ML20210N913
| ML20210N913 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/11/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20210N898 | List: |
| References | |
| 50-354-97-04, 50-354-97-4, NUDOCS 9708260179 | |
| Download: ML20210N913 (34) | |
See also: IR 05000354/1997004
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION l
Docket No:
50 354
License Nos:
NPF 57
- Report No.
50 354/97 04
Licensee:
Public Service Electric and Gas Company
Facility:
Hope Creek Nuclear Generating Station
Location:
P.O. Box 236
Hancocks Bridge, New Jersey 08038
1
Dates:
June 1,1997 July 12,1997
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Inspectors:
S. A. Morris, Senior Resident inspector
R. K. Lorson, Resident inspector
J. D. Noggle, Senior Radiation Specialist
J. D. Orr, Reactor Engineer
F. J. Laughlin, Emergency Preparedness Specialist
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Approved by:
James C. Linville, Chief, Projects taranch 3
Division of Peactor Projects
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9708260179 970811
PDR- ADOCK 05000354
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EXECUTIVE SUMMARY
Hope Creek Generating Station
NRC Inspection Report 50 354/97-04
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 conducted a focused evaluation of the solid radweste/ transportation
program area.
Ooerations
Good control of reactor power changes was evident during the report period, including
appropriate adherence to governing procedures. Response to transient conditions was
generally proper, and required NRC notifications were timely and accurate. Control of
refuel floor activities during new fuel receipt was generally in accordance with established
guidance. However, minor deficiencies were noted in documentation of operability-
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determinations, interpretation of procedural requirements, and dissemination of operational
guidance in the " night order" book (Section 01.1).
PSE&G's assessment of the waste sludge phase separator overflow event (s) was
comprehensive and thorough, yet at the same time highlighted significant weaknesses in
Individual attention to detail, and inter and intra departmental communications and
planning (Section 04.1).
Quality assurance oversight of operations was frequent and thorough. Several
performance improvements were instituted as a direct result of OA department activitles.
The Nuclear Review Board and Station Operations Review Committee continued to function
acceptably and in accordance with established requirements (Section 07.1).
Maintenance
Work week critiques continued to be very critical self assessments, and frequently
identified several areas for needed improvement. Housekeeping in the station remained
excellent. QA oversight of significant maintenance evolutions was good. This relatively
high number of emergent work activities (well above estabh:5ed goals) continued to
displace previously scheduled work items and often disrupted daily work plans (Section
M 1.1 ) .
Maintenance technicians exhibited inconsistent performance in the completion of assigned
work activities. While generally good coordination between departments and supervisory
oversight was evident for complex or high risk work, weak or inadequate adherence with
established guidance during the conduct of several routine maintenance activities resulted
in work delays and the need for unplanned engineering analyses. (Section M1.2)
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Executive Summary
Hope Creek personnel appropriately evaluated technical specification requirements for the
emergency diesel generator fuel oil system and consistently applied the requirements for
survolitance testing (Section M3.1).
The decisioa to remove the "A" and "C" resid ,al heat removal subsvstems from service
while contiriuing to question high pressure coolant injection valve operability was a poor
management decision and reflected a nonconservative safety focus (Section M4.1).
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Operators responded appropriately to a self-revealing fault in the rod position indication
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system. Good coordination between maintenance ar'd engineering departments was
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evident during system recovery efforts. Performance of troubleshooting activities was
mixed, but was ultimately successful in restoring the system to operability in a well
controlled manner (Section M4.2).
Enaineerina
The inspectors observed good overall coordination between engineering and other
departments during resolution of unplanned events and emergent corrective maintenance
activities. Departmental perior. nance indicators continued to evolve into more useful and
effective tools for management decision making (Section E1.1).
A management decision not to perform vendor-recommended inspecCons of recirculation
pump shafts (for craking) was based on sound engineering judgement and assessment of
other industry information regarding the potential coe.cern. A Nuclear Review Board
evaluation of this decision was appropriate (Section E2.1).
Though PSE&G management exhibited appropriate concern and acted promptly to resolve
individual service water system component failures, the frequency of emergent work and
number of repeat failures was high, which significantly increased overall system
unavailability time to levels well beyond established goals. This fact in part indicated that
previcus efforts to resolve repeat equipment failures (and improve system reliability) have
been less than fully effective (Section E2.2).
The operating experience review program functioned appropriately to ensure that
significant industry lessons learned were evaluated for applicability to the Hope Creek
station (Section 57.1).
Plant Suoaort
The licensee provided an effective waste mixing, sampling, characterization and waste
form processing program (Section R1.1).
The inspector noted that all of the solid radwaste was apptcpriately utored, locked, and
controlled by radiation protection technicians with required facility surveillances being met,
and that the radiation levels associated with t e low level radwaste stcrage f acility were
well below regulNy requirements (Section R2.1).
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Executive Summary
The licensee has an abandoned bituminous waste manufacturing system that has not yet
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been placed in a long-term laidup condition. Licensee actions appear appropriate to effect
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proper layup of the processing equipment (Section R2.2).
The solid radwaste/ transportation procedures generally met the new Department of
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Transportation requirements, however, one violation was identified in shipping paper
documentation for contaminated laundry shipmonts. Some improvements are needed in
the Process Control Program and implementing procedures to ensure that essential solid
radwaste processing parameters and criteria that are contained in federal, state and burial
license requirements are captured and properly referenced (Section R4.1).
The solid radwaste program oversight was good for offsite shipment review, however,
some improvements are needed for providing QA audits of radwaste processing vendors
(Section R7).
The inspectors noted marked improvements in the manner by which all ERO qualifications
were maintained, and increased sensitivity to current qualification status by ERO members
(Section PS.1).
The inspectors concluded that QA department efforts in the security area positively
influenced the ability of the site security organization to fulfillits function effectively
(Section S7),
As a result of actions taken in response to this and other self-identified issues, fire
protection and safety at the Hope Crcek site was notably improved (Section F7).
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TABLE OF CONTENTS
EX EC UTIVE S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
TABL E O F C O NT ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v
l . O p .i t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01
Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.1 General Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
04
Operator Knowledge and Performance . . . . . . . . , . . . . . . . . . . . . . . . . 3
04.1 Waste Sludge Phase Separator Overflow Event (s) . . . . . . . . . . . . 3
07
Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
07.1 Quality Assurance in Plant Operations . . . . . . . . . . . . . . . . . . . . 4
08
Miscellaneous Operations issues . . . . . , . . . . . . . . . . . . . . . . . . . . . . 5
08.1 (Closed) LER 50 354/97-08: Engineered safety feature
actuation: "C" service water pump auto-start . . .
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08.2 (Closed) Special Report 50-354/97-10: Overpown event due
to loss of f eedwater heaters . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
08.3 (Closed) LER 50-354/97 11: Failure to complete flood
protection action within required time frame
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08.4 (Closed) LER 50-354/97-012: Control rod scram during testing . . 6
08.5 (Open) LER 50 354/97-13: Unplanned high pressure coolant
injection system inoperability
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08.6 (Closed) URI 50-354/96-05 01: Reactivity mismariagement
SVellt . . . . . . . . . . . . . . .
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11. Maintenance . . . .
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M1
Conduct of Maintenance
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M1.1 General Comments
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M1.2 Maintenance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
M3
Maintenance Proceduras and Documentation . . . . . . . . . . . . . . . . . . . . 9
M3.1 Emergency Diesel Generator Surveillance Testing . . . . . . . . . . . . 9
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M4
Maintenance Stu! Knowledge end Performance . . . . . . . . . . . . . . . . . 10
M4.1 High Pressure Coolant injection System Feedwater injection
Valve Review
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M4.2 Rod Position Indication System Fault . . . . . . . . . . . . . . . . . . . . 12
111. Engineering
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Conduct of Engineering
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E1.1
General Comments
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E2
Engineering Support of Facilitias and Equiptrmt . . . . . . . . . . . . . . . . . 13
E2.1
Recirculation Pump Shaft Cracking . . . .
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E2.2 Station Service Water System Reliability and Availability . . . . . . 14
E7
Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . 16
E7.1
Engineering Review of Industry Operating Experience . . . . . . . . 16
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Table of Contents
I V. Pl a n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
R1
Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 16
R1.1
Liquid To-Solid Radwaste Processing . . . . . . . . . . . . . . . . . . . . 16
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Status of RP&C Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . 17
R2
R2.1
Onsite Radwaste Storage . . . . . . . . .
