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DOE OSTI2021

Dr. Cecil Lubitz (1925–2021)

Dr. Cecil Lubitz passed away on March 5, 2021 at his home in upstate New York. A graduate of the Class of 1945 from the Naval Academy, Cecil served in the US Navy until 1947 after which he earned a MS in electrical engineering (1949) and a PhD in Physics (1960) from the University of Michigan. Following graduate school, Cecil came to work at Knolls Atomic Power Laboratory (KAPL) on applied nuclear physics in the Naval Nuclear Propulsion Program (NNPP). Cecil worked at KAPL for 54 years and had an impactful career with many substantial contributions to the NNPP, national and global nuclear data communities.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

DOE OSTI2017

Criticality Scoping Calculations for a Multi-Function Dissolver Insert

As one of its missions, the Savannah River Site seeks to be a leader in the safe and productive disposition of special nuclear materials. One of the premier programs is the disposition of used nuclear fuel, primarily from research reactors, through dissolution, purification, and down blend into low enriched uranium for commercial power reactors. Prior to use, this fuel contained highly enriched uranium (HEU). In an effort to advance this expertise, this study has been performed to demonstrate the feasibility to disposition a number of special geometry HEU fuels. These fuels come from a number of different domestic and foreign research reactors which had geometries considerably different from round or square bundles of plates or rods. As such, they cannot be dispositioned in current H-Canyon dissolver inserts so a new, multi-function dissolver insert may be considered. While this is not a current mission, it is believed that this study's scoping calculations demonstrate the feasibility of the site taking such a mission. At the H-Canyon facility, a dissolver is a large stainless steel pot, loaded with a nitric acid solution into which uranium-alloy fuel is dissolved. The dissolver uses a device called an insert to feed the fuels to the acid solution in a controlled manner. Most fuel forms fit into one of the existing inserts which include one with ten {approx}5-inch diameter circular wells, one with four {approx}4.5 inch by {approx}15 inch slab wells, and one specifically designed for the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR). This study considered five bounding fuel candidates for scoping the potential new insert. The fuels include a generic annular plate fuel (APF), the Tower Shield Reactor (TSR) fuel elements, the Reactor A Haute Flux (RHF) core, Knolls Atomic Power Laboratory (KAPL) reactor fuel, and SLOWPOKE cores. These are expected to bound a range of proposed fuel candidates. Scoping studies were made to establish geometric restrictions for the new dissolver insert and to determine if these fuels could safely be dissolved. Homogenous models were used for the scoping study due to time constraints. Computational modeling was performed using KENOVI in the SCALE 6.1 code package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Measurements on the Diffusion Length of Thermal Neutrons in Water from 25 to 296°C

Here, the diffusion length of thermal neutrons has been measured in water over the range 25.9 to 295.3°C in the pressure vessel of the KAPL High Temperature Critical Assembly. The diffusion length was determined by fitting an exponential to the data found by activating manganese foils with neutrons from a small Sb-Be source. The temperature variation of the diffusion length could be fairly well represented by either of two simple approximations—either that the transport cross section of water has a 1/υ behavior, or that the transport cross section is determined from the Radkowsky prescription.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

VERA Core Physics Benchmark Progression Problem Results for VERA 3.6

This report contains a comparison of results obtained with VERA 3.6 Release Candidate 0 for the Core Physics Benchmark Progression Problems 1 through 9 to both computational tools and measured data from Watts Bar Unit 1 (WBN1). (Ref. 1 and Ref. 2) These problems progress from single pin cell analyses to full core depletion calculations. VERA results obtained for Problems 1 through 5 are compared against results obtained using the SCALE 6.2 Beta code KENO-VI. Results for Problem 6 and 7 are compared against those obtained using a coupled neutronic and thermal-hydraulic combination of the KAPL software MC21 and the CASL tool CTF. (Ref. 3, Ref. 4 and Ref. 5) VERA results from problems 5 and 9 are compared to data from WBN1. For all of the analyses contained in this report, VERA provided similar results to other state-of-the-art software programs and real reactor data. This indicates that the program is capable of modeling a full-scale PWR in a more discretized and computationally efficient manner than any other reactor physics program. Since the descriptions of the Core Physics Benchmark Progression Problems and the logic behind their selection is discussed in-depth elsewhere it is not covered in the scope of this document. Additionally, in-depth discussions of the software programs used in this analysis are not provided for the same rationale.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