Advanced Large Water Cooled Reactors A supplement to the IAEA’s Advanced Reactor Information System (ARIS)

September 2015 Advanced Large Water Cooled Reactors

PREFACE )

Despite the adverse ramifications from the accident at the Fukushima Daiichi nuclear power plant, nuclear power is receiving increasing global interest, particularly in Asia. A growing number IAEA of countries are considering building nuclear power plants to meet increasing energy needs of their growing economies while decreasing their greenhouse emissions. In his public lecture at the Singapore Energy Market Authority in January 2015, IAEA Director General Yukiya Amano stated that:

In the four years since the Fukushima Daiichi nuclear accident in Japan, huge ( gency

improvements have been made to nuclear safety all over the world, and there has also A been significant progress in treating and disposing nuclear waste. Remarkable research is being done on new generations of reactors which will be safer and generate less waste.

Member states, both those considering their first nuclear power plant and those with notions to nergy expand their existing programmes, are vitally interested in obtaining current information about E designs for reactors that are deployable now or in the near term. Fulfilling its mission as stated in the original IAEA Statute (Article III.A.3: “To foster the exchange of scientific and technical information on peaceful uses of atomic energy”) the Nuclear Power Division of the IAEA has regularly issued Page intentionally left blank publications on the status of nuclear reactor technology developments. Over the years the manner

of information dissemination has evolved and the latest rendition is offered online in the form of an tomic

online database that presents unbiased, detailed design descriptions for currently available nuclear A power plants. This Advanced Reactor Information System (http://aris.iaea.org) includes reactors of all sizes and types, from evolutionary nuclear plant designs for near term deployment to innovative concepts still under development. Hard copy supplements to the ARIS database focusing on Small Modular Reactors (under 700MWe capacity) and on Fast Reactors have already been published: Advances in Small Modular Reactor (SMR) Technology Developments https://www.iaea.org/NuclearPower/Downloadable/SMR/files/IAEA_SMR_Booklet_2014.pdf

Status of Innovative Fast Reactor Designs and Concepts

https://www.iaea.org/NuclearPower/Downloadable/FR/booklet-fr-2013.pdf nternational I Water Cooled Reactors have played a significant role in the commercial nuclear industry since its inception and currently account for more than 95% of all operating commercial reactors in the world. Of the 67 nuclear reactors now under construction, 64 are water cooled reactors. Therefore, it seems timely and imperative to offer this booklet, which provides an overview of the status of advanced, large (700 MWe or more) water cooled reactors. The objective is to provide Member States with a brief overview of the large nuclear power plants considered currently deployable. It should be regarded as a complementary publication to the ARIS database itself, as well as to the IAEA guidance document for evaluating nuclear power plants, Nuclear Reactor Technology Assessment for Near Term Deployment (IAEA NE Series NP-T-1.10). http://www-pub.iaea.org/books/IAEABooks/8950/Nuclear-Reactor-Technology-Assessment-for- Near-Term-Deployment The IAEA acknowledges the role and contributions of T. Vattappillil in the design, drafting and development of this booklet. The IAEA officer responsible for this publication is M.J. Harper of the Division of Nuclear Power. COPYRIGHT NOTICE Table of Contents ) ) All IAEA scientific and technical publications are protected by the terms of the Universal Copyright Convention as adopted in 1952 (Berne) and as revised in 1972 (Paris). The copyright has since been extended by the World Intellectual Property Organization (Geneva) to include electronic IAEA and virtual intellectual property. Permission to use whole or parts of texts contained in IAEA IAEA publications in printed or electronic form must be obtained form the IAEA Marketing and Sales Contents Unit. Guide 2

If not otherwise stated the IAEA owns the rights to distribute the images in this publication. Evolutionary Power Reactor — EPRTM 4 gency ( gency Open Source Image courtesy, attribution, sources and release licences are listed on page 31. gency ( gency A This document can be shared on non-commercial basis as long as credit is given to the IAEA China’s Advanced PWRs 6 A without implying IAEA endorsement of the resulting product. Advanced Power Reactor 1400b—bAP1000 8 For further information please contact the IAEA Marketing and Sales Unit. ATMEA1TM 10 nergy nergy

E b Advanced Pwr — APWR 12 E

Water-Water Energetic Reactor — VVER 14

Advanced Passive 1000 — AP1000 18 tomic tomic tomic tomic A KERENATM 20 A

Advanced Boiling Water Reactor — ABWR 22

Economic Simplified BWR — ESBWR 24

Enhanced CANDU 6 26 Marketing and Sales Unit, Publishing Section International Atomic Energy Agency Vienna International Centre Indian Pressurized HWR-700 — IPHWR-700 28 PO Box 100 nternational nternational nternational nternational

I Map 30 1400 Vienna, Austria I

Fax: +43 1 2600 29302 Image-Reference 31 Tel.: +43 1 2600 22417 Acronyms and Abbreviations 32 Email: [email protected] Website: http://www.iaea.org/books

Printed by the IAEA in Austria

September 2015 Table containing important reactor data Guide s Capacity: Gross electric output in MWe and reactor thermal output in MWt Design Life: Life of non-replaceable components Each Large, Advance Water Cooled Reactor Temperature: Averaged temperature between core inlet and core outlet under construction or deployable in the Pressure: Reactor operating pressure near future will have a two-page fold. Efficiency: Plant net efficiency Enrichment: Enrichment of new fuel in equilibrium at core reload PWR BWR HWR Full name of reactor design

Acronym for the reactor design Name of design organization

Figures depicting salient aspects of Introduction of reactor reactor design design with key features summarized

Information about type of fuel and fuel cycle related issues Illustration of the primary circuit with large components

Information on operation and construction status of design

Number of Units (as of Sept 2015)

In Operation

Under Flag of the country where Useful information of Construction units are in operation or Information on reactor systems that general nature under construction. relate to NSSS, safety, fuel and oper- Source: IAEA PRIS database ation Unit Names

2 3 4 AREVA upy o ia ncer ntuetto, oto, n operating and control, instrumentation, systems. nuclear vital to supply powerday’s additional an for available, batteries, DC are emergency with along generators diesel backup and days, three for fuel naturalandcirculation. Emergencyhavegeneratorsdiesel enough the systems, for core catchers, gravity driven CDF systems, double containment the severe accident, including concepts, coolant recirculation, containment spray a of unlikelycase very a of consequences the mitigate systemsto defence-in-depth with 10 be to calculated is EPR concert in safety installed the systems, all Considering g. 0.25 to up of acceleration ground with incidences seismic withstand to designed is EPR The buildings, constructed on concrete systems rafts, are related aircraft crash protected. nuclear the Twosafety hazards. external internaland ancillary four from the of protect to geographically installed and are independent separated, are which systems, redundant quadruple of set A robustness. physical and diversity,redundancy of principles general the with accordance in arranged and simple Safety related systems of the EPR are designed to be mechanically Safety of maintenance during plant’s the and activities cost life time. number of welds and their improved reduced geometry leads tohas the reduction RPV the Furthermore, envisioned. is years 60 about of life design plant a andminimized be to expected is vessel the to neutron reflectors insidethe RPV, the life-limiting irradiation damage operating acts and as an added safety transients currently feature during DBA. Through operational the addition of with diminish to compared helps PWRs, when conventional circuit, primary inventory increased the The of loops. primary the of one to connected pressurizer a by controlled is circuit primary the of overallpressure the and each containing a steam generator and a N4 reactor coolant loops, pump. The 4 of French consists circuit the primary The models. KONVOI following German concept plant nuclear 4-loop The EPR nuclear systems are designed according to a conventional Nuclear Systems design reviews underway in the US and the UK. additional with China, ofand France Finland, in impacts construction the delay underarereactors CurrentlyEPR scenarios. coremeltfouraccident and and reduce catchers systems, core retentions as i.e systems, mitigation accident Severe systems. related safety the the of availability ensures redundancy four-fold art, the of State fuels. fuel management strategies, which include the use of MOX flexible through products waste radioactive decrease and efficiency utilisation fuel increase to managing 1750MWe,about of output power a has EPR The errors. human for margins reduced and impact radiological limited systems, safety diversified and reliable PWR highly advanced with an system reactor of design the to led and This industries. countries various in nuclear German and French the of reactors experience operational nuclear 100 over of has design The France.benefited from Paris, AREVA’s in cumulative headquarters construction experience its with A multinational consortium specializing in nuclear energy energy nuclear in specializing consortium multinational EPR the of organization design REVA, Evolutionary Power Reactor -7 . There are several active and passive and activeseveral are There . TM i a large a is , Artist’s renditionofEPROlkiluoto3,Finland EPR ReactorVessel and PrimarySystems EPR Flamanville 3,France ( (Courtesy of AREVA) Courtesy of AREVA ) the computerizedoperatorfriendcontrolrooms aredesignedtominimizedhumanerrorfactors. packages maintenance consistent of ensuring a reliable process use for the evaluation of the systems. The I&C systems are also make fully standardised and components and systems plants the of standardization the consideration. into Additionally operation unplanned and planned taking >92% of factor availability high a has actinides is subsequently reduced. power density, core lowersaverage the neutronflux thermal aboutby 7-15% long livedof production the and raises the thermal efficiency of the plant. the Furthermore higher fuelfor burn-up given enrichment, due to low effectively which MWth, 4600 aroundof poweroutput larger a inventorypermits primary sizeand core larger The EPR. the 12.5%) of for option (upan enrichment burner, is PuO2 an to plutonium a as function to order in Due to a design that incorporates good availability/access for maintenance and in-service inspections, the EPR normal operationmodeanda25-60%loadfor“lessusual“mode. lcrct gi sses la floig aaiiis r dsge it te EPR, complex the modern into designed with are comply To capabilities 26%. following by load waste systems, produced grid total electricity the reducing while 37%, of order the in is Through innovative plant design, fuel cycle management and operational strategies the total plant net efficiency Operation UO use can which EPR,savings the infuel added generate leakage neutron reduce to reflectors neutron of addition and core larger The utilities. the of a cyclefuel range 12between to 24 months using ENU, ERU and MOX fuels,to needsaccording the specific flexibleconditions for irradiationdifferent lengthscosts. operate cycle can and Thefuel reactor low withcycle Design of the EPR core is characterized by considerable margins for fuel management optimization incorporating Construction of EPR Taishan,Construction China 60 1750 4590 Design Life LWR Capacity Capacity MWe MWt Yrs EPR 312.6 Light Water Temperature 15.5 36 Efficiency Pressure Coolant

