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Forwards Info Required for NRC to Review IST Program & Requests for Relief from ASME Code Section XI Requirements
	ML20125B950
Person / Time
Site:	Monticello
Issue date:	12/14/1977
From:	Mayer L; NORTHERN STATES POWER CO.
To:	; NRC
References
	NUDOCS 9212100225
	Download: ML20125B950 (134)
	v • d • e

{{#Wiki_filter:. MSP R!GUK02Y DOCET E COPY NORTHERN STATES POWEM COMPANY Q'_1 ' " /, \\ K Decembre 14, 1977 ?' N3([ l/l/p 7 Lv Q 4 Wcae ~ J U.s.socan,,1977 I Mr Victor 9tello, Di roi t or i Division of eperatIng Heactors 9 'g8 N c/o Distribution Services Branch, LP1. AD;t y 11 S Nucienr Eegulatory Consnission 4 k'a sh i n g t on. DC 20555

Dear Mr Stello:

MONTICELLO UltCl.;A CENERATING P! ANT Docket No. 50-263 Liccase No. DPR-22 Information Required for NPC Review of Inservice Irispection and Testing Program and Requesta for Relief from ASMF code Sec'ior XI Requirements The inservice inspection testing requirements for nuclear power t lant compon-ents delineated in 10CFR50, Section 50.55a, were changed by a rMston to the Regulations published on February 27, 1976. As a result of thf2 revi si on, Northern States Power Company must comply with the examinatio ard +est ing requirements contained in current e<litions of Sect ion XI of the AT:E l u t ter and Pressure Vessel Code. In a letter dated November 24, 1976 f rou Mr D L Ziemann, Chiet Operating Reactors Branch No. 2, (TSURC, Nort e n States Power Compar. 54; advised of the h procedure to follow in implementing the revi sions to 10CTR50, &c aon 50.55a. This letter a equested the submissic of a description of our ptanned inservice inspection a. d testing program and requests for relief f rom ASME Code require-ments. The attached report entitled, ASME C,de Section XI Inservice Inspection and Testing Program Beginning February 28, 1978 and Information Required for NRC Review of Requests for Relief f rom AS'E Code Section XI Requirements", provides all of the information requested in your November 24, 1977 letter with the exception of: a) Identification of components to be subjected to leakage, hydrostatic and pressure tests e) Identification of Quality c roup C coraponents to be visually inspected It is our intention to supply thie irfo e ion in the fonn of system drawings showing Quality Group boundaries. U< recific list of Quality Group C 9212100225 771214 PDR ADOCK 05000263 773490086 G PDR 4

1 NORTHEHN CTATED POWER COMPANY lir Victor Stello Page 2 December 14, 1977 (oraponents (iucluding supports and hangers) will be submitted. We believe these drawings will be sufficiently detailed for your review of our Quality Groun C Program. We are required to irnplement thi s p rogram on Februa ry 28, 1978. To pemit us to meet this schedule, we ask that you complete review of the attached report as soon at consible and issue the necessary Technical Specification changes requested in nur License Amendment Request dated August 30, 1977. You rs very t ruly, h-h&fV L 0 Mayer, PE Manager of Suelear Suppt ". 1xtt /ttet /deh cc: J G Keppler G Charnoff MPCA Attn: J W Ferman Attachment l

4 4 4 NORTilERN STATES POWER COMPANY MONTICELLO NUCLEAR CENERATING PLANT DOCKET No. 50-263 LICENSE NO. DPR-22 ASME CODE SECTION XI INSERVICE INSPECTION AND TESTING PROGRAM BEGINNING FEB".UARY 28, 1978 ANE INFORMATION. REQUIRED FOR NRC REVIEW OF REQUESTS FOR RELIEF FR04 ASME CODE SECTION XI REQUIREMENTS Submitted:. December 14, 1977 _., ~ - -. _ - - ~. - - - - - ~.., _. -.

f ? f .f l a ? f l-1 .i TABLE OF CONTEfffS l SECTION i I 1 Inservice Inspuction Program. Visual Obwrvation and Nondestructive Testing 2 Inservice Inspection. Program - I Pressure Tests 3 Inservice Testing of Pumps and Valves 3 and Valves 1 4 Requests for Relief from ASME Code Section XI Requirements Determined to be Impractical 5 Proposed Technical Specification Changes f i I D - I + b 4 .L. -ii - c_,.._,..._-. m.

!L SECTION 1 INSERVICE INSPECTION PROCP&1 VISUAL 03SERVATION AND IiOI: DESTRUCTIVE TESTINC f ASME Section XI Nondestructive Examination Progra 2 - Class 1 ASME Code Edition and Addenda: 1974 Edition through end including Surraer 1975 Addenda Program Period: February 28, 1978 to June 30,1981 (Third Inspection Period) J' NOTES: 1. The following tables identify the specific Class I cccponents and parts to be examined. These tables can be directly correlated with Table IWB-2500 and Table ITJB-2600 of Section XI identify the examination method for each listed item. The inspections that were completed during period one and period two are identified in the tables, along with the running percent completed during 'each of. these periods. No effort was 12ade to retrofit i i*. Items into the first two periods that tera not previously required for examination. The tables shcv the amount of items required to be examined during period three and the corresponding percentage that will have been con- . pleted by the end of this period. l 2. Repairs will be performed in accordance with the applicable requirements of the latest edition ar;d addenda of the ASME Code, Section XI. However, if rule.c for a particular repair are not specified in Section XI, the original design specification and Construction Code of the component or system, or .later editions of the Construction Code or ASME Code Section III, either in their entirety or portions thereof, may be used. LEGE!G Exam' nation method: V - visual U.T. - ultrasonic R.T. - radiography S - surface. examination, either liquid penetrant or magnetic particle Inspection" Period O*E . June 30, 1971 to October 30, 1975 IVO - Octobei 30, 1974 to February 28,.1978 THREE. February 281.1978 to June 30, 1981 1 r, .__..,__m_.

