ML20127G566

Responds to 730601 Request for TS Change Re Relief Valves & Contol Rod Scram Times.Ts Changes Encl

ML20127G566
Person / Time
Site:	Monticello
Issue date:	07/02/1973
From:	Skovholt D US ATOMIC ENERGY COMMISSION (AEC)
To:	Mayer L NORTHERN STATES POWER CO.
References
NUDOCS 9211170380
Download: ML20127G566 (15)

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Docket No. 50-263 4

Northern States Faser Company i

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'Mr. L. O. Meyer Director of Nealear Segport Services l

414 Nicollet M111 Wi

- ;:11a M4 - ta 55401 Gentlement License No. DPR-22 We have revisued your request (NSP letter dated June 1,1973) to change I

the Monticello Techaiaal specification requirements for relief valves 4

and contre 1~ tod scram tians. Based on the repet " Result of Transient Esanalysis for Monticello musisar Generating Flaat with End-of-Cycle 1 Core Dynands Characteristics" provided to the Directorsta of Lisenstag by NgP latter dated February 13, 1973, and the statannat in your June 1, 1973 letter that pre 14=i==ry calculations show that the End-of-Cycle 1 analysis passants the most limiting conditions espseted during the first 2250 mfd /STU saposers increment of cycle 2, y% have proposed that the required number of relief valves be incrassed from 3 to 4 and that the control rod scram time be changed to require fr. ster response and lasertian tiens over the first 2 seconds of the 5-secoed scram time interval. With primary system pressure relieved through t. relief valves 4==tand of 3 and the faster control rod scrs= laitiation 3 a have l

requestad, the maximum core coolant pressure according to the IOC 1 analysis follouring an assumed turbine trip and failure of the setonatic turbine bypass valve to open will remain at least 25 psi below the lowest of Le's 4 sciety valve set points, thereby satisfying a General Electric requiremer.t that the pressure margin to the sa'lety valve set point be 25 psi-,

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l We further, understand from telephone eenversations with MSF representatives that analyste.wt.11 acatisme for the period beyond

'1150 WD/sTU to the end of'eyela 2'and eyela 3 er until'the equilibrine fasi cycle. conditions are attained and that additional a Teekataal,spesifisation. changes may be messesarr'for the period 7' beyeed.1150 WD/8T0 to maintain relief and safety _ valve core.

emelsat system everpressure safety margins. Te assure that we 4

w have sMficient time to seaplete our review of your. analysis for

.operatf.02 beyond as incremental fuel exposure of 1150 WD/gTU

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during eycle 2, it is requested that your analysia.and.any

' -. proposed changes in technical specifications be submitted at least 30 days prior to the date at which ma exposure Amerement during eyele 2 of 2250 WD/5W would be achieved.

We are continuing our evaluation of the shape abanges in the scram

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resetivity surve and the aseessity for more restrictive technical specifications but agree that the Technical specifications changes

~t.you have proposed should be made now. All four relief ~ valves must be in servise where in the past is was possible to operata the 3

. reactor with one of the four Islief valves out of service.. Analysis

. of the overpressure transient following turbine trip without bypass

',,'will be bened on faster Technical specifications control rod meram

" "' ' ' times than'previously required but conservatively slower than actual

..maasured control rod scram tiana.

' We have concluded that the changes you have proposed by your June 1 1973 latter as modified by your letter of June 20, 1973, are required u-(to maintain acceptable ecre coolant system overpreeeurs safety margins 1

and do not present significant hasards emesiderations, and that there is reamonable assurance that the health and safety of:the publis will Toot be endangered by operation of the reester Asi the meaner propeeed 9

i to' an amposure increment in the second fuel cycis of 2150 WD/gTU.

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. Accordingly, pursuant to section 50.59 of 10 CFR Part 50, the p(

, Technical Specifications of Facility Operating License No. DPR-22 are

' hereby changed to replace existing pages 22, 24, 25, 26, 39, 79, 85, 45a, 36,112,119, and 134 with revised pages bearing the'same numbers.

The revised pages also include minor typographical corrections, p T.

. Sincerely,

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Eaclosuress Revised pages as stated above es w/manla=uress bec:

Donald 5. Nelson, Esquire Docket File

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Northera States Power Company Branch Reading 414 Niaallat Hall RP Reading

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Minneapolis, pt===ta 55401 JRBuchanan, ORNL TWLaughlin, DTIE Gerald Charnoff DJSkovholt, L:0R Shaw, Pittman, Potts, Trowbridge & Hadden TJCarter, L:0R 910 - 17th Street, N. W.

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Finnaapolia, Minnesota

$5415 NDube, L:0PS Steve Gadler, P. E.

2120 Carter Avenue St. Paul, Minnesota 55108 Harriett Lansing, Esquire Assistant City Attorney City of St. Paul 638 City Hall St. Paul, Finnamota

$5102 Kan Dzugan Minnesota Pollution Control Agency 717 Delaware Street, S. E.

u Minneapolis, Minnesota 55440 Environ =ntal Library of Minnesota 1222 S. E. 4th Street Minneapolis, Minnesota 55414 l

cc w/ enclosure and cy of NSP ltra dtd 2/13/73, 6/1/J) art.6/20/73:

d Warren R. Laween, M. D.

