Published May 1998
| Version v1
Miscellaneous
Benchmark calculation of CANDU end shielding system
Description
A shielding analysis was performed for the end shield of CANDU 6 reactor. The one-dimensional discrete ordinate code ANISN with a 38-group neutron-gamma library, extracted from DLC-37D library, was used to estimate the dose rate for the natural uranium CANDU reactor. For comparison, MCNP-4B calculation was performed for the same system using continuous, discrete and multi-group libraries. The comparison has shown that the total dose rate of the ANISN calculation agrees well with that of the MCNP calculation. However, the individual dose rate (neutron and gamma) has shown opposite trends between ANISN and MCNP estimates, which may require a consistent library generation for both codes
Additional details
Publishing Information
- Publisher
- KNS
- Imprint Place
- Taejon (Korea, Republic of)
- Imprint Title
- Proceedings of the Korean Nuclear Society spring meeting Vol. 2
- Imprint Pagination
- 1031 p.
- Journal Page Range
- p. 618-623
Conference
- Title
- 1998 spring meeting of the Korean Nuclear Society
- Dates
- 29-30 May 1998
- Place
- Seoul (Korea, Republic of)
INIS
- Country of Publication
- Korea, Republic of
- Country of Input or Organization
- Korea, Republic of
- INIS RN
- 33056805
- Subject category
- S61: RADIATION PROTECTION AND DOSIMETRY;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- A CODES; BENCHMARKS; CANDU TYPE REACTORS; COMPUTER CALCULATIONS; DOSE RATES; M CODES; NATURAL URANIUM; SHIELDING
- Descriptors DEC
- ACTINIDES; COMPUTER CODES; ELEMENTS; HEAVY WATER MODERATED REACTORS; METALS; POWER REACTORS; PRESSURE TUBE REACTORS; REACTORS; THERMAL REACTORS; URANIUM
Optional Information
- Notes
- 4 refs, 5 figs, 2 tabs