ML20024A898

Proposed Changes to Tech Spec Section 3.10 Resulting from New LOCA Analysis

ML20024A898
Person / Time
Site:	Prairie Island
Issue date:	06/24/1983
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20024A891	List: ML20024A890 ML20024A893 ML20024A898 ML20024A901
References
NUDOCS 8307010153
Download: ML20024A898 (9)

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EXHIBIT B License Amendment Request dated June 24, 1983 Exhibit B consists of revised pages of Appendix A Technical Specifications as listed below:

Pages TS-iv TS.3.10-1 TS.3.10-2 TS.3.10-9 TS.3.10-11 i Figure TS.3.10-5 Figure TS.3.10-7 i.

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"AP?ENDIX'A TECHNICAL SFECIFICATIONS.

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'_ Safety L'irits.. Reactor Core, Thermn2 and Hydraulic Two Loop -

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Operation

.I , 3.1-1 f(' <

Unit 1 anri Lait'2 Reactor Coolant System Heacup Lb.21tations 3fi 1.1-2 ji Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitaticus

$1O '3.1-3  ; (,'. Effeet-of Fluence-2nd-Copper Content on Shift of RTeg for Reactor 7psel Steels Exposed to '!<50* Te.tparature

' h\ [. 3. h4 Fast Veutron 71uence (E >l Mav) as s' Function of Full Power

Service-Life.

3.1-5 iDOSE EOUIVALCIT I-131 Primary Coolent Specific Activity Limit ,

, Y. . versus Percent of JP.ATED THER?t!.L e POWER with the Primary Coolant

. Specific Acivity >1.0 uC1/ggan DOSE EO.UIVALENT I-131 -

., 3.921 ?rairie Island Neleur Generating Dlant Site Soundary for Liquid

'f Effluents s ? 3.9-2 th.' ri.: Island'Euciaar Generatiog Plant Site Boundary for J u Casum Effluents

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h 3.10-1 Ragtiired Shutdown Reactivity Vs F.eactor Boron Concentration 3.10 Control Bank Insertion LimLt3

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• Insertion Linics 100 f,tep tNerlap with One Bottomed Rod L! O-t -

3.10 Insertion Limits 100 Step G-erlap with one Inoperable Rod

, c;P "4.10- 3 ' Hot Channel Factor Norms?.ited Operating Envelope

~ j, b a t c3.10-6 Deviation froc. Target. Flux Difference as a Function of Thermal l 4 ,

.\ Power 3 3 10-7 Normalized Expos' I re1ependant' Function BU(E ) tor Exxon Nuclear Compeny Tucl.

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V(Z) as a Function of Core :; sight j

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4. M-2 Prairie lfshnd Nt:close C-enarating alant Radiation Environmental

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, t 1.. 6. bl NSP Corporate Organizational Relati1nship t. On-site @perating J Organization '

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[ TS.3.10-1 REV 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Aeolicability Applies to the limits on core fission power distribution and to the limits on control rod operations.

Ob jeet tve To assure 1) core suberiticality after reactor trip, 2) acceptable core power distributions during power operation, and 3) limited potential reactivity in-sertions caused by hypothetical control rod ejection.  !

Specification A. Shutdown Reactivity The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown in Figure TS.3.10-1 under all steady-state operating conditions, except for physics tests, from zero to full power, including ef fects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be suberitical at hot shutdown conditions if all control rod assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron concentration. ,

l B. Power Distribution Limits '

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At all times, except dgring lp power physics testing, measured hot channel f actors, F" and F and in the bases, sha19 meet N e, following as definedlimits:

below F

q x 1.03 x 1.051 (2.32/P)* x K(Z) x BU(Ej )  !

F H* *

, i 1.55 x [l+ 0.2(1-P)]

where the following definitions apply:

(a) K(Z) is the axial dependence function shown in Figure TS.3.10-5.

l (b) Z is the core height location (c) is the imum pellet exposure in fuel rod j for ich the is being measured.

