ML19260A156
| ML19260A156 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/01/1977 |
| From: | Arnold R METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19260A155 | List: |
| References | |
| NUDOCS 7910290767 | |
| Download: ML19260A156 (9) | |
Text
1 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT CCMPANY AND PENNSYLVANIA ELECTRIC CCMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DFR-50 Docket No. 50-289 Technical Snecification Chance Request No. 6h This Technical Specification Change Request is submitted in support of Licensee's 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.
METROPOLITAN EDISON CCMPA.f T
By /,'
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' lice President Sworn and subscribed to me this
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1977
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s Notary Public RITA M. PC'.VEP3
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s Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Technical Stecification Change Recuest No. 6h The licensee requests that the attached revised pages replace pages h-ll and h-12 of the existing Technical Specifications, Appendix A.
In addition, pages h-13 to h-27 and Figure h.2-1 are being deleted. For completeness purposes, pages k-27a and h-28 are included.
Reasons for Pronosed Chance The reason for this proposed change is to revise those portions of the TMI-l Technical Specifications, Section h.2, where existing surveillance requirements are superseded by ASME Secticn XI inservice inspection and testing requirements.
Safety Analysis Justifyine Change The proposed change updates present surveillance requirements at TMI-l to make them consistent with ASME Section XI criteria. The change, therefore, does not reduce the efficacy or safety margins of present surveillance requirements. Based on the above, the licensee determines that the proposed change does not involve an unreviewed safety question in that (1) the protability of occurrence or the consequences of an accident or =alfunction of equipment important to safety previously evaluated in the safety analysis report is not increased; (ii) the possibility for an accident or talfunction or a different type than any evaluated previously in the safety analysis report is not created; and (iii) the margin of safety defined in the basis for any technical specificatien is not reduced.
1480 244
h.2 REACTCR COOLANT SYSTEM INSED.' LICE INSPECTICH Annlicability This technical specification applies to the inservice inspection of the reactor coolant system pressure boundary and portions of other safety oriented system pressure boundaries.
Ob.iective The objective of this inservice inspection program is to provide assurance of the continuing integrity of the reactor coolant system while at the same time minimizing radiation exposure to personnel in the performance of inservice inspections.
Specification h.2.1 Inservice Inspection of ASME Code Class 1, Class 2, and Cle3s 3 components shall be performed in accordance with Section :u of the ASME Boiler and Pressure Vessel Code and applicable Add. 2da as required by 10 CFR 50, Section 50 55a (s), except where specific written relief has been granted by the NRC.
h.2.2 Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50 55a (g), except where specific written relief has been granted by the NRC.
h.2.3 The reactor vessel material surveillance capsules re=cved frem TMI-1 during 1976 shall be incerted, irradiated in and withdrawn frcm the Three Mile Island Unit No. 2 reactor vessel in accordance with the schedule shown in Table h.2-2.
13ue licensee shall be responsible for the examinatien of these specimens and for submission of reports of test results in accordance with 10 CFR 50, Appendix H.
h.2.h The accessible portions of one reactor coolant pump motor flywheel asse=bly will be ultrasonically inspected within 3-1/3 years, two within 6-2/3 years, and all four by.the end of the 10 year inspection interval. However, the U.T. procedure is developmental and vill be used only to the extent tha+ it is shown to be meaningful. The extent of coverage vill be limited to those areas of the flywheel which are accessible without motor disassenbly, i.e., can be reached through the access ports. Also, if radiation levels at the lower access ports are prohibitive, only the upper access ports vill be
- used, 1480 245 h-11
h.2.5 The licensee shall submit a report or application for license amendment to the NRC within 90 days after the occurrence of either of the following:
1.
Failure of Three Mile Island Unit No. 2 to achieve ec==ercial operation at 100% power by October 1, 1978, or 2.
Beginning one year after attainment of co==ercial operation at 100% power, any time that Three Mile Island Unit No. 2 fails to
=aintain a cumulative reactor utilization factor of at least 65%.
