ML20137C535
| ML20137C535 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/31/1985 |
| From: | Hirst C, Lau F, Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17254A720 | List: |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR WCAP-11026, NUDOCS 8601160344 | |
| Download: ML20137C535 (74) | |
Text
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WCAP-11026 WESTINGHOUSE PROPRIETARY CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION R. E. GINNA REACTOR VESSEL FLUENCE AND RT PTS EVALUATIONS T. V. Congedo C. C. Heinecke T. E. Rens M. Weaver Work Performed fo'r Rochester Gas and Electric Corporation December 1985 APPROVED: )I ^'s *i APPROVED: IAN[ b F. L. Lau, Manager T. A. Meyer, Manager Structural Materials Radiation and Systems and Reliability Technology Analysis APPROVED:
fM W A
/C. W/ Hirst, Manager ~
Reactor Coolant System Components Licensing Although the information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its Licensees without the customer's approval.
WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR ENERGY SYSTEMS P. O. 80X 355 ,
PITTS8URGH, PENNSYLVANIA' 15230 860116034406011g 3897e:1d/120585 PDR Am O PDR P
3 TABLE OF CONTENTS PAGE TABLE OF CONTENTS i LIST OF TABLES ii LIST OF FIGURES iv
- 1. INTRODUCTION 1
- 1. The Pressurized Thermal Shock Rule 1
- 2. The Calculation of RTPTS 3 II. NEUTRON EXPOSURE EVALUATION 5
- 1. Method of Analysis 5
- 2. Fast Neutron Fluence Results 9 III. MATERIAL PROPERTIES 24
- 1. Identification and Location of Beltline Region Materials 24
- 2. Definition,of Plant Specific Material Properties 24 IV. DETERMINATION OF RTPTS VALUES FOR BELTLINE 29 REGION MATERIALS 1.' Status of Reactor Vessel Integrity in Terms of RTPTS 29 versus Fluence Results
- 2. Discussion of Results ,
31 V. CONCLUSIONS AND RfCOMMENDATIONS 33 VI. REFERENCES 35 VII. APPENDICES A. Power Distributions A -1 B. Weld Chemistry B -1 C. RTPTS Values of R. E. Ginna Reactor Vessel Beltline C -1 Region Materials r
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LIST OF TABLES
.P,a_3Le II.2-1 Fast Neutron (E>1.0 Mev) Exposure at the Pressure Vessel 10 Inner Radius - 0* Azimuthal Angle 11.2-2 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 11 Inner Radius - 14.5* Azimuthal Angle 11.2-3 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 12 Inner Radius - 30' Azimuthal Angle 11.2-4 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 13 Inner Radius - 44.5* Azimuthal Angle 11.2-5 Fast Neutron (E>1.0 MeV) Exposure at the 13* Surveillance 14 Capsule Center 11.2-6 Fast Neutron (E>1.0 MeV) Exposure at the 23' Surveillance 15 Capsule Center 11.2-7 Fast Neutron (E>l.0 MeV) Exposure at the 33* Surveillance 16 Capsule Center 111.2-1 R. E. Ginna Reactor Vessel Beltline Region Material Properties 26 IV.1-1 RTpys. Values for R. E. Ginna 32 A .1 -1 Core Power Distributions Used in the Plant Specific A-3 Fluence Analysis of R. E. Ginna B-1 R. E. Ginna Nozzle to Intermediate Shell Weld Chemistry B-2 from WOG Materials Database B-2 R. E. Ginna Intermediate to Lower Shell Weld Chemistry B-5 from WOG Materials Database C-1 RTpIS Values for R. E. Ginna Reactor Vessel Beltline C-2 Region Materials 9 Fluence = 1.0 x 1018 n/cm2 C-2 RTpIS Values for R. E. Ginna Reactor Vessel Beltline C-3 Region Materials 9 Fluence = 5.0 x 1018 n/cm2 C-3 RTpys Values for R. E. Ginna Reactor Vessel Beltline C -4 Region Materials 9 Fluence = 1.0 x 1019 n/cm2 C-4 RTpys Values for R. E. Ginna Reactor Vessel Beltline C-5 Region Materials 9 Fluence = 2 x 10 19 n/cm2 3897e:Id/121985 11
I LIST OF TABLES (Continued)
Page C-5 RT ys Values for R. E. Ginna Reactor Vessel Beltline C-6 Re ton Materials @ Fluence - 4 x 1019 n/cm2 C-6 RTPTS Values for R. E. Ginna Reactor Vessel Beltline C -7 Region Materials @ Fluence = 6.0 x 1019 n/cm2 C-7 RTpys Values for R. E. Ginna Reactor Vessel Beltline C-8 Region Materials @ Fluence = 7.0 x 1019 n/cm2 C-8 RTPTS Values for R. E. Ginna Reactor Vessel Beltline C -9 Region Materials @ Current Life (10.88 EFPY) - Plant
. Specific Fluence Values C-9 RTpys Values for R. E. Ginna Reactor Vessel Beltline C-10 Region Materials @ End of License (27.36 EFPY) -
Projected Fluence Values C-10 RTpis Values for R. E. Ginna Reactor Vessel Beltline C-11 Region Materials @ 32 EFPY - Projected Fluence Values 3897e:1d/120685 iii
LIST OF FIGURES PAGE 11.1-1 R. E. Ginne Reactor Geometry 6 11.2-1 Maximum Fast Neutron (E>l.0 MeV) Fluence at the Beltline 17 Weld Location as a Function of Full Power Operating Time - R. E. Ginna II.2-2 Maximum Current and Projected EOL Fast Neutron (E>1.0 MeV) 18 l Fluence at the Pressure Vessel Inner Radius as a Function i
of Azimuthal Angle - R. E. Ginna 11.2-3 Relative Radial Variation of Fast Neutron (E>1.0 MeV) Flux 19 and Fluence Within the Pressure Vessel Wall - R. E. Ginna 11.2-4 Relative Axial Variation of Fast Neutron (E>1.0 MeV) Flux 20 and Fluence Within the Pressure Vessel Wall - R. E. Ginna II I .1 -1 Identification and Location of Beltline Region Material 25 l
for the-R. E. Ginna Nuclear Plant Reactor Vessel IV.1 -1 R. E. Ginna Nuclear Plant RTpys Curves per PTS Rule 30 Method [1] Docketed Basemetal and WOG Data Base Mean Weld l
Materials Properties i A .1 -1 R. E'. Ginna Plant Core Description for Power Distribution A-2 1 .
Maps t
3897e:1d/121985 iv i
i
SECTION I INTRODUCTION The purpose of this report is to submit the reference temperature for pressurized thermal shock (RTPTS) values for the R. E. Ginna reactor vessel to address the Pressurized Thermal Shock (PTS) Rule.Section I discusses the Rule and provides the methodology for calculating RT PTS.Section II presents the results of the neutron exposure evaluation assessing the effects that past and present core management strategies have had on neutron fluence levels in the reactor vessel.Section III provides the reactor vessel beltline region material properties.Section IV provides the RT PTS calculations from 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 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 from the United States Nuclear Regulatory Commission (USNRC). PTS events have been shown from operating experience to be transients that 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 affecting the integrity of the vessel.
The Rule establishes the following requirements for all domestic, operating PWRs:
3897e:ld/120685 1
l Establishes the RTPTS (measure of f racture resistance) Screening Criterion for the reactor vessel beltline region l
270*F for plates, forgings, axial welds 300"F for circumferential weld materials
- 6 Months From Date of Rule: All plants must submit their present RT PTS values (per the prescribed methodology) and projected RT PTS values at ,
the expiration date of the operating license. The date that this submittal must be received by the NRC for plants tdth operating licenses is January 23, 1986.
- 9 Months From Date of Rule: Plants projected to exceed the PTS Screening Criterion shall submit an analysis and a schedule for inflementation of such flux reduction programs as are reasonably practicable to avoid reaching the Screening Criterion. The data for this submittal must be received by the NRC for plants with operating licenses by April 23, 1986.
- Requires plant-specific , PTS Safety Analyses before a plant is within 3 years of reaching the Screening Criterion, including analyses of alternatives to minimize the PTS concern.
- Requires NRC approval for operation beyond the Screening Criterion.
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 part of 10CFR Part 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" (RTNDT). For purposes of the Rule, RT NDT IS '
now defined 25 "the . reference temperature for pressurized thermal shock" (RTPTS) and calculated as prescribed by 10 CFR 50.61(b) of the Rule. Each USNRC licensed PWR must submit a projection of RT PTS values from the time of the submittal to the license expiration date. This assessment must be 3897e:ld/120585 2
submitted within 6 months after the effective date of the Rule, on January 23, 1986, with updat'es whenever changes in core loadings, surveillance measure-ments, or other information indicate a significant change in 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 va ues for W R. E. Mnna Nuclear Mant.
