ML20034D826

Reload Safety Evaluation,Cycle 19

ML20034D826
Person / Time
Site:	Kewaunee
Issue date:	02/02/1993
From:	Holby J, Mott L, Wanner D WISCONSIN PUBLIC SERVICE CORP.
To:
Shared Package
ML20034D819	List: ML20034D818 ML20034D826
References
NUDOCS 9302240224
Download: ML20034D826 (68)

Text

,

KEWAUNEE NUCLEAR POWER PLANT RELOAD SAFETY EVALUATION CYCLE 19 JANUARY 1993 WISCONSIN PUBLIC SERVICE CORPORATION WISCONSIN POWER & LIGHT COMPANY MADISON GAS & ELECTRIC COMPANY 05

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RELOAD SAFETY EVALUATION FOR KEWAUNEE CYCLE 19 i

Prepared By:

Aarr) /2 #/M-Date:

/.,2/-93 i)ldclear Fu'el Teclinician Reviewed By:

I Moh Date:

/- 2 ~L '! E Nuclear Fuelingineer Reviewed By:

bT Date:

/- 22-93 r

Nucl Fuel Analysis Supervisor

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Date:

PM-G Reviewed By: directof-Nujear Fuel Reviewed By:

A Date:

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'8 8 Superintendent-Nuclear Licensing and Systems Reviewed By:

b/N Date:

J -' / - f 3 Plant Operations Review Committee

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Approved By:

/ N / [.'r W.u-h

' Date:

22 ' Y 7 I

Vice I residentherg/ Suppl / '

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,7 TABLE OF CONTENTS l '.0 S U M M ARY...........................

......................I j

2.0 CORE DESIGN

..............................................-.3-i 2.1 Core Description..........................................

3.

1 1

2.2 Operating Conditions, Limits, and Design Objectives....................... 6 '

i 2.3 Scram Worth Insenion Rate..................................... 12.

2.4 Sh utdown Window......................................... 14 j

I 3.0 ACCIDENT EVALUATIONS....................

........ 16 I

3.1 Evaluation of Uncontrolled Rod Withdrawal from Suberitical................ 19 3.2 Evaluation of Uncontrolled Rod Withdrawal at Power....................21 3.3 Evaluation of Control Rod Misalignment

.......23 3.4 Evaluation of Dropped Rod

............. 25 '

3.5 Evaluation of Uncontrolled Boron Dilution........................... 27 3.6 Evaluation of Startup of an Inactive Loop..........

.... 29_

3.7 Evaluation of Feedwater System Malfunction.......................... 31 3.8 Evaluation of Excessive Load Increase.............................. 33.

3.9 Evaluation of Loss of Load.........................

............ 35 3.10 Evaluation of Loss of Normal Feedwater......

....................37-3.11 Evaluation of Loss of Reactor Coolant Flow Due to Pump Trip............... 38 3.12 Evaluation of Loss of Reactor Coolant Flow Due to Locked Rotor

...... 40 ~

3.13 Evaluation of Main Steam Line Break.............................. -. 42 '

3.14 Evaluation of Rod Ejection Accidents.............................. 45 -

3.15 Evaluation of Fuel IIandling Accident............................. 50 3.16 Evaluation of Loss of Coolant Accident...........

.................52 3.17 Power Distribution Control Verification........................... 54-

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TABLE OF CONTENTS (CONTINUED) i 4.0 TECHNICAL SPECIFICATIONS.................................... 56 5.0 STATISTICS UPD ATE......................................... 57

6. 0 REFEREN CES............................................... 60 t

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c LIST OF TABLES TABLE PAGE 2.1.1 Cycle 19 Fuel Characteristics

.....................................-4 t

2.4.1 Peaking Factor Versus Cycle 18 Shutdown Burnup...................... 15 3.0.1 Kewaunee Nuclear Power Plant List of Safety Analyses................... 17 t

3.0.2 Safety Analyses Bounding Values.................................. 13 3.1.1 Uncontrolled Rod Withdrawal from Subcritical......................... 20 3.2.1 Uncontrolled Rod Withdrawal at Power...,.......................... 22 3.3.1 Control Rod Misalignment......................................24.

t 3.4.1 D ropped Rod............................................ 26

[

3.5.1 Uncontrolled Boron Dilution............................

.......28 3.6.1 Startup of an '.nactive Loop....................

...............30 3.7.1 Feedwater System Malfunction

.... 32 3.8.1 Excessive Imad' Increase

.....................................34 3.9.1 Loss o f Load.................................

...........36 3.11.1 Imss of Reactor Coolant Flow Due to Pump Trip....................... 39 3.12.1 Loss of Reactor Coolant Flow Due to Locked Rotor

................41 3.13.1 Main Steam Line Break..................................

... 43 i

i 3.'4.1 Rod Ejection Accident at HFP, BOC............................... 46 jl 3.14.2 Rod Ejection Accident at HZP, BOC..........

....................47 3.14.3 Rod Ejection Accident at HFP, EOC................................ 48 1

t 3.14.4 Rod Ejection Accident at HZP, EOC..........

....................49 3.15.1 Fuel Handling Accident...................................... 51' 3.16.1 Loss of Coolant Accident....................................... 53 5.0.1 Reliability Factors 58 5.0.2 FQN Reliability Factors.....

............... 5 9 l

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- 111 -

LIST OF FIGURES

. FIGURE PAGE

-2.1.1 Cycle 19 Loading Pattern......................................

5 2.2.1 Hot Channel Factor Normalized Operating Envelope.....................

9 2.2.2 Control Bank Insertion Limits................................... 10 2.2.3 Target Band on Indicated Flux Difference as a Function of Operating Power Level (Typical)

...............................11 2.3.1 Scram Reactivity Insertion Rate.................................. 13 3.13.1 Variation of Reactivity with Core Temperature at 1000 PSIA for the End of. Life Rodded Core with One Rod Stuck (Zero Power).............. 44 3.17.1 Maximum (FQ

• P REL) vs Axial Core Height, Cycle 19................ 55 a

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1.0

SUMMARY

The Kewaunee Nuclear Power Plant is scheduled to shut down for the Cycle 18-19 refueling in March 1993. Startup of Cycle 19 is forecast for April 1993.

