ML20062H198

Plant Specific Fast Neutron Exposure Evaluations for First 20 Operating Fuel Cycles of Yankee Rowe Reactor

ML20062H198
Person / Time
Site:	Yankee Rowe
Issue date:	11/30/1990
From:	Shaun Anderson WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
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ML20062H197	List: ML20062H190 ML20062H198
References
NUDOCS 9012040171
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{{#Wiki_filter:_ _ _ _ _ _..... 9 PLANT SPECIFIC FAST HEUTRON EXPOSURE EVALUATIONS FOR THE FIRST 20 OPERATING FUEL CYCLES OF THE YANKEE ROWE REACTOR Stanwood L. Anderson Neverber 1990 Work performed under Shop order No. YAOP-450G WESTINGHOUSE ~ ELECTRIC CORPORATION Energy Systems Business Unit P. O. Box 355 ~ Pittsburgh, Pennsylvania 15230 901204o;7t 99yl2g hDR ADOQ' 05000029 PDC

1.0 BASEIJNE HEUTRON TRANSPORT CALCULATICHS The fast r.eutron fluence evaluations for the Yankee Rowe reactor were carried out using both or.a-and two-dimensional discrete ordinates techniques. Two-dimensional computations were completed with the DDT code (1) run in both the forward and adjoint mode. The ANISH discrete ordinates code (2) was utilized to perform the required ene-dimensional analyses. All of the two-dimensional cases were run in R, Theta geometry, while, the corresponding one-dimens!.cnal runs were carried out using the cylindrical geometry option. 1.1 Geometric Modeling The geometric models used in the fluence evaluations were based on design drawings of the plant with adjustments to the design based on available as-built meiJurements. In particular, as-built information was employed to establish the inner radius and thickness of the pressure vessel as well as the location and thickness of the thermal shield. h A plan view of the Yankee Rowe reactor geometry at the core midplane elevation is shown in Figure 1-1. This model was based on informatien from the followin:. Westinghouse reactor design drawings .;6J500, 646J614, 5490155, 540FB39, 540F857, 646J692, 673C510, and 673C511' on Babcock and Wilcox drawings 45109E, 71695E0 and 66042E; and on Stone & Webster drawings 9699-TV-2A and 9699-TV-2B. Also utilized 4-the development of thje model was additional as-built dimensional information provided by Yataea Atomic Electric Company. The analysis of the 1/8 core sector shown as 0-45 degrees on Figure 1-1 was used as the baseline for the transport evaluations. In terms of the baf fle support structure and thermal shield brackets, the chosen sector minimizes the amount of steel between the reactor core and the pressure vessel resulting in conservative evaluation of the vessel exposure. The mesh line schematic used in the R, Theta analysis for the 0-(* sector shown in Figure-1-1 is depicted in Figure 1-2. Tr9 radial and azimuthal mesh line dimensions corresponding to Figure 1-2 are given in Tables 1-1 and 1 2, respectively. In addition to the two-dimensional calculations, several one-dimensional computations were also carried out to establish adjustment factors for the final results and to perform sensitivity studies of several of the input variables. In establishing a geometric model for the one-dimensional calculations, the radial mesh from the first theta interval shown in Figure 1-14 were taken as a representative configuration. However, since, in the theta sense, they are not global to thu overall problem the baffle support structure and thermal shield strap were removed from the geometry. The redial mesh line schematic for the one-dimensional model is shown in Figure 1-3. The mesh line i dimensions for this model~were taken from Table 1-1. i

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1 TABLE 1-1 RADIAL MESH LINE DIMENSIchS FOR THE R, THETA MODEL 0F THE YANKEE RSWE REACTOR O - 45 DEGRE? SECTOR j LINE RADIUS LINE RADIUS LINE RADIUS LINE RADIUS A fem) A (en) A temi No (em) 1 55.14 36 99.78 71 120.33 106 143.54 2 60.00 37 100.52 72 121.25 107 144.43 3 65.00 38 101.26 73 122.18 108 145.31 ) 4 70.00 39 101.45 74 123.10 109 146.19 5 75.00 40 101.93 75 124.02 110 147.07 6 80.00 41 102.44 76 124.97 111 147.95 7 82.00 42 102.86 77 125.92 112 148.83 P 83.00 43 103.09 78 126.51 113 149.71 9 83.44 44 103.79 79 127.26 114 150.59 I 10 S4.96 45 104.34 80 128.01 115 151.48 11 86.57 46 104.98 81 128.76 116 152.36 12 87.08 47 105.46 82 129.51 117 153.24 13 87.44 48 106.10 83 130.26 118 154.12 14 87.91 49 106.78 84 131.01 119 155.00 15 88.31 50 107.35 85 131.76 120 155.88 16 88.73 51 107.72 86 132.51 12. 156.76 17 89.59 52 108.25 87 133.26 122 157.65 18 90.05 53 108.89 88 134.01 123 158.53 19 90.50 54 109.20 89 134.76 124 159.41 20 91.46 55 110.17 90 134.78 125 160.29 21 91.95 56 110.80 91 135.52 126 164.03 22 92.46 57 111.74 92 136.26 127 167.78 23 -92.75 58 112.08 93 137.00 128-171.52 24 93.45 59 112.93-94 137.38 129 175.26 25 93.73 60 113.77 95 137.77 130 175.90 '26 94.02 61 114.62 96 138.15 131 '176.53 27 94.62 62 115.25 97 138.54 132 177.17 28 95.47 63 115.89 98 138.92 133 178.00 29 95.87 64 116.52 99 139.31 134 180.00 30 96.47 65 117.16 100 139.69 135 183.00 31 96.82 66 117.79 101 139.97 136- '187.00 32 97.63 67 118.30 102 140.30 137 191.00 33-98.10 68 118.81 103 140.90 138 195.00 t 34 98.73 69 119.32 104 141.78 139 200.00 35 99.04 70 119.82 105 142.66 i k i i-l .A_ l

TABLE 1-3 AZIMUTRAL MESH LINE DIHENSIONS FOR THE R, THETA MODEL OF THE YANKEE ROWE REACTOR 0 - 45 DEGREE SECTOR LINE THETA THETA LINE THETA THETA ,,,,Hp,,,,, fdeci frevi ,,Hg, (deci frevi 1 0.00 0.00000 29 23.00 0.06389 2 1.00 0.00278 30 24.00 0.06667 3 2.00 0.00556 31 25.00 0.06944 4 3.00 0.00833 32 26.00 0.07222 5 3.85 0.01069 33 26.40 0.07333 6 4.00 0.01111 34 26.80 0.07444 7 5.00 0.01389 35 27.00 0.07500 8 6.00 0.01667 36 28.00 0.07778 9 7.00 0.01944 37 28.41 0.07893 10 8.00 0.02222 38 28.83 0.08008 11 9.00 0.02500 39 29.26 0.08128 12 9.28 0.02578 40 29.68 0.08246 13 10.00 0.02778 41 30.00 0.08333 14 11.00 3.03056 42 31.00 0.08611 15 12.00 0.03333 43 32.00 0.08889 16 13.00 0.03611 44 33.00 0.09167 17 14.00 0.03889 45 34.00 0.09444 18 15.00 0.04167 46 35.00 0.09722 19 16.03 0.04453 47 36.00 0.1000 20 16.34 0.04539 48 36.70 0.1019 21 16.66 0.04628 49 37.10 0.1031 22 16.97 0.04714 50 38.00 0.1056 23 18.03 0.05000 51 39.00 0.1083 24 19.00 0.05278 52 40.00 0.1111 25 20.00 0.05556 53 41.00 0.1139 26 21.00 0.05833 54 42.00 0.1167 27 21.60 0.06000 55 43.00 0.1194 28 22.00 0.06111 56 44.00 0.3422 57 45.00 0.1;50 -S-

