Enclosure 4 to Seria. RNP/93-1744 Page 1 of 1

H.B.ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 NRC DOCKET NO. 50-261/LICENSE NO. DPR-23

REQUEST FOR LICENSE AMENDMENT

PAGE CHANCE INSTRUCTIONS

Removed Page Inserted Page

V V

3.1-21 3.1-21

3.1-21a

3.1-22 3.1-22

3.1-22a

3.1-4 3.1-4

3.1-5 3.1-5

3.1-6 3.1-6

3.1-7 3.1-7

9309210295 930915 PDR ADOCK 05000261 P PDR

Enclosure 5 to Seria. RNP/93-1744 Page 1 of 1

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 NRC DOCKET NO. 50-261/LICENSE NO. DPR-23

REQUEST FOR LICENSE AMENDMENT

TECHNICAL SPECIFICATION PAGES

LIST OF FIGURES

Figure Title Page

1.1-1 Plant Site Boundary and Exclusion Zone 1-8

2.1-1 Safety Limits Reactor Core, Thermal, and Hydraulic Three Loop Operation, 100% Flow 2.1-4

3.1.4-1 Percent of Rated Thermal Power 3.1-15a

3.1-1 Reactor Coolant System Heat p Limitations Applicable for R rds-z Up to )R6 Efee-piEe iFill Pzer Years 24 3.1-21

3.1-2 Reactor Coolant System Cool own Limitations

APpico-b( -Apieabi1ity for Pe-4-ed- Up to X6 Effeetiye' Ru 0 ~ Y~r .4 3.1-22

3.10-1 (DELETED) 3.10-20

3.10-2 Shutdown Margin versus Boron Concentration 3.10-21

3.10-3 (DELETED) 3.10-22

3.10-4 (DELETED) 3.10-23

3.10-5 (DELETED) 3.10-24

6.2-1 Offsite Organization for H. B. Robinson 2

Management and Technical Support 6.2-3

6.2-2 Conduct of Operations Chart 6.2-4

v Amendment No.

3.1.2 Heatup and Cooldown

3.1.2.1 The reactor coolant pressure and the system heatup and cooldown

rates (with the exception of the pressurizer) shall be limited in

accordance with Figure 3.1-11and Figure 3.1-2d (fu szl qu-z

tip )for vessel

exposure up to +5 EFPY). -Tho-15-5EFP-cuves-may-be-used-o

ageratisfl-prior tethen d-f- 12.5EPPY . These limitations are as

follows:

a. .:Over the temperature range from cold shutdown to hot operating

'conditions, the heatup rate shall not exceed 60*F/hr. in any

one hour.

b. Allowable combinations of pressure and temperature for a

specific cooldown rate are below and to the right of the limit

lines for that rate as shown on Figure 3.1-2asr 3-.1-e2b (a

.apppreted-. This rate shall not exceed 100"F/hr. in any one

hour. The limit lines for cooling rates between those shown in

Figure 3.1-d FiPhaz 3.1 20 may be obtained by

interpolation.

c. Primary system hydrostatic leak tests may be performed as

necessary, provided the temperature limitation as noted on

Figure 3.1-1A:. Fig. z 3.1 . ap u is not

violated. Maximum hydrostatic test pressure should remain

below 2350 psia.

d. The overpressure protection system shall be operable whenever

the RCS temperature is below 350*F and not vented to the

containment. One PORV may be inoperable for seven days. If

the inoperable PORV has not been returned to service within 7

days, or if at any time both PORVs become inoperable, then one

of the following actions should be completed within 12 hours:

1. Cooldown and depressurize the RCS or

3.1-4 Amendment No. ,)-k

2. Heatup the RCS to above 350*F.

e. Operation of the overpressure protection system to relieve a

pressure transient must be reported as a Special Report to the

NRC within 30 days of operation.

3.1.2.2 The secondary side of the steam generator must not be pressurized

above 200 psig if the temperature of the vessel is below 120*F.

3.1.2.3 The pressurizer shall neither exceed a maximum heatup rate of

100*F/hr nor a cooldown rate of 200*F/hr. The spray shall not be

used if the temperature difference between the pressurizer and the

spray fluid is greater than 320*F.

