ML20043F029

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Proposed Tech Specs Re Cycle 5 Reload for Facility
ML20043F029
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 06/08/1990
From:
ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20043F017 List:
References
NUDOCS 9006140106
Download: ML20043F029 (45)
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Text

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DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition removal, relocation or movement of fuel, sources,incoreinstrumentsorreactIvitycontrolswithinthereactor pressure vessel with the vessel head removed and fuel in the vessel.

Normal movement of the SRMs. IRMs LPRMs, TIPS, or special movable detectors is not

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considered to be CORE ALTERATION.

Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

CRITICAL POWER RATIO

1. 8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in th) ]

assembly which is calculated by application of the M48torrelation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotoaic mixture of I-131, I-132, 1-133, 1-134, and 1-135 actually present.

The t1yroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

DRYWELL INTEGRITY 1.10 DRYWELL INTEGRITY shall exist when:

a.

All drywell penetrations required to be closed during accident conditions are either:

1.

Capable.of being closed by an OPERABLE drywell automatic isolation, system, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.4-1 of Specification 3.6.4.

b.

The drywell equipment hatch is closed and sealed, c.

The drywell airlock is in compliance with the requirements of Specification 3.6.2.3.

d.

The drywell leakage rates are within the limits of Specification 3.6.2.2.

e.

The suppression pool is in compliance with the requirements of Specification 3.6.3.1.

f.

The sealing mechanism associated with each drywell penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

9006140106 900608 ADOCK0500g6 DR GRAND GULF-UNIT 1 1-2 Amendment No. 35, l

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2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of

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rated flow.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel 1

steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than he during h.th wo loop operr C on,with th: r:::t:r ;;;;:1, :t::: i,':: pr:::;r: ;;r::t:r th:

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chereactorvesselsteamdomepressuregreaterthan785psigand core flow gi,..cer than 10% of rated flow l

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APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

i ACTION:

With MCPR less than the above limits and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUT 00WN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specifi-cation 6.7.1.

REACTOR COOLANT SYSTEM PRES $URE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel i

steam dome, shall not exceed 1325 psig.

1 APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

i With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

GRAND GULF-UNIT 1 2-1 Amendment No.16, _

1

A l 90/or l

2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Because fuel damage is not directly observable, i

a step-back approach is used to establish a Safety Limit for the MCPR.

MCPR i

greater than the applicable Safety Limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.

The integrity of this cladding barrier is related j

to its relative freedom frc. perforations or cracking.

Although some corrosion i

or use related cracking may occur during the life of the cladding, fission j

product migration from this source is incrementally cumulative and continuously measuraole.

Fuel cladding perforations, however can result from thermal stresses which occur from reactor operation signIficantly above design condi-tions and the Limiting Safety System Settings. While fission product migration 1

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 Safety Limit is defined with a MCPR margin to the conditions which would produce onset of transition boiling,ition of 1.0.

These conditions represent a significant departure from the cond intended by design for planned operation.

l 2.1.1 THERMAL POWER, low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calcula-tions at pressures below 785 psig or core flows less than 10% of rated flow.

Therefore, the fuel cladding integrity Safety Limitt46 established by other L

Q means.

Thin e done by establishing a limiting condition on core THERMAL POWER with the foe owing basis.

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi.

Analyses show that with a bundle flow of 28 x 103 lbs/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 greater than 28 x 102 lbs/hr.

Full scale 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 of more than 50% of RATED THERMAL POWER.

Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

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GRAND GULF-UNIT 1 B 2-1 Amendment No.16, 1

A/L fo/o r 2.1 SAFETY LIMITS BASES THERMAL POWER. Low P_ressure or Low Flow (Continued)

AM The Advanced Nuclear Fuels Corporation (ANF) M-Meritical power correla-ggg tion is ap)licable to the -i::d?cora ':staning ;ith :ycle 2.

The applicable y d range of tie W Fecorrelation is fo d ressures.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 ggwdevele 1 operation based on ATLAS test dat#.t Overall, be e of the design g pE thermal-hydraulic compatibility of the ANF-4w& fuel desi th the cycle 1 fuel, this justification and the associated low pressure a low flow limits B a-8 remain applicable for future cycles of cores containing these fuel designs.

Withregardtothelowflowrange,thecorkbypassregionwillbeflooded l

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 gg flux of oreater than 0.25 M1bs/hr-f t2 for all fuel assemblies which can approach critical heat flux.

Therefore, theN critical power correlation is appro-l priate for flows greater than 10% RATED CORE FLOW.

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

Since the OANFS M Hcorrelation is applicable at any pressure greater than 585 psig, the cycle 1 low pressure boundary of 785 psig remains valid for then M t __ _

correlation.

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SAFETY LIMITS i

l i

BASES l

l 2.1.2 THERMAL POWER, High Pressure and High Flow 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-rectlyobservableparameterinanoperatIngreactor.

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 onset 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).

The Safety Limit MCPR assures sufficient conservatism such that, in the 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 mon'toring the core operating statd One specific uncer m-tainty included in the safety limit is the uncertainty inherent in the4N4 MfMJ C(p) s critical power correlation.

ANF report XN-NF-524 %), Rev.rt. P ^^ " W -

%;r ";th:ti;;y ';r uilir,; W;t:r ";;;ttr;," 5.10% describes g Criti;;'

the methodology used in determining the Safety Limit MCPR.

OAHfB ThekMl4. 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-%)h. radial power factors and bounding flat local peaking distributions are l

used to estimate the number of rods in boiling transition.

Still further con-Oypg servatism is induced by the tendency of theeXN4 correlation to overpr l

number of rods in boiling transition.

These conservatisms and the inherent accuracy of th-d "-3 correlation provide assurance that during sustained opera-l tion at the Safety Limit MCPR there would be essentially no transition boiling

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in the core.

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REACTIVITY CONTR0L SYSTEMS ROD PATTERN CONTROL SYSTEM LIMITING CONDITION FOR OPERATION l

3.1.4.2 The rod pattern control system (RPCS) shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2*#

ACTION a.

With the RPCS inoperable or with the requirements of ACTION b, below, not satisfied and with:

1.

THERMAL POWER less than or equal to the Low Power Setpoint, control rod movement shall not be permitted, except by a scram.

2.

THERMAL POWER greater than the Low Power Setpoint, control rod withdrawal shall not be permitted.

b.

OPERABLE control rod movement may continue by bypassing control rod (s) in the RPCS** provided that:

i 1.

With one control rod inoperable due to being immovable, as a result of excessive friction or mechanical interference, or known to be untrippable this inoperable control rod may bebypassedintherodactloncontrpisystem(RACS)provided that the SHUTDOWN MARGIN has been determined to be equal to L

or greater than required by Specification 3.1.1.

2.

