ML20077C406
| ML20077C406 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/31/1994 |
| From: | Asgari M, Kam F K, Mcgarry E, Mike Williams LOUISIANA STATE UNIV., BATON ROUGE, LA, NATIONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERL, OAK RIDGE NATIONAL LABORATORY |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-W-6164 NUREG-CR-6206, ORNL-TM-12693, NUDOCS 9412020031 | |
| Download: ML20077C406 (74) | |
Text
~
NUREG/CR-6206 ORNL/TM-12693 Transport Calculations 0:?
Raciation Exposure to Vesse Su a oort Structures in the Trojan Reactor e si ri
- 1. L Williams, E II. K. Kam, E. D. McGarry Oak Ridge National Laboratory Prepared for U.S. Nuclear Regulatory Commission pbf$DO 05 0 344 p
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NUREG/CR-6206 ORNL/TM-12693 e-Transport Ca cu:ations of 4
Radiation Exposure to Vesse:.
Support Structures in t:ae Trolan Reactor l
l'repared by J
M. Asgari, M. L Williams, E IL K. Kam, E. D. McGany i
Oak Ridge National Laboratory Prepared for U.S. Nuclear Regulatory Commission I
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u-
ORNUTM-12693 Transport Calculations of Radiation Exposure to Vessel Support Structures in the Trojan Reactor l
Manuscript Completed: May 1994 Date Published: July 1994 Prepared by M. Asgari, M. L Williams 1, E H. K. Kam, E. D. McGarry2 Oak Ridge National Laboratory Managed by Martin Marietta Energy Systems, Inc.
Oak Ridge National Laboratory Oak Ridge, TN 37831-6050 Prepared for Division of Engineering Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC FIN W6164 Under Contract No. DE-AC05-840R21400
' Louisiana State University i
Nuclear Science Center Haton Rouge, LA 70803 rNational Institute of Standards and Technology BIdg. 235-A155 Gaithersburg, MD 20899
ABS'IRACT Neutron transport calculations and cavity dosimetry measurements were performed at the Trojan Reactor in order to determine the neutron environment within the ex-vessel region of the plant that contains the vessel support structure. The radiation embrittlement of a steel member that has a relative high degree of stress is a primary concern.
Comparison of transport calculations of the dosimeter activities with the experimental measurements shows that the values obtained with ENDF/B-VI cross section data overestimate the measured results for high-energy-threshold reactions in the cavity by up to 41%, and thermal reactions by up to a factor of 3.0. The transport calculations performed with the original SAILOR cross-section library (based on ENDF/B-IV data) overestimate measured threshold reactions by only 15% and the thermal reactions by about a factor of 2.50. These results are inconsistent with those obtained in earlier studies that compared transport calculations done with SAILOR vs ENDF/B-VI, which indicate that SAILOR tends to underestimate cavity dosimeter activities for threshold reactions, while the ENDF/B-VI values usually agree better with experimental results. One factor that probably contributes to the rather large discrepancy between the computed and measured activities is the core power distribution used in the transport calculations. Because of unavailability of plant-specific data, a generic power distribution provided by Westinghouse was used. Since the calculated cavity flux levels appear to be over-estimated, the results estimated for the exposure to the support structure should be conservative.
The ENDF/B-VI calculations give a fast fluence rate at the critical stress point equal to 6.90E+07 neutrons per em: per second. The thermal-to-fast-flux ratio at this point is computed to be equal to 46. The DPA rate at the critical point is calculated to equal 2.0E-13 displacements per atom per second.
1 iii
CONTENTS l
l ABSTRACT..................................................................
iii LIST O F FIG URES........................................................... vi LIST O F TAB LES.............................................................
vii ACKNOWLEDGMENTS.....
xi 1
1.
INTROD UCTION........................................................
1 2.
DESCRIPTION OF TROJAN PLANT AND EXPERIMENT CONFIGURATION........... 3 3.
DETAILS OF TRANSPORT CALCULATIONS...................................
5 4.
DESCRIPTION OF EXPERIMENTAL MEASUREMENTS.
13 l
l 5.
RESULTS..
15 1
6.
CONCLUSIONS.
52 l
7.
REFERENCES..
54 APPENDIX A. EXPERIMENTAL MEASUREMENTS OF DOSIMETER ACTIVITIES FOR TROJAN CYCLE 13....................................................
56 APPENDIX B. IMPACT OF 1/4 VS.1/8 CORE MODEL USED IN TRANSPORT l
CALCULATIONS 59 Y
l I
t I
i l
LIST OF FIGURES l
f.agg 2
l
- 1. Vessel support structure in the Trojan PWR...........................................
l 4
f
- 2. Ex-vessel geometry for Trojan...
6
- 3. DOT R-Theta model of Trojan used in transport calculations 7
- 4. DOT R-Z model of Trojan used in transport calculations..
8
- 5. DOT 1.D model of Trojan used in transport calculations................................
12 6.
Relative axial power distribution for Trojan Cycle 13....
si
54-A-LIST OF TA3LES
- f. ate 1.
Material compositions used in
- DOT' calculations................................... 9 2.
Assemblywise BOC, EOC burnup (MWD /r), and absolute power (MW*)
Trojan Cycle 13...
11 3.
Neutron dosimeters used in the cavity experiment
. 14 4.
Experimental results of Trojan cavity dosimetry measurements..............
. 14 5.
Absolute flux spectrum obtained with ENDF/B.VI Fe-O.H at the center of 31.5 surveillance capsule at 20.3 cm below the core midplane.......
16 6.
Absolute flux spectrum calculated with ENDF/B-VI Fe.O-H at the center of the 34.0
- surveillance capsule at 20.3 cm below the core midplane................... 17 7.
Absolute flux spectrum calculated with ORIGINAL SAILOR at the center of the 31.5' surveillance capsule at 20.3 cm below the core midplane.....
18 8.
Absolute flux spectrum calculated with ORIGINAL SAILOR at the center of the 34.0* surveillance capsule at 20.3 cm below the core midplane................... 19 9.
Calculated.aturated activities at the center of Trojan 31.5* surveillance capsule at the core midplane for Cycle 13.....................
20 10.
Calculated saturated activities at the center of Trojan 34.0* surveillance capsule at the core midplane for Cycle 13.................................. 21 11.
Spectrum averaged cross section at the center of the 31.5' surveillance capsule at the core midplane for Cycle 13................................... 22 12.
Spectrum averaged cross section at the center of the 34.0* surveillance capsule at the core midplane for Cycle 13................................... 23 13.
Comparison of calculated absolute saturated reaction rates for Trojan Cycle 13 with experimental and Westinghouse calculated results of Capsule X (removed April 1984)....................................... 24 14.
Calculated relative azimuthal variation of flux >1 MeV at four radial loca tions for Trojan Cycle 13.......................................... 25 i
15.
Calculated relative axial variation of flux >1 MeV at four radial locations for Trojan Cycle 13............................................ 27 vii
LIST OF TABLES (continued)
Radial variation of integrated parameters along RPV, calculated with 16.
29 ENDF/B-VI Fe-O-H at O' (20.3 cm below midplanc), for Trojan Cycle 13 Radial variation of integrated parameters along RPV, calculated with 17.
29 ENDF/B VI Fe-0.H at 45' (20.3 cm below midplanc), for Trojan Cycle 13 Radial variation of integrated parameters along RPV, calculated with 18.
29 ORIGINAL SAILOR Fe-O-H at 0* (20.3 cm below midplane), for Trojan Cycle 13..
Radial variation of integrated parameters along RPV, calculated with 19.
30 ORIGINAL SAILOR at 45' (20.3 cm below midplane), for Trojan Cycle 13........
Radial variation of cumulative fluence (n.cm'2) and DPA along RPV, 20.
30 calculated with ENDF/B-VI Fe-O-H at O*, at near core midplane for Trojan Cycle 13.
RPV peak cumulative fluence (n.cm.2) and DPA, calculated with ENDF/B-VI 21.
30 Fe-O-H for Trojan Cycle 13 Absolute flux spectrum obtained with ENDF/B-VI Fe-O-H, behind RPV 22.
........... 33 insulation at O*, R = 257.0 cm, and 20.3 cm below the core midplane Absolute flux spectrum obtained with ENDF/B-VI Fe-O H at the center of the 23.
34 instrumentation tube, R = 321.63 cm, O' and 20.3 cm below the core midplane.....
Absolute flux spectrum obtained with ORIGINAL SAILOR behind the RPV 24.
35 insulation at 0
- dosimeters, R = 257.0 cm and 20.3 cm below the core midplane.....
Absolute flux spectrum calculated with ORIGINAL SAILOR, at the center of 25.
36 the instrumentation tube, R = 321.63 cm,0 below the core midplanc.
Calculated saturated reaction rates at the 0* dosimeters k> cation,5 cm 26.
behind insulation (R = 257.0 cm), at a height 20.3 cm below the core midplane 37 for Cycle 13 Calculated saturated reaction rates at the dosimeters k> cation, center of the 27.
38 instrument tube, at a height 20.3 cm below the core midplane for Cycle 13 Absolute flux spectrum calculated with ENDF/B-VI Fe-O H, behind RPV 28.
insulation at 0*, R = 257.0 cm, and 121.19 cm above the core midplanc......... 39 Absolute flux spectrum calculated with ENDF/B-VI Fe-O-H, at the center of 29.
40 the instrumentation tube, R = 321.63 cm,0*, and 121.9 cm above the core midplane viii
LIST OF TABLES (continued) 30.
Absolute flux spectrum calculated with ORIGINAL SAILOR, behind RPV insulation at 0 *, R = 257.0 cm, and 121.9 cm above the core midplane.................... 41 i
31.
Absolute flux spectrum calculated with ORIGINAL SAILOR at the center of the instrumentation tube, R = 321.63 cm, O*, and 121.9 cm above the core midplane.. 42 32.
Calculated saturated reaction rates at the 0 cavity dosimeters location,5 cm behind insulation (R = 257.0 cm), at a height 121.90 cm above the core midplane l
fo r Cyc le 13........................................................ 43 I
33.
Calculated saturated reaction rates at the dcsimeters location, center of the instrumentation tube, at a height 121.90 cm above the core midplane for Cycle 13.... 44 I
34.
Spectrum averaged cross section at the cavity dosimeter location,5 cm behind the RPV insulation, at O ' of Trojan Cycle 13................................ 45 l
35.
Spectrum averaged cross section at the center of the instrumentation tube at 0-and 20.3 cm below midplane of Trojan Cycle 13............................. 46 36.
Non-saturation factors (h) for Trojan Cycle 47 37.
C/E values for Trojan dosimetry, based on ENDF/B-VI calculations............... 48 38.
C/E values for Trojan dosimetry, based on original SAILOR calculations 48 39.
Absolute flux spectrum calculated with ENDF/B VI Fe-O-H at location of support structure (i.e., e = 23 *,20.3 cm below core midplane,17.8 cm into the concrete).... 49 40.
Calculated saturated reaction rates at the location of the support structure (i.e., e = 23',20.3 cm below core midplane,18.0 cm into the concrete) 50 41.
Calculated results of neutron exposure at Trojan support structure location (i.e., e = 23 *,20.3 cm below midplanc) 51 A.1 NIST analyses of radiometric & SSTR dosimetry data measured by Westinghouse d uring Troj an Cycle 13.................................................. 58 B.1 Ratio of reflected 1/8 core results to non-symmetrical 1/4 core model.............. 59 ix
~
ACKNOWLEDGMENIS The authors wish to thank IIarish Manohara, a nuclear engineering graduate student at Imuisiana State University, for performing the transport calculations described in Appendix B. Thanks also to Pat Hileman for preparing this report for publication and to C. II. Shappert for cditing. We are also grateful to Arnie Fero, Parvin Lippincott, and Stan Anderson of Westinghouse Electric for providing important data used to model the Trojan reactor and for their numerous helpful comments. Finally, the authors gratefully acknowledge the programmatic support, encouragement, and comments from Al Taboada of the Nuclear Regulatory Commission.
i I
Xi 1
l 1 INTRODUCrlON The embrittlement of steel in reactor pressure vessels (RPVs) due to exposure to fast neutrons is well known and is constantly monitored during the operating lifetime of power reactors. In the past, the radiation embrittlement of ex-vessel steel components has been of less concern because 1
(1) the fast neutron levels beyond the RPV are much lower compared with levels at vessel welds and (2) most of the critical load-bearing members are far removed from neutron exposure.
However, a particular Westinghouse PWR design includes a portion of the vessel support i
structure embedded within the concrete shield surrounding the reactor cavity, as shown in Fig.1.
Note that the steel structure supporting the vessel consists of two vertical members connected by a short horizontal member near the core midplanc elevation. The vertical members are subjected to compressive forces and present no structural concerns due to embrittlement. Itowever, th.
horizontal beam is under stress from a bending moment and could be susceptible to loss of ductility. The maximum stress occurs at approximately 18 cm from the surface of the concrete shield wall around the cavity. Earlier analyses have indicated that the fast neutron (i.e., those with energies above 1 MeV) Oux present at the vessel support structure is well within the i
acceptable range that will not exceed allowable fluence limits during the reactor lifetime. 'nie fast neutron flux is low at these locations because the concrete surrounding the structure thermaliz,:s the high energy flux, thus lowering the fast neutron intensity while enhancing the thermal spectrum component. However, knowledge of the ratio of the thermtI to fast flux at the maximum stress location is of prime interest in this plant to obtain an stimate of the overall radiation embrittlement due to neutrons of all cr.ergies, including therr tal neutrons.'
The purpose of this study is to compute the neutron flux spectrum and nagnitude in the cavity and within the concrete shield surrounding the casity of the Trojan PWit, in order to provide an estimate for the neutron environment in the vicinity of the vessel support structure. An experimental program was also performed in conjui'ction with the calculational study to de,termin:
measured dosimetry reaction rates to verify the results from the transport calculations. This report summarizes results from the transport calculations and the experim:ntal measurements anc compares the computed dosimeter activities with the experimental values.
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2 DESCRWTION OF TROJAN PLANT AND EXPERIMENT CONFIGURATION Re Trojan Reactor is a Westinghouse PWR of the "three loop" design type that produces 3411 MW (thermal) of power. The plant has been operated by Portland General Electric Company since going critical in 1976. Because the same vessel support design as included in Trojan is also present in several other Westinghouse reactors that continue to or.erate in the United States, and because the experiment was already completed for this plant, this research project is being completed even though the reactor operation is being terminated.
Experimental dc ' metry was placed in the reactor cavity of the plant by Oak Ridge National Laboratory (ORNL) during Cycle 13, for the time period July 13,1990, to March 4,1991. A combination of radiometric foils, solid-state track recorders (SSTRs), and bead chain gradient dosimeters were utilized (See Section 4). During this cycle the fuel was loaded in a " low-leakage" pattern, which reduced the fluence at the vessel and beyond, compared with some of the earlier cycles. Cycle 7 was the first to use a low-leakage core configuration in the Trojan core.
However, the core loading for cycle 13 is somewhat unusual because highly burned assemblics are placed in the peripheral location, while fresh fuel is placed next to the periphal assemblies. This causes a steep gradient in the pinwise power distribution in the row of assemblies. As part of the routine RPV surveillance program, Westinghouse has analyzed plant dosimetry for some earlier cycles? but unfortunately, no in-vessel surveillance dosimetry was scheduled to be removed from the plant during this particular cycle. Thus only experimental results from beyond the vessel are available to validate the current transport calculations.
Figure 2 shows the ex-vessel geometry of interest in this project. As is typical for this reactor j
design, there are two large " detector wells" in the cavity at 0 and 45 *,respectively, that indent the concrete shield surrounding the cavity, extending nearly the entire core height. These detector wells will impact the thermal to fast flux ratio in this vicinity. Note also that there is an ex-core detector, normally used to monitor the reactor core, located in the cavity region at an azimuthal position of approximately 0 *. The cylindrical detector contains a larg: polyethylene segment (13.66 cm OD) to thermalize neutrons for detection, which significantly perturbs the flux in this region. The detector is movable into the recessed detector well within the concrete, as shown in Figure 2. A vertical instrument tube (5.08 cm OD) is positioned above the recessed compartment for access to the ex-core detector. The instrument tube channel through the concrete provided access for two sets of dosimeters that were placed in the cavity. De various activation foils and SSTRs are inserted in the instrument tube at distances of 20.3 cm below and 121.9 cm above the core midplane, respectively, at a radius of 321.6 cm from the core center. Unfortunately, this radial location in the cavity places the dosimeters behind the polyethylene segment of the ex-core detector with about 20 cm separating the dosimeters and the detector, making the geometry for the transport calculations more difficult to model accurately. Two more similar sets of dosimeters at the same axial positions were also placed in the cavity about 5 cm behind the thermal insulation that surrounds the RPV. These dosimeters are located at an azimuthal angle near 0*,
between the pressure vessel insulation and the ex-core detector. In addition to the dosimeter foils and SSTRs hung at 0*,two bead chain flux-gradient dosimeters were placed in the cavity at 0*
and 45 *,respectively.
3
I h
WATERTIGHT TO NUCLEAR
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ROLLER ASSEMBLY (21' TRAVEL) 1 Figure 2 Ex-vessel geometry for Trojan 4
4 3 DETAILS OF TRANSPORT CA1CULATIONS The methodology previously established at ORNL for RPV Duence analysis was used in this study. A three-dimensional distribution of the multigroup Dux in cylindrical geometry (R-Theta-Z) is computed by synthesizing Duxes determined from discrete-ordinates calculations in
+
two-dimensional R-Theta and R-Z and one-dimensional R coordinates, respectively.' The transport calculations were performed with the DOT 4.2 code using an Sa quadrature and a P3 Ixgendre expansion of the scattering cross sections. The multigroup structure was that of the 47 group SAILOR library.5 Most of the nuclear cross-section data were taken directly from the SAILOR library. This library 6
is collapsed from the original VITAMIN-C library that was processed from ENDF/B-IV.
However, two different data sources were used for the cross sections of hydrogen, oxygen, and iron. One set of transport calculations used the standard set of cross-section data contained in SAILOR for thcae three materials. A second set of calculations was based on new multigroup data for H-O-Fe that was recently processed from ENDF/B-VI evaluations.7 The newer i
ENDF/B-VI cross sections have been found to improve agreement between transport calculations
.l and measured dosimetry activities for cavity locations in some earlier studies because of an increase in the transmission of fast neutrons through the RPV."
Due to the symmetrical fuci loading pattern in the core, only one-eighth of the core is typically considered for the transport calculation. The one-eighth-core model corresponds to a 45
- slice with reflected boundary conditions at 45
- and 0 *, However, in this reactor design, the ex-core geometry is not exactly symmetrical by octant because the neutron pad is not symmetrically oriented about 45 *;it extends from 25' to 57.5 *. Hence, a quarter-core model is more appropriate for these types of reactors. The impact of using a quarter-core vs an eighth-core reactor model was examined, and it was found that the computed dosimeter activities at O' are not affected; the values at 45 *, however, differ by up to 12% The results of this comparison are summarized in Appendix B. The dosimeter results presented in the text were obtained from the one-eighth core calculation.
Figure 3 represents an octal section of the reactor at the horizontal midplane from the center of the core to beyond the biological concrete shield wall. The cross-sectional representation of the surveillance capsules, ex-core detector, instrument tube, and void indention in the concrete wall are included in the DOT-IV R-Theta calculation to account for the significant Dux perturbation effects from the presence of these components. The ex-core detector was modeled approximately based on information obtained from the manufacturer. Note that the e = 0* coordinate in the DOT model actually corresponds to the 90* azimuth in the actual reactor configuration, but the i
45' coordinates are consistent. The R-Z model, Figure 4. used in the R-Z calculation represents the portion of the reactor from the center of the core to beyond the shield wall radially, and from below the bottom of the active fuel up to above top of the active fuel axially. Figure 5 represents the 1-D model used in one-dimensional calculation, extending from the center of the reactor core to beyond the shield wall. Table 1 gives the material compositions used in all the DOT-IV calculations.
The calculation of neutron source for all DOT-IV runs were performed using the DOTSOR code.' This code generates the R-Theta and R-Z source for DOT-IV transport calculations from 5
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'1
a Table 1 Material compositions used in 'DCT' calculations MATERIAL CORE (WFSTINGHOUSE *)
U-235 = 237E-4 U-238 = 6.73E-3 O (FUEL) = 1393E-2 O(H O) = 1393E-2 H = 2.785E-2 Fe = 1.069E-4 2
Mn = 1.976E-6 Cs = 6.162E-5 Ni = 1.291e-4 Zr = 4348e-3 EXCORE WATER (WESTINGHOUSE *)
H = 5.007E-2 O = 2 503E-2 INCORE WATER H = 4.782E-2 O = 2391E-2 SS-304 (WESTINGHOUSE *)
Fe = 5.978E-2 Mn = 1.761E-3 Cr = 1_768E-2 Ni = 8.242E-3 A533B STEEL (WESTINGHOUSE")
C = 9.820E-4 Mn = 1.116E-3 Fe = 8.270E-2 Ni = 4.420E-4 POLYETHYLENE Density of.%3 G/CC C = 4.134E-2 H = 6.612E-2 INSULATION SS-304 @ 3% DENSITY CONCRETE TYPE 02-B H = 6.873E-3 O = 4329E-2 C = 1.153E-4 Mg = 1.239E-4 A1 = 1.741E-3 Si = 1.662E-2 K = 4.606E-4 Ca = 1.503E-3 Fe = 3.451E-4 Na = 9.641E-4
- Personal communication from E. P. Lippincott 9
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the given X-Y core power distribution. The X-Y core power distribution was determined from the given assemblywise BOC and EOC burnup data combined with the relative pinwise power distribution. Table 2 represents the assemblywise BOC, and EOC burnup, and the absolute power corresponding to the cyc:e averaged distribution normalized to the full power (3411 MW). The assembly burnup data were plant-specific values for Cycle 13, provided by Westinghouse."
