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Docket: 50-267 JUL 2 3 1985
Public Service Company of Colorado
ATTN: 0. R. Lee, Vice President
Electric Production
P. 0. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
Based upon the information provided by members of your staff during telephone
conferences on May 21 and 22, 1985, we have completed our review of Item 1.2
"Post Trip Review" contained in Generic Letter 83-28. We had provided our
preliminary review finding by letter dated May 17, 1985; the telephone calls
provided the needed clarification in some areas. The results of our review
are contained in the enclosed Safety Evaluation.
We have concluded that the post trip review data and information capabilities
for the Fort St. Vrain Station are adequate and that Item 1.2 has been
acceptably resolved.
If you have any questions or comments on this matter, please contact us.
Sincerely,
.:, e.
orwin R. Hunter, Chief
Reactor Safety Branch
Enclosure:
Safety Evaluation of
Item 1.2
cc:
See next page
851170
1n A r Ai-
2
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
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 Radiation Control Program Director
0.0110E0u, •
o UNITED STATES
Fes. 2,; NUCLEAR REGULATORY COMMISSION
�� , WASHINGTON,D.C.20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
PUBLIC SERVICE COMPANY OF COLORADO
FORT ST. VRAIN UNIT 1
DOCKET NO. 50-267
GENERIC LETTER 83-28, ITEM 1.2 POST-TRIP REVIEW
AND INFORMATION CAPABILITY)
(DATA
I, INTRODUCTION
On February 25, 1983, both of the scram circuit breakers at Unit 1 of the
Salem Nuclear Power Plant failed to open upon an automatic reactor trip
plantlstart rectorion system. This
tripped manuallyincident
yb the nccurred peratorduring about the
30
seconds after the initiation of the automatic trip signal . The failure of
the circuit breakers has been determined to be related to the sticking of the
under voltage trip attachment. Prior to this incident, on February 22, 1983,
at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was
generated based on steam generator low-low level during plant start-up. In
this case, the reactor was tripped manually by the operator almost
coincidentally with the automatic trip. Following these incidents, on
February 28, 1983, the NRC Executive Director for Operations (EDO), directed
the staff to investigate and report on the generic implications of these
occurrences at Unit 1 of the Salem Nuclear Power Plant. The results of the
staff's inquiry into the generic implications of the Salem unit incidents are
reported in NUREG-1000, "Generic Implications of the ATWS Events at the Salem
Nuclear Power Plant." As a result of this investigation, the Commission
(NRC) requested (by Generic Letter 83-28 dated July 8, 1983) all licensees of
operating reactors, applicants for an operating license, and holders of
construction permits to respond to certain generic concerns. These concerns
are categorized into four areas: (1) Post-Trip Review, (2) Equipment
Classification and Vendor Interface, (3) Post-Maintenance Testing, and
(4) Reactor Trip System Reliability Improvements.
The first action item, Post-Trip Review, consists of Action Item 1.1,
"Program Description and Procedure" and Action Item 1.2, "Data and
Information Capability." This safety evaluation (SE) addresses Action Item
1.2 only.
II. REVIEW GUIDELINES
The following review guidelines were developed after initial evaluation of
the various utility responses to Item 1.2 of Generic Letter 83-28 and
incorporate the best features of these submittals. As such, these review
guidelines in effect represent a "good practices" approach to post-trip
review. The staff has reviewed the licensee's response to Item 1.2 against
these guidelines:
- 2 -
A. The equipment that provides the digital sequence of events (SOE) record
and the analog time history records of an unscheduled shutdown should
provide a reliable source of the necessary information to be used in the
post-trip review. Each plant variable which is necessary to determine
the cause and progression of the events following a plant trip should be
monitored by at least one recorder (such as a sequence-of-events
recorder or a plant process computer) for digital parameters; and strip
charts, a plant process computer or analog recorder for analog (time
history) variables. Performance characteristics guidelines for SOE and
time history recorders are as follows:
° Each sequence of events recorder should be capable of detecting
and recording the sequence of events with a sufficient time
discrimination capability to ensure that the time responses
associated with each monitored safety-related system can be
ascertained, and that a determination can be made as to whether
the time response is within acceptable limits based on FSAR
Chapter 15 Accident Analyses. The recommended guidelines for the
SOE time discrimination is approximately 100 milliseconds. If
current SOE recorders do not have this time discrimination
capability, the licensee should show that the current time
discrimination capability is sufficient for an adequate
reconstruction of the course of the reactor trip and post-trip
events. As a minimum, this should include the ability to
adequately reconstruct the transient and accident scenarios
presented in Chapter 15 of the plant FSAR.
