ML20148B186
| ML20148B186 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/04/1988 |
| From: | Hawkins F, James Heller, Hooks K, Jury K, Greg Pick, Scott W, Shannon M, Weiss S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20148B044 | List: |
| References | |
| 50-321-87-31, 50-366-87-31, NUDOCS 8803210374 | |
| Download: ML20148B186 (14) | |
See also: IR 05000321/1987031
Text
.
,.
.
.
.
.
U.S. NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
Report No.:
50-321/87-31
Docket No.:
50-321
License No.:
50-366/87-31
50-366
Licensee:
Georgia Power Company
. Post Office Box 4545
Atlanta, Georgia 30302
'
Facility:
'Edwin I. katch Nuclear Power Plant, Units 1 and 2
Inspection At:
E.I. Hatch, Baxley, Georgia, November 30-December 11, 1987
Inspectors: "
A.4fg
3/V/IP
Kennethg. Hooks, Senior Operations Engineer
(D6te)
NRR (Team Leader)
g
61 A4
3/php
Wayne E/ Scott, Quality Operations Engineer
(Dite)
NRR (Assistant Team
adep
YdMw $
to$4
Yikf
Names K/ Heller, Resident Inspector
'( 7a te)
Region 'lIIw [. $Y
s/4bP
IL Keith R/ Jury, Reactor Inspector
'(06 te )
Region II
%>
3M///
Gregory A. Pick, Reactor Inspector
~( D'a te )
Region IV
d
[
3d/8
Melvin 7. Shannon, Reattor Inspector
'( 7 ate)
V Region II
i
Reviewed By:
Wf
5)LI196
F'. C. Hawkins, Chief, Quality Operations Sectien (Oate)
3
- f
Approved By:
l
S. H. Weiss, Chief, Quality Assurance Branch
' ( D'o t e )
i
8803210374 080316
ADOCK OJ000321
1
l
O
ppg
'
_ - .
_
-
..
--
.
-
.
.,
.
,
.
.
,,
.
.
Summa ry:
a.
Areas Inspected
This special, announced team inspection was the fifth in a series of NRC
Headquarters-directed Quality Verification Function Inspections (QVFis).
The~ inspection was performed to assess the licensee's quality veriff-
cation organizations' ability to identify, solve, and prevent safety-
significant deficiencies in the functional areas of plant operations and
modificaticns of plant' systems and components.
,
b.
Results
The NRC inspectors observed six conditions in the functional ares of
plant operations that were considered to be less-than-optimum and
three Potential Enforcement Findings in the functional area of modifica-
tions of plant systems and components.
The three Potential Enforcement
Findings are associated with a modification to the reactor water clean-
,
up (RWCU) system accomplished in 1985; they do not appear to be repre-
sentative of current quality verification activities.
The NRC
inspectors determined that the licensee's current quality verification
activities in the areas of operations and modifications are generally
adequate.
.
b
,
e
i
e
!
t
,
l
t
1
1
l
i
F
r
.
.,
2
J
1
\\-..
. ...
. _ . . . _ . ._ _ _ - - _ - _
, _ _ , . . _ . . _ _ _ _ . , . . , _ _ , _ . . . . - - _ _ _ _ _ . _ , , . . . - - _ _ .
-
.
,.
<.
"
1.0 INTRODUCTION
This special, announced team inspectio
n E.I. Hatch Nuclear Power Plant was
performed to assess the licensee's qual uy verification (QV) organizations'
ability to identify, resolve, and prevent safety-significant deficiencies in
various f.unctional areas.
If the QV organizations are technically credible,
they can and should help define identified deficiencies, provide insight into
the root cause of deficiencies, and approve and confirm the resolution of
deficiencies in a technically meaningful way. The inspection also assessed
line management's ability to ensure that identified deficiencies are dealt
with promptly and completely.
The inspection was the fifth in a series of NRC Headquarters-directed
inspections perfonnid under the guidance of NRC Inspection Manual Temporary
Instruction (TI) 2515/78, "Inspection of Quality Verification Functions."
