Reevaluation of fuel enthalpy in NSRR test for high burnup fuels

This paper describes the recent procedure of evaluation of the fuel enthalpy in the reactivity initiated accident (RIA) simulating tests performed at the nuclear safety research reactor (NSRR), and reports some important updates of the fuel enthalpies in the tests with high burnup PWR fuels. Previously, the fuel enthalpy had been evaluated by the procedure based on the short-life fission product measurement, i.e. a pellet slice was sampled from the test fuel rod after the NSRR test, a chemical separation process was applied to the solution of the pellet slice to separate barium, and the amount of Ba-140 was measured by gamma spectrometry on the separated barium. But a part of the results showed significant scattering even within the similar tests with similar fuels, which should have showed similar fuel enthalpies. The scattering appears to indicate the difficulty in treatment of the short-life nuclides after the completion of the NSRR test and unsuccessful measurement of the amount of fuel dissolved in the specimen preparation. Another difficulty of the procedure is that it is not repeatable for a specimen and so double check of an evaluation is not possible. Hence, an alternative procedure, which is based on the total amount of fissile materials evaluated by mass analysis, was developed and has been applied for the tests after 2003; the amount of fissile materials is input to a well-verified neutron transport calculation model for the NSRR reactor core to calculate a coupling factor of power densities between the test fuel rod and the NSRR driver fuel rods. This procedure does not require quickness and is repeatable, so it is applicable even many years later if the fuel sample is available. The recent procedure was thus applied to the tests before 2003, whose burnups are below 60 GWd/tU. It was shown that the fuel enthalpy had been significantly underestimated in the tests with high burnup PWR fuels: the test series HBO and TK. In this paper, the procedure adopted in the present reevaluation is described and the result is discussed; how the correction made can affect the PCMI failure limit and interpretation of other fuel behaviors, derived from the NSRR test results, such as cladding temperature, cladding deformation, and fission gas release under RIA conditions. (author)