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R2.2 Laidup Radwaste Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . 18
R3
RP&C Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . 19
R3.1
Radwaste/ Transportation Procedures . . . . . . . . . . . . . . . . . . . . 19
R4
Staff Knowledge and Performance in RP&C . . . . . . . . . . . . . . . . . . . . 20
R4.1
Radioactive Material Shipping Documentation Review . . . . . . . . 20
R5
Staff Training and Qualification in RP&C . . . . . . . . . . . . . . . . . . . . . . 21
R5.1 Radwaste Transportation Training for Shippers and RP
Te c h nic ia n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
R7
Quality Assurance in RP&C Activities
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R8
Miscellaneous RP&C lssues
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R8.1
Updated Final Safety Analysis Report (UFSAR) . . . . . . . . . . . . . 22
P3
EP Procedures and Documentation
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P3.1
Ir Office Review of Licensee Procedure Changes . . . . . . . . . . . 22
P5
Staff Training and Qualification in EP . . . . . . . . . . . . . . . . . . . . . . . . . 22
P5.1
Respirator Qualifications in Support of Emergency
Preparedness
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S7
Ouality Assurance in Security and Safaguards Activities . . . . . . . . . . . C3
F7
Quality Assurance in Fire Protection Activities . . . . . . . . . . . . . . . . . . 23
V. M a n a g e m e nt M e e ting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
X1
Exit M e eting Su mm ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
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1. Operations
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Conduct of Operations
,Qld General Observations
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a.
insnectica Scooe (71707)
The inspectors conducted frequent observations of ongoing plant operations,
including control room walkdowns, station operator and manager interviews, and
procedure reviews.
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b.
Observations and Findinos
Control of Routine Activities
Numerous minor power changes w re required during the report ,oeriod in response
to both planned and unplanned betivities, which included the minor reactivity
manipulations needed to compensate for fuel depletion as the end operating cycle
neared. The inspectors observed several power changes and noted that strict
compliance with procedures was maintained, and good oversight by both shift
supervision and reactor engineering was evident. Limitations in maximum
recirculation loop flow were well documented, understood and followed. Repeat
problems with " sticking" and " double-notching" control rods were experienced
following individual control rod scram time testing; however, operators appropriately
consulted the governing abnormal operating procedures and documented the
occurrences in control room logs and the corrective action program as necessary.
Ultimate Heat Sink (UHS - Delaware River) temperature increated more than 15
degrees during the six week period to a maximum of 78 degrees on July 12,1997.
The inspectors noted appropriate concern for the rising trend by operations and
management staff, particularly in light of recent problems with Station Service
Water (SSW) system reliability and unavailability (see section E2.2) and the
previously established admin:strative limit on maximum UHS temperature .
Operators demonstrated a good awareness of the necessary compensatory actions
and administrative controls required by an operability determination for the UHS.
Hope Creek has been operated under an administrative limit on maximum UHS
temperature since early 1996, when station engineering personnel determined that
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the current technical specification limit was non-conservative.
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With two exceptions, the inspectors observed good recognition of plant conditions
requiring entry into technical specification (TS) action statements. Frequent and
appropriate use of design basis information in the UFSAR and TS bases was
e/ident; examples included documentation of rod position indication system (RPIS)
anomalics and reactor manual control system " lockups." Exceptions to this
typically good performance included a reluctance to incorporate an engineering
assessment of the impact of self-identified control rod speed concerns into an
operability determination, and a more significant issue involving a failure to enter the
TS 3.8.1.1 action statement for an inoperable emergency diesel generator (EDG). In
the latter case, which occurred on July 8,1997, operators failed to recognize the
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need to enter the TS action statement when the EDG inoperability time
unexpectedly exceeded 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> while attempting to "barr" the engine over prior to a
routine monthly surveillance test. PSE&G management implemented reasonable
corrective actions for this self identified event, which included removing the
responsible supervisor from shift duties and re sensitizing the entire operations
department on the need to remain cognizant of the status of all safety-related
equipment. Additionally, once this error was discovered by the relieving shift, the
required actions were completed. This licensee identified and corrected violation is
being treated as a Non Cited Violation, consistent with Section Vll.B.1 of the NRC
Enforcement Policy. This item is closed. (NCV 50-354/97-04 01)
During a routine control room tear the inspectors questioned a night order book
entry that detailed specific operator actions for a degraded recirculation pump seal,
and for an electro-hydraulic control system leak. The inspector noted that the some
of the actions described in the night order book would be more appropriately
addressed in a formal procedure. The inspector discussed this issue with the
Operations tianager who agreed with the observation and subsequently revised the
affected abnormal operating procedures. The inspector concluded that this issue
was properly addressed.
Control of Emergent and Unplanned Activities
As has been documented in other recent NRC assessments, the inspectors
continued to observe good overall performance in operator response to transient
conditions and unplanned events. This was particularly apparent during the period
between July 5 and July 10, when the RPIS anomalies first surfaced (see section
M.4.2). Additionally, operators exhibited a good questioning attitude during control
rod manipulations and identified a repeat issue involving excessive rod speeds. - By
promptly involving engineering support perr.,onnel, operators were able to promptly
validate and document the concern, and enlist maintenance support to resolve the
problem in a timely manner. Response to a single rod scram event on June 13 was
also judged to be good (see section 08.4). Two 10 CFR 50.72 non-emergency
event reports (single rod scram and high pressure coolant injection inoperability)
were both timely and accurate.
Preparations for Refueling
The inspectors observed 'Aeveral activities on the refuel floor involving preparations
for the upcoming refueling outage. Specifically, the inspectors witnessed new fuel
receipt inspections and subsequent movement into the spent fuel pool, as well as
fuel channel receipt, inspection, and preparation. Foreign material exclusion
practices were also evaluated and judged to be acceptable. The inspectors-
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observed good use of governing procedural guidance and good oversight and
monitoring by operations, maintenance, and angineering department staff and
supervision. Corrective actions from previous events (in 1994 and 1995) involving
dropped new fuel bundles remained in place and were effective.
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The inspectors questioned an coparent discrepancy in the manner two different
operating shifts interpreted standard cperating procedure HC.OP-SO.KE-0001(Q)
guidance for the level of supervision is required on the refueling bridge while moving
new fuel bundles. Specifically, two different shifts identified an inconsistency in
the noted procedure; one shift opted to put a senior reactor operator (SRO) on the
bridge until the procedure could be clarified while the previously on-duty shift
elected to simply follow the guidance consistent with past practice (and did not
station an SRO on the bridge). Additionally, the latter-described shift did not initiate
a procedure revision request to clarify the procedure, which was inconsistent with
PSE&G management expectations. Until this issue was discussed with operations
department management, no action request had been generated to document this
concern. Operations management agreed with the inspectors findings in this
regard.
c.
Conclusions
Good control of reactor power changes was evident throughout the mar y
manipulations made during the report period, including appropriate adherence to
governing procedures. Response to transient conditions was generally proper, and
required NRC notifications were timely and accurata. Control of refuel floor
activities during new fuel receipts was generally in accordance with established
guidance. However, minor deficiencies were noted in documentation of operability
determincions, interpretation of procedural requirements, and dissemination of
operational guidance in the " night order" book.
04
Operator Knowledge and Performance
04.1 Waste Sludae Php_sylL poerator '- srflow Event (s)
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a.
Insoection Scoce (71707,71750)
The inspectors observed and evaluated PSE&G's response to an event in which a
radioactive waste collecting tank inadvertently overflowed during a routine
condensate polisher resin transfer,
b.
Observations and Findinas
On June 26,1997, during a routine condensate polisher resin transfer evolution,
several hundred gallons of contaminated wastewater and sludge were overflowed
from the waste sludge phase separator (a collecting tank) into the diked area around
the tank. While no radiological contaminants were released from the f acility as a
result of this " spill," significant effort was required to analyze the cause(s) of the
event and to develop and implement a plan for prompt cleanup.
The inspectors evaluated PSE&G's followup of this event, both the efforts to
determine cause(s) and institute corrective actions as well as to recover from the
spill. PSE&G's investigation concluded that the principal causes were a f ailure of a
motor operated valve in resin transfer flow path to reposition upon demand, as well
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as a failure by the chemistry technician conducting the evolution to ensure that the
valve was in the proper position per the governing procedure before initiating
subsequent steps. Corrective actions for these issues were judged to be
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appropriate.
