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Cryogenic distillation (CD) of hydrogen in combination with Liquid Phase Catalytic Exchange (LPCE) or Combined Electrolytic Catalytic Exchange (CECE) process is used for tritium removal/recovery from tritiated water from fission rectors like CANDU and is the proposed technology for ITER/DEMO. Tritiated water is being obtained after long time operation of CANDU reactors, or in case of ITER mainly by the Detritiation System. The cryogenic system of the Experimental Pilot plant for tritium and deuterium separation from ICSI Rm. Valcea consists of four distillation columns and significant effort is required in various batch mode operations for achieving high tritium concentration. Some problems have been experienced with the fourth column of the cascade regarding the heat transient transfer during start-up. This paper intent is to present a CAE/FEA (Computer Aided Engineering/Finite Element Analysis) thermal simulation of the column during the transitory thermal regime in order to investigate the temperature distribution along the column and its connections to other equipment. The simulation is done on the as-built design of the column and takes in consideration the actual configuration. This work provides an important platform to understand the thermal phenomena during cool down of a cryogenic distillation column besides finding the failures in thermal insulation or design flaws. The results of the simulation can be used as lessons-learned for future design work of cryogenic systems like Tritium Removal Facility from Cernavoda Nuclear Power Plant or ITER/DEMO Isotope Separation System. [ABSTRACT FROM AUTHOR]

• Cernavoda Tritium Removal Facility as tritium supplier for fusion activities is in line with ITER plans for construction. • Tritium quantities produced by CTRF are lower than those required by fusion activities, but CTRF can be considered as an important contributor. • Position of CTRF in EU may be considered as an advantage, all requirements for tritium management (storage for transportation, transfer arrangements, safety) can be easily met. National R&D for Cryogenics and Isotopes Separation (ICSI) completed the conceptual design of the Cernavoda Tritium Removal Facility (CTRF) in 2015. CTRF is located at CNE 1 1 Centrala Nucleara Electrica - NPP Cernavoda and is sized to process 40 kg/h heavy water from 2 CANDU reactors, with a detritiation factor of 100, over 40 years design life. CTRF removes tritium using liquid phase catalytic exchange (LPCE) paired with cryogenic distillation (CD). The design of CTRF uses expertise from ICSI (Pilot Plant for Tritium and Deuterium Separation) and from Canada (Kinectrics), together with experience from Wolsung TRF project, research tritium laboratories and industry. The CTRF design uses the most advanced TRF technology, including recent safety requirements for a tritium industrial facility as specified by the Romanian Regulator. Construction of CTRF is planned to start in 2020 and the detritiation of heavy water from first reactor is scheduled to begin in 2026. After this date tritium stored will become available to be used for fusion research and industrial facilities. First phase of detritiation process is to reduce the moderator tritium content from 65 Ci/kg to 10 Ci/kg, second phase to maintain the tritium concentration up to 10 Ci/kg and third phase to use CTRF to reduce tritium content as low as possible before decommissioning of the site. This paper will present the technologies used, provide a prediction of tritium production, and considerations for possible use of CTRF for supplying He-3, as a by-product of the detritiation process. [ABSTRACT FROM AUTHOR]

• Oxide wedging crack growth postulated for CANDU pressure tube manufacturing flaws. • Compressibility of oxide wedge is insignificant deviation from displacement control. • Flaw stress intensity factor effectively constant under cyclic operating stresses. • Fatigue crack growth cannot occur while oxide wedging is active mechanism. In-service inspections of Zr-2.5Nb pressure tubes in CANDU nuclear reactors have detected inner diameter surface breaking flaw indications which are classified as manufacturing flaws. It is considered prudent to evaluate potential growth by means of a postulated oxide wedging crack growth mechanism. An oxide wedging crack growth model has previously been developed and applied to predict future flaw sizes. Fatigue crack growth due to cyclic operating stresses is calculated for these flaws, and this is added to the postulated oxide wedging crack growth for fitness-for-service evaluations. This established approach is overly conservative since fatigue crack growth cannot occur while the oxide wedging crack growth mechanism is active, due to the essentially displacement-controlled loading of the flaw by the oxide wedge, which causes the stress intensity factor for the flaw to be effectively constant under cyclic operating stresses. This paper addresses this conservatism by evaluating crack growth due to postulated oxide wedging, and demonstrating that fatigue crack growth is not possible under oxide wedge loading. [ABSTRACT FROM AUTHOR]

• Environmental conditions were measured during a leak rate test (LRT) of a reactor building (RB). • The environmental conditions were used to evaluate thermal strains on RB. • Internal pressure-strain relationships of four LRTs were obtained after removing thermal strains. • The RB has shown a very small decrease in the pressure-strain relationship over 18 years. With nuclear reactor power plants, Reactor Building (RB) Leak-Rate Tests (LRTs) are conducted at regular intervals to confirm the air leak tightness of a containment structure. These LRTs provide an excellent opportunity to study the containment structure's behaviour when subjected to internal pressure, thereby evaluating the structural performance of the containment structure. However, significant time-dependent fluctuations are observed in the measured strains due to the contribution of thermal strain induced by the surrounding environment. In order to accurately study the mechanical behavior of the containment structure, the thermal strain must be separated from the measured total strain. In this study, environmental data was collected during an RB LRT. Finite element analyses were carried out by utilizing the environmental conditions to predict the thermal strain of the containment structure. The predicted thermal strain was then subtracted from the measured total strain, so the underlying mechanical strain induced by the internal pressure could be approximated. The mechanical behavior of a containment structure during LRTs over an eighteen-year period is compared to understand the overall aging effects of a CANDU® 1 1 CANDU (CANada Deuterium Uranium Reactor) is a registered trademark of the Atomic Energy of Canada Limited (AECL), used under exclusive license by Candu Energy Inc. containment structure. Only a slight change in structural stiffness was observed over eighteen years though further study is suggested to confirm this. This paper outlines how field data collected during LRTs can be used to provide insight into understanding the long-term behavior of a containment structure. [ABSTRACT FROM AUTHOR]

