V.S. Smirnov FSUE RDIPE, Moscow, Russia A. V. Lopatkin FSUE RDIPE, Moscow, Russia

Transcription

1 11th International Conference on Nuclear Engineering Tokyo, Japan, April 20-23, 2003 ICONE EXPERIMENTAL AND CALCULATION INVESTIGATIONS OF NEU- TRON-PHYSICAL CHARACTERISTICS OF BREST-OD-300 REACTOR V.V. Orlow FSUE RDIPE, Moscow, Russia Fon(095) Fax:(095) A.I. Filin FSU RDIPE, Moscow, Russia V.S. Khomyakov RF SRC IPPE, Obninsk, Russia I.P. Matveenko RF SRC IPPE, Obninsk, Russia SUMMARY Efforts in the field of experimental and calculation studies of BREST-OD-300 reactor physics in the whole are focused on: Creation of nuclear constant data and calculation software verified and certified by GAN RF; Justification of reliability and precision extent of calculation prediction of main neutron-physical characteristics of BREST- OD-300 reactor core; Confirmation of nuclear and radiation safety of BREST-OD-300 reactor facility and plant fuel cycle; Correction and refinement of design parameters, control and regulation systems, composition of starting and subsequent critical loadings, planning of reloading schedules. The main directions of activity development of nuclear constant data and calculation software, verification and certification, experiments at BFS stand, calculation analysis of experimental results, elaboration of international benchmark models of BREST-OD-300 reactor, estimation and analysis of prediction precision for the main neutron-physical parameters of BREST-OD-300 reactor. The report reviews the main activities on the above mentioned directions, gives a number of results obtained, and concludes on the compliance extent of required and achieved calculation precision of neutron-physical characteristics of BREST- OD-300 reactor. 1. DEVELOPMENT OF CONSTANT SYSTEM For the purpose of BREST reactor physics justification, the problem-oriented nuclear constant library ABBN/BREST is being created based on the Russian constant system BNAB-93[1] and the best modern evaluated nuclear data from ENDF/B-VI, A. V. Lopatkin FSUE RDIPE, Moscow, Russia V.S. Smirnov FSUE RDIPE, Moscow, Russia sval@entek.ru V.G. Muratov FSUE RDIPE, Moscow, Russia A.L. Kochetkov RF SRC PE, Obninsk, Russia A.M. Tsiboulia RF SRC IPPE, Obninsk, Russia JENDL-3.2 and a number of other files. Though this library is the main one, at different stages of studies it is acceptable to use directly BNAB-93, for instance, for verification, analysis of experiments, precision estimation, comparison of BREST and BN reactors base characteristics. BNAB constant preparation system allows to perform calculations using (1) engineer 3D codes in diffusion approximation, (2) precise deterministic codes, and (3) Monte-Carlo codes, with the uniform constant basis. For doing so, the main code CONSYST [2] is developed, which prepares constants taking into account resonance self-shielding effects. This code is accompanied by a number of specialized modules: PRECONS preparation of macro-constants for calculation in diffusion approximation, FORAN provision of precise calculations using international format ANISN, FORMCNP preparation of constants in group approximation using MCNP code, SUBGRAN for performing calculations in sub-group approximation, and some other. Such a representative set of codes, besides the immediate calculations of BREST-OD-300 reactor, allows to perform analysis of integral experiments and a wide range of verification studies. ABBN/BREST library is supplied with covariance matrixes of neutron data errors, allowing to estimate and to minimize on the integral experiments the constant component of error in main neutron-physical characteristics of BREST- OD-300 reactor. Works on precision estimation supported by specialized computer system IN- DEX[3]. At present composition of ABBN/BREST system is competed, the system is on the stage of verification. 2. CODES EVALUATION AND VERIFICA- TION Justification of BREST reactor neutronics assumes application of a collection of computer 1 Copyright by JCME

