1. 1st Meeting of WPEC Subgroup 26 (May 3, 2006) Proposed Agenda 9:00-9:30 Objectives of the meeting, methodology and relation with HPRL (M.Salvatores, G.Palmiotti, A.Plompen) Objective: Confirm chosen methodology and define relation with HPRL 9:30-10:30 Design Target Accuracies (15’ presentations + discussion) -ANL data (G.Palmiotti) -CEA data (G.Rimpault) -Comments on data provided by M.Ishikawa for JAEA (T.Kawano, M.Salvatores) Objective: Agree on a preliminary list of reference designs and parameters and related design target accuracies. 10:30-12:15 Covariance data: Status and perspectives for Subgroup objectives (15’ presentations + discussion) -LANL Data (P.Talou) -Data at ECN-Petten (A.Koning) -Data at NEA DataBank (E.Sartori) -Comments on JENDL data (T.Kawano) -BNL data (D.Rochman) Objective: Define strategy and time schedule for uncertainty analysis according to data availability 12:15-12:45 Sensitivity studies and tools: needs for Subgroup objectives (All) Objective: Agree on a strategy to evaluate sensitivity coefficients 12:45-13:00 Conclusions and next steps (All)

2. WPEC subgroup on „Nuclear Data Needs for Advanced Reactor Systems” Scope A systematic approach to define data needs for Gen-IV and, in general, for advanced reactor systems is needed in order to harmonize requests coming from different communities, to establish priorities and credible quantitative goals and timeframes, to define the respective and complementary roles of new data evaluations and of differential and integral experiments. A strong interaction and synergy among reactor designers, reactor physicists and nuclear physicists is a necessary prerequisite of this activity. Objectives The objectives of the subgroup are -Compilation of an agreed set target accuracies on relevant design parameters for the Gen-IV concepts. Required target accuracies should be justified in terms of impact on different phases of a specific design (feasibility, preconceptual and conceptual design etc.) -Definition of a set of data uncertainties and covariance data. These data should be as complete as possible. At this stage, it is not expected to have a “final” set, in particular of covariance data, but an agreed “first iteration” set. -Production of a set of quantitative data needs by isotope, reaction type, energy range. -Proposal for an approach to meet the needs and relative timeframe

3. Method of work It is proposed to split the work in two phases: -First phase (September 2005-September 2006): Consensus on the proposed methodology Definition of design target accuracies Definition of a “first iteration” set of data uncertainty values Evaluation of data needs -Second phase (September 2006-May 2007): Indication of differential and integral experiments needed to meet the needs: their respective role, use of existing experiments, definition of selected new experiments, experimental accuracies required, facility availability. Two reports will be produced, one at the end of each phase.

4. The approach 1- Sensitivity analysis is performed, via GPT (Generalized Perturbation Theory), on performance parameters (core, fuel cycle) of representative models of the systems of interest. 2- Uncertainty (e.g. nuclear data covariance) propagation and assessment 3- Once the sensitivity coefficient matrix S and the covariance matrix D are available, the uncertainty on any integral parameter can be evaluated: Impact on design and target accuracy requirements can then be specified.

5. Target accuracy requirements To establish priorities and target accuracies on data uncertainty reduction, a formal approach can be adopted: define target accuracy on design parameter and find out required accuracy on data (the “inverse” problem). The unknown uncertainty data requirements d i can be obtained solving the following minimization problem : i = 1... I with the following constraints n = 1... N where S ni are the sensitivity coefficients for the integral parameter Q n, and are the target accuracies on the N integral parameters. i are “cost” parameters related to each  i and should give a relative figure of merit of the difficulty of improving that parameter (e.g., reducing uncertainties with an appropriate experiment).

7. Target Accuracy of FBR Core Design M.Ishikawa (JAEA) Criticality Target → ±0.3%Δk (1σ) Traditional design error 0.5-1.0%Δk → These error values correspond to the number of peripheral fuel S/As of 10 – 20. This results the costly design due to heavy control rod system, or change of Pu enrichment, etc. Power distribution Target → ±3% (2σ) Traditional design error 5% → This forces to set allowance of 20 W/cm for the maximum linear power rate, which severely affects the design criteria of non-melting fuel. This results too much safety guard system, or too low fuel linear power rate, that is, too large core sizes, or too many fuel pin numbers. Doppler Reactivity : Target → ±14% (2σ) Traditional design error 20-30% → Since it is most fast and effective negative feedback in the accident condition, the error value directly affects the requirement of response time of the detector and control system. Sodium Void Reactivity Target→ ±20% (2σ) Traditional design error 40-50% → The ULOF evaluation of Monju was OK, but the large FBR core expects more severe results. ※ Ref: FBR R&D Committee under STA, Japan （ Core, Fuel ）（ April, 1996 ）

8. Target accuracies assumed for integral parameters G.Aliberti, G.Palmiotti, M.Salvatores K eff Power Peak Temperature React. Coeff. Void React. Coeff. Burnup Δρ Transmutation Target Accuracy ±0.5%±3%±10% 300 pcm (fast reactors) 500 pcm (thermal reactors) ±5%

9. Target accuracies for GEN-IV neutronics characteristics G.Rimpault (CEA, Cadarache) The design of the cores and fuel cycles of the Gen IV systems relies on some neutronic characteristics. Target accuracies are requested at the different stages of the design studies (1 st stage: viability; 2 nd stage: performance).

