ML20205J354
| ML20205J354 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 12/31/1985 |
| From: | Hirst C, Lau F, Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20205J329 | List: |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR WCAP-10998, WCAP-10998-R01, WCAP-10998-R1, NUDOCS 8601300161 | |
| Download: ML20205J354 (51) | |
Text
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WCAP-10998 Revision 1 WESTINGHOUSE PROPRIETARY CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION VIRGIL C. SUMMER. NUCLEAR STATION UNIT 1 REACTOR VESSEL FLUENCE AND RT EVALUATION PTS A. H. Fero C. C. Heinecke M. A. Weaver DECEMBER 1985 WORK PERFORMED FOR SOUTH CAROLINA ELECTRIC AND GAS dNDER SHOP ORDER NUMBER VCSP-1040 Approved: A F.'L. Lau, Manager Approved: -
bro T. A. Meydr, Manager Radiation and Systems Structural Materials and Analysis Reliability Technology Approved: dN C. W. Hirst, Manager RCS_ 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. Box 355 Pittsburgh, Pennsylvania .15230 8601300161 360123
PDR ADOCK 05000395 P PDR 3830e:ld/123085
TABLE OF CONTENTS Section Title Page 1
SUMMARY
OF RESULTS 1-1 2 INTRODUCTION 2-1 3 NEUTRON EXPOSURE EVALUATION 3-1 3-1. Method of Analysis 3-1 3-2. Fast Neutron Fluence Results '3-4 4 EFFECT ON REACTOR VESSEL INTEGRITY 4-1 4-1. Licensing Basis for. Pressurized Thermal Shock 4-1 4-2. Identification and Location of Beltline Region Materials 4-4
) 4-3. Definition of Plant Specific Material Properties 4-4 4-4. Status of Reactor Vessel Integrity in Terms of RT PTS 4-5 5 CONCLUSIONS 5-1 6 REFERENCES 6-1 7 -APPENDICES A. Core Power Distributions A-1 B. RT PTS Values for the V. C. Summer Reactor Vessel l Beltline Region Materials B-1 l
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i LIST OF TABLES Table Title Page 3-1 Fast (E > l.0 MeV) Neutron Exposure at the Reactor Vessel Inner Radius - Azimuthal Angle of 0* 3-7.
3-2 Fast (E > 1.0 MeV) Neutron Exposure at the Reactor Vessel Inner Radius - Azimuthal Angle of 12' 3-8 3-3 Fast (E > l.0 MeV) Neutron Exposure at the Reactor Vessel Inner Radius - Azimuthal Angle of 21* 3-9 3-4 Fast (E > l.0 MeV) Neutron Exposure at the Reactor Vessel Inner Radius - Azimuthal Angle of 30' 3-10 3-5 Fast (E > 1.0 MeV) Neutron Exposure at the Reactor Vessel Inner Radius - Azimuthal Angle of 45* 3-11 3-6 Fast (E > 1.0 MeV) Neutron Exposure at the 17' Surveillance Capsule Center 3-12 3-7 Fast (E > 1.0 MeV) Neutron Exposure at the 20' Surveillance Capsule Center 3-13 J 4-1 V. C. Summer Reactor Vessel Beltline Region Material.
Properties 4-6 4-2 V. C. Sunner Reactor Vessel Beltline RT PTS and Fluence Values .4-7
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LIST OF TABLES (Continued) ;
i l Table Title Page i
A-1 Core Power Distributions Used in the V. C. Sunumer Fluence Analysis A-3 B-1 RT Values for the V. C. Sunumer Reactor Vessel PTS Beltline Region Materials at Various Fluences B-2 I
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LIST OF FIGURES
. Figure Title Page 3-1 V. C. Summer Reactor Geometry 3-14 3-2 V. C. Sununer Reactor Geometry - 15' Neutron Pad 3-15 3-3 V. C. Summer Reactor Geometry - 26' Neutron Pad 3-16 i
3-4 Plan View of a Dual Reactor Vessel Surveillance Capsule 3-17 3-5 Maximum' Fast (E > 1.0 MeV) Neutron Fluence at the Beltline Material Locations as a Function of Full Power Opetating Time 3-18 3-6 Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Center of the Surveillance Capsules as a Function of Full Power Operating ~ Time 3-19 3-7 Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Reactor Vessel Inner Radius as a Function of Azimuthal Angle 3-20 3-8 Relative Radial Distribution of Fast (E > 1.0 MeV) !
Neutron Flux and Fluence Within the Reactor Vessel Wall 3-21 3-9 Relative Axial Variation of Fast (E > 1.0 MeV) Neutron Flux and Fluence Within the Reactor Vessel Wall 3-22 4-1 Identification and Location of Beltline Region Materials for the V. C. Summer Reactor Vessel 4-8 4-2 V. C. Summer RT Curves per PTS Rule Method 4-9 PTS A-1 V. C. Sunener Core Description for Power Distribution Table A-4 3830e:ld/010386 vil
Revisten 1 has been issued to correct a typographical error.
On pages 1-1, 3-5, and 5-1, "18 percent" was changed to "17 percent".
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SECTION 1
SUMMARY
OF RESULTS Westinghouse derived adjoint importance functions have been used to assess the effects that past, present, and projected fuel cycles have on the fast neutron exposure of the V. C. Summer reactor vessel.
Plant specific evaluations for fuel cycles using out-in fuel management have demonstrated that the maximum fast (E > 1.0 MeV) neutron flux incident on the reactor vessel was, on the average,17 percent less than predictions based on l design basis core power distributions. Evaluations for the in-place low leakage core loading strategy also demonstrated that the reactor vessel maximum fast (E > 1.0 Mev) neutron flux was further reduced, by approximately 35 percent, relative to that existing prior to the implementation of low leakage.
Excellent agreement was demonstrated between measured data f rom the first withdrawn surveillance capsule and the value calculated using the adjoint importance functions. The magnitude of the-neutron flux at the surveillance capsule locations and the. lead factors relating capsule exposure to maximum reactor vessel exposure have been affected by the in-place low leakage loading.
