ML17325A034
| ML17325A034 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/03/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17325A033 | List: |
| References | |
| NUDOCS 8704100018 | |
| Download: ML17325A034 (14) | |
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UNITED STATES NUCLEAR R EGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
90 TO FACILITY OPERATING LICENSE NO. DPR-74 INDIANA AND MICHIGAN ELECTRIC COMPANY DONAI.D C.
COOK NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-316 INTRODUCT ION I ~
I'c By letter dated March 30, 1987, the licensee has requested a one-time change to Technical Specification (TS} 4.6.5. 1.b.2 concerning weighing of ice baskets.
This technical specification item requires that a representative sample of at least 144 ice baskets be weighed to verify a minimum of 1220 lbs of ice per basket.
The Adhnical specification also requires that the representative sample include 6 baskets from each of the 24 ice condenser bays and be con-stituted of one basket each from radial rows 1, 2, 4, 6, 8 and 9 within each bay.
The licensee is therein required to demonstrate a minimum acceptable weight per basket, or minimum average weight based on a larger sample, for each bay.
The licensee is also required by this technical specification to evaluate the ice weight for 18 radial row groups of baskets (as opposed to bays).
With regard to this change request the licensee specifically has requested a
change to the technical specifications that would allow the weighing of three row 8 baskets in lieu of three adjacent row 9 baskets.
The request is neces-sitated by the inability to weigh the three required baskets in row 9 as they are apparently frozen in place and cannot be lifted for weighing, even after diligent efforts to do so.
As an alternative to requesting a change in the technical specification the licensee could satisfy the existing technical specification by emptying 144 row 9 baskets and refilling them.
The licensee has evaluated this option and rejected this approach on the basis that it would result in an unnecessary delay in the return to power, due to the time consum-ing nature of the task (about 25 days), with associated costs of approximately
$ 195,000 for each day's delay in returning to power.
Power operation is scheduled to resume on or about April 12, 1987.
The licensee, therefore, has opted to request a one-time change in the technical specification and provided justification for the adequacy of this approach.
Furthermore the licensee has committed, as an additional compensatory
- measure, to conduct an additional weighing of'ows 2, 4, 6 and 8 in each bay approximately 4 months after re-turn to power.
EVALUATION The purpose of the technical specifications concerning ice basket weights is to insure that the total ice mass, with allowance for sublimation, is con-sistent with assumptions in the safety analysis and that the distribution of 41000ZB ~040 8'70'
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ice, both circumferentially and radially, is reasonably
- uniform, Thus, the technical specifications require the compliance of ice weight criteria on a
per bay basis and also on a grouping which can be considered as row groups.
Meeting requirements for each bay provides assurance that the ice weight is sufficiently uniform circumferentially and the requirement for row-groups assures uniformity of ice weight in the radial direction.
With regard to row 9 baskets there are 3 groups with 8 baskets in each group; Group 1-bays 1 through 8, Group 2-bays 9 through 16, and Group 3-bays 17 through 24.
As noted above the licensee has proposed to substi,tute weighing three row 8 baskets for adjacent row 9 baskets.
This is due to the inability to weigh one row 9 basket in Group 1 and two row 9 baskets in Group 2.
All of the re-quired baskets in Group 3 have been weighed.
In order to justify that the row 8 basket ice weights are representative of row 9 basket ice weights the licensee has analyzed the results of recent ice basket weighing, conducted in March 1987.
Based on recent ice weighing, several notable observations can be made.
All of the row.9 baskets that could be weighed contained over the 1220 lb/basket minimum and the average ice weight for all the row 9 baskets was 1382 lbs, well over the minimum required by the technical specifications.
Visual observation of the row 9 baskets which cannot be weighed did not reveal any apparent differd'nces in ice loading from the other baskets.
The licensee has also performed limited statistical analysis which suggests that row 8 and row 9 baskets are equivalent.
As further justification for the technical specification change the liceasee has proposed to reweigh, four months after return to power, ice baskets in rows 2, 4,. 6, and 8.
The staff has considered the arguments provided by the licensee and concurs that the one-time change to the technical specification is warranted and does not present a significant safety threat.
