ML17059A447

Proposed Tech Specs Section 3.2.2, Minimum Reactor Vessel Temperature for Pressurization, & Associated Bases

ML17059A447
Person / Time
Site:	Nine Mile Point
Issue date:	09/01/1994
From:	NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML17059A448	List: ML17059A447 ML18040A135
References
NUDOCS 9409120247
Download: ML17059A447 (32)

Text

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. During in-service hydrostatic and leak testing, In order to generate additional plant-specific the reactor vessel temperature and pressure shall data, a capsule containing irradiated and satisfy the requirements of Figure 3.2.2.e if the unirradiated material will be re-inserted at the B core is not critical. capsule location. Re-insertion capsules have already been installed at the A and C locations.

d. The reactor vessel head bolting studs shall not A prime (') is used to indicate a re-insertion be under tension unless the temperature of the capsule. The withdrawal schedule for the re-vessel head flange and the head are equal to or insertion capsules is as follows:

greater than 100'F.

Fourth capsule (A') - 24 EFPY Fifth capsule (C') - 32 EFPY Sixth capsule (B') - 40 EFPY 9DN UA XO DM QO DM C)W 04 Vl4 OQ 04 mao AtQ~

XM 0

AMENDMENT NO. iQE 84

l HEATUP CORE NOT CRITICAL 1600 LU O

E5 CL O

l-z 't000 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ f ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

hJ I

CC CO lit 0" 500 ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ '\ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ A ~ ~ ~ ~ ~ ~ ~ ~ A r 363 0

303 Minimum Temperature for Boltup 100 F 0

0 50 100 155 200 250 300 350 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (F)

(reactor vessel beltline downcomer water temperature is measured at recirculation. loop suction)

FIGURE 3.2.2.a MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING HEATUP AND LOW-POWER PHYSICS TESTS (CORE NOT CRITICAL)

(HEATING RATE 5 100'F/HR) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. /Pg 85

REACTOR VESSEL BELTLINE REACT R PRESSURE i DOWNCOMER WATER IN TOP DOME TEMPERATURE F a 303 100 303 110 303 120 303 130 303 140 303 150 363 160 394 170 431 180 476 190 529 200 591 210 664 220 748 230 847 240 960 250 1092 260 1243 270 1417 280 Reactor vessel beltline downcomer water temperature is measured at recirculation loop suction TABLE 3.2.2.a MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING HEAT-UP AND LOW-POWER PHYSICS TESTS (CORE NOT CRITICAL)

(HEATING RATE ~100 F/HR) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 73k 86

COOLDOWN CORE NOT CRITICAL 1500 LLI O

O CL 0l-1000 ~ ~ ~ ~ ~ ~ ~ ~ ~ Og I~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \~~~~~~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

LU K

I 03 LIJ CC 579 500 ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ A~ ~ A ~ ~ ~ ~ r ~

0 293 Minimum Temperature for Boltup 100 F 0

0 50 100 't58 200 250 300 360 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE 0=)

(reactor vessel beltline downcomer water temperature is meaered at recirculation loop suction)

FIGURE 3.2.2.b MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN AND LOW-POWER PHYSICS TESTS (CORE NOT CRITICAL)

(COOLING RATE 5 100'F/HR) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. @g 87

e REACTOR VESSEL BELTLINE REACTOR PRESSURE si DOWNCOMER WATER IN TOP DOME TEMPERATURE F 293 100 293 110 293 120 293 130 293 140 293 150 579 160 655 170 739 180 822 190 907 200 1004 210 1116 220 1245 230 1394 240 Reactor vessel beltline downcomer water temperature is measured at recirculation loop suction TABLE 3.2.2.b MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN AND LOW-POWER PHYSICS TESTS (CORE NOT CRITICAL)

(COOLING RATE ~100 F/HR) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. )ME 88

ce i

e

HEAT UP CORE 'RITICAL 1500 LLI O

Cl CL O

I-z 1000 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ j ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

Water Level Must Be ln Range For LU Power Operation lf Core ls Critical CO Below 219 F CO LLI 500 ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

CC 428 303 Minimum Temperature for Boltup 100 F 0

0 50 100 150 200 250 300 350 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (F)

(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

FIGURE 3.2.2.c MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING HEATUP (CORE CRITICAL)

(HEATING RATE 5 100'F/HR) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. PPg 89

