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)

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

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.

The approachThe approach1- 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:

RR20 DSSR

Impact on design and target accuracy requirements can then be specified.

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 di can be obtained solving the following minimization problem :

mind/ 2ii

i i = 1 ... I

with the following constraints

Tn

2i

2ni

iQd/S n = 1 ... N

where Sni are the sensitivity coefficients for the integral parameter Qn,

and are the target accuracies on the N integral parameters. TnQ

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).

Target Accuracies for Gen IV Reactor Physics Parameters

(T.Taiwo, ANL) Systems Development Phase (1??)

Viability Performance

Multiplication factor, k-eff <0.5% <0.2%

Relative Power density Peak ~3% ~1% Distribution 7% 3%

Control rod worth Element 10% 5% Total 5% 2%

Burnup reactivity swing 3% <2% (of reactivity value) or 0.5%k or 0.5%k

Breeding gain 0.05 0.02

Reactivity coefficients Large effects 10% 5% Small effects 20% 10%

Kinetics parameters 5% 2%

Local nuclide densities Major constituents 5% 1% Minor constituents 10-20% 2-5%

Target Accuracy of FBR Core DesignM.Ishikawa (JAEA)

Criticality Target → ±0.3%Δk (1σ) Traditional design error 0.5-1.0%Δk → These error values correspondto 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/cmfor the maximum linear power rate, which severely affects the design

criteria of non-melting fuel. This results too much safety guard system, ortoo 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 effectivenegative 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 wasOK, but the large FBR core expects more severe results.

※Ref: FBR R&D Committee under STA, Japan（ Core, Fuel）（ April, 1996）

Target accuracies assumed for integral parametersG.Aliberti, G.Palmiotti, M.Salvatores

KeffPowerPeak

TemperatureReact. Coeff.

VoidReact. Coeff.

BurnupΔρ Transmutation

Target Accuracy ±0.5% ±3% ±10% ±10%

300 pcm(fast reactors)

500 pcm(thermal reactors)

±5%

Target accuracies for GEN-IV neutronics characteristicsTarget 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 (1st stage: viability; 2nd stage: performance).

Uncertaint ies at 1 System Development PhaseParameter Viability Perf ormanceMult iplicat ion f actor, keff BOL < 0.7% < 0.3%Local power density < 5% < 3%Structure Damage < 15% < 9%React ivity Swing (keff EOL) (<1.0%) (< 0.5%)Breeding Gain <+/-0.06 <+/-0.04Void Reactivity Eff ect on each component (leakage; non-leak.) < 16% < 10%Doppler React ivity Eff ect < 16% < 10%Delayed Neutron Fraction < 13% < 7%Control Rod Worth < 16% < 10%g heat ing < 16% < 10%

Viability Performance

Keff <0.5%(T); <0.7%(R) <0.2%(T); <0.3%(R); 0.3% (I)0.5%? 0.3%?

Peak Power 3% (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 ?? ??keff 0.7%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

Systems Investigated

Fast systems:

GFRLFRSFREFR

Thermal systems:

VHTRExtended BU PWR

For these systems, sensitivity profiles are available in a 15 energy group structure, for large set of integral parameters.

GFRThe 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 Zr3Si2 and coolant (60%-40% for the axial reflector and 80%-20% for the radial reflector)

LFRThe 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.

SFRThe 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).

EFRThe 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.

VHTR

Inner, Central and Outer Fuel

Isotope BOC EOC (90GWd)

U235 2.49E-05 1.09E-05

U238 1.51E-04 1.41E-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

C 6.40E-02 6.40E-02

O 2.64E-04 2.64E-04

Si 5.23E-04 5.23E-04

Compositions [1024 at/cm3]

