Text

Submittal of Published Versions of Approved Topical Reports, Table of Contents - Safety Evaluation
	ML030590324
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	02/13/2003
From:	Tuckman M; Duke Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	-nr, TAC MB3222, TAC MB3223, TAC MB3343, TAC MB3344 DPC-NE-1003-A, Rev 1
	Download: ML030590324 (208)
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Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Rod Swap Methodology Report For Startup Physics Testing DPC-NE-1003-A Revision 1 SER Dated October 1, 2002 Nuclear Generation Department Nuclear Engineering

Revision History Revsion Description DPC-NE-1003, Originally submitted to the NRC for approval in October Original Issue 1986. Additional submittals were made to the NRC supplying additional data and reponses to requests for additional information.

DPC-NE-1003A, NRC approved version issued in May 1987.

Original Issue DPC-NE-1003, Submitted to the NRC for approval in August 2001.

Revision 1 This revsion updates the report for completeness to indicate the use of NRC approved methods approved subsequent to the implementation of the original issue including the use of CASMO-3/SIMLUATE-3 reactor physics methods.

This revison also reflects a refinement in the rod swap to make use of two test banks.

This revision also reflects the ability of the SIMULATE-3P computer code to iterate numerical results in the determination of the reference bank critical height.

Finally, various editorial changes are made, such as reformatting tables and adding a Table of Contents, a List of Tables, and page numbers.

Changes associated with this revision are denoted by revision bars, except for the editorial changes mentioned above.

DPC-NE-1003-A NRC approved. SER issued October 1, 2002.

Revision 1 i

Table of Contents Secti on Page

1. Introduction ........................................................ 1

2. Definitions ......................................................... 2

3. Measurement Procedure ............................................... 3

4. Calculational Procedure ............................................... 4

5. Results ............................................................. 5

6. Conclusion .......................................................... 6

7. Refeneces .............................. *............................. 20 Appendix A - DPC/NRC correspondence including DPC responses to NRC requests for additional information.

Appendix B - Original issue NRC SER Appendix C - Revision 1 NRC SER List of Tables Section Page (s)

I. Duke Predicted and Inferred Bank Worth ............................ 7-11

2. Summary of Duke Predicted and Inferred Bank Worth ................. 12

3. Duke Predicted and Measured Critical Heights ...................... 13-17

4. Summary of Duke Predicted and Measured Critical Heights ........... 18

5. a Factors ......................................................... 19 ii

1. Introduction This report describes the calculational procedure used to develop the rod swap constants and describes the measurement procedure used to determine the inferred bank worths. This paper also presents a comparison between the calculated and inferred bank worths for McGuire 1 Cycles 2, 3 and 4, and McGuire 2 Cycles 2 and 3.

In order to perform the "Control Rod Worth Measurement - Rod Swap Test Procedure, (2), the following information must be provided to the station.

This information shall include the bank worths, critical heights and a's.

The critical heights and a's are used to calculate the inferred bank worth of each control and shutdown bank, as reduced from information following the iso-reactivity interchange with the reference bank.

This report presents the calculated procedures used to derive these parameters. The calculations as performed in this procedure utilize the approved physics codes and methodologies described in References 1 and 3.

The rod swap procedure is one of the methods available for determining total rod worth and individual bank worths during zero power physics testing.

1

2. Definitions The following is a list of the constants needed by the plant, to perform the rod swap procedure. These include:

"*Wex - Predicted reactivity worth of each control and shutdown bank, when inserted individually into an otherwise unrodded core.

" hp, - Predicted critical position of the reference bank after interchange with bank x, starting with the reference bank at 0 steps and bank x fully withdrawn.

"* ax - A correction factor which accounts for the effect of bank x on the partial integral worth of the reference bank, equal to the ratio of the integral worth of the reference bank from hpx to the fully withdrawn position with and without x in the core.

In addition, included is a list of constants and their definitions as used in this report.

WIX - Measured rod bank worth of bank x from rod exchange

WRe* - Measured rod bank worth of reference bank

(P2)X - The measured integral worth of the reference bank from the measured critical position (h mx) to the fully withdrawn position.

hm. - The measured critical position of the reference bank after interchange with bank x.

(hmx)o - The initial critical position of the reference bank before exchange with bank x.

(Apl) - The measured integral worth of the reference bank from 0 steps to (h mo)o-2

3. Measurement Procedure With an initial configuration of all rods out, hot zero power, the integral worth of the reference bank is measured using the standard boration/dilution technique. The reference bank is the bank that is predicted to have the highest integral worth. All other banks are then exchanged with the reference bank or other test banks at constant boron conditions until the measured bank is fully inserted.

The worth of each bank is then the amount of reactivity change caused by the withdrawal of the reference bank to its new critical height.

The rod bank worth is inferred from the measured reference bank worth and the measured reference bank height using the following equation:

Wx = Wre.f - ax (AP 2 )x - (AP,}

where the above terms are defined in Section 2.0 of this report.

3

4. Calculational Procedure This calculation is performed using EPRI-NODE-P or SIMULATE-3P to model core conditions during the rod swap procedure. The following procedure describes the method of data generation:

1. Calculate the integral bank worth at HZP, ARO critical boron. Insert one bank at a time with no overlap and calculate the bank worth as the difference between ARO and the bank fully inserted condition.

(The calculated highest worth bank will be considered the reference bank.)

2. With the reference bank fully inserted, calculate the critical boron concentration. (The reference bank in boron concentration is used in predicting the predicted rod worth - WY).

3. Using the above calculated critical boron concentration for the reference bank, the new integral bank worths at HZP are determined.

These values correspond to the predicted worth for each bank (Wpx).

The reference bank should be inserted in approximately six (6) step increments such that a plot of the integral worth of the reference bank can be obtained. (As should be noted, the Keff with the reference bank inserted, is referred to as the base Kff).

4. In order to calculate the critical height, the core is modeled with the measured bank fully inserted. The critical height (hp.) of the reference bank is then determined by adjusting the reference bank position until the Keff matches the base Keff.

5. In order to calculate a for each bank position, the following expression is used:

Integral Worth of the reference bank from h? to the fully withdrawn position with bank x inserted in the core Cc=

Integral worth of the reference bank from hP to the fully withdrawn position without bank x inserted in the core 4

5. Results Tables 1 and 2 present a comparison between Duke's predicted and inferred bank worths. A review of the available data from McGuire I Cycles 2, 3, and 4, and McGuire 2 Cycles 2 and 3, identifies a mean difference of 5.27 pcm or 0.66% between Duke's predicted and inferred bank worths.

Tables 3 and 4 identify a comparison between measured and predicted total critical heights. The standard deviation of the differences between the measured critical heights and Duke's calculated critical heights is 12.63.

Table 5 presents some typical a values as calculated for McGuire 1, Cycle 3.

Additional benchmarking of predicted and measured rod worth data using SIMULATE-3P can be found in Section 3.2 of Reference 3.

5

6. Conclusion Reference to the Rod Swap Test Procedure (2) identifies the specific acceptance criteria. In order to satisfy this procedure the following conditions must be met:

(a) The absolute value of the percent difference between the measured and predicted integral worth for the reference bank is < 15%.

(b) The sum of the measured/inferred worth of all the rods must be > 90%

of the predicted rod worth.

(c) For all RCC banks other than the reference bank, either:

(i) the percent difference between the inferred and predicted worth for each individual bank is < 30%

or (ii) IW'X-W',J < 200 pcm for each bank, whichever is greater.

These criteria were found acceptable using Duke's predicted values.

Based on the predicted and measured data presented in this report the rod swap method described has been verified to be accurate for use in startup physics testing.

6

Table 1 Duke Predicted and Inferred Bank Worth Duke Predicted Duke Inferred Difference Difference Unit/Cycle Bank Worth (PCM) Worth (PCM) (PCM) (%)

1/2 CA 289 301 -12 -4.0 CB 557 606 -49 -8.1 CC 786 788 -2 0.3 CD 616 566 50 8.8 SA 473 546 -73 -13.4 SB 443 479 -36 -7.5 SC 370 354 16 4.5 SD 362 374 -12 -3.2 SE 223 237 -14 -5.9 Total 4119 4251 -132 -3.1 Mean -14.67 -3.17 Standard Deviation 35.94 6.80 Difference (PCM) = Predicted - Inferred Difference (%) = x 100 WI 7

Table 1 (Cont.)

Duke Predicted and Inferred Bank Worth Duke Predicted Duke Inferred Difference Difference Unit/Cycle Bank Worth (PCM) Worth (PCM) (PCM) (%)

1/3 CA 311 305 6 2.0 CB 657 609 48 7.9 CC 789 745 44 5.9 CD 488 466 22 4.7 SA 269 303 -34 -11.2 SB 856 779 77 9.9 SC 394 373 21 5.6 SD 395 383 12 3.1 SE 429 392 37 9.4 Total 4588 4355 233 5.4 Mean 25.89 4.14 Standard Deviation 31.16 6.34 Difference (PCM) = Predicted - Inferred Difference (%) = x 100 WI 8

Table 1 (Cont.)

Duke Predicted and Inferred Bank Worth Duke Predicted Duke Inferred Difference Difference Unit/Cycle Bank Worth (PCM) Worth (PCM) (PCM) (%)

1/4 CA 301 313 -12 -3.8 CB 656 677 -21 -3.1 CC 775 778 -3 -0.4 CD 581 556 25 4.5 SA 293 307 -14 -4.6 SB 746 750 -4 -0.5 SC 381 377 4 1.1 SD 382 314 68 21.7 SE 473 471 2 0.4 Total 4588 4543 45 1.0 Mean 5 1.7 Standard Deviation 27 .04 8.0 Difference (PCM) = Predicted - Inferred Difference M%) = x 100 WI 9

Table 1 (Cont.)

Duke Predicted and Inferred Bank Worth Duke Predicted Duke Inferred Difference Difference Unit/Cycle Bank Worth (PCM) Worth (PCM) (PCM) (%)

2/2 CA 437 459 -22 -4.8 CB 413 452 -39 -8.6 CC 858 871 -13

-1.5 CD 654 664 -10

-1.5 SA 327 430 -103 -24.0 SB 425 480 -55 -11.5 SC 354 375 -21 -5.6 SD 355 374 -19 -5.1 SE 270 292 -22 -7.5 Total 4093 4397 -304 -6.9 Mean -33.78 -7.79 Standard Deviation 29.42 6.87 Difference (PCM) = Predicted - Inferred Difference (%) = x 100 Wi 10

Table 1 (Cont.)

Duke Predicted and Inferred Bank Worth Duke Predicted Duke Inferred Difference Difference Unit/Cycle Bank Worth (PCM) Worth (PCM) (PCM) (%)

2/3 CA 344 314 30 9.6 CB 698 668 30 4.5 cc 869 787 82 10.4 CD 591 530 61 11.5 SA 381 404 -23 -5.7 SB 906 842 64 7.6 SC 438 378 60 15.9 SD 440 406 34 8.4 SE 481 424 57 13.4 Total 5148 4753 395 8.3 Mean 43.89 8.40 Standard Deviation 30.70 6.23

- - This was the reference bank used because vendor supplied data was used for the official rod swap calculation.

Difference (PCM) = Predicted - Inferred Difference (%) = x 100 WI 11

Table 2 Summary of Duke Predicted and Inferred Bank Worth Duke Calculated Difference Difference (PCM) ( % )

Mean 5.27 .66 Standard Deviation 40.72 8.69 Difference (PCM) = Predicted - Inferred Difference (%) = x 100 Wi 12

Table 3 Duke Predicted and Measured Critical Heights Critical Height (Steps) Difference Unit/Cycle Bank Measured Predicted (Steps) 1/2 CA 83 88 -5 CB 197 195 2 CD 183 196 -13 SA 191 187 4 SB 156 157 -1 SC 144 158 -14 SD 147 156 -9 SE 86 92 -6

-42 I of Absolute Value 54 Standard Deviation 6.63 Difference (Steps) = Measured - Predicted 13

Table 3 (Cont.)

Duke Predicted and Measured Critical Heights Critical Height (Steps) Difference Unit/Cycle Bank Measured Predicted (Steps) 1/3 CA 127 117 10 CB 180 172 8 CC 224 201 23 CD 163 156 7 SA 127 i1 16 SC 139 133 6 SD 141 133 8 SE 132 126 6 84 I of Absolute Value 84 Standard Deviation 6.00 Difference (Steps) = Measured - Predicted

_J 14

Table 3 (Cont.)

I Duke Predicted and Measured Critical Heights Critical Height (Steps) Difference Unit/Cycle Bank Measured Predicted (Steps) 1/4 CA 108 121 -13 CB 201 203 -2 CD 179 191 -12 SA 136 149 -13 SB 218 216 2 SC 147 161 -14 SD 136 161 -25 SE 151 163 -12

-89 I of Absolute Value 93 Standard Deviation 8.15 Difference (Steps) = Measured - Predicted 15

Table 3 (Cont.)

Duke Predicted and Measured Critical Heights Critical Height (Steps) Difference Unit/Cycle Bank Measured Predicted (Steps) 2/2 CA 153 146 7 CB 190 191 -i CD 202 205 -3 SA 198 186 12 SB 194 183 11 SC 185 182 3 SD 184 182 2 SE 149 141 8 z 39 Sof Absolute Value 47 Standard Deviation 5.49 Difference (Steps) = Measured - Predicted 16

Table 3 (Cont.)

I Duke Predicted and Measured Critical Heights Critical Height (Steps) Difference Unit/Cycle Bank Measured Predicted (Steps) 2/3 CA 99 112 -13 CB 173 191 -18 CD 158 179 -21 SA 123 145 -22 SB 228 228 0 SC 130 159 -29 SD 131 159 -28 SE 131 147 -16

-147 1 of Absolute Value 147 Standard Deviation 9.24 Difference (Steps) = Measured - Predicted 17

Table 4 Summary of Duke Predicted and Measured Critical Heights Duke Calculated S(Differences) -155 I (Absolute Value of Differences) 425 Standard Deviation (of the Differences) 12.63 Difference (Steps) = Measured - Predicted 18

Table 5 a Factors Unit/Cycle Bank Calculated 1/3 CA 1.042 CB 0.877 CC 0.870 CD 1.161 SA 1.060 SC 1.052 SD 1.050 SE 0.903 19

7. References

1. Duke Power Company, 'Nuclear Physics Methodology for Reload Design", DPC NF-2010A, June 1985.

2. Duke Power Company, McGuire Nuclear Station, 'Control Rod Worth Measurement: Rod Swap Test Procedure", PT/O/A/4150/lIA, April 1984.

3. Duke Power Company, 'Nuclear Design Methodology Using CASMO-3/SIMULATE-3PI, DPC-NE-1004A, Revision 1, SER Dated April 26, 1997.

20

APPENDIX A NRC/DPC Correspondence Including DPC Responses to NRC Requests for Additional Information

D-UHE POWER GOMP.,,LNY P.O. Box 33189 cHAjRLOTTE. N.C. 28242 HAL B TUGKER TPLEPHO%-I.

%t4 r P., rog~r" (70.4) 37,,.-4.n31 "NtCLIEAA PtUO C"IOv February 11, 1987 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

McGuire Nuclear Station Docket Nos. 50-369/370 Catawba Nuclear Station Docket Nos. 50-413/414 Determination of Rod Worth Using Rod Swap Methodology Gentlemen:

By letter dated December 4, 1986, Duke submitted for information to NRC a descrip tion of the method by which bank worths are determined in startup physics testing.

By letter of January 12, 1987, the Staff responded to the submittal with a request for additional information. Attached are the responses to the Staff's questions.

It is intended that the methodology described in the December 4, 1986 submittal will be used for the next reloads of Duke's Westinghouse plants; the first of which is scheduled for May 1, 1987.

Very truly yours, Hal B. Tucker SAG/54/jgm Attachment

Document Control Desk February 11, 1987 Page 2 xc: Mr. Darl Hood, Project Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. J. Nelson Grace, Regional Administrator U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street NW - Suite 2900 Atlanta, GA 30323 Mr. W.T. Orders NRC Resident Inspector McGuire Nuclear Station p

-J

Document Control Desk February 11, 1987 Page 3 bxc: w/o attachment R.H. Clark

-'-1 M.S. Kitlan E.O. McCraw R. Van Namen N.A. Rutherford R.L. Gill MC-801.02 (7)

Page 1 ATTACHMENT QUESTION 1: Are all the rod worth calculations done with the EPRI-NODE-P Code, including both rod swap and rod worth for shutdown margin?

RESPONSE: Shutdown Margin calculations are performed according to the methodology approved in DPC-NF-2010A. Rod worths for both the shutdown margin calculation and the rod swap calculations are done using EPRI-NODE-P.

NOTE: See Section 5.4 of DPC-NF-2010A for the procedure for shutdown margin calculations.

QUESTION 2: Section 3, "Measurement Procedure": submit detailed procedures for the measurements. Include the actual boron dilution rate and the flux level for each of the tests included in the report.

RESPONSE: The most current procedures used in the rod swap measurements p are enclosed as Attachments 1, 2, and 3.

A summary of the reactivity insertion rates and flux levels for each of the tests in the reference is presented below.

Flux are values as measured on the reactivity computer picoammeter.levels REACTIVITY INSERTION TEST RANGE POINT OF ADDING UNIT/CYCLE RATE (PCM/HR) (AMPS) NUCLEAR HEAT (AMPS)

MIC2 450 I E-8 TO i E-7 1.4 E-6 MIC3 460 1 E-8 TO 1 E-7 4.25 E-7 MIC4 420 1 E-8 TO I E-7 5.1 E-7 M2C2 480 I E-7 TO 1 E-6 1.6 E-6 M2C3 720 1 E-7 TO 1 E-6 1.65 E-6 QUESTION 3: Section 4, "Calculational Procedure" - under 5: How many calculations are performed for each bank and at what positions.

RESPONSE: One . is calculated for each bank (except for the reference bank) at the predicted critical height. These calculations use the results of cases performed for Sections 4.3 and 4.4 of the reference. Cases are done with the reference bank being inserted in approximately 6-step increments both by. itself and in the presence of the bank being predicted.

Page 2 ATTACHMENT QUESTION 4: Table 3, "of.'s": Are the values given at the predicted heights?

RESPONSE: Alpha (0c) is the ratio of the reference bank worth from the predicted critical height to out of the core with and without bank X in the core. Values for are given at the predicted critical heights. However, the ratio of the reference bank worth with and without bank X in the core is insensitive to variations in the predicted critical heights and wil1ý have no significant impact on the inferred worth.

QUESTION 5: Submit a copy of Reference 2.

RESPONSE: Reference 2: Duke Power Company McGuire Nuclear Station, "Control Rod Worth Measurement: Rod Swap Test Procedure," PT/O/A/4150/11A, April, 1984 test procedure is enclosed as Attachment 4.

QUESTION 6: Provide data for at least 2 sets of side-by-side comparisons of boron dilution and rod swap data - predicted and measured. The data may be either for your plants or measured data from another plant and predictions by Duke.

RESPONSE: Table with requested data is provided below. All rod worths are given in units of PCM.

UNIT/ PREDICTED BOR/DILUTION I DIFF ROD SWAP Z DIFF CYCLE BANK WORTH MEAS WORTH ((P-M)/M)*100) INF WORTH ((P-I)fI)*100 MIC2 CD 616 566 8.8 586 5.1 MIC3 CD 488 483 1.0 466 4.7 MIC4 CD 581 580 0.2 556 4.5 M2C2 CD 654 665 -1.7 664 -1.5 M2C3 CD 591 556 6.3 530 11.5 MEAN 2.9 4.9 STANDARD DEVIATION 4.4 4.6

Page 3 ATTACHMENT QUESTION 7: What Organization does the safety analysis for the Duke Plants?

When this is not done by Duke, what is done (e.g. tests, comparisons, etc.) to show that the startup test results adequately represent the plant features and assumptions used in the safety analyses? J RESPONSE: Cycle specific safety reviews and any safety re-analyses required for McGuire and Catawba are performed by Westinghouse, the current fuel vendor. Assuming all startup tests meet acceptance criteria, transmittal of the results to Westinghouse is formally accomplished

'by providing them a copy of the startup report prepared for the

'-I NRC. If any review or acceptance criteria are exceeded, the the action statements in the procedure are followed. Actions required usually include review of the test data and predicted values, assessment of impacts on safety analyses and technical specification limits, etc. Groups ifivolved in these reviews include the Site Reactor Group, the General Office Nuclear Design Group and, as necessary, Site Compliance, G.O. Licensing, G.O. Safety Analysis, and Westinghouse.

The main safety analysis assumption verified by the rod swap procedure is that the plant will maintain adequate sjiutdown margin per technical specifications. One of the purposes of rod swap measurements and comparisons is to verify the accuracy of the total rod worth prediction used as an input to the shutdown margin calculation. An independent Duke Power shutdown margin is evaluated for each cycle using methods approved by the NRC in DPC-NF-2010A. The N-I rod worth used in this prediction is reduced by 10% for conservatism. Acceptance criteria listed in the procedure indicate that the total inferred rod worth as measured in the rod swap testing must be within 10% of the total predicted worth. If the total measured rod worth is less than the predicted worth by more than 10%, a review of the shutdown margin is made to determine if the current rod insertion limits provide adequate shutdown margin. If the shutdown margin is adequate, then no revision of the limits is.necessary. However, if the margin is not maintained, then Duke will notify Westinghouse, revise the rod insertion limits, and submit any necessary changes in the technical specifications to the NRC.

Reference McGuire Nuclear Station, Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testing, DPC-NE-1003, Rev. 1, December 1986.

2F5Arc'X-' ,

Attachment 1

orm 34731 (R9-86) (1) ID No.PT/01/AI/..4j0,3O 1 FoqRIN TAV41T'. r -

Change(s) 31 0 10

-Incorporated PREPARATION (2) STATION McGUIRE PROCEDURE TITLE POST REFUELING CONTROLLING PROCEDURE FOR CRITICALITY, ZERO POWER PHYSICS, i*(3)

AND POWER ESCALATION TESTING (4) PREPARED BY .AA -ye DATE

' (5) REVIEWED VY DATE e/7 CrossDiscilinar ie yN/R 1-71fi, (6) TEMPORARY APPROVAL (If Necessary)

-7 (SRO) DATE By DATE (7) APPROVED BY.. DATE (8) MISCELLANEOUS Reviewed/Approved By flATi Reviewed/Approved By flATC (9) COMMENTS (For procedure reissue indicate whether additional changes, other thae" previously approved changes, are included.

Attach additional pages. if necessary.) ADDITIONAL CHANGES INCLUDED. C Yes 2.qo (10) COMPARED WITH CONTROL COPY- YT(A 4 J ,.A DATE h/'-

/ 7 COMPLETION (11) DATE(S) PERFORMED (12) PROCEDURE COMPLETION VERIFICATION Z Yes 0 N/A Check lists and/or blanks properly initialed, signed, dated or filled in N/A or N/R. as appropriate?

Z Yes 03 N/A Listed enclosures attached?

C Yes 0 N/A Data sheets attached, completed, dated and signed?

r_,Yes C NI/A Charts. graphs, etc. attached and properly dated, identified and marked?

Yes [ N/A Acceptance criteria met?

VERIFIED VERIFIED BY DATE (13) PROCEDURE COMPLETION APPFlOVED DATE (14) REMARKS (Attach additional pages. if necessary.)

PT/O/A/4150/21 2age 1 of 15 DUKE POWER COMPANY McGUIRE NUCLEAR STATION POST REFUELING CONTROLLING PROCEDURE FOR CRITICALITY, ZERO POWER PHYSICS, AND POWER ESCALATION TESTING 1.0 Purpose 1.1 To provide a sequence of tests for the orderly startup of the unit after refueling.

1.2 To perform nuclear instrumentation overlap verification.

1.3 To determine the point of nuclear heat.

1.4 To establish the neutron flux levels corresponding to the Zero Power Physics Test Band.

1.5 To perform a checkout of the reactivity computer.

2.0 References 2.1 McGuire Nuclear Station Technical Specifications 2.2 WCAP-9648; Post-Refueling Nuclear Testing Program Criticality to Full Power.

2.3 The appropriate unit and cycle Nuclear Design Report.

3.0 Time Required 5 days, 2 engineers per shift - 3 shifts 4.0 Prerequisite Tests Initial/Date 4.1 PT/O/A/4600/14B, NIS Intermediate Range Calibration Functional Test (see Step 7.4).

4.2 PT/O/A14600/14A, NIS Power Range Calibration Functional Test (see Step 7.5)

NOTE: The tests in 4.1 and 4.2 must be completed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to beginning Physics Testing. Physics testing is defined as beginning when Control Rods are being withdrawn to achieve criticality. This occurs in Step 12.9 of PT/0/A/4150/28, Criticality Following a Change in Core Nuclear Characteristics.

5.0 Test Equipment 5.1 Reactivity Computer connected to one power range detector (Enclosure 13.6) (See Step 8.2 for installation step.)

"T/O/A/4150/21 Page 2 of 15 5.2 Chart recorders to display reactivity, flux, pressurizer level, and Tavg 5.3 Stopwatch or timer 5.4 Communications between Control Room operators and testing work station.

6.0 Limits and Precautions 6.1 The startup rate is administratively limited to 0.5 DPM.

6.2 During the Zero Power Physics Tests (Steps 12.3 - 12.10.20)

Special Test Exception 3.10.3 will be invoked. The appropriate Surveillance Requirements will be monitored by Operations.

6.3 Notify Westinghouse if any incore tilts exceed 2%.

6.4 The primary indication of core power will be AT, which should be cross checked with the NIS and the Thermal Power calculation on the OAC. If the thermal power and Power Range NIS disagree by more than 2%, then adjustment is necessary per Tech Spec 3/4.3.1, Table 4.3-1, notation 2. (CP/0/A/3007/17) 6.5 If the excore power indications are conservative, use caution when increasing power to avoid the high level trip setpoints.

6.6 Observe the Fuel Maneuvering Limits as outlined in Data Book Section 1.3.

7.0 Required Unit Status Initial/Diate

/ 7.1 The unit is in Mode 3 - Hot Standby

/ 7.2 The points listed on Enclosure 13.1 are being logged on OAC Gen.

24 program once per 6 minutes printed every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

7.3 Record the unit and cycle to which this procedure is being applied, in the test log.

/ 8.0 Prerequisite System Conditions 8.1 All RCC control banks and shutdown banks are fully inserted.

8.2 Begin to install the reactivity computer per Enclosure 13.6.

The reactivity computer shall be installed before beginning Step 12.4.

PT/O/A/4150/21 Page 3 of 15

-I 8.3 An evaluation of the impact of the core alterations on the excore detector sensitivity has been made. Document the results in the test log. Attach to this procedure any correspondence from offsite personnel on this subject.

8.4 Perform Enclosure 13.10 to demonstrate adequate Shutdown Margin at the zero power insertion limits per Tech Spec 4.1.1.1.1d.

8.5 Perform Enclosure 13.9 to verify adequate Shutdown Margin during Rod Swap.

/ 8.6 Provide I&E 7300 Systems Engineer with the new cycle 100% F.P.

predicted value of Reactor Vessel Tave.

/ 8.7 I&E 7300 Systems Engineers have set AT values to conservative numbers as necessary in the protection cabinets. Record in the test log the values which have been set in the cabinets.

9.0 Test Method

'-I The reactor is brought critical with the procedure for criticality.

Then, the Intermediate Range (I/R) NIS overlap data is recorded, the point-of-nuclear-heat flux level determined, and the Zero Power Physics Test (ZPPT) band is established. Also, the reactivity computer is verified to be set up correctly by "making reactivity changes and comparing the computer response to the calculated reactor period.

Next, the ZPPT's are performed to measure the ARO boron concentration, control rod worths, moderator temperature coefficients, and the low-power core power distribution (if necessary).

Finally, power escalation is begun, with a full core flux map between 10% and 50% full power. During the escalation above 50% full power, data is taken for the Power Range NIS calibrations. At ^80% full power, the P/R NIS is calibrated, then power is increased 100% full power. At 100% full power, the core power distribution, the NIS calibration, the thermal power output program, and the reactivity anomolies are all checked. Also, the target flux difference is measured, and Reactor Coolant System Flow Test is performed.

PT/O/A/4150/2 1 Page 4 of 15 10.0 -Data Required 10.1 Nuclear instrumentation overlap will be recorded on Enclosure 13.2.

10.2 The point of nuclear heat will be recorded on Enclosure 13.3.

10.3 The reactivity computer checkout results will be recorded on Enclosure 13.4.

10.4 Output of OAC Gen. 24 program per Enclosure 13.1.

10.5 Intermediate range high level trip setpoints on Enclosure 13.7.

10.6 Verification of adequate Shutdown Margin at the zero power insertion limits on Enclosure 13.10.

10.7 Verification of Shutdown Margin during Rod Swap on Enclosure 13.9.

11.0 Acceptance Criteria 11.1 There is at least one decade overlap on the NIS between the Source and.Intermediate Ranges, and between the Intermediate and Power Ranges (NOTE: Power Ranges are calibrated to Thermal Power, Best Est. (P1385). Use P1385 for Power-Range overlap data).

11.2 The value of the reactivity measured by the reactivity computer is within .04 "(4%) or I PCM, whichever is greater, of the reactivity inferred from the reactor period, or doubling time.

c -A2AS.04 (4%) or 1 pc 11.3 All acceptance criteria in each test procedure for the tests contained in this controlling procedure have been met.

PT/O/A/4150/21 Page 5 of 15 12.0 Procedure Initial/Date 12.1 Attach as Page 2 of Enclosure 13.4 the table of "reactivity and doubling time as a function of stable reactor period at BOL, HZP conditions" for the appropriate unit and cycle. Also attach as Page 3 of Enclosure 13.4 the curve (if provided) "Reactor Period and Doubling Time as a Function of Reactivity at BOL, HZP, No Xenon" for the appropriate unit and cycle.

/ 12.2 Inform the Operations Shift Supervisor that Special Test Exception Tech Spec 3.10.3 will be entered during criticality and Zero Power Physics Testing (Steps 12.3 - 12.10). Operations shall monitor the appropriate Surveillance Requirements during these Steps.

1 12.3 Complete PT/O/A/4150/28, Criticality Following a Change in Cori Nuclear Characteristics. It is permissible to sign off this step prior to signing off Steps 12.18 and 12.19 in PT/0/A/4150/28.

NOTE: Section 7.0 of this procedure will have been completed earlier.

NOTE: See Step 4.1 and 4.2.

/ 12.4 Begin PT/0/B/4600/55, Reactivity Computer Periodic Test approximately 4-6 hours prior to Step 12.6.

.1 12.5 Record the IR high level trip setpoints on Enclosure 13.7.

-I 12.6 With a Source Range reading of -103 cps and the reactor just critical withdraw Control Bank D or add demineralized water, to establish a slow positive startup rate (<50 pcm). When the Intermediate Range indication comes on scale, halt the flux level increase, establish just critical conditions, and record data as required by Enclosure 13.2, Page 1 of 2.

/ 12.7 Continue to increase the flux level, stopping, establishing just' critical conditions, and recording data with each decade increase in the Intermediate Range until the Source Range is blocked.

CAUTION: Do not exceed 105 cps on the Source Range unless the Source Range is blocked, as a reactor trip will occur.

II PT/OIAf 4150/21 Page 6 of 15 CAUTION: I/R high level trip setpoints are on Enclosure 13.7; do not exceed these values.

12.7.1 Verify from Enclosure 13.2 Page 1 of 2 that a minimum of one full decade of overlap exists between the Source Range and Intermediate Range before the Source 5

Range reaches 10 cps.

12.8 Determine the flux level at which the point of nuclear heat occurs by the following steps.

__/_ 12.8.1 Set up 1, 2 pen strip chart recorder with Tavg and reactivity, another 2 pen strip chart recorder with pressurizer level and flux signal.

/ 12.8.2 Establish just critical conditions with reactivity computer picoameter reading of about 1 x 10"8 amps.

Adjust the scale setting on the reactivity computer picoammeter (if necessary) such that the indicator is on scale and indicating a value near the low end of

the scale. Record start values on Enclosure 13.3.

NOTE: Stop increase if nuclear heat is observed prior to reaching this level, and repeat Step 12.5.2 from 1 X 10-9 amps on the reactivity computer picoameter.

/ 12.8.3 Establish a slow positive startup rate by rod withdrawal of about 20 pcm and allow the flux level to increase until nuclear heat is observed. At this time, re-establish just critical conditions by Control Bank D adjustment. Record Nuclear Heat Data on Enclosure 13.3.

NOTE: Nuclear heat can be best observed as an increase Tavg accompanied by a change in the reactivity trace and an increase in pressurizer level.

NOTE: It is permissible to also trend pressurizer level, Intermediate Range Level, and NC Loop Highest Average Temperature on the OAC to aid in the determination of nuclear heat.

I 12.8.4 Repeat Steps 12.8.2 and 12.8.3 a second time and record all data as requested on Enclosure 13.3.

