ML17334B471

Application for Amend to License DPR-74,requesting Relief from TS Surveillances Until Refueling Outage,Currently Scheduled to Begin on 940806

ML17334B471
Person / Time
Site:	Cook
Issue date:	04/16/1993
From:	Fitzpatrick E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17331A288	List: ML17331A287 ML17334B471
References
AEP:NRC:1181, NUDOCS 9304220175
Download: ML17334B471 (207)

Text

ACCELERATF/i DOCUii'll 5'l'ISTRLBUTlOY SYSTEM REGULA INFORMATION "DZSTRZBUTZOtOYSTEN (RZDSl ACCESSION NBR:9304220175 DOC.DATE: 93/04/16 NOTARIZED: YES DOCKET I FACIL:50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana M 05000316 AUTH. NAME . AUTHOR AFFILIATION FITZPATRICK,E. Indiana Michigan Power Co. (formerly Indiana !; Michigan Ele RECIP.NAME RECIPIENT AFFILIATION MURLEYFT ED Document Control Branch (Document Control Desk)

SUBJECT:

Application for amend to License -DPR-74,requesting relief from TS surveillances until refueling outage, currently scheduled to begin on 940806.

DISTRIBUTION CODE: AOOID TITLE: OR COPIES RECEIVED:LTR Submittal: General Distribution J ENCL g SIZE: e~j NOTES:

J f p~.

RECIPIENT

~p k COPIES 7~

RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 LA 1 1 PD3-1 PD 1 1 DEAN,N 2 2.,

INTERNAL: NRR/DE/EELB 1 1 NRR/DORS/OTSB 1 1 NRR/DRCH/HICB 1 1 NRR/DSSA/SCSB 1 0 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OC 1 0 OGC/HDS2 1 0 E 01 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 NOTE TO ALL"R1DS" REClPlEHTS:

DOCU).'EYT COii'TR(>! D!:.~V, PLEASE HELP US fO REDUCE WASTE! CONl ACl Tl lE ROOM P1.37 (EXT. 504-20o5) TO EL!M IYATE YOUR HA." lE FROl i D]STRiliUTlON LiSTS FOR DOCU!4EY l'S YOU DOi"l'T )REED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 16 ENCL 13

l Indiana Michigan Power Company P.O. Box 16631 Columbus~".iH 43216 IIHtMSlHSl svaamramaue POUFFE Donald C. Cook Nuclear Plant Unit 2 AEP:NRC:1181 Docket No. 50-316 License No. DPR-74 SURVEILLANCE INTERVAL EXTENSION FOR UNIT 2 CYCLE 9 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555 Attnr T. E. Murley April 16, 1993

Dear Dr. Murley:

This letter constitutes an application for amendment to the Technical Specifications (T/Ss) for the Donald C. Cook Nuclear Plant Unit 2. Specifically, we request an extension for certain surveillances which the T/Ss require to be performed beginning January 2, 1994. We are requesting relief from these T/S requirements until the Unit 2 refueling outage, which is currently scheduled to begin August 6, 1994. Many of these surveillances can only be performed during shutdown; therefore, to avoid unnecessary shutdown of the plant, we ask that your review of this request be performed on an expedited basis and that you respond to us by December 1, 1993.

A description of the proposed changes and our analysis concerning significant hazards considerations are contained in Attachment 1 to this letter. The proposed, revised T/S pages are contained in Attachment 2. The existing T/S pages, marked to reflect the proposed changes, are contained in Attachment 3.

All of the requested surveillance extensions are associated with surveillances normally performed during refueling outages. The current cycle will be lengthened approximately five months due to a planned power reduction to approximately 70% of rated thermal power, which is to begin in May 1993 and remain in effect until the end of the. cycle. The purpose of extending the cycle is to separate the refueling outages between Unit 1 '(which is scheduled for refueling in January, 1994) and Unit 2.

9304220i75 9304ib g0 PDR ADOCK 050003ib l

Dr. T. E. Murley AEP:NRC:1181 During the last refueling outage, which was extended approximately six months due to turbine-generator rotor vibrations, an effort was made to re-perform as many surveillances as possible. A significant number of T/S surveillances (approximately 70) were re-performed, reducing the number of surveillances for which we are requesting extensions. However, our efforts were constrained because Unit 1 was in a refueling outage at the same time.

Some of the Technical Specification pages affected by this submittal are pages for which changes are pending due to prior submittals. The proposed changes contained in this submittal are in addition to our previous requests and do not supersede them.

The pages included in this category and the applicable prior submittals which have not yet been processed are provided in the table below:

Letter Number Date T S Pa e Numbers AEP:NRC:1131A April 19, 1991 3/4 4-33 AEPsNRC:1178 September 24, 1992 3/4 6-14 AEP:NRC:1143 May 1, 1992 3/4 7-20 6 3/4 7-40 In accordance with 10 CFR 50.92(c), our evaluation of the changes indicates no significant hazards because these changes do not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibil'ity of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in the margin of safety.

These proposed changes have been reviewed and approved by the Plant Nuclear Safety Review Committee and by the Nuclear Safety and Design Review Committee.

In compliance with the requirements of 10 CFR 50.91(b)(1), copies of this letter and its attachments have been transmitted to the Michigan Public Service Commission and to the Michigan Department of Public Health.

Dr. T. E. Murley AEP:NRC:1181 This letter is submitted pursuant to 10 CFR 50.54(f) and, as such, an oath statement is enclosed.

Sincerely, E. E. Fit patrick Vice President dr Attachments cc! A. A. Blind Bridgman J. R. Padgett G. Charnoff A. B. Davis - Region III NRC Resident Inspector Bridgman NFEM Section Chief

~ '

Dr. T. E. Murley AEPsNRC:1181 bc: S. J. Brewer D. H. Malin/K. J. Toth M. L. Horvath Bridgman J. B. Shinnock W. G. Smith, Jr.

W. M. Dean, NRC Washington, D, C.

AEP:NRC!1181 DC-N-6015.1

~ C ~

STATE OF OHIO COUNTY OF FRANKLIN E. E. Fitzpatrick, being duly sworn, deposes and says that he is the Vice President of licensee Indiana Michigan Power Company, that he has read the forgoing Technical Specifications Changes Proposed in Letter AEPsNRC:1181 and knows the contents thereof; and that said contents are true to the best of his knowledge and belief.

Subscribed and sworn to before me this day of 19 NOTARY PUBLIC RITA D. HILL NOTARY PUBLIC. STATE OF OHIO

005 ~EYER TENSILE WOL WOL WOL TENSILE CNARPY CNARPY I.OCR CHARPY CNARPY CHARPY CHARPY CHARPY I CHARPY CHARPY CHARPY CNARPY NT-12 IRf-l2 Nf.72 NT-72 IOf-70 NT 70 0 68 HT-68 IOf-66 NT-66 W Cfl HT 69 fN-62 HT-62 HH-72 NL-48 NH-70 -f16 KH-66 HL~2 HK Hfl-62 CAPSULE U

NT ll IOI-71 NT-71 10f 69 HT 69 M HT-67 10f-65 NT-6 10f-63 HT-63 Of-61 NT-61 NK-71 NL&7 NII-69 -f15 HK-67 KL&3 KH-65 KL-91 III 63 HH-61 SPECIMEN CODE: MT - PLATE C552I-2 (TRANSVERSE)

KL - PLATF. C552I-2 (LONGITUDINAL)

MW - WELD METAL MH - WELD HEAT AFFECTED ZONE Figure 4-2 Capsule U Diagram Showing Q Location of Specimens, gyERYURE Thermal Monitors and CARO Dosimeters

>iso available 0~

Aperture CaE'd

SECTION 5.0 TESTING OF SPECIHENS FROM CAPSULE U 5.1 Overview The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center hot cell with consultation by Westinghouse Power Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H~ ~, ASTH Specification E185-82~ ~, and Westinghouse Remote Hetallographic Facility (RHF) Procedure RHF 8402, Revision 2 as modified by RHF Procedures 8102, Revision 1'and 8103, Revision 1.

Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-8512~ ~. No discrepancies were found.

Examination of the two low-melting point 579'F (304'C) and 590'F (310'C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579'F (304'C).

The Charpy impact tests were performed per ASTH Specification E23-88~ ~ and RHF Procedure 8103, Revision 1 on a Tinius-Olsen Hodel 74, 358J machine. The tup (striker) of the Charpy machine is instrumented with a GRC 830I instrumentation system, feeding information into an IBH XT computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve shown in Appendix A, the load of general yielding (PGy), the time to general yielding (tGY), the maximum load (PH), and the time to maximum load (tH) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (PF), and the load at which fast fracture terminated is identified as the arrest load (PA).

5-1

The energy at maximum load (EM),was determined by comparing the energy-time record and the load-time record. The energy at maximum load is roughly equivalent to the energy required to initiate a crack in the specimen.

~

• Therefore, the propagation energy for the crack (E ') is the difference .

between the total energy to fracture (ED) and 'the energy at maximum load.

The yield stress (ay) was calculated from the three-point bend formula having the following expression:

ay = PG~

• [L/[B*(W-a) *C])

where L distance between the specimen supports in the impact testing machine; B the width of the specimen measured parallel to the notch; W height of the specimen, measured perpendicularly to the notch; a - notch depth. The constant C is dependent on the notch flank angle (P), notch root radius (p), and the type of loading (i.e., pure bending or three-point bending).

In three-point bending a Charpy specimen in which P 45'nd p-0:010", Equation 1 is valid with C 1.21. Therefore (for L - 4W),

oy = PGy * [L/[B*(W-a) *1.21]) - [3.3PGyW]/[B(W-a) ] (2)

For the Charpy specimens, B = 0.394 in., W - 0.394 in., and a - 0.079 in.

Equation 2 then reduces to:

Gy = 33.3 x PGy (3) where oy is in units of psi and PGy is in units of lbs. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.

Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTH Specification A370-89[ ].

The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

5-2

Tension tests were performed on a 20,000-pound Instron Model 1115, split-console test machine, per ASTM Specification E8-89b~ ~ and E21-79 (1988)~ ~, and RMF Procedure 8102, Revision I. All pull rods, grips, and pins were made of Inconel 718 hardened to HRC45. The upper pull rod was connected thr ough a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.

Extension measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-85~II~.

Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air.

Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature.

Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. In the test configuration, with a slight load on the spec'imen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range of room temperature to 550'F (288'C). The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures. Experiments indicated that this method is accurate to +2'F.

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

5-3

5.2 Char -Notch Im act Test Results The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule U, which was irradiated to 1.58 x 10 n/cm (E > 1.0 MeV), are presented in Tables 5-1 through 5-4 and are compared with unirradiated results~ ~ as shown in Figures 5-1 through 5-4. The transition temperature 'increases and upper shelf energy decreases for the Capsule U materials are summarized in Table 5-5.

Irradiation of the reactor vessel intermediate shell plate C5521-2 Charpy specimens oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation) to 1.58 x 10 n/cm (E > 1.0 MeV) at 550'F (Figure 5-1) resulted in a 30 ft-lb transition temperature increase of 95'F and in a 50 ft-lb transition temperature increase of 110'F. This resulted in a 30 ft-lb transition temperature of 120'F and a 50 ft-lb transition temperature of 165'F (longitudinal orientation).

The average Upper Shelf Energy (USE) of the intermediate shell plate C5521-2 Charpy specimens (longitudinal orientation) resulted in a energy decrease of 16 ft-lb after irradiation to 1.58 x 10 n/cm (E > 1.0 MeV) at 550'F.

This results in an average USE of 111 ft-lb (Figure 5-1).

Irradiation of the reactor vessel intermediate shell plate C5521-2 Charpy specimens oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation) to 1.58 x 10 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-2) resulted in a 30 ft-lb transition temperature increase of 130'F and in a 50 ft-lb transition temperature increase of 135'F. This resulted in a 30 ft-lb transition temperature of 160'F and a 50 ft-lb transition temperature of 205'F (transverse orientation).

5-4

The average USE of the intermediate shell plate C5521-2 Charpy specimens (transverse orientation) resulted in an energy decrease of 14 ft-lb after irradiation to 1.58 x 10 n/cm (E > 1.0 MeV) at 550'F. This resulted in an average USE of 72 ft-lb (Figure 5-2).

Irradiation of the reactor vessel core region weld metal Charpy specimens to 1.58 x 10 n/cm (E > 1.0 MeV) at 550'F (Figure 5-3) resulted in a 75'F increase in 30 ft-lb transition temperature and a 50 ft-lb transition temperature increase of 40'F. This resulted in a 30 ft-lb transition temperature of 85'F and the 50 ft-lb transition temperature of 110'F.

The average USE of the reactor vessel core region weld metal resulted in an energy decrease of 6 ft-lb after irradiation to 1.58 x 10 n/cm (E > 1.0 MeV) at 550'F. This resulted in an average USE of 71 ft-lb (Figure 5-3).

Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens to 1.58 x 10 n/cm (E > 1.0 MeV) at 550'F (Figure 5-4) resulted in a 30 ft-lb transition temperature increase of 105'F and a 50 ft-lb transition temperature increase of 110'F. This resulted in a 30 ft-lb transition temperature of 45'F and the 50 ft-lb transition temperature of 80'F.

The average USE of the reactor vessel weld HAZ metal experienced an energy decrease of 33 ft-lb after i} radiation to 1.58 x 10 n/cm (E > 1.0 MeV) at 550'F. This resulted in an average USE of 82 ft-lb (Figure 5-4).

The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and show an increasingly ductile or tougher appearance with increasing test temperature.

A comparison of the 30 ft-lb transition temperature increases and upper shelf energy decreases for the various D. C. Cook Unit 2 surveillance materials with predicted values using the methods of NRC Regulatory Guide 1.99, Revision 2~ ~ is presented in Table 5-6. Comparison of the 30 ft-lb transition 5-5

temperature increase for'he intermediate shell plate C5521-2 (transverse orientation) is 33'F greater than the Regulatory Guide prediction.

However, the NRC Regulatory Guide 1.99, Revision 2 requires a 2 sigma allowance of 34'F be added to the predicted reference transition temperature to obtain a conservative upper bound value. Thus, the reference transition temperature increase is bounded by the 2 sigma allowance for shift prediction.

This comparison indicates that the transition temperature increases and the upper shelf energy decreases of the Intermediate Shell Plate C5521-2 (longitudinal orientation) and surveillance weld resulting from irradiation to 1.58 x 10 n/cm (E > 1.0 MeV) are less than the Regulatory Guide predictions. This comparison also indicates that the upper shelf energy decrease of the intermediate shell plate C5521-2 (transverse orientation) resulting from irradiation to 1.58 x 10 n/cm (E > 1.0 HeV) is less than the Regulatory Guide prediction.

The end of license (32 EFPY) RTNDT values for all the D. C. Cook Unit 2 beltline region materials are shown in Table 5-7. These values were predicted using Regulatory Guide 1.99, Revision 2 methodology and are projected to be within the Regulatory limits.

Photographs of the charpy and tensile specimens before testing are shown in Appendix B.

5.3 Tension Test Results The results of the tension tests performed on the various materials contained in Capsule U irradiated to 1.58 x 10 n/cm (E > 1.0 HeV) are presented in Table 5-8 and are compared with unirradiated results~ ~ as shown in Figures 5-9 and 5-10.

5-6

r The results of the tension tests performed on the intermediate shell plate C5521-2 (transverse orientation) indicated that irradiation to 1.58 x 10 n/cm (E > 1.0 MeV) at 550'F caused less than a 18 ksi increase in the 0.2 percent offset yield strength and less than a 16 ksi increase in the ultimate tensile strength when compared to unirradiated data~ ~ (Figure 5-9).

The results of the tension tests performed on the reactor vessel core region weld metal indicated that irradiation to 1.58 x 10 n/cm (E > 1.0 MeV) at 550'F caused less than a 9 ksi increase in the 0.2 percent offset yield strength and less than a 8 ksi increase in the ultimate tensile strength when compared to unirradiated data~ ~ (Figure 5-10).

The fractured tension specimens for the Intermediate Shell Plate C5521-2 material are shown in Figure 5-11, while the fractured specimens for the weld metal are shown in Figure 5-12.

The engineering stress-strain curves for the tension tests are shown in Figures 5-13 and 5-14. 0 5.4 Wed e 0 enin Loadin S ecimens Per the surveillance capsule testing program with the Indiana Michigan Power Company, the WOL specimens will not be tested and will be stored at the Westinghouse Science and Technology Center.

5-7

TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE D. C. COOK UNIT 2 INTERMEDIATE SHELL PLATE C5521-2 IRRADIATED AT 550 F, FLUENCE 1.58 x 10 ngcm2 (E > 1.0 Mey)

Temperature Impact Energy Lateral Expansion Shear

~Sem ie No. ~F ~C ~ft-ib ~J ~mile ~mm Lon 'tudinal Orientation ML45 75 ( 24) 14 ( 19) 10 (0.25) 10 ML48 100 ( 38) 27 ( 37) 23 {0.58) 20 ML42 125 ( 52) 35 ( 47) 27 '2 (0.69) 30 ML44 175 ( 79) 62 ( 84) (1.07) 50 ML41 200 ( 93) 55 { 75) 42 {1.07) 50 ML43 225 (107). 103 (140) 50 (1.27) 90 ML46 250 (121) 114 (155) 79 (2.01) 100 ML47 300 (149) 115 (156) 82 (2.08) 100 Transverse Orientation MT62 25 ( -4) 5 (7) 2 (0.05) 5 MT61 50 ( 10) 14 ( 19) 12 (0.30) 10 MT66 75 ( 24) 17 ( 23) 10 (0.25) 15 MT71 125 ( 52) 19 ( 26) 16 (0.41) 30 MT64 150 ( 66) 25 ( 34) 21 (0.53) 35 MT72 175 ( 79) 33 ( 45) 28 (0.71) 40 MT70 200 ( 93) 37 ( 50) 30 (0.76) 45 MT69 215 (102) 39 ( 53) 33 (0.84) 65 MT63 225 (107) 63 ( 85) 52 (1.32) 95 MT68 250 (121) 67 ( 91) 54 (1.37) 100 MT67 275 (135) 72 ( 98) 50 (1,27) 100 MT65 300 (149) 77 (104) 58 (1,47) 100 5-8

TABLE 5-2 CHARPY V=NOTCH IMPACT DATA FOR THE D. C. COOK UNIT 2 REACTOR VESSEL WELD METAL AND HAZ METAL IRRADIATED AT 550'F, FLUENCE 1.58 x 10 n/cm (E > 1.0 MeV)

Temperature Impact Energy Lateral Expansion Shear

~Sam le No. ('F) ('C) (R-1b) ( J) (mils) (mm) (~c)

Weld Metal MW70 -10 (-23) 23 ( 31) 16 (0.41) 25 MW64 0 (-18) 29 ( 39) 26 (0.66) 35 MW71 25 (-4) 17 ( 23) 14 (0.36) 30 iiIW68 50 ( 10) 21 ( 28) 21 (0.53) 40 MKV63 75 ( 24) 26 ( 35) 20 (0.51) 60 MAV61 100 ( 38) 42 ( 57) 37 (0.94) 80 MW72 125 ( 52) 60 ( 81) 50 (1.27) 85 MW65 150 ( 66) 71 ( 96) 57 (1.45) 100 MW66 175 ( 79) 47 ( 64) 39 (0.99) 85 MW62 185 ( 85) 62 ( 84) 49 (1.24) 100 MW67 200 ( 93) 78 . (106) 62 (1.57) 100 MW69 250 (121) 74 (100) 62 (1.57) 100 HAZ Metal MH67 -25 (-32) 4 (5) 2 (0.05) 10 MH63 25 (- 4) 21 ( 28) 10 (0.25) 35 MH71 50 ( 10) 48 ( 65) 30 (0.76) 55 MH69 65 ( 18) 18 ( 24) 19 (0.48) 30 MH72 75 ( 24) 64 ( 87) 44 (1.12) 90 MH70 100 ( 38) 39 ( 53) 32 (0.81) 50 MH62 125 ( 52) 118 (160) 71 (1.80) 100 MH64 150 ( 66) 72 ( 98) 55 (1.40) .95 MH61 175 ( 79) 83 (113) 61 (1.55) 100 MH65 200 ( 93) 103 (140) 71 (1.80) 100 MH68 225 (107) 74 (100) 53 (1.35) 100 MH66 250 (121) 69 ( 94) 66 (1.68) 100 5-9

TABLE 5-3 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE D. C. COOK UNIT 2 INTERMEDIATE SHELL PLATE C5521-2 IRRADIATED AT 550'F, FLUENCE 1.58 x 10 n/cm (E > 1.0 MeV)

Normalized Ener ies Test Charpy Charpy Maximum Prop. Yield Time Maximum Time to Fracture Arrest Yield Floe Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress Sample Ncehes Temp Energy

~s'ft-1h ft-lb in ~lbs ~csee ~lhs ~csee ~lhs ~lhs ~ks i ~ks i Lon itudinal Orientation 75 14 113 64 48 3722 0.14 3929 0:20 3929 638 124 127 ML45 27 217 167 50 3670 0.14 4600 0.38 4600 858 122 137 ML48 100 35 282 124 157 3945 0.29 4258 0.38 4258 384 131 136 ML42 125 175 62 499 241 258 3467 0.14 4735 0.53 4563 792 115 136 ML44 443 122 321 3731 0.28 4139 0.38 3887 1340 124 131 ML41 200 55 829 308 521 3262 0.14 4507 0.54 *

• 108 129 ML43 225 103 114 918 234 684 3265 0.14 4546 0.54

• 108 130 ML46 250 115 926 147 779 3126 0.14 4071 0.38 104 120 ML47 300 Transverse Orientation 40 12 28 1107 0.10 1319 0. 14 1319 55 37 40 MT62 25 5 14 113 77 35 3600 0.14 4154 0.22 4154 126 120 129 MT61 50 137 88 49 3825 0.14 4273 0.25 4273 513 127 134 MT66 75 17 153 48 105 2969 0.21 3187 0.24 3187 202 99 102 MT71 125 19 201 124 77 3381 0. 14 4194 0.32 4194 1577 112 126 MT64 150 25 266 156 110 3438 0.13 4273 0.38 4273 2083 114 128 MT72 175 33 37 298 128 170 3681 0.28 4096 0.39 4096 2151 122 129 MT70 200 314 153 161 3289 0.14 4158 0.38 4158 2533 109 124 MT69 215 39 .