With up to eight control rods inolerable for causes other than addressed in ACTION b.1, above, tiese inoperable control rods l

may be bypassed in the RACS provided that:

'a)

The control rod (s) to be bypassed is inserted and the directional control valves are disarmed either:

1)

Electrically, or 2)

Hydrauiicallybyclosingthedrivewaterandexhaust water isolation valves.

  • b)

All inoperable control rods are separated from all other inoperable control rods by at least two control cells in all directions.

c)

There are not more than 3 inoperable control rods in any RPCS group.

3.

Control rods may be bypassed in the Rod Action Control System i

(RACS) at any time.

However, if THERMAL POWER is less than o I

equal toti W of RATED THERMAL POWER:

"See Special Test Exception 3.10.2

  1. Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RPCS prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
    • Bypassing control rod (s) in the RPCS shall be performed under administrative control.

GRAND GULF-UNIT 1 3/4 1-16 Amendment No. 20, l

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REACTIVITY CONTROL SYSTEMS.

BASES CONTROL R005 (Continued) l Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position fea-ture provides the only positive means of determining that a rod is properly j

coupled and therefore this' check must be performed prior to achieving criti-cality after completing CORE ALTERATIONS that could have affected the control J

rod coupling integrity.

The subsequent check is performed as a backup to the initial demonstration.

In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a con-

. trol rod to less than 3 inches in the event of a housing failure.

The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any dam-age to the primary coolant, system.

The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

3/4.1.4 CONTROL kOD PROGRAM CONTROLS The rod withdrawal limiter system input power signal orginates from the first stage turbine pressure.

When operating with the steam bypass valves open, this signal indicates a core power level which is less than the true

, core power.

Consequently,'near the low power setpoint and high power setpoint of the rod pattern contro1' system, the potential exists for nonconservative control rod withdrawals.

Therefore, when operating at a sufficiently high power level, there is a small probability of violating fuel Safety Limits dur-ing a licensing basis rod withdrawal error transient. To ensure that fuel i

Safety Limits are not violated, this specification prohibits control rod with-drawal when a biased power signal exists and core power exceeds the specified level.

Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident.

The specified sequences are characterized by f o ),

homoceneous scattered patterns of control rod withdrawal. When THERMAL POWER is greater than t M of RATED THERMAL POWER, there is no possible rod worth l

e which, if dropped at the design rate of. the velocity limiter, could result in a peak enthalpy of 280 cal /gm.

Thus requiring the rod pattern controller

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function to be OPERABLE when THERMAL POWER is less than or equal toJ05 of l

RATED THERMAL POWER prcvides adequate control.

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

20,._j l

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REACTIVITY CONTROL SYSTEMS

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BASES

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CONTROL ROD PROGRAM CONTROLS (Continued)

The RPCS provides automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

A rod is out of sequence if it does not meet the criteria of the Banked Position Withdrawal Sequence 4as described l

in the FSAR.

The RPCS function is allowed to be bypassed in the' Rod Action Control System (RACS) if necessary, for example, to insert an inoperable con-trol rod, return an out-of-sequence control rod to the proper in-sequence position or move an in-sequence control rod to another in-sequence position.

The requirement that a second qualified individual verify such bypassing and positioning of control rods ensures that the bases for RPCS limitations are not exceeded.

In addition, if THERMAL POWER is below the low power setpoint, additional restrictions are provided when bypassing control rods to ensure l

operation at all times within the basis of the control rod drop accident analysis. Q Thebanalysis of the rod drop accident is presented in Section 15.4 of the FSAR and the technioues of the analysis are presented in - '-

Reference y.

t: - - - - ' - - - " ^-'-- ---- ' - " '

4 The RPCS is also designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during higher power operation.

A dual channel system is provided that, above the low power setpoint, restricts the withdrawal distances of all non peripheral control rods.

This_

restriction is greatest at highest power levels.

~

f 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM Q'/f,l-T, I N.,j, / 6hk The standby liquid control system provides a backup capability for bring-ing the reactor from full power to a cold, xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern.

To meet this objective it is necessary to inject a quantity of boron which produces p,g concentration of 660 ppm in the reactor core in approximately 90 to 120 min-gpsg_}gjygp3ggjg}jj. g j g, g yg g g g yn}gggyj_A 4 i .._..m..m... _m..._........,,. There is an additional allowance of 165 ppe in the reactor core to account for. imperfect mixing and leakage. The time requirement was selected to override the reactivity insertion rate due to cooldown following the xenon poison peak and the required pumping rate is 41.2 gpm. The minimum storage volume of th : & W r is established to allow for the portion below the pump suction that cannot be inserted. he tempera-ture requirement is necessary to ensure that the sodium pentab rat emains in

solution, x c e r-g.

x 1. TNatone, R. C. Stirn and J. A. Woolley, " Rod Drop Ac nalysis for Large BWR4 " G. E. Topical Report NEDO-10 rch 1972 2. C. J. Paone, R. d-R g, Supplement 1 to NED0-10527, l July 1972 3. J. M. Paone and R. C. Stirn, Addendum ed Cores," ement 2 to NED0-10527, January 1973 GRAND GULF-UNIT 1 B 3/4 1-4 Amendment No. 41, 1

m-- l A/L Mc/o f" i INSERT A to Pane B 3/4 1-4 To meet the 3% shutdown requirement, the minimum required solution i concentration at the design volume of 4530 gallons is 14.4 weight percent. In order to establish this minimum concentration, it is necessary to have a l sinimum weight of $803 pounds of sodium pentaborate. i INSERT B to Pane 3 3/4 1-4 The sodium pentaborate solution is required to be maintained above the minimum required concentration and below the maximum allowable concentration on Figure 3.1.5-2. Th e s e sh s ens h paj e /3 3/V t y we+e p e r v o'o ut ig su bdite.f y b A &c bi-9 */ o n t a.. p 4 I i + ,e" w,, w e -,ew-.- .,,,,m

