Nuclear Data Needs for Generation IV Nuclear Energy Systems

Nuclear Data Needs For Generation IV Nuclear Energy Systems Proceedings of the International Workshop

Editor Peter Rullhusen

Nuclear Data Needs For Generation IV Nuclear Energy Systems Proceedings of the International Workshop

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Nuclear Data Needs For Generation IV Nuclear Energy Systems Proceedings of the International Workshop

Antwerpen, Belgium

5-7 April 2005

Editor

Peter Rullhusen European Commission, Joint Research Centre, Institute for Reference Materials and Measurements, Belgium

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NUCLEAR DATA NEEDS FOR GENERATION IV NUCLEAR ENERGY SYSTEMS Proceedings of the International Workshop Copyright © 2006 by World Scientific Publishing Co. Pte. Ltd. All rights reserved. This book, or parts thereof, may not be reproduced in any form or by any means, electronic or mechanical, including photocopying, recording or any information storage and retrieval system now known or to be invented, without written permission from the Publisher.

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INTERNATIONAL ADVISORY COMMITTEE J. Chang M. E. Dunn P. Finck A. Hasegawa R. Jacqmin J. Katakura A. Koning K. Kozier

KAERI ORNL ANL JAERI CEA JAERI NRG AECL

L, Leal A. Nichols D. W. Nigg C. Nordborg P. Oblozinsky E. Pitcher A. Plompen M. Salvatores

LOCAL ORGANISER COMMITTEE

G. Giorginis F.-J. Hambsch G. Lovestam W. Mondelaers S. Oberstedt A. Plompen P. Schillebeeckx P. Siegler

ORNL IAEA-INDC INEEL OECD-NEA BNL LANL IRMM CEA

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PREFACE

On 6 March 2003 a Technical Exchange and Cooperation Arrangement in the field of Nuclear-Related Technology Research and Development was signed between the Department of Energy of the United States of America and the European Atomic Energy Community. Several joint programs and co-operative research projects have been agreed in the mean time between DOE and Euratom. At the first Steering Committee meeting between DOE and IRC on 23 February 2004 in Brussels, it was decided to organise a workshop on Nuclear Data Needs for Generation IV Nuclear Energy Systems. This workshop was organised jointly by JRC on behalf of Euratom, DOE and AECL on 5-7 April 2005 in Antwerp, Belgium. The workshop was attended by 70 participants from 20 countries. During three consecutive days recent achievements were presented on sensitivity analysis, model calculations, estimates of uncertainties, and the present status of nuclear data bases. Although detailed work on the different Generation IV systems did not yet start at the time of the workshop, it was impressive to see the amount of work and the results already achieved in connection with the development of these new systems. The local organisers are very grateful to all participants contributing with interesting presentations to the workshop and to the session chairmen for monitoring the lively discussions. I would like to thank the international advisory committee and especially Hussein Khalil (DOE) and Robert Speranzini (AECL) for their continuous support.

Peter Rullhusen Conference Chair

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CONTENTS Organising and Scientific Advisory Committees Preface

v vii

Nuclear Data Needs for Generation IV Systems Future of Nuclear Energy and the Role of Nuclear Data P. Finck

3

Nuclear Data Needs for Generation IV Nuclear Energy SystemsSummary of U.S. Workshop T. A. Taiwo, H. S. Khalil

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Nuclear Data Needs for the Assessment of Gen. IV Systems G. Rimpault

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Nuclear Data Needs for Generation IV- Lessons from Benchmarks S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn

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Core Design Issues of the Supercritical Water Fast Reactor M. Mori, A. Rineiski, W. Mashek, V. Sinitsa

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GFR Core Neutronics Studies at CEA J. C. Bosq, V. Brun-Magaud, G. Rimpault, J. Tommasi, A. Conti, J. C. Gamier

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Comparative Study on Different Phonon Frequency Spectra of Graphite in GCR Young-Sik Cho, Kang-Seog Kim, Do Heon Kim, Young-Ouk Lee, Jonghwa Chang

58

Innovative Fuel Types for Minor Actinides Transmutation D. Haas, A. Fernandez, J. Somers

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IX

X

The Importance of Nuclear Data in Modeling and Designing Generation IV Fast Reactors K. D. Weaver

70

The GIF and Mexico- "Everything is Possible" C. Arredondo Sanchez

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Benchmarks, Sensitivity Calculations, Uncertainties Sensitivity of Advanced Reactor and Fuel Cycle Performance Parameters to Nuclear Data Uncertainties G. Aliberti, G. Palmiotti, M. Salvatores, T. K. Kim, T. A. Taiwo, I. Kodeli, E. Sartori, J. C. Bosq, J. Tommasi

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Sensitivity and Uncertainty Study for Thermal Molten Salt Reactors A. Bidaud, T. Ivanona, V. Mastrangelo, I. Kodeli