3.1.2.4 Figures 3.1-1,/and 3.1-ishall be updated periodically in

accordance with the following criteria and procedures before the

calculated exposure of the vessel exceeds the exposure for which the

figures apply.

a. At least 60 days before the end of the integrated power period

for which Figures 3.1-l and 3.1-2, apply, the limit lines on

the figures shall be updated for a new integrated power period

utilizing methods derived from the ASME Boiler and Pressure

Vessel Code, Section III, Appendix C and in accordance with

applicable appendices of 10CFR50. These limit lines shall

reflect any changes in predicted vessel neutron fluence over

the integrated power period or changes resulting from the

irradiation specimen measurement program.

b. The results of the examinations of the irradiation specimens

and the updated heatup and cooldown curves shall be reported to

the Commission within 90 days of completion of the

examinations.

3.1-5 Amendment No.)*<

steels such as ASTM A302 Grade B parent material of the H. B. Robinson Unit

No. 2 reactor pressure vessel are well documented in the literature.

Generally, low alloy ferritic materials show an increase in hardness and other

strength properties and a decrease in ductility and impact toughness under

certain conditions of irradiation. Accompanying a decrease in impact strength

is an increase in the temperature for the transition from brittle to ductile

fracture.

A method for guarding against fast fracture in reactor pressure vessels has been

presented in Appendix G, "Protection Against Non-Ductile Failure," to

Section III of the ASME Boiler and Pressure Vessel Code. The method utilizes

fracture mechanics concepts and is based on the reference nil-ductility

temperature, RTNDT.

RTNDT is defined as the greater of: 1) the drop weight nil-ductility

transition temperature (NDTT per ASTM E-208) or 2) the temperature 60'F less

than the 50 ft-lb (and 35 mils lateral expansion) temperature as determined

from Charpy specimens oriented in a direction normal to the major working

direction of the material. The RTNDT of a given material is used to index

that material to a refeence stress int it factor curve (KIR curve) which

appears in Appendix G of A the ASMECde. The KIR curve is a lower bound of

dynamic, crack arrest, and 6tatic fracture toughness results obtained from

several heats of pressure vessel steel. When a given material is indexed to

the KIR curve, allowable stress intensity factors can be obtained for this

material as a function of temperature. Allowable operating limits can then be

determined utilizing these allowable stress intensity factors.

The Certified Material Test Reports (CMTRs) for the original steam

generators provided records of Charpy V-notch tests performed at +10*F.

Acceptable Charpy V-notch tests of +10*F indicate RTNDT is at or below

this temperature. The steam generator lower assemblies were replaced in

1984 and the material tests results indicate the highest RTNDT is 60*F or

below. The ASME code recommends that hydrostatic tests be performed at a

temperature not lower than RTNDT plus 60F*, thus the pressurizing

temperature for the steam generator shell is established at 120*F to

provide protection against nonductile failure at the test pressure.

3.1-6 Amendment No.)9f

V-notch 30 ft-lb temperature (A RT ) due to irradiation is added to the

original RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted

RTNDT (RTNDT initial +ARTNDT) is utilized to index the material to the KIR

curve and in turn to set operating limits for the nuclear power plant which

take into account the effects of irradiation on the reactor vessel

materials. Allowable pressure-temperature relationships for various heatup

and cooldown rates are calculated using methods (2) derived from Appendix G to

Section III of the ASIME Boiler and Pressure Vessel Code. The approach

specifies that the allowable total stress intensity factor (KI) at any time

during heatup or cooldown cannot be greater than that shown on the KIR curve

in Appendix G for the metal temperature at that time. Furthermore, the

approach applies an explicit safety factor of 2.0 on the stress intensity

factor induced by pressure gradients.

Following the generation of pressure-temperature curves for both the steady

state and finite heatup rate situations, the final limit curves are produced

in the following fashion. First, a composite curve is constructed based on a

point-by-point comparison of the steady state and finite heatup rate data. At

any given temperture, the allowable pressure is taken to be the lesser of the

two values taken from the curves under consideration.. The composite curve is

then adjusted to allow for possible errors in the pressure and temperature

sensing instruments.