PTS 4
1.2 THE CALCULATION OF RT PTS In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determin),g plant-specific values of RT PTS at a given time.
The prescribed equations in the PTS rule for calculating RT PTS a're actually one of several ways to calculate RT For the purpose of comparison with NDT.
the Screening Criterion, the velue of RT PTS f r the reactor vessel must be calculated for each weld and plate, or forging in the beltline region as given below. For each material, RT is the lower of the results given by PTS Equations 1 and 2.
Equation 1:
RTPTS = 1 + M + (-10 + 470(Cu) + 350(Cu)(NO] f Equation 2:
0 RT PTS = 1 + M + 283 f .194 where I = the initial reference transition temperature of the unirradiated material measured as defined in the ASME Code, NS-2331. If a measured value is not available, the following generic mean values must be used: 0*F for welds made with Linde 80 flux, and -56*F for welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes.
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M = the margin to be added to cover uncertainties in the values of initial RTNOT, copper 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.
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.
Note, since the chemistry values given in equations 1 and 2 are best estimate mean values, and the margin, M, causes the RT PTS values to be upper bound
. predictions, the mean material chemistry values are to be used, when available, so as not to compound conservatism.
1 3897e:ld/120585 4
SECTION II NEUTRON EXPOSURE EVALUATION The Westinghouse Adjoint Flux Program provides a cost ef fective tool to assess the effects that past and present core management strategies have had on neutron fluence levels in the reactor pressure vessel, along with several other considerations. This section presents the results from the application of the adjoint flux program to the R. E. Ginna reactor vessel for the Rochester Gas and Electric Company. This plant has recently operated using non-design basis core management.
11.1 METH00 0F ANALYSIS A plan view of the R. E. Ginna reactor geometry at the core midplane is shown in Figure 11.1-1. Since the reactor exhibits 1/8th core symmetry only a 0*-45* sector is depicted. Six irradiation capsules attached to the thermal shield are included in the design to constitute the reactor vessel surveillance program. Two capsules are located )BO' symmetrically at azimuthal positions of 13', 23', and 33' f rom the reactor core cardinal axes as shown in Figure 11.1-1.
In performing the fast neutron exposure evaluations for the reactor geometry ,
shown in Figure 11.1-1, two sets of transport calculations were carried out.
The first, a single computation in the conventional forward mode, was utilized to provide baseline data derived f rom a design basis core power distribution against which cycle by cycle plant specific calculations can be compared. The second set of calculations consisted of a series of adjoint analyses relating the response of interest at several selected locations within the reactor geometry to the power distributions in the reactor core. These adjoint importance functions when combined with cycle specific core power distribu-tions yield the plant specific exposure data for each operating fuel cycle.
The forward transport calculation was carried out in R,0 geometry using the 00T discrete ordinates code (2) and the SAILOR cross-section library (3). The 3897e:Id/120585 5
1 Figure II.1-1 PRESSURE VESSEL SURVEILLANCE CAPSULE 0*
13' (CAPSULES V, R) 23* (CAPSULES T, P) 1 eem < ]N SHIELD d,/ ,# '
33 ,C PSuleS S.~,
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4 c., .
mm,,I.,/f,j /// z7,,
? 't?, i ai ,: o,<,, 4$'
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l REACTOR COPE I /,i
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R. E. GINNA REACTOR GEOMETRY 6
l SAILOR library is a 47 group ENDF/8-IV based data set produced specifically for light water reactor applications. Anisotropic scattering is treated with aP expansion of the cross-sections.
3 The design basis core power distribution utilized in the forward analysis was derived f rom statistical studies of long-term operation of Westinghouse 2-loop plants. Inherent in the development of this design basis core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal
+2a level for a large number of fuel cycles, the use of this design basis distribution is expected to yield somewhat conservative results.
The adjoint analyses were also carried out using the P3 cross-section approximation f rom the SAILOR library. Adjoint source locations were chosen at the center of each of the surveillance capsules as well as at positions l along the inner diameter of the pressure vessel. Again, these calculations t
were run in R,e geometry to provide power distribution importance functions i for the exposure parameters of interest. Having the adjoint importance functions and appropriate core power distributions, the response of interest is calculated as:
RR,0 " IR Ie 1 (R,e) F (R,0) R dR de where:
R R,e
= Response of interest (+ (E > 1.0 MeV), dPa, etc.) at radius R and azimuthal angle e.
1 (R,e) = Adjoint importance function at radius R and azimuthal angle e.
F (R,0) = Full power fission density at radius R and azimuthal angle 9.
3897e:1d/122085 7
It should be noted that as written in the above equation, the importance function I (R e) represents an integral over the fission distribution so that the response of interest can be related directly to the spatial 4
distribution of fission density within the reactor core.
, The plant specific core power distributions used for Cycles lA through 13 were derived from burnup data supplied by Rochester Gas and Electric [9). The Cycle 14 power distribution was taken from the nuclear design data given in WCAP-10505 [8). As with the power distribution data the full power incremental irradiation times for Cycles l A through 13 were obtained f rom the cycle specific burnup data. The irradiation ' time for Cycle 14 was computed f rom the design burnup of 8800 MWO/MTU specified in WCAP-10505. The specific power distribution data used in the analysis is provided in Appendix A of this report. The data listed in Appendix A represent cycle averaged relative assembly powers. Therefore, the adjoint results are in terms of fuel cycle averaged neutron flux which when multiplied by the fuel cycle length yields ;
4 the incremental fast neutron fluence.
The projection of reactor vessel fast neutron fluence into the future to the expiraticn date of the operating license requires that a number of key assumptions be made. The present operating license for R. E. Ginna expires April 25, 2006 (forty years af ter the construction permit was issued). Based on current burnup, and projections assuming an 80 percent capacity factor, this expiration date corresponds to 27.36 Equivalent Full Power Years (EFPY) of operation. This report includes fluence projections to both the operating i license expiration date and 32 EFPY, representing 40 years of operation at a capacity factor of 80 percent.
The transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the Oak Ridge National
- the Westinghouse power reactor surveillance capsule data base (4]. The benchmerking 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%. When plant specific power distributions are used with ,
1 l
l 3897e:ld/121885 8 l 1
l
the adjoint importance functions, the benchmarking studies show that fluence l
predictions are within i 15% of measured values at surveillance capsule locations.
11.2 FAST NEUTRON FLUENCE RESULTS l
Calculated fast neutron (E >1.0 Mev) exposure results for R. E. Ginna are presented in Tables 11.2-1 through 11.2-7 and in Figures 11.2-1 through 11.2-4. Data is presented at several azimuthal locations on the inner radius of the pressure vessel as well as at the center of each surveillance capsule, j
In Tables II.2-1 through II.2-4 plant specific maximum neutron flux and fluence levels at 0*, 14.5*, 30*, and 44.5* on the pressure vessel inner radius are listed for the first 14 completed fuel cycles of R. E. Ginna, as well as for present (10.88 EFPY), End of License (27.36 EFPY), and 32 EFPY irradiation times. The incremental irradiation time listed is from the end of the previous cycle to the end of the currently listed cycle; thus, the duration of Cycle 13 was 2.32 x 10IEFPS, and f rom the e'id of Cycle 14 to 10.88 EFPY was an irradiation increment of 1.44 x 10 EFPS. Also presented are the design basis fluence levels predicted using the generic 2-loop core power distribution at the nominal + 2a level. Similar data for the center of surveillance capsules located at 13*, 23*, and 33' are given in Tables 11.2-5 through 11.2-7, respectively.
1 l
l l
3897e:1d/121885 9 L.
