lattice physics analysis of measured rod-by-rod fp inventory distributions of burnt bwr 9 × 9 uo ...
TRANSCRIPT
This article was downloaded by: [The UC Irvine Libraries]On: 08 November 2014, At: 14:43Publisher: Taylor & FrancisInforma Ltd Registered in England and Wales Registered Number: 1072954 Registered office: MortimerHouse, 37-41 Mortimer Street, London W1T 3JH, UK
Journal of Nuclear Science and TechnologyPublication details, including instructions for authors and subscription information:http://www.tandfonline.com/loi/tnst20
Lattice Physics Analysis of Measured Rod-by-RodFP Inventory Distributions of Burnt BWR 9 × 9 UO2
AssembliesToru YAMAMOTO a , Masaru SASAGAWA b & Yuichiro KANAYAMA ca Japan Nuclear Energy Safety Organization , TOKYU REIT Toranomon Bldg., 3-17-1Toranomon, Minato-ku, Tokyo , 105-0001 , Japanb Global Nuclear Fuel-Japan Co., Ltd. , 2-3-1 Uchikawa, Yokosuka-shi, Kanagawa ,239-0836 , Japanc Nuclear Fuel Industries, Ltd. , 3135-41 Muramatsu, Tokai-mura, Naka-gun, Ibaraki ,319-1196 , JapanPublished online: 19 Mar 2012.
To cite this article: Toru YAMAMOTO , Masaru SASAGAWA & Yuichiro KANAYAMA (2009) Lattice Physics Analysis of MeasuredRod-by-Rod FP Inventory Distributions of Burnt BWR 9 × 9 UO2 Assemblies, Journal of Nuclear Science and Technology,46:7, 653-664, DOI: 10.1080/18811248.2007.9711572
To link to this article: http://dx.doi.org/10.1080/18811248.2007.9711572
PLEASE SCROLL DOWN FOR ARTICLE
Taylor & Francis makes every effort to ensure the accuracy of all the information (the “Content”) containedin the publications on our platform. However, Taylor & Francis, our agents, and our licensors make norepresentations or warranties whatsoever as to the accuracy, completeness, or suitability for any purpose ofthe Content. Any opinions and views expressed in this publication are the opinions and views of the authors,and are not the views of or endorsed by Taylor & Francis. The accuracy of the Content should not be reliedupon and should be independently verified with primary sources of information. Taylor and Francis shallnot be liable for any losses, actions, claims, proceedings, demands, costs, expenses, damages, and otherliabilities whatsoever or howsoever caused arising directly or indirectly in connection with, in relation to orarising out of the use of the Content.
This article may be used for research, teaching, and private study purposes. Any substantial or systematicreproduction, redistribution, reselling, loan, sub-licensing, systematic supply, or distribution in anyform to anyone is expressly forbidden. Terms & Conditions of access and use can be found at http://www.tandfonline.com/page/terms-and-conditions
Lattice Physics Analysis of Measured Rod-by-Rod FP Inventory
Distributions of Burnt BWR 9� 9 UO2 Assemblies
Toru YAMAMOTO1;�, Masaru SASAGAWA2 and Yuichiro KANAYAMA3
1Japan Nuclear Energy Safety Organization, TOKYU REIT Toranomon Bldg.,3-17-1 Toranomon, Minato-ku, Tokyo 105-0001, Japan
2Global Nuclear Fuel-Japan Co., Ltd., 2-3-1 Uchikawa, Yokosuka-shi, Kanagawa 239-0836, Japan3Nuclear Fuel Industries, Ltd., 3135-41 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1196, Japan
(Received November 11, 2008 and accepted in revised form March 11, 2009)
Fuel rod gamma-ray spectrometry was performed for five-cycle irradiated BWR 9� 9-9 and 9� 9-7fuel assemblies, and relative rod-by-rod FP inventory distributions of 137Cs, 134Cs, 106Ru, 95Zr and 154Euwere obtained for lower and upper axial nodes of the assemblies. The inventories of these nuclidesrepresent burnups, burnups multiplied by a thermal neutron flux, a buildup of Pu, and a fission rate at theend of a fuel discharge cycle. The measured distributions were compared with those calculated by ageneral purpose neutronics code system SRAC and a continuous energy Monte Carlo burnup calculationcode MVP-BURN with infinite assembly models. The calculations adopted nuclear data libraries JENDL-3.2, -3.3, and ENDF/B-VI.8. The calculated results generally well reproduce the measured distributions,and the root mean square errors (RMSs) of the calculated results against the measurements are 1 to 2% forthe inventory distributions of 137Cs, 2 to 4% for 134Cs, 2 to 5% for 106Ru, 3 to 4% for 95Zr, and 4 to 5% for154Eu, which include measurement and analysis errors. Small differences in the calculated distributionsbetween SRAC and MVP-BURN were observed for 134Cs, 106Ru, and 154Eu for some of the corner rods,Gd2O3-UO2 rods, and other rods in the 9� 9-7 fuel assembly.
KEYWORDS: rod-by-rod FP inventory distribution, 137Cs, 134Cs, 106Ru, 95Zr, 154Eu, gamma-rayspectrometry, BWR, 9� 9 fuel assembly, SRAC code system, MVP-BURN code
I. Introduction
Rod-by-rod power distributions in fuel assemblies areessential parameters in core design and analysis especiallyrelated to thermal margins such as maximum linear heatgeneration rates and minimum critical power ratios forBWR cores. Verification work of the distributions has com-monly been performed using measured data of power dis-tributions of zero-power core physics experiments withfresh mock up fuel assemblies.1–3) However, studies on burntfuel assemblies are sparse, since such measurements usuallyhave to be done in postirradiation examination facilities.Misu et al.4) have reported experimental data of rod-by-rod distributions of 140Ba on irradiated BWR 9� 9-1 UO2
and MOX assemblies, which represent fission rate distribu-tions. They obtained the data by gamma scanning in a burntfuel storage pool of a 1,300MWe BWR plant. Yamamotoand Yamamoto5) have reported a study on the analyses of therod-by-rod inventory distributions of 137Cs, 134Cs, 106Ru, and95Zr of 8� 8-4 UO2 assemblies irradiated in a 1,100MWeBWR plant.
