TY - GEN
T1 - Development of INSPCT-S for inspection of spent fuel pool
AU - Walters, William
AU - Haghighat, Alireza
AU - Sitaraman, Shivakumar
AU - Ham, Young
PY - 2012
Y1 - 2012
N2 - In this paper, we discuss an accurate and fast software tool (INSPCT-S, Inspection of Nuclear Spent fuel-Pool Calculation Tool, version Spreadsheet) developed for calculation of the response of fission chambers placed in a spent fuel pool, such as Atucha-I. INSPCT-S is developed for identification of suspicious regions of the pool that may have missing or substitute assemblies. INSPCT-S uses a hybrid algorithm based on the adjoint function methodology. The neutron source is comprised of spontaneous fission, (α, n) interactions, and subcritical multiplication. The former is evaluated using the ORIGEN-ARP code, and the latter is obtained with the fission matrix (FM) formulation. The FM coefficients are determined using the MCNP Monte Carlo code, and the importance function is determined using the PENTRAN 3-D parallel Sn code. Three databases for the neutron source, FM elements, and adjoint flux are prepared as functions of different parameters including burnup, cooling time, enrichment, and pool lattice size. INSPCT-S uses the aforementioned databases and systems of equations to calculate detector responses, which are subsequently compared with normalized experimental data.
AB - In this paper, we discuss an accurate and fast software tool (INSPCT-S, Inspection of Nuclear Spent fuel-Pool Calculation Tool, version Spreadsheet) developed for calculation of the response of fission chambers placed in a spent fuel pool, such as Atucha-I. INSPCT-S is developed for identification of suspicious regions of the pool that may have missing or substitute assemblies. INSPCT-S uses a hybrid algorithm based on the adjoint function methodology. The neutron source is comprised of spontaneous fission, (α, n) interactions, and subcritical multiplication. The former is evaluated using the ORIGEN-ARP code, and the latter is obtained with the fission matrix (FM) formulation. The FM coefficients are determined using the MCNP Monte Carlo code, and the importance function is determined using the PENTRAN 3-D parallel Sn code. Three databases for the neutron source, FM elements, and adjoint flux are prepared as functions of different parameters including burnup, cooling time, enrichment, and pool lattice size. INSPCT-S uses the aforementioned databases and systems of equations to calculate detector responses, which are subsequently compared with normalized experimental data.
UR - http://www.scopus.com/inward/record.url?scp=84867905128&partnerID=8YFLogxK
UR - http://www.scopus.com/inward/citedby.url?scp=84867905128&partnerID=8YFLogxK
U2 - 10.1520/JAI104070
DO - 10.1520/JAI104070
M3 - Conference contribution
AN - SCOPUS:84867905128
SN - 9780803175365
T3 - ASTM Special Technical Publication
SP - 690
EP - 705
BT - Reactor Dosimetry
PB - ASTM International
T2 - 14th International Symposium on Reactor Dosimetry
Y2 - 22 May 2011 through 27 May 2011
ER -