ML20205H340

Rev 3 to Reactor Test Program,Kewaunee Nuclear Power Plant

ML20205H340
Person / Time
Site:	Kewaunee
Issue date:	01/12/1987
From:	Hintz D WISCONSIN PUBLIC SERVICE CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-87-40 PROC-870112, NUDOCS 8704010029
Download: ML20205H340 (29)

Text

. . . _ . .. .. .

.- 4 MSL1.1-4' REACTOR TEST PROGRAM KEWAUNEE NUCLEAR POWER PLANT j Wisconsin Public Service Corporation Wisconsin Power & Light Company Madison Gas & Electric Company Rev. 3 January 12, 1987 I

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. MSL1.2 r

TABLE OF CONTENTS PAGE 1.0 Introduction 1 2.0 Low Power Tests 1 2.1 Rod Drop Time 2 2.2 Initial Criticality 3 2.3 Determination of Maximum Flux Level 4 for Low Power Tests 2.4 Reactivity Computer Checkout 5 2.5 Isothermal Temp. Coefficient Measurement 5 2.6 Deleted 6 2.7 Rod Bank Worth Verification 7 1-12-87 3.0 Power Escalation Tests 10 4.0 Review and Remedial Action 13 5.0 Revisions 13 References 18 Appendix A: Verification of Rod Swap Methods A-1 l

MSL1.3 LIST OF FIGURES PAGE Figure 2.1-1. Typical Strip Chart Trace for 14 Rod Drop Test Figure 2.5-1 Isothermal Temperature Coefficient 15 Determination 1-12 Figure 2.6-1 Location and Identification Numbers of 16 Moveable In-Core Fission Chambers at Kewaunee Nuclear Power Plant

. .MSL1.4

.4.

LIST OF TABLES PAGE Table 1 Acceptance Criteria for Reactor Tests 17 Table A.1 Rod Worth Measurements, BOC IV A-4 1-12-87 Table A.2 Rod Worth Calculation Comparisons, A-5 ENC vs WPS h

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MSL1.5 1.0 Introduction This report describes the Reactor Test Program at the Kewaunee Nuclear Power Plant for the start-up of a reload core. Included are the test objectives, descriptions, and review and acceptance criteria.

The objective of the reactor test program is to verify that the reload core, and hence the reactor, is safe and can be operated in a safe manner. Furthermore, the test program verifies the reliability and accuracy of the computer codes used to analyze the reload core.

Appendix A contains the necessary information for approval of the rod swap method of measuring rod bank worths. This includes a comparison of the cycle IV results obtained independently by WPS and Westinghouse, and cycle V predictions from WPS and Exxon Nuclear Corporation.

This report offers a brief description of the Kewaunee Plant test program and is not intended to provide a detailed specification of the future test programs for use in a compliance inspection.

2.0 Low Power Tests The tests described in this section are to be performed at " low power". For the purposes of this report, low power is defined as the power range below the point of adding nuclear heat.

1-12-87

MSL1.6

~

All measurements taken during these tests and all predictions in-clude corrections for uncertainties, such as measurement and pre-diction accuracy. Extreme care is taken to maintain steady state conditions wherever practical in the tests, to assure that the parameter under surveillance can be measured as accurately as practical.

2.1 Rod Drop Time ,

The objective of the rod drop time test is to verify the mobility and minimum reaction time of the reds, thus assuring the capability to safely shut down the reactor, if necessary.

The test is performed at normal operating temperature with both reactor coolant pumps running. This test will be con-ducted prior to initial criticality.

The stationary gripper coil signal, the RPI produced rod drop signal and the 60 Hz reference time base are monitored and re-corded on a five point brush recorder for each rod drop.

l The desired bank is withdrawn to the full out position.

l Selected rods are then dropped by first removing the fuse in the moveable gripper coil, and then removing the fuse in the i

stationary gripper coil. This test is repeated until all 1

rods have been tested.

! Rod drop times are then dete'rmined'from the strip chart in-dications. For conservatism, the initiation of the event is assumed to be that point in time when the signal from the stationary gripper coil first starts to decay. The end of l

. MSL1.7 the event is chosen as the point when the rod enters the dashpot. Figure 2.1-1 shows a typical strip chart trace for this test.

The acceptance criterion for this test is Technical Specifi-cation 3.10.h. If this specification is not met, the rod shall be declared inoperable.