° Each analog time history data recorder should have a sample
interval small enough so that the incident can be accurately
reconstructed following a reactor trip. As a minimum, the
licensee should be able to reconstruct the course of the transient
and accident sequences evaluated in the accident analysis of
Chapter 15 of the plant FSAR. The recommended guideline for the
sample interval is 10 seconds. If the time history equipment does
not meet this guideline, the licensee should show that the time
history capability is sufficient to accurately reconstruct the
transient and accident sequences presented in Chapter 15 of the
FSAR. To support the post-trip analysis of the cause of the trip
and the proper functioning of involved safety related equipment,
each analog time history data recorder should be capable of
updating and retaining information from approximately 5 minutes
prior to the trip until at least 10 minutes after the trip.
° All equipment used to record sequence of events and time history
information should be powered from a reliable and
non-interruptible power source. The power source used need not be
safety related.
- 3 -
B. The sequence of events and time history recording equipment should
monitor sufficient digital and analog parameters, respectively, to
assure that the course of the reactor trip and post-trip events can be
reconstructed. The parameters monitored should provide sufficient
information to determine the root cause of the unscheduled shutdown, the
progression of the reactor trip, and the response of the plant
parameters and protection and safety systems to the unscheduled
shutdowns. Specifically, all input parameters associated with reactor
trips, safety injections and other safety-related systems as well as
output parameters sufficient to record the proper functioning of these
systems should be recorded for use in the post-trip review. The
parameters deemed necessary, as a minimum, to perform a post-trip review
that would determine if the plant remained within its safety limit
design envelope are presented in Table 1 . They were selected on the
basis of staff engineering judgment following a complete evaluation of
utility submittals. If the licensee's SOE recorders and time history
recorders do not monitor all of the parameters suggested in these tables,
the licensee should show that the existing set of monitored parameters
is sufficient to establish that the plant remained within the design
envelope for the accident conditions analyzed in Chapter 15 of the plant
FSAR.
C. The information gathered by the sequence of events and time history
recorders should be stored in a manner that will allow for data
retrieval and analysis. The data may be retained in either hardcopy,
(e.g. , computer printout, strip chart record) , or in an accessible
memory (e.g. , magnetic disc or tape) . This information should be
presented in a readable and meaningful format, taking into consideration
good human factors practices such as those outlined in NUREG-0700.
D. Retention of data from all unscheduled shutdowns provides a valuable
reference source for the determination of the acceptability of the plant
vital parameter and equipment response to subsequent unscheduled
shutdowns. Information gathered during the post-trip review is to be
retained for the life of the plant for post-trip review comparisons of
subsequent events.
III. EVALUATION AND CONCLUSION
By letter dated November 4, 1983, Public Service Company of Colorado provided
information regarding its post-trip review program data and information
capabilities for Fort St. Vrain Unit 1. The staff has evaluated the licensee's
submittal against the review guidelines described in Section II. Licensee
deviations from the Guidelines of Section II were reviewed with the licensee
by telephone on May 21 and 22, 1985. A brief description of the licensee's
responses and the staff's evaluation of the response against each of the
review guidelines is provided below:
- 4 -
A. The licensee's submittal described the performance characteristics of
the equipment used to record the sequence of events and time history
recorder data needed for post-trip review. Based on the staff's review,
the staff finds that the sequence of events and time history recorder
characteristics conform to the guidelines described in Section II A, and
are acceptable.