In order to extend the validation data on core design andanalyses of burnt fuel assemblies, rod-by-rod FP inventorydistributions were measured by gamma-ray spectrometry fortwo burnt BWR 9� 9 UO2 fuel assemblies, which are 9� 9-9 and 9� 9-7 fuel assemblies irradiated in a 1,100MWeBWR plant. The former fuel assembly has a water channelreplacing nine fuel rods in the center of the assembly, andthe latter two water rods replacing seven fuel rods. Burnupcalculations with infinite assembly models were performedby taking into account the irradiation histories of the assem-blies, and the calculated results were compared with themeasured data. The first part of the present paper givesdescriptions about the data of the rod-by-rod FP inventorydistributions, the second part shows the comparison of thedata with the calculated results, and the third part presentsdiscussion and conclusions.
II. FP Rod-by-Rod FP Inventory Distribution of9� 9 Fuel Assemblies
Major specifications of the 9� 9-9 and 9� 9-7 fuelassemblies6,7) are shown in Table 1. They were inserted toUnit 1 of Fukushima Power Station 2 (2F1: 1,100MWeBWR plant) as lead-use assemblies of high-burnup 9� 9
�Atomic Energy Society of Japan
�Corresponding author, E-mail: [email protected]
Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 46, No. 7, p. 653–664 (2009)
653
TECHNICAL MATERIAL
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
UO2 fuel in July 1996 and discharged in January 2003for postirradiation examinations.6) Figure 1 shows loadinglocations of the assemblies.6) Both assemblies had not beenloaded in the peripheral region of the core during irradiation.The assembly average burnups are 53.5 and 53.0 GWd/t forthe 9� 9-9 and 9� 9-7 fuels, respectively. The postirradia-tion examinations include axial gamma-ray spectrometry ofspecific gamma rays from FP nuclides with Ge detectors.The numbers of fuel rods under the examinations weretwelve and twenty-five fuel rods for the 9� 9-9 and 9� 9-7 fuels, respectively. Figure 2 shows enrichment distribu-tions of the assemblies6,7) and the fuel rod locations in the
assemblies for the gamma-ray measurements.6) Since the9� 9-9 and 9� 9-7 fuels have one-eighth and one-fourthsymmetries in nuclear design on enrichment and Gd2O3-UO2 rod distributions, the twelve and twenty-five rods wereselected for the gamma-ray spectrometry, respectively. Thegamma-ray measurements were performed in September andOctober in 2004 for the 9� 9-9 fuel at the Japan AtomicEnergy Agency and April 2005 in the Nippon Nuclear FuelDevelopment Co., Ltd. for the 9� 9-7 fuel.6) Table 2 showsthe FP nuclides and the relevant specific gamma rays for themeasurements. For each fuel rod, gamma rays were scannedin an axial direction by a Gd detector installed in a shielding
Table 1 Major specifications of the 9� 9-9 and 9� 9-7 fuelassemblies6,7)
Fuel assemblyLattice 9� 9-9 9� 9-7Number of fuel rods 72 74
(Part length rod: 8)Assembly av. 235Uenrichment (wt%)
3.4� 3.4�
Fuel RodOuter diameter (mm) 11.0� 11.2�
Cladding thickness (mm) 0.70� 0.71�
Cladding material Zry-2 Zry-2Pellet diameter (mm) 9.4� 9.6�
Pellet-cladding gap (mm) 0.20� 0.20�
Pellet density (%TD) 97� 97�
Pellet material UO2, UO2-Gd2O3 UO2, UO2-Gd2O3
Water channel/rod Square tube Round tubeDimension (mm) 38.5� 24.9�
Number 1 2
�Rounded number
Number: irradiation cycle
9 × 9-9 fuelassembly
9 × 9-7 fuelassembly
Fig. 1 Loading locations of the 9� 9-9 and 9� 9-7 fuel assem-blies
Rod number: order of enrichment for each assembly in descending order,G: Gd2O3-UO2 fuel rod, P: part length rod, W: water channel or rod
9 × 9-9 fuel assembly 9 × 9-7 fuel assembly
a b c d e f g h j A B C D E F G H J
5 4 3 2 22 3 4 5 1 9 8 5 3 5 3 5 8 9
4 3 4G 2 4G 2 4G 3 4 2 8 4P 7G 6 2P 7G 6 4P 8
3 4G 1 2 22 1 4G 3 3 5 7G 8 1 6 3 7G 6 5
2 2 2 2 2 2 4 3 6 1 3 7G 3
2 4G 2 2 4G 2 5 5 2P 6 6 2P 5
2 2 2
W
2 2 2 6 3 7G 3
W
W 1 6 3
3 4G 1 2 2 2 1 4 3 7 5 6 7G 3 6 1 8 7G 5
4 3 4G 2 4G 2 4G 3 4 8 8 4P 6 7G 2P 6 7G 4P 8
5 4 3 2 2 2 3 4 5 9 9 8 5 3 5 3 5 8 9
8
8
Fig. 2 Enrichment distributions of the 9� 9-9 and 9� 9-7 fuel assemblies and rod locations (shadowed rods) of fuelrods for gamma-ray spectrometry of FP nuclides
654 T. YAMAMOTO et al.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
collimator. The fuel rod was arranged such that the detectoralways saw the fuel rod in the direction from ‘‘j’’ to ‘‘a’’ inFig. 1 for the 9� 9-9 fuel, and the detector saw the rodsurface that gave the average intensity of gross gamma raysmeasured in the azimuthal direction of the fuel rod for the9� 9-7 fuel.
