IR 05000354/1998007
| ML20237E501 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/25/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20237E496 | List: |
| References | |
| 50-354-98-07, 50-354-98-7, NUDOCS 9809010008 | |
| Download: ML20237E501 (41) | |
Text
- - _ _ _ _ - _ - _ _ - - - _ - _ _ _ _ _ _
.
.
U. S. NUCLEAR REGULATORY COMMISSION REGION i I
Docket No:
50-354 ~
License No:
NPF-57 l
Report No.
50-354/98-07 l
,
Licensee:
'Public Service Electric and Gas Company l
Facility:
Hope Creek Nuclear Generating Station Location:
P.O. Box 236 Hancocks Bridge, New Jersey 08038 i
Dates:
June 28,1998 - August 8,1998 l
Inspectors:
S. M. Pindale, Senior Resident inspector J. D. Orr, Resident inspector l
J. D. Noggle, Senior Radiation Specialist G. C. Smith, Safeguards Specialist P. R. Frechette, Physical Security inspector Approved by:
James C. Linville, Chief, Projects Branch 3 Division of Reactor Projects
!
I l
.
!
l i
!
i 9009010008 980825
-
PDR ADOCK 05000354 i
G PDR
'
I l
l
.
. - ___ - __ - - - _-_
_____
_ - _ _ _ _ _ _ _ _. _ - _ - _ _ - _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _
.
.
EXECUTIVE SUMMARY Hope Creek Generating Station NRC Inspection Report 50-354/98-07 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covered a six-week period of resident inspection; in addition, it included the results of announced inspections by three regional inspectors, one that reviewed radioactive waste processing and material shipping activities and two that reviewed security and safeguards activities.
Operations J
. Hope Creek operators exhibited a proper safety focus and responded appropriately during i
routine and off-normal conditions (ultimate heat sink elevated temperature, safety system
. valve failure, feedwater system transient). However, the control room operators exhibited a knowledge weakness in determining whether control room annunciators have reflash capability. (Sections 01.1 & M4.2)
l PSE&G did not properly restrain scaffold material in the torus room in accordance with.
administrative procedures. Although no equipment operability concerns were identified, this identification by the inspectors demonstrated that housekeeping in the torus room was not closely monitored by Hope Creek supervisors. (Section 02.1)
!
Due to inadequate self-checking, weak supervisory oversight, and inattention to detail, several human performance errors by operations personnel occurred. These included inadvertent bumping and actuation of components (radiation monitor and safety relief valve acoustic monitors) and manipulation of an incorrect fuse during an equipment tagout.
l PSE&G management responded appropriately to these minor self identified errors by l
initiating a common cause analysis to correct this performance weakness. (Section 04.1)
l^
l The administrative program to track operator burdens was not being consistently updated or reviewed by operators. The specific items that were in the program indicated that the actual number and plant impact of the operator burdens was not noteworthy. (Section 04.2)
Maintenance -.
[
PSE&G conducted safety-related surveillance in a safe and deliberate manner with one L
exception. Problems encountered during the 'A' rod block monitor channel calibration procedure were not understood before steps were re-performed out of sequence with the j
procedure. -(Section M1.1)
Preventive maintenance for the 'A' filtration recirculation ventilation system ventilation fan
]
was properly. conducted. '.However, some problems with control of redundant safety-related equipment were observed. The inspectors determined that PSE&G was not il j
,
_
__
_
_
___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
l
.
.
l effectively posting protected equipment nor was the protected equipment consistently j
communicated to working groups. (Section M1.2)
j
{
Due to poor communications and poor work controlimplementation, maintenance personne.1 failed to follow the established equipment tagging process and worked on a f
reactor building ventilation supply fan without the appropriate tagging and controls in place as required by station procedures. An operations supervisor demonstrated a good questioning attitude when he identified this deficiency during a plant tour. (Section M4.1)
There were two instances where plant transients resulted due to an intermittent problem with a feedwater heater relay card in the trip circuit. Maintenance technicians did not identify and correct the first occurrence of this equipment problem when a specific relay was replaced. As a consequence, operators were unnecessarily challenged a second time.
The operators responded promptly and appropriately during both transients, and plant response was normal. (Section M4.2)
Enaineerina PSE&G had not corrected a known deficiency in verifying main steam isolation valves (MSIV) to be fully closed during " springs-only" full stroke closing tests by performing a
. local position indication observation. Also, PSE&G had not completed an intended Updated Final Safety Analysis Report change related to MSIV operation and design. Although no operability issues resulted, PSE&G failed to accurately track and resolve these items in its corrective action program. (Section E2.1)
PSE&G management conducted a detailed review and assessment of system engineering, reactor engineering, and other technical support departments in response to recent deficiencies related to performance monitoring and trending. This review was sufficiently critical and identified several areas of weakness requiring correction. In response, engineering management developed appropriate short and long term corrective action plans to improve performance in these areas. (Section E7.1)
Plant Suocort Hope Creek spent resin wastes were effectively sampled, packaged, de-watered and placed in temporary storage in a condition ready for shipment and disposal. (Section R1.1)
From early 1995 until this inspection, the Hope Creek evaporators were not effectively controlled by radwaste operators to ensure that abandoned radwaste processing equipment was properly layed-up in a drained and tagged out condition. Two waste evaporators were known to have through-wall cracks and were not tagged out and one of the cracked evaporators was not drained. The abandoned equipment did not adversely impact operations and did not result in any release of materials to the environment and therefore, was not risk significant. (Section R1.2)
For approximately two years, Hope Creek har been accumulating spent resin wastes in the onsite low level radioactive waste storage facility (LLRWSF). Offsite shipments are
>
iii
_ _ _ _ _ _ _ _ _
__
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
planned to resume during the summer of 1998 and the resolution of administrative requirement discrepancies relative to onsite storage is planned. (Section R1.3)
The Hope Creek radioactive waste processing and radioactive material shipping procedures l
were of good quality and effectively implemented regulatory requirements. The Hope Creek process control program continued to reference the asphalt radwaste solidification system, which is no longer in use. (Section R3.1)
l l
The ALARA and laboratory techniques used by a Hope Creek chemistry technician while performing a reactor coolant chemistry sample were very good. (Section R4.1)
PSE&G conducted security and safeguards activities in a manner that protected public health and safety in the areas of alarm stations, communications, protected area access control of personnel and packages. This portion of the program, as implemented, met l
PSE&G's commitments and NRC requirements. PSE&G's security facilities and equipment l
in the areas of protected area assessment aids, protected area detection aids, and personnel search equipment were determined to be well maintained and reliable, and were able to meet PSE&G's commitments and NRC requirements. (Sections S1 & S2)
The review of PSE&G's audit program indicated that the audits were comprehensive in j
scope and depth, that the audit findings were reported to the appropriate level of management, and that the program was being properly administered. In addition, a review of the documentation applicable to the self-assessment program indicated that the program was being effectively implemented to identify and resolve potential weaknesses. (Section S7)
iv i
_ _
_ _ _ _ _ - _ _ _ _ _ _ _ _ _ -. _ _ _ _ _ _ _ _ _ _ - _ _ _ _ -
- __
a
- _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _. ___. - _. _
.
.
l TABLE OF CONTENTS EX EC UTIVE S U M M A RY.............................................. ii
,
i i
l-TA B LE O F C O NTENTS.............................................. v l
1. Ope ra tio n s :..................................................... 1
' 01 '
. Conduct of Operations.................................... 1 l
01.1 General O observations................................ 1
O2-Operational Status of Facilities and Equipment
...................2
'
02.1 Material Condition and Housekeeping Weaknesses........... 2 j
Operator Knowledge and Performance......................... 3
)
04.1. Multiple Personnel Errors by Operations Personnel............ 3 04.2 Operator Barriers Program............................. 5 11. M ai n te n a nc e................................................... 6 M1 Conduct of Maintena nce................................... 6 M1.1 ' Conduct of Surveillance
.............................6
!
M1.2
'A' Filtration Recirculation Ventilation System Maintenance and Protected Equipment Control........................... 7 M4 Maintenance Staff Knowledge and Performance.................. 8 i
M4.1 Maintenance Work in Progress Without Adequate Safety Tagging. 8
-
M4.2 Feedwater Heater Trip Due to Equipment Failure............ 10 j
M8 Miscellaneous Maintenance issues............................11
M8.1 (Closed) Violation 50-3 54/9 7-03-01..................... 1 1 l
M8.2 (Closed) Violation 50 3 54/9 7-07-03..................... 12
)
M8.3 (Closed) Violations 50-354/97-10-02,04, & 09.............
I l l. Eng i ne e ri ng...................................................
1 3 E2-Engineering Support of Facilities and Equipment
.................13 E2.1 Inadequate Corrective Actions for Main Steam Isolation Valve Operation and Testing
..............................13 E2.2 Evaluation of Potential Generic Pump Design Deficiency.......
E7 Quality Assurance in Engineering Activities.....................
E7.1. System Engineering Review and Assessment by PSE&G Management j
..............................................17 E8 Miscellaneous Engineering issues............................
E8.1 (Closed) LER 50-3 54/98-03........................... 18
.
E8.2 (Closed) Violation 50-3 5 4/9 7-10-0 5.....................
l V. Pla n t S u ppo rt................................................. 2 0 R1 Radiological Protection and Chemistry (RP&C) Controls............ 20 R1.1 Hope Creek Solid Radwaste Processing
..................20
!
R1.2 Review of Abandoned Solid Radwaste Processing Equipment... 20 l
R1.3 Solid Radioactive Waste Storage....................... 22 i
R3 RP&C Procedures and Documentation........................ 23 v
i
!
t
_
z________________________._________________.__1.___._
_ _ _. _ _. _ _ _ _ _ _ _ _ _ _ _. _ _ _. _
... _ _
.. _ _ _ _ _ _ _ _
_ _
.__..___J
i
.
.
i I
!
R3.1 Radioactive Material Shipment Procedures................ 23 R4 Staff Knowledge and Performance in RP&C....................24 R4.1 Reactor Coolant Chernistry Sampling and Analysis........... 24 R5 Staff Training and Qualification in RP&C,...................... 24 R5.1 Radioactive Material Shipment Training
..................24 l
R7 Quality Assurance in RP&C Activities......................... 25
,
l R7.1 Radioactive Material Shipping Audit..................... 25
'
S1 Conduct of Security and Safeguards Activities.................. 26 l
S2 Status of Security Facilities and Equipment..................... 27 S3 Security and Safeguards Procedures and Documentation........... 28 S4 Security and Safeguards Staff Knowledge and Performance......... 28 SS Security and Safeguards Staff Training and Qualification........... 29 S6 Security Organization and Administration...................... 29 S7 Quality Assurance in Security and Safeguards Activities........... 30 S8 Miscellaneous Security and Safeguards issues
..................31 S8.1 (Closed) Violation 50-354/E 97 422-01013................ 31 F8 Miscellaneous Fire Protection issues
.........................31 F8.1 (Closed) LER 50-3 54/9 7-017.......................... 31 -
!