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Additional issues were raised by this event. Spec"!cally, communications between
the radioactive waste control room and the technician in the field were ineffective
and resulted in a delay in terminating the overflow evant. Following the event, the
cognizant department supertisors were not promptly informed. Control room
operators did not learn of the occurrence until an action request documenting the
issue was submitted for approval several hours later. These issues highlighted
inconsistent application of established management expectations and good
operating practice.
The inspectors reviewed and monitored a portion of the cleanup effort, which
involved the use of a portable submersible pump discharging to an opened manway
at the top of the tank. Cleanup activities were largely completed by July 3, seven
days after the event, but were suspended prior to the long holiday weekend so that
more localized decontamination could be completed the following week. However,
technicians failed to remove the pump's discharge hose from the tank prior to
departing for the weekend, which resulted in several hundred more gallons of waste
being siphoned back out of the tank and into the diked area. The room was again
pumped dry, decontaminated and the system restored, resulting in additional
exposure to workers in the high radiation tank area,
c.
Conclusions
PSE&G's final assessment of the waste sludge phase separator overflow event (s)
was comprehensive and thorough, yet at the same time highlighted significant
weaknesses in individual attention to detail, inter- and intra-departmental
communications and planning, and maintaining radiation exposure as low as
reasonably achievable.
07
Quality Assurance in Operations
07.1 Quality Assurance h Plant Ooerations
a.
Insocction Scoce (71707. 61726)
The inspectors reviewed quality assurance (QA) activities throughout the period,
including a recently completed QA departr,ent audit of Hope Creek operations, as
well as observed Nuclear Review Board (NRB) proceedings during which operational
items were discussed. Additionally, the inspectors attended several station
operations review committee (SORC) meetings to evaluate safety perspectives.
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b.
Observations and Findinas
The inspectors observed a portion of the in-progress OA audit of operations
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department activities, and judged the scope and quality of the effort to be good.
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Several significant action requests were generated as a result of this QA evaluation
which enhanced the safety of operations department activities. Specific examples
included the identification of individual f ailures to adhere to working hour (overtime)
requirements, plant labeling deficiencies, temporary modification implementation,
and, most significantly, the discovery that platng service water or safety auxiliaries
cooling system pumps in " manual" prevented their automatic operation in response
to certain engineered safety feature actuation signals. This latter issue resulted in a
thorough re-evaluation of current operations department procedures to ensure that
normal operating practices did not inadvertently cause safety subsystems to
become inoperable.
The NRB convemd for a two day session during the period and critically assessed
PSE&G's management and safety focus in the operation of the Hope Creek facility.
The inspectors witnessed thorough and probing questions of licensee management,
which indicated that the NRB members had adequately prepared for the meeting by
reviewing required materials. The inspectors noted that NRB members were
particularly concerned with recent negative trends in human performance,
unplanned technical specification action statement entries, emergent corrective
maintenance, organizational stability, and readiness for the upcoming refueling
outage.
The inspectors attended several SORC meetings during the report period and noted
generally good safety focus during the group's review of required documents. One
example involved a thorough questioning of a needed change to the minimum
critical power ratio limit (from 1.26 to 1.31) in order to permit extended reactor
operation to the planned refuel outage start date in September 1997. Additionally,
the SORC members also frequently reviewed non-required materials (e.g. operability
determinations, management decisions, etc.), which indicated a desire to ensura
safety and quality in allimportant station activities. The inspectors evaluated a
sample of approved SORC meeting minutes for accuracy and judged them to be
acceptable,
c.
Conclusions
Quality assurance activities in operations were frequent and thorough. Several
performance improvements were instituted as a direct result of QA department
activities. NRB and SORC continued to function acceptably and in accordance with
established requirements.
08
Miscellaneous Operations issues
08.1 (Closed) LER 50-354/97-08: Engineered safety feature actuation: "C" service water
pump auto-start. This event was discussed in NRC inspection report 50-354/97-03.
The inspectors judged that the committed corrective actions stated in LER, i.e.
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enhancing associated maintenance procedures and increasing the frequency of
transmitter line back flushes, were reasonable given the relatively minor significance
of the event.
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08,2 (Closed) Snecial Reoort 50-354/97-10: Overpower event due to loss of feedwater
heaters. This event was discussed in NRC inspectior' report 50 354/97-03. No
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new issues were revealed by this LER.
08,3 (Closed) LER 50-354/97-11: Failure to complete flood protection action within
required time frame. This LER was a minor issue and was closed.
08.4 (Closed) LER 50-354/97-012: Control rod scram during testing. The control room
operators responded properly to an unexpected single control rod scram during
reactor protection system (RPS) logic testing on June 13. PSE&G performed an
investigation and determined that the scram occurred due to a fuse failure that de-
energized the "A" scram solenoid for Control Rod 26-27 prior to the "B" scram
solenoid being de-energized for testing. Operators stabilized reactor power
following the rod scram, made a four hour report to the NRC per 10 CFR 50.72, and
safely recovered the rod. The inspectors reviewed the core thermallimits reports
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and noted that core thermallimits were not exceeded during the event or
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subsequent recovery. The event was adequately described in this LER and no new
issues were revealed.
08.5 (Ocen) LER 50-354/97-13: Unplanned high pressure coolant injection system
inoperability. This issue is described in detailin Section M4.1 of this report. As
stated in the noted section, this LER did not provide an adequate level of detail
regarding the event. PSE&G management acknowledged the inspectors'
'
assessment of the LER, and stated that an additional review of the document would
be conducted.
08.6 (Closed) URI 50-354/96-05-01: Reactivity mismanagement event. The licensee
determined that the root cause for a reactivity mismanagement event, involved the
lack of self-checking by a reactor engineer and an orgunizational/ programmatic
weakness. Corrective actions planned and taken by the licensee included:
enhancements to the applicable procedure HC.RE-FM.ZZ-0001(O), Revision
10, " Guidelines For Control Rod Movement" that provided very good
guidance for control rod pulls;
performance of additional training for all reactor engineers that addressed
e
self checking skills, management expectations for on-shift support with
r
double verification duties, and the formalization of three-point-
communications;
modifications to rod pull sheets for clarity; and
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7
the addition of a new requirement for reactor engineering personnel to
develop (and nuclear shift supervision to review) deviation plans for rod pulls
prior to changing core conditions.
The inspectors found that the licensee performed effectiveness reviews of their
correctiva actions on two separate occasions with veiy positive results. The
inspectors determined that the licensee's root cause analysis was thorough and
corrective actions were effective and very wellimplemented. Based on the
inspectors' review of the licensee's root cause analysis and corrective actions
taken, the absence of any safety consequence related to this event, and the
absence of any repeat occurrence of this event, the inspectors concluded that this
unresolved item is closed,
ll. Mainter} gag.a
M1
Conduct of Maintenance
ML1, General Comments
The inspectors attended numerous maintenance planning meetings and management
briefings, reviewed several post work week critique documents and performance
indicators, and interviewed work control center staff to determine the effectiveness
of maintenance planning and implementation. As a result of this effort, the
inspectors noted several positive and negative indications of PSE&G's performance
in this area. Specifically, positive indications included an increase in the number of
maintenance procedure revision requests, evidence that technicians closely
scrutinized the quality of procedures. Work week critiques continued to be very
critical self assessments, and frequently identified several areas for needed
improvement. Interviews with individual technicians indicated that recently re-
emphasized management expectations for procedural adherence and use of self-
check principles were generally well received and understood. Housekeeping in the
,
plant continued to be excellent. QA oversight of maintenance evolutions was good.
In spite of the noted positive observations, the inspectors questioned several weak
performance attributes. Of particular note were the procedural adherence issues
highlighted in section M1.2 below. Additionally, the relatively high number of
emergent work activities (well above established goals) continued to displace
previously scheduled work items and often disrupted daily work plans. Time spent
in unplanned technical specification action statements, as well as the number of
individual issues, remained above PSE&G goals, and frequently resulted in
challenges to schedule adherence. Discussions with various licensee managers
indicated recognition of these issues and that action plans had been developed and
implemented to address each of the concerns.
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4
8
Mil Maintenance Observations
a.
