ML20210V558

Proposed Tech Specs Reflecting Decrease of Two Recirculation Loop SLMCPR Limit to 1.11 & Single Recirculation Loop SLMCPR Limit to 1.12

ML20210V558
Person / Time
Site:	Grand Gulf
Issue date:	09/18/1997
From:	ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20046D790	List: ML20210V546 ML20210V558
References
GNRO-97-00087, GNRO-97-87, NUDOCS 9709240044
Download: ML20210V558 (6)

Text

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1 Attachment 3 to GNRO 97/00087 i i

i GGNS PCOL-97/003 Mark-up Pages of the affected Technical Specifications Technical Specification Bases I

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P PDR

SLs 200 i 2 0 5AFETY LIMITS JSLs) i 1

2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam done pressure < 785 psig or core flow < 10% rated core flow:

THERMAL POWER shall be s 25% RTP. l

• 2.1.1.2 With the reactor steam dome pressure m 785 psig and core I  !

flow a 10% rated core flow:

1. L MCPR shall be JE for two recirculation loop operation or t l.lf)for single recirculation loop operation. 1.12.

2.1.1.3 Reactor vessel water level shall be greater than the top  :

of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reacto.* steam dome pressure shall be s 1325 psig.

2.2 SL Violations

• With any SL violation, the following actions shall be completed: .

2.2.1 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance with 10 CFR 50.72.

2.2.2 Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.2.1 Restore compliance with all SLs; and 2.2.t.2 Insert all insertable control rods.

2.2.3 Wit'nin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the plant manager and the corporate executive responsitle for overall plant nuclear safety.

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• MCPR values in T.S. 2.1.1.2 are applicable only for cycidE[) operation, i (continued)

GRAND GULF 2.0-1 Amendment No. 430, 131

Reporting Requirements 5.6 5.6 .Repcrting Requirements i

5.6.5 CoreOperatinalimitsReport(COLR)(continued)

. 10.XN-NF-85-74(P)(A),"R00EX2A(BWR): Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA.

)

11.XN-CC-33(P)(A),"HUXY: A Generalized Multirod Heatup Code l with 10CFR50 Appendix K Heatup Option," Exxon Nuclear Company, Inc., Richland, WA.

12. XN-NF-825(P)(A), "BWR/6 Generic Rod Withdrawal Error for Plant Operation Within the Extended Analysis, Operating Doma MCPR,in," Exxon Nuclear Company, Inc., Richland WA.

13. an XN-NF-81-51(lea Exxon Nu: (r Company BWR Jet Pump Fuel Assembly Exxon Nuclear Company, Inc., Richland, WA.

14. XN-NF-84-97(P)(A), "LOCA Seismic Structural Response of an ENC 9x9 BWR Jet Pump Fuel Assembly," Advanced Nuclear fuels Corporation, Richland, WA.

15. XN-NF-86-37(P), " Generic LOCA Break Spectrum Analysis for BWR/6 Plants," Exxon Nuclear Company, Inc., Richland, WA.

16. XN-NF-82-07(P)(A), " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Inc.,

Richland, WA.

17. XN-NF-80-19(A), Volumes 2, 2A, 2B, & 2C, " Exxon Nuclear Methodology for Boiling Water Reactors EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA.

18. XN-NF-79-59(P)(A), " Methodology for Calculation for Pressure Drop in BWR Fuel Assemblies," Exxon Nuclear Company, Inc., Richland, WA.

• 19. NEDE-240ll-P A, General Electric Standerd Ap)11 cation for Reactor Fuel (GESTAR-II) with exception to tie misplaced fuel bundle analyses as discussed in GNRO 96/00087 and i

the generic MCPR Safety Limit analysis as discussed in GNRO-96/00100, letters from C. R. Hutchinson to USNRC.

( *20. Jll 02863SLMCPR, Revision 1, "GGNS Cycle 9 Safety Limit l MCPR Analysis."

(continued)

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• ltems 19 and 20 of TS 5.6.5.b are applicab/le only for Cycle @peration.

GRAND GULF 5.0 20 Amendment No. 440, 131 y 9 y 'e,- r+- - -- v -- p. e,,_,.w -,-----e.- wa--- , , + - -

Reactor Core SLs ,

B 2.1.1  ;

. BASES APPLICA8LE 2.1.1.1 Euel Claddina Intearity (continued)

SAFETY ANALYSES i ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With  ;

the design peaking factors, this corresponds to a

THERMAL POWER > 55 RTP. Thus a THERMAL POWER limit

of 2 5 RTP for reactor pressure < 785 psig is conservative. Because of the design therwal hydraulic cc=ptibility of the reload fuel designs with the cycle 1 fuel, this justification and the associated low pressure and low flow limits remain applicable for r

.i $.

future cycles of cores containing these fuel designs.

2.1.1.2 528 I  :

Y Bf The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an A00 from the limiting 5' ] condition of operation, at least 99.5 of the fuel rods in 3 the core would be ex ecte.:

j(f a

margin-betweencalcu$atedboilingtransition(i.e.,to The avoid b ,

MCPR = 1.00) and the MCPR SL is based on a detailed

• m I statistical procedure that considers the uncertainties in .

monitoring the core operating state. One specific j uncert

{

in the in

g=, includedintheSListheuncertainty?

ritica'l ower correlation. Reference 4 I descri e methodo ogy used in determining the MCP .

g % 'I .g '

-l y - aL The fuel vendor's critical power correlations are based on a 1 significant body of practical test data, providing a high 1

jI( degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual

]\ critical power being estimated. As long as the core

j. i,1g .g (I pressure and flow are within the range of validity of the correlations, the assumed reactor conditions used in t

I defining the SL introduce conservatism into tha limit

• because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition.- These conservatisms and the inherent accuracy of the fuel vendor's correlation provide a reasonable degree of assurance that 99.5 of the~ rods in the .

core would not be susceptible to transition boiling during -

sustained operation at the MCPn SL. If boiling transition were to occur, there is reason ') believe that the integri+y I

.l feontinued)

GRAND G')LFJ B 2.0-3 Revision No. 2

o I MCPR 8 3.2.2 l

BASES (continued) k.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is a 25% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation, The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER reaches a 25% RTP is acceptable given the large inherent margin to operating limits at low power levels.

REFERENCES 1. NUREG 0562, " Fuel .911ures As A Consequence of Nuc1 sate Boiling or Dry Out," June 1979.

2. NEDE-240ll P A, General Electric Standard Application for Reactor Fuel (GESTAR-!!).

3. UFSAR, Chapter 15, Appendix 158,

4. UFSAR, Chapter 15, Appendix 15C.

5. UFS/Ji, Chapter 15, Appendix 150.

6. NEDE-30130-P-A, Steady State Nuclear Methods.

l

7. NEDO 24154, Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors. l

8. Deleted i 0, Amendment 131 to the Operating License.

- l GRAND GULF B 3.2-8 Revision No. 2 l

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. e Attachment 4 to GNRO 97/00087 i t

GGNS PCOL-97/003 .

Additional Information Regarding 1.11 Cycle Specific SL MCPR for Grand Gulf-Cycle 10 I

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