507 236 272 3359 0.15 4457 0.54 4225 3236 112 130 MT63 225 63 540 221 319 3288 0.14 4212 0.54 109 125 MT68 250 67 .

580 222 358 3051 0.16 4249 0.54

• 101 121 MT67 275 72 77 620 223 397 3153 0.14 4251 0.54

• 105 123 MT 65 300

• Fully ductile fracture. No arrest load.

0

TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE D. C. COOK UNIT 2 MELD METAL AND HEAT-AFFECTED-ZONE (HAZ) METAL, IRRADIATED AT 550'F, FLUENCE 1.58 x 10 n/cm (E > 1.0 MeV)

Normalized Ener ies Test Charpy Charpy Maximum Prop Yield Time Maximum Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Load Stress Stress Number ~r ~ft-1b ft-lb in ~lbs ~mesc ~lbs ~lbs ~ksi ~ksi Weld Metal MW70 -10 23 185 133 52 3898 0. 14 4576 0.31 4576 778 129 141 MW64 0 29 234 106 127 3891 0. 14 4482 0.28 4482 467 129 139 MW71 25 17 137 83 54 3776 0.14 4105 0.23 4105 978 125 131 MW68 50 21 169 85 84 3847 0.26 4021 0.30 4021 941 128 131 MW63 75 26 209 131 79 3655 0.14 4430 0.32 4430 1618 121 134 MW61 100 42 338 151 187 3712 0.26 4359 0.42 4359 463 123 134 MW72 125 60 483 239 244 3323 0 '4 4581 0.54 4581 1252 110 131 MW65 150 71 572 235 337 3511 0.14 4535 0.54 117 134 MW66 175 47 378 221 158 3397 0 '4 4404 0.50 4404 810 113 130 MW62 185 62 499 231 269 3280 0.13 4232 0.54 109 125 MW67 200 78 628 250 378 3754 0.28 4440 0.62 125 136 MW69 250 74 596 231 365 3283 0.14 4273 0.54 109 125 HAZ Metal MH67 -25 4 32 15 17 964 0.10 1133 0. 17 1133 146 32 35 MH63 25 21 169 47 122 3327 0.21 3486 0.23 3486 858 111 113 MH71 50 48 387 166 221 3797 0.14 4557 0.38 4420 2126 126 139 MH69 65 18 145 58 87 3653 0.14 3834 0.20 3834 971 121 124 MH72 75 64 515 162 354 3910 0.14 4530 0.37 4500 3802 130 140 MH70 100 39 314 114 200 2970 0.21 3618 0.38 3618 954 99 109 MH62 125 118 950 301 649 3352 0.16 4355 0.53

• 111 128 MH64 150 72 580 240 340 3991 0.22 4756 0.54 4756 4056 133 145 MH61 175 83 668 243 425 3622 0.14 4662 0.54

• 120 138 MH65 200 103 829 253 576 3642 0.15 4799 0.54

• 121 140 MH68 225 74 596 187 409 3066 0.17 3981 0.50 *

• 102 117 MH66 250 69 556 228 328 3330 0.14 4287 0.52

• 111 126 5-11

• Fully ductile fracture. No arrest load.

TA8LE 5-5 EFFECT OF 550 F IRRADIATION TO 1.58 x 10 n/cm (E > 1.0 HeV)

ON THE NOTCH TOUGHNESS PROPERTIES OF THE D. C. COOK UNIT 2 REACTOR VESSEL SURVEILLANCE HATERIALS Average 30 ft-lb( 'I)

Average 35 mil Average 50 ft-lb( '1)

Average Energy Transition Lateral Expansion Transition Absorption at Temperature ( F) Temperature ( F) Temperature ( F) Full Shear (ft-lb)

Haterial Unirradiated Irradiated AT Unirradiated Irradiated 4T Unirradiated Irradiated AT Unirradiated Irradiated d(ft-lb)

P late C5521-2 25 120 95 50 150 100 55 165 110 127 111 - 16 (Longitudinal)

P late C5521-2 30 160 130 70 190 120 70 205 135 86 72 - 14 (Transverse)

Meld Hetal 10 85 75 50 100 50 70 110 40 77 71 - 6 HAZ Hetal - 60 45 105 - 40 85 125 - 30 80 110 115 82 - 33 (1) "AVERAGE" is defined as the value read from the curve fitted through the data points of the Charpy tests (Figures 5-1 through 5-4).

y5-12

TABLE 5-6 COMPARISON OF THE D. C. COOK UNIT 2 SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS 30 ft-lb Transition Tem . Shift U er Shelf Ener Decrease Fluence Predicted (a) Measured Predicted (a) Measured Material Capsule 10 n/cm ('F) ('F) (>) (>)

Plate C5521-2 0.264 55 55 16 13 (Longitudinal) 0.683 77 90 20 19 1.06 88 95 22 19 1.58 97 95 24 13 Plate C5521-2 0.264 55 80 16 14 (Transverse) 0.683 77 100 20 20 1.06 88 103 22 19 1.58 97 130 24 16 Weld Metal 0.264 45 40 15 4 0.683 63 50 18 9 1.06 72 70 21 10 1.58 80 75 23 8 HAZ Metal 0.264 50 13 0.683 70 21 1.06 72 12 1.58 105 29 (a) Regulatory Guide 1.99, Revision 2 5-13

TABLE 5-7 PROJECTED END OF LICENSE (32 EFPY) RTNDT AND UPPER SHELF ENERGY VALUES FOR D. C. COOK UNIT 2 BELTLINE REGION MATERIALS PER REGULATORY GUIDE 1.99, REVISION 2 MATERIAL DESCRIPTION RTNDT F UPPER SHELF ENERGY ft-lbs Intermediate Shell Plate, C5556-2 216 76 Intermediate Shell Plate, C5521-2 171 (171) 66 Lower Shell Plate, C5540-2 100 47 Lower Shell Plate, C5592-1 128 52 Intermediate Shell Longitudinal Welds 90 (56) 62 (located at 10'zimuth)

Lower Shell Longitudinal Welds 84 (50) 62 (located at 90'zimuth)

Circumferential Meld 102 (67) 62 Note numbers in () are based upon surveillance capsule data.

5-14

TABLE 5-8 TENSILE PROPERTIES FOR THE D. C. COOK UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS IRRADIATED AT 550'F TO 1.58 x 10 n/ctn (E ) 1.0 MeV)

Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp. Strength Strength Load Stress Strength Elongation Elongation in Area Material Number ~F ~ksi ~ksi. ~ki ~ksi. ~ksi.

Plate MTll 150 79.5 97.4 3.60 172.0 73.3 10.5 20.1 57 Plate MT12 550 70.3 93.7 3.65 131.8 74.4 9.6 16.7 44 Weld MW11 '25 82.5 96.8 3.20 173. 9 65.2 9.0 19.5 63 Weld MW12 550 75.9 92.7 3.50 190.2 71.3 8.1 17.1 58 5-15

(~C)

-150 -100 -50 0 50 100 150 200 100 80 60 0 3

c% 40 100 80 0 0

60 1.5

>C 40 I0tl'F ~ 1.0 20 0.5 0 0 0 UNIRRADIATED

~ IRRADIATED <550'F), FLUENCE I58 x IO n/cd 160 200 140 0 160 120 ~ ~

100 120-80 0 5 60 No'F ~ 80 40 40 20

-200 -100 0 100 200 300 400 TEMPERATURE ('f)

Figure 5-1. Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Longitudinal Orientation) 5-16

(~C)

-150 -100 -50 0 50 100 150 200 100 80 0

60 0 40 20 0

100 2.5 80 2.0 60'C 1.5 40 Ic9'F

~ t 1.0 20 0.5 0

0 NIIRRADIATED

~ IRRADIATED (550'F), FLUEHCE l58 x IO n/cn~

120 160 100 120 80 I

60 80 l35'F 40 0 13I'F 40 20

~~

-200 -100 0 100 200 300 400 500 TEMPLRATURL ('f)

Figure 5-2. Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) 5-17

( C)

-150 -100 -50 0 50 100 150 200 100

~ ~

80 60 0

4p 0 ~0 20 0

100 2.5 80 60 1.5

>C 50F 0 ~

40 1.0 20 0.5 0

0 UNIRRADIATED

~ IRRADIATED (550'F), FLUENCE !50 x IO n/cn 120 160 100 120 80 I 0 ~

60 0 80 i0'F 40 0 0 T5F 4p

-200 -100 0 100 200 300 400 500 TEMPERATURE ('F')

Figure 5-3. Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel Surveillance Weld Metal 5-18

('C)

-150 -100 -50 0 50 100 150 200 100 0 0 80 8

K 60

~ 40 100 80 0 0 60 2 1.5 OC o S ~

4 ~

40 O l25'I' 1.0

~

20 0 0,5 0

0 0 UNIRRADIATED

~ IRRADIATED (550'F), ELUEhEE I$8 x IO n/cn~

160 200 140 0

0 120 160 100

~

0 0 120 80

~

00 0~

~

60 ll9'I' 80 LJ 40, 40 20 o

-200 -100 0 100 200 300 400 TEMPERATURE ( f)

Figure 5-4. Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel Weld Heat-Affected-Zone Netal 5-19

ML45 ML48 ML42 ML41 ML43 ML46 ML47 Figure 5-5. Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Longitudinal Orientation) 5-20

MT62 MT61 MT66 MT71 tf I

• r

'i MT64 MT72 MT69 i

~ 4~AM+%38!!!!!!h

&NWBSNWWsst<. q I I t qw (!

MT63 MT68 MT67 MT65 Figure 5-6. Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) 5-21

MW70 MW64 MW71 MW68 MW63 MW61 MW72 MW65 S<

h I

-ca) i

)

MW66 MW62 MW67 MW69 Figure 5-7. Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel Surveillance Weld Netal 5-22

MH67 MH63 MH71 MH69 MH72 MH62 MH64 tycho l

a MH61 MH65 MH68 MH66 Figure 5-8. Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal 5-23

('C) 50 100 150 200 250 300 120 800 110 700 100 ULTIMATE TEHSILE STRENGTH 600 g 80

'00 cn 70 02% YIELD STRENGTH 60 400 50 300 40 Q Q UNIRRADIATED 4 ~ IRRADIATED AT 550'F, FLUENCE L50 xIO g'cn 80 70 REDUCTION IN AREA 60 9 50

- 40 g 30 TOTAL ELONGATION 20 10 UNIFORH ONGATION 100 200 300 400 500 TEMPERATURE ('F)

FigUre 5-9. Tensile Properties for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) 5-24

('C) 50 100 150 200 250 300 120 800 110 700 100 Il.TIMATE TENSILE STRENGTH Q 90 600 g 80 2+ 500 cn 70 D2% YIELD STRENGTH 400 300 40 b, 0 UNIRRADIATED 4 ~ IRRADIATED AT 550 F, FLUENCE L58 xI019n/cn2 80 REDUCTION IN AREA 60

~ 50 UNIFORM ELONGATIII 10 TOTAL E TION 100 200 300 400 500 TEMPERATURE ('F)

Figure 5-10. Tensile Properties for D. C. Cook Unit 2 Reactor Vessel Surveillance Weld Hetal 5-25

8 8 9 Specimen MT11 150'F Specimen MT12 550'F Figure 5-11. Fractured Tensile Specimens from D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) 5-26

Specimen MW11 125'F Specimen MW12 550'F Figure 5-12. Fractured Tensile Specimens from D. C. Cook Unit 2 Reactor Vessel Surveillance Weld Metal 5-27

100.00 90.00 80.00 70.00 6o.oo 50.00 h 40.00 30.00 MT11 20.00 150 F 10.00 0.00 0.0 0.10 0.20

.STRAIN, IN/IN 100.00 90.00 80.00 70.00 60.00 (0

50.00 IM 40.00 30.00 MT12 20.00 10.00 550 F 0.00 0.0 0.04 0.08 0.12 0.16 STRAIN, IN/IN Figure 5-13. Engineering Stress-Strain Curves for Plate C5521-2 Tensile Specimens NTll and HT12 (Transverse Orientation) 5-28

100.00 90.00 80.00 70.00 60.00 5000 IM 40.00 30.00 20.00 MW11 10.00 125 F 0.00 0.0 0.04 0.08 0.12 0.16 0.20 STRAIN, IN/IN 100.00 90.00 80.00 70.00 6o.oo 50.00 le 40.00 30.00 MW12 20.00 10.00 550 F 0.00 0.0 0.04 0.08 0.12 8.16 0.20 STRAIN, IN/IN Figure 5-14. Engineering Stress-Strain Curves for Weld Metal Tensile Specimens NWll and MW12 5-29

SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 Introduction Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known.

Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from analysis.

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.

Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTH Standard Practice E853, "Analysis and Interpretation of Light Water Reactor 6-1

Surveillance Results," recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference.

The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, "Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide 1.99, "Radiation Damage to Reactor Vessel Materials."

This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance Capsule U. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0. 1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history.

The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used =to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the projected exposure of the pressure vessel are provided.

6.2 Discrete Ordinates Anal sis A plan view of the reactor geometry at the core midplane is shown in Figure 4-1. Eight irradiation capsules attached to the thermal shield are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 4.0 , 40.0', 140.0 , 176.0',

184.0 , 220.0 , 320.0 , and 356.0 relative to the core cardinal axes as shown in Figure 4-1.

A plan view of a surveillance capsule holder attached to the thermal shield is shown in Figure 6-1. The stainless steel specimen containers are 1.0 inch square and approximately 38 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 3 feet of the 12-foot high reactor core.

6-2

From a neutron transport standpoint, the surveillance capsule structures are significant. They have a marked effect on both the distribution of neutron flux and the neutron energy spectrum in the water annulus between the thermal shield and the reactor vessel. In order to properly determine the neutron environment at the test'pecimen locations, the capsules themselves must be included in the analytical model.

In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters (P(E > 1.0 Mev), P(E > =0. I Mev), and dpa) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e.,

dpa/P(E > 1.0 MeV), within the pressure vessel geometry. The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e.,

the I/4T, I/2T, and 3/4T locations.

The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each cycle of irradiation; and established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core; but, also accounted for the effects of varying neutron yield 6-3

per fission and fission spectrum introduced by the build-in of plutonium as the burnup of individual fuel assemblies increased.

The absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra and radial distribution information from the forward calculation provided the means to:

t

l. Evaluate neutron dosimetry obtained from surveillance capsule locations.

2. Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall.

3. Enable a direct comparison of analytical prediction with measurement.

4. Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.

The forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R, 8 geometry using the DOT two-dimensional discrete ordinates code~ 5~ and the SAILOR cross-section library~ ~. The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light water reactor applications. In these analyses anisotropic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an SB order of angular quadrature.

The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 4-loop plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal +2u 6-4

level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.

All adjoint analyses were also carried out using an S8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library.

Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as the geometric center of each surveillance capsule. Again, these calculations were run in R, 8 geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, P (E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as:

R (r, 8) - fr f8 fE )(r, 8, E) S (r, 8, E) r dr d8 dE where: R (r, 8) = P (E > 1.0 MeV) at radius r and azimuthal angle 8 Adjoint importance function at radius, r, azimuthal angle 8, and neutron source energy E.

S (r, 8, E) - Neutron source strength at core location r, 8 and energy E.

Although the adjoint importance functions used in the analysis were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading patterns significantly impact -the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for,a given location the ratio of dpa/P (E > 1.0 MeV) is insensitive to changing core source distributions. In the application of these adjoint importance functions to the D C Cook Unit 2 reactor, therefore, the iron displacement rates (dpa) and the neutron flux (E > 0. 1 MeV) were computed on a cycle specific basis by using dpa/It) (E.> 1.0 MeV) and P (E > 0. 1 MeV)/P (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific Il') (E > 1.0 MeV) solutions from the individual adjoint evaluations.

6-5

The reactor core power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design reports for the first eight operating cycles of D C Cook Unit 2[17 through 19]

Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation period and provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall.

In Table 6-1, the calculated exposure parameters [P (E > 1.0 MeV),

4(E > 0. 1 MeV), and dpa] are given at the geometric center of the two symmetric surveillance capsule positions for both the design basis and the plant specific core power distributions. The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis. The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared. Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the Cycles 1 through 8 plant specific power distributions. It is important to note that the data for the vessel inner radius were taken at the clad/base metal interface; and,'hus, represent the maximum exposure levels of the vessel wall itself.

Radial gradient information for neutron flux (E > 1.0 MeV),

neutron flux (E > 0. 1 MeV), and iron atom displacement rate is given in Tables 6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5.

6-6

For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the is given by: 45'zimuth PI/4T(45') .i $ (220.27, 45') F (225.75, 45')

where: p~/4T(45') Projected neutron flux at the 1/4T position on the 45'zimuth 4 (220.27,45') Projected or calculated neutron flux at the vessel inner radius on the 45'zimuth.

F (225.75, 45') = Relative radial distribution function from Table 6-3.

Similar expressions apply for exposure parameters in terms of P (E > O.l MeV) and dpa/sec.

6.3 Neutron Dosimetr The passive neutron sensors included in. the D C Cook Unit 2 surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest [III (E > 1.0 Mev), P (E > O.l MeV), dpa].

The relative locations of the neutron sensors within the capsules are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule.

6-7

The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest.

Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known.

In particular, the following variables are of interest:

The specific activity of each monitor .

The operating history of the reactor.

The energy response of the monitor.

The neutron energy spectrum at the monitor location.

The physical characteristics of the monitor.

The specific activity of each of -the neutron monitors was determined using established ASTM procedures [20 through 33]. Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation history of the D C Cook Unit 2 reactor during Cycles I through 8 was obtained from NUREG-0020, "Licensed Operating Reactors Status Summary Report" for the applicable period.

The irradiation history applicable to Capsule U is given in Table 6-7.

Measured and saturated reaction product specific activities as well as measured full power reaction rates are listed in Table 6-8. Reaction rate values were derived using the pertinent data from Tables 6-6 and 6-7.

Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code ~ ~. The 6-8

FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data. The exposure parameters along with associated uncertainties where then obtained from the adjusted spectra.

In the FERRET evaluations, a log-normal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations. In general, the measured values f are linearly related to the flux P by some response matrix A:

f(s, )

Z A (s) (~)

g ig g where i indexes the measured values belonging to a single data set s, g designates the energy group and ~ delineates spectra that may be simultaneously adjusted. For example, R Z 1 g ig relates a set of measured reaction rates R; to a single spectrum P by the multigroup cross section o;g. (In this case, FERRET also adjusts the cross-sections.) The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.

In the FERRET analysis of the dosimetry data, the continuous quantities (i.e.,

fluxes and cross-sections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code ~ ~. This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620-point .spectrum was then easily collapsed 6-9

to the group scheme used in FERRET.

The cross-sections were also collapsed into the 53 energy-group structure using SAND II with calculated spectra (as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF/B-V dosimetry file.

Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section. Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant.

For each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight. In some cases, as for the cross sections, a multigroup covariance matrix is used. More often, a simple parameterized form is used:

2 RN + Rg Rg Mgg Pgg where RN specifies an overall fractional normalization uncertainty (i.e.,

complete correlation) for the corresponding set of values. The fractional uncertainties R specify additional random uncertainties for group g that are correlated with a correlation matrix:

P, 99 (I 8) b',

99

+ 8 exp [~~~]

272 I 2 The first term specifies purely random uncertainties while the second term describes short-range correlations over a range 7 (8 specifies the strength of the latter term).

For the a priori calculated fluxes, a short-range correlation of 7 6 groups was used. This choice implies that neighboring groups are strongly correlated when 8 is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R.E.

Maerker~ ~. Maerker's results are closely duplicated when 7 6. For the integral reaction rate covariances, simple normalization and random uncertai.nties were combined as deduced from experimental uncertainties.

6-10

Results of the FERRET evaluation of the Capsule U dosimetry are given in Table 6-9. The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 1.58 x 10 n/cm2 (E > 1.0 MeV) with an associated e uncertainty of + 8X. Also reported are capsule exposures in terms of fluence (E > 0. 1 MeV) and iron atom displacements (dpa). Summaries of the I

fit of the adjusted spectrum are provided in"-Table 6-10. In general, excellent results were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates. The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure.

A summary of the measured and calculated neutron exposure of Capsule U is presented in Table 6-12. The agreement between calculation and measurement falls within + 8X for all fast neutron exposure parameters listed. The thermal neutron exposure calculated for the exposure period undepredicted the measured value by approximately a factor of two.

Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13. Along with the current (8.65 EFPY) exposure derived from the Capsule U measurements, projections are also provided for an exposure period of 16 EFPY and to end of vessel design life (32 EFPY). In the evaluation of the future exposure of the reactor pressure vessel the exposure rates averaged over the first eight cycles of operation were employed.

In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the D C Cook Unit 2 reactor coolant system, exposure projections to 16 EFPY and 32 EFPY were also evaluated. Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14. In order to access RTNDT vs. fluence 6-11

trend curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations

~ 'I/4/) = 'I (I ( ) (44 (I ( ))

(I/4/) = 'I (I I ) (I (I ( ))

Using this approach results in the dpa equivalent fluence values listed in Table 6-14.

In Table 6-15 updated lead factors are listed for each of the D C Cook Unit 2 surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules.

In order to provide a consistent data base for comparison with measured shift data, the dosimetry sets from previously withdrawn surveillance capsules (X, Y, and T) were re-evaluated using the previously described least squares adjustment methodology along with current reaction cross-sections and nuclear data. The results of those re-evaluations were as follows:

FLUENCE [E > 1.0 MeV]

Capsule X 1.O6 X 10>>

Capsule Y 6.83 X 10 Capsule T 2.64 X lol8 The lo uncertainty associated with each of these fluence evaluations is SX.

