ML17320B075

Proposed Tech Spec Changes Re Addl Limitation on Nuclear Enthalpy Rise Hot Channel Factor Due to New Loca/Eccs Analysis in Support of Cycle 5 Reload

ML17320B075
Person / Time
Site:	Cook
Issue date:	05/21/1984
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17320B074	List: ML17320B073 ML17320B075
References
NUDOCS 8405290296
Download: ML17320B075 (17)

Text

Table 2.1 O.C. Cook Unit 2 l.OCA/ECCS Analysis Su@nary Results for the C cle 5 Core Conf i uration 85$ ENC Fuel Peak Rod Average 8urnup (HMD/kg) 2.0 10.0 47.0 FT 2.04 2.04 2.04 Q

T 1.415 1.415 1.415 Peak Cladding Temperature (oF) 2198 2190 '2096 Maximum Local Er-H20 Reaction (5) 7.4 7.3 5.7 Total Zr-H20 Reaction < 1.0 < 1.0 (1.0

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Table 3.2 1.0 DECLG Break Analysis Parameters Peak Rod Average Burnup (HID/kg) 2.0 10.0 47.0 Total Core Power (NMt)* 3411 3411 3411 T 2.04 Total Peaking (F~) 2.04 2.04 Fraction Energy Deposited in Fuel

~ Fully Moderated Core 0.974 0.974 0.974

~ Voided Core 0.954 0.954 0.954 C cle 5 85% ENC Fuel)

Peak ing

~ Axial x Engineering 1. 442 1. 442 l. 442 T

~ Enthalpy Rise (F~H) 1.415 1.415 1.415

• 2% power uncertainty is added to this value in the LOCA analysis.

Table 3.4 1.0 DECLG Break Fuel Response Results for Cycle 5 Peak Rod Average Burnup (NMD/kg) 2.0 10.0 47.0 Initial Peak Fuel Average Temperature (oF) 2151 2060 1629 Hot Rod Burst

~ Time (sec) 60. 9 61.7 67.9

~ Elevation (ft) 6.50 6.50 7.00

~ Channel Blockage Fraction .24 .27 .47 Peak Clad Temperature

~ Time (sec) 227 227 241

~ Elevation (ft) 8.63 8.63 8.88

~ Temperature (oF) 2198 2190 2096 Zr-Steam Reaction

~ Local Maximum Elevation (ft) 8.63 3. 63 8. 88

~ Local Max imum (g)* 7.4 7.3 5.7

~ Core Max imum <1.0 <1.0 <1.0

• Values 400 sec into LOCA transient.

i FQ~2- 0 I i F~1. < 150Re P NQO/KG PCT HODE (HOOE 14 AT I.Ct FT. l t- RLN'THEO HOOE

( HOOE 5 AT C.$ 0 FT- )

I Cl Clg Q,~

d 44.0 e0.0 Lt0.0 140 0 t00 0 t)0 0 tt0 4 %0.0 )Cl. ~

TIHE SE'CONDS Fiqure 3.41 TOODEE2 Cladding Temperature versus Time, 1.0 DECLG Break, 2. MWD/Kg Case

i FQ~R,. 04 i FOH~1'15DRi 17HW/K

1. PCT HODE tNOK 1$ AT l.lf FT. )

t. M TQKt) NSE tHOOE u. AT 1.00 FT )

44.0 t0.0 tt0.0 iC0.0 5$ .0 t.td 0 S0.0 05 0 Xb. ~

TIHE' SECONDS ci~nr~ l 4R TOODEE2 Claddino Temperature versus Time, 1.0 DECLG Break, 47. MWD/Kg Case

> FW2. 0 4 o FOWl. < 150R o 10HNOI K 1.. PCT NHK f NODE 19 AT a.et FT. )

t- RLPTllKD t400E (HATE 0 AT 0.$ 0 FT. )

i0.0 e0.4 1t0.0 140.0 'O0.0 t40.0 tl0.4 KJt.0 ~- ~

TINE SECONDS Figure 3.42 TOODEE2 Cladding Temperature versus Time, 1.0 DECLG Sreak, 10. NWD/Kg Case

Mr. Haro1d R. Den <<4 AEP: NRC: 0860K ATTACHMENT 2 PROPOSED REVISION TO TECHNICAL SPECIFICATION PAGES

CS FLO RATE AND NUC EAR ENT Y E 0 0 ERATIO 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation shown on Figures 3.2-4 and 3.2-5 for 4 and 3 loop operation, respectively.