NL 9o/cr REACTIVITY CONTROL SYSTEMS i BASES STANDBY LIQUID CONTROL SYSTEM (Continued) With redundant pumps and explosive injection valves and with a highly { reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable. Relief valves are provided on the SLCS pump discharge piping to protect the SLCS pump and piping from overpressure conditions. Testing of the relief valve setpoint and verifying that the relief valve does not open during steady state operation of the SLCS pumps demonstrates OPERABILITY of the relief valve. The relief valves are ASME Class 2 valves and, as such, have a i 3% tolerance in the opening pressure from the set pressure, per the ASME Code (Section III - Division 1 Subsection NC-7614.2(b), 1974 Edition). Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours assures that the solution is available for use. Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges. Compliance with the NRC ATWS Rule 10CFR50.62 has been demontirated by means of the equivalent control capacity concept using the plant specific minimum parameteis. This concept requires that each boiling water reactor must have a standby liquid control system with a minimum flow capacity and baron content equivalent in control capacity to 86 gpm for 13% weight sodium pentaborate solution (natural boron enrichment)4. The described minimum sys-l used for the 251-inch diameter reactor vessel studied in NEDE-24222. Reference g tem parameters (82.4 gpm, 13.6% weight with natural boron enrichment) provides an equivalent control capacity to the 10CFR 50.62 requirement. The techniques of the analysis are presented in a licensing topical report NEDE-31096-P, Oy Referenced. Only one subsystem is needed to fulfill the system design basis, and two subsystems are needed to fulfill ATWS rule requirements. An SLCS subsystem consists of the storage tank, one divisional pump, explosive type valve, and associated controls, and other valves, piping, instrumentation, and controls necessary to prepare and inject neutron absorbing solution into the reactor. k "( W. % d #aa.e4 I (4 p e A e," d 6 7.p R C. 7 P.. a e/ b o 3 n.,k e s /.e.' t *. a rep r. M v t v,.r- . n i ny. 1979," Assessment of BWR Mitigatioii of ATWS, Volume II," NEDE-24222, December h L. B. Claasen et al., " Anticipated Transients Without Scram, Response to l NRC ATWS Rule 10CFR50.62," G. E. Licensing Topical Report prepared for the BWR Owners' Group, NEDE-31096-P, December 1985. GRAND GULF-UNIT 1 B 3/4 1-4a Amendment No. 41, _.l

A L 9t/or l i 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE i LIMITING CONDITION FOR OPERATION (D 3.2.1 During two loop operation <all AVERAGE PLANAR LINEAR HEAT GENERATION l RATES (APLHGRs) for each type of fuel as a f shallnotexceedthelimitsshowninFigure\\unctionofAVERAGEPLANAREXPOSURE 3.2.1-12, 3.2.1-15, 3.2.1-1:,p 3.2.1-id, Or 3.2.1-h :: iti;1i;d be th: :=ihr ;f ith:r th: =-d:;:nt:nt ""FL"0". fu t:r ("*PT".0 ) Of Ti;;r; 3.2.1-2, er the p;.;;r-d;;;; dent ".^.FL = 7 ft:ter ("^ar'.O ) f rig"re ?.?.1-2. p During single loop operation, the APLHGR for each type of fuel as a function of AVhRAGE PLANAR EXPOSURE shall not exceed th: '!:it: = d:t:=f =d 5:h. .-+e. multiplied by if:n \\ the limit shown in Figure 3.2.1-1

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=il;r ;f ;ither "'.FT'.0

, "*Fr'O er 0.00. f p APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. _CTION: A Duringtwoloopoperationor$singleloopoperation,withanAPLHGRexceedin thelimitsofFigure\\3.2.1-

3. 2.1-h, 3. 2.1-lb, 3. 2.1-k, 3. 2.1-1d er 3.2.1--h-as corrected by the appropriate multiplication factor.f= =9 t=:
f f;;'

initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. I 1 SURVEILLANCE REQUIREMENTS l 4.2,1 All APLHGRs shall be verified to be equal to or less than the required limits: a. At least once per 24 hours, b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and 1 l 1 c. Initially and at least once per 12 hours when the reactor is i L operating with a LIMITING CONTROL ROD PATTERN for APLHGR. 1 l l 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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go.oii44 ar . ~ teFUEL i =4 t I t t g . -- - _ -_I__.-__.-_.-. t 9.0.12 4 g0.o.12.4 l t i i t I-i as4 FUEL 1 l s i a ii m---~--------- - - - - - - d - :-- --- o 5 3 i' n_ h u l 1 es.o.s.cl i m, .___i_ - _ _ _ _.. + - - _ _ _.. . ~ - _ _ i ~ t pe.o.7J) 1. I i 7 o io a m ao - so so AVERAGE PLANAR EXPOSURE (GWdMT) N1 FIGURE 3.2.1-1 MAPLHGR vs AVERAGE PLANAR EXPOSURE FOR ANF FUEL 2= e+ I e g --,. ~,, - ,f ~ y [,~ q,, ,,.W.e-ve~. ..____________,_______,_____._.,m_z__.____. .___m_______,m.m.

4+ + gg Q. O b E .,a G i j i i ' i' 'i-l i I l ). 3 i I w f 14 .3 -i x. t l l i x-l [ j / i 13 i i I 12.84 1 All 12.68 i I I 12.40 \\ i 3 12< ~ I to d5 12.es I i i y k4 g Sl I i i / l l 1s.44 1 1 l l l (N 10 Af -[ l I I i e I i ST I I l s.st 1 I, I, i, i I I I i 1 4 e i E o s 10 1s 20 25 30 ss' 40 4s so g A'terage Planar Exposwe (GWd/ST) FIGURE 3.2.1-1 MAPLHGR vs AVERAGE PLANAR EXPOSURE FOR SINGLE LOOP OPERATION,8X8 FUEL

y1 l~ . +.. .m.,. i - g

j 15

_. g: g _g g g. ] [ j. [ +- z O ~ s. g 14.2 6 ' Q r-- 34 I2 ',0 14 G. - 3 E 3 13.8 9 H-- / g 1354 13.3 0 / - 13.3 8 - y 3 05 t 20 --gio 13 13.2 0 (25 4 ZD 4 tr Wz y@ i2 ,,,y mm wh .RN o m A 3F 11 4 o 2gE 10.4 8 - )& KM E 10 i J I th i' -i 9.15 W I . 9 J m 8.61 E kr ) ih i I I i i 1 - l 1 i hI# O 5,000 10,000 15,000 20,000 25,000 30,000 35,000 ~40.000 4, is AVERAGE PLANAR EXPOSURE ( mwd / MT 1 ' z FIGURE 3.2.1-10 MAXIMUM AVER /lGE PLANAR LINEAR HEAT GE14ERATION RATE - 5 (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE i OR ANF FUEL-. g -TYPE ANF299E5G3S8 ~-,w a v ng,- 7 whan-y g - y opp,,,. y 9- -e sr ~w 4

s + l-I I I 1 I I l y g n: o 74,0 3. 7 .7 14. N (( ~ I 13.61 g G33,3,83 54 13.2 o 1232 12.85 trto 13 @Q 12.82 9-l gz i N$ i2 11.6 5 @!5 l w m@ 2 R l P 2 to a er 2 10.4 4 j C> l2% 19 % M 10 \\'N % \\ l i ii M< ~ 9.17 I, b $\\ g g s \\ R / 8.64 !p k' . I I I I .l. I I I ,O 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 000 p AVERAGE PLANAR EXPOSURE (mwd /MT1 l m FIGL.RE 3.2.1-lb MAXIMUM AVERAGE PLANAR LINEAR HEAT GEfERATION RATE I (MAPI_HGR) VERSUS AVERAGE PLANAR EXPOSURE FOR ANF FUEL TYPE ANF32tE6G4S8 M i f-- - r

