HomeMy WebLinkAbout851154.tiff ``tot A ECy9 UNITED STATES
0'= NUCLEAR REGULATORY COMMISSION
REGION IV
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611 RYAN PLAZA DRIVE, SUITE 1000
2,7 �a°y ARLINGTON,TEXAS 76011
MAY 3 7 1985
In Reply Refer To: Fln r^ TY r.^:MtdlSSIONfPS
Docket: 50-267/85-06 _
MAY 91985
Public Service Company of Colorado 111ATTN: 0. R. Lee, Vice President
Electric Production , « '.. COW.
P. O. Box 840
Denver, Colorado 80201
Gentlemen:
This refers to the inspection conducted by Mr. M. E. Skow of this office during
the period March 5-8, 1985, of activities authorized by NRC Operating License
DPR-34 for Fort St. Vrain Nuclear Station, and to the discussion of our
findings with Mr. Singleton and other members of your staff at the conclusion
of the inspection.
Areas examined during the inspection included receiving and maintenance with
specific focus on the control rod refurbishment. The inspection in these areas
is a continuation from NRC Inspection Report 50-267/85-01. Within these areas,
the inspection consisted of selective examination of procedures and
representative records, interviews with personnel , and observations by the
inspector. These findings are documented in the enclosed inspection report.
During this inspection, it was found that certain of your activities were in
violation of NRC requirements. Consequently, you are required to respond to
these violations, in .writing, in accordance with the provisions of
Section 2.201 of the NRC' s "Rules of Practice," Part 2, Title 10, Code of
Federal Regulations. Your response should be based on the specifics contained
in the Notice of Violation enclosed with this letter.
Should you have any questions concerning this inspection, we will be pleased to
discuss them with you.
Sincerely,
E. H. Johnson, Chief
Reactor Project Branch 1
Enclosures:
1. Appendix A - Notice of Violation
2. Appendix B - NRC Inspection Report
50-267/85-06
cc w/enclosures: (cont. on next page)
851154
Public Service Company of Colorado -2-
Mr. D. W. Warembourg, Manager
Nuclear Engineering Division
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
Mr. David Alberstein, 14/159A
GA Technologies, Inc.
P. 0. Box 85608
San Diego, California 92138
Kelley, Stansfield & O'Donnell
Public Service Company Building
550 15th Street, Room 900
Denver, Colorado 80202
Chairman, Board of County Comm.
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
Denver, Colorado 80203
Mr. H. L. Brey, Manager
Nuclear Licensing/Fuels Div.
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
J. W. Gahm, Manager, Nuclear
Production Division
Fort St. Vrain Nuclear Station
16805 WCR 19}
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
Colorado State Department of Health
APPENDIX A
NOTICE OF VIOLATION
Public Service Company of Colorado Docket: 50-267
Fort St. Vrain Nuclear Station License: DPR-34
Based on the results of an NRC inspection conducted during the period of
March 5-8, 1985, and in accordance with the NRC Enforcement Policy
(10 CFR Part 2, Appendix C) , 49 FR 8583, dated March 8, 1984, the following
violation was identified:
10 CFR 50, Appendix B, Criterion XVI requires that conditions adverse to
quality such as deficiencies, deviations, and nonconformances are promptly
identified and corrected. The FSAR section B.5 implements the Appendix B
requirements and delegates QA program definition and implementation to
the Manager, Quality Assurance.
Contrary to the above, timely corrective action was not taken to promptly
revise administrative procedures Q-15 and Q-16 as noted in CAR 84-093.
This is a Severity Level IV Violation. (Supplement I) (50-267/8506-01)
Pursuant to the provisions of 10 CFR 2.201, Public Service Company of Colorado
is hereby required to submit to this office, within 30 days of the date of this
Notice, a written statement or explanation in reply, including: (1) the
corrective steps which have been taken and the results achieved; (2) corrective
steps which will be taken to avoid further violations; and (3) the date when
full compliance will be achieved. Consideration may be given to extending your
response time for good cause shown.
Dated: MAY 0 7 1985
APPENDIX B
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report: 50-267/85-06 License: DPR-34
Docket: 50-267
Licensee: Public Service Company of Colorado (PSC)
P. 0. Box 840
Denver, Colorado 80201
Facility Name: Fort St. Vrain Nuclear Station (FSV)
Inspection At: Platteville, Colorado
Inspection Conducted: March 5-8, 1985
Inspector: Pi/ /3/scr
M. E. Skow, Reactor Inspector, Special Projects Date
and Engineering Section, Reactor Project
Branch 1
Approved: ,e. 041- 1.4111 3d 7S
Ireland, ef, Specal Projects and to
Engineering Section, Reactor Project Branch 1
Inspection Summary
Inspection Conducted March 5-8, 1985 (Report 50-267/85-06)
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Areas Inspected: Routine, unannounced inspection of receiving and maintenance
with specific emphasis on the control rod refurbishment program. The
inspection in these areas is a continuation of NRC Inspection Report
50-267/85-01. The inspection involved 33 inspector-hours onsite by one NRC
inspector.
Results: Within the two areas inspected, one violation was identified (lack
of QA independence, paragraph 2).
-3-
DETAILS
1. Persons Contacted
*C. Fuller, Station Manager
*F. Novachek, Technical/Administrative Services Manager
*L. Singleton, Manager, Quality Assurance (QA)
*F. Borst, Support Services Manager
*T. Orlin, QA Service Manager
*R. Craun, Supervisor, Nuclear Site Engineering
R. Gappa, Refurbishment Shift Floor Manager
*P. Moore, Supervisor QA Technical Support
*S. Willford, Training Supervisor
G. Redmond, MQC Supervisor
J. Jackson, QA/QC Supervisor
W. Parsons, QC Receiving Inspector
*Denotes those present at exit interview.
2. Receiving,
The NRC inspector reviewed the following documents:
Number Title Issue
MRIM-1 General Receiving Inspection 4
P-5 Material Control 8
Q-15 Control of Nonconforming Items 3
G-2 FSV Procedure Systems 15
Q-1 FSV Organization and Responsibilities 5
The majority of the CRDOA refurbishment parts that had arrived on site had
completed receiving inspection. The receiving inspection packages for
several parts were reviewed by the NRC inspector. The packages appeared
complete. The receiving inspectors had up-to-date drawings and change
notices available as well as the purchase orders.
During discussions with QA/QC personnel , the NRC inspector was informed
that corrective changes were being prepared to certain Administrative
Q procedures. Members of QA also stated that issuing changes to their
procedures took a long time. The reason given was that organizations
outside QA which must approve the procedures often delay approval action
or require revisions to the proposed changes. The approval signatures
that are required on these Q procedures are specified in G-2. For 16 of
the 18 Q procedures, approval signatures are required from the managers
of the Nuclear Production, Nuclear Engineering, and Quality Assurance
Divisions; and for 8 of those 16 procedures, the Executive Staff
Assistant (now Manager, Nuclear Licensing and Fuels) as well .
-4-
An example referred to by QA personnel was Q-15. They stated that current
attempts to revise Q-15 began in the first half of 1984. Three-major
major revisions had been drafted but had not been approved outside of QA,
as of November 1984. Two complete rewrites have taken place thus far in
1985. The latest was to have gone to Engineering the week of this
inspection. QA personnel stated that the Nuclear Engineering Division
causes most of the delays in achieving approvals.
Some documentation was found in the corrective action request files which
support the statements by QA personnel . The documents related to a change
to Q-16, "Corrective Action System." CAR 84-093 was written to correct
items identified by NFSC Audit C-84-02. The two items referenced in
CAR 84-093 are CAAR's 633 and 712. CAR 84-093 points out that "the
requests were made over 'one year' ago." CAR-84-093 explains the
underlying cause for CAAR 633's delay as "the failure to obtain timely
approval (or disapproval ) of a procedure in order to get it revised and
issued." CAAR 633 is dated June 20, 1983. The revision to Q-16 was
finally issued October 12, 1984.
CAR 84-093 stated that the resolution to CHAR 712 "was a trivial change
and was determined not to be a factor in the resolution of the CAAR. . . .
The minor change requested in CAAR 712 is currently caught up in a major
revision to Q-15. . . ." The change to Q-15 was expected to be completed
by February 1, 1985, as stated on CAR-84-093, "Action to Correct
Discrepancy."
Corrective action timeliness was previously raised as an unresolved item,
50-267/8123-01. It was discussed further in inspection report 50-267/82-03
and was left open pending further review of a revised corrective action
program. The item was closed in inspection report 50-267/82-18 after Q-16
had been revised and the number of outstanding CAARs had been reduced.
However, the examples cited above illustrate that the QA organization is
still having difficulty revising their Q procedures to resolve CARs as
required by 10 CFR 50, Appendix B.
The three managers of Nuclear Production, Nuclear Engineering, and
Nuclear Licensing and Fuels who must also approve Q procedures have each
been QA managers. The NRC inspector felt this provides strong potential
for constructive input to the QA program. However, the NRC inspector
noted that the requirement in G-2 for three or four of these managers to
all sign for approval could be cause for Q procedure revisions not being
made in a prompt, timely manner.
This is a violation. (8506-01)
3. Maintenance
The first rod, to be refurbished CRDOA 21, was discussed in the maintenance
section of NRC Inspection Report 85-01. During this inspection, CRDOA 21
had already been refurbished and installed in the core. CRDOAs 26 and 2
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were nearly complete, and CRDOAs 15 and 4 were earlier in their refurbish-
ment programs. CRDOA 6 was removed from the core during this period and
began to be refurbished.
The number of procedure deviation reports being generated has diminished
since the previous inspection. This supports the observations by the NRC
inspector that the refurbishment program is now proceeding more smoothly.
The availability of parts has improved, as discussed in paragraph 2, and
the methods for dispensing them are continuing to evolve within procedural
constraints. The system is maintaining parts traceability. However, if a
replaced part is inadvertently left out of the list in the procedure, it
would be difficult, after the fact, to find documentation during an audit.
No violations or deviations were noted in this area.
4. Exit Interview
An exit interview was held March 8, 1985, with Mr. Singleton, Manager
Qualtiy Assurance and other PSC personnel as denoted in paragraph 1 of
this report. The NRC senior resident inspector and R. Farrell , a senior
resident inspector from another site in Region IV, also attended this
meeting. At this meeting the scope of the inspection and the findings
were summarized.
ERR REGp
c`. 49. UNITED STATES
4 a9 NUCLEAR REGULATORY COMMISSION
a 1 0
REGION IV
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4 yy 611 RYAN PLAZA DRIVE, SUITE 1000
„%*##H ARLINGTON,TEXAS 76011 "fill e^f,9r
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MAY 07 1S•85 _ +
In Reply Refer To:
Docket: 50-267/85-08 MAY 91985 /)
GRE4-1,g,y
cot°,
Public Service Company of Colorado
ATTN: 0. R. Lee, Vice President
Electric Production
P. 0. Box 840
Denver, Colorado 80201
Gentlemen:
This letter forwards the report of the Systematic Assessment of Licensee
Performance (SALP) Board for the Fort St. Vrain Nuclear Generating Station.
The SALP Board met on April 12, 1985, to evaluate the performance of the Fort
St. Vrain Nuclear Generating Station for the period October 1 , 1983, through
February 28, 1985. The performance analyses and resulting evaluations are
documented in the enclosed SALP Board Report. This report is being provided
in advance of a meeting to be held between the NRC and Public Service Company
of Colorado in late May or early June 1985. The specific time and place will
be announced later.
The performance of your facility was evaluated in the selected functional
areas identified in Section IV of the enclosed SALP Board Report.
The overall performance of Fort St. Vrain Nuclear Generating Station was
satisfactory but exhibited a continuation of the problems noted during the
previous SALP evaluation. Resources were strained or not effectively used such
that minimally satisfactory performance was achieved with respect to plant
operations; maintenance; licensing activities; quality programs and administrative
controls affecting quality; and design, design changes, and modifications.