REACTOR VESSEL BELTLINE REA T R PRESSURE si D WNCOMER WATER IN TOP DOME TEMPERATURE F 303a 100 303a 110 303a 120 303a 130 303a 140 303 150 303a 160 303 170 303a 180 303a 190 303a 200 303a 210 428 219 431 220 476 230 529 240 591 250 664 260 748 270 847 .280 960 290 1092 300 1243 310 1417 320 Water level must be in range for power operation if core is critical below 219'F Reactor vessel beltline downcomer water temperature is measured at recirculation loop suction TABLE 3.2.2.c MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING HEATUP (CORE CRITICAL)

(HEATING RATE ~100 F/HR) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. Ak 90

COOLDOWN CORE CRITICAL i 600 UJ O

D 0CL

\

1000 ~ ~ ~ ~ ~ ~ ~ ~ ~

4 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

Water Level Must Be In Range For LLI Power Operation 0" If Core Is Critical Below 185 F 03 CO A~ A UJ K 500 ~% ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

0 360 200 Minimum Temperature for Boltup 100 F 0

0 50 100 155 00 250 300 350 REACTOR VESSEL 'BELTLINE DQWNCOMER WATER TEMPERATURE (F)

(reactor vessel beltline downcomer water temperature is fneasured at recirculation loop suction)

FIGURE 3.2.2.d MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN (CORE CRITICAL)

(COOLING RATE 5 100'F/HR) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. f)4g 91

0 REA TOR VE EL BELTLINE REACTOR PRESSURE si DOWNCOMER WATER IN TOP DOME TEMPERATURE F 200a 100 221a 110 241 a 120 265a 130 293a 140 325a 150 360a 160 360a 170 360a 180 483 185 512 190 579 200 655 210 739 220 822 230 907 240 1004 250 1116 260 1245 270 1394 280 Water level must be in range for power operation if core is critical below 185'F Reactor vessel beltline downcomer water temperature is measured at recirculation loop suction TABLE 3.2.2.d MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN (CORE CRITICAL) (COOLING RATE x 100 F/HR)

FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. Ak 92

LEAKXHYDRQ TEST CORE NQT CRITICAL 1500 Ill O

O 0

0 I-z 1000 ~ ~ ~ ~ ~ ~ ~ I~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

CC 700 CO CO UJ 500 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ h ~ ~ ~ ~ ~ A CC CL 360 Minimum Temperature for Boltup 100 F 0 50 100 150 200 250 300 350 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (F)

(reactor vessel beltline downcomer water temperature is meaured at recirculation, loop suction)

FIGURE 3.2.2.e MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING IN-SERVICE HYDROSTATIC TESTING AND LEAK TESTING (CORE NOT CRITICAL) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NQ. PAP 93

REACTOR VESSEL BELTLINE REA TOR PRESSURE si D WNCOMER WATER IN TOP DOME TEMPERATURE F 360 100 360 110 360 120 700 130 747 140 801 150

'64 160 937 170 1021 180 1117 190 1229 200 1357 210 Reactor vessel beltline downcomer water temperature is measured at recirculation loop suction TABLE 3.2.2.e MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING IN-SERVICE HYDROSTATIC TESTING AND LEAK TESTING (CORE NOT CRITICAL)

FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. AE 94

BASES FOR 3.2.2 AND 4.2.2 MINIMUMREACTOR VESSEL TEMPERATURE FOR PRESSURIZATION Figures 3.2.2.a, 3.2.2.b, 3.2.2.c, and 3.2.2.d are plots of pressure versus temperature for heatup and cooldown rates of up to 100'F/hr.

maximum (Specifications 3.2.1, 3.2.2.a, and 3.2.2.b). Figure 3.2.2.e is a plot of pressure versus temperature for in-service hydrostatic testing and leak testing (Specification 3.2.2.c). These curves are based on calculations of stress intensity factors according to Appendix G of Section III of the ASME Boiler and Pressure Vessel Code 1980 Edition with Winter 1982 Addenda and on the requirements of 10CFRSO, Appendix G. In addition, the ASME reference stress intensity factor curve was shifted to higher temperatures to account for neutron fluence effects at eighteen effective full power years of operation. The temperature shift was calculated by adding the Charpy shift (indexed at 30 ft-Ibs of absorbed energy) plus a margin term to the limiting beltline material (Plate G-307-4) initial reference nil-ductility temperature (RTNDT) in accordance with the guidance provided in Regulatory Guide 1.99, Revision 2. A plant-specific Charpy shift model was developed for the Nine Mile Point Unit 1 limiting plate. Since the adjusted RTNDT (ARTNDT) for the limiting G-307-4 plate is larger than the ARTNDT of any other vessel material, the pressure-temperature limits provide protection against brittle fracture for the beltline and non-beltline reactor pressure vessel materials. Reactor vessel flange/reactor head flange boltup is governed by other criteria as stated in Specification 3.2.2.d. The pressure readings on the figures have been adjusted to reflect the calculated elevation head difference between the pressure sensing instrument locations and the pressure sensitive area of the core beltline region. The coolant temperature on the figures, which is measured in the recirculation pump suction line, is a conservative measure of the vessel inner wall surface temperature due to gamma heating and frictional heating due to coolant flow. Also, the instrument error has been included in a conservative manner.

