# Criticality and Safety Parameter Studies of a 3-MW TRIGA MARK-II Research Reactor and Validation of the Generated Cross-Section Library and Computational Method

## Abstract

This study deals with the analysis of some neutronics and safety parameters of the current core of a 3-MW TRIGA MARK-II research reactor and validation of the generated macroscopic cross-section library and calculational techniques by benchmarking with experimental, operational, and available Safety Analysis Report (SAR) values. The overall strategy is: (a) generation of the problem-dependent cross-section library from basic Evaluated Nuclear Data Files such as ENDF/B-VI and JENDL-3.2 with NJOY94.10+, (b) use of the WIMSD-5 package to generate a few-group neutron macroscopic cross section for all of the materials in the core and its immediate neighborhood, (c) use the three-dimensional CITATION code to perform the global analysis of the core, and (d) checking of the validity of the CITATION diffusion code with the MCNP4B2 Monte Carlo code. The ultimate objective is to establish methods for reshuffling the current core configuration to upgrade the thermal flux at irradiation locations for increased isotope production. The computational methods, tools and techniques, customization of cross-section libraries, various models for cells and supercells, and many associated utilities are standardized and established/validated for the overall neutronic analysis. The excess reactivity, neutron flux, power distribution, power peaking factors, determination of the hot spot, and fuel temperature reactivitymore »

- Authors:

- Publication Date:

- OSTI Identifier:
- 20822156

- Resource Type:
- Journal Article

- Journal Name:
- Nuclear Technology

- Additional Journal Information:
- Journal Volume: 130; Journal Issue: 2; Other Information: Copyright (c) 2006 American Nuclear Society (ANS), United States, All rights reserved. http://epubs.ans.org/; Country of input: International Atomic Energy Agency (IAEA); Journal ID: ISSN 0029-5450

- Country of Publication:
- United States

- Language:
- English

- Subject:
- 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; CRITICALITY; CROSS SECTIONS; ENRICHED URANIUM; GLOBAL ANALYSIS; MONTE CARLO METHOD; NEUTRON FLUX; NEUTRONS; NUCLEAR DATA COLLECTIONS; NUCLEAR FUELS; POWER DENSITY; POWER DISTRIBUTION; REACTIVITY COEFFICIENTS; RESEARCH REACTORS; SAFETY; SAFETY ANALYSIS; STEADY-STATE CONDITIONS; THERMAL HYDRAULICS; VALIDATION

### Citation Formats

```
Bhuiyan, S.I., Mondal, M.A.W., Sarker, M.M., Rahman, M., Shahdatullah, M.S., Huda, M.Q., Chakrobortty, T.K., and Khan, M.J.H.
```*Criticality and Safety Parameter Studies of a 3-MW TRIGA MARK-II Research Reactor and Validation of the Generated Cross-Section Library and Computational Method*. United States: N. p., 2000.
Web.

```
Bhuiyan, S.I., Mondal, M.A.W., Sarker, M.M., Rahman, M., Shahdatullah, M.S., Huda, M.Q., Chakrobortty, T.K., & Khan, M.J.H.
```*Criticality and Safety Parameter Studies of a 3-MW TRIGA MARK-II Research Reactor and Validation of the Generated Cross-Section Library and Computational Method*. United States.

```
Bhuiyan, S.I., Mondal, M.A.W., Sarker, M.M., Rahman, M., Shahdatullah, M.S., Huda, M.Q., Chakrobortty, T.K., and Khan, M.J.H. Mon .
"Criticality and Safety Parameter Studies of a 3-MW TRIGA MARK-II Research Reactor and Validation of the Generated Cross-Section Library and Computational Method". United States.
```

```
@article{osti_20822156,
```

title = {Criticality and Safety Parameter Studies of a 3-MW TRIGA MARK-II Research Reactor and Validation of the Generated Cross-Section Library and Computational Method},

author = {Bhuiyan, S.I. and Mondal, M.A.W. and Sarker, M.M. and Rahman, M. and Shahdatullah, M.S. and Huda, M.Q. and Chakrobortty, T.K. and Khan, M.J.H},

abstractNote = {This study deals with the analysis of some neutronics and safety parameters of the current core of a 3-MW TRIGA MARK-II research reactor and validation of the generated macroscopic cross-section library and calculational techniques by benchmarking with experimental, operational, and available Safety Analysis Report (SAR) values. The overall strategy is: (a) generation of the problem-dependent cross-section library from basic Evaluated Nuclear Data Files such as ENDF/B-VI and JENDL-3.2 with NJOY94.10+, (b) use of the WIMSD-5 package to generate a few-group neutron macroscopic cross section for all of the materials in the core and its immediate neighborhood, (c) use the three-dimensional CITATION code to perform the global analysis of the core, and (d) checking of the validity of the CITATION diffusion code with the MCNP4B2 Monte Carlo code. The ultimate objective is to establish methods for reshuffling the current core configuration to upgrade the thermal flux at irradiation locations for increased isotope production. The computational methods, tools and techniques, customization of cross-section libraries, various models for cells and supercells, and many associated utilities are standardized and established/validated for the overall neutronic analysis. The excess reactivity, neutron flux, power distribution, power peaking factors, determination of the hot spot, and fuel temperature reactivity coefficients {alpha}{sub f} in the temperature range of 45 to 1000 deg. C are studied. All the analyses are performed using the 4- and 7-group libraries of the macroscopic cross sections generated from the 69-group WIMSD-5 library. The 7-group calculations yield comparatively better agreement with the experimental value of k{sub eff} and the other core parameters. The CITATION test runs using different cross-section sets based on the different models applied in the WIMSD-5 calculations show a strong influence of those models on the final integral parameter. Some of the cells are specially treated with the Prize options available in WIMSD-5 to take into account the fine structure of the flux gradient in the fuel-reflector interface region. The hot spot is found physically at the fuel position C4 with a maximum power density of 1.044559 x 10{sup 2} W/cm{sup 3}. The calculated total peaking factor is 5.8867 compared to the original SAR value of 5.6325. The curve of {alpha}{sub f} with the temperature at zero burnup shows that the curve deviates somewhat with that reported in the original SAR for low-enriched uranium fuel. The MCNP calculations establish that the CITATION calculations and the generated cross-section library are reasonably good for neutronic analysis of TRIGA reactors. The results obtained from the neutronic analysis will be used to analyze the thermal-hydraulic behavior and the safety margins of the core both for steady-state and pulse-mode operations.},

doi = {},

journal = {Nuclear Technology},

issn = {0029-5450},

number = 2,

volume = 130,

place = {United States},

year = {2000},

month = {5}

}