For: Westinghouse Fuel for: Exxon Nuclear Company Fuel N N AH R -" f R = WWWWO ~ WW&W&W&WWW WWWWWO WWWW

1. 48 [1. 0 + 0. 2 (1. 0-P) ] 1.49 [1.0 + 0.2 (1.0 - P) ]

N And, F < H 5 1.36/P for Exxon Nuclear Company Fuel where:

RATED THERMAL POWER N

andFAH = measured values of F N< obtained by using the movable incore detectors to ob$ ain a power distribution map. The measured values of F ~ and flow, without additonal uncertainty allowance, shall be used to compare with limits.

MODE 1.

fZZLIQAPXJJX'QXXQK N

With F A H

above the allowable limit or with the combination of RCS total flow rate and R outside the region of acceptable operation shown on Figure 3.2-4 or 3.2-5 (as applicable):

a~ Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

l. Either restore F NA and the combination of RCS total flow rate and R to withPn the above limits, or

2. Reduce THERMAL POWER to less than 50$ of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint to J 55$ 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 />.

D.C, COOK - UNIT 2 3/4 2-9 AMENDMENT NO.

kGXZQK: (Continued)

b. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify through incore flux mapp1ng and RCS total flow rate compar1aon that F $ E and the comhinatlon of R and ROE total flow rate are restored %o within the above limits, or reduce THERMAL POWER to less than 5$ of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c~ Identify and correct the cause of the out-of'-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER Limit required by ACTION items a.2 and/or b above; subsequent POWER OPERATION may proceed provfded that F ) and the combination of R and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the reg1on of acceptable operation as defined above for F" and as shown on Figure 3.2-4 or 3.2-5 (as applicable) for RCS flow rate and R prior to exceed1ng the following THERMAL POWER levels:

1. A nominal 501 of RATED THERMAL POWER,

2. A nominal 75$ of RATED THERMAL POWER, and

3. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining g 954 of RATED THERMAL POWER.

4.2.3.1 The provisions of Specif1cation 4.0.4 are not applicable.

N F <H '.2.3.2 shall be determined to be within the above limits and the combination of indicated RCS total flow rate and R shall be determined to be within the region of acceptable operation of Figure 3.2-4 or 3.2-5 (as applicable):

a. Prior to operat1on above 75$ of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 Effective Full Power Days.

4.2.3.3 The RCS total flow rate indicators shall be sub)ected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.3.4 The RCS total flow rate shall be determined by measurement at least once per 18 months.

D.C. COOK - UNIT 2 3/4 2-10 AMENDMENT NO.

~ ~

48 e

Measurement Uncertainties of 3.5% for Flo~ and 4%

F JH are accounted for in the analysis which 46 supports this Figure.

44 C) 42 ACCEPTABLE OPERATION REGION 4

40

[

~ ~

~ =-= -- =-'=-: -UNACCEPTABLE OPERATION REGION (0.3=,:i. 77) .

34 0.90 0.94 0.98 1.02 1.06 a ~ a<

R*F~NH/1. 48[1.0+0. 2( l. 0-P ) l WESTINGHOUSE FUEL R*FNH/1.49{1.0 0.2(1.0-P)] EXXON NUCLEAR CO. FUE F I GURE 3. 2 4 RCS TOTAL Fl O'NRATE VERSUS R FOUR LOOP 5 IN OPERATION

0. C. COOK UNIT 2 3/4 2-11 Amendment No.

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'Htasureaent Unclrtalnt)es:

.'of 3.5% for Flow and i%

~ ~ ~ Kt should be noted that

.. <or F are accounted for three looo ooerat'on using

.: in the analysis which this curse is not curren ly allowed Ne changes supports this Figure.

