Nuclear data for advanced MOX fuels

Directorate-General For Research And Innovation European Commission

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Nuclear energy and safety

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• Publié par: DIRECTORATE-GENERAL-FOR-RESEARCH-AND-INNOVATION-EUROPEAN-COMMISSION
• Nombre de lectures: 24
• Langue: English
• Poids de l'ouvrage: 2 Mo

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European Commission
Community Research
Project report
Nuclear Science and Technology
Nuclear data for advanced
MOX fuels
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DG Research/D.11.3 - R & T programme 'Nuclear fission safety 1994-98'
Contact: Mr G.A. Cottone
Address: European Commission, rue de la Loi/Wetstraat 200 (MO 75 5/43),
B-1049 Brussels - Tel. [phone]; fax [phone] European Commission
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and
Nuclear data for advanced
MOX fuels
" S. PILATE, " T. MALDAGUE,3 R. JACQMIN,
21 Ch. CHABERT, 3 S. VAN WINCKEL, ? Ch. DE RAEDT,
41 J.C. KUIJPER, 9 G. NICOLAOU,6) A. VENTURA
" BELGONUCLEAIRE
Brussels, Belgium
?CEA
Cadarache, France
3SCK-CEN
Mol, Belgium
''NRG
Petten, The Netherlands
5 EC- JRC-ITU
Karlsruhe, Germany
*ENEA
Bologna, Italy
Contract No FI4I-CT95-0002
FINAL REPORT
Work performed as part of the European Atomic Energy Community's R&T Specific Programme
"Nuclear Fission Safety 1 994-1 998"
Area A.2: Exploring innovative approaches/Fuel cycle concepts
Directorate-General ..-'
Science, Research and Development
2000 EUR19126EN LEGAL NOTICE
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Cataloguing data can be found at the end of this publication
Luxembourg: Office for Official Publications of the European Communities, 2000
ISBN 92-828-8491-0
© European Communities, 2000
Reproduction is authorized provided the source is acknowledged
Printed in Belgium
PRINTED ON WHITE CHLORINE-FREE PAPER EXECUTIVE SUMMARY
Scope
This study on "Supporting Nuclear Data for Advanced MOX Fuels" was performed in the
framework of the specific programme on Nuclear Fission Safety of the Framework
Programme for the European Atomic Energy Community (1994-1998), under contract
number FI4I CT 95 0002, by BELGONUCLEAIRE, SCK'CEN. ECN (now NRG), CEA, ITU
and ENEA under the co-ordination of BELGONUCLEAIRE (Mr. S. Pilate).
It supports the larger study entitled "Evaluation of Possible Partitioning and Transmutation
Strategies and of Means for Implementing Them" performed in parallel under Contract
number FI4I CT 95 0006, which includes scenarios of actinide recycling in thermal reactors
(PWRs) and fast reactors.
This final report is subdivided into 3 chapters :
A. Evaluation of MOX Fuel Irradiations in Thermal Reactors
B.n of MOX Fuels in Fast Reactors
C. Evaluations of Fundamental Data Files.
A. Evaluation of MOX Fuel Irradiations in Thermal Reactors
Chapter A is primarily devoted to the evaluation of MOX fuel irradiations in thermal reactors
mainly done by BELGONUCLEAIRE for the PWRs BR3 (with the support of SCK'CEN) and
Beznau-1 (ARIANE Programme) and by CEA for the PWR of Saint-Laurent B1. The
reactors of Saint-Laurent B1 and Beznau-1 are present-day PWRs, while BR3 was
representative of highly moderated neutron spectra.
Major findings
Table A.XXII page 57 provides the calculation-to-experiment ratios obtained for the spent
masses (related to U238) of the U, Pu, Am and Cm isotopes from the evaluations made at
CEA and BELGONUCLEAIRE. The calculation methods used have been validated
numerically ; at CEA, this was made by comparison with Monte Carlo reference results.
While the 2-CT accuracy of the chemical analysis is quoted to be 2 % or less, deviations
between calculated and measured values can only be considered significative when they
exceed 4 % ; some systematical errors may indeed have influenced the results, as the range
of deviations also indicates.
For the uranium isotopes, consistent trends were found : the U235 residual masses were
rather correctly reproduced by calculation, while the U236 masses were underestimated.
CEA could reduce this underestimate using a recent U235 evaluation with revised capture
resonance data.
For the plutonium isotopes, the masses of the 3 major isotopes Pu239, Pu240 and Pu241
were satisfactorily reproduced by calculation. Pu242 was well predicted in the Beznau-1 and
BR3 evaluations, but SLB1 calculations underestimated its mass by about 4 %. The mass of Pu238 was always underestimated. In SLB1 and Beznau-1, representative of
