ML20206K536

Proposed Tech Specs Pages,Changing SLMCPR

ML20206K536
Person / Time
Site:	Grand Gulf
Issue date:	05/06/1999
From:	ENTERGY OPERATIONS, INC.
To:
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ML20137N459	List: ML20206K524 ML20206K536
References
NUDOCS 9905130193
Download: ML20206K536 (11)

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SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:

l THERMAL POWER shall be s 25% RTP.

p.l.l.2 With the reactor steam dome pressure a 785 psig and core flow a 10% rated core flow:

MCPR shall be a for two recirculation loop operation or a M2 for single recirculation loop operation. g 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 Reactor 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 I hour, 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: l l

2.2.2.1 Restore compliance with all SLs; and I l

2.2.2.2 Insert all insertable control rods. I I

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2.2.3 Within 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 responsible for overall plant nuclear safety.

l N

• MCPR va l

(continued)

GRAND GULF 2.0-1 Amendment No. -140r131,136 9905130193 990506 ~~

PDR P

ADOCK 05000416 PDR L_

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 Core Operatina Limits Report (COLR) (continued)

10. XN-NF-85-74(P)(A), "RODEX2A (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 with 10CFR50 Appendix K Heatup Option," Exxon Nuclear  ;

Company, Inc., Richland, WA. l

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

13. XN-NF-81-51(P)(A), "LOCA-Seismic Structural Response'of an Exxon Nuclear 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. I

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.

p9. NEDE-240ll-P-A, General Electric Standard Application for Reactor Fuel (GESTAR-II) with exception to the misplaced fuel bundle analyses as discussed in GNR0-96/00087 and the generic MCPR Safety Limit analysis as discussed in GNR0-96/00100, letters from C. R. Hutchinson to USNRC.

h0.J11-02863SLMCPR, Revision 1,"GGNSCycle9SafetyLimit MCPR Analysis."

(continued)

• Items 19 and 20 of TS 5 ycle 10 operation. l GRAND GULF 5.0-20 Amendment No. M e r M1, 136

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Claddina Intearity (continued)

SAFETY ANALYSES j 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 > 50% RTP. Thus a THERMAL POWER limit of 25% RTP for reactor ptessure < 785 psig is conservative. Because of the design thermal hydraulic compatibility of the reload fuel designs with the cycle 1 fuel, this justification and the associated low pressure and low flow limits remain applicable for future cycles of cores containing these fuel designs.

2.1.1.2 MCPR The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an A00 from the limiting ,

condition of operation, at least 99.9% of the fuel rods in I the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,

MCPR - 1.00) and the MCPR SL is based on a detailed ,

statistical procedure that considers the uncertainties in i monitoring the core operating state. One specific  !

uncertainty included in the SL is the uncertainty inherent in the critical power correlation. Referenc8)66 descri g e methodology used in determining The MCPR SL.

The ca culated MCPR safety limit is. reported to the customary three significant digits (i.e., X.XX); the MCPR operating limit is developed based on the calculated MCPR safety limit to ensure that at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.

The fuel vendor's critical power correlations are based on a significant body of practical test data, providing a high 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 pressure and flow are within the range of validity of the correlations, the assumed reactor conditions used in defining the SL introduce conservatism into the 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 (continued)

GRAND GULF B 2.0-3 LDC 98033

Reactor Core SLs B 2.1.1 BASES SAFETY LIMIT 2.2.5 VIOLATIONS i (continued) If any SL is violated, restart of the unit shall not comence until authorized by the NRC. This requirement i

ensures the NRC that all necessary reviews, analyses, and .

i actions are completed before the unit begins its restart to l l normal operation.

1 REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. XN-NF524(A), Revision 2, April 1989.

3. 10 CFR 50.72.

4. 10 CFR 100.

5. 10 CFR 50.73.

6. EDE-24011-P-A G UAR I l

7. GNRI-96/00230, Amendment 131 to the Operating License. l

)

GRAND GULF B 2.0-6 Revision No. 2 1

. . MCPR B 3.2.2 BASES (continued)

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS l 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 I 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 Failures As A Consequence of Nucleate Boiling or Dry Out," June 1979.

2. NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR-II).

3. UFSAR, Chapter 15, Appendix 15B.

4. UFSAR, Chapter 15, Appendix 15C.

I

5. UFSAR, Chapter 15, Appendix 150.

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

7. NE00-24154, Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors.

8. Deleted

9. GNRI-98/00058, Amendment 136 to the Operating License. l l

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I GRAND GULF B 3.2-8 LDC 98033 l

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Attachment 6 to GNRO- 99/00037 i

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Additional Information Regarding the l Calculated Cycle Specific SL MCPR j for i Grand Gulf-Cycle 11 )

(Non-proprietary Version) I

References ll] General Electric BWR Thermal Analysis Basis (GETAB): Data. Correlation and Design Application, NEDO-10958-A, January 1977.

[2] General Electric Standard Applicationfor Reactor Fuel (GESTAR II), N EDE-240 l 1-P-A-11, November 1995.

(3] General Electric Standard Applicationfor Reactor Fuel (GESTAR II), N EDE-240 l 1 -P-A- 13, August 1996.

[4] General Electric Fuel Bundle Designs, NEDE-31152-P, Revision 5, June 1996.

