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The nuclear fuel cycle
CHAPTER
The nuclear fuel cycle
6
Chapter outline 6.1 6.2 6.3 6.4 6.5
Introduction .....................................................................................................329 The Status of nuclear power in the world...........................................................330 Nuclear safety..................................................................................................357 Releases of effluents ........................................................................................360 Management of radioactive wastes ...................................................................364 6.5.1 Storage and disposal.......................................................................367 6.5.2 Processing of used nuclear fuel........................................................368 6.5.3 Smart use of nuclear waste..............................................................373 6.6 Research reactors ............................................................................................374 6.7 Advanced nuclear power plants ........................................................................377 6.8 Nuclear fusion .................................................................................................381 6.8.1 ITER..............................................................................................386 6.9 Nuclear batteries .............................................................................................388 References .............................................................................................................392 Further reading .......................................................................................................394
6.1 Introduction This subject is a part of a broader subject which could be called “a human need for energy”. According to the World Energy Council (WEC), the demand for energy is expected to increase further. The reasons for this are clear: 50% of the countries of the world are in the process of industrialization with economies growing rapidly, some at 10% or more. This calls for an ever-increasing supply of coal, oil and now natural gas. China, for instance, which sustained a 10% growth rate in its market economy for over a decade, burns 1.2 Gt of coal per year (1996 figures) and expects to burn 1.4 Gt of coal per year after 2000. In addition, it imports oil, has an embryonic nuclear industry, and is seeking to diversify its renewable energy supplies beyond hydroelectric power. The global situation is made more challenging by the expected rapid increase in the world population. UN figures suggest that the current population of nearly 6 billion will
Radioactivity in the Environment. https://doi.org/10.1016/B978-0-444-64146-5.00006-9 # 2019 Elsevier B.V. All rights reserved.
329
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CHAPTER 6 The nuclear fuel cycle
increase to over 10 billion by 2050; 80% of that population will be in developing countries where, to put matters into perspective, 50% of the population do not have an electricity supply connection (neither do they have safe drinking water) so their demand for the services provided by energy and an improved life-style is undeniable (Fells, 1998). There are three energy sources, all of which have enormous potential, which might give us confidence that the world’s energy needs will be met in the second half of the 21st century. These are nuclear energy, renewable energy and nuclear fusion. They have the advantage that they are also environmentally clean and do not emit large quantities of greenhouse gases. Nuclear fusion has enormous potential, but will be very difficult to achieve from an engineering point of view. It is prudent to rely, for the time being at any rate, on the growth in renewable and nuclear energy. They are not, incidentally, mutually exclusive but complementary, and it is very difficult to imagine any future scenario post-2050 without a large slice of energy from both sources. Nuclear power, in particular, plays an important role in controlling carbon-dioxide emissions. The nuclear future may well lie with the fast-breeder reactor which uses uranium some 60 times more efficiently than current fission reactors. Prototype breeder reactors are in operation in Russia and Japan and were so until recently in the United Kingdom and France. They will be required post-2030 or so if a major new nuclear programme is embarked on, as uranium resources are predicted to start running into short supply at about that time. The problem with the continuing growth in nuclear power is the public’s perception of its safety, particularly the safety of radioactive waste disposal. If the spent nuclear fuel is reprocessed, as happens in the United Kingdom and France, and the high-level radioactive waste classified and stored in metal cylinders in a dry rock store, there is general international consensus that it will be very safety contained. In addition, the method of reprocessing recovers plutonium and unused uranium, both of which can be recycled as fuel for fission reactors. The other method of dealing with spent nuclear fuel is to merely store it either in cooling ponds or directly in a dry store. This presents a less tidy legacy for the future but is the preferred method in some countries, fearful that reprocessing will lead to proliferation (Fells, 1998).
6.2 The status of nuclear power in the world The concept of a nuclear fuel cycle is an old one, almost dating back to the concept of controlled nuclear fission to generate electricity. At the time of the development of the first nuclear power plants, it was generally taken for granted that fuel from power reactors would be reprocessed and that the recovered uranium and plutonium would be recycled.
6.2 The status of nuclear power in the world
In those days, uranium ore was a scarce and expensive commodity and it was naturally assumed that economically available supplies would not meet the demands required by a widespread use of nuclear power. Consequently, the extraction of all the potential energy content of uranium-235 seemed to be essential. Such a complete exploitation of uranium resources requires reprocessing of the spent fuel and the extraction of plutonium for burning in specially designed “fast” reactors. The approach became more attractive with the concept of fast-breeder reactors, which could produce more fuel than they consumed. For such reasons, many countries during the 1960s attached high priority to the development of fast reactors, and it was anticipated that they would be widely deployed in the 1980s (Semenov and Oi, 1993). Until the early 1970s then, the nuclear fuel cycle was pictured as an orderly sequence of processes. It began with uranium mining, milling, and conversion, was followed by fuel enrichment, fuel fabrication and power generation and was finally completed by reprocessing, recycling of plutonium and uranium to fast reactors and final disposal of waste streams from reprocessing plants. In essence, closure of the fuel cycle meant the effective use of plutonium. Three different types of fuel cycle are commonly identified for nuclear power generation, depending on whether fuel is recycled and on the type of reactor used for electricity production. • •
•
The “once-through” fuel cycle: in this cycle, the spent fuel is not reprocessed but kept in storage until it is eventually disposed of as waste. The thermal reactor cycle: in this cycle, the spent fuel is reprocessed and the uranium and plutonium can be recycled in new fuel elements. It is also possible to recycle only the uranium and to store the plutonium, and vice versa. The fast-breeder reactor cycle: in this cycle, the spent fuel is similarly reprocessed and the uranium and plutonium fabricated into new fuel elements. However, they are recycled to fast-breeder reactors, in which there is a central core of uranium/plutonium fuel surrounded by a blanket of depleted uranium (uranium from which most of the uranium-235 atoms have been removed during the process of enrichment) or to burner reactors. This depleted uranium consists mostly of uranium-238 atoms, some of which are converted to plutonium during irradiation. By suitable operation, fast-breeder reactors thus can produce slightly more fuel than they consume, hence the name “breeder” (see Fig. 6.1).
