IA) system supply

check valves would have prevented an accumulator, designed to provide a minimum

of two hours of operation for the pneumatically operated RDTVBs, from performing

its intended safety function. A two hour period is assumed in design Sosis

calculations to provide adequate time for manual alignment of the backup pneumatic

supMy from the primary containment instrument gas (PCIG) system,

b. Observation and Findinas

,

The RBTVBs are designed to open during certain accident conditions to prevent a

subatmospheric condition from developing in the torus air space. Under certain

circumstances, a small differential pressure would develop between the reactor

building and torus, which could challenge the structuralintegrity of Me torus. Eithei

l

'

of two sets of RBTVB would open to equalize pressure, in the Hope Creek design,

there are two sets of RBTVBs (each comprised of s check valve and an air operated

isolation valve (AOV) in series). An accumulator associated with each RBTVB

assembly is designed to provide the capability of opening the RBTVB AOVs

(1GSHV 5029 & 5031) over a two hour period when offsite power is lost. Each

accumulator is supplied by two diUerent systems, the lA system (normal) and the

PCIG system (backup), via isolation check valves. The PCIG supply piping also

includes a normal ly closed manualisolation valve. The lA check valves (1GSV-081

& 1GSV 093) must provide a relatively leak tight boundary to ensure that the

RBTVB accumulator can perform its intended safety function.

On Novt,mber 12,1997, P3E&G engineers discovered that previous testing of these

lA supply check valves was inadequate since it verified only that the valves were

closed by monitoring for gross leakage from the system, in order to preserve

RBVTB design basis assumptions, proper testing should have periodicall/ quantified

actualleak rate conditions. This would ensure that any lA check valve leakage

would not reduce the RBTVB accumulator pressures to unacceptably low levels

within two hours following a loss of offsite power. At the time of this discovery,

the reactor plant was shutdawn for a refueling outage.

The allowable check valve leak rate was derived from PSE&G calculation H 1 GS-

MDC-1105, which demonstrated that the accumulator size of 19.0 cubic feet was

sufficient to ensure one stroke of the AOV since it needed a minimum of 17.01

cubic feet of _ air at 80 psig to operate. Thus, the maximum allowable leak rate

through the interfacing check valves for the two hour design period was calculated

to be 3345 secm. On November 21,1997, a leakage rate test was performed to i

measure the actualleakage through the lA check valves in both RBTVB trains. The

inspectors noted that both trains exhibited leakage above the calculated maximum

rates which challenged the operability of the RBTVBs. PSE&G initiated prompt

I

_ _j

_. . _ _.

.

.

16

actions to rcwork the seating surf aces of the deficient check valves, which was

proven successful by subsequent satisf actory leak rate tests. In addition to the

check valve rework and testing, Licenseo Event Report (LER) 07 030 00,lmproper

in Service Test Hesults in Inoperable Containment Atmospheric Control System

Vacuum Breaker Assemblies, indicates that the PSE&G testing program has been

upgraded to list the affected valves as requiring leakage rate testing overy two

years. The inspector verified that these actions were completed.

Additionally, the Hope Creek design basis assumes that a backup pneumatic supply

from PCIG will be manually aligned to the RBTVBs within two hours of the initial

loss of offsite power. However, this requirement was never incorporated inn

operating procedures. The inspectors judged that, as a result of the "as found" lA

supply check valve leakage, combined with the lack of operational guidance to align

a backup pneumatic supply, it was unlikely that the RBTVBs would have operated

beyond one hour following the design basis events for which they were instr.: led to

mitigate. This was a violation of 10 CFR 50, Appendix B, Criterion V which

requires the establishment of procedures for activities that affect quality. These

procedures must include appropriate quantitative and qualitative acceptance criteria

for determining that important activities have been satisfactorily accomplished.

These circumstances, which resulted in the simultaneous degradation of two

independent and redundant trains of safety related primary containment vacuurn

relief, both required to be operable per technical specification 3.6.4, have existed

since initial plant construction and operation. Appropriate testing, if conducted,

would have confirmed the existence of excessive lA check valve leakage which

would have necessitated repairs to the affected valves. This would have provided

reasonable assurance that the RBTVBs would have been maintained in an operable

status fer the two hour period specified in the design basis. Thus, PSE&G's self-

identified failure to perform required leakage rate tests of the lA supply check valves

associated with the RBTVB assemblies was a violation of 10 CFR 50, Appendix B,

Criterion XI, Test Control, which requires appropriata testing of systems bo

identified and conducted in accordance with applicable procedures that specify

acceptance limits derived from design documents. LER 97-030 notes that operating

procedures have been revised to include operator actions to place the backup

pneumatic r,upply in service to support RBTVB operability. The inspectors verified

that these actions were comp eted.