V. M a n eg em e nt Mee ting s........................................... 3 2 l
X1 Exit Meeting Summary.................................. 32 i-
!
!
l l-I-
,
,
,
!
!
,
i
!
Vi
!
,
- - _ - - _ _
_---_-_ _ - _ _ _
__
_
---
.
.
l'
!
l.
Reoort Details l
l Summarv of Plant Status t
Hope Creek was operated at or near full power for the duration of the inspection period.
l'
l. Operations
Conduct of Operations i.
01.1 General Observations a.
Inspection Scope (71707)
Using Inspection Procedure 71707,the inspectors performed routine inspections of ongoing plant operations.
' b.
Observations and Findinas -
'
, Late in' July 1998, the temperature of the Delaware River.(Hope Creek's ultimate.
heat sink) continued a slow trend upward due to seasonably warm ambient temperatures. The inspectors verified that control room operators were performing increased monitoring every two hours in accordance with technical specification
' 4.7.1.3. On July 20,1998, ultimate heat sink temperature reached 84'F. The control room operators completed the actions for river temperature approaching 85'F required by abnormal procedure HC.OP-AB.ZZ-0122(Q), Service Water
<
System Ma// unction. Hope Creek technical specification 3.7.1.3. requires additional actions, which were properly described in the abnormal procedure, when river temperature exceeded 85 F.~ On July 27,1998, due to lowering ambient conditions, river temperature remained below 82*F..The control room operators exited the actions required by the service water system abnormal procedure, but they continued the increased river water temperature monitoring. The inspectors determined that the control room operators were sensitive to river water temperature trends and the operators appropriately applied the technical specification requirements.
On July 28,1998, operators were performing in-service valve testing on the high pressure coolant injection (HPCI) system. The HPCI condensate storage tank (CST)
motor-operated valve, BJHV FOO4, had been closed in order to stroke the HPCI suppression chamber suction motor-operated valve BJHV-F042. However, the operators were unable to re-open BJHV-FOO4 after testing BJHV-F042. The operators left the HPCI suction aligned to the suppression chamber with BJHV-F042 opened and initiated an action request to troubleshoot BJHV-FOO4. The control room operators considered HPCI operable since technical specification 3.5.1.c p
stated that HPCI was capable of taking its suction from the suppression chamber, although HPCI is normally aligned to take a suction from the CST. This configuration is consistent with its licensing basis as described in the Updated Final Safety Analysis Report. Although HPCI was considered operable, control room operators also considered it degraded with the BJHV-F004 problems. The control
_ - _ _ _ - - _ -
_ - _ _ _ _ - - _ _
--_
_
_ _
_ _ -
_--
, _ _ - - - _ _ _ _ _ - - - _ - _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ _ _.__
.
.
room operators promptly initiated an operability determination. The inspectors reviewed the operability determination and concluded that it acceptably evaluated system operability.
l On August 5,1998, the inspectors asked control room operators if a main control l
.oom overhead annunciator, HPC//RHR Area Leak Temperature High, had reflash
!
capability for other alarm inputs. The annunciator was in solid for a known condition. The operators struggled with the answer. The inspectors noted that I
control room operators struggled with the same general question for other main
- ontrol room overhead annunciators. The inspectors discussed this finding with the operations manager, who stated operators should understand the reflash capability
. of annunciators that are in solid. The operations manager further stated that he believed the majority (if not all) annunciators have reflash capability, and that he intended to have operator training provided on annunciator reflash during the operator requalification program.
l l
c.
Conclusions
!.
, Hope Creek operators exhibited a proper safety focus during routine and off norme j.
conditions. However, the inspectors determined that control room operators-l:
exhibited a knowledge weakness in determining whether control room annunciators have reflash capability.
Operational Status of Facilities and Equipment 02.1 Material Condition and Housekeeoina Weaknesses a.
Inspection Scone (71707)
The inspectors made frequent tours of the reactor building, the auxiliary building, r
and the service water intake structure, b.
Observations and Findinas In general, the material condition and housekeeping in the safety-related areas of the plant were adequate, with one notable exception. The inspectors toured the torus room and discovered that large amounts of scaffold material were not properly restrained for seismic considerations. The inspectors noted scaffold planks restrained in only one horizontal direction,55 gallon drums that were completely full of unrestrained scaffold knuckles, unrestrained scaffold tubes in storage racks, dollies unrestrained, and a chain fall and chain unrestrained in the overhead near safety-related equipment. The inspectors discussed the torus room housekeeping issues with the operations manager. The inspectors were informed by the scaffold program supervisor that scaffolds had recently been disassembled in the torus
- room. The inspectors concluded that the scaffold disassembles were not carefully supervised.- The scaffold program supervisor indicated that removal of the scaffold l
material from the torus room had been delayed. See NRC Inspection Report 50-
!
354/98-05 Section O2.1 for details about initial discovery of problems with scaffold
- _ _ _ _ _ _ - - _
_ _ _ _ _
_ _ - _ _ _ _ _ _ _ _
._
_-
.
.
l
materialin the torus room. The operations manager and scaffold program supervisor indicated that PSE&G still intended to remove most of the scaffold material and to correct the housekeeping problems in the torus room. The L
inspectors determined that the scaffold _ material that was stored and not properly j
restrained in the torus room was a violation of minor significance and is not subject
'
to formal enforcement action.
c.
Conclusions PSE&G did not properly restrain scaffold materialin the torus room in accordance
,
H with administrative procedures. Although no equipment operability concerns were identified, this identification by the' inspectors demonstrated that housekeeping in j
the torus room was not closely monitored by Hope Creek supervisors.
04'
Operator Knowledge and Performance 04.1 Multiple Personnel Errors by Ooerations Personnel l
_a.
Linsoection Scope (71707)
. The inspectors reviewed PSE&G's response and evaluation for several occurrences l
this inspection period that were caused by human performance errors.. The inspectors interviewed station personnel and reviewed relevant documentation.
b.
Observations and Findinas The following three licensee identified human performance events occurred in the operations department over a relatively short time period, e
On July 9,1998, while taking readings on the north plant vent radiation monitoring system (RMS) under the direction of a licensed reactor operator, a licensed operator trainee inadvertently pressed an adjacent flow button. This button caused the RMS sample pump to trip and therefore, made the north plant vent radiation monitor inoperable. The licensed reactor operator promptly placed the sample pump back in service and restored the radiation monitor to an operable status. PSE&G evaluated this problem and concluded that it was caused by improper use of the self-check technique (STAR: stop, think, act, review). The licensed reactor operator responsible for the licensed operator trainee was not effective in preventing this occurrence.
o On July 13,1998, while installing a tagout for a lamp test and indication circuit associated with the 'B' channel of the remote shutdown panel, an operator inadvertently removed the wrong fuse from the panel. He removed
!-
the fuse directly above the one he intended to remove. Initially, the equipment operator was experiencing difficulty in removing the correct fuse due to a space restriction between the fuse and insulating block. He discussed this difficulty with another equipment operator and a supervisor, who were involved in the tagout activity. They decided that the past
_ _ _ _ _ - _ _ _ _ _ _ _ _ -
.
.
,
practice of using a pen to dislodge one end of the fuse would allow the equipment operator to then grip the fuse with the fuse puller. However, I
when he turned back to the terminal board, he failed to self-check and he j
pulled the wrong fuse. The fuse that was pulled had minimal consequence in l
that it provided a remote shutdown indication for a switchgear room cooler l
i and was associated with the same channel ('B'). After discussion with the
)
[
field supervisor, the wrong fuse was re-installed without incident. PSE&G determined that the primary root cause of this event'was the equipment operator's failure to self-check when he re-entered the fuse panel. Two
<
significant causal factors were identified by PSE&G, including the lack of an
{
adequate tool to perform the assigned task and inadequate supervisory oversight of the evolution.
o On July 16,1998, while training a recently licensed senior reactor operator on log-taking for safety relief valves (SRV), a licensed reactor operator inadvertently depressed a button inside a control room panel that de-i energized a group of eight SRV acoustic monitors. An associated control
{
l room alarm annunciated and showed an open position for the eight SRVs.
j
.
. Operators verified that the plant was stable and all SRVs were closed.. After
'
contacting maintenance, power was restored to the eight acoustic monitors and all indications returned to normal. PSE&G determined the cause for this
!
. occurrence was poor STAR techniques and inattention to detail.
PSE&G initiate.d individual condition resolution action requests to document, evaluate and ccrrect the specific issues.= However, due to the similar nature of these occurrences and the narrow time period, PSE&G appropriately decided to look at these events collectively to perform a common cause analysis so that appropriate corrective actions are implemented. The inspectors noted that shift meetings and other communications have been instituted to discuss these issues, the apparent l
causes and management's expectations and standards for performance. PSE&G recognized that these types of errors could result in significant plant transients or challengesi which provided the bases for conducting the common cause analysis, c.
Conclusions Due to inadequate self-checking, weak supervisory oversight, and inattention to i
detail, several human performance errors by operations personnel occurred. These included inadvertent bumping and actuation of components (radiation monitor and safety relief valve acoustic monitors) and manipulation of an incorrect fuse during an equipment tagout. PSE&G management responded appropriately to these minor l
self identified errors by initiating a common cause analysis to correct this l
performance weakness.
L l
i l
l
.
.
l
l 04.2 Operator Barriers Proaram a.
Insoection Scooe (71707)
The inspectors reviewed the content of the operations department barriers lists.
The inspectors also interviewed operators to determine if the program was effectively being used.
l b..
Observations and Findinas The inspectors reviewed the operations department work around list and the concerns list. Hope Creek operations department distinguished operator barriers with two lists. The work around list are those operator barriers that may affect plant transients, and the concerns list are those barriers that may distract operatore during routine operations.
The inspectors noticed that a recent problem with position indication on the 'D'
l torus-to-drywell vacuum breaker had not been included on either list. (See NRC l.