This report presents an evaluation of the Cycle 19 reload and demonstrates that the reload-will not adversely affect the safety of the plant. Those accidents which could potentially -

be affected by the reload core design are reviewed.

5 Details of the calculational model used to generate physics parameters for this Reload t

Safety Evaluation are described in References 1 and 15. Accident Evaluation j

t methodologies applied in this report are detailed in Reference 2. These reports have

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been previously reviewed and approved by the NRC as shown in References 3 and 4.

The current physics model reliability factors are discussed in Section 5 of this report.

i An evaluation, by accident, of the pertinent reactor parameters is performed by comparing the reload analysis results with the current bounding safety analysis values. The j

evaluations performed in this document employ the current Technical Specification j

r (Reference 5) limiting safety system setpoints and operating limits.

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it is concluded that the Cycle 19 design is more conservative than results of previously.

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-I docketed accident analyses and implernentation of this design will not introduce an

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unreviewed safety question since:

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the probability of occurrence or the consequences of an accident will not be increased, f

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2. the possibility for an accident or malfunction of a different type than any evaluated j

previously in the safety analysis report will not be created and, I

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3. the margin of safety as defined in the basis for any technical specification will not be reduced.

This conclusion is based on these assumptions: There is adherence to plant operating limitations and Technical Specifications (Reference 5), and Cycle 18 is shut down within a

+300 MWD /MTU, -250 MWD /MTU window of the nominal design End of Cycle (EOC) burnup of 11,000 MWD /MTU.

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L 2.0 CORE DESIGN 2.1 Core Description -

The reactor core consists of 121 fuel assemblies of 14 x 14 design. The core loading pattern, assembly identification, control rod bank identification, instrument thimble I.D., thermocouple I.D., and burnable poison rod configurations _for Cycle 19 are presented in Figure 2.1.1.

Thirty-two (32) new Siemens Power Corporation (SPC) standard assemblies enriched to 3.4 w/o U235 and four (4) new Westinghouse Electric OFA assemblies enriched to-3.1 w/o U235 will reside with 85 partially depleted SPC assemblies. Table 2.1.1

' displays the core breakdown by region, enrichment, and number of previous duty cycles. Reference 6 describes the SPC 14 x 14 design. References 16 and 17 describe the Westinghouse OFA design.

The Cycle 19 reload core will employ 24 burnable poison rod assemblies (BPRAs) i containing 144 fresh and 144 partially depleted burnable poison rods. Fuel assemblies j

with two or three previous duty cycles are loaded on the core periphery flat region to lower power in that region and reduce reactor vessel fluence (see Reference 14) in the critical reactor vessel locations.

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Table 2.1.1 Cycle 19 Fuel Characteristics

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Number of Region Initial Previous Region Identifier W/O U235 Duty Cycles Assemblies 13 M

3.400 3

1 17 S

3.500 3

8 I8 T

3.400 2

8 18 T

3.500 2

8 19 U

3.460 2

28 20 W

3.400 1

32 21 X

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32 21 X

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CYCLE 19 LOADING PATTERN

2.2 Operating Conditions, Limits, and Design Objectives Cycle 19 core design is based on the following operating conditions, limits, and design objectives.

2.2.1 Operating Conditions

- Power Rating 1650 MWTH

- System Pressure 2250 PSIA

- Core Average Moderator Temperature (HZP) 547 F

- Core Average Moderator Temperature (HFP) 562 *F 2.2.2 Operating Limits A. Nuclear peaking factor limits are as follows:

(i) FQ(Z) limits a)

For SPC standard fuel:

FQ(Z)1(2.28/P)

• K(Z) for P > 0.5 l

FQ(Z)14.56

• K(Z) for P i 0.5 l

K(Z) is the function given in Figure 2.2.1 Z is the core height b)

For Westinghouse OFA fuel, the FQ(Z) limit is the SPC standard fuel limit less 10% (Reference 19).

l (ii) FAH limits FAHN < l.55 (1 + 0.2(1-P))

Where P is the fraction of full power at which the core is operating:

l B. The moderator temperature coefficient at operating conditions shall be negative.

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C. With the most reactive rod stuck out of the core, the remaining control rods shall be able to shut down the reactor by a suf5cient reactivity margin:

1.0% at Beginning of Cycle (BOC) 2.0% at End of Cycle (EOC)

D. The power dependent rod insertion limits (PDIL) are presented in Figure 2.2.2. These limits are those currently specified in Reference 5.

E. The indicated axial flux difference shall be maintained within a i 5% band about the target axial flux difference above 90 percent power. Figure 2.2.3 shows the axial flux difference limits as a function of core power. These limits are currently specified in Reference 5, which also provides limits on temporary operation allowed within the line 3.10.b.11.a envelope at power icvels between 50 percent and 90 percent.

F. At refueling conditions a boron concentration of 2l00 ppm will be suf5cient to maintain the reactor suberitical by 5 percent ak/k with all rods inserted and will maintain the core suberitical with all rods out.._.

2.2.3 Design Objectives A. The fuel loading pattern shall be capable of generating approximately 11,200 MWD /MTU based on a nominal end of Cycle 18 burnup of 11,000 MWD /MTU.

B. Fuel duty during this fuel cycle will assure peak fuel rod burnups less than the maximum bumup recommended by the fuel vendors.

C. The fuel loading pattern shall be a " lower" neutron leakage design in order to reduce vessel fluence in critical reactor vessel locations.

D. The Westinghouse Electric OFA assemblies will not be limiting with respect to power distribution and LOCA analysis assumptions (Reference 18).