Ilowever, the pinwise power distribution provided by Westinghouse was " generic," because the plant-specific pinwise power was not readily available. Because of the unusually severe gradiant across the peripheral assemblies in this cycle, the generic pinwise power distribution will tend to cause the calculated dosimeter activities to be over estimated. In the R-Z calculation, the cycle-average relative axial power distribution, as shown in Figure 6, was used. The 1-D source is obtained by integrating the R-Z source over Z.
The activation cross sections for computing the dosimeter reactions were obtained from ENDF/B-VL The fission spectrum for the core source was represented by a weighted combination of the U " and 2
2 the Pu " spectra based on an ENDF/B-V Watt distribution. The relative contribution of the two isotopes was determined by the fractional contribution of each to the fission source in the peripheral assemblies, at the mid-cycle burnup of these assemblies.
One difliculty with using the SAILOR cross-section library in this study is that the SAILOR group stmeture includes only two groups below 0.414 eV; furthermore, no upscattering within the thermal energy range is accounted for in the scatter matrix of the SAILOR cross sections. Historically, virtually all RPV fluence studies have reported difliculty in estimating dosimeter reactions such as Co" capture that are sensitive to the thermal energy range, probably a result of the rather crude treatment of the thermal scattering in the commonly used libraries (e.g., S AILOR and ELXSIR"').
Often discrepancies of 2 or more are observed between the measured and computed activities.
< infortunately, no other appropriate library with multiple thermal groups was available at the time of shis study. Therefore, there is some uncertainty about the accuracy of the computed thermal flux that must be taken into account when assessing the reported results. The measured activities from the thermal neutron dosimetry can be used to improve the computed thermal flux estimate.
Personal Communications, E. P. Lippincott, Westinghouse Corporation 10
Table 2 Assemblywise BOC, EOC burnup (MWD /T), and absolute power (MW*) for Trojan Cycle 13" 17930.0 16084.0 13275.0 24670 0 9187.0 27890.0 0.0 33080.0 26223.0 25280.0 22882.0 33528.0 19483.0 36042.0 10115.0 36773.0 4.575 10.146 10.599 9.773 11.359 8.994 11.160 4.074 16082.0 8149.0 19147.0 9311.0 9187.0 28271.0 0.0 28588.0 25255.0 18233.0 28123.0 19582.0 19541.0 36401.0 10118.0 32438.0 10.120 22.251 19.806 22.664 22.847 17.939 22.326 8.495 13335.0 19483.0 20252.0 19738.0 9065.0 25190.0.
0.0 28944.0 22728.0 28400.0 28644.0 28782.0 19064.0 33454.0 10028.0 332749.0 10.363 19.676 18.517 19.956 22.063 18.235 22.127 8.396 24671.0 9306.0 20270.0 8161.0 20428.0 23181.0 0.0 27274.0 33634.0 19588.0 29240.0 18058.0 28889.0 31402.0 8935.0 29898.0 9.889 22.688 19.793 21.838 18.670 18.140 19.716 5.790 9158.0 9179.0 9083.0 20483.0 28916.0 0.0 20280.0 19473.0 19518.0 19038.0 28834.0 36880.0 10309.0 25522.0 11.380 22.814 21.966 18.427 17.573 22.747 11.567 27869.0 28257.0 26584.0 22174.0 0.0 0.0 34208.0 38035.0 36388.0 34680.0 30470.0 10320.0 8892.0 36872.0 11.216 17 942 17.864 18.306 22.772 19.621 5.878 0.0 0.0 0.0 0.0 20478.0 34173.0 10134.0 10141.0 10011.0 8830.0 25708.0 36834.0 11.181
'!2.377 22.090 19.484 11.540 5.872 33505.0 29163.0 29125.0 27349.0 37184.0 33178.0 32938.0 29986.0 4.059 8.859 8.414 5.819 BOC BURNUP EOC BURNUP ABS POWER 6 "Only one-eighth of core is used in R-O matel.
- Assembly power corresponds to cycle-averaged distribution in MW, normalized to full power (3411 MW).
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a k 4 DFSCRWDON OF EXPERIMENTAL MEASUREMENTS A dosimetry experiment based on several different threshold reactions was placed in the reactor cavity of the TROJAN plant during Cycle 13. The list of the different dosimeters used in this experiment, along with the type of reaction, half-life, and threshold energy, is given in Table 3. Four sets of these dosimeters were placed at different locations in the reactor cavity. Two sets of experimental measurements were made in the instrument channel located behind one of the ex-vessel detectors near 0* azimuth. Two other sets of experimental measurements were taken behind the RPV insulation. The coordinates of these four sets are as follows (coordinates are referenced to coordinate system used in DOT transport calculations):
- 1. CAPSULE A (bchind RPV insulation):
R = 257 cm, e = 0", Z = 121.9 cm above midplane
- 2. CAPSULE B (bchind RPV insulation):
R = 257 cm, e = 0", Z = 20.3 cm below midplane
- 3. CAPSULE C (instrument tube):
R = 321.63 cm, e = 0", Z = 121.9 cm above midplanc
- 4. CAPSULE D (instrument tube):
R = 321.63 cm, e = 0", Z = 20.3 below midplanc i In addition to the above radiometric foils and SSTR dosimeters, two gradient bead dosimeters were placed in the cavity at 0* and 45 *, respectively. These results have not been considered in i this study. The experimental measurements were performed by Westinghouse" and reviewed by i NIST, who made some corrections in the SSTR fission dosimetry results to account for "U 2 impurities. The saturated activities obtained for the various dosimeters at the four cavity k) cations are summarized in Table 4. More details of the dosimeter measurement results are given in Appendix A. i 13 l \\
Table 3 Neutron dosimeters used in the cavity experiment Threshold Dosimeter Reaction Half-life energy (MeV) type' "Fe(n,p) "Mn 312 days 2.20 RM "Ni (n,p) 58Co 71 days 1.60 RM "Cu(n,p) "Co 5.27 years 5.00 RM "Ti (n,p) "Sc 84.0 days 3.90 RM SU (n,f) "7Cs 30.1 years 1.50 SSTR,RM 0.20 SSTR mNp(n,f) "Co(n, ) "Co 5.27 years 'Ihermal RM 'RM = radiometric foil; SSTR = solid state track recorder. Table 4 Experimental results of Trojan cavity dosimetty measurements Saturated Activities (Reaction Per Atom Per Second) CAPSULE Reaction A B C D "Cu (n, ) 2.42E-19 3.20E-19 1.62E-19 2.11E 19 Mi (n,p) 3.44E-18 4.60E-18 2.35E-18 3.04E-18 "Fe (n,p) 1.56E-17 2.07E-17 1.06E-17 1.35E-17 "Ni (n,p) 231E-17 3.10E-17 1.55E-17 2.01E-17 SU (n f) 8.24E-17 1.08E-16 5.94E-17 7.29E-17 SNp (n.f) 1.16E-15 1.51E-15 1.03E-15 1.02E-15 "Co (n, ), Bare 1.95E-14 3.2SE-14 2.05E-14 2.86E-14 "Co (n, ), Cd 1.12E-15 1.49E-14 9.66E-15 1.25E-i4
- U (n,f), Bare 1.11E-13 2.12E-13 1.54E-13 2.68E-13 SU (n,f), Cd 3.29E-14 2.90E-14 2.61E-14 3.56E-14 14
5 RESULTS Neutronic transport calculations were performed with both the ORIGINAL SAILOR and the MODIFIED SAILOR library which contains ENDF/B-VI Fe-O-H data. The syrthesized absolute fluxes were obtained for both sets of calculations. Based on these calculations, Tables 5 and 6 show the absolute flux spectra at the center of the 31.5 and 34.0* smveillance capsules, respectively, obtained with MODIFIED SAILOR library. Tables 7 and 8 show the same quantity obtained with ORIGINAL SAILOR library. At these locations the ENDF/B-VI Fe-O-H increases the ORIGINAL SAILOR value of the flux above 1.0 MeV by about 10% Tables 9 through 12 show the calculated saturated activities and the spectrum-averaged cross section at the surveillance locations obtained with both libraries. No surveillance capsule measurements were performed for Cycle 13, and hence no in-vessel measurements to which the transport calculations could be compared for this cycle. However, Table 13 shows the comparison of the calculated dosimeter activities for Cycle 13 with the capsule X experimental results from several cycles back, removed Apnl 1984.2 It can be seen that Cycle 13 calculated values obtained by LSU with both libraries underestimate the capsule X experimental results. The difference in the power distributions of the two different dosimeter activation periods could cause the discrepancy. The aux above 1.0 MeV in the RPV of TROJAN Reactor was found to peak at an azimuthal angle of about 45 *. Table 14 illustrates the azimuthal variation of the flux above 1.0 MeV at the RPV OT,3/4T positions, and at the radial locations of the wire dosimeters and center of the instrumentation tube, based on the ENDF/B-VI Fe-O.H calculations. A factor of 1.87 maximum variation is present in the azimuthal dependence of the flux above 1.0 MeV at OT of the RPV. This factor has a value of 1.20 at the radial location of the center of the instrumentation tube. From the R-Z transport calculations, based on the cycle-averaged axial power shape, the flux above 1.0 MeV is found to peak at the core midplane. Table 15 illustrates the axial variation of the flux above 1.0 MeV. As shown, the flux above 1.0 MeV is almost constant within 38.1 cm about the core midplane for the cycle-averaged power shape. Tables 16 and 17 show the radial variation ofintegrated flux parameters through the RPV at O' and 45*, calculated with ENDF/B-VI Fe-O-H, respectively. The same quantities at the same locations calculated with the ORIGINAL SAILOR are shown in Tables 18 and 19. The flux above 1.0 MeV incident on the vessel wetted surface at the peak azimuthal and axial location calculated with ENDF/B-VI Fe-O-H is about 1.868E+10 cm.2,3-i. 'ni s value is only about 10% higher than the one calculated with ORIGINAL SAILOR library. Tables 20 and 21 show the radial variation of the cumulative fluence (cm-2) and DPA through the RPV at 0* and at the peak azimuthal location, respectively, calculated with ENDF/B-VI Fe-O-H. Based on the calculated peak flux above 1.0 McV, the fast fluence due to the 211.25 effective fuh-power days of Cycle 13 at the vessel wetted surface and RPV OT are 3.410E+17 and 3.314E+17 cm.2, 15
1 4 Table 5 Absolute flux spectrum obtained with ENDF/B-VI Fe-O-H at the center of the 21.5* surveillance capsule at 20.3 cm below the core midplane Upper Group Cumulative Group Cumulative Cumulative Group energy flux flux DPA DPA DPA fraction 1 1.733E+ 01 8.523E+06 8.523E+06 2.490E-14 2.490E-14 1380E-04 2 1.419E+01 3.487E+ 07 - 4339E+07 9.225E-14 1.172E-13 7.433E 04 3 1.221E+01 1.410E+ 08 1.844E+08 3398E-13 4.569E-13 2.899E 03 4 1.000E+01 2.933E+ 08 4.777E+08 6.508E-13 1.108E-12 7.028E-03 5 8.607E+ 00 5.150E+08 9.927E+ 08 1.075E-12 2.183E-12 1385E-12 6 7.408E +00 1.234E+ 09 2.227E + 09 2398E-12 4381E-12 2.906E-02 7 6.065E+00 1.882E+09 4.109E +09 3357E-12 7.937E-12 5.036E-02 8 4.966E+00 3.890E+09 7.999E+09 6.112E 12 1.405E-11 8.913E-02 9 3.679E+00 3.259E+ 09 1.126E+ 10 4.465E-12 1.851E-11 1.175E-01 10 3.012E+00 2.588E+09 1385E+ 10 3.289E-12 2.180E-11 1383E-01 11 2.725E +00 3.172E+09 1.702E+ 10 4.045E-12 2.585E-11 1.640E-01 12 2.466E+ 00 1.578E+09 1.860E+ 10 1.849E-12 ' 2.770E-11 1.757E-01 13 2365E+00 4.605E+08 1.906E+ 10 5.0475-13 2.820E-11 1.789E-01 14 2346E+00 2319E+09 2.137E+ 10 2.414 E-12 3.062E-11 1.942E-01 15 2.231E + 00 6.747E+09 2.812E+ 10 6.977E-12 3.759E-11 2385E-01 16 1.920E+ 00 8.744E+09 3.687E+ 10 7.112E-12 4.470E-11 '"46E-01 17 1.653E+ 00 1355E+ 10 5.041E+ 10 1.098E-11 5.568E-11 .v32E-01 18 1353E+00 3.053E+ 10 8.095E+ 10 1.711E-11 7.279E-11 4.618E 01 19 1.003d+00 2314E+ 10 1.041E+ 11 8.482E-12 8.127E-11 5.156E.01 20 8.208E-01 1.091E+ 10 1.150E+ 11 6.121E-12 8.739E-11 5344E.01 21 7.427E-01 3.749E + 10 1.525E+ 11 1355E-11 1.009E-10 6.404E-01 22 6.081E-01 3308E+10 1.356E + 11 9.736E-12 1.107E 10 7.021E.01 23 4.979E-01 3.617E+ 10 2.217E+ 11 1.434E-11 1.250E-10 7.931E.01 24 3.688E 01 3388E+ 10 2.576E + 11 7.420E-12 1324E-10 8.402E-01 25 2.972E-01 4.847E+ 10 3.061E+ 11 9.718E-12 1.421E-10 9.018E-01 26 1.832E-01 4.299E+ 10 3.491E+ 11 6.061E-12 1.482E-10 9.403E-01 27 1.111E-01 3.219E+ 10 3.813E + 11 4.162E-12 1.524E-10 9.667E-01 28 6.738E 02 2.706E+10 4.083E+ 11 1.753E-12 1.541E-10 9.778E 01 29 4.087E 02 9326E+09 4.177E+ 11 7.548E-13 1.549E-10 9.826E-01 30 .i.183E-02 3.824E+ 09 4.215E + 11 1.087E-12 1.560E-10 9.895E 01 31 2.606E-02 1.078E+10 4323E+ 11 2.174E-13 1.562E-10 9.909E-01 32 2.418E-02 6.666E+ 09 4389E+ 11 2.919E-14 1.562E-10 9.911E-01 33 2.188E-02 1.272E+ 10 4.516E+ 11 1A11E-13 1.563E-10 9.917E.01 34 1.503E-02 2.481E+ 10 4.765E+ 11 4.654E-13 1.568E-10 9.947E 01 35 7.102E-03 3.027E+ 10 5.067E+ 11 2.693E-13 1.571E-10 9.964E-01 36 3355E-03 2.748E + 10 5342E+ 11 9.674E-14 1.571E-10 9.970E-01 37 1.585E.03 4.514E+ 10 5.793E+ 11 7.592E-14 1372E-10 9.975E-01 38 4.540"-04 2.072E + 10 6.001E+ : 1 2.042E-15 1.572E-10 9.975E 01 39 2.144E-04 2.512E + 10 6.252E+ 1 3.602E 15 1.572E-10 9.975E-01 l 40 1.013E.04 3374E+ 10 6589E+.1 7.526E-15 1.572E-10 9.976E 01 l 41 3.727E-0*. 4.015E+ 10 6.991E+ 11 1.582E-14 1.573E-10 9.977E-01 42 1068E-05 2.190E+ 10 7.210E + 11 1.408E-14 1.573E 10 9.977E-01 43 5.043E-06 2.547E + 10 7.464E+ 11 2347E-14 1373E-10 9.979E 01 44 1.855E-06 1.535E+ 10 7.618E + 11 2360E-14 1.573E-10 9.981E 01 45 8.764E-07 1.216E + 10 7.739E+ 11 2.728E-14 1.573E 10. 9.982E-01 46 4.140E-07 7363E+09 7.813E+ 11 2.885E-14 1.574E-10 9.984E-01 47 1.000E-07 2.685E+ 10 8.082E A 11 2.510E-13 1.576E-10 1.000E+ 00 16
Table 6 Absolute flux spectrum calculated with ENDF/B-VI Fe-O-H at the center of the 34.0* surveillance capsule at 20 3 cm below the core midplane Upper Group Cumuladve Group Cumulauve Cumulative Group energy flux flux DPA DPA DPA fraction 1 1.733E + 01 8.758E+06 8.758E+ 06 2 559E-14 2.559E-14 1390E-04 2 1.419E+01 3.603E+ 07 4.479E+ 07 9.535E-14 1.209E-13 6.569E-04 3 1.221E + 01 1.475E+ 08 1.923E+ 08 3.553E-13 4.763E-13 2387E-03 4 1.000E+ 01 3.087E+ 08 5.010E+ 08 6.850E-13 1.161E-12 6307E-03 5 8.607E +00 5.470E + 08 1.048E+09 1.142E-12 2303E-12 1.251E-02 6 7.408E+ 00 1319E+09 2367E+ 09 2.563E-12 4.866E-12 2.643E-02 7 6.065E+ 00 2.040E+ 09 4.407E+ 09 3.640E-12 8.506E-12 4.620E-02 8 4.906E A 00 4316E+ 09 8.724E+ 09 6.780E-12 1.529E-11 8303E-02 9 3.679E + 00 3.682E+ 09 1.241E+ 10 5.044E-12 2.033E-11 1.104E-01 10 3.012E+ 00 2.927E+09 1.533E+ 10 3.720E-12 2.405E-11 1306E-01 11 2.725E +00 3.603E+ 09 1.894E+ 10 4.594E-12 2.864E-11 1.556E-01 12 2.466E+ 00 1.793E+ 09 2.073E+ 10 2.102E-12 3.075E-11 1.670E-01 13 2365E+00 5.233E+ 08 2.125E + 10 5.736E-13 3.132E-11 1.701E-01 14 2346E+ 00 2.638E+ 09 2389E+ 10 2.746E-12 3.407E-11 1.850E-01 15 2.?31E + 00 7.704E + 09 3.159E+ 10 7.966E-12 4.203E-11 2.283E.01 16 1.920E + 00 1.006E+ 10 4.166E+ 10 8.184E-12 5.022E-11 2.727E-01 17 1.653E+ 00 1.567E+ 10 5.733E + 10 1.270E-11 6.292E-11 3.417E-01 18 1353E+00 3377E+ 10 9310E+ 10 2.004E-11 8.296E-11 4.506E-01 19 1.003E + 00 2.735E + 10 1.204 E + 11 1.002E-11 9.298E-Il 5.050E-01 20 8.208E-01 1.287E+ 10 1333E+ 11 1.163E-10 1.002E-10 5.442E-01 21 7.427E-01 4.467E + 10 1.780E + 11 1.280E-10 1.163E-10 6319E-01 22 6.081E-01 3.956E+ 10 2.176E + 11 1.451E-10 1.280E-10 6.951E 01 23 4.979E-01 4321E+ 10 2.608E + 11 1.540E-10 1.451E-10 7.882E-01 24 3.688E-01 4319E+ 10 3.039E + 11 1.657E-10 1.540E-10 8367E-01 25 2.972E-01 5.794E+ 10 3.619E+ 11 1.729E-10 1.657E-10 8.998E-01 26 1.832E-01 5.148E+ 10 4.134E+ 11 1.779E-10 1.729E-10 9392E 01 27 1.111E 01 3.847E+ 10 4.518E+ 11 1.800E-10 1.779E-10 9.662E-01 28 6.738E-02 3.231E+ 10 4.842E + 11 1.809E-10 1E00E-10 9.776E-01 29 4.087E-02 1.106E+ 10 4.952E+ 11 1.822E-10 1.809E-10 9.824E-01 30 3.183E-02 4.488E+ 09 4.997E + 11 1.824E-10 1.822E-10 9.894E-01 31 2.606E-02 1.288E+ 10 5.126E + 11 1.825E-10 1.824E-10 9.908E-01 32 2.418E-02 8.011E + 09 5.206E + 11 1.826E-10 1.825E-10 9.910E-01 33 2.188E.02 1301E+ 10 5356E+ 11 1331E-10 1.826E-10 9.916E-01 34 1.503E42 2.914E + 10 SI>47E + 11 1E34E-10 1.831E-10 9.946E-01 35 7.102E-03 3.567E + 10 6.004E + 11 1.836E-10 1.834E-10 9.%3E.01 36 3355E-03 3.247E + 10 6329E+ 11 1.836E-10 1E36E-10 9.970E41 37 1.585 E-03 5350E+ 10 6.864E + 11 1.836E-10 1.836E-10 9.974E-01 38 4.540E-04 2.441E + 10 7.108E + 11 1.837E-10 1.836E 10 9.975E-01 39 2.144E-04 2.966E + 10 7.405E + 11 1.837E-10 1.837E-10 9.975E-01 40 1.013E-04 3.995E+ 10 7.804 E+ 11 1.837E-10 1.837E-10 9.975E-01 l 41 3.727E-05 4.756E+ 10 8.7A0E + 11 1.837E 10 1E37E-10 9.976E-01 42 1.068E-05 2.591 E + 10 8.539F + 11 1.837E-10 1.837E-10 9.977E-01 43 5.043E-06 3.003E+ 10 8.839E + 11 1.838E-10 1.837E-10 9.979E 01 44 1.855E-06 1.798E + 10 9.019E + 11 1.838E-10 1.838E-10 9.980E-01 [ 45 8.764E-07 1.412E + 10 9.160E+ 11 3.168E-14 1.838E-10 9.982E-01 i 46 4.140E-07 8.518E + 09 9.245E + 11 3337E-13 1.838E-10 9.984E-01 47 1.000E-07 3.175E + 10 9.563E + 11 2.968E-13 1.841E-10 1.000E +00 17 1
