IR 05000354/1998006
| ML20236T162 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/21/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236T152 | List: |
| References | |
| 50-354-98-06, 50-354-98-6, NUDOCS 9807270469 | |
| Download: ML20236T162 (25) | |
Text
-
l
.
.
.
.
.
U. S. NUCLEAR REGULATORY COMMISSION REGION 1 Docket No:
50-354 License Nos:
NPF-57 Report No.
50-354/98-06
Licensee:
Public Service Electric and Gas Company i
i Facility:
Hope Creek Nuclear Generating Station Location:
P.O. Box 236 Hancocks Bridge, New Jersey 08038 Dates:
May 17,1998 - June 27,1998 Inspectors:
S. M. Pindale, Senior Resident inspector J. D. Orr, Resident inspector
,
L.' A. Peluso, Radiation Physicist
!
Approved by:
James C. Linville, Chief, Projects Branch 3 Division of Reactor Projects 9807270469 990721
1 PDR ADOCK 05000354 1[
'
G PDR
-
.
.
-
.___m_m__________ _ _ _ _.. _ _ _ _ _ _.
_ _ _ _ _ _ _ _ _. _ _ _
.
.
.
.
EXECUTIVE SUMMARY
.
Hope Creek Generating Station NRC Inspection Report 50-354/98-06 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a six-week period of resident inspection. In addition, it includes the results of an announced inspection by a regional inspector who reviewed the Hope Creek Radiological Environmental Monitoring and Meteorological Monitoring Programs.
Operations I
Operators exhibited good performance during routine and off normal events with one notable exception, where non-licensed equipment operators failed to explicitly follow a procedure and subsequently perform a correct second verification. Specifically, the 'A'
station service water pump was improperly racked in when the normal and emergency trip coil fuses were left in the wrong position, and the second verifier checking the breaker i
lineup failed to identify that the fuse blocks were not properly oriented. (Section 01.1)
During frequent routine plant tours, the inspectors identified some instances of minor material condition and housekeeping deficiencies. Examples included excessive noise (elevated vibration) on an emergency diesel generator jacket water keep warm pump, burnt out indicating light bulbs, and immovable louvers on service water intake structure ventilation doors. Although equipment was not rendered inoperable by these items, the number of deficiencies identified by the inspectors indicated a need for increase worker, supervisor and manager focus during the conduct of routine plant tours. (Section 01.2)
Hope Creek operators properly initiated an operability determination for the excessive cycling of the 'D' torus to drywell vacuum breaker. The engineering department provided a thorough evaluation of the problem and it was consistent with the operator's initial operability determination. (Section O2.1)
)
There were two examples (inoperable indicating circuit for control room chiller circulating pump at the Remote Shutdown Panel, and abnormal response from an untabeled i
,
emergency diesel generator over-voltage relay during post-maintenance testing) where i
operators were slow to fully document and process the bases for equipment operability when degraded indications were apparent. In both instances, post-issue reviews
,
demonstrated that operability was not chat'enged. (Section O2.2)
l A quality assessment audit of corrective action effectiveness at Salem and Hope Creek was comprehensive in scope and content. Although the audit findings reflected a probing
>
review of the effectiveness of corrective actions, the nature of these findings were similar l
from prior PSE&G and NRC corrective action system reviews, and were indicative of continued weaknesses in implementing the established corrective action program. These
!
findings warrant increased management attention for effective and lasting resolution.
(Section 07.1)
ii l
L__________.___________________________._._________
l
.
.
.
Mei ltenance l
The preparations to replace three single cells on safety-related batteries were thorough and well developed. The battery cell replacements were completed in a timely fashion and without error. PSE&G took initial steps to improve its performance monitoring of the safety-related batteries, which was a recent performance weakness, as the batteries approach their end of service life. (Section M1.1)
Due to poor worker practice, scaffold was erected too close to a non-safety related electrical panel, which caused a vibration induced actuation of a low pressure feedwater heater isolation relay. This actuation resulted in an unnecessary challenge to control room operators and plant equipment. (Section M1.2)
Operators appropriately researched and thoroughly understood the risk assessment associated with concurrent unavailabilities of the 'B' reactor feed pump and the reactor core isolation cooling system. However, operators did not consult the probabilistic safety assessment group to quantify or confirm their assessment prior to placing the plant in a configuration that may have impacted overall plant safety. (Section M4.1)
Eglineerina Sntem engineering did not effectively monitor and trend particulate sample results for the emergency diesel generator (EDG) fuel oil corage tanks. As a result, increasing particulate concentrations were not questioned in a timely fashion, and two of the eight tanks subsequently were out of specification and the associated EDG was rendered inoperable.
Also, chemistry personnel failed to properly test and report particulate concentration results for new fuel oil received at the station via tanker trucks since October 1997, which was contrary to station procedures. PSE&G responded effectively to the degraded condition of tH tanks, and implemented actions to quickly restore the tanks to an operable condition consstent with a Notice of Enforcement Discretion, which the HRC granted on May 22,1998. (Section E2.1)
Contract engineering personnel failed to ensure proper design and configuration control for a reactor core isolation ev> ling (RCIC) system modification during the Fall / Winter 1997 refueling outage. Specifically, the modification incorrectly changed the RCIC turbine steam line stop valve logic by deleting it's sequenced opening feature. As a consequence, the RCIC turbine oversped rapidly during a post-maintenance RCIC overspeed test (RCIC pump and turbine were uncoupled). This condition did not render the RCIC system inoperable during the specific valve configuration because testing demonstrated operability of the system while in this condition. Station engineering responded appropriately to this design and testing error. (Section E2.2)
Plant Support Due to a chsmistry technician error, insufficient technician proficiency training, and ineffective supervisory oversight, unsatisfactory results were reached during the performance of a boron analysis for the standby liquid control si/ stem sodium pentaborate l
solution. PSE&G promptly confirmed via verification :emple av:ayses that the actual iii
'
'
s
.
.
,
<
.
.
sodium pentaborate concentration did not fall below technical specification allowable
-
values (the initial errant analysis yielded an inaccurate and out of specification result).
(Section R4.1)
!
iv
-
-_
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
.
.
TABLE OF CONTENTS
-
EX EC UTIVE S UM M ARY.............................................. ii TA BLE O F C O NTENTS............................................... v 1. O p e ra tion s...................................................... 1
.