B. The licensee has established and identified the parameters to be
monitored and recorded for post-trip review. The parameters identified
in the submittal of November 4, 1983, as being recorded by the sequence
of events and time history recorders, did not correspond to the
parameters specified in Section II B of this report. The missing
parameters were: (1) Reactor Building Temperature, (2) RSIV position,
(3) Hot Reheat Steam Pressure, (4) Hot Reheat Steam Temperature, and
(5) Hot Reheat Steam Activity. By telecon on May 21 and 22, 1985 Public
Service Company of Colorado verified that these parameters are recorded
and meet the criteria of Section II B in this report. Based on the staff's
review, the staff finds that the parameters selected by the licensee
include all of those identified in Table 1 and conform to the guidelines
described in Section II B and are, therefore, acceptable.
C. The licensee has described the means for storage and retrieval of the
information gathered by the sequence of events and time history
recorders, and for the presentation of this information for post-trip
review and analysis. Based on the staff's review, the staff finds that
this information will be presented in a readable and meaningful format,
and that the storage, retrieval , and presentation conform to the guidelines
of Section II C.
D. The licensee's submittal indicates that the data and information used
during post-trip reviews would be retained in an accessible manner for
the life of the plant. Based on the staff's review, the staff finds that
the licensee's program for data retention conforms to the guidelines of
Section II D, and is acceptable.
Based on this review, the staff concludes that the licensee's post-trip review
data and information capabilities for Fort St. Vrain Unit 1 are acceptable.
Principal Contributor:
R. G. Ramirez, DHFS
Date: June 27, 1985
_ 5 -
TABLE 1 HTGR POST-TRIP REVIEW PARAMETER LIST
SOE Time History
Recorder Recorder (2) Parameter/Signal
x Reactor Trip
x Turbine Trip
x Control Rod Position
x (1) x Neutron Flux, Power
x (1) x Reactor Building Temperature
x Reactor Building Radiation
x (1) x Primary System Pressure
x Core Region Outlet Temperature
x (1) x Circulator Inlet Temperature
x (1) Circulator Speed
x (1) Circulator Seal Malfunction
x (1) Circulator Bearing Water Malfunction
x (1) Circulator Drain Malfunction
x (1) Circulator Penetration Overpressure
x Primary System Flow
x (1) x Primary System Moisture
x MSIV Position
x RSIV Position
x (1) x Main Steam Pressure
x (1) x Main Steam Temperature
x (1) x Hot Reheat Steam Pressure
x (1) x Hot Reheat Steam Temperature
x (1) Hot Reheat Steam Activity
x (1) Steam Generator Penetration
Overpressure
x Feedwater Flow
x Steam Flow
x Emergency Auxiliary Feedwater System;
Flow, Valve Status
_ 6 _
SOE Time History
Recorder Recorder (2) Parameter/Signal
X AC and DC System Status
X Diesel Generator Status
(1) Trip parameters
(2) Parameter may be monitored by either an SOE or time history recorder
MEQ(i
O`'` 4? UNITED STATES
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NUCLEAR REGULATORY COMMISSION el! III n? "n .dn gp
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JUL 26 ig85
JUL 23 1985 L�' JI
Docket: 50-267 GREELEY. COLO.
Public Service Company of Colorado
ATTN: 0. R. Lee, Vice President
Electric Production
P. 0. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
We have reviewed the documents you have provided related to the construction
of Building 10 and its associated walk-over structure. We find that
additional information is needed for us to continue our review. Therefore, we
request that you provide the information requested in the enclosure within 30
days of your receipt of this letter.
If you have any questions on this request, please contact the NRC Project
Manager, P. Wagner, at (817)860-8127.
Since this reporting request relates solely to the Fort St. Vrain Station, OMB
clearance is not required under P.L.96-511.
Sincerely,
9orw40-n-P;-/E4
in R. Hunter, Chief
Reactor Safety Branch
Enclosure:
Request for Information
cc:
See next page
2
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
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 Radiation Control Program Director
REQUEST FOR ADDITIONAL INFORMATION
FOR REVIEW OF BUILDING 10 AT FORT ST. VRAIN
The following information is needed before an evaluation of Building 10 and
the Walkover Structure can be completed:
1. Engineering drawings showing the Building 10 and Walkover Structure layout
(plans and elevations) and its spatial relationship with the surrounding
buildings.
