ML19317G770
| ML19317G770 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/13/1977 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317G751 | List: |
| References | |
| NUDOCS 8004010598 | |
| Download: ML19317G770 (8) | |
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D UNITED STATES o
y NUCLEAR REGULA TORY COMMISSION
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%,,.....f SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO. 50-312 RANCHO SECO NUCLEAR GENERATING STATION AMENDMENT TO FAF.ILITY OPERATING LICENSE Amendment No, 11 License No. DPR-54 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Sacramento Municipal Utility District (the licensee) dated November 26, 1975, as revised October 5,1976, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Cecr.ission's rules and regulations set fcrth in 10 CFR Chanter I; B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and mgulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized i
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 CPR Part 51 of the Comission's regulations and all applicable mquirements have been satisfied.
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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this' license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-54 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B as revised through Amendment No.11, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR TriE NUCLEAR REGULATORY COMMISSION h
Robert W. Reid, Chief i
Operating Reactors Branch #4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuan:e: July 13,1977
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ATTACHMENT TO LICENSE AMENDMENT NO. 11
_ FACILITY OPERATING LICENSE NO. OPR-54 DOCKET NO. 50-312 Revise Appendix A as follows:
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Remove Pages_
4 4-20 4 4-20 Page 4-19 Changes on the revised pages are shown by marginal lines.
is unchanged and is included for convenience only.
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-s RANCHO SECO UNIT 1 TF.CHNICAL SPECIFICATIONS Survci11ance Standards provided local leakage ~ rate measurements are made before and af ter repair to demonstrate that the leakar,c rate reduction achieved by repairs reduces.che overall neasured integrated leak rate to an acceptabla value.
4. 4.1.1. 7 Report of Test Results Each integrated leak rate test will be the subject of a summary technical report which will include a description of test methods used and a summary of local leak 'etection tests. Suf ficient data and analysis shall be included to show that a stabilized leak race wss attained and to identif y all significant required corrcetion f actors such as those associated with humidity and barometric pressure, and all significant errors such as those associated with instrumentation sensitivitics and data scatter.
4.4.1.2 Local Leakate Rate Tests 4.4.1.2.1 Scope of Testing The local Icak rate shall be measured for each of the following components:
(1)
Persennel hatch D ff!ff3 ]))
(2)
E=crgency hatch
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(3)
Equipment hatch seals (4)
Fucl transfer tube scals 0Fyl q
(5)
Fuel transfer tube chroud bellows
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(6)
Reactor Building normal sump drain line i
(7)
Reactor coolant pump seal water outlet line (8)
Reactor coolant pump scal inlet line (9)
Reactor Building equalizing line (10)
Decay Heat renoval inlet lines l
(11)
Reactor Lui3 ding spray inlet lines (12)
High pressure injcetion lines (13)
Electrical pencerations (14)
Reactor tuf 3 ding purge inlet line (15)
Reactor Building purge outlet line (16)
Reacter Building atmosphere nample lines (17)
Letdown to purifiention deminerali ce line (3S)
Pressut izer relief tank ras caepic liue (19)
Resetor coo 3 ant system vent header Amendment No.11 4-17
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~ RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Surveillance Standards
. (20)
Pressuri:er relief tank nitrogen supply line-(21)
Pressurizer scmple line (22). Reactor coolant drain tank header 4.4.1.2.2 Conduct of Tests-(a)
Wi th the exception of the personnel and emergency hatches' door seals, all local leak rate tests shall be performed at a pressure' of not less than 52 psig.
See pa ragrapn 6.4.1.2.5c.
for personnel and emergency hatch dcor seal test pressure.
(b)
Acceptable methcds of testing are halogen gas detection, soap bubbles, pressure decay, hydrostatic ficw or equivalent.
4.4.1.2.3 Acceptance Criteria The total leakage fecm all penetraticns and isolation valves shall not exceed 0.06 percent of the Reactor Suilcing atmosphere per 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.
4.4.1.2.4 Corrective Action and Retest (a)
I f a t any t ime i t is determined that the criterion of 4.4.1. 2.3 above i s exceeded, repai rs sha l l be initiated immediately.
(b)
If conformance to the criterien of 4.4.1.2.3 is not demon-strated within 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> felicwing detection of excessive local leakage, the ' reactor shall be shut down and depres-surized until repairs are effected and the local le:kage meets the acceptance criterion as demonstrated by retest.
4.4.1.2.5 Test Frequency Local leak detection tests shall be perforned at a frequency of at least each refueling interval, except that:
(a) The equipnent hatch and fuel transfer tube saals shall be additicnally tested after each opening.
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(b) The personnel and e=ergency hatches shall be tested betwgen the inner and outer doors at a pressure not less than 52 psig semi-annually.
