IR 05000272/1988013

Insp Repts 50-272/88-13 & 50-311/88-13 on 880503-0606. Violations Noted.Major Areas Inspected:Followup on Outstanding Insp Items,Operational Safety Verification, Maint,Surveillance & ESF Walkdown

ML18093A957
Person / Time
Site:	Salem
Issue date:	07/06/1988
From:	Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML18093A954	List: IR 05000272/1988013 ML18093A952
References
50-272-88-13, 50-311-88-13, NUDOCS 8807200229
Download: ML18093A957 (12)

Text

Report No License No Licensee:

U. S. NUCLEAR REGULATORY COMMISSION 50-272/88-13 50-311/88-13 DPR-70 DPR-75

REGION I

Public Service Electric and Gas Company P. 0. Box 236 Hancocks Bridge, New Jersey 08038 Facility Name:

Salem Nuclear Generating Station - Units 1 and 2 Inspection At:

Hancocks Bridge, New Jersey Inspection Conducted:

May 3, 1988 - June 6, 1988 Inspectors:

Approved by:

P. D. s\\'./etiand, Chief, Reactor Proj9cts Section No. 2B, Projects Branch No. 2, DRP Inspection Summary:

7-C- ?'?

date Inspections on May 3, 1988 - June 6, 1988 (Combined Report Numbers 50-272/88-13 and 50-311/88-13)

Areas Inspected:

Routine inspections of plant operations including:

followup on outstanding inspection items, operational safety verification, maintenance, surveillance, engineered safety feature walkdown, and review of licensee event report Results:

The failure to adequately complete a technical specification required surveillance test is cited in this inspection report (paragraph 7).

Although licensee identified, this violation 1 s similarity to a previous violation indi-cates that continued emphasis is necessary in the surveillance test are NRC review of the licensee identified reactor protection system calibration inadeq~acies was completed during this report period (paragraph 7).

Based upon the low safety significance and the licensee's program for comprehensive dynamic testing of reator protection functions, this deficiency was classified as a licensee identified violation.

8807200229 88070~

PDR ADOCK 05000272 Q

PDC

DETAILS Persons Contacted Within this report period, interviews and discussions were conducted with members of licensee management and staff as necessary to support inspec-tion activit.

Followup on Outstanding Inspection Items (92701, 92703)

(Closed)

Unresolved Item 50-272/82-34-01; Discrepancies in recording filler metal type and proper heat numbe Inspection report 50-272/82-34 states that this item was resolved prior to issuance of the inspection repor No further inspection is required and this item is close (Closed)

In~pector Follow Item 50-272/84-09-01 and 311/84-09-01; QA audits of staff performanc This item is ciosed based on

-

NRC QA inspection (87-31/87-32) and non-licensed staff train-ing inspection (87-10/87-13) in which no related concerns were identifie *

(Closed)

Unrsolved Item 50-311/84-45-02; Visual examination of painted surface The inspector reviewed Deviation 1 to Procedure SWRI-NDT-900-1/50 which delineates when a VT examinatiJn over paint is allowabl This item is close (Closed)

Unresolved Item 50-311/84-45-03; NOE Qualification record The inspector reviewed supplemental certification data and had no further questions at this time'.

This item is close *(closed)

IE Compliance Bulletin No. 86-03 (50-272/86-BU-03; 50-311/86-BU-03); Minimum flow recirculation line failures 0f ECCS pump The inspector reviewed licensee responses dated November 20, 1986 and January 15, 1987, in which the licensee concluded that the intent of GDC 35 is satisfied for the Safety Injection, Residual Heat Removal, and Charging/Safety Injection System The inspector has no further questions at this tim This bulletin is closed for Salem Units 1 and (Closed)

Inspector Follow Item 50-272 and 50-311/86-02-05; Gas decay tank sample calibration standar The inspector verified that a gas calibration standard (in styrofoam media rather than gel)

is used to calibrate the Johnson bomb geometry for gas decay tank

.*

(Closed)

(Closed)

