ML20024G415

Application for Amend to License DPR-22,revising Tech Specs by Raising Permissible Setpoint of Eight Safety/Relief Valves Installed at Plant to 1,108 Psig

ML20024G415
Person / Time
Site:	Monticello
Issue date:	08/16/1978
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G412	List: ML20024G411 ML20024G415 ML20024G649 ML20024G650
References
NUDOCS 9102120411
Download: ML20024G415 (5)

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!";I fEli STATES NUClIAP RECUL1 Top,Y CD'C11 SS10.';

NORTHERN STATES POWER COWANY It0NTICELLO NUCLEAR GENERATING PLANT Dociet No. 50-263 REQUEST FOR A>1ENIt1EITT TO OPERAllNG LICENSE NO, DPR-22 (License Amendment Request Dated August 16, 1978)

';orthern States iwer Company, a flinnesota corporation, requests authoriz, tion for changes to the Technical Specifications I

as shown on the attachments labeled Exhibit A, Exhibit n, and Exhibit C. D:hibit A describes the proposed changes along with reasons for the chatq ,. D:hibit B is a set of Technical Speci fication pages incorpora t its th . preposed changes. Exhibit C is a sa fety cvaluati o .n artita; the changct T h in recuect contait < no rest r icted or other defense in f o rtna t u s NORTHEPd STATES POWER CG!PANY bi ' " L'J GAL 5)

Wachter ~

l Vice President, Power Production &

3 System Operation l On this 16th day of August , 1978 , before l

ne a notary public in and for said County, persona lly appea red L J Wacht er, Vice President, Pouer Production & System Operation, and first being duly sworn acknowledged that he is authorized to execute this document in be-half of Northern States Power Company, tha t he knows the contents thereof

and that to the best of his knowledge, infennation and belief, the state-l ments nade in it are true and that it is not intcrpcsed for dela

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EXHIBIT A MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-2 2 LICENSE AMENDMENT REQUEST DATED August 16, 1976 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS Pursuant to 10CFR50, the holders of Provisional Operating License DPR-2 2 hereby propose the following changes to the Appendix A Technical Specifications.

PROPOSED CHidGES

1. Increase the allowable setpoint for the safety / relief valves from 1080 psig to 1108 psig.

2. Eevise the minir:u" Ope rat ing MCPR Limit for both 8x8 and 8xFP tuel to 1.33.

3. Revise the F>ases to be consistent with the changes proposed in

< (2) atave.

REASON FOR CHANGES Fcr a number of years Northern States Power Company has been directly involved in inproving the ieliability of the eight safety / relief valves ins talled a t the Monticello Nuclear Generating Plant. A number of equipment modifications and procedural changes have been nade to greatly reduce the probability that a valve will fail to open .inen required. One cont inuing problem, howeve r, i s the t ende ncy for these valves to leas excessively during normal operation. At othe r f acili ties, spuricus opening or f ailure to reclose have been pr ob lem s . LTnile excessise leakage or spurious opening is not a se riou s safety problem, it does reduce availability by requiring the plant to go to cold shutdown for safety / relief valve repla ceme nt or repaird.

Ge nt r ,. . a c e t r i :. ne ce t e rmined tnat leakage and the probabili ty c.

a spont aneous valve opening or f ailure to reclose is strongly influenced by the saf ety / relief valve simme r ma rgin. Simmer ma rgin is defined as the pressure dif fe re nce between the valve setpoint and the normal system operating pressure. General Electric recommends inc reasing valve sicser mcrgin to the maximum permitted by safety ar n!' v e . kle propm e ir this License Amendment Req ues t to inc reas t the simmer cargin by 2e psi and believe that this increase will si.nific nt!- impreve the pe rf o rma nce of our valves.

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EXHlBII A-

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1 Ue are able to increase safety / relief valve setpoint because of the la rge amount of excess relief capacity ins talled at Monticello.

Monticello was originally designed with four safety valves discharg-ing directly into the dryvell atmosphere and four safety / relief valves with exhaust piping to the suppression pool. In 19 74, all safety valves were replaced with safety / relief valves of the same type as those originally installed. This significantly increased the ins talled relief capacity in two ways (references 1 and 2).

Firs t, each safety / relief valve provided a greater flow rate than the spring saf ety valve it replaced. Second, because the setpoints 01 the safety / relief valves are subs tantially lower than safety valve setpoints, the modified system provided an earlier negative void reactivity feedback which aided in reducing transient pressure during limiting pressuri.zation events. The NRC Staff gave credi t for the valve capacity but required additional time to

-evaluate analytical models before allowing credit for the latter phe nmx :a - .