Secretary & Eancutive Officar State Department of Health 717 Delaware street, S. E.

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Bases:

2.2 The reactor coolan system integrity is an important barrier in the prevention of uncontrolled release of fission products.

It is essential that the integacy of this system he protected by establishing a pressure limit to be observed for all operat ing conditions and-whenever there is irradiated fuel in the reactor vessel.

The pressure safety limit of 1335 psig as measured in the vessel steam space is equivalent to 1375 psig at the lowest elevation of. the reactor coolant system.

The 1375=psig value was derived from the design pressures of the: reactor pressure vessel, coolant piping, and recirculation pump casing.

The respective design pressures are 1250 psig at 575*F,1148 psig at 562"F, and 1400 psig at 575*F.

The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes: ASME Boiler and Pressure Vessel. Code Section III-A for the pressure vessel, 1

ASME Boiler and Pressure Vessel Code Section III-C for the recirculation pump casing, and the USAS Piping Code Section B31.1 for the reactor coolant system piping. The ASME Code permits pressure transients up to 10 percent over the vessel design pressure -(110% x 1250 = 1375 psig) and'the USAS Code permits pressure transients up to 20 percent over t.he' piping design pressure (120% x.1148 =

1378 psig).

The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety. pressure limit of 1375 psig. The vessel has been denigned for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig and temper-ature of 575*F;. this is more than a factor of 1.5 below the yield strength of 42,300 psi at this temperature. At the pressure limit of 1375 psig, the general membrane stress increases to 29,400 '

psi, still safely below the yield strength.

The reactor coolant system piping provides a comparable margin of protection at the established pre.ssure safety limit.

The normal operating pressure of the reactor coolant system is approximately 1025 psig. The turbine -

trip from rated power with failure of the bypass system represents the most severe primary system.

pressure increase resulting from an abnormal operational ransient. The peak pressure in'this transient is 1183 psig. The safety valves are sized assuming no direct scram during 1

e 2.2 BASES 24 rn w

Bases:

2.2 1sIV closure. The only scram assuced is from an indirect means (high flux) and the pressure at the bottom of the vessel is limited to 1283 psig in this case. Reactor pressure is continuously monitored in the control room during operation on a 1500 psig full scale pressure recorder.

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2.2 BASES 25 e

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,lpe break in the event of HPCI failure, by depressurising the reactor vessel rapidly enough to actuate the core i

sprays or LPCI. Either of the two core spray systems or ITCI provide sufficient flow of coolant to limit fuel clad temperatures to well below clad melt and to assure that core ccometry remains' intact.

Three of the four relier/ safety valves are included in the automatic pressure relief system.

Cf these three, only two are required to pravide sufficient capacity for the automatic pressure relief system.

See Section 4.4'and 6.2.5.3 FSAR.

F.

BCIC The RCIC system is girovided. to supply c,.nt.inuous nnkeup w. ster to the reactor core w; ten the react.or is isolated frca the turbine and when the feedwtter r.yst. cat is not available The pumping capacity of the RCIC system is sufficient to r.nintain the w st.cr level above the core without any other. water system in operation.

If tne water level in the reactor vessel decreases to the RCIC initiation level, the system autcmatically starts.

'Ihe system may also be annually initiated at any time.

The HPCI system provides an alternate method.cf supplying makeup water to the reactor should the normal feedwater become unavailable.

Therefore, the specification calls for an opera-bility check of the HPCI system should the RCIC system be foural to be inoperable.

e 3.5 EASES 112

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l 3.0 LIMITInc co:IDITIOIra FOR CPERATICH 4.0. CU:tVEILIA!!CE IlEQUUIDfEITIG tien of four safety / relief valves chall every two refueling outages.- The ncminal be operable. The solenoid activated popping point of the safety valves shall relief functicn of the safety / relief val-

'be set as follows:

ves shall be operable as required by Spec-thunber of Valves set Point -(psisd.

ification 3 5.E.

__1210 2

2

< 1220 1

2.

If specification 3.6.E.1 is not ::ct, ini-2; a.

A minin r:m of two carety/relier valves shall tiate 'an orderly shutdown and have coolant be Lench checked or trplteed vith n bench prescure and teiperature reduced to 110 checked valve each yerueling' outage. All-i psig or less and 345 F or less within 24 four valves shall be. checked or replaced hours.

every two refueling outages. The popping point of the cafety/relier valves shall be set as follows:

Thimber of Valven Set Point (psig).

'4

_< 1080 b.

At leact one of the cafety/ relief valves-shall be disassembled and inspected each refueling outage.

The integrity of the cafety/ relief valve-c.

bellows shall be continuously monitored.

d.

The operability of the bellows monitoring 3.6/4.6 119

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