The use of the composite curve is mandatory in setting heatup limitations

because it is possible for conditions to exist such that over the course of

the heatup ramp the controlling analysis switches from the O.D. to the I.D.

location; and the pressure limit must, at all times, be based on the most

conservative case. The cooldown analysis proceeds in the same fashion as that

for heatup, with the exception that the controlling location is always at the

I.D. position. The thermal gradients induc'ed during cooldown tend to produce

tensile stresses at the I.D. location and compressive stresses at the O.D.

position. Thus, the I.D. flaw is clearly the worst case.

As in the case of heatup, allowable pressure temperature relations are

generated for both steady state and finite cooldown rate situations.

Composite limit curves are, then constructed for each cooldown rate of

3.1-7 Amendment No. Rcv,;F4-iien 7 6

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Carolina HeatuP LIMitatlon:s pplicable Powr & Llqnt Carnoany I. b

IC h "Ical Speciticationpi'l U p T.o 15 EF

3.1-ZI-a Amendment No. 113

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ontrolling Materi Weld Metal Copper Content : 0.38 wt.% Phosphorus Co 'ant : 0.0 12 wt.%

RT Initial : aaF NOT

DE -L)R O Aftte@ EFPPY: 1/4 T. 290OF ND13/4 T, N1'1 F

Curve 0 a pli31cablo for cooldon r tes up to 100~ "HR for the service perl d up to 15 ,VPY.

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3.1-22.a~~ AmnmI 1 o. 11

LIST OF FIGURES

Figure Title Page

1.1-1 Plant Site Boundary and Exclusion Zone 1-8

2.1-1 Safety Limits Reactor Core, Thermal, and Hydraulic Three Loop Operation, 100% Flow 2.1-4

3.1.4-1 Percent of Rated Thermal Power 3.1-15a

3.1-1 Reactor Coolant System Heatup Limitations Applicable Up to 24 EFPY 3.1-21

3.1-2 Reactor Coolant System Cooldown Limitations Applicable Up to 24 EFPY 3.1-22

3.10-1 (DELETED) 3.10-20

3.10-2 Shutdown Margin versus Boron Concentration 3.10-21

3.10-3 (DELETED) 3.10-22

3.10-4 (DELETED) 3.10-23

3.10-5 (DELETED) 3.10-24

6.2-1 Offsite Organization for H. B. Robinson 2 Management and Technical Support 6.2-3

6.2-2 Conduct of Operations Chart 6.2-4

v Amendment No.

3.1.2 Heatup a Cooldown

3.1.2.1 The reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figure 3.1-1 and Figure 3.1-2 (for vessel exposure up to 24 EFPY). These limitations are as follows:

a. Over the temperature range from cold shutdown to hot operating conditions, the heatup rate shall not exceed 60aF/hr. in any one hour.

b. Allowable combinations of pressure and temperature for a specific cooldown rate are below and to the right of the limit lines for that rate as shown on Figure 3.1-2. This rate shall not exceed 100 0F/hr. in any one hour. The limit lines for cooling rates between those shown in Figure 3.1-2 may be obtained by interpolation.

c. Primary system hydrostatic leak tests may be performed as necessary, provided the temperature limitation as noted on Figure 3.1-1 is not violated. Maximum hydrostatic test pressure should remain below 2350 psia.

d. The overpressure protection system shall be operable whenever the RCS temperature is below 350oF and not vented to the containment. One PORV may be inoperable for seven days. If the inoperable PORV has not been returned to service within 7 days, or if at any time both PORVs become inoperable, then one of the following actions should be completed within 12 hours:

1. Cooldown and depressurize the RCS or

3.1-4 Amendment No.

2. Heatup the RCS to above 350'F.

e. Operation of the overpressure protection system to relieve a pressure transient must be reported as a Special Report to the NRC within 30 days of operation.

3.1.2.2 The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 1201F.

3.1.2.3 The pressurizer shall neither exceed a maximum heatup rate of 100 0F/hr nor a cooldown rate of 200*F/hr. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 3201F.

