Relevant bibliographies by topics / Water cooled reactors

Contents

  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Water cooled reactors'

Author: Grafiati
Published: 4 June 2021
Last updated: 1 August 2024

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Journal articles on the topic "Water cooled reactors"

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Cai, Jiejin, Claude Renault, and Junli Gou. "Supercritical Water-Cooled Reactors." Science and Technology of Nuclear Installations 2014 (2014): 1–2. http://dx.doi.org/10.1155/2014/548672.

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Kirillov, P. L. "Supercritical water cooled reactors." Thermal Engineering 55, no. 5 (2008): 361–64. http://dx.doi.org/10.1134/s0040601508050017.

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Chereshkov, Dimitar G., Mikhail Yu Ternovykh, Georgiy V. Tikhomirov, and Alexander A. Ryzhkov. "Nuclear data uncertainty on generation IV fast reactors criticality calculations analysis comparison." Nuclear Energy and Technology 9, no. 3 (2023): 157–62. http://dx.doi.org/10.3897/nucet.9.111919.

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The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO2, MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-
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Sridharan, Kumar. "Understanding How Materials Corrode in Nuclear Reactors." AM&P Technical Articles 172, no. 1 (2014): 17–20. http://dx.doi.org/10.31399/asm.amp.2014-01.p017.

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Abstract The corrosion of structural materials and control of coolant chemistry are key factors that impact the lifetime of nuclear reactors and the development of future reactors. This article reviews corrosion issues in four reactor concepts that use vastly different coolant types: light water reactors, lead fast reactors, high temperature gas-cooled reactors, and fluoride salt-cooled high temperature reactors.
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Sutopo, Catur Febriyanto, and Arifin M. Susanto. "Kajian pembentukan peraturan mengenai sistem pendingin reaktor dan sistem terkait untuk reaktor berpendingin gas." Jurnal Pengawasan Tenaga Nuklir 1, no. 2 (2021): 11–19. http://dx.doi.org/10.53862/jupeten.v1i2.014.

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IN 2021, BAPETEN, AS THE REGULATORY BODY, IS ESTABLISHING A BAPETEN REGULATION REGARDING THE REACTOR COOLANT SYSTEM AND RELATED SYSTEMS, WHICH CURRENTLY ARE NOT YET AVAILABLE. Therefore, it is crucial to establish the BAPETEN Regulation. Based on the reasons, before setting the BAPETEN Regulation, it is necessary to conduct a study that is expected to provide a more comprehensive description and provide recommendations on what things need to be regulated in the BAPETEN Regulation, especially for gas-cooled reactors. The method used in this study is a literature study from various relevant refe
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Zhang, Zhen Guo, and Jie Xie. "The Application of New Materials in New Nuclear Reactors." Advanced Materials Research 1030-1032 (September 2014): 197–200. http://dx.doi.org/10.4028/www.scientific.net/amr.1030-1032.197.

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This paper combines the technical features and operation condition of the new nuclear reactors, such as sodium cooled fast reactor (SFR), supercritical water cooled reactor (SCWR), very high temperature reactor (VHTR), molten salt reactor (MSR) and nuclear fusion reactor, and analyzes the performance requirements of materials, it introduces the current application and prospects of new materials in different new nuclear reactors.
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Huang, Hai, Chenxi Cao, Yue Wang, Youwei Yang, Jianning Lv, and Jing Xu. "Model-Based Analysis for Ethylene Carbonate Hydrogenation Operation in Industrial-Type Tubular Reactors." Processes 10, no. 4 (2022): 688. http://dx.doi.org/10.3390/pr10040688.

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Hydrogenation of ethylene carbonate (EC) to co-produce methanol (MeOH) and ethylene glycol (EG) offers an atomically economic route for CO2 utilization. Herein, aided with bench and pilot plant data, we established engineering a kinetics model and multiscale reactor models for heterogeneous EC hydrogenation using representative industrial-type reactors. Model-based analysis indicates that single-stage adiabatic reactors, despite a moderate temperature rise of 12 K, suffer from a narrow operational window delimited by EC condensation at lower temperatures and intense secondary EG hydrogenation
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Gužela, Štefan, František Dzianik, Martin Juriga, and Juraj Kabát. "Shell and Tube Heat Exchanger – the Heat Transfer Area Design Process." Strojnícky casopis – Journal of Mechanical Engineering 67, no. 2 (2017): 13–24. http://dx.doi.org/10.1515/scjme-2017-0014.

