ML20137L638

Reactor Vessel Ref Temp for Pressurized Thermal Shock

ML20137L638
Person / Time
Site:	Wolf Creek
Issue date:	01/23/1986
From:	KANSAS GAS & ELECTRIC CO.
To:
Shared Package
ML20137L629	List: ML20137L625 ML20137L638
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8601280043
Download: ML20137L638 (46)

Text

i Attcchment KMLNRC 86-015 January 23, 1986 l

s l

WOLF CEEEK GEIREEATING STATION Reactor Vessel Reference Temperature For Pressurised Thermal Shock 060120004386012g82 PDR ADOCK 05 % ,

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4

TABLE OF CONTENTS PAGE

- TABLE OF CONTENTS 1 LIST OF TABLES 11 LIST OF FIGURES 111 i

i I. INTRODUCTION 1

1. The Pressurized Thermal Shock Rule 1

2. -The Calculation of RTPTS 3

]

II. NEUTRON EXPOSURE EVALUATION 5 I 1. Method of Analysis 5

2. Fast Neutron Fluence Results 9 III. MATERIAL PROPERTIES 18 -

1. Identification and Location of Beltline Region Materials 18

2. Definition of Plant Specific Material Properties 18 IV. DETERMINATION OF RTpTS. VALUES FOR BELTLINE 22 i REGION MATERIALS

1. Status of Reactor Vessel Integrity in Terms of RTpys 22 versus Fluence Results

2. Discussion of Results 23 i

V. CONCLUSIONS 26 VI. REFERENCES 28 j VII. APPENDICES A. Power Distributions A-1

8. Weld Chemistry 8-1

, C. RTpy5 Values of Wolf Creek Reactor Vessel Beltline C-1 Region Materials l

i l

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1

LIST OF TARIES Paie 11.2-1 Fast Neutron (E>l.0 MeV) Exposure at the Reactor Vessel 10 Inner Radius - 0* Azimuthal Angle 11.2-2 Fast Neutron (E>l.0 Mev) Exposure at the Reactor Vessel 11 Inner Radius - 26.5* Azimuthal Angle I

II.2-3 East Neutron (E>l.0 MeV) Exposure at the Reactor Vessel 12 Inner Radius - 30' Azimuthal Angle 11.2-4 Fast Neutron (E>l.0 MeV) Exposure at the Reactor Vessel 13 Inner Radius - 45* Azimuthal Angle 111.2-1 Wolf Creek Reactor Vessel Beltline Region Material Properties 21 IV .1 -1 RTpTS Values for Wolf Creek 24 B-1 Intermediate to Lower Shell Circumferential, Intermediate B-2 Shell Longitudinal and Lower Shell Longitudinal Weld Chemistry WOG Data Base Mean l

C -1 Wolf Creek Material Properties and RTpis Values for the C-2 Reactor Vessel Beltlin'e Region 9 Fluence = 1.0x1016 n/cm2 C-2 Wolf Creek Material Properties and RTPTS Values for the C-3 Reactor Vessel Beltline Region 9 Fluence = 5.0x1018 n/cm2 C-3 Wolf Creek Material Properties and RTPTS Values for the C -4 Reactor Vessel Beltline Region 9 Fluence = 1.0x1019 n/cm2 C-4 Wolf Creek Material Properties and RTpis Values for the C-5 Reactor Vessel Beltline Region 9 Fluence = 2.0x1019 n/cm2 C-5 Wolf Creek Material Properties and RTpiS Values for the C-6 I

Reactor Vessel Beltline Region 9 Fluence - 4.0x1019 n/cm2 C-6 Wolf Creek Material Properties and RTPTS Values for the C-7 Reactor Vessel Beltline Region 9 Fluence = 6.0x10 9-n/cm2 1

C-7 Wolf Creek Material Properties and RTpTS Values for the C-8 Reactor Vessel Beltline Region 9 Fluence = 7.0x1019 n/cm2 C-8 Wolf Creek Material Properties and RTpTS Values for the C-9 i Reactor Vessel Beltline Region 9 Current Life (.32 EFPY) l C-9 Wolf Creek Material Properties and RTpis Values for the C-10 i Reactor _ Vessel Beltline Region 9 End-of-License (31.74 EFPY)

Projected Fluence Values 11

LIST OF FIGURES PAGE 11.1-1 Wolf Creek Reactor Geometry 6 1 .

. I I .1 -2 Reactor Geometry Peak Neutron Flux Sectors 7 11.2-1 Maximum Fast Neutron (E>l.0 MeV) Fluence at the Reactor 15 Vessel Inner Radius as a Function of Full Power Operating Time 11.2-2. Maximum Fast Neutron (E>l.0'MeV) Fluence at the Reactor 16 Vessel Inner Radius as a Function of Azimuthal Angle 11.2-3 Relative Axial Variation of Fast Neutron (E>l.0 Mev) Flux 17 and Fluence at the Reactor Vessel Inner Radius 111.1-1 Identification and Location of Beltline Region Material 19 for the Wolf Creek Nuclear Plant Reactor Vessel IV .1 -1 Wolf Creek Nuclear Plant RTpys Curves per PTS Rule 25 Method [1] Docketed Basemetal and WOG Data Base Mean Weld Materials Properties

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A-1 Design Basis Relative Assembly Power Distribution A-2 I

S 4 111

. . , . _ , , . . . - - - _ _ - - ~ , _ _

SECTION I INTRODUCTION The purpose of this report is to submit the reference temperature for pressurized thermal shock (RTPTS) values f r the Wolf Creek reactor vessel to address the Pressurized Thermal Shock (PTS) Rule.Section I discusses the Rule and provides the methodology for calculating RT PTS. Section 11 presents the results of the neutron exposure evaluation assessing the effect that design basis core power distributions have on neutron fluence levels at the reactor vessel.Section III provides the reactor vessel beltline region material properties.Section IV provides the RT ca culations from PTS present through the projected end-of-license fluence values.

I.1 THE PRESSURIZED THERMAL SHOCK RULE The Pressurized Thermal Shock (PTS) Rule [1] was approved by the U.S. Nuclear Regulatory Commissioners on June 20, 1985, and appeared in the Federal Register as an amendment to 10CFR 50.61(b) on July 23, 1985. The date that the Rule was published in the Federal Register is the date that the Rule

~

became a regulatory requirement.