(Closed)

sample analysi The inspector also reviewed subsequent intercomparison results delineated in NRC combined Inspection Report Noi. 272/88-10 and 311/88-1 This item is close Inspector Follow Item 50-311/86-03-01; Comparison of licensee analytical results to BNL results of water sample This item is closed based on analytical measurement data comparisons reported in NRC combined inspection reports 272/87-33; 311/87-34 and 272/88-11; 311/88-1 Inspector Follow Item 50-272/86-06-02; Engineering Evaluation for fire.door No. 121-The evaluation of auxiliary building ventilation balance is an ongoing long term actio To ensure operability of fire door 121-1 until the ventilation problems are resolved, a new door has been installed and fire protection verifies the integrity of the door once per shif This item is close Unresolved Item 50-272/86-11-03; Thimble tube wall thinnin Thimble tubes were replaced during the seventh refueling outage per design change package lEC-223 This item is close (Closed)

Inspector Follow Item 50-272/86-19-01; Heat shrinkable tubing test result Raychem test results were provided to the licensee on Oct6ber 16, 1986 and will be reviewed by NRC per TI 2500/17 (IEN 86-53).

This item is closed.*

(Closed)

Inspector Follow Item 50-311/86-33-01; Computer RWP form cluttere The inspector reviewed radiation protection procedure RP 202 "Radiation \\iJork Permits", Revision 1 dated February 19, 1988, and se.*::!ral radiation work permits in effec The inspector concludes that the RWP forms in use are satisfactorily organized to*facilitate worker comprehensio This item is close (Closed)

Unresolved Item 50-272/87-19-01 and 50-311/87-21-01; Data collection concerns for pump and valve testin The inspector reviewed the Operations Newsletter dated April 8, 1988, in which the operators were directed to document and maintain all data resulting from pump and valve surveillance tests regardless of the acceptability of the data and reasons therefor The inspectors will continue to monitor lic~nsee *

actions in this regard during routine inspection This item is closed.

4 Operational Safety Verification (71707, 71709, 71881) Inspection Activities On a daily basis throughout the report period, inspections were conducted to verify that the facility was operated safely and in conformance with regulatory requirement The licensee's management control system was evaluated by direct observation of activities, tours of the facility, interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions for operation, and review of facility record The licensee's compliance with the radio-logical protection and security programs was also verified on a periodic basis.* These inspection activities were conducted in accordance with NRC inspection procedures 71707, 71709 and 71881 and included weekend and backshift inspection.2 Inspection Findings and Significant Plant Events (93701)

3.2.l Unit 1 Unit 1 operated at 100% power throughout the inspection perio.2.2 Unit 2 Unit 2 began the report period*operating at 100% powe Based upon the continuing Salem electrical distribution system design review, the licensee concluded on May 9, 1988, that 37 electrical circuits did not have adequate containment pene-tration conductor overcurrent backup protectio Of the 96 circuits specified in Uni*: 2*Technicai Specification (T.S.)

3.8.3.1, 37 circuits had marginal or unacceptable coordination between the backup circuit breaker and the conax connector at the containment penetratio Under certain circumstances, the rating of the conax connector could have been reached prior to the backup breaker tripping open if a circuit fault were to develop and the primary breaker failed to tri In all cases, the primary circuit breakers (located between the backup breaker and the conax connector) had proper coordination and were operabl The purpose of the primary and backup contain-ment penetration conductor overcurrent protective devices is to ensure that containment electrical penetrations and containment integrity will not be adversely affected by a circuit failure inside of containmen The licensee entered T.S. action statement 3.8.3.la which required the affected circuits to be de-energized by tripping the backup circuit breakers within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> A review of the operational impact on completing the action statement for each of the 37 circuits i dent ifi ed 2 concerns. _First, ther.e were_ 5

circuits (21 and 22 containment sump pumps, RHR valves 21SJ44 and 22SJ44, and one reactor coolant drain tank pump) that were

• essential to continued plant operation Secondly, because the backup breakers are frequently main feeders to load centers, complying with the action statement would have had an adverse affect throughout the plan The first issue was resolved by implementing major design change 2SC-2001 "Backup Protection for Electrical Penetrations 11 which installed in-line molded case circuit breakers with acceptable trip ratings in each of the five circuit The inspector found this design change and related engineering evaluation (S-1-ZZ-XX-EEE-240-0 -

11 Penetration Circuit Protection Analysis") to be acceptabl To address the second issue, the licensee submit-ted an emergency Technical Specification change request on May 10, 198 This submittal proposed that with one or more of the containment penetration conductor overcurrent protective device~ inoperable, the affected circuit be de-energized by tripping 11either 11 the primary or backup protective devic The submittal reasoned that opening either protective device will completely de-energize that portion of the circuit passing through the containment penetratio After review of the licensee's submittal and conference calls on May 10, and May 12, between the licensee, NRR, and Region I, a temporar waiver of compliance to change action statement T.S. 3.8.3.la was issued by NRR on May 12, 198 The licensee verified com-pliance with the new action statement and initiated a weekly check of open primary breakers that same da In the request for amendment, dated May 10, 1988, the licensee committed to repair all of the remaining circuits prior to startup from the September 1988 refueling outag (IFI 311/88-13-01)

