ML17325A641
| ML17325A641 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/29/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17325A640 | List: |
| References | |
| NUDOCS 8803040362 | |
| Download: ML17325A641 (8) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
TO FACILITY OPERATING LICENSE NO.
DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C.
COOK NUCLEAR PLANT UNIT NO.
2 DOCKET NO. 50-316
1.0 INTRODUCTION
By letter dated January ll, 1988, the Indiana Michigan Power Company submitted a request for revision of the Technical Specifications (TSs),
Appendix A to Facility Operating License DPR-74 for D.
C.
Cook Nuclear Plant, Unit 2.
The proposed revision would extend the surveillance requirements for several items starting from March 2, 1988, to the next refueling outage currently scheduled to begin June 10, 1988.
This one-time extension was requested due to operation at 80K of rated thermal power and various unanticipated outages of up to 49-day duration which resulted in a lower rate of fuel burnup.
The affected TSs include:
TSs Affected (1)
- 4. 5. 2. d. 1 4.5.3.1 (2)
- 4. 7. 7. 1 (3)
- 4. l. 3. 3 (4)
Table 4. 3-1, Items 7 8 8
- 4. 3. 2. l. 2 Table 4.3-2, Item 4.d Table 4.3-10, Items 2, 3, 8
11 (5)
Table 4.3-2, Items l.a, 2.a 3.a.l, 3.b.l, 3.c.l, 4.a Descri tion of Chan e
Delay RHR auto-closure interlock testing Delay steam generator snubber functional testing Delay analog rod position indication functional testing Delay RTD calibrations Delay testing of ESF manual actuation switches 88Q304Q3b2 88022093th pop
- DOCII'DQQP03DRb P
(6)
Table 4.3-1, Items 7, 9
Table 4.3-2, Item l.d 4.3.2.1.2 Table 4.3-6, 4.4.11.l.b Delay pressur izer pressure
& 10 calibrations, interlock function
- testing, and PORV calibrations Item 2 In addition to the surveillance interval extensions, the amendment also proposes two minor editorial changes to correct errors in the present TS pages.
The first of these changes adds the word "by" between the words "OPERABLE" and "the" in TS 4,3.2. 1.1.
The second change deletes a redundant "the" from TS 4.5.2.g.
These changes are purely editorial in nature.
The Comission's staff has determined that failure to act in a timely manner to grant extensions for the residual heat removal (RHR) auto-closure interlock testing, steam generator snubber functional testing and rod position indication functional testing would result in shutdown of Unit 2.
Therefore, this Safety Evaluation deals only with these three extensions.
The remaining three extension requests will be evaluated by the staff at a later date under a
separate cover.
2.0 DISCUSSION The proposed amendment is the second of two submittals that request surveillance interval extensions for Unit 2, Cycle 6.
The changes requested in this proposed amendment supplement the extension requests submitted in the licensee's amendment request dated October 28, 1987.
Those changes were granted by the Commission's staff by Amendment No. 97 to Facility Operating License No. DPR-74.
In addition, this reouest for surveillance extension is very similar to a recent extension which was granted for D.
C.
Cook, Unit l.
The reasons for the extension and the equipment included in this request are similar.
In discussions between the licensee and the staff, the licensee has stated that the results of the Unit 1 extensions have been reviewed and no problems were discovered with operability or instrument drift for any of the items that are being requested for this Unit 2 extension.
The specific TS changes are addressed below.
(1)
RHR Auto-closure interlock The proposed amendment requests a 4-month extension for the RHR auto-closure interlock test required by TS 4.5.2.d.l.
An extension is also requested for TS 4.5.3. 1 since it references TS 4.5.2.
The RHR auto-closure interlock auto-matically isolates the RHR system from the Reactor Coolant System (RCS) if RCS pressure is above 600 psig.
In order to demonstrate operability of the auto-closure interlock, it is necessary to open the RHR isolation valves in the cooldown line from the hot leg in order to verify that the valves would auto-matically close with the RCS pressure above 600 psig.
This cannot be accomplished with the unit operating (i.e., with the RCS fully pressurized) because it would result in exposing the RHR system to pressures higher than the RHR safety valve setpoints.
Previous surveillance testing by the licensee has demonstrated that the auto-closure interlock is very reliable.
The previous test results give the staff no reason to believe the auto-closure interlock would be inoperable
e during the extension period.
The calibration for the RCS wide-range. pressure transmitters, which provide input into the interlock, can be done't power and will be performed by the March 2, 1988, due date.