PT/O/A/4150/21 Page 7 of 15 12.8.5 Determine the Zero Power Physics Testing Range from the reactivity computer picoameter flux levels on Enclosure 13.3. Record on Enclosure 13.3.

NOTE: The range for all Zero Power Physics Testing will be defined as the next lowest whole decade such that the upper end of the decade is not within 410 of nuclear heat.

EXAMPLE: If nuclear heat is found at 5 x 10.6 amps on the picoameter then 5 x 10- 6

= 1.5 x 10-6 and 410 the range for zero power testing is 1.0 x 10"7 to 1.0 x 106 amps.

NOTE: If the signal is not clear for the decade defined, evaluate the situation and if changes are needed to be made to the testing decade, fully document in the test log the reason for the change before continuing.

12.8.6 Insert Control Bank D slightly, allow the flux to decrease until the reactivity computer picoaameter reads near the low end within the Zero Power Physics Test range determined above, and level out again.

12.9 Perform a checkout of the reactivity computer.

- I 12.9.1 Withdraw Control Bank D until a reactivity gain of approximately +25 pcm is indicated by the reactivity computer.

12.9.2 Let the flux increase to a stable period and measure the doubling time at two or three different times over the decade using a stopwatch or timer. From the doubling time, calculate the period from the following equation and record on Enclosure 13.4, page 1:

period = DT 0.693

PT/O/A/4150/2 1 Page 8 of f5

_____/ 12.9.3 Using the table on Page 2 of Enclosure 13.4, or the curve (if provided) on Page 3 of Enclosure 13.4, convert the observed period to reactivity and record on page I of Enclosure 13.4.

_____/ 12.9.4 Record all data on Enclosure 13.4.

/ 12.9.5 Repeat measurement as needed until at least three checks have been performed.

_____/ 12.9.6 Repeat Steps 12.9.1 through 12.9.4 for a reactivity addition of +50 pca.

______/ 12.9.7 Repeat measurement as needed until at least three checks have been performed.

______/ 12.9.8 Verify the Acceptance Criteria of 11.2 has been met for the positive reactivity insertions only.

/ 12.9.9 Verify a negative reactivity insertion check has been performed satisfactorily on the reactivity computer 0 per PT/O/B/4600155, Reactivity Computer Periodic Test.

/ 12.9.10 Position Control Bank D at -220 steps by boration or dilution.

/ 12.10 Zero Power Physics Testing Complete the tests listed below. Normal operating procedures shall be used to reconfigure the plant to meet any prerequisites. All tests should be performed within the test band established in Step 12.8.5, except power will be increased up to 13-4% full power for the low power flux map if it is taken.

/ 12.10.1 Perform PT/O/A/4150/10, Boron Endpoint Measurement.

/ 12.10.2 Perform PT/O/A/4150/12, Isothermal Temperature Coefficient Measurement for the ARO case.

PT/O/A/4150/21 Page 9 of 15 12.10.3 Perform ?T/0/A/4150/31, Determination of Rod Withdrawal Limits to Ensure Moderator Temperatures Within Limits of Technical Specifications. Testing may continue under Special Test Exception Tech Spec 3.10.3; however, PT/O/A/4150/31 Section 12.1 must be performed prior to the completion of data gathering for the Rod Swap test of Step 12.10.5. If the MTC calculated in Step 12.10.2 is less than 0 pcm/°F, mark this step N/A.

12.10.4 Record on Enclosure 13.8 the Reference Bank, rod banks, and sequence to be measured by rod swap.

NOTE: If the predicted worth of any bank is close to the predicted worth of the reference bank, measure this bank last.

/ 12.10.5 Perform PT/O/A/4150/11A, Control Rod Worth P Measurement - Rod Swap. This measurement is to be done for the rod banks identified on Enclosure 13.8.

/ 12.10.6 Following Rod Swap Measurements swap Control Bank D with the reference bank until Bank D is fully inserted.

/ 12.10.7 If Section 12.1 of PT/O/A/4150/31, Determination of Rod Withdrawal Limits procedure indicates no rod withdrawal limits are needed mark Step 12.10.8, 12.10.9, and 12.10.11 as N/A and continue. If the indication is that rod withdrawal limits will be needed, perform Steps 12.10.8, 12.10.9 and 12.10.11.

NOTE: It is permissible to perform Steps 12.10.8 and 12.10.9 if desired even though it might not be required. In that case, N/A Step 12.10.11.

/ 12.10.8 Place the rods close to a D-in only configuration by borating the reference bank out.

/ 12.10.9 Perform PT/0/A/4150/12 Isothermal Temperature Coefficient Measurement for the D-in case.

/ 12.10.10 Perform PT/0/A/4150/I1 Control Rod Worth Measurement.

This measurement is to be done only for Control D as it is completely withdrawn by boration.

PT/O/A/415O/2 1 Page 10 of 15 12.10.11 Perform Section 12.2 of PT/0/A/4150/31, Determintaion of Rod Withdrawal Limits to Ensure Moderator Temperature Coefficient within Limits of Technical Specifications.

12.10.12 Perform the following steps to reset bank overlap once Control Bank D is about 215 steps withdrawn.

I 12.10.12.1 Go to the Master Cycler Cabinet and reset the Bank Overlap Digital Counter to 000 by pushing the reset button.

/ 12.10.12.2 Reset the Bank Overlap Counter to 345 plus the present Control Bank D position by pushing the button to count up from 000 to the desired value (one push of the button is one digit change on the display).

NOTE: Perform Steps 12.10.13 and 12.10.14 in any order or concurrently..

/ 12.10.13 Increase reactor power by dilution or Control D withdrawal so that both approximately 3-4% full power and Control D about 215 steps withdrawn are achieved.

NOTE: Control D may be placed in a configuration for power increase if Step 12.10.17 is to be marked N/A.

/ 12.10.14 Remove reactivity computer from the Power Range NIS Channel to which it is connected and return the Channel to OPERABLE status using Enclosure 13.6.

/ 12.10.15 Verify that Thermal Power, Best Est. reasonably agrees with the indicated loop AT's. Resolve any problems.

NOTE: Thermal Power should 75% be approximately:

[(loop avg &T(OF) ( 75 )], between 0-75% full power.

/ 12.10.16 Verify all power range channels are operable.

CAUTION: Do not continue until Step 12.10.16 is completed.

/ 12.10.17 Perform PT/O/A/4150/0ZA, Core Power Distribution if any rod swap acceptance criteria were not met in "PT/O/A/4150/1IA. Mark N/A here and also Step 12.10.19 if all criteria were met.

-J PT/O/AI4150/21 Page 11 of 15 MOTE: It is permissible to perform Step 12.10.17 in any case if desired. In that case do not mark Step 12.10.19 as N/A.

-I---- 12.10.18 Record the Intermediate Range NIS overlap data at 3-4%

full power on Enclosure 13.2.

12.10.19 Perform PT/O/A/4150/23, Quarter-Core Flux Map Qualification Test.

NOTE: Testing may continue here; however, PT/O/A/4150/23, if performed now, must be complete prior to starting Step 12.11.7.

I 12.10.20 Place Control Bank D at ,160 to 180 steps withdrawn to have sufficient reactivity to put the turbine on line.

/ 12.10.21 Verify the following:

/ .12.10.21.1 Acceptance criteria for each Zero Power Physics Test performed was met or any discrepancies have been resolved.

/ 12.10.21.2 All shutdown banks completely withdrawn and within + 12 steps of group step counter demand position.

/ 12.10.21.3 Control banks above insertion limits and within + 12 steps of group step counter demand position.

I 12.10.21.4 Verify that the rod withdrawal limits are in place if they were required.

/ 12.10.21.5 Verify NC lowest operating loop Tave

>551° F.

/ 12.10.22 Inform the Operations Shift Supervisor that Special Test Exception Tech Spec 3.10.3 is being left.

Appropriate surveillance can be stopped.

Enclosure 13.1 data trending can be discontinued.

NOTE: Do not exceed 5% full power prior to completing steps 12.10.21 and 12.21.22.

/ 12.10.23 Review Data Book curves 6.1 and 6.3A and reissue these as needed to reflect actual measured data.

I

PT/0/A/4150/21 Page 12 of 15 12.11 Power Escalation Testing 12.11.1 Reset Power Range high level trip setpoints to

/ 109% F.P. This step need not be completed prior to going on-line, only before -20% F.P.

NOTE: Prior to putting the turbine on-line, verify Control D bank at ^-160 to 180 steps. This will ensure the availability of reactivity which will be needed while placing the turbine on-line. Hake sure that

_ / I Control D bank is returned to a position >200 steps before reaching 20% F.P. per Data Book Section 1.3.

12.11.2 Verify the Power Range High Level Trip Setpoints are set to 109% full power and inform the Control Room operator of that fact. This step need not be completed prior to going on-line, only before

"^20% F.P.

12.11.3 Between 10% and 50% F.P., perform PT/O/A/4150/OZA, Core Power Distribution. (It is suggested to perform this at the 30% F.P. hold. for Chemistry.)

NOTE: Equilibrium xenon is not necessary for this flux map. Boron samples may be waived also.

12.11.4 Following the flux map, perform PT/0/A/4150/23 Quarter

/ Core Flux Map Qualfication Test. This Step can be marked N/A if it was performed in Step 12.10.19.

12.11.5 Begin increasing reactor power from 3-4% to 50% full power at a rate of approximately 2.5% per hour (not to exceed 3% per hour). See Limit and Precaution 6.6.

NOTE: A sugSested sequence for power increase is to increase load at I HWe/min for 30 minutes then hold for the remainder of the hour.

12.11.5.1 As power is increased and the unit goes on-line, check all inputs to the Thermal Power Calculation by using OAC program Nuclear 28 (Thermal Power Outputs Dump).

Resolve all problems prior to the 50% full power plateau.

PT1O/A/41501 2 1 Page 13 of 15

-- __/ 12.11.5.2 Record the Intermediate Range NIS overlap data at 10%, 20% and 25% full power on Enclosure 13.2.

-I 12.11.5.2.1 Complete Enclosure 13.5.

12.11.5.2.2 Complete new Data Book Table 2.2.1 from the data on Enclosure 13.5.

/ 12.11.5.2.3 Write a procedure change to place the new Table 2.2.1 in the appropriate unit's Data Book.

/ 12.11.5.2.4 Generate a work request to have IAE recalibrate N35 and N36 and calibrate bistables NC-203 and NC-206 using IP/O/A/3206/02K and new Data Book Table 2.2.1.

NOTE: DO NOT exceed 25% Full Power until IAE has ccmpleted calibrations of Step 12.11.5.2.4.

/ 12.11.5.3 When approximately 40-50% full power, if the excore quadrant tilts exceed 1.02, and it is expected that these tilts might not clear within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of exceeding 50% RTP, perform the data taking for PT/O/A/4600/02F, Incore and NIS Interim Recalibration with a QCFM while reactor power is between power increases. If the excore quadrant til;: are less than 1.02, or expected to be less than 1.02, mark this step N/A.

/ 12.11.5.4 Record the Intermediate Range NIS overlap data at 50% full power on Enclosure 13.2.

PT/O/A/4150/21 Page 14 of 15 12.11.6 Begin increasing reactor power from 50% to approximately 80% full power at a rate of approximately 2.5% per hour (not to exceed 3% per hour). See Limit and Precaution 6.6.

12.11.7 Perform PT/0/A/4600/0ZE, Incore and NIS Recalibration:

Post Outage, between 50% and 80% full power.

NOTE: Closely check the data acquired in Step 12.11.7 which is to be used for calibration for consistency since some of the data was acquired at <75% full power.

_ / _ 12.11.8 Record the Intermediate Range NIS overlap data at 75%

full power on Enclosure 13.2.

/ 12.11.9 Remain below approximately 80% full power until the recalibration work performed in Step 12.11.7 is p completed by I&E.

_ / _ 12.11.10 While holding at below 80% power call I&E 7300 System Engineer to take data on Thot and Tcold.

12.11.11 I&E has evaluated data gathered in Step 12.11.10 to ensure operation at 100% wIll be acceptable with respect to AT. Record in the log any I&E setpoint changes in 7300.

__ / _ 12.11.12 Begin increasing reactor power from 80% to 100% full power at a rate of 2.5% per hour (not to exceed 3% per hour). See Limit and Precaution 6.6.

12.11.13 At 100% full power, perform the following tests (steps) in any order (a suggested order is listed).

12.11.13.1 Perform PT/0/A/4150/03, Thermal Power Output Calculation.

12.11.13.2 Perform PT/0/A/4150/02A, Core Power Distribution.

12.11.13.3 Perform PT/0/A/4150/08, Target Flux Difference Calculation.

12.11.13.4 Perform PT/O/A/4600/02A, Incore and NIS

/ Correlation Check.

12.11.13.5 Perform PT/0/A/4150/04, Reactivity Anomolies Calculation.

PT/O/A/4150/21 -

Page 15 of 15 12.11.13.6 Record the Intermediate Range NIS overlap data at 100% full power on Enclosure 13.2 and forward a copy of the enclosure to the appropriate I&E engineer.

12.11.13.7 Perform PT/1 or 2/A/4150/13, NC Flow Test.

NOTE: Once Step 12.11.13.6 is complete, Step 12.11.14 may be performed.

NOTE: Perform the next two steps in any order.

12.11.14 I&E has received data from the NC Flow Test and has made a final AT evaluation for the cycle at I00% F.P.

13.0 Enclosures 13.1 PAO Data 13.2 Nuclear Instrumentation Overlap Data Sheet 13.3 Nuclear Heat Determination Data Sheet 13.4 Reactivity Computer Checkout Data Sheet 13.5 Intermediate Range Channels Worksheets 13.6 Connecting the Reactivity Computer*

13.7 Intermediate Range High Level Trip Setpoints 13.8 Sequence of Control Rod Banks for Rod Swap 13.9 Verification of Shutdown Martin During Rod Swap 13.10 Shutdown Margin at Zero Power

PT/O/A/4150/21 Page 1 of 1 Enclosure 13.1 PAO Data Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing P1393 Control Bank D Position A0819 Loop A Taw$

A0825 Loop B Tavg A0831 Loop C Taw$

A0837 Loop D Tare A1058 Loop A &T A1070 Loop B &T A1082 Loop C AT A1094 Loop D AT Al106 Reference Temperature Tref P1355 Rx. Thermal Power - Best Estimate P1385 Rx. Thermal Power - Best Estimate A1081 Generator Megawatts P1447 Primary Thermal Output %

P1445 Secondary Thermal Output %

P1469 P/R Avg. Level 1 Min. Avg.

Quad. 4 (N-44)

P1467 P/R Avg. Level 1 Min. Avg.

Quad. 2 (N-42)

P1466 P/R Avg. Level 1 Min. Avg.

Quad. 1 (N-43)

P1468 P/R Avg. Level 1 Min. Avg.

Quad. 3 (N-44)

A1006 Turbine Impulse Chamber Pressure I

PT/O/A/4150/2 1 Page I of 2 Enclosure 13.2 Nuclear Instrumntation Overlap Data Sheet Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing Picoameter amps Aftes one decade increase on IR Source Ran e Intermedia R*-te

-32 N-1 N-35 N-36

.ontrol Board CPS CPS amps amps IS Cabinet CPS CPS amps amps Picoammeter amps After one decade increase on IR N-31 Source Rane N-32 Intermedia N-35 Ra-_e N-36 Control Board CPS CPS amps amps IS Cabinet CPS CPS amps amps Picoaineter amps Readings when Source Range blocked I .... Source Ran e Intermedia Rane N- 32 M-35 M-36 Control Board CPS CPS acps amps NIS Cabinet CPS CPS amps amps Picoameter amps Recorded By Date Unit Cycle Checked By Date

PT/o/A/4150/21 Page 2 of 2 Enclosure 13.2 Nuclear Instrumentation Overlap Data Sheet Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing Volts Volts Thermal Power, Recorded By Power Level N-35 N-36 Best Est. (P1385) (Date/Time) 3%

10%

20%

25%

50%

75%

100%

Unit Cycle NOTE: Data at 20 and 25% are needed to complete Enclosure 13.5. All other data are for info only.

N 0ote: ZR voLtage data is to be taken inside each IR drawer.

Al OUT A2 OUT Take readings across terminals 2-t4 TP3 and TP4 as shown on schematic. Set Fluke to DC Volts, 0 to 10 volt scale.

A.3 OUT TP3 is a grey terminal and TP4 is a black terminal.

+15 V -

PT/O/A/4150/21 Page I of 1 Enclosure 13.3 Nuclear Heat Determination Data Sheet Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing Flux Levels (amps)

Reac. Comp. Picoaa Time meter from P.R. N-35 N-36 Try 1 Start Nuclear Heat Try 2 4.

Start Nuclear Heat Avg. of 2 nuclear heat readings Zero Power Physics Testing Range amps to amps on power range NI Recorded By Date Checked By Date McGuire Unit Cycle

Page I of" Enclosure 13.4 Recorded By Date Reactivity Computer Checkout Data Sheet Checked By Post Refueling Controlling Procedure for Criticality, Date HcGuire Unit Zero Power Physics, and Power Escalation Testing Cycle Measured Calculated Initial Flux Measured Calculated Reactivity Apc Reactivity pnT /'c - ADT Level (amps) Doubling Time Period (from computer) (from pAPDT period)

Date Time Picoammeter Seconds Seconds pcm

PT/O/A/4150/21 Page 1 of 3 Enclosure 13.5 Intermediate Range Channels Worksheet Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing Step 1: From Enclosure 13.2 record below the values of Thermal Power Best Estimate which most closely correspond to 20% and 25% power levels.

Step 2: From Enclosure 13.2 record below the voltage data given for the power levels above.

Step 3: Convert amp voltage, Eout, from Step 2 to Current, Iin, by using the following equation. Record values below on table.

Iin @ x 10"4) (1L0 25 " - 1 x 10-11 U

Step 3 Step 3 Step 2 Step 2 Iin Iin Step I Eout Eout Current Current Power N35 N36 N35 N36 Level Volts Volts Amps Amps a) b)

Step 4: Complete page 2 of 3 and 3 of 3 of this enclosure by linearly extrapolating above data to 15%, 20%, 25% and 30% power as indicated, and then converting to volts as indicated.

Calculated By Date Checked By Date

PT/O/A/4150/21 Page 2 of 3 Enclosure 13.5 Intermediate Range Channels Worksheet Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing NOTE: Data is from Enclosure 13.5 page I of 3.

1. 20% power current for N36 (Rod Stop) amps

2. 25% power current for N36 (High Flux Trip) amps

3. 30% power current for N36 (T.S. Allowable Value) amps

4. N36 Hi Flux Rod Stop Reset at 15% RTP amps

5. N36 Hi Flux Trip Reset at 20% RTP amps NOTE: Convert the values found in Step 1 through 5 from amps to volts using the following equation. p (Round to 3 decimal.places.)

Eout= 8.75 + 1.25 LogloI + Iid)

(0 Iref v where Iid = 1 x 10-11 amps Iref = I x 10-4 amps

6. N36 20% power voltage (Rod Stop) (Use Step 1.0 as Iin) volts

7. N36 25% power voltage (High Flux Trip) (Use Step 2.0 as Iin) volts

8. N36 30% power voltage (T.S. Allowable Value) (Use Step 3.0 as Iin) volts

9. N36 Hi Flux Rod Stop Reset at 15% Power (Use Step 4.0 as Iin) volts

10. N36 Hi Flux Trip Reset at 20% Power (Use Step 5.0 as Iin) volts Calculated By Date Checked By Date

PT/0/A/4150/21 Page 3 of 3 Enclosure 13.5 Intermediate Range Channels Worksheet Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing NOTE: Data is from Enclosure 13.5 page I of 3.

1. 20% power current for N35 (Rod Stop) a Mps

2. 25% power current for N35 (High Flux Trip) a, 3ps a

3. 30% power current for N35 (T.S. Allowable Value) = an xps

4. N35 Hi Flux Rod Stop Reset at 15% RTP = an Ips,

5. N35 Hi Flux Trip Reset at 20% RTP = a, 3ps NOTE: Convert the values found in Step 1 through 5 from amps to volts using the following equation.

(Round to 3 decimal places.)

Eout= 8.75 + 1.25 LoglO IineIid) I volts s

where Iid = 1 x 1011 amps Iref = 1 x 10-4 amps

6. N35 20% power voltage (Rod Stop) (Use Step 1.0 as Iin) v_ lt s.

7. N35 25% power voltage (High Flux Trip) (Use Step 2.0 as lit) v<lt

8. N35 30% power voltage (T.S. Allowable Value) :s (Use Step 3.0 as Iin) vc lt

9. N35 Hi Flux Rod Stop Reset at 15% Power (Use Step 4.0 as Iin) vc,lt "s

10. N35 Hi Flux Trip Reset at'20% Power (Use Step 5.0 as Iin) v olt :s Calculated By Date Checked By Date

PT/O/A/4150/21 Page 1 of 1 Enclosure 13.6 Connecting the Reactivity Computer Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing NOTE: Any one of the four power range channels may be used. For clarity NI-43 is chosen arbitrarily.

Initial/Date

__/__Iv 13.6.1 Have IAE place Channel NI-43 in the tripped condition with

-_/-IV input plugs removed by using the "Prerequisites" and "Removing Power Channel from Service" sections of IP/O/A/3207/03K (power range cal.) in their entirety.

NOTE: This procedure does not necessarily require that the channel be placed in the tripped condition, or that the input plugs be removed. Inform the technician that these things are necessary for Performance testing.

/ 13.6.2 Verify detector A and B input plugs and high voltage

/ IV plug have been disconnected.

/ 13.6.3 Clean all three cable connectors.

/ 13.6.4 Connect the A input plug to the'A connector, the B input plug

/ IV to the B connector, and the HV plug to the KV connector on the Reactivity Computer Black Box.

13.6.5 Connect the EV cable and P cable from the reactivity computer

/ IV to the KV and Det AB Signal terminals on the Black Box.

/ 13.6.6 Secure the Black Box to a rack mount with a tie wrap.

13.6.7 To return NI-43 to service, verify the high voltage power

/ supply and picoameter at the Reactivity Computer are off.

13.6.7.1 Inform Shift Supervisor you are returning NI-43 to

-/-IV service.

/ 13.6.8 Disconnect the A and B input plugs and the HV input plug from

/ Iv the Reactivity Computer Black Box.

/ 13.6.9 Clean all three connectors.

/ 13.6.10 Have IAE return Channel NI-43 to service by performing the

/ iv "Prerequisites" and "Returning Power Range Channel to Service" sections of IP/O/A/3207/03K (power range cal.) in their entirety.

PT/O/A/4150/21 Page 1 of I Enclosure 13.7 Intermediate Range High Level Trip Setpoints Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing N-35 trip setpoint (25% full power)

- amps N-36 trip setpoint C25% full power) amps p '-'

Recorded By Date Unit Cycle

PT/O/A/4150/21 Page 1 of 1 Enclosure 13.8 Sequence of Control Rod Banks for Rod Swap Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing Reference Bank First Bank Second Bank Third Bank Fourth Bank Fifth Bank Sixth Bank Seventh Bank Eighth Bank 0

NOTE: Some of the Banks may not be measured by rod swap; mark these Banks in the sequence N/A. Indicate justification in the test log if banks will not be measured.

Recorded By Date Unit Cycle

PT/0/A/4150/21 Page 1 of 1 Enclosure 13.9 Verification of Shutdown Margin During Rod Swap Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing

1. Inserted control rod worth at BOL and at zero power insertion limits (from Enclosure 13.10, Step 2)

2. Rod swap Reference Bank worth pcm

3. Step 1. > 1.10 - Step 2. Yes No (10% conservatism on the predicted Reference Bank Worth)

Recorded By Checked By Date Unit Cycle-

1 )/A/4150/21 Page I of 1 Enclosure 13.10 Shutdown Margin at Zero Power Post'Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing introl rod position at zero power insertion limits:

CB steps withdrawn CC steps withdrawn CD steps withdrawn

.serted control rod worth at BOL and at the zero power pcm

.sertion limits (Data Book Curve 6.3A)

L, HZP, no xenon total rod worth pcm p

ata Book Table 6.3.1) ailable rod worth at BOL and at zero power insertion limits tep 3 - Step 2) pcm rth of highest worth stuck rod at BOL pcm ata Book Table 6.3.2) ailable Shutdown Margin at BOL and at zero power insertion limits Step 4 - Step 5)

0.90] pcm quired Shutdown Margin pcm ep 6 > Step 7 Yes No Recorded By Checked By Date Unit Cycle

G'L4srrcV-' Attachment 2

[FORIMATIONO Form 34731 (10-81)

(Formerly SPD-1002-1 DUKE POWER COMPANY (1) ID No: PT/I0A/41.50/11 PROCEDURE PREPARATION Change (s) 0 to PROCESS RECORD 3 Incorporateid (2) STATION: McGuire (3) PROCEDURE TITLE: Control Rod Worth Measurement (4) PREPARED BY: f'1C.~..QS DATE:

(5) REVIEWED BY A ,\ , DATE:

Cross-Disciplinary Review By: N/R:

(6) TEMPORARY APPROVAL (IF NECESSARY):

By: (SRO) Date:

By: Date:

(7) 40 APPROVED BY: Date:_ Zve (8) MISCELLANEOUS:

Reviewed/Approved By: Date:

This copy has been compared with the Control Copy and is verified correct.

Initial Date Time

PT/O/A/4150/11 Page I of 5 DUKE POWER COMPANY McGUIRE NUCLEAR STATION CONTROL ROD WORTH MEASbREMENT 1.0 Purpose 1.1 To measure the differential and integral worth of any of the Controlling Banks or Shutdown Banks.

1.2 To measure the differential boron worth over the range being tested.

2.0 References 2.1 Rod and Boron Worth Measurements During Boron Dilution, DAP/DBP-SU-7.4.

3.0 Time Required 3.1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 2 engineer for each Rod Bank measured.

4.0 Prerequisite Tests None 5.0 Test Equipment 5.1 Reactivity Computer (with flux signal from top and bottom of.one power range channel).

5.2 Two pen strip chart recorder with reactivity and T avg signals.

5.3 Two pen strip chart recorder with pressurizer water level and flux signal.

6.0 Limits and Precautions 6.1 The NC System .temperature is controlled preferably by secondary steam bypass to the condenser or by secondary steam dump to the atmosphere. Temperature control may alternatively by affected by steam generator blowdown.

6.2 Normally all reactor, coolant pumps should be operating for "maximum mixing in the NCS. If all reactor coolant pumps are not operating, the operating pumps should be those on the NCS charging loops (A&D). See Tech Spec 3.4.1.1 and 3.10.4 if all reactor coolant pumps are not operating.

PT/O/AI'4150/11 Page 2 of 5 6.3 The rod insertion limit will be violated during this test. The operators should be made aware in advance and should anticipate the associated alarms. Technical Specification 3.10.3 allows for this.

6.4 Chart speeds for rod worth measurements should be about .2 to 1 in./min. The sawtooth of the reactivity trace should be kept at about a 450 angle.

7.0 Required Unit Status "nitial/Date 7.1 The unit is just critical in the Startup Mode (Mode 2) at zero power with the flux level in the required testing range.

7.2 Record in the log the unit to which this test applies.

8.0 Prerequisite System Conditions

/ 8.1 The reactor coolant system pressure is at 2235 -50 psig.

p NOTE: Maintain NCS pressure within +/-25 psig of established pressure during the test.

8.2 The reactor coolant system temperature is 557OF +1, -5 0 F.

NOTE: Maintain NCS temperature within +/-10 F of established temperature during the test.

8.3 o 'it'

flew

/ -, ep

.- LO L-. r

t heater capah14ty.

8.4 Test equipment is set up per Section 5.0.

8.5 The unit is sufficiently stable as determined by the test

/

/ coordinator.

8.6 The indicated core reactivity is less than +/-1 pcm.

_ / _ 8.7 Record the requested data on Enclosure 13.1 for this step.

8.8 The Control Rods are positioned as specified by the Test

/ Coordinator.

8.9 Complete Enclosure 13.4 only if no overlap data is to-be taken.

Mark this step, Step 8.9.1, and Enclosure 13.4 N/A if overlap

_ / _

data is to be taken.

8.9.1 Bank selector switch is positioned in bank select to '-i the bank being measured if 8.9 is not N/A.

PT/O/A/4150/11 Page 3 of 5 8.10 Complete Enclosure 13.5 only if overlap data is to be taken.

Mark this step, Step 8.10.1 and Enclosure 13.5 N/A if this is not the case.

8.10.1 Bank select switch is in overlap (manual) unless 8.10 is N/A.

9.0 Test Method With the RCCA's positioned as requested by the Test Coordinator, the amount of demineralized water/boric acid required to compensate for the forthcoming configuration adjustment is determined. A continuous boron concentration change is initiated at a rate of approximately 500 pcm/hr. The RCCA's are moved in discrete increments to compensate for the change in boron concentration. From the data gathered, the differential and integral worth of RCCAs being measured is determined.

10.0 Data Required 10.1 Rod positions and reactivity will be recorded on Enclosure 13.1.

10.2 The following data should be recorded on the strip charts:

(attach charts to this procedure) 10.2.1. RCCA positions before and after each discrete increment.

10.2.2 Parameter scale and chart speed should be written on the chart.

10.3 Plot of integral and differential rod worth on Enclosure 13.2.

10.4 Predicted data on Enclosure 13.4.

11.0 Acceptance Criteria 11.1 The rod worth of the rod or bank being measured is within +/-10%

of the predicted rod worth as given on Enclosure 13.4.

11.2 The integral rod worth of Control Banks A, B, D, D in overlap is within +/-4% of the total measured values of Control Banks A, B, C and D individually as given on Enclosure 13.5. This only applies if overlap data is to be taken.

'tgtFgFe9RJCJF t'6Fj6er4 mtsui'-wD Aei Fxer.tc.TFPIXý LL1.JJ7 NfoTE": "i+-S ACaepT-A^JC C-Lt'n-iFOJ DO65 NOT APTLY IF T SW 136,A)& ptEAS~.C14 IS 1-$ RPEFOl&J( S IoJr-Or. 9D W

PTIO/A/4150/11 Page 4 of 5 12.0 Procedure Itial/Date NOTE: The following steps explain the general method for performing rod worth measurements for single RCCA's, Groups of RCCA's, or Banks of RCCA's during either dilution or boration.

/_

12.1 Verify that the strip chart recorders specified in Section 5.0 are operable and set up as~required.

12.2 Determine the amount of demineralized water/boric acid to I compensate for the required configuration adjustment. Sed Enclosure 13.3 for an example of how to determine this.

12.3 Record the beginning boric acid and primary water integrator __

_ /

values in the test log. If possible, reinitialize readings to 0.0.

12.4 Using the reactivity computer, measure the worth of the bank p

being tested from its current position to the fully I withdrawn/inserted position. Record the data on Enclosure 13.1.

NOTE: This is similar to a Boron Endpoint Measurement.

12.5 Using the number obtaiined in Step 12.2, initiate the required boron concentration change at a rate -hich will not cause a reactivity rate of change of >500 pcm. hr.

NOTE: This guideline corresponds to a dilution rate of

_ / _

approximately 2500 gallons per hour (40 GPM) or a boration rate __

of approximately 250 gallons per hour (4 GPM) of 4 w/o boric acid. See Enclosure 13.3 for an example of this.

12.6 Insert/withdraw RCCA's in discrete increments in order to compensate for dilution/boration. These increments should be limited such that the resultant reactivity change are within the-*

_ / _ guidelines of approximately +/-20 pcm. During these measurements,.-*

record all relevant data on the strip charts in use. See Section 10.2.

12.7 Terminate the boron concentration change such that the desired rod configuration is achieved.

NOTE: A delay of some minutes (typically 15 minutes) is unavoidable between termination of the transient and stabilization. This delay should be anticipated.

PT/O/A/4150/11 Page 5 of 5 NOTE: Normally the desired rod configuration will be either overshot or undershot. The Test Coordinator must evaluate the effects of this situation on the results of the affected test.

If effects are unacceptable, the Test Coordinator can repeat Steps 12.2 through 12.5 to correct the situation.

NOTE: If there is any overshoot, the bank selector switch may be changed to the next bank.

NOTE: For rod swap measurements, terminate the boron concentration change such that the final position of the bank is almost to the fully inserted position.

12.8 Using the reactivity computer, measure the worth of the bank being measured from its current position to the fully inserted/withdrawn position.- Record the data in the test log

/ for later entry into Enclosure 13.1. Mark this step as N/A if this data is already obtained (i.e., overshoot to next bank).

NOTE: This is Similar to a Boron Endpoint Measurement.

12.9 Record the final primary water and/or boric acid integrator

/ values in the test log.

/ 12.10 Record the "FINAL" data requested on E-nclosure 13.1.

12.11 After the test is over, record the required data on

/ Enclosure 13.1 from the strip charts.

_ / _ 12.12 Verify the acceptance criteria has been met.

12.13 Using the data on Enclosure 13.1, complete the plot(s) on

/ Enclosure 13.2.