The ef fect of resulting neutron fluence level changes on reactor vessel integrity relative to pressurized thermal shock (PTS) has been evaluated in terms of reference nil-ductility transition temperature data (RTNDT). For the PTS rule, the reference temperature (RT as calculated by the method PTS described in the rule) is for comparison with the applicable screening criterion. The RT PTS values based upon actual plate and weld material ;
chemistry data remain well below the NRC screening values for PTS using actual and projected neutron fluence through both present and proposed operating license expiration dates (i.e., 26.5 and 35.1 effective full power years respectively).
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The results provided in this report provide South Carolina Electric and Gas Co.apany with the information needed to con 91y with the requirements of the NRC rule for PTS to calculate plant specific current and projected RT PTS values for the reactor vessel beltline materials.
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SECTION 2 INTRODUCTION This report presents the results of the application of Westinghouse derived adjoint importance functions to the calculation of the V. C. Summer reactor ves el fluence for South Carolina Electric and Gas Company (SCE&G). The use of adjoint importance functions provides a cost effective tool to assess the ef fects that past and present core management strategies have had on neutron fluence levels in the reactor vessel. The results provided in this report will enable South Carolina Electric and Gas Company to comply with the Nuclear
) Regulatory Commission (NRC) rule for Pressurized Thermal Shock (PTS) [1]. The rule requires reporting the current and end-of-life fast (E > 1.0 MeV) neutron fluence and RT va ues matedals in W beMu region of W reacW PTS vessel along with the basis used for the neutron exposure evaluation. The, plant specific fluence and RT PTS data given herein were generated to meet this NRC requirement.
Section 3 outlines the adjoint neutron transport methodology, discusses the fuel management strategies that have been used to dater, and presents the neutron fluence data'for both design basis and plant specific (actual) fuel cycles. Neutron fluence data are presented at five locations at the inner surface of the reactor vessel and at the centers of the surveillance capsules.
A discussion of the NRC PTS rule and the effect of resulting fluence level changes on reactor vessel integrity relative to PTS is presented in Section
- 4. Following a discussion of the beltline region material property data for the reactor vessel, reference pressurized thermal shock temperature (RTPTS) results are presented in accordance with the NRC PTS rule requirements.
Conclusions are given in Section 5, and the references for the report are given in Section 6.
Appendix A lists the core power distributions used in the neutron fluence analysis. Appendix B provides a tabulation of the input and results for the RT calculation for all beltline region materials for the V. C. Sunner PTS reactor vessel.
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l SECTION 3 NEUTRON EXPOSURE EVALUATION i
3-1. METHOD OF ANALYSIS A plan view of the V. C. Summer reactor. geometry at the core midplane is shown in Figure 3-1. In general, the reactor may be described with octant symmetry. Three octants, however, have elongated neutron pads with dual surveillance capsule holders attached. Figure 3-2 shows a zero-to-45-degree sectnr with a 15-degree neutron pad segment included. This is the geometry for which the reactor vessel neutron fluence calculations have been performed.
Figure 3-3 shows a zero-to-45-degree sector with a 26-degree neutron pad i segment and a dual surveillance capsale holder. This is the geometry for which surveillance capsule evaluations are performed. A plan view of the holder attached to the neutron pad is shown in Figure 3-4. The stainless I steel specimen containers are approximately 1-inch square and approximately 56 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 5 feet i 1
of the 12-foot-high reactor core. The overall length of the neutron pads is approximately 12 feet and they are also centered axially on the core midplane. i Two sets of neutron transport calculations were carried out in performing the fast neutron exposure evaluations for the reactor geometries shown in Figures I 3-2 and 3-3. The first, a single calculation in the conventional forward mode, was used to provide baseline data derived from a design basis core power distribution against which cycle by cycle'pla'tn ~specific calculations can be compared. The_second set consisted of a series of adjoint mode neutron transport calculations relating the response of interest at the centers of the surveillance capsules and at several reactor vessel azimuthal locations ta the
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power distributions in the reactor core. These adjoint importance functions, j when combined with cycle specific core power distributions, yield the plant J specific expcsure data for each operating fuel cycle. ]
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l The forward transport calculation was carried out in R,0 geometry using the )
00T two-dimensional discrete ordinates transport code [2] and the SAILOR -l cross-section library [3]. The SAILOR library is a 47 neutron energy group, i ENOF-B/IV based data set which was developed specifically for light water -l reactor applications. Anisotropic neutron scattering is treated with a P 3 expansion of the scattering cross-sections.
The design basis core power distribution used in the forward transport l calculation was derived from statistical studies of long-term operation of Westinghouse 3-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 core power distribution is expected to yield conservative results. The design basis core power distribution data used in the analysis is provided in Appendix A of this report. The data listed in Appendix A l
represents cycle averaged relative assembly powers.
The adjoint transport calculations were also carried out using the 47 neutron energy group, P cross-secdons kom Me SAMR Nam WoM socce 3
locations were chosen at the center of each of the surveillance capsules as well as at positions along the inner radius of the reactor vessel. These calculations were also run in R,0 geometry to provide core power distribution importance functions for the neutron exposure parameters of interest. Having the adjoint importance functions and appropriate core power distributions, the response of interest is calculated as l F F F I(R,0,E) F(R.e,E) dE de R dR l
RR ' A' " JR Je .) E.
where:
RR,' 0' Response of interest (+ (E > 1.0 PeV), dpa, etc.) at l
radius R' and azimuthal angle O'.
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l I(R,0,E) = -Adjoint importance function at radius R cid azimuthal angle O for neutron energy group E l F(R,0,E) = Full power fission neutron density at radius R and azimuthal j angle e for neutron energy group E .
The fission neutron density distributions used include the enrichment- and burnup-dependent effects of the fissioning of other actinides in addition to U-235.