36

,

contained in this table are 0 ~

for Reference only.

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~I ACCEPTAgLE X

CL OPERATION

,CD REGION 32 C7 30 I

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UNACCEPTAgLE

~ OPERAT'ON

( ~ ~'"IGt' 28 ~

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Z6

=:: . (0.971,26.'.=.)

24 1.06 1.10 14

0. 90 0. 94 0.98 1.02 R<F "H/l.48{ 1. 0+0. 2( 1. 0-P ) ] MES NGHOUSE FUEL R*F"H/1.49{1.0+0.2(I.O-P) ] EXXON NUCLEAR CO. FUEL FIGURE 3.2-5 RCS TOTAL FLOQRATE VERSUS R - THREE LOOPS IN OPERATION

0. C. COOK UNIT 2 3/4 2-12 Apgn4nent No.

p, shape.

AH'n The curves are based on a nuclear enthalpy rise hot channel factor, or i.49 and a reference cosine with a peak of t.55 i'or axial power An allowance is included for an increase in F< H at reduced power based the expression:

F>

- 1.48 [1 + 0.2 (1-P)] (Westinghouse Fuel)

H N

F< = 1.49 [1 + 0.2 (1-P)] (Exxon Nuclear Company Fuel)

H where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable oontrol rod insertion assuming the axial power imbalance is within the limits of the f ( A I) function of the Overtemperature trip. When the axial power imbalance ik not within the tolerance, the axial power imbalance effect on the Overtemperature A T trips will reduce the setpoints to provide proteotion consistent with core safety limits.

For Exxon Nuclear Company supplied fuel, an additional limitation on F < is applied to ensure compliance with ECCS acceptance criteria. This limitation is discussed in basis section 3/4.2.2 and 3/4.2.3 and does not affect the safety limit curve.

2.1.2 0 C The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110$ (2735 psig) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entir e Reactor Coolant System is hydrotested at 3107 psig, 125$ of

'design pressure, to demonstrate integrity prior to initial operation.

D.C. COOK - UNIT 2 B 2-2 AMENDMENT NO.

The specifications of'his section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the calculated DNBR in the core at or above design during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and oladd1ng mechan1cal properties to within assumed design criteria. In add1tion, limiting the peak linear power density dur1ng Condition I events provides assurance that the initial cond1t1ons assumed f'r the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.

The definitions of certain hot ohannel and peaking factors as used in these specifications are as follows:

F (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of' fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

Nuclear Enthalpy Rise Hot Channel Faotor, is defined as the ratio N

FAH of the integral of linear power along the rod with the highest integrated power to the average rod power.

N The limits on F (Z) and F < for Westinghouse supplied fuel at a core

-average power of 3431 MWt are 1.$ 7 and 1.48, respectively, which assure consistency with the allowable heat generation rates developed for a core average thermal power of 3391 MWt. The limits on F (Z) and F ". for ENC supplied fuel have been established for core thermal power oP 3411 MWt. The limit on F (Z) is 2.04. The limit on F >g is 1.36 for LOCA/ECCS analysis and 1.49 for DIIB analyses. The analyses supporting the Exxon Nuclear Company limits are valid for an average steam generator tube plugging of up to 5$ and a max1mum plugging of one or more steam generators of up to 10$ . In establishing the limits, a plant system description with improved accuracy was employed during the reflood portion of the LOCA Transient. With respect to the Westinghouse supplied fuel the minimum proJected excess margin of at least 10$

to ECCS limits will more than offset the impact of incr ease steam generator tube plugging.