TA8LE 11.2-1 FAST NEUTRON (E > 1.0 MeV1 EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE Incremental 2
Irradiation Cycle Avg. Cumulative Fluence (n/cm 1 Time. Flux Plant Design Cvele No. (EFPS) (n/cm -sec) SDeCific Basis ( } _
10 I 18 1A 2.09 x 10 3.88 x 10 8.10 x 10 1.01 x 10 ,
I 10 18 18 18 2.38 x 10 4.16 x 10 1.80 x 10 2.15 x 10 2 0.71 x 10 I 4.22 x 10 10 2.10 x 10 18 2.50 x 10 18 10 18 18 3 2.88 x 10 3.62 x 10 3.14 x 10 3.88 x 10 4 2.17 x 10 7 3.60 x 10 10 3.93 x 10 18 4.93 x 10 18 10 18 18 5 1.84 x 10 4.63 x 10 4.78 t 10 5.82 x 10 6 2.42 x 10 I 3.93 x 10 10 5.73 x 10 18 6.98 x 10 18 10 18 18 7 2.35 x 10 3.74 x 10 6.61 x 10 8.12 x 10 8 2.20 x 10 7
4.40 x 10 10 7.58 x 10 18 9.18 x 10 18 10 18 I9 9 2.62 x 10 4.13 x 10 8.66 x 10 1.04 x 10 10 2.51 x 10 I
3.97 x 10 10 9.65 x 10 18 1.16 x 10 I9 11 1.83 x 10 3.86 x 10 1.04 x 10 1.25 x 10 "
12 2.30 x 10 I 4.53 x 10 1.14 x 10" 1.37 x 10" 13 2.32 x 10 3.11 x 10 1.21 x 10 1.48 x 10 "
14 2.27 x 10 7 3.05 x 10 1.28 x 10 D 1.59 x 10 D Current (10.88 EFPY) 1.44 x 10 3.08 x 10 1.32 x 10" 1.65 x 10 5.20 x 10 8 10 2.93 x 10 D EOL (27.36 EFPY) 3.08 x 10 4.16 x 10 10 (32.0 EFPY) 1.47 x 10 3.08 x 10 3.38 x 10 " 4.87 x 10 10 2 a) +,99 = 4.82 x 10 n/cm -sec. I 3897e:1d/121885 10
TA8LE 11.2-2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 14.5* AZIMUTHAL ANGLE Incremental 2
Irradiation Cycle Avg. Cumulative Fluence (n/cm ),
Time Flux Plant Design 2
Cycle No. (EFPS1 (n/cm ,itc). Specific Basis (a) _
1A 2.09 x 10 7 2.34 x 10 10 4.89 x 10 II 6.11 x l'0II 10 18 18 2.38 x 10 2.48 x 10 1.08xlb18 1.31 x 10 2 0.71 x 10 7
2.57 x 10 10 1.26 x 10 18 1.52 x 10 18 10 18 18
, 3 2.88 x 10 2.22 x 10 1.90 x 10 2.37 x 10 10 18 18 i 4 2.17 x 10 2.38 x 10 2.42 x 10 -
3.01 x 10 10 18 18 5 1.84 x 10 2.81 x 10 2.94 x 10 3.55 x 10 10 18 18 l 6 2.42 x 10 2.47 x 10 3.53 x 10 4.27 x 10 10 18 18 7 2.35 x 10 2.48 x 10 4.12 x 10 4.96 x 10
~
2.67 x 10 10 I8 I8 8
2.20 x 10 7 4.70 x 10 5.61 x 10 10 18 18 9 2.62 x 10 2.59 x 10 5.38 x 10 6.38 x 10 10 2.51 x 10 7 2.46 x 10 10 6.00 x 10 18 7.12 x 10 18 I 10 18 18 11 1.83 x 10 2.39 x 10 6.44 x 10 7.66 x 10 12 2.30 x 10 7 2.61 x 10 10 7.04 x 10 18 8.34 x 10 18 0 18 18 13 2.32 x 10 1.93 x 10 7.49 x 10 9.03 x 10 10 7,94 x 10 18 9.70 x 10 18 14 2.27 x 10 7 1.97 x 10 0 18 I9 Current (10.88 EFPY) 1.44 x 10 1.95 x 10 8.22 x 10 1.01 x 10 1.95 x 10 10 1.83 x 10 I9 2.55 x 10 I9 8
EOL (27.36 EFPY) 5.20 x 10 I9 2.98 x 10 I9 6 10 (32.0 EFPY) 1.47 x 10 1.95 x 10 2.12 x 10 10 n/cm 2 ,3,c, a) +,yg = 2.95 x 10 3897e:1d/121885 11 l - . ._ _, . ___._ . - .
6 TA8LE 11.2-3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE Incremental 2
Irradiation Cycle Avg. Cumulative Fluence (n/cm 1 Time Flux Plant Design Cvele No. (EFPS) (n/cm M Specific Basis ("} _
1A 2.09 x 10 7 -
1.50 x 10 10 3.14 x 10 U 4.12 x 10" 10 18 2.38 x 10 1.53 x 10 6.78 x 10" 8.81 x 10" 10 18 2 0.71 x 10 1.75 x 10 8.02 x 10" 1.02 x 10 10 18 18 3 2.88 x 10 1.46 x 10 1.22 x 10 1.59 x 10 4 2.17 x 10 7 1.80 x 10 10 1.61 x 10 18 2.02 x 10 18 10 18 18 5 1.84 x 10 1.71 x 10 1.93 x 10 2.38 x 10 6 2.42 x 10 7 1.62 x 10 10 2.32 x 10 18 2.85 x 10 18 7 2.35 x 10 7 1.71 x 10 10 2.72 x 10 18 3.32 x 10 18 8 2.20 x 10 I 1.76 x 10 10 3.11 x 10 18 3.75 x 10 18 10 18 18 9 2.62 x 10 1.84 x 10 3.59 x 10 4.27 x 10 10 18 10 2.51 x 10 I
1.78 x 10 4.04 x 10 4.76 x 10 18 10 18 18 11 1.83 x 10 1.48 x 10 4.31 x 10 5.12 x 10 18 12 2.30 x 10 7 1.43 x 10 10 4.63 x 10 5.58 x 10 18 10 18 18 13 2.32 x 10 1.27 x 10 4.92 x 10 6.03 x 10 14 2.27 x 10 1.54 x 10 10 5.27 x 10 18 6.48 x 10 18 Current (10.88 EFPY) 1.44 x 10 1.41 x 10 10 5.47 x 10 18 6.76 x 10 18 l
EOL (27.36 EFPY) 5.20 x 10 8 1.41 x 10 10 1.28 x 10 ' 1.70 x 10 I9 1.99 x 10I
- 8 10 (32.0 EFPY) 1.47 x 10 1.41 x 10 1.49 x 10 '
10 a) + yg = 1.97 x 10 n/cm2 -sec.
3897e:Id/121885 12 4
TA8LE 11.2-4 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 44.5* AZIMUTHAL ANGLE Incremental Irradiation Cycle Avg. Cumulative Fluence (n/cm ).
Time Flux Plant Design 2
Cycle No. (EFPS) (n/cm g Specific Basis (a) _
IA -
2.09 x 10 ~ 10 1.29 x 10 2.70 x 10 II 3.64 x 10 lI -
10 II lI 18 2.38 x 10 1.31 x 10 5.81 x 10 7.78 , 10 2 0.71 x 10 7 1.60 x 10 10 6.95 x 10 I7 9.01 ? 10 II 10 8 3 2.88 x 10 1.30 x 10 1.07 x 10 1.40 x 10 4 2.17 x 10 1.62 x 10 10 1.42 x 10 18 1.78 x 10 18 18 8 5 1.84 x 10 1.36 x 10 1.67 x 10 2.10 x 10 10 6 2.42 x 10 1.30 x 10 1.99 x 10 18 2.52 x 10 18 0 18 8 7 2.35 x 10 1.38 x 10 2.31 x 10 2.93 x 10 I 10 18 18 8 2.20 x 10 1.58 x 10 2.66 x 10 3.31 x 10 10 18 18 9 2.62 x 10 1.68 x 10 3.10 x 10 3.77 x 10 10 2.51 x 10 7 1.63 x 10 10 3.51 x 10 18 4.21 x 10 18 1.45 x 10 10 3.77 x 10 18 4.52 x 10 18 7
11 1.83 x 10 12 2.30 x 10 I 1.39 x 10 10 4.09 x 10 18 4.92 x 10 18 10 18 18 13 2.32 x 10 1.05 x 10 4.34 x 10 5.33 x 10 10 14 2.27 x 10 7 1.54 x 10 4.69 x 10 18 5.72 x 10 18 ,
1.44 x 10 10 18 I9 Current (10.88 EFPY) 1.30 x 10 4.88 x 10 5.97 x 10 1.30 x 10 10 1.16 x 10 I9 1.50 x 10 I9 8
EOL (27.36 EFPY) 5.20 x 10 1.47 x 10 8
1.30 x 10 10 1.35 x 10 I9 1.76 x 10 D
(32.0 EFPY) 10 a) +,,g = 1.74 x 10 n/cm 2,3,c, 3897e:1d/121885 13
TABLE 11.2-5 FAST NEUTRON (E > 1.0 MeV1 EXPOSURE AT THE 13' SURVEILLANCE CAPSULE CENTER Incremental Irradiation Cycle Avg. Cumulative Fluence (n/cm2 )