Necessary corrections were performed for the decay ofradioactivities during measurement periods to normalize thegamma-ray intensity at a fixed date. In order to convert thegamma-ray intensity distributions to relative inventory dis-tributions for comparison with those of burnup calculations,corrections for differences in gamma-ray transmission fac-tors between the UO2 and Gd2O3-UO2 rods are necessaryand were performed similarly to that described in Ref. 5).Relative inventory distributions were obtained for two hor-izontal planes at the lower and upper axial heights of the fuelassembly, which exhibit a difference in in-channel voidfraction. The axial heights were selected to be the middleof the 5th and 6th nodes (5/6 node) and the 18th node for the9� 9-9 fuel assembly, and the 5/6 node and the middle ofthe 18th and 19th nodes (18/19 node) for the 9� 9-7 fuelassembly. Here, the axial length of the node is one-twenty-fourth of an effective fuel length of about 371 cm, and thefirst and 24th nodes are the bottom and top of the effective
fuel length, respectively. The measured relative FP inventorydistributions are shown in Figs. 3 and 4 for the 9� 9-9 and9� 9-7 fuel assemblies, respectively, where the measuredvalues are normalized so that their average value is 1.0 foreach node. Figure 5 schematically shows the relative FPinventory distributions for the 18th node of the 9� 9-9 fuelassembly as an example. The measurement errors based onthe gamma-ray counting statistics are 0.1 to 0.5% for 137Cs,0.2 to 0.5% for 134Cs, 0.6 to 1.2% for 106Ru, and 3 to 6% for95Zr in the measurements of the 9� 9-9 fuel,8) and 0.4 to0.5% for 137Cs, 0.4 to 0.5% for 134Cs, 1.2 to 1.5% for 106Ru,and 2.6 to 3.1% for 154Eu in the measurements of the 9� 9-7fuel.9)
III. Burnup Calculations of Rod-by-Rod FP Inven-tory Distributions
Burnup calculations were performed for the lower andupper nodes for each assembly, which correspond to themeasurements. The calculations adopted an infinite assemblymodel, taking into account the irradiation histories of therelevant axial nodes such as the node average power and thein-channel void fractions, which were provided by the utilityas records of a core monitoring system of the relevant plant.For the 5/6 and 18/19 nodes, irradiation histories wereobtained by averaging those of the 5th and 6th nodes, andthe 18th and 19th nodes, respectively.
1. Deterministic AnalysisBurnup calculations were made using a general purpose
neutronics code system SRAC10) with a 107 neutron energygroup and a collision probability module for neutron spec-trum calculations. Resonance cross sections were obtainedby a hyperfine energy group calculation module PEACO.10)
As the nuclear data libraries, JENDL-3.211) and JENDL-3.312) were used for the 9� 9-9 fuel, and JENDL-3.3 andENDF/B-VI.813) were used for the 9� 9-7 fuel. The burnupsteps in these calculations were selected not to be larger than0.25GWd/t until a burnable poison (Gd2O3) was burned out,
5/6 node 137Cs 18 node7 1.10 1.06 1.088 0.84 1.02 0.82 0.979 1.10 1.10 1.02 0.97 0.93
e f g h j
7 1.10 1.04 1.078 0.81 0.99 0.81 0.979 1.10 1.10 1.04 0.99 0.99
e f g h j
134Cs 7 1.13 1.05 1.048 0.81 1.00 0.78 0.949 1.13 1.14 1.04 0.98 0.96
e f g h j
7 1.13 1.03 1.018 0.76 0.95 0.75 0.949 1.13 1.14 1.07 1.02 1.06
e f g h j
106Ru7 1.05 0.98 0.938 0.92 0.96 0.88 0.969 1.05 1.04 1.03 1.03 1.16
e f g h j
7 1.05 0.97 0.918 0.88 0.92 0.86 0.979 1.06 1.05 1.04 1.06 1.23
e f g h j
95Zr7 1.09 1.01 1.028 0.94 1.04 0.94 0.939 1.07 1.07 0.97 0.95 0.98
e f g h j
7 1.06 1.01 1.118 0.94 1.03 0.90 0.959 1.01 1.11 0.94 0.97 0.97
e f g h j
Fig. 3 Measured relative FP inventory distributions for the 9� 9-9 fuel assembly
Table 2 FP nuclides and relevant specific gamma rays
FP nuclide T1=2Representative nuclear
characteristicsGamma-ray energy
(keV)
137Cs 30.07 y Burnups 661.7
134Cs 2.062 y Burnups � flux 604.7
106Ru 373.59 d Pu buildup 621.9 (106Rh)�
154Eu 8.593 y Burnups � flux 1274
95Zr 64.02 d Power of EOC 765.8 (95Nb)�
�Gamma rays in decay of 106Rh and 95Nb that are short-lived daughter
nuclides of 106Ru and 95Zr, respectively
Lattice Physics Analysis of Measured Rod-by-Rod FP Inventory Distributions of Burnt BWR 9� 9 UO2 Assemblies 655
VOL. 46, NO. 7, JULY 2009
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
which is the node average burnup of 15GWd/t. After15GWd/t, a burnup step of 1.0GWd/t was applied. Thecalculated results are shown in Table A1 in Appendix.
