Tritium Production and Permeation in High-Temperature Reactor Systems

Tritium (H13) is a radioactive isotope of hydrogen formed by ternary fission events (rare emissions of three nuclides rather than two during a fission) and neutron absorption (and subsequent decay) of predecessor radionuclides, particularly 6Li and 7Li. Also in fusion, the concept of breeding tritium during the fusion reaction is of significance for the future needs of a large-scale fusion power plant. Tritium is of special interest among the fission products created in next-generation nuclear reactors such as gas cooled reactors and molten salt reactors, because of the large quantities produced when compared with conventional light-water reactors (LWR) and the higher temperatures of operation for these systems enhances permeation. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for mitigation of permeation of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented along with mitigation strategies for tritium permeation.

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Nondestructive, Energy-Dispersive, X-Ray Fluorescence an Analysis of Actinide Stream Concentrations from Reprocessed Nuclear Fuel

Advances in X-ray Analysis ◽

10.1154/s0376030800006273 ◽

1979 ◽

Vol 23 ◽

pp. 163-176

Author(s):

D. C. Camp ◽

W. D. Ruhter

Keyword(s):

Nuclear Fuel ◽

Nitrate Solution ◽

Fission Products ◽

Analytical Techniques ◽

Fluorescence Analysis ◽

Energy Dispersive ◽

Light Water Reactors ◽

Spent Fuel ◽

X Ray ◽

Water Reactors

In the event that nuclear fuel from light water reactors (LWR) is reprocessed to reclaim the uranium or plutonium, several analytical techniques will be used for product accountability. Generally, the isotopic content of both the plutonium and uranium in the reprocessed product will have to be accurately determined. One plan for the reprocessing of LWR spent fuel incorporates the following scheme. After separation from both the fission products and transplutonium actinides (including neptunium and americium), part of the uranium and all of the plutonium in a nitrate solution will merge together to form a coprocessed stream. This solution will be concentrated by evaporation and sent to a hold tank for accountability. Input concentrations into the hold tank could be up to 350 g U/ℓ and nearly 50 g Pu/ℓ. The variation to be expected in these concentrations is not known. The remaining uranium fraction will be further purified and sent to a separate storage tank. Its expected stream concentration will be about 60 g U/ℓ. These two relatively high actinide stream concentrations can be monitored rapidly, quantitatively, and nondestructively using the technique of energy-dispersive x-ray fluorescence analysis(XRFA).

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Regulatory Viewpoint on Innovative VHTR Development in Korea

Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 2 ◽

10.1115/htr2008-58058 ◽

2008 ◽

Cited By ~ 2

Author(s):

Hun-Joo Lee ◽

Sang-Kyu Ahn ◽

Kju-Myeng Oh ◽

Chang-Ju Lee

Keyword(s):

Research Effort ◽

Policy Statement ◽

Light Water Reactors ◽

Event Frequency ◽

Very High Temperature Reactor ◽

Regulatory Approach ◽

High Temperature Reactor ◽

Water Reactors ◽

Gen Iv ◽

Very High

This paper addresses that major changes in the safety approach, for instance the increased use of Probabilistic Risk Assessment (PRA), have been made. All commercial reactors in operation today belong to the Generations II and III. Generation IV International Forum (GIF) has launched several programs aimed at developing the next generation of nuclear energy systems. Part of the research effort is focused on new reactor concepts, such as the Very High Temperature Reactor (VHTR), currently developing in Korea. In parallel to the design process of VHTR currently underway, regulatory approach is moving forward to define new licensing rules. So, Korea Institute of Nuclear Safety (KINS) is defining, as a goal to risk-inform, the regulation and developing the regulatory framework and licensing process more efficient, predictable, and stable. However, the licensing of NPPs has focused until now on Light Water Reactors (LWRs) and has not incorporated systematically insights and benefits from PRA. In the meantime, USNRC and IAEA have recently drafted a risk-informed regulation and technology-neutral framework (TNF) for new plant licensing along with the innovative Gen-IV system design. KINS also expects that advanced NPPs will show enhanced margins of safety, and that advanced reactor designs will comply with the national safety goal policy statement. In order to meet these expectations, PRA tools are currently being considered by KINS; among them are frequency-consequence (F-C) curves, which plot the frequency of having Consequence. This paper discusses the role and the usefulness of such curves in risk-informing the licensing process in Korea, and shows that the use of F-C curve allows the implementation of both structural and rational Defence-In-Depth (DID). This paper focuses on F-C curves as means to assess the licensing basis events (LBEs) from the regulatory viewpoint on the innovative small and medium reactor (SMR) sized VHTR deployment in Korea. The principle underlying the F-C curve is that event frequency and dose are inversely related, i.e., the higher the dose consequences, the lower is the allowed event frequency.