The reactor vessel head flange and vessel flange in combination with the double "0" ring type seal are designed to provide a leak-tight seal when bolted together. When the vessel head is placed on the reactor vessel, only that portion of the head flange near the inside of the vessel rests on the vessel flange. As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surfaces adjacent to the "0" rings of the head and vessel flanges. Both the head and vessel flanges have an NDT temperature of 40'F and they are not subject to any appreciable neutron radiation exposure. Therefore, the minimum vessel flange and head flange temperature for bolting is established at 40'F + 60'F or 100'F.

Figures 3.2.2.a, 3.2.2.b, 3.2.2.c, 3.2.2.d, and 3.2.2.e have incorporated a temperature shift due to the calculated fast neutron fluence.

The neutron flux at the vessel wall is calculated from core physics data and has been determined using flux monitors installed inside the vessel within the surveillance capsules. The curves are applicable for up to eighteen effective full power years of operation.

Vessel material surveillance samples are located within the core region to permit periodic monitoring of exposure and changes in material properties. The Nine Mile Point Unit 1 surveillance program was originally designed to and meets the requirements of ASTM E185-66. The surveillance specimens removed to date were tested in accordance with ASTM E185-82. The initial RT~pp of the beltline materials has been calculated in accordance with ASME Section III. A longitudinal-transverse (L-T) to transverse-longitudinal (T-L) adjustment factor has been applied to the surveillance data to meet the current 10CFR50, Appendix H requirements. Therefore, the current surveillance program for Nine Mile Point Unit 1 meets the requirements of Appendix H to 10CFR50. The material withdrawal schedule has been revised and is specified in Specification 4.2.2.b.

AMENDMENT NO. QE 95

ATTACHMENT B NIAGARA MOHAWK POWER CORPORATION LICENSE NO. DPR-63 DOCKET NO. 50-220 Su ortin Information and No Si nificant Hazards Considera ion Anal sis BACKGROUND The proposed revisions to Technical Specification Section 3.2.2 and the associated Bases reflect changes in the limits for minimum reactor vessel temperature for pressurization.

The previous pressure-temperature (P-T) limits were calculated using Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials." The proposed P-T curve revisions are also based on the guidance given in Regulatory Position 2.1, except the generic model for calculating the shift in the reference nil-ductility temperature QRTgpy) caused by irradiation given by Equation (2) in Regulatory Guide 1.99, Revision 2 has been replaced by a plant-specific model. Regulatory Guide 1.99, Revision 2 anticipates the development and use of plant-specific models: "To use the surveillance data from a specific plant instead of Regulatory Position 1, one must develop a relationship of ERTo~ to fluence for that plant."

At the time the Regulatory Guide 1.99, Revision 2 model was developed, there were few surveillance capsule test results available which had been irradiated to fluences in the boiling water reactor operating range. The current Regulatory Guide 1.99, Revision 2 generic model results in an overly conservative representation of material performance in the low fluence range for some materials. The Nine Mile Point Unit 1 plant-specific model is physically based and more accurately represents the reference nil-ductility temperature (Charpy) shift behavior of the Nine Mile Point Unit 1 beltline plates. The proposed changes to the P-T limits are based on the more accurate representation of the Charpy shift behavior of Nine Mile Point Unit 1 irradiated limiting plate material as provided by the plant-specific model.