2 codes. Among them, it is necessary to outline the main codes used for obtaining the data included in the documentation on BREST-OD-300 nuclear and radiation safety justification. Code FACT-BR (NIKIET)[4] for fast reactor 3D diffusion calculation, codes MCNP (USA)[5] and MMKKENO (SSC RF-IPPE)[6] for precise calculations using Monte-Carlo method. These codes are being prepared for certification in GAN RF. Besides the above mentioned codes, for the purpose of verification and studies of various methodical aspects, some other codes are used, for instance, codes TRIGEX (SSC RF-IPPE)[7] and JARFR(KI)[8]. The latter codes are the main ones in Russia for physics justification of fast sodium-cooled reactors of BN type, substantial experience is accumulated in their application and verification. TRIGEX is also used for analysis of experiments carried out at BFS stand. In this case in co-operates with cell code FFCP[9] which is used to take into account heterogeneity of BFS assemblies. Of course, the above mentioned precise calculation codes are used in calculation analysis of experiments as well. Practically all the codes mentioned were modified in course of work to take into account peculiarities of BREST-OD-300 reactor. Particularly, in Monte- Carlo codes, β eff calculation blocks were refined. MCNP code was reworked to take into account more correctly the resonance self-shielding in the area of unresolved group resonances. This work pays increased attention to the group option of MCNP code which generally is not used. At present, different methodical aspects of calculation for low-efficient CPS elements. It should be noted that the concept of small reactivity resources in BREST reactor specifies a new scale of allowable precision of calculation and justification of neutron-physical characteristics of BREST-OD-300. This circumstance defines increased requirements to both codes and constants, as well as to their verification extent. For the purpose of verification, international benchmark model of BREST-OD-300 reactor is elaborated. It is expedient to carry out calculation analysis of its characteristics with involvement of maximum number of Russian and foreign specialists, and using different neutron codes and nuclear data. International scientific community in invited to take part in this benchmark. At present a benchmark model for protection codes verification is being elaborated. Validation of nuclear data and calculation codes is based on the following integral experiments: measurements of neutron withdrawal crosssections below the U 238 (n,f) reaction threshold, defining the number of fissions in core in the fast area, based on which one can conclude on the reliability of inelastic scattering process description within the energy interval MeV; series of spherical critical experiments with uranium and plutonium fuel and lead reflector of different thickness, series of experiments on ROMB stand with pancakes made of high-enriched uranium and lead, series of experiments on BFS-1 and BFS-2 stands. Among those, a special role is played by the BFS experiments carried out within the framework of specially launched extensive plan of activities on experimental study of neutron-physical characteristics of BREST-OD-300 reactor and its modeling. 3. EXPERIMENTAL STUDIES ON BFS STANDS The program of experimental studies at BFS stands serves as a basis for solution of all the above posed tasks related to neutronics of BREST-OD- 300 reactor. Three main directions in experiment statement can be noted: experiments of «benchmark» type, which serve mainly to verify nuclear data and calculation codes used in justification of BREST- OD-300; non-full-scale modeling of individual base elements of BREST-OD-300 reactor physics; full-scale model of BREST-OD-300. The main stages of the program on experimental justification of BREST-OD-300 and explanations concerning interconnection of these stages can be presented as follows: 1. Critical assembly BFS-61 a pioneer benchmark-model of fast breeder reactor with heavy coolant. First calculation analysis discovered noticeable discrepancies in calculation and experimental characteristics, including K eff, inspiring further development of experimental studies; 2. Critical assembly BFS-77 benchmarkmodel of low-enriched zone of BREST-OD-300 reactor for studying spectral peculiarities of lead-cooled reactor core; 3. Series of critical assemblies BFS-85, -87 benchmark-assemblies for testing lead and bismuth nuclear constants, as well as for testing reflecting properties of heavy-metal reflector (Pb, Pb-Bi); 4. Critical assembly BFS -64 non-full-scale sector model of core and reflector in BREST-OD-300, destined mainly for studying of power field distribution and CPS elements prototype efficiency; 5. Critical assembly BFS-95 non-full-scale sector model of core in BREST-OD-300, 2 Copyright by JCME

