ML20236D991
| ML20236D991 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/24/1987 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20236D974 | List: |
| References | |
| NUDOCS 8707310171 | |
| Download: ML20236D991 (6) | |
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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT 1 s
Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No.174 This Technical Specification Change Request is submitted in support of Licensee's 4
request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1.
As a part of this request, proposed replacement pages for Appendix A are also included.
4 GPU NUCLEAR CORPORATION BY:
Vict PresiEent & Director, TMI-l Sworn and Subscribed to before me this.2O day of Q u_lzl 1987.
I A m P. L Notary Public tr!Mr. f BR0lm, f.MA!!Y PUntC 4!00tE70Wn B030, DAUPHiti CODNTY 3.Y C0%El3SIOil EXPIP,ES JUNE 12,1989 IAembes Panaspessia Association of WWwo 8707310171 870724 PDR ADOCK 0500 9
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UNITED STATES OF AMERICA i
NUCLEAR REGULATORY COMMISSION 1
IN'THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 GPU NUCLEAR CORPORATION 3
This is to certify that a copy of Technical Specification Change Request No.174-to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with' executives of Londonderry Township, Dauphin County, Pennsylvania; Dr.uphin County, Pennsylvania; and the Pennsylvania Department of Environmental Resources, Bureau of Radiation Protection, by deposit in the United States mail, addressed as follows:
Mr. Jay H. Kopp, Chairman Mr.L Frederick S. Rice, Chairman-Board of Supervisors of Board of County Commissioners-Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Courthouse Middletown, PA 17057 Harrisburg, PA 17120 Mr. Thomas Gerusky, Director PA. Dept. of Environmental Resources Bureau of Radiation Protection P.O. Box 2063 Harrisburg, PA 17120 GPU NUCLEAR CORPORATION BY:
s Vice President & Director, TMI-l DATE:
July 24. 1987
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i I.
TECHNICAL SPECIFICATION CHANGE REOUEST (TSCR) No.174 l
GPUN requests that the following changed replacement pages be inserted into the existing Technical Specification:
Revised pages:
vif, 3-3, 3-4, 3-5 Revised figures:
3.1 -1, 3.1 -2 H
^
These pages are attached to this change request.
I II, REASON FOR CHANGE l
This change is requested to provide updated Technical Specification I
reactor coolant system heatup and cooldown limits for operation to 10 EFPY to account for irradiation effects on the reactor pressure i
vessel's nil ductility temperature (NDT) as determined by the analysis of Surveillance Capsule TMI-1-C.
I f
III.
SAFETY EVALUATION JUSTIFYING CHANGE j
The proposed Techni_ cal Specifications provide updated allowable pressure and temperature combinations and require that the reactor coolant system be maintained within these limits'during normal heatup i
and cooldown, and inservice leak and hydrostatic testing.
The attached Babcock & Wilcox Report, BAW-1901, Supplement 1, June 1987, provides the design bases to support the conclusion that the proposed reactor coolant system pressure-temperature limits provide sufficient margin to account for neutron irradiation effects for 10 EFPY of operation.
Technical Specification Sections 3.1.2.1 and 3.1.2.4 are revised to l
indicate that limits are applicable for operations until 10 EFPY.
4 The Technical Specification Bases are ravised to indicate that the proposed limits are developed in accordance with the methods described in BAW-10046A, Rev. 2 and Regulatory Guide 1.99, Rev. 2, based on analysis of surveillance capsule TMI-1-C.
The Bases are al,o revised to reference BAW-1901, which contains the I
estimates of reactor vessel wall fast neutron. fluxes for Cycles 1 through 4 and the values of reactor vessel fluxes and corresponding fluence accumulation based on predicted fuel cycle design conditions s
for ten (10) full power years of operation, in lieu of specifying these values in the Technical Specification. This is an administrative change which has no impact on nuclear safety.
The Technical Specification Bases are revised to delete reference to 100*F/ hour maximum allowable heatup and cooldown rate. The limits for 10 EFPY have been determined for maximum allowable heatup rates of 50*F/ hour, and a maximum allowable cooldown rate of 100*F/ hour above 282*F and 50*F/ hour at or below 282*F. These heatup and cooldown rates are indicated on the revised Technical Specification Figures 3.1-1 and 3.1-2. __ _ ___-__ _ _________ __-_ _ ___ -______________-___
j The Technical Specification Bases are also revised to delete the interpretive statement that the heatup and cooldown rate limits are not intended to limit instantaneous rates of temperature change, but are intended to limit temperature changes such that there exists no one-hour interval in which a temperature change greater than the limit takes place.
The development of the revised limits is based on linear heatup and cooldown ramp rates which by analysis have been extended to accomodate 15'F step changes at any time with the appropriate soal: (hold) times. Also, an additional 15*F step change has been included in the analysis with no additional soak time to accommodate decay heat initiation at approximately 252*F.
The last sentence of the first paragraph in the Bases is revised to correct a typographical error by adding N (nil) to DTT. This is an editorial change which has no impact on nuclear safety.
The Technical Specification Bases are also revised to indicate that j
the pressure limits are established considering the maximum differential pressure between the point of system pressure measurement and the limiting reactor vessel region for the allowable i
operating pump combinations, in lieu of the maximum differential pressure between the point of system pressure measurement and reactor vessel inlet.
This revision reflects the change in the controlling region of the reactor vessel as discussed in BAW-1901, Supplement 1, Pressure-Temperature Limits.