Time Flux Plant Capsule Design Cvele No. (EFPS) (n/cm ,1gi), Specific Measurement Basis (#} .,
18 1A 2,09 x 10 I 1.19 x 10' 2.49 x 10 18 3.07 x 10 1.27 x 10 U 7 18 18 18 2.38 x 10 5.51 x 10 6.57 x 10 1.30 x 10 U 18 18 2 0.71 x 10 7 6.44 x 10 18 6.53 x 10 7.61 x 10 D
3 2.88 x 10 7 1.13 x 10" 9.69 x 10 18 1.02 x 10 D 1.18 x 10 4 2.17 x 10 7 1.17 x 10 U 1.22 x 10I ' 1.50 x 10 I9 7 I9 I 5 1.84 x 10 1.43 x 10' 1.49 x 10 1.77 x 10 '
~
D 6 2.42 x 10 I 1.24 x 10' 1.79 x 10 2.13 x 10 7 2.35 x 10 7 1.23 x 10 U 2,08 x 10 D 2.48 x 10 8 2.20 x 10 7
1.36 x 10 U 2.37 x 10 D 2.80 x 10 D 9 2.62 x 10 7 1.29 x 10" 2.71 x 10 0 3.18 x 10 D 10 2.51 x 10 7 1.23 x 10 U 3.02 x 10 D 3.55 x 10 D 11 1.83 x 10 7
1.22 x 10 U 3.24 x 10 D 3.82 x 10 D 12 2.30 x 10 7 1.36 x 10" 3.56 x 10 D 4.16 x 10 D 13 . 2.32 x 10 7
9.43 x 10 10 3.78 x 10 " 4.50 x 10 "
14 2.27 x 10 7
9.49 x 10 10 4.00 x 10 D 4.83 x 10 D Current (10.88 EFPY) 1.44 x 10 9.46 x 10 10 4.14 x 10" 5.05 x 10" EOL (27.36 EFPY) 5.20 x 10 8
9.46 x 10 10 9.06 x 10 D 1.27 x 10 20 (32.0 EFPY) 1.47 x 10 8
9.46 x 10 10 1.04 x 10 20 1.48 x 10 20 Il 2 a) +,yg = 1.47 x 10 n/cm ,,,c, i
3897e:1d/121885 14
TA8LE 11.2-6 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 23' SURVEILLANCE CAPSULE CENTER Incremental Irradiation Cycle Avg. Cumulative Fluence (n/cm2 )
Time Flux Plant Capsule Design Cycle No. (EFPS) (n/cm2-igg.), Specific Measurement 8 asis (a)_
1A' .2.09 x 10 I
6.86 x 10 10 1.43 x 10 18 1.86 x 10 18 18 18 18 2.38 x 10 7.14 x 10 10 3.13 x 10 3.98 x 10 2 0.71 x 10 I 7.80 x 10 10 3.69 x 10 18 4.61 x 10 18 18 18
. 3 2.88 x 10 6.61 x 10 10 5.59 x 10 7.18 x 10 4 2.17 x 10 7 7.82 x 10 10 7.29 x 10 18 9.11 x 10 18 I 10 18 l
5 1.84 x 10 8.14 x 10 8.79 x 10 1.08 x 10 6 2.42 x 10 7 7.50 x 10 10 1.06 x 10 D 1.29 x 10 D t
7 2.35 x 10 7 7.82 x 10 10 1.24 x 10 , 1.50 x 10 D 1.42 x 10 I9 1.70 x 10 D 10 r
8 2.20 x 10 7 7.95 x 10 l 9 2.62 x 10 8.07 x 10 10 1.63 x 10 1.78 x 10" 1.93 x 10 "
10 2.51 x 10 7 7.74 x 10 10 1.82 x 10 2.15 x 10 2.32 x 10 D 10 11 1.83 x 10 7.18 x 10 1.96 x 10" .
10 12 2.30 x 10 6.85 x 10 2.11 x 10 2.52 x 10 10 13 2.32 x 10 7 5.95 x 10 2.25 x 10 " 2.73 x 10 14 2.27 x 10 6.55 x 10 10 2.40 x 10 " 2.93 x 10" 0
Current (10.88 EFPY) 1.44 x 10 6.25 x 10 2.49 x 10 3.06 x 10 "
8 10 EOL (27.36 EFPY) 5.20 x 10 6.25 x 10 5.74 x 10 " 7.69 x 10" 8 10 (32.0 EFPY) 1.47 x 10 6.25 x 10 6.66 x 10" 9.00 x 10 10 2 a) Oavg - 8.91 x 10 n/cm ,3,,,
3897e:1d/121885 15
TABLE 11.2-7 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 33*
SURVEILLANCE CAPSULE CENTER Incremental 2
Irradiation Cycle Avg. Cumulative Fluence (n/cm 1 Time Flux Plant Design Cycle No. (EFPS) (n/cm -sec) .SDecific Basis (a)_
1A 2.09 x 10 7 6.06 x 10 10 1.27 x 10 18 1.68 x 10 18 18 2.38 x 10 7
6.14 x 10 10 2.73 x 10 18 3.60 x 10 18 2 0.71 x 10 I 7.24 x 10 10 3.24 x 10 18 4.18 x 10 I8 3 2.88 x 10 7
5.95 x 10 10 4.96 x 10 18 6.50 x 10 18 4 2.17 x 10 7
7.40 x 10 10 6.56 x 10 18 8.25 x 10 18 18 I8 5 1.84 x 10 7
6.75 x 10 10 7.80 x 10 '9.73 x 10 6 2.42 x 10 7 6.45 x 10 10 9.37 x 10 18 1.17 x 10 7 2.35 x 10 7 6.86 x 10 10 1.10 x 10 D 1.36 x 10 D 8 2.20 x 10 7 7.22 x 10 10 1.26 x 10 D 1.53 x 10 D I D 1.75 x 10 9 2.62 x 10 7.61 x 10 1.46 x 10 7
1.64 x 10 D 1.95 x 10 0
10 2.51 x 10 7.39 x 10 1.75 x'10 D I 0 11 1.83 x 10 6.03 x 10 2.10 x 10 12 2.30 x 10 7
5.78 x 10 1.88 x 10 D 2.28 x 10 "
7 D 2.47 x 10 "
13 2.32 x 10 5.00 x 10 2.00 x 10 14 2.27 x 10 7 6.38 x IO N 2.14 x 10 D 2.65 x IO D Current (10.88 EFPY) 1.44 x 10 I 5.69 x 10 0 2.22 x 10 D 2.77 x 10 0 EOL (27.36 EFPY) 5.20 x 10 8 5.69 x 10 5.18 x 10 6.96 x 10 D (32.0 EFPY) 1.47 x 10 8 5.69 x 10 10 6.01 x 10 0 8.14 x 10 D 0 n/cm 2 ,3,c, a) g yg = 8.06 x 10 3897e:Id/121885 16
16096.1 1020 g.
s - p#
, s*
,/
- 10 19 "a : -
o ,
g W .
O ~
s l -
- 3 g . _
z O -
E 10 18 -
1 z -
1 . -
s -
l ! ! i l l 10 37 O 5 10 15 20 25 30 35 10.88 27.36 32.0 1
LICENSE EXPIRATION OPERATING TIME (EFPY)
FIGURE II.2-1. MAXIMUM FAST NEUTRON (E>1.0 MeV) FLUENCE AT THE BELTLINE WELD LOCATION AS A FUNCTION OF FULL POWER OPERATING TIME -
R.E. GINNA 17
1 G096.2 l O.20 _
r -
3
^ l N 3 -
\
E Me '
32.0 5FPY w
z 27.36 EFPY W lO l9 (LICENSE 3
u.
, [ EXPIRATION) g i _
o . _
lO.88 EFPY s _
m l z 4 -
t _
I 1
iO l8 l O 10 20 30 40 50 AZIMUTHAL ANGLE (deg.)
FIGURE 11.2-2. MAXIMUM CURRENT AND PROJECTED EOL FAST NEUTRON (E>1.0 MeV) FLUENCE AT THE PRESSURE VESSEL INNER RA010S AS A FUNCTION OF AZIMUTHAL Al GLE - R.E. GINNA 18
1.0 N r
- g. _
%T d
g %T
- ~
E v E
%T 0.1 __
. l l l l l l l- l 1 0 2 4 6 8 10 12 14 16 18 20 DEPTH INTO THE PRESSURE VESSEL (cm)
FIGURE 11.2-3. RELATIVE RADIAL VARIATION OF FAST NEUTR7N (E>1.0 MeV) FLUX AND FLUENCE WITHIN THE PRESSURE VESSEL - R.E. GINNA 19
l 1.0 _
a
!e 0.1 g . _
g . _
g . _
g , _
p 5 . _
E 0.01 _
0.001 l l I I I I L
- 300 - 200 - 100 0 100 200 300
, DISTANCE FROM CORE MIDPLANE (cm)
FIGURE 11.2-4. RELATIVE AXIAL VARIATION OF FAST NEUTRON ,
(E=1.0 MeV) FLUX AND FLUENCE WITHIN THE '
PRESSURE VESSEL WALL- R.E. GINNA i
I 20
In addition to the calculated data given for the surveillance capsule locations, measured fluence data from previously withdrawn surveillance capsules are also presented for comparison with analytical results. Capsules were removed from the 13* location at the end of cycles 2 and 3, and from the 23' location at the end of cycle 9.
