ML20024G292

Proposed Tech Specs Re MCPR Limit

ML20024G292
Person / Time
Site:	Monticello
Issue date:	07/29/1977
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G287	List: ML20024G286 ML20024G289 ML20024G292
References
NUDOCS 9102080493
Download: ML20024G292 (3)

Text

- -- -

y  : y ~

4 .

t -

9 3.0 I,IMITING CONDITIONS FOR OPERATION 4.0 SURVERLANCE REOUIRDENTS .

y n, (iv) The rod block function of the O- rod worth minimizer shall be

• O r; verified by attempt! ut to with-draw an out-of-sequence control

>@ rod beyond the block point.

co Os 04 7 tJ (b) Whenever the reactor is in the (b) If the rod worth minimizer is inoperable ou startup or run mode below 107 while the reactor is in the startup or y rated thermal power, no centrol run mode below 10I rated thermal power g og rods shall be moved unless the and the second independent operator or engineer is being used, he shall c ra a rod worth minimizer is operable

"$ or a accond independent operator or egineer verifies that the verify that all rod positions are correct prior to commencing withdrmaal operator at the reactor console or insertion of each rod group.

is following the control rod program. The second operator may be used as a substitute for an inoperable rod worth minimizer during a startup only if the rod i vorth minimizer fails af ter with-drawal of at least twelve control '

rods. i

4. Control rods shall not be withdrawn 4 Prior to control rod withdrawal for for startup or refueling unless at startup or during refueling verify i least two source range channels have an observed ccunt rate equal to or that at least two source range channels have an observed count rate h

greater than three counts per second, of at least three counts per second.

5. Whenever the Engir.eer, Nuclear, deter- 5. Whenever the Engineer, Nuclear, deter-mines that a limiting control rod mines that a limiting control rod pattern pattern exists, withdrawal of desig- exists, an instrtment functional test nated control rods shall be permitted I of the RBM shall be performed prior to l only when the REM system is operable. withdrawal of the designated rod (s) and daily thereafter.

I i

3,3/4.3 7g REV i

.n - - - - _ -_ -_ -- .__-- -- - . - - - -w- . - - -- -- r 3.0 LIMITING CONDITIONS FOR OPERATI(N 4.0 SuRVEILIANCE REQUIRBENTS C. Minimum Critical Power Ratio (MCPR) C. Ministan Critf cal Power Ratio (MCPR)

~

During power operation, MCPR shall be determined daily dur-the Operating EPR Limit shall be ing reactor power operation at 2t257.

l ,

> 1.46 for 8x8 fuel and 21.37 for rated thermal powse and following 7x7 fuel at rated power and flow. any change in power level or distri-If at any time during operation it is detemined that the limiting bution which has the potential cf bringing the core to its operating h

value for EPR is being exceeded, E PR limit.

action shall be initiated within 15 minutes to restore operation to with-in the prescribed limits. Surveil-lance and corresponding tetion shall continue until reactor operation is within the prescribed limits. If the steady state EPR is not returned to within the prescribed limita with-in two (2) hours, the reactor shall ha brought to the Cold Shutdown con-dition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. For core flows other than rated the Operating EPR Limit shall be the above applicable MCPR value times Kr where Kr is as shown in Figure 3.11.3.

189D 3.11/4.11 RW

_ _ __ _ o_ ----

.. ;

• g a Bases 3.11 (continued) .I.,

C. Minimum Critical Power Ratio ('.fCPR) ., .

The ECCS evaluation presented in Reference 4 assumed the steady state MCPR prior to the postulated loss-of-coolant accident to be 1.18 for all fuel types. In addition, the ECCS analysis presented in Reference 6 assumed an initial MCPR of 1.24 for reduced flow con-I ditions. The Operating MCPR Limit of 1.46 for 8x8 fuel and 1.37 for 7x7 fuel is determined from the analysis of transients discussed in Bases Sections 2.1 and 2.3. By maintaining an operating MCPR above these limits, the Safety Limit of.1.06 (T.S.2.1.A) applicable to all fuel types is maintained in the event of the most limiting abnormal operational transient.

For operation with less than rated core flow the Operating MCPR_ Limit is adjusted by multiplying the above limit by Kg. Reference 5 discusses how the transient analysis h done at rated conditions encompasses the reduced flow situation when the proper Kg factor ,

is applied.

