ML20207S417
| ML20207S417 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 03/12/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20207S410 | List: |
| References | |
| 50-354-85-98, NUDOCS 8703190414 | |
| Download: ML20207S417 (72) | |
See also: IR 05000354/1985098
Text
, ,
.
ENCLOSURE
SALP BOARD REPORT
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
_
SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE
INSPECTION REPORT 50-354/85-98
PUBLIC SERVICE ELECTRIC AND GAS COMPANY
HOPE CREEK NUCLEAR GENERATING STATION
ASSESSMENT PERIOD: NOVEMBER 1, 1985 - NOVEMBER 30, 1986
BOARD MEETING DATE: JANUARY 28, 1987
.
h31}gg f5bb y4
.
f @
TABLE OF CONTENTS
PAGE
I. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . 1
A. Purpose and Overview. . . . . . . . . . . . . . . . . . . 1
8. SALP Board Members. . . . . . . . . . . . . . . . . . . . 1
C. Background. . . . . . . . . . . . . . . . . . . . . . . . 1
II, CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
I I I . S UMMARY O F RES U LT S . . . . . . . . . . . . . . . . . . . . . . 11
3.1 Overall Facility Evaluation . . . . . . . . . . . . . . . 11
3.2 Facility Performance. . . . . . . . . . . . . . . . . . . 12
IV. PERFORMANCE ANALYSIS . . . . . . . . . . . . . . . . . . . . . 13
A. Plant Operations. . . . . . . . . . . . . . . . . . . . . 13
8. Radiological Controls and Chemistry . . . . . . . . . . . 17
C. Maintenance . . . . . . . . . . . . . . . . . . . . . . . 21
D. Surveillance. . . . . . . . . . . . . . . . . . . . . . . 23
E. Emergency Preparedness. . . . . . . . . . . . . . . . . . 26
F. Security and Safeguards . . . . . . . . . . . . . . . . . 28
G. Outages . . . . . . . . . . . . . . . . . . . . . . . . . 32
H. Preoperational and Startup Testing. . . . . . . . . . . . 34
I. Licensing Activities. . . . . . . . . . . . . . . . . . . 38
J. Training and Qualification Effectiveness. . . . . . . . . 41
K. Assurance of Quality. .................. 45
V. SUPPORTING DATA AND SUMMARIES. . . . . . . . . . . . . . . . . 49
A. Investigations and Allegations Review . . . . . . .... 49
B. Escalated Enforcement Actions . . . . . . . . . . . . . . 49
C. Management Conferences. . . . . . . . . . . . . . . . . . 50
D. Licensee Event Reports. . . . . . . . . . . . . . . . . . 51
TABLES
Table 1 - Tabular Listing of LERs by Functional Area ....... 53
Table 2 - LER Synopsis ...................... 54
Table 3 - Inspection Hours Summary ................ 60
Table 4 - Enforcement Summary . . . . . . . . . . . . . . . . . . . 61
Table 5 - Inspection Report Activities .............. 64
Table 6 - Unplanned Automatic Scrams and Shutdowns ........ 69
, ,
I. INTRODUCTION
A. Purpose and Overview
The Systematic Assessment of Licensee Performance (SALP) is an
integrated NRC staff effort to collect observations and data on a
periodic basis and to evaluate licensee performance. The SALP
process is supplemental to the normal regulatory processes used to
ensure compliance to NRC rules and regulations. It is intended to be
sufficiently diagnostic to provide a rational basis for allocating l
NRC resources and to provide meaningful guidance to licensee
management in order to improve th. quality and safety of plant
operations.
An NRC SALP Board, composed of the staff members listed in Section B
below, met on January 28, 1987 to review the collection of performance
observations and data in order to assess the licensee's performance
at the Hope Creek Generating Station. This assessment was conducted
in accordance with the guidance in NRC Manual Chapter 0516,
" Systematic Assessment of Licensee Performance". A summary of the
guidance and evaluation criteria is provided in Section 2.0 of this
report.
This report is the SALP Board's assessment of the licensee's safety
performance at the Hope Creek Generating Station for the period
November 1, 1985 through November 30, 1986. The summary findings and
totals reflect a thirteen month assessment period.
B. SALP Board Members
Chairman
W. Kane, Director, Division of Reactor Projects
Members
W. Johnston, Deputy Director, Division of Reactor Safety
P. Eselgroth, Chief, Projects Branch No. 2, DRP
L. Bettenhausen, Chief, Operations Branch, DRS
L. Norrholm, Chief, Reactor Projects Section 28, DRP
R. Borchardt, Senior Resident Inspector, Hope Creek
E. Adensam, BWR Project Directorate, NRR
D. Wagner, Licensing Project Manager, NRR
Other Attendees
D. Allsopp, Resident Inspector, Hope Creek
R. Gallo, Chief, Reactor Projects Section 2A, DRP
R. Summers, Project Engineer, Section 28, DRP
M. Shanbaky, Chief, Facilities Radiation Protection Section, DRSS
T
t
. - . . _ . . . . . _ _ . -- - - - .-
7 ej
2-
'
W. Lazarus, Chief, Emergency Preparedness Section, DRSS
L. Wink, Reactor Engineer, Test Programs Section, DRS
4
R. Keimig, Chief, Safeguards Section, DRSS
C. Background
C.1 Licensee Activities
j. The licensee began the evaluation period with construction
3 activities essentially complete and preoperational testing
approximately 32% complete. On December 2, 1985, the scheduled
initial fuel load date was revised from December 2, 1985 to
t
February 15, 1986, and overall project completion responsibility
i was transferred from the Vice President - Engineering to the
!
Vice President - Nuclear. The Assistant General Manager for
! Hope Creek operations was assigned the duties of project com-
! pletion manager. Many of the more significant preoperational
i tests were conducted during the months prior to fuel load. In
,
addition to the system testing conducted during the early part
of this assessment period, significant licensee resources were
L dedicated to the completion of administrative functions. :These
r- administrative functions included the writing and issuance of
station administrative procedures, surveillance test procedures,
maintenance procedures, and department operating procedures. A
large effort was also directed toward reducing the number of
- . outstanding NRC inspection items.
On April 11, 1986, Facility Operating License NPF-50 was issued
! to PSE&G authorizing operation of the reactor at power levels
i not to exceed 5% power. Fuel load activities commenced on
l April 15. Except for a two day delay caused by a faulty refuel-
4
ing bridge power supply cable, fuel loading progressed without
i a major delay until it was completed on April 27, 1986.
l On May 2, 1986, an alert was declared when offsite power was
,
lost to the four vital buses and only two of four emergency
diesel generators (EDG) were available for loading. Although
! only 2 of 4 EDGs are required to satisfy Technical Specifica-
tions in operational conditions 4 and 5, the emergency
classification guide.left no room for interpretation and
,
required the declaration of an alert. The licensee made the
r required notifications, restored power to the vital buses and
-
terminated the alert within one half hour.
j Initial criticality was achieved on June 28, 1986, and was
!. followed by a full core shutdown margin demonstration and a
!
source range non-saturation demonstration,
i
i
!
\
i
. . - , _ y ,,,,---_-m._m_,_,,,..____.,_m,,,, _.,,,,,,.,___.._,..,_m_, - . _ . . - _ .
.. _ . _ . . _ . _ . __. . .. _ _ _ _ _
-
.
. ,- .,
.;
'
3
i
10n June 29, 1986, the reactor scrammed on high intermediate
- range' monitor (IRM) flux. The reactor was in a non-coincidence
i'
reactor protection system logic mode at the-time (shorting links
removed). . Having just completed the necessary tests to install
,
the' shorting links, the reactor.had been placed in a sub-
!
P
critical condition. Due to decreasing neutron' counts, the
operator was downranging'IRMs. The' operator intended.to down-
1 range IRM "B", however he incorrectly selected IRM "D", which
4
then exceeded the RPS trip point and a reactor scram resulted.
[ The licensee reinstalled the shorting links, restarted the
- reactor and resumed startup program testing.
.
F
The licensee manually scrammed-the reactor on June 30, 1986, to
repair the reactor manual control system (RMCS), which had been
- inserting a continuous rod motion block for unknown reasons.
- The licensee and General Electric representatives diagnosed the '
! problem as a failed RMCS power supply. The licensee completed
1 RMCS power. supply replacement and the unit went critical on July
1, 1986.
1
,
On July 4, 1986, the reactor scrammed during-heatup for power
L ascension testing. The scram occurred when an average power
- range monitor (APRM) channel "E" high ~ upscale neutron trip was
coupled with a half scram manually inserted due to' narrow range
- level perturbations. The shift carried out the scram procedure
j and the plant was placed in a shutdown condition. The APRM'
5
channel "E" high upscale neutron trip was attributed.to a
- failed local power range monitor (LPRM) which was subsequently
l bypassed. The reactor was taken critical on July 7,1986.
4
i At 6:30 p.m. on July 6,1986, an alert was declared when
. tampering was considered a possible cause for the initiation of
- the diesel generator (DG) building fire suppression system. No
l water was actually released since the fusible links remained
intact. Subsequent investigation revealed that an area deter.or
, (heat sensor) malfunction caused the system initiation and that
}' tampering was not the cause. The alert was terminated at 7:30
, p.m. on July 6, 1986,
i
j On July 12, 1986, the licensee inserted a manual scram when
!- both "C" and "D" steam flow transmitters in main steam line "B"
l sensed high steam flow and shut all main steam isolation valves
.
i
(MSIVs). The high steam flow indication was attributed to
transmitter drift. Both "C" and "D" transmitters were replaced
l and the reactor taken critical on July 13, 1986 to continue low
l
power testing.
!
, During the period from July 15 to July 20, 1986, the unit
l experienced four separate automatic initiations of the high
l pressure coolant injection (HPCI) system. During each of the
- events, the HPCI turbine was tripped before any water was
i
[
__ . . _ _ _ . _ _ . _ . _ ~ - . _.__ _ _ _ _ _ _ _ _ __ _ _ _ _ _-___
, e
4
injected into the reactor vessel. A review of plant conditions
prior to, and after the actuations showed that reactor vessel
water level remained within the normal range and that the HPCI
system should not have received an actuation signal. The
licensee's investigation and a subsequent test conducted on July
20, 1986, established the most probable cause for three of these
spurious actuations to be workers in the drywell bumping into
reactor vessel level sensing lines. For the actuation on July
16, 1986, the cause was determined to be an instrument and con-
trols (I&C) technician valving error. In an effort to prevent
further spurious actuations, the licensee placed more stringent
controls on access into the drywell and reinforced the impor-
tance of proper valve operations to I&C technicians.
On July 19, 1986, the reactor scrammed from approximately 0.5%
power due to an operator error in the manipulation of the "B"
and "G" IRM range switches. The reactor was taken critical on
July 19,1986, for continuation of the low power test program.
On July 25, 1986, a reactor scram occurred from 3% power due to
reactor vessel low water level. Surveillance testing was in
progress on the turbine stop and control valves when an operator
erroneously shut the valves to start turbine chest warming.
This resulted in all bypass valves opening and a reactor high
water level due to swell which tripped the two operating feed
pumps. Feedwater was not restored before the reactor scrammed
on low level. All systems responded normally to the scram.
Following a SORC review of the event, the reactor was made
critical at 7:48 a.m. on July 26, 1986.
On July 30, 1986, the reactor scrammed while troubleshooting the
-22 volt DC portion of the electro-hydraulic control (EHC) logic
system. During troubleshooting, the -22 volt DC supply failed
and all bypass valves went full open causing a reactor vessel
high water level which tripped all feed pumps. The feed pumps
could not be restarted prior to receiving a low water level
reactor scram. The licensee commenced a reactor startup at 3:15
a.m. on July 31, 1986, and terminated startup at 4:45 a.m. when
the rod position indication system (RPIS) failed. The reactor
! was maintained sub-critical until RPIS troubleshooting was
l complete and the reactor taken critical later that day.
On August 8, 1986, the licensee declared an unusual event when
it was discovered that the reactor building to torus vacuum
breaker butterfly isolation valves were inoperable and would
have prevented the vacuum breakers from fulfilling their safety
function. The plant was shutdown and separate investigations
by the plant staff and the offsite safety review committee
l commenced. It was determined that the differential pressure
transmitter sensing lines were connected backwards, such that
'
l the isolation valves would close as a vacuum was created in the
i
L
. .
5
torus instead of open as required. The licensee's corrective
actions included a complete walkdown of the vacuum breaker
system and a verification that similar problems did not exist
in other plant systems.
On August 13, 1986, the reactor was placed in operational con-
dition 1 and the main generator was synchronized to the grid.
The shutdown from outside the control room test was conducted
on August 22, 1986.
On August 31, 1986, the reactor scrammed on low vessel level.
Loss of level control occurred during plant startup when a
secondary condensate pump was started. The startup control
system was controlling reactor level since the feedwater con-
trol system does not have single element control capability.
The licensee reviewed the event, took the reactor critical,
and entered operational condition 1 on September 2, 1986.
On September 6, 1986, the unit scrammed from 38% power due to
low water level in the reactor vessel. The low water level
condition occurred during reactor feed pump (RFP) minimum flow
valve (MFV) response tuning. In preparation for tuning the "C"
RFP MFV, the "C" RFP was paralleled with the running RFP. While
paralleling RFPs, the "C" MFV began oscillating which resulted
in reactor level oscillations and a low level reactor scram.
This scram was caused by a combination of operator inexperience
and a lack of feed system tuning. The unit returned to power
operation on September 7, 1986.
On September 11, 1986, a " Loss of Offsite Power" (LOP) test was
commenced as part of the power ascension test program. This
test simulated a total loss of offsite power by simultaneously
opening the appropriate circuit breakers on the 13.2 KV ring bus
and tripping the main turbine. The plant's automatic response
was then evaluated, including the fast transfer of selected
buses to emergency DC power, the starting and loading of all
four emergency diesel generators (EDG), and the automatic
sequencing of loads needed to respond to the resulting scram.