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10 20 30 INDICATED AXfAL FLUX DiF F E R E NCE Target Band on Indicated Flux Difference As a function of Operating Power Level (Typical) _ - - - _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _

2.3 Scram Worth Insertion Rate The most limiting scram curve is that curve which represents the slowest trip reactivity insertion rate normalized to the minimum shutdown margin. The Cycle 19 minimum shutdown margin is 2.28 percent at end of cycle hot full power conditions.

Figure 2.3.1 compares the Cycle 19 minimum scram insertion curve to the current bounding safety analysis curve.

It is concluded that the minimum trip reactivity insertion rate for Cycle 19 is conservative with respect to the bounding value. Thus, for accidents in which credit is taken for a reactor trip, the proposed reload core will not adversely affect the results of the safety analysis due to trip reactivity assumptions. L.-

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SCRAM REACTIVITY INSERTION RATE 3 E3df,M".gsi.

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2.4 Shutdown Window An evaluation of the maximum full power equilibrium peaking factors versus EOC 18 burnup is presented in Table 2.4.1. The values shown have conservatisms applied in accordance with References I and 7.

It is concluded that if the refueling shutdown of Cycle 18 occurs within the burnup window, the Cycle 19 peaking factors will not be significantly affected and will not exceed their limiting values. -

e Table 2.4.1 Peaking Factor Versus Cycle 18 Shutdown Burnup Fall FQ Cycle 19 Limit Cycle 19 Limit EOC 18 - 250 MWD /MTU l.54 1.55 2.14 2.28 EOC 18 Nominal 1.52 1.55 2.14 2.28 EOC 18 + 300 MWD /MTU 1.51 1.55 2.15 2.28 _ _ -.

3.0 ACCIDENT EVALUATIONS Table 3.0.1 presents the latest safety analyses performed for the accidents which are evaluated in Sections 3.1 through 3.16 of this report. The bounding values derived from these analyses are shown in Table 3.0.2 and will be applied in the Cycle 19 accident evaluations.

( _ _ _ _. _..

Table 3.0.1 Kewaunee Nuclear Power Plant List of Safety Analyses Accident Current Analysis Ref. No.

Uncontrolled RCCA Withdrawal From a 2/78 (Cycle 4-RSE) 9 Suberitical Condition Uncontrolled RCCA Withdrawal at Power 2/78 (Cycle 4-RSE) 9 Control Rod Drop 7/1/91 (Rev. 9-USAR) 8 RCC AssemL y Misalignment 7/1/91 (Rev. 9-USAR) 8 CVCS Malfunction 1/27/71 (AM7-USAR) 8 Startup of an Inactive RC IAop 1/27/71 (AM7-US AR) 8 Excessive Heat Removal Due to FW System 1/27/71 (A M7-US AR) 8 Malfunctions Excessive lead increase Incident 1/27/71 (AM7-USAR) 8 Loss of Reactor Coolant Flow Due to Pump Trip 3/73 (WCAP-8092) 10 Due to Underfrequency 7/88 (Rev. 6-USAR) 8 Imeked Rotor Accident 2/78 (Cycle 4-RSE) 9 Loss of External Electrical Load 1/27/71 (AM7-US AR) 8 Loss of Normal Feedwater 8/31/73 (AM33-USAR) 8 Fuel Handling Accidents 1/27/71 (AM7-US AR) 8 Rupture of a Steam Pipe 4/13/73 (AM28-USAR) 8 Rupture of CR Drive Mechanism Housing 2/78 (Cycle 4-RSE) 9 RC System Pipe Rupture (LOCA) 12/10/76 (AM40-US AR) 8 Westinghouse Zire - Water Addendum 12/14/79 11 Clad Hoop Stress Addendum 1/8/80 12 Advanced Nuclear Fuels Corporation 10/01/84 (XN-NF-84-31, 13 Rev.1).

Table 3.0.2 Safety Analyses Bounding Values Parameter Lower Bound Upper Bound Units Moderator Temp. Coefficient

-40.0 0.0 pcm/*Fm Doppler Coefficient

-2.32

-1.0 pcm/ Ff Differential Boron Worth

-11.2

-7.7 pcm/ ppm Delayed Neutron Fraction

.00485

.00706 Prompt Neutron Lifetime 15 N/A psec Shutdown Margin 1.0 (BOC)

N/A

% Ap 2.0 (EOC)

N/A Differential Rod Worth of 2 N/A 82 pcm/sec Banks Moving Ejected Rod Cases HFP,BOL Beff

.0055 N/A Rod Worth N/A

.30

% Ap FQ N/A 5.03 HFP,EOL Beff

.0050 N/A Rod Worth N/A

.42

% Ap FQ N/A 5.1 H7.P,BOL Beff

.0055 N/A Rod Worth N/A

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% Ap FQ N/A 11.2 HZP,EOL Beff

.0050 N/A Rod Worth N/A

.92

% Ap FQ N/A 13.0

3.1 Evaluation of Uncontrolled Rod Withdrawal from Suberitical An uncontrolled addition of reactivity due to uncontrolled withdrawal of a Rod Cluster Control Assembly (RCCA) results in a power excursion.

The most important parameters are the reactivity insertion rate and the doppler coefficient. A maximum reactivity insertion rate produces a more severe transient while a minimum (absolute value) doppler coefficient maximizes the nuclear power peak. Of lesser concern are the moderator coefficient and delayed neutron fraction which are chosen to maximize the peak heat flux.

Table 3.1.1 presents a comparison of Cycle 19 physics parameters to the current safety analysis values for the Uncontrolled Rod Withdrawal from a Suberitical Condition.

Since the pertinent parameters from the proposed Cycle 19 reload core are conservatively bounded by those used in the current safety analysis, an uncontrolled rod withdrawal from subcritical accident will be less severe than the transient in the current analysis. The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant. -

Table 3.1.1 Uncontrolled Rod Withdrawal From Suberitical Reload Safety Parameter Evaluation Curreut Values Safety Analysis Units A) Moderator Temp.

