Text

Amend 57 to License NPF-29,changing Tech Specs Re Safety Limits,Power Distribution Limits & Fuel Assemblies to Support Fuel Cycle 4 Load
	ML20236C408
Person / Time
Site:	Grand Gulf
Issue date:	03/13/1989
From:	Reeves E; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20236C411	List: ML20236C408; ML20236C424;
References
	NUDOCS 8903220135
	Download: ML20236C408 (18)
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UNITED STATES

[

k NUCLEAR REGULATORY COMMISSION s

WASHINGTON, D. C. 20006 y.....)

I SYSTEM ENERGY RESOURCES INC., et al

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2 DOCKET N0. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 1

AMENDMENT TO FACILITY OPERATING LICENSE-Amendment No. 57 License No. NPF-29 1.

The Nuclear Regulatory Commission (the Commission) has found that A.

The application for amendnent by System Energy Resources, Inc.,

(the licensee), dated December 6, 1988, as supplemented December 30, 1988 and January 31, 1989, complies with the standards.

and requirements of the Atomic Energy Act of 1954, as amended (the.

Act), and the Consission's rules and regulations set forth in 10 CFR Chapter I; f

)

8.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Connission; C.

Thereisreasonableassurance(1)thattheactivitiesauthorizedby l

this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of the public; ar.d E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all-applicable requirements have been satisfied.

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'Accordingly, the' license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; j

andparagraph2.C.(2)ofFacilityOperatingLicenseNo.NPF-29.ishereby 1

amended to read as follows:

(2) Technical Specifications The Technical Specifications contained'in Appendix A and.the.

Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 57, are hereby incorporated into this f

~ license. System Energy Resources, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental i

Protection Plan.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY. COMMISSION ex Edward A. Reeves, Acting Director Project Directorate 11-1 Division of Reactor Projects I/II i

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical.

Specifications Date of Issuance: March 13,1989 s

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, f ATTACHMENT TO LICENSE AMENDMENT NO. 57 FACILITY OPERATING LICENSE NO. NPF-29 DOCKET NO. 50-416 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert i

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B 3/4 2-1 B 3/4 2-1 8 3/4 2-4 B 3/4 2-4 L

B 3/4 2-6 8 3/4 2-6 5-5 5-5 4

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2.1 SAFETY LIMITS BASES THERMAL POWER. Low Pressure or Low Flow (Continued)

The Advanced Nuclear Fuels Corporation (ANF) XN-S critical power correla-l tion is applicable to the mixed core beginning with cycle 2.

The applicable range of the XN-3 correlation is for pressures above 585 psig and bundle mass flux greater than 0.25M1bs/hr-ft2 For low pressure and low flow conditions, a THERMAL POWER safety limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig and below 10% RATED CORE FLOW was justified for Grand Gulf cycle 1 operation based on ATLAS test data.

Overall, because of the design thermal-hydraulic compatibility of the ANF 8x8 fuel design.with the cycle 1 l

fuel, this justification and the associated low pressure and ' low flow limits remain applicable-for future cycles of cores containing these' fuel designs.

With regard to the low flow range, the core's bypass region'will.be flooded' at any flow rate greater than 10% RATED CORE FLOW.

With the bypass region flooded, the associated elevation head is sufficient to assure a bundle mass flux of greater than 0.25 M1bs/hr-ft2 for all fuel assemblies which can approach critical heat flux.

Therefore, the XN-3 critical power correlation is appro-priate for flows greater than 10% RATED CORE FLOW.

The low pressure rcnge for cycle 1 was defined at 785 psig.

Since the XN-3 correlation is applicable at any pressure greater than 585 psig, the cycle 1 low pressure boundary of 785 psig remains valid for the XN-3 i

correlation.

l i

GRAND GULF-UNIT 1 B 2-la Amendment No. 57 e _ ____ __. __

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SAFETY LIMITS BASES 2.1.2 THERMAL POWER High Pressure and High Flow l

The' onset of transition boiling results in a decrease in heat transfer from the clad, elevated clad temperature, and the possibility of clad failure.