The Rule outlines regulations to address the potential for pressurized thermal shock (PTS) of pressurized water reactor (PWR) vessels in nuclear power plants that are operated with a license f rom the United States Nuclear Regulatory Commissi'n (USNRC). PTS events have been shown f rom operating experience to be transients th3t result in a rapid and severe cooldown in the primary system coincident with a high or increasing primary system pressure. The PTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may produce the propagation of flaws postulated to exist near the inner wall surface, thereby potentially af fecting the integrity 4

of the vessel.

I l

I l

I The Rule establishes the following requirements for all domestic, operating PWRs:

4 o Establishes the RTPTS (measure of fracture resistance) screening ,

criterion for the reactor vessel beltline region 270*F for plates, forgings, axial welds 300*F for circumferential weld materials o 6 Months From Date of Rule: All plants must submit their present RT PTS

! values (per the prescribed methodology) and projected Ri py3 values at

the expiration date of the operating license. The date that this j submittal must be received by the NRC for plants with operating licenses j is January 23, 1986, and must be updated whenever changes in core loadings, surveillance measurements or other information indicates a significant change in projected values.

o 9 Months From Date of Rule: Plants projected to exceed the PTS Screening I Criterion shall submit an analysis and a schedule for implementation of such flux reduction programs as are reasonably practicable to avoid

]

reaching the screening criterion. The data for this submittal must be i received by the NRC for plants with operating. licenses by April 23, 1986.

o Requires plant-specific PTS Safety Analyses before a plant is within l

3 years of reaching the screening criterion, including analyses of I

alternatives to minimize the PTS concern.

o Requires NRC approval for operation beyond the screening criterion.

4 For applicants of operating licenses, values of the projected RT PTS are to be provided in the Final Safety Analysis Report. This requirement is added as 1 *

) part of 10CFR 50.34 In the Rule, the NRC provides guidance regarding the calculation of the toughness state of the reactor vessel materials - the " reference temperature .

for nil ductility transition" (RTNOT). For purposes of the Rule, RTNDT 5 now defined as "the reference temperature for pressurized thermal shock" i 2 l

i (RTPTS) and calculated as prescribed by 10CFR 50.61(b) of the Rule. Each USNRC licensed PWR must submit a projection of RT PTS values fr a the time of the submittal to the license expiration date. This assessment must be

, submitted within 6 months after the effective date of the Rule, on January 23, 1986, with updates whenever changes occur af fecting projected values. The calculation must be made for each weld and plate, or forging, in the reactor

vessel beltline. The purpose of this report is to provide the RT py3 values i for the Wolf Creek Nuclear Plant.

I 1.2 THE CALCULATION OF RT PTS I

In the PTS Rule, the NRC Staff has selected a conservative and unitorm method for determining plant-specific values of RTPTS * *9 " '

l The prescribed equations in the PTS rule for calculating RTPTS are actually '

one of several ways to calculate RT NOT. For the purpose of comparison with the screening criterion, the value of RT py3 for the reactor vessel must be

calculated for each weld and. plate, or forging in the beltline region as given

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below. For each material, RT py3 is the lower of the results given by l

! Equations 1 and 2.

1' Equation 1:

RTPTS = 1 + M & (-10 + m(Cu) + 350(Cu)(M)] f Equation 2:

0 RT PTS = 1 + M + 283 f .m i

j where

.  ! = the initial reference transition temperature of the unirradiated material i measured as defined in the ASME Boiler and Pressure Vessel Code,Section III, f Paragraph NB-2331. If a measured value is not available, the following l ,

i 3 i __ _ _ _ ._-~ - . - _ _ _ . . _ _ _ _

! generic mean values must be used: 0*F for welds made with Linde 80 flux, and i

-56*F for welds made with Linde 0091,1092 and 124 and ARCOS B-5 weld fluxes.

M = the margin to be added to cover uncertainties in the values of initial

• RTNDT, c pper and nickel content, fluence, and calculation procedures. In Equation 1, M=48'F if a measured value of I was used, and M-59'F if the generic mean value of I was used. In Equation 2 M 0*F if a measured value of I was used, and M 34*F if the generic mean value of I was used.

i J Cu and Ni = the best estimate weight percent of copper and nickel in the material.

f = the maximum neutron fluence, in units of 10I9n/cm2 (E greater than or -

, equal to 1 MeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence for the period of service in question.

j Note, since the chemistry values given in equations 1 and 2 are best estimate mean values, and the margin, M causes the RT values to be upper bound PTS predictions, the mean material chemistry values are to be used, when

available, so as not to compound conservatism. '

1 4

a 1

l 4

SECTION 11 NEUTRON EXPOSURE EVALUATION This section presents the results from the application of the design basis core power distribution to the Wolf Creek reactor vessel for the Kansas Gas and Electric Company.

11.1 METHOD OF ANALYSIS A plan view of the Wolf Creek reactor geometry at the core midplane is shown in Figure 11.1-1. The reactor core exhibits 1/8th symmetry. Four neutron pads, differing in width, are attached to the outer surface of the core barrel. The pads are asymmetrically located, positioning the surveillance capsules at 58.5*, 61.0*, 121.5*, 238.5*, 241.0* and 301.5*.

A plan view of the 45* sectors representative of the reactor vessel peak neutron flux locations is depicted in Figure !!.1-2. The 0* azimuth of Figure 11.1-2 is representative of the reactor vessel longitudinal weld at 90*. The 30* azimuth is representative of the longitudinal welds at 210* and 330*. The 26.5* azimuth is representative of the peak flux locations (26.5*,153.5*,

206.5* and 333.5*) for the reactor vessel circumferential weld and plate material.

In performing the fast neutron exposure evaluations for the reactor geometry shown in Figure 11.1-2, a single computation was utilized to provide baseline data derived from a design basis core power distribution. A forward transport calculation was carried out in R,e geometry using the 00T discrete ordinates code [2] and the SAILOR cross-section library (3). The SAILOR library is a 47 group ENDF/B-IV based data set produced specifically for light water reactor applications. Anisotropic scattering is treated with a P 3expansion of the cross-sections. An 56 angular quadrature was used in the calculation.

