On the Strategy and Requirements for Neutronics Testing in ITER

Journal Article · Sun May 15 04:00:00 EDT 2005 · Fusion Science and Technology

OSTI ID:20854046

Youssef, M Z ^[1]; Sawan, M E ^[2]

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1. University of California-Los Angeles (United States)
2. University of Wisconsin-Madison (United States)

Neutronics testing is among the several types of fusion technology testing scheduled to be performed in ITER. The three ports assigned for testing will test several blanket concepts proposed by the various parties with test blanket modules (TBM) that utilize different breeders and coolants. Nevertheless, neutronics issues to be resolved in ITER-TBM are generic in nature and are important to each TBM type. Dedicated neutronics tests specifically address the accuracy involved in predicting key neutronics parameters such as tritium production rate, TPR, volumetric heating rate, induced activation and decay heat, and radiation damage to the reactor components. In this paper, we address some strategies for performing the neutronics tests. Tritium self-sufficiency cannot be confirmed by testing in ITER, however, the testing can provide valuable information regarding the main parameters needed to assess the feasibility of achieving tritium self-sufficiency. The paper also addresses the operational requirement (i.e. flux and fluence) as well as the geometrical requirement of the test module (i.e. minimum size) in order to have meaningful and useful tests. Measured neutronics data require high spatial resolution. This necessitates that the measured quantity be as flat as possible in the innermost locations inside the test module. This requirement has been confirmed in the present work based on results from two-dimensional calculations. The US and Japan solid breeder test blanket modules are placed inside half a port in ITER. The R-{theta} model used accounts for the presence of the ITER shielding blanket and the surrounding frame of the port.

• OSTI ID: 20854046
• Journal Information: Fusion Science and Technology, Journal Name: Fusion Science and Technology Journal Issue: 4 Vol. 47; ISSN 1536-1055
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
ACCURACY
COOLANTS
HEATING RATE
ITER TOKAMAK
JAPAN
NEUTRON FLUENCE
PERFORMANCE TESTING
PHYSICAL RADIATION EFFECTS
SHIELDING
SPATIAL RESOLUTION
TRITIUM