However, the existence of critical power, or boiling transition, is not a di-rectly observable parameter in an operating reactor.

Therefore, the margin.to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution.

The mar-gin for each fuel assembly is characterized by the critical power ratio (CPR),

which is the ratio of the bundle power which would produce enset of transition boiling divided by.the actual bundle power.

The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR).

l The Safety Limit MCPR assures sufficient conservatism such that, in the l

event of a sustained steady state operation at the MCPR safety limit, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transi-tion.

The margin'between calculated boiling transition (MCPR.= 1.00) and the-Safety Limit MCPR is based on a detailed statistical procedure which considers the uncertainties in monitoring the core operating state.

One specific uncer-l tainty included in the safet critical power correlation. y limit is the uncertainty inherent in the XN-3 ANF report XN-NF-524(A), Rev.1, " Exxon Nuclear l.

Critical Power Methodology for Boiling Water Reactors," Nov. 1983, describes the methodology used in determining the Safety Limit MCPR.

The XN-3 critical power correlation is based on a significant body of l

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.

The assumed reactor conditions used in defining the safety limit introduce conservatism into the' limit because bound-ing high radial power factors and bounding flat local peaking distributions are used to estimate the number nf rods in boiling transition.

Still further cor.-

servatism is induced by the tendency of the XN-3 correlation to overpradict the number of rods in boiling transition.

These_ conservatives and the inherent accuracy of the XN-3 correlation provide assurance that during sustained opera-tion at the Safety Limit MCPR there would be essentially no transition boiling in the core.

i l

1 L

GRAND GULF-UNIT 1 8 2-2 Amendment No. 57

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1 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 Ouring two loop operation all AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-la, 3.2.1-1b, 3.2.1-1c, 3.2.1-1d, or 3.2.1-le as multiplied by the smaller of either the' flow-dependent MAPLHGR factor (MAPFAC ) of Figure 3.2.1-2, or the power-dependent MAPLHGR f

factor (MAPFAC ) of Figure 3.2.1-3.

p During single loop operation, the APLHGR for each type of fuel as a ftnction of AVERAGE PLANAR EXROSURE shall not exceed the limits as determined below:

a) for 8x8 ANF fuel types - the limit shown in Figure 3.2.1-1 l

as multiplied by the smaller of either MAPFAC, MAPFAC or 0.86; f

p and b) for 9x9 ANF ft.41 type the limit determined in Figure 3.2.1-le as multiplied by the smaller of either MAPFAC, MAPFAC or 0.86.

f p

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than '

or equal to 25% of RATED THERMAL POWER.

ACTION:

Ouring two loop operation or single loop operation, with an APLHGR exceeding the limits of Figures 3.2.1-1, 3.2.1-la, 3.2.1-lb, 3.2.1-Ic, 3.2.1-1d or 3.2.1-le as corrected by the appropriate multiplication factor for each type of fuel, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hout s or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGks shall be verified to be equal to or less than the required limits:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 t)ours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

GRAND GULF-UNIT 1 3/4 2-1 Amendment No. 57

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1 POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENI: RATION RATE (LHGR) shall not exceed the limits shown in Figure 3.2.4-1.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With the LHGR of any fuel rod exceeding the limit of Figure 3.2.4-1, initiate.

l corrective action within 15 minutes and restore the LHGR to within the-limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

t SURVEILLANCE REQUIREMENTS 4.2.4 LHGR's shall be determined to be equal to or less than their allowable l

limits:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERML POWER, Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is c.

operating on a LIMITING. CONTROL R00 PATTERN for LHGR, and d.

The provisions of specification 4.0.4 are not applicable.

i GRAND GULF-UNIT 1 3/4 2-7 Amendment No. 57-

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o 3/4.2' POWER DISTRIBUTION LIMITS j

BASES The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not I

exceed the 2200*F limit specified in 10 CFR 50.46.

l 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of coolant accident will not exceed the limit

)

l specified in 10 CFR 50.46.

j The peak cladding temperature (PCT) following a postulated loss-of coolant i

accident is primarily a function of. the average heat generation rate of all f

i the rods of a fuel assembly at any axial location and is dependent 'only secondar-ily on the rod to rod power distribution within an assembly. The peak clad _ tem-perature is calculated assuming a LHGR for the highest powered rod which is 1

equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady-state gap conductance and rod-to-rod local peaking factor.' The Technical Speci-

]

fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of' s

the highest powered rod divided by its local peakin factor.