Recent ice basket weighings provide evidence that row 9 baskets contain adequate ice and that substitution of three row 8 baskets for adjacent row 9 baskets is acceptable.
Furthermore, the ice weighing surveillance conducted in March 1987, resulting in the weighing of approximately 220 baskets, provides assurance that the ice condenser will per-form its intended function.
Basis for Emer enc Technical Specification Chan e
The Unit 2 is currently in a forced outage caused by steam generator tube leakage.
The ice condenser surveillances were initiated during this forced outage to prevent having to shut down the unit when the surveillances would have normally come due on April 19, 1987.
The licensee did not anticipate any unusual problems during the current ice basket weighings,
- however, row 9 baskets which were required to be weighed have not been freed for weighing and many of the baskets appear to be frozen in place.
-The licensee has used various methods to free the frozen baskets and will continue to do so until the unit is ready to start up about April 7, 1987.
The one remaining Dro-cess to free the baskets would require unloading and reloading with ice and would further extend the outage for about 25 days.
t The licensee has exhausted the reasonable and available methods for freeing the baskets.
On March 24, 1987, the licensee notified the Office of Nuclear Reactor Regulation that it would be necessary to extend the outage or request
an emergency technical specification change.
At that time, the startup was scheduled for April I, 1987 and clearly there was insufficient time to prop-erly pre-notice any such proposed action.
On March 27, 1987, the licensee advised the NRR that the steam generator examinations and repairs would extend the outage and that startup would begin with Mode 4 being reached about April 7, 1987.
This still leaves insufficient time to properly pre-notice the proposed license amendment and finding of no significant hazards consideration in the Federal Re ister.
The change was needed at the facility to prevent schedule sT>ppage an a
ow plant restart following the Unit 2 outage.
We have determined that the licensee has been responsive in the notification to NRR and in the submittal of the proposed Technical Specifi-cation change.
Based on our review, we do not believe the licensee delayed their notification or their submittal to create an emergency situation and take advantage of the post notice situation.
Discussions with the State of Michi an On March 31, 1987, the proposed Technical Specification
- change, the conditions requiring an emergency amendment of the license, and the staff's final no significant hazards consideration were discussed with the State of Michigan contact for licensing matters.
It was agreed that the efforts by the licensee to free the baskets'.and meet the requirements of the technical specifications were appropriate, that the substitution of row 8 for row 9 weights provides sufficient assurance for ice condenser operability for restart, and that the licensee's commitment to check weights at mid cycl.e would offer continued pro-tection for public health and safety during this period.
The State of Michigan understands the Commission's actions and has no further coments.
Final No Si nificant Hazards Determination In our review of the ice condenser operability with the licensee's inability to weigh many of the row 9 baskets, we determined that the required amount of ice is available in the baskets that could be weighed and that substitution of weights in row 8 for row 9 baskets was appropriate for a one-time basis.
Therefore, in the unlikely event of an accident, the frozen baskets in row 9 would perform as required.
The frozen baskets have not produced any detri-mental effects, outside of not being able to weigh them, and unloading the frozen baskets to free them would require time to replace the ice.
We con-sidered removing the ice from at least one frozen basket to determine the extent or cause of freezing between the baskets but because this would not offer any improvement in the ice condenser operability and would 'be time con-
- suming, we agreed with the licensee that freeing all the baskets could best be accomplished at the next refueling.
The licensee has coranitted to a mid-cycle test on ice basket weight to assure no abnormal weight loss.
With the ice available now in the baskets, this mid-cycle test would be a reasonable check on the longer term effects of ice basket freezing as well as additional assurance against any unforeseen loss of ice from the baskets.
The Comnission's standard for determining whether a significant hazard consider-ation exists is stated in 10 CFR 50. 92.
A proposed amendment to an operating license for a facility involves no significant hazards consideration if oper-ation of the facility in accordance with a proposed amendment would not (1) involve a significant inc~ease in the probability or consequences of an accident previously evaluated, (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 most significant accident for which the ice condenser must remove heat from the containment is the loss of coolant acci-dent (LOCA).
