iB1350: A Generation III.7 Reactor After the Fukushima Daiichi Accident

Author(s):  
Takashi Sato ◽  
Keiji Matsumoto ◽  
Kenji Hosomi ◽  
Keisuke Taguchi

iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and Western European Nuclear Regulation Association safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged station blackout without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven Advance Boiling Water Reactor (ABWR) design. The nuclear steam supply system is exactly the same as that of the current ABWR. As for safety design it has a double cylinder reinforced concrete containment vessel (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double fission product confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a severe accident (SA). It has a large volume to hold hydrogen, a core catcher, a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a design basis accident, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. The IC/PCCS pools have enough capacity for 7-day grace period. The IC/PCCS heat exchangers, core and spent fuel pool are enclosed inside the containment vessel (CV) building and protected against a large airplane crash. The iB1350 can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. The CV diameter is smaller than that of most PWR resulting in a smaller R/B. Each active safety division includes only one emergency core cooling system (ECCS) pump and one emergency diesel generator (EDG). Therefore, a single failure of the EDG never causes multiple failures of ECCS pumps in a safety division. The iB1350 is based on the proven ABWR technology and ready for construction. No new technology is incorporated but design concept and philosophy are initiative and innovative.

Author(s):  
Takashi Sato ◽  
Keiji Matsumoto ◽  
Toshikazu Kurosaki ◽  
Mitsuhiro Maida

The paper presents study result on an innovative, intelligent and inexpensive BWR (iBR). It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and WENRA safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iBR can survive passively such devastation and a very prolonged SBO without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven ABWR design. NSSS is exactly the same as that of the current ABWR. As for safety design it has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double FP confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a SA. It has a large volume to hold hydrogen, a core catcher, a passive flooding system and a passive containment cooling system (PCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a DBA, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the PCCS for a SA. IC/PCCS pools have enough capacity for 7 day grace period. The IC/PCCS, core and spent fuel pool are enclosed inside the CV building and protected against a large airplane crash. The iBR can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. CV diameter is smaller than that of most PWR resulting in the smaller R/B. The iBR is based on the proven ABWR technology and ready for construction. No new technology is incorporated but design concept and philosophy are initiative and innovative.


Author(s):  
Takashi Sato ◽  
Keiji Matsumoto ◽  
Kenji Hosomi ◽  
Keisuke Taguchi

iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the only Generation III.7 reactor incorporating Fukushima lessons learned and complying with Western European Nuclear Regulation Association (WENRA) safety objectives. It is about twice safer than any existing Gen III.5 reactors. It has 7-day grace period for SBO and SA without containment venting. It enables no evacuation and no long-term relocation in SA. It, however, is based on the well-established proven ABWR. The NSSS and TI are exactly the same as those of the existing ABWR. The iB1350 only enhanced the ABWR safety by adding an outer well (OW) as additional PCV volume, built-in passive safety systems (BiPSS) for SA, DEC systems and an APC shield dome over the containment. The BiPSS include an isolation condenser (IC), an innovative passive containment cooling system (iPCCS), in-containment filtered venting system (IFVS), and innovative core catcher (iCC). All the BiPSS are embedded and protected in the containment building against APC. No specialized safety features remote from the R/B are necessary, which reduces plant cost. The primary system has only one integrated RPV. There are no SGs, no pressurizer, no core makeup tanks, no accumulators, no hot legs, and no cold legs. The iB1350 consists of only one integrated RPV and passive safety systems inside the containment building. This configuration is simpler than the simplest large PWR and as simple as SMR. While SMR have rather small outputs, the iB1350 has 1350 MWe output. It is simple, large and economic. As for the safety design it has an in-depth hybrid safety system (IDHS). The IDHS consists of 4 division active safety systems for DBA, 1 or 2 division active safety systems for DEC and the built-in passive safety systems (BiPSS) for SA. The IDHS is originally based on the four levels of safety that have provided an explicit fourth defense level against devastating external events even before 3.11. It also can be explained along with WENRA Defense in Depth (DiD). It is said that independence between DiD levels are important. However, there are many exceptions for independence between DiD levels. For example, SCRAM is used in DiD2, DiD3a and DiD3b. Any DiD that allows exceptions of independence of DiD levels is fake. The iB1350 is rather based on the three levels of safety proposed by Clifford Beck (AEC, 1967). There is complete independence between level 2 (core systems) and level 3 (containment systems) without any exceptions of independence. DiD without exceptions of independence is a real DiD. Only passive safety reactors can meet the real DiD.


Author(s):  
Tanaka Go ◽  
Sato Takashi ◽  
Komori Yuji ◽  
Matsumoto Keiji

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.


