• Nuclear Engineering and Technology
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• v.17 no.2
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• pp.98-104
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• 1985

A number of reactivity devices are distributed in a CANDU-PHWR core to control the power distribution and excess reactivity. The effects of these devices are represented by incremental cross sections in core analysis. The incremental cross sections are generated by the SUPERCELL code using the two-group constant set calculated by the lattice code, WIMS. The incremental cross sections are then assessed for adjusters and zone controller by core simulation. Reactivity worth and channel powers are compared to the reference values. The deviation of reactivity worth and the maximum channel power are less than 0.97% and 0.6%, respectively, for the initial and equilibrium core.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.817-825
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• 2022

A recent benchmarking effort, under the auspices of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA), has been made to evaluate the current state of modeling and simulation tools available to model fluoride salt-cooled high temperature reactors (FHRs). The FHR benchmarking effort considered in this work consists of several cases evaluating the neutronic parameters of a 2D prismatic FHR fuel assembly model using the participants' choice of simulation tools. Benchmark participants blindly submitted results for comparison with overall good agreement, except for some which significantly differed on cases utilizing a molybdenum-bearing control rod. Participants utilizing more recently updated explicit isotopic cross sections had consistent results, whereas those using elemental molybdenum cross sections observed reactivity differences on the order of thousands of pcm relative to their peers. Through a series of supporting tests, the authors attribute the differences as being nuclear data driven from using older legacy elemental molybdenum cross sections. Quantitative analysis is conducted on the control rod to identify spectral, reaction rate, and cross section phenomena responsible for the observed differences. Results confirm the observed differences are attributable to the use of elemental cross sections which overestimate the reaction rates in strong resonance channels.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.791-803
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• 2015

The effective cross sections (XSs) in the direct whole core calculation code nTRACER are evaluated by the equivalence theory-based resonance-integral-table method using the WIMS-based library as an alternative to the subgroup method. The background XSs, as well as the Dancoff correction factors, were evaluated by the enhanced neutron-current method. A method, with pointwise microscopic XSs on a union-lethargy grid, was used for the generation of resonance-interference factors (RIFs) for mixed resonant absorbers. This method was modified by the intermediate-resonance approximation by replacing the potential XSs for the non-absorbing moderator nuclides with the background XSs and neglecting the resonance-elastic scattering. The resonance-escape probability was implemented to incorporate the energy self-shielding effect in the spectrum. The XSs were improved using the proposed method as compared to the narrow resonance infinite massbased method. The RIFs were improved by 1% in $^{235}U$, 7% in $^{239}Pu$, and >2% in $^{240}Pu$. To account for thermal feedback, a new feature was incorporated with the interpolation of pre-generated RIFs at the multigroup level and the results compared with the conventional resonance-interference model. This method provided adequate results in terms of XSs and k-eff. The results were verified first by the comparison of RIFs with the exact RIFs, and then comparing the XSs with the McCARD calculations for the homogeneous configurations, with burned fuel containing a mixture of resonant nuclides at different burnups and temperatures. The RIFs and XSs for the mixture showed good agreement, which verified the accuracy of the RIF evaluation using the proposed method. The method was then verified by comparing the XSs for the virtual environment for reactor applicationbenchmark pin-cell problem, as well as the heterogeneous pin cell containing burned fuel with McCARD. The method works well for homogeneous, as well as heterogeneous configurations.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2008.04a
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• pp.69-75
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• 2008

기포반응도는 월성발전소를 비롯한 CANDU형 원자로의 주된 안전성 쟁점사안으로 끊임없이 논의되어 왔다. 이는 설계기준사고가 노심에서 열에너지 불균형이 원인이 되어 기준이상의 핵연료 파손과 방사성물질 누출로 발전할 위험이 있는 사건들로 정의될 때, 사건 진행 과정에 기포반응도 증가는 조기에 운전중단을 실패할 경우 출력폭주로 이어지므로 사건의 결말이 중대사고로 전환될 위험이 크기 때문이다. 본 연구는 공개된 최신 핵자료인 ENDF/B-VII.0를 NJOY.99로 처리한 연속에너지 반응단면적 라이브러리를 구축하고 MCNP-4C에 접속하여 37봉 천연우라늄 핵연료다발의 표준노심격자에 대한 기포반응도를 시뮬레이션하여, 지금까지 각종문헌에 제시된 값들과 비교, 종합하므로 내제된 불확실성을 추정하는 내용이다. ENDF/B-VII.0 기반 MCNP-4C의 CANDU 노심격자 모델은 동일한 핵자료와 핵종농도를 사용한 WIMS-IAEA 모델과 비교할 때, 초기 노심의 임계도 오차 약 3.51mk가 연소 진행에 따라 $7.5\times10^{-4}mk$/MWD/teU의 비율로 감소하는 것으로 나타났다. 또한 MCNP-4C 예측기포반응도는 초기노심에서 기포율 50% 및 100%에 대해 각각 8.38 및 15.96mk, 평형노심에서 7.68 및 14.72mk로 계산된다. 이는 월성 2, 3, 4 FSAR의 초기노심 및 평형노심에서 100% 기포상태에 대한 값, 약15.0 및 10.6mk와 비교할 때, 초기노심은 약 1.0mk 평형노심은 약4, 1mk 보수적이지만, 다른 연구결과들과는 최대오차 ${\pm}1{\sim}2mk$ 이내에서 잘 일치하는 것으로 평가되었다. 본 연구는 CANDU 노심의 기포반응도 불확실성 요인의 규명 및 영향평가를 위한 노력의 일부로서 앞으로 감속재의 붕산농도 변화, 감속재 및 냉각재의 중수 순도 변화, 기기노화에 의한 격자 구조 및 물성 변화, 중성자속 및 출력 분포 불균형, 반응도조절장치의 위치, 등 주요 설계변수의 변화에 대한 반응도영향 분석연구를 계속할 계획이다.

• The Journal of The Korea Institute of Intelligent Transport Systems
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• v.21 no.5
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• pp.57-66
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• 2022

The Road Management Administration is cracking down on overloaded vehicles by installing low-speed or high-speed WIMs at toll gates and main lines on expressways. However, in recent years, the act of intelligently evading the overloaded-vehicle control system of the Road Management Administration by illegally manipulating the variable axle of an overloaded truck is increasing. In this manipulation, when entering the overloaded-vehicle checkpoint, all axles of the vehicle are lowered to pass normally, and when driving on the main road, the variable axle of the vehicle is illegally lifted with the axle load exceeding 10 tons alarmingly. Therefore, this study developed a technology to detect the state of the variable axle of a truck driving on the road using roadside camera images. In particular, this technology formed the basis for cracking down on overloaded vehicles by lifting the variable axle after entering the checkpoint and linking the vehicle with the account information of the checkpoint. Fundamentally, in this study, the tires of the vehicle were recognized using the Mask RCNN algorithm, the recognized tires were virtually arranged before and after the checkpoint, and the height difference of the vehicle was measured from the arrangement to determine whether the variable axle was lifted after the vehicle left the checkpoint.