Those abnormal operational transients, analyzed in FSAR Section 14.5, which result in an automatic reactor scram are not considered a violation of the LCO. Exceeding MCPR limits in such cases need not be reported.

Keferences

1. " Fuel Densification Ef fects in General Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEDM-10735, August, 1973.

2. Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 14,1974 (USAEC Regulatory Staff)

3. Communication: VANbore to IS Mitchell, " Modified CE Model for Fuel Densification,"

Docket 50-321, March 27, 1974.

4. "Monticello Nuclear Generating Plant Loss-Of-Coolant Accident Analysis Conformance with 10 CFR 50 Appendix K, August 1974," L 0 Mayer (NSP) to J F O' Leary. August 20. 1974

5. " General Electric BWR Generic Reload Application for 8 x 8 Fuel." NEDO-20360, Revision

1. November 1974 l

l 6. " Additional Effects of Core Flow on ECCS LOCA Analysis," A. Levine (CE) to Z. Rosztoczy j (NRC) August 24, 1976.

t 3.11 BASES 189F P1V '

7. T O u.s. Nucle An maovcAtony c'okwesioN n stNu : a NRC DISTRIBUTION Pon PART 50 DOCKET MATERIAL TO: F ROM: DAte oP pocuusNr Mr. Victor Stello N rthern States Power company 7/29/77 Minneapolis, Mn oats m6ceivso f..O. Mnver R /1/77 NTTs m TNotomi2:0 P R oP INPUT PomW NVWssM OP COPiss meCalvso o o*NAL QuNc LAssiPis o 4y Q .

oescairTsow sNcLosvan License No. DPR-22 Application for Amendt' Tech specs proposed change concerning correction of an editorial error E seeks d(# t a more restrictive MCPR limit.

w;f:~ ;-nn q, Hun..LaLLu n u.

j .

JO 10~ REM 0E

,, PLANT NAME:

i Monticello Nuclear Plant Generating Plant enclosures: 13p, 13p, 13p.

VT 8/P/77 (1-P)

. RAFETy FOR ACTION /INFORMATION ENVIPENMENTAL

, ASSICNED ADt ASSIGNED ADt V. HOORE (LTR)

/ RRANCH CHIEF: (7/ Dd V/5 '

BRANCH CHIEFt PROJECT MANAGERt PROJECT MANAGER:

, LICENSING ASSISTANTt LICENSING ASSISTANT:

B. MAELESS

, INTERNAL DISTRIBUTION JMme FTt ?9 3 SYS72NS SAFETY PLANT SYSTEMS SITE SAFETY &

C Muc rux HEINt: MAN TEDESCO ENVIRON ANALYSIS

/

/ T AF nFt fi

/1) SCHROEDER BENAROYA fATNAM DEN'IVN - & MULLER c?f fTCHFT Ft.n

/ COSSICK & STAFF ENGINEERING IPPOLITO

/ MAMiffED MNTCMT F. ROSA ENVIRO TECH.

vTpc BOSNAM ERNST J cA<F . -W SIHWELL OPERATING REACTORS BALLARD RtWn PAWLICMT / STEtLO YOUNGPLOOD 1- / FT9ENMtfT

! 990 TFCT WANACUTNT REACTOR SAFEIY / S~r8.A0 o*Svnnt.T ROSS / EAER P. COf.1TN9 NOVAK / BUTLER GAMMILL (2)

MottsmN RO S Z'IUCZY # CRIES

. Mrt T2 / CHECK SITE ANALYSIS E- MrtTEMES VOLLMER tr AT&I / BUNCH .

SALTZMAN J. COLLINS RUTBERG _

KREGER

EXTE RN AL DISTT.180 TION CONTROL NUMBE R j '

LPDRs_ Elinnea' DO!Ji' .Hf)I ,

[

/ TIC NSIC ,

'~

l NAT MB ~

f REC TV (J. HANCHETT) 77215c>222 ./

[ / 16 CYS ACRS SENT CATEGORY 5

, N AC P QaM 196 f2 73) w-y'I- pm- 1-y4 -g,e-.- .s-.-i 9 ry- goy 7 y- - "- **Mf' w 41p"-vf"-'-~-W----'**w ' =*- *- ' - - ++ F**