Salem Unit 1 does not have T.S. requirements on electrical containment penetration protection, however this issue will be reviewed on Unit 1 after the Unit 2 review is complete (IFI 272/88-13-01)

On May 13, 1988, the Salem Unit 2 control room operator was inserting control rods (Bank D) for a* turbine valve test load reductio At approximately 213 steps, the reactor tripped on power range high negative neutron flux rat All plant systems responded to the trip as designe Safety parameter display

• system (SPDS) data indicated that control rod 103 dropped during rod insertio Investigation and testing of the rod control system by the licensee did not reveal any equipment deficiencies that could have caused the tri However, a 16 volt power supply in the rod control logic cabinet and a power cabinet alarm circuitry card were replaced as precautionary measure The inspector reviewed the SPDS data, the licensee's test data and results, and Licensee Event Report 88-009-00 and agrees with the licensee's conclusions that control rod 103 dropped during rod insertion and that the test results are inconclusive as to the cause of ~he. dropped rod and subsequent tri The reactor ~as

..

. ' *

3. *

returned to criticality on May 15, 1988; The unit remained at 100% power throughout the rest of the report perio No further problems with the rod control sy~tem were experience Both Units On May 19, 1988, the licensee identified non-seismically qualified. pneumatic-electric (PE) relays associated with the emergency diesel generator (EOG) fire protection equipmen~. The licensee notified the NRC of the deficiencies via the Emergency Notification System (ENS) and informed the resident inspector The deficiencies were identified during the detailed design

~hase for a modification required for previously identified* fire protection system problem The mercury contacts associated with the PE relays interface with Class lE circuits for the fans in the diesel generator are In a seismic event the mer~oid relays in question could chatter and prevent the EOG room exhaust fans from operatin Diesel generator operability.is contingent upon diesel generator area ventilation to cool the generato Previous analyses indicate the EDG's can remain operable without ventilation for at least 20 minute SORC reviewed the deficiency and concluded that diesel and diesel ventilation operability was ensured by the fact that operations and/or fire protection operators were available to reset the PE relay Fire protection personnel routinely perform a surveillance procedure which delineates, in part, how to reset the relay SORC's immediate short term corrective actions directed that specific relay resetting instructions be provided for the operations departmen The inspector reviewed the implementation of the SORC directiv Operations supervision was given a copy of the fire protection surveillance procedure and told which steps to follow to reset the rela Night shift operators received training (on resetting the relay) prior to assuming their watch, but day shift operators did not receive training until several* hours into theit shif In addition, operations supervision was unclear as to whether operators or fire protection personnel had primary responsibility for resetting the relay in a seismic even The inspector requested that the li_censee demonstrate that the relay could be reset by an operator within the 20 minute time frame, using the fire protection surveillance procedure and prior to formal trainin This was accomplished successfull However, the inspector was concerned with the informality of implementation of the corrective actions intended by SOR It appears that improvements are warranted in this area to ensure accurate and timely implementation of immediate short term corrective

.*

action This issue will be reveiwed further during subsequent routine inspection Design Change Package (OCP) ISC/2SC-1609 11Modification of the Fire Protection System for the Diesel Generator and Control Areas to Comply 11'/ith 10CFR50 Appendix R

will eliminate the PE relay seismic concern These DCP 1 s are

. scheduled for installatjon during June, 198 No violations were identifie.

Maintenance Observations (62703)

The inspector reviewed the following safety related maintenance activities to verify that the activities were conducted in accordance with approved procedures, Technical Specifications, NRC regulations, and industry codes and standard Work Order Number 880509128 Description Troubleshoot No. 21 Waste Gas Compressor controls and valves The inspector observed poor radiological controls and housekeeping demonstrated by workers performing this work orde Tools, cement chips, and used anti-contamination clothing were strewn across the contaminated area boundar The inspectors concerns were brought to the attention of radiation protection supervision and correcte Work Order 870903044 Various Design Change Package 2SM0362 2SC-2001 Description

>~ sta ii at ion of Thermo coup 1 es (4) on 23 Auxiliary Feedwater Pump discharge lines for backleakage monitorin Installation of backup protection (breakers) for containment electrical penetration (21 and 22SJ44 1 s, 22 reactor coolant drain tank pump, 21 and 22 containment sump pumps).