MPa % % °C TM ( Courtesy of AREVA 4.95 Light Water Enrichment 24 Moderator Fuel Cycle LEU Fuel 2 fuel with an enrichment level of up to 5 wt% of U of levelenrichment wt% an 5 to with upof fuel Mos % Fuel ) EPR reactorduring60yearsofoperation(Courtesy AREVA) Representation oftotalradioactivewasteproducedbyone Flamanville 3 Taishan 1,2 Olkiluoto 3

ih 0 t 10 la for load 100% to 60% with 235 . Additionally, . TM

5 China’s Advanced PWRs hina has the largest number of reactor units under Hualong One (HPR1000) Cconstruction in the world due to their large scale The Hualong One, now officially called the HPR1000 by the investment into the industry to develop a sustainable Chinese government, is the result of China National Nuclear Coolant Moderator energy mix. As of 2015, it has 26 commercial nuclear Corperation (CNNC) and CGNPC merging their design as LWR Light Water Light Water power reactors in operation as well 24 new reactor units suggested by the Chinese National Energy Administration. under construction at various sites across the country. The Hualong uses systems from CNNC’s ACP-1000 and Capacity Temperature Fuel More reactor units are in various phases of planning. The CGNPC’s ACPR-1000. Both reactors were conventional Chinese reactor portfolio includes a variety of reactor 3-loop PWRs, but ACP-1000 core design was finally adopted 1150 MWe 301 °C LEU types that are available on the world market, including to be placed in the Hualong One. The design incorporates CANDU6-, EPR-, AP1000- and VVER-units. The country the latest safety systems following internationally accepted Capacity Pressure Enrichment has also successfully developed domestic PWR designs standards, including backup passive safety systems, SA through the construction experience and technology mitigation systems and enhanced seismic protection. Future 3050 MWt 15.7 Mpa <5 % transfer from companies such as CANDU Energy Inc., reactors will be deployed by both companies separately, Westinghouse and AREVA. The most important domestic maintaining much of their supply chains, but each version will Design Life Efficiency Fuel Cycle Chinese PWR designs include the CPR-1000, ACPR-1000, have slight differences concerning the safety systems. CAP-1400 and Hualong One, also known as HPR1000. Artist’s rendition of CAP1400 60 Yrs 36.6 % 18 Mos CPR-1000 Coolant Moderator Fuqing 5 CNNC’s CPR-1000 (‘Improved Chinese PWR’) design, with LWR an electric output of 1000MWe, is based on the 900MWe Light Water Light Water 3- loop French M310 plants and make the largest proportion CAP-1400 Capacity Temperature of current reactors under constructions. Even though Fuel The CAP-1400 is a large 1400MWe PWR design developed Coolant Moderator China has a nearly complete domestic supply chain for the through cooperation between China’s State Nuclear Power MWe °C CPR-1000, the intellectual property rights are retained by 1030 311 LEU Technology Corporation (SNPTC) and Westinghouse. The LWR Light Water Light Water AREVA and thus restricts its use to domestic market. The design is based on the AP1000 reactor which utilize passive first CPR1000 was connected to the Chinese grid in 2010, Capacity Pressure Enrichment safety systems and the simplified systems design philosophy Capacity Temperature Fuel but the granting of new construction licences have been to increase safety and operational flexibility. The reactor suspended in favour of other advanced domestic designs. 3000 MWt 15.5 MPa <5 % safety designs have been improved after the Fukushima 1500 MWe 304 °C LEU Daiichi accident to accommodate enhanced seismic design Fangchenggang 1, 2 Design Life Efficiency Fuel Cycle and enhanced response capacity under BDB type events. Hongyanhe 1, 2, 3, 4 Capacity Pressure Enrichment Ningde 1, 2, 3, 4 The CAP-1400 possesses MOX fuel loading capabilities. Yangjiang 3, 4 Yrs Mos Ling Ao 3, 4 60 32.9 % 18 The valuable lessons learned from the construction of 4058 MWt 15.5 MPa 4.95 % AP1000 units in China have further helped to reduce issues faced during the construction process. Modularization and Design Life Efficiency Fuel Cycle Fangjiashan 1, 2 advanced construction techniques have helped minimize Fuqing 1, 2, 3, 4 delays during construction. The first CAP-1400 will be 60 Yrs 34.5 % 18 Mos Yangjiang 1, 2 constructed at Shidaowan site in Shandong Province, where site preparations have been ongoing since early 2014. ACPR-1000 Coolant Moderator The ACPR-1000 (Advanced Chinese PWR) was designed by the China Guangdong Nuclear Power Corporation based on LWR Light Water Light Water the CPR-1000 (CGNPC) with full Chinese intellectual property rights. The design focuses on the safe performance of the Capacity Temperature Fuel reactor while improving the economic efficiency. This enhanced version of the 3-loop CPR-1000 has higher seismic standards, 1150 MWe 311 °C LEU features a double containment and a reactor core catcher for SA mitigation purposes. The reactors is designed with SAMS including such in-vessel retention and spray systems Capacity Pressure Enrichment to for containment heat removal. Furthermore the ACPR-1000 MWt MPa % meets major post-Fukushima safety requirements. 3500 15.6 <5

Design Life Efficiency Fuel Cycle Yangjiang 5, 6 Construction of CPR-1000 at Construction of CPR-1000 and Site preparation for CAP-1400 Hongyanhe 5 60 Yrs 33.0 % 18 Mos Fangchenggang, China ACPR-1000 at Yangjiang, China Shidaowan, Shandong Province, China 6 7 8

K E P C O Advanced Power Reactor 1400 iiain ytm ae ntle t cnrl h hdoe buildup hydrogen the inside containment the during asever accident. control to installed are Systems Mitigation purposes. LBLOCAs. during control flowrate for mechanism control redundancy for trains Fluidic electrical Devices (FD) are installed in the SIS, enabling a independent passive flow 2 with OPR1000 SIS designs and has Injection four trains of mechanical equipment Safety The corium. the of retention in-vessel aid systems flooding cavity and cooling vessel the environment. In the case of a severe accident to with core melt, radioactivity the of amount significant releasing without primary the of (SDS) System Depressurization Safety the through relieved be can transients from resulting containment the in buildup Pressure LOCA. a of case the in tanks water dedicated from water cooling emergency with Vessel(DVI) core Direct Injection the The supplies are designedwhich to prevent melt and radiation core releases. that systems mitigation eventsevereseveral are there accident an of unlikelycase the In designers. reactor the byquoted are 0.3g of the APR1400. A core damage frequency forof systems 10mitigation accident and operational safe of design the Active, passive and inherent safety features were combined to satisfy Safety higher designmargins. coolant and size inventory of the pressurizer and increased steam generators have resulted The in elevation. higher at placed are SGs the purposes recirculation natural For manner. symmetrical relatively a in arranged is circuit primary The legs. hot the of one to connected pressurizer single a with legs, cold the of mounted each are on pumps coolant reactor The legs. cold two and leg hot single generator,a steam large one of consisting each loops, coolant two has system primary The occurrences. operational transients or alleviate and margins safety enhance to added been have Korea. features South Additional in decade a over of history the to Korean similar OPR1000 very type reactors, are which systems have a nuclear well overall proven APR1400’soperating The Nuclear Systems public and the environment. experimental of series research projects with the purpose of protecting a workers, the from results of incorporation the allowed process design and endeavours.research parallel A previous their from learned lessons incorporate to industry construction, operations and decommissioning in the nuclear company has drawn from its extensive The requirements.experience utility in areasof of variety a to tailored be can that design deployable worldwide a in resulted economics hand-in-hand go and safety that resistance philosophy seismic KEPCO’s and features. safety higher with 80+ System US the and safe reactor operating conditions. The design performance highprovide systemstopassivesafety evolved as well from nuclear of This Korea. reactor is a 1400 MWe South PWR that utilizes innovative of reliability active Republicas the in increase generation electricity and economics, safety,improve enhance to improvements engineering of variety a K P10, n vltoay eco wih incorporates which reactor evolutionary an APR1400, orea Electric Power Company (KEPCO) designed the s ipe ta te previous the than simpler is -5 and a seismic design

Hydrogen Construction ofShin-Hanul1,Republic of Korea Primary systemsof AP1400 Construction at ShinKori-3 Republic ofKorea (Courtesy ofKEPCO)

In addition, a one-piece reactor head reduces the need for in-service inspections over inspections lifetime the of reactor. reactor the head need for the in-service reduces In addition, aone-piece radiationexposurewellrefuellingas asduring relatedworkerslengths refuelling of activities. outage reduce to help refuelling in conventional PWRs into a single structure. This for and used the components separateinstallation of number of a instrumentationin-core combines Assembly cable Head traysIntegral An 90%. around be to projected is of currently operating PWRs, without endangering availability, than the consequences plantsafe operation higher of relative the plant. predict The plant mechanisms availability These for an trips. APR1400 reactor related and non-safety of likelihood the lowers mechanism progression trip reactor reduced a and the transients operational alleviates components secondary and primary the of volume large The purposes. following The APR1400 is designed to accommodate utility needs for daily load- Operation of the reactor with a core consisting of one third MOX fuel is also is fuel MOX minor some with possible modifications. third one of consisting core a with reactor the of during integrity enhances safety structural and operational of performance the plant. Operation and operation in performance other PWR reactors. The 10% thermal margin for the fuel nuclear hydraulic, PLUS7have assemblieswhich fuel demonstrated thermal enhanced WestinghouseAPR1400uses The conditions. optimal under 60GWd/t of maximum a reach can but 55GWd/t is the form of in gadolinium can be poisons used as of part an Burnable improved months. fuel management18 strategy. of length The averagecycle BU of the fuel reactor average an with UO2 is core the for fuel standard The the of demands South load-following Korean grid, high but it the has been adapted to for diversetailored fuel management specifically strategies for was the global export business.design core and fuel APR1400 the Initially, Fuel • • double containment and acore-catcher. a and standards larger powercore output. The APR+ also features a safety stricter accommodating a shorter construction time than the APR1400, while design is very similar to the APR1400, and envisions the APR+, which will generate 1500MWe. Its overall In 2007 the Korean government Regulatory launched designs for Nuclear US Commission. the by design review through going licensing currently is APR1400 The 1400 3983 Design Life 60 LWR Capacity Capacity MWe MWt Yrs APR1400 307.2 35.1 Light Water Temperature 15.5 Efficiency Pressure Coolant MPa

°C % Light Water 4.09 Enrichment 18 Moderator Fuel Cycle LEU Fuel Mos % APR1400 fuelassemblystructure(KEPCO) AP1400 MainControlRoom(KEPCO) Shin-Hanul 1,2 Barakah 1,2,3 Shin-Kori 3,4 9 ATMEA1TM TMEATM is a joint venture of AREVA and Mitsubishi ATMEA1TM AHeavy Industries (MHI) with the goal to design, TM market and deploy an efficient and safe mid-range PWR on the world-market. Using innovative and proven nuclear Coolant Moderator technologies from the well-developed nuclear power LWR Light Water Light Water sectors in France and Japan, ATMEA1TM accords with compliance across a broad set of regulatory and commercial Capacity Temperature Fuel requirements worldwide. This design incorporates top- ATMEA level safety systems, high thermal efficiency and a flexible 1150 MWe 308.5 °C LEU fuel cycle while simplification reduces projected capital and construction costs. Additionally a simplified effective design and standardizations of components forecast Capacity Pressure Enrichment reductions in operational and maintenance costs over the 3150 MWt 15.5 MPa <5 % 60 years of design life. Higher plant efficiencies, innovative fuel management strategies and effective surveillance Design Life Efficiency Fuel Cycle programmes lead to less waste production and help to minimize the impact on the environment. ATMEA1 plant layout 60 Yrs 36 % 24 Mos (Courtesy of ATMEA) Nuclear Systems ATMEA1 Reactor Pressure Vessel and The primary system of the ATMEA1 is mix between AREVA’s and Core Catcher (Courtesy of ATMEA) MHI’s previous 2-loop primary designs (KONVOI and N4) and their Fuel new evolutionary 4-loop designs (APWR and EPR). The ATMEA1 Nuclear safety, fuel management capabilities and fuel economy were significant factors that went into the design uses a 3-loop primary system with a hot, a cold and a crossover concept of the ATMEA1. The reactor provides a flexible 12-24 month fuel cycle using a low enriched core (<5 % 235 leg in each loop. A steam generator is dedicated to each loop and U ) of UO2. Burnable poisons in the form of gadolinium pellets in the fuel are used to control reactivity when fresh one single pressurizer attached to one of the hot legs controls the fuel is loaded into the core. Load-following capabilities (25%-100%) are part of the standard core design, with the pressure of the entire primary circuit. The reactor produces around possibility to use cores consisting of up to one third MOX-fuel. With a few minor design modifications the ATMEA1 3150 MWth with a projected net electricity production of 1100 MWe. can support a 100% MOX core for future plutonium burning purposes. Using heavy neutron reflectors with smart fuel Heavy neutron reflectors are placed around the core to improve management is expected to allow 10% less fuel consumption and radioactive waste generation per MW produced neutron economy and fuel efficiency as well as to reduce irradiation when compared to current PWRs. To ease the fuel handling and refuelling, the fuel pool storage area is located to the vessel ensuring the intended 60 years of design life. All the outside the reactor building. This, combined with innovative high speed refuelling machine, reduces the planned major nuclear and safety systems are designed to withstand 0.3 g duration of normal refuelling outages to about 16 days, for an anticipated overall availability factor of around 92%. of ground seismic activity and external and internal hazards, such as fires, aircraf t crash and floodings. The design of the components Operation and safety systems reduce core damage frequencies to around Digital instrumentation and control systems are used extensively in the operation of the ATMEA1 to improve the 10-6 /RY. human-system interface with the goal to reduce the likelihood of operator errors. By encompassing the sources of Primary Configuration of ATMEA1 human error in the design, testing and maintenance of the I&C systems, the physical demands on the operators Safety are lowered and permissible response times for critical operator actions are increased. Additionally top mounted Safety systems of the ATMEA1TM are designed to meet current (Courtesy of ATMEA) I&C control systems on the vessel give valuable information about the operations of the core at all times. A national and international sets of regulatory requirements including the remote shutdown station exists when the main control room becomes inaccessible. Design and configuration US-NRC, France’s ASN, Japan’s NRA and Euroatom. The safety and of components have been placed to ensure easy access for future maintenance and surveillance programs. design of the ATMEA1 is based on deterministic analyses of defence- Furthermore they have be designed to eliminate certain phenomena, such as stress corrosion cracking or fatigue in-depth aided by probabilistic analyses. The design considers various cracking, by removing certain structures from high neutron fluence areas. Shielding in the reactor building allows additional site-specific safety requirements such as seismic resistance, extensive online maintenance capabilities which help to reduce unnecessary plant unavailability. Furthermore flooding and tsunamis. Safety functions are based on active operator the maximum collective doses exposure for average radiation work is set to less than 0.5 man-Sv/y. initiated active systems with passive backup systems. All the essential systems assigned to protection and safety of the reactor have full, triple redundancy (and one additional safety train for support systems) with physically separated trains. Diversity and redundancy are designed • Currently undergoing licensing in various countries into the plant in order to provide two heat sinks, autonomy of 30 days, • Adaptability to different grid requirements (50 or 60Hz) and access to diverse water sources on site, among other things. • Mid-rage power output make this reactor adaptable to MHI advanced accumulators are used as passive systems combining countries with smaller or less developed electric grids safety injection and LPIS. In the case of an ex-vessel severe accident, • AREVA and MHI construction expertise and combina- a dedicated core catcher system with corium retention and cooling tion of their supply chain systems are available to mitigate the accident progression as well as to • SAMS: core catcher, cooling systems and hydrogen re-combiners for long-term containment integrity minimize the radiological release. Passive hydrogen control provides continuous service, and a pre-stressed concrete containment vessel ATMEA1 is partly based on the • Provisions for off site emergency equipment utilization with steel liner helps protect against large commercial airplane crash. Japanese PWR Tomari 3 N4 Type reactors - Chooz B 1, 2 France [1] 10 11 12 Mitsubishi Heavy Industries emergency situations. water are or stored in multipurpose water storage tanks accidents for use during severe of of reactor cavity is possible to effects aid corium cooling. Plain and borated the floodingex-vessel Foroverpressurization.accidents, containment mitigate to designed are sprays vent containment alternate and vents, containment filtered Air Vessel ContainmentRecirculation System, The hydrogen equipment. ignitors, the autocatalytic recombiners, streamline injection to low-pressure tank and tank accumulator conventional a as serves that tank accumulator high-performance passive the into low the of combined haveaccumulatorssystembeen and injection pressure function The trains. electrical independent 2 least at with System Cooling Core Emergency comes mechanical train APWR 4 a the with of safety, versions All different requirements. commercial have or APWR regulatory to specific are the that features redundancy and of diversity The versions situations. different emergency and three safe operation inherently and normal passive for active, of systems mix a uses APWR The Safety are designed limits withstand ground of acceleration 0.3 g. 10 of frequency damage core a and situations accident severe during core the into water of injection for available Portable are connections pump APWR. the in used are resistance corrosion greater for tubes 690 Inconel and capacity increased with generators steam lowering vessel fluence to endure a design life of60 years. as well as economy neutron Innovative for added are reflectors Neutron MWt. the pressure controls of the circuit primary which has a pressurizer thermal ofcapacity 4451 single A leg. crossover one and leg cold one leg, hot one of consists loop Each loops. the between pumps and coolant reactor 4 capacity and generators steam 4 with design PWR core 4-loop increased with but Japan modernized digital safety in related systems. PWRs The primary is a of traditional fleet The nuclear systems of the areAPWR largely based on the existing Nuclear Systems diverse and redundant plant protection systems. provide margins, providing environment by the protect and safety options fuel flexible operational high establish to used Mitsubishi’sfacilitiesweretheresearch own fromtechnology rangePWRs1538from MWe to1700 MWe. of-the-art State- large envisioned of family this for uniquely capacities Generating requirements. each utility and regulatory country-specific fit EU-APWR, to designers the and by US-APWR, JP- APWR, standard version: three in comes reactor The version. specific country/regulatory of a variety to diversified been has The costs. and period use in for domestic Japan, designed was but originally APWR construction the minimizing while reactoroperatingreliably a 1970s,build theto since projects and construction knowledge design of Mitsubishi, the involved with in experience 26 PWR operational vast their combined electiricity production make to was safer, design more reactor reliable new the and for economical. vision main The companies Their PWR. large a developand design to 1990s the during M -7 -7 /RY is quoted by the designers. Additionally, nuclear systems nuclear Additionally, designers. the by quoted is /RY in PWR electricity generation, came together in Japan in together came generation, electricity PWR in itsubishi and five national electric companies involved companies electric national five and Advanced Pwr

