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The accurate calculation of pressurized water reactor burned fuel neutron source intensity plays a significant role in the subcritical reactivity measurement. The calculation methods of both spontaneous fission source and (α, n) neutron source were researched in this paper. There were multiple sources of alpha particle in reactor core, including 238Pu, 239Pu, 240Pu, 241Am, 242Cm, 244Cm, etc. Among these sources, the dominant source was coming from 242Cm decay. 242Cm approximation method as well as ratio fitting method were proposed for obtaining (α, n) neutron source intensity. In 242Cm approximation method, all alpha particle sources were neglected with the exception of 242Cm. The ORIGEN code was used to analyze the ratio of (α, n) to the spontaneous fission neutron of 242Cm for different enrichment assemblies at different shutdown time. It is found that the ratio of (α, n) to the spontaneous fission neutron of 242Cm is fixed around 0.191. Therefore, the neutron number density of spontaneous fission neutron is multiplied by the factor of 0-191 to obtain (α, n) neutron source intensity due to 242Cm decay. On account of neglecting all (α, n) neutron source intensity with the exception of 242Cm, it is foreseeable that 242Cm approximation method would underestimate (α, n) neutron source intensity. In ratio fitting method, the ratio of (α, n) to the total neutron source intensity was analyzed. It is found that the ratio of (α, n) to the total neutron source intensity is linearly related to the burunp when the burnup is greater than 20 000 MW·d/tU. Therefore, two fitting methods were proposed to obtain the ratio of (α, n) to the total neutron source intensity. The polynomial fitting method was used with burnup less than 20 000 MW·d/tU, on the other hand, the liner fitting method was adopted with burnup greater than 20 000 MW·d/tU. Based on the in-house core design code package, the burned fuel neutron source calculation module was developed. In order to account for reactor core space effect and burnup history effect, the micro-depletion correction method in core design code package was used to obtain accurate number densities of actinide nuclides. The developed module was verified by calculating fuel assembly examples. Comparing with the reference code, the maximum errors of total source intensity is about 5% provided by both 242Cm approximation method and ratio fitting method. With the developed burned fuel neutron source intensity calculation module, both radial and axial distributions of burned fuel neutron source intensity in the reactor core are obtained. These distributions are of significant importance to ex-core detector signal measurements. The work in this paper is an important technology support for subcritical rod worth measurement development. |