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Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

Science and Technology of Nuclear Installations ◽

10.1155/2017/1450648 ◽

2017 ◽

Vol 2017 ◽

pp. 1-25 ◽

Cited By ~ 1

Author(s):

Bruno Gonfiotti ◽

Sandro Paci

Keyword(s):

Fission Products ◽

Severe Accident ◽

Sensitivity Analyses ◽

Light Water Reactors ◽

Control Rod ◽

Light Water ◽

Severe Accidents ◽

Fuel Bundle ◽

Water Reactors

The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0) employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

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Computational Study for Inventory Estimation of Se-79, Tc-99, Sn-126, and Cs-135 in High-Level Radioactive Wastes From Spent Nuclear Fuels of Light Water Reactors

ASME 2011 14th International Conference on Environmental Remediation and Radioactive Waste Management, Parts A and B ◽

10.1115/icem2011-59087 ◽

2011 ◽

Author(s):

Keisuke Okumura ◽

Shiho Asai ◽

Yukiko Hanzawa ◽

Tsutomu Okamoto ◽

Hideya Suzuki ◽

...

Keyword(s):

Parametric Study ◽

Computational Study ◽

Estimation Method ◽

Fission Products ◽

Radioactive Wastes ◽

Light Water Reactors ◽

Nuclear Fuels ◽

Light Water ◽

Water Reactors ◽

High Level

Inventory estimation of long-lived fission products (LLFPs) in high-level radioactive wastes (HLW) from spent nuclear fuels of light water reactors is important for a safety assessment of their disposal. In order to develop an inventory estimation method of difficult-to-measure LLFPs (Se-79, Tc-99, Sn-126, and Cs-135), a parametric study was carried out by using a sophisticated burnup calculation code and data. In the parametric study, fuel specifications and irradiation conditions are changed in the conceivable range. The considered parameters are fuel assembly types (PWR / BWR), U-235 enrichment, moderator temperature, void fraction, power density, and so on. From the calculated results, we clarify the burnup characteristics of the target LLFPs and their possible ranges of generations. Finally, candidates of the key nuclide are proposed for the scaling factor method of HLW.

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Escape of fission products from defective fuel rods of light water reactors

Nuclear Engineering and Design ◽

10.1016/0029-5493(81)90033-9 ◽

1981 ◽

Vol 64 (1) ◽

pp. 81-85 ◽

Cited By ~ 19

Author(s):

E. Schuster ◽

F. Garzarolli ◽

A. Kersting ◽

K.H. Neeb ◽

H. Stehle

Keyword(s):

Fission Products ◽

Light Water Reactors ◽

Fuel Rods ◽

Light Water ◽

Water Reactors

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Analysis of Tritium Production and Distribution in HTR-PM

18th International Conference on Nuclear Engineering: Volume 3 ◽

10.1115/icone18-30147 ◽

2010 ◽

Author(s):

Yuanyu Wu ◽

Jianzhu Cao

Keyword(s):

Heat Exchanger ◽

Tritium Concentration ◽

Ternary Fission ◽

Tritium Production ◽

Limit Value ◽

Secondary Loop ◽

Production And Distribution ◽

Calculation Results ◽

Main Production ◽

Tritium Permeation

The research on radiological impact of tritium is highly concerned in high temperature gas-cooled reactors. In order to better assess the environmental performance of HTR-PM (HTR demonstration project with 2 × 250MW plants), analysis of tritium behavior in HTR-PM is conducted in this paper. The main production sources of tritium are the ternary fission in the fuel and neutron capture reactions of some nuclides. Based on the tight interactions between tritium sources and sinks, differential equations are built to describe tritium behavior in primary and secondary loop. Specific analysis is conducted to tritium permeation through heat exchanger walls to secondary loop, considering the oxidation of alloys used for heat exchanger. Applied with the parameters of HTR-PM, tritium concentration in primary and secondary loop is calculated, and the amount of tritium released to the environment is evaluated. The evaluation shows that the amount of tritium released to the environment is less than the limit value prescribed by Chinese regulation on radiation protection. The calculation results can also be applied to the safety analysis and used to guide the design of relevant systems and equipments for the HTR-PM.

