3 Concluding Remarks
8. ORIGEN-2 Libraries Based on JENDL-3.2 for PWR-MOX Fuel
Hideki MATSUMOTO, Masaaki ONOUE, Yoshihisa TAHARA
MITSUBISHI HEAVY INDUSTRIES, LTD
1. INTRODUCTION
ORIGEN-21 code and its libraries are provided from Oak Ridge National Laboratory for several core calculations. However, the problems have been pointed out, for example, the libraries are based on old cross section data, the libraries are not consistent with the current high-enriched fuel, and so on. In 1999, the newest version of ORIGEN-2 libraries2 was developed based on JENDL3.23 for LWR-UO2 fuel and FBR fuel in the Working Group on Evaluation of Nuclide Production, Japanese Nuclear Data Committee.
The PWR-UO2 libraries were verified through the comparison of nuclide contents between measured and calculated results.
In 2000, a set of ORIGEN-2 libraries4 for PWR-MOX fuel was developed based on JENDL-3.2 in the Working Group of the committee. This report describes the development of PWR-MOX libraries for ORIGEN-2.
2. MODEL FOR GENERATING PWR-MOX LIBRARY
Three parameters mainly affect the ratios of plutonium ]E+oi isotopes, so called Pu-Vectors. The first one is the type
of core in which the plutonium is produced. The second one is the discharged burnup. The third one is the cooling time that mainly affects Pu-241 content. In MOX cases, spectrum effect on one-group cross-section has to be also taken into account so as to increase the accuracy of ORIGEN calculation. So many libraries need to be provided because ORIGEN-2 treats the neutron spectrum only through its libraries. However, less libraries are preferable for ORIGEN-2 users.
Therefore, some efforts to reduce the number of libraries for PWR-MOX were done.
1E+00
1
I
1E-01
1E-02
UO2 PuVectorl Pu Vector2 PuVector3 Pu Vector4
1E-02 1E+00 1E+02 1E+04 1E+06
Energy (eV)
Fig.l Neutron Spectrum vs Pu Vectors (10wt%Pu-t. MOX & 4.1wt%U-235)
JAERI-Conf 2001-009 Pu Vector 1
4.1 Pu Vector 2
2.1
Pu Vector 3 1.9
Pu Vector 4 0.0
The effect of Pu-Vector on neutron spectrum was first investigated. Each plutonium isotope is a strong absorber, so the neutron spectrum would be strongly dependent on Pu total content. Fig.l shows the comparison of neutron spectrum within fuel lumps among the 4 fuel types of MOX and UO2. The Pu-vectors are shown in Table-1. They are selected to cover the range of PWR MOX utilization in Japan.
From the Fig.l, it can be seen that the 4 neutron spectrum profiles of MOX fuel are not so different. So, Pu-vector2 and Pu-Vector3 can be treated as the same Pu-Vector for generating the ORIGEN-2 cross-section sets.
Table 2 Specification of new set of ORIGEN-2 Library for PWR MOX Library ID Pu Content (wt%Pu-t) Pu Content (wt%Pu-f) Enrich (wt%U-235)
Three sets of Pu content, 5%Pu-t for low content MOX(PWRM0205), 10%Pu-t for middle content MOX(PWRM0210) and 13%Pu-t for high content MOX(PWRM0213), were selected as shown in the middle column of Table-2. The Pu-vector in the middle column of Table 2 stands for the typical one to be used in Japanese PWR. 13%Pu-t for PWRMOl 13 were chosen as the limit of Pu content. And 5%Pu-t for PWRM0305 were selected as the best Pu Vector come from gas-cooled reactor
The pin cell model that had been used for UO2 libraries was employed to generate the new ORIGEN-2 libraries for PWR-MOX, too. To check the applicability of the pin cell model in MOX cases, the comparison of nuclide production among the pin cell model, two types of the unit assembly model and the four-assembly model was carried out. Objectives of those models are (1) effect of water holes, (2) effect of Pu content distribution and (3) effect of UO2 assembly neighboring MOX assembly, as shown in Fig.2 and Table 3.