The design basis core power distribution utilized in the analysis was derived from statistical studies of long-term operation of Westinghouse 4-loop 5

16173 1 A OR VESSEL O-CORE BARREL i

(301.5') Z CAPSULE U (58.S*)

l I v (61.0-)

s J s8.s-"l[s8.s l- f f

270' <--- -

--- 90' 35' 2.s'

,/ -

/ N 50'13 75*

(241 O') Y  ! '\ '

/

(238.5') X ,

w (121.5-)

210' 180' PLAN VIEW Figure 11.1 1. Wolf Creek Reactor Geometry .

6 1

16172 5 360*

180*

180*

O' REACTOR VESSEL I'

l 1 26.5*

l / 32.5* 315' i / 225' i 135*

l 45' i PEAK ,

V////

I FLUX

/

i /

I I

I mxxmmxxy i

N

\

1 xxxxxxxx l

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( / NEUTRON PAD i

I I / CORE BARREL I /

1 I /

j / BAFFLE l

I /

I /

I /

I /

! l /

/

I >

I /

l

/

1 /

I -

I '

/

p' I

i Figure 11.12. Reactor Geometry Peak Neutron Flux Sectors i

i 7 i

plants. Inherent in the development of this design basis core power j distribution is the use of an out-in fuel management strategy; i.e., fresh l fuel on the core periphery. Furthermore, for the peripheral fuel assemblies,

a 2o uncertainty derived from tne 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 l +2a level for a large number of fuel cycles, the use of this design basis

! distribution is expected to yield conservative results.

The specific power distribution data used in the analysis is provided in

! Appendix A of this report. The data depicted in Appendix A represent cycle i averaged relative assembly powers. Therefore, the results are in terms of fuel cycle averaged neutron flux which when multiplied by the fuel cycle length yields the incremental fast neutron fluence.

l The current neutron fluences, based on irradiation times computed f rom month 13 gross thermal energy production (MW-hrs) pubitshed in the NRC Grey Book (NUREG-0020), are defined as of November 30, 1905. The projection of reactor i vessel fast neutron fluence into the future to the expiration date of the I operating license requires that a few key assumptions be made. First, the present operating license for Wolf Creek expires on March 11, 2025 (forty I years after the operating license was issued). Secondly, the projection j assumes an 80 percent capacity factor. The current and license expiration l

dates correspond to 0.32 and 31.74 Equivalent Full Power Years (EFPY) of 1 operation, respectively.

I I The transport methodology, using the SAILOR cross-section library has been benchmarked against the Oak Ridge National Laboratory (ORNL) Poolside Critical l Assembly (PCA) facility as well as against the Westinghouse power reactor j surveillance capsule data base (4]. The benchmarking studies indicate that the use of SAILOR cross-sections and generic design basis power distributions produces flux levels that tend to be conservative by 7-22%.

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8 l l l E ___ _ . _ _ _ _ _ _ - _ . _ _ _ _ - _ _ _ - _ . _ _ _ . -. _'

11.2 FAST NEUTRON FLUENCE RESULTS I.

Calculated fast neutron (E >1.0 MeV) exposure results for Wolf Creek are presented in Tables !!.2-1 through 11.2-4 and in Figures !!.2-1 through 11.2-3. Data is preser.ted at several azimuthal locations on the inner radius of the reactor vessel.

In Tables 11.2-1 through !!.2-4 design basis maximum neutron flux and fluence i

) levels at 0*, 26.5* (peak flux location), 30*, and 45' on the reactor vessel i a

inner radius are listed for present (0.32 EFPY), and end of license (31.74

]

EFPY) irradiation times. The design basis fluence levels are based on generic 4-Ioop core power distributions at the nominal + 2o level.

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._ ._ _ _ ____ _ _ _ - - _ _ _ _ _ _ _. _ _ _ __ _. _ __ _ ___m _

i TA8LE 11.2-1 FAST NEUTRON (E > 1.0 MeV) EXPOSbnE AT THE REACTOR I VESSEL INNER RADIUS - O' AZIMUTHAL ANGLE III ,

r 1

Elapsed Design Basis Beltline Region l Irradiation Irradiation Flux (b) Cumulative Fluence Interval Time (EFPV) (n/cm2-sec) (n/cm2)

CY-1 (11/30/85)(C) 0.32 1.83 x 10 10 1.85 x 10 II i

10 I 11/30/85 - 3/11/2025 Id) 31.74 1.83 x 10 1.83 x 10 '  !'

I 1

! (a) Applicable to 0*, 90', 180* & 270' azimuthal locations on the core I beltline.  !

(b) Design basis flux at 3411 MWth.

) (c) 11/30/85 is the date at which the current neutron fluences are defined.

) (d) Exposure period to the license expiration date of 3/11/2025.

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f i <

1 i

4 l i I

j to i

TA8LE 11.2-2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE REACTOR VESSEL INNER RADIUS - 26.5' AZIMUTHAL ANGLE I *)

Elapsed Design Basis Beltline Region Irradiation irradiation F1 x(b) Cumulative Fluence Interval Time (EFPY) (n/c -sec) (n/cm2) 10 II CV-1 (11/30/85)(*) 0.32 3.14 x 10 3.17 x 10 10 11/30/85 - 3/11/2025 II 31.74 3.14 x 10 3.14 x 10" (a) Applicable to 26.5', 153.5', 206.5* and 333.5* azimuthal locations on the core beltline.

(b) Design basis flux at which the current neutron fluences are defined.

(c) 11/30/85 is the date at which the current neutron fluences are defined.

(d) Exposure period to the license expiration date of 3/11/2025.

11

i 1

i TABLE 11.2-3 i FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE REACTOR VESSEL INNER RADIUS - 30* AZIMUTHAL ANGLE

• i l

Elapsed Design })s Beltline Region  ;

Irradiation i Cumulative Fluence i Irradiation Flyn i Interval Time (EFPY) (n/cW -sec) (n/cm2) 10 II f CY-1 (11/30/85)(D) 0.32 3.02 x 10 3.05 x 10

.i i 11/30/85 - 3/11/2025(') 31.74 3.02 x 10 10 3.02 x 10I '

l 1  !

I j (a) Design basis flux at 3411 MWth.

i j (b) 11/30/85 is the date at which the current neutron fluences are defined.

1 l (c) Exposure period to the license expiration date of 3/11/2025.

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a 1

i t  !