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AbstractNowadays, the operating nuclear reactors are able to utilise only 1 % of mined out uranium. An effective exploitation of uranium, even 60 %, is possible to achieve in so-called fast reactors. These reactors commercial operation is expected after the year 2035. Several design configurations of these reactors exist. Fast reactors rank among the so-called Generation IV reactors. Helium-cooled reactor, as a gas-cooled fast reactor, is one of them. Exchangers used to a heat transfer from a reactor active zone (i.e. heat exchangers) are an important part of fast reactors. This paper deals wi
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Marcus, Gail H. "Heeding the Lessons of History." Mechanical Engineering 134, no. 07 (2012): 28–33. http://dx.doi.org/10.1115/1.2012-jul-1.

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This article focuses on learning from the successes and failures of the first-generation reactor development. Reactor designs have evolved over time to meet increasingly rigorous demands for safety and to take advantage of technological developments to improve their economics, but these changes have been piecemeal. Although light-water reactors are the most common reactor technology in use today, heavy-water reactors were actually developed earlier. The earliest demonstration of a heavy-water moderated and cooled reactor took place in May 1944 at Argonne. The reasons for the domination of wate
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Englert, Matthias, and Christoph Pistner. "Technological readiness of alternative reactor concepts." Safety of Nuclear Waste Disposal 2 (September 6, 2023): 209. http://dx.doi.org/10.5194/sand-2-209-2023.

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Abstract. Some alternative reactor concepts are frequently discussed in the literature as relevant for the disposal of high-level radioactive waste. We present results on the technological readiness of and research and development on those reactor concepts from a project commissioned by the Federal Office for the Safety of Nuclear Waste Management on novel reactors (Pistner et al., 2023). The focus is on 7 technology lines and 10 specific reactor concepts within those lines. These are sodium-cooled fast reactors (SFRs), lead-cooled fast reactors (LFRs), gas-cooled fast reactors (GFRs), very hi
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Dissertations / Theses on the topic "Water cooled reactors"

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Zhao, Jiyun Ph D. Massachusetts Institute of Technology. "Stability analysis of supercritical water cooled reactors." Thesis, Massachusetts Institute of Technology, 2005. http://hdl.handle.net/1721.1/34651.

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Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2005.<br>Includes bibliographical references (p. 269-277).<br>The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500°C average core exit). The high coolant temperature as it leaves the reactor core gives the SCWR the potential for high thermal efficiency (45%). However, near the supercritical thermodynamic point, coolant density is very sensitive to temperature which raises concerns about instabilities in the supe
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Chan, Ping-lam. "Development of a simulation model for PWR reactor coolant system /." [Hong Kong] : University of Hong Kong, 1989. http://sunzi.lib.hku.hk/hkuto/record.jsp?B1273617X.

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Mowrey, James A. "Control system modeling for a boiling water reactor." Thesis, Georgia Institute of Technology, 1995. http://hdl.handle.net/1853/17083.

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陳炳林 and Ping-lam Chan. "Development of a simulation model for PWR reactor coolant system." Thesis, The University of Hong Kong (Pokfulam, Hong Kong), 1989. http://hub.hku.hk/bib/B31209130.

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Khamis, Ibrahim Ahmad. "Simulation of nuclear power plant pressurizers with application to an inherently safe reactor." Diss., The University of Arizona, 1988. http://hdl.handle.net/10150/184378.