2. A comprehensive list of the structural codes and revision dates
including the section of the code used in the analysis. Several
versions of the codes are referenced in the enclosure. Their specific
use in the analysis should be clarified.
3. The attachment V referenced in the document CN 1460 page 89 was not
included in the documents provided.
4. Provide the results of the seismic analysis including, but not limited
to, the displacements of Building 10, the Walkover Structure and the
surrounding structures.
5. Provide the effect of the tornado depressurization loads on the
structure along with the effect of the tornado missile spectrum.
1,0 RFGu,_r UNITED STATES Lyt q tool rnvonot{f 7S
G9. NUCLEAR REGULATORY COMMISSION
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6REr LEY. COLD.
In Reply Refer To:
Docket: 50-267
Public Service Company of Colorado
ATTN: 0. R. Lee, Vice President
Electric Production
P. 0. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
The Safety Evaluation enclosed in our July 12, 1985, letter contained some
issues for which NRC review was not complete. We have now completed our
review; the results are contained in the enclosed Supplemental Safety
Evaluation. We have concluded that all of the open items related to our
July 12, 1985, evaluation have been acceptably resolved.
The commitments made by Public Service Company of Colorado, as discussed in
the enclosure, are being included in the listing of commitments being
confirmed in relation to the authorization of Fort St. Vrain restart.
Sincerely,
E. H. Johnson, Chief
Reactor Project Branch
Enclosure:
Supplemental Safety Evaluation
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
(cont. on next page)
7/N/sI
Public Service Company of Colorado -2-
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 193
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
Colorado Radiation Control Program Director
SUPPLEMENTARY SAFETY EVALUATION
BY THE NUCLEAR REGULATORY COMMISSION
PUBLIC SERVICE COMPANY OF COLORADO
FORT ST. VRAIN NUCLEAR GENERATING STATION
DOCKET NO. 50-267
CONTROL ROD DRIVE AND POSITION INDICATION
I. INTRODUCTION
The NRC provided the results of the reviews of various Assessment Report
issues in a Safety Evaluation (which included three additional evaluations
as attachments) transmitted to Public Service Company of Colorado (PSC or
the licensee) by letter dated July 12, 1985. Included in the issues
discussed were control rod drive mechanisms (SE dated May 21, 1985) , and
the CRDM instrumentation (SE dated May 17, 1985). There were areas of
concern, in both of the above issues, for which our review was not
completed at the time of issuance of the SE. These areas of concern have
now been resolved.
II. EVALUATION
1. Supplemental Evaluation of Control Rod Drive Mechanisms
As discussed in our initial safety evaluation on the Control Rod
Drive Mechanisms dated May 21, 1985, the staff required the
following actions by the licensee in order to conclude that the CRDMs
were acceptable for plant restart. These are:
o The licensee must provide a commitment to operate the plant
within the CRDM temperature limits accepted by the NRC. The
temperature limits cannot be changed without NRC approval of new
temperature limits or alternative methods of assuring CRDM
operability; and
o The licensee must provide a commitment to submit an improved
CRDM surveillance and preventative maintenance program within 6
months of plant restart.
By letters dated June 7, 1985 (P-85180) and July 3, 1985 (P-85233),
the licensee submitted draft Technical Specifications for control
rods. Technical Specification 3.1.1 in these submittals was a
limiting condition for operation requiring that the CRDM motor
temperature be less than or equal to 250°F. Additionally, Technical
Specification 4.1.1 (P-85223) is a surveillance requirement for
control rod operability. Section 4.1.1.A.1.a of this specification
requires the licensee to record CRDM motor temperatures above 215°F
and report the results to the NRC on a monthly (31 day) basis.
-2-
Additional clarification was provided in the resubmittal dated
July 10, 1985 (P-85242) which specifies how the CRDM temperatures
will be monitored.
The licensee has committed (P-85180 and P-85242) to operate the
facility in accordance with procedures based upon these draft interim
Technical Specifications until formal Technical Specifications are
approved and implemented. The operating procedures that implement
these Specifications would be in place prior to plant restart. By
letter dated June 14, 1985 (P-85199) , the licensee committed to
submit an improved CRDM surveillance and preventative maintenance
program within 6 months of plant restart.
