American Journal of Modern Physics
Volume 5, Issue 5, September 2016, Pages: 135-141

Substantiation of Data Files of JEFF-3.1.2 for Safety Analysis of TRIGA Mark-II Reactor through the Scrutiny of Integral Parameter of Benchmark Lattices TRX and BAPL

Md. Mominul Islam1, *, Md. Mahbubul Haque2, S. M. Azharul Islam3

1Department of Physics, Hajee Mohammad Danesh Science & Technology University, Dinajpur, Bangladesh

2Materials Science Division, Atomic Energy Centre Dhaka, Bangladesh Atomic Energy Commission, Dhaka, Bangladesh

3Department of Physics, Jahangirnagar University, Savar, Dhaka, Bangladesh

Email address:

(Md. M. Islam)

*Corresponding author

To cite this article:

Md. Mominul Islam, Md. Mahbubul Haque, S. M. Azharul Islam. Substantiation of Data Files of JEFF-3.1.2 for Safety Analysis of TRIGA Mark-II Reactor through the Scrutiny of Integral Parameter of Benchmark Lattices TRX and BAPL. American Journal of Modern Physics. Vol. 5, No. 5, 2016, pp. 135-141. doi: 10.11648/j.ajmp.20160505.13

Received: August 4, 2016; Accepted: August 22, 2016; Published: September 10, 2016


Abstract: The aim of this analysis is to bear out the nuclear data files of JEFF-3.1.2 for theoretical safety analysis of a 3 MW TRIGA MARK-II research reactor is custom-made at AERE, Dhaka, Bangladesh through the study of integral parameters of benchmark lattice TRX and BAPL of thermal reactor. The basic evaluated nuclear data files of JEFF-3.1.2 are selected for TRIGA reactor and processed by using nuclear data processing code NJOY99.0. Different cross-sections of U-235 and U-238 are computed from the NJOY output of the evaluated nuclear data library. The 69 group cross-section library is engendered from the processed file for reactor code WIMSD-5B. From the generated 69 group cross-section library, the integral parameters of yardstick lattices TRX and BAPL are premeditated by using cell code WIMSD-5B. The calculated integral parameters are compared to the deliberated values as well as the consequences of Monte Carlo Code MCNP. From the assessment it is found that all the integral parameters are in good concurrence with some suspicions. Through benchmarking the integral parameters of TRX and BAPL lattices this analysis reflects the support to the evaluated nuclear data files of JEFF-3.1.2 for safety analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.

Keywords: BAPL, JEFF-3.1.2, NJOY99.0, TRIGA MARK-II, TRX and WIMSD-5B


1. Introduction

The Evaluated Nuclear Data Files (ENDF) system was built up for the storage and repossession of evaluated nuclear data to be used for applications of nuclear technology [1]. A vital corollary of each appraisal must be completed for its intended purpose. These applications run many features of the system including the selection of materials to be comprised, the information used, the formats used and the testing is required before a library is released. If the entailed data are not obtainable for various finicky reactions, the evaluator should provide them by using nuclear models [2]. The evaluated data sets are prepared in ENDF format and converted into forms appropriate for testing and actual applications using processing codes. Processing codes that generate point-wise and group averaged cross sections for use in neutronics calculations from an ENDF library are available. The basic data formats for an ENDF library are developed in such a manner that few constraints are placed on using the data as input to the codes that generate any of the secondary libraries [3]. The computer code NJOY99.0 [4] is used for converting nuclear data in ENDF-6 format into libraries as Joint Evaluated Fission and Fusion Data Library (JEFF) in Europe [5], Japanese evaluated nuclear data library (JENDL) [6], Chinese evaluated nuclear data library (CENDL) [7], ENDF/B-VI in USA [8] and BROND in Russia [9]. Joint Evaluated Fission and Fusion Data Library is high quality nuclear data libraries for accessible and prospective nuclear energy systems and this library involves evaluation efforts that cover the main nuclear data needs in the fields of fission and fusion applications [10]. The updated version JEFF-3.1.2 which is modified edition of JEFF-3.1.1 [11] is our interest in this present research. To study the safety analysis of Triga Mark –II research reactor by JEFF-3.1.2 data library, a careful verification is required for the usable data files of that library. The WIMS code is a freely accessible thermal reactor physics lattice-cell code used widely especially by scientists for thermal research and power reactor applications. In the present work the 69 group cross-section library is generated by using computer program NJOY99.0 and WIMSD-5B [12, 13]. The bench mark lattice TRX and BAPL is used for benchmarking the integral parameter for the generated library. The calculated integral parameters are compared to the standards values.