2. Monte Carlo AnalysisBurnup calculations were also performed using a contin-
uous energy Monte Carlo burnup calculation code MVP-BURN14) for the same nodes of the SRAC calculations. Thenuclear data libraries used are the same as those used in the
SRAC calculations. Applying the predictor-corrector option,a burnup step was set to be 1.0GWd/t until the node averageburnup of 15GWd/t and 5GWd/t after this burnup. Foreach burnup step, fifty batches with 10,000 neutron historiesper batch were calculated and the initial ten batches wereskipped for a statistical process. This makes a statistical error(1�) of infinite multiplication factors of around 0.1%dk anda pellet average burnup of about 0.1%.15) The calculatedresults are shown in Table A2 in Appendix.
IV. Comparison between Calculated and MeasuredResults
1. 9� 9-9 Fuel AssemblyFigure 6 shows a comparison between the calculated re-
sults by SRAC with JENDL-3.2 and the measurements forthe relative distributions for the 9� 9-9 fuel. The pointsaround the measured relative peaking around 0.8 for 137Csand 134Cs correspond to the Gd2O3-UO2 rods with relativelylower enrichments and lower burnups. The figure shows thatthe measured relative distributions are well reproduced bythe burnup calculations; however, the deviations from thelines with the calculations equal to the measurements in-crease with measurement error. Figure 7 also shows thesame comparison for the calculated results of MVP-BURNwith JENDL-3.2 and Table 3 shows root mean square errors(RMSs) in the differences between the calculated and meas-ured results.
1
2
3
S1S2
S3S4
S5
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1
2
3
S1S2
S3S4
S5
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1
2
3
S1S2
S3S4
S5
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1
2
3
S1S2
S3S4
S5
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
137Cs 134Cs
106Ru 95Zr
Gd2O3-UO2
fuel rod
Fig. 5 Relative FP inventory distributions for a radial plane of the18th node from the bottom of the 9� 9-9 fuel assembly
5/6 node 137Cs 18/19 node A B C D E F G H J A B C D E F G H J
1 1.05 1.03 1.08 1.16 1.05 1.16 1.09 1.01 1.03 1.06 1.02 1.09 1.14 1.04 1.13 1.09 1.01 1.032 0.98 0.85 0.88 1.02 0.86 0.89 1.03 0.83 0.95 0.82 0.963 0.86 0.86 1.09 0.91 1.06 0.85 0.84 0.85 1.06 1.01 1.04 0.824 0.95 0.9556 1.12 1.13789 1.11 1.14
134CsA B C D E F G H J A B C D E F G H J
1 1.11 1.08 1.14 1.21 1.10 1.19 1.13 1.08 1.12 1.14 1.08 1.12 1.18 1.08 1.17 1.12 1.07 1.152 0.99 0.82 0.85 0.99 0.82 0.86 1.04 0.78 0.91 0.76 0.933 0.81 0.81 1.06 0.87 1.04 0.81 0.79 0.79 1.03 0.99 1.01 0.784 0.97 0.9556 1.05 1.08789 1.04 1.09
106Ru A B C D E F G H J A B C D E F G H J
1 1.26 1.09 1.03 1.06 1.03 1.08 1.15 1.16 1.28 1.31 1.09 1.03 1.01 0.99 1.08 1.08 1.08 1.182 0.87 0.86 0.92 0.95 0.93 1.02 0.99 0.84 0.89 0.85 0.953 0.82 0.86 0.91 0.95 0.94 0.86 0.79 0.87 0.99 1.06 0.95 0.864 0.94 0.9856 1.01 1.03789 1.05 1.09
154Eu A B C D E F G H J A B C D E F G H J
1 1.00 1.03 1.03 1.18 1.05 1.14 1.13 0.97 0.90 1.00 1.02 1.11 1.10 1.01 1.20 1.03 1.03 0.932 0.94 0.90 0.98 1.03 0.89 0.94 0.98 0.90 0.96 0.93 0.983 0.91 0.98 1.11 0.87 1.06 0.90 0.89 0.89 1.06 1.03 1.02 0.804 0.95 0.9756 1.06 1.09789 1.06 1.03
Fig. 4 Measured relative FP inventory distributions for the 9� 9-7 fuel assembly
656 T. YAMAMOTO et al.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
Table A1 in Appendix gives a comparison of the calcu-lated results between SRAC and MVP-BURN. Figure 8schematically shows examples of the comparison for 137Csand 106Ru. The comparison indicates that the differences inthe calculated results between SRAC and MVP-BURN are
less than 1.5% for most FP nuclides. The deviations of 1.5 to2.0% are exceptionally observed in the 5/6 node of a cornerrod (j, 9) for the four FP nuclides; however, the calculatedresults of MVP-BURN are not always closer to the measure-ments than those of SRAC.
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
137Cs
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
134Cs
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
106Ru
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
95Zr
: 5/6 node, : 18 node
Fig. 6 Comparison between calculated results of SRAC (JENDL-3.2) and measured ones for relative FP inventorydistributions of the 9� 9-9 fuel assembly
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
137Cs
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
134Cs
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
106Ru
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
95Zr
: 5/6 node, : 18 node
Fig. 7 Comparison between calculated results of MVP-BURN (JENDL-3.2) and measured ones for relative FPinventory distributions of the 9� 9-9 fuel assembly
Lattice Physics Analysis of Measured Rod-by-Rod FP Inventory Distributions of Burnt BWR 9� 9 UO2 Assemblies 657
VOL. 46, NO. 7, JULY 2009
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
On the other hand, the differences in the calculated re-sults between JENDL-3.2 and JENDL-3.3 are less than 1.0%and negligibly small even though the comparison is notshown.