Several observations regarding the data presented in Tables 11.2~-1 through II.2-7 are worthy of note. These observations may be summarized as follows:
- 1. Calculated plant specific fast neutron (E > 1.0 MeV) fluence levels at the surveillance capsule center are in excellent agreement with measured data. The maximum difference between the plant specific calculations and the measurements is less than 8.5%. Differences of this magnitude are well within the uncertainty of the experimental results.
- 2. Low leakage fuel management introduced following cycle 12 has reduced the peak flux on the pressure vessel by about 24%.
- 3. The maximum neutron flux incident on the pressure vessel (O' azimuthal position) during the fuel cycles using out-in fuel management (cycles 1 A-12) was, on the average, approximately 16% less than predictions based on the design basis core power distributions.
A graphical presentation of tLe plant specific fast neutron fluence at the peak beltline weld location on the pressure vessel is shown in Figure 11.2-1 as a function of full power operating time. These data are presented for the O' location on the circumferential weld (see Section 111.1). The solid portion of the fluence curve is based directly on the plant specific
!' evaluations presented in this report. The dashed portion of the curve, however, involves a projection into the future. Since R. E. Ginna is committed to a consistent form of low leakage fuel management, the average i neutron flux at the key locations over the low leakage fuel cycles was used for all temporal projections. In particular, the neutron flux average over cycles 13 and 14 was used to project all fluence levels beyond the end of Cycle 14.
3897e:Id/120685 21
The fluence projection in Figure 11.2-1 has been carried out to 32 effective full power years. However, since RT data corresponding to the license PTS expiration date must be supplied to the NRC in response to the Pressurized Thermal Shock Rule (10CFR50.61(b)(1)), the fluence corresponding to the license expiration date is indicated in Figure 11.2-1. An 80% capacity factor was assumed for future operation (i.e. beyond 10.88 EFPY).
It should be noted that implementation of a more severe low leakage pattern would act to reduce the projections of fluence at key locations. On the other hand, relaxation of the current low leakage patterns or a return to out-in fuel management would increase those projections. In any event the RT PTS assessment must be updated per 10CFR50.61(b)(1) whenever, among other things, l changes in core loadings significantly impact the fluence and RT PTS j proj ections.
In Figure 11.2-2, the azimuthal variati.on of maximum fast neutron (E > 1.0 MeV) fluence at the inner radius of the pressure vessel is presented as a function of azimuthal angle. Data are presented for both current and projected end-of-life conditions. In Figure 11.2-3, the relative radial variation of fast neutron flux and fluence within the pressure vessel wall is presented. Similar data showing the relative axial variation of f ast neutron flux and fluence over the beltline region of the pressure vessel is shown in Figure 11.2-4. A three-dimensional description of the fast neutron exposure of the pressure vessel wall can be constructed using the data given in Figure 11.2-2 through 11.2-4 along with the relation l
+(R, e,Z) = 4(e) F(R) G(Z) where: $ (R,0,Z) = Fast neutron fluence at location R, e, Z within the pressure vessel wall i
+ (e) = Fast neutron fluence at azimuthal location e on the pressure vessel inner radius f rom Figure 11.2-2 3897e:ld/120685 22
F (R) = Relative fast neutron flux at depth R into the pressure vessel from Figure 11.2-3 G (Z) = Relative fast neutron flux at axial position Z from Figure 11.2-4 Analysis has shown that the radial and axial variations within the vessel wall are relatively insensitive to the implementation of low leakage fuel management schemes. Thus, the above relationship provides a vehicle for a reasonable evaluation of fluer.ce gradients within the vessel wall.
'f 3897e:1d/120685 23 s
I I
SECTION III MATERIAL PROPERTIES
's For the RT PTS calculation, the best estimate copper and nickel chemical composition of the reactor vessel beltline material is necessary. The material properties for the R. E. Ginna beltline region will be presented in this section.
III.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 III.1-1 identifies the location of all beltline region materials for the R. E. Ginna reactor vessel.
III.2 DEFINITION OF PLANT-SPECIFIC MATERIAL PROPERTIES The pertinent chemical and mechanical properties of the beltline region plate i and weld materials of the R. E. Ginna reactor vessel are given in Table III.2-1. Although phosphorus is no longer used in the calculation of RT PTS with respect to the PTS rule [1], it is given for reference since it is currently used in the Regulatory Guide 1.99 trend curve [6].
Material property values for the shell plates, which have been docketed with the NRC in Reference 11, were derived from vessel fabrication test certificate results. The property data for the welds have also been docketed with the NRC in Reference ll, however, the weld properties cannot be used in the same direct manner as the properties for the plates.
3897e:1d/121985 24
FIGURE III.1 1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE R.E. GINNA NUCLEAR PLANT REACTOR VESSEL CIRCUMFERENTIAL SEAMS VERTICAL SEAMS 270*
2 I WR-43 r- 9 9
10,.0" -
k
_ 55 CORE ~
j ,
180*
I
- o.
CORE y
5 L-- d 144" j Forging
- 1255255 1
$ 90'
~
l Cg U_
! 14.8" g 4 : WR-19 270*
> N 7
- Forging j 15' "y 125P666 5
r E - CORE _
b 0*
I d I 180' ,
39.6" i '
j k J 90' I
25 i
i
TABLE 111.2-1 l R. E. GINNA REACTOR VESSEL BELTINE l
REGION MATERIAL PROPERTIES Cu Ni P I(a)
(wt %) (wt %) (wt %) (*F)
Intermediate Shell Forging, 125S255 Docketed Value .07 .69 .01 20 Lower Shell Forging,125P666 Docketed Value .05 .69 .012 40 Circumferential Weld - Nozzle to Intermediate Shell, WR-43, Wire Heat 71249:
WOG Data Base Pean (b) .26 .60 .019 0 Circumferential Weld - Intermediate to Lower Shell, WR-19, Wire Heat 61782:
WOG Data Base Mean I) .25 .55 .012 0 Docketed Value [11] .20 .39 .012 0 NRC Value [7] .25 .56 -
0 (a) The docketed initial RTPTS valves are estimated according to Branch Position MTEB 5-2 [5] and all other weld initial RTPTS values are generic mean values defined by the PTS rule [1].
(b) See Appendix B.
3897e:1d/121985 26
Fast neutron irradiation-induced changes in the tension, fracture, and impact properties of reactor vessel materials are largely dependent on chemical composition, particularly in the copper concentration. The variability in irradiation-induced property changes, which exists in general, is compounded by the variability of copper concentration within the weldments. .
l To address the variation in chemistry, Babcock & Wilcox (B&W) performed a reactor vessel beltline weld chemistry study of eight B&W vessels, including the R. E. Ginna plant, and reported the results in BAW-1799 [10] for the Westinghouse Owners Group (WOG). The scope of the work included collecting existing sources of chemistry data, performing extensive chemical analysis on the available archive reactor vessel weldments, and developing predictive methods with the aid of statistical analyses to establish the chemistry of the reactor vessel beltline weldments in question.
In addition to the B&W report BAW-1799, the WOG Reactor Vessel Beltline Region Weld Metal Data Base was used. The WOG data base, which was developed in 1984 and is continually being updated, contains information from weld qualification records, surveillance capsule reports, the B&W report BAW-1799, the Materials Properties Council (MPC) data base, and the NRC Mender MATSURV data base.
For each of the welds in the R. E. Ginna reactor vessel beltline region, a material data search was performed using the WOG data base. Searches were performed for materials having the identical weld wire heat number as those in the R. E. Ginna vessel, but any combination of wire and flux was allowed. For all of the data found for a particular wire, the copper, nickel, phosphorous and silicon values were averaged and the standard deviations were calculated.
Although phosphorous and silicon are not needed for the PTS Rule, they are provided for the sake of completeness. The information obtained from the data base searches is found in Appendix B.
When the results of the material data base searches were evaluated, it was found that a large scatter existed in the measured as-deposited copper values for the data obtained for weld wire 71249, which is associated with the R. E.
Ginna nozzle to intermediate shell circumferential weld. The chemical 3897e:ld/121985 27
composition values for heat 71249 have already been addressed via evaluations of the reactor vessel materials by Florida Power and Light for Turkey Point Units 3 and 4 (see reference 12).
The WOG Data Base mean values are used for calculation of weld RT PTS values. These values are similar to the values found in the aforementioned sources (references 7,11 and 12).
e 3897e:ld/121985 28
SECTION IV DETERMINATION OF RT VALUES FOR BELTLINE REGION MATERIALS PTS Using the methodology prescribed in Section I.2, the results of the fast neutron exposure provided in Section II, and the material properties discussed in Section III, the RT * "'* * * """ '"" " "
PTS RESULTS Using the prescribed PTS Rule methodology, RT PTS values were generated for all beltline region materials of the R. E. Ginna reactor vessel as a function of se.eral 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 limiting, circumferential weld and shell plate of the R. E. Ginna vessel in terms of RT PTS e sus 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 Screeninc. Criterion for PTS (i.e., RT va ues can PTS PTS be readily projected foi any options under consideration, provided fluence is known), and o to show the current (10.88 EFPY), end-of-license (27.36 EFPY) and end-of-life (32 EFPY) RT PTS values using actual and projected fluence.