2. Methods

The tools of this study are computer program: NJOY99.0, WIMSD-5B; evaluated nuclear data library: JEFF-3.1.2; benchmark lattices: Thermal Reactor-one region lattice (TRX) and Bettis Atomic Power Laboratory-one region lattice (BAPL).

2.1. Computer Code NJOY99.0

The nuclear data processing system NJOY having new version NJOY99.0 is a modular computer code used for data processing in ENDF-6 format. One of the common applications of NJOY99.0 is to generate 69 group cross-section library from basic nuclear data library. The 69 group cross-section library for WIMSD-5B code from basic data files of JEFF-3.1.2 is created by NJOY99.0.

Table 1. Properties of TRX benchmark lattice.

Segment External radius in cm Nuclei Concentration (E 24 atoms/cm3)
Fuel 0.4915 235U 6.2530E-04
238U 4.7205E-02
Void 0.5042 --------- ------
Clad 0.5753 Al 6.025E-02
Moderator * 1H 6.676E-02
16O 3.338E-02

*Lattices spacing are 1.8060 cm& 2.1740 cm in triangular arrays

Table 2. Properties of BAPL benchmark lattice.

Segment External radius in cm Nuclei Concentration (E 24 atoms/cm3)
Fuel 0.4864 235U 3.1120E-04
238U 2.3127E-02
Void 0.5042 --------- ------
Clad 0.5753 Al 6.025E-02
Moderator ** 1H 6.676E-02
16O 3.338E-02

**Lattices spacing are 1.5578cm, 1.6523cm and 1.8057 cm

Table 3. Concern isotope of TRIGA with the respective material ID.

SL. NO. Isotope Material ID.
01 1-H-1 125
02 5-B-10 525
03 6-C-12 625
04 7-N-14 725
05 8-O-16 825
06 13-Al-27 1325
07 14-Si-28 1425
08 24-Cr-52 2431
09 25-Mn-55 2525
10 26-Fe-56 2631
11 28-Ni-58 2825
12 40-Zr-91 4028
13 68-Er-166 6837
14 68-Er-167 6840
15 82-Pb-207 8234
16 92-U-235 9228
17 92-U-238 9237

2.2. Reactor Code WIMSD-5B

WIMS consisting of a lattice transport code and the associated library WILLIE is used to unravel various thermal reactor problems. The unique WIMSD structure is used with 14 fast group between 10 MeV and 9.11 keV; 13 resonance group between 9.118 keV and 4eV; and 42 thermal groups from 4 eV and 0 eV [14]. Rejoinder of U-235 and U-238 is taken to compute the integral parameters of benchmark lattices by using WIMSD-5B code.

2.3. Benchmark Lattices

The H2O- moderated uranium lattices TRX-1 and TRX-2 [15] and H2O-moderated uranium oxide critical lattices BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 is used for benchmarking of several integral parameter. BAPL-1, BAPL-2 and BAPL-3 used uranium oxide fuel enriched 1.311wt%; TRX-1, TRX-2 used uranium metal fuel in U-235 enriched to 1.305wt%. These five lattices are called benchmark lattice. The material and dimensional properties of TRX benchmark lattices are listed in Table-1 [16] and properties of BAPL lattices listed in Table-2 [17]. The interaction of U-235 and U-238 nuclei at 300K is used to compute the integral parameter of the benchmark lattices using the reactor code WIMSD-5B. The WIMSD-5B is also used to determine neutron cross-section in thermal as well as epithermal range of U-235 and U-238 isotopes for each benchmark lattices.

2.4. Calculation Techniques

The exactness of the processed Group-wise Evaluated Nuclear Data File (GENDF) is analyzed to demonstrate the worth of the previously evaluated data. The chain of NJOY99.0 modules [18], which have been used to generate the 69-group cross section library, is represented by flow chart in Fig.-1. The data strips are routed using NJOY99.0, which can rich the new quality feature of the database. The isotopes listed in Table-2 are concern to the TRIGA Mark-II at AERE, Dhaka, Bangladesh. These elements have been processed in RECONR- BROADR- UNRESR- THERMR- GROUPR- WIMSR cycle by Pentium-IV PC in DOS command mode [19]. Using the WILLIE and WIMSD-5B code 69-group cross-section library is generated from the processed isotope of JEFF-3.1.2. Fission cross-section, absorption cross-section, captured cross-section of U-235 and U-238 are computed for TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 lattices through the generated 69-group cross-section libraries of JEFF-3.1.2 by using WIMSD-5B. The integral parameters  and  of TRX and BAPL lattices are represented in equations 2 to 5 [20]. The effective multiplication factor is noted by equation 1. The integral parameter of TRX and BAPL lattices of thermal reactor are calculated using this equation. The evaluated values of the integral parameters have been compared with the experimental values by cross-section evaluated working group (CSEWG) [21]. The overall analysis is performed at the department of Physics, Jahangirnagar University, Bangladesh.