The report shall provide justification for continued operation of TMI-1 vith the reactor vessel surveillance progra= conducted at Three Mlle Island Unit No. 2, or the application for license amendment shall propose an alternative program for conduct for the TMI-1 reactor vessel surveillance program.
For the purpose of this technical specification, the definition of cet=ercial operation is that given in Regulatory Guide 1.16, Revision h.
The definition of cumulative reactor utilization factor is:
Cumulative reactor utilization factor = (Cu=ulative =egawatt hours (thermal) since attainment of cc==ercial operatien at 100% power x (100)) divided by (licensed power (MWt) x (Cwnulative hours since attainment of ec==ercial operation at 100% pcVer)).
h.2.6 In addition to the reports required by Specification h.2.7, a report shall be submitted to the NRC prior to Septe=ber 1, 1982, which sn mnrizes the first five years of operating experience with the TMI-1 integrated surveillance program performed at TMI-2.
If, at the time of submission of this report, it is desired to continue the surveillance program at TMI-2, such continuation shall be justified on the basis of the attsined operating experience.
Bases a.
Specificaticn3 h.2.1 & 2 ensure that inservice inspection of ASMZ Code Class 1, 2 and 3 cc=ponents and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50 55a (g). Relief frc= any of the above requirements has been provided in writing by the Cet=ission and is not a part of these technical specifications.
b.
Because of damage to the.aurveillance capsule holder tubes originally installed in TMI-1, irradiation of the TMI-1 capsules will be conducted in TMI-2 pursuant to 10 CFR 50, Appendix H, Section 11.C.h.
Because of the similarity of TMI-1 and TMI-2, irradiation in TMI-2 vill be substantially equivalent to irradiation in TMI-1, and appropriate adjustments and
=argins can be imposed in applying the irradiation data to account for such differences as do exist.
)kh h-12
Tne withdrawal schedule has been formulated to optimize the availabilit/
of irradiation data from the capsules of both Units 1 and e Eecause the irradiation program is dependent upon the su:cessful operation and a reasonable utilization of TMI-2, reporting requirements are included to permit reevaluation of the program if LII-2 does not achieve full power operation in a reasonable period of time or suffers extended outages after the first year of operation.
1480 247 h-13
Pages h_1h to h-27 Deleted 1480 248 h-lh
TABLE h.2-2 SURVEILLA: ICE CAPSULE I?!SERTIO:I & WITEDPRAL SCFEDULE AT EII-2 Schedule Caccule Desienation Insertion Withdrawal TMI-1A TMI-2 Start-up End of 3rd Cycle D!I-1B End of 1st Cycle End of 6th C cle
/
E1I-lC End of 3rd Cycle End of lith C/cle DiI-1D End of 6th Cycle End of 15th Cycle TMI-1E Removed end of 1st Cycle of EfI-l D1I-l?
End of 10th Cycle End of 2hth Cycle 1480 249 h-27 a 10!. 2 9 h/2S/77 8 **
'W Mhm w
a.
h.3 TESTI:!G FOLLCWI: G OPE:!IIIS OF SYSTDi ~
Atelicability Applies to test requirements for Reactor Coolant System integrity.
Objective To assure Reactor Coolant System integrity prior to r' turn to criticality following normal opening, modification, or repair.
Scecificatien 4.3.1 When Reactor Coolant System repairs or modifications have been made, these repairs or modifications shall ce inspected and tested to meet all applicable code requirements prior to the reactor being made critical.
L.3.2 Following any opening of the Reactor Coolant System, it shall be leak tested at not less than 2285 psig prior to the reactor being made critical.
h.3.3 The limitations of Specification 3.1.2 shall apply.
Rases Repairs for modifications made to the Reactor Coolant System are inspectable and testable under applicable codes, such as 3 31.7, and ASME Boiler and Pressure Vessel Code,Section IX, IS h00.
For nor:al opening, the integrity of the Reactor Coolant System, in terms of strength, is unchanged. If the system does not leak at 2285 psig (operating pressure +100 psi; 50 psi is normal gystem pressure fluctuation), it will be leak tight during normal operation.
ll)
Reference (1) FSAR, Section h 1480 250 h-28
Figure h.2-1 Deleted 1480 251
.