- The EFPY Sin be determined using Figure 11.2-1.
3897e:ld/120685 29
-- a
l lll 1lll l 1ll tl
~
K 5m 1 1 1 1 2 2 2 3 3 3 2 5 7 0 2 5 7 2 5 7 0 2 5
- 5 0 5 0 5 0 5 5 0 5 0 5 0 1
D R= 0 - - - - - - ~ - - - - - -
e T 1 sP C 8
N R
- N iT u C R gS r 0 n r C 0 Ve R R C B an pT K .R a lt y pT EE s u SN $ y T.
ieL ho s E s si f 9 ez S DGI F Ue
( lz C S 1 le R C BN l s 7 l E R AN u i(
5 W E E SA e n1 ) et M E E ng 0 N MN c
lo I EU e P8 d N I TC l I G N l8 '
n G AL V a t _ AV LE a nE e V A l tF r _
,. L A &R u P i m U L e SY s p) e U WP d OL e F GA ca in L
U I
(
i a
t 3_
_ P E -
DT F N
fd 2 e _ L C A I i E 0 A I TR G c N I 9 T R AT U C ) E C P R aA n I A
U M
b E d= .E N F k
S I I D E ECU V PE R rn N I L E MR 1
3 0
od j -
E 1 O N EV -
U 0 l N T AE 1 eo T G I NS cf 1 I A t
eL di Pe f
R 9 O
N S
(
1 1
- (
T U
D l
H L
W E
L D
WP EE LR D
P l
6 2 A MT a( ) 2 L AS n3 / 7 T t2 W ER C E RU SE 0 L IL p F W D AE eP 2 LI S L cY on SN i) wt E f ee PT i
rnn RH c 9 OO Se PD hd E
( ei R[
FL 1 la T1 9
oo 2 lt I]
E rw ge 4 e ( S
) W ir et 2 9 n gS lo 7 5
d )
h 9 e l
l I
I I
1
' 0 2
0
'l l ll
Table IV.1-1 provides a summary of the RT values for all beltline region PTS materials for the lifetime of interest.
IV.2 DISCUSSION OF RESULTS As shown in Figure IV.1-1, the welds are the governing locations relative to PTS. All the RT PTS values remain below the NRC screening values for PTS using the projected fluence values through end-of-life (32 EFPY).
3897e:ld/120685 31
l f
TABLE IV.1-1 RT VALUES FOR R. E. GINNA PTS i
Present End-of-License End-of-Life Location Vessel Material (10.88 EFPY) (27.36 EFPY) (32 EFPY) i Nozzle to Intermediate 175 204 209 Shell Circumferential Weld WR-43 2 Intermediate to Lower 227 267 275 Shell Circumferential Weld WR-19 3 Intermediate Shell 111 121 123 Forgi'ng 125S255 4 Lower Shell Forging 116 122 124 125P666 l
I l
3897e:1d/120685 32
SECTION V CONCLUSIONS AND RECOMMENDATIONS Calculations have been completed in order to submit RT values for the PTS R. E. Ginna 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
- 1"
RTPTS Detailed fast neutron exposure evaluations using plant specific cycle by cycle core power distributions and state-of-the-art neutron transport methodology have been completed for the R. E. Ginna pressure vessel. Explicit calculaticns were performed for the first fifteen operating cycles of the unit (i.e. lA thru 14). Projection of the fast neutron exposure beyond the end of Cycle 14 was based on continued implementation of low leakage fuel management, utilizing the flux average over Cycles 13 and 14.
In regard to the low leakage fuel management already in place at the Ginna plant, the plant specific evaluations have demonstrated that for the low leakage case the average fast neutron flux at the 0* azimuthal position has been reduced by about 24% relative to that existing prior to implementation of low leakage. In particular, the following data applies at the O' location.
e (n/cm2-sec)
Out-In Pattern 4.05 x 10 10 Low Leakage Pattern 3.08 x 10 10 This location represents the maximum fast neutron flux incident on the reactor pressure vessel. At other locations on the vessel, as well as at the surveillance capsules, the impact of low leakage will differ from the data presented above.
3897e:1d/120685 33
I It should be noted that significant deviations from the low leakage scheme already in place will affect the exposure projections beyond the current operating cycle. A move toward a more severe form of low leakage (lower relative power on the periphery) would tend to reduce the projection. On the other hand, a relaxation of the loading pattern toward higher relative power on the core periphery would increase the projections beyond those reported.
As each future fuel cycle evolves, the loading patterns should be evaluated to determine their potential impact on projections made in this report.
The fast neutron fluence values from the plant specific calculations have been j compared directly with measured fluence levels derived from neutron dosimetry contained in three surveillance capsules withdrawn f rom the Ginna plant. The ratio of calculated to measured fluence values ranges from 0.916 to 0.986 for the three capsule data aints. This excellent agreement between calculation and measurement supports the use of this analytical approach to perform a plant specific evaluation for the R. E. Ginna reactor.
Material properties for the R. E. Ginna 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 f rom the WOG materials data base. These values are similar to those found in other sources.
Using the prescribed PTS Rule methodology, RT PTS values were generated for all beltline materials of the R. E. Ginna 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 l fluence exposure through 32 EFPY. The most limiting value at end-of-license (27.36 EFPY) is 267'F at the intermediate to lower shell circumferential weld.
The results provided in this report should enable Rochester Gas and Electric Company to comply with the initial 6 months submittal requirements of the US )
NRC PTS Rule.
l l
3897e:ld/122085 34 l
l-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. 8enchmark Testing of Westinghouse Neutron Transport Analysis Methodology -
to be published.
, S. NUREG-0800 - U.S. NRC Standard Review Plan, 8 ranch Technical Position 5-2, Revision 1, July 1981.
- 6. " 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.
- 7. NRC Policy Issue " Pressurized Thermal Shock," SECY-82-465, November 23,
, 1982.
- 8. WCAP-10505, "The Nuclear Design and Core Management of the R. E. Ginna Nuclear Reactor Cycle 14", P. W. Robertson, Y. A. Chao, J. L. Cole, Westinghouse Electric Corporation Nuclear Fuels Division, March,1984.
- 9. Letter of March 2,1984, f rom John D. Cook of Rochester Gas and Electric .
Corporation to Stan Anderson of Westinghouse Electric Corporation.
[
, 3897e:1d/120985 35
.m O
- 10. B&W Owners Group Report, BAW-1799, "BAW 177-FA Reactor Vessel Beltline Weld Chemistry Study", July 1983.
- 11. Rochester Gas & Electric letter from L. D. White Jr. to A. Schwencer, Chief, Operating Reactors Branch #1, dated September 13, 1977. I
- 12. NRC Letter Docket Nos. 50-250 and 50-251, " Evaluation of Reactor Vessel Materials Data for Turkey Point Plants Units 3 and 4 Reactor Vessels", I from S. A. Varga to J. W. Williams, Jr., of Florida Power and Light Company, April 26, 1984.
3897e:1d/121985 36
l APPENDIX A POWER DISTRIBUTIONS Core power distributions used in the plant specific fast neutron exposure analysis of the R. E. Ginna pressure vessel were derived from burnup data supplied by Rochester Gas and Electric Company for Cycles l A through 13, and from Reference 8 for Cycle 14.
A schematic diagram of the core configuration applicable to the R. E. Ginna plant is shown in Figure A.1-1. Cycle averaged relative assembly powers for each operating fuel cycle of 'R. E. Ginna are listed in Table A.1-1.
On Figure A.1-1 and in Table A.1-1 an identification number is assigned to each fuel assembly location; and three regions consisting of subsets of fuel assemblies are defined. In performing the adjoint evaluations, the relative power in assemblies comprising Region 3 has been adjusted to account for known biases in the analytical or design prediction of power in the peripheral assemblies while the relative power in assemblies comprising Region 2 has been maintained at the cycle average value. Due to the extreme self-shielding of the reactor core, neutrons born in fuel assemblies comprising Region i do not contribute significantly to the neutron exposure either at the surveillance capsules or at the pressure vessel. Therefore, power distribution data for assemblies in Region 1 are not listed in Table A.1-1 (and, by the same reasoning, Assembly 12 of Region 2).
l In each of the adjoint evaluations, within assembly spatial gradients have been superimposed on the average assembly power levels. For the peripheral assemblies (Region 3), these spatial gradients also include adjustments to
- account for analytical deficiencies that tend to occur near the boundaries of the core region.
3897e:1d/120685 A-1
16096.3 O*
i l
I 45' I 2 /
6 7 3 4 12 8 9 10 5 l
13 14 15 Ii 16 17 18 REGION I: ASSEMBLIES 12-21 REGION 2: ASSEMBLIES 6-il ..