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These drawings will be submitted as a supplement to this report.

a ASME Section XI Pressure Testing Pror, ram rd ASME Code. Edition and Addenda: 1974 Edition through and including Su=:2er 1975 Addenda Program Period: February 26, 1976 through october 26, 1979 g ]

4 6

< rwah and including Summer 1975 Addenda

seen, AO 2-80C 80-C Phin Steam Isolation 1

n. ((y CS-13-2 14B Core Spray Block 1

(

a I

alve Lineup RCIC RCIC-32 Slone Pump Suction Block 2

ontain' n, AO-2379 None Torus Vac Nx Isolation 2

ontainn.

ontainm.

ontainn.

onta iren.

Z Primary,

ond Serv System Dif-53 None Drvwell Denin Ktr Iso 2

4. 12 1PCCW 50-1426 BD-1426 Drvwell RPCCN Isolation 2

4. 6 1hRII 30-2397 30-2397 Pump Suction Isolation 1

omp Air CV-1478 CV-1478 Drvwell Como Air Iso 2

orn Air AS-39 None Service Air Iso 2

f.

y. y,-,

10. RBQUEST FOR RELIEF Applicable ASE Valve Conponent Function Code Class Category ID-7 Imediately stop the steam flow to the 2

11. REQJEST FOR RELIEF Applicable ASME Valve Component Function Code Class Category RIR SW-17.

12. REQUFAT FOR RELIEF Applicable ASME Valve Component Function Code Class Category IM-58 Shutoff demineralized water to drywell.

13. REQUEST FOR RELIEF Applicable ASIE Valve Component Function Code Class Category XP-6 Prevent reversal of flow of reactor.,

2) Iongitudinal and Circumferential. Welds in Shell (other than 1

1) and 2) Volumetric examination of 107. of each longitudinal weld and 57. of each circumferential. weld

3) Volumetric examination.of nozzle-to-vessel wald and inside radius will not be perfonned l

and Pressure Vessel Code,.1965 Edition, including Addenda through Summer 1966, were satisfied.just as if the 1

vessel'were shop fabricated. :In addition, additional. requirements more ' stringent than those ' required ! by the Code were applied by General Electric due'to the unique circumstances. surrounding the vessel fabrication.

4p-+4 m-

t V

1) Longitudinal and Circumferential Welds in Shell: VLCB-1, 1

1. I*

2) Integrally Welded Vessel Stabilizer Lugs 1

1) Volumetric examination of 107. of each longitudinal veld and 5% of each circumferential veld will not be performed as required by Exem Category B-B.

2) Volumetric examination of the vessel stabilizer lugs will not be performed as required by Exam Category B-H.

1

s. )

23. REQUEST FOR RELIEF 8

C-E-2 4

25. REQUEST FOR RELIFS ASEE COMPONENT FUNCTION Code Viv Class Cat

.s
o t P. Zick v.c. cru,mua W.L HARDING

$vT 56 Prospect Street H

Subject:

ASMC File #BC-76-418 H M CAN AVAN Section XI, Division 1, System Isressure Tests H.J Lts'LUCH , LJ cHOCKif

Dear Mr. Duba:

WI. cooper W O. DOTY G [ FR ATcHER Your inquiry of February 24, 1976 has been considered by the 3, gannison cogninant comrrit tee. We are responding to the following question: E.J HEM 2Y w.p. JOHN %oN yg g3m. E.t KEMMLEn-E.L MVE J E. t ATTAN

1. Are System leakage tests required for Class 2 and Class 3' systems 7-J L',co

2. Are systiem hydrostatic tests required for Class 2 and 3 systems?

y $1 H LioE LLt H Tf NOUHUr REPLY: ct.nAwuNs $ " $$$t is*"'

l. System Icakage tests are not required for Class 2 'and 3, components.

2. System hydrostatic tests are required for. Class 2'and 3 components at or near the end o,f each inspection-interval. In addition, a system hydrostatic test is required on components which have been repaired by welding prior to returning the plant to service.

Very truly y'ours, m' unt June Ling Nuclear Engi ing Administrator JL:lc R E C E IV E Da ENomrRwo i i9 2 ?. nib [ f. E. L KEMu.t.t,1 ".r Dr. g.4,,.. Member of Engincris Council for Professional Development and Engineers Joint Council 4-30 m 9

.m W -

e.es 9 w r.2,vv ,-s-v e W r-* e e-e u-w v'e-,, v v-'s -w-

1

26. RMUEST FOR I0'LTU ASME COMPONENT FUNCTION Code Viv Class Cat i

All Class 1, 2, and 3 Components pressure retaining 1,2,3 CODE REQUIREMENT Not all system pressure tests will be conducted for a minimum of four hours as required by IWA-5210(a). BASIS g This requirement is not practical nor meaningful'when perforlaing pressure tests of areas that are exposed for visual i -i examination. The four hour requirement is based on detection of leakage from insulated areas. Where areas are exposed for visual examination a shorter time period is justified. ALTERNATE INSPECTION (TESTIPC) Where areas of examination are not exposed, the test pressure and temperature will be maintained for a minimumi of four hours as_ required by IWA-5210(a). I Where areas to be examined are exposed for visual examination, the test pressure and temperature will be maintained for a minimum of ten minutes as established by IWA-5210(a) - Winter 1975 Addenda. SCHEDULE FOR IMPLDfENTATION February.28, 1978

5 SECTION 5 - PROPOSED TECHNICAL SPECIFICATION CHANGE Reproduced in this Section are the proposed Technical Specification changes included in Northern States Power Company's License Amendment Bequest _ dated August 30, 1977. Pages

11
iii
v
vii
1x 88 89 90 94 96 97 98 99 100 101 102 103 104 105 106
107 108A ilh 119 120
122 - 129 (renumbered pages 121A - 123) 135 136 137
138 lk6 151 152 189R (new page) 189S (new page) 189T (new page) i l'

l-j

Tnese pages have not been reproduced.

