ML20212M841
| ML20212M841 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 03/31/1981 |
| From: | Shaun Anderson WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20212M840 | List: |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8608270252 | |
| Download: ML20212M841 (23) | |
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ANALYSIS OF FAST NE(ITRON FLUX LEVELS AND END-0F-LIFE EXPOSURE FOR THE YANKEE R0WE REACTOR PRESSURE VESSEL l l S. L. Anderson l Westinghouse Electric Corporation i Nuclear Technology Division i i March 1981 l l 8600270252 860812 PDR ADOCK 05000029 [V~ [ P PDR fl l
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INTRODUCTION For all operating light water reactors, calculated neutron flux distributions within the pressure vessel wall are used, in conjunction with damage trend curves, to predict the degree of embrittle' ment of the reactor vessel steel over its service life. Embrittlement gradients are in turn used to determine pressure-temperature limita-tions for normal plant operation as well as to evaluate the effect of various heatup/cooldown transients on vessel conditions. Since the initial design of the Yankee Rowe reactor, significant improvements have been made in neutron transport methodology, in the application of that methodology to large PWR's, and in the cross-section data bases which are used as input to the computations. This report des-cribes a discrete ordinates Sn transport analysis using current method-ology which was performed for the Yankee Rowe reactor in order to pro-vide updated distributions of fast neutron (E > 1.0 Mev) fluence within the pressure vessel. s I i l
METHOD OF ANALYSIS A plan view of the Yankee Rowe reactor geometry at the core midplane is shown in Figure 1. Since the reactor exhibits 1/4 core symmetry only a 00 - 900 sector is depicted. The model shown in Figure 1 was developed from Westinghouse drawings 646J500, 549D155, 646J692, 540F839, and 540F857; and from Babcock and Wilcox drawing 45109E. In the analysis of the neutron environment within the geometry shown in Figure 1, predictions of the neutron flux magnitude ar.d energy spectrum were made with the D0T II) two dimensional discrete ordinates code. The computation was perfonned in the R,0 mode and was nonnalized to the time averaged power density at the reactor core midplane. The analysis employed 21 neutron energy groups, an 56 angular quadrature, and a P) expansion of the scattering cross-sections. The cross-sections were generated via the Westinghouse GAMBIT (2) code system with fine group to broad group processing carried out by the APPROPOS(3) and ANISN(41 codes. The final broad energy group structure used in the analysis is listed in Table 1. Reactor core power distributions for use in this analysis were supplied by Yankee Atomic Electric Company (5) . The supplied information included radial assembly power distributions for core cycles XII, XIII, XIV, and XV as well as mid-cycle pin power distributions for those assemblies located on the periphery of the core. To develop a power distribution for input to the DOT calculation the assemblywise power distributions were linearly averaged over the four core cycles. The assembly powers for each cycle as well as the four cycle average are listed in Table 2 and superimposed on Figure 1. To provide power density gradients in the peripheral fuel assemblies, the pin power distributions from cycle XV were employed. An axial peak to average ratio of 1.2 was then applied to the input power distributions.
. Although the Yankee Rowe reactor is truly 1/4 core symmetric, an examina-
' tion of Figure i shows that the deviation from 1/8 core symetry is quite small. Power density differences are on the order of 3% and geometric differences occur only locally in the areas of the core baffle and thermal shield support structures.
Based on these small differences, the decision was made to perform the transport calculations based on 1/8 core symmetry. This approach pennits the use of a much finer spatial mesh to describe the reactor internals which in turn produces better results. To insure conservatism in the 1/8 core mockup, the geometry in the 00 - 45 0 0 sector was employed with the power densities from the 45 - 90 sector. That is, the higher power densities were used in conjunction with the geo-metry having the least steel between the core and the pressure vessel. The net effect of this approximation should be no more than 3 - 5 percent. A schematic of the actual geometry and power distribution used in the analysis is shown in Figure 2. The geometric model used in the transport calculations is described in I Figure 3 and in Tables 3 and 4. Figure 3 shows the detailed mesh cell description of the geometry while the dimensions of the radial and azimuthal mesh boundaries are provided in Tables 3 and 4. respectively. In all, a 159 radial by 51 azimuthal mesh grid was employed to describe the problem geometry. m e
RESULTS OF ANALYSIS Results of the neutron transport analysis of the Yankee Rowe reactor are sunnarized in Figures 4 and 5 as well as in Table 5. In Figure 4, the calculated maximum fast neutron (E > 1.0 Mev) flux levels at the pressure vessel inner radius,1/4 thickness and 3/4 thickness loca-tions are presented as a function of azimuthal angle. As stated in the preceding section, these flux values are based on a reactor core power of 600 MWt and,an axial peak to average ratio of 1.2. In Figure 5, the predicted maximum and of life fast neutron (E > 1.0 Mev) fluence for the Yankee Rowe reactor is given as a function of radial position within the vessel wall. This data was based on an assumed operating lifetime of 24 effective full power years. That is, a design life of 30 calendar years with an assumed capacity factor of 80%. Again, an axial peaking factor of 1.2 has been applied to the calculation. In Table 5, the fast neutron (E > 1.0 Mev) fluence at the pressure vessel surface,1/4 thickness location, and 3/4 thickness location is summarized as a function of full power operating time.