These multi-discipline team inspections use interviews of licensee personnel,
direct observations of in-process activities, and review of work documents to
evaluate the effectiveness of quality verification organizations and
management.
Quality Verification Function Inspections (QVFIs) are not intended to verify
licensee compliance to administrative controls; they are intended to verify
the technical adequacy of safety-related activities.
If deficiencies are
found in these activities, the underlying procedures and administrative con-
trols are reviewed. The results of these inspections will be improvements
in operational safety through inspection processes that are focused on
activities that affect plant reliability and safety.
This QVFI at Hatch emphasized plant operations and mudifications of plant
systems and components. Selective samples were reviewed in these and closely
associated areas to identify safety-significant problems to be used as the
vehicles for assessing the effectiveness of quality verification. The details
and findings of these reviews follow. The more significant findings of the
inspection team have been categorized as observations and Potential Enforce-
ment Findings.
Observations are items that do not violate any regulatory requirements
and may not violate plant procedures, but that appear to be less than
optimum. Potential Enforcement Findings (PEFs) are apparent violations of
regulatory requirements that will be further evaluated by NRC Region 11
management for possible enforcement action.
2.0 PLANT OPERATIONS
The NRC inspectors evaluated the licensee's quality verification activities
in the area of plant operations. The evaluation was perfomed through direct
obsirvation of activities, interviews with licensee personnel, and reviews of
doe nentation. Activities observed included control room routines, back
shs
performance of quality assurance (QA) audits and surveillances,
Technical Specification (TS) surveillance tests, and various plannino and
status meetings. Documentation that was reviewed included selected Deficiency
Cards, QA audit and surveillance reports, Licensee Event Reports (LERs),
and Plant Review Boar d (PRB) meeting minutes.
Interview * were conducted with
3
-
-
-
_ _ .
.,
.___
. _ -
,
-- _
..
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _
.
..
.
.
..
-
.
personnel from the Departments of Nuclear Safety and Compliance (NSAC),
' Quality Assurance, and Operations.
2.1 Control Room Operations
s.
The NRC inspectors observed activities, interviewed personnel, and reviewed
documents in the combined control rooms. Activities observed included normal
shift decorum, shift turnover, a montnly test of switchyard breakers, and an
actual loss of automatic feedwater control.
Personnel interviewed included
On-Site Operations Supervisors (OS0Ss), Shif t Supervisors (SSs), and various
other shift personnel.
Documents were reviewed on the subjects of standing
orders, clearances, limiting conditions for operations (LCOs), surveillances,
operator aids, and emergency operating procedures (E0Ps).
Plant personnel appeared knowledge"able and professional and conducted them-
selves with 6ppropriate decorum.
This was esper.ially evident during a
- The E0Ps, howe"er, did not give the operators satisfactory assistance. The
NRC inspectors found that the flowchart-format E0Ps in the control room were
not adequately legible (Item No. 87-31-01, Observation).
The licensee's site
QA Audit 87-P0-2A of OctoT6r 15,1987, also identified legibility problems
with those E0Ps
The "muster" E0Ps which are largg and unwieldy, are
e
legible, although the flowchart arrows are sometimes difficult to follow.
However, the smaller versions of the master E0Ps that are placed at various
locations in the control room have been reduced fri size to tne point that
clarity of wording and logic flow paths is lost. As a consequence, they are
not s3tisfactory for use .in effecting a safe recovery from a plant casualty.
The inspectors also noted that, although there is a notebook of Annunciator
Response Procedures in a three-ring binder at each control panel, there is
no convenient way to free the operator's hands so he can follow a procedure
,
!
and still operate the par el. That is, the operator cannot lay the book on
I
the panel while he operates switches, and there is nowhere else to put the
(
book whare it can se read.
2.2 Licensee Event Reports
{
The NRC inspectors reviewed the site's 1987 Licensee Event Reports (LERs).
Two recurring problems were noted.
The first was overheating of vital power
Inverters, which caused three reactor trips (two at power) on low reactor
vessel water level.
The licensee attributed the overheating, in part, to
the fact that the ambient river water (service water) temperature reached
.
95'F and, in part, to equipment aging.