keff= (neutron production from fission in one generation) / (neutron absorption in the preceding generation + neutron leakage in the preceding generation)       (1)

ρ28 = Ratio of epithermal to thermal neutron captures cross-section of 238U = (c)38epth/(c)38th= (a-f)38epth/(a-f)38th    (2)

δ25 = Ratio of epithermal to thermal neutron fission cross section of 235U = (f)35epth/(f)35th                (3)

δ28 = Ratio of 238U fission to 235U fission = (ft)38 / (tf)35                 (4)

C*= Ratio of 238U captures to 235U fissions = (ct)38/(tf)35=(ta-tf)38/(tf)35           (5)

Figure 1. Flow chart of nuclear data processing code NJOY99.0.

3. Results

The NJOY output of the two isotopes U-235 and U-238 are compared in Tables 4 to 5 within the thermal range. The calculated epithermal absorption cross section, thermal absorption cross section, total absorption cross section, epithermal fission cross section, thermal fission cross section, epithermal capture cross section and total fission cross section for neutrons on U-235 and U-238 of benchmark lattices TRX and BAPL are plotted in Figs. 2 to 6. The values of effective multiplication factor keff for TRX and BAPL lattices are calculated and listed in Tables 6 & 7. The calculated values of other integral parameters ρ28, δ25, δ28 and C* for TRX and BAPL lattices are summarized in Tables 8 & 9 and compared with experimental values by CSEWG.

In the horizontal axes of graphs from Figs. 2 to Fig. 6, the symbols a, b, c, d, e, f and g represent epithermal absorption cross-section, thermal absorption cross-section, total absorption cross section, epithermal fission cross-section, thermal fission cross-section, epithermal capture cross-section and total fission cross-section, respectively.

Figure 2. Cross-section of U-235 & U-238 for TRX-1 lattice.

Figure 3. Cross-section of U-235 & U-238 for TRX-2 lattice.

Table 4. Absorption & Fission cross-section from NJOY out of U-235 & U-238 in JEFF-3.1.2.

Energy group no. Thermal energy (eV) Absorption cross-section in barn Fission cross-section in barn
U-235 U-238 U-235 U-238
28 4.000 43.533 0.64026 26.933 3.1331E-06
29 3.300 30.120 0.51490 23.401 3.1620E-06
30 2.600 14.579 0.46582 11.295 3.1373 E-06
31 2.100 26.410 0.45476 15.919 3.6127E-06
32 1.500 24.804 0.46632 19.098 3.9454E-06
33 1.300 73.260 0.47913 53.315 4.1586E-06
34 1.150 136.64 0.48902 106.58 4.2940E-06
35 1.123 113.36 0.49210 105.38 4.3363E-06
36 1.097 116.77 0.49512 96.053 4.3779E-06
37 1.071 101.49 0.49814 85.271 4.4194E-06
38 1.045 89.239 0.50110 76.214 4.4601E-06
39 1.020 80.565 0.50397 69.443 4.4995E-06
40 0.996 74.315 0.50766 64.841 4.5463E-06
41 0.972 69.948 0.51165 61.456 4.5954E-06
42 0.950 66.089 0.51704 58.452 4.6618E-06
43 0.910 62.608 0.52605 55.765 4.7719E-06
44 0.850 61.428 0.54052 54.969 4.9416E-06
45 0.780 65.862 0.57150 58.980 5.2942E-06
46 0.625 81.247 0.62384 72.252 5.8672E-06
47 0.500 111.78 0.68368 97.674 6.5038E-06
48 0.400 156.80 0.73736 133.49 7.0642E-06
49 0.350 198.13 0.77403 164.84 7.4432E-06
50 0.320 225.07 0.80114 184.35 7.7218E-06
51 0.300 238.33 0.82589 193.20 7.7752E-06
52 0.280 236.98 0.68145 191.04 8.3380E-06
53 0.250 222.38 0.91109 180.28 8.8425-06
54 0.220 212.86 0.98454 175.58 9.8585E-06
55 0.180 224.93 1.0985 189.27 1.0707E-05
56 0.140 264.47 1.2605 225.68 1.2358E-05
57 0.100 315.69 1.4407 271.11 1.4159E-05
58 0.080 359.60 1.5890 309.48 1.5637E-05
59 0.067 398.60 1.7189 343.13 1.6930E-05
60 0.058 437.03 1.8462 376.44 1.8196E-05
61 0.050 482.50 1.9984 415.38 1.9709E-05
62 0.042 536.91 2.1822 461.67 2.1536E-05
63 0.035 592.86 2.3730 508.92 2.3430E-05
64 0.030 652.42 2.5769 558.88 2.5453E-05
65 0.025 730.59 2.8459 624.01 2.8121E-05
66 0.020 839.86 3.2255 714.61 3.1885E-05
67 0.015 1007.7 3.8121 853.53 3.7699E-05
68 0.010 1318.1 4.9044 111.15 4.8481E-05
69 0.005 221.59 8.0839 1863.2 8.0005E-05