Conduct of Operations.................................... 1 01.1 General Observations................................ 1 01.2 Multiple Minor Equipment Deficiencies.................... 2 O2 Operational Status of Facilities and Equipment................... 3 O2.1
'D' Torus to Drywell Vacuum Breaker Cycling............... 3 O2.2 Operability Conclusions Associated with identified Deficiencies.. 5
Quality Assurance in Operations............................. 6 07.1 Quality Assessment Corrective Action Audit................ 6 ll. M ai nt e n a nc e................................................... 7 M1 Conduct of M aintenance................................... 7 M1.1 Class 1E 125Vdc Battery Single Cell Changeouts............ 7 M 1.2 Feedwater Heater Trip............................... 8 M4 Maintenance Staff Knowledge and Performance.................. 9 M4.1 Safety impact of Concurrent Reactor Core isolation Cooling and 'B'
Reactor Feed Pump System Outages..................... 9 lil. Engine ering................................................... 10 E2 Engineering Support of Facilities and Equipment................. 10 E2.1 Emergency Diesel Generator Fuel Oil Contamination......... 10 E2.2 Reactor Core Isolation Cooling System Steam Turbine Overspeed Due to Design Error
...................................13 E8 Miscellaneous Engineering issues............................ 16 E8.1 (Closed) Licensee Event Report 50-3 54/9 8-04.............. 16 I V. Pl a nt S u p p ort................................................. 1 6 R1 Radiological Protection and Chemistry (RP&C) Controls............ 16 R4 Staff Knowledge and Performance in RP&C.................... 16 R4.1 Standby Liquid Control System Boron Sampling Problems...... 16 V. Management Meeting s........................................... 18 X1 Exit Meeting Summ ary................................... 18 f
t v
i I
L J
- - - _ _
.
.
Reoort Details
-
i Summarv of Plant Status
Hope Creek was operated at or near full power for the duration of the inspection period.
1. Operations
Conduct of Operations 01.1 General Observations
)
a.
Insoection Scooe (71707)
i Throughout the report period, the inspectors observed operator performance in the areas of routine activities and surveillance, technical specification (TS) application and interpretation, and response to an abnormal condition from the fuel pool seal gates, b.
Observations and Findinas On June 1,1998, the inspectors observed the performance of a routine monthly surveillance test on the 'A' emergency diesel generator (EDG). The surveillance test was being performed by a reactor operator (RO) license candidate under the direct supervision of a licensed reactor operator. The trainee performing the task was knowledgeable of the assigned task and he demonstrated the skills and operations department standards expected of licensed ROs. The trainee was closely supervised by the licensed RO. The inspectors also observed the 'A' EDG operate locally. The inspectors noted that a non-licensed equipment operator thoroughly
walked down the EDG while it operated.
On June 1,1998, the inspectors observed the control room operators respond to an unexpected control room annunciator for high leakage past the fuel pool gate seals.
The control room supervisor promptly exhibited command and control and he directed actions in accordance with the alarm procedure. The equipment operator was responsive and completed the actions in the field. The operators thoroughly completed all the necessary actions for tha a! arm and verified that the alarm was not due to any leakage past the fuel pool gate seals.
i On June 4,1998, control room operators started the 'A' station service water (SSW) pump to perform post maintenance testing after a scheduled pump outage.
The control room operators immediately noticed that a 125Vdc overhead annunciator and 'A' SSW pump inoperable bezel light illuminated concurrent with the pump start. An equipment operator was dispatched to the pump 4.16kV breaker. The equipment operator discovered that the normal und emergency 125Vdc trip coil fuses were in the OFF position. The equipment operator was directed to locally trip the 'A' SSW pump at its 4.16kV breaker. PSE&G, in its investigation of the problem, determined that the equipment operator that performed the breaker restoration failed to follow Hope Creek operating procedure, HC.OP-
!
l l
_ _ _ _ _
_ _ _
_ _ _ _ _
____
_-
.
.
SO.PB-0001(Q),4.16kVSystem Operation, in that the normal and emergency trip fuses were not placed in the ON position as explicitly described in the procedure.
The second verifier for the tagout restoration had also failed to identify that the fuses were not positioned to the ON position. PSE&G determined that in addition to the lack of attention to detail and failure to follow procedures, the individuals involved were unfamiliar with fuse arrangements in the particular 4.16kV class 1E breaker. The inspectors determined that the control room operators were prompt in resolving the unusual indications after the 'A' SSW pump start and that PSE&G had adequately understood the problem before the 'A' SSW pump was restarted again for post maintenance testing. The inspectors also determined that PSE&G management was thorough in its corrective actions and proposed corrective actions to prevent problem recurrence. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Poliev. (NCV 50 354/98-06-03)
On June 15,1998, the inspectors locally observed the start of a routine monthly surveillance test on the 'B' EDG. The inspectors questioned the non-licensed equipment operator about some of the indications and alarms that were received shortly after the EDG start.' The inspectors determined that the equipment operator
.was knowledgeab!e about the expected alarms during the EDG start and that his -
~ follow-up actions to all the alarms were appropriate..
c.
Conclusions
. Operators exhibited good performance during routine and off normal events with one notable exception, where non-licensed equipment operators failed to explicitly follow a procedure and subsequently performed an incorrect second verification.
Specifically, the 'A' SSW pump was improperly racked in when the normal and L
emergency trip coil fuses were left in the wrong position, and the second verifier l
checking the breaker lineup failed to identify that the fuse blocks were not properly oriented.
01.2 Multiole Minor Eauioment Deficiencies a.
Insoection Scoos (71707)
Throughout the inspection period, the inspectors conducted frequent tours of the facility, including the' reactor building, auxiliary building service water intake structure and the fire pump house.
b.
Observations and Findinas p
'
During tours, the inspectors identified several minor items which warranted supervisory or management attention. These included 1) certain safety parameter display system (SPDS) meteorological data indicated flashing white question marks, 2) louvers on doors at the service water intake structure could not be moved, but signs on the doors provide instructions on how and when to operate them,3)
reading material at the auxiliery boiler building was inappropriate,4) protective
j
.
-
.-
.
.
..
clothing was placed inside a plant paging system intercom, 5) there were several burnt out light bulbs on equipment indication panels, and 6) a torque wrench had apparently dropped in a cable box in the 'C' core spray pump room.
The inspectors also identified two additional items. An office chair was found in a s
- dimly lit area of the auxiliary building on 77' elevation. PSE&G response to this item was thorough and subsequently identified this to be an inappropriate break location for security personnel during routine rounds. There was no indication that required rounds or measures were missed. In response, PSE&G stressed the need for station management to identify, question and pursue these types of abnormal or suspicious conditions in the plant.
The inspectors also informed an operations supervisor of a loud rumbling noise emanating from the 'D' emergency diesel generator Jacket water keep warm pump / motor.' in response, vibration readings were taken, which were found to have been elevated on the pump, but within acceptable values. Operations initiated an
. Action Request (980614071)to evaluate the condition. Operators vented the Jacket water keep warm system, wh.ch improved the vibration and noise levels.
PSE&G initiated a work order to further investigate this pump during a subsequent system outage. The inspectors determined that the evaluation was appropriate, and emergency diesel generator operability was not adversely affected.
