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Research Papers

Expected Safety Performance of the Supercritical Water Reactor Fuel Qualification Test

[+] Author and Article Information
Manuel Raqué

Karlsruhe Institute of Technology,
Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen, Germany
e-mail: manuel.raque@ensi.ch

Thomas Schulenberg

Mem. ASME
Karlsruhe Institute of Technology,
Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen, Germany
e-mail: thomas.schulenberg@kit edu

Tobias Zeiger

Karlsruhe Institute of Technology,
Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen, Germany
e-mail: tzeiger@bfs.de

1Present address: Eidgenössisches Nuklearsicherheitsinspektorat ENSI, Industriestr. 19, CH-5200 Brugg, Switzerland.

2Present address: Bundesamt für Strahlenschutz, Willy-Brandt-Str. 5, D-38226 Salzgitter, Germany.

Manuscript received April 10, 2015; final manuscript received June 23, 2015; published online December 9, 2015. Editor: Igor Pioro.

ASME J of Nuclear Rad Sci 2(1), 011005 (Dec 09, 2015) (9 pages) Paper No: NERS-15-1052; doi: 10.1115/1.4030917 History: Received April 10, 2015; Accepted June 25, 2015

Abstract

The supercritical water reactor (SCWR) fuel qualification test is an in-pile test of a four-rod fuel assembly at supercritical pressure inside a research reactor, which is operated at atmospheric pressure. The risk of radioactive release from this new test facility should not exceed the accepted risk of the existing research reactor. A large number of safety analyses have been performed to assess this risk, which are summarized in this paper. Among them are studies of design basis accidents, assuming different failure modes of the high-pressure system, as well as an assessment of consequences of postulated accidents beyond the design basis. Results show that the safety objectives can be met.

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Figures

Fig. 1

Cross sections of the pressure tube for the fuel qualification test: left, heated test section; right, recuperator section

Fig. 2

Sketch of the fuel qualification test loop and its safety systems [3]

Fig. 3

Coolant pressure, mass flow rates, and temperatures in the test section after a sudden break of the emergency cooling line L3 [3]

Fig. 4

Coolant pressure (left), mass flow rates in supply lines of the safety system, and pressure in the test section (right) after a sudden break of the feed line L1, close to the headpiece of the pressure tube [3]

Fig. 5

Power, mass flow rate, pressure, and coolant temperature in the test section after a trip of the recirculation pump HCC [3]

Fig. 6

Water temperatures inside depressurization tank BN and inside emergency reservoir HN1 as well as flow rate through pump HC3 (left). Coolant temperatures along the fuel rods over the total simulation period of 500,000 s (right) [3].

Fig. 7

Original (left) and modified attachment (right) of the pressure tube [8]

Fig. 8

Simulation of a 30-mm square fragment, total length 60 mm (dry), 70 mm (dry), and 60 mm (surrounded by water). Gray value indicates plastic strain [8].

Fig. 9

Lengthwise split of the pressure tube: dry (left) and surrounded by water (right, water invisible). Gray value indicates plastic strain [8].

Fig. 10

Coolant temperatures along the fuel rods for the first 600 s (left) and cladding temperatures during the total simulation time of 200 min (right). Inlet and outlet denote the lower and upper end of the assembly box. Segments 1–4 are in between [3].

Fig. 11

Progression of heating power (left) and rod centerline temperature (right) used as a basis for the source-term evaluation (right) [3]

Fig. 12

Source term of iodine-131: fuel rod centerline temperature and I-131 release per time step and total release with converted effective dose [3]

Fig. 13

Steam temperature predicted for the accident postulated in Section 3.2. Channel numbers refer to Fig. 1.

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