ML17321A820
| ML17321A820 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/09/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17321A816 | List: |
| References | |
| NUDOCS 8508190491 | |
| Download: ML17321A820 (5) | |
Text
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~y UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 88 TO FACILITY OPERATING LICENSE NO. DPR-58 INDIANA AND MICHIGAN ELECTRIC COMPANY DONALD C.
COOK NUCLEAR PLANT UNIT NO.
1 DOCKET NO. 50-315 Introduction By letter dated February 14, 1985, the Indiana and Michigan Electric Company submitted an application to amend the Technical Specifications for the Donald C.
Cook Nuclear Plant Unit Nos.
1 and 2.
The proposed changes would revise the heatup and cooldown curves based on the recent capsule data test results.
On June 27, 1985 in Amendment No.
69 to Facility Operating License No. DPR-74, the Unit 2 Technical Specification change was approved.
In response to questions by the staff at that time, the licensee again researched the data files to assure the review was based on the best available information on material composition.
On July 3,
- 1985, the licensee submitted the revised material information and on July 18, 1985, the licensee submitted an application for amendment which supersedes the Unit 1 application in the February 14, 1985 letter.
In addition to the proposed changes to the heatup and cooldown curves, the licensee submitted an application on July 19, 1985 to amend the Technical Specifications to revise the low temperature (cold) overpressurization limits.
These limits are related to the proposed heatup and cooldown curves.
Discussion
- 1) Heatup and Cooldown Curves The materials information submitted by letter of July 3 enclosed a letter fi'om Westinghouse Electric Corporation which gave eight measured values of Cu, N> and P content from welds made using the same weld wire heat 850819049i 850809 H PDR ADOCK 050003ii5 P
PDRJ
~ l
'V ~ number as the circumferential beltline weld in the Unit-1 vessel, It was recommended that the average values (0.32K Cu, 0.74K Ni, and 0. 16K P) be used, and the submittal to the NRC did so.
We agree that this is reasonable, because the uncertainty in composition is covered by the margin added to the calculated value of RTNDT.
The surveillance weld for Unit-1 was made with one of the weld wire heat numbers used in making the longitudinal welds.
Its copper and
=. nickel content were 0.27K and 0.74K respectively.
The measured shift reported from Capsule Y was less than that. predicted by Regulatory Guide 1.99, Revision 1 but was very consistent with the results from recent surveillance data.
Nevertheless, the surveillance result was not used directly, because the girth weld material, which has a higher copper content, is expected to be controlling.
The surveillance report for Capsule Y was used as the source of the fluence value.
The reported calculated maximum fluence at the vessel I.D. surface was 3.6 x 10'" n/cm'E
> 1 MeV), corresponding to 4.94 EFPY.
There being no announced plans for flux reduction, these numbers were ratioed and multiplied by 12 to obtain an I. D. surface fluence of 8. 8 x 10'" n/cm~
for the desired period of serviceability of the new P-T limits (12 EFPY).
We accept this value as the basis for the calculation of P-T limits.
To calculate RT the proposed revised Technical Specifications reference NDT'he surveillance report for Capsule Y, prepared by Southwest Research Institute (SRI), and the methods found acceptable by Regulatory Guide 1.99, Revision 1.
This is a satisfactory basis.
In our check of the P-T limits, we used both Revision 1 and some more recent information collected to provide a basis for updating Revision 1.
We found the two were in agreement within + 10'F.
To clarify the basis for calculating the RTNDT, the licensee at our request submitted revised bases pages 8 3/4 4-6 'thru 4-11.
The submittal dated August 1, 1985 contains no new information on the proposed amendment.
The revised pages are acceptable.
0 In conclusion, the staff has used the method of calculating pressure-temperature limits in USNRC Standard Review Plan 5.3.2, NUREG-0800, Rev.
1, July 1981 to evaluate the proposed pressure-temperature limits.
The amount of neutron irradiation damage to the limiting bel tline material was calculated using the method recommended in Regulatory Guide 1.99, Revision l.
Our conclusion is that the proposed pressure temperature limits meet the safety ma'rgins of Appendix G, 10 CFR 50 for twelve (12)
EFPY and may be incorporated into the Unit-1 technical specifications.
2)
Low Temperature (Cold) Overpressurization Limits In the July 19, 1985 letter, the licensee stated that the proposed cold overpressurization limits are to conform to the updated heatup and cooldown curves with respect to the temperature vs pressure limits for the
'eactor coolant system.
The licensee has proposed to reduce the Low Temperature Overpressure Protection System temperature to 170'F, and the low pressure settings on the pressurizer power operated relief valves to 400 psig.
The former settings were 188'F and 435 psig.
Since the allowable pressure on the new isothermal curves at 170'F is over 500 psig, we find these new settings and the proposed Technical Specifications acceptable.
Final Determination - No Si nificant Hazards Determination In our review of the heatup and cooldown curves and the related low temperature overpressurization limits, we have determined that the licensee has used acceptable methods and materials information and that the proposed pressure temperature limits meet the safety margins of Appendix G, 10 CFR 50 for twelve (12) effective full power years of reactor operation.
The proposed revision reflects conservative values of the Reference Nil-DuctilityTransition Temperature (RTNDT) as calculated by the method recomnended by Regulatory Guide 1.99, Revision 1.
Therefore, the proposed amendment does not change the consequences or probabilities of accidents previously evaluated and operation with the new limits assures
that the margin of safety is maintained.
The changes do not create a
new or different kind of accident from any previously evaluated.
On this basis, the staff has made a final determination that the proposed heatup and cooldown curves and the low temperature overpressurization limits involve no significant hazards consideration.
Environmental Consideration This amendment involves a change in the installation or use of a facility
= " component located within the restricted area as defined in 10 CFR Part 20.
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 Comoission has previously issued a proposed finding that this amendment involves no significant hazards ~nsideration and there has been no public cogent on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 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 this 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 this amendment will not be inimical to the common defense and security or to the health and safety of the public.
0ated:
August 9, 1985 Princi al Contributors:
E. Lantz P.
N. Randall
- 0. Wigginton