3.1.2.4 Figures 3.1-1 and 3.1-2 shall be updated periodically in accordance with the following criteria and procedures before the calculated exposure of the vessel exceeds the exposures for which the figures apply.

a. At least 60 days before the end of the integrated power period for which Figures 3.1-1 and 3.1-2 apply, the limit lines on the figures shall be updated for a new integrated power period utilizing methods derived from the ASME Boiler and Pressure Vessel Code, Section III, Appendix G and in accordance with applicable appendices of 1OCFR50. These limit lines shall reflect any changes in predicted vessel neutron fluence over the integrated power period or changes resulting from the irradiation specimen measurement program.

b. The results of the examinations of the irradiation specimens and the updated heatup and cooldown curves shall be reported to the Commission within 90 days of completion of the examinations.

3.1-5 Amendment No.

steels such as ASTM 2 Grade B parent material of t H. B. Robinson Unit No. 2 reactor pressure vessel are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and other strength properties and a decrease in ductility and impact toughness under certain conditions of irradiation. Accompanying a decrease in impact strength is an increase in the temperature for the transition from brittle to ductile fracture.

A method for guarding against fast fracture in reactor pressure vessels has been presented in Appendix G, "Protection Against Non-Ductile Failure," to Section III of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature, RTNOT.

RTNDT is defined as the greater of: 1) the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or 2) the temperature 60aF less than the 50 ft-lb (and 35 mils lateral expansion) temperature as determined from Charpy specimens oriented in a direction normal to the major working direction of the material. The RTNDT of a given material is used to index that

material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G of Section III of the ASME Boiler and Pressure Vessel Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture

toughness results obtained from several heats of pressure vessel steel. When

a given material is indexed to the KIR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

The Certified Material Test Reports (CMTRs) for the original steam generators

provided records of Charpy V-notch tests performed at +10oF. Acceptable

Charpy V-notch tests of +10aF indicate RTNDT is at or below this temperature. The steam generator lower assemblies were replaced in 1984 and the material tests results indicate the highest RTNDT is 601F or below. The ASME code

recommends that hydrostatic tests be performed at a temperature not lower than

RTNDT plus 60Fo, thus the pressurizing temperature for the steam generator shell is established at 1201F to provide protection against nonductile failure at the test pressure.

3.1-6 Amendment No.

V-notch 30 ft-lb tem prature (A RTNDT) due to irradiaon is added to the original ARfNoT to adjust the RTNDT for radiation embrittlement. This adjusted RTNDT.(RTNDT initial + ARTNOT) is utilized to index the material to the KIR curve and in turn to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials. Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods (2) derived from Appendix G to Section III of the ASME Boiler and Pressure Vessel Code. The approach specifies that the allowable total stress intensity factor (K,) at any time during heatup or cooldown cannot be greater than that shown on the KR curve in Appendix G for the metal temperature at that time. Furthermore, the approach applies an explicit safety factor of 2.0 on the stress intensity factor induced by pressure gradients.

Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced in the following fashion. First, a composite curve is constructed based on a point-by-point comparison of the steady state and finite heatup rate data. At any given temperture, the allowable pressure is taken to be the lesser of the two values taken from the curves under consideration. The composite curve is then adjusted to allow for possible errors in the pressure and temperature sensing instruments.

The use of the composite curve is mandatory in setting heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling analysis switches from the 0.D. to the I.D. location; and the pressure limit must, at all times, be based on the most conservative case. The cooldown analysis proceeds in the same fashion as that for heatup, with the exception that the controlling location is always at the I.D. position. The thermal gradients induced during cooldown tend to produce tensile stresses at the I.D. location and compressive stresses at the 0.0.

position. Thus, the I.D. flaw is clearly the worst case.

As in the case of heatup, allowable pressure temperature relations are generated for both steady state and finite cooldown rate situations. Composite limit curves are then constructed for each cooldown rate of

3.1-7 Amendment No.

law MATERIALS PROPERTIES BASE

Controlling Material Lower Circumferential Weld Curves applicable for heatup Copper Content 0.20 wt.% rates up to 60*F/Hr for the Nickel Content 1.06 wt.% service period up to 24 EFPY. RTNDT Initial 80F Includes +10'F and -60 PSIC RT NDTAfter 24 EFPY 1/4 T. 207.83'F allowance for instrumentation

3/4 T. 137.15'F error.