DESCRIPTION OF PROPOSED CHANGES The proposed amendment replaces existing Figures 3.2.2.a, b, c, d, and e and associated Tables 3.2.2.a, b, c, d, and e, that define the limits for minimum reactor vessel temperature for pressurization up to 18 effective full power years (EFPY), with new figures and tables also applicable for up to 18 EFPY. The new P-T limits were developed based on a plant-specific Charpy shift model for Nine Mile Point Unit 1 and in accordance with 10CFR50, Appendix G, Appendix G to Section III of the ASME Code, and Regulatory Guide 1.99, Revision 2. An editorial change is being made to Specification 3.2.2.c to clarify that Figure 3.2.2.e also applies to in-service leak tests as noted on the figure. Editorial changes have also been made to the figures and tables to improve their clarity and provide consistency with the terminology of 10CFR50, Appendix G. Also included are associated Bases changes in support of the new P-T limits.

EVALUATION Nine Mile Point Unit 1 is operated so that the reactor primary coolant system does not experience any substantial mechanical or thermal loading unless the reactor pressure Page 1 of 4

vessel (RPV) materials are at a temperature well above the reference nil-ductility temperature (RT<p7) of the limiting RPV material (beltline plate G-307-4). Protection against brittle fracture is further ensured by postulating a defect with a depth ~/i of the RPV wall thickness and a length 1 /~ times the wall thickness, and calculating the allowable pressure loading as a function of temperature using linear elastic fracture mechanics in accordance with Appendix G to Section III of the ASME Code. The ASME Code,Section III, Appendix G requirements have been met and include a safety factor of 2 applied to the membrane stress intensity factor (K~M) for heatup and cooldown, and a safety factor of 1.5 for leakage/hydrostatic testing. In addition, the ASME reference stress intensity factor (K,) curve is based on the lower bound of static, dynamic and crack arrest fracture toughness tests conducted on similar RPV steels.Section IV of Appendix G to 10CFR50 specifies that when the pressure exceeds 20% of the pre-service system hydrostatic test pressure, the temperature of the closure flange regions must exceed the RT<p7 of the flange region materials by 120'F for normal operation and 90'F for leakage/hydrostatic tests. When the core is critical, the temperature limits for non-critical operation are further increased by 40 F to produce the core operation P-T limits. The net effect of the 10CFR50, Appendix G and the ASME Section III, Appendix G P-T curve calculative procedures is to produce very conservative P-T curves. These procedures have been applied in the calculation of the proposed Nine Mile Point Unit 1 P-T limits. Detailed discussions concerning the calculative procedures for generating the new P-T limits are enclosed in the Attachment C report entitled, "Nine Mile Point Unit 1 Pressure-Temperature Operating Curves," MPM-69437, August 1994.

Neutron damage during plant operation is accounted for in the allowable pressure loading by calculating an adjusted reference nil-ductility temperature (ARTp7). Regulatory Guide 1.99, Revision 2, defines the ARTp7 as the sum of the reference nil-ductility temperature (RT~p7) plus the shift in the reference nil-ductility temperature caused by irradiation (BRTgp7) plus a margin term. The proposed amendment replaces the generic model for calculating BRTp7 given by Equation (2) of Regulatory Guide 1.99, Revision 2, with an accurate plant-specific model. The plant-specific model results in a BRT+p7 which is reduced as compared with the previous BRT~p7 determined using Equation (2) and Regulatory Position 2.1. The lower BRTp7 is a result of the more accurate representation of the Nine Mile Point Unit 1 RPV plate behavior as a function of neutron exposure.

However, the b,RTp7 is intended to be an accurate representation of the Charpy shift as a function of fluence. Since the ASME Section III, Appendix G safety factors have been maintained and the Regulatory Guide 1.99, Revision 2, margin term specified in Regulatory Position 2.1 has been applied in the same manner as for earlier P-T curve calculations, no significant reduction in the margin of safety has resulted.

The proposed changes to Specification 3.2.2.c and the labeling of Figures 3.2.2.a, b, c, d and e and associated Tables 3.2.2.a, b, c, d and e are strictly editorial to improve their clarity and provide consistency with the terminology of 10CFR50, Appendix G.

The proposed changes to the Bases for 3.2.2 and 4.2.2 provide a more comprehensive discussion of the calculative procedure used in the development of the P-T operating limits. Additional information regarding the Nine Mile Point Unit 1 surveillance program has also been included in the Bases for clarification purposes.

CONCLUSION Operation of Nine Mile Point Unit 1 in accordance with the proposed P-T operating limits will ensure station operations are conducted with the reactor vessel materials above the nil-ductility transition temperature. Operation in accordance with the proposed P-T Page 2 of 4

operating limits will preclude brittle failure of the reactor vessel material, since safety ma)gins specified in 10CFR50, Appendix G and Appendix G to Section III of the ASME Code will be maintained. The proposed changes to the Bases for 3.2.2 and 4.2.2 are administrative for clarification purposes and do not affect plant systems or operations.