Advanced accumulator flowmechanism APWR plantlayout(CourtesyofMHI) APWR PrimarySystems (MHI) reduce operationalreduce radiation areas in controlled these for radiation workers.to likely are areas radiation sensitive in equipment handling Remote plant. the of operation reliable more in reduce to manner viewed easily interpretation a and communication in errors. displayedOperators’ plantwork load the and errors of are thus state reduced, the potentially about resulting information vital with digital fully foreignanddemands. systemshome Instrumentationrequiredwith Control and The tocomply plantof the are designed into thus the APWR, giving the plant the ability to operate in a range of is15%-100% operation Load-following of full power when90%. over of availability predicted a have and plants Japanese current than cost less third a about for electricity produce to predicted is APWR large the scale of economies Through factors. economic the negating not while priorities top as reliability high and safety with reactors evolutionary in need improvements on information vital provided has reactors Japanese of experience operational extensive The Operation without any the to nuclear major modifications systems. 62 GWd/t, but there are plans to increase this value to 70 GWd/t. It is also possible to use one third MOX cores of maximum a with GWd/t 55 anticipatedaverageis The refuelling BUoutages. of length the systemsshorten handling fuel speed high and automatic with combined cycles fuel safety.improvedLonger for margin design achieve lower power densities, and a 24 month fuel cycle to enhanced fuel economy with the addition of larger toassembly slighter fuel a longer hasversion uraniumrequirements. the EU/US-APWR The reduces core the thus PWRs, Japanese in Oxide (UO2) pellet operating with low enriched uranium (max currently of 5 % of U235). in The addition of heavy neutron as reflectors in fuel same the capitalizing is on years of APWR previous the operating experience and in design improvements. used It fuel consists of standard sintered Uranium The Fuel Safety tems Emergency power sys Fuel assemblies Electric output transfer area 1540 4466 - Design Life 60 LWR Capacity SG heat Capacity BDB Mechanical Electrical Yrs MWe MWt APWR Light Water 307 34.4 15.4 Temperature Efficiency Pressure Coolant

MPa °C % Diesel generator 257 x3.65m 1538 MWe APWR Light Water 6500m 4 trains 2 trains 4.5 Enrichment 24 Moderator Fuel Cycle LEU - Fuel 2 Mos % Diesel generator US-APWR Alternate AC 257 x3.65m 1700 MWe 8500m 4 trains 4 trains 2 Comparison ofcoresize ATWS, multipleSGTR, Gas turbinegenerator MSLB +SGTR, SBO EU-APWR 257 x3.65m 1700 MWe 8500m 4 trains 4 trains 2 13 14 ROSATOM scenarios. the accident in products fission of releases slowand progression initiation, features These transients. operational for mechanism dampinga asinventory act primary increased circulation and abovecore the volume natural coolant large via the as core such features the Design of mechanisms. capability cooling the volumes aid feedwater to secondary increased with volumes coolant large on depend Today’s mechanisms.VVERs protection reactor being adopted today across the entire nuclear as industry inherent design features that wereofreactor part earlier designsVVER are distinctive The environment. of the to risk releases radiological the harmful minimize to concepts defence-in-depth and multiplebarriers use VVERs The regulation. national to tailored are that they can be distinguished by their electric output and safety VVER units systems,are designed in various sizes and models, but generally Main Safety Features ogtdnl ed, hs nbig n es rqet maintenance testsand surveillance RPV. the on frequent less in enabling without thus forged welds, are longitudinal vessel the of shells The envisioned. is melt reactor core increasing or degradationsevere situations where in core integrity their head, VVER lower reactor rail. the in by penetrations lack equipment vessels main the of ensure to made transportability been have efforts components, large some a design of the with During alloy. assemblies zirconium-niobium of fuel cladding rod hexagonal fuel utilize reactors type these addition, In replacement. for need the without years 40 over for operating successfully been have Steam plants steel. Russian stainless in generators of made tubes generator steam VVER walled containing Plants thick relatively designs. with generators steam horizontal employ reactors VVER are the materials construction in and incorporated concepts unique of number A Main DesignFeatures Water Water Energetic Reactor (VVER) to regulatoryorutilityneeds. AES-2006. MWe 1200 Each reactor model can be fitted with the components specific and AES-92 MWe 1000 the include VVERs the of models type PWR large, available the effective use of VVERs in closed fuel cycles. Currently implement to try designers the Furthermore capabilities. efficiency load-following implementing as well increasing as factors, load and periods, construction current shortening the face to designed challenges of the are nuclear industry by units reducing capital costs, VVER Modern companies proven. and tested domestically Russian been has that technology by constructed nucleara exporting mid-1960s, the since globe the across been have units VVER 60 Over Novgorod Atomenergoproekt. Nizhniy and Atomenergoproekt, Saint-Petersburg Atom-energoproekt, Moscow ROSATOM: within organizations design the by VVERs. The plants housing the VVER reactors are created R a class of Russian Pressurized Water Reactor called Reactor Water Pressurized Russian of class a osatom subsidiary OKB Gidropress is the designer of Artist’s versionofNovovoronezhNPP II, Schematic viewofVVER1000Primary Russia, housingaVVER-1200unit. Novovoronezh NPP II,Russia Construction of VVER1200 (CourtesyofRosatom) (Courtesy ofRosatom) [2]

VVER-1000 (AES-91& AES-92) construction in RussiaatLeningradII andBaltic. developed for Tianwan, China. VVER-1200/491 plants are under Atomenergoproekt by developed was designs (VVER-1200/491) being are reactors The of AES-2006 family other The design. Russia. II, Novovoronezh at built AES-92 the of basis the TheVVER1200/ on Atomenergoproekt Moscow by standards. developed was version regulatory V392M national different to tailored are which designs AES-2006 of families two are There and safecoolingfor72hourswithoutexternalpower supply. steam and containment BDBAof case independence unlikely allowing the in generators the for HRS passive 54 has the AES-2006 AES91/92 around the of feature to safety the to times addition In months. construction predicted optimized have techniques construction Modular GWd/t. 60-70 of BU average an with months 12-18 years. between set 60 be of can life outages Refuelling service expected an and 90% over of factor reactors are expected to have an efficiency of 34%, an availability series, consisting of a 4 loop design. The 1170 MWe VVER1200 10 of the European Utility Requirements. It is predicted to have a CDF the Russian Regulatory Standards and is also in compliance with systems. safety The passive to according designed is AES2006 construction and and active using standards safety high maintaining capital while costs, lower to series VVER1000 technologies the proven from previously utilize These to design. designed PWR are plants deployable advanced most their 2006, VVER1200/AES- the for designs the finalized Rosatom 2006 In VVER-1200 / AES-2006 one unitstillunderconstruction. there is one AES92 unit in operation in Currently India. for designed was (VVER-1000/V412) AES92 under currently are units China. Tianwan,in Twoconstruction Finland. in construction for designed originally was it as Standards, Safety European the with complies (VVER-1000/V-428M) AES91 the called model first The sites. different two at world the around constructed lifeof 33.7% design being currently reactors a with VVER1000 three are years. There 50 at operate factor 90%availability a to with designed efficiency is reactor VVER1000 The with passiveECCSandHRSforreactorprotection. rates for radiation workers. Modern VVER1000s are equipped lower neutron activation led to significant reduction of exposure with materials cobalt-free new of use steam The designs. generator horizontal improved with design 4-loop conventional a on based were reactors big these of systems primary The larger requirements associated with the Proven higher thermal output. history. operating of design features from years the smaller plants were scaled from up to fit the drawing designs, VVER440 older the on based were reactors type VVER1000 The protection. reactor and systems safety passive for forces during the 1980’s. It was one produced of design the PWR first1000MWe designsa tois VVER-1000 utilizeearly natural The -6 /RY. Primary systems are very similar to the VVER1000 VVER1000 the to similar very are /RY.systems Primary on the basis of the AES-91 originally originally AES-91 the of basis the on The other model is called the called is model other The Kudankulam Saint-Petersburg Saint-Petersburg , India and Leningrad II 1,2 1060 3000 3200 1170 Design Life Design Life 60 60 LWR LWR Capacity Capacity Capacity Capacity Baltic 1 V-491 MWe MWt Yrs Yrs MWe MWt 313.5 305 Light Water Light Water 33.7 Temperature Temperature 33.9 15.7 16.2 Kudankulam 1,2 Efficiency Efficiency Pressure Pressure Coolant Coolant Tianwan 3,4 V-428M Novovoronezh V-412

Mpa Mpa % % % °C °C V-392M 4.79 Light Water Light Water 4.55 Enrichment Enrichment Moderator Moderator Fuel Cycle Fuel Cycle 18 18 LEU LEU Fuel Fuel

II1,2 Mos Mos % % 15 16 ROSATOM ranging 12 between and 24 months. Maximum fuel burnup is 70 GWd/t with a fuel cycle duration design lifetime of 50-60 years and errors, large human to operational tolerance systems, extensiveflexibility. safety passive of include use features design Major projects. earlier operational reliability through the operational feedback from the global energy markets while maintaining high safety and on ability competitive high are design the of aspects Main as Russian the with and Euroatom standards. regulatory with modern safety regulation, andcodes standards as well complies design new The 1560MWe. of output electric an trend increasing the of from the economics of scale, a large PWR was anticipation designed In with of consumption electricity around the world and the benefits series. VVER1200 and VVER1000 their of construction successful worldwide the from greatly benefited has design recent most Rosatom’s VVER1500 Construction Time (mos) Occupational Radiation Thermal Output(MWt) Electric Output(MWe) Operator Action Time AES92 (VVER-1000/V412) Availability Factor Exposure (Sv/RY) Kudankulam Design Life(yrs) Plant Efficiency Load Following CDF (/RY) 1,2-India VVER1000 0.6×10 AES-2006 ( AES-2006 33.7% >90% 1070 3000 6 hrs Yes 0.5 46 50 Novovoronezh II1-Russia VVER -6 VVER1200/V392M) VVER1200/V392M) VVER1200 1560 4250 Design Life LWR 33.9% >92% 50 1200 3200 6 hrs <10 Capacity 0.39 Capacity Yes 54 60 [2] -6 Yrs MWe MWt AES91 (VVER-1000/V-428M) Light Water 316 Temperature 35.7 15.7 Efficiency Pressure Coolant Tianwan 3,4-China