Insoection Scone (62707)
The inspectors observed numerous preventive and corrective maintenance activities
l
during the report period, with more in-depth assessment of the following work:
1
HCU 22 59/38-59 Replace Scram Outlet Valve Seats
HCU 38-03 Repair Broken Flex Conduit
D" Emergency Diesel Generator Extended Outage
e
"D" Station Service Water Pump Repack
t
Non-1E Bailey Panel Power Supply Replacement
New Fuel Receipt and Inspection
b.
Observations and Findinas
The inspectors observed generally good work coordination between the operations,
radiation protection (when necessary), maintenance and engineering departments in
the conduct of the above activities. Radiation protection personnel worked closely
with mechanical maintenance technicians at the control rod drive (CRD) hydraulic
control unit (HCU) job sites and ensured good ALARA practices were maintained.
Operators performed all CRD manipulations and scram time post maintenance
testing without error. Good coordination between operations and mechanical
maintenance ensured that the "D" SSW pump packing was replaced and adjusted in
a timely fashion. The non 1E Bailey system power supply replacement job was
thoroughly pre-brinfed and closely monitored by supervisory personnel because of
the work's potential to induce a loss of turbine auxiliarias cooling system flow.
Several problems were noted as well, most of which involved inconsistent
application of established procedures, work orders, or management expectations.
One example involved a July 6 activity in which maintenance technicians were
receiving and staging new fuel channels on the refuel floor of the reactor building.
On this occasion, the technicians stacked channels six high, contrary to work order
guidance that the stacks be limited to five. Technicians interviewed following
identification of this issue stated that stacking channels higher than five had been
done previously without consequence, implying that the specified limit of five was
arbitrarily established. Additionally, once identified, the reactor engineering staff
rather than maintenance initiated the action request to document the event,
contrary to management expectations. This event caused unplanned work for
engineering personnel in that an evaluation to determine the impact on the fuel
channels at the bottom of each stack was necessitated.
Other examples of weak maintenance practices were evident. On July 1, plant
operators attempted to verify proper operation of the refueling bridge following
recently comp'.eted preventive maintenance (PM) activities. However, operators
discovered that a bulb was missing from the " grapple up" indication, which called
into questie the effectiveness of the previously completed PM. Subsequent
investigatkn determined that technicians had racognized the condition, but forgot to
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9
replace the burned out bulb upon initial discovery. When bulb replacement was
eventually performed several hours later, a piece of electrice!insulcting material
from the bulb socket was inadvertently dropped into the spent fuel pool causing a
foreign material exclusion discrepancy. On July 6 maintenance supervision
identified that technicians on the previous shift had improperly performed PM on the
wrong ventilation fan motor, which indicated a lack of attention to detail. On July 7
the QA department determined that maintenance technicians had performed repairs
to degraded seating surfaces on "D" emergency diesel generator air start valves
without using a procedure or work order, and before an engineering evaluation to
determine the amount of seat grinding that could be permitted was completed.
c.
Conclusions
Tht, inspectors observed inconsistent performance by maintenance technicians in
the completion of assigned work activities. While generally good coordination
between departments and supervisori oversight was evident for complex or high
risk work, weak or inadequate adherence to established guidance during the
conduct of several routine maintenance activities resulted in work delays and the
need for unplanned engineering analyses.
M3
Maintenance Procedures and Documentation
M3.1 Emeroency Diesel Generator Surveillance Testina
a.
Instection Scoce (61726)
The inspectors reviewed the adequacy of Hope Creek's monthly technical
specification (TS) surveillance testing of the "B" emergency diecel generator (EDG)
in accordance with procedure HC.OP-ST.KJ-0002(O).
b.
Observations and Findinas
On June 23, the "B" EDG failed its monthly operability surveillance test. During the
test operation of the "B" EDG, the fuel oil day tank low level alarm was received.
By procedure, absence of a day tank low level alarm is used to verify that the day
tank capacity meets TS requirements. The diesel fuel oil transfer pumps should
operate automatically to maintain day tank level above the low level alarm.
However, the inspectors learned that the "B" EDG has recently experienced an
intermittent problem in which the day tank low level alarm is received and both
diesel fuel oil transfer pumps start on the back up level switch.
PSE&G evaluated its current surveillance test acceptance criteria with respect to TS
requirements for the diesel fuel oil system. Licensee personnel concluded that the
cunent acceptance criteria (i.e. absence of a day tank low level alarm), although
satisfactory, was unnecessarily restrictive.
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10
The licensee revised the "B" EDG monthly test procedure appropriately to revise day
tank level acceptable criteria. At the conclusion of the repcrt period, the licensee
was tracking resolution of the intermittent problem with the narrow range level
switch in the corrective action program, and was planning similar changes to test
procedures for the other three EDG's,
c.
Conclusions
Hope Creek personnel appropriately evaluated technical specification 1guirements
for the emergency diesel generator fuel oil system and consistently applied the
requirements for surveillance testing.
M4
Maintenance Staff Knowledge and Performance
M4.1 Hiah Pressure Coolant inlection System Feedwater Inlection Valve Review
a.
Inspection Scone (62707)
The inspector reviewed the licens:a's response to an inoperable HPCI feedwater
injection valve.
b.
Observations and Findinas
The HPCI test return valve (1BJHV F008) failed to stroke open during surveillance
testing on June 16. The Work it Now (WIN) team performed troubleshooting and
attributed the F008 valve problem to loss of the open permissive interlock signal
supplied from the HPCI feedwater injection (8278) valve. HPCI injection flow is
directed to both a feedwater and a core spray header. The interlock signalis
provided through contacts (LS13) operated by rotor assembly 4 in the 8278
injection valve actuator. Rotor assembly 4 also operates the LS16 (F008 " auto
close" interlock signal when HPCI feedwater valve is not shut) and the LS15
.
contacts. The LS15 contacts protect the 8278 injection valve from excessive
seating forces following an autematic HPCI system trip. The LS15 contacts are
required to operate properly to ensure operation of the 8278 injection valve and
HPCI system operability.
The WIN team conducted additional troubleshooting on June 16 that indicated the
LS16 contacts were not functioning properly. Specifically, the resistance across the
LS16 contacts was low (2.4 ohms) when the 8278 injection valve was shut,
however, the LS16 contacts should have been open (high resistance) with the 8278
injection valve shut. The inspector determined during a followup review of the
event that the problem with the LS16 contact resistance data combined with the
known LS13 deficiency provided sufficient information to question the operability of
rotor assembly 4 and the LS15 contacts on June 16.
Operations management incorrectly determined that the LS13 problem did not affect
the other contacts on rotor assembly 4 based on successful stroking of the 8278
injection valve, and misinterpretation of the test and troubleshooting data.
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11
l.
Operations management then concluded that the 8278 injection valve was operable,
and removed the "A" and "C" residual heat removal (RHR) loops from service on
June 17 at 0520 for a planned maintenance outage.
Operations and engineering personnel continued to review the June 16
troubleshooting data, and by June 18 determined that additional testing was
necessary to confirm the operability of the 8278 injection valve. Additionally,
operations management terminated the "C" RHR outage (the "A" RHR loop had
been restored on June 18 at 0040)in response to the 8278 injection valve
operability questions. The June 18 testing Indicated that rotor assembly 4 on the
8278 injection valve was not functioning properly, and the HPCI system was
declared inoperable, and the required four hour notification was performed in
accordance with 10 CFR 50.72.
The licensee developed an integrated plan, which included a plant power reduction
and steam tunnel entry, to troubleshoot and repair the 8278 injection valve.
Maintenance personnel determined that loose limit switch finger base assembly
mounting screws caused the 8278 injection valve problems. The mounting screws
were tightened and the 8278 injection valve was successfully tested.
The initialindications that the HPCI system was ine srable occurred on June 16
when the F008 valve f ailed to operate. The P.PCI:
tem remained inoperable until
the 8278 injection valve was repaired and tested c
une 21. For approximately 20
hours between June 17 and June 18, the HPCI sy
a was inoperable while the -
"A" and "C" RHR loops were inoperable for mainte
ice. Had the system been
required to operate during this period, partialinjecti a flow would have been
available through the core spray injection line. Thb. condition is not addressed by
plant TS for power operation and therefore_TS 3.0.3 applied which required a plant
shutdown to be completed within seven hours. The licensee did not recognize that
the plant was in TS 3.0.3 and therefore failed to perform a plant shutdown as
required. Appendix B, Criterion XVI requires that conditions adverse to quality such
as equipment failures be promptly identified and corrected. Contrary to the above,
the 8278 injection valve problems were not promptly identified which resulted 'n the
plant exceeding the TS 3.0.3 action statement time limits. This is a violation of 10
'CFR 50, Appendix B, Criterion XVI. (VIO 50-354/97-04-02)
The licensee submitted Licensee Event Report (LER) 97-013-00 which described the
event. The inspector noted that the LER did not provide an adequate level of detail
regarding the event. Specifically, the LER did not state that the _TS 3.0.3 action
statement time limits were exceeded, and also did not provide detailed corrective
actions to address the initially incorrect HPCI system operability determination.