1 (d) BU(E.) is the normalized exposure dependence function for Exxed Nuclear Company fuel .shown in Figure TS.3.10-7. For Westinghouse fuel, BU(E.) = 1.0 3

(e) P is che g. fraction of full power at which the core is operating.

In the F limit determination when P 10.50, set P = 0.50.

• (2.21/P) shall be used for Westinghouse assemblies a _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . J

f. .

TS.3.10-2 REV (f) F or F is defined as the v3th the smallest margin or measured F or F or greatest exhess R' ltatt

.re8pectivelF ,

(g) 1.03 is the engi ering hot channel f actor, F , applied to the measured to account formanufacturiOgtolerance.

(h ) 1.05 is applied to the measured FN to account for measure-q ment uncertainty i

'(i) 1.04 is applied to the measured F to account for measure- j ment uncertainty

2. Hot channel f actors, F and F AH, shall be measured and the 9

target flux difference determined, at equilibtium conditions according to the following conditions, whichever occurs first:

(a) At least once per 31 effective full power days in conjunction with the target flux difference determination, or (b) Upon reaching equilibrium conditions af ter exceeding the reactor power at which target flux difference was last determined, by 10 '

or r. ore of rated power.

F the following limit for the middle axial 80 o9(equil)shallmeet the core:

F l (equil) x V(Z) x 1.03 x 1.05 < (2.32/P) x K(Z) x BU(E))

where V(Z) is defined in Figure 3.10-8 and other terms are defined in 3.10.B.1 above.

3. (a) If either measured hot channel factor exceeds its limit specified in 3.10.B.1, reduce reactor power and the ghij neutgon flux trip setpoint by 1 for each percent that the measured F o exceeds the 3.10.B.1 limit. Then follow 3.10.B.3.9c)r F" .

(b) If the measured F (equil) exceeds the 3.10.B.2 limits but not the 3.10.B.1 limit, tSke one of the following actions:

1. Within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> place the reactor in an equilibrium configuration for which Specification 3.10.B.2 is satisfied, or

2. Reduce reactor powe'r and the high neutron lux trip setpoint by 1: for each percent that the measured (equil) x 1.03 x 1.05 x V(Z) exceeds the (2.32/P) x K(Z) BU(E.) limit. i 3 i u J

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TS.3.10-@

REV 4

mechanical properties to within assumed design criteria. in addition, limiting the peak linear power density during Condition 1 events pro-vides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.

During opeJacion the plant staff compares the measured hot channel l 4

factors, Ff and , (described later) to the limit determined l in the transient d LOCA analyses. The limiting F (Z) includes measurement, i engineering, and calculational uncertainties. The 9erms on the right I side of the equations in section 3.10.5.1 represent the analytical limits. '

Those terms on the lef t side represent the measured hot channel factors l corrected for engineering, calculational, and measurement uncertainties.

F (Z), Height Deoendent Heat Flux Hot Channel Factor, is defined as the l inum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing q tolerances *on fuel pellets and rods. The maximum value of Fg (Z) is 2.32/P l for the Prairie Island reactors. This value is restricted fcrther by the K(Z) i and BU(E ) functions described below. The product of these three factors is

. Fq (Z).

The K(Z) function shown in Figure TS.3.10-5 is a normalized function that limits F n(Z) axially for three reasons. The K(Z) specified for the lowest sIx (6) feet of the core is based on large break LOCA analyses. .

Above this region the K(Z) value is based on DNBR requirements since the minimum DNBR would be expected in this region of the core, based on power, pressure, and temperature. The K(Z) value in the uppermost region of the core is based on the small break LOCA analyses. Fg(Z) in the uppermost region is limited to reduce the PCT expected during a small break LOCA since this region of the core is expected to uncover temporarily for some small break LOCA's.

The BU(E.) function shown in Figure TS.3.10-7 is a normalized function that lialts F analyses for the ENC fuel.