The licensee initiated corrective
l
action that included repair of the inverters and addition of portable air
conditioners.
The NRC inspectors interviewed plant engineers and four.d that
they were aware of industry problems with aging and overheating of plant
equip:nent .
The licensee's proposed long-term corrective action, which is
still under evaluation, consists of deletion, modification, or replacernent/
modification of the inverters.
Pending implementation of long-tem corrective
,
!
detion, the potential for further, similar challenges to the units' safety
systems remains.
4
.
.
.
_
_ _ _ _ _ _ _ _ _ _
_ _
__
__
__ ,
.
.
!
.
,.
.
The second recurring problem noted by the NRC inspectors dealt with plant
procedures.
In all cases, the licensee's corrective action included refer-
ence to the Procedure Upgrade Program (PUP).
Fundamentally, the PUP consists
of three phases.
The first phase was writing Administrative Procedure 10
AC-MGR-003-03, which established the controls and methods for procedure
development, revision,' review, and approval.
The second phase is the
rewriting of plant procedures.
That is being dune by a contractor, with
strong interaction by plant personnel.
The third phase will be to replace
the contractor with plant personnel.
Based on their reviews of these two problem areas, the NRC inspectors con-
cluded that the licensee has a strong interest in improving plant safety by
correcting and improving procedures.
'
2.3 Surveillances and Audits.
<
The licensee's surveillance schedule is generated quarterly and is based on
QA management's perception of plant problems.
A sumary of surveillances
performed, which is addressed to the Plant Manager and the Plant Support
Manager, is issued monthly.
The NRC inspectors reviewed the surveillance
sumaries from January through October 1987. The summaries showed that, in
the monthly system wa kdowns, the QA Department identified unsatisfactory
i
material conditions, discrepancies between installed equipment (as-built
condition) and the piping and instrument drawings, procedure weaknesses,
equipment requiring repairs, and components that were not properly identified.
Surveillance findings are forwarded to plant personnel by memorandum and are
included in the QA trendi,ng system.
The memorandum requires that the
addressees respond by outlining corrective actions.
The auditor of record
has the opportunity to evaluate the corrective action when the findings are
Deing closed.
.
A selected sample of QA audit reports and trend reports was reviewed by the
NRC inspectors.
The audits appeared adequate, although they are more procedure-
oriented than the surveillances. The trend reports provide an overview for
management but little in the way of details.
However, the details to back up
the trend results 3re available to management upon inquiry.
The NRC inepectors also reviewed a selected system surveillance check list and
found the check list to be adequately detailed, requiring the surveillance
team to look at installation of shielding, cabling, electrical equipment,
pumps, valves, motors, instrumentation, piping, hangers, snubbers, and
supports.
The surveillance check list appears to reflect lessons learned
from past plant problems and industry problems.
Interviews with licensee
auditors indicated that the check list is a living document and can be
5
enanged as ne::essary.
The NRC inspectors accompanied an auditor during an extensive backshift
(off-heurs) material condition surveillance, documented in Surveillance
Report 87-0RM-113.
The auditor was thorough, professional, and knowledgeable
of the check list and the plant procedures.
In addition, the auditor
prorrptly identified to the appropriate cognizant individuals the items that
needed near-term attention,
for example, three containers of coated weld
5
.
_ _ _ _ _ _ _
.
..
.
,
.
.
rods were found improperly stored in a tool box. Before leaving the site,
the auditor accompanied the on-shift maintenance manager to the scene, where
they reviewed the safety aspects of the condition and discussed possible
enurses of action for resolution of the condition and prevention of its
recurrence.
2.4 Plant Review Board Meetings
The NRC inspectors revieweo the minutes of Plant Review Board (PRB) meetings.
In particular, the September 18, 1987, writeup for deficiency card 1-87-834
was evaluated.
The deficiency card reported that data from an in-service
inspection (ISI) test:showed that a' service water pump was inoperable; however,
the crew performing the test did not recognize the inoperability.
The
inoperability was identified approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> later by the ISI engineer.
The PR8 minutes (page 5) imply tha't the action statement begins after the
determination of inoperability is made.