Sble 7 Absolute flux spectrum calculated with ORIGINAL SAILOR at the center of the 31.5' surveillance capsule at 20.3 cm below the core midplane Upper Group Cumulative Group Cumulative Cumulative Group enercy Oux flux DPA DPA DPA fraction 1 1.733E+ 01 7.643E+ 06 7.643E+ 06 2.233E-14 2.233E-14 1.550E-04 2 1.419E +01 3389E+ 07 4.153E + 07 8.%7E-14 1.120E-13 7.775E 04 3 1.221E+ 01 1378E+08 1.793E +08 3320E-13 4.440E-13 3.082E-03 4 1.000E + 01 2.768E +08 4.562E +08 6.143E-13 1.058E-12 7347E-03 5 8.607E+ 00 4.851E+ 08 9.413E+ 09 1.012E-12 2.071E-12 1.437E-02 6 7.408E + 00 1.129E+ 09 2.070E+ 09 2.194E-12 4.265E-12 2.960E-02 7 6.065 E+ 00 1.570E+ 09 3.640E+ 09 2.801E-12 7.065E-12 4.905E42 8 4.966E+ 00 3.1361i+ 09 6.777E + 09 4.927E-12 1.199E-11 8325E 02 9 3.679E + 00 2.844 E+ 09 9.621E + 09 3.8%E-12 1.589E-11 1.103E-01 10 3.012E + 00 2380E+09 1.200E + 10 3.025E-12 1.891E-11 1313E-01 11 2.7251i+ 00 2 884E+09 1.488E + 10 3.677E-12 2.259E-11 1.568E-01 12 2.466E + 00 1.440E + 09 1.632E+ 10 1.687E-12 2.428E-11 1.685E-01 13 2365E+ 00 4.475 E+ 08 1.677E+ 10 4.904E-13 2.477E-11 1.719E-01 14 234611+00 2.231E+ 09 1.900E+ 10 2323E-12 2.709E-11 1.881E-01 15 2.231E+ 00 5.940E+ 09 2.494E + 10 6.142E-12 3323E-11 2307E-01 16 1.920E +00 8 067E+ 09 3301E+ 10 6.561E 12 3.979E-11 2.762E-01 17 1.653E + 00 1.238E + 10 4.539E + 10 1.003E 11 4.983E-11 3.459E-01 18 1353E+00 2.769E+ 10 7309E+ 10 1.552E-11 6.534E-11 4.536E 01 19 1.003E + 00 2.146E+ 10 9.455E + 10 7.866E-12 7321E-11 5.082E-01 20 8.208E-01 9.770E + 09 1.043E+ 11 5.480E-12 7.869E-11 5.462E-01 21 7.427E-01 3.585E + 10 1.402E+ 11 1.2%E-11 9.165E-11 6362E-01 22 6.081E-01 2.970E+ 10 1.699E + 11 8.742E-12 1.004E-10 6.969E-01 23 4.979E-01 3399E + 10 2.039E + 11 1348E 11 1.139E-10 7.904E-01 24 3.688E-01 3.476E + 10 2386E + 11 7.189E-12 1.211E-10 8.403E-01 25 2.97211-01 4.224 E + 10 2.809E + 11 8.469E-12 1.295E-12 8.991E-01 26 1.832E-01 4.177E+ 10 3.226E+ 11 5.890E-12 1354E-10 9.400E-01 27 1.11 ]E-01 3.174E + 10 3.544E + 11 4.104E-12 1395E-10 9.685 E-01 28 6.738E-02 2.545 E+ 10 3.798E + 11 II,48E-12 1.412E-10 9.799E-01 29 4.087E-02 7.709E+ 09 3.875I1+ 11 6.239E-13 1.418E-10 9.842E-01 30 3.183E-02 2.761E+ 09 3.903E+ 11 7.850E-13 f.426E 10 9.897E-01 37 2.606E-02 1.043E + 10 4.007E + 11 2.102E-13 1.428E-10 9.912E-01 32 2.418E-02 6.821E+ 09 4.075E+ 11 2.987E-14 1.428E-10 9.914E-01 33 2.188E42 1.257E+ 10 4.201E + 11 1.030E-13 1.429E-10 9.921E-01 34 1.503E-02 2 069E+ 10 4.408E+ 11 3.881E-13 1.433E-10 9.948E-01 35 7.102E-03 2.780E + 10 4f>86E+ 11 2.473E-13 1.435E-10 9.%5E-01 36 3355E-03 2.540E + 10 4.940E+ 11 8.939E 14 1.436E-10 9.971E-01 37 1.585E43 3.853E+ 10 5325E+ 11 6.481E 14 1.437E-1P 9.976E-01 38 4.540E-04 1.894E+ 10 5.515 E + 11 1.872E-15 1.437E-10 9.976E-01 39 2.144E 04 2.291E + 10 5.744E + 11 3.285E-15 1.437E-10 9.976E41 40 1.013E-04 3.095E+ 10 6.054E + 11 6.904E-15 1.437E 10 9.976E-01 41 3.727E-05 3.673E+ 10 6.421E + 11 1.447E-14 1.437E-10 9.977E-01 42 106SE-05 2.013E + 10 6.622E + 11 1.293E-14 1.437E-10 9.978E-01 43 5.043E-06 2329E+ 10 6.855E + 11 2329E-14 1.438E-10 9.980E.01 44 1.8551!-06 1.420E + 10 6.997E + 11 2.183E-14 1.438E-10 9/381E-O! 45 8.7M1i-07 1.107E + 10 7.108E + 11 2.483E-14 1.438E-10 9/>83E-01 46 4.140E-07 1.535E + 10 7.261E + 11 6.013E-14 1.439E-10 9.987E41 47 1.000E47 1.956E + 10 7.457E + 11 1.828E-13 1.44 t E-10 1.000E+ 00 18
i Table 8 Absolute flux spectrum calculated with ORIGINAL SAILOR at the center of the 34.0* surveillance capsule at 20.3 cm below the core midplane Upper Group Cumulative Group Cumulative Cumulative Group energy flux flux DPA DPA DPA fraction 1 1.733E+ 01 7.874E +06 7.874E+ 06 2301E-14 2301E-14 1367E-04 2 1.419E+ 01 3.503E + 07 4.291E+ 07 9.270E-14 1.157E-13 6.876E-04 j 3 1.221E+ 01 1.438E+ 08 1.868E+08 3.465E-13 4.622E-13 2.747E-03 4 1.000E+01 2.907E +08 4.774E+ 08 6.450E-13 1.107E-i2 6.580E-03 5 8.607E+ 00 5.139E + 08 9.913E +08 1.073E-12 2.180E-12 1295E-02 6 7.408E + 00 1.205E + 09 2.1%E + 09 2341E-12 4.521E-12 2.687E-02 7 6.065E+ 00 1.699E +09 3.895E+ 09 3.031E-12 7.552E-12 4.487E-02 8 4.966E + 00 3.470E+ 09 7365E+09 5.451E12 1300E-11 7.727E-02 9 3.679E+ 00 3.216E+09 1.058E+ 10 4.406E-12 1.741E-11 1.034E-01 10 3.012E + 00 2.694E + 09 1327E+ 10 3.424E-12 2.083E-11 1.238E.01 11 2.725E + 00 3.277E+ 09 1.655E + 10 4.178E-12 2.501E-11 1.486E-01 12 2.466E+ 00 1.636E + 09 1.819E+ 10 1.918E-12 2.693E-11 1.600E-01 13 2365E+00 5.086E+ 08 1.870E+ 10 5375E13 2.749E-11 1.633E-01 14 2346E+00 2.539E+ 09 2.124E+ 10 2.643E-12 3.013E 11 1.790E 01 15 2.231E + 00 6.777E+ 09 2.801E+ 10 7.007E-12 3.714E-11 2.207E.01 16 1.920E + 00 9.286E+09 3.730E+ 10 7.552E-12 4.469E-11 2.655E-01 17 1.653E+ 00 1.431E + 10 5.161E + 10 1.159E-11 5.628E-11 3344E-01 i 18 1353E+00 3.244E + 10 8.405E+ 10 1.818E-11 7.446E-11 4.424E-01 19 1.003E + 00 2.536E + 10 1.094E+ 10 9.295E-12 8375E-11 4.977E-01 20 8.208E-01 1.152E + 10 1.209E+ 11 6.464E 12 9.022E-11 5361E-01 21 7.427E01 4.273E + 10 1.637E+ 11 1.544E 11 1.057E-10 6.278E-01 22 6.081E-01 3.549E + 10 1.992E + 11 1.045E-11 1.161E-10 6.899E-01 23 4.979E.01 4.067E + 10 2398E+ 11 1.612E-11 1322E-10 7.857E-01 24 3.688E-01 4.179E + 10 2.816E+ 11 8.642E-12 1.409E-10 8371E-01 25 2.972E-01 5.035 E+ 10 3320E+ 11 1.009E-11 1310E-10 8.970E-01 26 1.832E-01 4.997E + 10 3.189E+ 11 7.N6E-12 1380E-10 9389E 01 27 1.111E 01 3.795E+ 10 4.199E+ 11 4.908E-12 1.629E-10 9.681E 01 28 6.738E-02 3.035E+ 10 4.502E+ 11 1.966E-12 1.M9E-10 9.798E-01 29 4.087E-02 9.107E + 09 4.593E + 11 7370E-13 1.656E-10 9.841E-01 30 3.183E-02 3.236E+09 4.626E+ 11 9.200E-13 1.665E-10 9.896E01 31 2.606E-02 1.245 E + 10 4.750E+ 11 2.511 E-13 1.668E-10 9.911E-01 32 2.418E-02 8.204E+ 09 4.832E+ 11 3.593E-14 1.668E-10 9.913E-01 l 33 2.188E-02 1.487E + 10 4.981E+ 11 1.218E-13 1.669E-10 9.920E41 34 1.503E-02 2.419E+ 10 5.223E+ 11 4.538E-13 1.674E-10 9.947E-01 35 7.102E 03 3.272E + 10 5.550E + 11 2.910E-13 1.677E-10 9.965E01 36 3355E-03 3.001E + 10 5.850E+ 11 1.056E13 1.678E-10 9.971E-01 37 1.585 E-03 4357E+ 10 6306E+ 11 7I66E-14 1.679E-10 9.975E-01 38 4340E-M 2.234E + 10 6.529E + 11 2.203E-15 1.679E-10 9.976E01 39 2.144E-04 2.703E + 10 6.800E+ 11 3.876E-15 1.679E-10 9.976E-01 40 1.013E-04 3f62E+ 10 7.166E+ 11 8.170 6 15 1.679E-10 9.976E 01 41 3.727E-05 4347E + 10 7.601E + 11 1.712E-14 1.679E-10 9.977E-01 42 1.06SD05 2379E + 10 7.839E+ 11 1.529E-14 1.679E-10 9.978E-01 43 5.N3E-06 2.744E+ 10 8.113E+ 11 2.744E-14 1.679E-10 9.980E-01 44 1.855E06 1.663E+ 10 8.279E + 11 2356E-14 1.680E-10 9.981 E-01 45 8.764E-07 1.285 E + 10 8.40SE + 11 2.883E-14 1.680E-10 9.983E-01 46 4.140E-07 1.776E + 10 8385E+ 11 6.959E-14 1.681E-10 9.987E-01 47 1.000E 07 2314E+ 10 8.817E + 11 2.164E-13 1.683E-10 1.000E+ 00 8 19
l Table 9 Calculated saturated activities at the center of Trojan 31.5" surveillance capsule at the core midplane for Cyc!c 13 Reaction per atom per second W/ENDF/B VI Fe-O-H' W/ ORIGINAL SAILOR
- Dosimeter R = 207.33 cm R = 207.33 cm "Fe(n p)"Mn 4.851E-15 4.233E-15 "Ni (n p) "Co 6.813E-15 5.985E-15 "Cu(n,a) "Co 4.409E-17 4.051E-17 23'Np(n,f)"7Cs 2.506E-13 2.294E-16 U(n,f)"7Cs 2.623E-14 2.361E-14 238
- Ti(n,p) "Sc 7.094E-16 6.281E-16 "Co(n,y) "Co 3.443E-12 3.092E-12 4 (E) cm-2 s
Flux 4 (E > 1.0 MeV) 8.132E+10 7.343E+10 4 (E > 0.1 MeV) 3.491E+11 3.226E+ 11 4 (E < 0.4 eV) 3.421 E+ 10 3.491E+ !0 DPA rate dpals Total DPA rate 1.576E-10 1.441E-10 'ENDF/B-VI dosimetry cross sections were used in both calculations. 20
Table 10 Calculated saturated activities at the center of Trojan 34.0* surveillance capsule at the core midplane for Cycle 13 Reaction per atom per second W/ENDF/B-VI Fe-O-H" W/ ORIGINAL SAILOR' Dosimeter R = 207.33 cm R = 207.33 cm l "Fe(n,p)"Mn 5.358E-15 4.669E-15 "Ni (n,p) "Co 7.566E-15 6.639E-15 63Cu(n,a) "Co 4.693E-17 4.301E-17 "Ti (n,p) "Sc 2.928E-13 2.681E-13 238U (n,f) 237Cs 2.977E-14 2.679E-14 237Np(n,f)'37Cs 7.643E-16 6.745E-16 "Co(n,y) "Co 4.062E-12 3.643E-12 Flux 4 (E) cm.2 3 .i 4 (E > 1.0 MeV) 9.354E-10 8.445E+ 10 o (E > 0.1 MeV) 4.134E-11 3.819E+ 11 4 (E < 0.4 eV) 4.027E-10 4.091E+ 10 DPA rate dpa/s Total DPA rate 1.841E-10 1.683E-10 " ENDF/B-VI dosimetry cross sections were used in both calculations. 21
l l Table 11 Spectrum averaged cross section at the center of the 31.5* surveillance capsule at core midplane of Trojan Cycle 13 W/ENDF/B-VI Fe-O-H' W/ ORIGINAL SAILOR' E>1 E >.11 E>1 E >.11 Reactnan a (bams t. 5 (barnsy 5 (bantst 5 (barnsY "Fe(n,p)"Mn 5.965E-02 1.390E-02 5.764E-02 1.312E-02 58Ni (n,p) 58Co 8.379E-02 1.952E-02 8.150E-02 1.855E-02 "Cu(n,a) "Co 5.422E-04 1.263E-04 5.517E-04 1.256E-04
- Ti (n,p) "Sc 8.724E-03 2.032E-03 8.553E-03 1.947E-03 238U (n,f) "'Cs 3.226E-01 7.514E-02 3.215E-01 7.318E-02 2"Np(n,f)"7Cs 3.082E+00 7.178E-01 3.124E-00 7.111E-01 "Co(n,y) "Co 4.234E+01 9.862E-00 4.210E+01 9.583E-00
' ENDF/B-VI ds ietry cross sections were used in both calculations. 4 [ o(E)4(E)dE 0 i. [4(E)dE 1 [ o(E)4(E)dE 0 .g. [ 4(E)dE 0.11 22 l
Table 12 Spectrum-averaged cross section at the center of the 34.0 surveillance capsule at core midplane of Trojan Cycle 13 W/ENDF/B-VI Fe-O-H' W/ ORIGINAL SAILOR' E>1 E >.11 E>1 E >.11 Reaction 5 (bams)* 5 (barns)* 5(barns)* 5 (bamsY HFe(n,p)"Mn 5.728E-02 1.2%E-02 5.528E-02 1.222E-02 "Ni (n,p) 58Co 8.089E-02 1.830E-02 7.861E-02 1.738E-02 Cu(n,a) "Co 5.017E-04 1.135E-04 5.093E-04 1.126E-04 "Ti (n,p) "Sc 8.171 E-03 2.849E-03 7.986E-03 1.766E-03 2"U (n,f) 7Cs 3.183E-01 7.202E-02 3.172E-01 7.014E-02 237Np(n,f)7Cs 3.130E+00 7.082E-01 3.174E-00 7.019E-01 "Co(n,y) "Co 4.343E+01 9.826E-00 4.314E+01 9.540E-00
- ENDF/B-VI dosimetry cross sections were used in both calculations.
l i famuE)ss o lME)dE 1 I .g. 0.11 23
1 I l Table 13 Comparison of calculated absolute saturated reaction rates for Trojan Cycle 13 with experimental and Westinghouse calculated results of Capsule X (removed April 1984) LSU Calculated (for Cycle 13) Westinghouse (for Capsule X) w/ENDF6 w/ ORIGINAL Reaction Experimental Calculated Fe-O-H SAILOR 5'Fe(n,p) 54Mn 4.16E6 4.97E6(1.20)' 3.35E6(0.81)" 2.92E6(0.70)' Cu(n,a) "Co 4.09E5 4.31E5(1.05) 3.08E5(0.75) 2.82E5(0.69) '5 58Ni(n,p) 58Co 6.28E7 7.47E7(1.19) 5.31E7(0.85) 4.65E7(0.74) 23'Np(n,f)"7Cs 7.30E7 7.46E7(1.02) 4.67E7(0.64) 4.27E7(0.59) 238U(n,f) "7Cs 7.68E6 6.47E6(0.84) 4.53E6(0.59) 4.07E6(0.53) " Ratio of calculated to capsule X experimental values. 24
Table 14 Calculated relative azimuthal variation of Dux > 1 MeV at four radial locations for Trojan Cycle 13 J E OT 3/4 T R' = 257.0 cm R' = 321.6 cm 1 5 0000E43 0.5328 0.5809 0.6963 0.8346 2 2.0601 E.01 0.5364 0.5796 0.6968 0.8282 3 4.2601 E.01 0.5387 0.5783 0.6974 0.8225 4 7.2500E-01 0.5354 0.5800 0.6994 0.8037 5 1.1250E +00 0.5389 0.5809 0.7008 0.7500 6 1.2530E+00 0.5441 0.5802 0.7003 0.7249 7 13100E + 00 0.5377 0.5798 0.7005 0.7187 8 1.7820E + 00 0.53 % 0.5822 0.6989 0.6899 9 2.6000E + 00 0.5448 0.5871 0.7015 0.6005 10 3.2500E + 00 0.5485 0.5916 0.7064 0.5000 11 3.5565E + 00 0.5536 0.5952 0.7087 0.4392 12 3.8140E + 00 0.5534 03985 0.7118 03667 13 4.1326E + 00 0.5597 0.6001 0.7145 03038 14 43990E +00 0.5609 0.6028 0.7169 0.2755 15 4.7740E+00 03663 0 6094 0.7201 0.2556 16 5.1950E + 00 0.5725 0.6130 0.7231 0.2374 17 6.0700E + 00 0.56 0.6274 0.7352 0.2219 18 7.5001 E+ 00 0.6112 0.6533 0.754( 0.2153 19 9.0000E+ 00 0.6432 0.6858 0.7759 0.2181 20 1M35E+ 01 0.6781 0.7211 0.8001 0.2243 21 1.1310E + 01 0.7017 0.7418 0.8177 0.2265 22 1.1750E+ 01 0.7162 0.7558 OR264 0.2294 23 1.2375E + 01 0.7320 0.7714 OR387 0.2308 24 1.2994E + 01 0.7490 0.7890 0.8515 0.2335 25 13619E + 01 0.7702 0.8052 0.8626 02345 26 1.4375E4 01 0.7906 0.8255 0 8757 0.2%7 27 1.5255E+01 0.8120 0.8494 0.8897 0.2378 28 1.6380E + 01 0.8461 0E740 0.9044 0.2382 29 1.7500E +01 0.8722 0.8975 0.9158 0.2381 ?O 1 A435E +0i 0.8910 0.9103 0.9221 0.2372 31 1.9435E+ 01 0.9099 0.9158 0.9235 0.2371 32 2.0375E + 01 0.9149 0.9127 0.9212 0.2362 33 2.1037E+ 01 0.9050 0.9043 0.9172 0.2%8 34 2.1912E +01 03884 0.8834 0.9094 0.2350 35 23000E+ 01 0.8472 0.8537 0.8968 0.2339 36 23875E4 01 0.8062 0.8262 0.8848 0.2325 37 2.4444E+ 01 0.7850 0.8063 0.8757 0.2317 38 2.4819E+01 0.7761 0.7946 0.86 % 0.2311 39 2.5017E+01 0.7744 0.7873 0.8669 0.2314 40 2.5452E+01 0.7480 0.7711 0.8586 0.2311 41 2.6385E + 01 0.7012 0.7333 0.8428 0.2297 42 2.7450E + 01 0.6434 0.6R69 0.8265 0.2295 43 2.8301E+ 0! 0.5970 0.6523 0.8156 0.2290 44 2.8710E + 01 0.5814 0.6372 0.8109 0.2291 45 2 9174E + 01 0.5739 0.6242 0.8062 0 2298 46 3.0015E + 01 0.5565 0.6102 0.7992 0.2311 47 3 0650E + 01 0.5403 0.6030 0.7954 0.2318 48 3.1500E +01 0.5441 0.6046 0.7973 0.2313 49 3.2300E+ 01 0.5580 0.6083 0H005 0.2312 50 3.2d71E + 01 03619 0.6102 0.8027 0.2314 51 12821E4 01 05676 0.6195 0.8063 0.2310 52 33200E + 01 0.57R4 0.6294 0.8097 0.2306 53 33995E + 01 0.6088 0.6611 0.8214 0.2303 54 3.4695E+ 01 0.63 % 0.6912 0.8332 0.2293 25
[ \\ l l l Table 14 (continued) J 5 0T 3/4 T R' = 257.0 cm R' = 321.6 cm 0.2299 55 3.4850E+ 01 0.6570 0.6978 0.8354 56 3.5500E+01 0.7066 0.7277 0.8487 0.2309 0.2321 57 3.6500E+ 01 0.7743 0.7740 0.8722 0.2349 58 3.7375E + 01 0.8241 0.8142 0.8942 0.2392 59 3.8124E+ 01 0.8664 0.8466 0.9133 0.2464 60 3.8788E + 01 0.8913 0.8747 0.9281 0.2536 0.84;0 0.8950 0.9368 0.2668 61 3.9289E+ 01 0.91I3 0.9453 62 3.9749E + 01 0.9132 0.2801 63 4.0100fi+ 01 0.9287 0.9221 0.9539 0.2827 64 4 0237E+ 01 0.9370 0.9254 0.9574 65 4.0340E + 01 0.9 349 0.9285 0.95 % 0.2859 03320 66 4.0703E+ 01 0.9421 0.9378 0 9667 0.4697 67 4.1193fi+ 0) 0.9544 0.9506 0.9753 0.6625 68 4.1443E + 01 0 9585 0.9561 0.9791 0.7939 69 4.1751E + 01 0.9642 0.9626 0.9824 0.8947 70 4.2251E+ 01 0.9726 0.9710 0.9888 0.9571 71 4.2750E + 01 0.9792 0.9789 0.9939 72 43249E +01 0.9891 0.9860 0.9974 0 9692 0.9852 73 43749E+ 01 0.9944 0.9912 0.9998 0.9941 74 4.4249E+ 01 OW12 0.9955 1.0000 0.9975 75 4.4744E4 01 1.0000 0.9983 0.9983 76 4.4995 E + 01 0.993i 1.0000 0.9%7 1.0000 'R = 257.0 cm is the radial location of the dosimeters behind RPV insulation.