3 destined for studying the influence of plutonium isotopic composition on the main neutron-physical characteristics of BREST-OD- 300; 6. Non-full-scale model (planned) of core with nitride fuel for studying peculiarities of forming neutron spectra and respective spectral characteristics; 7. Full-scale model of BREST-OD-300 (planned), reproducing all the main elements of BREST-OD-300 reactor physics. By now, measurements performed on the assemblies BFS-61, -77, -85, -87, -64. Some details of these experiments follow. BFS-61. Series consisted of three critical configurations: BFS-61-0 core height ~ 85 cm and 90 cm in diameter is surrounded with reflector made of lead, steel, and depleted uranium dioxide; BFS-61-1 core similar to BFS-61, in radial reflector steel is substitutes by depleted uranium dioxide; BFS-61-2 core diameter ~100 cm, radial blanket - thickness 45 cm, composed of depleted uranium dioxide. The program of studies, besides criticality measurements, included measurements of spectral indices, reaction velocity distributions, absorber efficiency, and β eff. BFS-77 had a central insert ( ~ 85 cm), being close on its properties to the core of BREST- OD-300 (coolant lead, BR close to 1); experimental program was aimed mainly at measurement of central functionals (spectral indices, capture rate in uranium with respect to fission rate in plutonium, central reactivity coefficients, void reactivity effect, etc.). BFS-77 layout is presented on Figure 2. Critical assembly BFS-64 was used to model sector of BREST-300 reactor with radial reflector Pb-Bi (central sector ~ ); core diameter > 2 m, height ~ 1 m. Figure 2 BFS-77 layout. The program of experiments included measurements of efficiency of CPS elements prototypes made of different materials, spatial distribution of fission reaction rates, radial dependence of void reactivity effect. BFS-64 layout is presented on Figure 3. Figure 3 BFS-64 layout Figure 1 BFS-61 layout BFS-85 (uranium oxide core with enrichment ~ 60%, two options of radial reflector and central insert ( ~ 25 cm Pb, then Pb-Bi); the program of experiments included measurements of ration of fission reactions rates by fission chambers 3 Copyright by JCME

4 with U-238, U-235, Pu-239. Layout of critical assembly of BFS-85 series is presented on Fiure 4. BFS-87 (mixed U-Pu fuelled core with Pu content in fuel ~ 30%, two options of reflector Pb and Pb-Bi); the program of experiments included measurements of central spectral indices, fission reaction rates distribution, and β eff. BFS-87 series critical assembly layout is presented on Figure 5. Figure 4 BFS-85 layout Figure 5 BFS-87 layout 4. MAIN RESULTS OF INVESTIGATIONS 4.1. Analysis of criticality data. results of multiplication coefficient value for critical assemblies are presented in Tables 1 and 2. Table 1 demonstrates comparison of the results obtained with international benchmark results from the Handbook [10]. The experiments with designated specification are critical experiments in spherical geometry with bsv@entek.ru lead reflectors. Table 2 analyses criticality of BFS assemblies. In order to understand the possible scale of criticality calculation errors related to neutron data, the table incorporates calculation results using constants for lead taken from the one of the most modern estimates JENDL-3.2. It should be noted, that cross-sections of Pb in the standard version of BNAB library used in MMKKENO code calculations were obtained using JENDL-3 estimate. As one can see from the results obtained, the estimate JENDL-3.2 gives rather larger difference with the experiment. One more thing to be mentioned is that for the assemblies containing lead only in the core (BFS-61, BFS-77) the latest estimate produces lower values of K eff as compared to BNAB-93, and for the experiments with lead reflectors higher ones (see Table 1). In the whole, difference only due to constants of Pb can reach 0.5% for BFS assemblies, and for the small systems with high neutron leakage even more (up to 1%). Table 1 Spherical assembly criticality calculations Assembly Fuel Reflector, Cm MMK- KENO BNAB-93 MMK- KENO JENDL3.2 HMF- U α U PMF- Pu a Pu Table 2 BFS assemblies criticality calculations. BFS assembly MCNP-4B ММКKENO ENDF/B-VI ABBN-93 ABBN-93+ JENDL-3.2 for Pb (5) (5) (5) (5) (5) (5) (13) (6) (5) 77-1а (13) Remark: in brackets the statistical calculation error is given Analysis of spectral characteristics. results of the main spectral indices (F8/F5, F9/F5, C8/F5) are in a good consistency with experimental data, testifying in general the correctness of neutron spectrum calculation. Result 4 Copyright by JCME