2.2 Initial Criticality The purpose of this test procedure is to provide a safe and controlled method of achieving initial criticality.

The initial conditions are: The reactor coolant system temperature and pressure is nominally 547F and 2235 psig.

Both Reactor coolant pumps are operating, all full length rods are inserted, and rod drop tests for all rods have been com-pleted satisfactorily. The power range trip setpoint is set at 85% of full power.

The approach to criticality will be performed by baron dilution with the rods in the nearly full out position. An acceptable base count is established on the source range instrumentation 1-12-87 for the Inverse Count Rate Ratio (ICRR). An initial boron concentration is also determined from a reactor coolant system sample.

The rods are then pulled out of the reactor in specified in-crements, until they are in the nearly full out position.

After each increment the count rate is recorded and a plot of ICRR vs Rod Position is maintained.

~ .- - _ _ . _. - _ _ _ _ , _ . . . _ . _ . _ - . _ _... ._ - - . _ - .

.~  :- MSL1.8 - -

The reactor coolant is sampled every 15 minutes to determine

.the boron concentration. The pressurizer is sampled every 30 minutes to assure homogeneous distribution of boron in the reactor coolant. Boron dilution begins after rod withdrawal

-stops. Plots of ICRR vs dilution time, gallons of reactor

-q. 1 makeup. water added and boron concentration are maintained.

When criticality is achieved boron dilution is secured, and 2 the neutron flux is stabilized about'two decades above the-

! initial critical level. The neutron flux is stabilized using.

RCC group D. With the reactor just critical, reactor coolant i i temperature and pressure, RCC positions, boron concentration, , ,

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\' nuclear instrumentation readings and the date and time of t

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-\ initial criticality are recorded.

There are no specific acceptance or review criteria for this

test, as the following tests include boron concentratica ac-ceptance criteria.

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, 2.3 Determination of the Maximum Flux Level for Low P wer Tests The purpose of this procedure is to establish an upper limit and,the operating level of the zero power neutron flux level. . .

ThdPeactorcoolantsystemisatnormaloperatingpressure and temperature. The reactor is critical with bank D with-i x -

, drawn to the near full out position. Both reactor' coolant

\

pumps are operating.

Y  !' 'A nominal start-up rate of .25 Decades per Minute (DPM) is f

established by rod withdrawal, and the neutron flux level is I

allowed to increase until nuclear heating is observed. The

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MSLI.9 reactor is then brought to a steady state critical condition just before the point of nuclear heat addition. A plot of reactivity vs. flux is obtained by alternately withdrawing and inserting bank D in small amounts. The range of this plot is two to three decades of flux, with the point of nuclear heat addition as the maximum.

The low power physics tests will be performed at flux levels below the point of nuclear heat. The maximum level will be about one decade below the first indication of reactivity feedback.

f.4 Reactivity Computer Checkout The purpose of this procedure is to prepare and check out the reactivity computer for low power physics tests.

The reactor is just critical and the 20 reactivity constants have been entered into the reactivity program. Approximately 75 pcm of rod worth is inserted into the reactor core.

The computer is then calibrated at three reactivity values, approximately 25, 50 and 75 pcm; these positive reactivity insertions are obtained by rod withdrawal and measured via doubling time.

l A review cf the results is initiated if the agreement between the computer and actual values is not within 2% (norr.inally).

l 2.5 Isothermal Temperature Coefficient Measurement ,

l The purpose of this test is to determine the temperature l

coefficient of reactivity for the reactor core due to mod-l erator and doppler contributions.

l - -- .. -. . --- . - - .- - - . . .

MSL1.10- i The initial conditions are stable plant conditions with the boron concentration of the pressurizer, reactor coolant loops and volume control tank as near to the same concentration as is practical. The reactor is just critical with bank D in the near full out position.

The reactor coolant system temperature is increased or de-creased at a rate of approximately 20F per hour by manually adjusting the steam dump. Normally the cooldown is performed 1-12-87 first, and both a heatup and a cooldown are desired.

A plot of reactivity vs Tave is maintained during the heatup and cooldown. The isothermal temperature coefficient is the slope of the trace on this plot. See Figure 2.5-1.

The acceptance criterion for this test is Technical Specifi-cation 3.1.f. A review of the analytical data is performed if the measured isothermal temperature coefficient differs by i 3pcm/F from the predicted value.

