CALCULATION OF ISOTOPIC COMPOSITION DURING CONTINUOUS IRRADIATION AND SUBSEQUENT DECAY IN BIOLOGICAL SHIELD OF THE TRIGA MARK ΙΙ REACTOR

Transcription

1 International Conference Nuclear Energy for New Europe 2002 Kranjska Gora, Slovenia, September 9-12, CALCULATION OF ISOTOPIC COMPOSITION DURING CONTINUOUS IRRADIATION AND SUBSEQUENT DECAY IN BIOLOGICAL SHIELD OF THE TRIGA MARK ΙΙ REACTOR ABSTRACT Matjaž Božič Nuclear Power Plant Krško Vrbina 12, SI-8270 Krško, Slovenia Tomaž Žagar, Matjaž Ravnik Jožef Stefan Institute Reactor Physics Division Jamova 39, SI-1000 Ljubljana, Slovenia The calculation of detailed radioisotopic composition in biological shield and verification with experimental results is presented. Neutron fluxes in different spatial locations in biological shield are obtained with TORT code. Libraries used with TORT code are BUGLE-96 (coupled 47 neutron groups and 20 gamma groups) and library derived from VITAMIN-B6 (coupled library with 199 neutron groups and 42 gamma groups). Isotopic composition is calculated for various irradiation intervals and also for several decay intervals from different initial irradiation intervals. Neutron fluxes obtained with TORT code are used as input for activation calculation in several locations of the biological shield. Several neutron spectrum constants from known neutron flux have to be derived before isotopic calculation. SCALE code package is used for isotopic calculation. Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in beam port 4 (irradiation channel 4) in TRIGA Mark ΙΙ reactor. These experimental results are used for verifying the calculation models. The agreement between the measured and calculated activities is within the interval of experimental error. 1 INTRODUCTION Determination of the residual activity for reactor structure materials is one of the main tasks to solve before reactor decommissioning. One of the most important activated parts is the biological shield (concrete). Activation depends on the material properties and neutron flux. Neutron flux can be obtained either with experimental determination or using calculation methods. Experimental neutron activation data must be measured due to verification of the results from calculations. Calculation of neutron fluxes in biological shield (concrete) of the TRIGA Mark II research reactor in Ljubljana is performed with computer code TORT [2]

2 (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). and BUGLE-96 library [3]. In the first place comparison is made between calculated and measured saturated activity. The next step was calculation of the radioisotopic composition in different spatial points within biological shield with various irradiation times and different decay time intervals. Known calculation neutron fluxes with established neutron flux spectra factors (THERM, RES and FAST, derived from neutron flux spectra, BUGLE-96 library with 47 neutron energy groups) are used with code ORIGEN-S [7] from SCALE code package. Main long-lived radioactive isotopes generated after irradiation process in different concrete samples with their generating reactions [9] are presented in Table 1. Table 1: Long-lived isotopes identified in both types of concrete Element Symbol Atom. Numb. REACTION Cobalt Co 27 Co59 (n,γ) Co60 THERMAL Ni69 (n,p) Co60 FAST Cu63 (n,α) Co60 FAST Mangan Mn 25 Mn55 (n,2n) Mn54 FAST Fe54 (n,p) Mn54 FAST V51 (α,n) Mn54 CHARG. PART. Zinc Zn 30 Zn64 (n,γ) Zn65 THERMAL Zn66 (n,2n) Zn65 FAST Cesium Cs 55 Cs133 (n,γ) Cs134 THERMAL Ba134 (n,p) Cs134 FAST yield from fission Europium Eu 63 Eu151 (n,γ) Eu152 THERMAL Gd152 (n,p) Eu152 FAST Eu153 (n,2n) Eu152 FAST Europium Eu 63 Eu153 (n,γ) Eu154 THERMAL Gd154 (n,p) Eu152 FAST yield from fission Calcium Ca 20 Ca40 (n,γ) Ca41 THERMAL Ca42 (n,2n) Ca41 FAST Calcium Ca 20 Ca44 (n,γ) Ca45 THERMAL Ti48 (n,α) Ca45 FAST Sc45 (n,p) Ca45 FAST Potassium K 19 K39 (n,γ) K40 THERMAL Ca40 (n,p) K40 FAST % natural abundance Argon Ar 18 Ar38 (n,γ) Ar39 THERMAL Ar40 (n,2n) Ar39 THERMAL K39 (n,p) Ar39 FAST Ca42 (n,α) Ar39 FAST Barium Ba 56 Ba132 (n,γ) Ba133 THERMAL Ce136 (n, α) Ba133 FAST Ba134 (n,2n) Ba133 FAST Iron Fe 26 Fe54 (n,γ) Fe55 THERMAL Ni58 (n,α) Fe55 FAST HALF-LIFE Co60: y Mn54: d Zn65: d Cs134: y Eu152: 13.6 y Eu154: 8.8 y Ca41: 103*10 3 y Ca45: d K40: 1277*10 6 y Ar39: 269 y Ba y Fe55: y

