Testing the SAS2H SCALE Control Module on VVER Type Fuel

JP0450131

JAERI-Conf 2003-019

Testing the SAS2H SCALE Control Module on VVER Type Fuel Yevgen BILODID, Maxim YEREMENKO, Yuriy KOVBASENKO State Scientific and Technical Centre on Nuclear and Radiation Safety (SSTC NRS), Radgospna st. 35-3 7 03142 Kyiv, Ukraine Description and comments concerning validation control module SAS2H SCALE in NPPs experiments and international VVER benchmarks are provided. While testing, the experimental data from the NPP as well as calculation benchmarks CB-2 (VVER440) and OECD/NEA (VVER-1000) were used. Here, the special attention was paid to isotopes, which are important in respect to criticality analysis of spent nuclear fuel management system. The results listed below show compliance of the received data and those from the base. For the calculation of VVER reactor ftiel isotope composition it is recommended to consider the fuel pins position in fuel assembly. KEYWORDS. Nuclear Power Plant, VVER reactor, spentfuel, burnup credit, SCALE, SAS2H, benchmark

1. Introduction In the case of the bumup credit criticality assessment as to the spent fuel management system the methodology of bumup credit criticality calculation should be checked up thoroughly, and safety margins studied. This research became possible with the availability of computer and experimental methods that permitted modeling and monitoring isotopic composition of spent nuclear fuel with high accuracy. Due to specifics of the Western computer codes in the Westem-reactor orientation, until now, insufficient information was available concerning the applicability of the SCALE code package') to miclide inventory calculations of systems with fuel of VVER reactors. Consequently, it was necessary to conduct testing using experimental or operational results. This work is continuation of the comprehensive testing complex SCALE on VVER ftiel. The first stage consists of testing Control Module of criticality calculations by means of code KENO-VI, and has been made out as NUREG/CR-6736 . The second stage provides testing isotopic calculations using SAS2H and ORIGEN-ARp 4 complexes. This article presents the results of isotopic calculations using Control Module SAS2H. One of the best-known up-to-date computer prograins for calculating isotopic composition of spent nuclear fuel is the ORIGEN code. The disadvantage of this code is that its input libraries of neutron-physical constants are applicable to the Western types of reactors - PWR, BWR, CANDU. The use of this program as a constituent of the SCALE code package permits overcoming this disadvantage and preparing the problem-oriented libraries of neutron-physical constants. SCALE Control Module SAS2H is designed to calculate changing isotopic composition of fuel as it bums up. Control Module SAS2H prepares libraries of the neutronic constants for each separate task with the

use of additional programs included into the SCALE package. The aim of this article is to analyze the applicability of the ORIGEN-S code 5) as a constituent of the SCALE code package to the calculation of changes in isotopic composition of VVER ftiel depending on its burnup. This was done in two parts: 0 modeling the content of individual isotopes in spent nuclear fuel of VVER-440 and VVER-1000 by means of the programs for reactor cell calculation with subsequent comparison of its results with the results of SCALE-4.3 and SCALE-4.4a; modeling the isotopic composition of VVER440 and VVER-1000 spent nuclear fuel by means of the SCALE code package Control Modules SAS214 - to analyze and compare the obtained results and to make the conclusions on the applicability and errors of the control module CSAS26 of the SCALE4.3 and SCALE4.4a code packages in the calculation of isotopic composition of VVER spent nuclear fuel. As the experimental data, the results of studying isotopic content of VVER-440 and VVER I 000, SNF completed in different years were used. While testing Control Module SAS2H, the special attention was paid to those isotopes which are suggested to be used while undertaking the work on substantiation of nuclear safety taking spent fuel burnup into account: - first of all, these are fuel isotopes, i.e. those participating in chain reaction: 235 U, 236u, 238u, Ta 231pu, 240pu, 24 PU;

- secondly, actinides: 234u, 238PU, 242PU124

'Am , 243 Am , 237NP;

- as well as 18 main products of ftiel isotopes fission caused by chain reaction:

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JAERI-Conf 2003-019 95 MO, 99Tc, 133cS, 135cS, 143 Nd, 152SM, 153 Eu,

101RU, 103R-h,

145Nd,

105 Pd,

18Pd, 109Ag,

147SM, 149SM, 150SM, 15ISM,

0

155W

0

Here, the results of completed testing are presented. Upon which base the conclusion is made concerning possibility of using the control module SAS2H to determine the isotopic content of VVER-440 and VVER-1000 fuel. Errors of identifying concentration of particular isotopes are being determined.

