ML20086K440

Proposed Tech Specs 2.1.2 & 3/4.2.1,supporting Cycle 6 Reload

ML20086K440
Person / Time
Site:	Grand Gulf
Issue date:	12/05/1991
From:	Cottle W ENTERGY OPERATIONS, INC.
To:
Shared Package
ML19353B429	List: ML20086K407 ML20086K440 ML20086K461 ML20086K465 ML20086K474 ML20086K499 ML20086K511 ML20086K519
References
NUDOCS 9112130105
Download: ML20086K440 (19)

Text

-

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2.0 SAFETYJ MITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or low Flow 2.1.1 (HERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure leis than 785 psig or core flow less ti.an 10% of rated flow.

APPLICABILITY:

OPr. RATIONAL CONDITIONS 1 and 2.

ACTION:

dith THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome r essure lebs 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"i during both two loop operatica and single loop operation with the reactof_ l.07 dor %

3 vessel steam dome pressure greater than 785 psig and core flow greater tha 10% of rated flow.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 dnd 2.

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 SHUTOOWN 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.

REACTUR C006 ANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the reactor cooiant sytten pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTCOWN 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 Spccification 6.7.1.

GRAND GULF-UNIT 1 2-1 Amendment No.73,

+1intactos of1205

{DR ADOCKOS00g6

2.1 SAFEff 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 L4mits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuei 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, a step back approach is used to establish a Safety Limit for the MCPR.

MCPR 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 t.raterials 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 measurable.

Fuel cladding perforation, however, can result from thermal stresses which occur from reactor operation si tions and the Limiting Safety System Settings.gnificantly above desigecondi-While fission produr'..gration f-om cladding perforation is just as measurable as that from use relued 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 margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a significant departure from the condition intended by design for planned eperation.

2. L 1

[ h nd THERMAL POWER. Low Pressure or Low Flow

[Tirj fB;m f

The Wanad Nuclear Fve+s, Corporation (W) ANFB critical power correla-tion is applicable to the_fANF core.

The applicable range of the ANFB correla-PHP l tion is for pressures above 585 psig and bundle mass flex greater than 0.25M1bs/

hr-fta.

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 Gu?f cycle 1 operation ba:ed on ATLAS test data and the GEXL correlation.

The use of the GEXL-correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% t' rated flow, Therefore, the fuel cladding integrity Safety Limit was estair1shed by other means.

This was done by establishing 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 flows will always be_ greater than 4.5 psi.

l i

GRAND GULF-UNIT 1 B 2-1 Amendmer.t ho., 73 _{

a s

=-. - - - - - - - - -

---

~~- ~~~~~~

' ~~

3.1 5Afg,7YL1H!?S N 'N / 0 BASES

~~

'iHERMAL POWER, Low Pressure or Low Flow (Continued)

Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and his a value of 3.5 psi.

Thus, the bund'e flow with a 4.5 psi driving head will be greater than 28 x 108 lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indi-cate that the fuel assembly critical power at this flow is approximately 3.35 With the design peaking factors, this corresponds to a THERMAL POWER of MWt.

more than 30% of RATED THERMAL POWER.

Thus, a THERMAL POWER limit of 257, of

' RATED THERMAL POWER Nr reactor pressure below 785 psig is conservative.

Overall, because of tua design thermal-hydraulic _ compatibility _off_MLej AW fuel designs.ith the cycle 1 fuel, this justification and the associated low pres-l sure and low flow limits remain applicable for future cycles of cores contain-ing these fuel designs.

With regard to the low flow range, the core 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-fta for all fuel assemblies which can approach critical heat flux.

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

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

Since the ANFB correlation is applicable at any pressure greater than 585 psig, the cycle 1 low pressure boundary of 785 psig remains valid for the ANF8 correlation.

GRAND GULF-UNIT 1 B 2 la Amendment No.73, -

SAFETV LIMITS A / / ' 7 /

(o BASES i

i 2.1.2 THERMAL POWER, High Pressure and High Flow i

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, tho exhtence of critical power, or boiling transition, is not a di-rectly observ2bt) parameter in an operating reactor.

Therefore, the margin to boiling tra',$it 'n is calculated from plant operating parameters such as core power, core _ fW. feedwater temperature, and core power Wstribution.

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 suf ficient conservatism such that, in the event of a sustained steady state operation at the MCPR safety licit, 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 and includes the effects associated with channel bow.

One specific uncertainty included in the safety limit is the uncertainty inherent in the ANFB critical power correlation.

SNP % report XII I4P S24(E), Rev. 2, " Advanced Nuclear Fuels Corporation Critical l

t Power Methodology for-Boiling Water Reactors " April 1989, including Swk-

! mQ

4. rat.-1, describes the methodology used in determining the Safety Limit MCPR.

The ANFB critical power correlation is based on a significant body of

%4gemt(tical test data, providing a high degree of assurance that the critical prac 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 radial _ power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition.

Still further con-servatism is induced by the tendency of the ANFB correlation to overpredict the number of rods in boiling-transition.

These conservatisms and the inh:trent accuracy of the ANF8 correlation provide assurance that during sustained opera-tion at the Safety Limit MCPR there would be essentially no transition boiling i

in the core.

1-L

?

GRAND GULF-UNIT-1 8 2-2 Amendment No. 7{-- I J-..-.=..--.-.----.~-------~~-~--"

A//-7/[/b 3/4.2 POWER DISTRIBUT20N LXMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION During two loop operation, all AVERAGE PLANAR LINEAR HEAT GENERATION 3.2.1 RATES (APLHGRs) for each type of fuel as a fun: tion of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure 3.2.1-1.

During single loop operation, the APLHGR fei each type of fuel as a function

% of AVERAGE PLANAR EXPOSURE shell not exceed the limit shown in Figure 3.2.1-1 0,36 [ multiplied bygD *.

APPLICABILITY:

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

ACTION:

During two loop operation or single loop operation, with an APLHGR exceeding the limits of Figure 3.2.1-1 as corrected by the appropriate multiplication factor, initiate corrective action within 15 minutes 2nd restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a.

b.

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

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. 73,

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3 AVERAGE PLANAR EXPOSURE (GWd/MT)

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CORETHERMAL POWER (% RATED) 2g I'

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EXPOSURE CYCLE EXPOSURE E

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OWWDR1 pt are 3.2. Y ~E GRAND GULF-UNIT 1 3/4 2-7b Amendment No. 73, f

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2, CORE THERMAL POWER (% RATED) a

~3 e,

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2 FIGURE 3.2.4-3 LHGRFAC o

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CORETHERMAL POWER (% RATED)

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rIGURE 3.2.4-3 LHGRFAC P

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M/. - 7

/d 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temper-ature following the postulatec design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGF 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.

j 1

The peak cladding temperature (PCT) following a postulated loss-of-coolant 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 secondarily on the rod to rod power distribution within an assembly.

Planar Linear Heat Generation Rate (MAPLHGR) limits of Figure 3.2.1-1 areThe Maxim applicable to two loop operation.

}

For single-loop operation, a MAPLHGR limit corresponding to the product O,$(o Jof the MAPLHGR, Figure 3.2.1-1, andsO-& can be conservatively used to ensure

~that the PCT for single loop operation is bounded by the PCT for two loop operation.

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-tien shifts are very slow when there have not been significant pwer or control GRAND GULF-UNIT 1 8 3/4 2-1 Amendment No. 73; I

. -.