The inspector witnessed portions of installation and testing associated with these DCP 1 s and found the activities to be acceptabl No vfolations were identified.

.

.

~~-

.. ;:=**-*=-*=~*>.'~*****,,'?,',:*r:~";;:...'\\t'f4!'... ~~**.t.,4'!*l.L!S.p::.. ~'C,.:-*:';,'\\'!!°"'-~,-,*-..~.. ~

1.C.\\.te{*~*'!'.*.""'"*',i~**.-*.M..

,_i.l*i:i>*y:c,~.~'.':.-:_*.*

~

~--

1 '-'**-**'"'~~:d§_.. <.* *

... 1 ** ~J -. --~~-~-~*

~~~-.. *:-.;-::-_.-~~~

• • ~.

~-"~

• 8 Surveillance Observations (61726) Inspection Activity During this inspection period, the inspector performed detailed technical procedur~ reviews, witnessed in-progress surveillance testing, and reviewed completed surveillance package The inspec-tor verified that the surveillances were performed in accordance with Technical Specifications, licensee approved procedures, and NRC regulation These inspection activities were conducted in accordance with NRC inspection procedure 6172 The following surveillance tests were reviewed, with portions witnessed by the inspector:

2PD-16.2.006 PI/S-AF-3 OP-TEMP-8805-2 M2B 2IC-2.6.020 Source Range At Power Channel Functional Test Auxiliary Feed \\*later Backleakage Stroke Test Valve 22SJ44 following installation of backup breake Fuel Handling Crane Periodic Inspection and Operational Tests-Pressurizer Level Transmitter 2LT-459 Functional Test The inspector concluded that these surveillance tests were properly conducte No violations were identifie.

Engineered Safety Feature (ESF) System Walkdown (71710) Inspection Activity The inspectors independently verified the operability of selected ESF systems by performing a walkdown of accessible portions of the system to confirm that system lin~up procedures match plant dr~wings and the as-built configuratio The ESF system walkdown was also condµcted to identify equipment conditions that might degrade performance, to determine that instrumentation is calibrated and functioning, and to verify that valves are properly positioned and locked as appropriat This inspection was conducted in accordance with NRC inspection

'. *

. *'...

procedure 7171 The Units 1 and 2 Intermediate Head Safety In ec-tion (SI) Systems were inspecte The inspector noted boric ac d crystals built up around the stems of SI pump mini-flow isolation valves lSJ67, 2SJ67 and 2SJ6 These conditions were brought to the attention of operations shift supervision who directed the valve stems to be cleaned of These valves are normally open with the power to the Limitorque motor operator locked ou The valves must be closed before the SI pumps can be fed from the residual heat removal system (RHR) during the recirculation phases of emergency core coolin The inspector reviewed the previous inservice testing procedure 4.0.5 V-SJ-5 completed for each valve (done in Mode 5 prior to startup) and verified* that the valves stroked within the required time (10 seconds).

Overall system conditions were found to be acceptabl No violations were identifie.

Review of Licensee Reoorts (90712, 90713, 92700)

Upon receipt, the inspector reviewed licensee event reports (LERs) as well as other periodic and special reports submitted by the license The reports were reviewed fer accuracy and timely submissio Additional followup performed at the discretion of the inspector to verify corrective action implementation and adequacy is detailed with the applicable report summar The following reports were received and revie~ed during the *

inspection period:

Unit 1 Monthly Operating Report - April, 1988 Li.nit 2 Monthly Operating Report - April, 1988 Unit 1 Special Report 88-2 Fire Barriers Impaired For Greater

• Than Seven (7).Days On April 29, 1988, as part of an 18 month surveillance and in con-junction with a licensee initiated Penetration Seal Review Program, Fifty-two (52) fire barrier electrical and mechanical penetrations (of approximately 6000 totql penetrations) were found missing or degrade Hourly fire watches had already been established in the applicable areas for previous fire protection deficiencie PSE&G letter NLR-N88037 dated March 4, 1988 to the NRC delineates the licensee's schedule for completion of the Penetration Seal Review Program including repair of degraded seal The inspector had no further questions at this time.

• • *

Unit 1 LER Supplement 87-018-01 Improperly Calibrated Lead/Lag and Derivative Amplifiers This event was discussed in combined NRC Inspection Report 50-272/