These analyseh(Z) based consider oninternal pin exposure dependent pressure uncertainties, fuel swelling, t

rupture pressures, and flow blockage.

F is the measured Nuclear Hot Otannel Factor, defined as the

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mSximumlocal heat flux in the core divided by the average heat flux in the core. Heat fluxes are derived from measured neutron fluxes.

V{Z)isanaxjallydependent function applied to the equilibrium measured F to bound F 's that could be measured at non-equilibrium conditions.

Tkisfunctionisbasedonpowerdistributioncontrolanalysesthat S

eval-usted the effect of burnable poisons, rod position, axial effects, and xenon worth.

FE , Engineering Heat Flux Hot Channel Factor, is defined' as the a91ovanceonheat flux required for manufacturing tolerances. The ,

engineering f actor allows for local variations in enrichment, pellet I density and diameter, surf ace area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect i is a factor of 1.03 to be appl 2d to fuel rod surface heat flux.

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l TS.3.10-11 i REV inches from the bank demand position. An accidental misalignment limit of 13 steps precludes a rod misalign-ment greater than 15 inches with consideration of maximum instrumentation error.

2. Control rod banks are sequenced with overlapping banks as described in Technical Specification 3.10.

3. The control bank insertion limits are not violated.

4. Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion i

limits are observed. Flux difference refers to the difference in signals between the top and bottom halves of two-section excore. neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.

The permitted relaxation in F and F allows for radial power H

shape ch.usges with rod insertson to the insertion limits. It has been determined that provided the above conditions 1 through 4 are obs ed, these hot channel f actor limits are met. In specification 3.10, is arbitrarily limited for P < 0.5 (except for low power physics test ).

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The procedures for axial power distribution control referred to above are designed to minimize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers. Bas ically control of flux difference is required to limit the difference between i

the current value of Flux Difference ( dI) and a reference value

which corresponds to the full power equilibrium value of Axial Offset

CAxial Of fse't = dI/ fractional power). 1he reference value of . flux

! difference varie,s with power level and burnup but expressed as axial

{ of fset it varies only with burnup.

The technical specifications on power distribution control assure that 4

the F l 2 TS.3.0(Z) r0-7 is upper bound envelope not exceeded and xenon of 2.32 /P times Figures distributions are notTS.3.10-5 and developed i' which at a later time, would cause greater local power peaking even though the flux difference is then within the limits specified by the procedure.

1 The target (or reference) value of flux difference is ' determined as i follows: At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with the full length rod control rod baak more than 190 steps withdrawn (i.e. ,

normal full power operating position appropriate for the time in life, usually withdrawn farther as burnup proceeds). This value, divided by the fraction of full power at which the core was operating is the full power value of the target flux difference. Values for all other core power levels are obtained by multiplying the full power value by the fractiona'l power. Since the indicated equilibrium was noted, no allowances for excore detector error are necessary and indicated deviation of +5 percent dI are permitted from the indicated reference value. Figure TS.3.10-6 shows the allowed deviation from the target flux difference as the function of thermal power.

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FIGURE TS.3.10.7 REV Normalized Exposure Dependent ,

Function BU(E)) for Exxon Nuclear Company Fuel 4-.. ....p._.__-.___

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{ EXHIBIT C 2

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PRAIRIE ISLAND NUCLEAR GENERATI?C PLAlff

}-

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I j License Amendment Lequest - Dated June 24, 1983 l

i- XN-NF-83-38 4

f Prairie Island Units 1 and 2 4

Limiting Break LOCA-ECCS Analysis 4

Using EXEM/PWR 4

a 1

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XN NF 83 38 l

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PRAIRIE ISLAND UNITS 1 AND 2 l LIMITING BREAK LOCA-ECCS ANALYSIS l USING EXEM/PWR i

MAY 1983 l

l E(ON NUCLEAR COMPANY,Inc.

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