Although this is generally true, in
cases where sufficient information exists to determine inoperability at the
time a test is performed and shift personnel err in not determining
- inoperability, the inoperability must be treated as if it had been properly
determined wben that sufficient information was first aailable.
The NRC
inspectors discussed this issue with the licensee and verified that the
licensee understands NRC's position with respect to such issues.
The NRC
inspectors verified that the proper Limiting Condition of Operation (LCO) was
'
adhered to in this case.
2.5 QA Personnel
'
The NRC inspectors interviewed a number of personnel from the licensee's QA
organization.
They appea' red knowledgeable and professional. However, the
inspectors noted that the single QA auditor with plant operations experience
was due to leave the site shortly after the inspection. This lack of organi-
zational operations experience is considered a potential weakness (Item No.
87-31-02, Observation).
The inspectors noted that the lionsee has advertised
to fill the expected vacancy with a person who hus an operational background.
The inspectors encouraged the licensee to staff the QA organization with
more personnel experienced in operations to enable the QA organization to
perform more meaningful activities in monitoring plant operations.
The NRC inspectors reviewed the Quality Assurance Department Training Guide,
which establishes the QA training process and curriculum.
Requirements for
the various positions and levels are detailed, and trainees are required to
obtain appropriate approvals for the various steps of the qualification
processes.
The inspectors discussed the Quality Checker Program with licensee personnel.
Under this program, employees are assigned to perform quality surveillances
within their own departments for 1 month. They receive one day of familiari-
zation training on the requirements of Appendix 8 to 10 CFR 50 and the plant
implementing documents.
They then perform as quality checkers and gain
awareness of quality requirements as well as of quality problems, processes,
and tracking systems at the site. This program appeared to be an excellent
tool for sensitizing line personnel to the concepts of quality attainment and
verification and ensuring that they contribute meaningfully to plant relia-
bility and safety.
6
.-__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
.
.
.
. -
2.6 Health Physics
The licensee requires personnel to use the hand-and-foot radiation monitor
locatej at the entry point to the control room.
During one entrance to the
control room, the NRC inspectors alarmed the monitor.
The licensee's General
Employee Handbock requires an individual to notify health physics (HP) per-
sonnel immediately 1f the instrument indicates contamination. However, at
this location, there is neither a phone nor any other mechanism by which to
notify HP. The need for a phone or other commur.ication mechanism at the
monitoring station was taken under advisement by the licensee's management
(Item No. 87-31-03, Observation).
In this instance, the' licensee employee accompanying the NRC inspectors went
.
to HP and was instructed to take plastic shoe coverings to the inspectors and
escort the inspectors.to HP-for an additional contamination check.
During
later discussions with the HP/ Chemistry Superintendent, the inspectors were
informed that HP personnel should always go to the potentially contaminated
individual to guard against further spread of ::ontamination.
The HP/ Chemistry
. Superintendant stated that steps would be taken to ensure that on-shift HP
personnel are aware of the policy (Item No. 87-31-04, Observation).
The hand-and-foot monitors alarm frequently, apparently reacting to
radioactive noble gasses from fission product leaks that cling to clothing.
In the process of dealing with these frequent alarms, the inspectors recog-
nized a traffic pattern that has the potential for. spreading contamination.
People entering the reactor building must pass through the same relatively
narrow hallway and HP access point that people exiting the building must also
use.
People exiting go beyond the HP check point, use the hand-and-foot
monitors, and then return to the HP check point if the monitor indicates
contamination.
Passing near the monitors are otherwise uncontaminated people
moving into and out of the access point.
If a person were contamintted, the
potential exists for the traffic to spread the contamination both into the
.'
plant and out among the various offices.
This could continue until the per-
,
son alarmed a monitor at the plant gate (Item No. 87-31-05, Observation).
,
2.7 Storage and Contml of Equipment and Compounds
The inspectors found three safety-related service water pumps, both plan'c
service water and residual heat removal (RHR) service water, stored in the
open area adjacent to the intake structure.
One was rusty and its components
were separated.
A second, also not intact, was nearby.
