ML20116F021

Proposed Tech Specs Required to Support Cycle 9,reload 8

ML20116F021
Person / Time
Site:	Grand Gulf
Issue date:	07/31/1996
From:	ENTERGY OPERATIONS, INC.
To:
Shared Package
ML19311C160	List: ML20116F014 ML20116F021
References
GNRO-96-0087, GNRO-96-87, NUDOCS 9608060203
Download: ML20116F021 (14)

Text

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l Attachment 3 to GNRO-96/0087 I

l Mark-up Pages of the Affected l

Technical Specifications

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Technical Specification Bases I

(REVISION 1)

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l 9608060203 960731 PDR ADOCK 05000416 P PDR

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:

THERMAL POWER shall be s 25% RTP.

2.1.1.2 With the reactor steam done pressure a 785 psig and core flow a 10% rated l MCPR shall be m +:4Vfor two recirculation loop operation or a for single recirculation loop operation.

- A.LA 2.1.1.3 Reactor vesse 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 l 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 l 2.2.2.1 Restore compliance with s11 SLs; and j 2.2.2.2 Insert all insertable control rods. )

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, i l 4

i

1 (continued)

GRAND GULF 2.0-1 Amendment No. 120 i

Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref.1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (A00s).

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fue cladding integrity.

The fuel cladding is one of the physical barriers that l separate the radioactive materials from the environs. The j integrity of this cladding barrier is related to its  !

relative freedom from perforations or cracking. Although j some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously ,

measurable. Fuel cladding perforations, however, can result i from thermal stresses, which occur from reactor operation significantly above design conditions.

1 While fission product migration from cladding perforation is l just as measurable as that from use related cracking, the  ;

thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL ensures that during normal operation and during A00s, at i least 99.9%, of the fuel rods in the core do not experience transition boiling.

(continued)

GRAND GULF B 2.0-1 Revision No. 0

\

f 1 l

Reactor Core SLs B 2.1.1 BASES I

I APPLICABLE 2.1.1.1 Fuel Claddino Integrity (continued) i SAFETY ANALYSES I ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical j power at this flow is approximately 3.35 MWt. With 1 the design peaking factors, this corresponds to a THERMAL POWER > 50% RTP. Thus a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig is conservative. Because of the design thermal hydraulic compatibility of the reload fuel designs with the 1 cycle I fuel, this justification and the associated I low pressure and low flow limits remain applicable for  !

future cycles of cores containing these fuel designs.

l 1

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 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 ANFB critical power correlation. Reference ? @

describes the methodology used in determining the MCPR SL.

lvutt wasoong The 2., critical power correlatio ased on a significant body of practical test ata, 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 ANF6 correlaticjiDthe assumed reactor cowditions used in defining the SLTntroduce 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. Still furth:r conserv:ti:=

4s-4aduced by the tendency of the ?fB correktien te everprM4ct the Per Of red: in beiling tr:::ition._ These conservatisms and the inherent accuracy of the-ANFBMutt iswM'O l

'99 405 ms @ correlation provide a reasonable degree of assurance that m maeeu.weath th;r; ::;1d be-no-transition-belling-in-the-c+ee- during

! *T ru sussime.tE l sustained operation at the MCPR SL. If boiling transition j Q,7f{ oui " were to occur, there is reason to believe that the integrity

' (continued) i l

GRAND GULF B 2.0-3 Revision No. O

l l

Reactor Core SLs B 2.1.1 8 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, norwal operational transients, and anticipated operational occurrences (A00s).

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit I is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously l measurable. Fuel cladding perforations, however, can result ,

from thermal stresses, which occur from reactor operation significantly above design conditions.  ;

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL ensures that during normal operation and during A00s, at least 99.9% of the fuel rods in the core do not experience transition boiling.

(continued)

GRAND GULF B 2.0-1 Revision No. O

- - _ . .- - - _ . . - ~ _ . - - - - . . . - . - - . - - - - - - _ - _ - - . -

Reactor Core SLs B 2.1.1 BASES BACKGROUND Operation above the boundary of the nucleate boiling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES nomal operation and A00s. The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR SL is to be established, such that at least 99.9%,of the fuel rods in the core would not be expected to experience the onset of transition boiling.

The Reactor Protection System setpoints (LC0 3.3.1.1,

" Reactor Protection System (RPS) Instrumentation"), in combination with other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR SL.