101

Integral Reactor Physics Benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) J. B. Briggs, D. W. Nigg, E. Sartori

113

Computer Model of an Error Propagation Through Micro-Campaign of Fast Neutron Gas Cooled Nuclear Reactor E. lvanov

128

Combining Differential and Integral Experiments on 239Pu for Reducing Uncertainties in Nuclear Data Applications T. Kawano, K. M. Hanson, S. C. Frankle, P. Talou, M. B. Chadwick, R. C. Little

138

Sensitivity of Activation Cross Sections of the Hafnium, Tanatalum and Tungsten Stable Isotopes to Nuclear Reaction Mechanisms V. Avrigeanu, M. Avrigeanu, F. L. Roman, R. A. Forrest, R. Eichin, H. Freiesleben, K Seidel

145

Generating Covariance Data with Nuclear Models A. J. Koning

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Sensitivity of Candu-SCWR Reactors Physics Calculations to Nuclear Data Files K. S. Kozier, G. R. Dyck

163

The Lead Cooled Fast Reactor Benchmark BREST-300: Analysis with Sensitivity Method V. Smirnov, V. Orlov, A. Mourogov, D. Lecarpentier, T. Ivanovo

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Sensitivity Analysis of Neutron Cross-Sections Considered for Design and Safety Studies of LFR and SER Generation IV Systems K. Tucek, J. Carlsson, H. Wider

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Experiments INL Capabilities for Nuclear Data Measurements Using the Argonne Intense Pulsed Neutron Source Facility J. D. Cole, M. W. Drigert, R. Aryaeinejad, D. W. Nigg, R. V. F. Janssens, B. J. Micklich, G. Ter-Akopian

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Cross-Section Measurements in the Fast Neutron Energy Range A. Plompen

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Recent Measurements of Neutron Capture Cross Sections for Minor Actinides by a JNC and Kyoto University Group H. Harada, H. Sakane, S. Nakamura, K. Furutaka, J.-I. Hori, T. Jujii, H. Yamana

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Determination of Minor Actinides Fission Cross Sections by Means of Transfer Reactions M. Aiche, G. Barreau, S. Boyer, S. Czajkowski, D. Dassie, C. Grosjean, A. Guiral, B. Haas, B. Jurado, B. Osmanov, E. Bauge, M. Petit, E. Berthoumieux, F. Gunsing, L. Perrot, C. Theisen, F. Michel-Sendis, A. Billebaud, J. N. Wilson, I. Ahmad, J. P. Greene, R. V. F. Janssens

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Evaluated Data Libraries Nuclear Data Services from the NEA H. Henriksson, Y. Rugama

235

Nuclear Databases for Energy Applications: An IAEA Perspective R. Capote Noy, A. L. Nichols, A. Trkov

244

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Nuclear Data Evaluation for Generation IV G. Noguere, O. Bouland, A. Courcelle, E. Dupont, O. Serot, J. C. Sublet

253

Improved Evaluations of Neutron-Induced Reactions on Americium Isotopes P. Talou, T. Kawano, P. G. Young, M. B. Chadwick, E. J. Pitcher

262

Using Improved ENDF-Based Nuclear Data for Candu Reactor Calculations /. Prodea

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A Comparative Study on the Graphite-Moderated Reactors Using Different Evaluated Nuclear Data Do Heon Kim, C.-S. Gil, Y.-S. Cho, Y.-O. Lee, J. Chang

278

Author Index

285

NUCLEAR DATA NEEDS FOR GENERATION IV SYSTEMS

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FUTURE OF NUCLEAR ENERGY AND THE ROLE OF NUCLEAR DATA P. FINCK Argonne National Laboratory

1. Introduction Nuclear energy plays a major role to secure energy supply in future. In fact, nuclear energy provides a unique and cost-effective answer to the problem of C0 2 emission and to the requirement of a sustainable development (see Fig. 1) in particular in view of a potential growth in energy demand. In Fig. 2 the per capita GDP is shown as the per capita energy consumption, with evidence for a potential spectacular growth in energy demand due to population and GDP increase in developing countries (e. g. China, India). At present, there are indications for a nuclear energy revival, which is based on the consolidation and evolution of present reactor concepts, but also on the definition of new requirements for both future reactors and their associated fuel cycles. The key issues for nuclear energy development are: • Economics • Safety • Proliferation resistance • Waste minimization • Supplies of uranium These last two points are crucial to insure sustainability. However, the current once through cycle is not sustainable. Advanced technologies (closed fuel cycles) are required. As one example, in Fig. 3 it is shown the impact on spent fuel stockpiles, of the introduction of advanced technologies, leading to the deployment of closed fuel cycles based on fast reactors (LMR's), according to different hypothesis on energy demand growth. In this respect, the first phase of the Generation IV initiative has produced a wide international consensus around six preferred concepts and has underlined the need for improved solutions for the back-end of the fuel cycle, and in particular the essential benefits of the closed fuel cycle option, associated to the development of fast reactors. 3

• No COj emissions and no contribution to Global Warming

An already competitive energy source Safety improvements in 3rd Gen reactors are already significant

Promising assists to produce Hydrojer

Fig. 1 Benefits from using nuclear energy.

