IR 05000155/1986001
| ML20211G473 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/17/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20211G425 | List: |
| References | |
| 50-155-86-01, 50-155-86-1, NUDOCS 8702250397 | |
| Download: ML20211G473 (13) | |
Text
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SALP 6
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APPENDIX
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SALP BOARD REPORT.
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U.S. NUCLEAR REGULATORY COPWISSION
REGION III
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SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE
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50-155/86001 l
Inspection Report
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Consumers Power Company
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Name of Licensee
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Bic Rock Point Plant
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Fame of Facility
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November 1, 1984 through March 31, 1986
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Assessment Period
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8702250397 870217
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ADOCK 05000155
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Big Rock Point Plant
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Meeting Summary
The findings and conclusions of the SALP Board are documented in Inspection
Report No. 50-155/86001.
They were discussed with the licensee on July 21,
1986, at the Region III office in Glen Ellyn, Illinois.
The licensee's
regulatory performance was presented in each functional area.
Overall
performance and performance in each functional area was found to be acceptable.
The performance rating improved in the area of Licensing Activities.
Performance in the areas of Plant Operations and Surveillance and Inservice
Testing declined based on increased frequency of personnel error and problems
encountered in implementing administrative controls for the Plant Operations
area, and missed surveillances for the Surveillance and Inservice Testing area.
While the performance rating in the new area of Outages was given a Category 3
based on the Severity Level III violation received during the middle of the
SALP period, there has been no opportunity to evaluate the effectiveness of
your corrective measures.
The Emergency Preparedness and Security areas
continued to have a high level of performance.
While this meeting was primarily a discussion between the licensee and the
NRC, it was open to members of the public as observers.
The following licensee and NRC personnel were in attendance on July 21, 1986.
Consumers Power
J. Reynolds, Executive Vice President
F. W. Buckman, Vice President, Nuclear Operations
G. B. Slade, Executive Director, Nuclear Assurance
K. W. Berry, Director, Nuclear Licensing
D. Hoffman, Plant Superintendent
R. R. Frisch, Senior Licensing Analyst
T. C. Bordine, Staff Engineer
B. Alexander, Technical Engineer
U.S. Nuclear Regulatory Commission
A. B. Davis, Deputy Regional Administratcr
E. G. Greenman, Deputy Director, Division of Reactor Projects
D. C. Boyd, Chief, Reactor Projects Section 20
S. Guthrie, Senior Resident Inspector, Big Rock Point
R. B. Landsman, Project Manager, Section 2D
J. Bauer, Technical Staff
NRC Headquarters
T. S. Rotella, NRR Project Manager
J. A. Zwolinski, Director, BWR Project Directorate No.1
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ERRATA SHEET
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Facility: Big Rock Point
SALP Report No.
50-155/86001
Page
Line
Now Reads
Should Read
39-43
The licensee... to
Delete
shutdown.
Basis for Change: Additional information provided by licensee subsequent to
SALP issuance.
45
no training was provided with the exception of
until February 1986,
some I&C classes and
certain skill training,
no training was provided
during this SALP period
until February 1986
Basis for C(a ge:
The phrase incorrectly implied that training was never
provided and should have stated only that during most of
the SALP it wasn't.
26-27
As of this date only
Delete
'
15 have been checked
Basis for Change: Additional information provided by licensee subsequent to
SALP issuance.
38-39
Forced retirement of
Untimely retirement of
i
several older key
several older key members
members of the licensee
of the licensee staff,
staff
which was honored by
licensee management
Basis for Change: The phrase was not meant to mean that the employees were
forced out, only that they were encouraged.
,
I
26
It is noted however that
,
some relief in the form
-
of additional QA personnel
i
from the Palisades plant
'
.
was provided in September
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.
1985.
t
B4 sis for Change: Additianal information provided by licensee subsequent to
,
i
SALP issuance.
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C.
Maintenance / Modifications
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1.
Analysis
Portions of eight routine inspections by the Resident In ector
reviewed maintenance activities. One violation discuss d in
Section IV.H Outages, reflects on the licensee's abi ty to
conduct maintenance work during outages.
In additio, two
Regionally based inspections were performed.
The i spections
included reviews of normal maintenance and modiff
tion activities
j
to ensure that approvals were obtained prior to
itiating work,
activities were accomplished using approved to edures, post
maintenance testing was completed prior t
e rning components
or systems to service, and parts and ma
were properly
certified.