The inspectors informed station management that the relatively high number of
~
minor deficiencies reflected a lack of a questioning attitude by site workers, supervisors and management during their routine plant tours. Similar minor material condition and housekeeping weaknesses were identified in NRC inspection 50-354/98-05 (Section 02.1).
c.
Conclusions During frequent routine plant tours, the inspectors identified several instances of minor material condition and housekeeping deficiencies. Although equipment was not rendered inoperable by these items, the number of deficiencies identified by the inspectors indicated a need for increase worker, supervisor and manager focus during the conduct of routine plant tours.
O2 Operational Status of Facilities and Equipment l
'
02.1
'D' Torus to Drvwell Vacuum Breaker Cvelina I
a.
Insoection Scone (71707)
L The inspectors reviewed PSE&G's follow-up actions and operability determination l'
- (OD) regarding frequent cycling of the 'D' torus to drywell vacuum breaker.
!
,
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _. _ _ _. _ _ _ _ _ _ _ _ _ _
]
.
.
\\
-
\\
.
-
b.
Observations and Findinas On May 16,1998, and also on June 10,1998, the 'D' torus to drywell vacuum breaker briefly indicated an intermediate-open position. In one instance, only one of the two position switches indicated intermediate-open. Control room operators reviewed appropriate plant parameters and suspected that the 'D' torus to drywell vacuum breaker had come off its closed seat in response to a very slight, about 0.02-0.03 psid, differential pressure between the torus and drywell. Hope Creek typically operates near at zero differential pressure between the torus and drywell, I
but normal nitrogen makeup to the torus by primary containment instrument gas (PCIG) compressor operation may establish a slight negative differential pressure between the drywell and torus air spaces. The dual position indicators are designed to indicate intermediate-open when the vacuum breaker is no more than 0.010 inches open at valve centerline. The inspectors verified that PSE&G had recorded on October 7,1997, an acceptable closed limit switch setting for the 'D' torus to drywell vacuum breaker.
Between June 23 and June 26,1998, the 'D' vacuum breaker cycled similarly on about 15 occasions. Hope Creek operators developed an OD for the 'D' torus to drywell vacuum breaker considering its frequent partial cycling in response to minor-torus to drywell differential pressures about 0.02 psid.
The inspectors reviewed PSE&G OD, 980624117,and determined that it was thorough and consistent with technical specification requirements and the Updated Final Safety Analysis Report. The compensatory actions required by the OD were consistent with the technical specification action station requirements for an OPEN torus to drywell vacuum breaker. The compensatory actions required the control room operators to establish or maintain a differential pressure to close the 'D'
vacuum breaker and to remain in the technical specification action requirement for an cpen vacuum breaker until it is closed.
Control room operators have remained alert to plant evolutions that may raise torus pressure relative to the drywell and have initiated nitrogen makeup to the drywell to mitigate the differential pressure change. The operators' understanding of this problem has minimized the cycling of the 'D' torus to drywell vacuum breaker, but it has also introduced a task that potential!y can distract operators from other control i
room activities. Operations and station management were sensitive to this, and were continuing to evaluate this potantial distraction.
l The engineering department's follow-up assessment (FA) to the OD assessed that the frequent valve cycling, to the intermediate-open indication only, may be I
attributed to small amounts of dirt on the vacuum breaker magnets, or overly l
l conservatively set closed limit switch position indicators. Hope Creek also
I performed a seat leakage test of the combined vacuum breakers and yerified that
'
J drywell to torus bypass leakage was within acceptable limits.
The NRC inspectors noticed that the FA indicated that PSE&G plans to perform corrective actions on the 'D' torus to drywell vacuum breaker at the next refueling l
- _ _ _ _ _ _ _ - _ _ _ _ - - _ _ _ _ - _ - _ - _ _ -
.
.
outage or an outage of sufficient duration. PSE&G planning personnelinformed the inspectors that eight work orders (one for each vacuum breaker) will perform the 18 month preventive maintenance activity, which will set the limit switches at specific values to account for the conservative limit switch tolerance.
c.
Conclusions Hope Creek operators properly initiated an operability determination for the excessive cycling of the 'D' torus to drywell vacuum breaker. The engineering department provided a thorough evaluation of the problem consistent with the operator's initial operability determination.
02.2 Operability Conclusions Associated with Identified Deficiencies a.
Insoection Scone (71707)
The inspectors questioned operators regarding operability associated with certain issues, including the timeliness and completeness of operability conclusions.
b.
Observations and Findinas During this inspection period, the inspectors questioned the operators regarding component status with respect to operability. For example, Action Request (AR)
980523093was written to document an inoperable stop indication at the Remote Shutdown Panel (RSP) for BP-400, associated with the control room chiller circulating pump. The indication would not illuminate or test. The associated
,
operability screening stated that this indication is not required by technical specifications nor is it an acceptance criteria in the RSP channel check. Therefore, the screening concluded that the broken indication does not affect operability of the RSP. The inspectors questioned operations management regarding the basis for the
'
operability screening. Specifically, it was not clear whether the control power for the control room circulating pump was common with the associated indication circuit, and if so, may in fact have had an impact on operability. Operations management subsequently performed a detailed review of the circuit and concluded that there was no operability question. The inspectors agreed with this conclusion when provided with the additional facts and bases. In this case, the initial operability screening was narrowly focused and lacked sufficient bases for its (.
operability conclusion.
[
. In another instance, AR 980613122 identified a tripped and untabeled relay l
associcted with the 'C' emergency diesel generator (EDG) switchgear during EDG post-maintenance testing. The AR stated that this relay was noted to be tripped during testing but muld not be reset until after the EDG was shut down on two
- occasions. Initial investigation identified the relay to be an over-voltage relay.
Maintenance technicians also identified that the relay may not have actually tripped.
- Rather, the relay flag appeared to be dropping out during the EDG startup giving an indication that it had tripped. Additional investigation by operations and engineering personnel identified that this particular relay, associated with the instantaneous
- _ _ - _ _ _ _ _ - _ _ _ _ - _ - - _ - - _ - _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ -
.
.
g
6
over voltage, is not in service. Only a time delay over-voltage relay (alarm function)
is designed to be functional. One of the assigned AR actions is to confirm operation of this time delay relay and further investigate the circuit. The evaluation did not identify any EDG operability concerns.
Action Request 980613122was initiated on June 13,1998. However, it was not approved until three days later on June 16. Although subsequent review by PSE&G
. identified the relay in question and confirmed that there was no adverse operability impact, the inspectors expressed concern that the AR was not approved, and
' therefore, not formally assigned for detailed analysis in a timely fashion.
c.'