2. 9� 9-7 Fuel AssemblyFigure 9 shows a comparison between the calculated re-
sults by SRAC with JENDL-3.3 and the measurements forthe relative distribution for the 9� 9-7 fuel. Figure 10 alsoshows the same comparison for the calculated results ofMVP-BURN with JENDL-3.3. Table 4 shows root meansquare errors (RMSs) in the differences between the calcu-lated and measured results. The figures show that bothmeasured relative distributions are well reproduced by theburnup calculations; however, the RMSs are larger than themeasurement errors.
Table 3 RMSs in comparison between the calculated results(JENDL-3.2) and the measurements for the 9� 9-9 fuel assem-bly (unit: %)
Calculationcode
Nuclide 137Cs 134Cs 106Ru 95Zr
Measurement 0.1–0.5 0.2–0.5 0.6–1.2 3–6error (%)
SRAC 5/6 node 1.4 2.6 2.2 3.5
18 node 1.1 2.1 2.1 3.2
MVP-BURN 5/6 node 1.6 3.3 2.4 3.6
18 node 1.2 2.3 2.0 3.2
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Calculated peaking (MVP-BURN)
Cal
cula
ted
peak
ing
(SR
AC
)
137Cs
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Calculated peaking (MVP-BURN)
Cal
cula
ted
peak
ing
(SR
AC
)
106Ru
: 5/6 node, : 18 node
Fig. 8 Comparison of calculated results between SRAC and MVP-BURN with JENDL-3.2 for 137Cs and 106Ru of the9� 9-9 fuel assembly
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
137Cs
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
134Cs
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
106Ru
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
154Eu
: 5/6 node, : 18/19 node
Fig. 9 Comparison between calculated results of SRAC (JENDL-3.3) and measured ones for relative FP inventorydistributions of the 9� 9-7 fuel assembly
658 T. YAMAMOTO et al.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
Table A2 in Appendix gives the comparisons of thecalculated results between SRAC and MVP-BURN.Figure 11 schematically shows the examples of the com-parison for 137Cs and 106Ru. The comparison indicatesthat the differences in the calculated results between SRACand MVP-BURN are less than 1.5% for most cases. Theexceptions larger than 1.5% are observed for 134Cs, 106Ru,and 154Eu in some of the corner rods and Gd2O3-UO2
rods, and the rod (D, 4); the calculated results of MVP-BURN are generally closer to the measurements thanthose of SRAC. The point of the calculated peaking by
MVP-BURN equal to about 1.0 in the 18/19 node for106Ru corresponds to the rod (D, 4), and the calculationof SRAC gives underestimation. The neutron energy spec-tra of these fuel rods are largely different from those ofthe other rods since they adjoin the moderator of an out-channel gap or water rods, or contain a burnable poisonGd2O3.
On the other hand, the differences in the calculated resultsbetween JENDL-3.3 and ENDF/B-IV.8 are less than 1.0%and negligibly small even though the comparison is notshown.
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
137Cs
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
134Cs
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
106Ru
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Measured peaking
Cal
cula
ted
peak
ing
154Eu
: 5/6 node, : 18/19 node
Fig. 10 Comparison between calculated results of MVP-BURN (JENDL-3.3) and measured ones for relative FPinventory distributions of the 9� 9-7 fuel assembly
Table 4 RMSs in comparison between the calculated results and the measurements for the 9� 9-7 fuel assembly(unit: %)
Calculationcode
Nucleardata library
Nuclide 137Cs 134Cs 106Ru 154Eu
Measurement 0.4–0.5 0.4–0.5 1.2–1.5 2.6–3.1error (%)
SRAC JENDL-3.3 5/6 node 1.8 4.0 5.6 4.7
18/19 node 1.4 2.1 4.3 4.5
ENDF/B-VI8 5/6 node 1.8 4.0 5.5 4.7
18/19 node 1.5 2.1 4.3 4.5
MVP-BURN JENDL-3.3 5/6 node 1.5 3.7 5.0 4.4
18/19 node 1.7 2.5 4.5 4.4
ENDF/B-VI8 5/6 node 1.5 3.7 5.2 4.6
18/19 node 1.7 2.3 4.6 4.4
Lattice Physics Analysis of Measured Rod-by-Rod FP Inventory Distributions of Burnt BWR 9� 9 UO2 Assemblies 659
VOL. 46, NO. 7, JULY 2009
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
V. Discussion
The experimental errors mentioned in the previous sec-tions reflect the counting statistics of gamma rays. One of thepossible systematic errors would come from azimuthal dis-tributions of gamma-ray intensity. Figure 12 shows the
measurement data for normalized gross gamma-ray intensitydistributions in arbitrary azimuthal angles in the upperpart of typical fuel rods of the 9� 9-7 fuel. The largestdeviation from the average value (¼ 1:0) is about 3% forthe (D, 4) rod and 2% for the others. Even though a dif-ference exists between the countermeasures against these
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Calculated peaking (MVP-BURN)
Cal
cula
ted
peak
ing
(SR
AC
)
137Cs
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5
Calculated peaking (MVP-BURN)
Cal
cula
ted
peak
ing
(SR
AC
)
106Ru
: 5/6 node, : 18/19 node
Fig. 11 Comparison of calculated results between SRAC and MVP-BURN with JENDL-3.3 for 137Cs and 106Ru of the9� 9-7 fuel assembly
0.95
0.96
0.97
0.98
0.99
1.00
1.01
1.02
1.03
1.04
1.05
0 50 100 150 200 250 300 350
Arbitrary azimuthal angle (degree)
Nor
mal
ized
rel
ativ
e in
tens
ity
A1 B1 C1D1 D4
0.95
0.96
0.97
0.98
0.99
1.00
1.01
1.02
1.03
1.04
1.05
0 50 100 150 200 250 300 350
Arbitrary azimuthal angle (degree)
Nor
mal
ized
rel
ativ
e in
tens
ity
C2 C3 D3
Fig. 12 Gross gamma-ray distributions in arbitrary azimuthal angle for the 9� 9-7 fuel assembly
660 T. YAMAMOTO et al.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
distributions for the measurements of the 9� 9-9 and9� 9-7 fuels as mentioned in the previous section, thesedistributions would cause systematic errors in the measuredFP distributions.