12.14 In the log, calculate the differential boron worth over the bank being measured by divi-ding the measured rod worth by the difference in boron concentration over the rod worth measurement.

13.0 Enclosures 13.1 Rod Worth Measurement Data Sheet 13.2 Rod Worth Curves 13.3 Example of Determination of Dilution Rate 13.4 Predicted Rod Worth Data 13.5 Rod Worth Data if Worths in Overlap are to be Taken

PT/O/AI4150/ LI Page I of 2 Control Rod Worth Measurement Enclosure 13.1 Rod Worth Measurement Data Sheet Step 8.7 McGuire Unit Cycle Bank or RCCA Identification Boration E- Dilution] (Check zne Date Power (amps)

Initial Shutdown Bank Positions: A B C D E Initial Control Bank Positions: A_ B_ C_ D Overlap Measurement (yes or no)

Initial Final (Step 12 .10)

NCS Boron Concentration (ppm):

Pressurizer Boron Concentration (ppm):

NCS Temperature (T av) OF:

NCS Pressure , psig:

Step 12.9 (pcm)

So A REMARKS Recorded By Page of n~m r-

PT/0/A/4150/ili Page 2 of 2 Control Rod Worth Measuremenr Enclosure 13.1 (cotincd)Rod Step12. worth Measurement Step12.

(cotined)Data Sheet iRCC Position (Steps Withdrawn) Delta H Reactivity/ (pcm)

Timeia Final Average i (nh) , i ;, */T A l I I1

- AI,

I I I I !I I

S I I

I I II1 I

.

S

I I

' *I I' I I p *

I I Io S*

" I

I

I I

i °

*I I I i

SI

I I I I

I I

S I

S

I

I I

I I!

St 2

I

I

. A

I ULMARKS Recorded By Pase of

JV.geL IJI (Nt*I ' lud J Ra WorthIe*asureenenL EI'.c Iobure I1. 2 Hod Woitll C(urve-s Differential and InteRral RCC Bank (RCCA) Worth I d- 1300 McGulire Unit:

Ilaak: __________

12.0 1200 Date:

1I .0 1100 Test Conditions:

10.0 1000 1. RCC Bank Poaitlonu:

900 SDA 9.0 SUB 8.0 800 SDC SI)D 7.0 100 SI)E 6.it Hill m~~~~~i11111111 1 600 14:

0 CA CB 5 500 CC ________

. Ei E if l hE CD 4 400

2. Power Level:

300

3. NC Temp.:

2.0 200 Initial: _ _

1.0 100 Final: __

4. NC Press.:

.0 0 20 40 60 80 100 120 140 160 180 200 220 240 Initial: _ _

Final; ..

5. Avg. Core Bursaup:

I1CC Ilanak (ACCA) P*Lti on (Stelps WI thdrawni) I . HWII/HI'I

-.. ..... ..... 'I. l~4I (t Lt f ( ( I

- - - I I I PT/OIA/4150/11 Page I of I Control Rod Worth Measurement Enclosure 13.3 Example of Determination of Dilution Rate (Illustration Purposes Only)

It is desired to dilute Control Bank B from 223 to 0 steps at a rate not to exceed 500 pcm/hr.

1. The starting point is known: Initial Boron Concentration is 1130 ppm.

2. Go to Figure A.3 in the Core Design Report (or any other applicable document). The Integral Rod worth for Control Bank B from 223 to 0 is "909 pcm.

3. Go to Curve 6.2 in the Data Book at 1130 ppm BOL and get -10.7 pcm/ppm for the differential boron worth.

4. 990 pcm ÷ 10.7 pcm/ppm = -92.5 ppm change (dilute)

5. 1130 ppm - 92.5 ppm = 1037 ppm ending boron concentration.

6. Go to Figure 5.1 in the Data Book. To go from 1130 to 1037 ppm, add about 5656 gallons of demineralized water.

7. An alternate method is to use the Boron Predict Program on the OAC.

8. The maximum rate is 500 pcm/hr; therefore:

5656 gal. x 500 pcm x1 hour = 47.6 gpm 990 pcm hour 60 min.

9. To be conservative, go at 45 gpm.

10. Expect the time for the rod worth measurement to be 990 pcm *2 hours 500 pcm/hr

PT/O/A/4150/11 Page 1 of I Control Rod Worth Measurement Enclosure 13.4 Predicted Rod Worth Data Step 8.9 Complete one of the following two lines. Mark the other N/A.

Bank Being Measured (i.e., Control/Shutdown)

Rod Being Measured Predicted Rod Worth Value for the above condition.

pcm +/-10%

OR pcm +/- pCm p This information was transmitted to McGuire Nuclear Station by/in (listreference):

Reason for this test (refueling, etc.):

Recorded By Date McGuire Unit Cycle

PT/0/A/4150/11 Page 1 of I Control Rod Worth Measurement Enclosure 13.5 Rod Worth Data if Worths in Overlap are to be Taken Step 8.10 Individual Measured Rod Worth Values (not in overlap):

Control Bank A pcm Control Bank B pcm Control Bank C pcm Control Bank D pcm Sum of Control Bank A, B, C and D individual rod worths: pcm .+/-4%

OR pcm+/- pcm The above individual measured rod worth values were obtained from (list procedures):

which were performed on (list dates):

Recorded By Date McGuire Unit Cycle

-r1 3,4805 (7-81)

Sheet _. of _ _

Procedure No. PT-c!

C*! ,'c/ 11v Test Logger DUKE POWER COMPANY M tL.i r*c- NUCLEAR STATION TEST LOG Date Time Remarks Initial p

t I I 1 1* t t I 1* I r

Attachment 3 qC& _JA Faj5 clI lDNo PT/Q/A/-.-3)

Form 34731 (RS-851 i IT l'lR iUlMflt A 11 'I N L 0

CD'a rge a{s , 0 5ircor"- :c.

PREPARATION i2) STATIONC TrITE -r (3) PROCEDURE TITLE Control Rod Worth Measurement, Rod Swap (4) PREPARED BY DATE /'_/ZPzr_4&

(51 REVIEWED BY/

/ /

DATE he /Z Cross-Discialhnary Revi By ,N/R i6) TEMPORARY APPROVAL (if Necessary)

By (SRO) Date.

By Date 5L (7) APPROVED BY DATE (8) MISCELLANEOUS Reviewed/Approved By Date Reviewed/Approved By Ca:e

49) COMMENTS (For procedure reissue indicate whether additional changes, other -an previously approved changes, are includea.

Attach additional pages. if necessary.) ZADOITIONAL CHANGES INCLUDED.

COMPLETION

.1 -/--/1'tM/ DATE I fzR /

010) COMPARED WITH CONTROL COPY t01) DATE(S) PERFORMED (12) PROCEDURE COMPLETION VERIFICATION Z Yes Z N/A Check lists and/or blanks properly initialed. signed, dated or 'I led in N/A or N/R. as appropriate?

- Yes - N/A Listed enclosures attached?

- Yes - N/A Data sheets attach~ed. completed, dated and signed?

-Yes Z N/A Charts. graphs. etc. attached and properly dated. identified and marked?

- Yes = N/A Acceptance criteria met?

VERIFIED BY DATE (13) PROCEDURE COMPLETION APPROVED, -DATE (14) REMARKS (Attach additional pages. if necessary.)

FEB3 18987

PT/O/A/415011 lA Page 1 of 12 DUKE POWER COMPANY McGUIRE NUCLEAR STATION CONTROL ROD WORTH MEASUREMENT:

ROD SWAP 1.0 Purpose 1.1 To verify that the reactivity worth of the Reference RCC bank, as determined through reactivity computer measurement data, is consistent with design predictions.

NOTE: The reference RCC bank is the bank which has the predicted highest reactivity worth of all control and shutdown banks when inserted into an otherwise unrodded core.

1.2 To verify that the reactivity worth of each control and shutdown bank (except the reference bank), as inferred from data following iso-reactivity interchange with the reference bank, is consistent with design predictions.

2.0 References 2.1 Rod Bank Worth Measurements Utilizing Bank Exchange, WCAP-9863-A, May 1982.

2.2 Control Rod Worth Measurement, PT/0/A/4150/11 2.3 Post Refueling Controlling Procedure for Criticality, ZPPT, a nd Power Escalation Testing, PT/0/A/4150/21 2.4 Technical Specifications 3.4.1.1, 3.10.4, 3.10.3, and 3.10.2.

3.0 Time Required 3.1 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 1 engineer 4.0 Prerequisite Tests Initial/Da,te

/ 4.1 PT/O/A/4150/1O, ARO Boron Endpoiat Measurement NOTE: It is only necessary to obtain a value for ARO Boron Endpoint.

5.0 Test Equipment 5.1 Reactivity Computer (with flux signal from top and bottom of one power range channel).

PT/0/A/4150/11A Page 2 of 12 5.2 Two two-pen strip chart recorders. One chart recorder should have reactivity (on a scale of 10 pcm/inch, with 0 pcm being the center of the recorder sheet), and T avg from one loop (on a scale of 1 0F/inch for 556 to 558 0 F set up on one side of the recorder sheet). The other chart recorder should have flux (on a scale of 0 to the top end of the testing decade in amps) and pressurizer level (on a scale of 10% level/inch). Chart speeds should be I inch/min.

NOTE: The specifications in this step may be altered by the Test Coordinator as necessary to accommodate equipment limitations, as long as all four signals are recorded or trended.

6.0 Limits and Precautions 6.1 The NC system temperature is controlled preferably by steam dump to the condenser. Temperature control may alternatively be affected by steam generator blowdown.

6.2 Normally all reactor coolant pumps should be operating for maximum mixing in the NCS. If all reactor coolant pumps are not operating, the operating pumps shoulJ be those on the NCS charging loops (A and/or D). See Tech Spec 3.4.1.1 and 3.10.4 if all reactor coolant pumps are not operating.

6.3 The rod insertion limit and bank overlap sequence will be violated during this test. The operators should be made aware in advance and should anticipate the associated alarms.

Technical Specification 3.10.2 and 3.10.3 allows for this.

6.4 Maintain the flux level in the zero power test range established in Reference 2.3.

6.5 Prior to switching the rod control selector switch from one bank to another, verify both groups of the bank (if the bank has two groups) are at the same position in order to avoid group misalignment.

PT/O/A/4150/IIA Page 3 of 12 7.0 Required Unit Status Initial/Date

____I 7.1 The unit is just critical in the Startup Mode (Mode 2) at zero power with the flux level in the zero power test range established in PT/O/A/4150/21, "Post Refueling Controlling Procedure for Criticality, ZPPT, and Power Escalation Testing."

7.2 Record in the log the unit and cycle to which this test applies.

8.0 Prerequisite System Conditions NOTE: The following steps may be signed off in any order.

/ 8.1 The reactor coolant system temperature is 557*F +1, -5O*F.

NOTE: Maintain NCS temperature within +/-10 F of established temperature during the test.

/ 8.2 The difference between NC loop, pressurizer, and VCT boron concentrations is less than 20 ppm. List on Enclosure 13.3.

NOTE: Do not use the boronometer.

Boron samples are desirable but are not necessary for completion of test. Samples may be waived if reason is logged in the test log. Samples may be taken during the data taking at the test coordinator's request.

/ 8.3 Xenon worth rate is changing less than t.1 pcm/min.

/ 8.4 Test equipment is set up per Section 5.0.

/ 8.5 All available pressurizer heaters are on as needed, in order to improve mixing by maximizing the pressurizer spray.

8.6 All control and shutdown banks are fully withdrawn except Control Bank D which is at a position greater than about 215 steps withdrawn.

/ 8.7 The Rod Control Selector switch is in Bank Select Mode set on Control Bank D.

/ 8.8 Complete-Enclosure 13.1 with the predicted data. See Reference 2.3,Enclosure 13.8 for banks to be measured. See Enclosure 13.2 for an explanation of nomenclature used in this test.

NOTE: If any banks are not being measured mark the blanks on Enclosure 13.1 N/A.

PTIO/A/415O/I A Page 4 of 12 9.0 Test Method The RCC bank with the highest predicted value of reactivity worth is measured using the dilution technique per PT/0/A/4150/I1. This bank serves as a reference. The integral worth of the remaining RCC banks is implied from the difference in the critical rod position of the reference bank with and without the insertion of bank being tested.

The implied integral worths are then compared to predicted rod worths.

10.0 Data Required 10.1 The following conditions for the approximate time of criticality before each bank exchange, recorded on Enclosure 13.4:

Time Just critical height of reference bank 10.2 Nuclear design predictions on Enclosure 13.1.

10.3 Boron concentration information for the NCS and pressurizer on.

Enclosure 13.3. Boron samples are desirable but are not necessary for.completion of test. Samples may be waived if the reason is logged in the test log.

10.4 A copy of the rod positions and rod worths for the reference bank from Enclosure 13.1 of PT70/A/4150/11 when this test is complete.

10.5 The calculated, implied integral worth (W ) for each RCC bank x

except the reference bank. List data on Enclosure 13.4.

10.6 The percent difference between inferred and predicted worths for each individual RCC banks (e1 ) and for the sum of all banks (C2 )

on Enclosure 13.5.

PT/O/A/4150/1lA Page 5 of 12 11.0 Acceptance Criteria 11.1 The absolute value of the percent difference between measured and predicted integral worth for the reference bank Ls ý15%

(from Enclosure 13.5 (eI) 1 ý15%).

11.2 From Enclosure 13.5, the calculated value e2 =10%.

11.3 For all RCC banks other than the reference bank; either:

a) From Enclosure 13.5, z1 030% for each bank or x

or b) W - WP ý200 pcm for each bank, x x whichever is greater.

p

-5

PT/O/A/4150/ 1A Page 6 of 12 12.0 Procedure Initial/Date NOTE: See Enclosure 13.2 for an explanation of all nomenclature used in this test.

12.1 Measure the integral reactivity worth of the reference bank as follows:

NOTE: The reference bank is defined as that bank which is predicted to have the highest worth, of all control and shutdown banks, when inserted into an otherwise un-rodded core (see Enclosure 13.1 for the identity of this bank). In this procedure, all banks will be referred to by the bank number, except the reference bank. If the reference bank is currently positioned at less than 228 steps withdrawn (i.e., if it is Control Bank D), continue with step 12.1.5. Mark steps 12.1.1.

5 to 12.1.4 NA. If the reference bank is positioned at 228 steps withdrawn, continue on at Step 12.1.1.

12.1.1 Insert the reference bank until the indicated "reactivity is approximately -10 pcm."

/ 12.1.2 Withdraw the bank inserted below 228 until the indicated reactivity is approximately +10 pcm.

/ 12.1.3 Repeat steps f2.2.1 and 12.2.2 until the previously inserted bank is fully withdrawn.

/ 12.1.4 Adjust the position of the reference bank until the reactor is just critical. Record this position in the test log.

/ 12.1.5 Perform Control Rod Worth Measurement per PT/0/A/4150/11 on the reference Bank.

/ 12.1.6 Attach a completed copy of PT/O/A/4150/11 Enclosure 13.1 to this procedure.

/ 12.1.7 Record the total reference bank rod worth from PT/O/A/4150/11 Enclosure 13.1 on Enclosure 13.4 as shown.

/ 12.1.8 Ensure the reactor is critical at the same reference bank position as was obtained at the end of PT/0/A/4150/11.

PT/0/A/4150/1 lA Page 7 of 12 12.2 Measure the reactivity worth of the remaining control and shutdown banks, relative to the reference bank, as follows:

NOTE: The relative worth of each RCC bank is obtained from the critical position of the reference bank (initially nearly fully inserted) after full insertion of the bank being measured (initially fully withdrawn), at constant RCS boron concentration.

12.2.1 Record the initial critical bank configuration on (4) _ _ _

Enclosure 13.4 for the reference bank.

2 3 4 5 6 7 8 I 12.2.2 Insert bank 1 (identify this bank on top of Enclosure (4) __

13.5; i.e., Bank 1 is S/D E or Cont. B, etc.) until 2 3 4 5 6 the reactivity indicated by the reactivity computer tiS approximately -20 pcm.

7 8 I 12.2.3 Withdraw the reference Bank until the indicated (4) __

reactivity -isapproximately +20 pcm.

2 3 4 5 6 NOTE: Maintain the flux within the zero power test range established in Reference 2.4.

7 8 2 12.2.4 Repeat Steps 12.4.2 and 12.4.3 until bank 1 is fully (4) __

inserted. Keep the indicated reactivity within 2 3 4 5 6 +/-20 pcm.

7 8

/ 12.2.5 Adjust the position of reference bank until the (4) _reactor is just critical. Record the final critical 2 3 4 5 configuration data on Enclosure 13.4.

7 8

/ 12.2.6 Insert the reference Bank 1 until the indicated (4) -reactivity is approximately -20 pcm.

2 3 4 5 6 7 8

PT/0/A/4150/11A Page 8 of 12 12.2.7 Withdraw bank 1 until the indicated reactivity Ls

.1) approximately +20 pcm.

2 3 4 5 6 7 8

/ 12.2.8 Repeat Steps 12.4.6 and 12.4.7 until bank I is fully withdrawn.

2 3 4 5 6 7 8

/ 12.2.9 Adjust the position of reference Bank until the 4)- - -reactor is just critical. Record the critical 2 3 4 5 6 configuration data on Enclosure 13.4.

7 8

/ 12.2.10 Repeat Steps 12.4.2 through 12.4.9 for the remaining, unmeasured control and shutdown banks numbered 2 through 8 instead of bank 1. Identify the bank beside the bank number on Enclosure 13.4.

NOTE: If any banks are not being measured mark the blanks on Enclosure 13.4 and the check off blanks in stea 12.4.2 throuth 12.4.9 N/A.

12.3 Have Chemistry take a NC & pressurizer boron sample and write the results on Enclosure 13.2.

NOTE: The test may continue while waiting for the boron samples.

NOTE: This completes the data acquisition section of the test.

12.4 Compute the average of the reference bank critical-position on Enclosure 13.4.

PT/O/A/4150/11A Page 9 of 12-12.5 Compute the inferred worth for each control and shutdown bank (except the reference bank) as follows:

-I 12.5.1 Using the data from Enclosure 13.4, and the worth measurement data for the reference bank from Enclosure 13.1 of PT/0/A/4150/11, compute the value of (Aoix as described below and record on Enclosure 13.4:

=L°1(h")

o avg where:

R (h ) avg L 0 is the measured integral worth of the reference bank from 0-steps to (h0) from x oavg Enclosure 13.1.of PT/OIA/4150/11.

NOTE: Linearly interpolate if (h*), avg does not correspond to the steps on Enclosure 13.1 of PT/O/A/4150/11.

and (h0) avg is the average of the initial Xo0 and return critical positions of the reference bank before and after interchange with bank x as given on Enclosure 13.4.

-J

PT/0/A/4150/ 1A Page 10 of 12

/ 12.5.2 Using the data from Enclosure 13.4, the worth measurement data for the reference bank from Enclosure 13.1 of PT/O/A/4150/11 and the design data of Enclosure 13.1, compute the value of ax (o2 )x as described below and record on Enclosure 13.4:

228 a x(4°2) x = arx [OR]

hM x

is the measured integral worth of the where: Ri hK 228 x

reference bank from hM to the fully

withdrawn position from PT/O/A/4150/11, Enclosure 13.1. Linearly interpolate if {o/L"

h x does not correspond to the'steps on L PT/O/A/4150/11 Enclosure 13.1. *ec

-hM x is the measured critical position of the Y reference bank after interchange with bank x from Enclosure 13.4.

and is a correction factor from Enclosure 13.1 x

to account for the influence of bank x on the worth of the reference bank.

NOTL¶. j;

-F C1,..O(.

(.l2-3L'.

ePt(-G4.,L &-t ) e.0 I-~

PT/0/A/4150/ 1A Page 11 of 12 I___ 12.5.3 Compute the inferred integral worth of each bank x, WI, as described below and record on Enclosure 13.4:

Wx = WM - (AP )* (A- )

x R 1x x 2x where: WR is the measured total integral R

reference bank worth from PT/O/A/4150/1I Enclosure 13.1.

(Aoi)x is from step 12.5.1.

and a x(02)X is from step 12.5.2 p

C..I,% off S)'B. O*..1 C I

-,~ **, .3-..

~ c c*,-.,oC 'Je. G-%4;!_~s.- .

PT/O/A/415O/11A Page 12 of 12

/ 12.5.4 Compute the percent difference between inferred and predicted worths for each individual RCC bank and the sum of all banks described below.

W-I w/1 x 100, in %

2 = N i

i=1 Fill in all blanks and summarize the calculations on Enclosure 13.5.

/ 12.6 Verify all acceptance criteria have been met..

13.0 Enclosures 13.1 Nuclear Design Predictions for Rod Interchange Measurements 13.2 Nomenclature 13.3 Log of Boron Concentrations 13.4 Critical Configuration Data 13.5 Comparison of Inferred Bank Worths with Design Predictions*

13.6 Letter on Rod Swap

PT/O/A/4150/ 1A Page 1 of 1 Control Rod Worth Measurement: Rod Swap Enclosure 13.1 Nuclear Design Predictions for Rod Interchange Measurements McGuire Unit Cycle (a) Reference bank - the bank with the highest predicted integral worth.

(b) Reference bank critical position after interchange with bank x.

(c) Ratio of integral worth of the reference bank from hP to the fully withdrawn position with and without x in the core.

+ Control Bank C, Shutdown Bank E, etc.

NOTE: See Enclosure 13.2 for a complete listing of nomenclature used in this test.

Recorded By Date This data came from (list source):

PT/O/A/4150/IlA Page 1 of 1 Control Rod Worth Measurement: Rod Swap Enclosure 13.2 Nomenclature Predicted reactivity worth of each control and shutdown bank when inserted individually into an otherwise unrodded core.

X The calculated, implied rod bank worths of bank x from rod

2. WI exchange R Measured rod bank worth of reference bank

4. a A correction factor which accounts for the effect of bank x x

on the partial integral worth of the reference bank, equal tp the ratio of the integral worth of the reference bank from hx to the fully withdrawn position with and without x in the x core.

5. (Ao)X The measured integral worth of the reference bank from h. to the fully withdrawn position.

6. hP The predicted critical position of the reference bank after x

interchange with bank x starting with reference bank at 0, bank x fully withdrawn.

7. hK The.measured critical position of the reference bank after X

interchange with bank x.

M

8. [WRM] (hx~ v Is the measured integral worth of the reference bank from 0 steps to (hx)° avg

9. (hM)o avg Is the average of the initial and return critical positions x 0 avg of the reference bank before and after interchange with bank x.

-r .1228

10. Is the measured integral worth of the reference bank from hM oh X to the fully withdrawn position. x

PT/O/A/4150/1IA Page 1 of 1 Control Rod Worth Measurement: Rod Swap Enclosure 13.3 Log of Boron Concentrations NCS Boron Concentrations Time Recorded Date Sample Taken +VCT NCS Press. Comments By p

McGuire Unit Cycle NOTE: VCT sample needed only once at start of the test. hark this block as N/A after this.

"PT/O/A/4150/1 IA Page I of I Init Cycle Control Rod Worth Measurement: Rod Swap Enclosure 13.4 Critical Configuration and Worth Calculation Sheet M.JId Bank (x) Datef Time (it) (steps) j b (steps) J (AP ) Of(AP) Wx I I V. ex Di...... I pcms) (pcm) (pcot)

No. Ident. (step (step (step 12.4) (step 12.2.5) (step 12.5.1) (step 12.5.2) (Step 12.2.1) 12.2.9) 5.3) 12.

2 S....3)

I _ _ _ _ _ _ _ _ _ _ _ I_ _ _ ___ _

,p 4

5 6

7o , _

C U

-' 1 I

_fIkrc ItI&,, RE.-,- I-C^.IJALS H1.b A wuv.r" Recorded By DdLeC Checked By _D_ lite Step 12.1.7 W" -PCM 10.

5'

IT/O/A/4150/I1A

!age I of 1 Control Rod Worth Measurement: Rod Swap Enclosure 13.5 Comparison of Inferred Bank Worths With Design Predictions Unit Cycle Date Bank (x) x x No. Ident. (pcm) (pcm) (M) reference 1

2 3

4 5

6 8I 7

wx (pcm) (pcm) 2 (2)

.4

_______________________ a tep 12.2.7: Record the measured worth of the reference bank here.

rom Enclosure 13.1 from Enclosure 13.4 Recorded By __Date Checked By Date

Enclo=,tre 13.6 1 J1 A/- 1--U/ LL Letter Rod Swap Page I LANGFORD.F.L (WESt974) Posted: Wed 10 ,P"-'d5 9:10 EST s.s 4y.

,jSect Rod Swap Information for Mike Kitlin4 in, please forward this to Mike Kitlan.

s is to docuam1mt our telecon on 4/9/85 on the actions to be taken in the test bank could be worthore than the reference bank far the I Swap bank worth measureeent. When this occurs, the following fnts should be noted:

I) Do not change the reference bank designation.

2) Exchange the highest worth test bank last.

-3) With the reference bank fully out and the test bank nearly fully inserted, measure the remaining worth of the test bank by one of two methods.

a) Perform an "endpoint type" maneuver and insert the test bank from the critical position to zero steps and measure the reactivity worth using the reactivity computer.

b) If the remaining worth of the test bank is larger than approximately 50 pcm, then dilute the test bank in from the critical position to zero steps and measure the reactivity worth using the reactivity computer. This will render the measurement of the just critical position of the reference bank alohe after the. swap N/A.

4) The worth of the test bank will be equal L- the total worth of the reference bank plus the measured remat .,ng worth of the test bank minus the worth of the referenci bank from just critical to zero steps. Or in equation fc. .n:

WX = WR + WE - WRo where: WX is the worth of the test b.ank, WR is the total worth of the "eference bank, WE is the remaining worth of -he test bank with

"- the reference bank fully .trdrawn, and WRo is the worth of the refer. ce bank from the just critical position to ully inserted (Delta-Rho-I in the procem.re).

Note that Alpha-'4 times Oelta-Rho-2-x (pro.:adure notation) is not used since Delta-Rho-2-x is zero.

zpefully this meets your documentation requirements for this unique

-oblem. Also note that this was done at Zion last year without any roblems.

egards, 0 R. Grobmyer estinghouso NTD uclear Operations

Attachment 4 Form 34731 (10-81)

(Formerly SPD-1002-1)

DUKE POWER COMPANY (1) ID NOo:PT/0/A/-130/1i.A PROCEDURE PREPARATION Change(s) 0 to PROCESS RECORD 0 Incorporated (2) STATION: McGuire (3) PROCEDURE TITLE: Control Rod Worth Measurement: Rod Swap (4) PREPARED BY: I"S *- trx DATE: 444 (5) REVIEWED BY: DATE:

p N/R:_____

Cross-Disciplinary Review By:

(6) TEMPORARY APPROVAL (IF NECESSARY):

By: (SRO) Date:

By: Date:

(7) APPROVED BY: Date:

(8) MISCELLANEOUS:

Reviewed/Approved By: Date:

Form 34731 (10-81)

(Formerly SPD-1 002-1)

DUKE POWER COMPANY (1) ID No:PT/O/A/4150A/IA PROCEDURE PREPARATION Change(s)_o.to PROCESS RECORD 0 Incorporated (2) STATION: Mfcrre (3) PROCEDURE TITLE: Control Rod Worth Measurement: Rod Swap (4) PREPARED BY: M1,610A.. S. K~a. on DATE: 4/4/18+

(5) REVIEWED BY: DATE:

p Cross-Disciplinary Review By: N/R:

(6) TE-MORARY APPROVAL (IF NECESSARY):

By: (SRO) Date:

By: Date:

(7) APPROVED BY: Date:

(8) MISCELLANEOUS:

Reviewed/Approved By: Date:

I

Form 34634 (4-81) SPD-1001-2 DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECK LIST Yn C_ ou. .e UNIT: i X 2 X 3 (1) STATION:

OTHER:

4 T!= 'I-1-=

(2) CHECK LIST APPLICABLE TO: I R1 (f, 5 r-sx-(3) SAFETY EVALUATION - PART A The item to which this evaluation is applicable represents:

Yes No >* A change to the station or procedures as described in the FSAR or a test or experiment not described in the FSAR?

If the answer to the above is "Yes", attach a detailed description of the item being evaluated and an identification of the affected sect!on(s) of the FSAR.

(4) SAFETY EVALUATION - PART B Yes __No X_ Will this item require a change to the station Technical Specifications?

If the answer to the above is "Yes," identify the specification(s) affected and/or attach the applicable pages(s) with the change(s) indicated.

(5) SAFETY EVALUATION - PART C As a result of the item to which this evaluation is applicable:

Yes No

Will the probability of an accident previously evaluated in the FSAR be increased?

Yes No X Will the consequences of an accident previously evaluated in the FSAR be increased?

Yes No . May the possibility of an accident which is different than any already evaluated in the FSAR be created?

Yes No ) Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

Yes No .Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

Yes No X May the possibility of malfunction of equipment important to safety different than any already evaluated in the FSAR be created?

Yes No Will the margin of safety as defined in the bases to any I Technical Specification be reduced?

If the answer to any of the preceding is "Yes", an unreviewed safety question is involved. Justify the conclusion that an unreviewed safety question is or is not involved. Attach additional pages as necessary.

(6) PREPARED BY: rM, t'6 04 DATE: '

(7) REVIEWED BY: DATE:

(8) Page 1 of -I--

PT/O/A/4150/ IA Page 1 of 11 DUKE POWER COMPANY McGUIRE NUCLEAR STATION CONTROL ROD WORTH MEASUREMENT:

ROD SWAP 1.0 Purpose 1.1 To verify that the reactivity worth of the Reference RCC bank, as determined through reactivity computer measurement data, is consistent with design predictions.

NOTE: The reference RCC bank is the bank which has the -:

predicted highest reactivity worth of all control and shutdown banks when inserted into an otherwise unrodded core.

1.2 To verify that the reactivity worth of each control and shutdown bank (except the reference bank), as inferred from data following iso-reactivity interchange with the reference bank, is consistent with design predictions.

2.D References 2.1. Rod Bank Worth Measurements Utilizing Bank Exchange, WCAP-9863-A, May 1982.

2.2 Control Rod Worth Measurement, PT/0/A/4150/11 2.3 Post Refueling Controlling Procedure for Criticality, ZPPT, and Power Escalation Testing, PT/0/A/4150/21 2.4 Technical Specifications 3.4.1.1j 3.10.4, 3.10.3, and 3.10.2.

3.0 Time Required 3.1 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , I engineer 4.0 Prerequisite Tests itial/Da te 4.1 PT/O/A/4150/10, ARO Boron Endpoint Measurement 5.0 Test Equipment 5.1 Reactivity Computer (with flux signal from top and bottom of one power range channel).

PT/O/A/4150/1lA Page 2 of 11 5.2 Two two-pen strip chart recorders. One chart recorder shouid have reactivity (on a scale of 10 pcm/inch, with 0 pcm being the center of the recorder sheet), and T avg from one loop (on a scale of 1F/inch for 556 to 558OF set up on one side of the recorder sheet). The other chart recorder should have flux (on a scale of 0 to the top end of the testing decade in amps) and pressurizer level (on a scale of 10% level/inch). Chart speeds should be 1 inch/min.

NOTE: The specifications in this step may be altered by the Test Coordinator as necessary to accommodate equipment limitations, as long as all four signals are recorded or trended.

6.0 Limits and Precautions 6.1 The NC system temperature is controlled preferably by steam dymp to the condenser. Temperature control may alternatively be affected by steam generator blowdown.

6.2 Normally all reactor coolani pumps should be operating for maximum mixing in the NCS. If all reactor coolant pumps are not operating, the operating pumps should be those on the NCS charging loops (A and/or D). See Tech Spec 3.4.1.1 and 3.10.4 if all reactor coolant pumps are not operating.

6.3 The rod insertion limit and bank overlap sequence will be violated during this test. The operators should-be made aware in advance and should anticipate the associated alarms.

Technical Specification 3.10.2 and 3.10.3 allows for this.

6.4 Maintain the flux level in the zero power test range established in Reference 2.4.

6.5 Prior to switching the rod control selector switch from one bank to another, verify both groups of the bank (if the bank has two groups) are at the same position in order to avoid group misalignment.

PT/0/A/4150/IlA Page 3 of 11 7.0 Required Unit Status Initial/Date

/ 7.1 The unit is just critical in the Startup Mode (Mode 2) at zero power with the flux level in the zero power test range established in PT/O/A/4150/21, "Post Refueling Controlling Procedure for Criticality, ZPPT, and Power Escalation Testing."

/ 7.2 Record in the log the unit to which this-test applies.

8.0 Prerequisite System Conditions NOTE: The following steps may be signed off in any order.

/ 8.1 The reactor coolant system temperature is 557'F +1, -5*F.

NOTE: Maintain NCS temperature within +/-I*F of established temperature during the test.

I 8.2 The difference between NC loop, pressurizer, and VCT boron concentrations is less than 20 ppm. List on Enclosure 13.3.

NOTE: Do not use the boronometer.

