ML17321A096

Summary of 840518 Meeting W/Util & Exxon Nuclear Co Re Open Issues on Cycle 5 Reload Review.List of Attendees Encl

ML17321A096
Person / Time
Site:	Cook
Issue date:	06/06/1984
From:	Wigginton D Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8406210180
Download: ML17321A096 (17)

Text

Docket No. 50-316 JUN 6

)984 LICENSEE: Indiana and Michigan Electric Company FACILITY: Donald C.

Cook Nuclear Plant, Unit No.

2

SUMMARY

OF MEETING HELD ON MAY 18, 1984 WITH INDIANA AND MICHIGAN ELECTRIC COMPANY (IMEC) AND EXXON NUCLEAR COMPANY TO DISCUSS THE CYCLE 5 RELOAD REVIEW The staff met with the licensee and Exxon on May 18, 1984 to review any open issues on the Cycle 5 reload review for Donald C.

Cook Nuclear Plant,

, Unit No. 2.

The list of attendees is attached as Enclosure 1.

As a result of discussion with the reviewers, the current ECCS analysis performed by Exxon is undergoing revisions and will not be completed in time for the restart.

The Exxon representatives have devised an alternative Technical Specification on the nuclear enthalpy rise hot channel f((ctor (Fa

) which will allow Cook 2 to operate with the flow dependent Fa versus powI(r to be both LOCA and departure from nucleat boiling (ONB) limi(ed.

The Technical Specificatio)) will define acceptable operation limits based on the measured flow and FzH for various power levels.

This method is acceptable to the staff but may cause some power derating initially and further power deratings after about 8 months.

The proposed Technical Specifications, Enclosure II, were discussed but some revision will be forthcoming from the licensee.

The licensee proposes to eliminate the possibility of further power deratings by use of a Z equivalent method for adjusting FLECHT based heat transfer coefficients to the D.

C.

Cook axial power shape.

The licensee will provide more information on the Z equivalent method if it becomes necessary to prevent the further power derating of the plant.

Enclosure:

As stated David L. Wigginton, Project Manager Operating Reactors Branch ¹1 Division of Licensing cc w/enclosure:

See next page OR:

L D i ginton;ps

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Indiana and Michigan Electric Company Donald C.

Cook Nuclear Plant, Units 1 and 2

cc:

Mr. M. P. Alexich Vice President Nuclear Engineering American Electric Power Service Corporation 1 Riverside Plaza

Columbus, Ohio 43215 Mr. William R.

Rustem (2)

Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 Mr. Wade Schuler, Supervisor Lake Township

Baroda, Michigan 49101 W.

G. Smith, Jr., Plant Manager Donald C ~

Cook Nuclear Plant Post Office Box 458 Bridgman, Michigan 49106 U.S. Nuclear Regulatory Commission Resident Inspectors Office 7700 Red Arrow Highway Stevensvi 1 1 e, Michigan 49127 Gerald Charno ff, Esquire Shaw, Pittman, Potts and Trowbridge 1800 M Street, N.W.

Washington, DC 20036 Honorable Jim Catania, Mayor City of Bridgman, Michigan 49106 U.S.

Environmental Protection Agency Region V Office ATTN:

EIS COORDINATOR 230 South Dearborn Street

Chicago, IL 60604 Maurice S.

Reizen, h1.D.

Director Department of Public Health Post Office Box 30035

Lansing, Michigan 48109 The Honorable Tom Corcoran United States House of Representatives Washington, DC 20515 James G.

Keppl er Regional Administrator - Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137 J. Feinstein American Electric Power Service 1 Riverside Plaza

Columbus, Ohio 43216

ll ENCLOSURE 1

D.

C.

COOK 2 RELOAD MEETING MAY 18, 1984 Name D. L. Wigginton G.

Requa J.

M. Cleveland V. Vanderburg G.

F. Owsley H. Y. Fouad M. Duneni'el d G.

N. Lauben J.

Guttmann R. Copeland James G. Feinstein R.

C. Jones W.

Kemper Or anization NRR-ORB¹1 NRR-ORB¹1 AEPSC -

NMFM AEPSC -

NHFH Exxon Nuclear AEPSC - NSL NRR-CPB NRR-RSB NRR-RSB Exxon Nuclear AEPSC - Mgr.

NSSL NRC-RSB ENC

DRAFT

'Ehl psUeg 0

ITO 0

OERATON 3.2.3 The more restrictive of the following limit on F< H and the limit on N

the combination of indicated Reactor Coolant System (RCS) total flowrate and R

shall apply to the indicated fuel types.

a.

For all fuel types, the combination of indicated 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.

Exxon Nuclear Company Fuel N'H 1.48 f1.0 + 0.2 (1.0-P)3 1.49 [1.0 + 0.2 (1.0 - P) 3 and RATED THERMAL POWER b.