0

0

0

0

0

0

0

0

000400 0

0 0

The report was prepared within the framework of project Validation of ORIGEN-ARP and SCALE SAS2H for VVER and RBMK Fuel Designs)).

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4 4

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0004D

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2 I. Modeling VVER-440 Fuel Assembly

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004

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2. Analysis of data used for testing

4

127

I- central tube; 2- fuel rod; 3- fuel assembly shroud

The main technical parameters of the VVER-440 fuel assembly (FA) are provided in reference 67). The SAS2H control module is based on the ORJGEN-S code modeling fuel in one-dimensional approximation. A fuel assembly is described as infinite lattice of fuel rods. For VVER, this lattice is hexagonal. However, a number of elements disturbs the regularity of this lattice. For VVER-440 fuel, the components that disturb the regularity of the fuel rod hexagonal lattice are central tube, water gap between assemblies, and fuel assembly shroud (Figure 1). These components essentially impact on the neutron spectrum and fuel irradiation conditions, therefore, the possibility of their modeling should be provided in calculations. The SAS2H capabilities allow consideration of some differences between the actual fuel assembly design and a standard one. This is done through the INPLEVEL command permitting different options in modeling a fuel assembly. For example, if a completely regular fuel lattice is modeled, the command INPLEVEL = I is used. If we need to model deviations from this regular structure, commands INPLEVEL=2 or 3 should be used. Possibility of modeling VVEk-440 fuel

Fig.

VVER-440 Fuel Assembly Lattice

Uranium-water mixture 4) Zirconium central tube 2) or assembly -shroud 6) Water 1,3,5,7) 0

components (Figure 2 that disturb the regularity of the fuel rod hexagonal lattice should be provided in SAS211 calculations. All the above-mentioned is related to the conditions of modeling a fuel assembly as a whole, i.e. calculations of isotopic content averaged by FA are

Fig. 2 Model of VVER-440 fuel for control module SAS2H

concerned. Experimental data of isotopic content were obtained for individual fuel rods; in so doing, location of these fuel rods in the FA lattice impacts on their

1. fuel rods located near the central tube (conditions of their irradiation are resulted from additional impact of water in the tube);

The VVER-440 fuel assembly can be conditionally divided into three main zones:

isotopic composition.

2. fuel rods located in te

FA central part (their

lattice is regular); 3. fuel rods located at the FA periphery (conditions of their irradiation are resulted from additional impact of water in the interassembly gap).

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JAERI-Conf 2003-019

In the above regard, experiments should be modeled taking into account actual location of the fuel rod selected for the further analysis.

loading diagrams during the campaigns were determined with the DERAB code9) for modeling the reactor core with fine mesh. They have been

2.2. Modeling VVER I 000 Fuel Assembly

successfully used for the recent six years to model fuel campaigns of all VVER-440 and VVER-1000 reactors operating at Ukrainian NPPs.

VVER-1000 reactor fuel (Figure 3 is something more difficult in comparison with VVER-440 fuel from the point of view on geometrical model development for the calculational sequence SAS2H. The VVER-1000 assemblies are hexagonal in design and consist of one central tube, 312 fuel pin locations and 18 guide tubes. VVER-1000 assembly is without shroud. But, in general, the similar fuel pins separation into groups according to their allocation in the assembly as for VVER-440 Figure 4 is used.