19 20 REGION 3: ASSEMBLIES l-5 21 ________________90 j l
l l
FIGURE A.1-1. R.E. GINNA PLANT CORE DESCRIPTION F03 POWER DISTRIBUTION MAPS A-Z i
TA8LE A.1-1 CORE POWER DISTRIBUTIONS USED IN THE PLANT SPECIFIC FLUENCE ANALYSIS OF R. E. GINNA Fuel Cycle Assemb1v 1A 18 2 3 4 5 6 7 8 9 10 11 12 13 14 1 0.74 0.81 0.81 0.66 0.64 0.88 0.73 0.63 0.84 0.78 0.77 0.69 0.90 0.66 0.61 2 0.60 0.63 0.65 0.57 0.56 0.72 0.62 0.62 0.69 0.65 0.61 0.62 0.70 0.40 0.39 3 . 0.76 0.88 0.90 0.74 1.02 1.05 0.95 1.00 0.95 0.97 0.94 0.94 0.98 0.98 1.00 4 0.57 0.56 0.68 0.55 0.68 0.64 0.62 0.67 0.67 0.71 0.70 0.43 0.39 0.39 0.47 5 0.54 0.61 0.77 0.61 0.76 0.52 0.46 0.52 0.76 0.80 0.79 0.81 0.79 0.40 0.86 6 0.95 1.02 1.07 0.94 1.10 1.14 0.92 0.91 1.05 0.98 1.19 0.95 1.15 0.88 0.90 7 1.11 1.18 1.15 1.08 1.10 1.26 1.11 1.15 1.13 1.07 0.89 1.14 0.97 1.00 1.21 8 1.06 1.08 1.01 1.11 1.14 1.06 1.09 1.13 1.12 1.19 1.13 1.67 1.31 1.04 1.33 9 1.14 1.05 1.01 1.08 1.19 1.19 1.21 1.23 1.01 1.08 1.04 1.11 1.05 1.13 0.99 10 0.97 0.87 1.07 0.87 1.13 1.00 1.01 1.01 1.07 1.18 1.06 1.17 1.13 1.01 1.21 11 1.18 0.99 1.11 1.22 1.11 0.95 1.16 0.99 1.05 1.05 1.06 1.00 1.10 0.89 1.21 3897e:1d/120685 A-3
APPENDIX B WELD CHEMISTRY Tables B-1 and B-2 provide the weld data output from the WOG Material Data Base. Given are the searches of all available data for the wire heat in the R. E. Ginna 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 PTS analysis.
Weld Chemistry Data Source and Plant:
BAW-1799 - Babcock & Wilcox Report Number B&W -
Babcock & Wilcox Cu -
Emission Spectrographic Analysis FLA -
Turkey Point 4 FPL -
Turkey Point 3 .
Ni -
Weight % of Nickel OC1 -
Oconee 1 P -
Weight % of Phosphorous RGE -
Robert Emmett Ginna SC -
Surveillance Capsule Si -
Weight % of Silicon .
WQ -
Weld Qualification 3897e:ld/120685 B-1
TABLE B-1 R. E. GINNA N0ZZEL TO INTERMEDIATE SHELL WELD CHEMISTRY FROM WOG MATERIALS DATABASE l
SELECT REPDRT ID WIRE WIRE FLUI FLUI WELDCHEM Cu Ni P Si PLANT DESCRIPTION HEAT TYPE TYPE LOT CATA SOURCE 0219 71249 MN-MD-Ni LINDE 80 8738 BW,WQ 0.190 0.660 0.021 0.450 00M INTER TO LOWER SHELL CRI N0lILE TO INTER SHELL 0223 71249 MN-MD-Ni LINDE B0 8445 BW WG 0.210 0.570 0.021 0.520 FLA INTER TO LDWER SHELL FPL INTER TO LOWER SHELL
. FPL SURVEILLANCE WELD RSE N0!!LE TO INTER SHELL WEP INTER TO LOWER SHELL 0241 71249 MN-MD-N! LINDE 80 8669 PW,WE 0.210 0.550 0.012 0.410 0273 71249 MN-MD-N! LINDE B0 8457 0.230 0.550 0.020 0.510 FLA SURVEILLANCE WELD 0296 71249 MN-MD-NI LINDE 80 8445 BW[WQ FP ,5C 0.310 0.570 0.011 0.660 FLA INTER TO LOWER SHELL FPL INTER TO LOWER SHELL FPL SURVEILLANCE WELD RSE N0llLE TO INTER SHELL WEP INTER TO LOWER SHELL 0297 71249 MN-MO-N! LINDE 80 8457 FLA BAW,SC 0.300 0.600 0.014 0.500 FLA SURVEILLANCE WELD 0454 71249 MN-MD-N! LINDE 80 B445 1799,ESA 0.160 0.550 0.019 0.540 FLA INTER TO LOWER SHELL FPL INTER TO LOWER SHELL FPL SURVEILLANCE WELD RSE N0!!LE 10 INTER SHELL WEP INTER TO LOWER SHELL 0455 71249 MN-MO-N! LINDE 80 8445 6AW-1799,ESA 0.150 0.540 0.018 0.550 FLA INTER TO LOWER SHELL FPL INTER TO LOWER SHELL FPL SURVEILLANCE WELD
. RSE N0!!LE TO INTER SHELL WEP INTER TO LDWER SHELL 0456 71249 MN-MD-N! LINDE 80 8445 BAW-1799,ESA 0.180 0.550 0.019 0.540 FLA INTER TO LOWER SHELL FPL INTER TO LOWER SHELL FPL SURVEILLANCE WELD RSE N0llLE TO INTER SHELL WEP INTER TO LOWER SHELL 0457 71249 MM-M0-N! LINDE 80 8445 BAW-1799,ESA 0.190 0.540 0.019 0.610 FLA INTER TO LOWER SHELL FPL INTER TO LOWER SHELL FPL SURVEILLANCE WELD RGE N0ZILE 10 INTER SHELL WEP INTER TO LDWER SHELL B-2
TABLE B-1 (CON'T)
BAW-1799,ESA 0.150 0.550 0.020 0.600 FLA lhTER TO LOWER SHELL 0458 71249 M-M0-NI LIhDE 80 8445 FPL INTER TO L0 DER SHELL FPL SURVEILLMCE ELD R6E N0llLE TO INTER SHELL EP INTER TO LOWER SHELL 8445 BAW-1799,ESA 0.170 0.540 0.019 3.620 FLA INTER TO L0ER SHELL 0459 71249 M-MO-NI LINDE 80 FPL INTER TO LOWER SHELL FPL SURVEILLMCE ELD R6E N0llLE TO INTER SHELL WEP INTER TO LOWER SHELL LINDE 80 8445 BAW-1799,ESA 0.200 0.540 0.020 0.630 FLA INTER TO LDER SELL 0460 71249 M-MO-N!
FPL INTER TO LOWER SHELL FPL SURVEILLANCE WELD R6E N0llLE TO INTER SHELL WEP INTER TO LOWER SHELL 8445 BAW-1799,ESA 0.200 0.540 0.019 0.630 FLA INTER TO LOWER SHELL 0461 71249 M-MO-N! LINDE 80 FPL INTER TO LOWER SHELL FPL SURVEILLANCE WELD RGE N0llLE TO INTER SHELL WEP INTER TO LOWER SHELL BAW-1799,ESA 0.230 0.520 0.017 0.620 FLA INTER TO LOE R SHELL 0462 71249 M-M0-Ni LINDE 80 8445 FPL INTER TO LOWER SHELL FPL SURVEILLANCE WELD R6E N0llLE TO INTER SHELL WEP liiTER TO LOWER SHELL 0463 71249 MN-MO-N! LINDE 80 8738 BAW-1799,ESA 0.310 0.640 0.021 0.550 COM INTER TO LOWER SHELL CR3 N0llLE TO INTER SHELL LINDE 80 8738 BAW-1799,ESA 0.270 0.640 0.021 0.550 CDM INTER TO LOWER SHELL 0464 71249 M-M0-WI CR3 N0ZILE TO INTER SHELL MN-MD-N! LINDE 80 8738 BAW-1799,ESA 0.270 0.640 0.020 0.560 COM INTER TO LOWER SIELL 0465 71249 '
CR3 N0llLE TO INTER SHELL LINDE 80 8738 BAW-1799,ESA 0.280 0.640 0.021 0.560 COM INTER TO LCER SHELL 0466 11249 MN-MO-WI CR3 N0llLE TO INTER SHELL 0467 71249 M-MD-kl LINDE 80 8738 BAW-1799,ESA 0.280 0.640 0.021 0.550 00M INTER TO LOWER SHELL CR3 N0llLE TO INTER SHELL 0468 71249 MN-MD-NI LINDE 80 8738 BAW-1799,ESA 0.310 0.640 0.020 0.560 COM INTER TO LOWER SHELL CR3 N0llLE TO INTER SHELL 0469 71249 M-50-NI LINDE 80 8738 BAW-1799,ESA 0.280 0.640 0.020 0.550 COM INTER TO LOWER SHELL CR3 N0llLE TO INTER SHELL 0470 71249 . MN-M0-NI LINDE 80 8738 BAW-1799,ESA 0.280 0.630 0.021 0.550 COM INTER TO LOWER SHELL CR3 NullLE TO INTER SHELL l
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APPENDIX C RT A 0F R. E. GINNA PTS REACTOR VESSEL BELTLINE REGION MATERIALS Tables C-1 through C-10 provide the RT values, as a function of both
, PTS constant fluence and constant EFPY (assuming the projected fluence values),
for all beltline region materials of the R. E. Ginna reactor vessel. The RT values are calculated in accordance with the PTS rule, which is PTS 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 111.2-1 of the main report.