Abstract: A theoretical analysis has been developed to predict the critical height of the onset of gas entrainment (OGE) during dual and triple discharge from a stratified two-phase region. The two and three discharge branches are mounted on a circular wall, resembling a circular reservoir of a CANDU header–feeder configuration. A point sink model has been developed to predict the critical height and to map the velocities and acceleration flow fields during OGE. The model was verified by comparing the theoretically predicted critical height with the available experimental results. The theoretically predicted critical height is found to be a function of the branch Froude number, the location of the secondary branch with respect to the primary branch, and the angle between the branches. The effect of these variables on the predicted OGE height was investigated and is presented in this paper. Predictions of the critical height were shown to be within 25% of experimental values in both dual and triple discharge. [Copyright &y& Elsevier]

Delayed hydride cracking (DHC) in Zr–2.5Nb alloy material is of interest to the CANadian Deuterium Uranium (CANDU)(1) industry in the context of the potential to initiate DHC at a blunt flaw in a CANDU reactor pressure tube. The existing CANDU blunt flaw DHC evaluation procedure is based on a threshold peak flaw-tip stress for DHC initiation that is independent of flaw geometry. Work is underway to improve the existing blunt flaw DHC evaluation procedure by developing a methodology that takes into account the effect of flaw geometry parameters. The methodology is based on representing the stress relaxation due to hydride formation and crack initiation by a process zone. A process-zone model was used to develop failure assessment diagrams in the conventional format of R6 as described in this paper, although the coordinates of the diagram have a different physical meaning. The failure assessment diagrams were shown to be independent of the scale of the configuration as well as specific values of material properties. The failure assessment curves as derived in the conventional R6 format exhibit a strong dependence on the geometry of the flaw due to the stress concentration and associated stress gradient. By redefining the ordinate of the failure assessment diagram in terms of peak flaw-tip stress the geometry dependence of the failure assessment curves was reduced significantly. This minimal geometry dependence of the failure assessment curves is valuable with regards to practical engineering flaw evaluations. Agreement between the failure assessment diagram predictions and experimental results is reasonable. [Copyright &y& Elsevier]

• ISF with four distillation columns cascade to separate the deuterium-tritium mixture. • Control systems: Plant Control System (SCADA) and Safety Instrumented System (SIS). • SCADA – keep the process within the accepted limits - 1st layer of defence in depth. • SIS – performs the assigned security functions - 2nd defence layer in depth. Isotopes separation technologies are used in both Tritium Removal Facilities associated with CANDU reactors (to remove tritium from heavy water) and fusion applications (eq. ITER Isotope Separation System) to recover and enrich tritium for further operation. The expertise gained during design, manufacturing, building, commissioning and operation of the Cryogenic Distillation System (CDS) with proprietary four column design built at ICSI Rm. Valcea, could be applied for various tritium recovery applications, regardless of the type of front-end process used, Liquid Phase Catalytic Exchange or Combined Electrolysis and Catalytic Exchange. Cryogenic Distillation feed is a mixture of hydrogen isotopes which presents hazards in operation (explosion and tritium contamination), therefore the CDS of PESTD was designed to meet industrial and nuclear applicable standards. In this paper we present an overview of the CDS focusing on the process control and safety approach which meet specific conditions of Romanian codes and standards required for operation. However, specific information may be used later as reference for design consideration for other CDS. The cryogenic distillation process from ICSI Rm. Valcea is designed as four columns cascade inside a cold-box which acts as second barrier against hydrogen release and thermal insulation. The CDS process control is based on a Programmable Automation Controller (PACs), a Supervisory Control and Data Acquisition (SCADA) and specific low temperature field equipment and devices. The control room operator supervises the process using a specific designed Human Machine Interface (HMI) which permits to monitor process parameters, alarms acknowledgment and acts to adjust the process or override normal PACs control. [ABSTRACT FROM AUTHOR]

• Fluctuation of strain during LRT of a containment structure is due to temperature. • A method has developed to run thermo-mechanical analysis of a containment structure. • Strain history from the numerical analysis agrees well with field measurements. • The proposed method can be potentially used to remove noise in measured strain. A strategy for isolating the thermal strain due to solar radiation from the total measured strain during the Integrated Leak Rate Test of a nuclear containment structure is presented in this paper. In the proposed method, a series of numerical models are developed to estimate the temperature change on the prestressed concrete containment structure and to predict the strain due to temperature and applied pressure. The effects of the solar radiation, heat exchange with internal and external air and heat conduction are considered in the proposed method. The developed method is validated by reproducing the total strain profiles measured from the leak rate tests conducted at the Point Lepreau CANDU® 1 1 CANDU (CANada Deuterium Uranium Reactor) is a registered trademark of the Atomic Energy of Canada Limited (AECL), used under exclusive license by Candu Energy Inc. 6 containment structure. [ABSTRACT FROM AUTHOR]