Performance in these functional areas declined or showed no improvement since
the last SALP period. A decline was also noted in emergency preparedness.
Major strengths were noted in the areas of refueling and radiological controls,
with these areas showing improvement since the last SALP period.
We are concerned about the weaker performance in the five functional areas
identified above, and request that you develop an expedited response to Region IV
on the SALP Board's recommendations. Your responses should emphasize steps
that have been taken for prompt improvement and recent accomplishments; i .e,
plans implemented and results achieved since the latter part of the SALP
evaluation period. You should be prepared to discuss these plans and actions
at the meeting in late May or early June 1985, or earlier if practicable.
Public Service Company of Colorado -2-
Any comments which you may have regarding our evaluation of the performance of
your facility should be submitted to this office within 30 days of the date
of the formal meeting. Your comments , if submitted, and our disposition of
them, will be issued as appendices to the SALP Board Report.
Comments which you may submit are not subject to the clearance procedures of
the Office of Management and Budget as required by the Paperwork Reduction
Act, PL 96-511.
Should you have any questions concerning this letter, we shall be pleased to
discuss them with you.
Sincerely,
rfglit
obert D. Martin
Regional Administrator
Enclosure:
Appendix - SALP Board Report 50-267/85-08
Attachment 1 - NRR Licensing Input
Attachment 2 - AEOD Input
cc:
Mr. D. W. Warembourg, Manager
Nuclear Engineering Division
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
Mr. David Alberstein, 14/159A
GA Technologies, Inc.
P. 0. Box 85608
San Diego, California 92138
Kelley, Stansfield & O'Donnell
Public Service Company Building
550 15th Street, Room 900
Denver, Colorado 80202
Chairman, Board of County Comm.
of Weld County, Colorado
Greeley, Colorado 80631
(cont. on next page)
Public Service Company of Colorado -3-
Regional Representative
Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
Denver, Colorado 80203
Mr. H. L. Brey, Manager
Nuclear Licensing/Fuels Div.
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
J. W. Gahm, Manager, Nuclear
Production Division
Fort St. Vrain Nuclear Station
16805 WCR 194
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
SALP BOARD REPORT
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
SYSTEMATIC APPRAISAL OF LICENSEE PERFORMANCE
Inspection Report 50-267/85-08
Public Service Company of Colorado
Fort St. Vrain Nuclear Generating Station
October 1, 1983 - February 28, 1985
I. INTRODUCTION
The Systematic Assessment of Licensee Performance (SALP) program is an
integrated NRC staff effort to collect available observations and data on
a periodic basis and to evaluate licensee performance based upon this
information. SALP is supplemental to normal regulatory processes used to
ensure compliance to NRC rules and regulations. SALP is intended to be
sufficiently diagnostic to provide a rational basis for allocating NRC
resources and to provide meaningful guidance to the licensee' s management
to promote quality and safety of plant construction and operation.
An NRC SALP Board, composed of the staff members listed below, met on
April 12, 1985, to review the collection of performance observations and
data to assess the licensee performance 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 II of this report.
This report is the SALP Board's assessment of the licensee' s safety
performance at Fort St. Vrain Nuclear Generating Station for the period
October 1 , 1983, through February 28, 1985.
The SALP Board was composed of the following members of the NRC staff:
R. Denise Director, Division of Reactor Safety and Projects
RIV
R. Bangart Director, Division of Radiation Safety and
Safeguards, RIV
E. Johnson Chief, Reactor Project Branch 1, RIV
J. Miller Chief, Operating Reactors Branch 3, NRR
R. Ireland Chief, Special Projects and Engineering Section, RIV
G. Plumlee Senior Resident Inspector, RIV
P. Wagner Project Manager, RIV
Attendees at all or part of the SALP Board meeting were:
D. Powers Regional Technical Reviewer, RIV
W. Seidle Technical Assistant, RIV
J. Baird Chief, Emergency Preparedness Section, RIV
J. Everett Chief, Nuclear Materials Safety Section, RIV
B. Murray Chief, Facilities Radiological Protection
Section, RIV
II. CRITERIA
Licensee performance is assessed in selected functional areas, depending
whether the facility is in a construction, preoperational , or operating
phase. Each functional area normally represents areas significant to
nuclear safety and the environment, and are normal programmatic areas.
Some functional areas may not be assessed because of little or no licensee
activities or lack of meaningful observations. Special areas may be added
to highlight significant observations.
-2-
One or more of the following evaluation criteria were used to assess each
functional area.
1. Management involvement and control in assuring quality
2. Approach to resolution of technical issues from a safety standpoint
3. Responsiveness to NRC initiatives
4. Enforcement history
5. Reporting and analysis of reportable events
6. Staffing (including management)
7. Training effectiveness and qualification
However, the SALP Board is not limited to these criteria and others may
have been used where appropriate.
Based upon the SALP Board assessment, each functional area evaluated is
classified into one of three performance categories. The definitions of
these 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 or
construction 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 are
reasonably effective so that satisfactory performance with respect to
operational safety or construction 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 or construction is being
achieved.
The SALP Board has also categorized the performance trend over the course
of the SALP assessment period. The trend is meant to describe the
general or prevailing tendency (the performance gradient) during the SALP
period. This categorization is not a comparison between the current and
previous SALP ratings; rather the categorization process involves a
review of performance during the current SALP period and categorization of
the trend of performance during that period only. The performance trends
are defined as follows:
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Improved: Licensee performance has generally improved over the course of
the SALP assessment period.
Same: Licensee performance has remained essentially constant over the
course of the SALP assessment period.
Declined: Licensee performance has generally declined over the course of
the SALP assessment period.
III. SUMMARY OF RESULTS
1. Strengths
Major station design changes in progress or previously completed and
the aggressive control-rod-drive refurbishment program preparations
identified at the end of this appraisal period indicate management' s
attention to upgrading FSV to a more efficient, reliable, and safe
plant. Strong ALARA controls are evident during the performance of
all work activities.
2. Weaknesses
Management controls in all functional areas indicate weaknesses. The
most evident weaknesses were noted in the functional areas of plant
operations, maintenance, licensing activities, quality assurance
programs and administrative controls affecting quality and design,
design changes, and modifications. The SALP Board determined that
the examples of: failures to follow procedures, inadequate licensee
technical reviews and responses, inadequate cleanliness controls,
inadequate quality assurance program, and apparent breakdown in
design change controls were the strongest indicators of weaknesses in
management control .
Significant weaknesses were identified both during routine NRC
inspections and during the special NRC assessment of FSV in the area
of plant operations and maintenance which reflect an operating
philosophy that subscribes to a less formal mode of operation than is
common at other commercial nuclear power plants.
The areas of quality assurance and design control indicate
significant weaknesses concerning implementation of program
requirements which indicates a significant reduction in performance
level over the last appraisal period.
3. Performance Category
A summary of the licensee' s performance, as determined during the
SALP Board meeting, is shown in the table below. Also included is
the summary from the previous evaluation period.
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Previous Present Trend During
Performance Category Performance Category Latest SALP
Functional Area (9/1/82 to 9/30/83) (10/1/83 to 2/28/85) Period
A. Plant Operations 3 3 Same
B. Radiological Controls 1 Improved
1. Radiation Protection 2
2. Confirmatory Measure- 2
ments, Chemistry/
Radiochemistry
3. Radwaste Systems, 2
Effluent Release,
Effluent Monitoring
4. Transportation/Solid 1
Radwaste
5. Environmental Sur- 2
veillance
C. Maintenance 2 3 Declined
D. Surveillance 2 2 Same
E. Fire Protection *** 2 Improved
F. Emergency Preparedness 1 2 Same
G. Security and Safeguards 2 2 Declined
H. Refueling 2 1 Same
I. Licensing Activities 3 3 Same
J. Training 2 2 Same
K. Quality Programs 2/3** 3 Declined
and Administrative Con-
trols Affecting Quality
L. Design, Design Changes, 3 3 Declined
and Modifications
* Individual areas under radiological controls were not assigned a rating.
** Previously evaluated as two functional areas (quality assurance and
management controls).
*** No category assignment this period.
-5-
The total NRC inspection effort during this SALP evaluation period
consisted of 43 inspections involving a total of 3248 hours onsite by NRC
inspectors and contractors/consultants. A special audit of FSV operations
was also conducted during July-August 1984.
IV. PERFORMANCE ANALYSIS
A. Plant Operations
1. Analysis
This area has been inspected on a continuing basis by the NRC
senior resident inspector (SRI) and region-based personnel and
was examined by the NRC staff during the special assessment.
The four violations below involve activities in the functional
area of plant operations. These violations represent failure to
follow procedures.
• Severity Level IV Violation (50-267/8414-01). The nitrogen
blanketing subsystem for the prestressed concrete reactor
vessel (PCRV) cooling water surge tanks had been in a
deviation status without operation' s knowledge.
• Severity Level V Violation (50-267/8414-03). Reactor power
was increased above 2% without having completed the Overall
Plant Operating Procedure OPOP I.C. Master Check List.
• Severity Level IV Violation (50-267/8415-02). The
emergency firewater pump house fans were in a deviation
status without operation' s knowledge.
• Severity Level IV Violation (50-267/8416-01). No
administrative procedure for shift and relief turnover was
in effect.
The six licensee event reports (LER) listed below can be
attributed to plant operations:
• Following a turbine trip and reactor trip, primary coolant
moisture level increased. After the subsequent startup
with outlet temperature greater than 1200°F, the
concentration of total primary coolant oxidants exceeded
10 PPM. Operation continued in degraded mode of LCO 4.2. 10
for 1.6 hours. (LER 83-049)
• PCRV cooling water system outlet temperature exceeded the
120°F limit of LCO 4. 2.15(b). The normal audible/visual
alarm was isolated. Personnel failed to monitor
temperature and return cooling water flow to the heat
exchangers. (LER 83-052)
-6-
• Insertion of the neutron startup source fuel block into
Region 22 increased startup rate Channel II count rate and
initiated an automatic actuation of the reactor trip
circuitry. (LER 84-003)
• Loop I shutdown and two loop trouble trip occurred during
rod withdrawal for startup when a high level moisture
monitor tripped due to high moisture in the primary
coolant. Two of three low level moisture monitors for
Loop I were already manually tripped because they were
inoperable. (LER 84-006)
• Loop I shutdown occurred at 2% power when the rapid rise
relay tripped; this tripped the 4160/480 volt transformer;
this in turn tripped off the helium circulators due to loss
of circulator bearing water. The cause was apparently a
spurious actuation of the rapid rise relay contact.
(LER 84-007)
• A plant trip occurred from high pressure caused by icing in
the coolant purification system caused by high coolant
moisture content. The moisture came from automatic trip of
helium circulator caused by loss of bearing water from trip
of auxiliary transformer caused by trip of rapid rise
relay. (LER 84-008)
With respect to operational safety, satisfactory performance is
being achieved, but weaknesses are evident:
• Housekeeping has continued to be a problem as identified in
NRC Inspection Reports 83-25 (Open Item 8325-03), 84-14,
84-15 (Open Item 8415-03), 84-16, 84-18, 84-29, and 84-34.
• Weaknesses in the licensee' s startup procedures (Overall
Plant Operating Procedure) have been identified by the SRI
in NRC Inspection Reports 83-31 (Unresolved Item 8331-01),
84-13 (Open Item 8413-05) , and 84-14 (Violation 8414-03).
This was addressed as an item of concern due to the reactor
operators confusion over what the requirements were for
completing certain sections of the startup book and when to
complete these sections.
• Operator knowledge of equipment/plant status and
attentiveness to annunciators has been addressed as a
weakness as identified in NRC Inspection Reports 83-25
(Open Item 8325-01), 84-14, 84-15 (Unresolved Item 8415-01,
Violation 8415-02, and Open Item 8415-04) , 84-16
(Violation 8416-01) , 84-22, 84-26, 84-29, and 84-34.