,.-.s a ^ 15 .g g g g g g l U 4.. Q t C l . l4,0 T 14 13.8 7 i 3 c ^93 l 2 %.s 13.4 2 - q h. 33 S 13 1309 13.0 131u 13.9I v g w 13 i g ~ I2.98 ZD 13.0 <t m - d2 O w o Q 12 yg Il.75 - 68 wm $5 to CY O lo q\\ 3Q 11 a 1 S 2w A-RI 10.4 6 N i h* 's% h io r 1 y'*)h l I 9.18 i D 9 8.64 >E" 1 1 1 I .1. I 1 -1 h O 5,000 .10,000 15,000 20,000 25,000 30,000 35,000 40,000 KS,000 (T) AVERAGE R.ANAR EXPOSURE (mwd /MT) N D FIGURE 3.2.1-Ic MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLAN AR EXPOSURE FOR ANF-FUEL b TYPE ANF321E8G4SS e .~ ~ .,__,.a

.z.,

.~ 1

m
g 4

O 15 3 r i l i N. i i O T I i g .I i i l 1 i ~Z l \\ I3 II 14 13.84 r 13.8 13.30 12.94 13 33 13 g l13.93 ['jy,gg l I i i 12.98 i i O l f h E I N i 33:12 + / l w fh as o i i i n.n l e 1 W% i i o l ). i z l T i ~ J 11 I g 10 g, f+- N x i . f I l l l i iir l i i I l = ,~ i l l l I 1 8 I I O 5 10 .15 20 25 30 35 40 45 50 g . Average' Planar Exposure (GWd/MT) m FIGURE 3.2.1-1d MAPLHGR vs AVERAGE PLANAR EXPOSURE l ' FOR ANF FUEL TYPE ANF361E8GZS8

= g g a 4 i i 5 l i \\ } i H N I I / 14 _3 / i '8 7 I l l / I I l-1 l i l I g tr.is l 1 -i s 3 12 t Il Is i g l j11.99 ty ti.es 11.5s, l i 1 m2 g g ti.ri

  • l I

l ? NS 3 h $s' ','.,',' 11.14 '/ l \\ i I 2' ,F (f i is.,s l ~ 4 l l [ l 1s.ir ja l 1 g i ~n I l t l k i) / I t i ,F j/ l l \\ s. 3 .M / l N = E / \\ y s to is to 2s so ss 40 as F Average Planar Exposure (GWd/MT) N 5 FIGURE 3.2.1-1e MAPLHGR vs AVERAGE PLANAR EXPOSURE FOR ANF 9x9-5 LEAD TEST ASSEMBLY ~ 4 ~._ . ~, - .v_ --4 ..3,r~

ap =.

g.,

?;, ? I -[_.,, l l l y s 4 =o g-g 1.10 g b l h 5 1 = i 1.00-N+ I I s j i N O.90 LOOP MANUAL.- D_ an. [ t I T i j 6 \\ l i 1 0.80 i r { .q l O 1 DNE LOOP OPERATION p y ta-sus SLO / g l m I l 4 ; p ;- /- s_<_a_L...<, 3 i s q 0.s0 7 7 s i i i i i F .l i I O.50 g /0 20 40 SO 80 100-12 3 / CORE FLOW (X RATED)- l E O FIGURE 3.2.1-2 MAPFACf .. =

All. 90/ot~' g. 4.4 N / 1.0 \\ / / i N / / L .'N / / [ l 0.8 \\ / ,, m, 4

  • OR 25 %C P440%;

DLRDG ole LDCP CPDMTION CORE FLOW F $.90% l l g g<h'00%) \\ I 0.7 n / / N ~ 0.6 1 \\ N POR 5%IPI40%; \\ 0.5 CORE FLOW F > SC% N g j / \\ / \\T .c. /

  • I 0.4 0

to 40 60 80 40 0 120 CORE TPDthEL PCNER (% RATED)P b E L E rf ficu Af !I FIGURE 3.2.1-3 MAPFAC P GRAND GULF-UNIT 1 3/4 2-3b Amendment No. 39 l .c-, - - ~ - -,

Al L. 90/o f~ ' POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO t LIMITING CONDITION FOR OPERATION (M[{MMAp, NCf ed *U Mc P Ah ~ 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than' h th ""PP d "CP" limits at indicated core flo one THERMAL POWE as shown in FigurTs=3.2.3 &ene 3.2.3-2 F Wsh.'s D APPLICABILITY: OPERATIONAL CONDITIO 1, when THERMAL POWER is greater than or equal to 25fof RATED THERMAL POWER. ACTION: l With MCPR less than the applicable MCPR limits determined from Figures 3.2.3-1 = and 3.2.3-2,. initiate corrective action within 15 minutes and restore MCPR to l within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED _ THERMAL POWER within the next 4 hours. ['1 M 3 SURVEILLANCE REQUIREMENTS 4.2.3 MCPR shall be determined to be equal to or greater than the applicable MCPR limits determined from Figures 3.2.3-1 3.2.3 a. At least once per 24 hours, b.- Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL R0D PATTERN for MCPR. d. The provisions of Specification 4.0.4 are not applicable. 4 l l I' GRAND GULF-UNIT 1 3/4 2-4 A m e s-< J Ale.

.t q-r- NL T */o f" .t -t n R l f [ l 4 r y. j i. i t g f l ..l ) t i i g t 1 ,L 3 i i j i t _ _ _. + _ _ i ._ 4 _. _q _,u g l I 4 i-l I I i i g .f- -._._..f__..-__[_...,..-._.j...._,_ g* Ed ,i g n 4 W .\\. j 2 s. .._...__.1, d 4._. _ _. g g l l l g .1~ i i i l i i I t a y-. e--- _.._m... -L-._.. g i i i t i I i l i j o t a m m vi t n n I WdOW GPAND GULF-UNIT 1 3/4 2-5 Amendment No. l .m i.'d - 4 

+

c3 g w o Si 4 ) t e .so \\ . I. I o l 1-I \\ !,v 1.80 + g g ~ l \\h / 1.50 I l b I l w1 k g I ~ ? ( )% N_0 K-L OO P MANUAL MODE' + i i \\g( E a i. LOOPNANUAL[ MODE i-I ~ E' 4 ~ N I i W 120 \\

\\g f

l / t / I g % ( 1.10 x L r+ i RATED WCPR OPERATINC LINIT = 1.18 g l i l' i 0 20 40 SO SO. 100 12 CORE FLOW (I RATED) [ FIGURE 3.2.3-1 'NCPR 9 =..