Conclusions
^ There were two examples (inoperable indicating circuit for control room chiller circulating pump at the Remote Shutdown Panel, and abnormal response from an unlabeled emergency diesel generator over-voltage relay during post-maintenance testing) where operators were slow to fully document and process the bases for equipment operability when degraded indications were apparent, in both instances,
- post-issue reviews demonstrated that operability was not challenged.
Quality Assurance in Operations 07.1 Quality Assessment Corrective Action Audit a.-
Inspection Scone (40500)
The inspectors reviewed a quality assessment corrective action audit. The inspectors discussed the audit findings with quality assessment personnel, reviewed.
the audit report, and reviewed the action items initiated as a result of the audit.
. b.
Observations and Findmas Quality assessment personnel performed audit 98-190-1, Corrective Action Effectiveness, during the period April 20 to May 1,1998. This audit was common to Salem and Hope Creek. Quality assurance performed this audit at the request of the Nuclear Review Board, and this audit was in addition to the corrective action j
audit currently scheduled for November 1998.
!
i The inspectors reviewed the details contained in audit report 98-1901, dated June l
25,1998. The audit identified several weaknesses related to the corrective action l
program implementation., for which specific Action Requests were initiated to effect
{
L
. resolution. One of the key audit findings identified weaknesses in the process by l
which corrective action effectiveness is reviewed, monitored and tracked. For example, procedure NC.NA-AP.ZZ-0006(O), Corrective Action Progrem, describes the use of a Corrective Action Review Board (CARB) to eveluate the effectiveness i
of Condition Report corrective actions. However, the audit identified that the j
majority of CARB effectiveness reviews consisted of verification that corrective
actions have been completed rather that verifying that the corrective actions have
!
l i
l
_ _ _ _ _ _ _ _ _ _ _ _ _ _
..
- _ _ _ - _ _ _
.
.
.
-
been effective in preventing event recurrence. The audit also identified that there was a large backlog awaiting management review by the CARB. Quality assessment personnel initiated several Action Requests to address the various audit findings.
The inspectors found the audit to be comprehensive in scope and content. While the audit findings were reflective of a probing corrective action effectiveness review, quality assessment and the NRC had previously identified similar concerns.
The inspectors concluded that aggressive management attention is warranted to effect lasting and effective corrective actions.
c.
Conclusions A quality assessment audit of corrective action effectiveness at Salem and Hope Creek was comprehensive in scope and content. Although the audit findings reflected a probing review of the effectiveness of corrective actions, the nature of these findings were similar to prior PSE&G and NRC corrective action system reviews, and were indicative of continued weaknesses in implementing the j
established corrective action program. These findings warrant increased management attention for effective and lasting resolution.
'
l 11. Maintenance M1 Conduct of Maintenance l
M 1.1 Class 1E 125Vdc Batterv Sinale Cell Chanaeouts
l a.
Insoection Scope (62707)
On separate occasions the inspectors observed single cell changeouts on the l
1CD447 and 1 AD411125Vdc safety-related batteries.
I b.
Observations and Findinas On May 27,1998, the inspectors ob:;erved pre-job briefs and maintenance to replace battery cell no.3 and battery cell no. 50 on the 1CD447125Vdc safety-related battery. Problems with these cells were discussed in NRC Integrated inspection Report 50-354/98-05Section E2.3. When this problem was described in NRC Inspection Report 50-354/98-05,PSE&G had not determined how it would i
restore these two battery cells that had fallen below technical specification
,
individual cell voltage values. PSE&G since determined that single cell changeouts l
would be appropriate to restore the battery to a fully operable status.
The inspectors verified that PSE&G had purchased battery cells that were tested in accordance with the Updated Final Safety Analysis Report battery design requirements. The inspectors also verified that the post maintenance testing activities were adequate. The inspectors observed the pre-job brief for the battery i
I
_ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -. _ _ _ _ _ _ - _
_ _ _ _ - _ - _ _ _ _
__-_ _ _ - _ _ _ _ -
_ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ _ - _ _ _
.
e
!
- cell changeods to be thorough and appropriately attended by all individuals that l
were responsible for successfully completing the activity within two hours. Hope Creek technical specifications (TS) only allowed two hours with battery 1CD447 inoperable before actions must be taken to shut down the unit. The maintenance
- technicians were well prepared and carefully completed the evolution allowing j
1CD447 battery to be restored to a fully operable status before the two hour TS action statement expired.
..
On June 5,199' 8, the. inspectors observed and performed similar inspection
' activities for a single cell replacement on the 1 AD411 125Vdc safety-related battery for cell no. 5.- Cell no. 5 was not below any TS limits, but its performance trend data indicated that it my become limiting. PSE&G proactively decided to replace cell no. 5 before it became degraded below TS limits. The inspectors observed similar good performance during the cell no. 5 replacement to what was observed on the 1CD447 battery cell replacements.
!
NRC Inspection Report 50-354/98-05 indicated that Hope Creek had vacated its DC system manager position. PSE&G has since assigned a new DC system manager and has developed a battery system team common to both Salem and Hope Creek -
facilities. The inspectors were informed that Hope Creek intended to replace all the -
. safety-related batteries before their service life expires and to schedule the replacements based on each battery's performance and trend data.
H c.
Conclusions
. The preparations to replace three single cells on safety-related batteries were thorough and well developed. The battery cell replacements were completed in a timely fashion and without error. The inspectors determined that PSE&G had taken initial steps to improve performance monitoring of the safety-related batteries, which was a recent performance weaknesses, as the batteries approached their end of service life.
M1.2 Feedwater Heater Trio
l a.
Inspection Scone (62707. 71707)
)
The inspectors reviewed PSE&G's response and investigation following an unexpected trip of two low pressure feedwater heaters. The inspectors reviewed control room recorder traces for various parameters and interviewed personnel, b..
Observations and Rndinos l
On June 17,1998, with the unit operating at full power, control room operators received an alarm indicating that the '1 A' and '2A' low pressure feedwater heaters had tripped. Reactor power, pressure, and level remained stable. PSE&G investigation identified that workers had built scaffold to support a modification in the turbine building too close to a non-safety related electrical cabinet. Vibration from operating equipment apparently caused the scaffold that was in contact with the panel to actuate the '1 A' and '2A' feedwater heater isolation relay.
i
_ _ _. _.. _ _ _ _ _ _ _ _ _ _ _. _ _
_
J
_ _ _ _ _ _ _ _
____ ___ _ _ _ _ _
,
.
..
-
PSE&G's follow-up investigation found that the purpose of the scaffold was to protect the electrical panel during iron pre-filter modification installation activities.
Although procedure SH.MD-AP.ZZ-OO23(O), Scaffold Erection, Modification and Dismantling Guidelines; did not require a minimum clearance for non-safety related equipment, placing the scaffold in contact with the panel was a poor worker practice. in addition, this panel did not provide any precaution that it contained sensitive equipment, which is common throughout the station. PSE&G's response to the poor worker practices was acceptable, and included the initiation of procedure and training enhancements for scaffold erection. The scaffold was re-
. installed with adequate clearance from the panel.
c.
