ML111651086

Forwards Reactor Test Program

ML111651086
Person / Time
Site:	Kewaunee
Issue date:	05/22/1979
From:	Mathews E Wisconsin Public Service Corp
To:	Schwencer A Office of Nuclear Reactor Regulation
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ML111651087	List: ML111651086
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NUDOCS 7905300028
Download: ML111651086 (31)

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REGULATORY IJ RMATION DISTRIBUTION SYS 10 (RIDS)

ACCESSION NBR:7905300028 DOC.DATE: 79/05/22 NOTARIZED: NO FACIL:50-305 KEWAUNEE.NUCLEAR POWER PLANT, WISCONSIN PUBLIC SERVIC AUTHNAME AUTHOR AFFILIATION MATHEWSER, WISCONSIN PUBLIC SERVICE CORP, RECIPNAME RECIPIENT AFFILIATION SCHWENCER,A, OPERATING REACTORS BRANCH I DOCKET #

05000305

SUBJECT:

FORWARDS "REACTOR TEST PROGRAM,"

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• WISCONSIN PUBLIC SERVICE CORPORATION j

W P.O. Box 1200, Green Bay, Wisconsin 54305 May 22, 1979 Mr. A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors U. S.

Nuclear Regulatory Commission Washington, D. C.

20555 Gentlemen:

Kewaunee Nuclear Power Plant Docket No.

50-305 Operating License DPR-43 Letter to Mr. Schwencer from Mr. Mathews dated May 8, 1979 transmitting "Reactor Test Program, Kewaunee Nuclear Power Plant" Enclosed please find forty (40) copies of the revised Wisconsin Public Service Corporation Topical Report, "Reactor Test Program, Kewaunee Nuclear Power Plant," May, 1979.

This topical report was originally transmitted by letter dated May 8, 1979.

The revisions are based upon discussions held with the NRC staff in a meeting on May 10, 1979. The wording of these revisions was agreed upon via phone converation of May 17, 1979. Please modify the referenced topical report by replacing those revised pages.

WPSC will utilize both the rod swap technique and boron dilution methods to verify rod bank worths for the Cycle V startup tests this summer.

The verifica tion will include the four control banks for the boron dilution method and all six (6) banks for rod swap.

Based on our discussions with the NRC staff, WPSC understands that the successful completion and correlation of rod swap and dilution methods will fulfill the requirements for obtaining approval to use the rod swap technique for rod worth verification in the future.

Very truly yours, E. R.sthews, Vice President Power Supply & Engineering snf Enc.

ti,9O53OO021

9 REACTOR TEST PROGRAM KEWAUNEE NUCLEAR POWER PLANT Wisconsin Public Service Corporation Wisconsin Power & Light Company Nadison Gas & Electric Company Rev. 1 May 221 1979 7905300030

9 TABLE OF CONTENTS Introduction Low Power Tests 2.1 Rod Drop Time 2.2 Initial Criticality 2.3 Determination of Maximum Flux Level for Low Power Tests 2.4 Reactivity Computer Checkout 2.5 Isothermal.Temp. Coefficient Measurement 2.6 Zero Power Flux Distribution Measurement 2.7 Rod Bank Worths Verification Power Escalation Tests Remedial Action Appendix:

Verification of Rod Swap Methods for measuring Bank Worths

.. 4 1.0 2.0 3.0 4.0 PAGE 1

1 2

3 4

5 5

6 7A 10 12 A-I

9 LIST OF TABLES Acceptance Criteria for Reactor Tests Rod Worth Measureients, BOC IV Rod Worth Calculation Comparisons, ENC vs WPS Table 1 Table A.1 Table A.2 PAGE 16 A-4 A-5

9 LIST OF FIGURES Figure 2.1-1 Figure 2.5-1 Figure 2.6-1 Typical Strip Chart Trace for Rod Drop Test Isothermal Temperature Coefficient Determination Location and Identification Numbers of Moveable in-core fission Chambers at Kewaunee Nuclear Power Plant PAGE 13 14 15

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, 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 describes the test program as a minimum and is not in tended to detail specifications 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. One ex ception may be the zero power flux distribution measurement. The power level may be raised to a maximum of 5% of full power at the discretion of the test engineer to obtain better data.