- R = 321.6 cm is the center of the instrumentation tube.
26 I
Table 14 Calculated relative azimuthal variation of flux > 1 MeV at four radial locations for Trojan Cycle 13 J E OT 3/4 T R' = 257.0 cm R = 321.6 cm 6 1 5.0000E43 0.5328 0.5809 0.6963 0.8346 2 2.0601E.01 0.5364 0.57 % 0.6968 0.8282 3 4.2601 E.01 0.5387 0.5783 0.6974 0.8225 4 7.2500E.01 0.5354 0.5800 0.6994 0.8037 5 1.1250E+ 00 0.5389 0.5809 0.7008 0.7500 6 1.2530E + 00 0.5441 0.5802 0.7003 0.7249 7 13100E + 00 0.5377 0.5798 0.7005 0.7187 8 1.7820E+ 00 0.5390 0.5822 0.6989 0.6899 9 2.6000E + 00 0.5448 0.5871 0.7015 0.6005 10 3.2500E + 00 0.5485 0.5916 0.7064 03000 11 3.5565E4 00 0.5536 0.5952 0.7087 0.4392 l 12 3.8140E + 00 0.5534 0.5985 0.7118 03667 13 4.1326E + 00 0 5'97 0 6001 0.7145 03038 14 43990E + 00 0.5609 0.6028 0.7169 0.2755 15 4.7740E + 00 03663 0 6094 0.7201 0.2556 16 5.1950E+ 00 0.5725 0.6130 0.7231 0.2374 17 6.0700E + 00 0.5848 0.6274 0.7352 0.2219 18 7.5001 E + 00 0.6112 0.6533 0.7546 0.2153 19 9.00(K)E + 00 0.6432 0.6858 0.7759 0.2181 20 1.04351? + 01 0.6781 0 7211 0.8001 0.2243 21 1.1310E+01 0.7017 0.7418 0.8177 0.2265 22 1.1750E+ 01 0.7162 0 7558 0.8264 0194 23 1.2375E+ 01 0.7320 0.7714 0.8387 0.2308 24 1.2994E+ 01 0.7490 0.7890 0.8515 0.2335 25 13619E + 01 0.7702 CA052 OM26 0.2345 26 1.4375E + 01 0.79 % 0.8255 0.8757 0.2367 27 1.5255E +01 0.8120 0.8494 0.8897 0.2378 28 1.6380E4 01 0.8461 0.8740 0.9044 0.2382 29 1.7500E+01 0.8722 0.8975 0.9158 0.2381 30 IB435E+01 0.8910 0.9103 0.9221 0.2372 31 1.9435E+ 01 0.9099 0.9158 0.9235 0.2371 32 2.0375E + 01 0.9149 0.9127 0.9212 0.2362 33 2.1037E+ 01 0.9050 0.9043 0.9172 0.2368 34 2.1912E4 01 0.8884 0.8834 0.9094 0.2350 35 23000E + 01 0.8472 0.8537 0.8%8 0.2339 36 23875E+01 0.8062 0.8262 0.8848 0.2325 37 2.4444E+ 01 0.7850 0.8063 0.8757 0.2317 38 2.4819E+ 01 0.7761 0.7946 0.86 % 0.2311 39 2.5017E 401 0.7744 0.7873 0.8669 0.2314 40 2.5452E +01 0.7480 0.7711 0.8586 0.2311 l 41 2.63R5E +01 0.7012 0.7333 0.8428 0.2297 42 2.7450E+ 01 0.6434 0.6869 0.8265 0.2295 43 2.a301E + 01 0.5970 0.6523 0.8156 0.2290 44 2.8710E+ 01 0.5814 0.6372 0.8109 0.2291 45 2.9174E+ 0! 0.5739 0.6242 0.8062 0.2298 46 3.0015 E + 01 03565 0 6102 0.7992 0.2311 47 3.0650E + 01 03403 0.6030 0.7954 0.2318 48 3.1m)E+ 01 0.5441 0.6046 0.7973 0.2313 49 3.2300E + 01 0.5580 0.6083 0.8005 0.2312 50 3.2471 E+ 01 03619 0.6102 0.8027 0.2314 l 51 3.2821E + 01 03676 0.6195 0.8063 0.2310 52 33200E+ 01 0.5784 0 6294 0.8097 0.2306 53 33995E +01 0.6088 OM11 0.8714 0.2303 54 3 4695E + 01 0.6396 0.6912 0.8332 0.2293 25 l l
= Table 14 (continued) 6 J s 0T 3/4 T R* = 257.0 an R = 321.6 an 55 3.4850E+01 0.6570 0.6978 0.8354 0.2299 56 3.5500E+ 01 0.7066 0.7277 0 8487 0.2309 57 3.6500E+01 0.7743 0.7740 0.8722 0.2321 58 3.7375E+ 01 0.8241 0.8142 0.8942 0.2349 59 3.8124E+ 01 0.8664 0.8466 0.9133 0.2392 60 3.8788E+ 01 0.8913 0.8747 0.9281 0.2464 61 3.9289E+01 0.8990 0.8950 0.9368 0.2536 62 3.9749E + 01 0.9132 0.9113 0.9453 0.2668 63 4 0100E+ 01 0.9287 0.9221 0.9539 0.2801 64 4.0237E+01 0.9370 0.9254 0.9574 0.2827 65 4.0340E + 01 0.9349 0 9285 0.9596 0.2859 66 4.0703E+ 01 0.9421 0.9378 0.9667 0.3320 67 4.1193E +01 0.9544 0.9506 0.9753 0.4697 68 4.1443E+ 01 0 9585 0.9561 0.9791 06625 69 4.1751E+ 01 0.9642 0.% 26 0.9824 0.7939 70 4.2251E+01 0.9726 0.9710 0.9888 0.8947 71 4.2750E+ 01 0.9792 0.9789 0.9939 0.9571 72 43249E+ 01 0.9891 0.9860 0.9974
- 0. % 92 73 43749E+ 01 0.9944 0.9912 0.9998 0.9852 74 4.4249E+01 0 9992 0.9955 1.0000 0.9941 75 4.4744E+01 1.0000 0.9983 0.9983 0.9975 76 4.4995E+ 01 0.9931 Ifh00 0.9967 1.0000 "R = 257.0 cm is the radial k) cation of the dosimeters behind RPV insulation.
R = 321.6 cm is the center of the instrumentation tube. 6 26
Table 15 Calculated relative axial variation of Dux > 1 MeV at four radial locations for Trojan Cycle 13 J E' OT 3/4 T R = 257.0 cm R' = 321.6 cm 6 1 2.159E + 00 0.0407 0.0458 0.1581 0.1448 2 6.477E + 00 0.0636 0.0759 0.17 % 0.2017 3 1.079E+01 0.0882 0.1044 0.2058 0.2373 4 1.511E+01 0.1188 0.1361 0.2343 0.2666 5 1.943E+ 01 0.1544 0.1716 0.2649 0.2952 6 2381E + 01 0.1952 0.2115 0.2981 03245 7 2.826E+01 0.2411 0.2550 03325 03549 8 3.231E+ 01 0 2862 0.2973 03645 03818 9 3.780E+ 01 03480 03570 0.4104 0.41 % 10 4.511E+ 01 0.4360 0.4399 0.4737 0.4695 11 5.243E + 01 0.5276 0.5247 0.5376 0.5197 12 5.974 E+ 01 0.6170 0.6079 0.5997 0.5687 13 6.706E + 01 0.6972 0.6848 0.6585 0.6162 14 7.437E + 01 0.7651 0.7516 0.7124 0.6636 15 8.169E + 01 0.8201 0.8069 0.7603 0.7078 16 8.900E + 01 0.8627 0.8511 0.8021 0.74 % 17 9.632E + 01 0.8954 0.8856 0.8375 0.7878 18 1.036E + 02 0.9211 0.9123 0.8675 0.8213 19 1.109E+02 0.9401 0.9329 0.8925 0.8507 20 1.183E + 02 0.9541 0.9485 0.9134 0.8765 21 1.256E+02 0.9649 0.9004 0.9308 0.8991 22 1329E + 02 0.9730 0.96 % 0.9453 0.9191 23 1.402E + 02 0.9793 0.9767 0.9574 0.9362 24 1.475E + 02 0.9843 0.9823
- 0. % 73 0.9507 25 1.548E + 02 0.9881 0.9867 0.9755
- 0. % 28 26 1.622E +02 0.9913 0.9903 0.9821 0.9728 27 1.695E+ O2 0.9938 0.9932 0.9876 0.9809 28 1.768E + 02 0.9959 0.9955 0.9918 0.9876 29 1.841E + 02 0.9976 0.9973 0.9951 0.9925 30 1.914 E + 02 0.9987 0.9987 0.9975 0.9%4 31 1.987E + O2 0.9996 0.9996 0.9991 0.9987 32 2.060E4 02 1.0000 1.0000 1.0000 0.9999 33 2.134E+ O2 1.0000 1.0000 1.0000 1.000()
34 2.207E + 02 0.9995 0.9994 0.9993 0.9988 35 2.280E + 02 0.9986 0.9984 0.9978 0.9967 36 2353E + 02 0.9971 0.9968 0.9955 0.9932 37 2.426E + 02 0.9951 0.9947 0.9923 0.9886 38 2.499E + 02 0.9927 0.9921 0.9881 0.9824 39 2.573E + 02 0.9896 0.9888 0.9829 0.9747 40 2.646E+ O2 0.9860 0.9348 0.9765 0.9654 41 2.719E + 02 0.9815 0.9199 0.9688 0.9543 42 2.792E + 02 0.9762 0.9740 0.9595 0.9411 43 2.865E+ 02
- 0. % 97 0.9669 0.9483 0.9256 44 2.938E + 02 0.9615 0.9580 0.9349 0.9074 45 3.01111+02 0.9514 0.9468 0.9191 0.8865 46 3.0RS E + 02 0.9382 03127 0.9002 0.8628 47 3.158E+ 02 0.9216 0 9146 0.8779 0.8.W) 48 3.231E+ 02 0.9000 0.8915 0.8515 0.8059 49 3.304E + 02 0.8714 0.8622 0.8205 0.7721 50 3.377E 4 02 0.8359 0.8256 0.7843 0.7339 51 3.450E + 02 0.7915 0.7802 0.7426 0.6928 27
Table 15 (continued) J Z" 0T 3/4 T R = 257.0 cm R' = 321.6 cm 6 52 3.523E+02 0.7363 0.7248 0.6951 0.6484 53 3.597E+ 02 0.6696 0.6588 0.6423 0.6028 54 3.670E + 02 0.5909 0.5837 0.5852 0.5558 55 3.743E+02 0.5043 0.5031 0.5252 0.5075 56 3.816E+ 02 0.4164 0.4212 0.4645 0.4587 57 3.889E + 02 03321 03420 0.4045 0.4110 58 3.944E + 02 0.2732 0.2854 03613 03752 59 3.995E+02 0.2204 0.2363 03227 03455 60 4.061E + 02 0.1615 0.1800 0.2759 03060 61 4.127E+ 02 0.1138 0.1328 0.2357 0.2669 62 4.194E + 02 0.0778 0.0951 0.2001 0.2321 63 4.260E + 02 0.0521 0.0664 0.1683 0.1969 64 4319E+ 02 0.0355 0.0468 0.1435 0.1678 65 4371E+ 02 0.0245 0.0333 0.1244 0.1401 66 4.424 E+ 02 0.0149 0.0197 0.1089 0.0978 'Adal height in cm (reactor core 30.48 - 3%.24 cm, core midplane = 213.36 cm).
- R = 257.0 cm is the radial kx:ation of the dosimeters behind the RPV insulation.
'R = 321.6 cm is the center of the instrumentation tube. 1 l l l l 28
Table 16 Radial variation of integrated parameters along RPV, calculated with ENDF/B-VI Fe-O-H at 0* (20.3 cm below midplanc), for Trojan Cycle 13* 1.ncation Fast flux Flux >0.1 Thermal Thermal / DPA rate >1.0 MeV MeV flux < 0.4 eV fast (DPA/S) RPV liner 1.005E+ 10 1.985E+ 10 2.429E+ 10 2.416 1.553E-11 0-T 9.736E+09 2.020E+ 10 1.499E+ 10 1.540 1.506E-11 1/4-T 5.606E+09 1.774E+10 2.535E+08 0.045 9.674E-12 3/4-T 1.302E+09 8.883E+09 2.785E+07 0.021 3.432E-12 'RPV liner starts at 219.71 cm, RPV extends from 220.27 to 242.17 cm. NOTE: Negligible difference between midplane values and 20.3 cm below midplane. Table 17 Radial variation of integrated parameters along RPV, calculated with ENDF/B-VI Fe-O-H at 45' (20.3 cm below midplane), for Trojan Cycle 13' 1.ocation Fast flux Flux >0.1 Thermal Thermal / DPA rate > 1.0 MeV MeV flux < 0.4 eV fast (DPA/S) RPV liner 1.868E+ 10 4.430E+ 10 1.331E+ 11 7.123 2.992E-11 0-T 1.816E+ 10 4.500E+ 10 8.163E+ 10 4.495 2.892E-11 1/4-T 1.027E+ 10 3.803E+ 10 1.203E+09 0.117 1.859E-11 3/4-T 2.243E+09 1.750E+ 10 3.432E+07 0.015 6.420E-12 "RPV liner starts at 219.71 cm, RPV extends from 220.27 to 242.17 cm. NOTE: Negligible difference between midplane values and 20.3 cm below midplane. Table 18 Radial variation of integrated parameters along RPV, calculated with ORIGINAL SAILOR at 0* (20.3 cm below midplane), for Trojan Cycle 13* 1.acation Fast flux Flux >0.1 Thermal Thermal / DPA rate > 1.0 MeV MeV flux < 0.4 eV fast (DPA/S) RPV liner 9.858E+09 1.925E+ 10 2.255E+ 10 2.287 1.521E-11 O-T 9.525 E+09 1.%1E+ 10 1.416E+ 10 1.486 1.472E-11 1/4-T 5.241E + 09 1.710E+ 10 2.823E~t08 0.054 9.128E-12 3/4-T 1.116E+09 8.065E+ 09 3.354E+ 07 0.030 3.044E-12 "RPV liner starts at 219.71 cm, RPV extends from 220.27 to 242.17 cm. NOTE: Negligible difference between midplane values and 20.3 cm bekyw midplanc. 29
1 Table 19 Radial variation of integrated parameters along RPV, calculated with ORIGINAL SAILOR at 45* (20.3 cm below midplane), for Trojan Cycle 13* Iecation Fast flux Flux >0.1 Dermal Thermal / DPA rate >1.0 MeV MeV flux < 0.4 eV fast (DPA/S) RPV liner 1.709E+10 4.027E+ 10 1.187E+ 11 6.942 2.725E-11 0-T 1.658E+ 10 4.0%E+10 7.378E+ 10 4.450 2.631E-11 1/4-T 9.016E+09 3.439E+ 10 1.246E+09 0.138 1.647E-f l 3/4-T 1.807E+09 1.4%E+10 4.083E+07 0.023 5.373E-12 "RPV liner starts at 219.71 cm, RPV extends from 220.27 to 242.17 cm. Negligible difference between midplane values and 20.3 cm below midplane. NOTE: Table 20 Radial variation of cumulative fluence (n.cm-2) and DPA along RPV, calculated with ENDF/B-VI Fe-O-H at 0*, at near core midplane for Trojan Cycle 13" 32 EFPY Cycle-13' Location Fluence DPA Fluence DPA RPV liner 1.834E+ 17 2.835E-04 1.016E+19 1.570E-02 0-T 1.777E+17 2.749E4M 9.843E+ 18 1.523E-02 1/4-T 1.023E + 17 1.766E-04 5.668E+ 18 9.780E4)3 3/4-T 2.376E+ 16 6.264E-05 1.316E+ 18 3.470E413 "RPV liner starts at 219.71 cm, RPV extends from 220.27 to 242.11 '211.25 full power days = 1.8252E+07 seconds. '32 EFPY amounts to 11,6% full-power days = 1.011E+0'3 seconds. Table 21 RPV peak cumulative fluence (n.cm'2) and DPA calculated with ENDF/B-VI Fe-O-H for Trojan Cycle 13' Cycle-13' 32 EFPY location Fluence DPA Fluence DPA RPV liner 3.410E+ 17 5.461E-04 1.889E+ 19 3.025E412 0-T 3.314E+ 17 5.278E-04 1.836E+ 19 2.924E-02 1/4-T 1.874E+ 17 3.393E 04 1.038E+ 19 1.879E-02 3/4-T 4.094E+ 16 1.172E-04 2.268E+ 18 6.491E-03 "RPV liner starts at 219.71 cm, RPV cxtends from 220.27 to 242.17 cm.
- 211.25 full.powcr days = 1.8252E+07 seconds.