. _ Inspection Report 50-354/98-06 Section O2.1.) The inspectors also determined -
!
through several interviews with senior reactor operators, that they were unaware if the 'D' torus-to-drywell vacuum breaker was included on the list. Each of the senior reactor operators indicated that the 'D' torus-to-drywell vacuum breaker should be considered for the work around list or the concerns list.
The inspectors determined that the work around list and the concerns list were not
!
readily accessible for the operators. Further, control room operators were unfamiliar with recent entries to the work around or concerns lists. Members of the operations staff maintained the operator barrier lists, and items are typically added as directed by operations supervisors. The inspectors reviewed both lists and verified that they were accurate and inclusive, except for the 'D' torus-to-drywell vacuum breaker indication problem.
The inspectors discussed their findings with the operations manager who acknowledged that the operations barrier program had not recently received the appropriate amount of focus. The operations manager intended to improve the application of the operations barrier program. The 'D' torus-to-drywell vacuum breaker problem was added to the operator concerns list.
c.
Conclusions The administrative program to track operator burdens was not being consistently updated or reviewed by operators. The specific items that were in the program indicated that the actual number and plant impact of the operator burdens was not noteworth i
,-
.
!
l j
!
11. Maintenance
)
M1 Conduct of Maintenance
!
M1.1 Conduct of Surveillance a.
Insoection Scope (61726. 71707)
i L
l The inspectors used inspection procedures 61726 and 71707 to observe the
- conduct and performance of surveillance by equipment operators and maintenance technicians.
'
'b.
Observations and Findinas On July 22,1998, the inspectors observed operators perform inservice testing of l.
the 'G' and 'H' emergency diesel generator fuel oil transfer pumps. The inspectors L
determined that the operators were knowledgeable of the equipment operability
[
requirements for the asso_ciated 'D' emergency diesel generator. The operators
_... were well prepared for the surveillance and careful use of the procedure ensured.
that both diesel fuel oil transfer pumps were restored to a standby lineup in an
. expeditious fashion.
On July 30,1998, the inspectors observed the performance of a high pressure l
coolant injection (HPCI) surveillance test HC.OP-IS.BJ-OOO1(O), HPC/ Pump l
Ouarterly /ST, from the HPCI pump room. Equipment operators locally monitored L
the pump and turbine operation and collected data for the surveillance test; Maplewood Laboratory technicians were present to read and record pump and
'
turbine vibration data; a radiation protection technician was present for ALARA support; and a system manager was present to monitor the HPCI system operation.
_
The inspectors determined that each assigned individual was knowledgeable of their assignments and that each person was sensitive to the fact that torus bulk
. temperature technical specification limits would only support about a fifteen minute
,
[.
operation of the HPCI system. The HPCI turbine exhausts steam to the torus and it l;
consequently raises torus bulk water temperature. Communication was well
_
established between the main control room and the equipment operators in the field.
l The inspectors questioned the Maplewood Laboratory technicians' ALARA
. practices. Although the technicians received a minimal radiation dose, about 4 or 5
_.
_ mrem each, the inspectors questioned if the vibration monitoring equipment could j
'
be located in a lower dose field or even outside the HPCl pump room if longer cables are used from each temporarily installed vibration monitoring sensor. The technicians indicated that longer cables would be technically acceptable. The inspectors later discussed this with the ALARA supervisor who indicated that the
'Maplewood technicians would even consider this dose saving technique during other vibration monitoring evolutions as well.
On July 31,.1998, the inspectors observed instrument technicians perform an
.
eighteen month channel calibration of the 'A' rod block monitor (RBM), HC.lC-CC.SE-0019(0), Channel Calibration Nuclear Instrumentation System - Nondivisional
!
I L
(.
.e.
- l
Channe/ A RodB/ock Monitor. At procedure step 5.2.2, the technicians noticed a j
disagreement between the actual 'A' RBM response and the expected response.
l The technicians verified that the 'A' RBM was in a safe condition and discontinued the surveillance; The technicians notified the main control room operators and their supervisor of the problem. The supervisor and the technicians could not exactly explain the results, but they agreed to re-perform the affected steps of the surveillance. The inspectors determined that the technicians and the supervisor continued the surveillance without receiving any guidance from a system manager
,
who may have been able to provide technical expertise.
l The instrument technicians re-performed the affected steps but received unexpected i
results in a different portion of the surveillance. The technicians recognized that proper test conditions were not established and the procedure had not been performed in sequence. The surveillance was stopped by the technicians who restored the 'A' RBM to a normal alignment. Failure to perform the procedure in sequence constituted a violation of minor significance and was not subject to formal enforcement action.
. Technical expertise within PSE&G was subsequently contacted and an on the spot.
.,
change was incorporated into the 'A' RBM channel calibration procedure. PSE&G engineers determined that test citages established early in the procedure may drift slightly and later effect results. The on the spot change verified that the test voltages were properly set immediately before performance of the affected step.
The 'A' RBM channel calibration procedure was completed on August 1,1998, without any further complications.
c.
Conclusions PSE&G conducted safety-related surveillance in a safe and deliberate manner with one exception. Problems encountered during the 'A' rod block monitor channel calibration procedure were not understood before steps were re-performed out of sequence with the procedure.
M1.2
'A' Filtration Recirculation Ventilation System Maintenance and Protected Eauioment Control a.
Insoection Scoce (62707)
The inspectors observed preventive maintenance being performed on the 'A'
filtration recirculation ventilation system (FRVS) ventilation unit. Methods employed to administratively control or protect redundant equipment were also inspected, b.
Observations and Findinas The inspectors observed mechanical maintenance technicians perform a 36 month fan inspection on the 'A' FRVS ventilation fan. The technicians were j
knowledgeable about the tasks involved. The inspector determined that the work l
order and procedure were being properly used.
l l
!
i I
.
.
The inspectors noticed that the 'B' FRVS ventilation fan room was being used as an area to store dollies and tools for maintenance activities. As identified in the maintenance plan, PSE&G had established the 'B' FRVS ventilation fan as protected equipment while the 'A' FRVS ventilation fan was inoperable. The inspectors notified the work control supervisor. The dollies and tools were promptly removed l
and the mechanical maintenance superintendent was notified by the work control
[
supervisor.
l The inspectors also noticed that PSE&G had inconsistently placed signs in some l
areas identifying protected equipment areas. The 'B' FRVS ventilation fan room l
was not physically identified. Many doors to equipment areas were posted, but in fact the areas were not intended to be controlled as protected equipment areas.
l-The inspectors discussed the equipment left unattended and unrestrained in the 'B'
l FRVS ventilation fan room, and also their general observations of control and posting protected equipment areas, with the mechanical maintenance superintendent and the operations manager. The mechanical maintenance superintendent and the operations manager stated that better communication to the working groups and better control of postings needed to be established. The mechanical maintenance superintendent and operations manger intended to improve identification of protected equipment.
c.
Conclusions
,
Preventive maintenance for the 'A' filtration recirculation ventilation system ventilation fan was properly conducted. However, some problems with control of redundant safety-related equipment were observed. The inspectors determined that PSE&G was not effectively posting protected equipment nor was the protected equipment consistently communicated to working groups.
M4 '
Maintenance Staff Knowledge and Performance
~ M4.1 Maintenance Work in Proaress Without Adeauate Safety Taaaina a.
Inspection Scoce (62707)
The inspectors reviewed PSE&G's actions following their identification that maintenance workers wera conducting work on a reactor building ventilation fan without the' appropriate tags and controls in place. The inspectors interviewed maintenance personnel, reviewed relevant procedures and documentation, and assessed PSE&G's evaluation and corrective actions.
l.
b.
Observations and Findinas On July.15,1998, during a plant tour, an operations supervisor questioned maintenance workers regarding work activities on reactor building ventilation supply fan 18-VH3OO. The operations supervisor determined that the fan was not properly tagged out of service in accordance with the safety tagging program. The maintenance workers stated that they believed the required tags were in place when i
__
_-__
_ _ _ - _ _._- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
I
,
.
l
9 i
in fact the tags had been temporarily released for periodic test runs of the
ventilation fan by different maintenance workers. Upon discovery, operators
!
stopped the work, ensured the tags were subsequently re-applied, and documented j
the problem (Action Request 980715235)to initiate an investigation.
l l
Electricians were conducting work on the fan using a group taggout. During that j
work, the electricians temporarily released the tags several times in accordance with procedure NC.NA AP.ZZ-0015(O), Safety Tagging Program, to periodically run the
,
!
. fan. After the electricians completed these fan runs on July 15,1998, mechanical maintenance was to install the belts and perform an alignment. Since the
mechanical maintenance supervisor knew this work would continue into the p
following day (he was going to be off the next day), he arranged to have his i'
mechanics work for the electrical maintenance supervisor, who was already signed on the taggout. PSE&G's Action Request follow-up review identified that the mechanical maintenance supervisor did not want to take the time to develop and i
!
process a new taggout, and then turn it over to another mechanical maintenance supervisor within a few hours. While this practice is not prohibited by NC.NA-l AP.ZZ-0015(Q), this transfer of responsibility was not properly communicated or.
I controlled.
!:
PSE&G's evaluation found that the electrical supervisor informed the mechanical
supervisor that the tags were being re installed, but the electrical supervisor mis-understood from the communication that the tags had already been re-installed.
-
-Subsequently, the mechanical supervisor conducted a pre-job brief with the mechanics (although he had transferred responsibility to the electrical supervisor)
and instructed them to perform the work that afternoon (July 15). The electrical supervisor was not aware that the mechanics were assigned to the job that afternoon nor that the mechanical supervisor retained some responsibility for the
. mechanics' work (conducted pre-job brief).
!
The control switch for the supply fan was in the off position. However, the fan could have automatically started upon receipt of a low flow signal from either of the i
l.
two other operating supply fans. NC.NA-AP.ZZ-0015(O) requires that the first line
'
. supervisor (responsible individual) ensure the adequacy of blocking points for the work performed, and signify verification of the physical placement of the safety tags that indicate proper equipment isolation from all energy sources. In addition, NC.NA-AP.ZZ-0015(O) requires the job technician to obtain a pre-job brief from the
!
job supervisor to include identification of job boundaries and work scope, physical verification of the tags, and that testing for absence of energy sources be witnessed.
!