PSE&G management acknowledged the inspectors' observations and stated the an
additional review would be conducted.
c.
Conclusions
Initial troubleshooting activities performed following a failure of a system test return
valve provided sufficient information to question the operability of a high pressure
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12
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injection system injection valve. The decision to remove the "A" and "C" residual
heat removal subsystems from service while continuing to question injection valve
,
operability was a poor management decision and reflected a nonconservative safety
focus.
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M4,2 Rod Position Indication System Fault
a.
Inspection Scooe (71707,62707. 61726,37551)
The inspectors witnessed PSE&G's response to a July 5 event involving a failure of
the RPIS to operate reliably, causing station operators to declare the system
inoperable and enter a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> hot shutdown TS action statement. Followup
i
activities to restore the system to service were also evaluated,
b.
Observations and Find.irigg
On July 5, operators observed erratic RPIS indications on a significant portion of the
full core display, the four rod display, and the NSSS computer system. As a result,
operators appropriately declared the system inoperabla and entered the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> hot
shutdown action statement in accordance with TS 3.1.3.7.
Maintenance and
engineering personnel promptly developed and implemented an initial
troubleshooting plan, and quickly deduced that the problem could be alleviated by
simply reducing the !nternal temperature of the cabinet housing the RPIS circuitry.
As a result of this initial effort, and implementation of compensatory measures to
ensure cabinet temperature was maintained at a value below which degradation
would occur, operators completed an RPIS operability determination (subsequently
reviewed and approved at an emergency SORC meeting) and exited the noted
action statement just before the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> time limit expired. The inspectors
observed all of the aforementioned activities and judged them to be within both
PSE&G and regulatory guidelines. A conference call was held between senior NRC
and PSE&G management on July 5 to discuss the licensee's completed and planned
course of action.
The quality and control of troubleshooting activities that followed the declaration of
RPIS operability were mixed. Initial efforts to identify the degraded system
components through the evening of July 5 into the following morning yielded little
information about the source of the fault, but caused operators to re-enter the noted
TS shutdown action statement three separcte times as the problem was
intentionally reproduced. The inspectors judged that the troubleshooting plan
employed during these efforts was not well developed.
Station management deferred any subsequent troubleshooting efforts until the RPIS
vendor arrived at the site to aid in the problem diagnosis. The inspectors noted that
an excellent troublesh;oting plan was developed after three days of expert
consultation, print reviews, and design analysis. Contingency measures were
planned and briefed. A test manager and engineer were exclusively assigned to
supervise the efforts. All significant station maintenance was suspended during the
'
actual RPIS work in progress. As a result, on July 10, the faulty components in the
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13
RPIS were identified and validated in the shop by bench testing, and the system
was restored to an operable condition. RPIS operated reliably through the end of
the report period,
c.
Conclusions
- Operators responded appropriately to a self revealing fault in the rod position
indication system. Good coordination between maintenance and engineering
departments was avident during system recovery efforts. Performance of
troubleshooting activities was mixed, but was ultimately successful in restoring the
system to operability in a well-controlled manner.
Ill. Enaineerina
E1
Conduct of Engineering
EL1 General Comments
The inspectors observed good overall coordination between engineering and other
Hope Creek departments during resolution of unplanned events and emergent
corrective maintenance activities. Examples included service water system
c'orrective maintenance for repeat traveling screen shear pin breakage and pump
packing replacement, rod position indication system troubleshooting, component
inservice test failures, and excessive control rod speeds. Routine and planned
evolutions also received an adequate level of engineering oversight and
participation, especially during new fuel receipt and inspection conducted
throughout the report period.
Additionally, the inspectors observed that departmental performance indicators
continued to evolve into more useful and effective tools for management decision
making. Data to evaluate the engineering department contribution to recent
increases in site wide human error rates was effectively utilized. Recent
management focus on improving the quality and number of department staff was
evidenced by the fact that several new employees including temporary contract
personnel were hired to help reduce the sizable engineering work backlog, prepare
for the upcoming refueling outage, and conduct routine system monitoring.
E2
Engineering Support of Facilities and Equipment
E2.1 Recirculation Pumo Shaft Crackino
a.
insoection Scooe (37551)
The inspectors reviewed a Hope Creek management decision not to perform
recirculation pump shatt inspections during the upcoming refueling outage (RFO), in
spite of vendor recommendations to the contrary. The inspectors interviewed
responsible licensee management md system engineering personnel during the
course of the review, and attended a NRB meeting at which the basis for this --
decision was discussed.
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14
b.
Observations and Findinas
On June 12, the Hope Creek General Manager issued a "wiilte paper" to document
a decision not to perform General Electric (GE) recommended inspections of both
]"
recirculation pump shafts during RFO7. GE issued a Service Information Letter (SIL)
No. 459 Supplement 2 in October 1991 to inform Boiling Water Reactor (BWR)
owners of observed shaft cracking in recirculation pumps manufactured by the
Byron Jackson company. This SIL recommended shaft inspections at the earliest
opportunity following 80,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of pump operation.
The inspectors learned that, at the conclusion of the current operating cycle, the
Hope Creek recirculation pumps will have been operated nearly 84,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. As
such, system engineering personnel recommended adherence to the GE inspection
guidance, particularly since the noted SIL states that several plants which
conducted these inspections observed significant cracking which forced shaft
replacements.
The inspectors questioned the PSE&G management's safety perspective with regard
to the decision not to perform the GE-recommended inspectiwa, especially given
the fact that both recirculation pump shaft seals were already planned for
replacement in tne RFO7 schedule. PSE&G management stated that the decision
was based on several factors, including: (1) plants that have had shaft failures
reported unusual vibration indications one to three days prior to failure, (2) vibration
instrumentation currently exists to detect unusual trends, (3) seven " sister" BWR-4
plants have more pump operating hourt than Hope Creek, and (4) procedures exist
to direct actions regarding pump f ailure events. Additionally, the inspectors noted
that recirculation pump shaft breakage events are analyzed occurrences in the
facility UFSAR. Further, the white paper lists several actions to be implemented to
ensure early detection of any potential pump degradation and to reinforce operator
training regarding response to pump related events. Lastly, the NRB critically
reviewed this decision in a recent meeting and were satisfied that no safety concern
was created by deferring the inspections,
c.
Conclusions
Hope Creek management's decision not to perform vendor-recommended
inspections of recirculation pump shafts (for cracking) was based on sound
engineering judgement and assessment of other industry information. A Nuclear
Review Board evaluation of this decision was appropriate.
E12 Station Service Water System Reliability and Availability
a.
Insoection Scone (37551)
The inspectors conducted a general review of the overall reliability and availability of
the SSW system. Engineering evaluations, action requests, maintenance rule
indicators, system walkdowns, and interviews with cognizant engineering personnel
were used in formulating the assessment.
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15
b.
Observations and Findinos
Several examples of SSW system reliability prob l ems surfaced during the report
period, many of which had an impact on subsyrtem availability. The inspectors
were especially interested in the quality and timeliness of PSE&G's response to
these recent issues in light of the rising trend in ultimate heat sink temperature. On
two occasions, June 1 and June 10, equipment operators in the field discovered
failures of the "B" SSW traveling screen due to broken shear pins. Both events
required unplanned entries into the technical specification 3.7.1.2 action statement
and increased overall system unavailability. Following the June 10 f ailure, the root
cause of the repeat shear pin failure was attributed to a screen misalignment (due to
wear). However, this problem was not entirely resolved in that the previously
accepted practice of rotating a screen one full revolution with the smaller diameter
test shear pin installed following maintenance could not be completed due to minor
binding.
