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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Direct Processing of Lithium Chloride Based Waste Salt in a Ceramic Waste Form

Ceramic waste form materials were fabricated with a LiCl based salt mixture to demonstrate the suitability of direct processing of waste salt from an all LiCl pyroprocessing flow sheet on the formation, microstructure, and durability of ceramic waste forms. Materials were successfully synthesized by using the simplified direct processing technique with salt loadings of 5, 7.5, and 10 wt% at the laboratory scale. Materials fabricated with LiCl based salt indicated full conversion of zeolite to sodalite occurred to form products with low open porosity. This indicates that the Na 2 O content of NBS4 glass is suitable for processing salt chemistries which do not contain NaCl. Generated sodalite was microencapsulated by the glass binder, halite occlusions were detected within sodalite domains at all salt loadings, and a Cs-rich phase believed to be Cs-pollucite was detected in the material made with the lowest salt loading (5 wt%). Salt inclusions were encapsulated within the binder glass at all loadings, and found to consist primarily of NaCl with Cs present. A small amount of salt was found at the surface of the wasteform made with the highest salt loading of 10 wt%, which may indicate an upper limit to the amount of salt that can be accommodated or incomplete mechanical mixing of the reagents. Durability testing by using the ASTM C1308 method indicated that initial rapid dissolution of exposed halite inclusion phases was directly correlated to salt loading. Wasteform degradation during all tests occurred primarily due to dissolution of the sodalite phase. Durability of the binder glass was found to increase as the waste salt loading increased. Results for the release of cations from the salt indicate the changes in wasteform durability and halite inclusion chemistry are strongly influenced by ion exchange phenomena between the salt and the glass. This appears to be the result of the relatively stronger affinity for the glass phase of Li, compared to that of other cations such as Na and Cs. The effects from cation exchange were diminished at lower salt loadings.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Effects of gaseous NH{sub 3} and SO{sub 2} on the concentration profiles of PCDD/F in flyash under post-combustion zone conditions

Highlights: Black-Right-Pointing-Pointer Influence of NH{sub 3} and SO{sub 2} on 2378-PCDD/F in flyash and flue gases was investigated. Black-Right-Pointing-Pointer NH{sub 3} decreased the concentration of PCDD and PCDF by 34-75% in the flyash. Black-Right-Pointing-Pointer NH{sub 3} decreased the concentration of PCDD and PCDF by 21-40% from the flue gases. Black-Right-Pointing-Pointer SO{sub 2} led to 99% PCDD and 93% PCDF reductions in the flyash. Black-Right-Pointing-Pointer SO{sub 2} led to 89% PCDD and 76% PCDF reductions in the flue gases. - Abstract: The influence of gaseous ammonia and sulphur dioxide on the formation of 2378-substituted PCDD/F on a reference flyash from a municipal waste incinerator has been investigated using a laboratory scale fixed-bed reactor. The reference flyash samples (BCR-490) was reacted under a simulated flue gas stream at temperatures of 225 and 375 Degree-Sign C for 96 h. The experiments were carried out in two series: first with simulated flue gas alone, and then with injection of NH{sub 3} or SO{sub 2} gas into the flue gas just before the reactor inlet. It was found that the injection of gaseous ammonia into the flue gas could decrease the concentration of both PCDD and PCDF by 34-75% from the solid phase and by 21-40% from the gas phase. Converting the results to I-TEQ values, it could reduce the total I-TEQ values of PCDD and PCDF in the sum of the flyash and exhaust flue gas by 42-75% and 24-57% respectively. The application of SO{sub 2} led to 99% and 93% reductions in the PCDD and PCDF average congener concentrations, respectively in the solid phase. In the gas phase, the total reductions were 89% and 76% for PCDD and PCDF, respectively. Moreover, addition of SO{sub 2} reduced the total I-TEQ value of PCDD and PCDF in the flyash and exhaust flue gas together by 60-86% and 72-82% respectively. Sulphur dioxide was more effective than ammonia in suppressing PCDD/F formation in flyash under the conditions investigated.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2020