VVER1500 Mpa °C % % 35.7% >6 hrs >93% 1560 4250 <10 TBD Yes 54 50 Light Water 4.92 Enrichment -6 Moderator Fuel Cycle 24 LEU Fuel Mos % [3] latest setof requirements mandated byregulators. gone through safety upgrade processes, fitting the plants with the have units Several countries. various in renewed/extended been have licences operating programmes, management life plant of implementing Through operation. under still are and years 30 of A number of VVER units have surpassed their intended design life are VVER1500-series) predicted toservefor50 or60years. and VVER1200- (VVER1000-, models and operational practices, the new generation of advanced reactor material in improvements continuous through but years, 30 were models VVER older of lifetime design efficiency. The operational increasing and designs their years improve to the Rosatom benefited over has VVERS operating from gained experience The years.reactor era. 1200 exceeds time operation cumulative Their Soviet the during Russia various with starting in 1960s the constructed since countries been have reactors VVER 67 of total A Operation operation. Slight modifications are introduced into the design of design the assemblies fuel toPWR facilitate 3, 4, 5years or of life. fuel into introduced are modifications Slight during operation. stability and mechanical heat removal sufficient flow, fluid functions key to the stability and safety of the fuel, including optimal facilitate to design recent the in used all are filters anti-debris and grids anti-vibration grids, reactor. Spacer the into loaded fuel fresh of reactivity excess the control to used be can oxide)(gadolinium poisons neutron Burnable U235. wt% 4.95 to having up are of enrichment assembliesan rods fuel fuel hexagonal-shaped into These together cladding. bundled zirconium-alloy fuel from into fitted made pellets rods fuel UO2 sintered enriched utilize VVERs anticipated and burnup. fuel modes operational reactor, the of output power length, cycle in accordance with new reactor designs and fuel cycles such as fuel assemblies have increased technical and economical characteristics the respective reactor model and the operational need. fuel Modern VVER1200 and ofdimensions the to tailored be must assembly VVER1000 fuel Each reactors. VVER440, the including fleet VVER Rosatom entire the for fuel manufactures and develops company fuel The TVEL. subsidiary Rosatom the by produced is fuel VVER Fuel Fuel cyclelength(mos) Avg. reloadenrichment Avg. coreheight (m) Core diameter(m) density (KW/kgU) Avg. corepower Avg. BU(GWd/t) density (MW/m Avg. fuelpower (wt%) 3 ) VVER1000 4.45% 35.8 3.16 3.53 108 53 18 VVER1200 4.79% 12-18 108.5 36.8 3.16 3.75 60 Hexagonal VVERfuelassemblies. Novovoronezh NPP, Russia VVER Control Room VVER1500 4.92% 12-24 3.85 57.2 4.2 - - [6] [4] [5] 17 18 Toshiba - Westinghouse CDF ofCDF 1× 10 low a in resulting g, 0.3 of design seismic earthquake shutdown safe a with margins, large with criteria probabilistic-risk and deterministic-safety piping. primary The AP1000 is designed in with accordance the US NRC accidents, for principle” LOCA-type “leak-before-break the by in mitigated additionally are uncovering which core extended of likelihood the reduces core reactor the above penetration of placement the Moreover the risk of PV failure that could result in high-energetic corium interactions. features reduce 72-hourvesselandcooling mechanismvessel-retention in- the scenario, SA a of case systems. the safety-relatedpassiveIn the of actuation unnecessary eliminate to used are systems active safety/control proven Previously DBAs. postulated mitigate to action operator HVAC systems) are excluded from the through design, or thus water operaterequiringwater, service cooling no power, (ACcomponent systems immediate that support generators) diesel or fans (pumps, components power. AC Likewise,off-site active and on-site all of loss a during action the reactor will safely shut down and remain cooled even with no operator that ensure and functions related safety serve toutilized are convection, flow,pressurizedgas,gravityas such forces flow, naturalcirculation Natural and AP1000. the of development the in used are technology PWR A conservatively based design, decades of reactor operations and proven Safety design component lifetimes. in-core longer barrel help to attenuate neutron and gamma fluxes, contributing to trips. reactor unwanted Various reduce components inside and the RPV such as core shroud and margins core operating the to helps increase pressurizer The pipes. leg hot the of is one circuit, to connected primary the of temperature and pressure the controls pressurizer,A relatively large which plant. the of overallefficiency the increases and loop the along drops pressure reduces pump coolant reactor and generator steam between leg crossover a of of steam each to bottom the generator.connected elimination The directly are that pumps coolant reactor two and generator steam a a having PWR 2-loop conventional design. Each loop consists class of a 1100MWesingle hot leg, two a cold legs, is AP1000 The Nuclear Systems may result from potential accidents. featurestoprevent limitor theescapeoffission products that cooling and isolationcontainmentunique are design the into incorporated Also 2005. in Toshiba company Japanese the by based bought was which US LLC,Company ElectricWestinghouse the by experience operational and construction designs. This reactor model builds on 50+ years of successful buildingpreviousmaterialWestinghousethanless and PWR resultssimplerproceduresinfewertoO&M due components required by standards the high industry the and maintaining while regulators. costs, Plant capital simplification and time also plant construction minimize systems, safety of smaller complexity the lower simplification, which of all techniques, plant construction modular and footprint, extensive philosophy design on AP1000 The focuses reactor. the safe of the facilitate operation that functions related safety out carry to circulation natural or gravity as such forces natural on depend T mhss n asv sft sses Tee systems These systems. safety passive on emphasis oshiba’s -7 /RY. /RY. Advanced Passive 1000 AP1000 is a 1100 MWe PWR designed with

First Concrete atVCSummer2-USA Construction SiteSanmen1,2-China Primary Configurationof AP1000 [7] testing purposes. ALARA principles have principles to been enforced keep worker at doses lowtesting levels. ALARA purposes. lengthy the reduces training required workersfor as well as lengthy instrument periodic for positioning/calibration work for radiation workers. The dedicated standardization of components with across plants with components built-in-testing apparatus further of number reduced and platforms devicesaccess lifting at key A locations facilitate shorter, safer and costs. more reliable maintenance/repair generation electricity forecasted the lowering plant, nuclear the of life entire the for costs operational the of overallreduction Reduced on in resultrequirements maintenanceaspects. safety operational on compromising without operation plant reliable and stable a predicting maintenance andforced included). increased The operational margins unwanted helpreduce thus trips, reactors A high AP1000.degree of the reliability of with low design maintenance the requirement during can result considered in a were high that availability aspects factor of important ~93% (planned were maintenance and Operation Operation in contained is fuel The ZIRLO facilities. dedicated in testing extensive undergone have and features design proven ROBUST Westinghouse the of are assemblies fuel standard The soluble boron concentration, since this is considered more economical and efficient,changing andreduces than wasterather products. grey-rods using provided are capabilities following strategy.Load burnable management following fuel 60GWd/t good fuel of BU average Integral an have to U235. predicted is of assembly fuel % A needs. utility 4.8 special accommodate to 2.3 from enrichments varying to used be handle excessive and can fuel of of coating fresh reactivity (IFBA) surface absorbers thin boride contain rods fuel have can core reactor (UO Dioxide Uranium enriched alternatingcyclelength. fuel Following kept be refuellingoptimal a conditions, can outage to 17 days. use standardA to designed primarily is AP1000 The Fuel • • • • • • • waste. Reduced radiation exposure, radioactive less years design60 life 93% with life time availability. MOX Fuel capabilities. In-vessel retention features for DBA. of Corium found in currentlyoperating Westinghouse plants. Systems and are Components improved versions threats. security measures against range wide of natural and toDesigned physical accommodate protection times. scheduling to minimizeconstruction and lead- costs Toshiba’s innovative, modular, and parallel TM 1200 3200 Design Life LWR 60 tubing large with gas spaceto for gas increasedhigher accommodate production fission burnups. Capacity Capacity Yrs MWe MWt AP1000 Light Water 303 Temperature 15.5 32 Efficiency Pressure Coolant

% % MPa °C TM Light Water 4.8 Enrichment Moderator Fuel Cycle 18 LEU Fuel Mos % Westinghouse Plants(Courtesyof Toshiba) Comparison between AP1000 andexisting TM design, which have a number of previously of number a have which design, 2 ) fuel with an 18- or 16-20-month 16-20-month or 18- an with fuel ) VC Summer2,3 Sanmen 1,2 Haiyang 1,2 Vogtle 3,4 19 KERENATM TM REVA and major European utilities including E.ON, KERENA AVattenfall, EDF and TVO developed an advanced TM BWR called SWR10 0 0 in 20 08 but renamed to KERENA Coolant Moderator in 2009. Through the cooperation with multinational BWR utility companies, system simplifications and the use Light Water Light Water