Thus, the only portion of the interlock for which surveillances will not be current is the portion from the bistable of the RHR suction valves through valve operation.
To meet the
~
single failure criterion, all active components of the RHR System, including isolation valves, are duplicated.
Therefore, any undetected failure which might result from the lengthening of the surveillance interval will likely be offset by this built-in redundancy.
Also, the general fail-safe design of the systems offers an additional level of protection.
Therefore, the Commission's staff finds the licensee's request for one-time surveillance extension acceptable.
(2)
Steam Generator Snubbers The proposed amendment would delay functional testing of steam generator snubbers required by TS 4.7.7. 1.c.
The extension is needed from March 9, 1988, until the refueling outage.
The steam generator snubbers for which an extension is being requested are those numbered 91 and 92 in TS Table 3.7-9.
The extension is requested because these snubbers are inaccessible during power operation and because TS 4.7.7. l.c specifically requires the testing to be performed during shutdown.
Both snubbers required to be tested were selected randomly, i.e., neither of them are being tested as a result of a previous failure.
Thirteen of the 32 snubbers in Units 1 and 2 have been functionally tested, and of the 13 tested only one failed, that being a
failure to lock-up in compression.
The problem was not generic, and the snubber passed the subsequent retest in 1985.
Visual inspection of the steam generator snubbers per TS 4.7.7. 1.a is not required until after the scheduled outage start date, and for this reason, no extension for TS 4. 7. 7. l. a is requested.
Visual inspections have been performed on steam generator snubbers at the Cook Nuclear Plant since 1975.
These inspections are performed at least once per refueling cycle.
No problem or potential problem has been revealed by these inspections.
All snubbers have been found to be acceptable and no generic problems have developed.
A similar request for an extension for Unit 1 snubber surveillances was approved by the Commission on December 20, 1986, via Amendment 100 to the Unit 1 TSs.
The Safety Evaluation for that amendment required that the snubber functional testing surveillance requirements be revised to increase the snubber testing sample size at least in proportion to the increase in the length of the refueling cycle beyond 18 months.
The licensee intends. to impose this requirement on themselves for the Unit 2 steam generator snubbers as well.
This will require the licensee to perform functional testing of at least one more steam generator snubber during the upcoming refueling outage.
On the basis of the history of D.
C.
Cook Unit 2 snubber testing and inspection
- results, there is high confidence in the operability of the D.
C.
Cook Unit 2
- snubbers, and operation for approximately four additional months past the due date for snubber functional testing will not result in a significant decrease in plant safety.
Therefore, plant shutdown to perform snubber functional testing at the due dates indicated above would be unwarranted and the licensee's requested extension is acceptable.
C e
(3)
Rod:.Posi.ti.on Indication The proposed change would delay functional testing of the rod position indicator (RPI) channels required every 18 months by TS 4. 1.3.3.
The extension is needed from March 21, 1988, until the refueling outage expected in June 1988.
Although TS 4. 1.3.3 is only applicable in Modes 3, 4, and 5, the licensee believes relief is needed from this TS to continue operation in Modes 1 and 2 since TS 3/4. 1.3.2 requires the RPI channels to be operable in these modes.
The surveillance the licensee performs to satisfy TS 4. 1. 3. 3 is far more stringent than the channel functional testing that the TS requires.
The test is actually a calibration of the RPI channels over the rod insertion range.
Since rods must be inserted to perform the calibration, it cannot be performed at power because to do so would violate the rod insertion limits of TS
- 3. 1. 3. 6.
The operability of the RPI channels is functionally verified once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per TS 4. 1. 3. 2 by comparison to the demand position indication system.
These comparisons would be expected to indicate significant degradation in the RPI channels.
Indication that the core is performing as designed is provided by the incore flux maps, which are taken at least once every 31 effective fIIll power days to satisfy the requirements of TSs 4.2.2.2 (FA(Z)) and 4.2.3 (F< ).
Core performance is also indicated by the excore detectors, which are Ised to measure the quadrant power tilt ratio per TS
- 4. 2. 4.
These surveillances would also be expected to indicate discrepancies between indicated and actual rod position.
Therefore, the staff finds the licensee's request for a one-time surveillance extension acceptable.
- 3. 0 EMERGENCY. CIRCUMSTANCES The Commission has'etermined that emergency circumstances exist in that swift action is necessary to avoid shutdown of D.
C.
Cook Unit 2.