Panel Session 120: Roundtable: Waste Management - Energy Facilities Contractor Operating Group (EFCOG WM)

This Roundtable focused on the WM EFCOG working group as they met to discuss a variety of issues of importance to US DOE waste management operations. The purpose of the WM EFCOG is to seek out and promote the best management and operating practices, cost effective technologies and disposal options for all waste streams generated at US DOE facilities whether destined for US DOE or commercial facilities. Each site representative provided a radioactive waste management lessons learned briefing. Panelists with presentations: Savannah River Site (SRS) Solid Waste Program Update (Kerri Crawford); Idaho Clean-up Project Status (Frederick Hughes); West Valley Demonstration Project Update - EFCOG Waste Management Working Group (John Rendall); Key Project Challenges and Successes (Kalli Shupe); EFCOG WMWG Challenging Waste Subgroup Update (Tammy Monday); CH TRU Difficult Waste Stream Processing Challenges (Daniel Pancake); Argonne Building 350 Clean-up Success Using Effective Work Control (Daniel Dilday); Problematic Waste Streams on the Oak Ridge Reservation (John Wrapp)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Exploring Engineering Control through Process Manipulation of Radioactive Liquid Waste Tank Chemical Cleaning

One method of remediating legacy liquid radioactive waste produced during the cold war, is aggressive in-tank chemical cleaning. Chemical cleaning has successfully reduced the curie content of residual waste heels in large underground storage tanks; however this process generates significant chemical hazards. Mercury is often the bounding hazard due to its extensive use in the separations process that produced the waste. This paper explores how variations in controllable process factors, tank level and temperature, may be manipulated to reduce the hazard potential related to mercury vapor generation. When compared using a multivariate regression analysis, findings indicated that there was a significant relationship between both tank level (p value of 1.65x10 -23 ) and temperature (p value of 6.39x10 -6 ) to the mercury vapor concentration in the tank ventilation system. Tank temperature showed the most promise as a controllable parameter for future tank cleaning endeavors. Despite statistically significant relationships, there may not be confidence in the ability to control accident scenarios to below mercury’s IDLH or PAC-III levels for future cleaning initiatives.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Capability to Recover Plutonium-238 in H-Canyon/HB-Line - 13248

Plutonium-238 is used in Radioisotope Thermoelectric Generators (RTGs) to generate electrical power and in Radioisotope Heater Units (RHUs) to produce heat for electronics and environmental control for deep space missions. The domestic supply of Pu-238 consists of scrap material from previous mission production or material purchased from Russia. Currently, the United States has no significant production scale operational capability to produce and separate new Pu-238 from irradiated neptunium-237 targets. The Department of Energy - Nuclear Energy is currently evaluating and developing plans to reconstitute the United States capability to produce Pu-238 from irradiated Np-237 targets. The Savannah River Site had previously produced and/or processed all the Pu-238 utilized in Radioisotope Thermoelectric Generators (RTGs) for deep space missions up to and including the majority of the plutonium for the Cassini Mission. The previous full production cycle capabilities included: Np- 237 target fabrication, target irradiation, target dissolution and Np-237 and Pu-238 separation and purification, conversion of Np-237 and Pu-238 to oxide, scrap recovery, and Pu-238 encapsulation. The capability and equipment still exist and could be revitalized or put back into service to recover and purify Pu-238/Np-237 or broken General Purpose Heat Source (GPHS) pellets utilizing existing process equipment in HB-Line Scrap Recovery, and H-Canyon Frame Waste Recovery processes. The conversion of Np-237 and Pu-238 to oxide can be performed in the existing HB-Line Phase-2 and Phase- 3 Processes. Dissolution of irradiated Np-237 target material, and separation and purification of Np-237 and Pu-238 product streams would be possible at production rates of ∼2 kg/month of Pu-238 if the existing H-Canyon Frames Process spare equipment were re-installed. Previously, the primary H-Canyon Frames equipment was removed to be replaced: however, the replacement project was stopped. The spare equipment is stored and still available for installation. Out of specification Pu-238 scrap material can be purified and recovered by utilizing the HB-Line Phase- 1 Scrap Recovery Line and the Phase-3 Pu-238 Oxide Conversion Line along with H-Canyon Frame Waste Recovery process. In addition, it also covers and describes utilizing the Phase-2 Np-237 Oxide Conversion Line, in conjunction with the H-Canyon Frames Process to restore the H-Canyon capability to process and recover Np-237 and Pu-238 from irradiated Np-237 targets and address potential synergies with other programs like recovery of Pu-244 and heavy isotopes of curium from other target material. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Prospects for applications of electron beams in processing of gas and oil hydrocarbons