6-12

~ 0 0

~ 0 g

~ ~

g

TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVEILLANCE CAPSULE CENTER ItI(E > 1.0MeV) P(E > O.IMev) Iron Displacement Rate

~ncm 2

~sec ~ncm 2~sec d a sec 4.0'0.0'.0'0.0'.0'0.0'ESIGN BASIS 2.82 X 1010 9.05 X 1010 8.15 X 10 3.04 X 1011 4.58 X 10 ll l.ss x lo-'0 CYCLE 1 2.12 X 10 6.68 X 1010 6.13 X 1010 2.24 X 1O" 3.43 X 10 11 1.14 x lo-10 CYCLE 2 2.29 X 10 6.62 x lo'0 6.62 X 1010 2.22 X 1011 3.71 X 10 11 1.13 X 10 10 CYCLE 3 2.21 X 10 5.46 x lo'0 6.39 X 1010 1.83 X 1011 3.58 X 10 11 9.34 X 10-11 CYCLE 4 2.19 X 10 5.47 x lo'0 6.33 X 1010 1.84 X 1O" 3.55 X 10 11 9.35 X 10-11 CYCLE 5 2 24 X 1010 4 93 X 1010 6.47 x lo'0 1.66 X 1O" 3.63 X 10-11 8.43 X 10 11 CYCLE 6 1 93 X 1010 5 19 x lolo 5.58 X 1010 1.74 X IO" 3 13 X 10-11 8.87 X 10 11 CYCLE 7 2.12 x lo'.0 4.7o X 1010 6.13 X 1010 1.58 X 1011 3.43 X 10 11 8.04 X 10 1

CYCLE 8 1.66 X 1010 4.76 x lo'0 4.80 X 1010 1.60 X 1011 2.69 X 10 11 8.14 X 10-11

TABLE 6-2 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD/BASE METAL INTERFACE E > I.OMeV n c ~sec 0.0'5.0'0.0'5.0'ESIGN BASIS 6.43 X 1009 1.36 X 1010 1.72 X 1010 2.60 X 1010 CYCLE 1 6.34 X 1009 1.O1 X 1O'0 1.26 X 10 1.94 X 1010 CYCLE 2 6.74 X 1009 1.03 X 1010 1.24 X 1010 1.92 X 1010 CYCLE 3 6.43 X 1009 9.59 X 10 1.06 X 1010 1.60 X lolo CYCLE 4 6.58 x lo09 1 04 X 1010 1.11 X 1010 1.63 X 1010 CYCLE 5 6.50 X 1009 9.54 X 1009 9.83 X 10 1.46 X 10 CYCLE 6 5 87 x lp09 9.46 X 1009 1.02 X 10 1.53 X 1010 CYCLE 7 6 39 X 10 9.99 X 1009 9.63 X 1009 1.41 X 1010 4.92 X 10 9 lo09 CYCLE 8 7.4o x 9.53 X 10 1.42 X 10 E > 0.1MeV n cm2~sec 0.0'5.0'0.0'5.0'ESIGN BASIS 2.11 X 10 3.41 X 10 4.34 X 10 6.96 X 10 CYCLE 1 1.59 X 1010 2.54 X 1010 3.18 X 10 5.04 X 10 CYCLE 2 1.69 X 1010 59 X lpl0 3 12 X lplp 4 99 X lplp CYCLE 3 1.61 X 10 2.41 X 1010 2.67 X 1010 4.16 X 1010 CYCLE 4 1.65 X 1010 2.61 X 10 2.80 X 10 4.24 X 10 CYCLE 5 1.63 X 1010 2.39 X 1010 2.48 X 1010 3.80 X 1010 CYCLE 6 1.47 X 101 2.37 X 1010 2.57 X 1010 3.98 X 1010 CYCLE 7 1.60 X 10 2 51 X 1010 2 43 X 1010 3 67 X 1010 CYCLE 8 1.23 X 1010 1 86 x lplo 2 4p x lplp 3 69 X lpl0 6-15

TABLE 6-2 (Continued)

CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD/BASE METAL INTERFACE Iron Atom Dis lacement Rate d a sec 0.0'5.0'0.0'5.0'ESIGN 1.37 11 2.19 11 BASIS X 10 X 10 2.73 X 10 4.26 X 10 CYCLE 1 1.03 X 10 11 1.63 X 10-11 pp X lp-ll 3 p8 X lp-ll CYCLE 2 1.10 X 10 1.66 X 10 11 g7 X lp-ll 3 p5 X lp-ll CYCLE 3 p5 X lp-ll 1 54 X lp-ll 1.69 X 10 11 2.54 X 10 CYCLE 4 1.07 X 10 11 1.67 X 10 11 76 X lp-ll 2 5g X lp-ll CYCLE 5 p6 X lp-ll 1 54 X lp-ll 1.56 X 10 1

2.32 X 10 11 CYCLE 6 g 57 X lp-12 1 52 X lp-ll 1.62 X 10 11 2.43 X 10 11 CYCLE 7 1.04 X 10 11 1.61 X 10 11 53 X lp-ll 2 24 X lp-ll 8 8.02 2 1.19 11 11 CYCLE X 10 X 10 1.52 X 10 2.26 X 10 6-16

TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV)

WITHIN THE PRESSURE VESSEL WALL Radius

~cm 0'5'0'5'20.27(1) 1.00 1.00 1.00 1.00 220.64 0.977 0.978 0.979 0.977 221.66 0.884 0.887 0.889 0.885 222.99 0.758 0.762 0.765 0.756 224.31 0.641 0.644 0.648 0.637 225.63 0.537 0.540 0.545 0.534 226.95 0.448 0.451 0.455 0.443 228.28 0.372 0.373 0.379 0.367 229.60 0.309 0.310 0.315 0.303 230.92 0.255 0.257 0.261 0.250 232.25 0.211 0.212 0.216 0.206 233.57 0.174 0.175 0.178 0.169 234.89 0.143 0.144 0.147 0.138 236.22 0.117 0.118 0.121 0.113 237.54 0.0961 0.0963 0.0989 0.0912 238.86 0.0783 0.0783 0.0807 0.0736 240.19 0.0635 0.0632 0.0656 0.0584 241.51 0.0511 0.0501 0.0519 0.0454 242.17(2) 0.0483 0.0469 0.0487 0.0422 NOTES: 1) Base Meta1 Inner Radius

2) Base Meta1 Outer Radius 6-17

TABLE 6-4 h

RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 HeV)

WITHIN THE PRESSURE VESSEL WALL Radius

~cm po 15'0'5'20.27(1) 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 221.66 1.00 0.996 1.00 0.994 222.99 0.965 0.958 0.968 0.953 224.31 0.916 0.906 0.919 0.898 225.63 0.861 0.849 0.865 0.838 226.95 0.803 0.790 0.809 0.777 228.28 0.746 0.732 0.752 0.717 229.60 0.689 0.675 0.695 0.657 230.92 0.633 0.619 0.640 0.600 232.25 0.578 0.565 0.586 0.544 233.57 0.525 0.513 0.534 0.490 234.89 0.474 0.463 0.483 0.437 236.22 0.424 0.414 0.433 0.387 237.54 0.375 0.367 0.385 0.338 238.86 0.328 0.322 0.338 0.291 240.19 0.283 0.277 0.292 0.244 241.51 0.239 0.232 0.245 0.196 242.17(2) 0.229 0.220 0.232 0.183 NOTES: 1) Base Metal Inner Radius

2) Base Petal Outer Radius 6-18

TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa)

WITHIN THE PRESSURE VESSEL WALL Radius 00 15'0'5'20.27(1) 1.00 1.00 1.00 1.00 220.64 0.983 0.983 0.984 0.983 221.66 0.913 0.914 0.918 0.915 222.99 0.818 0.819 0.827 0.820 224.31 0.728 0.728 0.739 0.730 225.63 0.647 0.646 0.659 0.647 226:95 0.574 0.573 0.587 0.573 228.28 0.510 0.507 0.523 0.507 229.60 0.453 0.450 0.466 0.449 230.92 0.402 . 0.399 0.414 0.397 232.25 0.356 0.353 0.368 0.349 233.57 0.315 0.312 0.327 0.307 234.89 0.277 0.275 0.289 0.269 236.22 0.243 0.241 0.254 0.233 237.54 0.212 0.210 0.222 0.201 238.86 0.182 0.181 0.192 0.170 240.19 0.155 0.154 0.164 0.141 241.51 0.131 0.128 0.137 0.113 242.17(2) 0.125 0.122 0.130 0.106 NOTES: 1) Base Meta1 Inner Radius

2) Base Meta1 Outer Radius 6-19

TABLE 6-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission Monitor of Weight Response Product Yield Material Interest Fraction ~Ran e Half-Life ~X Copper Cu (n,e)Co 0.6917 E > 4.7 MeV 5.272 yrs Iron Fe 4(n,p)Hn 0.0582 E > 1.0 MeV 312.2- days Nickel Ni58(n,p)Co58 0.6830 E > 1.0 MeV 70.90 days Uranium-238* U238(n f)Cs137 1.0 E > 0.4 HeV 30.12 yrs 5.99 Neptunium-237* Np (n,f)Cs 1.0 E > 0.08 MeV 30.12 yrs 6.50 Cobalt-Aluminum* Co (n,y)Co 0.0015 0.4ev>E> 0.015 HeV 5.272 yrs Cobalt-Aluminum Co (n,y)Co 0.0015 -:

E > 0.015 MeV 5.272 yrs

• Denotes that monitor is cadmium shielded.

TABLE 6-.7 MONTHLY THERMAL GENERATION DURING THE FIRST EIGHT FUEL CYCLES OF THE D C COOK UNIT 2 REACTOR THERMAL THERMAL THERMAL THERMAL GENERATION GENERATION GENERATION GENERATION

~MONT NW-h IIONTII MW-h r MONTH MW-h 80NTH ~MII-h 3/78 53096 9/81 2430714 3/85 2461488 9/88 0 4/78 521821 10/81 284784 4/85 2449523 10/88 0 5/78 653969 11/81 2435848 5/85 2532441 11/88 0 6/78 1365478 12/81 2517865 6/85 2451623 12/88 0 7/78 1247083 1/82 2295944 7/85 1049002 1/89 0 8/78 1529472 2/82 2196190 8/85 60639 2/89 0 9/78 2178779 3/82 833555 9/85 0 3/89 775260 10/78 2231119 4/82 2391274 10/85 163249 4/89 2288800 11/78 848238 5/82 2516937 11/85 1372641 5/89 2530450 12/78 2476056 6/82 2331168 12/85 2019347 6/89 1297315 1/79 2240714 7/82 2496782 1/86 2043640 7/89 2508038 2/79 2220562 8/82 1011517 2/86 1360957 8/89 2155830 3/79 2483455 9/82 2241332 3/86 0 9/89 2452143 4/79 2164269 10/82 2293400 4/86 0 10/89 2493553 5/79 1449347 11/82 1575311 5/86 0 11/89 2355817 6/79 0 12/82 0 6/86 0 12/89 2454307 7/79 2258164 1/83 341534 7/86 980980 1/90 861130 8/79 2513690 2/83 2242228 8/86 2034055 2/90 2282581 9/79 2266726 3/83 2533602 9/86 1973118 3/90 2534067 10/79 1522346 4/83 2428234 10/86 2014579 4/90 2452883 ll/79 0 5/83 2461540 11/86 1974208 5/90 2165049 12/79 0 6/83 1851461 12/86 2039056 6/90 1767222 l/80 584404 7/83 1711373 1/87 2039325 7/90 0 2/80 2209403 8/83 2 343637 2/87 1776049 8/90 0 3/80 2418799 9/83 2 269321 3/87 159603 9/90 0 4/80 2354329 10/83 1188161 4/87 456573 10/90 0 5/80 2483250 11/83 453959 5/87 2080553 ll/90 1338296 6/80 2187611 12/83 2 383687 6/87 1849770 12/90 1887935 7/80 1408949 1/84 2 435731 7/87 1409763 1/91 2533040 8/80 2496594 2/84 2 235476 8/87 1485651 2/91 2110364 9/80 2393783 3/84 733977 9/87 0 3/91 2241782 10/80 1414143 4/84 0 10/87 1258092 4/91 2447136 11/80 0 5/84 0 11/87 1973726 5/91 2495857 12/80 1488758 6/84 0 12/87 1995088 6/91 2410166 1/81 2505373 7/84 1 417277 1/88 2039814 7/91 2435150 2/81 2271684 8/84 2 341526 2/88 1900060 8/91 590907 3/81 1053877 9/84 2 325725 3/88 2038466 9/91 2357045 4/81 0 10/84 2 423846 4/88 1432639 10/91 2519131 5/81 449803 ll/84 2 056182 5/88 0 ll/91 1912720 6/81 2374202 12/84 1 059199 6/88 0 12/91 2474282 7/81 1775877 1/85 1 361108 7/88 0 1/92 2442527 8/81 2338703 2/85 2 271484 8/88 0 2/92 1219168 6-21

TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES Capsule Center Measured Saturated Reaction Monitor and Activity Activity Rate xial Location dis sec- m dis sec- m ~IIP NU LiU Cu-63 (n,a) Co-60 Top-Middle 1.23 x 105 2.67 x 105 Middle 1.23 x 105 2.67 x 105 Bottom-Middle 1.24 x 105 2.70 x 105 Average 1.23 x 105 2.68 x 105 3.94 x 10 17 Fe-54(n,p) Mn-54 Top 9.19 x 105 2.15 x 106 Top-Middle 9.45 x 105 2.21 x 106 Middle 9.50 x 105 2.22 x 106 Bottom-Middle 8.59 x 105 2.01 x 106 Bottom 9.30 x 105 2.18 x 106 Average 9.21 x 105 2.15 x 106 3.64 x 10-15 Ni-58 (n,p) Co-58 Top 3.93 x 106 3.21 x 107 Middle 3.93 x 106 3.21 x 107 Bottom 3.96 x 106 3.24 x 107 Average 3.94 x 106 3.22 x 107 5.41 x 10 15 U-238 (n,f) Cs-137 (Cd)

Middle 4.88 x 105 2.94 x 106 1.94 x 10 14 6-22

TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES cont'd Capsule Center Measured Satur ated Reaction Monitor and Activity Activity Rate xial Location dis sec- m dis sec- m ~IIP NUCLEUS Np-237(n,f) Cs-137 (Cd)

Middle 4.26 x 106 2.57 x 107 1.61 x 10 13 Co-59 (n,y) Co-60 Top 1.88 x 107 4.09 x 107 Bottom 1.74 x 107 3.78 x 107 Average 1.81 x 107 3.94 x 107 2.70 x 10 12 Co-59 (n,y) Co-60 (Cd)

Bottom 32 x 106 1.59 x 107 1 20 x 10-12 6-23

TABLE 6-9

SUMMARY

OF NEUTRON DOSIMETRY RESULTS TIME AVERAGED EXPOSURE RATES P (E > 1.0 MeV) {'n/cm2-sec} 5.78 x 1010 4 (E > 0. 1 MeV) (n/cm2-sec} 2.00 x 1011 + 15X dpa/sec 9.78 x 10-11 + 10X 4 (E < 0.414 eV) (n/cm2-sec} 6.29 x 1010 + 20X INTEGRATED CAPSULE EXPOSURE (E > 1.0 MeV) (n/cm2} 1.58 x 10 4 (E > 0. 1 MeV) (n/cm2} 5.46 x 10 + 15X dpa 2.67 x 10 2 + 10X 4'E < 0 414 eV) (n/cm2} 1.72 x 1019 + 20X NOTE: Tota1 Irradiation Time = 8.65 EFPY 6-24

TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER Adjusted Reaction peas~ed Calcu ation Cu-63 (n,a) Co-60 3.94xl0 17 3.94x10 17 1.00 Fe-54 (n,p) Mn-54 3.64xl0 15 3.74x10 15 1.03 Ni-58 (n,p) Co-58 5.41xl0 15 5.28xl0 15 0.98 1.94xlo-'4 14 U-238 (n,f) Cs-137 (Cd) 1.93xl0 1.00 Np-237 (n,f) Cs-137 1.6lx10 13 1.61xl0 13 1.00 (Cd)

Co-59 12 2.69X10-12 1.00 (n,y) Co-60 2.70x10 Co-59 (n,y) Co-60 (Cd) 1.20xl0 12 1.20x10 12 1.00 6-25

TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER Energy Adjusyed Flux Energy Addus/ed Flux Group (Mev) (n/cm -sec) Group (Mev) (n/cm -sec) 1 1.73xl01 3.36x106 28 9.12x10 3 8.36xl09 2 1.49xl01 8.20xl06 29 5.53x10 3 1.05xl010 3 1.35xl01 3.76x107 30 3.36x10 3 3.25xl09 4 1.16xl01 9.5lx107 ~

31 2.84x10 3 3.09x109 5 l.ooxlol 2.29xl08 32 2.40xlO 3 2.98x109 6 8.61x100 4.llxl08 33 2.04xlO 3 8.60xl09 7 7.4lx100 9.86xl08 34 1.23xlO 3 8.31xl09 8 6.07xl00 1.42xl09 35 7.49xlO 4 7.99xl09 9 4.97x100 2.92x109 36 4.54xlO 4 7.73x109 10 3.68xl00 3.69xl09 37 2.75x10 4 8.15xl09 ll 2.87xl00 7.37xl09 38 1.67xlO 4 8.85xl09 12 2.23xl00 9.24x109 39 1.0lxlO 4 8.8lxl09 13 1.74xl00 , 1.21x1010 40 6.14x10 5 8.79xl09 14 1.35xl00 1.22xl010 41 3.73xlO 5 8.66xl09 15 l.llxl00 2.08xl010 42 2.26xlO 5 8.48xl09 16 8.2lx10 1 2.22x1010 43 1.37xlO 5 8.31x109 8 32xlO 6 1

17 6.39xlO 2.18x1010 44 8.02xl09 18 4.98x10 1 1.55xl010 45 5.04xlO 6 7.55xl09 19 3.88x10 I 2.05xl010 46 3.06xlO 6 7.18x109 20 3.02xlO 1 2.27x1010 47 1.86xlO 6 6.72xl09 21 1.83x10 1 2.16xl010 48 1.13x10 6 5.29xl09 22 l.llxlO 1 1.71xl010 49 6.83xlO 7 6.28x109 2

23 6.74xlO 1.23xl010 50 4.14x10 7 1.03xl010 2

24 4.09x10 7.42xl09 51 2.51x10 7 1.05xl010 2

25 2.55xlO 8.90xl09 52 1.52x10 7 1.05x1010 2

26 1.99xlO 4.83xl09 53 9.24xlO 8 3.16x1010 27 1.50xlO 2 6.52xl09 NOTE: Tabulated energy levels represent the upper energy of each group.

6-26

TABLE 6-12 0 COMPARISON OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE U Calculated Measured ~CM C(E > 1.0 MeV) (n/cm ) 1.49 x 1019 1.58 x 1019 0.94 4(E > O. 1 MeV) (n/cm ) 5.01 x IO>> 5.46 x 1019 0.92 dpa 2.55 x 10 2 2.67 x 10 2 0.96

%(E < 0.414 eV) {n/cm ) 8.84 x 1018 1.72 x 1019 0.51 6-27

TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD/BASE HETAL INTERFACE 8.65 EFPY (E > 1.0 Hev) 45'.65 X 10'8

[n/cm2]

(E > O.l HeV) 4.48 X 101 6.12 X 101 6.95 X 101 7.79 X 10 1.21 X 101

[n/cm2]

Iron Atom Displacements 2.92 X 10 3.93 X 10 4.46 X 10 4.91 X 10 7.39 X 10

[dpa]

16.0 EFPY 15'0'5'.31 (E > 1.0 Hev) X 1018 4.51 X 1018 5.12 X 1018 5.72 X 1018 8.58 X 1018

[n/cm2]

C (E > O.l HeV) . 8.28 X 1018 1.13 X1019 1.29 X 10 1.44 X 1019 2.23 X 1019

[n/cm2]

Iron Atom Displacements 5.40 7.26 3 X 10 X 10 8.24 X 10 9.09 X 10 1.36 X 10

[dpa]

32.0 EFPY (E > 1.0 Hev) 9.02 1.02 1.14 45'.63 X 10 X 10 X 10 X 10 1.71 X 10

[n/cm2]

(E > O.l HeV) 1.66 X 10 2.26 X 10 2.56 X 10 2.87 X 10 4.45 X 10

[n/cm2]

Iron Atom Displacements 1.08 2 2 2 2 X 10 1.45 X 10 1.64 X 10 1.81 X 10 2.72 X 10

[dpa]

TABL 14 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOWN CURVES 16 EFPY NEUTRON FLUENCE E > 1.0 HeV SLOPE ~da SLOPE (n/cm ) (equivalent n/cm2)

Surface ~14 T ~34 T Surface ~14 T ~34 T 00 x 1018 1.75 x lp 3.61 x 1017 3.31 x 1018 '.12 x 1018 7.69 x 1017 4.51 x lp 2.40 x 1018 4.96 x 1017 4.51 x lp 2.88 x 1018 1.04 x 1018 10'5'0'5'.31 5.12 x 1018 x 1018 5.63 x 1017 2.72 5.12 x 1018 3.27 x 1018 1.18 x 1018 5.72 x 10 3.07 x 1018 6.46 x 1017 5.72 x 10 3.73 x 10 1.38 x 1018 8.58 x 1018 4.51 x 10 9.01 x 1017 8.58 x 10 5.49 x lpl 1.90 x 10 32 EFPY NEUTRON FLUENCE E > 1.0 HeV SLOPE ~da SLOPE (n/cm ) (equivalent n/cm2)

Surface ~14 T 34T Surface 1 4T ~34 T 00 6.63 x 10 3.51 x 1018 7.22 x 1017 6.63 x 10 4.24 x 1018 1.54 x 1018 9.02 x 10 4.80 x 1018 9.92 x 1017 9.P2 x 1P18 x 1P18 2.07 x 1018 10'5'0'5 1.02 x 10 5.45 x lp 1.13 x 1018 1.02 x 10'9 6.54 x 1018 2.35 x 1018 1.14 x 10'9 6.15 x 1018 1.29 x 1018 1.14 x 1019 7.46 x 1P18 2.77 x 1018 1.71 x 10 9.02 x 1018 1.80 x 1018 1.71 x 10 1.10 x 10'9 3.79 x 1018 6-29

TABLE 6-15 UPDATED LEAD FACTORS FOR D C COOK UNIT 2 SURVEILLANCE CAPSULES

~Ca sule Lead Factor T 3 44(a)

X 3 41(c)

U 3 40(d)

Y 3 44(b)

S 1 30(d)

V 1.30(d)

W 1.30(d)

Z 1.30(d)

(a) Plant specific evaluation based on end of Cycle 1 calculated fluence.