Another possible systematic errors would come from theeffect of the adjacent fuel assemblies loaded with the rele-vant fuel assemblies. The measurement data include someerrors attributed to this effect since these data were directlycompared with the burnup calculations with the infiniteassembly model, which neglects this effect. For the 9� 9-7 fuel, the (F,1) and (A,6) rods are symmetrically locatedfrom each other in the nuclear design as seen in Fig. 2, andthe calculations give the same FP inventories; however, themeasurements in Fig. 4 show a systematical trend indicatingthat the (F,1) rod gives larger inventories than the (A,6) rodfor all FP nuclides in the 5/6 node and part of nuclides inthe 18/19 node. This is probably due to the effect of theadjacent fuel assemblies.
VI. Conclusions
The calculated results of SRAC and MVP-BURN gener-ally well reproduce the measured rod-by-rod inventory dis-tributions of 137Cs, 134Cs, 106Ru, 95Zr, and 154Eu of the five-cycle irradiated 9� 9-9 and 9� 9-7 fuel assemblies. TheRMSs of the calculated results against the measured ones are1 to 2% for the inventory distributions of 137Cs, 2 to 4% forthose of 134Cs, 2 to 5% for those of 106Ru, 3 to 4% for thoseof 95Zr, and 4 to 5% for those of 154Eu, which include themeasurement and analysis errors. Small differences in thecalculated distributions between SRAC and MVP-BURNwere observed for 134Cs, 106Ru, and 154Eu in some of thecorner rods, Gd2O3-UO2 rods, and other rods for the 9� 9-7fuel.
Acknowledgement
The authors would like to express their thanks to themeasurement teams involved in the postirradiation examina-tions of the 9� 9-9 fuel in the Japan Atomic Energy Agencyand the 9� 9-7 fuel in the Nippon Nuclear Fuel Develop-ment Co., Ltd.
References
1) E. Saji, H. Shirayanagi, ‘‘Analysis of boiling water reactormixed-oxide critical experiments with CASMO-4/SIMULATE-3,’’ Nucl. Sci. Eng., 121, 52 (1995).
2) K. Ishii, Y. Ando, N. Takada et al., ‘‘Analysis of high mod-eration full MOX BWR core physics experiments BASALA,’’Trans. At. Energy Soc. Jpn., 4[1], 45 (2005), [in Japanese].
3) M. F. Murphy, M. Plaschy, F. Jatuff et al., ‘‘Fission andcapture rate measurements in a SVEA-96 Optima2 BWR as-sembly compared with MCNPX predictions,’’ Proc. Int. Conf.on PHYSOR 2006, Vancouver, Canada, Sep. 10–14, 2006(2006).
4) S. Misu, H. Spieling, H. Moon, A. Koschel, ‘‘Pin-by-pin gam-ma scan measurement on MOX and UO2 fuel assemblies andevaluation,’’ Proc. Int. Conf. on PHYSOR 2000, Pittsburgh,Pennsylvania, USA, May 7–12, 2000 (2000).
5) T. Yamamoto, M. Yamamoto, ‘‘Analysis of rod-by-rod FPinventory distributions in BWR 8� 8 UO2 assemblies usinglattice physics method,’’ J. Nucl. Sci. Technol., 45[1], 1(2008).
6) Fiscal 2006 Report of Demonstration Program on Reliabilityof High Burnup 9� 9 Fuel, Japan Nuclear Energy SafetyOrganization (2007), [in Japanese].
7) Application for Establishment Permit of Modification of Unit 1and 2 of Fukushima Power Station 2, Tokyo Electric PowerCompany (1994), [in Japanese].
8) Fiscal 2005 Report of Demonstration Program on Reliabilityof Nuclear Design Methods of Full MOX Cores, Japan NuclearEnergy Safety Organization (2006), [in Japanese].
9) Fiscal 2006 Report of Demonstration Program on Reliabilityof Nuclear Design Methods of Full MOX Cores, Japan NuclearEnergy Safety Organization (2007), [in Japanese].
10) K. Okumura, K. Kaneko, K. Tsuchihashi, SRAC95: GeneralPurpose Neutronics Code System, JAERI-Data/Code 96-015(JAERI) (1996), [in Japanese].
11) T. Nakagawa, K. Shibata, S. Chiba et al., ‘‘Japanese evaluatednuclear data library version 3 revision-2,’’ J. Nucl. Sci. Tech-nol., 32, 1259 (1995).
12) K. Shibata, T. Kawano, T. Nakagawa et al., ‘‘Japanese eval-uated nuclear data library version 3 revision-3: JENDL3.3,’’J. Nucl. Sci. Technol., 39[11], 1125 (2002).
13) P. F. Rose, C. L. Dunford, ENDF-102 Data Formats andProcedures of the Evaluated Nuclear Data File ENDF-6,BNL-NCS-44945, Rev. 2, Brookhaven National Laboratory(1997).
14) K. Okumura, T. Mori, M. Nakagawa et al., ‘‘Validation of acontinuous-energy Monte Carlo burn-up code MVP-BURNand its application to analysis of post irradiation experiment,’’J. Nucl. Sci. Technol., 37[2], 128 (2000).
15) T. Yamamoto, Y. Kanayama, ‘‘Lattice physics analysis ofburnups and isotope inventories of U, Pu, and Nd of irradiatedBWR 9� 9-9 UO2 fuel assemblies,’’ J. Nucl. Sci. Technol.,45[6], 547 (2008).