The core power distributions for use in the V. C. Summer plant specific fluence evaluations were taken from the nuclear design reports for each operating fuel cycle to date. The specific core power distribution data used in the analysis are provided in Appendix A o 'f this report. The data listed in Appendix A are cycle averaged relative assembly powers. Therefore, the adjoint results are in terms of fuel cycle seraged neutron flux which when multiplied by the fuel cycle length yields the incremental fast neutron fluence.
The project-ion of reactor vessel fast neutron fluence into the future to the expiration date of the operating license requires that a number of key assumptions be made. The present operating license for V. C. Summer expires March 21, 2013 (forty years af ter the construction permit was issued). South Carolina Electric and Gas Company has applied for an extension of the operating license for V. C. Sumer to October 22, 2022 (forty years after initial criticality). This report includes fluence projections to both operating license expiration dates (present and proposed). In keeping with SCE&G's stated goal of achieving a lifetime capacity factor of 90 percent, this value was assumed in making the neutron fluence projections. SCE&G has also committed to low leakage loading patterns as a fuel management strategy. j Consistent with this commitnent, the neutron fluxes used for neutron fluence projection are a cycle-length-weighted average of neutron fluxes calculated for Cycle 2 and Cycle 3 which reflect low leakage loading pattern fuel management.
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l The Westinghouse neutron transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the ORNL PCA facility, the Venus PWR engineering mockup and the Westinghouse power reactor surveillance capsule data base [4]. The benchmarking studies show that the use of the SAILOR cross-sections and design basis core power distributions produces neutron fluxes that tend to be conservative, with calculations exceeding measurements by 10 to 25 percent. When plant specific core power distributions ~are used with the adjoint importance f unctions, the benchmarking studies show that neutron fluence predictions are distributed within i 15 percent of measured values at surveillance capsule locations.
3-2. FAST NEUTRON FLUENCE RESULTS Calculated f ast (E > l.0 MeV) neutron exposure results for V. C. Summer are presented in Tables 3-1 thrcogh 3-7 and in Figures 3-5 through 3-9. Data are presented at several azimuthal locations on the inner radius of the reactor vessel as well as at the center of each surveillance capsule. The fluence levels are based on a reactor therma'l power level of 2775 MW.
Tables 3-1 through 3-5 list plant specific maximum neutron flux levels at 0*,
12", 21*, 30*, and 45" on the reactor vessel inner radius for the first two operating cycles, and projected to the expiration date (present' and proposed) of the operating license. Plant specific beltline cumulative fluence levels for the two completed fuel cycles, and design basis cumulative fluence levels l based on a design basis 3-loop core power distribution (at the nominal &2o level) are presented for each completed fuel cycle. Similar data for the ,
i center of the surveillance capsules located at 17' and 20* are given in Tables 3-6 and 3-7, respectively. The measured f ast neutron fluence for surveillance capsule, U withdrawn at the end of Cycle 1, is also presented b Table 3-6 for comparison with analytical results (5].
Several observations regarding the data presented in Tables 3-1 through 3-7 are worthy of note. These observations are summarized as follows:
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- 1. The calculated plant specific fast (E > 1.0 MeV) neutron fluence at the center of surveillance capsule U is in excellent agreement with the measured data. The dif ference between the plant specific calculation and the measurement is approximately 2 percent. Differences of this magnitude are well within the uncertainty of the experhaental result.
- 2. The peak fast (E > 1.0 MeV) neutron flux incident on the reactor vessol (0* azimuthal position) during the fuel cycle using out-in fuel management (cycle 1) was, on the average, 17 percent less than predictions based on j the design basis core power distribution.
- 3. Low leakage fuel management introduced following cycle 1 has reduced the average peak fast (E > 1.0 MeV) neutron flux on the reactor vessel by about 35 percent relative to that existing prior to the implementation of low leakage.
The plant specific fast (E > 1.0 MeV) neutron fluence at key locations on the reactor vessel and at the center of the surveillance capsules is shown plotted in Figures 3-5 and 3-6 as a function of full power operating time. Reactor vessel data is presented for the 0* azimuthal location on the circumferential weld and for the beltline region on the longitudinal welds. Surveillance capsule data is presented for the 17* and 20' locations.
The solid portions of the fluence curves-in Figures 3-5 and 3-6 are based directly on the cycle 1 and 2 plant specific evaluations presented in this report. The dashed portions of these curves however involve a projection into the future. As described in Section 3-1, SCE&G has been committed to a consistent form of low leakage fuel management for V. C. Summer. As a result the average neutron flux at the key locations over the low leakage fuel cycles was used for all temporal projections. In particular, the time weighted average neutron flux for cycles 2 and 3 was used to project future fluence levels for V.C. Summer. The data presented on those curves represent the best available information upon which to base the future withdrawal schedules for surveillance capseles remaining in the V. C. Summer reactor.
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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.
The azimuthal variation of maximum fast (E > 1.0 MeV) neutron fluence at the inner radius of the reactor vessel is presented in Figure 3-7 as a function of azimuthal angle. Data are presented for both current and projected expiration-of-operating-license conditions. In Figure 3-8 .the relative radial variation of fast neutron flux and fluence within the reactor vessel wall is presented. Similar data showing the relative axial variation of fast neutron flux and fluence over the beltline region of the reactor vessel is shown in_ Figure 3-9. A three-dimensional description of the fast neutron exposure of the reactor vessel wall can be constructed using the data given in Figures 3-6 through 3-8 along with the relation 4(R,0,Z) = 4(O) F(R) G(Z) where: 4 (R,e,Z) = Fast neutron fluence at location R, e, Z within the reactor vessel wall 4 (0) = Fast neutron fluence at azimuthal-location e on the reactor vessel inner radius from Figure 3-6 F (R) = Relative fast neutron flux at radius R into the reactor vessel f rom Figure a
'-7 G (Z) = Relative fast neutron flux at axial position I f rom Figure 3-8 s
Analysis has shown that the relative radial and axial variations within the reactor vessel wall are relatively insensitive to the implementation of low leakage fuel management strategies. Thus, the above relationship provides a vehicle for a reasonable evaluation of neutron fluence gradients within the reactor vessel wall.