We find these commitments are acceptable to close the open issues
from our previous evaluation, and that the CRDMs are acceptable for
plant restart.
2. Supplemental Evaluation of Control Rod Instrumentation
As discussed in our May 17, 1985 Safety Evaluation on Control Rod
Position Instrumentation, all identified items were found to be
acceptable with the exception of the following:
o Confirmation of Cautions on Control Rod Overdrive
o Additional Technical Specifications on Control Rod Position
Instrumentation Operability
o Additional Surveillance Tests on Control Rod Position
Instrumentation
o Initiation of the Reserve Shutdown System
o Wattmeter Testing
By letter dated July 10, 1985, the licensee provided interim
Technical Specifications for Reactivity Control . The licensee
committed to operate the plant under procedures based on these
interim Technical Specifications until formal Technical
Specifications are approved and implemented. By letter dated
July 18, 1985, the licensee provided confirmation that all procedures
involving possible control rod overdrive contained appropriate
cautions and controls. We have evaluated the licensee's responses to
the above items as follows:
Confirmation of Cautions on Control Rod Overdrive
In Section 3.2 of our May 17, 1985 evaluation, we found that the
licensee has modified Procedure SOP 12-1 to preclude manual inward
overtravel of rods following a reactor scram. It was not clear that
this procedure covers all conditions in which manual inward motion of
-3-
a control rod could occur. By letter dated July 18, 1985, the
licensee confirmed that all procedures in which rod travel beyond the
full-in limit could occur contained appropriate cautions and controls
to reduce the likelihood of damage to potentiometers and their
couplings. Therefore, we find that the licensee has provided an
acceptable response to this concern.
Additional Technical Specifications on Control Rod Position
Instrumentation Operability
The licensee has proposed additional interim Technical Specifications
to address the staff's concerns on operability of the Control Rod
Position Instrumentation. Section 3.1.2 of these Technical
Specifications is consistent with the Limiting Conditions for
Operation described in Section 3.3 of our May 17, 1985 evaluation and
are therefore acceptable.
Additional Surveillance Tests on Control Rod Position Instrumentation
The licensee has proposed additional surveillance requirements to
address the staff's concerns on operability of the Control Rod
Position Instrumentation. The combination of the licensee's proposed
surveillance program and the additional surveillance stated in
Section 3.4 of our May 17, 1985, evaluation combine to provide
adequate assurance of RPI operability. The staff desired that the
licensee: (1) verify each full-in limit indication prior to startup
or during the first withdrawal of the rod from the full-in position;
(2) verify the analog and digital position indications prior to
startup or during the first withdrawal of the rod from the full-in
position; and (3) verify reasonableness of the analog and digital
control rod position indications. The proposed Technical
Specifications (Section 4.1.2) are consistent with the desired
concepts as described in Section 3.4 of our May 17, 1985, evaluation,
and are therefore acceptable.
However, we note that Section 3.4 of our evaluation acknowledges a
quarterly surveillance of the rod-out limit indication, based on the
licensee's proposal of January 31, 1985 (P-85040). The proposed
Technical Specifications are silent concerning a periodic check of
the out-limit indication. The out-limit indication operability
should be verified on a periodic basis, and can be accomplished in
conjunction with other tests (such as the partial scram test of fully
withdrawn rods required by Technical Specification 4.1.1.B). Plant
restart is acceptable without a specific surveillance requirement on
out-limit indication. However, this surveillance should be
explicitly incorporated in the final Technical Specification upgrade
program.
-4-
Initiation of the Reserve Shutdown System
The licensee has proposed additional Technical Specifications
concerning initiation of the reserve shutdown system. As specified
in Section 3.5 of our evaluation dated May 17, 1985, the licensee's
backup shutdown procedure must be revised to assure adequate shutdown
and certain verifications of rod positions shall be accomplished
following each reactor scram. The proposed Technical Specifications
(Section 4.1.6) provide consistent requirements for verification of
rod position following a reactor scram, and are therefore acceptable.
