ML20212B945
| ML20212B945 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 12/31/1986 |
| From: | Coen E, Holly J, Wozniak S WISCONSIN PUBLIC SERVICE CORP. |
| To: | |
| Shared Package | |
| ML20212B862 | List: |
| References | |
| NUDOCS 8612290349 | |
| Download: ML20212B945 (66) | |
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hj RELOAD SAFETY EVALUATION FOR KEWAUNEE CYCLE 13 s
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(q Prepared By: 4Ah.Th hd#v Date: 10 - Z 7- 76 Nuclear Fuel Engineer Reviewed By: e 7- //o ( , Date: /(8.2'/-8(y _
Nuc[earFuelAnalysisSupervisor ,
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-'NuclearFuciCydfcppervisor .., i
. Reviewed By: CfA4/ Date: //~ /7' Ib L 031ng & Systems Superintendent Reviewed By: d41# M Date /2 -d-d[o N K d6YOO Plant Oper ohsRefewCpfttee Approved By: A Cf 1. . Date [ 2 '/#7 b . _ ,,
Direct'or - Fuel f vices 0
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h TABLE OF CONTENTS t +
U 1
1.0 INTRODUCTION
3 2.0 CORE DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2.1 Core Description . . . . . . . . . . . . . . . . . . . . . . . 3 2.2 Design Objectives and Operating Limits . . . . . . . . . . . . 116 2.3 Scram Worth Insertion Rate . . . . . . . . . . . . . . . . . 12 2.4 Shutdown Window . . . . . . . . . . . . . . . . . . . . . . .
15
- 3.0 ACCIDENT EVALUATIONS. . . . . . . . . . . . . . . . . . . . . . .
3.1 Evaluation of Uncontrolled Rod Withdrawal from 18 Subcritical. . . . . . . . . . . . . . . . . . . . . . . . .
20 3.2 Evaluation of Uncontrolled Rod Withdrawal at Power . . . . . 22 3.3 Evaluation of Control Rod Misalignment . . . . . . . . . . . 24 3.4 Evaluation of Dropped Rod. . . . . . . . . . . . . . . . . .
26 3.5 Evaluation of Uncontrolled Baron Dilution. . . . . . . . . . 28 3.6 Evaluation of Startup of an Inactive Loop. . . . . . . . . .
3.7 Evaluation of Feedwater System Malfunction . . . . . . . . . 30 32 3.8 Evaluation of Excessive Load Increase. . . . . . . . . . . . 34 3.9 Evaluation of Loss of Load . . . . . . . . . . . . . . . . . 36 3.10 Evaluation of Loss of Normal Feedwater . . . . . . . . . . .
s 3.11 Evaluation of Loss of Reactor Coolant Flow Due to 37 P ump T rip . . . . . . . . . . . . . . . . . . . . . . . . . .
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k( 3.12 Evaluation of Loss of Recctor Coolant Flow Due to Locked Rotor . . . . . . . . . . . . . . . . . . . . . . . .
39 41 i 3.13 Evaluation of Main Steam Line Break. . . . . . . . . . . . . 44 3.14 Evaluation of Rod Ejection Accidents . . . . . . . . . . . . 49 3.15 Evaluation of Fuel Handling Accident . . . . . . . . . . . . 51 3.16 Evaluation of Loss of Coolant Accident . . . . . . . . . . . 53 3.17 Power Distribution Control Verification. . . . . . . . . . .
55 4.0 TECHNICAL SPECIFICATIONS. . . . . . . . . . . . . . . . . . . . .
56 5.0 STATISTICS UPDATE , . . . . . . . . . . . . . . . . . . . . . . .
59
6.0 REFERENCES
O V
LIST OF TABLES p- Table 2.1.1 Cycle 13 Fuel Characteristics . . . . . . . . . . . . . . 4 i 1 Table 2.4.1 Peaking Factor Sensitivity to Shutdown Window . . . . . . . . . . . . . . . . . . . . 14 Table 3.0.1 Kewaunee Nuclear Plant List of Safety Analyses . . . . . . . . . . . . . . . . 16 Table 3.0.2 Safety Analyses Bounding Values ........................17 Table 3.1.1 Comparison of Parameters for Uncontrolled Rod Withdrawal from Subcritical ...................19 Table 3.2.1 Comparison of Parameters for Uncontrolled Rod Withdrawal at Power .......................21 Table 3.3.1 Comparison of Parameters for Control Rod Misalignment . . . . . . . . . . . . . . . 23 Table 3.4.1 Comparison of Parameters for Dropped Rod Accident . . . . . . . . . . . . . . . . . . 25 Table 3.5.1 Comparison of Parameters for Uncontrolled Baron Dilution
(~ Accident . . . . . . . . . . . . . . . . . . . . . . . 27 Table 3.6.1 Comparison of Parameters for Startup of an Inactive Loop . . . . . . . . . . . . . . 29 Table 3.7.1 Comparison of Parameters for Feedwater System Malfunction . . . . . . . . . . . . . 31 Table 3.8.1 Comparison of. Parameters for Excessive Load Increase . . . . . . . . . . . . . . . . 33 Table 3.9.1 Comparison of Parameters for Loss o f Load . . . . . . . . . . . . . . . . . . . . . . . . 35 Table 3.11.1 Comparison of Parameters for Loss of Reactor Coolant Flow Due to Pump Trip . . . . . . . . . . . . . . . . . . . . . . . 38 Table 3.12.l Comparison of Parameters for Loss of Reactor Coolant Flow Due to Locked Rotor .....................40 Table 3.13.1 Comparison of Parameters for Main Steam Line Break. . . . . . . . . . . . . . . . . . . . 42 A-U i
,e Table 3.14.1 Comparison of Parameters for Rod f Ejection Accident at WP BOC .............45 i
Table 3.14.2 Comparison of Parameters for Rod Ejection Accident at HZP BOC .............46 Table 3.14.3 Comparison of Parameters for Rod 47 Ejection Accident at HFP EOC . . . . . . . . . . . . .