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i l i l i  ;

1  !

e I e

f f

f i

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. _ . . _ . _ _ _ . _ _ _ , ~ , - . _ _ _ _ . . . - . _ _ . . , , , . . _ - - . _ _ _ . - - , _ , _ , . . . . - _ . . . - - . _ , , _ _ - . . - - _ , . . - . , _ _ _ _ _ , _ . . . _ _ _ _ _ _ . _ , . - . . ,

TABLE 11.2-4 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE REACTOR VESSEL INNER RA010S - 45' AllMUTHAL ANGLE Elapsed Design Basis Beltline Region '

Irradiation irradiation Fi (n/cmgx(b) Cumulative Fluence Interval Time (EFPY) ~ -sec) (n/cm2) 10 I CV-1 (11/30/85)IDI 0.32 2.92 x 10 2.95 x 10 10 I9 11/30/85 - 3/11/2025(C) 31.74 2.92 x 10 2.92 x 10 .

i (a) Design basis flux at 3411 MWth.

(b) 11/30/85 is the date at which the current neutron fluences are defined.

(c) Exposure period to the license expiration date of 3/11/2025.

r I

e 13 l

l 1

I Graphical presentation of the design basis f ast neutron (E>1.0 Mev) fluence at ,

i key locations on the reactor vessel are shown in Figure !!.2-1 as a function ~

of full power operating time. Reactor vessel data is presented for the peak .

flux location (s) on the shell plates, circumferential weld, and longitudinal

welds. The fluence projections in Figure 11.2-1 have been carried out to 31.74 ef fective full power years (license expiration date of 3/11/2025. j i In Figure 11.2-2, the azimuthal variation of maximum fast neutron (E > 1.0 1

j MeV) fluence at the inner radius of the reactor vessel is presented as a  !

! function of azimuthal angle. Data are presented for both current and j projected end-of-life conditions. In Figure 11.2-3, the relative axial l variation of f ast neutron flux and fluence over the beltline region of the reactor vessel is presented. -

I f A two-dimensional description of the f ast neutron exposure of the reactor  !

l vessel inner radius can be constructed using the data given in Figures 11.2-2 and !!.2-3 along with the relation  ;

) +(e,Z) = +(e) G(Z) -

I where: + (e,Z) = Fast neutron fluence at location e, Z on the reactor i vessel inner radius i l [

! + (e) = Fast neutron fluence at azimuthat location e on the i <

i reactor vessel inner radius from Figure !!.2-2 l I ,

G (Z) = Relative fast neutron flux at axial position Z. from

Figure !!.2-3 L

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16172 7 1020 _

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6 LONGITUDINAL WELDS AT 21O' & 330*

4 -

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PLATE & CIRCUMFERENTIAL WELD AT 26.5*,153.5*,

2 -

206.5* a 333.5*

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N 1019 _

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LONGITUDINAL WELD, 90*

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1017 O 5 10 15 20 25 30 35 OPERATING TIME (EFPY)

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Figure 11.21 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Reactor VesselInner Radius as a Function of Full Power Operating Time ,

1 3 15

j .

16172-8 I020 _

8 -

6 -

4 -

3I.74 EFPY 2 -

N 1o19 _

$ e -

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6 -

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o.32 EFPY i

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I I I I I I 1017 30 40 60 70

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o lo 20 50 AZIMUTHAL ANGLE (DEGREE)

Figure 11.2 2. Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Reactor l VesselInner Radius as a Function of Azimuthal Angle

' 16

- - - .- ~ -

18172 6 10 _

- 8 _-

6 -

4 _

2 -

1.0 _

W 8 -

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l T O.01 _

8 -

6 -

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2 - CORE MIDPLANE O.001 ' I

-300 -200 -100 O I00 200 300 400 DISTANCE FROM CORE MIDLANE (cm)

Figure l1.2 3. Relative Axial Variation of Fast Neutron (E > 1.0 MeV) Flux and Fluence at the Reactor Vessel Inner Radius I

17

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SECTION III MATERIAL PROPERTIES _

For the RT calculation, the best estimate copper and nickel chemical PTS composition of the reactor vessel beltline material is necessary. The material properties for the Wolf Creek beltline region will be presented in this section.

111.1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS The beltline region is defined by.the Rule [1] to be "the region of the reactor vessel (shell material including welds, heat af fected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience suf ficient neutron irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage."

Figure 111.1-1 identifies the location of all beltline region materials for the Wolf Creek reactor vessel.

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III.2 DEFINITION OF PLANT-SPECIFIC MATERIAL PROPERTIES The pertinent chemical and mechanical properties of the beltline region plate and weld materials of the Wolf Creek reactor vessel are given in Table 111.2-1. Although phosphorus is no longer used in the calculation of RT PTS i with respect to the PTS rule [1], it is given for reference since it is used in the Regulatory Guide 1.99, Revision 1 trend curve (5].

Property values for the shell plates were derived from the vessel fabrication test certificates. Various studies have been conducted to determine the weld chemistry for the Wolf Creek Nuclear Power Plant. Values corresponding to two sources are presented in Table 111.2-1. The docketed values are those which were taken from vessel fabrication test certificates and can bo.found in reference 6 (the nickel values were not reported). The second source of weld 18

4 16172 4 CIRCUMFERENTIAL SEAMS VERTICAL SEAMS

• 270* -

R2OOS-4 101-124B IOI-124C 20.9" t

180* -- --

O' CORE \ 'l l N

4 4 R2OO5-2 R2OOS-l j 144" 8 b IOl-124A '

90-

! (______ @

C 101-171 15.I" J .

J 270' R2508-I .

101-1428 IOI-142C 50.8" "

4l 180* --

y O*

i R2508-3 L J R2508-2 l01-142A 90' l

j Figure 111.1-1. Identification and Location of Beltline Region Material for the Wolf ,

Creek Nuclear Plant Reactor Vessel i

19 l

4 chemistry was obtained by performing a search of the WOG Materials Data Base.