Y i i ( pt g o/of-k- I 4 i i i t (' j 1 l I i 1 i \\' i \\ g _.___p.____. i i i j i -.._._..._..._____9-7 g _= I I a l l l l! 4 w g -f ._{ n l i ol A l n i y) y I 5 i i i-g .... }.___ l l I I N! l ) h .i 4 i l, w i - iu.._ _._ g .._. _ y _ _, i 1 i i t 4 h l i I N 4 4 .T N w D I.' y d WdOW l. GAAND GULF.UN!T 1 3/4 2 6 g,,,g,,,g g,, _g .I 7 I ~~ a

W ~ .g g a %.I 2.40 a g i 3 Q l THERMAL' POWER.251 < P < 40I 2.20 CORE FLOW > 50I THERMAL; POWER 25I <P < 40 \\\\ CORE-fLOWf50! 5 I i 2M s .t f I i b x i I D ns i I i () b A-R, MAL POWER 401 <P<

702, 1.00 ORE _ NOWS

'\\ ( l YHERMAL POWER P >' 701 y l

  • \\

f f I fgp l [^ ALL CORE FLOWS w / N l 1.20 F i i i i S I' 0 20 40 SO 80 100 12 = P CORE THERMAL POWER (Z RATED) I .~ FIGURE 3.2.3-2 MCPRp

o... 4 b e... . w l. 4 0 l .N s l 1.5 - 3 7 l E 2 1.4 1.3 1.29 of O' 0-E w2 1.2 1 19 ? e P i 1.1 4 4 1.0 i i i [ eoc Eoc-sooonanour Eoc necomuu ocmsino c cycteexposuRe a E FIGURE 3.2.3 3 MCPR, 4 -ge-, +c ,m,, u n ._e,e_____w___a-

& L ' 90/W POWER DISTRIBUTION LIMITS 3/4.2.4: LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the limits shown inFigure3.2.4-( l } 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.h nitiate l' i corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER i within the next 4 hours. -e k " I I r ' * * " b ' k I l as cs,-e n te.J by fAe aff < i-e _ _"p r e' 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 hours, b. Within 12 hours after completion of.a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, c. Initially and at least once per 12 hours wnen the reactor is operating on a LIMITING CONTROL R00 PATTERN for LHGR, and d. The provisions of Specification 4.0.4 are not applicable, y__ euHyhed 6 tse sm. flee at' e:+4e+ +Ae $lo w-as 7 depssde-t LNG /t 4k do e (L HG/2 /=r4 Cp.) 4 foy+ e

3. 2.4 - 2 o v-6e p o v e< - depes de s t L H G /2 4% e +. s c~c m )
  1. y 3.y GRAND GULF-UNIT 1 3/4 2-7 Amendment No. 57,

l' N 4-6 o o p 18 F i -i ~ ? I e j 88 '** J-lesFuet w t -- j - g talk 4.14.1) 14

- t 98.111) l (i n 1 11)

{ i _. _.____=_._.7 O med FUEL / { 3 ~ .l w gg

  • t u

'9 eso.no -- ---- -+- - - - --- I a i I L i

  • Ono.nen e

i i I' O 10 20 .M 40 50 00 AVERAGE PLANAR EXPOSURE (GWd/MT) FIGURE 3.2.4-1 LHGR vs AVERAGE PLANAR EXPOSURE FOR ANF FUEL 3 a ? ^ b e t 4 -,s .r., .m-- ._s. ~ s

g 1 n a w a

4 lE N

I I/ N-( r = N l 1 i I U N I / i 'l l x(e..is.s) x a Y,H" ' ".') i 3,s F I I ,, to..u.i) g,3,3,,, I l j N '/ l j E12 N )} k i 9x9-5 fuel,,/! f .i x i i i i i ~ \\ kk k{ E \\ (" 8'43 N I 5 /[ 2 l [4 .a /l l l i i I l L' / I I t< i / I i l l l I } 2 I i 2 ? es\\ o s to 15 20 2s so as 40 4s so ss so Average Planar Exposure (GWd/ BIT) N FIGURE 3.2.4-1 LINEAR HEAT GENERATION RATE (LHGR) LMIIT VERSUS AVERAGE PLANAR EXPOSURE FOR ANF FUEL' .. ~..

s, pt pojor ,1 1 i I l l l i. ._ _..._ _j _. _ _... [; _.-_.._ _. _1.___..___. g l { i i i - i != 4 .i . y -.-- g b / i l 3 8' { + i .d I g i i l I l R I' t I i h. l^ 1 l t O 3 3 2 3 a 1: I OYdWOH1 GRAND GULF-UNIT 1 3/42*7b Amendment No, l

=. -

'^ e :, ^ s ur- '+ 4 .l' g-o N t. '.T 1.1 l T .E. i M j .~4 ---4..-. -.--~ ~j-I a _. _[' _.. _- l O gg _.__._L__ 4.u_..__ COFEFU)W h 50%-' i I tae m.COnenows O ( b &7 ---?- M e M O - gg __.----t--~- I COfEROW > 50% og -J a4 0 20 40 e 80 100 12 CORE N N M M W .F RGURE 3.2.44 LHGMAC = P 3e* 2 .O 1 e . ~, -.e ye e pw ,s. -.y., a ,m.'- -, - -. + - w: -v -~n w-,4 ,,w= eg-w-e m,.+-.-w. y ~ -- + w Jw--

-.-- -~ E. .. j ' A)/. ;4 o/o S* 3/4.2 POWER OISTRIBUTION LIMITS BASES s 1 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. 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 specified in 10 CFR 50.46. The peak cladding temperature (PCT) following a postulated loss-of-coolant s -accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondar- -__ily~on~the rod to rod power distribution,.w.ithin an assembly. Th p-d '.t d +- ys....__

4..,..= 1.- i.1.

. J... _2m = = l UP n .k. ki L... -..._m.A m.J ok4 k 1, ..i.. ..y r..'".'..'""A..WJ.'=.,.I.,.I.."4.." r'"- " ' ' k "l.,' ... a. i. 1. ,,. k.. .k. A.., 4... l U00 ..6..

f...

Y f ue n r y.. n. &. J. m_w._.. A.9

2...

2. .. J .k. k...... .A. .i..,. utek .k. .u..... A....A... ,6..Au i v. i. .'.A. * ' '.' A. & n. e n d---- =-a =--r , J..m

a. l..m -,A,U..E B A P E 1neal
n..' ir {'na"f me e n u.- ~r - - - ' T" ha

--r -' --'4 e m 1 Cnar 4. T.ehn - - - -, et.... ..A..e... mi .... wr - - - - - - - - - - -n i A W A n' i f u..f.A..n,u..f.AT.---er' uf D A.T. T A.u. r---- D.Ar ' A D.l ue n \\, - - - - 1r r-- t /.......

4. e. k. 4.e l ut'D' s

u....._u..._ u... 2 4.. a. 2. s k,.,

4..,
u..,.,

..u,.,... The aximum Aver. ...__J __J a r...... i age,,Pl,anar,L,ine,_ar Heat, Gene,_ ration Ra,te (MAPLHG,R) limits of Fi u,rek 3.2.1-1g i ,u e,,_ .2 ,e,... , _,_ __ _.. s.

2. r_.._ Jv,..

___s.s.__ .m ,w ... i, [ . #.. {. k..... k..

  1. . 3..u. A. m.,. = J *..

