ML20098B003
| ML20098B003 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 02/14/1973 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20094C015 | List: |
| References | |
| FOIA-95-262 NUDOCS 9510020029 | |
| Download: ML20098B003 (84) | |
Text
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_?AFE'"? E"AL*JA~~0:I 3Y EE 2* REC *0RA'"E 07 ' ::E:S!' C
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- 5. A'"CMIC DTERCY C0!^d.ISE* !!
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_?23RASKA FU3LIC ?OWER JI5 2!CT 3
C00PDt !!UCLEAR STAT *C?T
'EMARA CCU? m, E3RASEA DOCKE"' :D. 50-298 03 Lf1 N
O ISSUED: TIERUARY 14, 1973 9510020029 950i
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Criterion 66 -- Prevention of Fuel Storage Criticality Appropriate station fuel handling and storage facilities are provided to preclude accidental criticality for spent fuel.
The new fuel storage vault racks (located inside the reactor building) are top entry, and are geometrically designed to prevent an accidental critical array, even in the event the vault becomes flooded.
Vault drainage is provided to prevent possible water collection.
References:
Subsections VII-6, X-2 and X-3.
Criterion 67 -- Fuel and Waste Storage Decay Heat The spent fuel pool cooling system is designed to remove decay heat to maintain the pool water temperature.
The fuel storage pool contains sufficient water so that in the event of the failure of an active system component, sufficient time is available to either repair the component or provide alternate means of cooling the storage pool.
References:
Subsection X-5.
Criterion 68 -- Fuel and Waste Storage Radiation Shielding The handling and storage of spent fuel is done in the spent fuel storage pool.
Water depth in the pool is maintained at a level to provide sufficient shielding for normal reactor building occupancy (10CFR20) by operating personnel.
The spent fuel pool cooling and demineralizer system is designed to control water clarity (tc allow safe fuel movement) and to reduce water radioactivity.
Access-ible portions of the reactor and radwaste buildings have sufficient shielding to maintain dose rates within the limits of 10CFR20.
References:
Subsections IX-1 through IX-4, X-3, X-5, XII-2 and XII-3.
Criterion 69 -- Protection Against Radioactivity Release From Spent Fuel and Waste Storage The consequences of a fuel handling accident are presented in Subsection XIV-6 of the CNS-SAR.
In this analysis, it is demonstrated that undue amounts of radioactivity are not released to the public.
All spent fuel and waste storage systems are conservatively designed with ample margin, to prevent the possibility of gross mechanical failure which could release significant amounts of radioactivity.
Backup systems such as floor and trench drains are provided to collect potential leakages.
The fuel handling and waste disposal systems are described ir..ections X and IX, respectively.
Operators are rigorously trained and administrative procedures are strictly followed to re-duce the potential for human error.
The radiation monitoring system as described in Subsections VII-12 and VII-13 of the CNS-SAR is designed to provide station personnel with early indication of possible station malfunctions.
References:
Subsections V-1, V-2, V-3, IX-2 through IX-4, X-2, X-3, X-5, X-14, XII-1, XII-2, and XIV-6.
F-2-17
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T.A3LE OF CONTENTS (cone.)
Ps ee 6.0 ENGINEERED SAFETT FEAIURIS.........................
f 6.1 Ge n e ra1....................................
6.2 6-1 l'
Co n tainmen t S ys t e ms.......................................
t
- 6. 2.1 6-1 Tunc tio nal 0as ign.................................
l 6.2.2 6-1 Iso la cio n Sys c a =s......................
6.2.3 Laaka ge Testin g Pro gram...........................
5-*
6.2.4 6-7 A c:io s ph ere Con t ro1................................
6-8 6.2.5 Va cuum B rea ke r Va1ve s.............................
6 6.2.6 Sec ondary Con tainment...........................
t
- 6. 3 0-12 Emergency Co re Cooling Systems i
Gen e ra1.................(.IC CS ).............
[
6.3.1 g
6.3.2 6-14 aigh Pressure Coolanc Inf eccion System (HPCIS).....
6.3.3 6-17 Automatic Depres surizacion System (ADS)............
6-18 3
6.3.4 Low Pressure Coolant Injection (LPCI) S 0-18 6.3.5 Co re S pray Syst em...................... ys tem.......
- h 6.3.6 6-19 Discussion o f ECCS Ray 1ev..........................
- 6. 4 6-19 S cand by Ga s "re a cma n c S ys t e m..............................
O 6-22 7.0 riSTRUMENTATIO N AND CONTROLS....................
7.1 Genera 1....................................................
7.2 Plant P ro cac cion and Co ntrol Syste =s.......................
7-1 4
7.2.1 7-1 In t ro duc t io n......................................
~
7.2.2 7-1 Incident and Accident Surre111ance Inscrumentation.
7.2.3 7-2 Reacco r High Water Level Isolation Signal..........
O 7.2.4 7-3 Average Power Range. Monitor (APRM) Reactor Trip in m
S ea m, aan g.....................................
7.2.5 Diversification of Signals to Iniciace Emergency 7-3 FE N
Cor. Coo 11ns.....................................
r1 L
7.2.6 7-4 Arznunciation of Engineered Safety Feature Bypasses.
O 7.2.7 7-4 Scandby Gas Treatmenc System (SGTS) Instrumentation 7-4 l
7.2.8 Anticipated Transient Without Sc ram (AIWS).....
7-5 l
7.2.9 Cond ens ar Lo w 7acuum Tr1p.........................
-*~
7.2.10 7-6 Turbina to Reactor Procaction System (RPS)
In te rf a c t...................
t 7.3 Se pa ra tio n C ritar1a.......................................
7-4
_a 7.4 I~
Environmental Qualifications..............................7-6 7.4.1 7-7 Saismic Qualific atio ns............................
m Radiation Qualifications..
7.4.2 7-7 7.4.2 m ar E m _ m.nce cuum.........................
7-7 camons.................
7-e 8.0 ELECTRIC P0VER...................................................
g 8.1 Genera 1......................
w S.2 O ff s ite Po we r.............................................
3-1 8.3 On s i c a Po wer..............................................
8-1 8-3 i
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1 f ailure of the =ain scea:tlines outside che containmanc or the l
curbine-condenser and because of the conservative nature of the i
scaff's analysis of the dose consecuences, yelleving our review I
and approval of the i=oreved MSL 7 surveillance cesc program, l
1 the appropriace porcions of that case program will be included l
in the Technical Specificacions.
5.'. 2
- aakare Tescing ?reers=
_ "he pri=arv concM n-anc and comoonents vnich will be suojecced _
co concain=ene cast concicions were designed so chat per:,cdic ince-graced leakage race ces cing can be conducted ac peak calculaced accident pressure and reduced pressures.
e have reviewed che pro-posed cesc procedures for determinacion of the pr1=ar7 containmenc overall leakage, as well as penetracion and isolacion valve leakage, for both preservice and inservice containment lea 4 age ces es.
Penetrations, including personnel and equip =ent hatches. and C3 l.
airlocks, and isolacion valves, have generally been designed with che i
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capability of being individually leak cested at peak calculated acci-L.
c3 donc pressure.
Large hatches have been strengthened structurally to sus tain the pressures of individual leak casts. Syste=s designed prior to che i=nlementation of Appendix J, such as the control rod drive penetrations and scandby liquid control system, do not have design provisions for individual leak cases; however, che nor=al functional testing of chase systems ensure chair operability and chance the necessary concMnnent integrity.
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S 6-8
'Je conclude that design of the primary containment system will permit the conduct of a containment leakage testing program in compliance with the requirements set forth in proposed Appendix J co 10 CTR Part 50, "Raacter Containment Laakage Testing for Water Cooled Power Reactors" (36 Fed. Reg. 17053, Aug. 27,1971).
6.2.4 At osenere Centrol As an operational technique to preclude flammable gas concentra-tions, the primary containment will be operated with an inert nitroge atmosphere. The system will maintain the oxygen content of the containment at=esphere below 4 volu=e percent and we find it acceptah O
Following a less-of-coolant accident (LOCA), (a) hydrogen gas could be generated inside the prinarv containment from a chemical reaction between the fuel rod cladding and steam (metal-water reaction), and (b) both hydrogen and oxygen would be generated as a resu.1: of radiolytic decomoosition of recirculating water.
If a s uf ficient amount of the hydrogen is generated and oxygen is avail-g N
able in stoichiometric quantities, the subsequent reaction of C
hydrogen with oxygen can occur at rates rapid enough to lead to a signifi cant pressure increase in the containment. 31s could cause damage to the containment and could lead to failure of the containmen to maintain low leakage integrity.
General Design Criterion 41 of Appendix A to 10 CTR Part 50 requires that systems to control hydrogen, oxygen and other substance
i 7 / 5' CNS ENFORCEMENT CONFERENCE S
PREBRIEF CONTAINMENT ?.MTE6!!Irf ATTACHMENT A Draft NOV ATTACHMENT B NRC Inspection Report 50-298/94-14 ATTACHMENT C Enforcement History ATTACHMENT D Systematic Assessment of L1centec Per#ormance.
ATTACHMENT E Technical Specification snd T.S. Basis ATTACHMENT F USAR and Related Commitm nt:
ATTACHMENT G CNS Flow Diagram 2028 an') Walkdown S; eats ATTACHMENT H LLRT Results through 7/11/24 ATTACHMENT I Licensee Event Report 94-C" ATTACHMENT J GE Design Specification i
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DRAFT m.
COOPER NUCLEAR STATION 9414 01 10 CFR Part 50. Appena1x B. Criterion III, states, in part, that "[m]easures shall be established to assure that
.. the design basis are correctly translated into
. specifications. drawings These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled."
1.
The Cooper Nuclear Station Updated Safety Analysis Report. Appendix F.
"Conformance to AEC General Design Criteria." sates. in part. the "
the purpose of this appendix [is] to show that the design and construction of the Cooper Nuclear Station has been performed in accordance with these gerieral design criteria."
Contrary to the above. Flow Diagram No. 2028. " Reactor Building and Drywell Ecuipment Drain System." contained safety-related isolation valves but..as not included on the safety-related drawing list as of July 1. 1994. and some safety-related components were not included on the drawing.
2.
Draft General Design Criteria. Criterien 53. July 1967. in accordance with Appendix F to the USAR. states that "[a]ll lines which penetrate the primary containment and which communicate with the reactor vessel or the primary containment free space [were] provided with at least two isolation valves (or equivalent) in series."
1.
Contrary to the above. as of May 14. 1994 many penetrations were identified without redundant valving.
These penetrations incluaed. but '. sere not limited to. penetrations X-21. X-22. X-25.
X-29E X-30E/F, X-33E/F. X-209A/B/C/D. and X-218.
2.
Contrary to the above, as of February 22. 1994 ten manual operated vents. drains. or test connections had single manual valves for containment isolation.
3.
Draft General Design Criterion 1. in accordance with Appendix F to the Updated Safety Analysis Report, states that "... those systems and components of the station which [had] a vital role in the prevention or "11tigaticr " ::nsecuences of 3ccidents affecting the public health and safety [e.erej cesignea and ccnstructed to high quality standards General Electric Design Specification No. 22A1153. " Codes and Industrial Standarc.
Revisicn 1. states. in Note 3 of the Appendix, that
"[p]iping..snich is an integral part of the primary containment for isolaticn,:m roses. shall have at least the same quality and levels of assurance as the primary Containment."
Contrary to the above, the licensee failed to design, fabricate and erect approximately 300 containment penetrations to the standards specified in USAS B31.7-1969.
9414 02 Technical Specification 4.7. A.2.f.1 states, in part that " local leak rate tests (LLRT's) shall be performed on the ]rimary containment testable penetrations and isolation valves T1e total acceptable leakage for all valves and penetrations other than the MSIV's is 0.60 La."
1.
Contrary to the above, as of May 14, 1994 the licensee failed to provide for Type C local leak rate testing of 68 components passing through 54 containment penetrations.
2.
Contrary to the above. as of July 11. 1994, the total leakage for the valves and penetrations that had never been tested, with three tests remaining. exceeded the 0.60 La limit allowed by Technical Speci fications.
The 0.60 La limit was 5.37 scmh (189.60 scfh) and the leakage for the valves that had never been tested was in excess of 17.66 scmh (623.57 scfh).
3.
Contrary to the above. several instrument pressure switches had not had local leak rate testing performed after being isolated from the containment integrated leak rate test.
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' UNITED STATES.
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611 RY AN PL AZA ORIVE, SulTE 400 A R LINGTON,, T E X AS 76011 8064 N
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Docket:
50-298 License:
Nebraska' Public Power District ATTN: -Guy R. Ho'rn. ! ice President - Nuclear P.O. Box 499 Columbus. Nebraska 68602-0499 SUBJECTi NRC INSPECTION REPORT 50-298/94-14 This refers to tne inspection conducted by Ms. P. A. Goldberg and Mr. C. J. Paulk. of this office. and Mr. G. Cha, an NRC consultant. on June 13 i
througn August 'Z.
1994.
The inspection included a review of activ1tles authorized for your Cooper Nuclear Station facility.
At the conclusion of the inspection, the findings were discussed with you and those members of your
. staff identiflec in the enclosed report.
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Areas examined during the inspection are identified in the report.
Within these areas. the ansoection consisted of selective examinations of procedures
]
and representative records interviews with personnel, and observation of i
activities in oregress.
The purpose of the inspection was to determine i
whether activities autnorized by the license were conducted safely and in accordance with URC reau1rements.
4 Based 1x1 the resuits of this inspection, two apparent violations were identified and are being considerea for escalated enforcement action in accordance with the " General Statement of Policy and Procedure for NRC l
Enforcement Act :ns- (Enforcement Folicy). 10 CFR Part 2. Appendix C.
Accordingly. no.otice of Violation is presently being issued for these
. inspection findings.
Please be advised that the number and characterization
- of apparent vloiations described in the enclosed inspection report may change as a result 3f 's-ther "RC eview The-aooarent '/ations are of concern because it is aooarent that the crimary j
containment.sas inoperaole.ror an unaeterminea perloa of time.
Additionally.
it is apparent ~ rat there was a breakdown in your design control program.
dating back to :nitial construction, which you have had numerous opportunities
.to icentirj m u u m..:c;.
.ne apparent oreasccan in cesign controi contr1Dutea to.the problems issociated with the primary containmer' as well as other recem. ~. wen
.:d, ; a..:
- ne ;cper Nuciear Stat 1uti.
I in e'dc"ce:9?".
ecance.; cisc;ss these apparent vioiations nas Deen
'scneauled for. ;e::e cer '6.1994.
Tne decision to hold an enforcement
'.- s conTerence coes
- t'mean tnat the kRC has cetermined that a violation has
'occurrea cr "u: ""orcement acticn. vill te taken.
The purposes of this t
Nebraska Public Power District conference are to ciscuss tne apparent violations, their causes and safety significance: to provide you the opportunity to point out any errors in our inspection report; and to provide an opportunity for you to present your proposeo corrective actions.
In aodition, tnis is an opportunity for you to provide any information concerning your oerspectives on (1) the severity of the violation (s). (2) the application of the factors that the NRC considers when it determines the amount of a civil penalty that may be assessed in accordance with Section VI.B.2 of the Enforcement Policy. and (3) any other application of tne Enforcement Policy to this case. including the vercise of discretion in accordance with Section VII.
You will be advised by separate correspondence of the results of our deliberations on this matter.
No response regarcing ;nese apparent violations is required at this time.
This enforcement conference, whicn will also aooress issues involving the control room filtration system (EA 94-164) and the electrical distribution
- system (EA 94-166), aill be open to public observation in accordance with the Commission's cent 1 ruing trial program as discussed in the enclosed Federal Register Notices (Enclosure 2).
Although not required. we encourage you to provide your comments on how you believe holding this conference open to public observation affectea your presentation and your communications with the NRC.
In accordance with 10 CFR 2.790 of the NRC's " Rules of~ Practice." a copy of this letter and 'ts enclosures will be placed in the NRC Public Document Room.
Should you have any questions concerning this inspection, we will be pleased to discuss them with ycu.
Sincerely.
Thomas P. Gwynn. Director Division of Reactor Safety
Enclosures:
1.
Sooendix
- NRC Inseection Reece 50 298/94-14 2.
Federal Register Notices cc w/ enclosures:
Nebraska Public Power District ATTN:
G. D. Watson. General Counsei P.O. Box 499 Columous. 1eorasta docu2-0499
Nebraska Public Power District-Nebraska Public Pc..er District ATTN: Mr. John H. Mueller. Site Manager.
P.O. Box 499 Columbus. Nebraska 68602-0499 Lincoln Electric System ATTN:
Mr. Ron Stoddard lith and 0 Streets Lincoln. Nebraska 68508 4
Nebraska Department of Environmental Quality KATTN:
Randolph Wooa. Director P.O.. Box 98922-Lincoln. Nebraska 68509-8922 Nemaha County Boarc of Ccmmissioners ATTN: Larry Bohlken, Chairman Nemaha County Courtnouse 1824 N Street Auburn. Nebraska 68305 Nebraska Department of Health ATTH:
Harold Borcrert. Director i
Division of, Radiological Health i
301 Centennial Mai!. Soutn P.O. Box 95007 Lincoln, Nebraska 68509-5007 Department of Natural Resources ATTN:
R. A. Kucera. Department Director-1 of Intergc.ernmental Cooperation i
P.O. Box 176 Jefferson City, Missouri 65102
. Midwest Power i
ATTN:
Mr. James C. ?arker. Sr. Engineer 1
907 Walnut Street r.0,160x 657 Des Moines. Iowa 50303 Kansas Radiation Ccntrol Program Director
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i
Nebraska'PublicPower; District- -
E-Mai1Lreport.:: D. Sullican'(DJS) bcc to DMB (IE01)-
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bcc distrib. by RIVi Resident ~ Inspector.