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Stainless Steel /Ca7 ben Steel Tor the reactor internals, pressure vessel, and shield tank wall, nominal stainless steel and carbon steel compositions were employed. The number densities used to generate macroscopic cross-sections for these materials were as follows: Egnker Density (eter/ barn-emi Stainless Carbon Steel Steel Fe 5.943E-02 8.270E-02 Mn 1.762E-03 1.116E-03 Cr 1.768E-02 Ni 7.665E-03 4.420E-04 Si 1.723E-03 C 3.222E-04 9.819E-04 Water In the various transport calculations, water cross-sections were required for the following conditions. Temperature pressure Density fdec Fi fesi) (c/cm3) 535 2000 0.7629 525 2000 0.7725 500 2000 0.7951 450 2000 0.8345 72 (atm) 1.00 The corresponding number densities computed for these pressure / temperature conditions were as follows: Number density fatem/ barn-cm) 535 P 525 P 500 F 450 F 72 F H 5.104E-02 5.168E-02 5.320E-02 5.582E-02 6.690E-02 0 2.552E-02 2.584E-02 2.660E-02 2.791E-02 3.345E-02 Macroscopic cross-sections for use in the fluence evaluations were developed from the SAI14R library (4) using the P3, 47 group data. The SAILOR library is an ENDFB-IV data set developed explicitly for light water reactor applications. The data set is available from the CRNL Radiation Shielding Information Center (RSIC) as data set DLC .'G.

l.3 Houtron Transport Calculation TOPwAPD CALCULATIONS In performing the plant specific fluence evaluations, several sets of transport calculations were carried out. The first series of calculations consisted of three forward computations with variable water temperature in the downcomer regions. The purpose of these calculations was to provide azimuthally dependent correction facters to account for the impact et reduced temperature operation that occured during coastdown periods at the end of each fuel cycle. The forward calculations utilized the geometric representation of the 0-45 degree sector of the reactor geometry described in Section 1.1. The calculations were run in R, theta geometry using a P3 cross-section expansion and an SB order of angular quadrature. The core source utilized in each of the computations was representative of the burnup weighted average of 20 cycles of source data supplied by Yankee Atomic Electric Company (3). A summary of.the core power distribution information used in the calculations is provided in Table 1-3. In regard to Table 1-3, the fuel assembly identification numbers correlate with those specified in Figure 1-1. Also shown on the table are the corc coordinates specified in reference 3. The power distribution data itself represents the relative power in each fuel assembly averaged over the individual fuel cycles. The burnup weighted relative assembly powers from Table 1-3 were employed to provide the base case source distribution for use in the forward transport calculations. In developing the source distribution for the transport calculations, the within assembly spatial gradients were approximated with representative pin by pin power distributions also supplied by Yankee Atomic Electric Company (5). As noted in reference 5, these pin by pin distributions were taken from the core design analysis for fuel cycle 19. These relative spatial gradients were taken as representative of operation over the entire life of the unit. In addition to the relative assembly powers and within assembly gradients, Yankee Atomic also supplied axial peaking factors for each fuel cyclo (3). These peaking factors, representing the maximum flux location, are also listed on Table 1-3. The burnup weighted axial peaking factor of 1.22 was used in the computation of the normalization f actor for the forward DOT runs along with a nominal core power level of 600 MWt. Using the flux (E > 1.0 Mev) response from a series of three DOT runs with downcomer temperatures of 450, 500, and 525 degrees F permits the determination of the impact of downcomer water temperature on vessel exposure rates as a function of azimuthal angle. The results of this evaluation are provided in Figure 1-4 and Table 1-4. In this relative comparison, the 500 F case was taken to be the baseline and data ~for intermediate temperatures were taken fr:n the curves shown in Figure 1-4. The data extracted from the DOT runs to generate the temperature sensitivity study were taken from the vessel cladding at 14.5 and 44.5 degree azimuthal locations. These particular azimuths were taken as representative of the maximum and minimum exposure of the vessel. l _ _ - - - - _ - _ _ _ _ - - _ - _ - - _

TABLI 1-3

SUMMARY

OF RELATIVE CORE POWER DISTRIBUTICHS FOR THE YANKEE R0WE REACTOR FUrt ASSEMBLY TD AVERAGE CYCLE BURNUP H6 H7 HB H9 J6 J7 38 K6 K7 No, (MWD /MTU) 1 2 3 4 5 _6 7 8 9 Fa 1 8470 1.320 1.290 1.070 0.570 1.020 0.930 0.590 0.580 0.440 1.250 2 7866 1.400 1.330 1.040 0.520 1.020 0.970 0.560 0.600 0.420 1.250 3-6329 1.252 1.223 1.096 0.668 1.060 1.014 0.718 0.689 0.542 1.250 4 8734 1.314 1.245 1.097 0.651 1.111 1.023 0.651 0.656 0.504 1.250 5 8893 1.183 1.179 1.213 0.777 1.097 1.145 0.777 0.770 0.615 1.250 6 12419 1.316 1.221 1.115 0.655 1.100 1.041 0.655 0.657 0.501 1.250 7 11963 1.231 1.230 1.105 0.748 1.157 1.114 0.748 0.747 0.585 1.250 8 10142 1.241 1.241 1.148 0.727 1.152 1.086 0.727 0.727 0.574 1.250 9 11946 1.206 1.216 1.200 0.677 1.125 1.149 0.654 0.849 0.503 1.250 l 10 15148 1.115 1.164 1.219 0.864 1.110 1.127 0.871 0.861 0.666 1.250 l 11 .12869 1.166 1.195 1.222 0.770 1.165 1.192 0.804 0.824 0.653 1.250 ~ 12 14879 1.101 1.203 1.173 0.732 1.224 1.144 0.775 0.787 0.620 1.250 13 12890 1.079 1.151 1.190 0.789 1.102 1.133 0.793 0.801 0.640 1.200 I 14 16114 1.132 1.175 1.143 0.708 1.207 1.100 0.719 0.749-0.583 1.200 15 13000 -1.101 1.174 1.156 0.760'1.125 1.099 0.765 0.775 0.618 1.200 l 16 14168 1.095 1.150 1.149 0.746 1.200 1.116 0.757 0.778 0.624 1.175 17 15116 1.158 1.062 1.158 0.756 1.080 1.129 0.752 0.806 0.628 1.175 l 18 16020 1.073 1.058 1.138 0.743 1.222 1.109 0.739 0.814 0.626 1.175 19 15518 1.153 1.053 1.159 0.782 1.068 1.130 0.787 0.818 0.654 1.175 20 16850 1.043 1.031 1.140 0.765 1.203 1.113 0.760 0.831 0.650 1.175 AVG. 249334 1.165 1.165 1.152 0.732 1.137 1.103 0.740 0.770 0.595 1.218 i 1 i Tho eniculcted valu s for flux (E > 1.0 M v) ct the vecsol cladding from tho 500 F base case are shown graphically as a function of azimuthal angle in Figure 1-5. From Figure 1-5, it is noted that the maximum fast flux occurs in the 12 to 17 degree range relative to the cardinal axes and the minimum occurs at the 45 degree location. ADJOINT CALCULATIONS In order to perform the fuel cycle specific evaluations, adjoint transport calculationu based on source points in the vessel cladding zone were carried out for azimuthal locations representing the maximum and minimum exposure of the pressure vessel wall. These azimuthal locations were 14.5 degrees for the maximum and 44.5 degrees for the minimum. Both calculations were carried out using the DOT code with P3 scattering cross-sections and an S8 angular guadrature. The geometric model used in these evaluations was the same as that shown in Figure 1-2. The adjoint source term used in both calculations was the fast neutron flux (E > 1.0 MeV). Following completion of the adjoint DOT computations, the output adjoint fluxes were processed to provide importance function input to be used in evaluating each individual fuel cycle. Using the importance functions developed from the adjoint dot runs along with the power distributions for each fuel cycle, calculations were then carried out to determine the cycle specific neutron exposure of the reactor vessel. In performing the cycle specific fluence calculations, each fuel cycle was subdivided into a period of power operation and a coastdown interval resulting in a total of 40 operating periods. The calculations made use of power distributions given in Table 1-3 with the same power distribution assumed to apply for both subdivisions of each cycle. Temperature corrections.from Figure. 1-4 were used in conjunction with downcomer temperatures from reference 3 in computing the normalization of the fluence results for each period of operation. Cycle specific values of axial peaking f actors were taken from Table 1-3; and, again, the individual factors were assumed to apply to both power operation and coastdown periods. A radial translation factor of 0.987, developed from the 500 F forward DOT run was also used to translate results from the cladding mid-mesh to the actual clad / base metal interface. Tne impact of the stainless steel clad core was evaluated with two one-dimensional transport calculations using the ANISH code in cylindrical geometry. The one-dimensional geometric model was as shown in Figure 1-3. Thei calculations were carried out using the same p3 cross-section sets as w".re used' in the DOT calculations and an S8 order of angular quadrature. The pr,blems were normalized to a core source of 1.0 n/cm-sec and utilized the radial spatial distribution of source from the O degree azimuth in the DOT calculations. A comparison of the ANISH results at the pressure vessal clad l l l-t l FIGURE 1-4 EFFECT OF DOWNCOMER WATER TEMPERATURE ON THE FAST NEUTRON FLUX (E > 1.0 MeV) INCIDENT ON THE PRESSURE VESSEL INNER RADIUS .. i. n si 6 e4 .l. { $.. i!.. 4 .g. + I ' ",,- +.l.. .,i l 44 .,+ "j" .1 .8 . I. + l e' ' nexj ,, m*J 3