-0.23 s

10.0 pcm/ Fm Coefficient B) Doppler Temp.

-1.31 s

-1.0 pcm/*Ff Coefficient l

C) Differential Rod Worth

.044 s

.116

$/sec of Two Moving Banks D) Scram Worth vs. Time See Section 2.3 E) Delayed Neutron

.00632 s

.00706 Fraction F) Prompt Neutron 28 2

15 see Lifetime 3.2 Evaluation of Uncontrolled Rod Withdrawal at Power An uncontrolled control rod bank withdrawal at power results in a gradual increase in core power followed by an increase in core heat flux. The resulting mismatch between core power and steam generator heat load results in an increase in reactor coolant temperature and pressure.

The minimum absolute value of the doppler and moderator coefficients serves to maximize peak neutron power, while the delayed neutron fraction is chosen to maximize peak heat flux.

Table 3.2.1 presents a comparison of the Cycle 19 physics parameters to the current safety analysis values for the Uncontrolled Rod Withdrawal at Power Accident.

Since the pertinent parameters from the proposed Cycle 19 reload core are conservatively bounded by those used in the current safety analysis, an uncontrolled rod withdrawal at power accident will be less severe than the transient in the current analysis. The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kehaunce Plant.

4 Table 3.2.1 Uncontrolled Rod Withdrawal at Power Reload Safety Parameter Evaluation Current Values Safety Analysis Units A) Moderator Temp.

-0.23 s

0.0 pcm/ Fm Coefficient B) Doppler Temp.

-1.31 s

-1.0 pcm/*Ff Coefficient C) Differential Rod Worth

.044 s

.116

$/sec of Two Moving Banks D) FaHN 1.52 s

1.55 E) Scram Worth vs. Time See Section 2.3 F) Delayed Neutron

.00632 s

.00706 Fraction -

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t 3.3 Evaluation of Control Rod Misalignment t

The static misalignment of an RCCA from its bank position does not cause a system transient; however, it does cause an adverse power distribution which is analyzed to j

show that core Departure from Nuclear Boiling Ratio (DNBR) limits are not

.f exceeded.

The limiting core parameter is the peak FAH in the worst case misalignment of Bank D fully inserted with one ofits RCCAs fully withdrawn at full power.

l Table 3.3.1 presents a comparison of the Cycle 19 FAHN versus the current safety analysis FAH limit for the Misaligned Rod Accident.

Sincc the pertinent parameter from the proposed Cycle 19 reload core is conservatively bounded by that used in the current safety analysis, a control rod misalignment accident will be less severe than the transient in the current analysis.

The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.

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Table 3.3.1 Control Rod Misalignment Reload Safety Current Parameter Evaluation Value Safety Analysis A) FAHN 1,gl 5

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3.4 Evaluation of Dropped Rod i

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The release of a full length control rod, or control rod bank by the gripper coils while I

i the reactor is at power, causes the reactor to become suberitical and produces a l

mismatch between core power and turbine demand. The dropping of any control rod bank will produce a negative neutron flux rate trip. with no resulting decrease in 3

i thermal margins. Dropping of a single RCCA or several RCCA's from the same bank i

may or may not result in a negative rate trip, and therefore the radial power distribution must be considered.

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Table 3.4.1 presents a comparison of the Cycle 19 physics parameters to the current i

safety analysis values for the Dropped Rod Accident.

i Since the pertinent parameters from the proposed Cycle 19 reload core are I

conservatnely bounded by that used in the current safety analysis, a dropped rod ~

accident will be less severe than the transient in the current analysis. The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.

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Table 3.4.1 -

Dropped Rod t

Reload Safety Current i

Parameter Evaluation Value Safety Analysis Units A) FAHN 1.52 s

1.55 B) Doppler Temp.

-1.31 s

-1.0 pcm/*Ff Coefficient C) Delayed Neutron

.00632 s

.00706 Fraction D) Excore Tilt

.82 2

.80 (Control)

E) Full Power Insertion 305 s

400 pcm Limit Worth (BOL)

I F) Full Power Insertion 430 s

450 pcm Limit Worth (EOL)

G) Moderator

-6.23 s

0.0 pcm/ Fm Temperature Coefficient (BOL)

11) Moderator

-20.01 s

-17.0 pcm/ *Fm Temperature Coefficient (EOL) s

> r

... =

i 1

?

3.5 Evaluation of Uncontrolled Boron Dilution i

The malfunction of the Chemical and Volume Control System (CVCS) is assumed to deliver unborated water to the Reactor Coolant System (RCS).

i i

Although the boron dilution rate and shutdown margin are the key parameters in this j

t event, additional parameters are evaluated for the manual reactor control case. In this-j case core thermal limits are approached and the transient is terminated by a reactor trip on over-temperature AT.

j t

Table 3.5.1 presents a comparison of Cycle 19 physics analysis results to the current safety analysis values for the Uncontrolled Boron Dilution Accident for refueling and.

i full power core conditions.

Since the pertinent parameters from the proposed Cycle 19 reload core are conservatively bounded by those used in the current safety analysis, an uncontrolled boron dilution accident will be less severe than the transient in the current analysis.

l The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.

-i i

i

.)

i 1

t i

Table 3.5.1 Uncontrolled Boron Dilution i

Reload Safety

' Current Evaluation Safety Parameter Values Analysis Units i

i) Refueling Conditions A) Shutdown Margin 11.I 2:

5.0 ii) At-Power Conditions t

A) Moderator Temp. Coefficient

-0.23 s

0.0 pcm/*Fm

_j B) Doppler Temp. Coefficient

-1.32

.s

-1.0 pcm/*Ff l

C) Reactivity Insertion Rate by Boron

.0022-s

.0023 S/sec l

D) Shutdown Margin 2,28 2

1.00 t

E) FAHN 1.52 s

1.55 i

F) Delayed Neutron Fraction

.00632 s

.00706 l

l

~

t l

l !

c.

l 3.6 Evaluation of Startup of an Inactive Loop The startup of an idle reactor coolant pump in an operating plant would result in the' q

injection of cold water (from the idle loop hot leg) into the core which causes a rapid f

reactivity insertion and subsequent core power increase.