/ M,.A.n f. n.,, \\, A.L.. -. _ _ MAniUPn ,..e AP J ___J .y.....-.a. ...................... s .i y... ua nt h s.n n \\uun PAP wvu4 w.yv...J1.._.y.

6. v.

w e I A. J. m.y. _ J. _ _.. _ _

  • e. &.
6. w. --

I u ,__A_._ ruan \\.___ ____ m2 1 __ ,1... i auns wn #uwbyi Inw / wwa w q u vv. i.. yv.w. e1 ...........u...,,,.Ak.._..... - y_ wr ........ u.. _<..a...k. 4.., i. A..

4.,

-_... a... :.. u.. it. 4 4. k .x._... .oy -.. i -. .y G.,.e m-ne.A sle --& M I w= m @.m*n.,, I . -.... - ~.. P For sinale-loo) o ) era tion.;' t.'. ^"" 4 f d. a MAPLHGR limit corresponding to the_ product of tie 1APLHGR, Figure 3.2.1-1, and".ti.

.p;,repriet; ",'^"'" can A8 be conservatively used T. k.

.mi.i. mm.k u.. u Ant ur.n...'..-. c 4.... w

2. -

e a .o .w.. .. u.. u... J. .J.. _ 2. _ P.y... u. a. . n., n_n ,.. a. .-..J u.i.w r.y...,. u.e e,i n .....i.. .. i..y .i .y y. i s* . t.w ' u A n.i UP n /D,.....- T. k.w. man.iUPn i. 2.. 4.. .....f C\\ ,_ Au nun _E , i.i.1

2..
e. k.

r....-6 J.. ........ se..... ~. s .k. I .... M A.n. i.u.t* n

4.. E f.....

,k u-e .__.1-- v. 4 1. _1....J &. L.w. ---- - J._ A.w. u A h.r A.P T L.w-wn... -. u i iy. i.. i .yy..y s s.u n. ei w nim.m 1-__ ,s 3 M A n E A.P..J _4. -. J..y 1. w. .y - _i s. e. f. _ A.ww i.i i 1 a.y.. -.. i.u .r i .y . s avvy m J. of,_ 11 s via s. v i vs e wa ys AP 0, ,M.A. n f. m 7. J.w. &.._.. 4.._.w_.J....,f,- &. L.w* &. L. _. w. _ J J._... J.._ 1 n,u. n. J. _.... i. s. .J_ A_ .uw ..im ....i... m.. ......i.,.,,., u..m. <u...................,4...., v.. .....__.4..J,.2 k.. a. a.... i l ........i... y... _..a_.._ w w.J i u. ,u.........&.....u... ,. Aue ,..a. ,.. a u..... u.....,.. m.n i.e..m swsw a svw v.uv. ..ii. n. u ,.,..-' _ _ u..y_ u - - - - - - -- -r u__ . __a...i..., vyv...,._ y u,w_ u.. _ _ _..,..,.. us.y_ u vy i... nv o i.vvy _m. .ue . a wi.. is ,s.. v. u. aw

u.....,s.4..

v. s u.i w vs .i s- .v. ..,...... 4... 4...,. .h..................,a.,......u...-- u._.._... .m k.a .. _ 4..... u.. u... u... ....2,..

4. a. _. e m.............. a.u a.,.m u.

... i. .. u., _.. 2 _,J. n....,u...i..._..,.,............, u. u... u..,...k..... -+4 a.. .y... e..... l. i.........i1. ..,... e....

  1. i..u.

ui.n e.i.r.p, ..,,1......a-i .....,+ .k. ..k.. r----- -- ~~ - r-- u.. ,u,... u... .iv,. 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 a,re very slow when there have not been significant power or control m- _ loop _ d fa enu e +% f +4 e PCY & Je op e n W .dq/oep Pc Y fo e 1%e 6o~ dej_by +4e -- -_'- _-_~- ar e.m h' _. B 3/4 2-1 Amendment No. 57, l GRAND GULF-UNIT 1

h AU. 9 o/of" 3/4.2 POWER DISTRIBUTION LIMITS BASES .l AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) rod charv,es. The requirement to calculate APLHGR within 12 hours after the completion of a THERNAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating APLHGR after initially determining a LIMITING CONTROL R00 PATTERN exists ensures that APLHGR will be known following a change in THERMAL-POWER or power shape, that could place operation exceeding a thermal limit. The calculational procedure used to establish the APLHGR limits is based on a loss-of-coolant accident analysis. The analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. Thesemodelsaredescribedinreference\\If 0,...J 0. l 3/4.2.2 [ DELETED) r A 1 h f L I L S i L GRAND GULF-UNIT 1 B 3/4 2-2 Amendment No. 23, I

~ J pt. '9c/or j + POWER DISTRIBUTION' LIMITS BASES-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-J ding integrity Safety Limit MCPR, and an analysis of abnormal operational tran-sients.- For any abnormal operating transient analysis evaluation with the J - initial' condition of the reactor being at the steady state operating limit it isrequiredthattheresultingMCPRdoesnotdecreasebelowtheSafetyLimIt .MCPR at any time during the transient assuming instrument trip setting given in Specification'2.2. To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal opeaational transient, the most limiting tran-sients have been analyzed to deter::,ine which result in the largest reduction l in CRITICAL POWER RATIO (CPR. The type of transients evaluated were loss of flow, increase in pressire an)d power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta CPR. When added to the Safety Limit MCPR, the required operating limit MCPR of Specification 3.2.3 is obtained. The power-flow map of Figure B 3/4 2.3-1

defines the analytical basis for generation of the MCPR operating limits p g ("cfr::: 9 ).

( ae fed ces ~z.. 7p .g. The ' purpose of the MCPR and MCPR is to define operating limits at other ~ f p -- -: = l than_ rated core flow and power conditions. c,p ey== y 4.Ilexe m n he l +se The.MCPR s are established to protect the, core from inadvertent core flow 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. L: during loop Manual operation is the runout of one loop because the two recirculation loops are under independent control. Runout of both loops is l 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 assuined to be in equilibrium, and (b) core xenon concentration GRAN0' GULF-UNIT 1 B 3/4 2-4 Amendment No. 57,

t WL ;9 o/*f l E INSERT "B" MCPR' operating. limits are defined as functions of exposure (MCPR,), flow (MCPR ), and power (MCPR ). The limit to be used at a given operating 7 p state is the highest of-these three limits. The purpose of-the MCPR, is to define operating limits for all anticipated exposures during the Cycle. The MCPR, limits are established for a set of exposure intervals. The limiting transients are analyzed at .the limiting exposure for each interval. The MCPR, operating limits are established based on the largest delta-CPR calculated at the limiting exposure and ensure that the MCPR safety limit - will not be exceeded during the most limiting transient in.each of.the. exposure intervals. L s i

l-on guespu m I4 NI I I I!M-J1(10 QNYNS PERECENT OF RATED THERMAL POWER t 8 8 8 8 8 8-3 3 .g g g_ i i i i i i i i i i 4e l' 3-g

e x NT>

4,- ?- LIl '\\ \\ \\o $a i s llI ia-8g. Il e j g. g g, 5 I io now m. g \\ in now m. L g l 1 gofos ytv

9 */* f E = 3Ni1 mold % sol N \\ g B-s 5 I 1, Ys 3-k ~ 4-bh R = g

  1. o, I

I 8 glg g e e El g ! li ,1:l I y58 \\ \\v g oxx 005

  • X \\

'\\ 8 keg ax c a i l ll al E 1 18.

f.