These licensee identified and corrected violations of 10 CFR 50 Appendix B, Criteria

V and XI, are being treated as Non Cited Violations, consistent with Section Vll.B.3

of the NRC Enforcement Policy. (NCV 50 354/97-1010and 971011)

a. Conclusions

PSE&G failed to establish adequate testing and procedure controls to ensure that

the operability of the reactor building to-torus vacuum breaker assemblies was

maintained. Specifically, calculational assumptions made to encure their operation

following design basis events were not adequately translated into plant operating

and testing procedures. However, this deficiency was self identified during a non-

- _ _ - _ ___ __ _____ _ _ - - - _ .

.

.

17

routine comprehensive inservice test program review, and resulted in the prompt

development and imp!ementat;on of appropriate corrective actions.

E8 Miscallaneous Engineering issues

fil [ Closed) P2150 354/9511: deficiency in Anchor Dariing swing check valves. This

potentially-generic issue, and the events which led to its discovery at the Hope

Creek station, were discussed in detailin NRC Inspection Reports 50 354/94 26

and 95 03, PSE&G made an initial NRC notification of this issue, which involved a

design and configuration control deficiency with sixteen pump discharge check

valves (twelve safety related), on March 29,1995. PSE&G subsequer:tly completed

a formal 10 CFR 21 notification in a letter dated April 28,1995. Because of the

reported deficiencies, the moving valve disks and hinge arms were contacting and

deforming the valve bodie . The inspectors reviewed the corrective actions

proposed in PSE&G's report document, v'hich included internalinspections of each

susceptible check valve. Based on the inspections, several hinge arms and disks

were replaced or repaired by weld buildup. The inspectors verified by on site

inspection that internal valvo inspections completed subsequent to these repairs

proved that the repairs were effective in preventing further valve damage.

1

EfL2 (Closed) VIO 50 354/96 07 01:two examples of a failure to evaluate changes mada

,

to the facility in accordance with 10 CFR 50.59. The first example involved

operation of a safety auxiliaries cooling system (SACS) valve (1EGHV 2522E)in a

manner not permitted by station operating procedures. S%cifically, this valve,

which the UFSAR describes as providing an automatic isolation function between

the turbine auxillaries cooling systern and the safety-related SACS following a pipe

break, was failed open af ter discovery of an internal valve actuator oil leak. The

second example involved the placement of non safety related video camera cables

across multiple trains of safety related cabling, which defeated required channel

separation.

The inspectors reviewed PSE&G's March 31,1997 ietter which responded to this

violation. Corrective actions specified for the first issue included a procedure

revision which allowed operators to fail open 1EGHV 2522E (or F) under abnormal

circumstances. However, PSE&G engineers conducting a design basis review and

validation project for the service water and SACS systems later determined that the

10 CFR 50.59 safety evaluation used to loatify the procedure change was also

inadequate. Specifically, the evaluation , # 2.0 provide adequate assurance that

f alling open either one of the two noted au.wnatic isolation valves preserved the

safety function of SACS under all postulated accident conditions. The inspectors

reviewed the revised safety evaluation for this precedure cbenge and judged that its

conclusions were well founded, and appropriately focused on the safety implications

of the change.

The inspectors also assessed PSE&G's resolution of the second violation example.

The inspectors verified that PSE&G changed the governing procedures for

installation of temporary cables by including specific guidance and criteria for cable

routing. Additionally, focused training on this topic was provided to both fire

-

_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _

. _ _ _ _ _ - - - _ _- _-_ .

.

.

.

18

protection and radiation control personnel since these departments use temporary

video cameras and cabling most frequently. No further examples of this issus have

been identified.

