HTGR Knowledge Base

Conference Article: Nuclear design code development for fuel management and safety analysis of HTGR in Indonesia

Abstract

The development works and some results on the nuclear design codes for fuel management and safety analysis of HTGRs conducted in Batan is reported. Batan-MPASS code, an incore fuel management code, have been developed and verified to simulate the continuously flow of the pebble fuel elements in a pebble-bed type HTGR core for both once-through-then-out (OTTO) as well as multipass fueling schemes. One important feature of the code is that it can search directly the equilibrium core condition without simulating the transient cores. A similar code, Batan-PEU code, has also been built to simulate the peu a peu fueling scheme. These codes are equipped with in-core thermal-hydraulics modules for estimating the pebble fuel temperature. For prismatic/block type HTGR, Batan-FUEL code has been compiled where originally the code was developed for in-core fuel management of a research reactor with ordinary batch refueling scheme. The thermal-hydraulic modules for Batan-FUEL are planned to be developed in the future. The diffusion calculation module within Batan-FUEL code has already successfully used for the benchmark problems of the Japanese HTTR's start-up core physics experiments. These codes are based on 2-D, 3-D few group diffusion theory and the required cross section libraries were compiled using V.S.O.P and DELIGHT-7 cell calculation codes, developed by KFA and JAERI, respectively. Some application results of the codes for the modular HTR-M 200 MWth design are reported. The accomplishment of these codes are expected to contribute for assessing the techno-economic of small and medium-scale modular HTR design presently being conducted by Batan, especially for providing the fuel cost estimation.

For safety and accident analysis, two codes have been developed to simulate the depressurization and reactivity accidents, respectively, in a pebble-bed type HTGR. For the depressurization accident, the decay heat generated in the core is calculated based on the core composition prior to the accident using the JNDC Fission Product and Yield Data compiled by JAERI, taking into account also the contribution of decay heat of heavy metals. For the reactivity accidents, the neutron dynamics is simulated in the code with the improved quasi-static model under 2-D, RZ geometry, neutron diffusion theory. The application of the two codes for assessment of the modular HTR-M 200 MWth safety under such accidents is reported.

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Reference:

Proceedings of a Technical Committee Meeting held in Beijing, People's Republic of China, 2-4 November 1998

International Atomic Energy Agency, International Working Group on Gas-Cooled Reactors, Vienna (Austria)

IAEA-TECDOC--1210, pp:85-99