/a 1 i l I I I I I I o a a e s 8 8 a 2 g/ g W3 mod wntf3H1.031VW go powed GRAND GULF-UNIT 1 B 3/4 2-5 Amendment No.16 l

U A)L 90/or } J POWER DISTRIBUTION LIMITS l I 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 i l -decreasebelowA96. p pg 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 (3.3.6). These setpoints allow greater control rod withdrawal at: lower core powers 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-i not violated. The analyses that establish the. power dependent MCPR require-ments that support the RWL system are presented in '"" r:;;rt, " "" "5 (P)(A),- R;; n=t 3 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 s valve fast closure are bypassed, separate sets of MCPR limits are provided for l P high and low core flows to account for the significant sensitivity to initial l core flows. For core power above 40% of RATED THERMAL POWER, bounding power-L dependent MCPR limits were developed. The abnormal operating transients anal-yzed for single loop operation are discussed in Reference (, No change to the l-MCPR operating limit is required for single loop operation. At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, L the reactor will be operating at minimum recirculation pump speed and the modera-L tor void content will be very small. For all designated control rod patterns i which may be employed at this point, operating plant experience indicates that l the resulting MCPR value is in excess of requirements by a considerable margin. - - ~ - ().g f>gr ae c e y

  • "d

% e =p nf H< h e nle - sp e c M,'< .s, e n e - +r l GRAND GULF-UNIT 1 8 3/4 2-6 Amendment No. 57, y l

A)/. to/or ~ POWER DISTRIBUTION LINITS I

BASES MINIMUM CRITICAL POWER RATIO (Continued)

During initial start-up testing of the plant, a McPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. -The MCPR margin will thus be demonstrated such that future MCPR evaluation-t below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATE 0' THERMAL POWER is sufficient since power distribution ~ shifts are very a; slow when there have not been significant power or control rod changes. The. requirement to calculate MCPR within 12 hours after the completion-of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time'for' the power distribution to stabilize. The requirement for calculating MCPR after initially determining a LIMITING CONTROL R00 PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape, that 5 could place. operation exceeding a thermal limit. 3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in -any rod is less than the design linear heat generation even if fuel pellet l A y,, p ' .densification is postulated. [ 'c' The ' daily requirement for calculating LHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distri-bution shifts are very slow when there have not been.significant power or - control rod changes. The requirement to calculate LHGR within 12 hours after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while= still L allotting time for the power distribution to stet:llize. The requirement for p calculating LHGR after initially determining a LIMITING CONTROL R00 PATTERN exists = ensures that LHGR will-be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit. ,g:

References:

l General Electric Company Analytical Model for Loss-of-Coolant Anal ge h in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 2. [0 0) 3. [0ELETED) 4 4. [0ELETED). 5. GGNS Reactor Performan, aprovement gram, Single Loop Operation Analysis. General Electric al ort, February 1986. l 6. General Electric Company A ic odel for Loss-of-Coolant Analysis in Accordance with 10 50, Append Amendment 2, One Recircula-L tion Loop Out-of-S ce, NE00-20566-2, R fon 1, July 1978. 7. General Ele c Company, " Maximum Extended Oper ,g Domain Analysi arch 1986. 8. X 19(A), Volume 2 " Exxon Nuclear Methodology for Bo Water eactors: EXEM BWR ECCS Evaluation Model," Exxon Nuclear Compa September 1982. GRAND GULF-UNIT 1 8 3/4 2-7 Amendment No. 23, l

r NL 4 */of I INSERT 'C' l t t The LHGR limits of Figure 3.2.41 are multiplied by the smaller of either the flow dependent LHGR factor (LHGRFAC ) or the power dependent LHGR g factor (LHGRFAC ) corresponding to the existing core flow and power state p to ensure adherence to the fuel mechanical design bases during the limiting transient. LHGRFAC 's are generated to protect the core from slow flow g runout transients. Two curves are provided based on the maximum credible flow runout transient for either Loop Manual or Non Loop Manual operation. The result of a single failure or single operator error during operation in Loop Manual is the runout of only one loop becaust, both recirculation loops are under independent control. Non Loop Manual operational modes allow simcltaneous runout cf both loops because a single controller regulates core flow. LHGRFAC 's ar? cenerated to protect the core from plant p transients other than core flow increases. e i i l L L - - - -, ~. ....-,,-.v., --.n,---..--

+, i / )L. 90 f oS~ \\ s i INSERT "D" 1. XN NF 8019(A), Volume 2. " Exxon Nuclear Methodology for 8 oiling Water Reactors: EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, September 1982. 2. General Electric Company, ' Maximum Extended Operating Domain Analysis," 3 March 1986. 3. AECM 86/0066, " Final Summary Startup Test Report 12," Letter, O. D. Kingsley, MP&L, to J. N. Grace, NRC, February 1986. 4. XN NF-825(P)(A), Suppleoent 2, '8WR/6 Generic Rod Withdrawal Analysis; MCPR, for All Plant Operations Within the Extended Operation Domain," Exxon Nucloat Company, October 1986. 5. GGNS Reactor Performance Improvement Program, Single Loop Operation Analysis, General Electric Final Report, February 1986, i

1 I //l. 9 0/0 5~ 3/4.4 REACTOR COOLANT SYSTEM l 3/4.4.1 RECIRCULATION SYSTEM i RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 The reactor coolant recirculation system shall be in operation with either: Two recircuiation loops operating with limits and setpoints per a. Specifications 2.1.L 2.2.1, 3.2.1, and 3.3.6, or b. A single recirculation loop operating with: 1. A volumetric loop flow rate less than 44,600 gps, and 2. The loop recirculation flow control in the manual mode, and 3. Limitt and setpoints per Specifications 4r4re._2.2.1, 3.2.1, and 3.3.6. Operation is not permissible in Regions A, B or C as specified in Figure 3.4.1.1-1 except that operation in Region C is permissible during control rod withdrawals for startup. _APPLICABILIH: OM. RATIONAL CONDITIONS 1* and 2*. ACTION: L With no reactor coolant system recirculation loops in operation and the resctor mooe switch in the run pnsition, immediately place the reactor mode switch in the shutdown positinn. b. W0th operation in Region A as specified in Figure 3.4.1.3 1, immediately place the reactor mode switch in the shutdown position, c. With operation in regions B or C as specified in Figure 3.4.1.1-1, observe the indicated APRM, neutron flux noise level. With a r l sustained APRM neutron flux noise level greater than 10% peak-to peak of RATED THERMAL POWER, immediately place the reactor mode switch in the shutdown position. d. With operation in Region B as specified in Figure 3.4.1.1-1, immediately initiate action to either reduce THERMAL POWER by inserting control rods or increase core flow if one or more recirculation pumps are on fast speed by opening the flow control valve to within Region 0 of Figure 3.4.1.1-1 within 2 hours. e. With operation in Region C as specified in Figure 3.4.1.1-1, unless operation in this region is for control rod withdrawals during startup, immediately initiate action to either reduce THERMAL POWER or increase core flow to within Region D of Figure 3.4.1.1-1 within 2 hours. f. During single loop operation, with the volumetric loop flow rate greater than the above limit, immediately initiate corrective action to reduce flow to within the above limit within 30 minutes. "See Special Test Exception 3.10.4. GRAND GULF-UNIT 1 3/4 4-1 Amendment No. 62,