Table 3.14.4 Comparison of Parameters for Rod 48 Ejection Accident at HZP EOC . . . . . . . . . . . . .
Table 3.15.1 Comparison of Parameters for Fuel 50 Handling Accident . . . . . . . . . . . . . . . . . . .
'4 Table 3.16.1 Comparison of Parameters for Loss 52 of Coolant Accident . . . . . . . . . . . . . . . . . .
57 Table 5.0.1 Reliability Factors . . . . . . . . . . . . . . . . . .
58 i Table 5.0.2 FQN Reliability Factors . . . . . . . . . . . . . . . .
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LIST OF FIGURES
.O Figure 2.1.1 Cycle 13 Loading Pattern ................5 Figure 2.2.1 Hot Channel Factor Normalized Operating Envelope ...................8 Figure 2.2.2 Control Bank Insertion Limits . . . . . . . . . . . . . . 9 Figure 2.2.3 Target Band on Indicated Flux Difference ......................10 12 Figure 2.3.1 Scram Worth Versus Time . . . . . . . . . . . . . . . .
Figure 3.13.1 Variation of Reactivity with 43 Core Temperature at 1000 PSIA . . . . . . . . . . . . .
Figure 3.17.1 Maximum FQ Versus Axial Height, Power 54 Distribution Control Verification . . . . . . . . . . .
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1.0 Introduction The Kewaunee Nuclear Power Plant is scheduled to shutdown for the Cycle
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12-13 refueling in March of 1987. Startup of Cycle 13 is forecast for April 1987.
This report presents an evaluation of the Cycle 13 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.
Details of the calculational model used to generate physics parameters for this Reload Safety Evaluation are described in Reference (1). Accident Evaluation methodologies applied in this report are detailed in Reference (2). These reports have been previously reviewed (3). The current physics model reliability factors are discussed in Section 5 of this report.
An evaluation, by accident, of the pertinent reactor parameters is per-
{} formed by comparing the reload analysis results with the current bounding safety analysis values. The evaluations performed in this document employ the current Technical Specification (4) limiting safety system setpoints and operating limits.
It is concluded that the Cycle 13 design is more conservative than results of previously docketed accident analyses and implementation of this design will not introduce an unreviewed safety question since:
- 1) the probability of occurrence or the consequer.ces of an accident will not be increased,
- 2) the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report will not be created and, f- g D :
- 3) the margin of safety as defined in the basis for any technical specifi-cation will not be reduced. i This conclusion is based on these assumptions; Cycle 12 is shutdown within a
+ 500 MWD /MTU window of the nominal design EOC burnup, and there is adherence to plant operating limitations and Technical Specification (4).
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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, RCCA bank identification, instrument thimble I.D., thermocouple I.D., and burnable poison rod con-figurations for Cycle 13 are presented in Figure 2.1.1.
Forty new Exxon assemblies enriched to 3.4 w/o U235 will reside with eighty-one partially depleted Exxon assemblies. Table 2.1.1 displays the core breakdown by region, enrichment and previous cycle duty.
The Cycle 13 reload core will employ 24 burnable poison rod assemblies (BPRA'S) containing 160 fresh and 128 partially depleted burnable poison rods.
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Cycle 13 Fuel Characteristics i i i
Number of Number of 1 Initial Initial Previous Region Vendor W/0 U235 Duty Cycles Assemblies ,
3.2 3 1 10 ENC i ,
ENC 3.2 2 8 l 10 ENC 3.2 2 8 12 ,
ENC 3.4 2 32 13 i ENC 3.4 1 32 1 14 15 ENC 3.4 0 40 (FEED)
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FIGURE 2.1.1
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2.2 Design Objectives and Operating Limits Power Rating 1650 MWTH Sy:: tem Pressure 2250 PSIA Core Average Moderator Temperature (HZP) 547' F Core Average Moderator Temperature (HFP) 561* F Cycle 13 core design is based on the followng design objectives and operating limits.