/ 8.3 Xenon worth rate is changing less than +/-.1 pcm/min.

/ 8.4 Test equipment is set up per Section 5.0.

/ 8.5 All available pressurizer heaters are on as needed, in order to improve mixing by maximizing the pressurizer spray.

/ 8.6 All control and shutdown banks are fully withdrawn except Control Bank D which is at a position greater than about 215 steps withdrawn.

/ 8.7 The Rod Control Selector switch is in Bank Select Mode set on Control Bank D.

I 8.8 Complete Enclosure 13.1 with the predicted data. See Enclosure 13.9 for an explanation of nomenclature used in this test.

/ 8.9 Trend the points listed in Enclosure 13.2 every 15 minutes or less on the OAC.

9.0 Test Method The RCC bank with the highest predicted value of reactivity worth is measured using the dilution technique per PT/0/A/4150/11. This bank serves as a reference. The integral worth of the remaining RCC banks is implied from the difference in the critical rod position of the reference bank with and without the insertion of bank being tested.

The implied integral worths are then compared to predicted rod worths.

\

PT/O/A/4150/11A Page 4 of 11 10.0 Data Required I0.1 The following conditions for the approximate time of criticality before and after each bank exchange, recorded on Enclosure 13.5:

Time NCS Tavg NCS Boron Concentration Just critical height of reference bank 10.2 Nuclear design predictions on Enclosure 13.1.

10.3 Boron concentration information for the NCS and pressurizer on Enclosure 13.3.

10.4 A copy of the rod positions and rod worths for the reference bank from Enclosure 13.1 of PT/O/A/4150/11 when this test is complete.

10.5 The calculated, implied integral worth (W ) for each RCC bank.

except .the reference bank. List data on xEnclosure 13.7. Sp 10.6 The percent difference between inferred and predicted worths for each individual RCC banks (eI) and for the sum of all banks (C2) on Enclosure 13.8.

11.0 Acceptance Criteria 11.1 The absolute value of the percent difference between measured and predicted integral worth for the reference bank is 515%

(from Enclosure 13.8 (Ci) i ý15%).

11.2 From Enclosure 13.8, the calculated value &2 <10%.

11.3 For all RCC banks other than the reference bank; either:

a) From Enclosure 13.8, ce1 30% for each bank or x

b) WI - WP <200 pcm for each bank, whichever is greater.

x x

PT/O/A/4150/IIA Page 5 of 11 12.0 Procedure Initial/Date NOTE: See Enclosure 13.9 for an explanation of all nomenclature used in this test.

_ / _ 12.1 Request NCS and Pressurizer samples to be taken at approximately 15-20 minute intervals until all banks are measured.

NOTE: The Boronometer may not be used for NC loop concentrations.

NOTE: Notify Chemistry that the unused portions of the samples should be retained in appropriately labeled containers, for possible future re-analysis, until all acceptance criteria are met or as specified by the test coordinator.

12.2 Measure the integral reactivity worth of the reference bank as follows:

NOTE: The reference bank is defined as that bank which is predicted to have the highest worth, of all control and shutdown banks, when inserted into an otherwise un-rodded core (see Enclosure 13.1 for the-identity of this bank). In this procedure, the banks will be referred to by the bank number, the reference bank being number 1. If the reference bank is currently positioned at less than 228 steps withdrawn (i.e., if it is Control Bank D), continue with step 12.2.5. Mark steps 12.2.1 to 12.2.4 NA. If the reference bank is positioned at 228 steps withdrawn, continue on at Step 12.2.1.

12.2.1 Insert the reference bank I until the indicated

/ reactivity is approximately -10 pcm.

12.2.2 Withdraw the bank inserted below 228 until the indicated reactivity is approximately ÷10 pcm.

12.2.3 Repeat steps 12.2.1 and 12.2.2 Until the previously inserted bank is fully withdrawn.

12.2.4 Adjust the position of the reference bank I until the reactor is just critical. Record this position in the test log.

-J 12.2.5 Perform Control Rod Worth Measurement per PT/0/A/4150/11 on the reference Bank 1.

PT/OfA/4150/11A Page 6 of 11 I 12.2.6 Attach a completed copy of PT/O/A/4150/11 Enclosure 13.1 to this procedure.

I 12.2.7 Record the total reference bank rod worth from PT/O/A/4150/11 Enclosure 13.1 on Enclosure 13.7 and 13.8 as shown.

/ 12.2.8 Ensure the reactor is critical at the same reference bank position as was obtained at the end of PT/O/A/4150/11.

12.4 Measure the reactivity worth of the remaining control and shutdown banks, relative to the reference Bank 1, as follows:

NOTE: The relative worth of each RCC bank is obtained from the critical position of the reference bank (initially nearly fully inserted) after full insertion of the bank being measured (initially fully withdrawn), at constant RCS boron concentration.

NOTE: Perform rod swap measurements on Control Bank D last if possible.

/ 12.4.1 Record the initial critical bank configuration on (4)__ Enclosure'13.4 and 13.5"for the reference bank. Also 3 4 5 6 7 record the initial NC boron concentration and average T on Enclosure 13.5.

avg 8 9

/ 12.4.2 Insert bank 2 (identify this bank on top of Enclosure (4) __ 13.5; i.e., Bank 2 is S/D E or Cont. B, etc.) until 3 4 5 6 7 the reactivity indicated by the reactivity computer is approximately -20 pcm.

8 9

/ 12.4.3 Withdraw the reference Bank I until the indicated (4) _ reactivity is approximately +20 pcm.

3 4 5 6 7 NOTE: Maintain the flux within the zero power test range established in Reference 2.4.

8 9

PT/0/A/4150/ 1A Page 7 of 11 12.4.4 Repeat Steps 12.4.2 and 12.4.3 until bank 2 is fully (4) inserted. Keep the indicated reactivity within 3 4 5 6 7 +/-20 pcm.

8 9 12.4.5 Adjust the position of reference bank I until the (4) reactor is just critical. Record the final critical 3 4 5 6 7 configuration data on Enclosure 13.5. Also record the final NC Boron Concentration and average T avg on S 8 9 Enclosure 13.5.

NOTE: If time permits, measure the differential reactivity worth of reference Bank 1 with the reactivity computer by sequential bank insertions and withdrawals around the critical position. Record information in the test log if this is performed.

(Analysis may be performed at a later time.)

12.4.6 Insert the reference Bank I until the indicated (4) reactiviiy is approximately -20 pcm.

3 45 67 89 12.4.7 Withdraw bank 2 until the indicated reactivity is (4) approximately +20 pcm.

3 4 5 67 8 9

/ 12.4.8 Repeat Steps 12.4.6 and 12.4.7 until bank 2 is fully (4) -

withdrawn.

3 4 5 6 7 8 9

/ 12.4.9 Adjust the position of reference Bank 1 until the (4) .'eactor is just critical. Record the critical 3 4 5 6 7 configuration data on Enclosures 13.4 and 13.5.

8 9

PT/O/A/4150/1IA Page 8 of 11 12.4.10 Repeat Steps 12.4.2 through 12.4.9 for the remaining, unmeasured control and shutdown banks numbered 3 through 9 instead of bank 2. The banks may be measured in any order except that Control Bank D should be measured last. Identify the bank beside the bank number on Enclosures 13.4, 13.6, 13.7 and 13.8.

NOTE: If a Control Bank D-in ITC measurement is to be made, perform Steps 12.5, 12.5.1 and 12.5.2. If not, proceed directly to Step 12.5.3 and mark Steps 12.5, 12.5.1 and 12.5.2 as N/A.

/ 12.5 After all rod measurements have been made, again swap Control Bank D for the reference bank 1.

/ 12.5.1 By NC Boron Adjustment, reposition Control Bank D and the reference Bank I such that Control Bank D is almost fully inserted into the core and the reference bank 1 fully withdrawn from the core. (It is P acceptable to have Control Bank D fully inserted and the reference bank 1 almost fully withdrawn.)

/ 12.5.2. Perform PT/O/A/4150/12B, Moderator Temperature Coefficient of Reactivity During Startup Mode.

/ 12.5.3 By NC boron adjustment, reposition control and shutdown banks to the desired normal operating configuration of Control Bank D at about 215 steps withdrawn. Do not go out of Bank Control.

12.6 Perform the following steps once Control Bank D is about 215 steps withdrawn.

/ 12.6.1 Go to the Master Cycler Cabinet and reset the Bank Overlap Digital Counter to 000 by pushing the reset button.

/ 12.6.2 Reset the Bank Overlap Counter to 345 plus the present Control Bank D position by pushing the button to count up from 000 to the desired value (one push of the button is one digit change on the display).

/ 12.6.3 Place rod control to manual.

NOTE: This completes the data acquisition section of this test.

/ 12.7 If boron samples are no longer needed to be gathered, notify Chemistry.

PT/O/A/4150/1IA Page 9 of 11 12.8 Compute the average reference bank critical position on Enclosure 13.4.

12.9 Compute the inferred worth for each control and shutdown bank (except the reference bank 1) as follows:

12.9.1 Using the data from Enclosure 13.4, and the worth measurement data for the reference bank from Enclosure 13.1 of PT/O/A/4150/11, compute the value of as described below.

[](h*) avg where:

(h avg _

0 is the of reference integral the measured worth bank from 0 steps to (h)o av from Enclosure 13.f gf PT/O/A/4150/11.

and (h avg is the average of the initial and return critical positions of the reference-bank before and after interchange with bank x as given on Enclosure 13.4.

Fill in all blanks and complete the calculations on Enclosure 13.4 in the appropriate column.

PT/O/A/4150/I A Page 10 of 11

/ 12.9.2 Using the data from Enclosure 13.5, the worth measurement data for the reference bank from Enclosure 13.1 of PT/O/A/4150/11 and the design data of Enclosure 13.1, compute the value of ax (L2)px as described below:

ax (A2(AP 2)x =h I*I 228 where: Ml 228 is the measured integial worth of the R h "I reference bank from h to the fully withdrawn position from PT/O/A/4150/11, Enclosure 13.1. 0 hM is the measured critical position of the

x. reference bank after interchange with bank x from Enclosure 13.5.

and is a correction factor from Enclosure 13.1 x to account for the influence of bank x on the worth of the-*reference bank.

Fill in all blanks and complete the calculations on Enclosure 13.6.

/ 12.9.3 Compute the inferred integral worth of each bank x, x as indicated on Enclosure 13.7.

W1,

PT/O/A/4150/1lA Page 11 of 11 12.9.4 Compute the percent difference between inferred and predicted worths for each individual RCC bank and the sum of all banks described below.

WI- WP (Ci)x x x x 100, in N N 7- . W.

= i=1 =1- x 100, in %

2 =N 7-WP i= I Fill in all blanks and summarize the calculations on Enclosure 13.8.

_/ 12.10 Verify all acceptance criteria have been met.

/ 12.11 Inform Chemistry to discard the Chemistry samples they have saved, once all results of this.test are acceptable. _

13.0 Enclosures 13.1 Nuclear Design Predictions for Rod.Interchange Measurements 13.2 PAO Data 13.3 Log of Boron Concentrations 13.4 Calculation of (AP1 )x 13.5 Critical Configuration Data 13.6 Calculation of a x(Ap2 )x 13.7 Calculation of Inferred Integral Bank Worths 13.8 Comparison of Inferred Bank Worths with Design Predictions 13.9 Nomenclature

PT/O/A/4150/11A Page 1 of 1 Control Rod Worth M1easurement: Rod Swap Enclosure 13.1 Nuclear Design Predictions for Rod Interchange Measurements McGuire Unit Cycle (b) (c)

Bank Bank WP h No. rdentity Wx) + (pcm) (steps)

(a) 1 (Reference) 2 3

4 5

6 7

8 9

(a) Reference bank - the bank with the highest predicted integral worth.

(b) Reference bank critical position after interchange with bank x.

(c) Ratio of integral worth of the reference bank from hP to the fully withdrawn with and without x in the core.

position

+ Control Bank C, Shut- .n Bank E, etc.

NOTE: See Enclosure 13.5 .or a complete listing of nomenclature used in this test.

Recorded By Date This data came from (list source):

PT/O/A/4150/IlA Page 1 of I Control Rod Worth Measurement: Rod Swap Enclosure PAO Data13.2 A0821 Boric Acid Makeup Blended Flow P1390 Control Bank A Position P1391 Control Bank B Position P1392 Control Bank C Position P1393 Control Bank D Position P1546 Shtudown Bank A Position P1547 Shutdown Bank B Position P1548 Shutdown Bank C Position P1549 Shutdown Bank D Position P1550 Shutdown Bank E Position A0632 Intermediate Range Channel N35 A0633 Intermediate Range Channel N36 A0819 Loop A T

-A0825 Loop B T avg" A0831 Loop C T avg A0837 Loop D Tavg A1058 Loop A AT A1070 Loop B AT A1082 Loop C AT A1094 Loop D AT A1118 Pressurizer Pressure A0602 Boronometer Al124 Pressurizer Level _

P1461 NC Avg. T avg

P0828 Avg. Incore T/C Temperature __

A0603 Boric Acid Flow to Blender _

A1064 NC Loop A NR Cold Leg Temperature A1076 NC Loop B NR Cold Leg Temperature A1088 NC Loop C NR Cold Leg Temperature \_j Al100 NC Loop D NR Cold Leg Temperature ý-j

PT/O/A/4150/IIA Page 1 of 1 Control Rod Worth Measurement: Rod Swap Enclosure 13.3 Log of Boron Concentrations NCS Boron Concentrations Time Recorded Date Sample Taken +VCT NCS Press. Comments By IL-j McGuire Unit Cycle NOTE: VCT sample needed only once at start of the test. Mark this block as N/A after this.

PT/O/A/4150/I A Page I of I Control Rod Worth Measurement: Rod Swap Enclosure 13.4 Calculation of (1pOx Unit Cycle Date Bank(x)) *1" Bank W (steps) (AP No. Ident. +Initial *Return -,--Average (pcm) 2 3

4 5

6 7

8 9

+Step 12.4.1 - reference bank initial critical position.

Step 12.4.9 - reference bank final critical position upon exchange with bank x (bank x if out of core).

+'2,tp 12.9.1 *Step 12.8 i;cirded By

LCCCCCCCCCCC ( ((CCCCCCC(((CCCCCCCCC-CCCCCCCCCC( ( C C PT/O/A/4150/I IA Page I of I Control Rod Worth Measurement: Rod Swap lit fcle Enclosure 13.5 3te .'Critical Configuration Data NC NC Reference Bank RCC Batik Positions Time' T Boron Position (steps) aNNNConc. No. 4 No. 5 No. 6 No. 7 No. 8 No. 9 (hHo (itHx No. 2 No. 3 (ppm) 0 x ( (

(h rs". ) (OF)

+ N/A 228 228 228 228 228 228 228 228 0 228 228 228 228 228 228 228 N/A _

__ N/A 228 228 228 228 228 228 228 228

228 0 228 228 228 228 228 228 N/A N/A N 228 228 228 228 228 228 228 228 228 0 228 228 228 228 228 N/A *I. 228 228 228 228 228 228 228 228 228

_ N/A 228 228 228 _0 228 228 228 228 N/A

_-_" _ N/A 228 228 228 228 228 228 228 228 2128 228 228 228 0 228 228 - 228 N/A

_,___ N/A 228* 228 228 228 228 228 228 228

228 228 228 228 .. 228 . 0 228 228 N/A

,':N/A

-*_ 228 228 228 228 228 228 228 228 228 228 228 228 228 228 0 228..

_N/A *

.1 NA 2281 228 228 228 228 228 228. 228..

N____.___

A . 228 228 228 228 228/ ... 228 228 (

""N N/A 228 228 228 228 228 228 228 228-ý L":'

Step 12.4.5 - final critical bank position SLtp 12.4.9 Step 12.4.1 - initial critical bank position ecorded By

PT/0/A/4150/llA Page 1 of 1 Control Rod Worth Measurement: Rod Swap Enclosure 13.6 Calculation of a (A42)

Unit Cycle Date (1) (2) (1) x (2)

+from Enclosure 13.5

from Enclosure 13.1 Recorded By

PT/O/A/4150/11A Page 1 of I Control Rod Worth Measurement: Rod Swap Enclosure 13.7 Calculation of Inferred Integral Bank Worths Unit Cycle Step 12.2.7 = _______pcin Date 2) x( O (Ix (a)

Bank (x) +a No. Ident. (pcm) (pcm) (pcm) 2 3

4 5

6 7

8 9

(a) WI = WR - (A° ) - axW°)

+from Enclosure 13.4

'from Enclosure 13.6 Recorded by

PT/O/A/4150/11A Page 1 of 1 Control Rod Worth Measurement: Rod Swap Enclosure 13.8 Comparison of Inferred Bank Worths With Design Predictions Unit Cycle Date Bank (x) WI x (e) x No. Ident. (pcm) (pcm) 1 +

reference 2

3 4

5 6

7 8

9 W (pcm) W Pm

+Step 12.2.7: Record the measured worth of tf- :tference bank here.

from Enclosure 13.1

++from Enclosure 13.7 Recorded By

DuKE PowER COMPANY P.0. DOX 33180 OCtAULOT~r, N.O. 28249 HAL IL TUCKER yim -ONUO (7o4) 0*7-4881 WiuA, ,iG March 11, 1987 U.S. Nuclear Regulatory Comission Document Control Desk Washington. D.C. 20555

'-I

Subject:

McGuire Nuclear Station Docket Nos. 50-369/370 k-i Catavba Nuclear Station Docket Nos. 50-413/414 "k-i Determination of Rod Worth Using Rod Swap Methodology Gentlemen:

Pursuant to telecons of February 19, 1987 and March 10, 1987 between D.S. Hood (ONRR) at. al., and S.A. Gewehr (DPC) et. al., attached are revised responses to Questions 6 and 7 of D.S. Hood's request for information dated January 12, 1987.

Very truly yours,

,-.- '-I

"'V Hal B. Tucker SAC/61/jgm Attachment

,)

k-I 870318008B 870311 PDR ADOCK 05000369 Cb =V%

V.-)

Page I ATTACHHENT QUESTION 6: Provide data for at least 2 sets of side-by-side comparisons of Boron dilution and Rod Swap Data - predicted and measured. The data may be either for your plants or measured data from another plant and predictions by Duke.

RESPONSE

In the original Nuclear Physics Methodology Topical, DPC-NT-2010A, Duke Power Company benchmarked its methods for predicting rod worths against measurements made during the startup testing for both initial cores at the McOuire Nuclear Station. These measurements were made using the boration/ dilution technique for determining rod worths in sequential insertion. In its review of this topical, the NRC accepted the capability of Duke Power to adequately predict control rod vorths and shutdown margin using the outlined methodology.

In the Rod Swap Methodology Report recently sent to the Commission, Duke Pover benchmarked its'methodology for predicting rod vorths using the rod swap technique against 5 cycles of actual rod swap measurements. This methodology utilized the same computer codes previously benchmarked in DPC-NF-2010A. All predictions. when compared to the measured results, met the acceptance criter ia as outlined in the rod swap plant procedure.

It has been noted in previous conversations with the NRC that the two bench marking studies noted above do not make comparisons of the same units for the same cycles. It is Duke Power's position that there is really no benefit from this type of comparison. A valid comparison cannot be expected since boration/

dilution is a sequential measured worth calculation and rod swap consists of a summation of the wortha of each rod Individually inserted into an otherwise unrodded core. It is therefore impossible to make direct comparisons between wortha of the two methods. The only thing that can be looked at is the percent difference between measured and predicted for th two methods. When looking at percent differences between measured and predicted. one does not have to look at the same unit and cycle to verify methodologies are correct.

Comparisons of predicted and measured rod worths done using boration/ dilution and rod swap on the two Catawba units are enclosed. The boration/dilution technique was used to measure rod worths in sequential bank insertion for the Catawba 1 Cycle I core while Catawba 2 Cycle 1 measurements were done using the rod swap technique (Table 1). From a neutronics standpoint, the two cores are almost identical. This assumption can be justified by examining the core loadings and the results of the Zero Power Physics Testing for each of the units. Several key parameters concerning the core are shown in Table 2. Also enclosed are the quarter core loading pattern (Figure 1) and a comparison of the quarter core assembly power distribution from the zero power map taken during the startup physics testing (Figure 2).

It should also be pointed out that the rod vorths from the rod g-jap predic tions are not the worth. used to calculate the shutdown margin. Rod swap only verifies the code's ability to predict rod worths. The rod worth used in the shutdown margin calculation is the N-1 worth. .A Duke Power has provided a total of nine cycle of predicted rod worth comparl sons to measured data with good to excellent results. This demonstrates the A_

ability of the codes and methods used to adequately model reactivity effects due to control rods in any configuration. Therefore, the use of Duke Power predictions in the verification of shutdown margin with appropriate factors of conservatism applied to the calculation as outlined in DPC-NF-2010A Section 4.2.2.2 is justified. k'a

.S,

Page 2 ATTAC1ENT-1 QUESTION 7: What Organization does the safety analysis for the Duke Plants?

When this is not done by Duke, what is done (e.g. tests, comparisons, etc.) to show that the startup test results adequately represent the plant features and assumptions used in the safety analyses? 4_

RESPONSE: J The safety analyses for the MeCuire and Catawba Nuclear Stations have been performed by the current fuel vendor. The analyses utilized NRC-approved codes and methodologies and conservative input assumptions inclujing values for key nuclear physics parameters such as reactivity coefficients, core power distributions, and shutdown margins, which are expected to bound the actual values of these parameters for current and future reload cor'es. An evaluation is performed for each reload cycle which consists of comparing nuclear design predictions to the safety analyses assumptions to ensure the safety analyses remain bounding. The cycle-specific evaluation process Is described in WCAP-9272,

"Westinghouse Reload Safety Evaluation Metho!ology." Core physics testing performed for each cycle verifies the nuclear design predictions and ensures the ý_j actual core physics parameters are conservative with respect to the safety analyses.

The main safety analysis assumption verified by the rod swap procedure is that the

plant will maintain adequate shutdown a"rgin per Technical Specifications. One of the purposes of rod swap measurements and comparisons to predicted values is to verify the accuracy of the total rod worth prediction used so an input to the shutdown margin calculation. An independent Duke Power shutdown margin is evaluated for each cycle using methods approved by the NRC in DPC-NF-2010A. The 4 N-i rod worth used in this prediction is reduced by IOX for conservatism.

Acceptance criteria listed in the procedure indicate that the total inferred rod

"'*-1 worth as meaaured in the rod swap testing must be within laZ of the total predicted worth, If the total measured rod worth is less rhan the predicted worth by more than 10%, a review of the shutdown margin is made to determine if the current rod insertion limits provide adequate shutdown margin. If the shutdown margin is adequate. then no revision of the limits is necessary. However, if the "margin is not maintained, then Duke will notify Westinghouse, revise the rod insertion limits, and submit any necessary changes to Technical Specifications to the NRC. i In order to tie the rod swap measurements to the verification of inputs to the "safety analysis, Duke Power will parform an independent shutdown margin for each reload cycle using methods approved by the NRC in DPC-NF-2010A. In addition, for each cycle where Duke generates the rod swap prediction but the safety analysis has been performed by a vendor, a comparison between the Duke and vendor predicted total rod worth will be made at beginning-of-cycle, hot zero power conditions. ir Any significant discrepancies will be documented, reviewed, and resolved prior to startup physics testing.

Reference McGuire Nuclear Station, Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testing, DPC-NE-1003, Rev. 1, December 1986.

TABLE I Rod Worth Measurement Data Comparison of Rod Swap and Boration/Dilution Techniques Boron/Dilution Rod Swap Integral Worths Integral Worths Predicted Measure d Predicted Measured '- 4 Bank (PCM) (PCM) Z Diff** (PCM) (PCM) Z Diff**

1 .

D 772 794 -2.85 773 788 -1.94 C 790 849 -7.47 1214 1203 0.91 B* 852 882 -3.52 1190 1171 1.60 A 249 250 -0.40 572 548 4.20 SE 377 385 -2.12 508 460 9.45 SD 497 525 -5.63 755 772 -2.25 SC 497 522 -5.03 1098 1099 -0.09 SB 765 834 -9.02 SA 674 706 -4.75 N-1 - 7370 7414 -. 60 N 5473 5747 -5.01 C-

Reference Bank

- Z Diff - [(P-M)/P]*100 x.

TABLE 2 Catavba 1 Cycle 1 and Catawba 2 Cycle 1 Comparison of Core Parameters Unit I Unit 2 KG U/ASSY Batch 1 1.6 424.169 424.623 Batch 2 2.4 423.508 425.805 Batch 3 3.1 423.676 424.519 AVE ER Batch 1 1.6101 1.6104 Batch 2 2.3999 2.4014 Batch 3 3.1022 3.0954 ARO BORON ENDPOINT (PPM!B) 975 975 A

ISO. TEMP.

COEFF (PCH/8F) -1.745 -1.81

Figure 1 CATAWBA I CYCLE I AND CATAWBA 2 CYCLE 1 QUARTER CORE LOADING PATTERN H 6 F E 0 C B A 1 4 4 0 I 4

4 8

1.69 # 2.49 6 1.68

2.41

1.68

2.40

1.68

3.19 *

a 16

12.

16 .* 6

9 a 2.41 0 1.69

2.49

1.60

2.43

1.69 # 3.18

3.19 *

16

12 *a 12 #

20 4

4 4 . 4 4 . 4 a 1

1.69

2.49

1.68 . 2.48

1.60 0 2.49

1.68 6 3.10 #

0

' 12'

12.

16*

6 4 11 w 2.40 0 1.63 a 2.49 . 1.698 2.40 o 1.60

3.19 & 3.190

12.

12. i&

16 16 **

i tlli t~i ti ii ~tt~i It*,ll~t~t*fi

l i t 12

1.69 4 2.43 I 1.63

2.49 . 2.48 a 2.48

3.19 .

a 0 12 .0 16 #

I 16 I I

4 4 4 4 4 4 13 P 2.43 I 1.63 a 2.48

1.68 4 2.490 3.18

3.19 *

16 ' 16.

16 15. a 14

1.638 3.13 a 1.69 3.11 3.1 ,

3.19 4 ENRICHMENT

23

2 16

4 NUMBER OF BA'S PER ASSY 15 6 3.19 I 3.19 0 3.18 0 3.13 #

64* 0 6 4 I

II*I*I*04*II*****4**I*.a*Ia.III

Figure 2 CICI AND C2CI HZP POWER DISTRIBUTIONS HWD/lNTU ARO. HZP. NO XE, Nd Sfl H F E D C B A '

.72 0 .82 # .84 # 1.98 # .69 # .97 * .95 a .97 '

a * .72 # .82 # .82 # .97 # .87 * .94 * .93 # .95 *

0.g9 f 9.99 # 2.44 p 3.09

2.38 # 3.J9

2.15 # 2.11 *

.aeeeii**. II4*4**e45I4*I*I*I*I**,*.I..,.*..*...,..**..........ei...,,..i 0 .92 4 .79 a .95 * .88 # 1.12 * .91

1.11 # 1.91 # ,--'

9 # .02 8 .78 # .95 & .89

1.81 * .99

1.19

1.99 *

0.89

1.28

9.88 0 9.09 .99 4 2.25 # .91 4 1.00

.93 a .96 # .89 a 1.94 4 .93 # 1.01 # .93 * .90 & "_

19 f .83 0 .95 0 .99 # 1.015 .93 # 1.01 0 .93 & .69 a' f 0.99

1.85 a 9.al8l -. 95 0.08 8 a 9.990 # 8.80 1.12 ,

1.91 # .98 & 1.96 * .99 a 1.12

1.03 # 1.16 4 .73 .

31 1.90 # .89 # 1.e6 0 .99 a 1.12

1.93

1.17 * .74

1.38

1.12 , 1.90 , 9.e1 , e.e0 , 8.98 a -. 65 # -1.35 *

.91

1.94 * .94

1.121 1.43 a 1.195 1.22 &

12 F .91 4 1.04 a .93

1.13

1.44 a 1.19

1.23 #

9.99 09.81

1.09 * -. 89 a -. 69

9.80 a -. 81 .

.98 # .92 a 1.83

1.04 4 1.19

1.22 4 .69 :

13 # 1.91 0 .94 # 1.84 # 1.06 a 1.20

1.21 * .68 0 -2.97 0 -2.13 * -. 06 * -1.69 # -. 63 * .03

1.14 a 4 .97 v 1.14 # .96

1.16 0 1.22 .86

14 # .96 # 1.16 # .97

1.19

1.22 * .9 *9

-1.32 4 -1.72 0 -1.93 # -. 84 # 9.09 * -1.12 ,

4 1.18 4 1.34 * .92 * .74

CICI 15 . 1.98 # 1.84 * .93 a .76 p C2C" 9.90 8 # 8.99 * -1.98 * -2.63 a Z 01FF CICI CORE AVERAGE 1.09 C2Cl CORE AVERAGE 1.89 *.

Z DIFF CORE AVERAGE -. 95 T-%

CICl MAXIMUM NAGNITUDE IS 1.4Z AT ASSEMBLY D - 12 C2C1 MAXIMUN NAGNITUDE IS 1.44 AT ASSEMBLY 0 - 12 Z 0IFF MAXIMUM MAGNITUDE IS 3.19 AT ASSEMBLY C - 8 PERCENT ERROR BETWEEN THE MAXIMUM VALUES IS -. o9 AVERASE ABSOLUTE RELATIVE ERROR .98 PERCENT ROOT MEAN SgUARE OF THE RELATIVE ERROR 1.22 PERCENT ROOT MEAN SQUARE OF THE DIFFERENCE 1.16 PERCENT

"0 UNITED STATES

. oNUCLEAR REGULATORY COMMISSION 0" WASHINGTON, D.C. 20555-0001 Z mJune26, 2002 Mr. M. S. Tuckman Executive Vice President Nuclear Generation Duke Energy Corporation 526 South Church St Charlottte, NC 28202

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2 AND MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 RE: REQUEST FOR ADDITIONAL INFORMATION - APPLICATION FOR CHANGES TO TECHNICAL SPECIFICATIONS (TAC NOS. MB3343, MB3344, MB3222 AND MB3223)

Dear Mr. Tuckman:

The Nuclear Regulatory Commission is reviewing your application dated October 7, 2001, entitled "License Amendment Request applicable to Technical Specifications 5.6.5, Core Operating Limits Report; Revisions to Bases 3.2.1 and 3.2.3; and Revisions to Topical Reports DPC-NE-2009-P, DPC-NF-2010, DPC-NE-201 1-P, and DPC-NE-1 003" and has identified a need for additional information as identified in the Enclosure. These issues were discussed with your staff on June 6, 2002. Please provide a response to this request within forty-five (45) days of receipt of this letter so that we may complete our review.

Sincerely, betE ari, Senior Project Manager, Section 1 Project Directorate Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-413, 50-414, 50-369 and 50-370

Enclosure:

Request for Additional Information cc w/encl: See next page

McGuire Nuclear Station cc:

Ms. Lisa F. Vaughn Ms. Karen E. Long Legal Department (PBO5E) Assistant Attorney General Duke Energy Corporation North Carolina Department of 422 South Church Street Justice Charlotte, North Carolina 28201-1006 P. 0. Box 629 Raleigh, North Carolina 27602 County Manager of Mecklenburg County Mr. C. Jeffrey Thomas 720 East Fourth Street Manager - Nuclear Regulatory Charlotte, North Carolina 28202 Licensing Duke Energy Corporation Michael T. Cash 526 South Church Street Regulatory Compliance Manager Charlotte, North Carolina 28201-1006 Duke Energy Corporation McGuire Nuclear Site Elaine Wathen, Lead REP Planner 12700 Hagers Ferry Road Division of Emergency Management Huntersville, North Carolina 28078 116 West Jones Street Raleigh, North Carolina 27603-1335 Anne Cottingham, Esquire Winston and Strawn Mr. Richard M. Fry, Director 1400 L Street, NW. Division of Radiation Protection Washington, DC 20005 North Carolina Department of Environment, Health and Natural Senior Resident Inspector Resources c/o U.S. Nuclear Regulatory Commission 3825 Barrett Drive 12700 Hagers Ferry Road Raleigh, North Carolina 27609-7721 Huntersville, North Carolina 28078 Mr. T. Richard Puryear Dr. John M. Barry Owners Group (NCEMC)

Mecklenburg County Duke Energy Corporation Department of Environmental 4800 Concord Road Protection York, South Carolina 29745 700 N. Tryon Street Charlotte, North Carolina 28202 Mr. Peter R. Harden, IV VP-Customer Relations and Sales Westinshouse Electric Company 5929 Carnegie Blvd.