The inspectors reviewed PSE&G's response and evaluation of this event. The (
operations supervisor demonstrated a good questioning attitude when he identified L
,this issue. Maintenance initiated prompt actions, including the development of a
[
plan to present this issue and lessons learned to the maintenance department. In addition, this event will be considered for overall corrective actions as part of a broader common cause analysis as related to human performance (see Section 04.1 of this report). This issue, which was a violation of NRC requirements, will be
, _ _ - _ _ _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _
_ - _ _ _ - _
_ _ _ _ _ -
.
.
dispositioned when PSE&G completes its common cause analysis and corrective actions. (eel 50-354/98-07-01)
j c.
Conclusions
!
Due to poor communications and poor work control implementation, maintenance personnel failed to follow the established equipment tagging process and worked on a reactor building ventilation supply fan without the appropriate tagging and
!
' controls in place as required by station procedures. An operations supervisor
!
demonstrated a good questioning attitude when he identified this deficiency during a plant tour.
I
!
M4.2 Feedwater Heater Trio Due to Eauioment Failure
!
I a.
Insoection Scope (71707,62707)
l l
The inspectors reviewed PSE&G's response and follow-up after feedwater heaters tripped on July 9 and 24,1998. The inspectors reviewed plant parameters and documentation, interviewed operations and maintenance personnel, observed initial l
investigative efforts, and observed portions of feedwater heater restoration.
b.
Observations and Findinas On July 9,1998, while Hope Creek was operating at 100% power, the '2B'
feedwater heater tripped, followed by a trip of the downstream '3B' and '4B'
feedwater heaters. Operators entered the appropriate abnormal operating procedures related to a loss of feedwater heating and promptly reduced power to 78% power. After the plant was stabilized at this power level, operations and maintenance personnel initiated an investigation to determine the cause for the heater trips.
The investigation concluded that the feedwater heater trip was not due to any of
' '
the modification work that was in progress in the turbine building (near the feedwater heater panels). Maintenance troubleshooting conducted following the event concluded that a relay associated with the '2B' heater caused the trip, and the relay was replaced. After relay replacement, operators returned the unit to full power operation. PSE&G determined that the '3B' and '4B' heaters tripped due to
the feedwater system transient following the '2B' heator trip.
On July 24,1998, the '2B' feedwater heater tripped again with the plant operating at full power. No other feedwater heaters tripped following this occurrence.
L Operators reduced plant load to 95% power. After this occurrence, maintenance technicians installed diagnostic monitoring equipment in the '2B' feedwater heater panel to assist troubleshooting efforts. The technicians found that a relay card in the trip circuit was bad and was the cause for this (and probably the July 9)
transient. Maintenance technicians replaced the card, and the feedwater system was restored to normal on July 25. Operators then returned the unit to full power operation.
!
I
_
_
_ _ _ - _
_
_ _ _ _ _
__
.
.
j PSE&G initiated an Action Request to document and evaluate the July 9 and 24, 1998, feedwater heater transients. The evaluation determined that the maintenance troubleshooting conducted following the July 9 occurrence was inadequate because only a symptom was treated, not the cause of the problem. PSE&G management used this deficiency as an example of a troubleshooting weakness, and discussed this issued with maintenance personnel. The inspectors confirmed that the plant response for both transients was normal.
c.
Conclusions l
There were two instances where plant transients resulted due to an intermittent problem with a feedwater heater relay card in the trip circuit. Maintenance
.
technicians did not identify and correct the first occurrence of this equipment problem. As a consequence, operators were unnecessarily challenged a second time. The operators responded promptly and appropriately during both transients, and plant response was normal.
l M8 Miscellaneous Maintenance issues M8.1 -(Closed) Violation 50-354/97-03-01: Maintenance Technician Procedure Problems +
a.
Inspection Scope (92902)
The inspectors verified through on-site inspection that PSE&G's corrective actions
{
+
for maintenance violations were appropriate and that selected corrective actions
-
were completed.
,
b.
Observations and Findinas
l The subject violation involved two examples of maintenance technician failure to
!
follow written procedures. On May 22,1997, maintenance technicians performed
. preventive maintenance on a service water pump discharge check valve without referring to the governing procedure, in addition, the maintenance procedure did l:-
not contain specific acceptance criteria for the check valve inspection. The second l
violation occurred on May 17,1997, when maintenance technicians failed to
!
properly restore to service a high pressure coolant injection minimum flow valve differential pressure transmitter after a calibration activity.
The inspectors verified that the maintenance procedure, HC.MD-GP.ZZ-OO46(Q),
was revised to contain specific acceptance criteria for check valve inspections. The inspectors had also observed service water pump discharge check valve maintenance activities subsequent to this violation (see NRC Inspection Report 50-354/98-02)and determined those activities to be properly controlled in accordance with preventive maintenance procedures. NRC inspectors had also observed a maintenance in-Service Day on March 18,1998, and determined that the training presented provided relevant information to improve the performance of maintenance crews. This violation is closed.
l, i
b t
-
- _ - - - - _ - - -
-
,
- _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _
_ - - _ _ _ -.
.
.
c.
Conclusions Corrective actions for two examples of procedure violations involving maintenance j
technicians were appropriate and timely.
l M8.2. (Closed) Violation 50-354/97-07-03: Maintenance Technician Procedure Problems l
'
a.
Insoection Scope (92902)
The inspectors verified through on-site inspection and in-office review that PSE&G's corrective actions for maintenance violations were appropriate and that selected corrective actions were completed.
l-b.
Observations and Findinas The subject violation involved two examples of maintenance technician failure to follow written procedures. Technicians deviated from a procedure written for a 4160 Vac vital bus relay test on September 18,1997, and technicians performed a feedwater system flow transmitter calibration check and adjustment without completing the applicable sections of the governing procedure.
-
l The inspectors verified that a revision had been completed to enhance the l
procedure used during vital bus relay testing. The inspectors also interviewed the l
relay maintenance department supervisors and determined that appropriate training l
- had been completed and had reinforced PSE&G's standards for procedure l
adherence. The inspector verified through in-office review that PSE&G's corrective l
actions for the feedwater system flow transmitter calibration problems were
~
!
appropnate. This violation is closed.
l l
c.
' Conclusions -
!
Corrective actions for two examples of procedure violations by maintenance technicians were appropriate and timely.
M8.3 (Closed) Violations 50-354/97-10-02,04, & 09, Three Examoles of Failures to l,
. Either Establish or imolement Reauired Procedures a.
Insoection Scooe (92902)
The inspectors performed an onsite inspection and reviewed corrective actions described in PSE&G's response to the Notice of Violation for maintenance related procedure problems, b.
Observations and Findinas The three examples involved a failure to follow procedure instructions for a gain adjustment on a safety-related chiller pressure control valve controller, a failure to provide adequate guidance in a procedure for resetting the reactor core isolation
!
-
.__
_ _ _ _.
_ _ _ _
_ _. - - - - - - - - - _ - _ - - _ - - - - -, _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _,
!
{
r
cooling (RCIC) turbine overspeed device, and acceptance criteria in a power range neutron monitoring system (PRNMS) electrical protection assembly (EPA) channel
- calibration procedure had been recently revised, but it conflicted with technical specification requirements.
The inspectors verified that PSE&G had completed appropriate procedure revisions for the RCIC turbine overspeed device and the PRNMS EPAs. The inspectors determined that PSE&G had also addressed generic implications for technical specification related maintenance procedures that had recently been revised after
!
the Hope Creek Technical Surveillance improvement Program. PSE&G used its disciplinary policy to address the human performance issues with the safety-related chiller pressure control valve controller problems.
I c.
Conclusions
)
l PSE&G appropriately completed corrective actions for maintenance procedure
'
l problems. Generic implications were considered and applied.
I 111. Enaineerina E2 Engineering Support of Facilities and Equipment
!
l E2.1 Inadeouate Corrective Actions for Main Steam isolation Valve Ooeration and Testino l
l a.
Insoection Scooe (37551)
l The inspectors reviewed the design basis information for main steam isolation
!
valves (MSIVs) and reactor coolant safety / relief valves (SRVs) to determine if l
PSE&G had developed adequate testing to ensure continued operability.
b.
Observations and Findinas i
The inspectors verified that PSE&G had developed leak rate testing on the SRV
'
accumulators to ensure continued operability of the SRVs after a loss of power or loss of pneumatic supply event. The acceptable leak rate values were consistent with the design basis of the SRVs. The inspectors verified that PSE&G had also
,
'
developed procedures and incorporated into its test program a qualitative leak check of the MSIV accumulators. The inspectors determined that the qualitative leak check compared to an actual leak rate measurement for the MSIV accumulators was appropriate.
MSIVs are equipped with two independent energy sources for closure. Springs are compressed in the valve open position and stored air from an accumulator is available to the top of the MSIV piston cylinder during a close cycle. During normal operation a pneumatic supply is available to open and close the MSIV through its piston / cylinder arrangement.
l
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
The inspectors reviewed General Electric Nuclear Energy Service information Letter (SIL) No. 477 dated December 13,1988. GE Sll No. 477 was issued to Hope l
Creek, as well as other General Electric boiling water reactors (BWR), to inform
)
'
BWR owners that an MSIV had failed to close completely during normal plant operation after its air supply line became disconnected from the MSIV. The MSIV's failure to close completely with springs alone (without assistance from its accumulator air supply) was determined to be related to a recent maintenance activity that repacked the MSIV. Excessive tightening of the MSlV's gland flanges introduced excessive friction forces and prevented the MSIV from closing with spring force alone. GE SIL No. 477 was intended to alert BWR owners to this potential maintenance induced failure and the importance of an adequate " springs-only" test after similar maintenance activities or for testing. " Springs-only" testing of the MSIV would stroke the valve full closed with air vented from its associated accumulator.
GE SIL No. 477 also described a related concern in that primary containment over-
,
pressure during an accident could adversely affect the closing force achieved by the MSIV located within the primary containment. GE Nuclear Energy's recommended l
five actions for BWR owners were summarized as follows:
1.
Review packing chamber maintenance practices to assure that the valve stem is not subjected to excessive friction forces;
-2.
Verify that " springs-only" testing is performed using actual stem travel and not position switches. Closed position switches are typically set at 90%
closed and do not verify full closure coincident with the weakest part of the spring during stem travel; 3.
Perform a force balance calculation considering design basis accident (DBA)
containment overpressure to ensure that MSIV performance is not adversely affected; I
l.
l 4.
Test MSIV accumulators for leak tightness; and 5.
Modify applicable licensing basis documents (specifically the Updated Final Safety Analysis Report) so that the MSIV closure capability is consistent with the plant's actual closure capability during a DBA.