'32 EFPY amounts to 11,696 full. power dap = 1.OllE+09 seconds. 30
respectively. The projected cumulative fluence after 32 EFPY for the same locations are about 1.889E+19 and 1.836E+19 n.cm-2, respectively. Tables 22 and 23 show the absolute flux spectra at the O' cavity dosimeter location (5 cm behind insulation) and at the center of the instrumentation tube, at 20.3 cm below the core midplane, while Tables 24 and 25 show the values calculated with ORIGINAL SAILOR. Tables 26 and 27 show the calculated saturated reaction rates obtained with both libraries at the above locations, and Tables 28 through 33 show the same quantities at the same dosimeters radial location, but at the axial location of 1.219 meters above the core midplane. The spectrum averaged cross sections for several reactions calculated at the radial locations of the two dosimeter capsules, at 20.3 cm below the core midplanc are given in Tables 34 and 35, respectively. At the axial location of 121.9 cm above the core midplane, the values of the spectrum averaged cross section for the same reactions differ by less than 21% compared with the values at 20.3 cm below midplane, indicating little axial variation in the neutron energy spectrum. Table 36 shows the non-saturation factors for each dosimeter calculated from the power time history of Cycle 13. These factors correspond to the fraction of saturation obtained by each dosimeter at the time of removal from the core. Table 37 gives the ratio of calculated to experimental (C/E) results for the various dosimeters, obtained with the ENDF/B-VI data; Table 38 presents similar results obtained with the ORIGINAL SAILOR library. Note that the ENDF/B-VI results for the threshold dosimeters are about 30 to 40% higher than the measured results at the RPV insulation location (i.e., Capsules A & B), and they are about 50 to 70% higher at the instrument tube location (Capsules C & D). The poorest agreement at the instrument tube location is not surprising, since these dosimeters are located within the detector well behind the ex-core detector. This region cannot be modeled well with two-dimensional geometty. The ORIGINAL SAILOR results are about 20% higher than the experimental values at the RPV and are 20 to 40% high within the instrument tube. In contrast to other studies (e.g., the HBR Cycle 9 analysis)'2'S which indicate that the ORIGINAL SAILOR library tends to underestimate the fast neutron reaction rates in the cavity, these Trojan calculations performed with the ORIGINAL SAILOR overpredict the reaction rates compared to the measurements. Recall that the surveillance capsule X dosimeter measurements are underpredicted by the Cycle 13 transport calculations. Consistent with the earlier studies, the new ENDF/B-VI calculations are about 20 to i 25% higher than the ORIGINAL SAILOR results. However, since the ORIGINAL SAILOR fast-neutron-dosimetry results obtained here are already too high, the ENDF/B-VI values come out very high compared with the experiment. Thus the ENDF/B-VI calculations agreement is substantially poorer with the measurements than the ORIGINAL SAILOR calculations for the Trojan cavity dosimeters. We do not presently understand why these Trojan C/E values for ENDF/B-VI are so different from those obtained earlier in H. B. Robinson'$ and those projected for the ANO-1 reactor.8 However, the generic pinwise power distribution used in the transport calculations is at least partially responsible for the excessively high computed activities. A constant factor of ~0.75 applied to all calculated results would bring the C/E values in line with the earlier results. Both the ORIGINAL SAILOR and the ENDF/B-VI overestimate the bare Co(n,y) reaction by up to a factor 31
e to of 2, and the "U(n,f) is overestimated by a factor of 2 to 3. The latter two dosimeters are sensitive 2 to thermal neutrons and can be impacted by a number of factors with high uncertainties, such as the amount of water and impurity materials present in the concrete shield around the cavity. The C/E's of the thermal dosimeters in Capsule A appear to be inconsistent with the other capsules. Based on the Trojan cavity dosimeter analysis for Cycle 13, it appears that there is an unexplained inconsistency between the calculated and measured dosimetry results. Comparison with in-vessel dosimeter measurements for this cycle, if available, would help to narrow the possible causes for the observed discrepancies. Also, obtaining and utilizing the plant-specific pinwise power would eliminate this contribution to the discrepancy. Since the ENDF/B-VI calculations produce the higher estimates for the cavity flux, we recommend for now that these results be used in the radiation damage assessments, with the understanding that these fluxes should be conservative by about a factor of 2 for the thermal range and by 50% in the fast-energy range. Of primary concern in Trojan and other similarly designed reactors is the radiation embrittlement of the pivot joint in the horizontal support structure component embedded in the concrete shield. This steel member carries a large bending moment, and if it fails, the reactor support could be compromised. The location of this critical point is approximately 18 cm into the concrete, near e = 23, and about 20.3 cm below the core midplane. Table 39 shows the calculated saturated reaction rates for several dosimeters at this location. Table 40 gives results computed for the flux above 1 MeV, above 0.1 MeV, DPA rate in iron, cumulative fluence, and the thermal flux at this location as well as two other locations of 13 cm and 23 cm into the concrete. The ratio of thermal to fast flux at the 18 cm depth is calculated to be 46.13. Based on the calculated flux above 1.0 MeV, the total fluence at this location due to 211.25 effective full-power days of Cycle 13 is about 1.258E+15 n.cm;2 Based only on the fluence rate computed for Cycle 13, the projected cumulative fast fluence at the joint after 32 EFPY is about 6.971E+16 cm. Table 41 shows the absolute flux spectrum at the critical point. 2 The results are based on the ENDF/B-VI calculations, since they are higher than the ORIGINAL SAILOR values, and are therefore conservative. These are absolute results, corresponding to full-power conditions with the power distribution of Cycle 13. The actual structural components are not included in the transport model; hence, the values in Table 40 reflect the results calculated within the concrete with no correction for perturbations due to the steel itself. This fact, along with considerable uncertainty in the water content of the concrete, and the lumped thermal group approximation introduces a large uncertainty into the calculated value for the thermal flux exposure at the support structure. 32 l
Table 22 Absolute Oux spectrum obtained with ENDF/B-VI Fe-O-H, behind RPV insulation at 0*, R = 257.0 cm, and 20.3 cm below the core midplane { Upper Group Cumulative Group Cumulative Cumulative Group encryv Dux Hux DPA DPA DPA fraction 1 1.733E + 01 1.745E+ 05 1.745E+ 05 5.099E-16 5.099E 16 3.450E-04 2 1.419E+ 01 6.210E + 05 7.955 E+ 05 1.643E-15 2.153E-15 1.457E-03 3 1.221E+ 01 1.883E + 06 2.679E+ 06 4.537E-15 6.690E-15 4.527E-03 4 1.000E + 01 3385E+ 06 6M4E+06 7.511E-M 1.420E-14 9.609E-03 5 8.607E + 00 4.788E + 06 1.085E + 07 9.992E-1.s 2.419E-14 1.637E-02 6 7.408E + 00 9.574 E+ 06 2.043E+ 07 1.860E-14 4.279E-14 2.8%E-02 7 61)65E+00 1.200E +07 3.242E + 07 2.140E-14 6.420E-14 4344E-02 8 4.966E+00 1.939E +07 5.181E+ 07 31M6E-14 9.466E-14 6.405E-02 I 9 3.679E + 00 1.492E + 07 6.674E+ 07 2.044E-14 1.151E-13 7.789E-02 l to 3.012E+ 00 1.159E + 07 7.833E+ 07 1.473E-14 1.298E-13 8.786E42 l 11 2.725E + 00 1.414E+ 07 9.247E + 07 1203E-14 1.479E-13 1.001E-01 12 2.466E+ 00 7.140E + 06 9.%1E+ 07 8368E-15 1.562E-13 1.057E-01 l 13 2365E+ 00 2.230E +06 1.018E+ 08 2.444E-15 1.587E-13 1.074E-01 14 2346E+00 1.122E + 07 1.131E+08 1.168E-14 1.704E-13 1.153E.01 15 2.231 E + 00 3.136E+ 07 1.444E + 08 3.242E-14 2.028E-13 1372E-01 1 16 1.920E+ 00 4.577E+ 07 1.902E + 08 3.723E-14 2.400E-13 1.624E 01 17 1.653E + 00 7.436E + 07 2.646E + 08 6.026E-14 3.003E-13 2.032E-01 18 1353E+ 00 1.894E+ 08 4540E+08 1.061E-13 4.064E-13 2.750E-01 19 1.003E+ 00 1.843E + 08 6383E+ 08 6.755E-14 5.740E-13 3.207E-01 20 8.208E-01 8.402E +07 7.223E+ 08 4.713E-14 5.211E-13 3.526E-01 21 7.427E-01 4.279E + 08 1.150E + 09 1.547E-13 6.757E 13 4.572E-01 22 6.081E-01 3.782E + 08 1.528E + 09 1.113E-13 7.871E-13 5326E-01 23 4.979E-01 4365E+ 08 1.%5 E+ 09 1.730E 13 0.601E-13 6.4%E-01 24 3188E-01 6.108E + 08 2.576E+ 09 1.263E-13 1.086E-12 7351E 01 25 2.972E-01 7.795E + 08 3355E+09 1.563E-13 1.243E-12 8.409E-01 26 1.832E-01 7.054E + 08 4.061E + 09 9.946E-14 1342E-12 9.082E-01 27 1.111 E-01 4.606E + 08 4.521 E + 09 5.956E-14 1.402E-12 9.485E-01 28 6.738E-02 3.299E + 08 4.851E + 09 2.137E-14 1.423E-12 9.629E-01 29 4.087E 02 1.13711+ 08 4.%5E + 09 9.199E-15 1.432E-12 9.692E 01 30 3.183E-02 6.025E + 07 5.025E + 09 1.713E-14 1.449E-12 9.807E-01 31 2.606E-02 2.018E+ 08 5.227E + 09 4.068E-15 1.453E-12 9.835E-01 32 2.418E-02 1.198E + 08 5347E+ 09 5.246E-16 1.454E-12 9.838E41 33 2.188E-02 2.295E + 08 5.576E+ 09 1.879E-15 1.456E-12 9.851E-01 M 1.503E-02 3.120E + 08 5.888E + 09 5.853E-15 1.462E-12 9.891E-01 35 7.102E-03 3.152E+08 6.203E + 09 2.804E-15 1.465E-12 9.910E-01 36 3355E-03 2.731E + 08 6.476E+ 09 9.612E-16 1.465E-12 9.916E-01 37 1.585E-03 4.193E+ 08 6.8%E + 09 7.052E-16 1.466E-12 9.921E-01 38 4.540E-04 2.083E + 08 7.104E + 09 2.054E-17 1.466E-12 9.921E-01 39 2.144E 04 2.153E+ 08 7319E+09 3.088E-17 1.466E 12 9.921E4)1 40 1.013E-04 2.710E + 08 7.590E + 09 6.047E-17 1.466E-12 9.922E 01 41 3.727E-05 3.119E + 08 7.902E + 09 1.229E 16 1.466E-12 9.923E-01 42 1.068E-05 1.716E + 08 8.074E + 09 1.103E 16 1.467E-12 9.923E 01 43 5.043E-06 2.048E &08 8.279E + 09 2.048tL16 1.467E-12 9 925E-01 44 1.855 E-06 1343E+ 08 8.413E + 09 2M4E-16 1.467E-12 9.92f>E 01 45 8.764E4)7 1.200E+ 08 8.533E+ 09 2.694E-16 1.467E-12 9.928E-01 46 4.140E417 1.486E + 08 8 682E + 09 5.824E-16 1.468E-12 9.932E-01 47 1.000E-07 1.077E + 09 9.759E+ 09 1.007E-14 1.478E-12 1.000E + 00 33
Table 23 Absolute flux spectrum obtained with ENDF/B-VI Fe-O-H at the center of the instrumentation tube, R = 321.63 cm,0* and 20.3 cm below the core midplane Upper Group Cumulative Group Cumulative Cumulative Group energy flux flux DPA DPA DPA fraction 1 1.733E + 01 1367E+01 1367E+05 3.995E 16 3.995E-16 3.102E-04 2 1.419E +01 4.884E+05 6.251E+ 05 1.292E-15 1.692E-15 1313E-03 3 1.221E + 01 1.488E+ 06 2.113E+06 3.585E-15 5.277E-03 4.097E-03 4 1.000E +01 2.682E+06 4.795E + 06 5.951E-15 1.123E-14 8.717E-03 5 8.607E+ 00 3.807E +06 8 603E+06 7.946E-15 1.917E-14 1.489E-02 6 7.408E+00 7.657E+ 06 1.626E +07 1.488E-14 3.405E-14 2.644E-02 7 6.065E+ 00 9.736E+ 06 2.600E+07 1.737E-14 5.142E-14 3 993E-02 8 4.966E +00 1.604E + 07 4.203E+ 07 2.520E-14 7.662E-14 5.949E-02 9 3.679E+ 00 1.247E+ 07 5.450E + 07 1.708E-14 9370E-14 7.275E-02 10 3.012E+ 00 9.700E4 06 6.420E+ 07 1.233E-14 1.060E-13 8.232E-02 11 2.72SE+00 1.196E+ 07 7.616E+ 07 1.525E-14 1.213E-13 9.416E-02 12 2.466E + 00 6.106E + 06 S.226E+ 07 7.156E-15 1.284E-13 9.971E-02 13 2365E+00 1.889E+ 06 8.415E+ 07 2.071E-15 1305E-13 1.013E-01 14 2346E+00 9.544E+ 06 9370E+07 9.935E-15 1.404E-13 1.090E-01 15 2.231E + 00 2.608E +07 1.198E+08 2.697E-14 1.674E-13 1300E-01 16 1.920E+ 00 3.7%E + 07 1.577E +08 3.087E-14 1.983E 13 1.539E-01 17 1.653E+00 6.272E+07 2.205E + 08 5.082E-14 2.491E-13 1.934 E-01 18 1353E+00 1.563E+ 08 3.767E+ 08 8.756E-14 3366E-13 2.614E-01 19 1.003E + 00 1.511E+ 08 5.278E+ 08 5.537E-14 3.920E-13 3.044E-01 20 8.208E-01 7.829E+ 07 6.061E+ 08 4391E-14 4359E-13 3385E-01 21 7.427E 01 3.548E+08 9.609E + 08 1.282E-13 5.642E-13 4380E-01 22 6.081E-01 3.269E +08 1.288E+ 09 9.622E-14 6.604E-13 5.127E-01 23 4.979E-01 3.668E+08 1.655E+ 09 1.454E-13 8.058E-13 6.256E-01 24 3.688E-01 5.073E+ 08 2.162E + 09 1.049E-13 9.107E-13 7.071E-01 25 2.972E-01 7384E+08 2.900E+09 1.481E-13 1.059E-12 8.220E-01 26 1.832E-01 6.788E + 08 3.579E+ 09 9.571E-14 1.154E 12 8.%3E.01 27 1.111 E.01 4.484E +08 4.027E+09 5.798E-14 1.212E 12 9.414E-01 28 6.738E-02 3337E+ 08 4361E+09 2.162E-14 1.234E-12 9.581E-01 29 4.087E-02 1.197E +08 4.481E + 09 9.685E-15 1.244E-12 9.657E-01 30 3.183E-02 6.537E+ 07 4.546E+ 09 1.858E-14 1.262E 12 9.801E-01 31 2.606E-02 1.657E + 08 4.712E + 09 3341E-15 1.266E-12 9.827E 01 32 2.418E-02 1.043E+ 08 4.816E+ 09 4.569E-16 1.266E-12 9.830E-01 33 2.188E-02 2.203E + 08 5.037E+ 09 1.8ME-15 1.268E-12 9.844E-01 34 1.503E-02 3336E+08 5370E+09 6.257E-15 1.274E-12 9.893E-01 l 35 7.102E-03 3355E+08 5.706E+ 09 2.984E-15 1.277E-12 9.916E-01 i 36 3355E-03 3.063E + 08 6.012E + 09 1.078E-15 1.278E-12 9.924E-01 37 1.585E-03 4.660E + 08 6.478E+09 7.838E-16 1.279E 12 9.931E-01 38 4.540E-04 2390E+ 08 6.717E+09 2356E-17 1.279E-12 9.931E 01 39 2.144E-04 2.461E+ 08 6.963E +09 3.530E 17 1.279E-12 9.931E-01 40 1.013E-04 3.112E + 08 7.274E + 09 6.943E 17 1.2^/9E-12 9.931E-01 41 3.727E-05 3.600E+ 08 7.634E+ 09 1.418E-16 1.279E-12 9.933E-01 42 1.068E 05 1.983E + 08 7.833E+ 08 1.274E-16 1.279E-12 9.97,4 E-01 43 5.043E-06 2370E+08 8.070E + 09 2370E-16 1.280E-12 9.935E-01 44 1.855E-06 1.543E +08 8.224E4 09 2371E 16 1.280E-12 9.937E-01 45 8.764E-07 1347E+08 8358E+ 09 3.023E-16 1.280E-12 9.940E-01 46 4.140E-07 1.258E + 08 8.484E+09 4.928E-16 1.281E-12 9.943E-01 47 1.000E-07 7.796E+ 08 9.264E + 09 7.28SE-15 1.288E-12 1.000E + 00 34
