ML20024G227

Application for Amend to License DPR-22,changing Tech Specs, Involving Main Steam Line Low Pressure Setpoint & MCPR Limit

ML20024G227
Person / Time
Site:	Monticello
Issue date:	12/01/1975
From:	Wachter L NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G225	List: ML20024G224 ML20024G227 ML20024G229
References
NUDOCS 9102080370
Download: ML20024G227 (5)

Text

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o O Regulatory Docket Filg UNITED STATES NUCLEAR REGULATORY COMMISSION m tivintf Ad$AM.w ... -

NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT l

Docket No. 50 263

• k REQUEST FOR AMENIMDiT TO OPERATING LICENSE NO. DPR- 22

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(License Amendment Request Dated December 1,1975)

Northern States Power Company, a Minnesota corporation, requests authorization for changes to the Technical Specificatio'is as shown on the attachments labeled Exhibit A and Exhibit B. Exhibit A describes the proposed changes along with reasons for the change. Exhibit B is a set of Technical Specification pages incorporating the proposed changes.

This request contains no restricted or other defense information.

NORTHERN STATES POWER CCEPANY By C/ fu/$

1. L J Wachter Vice President, Power Production &

System Operation On this 1st day of Dece=ber , 1975 , before ce a notary public in and for said County, personally appeareo L J Wachter, Vice President, Power Production & System Operation, and first being duly sworn acknowledged that he is authorized to execute this document in behalf of Northern States Power Company, that he knows the contents thereof and that to the best of his knowledge, information and belief, the statements c.ade in it are true and that it is not interposed for delay.

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, DENiSE E. BRANAU '

h0TARY PUbuC- NINMDCTA :

e HtNNtPIN COUNTY  :

bry tomardssion Dpires Oct. IC,1981 ;

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910?O00370 701201 1 PDR ADOCK 05000263 l P PDR '

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Fgulatory Docket File CalIBIT A MONTICELLO NUCLEAR GENERA'i1NG PIANT DOCKET NO. 50-263 .

LICENSE AMENDMENT REQUEST DATED DECDiBER 1,19hg gl{*1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS APPENDIX A 0F PROVISIONAL OPERATING LICENSE NO. DPR-22 Pursuant to 10 CFR 50.59 the holders of Provisional Operating License DPR-22

) hereby propose the following changes to Appendix A, Technical Specifications.

1. MAIN STEAM LINE LOW PRESSURE SET POINT PROPOSED CHANGES 4

On pag *e 50, change the trip setting in item 1.e of Table 3.2.1 from "850" to'825 psig.

l On page 66, Bases, change the two values in the fourth paragraph from "850 psig" to "825 psig " In the same paragraph replace the words,

" Reference Section 14.5.4.1 FSAR." with " Reference License Amendment Re-quest Dated December 1,1975 from L. O. Mayer (NSP) to R. S. Boyd (USNRC)."

REASON FOR CHANGES Ihe main steamline low pressure isolation setting of 850 psig was chosen for initial plant design and analysis on the basis that it allowed suf-ficient margin from operating pressure, Experience has shown that it is desirable to reduce the setpoint. To avoid Technical Specification vio-lations the pressure switches must be set to trip well above the speciaed limit to provide sufficient margin for instrument drif t and normal variation.

By setting the switches to trip at too high a pressure, any pressure fluct-uation in hydraulic sensing lines or main steamlines may result in a spurious isolation and scran.. Since the safety evaluation shws that adequate pro-tection ic provided at a lower trip setting, it is prudent to make such a change.

SAFETY EVALUATION The FSAR discusses events which might potentially result in a decrease of reactor coolant inventory. Failure of the inital pressure regulator is one such event analyzed. In the most severe failure the turbine controller will call for 1107, of design flow which commences a reactor pressure re-duction. The depressurization causes a corresponding increase in the mod-erator voiding fraction which reduces reactor power. The depressurization continues until the main steamline low pressure trip causes a reactor isolation and scram. Reactor pressure increases after isolation due to decay heat

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. . generation. Pressure is then controlled by relief valve operation, if '

nece s sa ry. The transient is very mild and is bounded by more severe operational transients that are routinely analysad.  ;

The analysis reported in FSAR Section 14.5.4.1 is based on a 100 psi  : !

margin between the isolation setpoint and the turbine inlet pressure, /"

corresponding to an isolation setpoint of 850 peig. However, the mild progression of this transient can be used as a basis for determining the effects of a range of isolation settings. Iowering of the Technical

, Specification limit to 825 psig will not invalidate the transient safety analysis reported in the FSAR and will result in a negligible added re-quirement in terms of fuel duty and vessel cooldown. Werefore, lowering of the existing isolation setpoint as described above will not degrade '

the degree of protection of fered by this safety system.