3000

2500

01 2000 Leak Test Limit

LiIi 1500

Crtcoiyii i

.. .. temperature (327F) Z for the service period

500

IIII

50 100 150 200 250 300 350 400 450 500

INDICATED TEMPERATURE ('F)

H.B. Robinson Unit *2 Reactor Coolant System . ..FIGURE

Heatup Limitations CAROLINA POWER & LIGHT COMPANY

Technical Specifications Applicable Up To 24 EFPY 311

3.1-21 Amendment No.

0 0 MATERIALS PROPERTIES BASE

Controlling Material Lower Circumferential Weld Curves applicable for cooldown Copper Content 0.20 wt.% rates up to lOOT/Hr for the Nickel Content 1.06 wt.% service period up to 24 EFPY. RTNDT Initial -80F Includes +10' and -60 PSIG

RTNDTAfter 24 EFPY 1/4 T. 207.83F allowance for instrumentation 3/4 T. 137.15'F error.

3000

2500

U)2000 1 . 1 1 il 111

(I) CL) w 1500

0

1000

50 100 150 200 250 300 350 400 450 500

INDICATED TEMPERATURE ('F)

H.B. Robinson Unit 2 Reactor Coolant Syster

.- F . .FIGURE Cooldown Limitations

CAROLINA POWER & LIGHT COMPANY Technical Specifications Applicable Up To 24 EFPY 3.1-2

3.1-22 Amendment No.

Enclosure 6 to Seri*. RNP/93-1744 Page 1 of 3

Pressurized Thermal Shock

In accordance with the revised Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61, effective June 14, 1991, the reactor vessel beltline materials were evaluated to determine the impact of the revised rule. A summary of the pertinent data required by revised Section 50.61(b)(1) for both the current date and the HBR2 Operating License expiration date is included in the table of this Enclosure 6.

The PTS rule requires calculation of RTPTS values for all beltline materials for both (1) End-Of-License (EOL) and (2) the time of submittal. August of 1993 was chosen as the time of submittal to coincide with the planned Technical Specification submittal of updated Pressure - Temperature limit curves.

The calculation results show the upper circumferential weld joint (Weld 10-273) to be the most limiting beltline material for PTS considerations. However, as shown in the attached table, none of the calculated RTPTS values exceed the 10 CFR 50.61 screening criterion. Therefore, as concluded in 1991, a response to the NRC by December 16, 1991, in accordance to the PTS rule, was not required.

The methods for calculation of RTPTs do not currently permit the use of plant specific materials surveillance program data, as allowed by the methods for calculating RTNDT . Because the upper weld (10-273) has been demonstrated as applicable to the surveillance program, plant specific data was used for RTNDT calculations. This information makes the lower circumferential weld (11-273) most limiting for RTNDT purposes and the development of Pressure-Temperature curves, even though the upper circumferential weld (10-273) is calculated to be most limiting for PTS considerations.

Enclosure 6 to Serial: RNP/93-1744 Page 2 of 3

PREDICTED REFERENCE PTS TEMPERATURES FOR H. B. ROBINSON UNIT 2 BELTLINE MATERIALS (BASED ON 10 CFR 50.61, AS ISSUED MAY 1991)

PLATE %Cu %Ni CHEMISTRY INITIAL MARGIN OF ID SURFACE SHIFT IN ADJUSTED PTS RULE OR WELD FACTOR REFERENCE UNCERTAINTY FLUENCE (f) REFERENCE REFERENCE SCREENING

(CF) TEMP. (I) (m) (XlO 9n/cm2 ) TEMP. TEMP. CRITERION (NOTE 1) (ARTpTs) (RTpTs) TEMP.

(NOTES 1, 2) (NOTES 1,2)

Weld 0.34 0.66 217.7 -56 OF 66 OF 1.83 254 OF 264 oF 300 0F 10-273 1 (.951) (215 -F) (225 oF)

Weld 11- 0.17 0.92 197.8 -80 OF 56 OF 2.01 235 OF 211 OF 300 OF 273 1 1 1 (1.61) (224 OF) (200 OF) 3

Weld 0.22 .054 101.05 -56 oF 66 OF .63 88 oF 98 OF 270 OF 1-273A (.327) (70 OF) (80 -F)