In addition, work plannin
cheduling was
reviewed as well as the effectiveness
ministrative controls
toensureproperpriorityisassigneg
violations or
daviations noted.
During the evaluation period the 44,
see interrupted plant
!
operations for nine unscheduled
tenance outage periods
ranging from one to 11 days,
outages were required to
repair Reactor Depressurizat
stem (RDS) valves due to the
!
degraded condition of the
preventing successful performance
of quarterly surveillances.
ese included one forced shutdown
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i
required by Technical
ecif cations unidentified leak rate
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limitations. Two ou
(p iods of one day each were required
to successfully repa
K DB, seal leakage to heat exchanger
for Reactor Recir 1
Pump No. 2.
Also, two outages of three
.
and four days eac
required to diagnose and correct steam
leakage from the reac r vessel head o-rings. One outage period
of four days was us
to replace a recirculation pump seal, and
a one day outage w
required to correct steam leaks associated
i
with the plant sc m un December 7,1985.
Proper plannin and outage control was generally evident for the
.
nine unschedu ed outages. Although unplanned, the licensee in
the case of he RDS and recirculation pump outages had sufficient
'
warning to lan activities, prepare parts and procedures, and
perform o er maintenance work that fell within the scope and
time li tations of the forced outage.
Repair to RDS valve top
!
assemb es have become commonplace to the point that the licensee
'
routi ely overhauls spare top assemblies.
The licensee did not
ove aul the spare recirculation pump seal in advance of the
l
ou ge and was still rebuilding the seal as the plant was being
.
utdown to perform the replacement, even though the pump had been
died for two weeks prior to shutdown.
The licensee made
extensive use of vendor consultants and pump experts from the
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C.
Maintenance / Modifications
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1.
Analysis
Portions of eight routine inspections by the Resident Inspector
reviewed maintenance activities. One violation discussed in
Section IV.H. Outages, reflects on the licensee's ability to
conduct maintenance work during outages.
In additional, two
Regionally based inspections were performed. The inspections
included reviews of normal maintenance and modification
activities to ensure that approvals were obtained prior to
initiating work, activities were accomplished using approved
procedures, post maintenance testing was completed prior to
returning components or systems to service, and parts and
materials were properly certified.
In addition, work planning
and scheduling was reviewed as well as the effectiveness of
administrative controls to ensure proper priority is assigned.
No violations or deviations noted.
During the evaluation period, the licensee interrupted plant
operations for nine unscheduled maintenance outage periods
ranging from one to 11 days. Three outages were required to
repair Reactor Depressurization System (RDS) valves due to the
degraded condition of the system preventing successful
performance of quarterly surveillances. These included one
forced shutdown required by Technical Specifications
unidentified leak rate limitations. Two outage period of one
day each were required to successfully repair IA-60B, seal
leakage to heat exchanger for Reactor Recirculation Pump
No. 2.
Also, two outages of three and four days each were
required to diagnose the correct steam leakage from the reactor
vessel head o-rings. One outage period of four days was used
to replace a recirculation pump seal, and a one day outage was
required to correct steam leaks associated with the plant scram
on December 7, 1985.
Proper planning and outage control was generally evident for
the nine unscheduled outages. Although unplanned, the licensee
in the case of the RDS and recirculation pump outages had
sufficient warning to plan activities, prepare parts and
procedures, and perform other maintenance work that fell within
the scope and time limitations of the forced outage.
Repair to
RDS valve top assemblies have become commonplace to the point
that the licensee routinely overhauls spare top assemblies.
The licensee made extensive use of vendor consultants and pump
experts from the General Office for the seal replacement,
resulting in a refined and useful procedure for rebuilding and
installation. Outages for RDS and recirculation pump repairs
were well planned and executed. Outages to repair IA-60B
represented an operational situation that offered little
warning and first attempts at repairs were unsuccessful. The
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General Office for the seal replacement, resulting in a refined
and useful procedure for rebuilding and installation. Outages
for RDS and recirculation pump repairs were well planned and
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executed. Outages to repair IA-60B represented an operationa
situation that offered little warning and first attempts at
repairs were unsuccessful. The reactor vessel o-ring offer d
no warning prior to failure, but successful repairs were. layed
when the problem was misdiagnosed. Once the decision wa made
to perform the vessel head removal and ring replacemen the
physically demanding job was successfully completed w' h
conservative consideration to ALARA and personne sa ety.