For this accident, the amount of ice flow is of importance.
From the above discussion, we ag~ee that the weights, even with the substitution, remain above the minimum requirements.
Therefore the proposed amendment does not involve a significant increase in the probability or consequences of the most significant accident previously analyzed; the LOCA.
The freezing between the baskets makes them impossible to weigh but the freezing has not produced any other detrimental effect since the baskets are already held in place together by clevice pins for seismic consideration.
The baskets are not free for any exten-sive movement when not frozen together.
The row 9 baskets that are frozen together are likewise not available for excessive movement but are operational as required.
We agree that the frozen baskets and the change to substitute row 8 for row 9 weights does not involve a new or different kind of accident from any previously analyzed.
The substitution of row 8 for row 9 weights is valid from the weight observed in all the other rows and in the row 9 baskets that could be weighed.
The adjacent row 8 baskets provide sufficient assurance that the required total ice is available since gross local ice loss is not observed.
The proposed amendment to substitute row 8 for row 9 basket weights does not in-volve a significant reduction in a margin 'of safety.
Therefore, based on these considerations, the Commission has made a final determination that the amendment request involves a no significant hazards consideration.
ENVIRONMENTAL CONSIDERATION This amendment involves a change in the installation or use of the facilities'omponents located within the restricted areas as defined in 10 CFR 20.
The staff has determined that 'the amendment involves no significant increase in th
- 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 determined that the amendment involves no significant hazards consideration.
Accordingly, the amendment meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR Sec 51.22(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 the amendment.
CONCLUSION We have 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 pub'lic.
Principal Contributors:
D. Wigginton C. Tinkler April 3, 1987
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'g Docket No.
50-316
.:fcENSE::AUTMORITY FILE Cop%
UNITEDSTATAS NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 JUg 0 6. 1986 DO NOT REMOVE P~+od.
~, '3> M Z>PL -74 Mr. John Dolan, Vice President Indiana and Michigan Electric Company c/o American Electric Power Service Corporation 1 Riverside Plaza
- Columbus, Ohio 43216
Dear Mr. Dolan:
In our letter to you dated May 21, 1986, we transmitted Technical Specifications for the Donald C.
Cook Nuclear Plant, Unit No.
2 as a result of Amendment 82 to Facility Operating License, DPR-74.
The Technical Specifications con-tained inadvertent errors which do not affect the results of our review.
The attached Technical Specifications should be substituted in part for those in the May 21, 1986 letter.
We regret any inconvenience this may have caused.
Sincerely, D.L. Wiggi o
, Project Manager PWR Project Directorate
$4 Division of Licensing-A
Mr. John Dolan Indiana and Michigan Electric Company Donald C.
Cook Nuclear Plant CC:
Mr. M. P. Alexich Vice President Nuclear Operations American Electric Power Service Corporation I Riverside Plaza
- Columbus, Ohio 43215 Attorney General Department of Attorney General 525 West Ottawa Street Lansing, Michigan 48913 Township Supervisor Lake Township Hall Post Office Box 818 Bridgman, Michigan 49106 W. G. Smith, Jr., Plant Manager Donald C.
Cook Nuclear Plant Post Office Box 458 Bridgman, Michigan 49106 U.S. Nuclear Regulatory Commission Resident Inspectors Office 7700 Red Arrow Highway Stevensville, Michigan 49127 Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbr idge 1800 M Street, N.W.
Washington, DC 20036 Mayor, City of Bridgeman Post Office Box 366 Bridgman, Michigan 49106 Special Assistant to the Governor Room 1 - State Capitol Lansing, Michigan 48909 Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3500 N. Logan Street Post Office Box 30035 Lansing, Michigan 48909 The Honorable John E. Grotberg United States House of Representatives Washington, DC 20515 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 J. Feinstein American Electric Power Service Corporation I Riverside Plaza
- Columbus, Ohio 43216
G COND 3.2.2 F
) shall be limited by the following relationships:
W t
ous~
ue F (Z) ~~ [K(Z))
Q P
F (Z) < [3 94] [X(Z)]
xo
'uc e
o ue Fq<Z) 6 p
[Y(Z)l F (Z) < [4.20] [K(Z)]
p) 0.5 P < 0.5 Ob RATED ERMAL POWER (Z) is the easured hot channel factor including a.3X manufac-
~ turing toler ce'uncertainty"and a 5Z measuremenh uncertainty.