Sweden o

Norway

o

.Malaysia -Turkey Thailand • Trie Philippines

Per-capita energy consumption (tor»s|of oWpmstm) Fig. 2 Energy use vs. GDP

Innovative reactor concept and fuel cycles are a potential source for an increased role of reliable and well validated nuclear data, both at the stage of feasibility studies and for more detailed design assessments. However, no new requirement in the nuclear data field will be credible, without a sound assessment of uncertainties and their impact. This type of

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assessment, which is at the heart of the present workshop, has already started in some laboratories, and first results will be presented. 120,000 -i

2000

.

2010

•

2020

•

. y

2030

1

2040

2050

Fig. 3 Estimated Stockpile of Spent Fuel (MT/HM)

2. The role of nuclear data in developing future generation nuclear energy Some examples have been studied using a simplified approach. The systems chosen for the analysis are: • Extended burnups for LWR's • Very High Temperature Reactors • Fast Reactors: GFR, SFR, LFR • Accelerator Driven Systems The fuel cycle performances were also analyzed. As far as uncertainties, it has to be stressed that since credible and complete covariance data were not available, estimates of these data have been used, based on physics judgement and on the performances of nuclear data in the analysis of selected, high precision, integral experiments. Uncertainties were propagated through standard static and depletion codes using sensitivities coefficients calculated with Generalized Perturbation Theory (GPT).The results should not be used in an absolute sense, but have a relative value, as a first indication of major trends. A summary of the most significant data are given in the following tables (I through III).

6 Table I. Extended Burnup for LWR's Uncertainties on Kerr at Beginning of Cycle (BOC) Fuel Burnups in LWR's Total uncertainty is -510 pcm have been slowly (U235: 350 pcm; U238: 360 pcm) increasing in order to Uncertainties on Kefr at End of Cycle (EOC) reduce costs. Total uncertainty is -1220 pcm Current plans indicate (Pu239: 620 pcm; Pu241: 320 pcm; that burnup might be Pu 240: 620 pcm; U238: 690 pcm) increased in the [50-100] GWd/ton range. Uncertainty on Burnup Swing -2240 pcm, dominated by Pu240 As burnup increases, the Uncertainties on isotopics neutronic contributions of Largest (-6%) on Np237 Pu238 Pu240 transuranics become Am and Cm isotopes predominant Due to capture of U236, Pu240, Am241, 243 and Cm242, 244 Table n. The Very High Temperature Reactor Graphite moderated Uncertainties on Kefr at BOC U235 enrichment > 10% Total uncertainty is ~580 pcm Very High Burnup (U235: 360 pcm; U238: 430 pcm) Very High Thermal Uncertainties on Ketr at EOC Efficiency Total uncertainty is -1070 pcm High Outlet Temperature (Pu240: 630 pcm; Pu239: 570 pcm; U238: 550 pcm) Uncertainty on Burnup Swing -1749 pcm, dominated by Pu240 Note that other data uncertainties can play a significant role: graphite S(a,P); lower resonances of minor actinides if VHTR's are conceived in a burner mode. Table m. Fast Breeder or Burner modes. Fuels, structures, and reflectors might contain new materials (Zr, Si), as in the case of GFR's

Reactors Uncertainties on Ketr: ~2000pcm U238 (inelastic, capture), Pu239 and Pu241 (fission) still predominate but Si (inelastic) contributes 430 pcm in GFR Uncertainties on void worth: -12-20% He void worth: small absolute value < uncertainties. Sign is unclear Uncertainty on Doppler worth: -5-10% Uncertainty on Burnup Swing: less than 1000 pcm

3. Nuclear Data Needs As compared to current concepts, these examples show a slight increase in overall uncertainties. However, they do not indicate an urgent need for large amounts of new data, only isolated data might be needed urgently. But we must recall that predictive codes for existing reactors achieve very low uncertainty, not because of the quality of the nuclear data, but thanks to a series of "adjustments" to their data libraries. These adjustments were made possible by the accumulation of integral data from measurements in reactors and mock-up facilities. These were lengthy and costly, and might not be available in the future. The need to reduce costs and research and development schedules will

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lead to use powerful simulation techniques to (partially) replace experiments. The required ingredients are: • Better codes, based as far as possible on first principle description of phenomena • Better data. This in turn means: - To reduce uncertainties on key principal reactions - To provide reliable, complete, and systematically derived covariance files. These data are absolutely needed and should be developed in a consistent way. - The need for a few "clean" integral experiments (simple and well documented). 4. Conclusions It is recognized that nuclear energy has a key role in future energy supply. As far as nuclear data, specific nuclear data needs for existing and new concepts can be justified. Covariance files are needed to move the reactor design community towards a more efficient R, D & D model. A systematic approach for defining needs and for fulfilling the related requirements has to be used. In particular, new experiments should address effectively the problem of reducing uncertainties. In fact, required target accuracies can be sometimes smaller than achievable accuracies in current differential cross section measurements and new experiments as well as new approaches should be envisaged and implemented.