Conclusions Due to poor worker practice and supervisory oversight, scaffold was erected too close to a non-safety related electrical panel, which caused a vibration induced actuation of a low-pressure feedwater heater isolation relay. This actuation resulted in an unnecessary challenge to control room operators and plant equipment.
M4 Maintenance Staff Knowledge and Performance M4.1 Safety Imoact of Concurrent Reactor Core Isolation Coolina and 'B' Reactor Feed
. Pumo System Outanes a.
Insoection Scone 171707,617261 l '
The inspectors reviewed PSE&G's risk assessment of concurrent maintenance
_
system outagtes on the reactor core isolation cooling (RCIC) system and 'B' reactor feed pump (RFP).
l.'
b.-
Observations and ~Findinas On June 15,1998, prior to rendering the RCIC system unavailable for scheduled maintenance, the Operations Superintendent (OS) reviewed the risk matrix manual
^ to determine the overall plant safety impact with the 'B' RFP already unavailable.
i
.
The 'B' RFP was tagged out of service on June 14,1998, prior to the start of the RCIC system work in accordance with the plant maintenance schedule. The RFP was isolated from the main steam system and the main condenser for steam feak repairs. The OS thought that a RED, or high risk, condition would exist with both systems not available. The matrix also indicated an asterisk for this condition, but p
the asterisk was not defined. The OS appropriately clarified the actual risk associated with the concurrent RCIC system and the 'B' RFP unavailability through
,
l discussions with the probabilistic safety assessment (PSA) group and the operations manager.' The PSA group clarified that the #ED condition only existed if RCIC was already not available while the 'B' RFP was being isolated, with a seal being established on the RFP turbine (RFPT) to maintain main condenser vacuum. (A similar RED risk condition would exist when restoring the 'B' RFP and securing the vacuum seal from the main condenser.) The PSA group clarified that the actual risk would be YELLOW with the 'B' RFP already isolated with a vacuum seal established
. _ _
.
.
..
.
.
concurrent with making the RCIC system unavailable. The OS allowed the RCIC system to be made unavailable for maintenance after he understood the actual risk assessment in this condition and after all other contingencies were established in
-
accordance with the limiting condition for operation maintenance plan. The OS initiated an action request to clarify the risk matrix manual.
On June 17,1998, the 'B' RFP was returned to service prior to making the RCIC system available. PSE&G had originally intended and had scheduled to make RCIC available prior to restoring the 'B' RFP considering the RED risk condition that was described in the previous paragraph. The NRC inspectors questioned the OS about the decisions involved before intentionally entering a RED risk condition while the
'B' RFP was restored. The OS indicated that emergent problems encountered with RCIC during overspeed testing delayed its return. The operations manager and the OS considered that entering a RED risk condition during the short duration required to restore the 'B' RFP would be justified because it would result in an overall
- improvement to plant safety (after the 'B' RFP became available). The NRC
!
- inspectors interviewed the PSA group to determine if the decision was appropriate and to determine if the operations department had consulted the risk assessment group concerning this issue. The PSA group did not see a problem with the
.)
operations department decision but they indicated that they were not informed I
before operations made the decision to restore the 'B' RFP. The NRC inspectors
' determined that this did not meet management expectations in that the PSA group should have been consulted if activities that are identified as RED risk can not be
>
avoided, c.
Conclusions r
Operators appropriately researched and thoroughly understood the risk assessment
. associatad with concurrent unavailabilities of the 'B' reactor feed pump and the reactor core isolation cooling system prior to entering that activity. However, operators did not consult the probabilistic safety assessment group to quantify or confirm their assessment prior to placing the plant in a configuration that may have impacted plant safety.
Ill. Ennineerina E2 Engineering Support of Facilities and Equipment E2.1 Emeroency Diesel Generator Fuel Oil Contamination a.
Insoection Scone (37551. 71707)
,
L
- The inspectors reviewed PSE&G's response and follow-up after sample analyses yielded unacceptable results for the particulate concentration in the 'C' and 'D'
emergency diesel generator (EDG) fuel oil storage tanks (FOST). The inspectors observed sampling activities, FOST cleaning and maintenance, and interviewed station personnel. The inspectors also verified PSE&G's contingency actions after a j
i I'.g
_--
__
__ __ _ _ _ - - -_ _ __ _ _ ___ - - - - - __ - __-_ - __
.
,
.
i. m
.NRC Region I granted a notice of enforcement discretion for the EDG technical specification.
i b.
Observations and Findinas
'
On May 19,.1998, PSE&G sampled the 'C' and 'D' EDG FOSTs, as required by
' technical specification 6.8.4.e and determined that the particulate was greater than
.the 10 milligram / liter (mg/l) acceptance criterion. The 'C' and 'D' FOSTs are associated with the.'B' EDG. There are two FOSTs associated with each of the
. four Hope Creek EDGs As a result of the unacceptable sample, operators declared the 'B' EDG inoperable at 8:20 p.m., and entered a 72-hour allowed outage time as per technical specification 3.8.1.1.
During the next few days,- PSE&G developed and implemented an action plan to identify the cause and source of the contamination and to identify corrective actions. PSE&G sampled and analyzed the 'C' and 'D' FOSTs several times, with similar results. Then, operators drained the 'D' FOST and refilled it with new fuel oil. This new fuel oil was received onsite via several tanker trucks and sampled per l.
technical specification requirements, as well as particulate concentration analysis.
L Although the fuel oil in the tankers was well within specification, the particulate i
results for the 'D' FOST, after it was refilled, was again out of specification.
I
. After the 'D' FOST yielded unsatisfactory results following refilling, PSE&G
continued to investigate the impact of the particulate results. The EDG fuel oil
'
system has an in line system of strainers and filters. The selectable duplex strainers are designed to remove 40 micron particles and the selectable duplex filters remove
' 5 micron particles. PSE&G contacted the EDG vendor and confirmed that
,
p particulate less than 5 microns will not adversely affect EDG operation.
j'
PSE&G performs particulate sampling as required by technical specification 6.8.4.e, j-which requires that a quarterly particulate concentration of the stored fuel is less than or equal to 10 mg/l. Stored fuel refers to the fuel contained in the FOSTs, not
.the new fuel oil received from tankers. This is to done per ASTM D-2276, (
" Standard Test Method for Particulate Contaminant in Aviation Fuel by Line l
Sampling." Technical specification 6.8.4.e allows PSE&G to use 3.0 micron filters
!
instead of the 0.8 micron filters specified in ASTM D-2276.