A. Nuclear peaking factor limits are as follows:
(1) FQ(Z) limits
< (2.23/P)
- K(Z) for P > 0.5 FQ(Z)34.46*K(Z)forP<0.5 FQ(Z)
(ii) FoH limits .
F&N < 1.55(1 + 0.2(1-P))
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v Where P is the fraction of full power at which the core is operating:
K(Z) is the function given in Figure 2.2.1 Z is the core height location FQ B. The moderator temperature coefficient at operating conditions shall be negative.
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 sufficient reactivity l
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E. The power dependent rod insertion limits (PDIL) are presented in Figure 2.2.2. These limits are those currently specified in Reference 4.
F. The indicated axial flux difference shall be maintained within a + 5% _
band about the target axial flux difference above 90% power. Figure 2.2.3 shows the axial flux difference limits es a function of core power. These limits are currently specified in Reference 4.
G. A refueling boron concentration of 2100 ppm will be sufficient to main-tain the reactor subcritical by 10% Ak/k in the cold condition with all rods inserted c.nd will maintain the core subcritical with all rods out of the core.
H. Fuel duty during this fuel cycle will assure peak fuel rod burnups less than the maximum burnup recommended by the fuel vendor.
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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 13 minimum shutdown margin is 2.50% at end of cycle hot full power conditions. Figure 2.3.1 compares the Cycle 13 minimum scram inser-tion curve to the current bounding safety analysis curve.
It is concluded that the minimum trip reactivity insertion rate for Cycle 13 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 reac-tivity assumptions.
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1 5
1 FIGURE 2.3.1 1rm Scram Worth Versus Time
' b t
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. 2.4 Shutdown Window i
An evaluation of the full power equilibrium peaking factor variation at 80C 13 versus EOC 12 burnup is presented in Table 2.4.1. The values shown have conservatisms applied in accordance with references 1 and 9.
It is concluded that if refueling shutdown of Cycle 12 occurs within the burnup window the Cycle 13 peaking factors will not be significantly affected and will not exceed their limiting values.
i O
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--= - ..-~.-_ - _ .. - .---- .--.-.-.-- _ _-. -- _ _ -..,_.- - ..,- -..
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Table 2.4.1 Peaking Factor Versus Cycle 12 Shutdown Burnup I
FAH FQ Cycle 13 Limit Cycle 13 Limit 1.53 1.55 2.11 2.23 EOC 12 - 500 MWDA4TU 1.51 1.55 2.13 2.23 EOC 12 Nominal 1.50 1.55 2.15 2.23 EOC 12 + 500 MWDA4TU 0
I O
i O > o acc oew' eva'ua'1o s Table 3.0.1 presents the latest safety analyses performed fcr 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 13 accident evaluations.
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'( 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) 7 Suberitical Condition Uncontrolled RCCA Withdrawal at Power 2/78 (Cycle 4-RSE) 7 Control Rod Drop 1/27/71 (AM7-FSAR) 6 RCC Assembly Misalignment 1/27/71 (AM7-FSAR) 6 CVCS Malfunction 1/27/71 (AM7-FSAR) 6 Startup of an Inactive RC Loop 1/27/71 (AM7-FSAR) 6 Excessive Heat Removal Due to FW 1/27/71 (AM7-FSAR) 6 System Malfunctions Excessive Load Increase Incident 1/27/71 (AM7-FSAR) 6 1
Loss of Reacter Coolant Flow 3/73 (WCAP-8903) 8 Locked Rotor Accident 2/78 (Cycle 4-RSE) 7 Loss of External Electrical Load 1/27/71 (AM7-FSA9) 6 f Loss of Normal Feedwater 8/31/73 (AM33-FSAR) 6 Fuel Handling Accidents 1/27/71 (AM7-FSAR) 6 Rupture of a Steam Pipe 4/13/73 (AM28-FSAR) 6 Rupture of CR Drive Mechanism Housing 2/78 (Cycle 4-RSE) 7 RC System Pipe Rupture (LOCA) 12/10/76 (AM40-FSAR) 6 Westinghouse Zirc - Water Addendum 12/14/79 10 Clad Hoop Stress Addendum 1/8/80 11 Exxon 10/01/84 (XN-NF-84-31, Rev.1) 12 l
l
_r _ .._ _ _ ___ _ _ , _._ _ _ ,.~.__ _ _. . . - . . . _ . _ _ _ _ _ _
'l I l
l l Table 3.0.2
%f-t'y !bulyses Bounding Values Lower Upper Parameter' Bound Bound Units Moderator Temp. Coefficient -40.0 0.0 pcm/*F Doppler Coefficient -2.32 -1.0 pcm/'F Differential Baron Worth -11.2 -7.7 pcm/ ppm Delayed Neutron Fraction .00485 .0071 Prompt Neutron Lifetime 15 N/A y sec Shutdown Margin 1.0 (BOC) N/A Ep 2.0 (EOC) N/A Differential Rod Worth of 2 Banks Moving N/A 82 pcm/sec Ejected Rod Cases HFP, BOL Seff .0055 N/A N/A .30 Ep Rod Worth FQ N/A 5.03 HFP, EOL Seff .0050 N/A N/A .42 Ep Rod Worth FQ N/A 5.1 HZP, BOL Seff .0055 N/A N/A .92 Ep Rod Worth FQ N/A 13.0 HZP,EOL Seff .0050 N/A N/A .92 Ep Rod Warth FQ N/A 13.0 0
i l
l l
3.1 Evaluation of Uncontrolled Rod Withdrawal from Subcritical 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 maxirrun reactivity insertion rate produces a more severe transient while a minimum (absolute value) doppler coefficient maxi-mizes the nuclear power peak. Of lesser concern are the moderator coef-ficient and delayed neutron fraction which are chosen to maximize the peak heat flux.