Section 3.1.6 of the proposed Technical Specifications requires
insertion of reserve shutdown material if within 1-hour, more than
two control rods are not verified to be fully inserted. The staff
requirement was to insert reserve shutdown material if more than one
control rod cannot be verified to be fully inserted. The analysis of
Section 3.5.3.1 of the Updated Final Safety Analysis Report
demonstrates that adequate shutdown margin exists with two rods
withdrawn and that the requirements of Technical Specification 3.1.4
(Shutdown Margin) can be met. Consequently, proposed Technical
Specification 3.1.6 is acceptable.
Wattmeter Testing
In Section 3.6 of the May 17, 1985 evaluation, the staff concluded
that the wattmeter test proposed by the licensee is an adequate
method of verifying the rod full-in position provided that the ease
of interpretation of the data is increased and the level of reliance
on judgment and interpolation is decreased through the use of a more
appropriate choice of wattmeter range and recorder speed. Although
the staff expresses some concern over the licensee's method and
procedures, this item requires no further action prior to facility
restart. However, the licensee should confirm that they have
addressed this recommendation as a long term item.
III. CONCLUSION
The staff has reviewed the interim Technical Specifications submitted by
the licensee by letter dated July 10, 1985 (P-85242) , and the additional
confirmation concerning control rod overdrive procedures submitted by
letter dated July 18, 1985, and finds that these submittals address the
concerns in the staff's SE dated July 12, 1985. We conclude that the
control rod drive mechanisms and associated position instrumentation is
acceptable for plant restart and continued operation.
-5-
The staff has identified two items which should be pursued as long-term
improvements in Technical Specifications and Plant procedures. These
items are a specific surveillance test for the control rod out-limit
indication and improvements to the wattmeter tests.
Date: July 19, 1985
Principal Contributor: K. Heitner
RFGU UNITED STATES
``EpP 4 a
NUCLEAR REGULATORY COMMISSION
a▪ t I� ° REGION IV
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° y 611 RYAN PLAZA DRIVE,SUITE 1000
r,• •o e`^"`` No ARLINGTON,TEXAS 75011 WELD COUNTY COMMISSIONERS
JUL 19 1985 _
Docket: 50-267 JUL 2 41985
Mr. 0. R. Lee, Vice President
Electric Production
Public Service Company of Colorado
P.O. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
We have reviewed the various submittals you have made in response to our
October 16, 1984, "Preliminary Report Related to the Restart and Continued
Operation of the Fort St. Vrain Nuclear Generating Station," and other submittals
related to subjects which have required resolution prior to the resumption of
operations at Ft. St. Vrain (FSV). We have provided the results of our reviews
of the various areas of concern in a number of recently issued Safety Evaluations
(SEs). These safety evaluations contain the NRC findings on all issues
associated with the above report. In addition, we have approved some of your
proposals for resolving some of our concerns which will require long-term
actions. A listing of our SEs and approvals is contained in Enclosure 1, a
listing of your various commitments is contained in Enclosure 2.
Our June 26, 1984, confirmatory action letter (CAL) confirmed your commitment
that FSV would be maintained in a shutdown condition until the NRC authorized a
different status. Our reviews have now progressed to such a point that we have
determined that you have satisfied the commitments contained in our June 26,
1984, letter and have addressed the issues raised in the October 16, 1984,
report. We find it acceptable, therefore, for FSV to be operated in a
"dry-out" mode to aid in the removal of moisture. Such an operation would hold
reactor power below a level at which boilout in the economizer-evaporator-
superheater section of the steam generator would be achieved, but in no case
greater than 15% Rated Thermal Power until such time as certain equipment
qualification questions are resolved. Our evaluation of the acceptability of
operating FSV in this mode is contained in the enclosed Safety Evaluation
(Enclosure 3).
In addition, we have reviewed your proposed compensatory measures in the area
of fire protection, contained in your letter dated July 11, 1985, (P-85245) and
find them to be an acceptable basis to allow interim operations until our
technical review is completed. We have confirmed your commitments in the
listing contained in Enclosure 2; a SE of our finding will be provided under
separate cover.
