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Title: Fuel System Compatibility Issues for Prometheus-1

Abstract

Compatibility issues for the Prometheus-1 fuel system have been reviewed based upon the selection of UO{sub 2} as the reference fuel material. In particular, the potential for limiting effects due to fuel- or fission product-component (cladding, liner, spring, etc) chemical interactions and clad-liner interactions have been evaluated. For UO{sub 2}-based fuels, fuel-component interactions are not expected to significantly limit performance. However, based upon the selection of component materials, there is a potential for degradation due to fission products. In particular, a chemical liner may be necessary for niobium, tantalum, zirconium, or silicon carbide-based systems. Multiple choices exist for the configuration of a chemical liner within the cladding; there is no clear solution that eliminates all concerns over the mechanical performance of a clad/liner system. A series of tests to evaluate the performance of candidate materials in contact with real and simulated fission products is outlined.

Authors:
; ;
Publication Date:
Research Org.:
Bettis Atomic Power Laboratory (BAPL), West Mifflin, PA
Sponsoring Org.:
USDOE
OSTI Identifier:
884677
Report Number(s):
B-MT(SRME)-55
TRN: US0603671
DOE Contract Number:
DE-AC12-00SN39357
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 42 ENGINEERING; COMPATIBILITY; CONFIGURATION; FISSION; FISSION PRODUCTS; FUEL SYSTEMS; LINERS; NIOBIUM; PERFORMANCE; SILICON; TANTALUM; ZIRCONIUM; NRPCT

Citation Formats

DC Noe, KB Gibbard, and MH Krohn. Fuel System Compatibility Issues for Prometheus-1. United States: N. p., 2006. Web. doi:10.2172/884677.
DC Noe, KB Gibbard, & MH Krohn. Fuel System Compatibility Issues for Prometheus-1. United States. doi:10.2172/884677.
DC Noe, KB Gibbard, and MH Krohn. Fri . "Fuel System Compatibility Issues for Prometheus-1". United States. doi:10.2172/884677. https://www.osti.gov/servlets/purl/884677.
@article{osti_884677,
title = {Fuel System Compatibility Issues for Prometheus-1},
author = {DC Noe and KB Gibbard and MH Krohn},
abstractNote = {Compatibility issues for the Prometheus-1 fuel system have been reviewed based upon the selection of UO{sub 2} as the reference fuel material. In particular, the potential for limiting effects due to fuel- or fission product-component (cladding, liner, spring, etc) chemical interactions and clad-liner interactions have been evaluated. For UO{sub 2}-based fuels, fuel-component interactions are not expected to significantly limit performance. However, based upon the selection of component materials, there is a potential for degradation due to fission products. In particular, a chemical liner may be necessary for niobium, tantalum, zirconium, or silicon carbide-based systems. Multiple choices exist for the configuration of a chemical liner within the cladding; there is no clear solution that eliminates all concerns over the mechanical performance of a clad/liner system. A series of tests to evaluate the performance of candidate materials in contact with real and simulated fission products is outlined.},
doi = {10.2172/884677},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Jan 20 00:00:00 EST 2006},
month = {Fri Jan 20 00:00:00 EST 2006}
}

Technical Report:

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  • This report contains a review of design studies for inertial confinement reactors. The first of three volumes briefly discusses the following: Introduction; Key objectives, requirements, and assumptions; Systems modeling and trade studies; Prometheus-L reactor plant design overview; Prometheus-H reactor plant design overview; Key technical issues and R&D requirements; Comparison of IFE designs; and study conclusions.
  • The stability of the SiC layer in the presence of free nitrogen will be dependent upon the operating temperatures and resulting nitrogen pressures whether it is at High Temperature Gas-Cooled Reactor (HTGR) temperatures of 1000-1400 C (coolant design dependent) or LWR temperatures that range from 500-700 C. Although nitrogen released in fissioning will form fission product nitrides, there will remain an overpressure of nitrogen of some magnitude. The nitrogen can be speculated to transport through the inner pyrolytic carbon layer and contact the SiC layer. The SiC layer may be envisioned to fail due to resulting nitridation at the elevatedmore » temperatures. However, it is believed that these issues are particularly avoided in the LWR application. Lower temperatures will result in significantly lower nitrogen pressures. Lower temperatures will also substantially reduce nitrogen diffusion rates through the layers and nitriding kinetics. Kinetics calculations were performed using an expression for nitriding silicon. In order to further address these concerns, experiments were run with surrogate fuel particles under simulated operating conditions to determine the resulting phase formation at 700 and 1400 C.« less
  • Environmental and endurance tests were conducted to evaluate the performance of typical fuel-system components when exposed to high-density aviation turbine engine fuel. The environment tests simulated the extreme high and low temperatures encountered in hot- and cold-day missions. The results revealed that the high-density fuel (HDF) would not have any fuel boiling or freezing problems but the pump power required for HDF was higher than for JP-4 fuel as was expected and the lower heat capacity of HDF resulted in noticeably higher heat-exchanger discharge temperatures. The endurance tests revealed that the HDF would not cause abnormal wear or component leakage.more » Nothing in the test results suggested that current inputs to fuel-system life-cycle-cost models should be modified if HDF is used.« less
  • Ni-base alloys were considered for the Prometheus space reactor pressure vessel with operational parameters of {approx}900 K for 15 years and fluences up to 160 x 10{sup 20} n/cm{sup 2} (E > 0.1 MeV). This paper reviews the effects of irradiation on the behavior of Ni-base alloys and shows that radiation-induced swelling and creep are minor considerations compared to significant embrittlement with neutron exposure. While the mechanism responsible for radiation-induced embrittlement is not fully understood, it is likely a combination of helium embrittlement and solute segregation that can be highly dependent on the alloy composition and exposure conditions. Transmutation calculationsmore » show that detrimental helium levels would be expected at the end of life for the inner safety rod vessel (thimble) and possibly the outer pressure vessel, primarily from high energy (E > 1 MeV) n,{alpha} reactions with {sup 58}Ni. Helium from {sup 10}B is significant only for the outer vessel due to the proximity of the outer vessel to the BeO control elements. Recommendations for further assessments of the material behavior and methods to minimize the effects of radiation damage through alloy design are provided.« less
  • Ni-base alloys were considered for the Prometheus space reactor pressure vessel with operational parameters of {approx}900 K for 15 years and fluences up to 160 x 10{sup 20} n/cm{sup 2} (E > 0.1 MeV). This paper reviews the effects of irradiation on the behavior of Ni-base alloys and shows that radiation-induced swelling and creep are minor considerations compared to significant embrittlement with neutron ,exposure. While the mechanism responsible for radiation-induced embrittlement is not fully understood, it is likely a combination of helium embrittlement and solute segregation that can be highly dependent on the alloy composition and exposure conditions. Transmutation calculationsmore » show that detrimental helium levels would be expected at the end of life for the inner safety rod vessel (thimble) and possibly the outer pressure vessel, primarily from high energy (E > 1 MeV) n,{alpha} reactions with {sup 58}Ni. Helium from {sup 10}B is significant only for the outer vessel due to the proximity of the outer vessel to the Be0 control elements. Recommendations for further assessments of the material behavior and methods to minimize the effects of radiation damage through alloy design are provided.« less