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TABLE 3-1 FAST (E > 1.0 MeV) NEUTRON EXPOSURE AT THE REACTOR VESSEt. INNER RADIUS - AZIMUTHAL ANGLE OF 0*
8eltline Region Irradiation Cycle Average Cumulative Fluence (n/cm ) Cumulative Cycle Time Flux Plant Design Time No. (EFPS) (n/cm -sec) Specific Basis (a) (EFPY) 18 1 3.55 x 10 5.76 x 10 10 2.05 x 10 2.39 x 10 18 1.13-10 18 2(b) 2.11 x 10 4.58 x 10 3.01 x 10 3.81 x 10 18 j,79 3 ICI 3.70 x 10 7 4.08 x 10 10 _, ,, ,,
EOL 2013(d) 7.80 x 10 8
(4.26 x 1010)(f) 3.62 x 10 I9 5.63 x 10 I9 26.51 I9)
EOL 2022I ') 1.05 x 10 9 (4.26 x 1010)(f) 4.78 x 10 I9 7.46 x 10 I9 35.14 I9) 10 2 (a) Design Basis Flux = 6.73 x 10 n/cm -sec (b) End of Cycle 2 shutdown: October 5, 1985 (c) Fuel cycle projection (d) Present operating license expires: March 21,.2013 (e) Proposed operating license would expire: October 22, 2022 (f) Time weighted average neutron flux for Cycles 2 and 3, value in parentheses.
(g) A 90 percent capacity factor is assumed in projection to EOL.
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TABLE 3-2 FAST (E > 1.0 MeV) NEUTRON EXPOSURE AT THE REACTOR VESSEL INNER RADIUS - AZIMUTHAL ANGLE OF 12' Beltline Region Irradiation Cycle Average Cumulative Fluence (n/cm ) Cumulative Cycle Time Flux Plant Design. Time No. (EFPS) (n/cm -sec) SDecific Basis I (EFPY) 10 18 18 1 3.55 x 10 7
3.77 x 10 1.34 x 10 1.60 x 10 1.13 10 18 18 2(b) 2.11 x 10 3.27 x 10 2.03 x 10 2.55 x 10 ),79 10 3
IC) 3.70 x 10 3.00 x 10 ,_. __ ,_
7.80 x 10 8
(3.10 x 1010)(f) 2.62 x 10" 3.77 x 10 26.51 I9I EOL 2013(d) 9 5.00 x 10 35.14 I9I EOL 2022I ') 1.05 x 16 (3.10 x 1010)(f) 3.46 x 10 (a) Design Basis Flux = 4.51 x 1010.n/cm -sec (b) End of Cycle 2 shutdom: October 5, 1985 (c) Fuel cycle projection (d) Present operating license expires: March 21, 2013 (e) Proposed operating license would expire: October 22, 2022 (f) Time weighted average neutron flux for Cycles 2 and 3, value in parentheses.
(g) A 90 percent capacity factor is assumed in projection to EOL.
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TABLE 3-3 FAST (E > 1.0 MeV) NEUTRON EXPOSURE AT THE REAC10R VESSEL INNER RADIUS - AZIMUTHAL ANGLE OF 21' 8eltline Region Irradiation Cycle Average Cumulative Fluence (n/cm ) Cumulative Cycle Time Flux Plant Design Time No. (EFPS) (n/cm -sec) Specific Basis (a) (EFPY) 1 3.55 x 10 7 3.18 x 10 10 1.13 x 10 18 1.36 x 10 18 1.13 II 10 18 18 2 2.11 x 10 I 3.07 x 10 1.78 x 10 2.17 x 10 ),79 3(' 3.70 x 10 2.77 x 10 -- -- --
I9 E0L 2013(d) 7.80 x 10 8 (2.88 x 1010)(f) 2.42 x 10 3.21 x 10 I9 26.51 I9)
I9 I9 EOL 2022I '} 1.05 x 10 (2.88 x 1010)(f) 3.21 x 10 4.26 x 10 35.14 I9) 10 2 (a) Design Basis. Flux = 3.84 x 10 n/cm -sec (b) End of Cycle 2 shutdown: October 5, 1985 (c) Fuel cycle projection (d) Present operating license expires: March 21, 2013 -
(e) Proposed operating license would expire: October 22, 2022 --
(f) Time weighted. average .ieutron flux for Cycles 2 and 3 value in parentheses.
.(g) A 90 percent capacity factor is assumed in projection to E0L.
TABLE 3-4 FAST (E >-1.0 MeV) NEUTRON EXPOSURE AT THE REACTOR VESSEL INNER RADIUS - AZIMUTHAL ANGLE OF 30' 8elt11ne Region Irradiation Cycle Average Cumulative Fluence (n/cm ) Cumulative Time . Flux Plant Design Time Cycle No. (EFPS) (n/cm -sec) Specific BasisI #I (EFPY)
I 8 1 3.55 x 10 2.49 x'10 8.84 x 10 1.05 x 10 1.13 10 18 18 2(b) 2.11 x 10 2.31 x 10 1.37 x 10 1.68 x 10 3,79 10 3
IC 3.70 x 10 2.11 x 10 I '
7.80 x 10 8
(2.18 x 10 0)(f) 1.84 x 10 . 2.48 x 10" 26.51 I9I EOL 2013 9 3.28 x 10 35.14 I9I E0L 2022I 'I 1.05 x 10 - (2.18 x 1010)(f) 2.43 x 10 10 (a) Design Basis Flux = 2.96 x 10 nfc ,2,3,c (b). End of. Cycle 2 shutdown: October 5, 1985 (c) Fuel cycle projection-(d) Present operating license expires: March 2.1, 2013 (e) Proposed operating license would expire: October 22, 2022 (f) Time weighted average neutron flux for Cycles 2 'and 3, value in parentheses.
(g) A 90 percent capacity factor is assumed in projection to EC'_.
