Technical Brief

Thermal Hydraulic and Safety Assessment of First Wall Helium Cooling System of a Generalized Test Blanket System in ITER Using RELAP/SCDAPSIM/MOD4.0 Code

[+] Author and Article Information
S. P. Saraswat

Nuclear Engineering and Technology Programme, Indian Institute of Technology Kanpur,
Kanpur 208016, India
e-mail: satyasar@iitk.ac.in; satyasivam@gmail.com

P. Munshi

Nuclear Engineering and Technology Programme, Indian Institute of Technology Kanpur,
Kanpur 208016, India
e-mail: pmunshi@iitk.ac.in

A. Khanna

Nuclear Engineering and Technology Programme, Indian Institute of Technology Kanpur,
Kanpur 208016, India
e-mail: akhanna@iitk.ac.in

C. Allison

Innovative Systems Software,
Idaho Falls, ID 83406
e-mail: iss@cableone.net

Manuscript received January 30, 2016; final manuscript received August 11, 2016; published online December 20, 2016. Assoc. Editor: Akos Horvath.

ASME J of Nuclear Rad Sci 3(1), 014503 (Dec 20, 2016) (7 pages) Paper No: NERS-16-1012; doi: 10.1115/1.4034680 History: Received January 30, 2016; Accepted August 14, 2016

The key objective of the test blanket module (TBM) program is to develop the design technology for DEMO and future power-producing fusion reactors. The proposed first wall of the test blanket system (TBS) is a generalized concept for testing in ITER, an experimental fusion reactor under construction in France presently. The first wall of TBM (TBM FW) directly faces the plasma and is cooled by the first wall helium cooling system (FWHCS), which is considered as a critical component from an ITER safety point of view. The scope of this work comprises thermal hydraulic analysis of the FWHCS of a generalized TBS and the assessment of postulated initiating events (PIEs) on the ITER safety with the help of thermal-hydraulic code RELAP/SCDAPSIM/MOD4.0. The three reference accidents: in-vacuum vessel (VV) loss of coolant accident (in-vessel LOCA), ex-vessel LOCA, and loss of flow accident (LOFA) in FWHCS are selected for the safety assessment. This safety assessment addresses safety concerns resulting from FWHCS component failure, such as VV pressurization, TBM FW temperature profile, pressurization of port cell (PC) and Tokomak cooling water system vault annex (TCWS-VA), and passive decay heat removal capability. The analysis shows that the critical parameters are under the design limit and have large safety margins, in the investigated accident scenarios. A comparative analysis is also carried out with the previous results to validate the results.

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Grahic Jump Location
Fig. 1

Thermal hydraulic nodalization of TBM FWHCS

Grahic Jump Location
Fig. 2

Steady-state analysis FWHCS: (a) helium coolant temperature at inlet and outlet of TBM; (b) FWHCS mass flow rate at the inlet of TBM; (c) TBM FW temperature; (d) helium coolant pressure at inlet and outlet of TBM

Grahic Jump Location
Fig. 3

In-vessel LOCA analysis of FWHCS: (a) TBM FW temperature evolution during accident; (b) TBM FW temperature evolution for 10 days during accident (with decay heat)

Grahic Jump Location
Fig. 4

In-vessel LOCA analysis of FWHCS: (a) pressure profile of TBM FW and VV during accident; (b) break mass flow rate into VV from TBM FW; (c) integrated mass of helium into VV through TBM FW break

Grahic Jump Location
Fig. 5

Ex-vessel LOCA analysis of FWHCS: (a) integrated mass of helium into PC through hot leg break; (b) TBM FW temperature evolution for 10 days during accident (with decay heat); (c) mass flow rate through the hot leg of FWHCS loop; (d) TBM FW temperature evolution during accident

Grahic Jump Location
Fig. 6

Ex-vessel LOCA analysis: (a) pressure in PC during accident; (b) pressure in TCWS-VA; (c) pressure in PC and VV during ex-vessel LOCA followed by TBM FW failure

Grahic Jump Location
Fig. 7

LOFA analysis: (a) mass flow rate at the inlet of TBM; (b) TBM FW temperature evolution during normal operation, with and without FPSS; (c) TBM FW temperature evolution for 10 days during accident (with decay heat)




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