3 COMPARISON BETWEEN EXPERIMENTAL AND CALCULATED DATA Comparison is made for both types of concrete (Concrete Type 04 and Barytes Concrete). Generally, good agreement is observed in both cases. Concentrations of trace elements (Sr, Co, Cs, Eu, Zn) found in both types of concrete were determined with NAA method [10]. These trace elements are parent nuclides for several long-lived activation products found in activated concrete samples. Six long-lived isotopes are important according to experimental data. These isotopes are 54 Mn, 60 Co, 65 Zn, 134 Cs, 152 Eu and 154 Eu. Their parents and generating reactions are presented in Table 1. Only 54 Mn is produced by fast neutrons mostly in (n,p) reaction. Other five isotopes are produced by thermal neutrons mostly in (n,γ) reaction and contents of their parents are determined with NAA method. Experimental activity of 54 Mn is greater than calculated for approximately 30% for both types of concrete. For other five isotopes, 60 Co, 65 Zn, 134 Cs, 152 Eu and 154 Eu experimental activation is approximately five times greater than calculated data for reactions with thermal neutrons for both types of concrete. Reason for the discrepancy in both concrete cases is the group model of ORIGEN-S code and also the library BUGLE-96 used in transport calculation. BUGLE-96 library has only two energy groups below 0.5 ev. It has been produced for light water reactor shielding and reactor pressure vessel dosimetry applications. Neutrons above thermal range are well represented in both libraries. ORIGEN-S calculation is improved by calculating the spectrum weight factors (THERM, RES, FAST) and absolute value of thermal flux at each calculated spatial point in reactor body. These factors are all derived from calculated neutron flux with transport code TORT (using BUGLE-96 library). Thermal neutron spectrum is important for (n,γ) reactions which is the case for five isotopes, 60 Co, 65 Zn, 134 Cs, 152 Eu and 154 Eu (the fact is that experimental data is five times greater than calculated data for these five isotopes). Another reason for difference between experimental and calculated results is inaccurate composition data ( 54 Mn, 60 Co, 65 Zn, 134 Cs, 152 Eu and 154 Eu) measured by NAA method. Experimental and calculated saturated activities are presented in Figure 1 and 2. Neutron spectrum important for generating a particular isotop is denoted together with the isotope (TH, F). Experimental and calculated saturated activity [8] are determined in different depths of the concrete shield 2.1 Concrete Type 04 (ordinary concrete) Comparison with experimental saturated activity for Concrete Type 04 (ordinary concrete) is presented on Figure 1. ORIGEN-S code generates additional five long-lived isotopes 39 Ar, 40 K, 41 Ca, 45 Ca, 55 Fe which were not detected by experiments. 41 Ca, 45 Ca, 55 Fe are not gamma ray emitters. 39 Ar is in gaseous state. Isotope 40 K has half-time 1277x10 6 years with natural abundance %. All five isotopes are mostly generated by thermal neutron flux with (n,γ) reactions.