Therefore, there is reasonable assurance that operation of Nine Mile Point Unit 1 in the proposed manner will not endanger the public health and safety and that issuance of the proposed amendment will not be inimical to the common defense and security.

10CFR50.91 requires that at the time a licensee requests an amendment, it must provide to the Commission its analysis using the standards in 10CFR50.92 concerning the issue of no significant hazards consideration. Therefore, in accordance with 10CFR50.91, the following analysis has been performed:

The o eration of Nine Mile Point Unit 1 in accordance with the ro osed amendmen will not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.

Components of the reactor primary coolant system are operated so that no substantial mechanical or thermal loading is applied unless the reactor pressure vessel (RPV) materials are at a temperature well above the reference nil-ductility temperature (RTD~) of the limiting RPV material Protection against brittle fracture is further ensured by postulating a

~

defect with a depth /i of the RPV wall thickness and a length 1/~ times the wall thickness, and calculating the allowable pressure loading as a function of temperature using linear elastic fracture mechanics. Safety factors are applied to the allowable loading determination and lower bound fracture toughness properties are used to represent the material behavior. The net effect of the 10CFR50, Appendix G and the ASME Section III, Appendix G P-T curve calculative procedures is to produce very conservative P-T curves.

These procedures have been applied in the calculation of the proposed P-T limits.

Neutron damage during plant operation is accounted for in the allowable pressure loading by calculating an adjusted reference nil-ductility temperature (ARTo~). Regulatory Guide 1.99, Revision 2, defines the ARTSD~ as the sum of the reference nil-ductility temperature (RT~o~) plus the shift in the reference nil-ductility temperature caused by irradiation (b,RT~>), plus a margin. The proposed amendment replaces Equation (2) in Regulatory Position 2.1 with an accurate plant-specific model. The ARTzDz margin is the same as for earlier P-T curve calculations. Operation of NMP1 in accordance with the proposed P-T operating limits will preclude brittle failure of the RPV materials. Safety margins for brittle fracture are in accordance with those specified in 10CFR50, Appendix G and Appendix G to Section III of the ASME Code. Therefore, the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The o era ion of Nine !Ville Poin Unit 1 in accordance with he r osed amendment will not create the ossibilit of a new or different kind of accident from an acciden reviousl evalua d.

The proposed amendment incorporates P-T operating limits based on previously established calculative procedures described in 10CFR50, Appendix G, Appendix G to Section III of the ASME Code, and Regulatory Guide 1.99, Revision 2. The proposed changes to the P-T operating limits are based on analyses of the irradiated limiting plate material for Nine Mile Point Unit 1. The proposed changes do not modify any plant equipment nor do they create any potential initiating events that would create any new or different kind of accident. Operation in accordance with the proposed P-T operating limits will preclude brittle failure of the reactor vessel material, since safety margins specified in Page 3 of 4

t I

10CFR50, Appendix G and Appendix G to Section III of the ASME Code be maintained. Therefore, the proposed P-T limits will not create the possibility of a new or different kind of accident from any accident previously evaluated. ~

The o eration of Nine Mile Poin Uni 1 in accordance with he amendmen will no

~ ~ ~ ~ ~

~ ~

involve a si nificant reduction in a mar in of safe

~ ~ ~ ~

~

Operation in accordance with the proposed P-T operating limits will preclude brittle failure of the reactor pressure vessel since safety margins in 10CFR50, Appendix G and Appendix G to Section III of the ASME Code will be maintained. The plant-specific limiting material BRTgpy has been reduced as compared with the overly conservative b,Ropy used in previous P-T curve calculations as a result of the more accurate representation of the Nine Mile Point Unit 1 RPV plate behavior as a function of neutron exposure. However, the hRTD~ is intended to be an accurate representation of the Charpy shift (indexed at 30 ft-Ibs of absorbed energy) as a function of fluence. Since the ASME Section III, Appendix G safety factors have been maintained and the Regulatory Guide 1.99, Revision 2, margin term specified in Regulatory Position 2.1 has been applied in the same manner as in earlier P-T curve calculations, no significant reduction in the margin of safety has resulted from the use of a plant-specific b,RTD~ model. Therefore, the proposed amendment will not involve a significant reduction in a margin of safety.

Page 4 of 4

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