The situation has changed dramatically during the last 20 years. No closed fuel cycle of the type originally envisaged to be operational in the 1980s exists today. Although the closure of the nuclear fuel cycle has been experimentally demonstrated in France, Japan, Russia and the United Kingdom, it has not been demonstrated yet on a commercial scale. Current thinking is divided into two schools. One believes that plutonium as an energy source has no economic value and spent fuel should be disposed of in a safe
331
332
CHAPTER 6 The nuclear fuel cycle
Interim spent fuel storage Reprocessing plant
Spent fuel Uranium fuel
Thermal reactor
Enriched UF6
Reprocessing plant Recovered uranium
235
U enrichment
Spent fuel UF6
Pu oxide
Fuel fabrication
U/Pu (core) Max. fuel
Depleted
238
U (blanket)
Spent fuel
Fast reactor
Thermal reactor
Uranium fuel
(A)
U
Plutonium oxide
Max. fuel fabrication
UF6
U3O8
238
Pu
Uranium fuel fabrication
(B) Spent fuel and waste disposal
From enrichment plant
(C) Waste disposal
Waste disposal
FIG. 6.1 Schematic representations of (A) the once-through cycle, (B) the thermal reactor cycle and (c) the fast reactor cycle. U3O8 ¼ yellowcake, UF6 ¼ uranium hexafluoride and MOX ¼ mixed oxide fuel (uranium/plutonium). After Semenov, B.A. and Oi, N., 1993. Nuclear fuel cycles: adjusting to new realities. IAEA Bulletin 3, 2.
Extraction
Enrichment
Fuel fabrication
Nuclear power plant
Reprocessing Spent fuel Mining
HLW Conditioning
Waste disposal
FIG. 6.2 A schematic representation of the nuclear fuel cycle.
way (the “once-through” option). The other essentially adheres to the traditional nuclear fuel cycle (closed cycle option as illustrated in Fig. 6.2). The difference of opinions stems mainly from the predictions of nuclear electricity growth and the predicted availability of economical supplies of uranium, although it is influenced by political and environmental issues as well. We will return to this matter again. The present situation on the use of nuclear energy for electricity production is illustrated in Table 6.1 which lists power stations
6.2 The status of nuclear power in the world
Table 6.1 Nuclear power status in the world at the end of 2016
Argentina Armenia Belarus Belgium Brazil Bulgaria Canada China Czech Republic Finland France Germany Hungary India Iran Japan Korea, Republic of Mexico Netherlands Pakistan Romania Russian Federation South Africa Slovak Republic Slovenia Spain Sweden Switzerland Taiwan, China United Kingdom
In operation
Under construction
Nuclear electricity production in 2016
No. of units
No. of units
TW-h
Total net MWE
3 1
1632 375
7 2 2 19 36 6
5913 1884 1926 13,554 31,384 3930
4 58 8 4 22 1 42 25
2764 63,130 10,799 1889 6240 915 39,752 23,077
2 1 4 2 35
1552 482 1005 1300 26,111
2 4
1860 1814
1 7 10 5 6
688 7121 9740 3333 5052
15
8918
Total net MWE
1
25
2
2218
1
1245
21 2
21,622 1824
1 1
1600 1630
5
2990
2 3
% of total
7.7 2.2
5.1 31.4
41.4 15.0 15.1 95.7 197.8 22.7
51.2 2.6 35.0 14.0 3.5 29.3
2653 4020
22.3 386.5 80.1 15.2 35.0 5.9 17.5 154.3
33.7 72.8 13.0 51.3 2.6 2.1 1.8 29.2
3 1 7
2343 650 5520
10.3 3.7 5.4 10.4 184.1
3.5 3.4 5.2 16.5 18.1
2
880
15.2 13.7
6.5 54.1
2600
5.4 56.1 60.6 20.3 30.5
35.2 21.2 40.1 34.4 12.1
65.1
20.4
2
Continued
333
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CHAPTER 6 The nuclear fuel cycle
Table 6.1 Nuclear power status in the world at the end of 2016 Continued
Ukraine United Arab Emirates United States World total
In operation
Under construction
Nuclear electricity production in 2016
No. of units
Total net MWE
No. of units
TW-h
15
13,107
2 4
2070 5380
76.1
50.2
99
99,869
4
4468
804.9
19.6
448
391,116
61
61,264
2476.2
10.6
Total net MWE
% of total
After IAEA, 2017. Energy, Electricity and Nuclear Power Estimates for the Period up to 2050. Reference Data Series No. 1, 2017 Ed. International Atomic Energy Agency, Vienna, 2017.
Table 6.2 Nuclear power units by reactor type, worldwide, as of December 2015 Type code
Reactor type—full name
In operation
Shut down
Under construction
282
46
57
78 14
36 38
4 –
49
8
4
15
9
–
3
7
1
No. units PWR BWR GCR PHWR LWGR FBR
Pressurized light-water reactors Boiling light-water reactors Gas-cooled, graphite moderated reactors Pressurized heavy-water reactors; all types Graphite-moderated lightwater reactors Fast-breeder reactors
After IAEA, 2016. Nuclear Power Reactors in the World. Reference Data Series No. 2, 2016 Ed. International Atomic Energy Agency, Vienna, 2016.