Changes made to these pages l involve only layout, numbering, and_the table of contents. 5-1 o

-4 .-.c_. m. ?<. a, s 1 i e 1 3.0 LIMITING CONDITIONS FOR OPERATION .h.O SURVEILIANCE REQUIRD4ENTS - i j '3.4 STANDBY ' LIQUID CONTROL SYSTEM 4.h STANDBY LIQUID CONTROL SYSTD4 Applicability: Applicability: I AppliesL to the operating status of the Applies to the periodic testing require- -]' . standby liquid control system. ments for 'the' standby. liquid control system. ' 4 Objective: Obj ective: i 1 4-independent ' reaetivity control' mechanism. liquid control system. d To assure the availability of an To verify the operability of the ' standby 4 SPEX'IFICATION : SPECIFICATION: } A. Normal Operation A. 'Ihe opembility of the. standby liquidJ control. system shall be verified by' l ..1. The. standby liquid control system performance of the'following tests: .; g j. - shall be operable at-all times when J T < fuel -is in the reactor.and the-L:f reactor is not ' shutdown by -control.. 1. At.least once each operating cycle rods, except as.:specified in 3.4.B. manually initiate one of.the two ~ standby liquid control systems.and

2. Each standby liq tid control ' system pump Pump. demineralized. water. into the

.,1 . shall be. capable 3r delivering 24 gpm reactor vessel. Both systems shall-( - against -aIreactor pressure of 1275-psig, be tested and inspected in:the. 1 ~ course of two' operating cycles.:- ~3 ' The system pressure : relief valves shall' ~ be. operable with a setpoint.between. .s

4

.-1 50 and 1 50 psig.-

88-F 34/4 A LhEVI

j i 3.0 LIMITIIRG C0erDITIONS FOR OPERATION 4.0 SURVEILIANCE RDQUIRD4DffS

2. Inservice inspection and testirg of components shall be conducted in d

accordance with Specification 4.13 l 1 ) ~ 4 Y te B. Operation with Inoperable Componente B. Surveillance with Inoperable Componente Fmm and after the date that a mdundant When a component becomes inoperable, its component is ende or found to be inoperable, redundant coe snent shall be demonstruted Specification 3 4.A shall be conside md to be operable immediately and daily fulfilled, provided tint the component is the reafte r. returned to an operable condition within seven days. 89 34/4.4 REV

I i %O LIMITING CONDITIONS FOR OPEFJTION hO SURVEILLLNCE RKUIR&M3 i C. Volume-Concentration Requireseents i C. The availability of the proper boron The liquid poison tank shall conta' ~ bearing solution shall be verified by a boron bearing solution that satisfies perfor=nnee of the ro14cving tests: the' volume-concentration requirenents 1. At least once N r month - of Figure 3.h.1 and at all times when the standby liquid poison system is re-quired to be operable the tempersture Baron concentration shall be detervined. In cdditi x, the boron shall not be less than the solution te-P-erature presented in Figure 3.4.2. In concentration shall be determined addition, the heat trscing on the punp any time water or baron are added suction lines shall be operable whenever or if the solution temperuture drops the room tempersture is less than the below the limits specified by u Figure 3.k.2. solution temperature presented in Figure 3.4.2. 1 I 3.4/h.4 99 REV l

j Eases 3.4 and 4.4-V A. The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn con-trol rode can be inserted. To meet this objective, the liquid control system is designed to inject l a quantity of boron which produces a concentration of 900 ppm of boron in the reactor core in less than 125 minutes. 900 ppm boron concentration in the reactor core is required to bring the reactor from fuil power to a 37. o k suberitical condition considering the hot to cold reactivity swing, xenon poisoning and an additional 257. boron concentration margin for possible imperfect mixing of the chamaical solution in the reactor water and dilution from the water in the ecoldown circuit. A ministas net quantity of 1400 gallons of solution having a 21.47. sodium pentaborate concentration is required to meet this shut-down requirement. i he time requirement (125 minutes) for insertion of the borcn solution was selected to override the d rate of reactivity insertion due to cooldown of the reactor following the menon poison peak. The J msximum net storage volume of the boron solation is 2935 gallcas. ( 256 gallons are contained beJov the pump suction and, therefore, have not been used in the net quantities above.) 4 m 1 e \\ Boron concentration, solution te=perature, and volume (including check of tank heater and pipe l heat tracing system) are checked on a frequen, y to assure a high reliability of operation of the system should it ever be required. Experience with pu=p operability demoetrates that testin6 at a three-conth interval is adequate to detect if failures have occurred. Standby liquid control eyetms co ponente are insp-etad and tested in necordance with the requirewnts l of 10 CFR 50, Section 50 55a(g). These requirements are delineated in Specification 4.13 his inspection and testirc program, combined with the additional surveillance requirements contained in this section, provide a high degree of assurance that the standby liquid control system will perfor= as required when needed. The relief valves in the standby liquid control system protect the system piping and positive dis-- placement pumps which are nominally designed for 1500 psi from overpressure. The pressure relier valves discharge back to the standby liquid control solution tank. 3.4/4.4 BASES p arv

S 4 30 LIMITING CONDITIONS FOR OPERATION k.O SURVEIIJX;CE REQUIRDE2iTS 4 3.5 CORE AND CONTAIDMENT COOLING SYSTOG k.5 CORE AND CONTAIf0E2rf COOLING SYSTDG Applicability: Applicability-Applies to the operational status of the energency cooling syste=s. Applies to periodic testing of the emergency cooling syste=s. Objective: Objective: To insure adequate cooling capability for heat removal in the event of a loss of coolant To verify the operability of the emergency cooling systems. accident or isolation frac: the nort:2al reactor heat sink. im Specification: Specifiestion: Iov Pressure Core Cooling Capability Low Pressure Core Cooling Capabilf ty A. Core Spray System A. Surveillance of the core spray system shall be performed as follows: 1. Except as specified in 3 5. A.2., 3.5. A. 3., and 3 5. A.S. below, both core 1. Boutine Testing spray subsystems shall be operable when-- ever irradiated fuel is in the reactor 1 vessel ani reactor coolant water t mpera-

a. A simulated auto::ntic actuation test ture is greater than 21Y2F.

shall te conducted each refueling outag?.

b. Core spray headerap instrumentation shall be checked once each day, tested once each ::2enth, and calibrated once each 3-c:onth period.