AREVA of previously proven design concepts, risks associated with foreign licensing and construction can be further Capacity Temperature Fuel reduced. The design and the developing of the plant 1250 MWe 320 °C LEU aimed at producing an economically competitive plant for the future. Capacity Pressure Enrichment KERENA combines years of experience in reactor operation and design with innovative features proven in 3370 MWt 15.5 MPa 4.5 % dedicated research facilities. It is designed with a diverse mix of active and passive safety features to minimize the Design Life Efficiency Fuel Cycle operational safety risks and mitigate accident scenarios. Illustration of KERENA plant layout. 60 Yrs 36 % 24 Mos Operational flexibility and fuel cycle costs are optimized Emergency condensers: Left, during normal operation. Right, acting as ECCS when RPV water levels drops. (AREVA) by load-following capabilities, improved fuel cycle strategies and less complex maintenance procedures. Fuel AREVA plans to use modified ATRIUMTM10 for KERENA at low enrichments Nuclear Systems of UO2 (4.5 - 4.7 % of U235). The new fuel called ATRIUMTM12 would The nuclear systems of KERENA are based largely on the proven result in structurally larger, but fewer number of fuel assemblies in the core designs of AREVA’s 1300 MWe BWR plants, with the exception than in current BWR designs. Reducing the number of fuel assemblies of components belonging to passive safety systems. The core shortens the reactor’s standard postulated refuelling outages to about 11 produces 3370 MWt of thermal energy which is converted at 37% days. The innovative core design and fuel management strategy promises efficiency to 1250 MW. The coolant is recirculated by 8 reactor more fuel efficiency and a reduction of long-life radioisotope products of recirculation-pumps inside the RPV. Steam from the core passes about 15%. Through appropriate fuel management an average discharge through moister separators and steam dryers before entering the burnup of about 65 GWd/t can be expected. The flexiabel fuel cycle length high pressure-tubines via three main steam lines. Core thermal can be between 12 and 24 months. Additionally KERENA is capable of power is regulated through reactivity feedback by changing the flow using reprocessed uranium and up to 50% MOX cores without significant rate of the RR-Pumps or by the hydraulic control rod scram systems design change to systems and components. The spent fuel assemblies are through the bottom of the RPV. The reduced active core height and stored in the spent fuel pools, which contain residual heat removal systems large volume of coolant above the fuel reduce the probability of and shielding, inside the reactor building. uncovering the core during LOCA-type scenarios. KARENA design is based on German BWRs Operation Safety Gundremmingen 2,3, Germany [8] Through streamlined design solutions, low maintenance requirements and Redundancy and diversity of integral safety systems in the KERENA optimized fuel management strategies, the plant’s availability is trageted reactor are provided by two active and four passive qualified safety to be above 92%. KERENA load following capabilities (between 40-100% systems. An innovative approach of partially replacing active safety of full power) and frequency control make it a flexible tool in the energy systems with passive safety systems is used to lower the calculated mix of any utility. The reactor’s optimized recycling processes reduce the probabilities of accidents and radiological consequences of accidents overall environmental footprint of the plant during operation. Large water due to failure of I&C systems or human errors. Active systems volume in the primary of reactor helps transient control during operation employed for reactor protection are manually activated whereas the and avoids unnecessary reactor trips. Thorough significant simplification of passive systems act as backups. In many accident scenarios, passive system engineering, standardization and reduction of components, there systems provide an increased time frame for operators to choose are less periodic maintenance procedure to be carried out. These actions intervention procedures. Passive systems such as passive pressure improve the availability of the plant and the safety of the workers as well as minimizing the occupational dose to radiological workers. Artists rendition of KERENA’s pulse transmitters (PPPT) monitor the reactor water levels and RPV and internals design (AREVA) compare with the secondary pressure build-up to ensure operation within safety margins. Safety systems for accident control without • Designed with input from European BWR Forum, • Resistance to plane crashes and seismic vibration overheating/melting of the core include emergency condensers an organization consisting of Europe BWR vendors • Severe Accident Mitigation Systems and gravity driven core flooding system. In-vessel retention, drywell and utilities flooding, containment cooling and hydrogen recombiners all help to • 19 BWR Units providing construction and • Load following (40% and 100% of its nominal preserve the RPV and containment boundaries and to prevent the PPPTs: Left, normal operation. Right, upon operational experience of 60+ years power) release of harmful radioisotopes to the environment. pressure build up in reactor system with • Minimal construction time through modular • Enriched Uranium Fuels, Reprocessed Uranium or Automatic Depressurization (AD) initiated. construction methods MOX Fuel can be used 20 21 Advanced Boiling Water Reactor - ABWR E-Hitachi’s ABWR was designed to be an econom- ically viable method of electric production with high- ABWR erG standards of nuclear safety than current BWRs. The original was completed in 1991 by General Electric and its Coolant Moderator technical partners, Toshiba and Hitachi Ltd. with important BWR contributions from utilities from Japan and US. This plant, Light Water Light Water with a design life of 60 years, comes in three different ver- sions tailored to regulator and utility needs of the countries Capacity Temperature Fuel or regions in which they are envisioned to be deployed. Currently available versions are JP-ABWR, US-ABWR and 1420 MWe 283 °C LEU EU-ABWR. The term ABWR now refers collectively to the most advanced version of boiling water reactor designs, Capacity Pressure Enrichment whether the parent company is GE-Hitachi, Hitach-GE or Toshiba. Modularization techniques and equipment simpli- 3926 MWt 7.07 MPa 4 % fication, without compromising the safety of the plant, en- able shorter construction time and reduced up-front capital Design Life Efficiency Fuel Cycle costs. Load following capabilities and smart fuel managing strategies are intended to help provide a competitive elec- Kashiwazaki-Kariwa Nuclear Power Station, 60 Yrs % 24 Mos GE - Hitachi | Toshiba GE - Hitachi ABWR. (Kashiwazaki-Kariwa, Japan) 34.4 tricity genration option. In 2015 there were 4 operational Historical evolution of RPV in BWRs ABWR’s and 4 are under construction. Nuclear Systems Fuel Kashiwazaki- The NSSS of the ABWR uses a number of innovative design improve- The ABWR uses sintered UO2 as fuel at an average en- Kariwa 6, 7 ments that have been developed through monitoring the operation of richment of 4 wt% of U235. Increased fuel utilization, per- Shika 2 over 90 BWRs around the world. The ABWR uses 10 reactor internal formance and reliability are provided by the GNF2 fuel Hamaoka 5 pumps to circulate 3926 MWt of heat from the fuel through 4 main assemblies used in the ABWR. The fuel assemblies are steam lines to the turbines to produce 1420 MW of electricity at an ef- manufactured to have increased corrosion and debris re- ficiency of 34.4%. The elimination of recirculation pumps and piping in sistance in the BWR environment capable of operation Shimane 3 the primary, especially safety related RPV penetrations below the top at 120% power for up to 24 months. The average burnup Oma 1 of core, has helped to increase the overall safety of the reactor sys- is about 50 GWd/t due to the higher fuel mass and high tem, which has a CDF of 1. 6 ×10 -7. Reactivity control is maintained by enrichment compared to previous BWR technologies. A lower power density results in improved fuel cycle costs a combination of control rods, RIP flow rates and the amount of burn- Lungmen 1, 2 able poisons present in fuel. Simplification of components have made and a greater manoeuvrability for operation. the reinforced concrete containment vessel of the ABWR significantly Operation smaller than current MARK-Type BWR containments. Post-Fukushi- A 60 year design life is predicted by ABWR designers, accounting for continued functions like startups, ma adjustments have introduced multiple water injection connection shutdowns, automatic responses to load changes and minor transients during normal operation. Primary control points for portable pumps to the core if ECCS becomes unavailable. of the power plant remains within the control of the operator who monitors the status of individual systems as Safety well as the automation sequences instead of controlling individual equipment. ABWR uses active, passive and inherent features for reactor protec- ABWR core and coolant flow directions (Courtesy of GE Hitachi) Load-following capabilities of 70-100% of full power are provided through coolant flow of RIP pumps. This, tion. Country-specific versions of the plant vary in the type of safety combined with lower O&M costs through reduction and simplifications of systems, contribute to competitive systems, their redundancy/diversity or capacities in accordance to electricity generation. Having fewer total components also contributes to less maintenance activities, thus the regulatory requirements. Large volume of coolant in the primary resulting in a reduction of occupational exposure for radiation workers. alleviates or delays reactor pressurization during transients. Passive wetwell venting systems and HMS are in place to protect contain- JP/US - ABWR EU - ABWR ment integrity during postulated SA. Additional EDG and gas-tur- bine generators act as backup power for important systems during Electric output 1420 MWe 1600 MWe SBO initiated by extreme external events such as earthquakes, Thermal output 3926 MWt 4300 MWt floods or storms. Many systems, such as RHR, dual functions, both for shutdown cooling and for core/containment cooling during pos- 3 × 50 % Emergency Core Cooling 3 × 100 % of Emergency Core Cooling Reactor protec- tulated LOCAs. ECCS capabilities can also be reduced due to the (2 × HP-Core Flooding, 1 × RCIC, (2 × HP-Core Flooding, 1 × RCIC, tion systems elimination of recirculation in the primary loop. Three fold diversi- 3 × LP-Core Flooding) 3 × LP-Core Flooding) ty and separation for CRDM protect against ATWS type accidents. Post-Fukushima adjustments have resulted in the incorporation of Emergency power 3*EDG,CTG 3*EDG, 2*CTG, 4 Divisions additional external connections to diversify emergency water injec- Armored containment, divisional sep- Core melt spreading, corium shield, tion to the core and spent fuel pool. A core catcher is envisioned on Safety features aration, H -recombiners, core catcher, PCV vent 2 the floor of the lower drywell in the RCCV of the next Toshiba variant passive containment cooling of ABWR to help mitigate severe accidents involving core melts. Construction of Lungmen NPP- Taiwan, China 22 23 Economic Simplified BWR E-Hitachi’s latest design is the ESBWR which combines Geconomy of scale, improvements in safety and design ESBWR simplification to generate electricity in a reliable fashion. The company uses its 50+ years of experience in design, Coolant Moderator construction and fabrication in the nuclear industry to utilize BWR natural phenomenon and passive features for a safe operation Light Water Light Water of the BWR. Reduced O&M costs from simplifications leads to cost reductions which, in combination with shorter projected Capacity Temperature Fuel