ADEQUACY OF THE ANALYTICAL METHOD The accuracy of the calculations described in this report depends on the ability to define appropriate core power distributions, the adequacy of the cross-section libraries, and on the appropriateness of the transport model itself. Taken as a whole, these factors combine to yield an over-all uncertainty in the analytical results. The overall Westinghouse analytical methodology described in this report has been compared with measurements obtained a! a number of operating PWR's. These comparisons have been described in the open literature (6, 7)but will also be summarized here. Figures 6 and 7 depict a comparison of the calculated activity of four fast neutron detectors with measurements obtained from reactor vessel sur-veillance capsules removed from nine 2-loop PWR's. The calculations are in good agreement with the measured data. Since these monitors respond to different portions of the neutron energy spectrun, the agreement is an indication that both the energy spectra an- flux magnitudes are being determined well by the analytical method. Figures 8 and 9 also show comparisons of calculations with measurements of fast neutron flux. In this case, the data were obtained external to the reactor vessel in the air gap between the vessel and the biological shield. The data shown in Figure 8 were obtained at a 4-loop reactor; while the data in Figure 9 were obtained at a 3-loop reactor. Again, the agreement between calculation and measurement is excellent. Based on the comparisons presented in Figures 6 through 9, it is concluded that the Westinghouse design method is appropriate for the analysis of large PWR's and that the results presented in this report may be applied with confidence to the Yankee Rowe reactor. i ! r i
REFERENCES
- 1. Soltesz, R. G. et al, " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation - Volume 5 - Two-Dimensional Dis-crete Ordinates Transport Technique", WANL-PR-(LL)-034, August 1970.
- 2. Collier, E., et al, "Second Version of the GAMBIT Code", WANL-TME-1969, November 1969.
- 3. Soltesz, R. G. , et al, " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation - Volume 3, Cross-Section Generation and Data Processing Techniques", WANL-PR-(LL)-034, August 1970.
- 4. Soltesz, R. G. et al, ". Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation - Volume 4 - One-Dimensional Dis-crete Ordinates Transport Technique", WANL-PR-(LL)-034, August 1970.
- 5. Yankee Atomic Electric Company letter NED 80-506, R. J. Cacciapouti to S. L. Anderson, June 27, 1980.
- 6. McElroy, W. N. et al, " Development and Testing of Standardized Pro-cedures and Reference Data for LWR Surveillance", IAEA Specialists Meeting, Vienna, Austria, February 197^.
- 7. Anderson, S. L. and Capo, M. A. , " Measurements and Calculations of Neutron Streaming within a PWR Primary Shield Annulus", Trans. AM.
Nucl. Soc., 617 (1976). a
Table 1 Enerov Group Structure Used in the Transport Analysis of the Yankee Rowe Reactor Group LowerEneroy(Mev) 1 7.79* 2 6.07 3 4.72 4 3.68 5 2.87 6 2.23 4 7 1.74 8 1.35 9 1.05 10 0.821-11 0. 388 12 0.111 13 4.09 x 10-2 i 14 1. 50 x 10
-2 15 5.53 x 10-3 16 5.83 x 10-4 17 7.89 x 10-5 18 1.07 x 10-5 i 19 1.86 x 10-6 20 3.00 x 10 -7 21 0.0
- Upper energy of Group 1 is 10.0 Mev.
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- Table 2 Assembly Averaoe Power Distributions Avg. BOL; E0L Assembly Distributions bly ID CORE XII CORE XIII CORE XIV CORE XV AVG.