In authorizing the restart and the operation of Fort St. Vrain at power levels
not to exceed 15% power, we understand that you have committed to the following
actions:
.,,,TG -11 /8S
Public Service Company of Colorado -2-
1. Unless authorized by NRC to proceed to a higher power level , PSC will
restrict plant power level to no greater than 15% power. PSC shall notify
the NRC of the completion of aging and equipment operability studies for
equipment qualification under 10 CFR 50.49.
2. PSC will continue to implement those items listed as complete and will
resolve the longer range issues contained in Enclosure 2. In doing so,
the scope and schedule for resolution of these issues, that has been
previously agreed to with the NRC staff, shall not be altered without
prior agreement of the NRC staff.
If your understanding of these commitments is not the same as ours, please
contact our office within 24 hours at (817) 860-8100.
Since these reporting requirements relate solely to the FSV, OMB clearance is
not required under P.L. 96-511.
Sincerely,
/,o ei 7`
/ il)
7
Robert D. Martin
Regional Administrator
Enclosures:
1. List of SEs and Approvals
2. List of PSC Commitments
3. SE Related to Environmental Qualifications
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
(cont. on next page)
Public Service Company of Colorado -3-
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 Radiation Control Program Director
Enclosure 1
LIST OF SAFETY EVALUATION/APPROVALS ISSUED
RELATED TO THE RESTART OF FORT ST. VRAIN
Date Subject
July 8, 1985 Prestressed Concrete Reactor Vessel
Tendon Wire Corrosion Problem SE
July 9, 1985 Liquid Effluent Release from the Reactor
Building Sump.
July 10, 1985 Emergency Electrical Power Systems SE
July 12, 1985 Assessment Report SE (CROM, Technical
Specifications, and Management Controls)
July 12, 1985 CRDM Bearing SE letter
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,�EPp NEGL/(qO UNITED STATES Enclosure 3
NUCLEAR REGULATORY COMMISSION
!
w _, 3 REGION IV
-
yA 811 RYAN PLAZA DRIVE. SUITE 1000
'4. ° ARLINGTON.TEXAS 78011
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
PUBLIC SERIVCE COMPANY OF COLORADO
FORT ST. VRAIN NUCLEAR GENERATING STATION
DOCKET NO. 50-267
ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT IMPORTANT TO SAFETY
Background
As part of the ongoing review of licensees' conformance to the requirements
for environmental qualification, the staff raised a number of issues concerning
the adequacy of the Fort St. Vrain program. By letter dated January 28, 1985,
Public Service Company of Colorado (PSC) was requested to provide additional
information in order to determine the extent of compliance with the requirements
of 10 CFR 50.49, the rule on environmental qualification of electric equipment
important to safety. PSC provided a response to that request in a letter dated
March 28, 1985.
Additional information on environmental qualification was provided by the
licensee in a letter dated March 25, 1985 in response to Generic Letter 84-24,
Certification of Compliance to 10 CFR 50.49.
A meeting with the licensee to discuss the status of the Fort St. Vrain
equipment qualification was held in Bethesda, Maryland on April 3, 1985. As a
result of the information provided at this meeting and the licensee's letter
further clarifying information was requested by the staff in a letter dated
May 7, 1985. The response to that request is a letter from PSC dated
June 11, 1985.
On July 2, 1985 staff members from NRR and Region IV again met with the licensee
to discuss the current status of the Fort St. Vrain equipment qualification
program. A letter dated July 11, 1985 was sent by PSC to address followup issues.
Staff Evaluation
A summary of the results of the staff review and evaluation of the information
provided follows.
10 CFR 50.49 requires each holder of an operating license to establish a
program for qualifying electric equipment important to safety. The program
must include and be based on the time-dependent temperature and pressure
conditions at the location of the equipment resulting from the most severe
design basis accident during or following which the equipment is required to
remain operable. The program must also include and consider, if applicable,
the effects of humidity, chemical spray, radiation, aging, submergence, and
synergism.