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Pressurizer modeling for predicting the dynamic pressure of the PIUS system is presented. The transient behavior of this model for the PIUS system was investigated. The validity of this model for the PIUS system is limited to transients that are neither too large nor too long in duration. For example, the model is not capable of describing events following a complete loss of liquid for the pressurizer. However, the model can be used for qualitative prediction of the PIUS system behavior for a wide variety of severe transients. A review of pressurizer modeling indicates that the neglecting of t
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CARATIN, REINALDO L. "Estudo da imobilização de rejeitos radioativos em matrizes asfálticas e resíduos elastoméricos utilizando a técnica de microondas." reponame:Repositório Institucional do IPEN, 2007. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11574.

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Made available in DSpace on 2014-10-09T12:53:25Z (GMT). No. of bitstreams: 0<br>Made available in DSpace on 2014-10-09T14:09:32Z (GMT). No. of bitstreams: 1 12213.pdf: 5005382 bytes, checksum: c4bde457760b3a6d6f53b64c21e33010 (MD5)<br>Dissertação (Mestrado)<br>IPEN/D<br>Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Khoza, Samukelisiwe Nozipho Purity. "Characteristic behaviour of pebble bed high temperature gas-cooled reactors during water ingress events / Samukelisiwe Nozipho Purity Khoza." Thesis, North-West University, 2012. http://hdl.handle.net/10394/8706.

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The effect of water ingress in two pebble bed high temperature gas-cooled reactors i.e. the PBMR-200 MWthermal and the PBMR-400 MWthermal were simulated and compared using the VSOP 99/05 suite of codes. To investigate the effect of this event on reactivity, power profiles and thermal neutron flux profiles, the addition of partial steam vapour pressures in stages up to 400 bar into the primary circuit for the PBMR-400 and up to 300 bar for the PBMR- 200 was simulated for both reactors. During the simulation, three scenarios were simulated, i.e. water ingress into the core only, water ingress in
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UMBEHAUN, PEDRO E. "Desenvolvimento de um elemento combustível instrumentado para o reator de pesquisa IEA-R1." reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26935.

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Submitted by Marco Antonio Oliveira da Silva (maosilva@ipen.br) on 2016-12-21T17:48:50Z No. of bitstreams: 0<br>Made available in DSpace on 2016-12-21T17:48:50Z (GMT). No. of bitstreams: 0<br>Após o aumento de potência do reator IEA-R1 de 2 MW para 5 MW observou-se um aumento da taxa de corrosão nas placas laterais de alguns elementos combustíveis e algumas dúvidas surgiram com relação ao valor de vazão utilizada nas análises termo-hidráulicas. A fim de esclarecer e medir a distribuição de vazão real pelos elementos combustíveis que compõe o núcleo do reator IEA-R1, um elemento combustível pr
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MELLO, JOSÉ ROBERTO de. "Regulamentação do sistema elétrico do reator IEA-R1." reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26928.

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Submitted by Marco Antonio Oliveira da Silva (maosilva@ipen.br) on 2016-12-21T12:55:48Z No. of bitstreams: 0<br>Made available in DSpace on 2016-12-21T12:55:48Z (GMT). No. of bitstreams: 0<br>O reator IEA-R1 do Instituto de Pesquisas Energéticas e Nucleares (IPENCNEN/ SP) é um reator de pesquisa tipo piscina aberta, projetado e construído pela empresa norte-americana \"Babcock & Wilcox\", tendo, como refrigerante e moderador, água leve deionizada e berílio e grafite como refletores. Até cerca de 1988, os sistemas de segurança do reator recebiam alimentação de uma única fonte de energia. Nos a
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SANTOS, DIOGO F. dos. "Caracterização dos campos neutrônicos obtidos por meio de armadilhas de nêutrons a partir da utilização de água pesada (D2O) no interior do núcleo do reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23825.

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Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2015-07-23T11:04:17Z No. of bitstreams: 0<br>Made available in DSpace on 2015-07-23T11:04:17Z (GMT). No. of bitstreams: 0<br>Dissertação (Mestrado em Tecnologia Nuclear)<br>IPEN/D<br>Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Books on the topic "Water cooled reactors"

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Oka, Yoshiaki. Super light water reactors and super fast reactors: Supercritical-pressure light water cooled reactor. Springer, 2010.