Suite 500 Charlotte, North Carolina 28209

REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST APPLICABLE TO TECHNICAL SPECIFICATION 5.6.5, CORE OPERATING LIMITS REPORT, REVISIONS TO BASES 3.2.1 and 3.2.3 REVISIONS TO TOPICAL REPORTS DPC-NE-2009-P, DPC-NF-2010, DPC-NE-2011-P, AND DPC-NE-1 003 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 MCGUIRE NUCLEAR STATION, UNITS 1 and 2 DUKE ENERGY CORPORATION Topical Reports Numbered DPC-NE-2009-P Duke Power Company Westinghouse Fuel Transition Report and DPC-NF-2010-A, Duke Power Company McGuire Nuclear Station and Catawba Nuclear Station Nuclear Physics Methodology for Reload Design

1. Please provide a detailed qualitative technical justification for the requested changes to the topical reports (methodologies), DPC-NE-2011 and DPC-NF-2010. (i.e., why are these changes being made?).

2. To expedite the review process, please pro4 a qualitative and quantitative technical basis for each of the changes in these top* "*rts.

3. Please provide validation data that bench-marks the results of comparisons between the old and the new models (changes).

4. If the changes to these topical reports and methodologies impact the safe operation of the reactor core, please provide the safety significance (impact) of each of these changes.

5. Please provide the basis for why the proposed changes to the above stated topical reports should be found acceptable.

Topical Report Numbered DPC-NF-2010-A, Duke Power Company McGuire Nuclear Station and Catawba Nuclear Station Nuclear Physics Methodology for Reload Design

1. In the revision history section on page ii, the licensee provides the staff with the reason for the submittal. Since this is a licensing action, please list those Technical Specification(s), Bases, FSAR sections, conformance to regulatory documents, criteria, generic letters, etc. that are impacted by the request for these changes within the licensing framework.

2. Section 4.2.4.2, second paragraph. Please provide clarification of this change and the technical justification for it. Please provide a comparison between the old sentence and the new sentence.

3. In Attachment 7a, "Detailed Listing of the Changes to DPC-NF2010A," it is stated in many places, that "this change is made to avoid difficulties with the literal interpretation of the original description." Please provide clarification of this statement with a supporting example.

4. Section 4.2.4.4, fifth paragraph. Please provide clarification of this change and the technical justification for it. Please provide comparison between the old sentence and the new sentence.

5. Section 8.1, first paragraph. Is the added equation the same as that in the current version of the DPC-NF-2010A topical? If not, please provide technical justification for its use.

6. Section 9.1.5, first paragraph. Please provide clarification of this change and the technical justification for it. Please provide a comparison between the old sentence and the new sentence.

Topical Report Numbered DPC-NE-201 1-P-A, Duke Power Company Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors

1. The description of the transient conditions was changed in Tables 1 and 2, of Section 2.5. It is not clear to the staff exactly what was changed. Please clarify.

2. From section 6.1, please explain what is meant by "updated the equation."

3. From section 6.1, please provide further clarification of this statement.

4. Section 6.2, were is UMR listed in section 6.2? Please provide original definition and new definition for comparison.

Topical Report Numbered DPC-NE-1003, Revision 1 McGuire Nuclear Station and Catawba Nuclear Station Rod Swop Methodology Report for Startup Physics Testings, Revision 1

1. Appendix A of topical report DPC-NE-1 003, Revision 1, contains two versions of Duke Power Company's rod swap measurement procedure PT/O/NA4150/11 A: Attachment 3 (dated June 1986) and Attachment 4 (dated April 1984). There are differences in these two versions of the procedure. For example, in the Attachment 3 version, Steps 12.2.2 and 12.2.3, respectively, specify the insertion of bank 1 until the indicated reactivity is approximately -20 pcm, and the withdrawal of reference bank until the indicated reactivity is approximately +20 pcm; whereas in the Attachment 4 version, the insertion and withdrawal of bank 1 and reference bank, respectively, of steps 12.2.1 and 12.2.2 specify reactivity change of -/+ 10 pcm.

a. Since the Attachment 3 version of procedures is more recent, why is the Attachment 4 version referenced in Revision 1 of the topical report (Reference 2)?

b. Which of these two versions of rod swap measurement procedures will be used for McGuire and Catawba Units?

2. In the Attachment 3 version of rod swap measurement procedures PT/O/A/4150/11 A, Step 12.1.3 states that: "Repeat steps 12.2.1 and 12.2.2 until the previously inserted bank is fully withdrawn."

Is there a typographic error in the words "steps 12.2.1 and 12.2.2"? Should correct words be "steps 12.1.1 and 12.1.2"?

3. The equation in Section 3, Measurement Procedure, of the topical report for calculating the inferred rod worth of bank x is different from the equation in Step 12.5.3 of the Attachment 3 procedures. The difference appears to be due to the initial height of the reference bank for performing the rod swap measurement of the measured bank.

Clarify the exact procedure to be used in the rod swap test, and make all necessary corrections in the topical report and the procedures to be consistent.

4. The third sentence in Section 3 of the topical report is revised to read: "All other banks are then exchanged with the reference bank or other test banks at constant boron conditions until the measured bank is fully inserted." It is stated, in Attachment 9a, "Detailed Listing of Changes to DPC-NE-1003A," that the third sentence in Section 3 is revised to make the report consistent with current procedures. The "Revision History" in the topical report states that this revision [Revision 1] also reflects a refinement in the rod swap to make use of two test banks.

a. What are the current procedures? What is the date of the current procedures?

b. Are the current procedures the same or different from the ones in Attachment 3?

The Attachment 3 procedures do not include the exchange of a test bank with the other test bank.

c. If the current procedures are different from those of Attachment 3 or 4, provide a copy of the procedures, and appropriately reference them in the report.

d. Is the statement in "Revision History" referring to this revision? Please explain what the statement means.

PDuke Duke Energy Corporation I Power. S26 South Churh Stret A Duke Energ Company PO Box 1006 Charlotte, NC 28201-1006 (704) 382-2200 OFFICE (7(94)382-4361) FAX' Michael S. Tuckman Executive Vice Prevident Vuclear Generation August 7, 2002 U. S. Nuclear Regulatory Commission Washington D.C. 20555-0001 ATTENTION: Document Control Desk

Subject:

Duke Energy Corporation McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 370 Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 414 Response to NRC Request for Additional Information - TAC nos. MB3222, MB3223, MB3343, and MB3344) and License Amendment Request Supplement This purpose of this letter is to provide Duke Energy Corporation's (Duke) response to an NRC request for additional information (RAI) and to supplement a Duke license amendment request (LAR) previously submitted pursuant to 10CFR50.90.

Please note that some of the information contained in this submittal package has been determined to be proprietary and is being submitted pursuant to 10CFR2.790. This proprietary information is discussed below.

Duke submitted' a LAR applicable to McGuire and Catawba Technical Specifications (TS) 5.6.5.a and 5.6.5.b. Also included in this submittal were proposed revisions to the four Duke Topical Reports listed below.

'Reference 1: Letter, Duke Energy Corporation to U.S. Nuclear Regulatory Commission, ATTENTION: Document Control Desk, Dated October 7, 2001,

SUBJECT:

License Amendment Request Applicable to Technical Specification 5.6.5, Core Operating Limits Report; Revisions to Bases 3.2.1 and 3.2.3; and Revisions to Topical Reports DPC-NE-2009-P, DPC-NF-2010, DPC-NE-201 I-P, and DPC-NE-1003

U. S. Nuclear Regulatory Commission August 7, 2002 Page 2

"* DPC-NE-2009-P, Duke Power Company Westinghouse Fuel Transition Report, Revision 1;

" DPC-NF-2010, Duke Power Company McGuire Nuclear Station and __

Catawba Nuclear Station Nuclear Physics Methodology for Reload Design, Revision 1;

" DPC-NE-2011-P, Duke Power Company Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors, Revision 1;

"* DPC-NE-1003, McGuire Nuclear Station and Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testing, Revision 1.

The NRC RA12 asked questions on these topical reports. As described below, the Duke responses to these questions are included in the attachments to this letter.

In a subsequent submittal, 3 Duke proposed another LAR for McGuire and Catawba TS 5.6.5, but this LAR was only applicable to TS 5.6.5.b. The information contained herein explains the necessary coordination for changing TS 5.6.5.b for McGuire and Catawba. This LAR implements the provisions of an NRC approved Technical Specifications Task Force (TSTF) Standard Technical Specifications Traveler. 4 The NRC has approved and issued this LAR for both McGuires and Catawba.' Implementation of the 2 Reference 2: Letter, U. S. Nuclear Regulatory Commission to Duke Energy Corporation, Dated June 26, 2002,

SUBJECT:

Request for Additional Information, Application for Changes to Technical Specifications (TAC Nos. MB3222, MB3223, MB3343, and MB3344 3 Reference 3, Letter, Duke Energy Corporation to U.S.

Nuclear Regulatory Commission, ATTENTION: Document Control Desk, Dated December 20, 2001,

SUBJECT:

License Amendment Request Applicable to the Technical Specifications Requirements for the Core Operating Limits Report - Oconee, McGuire, and Catawba Technical Specification 5.6.5 4 TSTF-363, "Revise Topical Report References in ITS 5.6.5 COLR" SLetter, U. S. Nuclear Regulatory Commission to Duke Energy Corporation Dated July 10, 2002,

SUBJECT:

McGuire Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC Nos. MB3702 and MB3703) 6 Letter, U. S. Nuclear Regulatory Commission to Duke Energy Corporation Dated July 2, 2002,

SUBJECT:

Catawba Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC Nos. MB3728 and MB3729)

U. S. Nuclear Regulatory Commission August 7, 2002 Page 3 referenced industry traveler eliminates the need for the changes Duke proposed to McGuire and Catawba TS 5.6.5.b in Reference 1.

The LAR supplement transmitted herein deletes the proposed changes to McGuire and Catawba TS 5.6.5.b contained in Reference

1. The attached McGuire and Catawba TS pages (both marked and reprinted versions) update Reference lsuch that it contains the latest approved version of the affected TS pages and only applies to McGuire and Catawba TS 5.6.5.a. The affected TS pages are:

McGuire Units 1 and 2 Pages: 5.6-2, 5.6-3, B3.2.1-11, and B3.2.3-4; and Catawba Units 1 and 2 Pages: 5.6-3, B3.2.1-11, and B3.2.3-4.

As shown, conforming Bases changes have been made and the necessary Bases pages are also included.

The attachments to this letter are listed and described below.

"* Attachment 1 provides the Duke response to the NRC's general questions on Topical Reports DPC-NF-2010 and DPC NE-2011-P.

"* Attachment 2 provides the Duke response to the NRC's specific questions on Topical Report DPC-NF-2010.

"* Attachments 3a and 3b provide the Duke responses to the NRC's specific questions on Topical Report DPC-NE-2011-P.

Attachment 3a is the proprietary version and Attachment 3b is the non-proprietary version.

"* Attachment 4 provides the Duke response to the NRC's specific questions on Topical Report DPC-NE-1003.

" Attachment 5 provides the Duke response to an NRC concern on Topical Report DPC-NE-2009-P. This concern was not included in the NRC's RAI, 2 however it was discussed during an NRC/Duke telephone conference held on July 24, 2002.

U. S. Nuclear Regulatory Commission August 7, 2002 Page 4

"* Attachments 6a and 6b provide a marked copy of the existing approved Technical Specifications pages for McGuire Units 1 and 2 and Catawba Units 1 and 2, respectively. These marked copies show the proposed changes.

" Attachments 7a and 7b provide the reprinted Technical Specifications and Bases pages for McGuire Units 1 and 2 and Catawba Units 1 and 2, respectively.

Duke has determined that the revisions contained in this LAR supplement, as shown in Attachments 6a, 6b, 7a, and 7b have no __

impact on the determination of no significant hazards consideration that was included in Reference 1.

This submittal package contains information that Duke considers proprietary. This information is contained within the proprietary version of the response to the NRC questions on Topical Report DPC-NE-2011-P that is provided as Attachment 3a to this letter. In accordance with IOCFR2.790, Duke requests that this information be withheld from public disclosure. An affidavit that attests to the proprietary nature of this information is included with this letter. A non-proprietary version of this response is also provided as Attachment 3b to this letter.

Inquiries on this matter should be directed to J. S. Warren at (704) 382-4986.

Very truly yours, M. S. Tuckman

U. S. Nuclear Regulatory Commission August 7, 2002 Page 5 xc w/Attachments:

C. P. Patel (Addressee Only)

NRC Senior Project Manager (CNS)

U. S. Nuclear Regulatory Commission Mail Stop 0-8 H12 Washington, DC 20555-0001 R. E. Martin .(Addressee Only)

NRC Senior Project Manager (MNS)

U. S. Nuclear Regulatory Commission Mail Stop 0-8 H12 Washington, DC 20555-0001 L. A. Reyes U. S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 D. J. Roberts Senior Resident Inspector (CNS)

U. S. Nuclear Regulatory Commission Catawba Nuclear Site S. M. Shaeffer Senior Resident Inspector (MNS)

U. S. Nuclear Regulatory Commission McGuire Nuclear Site M. Frye Division of Radiation Protection 3825 Barrett Drive Raleigh, NC 27609-7221 R. Wingard, Director Division of Radioactive Waste Management South Carolina Bureau of Land and Waste Management 2600 Bull Street Columbia, SC 29201

U. S. Nuclear Regulatory Commission August 7, 2002 Page 6

ý-j M. S. Tuckman, affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

M. S. Tuckman, Executive Vice President Subscribed and sworn to me: me Date 7,*s 720O2_

Lt, Notary Public S22 2o1 My commission expires:

SEAL

U. S. Nuclear Regulatory Commission August 7, 2002 Page 7 bxc w/Attachments:

M. T. Cash C. J. Thomas G. D. Gilbert L. E. Nicholson K. L. Crane K. E. Nicholson J. M. Ferguson (2) -CN01SA L. J. Rudy G. A. Copp R. L. Gill P. M. Abraham

-G:-G* Pihl D. R. Koontz R. C. Harvey MNS Master File - MG01DM Catawba Master File - CN04DM NRIA/ELL Catawba Owners:

Saluda River Electric Corporation P. 0. Box 929 Laurens, SC 29360-0929 NC Municipal Power Agency No. 1 P. 0. Box 29513 Raleigh, NC 27626-0513 T. R. Puryear NC Electric Membership Corporation CN03G Piedmont Municipal Power Agency 121 Village Drive Greer, SC 29651

Attachment 4 Responses to Request for Additional Information Topical Report Numbered DPC-NE-1003, Revision 1, McGuire Nuclear Station and Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testings (TAC NOS. MB3343, MB3344, MB3222, MB3223)

General Subsequent to receiving the NRC RAI package, a clarification of Question 4.d. was obtained from the NRC during a conference call on Monday July 15, 2002. Responses to all questions in the NRC RAI are given below, and responses to Question 4.d. takes into account the clarification received from the NRC. Some of the responses require making revisions to the proposed version of this topical report. The revised pages are included at the end of this Attachment.

Question 1. Appendix A of topical report DPC-NE-1003, Revision 1, contains two versions of DPC's rod swap measurement procedures PTIO/A/415011 lA: Attachment 3 (dated June 1986) and Attachment 4 (dated April 1984). There are differences in these two versions of procedures. For example, in the Attachment 3 version, Steps 12.2.2.and 12.2.3, respectively, specify the insertion of bank 1 until the indicated reactivity is approximately -20 pcm, and the withdrawal of reference bank until the indicated reactivity is approximately +20 pcm; whereas in the Attachment 4 version, the insertion and withdrawal of bank 1 and reference bank, respectively, of steps 12.2.1 and 12.2.2 specify reactivity change of -/+ 10 pcm.

a. Since the Attachment 3 version of procedures is more recent, why is the Attachment 4 version referenced in Revision 1 of the topical report (Reference 2)?

b. Which of these two versions of rod swap measurement procedures will be used for McGuire and Catawba Units?

Response 1.a.

Appendix A of the submitted report is labeled "NRC/DPC Correspondence Including DPC Responses to NRC Requests for Additional Information." The information currently in Appendix A contains information provided by DPC in response to the NRC RAI (letter dated 1/12/87) associated with the original submittal of this report. The differences in Attachment 3 and Attachment 4 are due to the timing of the submittals of this topical report, NRC RAI, and DPC responses.

Attachment 3 contains the then most current versions of the procedures for rod swap measurements and were provided in response to Question 2 in the NRC RAI mentioned above. Attachment 4 is an earlier version of the Rod Swap procedure, and this procedure was provided in response to Question 5 of the NRC RAI mentioned above.

The reference list in the proposed version of this topical report was not updated, because the procedure is referenced in a general way and because some of the measured data used to perform the benchmark calculations was processed using the procedure referenced in the original submittal.

A4-1

Attachment 4 Responses to Request for Additional Information Topical Report Numbered DPC-NE-1003, Revision 1, McGuire Nuclear Station and Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testings (TAC NOS. MB3343, MB3344, MB3222, MB3223)

Response 1.b.

Duke currently employs the Westinghouse Dynamic Rod Worth Measurement technique for determining rod worth during ZPPT; however, rod swap may be used as a contingency. The procedure to be used in the event the rod swap test is to be performed now is not the same as those shown in Attachments 3 and 4. An information only version of the current procedure is provided in Attachment 4 (see response to Question 4.c.)

Question 2. In the Attachment 3 version of rod swap measurement procedures PT/O/A14150/1 1A, Step 12.1.3 states that: "Repeat steps 12.2.1 and 12.2.2 until the previously inserted bank fully withdrawn." Is there a typographic error in the words "steps 12.2.1 and 12.2.2"? Should the correct words appear to be "steps 12.1.1 and 12.1.2"?

Response

Yes, this is a typographical error. This error is not in the current Rod Swap procedure.

Question 3. The equation in Section 3, Measurement Procedure, of the topical report for calculating the inferred rod worth of bank x is different from the equation in Step 12.5.3 of the Attachment 3 procedures. The difference appears to be due to the initial height of the reference bank for performing the rod swap measurement of the measured bank.

Clarify the exact procedure to be used in the rod swap test, and make all necessary corrections in the topical report and the procedures to be consistent.

Response

The difference is the initial height of the reference bank for measuring the other banks. In the situation where the reference bank only inserted critical position is 0 SWD, the results of the topical report equation and the procedure equation are the same. If the critical position of the reference bank only inserted is not 0 SWD, it is necessary to account for this small amount of reactivity. This situation may arise as a result of small temperature or boron changes during the test. The proposed topical report has been modified to reflect this, and the revised pages (Pages 2 and 3) are included at the end of this Attachment.

A4-2

Attachment 4 Responses to Request for Additional Information Topical Report Numbered DPC-NE-1003, Revision 1, McGuire Nuclear Station and Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testings (TAC NOS. MB3343, MB3344, MB3222, MB3223)

Ouestion 4. The third sentence in Section 3 of the topical report is revised to read: "All other banks are then exchanged with the reference bank or other test banks at constant boron conditions until the measured bank is fully inserted." It is stated, in Attachment 9a

- Detailed Listing of Changes to DPC-NE-1003A, that the third sentence in Section 3 is revised to make the report consistent with current procedures. The "Revision History" in the topical report states that this revision [Revision 1] also reflects a refinement in the rod swap to make use of two test banks.

a. What is the "current procedures"? What is the date of the current procedures?

b. Are the current procedures the same or different from the one in Attachment 3? The Attachment 3 procedures did not include the exchange of a test bank with other test bank.

c. If the "current procedures" are different from that of Attachment 3 or 4, provide a copy of the procedures, and appropriately reference it in the report.

d. Is the statement in "Revision History" referring to this revision? Please explain what the statement means.

Response 4.a.

The current McGuire procedure is PT/0/A/4150/1 1A, dated 1/19/96.

Response 4.b.

The current procedure is not the same as Attachment 3. The current procedure allows for the exchange of two test banks, namely of the bank to be measured and the bank just This exchange takes place while moving the test bank to be measured into the fullymeasured.

inserted position.

Response 4.c.

An information only copy of the current McGuire procedure is included in Attachment 4 of this response package. The topical report only makes a general reference to the plant procedure.

Response 4.d.

The statement "This revision also reflects a refinement in the rod swap to make use of two test banks." in the Revision History of this topical report does apply to this proposed revision.

The statement refers to the description of intermediate steps of exchanging two test banks after measuring the worth of one test bank and before measuring the worth of the next test bank.

The test bank to be measured is moved into the fully inserted position by exchanging first with the previous test bank and then with the reference bank as necessary. The final test bank/reference bank configuration, and therefore measured worth of the test bank, is the same whether it is exchanged with the reference bank or with the previous test bank. This evolution is shown pictorially on the next page.

Clarification of Appendix A An additional correspondence between DPC and the NRC became known subsequent to the submittal of the proposed version of this topical report. Appendix A of the proposed version of this topical report has been modified to include this additional correspondence.

The pages to be added to Appendix A are provided at the end of Attachment 4 A4-3

Attachment 4 Responses to Request for Additional Information Topical Report Numbered DPC-NE-1003, Revision 1, McGuire Nuclear Station and Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testings (TAC NOS. MB3343, MB3344, MB3222, MB3223)

Rod Swap Rod Exchange with Two Test Banks Step 1 Step 2 Measure Testi Rod Swap Exchange Test1 and Test2 ref test1 test2 ref test1 test2 ARO ARO critical height I1, hx = 90 swd ARI ARI Step 3 Step 4 Exchange Test2 and Reference Measure Test2 by Rod Swap ref test1 test2 ref test1 test2 ARO ARO critical height hx = 150 swd ARI A4-4

Attachment 4 Responses to Request for Additional Information Topical Report Numbered DPC-NE-1003, Revision 1, McGuire Nuclear Station and Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testings (TAC NOS. MB3343, MB3344, MB3222, MB3223)

The following pages of this Attachment contain the marked up and reprinted pages that are revised from the proposed version of this topical report. These pages are being provided in response to Question 3.

2. Definitions The following is a list of the constants needed by the plant, to perform the rod swap procedure. These include:

"* WPx - Predicted reactivity worth of each control and shutdown bank, when inserted individually into an otherwise unrodded core.

" hp. - Predicted critical position of the reference bank after interchange with bank x, starting with the reference bank at 0 steps and bank x fully withdrawn.

"*ax - A correction factor which accounts for the effect of bank x on the partial integral worth of the reference bank, equal to the ratio of the integral worth of the reference bank from hP to the fully withdrawn position with and without x in the core.

In addition, included is a list of constants and their definitions as used in this report.

0 W1x - Measured rod bank worth of bank x from rod exchange o WmRef - Measured rod bank worth of reference bank

(Ap). - The measured integral worth of the reference bank from the measured critical position (h7.) to the fully withdrawn position.

o h7 - The measured critical position of the reference bank after interchange with bank x.

'4ýýI c-,',

1,c- *,hc. c,\\A CL 0 h.Mu-AL)t~c k. 7.. ~i e-L)O%-

~ ILC 2

3. Measurement Procedure With an initial configuration of all rods out, hot zero power, the integral worth of the reference bank is measured using the standard boration/dilution technique. The reference bank is the bank that is predicted to have the highest integral worth. All other banks are then exchanged with the reference bank or other test banks at constant boron conditions until the measured bank is fully inserted.

The worth of each bank is then the amount of reactivity change caused by the withdrawal of the reference bank to its new critical height.

The rod bank worth is inferred from the measured reference bank worth and the measured reference bank height using the following equation:

Wwx = Waoref - ax (Ap)d x

where the above terms are defined in Section 2.0 of this report.

3

2. Definitions The following is a list of the constants needed by the plant, to perform the rod swap procedure. These include:

"*W - Predicted reactivity worth of each control and shutdown bank, when inserted individually into an otherwise unrodded core.

"*h. - Predicted critical position of the reference bank after interchange with bank x, starting with the reference bank at 0 steps and bank x fully withdrawn.

"* x - A correction factor which accounts for the effect of bank x on the partial integral worth of the reference bank, equal to the ratio of the integral worth of the reference bank from hPX to the fully withdrawn position with and without x in the core.

In addition, included is a list of constants and their definitions as used in this report.

WX - Measured rod bank worth of bank x from rod exchange

W'ef - Measured rod bank worth of reference bank

(Ap2 ). - The measured integral worth of the reference bank from the measured critical position (h!') to the fully withdrawn position.

"* h. - The measured critical position of the reference bank after interchange with bank x.

"* (hx). - The initial critical position of the reference bank before exchange with bank x.

"* (Ap1 ) - The measured integral worth of the reference bank from 0 steps to (h1) 0 .

2

3. Measurement Procedure With an initial configuration of all rods out, hot zero power, the integral worth of the reference bank is measured using the standard boration/dilution technique. The reference bank is the bank that is predicted to have the highest integral worth. All other banks are then exchanged with the reference bank or other test banks at constant boron conditions until the measured bank is fully inserted.

The worth of each bank is then the amount of reactivity change caused by the withdrawal of the reference bank to its new critical height.

The rod bank worth is inferred from the measured reference bank worth and the measured reference bank height using the following equation:

W'x = WNref - a. (AP 2 )x - (AP,)

where the above terms are defined in Section 2.0 of this report.

3

'-I

Attachment 4 Responses to Request for Additional Information Topical Report Numbered DPC-NE-1003, Revision 1, McGuire Nuclear Station and Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testings (TAC NOS. MB3343, MB3344, MB3222, MB3223)

The following pages of this Attachment contain the additional DPC correspondence to be included in Appendix A of the proposed version of this topical report.

DUKE PowER ComPANY P.06 BOX 33189 G)IARLOTMZ 79.0. 28242 HAL. I TUCKER TUZLP230wx mmA* V.0m840 (0") 070-4=1 March II, 1987 U.S. Nuclear Regulatory Comission Document Control Desk Washington, D.C. 20555

Subject:

HcGuire Nuclear Station Docket Nos. 50-369/370 Catawba Nuclear Station Docket Nos. 50-413/414 Determination of Rod Worth Using Rod Swap Methodology Gentlemen:

Pursuant to telecons of February 19, 1987 and March 10, 1987 between D.S. Hood t,-J (ONRR) at. al., and S.A. Gevehr (DPC) et. al., attached are revised responses to Questions 6 and 7 of D.S. Hood's request for information dated January 12, 1987.

Very truly yours, Hal B. Tucker SAG/61/j gm Attachment

.0--,

8703180068 870311 PDR ADOCK 05000369 CD C2 TC

Page 1 ATTACHKENT QUESTION 6: Provide data for at least 2 sets of side-by-aide comparisons of boron dilution and Rod Swap Data - predicted and measured. The data may be either for your plants or measured data from another plant and predictions by Duke.

RESPONSE

In the original Nuclear Physics Methodology Topical, DPC-MF-2010A, Duke Power Company benchmarked its methods for predicting rod worths against measurements made during the startup testing for both I4nitial cores at the McGuire Nuclear Station. These measurements were made using the boration/ dilution technique for determining rod worths in sequential insertion. In its review of this topical, the NRC accepted the capability of Duke Power to adequately predict control rod worths and shutdown margin using the outlined methodology.

In the Rod Swap Methodology Report recently sent to the Comission, Duke Power benchmarked its methodology for predicting rod wortha using the rod swap technique against 5 cycles of actual rod swap measurements. This methodology utilized the same computer codes previously benchmarked in DPC-NF-2010A. All predictions. when compared to the measured results, met the acceptance criter ia as outlined in the rod swap plant procedure.

It has been noted in previous conversations with the NRC that the two bench marking studies noted above do not make comparisons of the same units for the same cycles. It is Duke Power's position that there is really no benefit from this type of comparison. A valid comparison cannot be expected since boration/

dilution is a sequential measured worth calculation and rod swap consists of a summation of the vorths of each rod individually inserted into an otherwise unrodded core. It is therefore impossible to make direct comparisons between wortha of the two methods. The only thing that can be looked at is the percent difference between measured and predicted for th two methods. When looking at percent differences between measured and predicted, one does not have to look at the same unit and cycle to verify methodologies are correct.

Comparisons of predicted and measured rod vorths done using boration/ dilution and rod swap on the two Catawba units are enclosed. The boration/dilution technique was used to measure rod worths in sequential bank insertion for the Catawba I Cycle 1 core while Catawba 2 Cycle 1 measurements were done using the rod swap technique (Table 1). From a neutronics standpoint, the two cores are almost identical. This assumption can be justified by examining the core loadings and the results of the Zero Power Physics Testing for each of the units. Several key parameters concerning the core are shown in Table 2. Also enclosed are the quarter core loading pattern (Figure 1) and a comparison of the quarter core assembly power distribution from the zero power map taken during the startup physics testing (Figure 2).

It should also be pointed out that the rod vorths from the rod swap predic tions are not the worths used to calculate the shutdown margin. Rod swap only verifies the code's ability to predict rod vortha. The rod worth used in the shutdown margin calculation is the N-1 worth.

Duke Power has provided a total of nine cycle of predicted rod worth compari sons to measured data with good to excellent results. This demonstrates the ability of the codes and methods used to adequately model reactivity effects due to control rods in any configuration. Therefore, the use of Duke Power predictions in the verification of shutdown margin with appropriate factors of conservatism applied to the calculation as outlined in DPC-NF-2010A Section 4.2.2.2 is justified.

Page 2 AAqn Page 2ATTACHMENT QUESTION 7: What Organization does the safety analysis for the Duke Plants?

When this is not done by Duke. what is done (e.g. tests, coMparisons, etc.) to show that the startup test results adequately represent the plant features and assumptions used in the safety analyses? T__

RESPONSE

The safety analyses for the HcGuire and Catawba Nuclear Stations have been performed by the' current fuel vendor. The analyses utilized NRC-approved codes and methodologies and conservative input assumptions inclujing values for key nuclear physics parameters such as reactivity coefficients, core power distributions, and shutdown margins, which are expected to bound the actual values of these parameters for current and future reload cores. An evaluation is performed for each reload cycle which consists of comparing nuclear design predictions to the safety analyses assumptions to ensure the safety analyses remain bounding. The cycle-specific evaluation process Is described in WCAP-9272, "Westinghouse Reload Safety Evaluation Methodology." Core physics testing performed for each cycle verifies the nuclear design predictions and ensures the actual core physics parameters are conservative with respect to the safety analyses.

The main safety analysis assumption verified by the rod swap procedure is plant will maintain adequate shutdown r"rgin per Technical Specifications. that the the purposes of rod swap measurements and comparisons to predicted One of values is to verify the accuracy of the total rod worth prediction used as an input to the shutdown margin calculation. An independent Duke Power shutdown margin is evaluated for each cycle using methods approved by the NRC in DPC-NF-2010A. The N-1 rod worth used in this prediction is reduced by 10 for conservatism.

Acceptance criteria listed in the procedure indicate that the total inferred rod worth as meazured in the rod swap testing must be within lia of the total predicted worth. If the total measured rod worth is less Than the predicted worth by more than 101, a review of the shutdown margin is made to determine if the current rod insertion limits provide adequate shutdown margin.

margin is adequate. then no revision of the limits is necessary. If the shutdown However, if the margin is not maintained, then Duke will notify Westinghouse, revise the rod insertion limits, and submit any necessary changes to Technical Specifications to the NRC.

In order to tie the rod swap measurements to the verification of inputs to the analysis, Duke Power will perform an independent shutdown margin reload cycle using methods approved by the NRC in DPC-NF-2010A.

Ssafety for each In addition, for each cycle where Duke generates the rod swap prediction but

" the safety analysis has been performed by a vendor, a comparison between the Duke and vendor predicted

. total rod worth will be made at beginning-of-cycle, hot zero power conditions.

Any significant discrepancies will be documented, reviewed, and resolved prior to -_*

startup physics testing.

"Reference HcGuire Nuclear Station. Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testing, DPC-NE-1003, Rev. 1, December 1986.

i )

Q g 41 Q

TABLE 1 Rod Worth Measurement Data Comparison of Rod Swap and Boration/Dilution Techniques Boron/Dilution Rod Swap Integral Worths Integral Worths Predicted Measured Predicted Measured Bank (PCI) (PCM) Z Diff** (PCM) (PCH) Z Diff**

D 772 794 -2.85 773 788 -1.94 C 790 849 -7.47 1214 1203 0.91 B* 852 882 -3.52 1190 1171 1.60 A 249 250 -0.40 572 548 4.20 C,;

SE 377 385 -2.12 508 460 9.45 SD 497 525 -5.63 755 772 -2.25 -:° SC 497 522 -5.03 1098 1099 -0.09 SB 765 834 -9.02 SA 674 706 -4.75 N-1 7370 7414 -. 60 N 5473 5747 -5.01

Reference Bank

- X Diff - [(P-M)/PI*100 4d

÷1 "4° Li Li Li Lila L

TABLE 2

'.-Z Catawba 1 Cycle 1 and Catawba 2 Cycle 1 Comparison of Core Parameters Unit I Unit 2 KG U/ASSY Batch 1 1.6 424.169 424.623 Batch 2 2.4 423.508 425.805 Batch 3 3.1 423.676 424.519 AVE ENR Batch 1 1.6101 1.6104 Batch 2 2.3999 2.4014 Batch 3 3.1022 3.0954 ARO BORON ENDPOINT (PP.fB) 975 975 ISO. TEMP.