.L J.l Callan '
Leah Tremper. 0C/LFDCB-. MS: MNBB 4503 Branch Chief-(DRP/C)
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DRSS-FIPB MIS' System-.
. Project Engineer-(DRP/C).
Branch Chief--(CRP/TSS)
RIV File Senior Resident. Inspector - River Bena
.J-Lieberman. DE. MS: 7-H-5
- T. F. Westerman P. Goldberg C; Paulk A. Howell
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08/08/94-08/05/94 08/18/9a 08/19/94 08/23/94 09/01/94
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1
l APPENDIX U.S. NUCLEAR REGULATORY COMMISSION
- EGION I'!
Inspection Report:
50-298/94-14 EA No..94-165 License:
DPR-46 Licensee:
Nebraska Public Power District P.O. Box 499 Columbus. Nebraska Facility Name:
'ccoer 'Juclear Station Insoection At:
Brc..nville. Nebraska Inspection Conducted:
June 13 through August 12. 1994 Inspectors:
P. A. Goldberg. Reactor Inspector. Engineering Branch Division of Reactor Safety C. J. Paulk. Reactor Inspector. Engineering Branch Division of Reactor Safety accompanied By:
- 1. Cha. Consultant Approved:
I. F. Westerman. Chler Engineering Brancn Date Division of Reactor Safety Insoection Rummarv Areas Inscectod-Reactive. announced inspection of the l'msee's actions concerning containment oenetration oroblems found as the t of reviews and Inspections performe0 Dy the licensee.
In addition. 1ssues telated to motor-operated valves and switch calibration f^r drvwell instrmontat,en pro revieweo.
Results-4 3 esu1*
- r ecti ce 3 ::Ons for 3 previcu;b 'dantified vicl3ticn.
the licensee was reviewing the design function of all pipincj and ecu::T.ent :rs: Ore carts to cetermine if they.,ere properly classified.
This effort c.as scneauleo to be completed in October 1994 and will be gyp,.3 tea c....7 <-n
.,uo c' in'c cement Acticn 93-137 (Section 2.1).
n l
l 2
The licensee preparea 15 design change packages to bring the containment penetrations into compliance with the draft General Design Criteria.
Criterion 53. July 1967. as stated in Appendix F to the Updated Safety Analysis Report. and 10 CFR Part 50. Appendix J.
Seven of these design change packages were reviewed and no concerns were identified (Section 2.1.1).
During the inspection. the inspectors found that Flow Diagram No. 2028.
which depicted 80 safety-related components. was not accurate since it failed to include some safety-related components.
The failure to include containment isolation valves on the drawing and the failure to identify the drawing as safety-related was identifled as an apparent violation or
'.0 :FR Part 50. Appendix B. Criterion III (Section 2.1.2)
The licensee cetermined that the containment isolation valves in 54 penetraticns naa not had Type C local leak rate tests performed on 68 of the cc.Tocrents ; ass 1nc ?.rcugn the penetrations.
The systems associated witn these valves were classified as nonessential.
- However, the containment isolation valves were required to function to prevent the release of the post accident containment atmosphere.
The failure to perform Type C local leak rate tests was identiflea as an apparent violation of Tecnnical Speci fication 4.7. A.2.f.1 (Section 2.1.3).
The total leakage of the local leak rate tests performed on components previously ^n Iasted exceeced the Technical Specification limit for leakage to enr.ure containment integrity.
This was identified as an apparent siciation of Tecnnical Specification 4.7.A.2.f.1.
(Section 2.1.3),
The licensee 'dentified a number of examples where penetrations were found to lack redunaant containment isolation.
The failure to have redundant containment 1 solation barriers was identified as an apparent violation of 10 CFR Part 50. Appendix B. Criterlon III (Section 2.1.4).
The licensee identiflea approximately 300 examples of components associated with containment penetrations which were not classified as essentlai.
~~.e ' allure u :esign. fabricate, and erect the containment isolaticn barriers to quality standards that reflected the importance of the safety incticn c.as ident1fio: : an apparent 'ncl tion of 10 CFR 3
Part 50. Appena1x B. Criterion III (Section 2.1.5).
The licensee cetermined that Containment Isolation Valve RHR-MOV-M027B was not capaole of passing its local leak rate test.
The licensee decided *:
.o
"'o crinar" contair - isolation ' unction &cm the leaking vai,e to anotner vaive.
This cnange was accomplishea by use of
3 a safety evaluation that was cerformed in accordance with 10 CFR 50.59.
It was conciuaea that tne licensee's change of primary containment isolaticn ocuncary was adequately justified and appropriate procedural controls..4ere identi fied (Section 2.2).
During a review of the licensee's actions concerning the lack of cleanliness inside motor-operated valve limit switch compartments, it was found that the licensee had not entered the recommended corrective actions into the corrective action tracking system.
This was a concern because of the lengthy amount of time allowed to pass before the corrective actions were due v.nich increased the chances for similar events to occur (Section 2.2)
The failure to perform local 'eak rate testing for several instrument pressure switches was identified as an apparent violation of Technical Speci fication 4. 7.4.2. f.1 (Section 2.3).
Unresoivea item 298/9403-01. :encerning ten valves used as single manual valves for containment isolation. was closed.
These ten single isolation valves without a second barrier were identified as another j
example of an apparent violation of 10 CFR Part 50. Appendix B.
Criterien !!! (Section 2.4).
Summarv of Inscectwn Finriinas-Example i of apparent Violation 298/9414-01 was identified (Section 2.1.2)
Example 2 of apparent Violation 298/9414-01 was identified (Sections 2.1.4 ano 2.4).
Example 2 of apparent Violati:n 298/9414-01 was identified (Secticn 2.1.5).
Example of apparent Violation 298/9414-02 was identified (Section 2.1 3).
Example 2 of apparent Violation 298/9414-02 was identified (Secticn 2.1.3).
Examole 2 cf apparent Violat':n 298/9414-02 was identified (Secticn 2 3).
- n:cact":- : llc..uc ::5-2?S 3214-03..as openec :3ecc on 2.2.2).
Ur e:c'.:: :t r 2?:.22:2-:1.as cicsea ::ec:1sn 2.4).
..s i + +, c h m.w Attacnmer:
Derscns 2:ntactea and Exit Meeting
.a.
DETAILS 1 PLANT STATUS During this inspection perica. Cooper Nuclear Station was shutdown.
1 2 ENGINEERING (37550 and 92903)
This inspection..as conductea to review Cooper Nuclear Station's actions concerning problems found with ccntainment penetrations.
In addition, licensee ~s actions concerning dirty torque switches on motor-operated valves and time-delay relays for tre emergency diesel generators were reviewed.
The inspectors reviewed the licensing basis for the Cooper Nuclear Station in order'to evaluate tne problems associated with the containment penetrations 4
against the apprcoriate criteria.
The inspectors found that the licensee was committed to the araft " General Design Criteria for Nuclear Power Plant Construction Permits.' 1ssuea in July 1967.
This commitment was documented in the Updated Safety Analysis Report (USAR). Appendix F.
The licensee was evaluated and licensed to the draft General Design Criteria. July 1971 and 10 CFR Part 50. Appena1x J. as stateo in Sections 3.1 and 6.2.3. respectively. of
" Safety Evaluaticn oy the Directcrate of Licensing U. S. Atomic Energy Commission in the Matter of "ebraska Public Power District. Cooper Nuclear Station. Nemaha County. Nebraska. Docket No. 50-298." dated February 14. 1973.
The inspectors a'.sc fcuno tnat tne licensee acknowledgea the applicaoliity of the draft General Design Criteria in the draft design criteria document prepared for the :;ntainment systems.
With regara to tac _rolicaD1iity of 10 CFR Part 50. Appendix B. to Cooper Nuclear Station. 10 CFR 50.54(a)(1) requires. that eacn plant licensed suDject to the quality assurance criteria in Appendix B shall implement pursuant to 10 CFR 50.34(b)(5:
the cuality assurance program described or referenced in the safety analysis report.
The final 10 CFR Part 50. Appendix B. rule was issuea on June 27 1970. ano the coerating license for Cooper Nuclear Station was issued on January 18. 1974.
On the basis of *ce 9.tove. : e ins;ectors reviewed the containment penetration issues against the craft General Design Criteria. July 1967. as described in the USAR. Appenan ?
10 CFF Part 50. Appendix B: 10 CFR Part 50. Appena1x J:
and, applicable licensee procedures. design specifications, and Technical Speci fications.
2.1 Containment Panetrations The licensee creoarea Special Procedure 94-202 dated May 17, 1994.
" Containment c.alk :<.n.' :: ' sce:t eacn primary containment penetration and the piping to tne cutboara ccntainment isolation barrier.
The purpose of the inspection'<.as t:
..coort cr.>elocrent of tne containment design criteria
5-document: to comply with a commitment. made in response to a violation in NRC Inspection Report 50-298/93-17. to review all containment penetrations: and, to support the upgrade of the licensee's program for primary reactor containment leakage testing in accordance with 10 CFR Part 50. Appendix J.
The inspectors reviewed Special Procedure 94-202 and found that licensee inspection of each primary containment penetration. and components which were in the containment isolation system was required.
This inspection was also in support of the preparittion of as-built drawings.
.he inspectors concluded that the procedure was adequate.
During the inspections. the licensee oetermined that 46. of the 255 primary containment penetrations inspectea. had been incorrectly classified
' nonessential at the time of plant construction and were not contained in the inservice inspection program.
In addition:
The licensee :etermined that a number of penetrations had not had local leak rate tests performed in accoraance with the requirements of 10 CFR Part 50. Appendix J.
A number of penetrations did not have two containment barriers outside of the primary containment in accordance with draft General Design Criteria. :r::aricn 53. July 1971.
A number of instrument lines and valves within the containment pressure boundary were classified as nonessential.
294 welds in the containment isolation barriers were found to either never nave hac nondestructive examinations performed or the qualification records could not be located.
Many penetrations..ere improperly classified during the construction of the plant.
The inspectors attempted to determine how such a problem occurred.
While no definite answer was provided the licensee stated that the architect engineer apparently had missed a note in the General Electric design specification which resulted in the improper classification of containment penetrations and asscciated components.
The inspectors founo that eculpment and components classified as essential were designeo. fabricatea. Installed, and tested in accordance with USAS B31.7-1969. " Nuclear Power Piping."
Equipment and components classified as nonessential ciere deslaned. fabricated. installed. and tested in accordance with USAS C31.1.0-1967. " Power Piping."
On the basis of these codes, the architect engineer designed the equipment and components.
The arcnitect engineer however, apparently missed a note in General Electric Design Scecification 22A1153. " Codes and Industrial
. fy.
Standard." Revisico 1.
Note 3 of five. to this specification. stated that
"[p]iping, wnico [.vas] an integral part of the primary containment for isolation purposes. shall have at least the same quality and levels of assurance as the crimary containment."
In Appendix A of tr.e Upaated Safety Analysis Report. the licensee provloed definitions for the classification of piping and equipment pressure parts.
Class C was assignea for "[p]iping ano equipment pressure parts for a high integrity system." Such as the containment vessel.
To meet this classification, the licensee applied the reau1rements of USAS B31.7-1969 for Class 11 piping.
rherefore. the penetration piping and equipment pressure parts should have oeen designed, fabricated. installed, and inspected accordingly.
As a result of corrective actions for a previously identified violation. the licensee was revleenng the design function of all piping and equipment pressure parts to :etermine if they <.ere properly classified.
This effort was scheduled to be completed in October 1994 and will be evaluated during followup of Escalated Action 93-137 for violations cited in NRC Report 50-298/93J.~
The inspectors coservec 17 liquid penetrant tests of welds that were originally desagr.ea. fabricated. installed. and tested in accordance with USAS B31.1.0-1967 rather than USAS B31.7-1969.
The inspectors observed one weld that exn101 tea inoication or weia slag.
The licensee rejected that weld.
Subseauently. the ':censee chipped the. veld slag off and retested the wela satisfactcrily.
The licensee como:etea tne liquid penetrant testing on 260 welds that had been improperly classified without identifying any other weld that was questionable.
The inspectors concluded that the licensee had performed testing in accorcarce.vith USAS B31.7-1969 for the weids that had no documentation of such inspection during the construction of the plant.
2.1.1 Design Moci fications To address the ccncerns identified by the licensee's inspections of primary containment penetrations. design changes were prepared.
The inspectors reviewed the ~ _:s.@ sc age pacs 6ges. ;15c a cc f;110 wing sections.
out of a total of 15 anich the licensee was preparing to bring the penetrations in:c _.~... _
.~....;
. n : ar. era. _esign Criteria.
Criterion 53. as stated in the USAR. Appendix F. and 10 CFR Part 50.
Appendix J.
During the licensee's verification and validation of the draft desian criteria :::ument for the primary containment. the identification of problems led to a compiete scrutiny of all penetrations (approximately 300).
As a result of !ne licensee's efforts. 99 penetrations were identified with oroblenis other van classification The problems were categorized into 11 types, wnicn rangec frcm missing caps to inaaequate design.
.7-
'2.1.1;1 Design Change 94-212 Torus Penetration X-218 Modification Penetration X-218. as-found. consisted of'a ball valve on the torus shell with eight thermocouples routed through it.
A sealant of unknown composition filled the vold and acted as a containment barrier.
The thermocouples were installed under Design Change 76-l. Revision 2. but were never placed 1r.
service.
The cesign change was later voided because there were no provisions to calibrate tr.e temperature elements ana the equipment was abandoned in place.
The penetration was not local leak rate testable, and was not on the local leak rate test list.
The design change consisted of removing all thermocouple hardware and the ball valve. and installing a 5.08 cm (2 in) socket welded cap. whicn would function as a primary containment 00undary. hence the penetration would be restored to its original design.
The design change was classified as essential and i
Seismic Class IS.
The 5.08 cm (2 in) socket welded cao was purchased as essential materiai.
The applicable design code for fabrication and installation was USAS B31.7-1969.
eeld integrity was checked by 100 percent liquid penetrant nondestructive examination and pneumatically tested to 1.25 of design pressure.
The results of the liquia penetrant tests were alscussed in Section 2.1 of this report.
The inspectors cia not identify any concerns with this design change.
2.1.1.2 Design Change 94-212A Electrical Penetrations X-209A through D Modi n caticns Design Change 94-212A consisted of two parts:
the first part, associated with Penetrations X-209A and X-209C. involved modifying the two thermocouple penetraticns :: cer"'it per'cdic local leak rate testing as required by 10 CFR Part 50. Accendix.;: and the second part. associated with Penetrations X-209B and X209D. involved permanently capping the two penetrations.
The inspectors aid not identify any concerns with this modification.
The inspectors reviewea Design Change 94-212A in its entirety, verified the design chances durinc *he walkdcwn. and concluded that
't was acceotable.
2.1.1.3 Design Change 94-212B Penetrations X-43 and X-44 Testable Flanges This design cnange replacea two fiangea piping joints near Penetrations X-43 and X-44 witn flanges incorporating a testable. double o-ring design.
The new design permitted Inese joints. which were Dart of the crimary containment
'bounaary. to c= pericalcally testeo in accoraance with~10 CFR Part 50.
Appendix J.
1 l
3-1 The design change was classified essential and Seismic Class IS.
All pressure I
retaining material e.as procurea essential.
The inspectors did not identify any concerns with tnis moalfication.
2.1.1.4 Design Change 94-212D Penetration X-21 and X-22 Upgrade The purpose of Design Change 94-212D was to enhance the isolation capacity for both the service 61r and instrument air heaaers, upstream of Penetrations X-21 and X-22 respectively.
Additionally. the modification provided test connections for ceriodically performing local leak rate tests of the containment isolation valves in accordance with 10 CFR Part 50. Appendix J recuirements.
LThe inspectors ma at :centify any concerns witn this moaification.
2.1.1.5 Design Change 94-212E Primary Containment Integrity Issues Design Change 92-212E consistea of three parts.
The first part removed Swagelok caps and installed valves and caps at ten test connections for instruments whicn <.ere in direct communication with primary containment.
The ten test connections. vere PC-PT-182. -4B2. -582. -1A1. -4A1. -5A1. -2104A.
-21048: and PC-PI-2104AG. -2104BG.
Also, at PC-DPT-3A1.
Drain Valve PC-V-243 was missing ana was reinstalled.
The second part of the modification removed unnecessary tees lccated in instrument lines which communicated directly with primary containment and replaced them with unions. elbows or installed welded caps.
The th1:a part of the moaification cut and tapped 14 instrument lines which penetratea crimary containment and had previously been spared out.
The valves were removeo ano neidea caos installed on the lines at the penetrations.
The inspectors aid not identify any concerns with this modification.