Appendix
Tables A1 and A2 show the numerical data of the calcu-lations with JENDL-3.2 for the 9� 9 fuel assembly andJENDL-3.3 for the 9� 9-7 fuel assembly, respectively.The tables also show the comparison of the calculated resultsbetween SRAC and MVP-BURN.
Lattice Physics Analysis of Measured Rod-by-Rod FP Inventory Distributions of Burnt BWR 9� 9 UO2 Assemblies 661
VOL. 46, NO. 7, JULY 2009
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
Table A1 Calculated FP rod-by-rod distributions of SRAC and MVP-BURN with JENDL-3.2 for the 9� 9-9 fuelassemblies
137Cs 134Cs 106Ru 95Zr
5/6 node Fuel rod SRACMVP S/M-1
SRACMVP S/M-1
SRACMVP S/M-1
SRACMVP S/M-1
-BURN (%) -BURN (%) -BURN (%) -BURN (%)
e,7 1.09 1.08 0.56 1.11 1.09 1.46 1.03 1.03 0.78 1.04 1.03 0.48
f.7 1.07 1.06 0.85 1.07 1.06 1.32 1.01 1.00 1.10 1.04 1.04 0.10
g.7 1.10 1.10 0.55 1.07 1.08 �0:37 0.95 0.95 0.95 1.09 1.08 0.65
e,8 0.81 0.82 �0:25 0.77 0.77 �0:26 0.88 0.89 �0:56 0.92 0.92 �0:22
f,8 1.02 1.02 0.49 0.99 0.98 1.02 0.95 0.94 0.64 1.04 1.03 0.58
g,8 0.81 0.82 �0:73 0.76 0.77 �1:16 0.88 0.89 �1:35 0.91 0.92 �0:87
h,8 0.98 0.98 �0:10 0.96 0.98 �1:23 0.97 0.97 �0:10 0.97 0.97 �0:10
e,9 1.08 1.08 0.46 1.11 1.09 1.38 1.03 1.02 0.88 1.03 1.03 0.59
f,9 1.09 1.08 0.18 1.11 1.10 0.73 1.03 1.02 0.59 1.03 1.02 0.79
g,9 1.03 1.03 �0:10 1.05 1.05 �0:28 1.04 1.04 �0:10 0.99 0.99 �0:10
h,9 0.98 0.99 �0:81 1.01 1.03 �1:36 1.07 1.08 �0:83 0.96 0.97 �0:62
j,9 0.94 0.95 �1:58 0.99 1.01 �1:89 1.16 1.18 �1:69 0.98 1.00 �1:70
18 node
e,7 1.09 1.09 0.55 1.11 1.11 0.18 1.04 1.03 0.87 1.05 1.04 0.48
f.7 1.06 1.06 0.38 1.06 1.05 0.57 1.00 1.00 0.50 1.05 1.05 0.10
g.7 1.08 1.07 0.56 1.03 1.02 0.59 0.93 0.93 0.54 1.09 1.09 0.28
e,8 0.80 0.80 �0:62 0.74 0.74 �0:14 0.86 0.86 �0:81 0.90 0.90 0.22
f,8 0.99 0.99 0.30 0.94 0.94 0.11 0.92 0.91 0.33 1.03 1.03 0.49
g,8 0.80 0.80 �1:00 0.74 0.75 �1:21 0.85 0.87 �1:73 0.89 0.90 �1:11
h,8 0.97 0.97 �0:10 0.95 0.95 0.32 0.95 0.96 �0:10 0.97 0.96 0.42
e,9 1.08 1.08 0.18 1.11 1.10 0.55 1.04 1.03 0.39 1.05 1.05 0.19
f,9 1.09 1.09 0.28 1.12 1.11 0.54 1.04 1.04 0.48 1.05 1.05 0.00
g,9 1.04 1.04 0.19 1.07 1.07 0.00 1.06 1.05 0.48 1.00 1.00 0.10
h,9 1.01 1.01 �0:49 1.06 1.07 �0:66 1.10 1.10 �0:45 0.96 0.97 �0:62
j,9 0.99 1.00 �0:80 1.09 1.10 �1:18 1.22 1.23 �0:41 0.97 0.98 �0:72
S/M � 1 (%): ((SRAC/MVP-BURN) � 1) � 100
662 T. YAMAMOTO et al.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
Table A2 Calculated FP rod-by-rod distributions of SRAC and MVP-BURN with JENDL-3.3 for the 9� 9-7 fuelassembly
137Cs 134Cs 106Ru 154Eu
5/6 node Fuel rod SRAC MVPS/M-1
SRAC MVPS/M-1
SRAC MVPS/M-1
SRAC MVPS/M-1
(%) (%) (%) (%)
A,1 1.01 1.02 �0:98 1.08 1.11 �2:70 1.19 1.21 �1:65 0.96 0.96 0.00
B,1 1.00 1.01 �0:99 1.04 1.06 �1:89 1.09 1.10 �0:91 0.97 0.98 �1:02
C,1 1.08 1.08 0.00 1.11 1.11 0.00 1.05 1.05 0.00 1.05 1.07 �1:87
D,1 1.13 1.12 0.89 1.16 1.14 1.75 1.04 1.03 0.97 1.10 1.09 0.92
E,1 1.06 1.06 0.00 1.09 1.08 0.93 1.03 1.03 0.00 1.06 1.06 0.00
F,1 1.13 1.12 0.89 1.16 1.15 0.87 1.04 1.03 0.97 1.10 1.09 0.92
G,1 1.08 1.09 �0:92 1.12 1.13 �0:88 1.05 1.06 �0:94 1.05 1.05 0.00