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Mori, Hideo, and Yoshiaki Oka. Supercritical-pressure light water cooled reactors. Springer, 2014.

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Oka, Yoshiaki, and Hideo Mori, eds. Supercritical-Pressure Light Water Cooled Reactors. Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55025-9.

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Sapra, M. K. Design, development, and testing of 25 NB size accumulator isolation passive value (AIPV) for advanced heavy water reactor (AHWR). Bhabha Atomic Research Centre, 2010.

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Ross, W. E. CATHENA model of shield circuit ITER. CFFTP, 1992.

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Ross, W. E. Thermosyphoning analysis with the CATHENA model of the blanket & first wall cooling loop for the SEAFP reactor design. CFFTP, 1994.

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Ross, W. E. Analysis of loss of electrical power with the CATHENA model of the blanket & first wall cooling loop for the SEAFP reactordesign. CFFTP, 1994.

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J, Elliot A., and International Conference on Water Chemistry of Nuclear Reactor Systems. (6th : 1992 : Bournemouth, England)., eds. The temperature dependence of G-values for aqueous solutions irradiated with a 23 MeV 2H+ beam. System Chemistry & Corrosion Branch, Chalk River Laboratories, 1992.

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Kveton, O. K. ITER ISS system alternative specification study revision 1.0. Canadian Fusion Fuels Technology Project, 1990.

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Ross, W. E. Modified Cathena Model of a Shield Circuit for ITER. CFFTP, 1993.

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Book chapters on the topic "Water cooled reactors"

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Nuttall, William J. "Water-Cooled Reactors." In Nuclear Renaissance, 2nd ed. CRC Press, 2022. http://dx.doi.org/10.1201/9781003038733-9.

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Schulenberg, Thomas. "Innovative Water-Cooled Reactors." In The fourth generation of nuclear reactors. Springer Berlin Heidelberg, 2022. http://dx.doi.org/10.1007/978-3-662-64919-0_3.

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Kale, R. D., and B. K. Sreedhar. "Main coolant pumps for water-cooled reactors." In Centrifugal Pumps for Sodium Cooled Reactors. CRC Press, 2023. http://dx.doi.org/10.1201/9781003460350-9.

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Oka, Yoshiaki, and Shinichi Morooka. "Reactor Design and Safety." In Supercritical-Pressure Light Water Cooled Reactors. Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55025-9_2.

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Oka, Yoshiaki. "Introduction and Overview." In Supercritical-Pressure Light Water Cooled Reactors. Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55025-9_1.

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Mori, Hideo, Yoshinori Hamamoto, Koichiro Ezato, Kazuyuki Takase, and Takeharu Misawa. "Thermal Hydraulics." In Supercritical-Pressure Light Water Cooled Reactors. Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55025-9_3.

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Watanabe, Yutaka, Hiroshi Abe, and Yoshiaki Oka. "Materials." In Supercritical-Pressure Light Water Cooled Reactors. Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55025-9_4.

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Katsumura, Yosuke, and Yusa Muroya. "Material–Coolant Interactions." In Supercritical-Pressure Light Water Cooled Reactors. Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55025-9_5.

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Vaidyanathan, G. "Modeling of Large Sodium–Water Reaction." In Dynamic Simulation of Sodium Cooled Fast Reactors. CRC Press, 2022. http://dx.doi.org/10.1201/9781003283188-13.

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Cox, B. "Modelling the Corrosion of Zirconium Alloys in Nuclear Reactors Cooled by High Temperature Water." In Modelling Aqueous Corrosion. Springer Netherlands, 1994. http://dx.doi.org/10.1007/978-94-011-1176-8_9.

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Conference papers on the topic "Water cooled reactors"

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Duffey, R., L. K. H. Leung, D. Martin, B. Sur, and M. Yetisir. "A Supercritical Water-Cooled Small Modular Reactor." In ASME 2011 Small Modular Reactors Symposium. ASMEDC, 2011. http://dx.doi.org/10.1115/smr2011-6548.