Table 3.1.1 presents a comparison of Cycle 13 physics parameters to the current safety analysis values for the Uncontrolled Rod Withdrawal from a Subcritical Condition.
Since the pertinent parameters from the proposed Cycle 13 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 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.
'J
i
/~h Table 3.1.1
-Q Uncontrolled Rod Withdrawal From Subcritical Reload Safety Current Evaluation Values Safety Analysis Units Parameter A) Moderator Temp. pcm/*Fm 0.52 1 10.0 Coefficient B) Doppler Temp. pcm/*Ff Coefficient -1.6 1 -1.0
. C) Differential Worth .116 $/sec of Two Moving Banks .052 5 D) Scram Worth vs.
Time See Section 2.3 E) Delayed Neutron
.00546 > .00485 Fraction _
f O
-' 3.2 Evaluation of Uncontrolled Rod Withdrawal at Power An uncontrolled control rod bank withdrawal at power results in a gradual The increase in core power followed by an increase in core heat flux.
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 cho-sen to maximize peak heat flux.
Table 3.2.1 presents a comparison of the Cycle 13 physics parameters to the current safety analysis values for the Uncontrolled Rod Withdrawal at Power Accident.
G The application of the reliability factor to the moderator coefficient calculated at HZP, BOC, no xenon core conditions results in a slightly positive value. It is anticipated that BOC Startup Physics Test measure-ments will demonstrate that the moderator coefficient will be negative at operating conditions.
l Since the pertinent parameters from the proposed Cycle 13 reload core are i
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 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.
O
(
Table 3.2.1
. Uncontrolled Rod Withdrawal at Power Reload Safety Current Parameter Evaluation Values Safety Analysis Units A) Moderator Temp.
Coefficient 0.52* 1 0.0 pcm/*Fm ,
B) Doppler Temp.
Coefficient -1.1 1 -1.0 pcm/*Ff C) Differential Rod Worth Of Two Moving Banks .052 1 .116 4/sec D) FNN 1.54 1 1.55 E) Scram Worth vs. See Section 2.3 Time 4
F) Delayed Neutron Fraction 0.00546 > 0.00485
- Moderator Temperature Coefficient will be verified negative at Startup Testing.
J e
?
I O
3.3 Evaluation of Centrol Rod Misalignment 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 show that core DNOR limits are not exceeded.
The limiting core parameter is the peak FM in the worst case misalignment l
of Bank D fully inserted with one of its RCCAs fully withdrawn at full power.
Table 3.3.1 presents a comparison of the Cycle 13 F&N versus the current safety analysis FM limit for the Misaligned Rod Accident.
Since the pertinent parameter from the proposed Cycle 13 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 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.
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Table 3.3.1 4_
Control Rod Misalignment Accident 4
Reload Safety Current Evaluation Value Safety Analysis Parameter t
'1.84 < 1.92 -
A) F M4 1
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() 3.4 Evaluation of Dropped Rod The release of a full length control rod, or control rod bank by the gripper coils while the reactor is at power, causes the reactor.to become
-6 subcritical and produces a 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 thermal margins. Dropping of a single RCCA may or may not result in a negative rate trip, and therefore the radial power distribution must be considered.
A comparison of the Cycle 13 FAHN to the current safety analysis F6HN limit for the Dropped Rod Accident is presented in Table 3.4.1.
Since the pertinent parameter from the proposed Cycle 13 reload core is
-() conservatively 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 13 reload core design, there-fore, will not adversely affect the safe operation of the Kewaunee Plant.
1 I
- . . .- .. .-- . . . _ - - . . . . - . - . . . . . - . - - . - . . - -. _ . . _ . - . _ . - _ . _ - =
Table 3.4.1 Dropped Rod Accident I
i Reload Safety Current Safety Analysis Parameter Evaluation Value
' A) F&N 1.66 < 1.92 f
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V,y 3.5 Evaluation of Uncontrolled Baron Dilution The malfunction of the Chemical and Volume Control System (CVCS) is assumed to deliver unborated water to the reactor coolant system.
Although the boron dilution rate and shutdown margin are the key parameters in this event, additional parameters are evaluated for the manual reactor control case. In this case core thermal limits are approached and the.
transient is terminated by a reactor trip on over-temperature AT.