Maintenance work (including mechanical, electr
, and
<
instrument / control) at Big Rock Point is per
by generally
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competent repairmen who exhibit craftsmans
d a general
familiarity with the facility and the equip
The amount
.
of unsuccessful repair attempts resulting W ework is generally
small.
Repairmen generally are cognizant o procedural require-
ments associated with their assigned
, communicate effec-
tively with operators and health ph
echnicians, and reflect
concern for ALARA considerations.
the input repairmen
provide to machinery history is o
with co-workers and supervisors ' g.arginal, communication
N
tes genuine interest in
continued safe and successful
on of the reactor. The
mechanic who performs the work jo example, often participates
in post maintenance testin. W
e the retirement of older,
experienced maintenance de
tm t personnel during the period
had a negative impact on
,ance as documented further
in Section IV.H, Outage
e naintenance staff demonstrated
flexibility and dedica i
roughout the evaluation period.
The size of the maintenon
staff is generally adequate for all
periods other than major
> fueling outages. A gradually
increasing backlog of ma ntenance orders over the period is
explained in part by i
reased emphasis on skills training which
over the short term r uces staff size availability.
Like the Operations Department the loss of older experienced
l
personnel due to etirement or other duties has altered composi-
l
tion of the main enance staff. While the I & C group remained
unchanged, in
e mechanical maintenance group of 12 men, five
'
were added dur ng the assessment period. Because hiring and
promotion is eavily influenced by Labor Relations agreements
that emphas'ze seniority, newly added staff members generally
l
have litt
or no experience with nuclear powered generating
plants i general or Big Rock Point specifically. Although the
I
license has long recognized the need for maintenance staff
l
traini g, no training was provided until February 1986, when a
i
regu'
r program of skills training offsite was initiated. The
skiy s training is general in nature and is not nuclear plant
splcific. No nuclear plant system or concepts training is
provided.
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reactor vessel o-ring offered no warning prior to failure, but
successful repairs were delayed when the problem was misdiagnosed.
Once the decision was made to perform the vessel head removal and
-
'
ring replacement the physically demanding job was successfully
completed with conservative consideration to ALARA and
personnel safety.
Maintenance work (including mechanical, electrical, and
instrument / control) at Big Rock Point is performed by generally
competent repairmen who exhibit craftsmanship and a general
familiarity with the facility and the equipment.
The amount
of unsuccessful repair attempts resulting in rework is generally
small.
Repairmen generally are cognizant of procedural require-
ments associated with their assigned task, communicate effec-
tively with operators and health physics technicians, and reflect
concern for ALARA considerations. While the input repairmen
provide to machinery history is often marginal, communication
with co-workers and supervisors indicates genuine interest in
continued safe and successful operation of the reactor. The
mechanic who performs the work, for example, often participates
in post maintenance tasting. While the retirement of older,
experienced maintenance department personnel during the period
had a negative impact on performance as documented further
in Section IV.H, Outages, the maintenance staff demonstrated
flexibility and dedication throughout the evaluation period.
The size of the maintenance staff is generally adequate for all
periods other than major refueling outages. A gradually
increasing backlog of maintenance orders over the period is
explained in part by increased emphasis on skills training which
over the short term reduces staff size availability.
Like the Operations Department the loss of older experienced
personnel due to retirement or other duties has altered composi-
tion of the maintenance staff. While the I & C group remained
unchanced, in the mechanical maintenance group of 12 men, five
were added during the assessment period. Because hiring and
promotion is heavily influenced by Labor Relations agreements
that emphasize seniority, newly added staff members generally
have little or no experience with nuclear powered generating
plants in general or Big Rock Point specifically. Although the
licensee has long recognized the need for maintenance staff
training, with the exception of some I&C classes and certain
skill training, no training was provided during this SALP
period until February 1986, when a regular program of skills
training offsite was initiated. The skills training is general
in nature and is not nuclear plant specific.
No nuclear plant
system or concepts training is provided.
.
First line supervision in the maintenance department reflects
adequate technical skills and managerial competence. During the
1985 outage, the maintenance department overcame the loss of
staff experience, inadequate outage planning, and parts proc
e-
ment to accomplish a relatively large number of modificati
s,
repairs, and preventive maintenance tasks.