1

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 rods, thus assuring the capability to safely shutdown 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.

The desired bank is withdrawn to the full out position.

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

This test is repeated until all rods have been tested.

Rod drop times are then determined 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 2

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 3.1O.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 boron dilution with the rods in the nearly full out position. Initial ten minute counts are taken on the source range instrumentation to establish a base 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.

3

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 makeup water added and boron concentration are maintained.

When criticality is achieved boron dilution is secured, and 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 temperature and pressure, RCC positions, boron concentration, nuclear instrumentation readings and the date and time of initial criticality are recorded.

There are no specific acceptance or review criteria for this test, as the following tests include boron concentration ac ceptance criteria.

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

The reactor coolant system is.at normal operating pressure and temperature. The reactor is critical with bank D with drawn to the near full out position. Both reactor coolant pumps are operating.

A nominal start-up rate of.25 Decades per Minute (DPM) is established by rod withdrawal, and the neutron flux level is allowed to increase until nuclear heating is observed. The 4

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.

2.4 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.

A review of the results is initiated if the agreement between the computer and actual values is not within 2% (nominally).

2.5 Isothermal Temperature Coefficient Measurement The purpose of this test is to determine the temperature coefficient of reactivity for the reactor core due to mod erator and doppler contributions.

5

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 heatup is performed first, and both a heatup and a cool down are desired.

A plot of reactivity vs Tave is maintained during the heatup and cool down. 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 3pcm/F from the predicted value.

2.6 Zero Power Flux Distribution Measurement The purpose of taking a zero-power flux map is to verify that the flux profile agrees with predictions, to assure that the core-is symmetric and that no loading errors have occurred.

The flux map is obtained via the moveable in-core instrumen tation system, which utilizes 36 locations (thimbles) throughout the core (See Figure 2.6-1).

At least 75% of the locations must be available to have a valid map. Fission chambers are used 6

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.

Addi tionally, the measured quadrant tilt is checked and reaction rate integrals are compared to predictions.

Because of the low flux levels and consequently the absence of feedback in the core, it is difficult to predict actual flux distributions at this level. Therefore, there is no acceptance criterion applicable. The review criteria for this test are:

1)

The measured normalized reaction rate difference in sym metric thimbles is less than 5%.

2)

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

3)

The calculated quadrant tilt is less than 5%.

7

2.7 Rod Ba" orth 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 prescribed zero power limits.

During this phase of the test a record is kept of rod 7A

position, boron concentration and reactivity scale on the reactivity meter. This information is then used 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 ment is 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 conficu ration is acceptable if the measured worth differs by less than 100 ppm from predicted. A review will be performed if the worth differs by more than -

50 ppm from the predicted value.

2.7.3 Rod Worth Measurement by Boron Dilution 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 8

with it (See section 2.7.2 for test description).

After the integral and differential worths are deter mined, the worths of the remaining 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. Although this is the only bank worth requiring measurement by dilution, the remaining bank worths may be verified by dilution in the event that the results of the rod swap method fail to meet the acceptance criteria.

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 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 position, 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 9

the spatial effects of bank X on the worth of the 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 measured worths less the sum of the predicted worths for all rod banks measured is +/- 10%.

ii)

The measured worth of the reference bank is 10%

of its predicted value.

iii) The inferred worth of an individual bank is -

15%

of its predicted value.

The acceptance criterion for rod 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.

3.0 Power Escalation Tests The purpose of the power escalation tests is to obtain reactor characteristics to verify physics design parameters. The tests shall include as a minimum incore flux maps at 75% and 100% full power, Nuclear instrumentation calibration, and critical boron concentration measurement at equilibrium xenon.

3.1 Power Profile Determination The power profile is determined by incore flux maps and the re sults are reviewed as described in section 2.6.

These maps verify that the flux profile is symmetric.and consistent with predictions.

10

The review criteria for the power profile test are:

i)

The measured normalized reaction rate integral difference in symmetric thimbles is less than 3%.

ii)

The standard deviation of the per cent difference of the measured 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.

3.2 Nuclear Instrumentation Calibration The nuclear instrumentation calibration is normally performed at 75% (nominal) power by performing flux maps over a range of axial offsets. The axial offsets are induced with control bank Do For each flux map and axial offset value, indicated power level from the power range instrumentation, upper and lower power range currents, and reactor output are recorded. The reactor output can be measured by secondary calorimetrics or the thermal output on the flux map summary.