5 comparison is presented in table 3. When analyzing spectral indices, account of heterogeneity of BFS assemblies is of significant importance. Table 4 demonstrates calculation and experimental results of С8/F5 spectral index distribution over the core cell of BFS-64 assembly. From the viewpoint of correctness of fuel reproduction prediction, the precision of this very index plays the key role. In the majority of cases, the divergence fall below the level of 3%, however in some cases differences reach 6% and even 8%. For some actinides ( 238 Pu, 240 Pu and 241 Am), the divergences can reach up to 15%, reflecting, evidently, contemporary precision of knowledge of fission cross-sections values for these nuclides in the area typical for fast reactors with heavy coolant. Table 3 Ratio of calculation values to experimental data for spectral indices on BFS-77 assembly. Index BFS-61 BFS-77 BFS-77 Diffusion calculation TRIGEX Diffusion calculation TRIGEX MCNP-4B C238/F ± ± ±0.050 F238/F ± ± ±0.030 F239/F ± 0.996± ±0.014 F240/F ± ± ±0.033 FNp 237 /F ± ±0.042 Fpu 238 /F ± ±0.022 FAm 241 /F ± ±0.039 FAm 243 /F ± ±0.080 Table 4 Difference between calculation and experimental data (%) for C8/F5 index on BFS-64 assembly. Measurement location Diffusion calculation TRIGEX MCNP-4B Fuel cell, p.1 +3±3 +6±4 Fuel cell, p.2 +5±3 +2±4 Fuel cell, p.3 +8±3 +2±4 Fuel cell, p.4-1±3 0±4 Fuel cell, p.5-1±3 0±4 Empty tube +5±3 +5± Efficiency of fsa and cps rods. The CPS of BREST-OD-300 reactor possesses some peculiarities which make it principally different from the analogous systems in other types of reactors. In particular, one of the conceptual provisions is low physical weight of reactivity compensating system, ensuring the impossibility of occurrence of positive reactor reactivity exceeding the value of delayed neutrons effective fraction. Within the CPS, several types of elements are envisaged which differ in composition, location, and respectively, efficiency. Especially difficult to calculate the CPS rods located in radial reflector. Diffusion mesh codes traditionally are not suitable for calculations of such type. Problems also arise in case of direct calculations using Monte-Carlo method, because of impossibility to reach the high statistical precision. The only reliable way to bring to light the errors of rods efficiency calculation is the comparison of calculation results with the experimental data. Table 5 shows comparison of experimental and calculation values of efficiency for fuel rods, reflector rods, and absorber rods. Table 6 demonstrates analogous comparison for several prototypes of BREST-OD-300 CPS elements, efficiency of which was measured ob BFS-64 assembly. In general, both relative conformity and distinct discrepancies (up to 30% and more) between calculation and experimental results can be observed. Sufficiently good (better than statistical precision) conformity of MCNP calculations with experimental data is a very pleasant fact, but it does not resolve all the questions. Table 5 Rods efficiency (β eff ), BFS-77 Rod type Experiment Diffusion calculation TRIGEX Test core BREST-300 Test core MCNP ± ±0.05 BREST-300 Driver ±0.05 Driver ±0.05 Absorber B 4 C ± rods with B 4 C ± rods with B 4 C ±0.05 Reflector Reflector Copyright by JCME

6 Table 6 Efficiencies of CPS elements prototypes (β eff ), BFS CPS element prototype Single ПАЗ (with B 4 C) Group of ПАЗ (with B 4 C) Group of ПАЗ (W) Single АЗ (B 4 C) Single АЗ( B 4 C) Single АР (Al 2 O 3 ) Single ПР (Al 2 O 3 +C) Channel ПОС Channel ПОС Group of 2 channels ПОС Location Core 0.321± (150/138) Core 1.164± Core 0.209± Reflector 0.062± (157/121) Reflector 0.058± (143/121) Reflector 0.028± (157/121) Reflector 0.027± (157/121) Reflector 0.030± (143/121) Reflector 0.030± (157/121) Reflector 0.058± (157/ /121) Experiment Diffusion calculation TRIG EX MCNP ± ± ± ± ± ± ± u238-exp 0.4 u238-calc Figure 6 Radial distribution of U-238 fission rate in BFS Pu239-exp 0.4 pu239-calc Figure 7 Radial distribution of Pu-239 fission rate in BFS Analysis of reactions rate distributions. The experimental program of studies on BFS assembly series envisaged a large series of measurements of fission reaction rate distributions on the isotopes Pu-239, U-238 and U-235. Some of the calculation results obtained and comparison with the measured values are presented below on the Figures 6-9. In the whole, the comparison demonstrates rather good conformity of the experiment and the calculation within the limits of experimental errors. However, in some cases larger differences are observed: 10% - for Pu-239 fission rate distribution, and more than 10% - for U-238 in BFS-64 assembly. It should be noted, however, that the mentioned assembly had the most complicated configuration. F9 1,0 0,8 0,6 0,4 0,2 Exp. Calc R, мм Figure 8 Radial distribution of Pu-239 fission rate in BFS Copyright by JCME