4 Saturated Activity Density (Bq/g) - LOG Concrete Depth (cm) meas - Mn 54 meas - Co 60 meas - Zn 65 meas - Cs 134 meas - Eu 152 meas - Eu 154 Mn 54: F Co 60: TH Zn 65: TH Cs 134: TH Eu 152: TH Eu 154: TH Figure 1: Comparison of calculated with experimental data for Concrete Type 04 at various depths measured radially from the inner wall 2.2 Barytes Concrete (heavy concrete) Barytes Concrete is also used as biological shield in TRIGA Mark II reactor. Its composition will be used in final calculations with decay time dependence after different irradiation intervals. The most important long-lived isotope is 133 Ba. It is created mostly by thermal neutron flux with (n,γ) reactions. It has the biggest magnitude of the saturated activity among all isotopes. Comparison with experimental data for Barytes Concrete is presented in Figure 2. In addition to experimentally observed isotopes, ORIGEN-S code generates three long-lived isotopes. 41 Ca, 45 Ca, 55 Fe which were not detected by experiments. 41 Ca, 45 Ca, 55 Fe are not gamma ray emitters. Isotopes 39 Ar, 40 K are not generated with ORIGEN-S code due to absence of K in composition description for Barytes Concrete used for calculations.

5 ,00E+06 Saturated Activity Density (Bq/g) - LOG Concrete Depth (cm) Mn 54: F Co 60: TH Zn 65: TH Ba 133: TH Eu 152: TH Eu 154: TH meas - Mn 54 meas - Co 60 meas - Zn 65 meas - Cs 134 meas - Ba 133 meas - Eu 152 meas - Eu 154 Cs 134: TH Figure 2: Comparison of calculated with experimental data for Barytes Concrete at various depths measured radially from the inner wall 3 ACTIVITY OF BIOLOGICAL SHIELD Calculations are performed in different spatial points (different concrete depths in the reactor midplane) in biological shield of the TRIGA Mark II reactor. Biological shield is made of Barytes Concrete. Activity (Bq/g) is calculated for different irradiation times: 1 year, 10 years, 20 years, 40 years. ORIGEN-S code is used for calculations. Factors THERM, RES, FAST and absolute thermal flux are obtained with transport code TORT [6]. Activity is calculated for different decay time intervals equal for each irradiation interval. Decay intervals are 1 year, 10 years, 30 years, 100 years, 300 years, 1000 years, 3000 years, years, years, years. Generally, total activity (Bq/g) is almost the same at irradiation times 20 years and 40 years. Total activities at irradiation times 1 year and 10 years and subsequent decay compared with previous two irradiation times are smaller due to not saturated concentrations of the main isotopes which are the major contributors to the total activity ( 133 Ba, 41 Ca, 39 Ar, 14 C). Total activity (Bq/g) for all four different irradiation times and selected decay times calculated at different depths are presented in Figure 3 below:

6 ACTIVITY (Bq/g) 1,00E-02 1,00E-05 1,00E-04 1,00E-03 1,00E-02 DECAY TIME (year) irradiation time: 1y irradiation time: 10y irradiation time: 20y irradiation time: 40y Figure 3: Specific activity of Barytes Concrete at depth 4.1 cm at various irradiation times As it is mentioned total activity for 20 years and 40 years of irradiation is practically the same. Total activity for 10 years and 1 year is smaller than total activities described above. ACTIVITIY (Bq/g) 1,00E-02 1,00E-03 1,00E-04 1,00E-05 1,00E-06 1,00E-07 1,00E-08 1,00E-09 1,00E-10 1,00E-05 1,00E-04 1,00E-03 1,00E-02 DECAY TIME (year) 4.1 cm 34.1 cm 64.1 cm 74.4 cm cm cm Figure 4: Total activity as a function of concrete depth at 40 years irradiation