COEFF (PCH/OF) -1.745 -1 .81

t )

Figure 1 CATAWBA I CYCLE I AND CATAMBA 2 CYCLE I QUARTER CORE LOADIN6 PATTERN H 6 F E D C B A 4 0 0 I 4 4 B # 1.69

2.46 1.69

2.49 # 1.69 # 2.48 0 1.69

3.19 *

4F 166 12 4 0 16 *

6 9 0 2.41 # 1.669 2.49 a 1.60 # 2.40 # 1.698 3.18

3.19 *

164

12 #

124

290

# # 44 1t 1.69

2.43 4 1.60 # 2.40

1.60 4 2.40

1.69 # 3.10 4 04 12 a. 12 # 16.

6

4 4 4 4 *

4 4 11

2.49 *4 1.69 : 2.48

1.61

2.49 ' 1.690 3.10 1 3.10 4

12 #

6 12 *

16 . ' 16.

*I~o**40*44414**40*43114444*4444*4*4 54 t.**4**4441416****41*4,444 12

1.68 6 2.41 1.61

2.49

2.49 ' 2.49 4 3.10 4

# 12 # 16. 16 13

2.46 1.63

2.48 . 1.60

2.40

3.10

3.19 .

4 16 4 16

16 # 15 4 14

1.638 3.11

1.69 . 3.11

3.16 4 3.160 ENRICHMENT o

21 6

16 # NUMBER OF BA'S PER ASSY 15

3.1 9 3.18 . 3.18a 3.19 .

0F 6 * # 6 G I

Figure 2 CICI AND C2CI HZP POWER DISTRIBUTIONS B HIND/TU ARO. HZP. NO XE, NO SH H a F E D C B A

.72 # .82 # .84

1.098. .89 * .97 a .95 a .97

CJ 9 0 .72 # .82 v .82 a .97 .897 # .94 # .93 0 .95 #

& 9.88 # 0.90 # 2.44

3.09 a 2.30

3.19 0 2.15 & 2.11 *

.82 , .79 * .95 * .86

1.92 # .91 1.11 a 1,01 9 a .92 6 .71 * .95 & .89

1.61 s

' .09 0 1.19

1.09 9.88 I 1.28

9.08 8.609 * .99

2.25 * .91 !.00 1 ,

0 .83 * .96 * .89 a 1.84 # .93

1.91 .93

& * .9g

18 * .93 * .95 0 .99

1.05 & .93 0 1.91 , .93 # .89

0.86 # !.05 8.98 a -. 95 a 0.00 a 8.39 0 4 .91

1.12 a

1.01 a .90

1.06 * .99

1.12 a 1.93

1.16 a .73 m 1.00 4 .89

1.06 *.99 a 1.12

1.93

1.17 * .74 a

1.98

1.12

0.90 a 0.00 1 0.09 8.00 0 * -. 85 * -1.35

.91 # 1.84 a .94

1.12

1.43

1.19 12

1.22 ,

' .91 6 1.04 * .93

1.13 & 1.44 # 1.19

1.23

9.099 # .00

1.90 * -. 88 * -. 69

0.90 4 -.81 S

.98 # .92 I 1.63

1.94 a 1.19

1.22 * .89

13 # 1.91 # .94 a 1.94 1.06 L

1.29

1.21 * .88

a -2.97 1 -2.13 4 -.0 a -!.89 a -. 83 * .03 # 1.14 4 .97 I 1.14 4 .96 # 1.1

1.22 * .88

14 # .98 I 1.16 .97

.

1.19 # 1.22 * .B9

s -1.02 * -1.72 a -I1.3 # -.84 a 9.0. a -1.12 aaaal**a****oaa*aaaa*a*,a***a*alaaa*,*aaa**4************ _

1.81 4 1.94 * .92 # .74

CIC 15 0 1.90

1.94 * .93 * .76 w C2CA 0 9.19 4 8.98 , -1.86 f -2.63 f z DIFF CIC1 CORE AVERAGE 1.90 C2C1 CORE AVERAGE 1.90 I

Z DIFF CORE AVERAGE -. 95 CScl MAXIMUM MAGNITUDE IS 1.43 AT ASSEMBLY D - 12 C2C1 MAXIMUM MAGNITUDE IS 1.44 AT ASSEMBLY D - 12 X DIFF MAXIMUM MAGNITUDE IS 3.19 AT ASSEMBLY C - 8 PERCENT ERROR BETWEEN THE MAXIMUM VALUES IS -. '9" AVERASE ABSOLUTE RELATIVE ERROR .88 PERCENT ROOT MEAN SOUARE OF THE RELATIVE ERROR -4 1.22 PERCENT ROOT MEAN SOUARE OF THE DIFFERENCE 1.16 PERCENT 4

9-

Attachment 4 Responses to Request for Additional Information Topical Report Numbered DPC-NE-1003, Revision 1, McGuire Nuclear Station and Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testings

- (TAC NOS. MB3343, MB3344, MB3222, MB3223)

The following pages of this Attachment contain an information only copy of the current rod swap procedure. This is being provided in response to Question 4.c.

rOrM J41JI thb-t#4)

"'-"uke Power Company '0 No. PT/O/A/4150/ 1A PROdhD:URE PROCESS RECORD "Change(s) 0 to S24 -incorporated PREPARATION (2) Station McGuire NuclearStation %gel\

(3) Procedure Title Control Rod Worth Meosurem l (4) Prepared By

________Date C/I o (5) Requires 10CFR50.59 evaluation?

Yes (New procedure or reissue with major changes)

J-J No (Reissue with minor changes ORFto incorporate previously approved changes)

(6) Reviewed By' Date f -,,l_ ,

Cross-Disciplinary Review By 4 Date (7) Additional Reviews Reviewed By Date -/

Reviewed By Date/

(8) Temporary Approval (if necessary)

By (SRO) Date By A 0 Date (9) Approved By I A I Date '/ -9 PERFORMANCE (compare with control copy every 14 calendar days)

(10) Compared with Control Copy Date Compared with Control Copy Date Compared with Control Copy Date (11) Date(s) Performed ____

Work Order Number (WO#) N COMPLETION (12) Procedure Completion Verification

[] Yes E N/A Check lists and/or blanks p initialed, signed, dated or filled in N/A or N/R, as appropriate?

L] Yes F1 N/A Listed enclosures attached?

0 Yes E] N/A Data sheets attached, completed, dated and signed?

1l Yes El N/A Charts, graphs, etc. attached and properly dated, identified and marked?

LI Yes El N/A Procedure requirements met?

Verified By Date (13) Procedure Completion Approved

-Date (14) Remarks (attachadditionalpages, if necessary)

PT/O/A/4150/lA Changes 0 to 24 incorporated Page 1 of 2 ATTACHMENT TO THE PROCEDURE PROCESS RECORD:

Procedure

Title:

Control Rod Worth Measurement: Rod Swap Changes included in the reissue:

Section 2.0 The following references were added:

- FSAR Section 14.3.2.3

- Technical Specification 3.10.3

- SER for Duke Power Rod Swap Methodology Report for Startup Physics Testing, May 22, 1987 The following Support Documents were added:

- PT/O/AI4150110, Boron Endpoint Measurement Section 3.0 Time requirements changed from six to eight hours Section 4.0 PTIO/AI4150/10, Boron Endpoint Measurement was removed as a prerequisite test.

Section 5.0 Step 5.1 was revised to better define the required reactivity computer.

Step 5.2 was added to define the scale of the 2 pen strip chart recorder for the reactivity computer setup.

Step 5.3 was added to recommend (optional) monitoring of Tave during testing.

Section 6.0 The following Limits and Precautions were added:

-If a stable startup rate of 0.5 DPM is achieved, insert rods to reduce startup rate to less than 0.5 DPM. If the startup rate is greater than or equal to 1.0 DPM, immediately trip the reactor.

- Avoid makeup to the VCT during rod swap evolution.

- Keep reactivity between - 50 pcm and 75 pcm during rod swap.

-Adjustments to procedure are required if any bank (other than Bank 8) is worth more than the reference bank.

Section 8.0 - Step 8.2 was deleted.

- Step 8.3 was changed to specify a pcm limit.

- Step 8.6 was revised to match the Startup Physics Test Program notation of 2235 + 50 psig in addition to providing the corresponding pressure range.

- Step 8.5 was revised to match the Startup Physics Test Program notation of 557 + 2 'F in addition to providing the corresponding temperature range.

The following Prerequisite System Conditions were added to ensure stable test conditions, to ensure NC system boron remains stable and to aid in setup of the reactivity computer:

- Test equipment setup per Section 5.0

- Rod Control System has been checked per Enclosure 13.10 Section 11.0 - Changes all references of Design Engineering to G.O. Nuclear Engineering Section 12.0 - Revised procedure to reflect change #24 throughout (20 to 40 pcm limit).

PT/OIA14150/11A Changes 0 to 24 incorporated Page 2 of 2 Section 12.0 - Added - NOTES:

(cont.) 1. The following steps ensure that no VCT makeup is required during rod swap.

2. To reduce pump head loss, all additions should occur directly to the suction of the NV pumps (through NV-175 or NV-265), NOT to the top of the VCT (through NV-171).

Auto Makeup Limit = 41.4% and Low Level Alarm = 15.7%.

- Added - CAUTION - Ensure that VCT pressure does not exceed 30 psig while performing Step 12.2. This may require batching the additions to allow the VCT pressure I-controller adequate time to operate. Failure to do so may result in misoperation of the boric acid transfer pump

- Added Step 12.2 - Ensure that the VCT level is sufficient such that makeup will not be required for approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

- Added Step 12.4 - Verify that drift in the reactivity trace over that last 30 minutes is less than 5 pcm.

- Added NOTE - Temporary signs will be provided for the OATC to assist in designating rod group being withdrawn and rod group being inserted.

- Added Step 12.16 - Any temporary signs provided for the bATC to assist in designating rod group being withdrawn and rod group being inserted should be removed from the Control Room upon completion of this test.

The procedure was updated to follow the current procedure writers guidelines for NOTES, CAUTIONS, IFs, etc. Additionally, the procedure steps were renumbered as needed.

,-.d

Li PT/0A/4150/1 1A Page I of 12 Li Li "DUKEPOWER COMPANY Li McGUIRE NUCLEAR STATION CONTROL ROD WORTH MEASUREMENT: ROD SWAP Li Li 1.0 PURPOSE Li Li NOTE: The reference bank is the bank which has the predicted highest reactivity worth of all control and shutdown banks when inserted into an otherwise unrodded core.

Li Li 1.1 To determine the worth of all control and shutdown banks, except the reference bank, as inferred from an iso-reactivity interchange with the reference bank.

Li Li 1.2 To verify that the reactivity worth of each control and shutdown bank (except the reference bank), as inferred from data following iso-reactivity interchange with the reference bank, is Li consistent with design predictions.

Li Li

2.0 REFERENCES

Li 2.1 Source Documents:

Li Li 2.1.1 Rod Bank Worth Measurements Utilizing Bank Exchange, WCAP-9863-A, May 1982.

Li Li 2.1.2 Duke Power Company, McGuire Nuclear Station, Catawbd Nuclear Station, Startup Physics Test Program, April 8 1988.

Li Li 2.1.3 Operating Experience Program Commitment 1-91-41-001A.

Li 2.1.4 Significant Event Report 90-15.

Li Li 2.1.5 FSAR Section 14.3.2.3 Li 2.1.6 Technical Specification 3.10.3 Li 2.1.7 NSD 213, "Conduct of Infrequently Performed Tests or Evolutions".

Li Li 2.1.8 SER for Duke Power Rod Swap Methodology Report for Startup Physics Testing, May 22, 1987.

Li Li 2.2 Support Documents:

Li Li 2.2.1 Control Rod Worth Measurement, PTIOIAI4150/11 Li 2.2.2 Post Refueling Controlling Procedure for Criticality, Zero Power Physics Li Test, and Power Escalation Testing, PT/O/A/4150/21 Li Li 2.2.3 MNS Technical Specifications Li Li Li Li Li

PTIOIAJ4150/1I A Page 2 of 12

"*Surveillance Requirement 3.1.1.1

"*Surveillance Requirement 3.10.4

"*Surveillance Requirement 3.10.3

"*Surveillance Requirement 3.10.2.

2.2.4 Duke Power Company, Startup and Operational Report for appropriate unit and cycle.

2.2.5 RODSWAP computer application Users Guide 2.2.6 PTIOIAI4150110, Boron Endpoint Measurement 3.0 TIME REQUIRED

3.1 'Duration

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.2 Personnel.Required: Two Engineers 4.0 PREREQUISITE TESTS None 5.0 TEST EOUIPMENT 5.1 Westinghouse Digital Reactivity Computer or equivalent, (with flux signal from top and bottom of one power range channel).

5.2 One 2 pen strip chart recorder with reactivity (+/-100 pcm scale) and flux signal inputs.

5.3 One strip chart recorder monitoring NC Tave (optional).

PT/01A14150/1 1A Page 3 of 12 6.0 *-'* LIMITS AND PRECAUTIONS 6.1 If a stable startup rate of 0.5 DPM is achieved, insert rods to reduce startup rate to less than 0.5 DPM. If the startup rate is greater than or equal to 1.0 DPM, immediately trip the reactor.

6.2 The NC system temperature is 557°F + 2*F (555°F to 559rF), and controlled preferably by steam dump to the condenser. Temperature may be,controlled by other methods as required by system conditions.

6.3 Normally all reactor coolant pumps should be operating for maximum mixing in the NCS.

U'If all reactor coolant pumps are not operating, the operating pumps should be those on the NCS charging loops (A and/or D). See Tech Spec 3.4.1.1 and 3.10.4 if all reactor coolant pumps are not operating.

LJ 6.4 The rod insertion limit and bank overlap sequence will be violated during this test. The Unit SRO and OATC should be made aware in advance and should anticipate the associated alarms. Technical Specification 3.10.3 allows for this violation.

6.5 Maintain the flux level in the zero power test range established in PT/0/A14150/21.

6.6 If bank has two groups, both must be at the same position prior to switching rod control selector switch between banks to avoid group misalignment.

6.7 If any unexpected, inadvertent drop of an RCCA(s) or Bank(s) of RCCAs occurs, recommend to Unit SRO immediate initiation of manual reactor shutdown.

(OEP Commitment 1-91-41-001A) 6.8 Avoid makeup to VCT during rod swap evolution.

6.9 Keep reactivity between - 50 pem and 75 pcm during rod swap.

6.10 Adjustments to procedure are required if any bank (other than Bank 8) is worth more than the reference bank.

7.0 REQUIRED UNIT STATUS

,Initial

___ Mode 2 with the flux level in the zero power physics test band established in PTIOIA14150/21.

1.-u

PT/OIA1415011 IA Page 4 of 12 8.0 PREREQUISITE SYSTEM CONDITIONS NOTES NOTES 3: 1) The following steps may be signed off in any order.

2) See Enclosure 13.1 for an explanation of nomenclature used in this test.

3) Banks should be measured in order of increasing predicted worth.

4) Step 8.6 may be performed prior to any other Section 8 step.

Initial Initial 8.1 Complete Enclosure 13.2.

8.2 Ensure the reactor is critical with all control and shutdown banks fully withdrawn except the Reference Bank which is < 50 pcm from fully inserted.

8.3 The Rod Control Selector switch is in Bank Select Mode set the Reference Bank.

8.4 Reactor coolant system temperature is 557 + 20F (555 OF to 559 0F).

8.5 Reactor coolant system pressure is 2235 + 50 psig (2185 to 2285).

8.6 Complete Enclosure 13.9.

8.7 IF available, start computer application, RODSWAP, as directed by reference 2.2.5.

8.8 Test equipment is setup per section 5.0 8.9 Rod Control System has been checked per Enclosure 13.10 Sections 7.0 and 8.0 Performed By/Date:

9.0 TEST METHOD The bank with the highest predicted value of reactivity worth has been measured using the dilution technique per PT/O/A/4150/1 1. This bank serves as a reference. The integral worth of the remaining banks is implied from the difference in the critical rod position of the reference bank with and without the insertion of bank being tested. The implied integral worths are then compared to predicted rod worths.

10.0 DATA REQUIRED 10.1 The following conditions for the approximate time of criticality after each bank exchange, recorded on Enclosure 13.3:

+ Time

+ Critical height of reference bank 10.2 Nuclear design predictions on Enclosure 13.2.

PTIO/A14150/1 1A Page 5 of 12

. .4:-.,

10.3 A copy of the rod positions and rod worths for the reference bank from Enclosure 13.2 of PTIO/A14150/1 1.

10.4 The calculated, implied integral worth WI for each RCC bank except the reference bank.

List data on Enclosure 13.3 OR from RODSWAP printout attached to Enclosure 13.6.

10.5 The percent difference between inferred and predicted worths for each individual RCC banks F-1 and for the sum of all banks E2 on Enclosure 13.7 OR on RODSWAP printout attached to Enclosure 13.6.

11.0 ACCEPTANCE CRITERIA NOTES: 1) The appropriate actions for failure of an acceptance or review criteria are as follows:

Acceptance Criteria: G.O. Nuclear Engineering shall:

"*Provide concurrence to continue testing.

"*Investigate and provide solution within 30 days of the test.

"*Submit a report of the findings to the NRC within 45 days of the test.

Review Criteria: G.O. Nuclear Engineering shall:

"*Investigate and provide solution within 60 days of the test.

"*Submit a report of the findings to the NRC within 75 days of the test.

2) For calculating percent differences, use CMeas X 100%

11.1 Acceptance Criteria 11.1.1 The sum of all banks (E2) >90% of predicted.

11.1.2 For all banks other than the reference bank, from Enclosure 13.7 either:

a) (E,), is +30% of predicted for each bank x OR b) (WP' - W I) is + 200 pcm of predicted for each bank x, whichever is greater.

11.1.3 All banks, both control and shutdown banks, are measured.

11.2 Review Criteria:

PTIOIA/4150/11 A Page 6 of 12 11.2.1 From Enclosure 13.7, the sum of all banks (82) is <110%.

11.2.2 For all banks other than the reference bank, from Enclosure 13.7, either:

a) (601 ) is +15% of predicted for each bank x OR b) (W'p - W) is+ 100pcm of predicted foreach bank x, whichever is greater.

PTIO/A/4150/11A Page 7 of 12 12.0 PROCEDURE Initial INOTE: All banks except reference bank are referred to by bank number identified on Enclosure 13.2.

12.1 Attach Enclosure 13.2 of PT/O/A/4150/11 and label as Enclosure 13.8.

NOTES: 1) Step 12.2 ensures that no VCT makeup is required during rod swap.

2) To reduce pump head loss, all additions should occur directly to the suction of the NV pumps (through NV-175 or NV-265), NO._T to the top of the VCT (through NV-171). Auto Makeup Limit = 41.4% and Low Level Alarm = 15.7%.

CAUTION: Ensure that VCT pressure does not exceed 30 psig while performing Step 12.2. This may require batching the additions to allow the VCT pressure controller adequate time to operate. Failure to do so may result in misoperation of the boric acid transfer pump.

12.2 Ensure that the VCT level is sufficient such that makeup will not be required for approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

12.3 Verify that drift in the reactivity trace over that last 30 minutes is less than 5 pcm.

INOTE: The first assigned bank on Enclosure 13.2 is referred to as Bank 1.

12.4 Measure integral worth of first assigned bank of Enclosure 13.2 as follows:

12.4.1 Record initial critical position of reference bank (hM ). on Enclosure 13.3.

CAUTION: 1) When switching from one bank to another, step counters for both groups, for banks with two groups, must be indicating the same step number to avoid rod misstepping.

2) During rod exchange, ensure limits and precautions per Step 6.8 are observed.

SNOTE: Temporary signs will be provided for the OATC to assist in designating rod group being withdrawn and rod group being inserted.

12.4.2 Direct Operations to insert bank 1 until indicated reactivity is approximately

- 40 pcm.

12.4.3 Direct Operations to withdraw reference bank until indicated reactivity is approximately + 40 pcm.

12.4.4 Repeat Steps 12.4.2 and 12.4.3 until bank 1 is fully inserted maintaining indicated reactivity at approximately +/- 40 pcm.

12.4.5 Direct Operations to adjust position of reference bank until reactor is critical.

PTIOIA14150/I A Page 8 of 12 12.4.6 Record final critical configuration data (hr) on Enclosure 13.3.

12.5 Measure integral worth of remaining assigned banks as follows:

NOTES: 1) The bank being measured is denoted as bank N.

2) The previously measured bank is denoted as bank N-I.

3) N =2 for the second assigned bank.

12.5.1 Direct Operations to insert bank N until indicated reactivity is approximately - 40 pcm.

12.5.2 Direct Operations to withdraw bank N-I until indicated reactivity is approximately + 40 pcm.

12.5.3 Repeat Steps 12.5.1 and 12.5.2 until bank N is fully inserted or bank N-I is fully withdrawn.

12.5.4 IF bank N is fully inserted before bank N-I is fully withdrawn, direct Operations to insert reference bank, compensating with withdrawal of bank N-I, maintaining indicated reactivity approximately +/- 40 pcm throughout, until critical conditions are achieved with bank N-I fully withdrawn.

12.5.5 IF bank N-I is fully withdrawn before bank N is fully inserted, direct Operations to withdraw reference bank, compensating with insertion of bank N, maintaining indicated reactivity approximately +/- 40 pcm throughout, until critical conditions are achieved with bank N fully inserted or reference bank is fully withdrawn.

12.5.6 IF bank N is not fully inserted and reference bank is fully withdrawn, mark Steps 12.5.7 and 12.5.8 N/A AND measure bank after others.

12.5.7 Adjust position of reference bank until reactor is critical.

12.5.8 Record final critical configuration data (hr) on Enclosure 13.3.

12.5.9 Repeat Steps 12.5.1 through 12.5.8 using Enclosure 13.4 for step signoffs to measure integral worths of assigned bank N = 3 through 7.

12.6 Measure integral worth of bank 8 as follows:

12.6.1 Direct Operations to insert bank 8 until indicated reactivity is approximately

- 40 pcm.

12.6.2 Direct Operations to withdraw bank 7 until indicated reactivity is approximately + 40 pcm.

12.6.3 Repeat Steps 12.6.1 and 12.6.2 until bank 8 is fully inserted fully withdrawn. OR bank 7 is I-

PTIOIAI4150/1AA Page 9of 12 12.6.4 IF bank 8 is fully inserted before bank 7 is fully withdrawn, insert reference bank, compensating with withdrawal of bank 7, maintaining indicated reactivity between +/- 40 pcm throughout until critical conditions.

12.6.3 IF bank 7 is fully withdrawn before bank 8 is fully inserted, withdraw reference bank compensating with insertion of bank 8 maintaining indicated reactivity between +/- 40 pcm throughout, until critical conditions are achieved with bank 8 fully inserted OR reference bank is fully withdrawn.

.12.6.6 IF bank 8 is fully inserted with reference bank not fully withdrawn, direct Operations to adjust reference bank position to critical and record hm on Enclosure 13.3.

12.6.7 IF bank 8 is NOT fully inserted and reference bank is fully withdrawn, perform the following:

12.6.7.1 IF remaining worth of bank 8 to be inserted is estimated to be less than approximately 50 pcm, measure remaining worth by inserting bank 8 to 0 steps and measure worth using reactivity computer. Record worth on Enclosure 13.3 in column for X.(AP2)x.

12.6.7.2 IF remaining worth of bank 8 to be inserted is estimated to be greater than approximately 50 pcm, perform the following:

a) Swap bank 8 for reference bank until bank 8 is fully withdrawn.

b) Record reference bank inserted, final critical'point (hm) final on Enclosure 13.5 and Enclosure 13.3 for bank 7.

c) On Enclosure 13.5, mark bank 8 drift as N/A and divide drift by 7 to get drift/bank.

d) Swap bank 8 for reference bank until reference bank is fully withdrawn.

INOTE: It is permissible to insert another bank to maintain the reactor critical.

e) Direct Operations to commence a slow NC system dilution and measure remaining worth of bank 8 using reactivity computer.

12.7 Direct Operations to insert reference bank until indicated reactivity is approximately - 40 pcm.

12.8 Direct Operations to withdraw bank 8 until indicated reactivity is approximately + 40 pcm.

PTI/O/A4150/11A Page 10 of 12 12.9 Repeat Steps 12.7 and 12.8 maintaining indicated reactivity approximately +/-40 pcm, until bank 8 is fully withdrawn and critical conditions are achieved.

12.10 IF Step 12.6.7.2 was NOT performed, perform the following:

Record (hM)o on Enclosures 13.3 and 13.5. .J

"*Divide through by 8 on Step 13.5.6 of Enclosure 13.5.

12.11 Complete Enclosure 13.5.

1 NOTE: If computer application, RODSWAP is used, Step 12.12 and any unused blanks on Enclosures 13.3, 13.5, and 12.12 Compute inferred worth for each control and shutdown bank (except reference bank) as follows:

12.12.1 Using data from Enclosure 13.3, and worth measurement data for reference bank from Enclosure 13.8, record value of (Ap,). on Enclosure 13.3.

12.12.2 IF bank being measured has a worth greater than reference bank worth, replace CC, (Ap 2 ). with worth measured by reactivity computer:

[wM 12.12.3 Using data from Enclosure 13.3, worth measurement data for reference bank from Enclosure 13.8 and data of Enclosure 13.2, compute value of a (Ap2), as described below and record on Enclosure 13.3:

FW M"1 where: [wMFh is the measured integral worth of the reference bank from hm to the fully withdrawn position from Enclosure 13.8.

Linearly interpolate if hm does not correspond to the steps on Enclosure 13.8.

Xhm is the measured critical position of the reference bank after interchange with bank x from Enclosure 13.3.

and CCx is a correction factor from Enclosure 13.2 to account for the influence of bank x on the worth of the reference bank.

PT/0/A/4150/11 A Page 11 of 12 12.12.4 IF bank being measured has a worth greater than the reference bank worth, compute the inferred integral worth of the bank and record on Enclosure 13.3:

w, =-wM +-W M FW (AM",)

where [w, w , is given inthe column marked

x. (Ap 2 ). on Enclosure 13.3.

- 12.12.5 Compute inferred integral worth of each bank x, W1, as described below and record on Enclosure 13.3:

Wx -(4R ).-(Ap2).

where: WR is the measured total integral reference bank worth from Enclosure 13.8.

(Ap,), is from step 12.12.1.

and Cc (Ap2 ), is from step 12.12.2 or 12.12.3.

12.12.6 Compute difference and percent difference between inferred and predicted worths for each individual RCC bank and the sum of all banks described below.

(F-1). =týwp 1- 1 N

F,2 x 100%

21'1 Fill in all blanks and summarize the calculations on Enclosure 13.6.

12.13 IF computer application, RODSWAP, is used, attach printout to Enclosure 13.6.

12.14 Complete Enclosure 13.7.

PTIO/A14150/11A Page 12 of 12 12.15 Verify all acceptance and review criteria have been met, or appropriate actions are being taken.

12.16 Any temporary signs provided for the OATC to assist in designating rod group being withdrawn and rod group being inserted should be removed from the Control Room upon completion of this test.

13.0 ENCLOSURES 13.1 Nomenclature 13.2 Nuclear Design Predictions for Rod Exchange Measurements 13.3 Critical Configuration and Worth Calculation Sheet 13.4 Additional Signoffs for Banks 3 through 7 13.5 Reference Bank Drift Evaluation 13.6 Comparison of Inferred Bank Worths with Design Predictions 13.7 Review Criteria Evaluation 13.8 Reference Bank Integral Worth 13.9 Requirements for.Infrequently Performed Tests 13.10 Rod Control Cabinet Group Select Light Checkout

PT/O1A/4150/1 IA Page 1 of 1 ENCLOSURE 13.1 NOMENCLATURE

1. Predicted reactivity worth of each control and shutdown bank when inserted individually into an otherwise unrodded core.

2. W1 The calculated, implied rod bank worths of bank x from rod exchange.

3. W;M Measured rod bank worth of reference bank.

4. (X. A correction factor which accounts for the effect of bank x on the partial integral worth of the reference bank, equal to the ratio of the integral worth of the reference bank from hp to the fully withdrawn position with and without x in the core.

5. (Ap 2 ). The measured integral worth of the reference bank from hm to the fully withdrawn position.

6. hp The predicted critical position of the reference bank after interchange with bank x starting with reference bank at 0, bank x fully withdrawn.

7. hm The measured critical position of the reference bank after interchange with bank x.

-(h,),

8.M The measured integral worth of the reference bank from 0 steps to (hM ).; equivalent to (Ap )X.

9. (hM)o Initial critical position of the reference bank before interchange with bank x.

10.

L oh

[FWV The measured integral worth of the reference bank from hm to the fully withdrawn position.

PTIO/A14150/1 IA Page I of I ENCLOSURE 13.2 NUCLEAR DESIGN PREDICTIONS FOR ROD EXCHANGE MEASUREMENTS McGuire Unit Cycle (b) (c)

Bank Bank hp a No.

o Identityx (pcm)

(x) (steps)

+

(a)

Reference N/A N/A I

2 3

4 5

6 7

8 (a) Reference bank - the bank with the highest predicted integral worth.

(b) Reference bank critical position after interchange with bank x.

(c) Ratio of integral worth of the reference bank from hp to the fully withdrawn position with and without bank x in the core.

+ Control Bank C, Shutdown Bank E, etc.

NOTE: See Enclosure 13.1 for a complete listing of nomenclature used in this test.

Recorded By Date This data came from (list source and document number):

PTIO/A/415011 1A Page I of I McGuire Unit _ Cycle ENCLOSURE 13.3 CRITICAL CONFIGURATION AND WORTH CALCULATION SHEET Bank (x) Date/Time (0). (h) (Ap 1 ,x a, (AP 2 )x WI No. Ident. N/A (steps) (steps) (pcm) (pcm) (pcm) 1 2 N/A 3 N/A 4 N/A 5 N/A 6 N/A 7

8

NOTE: IF bank being measured has a worth greater than the reference bank worth, these values will be as given by Enclosure 13.5 or Step 12.4.7.2.

Recorded By Date_

PT/0/A/415011 IA Page Ioof 1 C-,, *-. z<

ENCLOSURE 13.4 ADDITIONAL SIGNOFFS FOR BANKS 3 THROUGH 7 Bank 3 4 5 6 7 Stgn 12.5.1 12.5.2 12.5.3 12.5.4 12.5.5 12.5.6 12.5.7 12.5.8 Section 12.5 Performed By/Date: _______________________________

PT/0/A/4150/11A Page 1 of 1 ENCLOSURE 13.5 REFERENCE BANK DRIFT EVALUATION McGuire Umit - Cycle Step 13.5.1 Final Reference Bank Critical Position steps 13.5.2 Initial Reference Bank Critical Position steps 13.5.3 Reactivity worth of reference bank from 0 to position of Step 13.5.1 pcm 13.5.4 Reactivity worth of reference bank from 0 to position of Step 13.5.2. pcm 13.5.5 Difference of Step 13.5.3 and 135.4 (Circle correct sign) 13.5.3 - 13.5.4 = - - -_ = - pcm NOTE: Round Step 13.5.6 to the nearest pcm. I I I

13.5.6 Incremental drift for each bank (Circle correct sign and circle either 8 or 7 as appropriate)

(See Step 12.4.7.2.c)

Step 13.5.5/8 or7 = /8or7 ÷ pcm 13.5.7 (pl)x for banks:

bank 1 Step 13.5A pcm bank 2 Step 13.5.4 + 13.5.6 pcm bank 3 bank 2 + 13.5.6 pcm bank 4 bank 3 + 13.5.6 pcm bank 5 bank 4 + 13.5.6 pcm bank 6 bank 5 + 13.5.6 pcm bank7 bank 6 + 13.5.6 pcm bank 8 bank 7 + 13.5.6 pcm Recorded By Date Checked By Date

PT/O/A/4150/I IA Page 1 of I ENCLOSURE 13.6 COMPARISON OF INFERRED BANK WORTHS "WITH DESIGN PREDICTIONS McGuire Unit Cycle NOTE: Round rod worth numbers to the nearest pcm.

++

Bank (x)

No. Ident. Wxp W., (W.P - W.1) (6t)x (pcm) (pcm) (pcm) (%)

+

Reference 2

3 4

N 5

N 6

7 8

j W,(pcm) ZW. (pcm) 82 (%)

+Record the measured worth of the reference bank here.

from Enclosure 13.2

++from Enclosure 13.3 Recorded By Date_

Checked By .Date

PTr/0/A14150/11 A Page 1 of I ENCLOSURE 13.7 REVIEW CRITERIA EVALUATION 2'

McGuire Unit - Cycle NOTE: IF any of the below Review Criteria are checked "No", notify G.O. Nuclear Design by the next working day.

Yes No (4) (4)

Review Criteria 11.2.1: sum of all banks (e 2 ) from Enclosure 13.6 is <110%.

II. Review Criteria 11.2.2: for each bank x (S1 ),x from Enclosure 13.6 is +15% or (Wp -W') from Enclosure 13.6 is +100 pcm, whichever is greater.

Bank x No. Ident.