2.6 Deleted l 1-12-87 l

MSL1.11:

2.7- Rod Bank Worth Verification The purpose of this test is to determine the differential boron worth over the range of RCC bank insertion, to deter-mine the endpoint boron concentration and to infer the dif-ferential and integral worths of the RCC banks.

The initial conditions are normal operating temperature and pressure of the RCS, both reactor coolant pumps running, and the reactor is critical with the rods at the fully withdrawn position.

2.7.1 Boron Differential Worth Measurement The reactor _ coolant system is sampled at 15 minute intervals and the pressurizer is sampled at 30 minute intervals to determine the boron concentration. After dilution is initiated the RCC banks are inserted a specified number of steps as necessary to compensate for the reactivity change due to boron concentration changes, and to maintain the flux level within the l

prescribed zero power limits.

l During this phase of the test a record is kept of rod position, boron concentration and reactivity scale on the reactivity meter. This information is then used l

l

MSLI.12 with the traces on the strip chart to compute the dif-  ;

ferential boron worth over the range of RCC bank in-sertion. The dilution is terminated when the moving RCCA bank is near the full in position (i.e. within 100 pcm of the endpoint bank position).

2.7.2 Boron Endpoint Measurement After the system has stabilized, the endpoint concen-tration is determined by insertion of the RCC bank to the full in position. The incremental worth of the RCC bank is estimated by monitoring the flux and reactivity response via the reactivity computer. This last measure-mt..c n performed approximately three times, with the incremental worth taken as the average of the three measurements. The endpoint boron concentration is measured at the specified statepoint, with slight dif-ferences in system parameters accounted for.

The boron endpoint data for the all rods out configu-ration is acceptable if the measured endpoint differs by less than 100 ppm from predicted. A review will be per-formed if the endpoint differs by more than + 50 ppm from the predicted value.

2.7.3 Rod Worth Measurement by Baron Dilution i

The RCC bank predicted to have the greatest worth is measured by boron' dilution and the reactivity computer.

The procedure is identical to the differential boron worth determination, and can be performed concurrently

. _m

MSL1.13 -

with it (Sea secticn 2.7.2 for test dzscriptien).

After the integral and differential worths are deter-mined for the reference bank, the worths of the re-maining banks are inferred from the rod swap method.

Utilization of the rod swap method requires that the worth of the reference bank be measured by boron di-lution. The reference bank is defined as the bank predicted to have the highest worth. In the event that the results of the rod swap method fail to meet the acceptance criteria, all the remaining control bank worths and one of two of the shutdown bank worths will be verified by dilution.

2.7.4 Rod Worth Verification By Rod Swap Rod worth verification via rod swap techniques involves the measurement of several different statepoints of the reactor. These measurements are then compared to computer predictions of the same statepoints. Good agreement between the measured and predicted statepoint values indicates that the computer model can accurately predict parameters, such as shutdown margin and bank worths.

The remaining five bank worths are inferred in the following manner. The measured reference bank is initially in a full in, or almost full in, position with the reactor just critical. The bank to be measured (bank "X")

is then inserted to the full in pcsition, while the ref-erence bank is withdrawn to the critical position.

The worth of bank X can now be inferred from the worth of the reference bank. Corrections are made to account for

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y MSL1.14 the sp;tial effects of bank X on th2 worth of th2 ref-erence bank, and to account for the varying initial position of the reference bank.

The review criteria for rod worth verification via rod swap are:

i) The sum of the mear,ured worths less the sum of the predicted worths for all rod banks measured is 1-12-87 less than 1 10% of the total predicted worth, ii) The measured ~ worth of the reference bank is 1 10%

of its predicted value.

iii) The inferred worth of an individual bank is 1 15%

of its predicted value.

The accer:ance criterion for roo worth verification is that the sum of the predicted worths of the measured rods less the sum of the measured worths is less than 10% of the total predicted worth. l 1-12-87 3.0 Power Escalation Tests The purpose of the power escalation tests is to obtain reactor characteristics to verify flux symmetry and core power distribu-tions. The tests shall include es a minimum incore flux maps at 1-12-87 a power level below or equal to 30% and at power levels of 75%

and 100%. The tests shall also include nuclear instrumentation calibration, and critical bcron concentration measurement at equilibrium xenon.