NUCLEAR DATA NEEDS FOR GENERATION IV NUCLEAR ENERGY SYSTEMS - SUMMARY OF U.S. WORKSHOP T. A. TAIWO AND H. S. KHALIL Nuclear Engineering Division, Argonne National 9700 South Cass Avenue, Argonne, IL 60439,

Laboratory U.S.A.

A workshop on the data needs for Generation IV Nuclear Energy Systems was held in the U.S. in April 2003. A summary of this workshop is provided in this paper. Discussions during the workshop evolved along the traditional nuclear data topical areas of data needs, measurements, evaluations, processing and validation. Recommendations were made on how the Generation IV needs could be better defined and on approaches for resolving the needs.

1. Introduction The six advanced nuclear energy systems that have been selected for development under the Generation IV program target significant advances over current generation and evolutionary systems in the areas of sustainability, economics, safety and reliability, and proliferation resistance and physical protection. These nuclear systems are the Very High Temperature Reactor (VHTR), the Gas-Cooled Fast Reactor, the Lead-Cooled Fast Reactor (LFR), the Sodium-Cooled Fast Reactor (SFR), the Supercritical-Water-Cooled Reactor (SCWR), and the Molten-Salt-Cooled Reactor (MSR). As a group, the six systems employ a large spectrum of new fuels (particles, dispersion, solid solution), a variety of core geometries (pins, pebbles, hexagonal blocks, annular pins, plates), and various coolants (supercritical H20, sodium, lead alloys, C02, and molten salts). These systems introduce modeling challenges that differ from those of current generation Light-Water Reactor (LWR) cores. Therefore, while mature tools and data exist for the analysis of LWRs, the ability of these tools to model accurately the advanced systems has to be assessed systematically. Initial U.S. efforts directed to improving capabilities for modeling the physics of the Generation IV systems have focused on evaluating the applicability of existing tools and data, identification of the additional needs, and identification of integral benchmarks that can be used for verification and validation of existing and emerging tools. Several workshops were organized to

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identify analysis needs of Generation IV systems and relevant activities directed to meeting these needs. [1,2,3] Table 1. Topical Areas for the U.S. Workshop* Topical Area Title of Presentation Nuclear Data Needs for Gen IV Program Overview Needs for Gas-Cooled Reactors (VHTR, GFR) Generation IV Nuclear Data Needs for Liquid Metal Reactors (SFR and LFR) Energy Systems Perspective on Nuclear Data Needs for Advanced Nuclear Systems Nuclear Data Processing, Application of Covariance Data to Reactor Design Applications, and ORNL Cross-Section Processing Capabilities and Validation Experience for Advanced Reactor Applications Performance of New Los Alamos Actinide Evaluations Covariances in JENDL-3.3 and covariance evaluation with the KALMAN system Theory and Evaluation Status of Work on the ENDF/B-VII Library EMPIRE: Advanced Tool for Nuclear Reaction Data Evaluation Systematic Re-Evaluation of Neutron Resonance Parameters Development and Validation of Temperature Dependent Neutron Scattering Laws ORNL Tools for Cross Section Measurements and Nuclear Data Evaluation in Support of Gen IV Measurements BNEEL Perspective on Nuclear Data Issues for Gen-IV INEEL/ANL Collaborative Program for Advanced Nuclear Data Measurements at ANL/IPNS LANSCE Facility and Capabilities for Nuclear Data Measurements Differential and Quasi-Integral Measurements at the Gaerttner LINAC

This paper summarizes the results of a workshop on nuclear data needs held under the U.S. Generation IV program in 2003. The workshop, held at the U.S. National Nuclear Data Center (NNDC), Brookhaven National Laboratory, was attended by experts from the U.S. national laboratories and universities. [4] The remainder of this paper is divided into four sections, consistently with the organization of the workshop (see Table 1). These sections provide a summary of the conclusions and recommendations that emerged for the following areas: Nuclear Data Needs for Generation IV Nuclear Energy Systems (Section 2); Nuclear Data Processing, Applications, and Validation (Section 3); A working paper on Nuclear Data for Gen IV was submitted by A. J. Koning et al. of the Nuclear Research and Consultancy Group, Petten, The Netherlands. Additionally, a presentation on "Nuclear Data for AFC in LANL/T-16: Brief Overview for Gen-IV Workshop," was submitted by M. Chadwick.