The existing quarterly requirement for sampling the FOST particulate concentration was incorporated into the Hope Creek technical specifications in Amendment 100, dated July 24,1997. From July 1997 until May 19,1998, all FOST particulate concentration results were within the 10 mg/l acceptance criteria. Although technical specification 6.8.4.e does not require particulate sampling of new fuel oil lfrom' new tankus), the fuel oil testing program did require particulate sampling of new fuel oil as specified in chemistry procedure HC.CH-AP.ZZ-0014(Q), Hope Creek Generating Station DieselFuel Oil Testing Program. This sampling was initiated shortly after Amendment 100 became effective, however, it was inappropriately discontinued in October 1997 at the direction of the Diesel Fuel Oil Program manager. PSE&G continued to conduct the quarterly FOST particulate testing, required by technical specification 6.8.4.e.
i.
_
_
. - - _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
__O
_ _ _ __ _ _
_ - _ _ _ _ _ _ _
..-
e
PSE&G reviewed prior sampling results of the eight FOSTs since August 1997 and confirmed that all quarterly results were acceptable. However, this data showed increasing particulate concentration levels for the 'C' and 'D' FOSTs during the most recent sample results. For example, the particulate concentration levels in the
'C' and 'D' FOSTs were 7.5 mg/l and 3.2 mg/l, respectively, for the December 30, 1997, sample, and were 9.0 mg/l and 6.0 mg/l, respectively, for the February 24, 1998, sample. The inspectors found that PSE&G had not been trending FOST sample results, and responsibility for management of the Diesel Fuel Oil Testing Program was not clearly defined. Consequently, PSE&G did not identify or correct
- this adverse trend.
PSE&G further determined that fuel oil was added to the 'C' and 'D' FOSTs in January 1998 from a single tanker (2000 gallons to 'C' and 1200 gallons to 'D').
PSE&G suspected that this tanker was the source for particulate contamination.
After further chemical analysis, PSE&G determined that the cause for the particulate contamination was FOST contamination with lubrication oil within a fuel oil tanker.
The laboratory analysis showed the presence of additives not found in either EDG fuel oil or other lubrication oil used at Hope Creek. Based upon this information, PSE&G concluded that the January 1988 tanker was the source of the contamination.
Since the 72-hour allowed outage time was nearing expiration (expired at 8:20 p.m.
_on May 22) and PSE&G had not restored the EDG to an operable status, a -
,
conference call was held at 2:30 p.m. between PSE&G and NRC personnel on May 22,1998. The purpose of the call was for PSE&G to request enforcement discretion from the 72-hour allowed outage time of technical specification 3.8.1.1-
,
for the 'B' EDG. PSE&G requested an additional seven days to the allowed outage
, - time to permit draining, cleaning and refilling of the 'C' and 'D' FOSTs. This request was formalized via letter following the conference call on May 22,1998. The NRC granted a notice'of enforcement discretion (NOED) verbally on May 22,1998 at 3:45 p.m. The NRC formalized the NOED in a letter to PSE&G dated May 27, 1998.
As stated in PSE&G's May 22 request and the NRC's May 27 letter, PSE&G
. implemented several compensatory measures during the period of non-compliance with the 7-hour allowed outage time. These included maintaining the 'B' EDG available with additional supplies of fuel oil onsite (tankers) to support continuous operation, not performing elective work that could adversely impact the electrical distribution system, sampling new fuel oil supplies (tankers) before adding to the FOSTs, and maintaining spare oil filters on-site to support 'B' EDG availability with additional filters expected on-site by May 24. PSE&G also sampled the remaining lt six FOSTs on May 22. All were within specification although the 'H' FOST (associated with the 'D' EDG) was elevated and was just below the particulate concentration limit.
The inspectors verified the on-site supply of fuel oil on May 22 following the verbal authorization of the NOED. The inspectors also verified that PSE&G had the appropriate type and number of fuel strainers and filters.
~
'
. _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
.
During the next several days, PSE&G drained, cleaned, re-filled and sampled the 'C'
- and 'D' FOSTs, one tank at a time, as committed to in PSE&G's May 22,1998, letter. After both FOSTs were restored to normal, operators declared the associated
'B' EDG operable at 11:25 a.m. on May 27,1998.- PSE&G subsequently drained, cleaned, refilled and sampled the 'H' FOST, associated with the 'D' EDG due to an elevated, but acceptable, particubte concentration in that tank.
The inspectors observed that operations, maintenance, engineering and chemistry personnel were actively involved in investigating the out of specification limits on the FOSTs, and developing possible corrective actions. The inspectors also observed the analysis of numerous particulate concentration samples from FOSTs and tanker trucks. The sampling activities were conducted acceptably. PSE&G's response to this issue was comprehensive and conservative.
~ Technical Specification 6.8.1 requires that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of NRC Regulatory Guide 1.33, which includes chemical control procedures. PSE&G's failure to perform particulate concentration sampling of new fuel oil received from tanker trucks, as prescribed by procedure HC.CH-AP.ZZ-OO14(Q)is a violation. (VIO 50-354/98-06-04)
c.
Conclusions Chemistry personnel failed to properly sample and report particulate concentration results for new fuel oil received at the station via tanker trucks since October 1997, which was contrary to station procedures. PSE&G responded effectively to the degraded condition of the tanks, and implemented actions consistent with a Notice
,
'of Enforcement Discretion, which the NRC granted on May 22,1998.
"
,
System engineering did not effectively monitor and trend particulate sample results
' for the emergency diesel generator (EDG) fuel oil storage tanks. As a result,
' increasing particulate concentrations were not questioned in a timely fashion, and two of the eight tanks subsequently were out of specification rendering the associated EDG inoperable.
E2.2 Reactor Core Isolation Coolina Svstem Steam Turbine Oversoeed Due to Desian EEL 9L I
a.
Insoection Scone (37551,62707. 61726)
The inspectors reviewed PSE&G's response and corrective actions following an
-
unexpected overspeed of the reactor core isolation cooling (RCIC) system turbine.
L The inspectors observed troubleshooting, testing, modification and safety review activities performed by various PSE&G personnel.
'
,
_ _. _ _.. _ _. _. _ _. _ _. _ _ _ _.. _ _ _ _ _ _ _ _ _ _ _ _ _ _
___
.
..
___
__ - -_
. _ _ _ _ _ _ _ _ -
.
-
.
b.
Observations and Findinas On June 17,1998, operators performed a RCIC turbine overspeed test in accordance with procedure HC.MD-PM.FC-0001(Q), RC/C Steam Turbine /nspection and Preventive Maintenance, following RCIC system maintenance. The RCIC pump and turbine were uncoupled for the test. Upon initiation of the test, during which operators had planned to increase speed in discreet increments, the turbine rapidly accelerated and tripped on overspeed within several seconds. Operators terminated the test to begin an investigation.