Throughout the evaluation period several recurring pro ems were
not successfully repaired or adequately addressed. V ve M0-7067,
Turbine Bypass Isolation Valve, was not declare op able for
much of the evaluation period, based on diffic ti s with the
valve operator. Reactor Depressurization Sy
m RDS) valves
exhibit inherent design deficiencies that
sulted in three
forced shutdowns during the assessment p
1d a long history
of problems dating back to their instal
Management,
.
however, has not placed a high priority
comprehensive
solution and as a result the RDS system w
not improved over
the period.
Problems with the Emer M y iesel Generator (EDG)
fuel pump were allowed to continue nu
design change to the
pump mounting bracket scheduled
pletion during the 1985
refueling outage was deleted i
fort to return the plant to
an operable status. Shortly
ter the pump failed again,
placing the EDG in an actio
ment for the generator's
Limiting Condition for Oped]
Finally, the licensee made a
.
commitment to verify, prior
startup from the 1985 outage,
Limitorque Switch sett
so 18 Limitorque Valves the licensee
considered important
ty. As of this date only 15 have
been checked. The t
.ettings for valve M0-7067 have been
reset on three di
occasions, indicating a lack of decisive
direction on probl
th Limitorques Operators that goes back
to September, 1984, a was addressed in SALP 5.
SALP 5 expressed c cern that the Prevention Maintenance (PM)
program may be in dequate to address aging equipment. At the end
of this assessm t period the PM program continues to be reactive
in nature, rel
ng heavily on visual inspections that do not
involve disa embly or physical measurements, and on the obser-
vations of
erators monitoring noticeable changes in component
operating
aracteristics. There continues to be no program to
analyze
r trends in failures or any other measurable parameter
other
an pump capacity on certain pumps. The licensee has not
respo ed to NRC initiatives to upgrade the PM program to incor-
pora e vendor recommendations and industry experience. The plant
co
inues to rely on surveillance tests to identify problems that
m
be in some advanced stage of development due to aging
quipment. At the close of the assessment period the licensee
assigned an engineer to develop a program of predictive analysis
focusing on vibration and lubricating oil analysis.
Evidence of
problems associated with aging of plant equipment during the
assessment period included:
a.
Several examples of end of service life for solenoid valves
on the turbine stop valve, diesel fire pump (DFP), and the
exhaust ventilation downstream isolation valve.
.
First line supervision in the maintenance department reflects
adequate technical skills and managerial competence. During the
1985 outage, the maintenance department overcame the loss of
staff experience, inadequate outage planning, and parts procure-
ment to accomplish a relatively large number of modifications,
repairs, and preventive maintenance tasks.
Throughout the evaluation period several recurring problems were
not successfully repaired or adequately addressed. Valve M0-7067,
Turbine Bypass Isolation Valve, was not declared operable for
much of the evaluation period, based on difficulties with the
valve operator. Reactor Depressurization System (RDS) valves
exhibit inherent design deficiencies that have resulted in three
forced shutdowns during the assessment period and a long history
of problems dating back to their installation. Management,
however, has not placed a high priority on a comprehensive
solution and as a result the RDS system was not improved over
the period. Problems with the Emergency Diesel Generator (EDG)
fuel pump were allowed to continue and a design change to the
pump mounting bracket scheduled for completion during the 1985
refueling outage was, deleted in an effort to return the plant to
an operable status.
Shortly thereafter the pump failed again,
placing the EDG in an action statement for the generator's
Limiting Condition for Operation.
Finally, the licensee made a
commitment to verify, prior to startup from the 1985 outage,
Limitorque Switch settings on 18 Limitorque Valves the licensee
considered important to safety. The torque settings for valve
M0-7067 have been reset on three different occasions, indicating
a lack of decisive direction on problems with Limitorques
Operators that goes back to September,1984, as was addressed
in SALP 5.
SALP 5 expressed concern that the Prevention Maintenance (PM)
program may be inadequate to address aging equipment. At the end
of this assessment period the PM program continues to be reactive
in nature, relying heavily on visual inspections that do not
involve disassembly or physical measurements, and on the obser-
vations of operators monitoring noticeable changes in component
operating characteristics. There continues to be no program to
analyze for trends in failures or any other measurable parameter
other than pump capacity on certain pumps. The licensee has not
responded to NRC initiatives to upgrade the PM program to incor-
porate vendor recomendations and industry experience. The plant
continues to rely on surveillance tests to identify problems that
may be in some advanced stage of development due to aging
equipment. At the close of the assessment period the licensee
assigned an engineer to develop a program of predictive analysis
focusing on vibration and lubricating oil analysis. Evidence of
problems associated with aging of plant equipment during the
assessment period included:
a.