- 2 -
In accordance with 10 CFR 50.49, equipment for Fort St. Vrain may be qualified
to the criteria specified in either the DOR Guidelines or NUREG-0588, except
for replacement equipment. Replacement equipment installed subsequent to
February 22, 1983 must be qualified in accordance with the provisions of 10
CFR 50.49, using the guidance of Regulatory Guide 1.89, unless there are sound
reasons to the contrary.
The rule requires that operating reactors achieve environmental qualification
of all equipment by no later than March 31, 1985 unless an extension is
granted by the Director of the Office of Nuclear Reactor Regulation. A record
of qualification, including all necessary documentation, must be maintained in
an auditable form for the period of time the covered item is installed or
stored for future use.
The licensee's March 25, 1985 response to Generic Letter 84-24 concludes that
all equipment which is required to be qualified has been qualified and that
Fort St. Vrain is in full compliance with 10 CFR 50.49. This position was
reiterated in the letter of June 11, 1985. However, the staff review of the
information provided by the licensee in support of this claim leads to the
following observations:
1. The Fort St. Vrain equipment qualification program is based on the
environmental conditions which result from an assumption that a main
steam line break can be isolated within 4 minutes. If isolation
cannot be accomplished within 4 minutes, the temperature continues to
increase and could invalidate any claim of environmental qualification.
The adequacy of this assumption is under review by the staff and the
results could have an impact on the equipment qualification program.
2. For each equipment item, age-related degradation which could prevent
operation of the equipment must be identified and the equipment replaced
or repaired as necessary. The qualified life of the equipment cannot be
exceeded before corrective action is taken.
The licensee has stated that aging effects are not presently included in
the qualification program but an effort in this area has begun. One
necessary element which will establish the validity of the Fort St. Vrain
qualification program is that the results of the aging analysis must be
used to determine that equipment is currently qualified, and must be
factored into the surveillance/replacement intervals to assure that
equipment is maintained in a qualified state. Corrective action must be
taken before, not after, the equipment has exceeded its qualified life.
3. Operability times have not been established for specific pieces of
equipment in compliance with the requirement in 10 CFR 50.49 to
demonstrate the capability to perform the required function for the time
the equipment is required to remain operable.
The licensee has committed to perform a study to establish operability
time requirements. Qualification cannot be considered demonstrated until
the licensee determines that the equipment can perform its intended function
for the period of time it is required to operate during and following an
accident.
- 3 -
4. The licensee has identified a number of outstanding items to be completed
as part of the qualification program and has provided a schedule for
resolution of these items. The scheduled completion dates for these
items extend to March 1986 and are inconsistent with the schedule
requirements in 10 CFR 50.49.
Conclusion
As summarized above, a number of deficiencies are evident in the qualification
program and in the information provided to the NRC for review. Also, guidance
regarding the content and auditability of environmental qualification files
was provided to licensees and applicants in IE Information Notice 85-39, dated
May 22, 1985. Based on the information provided by PSC, it is clear that
documented evidence of qualification does not now exist for equipment in Fort
St. Vrain. Therefore, contrary to the assertions made by the licensee, the
NRC staff cannot at this time conclude that conformance to the requirements of
10 CFR 50.49 has been demonstrated for Fort St. Vrain.
As a result of the discussions during the July 2, 1985 meeting and the
licensee letter of July 11, 1985, the staff decided that the facility should
not be allowed to go to full power operation until all environmental
qualification issues are resolved to the extent that conformance with 10 CFR
50.49 can be demonstrated. The licensee is actively pursuing the means to
resolve these issues.
For the interim period, the licensee has proposed a hold on reactor power to
levels less than 15% of rated power to allow initiation of moisture dryout in
the reactor vessel stating that the lower energy release rates at these power
levels result in significantly lower peak equipment temperatures.
The staff has reviewed the information provided by the licensee in support of
the proposed low power operation. Based on the results of that review, we
find that operation pending resolution of these issues is allowable. The
staff agrees that operation of the plant at power levels not exceeding 15%
will preclude any significant effects on the equipment in question should an
accident occur and that such interim operation will not adversely affect the
health and safety of the public.
Date: July 19, 1985
Principal Contributor:
A. Masciantonio, DE
Hello