GE Hitachi construction time and lower costs, result in an economical MWe power plant. This design has active non-safety systems to 1520 281.6°C LEU handle operational transients, but all reactor safety systems are passive and thus do not require AC electrical power Capacity Pressure Enrichment nor operator actions for cooling for more than seven days. The predecessors of the ESBWR are the ABWR and the 4500 MWt 7.17 Mpa 4.2 % SBWR, the latter abandoned in the 1990s due to insufficient power output. Most systems and components in the ESBWR Design Life Efficiency Fuel Cycle have proven operating histories from the ABWR, but some Illustration layout of ESBWR plant layout Yrs Mos (Courtesy of GE Hitachi) 60 34 % 24 innovative features have been incorporated and tested by ESBWR design based on experience from ABWRs the designers in dedicated facilitates. ABWR unit Hamaoka - 5, Japan. [9] Fuel Nuclear Systems A key design feature of the ESBWR is its fuel design due to its large core power and innovative flow circulation The ESBWR is designed to operate at 35% efficiency, turning 4500 method. The fuel used is UO2 at an enrichment of 4.2 % of U235. An average BU of 55 GWd/t is predicted MWt into 1560 MWe. The NSSS of the ESBWR is quite similar to the with a flexible fuel cycle length of 12-24 months. The fuel assembly design is based on the proven GE14 ABWR which has 4 MSLs to transport the heat from the core to the assemblies, but has shorter active fuel heights to accommodate the use of natural circulation. The Global turbines. The significant difference is that the ESBWR is designed Fuel Group, a joint venture of GE, Hitachi and Toshiba, are currently working on improving parameters of the to operate using natural circulation to remove heat out of the core. In ESBWR core and fuel design. order to use natural circulation forces some changes were made to the primary systems, such as an increase in vessel height, installation Natural circulation significantly improves key performance parameters of the plant such as fuel efficiency and of a partitioned chimney above the reactor core, a decrease in active utilization. There are also a larger number of fuel bundles in the core due to the higher core power and shorter fuel height, and a taller, more open down-comer annulus that reduces active fuel heights. Larger bypass gaps between the fuel assemblies improve cold shutdown margins and core flow resistance and provides added driving head, pushing water to the thermal-hydraulic stabilities, resulting in milder responses from pressure transients. bottom of the core. The shorter fuel heights, improved steam separators Operation and tall chimney at the top of the vessel result in reduced pressure The comparatively low plant construction and development costs for the ESBWR with 60 years of design drops and higher overall plant efficiency, which translates into greater life and an availability factor of over 88% provide an economical way to produce low-carbon electricity. economy. The taller PV with its relatively larger volume of coolant, Standardization and simplification of the systems with a large proven design base can reduce licensing combined with the absence of hydraulic instabilities caused by forced and first-of-a-kind plant costs in many countries. Furthermore simple passive safety systems with minimal circulation, serve to enhance the operational flexibility of the reactor. Artists rendition of proposed ESBWR connections and welds and reduced maintenance activities decrease the O&M costs. This will in turn reduce North Anna 3, United States. the operational exposure of radiation workers performing online maintenance to less than 1 Sv/y. The fully Safety (Courtesy of GE Hitachi) digitized control room with self-diagnostic features puts fewer demands on plant operators who will be Some safety systems present in the ESBWR design were previously monitoring the automated systems to ensure correct procedure functionality and safe operation of the plan. used as backup systems to active safety. The passive safety systems are designed in a modular system, keeping them independent and BWR/4 ABWR ESBWR physically separated to curtail independent events from affecting the same physical aspects of the plant. Natural circulation eliminates the Power (MWth | MWE) 3293 | 1098 3926 | 1350 4500 | 1550 need for safety grade equipment like pumps and safety DG’s to be Vessel height | diameter (m) 21.9 | 6.4 21.1 | 7.1 27.7 | 7.1 present inside the containment and thus the size of the containment building can be reduced by about 30%, despite the increase in power Fuel bundles 764 872 1132 output of about 15%. The ESBWR design’s core damage frequency of 1.7 Active fuel height (m) 3.7 3.7 3 -8 × 10 per year is considered to be lowest of advanced reactors currently Power density (kW/l) 50 51 54 available. The GDCS and ADS act as the plant’s Emergency Core Recirculation pumps Cooling. Isolation condensers are used for HPI and decay heat removal 2 (external) 10 (internal) 0 in SBO conditions. LPI by gravity driven cooling systems is connected to Control rod drive type Liquid Pressure Fine Motion Fine Motion three water sources. In the case of a SA the PCCS, Containment- and Core Catcher Cooling System are in place to mitigate the progression Safety system pumps 9 18 0 RPV, Containment Structure and of the accident and limit radiological consequences. In accordance with Safety diesel generators 2 3 0 post-Fukushima lessons learned, emergency connections for DG and Safety Systems of the ESBWR water makeup to SGs are present as well as HMS. (Courtesy of GE Hitachi) CDF (/RY) 10-5 10-7 10-8 24 25 Enhanced CANDU 6 andu Energy Inc. (previously part of AECL, now a Csubsidiary of SNC Lavalin, a Canadian engineering EC6 company) designs and constructs CANDU reactors. Its most recent design is the Enhanced CANDU 6 (EC6), a 740 Coolant Moderator MWe heavy water moderated and cooled pressure tube HWR reactor which uses natural uranium as fuel. Using natural Heavy Water Heavy Water uranium as fuel gives the EC6 fuel cycle independence, avoiding the need for enrichment capability or complex fuel Capacity Temperature Fuel transactions. This reactor is designed from experience and feedback gained through the construction and operation of 2084 MWe 287 °C NU CANDU 6 plants, which have been deployed in five countries. The EC6 incorporates innovative features Capacity Pressure Enrichment and new technologies to enhance safety, operation and performance. Key passive safety systems are introduced in 740 MWt 11.1 MPa 0.7 % Candu Energy Inc. Energy Candu addition to the proven systems of the previous CANDU 6 reactors. Furthermore, the EC6 is very flexible with regard Design Life Efficiency Fuel Cycle to the choice of fuel cycle, allowing different countries to Illustration of a Twin EC6 plant layout 60 Yrs 35.5 % Online optimize for local fuel availabilities, national priorities, (Courtesy of Candu Energy Inc.) integration with other reactor technologies and grid sizes. Illustration of vessel and safety shutdown system (Courtesy of Candu Energy Inc.) Nuclear Systems Fuel The EC6 is a 740 MWe HWR that is moderated and cooled by heavy The EC6 fuel bundle consists of 37 elements, each approximately 0.5 water. The fuel is contained in 0.5-m long fuel bundles that reside in metres long containing sintered natural uranium oxide (UO2) pellets, 380 pressure tubes, which are situated in a low pressure, low tem- Zircaloy-4 sheath with Canlub (graphite) coating on the inside surface, perature tank called the calandria and is surrounded by a light-water and two end caps, which are held together with welded end-plates and filled concrete vault. The heat transport system circulates pressurized separated from each other and the pressure tube via appendages. heavy water in two interconnected “figure-of-eight” loops through the Several fuel bundles are exchanged on a daily basis using the automated 380 horizontal fuel channels. It is the same design as in the CANDU 6 online fuel handling system thereby keeping reactor conditions, such reactors, made up of one pressurizer, four steam generators, four heat as core-average reactivity and burnup, constant. transport pumps, and four inlet and outlet headers that connect to the Using NU as fuel permits national fuel cycle independence and pressure tubes via feeder pipes. Reactivity control is achieved auto- technology transfer for localizing fuel manufacture has been achieved 37 Element EC6 Fuel Bundle matically through liquid zone control units using light water and man- successfully in all countries operating CANDU 6 reactors. Several (Courtesy of Candu Energy Inc.) ually through control rods for large power reductions. An automated, other emerging fuel cycles can also be used, such as Recovered or Slightly-Enriched Uranium (RU/SEU - 0.9 – on-power fuel handling and storage system handles fresh fuel loading 1.2% enriched U235 from reprocessed commercial LWR fuel), Natural Uranuim Equivalent (NEU – blended from and spent fuel transfer and storage in an underwater storage bay. Primary components of EC6 HWR used LWR fuel and depleted uranium to obtain 0.7% NU enrichment) reprocessed high-burnup MOX fuel, or thorium Safety based fuel cycles which can benefit long-term energy security of a nation. CANDU 6 reactor plants have been augmented with additional passive, Operation accident resistance and core damage prevention features. Retained Based on experience from CANDU 6 reactors and design features include the two independent passive shutdown systems, each improvements for efficient operation and ease of maintenance, of which is 100% capable of safely shutting down the reactor and situated the EC6 reactor is predicted to have design life of 60 years (with in the low-pressure moderator region. One system uses spring-assisted, one mid-life refurbishment of certain critical equipment, such as gravity-driven shutoff rods while the other injects gadolinium nitrate the fuel channels and feeders) and an overall availability factor of solution from high-pressure tanks into the moderator. ECC provides 92%, achieved through online refuelling and periodic short-duration core refill and cooling by passive accumulator tanks at high pressure maintenance outages of 30 days once every 36 months. Designers and pumps at medium and low pressures. The low-pressure moderator expect potential deployment around the world due to EC6’s high serves as a passive heat sink during postulated accident scenarios and flexibility in terms of source and manufacturing of fuel The standard the large volume of light water surrounding the calandria provides a design of an EC6 nuclear power plant envisions a twin-plant complex, second (passive) core heat sink in case of core melt. Elevated water tank using open-top construction and pre-assembled modules, but a single located in the upper level of the EC6’s reactor building provides passive reactor can be built with no significant changes to the basic design. make-up cooling water via gravity feed to the calandria vessel and the With its comparatively small electrical output, single units could be calandria vault. The Emergency Heat Removal System is designed to deployed to countries with small and medium sized electrical grids. CANDU6 Reactors such as Qinshan NPP, provide an adequate long-term heat sink following an unavailability of Improved plant operability and maintainability, including reducing the HRS to the steam generators, ECC heat exchangers and the heat China are predecessors of the EC6 (Courtesy of Candu Energy Inc.) worker exposure, are achieved by automation of some standard transport system through the ECC piping. The EHRS is a modularized procedures, improved material and plant chemistry, a number of system, each module having its own independently powered, 100% capacity pump. The above systems and a low-flow Layout of primary system in CB health monitoring systems, and the ability to return to full power (Courtesy of Candu Energy Inc.) containment spray system comprise also the new Severe Accident Recovery and Heat Removal System (SARHRS) in immediately following grid interruption. The advanced control room the EC6, which operates with the gravity-driven passive water supply in the short term, the EHRS in the medium term, design and self-diagnostic systems requires a minimum of operator action for all phases of station operation. 26 and an independent diesel-powered pump-driven recovery circuit in the long term. 27 Indian Pressurized HWR-700 uclear Power Corporation of India Ltd. (NPCIL) is a IPHWR-700 Npublic sector enterprise under the Department of Atomic Energy of India, responsible for the design, construction, Coolant Moderator commissioning,and operation of nuclear power plants. HWR Their most recent design is the IPHWR-700, a HWR with Heavy Water Heavy Water