The LOP test was initiated from approximately 20% reactor power.
The reactor plant's response to the resulting transient was
within design limits. However, because cooling water flow to
the drywell coolers was lost, the Senior Nuclear Shift Super-
visor (SNSS) aborted the test and had offsite power restored to
the site distribution system. Cooling water flow was lost due
to the tripping of the reactor auxiliary cooling system (RACS)
pumps. In addition to the loss of RACS, other problems identi-
fied during the test included: the failure of the "C" EDG
output breaker to automatically close and supply power to the
"C" 1E bus, the sustained loss of power to the safety relief
valve acoustic monitor panel, the failure of the "B" safety
auxiliary cooling system (SACS) pump to restart, and the loss
of reactor building ventilation.
. ,
.
6
The unit remained shut down from September 12 to October 9,
1986. During this outage, the licensee conducted.an investi-
gation into the causes for the LOP discrepancies and took
corrective actions. In addition to the LOP related activities,
the service water pipe elbows at the SACS Feat exchangers were
-replaced due to erosion.
On September 24, 1986, a Confirmatory Action Letter (CAL No.
86-12) was issued to the licensee to confirm that an Augmented
Inspection Team (AIT) was being dispatched to the Hope Creek
site to assess the anomalies identified during the LOP tests.
After receiving authorization from the AIT team leader, a
non-critical LOP test was conducted on October 2,1986. This
test was successful in that it satisfied all Level 1 and Level 2
acceptance criteria. In addition to the original LOP test
scope, this test also verified the proper operation of a sample
of Bailey 862 logic module functions not previously tested.
Based upon the satisfactory test results, on October 7, 1986,
the CAL was modified and the NRC authorized a plant startup in
order to conduct an LOP test with the reactor critical.
The loss of offsite power (LOP) test was conducted from
approximately 20% power on October 11. All Level 1 and Level 2
acceptance criteria were met and although a number of
observations were made, the test results were determined to be
acceptable.
After successful completion of the LOP test, NRC Region I
authorized a plant restart for power ascension testing. The
unit was brought critical on October 12.
On October 18, 1986, the reactor scrammed on low reactor vessel
water level after an I&C technician installed a test box on the
"A" reactor feed pump (RFP) flow controller. The test box
caused both the "A" and "B" RFPs to run back to minimum flow
causing a decrease in water level. The licensee's investigation
determined that a wiring error had been made internal to the
test box. The test box wiring configuration was corrected prior
to the continuation of power ascension testing on October 19.
The facility attained 100% power on November 10.
On November 14, 1986, the reactor scrammed from 97% power after
receiving a reactor vessel high pressure signal. The high
pressure condition was caused by power ascension closure testing
of a main turbine control valve.
._ _ - ___ _
, .
7
The licensee entered cold shutdown on November 16, 1986, in
order to conduct various outage repairs. The reactor was taken
critical on November 28, for continuation of the power ascension
program. The unit remained in operational condition 1 through
the end of the assessment period.
C.2 Inspection Activities
Two NRC resident inspectors were assigned to the site throughout
the assessment period, and for a total of four months, there
were three resident inspectors on site. During this thirteen
month assessment period, 9170 hours0.106 days <br />2.547 hours <br />0.0152 weeks <br />0.00349 months <br /> of direct inspection were
performed, which equate to 8460 hours0.0979 days <br />2.35 hours <br />0.014 weeks <br />0.00322 months <br /> on an annual basis.
During the assessment period, five NRC team inspections were
conducted to examine the following areas:
-
As-built inspection in the areas of mechanical, electrical,
instrumentation and control, and structural systems as well
as a review of as-built equipment for selected emergency
procedures and Final Safety Analysis Report (FSAR) accident
analysis assumptions.
-
Technical Specification review to determine whether the
draft Technical Specifications and the Final Safety
Analysis Report were in agreement with the plant's as-built
condition. ,
-
Special inspection of the Hope Creek SAFETEAM program.
-
Post accident sampling and monitoring systems inspection
to verify the implementation of selected NUREG-0737,
Clarification of TMI Action Plan Requirements.
-
Operational As sessment Team Inspection to assess the
facility's operational effectiveness.
An Augmented Inspection Team was dispatched to the Hope Creek
site to review the anomalies that occurred during the power
ascension loss of offsite power tests.
Two special inspections were also conducted, as follows:
-
An investigaticn into the cause for the inoperability of
the reactor building to suppression chamber pressure relief
system.
-
An inspection in support of a licensing action related to
the deletion of the fire protection Technical
Specifications in accordance with Generic Letter 86-10.
8
This assessment report also discusses " Training and Qualifica-
tion Effectiveness" and " Assurance of Quality" as separate
functional areas. Although these topics, in themselves, are
assessed in the otner functional areas, through their use as
evaluation criteria, a synopsis of these two areas is provided.
For example, quality assurance effectiveness has been assessed
on a day-to-day basis by resident inspectors and as an integral
aspect of specialist inspections. Although quality work is the
responsibility of every employee, one of the management tools to
measure this effectiveness is reliance on quality assurance
inspections and audits. Other major factors that influence
quality, such as involvement of first-line supervision, safety
committees, and worker attitudes, are discussed in each area.
Due to limited inspection activities in the fire protection
area, it is not included as a separate functional area in this
report. Inspection activity that was performed in the area of
fire protection is included in the Plant Operations functional
area and related licensing activities are discussed in Section
IV.I.1.
Tabulations of inspection activities and associated enforcement
actions are contained in Tables 3, 4 and 5. The percentage of
total inspection time devoted to a functional area, tabulated
in Table 3, is included at the heading of each area analyzed in
Section 4.
. .
9
II. CRITERIA
Licensee performance was assessed in selected functional areas significant
to nuclear safety at operating facilities.
The following evaluation criteria were used, as appropriate, to assess each
functional area:
1. Management involvement in assuring quality.
2. Approach to resolution of technical issues from a safety standpoint.
3. Responsiveness to NRC initiatives.
4. Enforcement history.
5. Operational and construction events (including response to, analysis
of, and corrective actions for).
6. Staffing (including management).
7. Training effectiveness and qualification.
Based upon the SALP Board assessment, each functional area evaluated is
classified into one of three performance categories. The definitions of
these performance categories are:
Category 1. Reduced NRC attention may be appropriate. Licensee
management attention and involvement are aggressive and oriented toward
, nuclear safety; licensee resources are ample and effectively used so that
a high level of performance with respect to operational safety is being
achieved.
Category 2. NRC attention should be maintained at normal levels.
. Licensee management attention and involvement are evident and are
concerned with nuclear safety; licensee resources are adequate and
reasonably effective so that satisfactory performance with respect to
operational safety is being achieved.
Category 3. Both NRC and licensee attention should be increased.
Licensee management attention or involvement is acceptable and considers
nuclear safety, but weaknesses are evident; licensee resources appear to
be strained or not effectively used so that minimally satisfactory
performance with respect to operational safety is being achieved.
Trend. The SALP Board may determine to include an appraisal of the
performance trend of a functional area. Normally, this performance trend
will only be used when both a definite trend of performance is discernible
to the Board and the Board believes that continuation of the trend will
result in a change of performance level.
.. . ._ _ . - -
. - - . . - - - - - - - . - - -
_ _
,, _. _ .__..m. _ _ . _
.
_ .-
._
,
_.f . . ; _ _
-
.f i.; ,
1
- c
. . .. . : .
,
, .,
_. %. L ,
s , ,
j_ 10 i .s,
4 (
"
Improving: Licensee performance was determined to be improving near
the close of the assessment period.
- '
Declining: Licensee performance was determined to be declining near
-the close of the assessment? period. ,
!
.
, .
e s
%. .
s
r*
5
'
~ p
,
I
i .
- '%
- - . . . s
t'
/
,
,"
,
I 4 . g.
. e.#
1 ,%
}
}
,
,
i
l
d
.
3
- i
.
~e
(
.
4 s
p.
4
=_
%
/
9
.
I
L
,
I
- l
._[
<-
mt
I
L
, . \;
1 ' '
11
'
c
'
,
'I
,
s
').
.t ..
" "IM. SUMMARY OF RESULTS
/ 3.1 Overall Facility Evaluation
'te
'
The licensee completed the transition from a construction facility
u an operating nuclear power plant during this assessment period.
Trh plant progressed from a 90% complete construction status to
L
b beiag only a few weeks away from commercial operation in a thirteen
_.3 en+n period. A very ambitious schedule was established by manage-
ment., and, although not met for most milestones, it did provide good
<
, direction throughout the period. Despite the ambitious schedule, a
!c y good perspective on quality and nuclear safety was maintained.
,
f ; s Plant procedures and administrative programs are generally of high
q{'3 quality, due in part to the operating experience evaluation program.
Soma aspects of the radiation protection program, however, warrant
'
additional management attention. Efforts to improve administrative
activities without sacrificing quality are also needed. The incident
,e report program provides excellent feedback of operating experience to
all departments.
f ,
e'
Control room operations have been cor acted in a consistently pro-
fessional and safety conscious manner. Noise and access control,
6
especially during power ascension testing, have been excellent.
, ,, Except for two operator-error-induced scrams early in the test pro-
grem, the operators have performed well throughout the period. The
shift turnover meetings, work control group, and Technical Specifi-
cation interpretations promote good performance in the operations
area. Areas which warrant attention include: maintenance of control
room logs, reducing the number of alarming annunciators and reducing
the number of unplanned scrams and reportable events.
The organization is generally well staffed with qualified personnel.
The radiological and chemistry department vacancies which have been
'
recently created need to be filled promptly in order to provide the
necessary supervisory oversight. Approximately one-third of all
-reportable events were attributable to personnel error (mostly during
surveillance tests). The major contributor to these events has been
spurious initiation signals of the engineered safety features (ESF).
The occurrence rate of these events has been significantly reduced by
comprehensive corrective action programs.
Overall, a solid foundation has been established for the first cycle
of plant operation. Management support is evident, particularly in
the areas of emergency planning, security, and quality assurance.
The licensee recognizes the need for additional attention to support
programs, in particular, radiological controls.
-
-
_ _ _ _ - - - - - _ - _ _ _ . _
, ..
f-
.
..
-
,
12
,
Y 3.2 Facility Performance
'
U Functional Category > Category Recent
! Area Last Period This Period Trend
(11/1/84-10/31/85) (11/1/85-11/30/86)-
,
'
A. Plant Operations 1* 2
',
'
B. Radiological
Controls 2 2
C. Maintenance 2 1
D. Surveillance Not 2
Evaluated
E. Emergency _
Preparedness 2 1
F. Security and
Safeguards 1 1
G. Outages Not No Rating
Evaluated
H. Preoperational and
Startup Testing 2 2
I. Licensing Activities 2 1
,
J. Training and
Qualification Not 2 Improving
Effectiveness Evaluated
)
K. Assurance of Not 2
Quality Evaluated
This area was titled Operational Readiness in the previous SALP
f
,
,
I
l
t
?
e . _ _ _ - _ _ . _ _ - - _ _ _ _ _ _ - - _ . . _ . . _ . - _ . _ _ _ . - _ . _ _ .
. .
13
IV. PERFORMANCE ANALYSIS
A. Plant Operations (33%, 3030 Hours)
1. Analysis
The functional area of " operational readiness" was evaluated
to be Category 1 during the previous assessment period. Some
weaknesses were identified but the general conclusion was that
the transition from construction to operations was well control-
led, staffing was adequate and experienced, training programs
were effective, and administrative controls under development
appeared generally adequate. The SALP Board recommended that
the applicant provide NRC an operational readiness presentation,
based on a self appraisal, which was completed during April,
1986.
The operations area was under continual review by two resident
inspectors for the entire assessment period and by a third
resident for a total of four months. In addition to the
resident inspectors, this area was reviewed by preoperational
and startup program inspectors, the augmented inspection and
operational assessment teams, and senior NRC management during
numerous site visits. Two sets of initial operator licensing
examinations were given to a total of 25 candidates during
February and July,1986. Training and qualification effec-
tiveness is discussed in Section J of this report.
Plant operations have been conducted in a consistently
conservative and safety conscious manner. The transition of
project completion responsibility from the Vice President -
Engineering to the Vice President - Nuclear on December 2, 1985
(5 months prior to fuel load), helped change the focus from
construction completion to plant operations. Assigning the
assistant general manager for Hope Creek operations to the
position of project completion manager and providing him with
the resources necessary to do his job, significantly contrib-
uted toward establishing a high standard of performance and
emphasis on nuclear safety. A safety conscious attitude
was apparent throughout the entire Hope Creek operations
organization. Senior plant management is intimately involved
with the day-to-day operation of the plant. The station's
general manager and all department managers attend a daily
management meeting to discuss current issues and establish
priorities for future activities. The Vice President - Nuclear
occasionally attended these meetings. All work activities are
scheduled by the planning department based upon the priorities
established by management and input from the work group
superv, sors. This method of planning and scheduling has worked
well throughout the power ascension program and has ensured
that the " big picture" was maintained. The station operations
< - - - - -
,
. . i
14
review committee (SORC) has generally done a thorough job of
overseeing plant operations. The offsite safety review group
performed a number of in-depth reviews including an independent
investigation into the causes for the inoperability of the
reactor building to suppression chamber pressure relief system.
In addition to an accurate assessment, their recommended cor-
rective actions were timely and effective.
The licensee has been responsive to NRC concerns both prior to,
and since, plant licensing. Major NRC team inspections such
as the As-Built, Technical Specification, Augmented Inspection
Team, and the Operational Assessment Team inspections received
timely and effective support during the assessment period.
Prior to plant licensing, all appropriate NRC open items were
resolved. The licensee's commitment tracking system has ensured
prompt resolution of outstanding inspector concerns. Numerous
briefings were conducted for the NRC on spurious engineered
safety feature (ESF) actuations, Bailey 862 solid state logic
modules, the inoperable reactor building to suppression chamber
pressure relief system, and the loss of offsite power tests.