in individual countries. This is the status as of December 2016 as reported by IAEA. All these stations can be grouped by reactor type as shown in Table 6.2. Once a year Nuclear News publishes the world list of nuclear power plants which are operable, under construction, or an order (for power 30 MWe and over). The latest world list of nuclear power plants can be found in March 2018 issue of Nuclear News. Data on nuclear plants worldwide are operable, under construction, or on order as of 31 December 2017. Information are provided on net MWe, reactor type,
6.2 The status of nuclear power in the world
reactor model, initial criticality, commercial start, reactor supplier and the maps showing the location of each plant site together with GPS coordinates. We list the world power stations, shown in Table 6.3, based on information gathered from many sources, mainly national authorities listed in the table. It is of interest to refer to the article by Blix (1997) who was director general of IAEA for many years until 1997. At that time, oil, gas and coal—the fossil fuels— provided nearly 85% of the commercial energy that the world used: close to 37% for oil, 25% for coal and more than 21% for gas, with nuclear power and hydro power providing around 7% each, and commercial renewables such as solar, wind and biomass nearly 2.5%. (Noncommercial uses of renewable energy were estimated to provide another 10% of world energy consumption.) In China, coal supplied 75% of energy consumption, oil about 17%, nuclear and hydro 5%, and gas 2%. According to Blix (1997), the so-called renewable sources total a little more than 2% of world commercial energy. The bulk of that total comes from geothermal Table 6.3 World list of nuclear power plants Country
Authority
Power station
Argentina
Commision Nacional de Energia Atomica (CNEA)
Atucha 1 (Lima, Buenos Aires) Embalse (Rio Tercero, Cordoba) Armenia 2 (Metsa, pr. Armenia) Doel 1 (Doel, East Flanders) Doel 2 (Doel, East Flanders) Doel 3 (Doel, East Flanders) Doel 4 (Doel, East Flanders) Tihange 1 (Huy, Liege) Tihange 2 (Huy, Liege) Tihange 3 (Huy, Liege) Angra 1 (Itaorna, Rio de Janeiro) Angra 2 (Itaorna, Rio de Janeiro) Kozloduy 5 (Kozloduy, Vratsa) Kozloduy 6 (Kozloduy, Vratsa)
Armenia
Ministry of nuclear power
Belgium
Electrabel
Brazil
Bulgaria
Fumas Centrais Electricas SA
National Electric Co.
Net MWe
Type
335
PHWR
600
PHWR
400
PWR
392
PWR
392
PWR
985
PWR
1001
PWR
962 1008 1015 626
PWR PWR PWR PWR
1270
PWR
953
PWR
953
PWR Continued
335
336
CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
Canada
New Brunswick Power Corp.Ontario Hydro
Point Lepreau (Bay of Fundy, N.B.) Pickering 1 (Pickering, Ont.) Pickering 4 (Pickering, Ont.) Pickering 5 (Pickering, Ont.) Pickering 6 (Pickering, Ont.) Pickering 7 (Pickering, Ont.) Pickering 8 (Pickering, Ont.) Bruce 3 (Tiverton, Ont.) Bruce 4 (Tiverton, Ont.) Bruce 5 (Tiverton, Ont.) Bruce 6 (Tiverton, Ont.) Bruce 7 (Tiverton, Ont.) Bruce 8 (Tiverton, Ont.) Darlington 1 (Newcastle Twp., Ont.) Darlington 2 (Newcastle Twp., Ont.) Darlington 3 (Newcastle Twp., Ont.) Darlington 4 (Newcastle Twp., Ont.) Point Lepreau Gentilly 2 (Becancour, Que.) Qinshan 1 (Haiyan, Zhejiang) Qinshan 2 (Haiyan, Zhejiang) Qinshan 3 (Haiyan, Zhejiang) Qinshan 4 (Haiyan, Zhejiang)
New Brunswick Power Hydro Quebec China
Ministry of Nuclear Industry
Net MWe
Type
640
PHWR
515
PHWR
515
PHWR
516
PHWR
516
PHWR
516
PHWR
516
PHWR
769 769 785 785 785 785 881
PHWR PHWR PHWR PHWR PHWR PHWR PHWR
881
PHWR
881
PHWR
881
PHWR
635 638
PHWR PHWR
300
PWR
600
PWR
600
PWR
700
PHWR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Authority
Guangdong Nuclear Power Joint Venture Co. Ltd. Lingao Nuclear Co.
Cuba
Czech Republic
Finland
Ministry of Basic Industries
Czech Power Board
Imatran Voima Oy (IVO)
Teollisuuden Voima Oy (TVO)
Power station Qinshan 5 (Haiyan, Zhejiang) Qinshan 6 (Haiyan, Zhejiang) Guangdong 1 (Shenzhen, Guangdong) Guangdong 2 (Shenzhen, Guangdong) Lingao 1 (Lingao, Guangdong) Lingao 2 (Lingao, Guangdong) Lingao 2 (Lingao, Guangdong) Juragua 1 (Cienfuegos, Cienfuegos) Juragua 2 (Cienfuegos, Cienfuegos) Dukovany 1 (Trebic, Jihomoravsky) Dukovany 2 (Trebic, Jihomoravsky) Dukovany 3 (Trebic, Jihomoravsky) Dukovany 4 (Trebic, Jihomoravsky) Temelin 1 (Temelin, Jihocesky) Temelin 2 (Temelin, Jihocesky) Loviisa 1 (Loviisa, UUsimaa) Loviisa 2 (Loviisa, UUsimaa) TVO 1 (Olkiluoto, Turku-Pori) TVO 2 (Olkiluoto, Turku-Pori)
Net MWe
Type
700
PHWR
650
PWR
944
PWR
944
PWR
985
PWR
985
PWR
1000
PWR
417
PWR
417
PWR
428
PWR
412
PWR
4471
PWR
470
PWR
963
PWR
963
PWR
488
PWR
488
PWR
840
BWR
860
BWR Continued