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1, ~ ~ TABLE 1-4 EFFECT OF DOWNCOMER WATER TEMPERATURE ON THE TAST NEUTRON TLUX (E > 1.0 MeV) INCIDENT ON THE PRESSURE VESSEL INNER RADIUS r INLET MAXIMUM MINIMUM TEMPERATURE FLUX TLUX fDec F) (n/cm2-seci f n/cm2-sec). 450 0.855 0.823 455 0.868 0.839 460 0.880 0.853 465 0.892 0.871 470 0.908 0.888 475 0.921 0.904 480 0.937 0.922 485 0.951 0.941 490 0.969 0.960 495 0.983 0.980 500 1.000 1.000 505 1.018 1.023 510 1.036 1.045 515 1.055 1.070 i 520 1.074 1.094 I 525 1.096 1.122 -{ h 1 1 I fIj. t 1 L e l i I l 'y 5 l4 l-b . _ 13..

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.. _- --- - C _..____.__.. . _. -. - _..... _ _ _- -~ __._._._-...._----.5 2 8.f _...1 7. 6. 1 d '. V 3. ~~ 2. i 6 q l 4 e 8 i e i e + n 4 e i e i I ( i h I 4 l 1 I i I i i i L o Jo Ao 30 'to So Q L Morwe t M GLF. (des) l l 1 w- --qy.-

1 1 ~. yiold:d a vclu3 of 0.911 for the fuol Edjustm:nt factor for cyclos 1-9. Thic adjustment for the presence of stainless steel clad fuel was used for fuel cycles 1 through 9 which represent the cycles that included stainless fuel on the periphery of the core. Although stainless clad fuel was also precent i during cycles 10 and 11, the assemblies were interior to the core and no credit was taken for their presence, cycle lengths and downcomer temperatures were taken directly from reference 3. The individual cycle correction factors to account for axial peaking factors, temperature variations, and fuel composition are summarized in Table 1-5 Results of the individual adjoint calculations are summarized in Tables 1-6 and > l-7 for the 14.5 degree and 44.5 degree locations, respectively. A fast i neutron (E > 1.0 Mev) fluence profile was developed for the end of cycle 20 by normalizing the azimuthal flux shape from the 500 F base case transport calculation to the cycle specific calculations at the 14.5 and 44.5 degree locations. Since the forward DOT utilized the 20 cycle average core power distribution as input, the calculated azimuthal shape should be quite representative. The appropriate normalization factors were ccmputed as follows by dividing the cycle specific fluence by the DOT calculated flux in the vessel cladding. 14.5 DEG 44.5 DEG CYCLE SPECIFIC FLUENCE 2.13E+19 8.52E+18 DOT FLUX 3.26E+10 1.30E+10 NORMALIZATION 6.53E+08 6.55E+08 ) The agreement between the normalization factors calculated for the maximum and minimum vessel flux locations lends support to the use of the azimuthal shape taken from the forward DOT calculation. In developing the final fluence j profile for the vessel inner radius,-the average value of these normalization factors was used; i.e., 6.54E+08. A summary of the calculated azimuthal i fluence profile is presented in tabular form in Table 1-8 and illustrated graphically in Figure 1-6. The fluence values provided in Table 1-8 and Figure 1-6 are representative of I the clad / base metal interface at the axial location of the peak flux in the 0-45 degree sector. The actual axial location of the peak cannot be documented here, since Yankee Atomic supplied no axial profiles but, only 3 single peaking l factor for each cycle. Again, the data are applicable to the end of fuel cycle 20 and are representative of 6.77E+08 effective full power seconds ~ (21.45 ETPY) of operation. i In addition to the fluence at the' vessel inner radius, the distribution of fluence through the thickness of the pressure vessel wall is of interest. -p Tabulations of the fluence distribution at the maximum and minimum exposure i locations are given in Table 1-9. The radial distributions were obtained by i normalizing the radial flux profiles from the 500 F forward DOT to the relative j distributions given in Table 1-9. l j i 15 - 1 I

s, TABLI 1-3

SUMMARY

OF CYCLE DEPENDENT ADJUSTMENT FACTORS USED To DERIVE CYCLE SPECIFIC TAST NEUTRON TLUENCE AXIAL COOLANT TEMPERATURE PEAKING TUEL TEMP TACTOR CYetr PAcTOR FACTOR _DEG F 14.5 DEG 44.5 DEG 1P 1.250 0.911 500 1.000 1.000 1C 1.250 0.911 454 0.863 0.836 2P 1.250 0.911 496 0.988 0.983 2C 1.250 0.911 464 0.890 0.867 3P 1.250 0.911 501 1.001 1.003 3C 1.250 0.911 473 0.916 0.898 4P 1.250 0.911 507 1.025 1.032 1 4C 1.250 0.911 478 0.931 0.915 SP 1.250 0.911 505 1.018 1.022 SC 1.250 0.911 489 0.965 0.957 6P 1.250 0.911 505 1.018 1.022 i 6C 1.250 0.911 484 0.949 0.938 7P 1.250 0.911 505 1.018 1.022 7C 1.250 0.911 482 0.942 0.930 BP 1.250 0.911 507

1. L ',5 1.032

-8C 1.250 0.911 491 0.971 0.963 9P 1.250 0.911 507 1.025 1.032 l 9C 1.250 0.911 488 0.964 0.952 10P 1.250 1.000 503 1.011 1.013 10C 1.250 1.000 479 0.933 0.920 11P 1.250: 1.000 509 1.032 1.041 11C 1.250 1.000 493 0.978 0.972 12P 1.250 1.000 504 1.014 1.020 12C 1.250 1.000 492 0.973 0.968 1 13P 1.200 1.000 503 1.011 1.013 13C 1.200 1.000 499 0.996 0.995 l l 14P 1.200 1.000 512 1.043 1.053 14C 1.200 1.000 482 0.942 0.930 15P 1.200 1.000 510 1.036 1.045 15C 1.200 1.000 493 0.978 0.972 16P 1.175 1.000 511 1.040 1.050-16C-- 1.175 1.000: 504 1.014 1.019 17P 1.175 1.000 513 1.048 =1.060 17C 1.175 1.000 487 0.959 0.950 18P 1.175 1.000' 511 1.040' .1.050 18C 1.175 .1.000 493 0.978 -0.972 -19P 1.175' 1.000 509 1.032 1.041 19C 1.175 1.000 482 0.942 0.930 20P 1.175 1.000 512 1.043 1.053 20C 1.175 1.000 491 0.971 0.963 1 l l l 9 ! q 1