Plant procedures and administrative controls are thorough and
based upon a review of over 3000 documents such as IE bulle-
tins, circulars, information notices, INPO documents and vendor
recommendations. However, in an effort to incorporate these
numerous requirements, recommendations, and good practices, a
large administrative burden has been created for the plant
staff. Occasionally, this burden impacts negatively on the
implementation of the overall program. A review of the
equipment malfunction identification tagging (EMIT) system
identified a large percentage of tags on equipment in the
plant were no longer valid, and system walkdowns by the NRC
have identified a number of discrepancies in the tagging
request inquiry system (TRIS) valve lineups. It appears that
this administrative burden contributed to a month long delay
in determining the inoperability of the reactor building to
suppression chamber pressure relief system.
The Operations Department has a more than ample number of both
licensed and non-licensed operators to meet staffing require-
ments and man a 5 shift rotation with a minimum use of overtime.
The control room is consistently maintained in a professional
manner with very good access and noise control. Noise control
is especially aided by the plant page system design which pre-
vents routine pages from being heard in the control room. The
control room environment is also aided by the use of a work
control group that processes all work orders, surveillance
tests, and blocking permits outside of the control room with
the exception of the senior nuclear shift supervisor's (SNSS)
final approval.
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
.
. *
15
The narrative control room log needs to be improved. On
occasion, the logs have been found to lack detail and are
inconsistently maintained among different shifts. In addition,
there is significant duplication of information between the
SNSS, shift supervisor, and control room operator's logs. Also,
the control room alarm system needs to be improved. The large
number of overhead annunciators that are in alarm at any given
time, interferes with the ability to understand current plant
conditions. During full power operations, over 50 annunciators
in alarm have been observed.
With the exception of two violations identified shortly after -
initial licensing, no further Technical Specification adherence
problems have been identified in the operations area. The
establishment of a formalized TS interpretation log has aided
the operators in establishing a consistent and well thought out
approach to TS compliance.
Shift briefings conducted by the SNSS in the operations support
center are noted as a strength. Pre-shift briefs are conducted
for both the operators and all other support organizations.
Despite the pre-shift briefings, the interface between opera-
tions and chemistry needs improvement. A number of TS action
statement violations involving a failure to take a sample have
occurred, partially as a result of inadequate communications
between departments.
Control room operator errors directly caused, or may have con-
tributed to, 3 of the 14 unplanned scrams during the power
ascension program. (All plant scrams are described in Table 6
of this report.) Although higher than desired, this number
appears consistent with other recently licensed BWRs. There
have been 89 reportable events since low power license issuance
on April 11, 1986. Of these events, 47 can be categorized as
personnel errors and/or new procedure problems. The major
contributors to the reportable events are: 7 loss of coolant
accident (LOCA) signals, 9 engineered safety feature (ESF)
actuations, and 5 high pressure coolant injection (HPCI) system
actuations. The majority of these spurious signals share the
common root causes of valve misoperation, and inadvertent con-
tact with sensing lines during drywell work. The licensee
formed a task force to investigate these events and completed
the following corrective actions: (1) installed quick discon-
nects on instruments, (2) installed identification tags on
sensing lines, (3) installed protective cages around instrument
racks, and (4) blew back instrument lines to remove entrapped
air. Based upon recent performance, these corrective actions
have been effective.
-. .
_ _ _ _ _ _
. .
,
.
16
The licensee has implemented a strong housekeeping program
throughout the plant. An integral part of this program is the
plant management tours made on a routine basis and the follow-up
inspections to verity implementation of corrective action. A
plant painting program is being implemented that should also l
improve the plent's appearance. Considering the status of the l
plant during this assessment period, housekeeping and cleanli-
ness are adequate.
During the assessment period, two inspections were performed
to review the licensee's fire protection program, the system's
installation, and the FSAR and Technical Specifications for
compliance with Generic Letter 86-10. During these inspec-
tions, corporate and site management exhibited thorough and
aggressive involvement with, and control of, fire protection
program activities. It was also evident that priority was
given to problems requiring hardware solutions. The licensee
requested deletion of the fire protection Technical Specifica-
tions as reconnended in Generic Letter 86-10. The NRC deter-
mined that deletion of these Technical Specifications was in
accordance with the guidance provided in the Generic Letter
and that existing fire protection requirements ha u been
incorporated into plant procedures and equivalent adminis-
trative controls exist to control these activities. It was
concluded that adequate controls exist to evaluate fire pro-
tection program changes and ensure the ability to achieve and
maintain safe shutdown in the event of a fire. Staffing for
the fire protection program and training of personnel were
judged to be adequate.
In summary, the proper perspective on safety has been estab-
li2hed throughout the plant staff and station procedures.
Control rocm operator performance during plant transients and
events has been a noteworthy strength. Strong management
attention is evident in the day-to-day operation of the
facility.
2. Conclusion
Rating: Category 2
l Trend: None
3 Board Recommendations
Licensee: Evaluate methods to improve administrative
,
activities consistent with safe plant operations.
! NRC: None
!
l
L
. .
17 i
I
B. Radiological Controls and Chemistry (6%, 592 Hours)
1. Analysis
During the previous assessment period, the licensee's perform-
ance was evaluated as Category 2 in the area of Radiological
Controls and Chemistry. Weaknesses identified during the
period were: a lack of adequate licensee oversight and atten-
tion to detail in the development of the radiation protection
program; a need-to improve coordination and communication
between the operations, radiation protection, and chemistry
groups; and a lack of adequate justification to support
deferral of operability of certain process radiation monitors.
Inspection early in this period found a continuation of the
radiation protection program development problems identified
during the last period. These included inadequacies in the
radiation work permit program, high radiation area access
control program, and airborne radioactivity sampling and anal-
ysis program. A number of technical deficiencies in procedures
were also identified and were attributed to inattention to
detail during procedure reviews by the station and corporate
radiation protection group. These problems were attributed to
the lack of a thorough operational readiness review of the
program by the licensee. Although QA audits of selected ele-
ments of the radiation protection program were performed, they
focused primarily on procedure compliance and not on program
adequacy. While a limited operational readiness assessment of
station radiation protection program adequacy was performed by
the corporate radiation protection group, the assessment find-
ings were not tracked to resolution or verified closed by the
corporate group. The licensee initiated aggressive action to
resolve subsequent NRC findings. The findings were priori-
tized and contractor support was obtained to assist in their
resolution. Despite the number of findings, the licensee was
able to adequately resolve them to the satisfaction of the NRC
4
prior to issuance of the low power license. In order to fur-
ther upgrade the program, the licensee, after issuance of the
l
low power license, initiated a contractor review of the entire
program to identify other weaknesses. The findings are tracked
by computer to resolution and monitored by management. The
effectiveness of this review has yet to be verified.
A contributing factor to the lack of adequate program develop-
ment was a reorganization of the station radiation protection
group which resulted in the loss of some key supervisory
personnel and the lack of a fully staffed corporate radiation
protection group. The losses adversely affected the corporate
group's capability to provide normal program development
support. At the close of the assessment period some posi-
tions remained vacant and administrative procedures had not
l
l
.; ,'
18
'been updated to reflect revised _ reporting chains and personnel-
responsibilities. Experienced contractor personnel were effec-
)
tively used.to augment the organization. ,
Due to the low radiation and radioactive material. source terms,
the radiation protection program was not sufficiently challenged
to allow NRC to fully evaluate oversight and-control of in plant
radiological work activities. However, limited NRC review of
radiation protection technician' performance in'the field, and
review of an unplanned exposure to the hand of a technician
indicated _ weaknesses in the supervisory oversight of initial
program implementation and the training program for some
technicians. Also, the assignment of a junior technician'to
handle radioactive sources was considered inconsistent with the
goal of assuring that personnel are assigned to tasks commen-
surate with their training and experience.
A need to increase supervisory oversight of activities in the
radiation protection area was evidenced by the following: some
technicians using improper meters to perform radiation surveys,
inadequate documentation of radiation surveys, lack of consis-
tent performance of surveys, and use of inadequate radiation
work permits to control work with radioactive sources. The
licensee initiated appropriate action to review and resolve
the deficiencies associated with the identified problems.
Technicians were reinstructed regarding proper meter use and
documentation of surveys, source control was tightened, and
-reviews of program implementation were initiated.. The train-
_
ing program was permanently revised to address the identified
problems. In addition, supervisors were counseled and directed
not to assign individuals, including junior technicians, to
tasks for which they had not been qualified.
The special inspection to review implementation of NUREG-0737
post-accident sampling and analysis recommendations identified
a number of problems requiring licensee attention. Although
appropriate sampling and analysis equipment was installed and
operable, and procedures were in place where needed, NRC review
and observation during walkthroughs identified a lack of ade-
quate field testing of procedures, weaknesses in training and
qualification of personnel, and weak intragroup communications.
The weaknesses identified did not preclude collection of samples
but did delay their collection. The licensee initiated aggres-
sive and timely corrective action to address these NRC identified
problems. Regarding effluent monitoring and control, NRC review
determined that the licensee's recovery from delayed installa-
tion / testing of the process and effluent monitors, resulting
from the vendor going out of business, was well planned and
executed.
n _ _ _ _ _ _ -
- _______ - _______ _ _ __ ______
. -
19
Reviews of the ALARA Program found that a management commitment
to ALARA was evident. In addition, state-of-the-art techniques
are evaluated and adopted as appropriate. Radiation prctection
personnel have been placed in the planning and scheduling group
to provide for effective group interface and understanding of
planned work. Although a basic ALARA Program is in place, pro-
gram elements needing up grade were the ALARA goals program and
on going job reviews. These areas are being reviewed and eval-
ucted by the licensee in response to NRC concerns.
Reviews of radiation protection facilities and equipment found
them to be of acceptable. quality. Radiation protection equip-
ment was considered state of the art with ample supplies
available. The supplies were adequate to support plant opera-
tion, demonstrating adequate management attention to this
important area.
Resolution of effluent sample line loss issues associated with
the north and south plant vent monitors was delayed due, in
part, to the resignation of the Senior Radiation Protection
Supervisor-Radioactive Material Control and the subsequent
elimination of the position. Personnel were unable to locate
contractor line loss test reports and line loss test results
were not reviewed, evaluated and incorporated into plant
effluent surveillance procedures, demonstrating poor control
of records and inadequate evaluation and use of test results.
The water chemistry control program was reviewed and found to
conform to generally-accepted industry standards for controlling
contaminant ingress, activated product transport, and corrosion
of pressure boundary and heat transfer surfaces. Radiological
capability test standard interccmparisons showed all measure-
ments to be in agreement. However, comparisons of chemistry
measurements for metals and boron were in disagreement and
weaknesses in controlling, charting and trending chemical
measurements were noted. Resolution of these technical
issues was delayed, in part, by the resignation of the
Chemistry Engineer.
Reviews of preoperational/startup testing of radwaste systems
and initial implementation of the radwaste management program
indicated that management attention was directed to developing,
implementing, and maintaining a generally effective radwaste
management program. The licensee requested and received
approval for deferral of test completion for the gaseous and
solid radwaste systems into the startup phase. Preoperational
testing of the liquid radwaste system showed that the system
was able to perform its intended function. Tests were completed
in a timely manner and met generally-accepted industry standards
for such tests.
_--_ _ .
, -
20
The development of the packaging and shipping program was
delayed by discussions between the Hope Creek Generating
Station and the Salem Station regirding a unified packaging
and shipping program. No radwaste shipments from Hope Creek
Generating Station were completed during the assessment
period.
In summary, NRC reviews at the beginning of the period identi-
fied numerous programmatic deficiencies, particularly in the
area of radiation protection. These deficiencies were attrib-
uted to lack of a thorough review of program operational
readiness, reorganizations, and some staff vacancies. However,
the licensee was able to prioritize the NRC identified problems
and resolve them in a timely manner. The remaining problems
indicate a need to strengthen the internal audit program, sta-
bilize the organization, fill identified position vacancies and
improve inter- and intra group communication.
2. Conclusion
Rating: Category 2
Trend: None
3. Board Recommendations
Licensee: None
NRC: None
,
l
l
!
!
!
!
-
,
,
[+
l
21
C. Maintenance (5%, 445 Hours)
,
1. Analysis
The previous SALP evaluated the maintenance functional area
as a category 2. Noted strengths included the maintenance
traini_ng program and experienced supervisors and managers.
-
The majority of weaknesses identified were associated with the
transition from construction to operations, and the shift of
equipment responsibility from Bechtel to PSE&G. The SALP Board
recommended that this interface problem be resolved in order to
prevent problems during the operations phase. Early in this
assessment period, the station maintenance _ group assumed full
responsibility for the maintenance of all equipment.
During this assessment period, NRC inspectors conducted admin-
_
istrative program and procedure reviews, and observed a limited
number of corrective and preventive maintenance activities.
The mairitenance department is adequately staffed with experi-
enced personnel although the use of contractors is still
required to complete the required staffing in the instrument
and controls (I&C) area. There are approximately 60 personnel
in the mechanical and electrical-maintenance sections, all of
whom are permanent PSE&G employees. -Approximately one half of
the 80 I&C personnel are contractors. The reliance on con-
tractors is being reduced as new hires complete their required
training. These staffing levels appear to be adequate for the
plant work load since the number of outstanding corrective
maintenance work orders is maintained at approximately 800.
Less than 10% of.the outstanding corrective maintenance work
orders would be categorized as safety-related high priority.
The total outstanding ~ work order count is normally higher than
800 because all preventive maintenance (PM), and surveillance
tests (ST), are also given work order numbers by the inspection
order (IO) program. The 10 system appears to be an effective
management tool for the scheduling and tracking of periodic PM
and ST requirements.
The majority of maintenance department activity has been in.the
areas-of minor valve repair, gasket leaks, early life failure
replacements, preventive maintenance, and surveillance tests.