An interit Technical Specification was issued (page of the safety evaluation at tached to ref erence 2) which required seve n 'saf e tv / relief valves to be operable even though the transie analyses showed that only six valves were necessary. The practice ove r the four intervening years has been to license similar BUR's using the same analytical models for the number of operable safety /

reli ef valves assumed in the transient analyses. The Monticello Technical Specifications were neve r revised to remove the interim cons e rva tist imposed by the Staff. Rather than seek a reduction from seven to six operable safety / relief valves at this time, we pref er to take credit for the seventh safety / relief valve which permits a 2E psi increase in the setpoints of all valves while maintaining an acceptable transient peak vessel pressure.

Reload 6 is scheduled for the 19 78 Autumn ref ueling outage.- In additian to the sa f ety evaluation based on the cur rently authorized sarety/ relief valve setpoint of 1080 psig (reference 4), ad di ti ont.1 transient analyses have been perf ormed to justify an increase in sett >1nt to 1108 psig. The results of this analysis are presented in Exhibi t C.

As noted in Exhibit C, in the analysis of the turbine trip without bypass, it was found that the change in critical power ratio caused by thc increased safety / relief valve setpoint is ins ignif icant (0.002 delta CPR) . This change af fects the roundof f to two signifi-ca nt decimal places (the conventional roundof f adopted), therefore t he MCPR Ope rati ng Li It is increased by 0.01 ove r the limit reportec in reference (4).

EXillB1T A SAFETY EVALUATION The safety evaluation for _ Reload 6 was submitted for NRC review on August 10, '19 78 -(ref erence 4). Increasee.in'the maximum allowable safety / relief valve setpoint only af fect those events which result

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in valve scif-actuation. _The limiting events which have been reanalyzed are the most severe pressurization transient (turbine trip with f ailure of the bypass valve), vessel overpressure protec-tion analysis (closure of all main steam isolation valves with indirect sc ram f rom high neu tron flux), and the loss-of-coolant accident (sma11' break). In addition, the capability of the RCIC and HPCI systems were evaluated for ' the higher safety / relief valve setpoints. Refer to Exhibit C for the results of these analyses. A saa e ty / relief valve setpoint of 1108 psig f or all eight valves is clearly acceptable.

A stress analysis of. all four main steam lines and all eight safety / 2 relief valve discharge lines was completed and submitted to the NRC for review._when additional safety / relief valves were added (reference 1).

Incr easing ' the valve setpoint f rom 1080 to_1108 psig will result in a stect flow increase at setpoint pressure of less-than three pe rc ent . _Conserva tive assumptions we re used in de riving the transient.

loads'for the stress analysis reported in reference (1) making it valid f or the increased valve setpointr.. We will re-evaluate the torus discharge piping (including newly installed T-quenchers) for the 1106 psig setpoint using recent Monticello T-quencher test data.

All safety / relief valve discharge lines, main steam piping loads, and T-quenchers will be . re-evaluated for the Mark 1 Containment Long Term Program loads and for the ef fects of the increased setpoints.

This analysis will be initiated later this year when the discharge line loads. model is available from General . Electric. This reanalyris will allow for f urther increa.ces in saf ety/ relief valve simme r margin which may be justified in the future.

References

1. Le t te r f r om L 0 Maye r , N SP , t o J F O' Leary, US AEC, dated January 23, 1974, "Pe rma nen t Plant Changes to Accommac de Equilibrium Core bcram Reactivity Characteristics," with . rata dated March 19, 1974,

2. Le tte r f r om E R Goller, US AEC, to L 0 Pbyer, NSP, dated May 14, 19 74, Amendment No. 3 tc DPk-22.

3. 14 t te r f rom L 0 Maye r , NSP , to Direc tor, NRR, USNRC " License Amendment heq ues t . da ted March 21, 19 7E . "

4. Lc t te r f rom L 0 Maye r, NSP , to Director, NRR, USNRC, dated August 10, 1976, " Supplement No. I to License Amendment Request March 21, 1978."

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EXHIBIT B LICENSE AMENDMENT REQUEST DATED AUGUST 16, 197S This - exhibit consis ts _ of the following pages revised to incorporate all of - the proposed Technical Specification changes:

25 25

'26 119 n 134

, 189D*

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Revisions proposed in Supplement No. I t o . License knendment Reques t dated March 21,'1976 are also shown on these pages.- This supplement was subtitted for I;RC review on August 10, 1978.

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