(b) Plant specific evaluation based on end of Cycle 3 calculated fluence.

(c) Plant specific evaluation based on end of Cycle 5 calculated fluence.

(d) Plant specific evaluation based on end of Cycle 8 calculated fluence.

6-30

SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the D. C. Cook Unit 2 reactor vessel:

Capsule Estimated Location Lead Fluence Capsule (deg.) Factor Removal Time ( ) (n/cm2) 40 3.44 1.08 (Removed) 2.64 x 10 (Actual) 320 3.44 3.24 (Removed) 6.83 x 10 (Actual) 220 3.44 5.27 (Removed) 1.06 x 10 (Actual) 140 3.44 8.65 (Removed) 1.58 x 1019 (Actual) 4 1.30 32 2.22 X 10'9 356 1.30 Standby 184 1.30 Standby 176 1.30 Standby (a) Effective Full Power Years (EFPY) from plant startup.

7-1

SECTION

8.0 REFERENCES

1. J. A. Davidson, et al., "American Electric Power Company Donald C. Cook Unit No. 2 Reactor Vessel Radiation Surveillance Program", WCAP-8512, November 1975.

2. Code of Federal Regulations, 10CFR50, Appendix G, "Fracture Toughness Requirements", and Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.

3. Regulatory Guide 1.99, Proposed Revision 2, "Radiation Damage to Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, Hay 1988.

4. Section III of the ASME Boiler and Pressure Vessel Code, Appendix G, "Protection Against Nonductile Failure."

5. ASTM E208, "Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels."

6. ASTH E185-82, "Standard Practice for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF)."

7. ASTH E23-88, "Standard Test Methods for Notched Bar Impact Testing of Metallic Materials."

8. ASTH A370-89, "Standard Test Methods and Definitions for Mechanical Testing of Steel Products."

9. ASTM E8-89b, "Standard Test Methods of Tension Testing of Metallic Materials."

10. ASTH E21-79(1988), "Standard Practice for Elevated Temperature Tension Tests of Metallic Materials."

8-1

ll. ASTM E83-85, "Standard Practice for Verification and Classification of 12.

Extensometers."

SwRI Project No. 02-5928, "Reactor Vessel Material Surveillance Program 0

for Donald C. Cook Unit No. 2 Analysis of Capsule T", E. B. Norris, September 16, 1981.

13. SwRI Project No. 06-7244-002, "Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. 2 Analysis of Capsule Y", E. B.

Norris, February 1984.

14. SwRI Project No. 06-8888, "Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. 2 Analysis of Capsule X", P. K. Nair and M. L.

Williams, Hay 1987.

15. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, "Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5 Two-Dimensional Discrete Ordinates Transport Technique", WANL-PR(LL)-034, Vol. 5, August 1970.

1 6. "ORNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors".

17. AEP Letter RBB 88-005/4, R. B. Bennett (AEP) to H. C. Walls (Westinghouse), January 29, 1988.

18. AEP FAX, G. John (AEP) to S. L. Anderson (Westinghouse), October 6, 1992.

19. B. J. Johansen, et. al., "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit 2, Cycle 8)", WCAP-12651, October 1990. (Proprietary)

20. ASTM Designation E482-89, "Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTH Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

8-2

21. ASTM Designation E560-84, "Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTH Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

22. ASTH Designation E693-79, "Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

23. ASTM Designation E706-87, "Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTH Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

24. ASTM Designation E853-87, "Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

25. ASTH Designation E261-90, "Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

26. ASTM Designation E262-86, "Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in ASTH Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

27. ASTM Designation E263-88, "Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

28. ASTH Designation E264-87, "Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel", in ASTH Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

8-3

29. ASTM Designation E481-86, "Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver", in ASTH Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

30. ASTM Designation E523-87, "Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", in ASTH Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

31. ASTH Designation E704-90, "Standard. Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTH Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

32. ASTM Designation E705-90, "Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237", in ASTH Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

33. ASTM Designation E1005-84, "Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.

34. F. A. Schmittroth, FERRET Data Anal sis Core, HEDL-THE 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

35. W. N. HcElroy, S. Berg and T. Crocket, A Com uter-Automated Iterative Method of Neutron Flux S ectra Determined b Foil Activation, AFWL-TR-7-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.

36. EPRI-NP-2188, "Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al., 1981.

APPENDIX A Load-Time Records for Charpy Specimen Tests A-0

Wi P~ = MAXIMUMLOAD P - F RACTURE LQaO PGY.N GENERAL I YIE LO LQAO I I

I I

I I

I PA

• ARREST LQAO I I I I I I I

I I

.I .

I I

I I

I I tGYj mai rivE b'igure A-1. Ideali. oad-time record

v'E N2 "U" DQiL45 NL45 CI ag

~

a

~ D ~ 8 I~ 6 2+4 302 4. 0 TILE ( lCEC >

'iE N2 U" OCIRA8 N 48

~ D eB Io6 2e4 3I 2 4+0 TItK ( ttSEC >

Figure A-2. Load-time records for Specimens ML45 and ML48

'E N2 'U" OCttt 42 o r C9 CU wl

.D e8 I ~ 6 2e4 30 2 4+0 TIttE < ttSEC >

X I2 "u'Cttt.44

~ D I 6 2e4 30 2 4m 0 TItK C tCEC >

Figure A-3. Load-time records for Specimens ML42 and ML44

Cl eE 12 "U OCt1L.41 01.41 w ~

C9

~ D o8 1~ 6 2e4 30 2 4.0 T11K ( tSEC >

'A 02 'U" DCtL43 Pl g ~

O 0 (U 8

~ D ~ 8 1+6 2 4 3e2 4m 0 TI% ( NSEC >

Figure A-4. Load-time records for Specimens ML41 and ML43

E <<2 "U" OCttL46 og

~

R CI eD ~ 8 1~ 6 2e< 3e2 4.0 TIttE ( tSEC )

CI E I2 'U" g e

~ D ~ 8 1e6 2e4 302 4e0 TIttE C tSEC )

Figure A-5. Load-time records for Specimens ML46 and ML47

~CNT62 NT62 eD ~ 8 1.6 2,4 3e2 4,0 TINE < tSEC )

CI Q NUN OCNT6 I NT6I Cl

~ D ~ 8 I~ 6 2,4 3e2 4o0 TINE C tSEC )

Figure A-6. Load-time records for Specimens MT62 and MT61

lg I2 %US tt T66 Pl o~ r~

oD o8 1~6 2+4 3o2 4e0 Tlt1E < tCEC )

Cl X 12 'U" lO ag r ~

S 0

CL' Gi CI

~D ~ 8 1 ~ 6 2+4 30 2 4e0 T1tK C NSEC )

Figure A-7. Load-time records for Specimens MT66 and MT71

OCNT64 HT64

~ 8 I 6 Re4 30R 4+ 0 TilK C NSEC )

'iE IR "U i+6 Re4 30 R 4m 0 7itK ( ICEC )

Figure A-8. Load-time records for Specimens MT64 and MT72

E 02 OU>> OcttT70 ttTPO ar r

Q Ol P

Ol

~4

~

p ~ 8 I~ 6 L4 30 2 4e0 TIttE C tCKC )

vE N2 "U" OCttT69 ttT69 g co o

r Ol g

CO

~ D ~ 8 I~ 6 2e4 3t2 4e0 TItK . < tSEC )

Figure A-9. Load-time records for Specimens MT70 and MT69

'E N2 "U DMT63 tlT63 O CU CC D

D

~ D <<8 I~ 6 2e4 31 2 4e0 T IttE ( CEC )

CI a I2 "U OCttT68

%0 g EO o

W C9 I cU O

~ D ~ ~ 8 I~ 6 2+4 30 2 4,0 TIttE C HSEC )

Figure A-10. Load-time records for Specimens MT63 and MT68

~

E ge iu NT67 C)

~ D 1.6 Reh 30 8 ho 0 TINE < NSEC )

Xia "u NT65 og

~

C9 I cU o

~ D e8 Ie6 8+4 3og 4.0 TINE ( !CEC )

Figure A-ll. Load-time records for. Specimens M'Z67 and MT65

E IIQ NUN IIII70 X

ID o8 le6 Bo4 30 8 4e0 TltlE ( tlSEC )

18 1QN OCt0I64 IDI64 CI g co e~

X CI

~D ~ 8 1.6 8+ 4. 308 4m 0 TItIE < IISEC )

Figure A-12. Load-time records for Specimens MW70 and MW64

'E N2 "U" . OCtNl71.

e8 1~6 2o4 312 4m 0 TlttE ( NSEC )

X N2 'U"

<<8 1~6 2 4 30 2 4+0 TI% ( tCEC )

Figure A-13, Load-time records for Specimens MW71 and MW68

iE NR "U" OCMtt63 ttll63 o ~

Pl o

CU o

~ D ~ 8 1~6 Re4 3t 2 4e0 TittE ( tCEC )

XNR U OCtttt61 tNt61

~ D CO o 4o0

~ D e8 1~6 Ro4 30 2 TINE < tSEC )

Figure A-14. Load-time records for Specimens MW63 and MW61

O Q SU% OCtCl78 C9

~ D ~ 8 I+6 Re4 3o8 4e0 TItK < NSEC )

O X 18 U" og

~

X D ~ 8 1.6 Ro4 30 8 4+0 TIlK < tLXC )

Figure A-15. Load-time records for Specimens MW72 and MW65

'E tJR "U" OCt0J66 NJJ66 o

a AJ

.D .8 1.6 R.4 30 R 4. 0 TiNE ( NSEC >

uE ttR "U" OCNJJ6R AJ

~4 O

~ D .8 i+6 Ro4 30 R 4 0 TINE C NSEC )

Figure A-16. Load-time records for Specimens MW66 and MW62

Pl o

ID Cl 0a (U Cl

~ 8 I+6 2+4 3.2 4.0 TINE < ttSEC )

CI

'E <<2 U" OCtSI69 Nt69 oh cU Q

D CU

~ D ~ 8 I.6 2.4 30 2 4,0 TIlK < tCEC )

Figure A-17. Load-time records for Specimens MW67 and MW69

o lE IIQ NUN OCtIK67 tQt67 og

~

D~

D I AI Al

.D ~ 8 1.6 Rl4 30 8 4.0 T lttE ( ttSEC )

~ II UN ttK63 E IQ OcttK63 oa

~

Cl C9 D AI 5

lD .8 1.6 2.4 3.2 4.0 TIE ( ttSEC )

Figure A-18. Load-time records for Specimens MH67 and MH63

vE N2 "U" Pl oW ID C9 oD o8 I~ 6 2+4 3@2 4~ 0 TINE ( NSEC >

X N2 U" OCNH69 Ntt69 g ~

CU

~4 O

~ D o8 1.6 2,4 31 2 4.0 TINE C NSEC >

Figure A-19. Load-time records for Specimens MH71 and MH69

N2 "U" Oct1H72 or

~

CJJ C9 C-i CIJ EZ o

CU

,8 1.6 2o4 3.2 4e0 TIttE C tCEC )

ZN2 U og

~

m CJJ C9 O CU cZ CI CU

~ D .8 1,6 2.4 302 4. 0 T ltCE C tSEC )

Figure A-20. Load-time records for Specimens MH72 and MH70

v'E tt2 "U OCtN62 t%62 ox O

.D e8 1.6 2.4 3t 2 4. 0 TIttE ( NSEC )

'I E tt2 HU$ OCt%64 ttH64 a

C9 oD 8 l,6 2+4 3.2 4.0 TIttE ( ttSEC )

Figure A-21. Load-time records for Specimens MH62 and MH64

'eiE NB "U" OCNH61 NH61 olE a

0a IN 0

O

.D .8 1.6 2eh 30 8 4.0 TINE < NSEC )

'E NR "U" OCNH65 o

~ D ~ 8 Io6 R.h 308 4. 0 TINE ( t5EC )

Figure A-22. Load-time records for Specimens MH61 and MH65

g2 NUN 0Ct%68 W68 a

N a

oD .8 1~ 6 2.4 4.0 TllC < ICEC )

'E N2 "U" OCt%66 A<66 eD ~ 8 1~ 6 2e4 30 2 4o0 TllC ( NSEC )

Figure A-23. Load-time records for Specimens MH68 and MH66

APPENDIX B Photographs of Charpy, Tensile and WOL Specimens Prior to Testing B-0

Figure B-1. Charpy impact specimens ML45, ML48, ML42, and ML44 from Intermediate Shell Plate C5521-2 (longitudinal orientation) before testing.

B-I RM-28359

Figure B-2. Charpy impact specimens ML41, ML43, ML46, and ML47 from Intermediate Shell Plate C5521-2 (longitudinal orientation) before testing.

B-2 RM-28360

Figure B-3. Charpy impact specimens MT62, MT61, MT66, MT71, MT64, and MT72 from Intermediate Shell Plate C5521-2 (transverse orientation) before testing.

8-3 RM-28361

Figure B-4. Charpy impact specimens MT70, MT69, MT63, MT68, MT67, and MT65 from Intermediate Shell Plate C5521-2 (transverse orientation) before testing.

RM-28362

Figure B-5. Charpy impact specimens MW70, MW64, MW71, MW68, MW63, and MW61 from the weld metal, before testing.

B-5 RM-28363

Figure B-6. Charpy impact specimens MW72, MW65, MW66, MW62, MW67, and MW69 from the weld metal, before testing.

B-6 RM-28364

Figure B-7. Charpy impact specimens MH67, MH63, MH71, MH69, MH72, and MH70 from the heat-affected zone (HAZ), before testing.

B-7 RM-28365

Figure B-8. Charpy impact specimens MH62, MH64, MH61, MH65, MH68, and MH66 from the heat-affected zone (HAZ), before testing.

B-8 RM-28366

Figure B-9. Tensile specimens MTll and MT12 from D. C. Cook Unit 2 reactor vessel Intermediate Shell Plate C5521-2 (transverse orientation) before testing.

B-9 RM-28367

Figure B-10. Tensile specimens MWlland MW12 from D. C. Cook Unit 2 reactor vessel weld before testing.

B-10 RM-28368

Figure B-11. WOL specimens MW5, MW6, MW7 and MW8, from D. C. Cook Unit 2 reactor vessel. The specimens were not tested, but stored for future reference, B-1 1 RM-28369

APPENDIX C Heatup and Cooldown Limit Curves for Normal Operation C-0

TABLE OF CONTENTS Section Title ~acae 1 INTRODUCTION C-4 2 FRACTURE TOUGHNESS PROPERTIES C-4 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS C-5 4 HEATUP AND COOLDOWN LIMIT CURVES C-8 5 ADJUSTED REFERENCE TEMPERATURE C-10 6 REFERENCES C-24

LIST OF ILLUSTRATIONS

~

~Fi ere Title

~

Pa<ac 1 D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations C-16 (Heat up rate up to 60'F/hr) Applicable for the First 32 EFPY (Without Margins For Instrumentation Errors) 2 D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations C-17 (Heat up rate up to 60'F/hr) Applicable for the First 32 EFPY (With Margins of 10'F and 60 psig For Instrumentation Errors) 3 D. C. Cook Unit 2 Reactor Coolant System Cooldown (Cooldown C-18 Rates up to 100'F/hr) Limitations Applicable for the First 32 EFPY (Without Margins For Instrumentation Errors) 4 D. C. Cook Unit 2 Reactor Coolant System Cooldown (Cooldown C-19 Rates up to 100'F/hr) Limitations Applicable for the First 32 EFPY (With Margins of 10'F and 60 psig For Instrumentation Errors) 5 D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations C-20 (Heat up rate up to 60 F/hr) Applicable for the First 15 EFPY (Without Margins For Instrumentation Errors)

D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations C-21 (Heat up rate up to 60'F/hr) Applicable for the First 15 EFPY (With Margins of 10'F and 60 psig For Instrumentation Errors) 7 D. C. Cook Unit 2 Reactor Coolant System Cooldown (Cooldown C-22 Rates up to 100'F/hr) Limitations Applicable for the First 15 EFPY (Without Margins For Instrumentation Errors)

C-2

LIST OF ILLUSTRATIONS continued Ficiure Title Pacae 8 D. C. Cook Unit 2 Reactor Coolant System Cooldown (Cooldown C-23 Rates up to 100'F/hr) Limitations Applicable for the First 15 EFPY (With Hargins of 10'F and 60 psig For Instrumentation Errors)

LIST OF TABLES Table Title ~Pa e D. C. Cook Unit 2 Reactor Vessel Toughness Table C-11 (Unirradiated)

Summary of Adjusted Reference Temperature (ART) at 1/4T C-12 and 3/4T Location for 32 EFPY 3 Summary of Adjusted Reference Temperature (ART) at 1/4T C-13 and 3/4T Location for 15 EFPY Calculation of Adjusted Reference Temperatures for C-14 Limiting D. C. Cook Unit 2 Reactor Vessel Haterial-Intermediate Shell Plate, C5556-2 for 32 EFPY 5 Calculation of Adjusted Reference Temperatures for C-15 Limiting D. C. Cook Unit 2 Reactor Vessel Haterial-Intermediate Shell Plate, C5556-2 for 15 EFPY C-3

1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel. The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture toughness properties and estimating the radiation-induced hRTNDT.

RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction)'minus 60'F.

RTNDT increases as the material is exposed to fast-neutron radiation.

Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)~ ~. Regulatory Guide 1.99, Revision 2 is used for the calculation of RTNDT values at 1/4T and 3/4T locations (T is the thickness of the vessel at the beltline region).

2. FRACTURE TOUGHNESS PROPERTIES The unirradiated RTNDT values for the beltline region materials in the D. C.

Cook Unit 2 reactor vessel were established using the guidance provided in NUREG-0800, Branch Technical Position, NTEB 5-2~ ~, and subarticale NB-2331 of the ASHE Boiler and Pressure Vessel Code, Section III~ ~. The pre-irradiation fracture-toughness properties of the D. C. Cook Unit 2 reactor vessel are presented in Table l.

C-4

3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at that time. KIR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code~ ~. The KIR curve is given by the following equation:

KIR = 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)]

where KIR = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code~ ~ as follows:

C KIM + KIT < KIR (2) where KIM - stress intensity factor caused by membrane (pressure) stress KIT = stress intensity factor caused by the thermal gradients KIR = function of temperature relative to the RTNDT of the material

= 2.0 for Level A and Level B service limits

= 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT, for the C-5

reference flaw are computed. From equation 2, the pressure stress intensity factors are obtained and, from these, the allowable. pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during, cooldown, the reference flaw of Appendix G to the ASHE Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

During cooldown, the I/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the bT developed during cooldown results in a higher value of KIR at the I/O T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KIR exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the I/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown'amp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as C-6

finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K1R for the 1/4 T crack during heatup is lower than the K1R for the 1/4 T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K1R's do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the'1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant

-temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel, inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.

Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced by.

constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible'for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

C-7

Finally, the 1983 Amendment to 10CFR50~4~ has a rule which addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material RTNDT by at least 120'F for normal operation when the

'ressure exceeds 20 percent'f the preservice hydrostatic test pressure (621 psig without margins for instrumentation error and 561 psig with margins for D.

C. Cook Unit 2).

Table 1 indicates that the limiting initial RTNDT of 30'F occurs in the vessel flange of D. C. Cook Unit 2, so the minimum allowable temperature of this region is 150'F excluding margins for instrumentation error and 160'F with margins. These limits are shown in Figures 1 through 8 whenever applicable.

4. HEATUP AND COOLDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary reactor pressure vessel have been calculated using the methods discussed in Section 3. If pressure readings are measured at other locations than the limiting beltline region,'he pressure differences between the pressure transmitter and the limiting beltline region must be accounted for when using the pressure-temperature limit curves herein. The indicated pressure and temperature labels provided on the curves relate to the limiting beltline region of the reactor vessel.

Figures 1, 2, 5 and 6 contain the heatup curves for 60'F/hr. Figures 3, 4, 7 and 8 contain the cooldown curves up to 100'F/hr . Figures 1 and 3 are applicable for the first 32 EFPY of operation and include no margins for possible instrumentation errors. Figures 2 and 4 are applicable for the first 32 EFPY of operation and include margins of 10'F and 60 psig for possible instrumentation errors. Figures 5 and 7 are applicable for the first 15 EFPY of operation and include no margins for possible instrumentation errors.

Figures 6 and 8 are applicable for the first 15 EFPY of operation and include margins of 10'F and 60 psig for possible instrumentation errors.

The current D. C. Cook Unit 2 low temperature overpressure protection system (LTOP) setpoints are valid up to the 15 EFPY pressure-temperature limit curves.

C-8

The 32 EFPY pressure-temperature limit curves cannot be used with the current LTOP setpoints.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 1 through 8. This is in addition to other criteria which must be met before the reactor is made critical.

The leak limit curve shown in Figures 1, 2, 5 and 6 represents minimum temperature requirements at the leak test pressure specified by applicable codesI ~ ~. The leak test limit curve was determined by methods of References 2 and 4.

The criticality limit curve shown in Figure 1, 2, 5 and 6, specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 4. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inser vice hydrostatic test, and at least 40'F higher than the minimum pressure-temperature curve for heatup and cooldown calculated as described in Section 3. The maximum temperature for the inservice hydrostatic test for the D. C. Cook Unit 2 reactor vessel for 32 EFPY is 348'F with margins for instrumentation errors and 335'F without margins for instrumentation errors. A vertical line at 348 F and 335 F on the pressure-temperature curves (with and without margins), intersecting a curve 40'F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel. The maximum temperature for the inservice hydrostatic test for the D. C. Cook Unit 2 reactor vessel for 15 EFPY is 324'F with margins for instrumentation errors and 311'F without margins for instrumentation errors. A vertical line at 324'F and 311'F on the pressure-temperature curves (with and without margins),

intersecting a curve 40'F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures 1 through 8 define limits for ensuring prevention of nonductile failure for the D. C. Cook Unit 2 reactor vessel.