TABLI a-6

SUMMARY

OF CYCLE SPECITIC TAST NEUTRON TLUX AND TLUENCE AT THE 14.5 DEGREE VESSEL INNER RADIUS LOCATION CYCLE CYCLE CUMULATIVE AVERAGE CUMULATIVE CYCLE TIME TIME TLUX TLUENCE fefes) fn/cm2-seci (n/cq2) NO. fefen) __1.87E+07 2.32E+10 4.33E+17 l 1P 1.87E+07 1C 6.45E+06 2.52E+07 2.00E+10 5.62E+17 2P 1.76E+07 4.2BE+07 2.28E+10 9.63E+17 2C 6.04E+06 4.88E+07 1.04E+10 1.09E+18 3P 9.90E+06 5.87E+07 1.74E+10 1.36E+18 3C 9.15E+06 6.78E+07 2.51E+10 1.59E+18 4P 1.70E+07 8.48E+07 2.68E+10 2.04E+18 l 4C 9.03E+06 9.39E+07 2.42E+10 2.26E+18 j SP 2.19E+07 1.16E+0B 3.13E+10 2.95E+18 i 5C 4.90E+06 1.21E+08 2.96E+10 3.09E+18 6P 2.82E+07 1.49E+08 2.66E+10 3.84E+18 l 6C 9.26E+06 1.58E+08 2.46E+10 4.07E+18 7P 2.70E+07 1.85E+08 3.02E+10 4.88E+18 7C 8.78E+06 1.94E+08 2.80E+10 5.13E+18 SP. 2.43E+07 2.18E+08 2.98E+10 5.85E+18 8C 5.59E+06 2.24E+08 2.82E+10 6.01E+18 9P 2.97Evo7 2.54E+08 2.96E+10 6.89E+18 r 9C 6.24E+06 2.60E+08 2.78E+10 7.07E+18 10P 2.94E+07 2.89E+08 3.69E+10 8.15E+18 10C' 1.05E+07 3.00E+08 3 41E+10 8.51E+18 11P 2.92E+07 3.29E+08 3.70E+10 9.59E+18 11C 4.67E+06 3.34E+08 3.49E+10 9.75E+18 12P 3.68E+07 3.70E+08 3.49E+10 1.10E+19 12C 1.83E+06 3.72E+08 3.35E+10 1.11E+19 13P' 3.20E+07 4.04E+08 3.37E+10 1.22E+19 13C 1.01E+06 4.05E+08 3.34E+10 1.22E+19 14P 3.57E+07 4.41E+08 3.26E+10 1.34E+19 14C 5.66E+06 4.47E+0B 2.93E+10 1.35E+19 15P 3.00E+07 4.77E+08 3.35E+10 1.45E+19 15C 3.92E+06 4.80E+08 3.19E+10 1.47E+19 16P 3.51E+07 5.16E+08 3.33E+10 1.58E+19 .16C 1.43E+06 5.17E+08 3.25E+10 1.59E+19 17P 3.30E+07 5.50E+08 3.39E+10 1.70E+19 17C 5.47E+06 5.55E+08 3.10E+10. 1.72E+19-18P 3.63E+07. 5.92E+08 3.38E+10 1.84E+19 18C-4.37E+06 5.96E+08 3.18E+10 1.85E+19 19P 3.24E+07 6.29E+08 3.42E+10 1.97E+19 19C 6.36E+06 6.35E+08 3.13E+10-1.99E+19 20P 3.76E+07 6.72E+08 3.50E+10 2.12E+19 20C 4.44E+06-6.77E+08 3.23E+10 2.13E+19 l

TABLE 1-7

SUMMARY

OF CYCLE SPECITIC TAST NEUTRON TLUX AND TLUENCE AT THE 44.5 DEGREE VESSEL INNER RADIUS LOCATION CYCLE CYCLE CUMULATIVE AVERAGE CUMULATIVE CYCLE TIME TIME TLUX TLUENCE NO. (efesi (efes) (n/cm2-seci (n/er2) IP 1.87E+07 1.87E+07 9.80E+09 1.83E+17 1C 6.45E+06 2.52E+07 8.23E+09 2.36E+17 2P 1.76E+07 4.28E+07 9.43E+09 4.02E+17 2C 6.04E+06 4.88E+07

8. 2 8E+ 0 9 4.52E+17 3P 9.90E+06 5.87E+07 1.15E+10 5.66E+17 3C 9.15E+06 6.78E+07 1.03E+1C 6.60E+17 4P 1.70E+07 8.48E+07 1.11E+10 8.49E+17 4C 9.03E+06 9.39E+07 9.83E+10 9.38E+17 SP 2.19E+07 1.16E+08 1.28E+10 1.22E+18 5C 4.90E+06 1.21E+08 1.20E+10 1.28E+18 6P 2.82E+07 1.49E+08 1.11E+10 1.59E+18 6C 9.26E+06 1.58E+08 1.02E+10 1.68E+18 7P 2.70E+07 1.85E+08 1.23E+10 2 01E+18 7C 8.78E+06 1.94E+08 1.12E+10 2.11E+18 8P 2.43E+07 2.18E+08 1.22E+10 2.41E+18 8C 5.59E+06 2.24E+08 1.13E+10 2.47E+18 9P 2.97E+07 2.54E+0B 1.14E+10 2.81E+18 9C 6.24E+06 2.63E+08 1.05E+10 2.88E+18 10P 2.94E+07 2.80E+08 1.51E+10 3.32E+18 10C 1.05E+07 3.0CE+08 1.37E+10 3.46E+18 11P 2.92E+07 3.29E+08 1.48E+10 3.89E+18 11C 4.67E+06 3.34E+08 1.38E+10 3.96E+18 12P 3.68E+07 3.70t+08 1.40E+10 4.47E+18 12C 1.83E+06 3.72E+08 1.33E+10 4.50E+18 13P 3.20E+07 4.0.E+08 1.35E+10 4.93E+18 13C 1.01E+06 4.05E+08 1.33E+10 4.95E+18 14P 3.57E+07 4 41E+08 1.31E+10 5.41E+18 14C 5.66E+06 4.47E+08 1.16E+10 5.48E+18 15P 3.00E+07 4.77E+08 1.36E+10 5.89E+18 15C 3.92E+06 4.80E+08 1.25E+10 5.93E+18 16P 3.51E+07 5.16E+08 1.33E+10 6.40E+18 16C 1.43E+06 5.17E+08 1.28E+10 6.42E+18 17P 3.30E+07 5.50E+08 1.33E+10 6.86E+18 17C 5.47E+06 5.55E+08 1.20Ev10 6.92E+18 18P 3.63E+07 5.92E+08 1.31E+10 7.40E+18 18C 4.37E+06 5.96E+0B 1.21E+10 7.45E+18 19P 3.24E+07 6.29E+08 1.35E+10 7.89E+18 19C 6.36E+06 6.35E+08 1.20E+10 7.97E+18 20P 3.76E+07 6.72E+08 1.33E+10 8.47E+18 20C 4.44E+06 6.77E+08 1.22E+10 8.52E+18 i '

5'. [ TABLE 1-8 AZIMUTRAL~ DISTRIBUTION OF KAXIMUM FAST NEUTRON (E > 1.0 MeV) FLUENCE - AT THE PRESSURE VESSEL INNER RADIUS (END OF CYCLE 20 - 21.45 EFPY) THETA FLUENCE THETA FLUENCE (Deci (n/cm2) fDeci (n/er2) 0.50 1.95E+19 23.50 1.76E+19 1.50 1.96E+19 24.50 1.70E+19 2.50 1.99E+19 25.50 1.64E+19 0 3.43 2.01E+19 26.20 1.60E+19 3.93 2.02E+19 26.60 1.58E+19 o 4.50 2.04E+19 26.90 1.56E+19 5.50 2. 0 5 E+,19 27.50 1.52E+19 i 6.50 2.06E+19 28.21 1.48E+19 ~ 7.50 2.07E+19 28.62 1.45E+19 8.50 2.09E+19 29.05 1.42E+13 9.14 2.11E+19 29.47 1.39E+19 9.64 2.11E+19 29.84 1.37E+19 10.50 2.13E+19 30.50 1.33E+19 11.50 2.13E+19 31.50 1.28E+19 12.50 2.14E+19 32.50 1.22E+19 13.50 2.14E+19 33.50 1.17E+19 14.50' 2.13E+19 34.50 1.12E+19 r '15.52 2.10E+19 35.50 1.07E+19 16.19 2.08E+19 36.35 1.04E+19 s 16.50 2.08E+19 36.90 1.02E+19 16.82 2.08E+19 37.55 9.95E+18 17.50 2.08E+19 38.50 9.69E+18 18.50 2.04E+19 39.50 9.36E+18 19.50 1.99E+19 40.50 9.10E+18 20.50 1.94E+19 41.50 8.90E+18 21.30 1.89E+19 42.50 8.70E+18 21.80 1.86E+19 43.50 8.57E+18 22.50 1.82E+19 44.50 8.51E+18 k 4 b I i ? L 4 c1 ~b 1 1-I I l$ 9, r ..w-, e

4 F8GURE S-6 A;:IMUTHAL DISTRIBUTION OF MAXIMUM FAST NEUTRON (E > 1.0 MeV) FLUENCE AT THE PRESSURE VESSEL INNER RADIUS (END OF CYCLE 20 - 21.45 LFPY) l l l , gok- _.3........-'

• T..

