Arnold H. Fero

Neutron and Gamma Ray Flux Calculations for the Venus PWR Engineering Mockup

This paper describes the analysis of neutron and gamma ray fluxes in the VENUS PWR engineering mockup. This mockup experiment is unique in two ways. It is the first mockup to correctly represent the heterogeneities which exist in the PWR core peripheral fuel assemblies, core baffle, core barrel, and neutron pad. This is accomplished by using a representative PWR fuel assembly geometry (15 x 15) and full thickness Type 304 stainless steel reactor internals structures located with representative water gap spacing. The VENUS mockup also represents locally the stairstep geometry of the core periphery. Second, the VENUS mockup is extremely well characterized in terms of “as-built” dimensions, material compositions, and pin-by-pin core power distributions.

Ex-Vessel Neutron Dosimetry Analysis for Westinghouse 4-Loop XL Pressurized Water Reactor Plant Using 3D Parallel Discrete Ordinates Code RAPTOR-M3G

Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the beltline region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locations and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure.

Dosimetry analyses of the Ringhals 3 and 4 reactor pressure vessels

A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the 93Nb (n,n′) 93mNb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium.