Evaluation of tritium leakage rate into seawater in fusion DEMO cooling water system
Introduction
Considering the feasibility of Fusion DEMO, tritium handling is one of the important issues. It is desirable that tritium in atmosphere or in cooling water should be removed by proven technologies, and adverse effect on the environment caused by tritium should be suppressed. This paper focuses on tritium in cooling water system. In DEMO, large amount of tritium permeation is anticipated to blanket and divertor cooling water [1]. Thus, to control tritium concentration in the cooling water, part of the cooling water is bypassed and transported to a tritium removal facility (TRF). To assess whether existing technology can be applied or not, required DEMO TRF performance should be analyzed.
In Japan's DEMO, the thermal energy of blanket and divertor is used for power generation [2], and their cooling water systems are connected to power generation system via heat exchangers (HX) and steam generators (SG). Thus, tritium in the blanket and divertor cooling water will penetrate HX or SG. Finally, tritium will penetrate a condenser pipes and be leaked into seawater. In addition to DEMO TRF performance analysis, such tritium leakage rate analysis is necessary for a discussion of DEMO feasibility. In this case, tritium permeation is water to water via a metal cooling pipe, and there are several researches about the hydrogen permeation rate evaluation in such situation [3,4], however, effective evaluation method hasn't be established yet. Thus, in this research, with the gas to gas hydrogen permeation rate evaluation method, the conservative tritium permeation rate evaluation has been introduced. The construction of this paper is as follows. Section 2 describes the DEMO CWS and power generation system design and the required TRF performance. Section 3 explains the evaluation method of the tritium permeation. Section 4 shows the result of the calculation. Section 5 is the summary.
Section snippets
DEMO CWS design
Fig. 1 shows the concept of DEMO main CWS and power generation system (PGS) in Japan (the detail is shown in Ref. [2]). In Japan's DEMO, 5 in-vessel components are cooled. Blanket, divertor (it is divided into 2 parts), back plate and vacuum vessel. Divertor is divided into reduced-activation ferritic martensitic (RAFM) steel pipe part and Cu-alloy pipe part. As shown in Fig. 1, thermal enrgy of divertor (RAFM) is used for power generation. Divertor (Cu-alloy) thermal energy is used as heat
Evaluation concept
As shown in Section 2, the tritium concentration in the primary CWS can be controlled with known technology at 1TBq/kg. This tritium is anticipated to permeate the metal pipes in HX from the primary cooling water to the down stream cooling water. After that, turbine system tritium permeates the condenser pipes and finally, tritium is leaked into seawater. This tritium activity should be suppressed under the tolerance. In Japan, the tolerance of pressurized Bq/year water reactor is decided as
Calculation conditions
Using Eqs. (9)–(14), tritium permeation rate is calculated. In this research, 2 cases are considered shown in Fig. 2.
To evaluate the effect of HX, HX is used in case 1 and not used in case 2. The tritium concentration in the primary CWS ((a) and (d) in Fig. 2) are assumed to be constant at 1TBq/kg, and in (a), (b) and (d) in Fig. 2, [H2] is assumed to be controlled at 30Ncc/kg. This is the same value with the typical value of primary CWS in pressurized water reactors [9]. To evaluate the effect
Summary
In our previous research, Japan's DEMO CWS has been designed In this design, blanket and divertor CWS are anticipated to have several amount of tritium because of the tritium permeation from the core plasma [1]. Thus, a part of the cooling water should be bypassed and cleaned in TRF. From the designed CWS mass flow, the total required bypass flow rate to keep the tritium concentration at 1TBq/kg is calculated as 89.1 kg/h. This is under the typical performance of the existing TRF. This result
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