ML20003B531
| ML20003B531 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/22/1981 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20003B532 | List: |
| References | |
| NUDOCS 8102120355 | |
| Download: ML20003B531 (7) | |
Text
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4 UNITE 3 STATES j
NUCLEAR REG'JLATORY COMMISSION yw
{,10 Tl WASHINGTON. D. C. 20066 g
METROPOLITAN EDISON COMPANY JERSEY CKNTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendnent No. 60 License No. DPr.-50 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Metropolitan Edison Company, et al. (the licensee) dated April 11, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; I
and i
E.
The issuance of this amendment is in accordance with 10 CPR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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g1QS120 b
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2.
Accordingly, the license is amended by changes to the Technical I
Specifications as indicated in the attachment to this license l
amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-50 is hereby amended to read as follows:
(2) Technical Specitications The Technical Specifications contained in Appendices A and B. as revised through Amendment No. 60, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION d 4
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Robert W. Reid, Chief Operating Reactors Branch #4 Division of 1.icensing
Attachment:
Changes to the Technical Specifications Date of Issuance: Janua ry 22, 1981 4
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ATTACHMENT TO LICENSE AMENDMENT N0. 60 FACILITY OPERATING LICENSE N0. DPR-50 DOCKET NO. 50-289 Revise Appendix A as follows:
Renuve Pages Insert Pages 4-11 4-11 4-12 4-12 4-13 4-13 4-27a 4-27a
REAC"CR CCOIM:' SYSTD1 INSERVICE INSPECTION a.2 Applicability This technical specificatien applies to the inservice inspection of the reactor coolant system pressure boundary and portions of other safety oriented syste=
pressure boundaries as shown on Figure k.2-1.
Cb3ectiv The objective of this inservice inspection program is to provide assurance of the centinuing integrity of the reactor coolant syste= vhile at the sa=e ti=e =inimising radiation exposure to personnel in the perfor=ance of inservice inspections.
Srecification The inservice inspection program to be followed is outlined in Table k.2-1.
L.2.1 Except as provided for in this table and as discussed herein, the inservice inspection program is in accordance with the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Reactor Coolant Syste=s, dated January 1, 1970, as =odified by the Winter 1970 Addenda. Prior to initial plant operatien a preoperational inspection of the plant will be performed of at least the areas listed in the ASME Code; provided accessibility and the The necesspry inspection techniques are available for each of these areas.
only exception to this will be areas where the necessary base line data is
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already available and has been obtained by the sa=e techniques as vill be used during inservice inspectica.
The reactor vessel material surveillance capsules re=oved frc= TMI-l during 2.2.2 1976 shall be inserted, irradiated in and withdravan fro = Crystal River 3 (CR-3) in accordance with the schedule shown in Table 4.2-2.
Unit No.
(The insertion /vithdrawal schedule shown in Table 4.2-2 =ay be revised The licensee shall be at a later date pending the restart of TMI-2.)
responsible for the exa=ination of these speci= ens and for sub=ission of reports of test results in accordance with 10 CFR 50, Appendix H.
The accessible portions of one reactor coolant pu=p =oter flyvheel asse=bly h.2.3 vill be ultrasonically inspected within 3-1/3 years, two within e-2/3 years, and all four by the end of the 10 year inspecticn interval. However, the U.T.
it is I
l procedure is develop = ental and vill be used only to the extent that shown to be =eaningful. The extent of coverage vill be li=ited to these areas of the flywheel which are accessible without =otor disasse=bly, i.e.,
can be reached through the access pcrts. Also, if radiatien levels at the lever access ports are prohibitive, only the upper access ports vill be used.
The inspection schedule =ay be =0dified to coincide with those refueling 4.2.a l
or sintenance outages =ost closely approaching the inspectica schedule.
Sufficient records of each inspecticn shall te kept to allev cc=parison and
.2.
evaluatien of future inspections.
a.2.5 The inservice inspection shall be reviewed at the end of five years to concider incorporation of new inspection techniques and equipment vnich have been preven practical, and a possible extensien of the progra= to additional exa=inatien areas. The conclusiens of this review shall be sub=itted to the URC for evaluation.
4-11 Amendment No. 60 I
d L.2.7 Se licensee shall sub it a report or application for license a=endment to the 'IRC vithin 90 days after any time that Crystal River Unit Bree fails to =aintain a cu=ulative reacter utilization factor of at least ES%.
l 2e repcrt shall provide justification for centinued operation of M-1 l
vith the reactor vessel surveillance progrs= conducted at Crystal River Unit L:
3, or the application for license amend =ent shall propose an 4
alternste progrs= for conduct of the M-1 reacter vessel surveillance
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progrs=.
For the purpose of this technical specificatien, the definition of cc=ercial operation is that given in Regulatory Guide 1.16, Revisien h.
l Se definition of cu=ulative reactor utilization factor is: Cu=ulative l
reacter utilizatien factor = (Cu=ulative =egawatt hours (ther=al) since attai=ent of ec=ercial operation at 1005 power x (100)) divided by (licensed pcVer (.Wt) x (Cu=ulative hours since attai=ent of ec=ercial operation at 100% pover)).
l L.2.3 In addition to the reports required by Specification 4.2.7, a repert shall be sub=itted to the 'iRC prior to Septe=ber 1, 482, which s"--a-ices the first five years of cperating experience with the n!I-l integrated surveillance progrs= perfer ed at a host reactor.