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TABLE 3-5 FAST (E > 1.0 MeV) NEUTRON EXPOSURE AT THE REACTOR VESSEL INNER RADIUS - AZIMUTHAL ANGLE OF 45' 8eltline Region Irradiation Cycle Average Cumulative Fluence (n/cm ) Cumulative Cycle Time Flux Plant Design Time No. (EFPS) (n/cm -sec) Specific Basis (a) (EFPY) 10 1 3.55 x 10 1.77 x 10 6.30 x 10" 7.36 x 10 " 1.13 2
IDI 2.11 x 10 1.42 x 10 10 9.30 x 10" 1.17 x 10 18 j,79 IC) 10 3 3.70 x 10 1.42 x 10 .,_ ,, ,_
EOL 2013(d) 7.80 x 10 8
(1.42 x 1010)(f) 1.20 x 10 I9 1.73 x 10 I 26.51 I9)
EOL 2022I ') 1.05 x 10 9 (1.42 x 1010)(f) 1.59 x 10 I9 2.29 x.10 I9 35.14 I9) 10 (a) Design Basis Flux = 2.07 x 10 n/cm -sec (b) End of Cycle 2 shutdown: October 5, 1985 (c) Fuel cycle projection (d) Present operating license expires: March 21, 2013 (e) Proposed operating license would expire: October 22, 2022 (f) Time weighted average neutron flux for Cycles 2 and 3, value in parentheses.
(g) A 90 percent capacity factor is assumed in projection to E0L. .
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TABLE 3-6 FAST (E.> 1.0 MeV) NEUTRON EXPOSURE AT THE 17* SURVEILLANCE CAPSULE CENTER Beltline Region Irradiation Cycle Average Cumulative Fluence (n/cm )
Time Flux Plant Design Capsule "U" Cycle (EFPS) (n/ck -sec) Specific- Basis Data No.
I 18 18 1 3.55 x 10 1.76 x 10 6.25, x 10 .7.43 x 10 6.39 x' 10 18 I 2(b) 2.11 x 10 1.64 x 10 9.71 x 10 1.18 x 10 3
IC) 3.70 x 10 1.50 x 10' -- --
8 20 20 E0L 2013(d) 7.80 x 10 (1.55 x 10" )III 1.31 x 10 1.75'x 10 20 20 EOL 2022I ') 1.05 x 10 (1.55 x 10" )I I 1.73 x 10 2.32 x 10 2
(a) Design Basis Flux = 2.09 x 10' n/cm -sec (b) End of Cycle 2 shutdown: October 5, 1985 (c) Fuel cycle project'on (d) Present operating license expires: March 21, 2013 (e) Proposed operatinc license would expire: October 22, 2022' (f) Time weighted average neutron flux for Cycles 2 and 3, value in parentheses.
(g) A 90 percent capacity factor is assumed in projection to E0L.
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TABLE 3-7 FAST (E > i.0 MeV) NEUTRON EXPOSURE AT THE 20' SURVEILLANCE CAPSULE CENTER 8eltline Region Irradiation Cycle Average Cumulative Fluence (n/cm )
Cycle Time Flux Plant Design No. (EFPS) (n/cm -sec) SDecific Basis ( }
18 8 1 3.55 x 10 1.52 x l'0" 5.41 x 10 6.47 x 10 2(b) 2.11 x 10 1.48 x 10' 8.53 x 10 18 1.03 x 10 l9 3
IC) 3.70 x 10 1.33 x 10' -- --
8 20 0 E0L 2013(d) 7.80 x 10 (1.38 x 10' )I } 1.17 x 10 1.52 x 10 E0L 2022I ') 1.05 x 10 9 (1.38 x 10' )II) 1.54 x 10 20 2.02 x 10 20 (a) Design Basis Flux = 1.82 x 10" n/cm -sec (b) End of Cycle 2 shutdown: October 5, 1985 (c) Fuel cycle projection (d) Present operating. license expires: March 21, 2.013 (e) Proposed operating license would expire: October 22, 2022 (f) ' Time weighted average neutron flux for Cycles 2 and 3, value in parentheses.
(g) A 90 percent capacity factor is assumed in projection to E0L.
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Figure 3-1. V.C. Summer Reactor Geometry 3-14
16149 1 00 f4EUTRON PAD 1
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Figure 3-2. V.C. Summer Reactor Geometry - 150 Neutron Pad 3-15 l
16149-2 NEUTRON PAD 16.94 DEG. (CAPSULES U, V, X) 0 19.72 DEG. (CAPSULES W, Y, Z) l I
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SECTION 4 EFFECT ON REACTOR VESSEL INTEGRITY The effect of fast neutron fluence level changes, resulting from fuel manageraent strategies used for the V. C. Summcr reactor, on reactor vessel integrity relative to pressurized thermal shock (PTS) is presented in this section. This work is carried out in accordance with the NRC rule on PTS.
4 -1. LICENSING BASIS FOR PRESSURIZED THERMAL SHOCK The Pressurized Thermal Shock (PTS) Rule [1] was approved by the Nuclear Regulatory Commission (NRC) on June 20, 1985, and appeared in the Federal j 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 PTS rule is incorporated into the Code of Federal Regulations as 10 CFR 50.61.
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 9perated with a license from the Nuclear Regulatory Commission.
Operating experience has shown PTS events to be transients that result in a rapid and severe cooldown of 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 cause the propagation of flaws postulated to exist near the reactor vessel inner wall surface, thereby potentially affecting the integrity of the reactor vessel.
I The rule establishes the following requirements for all domestic PWRs for -i which an operating license has been issued: l l
l The rule establishes the RTPTS (measure of fracture resistance)
' screening criterion for the reactor vessel beltline region as 270*F for plates, forgings, and axial weld materials and as 300*F for circumferential weld materials.
I 3830e:1d/123085 4 -1
- All licensees must submit to the NRC their reactor vessel beltline
, material RT PTS values (calculated per the prescribed methodology for each material). Both current and projected RT PTS values at the.
expiration date of the operating license must be submitted. This submittal must be received by the NRC no later than January 23,1986 and must be updated whenever changes in core loadings, surveillance measurements or other information indicate a significant change in projer.ted values. The updated values must be reported to the NRC.