!
'
The inspectors reviewed PSE&G's follow up corrective actions to GE SIL No. 477.
The inspectors determined that PSE&G had appropriately addressed recommended actions Nos.1,3, and 4 listed above.
The inspectors reviewed Hope Creek test procedure, HC.OP-IS.AB-0103(Q)- Rev.
7, MSIV Loss of Power - Cold Shutdown -Inservice Test, for MSIV " springs-only" testing and determined that PSE&G did not locally verify that the MSIV was full closed as was recommended in GE SIL No. 477. The inspectors also verified that PSE&G had established its MSIV closed position switch indication at 90% of full closed. The inspectors reviewed PSE&G's corrective action program response to
_ - _ _ - _ _ _ _
.
i l
l
l l
GE SIL No. 477. Performance improvement request (PIR) No. 960604255 was the latest corrective action item in PSE&G's corrective action program to track-resolution of GE SIL No. 477. PSE&G had initiated its review of GE SIL No. 477 in
.
March 1989 under its older corrective action program system. The inspectors
!
determined that there were conflicting recommendations on how to resolve verifying the MSlV full closed. On September 19,1996, corrective action item No. 9 of PIR i
960604255lndicated that the method for verifying full closed had not yet been resolved. However, when the corrective action item to resolve all the issues l
associated with GE Sll No. 477 was closed out on May 11,1998, an adequate l
method for verifying valve full closure had not yet been incorporated into the
'
" springs-only" test. The inspectors determined that PSE&G's f ailure to complete its intended corrective action and incorporate a suitable method of verifying MSIV full closure during " springs-only" testing is a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action. (VIO 50-354/98-07-02)
l After the inspectors discussed this concern with PSE&G, the licensee initiated a l
new action request (AR 980810241)on August 10,1998, to readdress the technical issues in GE SIL No. 477. PSE&G completed an operability screening and considered that the MSIVs had not been locally verified full closed during " springs-only" testing. The operability screening concluded that the MSIVs were still completely operable. The operability determination was based on an actual MSIV friction force measurement that was performed during Hope Creek's latest refuel outage and an engineering calculation that used the friction force value to determine MSIV closure capability. The inspectors found this operability determination acceptable.
The inspectors also determined that PSE&G had initiated an Updated Final Safety Analysis Report (UFSAR) change as recommended by GE SIL No. 477. However, the UFSAR change was not yet approved or completed. The UFSAR change was submitted on December 11,1996, by Hope Creek engineers to PSE&G's Engineering Assurance Department for processing. A 10CFR50.59 applicability review, and not a 10CFR50.59 safety evaluation, was used to process the UFSAR change. Licensing engineers in PSE&G's Engineering Assurance Department determined that a 10CFR50.59 safety evaluation was necessary for the UFSAR change. The UFSAR change was rejected and returned to the valve engineering group to include a 10CFR50.59 safety evaluation. The inspectors determined that the UFSAR change and 10CFR50.59 safety evaluation were still not yet completed.
However, PSE&G had already closed its corrective action item for the UFSAR change under PIR 960604255once the UFSAR change was initially submitted. The corrective action item was closed out on the basis that the UFSAR administrative change process would ensure its completion. Based on PSE&G's recent operability determination, the inspectors found PSE&G's delay in completing a 10 CFR50.59 l
safety evaluation did not have any impact on plant safety. The inspectors determined that the failure to complete the appropriate UFSAR change was a violation of minor significance to 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, and is not subject to formal enforcement. The inspectors determined that under AR 980803115,PSE&G intended to submit a 10CFR50.59 safety evaluation to process the UFSAR change.
w__-____-___-_______________-
-
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
PSE&G had already self-identified examples of other corrective action items that had not been properly closed out. PSE&G initiated an evaluation on August 4,1998, (AR 580804142)to improve its method for ensuring proper completion of corrective action items. The inspectors determined that this AR was an appropriate corrective
action for the MSIV problems.
c.
Conclusions
'
The NRC inspectors identified that PSE&G had not corrected a known deficiency in verifying main steam isolation valves full closed during " springs-only" full stroke closing tests. The NRC inspectors also identified that PSE&G had not completed an intended Updated Final Safety Analysis Report change related to MSIV operation and design. The inspectors concluded that in part, PSE&G's failure to accurately track these items in its corrective action program, allowed the problems to remain for about nine years. These problems are NRC identified violations of 10 CFR 50,
,
L Appendix B, Criterion XVI, Corrective Action.
l l
E2.2 Evaluation of Potential Generic Pumo Desion Deficiency l
a.
Insoection Scooe (37551)
j i
The inspectors reviewed PSE&G's evaluation of a 10 CFR Part 21 report that was
)
l related to a potential safety hazard due to broken cast iron suction heads in a l
certain type of safety related pumps.
b.
Observations and Findinos
,
By letter dated July 9,1998, a vendor (Ingersoll-Dresser Pump Company) submitted
. a 10CFR21 report to the NRC related to APKD type pumps, whose cast iron suction heads could be subject to breakage during operation. The suction head is a cast l
iron, structural component that holds a bearing below the suction impeller. The bearing holder is connected to the outer portion of the suction head with several support ribs.
'
l The vendor submitted this report because of an event at another nuclear power
plant (Limerick Nuclear Generating Station) identified in May 1998. Specifically, l
while investigating clogging of one of the reactor recirculation system Jet pumps, a
!
piece of metal was found inside of the reactor vessel which was traced to the b
residual heat removal (RHR) pump. A subsequent remote visual inspection (boroscope) of the RHR pump, an APKD type design, identified that a portion of the suction head was missing. Follow-up investigation at Limerick revealed that an earlier failure of a non-safety related APKD type condensate pump experienced a suction head failure in December 1997, however, the pump may have been i
operated in the failed condition with no noticeable performance degradation.
The 10CFR21 identified that several nuclear facilities have been provided APKD type pumps for safety related applications, including the Hope Creek RHR pumps (4)
and core spray pumps (4).
..
l
_
_ _ - _
____ _ - _ ___ _____ _ _______ _ _ _ __ _ _ _
l.
- l PSE&G promptly evaluated the July 9,1998, report to determine whether the RHR
)
l or core spray pumps could experience similar damage. Maintenance engineers l
interfaced with the vendor and determined that, while the Hope Creek pumps'
i suction heads are similarly made from cast iron, the design is different. Specifically, l
the Hope Creek pumps have a pinned suction head design to the pump shell, which l
would positively contain the shaft sleeve and shaft upon a failure of the cast iron suction head. Therefore, PSE&G and the vendor concluded that, due to the design l
difference, Hope Creek would not experience a failure similar to the one that i
occurred at Limerick.
l Notwithstanding the design difference, PSE&G plans to perform periodic routine pump inspections, which would identify the existence of this type of suction head degradation, i
c.
Conclusions Maintenance engineers promptly and appropriately evaluated a 10 CFR Part 21 report that was related to a potentially generic safety-related pump design deficiency.
E7 Quality Assurance in Engineering Activities i
E7.1 - System Enaineerina Review and Assessment by PSE&G Manaaement
'
a.
insoection Scope (37551)
The inspectors reviewed PSE&G's plans to address recently identified system -
engineering performance weaknesses related to system monitoring and trending.
The inspectors interviewed system engineering management and reviewed system engineering performance monitoring for selected systems and parameters, b.
Observations and Findinas During recent NRC inspections, performance weaknesses were apparent regarding system monitoring and trending by the Hope Creek system engineers. Specifically, monitoring and trending deficiencies for safety related batteries and fuel oil storage for the emergency diesel generators contributed to delays in identifying and correcting adverse performance trends, and consequently, resulted in significant challenges to various station departments, in response to these deficiencies, system engineering management has implemented several short term corrective actions. In addition, PSE&G has been evaluating overall system engineering issues and developing plans to improve performance.
The inspectors discussed system engineering issues with the Director, System l
Engineering; who had evaluated the various departments, including Salem and Hope Creek System Engineering, Reactor Engineering, inservice Inspection / Testing, and Probabilistic Safety Assessment / Maintenance Rule. The Director, System Engineering assessed the performance of each department, identified specific
_ - - _ _
_- _ _ _ - _ - _ _ _ _ _ _ _ - _ _ - _ _. - _ - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - - _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ -
_ _ _ _ _ _ _ _ _ - _ _ _ _
_
..
i
.
department issues, and identified proposed issue resolutions. The inspectors noted that significant efforts were underway to self-identify and correct performance weaknesses.
One of the focus areas in PSE&G's self-assessment was performance monitoring.
Within the next few months, system engineers will be expected to develop a system performance monitoring notebook, and to develop a " system health report,"
which would illustrate system performance monitoring status. The first system health reports are expected to be issued in early November 1998.
The inspectors noted some improved performance relative to system monitoring.
Specifically, the inspectors reviewed trend date that the de system manager had i
been collecting for the Hope Creek safety-related batteries, in addition to data collected from technical specification required weekly and monthly surveillance, the system manager had required individual battery cells that exhibited early signs of degrading performance to be measured on at least a weekly basis. That trend data had been used to make timely decisions for starting battery equalizer charges or individual cell replacements. The inspectors determined that increased monitoring of cells number 27 and number 47 on the BD411 battery allowed the electrical maintenance department to prepare and replace the individual cells before they would have fallen below technical specification table 4.8.2.1-1 category C limits. Increased monitoring of cell number 22 on the CD447 battery allowed an equalizer charge to be promptly started to restore its individual cell voltage to above technical specification table 4.8.2.1-1 category B limits.
j c.
Conclusions
,
PSE&G management conducted a detailed review and assessment of system i
engineering, reactor engineering, and other technical support departments in response to recent deficiencies related to performance monitoring and trending.
This review was sufficiently critical and identified several areas of weakness
!
requiring correction. In response, engineering management developed appropriate i
short and long term corrective action plans to improve performance in these areas.
E8 Miscellaneous Engineering issues E8.1 (Closed) LER 50-354/98-03: Concurrent inocerability of the 'A' and 'B' Control Room Chillers: The inspectors performed an in-office review of licensee event
' report (LER) 50-354/98-03. The inspectors verified that the LER was accurate, timely, and consistent with the inspection details described in NRC Inspection Report 50-354/98-05 Section E2.1. No new information was presented in the LER.
NRC Violations 98-05-01,-02 & -03 will remain open to examine PSE&G's corrective action items in response to the Notice of Violation. This LER is closed.