The Data Base is a compilation of d large amount of weld chemistry data for many dif ferent wire / flux combinations. A search, as _ listed in Appendix B, was performed for materials having the identical weld wire heat number as in the Wolf Creek vessel, considering any flux lot, since the flux lot does not affect Cu, Ni or P content. The copper, nickel and phosphorous values are l averaged, for all the data found for a particular wire,- and the standard deviations are calculated. The WOG Data Base mean values are then used to calculate the weld RT PTS values.

e 20

TABLE III.2-1 WOLF CREEK REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni P 1*

(wt %) (wt %) (wt %) [*H Intermediate Shell Plate R2005-1:

Docketed Value .04 .66 .008 -20 Intermediate Shell Flate R2005-2:

Docketed Value .04 .64 .007 -20 Intermediate Shell Plate R2005-3:

Docketed Value .05 .63 .007 -20 Lower Shell Plate R2508-1:

Docketed Value .09 .67 .009 0 Lower Shell Plate R2508-2:

Docketed Value ,

.06 .64 .008 10 Lower Shell Plate R2508-3: -

Docketed Value .07 .62 .008 40 Intermediate (101-124A,B,C) and Lower Shell (101-142A,B C) Longitudinal Welds, Wire Heat 90146/Linde Flux 0091 (Lot 0842):

Docketed Value .04 .04 .006 -50

, 'WOG Data Base Mean(b) .04 .06 .006 ---

Intermediate to Lower Shell Circumferential Weld (101-171), Wire Heat 90146/Linde Flux 124 (Lot 1061):

Docketed Value .05 .05 .007 -50 WOG Data Base Mean(b) .04 .06 .006 ---

(a) All initial RTNOT values are actual values.

(b) See Appendix B 21

SECTION IV DETERMINATION OF RT VALUES FOR BELTLINE REGION MATERIALS PTS Using the methodology prescribed in Section 1.2, the results of the fast neutron exposure provided in Section II, and the material properties discussed in Section III, the RT PTS values for Wolf Creek can now be determined.

IV.1 STATUS OF REACTOR VESSEL INTEGRITY IN TERMS OF RT PTS VERSUS FLUENCE RESULTS Using the prescribed PTS Rule methodology, RT PTS va ues were generated for all beltline region materials of the Wolf Creek reactor vessel as a function of several fluence values and pertinent vessel lifetimes. The tabulated

, results from the total evaluation are presented in Appendix C for all beltline region materials.

Figure IV.1-1 presents the RT PTS values for the welds and limiting shell plate of the Wolf Creek vessel in terms of RT PTS versus fluence

• curves.

The curves in these figures can be used:

o to provide guidelines to evaluate fuel reload options in relation to the NRC RT PTS screening criterion for PTS (i.e., RT PTS values can be readily projected for any options under consideration, provided fluence is known), and o to show RT PTS values through end-of-license (31.74 EPFY) using actual and projected fluence.

• The EFPY can be determined using Figure 11.2-1.

22

Table IV.1-1 provides a summary of the RT values for all beltline region PTS materials for the lifetimes of interest.

IV.2 DISCUSSION OF RESULTS As shown in Figure IV.1-1, the base metal is the governing location relative to PTS. All the RT va ues ema n we e w eN screen ng va ues for PTS PTS using the projected fluence values through end-of-license (31.74 EFPY).

e f

I l

4 23

, . _ . , 7 _ , - _

i TABLE IV.1-1 RT VALUES FOR WOLF CREEK PTS

.RT VALUES (*F)

PTS Present End-of-License Screening Vessel Material (0.32 EFPY) (31.74 EFPY) Criteria Internediate Shell Plate 35 53 270 R2005-1 Intermediate Shell Plate 35 52 270 R2005-2 Intermediate Shell Plate 38 61 270 R2005-3 Lower Shell Plate R2508-1 69 121 270 Lower Shell Plate R2508 70 101 270 l

Lower Shell Plate R2508-3 103 140 270 Longitudinal Welds - 2 11 270 101 -1248, 101 -124C ,

! 101-1428 & 101-142C Longitudinal Welds - 1 9 270 101-124 A, 101-142A Circumferential Weld 101-171 2 11 300 24 l.

i

16172 3 350 l

-NRC RTPTS SCREENING VALUE (3OO'F) - CIRCUMFERENTIAL WELDS 300 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

/

250 i NRC RTPTS SCREENING VALUE (270*F) - PLATE AND LONGITUDINAL WELD E 200 tn

@ 150 - IMIT i00 -

50 -

CIRCUMFERENTIAL AND LIMITING LONGITUDINAL WELDS Ill)

O

' ' ' ' i t t i F I I I I 11 1018 1019 1020 NEUTRON FLUENCE (n/cm2)

LEGENO:

e = LIMITING SHELL PLATE RTPTS AT END OF LICENCE (31.74 EFPY)

A = CIRCUMFERENTIAL WELD RTPTS AT END OF LICENCE

5. = LIMITING LONGITUDINAL WELD RTPTS AT END OF LICENCE Figure IV.1-1. Wolf Creek Nuclear Plant RTPTS Curves per PTS Rule Method (1)

Docketed Basemetal & WOG Database Mean Weld Material Properties f

l 25 1

i

SECTION V CONCLUSIONS Calculations have been ccmpleted in order to submit RT PTS values for the gp Wolf Creek reactor vessel in meeting the requirements of the NRC Rule for Pressurized Thermal Shock [1]. This work entailed a neutron exposure evaluation and a reactor vessel material study in order to determine the RT values.

PTS Detailed fast neutron exposure evaluations using design basis core-power distributions and state-of-the-art neutron transport methodology have been completed for the Wolf Creek pressure vessel. Explicit calculations were performed for Cycle 1 through November 30, 1985. Projection of the fast neutron exposure to license expiration was based on continued implementation of the design ' basis core power distribution. Tabular and graphical representation of the design basis fast neutron (E>1.0 Mev) fluence at key beltline material locations are presented. The maximum fast neutron flux

~

occurs azimuthally at 26.5* , 153.5* , 206.5* and 333.5*. The maximum fast I9 2 fluence for shell plate and circumferential weldment is 3.14 x 10 n/cm at license expiration (31.74 EFPY). The maximum fast fluence for the longitudinal weldment is 3.02 x 10 I9 n/cm2 at license expiration.

Material properties for the Wolf Creek reactor vessel beltline region components were determined. The pertinent chemical and mechanical properties for the shell plates remain the same as those that were reported in the original vessel fabrication test certificates. The weld properties were mean values obtained from the WOG materials data base.