- For each PWR projected to exceed the PTS screening criterion, the licensee shall submit an analysis and schedule for implementation of such flux reduction programs as are reasonably practicable to avoid exceeding the screening criterion. This submittal must be received by the NRC by April 23, 1986.
- The rule requires that a plant-specific PTS safety analyses be submitted to the NRC at least 3 years before the PTS screening criterion is exceeded.
- The rule requires NRC approval for operation beyond the screening criterion.
Applicants for operating licenses are required to provide projected RTPTS values in the Final Safety Analysis Report. This requirement has been added to 10 CFR 50.34.
The NRC rule on PTS states that RT py3 values of ;70*F for plates, forgings, and axial weld materials, and 300*F for circumferential weld materials be used as screening criteria to determine the timing of plant specific evaluations of reactor vessel. integrity and of possible needed modifications to provide protection against PTS events. To this end, the NRC Staff has selected a conservative and uniform methou for determining plant-specific values of-RT py3 at a given time.
The equations given below are used to calculate RTPTS, the reference nil-ductility transition temperature for PTS, and are specified in 10 CFR 3830e:Id/123085 4-2
50.61(b)(2). For the purpose of comparison with the screening crit'erion, the value of RT PTS f r the reactor. vessel must be calculated as follows: The
. calculation must be made for each weld and plate, or forging, in the reactor vessel beltline. For each material, RT is the lower of the results given PTS by Equations 1 and 2.
l Equation 1: j RTPTS = I + r. + [-10 + 470(Cu) + 350(Cu)(N1)] f
- 1 l
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 Boiler and Pressure Vessel Code,Section III, paragraph NB-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.
M = the margin to be added. 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 N1 - the best estimate weight percent of copper and nickel in the
- material.
f = the maximum fast 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 reactor vessel'at the location where the material in question receives the highest fast neutron fluence for the period of service in question.
3830e:1d/123085 4-3
I Since the chemistry values given in equations 1 and 2 are best estimate mean values, and since the margin, M, causes the RT PTS values to be upper bound predictions, the use of mean material chemistry values is recommended, when available, so as not to compound conservatism. The basis for the Cu and Ni values used must be reported to the NRC. The relationship of the material on which any measurements were made to the actual material in the reactor vessel must be described.
4-2. IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS Figure 4-1 identifies the location of all beltline region materials for the V. C. Summer reactor vessel. The beltline. region is defined to be "the region of the reactor vessel (shell material including welds, heat af fected zones, j and plates or forgings) that directly surrounds the effective height of the l active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage.
^
4-3. DEFINITION OF PLANT-SPECIFIC MATERIAL PROPERTIES The pertinent chemical and mechanical properties of the beltline region plate and weld materials of the V. C. Summer reactor vessel are given in Table 4-1.
Also given in the table are weld material properties which were used in the 1982 Westinghouse Owner's Group (WOG) initial assessment of PTS [6). The latest nickel values are slightly lower than the WOG values which resulted from the averaging of two chemistry values from two weld qualifications with weld wire heat 4P4784. Although phosphorus is no longer used in the calculation of RTPTS, it is given for reference since it is used in the Regulatory Guide 1.99 Rev. I trend curve [7].
Material property values for the shell plates were derived from reactor vessel l fabrication test certificates. Weld material properties were derived from weld qualification records. All of the beltline region welds were manufactured using weld wire heat 4P4784 and flux Linde 124, flux lot 3930.
Since two Chicago Bridge & Iron weld qualification records were available for 3830e:ld/123085 4-4
l welds using this weld wire heat, these chemistry values were averaged to determine the best estimate Cu, Ni and P values. No additional data was found using the Westinghouse Owner's Group Reactor Vessel Materials Database [8].
4-4. STATUS OF REACTOR VESSEL INTEGRITY IN TERMS OF RT PTS Based on the methodology described in Section 4-1, RT values were PTS calculated for all bel.line region materials of the V. C. Summer reactor vessei as a function of fast neutron fluence through reactor vessel lifetime.
The tabulated results from the total evaluation are presented in Appendix 8 for all beltline region materials.
I Table 4-2 presents the current (end of Cycle 2) and projected to expiration of operating license (both present and proposed) values of.RT PTS for all of the beltline region materials. Figure 4-2 presents the RTp73 values for the most limiting base plate and weld material in terms of RT PTS as a function of fast (E > 1.0 MeV) neutron fluence. The curve shown for the intermediate ,
to lower shell circumferential weld in Figure 4-2 is also applicable to the I
intermediate and lower shell longitudinal welds, since the same copper and nickel values apply to these welds. The only difference is that the ,
longitudinal welds are exposed to a lower neutron fluence and would not have the same RT PTS values at a given time.
As shown in Figure 4-2, the intermediate shell plate is the governing location for the reacior vessel relative to PTS with a current RT PTS value of il8'F and with projected RT values of 155'F at the expiration of the present PTS operating license and 162*F at the expiration of the proposed operating l license. All of these values are well below the applicable NRC screening i criterion of 270*F for base metal. l l
f I I
J 3830e:Id/123085 4-5
Table 4-1 V. C. SUMMER REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Plate Trace Elements Code (Weicht Percent)
Reactor Vessel Beltline Material No,I1) Cu Ni P I(*F)(2)
Intermediate Shell Plate A9154-1 11-1 0.10 0.51 0.009 30 1
Intermediate Shell Plate A9153-2 11-2 0.09 0.45 0.006 -20 Lower Shell Plate C9923-1 10-2 0.08 0.41 0.005 10 Lower Shell Plate C9923-2 10-1 0.08 0.41 0.005 10 l Intermediate Shell Longitudinal Welds (3) 0.06 0.89 0.013 -44 1982 WOG Program Values (4) 0.06 0.91 0.013 -44 Intermediate to Lower Shell Girth Weld (3) 0.06 0.89 0.013 -44
, 1982 WOG Program values (4) 0.06 0.91 0.013 -44
, Lower Shell Longitudinal Welds (3) 0.06 0.89 0.013 -44
] 1982 WOG Program Values (4) 0.06 0.91 0.013 -44 I
i I
(I) Refer to Figure 4-1.