L
_ _ - - _ _ - _ _ _ _ _ _ _ _ - _ - _ _ - _ _ - _ _ - _ _ _
_ _ _ - - _ _
,
.
E8.2 (Closed) Violation 50-354/97-10-05: Reactor Core isolation Coolina System Governor Valve Stem Bindina a.
Insoection Scooe (92903)
.
The inspectors performed an onsite inspection and reviewed corrective actions l
described in PSE&G's response to the Notice of Violation for a reactor core isolation cooling (RCIC) system governor valve stem binding failure.
. b.
Observations and Findinas PSE&G replaced the RCIC governor valve stem in March 1997 with a stem manufactured of a different material. The replacement stainless steel material was used to address a previous binding problem caused by corrosion of the original equipment governor valve stem material. The design change package that installed the replacement governor valve stem failed to account for a different thermal expansion during RCIC system operation. On December 5,1997, the RCIC governor valve stem bound causing the RCIC turbine to overspeed during testing.
PSE&G determined that the binding was thermally induced from the replacement governor valve stem.
PSE&G subsequently re-sized the carbon spacers that surround the RCIC governor l
valve stem. Further testing of the RCIC system was successful.
E PSE&G determined that the root cause for the RCIC stem binding was personnel error in the development of the design change package. The corrective actions developed ensured that lessons learned from this event were communicated to engineering personnel and that maintenance procedures were updated to require a stem to-spacer _ clearance check.
l'
The inspectors verified that PSE&G completed its corrective action item for revising
!'
the vendor manual. The inspectors also verified that PSE&G had completed an appropriate procedure revision to the maintenance procedure, HC.MD-CM.FC-
0002(O) - Rev. 9, Reactor Core Isolation Cooling (RCIC) Turbine Steam Stop and Governor Valve Overhaul. The procedure revision required a physical clearance measurement between the governor valve steam and each carbon spacer. The inspector verified through interviews that PSE&G had communicated the lessons l.
learned from this event to engineering personnel.
!
c.
Conclusions PSE&G adequately implemented corrective actions for a reactor core isolation cooling system governor stem valve failur. _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
.
.
IV. Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Hope Creek Solid Radwaste Processina a.
Inspection Scooe (86750)
Plant tours were conducted to review the solid radwaste processing practices with respect to UFSAR descriptions and radwaste sampling, characterization, and waste classification requirements. These tours included the radwaste control room and the area outside of the radwaste drumming aisle used for resin de-watering activities.
b.
Observations and Findinas Hope Creek liquid radwaste processing results in spent bead resins and powdered resin sludge that are collected in a spent resin storage tank. Before sluicing, the tank contents are recirculated to ensure thorough mixing and then the spent resin is sampled and sluiced to a polyethylene liner and de-watered in the Hope Creek drumming aisle. During the inspection, a liner of spont resin was filled, de-watered, capped and transferred from the process building to the Low Level Radioactive Waste Storage Facility for temporary storage. The resin wastes were sampled during the sluicing operation and the de-watering procedure was effectively followed to ensure there was less than 1 % free standing water in the resin liner.
RP personnel effectively surveyed the liner and controlled personnel exposures and i
access during the transfer operation. Placing the resin liner in storage was remotely performed in the Low Level Radioactive Waste Storage Facility. Final building j
radiation and contamination surveying were performed as required following the waste placement work evolution.
c.
Conclusions Hope Creek spent resin wastes were effectively sampled, packaged, de-watered and placed into temporary storage in a condition ready for shipment and disposal in accordance with UFSAR requirements.
R1.2 Review of Abandoned Solid Radwaste Processina Eauioment a.
Insoection Scope (86750)
Hope Creek was designed to centrifuge waste resins and evaporate and crystallize l
waste liquids then to be mixed with hot asphalt resulting in a bituminous stable l
waste form. The production of bituminous wastes was discontinued permanently in early 1995. Two waste evaporators, a decon solution evaporator, and a crystallizer have been abandoned since that time. Walkdowns of this solid radwaste processing equipment was performed and interviews were conducted with i
f i
-_ ______ ___ -
_ _ _ _ - _ _
_ _ _ _ _
_-
.
.
I
.
'
applicable radwaste operations and chemistry personnel. This review was
'
performed to ascertain the current condition of this abandoned equipment.
b.
Observations and Findinas The history time line of the subject radwaste equipment indicates that during waste evaporator startup in early 1995, both waste evaporators were known to have been cracked; and after an April 1995 radioactive release from the decon solution evaporator, all of the evaporators were shutdown and have not been used since.
Late in 1995, some draining took place and tagouts were placed on the steam supply to the subject radwaste equipment. During 1996, a procedure was written
,
to lay-up all of the evaporators, however, the procedure was never approved or
!
implemented. During this inspection, the Hope Creek General Manager indicated
that the asphalt waste solidification system including the evaporators, was to be capped and isolated from existing plant systems through a design change, which is planned for 1999. (IFl 50-354/97-04-03).
Tours of the decon solution evaporator and crystallizer equipment cubicles did not a
reveal any outward signs of equipment deterioration. Tours of the two waste
.
r evaporators indicated through-wall cracking in both vessels with dry waste residue -
located on the lower portions of the vessel walls, bottom head piping and the floor.
Contamination levels of the waste residue were 100,000-150,000dpm/100 cm8
'
The 'A' waste evaporator was 40 mR/hr at contact and the 'B' waste evaporator was 300 mR/hr at contact.
!
Further investigation determined that the steam supply had been tagged out, however, the waste inlet and waste evaporator drain valves had not been tagged j
out. The 'A' waste evaporator was drained with the waste inlet valve closed and the drain valve opened. The 'B' waste evaporator valve alignment was reversed,
!
with the waste inlet valve open and the drain valve closed, it was determined that the 'B' waste evaporator contained an unknown amount of liquid. In the event any liquid may have leaked onto the floor, it would be collected in the floor drain system and processed through the liquid radwaste system.
in summary, neither waste evaporator had been tagged out of service. The 'B'
,
l_
waste evaporator was not drained, had significant through-wall cracking and the
'
waste inlet valve was open. The radwaste operations supervisor initiated an Action
[
Request and began corrective actions 1) to drain the "B" waste evaporator and
!
identify all valves associated with both evaporators (approximately 80) and 2) to tag out the system. In addition, the decon solution evaporator and the crystallizer are to be reviewed next to ensure that equipment was properly drained and tagged, as necessary.
I c.
Conclusions From early 1995 until this inspection, the Hope Creek evaporators were not effectively controlled by radwaste operations to ensure that abandoned radwaste processing equipment was properly layed-up in a drained and tagged out conditio _ _ _ _ - - - _ _ - _ - _ _ _ - _ _ _ _
_ _ _ _ - _ - _ - _ _ _ - _ _ _ _ _ _ _ - _ _ - _ - _ _ _ _.
.
.
I l
Two waste evaporators were known to have through-wall cracks and were not properly tagged out and one of the cracked evaporators was not drained. The abandoned equipment would not adversely impact operations and would not result in any release of materials to the environment and therefore, was not risk significant.
R1.3 Solid Radioactive Waste Storaae a.
Insoection Scope (86750)
l The inspectors toured Hope Creek plant areas to observe the condition of radioactive material storage areas. The Hope Creek and Salem common Low Level Radioactive Waste Storage Facility (LLRWSF) conditinn was also reviewed and l
included observations of a transfer of radwaste from Hope Creek to this facility.
!
,
I b.
Observations and Findinas During this inspection, there were limited amounts of stored contaminated equipment which were properly stored and posted in the plant. There was an o-inventory of 50 55-gallon drums of bituminous waste stored inside the Hope Creek >
l radwaste process building.
l I
'
PSE&G transferred a de-watered resin liner from the Hope Creek drum capping aisle into a shielded cask, and then to the LLRWSF by truck, where a remotely operated crane moved the resin liner into the storage vault. The inspectors observed these operations and found them to be effectively conducted and in accordance with procedural requirements. Inventory of the LLRWSF indicated that there were 34 individual shipments of Hope Creek generated radwaste backlogged in the facility
.
and an additional 3-4 shipments of bituminous waste still stored inside the radwaste I
processing building. Through a review of shipment records, a time line of shipping
'
history was determined. From May 1996 through July 1998, spent resin liners have been accumulating in the LLRWSF.
Procedure NC.RP-TI ZZ-0930(Q), Rev.1, /nterim Low LevelRadioactive Waste Transferand Storage, Section 3, states that when an offsite disposal facility is available, all radwaste containers will be shipped for disposal and not stored onsite.
During a recent Quality Assurance audit (July 1998, No.98-152), PSE&G identified the procedure discrepancy, which was documented in the corrective action system for cause determination and corrective actions. Corrective actions include:
reducing the current two year backlog of stored wastes (with significant financial resources allocated during this inspection) to begin offsite resin shipments during the summer of 1998; and resolving the procedure discrepancy to accommodate the current storage practice. Due to the low safety significance of the onsite radwaste
storage issue, the procedure discrepancy constitutes a violation of minor significance and is not subject to forrnal enforcement.
l
-
_ _ _ _ - _ _ _ _ - _ _ _ - - _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _
- _ _ _ _ -
.
.
I i
l
l c.
Conclusion
,
l l
For approximately two years, Hope Creek has been accumulating spent resin wastes
in the onsita low level radioactive waste storage facility (LLRWSF). Offsite I
shipments are planned to resume during the summer of 1998 and the resolution of l
administrative requirement discrepancies relative to onsite storage is planned.
l
.
R3 RP&C Procedures and Documentation
!
R3.1 Radioactive Material Shioment Procedures a.
Insoection Scope (86750)
,
i
.
The following procedures were reviewed with respect to DOT and NRC radioactive l
material transportation regulations.
l j
- Hope Creek Process Control Program, Rev. 2
- NC.RP-RW.ZZ-0906(O), Rev. 2, Shipment of Radioactive Material
- NC.RS-RW.ZZ-0911(O), Rev. O, Use of the 14-210or 14-215 Radioactive Material Shipping Package
- NC.RPTl.22-0930(0), Rev.1, Interim Low LevelRadioactive Waste Transfer and Storage
- HC.RP-Ti.ZZ-0803(O), Rev. 3, Collection, Segregation and Disposition of Dry
?
Active Waste and Used Protective Clothing l
- HC.RP-TI.ZZ-0804(O), Rev. 9, Labeling and Control of Radioactive Material l
.