(
Later in the period, on June 25, the "D" SSW pump packing overheated after six
hours of continuous pump operation. Maintenance technicians replaced the
packing, but it again overheated and failed within days of the work. Additionally,
technicians discovered that the "C" SSW pump packing failed to maintain shaft
leakage * 'qw an acceptable level. Several more individual SSW pump outages
were r
,ary throughout the remainder of the report period to address and
resolve pump packing degradation. System engineering personnel, in part based on
consultation with the SSW pump vendor, ultimately determined that minor
modifications to the packing material were, necessary to ensure proper operation. By
the end of the oeriod, the packing issues had been resolved, but not before a
significant amount of system unavailability time was accumulated.
A final example of system reliability problems involved repeat failures of SSW loop
vacuum breakers (two valves in parallel per SSW loop), normally energized solenoid-
operated valves which perform a specifically decribed design function in the loss of
offsite power accident sequence. The inspectors noted that these components have
repeatedly failed increased frequency inservice testing, evidence that supported a
prior PSE&G determination that the valve design was inadequata. However, in spite
of the historical poor performance, no design change package was ready (although
one is now planned) for implementation during the period to resolve the reliability
issue.
The inspectors reviewed 10 CFR 50.65 (" maintenance rule") required reliability and
availability data collected for the SSW system and noted that PSE&G had
appropriately categorized the system per section a(1) of the regulation, and
accordingly established goals for improving overall system performance. As of the
end of the report period, both indicators were " red," with unavailability time nearly
an order of magnitude greater than the performance criteria initially set for the
operating cycle.
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16
c.
Conclusions
Though PSE&G management acted promptly to resolve individual service water
system component failures, the frequency of emergent work and repeat failures was
high, which significantly increased overall system unavailability time to levels well
beyond established goals. This fact in part indicated that previous efforts to resolve
repeat equipment failures and improve system reliability have beea less than fully
effective.
E7
Quality Assuranct,in Engineering Activities
E2d Enaineerina Review of Industrv Ooeratina Exoerience
.
The inspectors observed generally good review, analysis and use of industry
operating experience (OE) during the report period. Detailed OE performance
indicators were developed and employed by station management to ensure that
potentially significant concerns were promptly addressed. Two notable examples
included the discovery of a conflict between the Hope Creek Safety Evaluction
Report and the station fuel receipt procedure regarding the use of the new fuel
storage vault. The issue was identified by engineering personnel during an
investigation of a OE information concerning 10 CFR 70.24 criticality monitor
requirements. This discovery led to a conservative decision by station management
not to use the new fuel vault until the discrepancy was resolved.
A second example, which was identified as a result of an issue at the Clin 3n
nuclear station, involved the discovery that reactor manual control system
,
transponder cards were potentially not receiving an adequate degree of post-
maintenance testing following replacement. The recirculation pump cracking
concern, addressed in detail in section E2.1 above, was also recognized following
receipt of OE information. The inspectors concluded that, based on the above and
other examples, the OE review program functioned appropriately to ensure that
significant industry lessons teamed were evaluated for applicability to the Hope
Creek station.
IV. Plant Sucoort
R1
Radiological Protection and Chemistry (RP&C) Controls
R 1.1
Liould-To-Solid Radwaste Processinc
a.
insoection Scoce (86750)
Through review of liquid radwaste processing documents, radwaste system
walkdowns, and interviews with licensee staff, the inspector reviewed Hooe Creek
Station liquid radwaste processing and resulting solids generation with respect to
requirements.
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17
b.
Observations and Findinos
Since early 1995, the Hope Creek spent resin / sludge solidification process for
conversion to a bituminous waste matrix has been shutdown. Currently, the station
,
processes the equipment and floor drain powdered resin wastes to the waste sludge
phase separator tank and the mixed bed resin wastes to the spent resin tank. The
reactor water cleanup powdered resin wastes are collected separately in the reactor
water cleanup phase separator tank. The wastes in each of these three tanks are
recirculated for mixing the wastes prior to sluicing the spent resin media to a
transport / burial liner. A sample from each liner is taken during the resin transfer for
chemistry gamma spectral analysis. The gamma-emitting constituents of the
sample are characterized directly from the chemistry sample analysis. Non-gamma
emitting radionuclides are derived from the applicable 10 CFR 61 waste stream
characterization provided by an outside laboratory analysis and included in the
waste liner radioactivity determination. The inspector verified fou. waste stream 10 CFR 61 char 4
'izations had been completed between January and February 1996
for bead resir ;,owdered resin, reactor . vater cleanup resin, and dry active waste.
s
The licensee utilizes vendor services in accordance with a vendor process control
program to dewater each waste liner prior to shipment or temporary onsite storage.
The solid radwaste burial volume for Hope Creek Station for 1996 was 3482 cubic
feet, down from 5155 cubic feet in 1995.
e
!
c.
Conclusions
'
The licensee provides an effective waste mixing, sampling, characterization and
waste form processing program. No discrepancies with regulatory requirements
were identified.
R2
Status of RP&C Facilities and Equipment
R2,1 Onsite Radwaste Storaae
a.
Insoection Scoce (86750)
The inspector reviewed the condition and radwaste storage inventory of the Hope
Creek process building and the low level radwaste storage facility (LLRWSF).
b.
Observations and Findinos
.The inspector noted an accumulation of 24 - 200 cubic foot resin liners being stored
in the LLRWSF. All of this material was generated by Hope Creek Station, in
addition, according to licensee records reviewed, there are currently 63 - 55 gallon
drums of bitumincus radwaste stored in the Hope Creek radwaste process building.
The inspector questioned the licensee about the significant amount of solid
radwaste stored at the facility (approximately 29 shipmants). The licensee indicated
that the radwaste truck bay was recently blocked during resin processing which
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18
prevented the removal of stored radwaste drums from the process building. During
the inspection, demobilization of resin liner processing activities in the radwaste
truck bay were being completed and the licensee indicated that the transfer of
drums out to the LLRWSF were targeted for July 1997. Ths licensee indicated that
financial negotiations with the low level radwaste burial f acility have recently been
completed, and plans were to resume shipping of waste this summer. Additionally,
molten metal volume reduction vendor process is expected to be used to process
low activity resin.
The inspector reviewed records indicating that quarterly surveillances of the resin
liners have been performed as specified by procedures. The inspector reviewed a
video tape of the latest LLRWSF vault inspection performed by robot on May 30,
1997. Although this method of surveillance is an excellent appicach, some
enhancements may be necessary in placement of resin liners to allow an access
path for the robot to enable complete vault inspections to be performed and
adjustment to the camera arm to allow viewing of the drainage trench located inside
the vault area. Review of the robot surveillance indicated no deterioration of the
resin liners stored in the f acility,
c.
Conclusions
The inspector noted that all of the solid radwaste was appropriately stored, locked,
and controlled by RP technicians with required facility surveillances being met, and
that the radiation levels associated with the LLRWSF were well below regulatory
requirements. Accumulated stored Hope Creek solid radwaste is targeted for a
shipment campaign for offsite processing and burial during the summer of 1997.
R2 2 Laiduo Radwaste Eauioment
a
a.
Insoection Scoce (86750)
The int:pector reviewed the status of formerly utilized radwaste processing
equipment for safe long-term lay up. This review was made with reference to the
UFSAR and interviews with licensee staff. Verification of equipment status was
made through observations in the plant,
b.
Observations and Findinas
The original Hope Creek bituminous wasto system has been abandoned since early
1995. Engineering evaluations are ongoing to determine a replacement for this
system. The licensee has determined that the original asphalt extruder will no
longer be utilized. A design change package is currently being drafted by the
licensee to cap and isolate the following abandoned equipment: two waste
evaporators, the decon solution evaporator, the crystallizer, and the two asphalt
extruders. The licensee has targeted the end of 1997 for completion of this work.
_ _ _ _ _ - _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
__
.
.
19
c.
Conclusions
The licensee has an abandoned bituminous waste manufacturing system that has
not yet been placed into a long-term laidup condition. The licensee projects this
work to be accomplished by the end of 1997. Licensee actions appear appropriate
to effect proper layup of the processing equipment. This is an inspector followup
item. (IFl 50 354/97 04 03)
R3
RP&C Procedures and Documentation
Elj, Radwaste/Transoortation Procedures
a.
Insoection Scoce (Tl 2515/133)
Due to a major revision of Title 49 CFR 171-178 effective April 1,1996, a
comprehensive procedure review was conducted at the Salem Nuclear Generating
Station.
b.