3 5/h.5 96 Rav

ei 2# -m - ~ s I 3 1 4 ) i 1 ( i 30 LIMITING CONDITIONS FOR OPERATION - k.O SURVEIILANCE REQUIRIMENTS j i i l 1

c. Inservice inspection and testing of components shall be conducted in accordsnee with Specification h.13 1

's i 4 3 1 'f' + w 3 ' 2. From and after the date that one of the 2. When it is determined that one core core spray systems is mde or found to be spray system is inoperable, the opec- [ inoperable for any reason, reactor opern-able core spray system and the IECI i tion is permissible only during the suc-mode cf the RER system ard the diesel l ceeding fifteen' days:unless such system generators required for operation of is sooner mde operable, provided that-such components (if no external source during such fifteen-days all active cow of power were available) ah=11 be l nents of tne other core spray system and demonstrsted to be ' operable immedia-I' the. LPCI mode of the RHR system and the tely. The operable core spray system i-diesel generators required for operation shall be demonstrated to be_ operable l of such components (if no extermi source. daily thereafter. i-of power were available) shall be operable. 4 3 From and after the date that both core 3 When 'it is de' ermined that both core t spray systems' are made or found to be spray systems are inoperable, the inoperable for any reason, reactor IECI mode of the RHR system and the j; 3 5/h.5 'W c l-REV _ ~.. -,. ..~.. -.

3 0. LIMITIIU CONDITICI!S FOR OPEPATIO'i L.O FtKEIIIAI:N FE;UIRD'.DITS operation is pemissible only during Ilesel generitors required fvr the succeedir4 seven days unless at operation of such coepenents (if least one of such systems is sooner no external source of power were mde operable, provided that during available) shall be deme:nstrated such seven days all active components to be cperable ir=ediately and of the LPCI mode of RHR system and the hily thereafter. diesel. generators required for operation of such components (if no extemal source of power were available) shall be opera-ble. k. Each core spray system shall be capable of delivering 3,020 gpm against a reactur Y pressure of 130 psig. If.this rate of delivery requirement cannot be met, the system shall be considered inoperable. S-If the requirements of 3 5.A.1 - 3 cannot be met, an ^ orderly shutdown of the reactor will be initiated and the reactor water temperstum shall be mduced to less than 212 0F vithin 2h hours. a 3 5/h.5 98 m

a L 30 LIMITING CONDITIONS FOR OPEPATION h.O SURVEILIANCE REQUIRIMDITS B. Lov Pressure Coolant Injection (LPCI) Subsystem B. Surveillance of the Low Pressure Coolant (LPCI mode of EHR cystem) Injection (LPCI) Subsystem (LPCI mcmie of EHR cystem) chall be perfor=ed as follows: f 1. Except as specified in 3.S.B.2 and 3 5.B.3 below, the LPCI shall be operable

1. Routine Testing whenever irradiated fuel is in the reactor

a. A simulated autocntic actuation test i

vessel and reactor coolant temperature is greater than 212 F. shall be conducted each refueltre outage. f 0 b. Inservice inspection and testirg of j components shall be conducted in accor: lance with Specification I.13 4 'T

c. During each five year period, an air test shall be performed on the drywell l

spray headers and nozzles. i 2. Frn: and after the date that one of the 2. When it is deterr.ined that one of the LPCI pumps or cdmission valves is made LPCI pumps is incperable, the rennining er found to be inopernble for any reason, active components of the LPCI and con-reactor operation is permissible orly tainment coolir4 subsystem, both core during the succeedirg thirty days unless spray systems and the diesel generators such pump or admission valve is sooner required for operation of such coe:ponents . rnde operable, provided that iurirg su-h (if na external source of power were l thirty dayu the r m ining active compenents wailable) chall be demonstrsted to be of the i D I and contairent x ; ling sub-uperable inmediately and the operable sya*,c=; and all a ctive e s p rmta >f both LPCI pumps laily thereafter. et.. spray rystems and tne dierel genera-tors required for operation ;f such com j ponents (if no external source of power [ were evn11able) shall be operable. 3 5/a 5 fgy

.. ~. ~ i 30 LIMITING CONDITION 3 FOR OPERATION k.O SURVEILI/ dice REQUIRD4CETs 3 From and after the date that two of the 3 Wen it is detemined that the IECI LPCI pumps or admission valves are made subsystem is inoperable, both core or found to be inoperable for any reason, spray systems, the containment coolina i reactor cperation is pemissible only subsystem, and the diesel generstors during the succeeding seven days unless required for operation of such com-such pu=ps or admission valves are made ponents (if no external source of operable sooner, provided that during such power were available) shall be seven days all active components of both demonstrated to be operable immedi-core spray systems, the containment ately and daily thereafter. cooling subsystem (including 2 LPCI pumps) and the diesel generators required for 4 operation of such components (if no o external source of power were available) shall be demonstrated to be operable at least once each day. h. A maximum of one dryvell spray loop (containment cooling mode of RHR) may be inoperable for 30 days when the reactor water temperature is greater than 212 F. Jr tha loop is not returned to service within 30 days, the orderly shutdown of the reactor will be initiated and the reactor water temperature chall be reduced tn less than 212 F. 0 5 Each LPCI cubsystem (RHR) pump shall be cepble af ieliverirg h,000 gpm against - a reactor pressure of 20 psig. If this 3 5/h.5 100 REV ^