337
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
France
Commissariat a l’Energie Atomique Centrale Nucleaire Europeene a Neutrons Rapides S.A. (NERSA) Electricite de France (EdF)
Phenix (Marcoule, Gard) Creys-Malville (Bouvesse, Isere) Chinon B1 (Chinon, Indre-et-Loire) Chinon B2 (Chinon, Indre-et-Loire) Chinon B3 (Chinon, Indre-et-Loire) Chinon B4 (Chinon, Indre-et-Loire) Saint-Laurent B1 (SaintLaurent-des-Eaux, Loir-et-Cher) Saint-Laurent B2 (SaintLaurent-des-Eaux, Loir-et-Cher) Bugey 2 (Loyettes, Ain) Bugey 3 (Loyettes, Ain) Bugey 4 (Loyettes, Ain) Bugey 5 (Loyettes, Ain) Fessenheim 1 (Fessenheim, Haut-Rhin) Fessenheim 2 (Fessenheim, Haut-Rhin) Dampierre 1 (Ouzouer, Loiret) Dampierre 2 (Ouzouer, Loiret) Dampierre 3 (Ouzouer, Loiret) Dampierre 4 (Ouzouer, Loiret) Gravelines B1 (Gravelines, Nord) Gravelines B2 (Gravelines, Nord) Gravelines B3 (Gravelines, Nord) Gravelines B4 (Gravelines, Nord) Gravelines B5 (Gravelines, Nord)
Net MWe 233 600
Type LMFBR
905 905
915
LMFBR PWR PWR PWR PWR PWR
915
PWR
910 910 880 880 880
PWR PWR PWR PWR PWR
880
PWR
890
PWR
890
PWR
890
PWR
890
PWR
915
PWR
915
PWR
915
PWR
915
PWR
915
PWR
905 905
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station Gravelines B6 (Gravelines, Nord) Tricastin 1 (Pierrelatte, Drome) Tricastin 2 (Pierrelatte, Drome) Tricastin 3 (Pierrelatte, Drome) Tricastin 4 (Pierrelatte, Drome) Blayais 1 (Blaye, Gironde) Blayais 2 (Blaye, Gironde) Blayais 3 (Blaye, Gironde) Blayais 4 (Blaye, Gironde) Paluel 1 (Veulettes, Seine-Maritime) Paluel 2 (Veulettes, Seine-Maritime) Paluel 3 (Veulettes, Seine-Maritime) Paluel 4 (Veulettes, Seine-Maritime) Cruas 1 (Cruas, Ardeche) Cruas 2 (Cruas, Ardeche) Cruas 3 (Cruas, Ardeche) Cruas 4 (Cruas, Ardeche) Saint-Alban 1 (Auberives, Isere) Saint-Alban 2 (Auberives, Isere) Flamanville 1 (Flamanville, Manche) Flamanville 2 (Flamanville, Manche) Cattenom 1 (Cattenom, Moselle) Cattenom 2 (Cattenom, Moselle)
Net MWe
Type
915
PWR
915
PWR
915
PWR
915
PWR
915
PWR
910
PWR
910
PWR
910
PWR
910
PWR
1330
PWR
1330
PWR
1330
PWR
1330
PWR
915 915 915 915 1335
PWR PWR PWR PWR PWR
1335
PWR
1330
PWR
1330
PWR
1300
PWR
1300
PWR Continued
339
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Germany
Authority
Bayernwerk AG Gemeinschaftskernkraft werk Grohnde GmbH (KWG) Gemeinschaftskernkraftwerk Neckar (GKN) Kernkraftwerk Brokdorf GmbH (KBR) Kernkraftwerk Brunsbuettel GmbH (KKB)
Power station Cattenom 3 (Cattenom, Moselle) Cattenom 4 (Cattenom, Moselle) Belleville 1 (Belleville s/Loire, Cher) Belleville 2 (Belleville s/Loire, Cher) Nogent s/Seine 1 (Nogent s/Seine, Aube) Nogent s/Seine 2 (Nogent s/Seine, Aube) Penly 1 (Saint-Martinen-Campagne, SeineMaritime) Penly 1 (Saint-Martinen-Campagne, SeineMaritime) Golfech 1 (Valence, Tarn et Garonne) Golfech 2 (Valence, Tarn et Garonne) Chooz B1 (Chooz, Ardennes) Chooz B2 (Chooz, Ardennes) Civaux 1 (Civaux, Vienne) Civaux 2 (Civaux, Vienne) Grafenrheinfeld KKG (Gragenrheinfeld, Ba.) Grohnde (Emmerthal, Nied.) Neckar 1 (Neckarwestheim, B.-W.) Neckar 1 (Neckarwestheim, B.-W.) Brokdorf (Brokdorf, S.-H.) Brunsbuettel (Brunsbuettel, S.-H.)
Net MWe
Type
1300
PWR
1300
PWR
1310
PWR
1310
PWR
1310
PWR
1310
PWR
1330
PWR
1330
PWR
1310
PWR
1310
PWR
1455
PWR
1455
PWR
1450
PWR
1450
PWR
1275
PWR
1360
PWR
785
PWR
1310
PWR
1370
PWR
771
BWR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Net MWe
Type
Isar 1 (Essenbach, Ba.) Isar 2 (Essenbach, Ba.) Kruemmel (Geesthacht, S.-H.) Hochtief/Hammers/ Heitkamp/Holzmann
878 1400 1260
BWR PWR BWR
Emsland (Lingen, Nied.)
1329
PWR
Philippsburg 1 (Philippsburg, B.-W.) Philippsburg 2 (Philippsburg, B.-W.) Gundremmingen B (Gundremmingen, Ba.) Gundremmingen C (Gundremmingen, Ba.) Unterweser (Rodenkirchen, Nied.)
890
BWR
1392
PWR
1284
BWR
1288
BWR
1345
PWR
Biblis A (Biblis, Hessen)
1167
PWR
Biblis A (Biblis, Hessen) Paks 1 (Paks, Tolna) Paks 2 (Paks, Tolna) Paks 3 (Paks, Tolna) Paks 4 (Paks, Tolna) Tarapur 1 (Tarapur, Maharashtra) Tarapur 2 (Tarapur, Maharashtra) Tarapur 3 (Tarapur, Maharashtra) Tarapur 4 (Tarapur, Maharashtra) Rajasthan 1 (Kota, Rajasthan) Rajasthan 2 (Kota, Rajasthan) Rajasthan 3 (Kota, Rajasthan) Rajasthan 4 (Kota, Rajasthan) Rajasthan 5 (Kota, Rajasthan)
1240 470 473 473 473 150
PWR PWR PWR PWR PWR BWR
150
BWR
490
PWHR
490
PWHR
90
PWHR
187
PWHR
202
PWHR
202
PWHR
202
PWHR
Authority
Power station
Kernkraftwerk Isar (KKI) Kernkraftwerk
Kruemmel Gmbh (KKK) Kernkraftwerke LippeEms GmbH (KKE) Kernkraftwerk Philippsburg (KKP) Kernkraftwerk RWE-Bayemwerk GmbH (KRB) Kernkraftwerk Unterweser GmbH (KKU) RWE Energie Aktiengesellshaft Hungary
Hungarian Power Companies, Ltd.