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Theory and Evaluation (Section 4); and Nuclear Data Measurements (Section 5). Perspectives on the path forward for the nuclear data activities in support of the Gen IV systems are summarized in Section 6. 2. Nuclear Data Needs 2.1. Workshop Conclusions Predictions of the performance of advanced nuclear systems are limited in their precision by the uncertainties in nuclear data for a variety of important nuclides relevant to reactor design. These uncertainties have bearing on the design margins required to ensure system safety and on the performance and economic competitiveness of the systems. Significant uncertainties exist in the data for minor actinides, as well as for some of the more common fissionable materials (in certain energy ranges) and non-fuel materials such as bismuth and lead in particular. Some of these uncertainties can be addressed with integral benchmarks, but additional direct cross section evaluations (and possibly measurements) are also needed. There is also a more general need for nuclear data covariances and for methods to process and utilize them in sensitivity and uncertainty analyses. Reduction of calculational uncertainty is also of interest in the U.S. Advanced Fuel Cycle (AFC) Program, and efforts in that program are being used to leverage the Generation IV program effort. Existing nuclear data are likely sufficient for preconceptual design of the VHTR because it is a thermal spectrum system and the enriched uranium fuel form is being proposed in current designs. Additional data that may be required (e.g., for high burnup and high enrichment designs) need to be identified through a systematic approach. Such an approach would consist of developing a reference configuration, defining target accuracies of relevant core and fuelcycle parameters, evaluating the impact of nuclear data uncertainties on these parameters, and defining necessary re-evaluations or measurements for data contributing a large fraction of the total uncertainty. Fast spectrum systems (GFR, LFR, SFR), which contain significant fractions of transuranics (particularly minor actinides) in their recycled fuel, require some additional data re-evaluations and possibly differential cross-section or integral measurements. For the GFR, such data evaluations might be necessary for the new fuel matrix and reflector materials that have been proposed. These efforts should be directed to obtaining nuclear data that yield better accuracy in the prediction of core and fuel-cycle related parameters (e.g., transmutation rate, criticality state, power distributions, decay heat, radiation doses, and neutroninduced damage, etc). Co variance data require particular attention since this

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information is largely unavailable in current evaluated nuclear data files. Additional covariance data are required in ENDF/B evaluations. In particular, only 46 out of 300-plus incident neutron evaluations in ENDF/B-VI (release 7) have uncertainty data. Cross section uncertainty data are available for only a few actinides (Am-241, Pu-240, Pu-241, Pu-242, Th-232, U-235, and U-238). The absence of uncertainty data for hydrogen should be remedied. Additionally, resonance-parameter uncertainty data are limited to the resolved energy region. The covariance data for secondary distributions are not addressed and require additional attention. Even where available, covariance data are not in a form readily applied in routine design calculations. Additional effort is therefore required to make this type of data available to core designers in the form useful for routine use. 2.2. Workshop Recommendations Specific nuclear data needs for the Generation IV systems should be further specified using the following systematic process: Preliminary reactor-core reference configurations should be developed and major fuel-cycle operation parameters should be preliminarily defined for each system. In addition to enabling prediction of typical system performance data, this effort should also produce an inventory of the materials in the entire fuel cycle (including those ultimately stored in a geologic repository), to facilitate the assessment of the nuclear data needs for entire systems. Viability phase target accuracies should be defined for core and fuel cycle design parameters based on key attributes of the reference configurations. More stringent target accuracies might be required in later phases of the system development activities, and can be defined later. A strong emphasis should be placed on sensitivity analysis tools that use covariance data and their application to determine the nuclear data uncertainties having greatest impact on performance, safety, and fuel cycle predictions. Uncertainties resulting from basic data (including those for plutonium, minor actinides and unconventional core materials) should be quantified for the selected reference configurations. Sources of the uncertainties should be identified and activities required to improve nuclear data should be defined. Additional experiments should be defined to address significant deficiencies that may exist in the available experimental database; desired accuracies for system performance parameters should be taken into account and additional effort should be made to specify these target accuracies. Nuclear data needs should be defined in more detail, as information from the concept point-design studies becomes available to identify more