C-9

5. ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1'.99'ev. 2 [1] the adjusted reference temperature (ART) for each material in the beltline is given by the following expression:

ART = Initial RTNDT + hRTNDT + Margin (3)

Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB2331 of Section III of the ASME Boiler and Pressure Vessel Code. If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

hRTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

NDT f

[CF] (0. 28-0. 10 1 og f) (4)

To calculate hRTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

-.24x (5)

(depth X) surface(

where x (in inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then put into equation (4) to calculate ARTNDT at the specific depth.

CF ('F) is the chemistry factor, obtained from Reference 1. All materials in the beltline region of D. C. Cook Unit 2 were considered for the limiting material. RTNDT at 1/4T and 3/4T are summarized in Tables 2 and 3 for 32 and 15 EFPY respectively. From Tables 2 and 3, it can be seen that'he limiting material is the intermediate shell plate C5556-2 for heatup and cooldown curves applicable up to 32 and 15 EFPY. Sample calculations for the RTNDT for 32 and 15 EFPY are shown in Tables 4 and 5.

TABLE 1 D. C. COOK UNIT 2 REACTOR VESSEL TOUGHNESS TABLE (Unirradiated)

CU NI I-RTNDT (a)

Material Description (X) (>) ('F)

Closure Head Flange, 4437-V-1 -20 (b)

Vessel Flange, 4436-V-2 30 (b)

Intermediate Shell, C5556-2 0.15 0.57 58 Intermediate Shell, C5521-2* 0.125 0.58 38 Lower Shell, C5540-2 O.ll 0.64 -20 Lower Shell, C5592-1 0.14 0.59 -20 Intermediate and Lower Shell 0.052 0.967 -35 Long. and Girth Weld Seams (Ht. S3986, Linde 124, Flux Lot No. 0934)*

• X weight copper and nickel content are mean values based on the available chemistry test results as indicated below

a. The initial RTNDT (I) values for the plates and welds are measured values based on actual data.

b. To be used for consideqgg flange requirements for heatup/cooldown curvesl ~.

Material Plate, C5521-2 Data Source Original Mill Test Report Copper

~wt 0.14 X. Nickel

~wt. X 0.58 Surveillance Program [1] 0.11 0.58 Mean value 0.125 0.58 Weld Original Mill Test Report 0.05 0.97 Surveillance Program [1] 0.055 0.97 Surveillance Program [1] 0.05 0.96 Mean value 0.052 0.967

TABLE 2

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURE (ART)

AT 1/4T and 3/4T LOCATION FOR 32 EFPY 32 EFPY

~Com onent 1//4T ', RTNpT at

~34T '

Intermediate Shell Plate, C5556-2 201 171 Intermediate Shell Plate, C5521-2 159 (158) 135 (129)

Lower Shell Plate, C5540-2 89 68 Lower. Shell Plate, C5592-1 114 86 Intermed. Shell Longitudinal Welds (a) 80 45 Lower Shell Longitudinal Welds (b) 92 68 Circumferenti al Weld 92 (64) 68 (40)

RTNpT numbers within ( ) are based on chemistry factor cal cul ated using capsule data.

(a) Intermediate shell longitudinal welds are located at 10'b)

Lower shell'ongitudinal welds are located at 90'-12

TABLE 3

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURE (ART)

AT 1/4T and 3/4T LOCATION FOR 15 EFPY 15 EFPY RTNDT at

~Com onent ~l4T ' ~34T F Intermediate Shell Plate, C5556-2 178 150 Intermediate Shell Plate, C5521-2 141 (137) 118 (110)

Lower Shell Plate, C5540-2 73 54 Lower Shell Plate, C5592-1 93 67 Intermed. Shell Longitudinal Welds (a) 52 20 Lower Shell Longitudinal Welds (b) 40 ll Circumferential Weld 77 (49) 41 (28)

RTNDT numbers within ( ) are based on chemistry factor calculated using capsule data.

I (a) Intermediate shell longitudinal welds are located at 10'b)

Lower shell longitudinal welds are located at 90'-13

TABLE 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING D. C. COOK UNIT 2 REACTOR VESSEL MATERIAL INTERMEDIATE SHELL PLATE, C5556-2 FOR 32 EFPY Re ulator Guide 1.99 - Revision 2 32 EFPY Parameter ~14 T ~34 T Chemistry Factor, CF ('F) 108.35 108.35 Fluence, f (10 n/cm ) ( ) 1.027 0.3703 Fluence Factor, ff 1.007 0.725 ARTNDT = CF x ff ('F) 109 79 Initial RTNpT, I ('F) 58 58 Margin, H ('F) (b) 34 34 Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 201

• 171

• ART = Initial RTNpT + ARTNDT + Margin (a) Fluence, f, is based upon fsurf (10 n/cm, E>1 Hev) = 1.71 at 32 EFPY. The D. C. Cook Unit 2 reactor vessel wall thickness is 8.5 inches at the beltline region.

(b) Margin is calculated as, H = 2 [ uI + e~ ] . The standard deviation for the initial RTNpT margin term, oI, is assumed to be O'F since the initial RTNpT is a measured value. The standard deviation for ARTNDT term, o>, is 17'F for the base metal, except that o~ need not exceed 0.5 times the mean value of hRTNDT.

• Limiting value used in development of heatup and cooldown limit curves.

TABLE 5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING D. C. COOK UNIT 2 REACTOR VESSEL MATERIAL INTERMEDIATE SHELL PLATE, C5556-2 FOR 15 EFPY Re ulator Guide 1.99 Revision 2 15 EFPY Parameter 1 4 T ~34 T Chemistry Factor, CF ('F) 108.35 108.35 Fluence, f (10 n/cm )( ) 0.483 0.1742 Fluence Factor, ff 0.797 0.537 ARTNDT = CF x ff ('F) 86 '8 Initial RTNDT I ( F) 58 58 Margin, M ('F) 34,.

34 Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 178

• 150

• ART = Initial RTNDT + hRTNDT + Margin (a) Fluence, f, is based upon fsurf (10 n/cm, E>l Mev) 0.804 at 15 EFPY. The D. C. Cook Unit 2 reactor vessel wall thickness is 8.5 inches at the beltline region.

(b) Margin is calculated as, H - 2 [ O'I + a~ ] . The standard deviation for the initial RTNDT margin term, o'I, is assumed to be O'F since the initial RTNDT is a measured value. The standard deviation for hRTNDT term, o~, is 17'F for the base metal, except that a< need not exceed 0.5 times the mean value of ARTNDT.

• Limiting value used in development of heatup and cooldown limit curves.

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEOIATE SHELL PLATE, C5556-2 LIMITING ART AFTER 32 EFPY: '/4T, 201'F 3/4T, 171'F 2500 LEAKTEST LMIT 2250 2000 1750 UNACCEPTABLE OPERATION 1500 ACCEPTABLE OPERATION 1250 HEATUP RATE UP to 60 'F/HR

a. 1000 CRlTICALlTYLIMIT BASED ON INSERVICE HYDROSTATIC TEST 750 TEMPERATURE (335 F}

FOR THE SERVICE PERIOD UP TO 32 EFPY 500 250

'0 50 100 150 200 250 300 350 400 450 500 INOICATEO TEMPERATURE (OEG.F)

Figure l. D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations (Heat up rate up to 60'F/hr} Applicable for the First 32 EFPY (Without Margins For Instrumentation Errors)

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE, C5556-2 LIMITING ART AFTER 32 EFPY: '/4T, 201'F 3/4T, 171'F 2500 LEAKTEST LIMIT 2250 2000 1750 1500 UNACCEPTABLE 1250 OPERATION 1000 CRITICALITYLIMIT g

HEATUP RATE UP BASED ON INSERVICE Ch,,

W I to 60 'F/HR HYDROSTATIC TEST TEMPERATURE (347 F) 750 FOR THE SERVICE PERIOD, UP TO 32 EFPY 500 ACCEPTABLE OPERATION 250 0 50 100 150 200 250 300 350 400 450 500 INOICATEO TFMPERATURE (OEG.F')

Figure 2. D. C.'ook Unit 2 Reactor Coolant System Heatup Limitations (Heat up rate up to 60'F/hr) Applicable for the First 32 EFPY (With Margins of 10'F and 60 psig For Instrumentation Errors)

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE, C5556-2 LIMITING ART AFTER 32 EFPY: 1/4T, 201'F I'-3/4T, 171'F 2500 2250 2000 1750 UNACCEPTABLE 1500 OPERA'EON w 1250 ACCEPTABLE OPERATION g 1000 O

LJ I

=750 O

COOLDOWN RATES 2: F/HR 0

500 20 40 60 250 100 0 50 100 150 200 250 300 350 400 450 500

. INDICATED TEMPERATURE (DEG.F)

Figure 3. D. C. Cook Unit 2 Reactor Coolant System Cooldown (Cooldown rates up to 100'F/hT) Limitations Applicable for the First 32 EFPY (Without Margins For Instrumentation Errors)

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE, C5556-2 LIMITING ART AFTER 32 EFPY: 1/4T, 201 F 3/4T, 171'F 2500 2250 2000 1750 1500 UNACCEPTABLE OPERATION w 1250 ACCEPTABLE L, 1000 OPERATION O

4J I

O 750 Z COOLD OWN F/HR 500 0

20 40 250 60 100 0

0 50 100 150 200 250 300 350 400 <50 500 INOICATEO TEMPERATURE (OEG.F')

Figure 4. D. C. Cook Unit 2 Reactor Coolant System Cooldown (Cooldown rates up to 100'F/hr) Limitations Applicable for the First 32 EFPY (With Margins of 10'F and 60 psig for Instrumentation Errors)

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE, C5556-2 LIMITING ART AFTER 15 EFPY: 1/4T, 178'F h ">'r 3/4T 150 F 2500 LEAKTEST LIMK 2250

'000 1750 UNACCEFIABLE 1500 OPERATION N

CL w 1250

a. 1000 HEAIUP RAKUP Ci to N'F1HR CRIIICALITYLIMIT LJ I BASED ON INSERVICE 750 HYDROSTATICTEST oX 'IEMPIHtATUREg1l F)

FOR THE SERVICE PERIOD UP TO 15 EFPY 500 ACCEFI'ABLE OPERA'IION 250 0 50 100 150 200 250 300 350 400 450 500 INDICATED,TEMPERATURE (DEG.F)

Figure 5. D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations (Heat up rate up to 60'F/hr) Applicable for the First 15 EFPY (Without Margins For Instrumentation Errors)

C-20

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE, C5556-2 LIMITING ART'FTER 15 EFPY: 1/4T, 178'F 3/4T, 150'F 2500 LEAKTEST LIMIT 2250 2000 1750 1500 UNACCEPrABLE OPERATION 1250 1000 O

I HEATUP RA1E UP CRITICALITYLIMIT 60 oF/HR BASED ON INSERVICE 750 Cl HYDROSTATICTEST

'IEMPERATURE (324 %)

FOR THE SERVICE PERIOD 500 UP 'IO IS EFPY ACCEPI'ABLE OPERA%ION 250 0 50 100 150 200 250 300 350 400 450 500 INO ICATED TEMPERATURE (BEG.F)

Figure 6. D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations (Heat up rate up to 60'F/hr) Applicable for the First 15 EFPY (With Margins of 10'F and 60 psig For Instrumentation Errors)

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE, C5556-2 LIMITING ART AFTER 15 EFPY: 1/4T, 178'F

'.'I3/4T, 150'F 2500 2250 2000 1750 1500 4I 1250 UNACCBFrABLB OPERATION

a. 1000 O

I 750 O

ACCEPTABLE 0 OPBRAIION 500 20 40 60 250 100 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEIAPERATURE (DEC.F)

Figure 7. D. C. Cook Unit 2 Reactor Coolant System Cooldown (Cooldown rates up to 100'F/hr) Limitations Applicable for the First 15 EFPY (Without Margins For Instrumentation Errors)

C-22

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE, C5556-2 LIMITING ART AFTER 15 EFPY: 1/4T, 178'F 3/4Ts 150 F 2500 2250 2000 1750 1500 UNACCEPI'ABLE 1250 OPERATION

a. 1000 750 Cl X ACCEFI'ABLE COOLDO%N RATES OPERATION 500 'F/HR 0

20 40 250 60 IOQ 0 50 100 150 200 250 300 350 400 450 500 INDICATKD TKLIPKRATURK (DKG.F')

Figure 8. D. C. Cook Unit 2 Reactor Coolant System Cooldown (Cooldown rates up to 100'F/hr) Limitations Applicable for the First 15 EFPY (With Margins of 10'F and 60 psig for Instrumentation Errors)

C-23

6. REFERENCES 1 Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May, 1988.

2 "Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

3 ASME Boiler and Pressure Vessel Code, Section III, Division 1-Appendixes, "Rules for Construction of Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure," pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York, 1986.

4 Code of Federal Regulations, 10CFR50, Appendix G, "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.,

Federal Register, Vol. 48 No. 104, May 27, 1983.

5 Letter Report, MT-SMART-090(89), "D. C. Cook Unit 2 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation", N. K. Ray, April 1989.

6 WCAP-8512, "American Electric Power Company Donald C. Cook Unit No. 2 Reactor Vessel Radiation Surveillance Program", J. A. Davidson, et al.,

November 1975.

7 "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events", 10 CFR Part 50, Vol. 58, No. 94, May 15, 1991.

C-24

. ATTACHMENT I DATA POINTS FOR HEATUP AND COOLDOWN CURVES (Without Margins for Instrumentation Errors)

The data points used in the development of the heatup and cooldown curves shown in Figures I and 3 are contained on the attached computer printout sheets.

C-25

AMP 60 DEG-F/HR HEATUP REG.GUIDE 1.99,REV.2 WITHOUT MARGIN 09/28/92 THE FOLLOWING DATA WERE CALCULATEOFOR THE INSERVICE HYDROSTATIC LEAK TEST.

MINIMUM INSERVICE LEAK TEST TEMPERATURE ( 32.000 EFPY)

PRESSURE (PSI) TEMPERATURE (DEG,F) 314 2485 335 PRESSURE PRESSURE STRESS 1.5 K1M (PSI) (PSI) (PSI SQ RT IN )

21444 89745 2485 26645 1 12505

AMP, 60 DEG-F/HR HEATUP REG.,GUIDE 1.99 ~,REV.2 WITHOUT, MARGIN,.'>; .: 09/28/92'

> C, COMPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2 HEATUP RATE(S) (DEG.F/HR) K 60.0 IRRADIATION PERIOD = 32.000 EFP YEARS FLAW DEPTH w (1 "AOWIN)T: -,

• 'NDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG.F) (r SI) (DEG,F) (PSI) 2 1 86.000 90.000 44~4.

Rh~ 8:

~

-';:-.--,::;.,",, 21 22

'85,000 -

190. 000 513.04 S24.65 40 41 280;000'h-':.

285:000 =- ':

?

'.102S. 52 978,92 .',

3 95. 000 537. 13 42 290.000 1075.56 100. 000 & 550.71 43 295.000 1129.28 4

5 105. 000 436z64- .-""i; 25 205. 000 565.38 44 300. 000 1 186. 80 6 "

110.000 ~ 26,," 210.000 . '681. 19 .. 45 '-, 305;,000 ", ',248. 53

', 310';.OQO..:h::" .1314.70 .

'...115.000 433.80":.'.:~::,';:,:~~~"27 215 000,'., ':, >698. 11 7

000: '; 6'16'45 46,h 8 ': 120. 000 '2~.>'. ',"'4~i'8 "" '

220

34 47,.; '31S..OQQ.':>-> -1385,56, 9 125. 000 435.53 29 225.000 636.06 48 320.000 1461. 55 10 130. 000 437.77 30 230.000 657.31 49 325.000 1542.87 135. 000 440.85 31 235.000 680. 15 50, 330.000 1629.85 1 1 12 . '40.000 "

444'.72'.-',,";-".:.'"32, '... '240.000.,;-

.704;60"...: 51'- .,;.'35'.000 -' ,

. 13 -,,146.

". -15Q;OQO 000 449.38 .':;;:=~33 ', .

"4> 250,000 245.000 '"., -',. -'730.85

"- ..::.?-'759.-27 34Q;OOQ;-';.:::;: 1822.90 1723:05'2:;

14 '454,73 ..5'c,'.34 .. -'. 53 ""'45'.OQQ:::'-.>'924.64 15 155. 000 460.84 35 255.000 789.61 54 350.000 2020.07 16 160. 000 467.63 36 260.000 822.35 55 355.000 2121.91 17 165. 000 475. 18 37 265.000 857.45 56 360.000 2230.63 18 '

170. 000 483.45',:,.'8. 270.000 -::.":-.,;.895. 10 57 .'- -'.;365.000',. 2346".74=

'75.000 19 . 492. 53 - ~ .39: 27S.-OOO -

. '";935 ~ 62 58 -,:.: 370;QOO",.:::, -2470,52-20 -

. 18Q.OQO 502.30 I

M C

?'P

?

-h =.

?

h>> '

X 5

DDLDOWN CURVES REG. GUIDE 1.99,REV.2 WITHOUT MARGIN 09/28/92 THE FOLLOWING DATA WERE PLOTTED FOR CDOLDOWN PROFILE 1 ( STEADY-STATE CDDLDDWN )

IRRADIATION PERIOD R 32.000 EFP YEARS FLAW DEPTH ~ ADWIN T INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG.F) (PSI) (DEG.F) (PSI) 1 85,000 492.98 20 180. 000 607.36 39 275.000 1058.28 2 90.000 495.89 21 185.000 618. 87 40 280 000

~ '1103.35 3 95.000 499.02 22 190. 000 631. 24 41 285. 000 1151. 78 4 100.000 502.38 23 195 . 000 644.42 42 290. 000 1203.56 5 105.000 505.99 24 200. 000 658.74 43 295. 000 1259.45 6 1 10. 000 509.88 25 205.000 674. 1 'I 44 300.000 1319. 45 7 115.000 514,06 26 210.000 690.49 45 305.000 1383.70 8 120.000 518. 55 27 215. 000 ?08,28 46 310.000 1452.62 9 125.000 523.38 28 220. 000 727.34 47 315. 000 1526.55 10 130.000 528.57 29 225. 000 747.75 48 320. 000 1605.82 11 135.000 534. 15 30 230. 000 769. 81 49 325. 000 1690.70 12 140.000 540.05 31 235. 000 V93.38 50 330.000 1781. VV 13 145.000 546,50 32 240. 000 e1&.e4 51 335.000 1&79.21 14 150.000 553.44 33 245,000 846.11 52 340.000 1983,42 15 155.000 560.90 34 250. 000 875.34 53 345. 000 2094.78 16 160.000 568.91 35 255. 000 907.00 54 350. 000 2214. 28 17 165.000 577.54 36 260. 000 940.84 55 355. 000 2341. 47 18 1VO.OOO 586.68 37 265.000 977. 14 56 360.000 247V.36 19 175.000 596.64 38 2?0. 000 1016. 15

AMP CODLOOWN CURVES REG. GUIDE 1.99,REV.2 WITHOUT MARGIN 09/28/92 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 2 ( 20 OEG-F / HR COOLDOWN )

IRRADIATION PERIOD ~ 32.000 EFP YEARS FLAW DEPTH ~ ADWIN T INDIGATED INDICATED INDICATED INDICATED ~ INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (r SI) (DEG.F) (PSI) (DEG.F) (r SI) 85.000 449. 12 15 155. 000 519.26 28 220,000 6S5.99 2 90. 000 452,03 16 160. 000 527,68 29 225,000 717.97 3 95. 000 455. 19 17 165. 000 536.76 30 230. 000 741. 44 4 100.000 458.58 18 170. 000 546.42 31 235. 000 766.88 5 105. 000 462.27 19 175. 000 556.96 32 240. 000 794.05 6 110.000 466.23 20 180.000 568.28 33 245.000 823.42 7 115.0QO 470-52 21 185. OOQ 580.49 34 250. 000 854.93 8 120.OQQ 475. 13 22. 1SO.QOO 593.50 35 255.000 888.77 9 125.000 480. 13 23 195. 000 607.65 36 260. 000 925. 12 10 130.000 485.49 24 200. 000 622.87 37 265. 000 964.44 11 135.000 491. 30 25 205. 000 639. 12 38 270. 000 1006.49 12 140.000 497.44 26 210.000 656.75 3S 275.000 1051.77 13 145.000 504. 19 27 215.000 675.73 40 280-000 1100. 35 14 150.000 511.43 I

PD

%O

0, ODLDOWN CURVES REG.

IRRADIATION PERIOD FLAW'EPTH ~ AOWIN T

=

GUIDE:1,99,REV.2 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 3 32.000 EFP YEARS WITHOUT MARGIN 0

( 40 DEG-F / HR COOLDOWN )

09/28/92 h

C, INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG. F ) (PSI } (DEG.F) (r SI) 2 85,QOO

.;...9Q.OOO 404.

407. 14 27,~,".".'5,-..;.:.:,15

';.]6,.

155.000

',160.000" 476. 97 485.84' 28, 29..

220. 000 225.00Q'0

665".35

-'88;7?

3 95.000 410.34 17 165.000 495.35 230.000 714. 14 4 100.000 413. 79 18 170.000 505. 68 31 235. 000 741.31 5 105. 000 4 17 . 56 19 175. 000 516. 85 32 240. 000 770.69 421.62 .,20 528.86 245. 000 802.22 7,

6 8

110.000 115,000

',. '120.00Q '- . 426.04 430. 80 ..

'.-','.," 21,

'.:"'22-..-.-

180.000

.185.000..