3

Fuel with gadolinium (benchmark NEA/OECD) or water (1) Uranium-water mixture 4) Zirconium, fuel pin cladding or central -tube 2) Water (3,5)s I fuel pin; 2 - guide tube for control rod or burnable absorber; 3 - central tube

Fig. 4 Model of VVER-1000 fuel for control module SAS2H

Fig. 3. VVER-1000 Fuel Assembly Lattice All the above-mentioned is related to the conditions of modeling a fuel assembly as a whole, ie. calculations of isotopic content averaged by FA are concerned. Experimental data of isotopic content were obtained for individual fuel rods; in so doing, location of these fuel rods in the FA lattice impacts on their isotopic composition. In the above regard, experiments sould be modeled taking into account actual location of the fel rod selected for the arther analysis.

These codes were used for determining loading diagrams of VVER-440 fuel rods with 36% initial enrichment, and then isotopic composition of VVER-440 spent nuclear fuel was calculated for them. The best results are obtained when the 27bumuplib library is used and the computer code is corrected to incorporate the fuel rod location in the assembly. Below (Table 1) the average errors in calculating concentrations of individual isotopes are presented. The Table uses the following notions:

3. Calculation results of experiments with SAS2H control module

27burnupli - SCALE package standard library without correction of the computer model to incorporate fuel rod position in the assembly. 27burnuplib (correction) - SCALE package

3. 1. VVER-440 experiment

standard library with correction of the computer model to incorporate fuel rod position in the assembly.

For this part of work, fuel operation in the core of Novovoronezh NPP VVER-440 during the three initial campaigns was modeled. The reactor performance was modeled with the PYTHIA code )

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JAERI-Conf 2003-019 As it can be seen from the Table 2 results, which were received with the help of different libraries, correlate good with each other in the majority of cases. Their difference is much less than the calculation and experimental data difference. On that ground, it can be

Table Errors in calculating concentrations of individual isotopes in VVER-440 spent fuel Isotope Effor(Calde r.-I)'100% 44groupndfS 27bumuplib 238groupndf 27bumuplib "ITU 235U

-9 I 5.0

-7.3 5.1

5 -9.6 6.7

(correction) -5.9 -2.2

236U MVU__ 7"rPu _ 7"'P-u 74'T-u-

-2.1 0.3 12.0 7.2 4.9

-2.1 0.2 -5.9 10.1 0.9

-2.5 0.2 -12.3 10.0 6.8

-2.6 0.4 -9.8 -4.6 -5.6

'241PU N7Pu TA-m-

7.3 7.1 67.9 -7.7 121.4

14.1 -3.3 54.4 4.0 82.2

9.8 7.8 68.5 -5.4 137.5

5.8 -0. 1 34.8 -1.2 78.2

242cM

arm

1 1

1 1

stated that the experimental data were insufficient or were received with the considerable fallibility. 4. Results of Benchmark Calculations with SAS2H Control Module 4. 1. VVER-440 Reactor. AER CB-2 Benchmark Calculations Works 11). 2) describe the calculational benchmark prepared by the international research group AER (Atomic Energy Research) that combines experts from the countries operating VVER. The benchmark was developed for the purpose of exploring the possibility to use different codes for determination of isotopic composition of VVER-440 spent nuclear fuel and the

3.2. VVER-1000 experiment Isotope composition of VVER-1000 reactor ftiel assembly was calculated according to the materials of references). Referencelo) presents the results of isotopic content measurement of spent fuel in WER- I 000 reactor. For .this measurement, one FA was used from VVER- I under operation at Zapori2hya NPP. Its in i enrichment was 392%. This was under ope during one campaign, and unloaded from the core 1999. Isotopic composition of three specimens with different bumup levels is presented. The concentrations of the most important transuranium elements that have the most significant influence on spent fuel characteristics were analysed. Table 2 presents the main calculation results, which were received with the different libraries of control module SAS2H. Accuracy of the relevant experimental data is accepted as 5 %.