For Exxon Nuclear Company fuel, N

F AH g M4/P awe

• Le Supp/<wd Myles]s The bases include a discussion of implementation of these limits.

MODE 1.

kQLM 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:

N l.

Either restore F> H and the combination of RCS total flow rate and R to withzn 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 g 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.

DRAFT

~egg:

(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 incor e flux mapp1ng and RCS total flow rate compar1son that F g H and the combination of R and RCS 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,ybove; subsequent POWER OPERATION may proceed provided that FP and the combination of R and indicated RCS total flow rate are demonstrated, through 1ncore flux mapp1ng and RCS total flow rate comparison, to be within the region of acceptable operation as defined above for F >N and as shown on 5'igure 3.2-4 or 3.2-5 (as appl1cable) for RCS flow rate and R prior. to exceeding the following THERMAL POWER levels:

1.

A nominal 50$ of'ATED 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'ttaining g 95$ of RATED THERMAL POWER.

NC E

4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

N 4.2.3.2 F < H shall be determined to be with1n the above limits and the comb1nation of 1ndicated 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 operation above 75$ of RATED THERMAL POWER after each fuel

loading, and b.

At least once per 31 Effective Full Power Days.

Where:

Westinghouse Fuel N

R 1.48 [1.0 + 0.2 (1.0 - P) ]

Exxon Nuclear Company Fuel N

F R

m w

w

~

1.49 [1.0 + 0.2 (1.0 - P) ]

D.C.

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

UT 0 DRAFT ~

(Continued)

N N

P

=

Measured values of F <

obtained by using the movable 1ncore detectors to obtain a power distribution map.

The measured values of F "

shall be used since Figures 3.2-4 and 3.2-5 include measurement uncerta1nty of 3.5$ for flow and Figures A H.

3.2-4 and 3.2-5 and the F A 2 limit also dnolup a measurement N

uncerta1nty of 4$ for incore measurement of F <

4.2,3.3 The RCS total flow rate ind1cators shall be sub)ected to a CHANNEL CALIBRATIONat 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 2-10a AMENDMENT NO.

46 Measurement Uncertainties of 3.5% for'1ow and 4X d'or F>~H are accounted for in the analysis which supports this Figure.

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33 D

F'CCEPTABLE OPERATION REGION

~ ~

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--:=-'= =UNACCEPTABLE OPERATION REGION (1.0, 37. 63) 36 E:

(0.9.=,3.. 77).,

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0. 90 1.10 0.94 0.98 1.02 1.06 R=F~H/1. 48L1.0+0. 2( l.0-P ) ] WESTINGHOUSE FUEL'=F H/1.49I1.(W0.2(1.0-P)]

EXXON NUCLEAR CO.

FUEL FIGURE 3.2-4 RCS TOTAL FLOMRATE VERSUS R -

FOUR LOOPS IN OPERATION D.

C.

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

DRAFT ~

38 36

. Measurement Uncertainties

- of 3.5% for Flow and 4X

-. for Incore Measurement of

-:.F$8 are Included in this

.: FTgure.

~

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Zt should be noted that three loon oneration using this curve is not currently=-..

al3.owed.

The changes contained in this table are "

'or Reference only.

34

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30 CD 28 ACCEPTABLE OPERATION REGION I

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UNACCEPTABLE OPERATION REGION

(..0,27.13) 26

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-'-:--- (0 971-6 i5)

=-----"---:

24

0. 90

0. 94 0.98 1.02 1.06 1.10 R~F"8/1.48[1.0+0.2(1.0-P) ] MES;NGHOUSE FUEL R=F~NH/1.49(1.0+0.2(1.0-P) ] EXXON NUCLEAR CO.

FUEL FIGURE 3.2-5 RCS TOTAL FLOXRATE VERSUS R -

THREE LOOPS IN OPERATION 0.

C.

COOK UNIT 2 3/4.2-12 Amendment No.

M TS DRA 3fLSJ'.S The curves are based on a nuclear enthalpy rise hot channel factor, F

N> H, of 1.49 and a reference cosine with a peak of 1.55 for axial power shape.

An allowance is included for an increase in F~< H at reduced power based on the expression:

N F< H

=

1.48

[1 + 0.2 (1-P)]

(Westinghouse Fuel)

N F<

=

1.49

[1 + 0.2 (1-P)]

(Exxon Nuclear Company Fuel) 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 control rod insertion assuming the axial power imbalance is within the limits of 'the f1

( A I3 function of the Overtemperature.trip.

When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature T trips will reduce the setpoints to provide protection consistent with core safety limits.

N For Exxon Nuclear Company supplied fuel, an additional limitation on F

is applied to ensure compliance with ECCS acceptance criteria.

This limifation 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

~CT~OC The restriction of this Safety Limit protects the integr ity 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 IIIof 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 entire Reactor Coolant System is hydrotested at 3107 psig, 125%

of'esign

pressure, to demonstr ate integrity prior to initial operation.