. 190 '000 ",

'.:541.74

. 555. 71, 33 34 35

-'250.000 255.000 872.54,

'36.07" 9 125.000 435.98 23 195.000 570.80 36 260-000 911.96 10 130.000 441.49 24 200. 000 586. 91 37 265.000 954.24 11 135.000 447.54 25 205.000 604.43 38 270 000 F 999.65 12 -; 140.000 ,454.06 ., ',"",:;.,"'".26,', ,210.000:.,'23.28, '

39 ,."275.000 i 1048.51 13 -'

':"145.000 -;-: 461.,12 " '.i",."27,"' .',215.'000:, 'i'643;47 . 4Q,

'&0;000,,  :" .'101':.03 14 15Q.OQO-' '68.73 nI ED

-;,'. ? c 5

AMP CODLDOWN CURVES REG. GUIDE 1.99,REV.2 WITHOUt MARGIN 09/28/92 THE FOLLOWING DATA WERE PLOTTED FOR COOLDDWN PROFILE 4 ( 60 DEG-F / HR CDOLDOWN )

IRRADIATION PERIOD = 32.000 EFP YEARS FLAW DEPTH ~ AOWIN T INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG.F) (PSI} (DEG.F) (PSI) 1 85,000 358.24 14 15Q. 000 425.29 27 215.000 611.88 2 90.000 361. 20 15 155.000 433.99 28 220.000 635.24 3 95.000 364. 41 16 160. 000 443.30 29 225. 000 660. 61 4 100.000 367.93 17 165. 000 453.48 30 230. 000 687.75 5 105.000 371. 80 18 170. 000 464.44 31 235. 000 717. 21 6 110.000 375.99 19 175.000 476.32 32 240. 000 748.73 7 115.000 380.56 20 180.000 489. 12 33 245.000 782.71 8 120.000 385.5Q 2'1 185.000 502.88 34 250.0QQ 819,44 9 125.000 390.89 22 190. 000 517. 80 35 255. 000 858.89 10 130.000 396. 71 23 195. 000 533.93 36 260. 000 901. 34 11 135.000 403.05 24 200. 000 551. 19 37 265. 000 947.04 12 140.000 409.83 25 205.000 569.96 38 270.000 S96.21 13 145,0QO 417.27 26 210.000 590,05 39 275,000 1049.09

COOLOOWN CURVES REG. GUIDE -1,99,REY.2. WITHOUT MARGIN 09/28/92 THE FOLLOWING DATA WERE PLOTTED FOR COOLOOWN PROFILE 5 ( 100 DEG-F / HR COOLDOWN)

IRRADIATION PERIOD < 32.000 EFP YEARS

",-.FLAM DEPTH AOWIN T

?

INDICATED INDICATED INDICATED INDICATED INDICATED INDICATEO TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE 1

2 -'

(OEG.F)

', 85,000 90.000 (PSI)

'62,'50, "::,,':;;;:;~:,14,

, .-'.265 58""':,"'".5."::i 15 F '

(OEG.F) 150;000 155.'000 (PSI)

'36. 15 ';

,';. 345,99,-';-'h,"."'28,;h

~

27 (DEG.F)

.h',2.15,000 '

-'.220;OOQ (PSI) 551'.05 578,41',

3 95.000 269.01 16 160.000 356.63 29 225.000 607.88 4 100.000 272.75 17 165.000 368. 16 30 230.000 639.62 276.90 170.000 380.66 235.000 674.05 5

6 7

105,000 110.000 115.000 .

281.41; ',:.":;::;.:- 19 286,36. ;;;.,-;,.":.".',20 18 175.000: '394.26;-

180.000 ; -. 408-.87'. --:

', 31 32, 33

. 240.000

'45.000 ,;750;89 -

'11;01 8, '120,000 '=,=: 291,76.-'..;",:,'.,'-';.2t" ;185.000 '".'.',.;;- 424.79-'..,':..',,.. 34 ;.,',,';:250..'000 -.793.84-,

9 125.000 297.70 ~

22 190.000 441.91 35 255.000 840.17 10 130.000 304. 13 23 195.000 460.53 36 260.000 890.04 135.000 200.000 480.63 265,000 943.83 11 12?;,:

13' "'.

'140.000:-.;.-.",.:,

'",,L45 OOO,:; ~

311. 18 318.82<:,h 327'2 "*.',"

24

-..?K~'.":i"'25 " '05;000'. '; ',

37

.502".30,:..:,.;,:':-" 38, '; j270,'OOO,

'"";"',,"26",', ','210 '000-","". ', .'25'6,""::? i'i":,'",'c, '",'"""",'.'.'; "".. ",'c."",.:.'

', '. 1001.60,';."'-

"'h>. '""',,'., 'l:i wc,

?.

h I

M h

5 h

ATTACHMENT 2 DATA POINTS FOR HEATUP AND COOLDOWN CURVES (With Margins of 10'F and 60 psig for Instrumentation Errors)

The data points used in the development of the heatup and cooldown curves shown in Figures 2 and 4 are contained on the attached computer printout sheets.

C-33

AMP 60 DEG-F/HR HEATUP REG. GUIDE 1.99. REV.2 WITH MARGINS 10/12/92 THE FOLLOWING DATA WERE CALCULATEDFOR THE INSERVICE HYDROSTATIC LEAK TEST.

MINIMUM INSERVICE LEAK TEST TEMPERATURE ( 32.000 EFPY)

PRESSURE (PSI) TEMPERATURE (DEG.F) 326 2485 347 PRESSURE PRESSURE STRESS 1.5 K1M (PSI ) (PSI) (PS I SO. RT. IN. )

22088 92529 2485 27288 115366

AMP 60 DEG-F/HR HEATUP REG. GUIDE 1.99, REV.2 WITH MARGINS 10/12/92 COMPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2 HEATUP RATE(S) (DEG.F/HR) i>0.0 IRRADIATION PERIOD = 32.000 EFP YEARS FLAW DEPTH > (1 "AOWIN)T INDIGATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG.F) (PSI) (DEG.F) (PSI) 1 85. 000 21 185.000 432.53 41 285.000 875.52 2 90.000 ,22 190.000 442.30 42 290. 000 918.92 3 95 . 000 23 195.000 453.04 43 295. 000 965.52 4 100. 000 37$ gO 24 200.000 464.65 44 300. 000 1015. 56 5 105. 000 25 205.000 477. 13 45 305.000 1069.28 6 1 10. 000 ~f669 26 210. 000 490. 71 46 310.000 1126.80 7 115.000 27 215.000 505.38 47 315.000 1188. 53 8 120. 000 28 220. 000 521. 19 48 320. 000 1254.70 9 125. 000 373.80 29 225. 000 538. 11 49 325. 000 1325.56 10 130. 000 374. 12 30 230. 000 556.45 50 330.000 1401. 55 11 135. 000 375.53 31 235. 000 576.06 51 335. 000 1482.87 12 140. 000 377.77 32 240. 000 597. 31 52 340. 000 1569.85 13 145. 000 380.85 33 245. 000 620. 15 53 345. 000 1663.05 14 '150. 000 384.72 34 250.000 644.60 54 350.000 1762.90 15 155. 000 389.38 35 255. 000 670.85 55 355. 000 1869.46 16 160. 000 394.73 36 260. 000 699.27 56 360. 000 1979.87 17 165. 000 400.84 37 265. 000 729.61 57 365. 000 2082.95 18 170. 000 407.63 38 270. 000 762.35 58 370. 000 2193. 34 19 175. 000 415. 18 39 275. 000 797.45 59 375. 000 2311.05 20 180. 000 423.45 40 280.000 835. 10 60 380. 000 2436.53

OOLDDWN CURVES REG. GUIDE 1.99. REV.2 WITH MARGINS 10/12/92 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 1 ( STEADY-STATE COOLDOWN )

IRRADIATION PERIOD < 32.000 EFP YEARS FLAW DEPTH  % AOWIN T INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG.F) (PSI) (DEG.F) (PSI)

85. OOQ 428.34 21 185. QQO 538.72 40 280.000 964.38 2 90. 000 430.89 22 190. 000 549.59 41 000 '85.

1006. 91 3 95. 000 433.55 23 195. 000 561. 28 42 290. 000 1052.62 4 100. 000 436.50 24 200. 000 573.70 43 295.000 1101.74 5 105. 000 439.67 25 205. 000 587.20 44 300. 000 1154.52 6 1 10. 000 443.08 26 210.000 601. 72 45 305.000 121'I. 18 7 1 15. OOQ 446.75 27 215.000 617.30 46 310.000 1271.92 8 120. 000 450.69 28 220. 000 633.95 47 315.000 1336.95 9 125.000 454.93 29 225. 000 651. 99 48 320. 000 1407. 01 10 130. 000 459.49 30 230. 000 671. 20 49 325. 000 1481. 90 11 12 135.000 140. 000 464.39 469.65 31

.32

'33 235.000 240.000 245.000 692.05 714. 40 738.34

'15052 330. 000 335.000 340. 000 1562.30 1648.49 1740.74 13 145. 000 475.32 .

14 150. 000 481. 30 34 250.000 764. 15 53 345.000 1839.56 15 155 . 000 487.85 35 255.000 791.84 54 350. 000 1945.08 16 160. 000 494.89 36 260.000 821. 50 55 355. 000 2058. 16 17 165. 000 502.45 37 265.000 853.59 56 360. 000 2178. 95 18 170. 000 510. 59 38 270.000 887.92 57 365. 000 2308.09 19 175. 000 519. 34 39 275.000 924.79 58 370. 000 2445.60 20 180. 000 528.62

AMP COOLDDWN CURVES REG. GUIDE 1.99, REV.2 WITH MARGINS 10/12/92 THE FOLLOWING DATA WERE PLOTTED FOR CODLDDWN PROFILE 2 ( 20 DEG-F / HR CODLDOWN )

IRRADIATION PERIOD ~ 32.000 EFP YEARS FLAW DEPTH  % ADWIN T INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE.

(DEG.F) (PSI) (OEG.F) (PSI) (DEG.F) (PSI)

85. 000 384.45 15 155. 000 445.63 29 225. 000 619.71 2 90. 000 386.97 16 160. 000 452.99 30 230.000 640.29 3 95. 000 389.72 17 165. 000 460.94 31 235. 000 662.57 4 100. 000 392.68 18 170. 000 469.47 32 240. 000 686. 41 5 105. 000 395.89 19 175. 000 478.59 33 245. 000 712. 18 6 1 10. 000 399.33 20 ~

180. 000 488.50 34 250.000 '?39.80 255.000 769.45 7

8 115.000 120.000 403.08 407. 10 ~ '-'2

-21 185.000 190. 000 499. 19 510.68 35 36 260.000 801. 54 9 125.000 411.46 23 195.000 523.07 37 265. 000 835.91 10 130.000 416. 14 24 200. 000 536.28 38 270. 000 872.80 11 135.000 421. 21 25 205. 000 550.64 39 275.000 912. 49 12 140.000 426.65 26 210. 000 566.08 40 280.000 955. 14 13 145.000 432.55 27 215.000 582.58 41 285 000 F 1001.02 14 150.000 438.79 28 220. 000 600.47 42 290.000 1050.33 I

Ca)

CODLODWN CURVES REG. GUIDE 1.99, REV.2 WITH MARGINS 10/12/92 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 3 ( 40 DEG-F / HR C(i::LDOWN )

IRRADIATION PERIOD = 32.000 EFP YEARS FLAW DEPTH ~ AOWIN T INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPI RATURE (PSI) (DEG.F) (PSI) (PSI)

PRESSUR1'ILG.F)

(DEG.F)

85. 000 339.64 15 155.000 402.68 29 225. 000 587.78 2 90. 000 342,16.-- 16 160.000 410. 40 30 230. 000 609.98 3 95. 000 344.92 17 165.000 418. 77 31 235. 000 633.75 4 100. 000 347.84 18 170.000 427.77 32 240. 000 659.48 5 105. 000 351. 09 19 175.000 437.43 33 245. 000 687.07 6 1 10. 000 354.60 20 180.000 447.91 34 250. 000 716.71 7 115.000 358.43 21 185.000 459.25 35 255. 000 748.86 8 120. 000 362.55 22 190.000 471.45 36 260.000 783.24 9 125 . 000 367.04 23 195.000 484.53 37 265. 000 820.25 10 130. 000 371.88 24 200.000 498. 71 38 270. 000 860.04 11 135. 000 377. 14 25 205.000 514.02 39 275.000 903.09 12 140. 000 382.73 26 210. 000 530.37 40 280.000 949. 17 13 145. 000 388.88 27 215.000 548. 16 41 285.000 998.74 14 150. 000 395.50 28 220. 000 567.27 42 290.000 1051. 84 I

CO

AMP COOLDDWN CURVES REG. GUIDE 1.99, REV.2 WITH MARGINS 10/12/92 THE FOLLOWING DATA WERE PLOTTED FOR COOLODWN PROFILE 4 ( 60 DEG-F / HR COOLODWN )

IRRADIATION PERIOD = 32.000 EFP YEARS FLAW DEPTH ~ AOWIN T INDICATED INDICATED INDICATEO INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG.F) (PSI) (DEG.F) (PSI) 1 85,000 293.63 15 155.000 358.94 29 225.000 556.53 2 90 000

~ 296. 15 16 160. 000 367.08 30 230.000 580.24 3 95. 000 298.93 17 165.000 375.92 31 235. 000 605.99 4 100. 000 301. 94 18 170. 000 385.38 32 240. 000 633.53 5 105. 000 305. 21 19 175. 000 395.72 33 245. 000 663.40 6 110.000 308.80 20 180. 000 406.85 34 250. 000 695.43 7 115.000 312. 73 21 185. 000 418.91 35 255.000 729.92 8 120.000 316. 99 22 190. 000 431.91 36 260 000

~ 767.02 9 125.000 321. 64 23 195.000 445.88 37 265.000 807.21 10 130.000 326.66 24 200.000 461. 02 38 270.000 850.25 11 135.000 332. 14 25 205.000 477.29 39 275.000 896.63 12 140.000 338.05 26 210. 000 494.93 40 280.000 946.47 13 145.000 344.50 27 215. 000 513. 98 41 285.000 '1000. 17 14 150.000 351.38 28 220. 000 534.37 I

CA LD

COOLDDWN CURVES REG. GUIDE 1. 99, REV. 2 WITH MARGINS 10/12/92 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 5 ( 100 DEG-F / HR COOLI .iN)

IRRADIATION PERIOD = 32.000 EFP YEARS FLAW DEPTH  % ADWIN T INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (PSI ) (DEG.F) (PSI) (DEG.F) (PSI) 1 85. 000 197. 87 15 155. 000 269. 06 28 220.00~ 470.82 2 90. 000 200. 43 16 160. 000 278.24 29 225.0Cii 496.49 3 S5. 000 203.30 17 165. 000 288.24 30 230.000 524.26 4 100. 000 206.44 18 170.000 299.05 31 235.000 554. 18 5 105. 000 209.94 19 .175.000 310. 76 32 240.000 586.40 6 1 10. 000 213.76 20 180.000 323.47 33 246.000 621. 20 7 115.000 217. 98 21 185.000 337.28 34 250.000 658.85 8 120.000 222.64 22 190.000 352. 12 35 265.000 699.35 9 125. 000 227.61 23 195.000 368.29 36 260.000 742.94 10 130.000 233. 10 24 200.000 385.67 37 265.000 789.96 11 135.000 239. 14 25 205.000 404.59 38 270.000 840.55 12 140. 000 245.69 26 210. 000 424.99 39 276.000 895. 10 13 145. 000 252.86 27 215. 000 447.00 40 280.000 953.54 14 150. 000 260.62 nI lD

ATTACHMENT 3 DATA POINTS FOR HEATUP AND COOLDOWN CURVES (Without Margins for Instrumentation Errors)

The data points used in the development of the heatup and cooldown curves shown in Figures 5 and 7 are contained on the attached computer printout sheets.

~ FOLLOWING DATA WERE PLOTTED FOR COOLOOWN PROFILE TH 1 ( STEADY-STATE COOLOOWN )

IRRADIATION PERIOD ~ 15.000,EFP,YEARS, INDICATED INDICATEO INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG F) (PSI) (OEG F) (PSI) 3 95.000 517.62 20 180.000 670.96 37 265.000 1193.06 21 185.000 687.09 38 270.000 1248.06 5,. 105.000 , , 527.50 22, , 190,000 ., 704,62 39 275.000 1307,03 9 125.000 552.01 25 210.000 788.49 43 295.000 1589.48 10 130.000 559. 36 27 215.000 813.61 44 300. 000 1673. 05 135.000 567.26 28 220.000 840.46 45 305.000 1763.06 15 155. 000 605. 15 32 240.000 969. 60 49 325. 000 2189. 57 16 160. 000 616. 50 33 245. 000 1008. 27 50 330. 000 2315. 42 17 , 165.000 . 628,.70 34 ,250,000 , .1049.65 ,51 335.000 2449.49

THE FOLLOWING DATA MERE PLOTTED FOR COOLDOWN PROFILE 2 ( 20 DEG-F / HR COOLDOWN )

,xIRRADIATION PERIOD ~h 15,000 EFP YEARS INDICATED INDICATED INDICATED INDICATED. INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE 3 95.000 475. 10 15 155.000 566.78 27 215.000 789.01 4 100.000 479'.99 16 160. 000 578. 79 28 220.000 817.93 5 ... 105,000 3, 485,30, 17 165,000, 591,61 29 225,000, 848,86 9 125.000 510.83 21 185.000 653.87 33 245.000 998. 19 10 130.000 518. 52 22 190.000 672.53 34 250.000 1042.74 11 135.000 ,, 526.82, 9....23., ,195.000 692,49 35 .,255,000 1090.58, n

I

\

b CYXvh982h h(4vX'M0hv wh4 Xh 2'X05080XNY(kX'4e/XkkvXh49khXNWhfNN)hvXXN 2(X'hh .0kX4077hXXXkvX' v774Xv

THE FOLLOWING DATA WERE PLOTTED FOR COOLOOWN PROFILE 3 ( 40 DEG-F / HR COOLDOWN )

, ., IRRADIATION PERIOD, 15,.000, EFP YEARS,,

INDICATED TEMPERATURE INDICATED PRESSURE INDICATED INDICATED, INDICATEO INDIGATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F,), , (PSI), .. . , (OEG,F), (PSI} ,, ( . ) (

3 95.000 431.68 15 155.000 528.06 27 215.000 765.79 4 100.000 436.74 16 160. 000 540. 69

..., 5 105.000,,442,17,. , 17,,165,000 554.46,29 225,000, 830.05 9 125. 000 468. 95 21 185. 000 620. 87 33 245. 000 990. 92 10 130.000 477.03 22 190.000 640.69 34 250. 000 1038. 91 11, 135.000,, 485.77 ,. 23 ,195,000 662.22 35 255.000 1090.58

THE FOLLOWING DATA WERE PLOTTED FOR COOLOOWN PROFILE 4 ( 60 DEG-F / HR COOLDOWN )

,, IRRADIATION PERIOD = 15.000 EFP YEARS..

INDICATED INDICATED INDICATEO INDICATED INDICATED INOICATEO TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F), .. .(PSI)... , ,. ,(OEG.F) ,(PSI} ,... (OEG.F,) (PSI) 3 95.000 387.30 15 155.000 489.00 26 210.000 712.86 4 100.000 392.56 16 160.000 502.47 27 215.000 743. 83 5 105.000 398 30

~ 17 165 000

~ 517 14

~ 28 220.000 777. 15 9 125.000 426.34 21 185.000 588.07 32 240.000 938.50 10 130.000 434.86 22 190.000 609.47 33 245.000 986.80 11 135.000 ,,444.03 . , 23 . ,,195,.000 632,54 34, 250.000 1038.71

THE FOLLOWING GATE WERE PLOTTED FOR OOOLOOWN PROFILE E ( IOO OEG-F / HR OOOLPOWN)

,IRRADIATION,PERIOD,>...15,000 EFP YEARS ...,

INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE

,(DEG,F)

PRESSURE

, L(PSI),, , .

TEMPERATURE

..(DEG,F)

PRESSURE (PSI,}

TEMPERATURE (DEG.F)

PRESSURE (PSI) 3 95.000 295.46 14 "

150.000 395.64 25 205.000 635.73 4 100. 000 301 . 22 15 155.000 409.97 26 210.000 669.41 5 105. 000 307. 55, 16,, 160. 000 425. 50 27 215. 000 705. 62 9 125. 000 338 . 85 20 180. 000 501 . 32 31 235. 000 880. 91 10 130.000, 348.43 21 185.000 524.32 32 240.000 933.55 11 135.000 358.88, , 22, ...,L190,000 549.02 33. 245.000 990.30,

COMPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2 HEATUP RATE(S) (DEG.F/HR) ~ 60.0 IRRADIATION PERIOD m 15.000 EFP,,YEARS INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE 3 95.000 464.74 21 185.000 572.86 39 275.000 1147.47 100. 000 460. 17 22 190.000 588. 71 40 280. 000 1206. 00 5,

4 105,000 457,67. 23 ., 195,000 .605,98 41 9 125.000 462 02

~ 27 215.000 689 27

~ 45 305.000 1569.44 10 130.000 466.00 28 220.000 714.32 46 310.000 1658. 28

,135,. 000.47,1,04 ,,29,, .,225;000 ..741; 08 .. 47, 3,15,. 000 1753,,27 15 155.000 500. 24 33 245.000 870. 13 51 335.000 2203. 37 16 160.000 509.84 34 250.000 908.57 52 340.000 2325.82 17, 165.000 , , 520.,41, 35 ,,255.,000 949,.85, , 53 ..., .345.000,, 2448.35 n

I

ATTACHMENT 4 DATA POINTS FOR HEATUP AND COOLDOWN CURVES (With Margins of 10'F and 60 psig for Instrumentation Errors)

The data points used in the development of the heatup and cooldown curves shown in Figures 6 and 8 are contained on the attached computer printout sheets.