Surveillance tests are further discussed in Section D. The
major activities observed during this assessment period include
control rod drive (CRD) seal replacement, repair and replace-
-
ment of service water elbows at the safety auxiliary cooling
system (SACS) heat exchangers, and replacement of the B resid-
ual heat removal (RHR) pump. Although these activities were
generally well controlled, some problems were identified. A
lack of procedural adherence and a failure to satisfy the
appropriate prerequisites was observed during the CRD seal
l
.. -
22
replacement. Also, the service water system was declared
operable following reassembly, even though a deficiency
report documenting questionable wall thicknesses had not been
dispositioned. The licensee has taken corrective action for
these problems, however, there has not been sufficient basis
to evaluate their long term effectiven*3s.
The plant management meetings, shift turnover meetings, and use
of a planning department to prioritize and schedule all work
activities has been an effective method of placing management's
plan into action. The maintenance planners are responsible for
developing a complete work package including special instruc-
tions, procedures, tool and parts requirements, and retest
requirements. This significantly reduces the administrative
burden on the worker in the field and ensures a consistency
among work packages.
Based upon a limited review in this area, good practices that
have been noted are a comprehensive preventive maintenance
program, the use of M0 VATS on all safety-related, motor operated
valves, the incorporation of the operational. experience evalu-
ation program findings into procedures, and the development of
a master equipment list.
In summary, based upon a limited amount of review, it appears
that a good foundation of procedures and programs has been
established in the maintenance area. Corrective actions have
been taken for procedure adherence and operability determination
problems which occurred early in the assessment period. There
has been limited activity in program implementation during the
period and the organization has not been fully challenged.
2. Conclusion
Ratino: Category 1
Trend: None
3. Board Recommendations
Licensee: None
NRC: Maintain normal inspection activity.
. -
23
D. Surveillance (9%, 823 Hours)
1. Analysis
The surveillance area was not evaluated during the previous
assessment period. Surveillance tests performed by the
licensee are the responsibility of several departments,
depending on the surveillance. The operations, maintenance,
chemistry, and site protection departments participate in
surveillance testing. This section addresses surveillance
tests performed without reference to the particular department
involved. Surveillance activities were routinely witnessed by
NRC inspectors. Because of problems encountered with the
review of preoperational test packages, an increased emphasis
was placed on the technical adequacy and performance of sur-
veillance tests during this assessment period. The surveillance
program is a well defined, computer based system that utilizes
technically adequate procedures. The use of the computerized
inspection order (IO) system for scheduling all periodic sur-
veillance tests allows for efficient and generally effective
management oversight of the approximately 5000 surveillance
tests performed on an annual basis.
Prior to the initial entry into each reactor operational con-
dition, the completion of mode change required surveillance
tests was frequently the critical path. Test progress normally
lagged the schedule for a number of reasons including:
- Not all surveillance procedures were fully written and
approved before needed.
- Technicians were not familiar with all procedures.
- Time delays for equipment failures were not factored into the
schedule.
Of the 89 reportable events during this assessment period,
26 are associated with the performance of surveillance tests.
,
Deficient surveillance procedures resulted in, or contributed
l to, the July 25 scram on low water level and the November 14
scram on high pressure. Schedule pressure and technician
unfamiliarity with surveillance procedures contributed to many
of the reportable events and to an NRC concern regarding the
use of unauthorized temporary procedure changes which altered
the intent of the procedure but had not been Station Operations
Review Committee (SORC) approved. Based upon recent perform-
ance, these problems have been corrected.
e -
24
Six instances of a failure to perform required surveillance
tests or take the action required by the appropriate technical
specifications were identified. The lack of effective communi-
cation between the operations and chemistry departments has
caused failures to obtain and analyze a number of samples
required by the technical specifications. It is noted that
many of these samples are situational in nature and cannot be
placed into the normal scheduling program. The licensee
recognizes that a problem exists and has taken steps to improve
the situation. There has not been sufficient basis to evaluate
the effectiveness of the corrective actions.
On numerous occasions during this assessment period, a single
channel loss of coolant accident (LOCA) signal was generated
from a not always apparent cause.. It appears likely that some
of the LOCA signals resulted froa valve operations on or around
the reactor pressure and level instrument racks which feed the
reactor protection and emergency core cooling system logic
schemes. However, because the exact cause for all of these
LOCA signals could not always he positively determined, the
licensee formed a task force to identify the root cause of
these LOCA signals. The investigation included a review of
all available data. Although no positive determination could
be made of the cause for the signals, a comprehensive action
plan was carried out. These actions included: blowing back
all instrument lines to remove entrapped air, installing
identification tags on all LOCA/ECCS instruments and sensor
lines, installing quick disconnects on LOCA/ECCS instruments,
technician training, review of all LOCA/ECCS surveillance
procedures, and installing cages around instrument racks.
These actions have apparently been effective since no spuri-
ous LOCA signals have been generated during the last four
months of the appraisal period.
Regarding effluent monitoring and control, our review found that
the licensee's recovery from delayed installation / testing of the
process and effluent monitors, resulting from the vendor going
out of business, was well planned and executed. However, the
simultaneous need for preoperational testing of the monitors and
continuous surveillance of those monitors to support early
operation led to occasional lapses in Technical Specification
surveillance tests. On two occasions, effluent monitors were
removed from service and necessary grab samples were not taken
resulting in self-identified failures to meet Technical Spec.1-
fication surveillance requirements. The failure to ensure
adequate communications among the various testing, operations
and technical support groups, and to clearly assign responsibil-
ity for declarations of operability /inoperability, contributed
to the problems noted.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
-
. .
25
The licensee implemented adequate local leak rate test (LLRT)
and containment integrated leak rate test (CILRT) programs.
The tests were conducted using acceptable procedures and
equipment, and the test personnel were knowledgeable and well
qualified. ,
In summary, the majority of difficulties experienced in the sur-
veillance area can be attributed to the self-imposed schedule
pressures associated with the plant entering the startup phase of
testing. The procedures and administrative controls in place
are adequate to implement an effective surveillance program.
Increased attention is needed to improve communications between
departments in order to reduce the number of missed nonroutine
surveillance tests. ,
1
2. Conclusion
Rating: Category 2
Trend: None
3. Board Recommendations
Licensee: None
NRC: None
!
!
-
I I B .
_m_________ _______ __________ m._ ____
_ _ _ _ _ _ - _ _ . _ _ _ _ - _ _ _ _ .
. .
26
,
E. Emergency Preparedness (5%, 454 Hours)
1. Analysis
During the previous assessment period, the licensee was eval-
uated as Category 2 in the area of Emergency Preparedness.
That assessment was based on the results of an Emergency
Preparedness Implementation Appraisal (EPIA) conducted on
r August 12-16, 1985, observation of the annual exercise held
on October 29, 1985, and two routine inspections. Several
critical emergency planning (EP) program areas were determined
to be incomplete and indications were that management attention
had been diverted from Hope Creek EP capabilities development
to (i) upgrading the Salem EP program and (ii) corporate
reorganization. The licensee's performance during the October
29, 1985 exercise was good with only a few weaknesses noted.
During this assessment period, there were two inspections.
One inspection was a follow-up emergency preparedness inspection
conducted February 3-6, 1986, of concerns identified during the
August 1985 EPIA. All but two of the concerns identified during
the EPIA had been resolved. One unresolved item related to
incomplete emergency preparedness training since sufficient
numbers of personnel had not been qualified to provide a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
emergency staffing capability. The licensee committed to com-
plete key personnel training and provide qualified staff prior
to exceeding 5% power. The licensee affirmed, in writing, on
April 8, 1986, that training had been completed and would be
maintained. Full staffing capability was satisfactorily
demonstrated during the November 12, 1986 full participation
exercise. A second unresolved item involved the radiation
monitoring system (RMS). The installation, calibration, func-
tional testing and operability of process and effluent monitors
has now been confirmed by reactor health physics inspections.
The RMS computer links were completed and computer capability
demonstrated. Functionality of the RMS during simulated emer-
gency conditions was confirmed during the November 1986
exercise.
The second inspection included observation of the November
exercise. The licensee satisfactorily demonstrated the ability,
within scenario limitations, to: identify accident conditions;
declare the correct emergency action level; notify governmental
authorities; activate and staff emergency response facilities;
take proper corrective actions; develop protective action recom-
.wndations; interface with governmental authorities including
the NRC Director of Site Operations; effectively plan recovery
operations; and adequately provide measures to protect public
health and safety. In addition, strong performance was noted
in the areas of personnel exposure control and radiation
sJrveys. No significant deficiencies were identified; and,
overall licensee performance during the exercise was adequate.
O. *
27
The En;ergency Preparedness Manager resigned and has been
replaced by a staff senior emergency planner promoted to fill
the vacancy. The Artificial Island emergency preparedness staff
which supports Hope Creek consists of twelve professionals.
Management has provided appropriate support of EP.
An Alert was declared on May 2 when offsite power was lost to
the vital buses and again on July 6 when tampering was con-
sidered a possible cause for a plant fire suppression system
actuation. In both cases, notifications were made promptly
and the emergency plan effectively implemented.
The licensee has installed a state-of-the-art siren system to
meet the requirement for an alert and notification system.
This system provides hard copy diagnostics of performance for
any one or all sirens. Additionally, an advanced surface water
clearing plan for Delaware River surface waters has been
developed and was satisfactorily tested during the November 1986
exercise.
The licensee and the State of New Jersey have negotiated an
agreement whereby the State receives 10 CFR 50.72 notifications
in the same time frame as the NRC as well as the follow-up 10
CFR 50.73 Licensee Event Reports.
FEMA will complete its review of the New Jersey State
Radiological Emergency Response Plan for Artificial Island
during 1987 to determine if approval is warranted per 44 CFR
350.12. "350" approval has been given to the Delaware Plan,
contingent upon a successful siren test.
In summary, the licensee has dedicated sufficient corporate
management attention and resources to establish an effective
, emergency preparedness program. Strong performance has been
noted during events and drills.
2. Conclusion
Rating: Category 1
Trend: None
3. Board Recommendations
Licensee: None
NRC: None
.
. .
28
F. Security and Safeguards (4%, 348 Hours)
1. Analysis
During the previous assessment period, the licensee was evalu-
ated as Category 1 in the area of Security and Safeguards. The
previous SALP assessment of this area was based on reviews of
pre-operational activity in the development of a site security
program. The licensee was effective in: integration with the
Salem security program, resolution of outstanding issues, and
training security personnel.
During this assessment period, the licensee completed both the
integration of the Hope Creek facility security program with
the Salem program and a major upgrade to the security program
that began several years ago. That upgrade included a combined
access control facility, installation of an integrated security
computer system and associated hardware, computerized access
control devices, state-of-the-art assessment aids and new
personnel search equipment. Those extensive activities were
completed by developing and implementing plans in a comprehen-
sive, well thought out and organized manner. Management
attention and oversight of the program was evident throughout
the period from the smooth transition and relatively trouble-
free implementation of program changes. The licensee provided
NRC with thorough and clear progress reports and prompt noti-
fications whenever changes to the plans were necessary.
'
The licensee aggressively addressed previously issued NRC
security related guidance during the development of the Hope
Creek program. The licensee demonstrated a clear understanding
of the safety and safeguards issues and effectively applied
l Salem program experiences to the Hope Creek program. Solutions
,
!
to technical safeguards problems were sound, timely and
conservative. Concerns identified by NRC were promptly and
effectively resolved by the licensee in a competent manner.
The NRC Site Evaluation Team was able to review and certify
the Hope Creek security program for implementation with minimal
difficulty and delay due to adequate records and preparation.
l
'
Aggressive corporate management attention to the development
and implementation of the security program aided in NRC
certification. The licensee has been effective in fostering a
highly professional attitude towards maintaining performance
! objectives of the NRC approved security plans by continued and
effective management. The performance of the security systems
and equipment has been sound and relatively trouble free since
the initial startup period. This performance results from the
extensive design, procurement and engineering effort expended
on program development. To date, the impact of integrating Hope
Creek into the Salem security program has been essentially
unnoticeable when viewed from an NRC regulatory perspective.
f
c
!
.- _
.- - . -.
. . . -- . . - - - . .- . . . - - . - - . -. .
.
w . . 1
29-
Corporate management's interest in establishing and maintaining
a strong security program was further demonstrated by the high
, quality of security force performance indicated during a
special NRC inspection of the security force training and quali-
'
fication program. That inspection was conducted to determine-
'
the quality'and effectiveness of the training program and to
measure the ability of security personnel to carry out their.
- assigned duties. The training is conducted by individuals who
are experienced and competent in their field and who are
~
assigned to security training only. . Training facilities have
.
adequate' classroom space and good training aids. Lesson plans-
4
are.well developed, thorough, and kept current through'. feedback
,
'
from supervisory personnel who perform on-the-job surveillance
of security personnel performance. The results of the special
inspection indicated that the security training program is broad
in scope, of high quality, and administered in a highly profes-
i . sional manner. The results indicated extensive corporate and
onsite licensee management involvement -in the training program
as well as a strong positive influence on'the part of the con- ;
- - tractor's site management and supervisory personnel.
.
'
The licensee's-security plans, procedures, and instructions are
clear,-concise and thorough. Letters and reports submitted to ,
NRC are also clear, promptly submitted, technically -accurate,
and seldom generate questions from the NRC.
The licensee's security management and contract security force
supervisors display a very positive and conservative attitude
i
towards plant security issues and compliance with regulatory
l requirements. These individuals are quick to understand issues .
>
that arise during simulated and actual security events and how '
those. issues can impact on plant security.
i
i The security program is strongly supported by the other plant '
i operating divisions on site and frequent interface is evident. ;
The maintenance staff detects unacceptable conditions with
! security equipment, and then aggressively pursues-corrective
[ action before they develop into major problems. When minor
,
problems were found during NRC inspections, security managers
i
'
were most often already aware of them and were in the process
of establishing corrective actions. This degree of cognizance
is creditable to a strong internal audit and surveillance
program and is further evidence of the licensee's desire to-
l- implement a high quality security program.