\\ 3.0 LIMITING CONDITIONS FOR OPEPJCIrn L.O SURVEILIAUCE RE71IFCC.TS rate of delivery requirement cannot be met, the pu=p chall be considered inoper-able. 6. If the requirments of 3 5.B.1-4 cannot be met, an onlerly shutdown of the reactor vill be initiated and the reactor water t _rature chall be reduced to less then 212 F vithin 24 hours. Containment Cooling Capability m e Containment Cooling Capability W H C. Residual Heat Removal (RHR) Service khter C. Surveillance of the RHR service water System system shall be perfomed as follows: 1. Except as specified in 3 5.C.2 and 3.S.C.3 below, both RHR service water system loaps 1. Inservice inspection and testing of shall be operable whenever irradiated fuel con:ponents shall be conducted in is in the react 3r vessel and reactar coolant acconiance with Specification 4.13 temperature is grmter than 2120F. 2. From and a f ter the date that une of the 2. BHR service viter system pumps is made or When it is determined that one RHR found to be inaperable for any reason, service water pu=p is inoperable, the redundant coc:ponents of the 101 3 5/4.5 .~. ~ - ~

Ja L 3 6 3.0 LDETIE CONDITION 3 FOR OPERATION k.O SURym1ANCE FREJIRDENTS t reactor operation is pemissible only reminir4 subsystem shall be during the succeeding thirty days unless demonstrated to be operable immedi-l such pump is sooner made operable, pro-ately and daily thereafter. vided.that during such thirty days all other active components of the RER service water system are opemble. 3 From and after the date that one of the 3 'w' hen one RHR service water system RER service water systems is made or found becomes inoperable, the opezeble ^ to be inoperable for any reason, reactor system shall be demonstruted to be' operstion is pemissible only during the opersble inmediately and daily t-succeeding seven days unless such system thereafter. is sooner made operable, provided that during such seven days all active co=po-y' nents of the operable EHR service water g system shall be demonstrated to be opers-ble at least once each day. h. To be considered operable, a RER service. vater ptanp shall be capable'or. delivering 3500 gpa against a head of 500 feet. 5 If the recuirements or 3 5.c.1-3 carmot i be met, an orderly shutdown of the reactor vill bc initiated and the reactor water ter:perature shall be reduced to less than 2120F vithin 24 hours. i' i 3 5/h.5 loe PAv 4 +-w, p m + ' +~, s e e N~-r m'- "w--,*-o-- v~ -er* "----w w

I l l 3.0 LIIIlTING CO!!DITICI:3 FCR CPEFICICII L.O CURVEILLADCE FE:,UIFE"E.75 High Pressure Core Cooling Capabilit. iiirh Fressure Core Cooling capability i D. High Pressure Coolant In. ject ion (HFCI) Synten D. 3nrveillince of EPCI Systen shall te p rfar ed as follows: 1. Except as specified in 3.S.D.2 below, 1. Foutine Testirs the HPCI systen shall be operatle then-ever the reactor pressure is greater than

n. A siz21sted automatic acttantion test 150 psig and irradiated fuel is in the shall be conducted each refueling reacter vessel.

outage.

b. Inservice inspection arxi testing of co::ponents shall be conducted

? in accordance with Sp=cification k.l~. Pw 2. From and after the data that tha HPCI 2. k' hen it is determined that HNI system is made or foun 1 to be ino;. arable system is inoperable, the RCIC system, for any reason, reactor operation is por-the LPCI rubsysten, an1 both af the care nissible only durin.- the succeedin ce.tn spray systens shsll be decionstrated days unless such :,ysten c: noonor rcile t o be gerable irred i ately. opera: Ic provi'Je_d th,*. !ur w h se. <_ n days all of the Aut', at i, Pr+ snure Relief system, the FCIC syrton, ieth of the core spray syatu r, wJ th-LPCI nl:.ystem 2n! contairc.ent coviinc r.O ie of the REF system are operable. 5.5/4.5 105 REV i

. -, ~ M ~' 4 q l ~ !0 LIMITING CONDITIONS FOR OPEFATION 'k.O StJRVEIIIANCE REQUIRC4E*;TS i. 4 3 To be considered operable, the HPCI system shall meet the following conditions: 1 The HPCI shall;be capable of delivering a. l 3,000 spa into the reactor vessel for ' the reactor pressure range of 1120 psig 1-to 150 peig. - b. He condensate storage tanks shall j. contain at least 75,000 gallons of l ]- condensate water. u l- ]; #. c. The controls for autoentic transfer of the HFCI pucip suction from the l' condensate storage tank to the ]. suppression chamber shall be operable. k. If the requirements of 3 5.D.1-2 cannot be met, an orderly reactor shutdown shall be initiated innsediately and the' reactor ) pressure shall be reduced to 150 psig i within 2k hours thereafter. t E 7 e i I a 104 l .3 5/4.5 RW I

t l l i I I I f -3.0 LIMITING CC'IDITIONS FOR OPERATIO;I 4.0 SUR'IEILLECE FEquIREMCITS L E. Automatic Pressure Belief System E. Surveillmace of the Automatic Pressure Belief System shall be performed as follows: j 1. Except as specified in 3 5.E.2 and 3 5.E.3 below, the entire auto :stie

1. Poutire Testing I.

pressure relief system shall be operable l at any time the reactor pressure is above

a. A simulated automatic actuation 150 psig and irradiated fuel is in the test shall be conducted each oper-reactor vessel.

ating cycle. j- [

b. Once each operating cycle, valve oper-2.