NPCIL horizontal pressure tubes, fuelled by natural uranium and using heavy water as both coolant and moderator. Still Capacity Temperature Fuel under construction it is the largest HWR in a series of Indian PHWR models, following the 220 and 540 MWe version that 700 MWe 288 °C NU are currently in operation. Their development began in the early 1980s with the construction of the 220 MWe reactor Capacity Pressure Enrichment plant. Increased flexibility of fuel sources and management procedures make the IPHWR-700 an attractive option for 2166 MWt 90 Bar 0.7 % India to cope with its growing electricity demand. Its design Illustration of Vessel and Safety Shutdown System is fitted with several active and passive safety systems to Design Life Efficiency Fuel Cycle mitigate the initiation, progression and consequences of 60 Yrs 29 % Online postulated accidents. Rajasthan Atomic Power Station, India [10] Rajasthan 7, 8 Kakrapar 3, 4 Nuclear Systems In contrast to LWRs, the high pressure, high temperature heavy water coolant of an HWR is kept separated from the low pressure, Fuel The IPHWR-700 uses natural uranium fuel with a Zircaloy-4 cladding, thus it has very little excessive reactivity low temperature heavy-water moderator. The core produces 2166 and during normal operation there is no need for neutron poison inside the fuel or the moderator. The lack of MWt of heat which is converted into 700 MWe at an efficiency of excess reactivity means that the reactor must be continuously refuelled during operation. Refuelling rates can 32%.Fuel is situated in an integral horizontal cylindrical calandria vary according to the previous full core reload, but in equilibrium operation stages (about 600 days after fresh with 392 horizontal fuel channels, with some channels reserved core load) approximately one fuel channel is reloaded daily. For optimal fuel utilization and to maximize power for instrumentation. The calandria is surrounded by alight-water- generation over a long period of time, it is possible to load the core with thorium or depleted uranium bundles, but filled concrete vault. The heat transport system consists of 2 loops, this complicates fuel management calculations and supply train issues. The average BU using natural uranium is each with 2 steam generators and 2 circulation pumps. Due to the about 7 GWd/t, but can be increased to 15 GWd/t using fuels with higher fissile content such as Slightly Enriched arrangement of the core with respect to the steam generators,natural Uranium (SEU), MOX, and ThO2. A maximum BU of nearly 30 GW/t has been achieved in the smaller IPHWR-220 circulation forces can be utilized for passive core cooling under test reactors. The relative short fuel lengths help to mitigate consequences of single bundle failures. shutdown conditions. The low temperature moderator and light water around the core act as passive safety heat sinks to mitigate Operation severe accident progression. The seismic design of the NSSS is High flexibility in terms of fuel cycle and manufacturing of fuel make the IPHWR-700 an economical reactor without supposed to safely handle ground acceleration up to 0.214 g and the getting into complex uranium transactions. From a fuel utilization perspective, operating the reactor at close to full CDF has been reduced to less than 10-5 /RY. Primary Configuration of IPHWR-700 power as much as possible would lead to optimal reactor performance. The reactor has an estimated design life Safety (Courtesy of NPCIL) of 40 years due to material changes from ageing and degradation effects from operation. The design to reduce The IPHWR-700 systems are designed taking into consideration the radiation exposure to occupational workers and environmental releases during operation. The plant layout and decades of feedback and experience from operation and construction shielding, through regulatory requirements for occupational radiation protection, minimizes the collective doses experience gained with the IPHWR-220 and IPHWR-540 series,a for plant workers. The main control room is designed to ease the sensory impact on operators and thus reduce total of 18 reactor units. Two independent methods provide total the probability for human errors affecting the availability, reliability, and safety of the plant. To perform a remote shutdown ability. The first consists of gravity driven shutdown shutdown or monitor critical parameters, an environmentally secure and radiation hardened backup control room is rods of cadmium and the second method employs the injection available, in the event the main control room becomes inaccessible. of a strong neutron poison (gadolinium nitrate)into tubes in the calandria. Normal control rods and a liquid zone control system are IPHWR-220 IPHWR-500 IPHWR-700 used for routine power changes. The ECCS is built on 2 ×100% redundancy with single failure criterion for each loop and consists of Power (MWth | MWe) 775 | 235 1730 | 540 2166 | 700 high pressure light water injection, which will absorb fast neutrons. Vessel Diameter (m) 6.4 7.1 7.1 In addition, long term coolant water recirculation is maintained to Avg BU (GWd/t) 6.7 7.5 7 reduce the likelihood of core damage. Part of the severe accident management in the double containment structure consists of the Active fuel length (m) 5.09 5.94 5.94 containment spray system and dedicated connection to the clean-up Avg linear heat rate (kW/m) 28.6 40.1 50.2 system. In the case of loss of external power, four emergency diesel Plant Efficiency (net %) 27.8 28.08 29.08 generators are available as backup power sources. Additionally Hydrogen recombiner test facility for future 4 4 4 there are independent injection connections from diesel driven improvement for HWR Steam Generators pumps to the steam generators, reactor vessel and SPF. (Courtesy of NPCIL) CDF (/RY) 10-5 10-5 10-5 28 29 30 International A tomic E nergy A gency ( IAEA ) Map associated withIAEA’s activities. they are nor publication this of content the for responsible not are entities these that note Please Open SourceImagesthroughout thebooklet: (Left toright,topbottom) Cover Page: would liketothanktheindividualsandorganizationthathaveprovidedtheirservices. image source Open [9] p.28Rajasthan Atomic PowerStation, India:NikhilPatidar(CCviaGoogle) [9] p.25 ABWR unitHamaoka-5,Japan:E-190(CCviaPanorima) [8] p.20 [7] p.18 [6] p.17VVERControlRoom:©RIA Novostiarchive(CCviaWikimediaCommons) [5] p.17Bottom:HexagonalVVERfuelassemblies:©RIA Novosti(CCviaWikimediaCommons) [4] p.17 Top: HexagonalVVERfuelassemblies:Dassteinerne Herz(CCviaWikimediaCommons) [3] p.16 Tianwan 3,4-China:风之清扬(CCviaWikimediaCommons) [2] p.16NovovoronezhII1-Russia:Hullernuc(CCviaWikimediaCommons) [1] p.14ChoozB1,2inFrance: Attribution: ©RaimondSpekking(CCviaWikimediaCommons) AP1000: EPR: APR1400: VVER1000: ABWR: Gundremmingen Units2,3,Germany: First ConcreteatVogtle 3-USA: IAEA PresentationChina2013 Public DomainImageviaWikimediaCommons IAEA Imagebank IAEA Imagebank IAEA Imagebank courtesy, attribution, sources and courtesy,sources attribution, Image Reference Image US-NRC(CCviaFlicker)

Felix König release licenses are listed below.IAEAlisted The are licenses release (CC viaWikimediaCommons) 31 International A tomic E nergy A gency ( IAEA ) Acronyms and Abbreviations Acronyms and abbreviations used in the booklet

AD Automatic Depressurization NPP Nuclear Power Plant ADS Automatic Depressurization System NSSS Nuclear Steam Supply System ALARA As Low As Reasonable Achievable O&M Operation and Maintenance ARIS Advanced Reactor Information System PC Primary Containment ATF Accident Tolerant Fuel PCCS Passive Containment Cooling System ATWS Anticipated Transient without Scram PORV Power Operated Relieve Valve BDB Beyond Design Basis PP Primary Pump BDBA Beyond Design Basis Accident PS Primary System BOP Balance of Plant PSIS Passive Safety Injection System BU Burn-up PWR Pressurized Water Reactor BWR Boiling Water Reactor RC Reactor Circulation CB Containment Building RCCV Reinforced Concrete Containment CDF Core Damage Frequency Vessel CRDM Control Rod Drive Mechanism RCIC Reactor Core Isolation Cooling CC Core Catcher RCP Reactor Coolant Pump CR Control Rods RCS Reactor Coolant System CS Containment Structure RHR Reactor Heat Removal CV Containment Vessel RHRS Residual Heat Removal System CVCV Chemical and Volume Control System RIP Reactor Internal Pumps DBA Design Basis Accident RP Reactor Plant DG Diesel Generators RPV Reactor Pressure Vessel Page intentionally left blank DID Defense in Depth RS Reactor System DVI Direct Vessel Injection RV Reactor Vessel ECCS Emergency Core Cooling System RY Reactor Years EDG Emergency Diesel Generator SA Sever Accident ESWS Essential Service Water System SBO Station Black-Out EUR European Utility Requirements SBWR Simplified Boiling Water Reactor FA Fuel Assembly SG Steam Generator FE Fuel Element SGTR Steam Generator Tube Rupture GDCS Gravity Driven Cooling System SIS Safety Injection System HEU High Enriched Uranium SLU Slightly Enriched Uranium HMS Hydrogen Mitigations System SPF Spent Fuel Pool HPI High Pressure Injection SS Safety System HRS Heat Removal System HVAC Heat, Ventilation, Air Conditioning IC Isolation Condenser I&C Instrumentation and Control LEU Low Enriched Uranium LBLOCA Large Break LOCA LOCA Loss of Coolant Accident LPI Low Pressure Injection LWR Light Water Reactor MOX Mixed Oxide MSL Main Steam Line MSLB Main Steam Line Break MWe Mega Watt electric MWt Mega Watt thermal

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