3
'
Security force personnel exhibit excellent morale because of
their recognized and respected role onsite, the excellent
support they are afforded by the management of all divisions
and the quality of the equipment they have been provided. As
- a result, they carry out their assigned duties and responsi-
4 bilities in a professional and dedicated manner,
i
!
!
!
,
m
,-,. , ,.. .,,.. ,,,-y-%-,m, v.,t,,- ,,m,m,,.,,,.,y . - - _ ,y.--,,,.,wm, ,. - ,,,-.,,,,_,.,,,,,,.,,,_ - --
_m-._ -. _. ,- m r -
.-,.,,-.,-,,,-em --,
, ..
30
Corporate security management is actively involved.in the
Region I Nuclear Security. Organization and other nuclear
industry groups engaged in security innovations and the
development of security program standards. This is evidence
of management support of the security program at a high level
in the licensee's organization.
To ensure continued effectiveness of the security program,
the licensee conducts in-house surveillances to monitor the
performance of the security organization. Experienced and
knowledgeable personnel perform these surveillances and the
findings are aggressively pursued to en:ure prompt and effec-
tive corrective action and feedback to the training program.
These surveillances are conducted in addition to the annual
security program audit required by the NRC.
Housekeeping of the access control facility and other security
areas is noteworthy. The general state of cleanliness demon-
strates a high degree of pride and morale on the part of the
security force.
The licensee submitted two security event reports pursuant to
10 CFR 73.71(c) during the assessment period. Both events were
bomb threats that were adequately responded to by the licensee
and were subsequently determined to be hoaxes.
During the assessment period, the licensee submitted a
temporary change (TC) applicable to both the Salem and Hope
Creek security plans. This TC identified compensatory measures
that would be implemented during modifications necessary to
consolidate the Salem and Hope Creek protected area. Prior to
the submittal of this change, the licensee contacted Region I
Safeguards personnel and requested a meetir.g onsite to review
and discuss the modification plans. The resulting TC fully
described the issues. The approach to and planning for this
modification is another indicator of the licensee's commitment
to maintain an effective and high quality security program.
In summary, close licensee management attention to this area
has resulted in an effective security program following a
smooth transition period during which the Salem features were
expanded to encompass the Hope Creek site.
2. Conclusion
Rating: Category 1
Trend: None
._. -_- _._ _ _ _ _.. .... _= .. . _ _ ..... _ - .. .. __ _ .._ -. _ . _ __
, . ..
31
1-
,
,
'
3. Board Recommendations
Licensee: None
NRC: Due-to the hiring of a new security force contractor,-
r
4
- maintain normal levels of inspection.
-
4
'
1
f' -f
{-
!
J
A
?
1
a
4
}-
.
!
h
i
i
i
t
!
ii. -
?
t
!
I
5
!
!
,
.
s w-e- ~
q w ~np --
-sw' e w g m , n-y, w, g,v_-,-yy,e g ,,w-e,, , ,re-m,,,pm, _pm.-- - n,,,. , - w ww w e w,,w.----nvw-s,v :. , -.r., wwe< staff questions and exhibited evidence of prior planning in
producing responses to NRC conderns. In most cases, responses
are sufficiently complete and timely. 'Decistor, making is done
at a level which assures adequate canegement review. Management
involvement is evident in P5E&G's responses to staff concerns as
most responses indicate awareness of pelicy, design and opera-
tional considerations. ~
During the current rating period all outstanding SER issues
were resolved, a number of exemptions to the regulations were ,
-
processed add granted, a compresstd power ascension test pro- /
gram was prpposed and submitted to the NRC, anj the low and full
power operating licenses were issued. In addition, following ,
!
licensing, a number of Technical Specification amendment >
,
'
requests have been submitted. In all cases, the licensee has
exhibited a clear understanding of the issues involved as ,
exemplified by the licensee's effort to " compress" the power .
ascension test program. For each test, the licensee' identified'
th9 purpose of the test, the proposed modifications, and pro - ,
, vided safety evaluations supporting the requested modifications.
During various conversations with the licensee regarding the
proposed modifications, the licensee exhibited a very clear
understanding of the issues involved. Similaris, the licensee
exhibited clear understanding of the issue: involved when it
i -
!
~ -- . . _ . . , , _ _
I
g lO
39
.
., submitted vari 6us exemption requests. Each request was accom-
panied by a detailed safety evaluation in support of_the
request, and the necessary findings under 50.12a. In each
y
'
case, it was the licensee's-responsibility to demonstrate to
,, the' staff's satisfaction-the acceptability'of the proposed
'A action' The licensee did so with clear knowledge and full
iS
"' understanding of the issues at hand and their implications on
planti operations.
,
-
Conservatism is routinely exhibited by the licensee when the
l' issue involves safety significance. Most of the licensee's
submittals have exhibited careful forethought, consideration
- of the_ proposed action, and technically sound responses.
J. In most cases, technically sound resolutions are proposed
initially; however, during this rating period,.one example ..
,
'
exists where this was not the case. In this instance, involving
'
.the _ testing of Bailey 862 solid state logic modules (SSLMs), the -
licensee proposed removing a fixed number of module sample popu-
- ;.lation on a regular basis for testing during power operations.
p Fe1. lowing discussions with the staff, the staff and licensee
t.
bota agreed that this was not an acceptable test method, and the
j proposal was superseded. -In this instance, the. licensee appeared
"
P
overeager to resolve the NRC concern without assuring itself
that a safety concern did not exist. Overall, however, sound
( resolutions are initially proposed.
V
'" (
In most cases, PSE&G was responsive to staff initiatives. With
3
5
the exception of not submitting the detailed control room design
review Summary Report II on the. schedule required by a license
-condition, most submittals met the deadlines. The licensee has-
'
provided timely responses to a number of Generic Letters during
j,~
this rating period. _PSE&G appears to make special' efforts in
' resolving issues in a timely fashion, and with full knowledge of
F the-issues at hand. The licensee's responses are usually tech-
I inically sound and thoroughly presented and supported. In the j
few cases where.the licensee has not provided sufficiently
detailed responses, upon notification of this, the licensee
has been very responsive in supplying the needed additional
i f information. Usually this evaluation is provided within twenty-
four hours. As noted earlier, acceptable resolutions to issues
~ '
are initially proposed in most cases.
..
Positions in the Hope Creek organization, including senior.-level
- management, are well defined. Positions and their associated
1
'
responsibilities are accurately described in the FSAR and Emer-
j' gency Plan and appear to be consistent with actual practice.
.
Since the last SALP cycle, PSE&G has filled the vacancies that
7
y existed in the organization. The staff has reviewed the quali-
fications of the individuals filling the previously vacant
~
-
.
'
ny 1
1
-
' '
-
,
h, .'~
. .
40
I '
positions and.found them acceptably qualified. The licensee
has maintained'a substantial and knowledgeable licensing staff
to assure timely and quality responses to NRC concerns.
In conclusion, corporate management is taking a very active role
in licensing matters and responses to NRC initiatives continue
to be timely, thorough, complete and conscious of safety
impacts.
2. Conclusion
Rating: Category 1 {
>
Trend: None .
3. Board Recommendations
'
Licensee.' None . .
' '
\~ > ,
i
'
NRC: None M '
T
.
- I
lN
..
V'
e
l'
I
e
t
'
fD
,e 'l-
.,
4 <
.I >
&,
,
-
j
_ . - - . - _ _ _ _ - _ - - _ _ - _ _ _ . _ _ - . _ _ _
. --
. .
_
41
J. Training ar.d Qualification Effectiveness (NA)
1. Analysis
During this assessment period, Training and Qualification
Effectiveness is being considered as a separate functional area
for the first time. Training and qualification effectiveness
continues to be an evaluation criterion for each functional
area.
The various aspects of this functional area have been considered
and discussed as an integral part of other functional areas and
the respective inspection hours have been included in each one.
Consequently, this discussion is a synopsis of the assessments
related to training conducted in other areas. Training effec-
tiveness has been measured primarily by the observed performance
of licensee personnel and, to a lesser degree, as a review of
program adequacy.
The licensee operates anc maintains well equipped training
facilities which provide training for all of the nuclear
departments including operations, I&C technicians, electricians,
mechanics, chemists, health physics technicians, machinists, and
welders. The Hope Creek training program is modeled after the
Salem program which has been INPG accredited in all ten training
areas.
The NRC administered two sets of initial operator licensing
examinations at Hope Creek during this SALP assessment period
(February 1986, and July 1986). A total of 25 Senior Reactor
Operator candidates were examined with 22 passing.
,
Weaknesses identified during the oral examinations included an
unfamiliarity with the flow signals to the APRM/RBM systems, ADS
logic, and fire protection equipment. It was also noted that
several candidates had a fundamental misconception about the
operation of the feed water control system (FWCS).
Two unplanned reactor scrams, early in the power ascension pro-
gram, were a result of control room operator errors. In both
cases, an IRM range switch was incorrectly downranged resulting
, in an IRM-high trip. A difference between the simulator and the
as-built feedwater system may have contributed to two other
scrams. The simulator does not accurately reflect the as-built
condition of the feedwater turbine reset logic and the actions
required to reset the turbine from the control room. Because of
these differences, the operators were slow to recover a tripped
feed pump and the reactor scrammed on low level. Prompt cor-
rective action in the form of shift briefings was taken and
simulator upgrades are planned.
- . ,. --- - , _ .
. *
42
Strengths observed during the oral examinations included the
candidates' familiarity with safety and major systems (with the
exception of the FWCS). Also, most candidates displayed a
responsible attitude toward their duties as licensed operators.
A weakness in the ability to interpret and apply the Technical
Specifications was noted during the grading of many of the SRO
written examinations. Also identified were weaknesses involving
the response of the FWCS (as mentioned above) and fire brigade
manning exemptions.
The Hope Creek full scope simulator is performing well and is
providing a . valuable tool for licensed operator training. The
simulator was also used to perform validations of all major
power ascension tests prior to actual in plant performance.
This significantly improved the quality of power ascension test
procedures and provided valuable training to both operators and
test engineers.
The plant operators, in general, have positive attitudes
towards the training program. They felt they have been ade-
quately trained on plant systems and system operations. They
also feel the lecture and simulator programs are excellent.
Although varying opinions were observed as to the technical
adequacy of the written training material, it was agreed that
the readability of these materials could be greatly improved.
Based upon discussions and direct observation, the performance
of licensed operators in the control room has been observed by
the NRC to be excellent. The operators are proficient in
recovering from plant transients and equipment malfunctions in
a competent and professional manner and have demonstrated a
consistently improving knowledge of Technical Specifications
as evidenced by daily discussions with NRC inspectors. Knowl-
edge of system operational characteristics, familiarity with
procedures, and actions on transient response were noted, and
are indicative of effective and valid training for licensed
operators.
The licensee's corporate and station management involvement in
training is good. Training review groups evaluate training on a
regular basis and provide feedback to the training program. The
training department is well staffed with experienced personnel.
Laboratory facilities are excellent and provide hands on train-
ing on such things as rebuilding circuit breakers, Limitorque
valve operators and motors. The 2 year assignment of licensed
operators to the training department is also a positive feedback
mechanism.
-__ - _ __-_-___-___ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ - _ - -
,. *
-
43
The training program has been responsive to the requests of
various departments on a timely basis. When it became apparent
that valving errors by I&C technicians were causing spurious
LOCA signals, the training department set up a training instru-
ment' rack on site and provided training to all technicians.
This training directly contributed to a reduction in spurious
signals. In addition, a modified SR0 training program is
planned for personnel designated as system engineers. It
appears that these changes will have a positive impact on the
performance of the engineering support groups.
Regarding training and qualification of radiation protection
personnel, a documented training and qualification program
for radiation protection personnel has been established and
implemented. The program consists of formal classroom and
on-the-job training. The program is not yet INPO accredited
and is. based on a job-task-analysis for the Salem Station.
Some findings this period (e.g., lack of adequate training for
individuals handling sources and improper use of radiation
survey instruments) suggest a need to perform a specific job
task analysis for radiation protection personnel at Hope Creek
and an upgrade of the program as appropriate. The licensee is
planning to do this as part of efforts to become INPO accredited
in this area.
Management's interest in establishing and maintaining a quality
security program was demonstrated by the high quality of secu-
rity force performance indicated during a special NRC inspection
of the security force training and qualification program. That
inspection was conducted to determine the quality and effective-
ness of the training program and to measure the ability of
security personnel to carry out their asugned duties. The
training is conducted by individuals who are experienced and
competent in their field and who are assigned to security trata-
ing only. Training facilities have adequate classroom space and
good training aids. Lesson plans are well developed, thorough,
and kept current through feedback from supervisory personnel
who perform on-the-job surveillance of security personnel
performance. The results of the special inspection indicated
that the security training program is broad in scope, of high
quality, and administered in a highly professional manner.
Also, the results indicated extensive corporate and onsite
licensee management involvement in the training program as well
as a strong positive influence on the part of the contractor's
site management and supervisory personnel.
In summary, based upon the high examination pass rate and
operators performance in the control roon, the licensed
operator training program is effective. Problems encountered
during plant operations were due to inexperienced personnel
more than training inadequacies.
__ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ -
. *
44
2. Conclusion
Rating: Category 2
Trend: Improving
3. Board Recommendations
Licensee: None
NRC: None
. *
45
K. Assurance of Quality (NA)
1. Analysis
Assurance of Quality is a new separate functional area for this
SALP period and is a summary assessment of management oversight
and effectiveness in implementation of the quality assurance
program and administrative controls affecting quality.
Activities affecting the assurance of quality as they apply
specifically to a functional area are addressed under each of
the separate functional areas. Further, this functional area
is not an assessment of the quality assurance department alone,
but is an overall evaluation of management's initiatives,
programs, and policies which affect or asrure quality.
