ML19210B129
| ML19210B129 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/23/1977 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19210B127 | List: |
| References | |
| NUDOCS 7911040064 | |
| Download: ML19210B129 (9) | |
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' *4 UNITED STATES NUCLEAR REGULATORY cOMMISslON o
4 f,% % i Ef WASHINGTON D. C. 20555 9.
\\... e" j METRODOLITAN ED150h COM ANY JiRSEY CENTRAL DOWEP AN LIGHT COMDANY PENNSYLVANIA ELECTRIC COMDANY DOCKET NO. 50-229 THREE MILE ISLAN NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No, 27 License No. DPR-50 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment by Metropolitan Edison Company, Jersey Central Power & Light Company, and Pennsylvania Electric Company (the licensees) dated February 18, 1977, comply 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 conforfaity with the applications, the provisions of the Act, and the rules and regulations of the Comission; 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 (ii) 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; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paraaraph 2.c.(2) of Facility Operating License No. DPR-50 is hereb.y amended to read. as follows:
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(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 27, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with tne Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION G
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Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: March 23,1977 1536 24i
ATTACHMENT TO LICENSE AMEND! TENT NO. 27 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Remove Pages Insert Pages 3-43 & 3-44 3-43 & 3-44 4-29 & 4-30 4-29 & 4-30 4-34a & 4-34b 4-34a & 4-34b The changed areas on the revised pages are shown by marginal lines.
Pages 3-43, 4-29, and 4-34b are unchanged and are included for convenience only.
a i536 242
for any reason, reactor operatien is per=issible for the succeeding seven days prcvided that during such sever days the operable diesel generater is tested L=cediately and daily.
In the event tv0 diesel generators are inoperable, the u=it shall be placed in hot shutdown in 12 hcurs.
If cne diesel is not operable within an additional 2h hour period the pla t shall be placed in ecid shutdevn within an additional 2L hours thereafter."
d.
If one Unit Auxiliary Transfer =er is incperable and a L160 volt tie frc= Unit 2 transfer:er cannet be placed in service and a diesel generator becc=es inoperable, the u=it vill be placed in het shutdevn within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If cne cf the abcve sources of pcVer is not =ade operable within an additional 2k hours the unit shall be placed in cold shutdevn within an additional 2k hours thereafter.
e.
If Unit 1 is separated frc= the syste= vhile carrying its ov.9 auxiliaries, or if cnly one 230 kv line is in service, continued reacter operation is per=issible provided one e=ergency diesel generator shall be staned and run continuously until two trans-
=ission lines are restored.
f.
The engineered safeguards electrical bus, switchgear, load shedding, and autc=atic diesel start syste=s shall be operable except as provided in Specification 3.7.2c above and as requirec for testing.
g.
One station battery may be re=oved frc= service for not =cre than eight hours.
Bas es The Unit Electric Power Syste= is designed to provide a reliable scuree of pcVer for balance of plant auxiliaries and a continucusly available pcVer supply for the engineered safeguards equipment. The availability cf the various co._penents of the Unit Electric Power Syste= dictate the per=issible =cde of station operation.
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3.8 FUEL LOADING A D RF rr?.,INO Anolicability Applies to fuel loading and refueling operations.
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Objective To assure that fuel leading and refueling operations are performed in a responsible manner.
Specification 3.8.1 Radiation levels in the Reactor Building refueling area shall be monitored by RM-G6 and RM-G7 Radiation levels in the spent fuel storage area shall be monitored by RM-G9 If any of these instru-ments beccme inoperable, portable survey instrumentation, having the appropriate ranges and sensitivity to fully protect individuals involved in refueling operation, shall be used until the permanent instrumentation is returned to service.
3.8.2 Core suberitical neutren flux shall be continuously monitored by at least two neut --
flux monitors, each with continuous indication available, whenever.are geometry is being changed.
When core geometry is not being changed, at least one neutron flux monitor shall be in service.
3.8.3 At least one decay heat removal pump and cooler shall be operable.
3.8.h During reactor vessel head removal and while loading and unloading fuel from the reactor, the boron concentration shall be maintained at not less than that required for refueling shutdown.