There are three primary steam valves located at the RCIC turbine, consisting of the steam line stop valve (F045), the turbine trip and throttle valve, and the governor valve. Hope Creek system procedures instructed the operators to hold the push button for F045 for as long as valve movement was desired (a " jog" control).
I However, PSE&G's troubleshooting of the turbine overspeed identified that F045 continued to move to its full open position without interruption when the push
~
button was released. This response was unexpected.
Further investigation identified that a modification (DCP 4EC-3638), performed during the most recent Hope Creek refueling outage (RFO7; Fall - Winter 1997)
inadvertently removed the jog feature of F045. The modification was designed to resolve generic concerns associated with electrical hot shorts for motor-operated
,
valves (MOV). In addition, the modification removed a normal / test switch from the F045 logic circuit. However, the contractor electrically jumpered the switch, which effectively turned the jog control for F045 into a seal-in feature. Further, this design error was not identified during the contractor's design review or during the post-modification testing.
PSE&G evaluated the overall RCIC system response during a system initiation, with consideration of the specific configuration of the F045 valve. RCIC turbine quick starts have been a design and operational challenge in the nuclear industry, in order to prevent RCIC turbine overspeed events, PSE&G had provided a timing sequence for F045 to make it ramp open for three seconds until the valve reaches 22% open, i
l and then to continue opening to 100% open after ten seconds from the initiation signal. PSE&G also implemented several other changes to the F045 valve design, j
such as valve trim and plug modifications to smooth the valve operation. The i
design error for the F045 control circuit affected the timing sequence described above such that it was bypassed and F045 would stroke fully open without l.
interruption, j
L l
PSE&G concluded that the RCIC system was able to fulfill the its safety related
!
function (inject water to the reactor vessel within 30 seconds of initiation) without j
tripping on overspeed with the F045 timed delay opening circuit bypassed. Further,
-the RCIC system was successfully functionally tested at the end of the last refueling outage, after this modification inadvertently removed the F045 timing sequence, and the RCIC system was periodically operationally tested satisfactorily during the l
operating cycle. PSE&G reviewed the overspeed test data from June 17 and related operational and design basis information, and concluded that the turbine overspeed had no adverse equipment impact.
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_
_ _ _
____.
_ _.
_
_. _ _
. _ _ _ _ _ _
_, _.
._ -
._
_
_
_ _ _ _
. _ - _ _ _
_ _ - _ _ _ _ - _ _ _ _ _ _
_ _ _ _ _ _ -
,.
.
.
.
!
PSE&G determined that the same contractor performed the hot short modification
,
for MOVs on Hope Creek and Salem 1 and 2 (51 at Hope Creek,34 at Salem 1, and 36 at Salem 2). In response to the design error on F045, engineering reviewed each MOV circuit for the modified valves to determine whether similar conditions exist on other MOV circuits. None were identified. The RCIC F045 control circuit is unique in its operation and configuration in addition to the engineering review of the modification, the quality assurance organization was performing an independent assessment of the hot short modifications. No additional similar concerns were identified by the end of the inspection period.
Engineering developed and implemented a change to DCP 4EC-3638 to restore the jog control of F045. The modification also restored the timing sequence for F045 upon an automatic start of the RCIC system.
The Station Operations Review Committee (SORC) reviewed the technical issue and operability determination on June 19,1998 (Meeting No. H98-038). During the review, the SORC identified two additional concerns requiring resolution. The concerns were related to the adequacy of response time testing for certain components and overall RCIC response time testing. These concerns were properly documented in Action Requests (980620075 and 980620162).
l The inspector reviewed PSE&G's response to the RCIC system design error and resulting testing anomalies. Efforts to identify the root causes of the design error
'
and to restore the system to its design basis were effective. The associated system testing that was performed after the jog feature was restored was conducted safely and conservatively, with appropriate attention on good communications and coordination.
PSE&G's failure to ensure proper design and configuration control, and the failure to adequately review and test DCP 4EC-3638 is a violation of 10CFR50, Appendix B, Criterion lli (Design Control). (VIO 50-354/98-06-05)
c,
.Qpr$dyfigng Contract engineering personnel failed to ensure proper design and configuration control for a reactor core isolation cooling (RCIC) system modification during the Fall / Winter 1997 refueling outage. Specifically, the modification incorrectly changed the RCIC turbine steam line stop valve logic by deleting it's sequenced opening feature. As a consequence, the RCIC turbine oversped rapidly during a post-maintenance RCIC overspeed test (RCIC pump and turbine were uncoupled).
l This condition did not render the RCIC system inoperable during the specific valve configuration because the other testing demonstrated the operability of the system l
while in this condition. Station engineering responded appropriately to this design and testing error.
l i
_ _ _ _ _
_ _ _ _
___ -_ _
_ _ _
-
. _ _
)
.
l l-l
l
!
E8 Miscellaneous Engineering issues E8.1 (Closed) Licensee Event Reoort 50-354/98-04: Emergency Diesel Generator Inoperability in Excess of allowed Outage Time Due to High Particulate i
i Concentration in Diesei Fuel Oil Storage Tanks. This issue, including enforcement l
sanctions, is discussed in detail in Section E2.1 of this report. The inspectors l
reviewed this LER and concluded that the report accurately described the event, l
determined the root causes, and developed appropriate corrective actions. The
inspectors considered this LER closed.
I l
IV. Plant Support j
R1 Radiological Protection and Chemistry (RP&C) Controls l
During the period June 8 - 12,1998, an NRC Region-based inspector reviewed the j
Hope Creek Radiological Environmental Monitoring and Meteorological Monitoring Programs. This inspection was a combined Salem and Hope Creek inspection, the i
results of which were documented in NRC Integrated Inspection Report 50-272 &
'
50-311/98-05, dated July 10,1998. That report identified violation No. 50-354/98-06-01.,which related to PSE&G's failure to establish, implernent, and j
maintain environmental sampling procedures to collect drinking water, fish, and
invertebrate samples. The report also identified inspector follow-up item No. 50-L 354/98-06-02,which was related to determining whether there was a requirement to perform channel calibrations for meteorological sensors.
R4 Staff Knowledge and Performance in RP&C l
R4.1 Standbv Liauid Control System Boron Samolina Problems l
l a.
Insoection Scooe (71750. 71707)
The inspectors reviewed the chemistry department identification and follow-up of l
unexpected sample results for the standby liquid control (SLC) system sodium
'
pentaborate solution. The inspectors reviewed documentation and interviewed responsible chemistry personnel.
b.
Observations and Findinas
!
On June 5,1998, Hope Creek chemistry personnelidentified that a routine SLC l
sodium pentaborate sample was lower than expected. The sample analysis result was 13.01 weight percent (w%) sodium pentaborate concentration, which was out l
j of specification. Technica! Specification (TS) 3.1.5 requires the SLC sodium i
pentaborate concentration to be between 13.6 and 14.4 w%. Chemistry personnel immediately believed the results to be invalid because an analysis performed two days earlier yielded a result of 13.98 w% and there were no changes in SLC tank level.