-7-
• The NRC' s special assessment of the operation of FSV
conducted during this appraisal period disclosed weaknesses
in the conduct of operations which confirmed previous
observations by the NRC. These weaknesses were attributed
to an operating philosophy that appears to subscribe to
less formality and less rigid control of operations in
terms of the use of procedures, detail and verification
steps in procedures, and adherence to procedures, than is
common at other commercial nuclear power plants. (Refer to
NRC Inspection Reports 84-18 and 84-22. )
• As identified in this assessment report and in NRC
Inspection Report 84-16, failure to follow procedures by
operators and other plant personnel has continued to be a
problem.
With respect to operational safety, the following improvements
have been identified:
• PSC has taken steps to enforce procedure usage which makes
it clear that disciplinary action will be taken in any
documented failure to follow procedure violation as
identified in NRC Inspection Report 84-16. A positive
effect of this improvement is not yet evident.
• As identified in NRC Inspection Reports 84-13 and 84-14,
the licensee has begun a program of walkdowns on systems
disturbed during an outage to verify that piping and
instrumentation drawings (P&ID) agree with the as-built
system, and that the standard operating procedure (SOP)
valve lineups also agree with the as-built system prior to
plant startup from refueling.
• PSC has increased shift manning by requiring a third senior
reactor operator in the control room acting in a
supervisory capacity. This additional senior reactor
operator has provided additional oversight for plant
operations, adherance to 'Technical Specification and
procedural requirements, review of plant logs, shift
turnover, and training.
2. Conclusions
The licensee is considered to be in Performance Category 3 in
this area.
During this period, there has been some improvement in the
performance of operators; however, the planned improvements at
the management level are not evident yet. Overall the general
trend has been the same.
-8-
3. Board Recommendations
a. Recommended NRC Actions
A high level of NRC attention in this area should continue
in an effort to assure that the licensee' s attention is
directed at improving performance in this area.
b. Recommended Licensee Actions
Increased and vigorous management attention is required to
improve performance in this functional area. Licensee
management should emphasize adherence to procedures and
Technical Specifications in an effort to reduce the number
of procedural violations. Management should increase
monitoring of plant operations until compliance is
achieved. Management should also continue efforts for
improving Technical Specifications and operating
procedures.
B. Radiological Controls
1. Analysis
Ten inspections were conducted during the assessment period by
region-based radiation specialists regarding radiological
controls. These ten inspections covered the following areas:
radiation protection-normal operations; radiation protection-
control rod drive repair outages; radwaste management, effluent
releases, and effluent monitoring; chemistry/radiochemistry and
confirmatory measurements; transportation of radioactive
materials/solid radwaste; and environmental monitoring. Three
violations and two deviations were identified:
• Severity Level IV Violation (50-267/8420-03) . Failure to
make 10 CFR Part 50.72(a)(2) report regarding liquid
effluents that exceeded two times MPC limits.
• Severity Level IV Violation (50-267/8420-02) . Failure of
continuous liquid effluent monitors to terminate releases
that exceeded Technical Specification limits.
• Severity Level V Violation (50-267/8404-01). Failure to
follow health physics procedures.
• Notice of Deviation (50-267/8328-01) . Failure to designate
individual responsible for low-level waste transportation
program.
• Notice of Deviation (50-267/8328-02). Failure to conduct
training regarding transportation activities.
-9-
The five LERs listed below can be attributed to the functional
area of radiological controls:
• With the reactor at ≥69% power, an unsampled radioactive
gaseous release was made from the reactor building via the
filtered and monitored ventilation exhaust stack. The
release was calculated to be less than associated maximum
permissable concentrations. The release was caused by a
crack in the casing of the analytical moisture monitor
allowing the escape of primary coolant into the reactor
building. (LER 83-046)
• Personnel failed to adjust the radioactive gaseous effluent
activity monitor nominal alarm/trip setpoints in accordance
with the 0DCM. (LER 84-001)
• Reactor building sump sample was found to be above MPC for
unknown beta emitters. (LER 84-009)
• HP personnel failed to place continuous sampler in service
resulting in a ≥10 gpm liquid effluent release from the
reactor building sump without the sampler in service.
(LER 84-010)
• Both hot service facility area radiation monitors were
inoperable for a period of time longer than that allowed by
Technical Specifications. (LER 84-013)
a. Radiation Protection
This area was inspected five -times during the assessment
period. These inspections involved: two inspections con-
cerning work associated with control rod drive repair; one
inspection during the refueling outage; one inspection of
the ALARA program; and one inspection during routine
operation.
The person-rem at FSV continues to be below the national
average for light water reactors. The total exposure
during 1983 at FSV was 0.95 person-rem compared to the
national average of 735 person-rem for light water reactor.
The FSV total for 1984 was 3 person-rem. The 1984
person-rem for light water reactors have not been
tabulated, but it is expected that the 1984 data will be
about the same as 1983.
-10-
The licensee implemented a comprehensive program to control
radiation protection activities during control rod drive
repair work. The licensee devoted considerable effort
during the advanced planning and preparation phase to
assure proper controls were implemented during actual work
activities. The radiation protection department maintains
an aggressive program that requires adherence to
established controls.
The radiation protection manager was promoted to the posi-
tion of support services manager. This new position in-
cludes responsibility for managing the licensed and
nonlicensed training programs along with previous manage-
ment responsibilities for the radiation protection and
radiochemistry programs and functioning as the radiation
protection manager. During this same period, the support
services manager was also assigned the responsibility for
managing the environmental monitoring and water chemistry
programs which were previously the responsibility of the
corporate office.
The staffing of the radiation protection staff has been
stable during the assessment period; the turnover rate has
been less than 10%. All of the radiation protection
technicians meet the ANSI N18. 1-1971 qualifications as
senior health physics technicians.
b. Chemistry/Radiochemistry and Confirmatory Measurements
This area was inspected once during the assessment period
which included onsite confirmatory measurements with the
Region IV mobile laboratory. Problems continue to exist
concerning acceptable agreement between the NRC' s and the
licensee' s results of radionuclides identified on prepared
charcoal cartridge standards. This same problem existed in
the previous assessment period. Radionuclide analyses of
all other sample media were in 100% agreement. The
radiochemistry staff has experienced a low turnover rate
during the past several years. As a result, a stable
program exists with an adequate number of experienced
technicians.
Several problem areas were identified with the water
chemistry program concerning analytical procedures, QA/QC
program, and a formal training program. Organization
changes occurring during the assessment period included
transferring the responsibility for supervising the water
-11-
chemistry program from the corporate office to the site
under the management of the support service manager.
c. Radwaste Management, Effluent Releases, and Effluent Monitoring
This area was inspected initially as part of the routine
inspection program and a second time during a special
inspection. The special inspection included a violation
concerning the failure of the liquid effluent monitors to
terminate releases that contained high concentrations of
beta activity. The liquid releases at FSV are unique when
compared to light water reactors in that beta activity has
been the predominant concern during liquid releases. The
installed monitors, which are described in the FSAR, were
designed to respond to gamma activity. Consequently, the
monitors failed to terminate releases, which included high
concentrations of beta activity.
The effluent sampling and analyses activities are well
defined with only minor changes occurring in the program.
There has been little turnover in the past several years
regarding personnel responsible for performing effluent
analyses.
d. Transportation Activities and Solid Radwaste
This area was inspected once during the assessment period.
The licensee had made shipments involving both spent fuel
and low-level waste. Two deviations concerning IE
Bulletin 79-19 were identified: (1) failure to designate a
person responsible for the low-level waste transportation
program; and (2) failure to conduct training on
transportation activities.
The FSV facility generates small amounts of low-level waste
when compared to a typical light water reactor. The
licensee has made about 10 low-level radwaste shipments
between 1973 through 1984. Most of these shipments were
made in 1983 and consisted of contaminated reflector
blocks. No shipments were made in 1984. During late 1984
and early 1985, the licensee accumulated considerable
amounts of low-level waste as a result of the control rod
drive repair work, which is additive to that already stored
on the site. Some of this low-level waste is scheduled to
be shipped in late-1985.
The licensee had revised their procedures to include the
July 1, 1983, update to the DOT regulations, but procedures
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have not been developed that address the revisions to
10 CFR 20.311, 10 CFR 61.55, and 10 CFR 61.56 which were
effective December 27, 1983.
e. Environmental Monitoring
The environmental monitoring program was inspected once
during the assessment period. No significant problems were
identified during the inspection. The licensee amended
their radiological effluent Technical Specification in
November 1983 to be in agreement with the format in
NUREGs 0472/0473. The responsibility for management and
implementation of the environmental monitoring program was
transferred from the corporate office to the site under the
supervision of the support services manager. Under the new
organization, the environmental monitoring program is
included as part of the chemistry/radiochemistry
organization.
2. Conclusions
The licensee' s person-rem values continue to be less than 1% of the
national average for light water reactors. The radiation protection
and radiochemistry programs are well managed with a high level of
technical competence. The organization changes involving the water
chemistry program should provide improvements in the areas of QA/QC
activities, implementing procedures, and training. The reassignment
of responsibilities for the environmental monitoring program should
provide better technical oversight of program activities. The
radiation protection manager (support services manager) has been
assigned management responsibility for the training, environmental
monitoring, and water chemistry programs. The additional work load
will reduce the amount of time he has available for radiation protec-
tion manager duties.
The liquid effluent monitors described in the FSAR do not provide
adequate monitoring for beta activity. Transportation procedures
have not been revised to include the December 27, 1983, update to
10 CFR 20.311, 10 CFR 61.55 and 10 CFR 61.56. Confirmatory
measurement results on prepared charcoal cartridge standards indicate
disagreement between the NRC' s and the licensee' s measurements.
The licensee is considered to be in Performance Category 1 in this
area.
Trend: Improved.
-13-
3. Board Recommendations
a. Recommended NRC Actions
The overall level of NRC inspection effort in this area can be
reduced but additional inspection related to water chemistry,
transportation, liquid effluent control will be performed.
b. Recommended Licensee Actions
Management attention is needed to assure that the radiation
protection manager has adequate time available to devote to
radiation protection activities. The licensee' s calibration
program regarding the analyses of charcoal cartridges should be
reviewed to verify the validity of their measurements.
Transportation procedures should be updated to assure the
requirements of 10 CFR 20.311, 10 CFR 61.55 and 10 CFR 61.56 are
included as part of the transportation program. Management
needs to take positive steps to ensure that unplanned and
unmonitored releases do not occur.
C. Maintenance
1. Analysis
This area was inspected on a continuing basis by the SRI and one
inspection was performed by region-based personnel . The three
violations below involve activities in the functional area of
maintenance. One violation (8422-01) consists of two parts
against maintenance with a third part involving activities in
the functional area of quality assurance. One violation
(8429-02) involved activities in both the functional area of
maintenance and quality assurance. These violations can be
attributed to the failure to follow procedures.
• Severity Level V Violation (50-267/8415-08) . A special
test was in progress without having the shift supervisor' s
signature documenting permission to initiate the test.
• Severity Level IV Violation (50-267/8422-01).
a. During a review of a design change to the steam generator
marmon flanges, the NRC inspector determined that weld data
sheets were not attached to the controlled work procedure
(CWP), weld data sheets did not contain the required
testing and visual inspection requirements, weld rod
control was not in accordance with Procedure WM-7, and weld
data reports were not completed.
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b. (This part of the violation is listed under the area of
quality programs and administrative controls affecting
quality. )
c. It was also determined during the above review that
electrode control was not recorded on the weld rod control
form.
• Severity Level IV Violation (50-267/8429-02). The assembly
of the shim motor and brake subassembly on control rod
drive (CRD) 21, an activity affecting quality, had a hold
point and procedural steps that were not signed off as
required.
The nine LERs listed below can be attributed to the functional
area of maintenance:
• Emergency diesel generator set taken out-of-service due to
blown air filter petcock disabling the air starter motor.
Reportable as a degraded mode of LCO 4.6.1(d) .