During the assessment period, four inspections were performed
in the area of quality programs and administrative controls
affecting quality. These inspections covered the following
areas:
- Administrative procedures, records, design control and
modification, review committees and staffing and
nonlicensing training for operations;
- Bulletins and Construction Deficiency Reports (CDRs); and
- Licensee actions concerning the Salem ATWS event.
In addition, the implementation of the Quality Assurance
(QA) program was reviewed by the resident and region based
inspectors in conjunction with other functional areas.
Overall, the licensee appears to have developed a strong
program for assuring quality during operations. The licensee
established a generally effective program for ensuring the
timely issuance of the plant administrative procedures.
These procedures are well written, complete and meet the
FSAR commitments.
The operating experience evaluation program's review of over
3000 industry documents from the NRC, INPO, and vendors has
had a positive impact on the quality of plant procedures.
In addition, the incident report program provides a rigorous
mechanism to ensure that Hope Creek's own operational experi-
ence is evaluated and changes made to procedures when required.
All occurrences meeting certain criteria, whether reportable to
the NRC or not, are documented and investigated. Each dispo-
sition is performed by the appropriate work group and includes
the correcth e action taken or planned. Station management is
required to review and approve the disposition of all incident
reports.
m
, -
46
In the design change and modification area, the licensee has
made major organizational changes with respect to engineering
support for plant operations. A new engineering manual has
been developed that is a distinct improvement on previous
procedures.
In the area of review committees, careful forethought and
planning by management in the establishment of the various
committees is evident. The Station Operations Review Committee
(SORC) has been extensively involved with the preparations for
operations since it became functional in July 1984. Since
then, the licensee has made significant changes in the SORC
review process to enhance the quality and timeliness of com-
mittee reviews. Other strengths include the Offsite Safety
Review Group initiative to be in the online review of pro-
posed design changes / modifications which exceeds 10 CFR 50.59
requirements. The Offsite Safety Review Group performed a
timely and in-depth review of the reactor building to
suppression chamber pressure relief system inoperability.
The licensee has implemented an effective program, with
adequate staffing to follow-up NRC bulletins, circulars,
information notices and CDRs. The evaluation, analysis and
resolution of problems and NRC initiatives have been effective
and timely.
In the area of licensee actions concerning the Salem ATWS
event, licensee management has been aggressive in taking an
active part to assura that the ATWS issue receives proper
emphasis. This aggressive approach is indicated by the Vice
President - Nuclear's letter to station personnel regarding
" Commitment Management," and by licensee procedures wnich have
been implemented including the following: Reliability and
Assessment Management, Response Coordination, Vendor Interface
and Reliability Monitoring. The licensee has established a
Response Coordination Team which is responsible for review,
approval and implementation of all vendor supplied information,
regulatory bulletins, industry standards, engineering recommen-
dations, and operational experiences as applicable.
During this assessment period, the implementation of the QA
program was judged as generally very good. Strong points
observed during review of other functional areas included
extensive QA review of preoperational test results and excel-
lent surveillance coverage of the containment integrated leak
rate test (CILRT). One weakness concerns timeliness of
addressing quality concerns. QA had previously identified a
deficiency involving the use of unapproved temporary proce-
dures for performance of surveillance tests, however, the
practice was not corrected until an NRC inspector identified
the same concern. Upon subsequent review, the NRC found that
_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. .
47
a large number of QA identified concerns were not responded to
in a timely manner by various departments. The licensee has
since increased the visibility of QA concerns and improved the
timeliness of corrective actions.
The licensee's philosophy on assuring quality at Hope Creek
keys on individual achievement of a high level of performance,
emphasizing personnel responsibility, accountability, and pride ;
of ownership. In keeping with this philosophy, programs to I
promote quality awareness and employee involvement have been
instituted during this SALP period and appear to be well
received by station personnel. Examples of these programs are:
- Plant Material Improvement Programs which include cleanup,
painting, and labeling activities in the plant.
- Employee Involvement Program facilitates management / worker
interfaces and awards for good performance.
- Quality Awareness Committee comprised of nuclear department
volunteers who periodically issue a " Quality Gram" promoting
improvements in quality performance.
- Quality Awareness Days are sponsored by individual depart-
ments and inform other departments of quality-improvement
activities in progress within the sponsor department.
- Quality Concerns Reporting Program enables plant personnel to
confidentially express quality concerns to be investigated by
licensee QA personnel.
l Due to the low radiation and radioactive material source term,
!
the radiation protection program was not sufficiently challenged
to allow NRC to fully evaluate oversight and control of in plant
i activities. However, a need to increase supervisory oversight
of activities in this area was evidenced by: technicians
'
repeatedly using improper meters to perform radiation surveys,
less than adequate documentation of radiation surveys, lack of
consistent performance of surveys, and use of inadequate radi-
ation work permits to control work with radioactive sources.
Although corrected by the licensee when brought to his atten-
tion, these example, demonstrate a lack of aggressive oversight
of in plant activities during initial program implementation.
A combination of these weaknesses resulted in a technician
receiving an unplanned exposure of 1.4 rads to his hands.
Reviews of the external and internal exposure controls program
prior to plant licensing found examples of deficient procedures
being established and implemented. Examples include a less
than adequate: radiation work permit (RWP) program, high radi-
ation area access control program and airborne radioactivity
__ .. . _ _ _ _ _ _ _ _ _ _ _ _ _
e .
48
sampling and analysis program. Although corrected in a timely
manner, these examples are indicative of lack of adequate
attention to detail during program development and a lack of
acceptable reviews.
Quality Assurance review of the technical program development
and implementation of the radiation protection program at Hope
Creek was limited. Technical evaluation of program procedures
was conducted solely by the Station Operations Review Committee.
Less than adequate procedures were generated due to insufficient
technical review.
Reviews of preoperational/startup testing of radwaste systems
and initial implementation of the radwaste management program
indicated that management attention was directed to developing,
implementing, and maintaining a generally effective radwaste
management program. Application of the Quality Assurance
program to preoperational tests of the radwaste systems was
thorough and demonstrated an effective identification, track-
ing and closure of test discrepancies. A contingency plan for
processing solid radwaste was developed using vendor-supplied
solidification equipment temporarily attached to the solid
radwaste system. Vendor procedures were reviewed and incor-
porated as controlled plant procedures and included inspection
hold points and other controls governing the vendor's process
control programs.
In summary, the licensee has established a generally effective
program for ensuring quality. The operating experience evalu-
ation program has had a positive impact on the quality of plant
procedures and management has frequently reinforced the role of
the individual in assuring quality. However, increased station
and corporate management attention is warranted in the radio-
logical controls area.
2. Conclusion
Rating: Category 2
Trend: None
3. Board Recommendations
Licensee: None
NRC: None
.
_ ___ _ _______ _ __ - -____ __- ___- __ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
. -
49
V. Supporting Data and Summaries
A. Investigations and Allegation Review
No investigations were conducted during the assassment period.
Five allegations were received during the assessment pariod.
-
Hiring impropriety
-
Crack or scratch in a main steam isolation valve (MSIV) poppet
assembly.
-
Member of Safety Analysis Group does not have a degree.
-
Improper drawing control, retests after maintenance,
performance of preoperational tests, setup and calibration of
radiation monitors.
-
Inadequate training in Chemistry Department, unqualified
supervisors.
All of the allegations were investigated and no significant safety
issues were identified.
B. Escalated Enforcement Actions
On September 24, 1986, a Confirmatory Action Letter (CAL No. 86-12)
was issued to the licensee to inform them.that an Augmented Inspec-
tion Team (AIT) was being dispatched to the Hope Creek site to
assess the anomalies related to the Loss of Offsite Power (LOP)
tests. The CAL also confirmed that the licensee would take the
following actions:
-
Defer any additional LOP integrated testing until the NRC AIT
team leader determines that such testing can continue.
-
Provide any LOP test procedures to the NRC AIT for their review
prior to implementation.
-
Make available to the NRC AIT relevant written material related
to deficiencies identified during the LOP tests conducted on
September 11 and 19, 1986, including:
- preoperational test results
surveillance test results
component installation and function test records
- _ _ _ _ _ _ . _________________ _ _______ _ _ - _ _ _ _ .
-, .
50
-
Provide a written report to the Regional Administrator prior to
restart that includes an analysis of the LOP testing conducted
on September 11 and 19, 1986.
-
Receive Regional Administrator authorization for unit startup.
On October 7,1986, the CAL was modified to allow a plant startup
in order to conduct a reactor critical LOP. The CAL was further
modified on October 16 to allow limited continuation of the power
ascension test program. Based upon the AIT findings, licensee
commitments made in an October 15, 1986 meeting, relating to
Bailey 862 modules, and discussions between NRC Region I and PSE&G
on October 17, 1986, a letter terminating the CAL was issued on
October 21.
On November 17, 1986, an enforcement conference was held to discuss
design deficiencies identified during the LOP test, Regulatory Guide
1.97 instrumentation, and the inoperability of the Reactor Building
to Suppression Chamber Pressure Relief System. Enforcement action
was under review at the conclusion of the assessment period.
C. Management Conferences
-
February 27, 1986: SALP management meeting
-
March 10, 1986: NRC/ Region I - PSE&G readiness for fuel
load
-
March 11, 1986: NRC/NRR - PSE&G readiness for fuel load
-
June 5, 1986: Spurious ECCS actuations, management
changes, lessons learned at similar plants
during startup, control of work practices
(
-
July 21, 1986: Commission meeting for Hope Creek full
! power license
-
July 24, 1986: Corrective action program to prevent
spurious ESF actuations
i
! -
September 19, 1986: LOP Test results
-
October 15, 1986: LOP Test results and Bailey 862 modules
-
November 17, 1986: Enforcement Conference, Design deficiencies,
LOP, Vacuum breaker operability, RG 1.97
instrumentation
,
f
O O
51
D. Licensee Event Reports (LERs)
1. Causal Analysis
Eighty-nine LERs, numbered 86-01 through 86-89, were reported
during this assessment period. These LERs are characterized in
Table 1 by cause for each functional area. Three common causal
chains were identified.
a. Emergency Core Cooling System (ECCS) Actuations
Nineteen LERs (354/86-2, 86-7, 86-10, 86-14, 86-19, 86-20,
86-21, 86-23, 86-24, 86-33, 86-39, 86-41, 86-42, 86-43,
86-46,86-53,86-54,86-59,86-61) describe actuations of
the ECCS due to low reactor vessel water level signals.
Seven ECCS actuations occurred as a result of personnel
error while conducting surveillance tests and nine have
unexplained causes. Investigations eventually discovered
that ECCS initiations could result when personnel in the
drywell stepped on or bumped reactor vessel level instru-
ment piping. While it could not be positively determined
that this explanation applied to all unexplained ECCS
actuations, the licensee has concluded that it is the most
probable cause.
b. Surveillance Testing
Fourteen LERs (354/86-8, 86-9, 86-17, 86-20, 86-21, 86-33,
86-38,86-43,86-52,86-53,86-57,86-62,86-87,86-89)
describe I&C technician personnel errors. Seven LERs
(354/86-2,86-6,86-13,86-15,86-49,86-55,86-66)
describe events initiated due to I&C procedural errors.
Seven of these LERs initiated ECCS equipment and are
identified in Section V.D.1.a. of this report.
c. Actuation of Control Room Emergency Filtration (CREF) System
Eight LERs (354/86-12, 86-16, 86-17, 86-25, 86-36, 86-47,
86-74,86-75) describe inadvertent actuations of CREF.
Six CREF actuations occurred due to drift of the high
voltage power supply to the ventilation duct radiation
monitors. One actuation resulted from an I&C technician
error during a surveillance test and another actuation
was a result of a design deficiency. The licensee has
replaced all high voltage power supplies which have
caused inadvertent CREF actuations with an upgraded model.
. -
52
2. AE00 Review
The Office for Analysis and Evaluation of Operational Data
(AE0D) assessed fifteen of the LERs submitted during the
assessment period using a refinement of the basic methodology
presented in NUREG-1022, Supplement 2. The results of this
evaluation, which was sent to the licensee by letter dated
January 9,1987, indicate that Hope Creek has an overall LER
score approximately equal with the industry average.
The principal weaknesses identified in the LERs, in terms of .
safety significance, involve the requirement to provide identi-
fication of failed components and the requirement to discuss
the safety consequence of the event. The failure to adequately
identify the manufacturer and model number of the components
that fail prompts concern that others in the industry won't
have immediate access to information involving possible generic
problems. Deficiencies in the safety assessment discussions
cause concerns about whether the potential safety consequences
of each event are being identified and evaluated.
A strong point for the Hope Creek LERs evaluated is the discus-
sion of the mode, mechanism, and effect of failed components.
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.. .
53
TABLE 1
TABULAR LISTING OF LERs BY FUNCTIONAL AREA
HOPE CREEK GENERATING STATION
(November 1, 1985 - November 20,1986)
Area Cause Code
A B C D E X TOTAL
A. Plant Operations 13 6 2 12 12 45
B. Radiological Controls 3 1 1 5
C. Maintenance 6 1 1 8
D. Surveillance 13 1 9 4 27
F. Security and Safeguards
G. Outage
H. Preoperational and
Startup Testing
I. Licensing Activities
J. Training and Qualification
Effectiveness
K. Assurance of Quality 1 1 1 3
Other 1 1
Totals 35 10 12 18' 14 89
Cause Codes: A. Personnel Error
B. Design, Manufacturing, Construction, or Installation Error
C. External Cause
D. Defective Procedure
E. Component Failure
X. Other
_. ._
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.. .