. 2::L: ' ':.~.: '. L... : - : L: i:2 ::L ~.2.'^ : -'_-.::I.~_^^^':~~: ~: " ::. ':T~.:::::~:::._^..:::. ' 7 ^ ~.:_,,,_-j^ 7 6.., k 5.._ 'E=28Mi=i -;=isfife=_.=N _ - =.t-=_._2?i==.F#:-5E:9.t==2 =ffi._ '.^1X_..=...Js=..s==:~:E-sh e.= lf2t+d-ssss u -g Mrg ;n._;_...... =.. _ _ _. _.. _.._. _ _.. _ _. _ _ _ _ _. _ _ _. g _ _.. =... J,, j== = =._ r g u 7 ~~ " " #2 F-3.. ~ w C. y gw ~ b N N d., i x \\! a i \\ x ] I ~. N i i 1){dg _ _ = =.. _ _ _... _ _... _.. _ _ _ _. . _. _. a e i i g N i i Q . _ _ _ _rin ue+;a m meau..:.:..=m25-s====_. - - .. _ _ _. _ _....._:.1. mim..__. _-.=w=w=5===mi-w+= 2 hat====m.=. s_.. ~.. g, 7.' p Cw 6. 2 ~ 1 5. +rgE-gig===t = =-m-r==== $ rz:. :_pE=a f====== -- =1 7:==r:.;; - ^ = :==t=== z_ _...,_._ J.._ - -.. _ ~. _ E 3. 2. 1 i s 4 9 h I s I i e i I tl N.! I O )o JLo ao go 5o 91.1 MUTMQL NdCrLE Q e.S ) 20 -

1 4 r TABLE 1-9 '~ i FAST NEUTRON TLUENCE (E > 1.0 MeV) AS A FUNCTION OF RADIAL POSITION WITHIN THE YANKEE ROWE PRESSURE VESSEL (END OF CYCLE 20 - 21 45 ETPY) NEUTRON TLUENCE (n/cm2) RADIUS fem) MAXIMUM MTNTMUM 139 97 2.14E+19 8.51E+18 i 140 14 2.12E+19 8.44E+18 140.60 2.03E+19 8.11E+18 141.34 1.88E+19 7.59E+18 142.22 1.70E+19 6.86E+18 ^ 143.10 1 51E+19 6.15E+18 143 99 1.34E+19 5.50E+18 144.87 1.19E+19 4.88T+18 145.75 1.05E+19 4.33E+18 146.63 9.21E+18 3.84E+18 147.51 0.08E+18 3.40E+18 148.39 7.16E+18 3.00E+18 149.27-6.26E+18 2.65E+18 150.15 5.49E+18 2.34E+18 151.04 4.81E+18 2.06E+18 I 151.92 4.21E+18 1.81E+18 152.80 3.68E+18 -1.59E+18 153.68 3.21E+18 1.39E+18 154.56 2.80E+18 1.22E+18 155.44 2.43E+18, 1.07E+18 156.32 2.11E+18 9.30E+17 156.76 1 97E+18 8.71E+17 P I y l a. ( L L \\ L 6 1, ~. 'f

3.0 ADDZTICHAL CALCULATIONS In addition to the the baseline calculations described in Section 1.0, several additional studies were carried out as a part of this evaluation. The first involved the determination of the ef fect that the extra baf fle structure and thermal shield hardware that is present in the 45-90 degree sector of Figure 1-1 would have on the incident flux on the pressure vessel. The second study was to determine the reduction in exposure due to the presence of spacers and the removal of fuel rods in the core quadrant opposite the sector described in the baseline computations. The third study investigated the impact of variations in the inner radius of the pressure vessel on the incident vessel flux. The fourth study was to determine the increase in vessel exposure rates that would occur following complete removal of the thermal shield. 2.1 Calculation of Vessel Exposure Rates in the 45-90 Degree Sector In Section 1.0, it was noted that the 0-45 degree sector chosen for the baseline transport calculations was a conservative representation of the i geometry of the reactor because the amount of steel structure between the core and the pressure vessel was minimized. To assess the impact of the extra baffle structure and thermal shield support hardware that is actually present in the 45-90 degree sector an additional R, Theta DOT calculation was run using the geometry representative of the 45-90 degree sector shown in Figure 1-1. In performing the dot run, the 500 F base case was adjusted to accomodate the extra structure in the geometric model. However, all other input (neutron cross-sections, quadrature) remained the same. The results of the

sources, 45-90 degree sector calculation are provided in Table 2-1 as a comparison of the-calculated flux at the vessel cladding from the two corresponding 1/8 core models.

In regard to the data presented in Table 2-1, it should be noted that the order of azimuthal angle runs from 0-45 degrees in the one case and 90-45 degrees in the other. This permits the comparison of data on a one to one basist i.e. 0.50 degrees in the 0-45 model is equivalent to 89.5 degrees in the 45-90 model. An examination of Table 2-1 shows that in the 45-90 degree sector incident flux levels at the pressure vessel clad are either equivalent to or somewhat less than corresponding values from the baseline computation. The maximum flux reduction afforded by the presence of the additional structural material is on the order of 5-7%. 2.2 Impact of Core Spacers and Fuel Rod Removal A schematic of the Yankee Rowe baf fle cavity is shown in Figure 2-1. The baselina transport calculations deceribed in Section 1.0 conservatively positioned the core in the upper right quadrant relative to the configuration depicted in Figure 2-1. This positioning placed fuel directly against the baffle plates on all peripheral fuel surfaces. In the opposite quadrant (lower lef t on Tipre 2-1), however, the presence of stainless steel spacers resulted in fuel being displaced from the baffle plates by one row of fuel. During the first 17 eycles of operation fuel bearing rods were in place at the end of each However, subsequent to cycle,17 these rods were removed creating an spacer. entire row of non-fuel bearing area adjacant to the core baffle. 1...