(LER 83-041)
• "lA" Boiler feedpump removed from service three times from
October 6-11, 1983, to resolve speed oscillation problems
while "1C" boiler feedpump was out for repairs. Plant at
power. Report required due to degraded mode of LCO 4.3.2.
(LER 83-042)
• The bearing water makeup pump was removed from service to
perform maintenance on leaking valves and filters
associated with the pump. Pump was removed from service
with reactor at power constituting a degraded mode of
LCO 4.2.2(d) . (LER 83-044)
• The Halon 1301 system and associated isolation dampers in
the three-room complex were removed from service to perform
modification work associated with CN 1722. (LER 83-045)
• One Class I hydraulic shock suppressor (snubber) found
inoperable. No visible oil level in reservoir due to
leaking reservoir gaskets. (LER 83-047)
• Two of three of the "10" helium circulator bearing water
pressure differential switches were found inoperable due to
an accumulation of dirt and oil on the switches.
(LER 83-048)
-15-
• Mounting bolt washout and concurrent vibration damaged air
intake filter and generator mounting assembly on the
ACM-DG. ACM-DG was out-of-service for almost 2' days.
(LER 83-051)
• Bearing water makeup pump (P-2105) removed from service to
perform maintenance on leaking valves. Constitutes
operation in a degraded mode of LCO 4.2.2(d) . (LER 83-053)
• On two successive days, the emergency feedwater supply
header to the Loop I helium circulators was isolated to
repair leaking valves. This constitutes operations in a
degraded mode of LCO 4.2.2(a) . (LER 83-054)
With respect to maintenance, satisfactory performance is being
achieved, but some weaknesses are evident:
• Problems with improper completion of plant trouble reports
(PTR) and station service requests (SSR) were identified in
NRC Inspection Reports 84-01, 84-13 (Open Item 8413-03) ,
and 84-34.
• As identified in NRC Inspection Report 84-16, the backlog
of PTRs has continued during this assessment period and was
most evident by the large number of "out-of-service tags"
on switches, alarm windows, and other equipment in the
control room.
• Poor maintenance practices were identified in NRC
Inspection Reports 84-16, 84-22, 84-29, and 84-34 (Open
Item 8434-01) .
• Poor housekeeping was also identified as an area of concern
in NRC Inspection Reports 84-29 and 84-30.
• The October 1984 NRC assessment of FSV maintenance
activities resulted in the following concerns:
1) Scheduling - no formalized priority system.
2) Preventive Maintenance - appeared to be a static,
nontechnical approach, rather than a dynamic,
technology-based, engineered method for ensuring equipment
readiness.
3) Spare Parts Management - no shelf-life program and
no special designation for safety-related items.
-16-
4) Maintenance Procedure - some procedures were not
precise.
5) Maintenance Testing - licensee does not endorse 100%
postmaintenance testing.
6) Backlog - no means to display the status of backlog
items to management.
With regard to maintenance, the following improvements have been
identified:
• The licensee has developed the Power Plant Maintenance
Information System (PPMIS) for maintenance documentation
gathering.
• The licensee has initiated a large—scale cleanup program in
response to the June 1984 special inspection (NRC
Inspection Report 84-16) by Region IV personnel .
2. Conclusions
Due to the increase in the number of violations in this
reporting period versus the last SALP, as well as the marked
decrease in the quality of maintenance activities as identified
in the above analysis over the last reporting period, the
licensee' s overall performance rating has decreased in this
functional area.
The licensee has initiated programs designed to improve their
performance in this area, however, as of the end of this
reporting period, no significant improvement has been achieved.
The licensee is considered to be in Performance Category 3 in
this area.
Trend: Declined.
3. Board Recommendations
a. Recommended NRC Actions
The NRC inspection effort in this functional area should be
increased. Emphasis should be placed in the areas of
housekeeping, scheduling, preventive maintenance, spare
parts management, maintenance procedures, maintenance
testing, and maintenance practices.
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b. Recommended Licensee Actions
Licensee management attention should be increased in the
areas of weaknesses identified above. Initiatives should
also be undertaken to strengthen overall management control
in this area.
D. Surveillance
1. Analysis
This area was inspected on a continuing basis by the SRI. The
one violation below involved activities in the functional area
of surveillance. This violation can be attributed to a failure
to follow procedures.
• Severity Level IV Violation (50-267/8329-01) . PCRV cooling
system subheader temperature alarms were not tested and
temperature readings were not checked as required.
The eight LERs listed below can be attributed to the functional
area of surveillance:
• Surveillance requirement for PCRV cooling water system
temperature scanner not performed within the required
interval . A temporary scheduling technician did not
deliver the test to the responsible department prior to the
"late" date. (LER 83-050)
• During surveillance test, one of eight helium circulator
penetration safety valves was found to relieve at a setting
below the minimum acceptable set pressure. (LER 83-055)
• Routine surveillance test identified one of three bearing
water pressure differential switches inoperable.
(LER 84-002)
• Surveillance test requirements for SR 5.2.23, firewater
booster pump surveillance were not performed in conjunction
with Surveillance Test SR 5.2.7a-A, water turbine drives
surveillance due to a procedure inadequacy caused by
personnel modifying the procedure without adequate
technical review of the change. (LER 84-004)
• Corrosion and failure of PCRV tendon wires. (LER 84-005)
• Surveillance test identified five of twelve UT detectors
for steam pipe rupture under the PCRV and one of twelve UT
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detectors outside the PCRV as out-of-tolerance limits.
(LER-011)
• Reserve shutdown hopper of Control-Rod-Drive-and- Orifice
Assembly 21 failed the functional test of SR 5. 1.2c-X when
only half of the reserve shutdown material was released.
(LER 84-012)
• Semiannual surveillance for loss of outside power and
turbine trip failed. Both diesel generator tie breakers
failed to close. (LER 84-014)
2. Conclusions
With respect to surveillance, satisfactory performance is being
achieved.
The licensee is considered to be in Performance Category 2 in
this area.
Trend: Same.
3. Board Recommendations
a. Recommended NRC Actions
The NRC should maintain a normal level of attention in this
area.
b. Recommended Licensee Action
The licensee management should take steps to ensure the
accuracy of surveillance procedures. The proposed program
for Technical Specification upgrade to incorporate a
surveillance for each limiting condition for operation
(LCO) should continue to be actively pursued.
E. Fire Protection
1. Analysis
This area was inspected on a continuing basis by the SRI . One
inspection was performed in this area by a region-based
inspector. The one violation and two deviations below involved
activities in the functional area of fire protection. The
violation can be attributed to the failure to follow procedures.
-19-
• Notice of Deviation (50-267/8412-09) . Public Service
Company of Colorado committed, in Section 2.0 of the fire
protection program review for Fort St. Vrain Nuclear
Generating Station in response to Branch Technical
Position 9.5-1, submitted by letters P-78167 and P-78182,
dated October 13, 1978, and November 13, 1978, to certain
planned improvement actions in order to meet the guidelines
of the Branch Technical Position 9.5-1. In deviation from
this commitment: (1) three 55-gallon drums of oil were
found stored in front of the Loop 2 buffer-helium panel
(S 2112) and one 55-gallon drum of oil was found stored
just under the end of the bearing water cooler (E 2105 S) ;
(2) the hydraulic oil mist detector had not been hooked up
and is inoperable; (3) the fire detectors to be installed,
according to Table 2.0-1, have either not been installed or
made operable; and (4) lubricating oil is still being
stored in drums in the fuel handling purge system equipment
room.
• Severity level IV Violation (50-267/8414-02) . Areas were
identified inside the reactor building where: unused
combustible material was not stored in covered flameproof
containers; used combustible material was not kept in
noncombustible bins or containers; equipment manuals
(procedures) were not stored in suitable cabinets; and work
areas were not controlled as required.
• Notice of Deviation (50-267/8430-04) . Installation of fire
detection equipment and an oil mist detection system had
not been completed by October 1984 as committed in
letter P-84194 and a request for change in schedule had not
been received by the NRC.
No LERs can be attributed to the area of fire protection.
A special fire protection safety inspection (NRC Inspection
Report 83-23) was conducted August 22-26, 1983, to determine
compliance with 10 CFR 50.48 and applicable sections of
10 CFR 50, Appendix R. As identified in the previous SALP
Report 83-30, the results of this inspection were under review
and evaluation. The results of this inspection indicated the
need to expeditiously resolve the issue of compliance of FSV to
the requirements of Appendix R.
FSV responded to NRC letters that they were in compliance with
Appendix R. The licensee relied on previous (BTP 9.5. 1
Appendix A) evaluations even though 10 CFR 50.48 and NRC letters
required a reassessment to ensure compliance with Items III .G. ,
-20-
III .3. , and III.0. of Appendix R. The staff requested that the
required reassessment be promptly completed so that compliance
determinations can continue. The licensee subsequently
committed to perform the reassessment which was to consist of a
four report submittal . As of the end of this reporting period,
the staff has received the first, second, and third reports
(i .e. , Shutdown Model , Electrical Reviews, and Fire Protection)
of the four-part evaluation performed by the Tenera Corporation
for PSC. The fourth report, to consist of proposed
modifications and exemption requests, was received April 1,
1985. A subsequent commitment was made to provide a fifth
report providing a BTP 9.5-1, Appendix A type evaluation for the
new Building 10 and is due June 1, 1985. The reports are
currently under review/evaluation by the staff. Findings from
this review will be factored into the next appraisal .
With respect to fire protection, satisfactory performance is
being achieved, but weaknesses are evident:
• Housekeeping, as identified in the violation and deviation
above as well as in other functional areas of this report,
has continued to be an area of concern during this
assessment period.
• The licensee did not agressively pursue the evaluation and
resolution of Appendix R issues until late in the
evaluation period.
2. Conclusion
The licensee has initiated programs to improve plant cleanliness
which has resulted in bringing the plant' s housekeeping to what
must be considered an average level . However, housekeeping
still continues as an area in need of further improvement.
The licensee is considered to be in Performance Category 2 in
this area.
Trend: Improved.
3. Board Recommendations
a. Recommended NRC Actions
The level of NRC inspection effort should remain about the
same with emphasis placed on the licensee' s completion of
10 CFR Part 50, Appendix R, commitments.
-21-
b. Recommended Licensee Actions
Licensee management must continue its involvement in
upgrading the fire protection system to meet
10 CFR Part 50, Appendix R, requirements. Also, licensee
management should continue to place emphasis on improving
housekeeping practices at FSV.
F. Emergency Preparedness
1. Analysis
During the assessment period, three emergency preparedness
inspections were conducted. The first inspection, conducted
October 24-28, 1983, was a review of the emergency preparedness
program in the areas of emergency detection and classification,
protective action decision making, emergency notification and
communication. No violations or deviations were identified, but
concerns were identified in regard to shift supervisors
capabilities in following emergency implementing procedures,
describing the emergency coordinator function, and the transfer
of authority during the course of the emergency. These NRC
concerns were not effectively addressed in a timely manner as
demonstrated by similar problems identified during the
licensee' s emergency exercise in August 1984.
The next inspection was conducted during the period February 27
through March 2, 1984. This inspection reviewed the licensee' s
program in the areas of emergency response organization staffing
and augmentation, emergency training, emergency worker
protection and changes to the emergency preparedness program.
No significant problems were identified during this inspection.
The last inspection was conducted during the period
August 13-17, 1984. The inspection included observation of the
licensee' s annual emergency exercise conducted on August 15,
1984. During this exercise, significant problems were observed
in regard to station staff following emergency plan implementing
procedures and demonstrating effective management control of the
emergency response facilities resources. In addition, a
violation was identified in regard to the distribution and
timeliness of submittal to NRC of amendments to the emergency
plan. The licensee' s response to these findings was considered
to satisfactorily address the NRC concerns, but had not been
verified by inspection by the end of the period.
During the August 15, 1984 exercise, an adequacy survey of the
prompt public alert and notification system (tone alert radios)
-22-
was conducted by the Federal Emergency Management Agency (FEMA) .