India
Atomic Energy Commission, Department of Atomic Energy
Continued
341
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Japan
Authority
Chubu Electric Power Co.
Chugoku Electric Power Co., Inc.
Hokkaido Electric Power Co.
Hokuriku Electric Power Co.
Japan Atomic Power Co. Ltd.
Power station Madras 1 (Kalpakkam, Tamil Nadu) Madras 2 (Kalpakkam, Tamil Nadu) Narora 1 (Narora, Uttar Pradesh) Narora 2 (Narora, Uttar Pradesh) Kakrapar 1 (Kakrapar, Gujarat) Kakrapar 2 (Kakrapar, Gujarat) Kaiga 1 (Kaiga, Karnataka) Kaiga 2 (Kaiga, Karnataka) Kaiga 3 (Kaiga, Karnataka) Hamaoka 3 (Hamaokacho, Shizuoka) Hamaoka 4 (Hamaokacho, Shizuoka) Hamaoka 5 (Hamaokacho, Shizuoka) Shimane 1 (Kashimacho, Shimane) Shimane 2 (Kashimacho, Shimane) Tomari 1 (Tomari-mura, Hokkaido) Tomari 2 (Tomari-mura, Hokkaido) Tomari 3 (Tomari-mura, Hokkaido) Shika 1 (Shika-machi, Ishikawa) Shika 2 (Shika-machi, Ishikawa) Tokai 2 (Tokai Mura, Ibaraki) Tsuruga 1 (Tsuruga, Fukui) Tsuruga 2 (Tsuruga, Fukui)
Net MWe
Type
155
PHWR
155
PHWR
202
PHWR
202
PHWR
202
PHWR
202
PHWR
202
PHWR
202
PHWR
202
PHWR
1056
BWR
1092
AEWR
1380
BWR
439
BWR
790
BWR
550
PWR
550
PWR
912
PWR
505
BWR
1358
AEWR
1056
BWR
341
BWR
1115
PWR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
Kansai Electric Power Co., Inc.
Mihama 1 (Mihama-cho, Fukui) Mihama 2 (Mihama-cho, Fukui) Mihama 3 (Mihama-cho, Fukui) Takahama 1 (Takahama-cho, Fukui) Takahama 2 (Takahama-cho, Fukui) Takahama 3 (Takahama-cho, Fukui) Takahama 4 (Takahama-cho, Fukui) Ohi 1 (Ohi-cho, Fukui) Ohi 2 (Ohi-cho, Fukui) Ohi 3 (Ohi-cho, Fukui) Ohi 4 (Ohi-cho, Fukui) Genkai 1 (Genkai, Saga)
Kyushu Electric Power Co., Inc.
Development Corp. (PNC) Shikoku Electric Power Co.
Tohoku Electric Power Co., Inc.
Genkai 2 (Genkai, Saga) Genkai 3 (Genkai, Saga) Genkai 4 (Genkai, Saga) Sendai 1 (Sendai, Kagoshima) Sendai 2 (Sendai, Kagoshima) Monju FBR (Tsuruga, Fukui) Ikata 1 (Ikata-cho, Ehime) Ikata 2 (Ikata-cho, Ehime) Ikata 3 (Ikata-cho, Ehime) Higashidori 1 (Higashidori, Aomori) Onagawa 1 (Onagawa, Miyagi) Onagawa 2 (Onagawa, Miyagi) Onagawa 3 (Onagawa, Miyagi)
Net MWe
Type
320
PWR
470
PWR
780
PWR
780
PWR
780
PWR
830
PWR
830
PWR
1120 1120 1127 1127 529
PWR PWR PWR PWR PWR
529 1127 1127 846
PWR PWR PWR PWR
846
PWR
280
LMFBR
538
PWR
538
PWR
846
PWR
1100
BWR
498
BWR
796
BWR
796
BWR Continued
343
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Kazakhstan
Korea
Net MWe
Type
Fukushima Daiichi 1 (Ohkuma, Fukushima) Fukushima Daiichi 2 (Ohkuma, Fukushima) Fukushima Daiichi 3 (Ohkuma, Fukushima) Fukushima Daiichi 4 (Ohkuma, Fukushima) Fukushima Daiichi 5 (Ohkuma, Fukushima) Fukushima Daiichi 6 (Ohkuma, Fukushima) Fukushima Daini 1 (Naraha, Fukushima) Fukushima Daini 2 (Naraha, Fukushima) Fukushima Daini 3 (Naraha, Fukushima) Fukushima Daini 4 (Naraha, Fukushima) Kashiwazaki Kariwa 1 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 2 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 3 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 4 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 5 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 6 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 7 (Kashiwazaki, Niigata) BN-350 (Aktau, Mangyshlak)
439
BWR
760
BWR
760
BWR
760
BWR
760
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1315
BWR
1315
BWR
135
LMFBR
Kori Kori Kori Kori
563 612 903 903
PWR PWR PWR PWR
Authority
Power station
Tokyo Electric Power Co.
Kazakh State Atomic Power Engineering and Industry Corp. (KATEP) Korea Electric Power Corp.