!
l t___
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. _ _ _ _ _ _ _
_ _ _ _ _ _ _. _ _ _ _ _ _ _. _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _. _
_ _. _
_ _.. _ _ _ _ _ _
_ _ _ _ _____ _ __ _
...
..
..
..
.
.,
u
..
. A chemistry supervisor was providing oversight of a substantial portion of the June 5 sample analysis. After the unexpected results were identified, the supervisor suspected that the chemistry technician made an error during the dilution phase of
the analysis, during which the supervisor was not present. Later, two additional independent samples were taken early in the afternoon on June 5. These results were acceptable and were 13.72 w% and 13.76 w%. Although these results were acceptable, the chemistry technicians typically add a mixture of borax and boric acid to increase the sodium pentaborate concentration whenever the sample result is less than 13.8 w% in order to maintain sufficient margin to the low acceptance criterion. The chemists added borax and boric acid to increase the sodium pentaborate concentration to about 14.2 w%. However, the subsequent sample, taken around 10:00 p.m. on June 5, measured 13.8 w%, which was acceptable but was not as expected.
The chemistry manager directed additional samples to be taken and analyzed. A sample taken at 12:50 a.m. on June 6 was analyzed to be 14.22 w%, and a subsequent sample taken at 8:15 a.m. on June 6 was analyzed to be 14.27 w%,
both within the expected range. Chemistry personnel believed that the 13.72, i
13.76 and 13.8 w% results may have been slightly low due to an insufficient tank mixing time, which is currently prescribed to be four hours.
The inspectors reviewed the associated Action Request (AR 980607046)that -
chemistry initiated for the sample analysis discrepancies. The AR appropriately identified that there were prior instances where SLC tank sodium pentaborate concentration was found to have been reduced after the performance of a SLC system surveillance..The inspectors noted that the AR contains an assignment to evaluate the SLC tank mixing time requirement following chemical addition to ensure accurate sample analyses are achieved.
PSE&G identified severalissues while evaluating this event. Although the chemistry
,
l technician was properly trained and qualified for this activity, he had not performed
'
' this analysis in eight years. The chemistry department previously recognized (about a year ago) a need to improve proficiency training for specific tasks, including boron
'
analysis, to ensure these tasks are performed on some minimum specified L
frequency. In response, the department developed a proficiency training cycle for
!
many tasks, to be completed over a two-year period. The individual performing the initial June 5 analysis had not received proficiency training for boron analysis for the current two-year proficiency training cycle (he was scheduled for this training in l
'
1999).
PSE&G also identified that supervisory oversight of the sample analysis and I
communication of analysis results with the operations department were weak. The supervisor did not provide oversight during a critical step in the boron analysis
. (dilution), and the chemistry department did not promptly inform the control room operators of the sample result even though it was suspected to be inaccurate.
i E
u______z____________._
. _ _ _ _ _ _
__
_
-___ ___ _ _ _ _ - _ _ _ _ _ _
o l
!
.
!
The inspectors confirmed that PSE&G has initiated appropriate corrective actions to l
address the identified deficiencies. The inspectors also confirmed that the sodium pentaborate concentration did not fall below the allowed TS value of 13.6 w%.
l l
c.
Conclusions Due to a chemistry technician error, insufficient technician proficiency training, and ineffective supervisory oversight, unsatisfactory results were reached during the performance of a boron analysis for the standby liquid control system sodium pentaborate solution. PSE&G promptly confirmed by additional sample analyses that the actual sodium pentaborate concentration did not fall below technical specification allowable values (the initial errant analysis yielded an inaccurate and out of specification result).
V. Manaaement Meetinas X1 Exit Meeting Summary i
The inspectors presented the inspection results to merr'. ars of licensee management at the j
conclusion of the inspection on July 16,1998. The licensee acknowledged the findings i
presented.
i l
.-
!
l L__.__________.____________________.----__..______
_
_
_
_
_ __._
_ _ -... - - - _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ -. _.
,
- - -
_ _ - -. - -
---_ _____________ _
,
.-
'.
INSPECTION PROCEDURES USED IP 37551:
Onsite Engineering l
lP 61726:
Surveillance Observations
!-
IP 62707:
Maintenance Observations IP 71707:
Plant Operations IP 71750:
Plant Support Activities IP 92901:
Followup - Plant Operations
!
IP 92903:
Followup - Engineering
!
IP 93702:
Prompt Onsite Response to Events at Operating Power Reactors
ITEMS OPENED AND CLOSED Opened l
50-354/98-06-01 VIO Failure to establish environmental sampling procedures to collect drinking water, fish, and invertebrate samples. (Section
,
l R1)
50-354/98-06-02 IFl Determine whether requirement exists to perform a channel calibration for meteorological sensors. (Section R1)
50-354/98-06-04 VIO
. Failure to follow procedures to sample and analyze emergency l
diesel generator new fuel oil for particulate concentration.
(Section E2.1)
l 50-354/98-06-05 VIO Failure to ensure proper design control / configuration, design I
review, and test control for the RCIC system. (Section E2.2)
i Opened / Closed 50-354/98-05-03 NCV Failure to properly restore the 'A' SSW pump breaker. (Section 01.1)
'
i Closed 50-354/98-04 LER Emergency Diesel Generator Inoperability in Excess of allowed
,
!
Outage Time Due to High Particulate Concentration in Diesel Fuel Oil Storage Tanks. (Section E8.1)
L_____________---
-_
--
_ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _. - _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ - _ - _ _ _
_ _. - _ _ -
_,
.
.
e
.
LIST OF ACRONYMS USED AR Action Request CARB Corrective Action Review Board DCP Design Change Package EDG.
.FA Followup Assessment FOST Fuel Oil Storage Tanks mg/l Milligram / liter MOV Motor-operated Valves
- NOED Notice of Enforcement Discretion NRC Nuclear Regulatory Commission OD Operability Determination OS Operations Superintendent PCIG Primary Containment instrument Gas PDR
- Public Document Room PSA Probabilistic Safety Assessment PSE&G Public Service Electric and Gas RCIC Reactor Core Isolation Cooling RFP Reactor Feed Pump RFPT Reactor Feed Pump Turbine RO Reactor Operator RP&C Radiological Protection and Chemistry RSP Remote Shutdown Panel
.SLC Standby Liquid Control SORC Station Operations Review Committee SPDS Safety Parameter Display System
,
SSW Station Service Water TS Technical Specification w%
Weight Percent I
%
i
'
_ - - - _ _ - - _ _ _ _ _ _ _ - - - _ _ _ _ -. - - - _ _ _ - - -