The third, which
appeared to have been rebuilt, was on blocks with the top of its shaft
wrapped and plastic covering the opening in the suction bell; however, the
plastic was ripped and the intake bell was open to the elements. Lying about
on the asphalt, in a bucket, and in a covered bin were large numbers of nuts
and bolts, impellers, and other items apparently associated with these and
other pumps. Additionally, a canvas bag of nuts and bolts was found lying 01
the grating over the intake structure.
The bag was still attached to a rope
apparently used to remove the material from the structure or to lower the
safety-related fasteners into the structure.
There was no way the inspector could ascertain whether all the equipment and
material had been stored and handled in a way that was appropriate to the
7
--_
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
'l
-
.
.
.
-
.
requirements of its safety-related status. Licensee personnel stated that
they were preparing a memorandum to the maintenance personnel to clarify that
discarded material snoulo not be left at the job site.
They also stated that
they would be ccnsidering establishing ready-for-use covered bins at that
location for both new ano reusable items (Item No. 87-31-06, Observation).
2.9 Suma ry
QA audits and surveillances were of adequate technical depth and generally
performance oriented, identifying real problems as well as procedural
discrepancies. QA personnel were knowledgeable and professional, although
the addition of personnel with operations experience would enhance the
organization. Control room personnel were knowledgeable and professional,
and the operations observed were adequate, with the possible exception of the
use of illegible E0P flow charts.
Several HP practices were observed that
appeared to be less than optimum, but HP performance was adequate overall.
'
A potential problem with the identification of safety-related and non-safety-
, related fasteners for the service water system was observed and reported to
licensee management.
In the functional area of plant operations, the NRC inspectors found the
licensee's quality verification organizations to be generally effective in
identifying, solving, and preventing safety-significant deficiencies, and
they found the licensee's management to be generally effective in dealing
promptly and completely with identified deficienci'es.
3
3.0 MODIFICATIONS OF PLANT SYSTEMS AND COMPONENTS
Tha NRC inspectors reviewed open and closed Design Change Requests (DCRs),
J
t
l
permanent and temporary plant modifications, jumpers, post-modification
tests, and the licensee's commitments to assess the effectiveness of the
licensee's quality verification organizations in identifying, resolving, and
'
l
preventing recurrence of safety-significant technical oeficiencies in the
modification of plant systems and components.
Various modification packages
'
were reviewed to ensure that design inputs, analyses, reviews, approvals, and
post-maintenance / modification testing were adequate. The NRC inspectors also
assessed the effectiveness of line management in ensuring that identified
deficiencies were dealt with promptly, completely, and correctly. Plant pro-
cedures were reviewed to the extent required to understand how the licensee
perfonns work in the areas being inspected.
3.1 Open DCRs
The NRC inspectors reviewed various DCRs that were categorized for implemen-
tation or voiding.
Open DCRc to be implemented were reviewed to determine
why they were open and to determine if any safety-significant DCRs were not
being implemented as necessary.
The DCRs that were categorized for voiding
were reviewed to determine if the modifications being voided were important
and, if so, why they were being voided after they had been approved by the
PRB for implementation.
The NRC inspectors found that DCRs can be voided for various reasons:
the
DCR is no longer applicable to plant configuration; the DCR is "old" and, if
necessary, will be replaced by a DCR under the new program; the work entailed
8
1l.
__
_ _ _ _ _ _ _ _
___._ _ _ ,.
_ _ _ _ _
_
_ _ _ , _ _
__ ,
_ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ ____
_
_ _ _ _ _ _
.';
-
.
.
-
.
in the DCR has been performed under a different DCR; or the DCR may never have
been needed.
Upon review of the DCRs, the NRC inspectors were concerned that
the "old" DCRs (e.g. DCRs78-105, 78-127, and 82-283) were, in fact, necessary
modifications and would increase the already large backlog of open DCRs.
Additionally, many of the DCRs scheduled to be voided (e.g. DCRs79-487,
82-136,82-137, 83-276,84-230, and 85-189) appeared to warrant further evalu-
ation or implementation.
The apparent misclassification of these DCRs by the
licensee is of conce'rn, because these DCRs might not be re-evaluated until
their scheduled voiding date and wculd not be rescheduled for the necessary
implementation until that time.