54
TABLE 2
LER SYNOPSIS
Hope Creek Generating Station
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION
86-001 2/15/86 A Damaging of "D" Diesel Generator
86-002 4/13/86 D Inadvertent "B" Channel LOCA
Signal During Surveillance Test
Performance
86-003 4/15/86 A Inadvertent RPS Initiation During
Performance of NMS Component
Troubleshooting Activities86-004 4/16/86 E Noncoincident Scram Signal
Resulting from Neutron Monitoring
System Component Failure
86-005 4/16/86 A FRVS Inoperability During Core
Alterations86-006 4/17/86 D Primary Containment Isolation
Resulting From a Procedural
Inadequacy
86-007 4/20/86 X B Channel Engineered Safety
Features Actuation
86-008 4/24/86 A Missed Surveillance During
Initial Core Loading Due to
Personnel Error
86-009 4/25/86 A Inadvertent RPS "A" Trip System
Initiation During Surveillance
Testing
86-010 4/26/86 X A Channel Engineered Safety
Feature Actuation
86-011 5/02/86 A Loss of Off-Site Power
86-012 5/4/86 E Control Room Emergency Filtration
Actuation Resulting From
Equipment Malfunction
. .
55
1
Table 2 (Cont'd)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION
86-013 5/6/86 D Inadvertent Isolation of RWCU
System During Surveillance Test
Performance
86-014 5/6/86 X D Channel Engineered Safety
Feature Actuation
86-015 5/6/86 D Spurious A Channel LOCA
Initiation
86-016 5/8/86 E A Control Room Emergency
Filtration Initiation
86-017 5/9/86 A Inadvertent Actuation of the "A'
Control Room Emergency Filter
Unit During Troubleshooting
86-018 5/12/86 B Failure of Service Water
Strainers86-019 5/13/86 X D Channel Engineered Safety
Feature Actuation
86-020 5/15/86 A D Channel Engineered Safety
Feature Actuation
86-021 5/15/86 A D Channel Engineered Safety
Feature Actuation
86-022 5/16/86 A Inadvertent Isolation of Reactor
Water Cleanup System
86-023 5/19/86 A B Channel Engineered Safety
Feature Actuation
86-024 5/25/86 D Inadvertent "D" Channel LOCA
Signal During Surveillance Test
Performance
86-025 5/30/86 A Power Supply Trip Causes Control
Room Emergency Filtration Chiller
Activation
86-026 5/30/86 E Automatic Start of a Control Room
Chiller
86-027 6/2/86 B Installation of Combustible Material
in the Traveling Screen Motor Room
.. .
56
Table 2 (Cont'd)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION
.86-028 6/7/86 X Spurious Actuation of the "A"
Channel of the Standby Liquid
Control System
86-029 6/11/86 E Automatic Start of "B" Control
, Area Chiller
86-030 6/18/86 A Automatic Start of "B" Control
Area Ventilation Train
86-031 6/29/86 A Reactor Scram Due to Personnel
. Error in Ranging IRMS86-032 6/30/86 E Initiation of Manual Scram for
Troubleshooting of Reactor Manual
Control System
4
86-033 7/3/86 A Inadvertent "B" Channel LOCA
- Signals During Instrument
Calibration Performance
- 86-034 7/12/86 E Main Steam Isolation Valve
Closure and Subsequent Manual Scram
86-035 7/4/86 E Reactor Scram Signal Originating
,
From The Neutron Monitoring
l System
,86-036 7/7/86 E Isolation of The "A" Control Room
Ventilation Unit Due to Radiation
4
Monitor Upscale Trip
86-037 7/12/86 A Failure to Comply With Technical
Specifications Action Statement
86-038 7/13/86 A Missed Channel Checks on Reactor
Protection and Isolation
Actuation Instrumentation
86-039 7/14/86 X "A" Channel LOCA Logic Actuation
86-040 7/9/86 A Inoperable RCIC Actuation
Instrumentation
.. -
57
Table 2 (Cont'd)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION
.,86-041 7/15/86 X Inadvertent HPCI System
-
Initiation
86-042 7/17/86 X Inadvertent HPCI System
Initiation
86-043 7/17/86 A Inadvertent HPCI System
Initiation Due to an I&C Error
86-044 7/25/86 E Reactor Scram on Low Level
.
Resulting from an EHC Transient
86-045 7/19/86 A Reactor Scram Due to IRM Ranging
Error
86-046 7/20/86 X Inadvertent HPCI System.
Initiation
86-047 7/29/86 E Actuation of the Control Room
Emergency Filtration System Due
To Radiation Monitor Spike
86-048 7/30/86 E Full Reactor Scram on Low Water
Level
86-049 8/1/86 D Missed Response Time Surveillance
Due to Procedure Inadequacy
86-050 8/1/86 D Reactor Water Cleanup System
Isolation on High Differential
Flow
86-051 8/3/86 E Reactor Water Cleanup Isolation
on Spurious High Temperature Trip
86-052 8/20/86 A Violation of the Surveillance
Requirements for the Suppression
Pool Temperature Monitoring
System
86-053 8/4/86 A "A" Channel LOCA Logic Actuation
86-054 8/4/86 A "A" Channel LOCA Logic Actuation
and Full Reactor Scram
- -
,- .
'
58
Table 2 (Cont'd)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION
86-055 8/5/86 D Primary Containment Isolation Due
To Procedure Inadequacy
86-056 8/8/86 A Inoperable Re ..or Building to
Torus Vacuum ;reakers86-057 8/8/86 A Inadvertent Actuation of the "A"
Channel NSSSS Isolation Logic
86-058 8/8/86 A Failure to Sample Results in
Technical Specification Violation
86-059 E/14/86 X "B" Channel ESF Logic Actuation
86-060 8/16/86 8 Violation of Suppression Pool
Level Technical Specification
86-061 8/22/86 D Inadvertent HPCI System
Initiation
86-062 9/20/86 A Failure to Satisfy TS
Surveillance Requirement for
Leakage Detection Monitors86-063 8/28/86 B ASCO Solenoid Valve Air Supply
Prassure Rating
86-064 8/31/86 A Reactor Scram on Low Level
86-065 9/6/86 X Full Reactor Scram on Low Reactor
Water Level 3
86-066 9/7/86 D Missed Surveillance: Turbine
!
Bypass Valve Testing
'86-067 9/15/86 B SRV Acoustic Monitors Inop:
Seals Missing
i
86-068 9/17/86 A Missed Surveillance: North Plant
Vent
86-069 9/24/86 D Reactor Scram - IRM/APRM
86-070 10/22/86 A "C" Core Spray Pump Discharge
Pressure Transmitter Isolated
- .- , . - - , . . - - . . . . - - , , ,-m---, - - ~ . , . . -. - , -
. *
59
Table 2 (Cont'd)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION
86-071 10/4/86 B PASS Sample Valves Installed in
Less Favorable Orientation
86-072 10/3/86 X Inoperable Reactor Building
Exhaust Radiation Monitoring
Instrument
86-073 10/3/86 B Electrical Penetration Assembly
Installation Error
86-074 10/2/86 A Inadvertent Actuation of "B"
Control Room Emergency Filtration
Unit when Connecting a Recorder
86-075 10/5/86 B Inadvertent Actuation of "B"
Control Room Emergency Filtration
Unit During Troubleshooting
'86-076 10/5/86 X Inadvertent Automatic Start of
"B" Emergency Diesel Generator
,
'86-077 10/10/86 E Inadvertent Isolation of Reactor
Water Cleanup System
86-078 11/11/86 E RWCU Isolation
86-079 10/19/86 E RWCU Isolation on High Differential Flow
86-080 10/18/86 B Full Reactor Scram on Low Reactor
Water Level 3
86-081 10/19/86 E Isolation of Reactor Cleanup
i 86-082 10/28/86 A High Pressure Coolant Injection
System Inoperative
'86-083 10/30/86 E ESF Actuation
86-084 10/30/86 A North / South Plant Vent Monitors
,86-085 11/14/86 A Reactor Scram on High Pressure
86-086 11/14/86 X Reactor Building Ventilation Isolation
86-087 11/17/86 A ESF-A Channel NSSSS Isolation
86-088 11/18/86- D Loss of RHR Room Cooling
86-089 11/19/86 B RWCU Isolation Due to Loose Wire
!
. . - _
-- .- . - - - - - _ _ . - _ . - - - . . . - _ . . -
. .. . . .. .
.. .
60
TABLE 3
INSPECTION HOURS SUMMARY (11/1/85 - 11/30/86)
HOPE CREEK GENERATING STATION
HOURS % OF TIME
A. Plant Operations. . . . . . . . . . . . . . . . 3030 33
B. Radiological Controls and Chemistry. . ..... 592 6
C. Maintenance. . . . . . . . . . . . . . . . . . . 445 5
D. Surveillance . . . . . . . . . . ........ 823 9
E. Emergency Preparedness . . . . . . . . . . . . . 454 5
F. Security and Safeguards. . . . . ........ 348 4
G. Outages. . . . . . . . . . . . . . . . . . . . . N/A
H. Preoperational and Startup Testing . . . . . . . 3478 38
I. Licensing Activities . . . . . . . . . . . . . . N/A
J. Training and Qualification Effectiveness . . . . N/A
K. Assurance of Quality . ............. N/A
Total 9170 100
.. .. .. .
.
. . . _
. , - _ _ . . _ . ._ __. . _. _ _ _ . , .. __ __.
... .-
61
-
TABLE 4
ENFORCEMENT SUMMARY (11/1/85-11/30/86)
Hope Creek Generating Station
i
'
SEVERITY LEVEL
AREA 1 2 3 4 5 DEV TOTAL
OPERATIONS 4 4
RAD PROTECTION
s. MAINTENANCE
SURVEILLANCE 4 4
EMERGENCY PREP.
SEC/ SAFEGUARDS
.
'
DUTAGES
TRAINING EFFECTIVENESS
1 LICENSING
{ ASSURANCE OF QUALITY :
- PREOP /STARTUP 1 8 2 11
.,
1
'
. TOTALS: 1 12 6 19
,
>
.
I
i
!
!
,
e
I
,
- - - - - , , -.4,_t- 7..,..w, ,--__~.,,e -,_., ,. -
. - , - , ...,---w--
. -
62
TABLE 4 (Cont'd)
ENFORCEMENT SUMMARY
INSPECTION VIOL. FUNCTIONAL
REPORT REQUIREMENT LEVEL AREA VIOLATION
354/85-61 APPENDIX B 4 PREOP / MANDATORY WITNESS POINT BYPASS
12/01/85 01/12/86 STARTUP DURING PRE 0P TEST
354/85-61 APPENDIX B 4 PRE 0P/ INADEQUATE QUALITY CONTROL
12/01/85 01/12/86 STARTUP INSPECTION
354/85-65 APPENDIX J 5 PREOP / VALVE IMPROPERLY OPERATED IN
2/23/85 01/03/86 STARTUP PREPARATION FOR INTEGRATED LEAK
RATE TEST.
354/86-03 APPENDIX B 4 PREOP / INITIAL CRITICALITY PROCEDURES
01/06/86 01/17/86 STARTUP
354/86-06 APPENDIX B 4 PREOP / FAILURE TO FULLY TEST CORE SPRAY
01/13/86 02/09/86 STARTUP LOGIC
354/86-10 APPENDIX B 4 PREOP / BYPASSING OF MANDATORY WITNESS
01/27/86 02/07/86 STARTUP POINTS.
354/86-10 APPENDIX B 5 PRE 0P/ INADEQUATE REVIEW OF TEST RESULTS.
01/27/86 02/07/86 STARTUP
354/86-20 TECH SPECS 4 OPERATIONS FRVS INOPERABLE DURING CORE
03/17/86 04/30/86 ALTERATIONS.
354/86-20 TECH SPECS 4 OPERATIONS MISSED SBLC SURVEILLANCE TEST.
03/17/86 04/30/86
354/86-30 LCO 3.0.4 & 4 OPERATIONS TECHNICAL SPECIFICATION VIOLATION:
3.7.4 RCIC INOPERABLE DUE TO NO AUTO-SWAP
06/10/86 07/14/86 0F SUCTION TO SUPPRESSION P00L.
354/86-32 APPENDIX B 5 SURVEILLANCE FAILURE TO PERFORM POWER ASCENSION
06/23/86 07/03/86 TEST IN ACCORDANCE WITH APPROVED
PROCEDURES
354/86-35 APPENDIX B 5 SURVEILLANCE USE OF A POWER ASCENSION PROCEDURE
07/07/86 07/24/86 INAPPROPRIATE TO THE CIRCUMSTANCES
i
i
l
i
. .. _ - - . -
. -
63
, TABLE 4 (Cont'd)
. INSPECTION VIOL. FUNCTIONAL
REPORT REQUIREMENT LEVEL AREA VIOLATION
354/86-35 10 CFR 50 5 SURVEILLANCE FAILURE TO FOLLOW PROCEDURE AND
'
l 07/07/86 07/24/86 FAILURE TO ADEQUATELY REVIEW TEST
RESULT
, 354/86-41 TECH SPEC 3 PRE 0P/ REACTOR BUILDING / TORUS VACUUM
>
3.6.4.2 STARTUP BREAKER ASSEMBLIES INOPERABLE
08/13/86 09/02/86
'
354/86-41 TECH SPEC 4 PREOP / ACOUSTIC MONITORS NOT POWERED
3.3.7.5 STARTUP FROM UNINTERRUPTIBLE SOURCE
08/13/86 09/02/86
354/86-40 TECH SPEC 5 SURVEILLANCE UNAUTHORIZE0 OPERATOR AIDS
6.8.1
1
08/12/86 09/08/86
354/86-48 TECH SPEC 4 OPERATIONS CORE SPRAY PRESSURE TRANSMITTER
6.8.1 ISOLATED
10/14/86 11/17/86
'
354/86-49 10 CFR 50 4 PREOP / FAILURE TO FOLLOW PROCEDURE FOR
10/11/86 10/16/86 STARTUP TORQUING POSEMOUNT TRANSMITTER
354/86-53 LICENSE 4 PREOP / FAILURE TO COMPLY WITH LICENSE
NPF-57 STARTUP CONDITION C10 - DID NOT PERFORM
10/27/86 10/31/86 TIMELY 50.59 REVIEW
d
I
k
i
l
- - , - - , . , , . - - - - . _ , - - . _ _ . - , _ - . .,n_ - , - . . - . - - . - - - . . , . - - , - . , - - - - . - -._ - ---.-- . , . . - -
-__ __ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. .