Table 3 Specification of Sensitivity Pu Content Distribution
Pu Vector Enrichment(wt%U-235) Pu Content (wt%Pu-f)
Cold
Remarks (Drawings in Fig.l)
Study Calculation for Model Unit Assembly
Same as PWRM02XX 0.2 shown in Table2
0.2
(a) Pin Cell (b) Pu content distribution (c) Average Pu content
®
-O|G|Cgioioic5 Op
MOX
(d) Four assembly
Fig.2 Calculation Model for sensitivity study with respect to nuclide production
JAERI-Conf 2001-009 Pu-242 •••••Am-241 |
A comparison of the main heavy nuclides is Table 4 Comparison of heavy nuclide contents (67GWd t) shown in Table 4 and Fig.3. From the
comparison between Pu content distributed model and smear model, it can be seen that the effect on nuclide production is negligibly small.
From the comparison between the pin cell model and the unit assembly model, it can be also seen that the single pin cell model
i
represents well the assembly model.
However, the agreement of the contents obtained with the pin cell model and with r* 8
the four-assembly model is not good, | « 14.1% for Am-241, 13.9% for Npr239 and | ,
•z
so on.
2
Table 5 and Fig.4 show the comparison of
0.
Fission Products (FP) among the four models. The effects of water hole and Pu content distribution are as negligibly small as in heavy nuclide case. Large discrepancy, 35%, about Gd-155 content can be seen between the pin cell model and the four-assembly model. Considering Gd-155 absolute content is very small as shown in Fig.4, the absolute error is not so large.
However, Gd-155 content is important for Burnup Credit evaluation. So, more investigation was done to select the model.
Solid Lines: Single Pin Cell Model Dashed Lines: Four-Assembly Model
30000 BurnupfMWd/i)
(jp-239
"^T-13.9%
60000 70000
Fig.3 Actinide Number Densities vs Burnup Table 5 Comparison of FP contents (67GWd/t)
Sm-149
Taking a careful look to the results that Np-239, Pu-239, Am-241, Cm-242 and Gd-155 contents obtained from the pin cell model did not agree well with those from the four-assembly model, it has been
found that they are the nuclides produced Fig.4 FP Number Densities vs Burnup mainly as decay product. So, the amount of each production is sensitive to the irradiation length. The irradiation length is strongly dependent on specific power. A nominal power is applied to the depletion
Sm-149
50.0 UO2-1 in Quadruple Model UO2-2 in Quadruple Mode]
Nominal Power
MOXUO2-1
calculation in the pin cell model. On the other hand, in the four-assembly model, the specific power of each assembly changes with depletion so that the average power corresponds the nominal power. Fig.5 shows the comparison of the power level between the pin cell model and the four-assembly model. It can be seen the MOX assembly power increases with burnup and it is larger than nominal power level. So, the pin cell Fig.5 Specific Power vs Burnup calculation was performed again with the adjusted specific power that corresponds the average specific power of MOX in four-assembly model during depletion. The comparison between the pin cell and four-assembly model is shown in Table 6.
Table 6 also shows the result with buckling adjustment. It was the case where the leakage of Bl calculation of pin cell model was adjusted to that of MOX in four-assembly model. It can be seen that the large error obtained in previous comparison in Table 4 and Table 5 are reduced to 10% or less except for Gd-155 (18.8%) and Sm-149 (11%) (see Table 6). With bucking adjustment, they are also within 10%. However, the buckling adjustment is case-wise problem and depends on UO2 assembly characteristics, numbers and so on.
In adjusted power case, Gd-155 and Sm-149 errors are still existing, but the absolute value of Gd-155 content is very small, so its error 18.8% is acceptable. The 11% error of Sm-149 is also acceptable for ORIGEN-2 calculation.