A plot of incore axial offset vs. excore axial offset is gener ated from the data accumulated.

This plot should be very close to a straight line; its slope is the incore-axial offset to ex core axial-offset ratio.

This ratio is calculated for each de tector and then used to calibrate the delta-flux meters. This calibration is normalized to 100% power by secondary calori metrics.

1 0A

The thermal power output of the steam generators is obtained using a mass and energy balance from data obtained using secondary system instrumentation. Steam Generator pressure, feedwater temperature and feedwater flow data are used to determine power by the relation Power (flow rate) LB/HR X (Ho-Hi) BTU/LB 3.412 X 106 BTU/MW-HR where Ho and Hi are the outlet and inlet enthalpies of the steam and feedwater.

No acceptance or review criteria are applicable for this reactor test.

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

11

The review criterion for critical boron concentration at hot full power is that the measured worth is + 50 ppm of the predicted worth. The acceptance criterion is 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 Plant Operations Review Committee (PORC) prior to reaching 100% power. If an acceptance criteria is exceeded a review will be performed and brought before the PORC prior to exceeding 5% reactor power. Reactor power shall not exceed 5% without verification of adequate shutdown margin.

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

12

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TABLE 1 ACCEPTANCE AND REVIEW CRITERIA FOR REACTOR TESTS

-16 REACTOR TEST REVIEW CRITERIA ACCEPTANCE CRITERIA Rod Drop Time Consistency with Past Results T.S. 3.10.h.:

Rod Drop Time 1.8 seconds Initial Criticality Not Applicable Not Applicable Max Low Power Flux Not Applicable Not Applicable Reactivity Computer Checkout 2% Accuracy Not Applicable Isothermal Temperature Measured.ITC 3 PCM of predicted ITC T.S.3.1.f.: ITC is negative in operating Coefficient Determination range Flux Map at Zero Power Measured normalized reaction route integrals None in symmetric thimbles less than 5%

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

Calculated Quadrant Tilt less than 5%

Rod Bank Worth Measurements ARO CB 50 ppm of predicted value ARO CB 100 ppm of predicted value (Measured means inferred the sum of the measured worths less the sum of the predicted worths of the if rod swap method the sum of the predicted worths for measured rods less the sum of the is applied) all rod banks measured is + 10% of the measured worths is less than 10% of the predicted sum total predicted worth.

The measured worth of an individual bank is 15% of its predicted value Additionally for Rod Swap Method; The measured worth of the reference bank is +/- 10% of its predicted value Power Profile Measurement at Measured normalized reaction rate integrals T.S.3.10.b.1:

Power distribution limits high power in symmetric thimbles is less than 3%

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 CalibrAtion Not Applicable Not Applicable Equilibrium ARO C 50 ppm of, predicted value ARO C B 100 ppm of predicted value

APPENDIX A VERIFICATION OF ROD SWAP METHODS

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 independently 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 boron 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

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 methodls, this section includes a table of comparisons between WPS and Exxon predictions concerning RCC Bank worths and reactivity requirements for cycle V.

A-2

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 attri7.

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

Table A.1 Rod Worth Measurements, BOC IV WPS Predicted Worth RESULTS BOC IV Inferred Differential Worths(3)

Integral WESTINGHOUSE RESULTS BOC IV Predicted Inferred Worths 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 prqxomation 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..

WS w$11 use the integral -method when the rod swap method 'is used for Rod Bank worth.verification, A-4

,RCC BANK

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

RCC BANK D

C B

A Shutdown Total Rod Worth Total Reactivity Requirements Excess Reactivity ENC Predicted WPS Predicted Worth(1 )

Worth(1) 731 1386 1012 1684 1512 695 1301 941 1588 1480 BOC(2)

EOC(3)

ENC()

WPS(5)

ENC(4) wpS 6325 2514 1555 6005 2010 1740 6658 2795 574 6528 2533 533 All worths in PCM Calculated with no Xenon Calculated at equilibrium Xenon XN-NF-79-27 KNPP Cycle.5 Safety Analysis Report April, 1979. Exxon Nuclear Co.

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

A-5

1.

2.

3.

4..

5.

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.

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