Using the prescribed PTS Rule methodology, RT PTS values were generated for all beltline materials of the Wolf Creek reactor vessel as a function of several fluence values and appropriate vessel lifetimes. All of the RT PTS values remain below the NRC screening values for, PTS using the projected fluence exposure through 31.74 EFPf. The most limiting value at

- s J

26

l end-of-license (31.74 EFPY) is 140*F at the lower shell plate which is well below the applicable 270*F' screening criterion. .

l The results provided in this report comply with the initial 6 months submittal ,

requirements of the USNRC PTS Rule and demonstrates that no further submittals are required unless changes in core loadings, surveillance measurements, or other information indicates a significant change in projected values [1].

O f

l i

l l

l I

l -

I .

27 l

t

SECTION VI REFERENCES

, 1. Nuclear Regulatory Commission,10CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," Federal Register, Vol. 50, No.141, July 23, 1985.

- 2. Soltesz, R. G., Disney, R. K., Jedruch, J. and Ziegler, S. L., " 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.

3. " SAILOR RSIC Data Library Collection DLC-76," Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray,3 P , Cross-Section Library for Light Water Reactors.

4. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology -

to be published. .

5. " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," Regulatory Guide 1.99 - Revision 1 U.S. Nuclear Regulatory Commission, Washington, April 1977.

6. Table 5.3-3, " Wolf Creek Unit 1 Reactor Vessel Material Properties", Wolf Creek FSAR.

28

APPENDIX A s

POWER DISTRIBUTIONS Core power distributions used in the design basis fast neutron exposure analysis of the Wolf Creek pressure vessel were supplied by Westinghouse's Nuclear Fuels Division (NFD). The values are supplied in'the form of a assembly-wire radial power distribution for all fuel assemblies in the core, rod-by-rod power distributions in the peripheral fuel assemblies, and an axial power distribution.

The assembly-by-assembly radial power distribution is derived from a statistical analysis of calculated distributions from all available independent cycles (including optimized fuel analyses). Calculations are available for 23 cycles in 10 four-loop reactors. On the average, the long tenn distribution represents an upper tolerance limit of beginning-of-cycle (B0C) and end-of-cycle (E0C) power in peripheral assemblies, based on a 95-percent probability with 95-percent confidence. The distribution is biased' to account for observed differences between calculated and measured power in peripheral assemblies. The rod-by-rod power distributions in the peripheral assemblies are derived from typical equilibrium EOC conditions. The axial power distribution is time averaged and is representative of 40 years of plant operation.

A schematic diagram of the core configuration applicable to the Wolf Creek plant is shown in Figure A-1. Design basis cycle averaged relative assembly powers for Wolf Creek are listed in Figure A-1.

A-1 l

16173-2 k

O.86 -- - -- -

1.07 1.06 0.87 0.94 1.09 0.90 1.04 1.12 0.92 1.09 1.06 0.88 1.10 1.04 1.05 0.87 0.87 1.07 1.00 1.05 '

l.02 1.10 1.00 1.05 1.15 0.75 B

1.06 1.09 1.01 0.81 k

l l

Figure A-1. Design Basis Relative Assembly Power Distribution l

A-2 l l

l --. - - - - - .. .. - _.

,.l

APPENDIX 8 9

WELD CHEMISTRY Table B-1 provides the weld data output from the WOG Material Data Base.

Given are the searches of all available data for the wire heat in the Wolf I

Creek reactor vessel beltline region. The pertinent material chemical compositions are given, along with the wire / flux identification. The mean chemistry values and the population standard deviation are then calculated.

The mean values of copper and nickel are used in the RT analysis.

PTS Weld Chemistry Data Source and Plant:

CE -

Combustion Engineering Cu -

Weight % of Copper Ni -

Weight % of Nickel P -

Weight % of Phosphorous SAP -

Wolf Creek SC -

Surveillance Capsule Si -

Weight % of Silicon TGX -

South Texas WQ -

Weld Qualification 4

B-1

- s a

. L L

L L s

s s . L E L u .

a s

s s

. E H l S l 6 E h

S ED S L g

s s E N mR E s N R WO 0E R W s s O E L LT E u s ! N E N d E . s s i G C L LI 0 C . s s P . L N LL L N s 1 sI . A EE 0 A s s

s E O L H H1 O L T L s s C T LSS l

i s S I E I s H

l u s

a s

K R E WEE R R L T R T W l i 8 l E nW s s

s Go s a

N uN O ouM I S I L t l S s

s -

W s .

s s

C n T a N . s -

. s Td A 1 a A s L s P P PPr1 P P A A A e6 A A S SS S1 S S s

s s

s

_ I 1 s s DS s s 3 1 3 3 s s

s l I s 3 8 s s

iP M s s i

-_ 0

.- 2 0

9 0

0 3 5 s 9 8 s

f. L s S 5 1 4 3 1 s E

l s  : s T i s 0 0 0 0 0 s a L s s 0 0 a t

s 0 0 s I

i s s

7 0  %

0 5

0 0

@0 1 0 0 s s

s L s P s

. 0 00 s L s 0 0 0 s i W s . 0 0 s s

t s .

s 0 0 0 0 B s f

s - 5 4 9 M5 s 1 L s 1 s N

. 0 G.

0 4 s 0 6 s 1 A s

. 0 0 0 2 s NN i I s

s

. - M. 0 s s

t s 0 0 0 0 0 s s 5 4 4 s il D s u 0 0 0 3 4 s

_ 1 l iJ t N s C s 0 0 0 3

3 Us s

- F l A s s

3 5 s 4 0 s I

l Hl E s 00 s s

UuN s s O 0 s s

CN s s s t.

i HOE s N s E E s

s r

f l LS s MA C .

s s

s AC A s EAT R U S u,

G W. s

~l LD s s

L B O E S E E E s

s s h C C C s LLtL.A s s

s s

s iHH. T t

1 A s

s s

t uT l O 1 2 I s

s 6 4 s i

t D s s

F L 0 6 M s 0 s H s s

1 I s

kL s s

s s

i L W s s s

WO s s 1 s HL s s

4 2

9 0

4 2

s s

'a t E 1 0 1 s s uP s i D s l V E E i .

s aj N s FI s

. t

. h!