- HC.RP-Tl.ZZ-0901(O), Rev. 4, Receipt and Inspection of Radioactive Material
- HC.RP-Tl.ZZ-0902(O), Rev. 3, Radioactive Waste Sampling and Classification I
- HC.RP-Tl.22-0913(0), Rev. 7, Preparation of Laundry for Shipment l
l b.
Observations and Findinos
'
The Hope Creek Process Control Program (PCP) still includes process controls for
,
(
the production of bituminous waste products, although this practice was abandoned l
over 3 years ago. It also provides for the resin dewatering practice that is currently in use. PSE&G indicated that when the asphalt waste solidification process is permanently isolated from plant equipment, the process control program will be updated.
Several previous radioactive material transportation procedures have been effectively combined and organized into a common procedure. Also, radioactive waste sampling and shipping procedures have been upgraded to indicate that they implement the station process control program. No significant procedure discrepwcies were identified.
c.
Conclusions The Hope Creek radioactive waste processing and radioactive material shipping procedures were of good quality and effectively implemented regulatory
,.
__ __ - - __ _ _- _ - - - ___-_ - - ___ - - _ _ _ _ -__-_-__ __ __ _ _ _ _ _
.___ _ _______ ______
\\
l^
i.
l
- g l
i requirements. The Hope Creek process control program continues to reference the asphalt radwaste solidification system, which is no longer in use.
R4 Staff Knowledge and Performance in RP&C R4.1 Reactor Coolant Chemistrv Samolina and Analysis
)
J a.
Inspection Scope (71750)
i
'
The inspectors observed a chemistry technician collect and analyze a reactor coolant chemistry sample for ionic concentration.
l b.
Observations and Findinas The inspectors observed a chemistry technician collect and analyze a reactor
'
,
.
coolant chemistry sample for chloride, nitrate, and sulfate ionic concentrations. The
!
chloride analysis was for technical specification surveillance requirements. The
' technician appropriately used the Hope Creek sampling procedure and he
. conscientiously applied ALARA principles..The technician was very proficient with-
.
'the laboratory equipment. The sample results were compared with previous sample results and the results were as expected. The inspectors determined through l
questioning, that the chemistry technician was familiar with technical specification
,
L chemistry requirements and other industry adopted chemistry guidelines adopted by
[
PSE&G.
l
'
l'
c.
Conclusions j
The inspectors considered the ALARA and laboratory techniques used by a Hope Creek chemistry technician while performing a reactor coolant chemistry sample to be very good.
l R5 Staff Training and Qualification in RP&C R5.1 Radioactive Material Shioment Trainina
,
a.
Insoection Scoce (86750)
ll
.The inspectors reviewed radioactive material shipping lesson plans and training attendance documentation, and interviewed cognizant licensee individuals with E
respect to 49CFR172 Subpart H and NRC IE Bulletin No. 79-19.
b.-
Observations and Findinas For both Hope Creek and Salem Stations, radioactive material shipments were accomplished by two authorized shippers while eight quality verification inspectors were available to provide independent reviews of each outgoing radioactive shipment (except for excepted package shipments). Training records were verified to be within three years for all 10 individuals. The licensee's in-house training
_ - - _ _ _ _ - - _ _ _ _ _ _ _ _ _ - _ _ _ - - _ _ _ - _ _ - _ _ _.
- _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _
--
- - - - _
._-___ _ _
. _ _ _ _ _ _ _ _
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - -
.
.
I l
'
l program contained only limited DOT regulation content, however, the Manager of l
Technical Services indicated that this training course would be alternated with a l
l-vendor course covering all DOT and NRC radioactive shipment regulations on a j
biennial basis.
'
c.
Conclusions
!
All principal radioactive shipment personnel have fulfilled the required DOT and NRC l
training requirements.
R7 Quality Assurance in RP&C Activities R7.1 Radioactive Material Shiocino Audit a.
Insoection Scope The licensee completed a radioactive material shipping program audit on July 7,
'
!
1998, and a draft document of this audit was reviewed. In addition, radioactive l
waste processing and transport vendor audits were reviewed in accordance with IE
)
'
b.
Observations and Findinos Quality Assurance audit No.98-152 consisted of two outside utility technical specialists and a team of auditors. The audit was of good depth and provided effective follow through on findings. Several offsite vendors supply transfer, packaging and transport of the licensee's radioactive waste and fall within the audit responsibilities of IE Bulletin 79-19. These include: Molten Metal Technology, Frank Hake, Hittman Transportation Services, Tri-State Motor Transport, and Kindrick Trucking. PSE&G's recent radioactive material shipping program QA audit identified a long standing issue with respect to vendor audits; and as a result of the program audit, corrective actions have been taken to ensure that the applicable l
radwaste vendors are periodically audited. All of the above mentioned vendors l
have either been audited by Nuclear Utilities Procurement Issues Council (NUPIC) or are scheduled for audits within the next six months.
!
Each radioactive material shipment with radioactivity exceeding excepted package I
concentrations require an independent review of the shipment by a quality l
verification inspector. Documentation of each radioactive shipment verification was
observed in each shipment record reviewed during the inspection.
t c.
Conclusions l
Quality assurance oversight of the radioactive material shipment program was effective through performance of a program audit and through independent shipment verifications.
l l
!
,
l
-_ __-__-
_ _ _
'
.
.
S1 Conduct of Security and Safeguards Activities a.
Insoection Scope (81700)
The inspectors determined whether the conduct of security and safeguards activities met the licensee's commitments in the NRC-approved security plan (the j
Plan) and NRC regulatory requirements. The security program was inspected during the period of August 3-6,1998. Are a inspected included: access authorization program; alarm stationg; communications; and protected area (PA) access control
- of personnel and packa'g"s',
s b.
Observations and Findinas Access Authorization Proaram. The inspectors reviewed implementation of the access authorization (AA) program to verify implementation was in accordance with applicable regulatory requirements and Plan commitments. The review included an evaluation of the effectiveness of the AA procedures, as implemented, and an
examination of AA records for 10 individuals. Records reviewed included both L
persons who had been granted and denied access. The AA program, as implemented, provided assurance that persons granted unescorted access did not
'-
constitute an unreasonable risk to the health and safety of the public. Additionally, L
. the inspectors verified, by reviewing access denial records and applicable procedures, that appropriate actions were taken when individuals were denied L
access or had their access terminated. Those actions included the availability of a i
formalized process that allowed the individuals the right to appeal the licensee's
.
'
decision.
Alarm Stations. The inspectors observed operations of the Central Alarm Station i
(CAS) and the Secondary Alarm Stations (SAS) and verified that the alarm stations were equipped with appropriate alarms, surveillance and communications capabilities. Interviews with the alarm station operators found them knowledgeable j
of their duties and responsibilities. The inspectors also verified, through observations and interviews, that the alarm stations were continuously manned,
'
independent and diverse so that no single act could remove the plants capability for detecting a threat and calling for assistance, and the alarm stations did not contain any operational activities that could interfere with the execution of the detection, assessment and response functions.
Communications. The inspectors verified, by document reviews and discussions with alarm station operators, that the alarm stations were capable of maintaining
continuous intercommunications, continuous communications with each on duty
'
security force member (SFM), and alarm station operators were testing i
communication capabilities with the local law enforcement agencies as committed
'
to in the Plan.
Protected Area Access Control of Personnel and Hand-Carried Packaoes. On August 4 and 5,1998, during peak activity periods, the inspectors observed -
personnel and package search activities at the personnel access portal. The
.
_ _ _ _ - - _ _ _ _
.
.
inspectors determined, by observations, that positive controls were in place to ensure only authorized individuals were granted access to the PA and that all personnel and hand carried items entering the PA were properly searched.
I c.
Conclusions PSE&G was conducting its security and safeguards activities in a manner that protected public health and safety and this portion of the program, as implemented, met the licensee's commitments and NRC requirements.
]
S2 Status of Security Facilities and Equipment a.
insoection Scoce (81700)
The inspectors reviewed PA assessment aids, personnel search equipment and testing, maintenance and compensatory measures.
b.
Observations and Findinas
Assessment Aids. On August 4,1998, the inspectors evaluated the effectiveness'
'
of the assessment aids, by observing, in both the CAS and SAS, on closed circuit television (CCTV), a SFM accompanied by an inspector conducting a walkdown of the PA. The assessment aid picture quality and zone overlap were generally good.
However, the inspectors noted that two monitors in the CAS and two monitors in the SAS were not functioning. When questioned by the inspectors about the impact the inoperable monitors had on alarm assessment, the operators were able to demonstrate that supplemental monitors could be used to view the alarm zone usually covered by the non-functioning monitors. The licensee indicated that a maintenance request had been initiated for the inoperative monitors. The inspectors verified that on August 5,1998, the inoperative monitors in both alarm stations had been replaced with new monitors.
Personnel and Packaae Search Eauioment. The inspectors observed both the
routine use and the weekly performance testing of the licensee's personnel and package search equipment in both the warehouse and the personnel access portal.
The inspectors determined, by observations and procedural reviews, that the search equipment performed in accordance with licensee procedures and Plan
'
commitments.
Testina, Maintenance and Comoensatorv Measures. The inspectors reviewed testing and maintenance records for the previous six months. The records indicate l
a good working relationship with both Instrumentation and Controls and Maintenance, as evidenced by the minimal requirement for compensatory measures due to repairs being accomplished in a timely manne. _ _ _ _ _ _ - _ - _ _ _ _._ ___ _ _ __-_-_
_ _ _. -______ _ __- _ ___ _ _ _ -
.
c.
Conclusions PSE&G security facilities and equipment were determined to be well maintained and reliable and were able to meet the licensee's commitments and NRC requirements.
S3 Security and Safeguards Procedures and Documentation a.
Inspection Scoce (81700)
i The inspectors reviewed implementing procedures and security event logs.
b.
Observations and Findinos Security Proaram Procedures. The inspectors verified that the procedures were consistent with the Plan commitments, and were properly implemented. The verification was accomplished by reviewing selected implementing procedures associated with PA access control of personnel and packages and the testing and
,
maintenance of personnel search equipment.
Security Event Loas. The inspectors reviewed the Security Event Logs for the previous six months. Based on this review, and discussion with security i
management,it was determined that the licensee had recently implemented a new j
program to track and trend loggable events. The new program was adequate for its intended function.
c.
Conclusions Security and safeguards procedures and documentation were being properly
implemented. Event Logs were being properly maintained and used to analyze,
)
track, and resolve safeguards events.
i S4 Security and Safeguards Staff Knowledge and Performance a.
inspection Scope (81700)
!