Figure 4. Cross-section of U-235 & U-238 for BAPL-.1.

Table 5. Transport & total scattering cross-section from NJOY out of U-235 & U-238 in JEFF-3.1.2.

Energy group no. Thermal energy (eV) Transport cross-section in barns Total scattering cross-section barn
U-235 U-238 U-235 U-238
28 4.000 55.107 9.1126 10.970 8.0167
29 3.300 41.786 9.2503 11.066 8.2884
30 2.600 26.642 9.3123 11.319 8.3011
31 2.100 38.782 9.4295 11.839 8.5922
32 1.500 37.621 9.5217 11.405 8.0688
33 1.300 86.391 9.5762 11.381 7.8842
34 1.150 149.68 9.6143 5.6979 3.9807
35 1.123 144.30 9.6237 5.5339 3.9413
36 1.097 129.65 9.6329 5.5622 3.9685
37 1.071 114.36 9.6422 5.6450 4.0163
38 1.045 102.13 9.6512 5.5540 3.9455
39 1.020 93.480 9.6541 5.4849 3.8996
40 0.996 87.236 9.6428 5.5140 3.8899
41 0.972 82.877 9.6317 5.2317 3.6973
42 0.950 79.058 9.6361 7.7636 5.4683
43 0.910 75.664 9.6589 9.4923 6.6504
44 0.850 74.599 9.6911 10.201 7.1011
45 0.780 79.212 9.7591 12.097 8.3351
46 0.625 94.853 9.9317 12.212 8.2745
47 0.500 125.65 9.8992 12.283 8.1709
48 0.400 170.83 9.9640 11.161 7.3494
49 0.350 212.21 10.006 9.6537 6.3468
50 0.320 239.14 10.037 8.0837 5.3218
51 0.300 25.238 10.0065 8.2297 5.4283
52 0.280 25.102 10.103 10.095 6.6648
53 0.250 236.48 10.179 10.371 6.8318
54 0.220 227.06 10.276 11.596 7.6010
55 0.180 239.27 10.381 12.016 7.7886
56 0.140 278.95 10.538 12.487 8.0119
57 0.100 330.28 10.710 11.110 7.0761
58 0.080 374.23 10.843 9.8740 6.2623
59 0.067 413.25 10.958 8.5780 5.4284
60 0.058 451.67 11.056 8.3389 5.2650
61 0.050 497.09 11.152 8.6543 4.4503
62 0.042 551.57 11.370 8.5157 5.3526
63 0.035 607.68 11.643 7.4212 4.6604
64 0.030 667.32 11.886 7.8852 4.9441
65 0.025 745.52 12.159 8.3875 5.2502
66 0.020 854.78 12.521 8.9734 5.6058
67 0.015 1022.6 13.056 9.6824 6.0312
68 0.010 1333.1 14.183 10.749 6.6879
69 0.005 2231.1 17.547 12.002 7.4511

Figure 5. Cross-section of U-235 & U-238 for BAPL-2 lattice.

Figure 6. Cross-section of U-235 & U-238 for BAPL-3 lattice.

Table 6. keff comparison of TRX benchmark lattices.