Observations and Findinas
The following procedures were reviewed:
e
Hope Creek Process Control Program, Rev. 2
Shipment of Radioactive Materials Excluding Waste for Burial, NC.RP-RW.ZZ-
e
0909(O), Rev.1
e
Shipment of Radioactive Waste for Burial, NC.RP-RW.ZZ-0906(Q), Rev.1
e
Shipment and Receipt of Laundry, NC.RP-TI.ZZ-0915(Q), Rev. 2
e
Liquid Radwaste-Filter Demineralizer System Operat. ion, HC.RW SO.HB-
0006(R), Rev. O
Solid Radwaste Asphalt Storage and Transfer System Operations, HC.RW-
e
SO.HC-0002(R), Rev. O
e
= Solid Radwaste Solidification System Operation, HC.RW SO.HC-0005(R),
Rev.0
e
Solid Radwaste-Drum Handling System, HC.RW-SO.HC-0006(R), Rev. O
Solid Radwaste Operation with Temporary Solidification / Dewatering Filtration
e
Systems, HC.RW SO.HC-0007(R), Rev. O
e
Dewatering Procedure for CNSI 24 Inch Diameter Pressure Vessels,
VHC.RW-SO.HB-0101(R), Rev.1
Operating Procedure for CNSI Liquid Processing System, VHC.OP SO.HB-
e
0102(R), Rev. 2
Setup and Operating Procedure for the RDS-1000 Unit at Hope Creek,
o
VHC.RW-SO.HC-0101(R), Rev.1
e
Bead Resin / Activated Carbon Dewatering Procedure for CNSI 14-195 or
Smaller Liners, VHC.RW SO.HC-0102(R), Rev.1
Setup and Operating procedure for Dewatering Precoat Media in a 21-300
e
Liner Using toe RDS-1000, VHC.RW-SO.HC-0103(R), Rev.1
Operating Guidelines for use of Polyethylene High Integrity Containers,
VHC.RW-SO.HC 0104(R), Rev.1
_____
_ _ _ _ _ _ _
_ _ - _ _ _ _ - _ _ _ _ _ _ - _ __
_-
,.
.
20
The Process Control Program did not specify the regulatory and buriallicense waste
processing parameters and criteria. Although the implementing radwaste
dewatering procedures provided appropriate methods, these procedures were not
defined as part of the Process Control Program and did not reference the regulations
- and burial license requirements that they were designed to meet. Most of the
above procedures were not affected by the revision in the DOT regulations.
Procedure errors were minor and were discussed with the licensee during a detailed
inspection debrief. One exception was procedure NC.RP-TI.ZZ-0915(O), Rev. 2,
" Shipment and Receipt of Laundry." This procedure did not include the new
requirement to classify LSA shipments as LSA-1, LSA II, or LSA lli, m addition, dose
conversion factors were poorly documented and referenced,
c.-
Conclusions
Radwaste transportation procedures were generally adequate, and revised DOT
regulations were effectively represented. The Process Control Program and
implementing procedure should be enhanced to ensure that waste form processing
parameters are consistently and properly controlled.
R4
Staff Knowledge and Performance in RP&C
R4.1 Radioactive Material Shiooina Documentation Review
b.
Observations and Findinas-
The inspector reviewed selected radioactive material shipping records since April 1,
1996. The shipping papers were well documented, included an independent quality
control inspection review (for all but excepted quantity shipments), and all
consignee licenses and shipping cask certificates of compliance were on file as
required. The inspector noted that all shipment records except contaminated
laundry shipment records were printed using the Radman computer code. These
shipping records were found to be compete and accurate. The contaminated
~ laundry shipping records were produced manually in accordance with procedure
NC.RP-TI.ZZ-0915(O), Rev. 2, and these shipping records failed to correctly specify
the LSA group specification on any of the laundry shipments since April 1,1996, as
required by 49 CFR 172.203(d)(11). Examples include: Hope Creek shipment nos.
9703,9708, and 9712. This is a violation of regulatory requirements. This
.
. violation was cited previously in NRC Inspection Report 50-272/97-12 and 50-
311/97-12 for PSE&G's Salem facility. The inspector also noted that the dose
conversion factors used to determine radioactivity content of the laundry shipments
were not included in the shipping papers nor was the date of derivation of these
conversion factors,
c.
Conclusions
The inspector's review of selected radioactive material shipments made during April
1
1996-1997 indicated that the shipments met regulatory requirements, except for
the laundry shipments which resuited in a violation.
. . .
. . . . ..
. . . . . . ..
.
3
0
21
R5
Staff Training and Qualification in HP&C
Bhd Radweste Transoortation Trainina for Shicoers and RP Technicians
b.
Observations and Findinos
The Hope Creek Station authorized radioactive material chippers, and QA personnel
attended a week long segulations traininc course in February of 1995 which fulfilled
regulatory training requirements, in addition, in January 1996, a one day revised
shipping regulations update workshop was well attended by PSE&G RP and QA
personnel. The RP technicians were also provided a radioactive material shipping
training course as part of continuing RP technician training during the third and
fourth quarters of 1096. Currently, the Services Group is planning for another
week long regulations training course to be held during the summer of 1997.
c.
Conclusions
The licensee met the training requirements for radioactive material shippers and RP
technicians.
R7
Quality Assurance in RP&C Activities
,
b.
Observations and Findinas
The licensee conducted a radioactive material control audit between July 22,1996
and August 13,1996. The audit included shipping activities from both Salem and
Hope Creek Stations, included offsite technical expertise and appeared to be a
thorough program review. However, the audit failed to identify the inadeqeacies in
the laundry shipment procedure. The QC orpanization continues to be active in
independently verifying all radioactive material shipments, except for the excepted
quantity shipments to ensure regulatory requirements are met.
QA audits of radwaste processing vendors were inconsistently performed with
regard to NRC Bulletin 79-19, which directs licensees to establish and implement a
management controlled audit function of all transfers, packaging at:d transport
activities.
c.
Conclusions
The licensee's audit program for the radwaste/ transportation program was found to
be adequate. The licensee stated that the vendor audit program for this program
area would be reevaluated and modified as required.
e;
e
22
R8
Miscellaneous RP&C lesues
BHJ Uodated Final Safety Analysis Reoort (UFSAR)
- a,
insoection Scoce
The inspector reviewed current Hope Creek Station practices with respect to
Sections 11.2 and 11.4 of the UFSAR.
b.
Observations and Findinos
The UFSAR continues to describe the bituminous waste processing system as well
as the resin dewatering system. When the bituminous waste processing system is
capped and isolated, Section 11,4 of the UFSAR will require updcting.
c.
Conclusions
The UFSAR description of solid radwaste processing was accurate and correctly
reflected current plant operations.
P3
EP Procedures and Documentation
'.S
E2d in-Office Review of Licensee Procedure Chanoes
An in-office review of revisions to the emergency plan and its implementing
procedures submitted by the licensee was completed. A list of the specific
revisions reviewed are included in Attachment 1 to this report. Based on the
licensee's determination that the changes do not decrease the overall effectiveness
of the emergency plan, and that it continues to meet the standards of 10 CFR 50,47(b) and the requirements of Appendix E to Part 50, NRC approval is not
-required for those changes Implementation of those changes will be subject to
inspection in the future,
P5-
Staff Training and Qualification in EP
E5d Resoirator Qualifications in Sucoort of Emeraency Preoaredness
in previous observations and inspection reports the inspectors noted some
reluctance on the part of emergency preparedness (EP) department staff to employ
PSE&G's corrective action program. This assessment was primarily derived from
the relatively few issues documented in action requests, During the current
inspection report period, the inspectors followed one self identified issue that
' indicated that use of the corrective action program was improving. Specifically,
over the course of the last few months, severallevel 3 action requests had been
initiated listing failures by specific licensee personnel assigned to emergency
response organization (ERO) positions to maintain current respirator qualifications.
As a result of a departmental self-assessment of ERO member qualifications and a
review of corrective action program trends, the EP staff initiated a significance level
.
9
.
23
1 condition report which clearly established the large scope of this particular
problem and forced increased management attention on the issue. As a result, the
inspectors noted marked improvements in the manner by which all ERO
qualifications were maintained, and an increased overall sensitivity to current
qualification status by ERO members.