Table 24 Absolute flux spectrum calculated with ORIGINAL SAILOR behind the RPV insulation at 0* dosimeters, R = 257.0 cm, and 20.3 cm below the core midplane j Upper Group Cumulative Group Cumulative Cumulative Group energy flux Dux DPA DPA DPA fraction 1 1.733E + 01 1.434E+05 1.434E+05 4.189E-16 4.189E-16 3.478E-04 2 1.419E+ 01 5.957E+05 7391E+05 1.576E-15 1.995E-15 1.657E43 3 1.211E+01 1.782E + 06 2.522E+ 06 4.294E-15 6.289E-15 5.221E-03 4 1.000E+01 3.025E+06 5.546E+06 6.712E-15 1300E-14 1.079E.02 5 8.607E +00 4.236E+ 06 9.783E+ 06 8.841E-15 2.184E-14 1.813E-02 6 7.408E+00 8.116E+06 1.790E + 07 1.577E-14 3.761E-14 3.123E-02 7 6.065E + 00 8.940E+ 06 2.684E +07 1.595E-14 5356E-14 4.447E-02 l 8 4.966E + 00 1380E+07 4.064E + 07 2.168E-14 7.524E 14 6.247E-02 l 9 3.679E+ 00 1.177E+ 07 5.241E+07 1.612E-14 9.137E-14 7.586E-02 10 3.012E+ 00 9.565E + 06 6.197E+ 07 1.216E-14 1.035E-13 8.595E-02 11 2.725E + 00 1.167E+ 07 7364E+ 07 1.488E-14 1.184E-13 9.830E-02 12 2.466E + 00 5.832E+06 7.947E +07 6.836E-15 1.252E-13 1.040E-01 13 2365E+00 1.993E+06 8.147E +07 2.185E-15 1.274E-13 1.058E-01 14 2346E+00 9.826E + 06 9.129E+ 07 1.023E-14 1376E-13 1.143E-01 15 2.231E+00 2.427E+ 07 1.156E+ C8 2.509E-14 1.627E-13 1351E-01 16 1.920E+00 3.851E + 07 1.541E + 08 3.132E-14 1.941E-13 1.611E-01 17 1.653E+ 00 5.812E+ 07 2.122E +08 4.709E-14 2.412E-13 2.002E-01 18 1353E+00 1.524E + 08 3.646E+ 08 8.539E-14 3.265E-13 2.711E-01 19 1.003E4 00 1.605E + 08 5.251E+08 5.883E-14 3.854E-13 3.200E-01 20 8.208E-01 6 501E+07 5.901E+08 3.647E-14 4.218E-13 3.502E-01 21 7.427E-01 3347E+08 9.248E +08 1.210E-13 5.428E-13 4.507E 01 22 6.081E-01 3.202E+ 08 1.245E + 09 9.424E-14 6370E-13 5.289E-01 23 4.979E 01 3338E+08 1.579E +09 1323E-13 7.694E-13 6388E-01 24 3.688E-01 4.940E +08 2.073E + 09 1.022E-13 8.715E-13 7.236E-01 25 2.972E-01 5.909E+ 08 2.664E+09 1.185E-13 9.900E-13 8.219E41 26 1.832E-01 6.458E+ 08 3309E+09 9.106E-14 1.081E-12 8.975E-01 27 1.111E41 4.186E + 08 3.728E+09 5.412E-14 1.135E-12 9.425E-01 28 6.738E-02 3.249E + 08 4.053E + 09 2.104E-14 1.156E-12 9399E-01 29 4.087E-02 1.012E + 08 4.154E+09 8.186E-15 1.164E-12 9.667E-01 30 3.183E.02 4.846E+07 4.203E +09 1378E-14 1.178E-12 9.782E-01 31 2.606E-02 2.131E+ 08 4.416E+09 4.296E-15 1.182E-12 9.817E-01 32 2.418E-02 1.561E + 08 4.572E+ 09 6.837E-16 1.183E-12 9.823E-01 33 2.188E 02 2.693E+ 08 4.841E+ 09 2.206E-15 1.185E-12 9.841E-01 34 1.503E-02 2.869E + 08 5.128E+ 09 5381E-15 1.191E-12 9.886E-01 35 7.102E-03 2.805E+ 08 5.408E+ 09 2.495E-15 1.193E-12 9.907E 01 36 3355E.03 2332E+08 5.642E+ 09 8.210E-16 1.194E-12 9.913E-01 37 1.585E-03 3.537E + 08 5.995E+ 09 5.949E-16 1.195E-12 9.918E 01 38 4.540E-04 1.807E+ 08 6.176E+ 09 1.782E-17 1.195E 12 9.919E-01 39 2.144E44 1.860E+ 08 6362E+09 2.667E-17 1.195E-12 9.919E-01 40 1.013E44 2351E+08 6.597E + 09 5.244E-17 1.195E-12 9.919E-01 41 3.727E-05 2.714E+ 08 6.868E + 09 1.069E-16 1.195E-12 9.920E-01 42 1.068E 05 1.494E+ 08 7.018E+09 9.603E-17 1.195E-12 9.921E-01 43 5.043E-06 1.795E + 08 7.197E+09 1.795E 16 1.195E-12 9.922E-01 44 1.855E.06 1.188E + 08 7316E+09 1.826E-16 1.195E-12 9.924E-01 45 8.764E-07 1.052E+ 08 7.421E + 09 2361E-16 1.1%E-12 9.926E.01 46 4.140E-07 2367E+08 7.658E+ 09 9276E-16 1.1%E.12 9.934E-01 47 1.000E-07 8364Et OS 8.514E+ 09 8.006E-15 1.204E-12 1.000E + 00 35
Table 25 Absolute Dux spectrum calculated with ORIGINAL SAILOR, at the center of the instrumentation tube, R = 321.63 cm,0* below the core midplane Upper Group Cumulative Group Cumulative Cumulative Group energy flux flux DPA DPA DPA fractkm 1 1.733E + 01 1.130E + 05 1.130E+ 05 3302E-16 3302E-16 3.152E 04 2 1.419E+01 4.686E+05 5.816E+ 05 1.240E-15 1.570E-15 1.499E-03 3 1.211E1 D1 1.413E+ 06 1.995E + 06 3.404E-15 4.974E-15 4.747E43 4 1.000E+01 2.401E +06 43%E+06 5329E-15 1.030E-14 9.833E-03 5 8.607E+00 3382E+06 7.778E+ 06 7.059 6 15 1.736E-14 1.657E-02 6 7.408E + 00 6.519E+ 06 1.430E+ 07 1.767E-14 3.003E-14 2.866E42 7 6.065E+ 00 7.276E+06 2.157E+ 07 1.298E-14 4301E-14 4.105E 02 8 4.966E +00 1.143E+07 3300E+07 1.795E-14 6.096E-14 5.818E-02 9 3.679E+ 00 9.789E+06 4.279E+ 07 1341E-14 7.437E-14 7.098E-02 10 3.012E+ 00 7.989E+06 5.078E + 07 1.015E-14 8 453E-14 8.067E-02 11 2.725E+00 9.836E + 06 6.061E+ 07 1.254E-14 9.707E-14 9.264E-02 12 2.466E + 00 4.996E+ 06 6.561E +07 5.856E-15 1.029E-13 9.823E-02 13 2365E+00 1.681E 606 6.729E +07 1.842E-15 1.048E-13 9.999E.02 14 2346E+00 8335E+ 06 7.563E+07 8.676E-15 1.134E-13 1.083E 01 15 2.231E + 00 2.012E + 07 9.575E+ 07 2.080E-14 1342E-13 1.281E-01 16 1.920E+00 3.172E + 07 1.275E+ 08 2 580E-14 1.600E-13 1.527E-01 17 1.653E + 00 4.884E+ 07 1.763E +08 3.957E-14 1.996E-13 1.905E-01 18 1353E+00 1.258E+ 08 3.021E+08 7.050E-14 2.701E-13 2.578E-01 19 1.003E + 00 1302E+08 4323E+08 4.771E-14 3.178E-13 3.033E-01 20 8.208E-01 6.069E + 07 4.930E +08 3.404E-14 3.519E-13 3358E-01 21 7.427E-01 2.764E + 08 7.694E+08 9.990E-14 4.518E-13 4312E-01 22 6.081E-01 2.735E +08 1.043E +09 8.048E-14 5322E-13 5.080E-01 23 4.979E-01 2.872E +08 1330E+09 1.138E-13 6.461E-13 6.166E-01 24 3.688E-01 4.069E +08 1.737E +09 8.414E-14 7302E-13 6 969E-01 25 2.972E-01 5.598E +08 2.297E +09 1.122E-13 8.425E-13 8.041E 01 26 1.832E-01 6.074E + 08 2.904E +09 8.565E-14 9.281E-13 8.858E-01 27 1.111E-01 4.041E+08 3308E+09 5.226E-14 9.804E-13 9357E-01 28 6.738E-02 3.16SE +08 3.625E +09 2.052E-14 1.001E-12 9.553E-01 29 4.087E 02 1.039E + 08 3.729E+09 8.409E-15 1.009E-12 9.633E-01 30 3.183E-02 5343E+07 3.782E+ 09 1.519E-14 1.024E-12 9.778E-01 31 2.606E-02 1.690E + 08 3.951E+ 09 3.407E-15 1.02SE-12 9.810E-01 32 2.418E-02 1.260E +08 4.077E+ 09 5.516E-16 1.028E-12 9.816E-01 33 2.188E 02 2.456E + 08 4323E+09 2.011E-15 1.030E-12 9.835E-01 34 1.503E-02 3.000E +08 4.623E+ 09 5.627E-15 1.036E-12 9.888E41 35 7.102E-03 3.004E+ 08 4.923E+ 09 2.672E-15 1.039E-12 9.914E-01 36 3355E-03 2.608E + 08 5.184E+ 09 9.181E-16 1.040E-12 9.923E-01 37 1.585E-03 3,920E + 08 5.576E+ 09 6.593E-16 1.040E-12 9.929E-01 38 4.540E-04 2.072E + 08 5.783E+ 09 2.043E-17 1.Gs0E-12 9.929E-01 39 2.144E 04 2.121E +08 5.995E+09 3.041E-17 1.040E-12 9.930E-01 40 1.013E-04 2.693E + 08 6.265E+ 09 6.008E-17 1.040E-12 9.930E-01 41 3.727E-05 3.120E + 08 6.577E+ 09 1.229E 16 1.041E-12 9.931E-01 42 1.068E-05 1.719E + 08 6.749E + 09 1.105E-16 1.041E-12 9.932E 01 43 5.043E-06 2.062E + 08 6.955E + 09 2.062E-16 1.041E-12 9.934E 01 44 1.855E 06 1353E +08 7.090E+ 09 2.080E-16 1.041E-12 9.936E-01 45 8.764E 07 1.177E + 08 7.20SE+ 09 2.640E-16 1.041E-12 9.939E-01 46 4.140E-07 2.251E 4 08 7.433F + 09 8.818E-16 1.042E-12 9.947E 01 47 1.000E 07 5.923E4 03 8.025E+ 09 5.537E-15 1.048E.12 1.000E + C0 36 I
Table 26 Calculated saturated reaction rates at the O' dosimeters location, 5.08 cm behind insulation (R = 257.0 cm), at a height 20.3 cm below the core midplane for Cycle 13 Reaction per atom per second W/ENDF/B-VI Fe-O-Ha W/ ORIGINAL SAILOR' Dosimeter R = 257.00 cm R = 257.00 cm 5'Fe(n,p) 5'Mn 2.911E-17 2.330E 17 Ni (n,p) 5sCo 4.059E-17 3.260E-17 58 63Cu(n,a) "Co 4.190E-19 3.668E-19 Np(n,f)'3'Cs 2.008E-15 1.632E-15 237 238U (n,f) 37Cs 1.439E-16 1.164E-16 Ti (n p) '6Sc 5.571E-18 4.654E-18 "Co(n,y) "Co 6.029E-14 5.108E-14 Flux & (E) cm.2,3.i & (E > 1.0 MeV) 4.569E+ 08 3.672E+08 $ (E > 0.1 MeV) 4.061 E +09 3.309E+09 ~ $ (E < 0.4 eV) 1.220E+09 1.093E+ 09 DPA rate dpa/s Total DPA rate 1.478E-12 1.204E-12 s ENDF/B-VI dosimetry cross sections were used in both calculations. l 1 4 9 37
= Table 27 Calculated saturated reaction rates at the dosimeters location, center of the instrument tube, at a height 20.3 cm below the core midplane for Cycle 13 Reaction per atom per second W/ENDF/B-VI Fe-O-H' W/ ORIGINAL SAILOR' Dosimeter R = 321.63 cm R = 321.63 cm "Fe(n,p)"Mn 2.379E-17 1.904E-17 "Ni (n,p) "Co 3.325E-17 2.670E-17 Cu(n,a) "Co 3.337E-19 2.928E-19 ^ 237Np(n,f)'87Cs 1.706E-15 1.383E-15 "U (n,f) 7Cs 1.195E-16 9.652E-17 "Ti (n,p) "Sc 4.472E-18 3.742E-18 "Co(n,y) "Co 5.251E-14 4.417E-14 Flux 4 (E) n cm.2 3-i 4 (E > 1.0 MeV) 3.791E+08 3.042E+08 4 (E > 0.1 MeV) 3.579E+09 2.904E+09 4 (E < 0.4 eV) 9.054E+08 8.174E+08 DPA dpa/s Total DPA 1.288E-12 1.048E-12 ' ENDF/B-VI dosimetry cross sections were used in both calculations. 38 l
Table 28 Absolute flux spectrum calculated with ENDF/B-VI Fe-O-H, behind RPV insulation at 0*,R = 257.0 cm, and 121.19 cm above the core midplane Upper Group Cumulative Group Cumulative Cumulative Group energy flux flux DPA DPA DPA fraction 1 1.733E +01 1389E+ 05 1389E+05 4.058E-16 4.058E-16 3/71E 04 2 1.419E + 01 4.950E+05 6339E+05 1310E-15 1.716E-15 1.467E-03 3 1.221E +01 1.509E+ 06 2.143E+06 3.634E-15 5350E-15 4.576E-03 4 1.000E+ 01 2.715E+ 06 4.858E+ 06 6.025E-15 1.137E-14 9.728E-03 5 8.607E+ 00 3.848E + 06 8.705E +06 8.030E-15 1.940E-14 1.660E-02 6 7.408E + 00 7.709E+06 1.641E+ 07 1.498E-14 3.438E-14 2.941E 02 7 6.065E+ 00 9.656E+06 2.607E+07 1.723E-14 5.161E 14 4.414E-02 8 4.966E + 00 1.557E +07 4.164E+07 2.446E-14 7.607E-14 6.506E.02 9 3.679E+ 00 1.1%E+ 07 5360E t07 1.638E-14 9.245E-14 7.907E-02 10 3.012E+00 9.288E+ 06 6.289E + 07 1.180E-14 1.043E-13 8.917E-02 11 2.725E+ 00 1.131E+ 07 7.419E+07 1.442E 14 1.187E-13 1.015E-01 12 2.466E + 00 5.721E+06 7.991E+ 07 6.705E 15 1.254E-13 1.072E-01 13 2365E+00 1.795E+ 06 8.171E+07 1.967E-15 1.273E-13 1.089E-01 14 2346E+00 9_004E+ 06 9.071E + 07 9373E 15 1367E 13 1.1690-01 15 2.231E+ 00 2.510E + 07 1.158E+ 08 2.595E-14 1.627E-13 1391E-01 16 1.920E+ 00 3.660E+ 07 1.524E+08 2.976E-14 1.924E-13 1.646E-01 17 1.653E+ 00 5.928E + 07 2.117E+ 08 4.80M-14 2.405E-13 2.057E-01 18 1353E+00 1.506E+08 3.623E+ 08 8.A 40E-14 3.249E-13 2.779E-01 19 1.003E +00 1.459E+08 5.082E+ 08 5348E-14 3.783E-13 3.236E 01 20 8.208E-01 6.634E + 07 5.746E+ 08 3.721E-14 4.156E-13 3.554E-01 21 7.427E-01 3375E+08 9.121E+ 09 1.22)E-13 5375E 13 4.598E 01 22 6.081E-01 2.989E +08 1.211E+ 09 8.797E 14 6.255E-13 5350E-01 23 4.979E-01 3.451E+08 1.556E+09 1368E-13 7.623E-13 6.520E-01 24 3.688E-01 4.783E+ 08 2.035E+ 09 9.895E-14 8.613E-13 7366E-01 25 2.972E.01 6.126E+ 08 2.647E+ 09 1.228E-13 9.841E-13 8.417E-01 26 1.832E-01 5.552E +08 3.203E+ 09 7.829E-14 1.062E-12 9.086E-01 27 1.111E.01 3.633E + 08 3366E+09 4.697E-14 1.109E-12 9.488E-01 28 6.738E-02 2.604E + 08 3.826E + 09 1.687E-14 1.126E-12 9.632E-01 29 4.087E.02 8.959E + 07 3.916E +09 7.251E-15 1.133E-12 9.694E-01 30 3183E-02 4.714E + 07 3.%3E + 09 1340E-14 1.147E-12 9.809E 01 31 2.606E-02 1.571E +08 4.120E + 09 3.166E-15 1.150E-12 9.836E-01 22 2.418E-02 9389E+07 4.214E+ 09 4.112E-16 1.150E-12 9.840E.01 33 2.188E-02 1.806E + 08 4394E+09 1.479E-15 1.152E-12 9.852E-01 34 1.503E 02 2 460E+08 4.641E + 09 4.615E-15 1.157E-12 9.892E-01 35 7.102E-03 2.491E+ 08 4.890E+ 09 2.216E-15 1.159E-12 9.911E-01 36 3355E-03 2.158E+ 08 5.105E+ 09 7.597E-16 1.160E-12 9.917E-01 37 1.535E-03 3316E+08 5.437E + 09 5.578E-16 1.160E-12 9.922E-01 38 4.540E.04 1.648E+ 08 5.602E+09 1.625E-17 1.160E-12 9.922E-01 39 2.144E-04 1.705E +08 5.772E + 09 2.446E-17 1.160E-12 9.922E 01 40 1.013E-04 2.148E + 08 5.987E + 09 4.793E 17 1.160E-12 9.923E 01 41 3.727E-05 2.474E+ 08 6.235E +09 9.747E-17 1.160E-12 9.924E-01 42 1.068E 05 1362E+08 6371E+09 8.753E-17 1.160E 12 9.924E-01 43 5.043E-06 1.625E+ 08 6.533E+ 09 1.625E-16 1.161E-12 9.926E-01 44 1.855E-06 1.064E + 08 6.640E + 09 1.636E-16 1.161E-12 9.927E 01 45 8.764E.07 9.503E+ 07 6.735E + 09 2.132E-16 1.161E-12 9.929E-01 46 4.140E-07 1.166E+ 08 6.851E + 09 4.567E-16 1.161E-12 9.933E-01 47 1.000E.07 8.400E+ 08 7.691E +09 7.853E-15 1.169E-12 1.000E +00 39
Table 29 Absolute Dux spectrum calculated with ENDF/B-VI Fe-O-H, at the center of the instrumentation tube, R = 321.63 cm,0* and 121.9 cm above the core midplane Upper Group Cumulative Group Cumulative Cumulative Group enerry flux flux DPA DPA DPA fraction 1 1.733E + 01 1.017E +05 1.017E +05 2.970E-16 2.970E-16 3.077E-04 2 1.419E+01 3.632E+05 4.649E+ 05 9.611E-15 1.258E-15 1303E 03 3 1.211E+01 1.11CE+06 1.575E +06 2.675E-15 3.933E-15 4.075E-03 4 1.000E+01 2.002E+ 06 3.577E + 06 4.441E-15 8375E-14 8.676E-03 5 8.607E+00 2 846E+06 6.423E +06 5.939E-15 1.431E-14 1.483E-02 6 7.408E4 00 5.719E+ 06 1.214E+ 07 1.111E-14 2542E-14 2.634E-02 7 6.065E+00 7.280E+06 1.942E + 07 1.299E-14 3 841E-14 3.980E-02 8 4.966E + 00 1.202E+07 3.144E + 07 1.888E-14 5.729E-14 5.935E-02 9 3.679E+00 9354E+06 4.079E + 07 1.282E-14 7.010E-14 7.263E-02 10 3.012E+ 00 7.257E+06 4.805E+ 07 9.224E-15 7.933E-14 8.219E-02 11 2.725E+ 00 8.946E+06 5.699E+ 07 1.141E-14 9.073E-14 9.400E-02 12 2.466E+ 00 4.556E +06 6.155E+07 5339E-15 9.607E-14 9.953E-02 13 2365E+00 1.407E +06 6.296E + 07 1.542E-15 9.762E 14 1.011E-01 14 2346E+00 7.118E+ 06 7.008E +07 7.409E-15 1.050E-13 1.088E-01 15 2.231E + 00 1.953E+ 07 8.960E+ 07 2.019E-14 1.252E-13 1.297E-01 1 16 1.920E + 00 2.846E + 07 1.181E + 08 2315E-14 1.484E-13 1.537E-01 17 1.653E + 00 4.702E +07 1.651E+ 08 3.810E-14 1.865E-13 1.932E 01 18 1353E+00 1.172E + 0B 2.823E + 08 6.567E 14 2.521E-13 2.612E-01 19 1.003E + 00 1.132E+ 08 3.955E + 08 4.148E-14 2.939E-13 3.042E-01 20 8.208E-01 5.865'6 +07 4341E+08 3.290E-14 3.265E-13 3383E 01 21 7.427E-01 2.665E + 0C 7.207E+08 9.632E-14 4.228E-13 4381E.01 22 6.081E-01 2.446E+ 08 9.653E+ 08 7.200E-14 4.948E-13 5.127E-01 l 23 4.979E41 2.734E+08 1.239E+ 09 1.084E-13 6.032E-13 6.250E-01 24 3.688E-01 3.795 E + 08 1.618E+ 09 7.849E-14 6.817E-13 7.063E 01 25 2.972E-01 5.545E + 08 2.173E + 09 1.112E-13 7.929E-13 8.215E-01 26 1.832E-01 5.098E +08 2.683E+ 09 7.188E-14 8.648E-13 8.959E-01 27 1.111E-01 3370E+08 3.020E + 09 4357E-14 9.084E-13 9.411E-01 28 6.738E-02 2.509E + 08 3.270E+ 09 1.625E-14 9.266E-13 9.579E.01 29 4.087E-02 9.000E + 07 3360E+09 7.283E-15 9319E-13 9.655E-01 30 3.183E-02 4.913E + 07 3.410E +09 1397E-14 9.459E-13 9.799E-01 31 2.606E-02 1.245E+08 3.534E + 09 2.510E-15 9.484E-13 9.825E-01 32 2.418E 02 7.859E+07 3.613E + 09 3.441E-16 9.487E-13 9.829E-01 33 2.188E-02 1.659E+08 3.779E +09 1359E-15 9.501E-13 9.843E-01 34 1.503E-02 2.513E + 08 4.030E+ 09 4.715E 15 9.548E-13 9.892E-01 35 7.102E-03 2.530E + 08 4.283E + 09 2.250E-15 9.570E-13 9.915E 01 36 3355E-03 2311E+08 4.514E + 09 8.133E 16 9.578E-13 9.923E.01 37 1385E-03 3.520E +08 4.866E+09 5.921E 16 9.584E-13 9.930E-01 38 4.540E-04 1.806E+ 08 5.047E+ 09 1.781E-17 9384E-13 9.930E 01 39 2.144E 04 1.862E + 08 5.233E + 09 2.671E-17 9385E-13 9.930E-01 40 1.013E-N 2357E+08 5.469E + 09 5.258E-17 9.585E-13 9.931E-01 41 3.727E-05 2.729E+08 5.742E+ 09 1.075E-16 9386E-13 9.932E-01 42 1.068E.05 1.504E + 08 5.892E + 09 9.((s9E-17 9.587E 13 9.933E-01 43 5.043E-06 1.800E + 08 6.072E + 09 1.800E-16 9389E-13 9.934E-01 l 44 1.855E 06 1.172E+08 6.189E+ 09 1.801E-16 9.591E-13 9.936E-01 45 8.7ME-07 1.024E+ 08 6.291E+ 09 2.298E-16 9.593E 13 9.939E+ 01 46 4.140E47 9.565E +07 6387E+09 3.748E-16 9.597E-13 9.943E-01 47 1.000E-07 5.928E+ 08 6.980E + 09 5.541 E-15 9.652E-13 1.000E + 00 l 40