1 2

3 4

5 6

7 8

Recorded by Date Checked by Date Dt

PTIO/A1415011 IA Page Iof I ENCLOSURE 13.9 REQUIREMENTS FOR INFREQUENTLY PERFORMED TESTS This test, which involves exchanging (swapping) a bank with either the Reference Bank and/or the previous bank to measure its reactivity worth, involves additional requirements and management involvement since it is an infrequently performed test. The guidance in this enclosure establishes an environment that places a high priority on preserving the plant's nuclear safety which is management's prime responsibility.

The Management Designee's responsibility is to ensure management expectations are met and that the evolution is controlled appropriately. The Management Designee can stop the evolution at any point that is deemed necessary or appropriate and provide the Operations Shift Supervisors with guidance for any recovery actions.

The Evolution Coordinator's responsibility is overall coordination of the evolution to ensure it is done in a safe controlled manner. The Evolution Coordinator can stop the evolution at any point that is deemed necessary or appropriate and provide the Operations Shift Supervisor with guidance for any recovery actions. (Reference SOER 91 01)

The Management Designee shall initial and date the steps below when completed.

1.0 Record the following:

Evolution Coordinator Management Designee 2.0 A pre-job briefing has been performed by the Management Designee.

PT/0/AJ4150111A Page lof 2 ENCLOSURE 13.10 i'ROD .. CONTROL CABINET Liw GROUP SELECT LIGHT CHECKOUT Li NOTE: Shutdown and control banks may be done in any order.

V)

L._

1.0 SHUTDOWN BANK A (SDA)

L. 1.1 Have OATC select SDA on "CRD BANK SELECT" 1.2 Verify that only "GRP SELECT" light "C" is illuminated on Rod Control Power Cabinets LiJ IAC and 2AC.

2.0 SHUTDOWN BANK B (SDB) 2.1 Have OATC to select SDB on "CRD BANK SELECT' Verify that only "GRP SELECT"' light "C' is illuminatedS2.2 bn Rod Control Power Cabinets IBD and 2BD.

_ 3.0 SHUTDOWN BANK C (SDC) 3.1 Have OATC to select SDC on "CRD BANK SELECT'

"-"3.2 Verify that only "GRP SELECT' light "A" is illuminated on Rod Control Power Cabinet SCDE.

Li Li 4.0 SHUTDOWN BANK D (SDD)

Li 4.1 Have OATC to select SDD on "CRD BANK SELECT" 4.2 Verify that only "GRP SELECT' light "B" is illuminated on Rod Control Power Cabinet Liw SCDE.

_ 5.0 SHUTDOWN BANK E (SDE) 5.1 Have OATC to select SDE on "CRD BANK SELECT' 5.2 Verify that only "GRP SELECT"' light "C'" is illuminated on Rod Control Power Cabinet SCDE.

Liw 6.0 CONTROL BANK A (CBA)

L6.1 Have OATC to select CBA on "CRD BANK SELECT' 6.2 Verify that only "GRP SELECT' light "A" is illuminated on Rod Control Power Cabinets Liw 1AC and 2AC.

7.0 CONTROL BANK B (CBB)

L..i

, , 7.1 Have OATC to select CBB on "'CRD BANK SELECT'

PT/O/A/4150/11A Page 2 of 2 ENCLOSURE 13.10 ROD CONTROL CABINET

- GROUP SELECT LIGHT CHECKOUT 7.2 Verify that only "GRP SELECT' light "A" is illuminated on Rod Control Power Cabinets 1BD and 2BD.

8.0 CONTROL BANK C (CBC) 8.1 Have OATC to select CBC on "CRD BANK SELECT' 8.2 Verify that only "GRP SELECT' light "B" is illuminated on Rod Control Power Cabinets IAC and 2AC.

9.0 CONTROL BANK D (CBD) 9.1 Have OATC to select CBD on "CRD BANK SELECT' 9.2 Verify that only "GRP SELECT' light "B" is illuminated on Rod Control Power Cabinet 1BD and 2BD.

10.0 IF any expected response is not received, contact Work Control Shift Work Manager to have E Work Order generated for troubleshoot/repair.

Performed By Date Verified By Date

APPENDIX B Original Issue NRC SER

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D C. 20555 k..:-,w

~MAY 22 1987 Docket Nos.: 50-369, 50-370 50-413, 50-414 Mr. H. B. Tucker, Vice President Nuclear Production Department Duke Power Company 422 South Church Street Charlotte, North Carolina 28242

Dear Mr. Tucker:

Subject:

ROD SWAP METHODOLOGY'REPORT FOR STARTUP PHYSICS TESTING, MCGUIRE AND CATAWBA NUCLEAR STATIONS, UNITS 1 AND 2 (TACs 62981, 62982, 62983, 62984)

By letter dated December 4, 1986, you submitted a report titled "Rod Swap Methodology Report for Startup Physics Testing," and you submitted additional information by letters dated February 11 and March 11, 1987. In addition, telephone discussions were held on May 1, 1987 with members of your company regarding conditions associated with our approval.

We have reviewed'the material submitted and find the iod swap methodology as described to be acceptable for rod worth measurement of reloaded cores for McGuire and Catawba Stations, Units 1 and 2. This approval recognizes your prior agreement to certain conditions listed in our enclosed Safety Evaluation Report.

Should you have any questions regarding the enclosure, contact me at (301) 492-8661 or K. Jabbour at (301) 492-7367. In any future correspondence regarding this approval, please include a reference to TACs 62981, 62982, 62983 and 62984.

I .

Sincerely, Darl Hood, Project Manager Project Directorate 11-3 Division of Reactor Projects I/Il

Enclosure:

Safety Evaluation Report cc w/encl See next page

. %UNITED STATES S-.NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 Enclosure SAFETY EVALUATION REPORT FOR DUKE POWER COMPANY'S "ROD SWAP METHODOLOGY REPORT FOR STARTUP PHYSICS TESTING" Introduction Duke Power Company (the licensee) submitted a report titled "Rod Swap Methodology

.Report for Startup Physics Testing" on December 4, 1986. Answers to NRC questions and additional information were submitted by letters dated February 11, 1987 (Ref. 2) and March 11, 1987 (Ref. 3). The report describes the rod swap methodology which Duke Power Company would like to use for rod worth measurement for the McGuire I and 2 and the Catawba I and 2 units after each reload. While the rod swap technique has been used on Duke plants in the past, the methodoloay was the Westinghouse methodology which NRC approved on "May 28, 1983. Due to the complexities of Rod Swap, the May 28, 1983 approval stated that the method was approved for use by Westinghouse only. Thus, it is necessary for Duke to obtain NRC approval before using the Duke Rod Swap methodology.

Background

The reactivity worth of the control rods is measured at the beginning of each cycle. Rod worth measurements are made in order to verify shutdown margin.

The measurement conditions are not those used in the accident analysis but comparison of measurement and predicted rod worths for a known set of conditions gives assurance that rod worths and the shutdown margin predicted for the worst conditions are accurate. For reload cores, usually, not all rod banks are measured. Normally, the control banks (approximately 4 banks, worth about half the total worth) are measured.

The traditional method of rod worth measurement is by boron dilution. Starting from an all rods out critical configuration, the bank is inserted a few steps at a time and the reactor is kept critical by diluting the boron concentration.

One control bank would be inserted until it Is all the way in and then the next bank would be started. A reactivity computer is also used to measure the reactivity change at each position. The reactivity worth of the baink is the sum of all the reactivity changes recorded by the reactivitycomputer. The worth of the bank is also equal to the difference in boron concentrations from the bank fully withdrawn to fully inserted positions.

Several years ago an alternative method of rod worth measurement called rod swap or rod exchange was proposed. In this method the highest worth bank, called the reference bank, is measured by boron dilution and remaining banks, called test banks, are measured by "swapping" the test bank with the reference bank. The critical position of each measurement is the reference bank position when the test bank is fully inserted. This method is an indirect method in that it does not measure the worth of banks in combination (i.e. banks D + C + B + A). Rod Swap does have some advantages over boron dilution, however. It does not require the large change in boron concentration and

subsequent processing of thousands of gallons of water. It is less time consuming and thus all banks can be measured in much less time than it would k.-' take to measure one half the banks by boron dilution.

Evaluation The Duke Report presents a minimal description of the methodology and a comparison of calculated and inferred worths for several cycles on McGulre I

- and 2. Additional information supplied more details of the procedure. The Duke methodology is very similar to the methodology NRC approved for use by Westinghouse. Duke will use previously approved physics codes and S methodologies as described in Reference 4 for the calculations of rod worths and critical heights.

s-_- As verification of the methodology, Duke supplied rod swap data for 5 cycles Unit 1, Cycles 2, 3 and and 4, McGuire Unit 2, Cycles 2 and 3). This data compares measured and predicted worth for each bank. In addition we have made s_- comparisons of this data with that presented in the Startup Reports for these (This data is different since Westinghouse did the calculations for these cycles). Examination of the data reveals that the greatest deviation on S- any one bank was 103 pcm or 24% on a small bank. The greatest deviation on-the S, total worth was 6.9Z for Unit. 2, Cycle 2. The average total difference was 4.94% which compares favorably with the 6.38% for the Westinghouse predictions.

While for some of the McGuire data the difference between measurement and prediction is greater than usually seen, it is still within the acceptable L_. range. Duke did not perform a side-by-side comparison of boron dilution and S, rod swap for the same cycle. However, Duke suDplled data from the initial startup of Catawba 1 and 2, Catawba I using boron dilution and Catawba 2 using L-. Rod Swap. The cores are essentially identical as confirmed by as built S, parameters and other physics test measurements. The rod worth measurements were within acceptable limits.

- Based on our review of the material submitted, we find the rod swap methodology as oroposed by Duke Power Company to be acceptable subject to the following conditions, to which Duke Power Company has agreed:

1) The boron dilution rate for measurement of the reference bank shall not exceed 500 pcm.

2) All banks, both control and shutdown banks, must be measured.

L_. 3) The review criteria are:

A. The absolute-vqlue of the percent differerce between measured and predicted integral worth for the reference bank is < 10 percent.

B. For all banks other than the reference bank, either (whichever is greater);

_ 1) the absolute value of the percent difference between inferred and predicted integral worths is -C 15 percent or

2) the absolute value of the reactivity difference between inferred and predicted integral worths is ( 100 pcm.*

C. The sum of the measured/inferred worth of all the rods must be

_.110 percent of the predicted worth.

4) The acceptance criteria are:

(1) The sum of the measured/inferred worth of all the rods must be

> 90 percent of the predicted rod worth.

(2) For all banks other than the reference bank, either (whichever is greater) a) the absolute value of the percent difference between inferred and predicted integral worth is < 30 percent bý the absolute value of the reactivity difference betweenor inferred and predicted integral worths is < 200 pcm.

(3) The absolute value of the percent difference between measured and predicted integral worth for the reference bank is <15 percent.

5) Additional testing is required if the refe-rence bank boron concentrations and reactivity computer worth do not agree.

action for failure of an acceptance or review criterior requireRemedial investigation and solution within 30 days (for acceptance or 60 days (for review criterion). The licensee must then criterion) report of the findings to the NRC within 45 days of the testsubmit a (for acceptance criterion) or within 75 days of the test (for review criterion).

A pcm is equal to 10-5

,k/k.

-4 REFERENCES

1) Letter, H. B. Tucker, Duke Power Company, to Harold R. Denton, NRC, December 4, 1986.

2) Letter, H. B. Tucker, Duke Power Company, to Nuclear Regulatory Commission, Document Control Desk, dated February 11, 1987.

3) Letter, H. B. Tucker, Duke Power Company, to Nuclear Regulatory Commission, Document Control Desk, dated March 11, 1987.

4) Duke Power Company, "Nuclear Physics Methodology for Reload Design,"

DPC-NF-2010A, June 1985.

Mr. 1t. B. Tucker Duke Power Company Catawba Nuclear Station cc:

A.V. Carr, Esq. North Carolina Electric Membership Duke Power Company Corp.

422 South Church Street 3400 Sumner Boulevard Charlotte, North Carolina 28242 P.O. Box 27306 Raleigh, North Carolina 27611 J. Michael McGarry, III, Esq.

Bishop, Liberman, Cook, Purcell Saluda River Electric Cooperative, and Reynolds Inc.

1200 Seventeenth Street, N.W. P.O. Box 929 Washington, D. C. 20036 Laurens, South Carolina 29360 North Carolina MPA-1 Senior Resident Inspector Suite 600 Route 2, Box 179N 3100 Smoketree Ct. York, South Carolina 29745 P.O. Box 29513 Raleigh, North Carolina 27626-0513 Regional Administrator, Region II U.S. Nuclear Regulatory Commission, L.L. Williams 101 Marietta Street, NW, Suite 2900 Area Manager, Mid-South Area Atlanta, Georgia 30323 ESSD Projects "Westinghouse Electric Corp.

MNC West Tower - Bay 239 P.O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. Heyward G. Shealy, Chief Bureau of Radiological Health South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 Ccunty Manager of York County York County Courthouse Karen E. Long York South Carolina 29745 Assistant Attorney General N.C. Department of Justice Richard P. Wilson, Esq. P.O. Box 629 Assistant Attorney General Raleigh, North Carolina 27602 S.C. Attorney General's Office P.O. Box 11549 Spence Perry, Esquire Columbia, South Carolina 29211 General Counsel Federal Emergency Manl agement Agency Piedmont Municipal Power Agency Room 840 100 Memorial Drive 500 C Street Greer, South Carolina 29651 Washington, D. C. 20'472 Mr. Michael Hirsch Federal Emergency Management Agency Office of the General Counsel Room 840 500 C Street, S.W.

Washington, D. C. 20472 Brian P. Cassidy, Regional Counsel Federal Emergency Management Agency, Region I J. W. McCormach POCH Boston, Massachusetts 02309

Mr. H. B. Tucker Duke Power Company McGuire Nuclear Station cc:

Mr. A.V. Carr, Esq. Dr. John M. Barry Duke Power Company Department of Environmental Health P. 0. Box 33189 Mecklenburg County 422 South Church Street 1200 Blythe Boulevard Charlotte, North Carolina 28242 Charlotte, North Carolina 28203 County Manager of Mecklenburg County 720 East Fourth Street Charlotte, North Carolina 28202 Chairman, North Carolina Utilities Commission Mr. Robert Gill Dobbs Building Duke Power Company 430 North Salisbury Street Nuclear Production Department Raleigh, North Carolina 27602 P. 0. Box 33189 Charlotte, North Carolina 28242 Mr. Dayne H. Brown, Chief Radiation Protection Branch J. Michael McGarry, III, Esq. Divisior of Facility Services Bishop, Liberman, Cook, Purcell Department of Human Resources and Reynolds 701 Barbour Drive 1200 Seventeenth Street, N.W. Raleigh, North Carolina 27603-2008 Washington, D. C.. 20036 Senior Resident Inspector c/o U.S. Nuclear Regulatory Commission Route 4, Box 529 Huntervilie, North Carolina 28078 Regional Administrator, Region II U.S. Nuclear Regulatory Commission, 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 L. L. Williams Area Manager, Mid-South Area ESD Projects Westinghouse Electric Corporation MNC West Tower - Bay 239 P. 0. Box 355 Pittsburgh, Pennsylvania 15230

APPENDIX C Revision 1 NRC SER

"11ý UNITED STATES

- NUCLEAR REGULATORY COMMISSION 0 WASHINGTON, D.C. 20555-0001 October 1, 2002 Ck' Mr. H. B. Barron Vice President, McGuire Site Duke Energy Corporation 12700 Hagers Ferry Road Huntersville, NC 28078-8985

SUBJECT:

McGUIRE NUCLEAR STATION, UNITS 1AND 2 RE: ISSUANCE OF AMENDMENTS (TAC NOS. MB3222 AND MB3223)

Dear Mr. Barron:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 208 to Facility Operating License NPF-9 and Amendment No. 189 to Facility Operating License NPF-17 for the McGuire Nuclear Station, Units 1 and 2. The amendments consist of changes to the Technical Specifications in response to your application dated October 7, 2001, as supplemented by letter dated August 7, 2002.

The amendments revise TS 5.6.5.a by adding a few parameter limits currently included in the Core Operating Limits Report. In addition to the license amendment request, you also submitted revisions to four previously approved topical reports for the Nuclear Regulatory Commission staff review and approval. The enclosed Safety Evaluation also addresses these topical reports.

A Notice of Issuance will be included in the Commission's biweekly FederalRegisternotice.

Sincerely, Robert E. Martin, Senior Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-369 and 50-370

Enclosures:

1. Amendment No. 208 to NPF-9

2. Amendment No. 189 to NPF-17

3. Safety Evaluation cc w/encls: See next page

McGuire Nuclear Station cc:

Ms. Lisa F. Vaughn Ms. Karen E. Long Legal Department (PBO5E) Assistant Attorney General Duke Energy Corporation North Carolina Department of 422 South Church Street Justice Charlotte, North Carolina 28201-1006 P. O. Box 629 Raleigh, North Carolina 27602 County Manager of Mecklenburg County Mr. C. Jeffrey Thomas 720 East Fourth Street Manager - Nuclear Regulatory Charlotte, North Carolina 28202 Licensing Duke Energy Corporation Michael T. Cash 526 South Church Street Regulatory Compliance Manager Charlotte, North Carolina 28201-1006 Duke Energy Corporation McGuire Nuclear Site Elaine Wathen, Lead REP Planner 12700 Hagers Ferry Road Division of Emergency Management Huntersville, North Carolina 28078 116 West Jones Street Raleigh, North Carolina 27603-1335 Anne Cottingham, Esquire Winston and Strawn Mr. Richard M. Fry, Director 1400 L Street, NW. Division of Radiation Protection Washington, DC 20005 North Carolina Department of Environment, Health and Natural Senior Resident Inspector Resources c/o U.S. Nuclear Regulatory Commission 3825 Barrett Drive 12700 Hagers Ferry Road Raleigh, North Carolina 27609-7721 Huntersville, North Carolina 28078 Mr. T. Richard Puryear Dr. John M. Barry Owners Group (NCEMC)

Mecklenburg County Duke Energy Corporation Department of Environmental 4800 Concord Road Protection York, South Carolina 29745 700 N. Tryon Street Charlotte, North Carolina 28202 Mr. Peter R. Harden, IV VP-Customer Relations and Sales Westinghouse Electric Company 6000 Fairview Road 12th Floor Charlotte, North Carolina 28210

"0 UNITED STATES NUCLEAR REGULATORY COMMISSION 0 WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 208 TO FACILITY OPERATING LICENSE NPF-9 AND AMENDMENT NO. 189 TO FACILITY OPERATING LICENSE NPF-17 DUKE ENERGY CORPORATION MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370

1.0 INTRODUCTION

By letter dated October 7, 2001, as supplemented by letter dated August 7, 2002, Duke Power Company, et al. (DPC, the licensee), submitted a request for changes to the McGuire Nuclear Station, Units 1 and 2, Technical Specifications (TS).

Revisions were proposed for TS 5.6.5.a, Item 1, to add the moderator temperature coefficient (MTC) 60 parts per million (ppm) surveillance limit. The specific value of the surveillance limit was previously relocated to the Core Operating Limits Report (COLR). A new item 12, "31 EFPD surveillance penalty factors for Specifications 3.2.1 and 3.2.2," is also proposed to be added to TS 5.6.5.a.

The initial submittal, dated October 7, 2001, proposed to change the dates and revision numbers for three of the Nuclear Regulatory Commission (NRC) approved analytical methods previously listed in TS 5.6.5.b, as listed below. The changes would reflect later versions of these topical reports that were also submitted with the October 7, 2001, submittal for NRC review and approval. As required by TS 5.6.5.b, only those methods listed within the TS as having been reviewed and approved by the NRC, can be used to determine the subject core operating limits. The subject core operating limits are listed in TS 5.6.5.a and their values are located in the COLR. A revision to a fourth report, DPC-NE-1003, was also submitted for NRC review and approval.

DPC-NE-2009, Revision 1, "Duke Power Company Westinghouse Fuel Transition Report," August 2001.

DPC-NF-201 0, Revision 1, "Duke Power Company McGuire Nuclear Station and Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," August 2001.

DPC-NE-201 1, Revision 1, "Duke Power Company Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors," August 2001.

DPC-NE-1 003, Revision 1, "McGuire Nuclear Station and Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testing," August 2001.

The licensee in its letter of October 7, 2001, stated that, once approved, the approved topical report revisions, except for DPC-1003, Revision 1, will be listed in Section 5.6.5.b of the McGuire TS, to replace their respective original versions, and that the approved version of DPC-NE-201 1-P, Revision 1, will also be listed in the references for TS Bases 3.2.1 and 3.2.3 to replace the existing reference to the original version, DPC-NE-2011-P-A.

However, on July 10, 2002, the NRC issued amendments numbered 203 and 184 to the McGuire Unit 1 and 2 operating licenses that effectively relocated the topical report revision numbers and dates from the TS 5.6.5.b list of approved methodologies to the COLR.

Amendments 203 and 184 were consistent with the NRC Technical Specification Task Force (TSTF) Standard TS Traveler TSTF-363, "Revise Topical Report References in ITS 5.6.5 COLR." Accordingly, since this portion of its request is no longer needed in view of amendments 203 and 184, the licensee's letter dated August 7, 2002, eliminated the requc*.ts to change TS 5.6.5.b and proposed revisions to BASES 3.2.1 and 3.2.3 to make its submittal consistent with the implementation of amendments 203 and 184 at the McGuire Nuclear Station. Nonetheless, this Safety Evaluation sets forth the NRC staff's. evaluation of the licensee's proposed changes to the topical reports listed above.

2.0 BACKGROUND

Title 10 of the Code of FederalRegulations (10 CFR) Section 50.36 (c)(2)(ii)(B), Criterion 2, specifies that a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier must be included in the TS limiting conditions for operation (LCO). Accordingly, the reactor operating parameters, which are the initial conditions for the safety analyses of the design basis transients and accidents, are included in the TS LCOs.

Since many parameter limits, such as core physics parameters, generally change with each reload core, licensees previously needed to request TS amendments to update these parameters for each refueling cycle. NRC Generic Letter (GL) 88-16 (Ref. 4) provides guidance for relocating the values of the cycle-specific core operating parameter limits from TS to the COLR, thus eliminating unnecessary burden on the licensees and the NRC to update these limits in the TS for each fuel cycle. The guidance includes adding the COLR in the TS administrative reporting requirement that also specifies (1) the cycle-specific parameters included in the COLR, and (2) the analytical methods that the NRC has previously reviewed and approved to be used to determine the core operating parameters limits.

The McGuire TS 5.6.5, "Core Operating Limits Report (COLR)," conforms to GL 88-16 guidance. TS 5.6.5.a lists a set of parameters, including the reference to the actual TS number for each specified parameter. TS 5.6.5.b specifies the topical reports that are used for the determination of the core operating limits.

The proposed TS changes in this license amendment request are to revise the parameters listed in TS 5.6.5.a. These revisions are based on the guidance of GL 88-16.

3.0 STAFF EVALUATION In this section, the staff will discuss the review of the revised versions of the four previously approved topical reports submitted for staff review, and the proposed TS changes.

3.1 Topical Reports Revisions The licensee requested the NRC to review revisions to four topical reports that were previously approved and listed in TS 5.6.5.b as the approved methodologies used for the determination of the parameter limits in the COLR. Since the staff has reviewed and approved the original versions of these topical reports, the staff review of these revised versions concentrated on the revisions made to the approved reports.

3.1.1 DPC-NE-2009, Revision 1 Topical report, DPC-NE-2009-P-A, (Ref. 5), provides general information about the Robust Fuel Assembly (RFA) design and describes methodologies used for reload design analyses to support the licensing basis for use of RFAs in the McGuire and Catawba reload cores. These methodologies include fuel rod mechanical reload analysis methodology and the core design, thermal-hydraulic analysis, and accident analysis methodologies. The NRC approved the report in September 1999.

Revision 1 of DPC-NE-2009, as amended by the August 7, 2002, letter (Ref. 2), consists of the following minor changes to its Chapter 6, "UFSAR Accident Analyses."

(A) Update of the reference list in Section 6.7 as follows:

"* Update reference 6-25, WCAP-1 0054-P-A Addendum 2, to Revision 1, dated July 1997.

"* Correct reference 6-35, WCAP-8354, with proprietary topical report number, and designate the second report as a non-proprietary report.

"* Add reference 6-39, Westinghouse letter NSD-NRC-99-5839, "1998 Annual Notification of Changes to the Westinghouse Small Break LOCA and Large Break LOCA ECCS Evaluation Models, Pursuant to 10 CFR 50.46(a)(3)(ii)," dated July 15, 1999 (Ref. 6).

(B) Addition of a paragraph to Section 6.5.1, "Small Break LOCA," to explain that the Westinghouse small break LOCA NOTRUMP Evaluation Model includes the error corrections and model enhancements described in a few Westinghouse annual notifications required by 10 CFR 50.46, including the 1998 annual notification referenced in Reference 39.

The first two changes in the reference list are editorial and merely provide the latest version of the approved topical report or identify the proprietary and non-proprietary versions of a topical report. Reference 6-39, Westinghouse letter NSD-NRC-99-5839, is the annual notification of the changes to the LOCA evaluation models during 1998. This notification documented the following error corrections or model enhancements to the NOTRUMP small break LOCA Evaluation Model:

-4

"* A programming error correction on the SBLOCTA rod-to-rod radiation model, that is not modeled in licensing basis analyses and therefore, has no impact on the small break LOCA results.

"* A logic simplification to the NOTRUMP droplet fall model that produces insignificant differences in results.

" A change in the reactor coolant pump heat in NOTRUMP that is not used in the evaluation model and therefore, has no impact on the small break LOCA results.

" A modification of NOTRUMP steam generator tube condensation heat transfer logic for a foreign plant that does not affect standard Westinghouse Pressurized Water Reactor _.

calculations.

" An extension of reactor coolant conditions to allow for the NOTRUMP point kinetics calculations to be performed for cases that experience core uncovery conditions prior to reactor trip. For typical small break LOCA analyses, the reactor trips long before any threat of core uncovery and therefore, the change has no impact on peak cladding temperature calculations.

"* A programming change in SBLOCTA code to allow for modeling of variable length blankets on either ends of the rod that involves no changes to the thermal-hydraulic fuel rod model, nor the solution technique.

Since the changes documented in the Westinghouse annual notice have insignificant impact on the small break LOCA analyses, the staff concludes the addition of Reference 6-39 is __

acceptable. Therefore, Revision 1 of DPC-NE-2009-P-A, as modified in the August 7, 2002, letter, is acceptable.

3.1.2 DPC-NF-2010, Revision 1 Topical Report DPC-NF-2010, (Ref. 7), describes DPC's Nuclear Design Methodology for McGuire and Catawba Nuclear Stations. The nuclear design process consists of mechanical properties used as nuclear design input, the nuclear code system and methodology that DPC intends to use to perform design calculations and to provide operational support, and the -

development of statistical factors.

Revision 1 of DPC-NF-2010, updates the report to permit the use of certain methods approved subsequent to the implementation of the original version, such as the use of CASMO-3/

SIMULATE-3P reactor physics methods (Ref. 8). Other changes are made to reflect revisions to the core design parameters such as shutdown margin, boron and control rod worth, axial and radial peaking factors, and cycle length, as well as numerous editorial changes.

During the review, the staff also identified a few discrepancies associated with administrative changes. In response to the staff's request for additional information (Ref. 2), the licensee provided further changes to Revision 1 of the topical report. These modifications include clarifications to revised sections and minor changes to equations. The NRC staff has reviewed the analyses associated with the changes to Topical Report DPC-NF-2010 and the responses to the requests for additional information pertaining to these changes. The staff has concluded

that the changes to this topical report consist mostly of administrative changes and clarifications to the original NRC approved topical report and that there are no unreviewed methodology or regulatory issues. Therefore, the staff finds the changes to be acceptable.

3.1.3 DPC-NE-2011, Revision 1 Topical Report DPC-NF-201 1, (Ref. 9), describes the methodology for performing a maneuvering analysis for four-loop plants, such as the McGuire and Catawba Nuclear Stations.

The licensee has developed this methodology as an alternate to the existing Relaxed Axial Offset Control (RAOC) Methodology. The licensee pointed out that this maneuvering analysis results in several advantages: more flexible and prompt engineering support for the operating stations, consistency with the methods of the licensee's nuclear design process, and potential increases in available margin through the use of three-dimensional monitoring techniques. The increase in margin occurs in limits on power distribution, control rod insertion, and power distribution inputs to the overpower delta-temperature and over-temperature delta-temperature reactor protection system (RPS) trip functions.

Revision 1 of DPC-NE-201 1, updates the report to include editorial changes, and to permit the use of certain methods approved subsequent to the implementation of the original version, such as the CASMO-3/SIMULATE-3P methodology (Ref. 8). Other changes are made to reflect revisions to the core design parameters such as power peaking factors, axial and radial power distributions, and cycle length, as well as numerous editorial changes.

In response to the NRC staff's request for additional information (Ref. 2), the licensee provided additional information regarding cycle depletion times to clarify issues associated with power peaking versus burnup as a function of cycle time. The licensee's amendment request also included clarifications to revised sections and minor changes to equations. The NRC staff has reviewed the analyses associated with the changes to Topical Report DPC-NE-201 1-A and the responses to the requests for additional information pertaining to the requested changes. Since the changes to this topical report consist mostly of administrative changes and clarifications to the original NRC approved topical report, the staff finds the changes to be acceptable.

3.1.4 DPC-NE-1003, Revision 1 Topical Report DPC-NE-1 003 (Ref.' 10), describes the measurement procedure used to determine the inferred bank worth and the calculation procedures used to develop the rod swap correction factor that accounts for the effect of a test bank on the partial integral worth of the reference bank. The NRC approved the report in May 1987 (Ref. 11) for rod worth measurement of reload cores for McGuire and Catawba Stations, Units 1 and 2.

Revision 1 of DPC-NE-1003 updates the report to permit the use of certain methods approved subsequent to the implementation of the original version, such as the use of CASMO-3/

SIMULATE-3P reactor physics methods (Ref. 8). Other changes are made to reflect the revision of the rod swap measurement procedures, and various editorial changes. In response to staff questions, the licensee, in its letter of August 7, 2002, provided the current version of the control rod worth measurement rod swap procedures, PT/O/A/4150/1 1A, dated January 19, 1996. The staff review of this current control rod worth measurement procedure has found it to be acceptable. The licensee, in the August 7, 2002, letter also modified the equation in Section 3 of the topical report for the calculation of the inferred rod bank worth from the

measured reference bank worth and bank height. This change is consistent with the equation described in step 12.12.5 of the current measurement procedures of January 19, 1996.

Therefore, Revision 1 of DPC-NE-1 003, as modified in the August 7, 2002, letter, is acceptable.

3.2 Proposed TS Changes This section addresses the staff's evaluation of the proposed changes to TS 5.6.5.a regarding the cycle-specific operating parameters specified in the COLR. The staff review of these TS changes are based on the guidance of GL 88-16.

TS 5.6.5.a provides a list of core operating limits that are established prior to each reload cycle, or prior to any remaining portion of a reload cycle. The values of the limits are located in the COLR. For McGuire Nuclear Station, Units 1 and 2, the licensee proposed to revise the list by:

(1) adding "60 ppm" to Item 5.6.5.a.1 regarding the moderator temperature coefficient (MTC) surveillance limit for Specification 3.1.3, and (2) adding Item 5.6.5.a.12, "31 EFPD surveillance penalty factors for Specifications 3.2.1 and 3.2.2."

These changes are evaluated below.

3.2.1 MTC 60 ppm Surveillance Limit McGuire TS LCO 3.1.3 specifies that the MTC be maintained within the LCO limits, which are based on the safety analysis assumptions. For verification that these LCO limits are met, the Surveillance Requirements of TS 3.1.3 also place surveillance limits for conducting the end of cycle MTC measurement at boron concentrations of 300 ppm and 60 ppm. The LCO limits and the 300 ppm and 60 ppm surveillance limits are specified in the COLR. However, TS Item 5.6.5.a.1 operating limits does not currently identify the 60-ppm surveillance limit.

The proposed change to the McGuire TS would add the 60 ppm surveillance limit in Item 5.6.5.a.1. The new TS would read "Moderator Temperature Coefficients BOL and EOL limits and 60 ppm and 300 ppm surveillance limit for Specification 3.1.3." The NRC approved incorporating the 60-ppm surveillance limits into the COLR during the Improved Technical Specifications conversion in 1998 (Ref. 12 and 13); however, reference to this surveillance was not included in TS Item 5.6.5.a.1 at that time. The proposed TS change to include the 60 ppm surveillance limit in TS Item 5.6.5.a.1 provides consistency with previously approved requirements and, therefore, it is acceptable.