S7
Quality Assurance in Security and Safeguards Activities
The inspectors reviewed the results of a recently completed QA department audit of
the site security organization and discussed the implications of the findings with QA
and security department management. As a result of this review, the inspectors
judged that QA had done a thorough job of evaluating the status of site security,
and identified severalissues which required improvement. One specific example of
a significant QA-identified issue involved the July 7 discovery that a file cabinet in
the security department training office outside the protected area, which contained
safeguards information, was unlocked with the immediate surrounding area
unmanned. QA promptly reported and appropriately documanted this concern. A
subsequent file cabinet inventory was completed which ac.aunted for all controlled
documents. As a result, security management embarked on an extensive review of
the need for this and other cabinets to reduce the number of safeguards information
repositories and to improve control of the information. The inspectors concluded
that QA department efforts in the security area positively influenced the ability of
,
the site security organization to fulfillits function effectively.
F7
Quality Assurance in Fire Protection Activities
Throughout the report period, the inspectors observed an increasing trend in the
number of self-identified issues raised by fire protection department personnel.
Following individual discussions with various department staff, the inspectors
judged this change to be attributed primarily to renewed management focus on
improved department efficiency, expectations regarding use of the corrective action
program, and adherence to established fire protection procedures. Examples of
significant self identified performance issues included: (1) recognition that numerous
fire protection system valves and components have not been entered into the
tagging request and information system, (2) documentation of a significant backlog
of fire protection related maintenance a#vities and the fact that these activities
were not included in the work week sc?edule, and (3) a specific July 3 discovery
that new fuel shipping containers (made of untreated wood) were being stored in
the reactor building during fuel receipt activities (as has been the historical practice).
Department management recognized that no ccmpensatory measures had been
established to mitigate a potential incident involving this increase in area fire
loading, and took prompt action to correct the situation. As a result of actions
taken in response to this and other self-identified issues, fire protection and safety
at the Hope Creek site was notably improved.
_. . _ _ _ _. _ . _ _ _- - -
-.. _ _ . . _ _ . . - . _ . . _ . . _ . _ _ . _ _ _ . _
. . _ . - -
.
e
i
,
i
24
V. Manaaement Meetinas
,
.X1
Exit Meeting Summary
- A recent discovery of a licensee operating their facility in a manner contrary to the UFSAR
description highlighted the need for a special focused review that compares plant practices,
procedures and/or parameters to the UFSAR descriptions. While performing the
-
i-
inspections discussed in this report, the inspectors reviewed the applicable portions of the
'
UFSAR that related to the areas inspected. The inspectors verified that the UFSAR
wording was consistent with the observed plant practices, procedures and/or parameters.
l
On July 18,1997, the inspectors presented their findings and conclusions to members of
j
licensee management. The licensee acknowledged the assessment presented.
The inspectors asked the licensee whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified.
.
4
3
4
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!
i
i
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4
5
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I'
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,_
. - -
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._
,
, _ . .
-
.
_.
- . - . .
._-
-
-
0
.
25
INSPECTION PROCEDURES USED
IP 37551:
Engineering
IP 61726:
Surveillance Observations
IP 62707:
Maintenance Observations
IP 71707:
Plant Operations
IP 71750:
Plant Support Activities
IP 86750:
Solid Radioactive Waste Management and Transportation of Radioactive
Materials
Tl 2515/133:
Implementation of Revised 49 CFR Parts 100179 and 10 CFR Part 71
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-354/97-04-02
high pressure coolant injection system feedwater
injection valve inoperable
50 354/97-04-03
IFl
disposition of two waste evaporators, decon
.,
mlution evaporator, crystallizer and two asphalt
extruders to a drained and isolated long term
layup
'
Closed
50 354/96-05 01
reactivity mismanagement event
ESF actuation
"C" SW pump auto-start
50-354/97-010
Special
overpower event due to loss of feedwater
heaters
Report
50-354/97-011
LER
failure to complete flood protection action within
required time frame
50-354/97-012
LER
control rod scram during testing
-50 354/97-04-01
failure to complete offsite power distribution
lineup
Discussed
unplanned HPCI system inoperability
i
e
g
t
.-
26
LIST OF ACRONYMS USED-
BWR=
Boiling Water Reactor-
-CRD
Control Rod Drive
-: ECCS -
Emergency Core Cooling Systern
-i
- EDG-
d
^EP
ERO-
~ Emergency Response _ Organization
'
GE -
HCU:-
Hydraulic Control Unit
-HPCI
High Pressura Coolant injection
LER
Licensee Ecant Report
1
- LLRWSF -
Low Level Radioactive Waste Storage Facility
- Low Specific Activity '.
NRB
Nuclear Review Board--
NRC-
Nuclear Regulatory Commission
OE-
Operating Experience -
Public Document Room
Preventive Maintenance
PSE&G
Public Service Electric and Gas
Quality Assurance
Refueling Outage
RHR-
- RP
Radiation Protection
- RP&C-
Radiological Protection & Chemistry Controls
RPIS ~
Rod Position Indication System .
Sll .
Service information Letter
SORC
Station Operations Review Committee
q
Senior Reactor Operator
'
' SSW
-Station Service Water
TS-
. Technical Specification
-UFSAR_
Updated Final Safety Analysis Report
WIN
-Work it Now
i;.
&
e
e
ATTACHMENT 1
REVIEWED LICENSEE DOCUMENTS -
NUCLEAR BUSINESS UNIT
EMERGENCY PLAN IMPLEMENTING PROCEDURES
EF4P 105H
Upprading Protective Action Recommendations
7, 8
EPIP 204H
En.ergency Response Callout/ Personnel Recall
35,36
EPIP 205H
TSC Post Accident Core Damage Assessment
1
EPIP 301H
RPT On-Shift Response
15
EPIP 309H
Dose Assessment
10
,
EPIP 313H
Control Point-Chemistry Response
5
EPIP 314H
Chemistry Supervisor-TSC Response
5
4
EVENT CLASSIFICATION GUIDE
Section 8.4
Control Room Evacuation
1
Section 11.7
Security / Emergency Response Capabilities
1
,
Attachment 6
Primary Communicator Log
1, 2
Attachment 7
Primary Communicator Log (GE)
1, 2
Attachment 9
Non-Emergency Notifications Reference
1, 2
,
Attachment 15
Environmental Protection Plan
1
4
Attachment 18
4 Hr Report Radiological Transportation Accident
1
Attachment 19
24 Hr Report-Fitness For Duty (FFD) Program Events
1
Attachment 24
UNUSUAL EVENT (Common Site)
1-
,
EVENT CLASSlFICATION GUIDE TECHNICAL BASIS
Section 3.1
Fuel Clad Barrier
1
,
.
Section 3.3
Containment Barrier
1
!
Section 7.1
Loss of AC Power Capcbilities
1
Section 11.2
Design Basis /Unanalyzed Condition
1
,
Section 11.7
Security / Emergency Response Capabilities
1
)
COMMON SITE
Document
Document Title
Revision
Section 3
Emergency Organization
8, 9
Section 5
Emergency Classification System
5
Section 8
Public information
5
Section 9
Emergency Facilities and Equipment
6
Section 16
Radiological Emergency Response Training
6, 7
.-
. . - -
_
.-
--
-
..
..
.
.
. - .-
. ._ -
- 9 L
-)-
o;
Attachment 1
2
ADMINISTRATLVE
-!
EPIP 1006
Emergency Equipment Inventory (Radiation Protection)
18
EPIP 1007
EOF / ENC Supply & Locker inventory
17
EPIP 1008
Emergency Communications Drills
.
14
EPIP 1013
Emergency Response Personnel Telephone List
36,37
EPIP 1016
Test Procedures for EOF Backup Generator,
3
Vent System and HVAC Filter Replacement
i
M
i
EPIP 404
Protective Action Recommendations
8, 9
1
EPIP 602
Radiological Dose Assessment
19
i
SECURITY
EPIP 902
Accountability / Evacuation
14
EMERGENCY NEWS CENTER
i
NC.EP EP.ZZ-0801(Q) Emergency News Center Operation
0
NC.EP EP.ZZ-006(O) ENC Evacuation and Activation of Back-up ENC
0,1
,
.
EPIP 801
Vold
10
4
'
EPIP 802
Vold
9
j
EPIP 803
Void
8
EPIP 804
Void -
6
EPIP 805
Vold
8
EPIP 806
Void
5
EPIP 807
Emergency News Center Telephone Directory
9,10
l
1
i
f
i
-