Waste-free processing of oil and oil gases can be based on electron-beam technologies. Their major advantage is an opportunity of controlled manufacturing of a wide range of products with a higher utility value at moderate temperatures and pressures. The work considers certain key aspects of electron beam technologies applied for the chain cracking of heavy crude oil, for the synthesis of premium gasoline from oil gases, and also for the hydrogenation, alkylation, and isomerization of unsaturated oil products. Electronbeam processing of oil can be embodied via compact mobile modules which are applicable for direct usage at distant oil and gas fields. More cost-effective and reliable electron accelerators should be developed to realize the potential of electron-beam technologies.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

Simulation of cooling and solidification of three-dimensional bulk borosilicate glass. Effect of structural relaxations

The modeling of the viscoelastic stress evolution and specific volume relaxation of a bulky glass cast is presented in this article and is applied to the experimental cooling process of an inactive nuclear waste vitrification process. The concerned borosilicate glass is solidified and cooled down to ambient temperature in a stainless steel canister, and the thermomechanical response of the package is simulated. There exists a deviant compression of the liquid core due to the large glass package compared to standard tempered glass plates. The stress load development of the glass cast is finally studied for different thermal load scenarios, where the cooling process parameters or the final cooldown rates were changed, and we found a great influence of the studied cooldown rates on the maximum stress buildup at ambient temperature.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Feasible conversion of solid waste bauxite tailings into highly crystalline 4A zeolite with valuable application

Highlights: • Concept to convert waste to valuable product is carried out in this study. • An industrially feasible and cost-effective approach was developed and optimized. • Highly crystalline and well-defined zeolite was produced under moderate conditions. • The zeolite derived from the bauxite tailings displayed high ion exchange capacity. • Bauxite tailings have potential application in heavy metal ions adsorbent. - Abstract: Bauxite tailings are a major type of solid wastes generated in the flotation process. The waste by-products caused significant environmental impact. To lessen this hazardous effect from poisonous mine tailings, a feasible and cost-effective solution was conceived and implemented. Our approach focused on reutilization of the bauxite tailings by converting it to 4A zeolite for reuse in diverse applications. Three steps were involved in the bauxite conversion: wet-chemistry, alkali fusion, and crystallization to remove impurities and to prepare porous 4A zeolite. It was found that the cubic 4A zeolite was single phase, in high purity, with high crystallinity and well-defined structure. Importantly, the 4A zeolite displayed maximum calcium ion exchange capacity averaged at 296 mg CaCO{sub 3}/g, comparable to commercially-available zeolite (310 mg CaCO{sub 3}/g) exchange capacity. Base on the optimal synthesis condition, the reaction yield of zeolite 4A from bauxite tailings achieved to about 38.43%, hence, this study will provide a new paradigm for remediation of bauxite tailings, further mitigating the environmental and health care concerns, particularly in the mainland of PR China.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Characterization of Non-pertechnetate Species Relevant to the Hanford Tank Waste