]

The moderator temperature coefficient is chosen to maximize the reactivity effect of -

i the cold water injection. Doppler temperature coefficient is chosen conservatively low l

(absolute value) to maximize the nuclear power rise. The power distribution (FAH) is a

L used to evaluate the core thermal limit acceptability.

~

i Table 3.6.1 presents a comparison of the Cycle 19 physics calculation results to the I'

current safety analysis values for the Startup of an inactive Loop Accident.

a Since the pertinent parameters from the proposed Cycle 19 reload core are conservatively bounded by those used in the current safety analysis,' the startup of an.

inactive loop accident will be less severe than the transient in the current analysis.

i The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.

l l

1 I

t Table 3.6.1 i

Startup of an Inactive Loop l

i i

t Reload Safety' Current Parameter Evaluation Values Safety Analysis Units i

A) Moderator Temp.

-33.21 2

-40.0 pcm/*Fm Coefficient i

B) Doppler Coefficient

-1.87 s

-1.0 pcm/*Ff I

C) FAHN 1.52 s

1.55 i

i I

e E

f e

p 9,

3.7 Evaluation of Feedwater System Malfunction The malfunction of the feedwater system such that the feedwater temperature is decreased or the flow is increased causes a decrease in the RCS temperature and an i

attendant increase in core power level due to negative reactivity coef5cients and/or control system action.

}

Minimum and maximum moderator coefficients are evaluated to simulate both BOC

?

i and EOC conditions. The doppler reactivity coefficient is chosen to maximize the nuclear power peak.

A comparison of Cycle 19 physics calculation results to the current safety analysis values for the Feedwater System Malfunction Accident is presented in Table 3.7.1.

i Since the pertinent parameters from the proposed Cycle 19 reload core are conservatively bounded by those used in the current safety analysis, a feedwater r

system malfunction will be less severe than the transient in the current analysis. The implementation of the Cycle 19 reload core design, therefore, will not adversely affect i

the safe operation of the Kewaunee Plant.

[

i r

t f

l i i

Table 3.7.1 Feedwater System Malfunction Reload i

Safety Current i

Evaluation Safety Parameter Values '

Analysis U.its l

i) Beginning of Cycle A) Moderator Temp. Coef5cient

-6.23 s

0.0 pcm/*Fm 7

3) Doppler Temp. CoefGcient

-1.32 s

-1.0 pcm/*Ff ii) End of Cycle A) Moderator Temp. Coef6cient

-28.26 2

-40.0 pcm/*Fm I

B) Doppler Temp, Coef5cient

-1.31

's

-1.0 pcm/*Ff iii) Beginning and End of Cycle C) FAHN 1.52 s

1.55 f

I e

h r

I l

32 -

l

4 Table 3.8.1 Excessive Load Increase i

Reload Safety Current Evaluation Safety Parameter Values Analysis Units i) Beginning of Cycle A) Moderator Temp. Coefficient

-6.23 s

0.0 pcm/ Fm t

B) Doppler Temp. Coefficient

-1.32

's

-1.0 pcm/"Ff ii) End of Cycle A) Moderator Temp. Coefficient

-28.26 2

-40.0 pcm/ Fm B) Doppler Temp. Coefficient

-1.31 s

-1.0 pcm/ Ff iii) Beginning and End of Cycle C) FAHN 1.52 s

1.55 3

[

C P

k i

4 f -

3.8 Evaluation of Excessive Load Increase An excessive load increase causes a rapid increase in steam generator steam flow.

The resulting mismatch between core heat generation and secondary side load demand' results in a decrease in reactor coolant temperature which causes a core power increase due to negative moderator feedback and/or control system action.

J This event results in a similar transient as that described for the feedwater system-

'I malfunction and is therefore sensitive to the same parameters.

-f Table 3.8.1 presents a comparison of Cycle 19 physics results to the current safety l

analysis values for the Excessive Load Increase Accident.

Since the peninent parameters from the proposed Cycle 19 reload core are conservatively bounded by those used in the current safety analysis, an excessive load increase accident will be less severe than the transient in the current analysis. The l

'I implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.

l l

9 l

I i,

f

.I

3.9 Evaluation ofless ofImad A loss of load is encountered through a turbine trip or complete loss of external electric load. To provide a conservative assessment of this event, no credit is taken for direct turbine / reactor trip, steam bypass, or pressurizer pressure control, and the result is a rapid rise in steam generator shell side pressure and reactor coolant system temperature.

1 Minimum and maximum moderator coefficients are evaluated to simulate both BOC and EOC conditions. The doppler reactivity coefficient is chosen to maximize the nuclear power and heat flux transient. The power distribution (FAH) and scram reactivity are evaluated to ensure thermal margins are maintained by the reactor protection system.

A comparison of Cycle 19 physics parameters to the current safety analysis values for the Imss of Load Accident is presented in Table 3.9.1.

l

)

1

-i Since the pertinent parameters from the proposed Cycle 19 reload core are conservatively bounded by those used in the current safety analysis, a loss of load i

accident will be less severe than the transient in the current analysis. The j

implementation of the Cycle 19 reload core design, therefore, will not adversely affect I

the safe operation of the Kewaunee Plant.