further use of this information in criticality calculations of fuel storage systems (application of the bum-up credit to VVER fuel). Taking into account that there is the experimental information insufficient or absent on the isotopic concentration in spent nuclear ftiel, comparison with the benchmark results permits us to conduct testing for the greater number of nuclides. The benchmark describes assembly selected, its loading modes during operation, and the results obtained. During benchmark calculations with the SAS2H control module, fuel design, loading modes during operation, and cooling modes were specified according to the benchmark data. Below, Table 3, the average errors are presented in calculating concentrations of individual isotopes given the use of the 27bumuplib library: Table 3 Deviation in calculating concentrations of individual isotopes in VVER-440 spent fuel

Table 2 Deviation in calculating concentrations of individual isotopes in VVER-1000 spent fuel Isotope "'_ '7'-U "T__ 238u M-Pu

... Pu M"P-uM-Pu 242PU

Am "7Am

Deviatio (Calc/exper.44groupndf5 238groupndf5 -13.37 -13.54 -0.50 -0.11 -18.49 -19.07 0.09 0.08 -29.92 -30.56 -14.91 -13.84 -7.76 -12.19 -17.20 -8.11 -11.75 -3.98 -91.41 -90.57 -33.08 -11.89

1

-27.69 -4.93

*100% 27bumuplib -12.04 -0.64 -18.31 1 0.08 -27.22 -11.19 -17.95 -5.60 -16.65 -90.51

Deviation 0.8%

236U

1.3% -0.4% 13.2% 6.1% 3.2%

23SU 237 Np

zj5pu 239PU Z40pU 241pu

-1.0%

242PU

3.9% -9.5% 4.3% -2.1%

-24

'Am

__243 Am

95mo 99

Tc 101Ru

-36.92

0.1 %

0.7% -1.0%

103P

LEL86__J I - 703 -

Isotope Z.33U

h 109Ag JJJCS

0.9% 11.0% 2.2%

JAERI-Conf 2003-019 14JNd 143Nd 141sM 149sM 150SM DISM

1.5YSm 153Eu l:'3Gd

0.4% 0.9% -0.7% 12.6% 4.7% 16.5% 7.4% -8.8% 89.7%

Several calculation states were included into the benchmark exercise. These states cover the operational states, and cold conditions. The requested results include ki,,f values, pin power distributions, and isotopic concentrations. The benchmark is completely calculated, and uncertainties are indicated in comparison with the calculation results of other participants. The modeling of this benchmark is enough

In general, the calculational results satisfactory correlate. However, the rather great discrepancy for 155 Gd should be noted. 4.2. VVER- I 000 Reactor. OECD/NEA Benchmark Calculations Benchmark OECD was chosen for calculations 13). It includes the calculations of uranium and MOX fuel for the VVER reactors. Below, the calculation results of uranium fuel only are presented. The benchmark model consists of assemblies that are typical for the advanced designs that are under active development in Russia for the VVER-1000 reactors. The VVER-1000 assemblies are hexagonal in design and consist of one central tube, 312 ftiel pin locations 12 of which are U/Gd rods), and 18 guide tubes. The clad and structural materials are composed of a Zr-Nb alloy. The UGD assembly is shown in Figure 5, and consists of fuel rods with 37 wt.% 235 enrichment. The 12 U/Gd pins have a U enrichment of 3.6 wt.% and a Gd2O3 content of 4.0 wt.%.

difficult because fuel assembly includes several types fuel elements, which results in necessity of simplification while modeling and in uncertainties while results processing, Tables 4 and .

Table 4 Uncertainties in calculating concentrations of individual isotopes in VVER-1000 spent fuel (assembly averaged) Isotope 2 35 U

Tj6Q TjgU 239PU 240PU

11-pu 742puT55G-d TrGd

Error alc/exper.-I) 100% 44groupndf5 238groupndf5 27bumuplib -1.10 -0.52 -0.49 3.36 2.89 3.74 0.11 0.10 0.07 -3.43 -0.58 2.87 1.12 4.02 -2.90 0.03 2.69 9.58 8.12 9.85 -2.19 -22.40 -20.70 -38.08 -36.98

Table Uncertainties in calculating concentrations of individual isotopes in VVER-1000 spent ftiel (comer fuel pin) Isotope U 7wU 7"U 239 Pu 240PU