2.1.1.6 Design Charge 94-212H Post Accident Sampling System Modifications and Penetration X-51F Upgrade The purpose of Design Change 94-212H was to replace the existing nonessential post-acc10ent ;amoling system Containment Atmosphere Sample Isolation Valve PAS-A0V-3AV n th teo qualified 1.27 cm (0.5 in) air-operated valves.
PC A0V-247AV anc PC 10V-248AV. at Penetration X-51F.
In addition. test connections aitn cacpea manual valves were providea.
The insoectors cic 'ot identify any concerns with this modification.
2.1.1.7 Mainterance c.ork Requests 94-2978 and 94-3116
~
These maintenance c.crs reauests instailea caps ana piugs to provice the secona barrier for containment isolation.
During the licensee's inspections, numercus
,e.;a.
as 2.
a.m ;est m;nnections naving 01 rect access to tne primary ccntainment c.1ere,found to lack a second barrier.
These were identitlea ano
- a; ;r piug.s6s acaea. Cepenalng on tne installation.
i
9 The inspectors reviewea Maintenance Work Requests 94-2978 and 94-3116 and concluded that both were acceptable.
2.1.2 Drawing Control During a review of the penetration walkdown packages, the ins)ectors noted that some of tne ccntainment isolation valves identified on t1e penetration drawings, and existing in the plant. <<ere not included on Flow Diagram No. 2028. " Reactor Building and Drywell Equipment Drain System." Revision N27.
The inspectors fcund that Flow Diagram No. 2028 was not included on the safety-related drawing hst in accordance with Cooper Nuclear Station Engineering Procedure 3.8. " Drawing Control Procedure." Revision 7.
The inspectors concluded that the drawing was inaccurately classified as a result of the problems associated with classification of components as discussed in Section 2.1. above.
Cooper Nuclear Station Engineering Procedure 3.8. " Drawing Control Procedure."
Revision 7. definea a safety-related drawing as "a drawing or schematic describing the features. characteristics, design or location of safety-related components, systems. Or structures." The procedure also stated that any new drawing, or porticn of a new drawing. classified as safety-related would be added to the safety-related drawing list.
During the inspecu en, the licensee initiated Condition Report 94-0309 in response to the inspectors' finding.
The condition report stated that the subject drawing aepictea a total of 80 safety-related components. but was not included on the safety-related drawing list.
In response to this condition report, the licensee identified an additional 13 drawings, with safety-related components, that c.ere not includea on the safety-related drawing list.
Additionally' Draft General Design Criteria. Criterion 1. July 1967. in accordance witn Appena1x F to the uSAR. stated that "
those systems ana components of the station which [had] a vital role in the prevention or mitigation of consequences of accidents affecting the public health and safety
[were] designed and constructed to high quality standards The inspectors identified five missing valves on Flow Diagram 2028.
These valves were associated with Penetrations X-18. X-30E X-30F. X-33E. and X-33F.
For Penetraticn 0 18. an unlabelea vent isolation valve downstream of Valve RW-254 was not on the drawing.
For Penetration X-30E, Valve NBI-502.
the manual containment 1 solation valve for the air-to-vessel flange leak off detection air-cperatea valse..sas not snown.
For Penetration X-30F.
Valve MS-900, the manual containment isolation valve for the air-to-reactor
, esse' teac.:r.t
.h cc; sr.c<.n.
r;r Fenetration X-33E. ialve MS-501, the manual containment dsolation valve for the air-to-vessel flange leak off Jetection air. dra:dQ <ai e. c.as rot snown.
For Penetration x-33F.
Valve MS-899. the manual containment isolation valve for the air-to-vessel
- -ad.er.t
.;; c ;. :...a
.:0-Appendix B to 10 CFR Part 50. Criterion III. requires that "[m]easures shall be establisheo to assure that the design basis
.. are correctly translated into drawings.
These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled."
The inspectors identified the licensee's failure to properly classify drawings as safety-related and the failure to include safety-related components on the drawing as Example 1 of an apparent violation of 10 CFR Part 50. Appendix B.
Criterion III (293/9414-01).
2.1.3 Local Leak Rate Tests The licensee determined that the containment isolation valves in 54 penetrations cid not have Type C local leak rate tests performed on 68 of the components passing through the cenetrations.
The systems associated with these valves v.ere classified as nonessential since they did not have to function post-accident.
Containment isolation valves, however, were required to function to prevent the release of the cost-accident containment atmosohere.
Containment 1 solation valves. as defined in 10 CFR Part 50. Appendix J. would be "any valve wnicn [was) relled upon to perform a containment isolation function."
Type C tests.erc recuired for containment isolation valves that " provide [d] a direct connecticn oetween tne inside and outside of the primary containment under normal operation
.. [were] required to close automatically upon receipt of a containment isolation signal and [were] required to operate intermittently under post-accident conditions."
In accordance with draft General Design Criteria. Criterion 57. July 1967. as stated in Appena1x F to the USAR. the licensee was required to demonstrate the functionai performance of containment system isolation valves and monitoring valve leakage."
Technical Specification 4.7. A.2. f.1 required that " local leak rate tests (LLRT's) shall be performed en the primary containment testable j
penetrations and Tsolation valves The inspectors icentified the failure to perform local leak rate tests as Example 1 of an apparent violation of Technical Specification 4.7. A.2.f.1 (298/9414-02).
The licensee had begun performing local leak rate tests on the identified components.
The 'nspectors attempted to review the results of this testing.
The licensee had not develooed a running total of the results of the as-found tests to cetermine tne status of Ine primary containment and its ability to perform as cesicnea.
The inspectors were informed that one penetration i.(-22). On June _3.
994. nad in excess of 17 scmn (600 scfh) leakage.
This significantlv mceeced Tcchnical Soecification 4.7. A 2.f 1 leakage limit of 0.60 La w.m scmn tis 9.o scih)/.
1
, The total leakage of the local leak rate tests being performed on containment isolation components not previously tested, v.ith three remaining to be tested.
was in excess of 17.66 scmh (623.57 scfh).
This value did not include any leakage from those components previously tested, nor did it reflect the actual leakage througn penetration X-22. which was listed only as greater than 17 scmh (600 scfh).
As noted above. Technical Specification 4.7. A.2.f.;
i established the limit for local leak rates to be 5.37 scmh (189.60 scfh).
This limit was established to ensure containment integrity.
The licensee haa mitiated a licensee event report on July 5.1994. to address the identifica = r cf cenetrations that had not been tested as requirea by 10 CFR Part 50. % cendix J.
The licensee stated that the causes would be addressea in a succiement to the report.
4
-On the basis of tre test results for the newly tested components, the inspectors concluaed that the licensee had exceeded the Technical Specification limit for leakage to ensure containment integrity for an extendea perloa c.itnout taking the required corrective actions.
As sucn. this is identified as Example 2 of an apparent violation of Technical Speci fication 2.~ - 2.'.1 (298/9414-02).
2.1.4 Redunaant 2ntainment Isolation Barriers The licensee mnected approximately 300 penetrations during the performance of Special Procecure 94-202.
A numoer of those penetrations were found to lack redundant containment Darriers.
The licensee laentifiea some penetrations with both 1 solation valves located outside the pr -arv containment.
However. between the containment wall and the first isolation valve outside containment. there existed a single vent.
drain, or test connection valve.
Examples of this type of single barrier were Penetrations v21.
-22. and X-25.
Some penetraticns c.ere ident1 fled by the licensee with only a single isolation valve outside of c:ntainment.
Penetrations X-29E. X-30E/F and X-33E/F were examples.
Penetrations X-215 and X-209A/B/C/D had thermocouple wires routed in piping through the penetrations.
On the outside of containment was an open valve.
incapable of c % -rg "ith an unidentified sealant that could not be determined to be qualified.
These penetrations were determined to have an unqualified barrier Appendix B t:.: ;9 Part 50. Criterion III, requires that "[m]easures snail be established to assure that the design basis
.. are correctly translatea.nto xecincations inese measures shall include provisions to assure that incroorlate cuality standards are specified and included in design cocument; s.a tnat ceviations from such standards are controlled."
3 -
Additionally. 'n accorcance with araft General Design Criteria. Criterion 53.
July-1967. as stateo ir. /,ppendix F to the USAR. "[a]ll lines which penetrate the primary containment and which communicate with the reactor vessel or the primary containment free soace [were] provided with at least two isolation valves (or eaulvaient) in series."
The inspectors 1centiflea tne failure to have reaundant barriers as Example 2 of an apparent /1olatitn tf 10 CFR Part 50. 'ppendix 8. Criterion III (298/9414-01).
2.1.5 Classificanon of Primary Containment Isolation Barriers The licensee 1centiflea tnat approximately 300 examples of components associated alth containment penetrations were not classified as essential.
Draft General Cesign Criteria. Criterion 1. July 1967. as stated in Appendix F to the USAR. recu1rea those systems and components of the station which
[had] a vital role in the crevention or mitigation of consecuences of accidents arrecting tne cuolic nealth ano safety [were] designed and constructed to high quality standards General Electric Des gn Scecificat~ n No. 22A1153. " Codes and Industrial Stancard." Revisicn
. states, in Note 3 of the Appendix, that "[p]iping, which is an integrai part of the primary containment for isolation purposes, shall have at ' east the same quality and levels of assurance as the primary containment."
In addition.10 CFR Part 59. Appendix B. Criterion III, requires that
"[m]easures snall ':e estabilshed to assure that the design basis.
are correctly translated into specifications These measures shall include provislens to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controllea.'
The licensee ccnciuced tnat those components not classified as essential were designed. faDricatea. ano erectea to quality standards that did not reflect the importance of tne safety function to be performed in accordance with 10 CFR Part 50. 'opendix 5. Criterion III: General Electric Design Specification No. 22A1153. Revision 1: or Appendix F to the USAR.
The failure t: dest:n. 'a r'cate end erect the containment isolation barriers to quality standaras tnat reflected the importance of the safety function was identified as E..ampie 3 ef an apcarent violation of 10 CFR Part 50.
Appenaix B. Criterion III (298/9414-01).
2.1.6 Containment Penetration Insoections The inspectors rev'e..ea a numoer of primary containment penetrations pre m us' nsc. 2;;
. tne 11censee.
For those penetrations, the inspectors concluded tnat the 'i ensee's insoection had been accurate and the marked-up cranings ref'ectec tre actual conoition in tne plant.
1 2.2 Motor-onerated "alvo !ssues On December 20. 1993. as documentea in NRC Inspection Report 50-298/93-29.
Valve HPCI-MOV-M017 failed to stroke.
The licensee formed a problem resolution team to investigate that failure.
The team issued a report on January 7.1994. that documented the team's findings.
Those findings were that the failure :as the result of fiberglass fragments between the limit switch contacts.
The team presented this report as the response to Nonconformance Report 93-270 in oroer to recommend corrective actions.
On March 14. 1994 as documented in NRC Inspection Report 50-298/94-09.
excessive leakage c.as noted during the venting of piping between Valves RHR-MOV-M025A and -M027A.
In this instance. the licensee determined that tne prcolem c.as relatea to foreign material deposited on the valve seat after maintenance that breached the residual heat removal system boundary.
On May 27. 1994. :ne licensee reported that Valve MOV-M016 was found
" partially deenergizea" after attempting to close the valve.
The licensee's I
investigation led to the identification of " particles" stuck between the contacts of the tcrque sitch.
On June 20. 1994. the licensee reported that Valve RHR-MOV-M0278 was not l
capable of passirg 'ts local leak rate test.
At the time of this inspection, the licensee had not determined a root cause for the failure.
The licensee had decided to move the primary containment isolation function from Valves RHR-MOV-M025A(B) and -M027A(B) to Valves RHR-CV-26CV(27CV).
RHR-MOV-M0274A(B). and -M025A(B).
The licensee performed this change by use of a safety evaluation that was performed in accordance with 10 CFR 50.59.
2.2 1 Safety Evaluaticn Review i
The inspectors reviewea the safety evaluation and found that the evaluation was thorough and in accordance with the requirements of 10 CFR 50.59.
The inspectors noted that. in oraer to accomplish this change the licensee had to change operating preceaures to prevent the opening of either RHR-MOV-M0274 valve and to ensure that the motor operator will remain deenergized when the reactor coolant temoerature was above 100 C (212 F).
Another change to the procedures was tnat shutdown cooling could only be initiated when the reactor pressure c.as 'a::
un M.' ' kPa (5 psig).
The inspectors concluded that the licensee's change of primary containment isolation boundary was adequately justified. and appropriate procedural controls were 1dentified.
2.2.2 Limit Switch Comoartment Cleanliness During review o' *e l!cersee's' actions related to the lack of cleanliness inslae the ilmit 2<.1 ten ccmpartments, f'e inspectors found that the licensee had proposed 3 :mieticn date of Seatt. >er 1994 for the corrective actions s
related to the ta;iure of '!alve HPCI-MOV-M017.
The licensee had not entered the corrective act cns into its tracking system.
-p-This was a concern to Ine insoectors for two reasons.
The lengthy amount of time allowea to cass oefore the corrective actions were due increased the chances for similar events to occur.
In this case, a similar event did occur when Valve MOV-M016 failed to ocerate properly.
The other concern was the failure to timely incorporate the corrective actions into the tracking system to assure that management is provided with an approorlate status of corrective actions.
The licensee nad indicatea that the failure to track was a backlog problem because of an aaministrative overload.
In each case, a condition report had been issued and initial corrective action initiated.
The inspectors were concerned that the licensee would have failed to perform these corrective acticns without the NRC inspection into the motor-operated valve issues.
The licensee alc arm a conaltion resolution team to review the failure of Valve MOV-M016.
N s team naa not issued its report. therefore. the inspectors did not review the licensee's actions for that failure.
The review of the licensee's actions 's considered to be an inspection followup item (298/9414-03).
2.2.3 Analysis n Other salve Concerns The failures of Valves RHR-MOV-M027A(B) presented other concerns.
One concern was related :: ~s ::nt :' -f fcreign materials wnen systems were breachea.
The inspectors noted that corrective actions had not been approved for the March event wnen..eia s.ag aas determined to be tne cause of the problem.
When questionea oy the 'nspectors. the responsible engineer stated that this issue had been given iczer criority ano, in essence. that there was a lack of personnel to ensure the corrective action process was timely. Another concern was that the licensee naa not considered any interim actions to prevent foreign materiai to get into systems other than a memorandum to maintenance personnel informrg them of recent problems and instructing them to be careful.
The inspectors concludea that management attention was warranted in the areas of foreign mater ii exclusion and the corrective action programs.
'e corrective action crogram e.as considered to warrant the attention tuse of the fact that it ac ceen implemented only recently and the inspeccors noted these concerns.
2.3 Switch Calibratinn The licensee icentifiea tnat severai instrument pressure switches in Racks 25-5 and 4 subject to drywell cressure, were isolated during the performance or :r.e ;ontainment integrated leak rate tests performed in accordance with Ine 2.SME Boller and Pressure Vessel Code.
The licensee stated that these instruments e.ere 1solated because the licensee's staff thought the test pressure tatoroximately 400 kPa (58 osi)) would damage the instruments.
i.ocal lean rate :est ng naa not ceen oerformea in ileu of opening the valves to the racks cur va 'ntegrated leak rate testing.
On July 8. 1994 Surveillance Prcceaure 6.3.1.1.2. Revision 0. " Primary Containment Instrument
at
--n Local Leak Rate 7ests." was issued to initiate local leak rate testing for these pressure sentches.
The pressure switches on Racks 25-5 and -6 include:
PC-PS-12A. B. C. ana D: PC-PS-101A B. C. and D: PC-PS-119A. B. C. and'D:
PC-PS-16: and PC 2T-512A and B.
The pressure switches perform scram.
containment isolation. and emergency core cooling system initiation upon receiving a dry'<. ell pressure signal of 13.7 kPa (2 psig) or greater.
The licensee contactea the instrument vendor and was notified that the instruments could withstana tne test pressure, but snould be calibrated after the test to ensure there was no shift in the operating characteristics of the instruments.
The licensee stated that thes'e instruments would be calibrated after being subjected to the cressure of the containment integrated leak rate test.
The failure to perfcr1 local leak rate tests is identified as Example 3 of an apparent violatun of Technical Specification 4.7. A.2.f.1 (298/9414-02).
2.4 (Closed) !'~ = &1vad N m c0-29H 9an3.nl:
Use of Sinole Manual Valve for Containmont Molation NRC Inspection Report 50-298/94-03 summarized the inspection conaucted during January 2 through February 12. 1994.
The report discussed the use of a single manual valve for :cntainment isolation. anich was determined to be an Unresolved item c298/9403-01) pending additional NRC review.
The valves in question were ali manual operated vents drains, or test connections; a total of ten valves c.ere affectec.
During this inspection, the inspectors determined that tne ten valves, identified in tre earlier insoettion. had been modified by means of a maintenance wort 'eauest.
The modification consisted of adding either a cap or plug. which actec as a second barrier for containment isolation.
This design philosoony c.as ccnsistent with draft General Design Criteria 53. as stated in Appena1x F to the USAR.
All material used in the maintenance work requests were classified essential. and their certification and traceability were available.
The licensee sucmitted its response to NRC Inspection Report 50-298/94-03 by letter dated May 31. 1994.
The response stated that the licensee was reconstituting tne cesign basis for the primary containment system and would evaluate the issue vithin that task.