2.1.1.1 Fuel Claddina Intearity

%n.ys.we,esn. c..amc AL t==em.g The use of theMeorrelation is valid for critical power mes gealculations at pressures > psig and bundle mass fluxe

> 4rf5- x 10' lb/hr-fts (Re . . For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flow will always be

> 4.5 psi. Analyses show that with a bundle flow of 28 x 108 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus the bundle flow with a 4.5 psi driving head will be > 28 x 108lb/hr. Full scale (continued)

GRAND GULF B 2.0-2 Revision No. O

l t

! Reactor Core SLs B 2.1.1 BASES l

APPLICABLE 2.1.1.1 Fuel Claddina Intearity (continued) i SAFETY ANALYSES l 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 pressure < 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 l

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 monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty inherent in the ANFB critical power correlation. Reference 2 @)

describes the methodology used in determining the MCPR SL.

g Evue.t. sanoeng" The R critical power correlatioOs 4-Ibased 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

-ANF9-correlatiOnfr the assumed reactor conditions used in defining the SL'Tntroduce 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. Sti" furth:r censerv:ti =

-1: induced by the tendency of-the-ANFS-arrelatten te

-overpredict the Wer Of reds in-bo444ng-trans4 tion._ These conservatisms and the inherent accuracy of the-ANFfK5utt. %%s'O qq.ge, r d correlation provide a reasonable degree of assurance that m ms c. owr wout.h ther: =uld-be-no-transition-boiling-in-the-core during ueT ru susesmusi sustained operation at the MCPR SL. If boiling transition Q ] p { W W were to occur, there is reason to believe that the integrity (continued) l i

GRAND GULF B 2.0-3 Revision No. O

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR (continued)

SAFETY ANALYSES of the fuel would not be compromised. Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach.

i Much of the data indicate that BWR fuel can survive for an extended period of time in an environment of boiling transition.

2.1.1.3 Reactor Vessel kater level During MODES I and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel l .during periods when the reactor is shut down, consideration l

must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top

> of the active irradiated fuel during this period, the

! ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes less than two-thirds of the core height.

The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

i SAFETY LIMITS The reactor core SLs are established to protect the i integrity of the fuel clad barrier to the release of l radioactive materials to the environs. SL 2.1.1.1 and l SL 2.1.1.2 ensure that the core operates within the fuel i design criteria. Sl 2.1.1.3 ensures that the reactor vessel water level is grenar than the top of the active irradiated I fuel in order to preunt elevated clad temperatures and resultant clad perforation.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

s (continued)

GRAND GULF B 2.0-4 Revision No. O

Reactor Core SLs l

B 2.1.1 BASES (continued)

]

SAFETY LIMIT 2.2.1 4

VIOLATIONS If any SL is violated, the NRC Operations Center must be notified within I hour, in accordance with 10 CFR 50.72 (Ref. 3).

2.2.2 Exceeding an SL may cause fuel damage and create a potential i

for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits'(Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. (The required actions for a violation of the reactor water level SL include manually initiating ECCS to restore water level and depressurizing l the reactor vessel, if necessary, for ECCS operation.) The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal. 1 2.2.3 If any SL is violated, the General Manager, Plant Operations and the Vice President, Operations GGNS shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provides time for plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to the senior management.

2.2.4 If any SL is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC in accordance with 10 CFR 50.73 (Ref. 5). The report will describe the applicable circumstances preceding the violation, the effect of the violation upon unit components, systems, or structures, and the corrective actions taken to prevent recurrence. A copy of the report shall also be submitted to the General Manager, Plant Operations and the Vice President, Operations GGNS.

(continued)

GRAND GULF B 2.0-5 Revision No. 1

/3

i Reactor Core SLs B 2.1.1 l

BASES 1

SAFETY LIMIT 2.2.5 VIOLATIONS (continued) If any SL is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation.

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.

l (t_ NSEE l_@ot {-% GE? TAR-h _

l GRAND GULF B 2.0-6 Revision No. I

. l MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Minimum Critical Power Ratio (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (A00s). Although fuel damage does not necessarily occur if a fuel rod actually experiences boiling transition (Ref.1),

the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e.,

the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the A00s to ~stablish the operating limit MCPR are presented in the UFSAR, Chapters 4, 6, and 15, and References 2, 3, 4, and 5. To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (E PR). When the largest E PR is added to the MCPR SL, the required operating limit MCPR is obtained.

(continued)

GRAND GULF B 3.2-5 Revision No. O

MCPR B 3.2.2 ,

1 l

BASES I APPLICABLE The MCPR operating limits derived from the transient SAFETY ANALYSES analysis are dependent on the operating core flow and power (continued) state (MCPR, and MCPR,, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Refs. 3, 4, and 5). Flow dependent MCPR limits are determined by steady state thermal hydraulic methods using the three dimensional BWR simulator code (Ref.

6).and the -"Itich:nn:1 ........ ..,.... ,, - m .. M .

MCPR, curves are provided based on the maximum credible flow runout transient for Loop Manual operation. The result of a single failure or single operator error during Loop Manual operation is the runout of only one loop because both recirculation loops are under independent control.

Power dependent MCPR limits (MCPR,) are determined by the three dimensional Epul simulator code and the one dimensional transient code (Ref94). The MCPR, limits are established j for a set of exposure intervals. The limiting transients )

are analyzed at the limiting exposure for each interval.

Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast I closure scram trips are bypassed, high and low flow MCPR, operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of the NRC Policy Statement.

LC0 The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The MCPR operating limits are determined by the.

larger of the MCPR, and MCPR, limits.

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a slow l recirculation pump speed and the moderator void ratio is j small. Surveillance of thermal limits below 25% RTP is i unnecessary due to the large inherent margin that ensures  ;

that the MCPR SL is not exceeded even if a limiting l transient occurs. l l

l (continued)

GRAND GULF B 3.2-6 Revision No. 0 ,

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_ .. -. . . -- . _ _ . . . . _ _ ~ - . - - - . - . - - - . . . _ - _ - .- __ .-- ._

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MCPR 8 3.2.2 BASES APPLICABILITY Studies of the variation of limiting transient behavior have (continued) been performed over the range of power and flow conditions.

These studies encompass the range of key actual plant 4

parameter values important to typically limiting transients.

The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in M00E 2 the intemediate range monitor (IM) provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern.

Therefore, at THERMAL POWER levels < 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LC0 is not required.

4 ACTIONS ad If any MCPR is outside the required limit, an assumption regarding an initial condition of the design basis transient analyses may not be met. Therefore, prompt action should be taken to restore the MCPR(s) to within the required limit (s) such that the plant remains operating within analyzed conditions. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is normally sufficient to restore the MCPR(s) to within its limit and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of i specification.

i Ed l If the MCPR cannot be restored to within the required limit within the associated Completion Time, the plant must be i

brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER '

sust be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed I Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

I I

(continued)

GRAND GULF 8 3.2-7 Revision No. 0 ,

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i l i l l

l MCPR l l

i B 3.2.2 i l l

I l BASES (continued)

! SURVE!LLANCE SR 3.2.2.1 l REQUIREMENTS.

The MCPR is required to be initially calculated within l 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 251 RTP and then every j

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 l 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 /> i

i Frequency is based on both engineering judgment and i recognition of the slowness of changes in power distribution ,

l 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 THERMe.L POWER reaches a 255 RTp is acceptable given the l

' large inherent margin to operating limits at low power 1evels.

l j

! REFERENCES 1. NUREG-0582, " Fuel Failures As A Consequence tf I l

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i 3. UFSAR, Chapter 15, Appendix 155.

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! 4. UFSAR, Chapter 15. Appendix 15C. i

5. UFSAR, Chaater 15. Aeoendix 4 -- .

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1 1 GRAND GULF B 3.2-8 Revision No. O s

a l

Reporting Requirements 5.6 5.6 Reporting Requirements l 5.6.5 Core Operatino limits Report (COLR) (continued)

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

GE4

11. XN-CC-33(P)(A), "HUXY: A Generalized Multirod Heatup Code

$ with 10CFR50 Appendix K Heatup Option," Exxon Nuclear g Company, Inc., Richland, WA. l

• o

'E g 12. XN-NF-825(P)(A), "BWR/6 Generic Rod Withdrawal Error I o Analysis, MCPR, for Plant Operation Within the Extended

)5 g Operating Domain," Exxon Nuclear Company, Inc., Richland, 1

x WA. l

.l 5 0 13. XN-NF-81-51(P)(A), "LOCA-Seismic Structural Response of

.g 8!

~

an Exxon Nuclear Company BWR Jet Pump Fuel Assembly,"

i Exxon Nuclear Company, Inc., Richland, WA.

, } ,s f 14. XN-NF-84-97(P)(A), "LOCA-Seismic Structural Response of ll an ENC 9x9 BWR Jet Pump Fuel Assembly," Advanced Nuclear A]3a Fuels Corporation, Richland, WA.  ;

5_ 3 ~i j y 15. XN-NF-86-37(P), " Generic LOCA Break Spectrum Analysis for

.a1. %g BWR/6 Plants," Exxon Nuclear Company, Inc., Richland, WA.

I j 16. XN-NF-82-07(P)(A), " Exxon Nuclear Company ECCS Cladding

&j Swelling and Rupture Model," Exxon Nuclear Company, Inc.,

j 0; Richland, WA.

$ l

17. XN-NF-80-19(A), Volumes 2, 2A, 2B, & 2C, " Exxon Nuclear

• Methodology for Boiling Water Reactors EXEM BWR ECCS

{ luation Model," Exxon Nuclear Company, Inc., Richland,

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

n

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety

analysis are met.

(continued)

GRAND GULF 5.0-20 Amendment No. 120 1

i *

, I Attachment 4 to GNRO-96/0087 i

4 General Electric l GGNS Cycle 9 SLMCPR Analysis (REVISION 1)

(GE Properietary Information. Not included for the distribution) l A -

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