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A new small modular reactor (SMR) is proposed for a 300 MW(e) nuclear generating station. It is referred as the SuperSafe© Reactor (SSR) and is a scaled down version of the Canadian supercritical water-cooled reactor (SCWR), which is designed to operate at supercritical conditions (pressure of 25 MPa and fluid temperature of 625°C) at the turbine inlet with high cycle efficiencies (∼45%). The supercritical turbine technology and associated components used in the balance of plant (BOP) are similar to those in existing supercritical fossil-fired plants. The reactor core consists of fuel channels
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Aminov, Rashid. "EMERGENCY COOLING SYSTEM FOR WATER-COOLED REACTORS." In 17th International Multidisciplinary Scientific GeoConference SGEM2017. Stef92 Technology, 2017. http://dx.doi.org/10.5593/sgem2017/41/s16.001.

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Aminov, Rashid. "EMERGENCY COOLING SYSTEM FOR WATER-COOLED REACTORS." In 18th International Multidisciplinary Scientific GeoConference SGEM2018. Stef92 Technology, 2018. http://dx.doi.org/10.5593/sgem2018/4.1/s16.002.

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Yu, Fei, Yueming Fu, Feng Zhu, and Mingqian Zhang. "Feasibility Study of Passive Water-Cooled Residual Heat Removal Scheme for Small Reactors." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-92779.

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Abstract With the development of nuclear energy and the need for environmental protection, some advanced small reactors have gradually entered the public’s field of vision. The passive natural circulation is an important part of the inherent safety feature for advanced small reactors. Based on the limitations of the site environment, the large heat exchange area of the air cooler, and the high cost of equipment construction for the air-cooled waste heat removal scheme of a small reactor, this paper studies the water-cooled passive residual heat removal scheme and analyzes the water-cooled wate
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Guo, Zihao, Guang Wang, Genghua Su, and Xiuan Shi. "Light Water-Cooled High-Performance Research Reactor Physics Design and Analysis." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-93256.

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Abstract Research Reactor can provide a platform for advanced reactors’ fuel and material irradiation test, medical isotopes production and neutron science research, which plays an important role in nuclear energy and technology domain. Nowadays, most of research reactor in worldwide are reaching the retirement and decontamination stage. For the increasing demand of the material irradiation and medical isotopes in the world, several new and advanced research reactors should be constructed to replace the void in the future. Differed from the design scheme of the traditional PWR, that of researc
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Mesina, George, and Robert Martin. "Recent RELAP5-3D Developments for Water and Gas-Cooled Reactors." In 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20). American Nuclear Society, 2023. http://dx.doi.org/10.13182/nureth20-40349.

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Grande, Lisa, Bryan Villamere, Adrianexy Rodriguez-Prado, Sally Mikhael, Leyland Allison, and Igor Pioro. "Thermal Aspects of Using Thoria Fuel in SuperCritical Water-Cooled Nuclear Reactors." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75969.

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SuperCritical Water-cooled Reactors (SCWRs) are one of six next-generation nuclear-reactor design options under consideration worldwide. These nuclear-reactor design options are included in the major international treaties such as: Generation IV International Forum (GIF) and INternational PROject on innovative nuclear reactors and fuels (INPRO). SCWR coolant is light water, which operates at supercritical pressures and temperatures. Typical SCWR coolant operating parameters are 25 MPa and 350–625°C. These SCWR operating conditions significantly increase the thermal efficiency of a SCW Nuclear
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Forsberg, Charles W., Per F. Peterson, James E. Cahalan, Jeffrey A. Enneking, and Phil MacDonald. "Refueling Liquid-Salt-Cooled Very High-Temperature Reactors." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89471.

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The liquid-salt-cooled very high-temperature reactor (LS-VHTR), also called the Advanced High-Temperature Reactor (AHTR), is a new reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. Depending upon goals, the peak coolant operating temperatures are between 700 and 1000°C, with reactor outputs between 2400 and 4000 MW(t). Several fluoride salt coolan
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Dragunov, Alexey, Eugene Saltanov, Igor Pioro, Brian Ikeda, Marija Miletic, and Anastasiia Zvorykina. "Investigation of Thermophysical and Nuclear Properties of Prospective Coolants for Generation-IV Nuclear Reactors." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16020.