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systematically the specific nuclides and energy ranges for which new differential measurements and/or independent confirmatory measurements are needed. Attention should be given to the potential need for additional integral experiments. This would be a natural complement to the integral data documentation efforts recommended above. Consideration should be given to creating a nuclear data advisory group to ensure that Generation IV resources are used effectively to address the data needs of the program. The formation of the group could be an outgrowth of future planned workshops. 3. Nuclear Data Processing, Applications, and Validation 3.1. Workshop Conclusions The design differences between the current generation reactors and the advanced nuclear systems dictate the need to quantify uncertainties in basic data, particularly for the minor actinides and unconventional core materials. The worldwide database of integral data represents a valuable resource for improving calculated design parameters, and as such, a high priority should be placed on identifying previous integral experiment measurements of greatest relevance to Generation IV systems and on documenting/preserving their specifications and measured results. Additionally, improved covariance data are needed for use in the methodology of data adjustments to reduce design uncertainties. It might be possible to use covariance data present in other data files when such data is absent in the nuclear data library being used for analysis. For example, covariance data are available in the Japanese nuclear data file, JENDL-3.3, and these data may be applicable with cross sections in other evaluated nuclear data files (e.g. ENDF/B-VI and -VII). The consistency checks that are performed by the tools generating the covariance data should ensure, in principle, the accuracy (pedigree) of the data. New actinide evaluations have been performed at LANL for uranium isotopes 232 to 241 (with the exception of U-236 and U-240), Pu-239, and Np237. The methods employed include new fits to experimental data, new theoretical calculations, and use of integral experiments to guide choices of data used in the evaluations. This activity has resulted in generally improved prediction of the criticality state for both fast and thermal systems. The predicted flux is a little soft for fast U-235 systems, and additional re-evaluation with latest methods might further improve the situation. There are integral measurements that are expected to be relevant to Generation IV systems. For instance, ANL has maintained a wealth of integral

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data from the ZPR/ZPPR fast reactor critical experiments program. The available data include keff (enrichment), reaction rate ratios (breeding ratio, burnup swing), spatial reaction rates (power distribution), control rod worths and sodium void reactivity (safety), Doppler worths and flux ratios. Also of note are the ANL activities on gas-cooled fast systems. Critical experiments were performed for the gas-cooled fast-spectrum systems at the ANL Zero Power Reactor (ZPR-9) facility in the mid to late 1970. Three gas-cooled fast reactor assemblies (GCFR-I, GCFR-II, and GCFR-III) were employed in these experiments. In addition to the traditional measurements on criticality, reflector worth, reaction rates distributions, material worths, control rod worths, etc., the measurements also addressed safety problems such as steam entry and analytical problems such as the impact of neutron streaming in gas-filled channels on various integral parameters. Pu/U oxide driver fuel and U02 blankets were used in some of the experiments. These experiments can be used for verification and validation of deterministic analysis methods and the Monte Carlo tools. 3.2. Workshop Recommendations Processing: Processing codes (e.g., NJOY) should be extended to provide covariances in multigroup format as needed for direct use in reactor system analysis. Data Validation: A high priority must be placed on identifying previous integral experiment measurements of greatest relevance to Generation IV systems and on documenting/preserving their specifications and measured results in a peer-reviewed format; the Generation IV program should coordinate its efforts within the framework of international activities in these areas coordinated by OECD/NEA. 4. Nuclear Theory and Evaluation 4.1. Workshop Conclusions In the U.S., the effort devoted to the development and maintenance of basic nuclear data library [Evaluated Nuclear Data File (ENDF/B)] has been in place since 1966. The work is organized within the Cross Section Evaluation Working Group (CSEWG). This library includes evaluated nuclear cross section data for all nuclides relevant for applied nuclear technology. The CSEWG activities cover all aspects of nuclear data, including, experimental measurements, data evaluation, storage formats, processing, and validation. These activities are ongoing mainly at the national laboratories (ANL, BNL, INEEL, LANL, LLNL, and ORNL). The CSEWG activities are well coordinated with the international

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nuclear data community through interactions with the IAEA and OECD/NEA. Work is underway by CSEWG to release version VII of the nuclear data file in 2005. This version would retain essentially the same format as version VI, so that current processing codes can be used in a straightforward manner. The U.S. national laboratories are also participants in international organizations involved in activities advancing nuclear data. For example, the Working Party on International Nuclear Data Evaluation Cooperation (WPEC) of the OECD/NEA promotes the exchange of information on nuclear data evaluations, measurements, nuclear model calculations, validation, and related topics, and provides a framework for cooperative activities between the participating projects. The Working Party also assesses needs for nuclear data improvements and addresses those needs by initiating joint evaluation and/or measurement efforts. The WPEC activities involve the evaluation projects in Japan (JENDL), United States (ENDF), Western Europe (JEF), and non-OECD countries (BROND, CENDL, and FENDL). Nuclear cross sections are typically evaluated using a combination of measured data and nuclear model calculations. There are ongoing activities in nuclear data theory and evaluation. Physics tools in this area are used for theoretical predictions needed to fill the gap that exists in experimental measurements of nuclear data. An example of such capability is the EMPIRE code system, which is an integrated package of nuclear reaction model codes, input parameter libraries, differential experimental data library and utility codes. EMPIRE is an advanced tool for nuclear reaction data evaluation for energies above 1 KeV. Other comparably robust tools are the GNASH/McGNASH code at LANL and the European THALYS capability. Of the three codes, EMPIRE is the only one already released. ORNL is also developing a cross section analysis and evaluation code (SAMMY) for simultaneous differential and integral data analysis and evaluation to reduce biases on the data. This code can produce covariance data for both the resolved resonance region (based on the Reich-Moore formulation) and the unresolved regions (based on single-level Breit-Wigner formulation). The incorporation of the covariance data produced by the SAMMY code in the ENDF data files would require changes to the ENDF format. Activities in data testing, including benchmark calculations are also ongoing at ORNL. 4.2. Workshop Recommendations Effort should be devoted to the development of tools for generating covariance data for the evaluated nuclear data files (e.g., the planned ENDF/B-VII file) and in multigroup format of direct use to reactor system designers. These tools must