3.2 Closed DCRs
Approximately 20 closed DCRs were reviewed. No discrepancies were identified
during this review.
The documentation for 10 CFR 50.59, unreviewed safety
,
questions, appeared-to have improved since the last Region II inspec: ion.
It also appeared that licensee management had taken the initiative to reduce
the backlog of open DCRs.
' 3.3 Open Temporary DCRs
Approximately five open temporary DCRs were reviewed.
With the exception of
DCR 85-007, which is discussed in Section 3.4, all were found to be adequate.
3.4 DCR 85-007, "Reactor Water Clean-Up (RWCU) System High Differential Flow"
In December 1984, on three separate occasions, RWCU system primary contain-
ment isolation valve IG31-F001 inside containment failed to close as required
upon receipt of a high differential flow signal with RWCU pump 1G31-C0018
running; however, the second (series) RWCU system isolation valve, IC31-F004,
did close. The failure of valve 1G31-F001 to close or, high differential flow
signal (isolate) was identified in LER 84-029 because the RWCU valves are
primary containment isolation valves; thus, the closure of IG31-F004 consti-
tuted an engineered safety feature (ESF) actuation.
This condition was also
reported under 10 CFR 50.72, because it was a failure of valve IG31-F001 to
close as required by the Hatch Unit 1 TS 3.7.D.1 and Table 3.7-1.
The corrective action of LER 84-029 consisted of backflushing the flow trans-
mitter sensing line to remove air that was determined to be the cause of the
pressure spikes that were the initiation signal (high differential flow).
However, the RWCU system was not tested before being placed back in operation
j
to verify that both isolation valves would close as required on a high differ-
ential flow signal (Item No. 87-31-07.a, Potential Enforcement Finding).
,
This post-maintenance testing oversight came to light un January 5, 1985,
l
and on January 10, 1985, when the running RWCU pump IG31-C0018 isolation
valve 1G31-F001 again failed to close upon receipt of a high differential
flow signal.
This condition was identified in LER 85-001. A review of the
instrumentation logic revealed that pump start permissive signals from relays
IG31-R616A and 8 were required for the valves to isolate properly; therefore,
if either one or both pumps were not running, the associated valve with the
i
idle pump would not isolate on high differential flow.
The licensee then
evaluated this condition as a design flaw in the instrumentation logic. As a
1
9
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. - _ - _ .
,
- .;
..
.
.
'
'
result, DCR 85-007 was generated to allow both RWCU system isolation valves
1G31-F001 and F004 to close on a high differential flow signal, with either
of the two RWCU pumps running, as required per the Technical Specifications 3.7,0.1 and Table 3.7-1.
DCR 85-007 also changed a 15-second time delay relay
to a 45-second time delay relay to mitigate the effect of pressure spikes in
the flow transmitter sensing lines.
DCR 85-007 was reviewed by the NRC inspectors, and several apparent discrepan-
cies were identified. One of these was in the area of safety evaluations.
10 CFR 50.59 requires the licensee to develop and retain a written safety
evaluation that demonstrates the baser, for the determination that the change,
test, or experiment does not involve an unreviewed safety question.
Contrary
to the 10 CFR 50.59 requirement, the safety evaluation for DCR 85-007
Revision 1 did not adequately detail the bases for the determination that the
addition of a 45-second delay relay to the RWCU flow logic was not an
unraviewed safety quefition.
The detemination did not consider the original
design basis for the actuation or various failure modes, nor did it document
any design-basis accidents that were reviewed for impact, or other systems
and components that could have been affected by the change (Item No.87-31-08,
~ Potential Enforcement Finding).
Post-modificat. ion functional testing was reviewed to veriff operability
following the modification by DCR 85-007.
ANSI 18.7, Section 5.2.7 states
that a suitable level of confidence in structures, Systems, or components
on which maintenance or modifications have been performed should be obtained
by appropriate inspecticn and performance testing. Maintenance work order
(MW0) 1-85-426 and surveillance procedure HNP-1-5261 indicate that the time
delay relays were bench tested before they were installed.