ESJ (Closed) VIO 50 354. J6125 02013:f ailure to promptly identify and correct a

condition adverse to quality. This violation involved fifteen pairr of high energy line

break backdraft isolation dampers which had been installed backwards in the

reactor building ventilation system. This condition was self identified in 1992 as a

design discrepancy, yet no action to resolve the discrepancy was taken until "re-

discovered" by NRC Inspectors during a restart assessment team inspsetion during

RF06. The inspectors reviewed PSE&G's November 22,1996 letter which

responded to this violation, and judged tha the root cause evaluation and proposed

correctiva actions were reasonable and comprehensive. PSE&G attributed the

cause oi this issue to non conservative decision making, an inadequate corrective

action program, and a poor appreciation for maintaining the plant in accordance

with the design and licensing bases. The inspectors verified that corrective actions,

which included reorientation of the reversed dampers, development of an improved

corrective action program, and focused training on design and licensing bases

maintenance, were implemented. PSE&G c!so performed an " extent of condition" c

review for several other important to safety plant systems in an attempt to identify

other potential design basis discrepancies. These efforts were discussed in detailin

both LER 50 354/96 06and NRC Inspection Report 50 354/96 80.

EQd (Closed) VIO 50 354/E96125-02023: failure to promptly identify and correct a

condition adverse to quality. This violation involved an NRC inspector discovery

that control rod speeds were not maintained in a manner consistent with design

l basis assumptions for a startup rod withdrawal error scenario. PSE&G reported this

issue to the NRC in LER 50 354/96-08 00and -01. The inspectors reviewed this

l LER and PSE&G's November 22,1996 letter which responded to the violation.

r-

PSE&G concluded that this issue was attributed to a failure to include UFSAR-

assumed control rod speed design values as acceptance criteria in the control rod

speed adjustment procedure, and that operations and engineering personnel were

-

not familiar with the associated design basis assumptions. Corrective actions

included: (1) validation that all post maintenance "as-left" control rod spaed values

were consistent with design basis assumptions, (2) procedure enhancements, and

(3) design and licensing basis training for both operator 5 and engineers. The

inspectors concluded that PSE&G's assessment and corrective actions were

reasonable. The inspectors also observed that control rod speed adjustments

performed during RFO7 were proper.

EQJ (Closed) VIO 50-354/E96-125-03013: failure to obtain NRC approval before making

a change to the facility. This issue involved an intentional change to the operation

' _

of the station service water (SSW) pump discharge strainers, in that backwash

valve operation was modified to allow a continuous (vs. cyclic) partial SSW flow

diversion from the safety auxillaries cooling system (SACS) heat exchangers. This

change was developed as a means to improve discharge strainer performance.

When implemented, the actual backwash flow diverted from SACS was much

greater than calculated (2000 gpm vs. 500 gpm). Engineering personnel then

. . . . .

.. . .

.

.

.

.

. .. .. .

..

. . _ _ _ _

_ _ _ _ _ _ _ _ _ _-

.

.

19

recognized that this modification invalidated the technical specification (TS) '

maximum allowed ultimate heat sink (UHS) temperature limit, but argued that the

use of administrative controls on maximum UHS temperature were sufficient to -

permit retention of the modification under 10 CFR 50.59.

In their November 22,1996 violation response letter, as well as during an

enforcement conference held on June 11,1990, PSE&G agreed that a violation

occurred, but did not agree with the NRC's assessment of the f acts. The inspectors

judged t' : since implementation of the modification rendered TS limits non-

conservative, it constituted a change to the facility requiring a change of TS.

However, PSE&G argued that the calculated vs. actual flow discrepancies qualified

as " discovery" following the modification, and that this discovery should have been

handled as a degraded condition (but was not), and characterized as " operable but 2

degraded"in accordance with NRC Generic Letter 91 18. As such, compensatory

measures such as placement of a.iministrative limits on system operating

parameters would have been permissible.

The inspectors verified that PSE&G Implemented several corrective actions +o

prevent recurrence of future similar issues. Specifically, PSE&G revised their 10

CFR 50.59 safety evaluation process to include recently issued industry and NRC

guidance in this area. Additionally, the operability determination process employed

at the station was completely revised, addmg more design and licensing basis

review requirements, as well as additional internal enp'neering and management

reviews. Extensive training for operations and engineering personnel was provided

, regarding maintaining the facility in accordance with design and licensing basis

'

assumptions. Finally, a comprehensive SACS and 60W design basis re validation

effort was completed in May 1997 to provide further assurance that plant operation

was consistent with license requirements.