Among radioactive constituents present in the tank waste stored at the U.S. DOE Hanford Site, technetium-99 (Tc), which is generated from the fission of 235 U and 239 Pu in high yields, presents a unique challenge in that it has a long half-life ( = 292 keV; T1/2 = 2.11105 y) and exists predominately in soluble forms in the liquid supernatant and salt cake fractions of the waste. In the strongly alkaline environments prevalent in most of the tank waste, its dominant chemical form is pertechnetate (TcO 4 - , oxidation state +7). However, attempts to remove Tc from the Hanford tank waste using ion-exchange processes specific to TcO 4 - only met with limited success, particularly when processing tank waste samples containing elevated concentrations of organic complexants. This suggests that a significant fraction of the soluble Tc can be present as low-valent Tc (oxidation state < +7) (non-pertechnetate). The chemical identities of these non-pertechnetate species are poorly understood. Previous analysis of the SY-101 and SY-103 tank waste samples provided strong evidence that non-pertechnetate can be comprised of [fac-Tc(CO) 3 ] + complexes containing Tc in oxidation state +1 (Lukens et al. 2004). During the last three years, our team has expanded this work and demonstrated that high-ionic-strength solutions typifying tank waste supernatants promote oxidative stability of the [fac-Tc(CO) 3 ] + species (Rapko et al. 2013a; 2013b; Levitskaia et al. 2014; Chatterjee et al. 2015). Obtained results also suggest possible stabilization of Tc(VI) and potentially Tc(IV) oxidation states in the high-ionic-strength alkaline matrices particularly in the presence of organic chelators, so that Tc(IV, VI) can serve as important redox intermediates facilitating the reduction of Tc(VII) to Tc(I). Designing strategies for effective Tc management, including separation and immobilization, necessitates understanding the molecular structure of the non-pertechnetate species and their identification in the actual tank waste samples, which would facilitate development of new treatment technologies effective for dissimilar Tc species. The key FY 2016 results are summarized below.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Remote Handled Hot Cell Facility and Other Research Transuranic Waste (Debris)

This acceptable knowledge (AK) Summary Report has been prepared for the Central Characterization Program (CCP) for remote-handled (RH) transuranic (TRU) waste generated and managed by Sandia National Laboratories/New Mexico (SNL/NM) in Albuquerque, New Mexico. The waste described in this report was predominately generated in the SNL/NM Hot Cell Facility (HCF) during the removal and packaging of experimental material and decontamination operations in Building 6580 at Technical Area (TA)-V. The waste stream also includes a very small amount of waste (estimated at less than 1 gram of fuel similar to that used for research at the HCF) originating from classified research at TA-I. In addition, the waste also includes secondary waste termed by SNL/NM as process generated waste (PGW), created during repackaging operations for this waste at the Auxiliary Hot Cell Facility (AHCF). The waste has been stored at the Sandia Pulsed Reactor (SPR) Dense Pack Storage Facility within TA-V or the Manzano Bunkers located at Manzano Base within Kirtland Air Force Base. All of the waste is being repackaged at the AHCF. This report was prepared in accordance with CCP-TP-005, CCP Acceptable Knowledge Documentation (Reference 1), to implement the AK requirements of DOE/WIPP-02-3214, Remote-Handled TRU Waste Characterization Program Implementation Plan (WCPIP) (Reference 2); Waste Isolation Pilot Plant Hazardous Waste Facility Permit, Waste Analysis Plan (WIPP-WAP) (Reference 3); and DOE/WIPP-02-3122, Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WIPP-WAC) (Reference 4).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 2, April-June 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 97 The Nuclear Community and the Public: Cognitive and Cultural Influences on Thinking About Nuclear Risk, M. A. Meyer; 109 Twenty-Third Water Reactor Safety Information Meeting, D. A. Copinger; ACCIDENT ANALYSIS: 126 Analysis of a PWR LBLOCA Without SCRAM, Trevor N. Tyler, Rafael Macian-Juan and John H. Mahaffy; DESIGN FEATURES: 139 Vulnerability of Multiple-Barrier Systems, N. C. Lind; ENVIRONMENTAL EFFECTS: 149 A Study of Wet Catalytic Oxidation of Radioactive Spent Ion Exchange Resin by Hydrogen Peroxide, Xingchao Jian, Tianbao Wu, and Guichun Yun; 157 A Comparison Study and Resolution of Differences Between Emergency Response and Safety Analysis Codes Used at the Savannah River Site, A. A. Simpkins; OPERATING EXPERIENCES: 164 Reactor Shutdown Experience, Compiled by J. W. Cletcher; RECENT DEVELOPMENTS: 167 Reports, Standards, and Safety Guides, D. S. Queener; 172 Proposed Rule Changes as of Dec. 31,1995; ANNOUNCEMENTS: 178 American Nuclear Society 1997 Annual Meeting; 178 American Nuclear Society Nuclear Criticality and Safety Division Topical Meeting; 176 The Authors.