(5} Methodology and Uncertaintiesfor Safety Limit MCPR Evaluations, NEDC-32601 P, Class 1II, December 1996.

[ 6] R-Factor Calculation Methodfor Gell, GE12 and GE13 Fuel, NEDC-32505P, June 1997.

Comparison of Grand Gulf Cycle 11 SLMCPR versus the Generic Gell Value Table I summarizes the relevant input parameters and results of the SLMCPR determination for the generic GElI core and the Grand Gulf Cycle 11 core. Both generic and plant / cycle specific evaluations are performed using the methods described in GETAB[l}. The uncertainties applied in this analysis are reported in Table 2. These evaluations yield different calculated SLMCPR values because the inputs that are used are different. The quantities that have been shown to have some impact on the determination of

)

the safety limit MCPR (SLMCPR) are provided. Much of this information is redundant but is pmvided in  !

this case because it has been provided previously to the NRC to assist them in understanding the differences between plant / cycle specific SLMCPR evaluations and the generic values calculated previously by fuel product line. ((

))

Prior to 1996, GESTAR 11[2] stipulated that the SLMCPR analysis for a new fuel design be performed for a large high power density plant assuming a bounding equilibrium core. The gel 1 product line generic SLMCPR value of 1.07 was determined according to this specification. Later revisions to GESTAR 11[3](see Section 11," Cycle MCPR Values", of Appendix A of U.S. Supplement to GESTAR-II, Revision 13) that have been submitted to the NRC to describe how plant / cycle specific SLMCPR analyses are used to confirm the calculated SLMCPR value on a plant / cycle specific basis using the uncertainties defined in Reference [4].

In comparing the generic Gell and Grand Gulf Cycle 11 SLMCPR values it is important to note that the Grand Gulf Cycle i1 core is not an equilibrium core. It is a mixed core with GElI and SPC 9x9-5 fuel. The Grand Gulf Cycle 11 core is loaded such that the fresh batch of Gell has the highest enrichment (( -)), as compared to a core average enrichment of(( )), as shown in Table 1.

' By way of comparison, the generic gel I equilibrium core has batch and core average enrichments of ((

)). Higher enrichment in the fresh fuel for the Grand Gulf Cycle 11 core (compared to the rest of the core) produces higher power in the fresh bundles relative to the rest of the core. These enrichment differences result in the Gell fresh fuel producing a higher relative share of the number of fuel rods calculated to be ausceptible to _ boiling transition (NRSBT). In fact, the Cycle 11 analysis determined that the NRSBT from the SPC fuel was (( )).

The core MCPR distribution for the Grand Gulf Cycle 11 core is somewhat simi!ar to the core MCPR distribution used in the gel 1 generic analysis. ((

)) Ilowever,((

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((

)) The Grand Gulf Cycle 11 core has a flatter core MCPR distribution than the generic Gell equilibrium core as well as a flatter bundle R-factor distribution.

((

))

((

))

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((

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The uncontrolled bundle pin-by-pin power distributions were compared between the Grand Gulf Cycle 1I bundles and the GElI bundle used in the generic SLMCPR analysis. Pin-by-pin power distributions are characterized in terms of R-factors using the methodology defined in Reference [ 6]. For the Grand Gulf Cycle 11 bundles, there is a slightly flatter distribution of uncontrolled R-factors for the highest power rods in each bundle, which in the calculation are the rods most likely to be susceptible to boiling transition. This fact is difficult to ascertain graphically since the relative flatnesses are similar and the rods that have an R-factor closer to the R-factor for the lead rod are statistically worth much more than those that have R factors that are further away, (( The flatness of the pin R-factor distribution within a particular bundle is characterized ((

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(( 11 i Table 1 Comparison of Gell Generic and Grand Gulf Cycle 11 Core and Bundle Quantities that l Impact the SLMCPR (( j i l l i

ll Table 2 Uncertainties Applied in the Grand Gulf Cycle 11 SLMCPR Evaluation (( I ll j 1 l l l l [ L

Summary Various quantities (( )) have been used over the last year to compare quantities that impact the calculated SLMCPR value. These other quantities have been provided to the NRC  ; previously for other plant / cycle speciGe analyses using a format such as that given in Table 1. These I other quantities have also been compared for this core / cycle ((

                 )). Note that regardless of what comparisons are made, the conclusion is easily reached that the Grand Gulf Cycle 11 core / cycle has a Datter core MCPR distribution ((                          )) and substantially Hatter in-bundles power distributions ((                          )) than what was used to perform the gel 1 generic SLMCPR evaluation.

4 The calculated 1.08 Monte Carlo SLMCPR for Grand Gulf Cycle 11 is consistent with what one would . expect (( l

                                                                                                    )) the 1.08 SLMCPR value is appropriate.

Based on all of the facts, observations and arguments presented above, it is concluded that the calculated SLMCPR value of 1.08 for the Grand Gulf Cycle 11 transition core is appropriate. It is reasonable that  ! this value is 0.01 higher than the 1.07 value calculated for the generic GElI equilibrium core. For single loop operations (SLO) the calculated safety limit MCPR for the limiting case is 1.09 ((  ;

                                                                   ))

Prepared by: VeriGed by: W. E. Russell h F. T. Bolg Technical Project Manager Nuclear Fue ngineering Grand Gulf Project}}