H,1 1.00 1.01 �0:99 1.05 1.06 �0:94 1.09 1.10 �0:91 0.97 0.97 0.00
J,1 1.01 1.02 �0:98 1.08 1.12 �3:57 1.19 1.21 �1:65 0.95 0.95 0.00
B,2 1.02 1.03 �0:97 1.01 1.02 �0:98 0.97 0.97 0.00 1.01 1.01 0.00
C,2 0.86 0.86 0.00 0.80 0.81 �1:23 0.87 0.88 �1:14 0.88 0.90 �2:22
D,2 0.88 0.87 1.15 0.83 0.83 0.00 0.93 0.92 1.09 0.93 0.93 0.00
E,2 1.04 1.04 0.00 1.01 1.00 1.00 0.94 0.93 1.08 1.05 1.04 0.96
F,2 0.86 0.86 0.00 0.80 0.81 �1:23 0.87 0.89 �2:25 0.89 0.89 0.00
G,2 0.89 0.89 0.00 0.85 0.85 0.00 0.95 0.94 1.06 0.93 0.92 1.09
H,2 1.03 1.03 0.00 1.02 1.02 0.00 0.98 0.98 0.00 1.01 1.00 1.00
B,3 0.86 0.86 0.00 0.80 0.81 �1:23 0.87 0.88 �1:14 0.88 0.90 �2:22
C,3 0.87 0.86 1.16 0.80 0.80 0.00 0.91 0.91 0.00 0.93 0.93 0.00
D,3 1.11 1.10 0.91 1.07 1.06 0.94 0.94 0.93 1.08 1.11 1.11 0.00
E,3 0.92 0.92 0.00 0.91 0.90 1.11 1.00 0.98 2.04 0.98 0.97 1.03
F,3 1.08 1.07 0.93 1.08 1.05 2.86 0.99 0.98 1.02 1.09 1.08 0.93
G,3 0.86 0.86 0.00 0.80 0.80 0.00 0.88 0.88 0.00 0.89 0.90 �1:11
D,4 0.97 0.96 1.04 0.99 0.98 1.02 1.06 1.03 2.91 1.01 1.00 1.00
A,6 1.13 1.12 0.89 1.16 1.15 0.87 1.04 1.03 0.97 1.10 1.09 0.92
D,9 1.13 1.12 0.89 1.16 1.15 0.87 1.04 1.03 0.97 1.10 1.09 0.92
continued on next page
Lattice Physics Analysis of Measured Rod-by-Rod FP Inventory Distributions of Burnt BWR 9� 9 UO2 Assemblies 663
VOL. 46, NO. 7, JULY 2009
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14
continued
18/19 node A,1 1.04 1.04 0.00 1.14 1.14 0.00 1.22 1.22 0.00 0.97 0.96 1.04
B,1 1.02 1.02 0.00 1.08 1.07 0.93 1.11 1.11 0.00 0.98 0.98 0.00
C,1 1.07 1.08 �0:93 1.11 1.11 0.00 1.05 1.05 0.00 1.04 1.05 �0:95
D,1 1.11 1.11 0.00 1.14 1.14 0.00 1.03 1.03 0.00 1.09 1.08 0.93
E,1 1.06 1.06 0.00 1.09 1.08 0.93 1.05 1.05 0.00 1.06 1.06 0.00
F,1 1.11 1.11 0.00 1.14 1.14 0.00 1.04 1.03 0.97 1.09 1.09 0.00
G,1 1.08 1.08 0.00 1.12 1.11 0.90 1.06 1.06 0.00 1.04 1.04 0.00
H,1 1.02 1.03 �0:97 1.09 1.10 �0:91 1.12 1.12 0.00 0.98 0.98 0.00
J,1 1.04 1.05 �0:95 1.14 1.14 0.00 1.22 1.23 �0:81 0.97 0.96 1.04
C,2 0.85 0.85 0.00 0.79 0.80 �1:25 0.86 0.87 �1:15 0.87 0.89 �2:25
D,2 0.96 0.96 0.00 0.91 0.90 1.11 0.91 0.91 0.00 0.99 0.99 0.00
F,2 0.84 0.84 0.00 0.78 0.79 �1:27 0.86 0.87 �1:15 0.87 0.88 �1:14
G,2 0.98 0.98 0.00 0.95 0.95 0.00 0.94 0.94 0.00 0.99 0.99 0.00
B,3 0.85 0.85 0.00 0.79 0.80 �1:25 0.86 0.87 �1:15 0.87 0.89 �2:25
C,3 0.84 0.84 0.00 0.77 0.78 �1:28 0.88 0.88 0.00 0.91 0.91 0.00
D,3 1.06 1.06 0.00 1.00 1.00 0.00 0.91 0.90 1.11 1.08 1.08 0.00
E,3 1.01 1.00 1.00 0.99 0.98 1.02 0.98 0.98 0.00 1.05 1.05 0.00
F,3 1.05 1.04 0.96 1.02 1.01 0.99 0.96 0.95 1.05 1.07 1.07 0.00
G,3 0.83 0.84 �1:19 0.76 0.79 �3:80 0.85 0.86 �1:16 0.88 0.89 �1:12
D,4 0.94 0.93 1.08 0.94 0.91 3.30 0.94 1.00 �6:00 1.00 0.98 2.04
A,6 1.11 1.11 0.00 1.14 1.14 0.00 1.04 1.03 0.97 1.09 1.09 0.00
D,9 1.11 1.11 0.00 1.14 1.14 0.00 1.04 1.03 0.97 1.09 1.09 0.00
S/M � 1 (%): ((SRAC/MVP-BURN) � 1) � 100
664 T. YAMAMOTO et al.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
Dow
nloa
ded
by [
The
UC
Irv
ine
Lib
rari
es]
at 1
4:43
08
Nov
embe
r 20
14