(2)lnitial reference transition temperature of the unirradiated material measured as defined in the ASME Boiler and Pressure Vessel Code, Section
! III Paragraph N8-2331.
(3) Chemistry values for welds averaged between two available Chicago Bridge and Iron weld qualifications with weld wire heat 4P4784.
(4)1982 WOG values are provided for reference only.
3830e:1d/123085 4-6
~ Table 4-2 V. C. SUMMER REACTOR VESSEL 8ELTLINE RTpys~AND FLUENCE VALUES RTp;S (dearees Ft Expiration Expiration Current - of Present of Ppoposed End of Cycle 2 Operating License Operating License Reactor Vessel Beltline Material (October 5.1985) (March 21. 2013) (October 22. 2022)
Intermediate Shell Plate A9154-1 118 155 162 l
l Intermediate Shell Plate A9153-2 62 94 99 i
Lower Shell Plate C9923-1 86 113 118 Lower Shell Plate C9923-2 86 113 118 Intermediate Shell Longitudinal Welds 23 42 46 Intermediate to Lower Shell Circumferential Weld 31 56 60 LowGr Shell Longitudinal Walds 23 42 46 Fast (E > 1.0 MeV) Neutron Fluence (n/cm2)_.
Expiratibe Expiration Current - of Present of Proposed End of Cycle 2 Operating License Operating License Reactor Vessel 8ein'iRMterial (October 5. 1985) (March 21. 2013) (October 22, 2022)
Intermediate Shell r!M*: 3.01 x 1018 3.62 x 1019 4.78 x 1019 Lower $h*!1 Plates 3.01 x 1018 3.62 x 1019 4.78 x 1019 Interteediate Shell Longitudinal Welds 9.30 x 1017 1.20 x 1019 1.59 x 1019 Intermediate to Lower Shell Circumferential Weld 3.01 x 1018 3.62 x 1019 4.78 x 1019 Lower Shell Longitudinal Weld 9.30 x 1017 1.20 x 1019 1.59 x 10 19 3830e:1d/123085 4 -7 l
16149-4 CIRCUMFERENTI AL WELDS LONGITUDINAL WELDS 00 11-2
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4-8
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A = CURRENT !l.8 EFPY) ATPTS VALUES e a PRESENT END-OF-LICENSE (?6.5 EFPY) RTPTS VALUES o PROPOSED END-OF-LICENSE (35.1 EFPY) RTPTS VALUES REFER TO FIGURE 3-5 FOR NEUTRON FLUENCE AS A FUNCTION OF FULL POWER OPERATING TIME (EFPY)
Figure 4-2. V. C. Summer RTPTS Curves Per PTS Ruta Method 49
SECTION 5 CONCLUSIONS Detailed fast neutron exposure evaluations using plant specific cycle by cycle core power distributions and state-of-the-art neutron transport methods have been completed for the V. C. Summer reactor vessel and surveillance capsules.
Explicit calculations were performed for the first three fuel cycles. The projection of the fast neutron exposure beyond cycle 2 was based on continued implementation of low leakage fuel managen, ant similar to that of cycles 2 and 3.
Plant specific evaluations have. demonstrated that during fuel cycles using out-in fuel management, the maximum fast (E > 1.0 MeV) neutron flux incident on the reactor vessel was, on the average,17 percent less than predictions l based on design basis core power distributions. With regard to the low leakage fuel management strategy in place at V. C. Summer, the plant specific evaluations have shown that the averaqe fast (E > 1.0 MeV) neutron flux at the 0* and 45' azimuthal positions was reduced by about 35 percent and 25 percent, respectively, relative to that which would be present without the implementation of low leakage.
Implementation of a more severe form of low leakage (lower relative power on the periphery) would tend to further reduce the neutron flux. On the other hand, a relaxation of the loading pattern toward higher relative power on the j core periphery would increase the neutron fluxes beyond those reported.
Implementation of low leakage loading patterns also affects the magnitude of the neutron flux at the surveillance capsule locations as well as the lead factors relating capsule exposure to maximum reactor vessel exposure.
Therefore', the assumptions of constant lead factors and surveillance capsule exposure rates over plant lifetime are not necessarily valid. Data depicting the actual surveillance capsule exposure as a function of full power operating i time has been presented in this report for V. C. Summer.
3830e:ld/Oll666 (Rev 1) 5-1
The fast neutron fluence values from the plant specific calculations have been compared directly with measured fluence levels derived from neutron dosimetry contained in surveillance capsule U withdrawn f rom V. C. Summer at the end of Cycle 1. The ratio of calculated to measured fluence was 0.98. This agreement between calculation and measurement supports the use of this analytical approach to perform plant specific neutron fluence evaluations for the V. C. Summer reactor.
The reactor vessel integrity evaluation was based upon actual base plate and ~
weld material chemistry data, and projected plant-specific neutron fluence values. The intermediate shell plate is the governing location for the reactor vessel relative to PTS with a current (end of cycle 2) RT PTS value of 118'F and with projected RT values f 155'F at the expiration of the PTS present operating license and 162*F at the expiration of the proposed operating license. All of these values are well below the applicable NRC screening criterion of 270*F for base metal.
3830e:1d/123085 5-2 1
SECTION 6 REFERENCES
- 1. " Analysis of Potential Pressurized Thermal Shock Events," US Nuclear Regulatory Comission Final Rule,10 CFR Part 50, 50.34, 50.61, Federal Register, Vol. 50, No. 141, pp. 29937-29945, July 23, 1985.
- 2. WANL-PR(LL)-034, Vol. 5, R. G. Soltesz, et al, " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation Vol. 5 - Two Dimensional, Discrete Ordinates Transport Technique," August 1970.
- 3. "0RNL RSIC Data Library Collection DLC-76," SAILOR Coupled 5,alf-Shielded, 47 Neutron, 20 Gama-Ray, 3P , Cross Section Library for Light Water Reactors."