In addition. the licensee advised that it was pursuing efforts to resolve NRC concerns involving the identification and control of unual primary containment isolation valves. or more appropriately tne aaministrative control of the valve and cao/ plug combination.
~"e 1censee stated that it planned to complete this effort by August 1994.
In addition to tne ten.-ai.es identifiea in Unresolvea item 298/9403-01.
additional manua' cents. drains and test !alves were cappea or plugged in accordance with '91ntenance e.'ork Recuest 94-2978 and its sucolement 94-3116.
This was aiscusie is a cart of the cesign changes in Section 2.1.1 of this report.
i i
l 15 In accercance
.ir- "ra' leneral Design Criter'a. C-'terion 53..'uly 1967. as stated in Appena1x ? :o tne USAR. "[a]Il lines which penetrate the primary containment anc rnicn communicate with the reactor vessel or the primary containment free space [e.ere] provided with at least two isolation valves (or equivalent) in series."
The ten single isolation valves without a secona barrier were identified as Example 2 of the a: parent '/1olation of 10 CFR Part 50. Appendix B.
1' Criterlon 111. identified in Section 2 1.4 (298/9414-01).
ATTACHMENT 1 PERSONS CONTACTED 1.1 Licensee Parsonnel
- R. Gardner. Plant Manager
- R. Godley. Manager. Nuclear Licensing and Safety
- G. Horn. Vice President. Nuclear
- S. Jobe. Acting Senior Nuclear Division Manager. Safety Assessment
- J. Lynch. Manager. Engineering
- E. Mace. Senior Manager. Site Support
- J. Mueller. Site Manager
- J. Sayer. Technical Assistant to Plant Manager
- R. Wilbur. Division Manager
- V. Wolstenholm. Division Manager. Quality Assurance 1.2 Other Pors-rno;
- H Berchert. Director. Division of Radiological Health. State of Nebraska
- J. Parker. Mid'<.est Pov.er
- R. Stoddard. Lincoin Eleccric System
- W. Turnbull. Midwest Power 1.3 NRC Personnel
- A. Beach. Direc:cr. Division of Reactor Projects
- L. Callan Regional Administrator. Region I'/
- P. Goldberg Reactcr Inspector. Engineering Branch
- C. Hackney. State Liaison Officer
- P. Harrell. Chief. Peactor Projects Branch C
- R. Kopriva. Senior Resident Inspector
- W. Walker. Resident Inspector In addition to tne personnel listed above. the inspectors contacted other personnel during this inspection period.
- Denotes personnel that attended the exit meeting on August 12. 1994.
2 EXIT MEETING An exit meeting c.as conducted on August 12. 1994.
During this meeting the scope and find' 7s :f the inspect'cn c.ere reviewed.
The licensee ackncwledged the inspection findings documented in this report.
The licensee did not identify as crc:-'etary a.ny information provided to. or reviewed by. the inspectors.
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LIMITING CONDITIONS FOR OPERATION" ~
~~ ~ URVEILLANCE REOUIREMENTS-
~~
S
.3.7 ~ CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS j
Applicability:
Applicability:
-Applies to the operating status of Applies to the primary and secondary the primary and secondary contain-containment integrity.
ment systems.
-Objective:
Objective:
To assure the integrity of the pri-To verify the integrity of the primary n
mary and secondary containment systems, and secondary containment.
l Specification:
Specification:
i l
A.
=
..1.
Suppression Pool 1.
Suppression Pool i
At any cie that the nuclear system
- a. The suppression pool water level i
is pressurized above atmospheric and temperature sh'll be checked a
pressure or work is being done once per day.
which has the potential to drain the vessel, the suppression pool
- b. Whenever there is' indication of water volume and temperature shall relief valve operation or testing j
l be maintained within the following which adds heat to the suppression l
limits except as specified in pool, the pool temperature shall J
3.7.A.2. and 3.5.F.5.
be continually monitored and also j
bserved and logged every 5 3
i a.
Minimum water volume - 87,650 ft minutes until the heat addition is terminated.
3 lb.
Maximum water volume - 91,100 ft i
- c. Whenever there is indication of c.
Maximum suppression pool temperature relief valve operation with the during normal power operation - 95 F.
temperature of the suppression f-pool reaching 160 F or more and d.
During testing which adds heat to the primary coolant system pres-i the suppression pool, the water sure greater than 200 psig, an i
temperature shall not exceed 10 F external visual examination of above the normal power operation the suppression chamber shall limit specified in c. above.
In be conducted before resuming connection with such testing, the power operation, j
pool temperature must be reduced to l
I below the normal power operation
- d. A visual inspection of the i~
limit specified in c. above within suppression chamber interior,
-24 hours.
including water line regions, j.
shall be made at each major
]
e.
The reactor shall be scrammed from refueling outage.
any operating condition if the pool j
temperature reaches 110 F.
Power i
operation shall not be resumed l
until the pool temperature is reduced below the normal power 4
operation limit specified in c.
above.
l
-159-C /1 T/O C N
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS
',7;A.1 (cont'd) 4.7.A (cont'd)
.f.
During reactor isolation conditions, 2.
Leak Rate Testine the. reactor pressure vessel shall be depressurized to less than 200 psig at a.
Integrated leak-rate test (ILRT's) normal cooldown rates if the pool shall be performed to verify primary temperature reaches 120*F.
containment integrity.
Primary containment integrity is confirmed if 2.
Containment Inteerity the leakage rate does not exceed the equivalent of 0.635 percent of the l. a.
Primary containment integrity shall be primary containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> maintained at all-times when the at 58 psig.
reactor is critical or when the reactor water temperature is above b.
Integrated leak rate tests may be 212'F and fuel is in the reactor performed at either 58 n-ig or 29 psig, vessel except while performing "open the leakage rate test
- i od, extending vessel" physics tests at power levels to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of ree-ined internal not to exceed 5 MW(t),
pressure. If it can be demonstrated to the satisfaction of those responsible b.
'Jhe n Coolant Temperature is above for the acceptance of the containment 212'F, the drywell and suppression structure that the leakage rate can be chamber purge and vent system may be accurately determined during a shorter in operation for up to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per
-test period, the agreed-upon shorter calendar year with the supply and period may be used, exhaust 24-inch isolation valves in one supply line and one exhaust line Prior to initial operation, integrated open for containment
- inerting, leak rate tests must be performed at 58 deinerting, or pressure control, and 29 psig (with the 29 psig test being performed prior to the 58 psig If venting or purging is through test) to establish the allowable leak Standby Gas for such operations, then rate. 4 (in percent of containment both Standby Gas Treatment Systems volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) at 29 psig as the shall be operable and only one Standby lesser of the following values.
Gas Treatment System is to be used.*
(L. is 0 635 percent)
Not applicable to valves open during venting or purging provided such 4-0.635 _*
venting or purging utilizes the 2-inch L,m bypass line(s) around the applicable inboard purge exhaust isolation for
- s 0.7 valve (s) with the inboard valve (s) in a closed condition.
h*
where L, - measured ILR at 29 esic L,
- measured ILR at 58 psig, and Lm r 1.0 L,m 4 - 0. 6 35 P,1/2 P,
-160-05/15/90
J i
LINITINOCONDITIONSh0ROPbTION l SURVEILLANCE REOUIREbhTS 3.7.A (cont'd.)
4.7 A.2.b. (cont'd.)
where P, = peak accident pressure, 58 psig l
P = appropriately measured test pres-E sures (psig) l for em > 0.7 L
am c.
The ILRT's shall be performed at the following minimum frequency:
1.
Prior to initial unit operation.
2.
A 'approximately three and one-third year intervals so that any ten-year interval would 4
include four ILRT's. These intervals may be extended up to eight months if necessary to coincide with refueling outage.
{hemeasuredleakageraces,Lgmand d.
am, ghall be less than 0.75 t and 0.75 a for the reduced pressure tests and peak pressure test respectively, e.
Except for the initial ILRT, all ILRT's shall be performed without any pre-liminary leak detection surveys and leak repairs immediately prior to the test.
If an ILRT has to be ter-minated due to excessive leakage through identified leakage paths, the leakage through such paths shall be determined by a local leakage test and recorded. After repairs are made another ILRT shall be conducted.
4 If an ILRT is completed but the acceptance criteria of Specification 4.7.A.2.d is not satisfied and repairs are necessary, the ILRT need not be 1
i
-161-09/21/84
i V
LIM 1TTNG CONDfTTONS FOR OPERATION SURVEILIANCE REOUIREMENTS 3.7.A (Cont'd)
A.7.A.2.e (cont'd).
repeated provided locally. measured leakage reductions, achieved. by-repairs, reduce the - containment's ove ra2.l' measured leakage rate sufficiently to meet the acceptance
- criteria, f.
Local Leak Rate Tests 1.
With the exceptions specified below, local leak rate tests (LLRT's) shall be performed on the primary-containment testable penetrations and isolation valves at a pressure of 58 psig during each reactor shutdown for refueling, or other convenient intervals, but in no case at intervals greater than two years, g
The test duration of all valves and penetrations shall be of sufficient length to determine repeatable results.
The total acceptable leakage for all valves and penetrations other than the MSIV's is 0.60 La.
2.
Bolted double-gasket seals shall be tested after each opening and during l.
each reactor shutdown for refueling, or other convenient intervals but in no case at intervals greater than two years.
3.
The main steam isolation valves (MSIV's) shall be tested at a pressure of 29 psig.
If a total leakage rate of 11.5 scf/hr for any one MSIV is exceeded, repairs and retest shall be performed to correct the condition. This is an exemption to Appendix J of 10CFR50.
4#
'r
.. - - -..... -. - -. - ~..
i i
LIMITI!'c CONDITIONS FOR OPERATION SURVEILIANCE REOUTREMENTS 3.7.A (Cont'd) 4.7.A.2.f (cont'd)
~
4.
Main steam line and feedwater line expansion bellows shall be tested by l
pressurizing between the laminations of the bellows -_ at ' a pressure of-5 psig. This is an exemption to Appendix J of 10CFR50.
5.
The personnel airlock shall be tested at 58 psig at intervals no longer than six months.
This j
testing may be extended'to the next I
refueling outage (not to exceed 24 months) provided that there have been no airlock openings since the last succes'sful test at 58 psig. In the event the personnel airlock is not opened between refueling outages, it shall be leak checked at 3 psig at intervals no longer than six months.
Within three days of opening (or every three days during periods of frequent opening) when containment integrity is required, test the personnel airlock at 3 psig.
This -is an exemption to Appendix J of 10CFR50.
The maximum allowable leakage at a test pressure of 58 psig is 12 scfh.
Leakage measured at test pressure less than 58 psig is adjusted to the equivalent value at 58 psig.
g.
Deleted h.
Drvwell Surfaces s
l The interior surfaces of the drywell and torus shall be visually l
inspected each operating cycle for evidence of torus corrosion or t
leakage.
4 l
t i
4 0
4 8
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.7.A (cont'd.)
4.7.A (cont'd.)
I
~
3.
Pressure Suppression Chamber -
3.
Pressure Suppression Chamber -
Reactor Building Vacuum Breakers Reaccor Building Vacuum Breakers a.
Except as specified in 3.7.A.3.b a.
The pressure suppression chamber-reactor below, two pressure suppression building vacuum breakers and associated chamber-reactor building vacuum instrumentation, including set points breakers shall be operable at all shall be checked for proper operation times when primary containment in-every three months, tegrity is required. The set point of the differential pressure instru-mentation.which actuates the pressure suppression chamber-reactor building air actuated vacuum breakers shall be 0.5 psid. The self actuated vacuum breakers shall open fully when subjected to a force equivalent to 0.5 psid acting on the valve disc.
b.
From and after the date that one of b.
During each refueling outage each the pressure suppression chamber-vacuum breaker shall be tested to reactor building vacuum breakers is determine that the force required made or found to be inoperable for to open the vacuum breaker does not any reason, the vacuum breaker switch exceed the force specified in shall be secured in the closed positica Specifications 3.7.A.3.a and each and reactor operation is permissible vacuum breaker shall be inspected only during the succeeding seven days and verified to meet design unless such vacuum breaker is sooner requirements, made operable, provided that the repair Procedure does not violate primary containment integrity.
4 Drvwell-Pressure Suppression Chamber 4.
Drvwell-Pressure Suppression Chamber Vacuum Breakers Vacuum Breakers 1
a.
When primary containment is required, a.
Each dryvell-suppression chamber vacuum all drywell-suppression chamber vac-breaker shall be exercised through an uum breakers shall be operable at the opening-closing cycle every 30 days.
1 0.5 psid setpoint and positioned in the fully closed position as indicated by the position indicating system except i
during testing and except as specified in 3.7.A.4.b and.c below.
b.
Three drywell-suppression chamber b.
When it is determined that a vacuum vacuum breakers may be determined breaker valve is inoperable for opening
~
to be inoperable for opening pro-at a time when operability is required vided they are secured in the fully all other vacuum breaker valves shall I
closed position or that the require-be exercised immediately and every 15 ment of 3.7.A.4.c is demonstrated to days thereafter until the inoperable be met, valve has been returned to normal service.
4/29/83
-163-
~
\\
LIMITING CONDITIONS FOR OPERATION l SURVEILLANCE REOUIREMENTS 3.7.A.4 (cont'd.)
- 4. 7. A.4 (cont 'd.)
c.
Once each operating cycle, each vacuum c.
The total leakage between the dry-well and suppression chamber shall breaker valve shall be visually in-be less than the equivalent leakage spected to insure proper maintenance through a 1" diameter orifice, and operation of the position indicatiot switch. The differential pressure set-i point shall be verified.
d.
Prior to reactor startup after each d.
If specifications 3.7.A.4.a, b or c, cannot be met, the situation shall refueling, a leak test of the dryvell be corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the to suppression chamber structure shall be conducted to demonstrate reactor will be placed in a cold shutdown condition within tho sub-that the requirement of 3.7.A.4.c is met.
sequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
5.
Oxygen Concentration 5.
Oxygen Concentration a.
The primary containment oxygen con-a.
After completion of the startup test centration shall be measured and program and demonstration of plant recorded at least twice weekly, electrical output, the primay con-tainment atmosphere shall be reduced to less than 4% oxygen with nitrogen gas during reactor power operation with reactor coolant pressure above 100 psig, except as specified in 3.7.A.S.b.
b.
The quantity of liquid nitrogen in b.
Within the 24-hour period subsequent to placing the reactor in the Run mode the liquid nitrogen storage tank shall following a shutdown, the containment be determined twice per week when the v lume requirements at 3.7.A.S.c are atmosphere oxygen concentration shall in effect, be reduced to less than 4% by volume and maintained in this condition.
De-inerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown, i
c.
When the containment atmosphere oxygen i
concentration is required to be less than 4%, the minimum quantity of liquid i
nitrogen in the liquid nitrogen storage tank shall be 500 gallons.
d.
.If the specifications of 3.7.A.5.a thru e cannot be met, an orderly shutdown shall be initiated and the reactor
)
shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
e.
The specifications of 3.7.A.5.a thru d are not applicable during a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> continuous period between the dates of March 22, 1982 and March 25, 1982.
i l
l 4/29/83
-164-
LIMITING CONDITf0N'FOR OPERATION SURVEILIMCE REOUIREMENT 3.7.A (cent'd.)
4.7.A (cont'd.)
6.
Low Low set Relief Function 6.
Low-Low Set Relief Function a.
The low-low set function of the
- a. The low-low set safety / relief valves safety-relief valves-shall be shall be tested and calibrated as operable when there is irradiated specified in Table 4.2.B.
fuel in the reactor vessel and the.
reactor -coolant temperature is 2 212*F, except*as specified in 3.7.A,6.a.1 and 2 below.
.l.
With the low-low function of one safety / relief valve (S/RV) inoperable, restore the inoperable LLS S/RV to OPERABLE within 14 days or be in the HOT STANDBY mode within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
With the low-low set function of both S/RVs inoperable, be in at least HOT STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b. The pressure switches which control the low-low set safety / relief valves shall have the following settings.
NBI-PS-51A Open Low Valve 1015 20 psig (Increasing)
NBI-PS-SlB Close Low Valve 875 20 psig (Decreasing)
NBI PS-51C Open High Valve 1025 20 psig (Increasing)
NBI-PS 51D Close High Valve 875 t 20 psig (Decreasing) 1.
At least once per operating cycle the following conditions shall be B.
Standbv Gas Treatment System demonstrated.
1.
Except as specified in 3.7.B.3
- a. Pressure drop across the combined below, both Standby Gas Treatment HEPA filters and charcoal adsorber subsystems shall be operable at all banks is less than 6 inches of water times when secondary containment at the system design flow rate.
integrity is required.
2.a. The results of the in-place cold DOP
- b. Inlet heater input is capable of leak tects on the HEPA filters shall reducing R.H.
from 100 to 70% R.H.
show 299% DOP removal. The results of the halogenated hydrocarbon leak 2.a. The tests and sample analysis of tests on the charcoal adsorbers Specification 3.7.B.2 shall be shall shev 299%
halogenated performed at least once every hydrocarbon removal.
The DOP and 18 months for standby service or halogenated hydrocarbon tests shall after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system be performed at a Standby Gas operation and following significant Treatment flowrate of $1780 CFM and painting, fire or chemical release at a Reactor Building pressure of in any ventilation zone s.25" Ug.
communicating with the system.