The need for certain surveillance extensions was identified in September 1987.
The dates that the surveillance intervals would expire fell into two groups.
Several were due during the brief period of time from December 31,
- 1987, through January 4, 1988.
The second group was not due until the period beginning March 2, 1988, and continuing into the Unit 2 refueling outage.
Since there was a gap of approximately two months between the groups, in discussions between the staff and the licensee, the decision was made that two requests for surveillance extensions would be appropriate.
Thus, attention could be focused on those requests which involved the most immediate need.
The licensee's January 11,
- 1988, proposed amendment is the second of the two submittals that requested surveillance interval extensions for Unit 2, Cycle 6.
The changes requested in the January ll, 1988 letter supplement the extensions requested by the letter dated October 28, 1987.
Those changes were granted by the Commission by Amendment No.
97 to Facility Operating License No.
DPR-74 dated December 28, 1987.
Inadvertently the licensee's January 11, 1988 request was not promptly noticed by the NRC.
Notice requesting comments on the Commission's proposed no significant hazards consideration determination was published in the Federal.
~Re ister on February 17, 1988 (53 FR 4795).'o comments have been
- 4. 0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Commission's regulations in 10 CFg 50.92 state that the Commission may make a final determination that a license amendment involves no significant hazards considerations if operation of the facility, in accordance with the amendment, would not:
(1)
Involve a significant increase in the probability or consequences of any accident previously evaluated; or
~
(2)
Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)
Involve a significant reduction in a margin of safety.
The requested amendment has been evaluated against the standards in 10 CFR 50.92 as follows:
RHR Auto-closure Interlock Test Criterion 1 The surveillance test history of the auto-closure interlock has shown that the system is highly reliable, and there is no reason to believe the equipment would be inoperable during an extension period.
The wide-range pressure transmitters, which provide input into the auto-closure interlock, will have a current calibration.
Additionally, when the RHR system is not in service, power is removed from the suction valve operators, thus preventing inadvertent valve opening and eliminating the need for the auto-closure interlock.
For these
- reasons, the extension is not expected to result in a significant increase in the probability or consequences of a previously evaluated
- accident, nor will it result in a significant reduction in a margin of safety.
Criterion 2 This extension will not result in a change in plant configuration or operation.
Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.
Criterion 3 (2)
See Criterion 1, above.
Steam Generator Snubbers Criterion 1 Thirteen steam generator snubbers have been functionally tested at the Cook Nuclear Plant since 1983 with only one failure, the cause of which was not generic.
Visual inspections have been performed on snubbers since 1975, revealing no problems or potential problems.
Based on this surveillance history, the steam generator snubbers are not expected to
be inoperable during the extension period.
Thus, it is believed that this "change will not result in a significant increase in the probability or consequences of a previously evaluated
- accident, nor will it significantly reduce a margin of safety.
Criterion 2 Delaying the snubber functional test will not result in a change in plant
" design or operation.
Therefore, the change will not create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated.
Criterion 3
(3)
See Criterion 1, above.
Rod Position Indication System Functional Testing Criterion 1 TS-required comparison of the RPI channels to the demand position indication system would be expected to indicate significant degradation in the RPI channels.
In addition, other surveillance, such as the determination of the quadrant power tilt ratio and incore flux mapping, provide a comparison of core performance to design and would be expected to indicate significant deviations of the control rods from their indicated position.
Also, the RPI channel surveillance history is good and provides no reason to believe the changes would be inoperable during the extension period.
For all these
- reasons, the change will not involve a significant increase in the probability or consequences of a previously analyzed accident and it will not involve a significant reduction in a margin of safety.
Criterion 2 The proposed change will not result in a change in plant configuration or operation.
Thus, the change should not create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated.
Criterion 3 See Criterion 1, above.
Therefore, based on these considerations and the three standards given above,.
the Commission has made a final determination that the requested changes involve no significant hazards consideration.
5.0 STATE CONSULATION In accordance with the Commission's regulations, efforts were made to contact the Michigan representative.
The state representative was contacted and had no commen'ts.
- 6. 0 ENVIRONMENTAL CONSIDERATION This amendment involves changes in th surveillance requirements.
The staff has determined that the amendment involves no significant increase in the
- amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has made a final no significant hazards consideration finding with respect to this amendment.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 551.22(c)(9).
Pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
7.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed
- manner, and (2) such activities will be conducted in compliance with the Commission s regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date:
February 29, 1988 Principal Contributor:
John F.
Stang