1 TABLE 3-1 TAST NEUT.h'N TLUX (E > 1.0 MeV) AT THE PRESSURE VESSEL CLAD IN THE 0-45 AND 45-90 DEGREE SECTORS CYCLE 1-4 AVERAGE POWER DISTRIBUTION - 500 T DOWNCOMER NET.'RON TLUX NEUTRON TLUX THETA _ (n/ im2-sec) THETA (n/cm2-sec) (Deci 0-4B 90-45 (Dec) 0-45 90-45 0.50 2.9BE+10 2.80E+10 23.50 2.69E+10 2.68E+10 1.50 3.00E+10 2.82E+10 24.50 2.60E+10 2.59E+10 2.50 3.04E+10 2.86E+10 25.50 2.51E+10 2.50E+10 3.43 3.07E+10 2.90E+10 26.20 2.45E+10 2.43E+10 3.93 3.09E+10 2.93E+10 26.60 2.41E+10 2.38E+10 4.50 3.11E+10 2.95E+10 26.90 2.38E+10 2.34E+10 5.50 3.13E+10 2.99E+10 27.50 2.32E+10 2.27E+10 6.50 3.15E+10 3.01E+10 28.21 2.26E+10 2.20E+10 7.50 3.17E+10 3.04E+10 28.62 2.22E+10 2.15E+10 8.50 3.19E+10 3.05E+10 29.05 2.17E+10 2.10E+10 9.14 3.22E+10 3.06E+10 29.47 2.13E+10 2.0$E+10 9.64 3.23E+10 3.07E+10 29.84 2.09E+10 2.01E+10 10.50 3.25E+10 3.06E+10 30.50 2.03E+10 1.94E+10 11.50 3.26E+10 3.04E+10 31.50 1.95E+10 1.06E+10 12.50. 3.27E+10 3.04E+10 32.50 1.86E+10 1.79E+10 13.50 3.27E+10 3.34E+10 33.50 1.78E+10 1.72E+10 14.50 3.26E+10 3.04E+10 34.50 1.71E+10 1.67E+10 15.52 3.21E+10 3.03E+10 35.50 1.64E+10 1.61E+10 16.19 3.18E+10 3.03E+10. 136.35 1.59E+10 1.57E+10 16.50 3.18E+10 3.04E+10 36.90 1.56E+10 1.54E+10 16.82 3.18E+10 3.01E+10 37.55 1.52E+10 1.51E+10 17.50 3.18E+10 2.96E+10 38.50 1.48E+10 1.47E+10 18.50 3.12E+10 2.90E+10 39.50 1.43E+10 1.43E+10 19.50 3.04E+10 2.85E+10 40.50 1.39E+10 1.39E+10 20.50 2.96E+10 2.81E+10 41.50 1.36E+10 1.36E+10 21.30 ,2.89E+10 2.76E+10 42.50 1.33E+10 1.33E+10 21.80 2.84E+10 2.68E+10 43.50 1.31E+10 1.31E+10 22.50 2.78E+10 2.59E+10 44.50 1.30E+10 1.20E+10 L 9 I b t -

In erd:r to caticato tho imptet of thic coro displaconOnt cdjoint cv01uations at the 44.5 degree azimuth were performed with the neutron soure:s in th;as peripheral rods set to zero. For cycles 1-17 the sources in the spacer locations were removed; while, for cycles 18-20 the rods adjacent to the ends of the spacers were also deleted from the calculation. The results of this adjoint evaluation are summarized in Table 2-2. An examination of the data provided in Table 2-2 and the corresponding information from Table 1-7 shows the following comparison. FLUENCE WITHOUT SPACERS 8.52E+18 FLUENCE WITH SPACERS 7.88E+18 Thus, to a first approximation the 44.5 degree vessel exposure in the octants containing spacers and deleted fuel rods would be lower than the baseline exposure projections by a factor of 0.925. 2.3 Parameter Study of Vessel Inner Radius The geometric model used in these fluence evaluations made use of as-built dimensions for the inner radius of the vessel cladding. The nominal radius was established as 139.69 cm. by averaging several measured values provided by Yankee Atomic. The span of these measured radii ranged from a minimum of 139.32 cm. to a maximum of 140.06 cm. Thus, the measurements varied from nominal by from -0.31 to +0.37 cm. To assess the uncertainty associated with these radial variations a series of one-dimensichal ANISH runs was performed using the model described in Section 1.0. The zirconium core calculation was taken as the base case and two additional runs were carried out with the minimum and maximum radii. The following data extracted from the cladding location of each computation demonstrate the impact of vessel radius on the incident exposure rate. Relative Min er Max Flux Nominal Vessel IR = 139.38 cm 2.237E-07 1.051 Vessel IR = 139.69 cm 2.129E-07 1.000 i Vessel IR = 140.05 cm 2.008E-07 0.943 As can be seen from the above tabulation, the uncertainty associated with vessel radius is on the order of +/- 5-6 %. l + + / / e e e ( TIGURE 3-1 SCHEMATIC OF YANKEE ROWE BATTLE CAVITY xxxxxxyxCsxxxxx-sy xxxx b-h s s s s T/,/"72 FT i Nxxxxy \\gx x x x x / N p s, s 1 mm t-xxxxo 8_ k b N 'e N s T //"71 i i i Ns A x x x NN s s f 4 s s N s I N 1 1 1 I i 1 F/, 2 N N -r s q d s s; N s h,, i i i r_--y 7 s N N N s g 7 ~ f N U/M i i i 1 7 s N s / 7 s m'm g Kg i r-i s s s d s p / N \\ d hR /f s e N s s / b1 sx 3 s s s N \\Y xsxs @ SM su u d I ( i s

) l TABLE 2-2

SUMMARY

OF TAST NEUTRON (E > 1.0 MeV) TLUX AND TLUENCE AT THE 44.5 DEGREE AZIMUTHAL LOCATION WITH PERIPHERAL TUEL RODS REMOVED CYCLE CYCLE CUMULATIVE AVERAGE CUMULATIVE CYCLE TIME TIME FLUX TLUENCE i NO. (efes) fefosi In/cr2-seci (n/cm2) IP 1.87E+07 1.87E+07 9 13E+09 1.71E+17 1C 6.45E+06 2.52E+07 7.67E+09 2.20E+17 2P 1.76E+07 4.28E+07 8.79E+09 3.75E+17 2C 6.04E+06 4.88E+07 7.73E+09 4.22E+17 3P 9.90E+06 5.87E+07 1.07E+10 5.27E+17 3C 9.15E+06 6.78E+07 9.55E+09 6.15E+17 4P 1.70E+07 8.48E+07 1.03E+10 7.90E+17 4C 9.03E+00 9.39E+07 9.15E+10 8.73E+17 SP 2.19E*07 1.16E+08 1.19E+10 1.13E+18 SC 4.90E+06 1.21E+08 1 11E+10 1.19E+18 6P 2.82E+07 1.49E+08 1.03E+10 1.48E+18 6C 9.26E+06 1.58E+08 9 45E+09 1.56E+18 7P 2.70E+07 1.85E+08 1.14E+10 1.87E+18 7C 8.78E+06 1.94E+08 1.04E+10 1.96E+18 8P 2.43E+07 2.18E+08 1.13E+10 2.24E+18 ? 8C 5.59E+06 2.24E+08 1.05E+10 2.30E+13 9P 2.97E+07 2.54E+08 1.07E+10 2.61E+18 9C 6.24E+06 2.60E+08 9.80E+09 2.68E+18 10P 2.94E+07 2.89E+08 1 39E+10 3.09E+18 10C 1.05E+07 3.00E+08 1.27E+10 3.22E+18 11P 2.92E+07 3.29E+08 1.37E+10 3.62E+18 11C 4.67E+06 3.34E+08 1.28E+10 3.68E+18 12P 3.68E+07 3.70E+08 1.30E+10 4.16E+18 12C 1.83E+06 3.72E+08 1.24E+10 4.18E+18 13P 3.20E+07 4.04E+08 1.06E+10 4.58E+18 13C 1.01E+06 4.05E+08 1.23E+10 4.59E+18 14P 3.57E+07 4.41E+08 1.22E+10 5.03E+18 14C 5.66E+06 4.47E+08 1.08E+10 5.09E+18 15P 3.00E+07 4.77E+08 1.26E+10 5.47E+18 15C 3.92E+06 4.80E+08 1.16E+10 5.51E+18 . 51E+07 5.16E+08 1.23E+10 5.95E+18 167 16C 1.43E+06 5.17E+08 1.19E+10 S t.96E+18 17P

3. DOE +07 5.50E+08 1.24E+10 6.37E+18 17C 5.4PE+06 5.55E+08 1.11E+10 6.43E+18 18P

3. 6:lE+ 07 5.92E+0B 1.18E+10 6.86E+18 18C 4.37E+06 5.96E+08 1.09E+10 6.91E+18 19P 3.24E+07 6.29E+08 1.22E+10 7.31E+18 19C 6.36E+06 6.35E+08 1.09E+10 7.37E+18 20P 3.76E+07 6.72E+58 1.20E+10 7.83E+18 20C 4.44E+06 6.77E+08 1.10E+10 7.88E+18 I'i I.