05 NUCLEAR FUELS↗
NASA NTRS1990

Technologies for ECLSS Evolution

Viewgraphs and discussion on technologies for Environmental Control and Life Support System (ECLSS) evolution are presented. Topics covered include: atmosphere revitalization including CO2 removal, CO2 reduction, O2 generation, and trace contaminant control; water recovery and management including urine processing, hygiene water processing, and potable water processing; and waste management. ECLSS technology schematics, process diagrams, and fluid interfaces are included.

Diamant, Bryce L.↗
NASA NTRS1993

Advanced collapsible tank for liquid containment

Tanks for bulk liquid containment will be required to support advanced planetary exploration programs. Potential applications include storage of potable, process, and waste water, and fuels and process chemicals. The launch mass and volume penalties inherent in rigid tanks suggest that collapsible tanks may be more efficient. Collapsible tanks are made of lightweight flexible material and can be folded compactly for storage and transport. Although collapsible tanks for terrestrial use are widely available, a new design was developed that has significantly less mass and bulk than existing models. Modelled after the shape of a sessible drop, this design features a dual membrane with a nearly uniform stress distribution and a low surface-to-volume ratio. It can be adapted to store a variety of liquids in nearly any environment with constant acceleration field. Three models of 10L, 50L, and 378L capacity have been constructed and tested. The 378L (100 gallon) model weighed less than 10 percent of a commercially available collapsible tank of equivalent capacity, and required less than 20 percent of the storage space when folded for transport.

Flanagan, David T.↗
DOE OSTI2012

US/UK Loan Account Project Status PMOD477

The viewgraphs describe the status of PMOD477 for LANL. The meeting will occur at DOE-HQ with NA-11 and Military Applications personnel in attendance. Serves to repatriate material with a balance to zero by December 2012. Phase 1 -- Establish formality of operations for War Reserve (WR): Complete surrogate taskings to A90 through a Materials Channel and perform US/UK lessons learned; Complete the US/UK agreed Quality Acceptance Plan, Materials Plan, Shipping procedure, and establish the formal UK/US point of contacts. Phase 2 -- Metal Manufacture (WR): Process material and store material as electrorefined metal (ER) rings, with initial assay and isotopic analysis, prior to manufacturing. Material is cast into accepted configuration and appropriate acceptance document for each aliquot will be generated. Phase 3 -- Intermediate Material Manufacture, Packaging and Shipping (WR): Continue processing of the material in accepted configuration with appropriate acceptance documentation for each aliquot. Provide an initial tasking of the material owed to UK including appropriate quality acceptance documentation. Phase 4 -- Complete Tasking (WR). Phase 5 -- Residue Processing (Non-WR): Complete processing of residue material and waste into accepted configuration with appropriate acceptance document for disposal.