1 (Kori, 2 (Kori, 3 (Kori, 4 (Kori,
Kyongnam) Kyongnam) Kyongnam) Kyongnam)
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Mexico
The Netherlands
Pakistan
Authority
Comision Federal de Electricidad
NV Elektriciteits Produktiemaatschappij Zuid-Nederland (NV EPZ) Pakistan Atomic Energy Commission
Net MWe
Type
Wolsong 1 (Kyongju, Kyongbuk) Wolsong 2 (Kyongju, Kyongbuk) Wolsong 3 (Kyongju, Kyongbuk) Wolsong 4 (Kyongju, Kyongbuk) Yonggwang 1 (Yonggwang, Chonnam) Yonggwang 2 (Yonggwang, Chonnam) Yonggwang 3 (Yonggwang, Chonnam) Yonggwang 4 (Yonggwang, Chonnam) Yonggwang 5 (Yonggwang, Chonnam) Yonggwang 6 (Yonggwang, Chonnam) Ulchin 1 (Ulchin, Kyongbuk) Ulchin 2 (Ulchin, Kyongbuk) Ulchin 3 (Ulchin, Kyongbuk) Ulchin 4 (Ulchin, Kyongbuk) Ulchin 5 (Ulchin, Kyongbuk) Ulchin 6 (Ulchin, Kyongbuk) Laguna Verde 1 (Laguna Verde, Veracruz) Laguna Verde 2 (Laguna Verde, Veracruz) Borssele (Borssele, Zeeland)
629
PHWR
700
PHWR
700
PHWR
700
PHWR
900
PWR
900
PWR
950
PWR
950
PWR
950
PWR
950
PWR
920
PWR
920
PWR
950
PWR
950
PWR
950
PWR
950
PWR
800
BWR
800
BWR
452
PWR
Kanupp-1 (Karachi, Sind) Chasnupp (Mianwali, Punjab)
125
PHWR
300
PWR
Power station
Continued
345
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
Romania
Romanian Electricity
Cernavoda 1 (Cernavoda, Constanta) Cernavoda 2 (Cernavoda, Constanta) Balakovo 1 (Balakovo, Saratov) Balakovo 2 (Balakovo, Saratov) Balakovo 3 (Balakovo, Saratov) Balakovo 4 (Balakovo, Saratov) Beloyarskiy 3 (BN-600) (Zarechnyy, Sverdlovsk) Bilibino-1 (Chukotka Autonomous Okrug) Bilibino-2 (Chukotka Autonomous Okrug) Bilibino-3 (Chukotka Autonomous Okrug) Bilibino-4 (Chukotka Autonomous Okrug) Kalinin 1 (Udomlya, Tver) Kalinin 2 (Udomlya, Tver) Kalinin 3 (Udomlya, Tver) Kola 1 (Polyarnyye Zori, Murmansk) Kola 2 (Polyarnyye Zori, Murmansk) Kola 3 (Polyarnyye Zori, Murmansk) Kola 4 (Polyarnyye Zori, Murmansk) Kursk 1 (Kurchatov, Kursk) Kursk 2 (Kurchatov, Kursk) Kursk 3 (Kurchatov, Kursk) Kursk 4 (Kurchatov, Kursk)
Authority (RENEL) Russia
Ministry of Atomic Power
Net MWe
Type
655
PHWR
650
PHWR
950
PWR
950
PWR
950
PWR
950
PWR
560
LMFBR
12
LWGR
12
LWGR
12
LWGR
12
LWGR
950 950 950 411
PWR PWR PWR PWR
411
PWR
411
PWR
411
PWR
925
LGR
925
LGR
925
LGR
925
LGR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Slovakia
Slovenia
Authority
Slovak Power Board
Nuklearna Elektrana Krsko
Power station Leningrad 1 (Sosnovyy Bor, St. Petersburg) Leningrad 2 (Sosnovyy Bor, St. Petersburg) Leningrad 3 (Sosnovyy Bor, St. Petersburg) Leningrad 4 (Sosnovyy Bor, St. Petersburg) Novovoronezhskiy 3 (Novovoronezhskiy, Voronezh) Novovoronezhskiy 4 (Novovoronezhskiy, Voronezh) Novovoronezhskiy 5 (Novovoronezhskiy, Voronezh) Smolensk 1 (Desnogorsk, Smolensk) Rostov-1 (Volgodonsk-1) Rostov-2 (Volgodonsk-2) Smolensk 2 (Desnogorsk, Smolensk) Smolensk 3 (Desnogorsk, Smolensk) South Urals 1 (Chelyabinsk, Chelyabinsk) South Urals 2 (Chelyabinsk, Chelyabinsk) Vk-50 (Dimitrovgrad, Ulyanovsk, RSFSR) Bohunice 3 (Trnava, Zapadoslovensky) Bohunice 4 (Trnava, Zapadoslovensky) Mochovce 1 (Mochovce, Zapadoslovensky) Mochovce 2 (Mochovce, Zapadoslovensky) Krsko (Krsko, Vrbina)
Net MWe
Type
925
LGR
925
LGR
925
LGR
925
LGR
385
PWR
385
PWR
950
PWR
925
LGR
950 950 925
PWR PWR LGR
925
LGR
750
LMFBR
750
LMFBR
50
PWR
408
PWR
408
PWR
420
PWR
420
PWR
676
PWR Continued
347
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
South Africa
ESKOM
Koeberg 1 (Melkbosstrand, Cape) Koeberg 2 (Melkbosstrand, Cape) Asco 1 (Asco, Tarragona) Asco 2 (Asco, Tarragona) Vandellos 2 (Vandellos, Tarragona)
Spain
Asociacion Nuclear Asco Endesa—Iberdrola S.A. Central Nuclear Vandellos II, A.I.E. Central de Trillo Central Nuclear de Almaraz Centrales Nucleares del Norte, SA Iberdrola SA
Sweden
OKG Aktiebolag
Vattenfall
Sydkraft AB
Trillo 1 (Trillo, Guadalajara) Almaraz 1 (Almaraz, Caceres) Almaraz 2 (Almaraz, Caceres) Santa Maria de Garona (Santa Maria de Garona, Burgos) Cofrentes (Cofrentes, Valencia) Oskarshamn 1 (Oskarshamn, Kalmar) Oskarshamn 2 (Oskarshamn, Kalmar) Oskarshamn 3 (Oskarshamn, Kalmar) Ringhals 1 (Varberg, Halland) Ringhals 2 (Varberg, Halland) Ringhals 3 (Varberg, Halland) Ringhals 4 (Varberg, Halland) Forsmark 1 (Forsmark, Uppsala) Forsmark 2 (Forsmark, Uppsala) Forsmark 3 (Forsmark, Uppsala)
Net MWe
Type
920
PWR
920
PWR
996
PWR
992
PWR
1045
PWR
1003
PWR
941
PWR
950
PWR
446
BWR
1063
BWR
445
BWR
605
BWR
1400
BWR
835
BWR
875
PWR
915
PWR
915
PWR
970
PWR
970
BWR
1155
BWR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
Switzerland
Bernische Kraftwerke AG (BKW) Kernkraftwerk GoesgenDaeniken AG Kernkraftwerk Leibstadt AG Nordostschweizerische Kraftwerk AG (NOK)
Muehleberg (Muehleberg, Bern) Goesgen (Daeniken, Solothurn) Leibstadt (Leibstadt, Aargau) Beznau-1 (Doettingen, Aargau) Beznau-2 (Doettingen, Aargau) Chinshan 1 (Chinshan, Taipei) Chinshan 2 (Chinshan, Taipei) Kuosheng 1 (Kuosheng, Wang-Li, Taipei) Kuosheng 2 (Kuosheng, Wang-Li, Taipei) Maanshan 1 (Herng Chuen) Maanshan 2 (Herng Chuen) Khmel’nitskiy 1 (Neteshin, Khmel’nitskiy) Khmel’nitskiy 2 (Neteshin, Khmel’nitskiyn Rovno 1 (Kuznetsovsk, Rovno) Rovno 2 (Kuznetsovsk, Rovno) Rovno 3 (Kuznetsovsk, Rovno) Rovno 4 (Kuznetsovsk, Rovno) South Ukraine 1 (Konstantinovka, Nikolaev) South Ukraine 2 (Konstantinovka, Nikolaev) South Ukraine 3 (Konstantinovka, Nikolaev)
Taiwan, China
Ukraine
Taiwan Power Co.
