Neutronic Issues and Ways to Resolve Them. P.A. Fomichenko National Research Center Kurchatov Institute Yu.P. Sukharev JSC Afrikantov OKBM,

Transcription

1 GT-MHR Project High-Temperature Reactor Neutronic Issues and Ways to Resolve Them P.A. Fomichenko National Research Center Kurchatov Institute Yu.P. Sukharev JSC Afrikantov OKBM,

2 GT-MHR PROJECT MISSION AND CURRENT WORK STAGE One of the most successful international projects in nuclear technologies is current cooperation between the RF and the USA in development of a gas-turbine modular helium reactor (GT-MHR). The GT-MHR mission declares that the GT-MHR is a project funded in cooperation by the US DOE and the Rosatom, aimed at the solution of one of the most important nonproliferation problems disposal of weapon-grade plutonium. The GT-MHR technology provides effective disposal of weapon-grade plutonium and could be used for development of new-generation reactors capable of effective production of electricity, hydrogen and disposal of actinides of spent fuel from light-water reactors. In 2002 the GT-MHR Preliminary Design was completed, approved by the Minatom Scientific-Technical Council and revised by the US experts assigned by the US Department of Energy. Since 2003 the R&D program aimed at demonstration of the Project key technologies has been realized for the most important concepts of the Project. 2

3 GT-MHR DISTINCTIVE FEATURES The GT-MHR design has a number of distinctive features related to its design : 1) the fuel composition based on weapons grade plutonium in combination with erbium burnable absorber; 2) fuel particles with multilayer coating dispersed in fuel compacts which are arranged in hexagonal graphite fuel blocks; 3) annular layout of reactor core with fuel blocks and reflector graphite blocks; Control rod stand pipe Upper reflector of graphite blocks Annular core of prismatic FA Inner reflector of graphite blocks Outer reflector of graphite blocks Lower reflector of graphite blocks Fuel particle Pyrocarbide PyC Silicon carbide SiC Porous carbonpyc Fuel kernel 3 Fuel compact FA I type FA II type 4) higher power density of the core (up to 6.5 MW/m 3 ), fuel composition high temperatures (maximum fuel temperature is 1250 ºC); 5) relatively small cross-sectional dimensions of annular core (inner and outer diameters are 2.9 and 4.8 m correspondently) and its axial elongation (core height amounts 8 m).

4 SCOPE AND TRENDS OF ACTIVITY The scope and trends of analytical l and experimental activities iti at studying neutronic characteristics of the GT-MHR are determined in compliance with the requirement for reliable validation applying certified software tools. Risk factor existing in analytical validation of the GT-MHR neutronic characteristics consists in that the applicable neutronic analysis codes, which have been used up to the present time, have been tested mostly within analytical and experimental studies of HTGRswith uranium fuel. Therefore, at the early stages of the GT-MHR nuclear design development it has already been determined that the experiments should be performed using critical test facilities and analytical benchmark-studies which are becoming quite relevant considering that plutonium fuel and erbium burnable poison are used. The report contains the validation program of the GT-MHR neutronic properties, the stages of its implementation and the main results obtained by now. 4

5 EXPERIMENTAL VALIDATION OF NEUTRONIC CHARACTERISTICS The analyses of data of previous Russian and foreign experiments for HTGR designs validation permits to make a conclusion that in the course of these experiments: - range of problems put within the framework of GT-MHR Project was covered incompletely; - neutron spectrum adequate with that of GT-MHR core was not simulated; - annular core, determining GT-MHR physics to a large extent, was not simulated; - effects of plutonium and erbium on the HTGR physics were not studied. At the same time, the Russian Federation has the experimental base to fulfill work on partial validation of GT-MHR neutronic characteristics. It includes, first of all, the following facility: - ASTRA critical facility at NRC KI designed for experimental investigation of neutronic characteristics of high-temperature gas-cooled thermal-neutron reactors (the purpose within the GT-MHR Project includes simulation of the layout and design decisions); - critical test facilities BPF (Big Physical Facility) at IPPE, intended for full-scale simulation of reactor cores (the purpose within the GT-MHR Project includes experimental validation of reactivity temperature e effects ec considering various plutonium u compositions o s in neutron spectrum which is close to neutron spectrum of the GT-MHR core); - multi-purpose research reactors SM-3 and RBT-6 at NIIAR (their purpose within the GT-MHR Project includes irradiation of coated fuel particles and PuO 2-x fuel compacts located in ampoule channels result is also the isotopic content of irradiated fuel). 5