Subsequent analysis of survey data led to a conclusion by FEMA
that the system performance did not provide reasonable assurance
that the system was adequate to alert and notify the public in
the plume exposure emergency planning zone in the event of an
accident at the station. Just prior to the end of the
assessment period, this deficiency was identified to the
licensee along with a request for corrective action and use of
compensating measures until the action is completed. This issue
had not been resolved by the end of the period.
Several organizational changes were made during the assessment
period in the areas of station management and emergency
preparedness which appeared to have the potential to strengthen
the emergency preparedness program.
There were no reportable events in the emergency preparedness
area received during this reporting period.
2. Conclusions
The licensee has maintained an acceptable level of emergency
preparedness during the period and demonstrated adequate
capability to protect the health and safety of the public during
an emergency exercise. However, the licensee' s program in this
area decreased in effectiveness during the period from the level
attained during the previous assessment period. Corporate and
station management appeared to be actively involved in the
program but management controls were not effective in
maintaining the level of preparedness and resolving NRC concerns
during the first two-thirds of the period. Organizational
changes made in the latter part of the period appeared to
strengthen the licensee' s management controls and provide more
attention to implementation of an effective emergency
preparedness program.
The licensee is considered to be in Performance Category 2 in
this area.
Trend: Same.
3. Board Recommendations
a. Recommended NRC Action
The NRC inspection effort in this functional area should be
conducted in accordance with the normal inspection program.
-23-
b. Recommended Licensee Action
The licensee should maintain a high level of attention in
this area to assure that the management changes made during
the latter part of the period are effective in addressing
the concerns identified by the NRC during the period and
reversing the downward trend observed during the first part
of the period.
G. Security—Safeguards
1. Analysis
The physical security staff performed five inspections during
this SALP period. NRC Inspection Reports 83-32, 84-05, 84-09,
84-23, and 84-35 document the findings of these inspections.
All inspections were routine. All inspections except 84-35 were
unannounced.
The two violations listed below involve activities in the
functional area of security:
• Severity Level IV Violation (50-267/8423-02) . Inadequate
key control .
• Severity Level IV Violation (50-267/8423-03) . Failure to
maintain operable detection aids.
A management meeting was held on February 19, 1985, to discuss
vital equipment withheld from Revision 14 to the physical
security plan that had previously been characterized as vital
equipment. The licensee was given the opportunity to provide
the technical basis for describing this equipment as nonvital or
place this equipment back in the physical security plan and
protect it in accordance with 10 CFR 73.55(c)(1). The
management meeting was called to discuss this inadequacy of the
physical security plan discovered in December and documented in
Inspection Report 84-35. The licensee was noted to lack
initiative in making improvements in the security program.
2. Conclusion
The licensee has now improved his security plan, and present
licensee management is more effective and responsive to the need
to improve the program. They still need to demonstrate
increased initiative. The licensee is considered to be in
Performance Category 2.
Trend: Declined.
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3. Board Recommendations
a. Recommended NRC Action
Continue basic program.
b. Recommended Licensee Action
The licensee management should take an aggressive approach
in resolving security-related issues.
H. Refueling
1. Analysis
This area was inspected on a periodic basis by the SRI during
the licensee' s third refueling outage. No violations or
deviations were identified. The conduct of refueling activities
reflected preventive maintenance, adequate staffing, and
training which resulted in a smooth and uneventful refueling.
No LERs can be attributed to the area of refueling.
2. Conclusion
With respect to refueling, satisfactory performance is being
achieved.
The licensee is considered to be in a Performance Category 1 in
this area.
Trend: Same.
3. Board Recommendations
a. Recommended NRC Actions
The level of NRC inspections should remain consistent with
the basic inspection program.
b. Recommended Licensee Actions
The licensee should maintain the present level of attention
in this area. The licensee should perform adequate
planning prior to the next refueling to ensure that the
fuel examination activities and the equipment necessary are
-25-
well defined and agreed to between the NRC and the
licensee.
I. Licensing Activities
1 . Analysis
The evaluation of the licensee' s performance in this functional area
was based on consideration of the seven attributes specified in NRC
Manual Chapter 0516, but was weighted in favor of the key attributes
of management involvement, resolution of technical issues, and
responsiveness. A discussion of the analysis of this functional area
is contained in the enclosure to this report.
2. Conclusion
Based on our evaluation of the attributes as they relate to the
licensing activities, an overall Performance Category 3 is
determined. Specifically, management attention and involvement
must be improved and resolution of NRC initiatives must be more
aggressively pursued.
3. Recommendations
In order to improve performance in the licensing activities
functional area, the following is recommended:
a. PSC nuclear department staffing levels should be
reevaluated to ensure that there is adequate attention to
licensing issues, and that personnel are made aware of and
knowledgeable in the light water reactor licensing
activities;
b. A program should be implemented that will keep PSC
management informed of current NRC initiatives in other
reactors, and provide for an assessment of how those
initiatives could affect Fort St. Vrain; and
c. PSC should identify an individual with responsibility to
ensure complete, timely, and correct responses to NRC
requests.
J. Training
1. Analysis
Two specific inspections in this functional area were performed
by NRC region-based inspectors. Eight inspections conducted in
-26-
other functional areas also assessed training quality and
effectiveness for that area. There were no violations or
deviations identified in the training areas. During the period,
there were numerous cases of personnel failing to follow
procedures.
No LERs can be attributed to the area of training.
Two senior reactor operator replacement retake examinations were
administered in February 1984. Both candidates passed the
examination. The Fort St. Vrain Requalification Program
was evaluated as satisfactory after administering examinations
to a sample of licensed personnel in July 1984. Overall
operator licensing performance, based on this small sample, has
been acceptable during the SALP review period.
2. Conclusions
The licensee continues to demonstrate a satisfactory training
program although weaknesses are evident by the performance noted
in several of the above functional areas. The licensee is
considered to be in Performance Category 2 in the functional
area of training.
Trend: Same.
3. Board Recommendations
a. Recommended NRC Actions
The level of NRC inspection effort in this functional area
should focus on training effectiveness.
b. Recommended Licensee Actions
The licensee should place emphasis on continuing to pursue
the INPO accreditation program. The licensee should
emphasize the importance of procedure compliance in the
training program. The licensee should prepare for training
on the revised operating procedures and technical
specifications.
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K. Quality Programs and Administrative Controls Affecting Quality
1. Analysis
This area was inspected on a continuing basis by the SRI and
periodically by region-based inspectors. The inspections
included organization and administration, safety review
committees, licensee audits, quality assurance (QA) program,
quality control program, procurement activities, and quality
materials receipt, handling, and storage. Included in the scope
of this functional area was the licensee' s utilization of the
quality control (QA/QC) organization and maintenance quality
control (MQC) organization.
The degree and success of administrative controls exerted by the
licensee over safety—related activities at FSV was not the
subject of specific inspections during this evaluation period,
but management involvement is considered during many inspection
activities. Included within the scope of this functional area
is management' s utilization of the plant operations review
committee (PORC) , and the nuclear facility safety committee
(NFSC).
The ten violations and one deviation below involved activities
in this functional area. One part of a three-part violation
(8422-01) involved activities in this functional area. One
violation (8429-02) involved activities in both this functional
area and maintenance. These violations can be attributed to the
failure to follow procedures and an inadequate QA program.
• Severity Level IV Violation (50-267/8401-01) . The required
receipt inspection had not been performed on safety-related
switchgear purchased for electrical system modifications to
be made during the third refueling outage.
• Severity Level IV Violation (50-267/8401-04) . A QA/QC
inspector failed to document nonconforming reserve shutdown
material purchased under Purchase Order (PO) N3554, and
failed to accumulate the QA PO record file for the purchase
as required to include all pertinent documents.
• Severity Level V Violation (50-267/8401-05) . MQC was not
reviewing all new/revised maintenance procedures as
required
• Severity Level IV Violation (50-267/8414-04) . Surveillance
SR 5.2.16f-RX, "PCRV Auxiliary System Penetration Check
Valve Test," and Maintenance Procedure MP 11-3,
"Repair/Replacement of Reactor Penetration Purge Flow and
RSD in Line Check Valves," had not been followed and
SR 5.2. 16f-Rx did not contain appropriate proof tests.
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• Notice of Deviation (50-267/8415-06) . In deviation from
PSC letter P-80028 in response to actions taken at FSV to
implement the TMI requirements, an annual independent check
of plant operations and specifically shift turnover
procedures had not occurred during 1983.
• Severity Level IV Violation (50-267/8422-01) .
a. (See the functional area of Maintenance. )
b. Postweld heat treatment specification data and report
sheets were not completed and/or attached to the control
work procedure (CWP), mandatory QA/QC inspection hold
points were not identified, and the PWHT charts were not
signed/dated by QA/QC.
c. (See the functional area of Maintenance. )
• Severity Level IV Violation (50-267/8422-02) . No
inspection points had been inserted in the CWP-Deviation
Reports for CWP 84-120; a hold point had not been assigned
to an item that required all work to stop in order to
perform the inspection; and the review of completed
CWPs 83-171 and 84-74 had not been performed.
• Severity Level IV Violation (50-267/8429-02) . The
procedure for the assembly of the shim motor and brake
subassembly on CRD 21, an activity affecting quality, had a
hold point and procedural steps that were not signed off as
required.
• Severity Level IV Violation (50-267/8430-02) . Applicable
portions of the licensee' s QA program, as defined in their
FSAR for the alternate cooling method (ACM) equipment, the
fire protection system, and the plant security system, were
not documented by written policies, procedures, or
instructions.
• Severity Level V Violation (50-267/8501-01) . Items
received for Supplement 2 to PO N-5868 were not inspected
using MRIM-2.4.
• Severity Level IV Violation (50-267/8503-01).
Quality-related-confirming orders for control-rod-drive and
orificing-assembly parts were processed without nuclear
engineering division and quality assurance reviewed
purchase requisitions.
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The one LER listed below can be attributed to this functional
area:
• GA Technology' s assumptions used in development of
Figure 4. 1.9-1 and 4. 1.9-2 of LCO 4.1.9 were not
conservative and could have led to reactor operation in a
manner less conservative than that assumed in the basis of
the specification. Plant was operating at about 66% power
when this error was identified. (LER 83-043)
With respect to QA, satisfactory performance is being achieved,
but weaknesses are evident:
• There has been a significant increase in numbers of
procedural violations during the appraisal period.
• The increase in failure to comply with procedures
describing the licensee' s QA program (Violations 8401-01,
8401-04, 8401-05, 8414-04, 8422-01, 8422-02, and 8429-02)
and the failure to document portions of this program
(Violation 8430-02) is considered a significant decrease in
performance.
• NRC Inspection Reports 84-26 (Violation 8426-02) and 84-29
(Violation 8429-04) indicate weaknesses in the licensee' s
corrective action program for which problems previously
identified by the licensee' s QA department were not
corrected in a timely manner.
• This appraisal period has identified numerous weaknesses in
the area of procurement and receipt inspection. (e.g. ,
Unresolved Item 8401-03, Open Item 8417-07, and Violations
8401-01, 8401-04, 8501-01, and 8503-01)
• Responses to QA audit findings frequently are not timely or
responsive.
• Inadvertent deletion of QA commitments to the NRC has been
identified as a weakness (Open Item 8415-07 and
Deviation 8415-06) .
• Inadequate MQC inspection, control of nonconforming
material , and traceability of safety-related parts have
been identified as a weakness (NRC Inspection
Reports 84-18, 84-30, and 85-03).
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• During the NRC assessment of FSV operations in July and
August 1984, concerns were identified with the licensee not
having a shelf-life program and not having a parts
management system with a special designation for
safety-related items. This review also concluded that PSC
should undertake initiatives designed to strengthen overall
management control .
• NRC Inspection Report 84-22 addressed the NRC' s concerns
over management controls affecting quality that permitted
the recurrence of the types of violations and activities as
noted in the report.