Energoatom
Net MWe
Type
355
BWR
970
PWR
1030
BWR
365
PWR
365
PWR
604
BWR
604
BWR
948
BWR
948
BWR
890
PWR
890
PWR
950
PWR
950
PWR
402
PWR
416
PWR
950
PWR
950
PWR
950
PWR
950
PWR
950
PWR
Continued
349
350
CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
United Kingdom
Authority
British Energy plc— Nuclear Electric plc
British Energy plc— Scottish Nuclear Ltd.
Power station Zaporozhye 1 (Energodar, Zaporozhye) Zaporozhye 2 (Energodar, Zaporozhye) Zaporozhye 3 (Energodar, Zaporozhye) Zaporozhye 4 (Energodar, Zaporozhye) Zaporozhye 5 (Energodar, Zaporozhye) Zaporozhye 6 (Energodar, Zaporozhye) Dungeness B1 (Lydd, Kent) Dungeness B2 (Lydd, Kent) Sizewell B (Sizewell, Suffolk) Hinkley Point B1 (Hinkley Point, Somerset) Hinkley Point B2 (Hinkley Point, Somerset) Hartlepool 1 (Hartlepool, Cleveland) Hartlepool 2 (Hartlepool, Cleveland) Heysham A1 (Heysham, Lancashire) Heysham A2 (Heysham, Lancashire) Heysham B1 (Heysham, Lancashire) Heysham B2 (Heysham, Lancashire) Hunterston B1 (Ayrshire, Strathclyde) Hunterston B2 (Ayrshire, Strathclyde) Torness unit A (Dunbar, East Lothian) Torness unit B (Dunbar, East Lothian)
Net MWe
Type
950
PWR
950
PWR
950
PWR
950
PWR
950
PWR
950
PWR
555
AGR
555
AGR
1188
PWR
585
AGR
585
AGR
575
AGR
575
AGR
550
AGR
550
AGR
625
AGR
625
AGR
575
AGR
575
AGR
625
AGR
625
AGR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
United States
Authority
Power station
Nuclear Decommissioning Authority
Wylfa-1 Wylfa-2 Oldbury-1 Oldbury-2 Palo Verde 1 (Wintersburg, AZ) Palo Verde 2 (Wintersburg, AZ) Palo Verde 3 (Wintersburg, AZ) Calvert Cliffs 1 (Lusby, MD) Calvert Cliffs 1 (Lusby, MD) Pilgrim (Plymouth, MA) Brunswick 1 (Southport, NC) Brunswick 2 (Southport, NC) Robinson 2 (Hartsville, SC) Perry 1 (North Perry, OH)
Arizona Public Service Co.
Baltimore Gas & Electric Co. Boston Edison Co. Carolina Power & Light Co.
The Cleveland Electric Illuminating Co. Commonwealth Edison Co.
Consolidated Edison Co. Consumers Energy Co.
Braidwood 1 (Braidwood, IL) Braidwood 2 (Braidwood, IL) Byron 1 (Byron, IL) Byron 1 (Byron, IL) Dresden 2 (Morris, IL) Dresden 3 (Morris, IL) LaSalle County 1 (Seneca, IL) LaSalle County 2 (Seneca, IL) Quad Cities 1 (Cordova, IL) Quad Cities 2 (Cordova, IL) Indian Point 2 (Buchana, NY) Palisades (South Haven, MI)
Net MWe
Type
490 490 217 217 1270
GCR GCR GCR GCR PWR
1270
PWR
1270
PWR
825
PWR
825
PWR
670 767
BWR BWR
754
BWR
683
PWR
1205
BWR
1120
PWR
1120
PWR
1105 1105 912 794 1078
PWR PWR BWR BWR BWR
1078
BWR
789
BWR
789
BWR
975
PWR
781
PWR Continued
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
Detroit Edison Co.
Enrico Fermi 2 (Newport, MI) Catawba 1 (Clover, SC) Catawba 2 (Clover, SC) McGuire 1 (Cornelius, NC) McGuire 2 (Cornelius, NC) Oconee 1 (Seneca, SC) Oconee 2 (Seneca, SC) Oconee 3 (Seneca, SC) Beaver Valley 1 (Shippingport, PA) Beaver Valley 2 (Shippingport, PA) Arkansas Nuclear One-1 (Russellville, AR) Arkansas Nuclear One-2 (Russellville, AR) Grand Gulf (Port Gibson, MS) Waterford 3 (Taft, LA) River Bend (St. Francisville, LA) St. Lucie 1 (Hutchinson Island, FL) St. Lucie 2 (Hutchinson Island, FL) Turkey Point 3 (Florida City, FL) Turkey Point 4 (Florida City, FL) Crystal River 3 (Red Level, FL) Edwin I. Hatch 1 (Baxley, GA) Edwin I. Hatch 2 (Baxley, GA) Alvin W. Vogtle 1 (Waynesboro, GA) Alvin W. Vogtle 2 (Waynesboro, GA)
Duke Power Co.