311/87-3 The licensee identified an error in calibrating the dynamic response (lead/lag and derivative functions) of certain reactor protection circuit The licensee concluded that the miscalibration of lead/lag circuits resulted in conservative tripping of the RPS circuits such as low steam line pressure and low pressurizer pressur Further evaluation was necessary to determine the effect of the improperly calibrated derivative amplifier On May 25, 1988, the licensee supplemented the original LER concluding that*the overpower delta-T (OPdT) trip function would have tripped non-conservatively only when reactor coolant system (RCS)

temperature was increasing rapidl The allowable trip setpoint would not be exceeded unless the RCS temperature rose at a rate greater than 9 degrees F per minut Further, the licensee noted that OPdT protection is a backup for the high neutron flux protection function and it also limits the required range for overtemperature delta T protectio These functions are not taken credit for in the Salem accident analyse NRC review of this event recognized the inoperability of the OPdT function under certain circumstance However, the licensee identified the problem and promptly corrected and reported the conditio No further corrective actions were identifie In consideration of the low safety significance as delineated above and to encourage further improvement in dynamic testing of RPS functions, NRC determined that no violation would be cited in accordance with lOCFR 2, Appendix C (272/311/88-13-02).

Unit 1 LER Supplement 88-001-0~

Diesel Generator Day Tanks Seismic Deficiency This event was discussed in combined inspection 272/311/88-0 Licensee investigation has attributed the root cause of the lack of anchoring of the Unit 1 day tanks to inadequate design and design review, in that anchoring was not specified in historical design record PSE&G civil drawings have been updated to address anchor requirement The inspector had no further questions at this tim Unit 2 LER 88-006-00 Reactor Trip/False No. 23 Reactor Coolant Loop Low Flow Signal This event was discussed in combined inspection 272/311/88-11 and the inspector had no further questions following review of the LE..

Unit 2 LER 88-007-00

Reactor Trip Resulting From Faulty Turbine Electro-hydraulic Controls (EHC) Response This event was reviewed in combined inspection 272/311/88-1 The inspector had no further questions following review of this LE Unit 2 LER 88-008-00 Fire Protection Containment Isolation Valve Missed Surveillance On April 26, 1988, the licensee identified that valve 2FP147 had not been surveilled within the Technical Specification required interval of every 92 day This valve is normally closed and fails close It is manually opened when deluge flow to the reactor coolant pump is required for fire suppressio The 2FP147 valve was subsequently tested satisfactorily. At the time the surveillan~e was scheduled, the licensee decided to postpone the surveillance on 2FP147 and do it later in the procedure along with check valve 2FP148 since test-ing of both valves requires manual isolation of the containment fire protection header to preclude inadvertent deluge of the reactor coolant pump When 2FP148 was tested and since the test data for the two valves is recorded on different checkoff sheets, testing on 2FP147 was forgotte Supervisory review of the com-pleted surveillance procedure failed to identify the mi_ssing dat The applicable surveillance ~rocedure SP(0)4:o.s V-MISC has been revised to perform surveillance of 2FP147 and 2FP148 at the same tim Corrective disciplinary action has been taken by the licensee for those personnel involved and the need to maintain attention to detail was reemphasized with operations personne The inspector has determined that this event is similar in nature to one for which a violation was previously cited (311/87-18-01)

in that personnel performed and management reviewed the completed procedure, failing to identify the i ncomp fote checkoff sheet and resulting in a portion of the surveillance not being accomplished within the required time.* This represents i licensee identi-fied violation of Technical Specifications 4.0.5 and 6.8.l (311/88-13-03)

Since appropriate corrective actions have already been taken by the licensee, no further action is require The following events are discussed in Section 3.2 of this report:

Unit 1 LER 88-010-00 Unit 2 LER 88-009-00 Diesel Generator Cardox PE Relay Not Seismically Qualified-Potential Loss of DIG Ventilation Reactor Trip/Dropped Control Rod

• 12 Exit Interview (30703)

The inspectors met with Mr. J. Zupko and other licensee personnel periodically and at the end of the inspection report to summarize the scope and findings of their inspection activitie Based on Region I review and discussions with the licensee, it was determined that this report does not contain information subject to 10 CFR 2 restriction *