Lattices JEFF-3.1.2 Experiment (CSEWG, 1986) Percentage of error
TRX-1 0.9853975 1.0000 1.46
TRX-2 0.9826511 1.0000 1.7

Table 7. keff comparison of BAPL benchmark lattices.

Lattices JEFF-3.1.2 Experiment (CSEWG, 1986) Percentage of error
BAPL-1 0.9828444 1.0000 1.7
BAPL-2 0.9849318 1.0000 1.5
BAPL-3 0.987897 1.0000 1.21

Table 8. Integral parameter comparison of TRX benchmark lattices.

Lattices Integral Parameter JEFF-3.1.2 Experiment (CSEWG, 1986) Percentage of error
TRX-1 ρ28 1.3466 1.3200 2
δ25 0.0958 0.0987 2.9
δ28 0.0949 0.0946 0.31
C* 0.78848 0.7970 1.06
TRX-2 ρ28 0.832 0.8370 0.59
δ25 0.05868 0.0614 4.4
δ25 0.0685 0.0693 1.1
C* 0.6322 0.6470 2.2

Table 9. Integral parameter comparison of BAPL benchmark lattices.

Lattices Integral parameters JEFF-3.1.2 Experiment (CSEWG, 1986) Percentage of error
BAPL-1 ρ28 1.4767 1.3900 6
δ25 0.0811 0.08400 3.4
δ28 0.7540 0.0780 3.3
C* 0.8250 …… ….
BAPL-2 ρ28 1.2101 1.1200 8.5
δ25 0.0661 0.0680 2.7
δ28 0.0649 0.0700 7.2
C* 0.7460 …. ….
BAPL-3 ρ28 0.9440 0.9606 1.7
δ25 0.0507 0.0520 2.3
δ28 0.0533 0.0570 6.4
C* 0.6616 ….

4. Discussion

Very recent, only the integral parameters of TRX and BAPL lattices of JEFF-3.1.1 have been compared and the values of keff are very close to the experiment [22]. In the present work NJOY output for the two isotopes U-235 and U-238 the group constants are consistent with each other. For each TRX and BAPL lattices the total absorption cross-section of U-235 is larger than that of U-238, epi-thermal fission cross-section of U-238 is marginal, but the thermal fission is completely absent. The captured cross-sections of U-235 are very low but thermal fission cross-section of U-235 is remarkably high. Captured cross-section of U-235 is very much lower than that of U-238. Moreover, the character cross-sections for each lattice are identical. From Tables 6 & Table 7, the calculated values of effective multiplication factor keff are very close to the experimental values, the maximum deviation is 1.7% for TRX-2 and BAPL-1 lattices. From Tables 8 & Table 9 it can be seen that the uncertainties of calculated values of the integral parameters of ρ28, δ25, δ28 and C*for TRX-1 & TRX-2 lattices do not deviate by more than 5% from the experimental values. Only the values of ρ28 in BAPL-1 and BAPL-2 lattices, values of δ28 in BAPL-2 and BAPL-3 show more than 5% inaccuracy but rest of the values of the integral parameters are very close to the benchmark values.

5. Conclusion

This theoretical study compares with the benchmark integral parameters of thermal reactor metallic uranium (TRX) and uranium oxide (BAPL) lattices with the nuclear data library JEFF-3.1.2 by using NJOY’99.0 and WIMSD-5B codes. The results from these computations are compared with experimental values by CSEWG, and it is found that there are no significant differences between calculated and experimental values. The cross-sections for epithermal and thermal neutrons at each lattice are practically identical. The integral parameters are almost equal to the experimental values, except for few values.

In the experimental result by CSEWG, the values of C* are present in TRX lattice but absent for BAPL lattice, therefore judgment for value of C* for BAPL benchmark lattice are not achievable. To conclude, this analysis provides a clear confirmation of the nuclear data library JEFF-3.1.2 for the neutronic calculation of TRIGA mark-II research reactor. Therefore JEFF-3.1.2 is completely reliable for safety analysis of the TRIGA reactor at AERE, Dhaka, Bangladesh.