Table 3.5.1 presents a comparison of Cycle 13 physics analysis results to the current safety analysis values for the Uncontrolled Baron Oilution Accident for refueling and full power core conditions.
The application of the reliability factor to the moderator coefficient calculated at HZP, BOC, no xenon core conditions results in a slightly positive value. It is anticipated that BOC Startup Physics Test measure-ments will demonstrate that the moderator coefficient will be negative at operating conditions.
Since the pertinent parameters from the proposed Cycle 13 reload core are conservatively bounded by those used in the current safety analysis, an uncontrolled boron dilution accident will be less severe than the transie in the current analysis. The implementation of the Cycle 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.
.g _ _-
^
q.
+ a
/ Y Table 3.5.1
.A_J Uncontrolled Baron Oilution Accident-6 .-
" Reload Safety Current Parameter Evaluation Values Safety Analysis Units
- 1) Refuelina Conditions A) Shutdown Margin 10.8 >,
10.0 %Ap
- 11) At-Power Conditions 0.52* < 0.0 pcm/*Fm A) Moderator Temp. _
Coefficient
-1.1 < -1.0 pcm/*Ff B) Doppler Temp.
Coefficient O C) Reactivity Insertion .0021 < .
.0023 $/sec Rate by Baron D) Shutdown Margin 2.50 >_ 1.00 %Ap 1.54 < 1.55 E) FAHN ,
F) Delayed Neutron 0.00546 > 0.00485 Fraction s
- Moderator Temperature Coefficient will be verified negative at Startup Testing.
.c 10 I
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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 injection of cold water (from the idle loop hot leg) into the core which causes a rapid reactivity insertion and subsequent core power
' increase.
The moderator temperature coefficient is chosen to maximize the reactivity effect of the cold water injection. Doppler temperature coefficient is chosen conservatively low (absolute value) to maximize the nuclear power L rise. The power distribution (FAH) is used to evaluate the core thermal limit acceptability.
Table 3.6.1 presents a comparison of the Cycle 13 physics calculation results to the current safety analysis values for the Startup of an Inactive Loop Accident.
Since the pertinent parameters from the proposed Cycle 13 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. The implementation of the Cycle 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.
4 3
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l O Table 3.6.1 l Startup of an Inactive Loop Accident Reload Safety Current Evaluation Values Safety Analysis Unit Parameter
-34.7 > -40.0 pcm/*Fm A) Moderator Temp. _,
Coefflent
-1.5 < -1.0 pcm/*Ff B) Doppler.
Coefficient 1.54 < 1.55 C) FoHN i
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3.7 Evaluation of Feedwater System Malfunction
~
The malfunction of the feedwater system such that the. feedwater tem-perature is decreased or the flow is increased causes a decrease in the RCS temperature and an attendant increase in core power level due to negative reactivity coefficients and/or control system action.
Minimum and maximum moderator coefficients are evaluated to simulate both BOC and EOC conditions. The doppler reactivity coefficient is cho-sen to maximize the nuclear power peak.
A comparison of Cycle 13 physics calculation results to the current safety analysis values for the Feedwater System Malfunction Accident is presented in Table 3.7.1. ,
Since the pertirent parameters from the proposed Cycle 13 reload core are conservatively bounded by those used in the current safety analysis, a feedwater system malfunction will be less severe than the transient in the current analysis. The implementation of the Cycle 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.
O
~ _ - _ . . _._ . ._. . _ ____ ___. . . - _ .
~
Table 3.7.1 Feedwater System Malfunction Accident Reload Safety Current Parameter Evaluation Values Safety Analysis Units
- 1) Begiming of. Cycle A) Moderator Temp.
Coefficient -5.1 1 0.0 pcm/*Fm B) Doppler Temp. pcm/*Ff Coefficient -1.42 > -2.32
- 11) End of Cycle A) Moderator Temp. -31.2 > -40.0 pcm/*Fm Coefficient
-1.1 pcm/*Fm B) Doppler Temp. 1 -1.0 Coefficient 1.54 5 1.55 ,
C) FAHN O
O' V
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.
This event results in a similar transient as that described for the feed-water system malfunction and is therefore sensitive to the same parameters.
Table 3.8.1 presents a comparison of Cycle 13 physics results to the current safety analysis values for the Excessive Load Increase Accident.
Since the pertinent parameters from the proposed Cycle 13 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 implementation of the Cycle 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.
, O P
Table 3.8.1 Excessive Load Increase Accident Reload Safety - Current Paramster Evaluation Values Safety Analysis Units
- 1) Beginning of Cycle A) Moderator Temp. -5.1 -
< 0.0 pcm/*Fm '
Coefficient B) Doppler Temp. -1.42 > -2.32 pcm/*Ff
- Coefficient 4
- 11) End of Cycle A) Moderator Temp. -31.2 >
-40.0 pcm/* Fit Coefficient B) Doppler Temp. -1.1 < -1.0 pcm/*Ff Coefficient 1.54 '< 1.55 C) F&N i
4 O
1 l
1
'3.9 Evaluation of Loss of Load 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.,
Minimumandmaximummoderatorcoefficientsareevaluatedtosimulatepth l 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 13 physics parameters to the current safety analysis values for the Loss of Load Accident is presented in Table 3.9.1.
i Since the pertinent parameters from the proposed Cycle 13 reload core are conservatively bounded by those used in the current safety analysis, a loss of load accident will be less severe than the transient in the current ana-lysis. The implementation of the Cycle 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.