2015 ◽  
Vol 83 ◽  
pp. 35-42 ◽  
Author(s):  
Jun Yang ◽  
Jaehyok Lim ◽  
Sung Won Choi ◽  
Doo Yong Lee ◽  
Somboon Rassame ◽  
...  

Author(s):  
Jue Yang ◽  
Xuenong Zhu ◽  
Xiangang Fu ◽  
Wei Cai ◽  
Jie Ye ◽  
...  

Developing the advanced nuclear power plant design to meet the demanding safety, efficiency and environmental goals of electric utilities requires great efforts. In this paper, a design of the safety systems for the large-power PWR units is introduced, which is deemed a optimal combination of the passive safety systems with the active safety systems. The typical design basis accidents are analyzed for this safety system design, such as the Small Break LOCA, SGTR, SLB and Loss of Flow Accidents (LOFA). The results show that the safety systems of the passives combined the actives can mitigate effectively these typical accidents in large-power PWRs. PSA results also show that the passive safety systems contributes to the reduction of the CDF. It is preliminarily concluded that the passive combined active safety system is designed in balance.


2003 ◽  
Author(s):  
M. Ishii ◽  
S. T. Revankar ◽  
T. Downar ◽  
H. J. Yoon Y. Xu ◽  
D. Tinkler ◽  
...  

Author(s):  
Guohua Yan ◽  
Chen Ye

In the entire history of commercial nuclear power so far, only two major accidents leading to damage of reactor core have taken place. One is Three Mile Island (TMT) accident (1979), which is caused by a series of human error, and the other is Chernobyl accident (1986), which is due to the combined reason of design defects and human errors. After TMI and Chernobyl accidents, in order to reduce manpower in operation and maintenance and influence of human errors on reactor safety, consideration is given to utilization of passive safety systems. According to the IAEA definition, passive safety systems are based on natural forces, such as convection and gravity, and stored energy, making safety functions less dependent on active systems and operators’ action. Recently, the technology of passive safety has been adopted in many reactor designs, such as AP1000, developed by Westinghouse and EP1000 developed by European vendor, and so on. AP1000 as the first so-called Generation III+ has received the final design approval from US NRC in September 2004, and now being under construction in Sanmen, China. In this paper, the major passive safety systems of AP1000, including passive safety injection system, automatic depressurization system passive residual heat removal system and passive containment cooling system, are described and their responses to a break loss-of-coolant accident (LOCA) are given. Just due to these passive systems’ adoption, the nuclear plant can be able to require no operator action and offsite or onsite AC power sources for at least 72h when one accident occurs, and the core melt and large release frequencies are significantly below the requirement of operating plants and the NRC safety goals.


Author(s):  
Luben Sabotinov ◽  
Borislav Dimitrov ◽  
Giovanni B. Bruna

The paper presents the methodology adopted to assess the Interim Safety Analysis Report (ISAR) of the Belene NPP in the framework of the contract between the Bulgarian Nuclear Regulatory Authority (BNRA) and RISKAUDIT (IRSN&GRS). It stresses the in-depth analysis carried-out for several relevant-to-safety issues and illustrates in some detail the investigation of the Large Break Loss of Coolant Accident (LB LOCA) with loss of power and failure of the active part of the Emergency Core Cooling System (High Pressure and Low Pressure Safety Injection pumps), performed with the French best estimate thermal-hydraulic code CATHARE. The role, problems and efficiency of the passive and active safety systems during the accident scenarios are discussed. Finally, the main conclusions of the safety evaluation of the Belene NPP project are summarized.


Author(s):  
Linsen Li ◽  
Feng Shen ◽  
Mian Xing ◽  
Zhan Liu ◽  
Zhanfei Qi

A small Pressurized Water Reactor (PWR) with compact primary system and passive safety feature, which is called Compact Small Reactor (CSR), is under pre-conceptual design and development. For the purpose of preliminary assessment of the primary coolant system and capability evaluation of the passive safety system, a detailed thermal-hydraulic (T-H) system model of the CSR was developed. Several design-basis accidents, including feedwater line break, double ended direct vessel injection line break (one of the small-break Loss Of Coolant Accidents, LOCA) and etc, are selected and simulated so as to evaluate and further optimize the design of passive safety systems, especially the passive core cooling system. The results of preliminary accident analysis show that the passive safety systems are basically capable of mitigating the accidents and protecting the reactor core from severe damage. Further research will be focused on the optimization of pre-conceptual design of the thermal-hydraulic system and the passive core cooling system.


2007 ◽  
Vol 237 (18) ◽  
pp. 1999-2005 ◽  
Author(s):  
Jose N. Reyes ◽  
John Groome ◽  
Brian G. Woods ◽  
Eric Young ◽  
Kent Abel ◽  
...  

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