The installation
post-modification testing perfomed under NW01-85-401 did not test the actual
installation of the 45-second delay timer. Based on their interpretation of
ANSI 18.7, the NRC inspectors do not consider the bench test so be adequate
87-31-07.b, Potential
performance testing)for this installation (Item No.
Enforcement Finding .
Unit 1 and Unit 2 Technical Specifications were reviewed to verify tL impact
of the 45-second delay timer in the containment isolation actuation circuitry.
The Unit 2 Technical Specifications assumed the isolation response time to be
43 seconds.
The addition of the 45-second delay timer appeared to be in
conflict with Technical Specifications.
In a letter dated December 22, 1987,
the licensee requested tempor6ry relief from TS 3/4.3.3 for the RWCU values
until the TS could be changed.
Region II granted this relief on December 24,
1987, with concurrence of NRR (Item No. 87-31-09, Potential Enforcement
Finding).
3.5 Field Verification of Maintenance Work Orders
The NRC inspectors conducted field verification of closed MWO 1-87-01780.
This MWO and its associated work process sheet (WPS) provided gt:idance for
the reinforcement and installation of pipe supports in the high pressure
coolant injection (HPCI) room, as required by NRC Bulletin 79-14. During the
walkdown, the NRC inspectors discussed the MWG package with the licensee's
j
cognizant mechanical QC inspector.
Based on those discussions, the NRC
inspectors concluded that QC verification activities associateo with MWO
,
1-87-01780 were detailed and thorough.
<
10
_
.
.
_
.
_
_- _
a
.
.
'
On three occasions, the NRC inspectors observed part of the licensee's
activities associated with WO 1-87-S632.
The WO covered the replacement of
six pneumatic seals in the fuel transfer canal anc a minor modification of
the seal leak detection system under DCR 87-099 and FCR 87-099-01.
The W0
appeared to be the result of thorough pre-planning.
The personnel performing
,
the seal removal and replacement were knowledgeable about the work and their
responsibilities, and there was considerable management review of the actual
work.
Extensive precautions were taken to avoid radiation exposure and con-
,
tamin6ticn.
3.6 K'lity Assurance Audits of the DCR Program
The NRC inspectors reviewed reports of two audits of the DCR system performed
by the licensee.
The 1986 audit, "Quality Assurance Audit of Design Change
and Modification Control (86-DCR-1)," and the 1987 audit, "Design Change
Request Program (87-DCR-1C)," were reviewed to determine the scope and depth
to which the licensee evaluates the design change process.
Additionally,
these audits were reviewed to detennine whether the audit findings were tech-
nical or administrative.
The 1986 audit addressed the general QA planning matrix areas of administra-
tion, implementation, testing / inspection, and system restoration.
The audit
report identified numerous problems with the design change process and modi-
fication control.
The audit findings and areas evaluated during the audit
encompassed many aspects of the design programs, and the findings identified
technical deficiencies with decign changes and spe'cific technical inadequacies.
The 1987 audit examined the DCR program for compliance with ANSI N18.7 and
site procedures.
This audit reviewed the processing of electrical and
mechanical DCRs and was coordinated with both Plant Yogtle QA and Southern
Company Services QA.
The primary weakness identified during the audit dealt
with an inattention to detail during the engineering review process, which
allowed minor errors in DCR packages to go undetected.
This audit appeared
to be compliance based and was not nearly as technically in depth as the 1986
audit.
3.7 Summary
The DCRs and associated W0s reviewed were generally adequate, although some
open DCRs appear to require further evaluation.
Although discrepancies were
'
identified in the activities associated with OCR 85-007, and these discrepan-
cies had not been previously iden'ified by tha licensee's quality verifica-
'
tion organizations, the NRC inspec. ors do not believe that this oversight is
representative of the present ogratizations. The direct observations and field
verification of WO activities indicated that licensee personnel were
knowledgeable and professional, and quality verification activities were
performed effectively.
The NRC inspectors found, based on the sample of hardware and documents
reviewed during this inspection, thet the present activities of the licensee's
quality serification organizations in the area of modification of plant
systems and components are generally adequate. Quality verification
activities in the design change control area were evaluated as effective.
