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MIGSHIELD: A new model-based interactive point kernel gamma ray shielding package for virtual environment

  • Li, Mengkun;Xu, Zhihui;Li, Wei;Yang, Jun;Yang, Ming;Lu, Hongxin;Dai, Xinyu
    • Nuclear Engineering and Technology
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    • v.52 no.7
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    • pp.1557-1564
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    • 2020
  • In this paper, 3D model-based interactive gamma ray shielding package (MIGSHIELD) is developed in virtual reality platform for windows operating system. In MIGSHIELD, the computational methodology is based on point kernel algorithm (PK), several key parameters of PK are obtained using new technique and new methods. MIGSHIELD has interactive capability with virtual world. The main features made in the MIGSHIELD are (i) handling of physical information from virtual world, (ii) handling of arbitrary shapes radioactive source, (iii) calculating the mean free path of gamma ray, (iv) providing interactive function between PK and virtual world, (v) making better use of PK for virtual simulation, (vi) plug and play. The developed package will be of immense use for calculations involving radiation dose assessment in nuclear safety and contributing to fast radiation simulation for virtual nuclear facilities.

Effects of Outside Repair Welding on the Crack Growth in the Surge Nozzle Weld on the Hot Leg Side in a Nuclear Power Plant (외면 보수 용접이 원전 고온관 밀림노즐에서의 결함성장에 미치는 영향)

  • Na, Kyung-Hwan;Yun, Eun-Sub;Park, Young-Sheop
    • Journal of Welding and Joining
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    • v.29 no.2
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    • pp.34-39
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    • 2011
  • Nickel-based austenitic alloys such as Alloy 82 and 182 had been employed as the weld metals in nuclear power plants (NPPs) due to their high corrosion resistance as well as good mechanical properties. However, since the 2000s, the occurrence of primary water stress corrosion cracking has been reported in conjunction with these alloys in domestic and oversea NPPs. In the present work, we assumed an imaginary crack at the inner surface of a surge nozzle weld that had previously experienced the outside repair welding, and constructed its finite element model. Finite element analysis was performed with respect to the heat transfer, and then to the residual stress for obtaining the total applied stress distributions. These stress distributions were finally converted to the stress intensity factors for estimating crack growth rate. From the comparison of crack growth rate curves for the cases of no repair welding and outside repair welding, it was found that the outside repair welding did not exhibit negative effect on the crack growth for the surge nozzle under consideration in this work; in both cases, the cracks stopped growing before they became the through-wall cracks.

Risk Assessment of Integrated Leak Rate Test(ILRT) Extension for Korea Standard Nuclear Power Plant (한국표준형원전의 격납건물종합누설률 시험 주기연장에 대한 리스크 평가)

  • Chi, Moon-Goo;Hwang, Seok-Won;Oh, Ji-Yong
    • Journal of the Korean Society of Safety
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    • v.26 no.5
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    • pp.99-104
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    • 2011
  • An ILRT Interval for a nuclear power plant in Korea was extended from once in five years to once in ten years. Therefore, it is necessary to evaluate risk impact for ILRT interval extensions. In this paper, input data were generated for the reference plants, KSNP, using raw data such as meteorological data, population distribution data and source term data. And, using MACCS II code the risk impact assessment was performed based on the two methodologies of NUREG-1493 and NEI Interim Report. The risk impact derived from an ILRT interval extension was identified not to be significant. It is considered to apply this study and results to making an accident management plan and safety goal, and to the field of public acceptance.

Reactivity balance for a soluble boron-free small modular reactor

  • van der Merwe, Lezani;Hah, Chang Joo
    • Nuclear Engineering and Technology
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    • v.50 no.5
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    • pp.648-653
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    • 2018
  • Elimination of soluble boron from reactor design eliminates boron-induced reactivity accidents and leads to a more negative moderator temperature coefficient. However, a large negative moderator temperature coefficient can lead to large reactivity feedback that could allow the reactor to return to power when it cools down from hot full power to cold zero power. In soluble boron-free small modular reactor (SMR) design, only control rods are available to control such rapid core transient. The purpose of this study is to investigate whether an SMR would have enough control rod worth to compensate for large reactivity feedback. The investigation begins with classification of reactivity and completes an analysis of the reactivity balance in each reactor state for the SMR model. The control rod worth requirement obtained from the reactivity balance is a minimum control rod worth to maintain the reactor critical during the whole cycle. The minimum available rod worth must be larger than the control rod worth requirement to manipulate the reactor safely in each reactor state. It is found that the SMR does have enough control rod worth available during rapid transient to maintain the SMR at subcritical below k-effectives of 0.99 for both hot zero power and cold zero power.