C-48

~ THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE.1

,IRRADIATIONPERIOD = 415,000 EFP YEARS

( STEADY-STATE COOLDOWN )

INDICATED INDIGATED INDICATED INDICATED INDICATED INDICATEO TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (PEG.F) (PSI) ,, (OEG,F) , (PSI) . , ,(OEG.F) (PSI,)

3 95 000 449 08 21 185 000 595 79 39 275 000 1133 06 4 100. 000 453. 20 22 190. 000 610. 96 40 280.000 1188.06 5 105.000,, 457162,,23 ,,195,000,,,, 627..0941285.000 1247,.03 9 125.000 478.81 27 215.000 705.28 45 305.000 1529.48 10 130. 000 485. 17 28 220.000 728.49 46 310. 000 1613. 05 1 1 135.000 . 492. 01 29 225. 000 ..753. 61 47 315.000 1703. 06 15 155. 000 524. 89 33 245. 000 873. 82 51 335. 000 2129. 57 16 160.000 534.59 34 250.000 909.60 52 340.000 2255.42 17 ,. 165.000, , 545. 15 , , 35 ...255.000 ,,948,.27, 53 345,.000 2389.,49

THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 2 ( 20 DEG-F / HR COOLDOWN )

IRRADIATION PERIOD, >,,15,.000,EFP,YEARS INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE

.. (OEG,.F,), (PSI) . .. (DEG.,F), (PSI) (DEG.F) (PSI) 3 95.000 406.34 16 160.000 495.62 28 220.000 702.27 4 100. 000 4 10. 54 17 165.000 506.78 29 225.000 729.01 5 105. 000 415. 10 18 170. 000 518. 79 30 230. 000 757. 93

~

9 125.000 437.06 22 190.000 576.49 34 250.000 896.74 10 130.000 443.68 23 195.000 593.87 35 255.000 938.19 11 135.000 ~

450.83 24 200.000 612.53 36 260.000 982.74 O

I

THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 3 ( 40 DEG-F / HR COOLDOWN )

,.IRRADIATION,PERIOD,=.,15,000,EFP YEARS INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (OEG.F) (PSI) , . (OEG.~F , (PSI) 2.6 16 160.000 456.22 28 220.000 676.86 17 165.000 468.06 29 225.000 705.79 5 , 105,000...,,,371;68....., 18,. 170.000, 480...,69 30 230.000, 736.73 9

10 11 ,

125.000 130.000 135.000 .

394.53 401. 45

, 408,,95,,

22 23 24 190.000 195.000 200,.000 ,

542.32 560.87

,580,.69. ..

34 35 36 250.000 255.000

,260,.000 , 886.20 930.92 978.91 Pl I

Vl 8

THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 4 ( 60 DEG-F / HR COOLOOWN )

IRRADIATION,PERIOD ~,15,000 EFP YEARS

'.;yg ':,'": .':': '~."":"'";","'

INDICATED

',;; -'~ ". "': """-""":,"~'" "'

INDICATED INDICATED

' "'~ "":..':" . g: . ' '

INDICATED

". ""'rS INDICATED

~i 4'NDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE

,, ,(DEG,F) ...(PSI,),,...,,...,.. . ,(DEG.F), , (PSI) . (DEG,F) .. ..(PSI),

3 95. 000 317. 96 15 155. 000 404. 78 27 215. 000 623. 97 4 100.000 322.43 16 160.000 416.41 28 220.000 652.86 5 105.000 327.30 17 165.000 429.00 29 225.000 683.83 9 125.000 351.18 21 185.000 489.93 33 245.000 833.66 10 130.000 358.45 22 'I90.000 508.32 34 250.000 878.50 135.000, , 366.34 ,, 23 195,000 528.07 35 255.000 926.80 I

Ql

~

THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 5 ( 100 DEG-F / HR COOLDOWN)

IRRADIATION PERIOD,,~ ,15,,000EFP YEARS,

~

INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (PEG.F) (PSI) , , ,(OEG,.,F) (PSI) <OEG.F

~

3 95.000 225.32 15 155.000 322.38 27 215.000 575.73 4 100. 000 230. 14 16 160.000 335.64 28 220.000 609.41 5 ... 105.,000 . 235.46 , , ,17 ,.,165,.000 349,97 29 9 125.000 261.91 21 185.000 420.16 33 245.000 820.91 10 130.000 269.99 22 190.000 441.32 34 250.000 873.55 11 135.,000 , . 278.85 23 195.000 464.32, 35 255,000 930.30

COMPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2 HEATUP RATE(S) (OEG.F/HR) 22 60.0

~ IRRADIATION,PERIOD < 15,0009 EFP, YEARS, INDICATEO INDIGATED INDICATED INDICATED INDIGATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE

, (DEG.F) (PSI). ... (DEG.F) (PSI) (DEG.F) (PSI) 3 95.000 420.98 22 190.000 498.06 40 280.000 1032.73 4 100.000 411.42 23 195. 000 512. 86 5 105. 000,, 404,74 ,,24,, 200,000 v 528,,71 v 41 42 285.000 290,.000,,1 1087. 47 146. 00 10 vv" 9

4 r;.

125.000

~ -:;

130.000 397.27 399.03 m" 5~ -v r 28 29 y "4 .. w"" . v.3.y40yr2v4r 220.000 225.000 V. 5vrr444yv4va 606. 15 629.27

" "y4 "

• 44vr 46 47

" ~

~ 310.000 315.000

" "... ~ .:.v, 1426.22 1509.44

~ ~ ~

.v.:: . awe r r 15 155.000 423.56 34 250.000 774.24 340.000 16 'l60.000 165. 000,

'31.62 35 255.000 810. 13 52 53 345.000 2019.60 2143. 37 17,, , 440124,0..6.,36,,260.,000.,. .4848...57v ,., 7544, 350,000 9,2265. 82, Vl

9304220175 Attachment 1 to AEP:NRC:1181 Reasons and 10 CFR 50.92 Significant Hazards Evaluations for Changes to the Technical Specifications for Donald C. Cook Nuclear Plant Unit: 2

,f to AEP:NRC:1181 Page 1 As discussed in the cover letter, the purpose of this proposed amendment is to prevent a surveillance outage before our next refueling outage, currently scheduled to begin August 6, 1994. This submittal requests extensions for surveillances that must be performed during shutdown or that present such operational difficulty that performing the surveillance is not practical at power. We propose to add the following Technical Specification (T/S) to Section 4.0 of the T/Ss.

4.0.8 By specific reference to this .section, those surveillances which must be performed on or before August 13, 1994, and are designated as 18-month or 36-month surveillances (or required as outage-related surveillances under the provisions of Specification 4.0.5) may be delayed until the end of the cycle 9-10 refueling outage. For these specific surveillances under this section, the specified time intervals required by Specificati'on 4.0.2 will be determined with the new initiation date established by the surveillance date during the Unit 2 1994 refueling outage.

We reference this Specification by footnote in all surveillances that require this extension. This footnote will be applicable to the following T/Ss with the indicated surveillance due date. Dates given include the grace period allowed by T/S 4.0.2.

Descri tion of Chan e Due Date (1) 4.3.1.1.3 Delay time-response testing for 01/02/94 4.3.2.1.3 reactor trip and engineered limiting safety features instrumentation due date (2) 4.5.1.d Delay testing for equipment 04/15/94 4.5.2.e response to ESF signals (safety limiting 4.6.2.1.c injection, containment pressure due date 4.6.2.2.c high-high, containment isolation 4.6.3.1.2 phase A and B and purge exhaust) 4.7.1.2.e 4.7.1.2.f 4.7.3.l.b 4.7.4.l.b 4.7.5.1.e.2 4.7.6.l.d.3 (3) Table 4.3-2<, Delay auxiliary feedwater system 05/05/94.

Item 6.d testing including channel 4.7 '.2.e functional testing of loss of 4.7.1.2.f main feedwater pump signal to AEP:NRC:1181 Page 2 Descri tion of Chan e Due Date (4) 4.8.1.1.2.e Delay diesel generator testing 03/25/94.

4.8.1.2 including relief valve testing limiting 4.4.11.3 and essential service water valve due date 4.7.4.1.b testing (5) Table 4.3-1>, Delay RTD calibrations 04/28/94 Items 7 & 8 4.3.2.1.2 (P-12)

Table 4.3-2>,

Item 4.d Table 4.3-6A Items 5, 6, 7 & 8 Table 4.3-10, Items 2, 3, 11 (6) Table 4.3-1>, Delay pressurizer pressure & 01/29/94 Items 7, 9, 10 & ll Table 4.3-2>, Item l.d level calibrations, interlock function testing, and PORV 4.3.2.1.2 (P-11) calibrations 4.4.11.1.b (7) Table 4.3-10, Item 16 Delay Reactor Vessel Level 04/20/94 Indication Syst: em Calibration (8) 4.1.3.3 Delay analog rod position 05/03/94 indication functional testing (9) 4.5.2. d. 1 Delay RHR auto-closure interlock 03/07/94.

4.5.3.1 testing (10) 4.7.7.l.a Delay visual inspection of 03/19/94 inaccessible snubbers Table 4.3-1<, Item 5 Delay intermediate range 01/17/94 4.3.1.1.2 (P-6) calibration and interlock functional testing (12) 4.6.5.9 Delay divider barrier seal 03/08/94 inspection (13) 4.7.9.2.b.l Delay RCP fire protection testing 03/30/94 (14) Table 4.3-10, Item 18 Delay containment water level 01/31/94 4.5.2.d.2 calibrations and"sump visual limiting 4.5.3.1 inspection due date (15) 4.2.5.2 Delay reactor coolant flow 01/28/94 Table 4.3-1>, calibrations Items 12 & 13 to AEP:NRC:1181 Page 3 Descri tion of Chan e Due Date (16) Table 4.3-2>, Delay ESF Manual Trip Actuating 04/15/94 Items 9.a, 9.b, Device Operational Test limiting 9.c & 9.d due date

> Tables 4.3-1 and 4.3-2 refer to T/S 4.3.1.1.1 and T/S 4.3.2.1.1, respectively.

A description of the proposed changes, the reasons for the changes, and our analyses concerning significant hazards considerations for each group of extension requests are given in the remainder of this attachment. It is worth noting that two similar extension requests for the Unit 2 Cycle 6-7 outage were approved by the NRC on December 28, 1987 and February 29, 1988 via Amendments 97 and 99, respectively. These two amendments grant:ed extensions for the T/Ss described in groups 1 through 9 and 16, above.

1 and 2 'eact:or Tri and ESF Res onse Testin We are requesting extensions for the time-response testing required by T/Ss 4.3.1.1.3 and 4.3.2.1.3 for the reactor trip and Engineered Safet:y Features (ESF) instrumentation in T/S Tables 3.3-1 and 3.3-3. In addition, we are requesting extensions for surveillance requirements involving equipment that actuates on an ESF signal (see table below). These surveillances in many cases involve the same equipment and are performed in part to satisfy the response time testing of T/Ss 4.3.1.1.3 and 4.3.2.1.3.

These additional surveillances, the affected components, and the respective ESF actuation signals are as follows:

It:em ~TS Com onents ~ESP Si nal 4.5.l.d accumulator isolation valves SI 2 ~ 4.5.2.e ECCS automatic valves SI centrifugal charging pump SI safety injection pump SI residual heat removal pump SI 4.6.2.1.c containment spray automatic containment pressure valves and pumps high-high 4, 4.6.2.2.c spray, additive system automatic containment=pressure valves high-high 4.6.3.1.2 containment isolation valves Phase A isolation Phase B isolation containment purge and exhaust containment purge and valves exhaust isolation

6. 4.7.1.2.e,f auxiliary feedwater automatic various valves and pump starting See T/S Group (3) to AEP:NRC:1181 Page 4 Item ~TS ~Gom onents ESF Si al

7. 4.7.3.1.b component cooling water SI automatic valves

8. 4.7.4.1.b essential service water SI automatic valves See T/S Group (4)

9. 4.7.5.l.e.2 control room ventilation SI Phase A isolation

10. 4.7.6.1.d.3 ESF ventilation containment pressure high-high The extensions are needed from January 2, 1994 (most limiting surveillance due date), until the Unit 2 refueling outage.

At the Cook Plant, response time testing is performed in several parts. The portions of circuitry from the transmitter to the bistable, from the bistable to the master relay contact, and from the master relay contact to equipment operation are tested separately. Testing of the complete portion from the transmitter to the master relay contact cannot be performed at power without violating the T/Ss or adversely impacting plant operation, i.e., reactor trip.

T/Ss 3.3.1.1, 3.3.2.1 and 3.0.3 require the plant to be shut down if sufficient reactor trip or ESF instrumentation is not operable. Both trains (all channels) of the function being tested must be taken out of service during this test because the same test signal goes into both trains, which generates a reactor trip signal or ESF actuation. Should they not be in test, each signal would initiate protective functions such as safety injection and containment spray.

Therefore, the portion of the time-response tests from the bistable up to the master relay must be done during shutdown. However, testing from the transmitter to the bistable can be performed at power and will be prior to its surveillance due date. The balance of the equipment, i.e., from the master relay contact to equipment operation, is tested as part of the surveillances listed in the table above. Of these surveillances, Items 2 through 8 are specifically required by T/Ss to be performed during shutdown. Items 1, 9 and 10 are not specifically prohibited by T/Ss from being performed at power. However, to do this testing (as well as the other testing listed in the table) would require us to remove an entire train of safety equipment from operation (with the exception of the specific equipment being tested). Because this removes a layer of protection built into the plant, and because it it involves operating the plant in an abnormal is not considered prudent to perform this testing at power.

configuration, The surveillance history of these ESF systems shows that we have no reason to believe that there may be any failures in meeting the T/S requirements due to equipment degradation during the extension period. Additionally, we note that the ESF and reactor protection system channels are subjected to a T/S required surveillance program of channel checks and channel functional tests. All required channel checks and channel functional tests will continue to be performed. We believe these additional tests provide 'ndication of the operability of the systems, and would provide indication of significant degradation.

to AEP:NRC:1181 Page 5 10 CFR 50.92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment, does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Criterion 1 Based on our review of past test data, and the fact that the equipment is subject to a surveillance program which includes channel checks and channel functional tests, we believe the extensions we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability of occurrence or consequences of a previously analyzed accident, but the results, of which are within limits established as acceptable.

For the reasons detailed above, we believ'e this change falls within the scope of this example. Therefore, we believe this change does not involve significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 6 3 Auxilia Feedwater Pum Testin T/S Table 4.3-2 Item 6.d requires a channel functional test of the motor driven auxiliary feedwater pump start on loss of main feedwater pump signal to be performed on an 18 month basis. To perform thi's testing during power operations would involve tripping at least one main feed pump, which would result in a reduction of power and cause a thermal transient to be imposed on the plant.

T/Ss 4.7.1.2.e & 4.7.1.2.f require testing to demonstrate that the motor- and turbine-driven auxiliary feedwater pumps start and that the associated automatic valves actuate to their correct position upon receipt of the appropriate signal as listed in T/S Table 4.3-2. Per T/Ss 4.7.1.2.e & 4.7.1.2.f, this testing must be performed during shutdown. These extensions are needed from May 5, 1994, until the Unit 2 refueling outage.

Based on the above, we cannot perform these surveillances while at power.

However, in practice, the essential portions of these T/Ss (that is, startup of the auxiliary feedwater pumps when required and movement of the valves to their correct position) occur when the unit trips. The last reactor trip occurred on July 2, 1992. Prior testing experience with regard to these surveillances has indicated no significant problems when the surveillance,was performed. Although we recognize that not all the actuation circuitry has been challenged as a result of the reactor trip, we feel that our recent experience, in conjunction with the excellent test history in this area, justifies our request to extend the surveillance interval.

10 CFR 50 92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

to AEP:NRC:1181 Page 7 Criterion 1 As discussed above, portions of the system have undergone a challenge due to a recent actuation (during a unit trip). This fact, coupled with our excellent test history for these surveillances, leads us to believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. Ve believe this change falls within the scope of this example, for the reasons cited above.

Thus, we believe this change does not involve significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 8 4 Diesel Generator Testin An extension of the surveillance interval is requested for= the surveillance requirements of T/S 4.8.1.1.2.e. These surveillances are required by T/Ss to be performed during shutdown. The requirements include subjecting the diesel to an inspection in accordance with manufacturer's recommendations, as well as testing to verify that the diesel generator and its associated circuitry are capable of energizing, sequencing and shedding the emergency loads upon receipt of the appropriate signal. An extension of the surveillance interval is also necessary for part of the requirements of T/S 4.8.1.2, since 4.8.1.1.2 is referenced there.

The extension is needed from March 25, 1994 (limiting due date), through the Unit 2 refueling outage.

During the four and a half month period from March 25 until the start of the outage, each diesel generator should accumulate 5 additional starts and 5-7 additional running hours. The affect that these additional starts would have on the diesel generators is believed to be insignificant based on the wear history of each machine. Thus, we believe the additional starts do not constitute sufficient need to perform the subject surveillances prior to the proposed extended date. The history of diesel generator repairs from the past few years do not indicate any problem areas which, in our judgement, would be significantly affected by the proposed surveillance interval extension. Furthermore, conditions which have required maintenance on the diesel generators have been corrected at the time of discovery and have not required deferral until an outage (i.e,, we should not be deferring any significant maintenance items through the extension period). Currently, we have a trending program for the parameters measured during our T/S required monthly testing. These trends are reviewed by our diesel generator system engineer. If an adverse trend began to develop, the preventive/corrective measures would be taken to prevent a significant problem from occurring. Also, a review of previous test results did not indicate any reasons to suspect that the diesel generator associated circuitry (i.e.,

energizing, sequencing, and shedding the various emergency loads) would not pass required surveillance tests with the surveillance interval extended. Based on the above, we believe that there is no reason to suspect that the diesel generators would not be capable of performing their safety functions as required by the T/Ss.

Two other extensions related to the diesel generators are also necessary to avoid a shutdown. These are for the requirements of T/Ss 4.4.11.3 and 4.7.4.l.b. T/S 4.4.11.3 requires testing of the emergency power supply for the power operated relief valves (PORVs) and their associated block valves. Since this testing involves cycling the PORVs and block valves, .it is generally performed during shutdown and in conjunction with the diesel generator testing requirements of T/S 4.8.1.1.2.e, as suggested by T/S 4.4.11.3. T/S 4.7.4.1.b involves testing automatic valves in the essential service water (ESW) system. Per T/Ss, this surveillance testing must be performed during shutdown. Since some of the ESW valves which are required to be tested involve cooling water for the diesel generator and its associated equipment, this testing is generally conducted in conjunction with the diesel generator testing of T/S 4.8.1.1.2. The extension for both the ESW valves and the PORV emergency power supply are needed for the period of April 15, 1994 through the Unit 2 refueling outage. Previous test to AEP:NRC:1181 Page 9 results do not indicate any reason to suspect that the valves and their associated circuitry would not pass the required surveillance with the extended interval.

10 CFR 50 92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Criterion 1 For the diesel-generator machinery, the extension will result only in approximately 5 additional starts and 5 to 7 additional run hours. This is considered insignificant with regard to the wear history of each machine. For the diesel-associated circuitry, the ESW automatic valves, and the PORV emergency power supply, our review of previous test data has not indicated any reason to believe the equipment would not pass the required surveillance tests with the extended interval. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will in a margin of safety.

it result in a significant reduction Criterion 2 This extension will not result. in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability of occurrence or consequences of a previously analyzed accident, but the results of which are clearly within the limits established as acceptable. We believe these changes fall within the scope of this example.

Therefore we believe this change does not involve significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 10 5 RTD Calibrations Extensions are requested for the calibration of resistance temperature detectors (RTDs). The extensions are needed from April 28, 1994, until the Unit 2 refueling outage. The T/S surveillances involving the RTD calibration are listed below.

Re'irement 4.3.1.1.1, Table 4.3-1, Item 7 OTAT Channel Calibration 4.3.1.1.1, Table 4.3-1, Item 8 OPAT Channel Calibration 4.3.2.1.2 (P-12) Total Interlock Function Testing 4.3.2.1.1, Table 4.3-2, Item 4.d Steam Flow in Two Steam Lines--

High Coincident with T,~--Low-Low Channel Calibration 4.3.3.5.1 Table 4 '-6A, Items 5 & 7 Calibration of Appendix R Remote Shutdown Monitoring Ins trumentat ion Reactor Coolant Loops (2 & 4) Temperature (Cold) 4.3.3.5.1 Table 4.3-6A, Items 6 & 8 Calibration of Appendix R Remote Shutdown Monitoring Instrumentation Reactor Coolant Loops (2 & 4) Temperature (Hot) 4.3.3.6 Table 4.3-10, Item 2 Calibration of Post-Accident Monitoring Reactor Coolant Outlet Temperature - Tao>

Channel 4.3.3.6 Table 4.3-10, Item 3 Calibration of Post-Accident:

Monitoring Reactor Coolant Inlet Temperature - Tco<z Channel 4.3.3.6 Table 4.3-10, Item 11 Calibration of Post-Accident Monitoring Reactor Coolant System Subcooling Margin Monitor Channel

to AEP:NRC:1181 Page 11 The extensions requested in this category are for the calibration of the sensors only. The calibration procedure requires data to be taken at RCS temperatures ranging from approximately 250'F through operating temperatures. This procedure cannot be performed at power because of the low temperatures necessary for the calibration and because isothermal conditions throughout the RCS are required.

The channels involved with the RTDs are subject to T/S required channel checks and/or channel functional tests. This testing, which will continue during the extension period, would be expected to provide indication of RTD drift. Also, since narrow range RTDs feed the hT circuits, comparisons of hT~ to the calorimetric calculated power or power range detectors should show drift in the narrow range RTDs. We have found RTDs at the Cook Nuclear Plant to be very stable, and have not experienced significant drifting problems. For all of these reasons, we have no reason to believe that the RTDs will not remain operable during the extension period.

10 CFR 50.92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1), involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Criterion 1 The RTDs at the Cook Nuclear Plant have traditionally been very stable. Several independent instruments are available which would allow us to notice drift of the RTDs. Also, channels involving the RTDs are subject to T/S required channel checks and/or channel functional tests, which will continue, to be performed during the extension period. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

to AEP:NRC:1181 Page 12 Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. We believe this change falls within the scope'f this example, for the reasons cited above.

