Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

Journal Article · Mon Sep 29 00:00:00 EDT 2014 · Annals of Nuclear Energy

DOI:https://doi.org/10.1016/j.anucene.2014.09.005· OSTI ID:1185510

George, Nathan Michael ^[1]; Terrani, Kurt A. ^[2]; Powers, Jeffrey J. ^[2]; Worrall, Andrew ^[2]; Maldonado, Ivan ^[1]

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1. Univ. of Tennessee, Knoxville, TN (United States)
2. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched ²³⁵U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the harder neutron spectrum in the system, causing more ²³⁹Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Furthermore, implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased ²³⁵U enrichment and the additional UO₂ pellet volume enabled by using thinner cladding.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• Grant/Contract Number: AC05-00OR22725
• OSTI ID: 1185510
• Alternate ID(s): OSTI ID: 1247582
• Journal Information: Annals of Nuclear Energy, Vol. 75; ISSN 0306-4549
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
nuclear
accident tolerant fuel
ATF
reactor physics
cladding
lattice physics
FeCrAl
SiC
SS304
SS310
alternate cladding
PWR