-165-3/11/92
LIMITING CONDITION FOR OPERATION SURVETT YANCE REOUTREMENT
- 3..B (cont'd) 4.7.B (cont'd)
- b. The results of laboratory carbon
- b. Cold DOP testing shall be performed sample analysis shall show 299%
after each complete or partial radioactive methyl iodide removal replacement of the HEPA filter bank with inlet conditions of: velocity or after any structural maintenance 227 FPM, 21.75 mg/m inlet methyl on the system housing.
3 iodide concentration, 270% R.H. and 530*C.
- c. Halogenated hydrocarbon testing shall be performed after each
- c. Each fan shall be shown to provide complete or partial replacement of the charcoal adsorber bank or after 1780 CMF 110%.
on the any structural maintenance system housing.
3.
From and after the date that one l
Standby Gas Treatment subsystem is made or found to be inoperable for
- d. Each subsystem shall be operated l
any reason, reactor operation is with the heaters on at least permissible only during the 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month, succeeding seven days unless such Test sealing of gaskets for housing j
subsystem is sooner made operable, e.
provided that during such seven days doors downstream of the HEPA filters all active components that affect and charcoal adsorbers shall be operability of the operable Standby.
performed at, and in conformance Gas Treatment subsystem, and its
- with, each test performed for associated diesel generator, shall compliance with Specification 4.7.B;2.a and Specification be operable.
3.7.B.2.a.
Fuel handling requirements are specified in Specification 3.10.E.
3.
System drains where present shall be inspected quarterly for adequate 4
If these conditions cannot be met, water level in loop-seals.
procedures shall be initiated per operating cycle immediately to establish reactor 4.a. At least once conditions for which the Standby Gas automatic initiation of each Standby Treatment System is not required.
Gas Treatment subsystem shall be l
demonstrated, b.
At least once per operating cycle manual operability of the bypass valve for filter cooling shall be demonstrated.
c.
When one Standby Gas Treatment subsystem becomes inoperable, the operable Standby Gas Treatment subsystem shall be verified to be J
operable immediately and daily thereafter.
A demonstration of diesel generator operability is not required by this specification.
C.
Secondarv containment C.
Secondary containment surveillance 1.
Secondary containment integrity shall b'e performed as indicated shall be maintained during all modes below:
of plant operation except when all of the following conditions are met.
-165a-3/11/92
LIMITING CONDITIONS FOR OPEPATION SURVEIT TANCE REOUIPWENTS 3.7.C (cont'd.)
4.7 C (cont'd.)
a.
The reactor is suberitical and a.
A preoperational secondary Specification 3.3.A is met.
containment capability test shall be conducted after isolating the b.
The reactor water temperature is reactor building and placing either below 212*F and the reactor coolant Standby Gas Treatment subsystem l system is vented.,
filter train in operation.
Such tests-shall demonstrate the c.
No activity is being performed which capability to maintain 1/4 inch of can reduce the shutdown margi.1 below water vacuum under calm wind that specified in Specification (2<E<5 mph) conditions with a filter 3.3.A.
train flow rate of not more than 100% of building volume per day, d.
No irradiated fuel is being handled (u-wind speed) in the secondary containment and no loads which could potentially damago b.
Additional tests shall be performed irradiated fuel are being moved in during the first operating cycle the secondary containment.
under an adequate number of different environmental wind e.
If secondary containment integrity conditions to enable valid cannot be maintained, restore extrapolation of the test results.
secondary containment integrity within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or; c.
Secondary containment capability to maintain 1/4 inch of water vacuum a.
Be in at least Hot Shutdown under calm wind (2<E < 5 mph) within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and conditions with a filter train flow in cold shutdown within the rate of not more than 100% of following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, building volume per day, shall be demonstrated at each refueling b.
Suspend irradiated fuel handling operations in the outage prior to refueling, secondary containment, d.
After a
secondary containment movement of loads which could potentially damage irradiated violation is determined, the Standbv Gas Treatment System will b'e fuel in the secondary operated immediately after the containment, and all core af fected zones are isolated from the alterations and activities remainder of the secondary which could reduce the containment to confirm its ability shutdown margin.
The to maintain the remainder of the provisions of Specification 1.0.J are not applicable, secondary containment at 1/4 inch of water negative pressure under calm wind conditions.
D.
Primary Containment Isolation Valves D.
Primary Containment Isolation Valves 1.
During reactor power operating 1.
The primary containment isolation conditions, all isolation valves valves surveillance shall be listed in Table 3.7.1 and all performed as follows:
instrument line flow check valves a.
At least once per operating cycle shall be operable except as the operable isolation valves that specified in 3.7.D.2.
are power operated and automatically initiated shall be tested for simulated automatic initiation and closure times.
-166-
LIMITING CONDITIONS FOR OPEPATION SURVEIIIANCE REOUIREMENTS 3.7.D (cont'd.)
4.7.D (cont'd.)
o b.
At least once per quarter:
(1) All normally open power operated isolation valves (except. for the main ste%1
' I..e power operated
)
isolac'.,2 va es) shall be ; fully closed and.copened.
l (2) IJith the reactor power less than 75%,
trip main. steam isolation valves
-individually and-verify closure time.
c.
At least once per operating cycle the ~ operability of the reactor coolant system instrument line flow check valves shall be verified.
d, At least once per operating cycle, while shutdown, the devices that limit the maximum opening angle to 60' shall be verified functional for the following valves:
PC-230MV, PC-231MV, PC-232MV, and PC-233MV.
2.
In the event any isolation valve 2.
tJhenever an isolation valve listed specified in Table 3.7.1 becomes in Table 3. 7.1 is inoperable, the inoperable, reactor power operation position of at least one other valve may continue provided at least one in each line having an inoperable valve in each line having an valve shall be recorded daily, inoperable valve shall be in the mode corresponding to the isolated l
condition.*
3.
If Specification 3.7.D.1 and 3.7.D.2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- Isolation valves closed to satisfy these requirements may be reopened on an ' intermittent basis under administrative control.'
4 I
f a
f P
COOPER NUCLEAR STATION-TABLE 3.7.1 (Page 1)
PRIMARY CONTAINMENT IS0lATION VALVES Number of Power Maximum Action On Operated Valves Operating Normal Initiating Valve & Steam Inboard outboard Time (Sec) (1)
Position (2)
Sirml 0)
Main Steam Isolation Valves MS-AO A,B,C, 6 D 4
3<T<5 0
GC MS-AO A,B,C, 6D 4
3<T<5 0
GC Drywell Floor Drain Iso. Valves 2
15 0
GC RW-AO-82, RW-AO-83 Drywell Equipment Drain 2
15 0
cc Iso. Valves RW-AO-94, RW-AO-95 Main Steam 1.ine Drain 1
1 30 0
cc 2
Valves MS-MO-74, MS-MO-/7 Reactor Water Sample Valves 1
1 15 0
cc RR-740AV, RR-741AV l
Reactor Water Cleanup System 1
1 60 0
cc Iso. Valves RWCU-MO-15, RWCU-MO-18 RllR Suction Cooling Iso.
I 1
40 C
Sc Valve RiiR-MO-17, RilR-MO-18 R11R Discharge to Radwaste 2
20 c
sc Iso. Valves RilR-MO-57, RHR-MO-67 Suppression Chamber Purge 6 2
15 c
sc Vent PC-24SAV, PC-230MV Suppression Chamber N2 Supply 2
15 C
SC PC-237AV, PC-233MV
i COOPER NUCLEAR STATION TABLE 3.7.1 (Page 2)
PRIMARY CONTAINMENT ISOLATION VAINES i
Number of Power Maximum Action On Operated Valves Operating Normal Initiating' j
Valve & Steam Inboard Outboard Time (Sec) (1)
Position (2)
Sienal (3)
Primary Containment Purge & Vent 2
15
'PC-246AV, PC-231MV C
.SC Primary Containment & N SuPP y 2
15 l
2 PC-238AV, PC-232MV C
SC Suppression Chamber Purge & Vent 1
40 C
PC-230MV Bypass (PC-305MV)
SC(4)
Primary Containment Purge & Vent 1
40 C
PC-231MV Bypass (PC-306MV)
SC(4)
Dilution Supply PC-1303MV, PC-1304MV 2
15 C
SC PC-1305MV, PC-1306MV 2
15 C
SC Dilution Supply PC-1301MV, PC-1302MV 2
15 0
CC PC-1311MV, PC-1312MV 2
15 0
GC Suppression Chamber Purge and Vent Exhaust 1
15 PC-1308MV C
SC Primary Containment Purge and Vent Exhaust 1
15 PC-1310MV C
SC i
i h
t
NOTES FOR TABLE 3.7.1 1.
Maximum valve operating times in seconds in the closed direction.
This is the direction required for Primary Containment isolation.
2.
Normal position indicates the normal valve _ position during power operations.
0 = Open C = Closed 3.
Action on initiating signal indicates the valve operation af ter the signal initiation.
GC = Goes Closed SC = Stays Closed 4.
PC-305MV & PC-306MV have override switches (key operated) which can be used to open valves when isolation signals are in.
l 2
-170-10/20/81
-~
_ - ~. -. -
j-3.7 & 4.7 BASES t
3.7.A & 4.7.A PRIMARY CONTAINMENT 3.7;.t.1 & 4.7.A.1 SUPPRESSION POOL The integrity of the primary containment and operation of the cara standby cooling system, in combination, limit the off-site doses to values less chan those suggested in 10CFR100 in the event of a break in the primary system piping.
Thus, containment integrity is specified whenever the potential for violation of the primary reactor i
system integrity exists.
Concern about such a violation exists whenever the reactor I
is critical and above atmospheric pressure. An exception is made to this requirement i
during initial core loading and while the low power test program is being conducted and ready access tc the reactor vessel is required.
There will be no pressure on the system at this time, thus greatly reducing the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Pro-
]
cedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not result in any fuel damage.
In addition, in the unlikely event that an j
excursion did occur, the reactor building and standby gas treatment system, which i
shall be operational during this time, offer a sufficient barrier to keep off-site 4
doses well below 10CFR100 limits.
I The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1035 psig.
Since a13 l
of the gases in the drywell are purged into the pressure suppression chamber air j
space during a loss-of-coolant accident, the pressure resulting from isothermal
{
compression plus the vapor pressure of the liquid must not exceed 62 psig, the 8
suppression chamber maximum pressure.
The design volume of the suppression cham-ber (water and air) was obtained by considering that the total volume of reactor l
coolant to be condensed is discharged to the suppression chamber and that the i
drywell volume is purged to the suppression chamber.
As a result of the Mark I Containment Program, the District has completed the evaluation and requalification of the various containment structures and compo-7 nents at CNS.
As a result of the requalification work, significant modifications were designed in accordance with the NRC acceptance criteria and installed. The l
Plant Unique Analysis Report, which was submitted on April 29,1982, and accepted i
on January 20, 1984, contains a detailed summary of the modifications installed.
1 The maximum and minimum watcr volumes of 91.100 and 87,650 were not altered, but the downcomers were shortened by l' O ", so that their nominal submergence is now l
3 feet and the initial volume of water in them is dec'reased proportionately.
The acceptability of this is proven in " Mark I Containment Program Downcomer Submer-gence Functional Assessment Report", Task 6.6, NEDE - 21885-P, Class III, June, 1978.
I Should it be necessary to drain the suppression chamber, this should only J
1 r
-176-5/13/85 i
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3.7.A & 4.7.A B ASES (cont'd)
' ~ ~ - ~ ' ~
be done when there is no requirement for core standby cooling systems operability as explained in bases 3.5.F.
Experimental data indicates that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160*F during any period of relief valve operation with sonic conditions ar the. discharge exit.
Spec-Ifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of poten-tially high suppression chamber loadings.
In addition to the limits on temperature of the suppression chamber pool water, op-erating procedures define the action to be taken in the event a relief valve inad-vertently opens or sticks open. This action would include:
(1) use of all avail-able means to close the valve, (2) initiate suppression pool water cooling heat ex-changers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.
Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally change very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.
By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of sig-nificant heat addition, the temperature trends will be closely followed so that ap-propriate action can be taken. The requirement for an external visual examination follow-ing any event where potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.
The maximum suppression pool temperature of 95*F is based on not exceeding the 200*F Mark I temperature limit as contained in NUREG-0661.
This 95'F limit also prevents exceeding LOCA considerations, or ECCS pump NPSH requirements.
The basis for these limits are contained in NEDC-24360-P.
3.7.A.2 & 4.7 A.2 CONTAINMENT INTEGRITY The maximum allowable test leak rate is 0.635%/ day at a pressure of 58 psig, the peak calculated accident pressure. Experience has shown that there is negligible difference between the leakage rates of air at normal temperature and a steam-hot air mixture.
Establishing the test limit of 0.635:/ day provides an adequate margin of safety to assure the health and safety of the general public.
It is further considered that the allowable leak rate should not deviate significantly from the containment design value to take advantage of the design leak-tightness capability of the structure over its service lifetime. Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operational leak rate.
The allowable operational leak rate is derived by multiplying the maximum allowable leak rate, La, or the allowable test leak rate, Lt. by 0.75 thereby providing a 25:
margin to allow for leakage deterioration which may occur during the period between leak rate tests.
The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak race remains within the specification. The leak rate test frequency is based on the NRC guide for developing leak rate testing and surveillance of reactor containment vessels.
Allowing the test intervals to be extended up to 8 months permits some flexibility needed to have the tests coincide with scheduled or unscheduled shutdown periods.
The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage
-177-5/13/85
3.7.A & 4.7.A BASES (cont'd.)
trends.
Whenever a bolted double-gasketed penetration is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly.
It is expec ted that the majority of the leakage from valves, penetrations and seals would be into the reactor building.
However, it is possible that leakage into other parts of the facility could occur.
Such leakage paths that may affect significantly the consequences of accidents are to be minimized.
l Certain isolation valves are tested by pressurizing the volume between the inboard and outboard isolation valres.
T'ais results in conservative test results since the inboard valve, if a globe valva, will be tested such that the test pressure is tending to lif t the globe of f ics seat. Additionally, the measured leak rate for such a test is conservatively assigned to both of the valves equally and not divided between the two.
The main steam and feedwater testable penetrations consist of a double layered metal bellows.
The inboard high pressure side of the bellows is subjected to drywell pressure.
Therefore, the bellows is tested in its entirety when the drywell is tested.
The bellows layers are tested for the integrity of both layers by pressurizing the void between the layers to 5 psig. Any higher pressure could cause permanent deformation, damage and possible ruptures of the bellows.
Surveillance requirements for integrity of the personnel air lock are specified in (Exemption) to the letter, D. G.
Eisenhut to J. M. Pflant, September 3, 1982. When the Personnel Air Lock Leakage Test is performed at a test pressure less than 58 psig, the measured leakage must be adjusted to reflect the expected leakage at 58 psig.
Equation A-3 of Enclosure 3 (Franklin Research Center Technical Evaluation Report) to the letter, D. G. Eisenhut to J. M.
Pilant, September 3, 1982, defines the method of adjustment.
The primary containment pre-operational test pressures are based upon the calculated primary containment pressure response in the event of a loss-of-coolant accident.
The peak drywell pressure would be about 58 psig which would rapidly reduce to 29 psig following the pipe break.
Following the pipe break, the suppression chamber pressure rises to 27 psig, equalizes with drywell pressure and therefore rapidly decays with the drywell pressure decay.
The design pressure of the drywell and suppression chamber is 56 psig.
Based on the calculated containment pressure response discussed above, the primary containment preoperational test pressure was chosen.
Also, based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than the individual components separately.
The design basis loss-of-coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 0.635%/ day at 58 psig. Calculations made by the NRC staff with leak rate and a standby gas treatment system filter efficiency of 90% for halogens and assuming the fission product release fractions stated in NRC Regulatory Guide 1.3, show that the maximum total whole body passing cloud dose is about 1.0 REM and the maximum total thyroid dose is about 12 REM at 1100 meters from the stack over an exposure duration of two hours.
The resultant doses reported are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident. These doses are also based on the assumption of no holdup in the secondary containment resulting in a direct release of fission products from the primary containment through the filters and stack to the environs.
Therefore, the specified primary containment leak rate and filter efficiency are conservative and provide margin between expected off-site doses and 10 CFR 100 guidelines.
The water in the suppression chamber is used for cooling in the event of an accident; i.e.,
it is not used for normal operation; therefore, a daily
-178-1/7/93
,r---i a,
ie--,
asi' mi l
l 3.7.A & 4.7.A BASES (cont'd.)
~
trends. Whenever a bolted double-gasketed penetration is broken and remade, the space between the gaskets is pressurized to determine that the seals are perform-ing properly.
It is expected that the majority of the leakage from valves, pene-trations and seals would be into the reactor building.
However, it is possible that leakage into other parts of the facility could occur. Such leakage paths that may affect significantly the consequences of accidents are to be minimized.
a Table 3.7.4 identifies certain isolation valves that are tested by pressurizing the volume between the inboard and outboard isolation valves. This results in conservative test results since the inboard valve, if a globe valve, will be tested such that the test pressure is tending to lift the globe off its seat.
Additionally, the measured leak race for such a test is conservatively assigned to both of the valves equally and not divided between the two.
The main steam and feedwater testable penetrations consist of a double layered metal bellows.