Weld 0.22 .054 101.05 -56 oF 66 oF 1.34 109 oF 119 oF 270 OF 1-273B (.696) (91 OF) (101 oF)

Weld 0.22 .054 101.05 -56 OF 66 OF .285 66 OF 76 OF 270 OF 1-273C 1 (.148) (51 OF) (61 OF)

Weld 0.22 .054 101.05 -56 OF 66 oF 3.93 137 oF 147 oF 270 OF 2-273A (.373) (74 OF) (84 OF)

Weld 0.22 .054 101.05 -56 OF 66 oF .752 93 oF 103 oF 270 oF 2-273B I I (.413) (76 OF) (86 oF)

Weld 0.22 .054 101.05 -56 oF 66 OF 1.66 115 OF 125 OF 270 OF 2-273C (.912) (98 OF) (108 OF)

Weld 0.22 .054 101.05 -56 OF 66 OF 2.01 120 OF 130 OF 270 OF

3-273A (1.61) (114 OF) (124 OF)

Weld 0.22 .054 101.05 -56 OF 66 OF .44 78 OF 88 oF 270 OF 3-273B 1 I (.353) (72 -F) (82 OF)

Weld 0.22 .054 101.05 -56 OF 66 OF .44 78 OF 88 OF 270 OF 3-273C (.353) (72 OF) (82 OF)

Enclosure 6 to Serial: RNP/93-1744 Page 3 of 3

PREDICTED REFERENCE PTS TEMPERATURES FOR H. B. ROBINSON UNIT 2 BELTLINE MATERIALS (BASED ON 10 CFR 50.61, AS ISSUED MAY 1991)

PLATE %Cu %Ni CHEMISTRY INITIAL MARGIN OF ID SURFACE SHIFT IN ADJUSTED PTS RULE OR FACTOR REFERENCE UNCERTAINTY FLUENCE (f) REFERENCE REFERENCE SCREENING

WELD (CF) TEMP. (I) (M) (XlO'9n/cm2) TEMP. TEMP. CRITERION (NOTE 1) (ARTPTS) (RTPTS) TEMP.

(NOTES 1,2) (NOTES 1,2)

Plate 0.13 0.11 63 69 OF 34 oF 1.83 73 OF 176 OF 270 OF W10201-1 (.951) (62

0F) (165

0F)

Plate 0.15 0.25 84.75 30 OF 34 oF 1.83 99 OF 163 OF 270 OF1 W10201-2 [ (.951) (84 oF) (148 OF)

Plate 0.11 0.08 51.8 36 OF 34 oF 1.83 60 OF 130 OF 270 OF W10201-3 (.951) (51 OF) (121 OF)

Plate 0.12 0.09 57.1 20 oF 34 OF 4.82 80 OF 134 OF 270 OF W10201-4 [ 1 (2.65) (72 OF) (126 OF)

Plate 0.10 0.12 51.2 20 OF 34 oF 4.82 71 OF 125 OF 270 OF W10201-5 (2.65) (65 OF) (119 oF)

Plate 0.09 0.09 44.2 45 OF 34 OF 4.82 62 OF 141 OF 270 OF W10201-6 (2.65) (56 OF) (135 OF)

Plate 0.12 0.10 58 50 OF 34 OF 2.01 69 oF 153 OF 270 OF W9807-3 (1.61) (66 OF) (150 OF)

Plate 0.15 0.10 70.5 33 OF 34 OF 2.01 84 oF 151 oF 270 -F W9807-5 (1.61) (80 OF) (147 OF)

Plate 0.14 0.15 70.5 9 OF 34 OF 2.01 84 OF 127 OF 270 OF W9807-9 (1.61) (80 OF) (123 OF)

NOTES: (1) Values are specified for both End-of-License and August 1993 (in parenthesis). End-of-License values are based on

maximum anticipated 29 effective full power years of operation through July 10, 2010.

(2) Predicted "Shift in Reference Temperature (ARTFTS)" and "Adjusted Reference Temperature (RTPTs)" were calculated using Equations 1 and 2 in 10 CFR 50.61 (shown below). Equation variables are as noted in the table headings, above.

ARTPTs = (CF) f(0 .2 8 - 0.10 log f)

RTPTS = I + M + ARTPTs