The inspectors reviewed security staff requisite knowledge.
!
b.
Observations and Findinas l
Security Force Reauisite Knowledae. The inspectors observed a number of SFMs in the performance of their routine duties. These observations included alarm station j
l operations and personnel and package searches. Additionally, the inspectors I
l interviewed SFMs and based on the responses to the inspectors' questioning, l
determined that the SFMs were knowledgeable of their responsibilities and duties, i
and could effectively carry out their assignments, i
'
l
--
.
(
--_ _ _ - _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ -.
.
.
!
c.
Conclusions The SFMs adequately demonstrated that they have the requisite knowledge necessary to effectively implement the duties and responsibilities associated with their position.
S5 Security and Safeguards Staff Training and Qualification a.
Insoection Scone (81700)
Areas inspected were security training and qualifications and training records.
I b.
Observations and Findinas Security Trainina and Qualifications. On August 4,1998, the inspectors randomly
{
selected and reviewed T&Q records of 10 SFMs. Physical and requalification i
records were inspected for armed and supervisory personnel. The results of the review indicated that the security force was being trained in accordance with the approved T&Q plan, in addition, on August 5 and 6,1998, the inspectors observed classroom training and firing range instruction of SFMs. The training was well presented, and was effective.
. Trainina Records. The inspectors were able to verify, by reviewing training records, that the records were properly maintained, accurate and reflected the current i
qualifications of the SFMs.
l c.
Conclusions Security force personnel were being trained in accordance with the requirements of
,
'
the T&Q plan. Training documentation was properly maintained and accurate and l
the training provided by the training staff was effective.
S6 Security Organization and Administration a.
Insoection Scoce (81700)
The inspectors reviewed management support, effectiveness and staffing levels.
.
I b.
Observations and Findinas l
Manaaement Sucoort. The inspectors reviewed various program enhancements l
made since the last program inspection, which was conducted in November 1997.
l These enhancements included the procurement of a new weapons (handguns, shotguns and rifles), upgrades to the security Access Authorization, Fitness and l
'
Training facilities and initiation of work on the installation of a new security computer.
I
1 l..
_ _ _ _ _ _ - _.
__ __
_
__
_
.
.
Manaaement Effectiveness. The inspectors reviewed the management organizational structure and reporting chain. The Security Manager's position in the organizational structure provides a means for making senior management aware of programmatic needs. Senior management's positive response to requests for equipment, training and resources, in general, has contributed to the effective administration of the security program. However, at the inspectors' request, the
,
licensee ran a card history on first line supervisors to determine the amount of time t
spent in the alarm stations providing oversight. The results of the card history run indicated that the time spent by first line supervisors in the alarm stations was minimal. Licensee management stated that the amount of time indicated on the card histories did not meet management expectations and that this issue would receive management attention.
l Staffina Levels. The inspectors verified that the total number of trained SFMs l
immediately available on shift met the requirements specified in the Plan, c.
Conclusions
.
.The level of management support was adequate to ensure effective implementation of the security program, and was evidenced by the hiring of additional security force members and the allocations of resources to support programmatic needs.
S7 Quality Assurance in Security and Safeguards Activities a.
Insoection Scope (81700)
The inspectors reviewed audits, problem analyses, corrective actions and effectiveness of management controls.
l b.
Observations and Findinas Audits. The inspectors reviewed the 1998 QA audit of the security and fitness-for-duty (FFD) program, conducted May 18-29,1998, (Audit No. A98-031/035). To enhance the effectiveness of the audits, both audit teams included independent technical specialists.
The inspectors determined that all audit findings were minor and not indicative of programmatic weaknesses, and that implementation of corrective actions for the findings would enhance program effectiveness. Inspectors' discussions with l
security management revealed that the responses to the audits were not yet I
complete.
i Problem Analyses. The inspectors reviewed data derived from the security department's self-assessment program. The self assessment program has been j
recently revised and the inspectors review disclosed that the revised program I
should provide meaningful data. The process to provide feedback information from the program to the implementing organizations such as training and operations was l
still being refined at the time of the inspection.
l l
.
.
,
.
Corrective Actions. The inspectors reviewed corrective actions implemented by the licensee in response to the QA audits and self-assessment programs. The corrective actions that had been completed were determined to be effective.
Effectiveness of Manaaement Controls. The inspectors observed that the licensee has programs in place for identifying, analyzing and resolving problems. They include the performance of annual QA audits, a departmental self-assessment program and the use of industry data such as violations of regulatory requirements identified by the NRC at other facilities, as a criterion for self-assessment, c.
Conclusions A review of PSE&G's audit program indicated that the Security and FFD security audit was comprehensive in scope and depth, that the audit findings were reported to the appropriate level of management, and that the program was being properly administered, in addition, a review of the documentation applicable to the self-
. assessment program indicated that the program, although recently implemented, i
was adequate to identify and resolve potential weakness.
f S8 Miscellaneous Security and Safeguards issues
'
S8.1 (Closed) Violation 50-354/E 97-422-01013: Safeauards Event Loa Not Maintained
~ Comolete and Accurate: This violation involved a willful failure to maintain complete and accurate information associated with an entry in the safeguards event log, and the violation was common to Salem and Hope Creek. NRC Inspection Report 50-272 & 311/98-06 for PSE&G's Salem station closed this issue (Section S8.1). Therefore, this item is also closed for Hope Creek.
F8 Miscellaneous Fire Protection issues F8.1 (Closed) LER 50-354/97-017: Missed Emeraencv Liahtina Surveillance a.
Inspection Scooe (92700)
The inspectors performed an onsite review of the subject licensee event report (LER), and independently verified selected corrective actions for completion, b.
Observations and Findinas j
On July 31,1997, PSE&G quality assurance personnel discovered that Hope Creek had not performed surveillance testing on its battery powered emergency lighting.
The surveillance testing was required by Hope Creek's fire protection program as described in the Updated Final Safety Analysis Report (UFSAR) and NFP-57 License Condition 2.C.7. The details of this problem and the Non-cited Violation are describad in NRC Inspection Report 50-354/97-05(Section F2.2).
The inspectors verified that PSE&G had completed all required surveillance testing for all the Appendix R battery powered emergency lights. The inspectors also
. _.
_ - _ _ _ - _ _
-_ _ ____-___ __
. _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ - _ _ _ _ _ - _ _
, _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _
.
.
'
verified that the emergency light surveillance tests were brought into Hope Creek's normal administrative process for scheduling technical specification requirement surveillance.
The inspectors verified that PSE&G had developed a preventive maintenance procedure that would reasonably assure that all the battery powered emergency lights would remain operable. The inspectors determined that recent revisions to the surveillance tests procedures adequately implemented the test requirements described in the UFSAR. This LER is closed.
c.
Conclusions PSE&G's corrective actions for missed battery powered emergency lighting unit surveillance were adequate and ensured the emergency lighting ud s would remain -
t in'a functional condition. This licensee identified problem was previously dispositioned as a Non-Cited Violation in NRC Inspection Report 50-354/97-05.
V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on August 17,1998. The licensee acknowledged the findings presented.
.The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified, t
.
l l
l l-
_ _ - _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ -
.
.
INSPECTION PROCEDURES USED IP 37551:
Onsite Engineering IP 61726:
Surveillance Observations IP 62707:
Maintenance Observations lP 71707:
Plant Operations IP 71750:
Plant Support Activities IP 81700:
Physical Security Program for Power Reactors IP 86750:
Occupational Radiation Exposure IP 92901:
Followup - Plant Operations IP 92902:
Followup - Maintenance IP 92903:
Followup - Engineering IP 93702:
Prompt Onsite Response to Events at Operating Power Reactors ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-354/98-07-01 eel Maintenance work in progress without adequate safety tagging. (Section M4.1)
50-354/98-07-02 VIO Inadequate corrective actions for MSIV operation and testing. (Section E2.1)
l Closed 50-354/97-03-01 VIO Maintenance technician procedure problems. (Section M8.1)
50-354/97-07-03 VIO Maintenance technician procedure problems. (Section M8.2)
50-354/97-10-02 VIO Failure to either establish or implement required procedures. (Section M8.3)
50-354/97-10-04 VIO Failure to either establish or implement required procedures. (Section M8.3)
50-354/97-10-05 VIO Reactor core isolation cooling system governor valve j
stem binding. (Section E8.2)
l
[
50-354/97-10-09 VIO Failure to either establish or implement required procedures. (Section M8.3)
50-354/E 97-422-01013 VIO Safeguards event log not maintained complete and accurate. (Section S8.1)
w_____-_______-___________________
. _ _ _ _ _ _
.
_._._______ ___ ___ ___ __ _ _______ _ _ _ __ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _,
.
.
-
50-354/97-017'
LER Missed emergency lighting surveillance. (Section F8.1)
50-354/98-03 LER Concurrent inoperability of the 'A' and 'B' control room chillers. (Section E8.1)
Discussed 50-354/97-04-03 IFl Layed-up radwaste equipment to be capped and isolated in 1999. (Section R1.2)
.,
i f
!
!
,
. _ -. _ _. _ _ _ _ _. _ _ _ _ _ _ _. _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _. _ _ - _ _ _ _. _ _ _ _ _ _ _ _
_w
-
!
o
.
LIST OF ACRONYMS USED
AA Access Authorization
]
AR Action Request
!
BWR Boiling Water Reactor CAS central alarm system CCTV closed circuit television
{
CST Condensate Storage Tank j
DBA Design Basis Accident DOT U.S. Department of Transportation j
FFD fitness-for-duty FRVS Filtration Recirculation Ventilation System HPCI High Pressure Coolant injection LER Licensee Event Report LLRWSF Low level radioactive waste storage facility MOV Motor Operated Valve MSIVs Main Steam isolation Valves-NRC Nuclear Regulatory Commission NUPIC Nuclear Utilities Procurement issues Council PA protected area PCP Process control program PDR Public Document Room PIR Performance improvement Request PSE&G Public Service Electric and Gas QA quality assurance RBM Rod Block Monitor RHR Residual Heat Removal RMS Radiation Monitoring System J
RP&C Radiological Protection and Chemistry
'
l SAS secondary alarm system SFM security force member SIL Service Information Letter SRVs Safety / Relief Valves T&Q training and qualification the Plan NRC-approved physical security plan UFSAR-Updated Final Safety Analysis Report
!
!
I