If, at the time of l
sub=ission of this report, it is desired to centinue the surveillance prcgrs:.. at a host reacter, such centinuation shall be justified en the basis of the attained operating experience.
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Amendment No. 60 4-12
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l Inspection Basen The nuclear plant was designed prior to the issuance of Section XI of the a.
ASME Code, Rules for Inservice Inspectics of Nuclear Reactor Coolant Syste=s dated January 1, 1970. However, sufficicnt accessibility was included in the design to perfor= =ost inspections discussed in the code. The proposed inspection program fonovs the code except that inspecticas are focused on areas which engineering analysis has indicated are subject to the =cre critical stress, radiation, or transient conditions. The areas selected for inspection en this basis are listed in Table 4.2-1.
These areas are expcsed to the = ore severe conditions (which are still vell within code li=its) in the reactor coolant system. Therefore, they are expected to indicate potential proble=s It before significant flaws develop in the selected areas or in other areas.
is considered that the focused approach specified herein vill result in a l
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=eaningful inspection progrs= in that it will provide assurance of continuing l
plant integrity.
In those areas where inspection methods are develep= ental, such as for re=ote inspection of the reactor vessel velds, reactor vessel no::le inside radii and velds, and ultrasonic inspection of pressurizer support bracket velds, the inspection =ethods win be developed and tested to the extent practicable during preoperatienal inspections. (Development of inspection techniques vill not be atte=pted on radioactive equip =ent unless necessary to explore a specified proble=.) A preoperational inspection is planned of areas listed in the ASME Code which are within the inservice inspection boundaries and which are accessible for inspection. However, as discussed above, in areas where inspection nethods are develop = ental, the inspections vill only be perfor=ed to the extent practicable. Once an inspecticn methed is selected for a particular inspection (e.g., U.T. for =cs volu=etric inspections), it is intended that all subsequent inservice inspecticas be perfor=ed using the identical =ethod and en the sa=e ce=ponent parts wherever practicable.
In addition to the above inspection, if any of the ec=penents within the ir. service inspection boundary are disasse= bled for =aintenance, the accessible perts vin be given a cor=al visual exa=inatice as part of the routine plant
=aintenance operations.
Because of da= age to the surveillance espsule holder tubes originany installed b.
in TMI-1, irradiation of the SG-1 capsules was to be conducted in DC-2 pursuant to 10 CFR 50, Appendix H, Section II.C.k.
Cne of the five re=aining OC-1 capsules (Capsule E had been withdrawn and tested earlier) was instaned in a holder tube in the OC-2 reactcr at the initial startup of SC-2.
The other four capsules were scheduled for later insertiens. Ecvever, :iue to the IMI-2 Incident, Unit 2 =ay be cut of operatica for a censiderably longer period of time than vill be OE-1.
So that CC-1 vin have an ongoing surveillance props =, a OC-1 capsule vin be inserted into a hol:ler tube in the Crystal i
I River Unit 3 (CR-3) reactor. Because si=ilarities exist between TMI-l and CR-3, appropriate adjust =ents and margins can be imposed to the surveillance capsule irradiation in CR-3 to account for such differences that may exist in the irradiation exposure of the TMI-l reactor vescal and the surveillance capsule.
- he withdrawal schedule has been for=ulated to optisi:e the availability of irradiation data frc= all the capsules being irradiated 'in the CR-3 reacter.
5ecause the irradiatica pre ga= is dependent upcn the successful Operation and a ressenable utilization of CR-3, reperting require =ents are included te pe._.it re-evaluation of che propa= if CR-3 suffers extended cutages.
'he reactor coolant pump ncter flywheel ultrasenic test procedure is being c.
developed to detect flaws of a s=an enough size to previde assurance cf centinued intepity based upon a censervative fracture rechanic's evaluatien.
4-13 Amendment No. 60
TABLE h.2-2 A.
SURVE!LLA'!CE CAPSULE INSERTION & '4ITEDRA'4AL SCHEDULE AT CE-2 (Note: This schedule vill be revised at a later date pending the restart schedules of 2C-1 and 2C-2)
Schedule Cansule Desi natien Insertien
'41thdraval 2C-1A 2C-2 Start-up End of 3rd Cycle T?C-13 End of ist C/cle End of 6th Cycle CC-1D End of 6th Cycle End of 15th Cycle CC-lE Re cved end of 1st Cycle of CC-1 2C-l?
End of 1Cth Cycle End of 2hth Cycle SURVEILLA' ICE CAPSULE I"SERTION L 'JISERA'JAL SCHEDULE AT CR-3 3.
Caesule Designation Insertien
'Ji ' " d ** val T:C-lc End of 2nd C/cle End of 5th Cycle 4-27a Amendtnent No. 60