Table 30 Absolute flux spectrum calculated with ORIGINAL SAILOR, behind RPV insulation at 0, R = 257.0 cm, and i21.9 cm above the core midplane Upper Group Cumulative Gmup Cumulative Cumulative Group enerry flux flux DPA DPA DPA fraction 1 1.733E + 01 1.144E + 05 1.144E+ 05 3342E 16 3342E-16 3300EN 2 1.419E+ 01 4.751E+05 5.895E+ 05 1.257E-15 1.591E-15 1.667E.03 3 1.211E + 01 1.430E+ 06 2.019E + 06 3.444E-15 5.035E-15 5.273Em 4 1.000E+ 01 2.429E+ 06 4.448E+ 06 5390E-15 1.042E-14 1.092E-02 5 8.607E+ 00 3.408E + 06 7.856E+06 7.113E-15 1.754E-14 1.837E-02 6 7.408E +00 6.540E+ 06 1.440E+ 07 1.271E-14 3.024E-14 3.167E-02 7 6.065E +00 7.200E + 06 2.160E+ 07 1.285E-14 4309E-14 4.513E-02 8 4.966E+ 00 1.109E+ 07 3.269E+ 07 1.743E-14 6.05'2E-14 6338E.02 9 3.679E+ 03 9.442E + 06 4.213E + 07 1.294 E-14 7.145E-14 7.693E42 10 3.012E + 00 7.676E+ 06 4.981E + 07 9.756E-15 8321E-14 8.714E-02 11 2.725E + 00 9341E+06 5.915E+ 07 1.191E-14 9312E-14 9 962E-02 12 2.466E+ 00 4.676E + 06 6383E+07 5.481E 15 1.006E-13 1.054E-01 13 2365E+ 00 1.607E+ 06 6343E+07 1.761E 15 1.024E-13 1.072E 01 14 2346E+00 7.891E+06 7332E+07 8.215E 15 1.106E-13 1.158E-01 15 2.231E+ 00 1.943E + 07 9.275E+ 07 2.009E 14 1307E-13 1368E-01 16 1.920E +00 3.082E+ 07 1.236E + 08 2.506E 14 1.557E-13 1.631E-01 17 1.653E+ 00 4.633E + 07 1.699E+ 08 3.754E-14 1333E 13 2.024E 01 18 1353E+ 00 1.214E+ 08 2.912E+ 08 6.799E-14 2.613E-13 2.736E-01 19 1.003E + 00 1.272E+ 08 4.185E+ 08 4.664E-14 3.079E-13 3.225E-01 20 8.20SE-01 5.142E + 07 4.699E + 08 2.884 E-14 3367E-13 3.527E 01 21 7.427E 01 2.648E+ 0S 7347E+ 08 9.569E 13 4324E-13 4.529E-01 22 6.081E.01 2.535 E+ 08 9.882E+08, 7.461E-14 5.070E 13 5310E-01 23 4.979E-01 2.645E + 08 1.253E + 09 1.049E-13 6.119E-13 6.40SE-01 24 3.688E41 3.890E+ 08 1.642E + 09 8.057E-14 6.925E 13 7.252E-01 25 2372E-01 4.656E+08 2.108E + 09 9335E-13 7.858E-13 8.230E-01 26 1.832E-01 5.096E + 08 2.618E+ 09 7.186E 14 8377E-13 8.982E-01 27 1.111E-01 3306E+ 08 2.948E4 09 4.275E 14 9.004E-13 9.430E-01 28 6.738E-02 2367E+ 08 3.205E +09 1.663E 14 9.170E-13 9.604E-01 29 4.087E-02 7.973E + 07 3.285E + 09 6.453E-15 9.235E-13 9.672E 01 30 3.183E-02 3.776E+ 07 3322E+09 1.074E-14 9342E-13 9.784 E-01 31 2.606E.02 1.664E + 08 3.489E + 09 3355E-15 9376E-13 9.819E 01 32 2.418E-02 1.218E+ 08 3.611E + 09 5334E-16 9381E-13 9.825E-01 33 2.188E 02 2.116E+ 08 3.822E + 09 1.733E-15 9399E-13 9.843E-01 34 1.503 E-02 2.260Ei O8 4.048E+ 09 4.240E-15 9.441E-13 9.887E-01 35 7.102E-03 2.215E + 08 4.270E+ 09 1.970E-15 9.461E-13 9.908E-01 36 3355E-03 1.841E + 08 4.454E + 09 6.482E-16 9.467E-13 9315E.01 37 1.585E-03 2.793E + 08 4.733 E + 09 4.697E 16 9.472E-13 9.920E-01 38 4340E-04 1.428E +08 4R76E+ 09 1.408E-17 9.472E-13 9.920E-01 39 2.144E-04 1.472E+ 08 5.023E + 09 2.110E-17 9.472E 13 9.920E-01 40 1.013EN 1.862E + 08 5.209E+ 09 4.153E-17 9.473E-13 9.920E-01 41 3.727E-05 2.151E+ 08 5.424E + 09 8.473E-17 9.473E-13 9921E-01 42 1.068E-05 1.185E+ 08 5.543E + 09 7.615 E-17 9.474E-13 9.922E-01 43 SDt3E-06 1.423E+ 08 5.685E + 09 1.423E-16 9.476E-13 9924E41 44 1.855E.06 9.413E + 07 5.779E + 09 1.447E-16 9.477E 13 9.925E.01 45 8.764E-07 8323E+07 5.862E+ 09 1.868E 16 9.479E-13 9927E-01 46 4.140E-07 1.863E + 08 6.049E+ 09 7.298E-16 9.486E 13 9.935E-01 47 1.000E-07 6.671E + 08 6.716E + 09 6.236E-15 9348E-13 1.000E+00 41
Table 31 Absolute flux spectrum calculated with ORIGINAL SAILOR at the center of the instrumentation tube, R = 321.63 cm,0* and 121.9 cm above the core midplane Upper Group Cumulative Group Cumiative Cumulative Group energy flux flux DPA DPA DPA fraction 1 1.733E +01 8.415E + 04 8.415E + 04 2.459E-16 2.459E-16 3.125E-04 2 1.419E+01 3.486E+ 05 4328E+05 9.225E-16 1.168E-15 1.485E-03 3 1.211E + 01 1.054E + 06 1.487E+ 06 2.540E-15 3.708E-15 4.712E-03 4 1.000E+ 01 1.793E+06 3.280E+06 3.978E-15 7.686E-15 9.767E-03 5 8.607E + 00 2.528E + 06 5.808E + 06 5.276E-15 1.296E-14 1.647E-02 6 7.408E+ 00 4.872E + 06 1.068E + 07 9.466E-14 2.243E-14 2.850E 02 7 6.065E+ 00 5.443E+ 06 1.612E +07 9.711E-15 3.214E-14 4.084E-02 8 4.966E+ 00 8.565E+ 06 2.469E+ 07 1346E-14 4.559E-14 5.793E-02 9 3.679E+ 00 7352E+ 06 3.204E+ 07 1.007E-14 5.567E-14 7.073E-02 10 3.012E+ 00 5.983E+ 06 3 802E+07 7.604E-15 6327E-14 8.040E-02 11 2.725E + 00 7364E+ 06 4.539E+ 07 9389E-15 7.266E-14 9.233E-02 12 2.466E + 00 3.729E + 06 4.911E + 07 4.370E-15 7.703E-14 9.788E-02 13 2365E+ 00 1.253E+ 06 5.037E+07 1373E-15 7.840E-14 9.%2E.02 14 2346E+00 6.219E+ 06 5.659E+ 07 6.474E-15 o.488E-14 1.079E-01 15 2.231E+ 00 1.507E+ 07 7.166E+07 1.558E-14 1.005E 13 1.276E-01 15 1.920E + CO 2381E+07 9.546E + 07 1.936E-14 1.198E-13 1.523E-01 l'i 1.653E+ :0 3.666E+ 07 1321E+ 08 2.971E-14 1.495E-13 1.900E-01 18 '353E+W) 9.453E+ 07 2.267E +08 5.296E-14 2.025E-13 2.573E 01 19 1003E+ 00 9.785E + 0 / 3.245E + 08 3 586E-14 2384E-13 3.029E 01 20 d.208E-01 4.554E+ 07 3.700E + 08 2.554E-14 2.639E 13 3353E-01 21 7.427E-01 2.082E +08 5.782E + 08 7.524E-14 3391E-13 4309E.01 22 6.081E-01 2.052E + 08 7.834E+ 08 6.040E-14 3.995E 13 5.077E 01 23 4.979E.01 2.150E + 08 9.984E + 08 8.521E-14 4.847E-13 6.160E-01 24 3.688E-01 3.046E + 08 1303E+ 09 6.299E-14 5.477E-13 6.960E-01 25 2.972E.01 4.213E+ 08 1.724E+ 09 8.447E-14 6322E-13 8.033E-01 '6 1.832E-01 4.576E + 08 2.182E+ 09 6.453E-14 6.%7E-13 8.853E-01 27 1.111E41 3.045 E + 08 2.486E +09 3.937E-14 7361E 13 9353E-01 28 6.738E-02 2388E+08 2.725E+09 1.547E-14 7.516E-13 9.550E-01 29 4.087E-02 7.833E+ 07 2.804E+ 09 6339E-15 7.579E-13 9.631E 01 30 3.183E-02 4.028E + 07 2.844E+ 09 1.145E 14 7.694E-13 9.776E-01 31 2.606E 02 1.274E+ 08 2.971E+ 09 2.568E-15 7.719E-13 9.809E-01 32 2.418E-02 9.512E + 07 3.066E+ 09 4.165E-16 7.723E-13 9.814E-01 33 2.188E-02 1.854E +08 3.252E+ 09 1.519E-15 7.739E 13 9.833E-01 I 34 1.503E-02 2.264E + 08 3.478E + 09 4.248E-15 7.781E-13 9.887E.01 I 35 7.102E-03 2.269E+ 08 3.705E + 09 2.018E-15 7.801E-13 9.913E-01 36 3355E03 1.970E + 08 3.902E+ 09 6.934E-16 7.808E 13 9.922E-01 37 1.585E-03 2.964E+ 08 4.198E+ 09 4.985E 16 7.813E-13 9.928E-01 [ 38 4.540E.04 1.568E+ 08 4355E+ 09 1.545E-17 7.813E-13 9.928E-01 39 2.144E-04 1.606E + 08 4.516E+ 09 2303E 17 7.814E-13 9.929E-01 (' 40 1,013E-04 2.041E+ 08 4.720E+ 09 4.554E-17 7.814E-13 9.929E-01 41 3.727E-05 2367E+08 4.957E + W 9324E-17 7.815E-13 9.930E-01 42 1.06SE 05 1305E+08 5.087E + 09 8387E-17 7.816E-13 9.931E-01 43 5.043E-06 1.567E +08 5.244E + 09 1.567E-16 7.817E-13 9.933E-01 44 1.855 E-06 1.029E + 08 5347E+ 09 1.582E-16 7.819E-13 9.935E-01 45 8.7ME-07 8.949E +07 5.436E+ 09 2.008E-16 7.821E-13 9.938E-01 46 4.140E-07 1.713E+ 08 5.608E + 09 6.712E-16 7.828E-13 9.946E-01 47 1.000E-07 4.506E +08 6.058E + 09 4.212E-15 7.870E-13 1.000E+ 00 h 42 \\ b
Table 32 Calculated saturated reaction rates at the 0* cavity dosimeters location, 5 cm behind insulation (R = 257.0 cm), at a height 121.90 cm above the core midplane for Cycle 13 Reaction per atom per second W/ENDF/B-VI Fe-O-H" ORIGINAL SAILOR" Dosimeter R = 257.00 cm R = 257.00 cm "Fe(n,p)"Mn 2.337E-17 1.872E-17 "Ni (n,p) "Co 3.256E-17 2.618E-17 63Cu(n,a) "Co 3.364E-14 2.947E-19 25'Np(n,f)"7Cs 1.592E-15 1.2%E-15 2"U (n,f) '27Cs 1.151E-16 9.324E-17 "Ti (n,p) "Sc 4.476E-18 3.742E-18 "Co(n,y) "Co 4.728E-14 4.003E-14 Flux 0 (E) n cm.2 s~' O (E > 1.0 MeV) 3.647E+08 2.933E+08 & (E > 0.1 MeV) 3.203E+09 2.618E+09 4 (E < 0.4 eV) 9.566E+08 8.534E+08 DPA dpa/s ~ Total DPA 1.169E-12 9.548E-13 ENDF/B-VI dosimetry cross sections were used in both calculations. 43
1 = Table 33 Calculated saturated reaction rates at the dosimeters location, center of the instrumentation tube, at a height 121.90 cm above the core midplane for Cycle 13 Reaction per atom per second W/ENDF/B-VI Fe-O-H' ORIGINAL SAILOR
- Dosimeter R = 321.63 cm R = 321.63 "Fe(n,p)"Mn 1.779E-17 1.425E-17 "Ni (n,p) "Co 2.488E-17 1.998E-17
~ Cu(n,a) "Co 2.491E-19 2.187E-19 237Np(n,f)'27Cs 1.279E-15 1.039E-15 238U (n,f) 237Cs 8.951E-17 7.232E-17 "Ti (n,p) "Sc 3.341E-18 2.797E-18 "Co(n,y) "Co ' 985E-14 3.354E-14 Flux c (E) n cm.2 3 .i o (E > 1.0 MeV) 2.841E+08 2.282E+ 08 & (E > 0.1 MeV) 2.683E+09 2.182E+ 09 4 (E < 0.4 eV) 6.884E+08 6.219E+08 DPA dpa/s Total DPA 9.652E-13 7.870E-13
- ENDF/B-VI dosimetry cross sections were used in both calculations.
44 1 i
i Table 34 Spectrum averaged cross section at the cavity dosimeter location,5 cm behind the RPV insulation, at 0* of Trojan Cycle 13 W/ENDF/B-VI Fe-O-H" W/ ORIGINAL SAILOR' E>1 E > 0.11 E>1 E > 0.11 Reaction & (barns)' & (barns)' & (barns)' & (barns)' 5dFe(n,p) 5dMn 6.370E-02 7.168E-03 6.346E-02 7.041E-03 58Ni (n,p) 58Co 8.882E-02 9.994E-03 8.878E-02 9.851E-03 "Cu(n,a) "Co 9.169E-04 1.032E-04 9.990E-04 1.108E-04 'Ti (n,p) "'Sc 1.219E-02 1.372E-03 1.267E-02 1.406E-03 23sU (n,f) 337Cs 3.149E-01 3.543E-02 3.171E-01 3.519E-03 237Np(n,f)'87Cs 4.394E +00 4.944E-01 4.446E-00 4.933E-01 5'Co(n,y) "Co 1.319E+02 1.485E-01 1.391E +02 1.544E-01 ENDF/B-VI dosimetry cross sections were used in both calculations. 6 ~ [o(E)$(E)dE = I a [4(E)dE 1 [o(E)$(E)dE ' o = [ 4(E)dE 0.11 45
r = Table 35 Spectrum averaged cross section at the center of the instrumentation tube at 0* and 20.3 cm below midplane of Trojan Cycle 13 W/ENDF/B-VI Fe-O-H" W/ ORIGINAL SAILOR
- E>1 E > 0.11 E>1 E > 0.11 Reaction 3(barns)6 u (barns)'
& (varns)6 3 (barns)M 5dFe(n.p) 5dMn 6.274E-02 6.646E-03 6.259E-02 6.557E-03 Ni (n,p) 58Co 8.771E-02 9.291E-03 8.776E-02 9.194E-03 58 Cu(n,a) "Co 8.801E-04 9.323E-05 9.626E-04 1.008E-04 63 "'Ti (n,p) "Sc 1.179E-02 1.249E-03 1.230E-02 1.288E-03 U(n,f)137Cs 3.153E-01 3.340E-02 3.173E-01 3.324E-03 23: 2nNp(n,f)'37Cs 4.501E+00 4.768E-01 4.546E-00 4.762E-01 "Co(n,y) "Co 1.385E +02 1.467E-01 1.452E+02 1.521E-01 ENDF/B-VI dosimetry cross sections were used in both calculations. a = [o(E)$(E)dE a.. 0.11 46 i I
Table 36 Non-saturation factors (h) for Trojan Cycle 13 Dosimeter h' 5'Fe(n.p) 5'Mn 3.780E41 ssNi (n,p) 58Co 6.501E-01 '3Cu(n,a) "Co 7.236E-02 5Co(n,y) "Co 7.236E-02 238U (n.f) '5'Cs 1.317E-02 237Np(n,f)'37Cs 1.317E-02
- Ti (n.p)'6Sc 6.411E-01
- h =[ P,(1 -e '*) e *'~9
) i l I 47 I i
f TABLE 37 C/E values for Trojan dosimetry, based on ENDF/B-V1 calculations Dosimetry set Reaction A B C D Cu (n,a) 1.39 1.31 1.53 1.58 63 'Ti (n.p) 1.30 1.21 1.42 1.47 "Fe (n,p) 1.50 1.41 1.68 1.76 Ni (n,p) 1.41 1.31 1.61 1.66 58 2"U (n,f) 1.40 1.33 1.51 1.64 Np (n.f) 1.37 1.33 1.24 1.67 237 "Co (n,y), Bare 2.42 1.85 1.94 1.84 U (n,f), Bare 4.435 2.% 2.41 1.82 235 TABLE 38 C/E values for Trojan dosimetry, based on original SAILOR calculations Dosimetry set Reaction A B C D Cu (n,a) 1.22 1.15 1.35 1.39 63 "Ti (n p) 1.09 1.01 1.19 1.23 "Fe (n.p) 1.20 1.13 1.34 1.41 S'Ni (n.p) 1.13 1.05 1.29 1.33 U (n,f) 1.13 1.07 1.22 1.32 238 Np (n,f) 1.12 1.08 1.01 1.35 237 "Co (n,y), Bare 2.05 1.57 1.63 1.54 U (n,f), Bare 3.71 2.49 1.99 1.51 235 48
Table 39 Absolute flux spectrum calculated with ENDF/B VI Fe-O-II at location of support structure (i.e., e = 23*,20 3 cm below core midplane,17.8 cm into the concrete) Upper Group Cumulative Group Cumulative Cumulative Gmup energy flux flux DPA DPA DPA fraction 1 1.733E + 01 2341E+04 2341E+04 6.841E-17 6.841 E-17 3370E44 2 1.419E+ 01 8.750E+ 04 1.109E+05 2315E-16 2.999E-16 1.477E-03 3 1.221 E + 01 2.568E + 05 3.677E + 05 6.186E 16 9.186E-16 4.524E-03 4 1.000E+01 4.908E + 05 8.586E + 05 1.089E 15 2.f08E-15 9.889E-03 5 8407E+ 00 7.144E + 05 1.573E + 06 1.491E-15 3.499E-15 1.723E-02 6 7.408E+00 1.761E+06 3334E+06 3.422E-15 6.920E-15 3.408E-02 7 6.065E + 00 2.474E +06 5.808E+ 06 4.413E-15 1.133E-14 5382E-02 8 4.966E + 00 432SE+06 1.014E +07 6.799E-15 1.813E-14 8.931E-02 9 3.679E+ 00 3.243E+ 06 1338E+07 4.443 E-15 2.258E-14 1.112E-01 10 3.012E+ 00 2 870E+ 06 1.625E + 07 3.648E-15 2.622E-14 1.292E-01 11 2.725E + 00 3.624E+ 06 1.987E + 07 4.621E-15 3.084E-14 1.519E-01 12 2.466E + 00 2344E+06 2.222E+ 07 2.748E-15 3359E-14 1h54E-01 13 2365E+ 00 7.911E + 05 2301E+ 07 8.671 E-16 3.446E-14 1.697E-01 1 14 2346E+00 3.414E +06 2.642E + 07 3.554E-15 3.801E-14 1.872E-01 15 2.231 E + 00 6.556E + 06 3.298E+ 07 6.779E-15 4.479E-14 2.206E-01 16 1.920E+ 00 7.466E + 06 4.044E + 07 6.072E-15 5.086E-14 2305E-01 17 1.653E+ 00 1.213E+07 5.257E+ 07 9.828E-15 6.069E-14 2.989E-01 18 1353E+00 1.620E + 07 6.877E+ 07 9.077E-15 6.977E-14 3.436E-01 19 1.003E + 00 1.098E + 07 7.975E + 05 4.023E-15 7379E-14 3.634E-01 20 8.208E-01 1.020E + 07 8.995 E + 07 5.723E-15 7.952E 14 3.916E-01 21 7.427E-01 3.089E+ 07 1.208E+ 08 1.116E-14 9.068E-14 4.466E-01 22 6.081E-01 3.483E+ 07 1.557E+08 1.025E-14 1.009E-13 4.971E-01 23 4.979E-01 2.775E+ 07 1.834E + 08 1.100E-14 1.119E-13 5313E-01 24 3.688E-01 2.920E + 0'7 2.126E+08 6.039E-15 1.180E-13 5.810E-01 25 2.972E41 6.129E +07 2.739E + 08 1.229E-14 1303E-13 6.415E41 26 1.832E-01 8.804E + 07 3.620E+ 08 1.241E-14 1.427E-13 7.027E-01 27 1.111E-01 7.829E+ 07 4.402E + 08 1.012E-14 1.528E-13 7.525E-01 28 6.738E-02 7h39E+ 07 5.166E+08 4.948E-15 1.577E-13 7.769E 01 29 4.087E-02 3.487E+ 07 5.515E+ 08 2.822E 15 1.606E-13 7.908E-01 30 3.183E-02 2.8%E+ 07 5.805E + 08 8.234 E-15 1.688E-13 8314E-01 31 2.606E-02 1.173E+ 07 5.922E + 08 2366E-16 1.690E 13 8325E-01 32 2.418E-02 1.520E+ 07 6.074E + 08 6.655E-17 1.691E-13 8329E-01 33 2.188E-02 5380E+ 07 6 632E+08 4.570E-16 1.6%E-13 8351E-01 34 1.503E-02 1.126E+ 08 7.758E + 08 2.113E-15 1.717E-13 8.455E-01 35 7.102E-03 1.105E+08 8.864E +08 9.832E-16 1.727E-13 8.504E-01 36 3355E-03 1.189E + 08 1.005E +09 4.185E-16 1.731E-13 8.524E-01 37 1.585 E-03 2.096E+ 08 1.215E+ 09 3325E-16 1.734E-13 8.542E-01 38 4.540E-04 1.296E+ 08 1344E+09 1.277E-17 1.734E-13 8342E-01 39 2.144E-04 1335E+08 1.478E + 09 1.915 E-17 1.735 E-13 8.543E 01 40 1.013E-04 1.817E+ 08 1660H + 09 4.054E-17 1.735 E-13 8.545E-01 41 3.727E-05 2308E+08 1.891E +09 9.091E-17 1.736E-13 8.550E-01 42 1.068E-05 1392E+ 08 2.030E + 09 8.947E-17 1.737E-13 8.554E-01 43 5.043 E-06 1.848E+ 08 2.214E+ 09 1.848E-16 1.739E-13 8.563E 01 44 1.855E46 1364E+08 2351E+09 2.097E-16 1.741E-13 8.573E.01 45 8.764E.07 1352E+ 08 2.486E + 09 3.034E-16 1.744E-13 8.588E-01 46 4.140E-07 1.973E + 08 2483E+09 7.731E-16 1.751E-13 8.626E-01 47 1.0(X)E-07 2.983E+ 09 5.667E+ 09 2.789E-14 2.030E-13 1.000E+ 00 49
Table 40 Calculated saturated i: action rates at the location of support structure (i.e.,0 = 23*,20.3 cm below midplane,18.0 cm into the concrete) Reaction per atom per second W/ENDF/B-VI Fe-O-H ORIGINAL SAILOR Dosimeter R = 325.76 cm R = 325.76 "Fe(n,p) "Mn 5.908E-18 4.928E-18 58Ni (n,p) 58Co 8.059E-18 6.683E-18 65Cu(n,a) "Co 6.715E-20 6.280E-20 237Np(n,f)'37Cs 2.919E-16 2.283E-16 U (n,f) 37Cs 2.777E-17 2.205E-17 23 Wi (n,p) "Sc 9.831E-19 8.775E-19 "Co(n,y) "Co 1.20SE-13 9.873E-14 Flux & (E) cm.2 3.i 4 (E > 1.0 MeV) 6.895E+07 5.383E+07 & (E > 0.1 MeV) 3.620E+08 2.723E+08 4 (E < 0.4 eV) 3.181E+09 2.686E+09 DPA dpa/s Total DPA 2.030E-13 1.581E-13 50
Table 41 Calculated results of neutron exposure at Trojan support structure location (i.e., 0 = 23 *, AND 20.3) I ) 12.7 cm into 17.78 cm into 22.86 cm into PARAMETERS Concrete Concrete Concrete 0 (E > 1.0 MeV), cm 2. s 8 1.160E+8 6.895E+7 4.242E+7 0 (E > 0.1 MeV), em-2. s 7.759E+8 3.620E+8 1.828E+8 0 (E < 0.4 eV), cm-2, 3 8 3.486E+9 3.181E+9 2.590E+9 DPA rate (displacements per atom per s.) 3.710E-13 2.030E-13 1.187E 13 i t Fluence' (Cycle-13), cm'2. s-3 2.117E+ 15 1.258E+ 15 7.742E+ 14 Fluence' (32 EFPY), cm-2. s' 1.173E+17 6.971E+ 16 4.289E+ 19
- Based on flux above 1.0 MeV and 211.25 full. power days.