7 F8 1 0,1 0,01 Exp. Calc. will be reached. However, it cannot be also stated unambiguously for such parameters like efficiency of CPS elements, reproduction coefficients. Perhaps, the planned precisions will be reached only at the stage of BREST-OD-300 reactor first criticality and in course of its trial operation. 1E R, мм Figure 9 Radial distribution of U-238 fission rate in BFS ESTIMATION OF UNCERTAINTY OF THE MAIN NEUTRON-PHYSICAL CHAR- ACTERISTICS OF BREST-OD-300 The integral numerical value which characterizes the progress in understanding the physics of BREST-OD-300 reactor is the estimate of prediction precision of its main neutron-physical parameters. table 7 shows the estimates with binding to some key dates and in comparison with assumed requirements. The year of 1990 is the starting point of the analysis, as this date characterizes almost absolute absence of macroscopic experiments. Concurrent state (2002) is characterized by the presence of several good benchmarks and the experiment on the non-full-scale model of this reactor. As one can see from the table, on many key parameters a noticeable progress and convergence of required and reached precisions is observed. In 2004 it is planned to complete at least the first stage of experiments on the full-scale model of BREST- OD-300 reactor at BFS stand. The estimates show that with the modern errors of experimental methods, the required precision on the parameters like criticality, reactivity distribution, reactivity effects, REFERENCES 1. G.N. Mantourov, M.N. Nikolaev, A.M. Tsiboulia. Group constant system BNAB-93. Part 1: Nuclear constants for calculation of neutron and photon emission fields. Issues of atomic science and technology, Series: Nuclear constants. Issue 1, M.,1996, p G.N. Mantourov. Annotation of CON- SYST code. In coll.: Issues of atomic science and technology, Series: Nuclear constants. Issue 1, M., 2000, p G.N. Mantourov. System of codes and archives INDEX. In coll.: Issues of atomic science and technology, Series: Nuclear constants. Issue 5(59), M., 1984, p. 20. Table 7 Correlation of required and reached uncertainties of calculation of the main neutronphysical characteristics of BREST-OD-300 reactor. Parameter Required Reached K eff 0.5% 2.5% 1.0% + Reproduction ? coefficient Efficiency of 5% 30% 30%? CPS elements Power distribution 2% 5% 3% + Void reactivity 0.2% 1% 0.4% + effect Doppler-effect 10% 20% 15% + 4. S.V.Barinov, A.V.Radkevitch Application of CONSYST/ABBN polygroup neutron data preparation system in program complex FACT-BR for three-dimensional neutron-physical calculations. Collection Algorithms and programs for neutron-physical calculations of nuclear reactors,1999 г., p MCNP-4B, Manual, LA-12625M, A.A. Blyskavka, G.N. Mantourov, M.N. Nikolaev, A.M. Tsiboulia. Program complex CON- SYST//MMKKENO for calculation of nuclear reactors by Monte-Carlo method in multi-group approximation with scattering indicatrisses in Р n approximation. Preprint IPPE 2887, 28 p., Obninsk, Copyright by JCME

8 7. A. S. Seregin. Annotation of TRIGEX program for small-group calculation of reactor in tree-dimensional hexagonal geometry. - Issues of atomic science and technology, Series: Physics and technology of nuclear reactors. Issue 4(33), M., 1983, p Yaroslavtseva L.N. JARFR complex for calculation of neutron-physical characteristics of nuclear reactors. - Issues of atomic science and technology, Series: Physics and technology of nuclear reactors. 1983, Issue 8(37), p A.A. Bezborodov, B.G. Ryazanov, M.M. Savos kin. of heterogeneity effects in critical subassemblies of fast reactors by ВПС method. In coll.: Issues of atomic science and technology. Series: Physics and technology of nuclear reactors. Issue 2, M., 1986, p International Handbook of Evaluated Criticality Safety Benchmark Experiments. - NEA/NSC/DOC(95)03. NEA OECD, Paris, Copyright by JCME