The Maximum Aver-

{

age Planar Linear Heat Generation Rate (MAPLHGR) li its of Figures 3.2.1-1, j

3.2.1-la, 3.2.1-lb, 3.2.1-1c, 3.2.1-1d or 3.2.1-le are multiplied by the smaller l

j of either the flow dependent MAPLHGR factor (MAPFAC ) or the power dependent MAPLHGR factor (MAPFAC ) corresponding to existing ore flow and power state to p

assure the adherence to fuel mechanical design bases during the most. limiting transient.

I For single-loop operation with ANF 8x8 fuel, a MAPLHGR limit corresponding f

to the product of the MAPLHGR, Figure 3.2.1-1, and the appropriate MAPFAC, can be conservatively used.

The allowable MAPLHGR shown in Figere 3.2.1-1 is a conservative bound during Cycle 4 for all 8x8 fuel types and the Cycle 3 SLO j

MAPLHGR (Reference 5).

The MAPLHGR limit for ANF 9x9-5 fuel is the product of i

the MAPLHGR shown in Figure 3.2.1-le and the appropriate MAPFAC.

The maxtmum j

MAPFAC during single loop operation is 0.86 for all fuel types.

1 MAPFAC 's are determined using the three-dimensional BWR' simulator code to f

maximum credible flow runout transient for ANF fuel for either Loop Manual or f

analyze slow flow runout transients.

Two curves are provided based on the Non loop Manual operation.

The result of a single failure or single operator I

error during operation in Loop Manual is the runout of only one loop because both recirculation loops are under independent control.

Non-Loop Manual t

operational modes allow simultaneous runout of both loops because a single i

controller regulates core flow.

^

1 MAPFAC 's are generated to protect the core from plant transients other il p

than core flow increases.

3 The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient ~since power distribu-tion shifts are very slow when there have not been significant power or control i

GRAND GULF-UNIT 1 B 3/4 2-1 Amendment No. 57 d

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POWER DISTRIBUTION LINITS BASES l

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel clad-ding integrity Safety Limit MCPR, and an analysis of abnorm 41 operational tran-sients.

For any abnonnal operating transient analysis evaluation with the I

1 initial condition of the reactor being at the steady. state operating limit, it is required that the resulting'MCPR does not decrease belaw the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in' Specification 2.2.

i To assure that'the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATI0'(CPR). The type of transients evaluated were loss of flow, increase in pressiire and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yie ks the largest delta CPR.

When added to the Safety Limit MCPR, the required' operating limit MCPR nf Speciff;:ation 3.2.3 is obtained.

The power-flow ' nap of Figure B 3/4 2.3-1 defines the analytical basis for generation of the MCPR operating limits (Reference 7).

I The purpose of the MCPR and MCPR is to define operating limits at other g

2 f

p than rated core flow and power conditions.

1 The MCPR s are established to protect the core from inadvertent core flow l

f increases such that the 99.9% MCPR limit requirement can be assured.

The ref-erence core flow increase event used to establish the MCPR is a hypothesized f

slow flow runout to maximum, that does not result in a scram from neutron flux overshoot exceeding the APRM neutron flux-high level ~~(Table 2.2.1-1 item 2).

Two flow rates have been considered.

The maximum credible flow during a runout transient depends on whether the plant is in Loop Manual or Non Loop Manual operation.

The result of a single failure or single operator error.

during Loop Manual operation is the runout of one loop because the two recirculation loops are under independent control.

Runout of both loops is possible during Non Loop Manual operation because a single controller regulates core flow. With this basis, the MCPR curves are generated from a f

series of steady state core thermal hydraulic calculations performed at several core power and flow conditions along the steepest flow control line.