6 PROBLEMS SOLVING AT CRITICAL FACILITY ASTRA U Fission Rates Distribution Experimental Data JAR Calculation Annular core simulating. Preparation and fulfillment of experiments at the ASTRA critical facility at room temperature and with heating of the critical assembly. Systematization and analysis of analyticalexperimental tests results. Obtain experimental results on: -criticality analysis - reaction rate distribution - space flux distribution -control rods worth in inner and side reflectors - temperature effect of reactivity - kinetics parameters estimation

7 STUDY OF REACTIVITY TEMPERATURE EFFECTS AT BPF Neutron Spectrum 1.E+00 compact E-01 1.E-02 1.E-03 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07 Ene r gy (e V ) Scheme of experiments on temperature coefficient 1. Sample outside the core Reflector ~ 400 mm 2. Sample in the core Active core ~ 1000 mm Cooling air inflow 48 mm Top view B A 51 mm Proving the possibility of performing reactivity temperature effects studies for the testt specimens in the central graphite insert at BPF. Creation of the facility at BPF with central graphite insert, where the GT-MHR reactor neutron spectrum is simulated. Performing the experiments on the reactivity temperature effects for the test specimens of different Pu isotopic compositions including the effects of Er poison. 7

8 BENCHMARK ANALYSES A lack of experimental data to validate neutronic characteristics ti of the core with plutonium fuel makes benchmark analysis very important. Now neutronic analyses within the GT-MHR Project use both engineering (WIMS-D, JAR- HTGR, UNK), and precision (MCNP, MCU, UNK-MC) codes. Analytical codes, procedures and models are validated and upgraded mainly on the basis of benchmark analysis results (alongside with comparison of calculation results with available experimental data). By the present methodological approaches and analytical models have been developed, which adequately reflect Project distinctive features associated with annular core, plutonium fuel in the form of coated particles and with erbium burnable poison. A considerable contribution in better understanding of the reactor physics was made by development and calculations of the stepwise benchmark analyses proposed for investigations within the framework of the GT-MHR project and, in particular, the IAEA Coordinated Research Programme IWGGCR CRP 5. The test provided sequential calculations, beginning from fuel compact cells, a fuel assembly cell and ending with a reactor 3D model. 8

9 BENCHMARK ANALYSES (stages of calculations) Stage 1. Fuel compact cells FCC R= cm Fuel block graphite R=0.625 cm Fuel compact graphite TRISO fuel particles FCC2 R= cm Fuel block graphite R=0.625 cm Homogenized mixture of fuel compact graphite and fuel kernels and their coatings FCC1 R= cm Homogenized mixture of fuel block graphite, fuel compact graphite and fuel kernels and their coatings Stage 2. Burnable poison cells BPChom R= cm Homogenized mixture of fuel, fuel block graphite and fuel compact graphite which is accounted for 1 burnable poison compact R= cm Fuel block graphite R=0.625 cm Homogenized mixture of burnable poison compact graphite and erbium oxide kernels R= cm Homogenized mixture of fuel, fuel block graphite and fuel compact graphite which is accounted for 1 burnable poison compact R= cm Fuel block graphite Erbium oxide kernels R=0.625 cm Burnable poison compact graphite BPC-het Stage 3. Fuel assembly cell 9 reflection reflection reflection reflection Heterogeneous setting of fuel kernels and their coatings Homogenization of burnable poison in a compact Detailed geometry of a fuel assembly Boundary conditions reflection at all surfaces Layer # Beginning End of of cycle cycle 1 2/3 3/ /3 3 2/3 3/3 4 1/3 2/3 5 1/3 2/3 6 1/3 2/3 7 1/3 2/3 8 2/3 3/ /3 10 2/3 3/3 0 fresh fuel 1/3 fuel irradiated during 1/3 of life-time 2/3 fuel irradiated during 2/3 of life-time 3/3 fuel irradiated during 3/3 of life-time (spent fuel) lea akage leakage leakage Detailed setting of the core arrangement Detailed setting of core components geometry Isotopic content at the discrete condition through the life time of the fuel assembly burnup Boundary conditions leakage at all outer surfaces Stage 4. Reactor 3D model lea akage

10 BENCHMARK ANALYSES (calculational models improving) 3D reactor analysis: - critical parameters - fuel depletion - power distribution - control rods efficiency - neutron fluence - temperature reactivity coefficients 10

11 VERIFICATION AND LICENSING OF NEUTRONIC ANALYSIS SOFTWARE To systemize work of formation of aset of reactor physics analysis software in the GT-MHR Project on the basis of regulatory documents requirements a generalized list of safety-related parameters, for which provided calculation errors should be specified in certificates of individual software, was developed. A draft generalized matrix of verification of calculation codes for GT-MHR neutronic characteristics was formulated. It includes, in particular, critical parameters, power distribution, control rods efficiency, reserve shutdown system efficiency, reactivity effects, and effect of moisture ingress, reactor kinetics parameters, and residual power. The proposed verification matrix type gives the possibility to use all available information obtained as of benchmark analysis of standard d problems is performed, and as new experimental data are accumulated and experimental data of previous studies are revised. 11