3.1 Flux _ Symmetry Tests The flux sy.nmetry test is conducted at a power level less tMn or equal tc 30% of full power. The test is provided to assure 1-12-87 that the flux profile agrees with predictions, that the core

MSL1.15 is symmetric, and that no loading errors have occurred. The test is accomplished by obtaining a flux map via the moveable in-core instrumentation system, which utilizes 36 locations (thimbles) throughout-the core (See Figure 2.6-1). At least 75% of the locations should be available to have a valid map.

Fission chambers are used to obtain 61 data points along the axial length of each of the 36 channels. The data is then reduced through the use of the INCORE computer program.

The results of the INCORE program are then used to determine if the loading is symmetric. This is done by comparing the measured normalized reaction rate integrals in symmetric thimbles. Additionally, the measured quadrant tilt is checked 1-12-87 and reaction rate integrals are compared to predictions.

The review criteria for this test are:

1) The measured normalized reaction rate difference in sym-metric thimbles is less than 10%.

2) The standard deviation of the per cent difference in the measured to predicted reaction rate integrals is less than 5%.

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3) The calculated quadrant tilt is less than 4%.

l The acceptance criterion for these tests is Technical Specifi-cation 3.10.b.

3.2 Pcwer Distribution Tests l

l The power distribution tests are conducted at power levels near i

j 75% and 100% and e e provided to determine if the measured and f predicted core power distributions are consistent.

i . - .- ,_

MSL1.16 The power distribution is . determined by incore flux maps as described in section 3.1.

The review criteria for the power profile ~ test are:

1) The measured normalized reaction rate integral difference in symmetric thimbles is less than 6%.

ii) The standard deviation of the per cent difference of the meas i .ed to predicted reaction rate integrals is less than 5%.

iii) The calculated quadrant tilt is less than 2%.

The acceptance criterion for power profile determination is Technical Specification 3.10.b.

s 3.3 Nuclear Instrumentation Calibration Calibration of Nuclear Instrumentation is an integral part of the overall reactor test program. Calibration is normally performed at 75% (nominal) power by using data from one or more flux maps in accordance with approved calibration procedures. 1-12-87 No acceptance or review criteria are applicable for this reactor test.

3.4 Critical Boron Concentration at Equilibrium Xenon The critical boron concentration is determined at hot-full-power at equilibrium Xenon, steady-state conditions. The l concentration is determined by chemical analysis of a reactor coolant system sample.

. MSL1.17 The review criterion for critical boron concentration at hot full power is'that the measured worth is 1 50 ppm of the predicted worth. The acceptance criterion is 1 100 ppm agreement.

4.0 Review and Remedial Action Each reactor test shall be reviewed by the test engineer for results within the review and acceptance criteria specified for the test. In the event of exceeding a review criteria, the data and predictions will be reevaluated in an' effort to identify any errors in data reduction or anomalies in calculational logic.

This review will be presented to the Plant Operations Review Committee (PORC) prior to reaching 100% power. If an acceptance criterion for a low power test is exceeded, a review will be performed and brought before PORC prior to exceeding 5% reactor power. Reactor power shall not exceed 5% without verification of adequate shutdown margin. The technical specifications provide limiting conditions for normal operation and physics testing; compliance with these specifications will be maintained at all times.

! The results of all reactor physics tests are reviewed by PORC.

5.0 Revisions Under the provisions of 10CFR50.59(a)(1)(iii), the Kewaunee Plant

( is permitted to make changes in the test program which are not i described in the FSAR without prior commission appr. oval, provided that the proposed revisions do not involve a change in technical specifications or any unreviewed safety question. A record of changes made to the program along with any applicable safety evaluations shall be maintained by the Kewaunee Plant.

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MSL1.18C * . -

TABLE 1 ACCEPTANCE AND REVIED CRITERRIA FOR REACTOR TESTS REACTOR TEST REVIEW CRITERIA ' ACCEPTANCE CRITERIA Rod Droo Time Consistency with Past Results -T.S. 3.10.h.: Rod Drop Time 1.8 seconds Initial Criticality Not Appilcable Not Applicable Max Low" Power Flux Not Applicable Not Appilcable Reactivity Computer Checkout 2% Accuracy Not Appilcable Isothermal Temperature Measured ITC + 3 PCM/F' of predicted ITC T.S.3.1.f.: ITC is negative in operating Coefficient Determination range Rod Bank Worth Measurements ARO Cg 1 50 ppa of predicted value ARO C8 1 100 ppa of predicted value (Measured means inferred if The sum of the measured worths less the sue The sum of the predicted worths of the rod swap method is applied) of the predicted worths for all rod banks measured rods less the sum of the measured is < + 10% of the total predicted measured worths is less than +10% of,

~

worth, the total predicted worth.