3.8.5 Direct co==unications between the control room and the refueling personnel in the Reactor Building shall exist whenever changes in core geometry are taking place.
3.8.6 During the handling of irradiated fuel in the Reactor Building at least one door on the personnel and emergency hatches shall be closed.
The equipment hatch cover shall be in place with a minimum of four bolts securing the cover to the sealing surfaces.
3.8 7 During the handling of irradiated fuel in the Reactor Building, each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
1.
Closed by an isolation valve, blind flange or manual valve, er 2.
Be capable of being closed by an operable e.utomatic containment purge and exhaust isolation valve.
3.8.8 If any of the above specified limiting conditions for fuel loading and refueling are not met, movement of fuel into the reactor core shall cease; action shall be initiated to correct the conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be made.
1586 244 Amendment No. 27 3 kh
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Applicability Appliet to centai::.snt le d age.
Obfective To veri.^t the.t lec.kage frc: the reacter building is tsintained rithin allc ac:e limits.
Specificetier k.k.l.1 Intecrated Leah ee Pzte Tests k.h.l.1.1 Design Pressure Leakage Eate (L ), frc= the reactor building at the 55 psig The design inte rated leaktge rate, d
desi n pressure, P, is.1 veight percent of the bui3 ding at csphere at that C
6 pressure per 2h hcurs.
L.k.1.1.2 f.11cvelle Integrated Leake.ge ?. ate The na;.itu: ellevable integrated leakage rate, (L ), frem the reactor building ti a
the cale':lt. ed pe@. reactor building internal pressure of 50.o psig (Pa) associt.ted with the design basis accident, shall not exceed.1 veight percent of the building atsosphere at that pressure per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
h.h.1.1.3 Testing at Relueed Pressure The governing criteria for the periodic integrated leakage rate tests to be per-formed e.t the reduced test pressure, Pt (of not less than 27 5 psig), is the maximun allovable containment test leakage rate, L, which shall be deter-ined t
as follows:
Prior to reactor operation the initial value of the integrated 1cchage a.
rate of the reactor building shall be =easured at design pressure and
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at the reduced pressure to be used in the periodic integrated leakege rate tests. The leakage rates thus measured shall be identified as Lg and L = respectively, and this same nomenclature shall apply to any t
subsequent periodic integrated leakage rate test ceasurements.
b.
Lt shall equal L "tm for values of t= below 0 7 a
L L
am a=
P L
c.
L shall equal L t for values of te above 0.7 g
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The value of Lt will be included in this specification after the initial leak rate test.
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L.L.1.1.L Conduct of Te.ts
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During the period between the initiation of the containment inspection a.
and the performance of a periodic integrated leakage rate test, no re-pairs or adjust =ents shall be =ade unless the inspection reveals structural deterioration which could affect the containment structural integrity or leak-tightness.
Such structural deterioration shall be corrected before performance of the test and a description of the deterioration and the corrective action taken shall be reported as part of the test report sub-
=itted in accordance with technical specification 4.4.1.1.Sw The containment test pressure shall be allowed to stabilize for a period b.
of net less than four hours prior to the start of a leakage rate test.
c.
The test duration shall be at least 2L hours unless experience frc= at
-least two prior tests provides evidence of the adequacy of a shorter test duration.
Test accuracy shall be verified by supple =entary =eans, such as =easuring d.
the quantity of air required to return to the starting point or by i= posing a known leak rate to de=onstrate the validity of measurements.
Closure of contain=ent isolation valves for the purpose of the test shall e.
be acco=plished by the means provided for normal operation of the valves without preliminary exercises or adjustment.
f.
Portions of th( following fluid syste=s will be drained and vented to contain=ent at=osphere prior to and during the integrated leakage rate tests:
1.
Parts of the reactor coolant pressure boundary open directly to containment atmosphere under post accident conditions. (beco=e an extension of contain=ent boundary) 2.
Portions of closed systems inside contain=ent that penetrate contain=ent and rapture as a result of a less of coolant accident.