References

  1. M. B. Chadwick, et al., "ENDF/B-VII.0: Next generation evaluated nuclear data library for nuclear science and technology", Nuclear Data Sheets, vol. 107, pp. 2931-3060, December, 2006.
  2. O. Allaoui, et al., "Validation of ENDF/B-VII.0 nuclear data library for shielding calculations using the Monte Carlo method", International Journal of Advanced Research, vol. 2, pp. 55-62, 2014.
  3. Cross Sections Evaluation Working Group, "ENDF-6 Formats Manual", BNL-90365-2009, Brookhaven National Laboratory, pp. 1-13, 2009.
  4. R. E. MacFarlane and D. W. Muir, "NJOY99.0: Code system for producing point-wise and multi-group neutron and photon cross sections from ENDF/B", RSICC Code Package PSR-480. Los Alamos National Laboratory, Los Alamos, New Mexico, USA, 1999.
  5. A. J. Koning, et al., "Status of the JEFF nuclear data library", Journal of the Korean Physical Society, vol. 59, pp. 1057-1062, August 2011.
  6. K. Shibata, et al., "JENDL-4.0: A new library for nuclear science and engineering", Journal of Nuclear Science and Technology, vol. 48, pp. 1–30, 2011.
  7. Z. Youxiang, L. Tingjin, Z. Jingshang and L. Ping, "CENDL-3- Chinese evaluated nuclear data library, version 3", Journal of Nuclear Science and Technology, vol. 39, pp. 37-39, 2002.
  8. ENDF-B/V-VI: The US Evaluated Nuclear Data Library, BNL-NCS-60496, Brookhaven National Laboratory, 1993.
  9. BROND-2. Library of recommended evaluated neutron data, VANT, Ser. Nucl. Const, N 2-3.13, 1991.
  10. A. J. Koning et al., "The JEFF evaluated nuclear data project", Proceedings of the International Conference on Nuclear Data for Science and Technology, ND2007, Nice, France, 22-27 April 2007.
  11. A. Santamarina, D. Bernard and Y. Rugama, "Validation Results from JEF-2.2 to JEFF-3.1.1", JEFF Report 22, 2009.
  12. J. R. Askew, F. J. Fayers and P. B. Kemshell, "A general description of the lattice code WIMS", Journal of the British Nuclear Energy Society, vol. 5, pp. 564, 1966.
  13. T. Kulikowska, "WIMSD-5B: A neutronic code for standard lattice physics analysis", Distributed by NEA Data Bank. Saclay, France, 1996.
  14. F. Leszczynski, "Description of Wims Library Update Project (WLUP)", 2002 International Meeting on Reduced Enrichment for Research and Test Reactors, Bariloche, Argentina, November 3-8, 2002.
  15. J. Hardy, Jr. D. Klein and J. J. Volpe, "A study of physics parameters in several water-moderated lattices of slightly enriched and natural uranium", Nuclear Science and Engineering, vol. 40, pp. 101-115, 1970.
  16. R. L. Hellens and G. A. Price, "Reactor physics data for water-moderated lattices of slightly enriched uranium", Reactor Technology Selected Reviews-1964, 529.
  17. J. R Brown, D. R. Harris, F. S. Frantz, J. J Volpe, J. C. Andrews and B. H. Noordhoff, Kinetics and buckling measurements in lattices of slightly enriched U or UO2 rods in H2O. WAPD-176, January, 1958.
  18. M. Halder and S. M. T. Islam, "Comparative study of generated wimsd-5b multi-group constants library based on JENDL-3.2 with JEFF-3.1.1 CENDL-3.0 and original WIMS and validation of generated library through some benchmark experiments analysis", IOSR Journal of Applied Physics, vol. 8, pp. 39-43, 2016.
  19. M. N. Uddin, M. M. Sarker, M. J. H. Khan and S. M. A. Islam, "Validation of CENDL and JEFF evaluated nuclear data files for TRIGA calculations through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors", Annals of Nuclear Energy, vol. 36, pp. 1521-1526, 2009.
  20. R. Sher and S. Fiarman, "Studies of Thermal Reactor Benchmark Data Interpretation: Experimental Corrections", EPRI NP-209, US, October 1976.
  21. Cross Section Evaluation Working Group (CSEWG), "Benchmark specifications with supplements", Brookhaven National Laboratory, National Nuclear Data Center, Upton, New York 11973, BNL-19302, II. ENDF-202, USA, November, 1986.
  22. K. Benaalilou, et al., "A comparative study of integral parameters for TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA MARK-II research reactor at CNESTEN using the cross-section ENDB-VII and JEFF3.1", International Journal of Current Research, vol. 6, pp. 4519-4523, 2014.

Article Tools
  Abstract
  PDF(1030K)
Follow on us
ADDRESS
Science Publishing Group
548 FASHION AVENUE
NEW YORK, NY 10018
U.S.A.
Tel: (001)347-688-8931