3.4 Thorm:1 Shiold R;I; val i The impact-of thermal-shield remova'l on the exposure of the reactor pressure vessel was determined by rerunning the baseline 500 T forward DOT calculation with the thermal shield (zone 5 or, Tigure 1-2) replaced with downcomer water at $00 T. The results of the evaluation are provided on Table 2-3. 1 Trom Table 2-3 it is noted that, depending on azimuthal angle, the-removal of i -the thermal shield would cause an increase in the fast neutron exposure rate at l the pressure vegnal. inner radius by a-factor of from 1.39 to 1.50. This 3 variation with angle is caused by the varying steel / water ratio, but in this 1 case does not exhibit a large swing from 0 - 45 degrees. ) i j r I j. t t [ I i i t t .N r s a i. L r 't 6 9 'i l i , 1 l

i

• E.

TABLE 3-3 i TAST NEUTRON TLUX (E > 1.0 HeV) AT THE PRESSURE VESSEL CLAD 1 J WITH AND WITHOUT THE THERMAL SHIELD IN PLACE CYCLE 1-20 AVERAGE POWER DISTRIBUTION - 500 F DOWNCOMER ' *VTRON TLUX NEUTRON TLUX (n/cm2-see) (n/em2-see) 1 1 WITH WITHOUT WITH WITHOUT THETA THERMAL THERMAL THETA THERMAL THEPp.AL JDect. SHIELD SHIELD P.ATIO (Dec) SHIELD SHIELD RATIO O.50 2.98E+10 4.15E+10 1.39 23.50 2.69E+10 3.86E+10 1.43 1.50 3.00E+10 4.20E+20 1.40 24.50 2.60E+10 3.75E+10 1.44 2.50 3.04E+10 4.27E+A0 1.40 25.50 2.51E+10 3.63E+10 1.45 3.43 3.07E+10 4.33E+10 1.41 26.20 2.45E+10 3.55E+10 1.45 3.93 3.09E+10 4.35E+10 1.41 26.60 2.41E+10 3.50E+10 1.45 4.50 3.11E+10 4.38E+10 1.41 26.90 2.38E+10 3.46E+10 1.45 5.50 3.13E+10 4.41E+10 1.41 27.50 2.32E+10 3.37E+10 1.45 6.50 3.15E+10 4.42E+10 1.40 28.21 2.26E+10 3.30E+10 1.46 i 7.50 3.17E+10 4.4SE+10 1.40 28.62 2.22E+10 3.24E+10 1.46 8.50 3.19E+10 4.50E+10 1.41 29.05 2.17E+10 3.18E+10 1.47 9.14 '3.22E+10 4.55E+10 1.41 29.47 2.13E+10 3.13E+10 1.47 9.64 3.23E+10 4.58E+10 1.42 29.84 2.09E+10 3.08E+10 1.47 10.50 3.25E+10 4.61E+10 1.42 30.50 2.03E+10 2.99E+10 1.47 11.50 3.26E+10 4.60E+10 1.41 31.50 1.95E+10 2.87E+10 1.47 12.50 3.27E+10 4.57E+10 1.40' 32.50 1.86E+10 2.75E+10 1.48 13.50 3.27E+10 4.56E+10 1.39 33.50 1.78E+10 2.63E+10- 1.48 t 14.50 3.26E+10 4.55E+10 1.40 34.50 1.71E+10 2.53E+10 1.48 15.52 3.21E+10 4.55E+10 1.42 35.50- 1.64E+10 2.43E+10 1.48 16.19 3418E+10 4.49E+10. 1.41 36.35 1.59E+10 2.36E+10 1.48 16.50 3.18E+10 4.48E+10 1.41 36.90 1.56E+10 2.32E+10 1.49 16.82 3.18E+10 4.46E+10 1.40 37.55 1.52E+10 2.27E+10 1.49 17.50' 3.18E+10 4.44E+10 1.40 38.50 1.48E+10 2.21E+10 1.49 18.50 3.12E+10 4.35E+10 1.39 39.50 1.43E+10 2.15E+10 1.50 19.50 3.04E+10 4.24E+10 1.39 40.50 1.39E+10 2.10E+10 1.51 { 20.50 2.96E+10 4.14E+10 1.40 41.50 1.36E+10 2.05E+10 1.51 21.30 2.89E+10 4.06E+10 1.40 42,50 1.33E+10 2.00E+10 1.50 21.80 2.84E+10 -4.01E+10 1.41 43.50 1.31E+10 1.97E+10 1.50 22.50 2.78E+10 3.95E+10 1.42 44.50 1.30E+10 1.95E+10 1.50-1 i 9 L

u>- o 3.0 - COMPARISON OF EXPOSURE CALCULATIONS WITH MEASUREMENTS Surveillance Caesule Measurements over the course of the operating lifetime of power reactors, surveillance capsules are periodically withdrawn to provide materials data as well as neutron dosimetry applicable to the specific reactor. These dosimetry resultr afford the opportunity for power reactor benchmarking that, in turn, permits an assessment of the accuracy of the analytical techniques used to calculate vessel exposures. The following is a summary of comparisons of plant specific I calculations with capsule measurements for a variety of Westinghouse reactors. Neutron Flux (n/cm2-soc) PLANT / CAPSULE CALCULATED MEASURED C/M A1 9.25E+10 1.01E+11 0.916 B1 1.08E+11 1.34E+11 0.806 C1 9.25E+10 1.06E+11 0.873 2 D1 9.5BE+10 1.01E+11 0.949 El 9.44E+10 1.05E+11 0.899 F1 1.07E+11 1.24E+11 0.863 G1 9.23E+10 9.47E+10 0.975 [ H1 9.51E+10 1.09E+11 0.872 Il 9.51E+10 1.09E+11 0.872 r J1 9.32E+10 1.30E+11 0.717 J2 8.94E+10 1.01E+11 0.885 K1 8.33E+10 1.04E+11 0.801 K2 9.21E+10 1.10E+11 0.837 I L1 9.52E+10 1.16E+11 0.821 L2 8.34E+10 9.05E+10 0.922 M1 1.10E+11 1.43E+11 0.769 M2 6.64E+10 8.56E+10 0.776 M3 1.10E+11 1.46E+11 0.753 N1 1.31E+11 1.61E+11 0.814 N2 1.19E+11 1.42E+11 0,838 N3 7.66E+10 8.27E+10 0.926 01 6.15E+10 6.92E+10 0.889 02 6.77E+10 7.30E+10 0.927 03 5.94E+10 6.28E+10 0.946 P1 5.64E+10 6.47E+10 0.872 P2 5.96E+10 6.84E+10 0.871 P3 5.41E+10 4.99E+10-1.084 i Q1 6.45E+10 7.47E+10 0.863 Q2 7.04E+10 8.43E+10 0.835 Q3 7.26E+10 7.12E+10 1.020 Q4 6.33E+10 -5.78E+10 1.095 g AVERAGE C/M RATIO FOR 31 SURVEILLANCE DATA POINTS 0.880 1 SIGMA STANDARD DEVIATION OF THE DATA BASE 0.085 7