As noted in NRC Inspection Report 84-34, the licensee has
contracted Gilbert Commonwealth, Inc. , to perform: (1) a
complete review of FSV licensing requirements, (2) clearly
define and provide procedural guidance for commercial grade
components, sole source suppliers, identical replacement parts,
and vendor qualification criteria, and (3) prepare a new
definition of "quality-related." Several of the contractor' s
resulting recommendations were beyond the original scope of the
proposed work, but were evaluated by PSC to be legitimate and
warranted further investigation. Gilbert Commonwealth' s review
indicated that significant weaknesses exist in the licensee' s QA
program within the areas of classification of items, FSV
administrative procedures, vendor evaluations and source
inspection, receiving inspection activities, and storeroom
separation and shelf-life.
As identified in the licensee' s letter dated February 28, 1985,
P-85066, the licensee contracted with NUS Operating Services
Corporation to perform an independent assessment of PSC' s
management controls for its nuclear activities. This
investigation confirmed the basic concerns that were identified
by the NRC and provided recommendations which, if aggressively
implemented on a timely basis, should correct the identified
weakness. PSC submitted an action plan constituting a
first-phase response to the NUS report.
2. Conclusions
Significant weaknesses have been identified during this
appraisal period in the areas of administrative controls and
oversight of nuclear activities, procedural violations,
corrective action program, procurement and control of
nonconforming materials, and parts management.
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The licensee has initiated programs designed to improve their
performance in this area, however, as of the end of this
reporting period, the programs are still being developed and no
significant improvement has been observed.
The licensee is considered to be in Performance Category 3 in
this area.
Trend: Declined.
3. Board Recommendations
a. Recommended NRC Actions
The NRC inspection effort in this functional area should be
increased. Emphasis should be placed in the areas of:
design changes; QA program; and followup of licensee
corrective actions, management' s program to ensure
procedural compliance and implementation of appropriate
Gilbert Commonwealth' s and NUS recommendations.
b. Recommended Licensee Actions
Licensee management attention must be increased to the
extent necessary to provide sufficient management oversight
of its nuclear activities. Emphasis should be increased
in: the QA department' s independence and capability to
provide timely corrective action; implementing, on a timely
basis, contractor recommendations to correct identified
weaknesses; and improving the quality of communications
between the PSC QA, production, and engineering
departments. The licensee should implement an increased
audit/monitoring activity to identify and correct weak
areas.
L. Design, Design Changes, and Modifications
1. Analysis
This area was inspected on a continuing basis by the SRI . One
inspection was performed in this area by region-based personnel .
The eight violations and one deviation identified below involve
activities in the functional area of design, design changes, and
modifications. The violations below can be classified as
failures to follow procedures and inadequate design controls.
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• Severity Level V Violation (50-267/8326-01) . The NRMCA
certificate was not available and the certification for the
batch plant had not been approved by the NRMCA.
• Severity Level V Violation (50-267/8412-01). The control
room copy of Drawing PI-21-10 had not been updated as
required.
• Notice of Deviation (50-267/8414-05). In deviation from
PSC letter P-83368, dated November 10, 1983, in response to
violations contained in NRC Inspection Report 83-24, "old"
CWP forms were still in use on May 14, 1984.
• Severity Level V Violation (50-267/8414-12). The process
of performing/controlling CWPs, procedure/inspection/test/
reports (PITR) , and deviation reports (DR), which are
activities affecting quality, were not prescribed by
instructions/procedures.
• Severity Level IV Violation (50-267/8422-03) .
a. CWP-DRs for CWP 84-120 affected the tagging boundaries
and were not approved by the shift supervisor; CWPs 83-171
and 84-74 were not processed, controlled, and implemented
as required; and the systems addressed in CWPs 83-171 and
84-74 were returned to service without the shift supervisor
performing the required verifications.
b. The shift supervisors were not following the corrective
actions in the licensee' s letter P-82049, in response to
Violation 8126-03, concerning the return of modified
systems to service.
c. The licensee' s administrative procedure revision in response
to Violation 8324-01 did not prevent the shift supervisor
from returning a system to service without performing the
required verifications.
• Severity Level IV Violation (50-267/8422-06) . Licensee
submittals to IE Bulletin 80-11 were not submitted under
oath or affirmation and did not provide a complete detail
of wall modifications with drawings as required.
• Severity Level IV Violation (50-267/8426-02) . The
controlled work procedure manual was not being used during
the preparation of CWPs.
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• Severity Level IV Violation (50-267/8429-03). Requirements
for controlling and documenting safety-related design
changes were not followed.
• Severity Level IV Violation (50-267/8429-04) . Design
changes on safety-related equipment were authorized based
on engineering judgement for which no design verification
or checking was performed.
No LERs can be attributed to this area.
With respect to design, design changes, and modifications,
satisfactory performance is being achieved, but weaknesses are
evident:
• NRC Inspection Reports 84-14 and 84-26 identified poor
planning/analysis in the area of design modifications.
• NRC Inspection Report 84-22 (Unresolved Item 8422-05) and
84-29 (Violation 8429-04) identified the use of
"engineering judgement" as a design change justification
without supportive design verification.
• As identified in the above violations and in the previous
SALP, the breakdown in the licensee' s design change program
regarding newly modified systems being returned to
operation without the required documentation and shift
supervisor verification continued to be a weakness.
• A significant weakness was identified concerning certain
portions of the licensee' s design program not being
prescribed by instructions/procedures.
• The same statement from the last two appraisal reports
should again be reemphasized, "This functional area
requires considerable coordination between the nuclear
engineering division and the nuclear production division on
a day-to-day basis." Coordination between the design
groups and the QA department has also been identified as a
weakness during this appraisal period.
2. Conclusions
The licensee is considered to be in Performance Category 3 in
this area.
Trend: Declined.
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3. Board Recommendations
a. Recommended NRC Actions
The level of NRC inspection effort in this functional area
should be increased with particular emphasis placed on the
licensee' s development and control of modifications.
b. Recommended Licensee Actions
Increased management attention in the area of modification
controls; coordination between NED, NED-site, QA, and
production; CN preparation; and increased emphasis on
implementation of 10 CFR 50.59. .
V. SUPPORTING DATA AND SUMMARIES
A. Licensee Activities
1. Major Outages
October 29, 1983 - Unscheduled outage for 62. 1 hours -
automatic scram due to moisture.
Outage extended to perform surveillance
testing.
November 1, 1983 - Continued outage from October to
complete scheduled surveillance testing
- 167 hours.
January 19, 1984 - Commenced scheduled refueling -
thru May 16, 1984 included turbine overhaul , routine
corrective and preventive maintenance,
"A" helium circulator changeout, PCRV
tendon surveillance.
June 22, 1984 thru - Unscheduled outage continued for
February 28, 1985 control-rod-drive-mechanism-
malfunction investigation and overhaul
of mechanisms. Helium
circulator investigation and
refurbishment.
2. Power Limitations
The reactor power level was limited to 2% power from April 19,
1984, through May 16, 1984, pending a final resolution of the
PCRV tendon wire corrosion problem. The licensee presently has
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a continuing administrative limit of 85% reactor power pending
completion of rise-to-power (B-0) testing.
3. License Amendments
Refer to the attached licensing enclosure.
4. Significant Modifications
Major modifications completed during this appraisal period were
Building 10 construction, which was completed in January 1984;
and the 480V system upgrade, which was completed during the
third refueling.
B. Inspection Activities
1. Violations
See Table 1
2. Major Inspections
The special inspections listed below were conducted during this
appraisal period:
• A special inspection of licensee activities occurred on
June 4-5, 1984, and is documented in NRC Inspection
Report 84-16.
• On July 9-13, 1984, the NRC staff audited the overall
operation of FSV. The staff defined as areas for review:
(1) the failure of 6 of 37 control-rod pairs to
automatically insert on a scram signal on June 23, 1984;
(2) the overall conduct of operations, including
maintenance and housekeeping; (3) assessment of existing
Technical Specifications; (4) the continued water ingress
problem; and (5) the construction and utilization of
Building 10. A further plant visit was conducted on
August 1-3, 1984, to audit the performance of control-rod
instrumentation in response to observed anomalies.
C. Investigations and Allegations Review
One investigation was conducted during this reporting period and
documented in NRC Inspection Report 50-267/84-11, dated March 20,
1984. This investigation was conducted as a result of the NRC' s
discovery that a contract security employee had not listed two felony
convictions on his personnel security questionnaire. Subsequent to
-36-
this investigation the employee voluntarily informed his employer.
His employment was terminated.
D. Escalated Enforcement Actions
1. Civil Penalties
There were no civil penalties issued during this evaluation
period.
2. Orders
No orders were issued relating to enforcement.
E. Management Conferences
1. Conferences
The following conferences were held during this appraisal
period:
• The ACRS meeting on May 19, 1984, as identified in NRC
Inspection Report 84-14.
• Various licensing related meetings and conferences as
identified in the attached licensing enclosure.
• A June 25, 1984, meeting in Region IV between the NRC
Region IV and PSC to discuss: (1) NRC commitments as a
result of BTP 9.5-1, Appendix A review; (2) Building 10
licensing requirements; (3) NRC notification of control-rod
failure to scram; (4) documentation of disciplinary
actions; (5) clearance tags; (6) radiological emergency
exercise scenario; and (7) temporarily installed controller
in LN2 System.
• A February 19, 1985, meeting in Region IV between the NRC
Region IV and PSC to discuss matters related to management,
security, and the offsite emergency warning system.
2. Confirmation of Action Letters (CAL)
Three CALs were issued during this appraisal period:
• On October 13, 1983, a CAL was issued confirming
commitments made in an Enforcement Conference on October
12, 1983, regarding diesel generator test requirements.
(Identified in previous SALP report) .
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• On June 26, 1984, a CAL was issued confirming commitments
made in a meeting of June 25, 1984, regarding the failure
of six control-rod pairs to properly insert on receipt of a
trip signal on June 23, 1984. (Refer to NRC Inspection
Report 84-15)
• On November 2, 1984, a CAL was issued confirming
commitments made in a telecommunication on October 25,
1984, regarding key control .
F. Review of Licensee Event Reports and 10 CFR Part 21 Reports Submitted
by the Licensee
1. Licensee Event Reports (LERs)
The SALP Board reviewed the LERs submitted by the Public Service
Company of Colorado for the period of October 1, 1983, through
February 28, 1985. This review included the following LERs:
• 83-041 through 83-055
• 84-001 through 84-014
The NRC Office for Analysis and Evaluation of Operational Data (AEOD)
performed reviews of licensee LERs, concentrating on the technical
accuracy, completeness, and clarity of the event reports. Refer to
Attachment 2 and 3 for details of those reviews.
2. Part 21 Report
One 10 CFR Part 21 report was made by the licensee on July 30, 1984,
as documented in NRC Inspection Report 84-22. The report concerned
defective 5/8-inch diameter threaded rod purchased under
specification ASTM A193, 8-16, from Texas Bolt Company of Houston,
Texas.
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TABLE 1
INSPECTION ACTIVITY AND ENFORCEMENT
FUNCTIONAL VIOLATIONS SEVERITY LEVELS DEVIATIONS
AREA V IV III II I
A. Plant Operations 1 3
B. Radiological Controls 3 2 2
C. Maintenance 1 2
D. Surveillance 1
E. Fire Protection 1 2
F. Emergency Preparedness 1
G. Security Safeguards 2
H. Refueling
I. Training
J. Licensing Activities
K. Quality programs and
Administrative Controls
Affecting Quality 2 8 1
L. Design, Design Changes ,
and Modifications 3 5 1
TOTAL 10 25 6
Docket No. 50-267 Attachment 1
FACILITY: Fort St. Vrain Nuclear Generating Station
LICENSEE: Public Service Company of Colorado
EVALUATION PERIOD: October 1, 1983 to February 28, 1985
PROJECT MANAGER: Philip C. Wagner
I . INTRODUCTION
This report contains an assessment of the licensee' s performance in the
functional area of licensing activities as input to the Systematic
Assessment of Licensee Performance (SALP) review for the Fort St. Vrain
Nuclear Generating Station (FSV) . The assessment of the licensee' s
performance was conducted according to NRR Office Letter No. 44, NRR
Inputs to SALP Process, dated January 3, 1984. This Office Letter
incorporates NRC Manual Chapter 0516, SALP.