In.the actual calculations a conservative highly steep generic representation of the 105% steam flow rodline flow control line has been used.

Assumptions used in the original calculations of this generic flow control line were consistent with a slow flow increase transient duration of several minutes:

(a) the plant heat balance was assuidd to be in equilibrium, and (b) core xenon concentration s

l l

1 GRAND GULF-UNIT 1 B 3/4 2-4 Amendment No. 57 1

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7

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POWER DISTRIBUTION LIMITS t

BASES MINIMUM CRITICAL POWER RATIO (Continued) was assumed to be constant.

The generic flow control line is used to define several core power / flow states at which to perform steady state core thermal-hydraulic evaluations.

Loop Manual and Non Loop Manual modes of operation were analyzed.

Consistent with the single failure / single operator error criterion, one loop runout was postulated for Loop Manual operation whereas two loop runout was postulated for Non Loop Manual operation.

The maximum core flow at loop runout was. assumed to be 110% of rated flow.

Peaking factors 'were selected such that the MCPR for the bundle with the least margin of safety would not decrease below 1.06.

The MCPR is established to protect the core from plant transients other p

than core flow increase including the localized rod withdrawal error event.

Core power dependent setpoints are incorporated (incremental control rod with-drawal limits) in the Rod Withdrawal Limiter (RWL) System Specification These setpoints allow greater control rod withdrawal at lower core powers (3.3.6).

where core thermal. margins are large. However,.the increased rod withdrawal requires higher initial MCPR's to assure the MCPR safety limit. Specification (2.'1.2) is.

a not violated.

The analyses that establish the power dependent MCPR require-1 ments that support the RWL system are presented in ANF report,.XN-NF-825 (P)(A),

~

Supplement 2.

For core power below 40% of RATED THERMAL POWER, where the EOC-RPT and the reactor scrams on turbine.stop valve closure and. turbine control valve fast closure are bypassed,. separate sets of MCPR limits are provided for high and low core flows to account for the significant sensitivity to initial P

core flows.

For core power above 40% of RATED THERMAL POWER, bounding power-dependent MCPR limits were developed.

The abnormal operating transients anal-yzed for single loop operation are discussed in Reference 5. -No change to the l

a MCPR operating limit is required for. single loop operation.

At THERMAL POWER levels less than or equal to 25% of RATED: THERMAL POWER, j

the reactor will be operating at minimum recirculation pump speed and the modera-j tor void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.

)

i

. GRAND GULF-UNIT 1 B 3/4 2-6 Amendment No. 57.

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBL?ES l

5.3.1 The reactor care shall contain 800 fuel assemblies.

Each fuel assembly shall contain fuel rods and water rods clad with Zircaloy cladding.

Each fuel rod shall have a design nominal active fuel length of 150 inches. The initial core loading shall have a design nominal enrichment of 1.708 weight percent U-235.

Reload fuel shall have mechanical, thermal-hydraulic and neutronic l

l characteristics compatible with the initial core loading.

l l

CONTROL R00 ASSEMBLIES 1

5.3.2 The reactor core shall contain 193 control rod assemblies, each consisting of a cruciform array of stainless steel tubes containing a design nominal 143.7 inches of boron carbide, B C, powder surrounded by a cruciform 4

shaped stainless stael sheath.

5.4 REACTORC00llflTSYSTEM DESIGN PRESSURE AND TEMPERATURE i

1 5.4.1 The reactor coolant system is designed and shall be maintained:

)

i In accordance with the code requirements specified ir. Section 5.2 a.

of the FSAR, with allowance for normal degradation pursuant to the f

applicable Surveillance Requirements, i

l b.

For a pressure of:

1.

1250 psig on the suction side of the ret'eculation pump.

2.

1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.

3.

1550 psig from the discharge shutoff valve to the jet pumps.

c.

For a temperature of 575'F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,000 cubic feet at a nominal T,y, of 533*F.

l l

GRAND GULF-UNIT 1 5-5 Amendment No. 57

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ML20236C408

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