The inboard high pressure side of the bellows is subjected to
?"
drywell pressure. Therefore, the bellows is tested in its entirety when the drywell is tested. The bellows layers are tested for the integrity of both layers by pressurizing the void between the layers to 5 psig.
Any higher pressure could cause permanent deformation, damage and possible ruptures of the bellows.
\\
Surveillance requirements for integrity of the personnel air lock are specified in Enclosure 1 (Exemption) to the letter D. G. Eisenhut to J. M. Pilant, September 3, 1982. When the Personnel Air Lock Leakage Test is performed at a i
test pressure less than 58 psig, the measured leakage must be adjusted to reflect the expected leakage at 58 psig.
Equation A-3 of Enclosure 3 (Franklin Researh Center Technical Evaluation Report) to the letter, D. G. Eisenhut to J. M. F11 ant, September 3, 1982, defines the method of adjustment.
The primary containment pre-operational test pressures are based upon the calculated primary containment pressure response in the event of a loss-of-l coolant accident. The peak drywell pressure would be about 58 psig which would rapidly reduce to 29 psig following the pipe break. Following the pipe break, the suppression chamber pressure rises to 27 psig, equalizes with drywell pressure and therefore rapidly decays with the drywell pressure decay.
The design pressure of the drywell and suppression chamber is 56 psig. Based on the calculated containment pressure response discussed above, the primary a
containment preoperational test pressure was chosen.
Also, based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than the individual components separately.
The design basis loss-of-coolant accident was evaluated at the primary con-tainment maximum allowable accident leak race of 0.635%/ day at 58 psig, Calculations made by the NRC staff with leak rate and a standby gas creat-j system filter efficiency of 90% for halogens and assuming the fission ment product release fractions stated in NRC Regulatory Guide 1.3, show that the maximum total whole body passing cloud dose is about 1.0 REM and the maximum total thyroid dose is about 12 REM at 1100 meters from the stack over an exposure duration of two hours.
The resultant doses reported are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident. These doses are also based on the assumption of no holdup in the secondary containment resulting in a direct release of fission products from i
che primary containment through the filters and stack to the environs.
Therefore, the specified primary containment leak rate and filter efficiency are conservative and provide margin between expected of f-site doses and 10 CFR 100 guidelines.
The water in the suppression chamber is used for cooling in the event of an accident; i.e., it is not used for normal operation; therefore, a daily
-178-09/no/R7
3.7.T T C7.A BASES (cont'd) be of the temperature and volume is adequate to assure that adequate heat removal capability is present.
The interiors of the drywell and suppression chamber are painted to prevent rusting.
The inspection of the paint during each major refueling outage, approximately once per year, assures the paint is intact. Experience with this type of paint at fossil fueled generating stations indicates that the inspection interval is adequate.
The intent of Specification 3.7. A.2.b is to reduce the probability of a LOCA occurrence when the 24-inch purge and vent valves are open in series. These valves are normally closed during power operation to minimize reliance on the valve operators to ensure containment integrity.
The requirements for Standby Gas is due to the damage the filters would experience from excessive difference pressure caused by a LOCA with the
(
24-inch exhaust valves open in series from the drywell or suppression chamber. This specification does allow venting with the inboard exhaust bypass valve and the outboard exhaust valve both open in series and the time does not count against the yearly limit.
The NRC has accepted the determination that due to the small size of the bypass valve, there is no chance of damage to the filters if a LOCA occurs while venting the containment through the bypass with a SBGT system on line.
The term " calendar year" is a period of time beginning on January 1 and ending on December 31 for each nunbered year.
3.7.A.3 & 4 and 4.7.A.3 & 4 VACUUM BREAKERS The purpose of the vacuum relief valves is to equalize the pressure between the drywell and suppression chamber and reactor building so that the structural integrity of the containment is maintained.
The vacuum relief system from the pressure suppression chamber to reactor buildir'g consists of two 100% vacuum relief breakers (2 parallel sets of 2 valves in series).
Operation of either system will maintain a pressure I
differential of less than 2 psi, the external design pressure.
One valve may be out of service for repairs for a period of 7 days.
If repairs cannot be completed within 7 days the reactor coolant system is brought to a condition where vacuum relief is no longer required.
The capacity of the 12 drywell vacuum relief valves are sized to limit the pressure differential between the suppression chamber and drywell during post-accident dry-well cooling operations to well under the design limit of 2 psi.
They are sized on the basis of the Bodega Bay pressure suppression system tests.
The ASME Boiler and Pressure Vessel Code,Section III, Subsection B, for this vessel allows a 2 psi differential; therefore, with three vacuum relief valves secured in the closed position and 9 operable valves, containment integrity is not impaired.
3.7.A.5 and 4.7.A.5 OXYGEN CONCENTRATION Safety Guide 7 assumptions for Metal-Water reaction result in h).
E n cor. entration in excess of the Safety Guide 7 flammability limit.
By keeping the oxygen concentration less than 41 by volume the requirements of Safety Guide 7 are satisfied.
The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-of-coolant accident upon which the specified oxygen concentration limit is based.
Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety of fered without significantly reducing the margin of safety.
Thus, to preclude the possibility of starting the reactor and operating for extended Period of time with significant leaks in the primary system is at or near rated operating temperature and pressure. The 24-hour period to provide inerting is judged to be sufficient to perform the leak inspection and establish the required oxygen concentration.
179-05/15/89
3.7.A & 4.7.A BASES (cont'd)
~
The primary containment is normally slightly pressurized during periods of reactor operation.
Nitrogen used for inerting could leak out of the containment but air could not leak in to increase oxygen concentration.
Once the containment is filled with nitrogen to the required concentration, no monitoring of oxygen concentration is necessary.
However, at least twice a week the oxygen concentration will be determined as added assurance.
The 500-gallon conskrvative limit on the nitrogen storage tank assures that adequate time is available to get the tank refilled assuming normal plant operation.
The estimated maximum makeup rate is -1500 SCFD which would require about 160 gallons for a 10 day makeup requirement.
The normal leak rate should be about 200 SCFD.
3.7.A.6 & 4.7.A.6 LOV-LOW SET RELIEF FUNCTION The low-low set relief logic is an automatic safety relief valve (SRV) control system designed to mitigate the postulated thrust load concern of subsequent actuations of SRV's during certain transients (such as inadvertent MSIV closure) and small and intermediate break loss-of-coolant accident (LOCA) events.
The setpoints used in Section 3.7 A.6.b are based upon a minimum blowdown range to provide adequate time between valve actuations to allow the SRV discharge line high water leg to clear, coupled with consideration of instrument inaccuracy and the main steam isolation valve. isolation setpoint.
The as-found setpoint for NBI-PS-51A, the pressure switch controlling the opening of RV-71D, must be 5 1040 psig.
The as-found closing setpoint for.NBI-PS-SlB must be at least 90 psig less than SLA, and must be 2 850 psig.
The as-found setpoint for NBI-PS-51C, pressure switch controlling the opening of RV-71F must be s 1050 psig.
The as-found closing setpoint for NBI-PS-51D must be at least 90 psig below SlC. and must be 2 850 psig.
This ensures that the analytical upper limit for the opening retpoint (1050 psig), the analytical lower limit on the closing setpoint (850 psig) and the analytical limit on the blowdown range (2 90 psig) for the Low-Low Set Relief Function are not exceeded. Although the specified instrument setpoint tolerance is i 20 psig, an instrument drift of i 25 psig was used in the analysis to ensure adequate margin in determining the valve opening and closing setpoints. The opening setpoint is set such that, if both the lowest set non-LLS S/RV and the highest set of the two LLS S/RVs drift 25 psig in the worst case directions, the LLS S/RVs will still control subsequent S/RV actuations.
Likewise, the closing seepoint is set to ensure the LLS S/RV closing setpoint remains above the MSIV low pressure trip.
The 90 psig blowdown provides adequate energy release from the vessel to ensure time for the water leg to clear between subsequent S/RV actuations.
3.7.B & 3.7.C STANDBY GAS TREATMENT SYSTEM AND SECONDARY CONTAINMENT i
The secondary containment is designed to minimize any ground level release of j
radioactive materials which might result from a serious accident.
The reactor building provides secondary containment during reactor operation when the drywell is sealed and in service.
The reactor building provides primary containment when the reactor is shut down and the drywell is open, as during refueling.
Because the secondary containment is an integral part of the complete containment system.
secondary containment is required at all times that primary containment is required as well as during refueling, and during movement of loads which could potentially damage irradiated fuel in the secondary containment.
Secondary containment may be l'
broken for short periods of time to allow access to the reactor building roof to perform necessary inspections and maintenance.
The Standby Gas Treatment System consists of two, distinct subsystems, each j
containing one exhaust fan and associated filter train, which is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions.
Both Standby Gas Treatment System fans are designed to l i
automatically start upon containment isolation and to maintain the reactor butiding pressure to the design negative pressure so that all leakage should be in-leakage.
Should one subsystem fail to start, the redundant subsystem is designed to start l automatically.
Each of the two fans has 100 percent capacity.
l 10A 4/11 169 l
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" INTENTIONALLY LEFT BLANK" i
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3.7.B & 3.7.C BASES (cont'd)
High efficiency pa;;iculate absolute (HEPA) filters are installed before and after the charcoal adso'rbe rs to minimize potential release of particulates to the environment and to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radiolodine to the environment. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and HEPA filters.
The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected accident conditions.
If the performance of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed.
i only one of the two Standby Gas Treatment subsystems is needed to cleanup the reactor building atmosphere upon containment isolation.
If one subsystem is found to be inoperable, there is no immediate threat to the containment system performance and reactor operation or refueling operation may continue while repairs are being made.
If both subsystems are inoperable, the plant is brought to a condition where the Standby Gas Treatment System is not required.
4.7.B & 4.7.C BASES Standby Gas Treatment System and Secondary Containment Initiating reactor building isolation and operation of the Standby Gas Treatment System to maintain at least a 1/4 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leak tightness of the reactor building and performance of the Standby Cas Treatment System.
Functionally testing the initiating sensors and I associated trip channels demonstrates the capability for automatic actuation.
Periodic testing gives sufficient confidence of reactor building integrity and I
Standby Gas Treatment System performance capability.
Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter.
A 7.8 kw heater is capable of maintaining relative humidity below 70%. Heater capacity and pressure 4
drop should be determined at least once per operating cycle to show system performance capability, The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated.
Tests of the charcoal 4
adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with ANSI N510-1980. The test canisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test.
Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable, all adsorbent in the system shall be replaced
-182-3/11/97
4.7.3 & 4.7.C BASES with an adsorbene qualified according to Table 3.1 of ANSI N509-1980.
The replacement tray for the adsorber tray removed for the test should meet the sam adsorbent quality. Tests of the HEPA filters with DOP aerosol shall be performed in f
accordance to ANSI N510-1980.
Any filters found defective shall be replaced with i
filters qualified pursuant to Regulatory Position C.3.d. of Regulatory Guide 1.52, j
Revision 2, March, 1978.
All elements of the heater should be demonstrated to be functional and operable during the test of heater capacity.
Operation of the heaters will prevent moisture buildup in the filters and adsorber system.
With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal.
Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated.
If system drains are present in the filter /adsorber banks, loop-seals must be used with adequate water level to prevent by-pass leakage from the banks.
If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use.
The determination of significance shall be made by the operator on duty at the time of the incident.
Knowledgeable staff members should be consulted prior to making this determination.
Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability.
If one Standby Gas Treatment subsystem is inoperable, the operable subsystem's operability is verified daily.
This substantiates the availability of the operable subsystem and thus reactor operation or refueling operation can continue for a limited period of time 3.7.D 6 4.7.D BASES Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.
Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident.
The maximum closure times for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.
The USAR identifies those testable primary containment valves that perform an isolation function, and testable penetrations with Double 0-Ring Seals, and testable penetrations with testable Bellows ensuring that any changes thereto receive a 10CFR50.59 review. In addition, plant procedures also identify containment isolation valves, and testablo penetrations with Double 0-Ring Seals, and testable penetrations with testable Bellows changes to these procedures and the USAR are controlled by Technical Specification 6.2.1.A.4 (Administrative Controls).
These valves are highly reliable, have a low s e rvi ce requirement, and most are no rmally closed.
The initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolation. The test interval of once per operating cycle for automatic initiation
3 L S-4-4.7.D BASES (cont'd) results in a failure probability of 1.1 x 10'7 that a line will not isolate.
More frequent testing for valve operability results in a greater assurance that the valve will be operable when needed.
In order to assure that the doses that may result from a steam line break do not exceed the 10CFR100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, including instrument delay, as long as 10.5 seconds.
The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system. Each instrument line contains a 0.25 inch restricting orifice inside the primary containment and an excess flow check valve outside the primary containment. A program for periodic testing and examination of the excess flow check valves is performed as follows:
1.
vessel at pressure sufficient to actuate valves. This could be at time of vessel hydro following a refueling outage.
2.
Isolate sensing line from its instrument at the instrument manifold.
4 3.
Provide means for observing and collecting the instrument drain or vent valve flow.
i 4.
Open vent or drain valve.
a.
Observe flow cessation and any leakage rate.
i b.
Reset valve after test completion.
5.
The head seal leak detection line cannot be tested in this manner. This valve will not be exposed to primary system pressure except under unlikely conditions of seal failure where it could be partially pressurized to reactor pressure.
Any leakage path is restricted at the source and therefore this valve need not be tested. This valve is in a sensing line that is not safety related.
6.
Valves will be accepted if a marked decrease in flow rate is observed and the leakage rate is acceptable.
The operators for containment vent / purge valves PC 230MV, PC-231MV, PC 232MV, and PC-233MV have devices in place to limit the maximum opening angle to 60 degrees. This has been done to ensure these valves are able to close against the maximum differential pressure expected to occur during a design basis LOCA.
om a sw 184 05/15/89 m.
~.
USAR APPENDIX A PRESSURE INTEGRITY OF PIPING AND EQUIPMENT PRESSURE PARTS s
s 1.0 SCOPE This-appendix provides additional information pertinent to the preceding sections concerning the pressure integrity of piping and equipment parts.
Piping and equipment pressure parts are classified according to service and location.
The design, fabrication, inspection, and testing requirements which are defined for the equipment of each classification assure the proper pressure integrity. This Appendix describes the requirements in effect at the time of the original installation of the piping and equipment pressure parts.
The evolution of industry codes and standards, regulatory requirements, fabrication, testing, and erection procedures; and supplementary requirements has resulted in parts of these requirements being superseded.
The new requirements generally result. in an improvement in quality and overall margins over the original requirement.
Upgrades or replacement of piping and equipment pressure parts are performed to these new requirements provided the safety design bases described in the USAR are maintained.
For the purpose of this appendix, the pressure boundary of the process fluid includes but is not necessarily limited to:
branch outlet nozzles or nipples, instrument wells, reservoirs, pump casing closures, blind flanges and similar pressure closures, studs, nuts and fasteners in flanged jointe between pressure parts and bodies and pressure parts of in-line components such as traps and strainers.
Specifically excluded from the scope of this appendix are pressure j
parts such as vessels and heat exchangers or any components which are within the scope of the ASME Pressure Vessel Code,Section III and VIII; and nonpressure parts such as pump motors, shafts, seals, impe11ers, wear rings, valve stems, gland followers, seat rings, guides, yokes, and operators; any nonmetallic material such as packing and gaskets; fasteners not in pressure part joints such as yoke studs and gland follower studs; and washers of any kind.
1.1 Codes and Specifications
(
The piping and equipment pressure parts in this station are designed, fabricated, inspected, and tested in accordance with recognized industrial codes and specifications.
In some cases supplementary requirements are applied to increase safety and operational reliability.
The application of the industrial codes and specifications is defined in this appendix as well as the application of the supplementary requirements. Where conflicts occur between the industrial codes and specifications and the supplementary requirements, the suppicmentary requirements take precedence.
United States of America Standards (USAS) referenced herein have been superceded by ANSI standards.
The edition of the USA standards in effect when bids were made for supplying and installing piping was:
USAS-B31.1.0 - Power Piping (1967)
USAS-B31.7 - Nuclear Power Piping (Feb. 1968) w/
Draft and Errata (June 1968)
A-1-1 07/22/87
CNS 2.0 CLASSIFICATION OF PIPING AND EQUIPMENT PRESSURE PARTS Fo'r the purpose of identification and association of requirements, piping and equipment pressure parts are classified in accordance with one of two basic principles.
2.1 GE Company Classification and Pressure Integrity Requirements Class A Piping and equipment pressure parts which cannot be isolated from the reactor vessel.
Class B Piping and equipment pressure parts, which can be isolated from the reactor vessel by only a single isolation valve.
Class C Piping and equipment pressure parts other than included in Classes A and B, for a high integrity system.
i Class D Piping and equipment pressure parts which serve as an exten-sion of containment and which operate at either pressures greater than 150 psig or temperatures greater than 2120F.
Class E Piping and equipment pressure parts which serve as an exten-sion of containment and which operate at pressures equal to or less than 150 psig or temperatures equal to or less than 2120F.
Class F Piping and equipment pressure parts which transport fibrous or particulate materials such as resins or filter aids and which operate at pressures equal to or less than 150 psig and temperatures equal to or less than 2120F.
Class G Piping and equipment pressure parts used for acids in concen-trations of 60 to 100 percent at ambient temperatures or caustics in concentrations of 50 percent or less at tempera-tures less than 1500F.