ELD (Closed) LER 50 354/96 22: condition alone that could have prevented removal of

residual heat - safety auxiliaries cooling system (SACS) deficiencies. The inspectors

reviewed the original and two supplemental revisions to this LER. The issues

described in this LER were also briefly discussed in NRC Inspection Reports 50-

354/9610 and 97 06. Three related self-identified issues were reported: (1) SACS

operability could not be assured under all allowable system configurations, (2) TS

allowed outage times for SACS, station service water (SSW), and the emergency

diesel generators (EDG) were based on insufficient engineering justifications, and (3)

phnt operating procedures permitted a SACS configuration unsupported by the

system's design basis. These issues were identified during an independent SACS

des!gn basis review and validation effort performed as a corrective action for

previously-identified similar issues (see LER's 50 354/95-37,96-09, and 9615).

The inspectors reviewed the root cause information provided in this LER, along with

the stated corrective actions. PSE&G concluded that inadoquate design and

licensing basis reviews and validations were performed during initial system startup

test and operating procedure development in the mid 1980's. These inadequate

reviews resulted in conditions which permitted system operation outside of design

basis assumptions. Inadequate personnel knowledge of system licensing basis

-

.

.

20

information contributed to the problems. When discovered, as an interim measure,

administrative controls and compensatory actions were developed as part of a

formal operability determination to assure that system operability was maintained.

PSE&G corrective actions included a comprehensive SACS system flow balance,

under all postulated operating configurations, to fully understand system flow

characteristics. The inspectors observed portions of this testing which was

conducted during RFO7. Additionally, the limiting conditions for operation for

SACS, SSW, and ultimate heat sink were revised in TS amendment 100 to account

for these and other identified design basis discrepancies such that the

compensatory measures specified by the noted operability determination were no

longer necessary. The inspectore verified by on site inspection that several other

corrective actions have been completed since these issues were first identified,

including design and licensing basis treir'ing for operations and engineering

personnel, operating procedure revisions,10 CFR 50.59 safety evaluation process

changes, and inservice testing program enhargements.

The inspectors concluded that PSE&G Implemented prompt and comprehensive

corrective actions following their self discovery of the noted issues. The stated root

causes for these concerns were sound, and not reasonably linked to current

performance. Since these issues were described earlier in NRC lospection Report

50 354/97 06,in which enforcement discretion was applied not h cite a violation

of technical specification requirements for SACS, these issues are Non-Cited

Violations consistent with Section Vll.B.3 of the NRC Enforcement Policy. (NCV

50 354/97 10 12)

IV. Plant Support

P2 Status of Emergency Preparedness Facilities, Equipment, and Resources

On December 19,1997, the inspectors conducted a walkdown of PSE&G's

Emergency Operations Facility located in Salem, New Jersey. The facility was well

maintained and secured. Telephone systems were operational, status boards were

cleaned of old information, and technical resource documents were available and

up-to-date. The inspectors concluded that the facility was appropriately maintained

in a standby condition for prompt activation.

S4 Security and Safeguards Staff Knowledge and Performance

On November 24,1997, security department personnel determined that guards at

- the access control f acility failed to perform an adequate search of an Individual

entering the protected area. The individual alarmed the metal detoctor and required

a hands-on search prior to entry in accordance with access control requirements.

However, according to a subsequent internal PSE&G investigation, the guards did

not maintain positive control of the individual nor did they conduct the necessary

search before the individual entered the protected area. This error was quickly

recognized and the individual was returned to the access control center within

seconds after he entered the protected area. He was properly searched before

_ _ _ - - _ _ _ _ _ _ . - _ - _ - _ _

.

.

21

being permitted to re enter. According to the PSE&G investigation, inadequate

communications, inattention to detail, and work site distractions contributed to this

event.

PSE&G performed a thorough review of this, event, and ;dentified eqveral corrective

actions. Included among these actions were individual disciplinary measures, re-

emphasis on maintaining positive control over individuals attempting to gain

protected area access, and enhancements to security post orders with respect to

personnel access control. The inspectors judged PSE&G actions following this

event as thorough. This licensee identified end corrected violation of technical

'

specification 6.8.1.e, which requires site security plan procedures be implemented,

is being treated as a Non Cited Violation, consistent with Section Vll.B.1 of the NRC

Enforcement Policy. (NCV 50 354/971013)

V. Manaaement Mettingt

X* Exit W.coting Summary

The inspectors presented the inspection results to members of licensee management at the

conclusion of the inspection on January 20,1998. The licensee acknowledged the

findings presented.