0 h

t t

s a

T A s . ] i s s

s . L L L a

s . s EL s s s s

I A s s s AN I I s EE s R P

._ s s

s Y s DD s s

I T W 4 4

4 s

s EL J s l B B s s s NI s s

s s

R EL I.

s s

_ s s s s TN s s

s s

NO T s s s s

I L A s ET ' s s

0 s R A  %  %  % s P s I E 1 s E s W N 1 0 01 0 . s

._ E s s 9 9 9 v s T s 3 6 6 e ss C s E s I 5 4 3 6

n d. s -

L s I 0 1 ad s E a 0 0 0 et s S s ss a o$

c

_ t .

l APPENDIX C RT VALUES OF WOLF CREEK PTS REACTOR VESSEL BELTLINE REGION MATERIALS Tables C-1 through C-9 provide the.RT values, as a function of fluence, PTS for all beltline region materials of the Wolf Creek reactor vessel. The RT PTS values are calculated in accordance with the PTS rule, which is Reference 1 in the main body of this report. The vessel location numbers in the following tables correspond to the vessel materials identified below and in Table III.2-1 of the main report.

Location Vessel Material 1 Intermediate Shell Plate - R2005-1 2 Intermediate Shell Plate - R2005-2 3 Intermediate Shell Plate-- R2005-3 4 Lower Shell Plate - R2508-1 1

5 Lower Shell Plate - R2508-2 6 Lower Shell Plate - R2508-3 7 Longitudinal Welds - Wire Heat 90146 8 Circumferential Welds - Wire Heat 90146 1

e t

C-1

l i l

! i

. InDtE C-I  !'

! WOLF CREEK MATERIAL PROFERTIES AND RTPTS VALUES FOR i THE REACTOR VESSEL BELTLINE REGION O FLUENCE =1E18 N/CMa*2 -

t  :

e

[

• i i  !

l !ID ! PLANT! CU  !. NI  ! P VALUE  ! TYPE ! FLUENCE ! RTPTS  !

j _________ - ____________ _ _- .__'_ _.______________________ . _ _ '

1 j 1. SAP O.040 O.660 O. Os * -2a At:IUAL B.M. O.10E+19 38 3

2 ' SAP O.040 O.640 O.007 -20 . ACTUAL B.M. O.10E+19 3G '

] 3 SAP O.050 0.630 0.007 -20 ACTUAL ' B.M. O.10E+19 41 i 4 SAP O.090 0.670 0.009 O ACTUAL B.M. O.10E+19 77 l ?r0 '5 SAP O.060 O.64O O.000 10 ACTUAtl B.M. O.10E+19, 75  ;

l 6- SAP O.070 O.620 O.000 40 ACTUAL B.M. O.10E+19 108  ;

-50 ACTUAL L.W.' O.10E+19 3

! 7 SAP O.040 0.060 0.006 O SAP O.040 O.060 O.606 -50 ACTUAL C.W. O.10E+19 3 i

< l i

Notes: 'ID = Location of vessel material (see page C-1)  !

t l

I =

Initial value of RTNDT actual or estimated

" ACTUAL" denotes that the listed values of initial RT NDT are actual rather than estimated values

.Value =  ;

3 B.M. =

Base Metal (Plates)

L.W. = Longitudinal Weld

C.W. = Circumferential Weld Reference Temperatures are in F '

r a . . . ., ..

. TADLE C-2 WOLF CREEK MATERIAL PROPERTIES AND RTPTS VALUES FOR THE REACTOR VESSEL BELTLINE REGION O FLUENCE =5E18 N/CM**2

!ID ! PLANT! CU  ! NI  ! P  ! I  ! VALUE  ! TYPE ! FLUENCE ! RTPTS !

1 SAP O.040 0.660 0.000 -20 ACTUAL B.M. 0.50E+19 43 2' SAP .O.040 O.640 0.007 -20' ACTUAL B.M. O.50E+19 43-3 SAP O.050 0.630 0.007 -20 ACTUAL B.M. 0.50E+19 ~4 8 4 SAP O.090 0.670 0.009 O ACTUAL B.M. 0.50E+19 92 p 5 SAP O.060 0.640 0.008 10 , ACTUAL B.M. 0.50E+19 84 w 6 SAP O.070 0.620 0.008 40 ACTUAL _p.M. 0.50E+19 120 7 SAP O.040 0.060 0.006 -50 ACTUAL L.W. 0.50E+19 e, 8 SAP O.94o G.06o O.006 AC TilAt. C.W. 4.SOE*19 c,

9

. TABLE C-3 WOLF CREEK MATERIAL PROPERTIES AND RTPTS VALUES FOR THE REACTOR VESSEL BELTLINE REGION O FLUENCE =1E19 N/CM**2

!ID ! PLANT! CU  ! NI  ! P  ! I  ! VALUE ! TYPE !' FLUENCE ! RTPTS '

1 SAP O.040 0.660 0.008 -20 ACTUAL B.M. O 10E+2O 46

2. SAP O.040 0.640 0.007 -20 ISCTUAL B.M. O.10E+20 46 3 SAP O.050 0.630 0.007 -20 ACTUAL B.M. O.10E+2O 53 4 SAP O.090 0.670 0.009 O ACTUAL B.M. O.10E+20 101 p 5 ' SAP O.060 0.640 0.008 10 ACTUAL B.M. O.10E+20 90 A 6 SAP O.070 0.620 0.008 40 ACTUAL B.M. O.10E+20 126 7 SAP O.040 0.060 0.006 -50 ACTUAL L.W. O.10E+20 G 8 SAP 0.040 0.060 0.006 -50 ACTUAL. C.W. O.10E+20 e

. lAblE C-4 WOLF CREEK MATERIAL PROPERTIES AND RTPTS VALUES FOR l

THE REACTOR VESSEL BELTLINE REDION G FLUENCE =2E19 N/CM*m2

. t CU  ! NI  ! F '

I ' VAtUE  ! TYPE ! FLUENCE ! RTPTS *

!ID ! PLANT!