The measured worth of an individual bank is

+ 15% of its predicted value Additionally for Rod Swap method: 1-12-87 The measured worth of the reference bank is +~

10% of its pr edicted value Measured normalized reaction rate difference TS 3.10.b.1 Power distribution limits Flux Symmetry Test in symmetric thimbles less than 10%

(less than or equal to Standard deviation of the % difference of 30% power) measured to predicted reaction rate integrals is less than 5%

Calculated quadrant tilt is less than 4%

Power Distribution Tests Measured normalized reaction rate Integrals in T.S.3.10.b.1: Power distribution limits symmetric thimbles is less than 6%

(near 75% and 100% power)

Standard deviation of the % difference of measured to predicted reaction rate integrals is less than 5%

Calculated quadrant tilt is less than 2%

Nuclear Instrumentation Not Appilcable Not Appilcable Calibration Equilibrium ARO CB + 50 ppa of predicted value ARO BC f 100 ppe of predicted value MSL1.19 REFERENCES Westinghouse Electric Corporation, " Rod Exchange Technique for Rod Worth Measurement" and " Rod Worth Verification Tests Utilizing RCC Bank Interchange", submitted on Docket 50-305 via letter from Mr.

E. W. James (WPSC) to Mr. A. Schwencer (NRC), May 12, 1978.

Westinghouse Electric Corporation, " Proprietary Version of Overhead Slides Used for Rod Exchange Techniques Presentation to NRC 9/29/78",

via letter NS-TMA-1973 from T. M. Anderson (Westinghouse) to P. S. Check (NRC), November 1, 1978.

t Exxon Nuclear Company, Inc., "Kewaunee Nuclear Plant Cycle 5 Safety Analysis Report", XN-NF-79-27, April,1979.

" Westinghouse Position Statement on Core Tilt", letter from R. S. Grimm (Westinghouse) to D. C. Hintz (WPSC), dated April 2, 1981.

x

MSL1.20 APPENDIX A VERIFICATION OF ROD SWAP METHODS i

l

MSL1.21 A.1 History Wisconsin Public Service Corporation utilized the Rod Swap Technique for measuring rod bank worths for cycle IV startup tests in May, 1978. The data reduction was done concurrently and indepen'ently d of Westinghouse Electric Corporation.

Although the WPS predictions agreed well with the measurements, and, in fact, did meet the acceptance criteria, the Westinghouse predictions were not as accurate. During the subsequent re-analysis by Westinghouse, an error was found in their work.

This eventually led to a new submittal to the NRC, via Westing-house transmittal letter NS-TMA-1973, November 1, 1978.

The Westinghouse submittal referenced above includes a description of the test methods and data reduction methodology. The Techni-cal justification for rod swap, including comparison to the baron dilution method of rod worth measurement, is included in the above referenced submittal and the submittal to the NRC en-titled " Rod Exchange Techniques for Rod Worth Measurement." This was submitted on docket 50-305 in a letter from Mr. E. W. James

, (Wisconsin Public Service Corporation) to Mr. A. Schwencer (NRC) dated May 12, 1978.

The WPS staff has recalculated all of the 1978 cycle IV rod swap data following the procedure outlined in the referenced West-inghouse submittal of November 1, 1978. The results of these calculations are included within this appendix.

A-1 l

. MSL1.22 To further demonstrate the reliability of the WPS calculational methods, section 3.0 of this appendix includes comparisons of predictions of rod worth for cycle V with the predictions of Exxon Nuclear Company. Although this comparison does not directly indicate the reliability of the WPS calculational models, the agree-ment in theory with ENC and Westinghouse, and the agreement with the measurements of Cycle IV, together demonstrate the reliability of the WPS calculational methods and models.

A.2 Cycle IV Results Due to the proprietary nature of the calculational methods, WPS references the Westinghouse submittal to the NRC via trans-mittal letter NS-TMA-1973, November 1978, for the details of the rod swap calculational methods.

Table A.1 includes the Westinghouse results and the WPS results for Kewaunee, BOC IV rod swap bank worth measurements. As can be seen by the table, the agreement between WPS and Westinghouse is very good.