The inspectors asked the licensee whether any materials examined during the inspection

should be considered proprietary. No proprietary information was identified.

ii

~

. _

.

.

22

INSPECTION PROCEDURES USED

IP 37551: Onsite Engineering

IP 61726: Surveillance Observations

IP 62707: Maintenance Observations

IP 71707: - Plant Operations

IP 71750: Plant Support

IP 90712: Inoffice Review of Written Reports

ITEMS OPENED, CLOSED, AND DISCUSSED

Qoened

50 354/97 10-02 VIO Failure to implement maintenance procedure.

50 354/97 10-03 VIO Failure to establish appropriate instructions

50 354/97 10-04 VIO Failure to establish appropriate procedure

50 354/97 10-05 VIO Insufficient corrective actions

50 354/97-10-09 VIO Failure to establish appropriate procedure.

? ,

l

Opened / Closed

50 354/97-10-01 NCV Failure to implement administrative procedure

'50 354/97-10-06 NCV Improperly performed surveillances for determining

specific gravity of the 125 Vdc and 250 Vdc batteries

50 354/97-10-07 NCV Failure to perform secondary containment isolation

actuation system surveillances.

50-354/97-10-08 NCV Removal of a RHR shutdown cooling suction line

snubber.

50-354/97 10 10 NCV Failure to establish appropriate procedure

50 354/97-10 11 NCV Failure to perform appropriate testing

50-354/97-10 12 NCV Failure to operate and maintain SACS in accordance

with design and licensing basis

50-354/97-10-13 NCV Failurs to perform an adequate search of an individual.

C10D.d

50 354/95 11 P21 Deficiency in Anchor Darling swing check valves.

50 354/96-07 01 VIO Failure to evaluate changes made to the facility in

accordance with 10 CFR 50.59.

50 354/E96-125 01013 VIO Failure to plan appropriate post maintenance TS

surveillance testing.

50 354/E96-125 01023 _VID

_

Failure to conduct appropriate post maintenance TS- -

surveillance testing.

50 354/E96-125-02013 VIO Failure to promptly identify and correct a condition

adverse to quality.

50 354/E96-125 02023. VIO Failure to promptly identify and correct a condition

adverse to quality.

,

.

.

23

50 354/E96125 03013 VIO Failure to obtain NRC Approval before making a change

to the facility.

50 354/97 04-02

'

VIO HPCI system feed water injection line inoperable.

50-354/97-07-01 VIO Failure to promptly identify inoperable electric fire water

pump

!

50 354/96 22 LER Condition alone that could have prevented removal of

residual heat - safety auxillaries system deficiencies.

50 354/97 27 LER Operation In a TS prohibited condition due to improperly

performed surveiilances for determining specific gravity

of the 125 Vdc and 250 Vdc batteries.

50-354/97 28 LER TS prohibited condition failure to perform secondary

containment isolation actuation system surveillances.

50 354/97 29 LER Operation in a TS prohibited condition due to removal of

a RHR shutduwn cooling suction line snubber.

50-354/97 30 LER Failed IST surveillance for reactor building to torus

vacuum breakers,

50 354/97 32 LER TS required shut down due to concurrent inoperability

of the HPCI and RCICs

- _ _ - _ _ _ _ _ _ _ _ _ - _ _ - ____- -.

.

\

.

24

LIST OF ACRONYMS USED

nOV Air Operated Isolation Valve

APRM Average Power Range Monitor

DCP Design Change Package

EDG Emergency Diesel Generator

EPA Electrical Protection Assemblies

HPCI High Pressure Coolant injection

IA Instrument Air

IST Inservice Test

LER Licensee Event Report

M&TE Measuring and Test Equipment

NRC Nuclear Regulatory Commission

PClG Primary Containment Instrument Gas

PDR Public Document Room

PRNMS Power Range Neutron Monitoring System

, PSE&G Public Service Electric and Gas

l QA Quality Assurance

l

'

RBTVB Reactor Building to Torus Vacuum Breaker

RCIC Reactor Core isolation Cooling

RG Regulatory Guide

RHR Residual Heat Removal

RPS Reactor Protection System

RWCU Reactor Water Cleanup

SACS Safety Auxillaries Cooling System

SRV- Safety Relief Valve

SSW Station Service Water

TS Technical Specification

UHS Ultimate Heat Sink

__ __

%