2. MCPR LIMITS PROPOSED CHANCES On page 189D, TS 3.11.C and on page 189F, first paragraph of Bases 3.11.C, change the operating MCPR limits from "1.41" to "1.38" and from "1.33" to "1.29" for 8 x 8 and 7 x 7 fuel respectively.

REASON FOR CHANGES ne present limits were derived from a transient analysis based on Cycle 4

, operation. An analysis based on Cycle 5 operation shows these limits to be bounding but overly conservative. W e proposed changes will provide additional margin to operating limits, thereby giving more operating flexi-y bility.

SAFETY EVALUATION The opetating MCPR limit is derived by adding tc the MCPR safety limit the maximum delta CPR of the abnocmal opetational transients analyzed.

For Cycle 5 it has been determined that the turbine trip with failure of the bypass valves .to open is the most limiting abnormal operating transient.

We delta CPR values were calculated to be 0.23 and 0.32 for 7 x 7 and 8 x 8 fuel respectively; the Technical Specificattun MCPR safety limit re-mains 1.06, resulting in operating MCPR limite of 1,29 and 1,38 for 7 x 7 s and 8 x 8 fuel respectively. The reactor transients and the thermal hydraulic responses were analyzed as described in the topical report, "GE/BWR Generic Reload Licensing Application for 8 x 8 Fuel, Revision 1, Suppicment 2, Msy 1975, NED0-20360." n e input and output parameters are presented in Tables 1, 2 and 3 and in Figure 1.

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Table 1 CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS i

,. 7x7 _8x8 Peaking factors (local, radial 1.24/1.46/1.40 1.24/1.53/1.40

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7, and axial)

R-factor 1.100 1.102

,' Bundle power f 4.932 MWt 5.165 l

Non-fuel power fraction 0.035 0.035 Core flow Bundle flow 57.6 M1bfhr 1.077x10 lb/hr 57.6M1bfhr 1.061x10 lb/hr Reactor pressure 1030 psia 1030 psia Inlet enthalpy 523.2 Btu /lbm 523.2 Btu /lbm q Initial MCPR 1.29 1.41 Table 2 i TRANSIENT INPUT PARAMETERS 1

Ihermal Power (MWt) 1670 100%

Rated Steam Flow (Ib/hr) 6.78x10 6 1007.

Rated Core Flow (1b/hr) 57.6x10 6 1007.

Dome Pressure psig 1025 Turbine Pressure psig 980 S/RV Set Point peig 1080 + 1%

I S/RV Capacity (at Set Point) #/% 6/68.76

'l S/RV Tilne Delay (msec) 400 S/RV Stroke Tizne (msec) 100 Void Coefficient (c/7.Rg) -8.688 Void Fraction (%) 36 Doppler Coefficient (C/0F) -0.208

] Average Fuel Temperature (01) 1153

] '

. Scram Reactivity Curve EOC-5 Scram Worth ($) 33.024 Table 3 TRANSIENT DATA RESULTS

' Coro A A A A Power Flow d Q/A Psi Pv ACPR Transien t (7.) (%) (%) Q (psig) (psig), (8x8/7x7) i.

l8 Torbine Trip w/o Bypass 100 100 390 118 1196 1221 0.32/0,23 l

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1. NEUTRON FLUX 1.?} VESSEL PRES' RISE (PSI)

, 2. PEAK FUEL CENTER TEMP _ 2.

• SM LINE PRES RISE (PSI) .

3. AVG. SURFACE HEAT FLUX 3. ~ TURBINE PRESS RISE (PSI)

4. FEEDWATER FLOW _4 CORE INLET SUB (BTU /LB) am 5. VESSEL STEAM FLOW

5. CORE AVG VOID FRACT (%)

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6. TURBINE STEAM FLOW (%)

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3. N.R. SENSED Lt. VEL (INCHES) 2, SURFACE HEAT FLUX isa 3.

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CORE INLET FLOW (%)

DRIVE FLOW 1 ') 4.

FLOW VARIABLE SCPJef ESTIMATED IIIEIciAL POWER

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