- 4. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology (to be published).
- 5. WCAP-10814. " Analysis of Capsule U f rom the South Carolina Electric and Gas Company Virgil C. Sumer Reactor Vessel Radiation Surveillance Program," R. S. Boggs, A. H. Fero, and W. T. Kaiser, June 1985.
- 6. Letter WOG-82-290, " Calculation of Operating and NT0L Vessel RT NDT Values," R. A. Muench, December 31, 1982.
- 7. Regulatory Guide 1.99 - Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Comaission, Washington, April 1977.
- 8. Westinghouse Owners Group Reactor Vessel Materials Database, Rev. O, January 1985.
- 9. WCAP-9685, "The Nuclear Design of the Virgil C. Sumer Power Plant Unit 1," R. J. Fabean, et al, March 1980 3830e:ld/123085 6-1
i
- 10. WCAP-10663, Rev.1 "The Nuclear Design and Core Management of the Virgil C. Sumer Power Plant Cycle 2," R. M. Smith, March 1985.
- 11. WCAP-10874, " Nuclear Design Report Plant Operations Package for Virgil C.
Sumer Nuclear Power Plant Cycle 3," D. W. Chung, et al, August 1985.
3830e:1d/123085 6-2
APPENDIX A CORE POWER DISTRIBUTIONS Core power distributions used in the plant specific fast neutron exposure analysis of the V. C. Summer reactor vessel were derived f rom the following fuel cycle nuclear design reports:
Fuel Cycle Nuclear Design Report 1 WCAP-9685 2 WCAP-10663, Rev.1 3 WCAP-10874 A schematic diagram of the core configuration applicable to V. C. Summer is shown in Figure A-1. Cycle averaged relative assembly powers for each fuel cycle for V. C. Summer are listed in Table A-1 along with the design basis core power distribution.
On Figure A-1 and in Table A-1 an identification number is ' assigned to each fuel assembly location. Three regions consisting of subsets of fuel assemblies are defined. In performing the adjoint evaluations, the relative power in the fuel assemblies comprising Region 3 has been adjusted to account for known biases in the prediction of power in the peripheral fuel assemblies while the relative power in the fuel 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 the fuel assemblies comprising Region 1 do not contribute significantly to the neutron exposure of either the surveillance capsules or the reactor vessel. Therefore, core power distribution data for fuel assemblies. in Region 1 are not listed in Table A-1.
In each of the adjoint evaluations, within assembly spatial gradients have been superimposed on the average assembly power levels. For the peripheral 3830e:1d/112285 A-1
assemblies (Region 3), these spatial gradients also include adjustments to account f or analytical deficiencies that tend to occur near the boundaries of the core region.
3830e:1d/ll2285 A-2
TABLE A-1 CORE POWER DISTRIBUTIONS USED IN THE V. C. SUMMER FLUENCE ANALYSIS Plant Specific Cycle Averaged Design Basis Relative Assembly Power Fuel Relative Fuel Cycle Assembly Assemb1v Power 1 2 3 1 1.00 0.744 0.548 0.420 2 0.83 0.592 0.424 0.442 3 1.21 0.882 1.021 0.938 4 0.86 0.613 0.548 0.442 5 0.92 0.676 0.424 0.455 6 0.98 0.963 0.865 0.967 7 1.10 0.970 1.232 1.032 8 1.00 1.077 1.013 1.164 9 1.05 1.030 1.177 1.206 10 1.08 0.957 1.065 1.064 11 1.06 1.050 1.198 1.229 12- 0.95 0.883 0.875 0.869 Note: Refer to Figure A-1 for fuel assembly location.
3830e:1d/112285 A-3
( . _ __ ____
1 1 2 6 7 J 4 13 8 9 10 5 (
14 15 16 11 12 17 18 19 20 21 22 23 REGION ASSEMBLIES 1 13 26 24 25 2 6 12 3 1-5 26
/
Figure A-1. V.C. Summer Core Description for Power Distribution Table A-4
APPENDIX B l
l RT VALUES FOR V. C. SUMMER PTS I REACTOR VESSEL BELTLINE REGION MATERIALS l
Table B-1 provides the RT values as a function of fast neutron fluence, PTS for all beltline region materials of the V. C. Summer reactor vessel. Refer to Figure 3-5 for the relationship between plant specific fast neutron fluence and effective full power years of operation. The RT PTS values are calculated in accordance' with the PTS rule, as described in Section 4-1. The reactor vessel location numbers in Table B-1 correspond to the reactor vessel materials identified below and in Figure 4-1 and Table 4-1 of this report.
Location Reactor Vessel Beltline Material 1 Intermediate shell plate A9154-1 2 Intermediate shell plate A9153-2 3 Lower shell plates C9923-1,2 4 Intermediate and lower shell longitudinal welds 5 Intermediate to lower shell circumferential weld 3830e:1d/112285 B-1 ,
TABLE B-1 RT VALUES FOR THE V.~C. SUMMER REACTOR VESSEL PTS BELTLINE REGION MATERIALS AT VARIOUS FLUENCES Values of RT PTS Location Cu Ni P RTNDTI Value Type f=0.1 f=0.5 f=1.0 f=2.0 f=3.0 j 1 0.10 0.51 0.009 30 Actual B.M. 107 123 133 144 152 2 0.09 0.45 0.006 -20 Actual B.M. 53 67 74 84 91 3 0.08 0.41 0.005 10 Actual B.M. 79 90 97 105 111 4 0.06 0.89 0.013 -44 Actual L.W. 24 35 41 48 54 5 0.06 0.89 0.013 -44 Actual C.W. 24 35 41 48 54 NOTES: a) Trace elements are listed in weight percent.
b) Reference temperatures are in degrees F.
c) Value = " Actual" denotes that the listed values of initial RT NDT are actual rather than estimated values d) B.M. = base material (plate) e) L.W. = longitudinal weld f) C.W. = circumferential weld I9 2 g) f = the fast (E > 1.0 MeV) neutron fluence in units of 10 n/cm 3830e:1d/122685 B-2 I