Comparison of UWCC MOX fuel measurements to MCNP-REN calculations
- Los Alamos National Lab., NM (United States)
The development of neutron coincidence counting has greatly improved the accuracy and versatility of neutron-based techniques to assay fissile materials. Today, the shift register analyzer connected to either a passive or active neutron detector is widely used by both domestic and international safeguards organizations. The continued development of these techniques and detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model, as it is currently used, fails to accurately predict detector response in highly multiplying mediums such as mixed-oxide (MOX) light water reactor fuel assemblies. For this reason, efforts have been made to modify the currently used Monte Carlo codes and to develop new analytical methods so that this model is not required to predict detector response. The authors describe their efforts to modify a widely used Monte Carlo code for this purpose and also compare calculational results with experimental measurements.
- OSTI ID:
- 298309
- Report Number(s):
- CONF-981106-; ISSN 0003-018X; TRN: 99:001940
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 79; Conference: American Nuclear Society winter meeting, Washington, DC (United States), 15-19 Nov 1998; Other Information: PBD: 1998
- Country of Publication:
- United States
- Language:
- English
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