Class H Piping and equipment pressure parts used for acids in con-centrations of 10 percent or less.
Class L Piping and equipment pressure parts which require materials considerations to maintain deionized water purity.
Class M Power piping and equipment pressure parts not otherwise class-ified and which are considered within the scope of USAS B31.1.0, Code for Power Piping.
Class N Miscellaneous piping and equipment not otherwise classified and not considered within the scope of USAS B31.1.0, Code for Power Piping.
2.2 Engineer - Constructor's Classification and Definition of Piping and In-Line Pressure Parts N
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CNS l
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For this project, all piping systems or subsystems and all in-line pres-g t
sure parts are functionally classified as IN, IIN IIIN, or IVP, and seismically classified as IS or IIS.
2.2.1 Functional Piping and Equipment Pressure Part Classifications 1.
Class IN nuclear piping and in-line pressure parts are those, whose loss or failure could cause or increase the severity of a nuclear incident.
2.
Class IIN nuclear piping and in-line pressure parts are those, whose loss or failure could cause a hazard to plant personnel, but would represent no hazard to the public.
3.
Class IIIN nuclear piping and in-line pressure parts, are those that normally would be Class IIN, except that the operating pressure does not exceed 150 psig and the operating temperature is below 2120F.
4.
Class IVP power piping and in-line pressure parts are those, which are ccnventional steam and service piping and equipment pressure parts.
2.2.2 Seismic Piping Classifications 1.
Class IS seismic piping and in-line pressure parts are those, whose failure would cause significant release of radioactivity or which are vital to a safe shutdown of the plant and removal of decay and sensible heat.
2.
Class IIS seismic piping and in-line pressure parts are those, which may be essential to the operation of the station, but which are not essential to i
l 2.3 Tabulation of Classification Equivalencies l
Classification in Accordance with Definitions of:
GE Company Engineer-Constructor A and B IN/IS C and D IIN/IS and IIN/IIS E and F IIIN/IS and IIIN/IIS i
F,G,H,L,M and N IVP/IS and IVP/IIS 2.4 Engineer-Constructor's Classification and Definition of Equipment Equipment is classified by seismic requirements as follows:
1.
Class I equipment is that whose failure would cause significant re-lease of radioactivity or which is vital to a safe shutdown of the plant and remov-al of decay and sensible heat.
]
l A-2-2
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2.
Class 11 equipment is that which may be essential to the operation
-g of the station but which is not essential to a safe shutdown.
A-2-3
CNS 3.0 DESIGN REQUIREMENTS 3.1 Piping Design All piping is designed in accordance with USAS B31.1.0," Power Piping".
Class IN/IS piping is also designed to meet the requirements of Appendix C which outlines loading criteria to be met for high reliability for piping designed to rational stress analysis techniques.
All other Class IS piping is designed to meet the supplementary requirements included in this appendix, Subsection A-3.1.1.
The terms utilized in this Subsection A-3.1 are either defined in the text, or per-tain to definitions of USAS B31.1.0.
Selection of design earthquakes is discussed gin Appendix A of the Cooper Nucicar Station PSAR.
3.1.1 Analysis 3.1.1.1 Primary Stresses (Sp)
Primary stresses are as follows:
1.
Circumferential Primary Stress (S )
R Circumferential primary stresses are below the allowable stress (S ) at the design h
pressure and temperature.
2.
Longitudinal Primary Stresses (S )
t The following loads are considered as producing longitudinal primary stresses:
internal or external pressures; weight loads including valves, insulation, fluids, and equipment; hanger loads; static external loads and reactions; and the inertia
- s load portion of seismic loads.
When the seismic load is due to the maximum probable earthquake (0.lg), the vectorial combination of all longitudinal primary stresses (S,) does t
not exceed 1.2 times the allowable stress (S )-
h When the seismic load is due to the hypothetical maximum possible earthquake (0.20g), the vectorial combination of all longitudinal primary stresses does not exceed 1.8 times the allowable stress (S )*
h 3.1.1.2 Secondary Stresses (S )
E Secondary stresses are determined by use of the maximum shearing stress TMax=1/2%df+4S'=1/2S' 2
e E
i
- where, SE=
jt 4S '
2 t
(See USAS B31.1.0)
The following loads are considered in determining longitudinal secondary stresses:
(a) thermal expansion of piping, (b) movemen* of attachments due to ther-mal expansion, (c) forces applied by other piping systems as a result of their expan-6 sion, (d) any variations in pipe hanger loads resulting from expansion of the system, j
I A-3-1 l
CNS 1
5.0 FABRICATION AND INSTALLATION REQUIREMENTS i
Fabrication and erection of piping and equipment pressure parts are in accordance with USAS B31.1.0, " Power Piping", and the supplementary requirements in schedules FIN, FIIN, FIIIN, and FIVP included herein.
These schedules are applied as follows:
Piping and Equipment Fabrication and Pressure Parts Classification Erection Schedules IN FIN IIN FIIN IIIN FIIIN IVP FIVP s
9se A-5-1 i
CNS 6.0 TESTING AND INSPECTION REQUIREMENTS
\\
Testing and inspection of piping and equipment pressure parts are in ac-cordance with USAS B31.1.0, " Power Piping". and the supplementary requirements in schedules TIN, TIIN, TIIIN, and TIVP included herein.
These schedules are applied as follows:
~
Piping and Equipment Inspection and Pressure Parts Classification Test Schedule IN TIN IIN TIIN IIIN TIIIN
~IVP TIVP 6.1 Methods, Techniques and Acceptance Standards 6.1.1 Radiography 6.1.1.1 Welds The radiography of welds, including acceptability standards, are in ac-cordance with the following:
Classification Criteria & Acceptance Standards IN & IIN ASME Boiler and Pressure Vessel Code,Section III, Paragraph N-624 IIIN & IVP ASME B&PV Code,Section I, para. PW-51 and Section VIII, para. VW-51 (a through k).
- 6.1.1.2 Castings Methods and Techniques The radiography of castings employ methods and techniques in accordance with ASTM E94, " Tentative Recommended Practices for Radiographic Testing", to the
, quality level in accordance with ASTM E142, " Standard Method for Controlling Qual-icy of Radiographic Testing".
Acceptance Standards Discontinuities are judged by comparison with ASTM E71, E186, and E280 as appropriate for section thickness.
Discontinuity types A through C of severity level 2 are acceptable; discontinuity _ types beyond C are not acceptable.
6.1.2 Ultrasonic Testine Ultrasonic examination of forgings in Class IN and IIN systems is done in accordance with the following:
s A-6-1
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6.1.2.1 Ultrasonic Examination Ultrasonic examination of pipe, plate and forgings shall be performed, cnd acceptance standards shall comply with the following applicable specifications:
(a) Pipe, (Seamless) ASTM E213.
Ultrasonic inspection of pipe and tubing for longitudinal discontinuities.
(b) Pipe Welded Without Filler Metal, ASTM E273.
Ultrasonic inspection of longitudinal and spiral welds of welded pipe and tubing.
(c)
Fo rgings, Bars, Bolting Materials and Plate, ASTM A388.
Ultrasonic testing and inspection of heavy steel forging.
In examination of plate or bars where the words " forging" or " forgings" appear they are considered to mean plate or bar material.
6.1.2.2 Normal Beam Examination General Acceptance Standards,
The materials shall be considered unacceptable based on the following test indications unless eliminated or repaired:
s (a)
Indications of discontinuities in the material that produce a com-
\\
plete loss of back reflection not associated with the geometric configuration of the piece.
(Complete loss in back reflection is assumed when the back reflection falls below 5 percent of full screen height.)
(b) Traveling indications of discontinuities 10 percent or more of the back reflection.
(A traveling indication is defined as an indication which dis-plays sweep movement of the oscilloscope pattern at a relatively constant amplitude as the search unit is moved along the part being examined.)
6.1.3 Liquid Penetrant Testing Methods, techniques and acceptance standards for liquid penetrant testing cre in accordance with the following:
Classification Criteria & Acceptance Standards 1%5 s l A &d.9 4 a b VM IN, IIN, IIIN ASME - Section III, Paragraph N-627 or ASME B&PV Code 6.1.4 Magnetic Particle Testing Methods, techniques and acceptance standards for magnetic particle test-ing are in accordance with the following:
s A-6-2
+w
CNS Classification Criteria & Acceptance Standards IN, IIN, IIIN ASME Section III, Paragraph N-626, Paragraph 1-724 for pipe and fittings.
IVP ASME B&PV Code,Section VIII.. Appendix VI on MS-1, RF-1 systems and 20% random testing on IS (seismic) portion of RCC-1 system.
6.1.5 Hydrostatic Testing Hydrostatic tests of piping and equipment pressure parts are conducted in accordance with the following:
Classification Criteria & Acceptance Standards
'IN, IIN-USAS B31.1.0 and the applicable sections of other IIIN, IVP published piping codes referenced in ASME Section III and applicable to nuclear power piping.
USAS B31.1.0, "Section 137".
6.2 Personnel Qualification Reouirements (Pressure containing components in General Electric BWR System Classifi-cations A B, C, D, E, and F.)
The manufacturer of pressure containing components shall be responsible to ensure that personnel who perform nondestructive examina-tions of pressure containing components meet the qualification requirements of Appendix IX, Paragraph IX-325,Section III of the ASME Boiler and Pressure Vessel Code. This shall apply to both the manufacturer's own employees and those of his subvendors.
A-6-3 I
1 CNS l
1 8.0 FABRICATION AND ERECTION SCHEDULE FIN & FIIN t
Paragraphs apply to both Schedule FIN and FIIN unless noted otherwise:
8.1 Welding -
Welding of piping and equipment pressure parts is accomplished according to the following requirements:
8.1.1 Qualification I
All welding, including fillet, seal, repair, and attachment welds, is performed in accordance with written welding procedures.
Procedure qualification and welder performance qualification are in accordance with Section IX of the ASME Boiler and Pressure Vessel Code.
8.1.2-Qualification Records Qualification records and application of welder's identification symbols are in accordance with Section 127.6 of USAS B31.1.0.
8.1.3 Butt Joints Joint design and welding procedures for longitudinal and girth butt joints larger than 2 inches in nominal pipe size are in accordance with General Electric Dwg. 209A4280.
8.1.4 Branch Connections Branch connections are made using fittings to USAS B16.9 8.1.5 Socket Welds Socket welds are employed for nominal pipe size 2 inches and smaller and are in accordance with USAS B31.1.0, Paragraph 127.4.4.
8.1.6 Attachment Welds Attachment of nonpressure-containing parts (such as supports and hangers) to pressure-containing components shall be by full penetration welds with inspection, heat treatment and welding per requirements for butt welds.
8.1.7 Fabrication Reinforcement for Openings Reinforcement is in accordance with the requirements of the applicable sections of published piping codes referenced in ASME Section III applicable to nuclear piping systems.
S.1.8 Welding Procedures and Processes (l)
(1) See Subsection A-8.8.1 on specific limitations on welding austenitic stainless
~~
steel.
A-8-1 s
CNS i
1.' Welding procedures 2.
Repair procedures 3.
Heat.. treatment procedures 4.
Cleaning procedures 5.
Quality Assurance Control Plan (as specified in Appendix D) 8.9 Inspection and Testing Inspection and testing of piping and equipment pressure parts, including completed welds, assemblies, and subassemblies, is performed as shown in the appli-cable schedule for the specific classification of piping and equipment pressure parts (see Subsection A-6.0).
d i
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CNS 13.0 INSPECTION AND TESTING SCHEDULE TIN
's Refer to Subsection A-6.0 for application of this schedule and for test methods, techniques, and acceptance standards.
13.1 Certification The manufacturer of the materials or components certifies that the require-ments for which he is responsible, including those of this appendix as well as those of the specific material specification, are fully satisfied.
13.2 Hydrostatic Tests Piping and equipment pressure parts are hydrostatically tested.
If any repairs are made, the piping or equipment pressure part is recested.
If any omis-sions or modifications of the test requirement are made, the deviation is shown valid before approval.
l 13.3 Nondestructive Testing 13.3.1 Welds Girth and longitudinal pressure containing complete penetration groove butt welds are 100% examined by radiography.
Accessible surfaces of the weld and s
adjacent base metal are examined by either liquid penetrant or magnetic particle methods.
Fillet welds, socket welds, and nonpressure containing attachment velds such as supports, lugs, anchors, and guides are examined on all accessibic surfaces by either liquid penetrant or magnetic particle methods.
Radiography is not re-quired.
Welds attaching branch connections larger than 4 inches in pipe size are 100% examined by radiography, and accessible surfaces of the weld and adjacent base metal are examined by either liquid penetrant or magnetic particle methods.
Welds attaching branch connections 4 inches and smaller are examined by either liquid penetrant or magnetic particle methods on the accessible surfaces of the weld and adjacent base metal.
Ultrasonic examination is performed whenever required in accordance with Subsection A-6.1.2.
13.3.2 Double-Welded Joints The back of the first side welded shall be ground or chipped to sound metal and visually inspected prior to welding the second side.
13.3.3 Castings Castings for pressure containing components larger than 4 inches are 100%
examined by radiography and all accessible surfaces, including machined surfaces A-13-1
i CNS and castings 4 inches and smaller are examined by either the magnetic particle or 4
the liquid penetrant method.
!w 13.3.4 Forgings Forgings for pressure containing components over 4 inches nominal dia-meter are. examined in the finished condition by ultrasonic inspection; components 4 inches and smaller on all accessible surfaces including machined surfaces, by i
either the liquid penetrant or the magnetic particle method.
13.4 Submittals j
Approval is required for the following inspection and test procedures:
1.
Radiography 2.
Ultrasonic testing 3.
Liquid penetrant testing 4.
Magnetic particle testing.
i i
i 5
6 A-13-2
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v
r CNS 14.0 INSPECTION AND TESTING SCHEDULE TIIN Refer to Subsection A-6.0 for application of this schedule and for test msthods, techniques and acceptance standards.
14.1 Certification The manufacturer of the materials or components certifies that the require-ments for which he is responsible including those included in this appendix as well as those of the specific material specification, are fully satisfied.
14.2 Jiydrostatic Tests Piping and equipment pressure parts are hydrostatically tested.
If any repairs are made, the piping or equipment pressure part is retested.
If any omis-sions or modifications of the test requirement are made, the deviation is shown valid before approval.
14.3 Nondestructive Testing 14.3.1 Welds Girth and longitudinal pressure containing complete penetration groove butt velds are 100% examined by radiography.
Fillet welds, socket welds, and nonpressure-containing attachment welds such as supports, lugs, anchors, and guides are examined on all accessible surfaces s
by either the liquid penetrant or the magnetic particle method.
Radiography is not required.
Welds attaching branch connections larger than 4 inches in pipe size are 100% examined by radiography, except where configuration does not permit effective radiography; then the root and final pass is examined by liquid penetrant or mag-notic particle methods.
Accessible surfaces of the weld and adjacent base metal of branch connec-tions 4 inches and less in pipe size are examined by either the liquid penetrant or the magnetic particle method.
Ultrasonic examination is not required.
14.3.1.1 Double-Welded Joints The back of the first side welded is ground or chipped to sound metal and visually inspected prior to welding the second side.
14.3.2 Castings Castings for pressure containind components larger than 4 inches are 100%
examined by radiography and in the finished condition on all accessible machined surfaces by either the liquid penetrant or the magnetic particle method.
2 1
A-] 4 -1
$ri.
y Castings for pressure'containing components 4 inches nominal size and smaller do not require special non-destructive testing beyond non-destructive test-
.ing per materials specification.
14.3.3 Forginas -
i Forgings for pressure containing components larger than 4 inches in nomi-nal pipe size are examined in the finished condition on all accessible surfaces including machined surfaces by either the liquid penetrant or the magnetic particle method.
14.A Submittals i
Approval is required for the following inspection and test procedurest 1
1.
Radiography 2.
Ultrasonic testing 3.
Liquid penetrant testing 4.
Magnetic particle testing T
4
\\
A-14-2
^'-
" APPENDIX F I
-CONFORMANCE TO AEC GENERAL DESIGN CRITERIA PAGE 1.0
SUMMARY
DESCRIPTION F-1-1 2.0 CRITERION CONFORMANCE F-2-1 2.1 Group I -- Overall Plant Requirements (Criteria 1-5)
F-2-1 2.2 Group II -- Protection by Multiple Fission Barriers (Criteria 6-10)
F-2-2 2.3 Group III -- Nuclear and Radiation Controls
_(Criteria 11-18)
F-2-3 2.4 Group IV -- Reliability and Testability of Protection Systems (Criteria 19-26)
F-2-5 2.5 Group V -- Reactivity Control (Criteria 27-32)
F-2-7 2.6 Group VI -- Reactor Coolant Pressure Boundary (Criteria 33-36)
F-2-8 2.7 Group VII -- Engineered Safety Features (Criteria 37-65)
F-2-10 2.7.1 General Requirements for Engineered Safety Features (Criteria 37-43)
F-2-11 2.7.2 Emergency Core Cooling Systems (Criteria 44-48)
F-2-12 2.7.3 Containment (Criteria 49-57)
F-2-13 2.7.4 Containment Pressure Reducing Systems F-2-15 2.7.5 Air Cleanup Systems F-2-16 2.8 Group VIII -- Fuel and Waste Storage Systems (Criteria 66-69)
F-2-16 2.9 Group IX -- Plant Effluents (Criterion 70)
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f-1-1
- . ')
CRITERION CONFORMANCE 2.1 Group 1 -- Overall Plant Reouirements (Criteria 1-5)
The purpose of these criteria'is to insure that those systems and compon-ents of the station which have a vital role in the prevention or mitigation of con-sequences of accidents affecting public health and safety are designed and construc-ted to high quality standards which include consideration of natural phenomena and fire.