Location Vessel Material 1 'Circumferential Weld - Nozzle to Intermediate Shell, WR-43 2 Circumferential Weld - Intermediate to Lower Shell, WR-19 3 Intermediate Shell Forging 1255255 i
l 4 Lower Shell Forging 125P666 .
3897e:1d/120685 C-1
i l
TABLE C-1 RTPTS VALUEE FOR THE R.E.GINNA REACTOR VESSEL BELTLINE REGION O FLUENCE =1E18 N/CM**2 NI ! P ! I ! VALUE ! TYPE ! FLUENCE ! RTPTS !
!ID ! PLANT! CU !
-- -----RGE -------------
O.260 0.600 --------------
0.019 -GENERIC 0 C.W. O.10E+19 O.10E+19-149 143 1
0.012 O GENERIC C.W.
2 RGE O.250 0.550 B.M. O.10E+19 89 0.690 0.010 20 ACTUAL 3 RGE O.070 B.M. O.10E+19 102 0.690 0.012 40 ACTUAL 4 RGE O.050 B.M.=
Base Metal (Forgings)
NOTES:
Longitudinal Weld l
L.W.=
l C.W.= Circumferential Weld Reference Temperatures are in F C-2 L --- - .. . _.
, TABLE C-2 RTPTS VALUES FOR THE R.E.GINNA REACTOR VESSEL BELTLINE REGION O FLUENCE =5E18 N/CM**2 4
!ID ! PLANT! CU ! NI ! P ! I ! VALUE ! TYPE ! FLUENCE ! RTPTS !
.)
1 RGE O.260 0.600 0.019 0 GENERIC C.W. 0.50E+19 197 2 RGE O.250 0.550 0.012 O GENERIC C.W. 0.50E+19 188 3 RGE O.070 0.690 0.010 20 ACTUAL B.M. 0.50E+19 101 4 RGE O.050 0.690 0.012 40 ACTUAL B.M. 0.50E+19 109 1
4 I
i t
l I
C-3
TABLE C-3 RTPTS VALUES FOR THE R.E.GINNA REACTOR VESSEL BELTLINE REGION O FLUENCE =1E19 N/CM**2
!ID ! PLANT! CU ! NI ! P ! I ! VALUE ! TYPE ! FLUENCE ! RTPTS !
20 1 RGE O.260 0.600 0.019 0 GENERIC C.W. O.10E+ 226 RGE O.250 0.550 0.012 O GENERIC C.W. O.10E+20 215 .
2 108 RGE O.070 0.690 0.010 20 ACTUAL B.M. O.10E+20 3* O.10E+20 114 4 RGE O.050 0.690 0.012 40 ACTUAL B.M.
e I.
t l
C-4
c O
TABLE C-4 RTPTS VALUES FOR THE R.E.GINNA REACTOR
-VESSEL BELTLINE REGION Gi! FLUENCE =2E19 N/CM**2 .
I CU NI ! P ! I ! VALUE ! TYPE ! FLUENCE ! RTPTS !
!ID ! PLANT! !
=
GENERIC C.W. O.20E+20 260 0.600 0.019 0 1 RGE O.260 GENERIC C.W. O.20E+20 247 RGE O.250 .O.550 0.012 .O 116 2
0.010 20 ACTUAL B.M. O.20E+20 3 RCE O.070 0.690 40 ACTUAL B.M. O.20E+20 119 4 RGE_ O.050 O.690 O.012 r
1 0
e
,3
- [
'i
. l P
. r C-5
'IABLE C-5 RTPTS VALUES FOR THE R.E.GINNA REACTOR VESSEL BELTLINE REGION @ FLUENCE =4E19 N/CM**2
!ID ! PLANT! CU ! NI ! P ! I ! VALUE ! TYPE ! FLUENCE ! RTPTS !
0 60 0 019 0 GENERIC C 1 0 40E+20 302 1 RGE O.260 O.250 0.550 0.012 O GENERIC C.W. O.40E+20 285 1 2 RGE 126 RGE O.070 O.690 O.010 20 ACTUAL B .' M . O.4OE+20 3 1.25 4 RGE O.050 O.690 C.012 40 ACTUAL B.M. O.4OE+20 C-6
\ ______ .
. . . - _ - -. - - . - .~.
TABLE C-6 t .
RTPTS YALUES FOR THE R.E.GINNA REACTOR VESSEL BELTLINE REGION O FLUENCE =6E19 N/CM**2 s ->
!ID ! PLANI! CU t- NI ! P ! I ! VALUE !. TYPE ! FLUENCE ! RTPTS ! --
1 RGE O.260 0.600 0.019 0 GENERIC C. W. . 0.60E+20 330 2 RGE O.250 0.550 0.012 O GENERIC C.W. 0.60E+20 311 3 RGE O.070 .O.690 O.010 20 ACTUAL B.M. O.60E+20 133*
4 RGE .O.050 O.690 O.012 40 ACTUAL B.M. O.60E+20 129 N No .
J F
C-7
TABLE C-7 RTPTS VALUES FOR THE R.E.GINNA REACTOR VESSEL BELTLINE REGION'O FLUENCE =7E19 N/CM**2
!ID '?LANT! CU ! NI ! P ! I ! VALUE ! TYPE ! FLUENCE ! RTPTS !
_________________=_ _ _ _ . _
1 RGE O.260 0.600 0.019 O GENERIC C.W. 0.70E+20 341 2 RGE O.250 0.550 0.012 O GENERIC C.W. 0.70E+20 322 3 RGE O.070 0.690 0.010 20 ACTUAL B.M. 0.70E+20 135 4 RGE O.050 0.690 0.012 40 ACTUAL B.M. 0.70E+20 131 C-8
a TABLE C-8 RTPTS VALUES FOR THE R.E.GINNA REACTOR . VESSEL BELTLINE REGION MATERIALS @ CURRENT LIFE (10.88 EFPY)-PLANT SPECIFIC FLUENCE VALUES
!ID ! PLANT! CU ! NI ! P ! I ! VALUE ! TYPE ! FLUENCE'! RTPTS !
-= = - - - - - - - - - - - _ .
O.26E+19 0.600 0.019 O GENERIC C.W. 175 1 RGE O.260 2 RGE O.250 0.550 0.012 O GENERIC C.W. O.13E+20 226 O.010 20 ACfUAL H.M. O.13E+20 111 3 RGE O.070 O.690 O.690 O 012 40 ACTUAL B.M. O.13E+20 115 4 RGE O.050 l
~
l
-~
r I
.s.
Y s
TABLE C-9 RTPTS VALUES FOR THE R.E.GINNA REACTOR VESSEL BELTLINE REGION MATERIALS GEND OF LICENSE (27.36 EFPY)-PROJECTED FLUENCE VALUES
!ID ! PLANT! CU ! NI ! P !- I ! VALUE ! TYPE ! FLUENCE ! RTPTS !
1 RGE O.260 0.600 0.019 O GENERIC C.W. O.60E+19 204 2 RGE O.250 O.550 O.012 O GENERIC C.W. O.29E+20 266 3 RGE O.070 O.690 0.010 20 ACTUAL D. M. O.29E+20 121 4 RGE O.050 O.690 O.012 40 ACTUAL B.M. O.29E+20 122 i
I 1
l C-10
TABLE C-10 RTPTS VALUES FOR THE R.E.GINNA REACTOR VESSEL BELTLINE REGION MATERIALS @ 32 EFPY-PROJECTED FLUENCE VALUES i
i
! P ! ! VALUE ! TYPE ! FLUENCE ! RTPTS !
l
!ID ! PLANT! CU ! NI I 0 GENERIC C.W. 0.67E+19 209 1 RGE O.260 0.600 0.019 O.250 0.550 0.012 0- GENERIC C.W. O.34E+20 276 2 RGE B.M. O.34E+20 123
.) RGE O.070 O.690 0.010 20 ACTUAL 4O ACTUAL B.M. O.34E+20 124 4 RGE O.050 0.690 O.012 i
i l
l l
l i
a C-11 1
4 v