II . SUMMARY
NRC Manual Chapter 0516 specifies that each functional area evaluated will
be assigned a performance category (Category 1, 2, or 3) based on a
composite of a number of attributes. The performance of the Public
Service Company of Colorado in the functional area of Licensing Activities
is rated Category 3.
III. CRITERIA
The evaluation criteria used in this assessment are given in NRC Manual
Chapter 0516 Appendix, Table 1, Evaluation Criteria with Attributes for
Assessment of Licensee Performance.
IV. METHODOLOGY
This evaluation represents the integrated inputs of the Operating Reactor
Project Manager (ORPM) and those technical reviewers who expended
significant amounts of effort on the Fort St. Vrain licensing actions
during the current rating period. Using the guidelines of NRC Manual
Chapter 0516, the ORPM and each reviewer applied specific evaluation
criteria to the relevant licensee performance attributes, as delineated in
Chapter 0516, and assigned an overall rating category (1, 2, or 3) to each
attribute. The reviewers included this information as part of Safety
Evaluation Reports transmitted to the Division of Licensing. The ORPM,
after reviewing the inputs of the technical reviewers, combined this
information with his own assessment of licensee performance and, using
appropriate weighting factors, arrived at a composite rating for the
licensee. This rating also reflected the comments of the NRR Senior
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Executive assigned to the Fort St. Vrain SALP assessment. A written
evaluation was then prepared by the 0RPM and circulated to NRC management
for comments which were incorporated in the final draft.
The basis for this appraisal was the licensee' s performance in support of
licensing actions that were either completed or had a significant level of
activity during the current rating period. These actions, consisting of
amendment requests, exemption requests, relief requests, responses to
generic letters, TMI items, and other actions, are classified as follows:
- 11 Multi-Plant Actions (4 completed). Included in this category are:
• Radiological Effluent Technical Specifications, A-02, Complete
• Snubbers, B-17 and B-22, Complete
• Control of Heavy Loads, Phase I , C-10, Complete (later reopened)
• Technical Specifications Affected by Changes to 50.72 and 50.73,
A-18, Complete
• Environmental Qualification (10 CFR 50.49) , B-60
• Masonry Wall Design (IEB 80-11) , B-59
• Instrumentation to Follow Accidents (RG 1.97) , A-17
• Control of Heavy Loads, Phase II, C-15
• Responses to G.L. 83-28, B-76 through B-79, and B-85 through
B-93
• NUREG-0737 Technical Specifications (G.L. 83-36 and 37), B-83
• Diesel Generator Reliability Technical Specifications
(G.L. 84-15) , D-19
- 35 Plant Specific Actions (17 Completed) . Included in this category
are:
• Administrative Controls Technical Specifications, Complete
• CRDM Hot Spots, Complete
• PCRV Hot Spots, Complete
• Plateout Probe Data Evaluation, Complete
• Plateout Probe Removal Schedule, Complete
• Steam Generator Tube Leak Investigation, Complete
• Moisture Monitor Relocation, Complete
• Disposal of Sulphur-35, Complete
• Use of H-451 Graphite Fuel Blocks, Complete
• Calibration Sources License Condition, Complete
• Neutron Detector Decalibration, Complete
• Moisture Monitor LCI , Complete
• Fuel Segment 2 Report Review, Complete
• Security Plan Revision 14, Complete
• Secondary Coolant Activity TS, Complete
• Steam Generator Tube TS, Complete
• Fire Hose Station Numbers, Complete
-3-
• Instrument Setpoints
• Fuel Block Crack Investigation
• Electrical Modifications and Technical Specifications
• Startup Test Program Changes
• LCO 4.1.9 Non Conservatisms
• ISI/IST Technical Specifications
• Appendix R Review
• PCRV Tendon Problems
• Changes to Circulator Overspeed Trip
• Seventy Percent Moisture Monitor Tests
• Building 10 Evaluation
• Restart Issue (October 16, 1984 Report)
• Control of Heavy Loads, Phase I Reevaluation
• Technical Specifications Upgrade Program
• Liquid Effluent Release Monitors
• Organizational Changes
• Security Plan Reevaluation
• Circulator Operability Technical Specifications
12 TMI (NUREG-0737) Actions (2 Completed) . Included in this category
are:
• Shift Manning, F-02, Complete
• High Range Radiation Monitor, F-22, Complete
• Inadequate Core Cooling Guidelines, F-04
• Abnormal Transient Operator Guidelines, F-05
• Post Accident Sampling, F-12
• Noble Gas Monitor, F-20
• Technical Support Center, F-63
• Operational Support Center, F-64
• Emergency Operations Facility, F-65
• Meteorological Data Upgrade, F-68
• Detailed Control Room Design Review, F-08 and F-71
• Safety Parameter Display System, F-09
V. ASSESSMENT OF PERFORMANCE ATTRIBUTES
The licensee' s performance evaluation is based on consideration of seven
attributes as specified in NRC Manual Chapter 0516. For most of the
licensing actions considered in this evaluation, only three of the
attributes were of significance. Therefore, the composite rating is
heavily based on the following attributes:
- Management Involvement and Control in Assuring Quality
• Approach to Resolution of Technical Issues from a Safety Standpoint
- Responsiveness to NRC Initiatives.
-4—
With the exception of Enforcement History, for which there was no basis
within NRR for evaluation, the remaining attributes of
Reporting and Analysis of Reportable Events
▪ Staffing (Including Management)
- Training and Qualification Effectiveness
were judged to apply only to a few licensing activities.
A. Management Involvement and Control in Assuring Quality
There were numerous instances during this assessment period in which
management involvement has not been apparent. There has been a lack
of corporate management involvement in site activities as evidenced
by the infrequent plant tours taken by these individuals and the poor
general housekeeping practices which existed prior to NRC special
inspections. Both management visibility and plant housekeeping
practices have been addressed by PSC and a significant plant cleanup
effort has been undertaken.
There have also been numerous instances when requested information
has not been submitted in a timely and thorough manner which
necessitated further questioning and, therefore, a delay in
completion of the given issue. An example of the delayed issues is
the resolution of NUREG-0737, Item II . F.1. 1 "Noble Gas Accident
Monitoring Instrumentation" which has required numerous pieces of
correspondence but is yet to be finalized. When NRC questioned the
acceptability of the installed "semiportable" monitor (NRC letter
dated April 3, 1984) , PCS responded that an evaluation of obtaining a
monitor of an acceptable range would be performed (PSC letter dated
May 3, 1984). PSC' s July 2, 1984 letter stated that their present
plans called for dilution of the sample stream to the existing
monitor with an additional clarification of the proposed installation
date made by letter dated July 12, 1984. NRC' s October 24, 1984
letter agreed that dilution was a practical approach but requested
that design information be provided together with an implementation
schedule. PSC responded to this request by letter dated November 28,
1984 that an evaluation of a new higher range and testable monitor
was to be undertaken and that NRC would be advised of either the new
monitor or the dilution system.
Another area of apparent weakness in management involvement relates
to the number of procedural violations which have occurred during
this reporting period. The specific violations are discussed
elsewhere in this report but are mentioned here because we believe
comments from the licensing organization are worthwhile. A specific
example of the type of problem we find is presented in LER 84-010
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wherein incorrect radioactive effluent monitoring occurred twice
during a single release and little corrective action was proposed.
In addition, the problem of coordination within PSC, which was
mentioned in both the 1982 and 1983 SALP reports, continues to exist.
This situation may be improved, however, by the recent formation of
the Nuclear Licensing and Fuel Division.
Based on the above considerations, a rating of Category 3 has been
assigned to this attribute.
B. Approach to Resolution of Technical Issues From a Safety Standpoint
The frequent delays in obtaining complete and thorough responses to
NRC questions and concerns, due to the positions taken by PSC, tends
to lower the rating for this performance element. The previous
two SALP Reports made mention of the uniqueness of the facility as
being a contributing factor in PSC' s ability to respond to NRC
requests (which are usually directed to Light Water Reactors) . While
this remains a factor, it appears that insufficient effort has been
expended by PSC in order to ensure timely resolution of issues. This
problem was discussed in some detail in the NRC' s, October 16, 1984,
assessment report related to PSC' s response to Generic Letter 83-28.
PSC had responded that because the plant is a one of a kind design,
the need for vendor interface does not exist. The NRC had proposed
the establishment of periodic communication with vendors to help
provide assurance of continued reliable equipment operation; PSC
chose not to establish a program without attempting to understand its
purpose or advantages.
PSC has, at times, delayed the completion of commitments when, in
their opinion, the requirement was not necessary. Examples of this
problem are the completion of the B-0 startup tests for Xenon
stability and full load rejection (or the arrangements necessary for
their completion) and the delay in establishing an enforceable fuel
element surveillance program. Both of these issues have required
numerous correspondence and neither has been resolved.
On the basis of the above indications, a rating of Category 3 is
assigned to this attribute.
C. Responsiveness to NRC Initiatives
While the uniqueness of the facility was mentioned in the previous
two SALP reports as also being a contributing factor in PSC' s ability
to promptly respond to NRC initiatives, a review of the evaluation
attributes contained in NRC Manual Chapter 0516 indicates that the
thoroughness and acceptability of responses should be considered. A
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review of the correspondence associated with numerous licensing
actions (e.g. NUREG-0737 items, Electrical System Technical
Specifications, Startup Testing Program, Appendix R, PCRV Tendon
Surveillance) indicates that responses frequently lack sufficient
detail to reach a conclusion, thereby requiring supplemental requests
and repeated submittals.
An example of PSC taking a nonconservative approach to an NRC
initiative, which has resulted in undue delays in reaching resolution
of a safety issue is the response to the control of heavy loads. PSC
took a literal interpretation of the definition of a heavy load and
defined a load as being 165 tons rather than the approximately one
ton load envisioned by the NRC. This has caused a complete
reevaluation to be required and could have been avoided by a
conservative approach which requested NRC clarification.
Another example of PSC taking the position that the uniqueness of
their design should relieve the facility from compliance with
requirements is in the area of fire protection. Although PSC had
provided a statement that the facility complies with the requirements
of Appendix R (by letter dated November 20, 1981) , an inspection in
August 1983 disclosed numerous deficiencies. Following numerous
contacts and discussions, PSC requested a general exemption to the
requirements of Appendix R, Items III.G, III .J, and III.L by letter
dated March 2, 1984. The NRC stated that the general exemptions were
not acceptable and provided further guidance. Three meetings have
been held to discuss the issues and numerous pieces of correspondence
have been exchanged. PSC has provided the first three of four
reports to ensure compliance of Appendix R requirements by letters
dated November 17 and December 17, 1984, and January 17, 1985. A
great deal of time could have been saved and compliance with the
requirements achieved much sooner had PSC sought a clearer
understanding of the issues and taken a conservative approach toward
resolution. In addition, the recommendation, contained in the past
two SALP Reports, that PSC be more assertive in keeping abreast of
NRC policy matters, does not appear to have been implemented.
Based on the above observations, a rating of Category 3 is assigned
to this attribute.
D. Enforcement History
No basis exists for a licensing evaluation of this attribute.
-7-
E. Reporting and Analysis of Reportable Events
This attribute is addressed as a separate functional area elsewhere
in the SALP report. We have reviewed the functional area and agree
with the conclusions.
F. Staffing
The delays encountered in obtaining an acceptable resolution to
various NRC initiatives as discussed in the preceding areas is
indicative of a marginally acceptable staff size. This same concern
was expressed in the two previous SALP reports. While the increase
in the number and significance of the problems encountered by PSC
during this assessment period has had an effect on their ability to
promptly respond to numerous NRC initiatives, delays are encountered
in almost all areas. PSC has established a Licensing Division which
should improve coordination of PSC efforts in dealing with the NRC,
but may not affect the delays being experienced if the technical
staff is not expanded.