From and after the date that one of the ability shall be verified by cycling i-automatic pressure relief system valves is tne valves and observing a compensating rede or. found to be incperable for any change in turbine bypass valve position w .L reason, reactor cperation is permissible m' only during the succeeding seven days Inservice inspection and testing of c. unless such ' valve is sooner :cade operable, components shall be conducted in i accordance with Specification 4.13. provided that during such seven days both remaining autccatJ relief system l valves and' the HPCI system are operable. 3 From and after the date that more: than one of the automatic pressure relier valves are cade or found to be inoperuble l 1 for any reason, reacto* cperation 'is permissible only during the succeeding Een it is determined that one or 2. 5 24 hours unless repairs are made and more automatic preeeure relief welves provided that during such time the EFCI of the Automatic Pressure Relief a j system 1 gerable. system is inoperable, the RPCI system shall be demonstrated to be operable i

k..If the requirements of 3.% E.1-3 cant.at I-diately and weekly thereafter.

i be met, an orderly re. actor shutdown shall i be initiated immediately and the reactor shall be' reduced to 150 psig within 24 3 1 hours thereafter. ] 3 5/4.5 105 REV 4 0 e.. m

3.0 LIMITING CCHDITION5'FGt OPEPATION k.O SURVEILIATE REQUIRDENTS F. -Reactor Core Isolation Cooling System (RCIC) F. Surveillance of Reactor Core Isolation Cooling System (RCIC) Surveillance of the RCIC System aball be perferned as follows: 1. Except as specified in 3.5.F.2 below, the

1. Routine Testing RCIC-syste= shall be operable whenever the reactor pressure is greater than 150

a. A simulated automatic actuation test

~ psig ani irradiated fuel is in the reactor shall be conducted each refueling outage. vessel. a. To be considered o;erable, the RCIC

b. Inservice inspection aal testing of components shall be conducted system shall be capable of delivering 4

LOO gpm into the rmetor vessel. in accordance with Specification 4.13 os 2. From and after the date that the RCIC sys-2. tem is taie or found to Le inoperable for k"nen it is determined that the RCIC sys-any reason, reactor operation is permissible tem is inoperable, the HPCI system shall only'during the succeeding 15 days unless be demonstrated to be operable inur.ediately such system is sooner cade oper able, provided and daily thereafter. that durin6 such 15 days all_ active compo-nents of the HPCI system are operable. 3. If the requirements of 3 5.F.1 - 2 cannot be met, an orderly ' shutdown of the reactor shall be: initiated immediately and the reactor pressure shall be reduced to 150 psig within 2h hours thareafter. g 35/4.5 RW

7 -_. 3 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREN.NTS 1. Recirculation System I. Recirculation System I 1. Except as specified in 3.5.1.2 below, whenever 1. Once per month, when irristed fuel is in the irradiated fuel is in the reactor, with reactor coolant temperature greater than 212*F and both reactor with reactor coolant temperature greater than 212 F and both reactor recirculation reactor recirculation pumps operating, the recirculation system cross tie valve inter 10cks pumps operating, the recirculation system crose shall be operable. tie valve interlocks shall be deemnetrated to be operable by verifying that the cros tie valves cannot be opened using the normel control 2. The recirculation system cross tie valve inter-switch. locks may be inoperable if at least one cross tie valve is maintained fully closed. w 2. When a recirculation system cross tie velve 4 interlock is inoperable, the position of at N least one fully closes cross tie velve shall 3. Valves in the equalizer' piping between the be recorded daily. recirculation loops shall be closed. Reactor operation with one loop shall be limited to 24 hours. 3 Inservice. inspection and testin6 of co:::ponents shall be conducted in accordance with Specification 1.13 4 3.5/4.5 108A REV

~ Bases 4.5: The testing interval for the core and containment cooling systems is based en a quantitative reliability analysis, judgment, and practicality. The core cooling systems have not been designed to be fully testable during operation. For exa~ pie, the core spray final admission valves do not open until reactor Pressure has fallen to 450 psig; thus, during operation even if hi h drywell pressure were simulated, the 6 final valves would not open. In the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel, which is not desirable. The systems can be automatically actuated during a refueling outage and this will be done. To increase the availability of the individual components of the core and containment coolicg systems, the coagnmeents idrich make up the system, i.e., instrumentation, puarps, valve operators, etc., are tested more frequently. The instrumentation will initially be fitnetionally tested once per month until a trend is established and thereafter according to Figure k.1 (see Section 3.1/4.1) with an interval not greater then three months. Core and containment cooling system components are inspected and tested in accordance with the requirements of 10 CFR 50, Section 50 55a(g). These requirements are delineated in Specification cn 4.13 This inspection and testing program, co=bined with the additional curveillance requirements contained in this section, provide a high degree of assurunce that the core and contairment cooling systems will perfom as required when needed. i With components or subsystems out-of-service,. overall core and containment cooling reliability is main-tained by demonstrating the operability of the remaining cooling equipment. The degree of operability to be demonstrated depends on the nature of the reason for the out-of-service equipment. For routine i out-of-service periods caused by preventative maintenance, etc., the pump and valve operability checks will be performed to demonstrate operability of the remaining components. However, if a failure, design deficiency, etc., caused the out-of-service period, then the demonstration of operability should be thorough enough to assure that a similar problan does not exist en the remaining components. For example, if an out-of-service period were caused by fnilure of a pump to deliver rated capacity due to a desigrt deficiency, the other pm ps of tnis type might be subjected to a flow rate test in addition to the operability enecks. 4 4.5 BASES 114 REV

~.- 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 StTRVEILIANCE REQUIREMENTS E. Safety / Relief Valves E. Safety / Relief Valves 1. Durin; power operating conditions and whenever 1. The inteerity of the safety / relief reactor coolant pressure is greater than 110 psig valve bellows shall be continuously and temperature is greater than 345 F : monitored. a. The safety valve function (self-actuation) o f