7'r'. 79P

Error aic/exper.-I) 100% +,+groupndf5 238grouvndf5 27bumuplib 0.73 -3.78 0.23 -14.98 -13.44 -24.27 -22.52

1.63 -4.32 0.18 -11.23 -10.69 -20.82 -20.29

0.58 -3.59 0.22 -13.19 -18.09 -20.39 -29.78

5. Conclusions Analysis of the results permits the following Cell types: 1. Central tube cell. 2. Fuel cell 3.7 wt.% LEU). 3. Guide tube cell. 4. Fuel cell 3.6 wt.% LEU with 4.0 wt.% Gd2O3). Fig. 5 UGD assembly configuration

conclusions: 1. SCALE code package SAS2H control module permits sufficiently accurate determination of isotopic composition of VVER spent nuclear fuel. 2. The most comprehensive testing based on independent sources including both experimental results and calculational benchmarks has been conducted for 231U, 236u, 238u, 239PU, 240PU, 241PU, 242pU fuel isotopes.

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JAERI - Corif 2003 - 019

3. More accurate are results received during calculation taking into account the disposition of experimental fuel pins, and their operational conditions.

RBMK-1000 Reactors. Russian Scientific Center "Kurchatov Institute", 2002 (in Russian). I 1) L.Markova. Calculation Burnup Credit Benchmark No.2 (CB2). 7th AER

4. The SAS2H sequence using is more appropriate in comparison with ARP sequence especially in difficult cases (presence of control or burnable absorber rods, etc.)

SYMPOSIUM on VVER Reactor Physics and Reactor Safety Hornitz near Zittau, Germany, Sept. 23-26,1997. 12) L.Markova. CB2 Result Evaluation (VVER-440 Burnup Credit Benchmark). 7th AER Symposium on VVER Reactor Physics and Reactor, Slovakia, October 48, 1999. 13 A VVER-1000 LEU and MOX Assembly Computational Benchmark. Specification and Results. OECD, 2002.

5. SCALE Libraries 27BURNUPLIB or 44GROUPNDF5 can be recommended for calculating isotopic composition of VVER spent nuclear fuel.

Acknowledgements The authors are grateftil to the US Government for financial support in realization of this work.

References 1) SCALE: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation, NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2R6), Vols. 1, 11, and III, May 2000. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-545. 2) Y.Kovbasenko, V.Khalimonchuk, A.Kuchin, Y.Bilodid, M.Yeremenko, O.Dudka "Validation of SCALE sequence CSAS26 for criticality safety analysis of VVER and RBMK fuel designs" NUREG/CR-6736, PNNL-13694,2002. 3) 0. W. Hermann, C. V. Parks. SAS2H A Coupled One-Dimensional Depletion And Shielding Analysis Module. Oak Ridge Natl. Lab., 2000. 4) S. M. Bowman, L. C. Leal. ORIGEN-ARP: Automatic Rapid Process For Spent Fuel Depletion, Decay, And Source Term Analysis. Oak Ridge Natl. Lab., 2000. 5) 0. W. Hermann, R. M. Westfall. ORIGEN-S: SCALE System Module To Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup And Decay, And Associated Radiation Source Terms. Oak Ridge Nati. Lab., 2000. 6) R.Z. Arninov, V.A. Khrustalyov, A.S.Dukhoven'sky, A.S.Ocadchy. NPP with VVER: Modes, Characteristics, Efficiency. Energoatomizdat, Moscow, 1990 (in Russian). 7) Design and Performance of WWER Fuel, Technical Report Series No. 379, LAEA, Vienna, 1996. 8) Programm PYTHIA zur Berechming des Makroabbrands in Druckwasserreaktoren vom Typ WWER, K.A.B.GmbH, Berlin, 1998. 9) Moller A. DERAB Code Manual, K.A.B.GrnbH, Berlin, 1998 10) A.L.Tataurov, V.M.Kwator. Report on RD Work: Calculated-Experimental Studying Nuclide Composition of Fuel Spent in VVER-1000 and - 705 -