I

Table 3.9.1 Loss of Load Reload Safety Current Evaluation Safety Parameter Values Analysis Units i) Beginning of Cycle i

l A) Moderator Temp. Coefficient

-6.23 s

0.0 pcm/*Fm B) Doppler Temp. Coefficient

-1.61 2

-2.32 pcm/*Ff i

ii) End of Cycle A) Moderator Temp. Coefficient

-28.26 2

-40.0 pcm/*Fm B) Doppler Temp. Coefficient

-1.60 2

-2.32 pcm/*Ff I

iii) Beginning and End of Cycle I

C) FAHN 1.52 s

1.55 D) Scram Worth Versus Time See Section 2.3.

3.11 Evaluation of Loss of Reactor Coolant Flow Due to Pump Trip The simultaneous loss of power or frequency decay in the electrical buses feeding the reactor coolant pumps results in a loss of driving head and a flow coast down. The effect of reduced coolant flow is a rapid increase in core coolant temperature.' The reactor is tripped by one of several diverse and redundant signals before thermal

)

hydraulic conditions approach those which could result in fuel damage.

l The doppler temperature coefficient is compared to the most negative value since this :

]

results in the slowest neutron power decay after trip. The moderator temperature coefficient is least negative to cause a larger power rise prior to the trip. Trip i

reactivity and FAH are evaluated to ensure core thermal margin.

l Table 3.11.1 presents a comparison of Cycle 19 calculated physics parameters to the i

current safety analysis values for the Loss of Reactor Coolant Flow Due to Pump Trip _

j Accident.

Since the pertinent parameters from the proposed Cycle 19 reload core are conservatively bounded by those used in the current safety analysis, a loss of reactor coolant flow due to pump trip accident will be less severe than the transient in the.

current analysis. The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant. _

3.10 Evaluation of Loss of Normal Feedwater A complete loss of normal feedwater is assumed to occur due to pump failures or valve malfunctions. An additional conservatism is applied by assuming the reactor coolant pumps are tripped, further degrading the heat transfer capability of the steam generators. When analyzed in this manner, the accident corresponds to a loss of offsite power.

The short term effects of the transient are covered by the Loss of Flow Evaluation (Sec. 3.11), while the long term effects, driven by decay heat, and assuming auxiliary feedwater additions and natural circulation RCS flow, have been shown not to produce any adverse core conditions.

The Loss of Feedwater Transient is not sensitive to core physics parameters and I

therefore no comparisons will be made for the Reload Safety Evaluation. -__

Table 3.11.1 Imss of Reactor Coolant Flow Due to Pump Trip Reload Safety Current Evaluation Safety Parameter Values Analysis Units A) Moderator Temp. Coefficient

-6.23 s

0.0 pcm/*Fm B) Doppler Temp. Coefficient

-1.61 2

-2.32 pcm/"Ff C) FAHN 1.52 s

1.55 D) Scram Worth Versus Time See Section 2.3 E) Fuel Temperatme 2030 s

2100

• F _ __

3.12 Evaluation of Imss of Reactor Coolant Flow Due to Locked Rotor This accident is an instantaneous seizure of the rotor of a single reactor coolant pump resulting in a rapid flow reduction in the affected loop. The sudden decrease in flow results in DNB in some fuel rods.

The minimum (absolute value) moderator temperature coefficient results in the least reduction of core power during the initial transient. The large negative doppler -

temperature coefficient causes a slower neutron flux decay following the trip as does the large delayed neutron fraction.

I Table 3.12.1 presents a comparison of Cycle 19 physics parameters to the current.

i safety analysis values for the Locked Rotor Accident.

j Sfnce the pertinent parameters from the proposed Cycle 19 reload core are I

conservatively bounded by those used in the current safety analysis, a locked rotor accident will be less severe than the transient in the current analysis. The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant. -

+

Table 3.12.1 Loss of Reactor Coolant Flow Due to Locked Rotor Reload Safety Current Evaluation Safety Parameter Values Analysis Units A) Moderator Temp. Coefficient

-6.23 s

0.0 pcm/*Fm B) Doppler Temp. Coefficient

-1.61 2

-2.32 pcm/ Ff C) Delayed Neutron Fraction

.00632 s

.00706 D) Percent Pins > Limiting FAHN 28.40 s

40.0 (DNBR = 1.3)

E) Scram Worth Versus Time See Section 2.3 F) FQ 2.15 s

2.28 G) Fuel Temperature 2030 s

2100

• F I

i........ _...

3.13 Evaluation of Main Steam Line Break The break of a main steam line inside containment at the exit of the steam generator causes an uncontrolled steam release and a reduction in primary system temperature and pressure. The negative moderator coefficient produces a positive reactivity insertion and a potential return to criticality after the trip. The doppler coe ficient is f

chosen to maximize the power increase.

Shutdown margin at the initiation of the cooldown and reactivity insertion and peak rod power (FAH) during the cooldown are evaluated for this event. The ability of the safety injection system to insert negative reactivity and reduce power is minimized by using the least negative boron worth coefficient.

Table 3.13.1 presents a comparison of Cycle 19 calculated physics parameters to the current safety analysis values for the main steam line break accident. Figure 3.13.1 compares core Keff during the cooldown to the current bounding safety analysis cun'e.

Since the pertinent parameters from the proposed Cycle 19 reload core are conservatively bounded by those used in the current safety analysis, a main steam line break accident will be less severe than the transient in the current analysis. The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.

Table 3.13.1 Main Steam Line Break Reload Safety Current Evaluation Safety Parameter Values Analysis Units A) Shutdown Margin 2.28 2

2.00

%Ap B) FAH 4.09 s

6.6 C) Doppler Temp. Coef6cient

-1.31 s

-1.0 pcm/ Ff D) Boron Worth Coefficient

-7.9 s

-7.7 pcm/ ppm -_

Figure 3.13.1 VARIATION OF REACTIVITY, WITH CORE TEMPERATURE AT 1000 PSIR FOR THE END OF LIFE RODDED CORE WITH ONE ROD STUCK (ZERO POWER) 5 N

.n

~ '

\\

~

A

\\

i

'N a

s x

b

\\

ll) kN N

(

O USAR O

WPS CYCLE 19 3'50.00 400.00 450.00 500.00 550.00 600.00 300.00 CORE AVERAGE TEMPERATURE

[DEG F) _ _ - - - - _ _ __ _ _ _ _

3.14 Evaluation of Rod Ejection Accidents The ejected rod accident is defined as a failure of a control rod drive pressure housing-followed by the ejection of a RCCA by the reactor coolant system pressure.