Neutron dose rate analysis of the new CONSTOR® storage cask for the RBMK-1500 spent nuclear fuel

  • Narkunas, Ernestas;Smaizys, Arturas;Poskas, Povilas;Naumov, Valerij;Ekaterinichev, Dmitrij
    • Nuclear Engineering and Technology
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    • v.53 no.6
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    • pp.1869-1877
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    • 2021
  • This paper presents the neutron dose rate analysis of the new CONSTOR® RBMK-1500/M2 storage cask intended for the spent nuclear fuel storage at Ignalina Nuclear Power Plant in Lithuania. These casks are designed to be stored in a new "closed" type interim storage facility, with the capacity to store up to 202 CONSTOR® RBMK-1500/M2 casks. In 2016 y, the "hot trials" of this new facility were conducted and 10 CONSTOR® RBMK-1500/M2 casks loaded with the spent nuclear fuel were transported to the dedicated storage places in this facility. During "hot trials", the dose rate measurements of the CONSTOR® RBMK-1500/M2 casks were performed as the dose rate is one of the critical parameter to control and it must be below design (and safety) criteria. Therefore, having the actual data of the spent nuclear fuel characteristics, the neutron dose rate modeling of the CONSTOR® RBMK-1500/M2 cask loaded with this particular fuel was also performed. Neutron dose rate modeling was performed using MCNP 5 computer code with very detailed geometrical representation of the cask and the fuel. The obtained modeling results were compared with the measurement results and it was revealed, that modeling results are generally in good agreement with the measurements.

Radioactive effluents released from Korean nuclear power plants and the resulting radiation doses to members of the public

  • Kong, Tae Young;Kim, Siyoung;Lee, Youngju;Son, Jung Kwon;Maeng, Sung Jun
    • Nuclear Engineering and Technology
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    • v.49 no.8
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    • pp.1772-1777
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    • 2017
  • Korean nuclear power plants (NPPs) periodically evaluate the radioactive gaseous and liquid effluents released from power reactors to protect the public from radiation exposure. This paper provides a comprehensive overview of the release of radioactive effluents from Korean NPPs and the effects on the annual radiation doses to the public. The amounts of radioactive effluents released to the environment and the resulting radiation doses to members of the public living around NPPs were analyzed for the years 2011-2015 using the Korea Hydro & Nuclear Power Co., Ltd's annual summary reports of the assessment of radiological impact on the environment. The results show that tritium was the primary contributor to the activity in both gaseous and liquid effluents. The averages of effective doses to the public were approximately on the order of 103mSv or 102mSv. Therefore, even though Korean NPPs discharged some radioactive materials into the environment, all effluents were within the regulatory safety limits and the resulting doses were much less than the dose limits.

Systems Engineering Process Approach to the Probabilistic Safety Assessment for a Spent Fuel Pool of a Nuclear Power Plant (사용후핵연료저장조의 확률론적안전성평가 수행을 위한 시스템엔지니어링 프로세스 적용 연구)

  • Choi, Jin Tae;Cha, Woo Chang
    • Journal of the Korean Society of Systems Engineering
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    • v.17 no.2
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    • pp.82-90
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    • 2021
  • The spent fuel pool (SFP) of a nuclear power plant functions to store the spent fuel. The spent fuel pool is designed to properly remove the decay heat generated from the spent fuel. If the cooling function is lost and proper operator action is not taken, the spent fuel in the storage pool can be damaged. Probabilistic safety assessment (PSA) is a safety evaluation method that can evaluate the risk of a large and complex system. So far, the probabilistic safety assessment of nuclear power plants has been mainly performed on the reactor. This study defined the requirements and the functional architecture for the probabilistic safety assessment of the spent fuel pool (SFP-PSA) by applying the systems engineering process. And, a systematic and efficient methodology was defined according to the architecture.