Thus, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 13 6 Pressurizer Pressure & Level Calibrations and PORV Calibrations We are requesting an extension for the performance of some of the pressurizer channel calibrations (pressurizer pressure instruments NPS-153 & NPP-153, and pressurizer level instrument NLP-153) and interlock testing involving the pressurizer pressure instrumentation. We are also requesting relief for the calibration of the PORVs. The extensions are needed from January 29, 1994, until the Unit 2 refueling outage. The affected T/Ss are as follows:

Re uirement 4.3.1.1.1, Table 4.3-1, Item 7 Calibration for OTAT Reactor Trip.

4.3.1.1 1, Table 4.3-1, Item

~ 9 Calibration for Pressurizer Pressure-Low Reactor Trip 4.3.1.1.1, Table 4.3-1, Item 10 Calibration for Pressurizer Pressure-High Reactor Trip 4.3.1.1.1, Table 4.3-1, Item 11 Calibration for Pressurizer Water Level-High Reactor Trip 4.3.2.1.1, Table 4.3-2, Item l.d

~ ~ ~ ~ ~ Calibration for Pressurizer Pressure-Low ESF Actuation 4.3.2.1.2 (P-11) Interlock Total Function Testing 4.4.ll.l.b Calibration of Power Operated Relief Valves Performance of this calibration is not considered to be prudent at power due to the configuration of the pressurizer pressure and level instrumentation. Two of the pressurizer pressure instruments (NPS-153 and NPP-153) share a common sensing line with one of the pressurizer level instruments (NLP-153). Calibrating either NPS-153 or NPP-153 poses the risk of perturbing the, input to the other transmitter, which could result in a trip. Calibrating NLP-153 poses the risk of perturbing the input to NPS-153 and NPP-153 transmitters,'hich also could result in a trip. The exemption 'for the PORVs is also needed because the calibrations make all three PORVs inoperable at the same time, which is contrary to the requirements of T/S 3.4.11.

As discussed in the previous paragraph, certain channels of pressurizer pressure and level instrumentation pose a threat to tripping the reactor. However, there are three channels of pressurizer level instrumentation and four channels of pressurizer pressure instrumentation of which two level and two pressure channels of instrumentation can, and will, be calibrated as required by the Technical Specifications. Thus, two of the three pressurizer level and two of the four to AEP:NRC:1181 Page 14 pressurizer pressure channels will satisfy the T/S surveillance requirements.

Also, the instrumentation channels for which we are requesting surveillance interval extensions are subject to T/S required channel functional testing and/or channel checks. The channel functional tests we perform are far more stringent than required. These tests not only demonstrate channel functionality, but also verify calibration of trip setpoints, actuations and alarms. The only portion of the channel that is not tested is the sensor, which is qualitatively verified during channel checks. Thus, the testing we will continue to perform would be expected to provide indication of the operability of the systems, and would provide indication of significant degradation. Lastly, we note that based on our review of the surveillance history, we believe this equipment will remain operable during the extension period.

10 CFR 50 92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

li (1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Criterion 1 Completing the required T/S surveillances on two of the three pressurizer level channels and two of the four pressurizer pressure channels will ensure that the maj ority of the equipment is calibrated as required. Also, the applicable channel functional tests and channel checks should ensure that these systems will perform as designed. Additionally, based on the surveillance history of the equipment, we believe that the equipment will remain operable during- the extension period. We therefore believe the extension we are requesting will not result in deterioration to the extent that the equipment cannot perform its intended function. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of .safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, t'e extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

to AEP:NRC:1181 Page 15 Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples.(48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes that may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. We believe this change falls within the scope of this example for the reasons cited above. Thus, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 16 7 Reactor Vessel Level Indication S stem An extension is requested for the channel calibration of the Reactor Vessel Level Indication System (RVLIS) required by T/S Table 4.3-10, Item 16. The required calibration cannot be performed at power because work must be performed in the lower volume of containment and reactor head area, which are only accessible when the unit is shut down. The extension is needed from April 20, 1994, until the Unit 2 refueling outage.

RVLIS has two trains of indication that are subjected to T/S required monthly channel checks which we will continue to perform during the extension period.

These channel checks provide indication of the operability of the system, and would be expected to provide indication of significant degradation of the system.

Our review of the maintenance history of the system gives us no reason to believe the system would be inoperable during the extension period. Additionally, indication of inadequate core cooling can be obtained by observing core exit thermocouple readings or by checking the subcooling margin monitor. These are the methods the operators would have used to assess inadequate core cooling prior to having RVLIS. We also note that there are annunciators which indicate failure of RVLIS. For these reasons, we believe that the extensions we are requesting will not adversely impact the ability of this equipment to perform its safety function.

10 CFR 50 92 Criteria

~

Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration~

if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Criterion 1 The equipment is subject .to normal surveillances which would be expected 'to provide indication of significant degradation. Also, other instrumentation is available which also provides indication of inadequate core cooling. Lastly, the past maintenance history of the equipment gives us no reason to believe that the equipment would be inoperable during the extension period. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

F to AEP:NRC:1181 Page 17 Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability of occurrence or consequences of a previously analyzed accident, but the results of which are within limits established as acceptable.

For the reasons detailed above, we believe'this change falls within the scope of this example. Therefore, we believe this change does not involve significant hazards consideration as defined in 10 CFR 50.92.

Attachment 1 to AEP:NRC:1181 Page 18 8 Rod Position Indication S stem

, This change would delay functional testing of the rod position indicator (RPI) channels required every 18 months by T/S 4.1.3.3. The extension is needed from May 3, 1994, until the Unit 2 refueling outage. Although T/S 4.1.3.3 is only applicable in Modes 3, 4, and 5, we believe relief is needed from this T/S to continue operation in Modes 1 and 2 since T/S 3/4.1.3.2 requires the RPI channels to be operable in these modes.

The surveillance we perform to satisfy T/S 4.1.3.3 is actually a calibration of the RPI channels over the rod insertion range. Since rods must be inserted to perform the calibration, it cannot be performed at power because to do so would violate the rod insertion limits of T/Ss 3.1.3.5 and 3.1'.3.6.

The operability of the RPI channels is functionally verified once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per T/S 4.1.3.2 by comparison to the demand position indication system. Also, during the 31 day surveillance to satisfy T/S 4.1.3.1.2, the rods are moved at least eight steps and the RPI meters are verified to track with the demand position. These comparisons would be expected to indicate significant degradation in the RPI channels. Surveillances that indicate the core is performing as designed are provided by the incore flux maps,'which are taken at least once every 31 effective full power days to satisfy the requirements of T/Ss 4.2.2.2 (Fo(Z)), 4.2.3 (F~H) and 4.2.1.4 (Axial Flux Difference Target Band).

Core performance is also indicated by the excore detectors, which are used to measure the quadrant power tilt ratio per T/S 4.2.4 and axial flux difference per T/S 4.2.1.1.a. These surveillances would be expected to indicate significant discrepancies between indicated and actual rod position. Lastly, since the T/S required surveillances used to verify operability of the RPIs will continue to be performed during the extension period, there is no reason to believe that we would be operating outside the bounds of T/S 3.1.3.2.

10 CFR 50 92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Attachment 1 to AEP:NRC:1181 Page 19 Criterion 1 T/S required comparison of the RPI channels to the demand position indication system would be expected to indicate significant degradation in the RPI channels.

In addition, other surveillances such as the determination of the quadrant power tilt ratio, axial flux difference and incore flux mapping surveillances, provide

... a comparison of core performance to design and would be expected to indicate significant deviations of the rods from their indicated position. Since operability of the RPIs will continue to be determined with our T/S required surveillances during the extension period, there is no reason to believe that we would be operating outside the bounds of T/S 3.1.3.2. For these reasons, we believe the extension we are requesting w'ill not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. We believe this change falls within the scope of this example, for the reasons cited above.

Thus, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 20 9 RHR Auto-Closure Interlock We are requesting an extension for the residual heat removal (RHR) auto-closure interlock test required by T/S 4.5.2.d.l. An extension is also requested for T/S 4.5.3.1 since it references T/S 4.5.2. The extensions are needed from March 7, 1994, until the Unit 2 refueling outage. The RHR auto-closure interlock automatically isolates the RHR system from the RCS if RCS pressure is above 600 it is psig. In order to demonstrate operability of the auto closure interlock, necessary to open the RHR isolation valves in the cooldown line from the hot leg in order to verify that the valves would automatically close with the RCS pressure above 600 psig. This cannot be accomplished with the unit operating (i.e., with the RCS fully pressurized) because it would result in exposing the RHR system to pressures higher than the RHR safety valves'etpoints; Previous surveillance testing has demonstrated that the auto-closure interlock is very reliable. The previous test results give us no reason to believe the auto-closure interlock would be inoperable during the extension period. The calibration for the RCS wide-range pressure transmitters, which provide input into the interlock, can be done at power and will be performed by its September 9, 1993 due date. Thus, the only portion of the interlock for which the surveillances will not be current is the portion from the bistable of the RHR suction valves through valve operation. Additionally, we note that when the unit is operating (i.e., not on RHR), the RHR suction valves are closed and procedures require power to be removed from the valve operators. This precludes inadvertent valve opening and thus alleviates the need for the auto-closure interlock to function.

10 CFR 50.92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

to AEP:NRC:1181 Page 21 Criterion 1 The surveillance test history of the auto-closure interlock has shown that the system is highly reliable, and gives us no reason to believe the equipment would be inoperable during the extension period. The wide-range pressure transmitters, which provide input into the auto-closure interlock, will have a current calibration. Additionally, we note that when the RHR system is not in service, power is removed from the suction valve operators, thus preventing inadvertent valve opening and eliminating the need for the auto-closure interlock. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. Ve believe this change falls within the scope of this example, for the reasons cited above.

Thus, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 22 10 Visual Ins ection of Inaccessible Snubbers This change would delay visual inspections of, inaccessible snubbers required by T/S 4.7.7.1.a. The extension is needed from March 19, 1994, through the Unit 2 refueling outage. The extension is required because, by definition in T/S 4.7.7.l.a and Table 3.7-9, these snubbers are inaccessible during reactor operation, thus requiring the inspections to be performed during shutdown. Note that functional testing of snubbers per T/S 4.7.7.1.c is not required until after the scheduled refueling outage start date.

In the past ten years of visual inspections on Unit 2 inaccessible snubbers, we have found only one inoperable snubber. The inoperable snubber was discovered during the steam generator outage in 1988. Since then, four visual inspections have been performed on the inaccessible snubbers, in which none have been found to be inoperable. Based on these inspection results, we are allowed to perform the inspections at the maximum allowed T/S frequency of 18 months (125X).

It should be noted that we submitted a request in our letter AEP:NRC:1143, dated May 1, 1992 to permanently change the surveillance intervals for snubber visual inspections. The submittal is based on guidance from Generic Letter 90-09, "Alternate Requirements for Snubber Visual Inspections Intervals and Corrective Actions." If we could apply the guidance of Generic Letter 90-09 or our proposed new Specifications on our current visual inspection results of inaccessible we would have up to the maximum 48 month interval allowed for our next 'nubbers, inspection. This would put our inspection due date beyond the scheduled refueling outage start date, thus eliminating the need for this extension. Based on the history of our inaccessible snubbers and on the guidance of Generic Letter 90-09, we believe the inaccessible snubbers will remain operable during the extension period.

10 CFR 50.92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

to AEP:NRC:1181 Page 23 Criterion 1 Our surveillance history of visual inspections on inaccessible snubbers has found only one inoperable snubber in the past ten years. Also, if Generic Letter 90-09 guidance is applied, our surveillance interval would be 48 months. Based on the above, we have no reason to believe the inaccessible snubbers will be inoperable during the extension period. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability of occurrence or consequences of a previously analyzed accident, but where the results are within the limits established as acceptable.

We believe this change falls within the scope of this example, for the reasons cited above. Therefore, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 24 11 Intermediate Ran e Detector Calibrations This change would delay the calibration of the intermediate range (IR) detectors required by T/S 4.3.1.1.1, Table 4.3-1, Item 5. Also, it would delay interlock functional testing of P-6 required by T/S 4.3.1.1.2. These extensions are needed from January 17, 1994, until the Unit 2 refueling outage.

The need for this extension is because the IR detectors cannot be calibrated while at power. The calibration requires that a test signal covering the range of 10 to 10 amps be superimposed over the existing current. Since the current of the IR detectors is in the 10 amps range during power operation, a 4

superimposed signal less than that could not be observed. Therefore, the IR detectors could not be calibrated below the actual current at which we are operating.

Past operating history in Unit 2 has shown that the IR detectors have performed without serious degradation. There is no reason to believe that the detectors would be inoperable during the extension period. The IR is subjected to a T/S required channel check every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The IR currents are trended daily and normalized to 25X power to ensure that the high flux at low power trip set points do not become nonconservative. Through the channel checks and trending program, it is expected that any degradation in an IR detector would be noticed. Lastly, the protection provided by these detectors is required while shut down, or at low power (approximately less than 10X) and not at our normal operating power.

10 CFR 50 92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluati'on of the proposed change with respect to these criteria is provided below.

to AEP:NRC:1181 Page 25 Criterion 1 Our operating history of the Unit 2 IR detectors have shown that they are highly reliable, and give us no reason to believe they would be inoperable during the extension period. Our channel checks and trending program would detect degradation in an IR detector. For these reasons, we believe the extension we .

are requesting will not result in a significant increase in the, probability or consequences of "'a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. We believe this change falls within the scope of this example, for the reasons cited above.

Thus, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 26 12 Divider Barrier Seal Ins ection T/S 4.6.5.9 requires a visual inspection of at least 95X of-the seal's entire length. Also, it requires that two test coupons be removed from the seal for testing to ensure the physical properties are within specified limits. Per this Specification, the inspection is to be performed while shut down. The extensions are needed from March 8, 1994, until the Unit 2 refueling outage.

The divider barrier seal is a passive design feature, thus it is not subjected to any outside forces other than the environment. During the cycle 7-8 refueling outage, we replaced 100X of the divider barrier seal. Our subsequent inspection, during the last outage, revealed no degradation of the seal. Also, when the test coupons were subj ected to the tensile strength and elongation tests, they satisfied the acceptable physical property requirements. Based on the facts that the divider barrier seal is passive, new, and has shown no degradation, we believe there is no reason to suspect that it would not be operable during the extension period.

10 CFR 50 92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Criterion 1 The divider barrier seal is a passive design feature which was entirely replaced in 1990. Our subsequent inspection revealed no degradation to the seal and the physical properties of the test coupons were acceptable. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

to AEP:NRC:1181 Page 27 Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. We believe this change falls within the scope of this example, for the reasons cited above.

Thus, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 28 13 Reactor Coolant Pum RCP Fire Protection T/S 4.7.9.2.b requires that the RCPs fire protection system be functionally tested every 18 months. In order to perform the test, the RCP fire detection instrumentation required per T/S Table 3.3-11 and the fire suppression system required by T/S 3.7.9.2 must be made inoperable, which is not considered prudent during operation of the RCPs. It is also noted that, since the RCPs are located in a high radiation area, a firewatch cannot be established per Action Statement B of T/S 3.7.9.2. Therefore, we would be forced to rely on closed 'circuit television coverage as a substitute for the continuous firewatch. In the event that a camera failed, we would be in non-compliance with the requirements of T/S 3.7.9.2. The extension is need from March 30, 1994, until the Unit 2 refueling outage.

Based on the past RCP sprinkler system surveillance history, there is no reason to believe that it would not be capable of performing it's intended safety function during the extension period. Also, we have seismically qualified oil collection systems on the RCPs, installed in accordance with 10 CFR 50, Appendix R. These systems are designed to mitigate the effects of a RCP lube oil leak.

10 CFR 50 92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Criterion 1 Based on the RCP fire protection system surveillance record there is no reason to believe that it would not be capable of performing it's intended safety function. Additionally, it is noted that the RCP oil collection system is designed to mitigate the effects of' RCP lube oil leak. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

to AEP:NRC:1181 Page 29 Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. We believe this change falls within the scope of this example, for the reasons cited above.

Thus, we believe'his change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 30 14 Containment Water Level Instrumentation and Sum Visual Ins ection T/S 4.3.3.6, Item 18 requires that the containment water level instrumentation be calibrated every 18 months. T/S 4.5.2.d.2 requires that the sump and its inlets be subjected to an 18 month visual inspection. An extension is also "

needed for T/S 4.5.3.1 since it references T/S 4.5.2. These surveillances cannot be performed during reactor operation since they require entry into the lower volume of containment. The extensions are needed from January 31, 1994 (calibrations) and March 21, 1994 (visual), until the Unit 2 refueling outage.

Our past history on containment water level instrumentation has not shown any significant degradation. This water level instrumentation is used to measure the amount of water on the containment floor above the sump. Normally, there is no water on the floor. The instrumentation consists of RTDs, which have shown stable operation in the past. Since the instruments have a "live" zero point on the scale, a reading of zero or greater indicates that the instruments are performing correctly. In addition, there are two redundant channels that are subjected to T/S required monthly channel checks and the channels can be compared to show if drift exists. There is no reason to believe that the containment water level instrumentation would be inoperable during the extension period.

The visual inspection is performed to ensure that we have a clean system prior to start up. During reactor operation, entry into the containment sump area is restricted. Also, we have very strict material control requirements for entry into containment and at the end of an outage, a "containment closeout tour" is performed to ensure that no material is left within containment. In addition, performance of visual inspections following reactor operation has shown that very little debris, ever accumulates in the sump. There is no reason to believe that the sump or it's inlets would become blocked.

10 CFR 50.92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Attachment 1 to AEP:NRC:1181 Page 31 Criterion 1 Our past history on containment water level instrumentation has not shown any significant degradation of these instruments. Typically, there is no water on the containment floor for the instruments to measure; however, the instrumentation is calibrated to read a "live" zero level. Also, we have two redundant channels that are subjected to monthly channel checks, which would show if drift exists. Therefore, there is no reason to believe that the level instrumentation would not perform its intended function during the containment'ater extension period. The likelihood of a significant amount of debris entering the sump is very low because we have very strict requirements for material control inside containment, restricted access into the containment sump area, and an inspection of containment is performed at the end of an outage. There is no reason to believe that the sump or its inlets could become blocked during the extension period. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction

, in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by. providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. We believe this change falls within the scope of this example, for the reasons cited above.

Thus, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 32 15 Reactor Coolant Flow Transmitter Calibrations T/S 4.2.5.2 and T/S 4.3.1.1.1, Table 4.3-1, Items 12 & 13 require the reactor coolant (RC) flow instrumentation for each loop to be calibrated every 18 months.

These calibrations should not be performed at power because of the possibility of a reactor trip. Each set of transmitters (3 per loop) has a common sensing line, which when valving in (or out) one of the transmitters could cause a reduced differential pressure in the .other two transmitters. This could cause a reactor trip on low flow in one loop since the two out of three trip -logic would be satisfied. The extension for these surveillance requirements are needed from January 28, 1994, until the Unit 2 refueling outage.

Past surveillance history has shown that the RC flow channels are very stable; very little or no drift is found during calibration and they have always been within their allowable range. Also, since there are three channels per loop, drift would be expected to be discovered during the T/S required shiftly channel checks or monthly functional checks. Since the channels have been very stable and we have three channels per loop to indicate drift, there is no reason to believe that continued operation during the extension period would cause the instrumentation to become inoperable.

10 CFR 50 92 Criteria Per 10 CFR 50.92, will not involve significant hazards consideration if thea proposed amendment proposed amendment does not:

a (1) 'involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of'ccident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Criterion 1 Our reactor coolant flow channels have been very stable in the past. Ve have three channels per loop and perform T/S required channel checks and functional tests which should show any indication of drift. Therefore, there is no reason to believe that the reactor coolant flow channels would not be operable during the extension period. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

to AEP:NRC:1181 Page 33 Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. We believe this change falls within the scope of this example, for the reasons cited above.

Thus, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

to AEP:NRC:1181 Page 34 16 Tri Actuatin Device 0 erational Testin ESF manual actuation Extensions are requested for the Trip Actuating Device Operational Testing (ESF manual actuation switches) specified in T/S 4.3.2.1.1, Table 4.3-2, Items 9.a, 9.b, 9.c, and 9.d. These tests cannot be performed at power since they would actuate the ESF functions associated with the switches (see table below). The extensions are needed from April 15, 1994, through the Unit 2 refueling outage.

Table 4.3-2 Item No Descri tion 9.a Safety injection (ECCS)

Feedwater Isolation Reactor Trip (SI)

Containment Isolation Phase A Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System 9.b Containment Spray Containment Isolation Phase B Containment Air Recirculation Fan 9.c Containment Isolation Phase A Containment Purge and Exhaust Isolation 9.d Steam Line Isolation The circuitry associated with manual actuation of ESF functions is subjected to T/S required channel functional tests, monthly or bi-monthly. The only portion of the channel not tested is the manual actuation switches. Previous surveillance testing of the switches have shown them to be highly reliable; in fact, there has never been a failure of any of the ESF manual switches detected during surveillance testing of the switches in either unit. Additionally, we note that the manual circuitry serves as a backup to automatic actuation channels, which initiate the same ESF functions. The automatic channels are subjected to T/S required channel checks and channel functional tests to verify operability.

to AEP:NRC:1181 Page 35 10 CFR 50.92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Criterion 1 The surveillance test history of the ESF manual switches is excellent, indicating no failures of the switches in either unit. The majority of the manual circuitry is subject to a channel functional test on a monthly or bi-monthly basis. The channel functional testing will continue to be performed during the surveillance extension period. Additionally, we note that the manual circuitry serves as a backup to automatic circuitry, which initiates the same ESF functions. For these reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will it result in a significant reduction in a margin of safety.

Criterion 2 This extension will not result in a change in plant configuration or operation.

Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.

Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by, providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable. We believe this change falls within the scope of this example, for the reasons cited above.

Thus, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.