Several examples of end of service life for solenoid valves
on the turbine stop valve, diesel fire pump (DFP), and the
exhaust ventilation downstream isolation valve.
.
Operations Department personnel performed fuel handling
operations for the 1985 refueling outage.
Fuel handling was
safely conducted by adequately trained individuals in accordance
-
with approved procedural requirements.
Staffing on both the
reactor deck and in the control room was adequate, and communi-
cation between the two areas was effective.
Management involve-
ment in refueling activities was evident.
Tool control and
status board mainteaance was adequate.
Licensee responsive ss
to NRC initiative was evident by their prompt action to co rect
procedural deficiencies in data recording and in relocat' n of
bagged equipment that had obstructed access to the refu
ling
deck status board.
During the 1935 refueling outage several incid
s ccurred
which demonstrated inadequate management con
er the outage
process.
The incidents involved:
Repeated examples of contractors and
see travel crew
personnel, not normally assigned to B
Rock Point,
performing work on the wrong com nen or system, pointing
to inadequate control over the
ties of travel crews
and contractors.
Repeated examples of superv
, maintenance, operations,
and engineering personnel p@a19d ravel crew personnel,
circumventing or failing
t
ere to administrative
~
requirements, particula)
ose related to component
tagging and isolation.
Repeated examples b
'n ividuals, throughout the
organization, o
ntion to detail and failure to
exercise suffic
re in performance of outage related
work to ensure p
safety.
Several factors contr' uted to the breakdown in the outage
management process:
Throughout
e facility, components, valves, and systems
identifica on was generally inadequate, with many compo-
nents un1
eled.
The licensee had not acted upon earlier
request from the Resident Inspector to improve component
identi ication and discounted warnings on the potential for
mish-s.
F ced retirement of several older key members of the
icensee staff, including the Operations Superintendent,
the coordinator of the ISI program, an experienced Shift
Supervisor, and a Maintenance Supervisor who in the past
had acted as a coordinator and single contact point for
control of travel crew personnel.
The impact of the loss
of these individuals two months prior to commencement of
[
.
Operations Department personnel performed fuel handling
operations for the 1985 refueling outage.
Fuel handling was
'
safely conducted by adequately trained individuals in accordance
with approved procedural requirements. Staffing on both the
reactor deck and in the control room was adequate, and communi-
cation between the two areas was effective. Management involve-
ment in refueling activities was evident. Tool control and
status board maintenance was adequate. Licensee responsiveness
to NRC initiative was evident by their prompt action to correct
procedural deficiencies in data recording and in relocation of
bagged equipment that had obstructed access to the refueling
deck status board.
During the 1985 refueling outage several incidents occurred
which demonstrated inadequate management control over the outage
process. The incidents involved:
Repeated examples of contractors and licensee travel crew
personnel, not normally assigned to Big Rock Point,
performing work on the wrong component or system, pointing
to inadequate control over the activities of travel crews
and contractors.
Repeated examples of supervisors, maintenance, operations,
and engineering personnel, and travel crew personnel,
circumventing or failing to adhere to administrative
requirements, particularly those related to component
tagging and isolation.
Repeated examples by individuals, throughout the
organization, of inattention to detail and failure to
exercise sufficient care in performance of outage related
work to ensure plant safety.
Several factors contributed to the breakdown in the outage
management process:
'
Throughout the facility, components, valves, and systems
identification was generally inadequate, with many compo-
nents unlabeled. The licensee had not acted upon earlier
requests from the Resident Inspector to improve component
identification and discounted warnings on the potential for
mishaps.
Untimely retirement of several older key members of the
licensee staff, which was honored by licensee management,
including the Operations Superintendent, the coordinator
of the ISI program, an experienced Shift Supervisor, and
a Maintenance Supervisor who in the past had acted as
a coordinator and single contact point for control of
travel crew personnel. The impact of the loss of these
individuals two months prior to commencement of
F
.