D.C.

COOK - UNIT 2 B 2-2 AMENDMENT NO.

Dnap The specif1cations of this section provide assurance of fuel integr ity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the calculated DNBR in the core at or above des1gn during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.

The definit1ons of certain hot channel and peaking factors as used in these specifications are as follows:

Fq(Z)

Heat Flux Hot Channel Factor, is defined as the maxirmm 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 tolerances on fuel pellets and rods.

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

A for Westinghouse supplied fuel at a core average power of 34II1 MWt are 1.$7 and 1.48, respectively, wh1ch 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 g core thermal power oP 3411 MWt.

The limit on F (Z) is 2.04.

The limit on F>

is k&f for LOCA/ECCS analysis and 1.49 for DR3 analyses.

The analyses supporting the Exxon Nuclear Company l1mits are valid for an average steam generator tube plugging of up to 5$ and a maximum 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 port1on of the LOCA Transient.

With respect'o the Westinghouse supplied fuel the minimum projected excess margin of at least 10$

to ECCS l1mits will more than offset the impact of increase steam generator tube plugging.

ke S uPP Idee'~

A~BIgsis The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound envelope is not exceeded during either normal operat1on oz in the event of xenon red1stribution following power changes.

The F (Z) upper bound envelope is 1.97 times the average fuel rod heat flux for WesPinghouse supplied fuel and 2.04 times 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 1n accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels.

The value of the D.C.

COOK - UNIT 2 B 3/4 2-1 AMENDMENT NO.

~

DRAFT 4 2 HEAT HOT 0

CHAN E TO 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 4.2.2 and 4.2.3.

This periodic surveillance 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 Specif ication 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 FL'UX DIFFERENCE, is maintained within the limits.

N F<H will be maintained within its limits provided conditions a. through d.

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 Au N

~

~

race 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 applies to Exxon Nuclear Company fuel.

N This F

< H limit, in combination with the F~(Z) limit, ensures compliance with the ECCS acceptance criteria.

An allowance is included for an increase in F

> H at reduced power based on the following expression:

N F AH+44k/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 associated F

< H limit of fkg results from the LOCA analysis.

The more restrictive of the flow dependent DNBR F

< H limit and the LOCA N

F A

limit for Exxon Nuclear Fuel Company fuel must be applied.

be svppj<<4 pew, a<<l q,z D.C.

COOK - UNIT 2 B3/4 2-4 AMENDMENT NO.

ST UT 0 M

~ a (Continued)

~

DRAFT Figure B 3/0 2-2 illustrates the implementation of the limits as a function of power.

A measured flow will result in a limiting value for R which must be obtgined from Figure 3.2-4 or Figure 3.2-5.

Prom this limiting R, a limiting F

< H can be obtained because:

Westinghouse Fuel F

A H=1.48 X

R X [1.0+0.2(1.0-P)]P Exxon Nuclear Company Fuel F

A H=1.49 X R X [1.0+0.2(1.0-P) I Where:

P THERMAL POWER RATED THERMAL POWER Figure B 3/4 2-2 displays two llmitigg DNBR P curves fear Exxon Nuclear Fuel N

Company fuel for flows of 36.77 2

'lO

cpm, an 37.63 2 10 gpm.

Also displayed on Figure B 3/4 2-2 is the limit oq FA which results from the LOCA analysis for Exxon Nuclear Company fuel F must be maintaiyd below and to the left of both the applicable DNBR F< H ltmxt and the LOCA F< H limit.

For Westinghouse fuel there is only one g limit. It must be obtained

-N from the applicable relationships among R, F< H, P,

and flow.

When an F measurement is taken, both experimental error and manufacturing tolerance mus3 be allowed for.

5C is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3$ is the appropriate allowance for manufacturing tolerance.

When RCS flow rate and F<

are measured, no additional allowances are N

necessary prior to comparison 4th the limits of Specification 3.2.3.

Measurement errors of 3.5$ for RCS flow total flow rate and 4$ for FN<~

have been allowed for in determination of the design DNBR value and in the determination of the LOCA/ECCS limit.

D.C.

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

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GUY'VC be revis~P bus~A ~

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20 40 60 80

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100 PERCENT OF RATED THERMAL POWER FIGURE 8 3/4 2-2 TYPICAL F Z H LIMIT VERSUS PERCENT THERMAL POWER fOR EXXON FUEL

0 (s

4

V 1

MEETING

SUMMARY

DISTRIBUTION Docket or Central File NRC PDR Local PRD ORBgl Rdg J. Partlow (Emergency Preparedness only)

Steve Varga Project Manager OELD F.. Jordan J.

N. Grace ACRS (10)

NSIC Gray Fi 1 e Plant Service List NRC Partici ants M. Dunenfeld G.

N. Lauben J.

Guttmann R.

C. Jones

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