64
TABLE 5
INSPECTION REPORT ACTIVITIES (11/1/85-11/30/86)
Hope Creek Generating Station
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
l
354/85-55 SPECIALIST 40 PREOP TEST PROGRAM
11/04/85 11/15/85
354/85-56 RESIDENT 251 ROUTINE RESIDENT INSPECTION
10/28/85 12/01/85
354/85-57 SPECIALIST 76 PRE 0PERATIONAL SECURITY PROGRAM REVIEW
11/12/85 11/15/85
354/85-58 SPECIALIST 829 AS-BUILT TEAM INSPECTION IN AREAS OF
12/02/85 12/13/85 MECHANICAL, ELECTRICAL, INSTRUMENTATION AND
CONTROL AND STRUCTURAL SYSTEMS
354/85-59 SPECIALIST 72 PREOPERATIONAL INSPECTION OF CHEMICAL AND
11/18/85 11/22/85 RADI0 CHEMICAL MEASUREMENT PROGRAM.
354/85-60 SPECIALIST 44 PRESERVICE INSPECTION PROGRAM
11/18/85 11/22/85
354/85-61 RESIDENT 265 ROUTINE RESIDENT REPORT. MAJOR FOCUS ON
12/01/85 01/12/86 PRE 0P TESTING.
354/85-62 SPECIALIST 105 STAFFING, TRAINING, QUALIFICATION OF
12/09/85 12/18/85 PERSONNEL AND LOCAL LEAK RATE TESTING.
354/85-63 SPECIALIST 34 CONSTRUCTION PROGRAM
12/16/85 12/23/85
354/85-64 RESIDENT 300 TECHNICAL SPECIFICATION REVIEW CONDUCTED BY
12/02/85 12/13/85 PARAMETER INC.
354/85-65 SPECIALIST 115 CILRT INSPECTION
12/23/85 01/03/86
354/85-66 SPECIALIST 63 FOLLOWUP ON GENERIC LETTER 83-28, QA
12/30/85 01/03/86 RECORDS AND MEASURING AND TEST EQUIPMENT.
354/86-01 SPECIALIST 142 FIRE PROTECTION AND FOLLOWUP ON
01/07/86 01/11/86 CONSTRUCTION PROGRAM 0"EN ITEMS.
354/86-02 SPECIALIST 146 PLANT PROCEDURES AND FOLLOWUP ON PREVIOUSLY
01/27/86 02/14/86 IDENTIFIED ITEMS.
.- .
65
Table 5 (Cont'd)
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
354/86-03 SPECIALIST 151 PREOP AND POWER ASCENSION PROGRAMS
01/06/86 01/17/86
354/86-04 SPECIALIST 74 QA PROGRAM OVERVIEW
01/06/86 01/16/86
354/86-05 SPECIALIST 95 PRE 0PERATIONAL WATER CHEMISTRY CONTROL
01/13/86 01/24/86 PROGRAM AND FOLLOWUP ON PREVIOUSLY
IDENTIFIED ITEMS.
354/86-06 RESIDENT 410 ROUTINE RESIDENT REPORT WITH EMPHASIS ON
01/13/86 02/09/86 PRE 0P TESTING.
354/86-07 SPECIALIST 50 RADIOLOGICAL CONTROLS INSPECTION
01/21/86 02/14/86
354/86-08 SPECIALIST 32 PREOPERATIONAL SECURITY PROGRAM REVIEW.
01/27/86 01/31/86
354/86-09 SPECIALIST 140 FOLLOWUP OF EMERGENCY PREPAREDNESS
02/03/86 02/03/86 IMPLEMENTATION APPRAISAL.
354/86-10 SPECIALIST 145 PRE 0PERATIONAL TEST PROGRAM IMPLEMENTATION.
01/27/86 02/07/86
354/86-11 SPECIALIST 44 RPV INTERNALS RECORD REVIEW
01/27/86 01/31/86 PRESERVICE INSPECTION PROGRAM.
354/86-12 SPECIALIST 103 PREOPERATIONAL AND STARTUP PROGRAM
02/10/86 02/21/86 IMPLEMENTATION.
354/86-13 SPECIALIST 82 FOLLOWUP ON OUTSTANDING ITEMS AND
02/10/86 02/14/86 MECHANICAL SNUBBER INSPECTION.
354/86-14 SPECIALIST 73 SAFETEAM INSPECTION
02/03/86 02/07/86
354/86-15 RESIDENT 501 ROUTINE RESIDENT INSPECTION WITH EMPHASIS
02/10/86 03/16/86 ON OUTSTANDING ITEMS FOLLOWUP AND
PRE 0PERATIONAL TESTING.
354/86-16 SPECIALIST 0 OPERATOR LICENSING EXAM.
02/24/86 03/24/86
354/86-17 SPECIALIST 36 MAINTENANCE AND I&C SURVEILLANCE
02/24/86 02/28/86 PROCEDURES.
- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
'
/
. 4 J
66 , ,
'
TABLE 5 (Cont'd)
,
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
354/86-18 ' SPECIALIST 75 PREOP, STARTUP', CILRT, AND SURVEIL 1.ANCE
03/03/86 03/14/86,' TEST INSPECTION.
']
,
j
354/86-19 SPECIALIST 35 FOLLOWUP ON OPEN ITEMS. 's -
03/03/86 03/06/86 -
.
354/86-20 RESIDENT 530 PbuTINERESIDENT ~
03/17/86 04/30/86 7
354/86-21 7 SPECIALIST 108 %REOPsAND STARTUP PROGRAM REVIEW.
03/12/86 03/21/86
354/86-22 SPECIALIST 128 INSPECTION BY 3 REGION-BASED INSPECTORS OF
03/31/86 04/11/86 PREVIOUS INSPECTION FINDINGS.
354/86-23 SPECIALIST 156 ROUTINE INSPECTIOM BY 5 REGION-BASED
.04/14/86 04/25/86 c INSPECTORS OF PAEVIOUS INSPECTION FINDINGS
354/86-24 c SPECIALIST 71 INSPECTION BY 2 REGION-BASED INSPECTORS OF
04/28/86 05/09/66 PREOPERATIONAL TESTING.
354/86-26 . RESIDENT 271 R,0VTINE RESIDENT INSPECTION J
05/01/86 06/09/86
354/86-27 SPECIALIST 100 INSPECTION FINDINGS OH PREVIOUS
5/19/86 5/30/86 INSPECTIONS.
.
354/86-28 SPECIALIST 65 SECURITY INSPECTION OF TRAINING PROGRAM FOR 4'
5/27/86 5/30/86 SECURITY PERSONNEL.
354/86-29 SPECIALIST .38 SPECIAL INSPECTION IN SUPPORT OF LICENSING
5/27/86 5/30/86 ACTION RELATED TO LICENSEE REQUEST DATED
5/13/86 TO DELETE FIRE PROTECTION TECH. '
SPEC.
354/86-30 RESIDENT 353 ROJTINE FOLLOWUP INSPECTION.
6/10/86 7/14/86
354/86-31 SPECIALIST 75 INSPECTION OF PREVIOUS INSPECTION FINDINGS,
6/9 /86 6/20/86 POWER ASCENSION TEST PROGRAM.
354/86-32 SPECIALIST 94 IhSPECTION OF POWER ASCENSION TEST PROGRAM t
6/23/86 7/3 /86 COVERING INITIAL CRITICALITY
354/86-33 SPECIALIST 28 UNANNOUNCED INSPECTION OF RADI0 ACTIVE WASTE
6/16/86 6/18/86 (RADWASTE) PROGRAM DURING INITIAL FUEL LOAD
ACTIVITIES.
L-_- .
_ -
-
4
,
,3
s
-
67
TABLE 5 (Cont'd)
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
354/86-34 SPECIALIST 0 OPERATOR LICENSING EXAMINATIONS
7/7/86 7/11/86
.354/86-35 SPECIALIST 90 INSPECTION OF OVERALL POWER ASCENSION TEST
~ -27/7/86 7/24/86 PROGRAM, QA/QC INTERFACES AND TOURS OF THE
FACILITY
354/86-36 i RESIDENT 204 ROUTINE RESIDENT INSPECTION
7/15/86 8/11/86
354/86-37 SPECIALIST 26 INSPECTION OF PREVIOUS FINDINGS IN
7/30/86 8/1 /86 RADIATION AREAS
'_ , ,' 354/86-38 SPECIALIST 45 POWER ASCENSION TEST PROGRAM, PROCEDURE
x8/11/86 8/22/86 REVIEWS, QA/QC INTERFACES AND TOURS OF THE
FACILITY.
-354/86-39 SPECIALIST 36 INSPECTION OF RADIOACTIVE WASTE PROGRAM
8/12/8S; 8/15/86
'
354/86-40 RESIDENT 92 ROUTINE RESIDENT INSPECTION
,
, 8/12/86 9/8/86
'
'
354/86-41 (RESIDENT 93 SPECIAL INSPECTION OF THE CAUSES FOR
8/13/86. 9/02/86
'
INOPERABILITY OF REACTOR BUILDING TO
SUPPRESSION CHAMBER PRESSURE RELIEF SYSTEM.
354/86-42' CANCELLED
354/86-43 SPECIALIST 31 INSPECTION OF OVERALL POWER ASCENSION TEST
9/2/86 9/5/86 PROGRAM.
354/86-44 SPECIALIST 72 ROUTINE INSPECTION OF SOLID RADIOACTIVE
9/08/86 9/12/86 WASTES (RADWASTE) PROGRAM DURING STARTUP
ACTIVITIES.
354/86-45 SPECIALIST 154 INSPECTION OF THE LICENSEE'S IMPLEMENTATION
9/22/86 9/26/86 AND STATUS OF NUREG-0737
354/86-46 SPECIALIST 60 INSPECTION OF OVERALL POWER ASCENSION TEST
9/11/86 9/19/86 PROGRAM
354/86-47 RESIDENT 283 ROUTINE RESIDENT INSPECTION
/'
9/9/86 10/13/86
-
354/86-48 RESIDENT 195 ROUTINE RESIDENT INSPECTION
10/14/86- 11/17/86
\<
,
f-
'
, *
,
. e
/
68
1
-. .
TABLE 5 (Cont'd)
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
354/86-49 SPECIALIST 50 INSPECTION OF OVERALL POWER ASCENSION TEST
10/11/86 10/16/86 PROGRAM
354/86-50 TEAM INSP 538 INSPECTION OF THE LOSS OF 0FFSITE POWER
9/25/86 10/3/86 TEST ON' SEPTEMBER 11, 1986 i
354/86-51 SPECIALIST 227 INSPECTION OF EMERGENCY PREPAREDNESS
11/10/86 12/1/86 PROGRAM AND IMPLEMENTATION
354/86-52 TEAM INSP 344 OPERATIONAL READINESS TEAM INSPECTION
10/20/86 10/31/86
354/86-53 SPECIALIST 37 INSPECTION OF OVERALL POWER ASCENSION TEST
10/27/86 10/31/86 PROGRAM
354/86-54 CANCELLED
al'
354/86-55 SPECIALIST 48 INSPECTION OF OVERALL POWER ASCENSION TEST e-
I/ O 11/10/86 11/19/86 PROGRAM
s
e
f %
s.
,
1
9 4 '
l$
a
!
= , - - , . -- _.
, e
69
TABLE 6
UNPLANNED AUTOMATIC SCRAMS AND SHUTDOWNS (11/1/85 - 11/30-86) I
i
HOPE CREEK GENERATING STATION I
Root Functional
Date Power Level Description Cause Area
1. 4/15/86 Shutdown IRM high scram due to Personnel Surveillance
bumping IRM cable. error
2. 4/16/86 Shutdown High APRM scram due to Equipment
failure of 1 LPRM input. failure -
random
3. 4/25/86 Shutdown IRM high scram due to Personnel Surveillance
bumping IRM cable. error
4. 6/29/86 less than 1% IRM high scram, caused by Personnel Operations
downranging the wrong IRM error
(non-coincident RPS mode).
6/29/86 Restart
5. 6/30/86 1% Manual scram to trouble- Equipment
shoot reactor manual failure -
control system. random
7/1/86 Restart
6. 7/4/86 2% High APRM trip due to a Equipment
momentary upscale spike of failure -
an LPRM. A half scram was random
already present due to
inoperable instrumentation.
7/5/86 Restart
7. 7/12/86 1% Manual scram after the Equipment
MSIVs were automatically failure -
closed due to steam flow random
transmitter drift.
7/13/86 Restart
8. 7/19/86 less than 1% IRM high scram caused by Personnel Operations
downranging vice upranging error
2 separate IRMs.
7/20/86 Restart
, < , -
70
TABLE 6 (Cont'd)
Root Functional
Date Power Level Description Cause Area
9. 7/25/86 3% Reactor vessel low level Inadequate Surveillance
Scram caused by the loss of procedure
feed flow after RFP trip
on swell induced high level.
7/26/86 Restart
10. 7/30/86 6% The EHC power supply failure Equipment
caused the bypass valves to failure -
open. The resulting swell random
tripped the feed pumps and
level could not be restored
prior to the low level scram.
7/31/86 Restart
11. 8/31/86 5% Reactor feed pumps tripped Equipment Preop /startup
due to level control dif- failure
ficulties as a result of
inadequate minimum flow
valve tuning. Operators
were unable to reset the
trip before low level scram.
9/1/86 Restart
12. 9/6/86 33% Low level scram while Personnel Operations
swapping feed pumps as a error
result of unstable control
of "C" RFP prior to
completion of tuning.
9/6/86 Restart
13. 10/18/86 50% Feedwater control test box Faulty test Preop /startup
had internal wiring errors box wiring
that caused RFP runback and
low level' scram.
10/19/86 Restart
14. 11/14/86 98% High pressure scram due to Procedure Preop /Startup
control valve closure test deficiency surveillance
exceedino maximum combined
flow limit.
11/28/86 Restart