11
.
. - . - - - ..
-
. -
-
-
_ _ _ _ __. _ _ _ _
' ;
..
.
.
.
'
'
Good attitudes toward quality, high knowledge levels, and an istproving trend
are evident.
4.0 EXIT INTERVIEW
The inspectors met with the licensee's representatives (included in the list
in Appendix A) on December 11, 1987.
The purpose, scope, and results of the
inspection were discussed.
..
.
%
0
0
4
6
,
l
t
!
l
!
l
12
!
I
1.
- _ _ _ _
.t
.*
.
.
.
.
-
.
APPENDIX A
Persons Contacted
Georoia Power Company (GPC)
- J. T. Beckham, Jr.,*Vice President, Plant Hatch
- S. J. Bethay, Nuclear Safety and Compliance (NASC) Supervisor
- J. K. Branum, Sr. Nuclear Engineer
E. Burkett - Supervisor, General Engineering
- C. L. Coggin, Manager, Training and Emergency Preparedness
R. L. Colson, QC Electrical Inspector
- G. M. Creighton, Regulatorg Specialist (NASC)
C. Dixon, QA Engineering Support ' Supervisor
- P. E. Fornel, Manager Maintenance
- 0. M. Fraser, Site QA Manager
- M. H. Googe, Manager, Outages and Planning
G. A. Goode, General Engineering Superintendent
'
- G. R. Goodman, Independent Safety Evaluation Group (ISEG)
J. Harrinonds, ISEG Supervisor
W. Hayden, Health Physics and Chemistry
R. L. Hayes, Deputy Operations Manager
- J. D. Heidt, Nuclear Licensing Manager - Hatch
F. A. Herrington - Senior Regulatory Specialist
R. L. Keck - Superintendent, Recctor Systems Engineering
- C
L. McDaniel, Acting Manager, General Support
C. Melchoir - System Engjneer
- C, T. Moore, General Manager, QA
- J. E. Newton, Maintenance Planning Supe. visor
- H. C. Nix, Plant Manager - Hatch
..
- T. R. Powers, Manager, Engineering Support
J. Robertsun - Supervisor, Reactor Systems Engineering
i
!
J. Shuman - Supervisor, Reactor Systems Engineering
l
D. Smith, Health Physics and Chemistry (HP/C) Supervisor
- H. L. Sumner, Manager of Operations
l
W. B. Thigpen, QA
- S. B. Tipps, Manager, Nuclear Safety & Compliance
l
i
R. Tracy - System Engineer
!
C. A. Tyre, Shif t Supervisor
E. 2. Wahab - Superintendent, Balance of Plant Engineering
!
A. Wheeler - Supervisor, Balance of Plant Engineering
D. Williams - System Engineer
i
- R. W. Zavadoski, Manager, Health Physics and Chemistry
l
l
l
- Denotes those .; tending the exit meeting on December ll,1987.
l
l
l
l
l
!
l
l
_
_
, - . _ . . __.,
.
,_
_ _ _ _ _
_ _ _ . _ . _
__ _
_ _ _ . _
_ . _ ,
_ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _
.
' ) .*
.
.
- .
-
.
NRC
- G. A. Belisle, NRC, Section Chief, QA, RII
- L. P. Crocker, Hatch Project Manager, NRR
- F. C. Hawkins Quality Operations Section Chief, NRR
- G. W. Lapinsky, NRR-
.
- J. E. Menning, Resident Inspector
- M.
V. Sinkule, Section Chief, RI!
.
OTHERS
'
,
- D. Dismukes, Ber.htel, Mecheical Supervisor
J. N. Keller, Bechtet, Engineering Supervisor
- G. D. McGaha, Southern Compa'ny Services (SCS), Design Project Manager
- R. W. Montross, Oglethorpe, Site Representative
, O. Prescott, SCS, Technical Aide
.
Other licensee employees contacted included operators, engineers, auditors,
technicians, mechanics, and office personnel.
.
l
Denotes those attending the exit meeting on December 11, 1987.
'
l
!
I
-
- - -
-
-
-