A.3 Cycle V Predictions Exxon Nuclear Company, the fuel supplier for KNPP Cycle V, has performed physics calculations on the KNPP reactor core indepen-dently of WPS calculations. To demonstrate the correlation of WPS methods, this section includes a table of comparisons between WPS and Exxon predictions concerning RCC Bank worths and reactivity requirements for cycle V.

A-2

. MSLI.23 Table A.2 compares predictions of total rod worth, total reactivity requirements and excess reactivity. Also included are the in-dividual RCC bank worths determined by computer simulation of boron dilution measurements by both ENC and WPS. The Exxon values used in this table are from Kewaunee Nuclear Plant Cycle 5 Safety Analysis Report, by Exxon Nuclear Company, Inc., April, 1979 (XN-NF-79-27).

The comparisons of these predictions (as shown by table A.2) indi-cates that the WPS calculational model conservatively predicts rod worths within 5% of those predicted by Exxon.

The differences between requirements and shutdown margin at BOL is attri-buted to the fact that the minimum shutdown condition determined by WPS occurred at Hot Zero Power, with the rods at the zero power insertion limits and a negatively skewed xenon distribution. This is being compared to an Exxon full power condition with conservative require-ments applied.

The minimum shutdown margin is predicted by both models to be at an end of life, hot full power condition. The respective shutdown mar-gins are 0.574% and 0.533% reactivity, respectively; the difference amounting to only 0.041% reactivity.

A-3

,. ~MSLI.24 Table A.1 Rod Worth Measurements, B0C IV WPS RESULTS BOC IV WESTINGHOUSE RESULTS BOC IV Predicted Inferred Worths (3) Predicted Inferred Worths RCC BANK Worth Differential Integral Worth Differential Integral CA 929 972 966 ~(1)- 974 976 SA 660 720 705 712 717 SB 660 716 710 716 722 CB 796 677 694 694 699 CD 683 702 678 702 696 CC(2) 1043 1025 1025 1025 1025

{ 4771 4812 4778 4822 4834

1. Westinghouse proprietary information. Refer to submittal of November 1, 1978 Westinghouse Transmittal letter NS-TMA-1973, from T. M. Anderson to Paul S.

Check. Information referenced is on " Summary Table (Revised)". No page num-ber is given.

2. Control bank C was chosen as reference bank, therefore, its worth was measured directly by boron dilution.

! 3. The difference between the integral and differential methods is in the ap-proximation of the influence of the inserted bank on the reference bank. The integral method uses a correction factor formed by the ratio of two integrals, the differential method forms the same factor by a ratio of differential worths.

WPS will use the integral method when the rod swap method is used for Rod Bank worth verification.

A-4

, MSL1.25 TABLE A.2 Comparisons of Predictions for Cycle V (WPS vs ENC)

ENC Predicted WPS Predicted RCC BANK Worth (1) Worth (1)

D 731 695 C 1386 1301 B 1012 941 A 1684 1588 Shutdown 1512 1480 80C(2) EOC(3)

ENC (4) WPS(5) ENC (4) WPS Total Rod Worth 6325 6005 6658 6528 Total Reactivity Requirements 2514 2010 2795 2533 Excess Reactivity 1555 1740 574 533

1. All worths in PCM

2. Calculated with no Xenon

3. Calculated at equilibrium Xenon

4. XN-NF-79-27 KNPP Cycle 5 Safety Analysis Report April, 1979. Exxon Nuclear Co.

5. Calculated at Hot Zero Power, negatively skewed Xenon distribution, Rods at ZPIL.

A-5

6 NRC-87-40 WPSC [414) 433-1234 TELEX 5101012698 WPSC GR8 TELECORER [414] 433-1297 EASYUNK 62891993 WNK:ONBIN PUBLIC SGMCE CORPORATION 600 North Adams + RO. Box 19002 + Green Bay, WI 54307-9002 March 25, 1987 10 CFR 50.59 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

Docket 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant Reactor Test Program - Revision 3 For your information and consistent with past practices, enclosed is a copy of Revision 3 to the Kewaunee Nuclear Plant Reactor Test Program.

Sinc rely, D. C. Hintz Vice President - Nuclear Power MSL/jms Enc.

cc - Mr. Robert Nelson, US NRC US NRC, Region III O

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