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_ 34 _
l
e Table 3.9.1 Loss of Load Accident Reload Safety Current Parameter Evaluation Values Safety Analysis Units
- 1) Beginning of Cycle A) Moderator Temp. -5.1 5 0.0 pcm/*Fm Coefficient B) Doppler Temp. -1.1 1 -1.0 pcm/*Ff Coefficient
- 11) End of Cycle
-31.2 >_ -40.0 pcm/*Fm A) Moderator Temp.
Coefficient O B) Doppler Temp. -1.42 > -2.32 pcm/*Ff Coefficient 4
1.54 5 1.55 C)FAHN 2
- 0) Scram Worth
! Versus Time See Section 2.3 l
t i
O t
i
4 e i 3.10 Evaluation of Loss of Normal Feedwater p V A complete loss of normal feedwater is assumed to occur due to pump failures or valve malfunctions. An additional conservatism is applied by i
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 therefore no comparisons will be made for-the Reload Safety Evaluation.
i O
l
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()
V 3.11 Evaluation of Loss of Reactor Coolant Flow Due to Pump Trip The simultaneous loss of power to the reactor coolant pumps results in a loss of drivirh 5ead 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.
The doppler temperature coefficient is compared to the most negative value ,
The since this results in the slowest neutron power decay after trip.
moderator temperature coefficient is least negative to cause a larger power rise prior to the trip. Trip reactivity and FAH are evaluated to ensure core thermal margin.
O Table 3.11.1 presents a comparison of Cycle 13 calculated physics parame-ters to the current safety analysis values for the Loss of Reactor Coolant l
Flow Due to Pump Trip Accident.
Since the pertinent parameters from the prcposed Cycle 13 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 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.
. O
. _ _ _ _ - _ - _ . ~ - . . - - _ _ . - _ - . _ - _ - -.
.=
w.
i Table 3.11.1 Loss of Reactor Coolant Flow Due to Pump Trip 1
Reload Safety Current Evaluation Values Safety Analysis Units Parameter
' A) Moderator Temp. pcm/*Fm
-5.1 5 0.0 Coefficient B) Doppler Temp.. >_ -2.32 pcm/*Ff Coefficient -1.42 J
i 1.54 1 1.55 C) F#N
- 0) Scram Worth
' Versus Time See Section 2.3 i
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o D 1 3.12 Evaluation of Loss of Reactor Coolant Flow Due to Locked Rotor s.
This accident is an instantaneous seizure of the rotor of a single reactor The coolant pump resulting in a rapid flow reduction in the affected loop.
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 J
folloving the trip as does the large delayed neutron fraction.
Table 3.12.1 presents a comparison of Cycle 13 physics parameters to the current safety analysis values for the Locked Rotor Accident.
Since the pertinent parameters from the proposed Cycle 13 reload core are O
b 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 13 reload core design, there-fore, will not adversely affect the safe operation of the Kewaunee Plant.
O
Table 3.12.1 s/
Loss of Reactor Coolant Flow Due to Locked Rotor Reload Safety Current Evaluation Values Safety Analysis Units Parameter A) Moderator Temp. pcm/*Fm Coefficient -5.1 1 0.0 B) Doppler Temp.
-1.42 > -2.32 pcm/*Ff Coefficient t
C) Delayed Neutron Fraction 0.00546 > 0.00485 D) Percent Pins >
Limiting F6HN 24.4 1 40.0 %
(DNBR=1.3)
E) Scram Worth Versus Time See Section 2.3 2.12 1 2.23 F) FQ i
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i) v 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 coef-ficient produces a positive reactivity insertion and a potential return to i
criticality after the trip. The doppler coefficient is 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 coef-ficient.
(
Table 3.13.1 presents a comparison of Cycle 13 calculated physics parame-ters 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 curve.
Since the pertinent parameters from the proposed Cycle 13 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 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Flant.
O 4
Table 3.13.1 v .-
Main Steam Line Break Accident Reload Safety Current Parameter Evaluation Value Safety Analysis Unit A) Shutdown Margin 2.50 1 2.00 Map 5.2 1 8.8 B) FAH C) Doppler Temp. -1.1 1 -1.0 pcm/*Ff.
Coefficient D) Boron Worth -7.7 1 -7.7 pcm/ ppm Coefficient i
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u FIGURE 3.13.1
's VARIATION OF REACTIVITY, WITH CORE TEMPERATURE AT 1000 PSIA FOR THE END OF. LIFE R000E0 CORE WITH ONE ROD STUCK (ZERO POWER)
, FIGURE 3.13.1
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FSAR O WPS CYCLE 13 900.00 3'50.00 400.00 450.00 500.00 550.00 600.00 CORE AVERAGE TEMPERATURE (DEG F)
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'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 thru 3.14.4 present the comparison of Cycle 13 calculated physics parameters to the current safety analysis values for the Rod Ejection Accident at zero and full power, BOC and EOC core conditions.