12 Neutronic characteristics EXEMPLE OF VERIFICATION MATRIX Calculation of test problems Verification basis Comparison with experimental data and reference codes results Critical parameters + + Space power distribution + + Space distribution of damaging neutron fluence + - Efficiency of control rods (separate and groups) + + Efficiency of reserve shutdown system + - Dependence of reactivity on temperature + + Dependence of reactivity on power + - Moisture ingress effect + + Fuel burnup + + Reactor kinetics parameters + + Residual power

13 BASIC RESULTS OF VERIFICATION OBTAINED BY NOW The investigations revealed the following factors determining sources and values of uncertainties: used nuclear data, effective models of consideration of double heterogeneity of fuel arrangement, the ratio of reaction rates of plutonium and erbium, detailing of space-energy description of neutron fields, consideration of spatial distribution of temperature and burnup over the core volume. In particular, it was shown that application of the procedures, developed specially for heterogeneous description of a fuel composition for a fuel block cell, in calculations using engineering g programs permitted to decrease the multiplication factor determination error from 5 to 0.5 %, and for isotope concentrations of plutonium and Er from 30 to 9 %, as compared with real-geometry reference calculations with detailed simulation of isotope kinetics in burnup. 13

14 BASIC RESULTS OF VERIFICATION OBTAINED BY NOW (continuation) The nuclear data effect on uncertainty of isotope composition calculations l and, as a result, on the temperature reactivity coefficient was analyzed using the complex MONTEBURNS MCNP5 ORIGEN 2.1 for the temperature range of K. It was obtained that application of different nuclear data could lead to the change of 239 Pu burnup dynamics and to the change of the sign of calculated reactivity temperature coefficient for the burnup value of 600 MW day/kg (see Table ) because of the change of the ratio between effects of plutonium and erbium isotopes. It was obtained that application of neutron constants from the files of estimated data JENDL3.3 3 is the most conservative approximation with regard to the meeting of the condition of negativity of the temperature reactivity coefficient. 14 Dependence of temperature reactivity coefficients on burnup calculated by MCNP5 code MW day/kg Burnup of 239 Pu, % Library JENDL 3.3 Library ENDF/B-5, Reactivity coefficient*, /K Library Library JENDL 3.3 ENDF/B-5,6 From to From to From to from to from to * The range of reactivity coefficient values is determined by a statistic error

15 BASIC RESULTS OF VERIFICATION OBTAINED BY NOW (continuation) The change of the ratio between effects of plutonium and erbium isotopes is shown in Fig. 1. Isotopes 239, 241, 240 Pu and 167 Er determine ratio of fissions and absorptions in the reactor system with pure plutonium fuel. Values of temperature reactivity coefficient could be expected negative at burnup lower than 600 MW day/kg, this corresponds to the ratio of the total sum 239 Pu, 241 Pu masses to the mass of 167 Er less than 25 (Fig. 2). Fig. 1: Dependence of a fraction of absorption of 239 Pu ( ), 240 Pu ( ), 241 Pu ( ), 167 Er ( ), and fission of 239 Pu ( ), 241 Pu ( ) on burnup height Fig. 2: Dependence of a ratio of total absorption ( ) and fission ( ) of 239 Pu and 241 Pu to absorption of 167 Er ( ) and their masses ( ) on burnup height. 15

16 BASIC RESULTS OF VERIFICATION OBTAINED BY NOW (continuation) The level of detail of a space-energy partition in the full-scale deterministic analysis of the reactor in the group diffusion approximation of the neutron transport equation influences considerably both to power distribution (Kr) (Fig. 3a) and efficiency of control rods (Fig. 3b). 96 points per fuel block and 51 group approximation is used here as a reference. Fig. 3 (a, b): Dependence of neutronic characteristics on calculation approximations 16

17 BASIC RESULTS OF VERIFICATION OBTAINED BY NOW (conclusion) The performed studies show that the total calculation uncertainties are: for a multiplication factor less than 1 %, for concentrations of basic isotopes of plutonium puo u and debu erbium to 10 %, for power ds distribution to 10 %, for control rods efficiency to 20 %, for temperature reactivity coefficient to 30 %. As follows from the results showed in previous figure, power peaking factor and control rods efficiency calculation uncertainties can be reduced by using detailed calculation grid. Now in design calculations 24 mesh-points per fuel block and 13 group approximation is used. Reduction of temperature reactivity coefficient calculation uncertainties is not offers possible yet. 17