- Based on Dux above 1.0 MeV and 32 EFPY = 11,6% full-power days, l
i e 51
a 6 CONCLUSIONS Results from neutron transport calculations based on two different sets of Fe-H-O cross-section data were found to differ by about the same amount in the Trojan analysis as observed in earlier studies comparing the newer ENDF/B-VI cross sections with the earlier multigroup data in the SAILOR library (based on ENDF/B-IV). The cavity dosimetry results for high-threshold reactions were calculated to be about 30% higher with the ENDF/B-VI cross sections, which is corisistent with the change recently found in the analysis of the H. B. Robinson PWR," and projected (based on one-dimensional calculation 1) for the Arkansas Nuclear One, Unit 1 PWR.8 A recent re-analysis of the PCA experiment performed earlier at ORNL also indicated that ENDF/B-VI would increase the neutron transmission through the simulated RPV in that system." However, unlike all of the earlier studies, the higher absolute reaction rates in the Trojan cavity results do no_t improve agreement with the experimental measurements, but rather give worse C/E values than obtained with the ORIGINA.L SAILOR library data. The midplane C/E values for the threshold dosimeters located behind the insulation were found to be from 1.01 to 1.15 with the ORIGINAL SAILOR, and from 1.21 to 1.41 with the ENDF/B-VI cross sections. At the instrament tube location the ORIGINAL SAILOR library gives C/E values at the midplane of 1.23 to 1.41 for threshold reactions and ENDF/B-VI gives 1.47 to 1.76. The dosimeters in the instrument tube are more difficult to calculate accurately because they are located behind an ex-core detector. The C/E values for the fast-flux dosimeters are typically 5 to 10% higher at the axial position j 121.9 cm above the midplane, compared with values at 20.3 cm below the midplane. Both cross-section libraries substantially overestimate the thermal reaction dosimeters at I all locations, by up to a factor of 3 near the midplane. It is not known at this time why the present Trojan plant should behave in a qualitatively different manner than the other systems that have been analyzed with the ENDF/B-VI data. The experimental values were analyzed by two independent organizations, and the model used in the transport calculations has been extensively checked by LSU as well as Westinghouse personnel. Use of a generic rather than a plant-specific pinwise power distribution in the calculations would tend to cause the computed results be too high due to an over estimate of the core leakage. The magnitude of this effect is not known, since the plant-specific distribution is unavailable. The approximation of assuming 45 symmetry in the transport calculations was tested by comparing these with results from a more rigorous 90 symmetrical model, and no substantial discrepancies were observed. An inconsistency in the dosimeier location could also cause a discrepancy of this magnitude, but no errors have been identified at this time. Because of the presence of the ex-core detector and detector wells, the flux synthesis approximation could be introducing some errors into the transport calculations for the dosimeters located in the instrument tube. A three-dimensional transport calculation of the cavity region could help to resolve this question. ( mparison with recent in-vessel dosimeter measurements would also be very
- Personal Comrnunications, L F. Miller and F. B. Kam, ORNL 52
1 useful in identifying the causes for the discrepancies between the computed and measured absolute reaction rates in Cycle 13. Surveillance dosimeters, still contained within the Trojan plant, could be analyzed for this purpose after the reactor is shut down. Note that if these ENDF/B-VI transport calculations are compared with the earlier surveillance capsule dosimeter measurements for Capsule X (removed from reactor in 1984), then C/E values less than unity are obtained, whereas the cavity measurements in this study give C/E considerably more than unity. Perhaps this inconsistency is merely the consequence of having a significantly different power distribution for Cycle 13 compared with the power distributions appropriate for Capsule X exposure. On the other hand, it could also indicate an inconsistency between the two sets of dosimeter measurements performed at different times. If the causes for the discrepancy between the calculated and measured results cannot be identified and eliminated, then a least-squares adjustment code could be utilized to unfold an absolute spectrum by combining the dosimeter measurements and transport calculations. However, this step was not performed in the present study. It is important to note that both the ORIGINAL SAILOR and the ENDF/B VI SAILOR library produce very similar values for the spectrum-averaged cross-sections of all the dosimeters. Thus, the discrepancies observed in the absolute reaction rates are mainly due to a difference in the overall flux magnitude (i.e., both libraries predict a similar neutron energy distribution). The C/E values for the various fast-neutron dosimeters are also fairly consistent, again indicating a possible discrepancy in the magnitude of the leakage from the core. Without additional information to resolve the apparent discrepancy between the calculated and measured dosimeters, the higher of the two transport calculations (i.e., ENDF/B VI) has been used to estimate the fast-neutron exposure to the support structure. The fast flux (E > 1 Mev) at the critical point of maximum stress in the support structure (-18 cm into concrete shield) is computed to be 6.90E+07 2 neutrons /cm -s based on the unadjusted ENDF/B-VI calculations. This valr: is about 30% higher than obtained with the ORIGINAL SAILOR library. The thermal-to-fast-flux ratio at this point is calculated with both sets of cross-section data to be about 46. The thermal neutron dosimeter measurements indicate that the transport calculations tend to significantly overestimate the magnitude of thermal reaction rates in the cavity; thus, this value is probably an overestimate. The DPA rate is calculated to be about 2.0E-13 displacements per atom per second. Based on these flux and DPA rates the 2 cumulative fast fluence after 32 EFPY is found to be 7.0E+16 neutrons /cm, and the cumulative displacements per atom is 2.0E-04. 53
= 7 REFERENCES 1. L K. Mansur, K. Farrell, "On Mechanisms by Which a Soft Neutron Spectrum May Induce Accelerated Embrittlement,"I. NucL Mater. 170, 236-245 (1990). 2. S. E. Yanichko, S. L Anderson, W. T. Kaiser, Analysis of Capsule Xfrom Ponland General Electric Company Trojan Reactor Vessel Radiation Surveillance Program, WCAP 10861, Westinghouse, June 1985. 3. J. M. Chicots et al., Analysis of Capsule Vfrom Ponland General Electric Company, 7>ojan Reactor Vessel Radiation Surveillance Programs, Rev. 2, WCAP 12868, Westinghouse, December 1991. 4. P. Chowdhury, M. L Williams, and F.B.K. Kam, Development of a Diree-Dimensional Flia Synthesis Pmgram and Comparison with 3-D Transport 771cory Results, NUREG/CR-4984, ORNIlrM-10503, U.S. Nuclear Regulatory Commission, Washington, D.C., January 1988. 5. G. L Simmons and R. Roussin, " SAILOR - Coupled, Self-shielded,47 Neutron, 20 Gamma-ray, P, Cross Se: tion Library for Light Water Reactors," RSIC Data 3 Library Collection (DLC-7:), Radiation Shielding Information Center, Oak Ridge, Tennessee,1985. 6. R. W. Roussin et al. VITAMIN-C: llie CTR Pmcessed Multigroup Cross-Section Libraryfor Neurmnics Studies, ORNIJRSIC-37, Oak Ridge National Laboratory, Oak Ridge, Tennessee,1980. 7. C. Y. Fu et al., "ENDF/B-VI Iron: Improvements over ENDF/B-V and Impacts on Pressure Vessel Surveillance Dosimetry," presented at the 7th ASTM-EURATOM Symposium on Reactor Dosimetry, Strasbourg, France, August 26-31, 1990. 8. M. L Williams et al., " Transport Calculation of Neutron Transmission thrcugh Steel Using ENDF/B-V, Revised ENDF/B-V and r:NDF/B-VI Iron Evaluation," Ann. NucL Energy, 18 (10) 549-565 (1991). 9. R. E. Maerker et al., RSIC Code Package PSR-277/LEPRICON, Radiation Shielding Information Center, Oak Ridge, Tennessee, March 1990. .54 l 1 i
i 1 10. M. L Williams et al., The ELXSIR Cross-Section Libraryfor LWR Pressure VesselIrradiation Studies: Part of the LEPRICON Computer Code System, EPRI NP-3654, Electric Power Research Institute, Palo Alto, California,1984. 11. F. H. Ruddy et al., Reactor Cavity Neutron Dosimetry Results for the Trojan Nuclear Power Plant, Westinghouse STC, STC Report 93-9TDO-TROJN-R1, March 19,1993. 12. R. E. Maerker, LEPRICON Analysis of Pressure Vessel Surveillance Dosimetry inserted into H. B. Robinson-2 dwing Cycle 9, NUREGICR-4439, ORNI/TM-10132, U.S. Nuclear Regulatory Commission, Washington, D.C., August 1986. 13. M. L Williams, M. Asgari, F.B.K. Kam, Impact of ENDF/B-VI Cross-Section Data on H. B. Robinson Cycle 9 Dosimetry Calculation, NUREGICR-6071, ORNI/FM-10132, Oak Ridge National Laboratory, Oak Ridge, Tennessee, October 1993. 14. F. H. Ruddy and E. D. McGarry, " Benchmark Referencing of Solid State Track Recorder Neutron Dosimeters in Standard Neutron Fields," in Proceedings of the 7th ASTM-Euratom Symposium on Reactor Dosimetry, Strasbourg, France, August 27-31, 1990, Kluwer Academic Publishers, London,1992, p. 825. l 55
= l APPENDIX A EXPERIMENTAL MEASUREMENTS OF DOSIMETER ACTIVITIES FOR TROJAN CYCLE 13 BENCHMARK REFERENCING AND PRELIMINARY ANALYSFS OF THE NEUTRON FLUENCE DOSIMETRY MEASUREMENTS The NRC has sponsored measurements to validate the transport calculations of fast-neutron-fluence exposure of certain critical support structures for the pressure vessel of the TROJAN nuclear power plant. The National Institute of Standards and Technology (NIST) was given the responsibility of seeing that benchmarked neutron fluence measurements were made using technology that was consistent with that in actual use by the nuclear power industry. This was carried out (1) by contracting with divisions of Westinghouse to supply dosimeters and make the measurements in the ex-vessel cavity of the TROJAN Reactor, and (2) by separately evaluating known-fluence irradiations of representative sets of their dosimeters in the standard neutron fission spectrum at NIST. Table A.] summarizes the NIST analysis of the Westinghouse measurements. The radiometric measurements employed selected isotopic reactions in which radioactivity is induced in disks and wires by primarily fast neutrons (E > 1 MeV), and the activity is analyzed by gamma counting. The measurements also employed solid-state track recorders (SSTR), where the exposure fluence is measured by optical counting tracks formed in mica deposits in contact with fissionable depositr of Np " and highly 2 depleted U2n, Because the SSTR technique is considerably newer, mass assay and optical efficiency measurements have been undergoing benchmark measurements in a cooperative program involving Westinghouse and NIST for a number of years and results pertinent to this series of measurements." The significant conclusion is that benchmark irradiation results are within a few percent of certified values depending on the treatment of a 3% to 5% bias, apparently associated with the masses of the ultralight deposits. The results from the benchmark-irradiated activation monitors are summarized by noting that the specific activities reported by the Waltz Mill Laboratory of Westinghouse were within 5% of those certified for the NIST irradiations, which were certified to 2.5%. Examination of spectral differences among the locations. Four sets of saturated specific activity radiometric results for different isotopic reactions are given in Part 1-A of Table A.I. All values are from Ref.11 and are in units of 10" reactions, or fissions, per second per target atom. Part 1-B of the table provides an 56
examination of the lack of spectral differences among the four locations. In particular, therein all reaction rates have been divided by those of Ni"(n,p) Co.58 The last two columns show averages and standard deviations, in percent, which suggest that within 5% each ratio species is the same for all locations. The 11.2% value associated with the US(n,f) Ru data results from an obviously incorrect value in location C. Since the ,l reaction-rate ratios for five isotopes, each of which responds in a different energy range I above 1 MeV, are the same within 5% for all four locations, the fast-neutron spectrum has to be essentially the same for all the locations. Determination of flux-magnitude differences among the locations. By dividing each reaction rate ratio to Ni"(n,p) by its average in all four locations from Part 1-B, we obtain relative results that can be averaged for each location to obtain a measure of how well their relative magnitudes are known. Part 1-C of the table gives such results, where Location B is (1.37/0.70) = 1.97 1.8% greater than Location C. As indicated previously, the problem with the US(n,f) Ru data has been circumvented by eliminating it from these latter analyses. e 57
~ Table A.1 NIST analyses of radiometric and SSTR dosimetry data measured by Westinghouse during Trojan Cycle 13 Reaction A B C D Avg. Std(%) i 1-A* j Cu"(n,a) 0.0242 0.032 0.0162 0.0211 0.0234 Ti ' (n,p) 0344 0.460 0.235 0304 03358 8 Fe"(n,p) 1.560 2.070 1.060 1350 1.5100 J S8 Ni (n,p) 2310 3.100 1.550 2.010 2.2425 28 U s(n,f)Ru U 8.830 11.200 7.060 6.860 8.4875 2 (n,f)Zr 8.020 11.100 5.580 7.020 7.9300 } 2 U s(n,f)Cs 7.760 10.600 5.600 7.090 7.7625 28 U SSTR 8.680 10.800 5.720 6.770 7.9925 Np'7 SSTR 124.00 162.00 81.20 110.00 11930 6 1-B Cu"(n,a) 0.01048 0.01032 0.01045 0.01050 0.0104 0.65 [ Ti ' (n,p) 0.14892 0.14839 0.15161 0.15124 0.1500 0.94 8 Fe"(n,p) 0.67532 0.66774 0.68387 0.67164 0.6746 0.88 S8 Ni (n,p) 1.0 1.0 1.0 1.0 2 U s(n,f)Ru 3.82251 3.61290 4.55484 3.41294 3.8508 11.21 28 U (n,f)Zr 3.47186 3.58065 3fm 00 3.49254 3.5363 1.55 28 U (n,f)Cs 335931 3.41935 3.61290 3.52736 3.4797 2.81 28 U SSTR 3.75758 3.48387 3.69032 3 36816 3.5750 437 Np'7 SSTR 53.6797 52.2581 523871 54.7264 53.263 1.90 1-C# Cu"(n,a) 1.03529 13 6898 0.69305 0.90267 Ti"(%p) 1.02457 137007 0.69993 0.90544 Fe"(n,p) 1.03311 13 7086 0.70199 0.89404 58 Ni (n,p) 1.03010 138239 0.69119 0.8 % 32 28 U (n,f)Ru 28 U (n,f)Zr 1.01135 13 9975 0.70366 0.88525 28 U (n,f)Cs 0.95"* 136554 0.72142 0.91337 28 U SSTR 1.08w2 135127 0.71567 0.84704 Np'7 SSTR 1.03940 135792 0.68064 0.92205 Avg. 1.0324 13708 0.7009 0.8958 Std (%) 2.44 1.02 1.47 2.25 ' Radiometric reaction and fission rates per nucleus => (where all values have an exponent of 10-17) < =. ' Ratios of above rates relative to Ni (n,p) reactions. S8 ' Ratios of rates relative to their averages in part-B. 58
APPENDIX B IMPACT OF ONE-FOURTH vs ONE-EIOHTH CORE MODEL USED IN TRANSPORT CALCULATIONS The neutron pad component in Trojan extends from 25' to 57.5* (in the coordinate system of the calculation models); therefore, unlike most reactor systems that contain a full thermal shield that completely surrounds the core, the ex. core geometry of Trojan is not symmetrical over 45' segments. Rather, the neutron pad layout exhibits 90* symmetry. In a one-eighth core Re model the neutron pad is represented as extending from 25 to 45 *,with a reflected boundary at 45 *. 'Ihis representation effectively reflects the pad into the interval 45' to 65
- in the reficcted octant. Thus, the azimuthal internal 57.5
- to 65
- is incorrectly occupied by the stainless steel pad, rather than water down, as in the true reactor geometry.
In order to investigate the impact of the one<ighth core symmetry approximation on cavity dosimeter calculations, a one-fourth core model with the correct asymmetrical neutron pad layout was used in a DOT Re transport calculation, and results were compared with those obtained from the reflected octant case. Table B.1 shows the ratios of various cavity reaction rates obtained in the one-eighth core model to those obtained in the one-fourth core model. It can be seen that at 0* there is very little difference in values computed with the two models. At 45' where the greatest impact is expected, the one-fourth core model gives about 12% higher reaction rates for the higher threshold dosimeters with less difference observed for lower threshold reactions. Table B.1 Ratio of reflected one-eighth core results to nonsymmetrical one-fourth core model 257 cm Radius 361.23 cm Radius Dosimeter 0* 45' 0* 45' Fe" 1.002 1.04 1004 1.115 Ni" 1.003 1.04 1.t,045 1.1099 Cu'5 1.0008 1.029 1.0017 1.11 Np " 1.005 1.024 1.0048 1.045 2 U 1.0039 1.038 1.005 1.092 Ti" 1.018 1.036 1.0033 1.118 Co" 1.009 1.008 1.0059 1.0064 59
NUREG/CR-6206 ORNI/rM-12693 Dist. Category RF INTERNAL DISTRIBUTION 1. C A. Baldwin 30. R. E. Stoller 2. B. R. Bass 31-34. J. A. Wang 3. W. R. Corwin 35. R. M. Westfall 4. S. K. Iskander 36. O. E. Whitesides 5-21. F.B.K.Kam 37. Central Research Library 22. R. K. Nanstad 38. ORNL Y-12 Research Library 23. C. E. Oliver Document Reference Section 24. W. E. Pennell 39-40. Laboratory Records Department 25. C E.Pugh 41. laboratory Records,ORNL(RC) 26. 1.Remec 42. ORNL Patent Office 27-29. C H. Shappert EXTERNAL DIS'IRIBUTION 43. M. Asgari, Louisiana State University, Nuclear Science Center, Baton Rouge, LA 70803 44. A. Hiser, U.S. Nuclear Pegulatory Commission, Division of Engineering, Mail Stop NL S 217 C, Washingica, DC 20555 45. E. P. Lippincott, Westinghouse Electric Corporation, M/S E4-33, Westinghouse Energy Center, P. O. Box 355, Pittsburgh, PA 15230-0355 46. R. E. Johnson, USNRC, Division of Engineering, Mail Stop NL S 217 C, Washington, DC 20555 47. E. D. McGarry, National Institute of Standards and Technology, Bldg. 235-A155, Gaithersburg MD 20899 48. E. Sajo, Louisiana State University, Nuclear Science Center, Baton Rouge, LA 70803 49. C Z. Serpan, Jr., U.S. Nuclear Regulatory Commission, Divirion of Engineering, Mail Stop NL S 217C, Washington, DC 20555 50. Al Taboada, U.S. Nuclear Regulatory Commission, Division of Engineering, Mail Stop NL S 217 C, Washington, DC 20555 51. M. L Williams, Imuisiana State University, Nuclear Science Center, Baton Rouge, LA 70803 52. Office of the Assistant Manger for Energy, Research and Development Department of Energy Oak Ridge Operations, Oak Ridge (DOE-ORO), P.O. Box 2001, Oak Ridge, TN 37831-6269 53 54. Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831 55-182. Given distribution as shown in category RF (NTIS-10) 61
NC FORM 335 U.S. NUCLEAR REGULATORY Commission
- 1. REPORT NUMBER L'a*!A no2 lMT=';,,^%*. Th '"i" ""-
220'.22o2 BIBLIOGRAPHIC DATA SHEET NUREG/CR-6206 rsee inern,ct,ons en tne,wruj ORNL/TM-12693
- 2. TITLE AND SUBTITLE Transport Calculations of Radiation Exposure to Vessel Support Structures in the Trojan Reactor.
3. DATE REPORT PUBLISHED MONTH l vtan July 1994
- 4. FIN oR GRANT NUMBER W6164 J
- 5. AUTHOR (S)
- 6. TYPE OF REPORT
.1 1 2 M. Asgari, M. L. Williams, F. B. K. Kam, E. D. McGarry Technical
- 7. PE R toD CoV E R E o finclauwe ostest i
8 PEM oXMING oRGANIZATloN - N AME AND ADDREbs it? Nnc. paarse owenson. orrece or neron. us Nucear nepuurory commonuon.amt mesas esoren;sr eentracuor, prone nema ennt me*Iang encores) Oak Ridge National Laboratory Louisiana State University National Institute of Oak Ridge, TN 37831-6050 Nuclear Science Center Standards and Technolo;y e Baton Rouge, LA 70803 Bldg. 235-A155 Gaithersburg, MD 20899
- 9. SPONSORING ORGANIZATION - NAME AND ADDR E5S ist mac. rype %m, as eac,e~;,rcone,sraer, pre de sac o.wm on. Otrace er meemn, ui wurner s.e,ase,y commmuon ome n n,.,m.o Division of Engineering Office of Nuclear Regulatory Research U.S. Regulatory Commission Washington, DC 20555-0001
- 10. SUPPLEMENT ARY NOTES
- 11. ABSTRACT rico.orm er auf Comparison of transport calculations of the dosimeter activities with the expenmental measurements shows that the values obtamed with ENDF/B-VI cross-section data overestunate the measured results for high-energy-threshold reactmas in the cavity by up to 41%, and thennal reactions by up to a factor of 3.0. The transport calculations performed with the original SAILOR cross-section library (based on ENDF/B-IV data) overestunate aaul threshold reactions by only 15% and the thermal reactions by about a factor of 2.50. These results are inconsistent with those oblamed in earlier studies that w r.oi transport calculations done with SAILOR vs ENDF/B-VI, which indicate that SAILOR tends to underestunate cavity donimeter activities for threshold reactions, while the ENDF/B-VI values usually agree better with expenmental results. One factor that probably contributes to the rather large discrepancy between the computed and ag.muoi activities is the core power distribution used in the transport calculations. Because of unavailability of plant-specific data, a generic power distributh provided by Wdn=hane was used. Since the calculated cavity flux levels appear to be over-estunated, the results estimated for the exposure to the support structure should be conservative.
- 12. x E Y wor DstDE scR.PT oR s <t,,..,,.
- raer, r.,,ne,*,,.,
i3..vait. civvstavaMi=1 ENDF/Il-VI, neutron transport calculations, neutron fluxes, dosimetry activities (Tann fa,rr Unclassified Ir. e e,1 Unclassified
- 15. NUMBER of PAGES
- 16. PRICE NRC 7 ORM 335 (2499
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