The application of the reliability factor to the moderator coefficient calculated at HZP, BOC, no xenon core conditions results in a slightly positive value. It is anticipated that BOC Startup Physics Test measure-ments will demonstrate that the moderator coefficient will be negative at operating conditions.
Since the pertinent parameters from the proposed Cycle 13 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 13 reload core design, there-
, fore, will not adversely affect the safe operation of the Kewaunee Plant.
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() Table 3.14.1 Rod Ejection Accidents HFP, BOC i
i Reload Safety Current Parameter Evaluation Values Safety Analysis Units 4
A) Moderator Temp.
Coefficient -5.1 ji 0.0 pcm/*Fm B) Delayed Neutron Fraction 0.00620 2; 0.00550 C) Ejected Rod Worth 0.07 ji 0.30 %Ao D) Doppler Temp.
-1.2 < -1.0 pcm/*Ff Coefficient E) Prompt Neutron 28.9 15.0 psec Lifetime 2; 2.24 j[ 5.03 F) FQN G) Scram Worth Versus Time See Section 2.3 f
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Table 3.14.2 Rod Ejection Accidents HZP, BOC Reload Safety Current Safety Analysis Units Parameter Evaluation Values A) Moderator Temp.
Coefficient 0.52* 1 0.0 pcm/*Fm B) Delayed Neutron Fraction 0.00620 > 0.00550 C) Ejected Rod Worth 0.43 $ 0.91 2 10 D) Doppler Temp. .
-1.9 1 -1.0 pcm/*Ff Coefficient O E) Prompt Neutron Lifetime 28.9 > 15.0 usec 4.88 1 11.2 F) FQN G) Scram Worth Versus Time See Section 2.3
- Moderator Temperature Coefficient will be verified negative at Startup Testing.
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Table 3.14.3 Rod Ejection Accidents HFP, EOC I
Reload Safety Current-l Parameter Evaluation Values Safety Analysis Units A) Moderator Temp.
Coefficient -17.9 1 0.0 pcm/*Fm
- 8) Delayed Neutron Fraction 0.00546 1 0.00500 C) Ejected Rod f
- Worth 0.11 1 0.42 %Ap D) Doppler Temp.
Coefficient -1.31 1 -1.0 pcm/*Ff E) Prompt Neutron 15.0 psec l Lifetime 31.9 1
) 2.68 5.1 F) FQN 1 G) Scram Worth Versus Time See Section 2.3 a
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Table 3.14.4 Rod Ejection Accidents HZP, EOC Reload Safety Current Evaluation Values Safety Analysis Units Parameter A) Moderator Temp.
0.0 pcm/*Fm Coefficient -4.6 5 B) Delayed Neutron Fraction 0.00546 > 0.00500 C) Ejected Rod Worth 0.70 5 0.92 %Ap D) Doppler Temp. pcm/'Ff Coefficient -2.83 5 -1.0 O E) Prompt Neutron 31.9 > 15.0 psec Lifetime _
8.25 5 13.0 F) FQN E) Scram Worth Versus Time See Section 2.3 0
p)
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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 restric-tions of the power distribution control procedures.
Table 3.15.1 presents a comparison of the Cycle 13 FQN, calculated at end of Cycle 13 less 2.0 GWD/MTU, to the current safety analysis FQN limit for the Fuel Handling Accident.
Since the pertinent parameter from the proposed Cycle 13 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 ana-lysis. The implementation of the Cycle 13 reload core design, therefore, will not adversely affect the safe operation of the Kewaunee Plant.
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3 Fuel Mandling Accident l 4
5 4
.i l Reload Safety Current Parameter Evaluation Values Safety Analysis {
i-A) FQN 1.97 -
< .2.53
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() 3.16 Evaluation of Loss of Coolant Accident The Loss of Coolant Accident 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 13 physics calculation
( results to the current safety analysis values for the Loss of Coolant Accident.
Since the pertinent parameters from the proposed Cycle 13 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 13 reload core design, there-fore, will not adversely affect the safe operation of the Kewaunee Plant.
I 4
,, _ _ _ _ . ~ _ _ _ _ , _ _ _ _ _ . . - _
. .. . . . . . - - _ _ _ _ = _ . . _ . . .- .-. .- -.- -.-.. _ . . . . .- . -
Table 3.16.1 e
i Loss of Coolant Accident Reload Safety Current i Parameter Evaluation Values Safety Analysis i
- - A) Scram Worth versus Time See Section 2.3 1 B) FQ See Section 3.17 C) FAH 1.54 < 1.55 4
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() 3.17 Power. Distribution Control Verification The total peaking factor FQT relates the maximum local power density to the tu core average pcwer density. The FQT is determined by both the radial and axial power distributions. The radial power distribution is relatively
' fixed by the core loading pattern design. The axial power distribution Js controlled by the procedures (9) 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 BOC to EOC. Conservative factors which account for potential power distribu-tion variations allowed by the power distribution control procedures, manu-facturing tolerances, and measurement uncertainties are applied to the calculated FQT(Z).