10. Viability Performance Keff<0.5%(T); <0.7%(R)<0.2%(T); <0.3%(R); 0.3% (I) 0.5%? 0.3%? Peak Power3% (T); <5% (R) 1% (T); <3% (R) 0.3%? 2%? Power Distribution 7% (T) 3% (T); 3% (I, 2  ) 5%? 2%? CR worth(individual) 10%(T); <16%(R) 5%(T); <10%(R) 10%? 5%? BU reactivity?? ??  keff0.7%  keff 0.3%  keff Reactivity coeff.10-20%(T); <16%(R)5-10%(T); <10%(R); 14-20%(I, 2  ) 10%? 7%? Kinetics param.5%(T); <13%(R) 2%(T); <7%(R) 5%? 3%? T: Taiwo (ANL); R: Rimpault (CEA); I: Ishikawa (JAEA) PROPOSED DESIGN TARGET ACCURACIES (1  ) FOR ALL Gen-IV FRs

11. Systems Investigated Fast systems: GFR LFR SFR EFR Thermal systems: VHTR Extended BU PWR For these systems, sensitivity profiles are available in a 15 energy group structure, for large set of integral parameters.

12. GFR The gas cooled fast reactor contains CERCER fuel which is a mixture (56%-44%) of a ceramic matrix material SiC and a ceramic heavy metal carbide fuel with 5% of Minor Actinides (MA). The materials of the core region are structure (20%), coolant (40%) and fuel (40%) and the average enrichment (PUC/(UC+PuC)) is 17%. The coolant is helium and the reflector is a mixture of Zr 3 Si 2 and coolant (60%- 40% for the axial reflector and 80%-20% for the radial reflector)

13. LFR The lead cooled fast rector, that is being also investigated in the frame of a benchmark problem prepared by KAERI and also adopted by IAEA, is a 900 MWth reactor loaded with U-TRU-Zr metallic alloy fuels (2% of MA). The core contains 192 hexagonal ductless fuel assemblies and it is surrounded by ducted lead reflector and steel shields.

14. SFR The small size transmuter sodium cooled fast reactor is an 840 MWth reactor loaded with U-TRU-Zr metallic alloy (10% of MA) and very low conversion ratio (<0.25).

15. EFR The large size sodium cooled reactor, whose specifications have been provided by the CEA, is a 3600 MWth reactor loaded with U-TRU oxide fuel (1% of MA). The core is surrounded by a blanket.

16. VHTR Inner, Central and Outer Fuel IsotopeBOCEOC (90GWd) U2352.49E-051.09E-05 U2381.51E-041.41E-04 Np237-1.48E-07 Pu238-4.86E-08 Pu239-2.52E-06 Pu240-7.57E-07 Pu241-8.91E-07 Pu242-2.360E-07 Am241-2.07E-08 Am242-4.24 E-10 Am243-3.38E-08 Cm242-7.58E-09 Cm243-1.34 E-10 Cm244-8.07E-09 Cm245-4.30E-10 C6.40E-02 O2.64E-04 Si5.23E-04

17. Extended BU PWR Enrichment:8.5% Burnup: 100 GWd/Kg

18. 15 Energy Group Structure

19.  All the sensitivity calculations in this study have been performed with the ERANOS code system, which allows to calculate homogeneous and inhomogeneous solutions of the Boltzmann equation and generalized importance functions, and to perform perturbation and uncertainty analysis.  The discrete ordinate module BISTRO in ERANOS has been used to perform flux and generalized importance function calculations. An S 4 P 1 approximation in RZ geometry has been proved accurate enough for this type of calculation.  Decay heat calculations have been performed with the ORIGEN code.  The time dependent perturbation calculations are performed using the NUTS code. NUTS solves the direct and adjoint time dependent Bateman equations and computes the perturbation integrals, taking into account power plant history and reprocessing losses for any type of nuclear fuel cycle.  Cross-sections of all fast systems have been processed with the ECCO code of ERANOS using JEF3.0 nuclear data. For thermal systems, the cross-sections were generated with the WIMS code in conjunction with JEF2.2 library. Calculation Tools