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Fuel Survey in the Light Water Reactors Based on the Activity of the Fission Products

Nuclear Technology ◽

10.13182/nt80-a17690 ◽

1980 ◽

Vol 49 (3) ◽

pp. 426-434 ◽

Cited By ~ 27

Author(s):

R. Beraha ◽

G. Beuken ◽

G. Frejaville ◽

C. Leuthrot ◽

Y. Musante

Keyword(s):

Fission Products ◽

Light Water Reactors ◽

Light Water ◽

Water Reactors

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Hydrogen Diffusion in FeCrAl Alloys for Light Water Reactors Cladding Applications

Volume 6B: Materials and Fabrication ◽

10.1115/pvp2016-63164 ◽

2016 ◽

Cited By ~ 1

Author(s):

Raul B. Rebak ◽

Young-Jin Kim

Keyword(s):

Hydrogen Diffusion ◽

Zirconium Alloys ◽

The United States ◽

Department Of Energy ◽

Light Water Reactors ◽

Alumina Layer ◽

Fecral Alloys ◽

Water Reactors ◽

Tritium Permeation ◽

Storage Pools

There is a worldwide effort to develop nuclear fuels that are resistant to accidents such as loss of coolant in the reactor and the storage pools. In the United States, the Department of Energy is teaming with fuel vendors to develop accident tolerant fuels (ATF), which will resist the lack of cooling for longer periods of times than the current zirconium alloy - uranium dioxide system. General Electric (GE) and its partners is proposing to replace zirconium alloys cladding with an Iron-Chromium-Aluminum (FeCrAl) alloy such as APMT, since they are highly resistant to attack by steam up to the melting point of the alloy. FeCrAl alloys do not react with hydrogen to form stable hydrides as zirconium alloys do. Therefore, it is possible that more tritium may be released to the coolant with the use of FeCrAl cladding. This work discusses the formation of an alumina layer on the surface of APMT cladding as an effective barrier for tritium permeation from the fuel to the coolant across the cladding wall.

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The CHF Enhancement Combining Nano-Fluid and Honeycomb Porous Plate in a Saturate Pool Boiling

Volume 2A: Thermal Hydraulics ◽

10.1115/icone22-30199 ◽

2014 ◽

Author(s):

Shoji Mori ◽

Suazlan Mt Aznam ◽

Kunito Okuyama

Keyword(s):

Large Scale ◽

Heated Surface ◽

Porous Plate ◽

Severe Accident ◽

Light Water Reactors ◽

Nanoparticle Deposition ◽

Large Scale Systems ◽

Nano Fluid ◽

Surface Diameter ◽

Water Reactors

Several studies have proposed the use of nanofluids to enhance the in-vessel retention (IVR) capability in the severe accident management strategy implemented at certain light-water reactors. Systems using nanofluids for IVR must be applicable to large-scale systems, i.e., infinite heated surfaces. However, the effect of the size of heater with nanoparticle deposition was revealed that the CHF is decreased with the increased heater size. On the other hand, the CHF using a honeycomb porous plate was shown experimentally to be more than twice that of a plain surface with a heated surface diameter of 30 mm, which is comparatively large compared to 10 mm. This enhancement is resulted from the capillary supply of liquid onto the heated surface and the release of vapor generated through the channels. In the present paper, in order to enhance the CHF of a large heated surface, the effects of a honeycomb porous plate and a nanofluid on the CHF were investigated experimentally. As a result, the CHF was enhanced greatly by the attachment of a honeycomb porous plate to the modified heated surface by nanoparticle deposition, even in the case of a large heated surface.

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