Also, there must be sufficient surveillance and record keeping during fab-rication and construction to ensure that these high quality standards have been met.
As the station consists of a single nuclear plant, Criterion 4, Sharing nf Systems, is not applicable.
It will be seen that the concerns of these criteria have becn properly considered throughout the design of the station.
Criterion 1 -- Quality Standards thorough quality assurance program has been undertaken during design a
construction of the station to ensure that highest quality standards were used.
2n:
A:plicaolo codes were used where they were sufficient and more stringent require-ments were placed on the design, where available codes were not sufficient.
The uality assurance program is presented in Appendix D.
The description of the
/arious systems and components includes the codes and standards that are met in the Jesien and.their adequacy.
References:
Subsections I-5, :-10, III-2 through III-8, IV-1 through
- 7-8. VII-2 through VII-5, Sections V, VI, VIII, and Appendix D.
Criterion 2 -- Performance Standards Conformance to the structural loading criteria presented in Appendix C insures that those systems and components affected by this criterion are designed and built to withstand the forces that might be imposed by the occurrence of the various natural phenomena mentioned in the criterion, and this presents no risk to
- ne health and safety of the public.
The phenomena considered and margins of safety are also given.
References:
Subsections I-5, XII-2 and Appendix C.
Criterion 3 -- Fire Protection As described in Subsection X-9, the materials and layout used in the station design nave been chosen to minimize the possibility and to mitigate the effects cf fire.
Sufficient fire protection equipment is provided in the unlikely event of a fire, and in no case will the ability of the station to be shutdown be compromised-by fire.
References:
Subsection X-9,Section XII.
Criterion 5 -- Records Recuirement l
Complete records of the as-built design of the station, changes during eperation and quality assurance records will be maintained throughout the life of the station.
emer F-2-1
CNS Criterion 9 -- Reactor Coolant Pressure Boundary (Nuclear System Process Barrier)
The nuclear system process barrier consists of the vessels, pipes, pumps,
- abes and similar process components that contain steam, water, gases, and radio-active materials coming from, going to, or in communication with the reactor core.
These are described primarily in Section IV " Reactor Coolant System".
The reactor coolant system is designed to carry its dead weight and specified live loads sep-arately or concurrently; these include pressure and temperature stresses, vibrations, and seismic loads prescribed for the station.
Provisions are made to control or shutdown the reactor coolant system in the event of malfunction of operating equip-ment or leakage of coolant from the system.
The reactor vessel and support struc-tures are designed, within tae limits of applicable criteria for low probability accident conditions, to withstand the forces that would be created by a full area flow of any vessel nozzle to the containment atmosphere witn the reactor vessel at design pressure concurrent with the station maximum earthquake loads.
References:
Subsections I-5, IV-2, IV-3, IV-4, IV-10, VII-8, XII-2, (IV-i. XIV-6, Appendix A and Appendix C.
Criterion 10 -- Containment Two containment systems are provided; the drywell suppression chamber rimarS containment and the reactor building (secondary containment).
These are described in Section V.
The primary containment system is designed, fabricated, and erected to accommodate without failure the pressures and temperatures resulting from or sub-sequent to the double-ended rupture or equivalent failure of any coolant pipe with-in the primary containment.
The reactor building, encompassing the primary contain-ment system, provides secondary containment when the primary containment is closed and in service, and provides for primary containment when the primary containment is open.
The two containment systems and such other associated engineered safe-guards as may be necessary are designed and maintained so that off-site doses re-sulting frem postulated design basis accidents are below the values stated in
'0CFR100.
References:
Subsections V-2, V-3, XIV-4, and XIV-6.
2.3 Group III -- Nuclear and Radiation Controls (Criteria 11-18)
These criteria identify and define the station instrumentation and control systems necessary for maintaining the station in a safe operational status.
This also includes determining the adequacy of radiation shielding, effluent monitoring, and fission process controls, and providing for the effective sensing of abnormal conditions and initiation of nuclear safety systems and engineered safeguards.
To satisfy the intent of these criteria the station is provided with a comorenensive control and instrumentation system, most of which is described in Section VII.
Control of the station is from a central control room.
Shielding and radiation protection are discussed in Subsection XII-3.
F-2-3
CNS initiate the necessary ocess control systems and overrides all other controls to Afety actions.
Subsections I-5, VI, VII-2 through VII-5, and VII-L2,
References:
Criterion 16 -- Monitoring Reactor Coolant Pressure Boundary _
The methods of detecting leakage through the reactor coolant pressure coundary, and the limits imposed on this leakage, are discussed in Subsection IV-10.
Subsections I-4, IV-10, V-2, VII-8, and X-14.
References:
Criterion 17 -- Monitoring Radioactive Releases The station process and area radiation monitoring systems and station parameters from specific sampling procedures are provided for monitoring significant station process systems and specific areas including the station effluents to the site environs and to provide alarms and signals for appropriate corrective actions.
These are described in Subsections VII-12 and VII-13.
References 1 Subsections I-4, VII-12, 'lII-13. IX-2 and IX-4 Criterion 18 -- Monitoring Fuel and Waste Storace fuel storage areas have been analyzed to determine The new and spent Control their safety, and instrumentation is provided for monitoring where needed.
and monitoring of waste storage is provided as described in S ection IX, S ubsection
!II-12 and X-5.
=<
Subsections I-5, VII-12, VII-13, IX-2, IX-4, and X-5.
References:
2.4 Group IV - Reliability and Testability of Protection Systems (Criteria 19-26)
The purpose of these criteria is to ensure that the systems used to pre-vent breach of the clad barrier will:
(1) function when needed in spite of the failure of a component within the system, (2) be designed such that a condition the proper functioning of that sys-requiring a protection system will not preventand (3) be designed so that each channel o Protection tem, of other channels within that system and the control systems.
ant system testability and detection of failures within the protection systems are nec-As seen in the de essary to ensure the reliability of these systems.
sufficient attention has been paid to component descriptions of these systems, independence and power supply, to ensure reliability, system testability and alarms, to these criteria.
The des-the protection systems are adequate with respect that cription of these systems appears largely in Section VII of the CNS-SAR.
Criterion 19 -- Protection Systems Reliability The components of the protectinn systems are designed to a high standard Each system is designed with provisions for testing which approx-of reliability.
imate very closely the functioning of the system under design conditions of that system.
F-2-5
l USAR Criterion 25 -- Demonstration of Functional Onerability of Protection Systems All of the protection systems contain sufficient test signals, bypasses and indicators to allow testing.of the system under simulated conditions closely approximating the actual condition for which the protective action is required.
Provisions are also included to automatically override any testing being carried on, should the channel under test be needed for a protective action.
References:
Subsections I-5, VI-7, VII-2 through VII-5, and VII 12.
Criterion 26
- Protection Systems Fail-Safe Desien Systems essential to the protection functions are designed to fail safe in their most probable failure modes.
- Thus, a systematic or environmentally caused failure will be indicated and will not compromise the protective function of the system.
References:
Subsections I-5, VI-l through VI-6, VII-2 through VII-5, VIII-4 and VIII-5.
2.5 Groue V -- Reactivity Control (Criteria 27-32)
Conformance to these six criteria provides assurance that the reactor core can be made and held suberitical from normal operation or from no rmal anticipated operational transients, by at least two reactivity control systems and that malfunction of a reactivity control system will not result in unacceptable damage to the fuel, rupture of the reactor coolant pressure boundary, or disrupt the core to the point of preventing core standby cooling if needed. Two systems, an operational control system, consisting of moveable control rods, and control by recirculation flow control; and a standby liquid control system are provided to meet the intent of these criteria. The moveable control rod system design is given in Subsection III-4 and control of the moveable rod system is described in Subsection VII-7; the nuclear design, including the control rod reactivity worths, is given in Subsection III-6; reactor coolant recirculation system flow control is described in Subsection VII-9; and the standby liquid control system is described in Subsection III 8.
Criterion 27 -- Redundancy of Reactivity Control The two reactivity control systems provided are completely independent and of different principal.
The operational control system accommodates fuel burnup, load changes and long-term reactivity changes. The standby liquid control system provides independent shutdown capability if it is needed.
References:
Subsection I-5, III-4, III-9, and VII-7.
l Criterion 28 -- Reactivity Hot Shutdown Canability 30th the control rod system and the standby liquid control system are capable of making and holding the core suberitical from any hot standby or hot operating condition up through full power. Consistent with current practice, this F-2-7 07/22/91 l
l
CNS coolant system design, described in Section IV and Subsection III-3, together with i
- he quality assurance program (Appendix D), show that these criteria have been prop-crly considered.
In-service inspection of components and parts inside this bound-ary is discussed in Appendix J.
l 4-Criterion 33 -- Reactor Coolant Pressure Boundary Capability As shoen in Section XIV, the consequences of the design basis rod drop accident cannot result in damage (either by motion or rupture) to the nuclear system l
process barrier.
This is due to the inherent safety features of the reactor core design combined with the control rod velocity limiter.
References:
Subsections I-5, III-3 through III-6, IV-2, IV-5 IV-6, and XIV-4 through XIV-6.
Criterion 34 -- Reactor Coolant Pressure Boundarv Rapid Propagation Failure Prevention The ASME and USASI Codes are used as the established and acceptable cri-teria for design, fabrication, and operation of components of the nuclear system
]
primary barrier.
The nuclear system primary barrier is designed and fabricated to i
meet the following, as a minimum:
1.
Reactor Vessel--ASME Boiler and Pressure Vessel Code,Section III Nuclear Vessels, Subsection A.
2.
Pumps--ASME Boiler and Pressure Vessel Code,Section III,1;uclear h=<
Vessels, Subsection C.
3.
Piping and Valves--USAS B.31.1, Code for Pressure Power Piping.
The brittle fracture failure mode of the nuclear system primary barrier components is prevented by control of the notch toughness properties of ferritic steel.
This control is exercised in the selection of materials and fabrication of i.
equipment and components.
In the design, appropriate consideration is given to the different notch toughness requirements of each of the various ferritic steel product forms, including weld and heat-affected zones.
In this way, assurance is provided that brittle fracture is prevented under all potential service loading temperatures.
References:
Subsections III-3, IV-2, IV-3, VII-8, Appendix A and Appendix D.
Criterion 35 -- Reactor Coolant Pressure Boundary Brittle Fracture Prevention The applicant's selected approach to brittle fracture prevention is to use a temperature based rule with modifications drawn from fracture mechanics techno-4 logy.
The approach, which is generally accepted by materials specialists, esta-blishes the requirements for brittle fracture prevention.
These requirements are less stringent, when measured in terms of NDTT requirement, for thin section mat-erials than the thick section materials assumed in the first draft of this criter-ion.
mm/
f F-2-9
1 CNS cipated and credible phenomena associated with the station operational transients or design basis accidents being considered.
While the first seven criteria are s
applicable to all of the engineered safety features, the remaining criteria fall into four groups:
emergency core cooling systems (Criteria 44-48); containment s
(Criteria 49-57); containment pressure reducing systems (Criteria 58-61); and air cleanup systems (Criteria 62-65).
Examination of each of these safety features will show that their design conforms to the Group VII Criteria.
2.7.1 General Requirements for Engineered Safety Features (Criteria 37-43)
Criterion 37 -- Engineered Safety Features Basis for Design
]
The normal station control systems maintain station variables within operating limits.
These systems are thoroughly engineered and b,cked up by a sig-nificant amount of experience in system design and operation.
Even if an improbable maloperation or equipment failure occurs (including a nuclear system process barrier break up to and including the circumferential rupture of any pipe in that barrier),
the nuclear safety systems and engineered safeguards limit the effects to levels well below those which are of public safety concern.
These engineered safety feat-ures include those systems which are essential to the containment, isolation, and
- ore standby cooling functions.
References:
Subsections I-5, III-3, III-4, IV-2, IV-4, IV-6, V-2, V-3, VI-l througn VI-7, VII-2 through VII-4, VIII-4 through VIII-6, and VIV-1 through XIV-7.
Criterion 38 -- Reliability and Testability of Engineered Safety Features The design of each of the systems essential to the engineered safety feat-ures includes the use of highly reliable components and provides for ready testa-bility of these systems.
Extensive analytical and experimental programs have shown that these systems are capable of performing their designated tasks.
References:
Subsections I-5, III-4, III-5, IV-6, V-2, V-3, VI-6, VII-2, VII-4, VII-3, VII-12, and VIII-4 through VIII-6.
Criterion 39 -- Emergency Power for Engineered Safety Features With the redundant, full capacity diesel generators and batteries and re-dundant sources of offsite power, adequate power sources to accomplish all required safety functions under postulated design basis accident conditions is assured.
Furthermore, each power source can be periodically tested for availability.
References:
Subsections VII-2, VII-3, VII-4, and VIII-2 through VIII-6.
Criterion 40 -- Missile Protection The systems and equipment which are required to function af ter design basis accidents or abnormal operational transients are designed to withstand the most severe forces and environmental effects, including missiles from station equip-ment failures anticipated from the accidents and missiles generated by tornadoes, without impairment of their performance capability.
References:
Subsections V-2, XII-2, and Appendix C.
F-2-11
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i 1
CNS l
Criterion 46 -- Testing of Emergency Core Cooling System Components To assure that the CSCS functions properly, if needed, specific provisions have been made for testing the operability and functional performance of each act-l ive component of each system.
References:
Subsections I-5, VI-6, and VII-4.
Criterion 47 -- Testing of Emergency Core Cooling Systems Specific provisions such as recirculation loops have been provided in the 4
CSCS design to allow periodic testing of the delivery capability of these systems with conditions as close to accident conditions as possible.
References:
Subsections VI-6, and VII-4 Criterion 48 -- Testing of Operational Sequence of Emergency Core Cooling Systems To assure that the CSCS functions properly, if needed, specific provisions nave been made for testing the sequential operability and functional performance of each individual system.
Testing of the systems is done in parts rather than testing of the entire operational sequence.
This is due to the unavailability of these systems during a complete operational test as described, particularly since it may be extremely difficult to perform such a test during reactor operation.
The design complications which will be required in order to permit such a test compli-cates an already complex system, which may be detrimental to safety.
v
References:
Subsections I-5, VI-4, VI-6, VII-4, VIII-5, VIII-6, and X-8.
2.7.3 Containment (Criteria 49-57)
Criterion 49 -- Containment Design Basis The primary containment structure, including access openings and penetra-tions, is designed to withstand the peak accident pressure and temperatures which could occur due to the postulated design basis loss-of-coolant accident.
The con-tainment design includes considerable allowance for energy addition from metal-water or other chemical reactions beyono conditions that could exist during the accident.
References:
Subsections I-5, IV-6, V-2, V-3, VI-1, VI-2, VI-5, VII-3, VII-4, XIV-2 through XIV-7, and Appendix C.
Criterion 50 -- NDTT Requirement for Containment Material The design of the containment and its material are described in Subsection V-2.
The criterion as stated is considered to be overly specific, considering the deneral nature of the other criteria.
In keeping with the intent of these criteria to serve as a general guide, this criterion is interpreted to mean that the contain-ment will be designed in accordance with applicable engineering codes.
References:
Subsections V-2 and V-3.
s l
F-2-13 i
f 1
j
CNS during. stat. ion lifetime.
Such tests will be made at a pressure which permits extra-polation of results to the design pressure condition, using relationships establish-ed Initially for comparative leakage at these low conditions."
Provisions have been included in the station design for periodic leakage rate testing as described above.
Reference:
Subsection V-2.
Criterion 56 -- Provisions for Testing of Penetrations Provisions are made to demonstrate leak tightness at design pressure of all resilient seals and expansion bellows on containment penetrations on an indivi-dual basis.
Reference:
Subsections V-2 and V-3.
Criterion 57 -- Provisions for Testing of Isolation Valves Provisions are also made for demonstrating the functional performance of containment system isolation valves and monitoring valve leakage.
References:
Subsections IV-6 IV-10, V-2, VII-3, and VII-12.
2.7.4 Containment Pressure Reducing Systems Criterion 58 -- Inspection of Containment Pressure Reducing Svstems v
The containment spray cooling system, an integral part of the residual heat removal system, is designed to allow periodic inspection of the pumps, pump motors, valves, heat exchangers, and piping of this system.
The torus and torus water and the spray nozzles may also be periodically inspected.
References:
Subsections IV-8, V-2, V-3, VI-4, VI-6, X-6, X-8, and XII-2.
Criterion 59 -- Testing of Containment Pressure Reducing Systems Components All of the valves and pumps of these systems can be tested periodically for operability and capability to perform as required.
References:
Subsections IV-8, V-2, VI-4, VI-6, VII-3, VII-4, X-6, and X-8.
1 Criterion 60 -- Testing of Containment Sprav Svstems The capability to test the functional performance of the containment spray cooling system is provided by inclusion in the design of appropriate test connec-tions.
References:
Subsections IV-8, VI-4, VI-6, and VII-7.
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