2. The opembility of the bellows seven safety / relief valves shall be operable, monitorirg system shall be demonstrated at least once every three months.

b. The solenoid activated relief function

3. Inservice inspection ar4 testirg of

( Automatic Pressure Reller) shall be coc:ponents shall be conducted in operable as required by Specification accordance with Epecification I+.13 [

3. % E.

e l 1 I I 1 3.6/4.6 119 REV ^ ~ l l

T^ 7' l I 3 0 - LIMITING CONDITIONS FOR OPEPATION k.0 SURVEILIANCE RB;lUIRDE:GS [ F. deleted F. deleted i + l i . [0. Jet Pumps G. Jet Pumps .o d. Whenever the reactor is in the Startup Whenever there is recirculation flow with the or Run modes, all Jet pumps shall be oper-reactor,in the Startup or Run modes, jet pump able. If it is ' determined that a jet pump is operability shall be checked daily by verify-inoperable, the plant shall be placed in a ing that all-the following conditions do not cold shutdown condition within 24 hours. occur simuftaneously: ~ 1. The two recirculation loop flows are unbalanced by 1% or more when the I recirculation pumps are operating at the same speed. 2. The indicated value of core flow. rate is 10% or more less than the value de-rived frors loop flow measurements. i 3,6/4.6 13 REV -w yr 4-' WY D-- P N

T'bb-

vr-- Tr 1& 1----=--y, vg

q 'l l Bases Continued 3.6 and h.6: h e safety / relief valves have two ibnctions; i.e. power ralief or self-actuated by high pressure. he solenoid actuated function (Autenatic Pressure Eelief) in which external instn=entation signals of coincident high dryvell pressure and low-low veter level initiate opening of the valves. This function is discussed in Specification 3 5.E. In addition, the valves een be operated manually. he safety function is performed by the same safety / relief valve with self-actuated integral bel. lows and pilot valve causing main valve cperation. Article 9 or the AmE Pressure vessel Code sec+ ion III Ihiclear Vessels requires that these bellows te menitored for failure since this would defeat the safety function of the safet.y/ relief valve. It is realized that there is no vay to repair or replace the bellows during creration and the plant must be shut down to do this. The thirty-day pariod to do this allovs the crerster flexibility to choose his time for shutdown; meenuhile, because of the redundancy present in the design and the continuin6 monitoring.of the integrity of 'the other valves, the overpressure pressure protection has not teen compro:mised. he auto-relief function vould not be impnited by a failure of the bellows. However, the self-actuated overpressure safety function vould be imp 91 red by such a f allure. w r0 P Provision also has been made to detect failure of the bellows monitoring system. Testin6 of this system quarterly provides assurance of bellows integrity. When the setpoint is being bench checked, it is prudent to disassenble one of the safety / relief ~ valves to examine for crud buildup, bendin6 of certain actuator members or other signs of possible deterioration. The program of earety/ relief valve testing confonts to the requirements of 10 CFR 50, rection 50 55a(g). hese requirements are delineated in Specification 4.13 . mis inspection and testing program, combined with the additional surveillance requirements contained in this section, provide a high degree of assurance that the safety / relief valves vill perform as required when needed. l 3.6/4.6 FAfEs 135 Pr.' I

A Awmh eJ.a+..a -si-v -~h 4 -v1_-+-,,eAma-ass ry,, --u--m-a4AL,e---Lb 6,- fa K,s,-2,L 4uAu AL w--m%4 + 4, 1,w-9,--, b:. m W l P ? s i. s$ t t I b f 4 b I B L 44 N t 3 e. T m o UW T w p i. 3 G C T l f3 are +3 Q O A - c to g 60 w - t 1-5-22 In

s - t j Bases Continued 3.6 and h.6: G. Jet pumps Failure of a ' jet pu=p nozzle assembly hold down mechanism, nozzle asse=bly and/or riser, would -increase'the cross-sectional flov area for blevdown following the design basis double-ended line break. 'lherefore, if a failure occurred, repairs must be made. The detection technique is as follows. With the two recirculation pu=ps talanced in speed to within + 5%, the flow rates in both recirculation locps vill be verified.by Control Roca =onitoring instru=ents. Tf the two. flow rate values do not differ by more than 10%, riser and nozzle assembly integrity has been verified. If they do differ by 10% or more, the core flow rate measured by the jet pu=p diffuser differential pressure system must be checked against.the core flow rate derived from the measured values of loop flow to core flov. correlation. If the difference betvaen measured and derived core flow rate is 10% or more (with the derived value higher) diffuser measurements vill be taken to define the location within the vessel of failed jet pu=p norzie (or riser) and the plant shut devn fer repairs. If the potential blevdown flow arco is increased, the system resistance to the recirculation picp is also reduced; hence, u r$) the affected drive pump vill '2mn out' to a substantially higher flow rate (approximately 115% to 120% for a single nozzle failura). If the two loops are balanced in flow at the same pu=p speed, the resistance characteristics cannot have changed. ' Any imbalance between drive loop flow rates vculd be indicated by the plant process -instr centation. In addition, the affected jet pep would provide a leakage path past - the core thus reducing the core flev rate. The reverse flow throuEh the insetive Jet pe p would still be indicated by a positive differential pressure but the_ riet effect vould be a slight decrease (3% to 6%) in the total core flow measured. 'Ihis decrease, together with the loop flov increase, would result in a lack of correlation between measured and derived core flev rate. -Finally, the affected jet pump diffuser ' differential pressure signal vould be reduced because the backflow would be less than the nomal forward flow. A nozzle-riser system failure could also generate the coincident failure of a jet pump body; benever, the converse is not true. The lack of any substantial stress in the jet pung body makes failure impossible - without an initial nozzle-riser system failure. 137 3 6/4.6 BASES FEV s %.m -r- -r m--- .,_m m