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also provide covariance information for the unresolved resonance range and for secondary distributions. The evaluated data file formats for the covariance data should be modified to accommodate the additional data. Because of similarities of modeling methods and commonality of much of the basic data, some economies in this effort may be realized by combining covariance data from one evaluation with cross sections from another evaluation, with some suitable renormalization. The overall effort must encompass validation of the covariance data. New evaluations should be performed as specified above in the section on Data Needs. These new evaluations should be incorporated into evaluated nuclear data libraries. 5. Nuclear Data Measurements 5.1. Workshop Conclusions Facilities for nuclear data measurements include the ORELA facility at ORNL. The facility uses a powerful electron accelerator-based neutron source to provide high neutron flux across the energy range of interest to nuclear reactor analysis (0.001 eV to 100 MeV), with very high resolution for cross-section measurement. Nuclear data activities, including cross section measurements, are also conducted at the LANL LANSCE facility in collaboration with other U.S. national laboratories (LLNL and INEEL), universities, and international collaborators. For higher energy neutrons (-0.1 < En < ~400 MeV), measurement capabilities include the GEANIE spectrometer (gamma and X-ray yield measurements), FIGARO (inelastic neutron scattering and fission neutron and gamma measurements), and a neutron-induced charged particle production measurement facility. For lower energy neutrons, (thermal < En < -100 keV), the DANCE spectrometer is used for total neutron capture cross section measurements, and a Lead Slowing-Down Spectrometer (LSDS) is under construction. Together, the LANSCE neutron sources cover all neutron energies from ultra-cold to -800 MeV. Sample sizes required for these facilities are quite small. For GEANIE and FIGARO, samples weighing on the order of grams are required, while milligram and nanogram samples are required for DANCE and the LSDS facilities, respectively. Both differential and quasi-integral measurements are performed at the Gaerttner LINAC facility in Rensselaer Polytechnic Institute (RPI). This facility is a major research laboratory used for basic and applied research, including neutron cross-section measurements. Nuclear data activities at the facility

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include transmission measurements (0.001 eV to several keV), capture and scattering measurements (0.005 eV - 0.50 keV), self-indication measurements (0.005 eV - 0.5 keV), and alpha (0.005 eV - 0.5 keV) and fission (0.1 eV - 100 keV) measurements in the lead slowing down spectrometer. 5.2. Workshop Recommendations A strong emphasis should be placed on maintenance of the required experimental capabilities and on development of an integrated collaborative effort, coordinated with relevant international activities that will provide the necessary measurements. This should be done in a manner that makes best use of what will almost certainly be limited financial resources. The available experimental facilities, equipment, accelerator targets, and personnel capabilities, which are required to support necessary differential nuclear data measurement activities, are believed to be sufficient to address the anticipated need for new data. A mechanism should be established to facilitate the acquisition, maintenance, storage, distribution, and community usage of sample targets especially in the case of purified stable isotopes and actinides of interest. Consideration should be given to the application of accelerator mass spectroscopy to the measurement of integral data for minor actinides. These data would augment those obtained from integral measurements (critical experiments or irradiated fuel characterization). 6. Conclusions and Suggested Path Forward A principal conclusion from workshop is that current nuclear data are generally adequate for the initial phases of the development of Generation IV systems. The evaluated nuclear data files provide a reasonably comprehensive source of basic data for use in Generation IV system analysis. Existing evaluations of nuclear data, such as ENDF/B-VI, are probably adequate for early preconceptual design development of the VHTR. However, high burnup operation of the VHTR would require re-evaluation of some transuranics data (cross sections, decay data, and fission yields) that have not typically been important in thermal reactor design analysis. Some new differential measurements may also ultimately be needed for selected nuclides in the case of the VHTR, depending on the detailed spectral conditions and specific fuel cycle strategy selected for the system. The fast spectrum systems (GFR, LFR, and SFR) to be deployed for actinide management within a closed fuel cycle, are expected to require additional data evaluation for transuranics, particularly minor actinides, as well as integral measurements for validation of the basic data and their processing tools. Non-conventional