Isotope BOC EOC

U-233 - 1.112E-12 U-234 - 1.331E-7 U-235 5.721E-4 6.398E-5 U-236 - 8.305E-5 U-237 - 1.138E-7 U-238 6.159E-3 5.759E-3 Np-237 - 9.534E-6 Np-239 - 5.805E-7 Pu-238 - 7.042E-6 Pu-239 - 5.070E-5 Pu-240 - 2.523E-5 Pu-241 - 1.706E-5 Pu-242 - 1.043E-5 Cm-242 - 3.530E-7 Cm-243 - 1.506E-8 Cm-244 - 2.240E-6 Cm-245 - 2.227E-7

O 2.744E-2 2.744E-2 H 2.794E-2 2.794E-2 Zr 4.282E-3 4.282E-3

FP (*) 2.708E-18 2.344E-3 (*) Fission Products

Extended BU PWREnrichment: 8.5%

Burnup: 100 GWd/Kg

15 Energy Group Structure

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 S4P1 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

GFR, LFR, SFR, EFR Nominal Values

Reactor Keff [pcm]

Power Peak

Temperature Reactivity Coeff.

[pcm]

Void Reactivity Coeff.

[pcm]

Burnup [pcm]

Decay Heat (s) [W]

Dose (t) [Sv]

Neutronic Source (u)

[n/s] GFR 1038.0 1.45 (a) 1549 (b) 350.1 (c) 1081.3 (d) 6.246E+3 1.028E+1 2.183E+11 LFR 22.9 1.29 (e) 228.1 (f) 6575.5 (g) -1464 (h) 4.616E+3 1.140E+1 2.275E+11 SFR 5015.4 1.53 (i) 231 (l) 1831 (m) -3981.1 (n) 1.400E+4 2.319E+1 8.537E+11 EFR 9786.5 1.63 (o) 1289 (p) 1934.5 (q) -9123.9 (r) 4.324E+3 1.235E+1 1.996E+11

(a) Center core radially and axially (b) Tfuel=300K - Tfuel=1263K (c) He loss in core and reflector (d) 415 days (s) 100 y after discharge (t) 100000 y after discharge (u) 2 y after discharge

(e) (R , Z) =(100.96 , 117.90)cm (f) Tfuel=300K - Tfuel=900K (g) Pb loss in core (h) 310 days

(i) (R , Z) =(66.59 , 143.03)cm (l) Tfuel=300K - Tfuel=850K (m) Na loss in core (n) 155 days

(o) (R , Z) =(153.24 , 125)cm (p) Tfuel=300K - Tfuel=1520K (q) Na loss in core and blanket (r) 1700 days

VHTR Nominal Values

Keff (a) Keff (b) Peak Power (c)

Peak Power (d)

Temperature (e) Reactivity Coeff.

[pcm]

Temperature (f) Reactivity Coeff.

[pcm]

Burnup (g) [pcm]

Decay Heat (h) [W]

Dose (i) [Sv]

Neutronic Source (l) [n/s]

1.37767 1.01610 1.95863 2.24966 2095.3 3416.3 -25829.4 1.670E+1 2.979E-2 2.227E+8

(a) BOC (b) EOC (c) BOC at (R , Z) =(147.62 , 556.5)cm (d) EOC at (R , Z) =(147.62 , 556.5)cm

(e) Tfuel=773K – Tfuel=1373K at BOC (f) Tfuel=773K – Tfuel=1373K at EOC (g) 845.63 days

(h) 100 y after discharge (i) 100000 y after discharge (l) 2 y after discharge

PWR Nominal Values

Keff (a) Keff (b) Temperature (c)

Reactivity Coeff. [pcm]

Temperature (d) Reactivity Coeff.

[pcm]

Burnup (e) [pcm]

Decay Heat (f)

[W]

Dose (g) [Sv]

Neutronic Source (h) [n/s]

1.498025 0.872310 695.2 1054.6 -47883.6 6.000E+2 1.128E+0 2.486E+10

(a) BOC (b) EOC (c) Tfuel=550K - Tfuel=900K at BOC (d) Tfuel=550K - Tfuel=900K at EOC

(e) 2773.5 days (f) 100 y after discharge (g) 100000 y after discharge (h) 2 y after discharge