1 SAP 0.040' O.660 0.vvu -20 AUlUAL B.M. O.20E+20 50 2 SAP O.040 0.640 0.007 -20 ACTUAL B.M. O.20E+2O 49 3 SAP O.050.0.630 0.007 -20 ACTUAL B.M. O.20E+20 58 4 SAP O'.090 0.670 0.009 O ACTUAL B.M. O.70E+20 112 5 SAP O.060 0.640 0.000 10 ACTUAL B.M. O.20E+20 96 6 SAP O.070 0.620 0.000 40 ACTUAL B.M. O.20E+2O 134 n 7 SAP 0.040 0.060 0.006 -50 ACTUAL. L.W. O.20E+20 10

[n Ei SAP 0.040 0.060 0.006 -50 ACTilAL C.W. 0.20F+2O 10

l

. TAL*LE C-5 WOLF CREEK MATERIAL PROPERTIES AND RTPTS VALUES FOR THE REACTOR VESSEL BELTLINE REGION O FLUCNCE=4E19 N/CM**2

!ID ! PLANT! 'CU  ! NI  ! P  ! I ' VALUE

• TYPE ! FLUENCE ! RTPTS

• 1 SAP O.040, 0.660 0.000 -20 ACTUAL B.M. O.40E+20 54 2 SAP O.040 0.640 0.007 -20 ACTUAL B.M. O.40E+20 54 3 SAP O.050 3. u!O O.007 -20 ACTUAL B.M. O.40E+2O 64 4 SAP O.090 0.670 0.009 O ACTUAL B.M. O.40E+2O 126 p 5 SAP O.060 0.640 0.000 10 ACTUAL B.M. O.40E+20 104 et 6 SAP O.070 0.620 0.008 40 ACTUAL B.M. O.40E+20 143 7 SAP O.040 0.060 0.006 -50 ACTUAL L.W. O.40E+20 12 u SAP 6.040 0.060 0.' J t6 -50 AC16'iti. C.W. 6. 41 E+20 12

. g . e e g

. TABLE C-o WOLF CRFEK MATERIAL PROPERTIES AND RTPTS VALUES FOR THE REACTOR VESSEL BELTLINE REGION O FLUENCE =6E19 N/CM**2

!ID ! PLANT! CU  ! NI ' P '

I  ? VALUE  ! TYPE ! FLUENCE ! RTPTS '

1 SAP O.040 0.660 n.OOR -20 ACTUAL B.M. 0.60E+20 57 2 SAP O.040 0.640 0.007 -20 ACTUAL. B.M. O.60E+2O 57 3 SAP O.050 0.630 0.007 -20 ACTUAL B.M. 0.60E+20 , 68 4 SAP O.090 0.670 0.009 O ACTUAL B.M. O.60E+20 135 7 5 SAP O.060 0.640 0.000 10 ACTUAL B.M. 0.60E+20- 109

'8 6 SAP O.070 0.620 0.000 40 ACTUAL B.M. 0.60E+20 150 7 SAP O.040 0.060 0.006 -50 ACTUAL L.W. 0.60E+2O 14 8 SAP O.040 0.060 0.006 -50 ACTUAL C.W. o.60E+20 14

l 1

. (AHLL C-7 WOLF CREEK MATERIAL Ph0PERTIES AND RTPTS VALUES FOR THE REACTOR VESSEL DELTLINE REGION -(* FLUENCE =7E19 N/CM**2

'ID ! PLANT! CU  !

• NI  ! P

• I  ! VALUE  ! TYPE ! FLUENCE ! RTPTS !

1 SAP O.040 0.660 0.000 -20 ACTUAL B.M. 0.70E+20 59 2 SAP O.040 0.640 0.007 -20 ACTUAL B.M. 0.70E+20 58 3 SAP O.050 0.630 0.007 -20 ACTUAL D.M. 0.70E+2O 69 4 SAP O.090 0.670 0.009 O ACTUAL B.M. 0.70E+20 138 7 5 SAP O.060 0.640 0.008 10 ACTUAL D.M. O.'70E+20 112 C' 6 SAP O.070 0.620 0.008 40 ACTUAL B.M. 0.70E+20 ,152 7 SAP O.040 O.060 0.006 -50 ACTUAL L.W. O.70E+20 14 8 GAP O.040 0.060 0.006 -50 ACTUAL C.W. 0.70E+20 14

. ~ - .

I t,

TABLE C-8 l

j WOLF CREEK MATERIAL PROPERTIES AND RTPTS VALUES FOR THE REACTOR VESSEL DELTLINE REGION @ CURRENT LIFE (.32 EFPY) l 1

1 i !ID ! PLANT! CU  ! NI  ! P  ! 1  ! VALUE  ! TYPE ! FLUENCE ! RTPTS !

l 1

1 SAP O.040 0.660 0.008 -20 ACTUAL B.M. O.32E+18 35

, 2 SAP O.040 0.640 0.007 -20 ACTUAL D.M. O.32E+18 35 l 3 SAP O.050 0.630 0.007 -20 ACTUAL B.M. O.32E+18 38'

! 4 SAP O.090 0.670 0.009 O ACTUAL B.M. O.32E+18 69

• p 5 SAP. O.060 0.640 0.008 10 ACTUAL B.M. O.32E+18 70 1 e 6 SAP O.070 0.620 ~O.008 40 ACTUAL B. M. O.32E+18 103 I

7 SAP O.040 0.060 0.006 -50 ACTUAL L.W. O.31E+18 2 8 SAP O.040 0.060 0.006 -50 ACTUAL C.W. O.32E+18 2 l I 1

i

_ _ . _ _ _ . _ .__-.__.._____.-_..____..-___.___._m.___._.. . . _ . . . _ _ _ . . - . _ . .

]

l 4

IAbLE C-9 WOLF CREEK MATERIAL PROPERTIES AND RTPTS VALUES FOR THE REACTOR VESSEL i

BELTLINE REGION GEND OF LICENSE (31.74 EFPY) PROJECTED FLUENCE VALUES 3

1 l

!!D ! PLANT! CU  ! NI  ! P  !  !  ! VALUE  ! TYPE ! FLUENCE ! RTPTS !

i ----------------------------------------------------------- - - - - - - - - - .

1 SAP O.040 O.660 O.008 -20 ACTUAL B.M. O.31E+20 52

2 SAP O.040 .O.640 0.007 -20 ACTUAL B.M. O.31E+20 52 .

j 3 SAP O.050 O.630 0.007 -20 ACTUAL B.M. O.31E+20 61 j n 4 SAP O.090 O.670 O.009 O ACTUAL B.M. O.31E+20 120 4 e 5 SAP O.060- 0.640 0.000 10 ACTilAL B.M. O.31E+20 101 i

E 6 SAP O.070 O.620 O.000 40 ACTUAL D.M. O.31E+20 140 '

7 SAP O.040 O.060 0.006 -50 ACTUAL L.W. o.30E+20 11 . .

8 SAP O.040 0.060 0.006 .-50 ACTUAL C.W. O.31E+20 11 0

s l

t i

5 l

l

. . .J , .- .

, - , --