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structural or fuel-matrix materials may also necessitate new evaluations or measurements of basic data. A systematic approach based on sensitivity and uncertainty analysis was recommended for further specifying data needs. Implementation of this approach requires estimates of covariance data that are largely missing from current evaluated nuclear data files (e.g., ENDF/B). This covariance information should be developed expeditiously and made available to analysts for the purpose of estimating key uncertainties. Several organizations are engaged in the efforts that meet nuclear data needs identified during the workshop. These efforts include nuclear data sensitivity and uncertainty studies for the Generation IV systems, development of an application oriented covariance matrix for uncertainty analysis, and collaboration in planned integral experiments (e.g., GFR physics experiments). Acknowledgments The authors are greatly indebted to the experts from U.S. national laboratories and universities who participated in the nuclear data workshop that formed the basis for this document. Special thanks are due to the U.S. National Nuclear Data Center at the Brookhaven National Laboratory for hosting the workshop. The work described in this document was supported by the U.S. Department of Energy. References 1. "Minutes of Workshop on Reactor Physics and Analysis Capabilities for Generation IV Nuclear Energy Systems, Argonne National Laboratory, Argonne, February 18-19, 2003." Issued on March 11, 2003. 2. "Minutes of Workshop on Thermal-Hydraulic and Safety Analysis Tools for Generation IV Nuclear Energy Systems, Idaho National Engineering and Environmental Laboratory, Idaho Falls, March 18-19, 2003." Issued on April 23, 2003. 3. "Minutes of International Workshop on Reactor Physics Advances for Design and Analysis of Generation IV Nuclear Energy Systems, Hyatt Regency Hotel, Chicago, IL, April 30, 2004." Issued on June 17, 2004. (http://www.physor2004.anl.gov/workshops.htm). 4. "Minutes of Workshop on Nuclear Data Needs for Generation IV Nuclear Energy Systems, Brookhaven National Laboratory, Upton, New York, April 24-25, 2003." Issued on May 14, 2003.

NUCLEAR DATA NEEDS FOR THE ASSESSMENT OF GEN IV SYSTEMS GERALD RIMPAULT Centre d'Etudes Nucleaires de Cadarache Commissariat a VEnergie Atomique (CEA) 13108 Saint-Paul-lez-Durance CEDEX, France Four of the six nuclear systems identified by the Gen-IV international forum are relying in fast reactors. The high performance required from these future FR's calls for very innovative core characteristics compared with conventional fast reactor designs, which in turns give rise to new challenges for the available neutronics methods and data. The ERANOS "formulaire" was developed for reliable, precise and efficient calculations of sodium-cooled fast neutron reactor cores. Such a "formulaire" enables the prediction of all the neutronic quantities of interest for reactor design, operation and safety studies, along with their corresponding uncertainties. Given its achievement in terms of accuracy for existing sodium cooled reactors, the way the ERANOS "formulaire" has been designed should be looked as an example for developing similar tools for GENIV fast cores. The methodology for defining nuclear data needs is briefly covered. Target accuracies for GEN-IV neutronic characteristics is an important point to start with. The covariance data is of significant importance for quantifying the needs and evaluators should provide them with their nuclear data. Hence, nuclear data requests should be associated to uncertainty values. Integral experiments have a complementary role to differential measurements for meeting some nuclear data needs (for instance Pu239 fission). High Priority Level Requests should include those nuclides accessible with current differential measurement technology and should include less important nuclides A list of potential requests is existing, quantifying their uncertainties remain a significant effort particularly when looking at fuel cycle quantities.

1. Introduction Four of the six nuclear systems identified by the Gen-IV international forum are relying in fast reactors. The high performance required from these future FR's calls for very innovative core characteristics compared with conventional fast reactor designs, which in turns give rise to new challenges for the available neutronics methods and data. The current paper is aiming at describing the methodology for defining nuclear data needs and the short list of high priority ones. In order to achieve such a demonstration, the following items are successively covered: • Target accuracies for GEN-IV neutronic characteristics • The covariance data

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The integral uncertainties with or without use of integral experiments Convergence between adjusted and evaluated Nuclear Data Uncertainties Requested for Nuclear Data Way to classify High and Low priority level Requests List of potential Nuclear Data Needs Conclusion and Perspectives 2. Target accuracies for GEN-IV neutronic characteristics The design of the cores and fuel cycles of the Gen IV systems relies on some neutronic characteristics. Target accuracies are requested for these neutronic characteristics and their values differ at the different stages of the design studies (1st stage: viability; 2nd stage: performance). Table 1. Target Uncertainties for Generic GFR Neutronic Characteristics Uncertainties at 1 a

System Development Phase

Parameter

Viability

Multiplication factor, keff BOL