Tables 3.14.1 through 3.14.4 present the comparison of Cycle 19 calculated physics parameters to the current safety analysis values for the Rod Ejection Accident at zero and full power, BOC and EOC core conditions.

Since the pertinent parameters from the proposed Cycle 19 reload core are -

conservatively bounded by those used in the current safety analysis, a rod ejection accident will be less severe than the transient in the current analysis. The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.........

Table 3.14.1 Rod Ejection Accident at HFP,BOC Reload Safety Current Evaluation Safety Parameter Values Analysis Units A) Moderator Temp. Coef6cient

-6.23 s

0.0 pcm/*Fm B) Delayed Neutron Fraction

.00595 2

.00550 i

C) Ejected Rod Worth

.06 s

0.30

%Ap 1

l D) Doppler Temp. Coef6cient

-1.32 s

-1.0 pcm/ Ff E) Prompt Neutron Lifetime 28.3 2

15.0 psec F) FQN 2.13 s

5.03 l

G) Scram Worth Versus Time See Section 2.3 -____ __

Table 3.14.2 Rod Ejection Accident at HZP,BOC Reload Safety Current Evaluation Safety Paranieter Values Analysis Units A) Moderator Temp. Coefficient

.23 s

0.0 pcm/ Fm B) Delayed Neutron Fraction

.00595 2

.00550 C) Ejected Rod Worth 0.43 s

0.91

%Ap D) Doppler Temp. Coefficient

-2.19 s

-1.0 pcm/*Ff E) Prompt Neutron Lifetime 28.31 2

15.0 sec F)FQN 4.32 s

11.2 G) Scram Worth Versus Time See Section 2.3 _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - - _ _ _

Table 3.14.3 Rod Ejection Accident at HFP,EOC Reload Safety Current Evaluation Safety Parameter Values Analysis Units A) Moderator Temp. Coefficient

-20.01 s

0.0 pcm/ Fm B) Delayed Neutron Fraction

.00537 2

.00500 C) Ejected Rod Worth 0.10 s

0.42

%dp D) Doppler Temp. Coefficient

-1.31 s

-1.0 pcm/'Ff E) Prompt Neutron Lifetime 31.0 2

15.0 psec F) FQN 2.52 s

5.1 G) Scram Worth Versus Time See Section 2.3 Table 3.14.4 Rod Ejection Accident at HZP,EOC Reload Safety Current Evaluation Safety Parameter Values Analysis Units A) Moderator Temp. Coefficient

-6.48 s

0.0 pcm/ Fm B) Delayed Neutron Fraction

.00537 2

.00500 C) Ejected Rod Worth 0.65 s

0.92

%Ap D) Doppler Temp. Coefficient

-2.65 s

-1.0 pcm/*Ff E) Prompt Neutron Lifetime 31.0 2

15.0 sec F) FQN 7.25 s

13.0 G) Scram Worth Versus Time See Section 2.3. - _ _ _ - _ - _ _ _ _ _ _ _

3.15 Evaluation of Fuel Handling Accident This accident is the sudden release of the gaseous fission products held within the fuel cladding of one fuel assembly. The fraction of fission gas released is based on a conservative assumption of high power in the fuel rods during their last six weeks of operation.

The maximum FQ expected during this period is evaluated within the restrictions of the power distribution control procedures.

Table 3.15.1 presents a comparison of the maximum Cycle 19 FQN calculated during the last 2.0 GWD/MTU of the cycle, to the current safety analysis FQN limit for the Fuel Handling Accident.

Since the pertinent parameter from the proposed Cycle 19 reload core is conservatively bounded by that used in the current safety analysis, a fuel handling accident will be less severe than the accident in the current analysis. The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.

Table 3.15.1 Fuel Handling Accident Reload Safety Current Evaluation Safety Parameter Values Analysis A) FQN 1.99 s

2.53

3.16 Evaluation of Loss of Coolant Accident The less of Coolant Accident (LOCA) is defined as the rupture of the reactor coolant system piping or any line connected to the system, up to and including a double-ended guillotine rupture of the largest pipe.

The principal parameters which affect the results of LOCA analysis are the fuel stored energy, fuel rod internal pressures, and decay heat. These parameters are affected by the reload design dependent parameters shown in Table 3.16.1.

The initial conditions for the LOCA analyses are assured through limits on fuel design, fuel rod burnup, and power distribution control strategies.

Table 3.16.1 presents the comparison of Cycle 19 physics calculation results to the current safety analysis values for the Loss of Coolant Accident.

Since the pertinent parameters from the proposed Cycle 19 reload core are conservatively bounded by those used in the current safety analysis, a loss of coolant accident will be less severe than the transient in the current analysis. The implementation of the Cycle 19 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant. _

Table 3.16.1 Loss of Coolant Accident Reload Safety Current Evaluation Safety Parameter Values Analysis A) Scram Wonh Versus Time See Section 2.3 B) FQ See Section 3.17 C) FAH 1.52 s

1.55 l

3.17 Power Distribution Control Verification The total peaking factor FQT relates the maximum local power density to the core average power density. The FQT is determined by both the radial and axial power distributions. The radial power distribution is nlatively fixed by the core loading pattern design. The axial power distribution is controlled by the procedures (Reference 7) described in Section 2.2 of this report.

Following these procedures, FQT(Z) are determined by calculations performed at full power, equilibrium core conditions, at exposures ranging from 13OC to EOC.

Conservative factors which account for potential power distribution variations allowed by the power distribution control procedures, manufacturing tolerances, and measurement uncertainties are applied to the calculated FQT(Z).

Figure 3.17.1 compares the calculated FQT(Z), including uncertainty factors, to the FQT(Z) limits. These results demonstrate that the power distributions expected during Cycle 19 operation will not preclude full power operation under the power distribution control specifications currently applied (Reference 5).. _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ __ - _ _ _ _.