the present industrial MOX irradiations, the underestimate was 4 to 8 %.
Concerning americium isotopes, while the CEA calculations reproduce well Am241 masses,
the BN ones overestimated them by some 13 % ; the situation is reversed for Am243, well
reproduced by BN and not by CEA (underestimate by 7 %, in connection with the
underestimate on Pu242). The reasons for these differences could not be identified.
Am242m was largely underestimated by both groups.
All curium isotopes wered : Cm243 largely (by 10 to 30 %), Cm244 by about
7 %, and Cm245 by about 10 %.
From the evaluation of these irradiations, which covered a large range of burnup values,
from 10GWd/t to nearly 100GWd/t (at the hot spot in BR3), no clear correlation of the
actinide mass prediction could be observed versus burnup.
In the highly moderated neutron spectrum of BR3 (where the equivalent moderation ratio is
4.5 versus the standard PWR value of 1.9), the discharged masses of Pu isotopes (except of
Pu238) and of Am241 were reproduced by calculation within the error margin of the
measurements, as was also the case in the SLB1 and Beznau-1 evaluations. The deviation
on Cm244 mass was about twice as high. Except for this, the C/E ratios were found to be
similar in overmoderated or in standard lattice PWRs.
Complementary calculations
Complementary calculations made at SCK'CEN and at CEA on the effect of method
approximations confirmed the validity of these results, obtained with 172-group cross-
sections derived from the JEF2.2 file and with transport theory using a fine geometrical
representation. It was also demonstrated that simpler calculation methods (ORIGEN 2) did
not yield satisfactory results for the highly moderated BR3 fuel. The use of the SAS2H
module of SCALE 4.3 led to improved, but still often insufficiently accurate, actinide
inventories.
Application to the evaluation of P&T strategies
In the parallel study contract FI4I CT 95 0006, "Evaluation of possible P&T Strategies and of
Associated Means to Perform Them", the following actinide recycling scenarios were
considered in thermal reactors (PWRs):
y a recycling of Pu only (current MOX) or of Pu+Am in MOX fuel (U02Pu02Am02);
> standard moderator-to-oxide volume ratio (1.9 as in most present PWR rod lattices), or
an enhanced ratio (3.5 with an enlarged rod lattice).
High moderation is a way to burn more effectively plutonium, especially Pu239.
The evaluations of MOX fuel irradiations give indications on the accuracy of recycling
strategies for what concerns mass balances, waste toxicities in the repository after long-term
storage, and shorter-term hazards.
According to the trends found in this study :
- the reduction calculated for total Pu masses would be well predicted ;
- the build-up of Am quantities would be rather correct;
- the strong increase in Cm quantities (mainly Cm244) could be underestimated by some
10%.
IV With respect to long-term waste toxicities, after 1000 years, as the waste toxicity is
dominated by the chain Pu241+Am241, the toxicity would be well predicted in the MOX
recycling cases considered so far.
After 10,000 years, the situation is more complex. As MOX fuel containing Pu, Am (and
possibly Cm), is recycled several times, the rise in Cm quantities makes it necessary to
calculate correctly also the spent masses of Cm243+Cm244+Cm245, which may represent,
depending on the case, 30 to 40 % of the total toxicity. If their contribution is underestimated
with the best methods by 15 to 30 %, the total radio-toxicity can be underestimated by nearly
10%.
Such a deviation may not be very important in itself, except if one precisely wishes to
compare one strategy versus another, for example overmoderation versus standard
moderation.
Sensitivity studies
At NRG, a code CSS1MAT was developed to calculate "density-to-density" and "one-group
cross section-to-density" sensitivity matrices for burnup schemes. It was applied to BR3
irradiations. It can be expanded to calculate a "multigroup cross-section-to-density"
sensitivity matrix.
In an earlier work at CEA, the impact of uncertainties in cross-sections and decay times on
discharged actinide mass balances had been assessed in a preliminary way for the multiple
recycling of Pu (only) in PWRs, either with standard lattice or with overmoderated lattice.
When comparing these uncertainties on discharged Pu, Am and Cm isoto