(c) *The personnel and energency hatches' inner and outer door 0-ring seals shall be tested within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> af ter each cpening when contain=ent integrity is required in Specification 3.6.1.
Test pressure for the personnel and eme_rgency hatgnes _0- ring seals shall be 10 psig.
- Exe=ption to Appendix J of 10 CFR 50.
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Anend=ent No. 11-n'
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS (1) The leah rate (L ) ectablished at the reduced pressure of 10 psig t
shall be extrapolated to the leak rate (L ) that will occur'at a
the calculated peak containment. pressure of 52 psig using the following formula:
La = 5. 2 L e (2) The extrapolated leak rate L will be added to the local leak a
rates established for the other components, and the total must meet the criterion of 4. 4.1. 2.3.
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Amendment No..Il
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m RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.4.1.3 Isolation Valve Functional Tests Every three months, remotely operated Reactor Building isolation valves shall be stroked to the position required to fulfill their safety function unless such operation is not_ practical during-plant operation. The latter valves shall be tested during the next shutdown of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> duration.
4.4.1.4 Annual Inspection A visual examination of the accessible interior and exterior surfaces of the containment strue:ure and its ce=ponents shall be perfor=ed annually and prior to any integrated leak test, to uncover any evidence of deterioration which may affect either the contain-ment's structural integri:y or leak-tightness. The discovery of any significan: deterioration shall be accompanied by corrective actions in accord with acceptable procedures, nondestructive tests, and inspections, and local tes:ing where practical, prior to the conduct of any integrated leak test.
Such repairs shall be reported as part of the test results.
- 4. 4.1. 5 Reactor Building Modifications Any major codification or replacenent of componen:s affecting the Reac:or Building integrity shall be followed by ei:her an in:egrated leak race test or a local leak test, as appropria e, and shall =eet the acceptance criteria of 4.4.1.1.5 and 4.4.1.2.3 respectively.
Bases The Reactor Building is designed for an internal pressure of 59 psig and a steam-air mixture te=perature of 286 F.
Prior to initial operation, the con-tainment will be strength tested at 115 percent of design pressura. The containment will also be leak tested prior to initial operation at, Pp and Pt (52 psig and 26 psig, respectively). These tests will verify that the leakage rate frem Reactor Building pressurization satisfies the relationships given in the specification.(1)(2)
The performance of a periodic integrated leakage race test during plant life provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of :he containment.
In order _to provide a realistic appraisal of the integrity of the contain=ent under accident conditions, this periodic test is to be performed without pre-
.liminary leak detection surveys or leak repairs, and containment isolation valves are to be closed in the normal =anner.
The reduced tes: pressure of 26 psig for the periodic integrated leakage rata tes: is sufficiently high to provide an accurate ceasure=ent of the leakage rate and i: duplicates the pre-operational leakage race test a: 26 psig. The specification provides a d
relationship for relating the =easured leakage of air at 26_psig to the potential leakage a: 52 psig. The =inimum of 24 heurs was specified for the
-integrated leakage rate test to help stabilize conditions and D. -
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m RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 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 on three major considerations. First is the low probability of leaks in the liner, because of conformance of the complete containment to a 0.10 percent leakage rate at 52 psig during pre-operational testing and the absence of any significant stresses in the liner during reactor operation. Second is the more frequent testing, at 52 of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value 0.06 percent of leakage that is specified as acgeptable from penetrations and isolation valves. Third is the tendon stress surveillance program which provides assurance that an important part of the structural integrity of the containment is =aintained.
More fr'.quent toting of various penetrations is specified as these locations are more susceptible,to leakage than the Reactor Building liner due to the mechanical closure involved. Particular attentica is given to testing those penetrations with resilient sealing materials, penetrations that vent directly to the Reactor Building atmosphere, and penetrations that connect to the reactor coolant system pressure boundary. The basis for specification of a total leakage rate of (0.075 percent) from penetrations and isolation valves is that approximately three quarters of the allowable integrated le.akage rate should be from those sources, in order to provide assurance that the inte-grated leakage rate would remain within the specified limits d 4ing the inter-vals between integrated leakage rate tests. Valve operability tests are specified to assure proper closure or opening of the Reactor Building isolation valves to provide for isolation of functioning of safety ' features systems. Valves will be stroked to the position required to fulfill their safety function unless it is established that such testing is not practical during eperations.
The airlock seals are tested at 10 psig because that is the manufacturer's recommended pressure for reverse flow through the seals. The extrapolation formula is derived assuming laminar, incompressible flow and provides con-servative leak rates.
This specification complies with the AppenZix J to 10CFR50 as published in the Federal Register on February 23, 1973, with the exemptions to Appendix J granted July 13, 1977.
REFERENCES (1)
FSAR, paragraph S.2.1.1.1 0
VV (2)
FSAR, section 14 Amendment No. 11 4-20
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