E1 .723 .789 .717 .756 .746 E2 1.011 1.105 1.004 1.123 1.061 E3 1.164 1.049 1.198 1.128 1.135 E4 1.168 1 .1 31 1.198 1.185 1.171 E5 1.088 1.046 1.127 1.104 1.091 D1 .563 .628 .562 .602 .589 D2 1.048 1.132 1.052 1.083 1.079 D3 1.197 1.164 1.193 1.167 1.180 D4 1.199 1.079 1.203 1.146 1.157 D5 1.166 1.137 1.195 1.161 1.165 C2 .722 .785 .721 .746 .744 C3 1.1 53 1.199 1.142 1.144 1.160 C4 1.226 1.170 1.175 1.189 1.190 C5 1.1 30 1.086 1.1 31 1.121 1.117 B3 .742 .786 .732 .747 .752 B4 1.123 1.138 1.101 1.086 1.112 B5 1. 216 1.115 1. 201 1.129 1.165 A4 .065 .6 31 .596 .603 .609 A5 .763 .791 .753 .757 .766
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Table 3 , Padial Mesh line Dimensions for Yankee Rowe Analysis Line Line Line Line No. Radius (cm) No. Radius (cm) No. Radius (cm) No. Radius (cm) 1 55.139 28 95.469 55 110.166 82 125.809 2 60.000 29 95.872 56 110.803 83 126.325 3 65.000 - 30 96.467 57 111.441 84 126.841 4 70.000 31 96.818 58 112.078 85 127.357 5 75.000 32 97.626 59 112.925 86 127.872 6 80.000 33 98.104 60 113.771 87 128.389 7 82.000 34 98.727 61 114.618 88 128.905 , 8 83.000 35 99.302 62 115.253 89 129.421 9 53.440 36 100.042 63 115.888 90 129.936 10 84.962 37 100.476 64 116.522 91 130.452 11 86.567 38 100.916 65 117.157 92 130.968 12 87.078 39 101.446 66 117.792 93 131.484 13 87.439 40 101.931 67 118.300 94 132.000 14 87.911 41 102.441 68 118.808 95 132.516 15 88.311 42 102.855 69 119.316 96 133.032 16 88.731 43 103.094 70 119.824 97 133.572 17 89.593 44 103.788 71 120.332 98 134.112 18 90.052 45 104.341 72 120.821 99 134.651 19 90.501 46 104.981 73 121.310 100 135.191 20 91.455 47 105.463 74 121.799 101 135.731 l 21 91.945 48 106.099 75 122.288 102 136.271 22 92.458 49 106.775 76 122.764 103 136.811 23 92.751 50 107.346 77 123.241 104 137.350
- 24 93.449 51 107.716 78 123.717 105 137.890 25 93.730 52 108.253 79 124.193 106 138.430 26 94.021 53 108.891 80 124.777 107 138.937 27 94.619 54 109.528 81 125.293 108 139.444 i
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Table 3 (continued) Line Line Line Line No. Radius (cm) No. Radius (cm) No. Radius (cm) No. Radius (crd 109 139.951 122 146.542 135 153.133 148 165.000 110 140.458 123 147.049 136 153.640 149 170.000 111 140.965 124 147.556 137 154.147 150 175.000 112 141.472 125 148.063 138 154.654 1 51 177.000 113 141.979 126 148.570 139 155.161 152 178.000 114 142.486 127 149.077 140 155.668 153 179.030 115 142.993 128 149.584 141 156.175 154 179.500 116 143.500 129 150.091 142 156.682 155 181.000 117 144.007 130 150.598 143 157.189 156 183.000 118 144.514 131 151.105 144 157.696 167 187.000 119 145.021 132 151.612 145 158.203 158 191.000 120 145.528 133 152.119 146 158.710 159 195.000 1 21 146.035 134 152.626 147 160.000 160 200.000 I
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Table 4 Azimuthal Mesh Line Dimensions for Yankee Rowe Analysis Line Line No. 0 (deo.) 0 (rev.) No. e (deo.) 0 (rev.) 1 0.00 0.000 27 23.00 0.06389
. 2 1.00 0.002778 28 24.00 0.06667 3 2.00 0.005556 29 25.00 0.06944 4 3.00 *0.008333 30 26.00 0.07222 5 3.85 0.01069 31 26.40 0.07333 6 4.00 0.01111 32 26.80 0.07444 7 5.00 0.01389 33 27.00 0.07500 8 6.00 0.01667 34 28.00 0.07778 9 7.00 0.01944 35 29.00 0.08056 10 8.00 0.02222 36 30.00 0.08333 11 9.00 0.02500 37 31 .00 0.08611 12 9.28 0.02578 38 32.00 0.08889 13 10.00 0.02778 39 33.00 0.09167 14 l'.00 0.03056 40 34.00 0.09444 15 12.00 0.03333 41 35.00 0.09722 16 13.00 0.03611 42 36.00 0.1000 17 14.00 0.03889 43 36.70 0.1019 18 15.00 0.04167 44 37.10 0.1031 19 16.00 0.04444 45 38.00 0.1056 20 17.00 0.04722 46 39.00 0.1083 21 18.00 0.05000 47 40.00 0.1111 22 19.00 0.05278 48 41.00 0.1139 23 20.00 0.05556 49 42.00 0.1167 24 21.00 0.05833 50 43.00 0.1194 25 21.60 0.06000 51 44.00 0.1222 26 22.00 0.06111 52 45.00 0.1250 7- c. -
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. =,. ,., Table 5 Fast Neutron Excesure of the Yankee Rowe Reactor Vessel as a Function of Full Power Operatina Time 2 Operating luence R > 1.0 Me/) { n/cm - sec } Time (EFPY) Surface 1/4 7 3/4 T 1 1.01 x 10 18 5.84 x 10 I7 1.39 x 10 I7 2.92 x 10 IO 6.95 x 10 I7 18 5 5.05 x.10 10 1.01 x 10 I9 5.84 x 10 18 1.39 x 10 8 13.3 1.34 x 10 I9 7.77 x 10 18 1.85 x 10 I8 15 1.52 x 10 I9 8.76 x 10 18 2.09 x 10 18 20 2.02 x 10 I9 1.17 x 10 I9 2.78 x 10 18 24 2.42 x 10 I9 1.40 x 10 I9 3.34 x 10 18 6
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1 _ t F)C-URE lo Activatica Neutron Flux P.onitor Saturated Activity'at the Dosimeter Block Location for Nine 2-Loop PWR Power Plants. (Normali:ed to 1650 MWT) 25
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