~ K(Z) is the funct n obtained from Figure 3.2-2 for Westinghouse fuel d Figure 3.2-2(a) for Exxon Nuclear'ompany fuel..
MODE 1 With F (Z) exceeding its limit:
a.
'educe THERMAL POWER at least 1% fo each 1% F (Z) exceeds the limit within 15 minutes and similarly reduce e Power Range Neutron, Flux-High Trip Setpoints withi the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total f 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the erpower AT Trip Setpoints have been reduced at least 1% for each
.1% F (Z) exceeds the limit.
b.
Identify and correct the cause of the out of imit condition prior to increasing THERMAL POWER above the reduced 1imit required by a, above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its smimit.
D. C.
COOK - UNIT 2 3/4 2-5 AMENDMENT N 82
BLE 3-3 Co tinued F ATURE CTUATION SYSTEM I STRUMENTATIO 0
3 ~
CONTAINMENT ISOLATION a.
Phase "A" Isolation TOTAL NO.
0 C
CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES
- 1) Manual 2
2)
Prom Safety 2
In)ection Automatic Actuation Logic b.
Phase "B" Isolation
- 1) Manual
- 2) Automatic Actuation Logic
- 3) Containment Pressure-Hi High c.
Purge and aust Isolatio II 1) nual
- 2) Containment.
Radioactivity-High Train A
- 3) Containment 3
Radioactivity-High Train B
1, 2, 3, 4
1, 2, 3, 4
1, 2, 3, 4
1, 2, 3, 4
1~
2~
3 1, 2, 3, 4
1, 2, 3, 4
1, 2,
3, 4
0
d 4
O Th OPERABILITY of the protective and ESF instrumentation systems and interloc ensure that 1) the associated ESF action and/or reactor trip will be initia qd when the parameter monitored by each channel or combination thereof exceeds its setpoint,
- 2) the specified coincidence logic ks maintained, 3~ sufficient redundancy is maintained to permit a channel to be out of service or testing or maintenance, and 4) sufficient system functional capab lity is available for protective and ESF purposes from diverse parameter The OPERABILI of these systems is required to provide the overall reliability, redundan and diversity assumed available in the facility design for the. protect n and mitigation of accident and transient conditions.
The integra ed operation of each of these systems is consisten" with the assumptions used n the accident analyses.
Protection has been pro ided for main feedwater system malfunctions in MODES 3 and 4.
This protecti n is required when main feedpumps are aligned to feed steam generators in MO ES 3 and 4.
The availability of feedwater isolation on high-high steam geAqrator level terminates the addition of cold water to the steam generators in~ ny main feedwater system malfunction.
The total volume that can be added to he steam generators by the main feedwater system in MODES 3 and 4 is limited this safeguards actuation and the fact that feedwater isolation on low T setpoint coincident with reactor trip can only be cleared above the low-low ste generator level trip setpoint.
avn The restrictions associated with b assing ESF trip functions below either P-11 or P-12 provide protection a
inst an increase in steam flow transient and are consistent with assumpt ns made in the safety analysis.
The surveillance, requirements spehifie for these systems ensure that the overall system functional capability is intained comparable to the original design standards.
The periodic surve llance tests performed at the minimum frequencies are sufficient to demonstra e this capability.
The measurement of response time at the spec fied frequencies provides assurance that the protective and ESF action funct on associated with each channel is completed within the time limit assumed n the accident 'analyses.
No credit was taken in the analyses for those channe s with response times indicated as not applicable.
Response
time may be demonstrated by any series of equential, overlapping or total channel test measurements provided at such tests demonstrate the total channel response time as defined.
nsor response time verification may be demonstrated by either 1) in place, o
te or offsite test measurements or 2) utilizing replacement sensors with c rtified response times.
V