During the evaluation period there was evidence that the site
staff was in danger of becoming overburdened by assignment of
-
several functions formerly performed by the corporate QA grc p.
Those added duties were subsequently completed or reassigned
elsewhere and the site staff appears adequate for the rem (ning
workload.
The site QA staff communicates effectively wi
i plant
management and is persistent in pressing for managemen action
,
to resolve audit findings. The Plant Review Committe (PRC)
'
considers the quality aspects of technical and safet issues.
In turn, plant management generally demonstrate th r regard
for the significance of findings and comments
the QA staff.
Site QC inspectors are generally thorougt, and
cientious and
draw heavily on their plant experience. Botgj
staff are responsive to NRC initiatives a W iries.
Licensee corporate management detracted
the effectiveness of
Programs and Administrative controls affe
ng quality. Examples
include:
a.
Licensee corporate management,
ransferring to the site
staff several significant Qu
Assurance functions with-
out a corresponding increas
vailable site resources,
placed a burden on the sta
ch resulted in QA reviews
thatwerelesscomprehensQa.,
ithdrawal of commitments to
..
support audit activitie
ff site, and a virtual elimination
of time available to au
s to review and observe activi-
ties in the plant.
Some
functions were performed by QC
inspectors.
The rel ta e of corporate management to
respond to the co
of the site QA Superintendent in
this regard and t
oor response to NRC initiatives to
address the iss
noted.
b.
Licensee Corporate anagement deleted entirely fifteen NODS,
the document in w ich the licensee staff can theoretically
be assured of fi ding all applicable code and regulatory
requirements c piled in one location.
The NODS are the
means by whic the licensee's Quality Assurance Program
Description or Operational Nuclear Power Plants (Topical
Report CPC-A) is implemented, and results from a commitment
l
made in t
licensee's Regulatory Performance Improvement
Program
bmitted in response to a March 9, 1981 Confirmatory
Order.
holesale deletion of the NODS without a review to
insur all of the quality requirements contained therein
'
were 1 ready addressed in existing administrative procedures
,
j
res ted in a period when the quality requirements were not
av lable to the NODS user.
Inspectors identified at least
o examples of cancelled NODS being referenced in other
,
rocedures.
.
.
.
..
e
.
During the evaluation period there was evidence that the site
QA staff was in danger in becoming overburden by assignment of
several functions formerly performed by the corporate QA group.
Those added duties were subsequently completed or reassigned
elsewhere and the site staff appears adequate for the remaining
workload. The site QA staff communicates effectively with plant
management and is persistent in pressing for management action
to resolve audit findings. The Plant Review Committee (PRC)
considers the quality aspects of technical and safety issues.
In turn, plant management generally demonstrates their regard
.
for significance of findings and coninents from the QA staff.
Site QC inspecto s are generally thorough and conscientious and
draw heavily on their plant experience. Both the QA and QC site
staff are respontible to NRC initiatives and inquiries.
Licensee corporate management detracted from the effectiveness
of Programs and Administrative controls affecting quality.
Examples include:
a.
Licensee corporate management, by transferring to the
site staff several significant Quality Assurance functions
without a correspor. ding increase in available site resources,
placed a burden on the staff which resulted in QA reviews
that were less comprehensive, withdrawal of commitments to
support audit activities off site, and a virtual elimination
of time available to auditors to review and observe
activities in the plant. Some QA functions were performed
by QC inspectors.
It is noted however that some relief in
the form of additional QA personnel from the Palisades
plant was provided in September 1985. The reluctance of
corporate management to respond to the concerns of the
site QA Superintendent in this regard and their poor
response to NRC initiatives to address the issue was noted,
b.
Licensee Corporate management deleted entirely fifteen N0DS,
the document in which the licensee staff can theoretically
be assured of finding all applicable code and regulatory
requirements complied in one location.
The N0DS are the
means by which the licensee's Quality Assurance Program
Description for Operational Nuclear Power Plants (Topical
Report CPC-2A) is implemented, and results from a commitment
made in the licensee's Regulatory Performance Improvement
Prograa submitted in response to a March 9, 1981
Confirmatory Order. Wholesale deletion of the N0DS without
a review to insure all of the quality requirements contained
therein were already addressed in existing administrative
procedures resulted in a period when the quality
requirements were not available to the N0DS user.
Inspectors identified at least two examples of
cancelled N0DS referenced in other procedures.
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