The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound envelope is not exceeded during either normal operation oF in the event of xenon redistribution following power changes. The F (Z) upper bound envelope is 1.97 times the average fuel rod heat flux for Weslinghouse suppl1ed fuel and 2.04 t1mes the average fuel rod heat flux for Exxon Nuclear Company supplied fuel.

Target flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near the1r normal posit1on for steady state operation at high power levels. The value of the D.C. COOK - UNIT 2 B 3/4 2-1 AMENDMENT NO.

2 HEAT L OT CTO The limits on heat flux hot channel factor, RCS flowrate, and nuclear, enthalpy rise hot channel factor ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) $ n the event of a LOCA the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 0.2.2 and 4.2.3. This periodic surveillanoe is sufficient to ensure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than + 12 steps from the group demand position.

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6.

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits, maintained within its limits provided oonditions a. through d.

N F<H will be above are maintained. As noted on Figures 3.2-4 and 3.2-5, RCS flow rate and F< may be "traded off" against one another (i.e., a low measured RCS flow ra@ is acceptable if the measured F< is also low) to ensure that the calculated DNBR will not be below thebesign DNBR value. The relaxation of F<H as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. The form of this relaxation for DNBR limits is discussed in Section 2.1.1 of the basis.

An ditional limitation on F <N applies to Exxon Nuolear Company fuel.

This F < H limit, in combination witk the F~(Z) limit, ensures compliance with the ECCS acceptance criteria. An allowance is included for an increase in F <N H

at reduced power based on the following expression:

F <N g 1.36 /P (Exxon Nuclear Company Fuel) where: P is the fraction of RATED THERMAL POWER.

The power dependence of this allowance is 1/P because the assooiated F <

H limit of 1 36 results from the LOCA analysis.

The more restrictive of the flow dependent DNBR F <N H limit and the LOCA F ~ limit for Exxon Nuclear Fuel Company fuel must be applied.

D.C. COOK - UNIT 2 B3H 2-4 AMENDMENT NO.

RISER- (Continued)

F~re B 3/4 2-2 illustrates the implementat1on of the limits as a f nctio of power. A measured flow will result in a limiting value for R which must be obtained from Figure 3.2-4 or Figure 3.2<<5. From this limiting R, a limiting F < can be obtained because:

H Westinghouse Fuel Exxon Nuclear Company Fuel F > H=1.48 X R X [1.0+0.2(1.0-P)3, F A H=1.49 X R X [1.0+0.2(1.0-P)]

THERMAL POWER Where: P RATED THERMAL POWER Figure B 3/4 2-2 displays two limitigg DNBR F N curves fear Exxon Nuclear Company fuel for flows of 36.77 X 10 cpm, an 37.63 X 10 gpm. Also displayed on Figure B 3/4 2-2 is the limit oqF< H which results from the LOCA analys1s i'or Exxon Nuclear Company fueQ p must be maintaiyd below and to the left of both the applicable DNBR F< H l mIIt and the LOCA F<

H limit.

For Westinghouse fuel there 1s only one N limit.

g It rust be obtained from the applicable relationships among R, F< H, P, and flow.

When an F measurement is taken, both experimental error and manufacturing tolerance must be allowed fore 5g is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3$ is the appropriate allowance for manufactur1ng tolerance.

When RCS flow rate and F <N are measured, no additional allowances. are necessary prior to comparison 4th the limits of Specif1cat1on 3.2.3.

Measurement errors of 3.5$ for RCS flow total flow rate and 4$ for FN<H have been allowed for in determination of the design DNBR value and in the determ1nation of the LOCA/ECCS limit.

D.C. COOK - UNIT 2 B 3/4 2-4a AMENDMENT NO.

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.35 0 20 40 60 80 100 PERCENT OF RATED THERMAL POWER FIGURE B 3/4 2-2 ILLUSTRATIVE EXAMPLE OF F~H LIMIT VERSUS PERCENT THERMAL POWER FOR EXXON FUEL B 3/4 2-4b