CONFIGURATION↗
DOE OSTI2025

Processes in Salt Repositories for Radioactive Waste Disposal

This document summarizes the key processes (thermal, hydrological, mechanical, and chemical; THMC) impacting the features of a deep geological repository for radioactive waste in salt. Some processes are natural and on-going whether the repository is there or not, and other processes are driven by the perturbation associated with the repository. The features considered here include both engineered and natural components of the repository system. The engineered barrier system (EBS) in a salt repository is quite different from those implemented for a repository in clay or crystalline rocks, because it is comprised mostly of granular salt and salt-compatible cements, rather than bentonite. When compared to other rocks (i.e., silicates), salt has unique properties that make it an excellent potential host rock. Openings and fractures in salt creep closed readily. Salt has high thermal conductivity, which can reduce peak temperatures. Additionally, far away from the excavations the porosity of salt is unconnected, which leads to essentially zero advective or diffusive transport. The small amount of hypersaline brine occurring in salt minimizes microbial activity, reduces colloid-assisted transport, and eliminates in-package criticality (i.e., chloride is a neutron poison). At the end of the report, we present a brief outline for a potential salt repository, including considerations avoided in previous repository disposal concepts. We propose considering higher-temperature processes in future disposal concepts, rather than trying to minimize the thermal perturbation of the repository. Since hot salt is drier, a dry repository would limit corrosion, gas generation, and solute transport. Openings and fractures creep shut faster in hot salt. Therefore, higher temperatures could be seen as beneficial, rather than something to minimize, through increased spacing between waste packages (increasing repository costs).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Use of the Waste-Incidental-to-Reprocessing Citation Process at the West Valley Demonstration Project - 12250

The West Valley Demonstration Project recently achieved a breakthrough in management of radioactive waste from reprocessing of spent nuclear fuel by taking advantage of lessons learned at other Department of Energy (DOE) sites in implementation of the waste-incidental-to-reprocessing citation process of DOE Manual 435.1-1, Radioactive Waste Management. This breakthrough involved a revision to the site procedure on waste-incidental to reprocessing. This procedure revision served as the basis for a determination by the DOE West Valley field office using the citation process that three secondary waste streams consisting of equipment that had once been contaminated by association with HLW are not HLW following decontamination and may be disposed of as low-level waste (LLW) or transuranic waste. These waste streams, which comprised much of the approximately 380 cubic meters of West Valley waste contaminated by association with HLW, included several vessels and certain tank farm equipment. By making use of lessons learned in use of the citation process by other DOE sites and information developed to support use of the citation process at the Hanford site and the Savannah River Site, the team developed a technical basis for showing that use of the citation process of DOE Manual 435.1-1 for the three new waste stream was appropriate and technically justified. The Waste Management Working Group of the EFCOG assisted in transferring lessons learned by drawing on experience from around the DOE complex. This process shared knowledge about effective implementation of the citation process in a manner that proved to be beneficial to the West Valley Demonstration Project and resulted in a technical basis document that could be used to determine that the three new waste streams were not HLW. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Experimental study of the tritium distribution in the effluents resulting from the sodium hydrolysis

Within the framework of the dismantling of fast breeder reactors in France several processes are under investigation regarding sodium disposal. One of them, called ELA (radioactive sodium waste treatment process), is based on the implementation of the sodium-water reaction, in a controlled and progressive way, to remove residual sodium. This sodium contains impurities such as sodium hydride, sodium oxide and tritiated sodium hydride. The hydrolysis of these various chemical species leads to the production of a liquid effluent, mainly composed of an aqueous solution of sodium hydroxide, and a gaseous effluent, mainly composed of nitrogen (inert gas), hydrogen and steam. The tritium is distributed between these effluents, and, within the gaseous effluent, according to its forms HT and HTO (tritiated water). HTO being 10,000 times more radio-toxic than HT, a precise knowledge of the mechanisms governing the phase distribution of tritium is necessary. This paper presents the first experimental results from a parametric study on the tritium distribution between the various effluents generated during hydrolysis operations. A series of experiments have been performed in order to study the influence of water flow rate, argon flow rate, initial mass and specific activity of the hydrolyzed sodium sample. An important influence of the total tritium concentration in the hydrolyzed sample has been highlighted. As for the phenomena suspected to be responsible for the phase change of tritiated water, in the studied range of parameters, vaporization induced by the heat of reactions seems to be dominant over the evaporation induced by the inert gas flow rate.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