Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

Abstract

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HP-ENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four hightemperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the preferred working fluid and the HP working temperature can be as high as 1300 K. It is feasible to achieve criticality and to maintain a nearly zero burn-up reactivity swing for at least 20 EFPY with an average linear heat generation rate (LHR) of 90W/cm. The preferred design utilizes nitride fuel made of natural nitrogen and loaded with depleted uranium and TRU from LWR spent fuel cooled for approximately 30 years. The preferred intermediate coolant is LiF-BeF ; its average outlet temperature is ~ 1040K. Effective heat transfer to the intermediate coolant is obtained with HPs extending out of the core less than 50 cm. The required reactor vessel height is significantly smaller than that of the reference ENHS: 9 vs. ~20 m. The vessel diameter is slightly larger: 4 vs. ~ 3.5 m. In conclusion, it appears feasible to design a HP-ENHS reactor to achieve its primary design objectives. The resulting HP-ENHS reactor concept is unique in offering sustainable proliferation-resistant nuclear energy that can be delivered at very high temperatures. A number of outsta...

Numerical and Experimental Studies on Small/Micro Nuclear Reactors

“…To improve the lifetime of the reactor, the KAIST proposed an 18 MW th H-MMR with UN fuel, which provides continuous operation for about 56 years in the case of 12.1% 235 U enrichment. In addition, Ehud (2008) proposed a 125 MWth HP-ENHS core, which can achieve at least 20 years of core lifetime by loading the depleted uranium and TRUs.…”

Review of reactor conceptual design and thermal hydraulic characteristics for heat pipe in nuclear systems

“…The reactor incorporated HPs as part of its design to enhance the passive safety mechanisms. These passive safety measures were intended to minimize the reliance on active systems and human intervention during abnormal events, ensuring a high safety level in the reactor operation (Greenspan, 2008). The Special Purpose Reactor (SPR) conceptualized by LANL was an innovative small nuclear reactor (SMR) that offers several advantages.…”

Analysis of Th and U breeding in a heat pipe cooled traveling wave reactor

“…To improve the lifetime of the reactor, the KAIST (Choi et al, 2020) proposed an 18 MW th H-MMR with UN fuel, which provides continuous operation for about 56 years in the case of 12.1% 235 U enrichment. In addition, Ehud (2008) proposed a 125 MWth HP-ENHS core, which can achieve at least 20 years of core lifetime by loading the depleted uranium and TRUs.…”