BAW-1901, Supplement 1, has also been added as Reference 5 to the Technical Specification Bases, i
Technical Specification Figures 3.1-1 and 3.1-2 are revised to indicate the updated reactor coolant system pressure-temperature limits applicable for 10 EFPY of operation.
The titles in the List of Figures is updated accordingly.
The pressure-temperature limits of the reactor coolant pressure boundary (RCPB) for TMI-l are established in accordance with the requirements of 10 CFR 50, Appendix G.
The limit curves for TMI-1 are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at the end of the tenth (10) EFPY, and tM most limiting reference temperature within 1
the beltline region is dsed. The methods and criteria employed to establish operating pressure and temperature limits are described in Babcock & Wilcox Topical Report BAW-10046A, Rev. 2, approved in NRC Safety Evaluation Report dated April 30, 1986.
j The major components of the RCPB have been analyzed in accordance with 10 CFR 50, Appendix G.
The closure head region, the reactor vessel outlet nozzle, and the beltline region have been identified as the only regions of the reactor vessel that control the pressure-temperature limits.
For 10 EFPY of operation, the maximum allowable pressure as a function of fluid temperature is obtained through a point-by-point comparison of the limits imposed by the closure head region, the outlet nozzle, and the beltline region. The maximum allowable pressure is taken to be the lowest of the three calculated pressures. - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _
The unirradiated impact properties were determined for the surveillance beltline region materials in accordance with 10 CFR 50, Appendices G and H.
For other beltline region and RCPB materialt for which the measured properties are not available, the unirradiated.
impact properties and residual elements, as originally established for the beltline region materials, are listed in Table 2-1 of the attached report BAW-1901, Supplement 1.
Figure 2-1 of BAW-1901, Supplement 1, illustrates the calculated peak neutron fluence at several locations, through the reactor vessel beltline region wall.
The neutron fluence values are the predicted fluences that have been demonstrated in BAW-1901, Section 6, to be conservative. Topical Report BAW-1901, " Analyses of Capsule TMI-1-C GPUN TMI-1", March 1986, was submitted to NRC May 5,1986 (GPUN letter 5211-86-2080).
The design curves of Regulatory Guide 1.99 Rev. 2 were used to predict the radiation-induced RTNDT values as a function of the material's copper and nickel content and neutron fluence.
i Protection against nonductile failure is ensured by maintaining the coolant pressure below the upper limits of the pressure-temperature l
limit curves. The maximum allowabic pressure has been reduced by the pressure differential between the point of system pressure measurement and the limiting region of the reactor vessel for all operating pump combinations.
Instrumentation errors for pressure and i
temperature are applied to the W corrected for sensor location.
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IV.
NO SIGNIFICANT HAZARDS CONSIDERATIONS GPUN has determined that the Technical Specification Change Request l
poses no significant hazards as defined by NRC in 10 CFR 50.92.
1.
Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated. The design basis event relate, to this change is nonductile failure of the reactor coolant iressure boundary.
The updated pressure-temperature limits have been established in accordance with the requirements of 10 CFR 50, Appendix G.
Extending the curves for applicability to l
10 EFPY is based on maintaining the design margin assumed in the original curves.
Operation of the facility in accordance with the proposed amendment provides assurance of protection against nonductile failure of the reactor coolant pressure boundary for operation of 10 EFPY. Therefore, operation in accordance with the proposed amendment does not involve a 4
significant increase in the probability of occurrence or consequences of an accident previously evaluated.
2.
Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any previously evaluated.
The design basis event related to the change is nonductile failure of the reactor coolant pressure boundary. The l,
proposed amendment provides assurance of protection against nonductile failure of f he reactor coolant pressure boundary for operation of 10 EFPY and is unrelated to the possibility of creating a new or differes.c kind of accident.
3.
Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. The proposed amendment revises the heatup and cooldown curves to reflect information gained from the analysis of Capsule TMI-1-C, and to extend the curves to cover 10 EFPY of operation.
The results of the analysis of this capsule were reported in GPUN letter to NRC dated May 5, 1986 (5211-86-2080). The report provided additional information on the neutron fluence and Reference Temperature Nil Ductility Temperature (RTNDT) for the reactor vessel material. The updated information is reflected in the revised heatup and cooldown curves.
Extending the curves for applicability to 10 EFPY is based on maintaining the margin of safety assumed in the original curves. This margin of safety is assured by the results of the surveillance capsule analysis. The updated pressure-temperature limits have been established in accordance with the requirements of 10 CFR 50 Appendix G.
Therefore, it is concluded that operation of the facility in accordance with the proposed amendment does not involve a significant reduction in a margin of safety.
The Commission has provided guidelines pertaining to the application of the three standards by listing specific examples in 48 FR 14870.
l The proposed amendment is considered to be in the same category as example (ii) of amendments that are considered not likely to involve significant hazards consideration in that the proposed change constitutes an additional control not presently included in the technical specifications. This change is similar in that it involves a change to the heatup and cooldown curves that places more restrictions on heatup and cooldown than previously existed. Thus, operation of the facility in accordance with the proposed amendment involves no significant hazards considerations.
V.
IMPLEMENTATION It is requested that the amendment authorizing this change become effective upon issuance.
Issuance is requested prior to November 1,1987, to allow sufficient time for implementation prior to exceeding 5 EFPY.
VI.
AMENDMENT FEE (10 CFR 170.21)
Pursuant to the provisions of 10 CFR 170.21, attached is a check for
$150.00. _ _ _ _ _ _ - -. - -
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