Lp g /\\)L %/W REACTOR COOLANT SYSTEM i 1 LIMITING CONDITION FOR OPERATION (Continued) g. During single loop operation, with the loop flow control not in the manual mode, place it in the manual mode within 15 minutes. i h. During single loop operation, with temperature differences exceeding the limits of SVRVEILLANCE REQUIREMENT 4.4.1.1.5, suspend the THERMAL POWER or recirculation loop flow increase. i. With a change in reactor operating conditions, from two recircula-l tion loops operatin two loop operation,g to single loop operation, or restoration of the limits and setpoints of Specifications 4r4rdL l 2.2.1, 3.2.1, and 3.3.6 shall be implemented within 8 hours or declare the associated equipment inoperable (or the limits to be "not satisfied"), and take the ACTIONS required by the referenced specifications. SVRVEILLANCE REQUIREMENTS 4.4.1.1.1 At least once per 24 hours, the reactor coolant recirculation system shall be verified to be in operation and not in Regions A, B or C as specified in Figure 3.4.1.1-1 except that operation in Region C is permissible during control rod withdrawals for startup. l 4.4.1.1.2 Each reactor coolant system recirculation loop flow control valve in an operating loop shall be demonstrated OPERABLE at least once per 18 months t by: Verifying that the control valve fails "as is" on loss of hydraulic a. pressure at the hydraulic unit, and i b. Verifying that the average rate of control valve movement is: 1. Less than or equal to 11% of stroke per second opening, and 2. Less than or equal to 11% of stroke per second closing. 4.4.1.1.3 During single loop operation, verify that the loop recirculation flow control in the operating loop is in the manual mode at least once per 8 hours. 4.4.1.1.4 During single loop operation, verify that the volumetric loop flow rate of the loop in operation is within the limit at least once per 24 hours. GRAND GULF-UNIT 1 3/4 4-la Amendment No. 62,__ l

t vm A)L 90/oS~ i 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor core coolant recirculation loop inoperable has j been evaluated and found to remain within design limits and safety margins pro-vided certain limits and setpoints are modified. The "GGNS Single Loop Opera-tiori Analysis" identified the fuel & dig St:;Hty hf:ty Li it, "?L"C" r ---+ 44a44 and APRM setpoint modifications necessary to maintain the same margin of safety for single loop operation as is available during two loop operation. Additionally, loop flow limitations are established to ensure vessel internal vibration remains within limits. A flow control mode restriction is also incorporated to reduce valve wear as a result of automatic flow control attempts and to ensure valve swings into the cavitation region do not occur, i An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by waitoring jet pump per-forma ce cu a prescribed schedule for significant degradation. During two loop operation, recirculation loop flow mismatch limits are in compliance with SCCS LOCA analysis design criteria. The limits will ensure an adequate core flow coastdown from either recirculation loop follewing a LOCA. In cases where the mismatch limits cannot be maintained, continued operation is per-mitted with one loop in operation. The power / flow operating map is divid a into four (4) regions. Regions A and B are restricted from operttions. They include the operating area above the 80% rod line and below 40% core flow. Region C includes the operating area above the 80% rod-line and between 40% and 45% core flow. Operation in Region C is allowed only for control rod withdrawals during startup for required fuel preconditioning. Region D consists of the rest of the operating map. No core thermal-hydraulic stability related restrictions are applied to Region D since the potential onset of core thermal-hydraulic ir. stabilities is not predicted within Region D. The definition of Regions A, B and C is based on BWR stability operational data and required operator actions. Although a large margin to onset of insta-bility was observed in Regions A, 8 and C during GGNS stability tests for typical operating configuration, a conservative approach is adopted in the specification. With no reactor coolant system recirculation loops in operation, and the reactor mode switch in the Run position an immediate reactor shutdown is required. ReactorshutdownisnotrequIredwhenrecirculationpumpmotorsare de-energized during recirculation pump speed transfers. Upon entry to Region A an immediate reactor shutdown is required. Upon entry to Region B or Region C, unless operation in Region C is for control rod withdrawals during startup, either a reduction of THERMAL POWER to below the 80% rod-line by control rod insertion or an increase in core flow to exit the region by opening the recirculation loop FCV is required. Per the specification, the APRM neutron flux noise level should be observed while in Regions B and C. In the unlikely event in which a sustained GRAND GULF-UNIT 1 B 3/4 4-1 Amendment No. 62,

3 $4 90/o s" SPECIAL TEST EXCEPTIONS 3/4.10.2 R0D PATTERN CONTROL SYSTEM l LIMITING CONDITION FOR OPERATION 3.10.2 The sequence constraints imposed on control rod groups by the rod pattern control system (RPCS) per Specification 3.1.4.2 may be suspended by means of the individual rod position bypass switches

  • for the following tests:

a. Shutdown margin demonstrations, Specification 4.1.1. b. Control rod scram, Specification 4.1.3.2. ' ' 7* c. Control rod friction measurements. d. Startup Test Program with the THERMAL POWER less than 4GE of l RATED THERMAL POWER. APh.!CABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With the requirements of the above specification not satisfied, verify that j the RMS is OPERADLE per Specification 3.1.4.2. SURVEILLANCE MOUIREMENTS 4.10.2 When the sequence constraints imposed on control. rod groups by the RPCS are bypassed, verify; a. Within 8 hours prior to bypassing any sequence constraint and at least once per 12 hours while any sequence constraint is bypassed, that movement of the control rods ' :: 75% "^D DE".SITY t th; '*CS lee rere retre!-t-is limited to the established control rod sequence for the specified ;est, and b. Conformance with this specification and test procedures by a second licensed operator or other technically qualified member of the unit technical staff, of /e ss fin., ey./ /o t o R, or of RAYE6 7WEMA t. /*u CR

  • Bypassing control rod (s) in the RPCS shall be performed under administrative control.

GRAND GULF-UNIT 1 3/4 10-2 4,w d-ed Ah _}}

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