1 Fron thic curvoillcnco cipsulo d8ta basa, it is CSon that et tha capsuls _ locations ca cu ated values using the carrent radiation transport l l _3 methodology tend to be low relative to measurement by about 12 %. Reactor Cavity Measurements over the course of the last decade many utilities, either because of a need for very accurate fluence evaluations for regulatory concerns or as a mode of data aquisition to establish a life extension data base, have installed neutron dosimetry in the annular space between the outer radius of the reactor vessel and the inner radius of the primary biological shield. Programs have been in place since 1983 and comparisons of caler '.ations with measured data from these programs provide an additional means to benchmark analytical capability against data obtained directly from power reactor facilities. The following.is a summary of comparisons-of plant specific calculations with cavity dosimetry measurements from a variety of Westinghouse reactors. Neutron Flux (n/cm2-sec) PLANT / DATA POINT CALCULATED MEAstUJ.Q_ C/M R1 6.86E+08 8.13E+08 0.844 R2 6.14E+0B 6.75E+08 0.910 l-R3 4.01E+08 3.7BE+08 1.061 R4 2.75E+08 2.99E+08 0.920 R5 5.65E+08 6.49E+08 0.871 l-R6 5.25E+08 6.37E+08 0.824 R7 2.97E+08 3.28E+08 0.905 R8 2.40E+08 3.17E+08 0.757 L S1 6.62E+08 8.22E+08 0.805 S2 3.44E+08 6.37E+08 1.011 l S3 6.60E+08 6.99E+08 0.944 RS 4 4.92E+08 5.43E+08 0.909 SS-5,07E+08 6.65E+08 0.762 S6 4.64E+08 5.82E+08 0.797 S7 4.23E+08 4.02E+08 1.052 S8 3.39E+08 3.70E+08 0.916 T1 5.36E+08 5.51E+08 0.973 i L T2 4.44E+08 4.57E+08 0.972 T3 3.33E+08 3.58E+08 0.930 T4 2.04E+08 2.34E+08 0.872 TS 5.25E+08 6.39E+08 0.822 T6 4.44E+08 5.10E+08 0.871 T7-3.83E+08 4.34E+08 0.882 T8' 2.56E+08 2.79L+08 0.918 l 5:, NGutron Flux (n/cm2-osc) PLANT / DATA POINT CALCULATED MEASURED C/M U1 4.76E+08 5.53E+08 0.861 U2 4.16E+08 5.12E+08 0.813 U3 3.70E+08 4.39E+08 0.843 U4 2.40E+08 2.94E+0B 0.816. V1 1.73E+09 1.87E+09 0.925 V2 1.45E+09 1.69E+09 0.858 V3 1.12E+09 1.23E+09 0.911 V4 9.28E+0B 1.10E+09 0.844 AVERAGE C/M RATIO FOR 32 CAVITY DATA POINTS 0.887 1 SIGMA STANDARD DEVIATION OF THE DATA BASE 0.073 From this' cavity dosimetry data base, it is seen that at the cavity sensor locations calculated values using the current radiation transport methodology tend to be low relative to measurement by about 11 %. This observation is fully consistent with the previous comparisons from the surveillance capsule data base. Maine Yankee Measurements In addition to'the data base comparisons on Westinghouse designed reactors, the current Westinghouse radiation transport methodology was i-used for an analysis of the Maine Yankee reactor that included a comparison of calculation with dosimetry results from both surveillance capsules and reactor cavity. That analysis with the comparative information,is documented reference 6. Pertinent calculation / measurement cor.psrisons from that evaluation are as-follows: Neutron Flux (n/cm2-sec) LOCATION CALCULATION MEASUREMENT C/M 35 DEG ACC. CAPSULI 3.45E+11 4.81E+11 0.717 263=DEG WALL CAPSULE 3.35E+10 3.89E+10 0.861 263 DEG CAVITY CYCLE 7 '6.37E+0B 7.98E+08 0.798 263 DEG CAVITY --CYCLE 8 5.16E+08 6.53E+08 0.790 f 0.792 AVERAGE C/M RATIO FOR 4 DATA POINTS The observed C/M ratios are slightly lower than the overall Westinghouse data base averages, but are still within approximately 1 sigma of those averages. Thus, the dosimetry comparisons indicate the same trend towards underprediction as was observed with the prior comparisons. I

4.0 REFERENCES

i e-1 - Soltesz, R. G. et. al., " Nuclear Rocket Shielding Methods, Modification, Updating, and Input Data Preparation - Volume 5 - Two-Dimensional Discrete ordinates Transport Technique", WANL-PR- ( LL) -0 3 4, August 1970. 2 - Soltest, R. G. et. al., " Nuclear Rocket Shielding Methods, Modification, Updating, and Input Data Preparation - Volume 4 - One-Dimensional Discrete ordinates Transport Technique", W ANL-PR- ( LL) - 03 4, August 1970. 3 - Yankee Atomic Electric Company letter # RP90-163, Kevin J. Morrissey to S. L. Anderson, June 1, 1990. 4 - SAILOR RSIC DATA LIBRARY COLLECTION DLC-76, " Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. 5 - Yankee Atomic-Electric Company letter 1 RP90-299, Kevin J. Morrissey to S. L. Anderson, October 5, 1990. 6 - Anderson, S. L. and Lippincott, E. P., " Summary of Fast Neutron Exposu.a Evaluations for the Maine Yankee Reactor Pressure Vessel", WCAP-11335, November 1986. o .6 Attachment B Results of the PTS Fracture Mechanics Analysis I

4 -. i t i fable 1 i Fluence Distribution for Beltline Materials P:ak fluence at 21.44 EFPY (1990)= 2.14 e19 n/cm2 A21MUTHAL VARIATION Axlal Welds 0 to 5 5 to 10 10 to 1515 to 20 20 to 25 25 to 30 30 to 35 35 to 40 40 to 45 40 to 45 Upper Plate 10 to 20 0.300 0.314 0.321 0.302 0.268 0.226 0.183 0.149 0.132 0.132 20 to 30 1.030 1.077 1.102 1.038 0.919 0.775 0.628 0.510 0.453 0.453 j 30 to 40 1.625 1.698 1.738 1.637 1.449 1.222 ^ 990 0.805 0.714 0.714 40 to 50 1.853 1.936 1.982 1.867 1.653 1.393 1.130 0.917 0.814 0.814 i % of Height 50 to 60 1.947 2.034 2.082 1.961 1.737 1.464 1.187 0.964 0.856 0.856 60 to 70 1.947 2.034 2.082 1.%1 1.737 1.464 1.187 0.964 0.856 0.856 70 to f0 2.001 2.091 2.140 2.016 1.785 1.504 1.220 0.991 0.880 0.880 80 to 90 1.965 2.053 2.101 1.980 1.753 1.477 1.198 0.973 0.864 0.864 90 to 100 1.947 2.034 2.082 1.%1 1.737 1.464 1.187 0.964 0.856 0.856 cire Weld 1.781 1.861 1.905 1.794 1.588 1.339 1.086 0.882 0.783 Lower Plate 35 to 40 0 to 10 1.781 1.861 1.905 1.794 1.588 1.339 1.086 0.882 0.783 0.882 % cf Height 10 to 20 1.401 1.*64 1.4 98 1.411 1.249 1.053 0.854 0.694 0.616 0.694 + 20 to 30 0.808 0.845 0.i45 0.814 0.721 0.608 0.493 0.400 0.355 0.400 30 to 40 0.140 0.146 0.150 0.1',11 0.125 0.105 0,085 0.069 0.062 0.069

)

i, s i l i i

-f m. k Tebte 2 Mean Dette RINDT Distribution for Belttine Materlets - Peak ituence et 21.44 EFPY (1990): 2.14 e19 n/cm2 A21MUTHAL YARIATION Axlet Welds 0 to 5 5 to 10 10 to 1515 to 20 20 to 25 25 to 30 30 to 35 35 to 40 40 to 45 40 to 45 Upper Plate 10 to 20 102 105 106 103 87 78 70 66 157 20 to 30 192 195 197 193 184 170 154 138 129 227 30 to 40 222 225 226 223 215 204 189 173 164 255 40 to 50 230 233 234 230 223 213 199 183 174 263 % of Height 50 to 60 233 236 237 234 226 216 202 187 178 267 60 to 70 233 236 237 234 226 216 202 187 178 267 70 to 80 235 237 239 235 228 217 204 189 180 268 80 to 90 234 236 238 234 227 216 203 188 179 267 90 to 100 233 236 237 234 226 216 202 187 178 267 . Circ Weld 312 315 316 313 305 295 282 269 261 Lower Plate 35 to 40 0 to 10. 308 310 312 308 301 290 276 260 251 269 % of Height 10 to 20 293 296 297 293 286 274 258 242 232 253 20 to 30 254 257 259 254 245 231 216 201 193 219 30 to 40 148 149 150 148 144 139 134 130 128 129 Notet To determine the reference tamperature, en inittel temperature of 30F for plates eno 10F for welds must be added to these mean reference temperatures. l

L, o Table 3 PTS Fracture Mechanics Results 1990 Peak Fluence Peak Reference Conditional Failure ('x 1E+19 n/cm2) Temperature (F) Probability Upper Plate 2.14 269 5.60E-04 . Lower Plate 1.9 342 4.00E-06 L Circ. Weld .1.9 326 1.30E-04 Upper Axial Weld 0.88 278 5.80E-03 Lower Axial Weld 0.88 279 2.00E-05 6.51E-03 M- = _}}