3.2.2 Relocation of Hot Channel Factors Surveillance Penalty Factors to COLR Surveillance Requirements in TS 3.2.1 and 3.2.2, respectively, require that the heat flux hot channel factor, Fq (x,y,z), and the enthalpy rise hot channel factor, Fh (x,y), be measured every 31 effective full power days (EFPD) during equilibrium conditions using the incore detector system to verify they are within the respective limits. To address the possibility that these hot channel factors may increase and exceed their allowable limits between surveillances, penalty factors are applied to these hot channel factors if their margins to the respective limits have decreased since the previous surveillance. These margin-decrease penalty factors are

calculated by projecting the limiting hot channel factors over the 31 EFPD surveillance intervals with the maximum changes at the limiting core location, and are based on reload core design.

In Section 8, "Improved Technical Specification Changes," of DPC-NE-2009, the licensee proposed to replace the penalty factors with tables of penalty value as a function of burnup in the COLR to facilitate cycle-specific updates. TS Item 5.6.5.b.14 lists topical report DPC-NE-2009-P-A that includes (in response to a staff question during the review of DPC-NE-2009) the approved methodology used to calculate these burnup-dependent penalty factors. The staff found the methodology and the inclusion of the burnup-dependent margin decrease penalty factors in the COLR acceptable, as stated in the staff's Safety Evaluation supporting license Amendment Nos. 188 and 169, respectively, for McGuire Nuclear Station, Units 1 and 2 (Ref. 14).

The proposed changes to the McGuire TS would add Item 5.6.5.a.12 that reads: "31 EFPD surveillance penalty factors for Specifications 3.2.1 and 3.2.2." The addition of TS Item 5.6.5.a.12 would make it consistent with the previous staff approval of including these surveillance penalty factors in the COLR and, therefore, this proposed change is acceptable.

4.0

SUMMARY

The staff has reviewed the revisions to four previously approved topical reports described in Section 1.0 of this Safety Evaluation, and the proposed changes to McGuire Nuclear Station, Units 1 and 2, TS 5.6.5.a related to the COLR. Based on our evaluation, described in Section 3 of this Safety Evaluation, the staff concludes that the these topical report revisions, as amended by the August 7, 2002, letter, and the TS changes are acceptable.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the North Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change recordkeeping, reporting, or administrative procedure requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (67FR 54680). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

I

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7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

1. Letter from M. S. Tuckman, Duke Energy Corporation, to US Nuclear Regulatory Commission, "Duke Energy Corporation; Catawba Nuclear Station Units 1 and 2, Docket Nos. 50-413, 50-414; McGuire Nuclear Station Units 1 and 2, Docket Nos. 50-369, 50-370; License Amendment Request Applicable to Technical Specifications 5.6.5, Core Operating Limits Report; Revisions to BASES 3.2.1 and 3.2.3; and Revisions to Topical Reports DPC-NE-2009-P, DPC-NF-2010, DPC-NE-201 1-P, and DPC-NE-1003,"

October 7, 2001.

2. Letter from M. S. Tuckman, Duke Energy Corporation, to US Nuclear Regulatory Commission, "Duke Energy Corporation; McGuire Nuclear Station Units 1 and 2, Docket Nos. 50-369 and 370; Catawba Nuclear Station Units 1 and 2, Docket Nos. 50-413 and 414; Response to NRC Request for Additional Information - TAC nos. MB3222, MB3223, MB3343 and MB3344) and License Amendment Request Supplement," August 7, 2002.

3. Letter from M. S. Tuckman, Duke Energy Corporation, to US Nuclear Regulatory Commission, "License Amendment Request Applicable to the Technical Specifications Requirements for the Core Operating Limits Report - Oconee, McGuire, and Catawba Technical Specifications 5.6.5," December 20, 2001.

4. Letter from Dennis Crutchfield, USNRC, to All Power Reactor Licensees and Applicants, "Removal of Cycle-Specific Parameter Limits from Technical Specifications (Generic Letter 88-16)," October 4, 1988.

5. DPC-NE-2009-P-A, "Duke Power Company Westinghouse Fuel Transition Report,"

December 1999.

6. Letter from J. S. Galembush, Westinghouse Electric Company, to US Nuclear Regulatory Commission, "1998 Annual Notification of Changes to the Westinghouse Small Break LOCA and Large Break LOCA ECCS Evaluation Models, Pursuant to 10 CFR 50.46(a)(3)(ii)," NSD-NRC-99-5839, July 15, 1999.

7. DPC-NF-201 OA, "Duke Power Company McGuire Nuclear Station and Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985..

8. DPC-NE-1004A, Revision 1, "Nuclear Design Methodology Using CASMO-3/

SIMULATE-3P," SER dated April 26, 1997.

9. DPC-NE-201 1, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors," March 1990.

10. DPC-NE-1 003, "Rod Swap Methodology Report for Startup Physics Testing,"

December 1986.

11. Letter from Dad Hood, USNRC, to H. B. Tucker, Duke Power Company, "Rod Swap Methodology Report for Startup Physics Testing, McGuire and Catawba Nuclear Stations, Units I and 2 (TACs 62981, 62982, 62983, 62984)," May 22, 1987.

12. Letter from Frank Rinaldi, USNRC, to H. B. Brown, McGuire Site, Duke Energy Corporation, "Issuance of Amendments - McGuire Nuclear Station, Units 1 and 2, (TAC Nos. M98964 and M98965)," September 30, 1998.

13. Letter from Peter Tam, USNRC, to G. R. Peterson, Catawba Nuclear Station, Duke Energy Corporation, "Issuance of Amendments - Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M95298 and M95299)," September 30, 1998.

14. Letter from Frank Rinaldi, USNRC, to H. B. Brown, McGuire Site, Duke Energy Corporation, "McGuire Nuclear Station, Units 1 and 2, Re: Issuance of Amendments (TAC Nos. MA241 1 and MA2412)," September 22, 1999.

15. Letter from Peter Tam, USNRC, to G. R. Peterson, Catawba Nuclear Station, Duke Energy Corporation, "Catawba Nuclear Station, Units 1 and 2, Re: Issuance of Amendments (TAC Nos. MA2359 and MA2361)," September 22, 1999.

16. Letter from Robert F. Martin, USNRC, to David L. Rehn, Catawba Site, Duke Power Company, "Issuance of Amendments - Catawba Nuclear Station, .Units 1 and 2 Cycle Specific Parameters to the Core Operating Limits Report (TAC Nos. M85472 and M85473)," March 25, 1994.

Principal Contributor Y. Hsii A. Attard Date: October 1, 2002

1--- Coa o UNITED STATES 1P o I ~ NUCLEAR REGULATORY COMMISSIO

-~ WASHINGTON, D.C. 20555-0001

ý-J lop October 1, 2002 ci - 8 Mr. G. R. Peterson Site Vice President REGULATORY CO,*ALIAý,-4E Catawba Nuclear Station Duke Energy Corporation 4800 Concord Road York, South Carolina 29745-9635

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2 RE: ISSUANCE OF AMENDMENTS (TAC NOS. MB3343 AND MB3344)

Dear Mr. Peterson:

The Nuclear Regulatory Commission has issued the enclosed Amendnient No. 202 to Facility Operating License NPF-35 and Amendment No.195 to Facility Operating License NPF-52 for the Catawba Nuclear Station, Units 1 and 2. The amendments consist of changes to the Technical Specifications (TS) in response to your application dated October 7, 2001, as supplemented by letter dated August 7, 2002.

The amendments revise TS 5.6.5.a by adding a few parameter limits currently included in the Core Operating Limits Report. In addition to the license amendment request, you also submitted revisions to four previously approved topical reports for the Nuclear Regulatory Commission staff review and approval. The enclosed Safety Evaluation also address these topical reports.

A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Chandu P. Patel, Project Manager, Section 1 Project Directorate II /RA/

Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-413 and 50-414

Enclosures:

1. Amendment No. 202 to NPF-35

2. Amendment No. 195 to NPF-52

3. Safety Evaluation cc w/encls: See next page

Catawba Nuclear Station cc:

Mr. Gary Gilbert North Carolina Electric Membership Regulatory Compliance Manager Corporation Duke Energy Corporation P. 0. Box 27306 4800 Concord Road Raleigh, North Carolina 27611 York, South Carolina 29745 Senior Resident Inspector Ms. Lisa F. Vaughn U.S. Nuclear Regulatory Commission Legal Department (PB05E) 4830 Concord Road Duke Energy Corporation York, South Carolina 29745 422 South Church Street Charlotte, North Carolina 28201-1006 Virgil R. Autry, Director Division of Radioactive Waste Management Anne Cottingham, Esquire Bureau of Land and.Waste Management Winston and Strawn Department of Health and Environmental 1400 L Street, NW Control Washington, DC 20005 2600 Bull Street Columbia, South Carolina 29201-1708 North Carolina Municipal Power Agency Number 1 Mr. C. Jeffrey Thomas 1427 Meadowwood Boulevard Manager - Nuclear Regulatory P. O. Box 29513 Licensing Raleigh, North Carolina 27626 Duke Energy Corporation 526 South Church Street County Manager of York County Charlotte, North Carolina 28201-1006 York County Courthouse York, South Carolina 29745 Saluda River Electric P. O. Box 929 Piedmont Municipal Power Agency Laurens, South Carolina 29360 121 Village Drive Greer, South Carolina 29651 Mr. Peter R. Harden, IV VP-Customer Relations and Sales Ms. Karen E. Long Westinghouse Electric Company Assistant Attorney General 6000 Fairview Road North Carolina Department of Justice 12th Floor P. O. Box 629 Charlotte, North Carolina 28210 Raleigh, North Carolina 27602 Elaine Wathen, Lead REP Planner Division of Emergency Management 116 West Jones Street Raleigh, North Carolina 27603-1335

Catawba Nuclear Station cc:

Mr. T. Richard Puryear Owners Group (NCEMC)

Duke Energy Corporation 4800 Concord Road York, South Carolina 29745 Richard M. Fry, Director Division of Radiation Protection North Carolina Department of Environment, Health, and Natural Resources 3825 Barrett Drive Raleigh, North Carolina 27609-7721 v.-_

-*, UNITED STATES

- NUCLEAR REGULATORY COMMISSION

- 0 WASHINGTON. D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 29? TO FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO. !95 TO FACILITY OPERATING LICENSE NPF-52 DUKE ENERGY CORPORATION, ET AL.

CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414

1.0 INTRODUCTION

By letter dated October 7, 2001, as supplemented by letter dated August 7, 2002, Duke Energy Corporation, et al. (DEC, the licensee), submitted a request for changes to the Catawba Nuclear Station, Units 1 and 2, Technical Specifications (TS).

Revisions were proposed for TS 5.6.5.a, Item 1, to add the moderator temperature coefficient (MTC) 60 parts per million (ppm) surveillance limit. The specific value of the surveillance limit was previously relocated to the Core Operating Limits Report (COLR). Two new items were also proposed to be added to TS 5.6.5.a. These two items are (1) Item 12, "31 EFPD surveillance penalty factors for Specifications 3.2.1 and 3.2.2," and (2) Item 13, "Reactor makeup water pumps combined flow rates limit for Specifications 3.3.9 and 3.9.2."

The initial submittal, dated October 7, 2001, proposed to change the dates and revision numbers for three of the Nuclear Regulatory Commission (NRC) approved analytical methods previously listed in TS 5.6.5.b, as listed below. The changes would reflect later versions of these topical reports that were also submitted with the October 7, 2001, submittal for NRC review and approval. As required by TS 5.6.5.b, only those methods listed within the TS as having been reviewed and approved by the NRC, can be used to determine the subject core operating limits. The subject core operating limits are listed in TS 5.6.5.a and their values are located in the COLR. A revision to a fourth report, DPC-NE-1003, was also submitted for NRC review and approval.

0 DPC-NE-2009, Revision 1, "Duke Power Company Westinghouse Fuel Transition Report," August 2001.

0 DPC-NF-201 0, Revision 1, "Duke Power Company McGuire Nuclear Station and Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," August 2001.

DPC-NE-201 1, Revision 1, "Duke Power Company Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors," August 2001.

DPC-NE-1003, Revision 1, "McGuire Nuclear Station and Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testing," August 2001.

The licensee in its letter of October 7, 2001, stated that, once approved, the approved topical report revisions, except for DPC-1003, Revision 1, will be listed in Section 5.6.5.b of the Catawba TS, to replace their respective original versions, and that the approved version of DPC-NE-201 I-P, Revision 1, will also be listed in the references for TS Bases 3.2.1 and 3.2.3 to replace the existing reference to the original version, DPC-NE-201 1-P-A.

However, on July 2, 2002, the NRC issued amendments numbered 199 and 192 to the Catawba Unit 1 and 2 operating licenses that effectively relocated the topical report revision numbers and dates from the TS 5.6.5.b list of approved methodologies to the COLR.

Amendments 199 and 192 were consistent with the NRC Technical Specification Task Force (TSTF) Standard TS Traveler TSTF-363, "Revise Topical Report References in ITS 5.6.5 COLR." Accordingly, since this portion of its request is no longer neede~d in view of amendments 199 and 192, the licensee's letter dated August 7, 2002, eliminated the requests to change TS 5.6.5.b and proposed revisions to BASES 3.2.1 and 3.2.3 to make its submittal consistent with the implementation of amendments 199 and 192 at the Catawba Nuclear Station. Nonetheless, this Safety Evaluation sets forth the NRC staff's evaluation of the licensee's proposed changes to the topical reports listed above.

2.0 BACKGROUND

Title 10 of the Code of FederalRegulation (10 CFR) Section 50.36 (c)(2)(ii)(B), Criterion 2 specifies that a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier must be included in the TS limiting conditions for operation (LCO). Accordingly, the reactor operating parameters, which are the initial conditions for the safety analyses of the design basis transients and accidents, are included in the TS LCO.

Since many parameters limits, such as core physics parameters, generally change with each reload core, licensees need to request TS amendments to update these parameters for each refueling cycle. NRC Generic Letter (GL) 88-16 (Ref. 4) provides guidance for relocating the values of the cycle-specific core operating parameter limits from TS to the COLR, and thus eliminates the unnecessary burden on the licensees and the NRC to update these limits in the TS each fuel cycle. The guidance includes adding the COLR in the TS administrative reporting requirement that also specifies (1) the cycle-specific parameters included in the COLR, and (2) the analytical methods that the NRC has previously reviewed and approved to be used to determine the core operating parameters limits.

The Catawba TS 5.6.5, "Core Operating Limits Report (COLR)," conforms to the GL 88-16 guidance. TS 5.6.5.a lists a set of parameters, including the reference to the actual TS number for each specified parameter. TS 5.6.5.b specifies the topical reports that are used for the determination of the core operating limits.

The proposed TS changes in this license amendment request are to revise the parameters listed in TS 5.6.5.a. These revisions are based on the guidance of GL 88-16.

3.0 STAFF EVALUATION In this section, the staff will discuss the review of the revised versions of the four previously approved topical reports submitted for staff review, and the proposed TS changes.

3.1 Topical Reports Revisions The licensee requested the NRC to review revisions of four topical reports that were previously approved and listed in TS 5.6.5.b as the approved methodologies used for the determination of the parameter limits in the COLR. Since the staff has reviewed and approved the original versions of these topical reports, the staff review of these revised versions will concentrate on the revisions made to the approved reports.

3.1.1 DPC-NE-2009, Revision 1 Topical report, DPC-NE-2009-P-A, (Ref. 5), provides general information about the Robust Fuel Assembly (RFA) design and describes methodologies used for reload design analyses to support the licensing basis for use of the RFA design in the McGuire and Catawba reload cores. These methodologies include fuel rod mechanical reload analysis methodology and the core design, thermal-hydraulic analysis, and accident analysis methodologies. The NRC approved the report in September 1999.

Revision 1 of DPC-NE-2009-A, as amended by the August 7, 2002, letter (Ref. 2), consists of the following minor changes to Chapter 6, "UFSAR Accident Analyses:"

(A) Update of the reference list in Section 6.7 as follows:

"* Update reference 6-25, WCAP-10054-P-A Addendum 2, to Revision 1, dated July 1997.

"* Correct reference 6-35, WCAP-8354, with proprietary topical report number, and designate the second report as a non-proprietary report.

"* Add reference 6-39 a Westinghouse letter NSD-NRC-99-5839, "1998 Annual Notification of Changes to the Westinghouse Small Break LOCA and Large Break LOCA ECCS Evaluation Models, Pursuant to 10 CFR 50.46(a)(3)(ii)," dated July 15, 1999 (Ref. 6).

(B) Addition of a paragraph to Section 6.5.1, "Small Break LOCA," to explain that the Westinghouse small break LOCA NOTRUMP Evaluation Model includes the error corrections and model enhancements described in a few Westinghouse annual notifications required by 10 CFR 50.46, including the 1998 annual notification referenced in Reference 39.

The first two changes in the reference list are editorial and merely provide the latest version of the approved topical report or identify the proprietary and non-proprietary versions of a topical report. Reference 6-39, the Westinghouse letter NSD-NRC-99-5839, is the annual notification of the changes to the LOCA evaluation models during 1998. This notification documented the following error corrections or model enhancements to the NOTRUMP small break LOCA Evaluation Model:

-4

" A programming error correction on the SBLOCTA rod-to-rod radiation model that is not modeled in licensing basis analyses and therefore, has no impact on the small break LOCA results.

"* A logic simplification to the NOTRUMP droplet fall model that produces insignificant differences in results.

"* A change in the reactor coolant pump heat in NOTRUMP that is not used in the evaluation model and therefore, has no impact on the small break LOCA results.

"* A modification of NOTRUMP steam generator tube condensation heat transfer logic to a foreign plant that does not affect standard Westinghouse Pressurized Water Reactor calculations.

" An extension of reactor coolant conditions to allow for the NOTRUMP point kinetics calculations to be performed for cases that experience core uncoIery conditions prior to reactor trip. For typical small break LOCA analyses, the reactor trips long before any threat of core uncovery and therefore, the change has no impact on peak cladding temperature calculations.

"* A programming change in SBLOCTA code to allow for modeling of variable length blankets on either ends of the rod that involves no changes to the thermal-hydraulic fuel rod model, nor the solution technique.

Since the changes documented in the Westinghouse annual notice hava insignificant impact on the small break LOCA analyses, the staff concludes the addition of Reference 6-39 is acceptable. Therefore, Revision 1 of DPC-NE-2009-P-A, as modified in the August 7, 2002, letter, is acceptable.

3.1.2 DPC-NF-2010A, Revision 1 Topical Report DPC-NF-201 OA, (Ref. 7), describes Duke Power Company's Nuclear Design Methodology for McGuire and Catawba Nuclear Stations. The nuclear design process consists of mechanical properties used as nuclear design input, the nuclear code system and methodology the licensee intends to use to perform design calculations and to provide operational support, and the development of statistical factors.

Revision 1 of DPC-NF-201 0A, updates the report to permit the use of certain methods approved subsequent to the implementation of the original version, such as the use of CASMO-3/SIMULATE-3P reactor physics methods (Ref. 8). Other changes are made to reflect revisions to the core design parameters such as shutdown margin, boron and control rod worth, axial and radial peaking factors, and cycle length, as well as numerous editorial changes.

During the review, the staff also identified a few discrepancies associated with administrative changes. In response to the staff's request for additional information (Ref. 2), the licensee provided further changes to Revision 1 of the Topical report. These modifications include clarifications to revised sections and minor changes to equations. The NRC staff has reviewed the analyses associated with the changes to Topical Report DPC-NF-2010A and the responses to the requests for additional information pertaining to these changes. The staff has concluded

that the changes to this topical report consist mostly of administrative changes and clarifications to the original NRC approved topical report and that there are no unreviewed methodology or regulatory issues. Therefore, the staff finds the changes acceptable.

3.1.3 DPC-NE-201 1, Revision 1 Topical Report DPC-NE-201 1, (Ref. 9), describes the methodology for performing a maneuvering analysis for four-loop plants, such as McGuire and Catawba Nuclear Station. The licensee has developed this methodology as an alternate to the existing Relaxed Axial Offset Control Methodology. The licensee pointed out that this maneuvering analysis results in several advantages: more flexible and prompt engineering support for the operating stations, consistency with the methods of the licensee's nuclear design process, and potential increases in available margin through the use of three-dimensional monitoring techniques. The increase in margin occurs in limits on power distribution, control rod insertion, and power distribution inputs to the overpower delta-temperature and over-temperature delta-temperature reactor protection system trip functions.

Revision 1 of DPC-NE-201 1, updates the report to include editorial changes, and to permit the use of certain methods approved subsequent to the implementation of the original version, such as the use of CASMO-3/SIMULATE-3P methodology (Ref. 8). Other changes are made to reflect revisions to the core design parameters such as power peaking factors, axial and radial power distributions, and cycle length, as well as numerous editorial changes.

In response to the NRC staff's request for additional information (Ref. 2), the licensee provided additional information to the staff regarding cycle depletion times to clarify issues associated with power peaking versus bumup as a function of cycle time. The licensee's amendment request also included clarifications to revised sections and minor changes to equations. The NRC staff has reviewed the analyses associated with the changes to Topical Report DPC-NE-201 1-A and the responses to the requests for additional information pertaining to the requested changes. Since the changes to this topical report consists mostly of administrative changes and clarifications to the original NRC approved topical report, the staff find the changes acceptable.

3.1.4 DPC-NE-1003, Revision 1 Topical Report DPC-NE-1 003 (Ref. 10) describes the measurement procedure used to determine the inferred bank worth and the calculation procedures used to develop the rod swap correction factor that accounts for the effect of a test bank on the partial integral worth of the reference bank. The NRC approved the report in May 1987 (Ref. 11) for rod worth measurement of reload cores for McGuire and Catawba Stations, Units 1 and 2.

Revision 1 of DPC-NE-1003 updates the report to permit the use of certain methods approved subsequent to the implementation of the original version, such as the use of CASMO-3/

SIMULATE-3P reactor physics methods (Ref. 8). Other changes are made to reflect the revision of the rod swap measurement procedures, and various editorial changes. In response to staff questions, the licensee, in its letter of August 7, 2002, provided the current version of the control rod worth measurement rod swap procedures, PT/O/A/4150/11 A, dated January 19, 1996. The staff review of this current control rod worth measurement procedure has found it acceptable. The licensee in the August 7, 2002, letter also modified the equation in Section 3

of the topical report for the calculation of the inferred rod bank worth from the measured reference bank worth and bank height. This change is consistent with the equation described in step 12.12.5 of the current measurement procedures of January 19, 1996. Therefore, Revision 1 of DPC-NE-1003, as modified in the August 7, 2002, letter, is acceptable.

3.2 Proposed TS Changes This section addresses the staff's evaluation of the proposed changes to TS 5.6.5.a regarding the cycle-specific operating parameters specified in the COLR. The staff review of these TS changes are based on the guidance of GL 88-16.

TS 5.6.5.a provides a list of core operating limits that are established prior to each reload cycle, or prior to any remaining portion of a reload cycle. The valves of the limits are in the COLR.

For Catawba Units 1 and 2, the licensee proposed to revise the list by:

(1) adding "60 ppm" to Item 5.6.5.a.1 regarding the moderator temperature coefficient (MTC) surveillance limit for Specification 3.1.3, (2) adding Item 5.6.5.a.12, "31 EFPD surveillance penalty factors for Specifications 3.2.1 and 3.2.2," and (3) adding Item 5.6.5.a.1 3, "Reactor makeup water pumps combined flow rates limit for Specifications 3.3.9 and 3.9.2."

These changes are evaluated below.

3.2.1 MTC 60 ppm Surveillance Limit Catawba TS LCO 3.1.3 specifies that the MTC be maintained within the LCO limits, which are based on the safety analysis assumptions. For verification that these LCO limits are met, the Surveillance Requirements of TS 3.1.3 also places surveillance limits for conducting the end of cycle MTC measurement at 300 ppm and 60 ppm boron concentration. The LCO limits and the 300-ppm and 60-ppm surveillance limits are specified in the COLR. However, TS Item 5.6.5.a.1 operating limits does not currently identify the 60-ppm surveillance limit.

The proposed change to the Catawba TS would add the 60-ppm surveillance limit in Item 5.6.5.a.1. The new TS would read "Moderator Temperature Coefficients BOL and EOL limits and 60 ppm and 300 ppm surveillance limit for Specification 3.1.3." The NRC approved incorporating the 60-ppm surveillance limits into the COLR during the Improved Technical Specifications conversion in 1998 (Ref. 12 and 13); however, reference to this surveillance was not included in TS Item 5.6.5.a.1 at that time. The proposed TS change to include the 60-ppm surveillance limit in TS Item 5.6.5.a.1 provides consistency with previously approved requirements and, therefore, it is acceptable.

3.2.2 Relocation of Hot Channel Factors Surveillance Penalty Factors to COLR Surveillance Requirements in TS 3.2.1 and 3.2.2, respectively, require that the heat flux hot channel factor, Fq (x,y,z), and the enthalpy rise hot channel factor, F., (x,y), be measured every 31 effective full power days (EFPD) during equilibrium conditions using the incore detector

system to verify they are within the respective limits. To address the possibility that these hot channel factors may increase and exceed their allowable limits between surveillances, penalty factors are applied to these hot channel factors if their margins to the respective limits have decreased since the previous surveillance. These margin-decrease penalty factors are calculated by projecting the limiting hot channel factors over the 31 EFPD surveillance intervals with the maximum changes at the limiting core location, and are based on reload core design.

In Section 8, "Improved Technical Specification Changes," of DPC-NE-2009, the licensee proposed to replace the penalty factors with tables of penalty value as functions of burnup in the COLR to facilitate cycle-specific~updates. TS Item 5.6.5.b.14 lists topical report DPC-NE-2009-P-A that includes (in response to a staff question during the review of DPC-NE-2009) the approved methodology used to calculate these burnup-dependent penalty factors. The staff found the methodology and the inclusion of the burnup-dependent margin decrease penalty factors in the COLR acceptable as stated in the staff's safety evaluation supporting license amendment Nos. 180 and 172, respectively for Catawba Units 1 and 2 (Ref. 15).

The proposed changes to the Catawba TS would add Item 5.6.5.a.12, that reads: "31 EFPD surveillance penalty factors for Specifications 3.2.1 and 3.2.2." The addition of TS Item 5.6.5.a.12 would make it consistent with the previous staff approval of including these surveillance penalty factors in the COLR and, therefore, this proposed change is acceptable.

3.2.3 Reactor Makeup Water Pumps Combined Flow Rates Limit The relocation of the reactor makeup water pumps combined flow rates limit for the boron dilution mitigation system from Catawba TS 3.3.9 and 3.9.2 to the COLR was approved by the NRC as described in a letter dated March 25, 1994 (Ref. 16). The reactor makeup water pumps flow rate limit is included in the Catawba COLR.

The proposed changes to the Catawba TS would add Item 5.6.5.a.13, "Reactor makeup water pumps combined flow rates limit for Specification 3.3.9 and 3.9.2," to TS 5.6.5.a. The addition of this item would make the TS 5.6.5.a list consistent with the core operating limits included in the Catawba COLR and is therefore, acceptable.

4.0

SUMMARY

The staff has reviewed the revisions of four previously approved topical reports described in Section 1.0 of this Safety Evaluation, and the proposed changes to Catawba Nuclear Station, Units 1 and 2, TS 5.6.5.a related to the COLR. Based on our evaluation described in Section 3 of this Safety Evaluation, the staff concludes that the these topical report revisions, as amended by the August 7, 2002, letter, and the TS changes are acceptable.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding [67 FR 54680]. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

1. Letter from M. S. Tuckman, Duke Energy Corporation, to US Nuclear Regulatory Commission, "Duke Energy Corporation; Catawba Nuclear Station Units 1 and 2, Docket Nos. 50-413, 50-414; McGuire Nuclear Station Units 1 and 2, Docket Nos. 50-369, 50-370; License Amendment Request Applicable to Technical Specifications 5.6.5, Core Operating Limits Report; Revisions to BASES 3.2.1 and 3.2.3; and Revisions to Topical Reports DPC-NE-2009-P, DPC-NF-201 0, DPC-NE-201 1-P, and DPC-NE-1 003,"

October 7, 2001.

2. Letter from M. S. Tuckman, Duke Energy Corporation, to US Nuclear Regulatory Commission, "Duke Energy Corporation; McGuire Nuclear Station Units 1 and 2, Docket Nos. 50-369 and 370; Catawba Nuclear Station Units 1 and 2, Docket Nos. 50-413 and 414; Response to NRC Request for Additional Information - TAC nos. MB3222, MB3223, MB3343 and MB3344) and License Amendment Request Supplement," August 7, 2002.

3. Letter from M. S. Tuckman, Duke Energy Corporation, to US Nuclear Regulatory Commission, "License Amendment Request Applicable to the Technical Specifications Requirements for the Core Operating Limits Report - Oconee, McGuire, and Catawba Technical Specifications 5.6.5," December 20, 2001.

4. Letter from Dennis Crutchfield, USNRC, to All Power Reactor Licensees and Applicants, "Removal of Cycle-Specific Parameter Limits from Technical Specifications (Generic Letter 88-16)," October 4, 1988.

5. DPC-NE-2009-P-A, "Duke Power Company Westinghouse Fuel Transition Report,"

December 1999.

6. Letter from J. S. Galembush, Westinghouse Electric Company, to US Nuclear Regulatory Commission, "1998 Annual Notification of Changes to the Westinghouse Small Break LOCA and Large Break LOCA ECCS Evaluation Models, Pursuant to 10 CFR 50.46(a)(3)(ii)," NSD-NRC-99-5839, July 15, 1999.

7. DPC-NF-2010A, "Duke Power Company McGuire Nuclear Station and Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985.

8. DPC-NE-1004A, Revision 1, "Nuclear Design Methodology Using CASMO-3/

SIMULATE-3P," SER dated April 26, 1997.

9. DPC-NE-201 1, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors," March 1990.

10. DPC-NE-1 003, "Rod Swap Methodology Report for Startup Physics Testing," December 1986.

11. Letter from Darl Hood, USNRC, to H. B. Tucker, Duke Power Company, "Rod Swap Methodology Report for Startup Physics Testing, McGuire and Catawba Nuclear Stations, Units 1 and 2 (TACs 62981, 62982, 62983, 62984)," May 22, 1987.

12. Letter from Frank Rinaldi, USNRC, to H. B. Brown, McGuire Site, Duke Energy Corporation, "Issuance of Amendments - McGuire Nuclear Station, Units 1 and 2, (TAC Nos. M98964 and M98965)," September 30, 1998.

13. Letter from Pater Tam, USNRC, to G. R. Peterson, Catawba Nuclear Station, Duke Energy Corporation, "Issuance of Amendments - Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M95298 and M95299)," September 30, 1998.

14. Letter from Frank Rinaldi, USNRC, to H. B. Brown, McGuire Site, Duke Energy Corporation, "McGuire Nuclear Station, Units 1 and 2, Re: Issuance of Amendments (TAC Nos. MA2411 and MA2412)," September 22, 1999.

15. Letter from Peter Tam, USNRC, to G. R. Peterson, Catawba Nuclear Station, Duke Energy Corporation, "Catawba Nuclear Station, Units 1 and 2, Re: Issuance of Amendments (TAC Nos. MA2359 and MA2361)," September 22, 1999.

16. Letter from Robert F. Martin, USNRC, to David L. Rehn, Catawba Site, Duke Power Company, "Issuance of Amendments - Catawba Nuclear Station, Units 1 and 2 Cycle Specific Parameters to the Core Operating Limits Report (TAC Nos. M85472 and M85473)," March 25, 1994.

Principal Contributor: Y. Hsii A. Attard Date: October 1, 2002

v t e Letter Sequence Other
TAC:MB3222, Revise Topical Report References in ITS 5.6.5, Colr (Approved, Closed) TAC:MB3223, Revise Topical Report References in ITS 5.6.5, Colr (Approved, Closed) TAC:MB3343, Revise Topical Report References in ITS 5.6.5, Colr (Approved, Closed) TAC:MB3344, Revise Topical Report References in ITS 5.6.5, Colr (Approved, Closed)
Initiation Acceptance... Supplement Administration Questions Answers Results Approval Other: ML022390110, ML022740527, ML022750119, ML022750206, ML023330372, ML030590158, ML030590174, ML030590245, ML030590251, ML030590324
MONTHYEAR2002-08-20 [Table View]

ML030590324

Text

Subject:

Subject:

RESPONSE

SUBJECT:

Dear Mr. Tuckman:

Enclosure:

Subject:

SUBJECT:

SUBJECT:

SUBJECT:

SUBJECT:

SUBJECT:

Response

Response

Subject:

RESPONSE

RESPONSE

Title:

2.0 REFERENCES

3.1 'Duration

Dear Mr. Tucker:

Subject:

Enclosure:

Background

SUBJECT:

Dear Mr. Barron:

Enclosures:

1.0 INTRODUCTION

2.0 BACKGROUND

SUMMARY

5.0 STATE CONSULTATION

6.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

8.0 REFERENCES

SUBJECT:

Dear Mr. Peterson:

Enclosures:

1.0 INTRODUCTION

2.0 BACKGROUND

SUMMARY

5.0 STATE CONSULTATION

6.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

8.0 REFERENCES

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