NOTE:
Syste=s that are required to maintain the plant in a safe condition during the tests and systems that are normally filled with water and operating under post-accident conditions need not be vented.
In addition, missile shielded lines outside the secondary shield vill not be vented.
The fluid block system shall be deactivated prior to and during the g.
integrated leakage rate tests.
h.
All containment components normally pressurized by the penatration pressurization system shall be at at=cspheric pressure during the integrated leakage rate tests.
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Amend =ent No. 27 h-30
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Feactor Bu 1 ding Ncdificatiens Any major =cdification or replace =ent of ec=penents affecting the reacter building' integrity shall be folleved by either an integrated leak rate test or Iccal leak test as appropriate, and shall =eet the acceptance criter:a cf 3
L. 4.1.1. 5 and L. L.1.2. 3, re spectively.
caner (3)
The reactor building is designed for an internal pressure Of 55 psig and a stea=-air =ixtare te=perature of 231 F.
prier to initial cperation, the contain=ent vill be strength tested at 115 percent of design pressure and les}.
rate tested at the design pressure.
The centainment vill also be leak tested pricr to initial operatien at approxi=ately 50 percent cf the design pressure.
These tests vill verify that the leakage rate frc= reacter building pressurization satisfies the relationships given in the specifications.
The perfor=ance of periodic integrated and local leakage rate tests during the l
Plant life provides a current assess =ent of potential leakage frc= the centain=e..:
in case of an accident that vould pressurize the interior of the contain=ent.
In crder to provide a realistic appraisal of the integrity of the contain=ent under accident conditions "as found" local leakage results =ust be docu=ented for correction of the integrated leakage rate test results.
Contain=ent isolation valves are to be closed in the nor=al =anner prior to local or integrated leakage rate tests.
The =inimu= test pressure of 27.5 psig fer the pt iodic integrated leakage rate test is sufficiently high to provide an accurate =easure=ent of the 3eakage rate and it duplicates the pre-operational leakage rate test at the reduced pressure. The specification provides a relationship for relating the measured leakage of air at the reduced pressure to the potential leakage of 55 psig.
The =inimu= of 2h hours was specified for the integrated leakage rate test to help stabilize conditions and thus improve accuracy and to better evaluate data scatter.
The frequency of the periodic integrated leakage rate test is keyed to the refueling schedule for the reactor, because these tests can best be performed during refueling shutdowns.
The specified frequency of periodic integrated leakage rate tests is based en three major censiderations.
First is the low probability of leaks in the liner, because of conformance of the complete contain=ent to a 0.10 percent leakage rate at 55 psig during pre-operational testing and the absence of any significant stresses in the liner during reactor operation.
Second is the
= ore frequent testing, at design pressure, of those portions of the contain=ent envelope that are =ost likely to develop leaks during reactor operation (penetrations and isolation valves which are not continuously pressurized by the penetration pressurization syste= or are not fluid blocked post-accident by the fluid block syste=) and the icv value (0.06 percent) of leakage that is specified as acceptable frc= penetrations and isolation valves.
Third is the tendon stress sur veillance progra= which provides assurance that an important part of the structural integrity of the containment is maintained.
Amendment No. 27 h-3ha 7 [;J
More frequent testing of various penetrations is specifie:i as these locatiens are tuore susceptible to leakage than the reactor building liner due to the mechanical closure involved.
particular attention is given to testing these penetrations and process lines not serviced by the penetration pressurizatien system or the fluid block system. The basis for speci^/ing a total leakage rate of 0.06 percent frc= those penetrations and isolation valves is that more than one-half of the allevable integrated leakage rate vill be fro = these sources.
Valve operability tests are specified to assure proper elesure or opening of the reactor building isolation valves to provide for isolation er functioning of Engineered Safety Features syste=s. Valves vill be stroked to the position required to fulfill their safety function unless it is established that such testing is not practical during operation.
Valves that cannot be full-streke tested vill be part-stroke tested during operation and full-strcke tested during each nor=al refueling shutdown.
REFERENCE (1) FSAR, Section 5 n,8 1500 a 4-3hb