Figure 3.17.1 MAX (F0 m P REL ) VS RXlRL CORE HEIGHT CYCLE 19 S3D 92336.1029 CORE HEIGHT (INCHES) 3 9 15 21 27 33 39 45 51 57 63 59 75 BI B7 93 99 105 Ill 117 123 124 135 141 g

o I

ru

~-

~

~,

~

()

(#

()

g3

(,

g)

y N

p O

o

()

(*

(1 o

($

0 I

X g

TT n

GE o

R T

(1

[

I?

8

.o 24 23 ?? ?! PD 19 18 17 16 15 14 13 12 11 10 09 DB D7 06 05 04 03 02 01 a

AX1AL POINT _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. __.

4.0 TECHNICAL SPECIFICATIONS No Technical Specification changes are required as a result of this reload.

5 _ _ - - - _ _ - _ - _ - _ - _ _ _ _ _

l.

5.0 STATISTICS UPDATE Measurements and calculations of Cycles 15,16, and 17 are incorporated into the FQN and FAH statistics data base. The moderator temperature coefficient statistics data base includes results from Cycles 13 through 18. The reliability and bias factors used for the Cycle 19 Reload Safety Analyses are presented in Tables 5.0.1 and 5.0.2. c_________

Table 5.0.1 Reliability Factors Parameter Reliability Factor Bias FQN See Table 5.0.2 FAH 4.06 %

0 Rod Worth 10.0 %

0 Moderator Temperature 2.3 pcm/*F 2.9 pcm/*F Coefficient Doppler Coefficient 10.0 %

0 Boron Worth 5.0 %

0 Delayed Neutron Parameters 3.0%

0 i _.....

Table 5.0.2 FQN Reliability Factors Core I2 vel oNode RF (%)

1 (Bottom)

.0775 13.39 2

.0587 10.25 3

.0219 4.61 4

.0272 5.34 5

.0217 4.59 6

.0225 4.69 7

.0229 4.74 8

.0228 4.73 9

.0241 4.91 10

.0180 4.13 11

.0174 4.06 12

.0173 4.05 13

.0174 4.06 14

.0176 4.08 15

.0173 4.05 16

.0186 4.19 17

.0208 4.47 18

.0200 4.37 19

.0262 5.19 20

.0243 4.93 21

.0462 8.23 22

.0337 6.29 23

.0836 14.41 24 (Top)

.0667 11.58

6.0 REFERENCES

s 1.

Wisconsin Public Service Corporation, Kewaunee Nuclear Power Plant, topical F

.i report entitled, " Qualification of Reactor Physics Methods for Application to Kewaunee," dated September 29,1978.

(

2.

Wisconsin Public Service Corporation, Kewaunee Nuclear Power Plant, topical report WPSitSEM-NP-A entitled, " Reload Safety Evaluation Methods for-Application to Kewaunee," Revision 2, dated October 1988.

-i 3.

Safety Evaluation Report by the Office of Nuclear Reactor Regulation:

" Qualification of Reactor Physics Methods for Application to Kewaunee,"-

October 22,1979.

1 4.

Safety Evaluation Report by the Office of Nuclear Reactor Regulation: " Reload Safety Evaluation Methods for Application to Kewaunec," April 1988.

i

.t 5.

Wisconsin Public Service Corporation Technical Specifications for the Kewaunee Nuclear Power Plant. Docket Number 50-305, Amendment No. 96, dated October 14, 1992.

l I

6.

Exxon Nuclear Company, " Generic Mechanical and Thermal Hydraulic Design for Exxon Nuclear 14 x 14 Reload Fuel Assemblies with Zircaloy Guide Tubes i

for Westinghouse 2-Loop Pressurized Water Reactors," November 1978.

i l

I 7.

R. J. Burnside and J.- S. Holm, " Exxon Nuclear Power Distribution Control for i

Pressurized Water Reactors, Phase II," XN-NF-77-57, Exxon Nuclear Company,

{

Inc., January 1978.

f

- h l

8.

Wisconsin Public 3ervice Corporation, Kewaunee Nuclear Power Plant, Updated i

Safety Analysis Report, Revision 9, dated July 1,1991.

9.

" Reload Safety Evaluation," for Kewaunee Nuclear Power Plant Cycles 2,3, and 4.

r

10. WCAP 8092, " Fuel Densification Kewaunee Nuclear Power Plant," March 1973.

11. ECCS Reanalysis - ZlRC/ Water Reaction Calculation. Ietter from E. R.

Mathews to A. Schwencer, December 14, 1979.

l

12. Clad Swelling and Fuel Blockage Models. Ixtter from E. R. Mathews to f

f D. G. Eisenhut, January 8,1980.

t

13. "Kewaunee High Burnup Safety Analysis: Limiting Break Loca & Radiological l

t Consequences," XN-NF-84-31, Revision 1, Exxon Nuclear Company, Inc.,

October 1,1984.

+

5

+

9..

I

14. NRC letter 89-061, from C. R. Steinhardt to U.S. NRC Document Control j

Desk, May 12,1989.

15. " Reload Safety Evaluation, Appendix A," for Kewaunee Nuclear Power Plant Cycle 17, February 1991.

16. I.etter from J. C. Miller, Westinghouse, to S. F. Wozniak, Wisconsin Public Service Corporation, dated November 24,1992, "Kewaunee (WPGQ)".

17. Westinghouse 'etter PSA-91-233 from D. J. Cesare to S. F. Wozniak dated i

November 11, 1991.

18. WPS letter from C. R. Steinhardt to U.S. Nuclear Regulatory Commission, t

Docket 50-305, dated June 19,1991 " Core Reloads of Advanced Design Fuel Assemblies".

+

r

19. SPC letter from H. G. Shaw (SPC) to S. F. Wozniak dated December 11, 1992,

" Disposition of LBLOCA Analysis for Kewaunce with Four Westinghouse Lead -

[

'l' Assemblies".

h r

i

?

L l

?