Figure 3.17.1 compares the calculated FQT(Z), including uncertainty fac-tors, to the FQT(Z) limits. These results demonstrate that the power distributions expected during Cycle 13 operation will not preclude full power operation under the power distribution control specifications 1
i currently applied (4).
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FIGURE 3.17.1 O .
MRX (F0 m P REL 1 VS RXIRL CORE HEIGHT CYCLE 13 S3D 862119.1029 ;
CORE HEIGHT (INCHES 1
, s e is 2: 27 33 se is si s7 as as 7s ei er es se i0s m in7 12s #2e iss sei E
y __
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4.0 TECHNICAL' SPECIFICATIONS j 1 4 No Technical Specification changes are required as a result of this reload.
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5.0 STATISTICS UPDATE In an effort to provide continuing assurance of the model applicability, 4
Cycle 11 measurements and calculations were added to the statistics cata 1
j base prior to model applications to the Cycle 13 Reload Analysis. The i
- reliability and bias factors applicable to Cycle 13 analyses are presented l in Tables 5.0.1 and 5.0.2.
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-me.=s--sd--Nc-+>-- + -.%w- -,-e-.---e--*--6-w+esi94 w-er-pe'=*- -w--w,-+ - -yy ,eg y g-yg _ w.-.--vgyy-- =- - - -%vs--e-*w+- ,-n w-.y,wr m yg-
. - . . - . _ . . . . - . - ~ . . - .- . -. -. _ -. . . -. - -..........--
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Table 5.0.1
-(
1 Reliability Factors Parameter Reliability Factor Bias j
I FQN See Table 5.0.2 FoH 3.6% 0 4
Rod Worth 10.0% 0 Moderator
- Temperature '
Coefficient 4.68PCW'F 1.1 PC W'F e
4' Doppler
! Coefficient 10.0% 0 ,
- 0 Boron Worth 5.0%
Delayed Neutron 2
Parameters 3.0% 0 -
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? Table 5.0.2
^
])'# . FQN Reliability Factors Core Level oNode RF (5) 1 (Bottom) .116 19.94 ,
2 .048 8.46 3 .031 5.82 4 .028 5.46 5 .030 5.74-6 .025 5.09 7 .024 4.92 8 .027 5.34 9 .025 5.06 10 .028 5.45 1
.024 4.93
() 11 12 .025 5.00 13 .021 4.55 14 .022 4.60 15 .021 4.53 16 .022 4.60 17 .023 4.69 18 .023 - 4.69 19 .028 5.42 i
20 .029 5.65 21 .046 8.17 22 .044 7.81 23 .088 15.09 k) 24 (Top) .091 15.57 4
,m
()
6.0 REFERENCES
- 1. Wisconsin Public Service Corporation, Kewaunee Nuclear Power Plant, topical report entitled, " Qualification of Reactor Physics Methods for Application to Kewaunee."
- 2. Wisconsin Public Service Corporation, Kewaunee Nuclear Power Plant, topical report entitled, " Reload Safety Evaluation Methods for Application to Kewaunee."
- 3. Safety Evaluation Report by the Office of Nuclear Reactor Regulation:
" Qualifications of Reactor Physics Methods for Application to Kewaunee," October 22, 1979.
- 4. Wisconsin Public Service Corporation Technical Specifications for the Kewaunee Nuclear Power Plant.
- 5. Exxon Nuclear Company, " Generic Mechanical and Thermal Hydraulic Design for Exxon Nuclear 14 X 14 Reload Fuel Assemblies with Zircaloy Guide Tubes for Westinghouse 2-Loop Pressurized Water Reactors," November 1978.
- 6. Wisconsin Public Service Corporation, Kewaunee Nuclear Power Plant, Final Safety Analysis Report.
- 7. " Reload Safety Evaluation," for Kewaunee Nuclear Powar Plant Cycles 2, 3 and 4.
- 8. WCAP 8093, " Fuel Densification Kewaunee Nuclear Power Plant," March 1973.
- 9. R.J. Burnside and J.S. Holm, " Exxon Nuclear Power Distribution Control For Pressurized Water Reactors, Phase II" XN-NF-77-57, Exxon Nuclear Company, Inc., January 1978.
- 10. ECCS Reanalysis - ZIRCAfater Reaction Calculation. Letter from E. R.
Mathews to A. Schwencer, December 14, 1979.
- 11. Clad Swelling and Fuel Blockage Models. Letter from E. R. Mathews to D.G. Eisenhut, January 8, 1980.
- 12. "Kewaunee High Burnup Safety Analysis: Limiting Break Loca t
Radiological Consequences", XN-NF-84-31, Revision 1, Exxon Nuclear i Company, Inc., October 1, 1934.
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