7 Dependence on the portion in the concrete (depth) is presented in Figure 4. In decay period between 1 year and 100 years 133 Ba is the main contributor to total activity at all depths. Main contributor to the total activity after 100 years is isotope 41 Ca. Half-life of 41 Ca is years. 41 Ca is produced mostly by thermal neutron spectrum with neutron reaction (n,γ) from parent isotope 40 Ca. Barytes Concrete contains 4.07 wt% of 40 Ca. Important contribution to the total activity between decay period 100 years an 1000 years is isotope 39 Ar. Barytes Concrete does not contain element K. Parent isotope for 39 Ar is isotope 42 Ca. The neutron reaction for generating 39 Ar is (n,α) with fast neutron spectra. Isotope 42 Ca is naturally abundant. Half-life of 39 Ar is 269 years. Considerable contribution to the total activity between decay period 1000 years an years is isotope 14 C. The neutron reaction for generating 14 C is (n,p) with fast neutron spectra. Parent isotope is 14 N. Isotope 14 N is naturally abundant. Half-life of 14 C is 5730 years. Observable contribution to the total activity after years of decay is by isotopes 10 Be with half-life 1600*10 6 years and 40 K with half-life 1277x10 6 years. Both isotopes are mostly produced by thermal neutrons by (n,γ) neutron recation. Activities for individual isotopes and also for total activity (Bq/g) are presented in figure 5. The values presented are for the spatial point in biological shield, 4.1 cm in concrete depth at core midplane with irradiation times of 40 years. ACTIVITY (Bq/g) 1,00E-02 1,00E-03 1,00E-04 1,00E-05 1,00E-06 1,00E-07 1,00E-08 1,00E-09 1,00E-10 1,00E-11 1,00E-12 1,00E-13 1,00E-14 1,00E-15 1,00E-05 1,00E-04 1,00E-03 1,00E-02 DECAY TIME (year) h 3 be 10 c 14 p 32 p 33 ar 37 ar 39 k 40 ca 41 ca 45 sc 46 cr 51 mn 54 fe 55 fe 59 co 60 ni 63 zn 65 sr 85 sr 89 y 89m cs134 ba133 eu152 eu154 eu155 gd153 total Figure 5: Isotopic activity (Bq/g) versus decay time at 4.1 cm concrete depth (core midplane) and 40 years of irradiation time Portions of individual isotopes to the total activity (Bq/g) are clearly visible from Figure 5.

8 CONLUSIONS Results of calculation of neutron activation of biological shield of TRIGA Mark II reactor are presented. Calculated results are compared to the experimental obtained after irradiation of samples in beam port 4 (Concrete Type 04 and Barytes Concrete irradiation samples). Calculated saturated activity (Bq/g) is compared. Saturated activity of the isotopes generated fast neutron reactions is in excellent agreement (within 30%). On the other hand saturated activity of the isotopes produced in thermal neutron reactions are approximately five times lower than in experimental values. The reason is lower absolute thermal flux used in input for code ORIGEN-S. Possible reason is also inaccuracy in composition for both types of concrete obtained by NAA method. Detailed calculation of isotopic composition is also presented with individual isotopic activity and total activity. Different irradiation times, various decay intervals at several different concrete depths are also presented. Total activity after the irradiation is observed up to 100 cm depth in Barytes Concrete used as biological shield. All main long-lived isotopes are determined through by using of SCALE code package. 5 REFERENCES [1] R. Jeraj, M. Ravnik, TRIGA Mark II Reactor U(20) Zirconium Hydride Fuel Rods In Water With Graphite Reflector, International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC/(95)03/III, Volume III, IEU-COMP-THERM- 003, 1995, p [2] W. A. Rhoades, D. B. Simpson, The TORT Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code, (TORT Version 3), OAK RIDGE NATIONAL LABORATORY, ORNL/TM-13221, [3] BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, OAK RIDGE NATIONAL LABORATORY, DLC-185,1996 [4] T. Žagar, M. Ravnik, "Neutron Activation Measurements in Research Reactor Concrete Shield", Proc. Int. Conf. Nuclear Energy in Central Europe 2001, Portorož, Slovenija, September, Nuclear Society of Slovenia, [5] T. Žagar, M. Ravnik, "Measurement of Neutron Activation in Concrete Samples", Proc. Int. Conf. Nuclear Energy in Central Europe 2000, Bled, Slovenija, September, Nuclear Society of Slovenia, [6] M. Božič, T. Žagar, M. Ravnik, "Calculation of Neutron Fluxes in Biological Shield of the TRIGA Mark II Reactor" Proc. Int. Conf. Nuclear Energy in Central Europe 2001, Portorož, Slovenija, September, Nuclear Society of Slovenia, 2001 [7] O. W. Herman, 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 NATIONAL LABORATORY, March 2000 [8] G. F. Knoll, Radiation Detection and Measurement, John Wiley & Sons, New York, USA, 1979, pp [9] G. Erdtmann, W. Soyka, The Gamma Rays of the Radionuclides, Verlag Chemie, Weinheim, New York, 1979, pp [10] T. Žagar, Activation of TRIGA Research Reactor Body, Doctoral Thesis, University of Maribor, 2002, pp