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Recently, a group of countries has initiated an international collaboration, the Generation IV International Forum (GIF), to develop the next-generation nuclear reactors. The GIF program has narrowed the design options of nuclear reactors to the following six concepts: 1) SuperCritical-Water-cooled Reactor (SCWR); 2) Sodium-cooled Fast Reactor (SFR); 3) Lead-cooled Fast Reactor (LFR); 4) Molten Salt Reactor (MSR); 5) Gas-cooled Fast Reactor (GFR); and 6) Very-High-Temperature Reactor (VHTR); The purpose of this paper is to compare main thermophysical, corrosion, and neutronic properties of the
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Zhou, Yifeng, Paul Ponomaryov, Cristina Mazza, and Igor Pioro. "Investigation of Thermal Efficiency of Generic Pressure-Channel Reactors With Steam Reheat." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60442.

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Currently, i.e., in 2016, 4361 nuclear-power reactors operate in the world. 96.6% of these reactors are water-cooled (373 reactors (280 PWRs, 78 BWRs and 15 LGRs are cooled with light water and 48 reactors — PHWRs are cooled with heavy water. 15% of all water-cooled reactors are pressure-channel or pressure-tube design, the rest — pressure-vessel design. All current NPPs with water-cooled reactors have relatively low thermal efficiencies within 30–36% compared to that of current NPPs with AGRs (42%) and SFR (40%) and compared to that of modern advanced thermal power plants: combined-cycle plan
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Reports on the topic "Water cooled reactors"

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Macdonald, Dgiby, Mirna Urquidi-Macdonald, and Jonathan Pitt. Electrochemistry of Water-Cooled Nuclear Reactors. Office of Scientific and Technical Information (OSTI), 2006. http://dx.doi.org/10.2172/890516.

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Forsberg, C. (Advanced technologies for water-cooled reactors). Office of Scientific and Technical Information (OSTI), 1988. http://dx.doi.org/10.2172/6888806.

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Pimblott, S. M. Effects of Water Radiolysis in Water Cooled Reactors. Technical progress report. Office of Scientific and Technical Information (OSTI), 2000. http://dx.doi.org/10.2172/761635.

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Macdonald, Digby, Mirna Urquidi-Macdonald, Yingzi Chen, Jiahe Ai, Pilyeon Park, and Han-Sang Kim. Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors. Office of Scientific and Technical Information (OSTI), 2006. http://dx.doi.org/10.2172/896213.

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Pimblott, S. M. Effects of Water Radiolysis in Water Cooled Reactors, NERI Proposal No.99-0010. Office of Scientific and Technical Information (OSTI), 2000. http://dx.doi.org/10.2172/761538.

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Pimblott, S. M. Effects of Water Radiolysis in Water Cooled Reactors, ''NERI Proposal No.99-001C''. Office of Scientific and Technical Information (OSTI), 1999. http://dx.doi.org/10.2172/762090.

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S. M. Pimblott. Effects of Water Radiolysis in Water Cooled Reactors - Nuclear Energy Research Initiative (NERI) Program. Office of Scientific and Technical Information (OSTI), 2000. http://dx.doi.org/10.2172/772760.

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Philip MacDonald, Jacopo Buongiorno, James Sterbentz, et al. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production. Office of Scientific and Technical Information (OSTI), 2005. http://dx.doi.org/10.2172/835712.

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Pimblott, S. M. Effects of Water Radiolysis in Water Cooled Reactors, NERI Proposal No.99-0010. Technical progress report. Office of Scientific and Technical Information (OSTI), 2000. http://dx.doi.org/10.2172/761609.

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Olsen, C. S. Non-fertile fuels development for plutonium and high-enriched uranium dispositioning in water cooled reactors. Office of Scientific and Technical Information (OSTI), 1994. http://dx.doi.org/10.2172/164456.

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