Based on the above considerations, a rating of Cateogry 3 is assigned
to this attribute.
E. Training and Qualification Effectiveness
During this assessment period, two Senior Reactor Operator
replacement retake examinations were administered with both
candidates passing. In addition, the Fort St. Vrain Requalification
Program was evaluated and found to be acceptable. The number of
procedural violations and the presented inability to determine the
applicability of some light water reactor requirements, however,
indicates possible deficiencies in the overall training program. The
area of training is discussed as a separate functional area elsewhere
in this report.
Based on these limited observations, a rating of Category 2 is
assigned to this attribute.
VI. Conclusion
Based on our evaluation of the attributes reviewed above, PSC performance
in the Licensing Activities functional area has been determined to be
Category 3.
VII . Recommendations
In order to improve performance in the Licensing Activities function area,
the following recommendations should be considered:
-8-
1. PSC nuclear department staffing levels should be reevaluated with
consideration given to including personnel knowledgeable of light
water reactor operations;
2. A program should be implemented that will keep PSC management
informed of current NRC initiatives and how those initiatives could
affect Fort St. Vrain; and
3. PSC should implement a policy of providing complete and candid
responses to NRC requests.
Supporting Data and Summary
1. Licensing Related Meetings*
January 17, 1984 SALP Meeting (Site)
February 29, 1984 Environment Qualification (Headquarters)
April 4, 1984 Cracked Fuel Problem (Headquarters)
May 2, 1984 Property Insurance (Headquarters)
May 17, 1984 ACRS (Site)
June 8, 1984 Appendix R (Headquarters)
July 9-11, 1984 Assessment Team (Site)
August 1-3, 1984 Assessment Team (Site)
August 30, 1984 Project Manager and PSC Licensing (Denver)
August 31, 1984 Project Manager and Plant Staff (Site)
October 1, 1984 Control Rod Problems (Site)
November 28 and 29, 1984 Control Rod Problems (Site)
December 12 and 13, 1984 Masonry Walls (Denver and Site)
January 15, 1985 Restart Issues (Region IV)
January 31, 1985 Appendix R (Headquarters)
February 20-22, 1985 Restart Issues (Site)
2. Site Visits*
May 21, 1984 Commissioner Gilinsky made a general plant
tour.
June 4 and 5, 1984 Licensing personnel from Region IV toured
facility to evaluate comments made by
Commissioner Gilinsky and to discuss the
situation with PSC
July 10 and 11, 1984 PM and others met with PSC personnel and
reviewed various in-progress maintenance
activities
*All meetings held at the site and each of the above site visits included a
general plant tour including the reactor building, turbine building, and the
control room by the project manager and others.
3. Commission Briefings
None.
4. Schedular Extensions Granted
None. However, an extension for the Emergency Exercise was requested.
When the exercise was postponed, a rule change made the request moot and
it was withdrawn.
5. Reliefs Granted
None.
-2-
6. Exemptions Granted
None. The Exemption Request for 10 CFR 50.54(w) , "Property Damage
Insurance," has been withdrawn.
7. License Amendments Issued
Amendment No. 36 Administrative Controls, Staffing, and STA,
October 13, 1983
Amendment No. 37 Radiological Effluent Technical Specifications,
November 23, 1983
Amendment No. 38 Plateout Probe Removal Schedule, January 3, 1984
Amendment No. 39 Snubbers-Hydraulic and Mechanical , January 25,
1984
Amendment No. 40 Use of H-451 Graphite, March 2, 1984
Amendment No. 41 Calibration Source Size Changes, March 8, 1984
Amendment No. 42 LER Rule Change to Technical Specifications,
June 4, 1984
Amendment No. 43 Moisture Monitor Changes to Technical
Specifications, June 5, 1984
Amendment No. 44 Secondary Coolant Activity Surveillance
Requirements, October 26, 1984
Amendment No. 45 Steam Generator Tube ISI Requirements,
November 9, 1984
Amendment No. 46 Fire Hose Station Numbering System Change,
January 3, 1985
8. Emergency Technical Specifications Issued
None.
9. Orders Issued
None.
-3-
10. Licensee Management Conferences
November 9, 1984 EDO, Director NRR and Administrator, Region IV
met with CEO and Vice President of PSC on Conduct
of Operations
``Epp AEC��q Attachment 2
UNITED STATES
a .I y, o
. NUCLEAR REGULATORY COMMISSION
WASHINGTON,D.C.20555
O ao
**Irk*
N0V 5 1984
MEMORANDUM FOR: Eric H. Johnson, Chief
Reactor Project Branch No. 1
Division of Resident, Reactor Project
and Engineering Programs, RIV
FROM: Karl V. Seyfrit, Chief
Reactor Operations Analysis Branch
Office for Analysis and Evaluation
of Operational Data
SUBJECT: EVALUATION OF LERs FOR FORT ST. VRAIN
AEOD INPUT TO SALP REVIEW COVERING THE
PERIOD FROM OCTOBER 1 , 1983, TO NOVEMBER 30, 1984
In support of the ongoing SALP reviews, AEOD has reviewed the LERs for
Fort St. Vrain. Our review concentrated on LER Form completeness and
the clarity, understandability, and adequacy of the event report contents.
From the LERs that were reviewed, we concluded that the licensee provided
adequate event reports during the assessment period. We found no signifi-
cant deficiencies and the reports complied with the guidelines of NUREG-0161
and NUREG-1022 in all reviewed categories.
The enclosure provides additional observations from our review of the LERs.
If you should have any questions regarding this report, please contact
either myself or Ted Cintula of my staff. Mr. Cintula can be reached at
FTS 492-4494.
v .22
/Karl V. Seyfr' , Chief
Reactor Operations Analysis Branch
Office for Analysis and Evaluation
of Operational Data
Enclosure:
As stated
cc: w/enclosure:
T. Colburn, NRR
G. L. Plumlee, RIV
AEOD INPUT TO SALP REVIEW FOR FORT ST. VRAIN
The Licensee submitted about 35 reports, plus updates, during the assessment
period from October 1 , 1983 to November 30, 1984. Our review included the
following LER numbers:
83-041 to 83-055
84-001 to 84-009
The LER review followed the general instructions and procedures of NUREG-0161
and NUREG-1022. The specific review criteria and our findings follow:
1 . LER Completeness
a) Was the information sufficient to provide a good understanding
of the event?
1983 LERs
The information in the two free-form narrative sections of the
LER Form was consistently brief and to the point. There were
a few instances of overrunning narratives, but they were of small
magnitude and would not be a problem for future abstracting. Our
review concluded the LERs provided sufficient information to
provide a clear and adequate description of the occurrence, the
direct consequences and the corrective action. The reports typical -
ly included specific details of the event such as valve identifica-
tion numbers, model numbers, number of operable redundant systems ,
the date of completion of repairs, etc. , to provide a good under-
standing of the event. The reports were easy to read and meaning-
ful .
1984 LERs
The abstract described the major occurrences of the event, including
all component or system failures that contributed to the event and
the significant corrective actions taken or planned to prevent
recurrence as stated in NUREG-1022.
b) Were the LERs coded correctly?
1983 LERs
We checked the codes that the licensee selected against the narra-
tive description of the event for accuracy. We agreed with the
licensees selection in all coded fields except for a few entries.
These disagreements were minor and did not detract from our overall
impression of a judicious selection of coded information.
- 1 -
1984 LERs
We agreed with the licensee' s selection in all coded fields.
c) Was supplemental information provided when needed?
1983 LERs
The licensee provided additional supplementary information with
every LER. The attachments typically provided plant specific
detailed information such as the limiting condition of operation,
the purpose of the system, all functions performed by the defective
component, etc. , which was useful in assessing the full impact of
the event rather than just a restatement of the original argu-
ments. The attachments were well organized with each topic of
discussion separated and titled. In addition, the licensee typi-
cally provided tabular information and simplified flow schematics
with the supplemental information. These aids greatly assisted in
explaining the event. In view of both the quantity and quality of
the supplemental information, we concluded that the licensee was
outstanding in this category.
1984 LERs
The narrative description in the attachments was very informative
and the new reports provided substantial detail about the events.
The licensee typically stated the purpose of the system and all
functions performed by the defective component. The safety analysis
often assumed the conservative loss of the complete system to describe
a worst case scenario. Some reports included diagrams and tables to
help explain the event and the narratives and diagrams were coded with
symbols so it was easy to follow all system/component interactions
of the event. We thought the supplemental information was excellent.
d) Follow-up Reports
1983 LERs
The licensee positively stated in each LER as to whether the LER would
be updated at some future date or that no further corrective action
was required. However, only one of the promised LERs was actually
updated in this assessment period (LER 83-050) . A review of the data
base showed that thirteen other older LERs were also updated in this
assessment period. A review of these LERs showed the updated reports
contained new narrative information and the codes were revised correctly
in accordance with the guidelines of NUREG-0161 . The portions of the
narratives that were revised were identified by a vertical line in the
left hand margin of the page so the extent of corrected information
was readily apparent to all readers.
- 2 -
1984 LERs
Three of the new reports have been updated so far. They were up-
dated correctly by the standards of NUREG-1022 and the above
comments would be applicable.
e) Were similar occurrences properly referenced?
1983 and 1984 LERs
Previous LER numbers of events of a similar nature were referenced
correctly. In addition, the licensee positively stated when there
have been no previous similar reports.
2. Multiple Event Reporting in a Single LER
The licensee submitted several LERs that combined multiple events of
component failures into a single report. These multiple events were
combined correctly into a single LER in accordance with the guidelines
of NUREG-0161 and NUREG-1022.
3. Prompt Notification Follow-up Reports
Only two PNs were issued in this SALP assessment period, the failure of
six of the 37 control rod pairs to insert on June 22, 1984 and an
excessive Beta radiation liquid release on July 20, 1984. Each of
these events were reportable, and they were reported as LERs 84-008 and
009. Therefore, it appears the licensee is reporting all events that are
required to be reported.
- 3 -
``ERR REGt,4
m° 'o9t UNITED STATES Attachment 3
a` I 5' o
NUCLEAR REGULATORY COMMISSION
m ° 3 WASHINGTON,D.C.20555
3
P'•t++#a
MAR 221986
MEMORANDUM FOR: Eric H. Johnson, Chief
Reactor Project Branch No. 1
Division of Resident, Reactor Project
and Engineering Programs, RIV
FROM: Karl V. Seyfrit, Chief
Reactor Operations Analysis Branch
Office for Analysis and Evaluation
of Operational Data
SUBJECT: EVALUATION OF LERs FOR FORT ST. VRAIN
AEOD INPUT TO SALP REVIEW COVERING THE
PERIOD FROM OCTOBER 1 , 1983, to FEBRUARY 28, 1985
In my memorandum to you dated November 5, 1984 (same subject) we provided
an evaluation of the LERs for Fort St. Vrain for the period October 1 , 1983
through November 30, 1984. In response to Regional Office Notice 0603,
we have evaluated the additional LERs through February 28, 1985. The
findings in our previous evaluation are valid for the extended assessment
period, but with one additional consideration.
We found in the most recent LERs, that the licensee needs to improve the
safety evaluation of the event in the LERs. In general , the analysis of the
event concluded without adequate bases that there was no potential effect on
the health and safety of the public. The evaluation of the safety consequences
and implications should also include the potential effects of the event had
it occurred at different operating conditions and whether it is an analyzed
event.
If you should have any questions regarding our evaluation, please contact either
myself or Wayne Lanning of my staff. Mr. Lanning can be reached at FTS 492-4433.
7i1/7C`i-
arl V. Seyfrit ief
Reactor Operations Analysis Branch
Office for Analysis and Evaluation
of Operational Data
cc: T. G. Colburn, NRR
G. L. Plumlee, R IV
R. E. Ireland, R IV
Hello