Table 6 Comparison between Pin Cell model(p) and Four-assembly model(f)
U-235
The appropriate power level (specific power) can be applied in ORIGEN-2 calculation as input, so it should not be considered in generating the ORIGEN-2 library. Therefore, the pin cell model is appropriate and selected to generate the set of ORIGEN-2 libraries for PWR-MOX, after all.
JAERI-Conf 2001-009
3. Verification
To ensure the new set of ORIGEN-2 libraries for PWR MOX, the contents obtained with ORIGEN-2 were compared to the PIE data5 that was published by CRIEPI. The PIE was carried out at ITU in Europe.
The two samples of PIE were irradiated in a 14X14 assembly in a commercial PWR in Europe. Its initial Pu content is 5.07wt%Pu-t and its isotopic ratio (Pu238/Pu239/Pu240/Pu241/Pu242) is 1.5/59.0/24.4/9.2/4.9/1.0 respectively. The discharged burnups of the samples are 46.04GWd/t (MOX1) and 47.50GWd/t (M0X2).
The summary results, which were shown in Ref. (2), are shown here in Table 7 again regarding 46.04GWd/t(MOXl). From Table 7, it can be found that the agreement between measured and calculated contents for U and Pu isotopes, minor actinides and fission products are within about 10%, about 20%
and about 20%, respectively. The discrepancies of some nuclides are beyond the estimated values.
However, there are many unknown parameters in both PIE and calculational model. Considering such a simple calculation as ORIGEN-2, the agreement is acceptable. If more sophisticated model is needed, SWAT6, SRAC7 and PHOENIX-P8 can be applied.
Table 7 Comparison between PIE (E) Nuclide
results and ORIGEN-2(C) results (C/E) Nuclide
* Results obtained with the unpublished version of ORIGEN-2 library that was generated using a 14X14-assembly configuration.
4.SUMMARY
A set of ORIGEN-2 libraries for PWR MOX fuel was developed based on JENDL-3.2 in the Working Group on Evaluation of Nuclide Production, Japanese Nuclear Data Committee. The calculational model
generating ORIGEN-2 libraries of PWR MOX is explained here in detail.
The ORIGEN-2 calculation with the new ORIGEN-2 MOX library can predict the nuclides contents within 10% for U and Pu isotopes and 20% for both minor actinides and main FPs.
5. Acknowledgement
This report describes the development of PWR-MOX libraries for ORIGEN-2 about JAERI-Data/Code 2000-036. One of the authors (H.M) is grateful to the other authors of JAERI-Data/Code 2000-036, Mr.Suyama, Mr.Sasahara and Mr.Katakura for their effort to develop the ORIGEN-2 libraries.
6. REFERENCES
1) A.G.Croff. "ORIGEN-2 - A Revised and Updated Version of the Oak Ridge Isotope Generatioin and Depletion". ORNL-5621, July 1980
2) K.Suyama, et.al. "ORLIBJ32:The Set of New Libraries of ORIGEN2 Code Based on JENDL-3.2".
JAERI-Data/Code99-003, February 1999
3) T.Nakagawa, et.al "Japanese Evaluated Nuclear Data Library Version 3 Revision-2:JENDL-3.2".
J.Nucle.Sci.Technol., Vol.32, pp.1259-1271, December 1995
4) K.Suyama, et.al. "ORIGEN2 Libraries Based on JENDL-3.2 for LWR-MOX Fuels", JAERI-Data/Code 2000-036
5) A.Sasahara, et.al. "Post irradiation examinations and the validity of computational analysis for high burnup UO2 and MOX spent fuels", CRIEPIT95012,1995
6) K.Suyama, et.al.JAERI-Data/Code97-047, November 1997 7) K.Okumura, et.al.JAERI-Data/Code96-015, March 1996
8) R.J.J. Stamm'ler, et.al. "Methods of Steady-State Reactor Physics in Nuclear Design", ACADEMIC PRESS, 1983.
JP0150770 JAERI-Conf 2001-009