n m ~ ~ a t 3.0 LIMITING CONDITIOIG FOR OPD1ATION h.0 SURVEILIRiCE PMUIRDETS 3 Pressure Suppression Chamber - 3 Pressure Suppression Charter - Reactor Building Vacuum Breakers Reactor Building Vacuum Breakers i 3 Except as specified in 3.7.A.3.b a. a. The pressure suppression chamber-reactor below, two pressure suppression building vacuum brerkers and associated in-chamber-reactor building vacuum strumentation including set point shall be i breakers shall be operable at all checked for proper operation every three times when the primary containment months. integrity is required. The set i point of the differential pressure b. Inservice inspection and testing of instrumentation which actuates the components shall be conducted in w pressure suppression cha*ar-reacter accordance with specification 4.13. /u building vacuum breakers chall be

~

o.5 psi. b. From and after the date that one of the pressure suppression chamber-reactor building vacuum breakers is made or found to be inoperable f or any reason, reactor operation is permissible only during the suceed-ing seven days unless such vacuum breaker is sooner made cperable, provided thst the repair procedure does not violate primary cent lintent integrity. Ik6 REV 3.7/4.7

~ ~ .. ~, e p .n s .) 4 3.0 LIMITING CONDITIONS FOR OPFRATION h.O SURVEILLAIICE FIQUIRENS d. The fbel cask or irradiated fuel is not being moved within the reactor building. D. Primary Containment Isolation Valves D. Primary Containment Isolation Valves 1. During reactor power operating conditions, 1. The primary contairment isolation valves all isolation valves listed in Table 3.7.1 surveillance shall de performed as follows: and all primry system instru lent line 4 flow check valves shall be operable except At least once per operatira; cycle the a. vi as specified in 3.7.D.2. operable isolation valves that are power operated and autocatically initiated shall be tested for simulated auto atic initiation and closure times, b. Inservice inspection and testing of components shall be conducted in accordance with Specification 4.13. 3.7/4.7 131 Rsv

j 3.0 LIMITDIG CO!iDITIO*iG FOR OPEPATION k.O SURETIJAICE FMUIREMDCS Y Pn 2. In the event any isolation valve specified 2. Whenever an isolation valve listed in in Table 3.7.1 becomes inope rab le, reac tor operation in the run mode may continue Table 3.7.1 is inoperable, the position or provided at least one valve in each line at least one ibily closed valve in each line having an inoperable valve is closed. having an inoperable valve shall be recorded daily. 3. If Specifica tion 3. 7.D.1 and 3.7.D.2 cannot be met, initiate norr.a1 orderly shutdown and have reactor in the cold shutdown condition within 24 hours. 3.7/4.7 IW av 4

J 30 LIMITIIU C0!;DITIO?C FOR OPERATION h.0 SUFIEILIANCE REQUIREMENI'S 3 13 INSERVICE IfCPECTION A?D TESTIIG h.13 INSEH7 ICE IIiSPECTION AND TESTIIU Applicability: Applicability: Applies to components which are part of Applies to the periodic inspection and the reactor coolant pressure boundary and testing of components which are part of their supports and other safety-related the raaetor coolant pressure boundary pressure vessels, piping, pu=ps, and and their supports and other safety-valves. related pressure vessels, piping, pumps, and valves. Objective: Objective: To assure the integrity of the reactor To verify the integrity of the reactor coolant pressure boundary ruxl the coolant pressure boundary and the w -6 operational readiness of safety-related operational readiness of safety-N pressure vessels, piping, pumps, and related pressure vessels, pipin6, pumps, valves. and valves. t Specification: Soecification: i A. Inservice Inspection A. Inservice Inspection la To be considered operable, Quality

1. Inservice inspection of Quality l

Group A, B, and C components shall Group A, B, and C components shall l satisfy the requiraments contained be perfomed in acconlance with l in Dection XI of the ASME Boiler the requirements for ASME Code Class l and Pressure Vessel Code and appli-1, 2, and 3 components, respectively, cable Mdend t for continued service contained in Section XI of the ASME of ASME Code Class 1, 2, and 3 compo-Boiler and Pressure Vessel (: ode and nents, respectively, except where applicable Addenda as required by relier has been requested from the 10 CFR 50, Section SO.55a(g), except Conmission pursuant to 10 CFR 50, where relief has been requested from Seetion 50 55a(g)(6)(1), the Comission pursuant to 10 CFR 50, Section 50 55a(g)(6)(1), 3 13/h.13 189R REV J l

A 30 LIMITUU 00NDITIONS FOR OPEFATION h.O SURVEIIJANCE IEQUIRD4ENIS B. Inservice Testing of Pu=ps and Valves B. Inservice Testirg of Pms and Valves 1. To be considered operable, Quality

1. Inservice testing of Quality Group Group A, B, and C pu=ps and valves A, B, and C pu=ps and valves shall be l 4 shall satisfy the requirements cou-perfomed in accordance with the o>

tained in Section XI of the AliME ik>iler requi m=ents for ASME Code Class 1, and Pressure Vessel Code and appli-2, and 3 pu=ps and valves, respectively, cable Addenda for operability of contained in Section XI of the ASME ASME Code Class 1, 2, and 3 pumps Boiler and Pressure Vessel Code and and valves, respectively, except applicable Addenda as required by where relief has been requested from 10 CFR 50, Section 50 55a(g), except the Commission pursuant to J O CFR 50, where relief has been requested from Section 50 55a(g)(6)(1). the Co= mission pursuant to 10 CFR 50, Section 50 55a(g)(6)(i). 313/h.13 189s REV

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ML20125B950

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Dear Mr Stello:

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Dear Mr. Duba:

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ML20125B950

Text

Dear Mr Stello:

Subject:

Dear Mr. Duba:

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