Duquesne Light Co.
Entergy Operations, Inc.
Florida Power and Light Co.
Florida Nuclear Corp. Georgia Power Company
Net MWe
Type
1139
BWR
1129 1129 1129
PWR PWR PWR
1129
PWR
846 846 846 810
PWR PWR PWR PWR
833
PWR
836
PWR
8965
PWR
1173
BWR
1075 936
PWR BWR
839
PWR
839
PWR
666
PWR
666
PWR
868
PWR
810
BWR
820
BWR
1162
PWR
1162
PWR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Net MWe
Type
Oyster Creek (Forked River, NJ) Three Mile Island 1 (Londonderry Twp., PA) South Texas Project 1 (Palacois, TX) South Texas Project 2 (Palacois, TX) Clinton (Clinton, IL) Donald C. Cook 1 (Bridgman, MI) Donald C. Cook 2 (Bridgman, MI) Duane Arnold (Palo, IA) Cooper (Brownville, NE)
619
BWR
786
PWR
1250
PWR
1250
PWR
930 1020
BWR PWR
11,080
PWR
538 764
BWR BWR
James A. Fitzpatrick (Scriba, NY) Indian Point 3 (Buchanan, NY) Nine Mile Point 1 (Scriba, NY) Nine Mile Point 2 (Scriba, NY) Seabrook (Seabrook, NH) Millstone 2 (Waterford, CT) Millstone 3 (Waterford, CT) Monticello (Monticello, MN) Prairie Island 1 (Red Wing, MN) Prairie Island 1 (Red Wing, MN) Fort Calhoun (NE)
780
BWR
965
PWR
610
BWR
1080
BWR
1150
PWR
875
PWR
1149
PWR
536
BWR
503
PWR
500
PWR
478
PWR
Diablo Canyon 1 (Avila Beach, CA) Diablo Canyon 2 (Avila Beach, CA)
1130
PWR
1160
PWR
Authority
Power station
GPU Nuclear Corp.
Houston Lighting & Power Co.
Illinois Power Co. Indiana/Michigan Power Co. IES Utilities, Ine. Nebraska Public Power District New York Power Authority Niagara Mohawk Power Corp.
North Atlantic Energy Service Corp. Northeast Utilities
Northern States Power Co.
Omaha Public Power District Pacific Gas & Electric Co.
Continued
353
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Net MWe
Type
Susquehanna 1 (Berwick, PA) Susquehanna 2 (Berwick, PA) Limerick 1 (Pottstown, PA) Limerick 2 (Pottstown, PA) Peach Bottom 2 (Delta, PA) Peach Bottom 3 (Delta, PA) Hope Creek (Salem, NJ) Salem 1 (Salem, NJ) Salem 2 (Salem, NJ) R.E. Ginna (Ontario, NY)
860
PWR
860
PWR
1055
BWR
1055
BWR
1159
BWR
1035
BWR
1031 1106 1106 470
BWR PWR PWR PWR
Virgil C. Summer (Parr, SC) San Onofre 2 (San Clemente, CA) San Onofre 3 (San Clemente, CA) Joseph M. Farley 1 (Dothan, AL) Joseph M. Farley 2 (Dothan, AL)
885
PWR
1070
PWR
1080
PWR
860
PWR
860
PWR
Browns Ferry 1 (Decatur, AL) Browns Ferry 2 (Decatur, AL) Browns Ferry 3 (Decatur, AL) Sequoyah 1 (SoddyDaisy, TN) Sequoyah 2 (SoddyDaisy, TN) Watts Bar 1 (Spring City, TN)
1065
BWR
1118
BWR
1118
BWR
1148
PWR
1148
PWR
1177
PWR
Authority
Power station
Pennsylvania Power & Light Co.
PECO Energy Co.
Public Service Electric & Gas Co. Rochester Gas & Electric Corp. South Carolina Electric & Gas Co. Southern California Edison Co. and San Diego Gas & Electric Co. Southern Nuclear Operating Co.
Tennessee Valley Authority
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Net MWe
Type
Comanche Peak 1 (Glen Rose, TX) Comanche Peak 2 (Glen Rose, TX) Davis-Besse (Oak Harbor, OH) Callaway (Fulton, MO) Vermont Yankee (Vermont, VT) North Anna 1 (Mineral, VA) North Anna 2 (Mineral, VA) Surry 1 (Gravel Neck, VA) Surry 2 (Gravel Neck, VA) WNP-2 (Richland, WA)
1150
PWR
1150
PWR
877
PWR
1171 504
PWR BWR
893
PWR
897
PWR
801 801 1157
PWR PWR BWR
Point Beach 1 (Two Rivers, WI) Point Beach 2 (Two Rivers, WI) Kewaunee (Carlton, WI)
485
PWR
485
PWR
503
PWR
Wolf Creek (Burlington, KS)
1160
PWR
Authority
Power station
Texas Utilities Electric Co.
Toledo Edison Co. Union Electric Co. Vermont Yankee Nuclear Power Corp. Virginia Power
Washington Public Power Supply System Wisconsin Electric Power Co.
Wisconsin Public Service Corp. Wolf Creek Nuclear Operating Corp.
installations, new wind and solar technologies and biomass plantations. This share could increase, but only to a limited extent. The estimate made by the WEC for new renewable supplies in the medium term is that with adequate support, the share of new renewable energy supplies, currently only 2%, could reach 5–8% of increased world energy supply by 2020. The argument is related to the energy density which is so variable. For example: • • • • •
1 kg 1 kg 1 kg 1 kg 1 kg
firewood produces about 1 kWh of electricity, of coal produces about 3 kWh of electricity, of oil produces about 4 kWh of electricity, of natural uranium produces about 50,000 kWh of electricity and of plutonium produces about 6,000,000 kWh.
The low energy density of the renewable sources means that if you want significant amounts of energy (electricity) from them, you must “harvest” them over large
355
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CHAPTER 6 The nuclear fuel cycle
areas—and this is expensive. It has been calculated that to achieve the electricity generating capacity of a 1000-MWe power plant, an area of 50–60 km2 would be needed to install solar cells or windmills, or an area of 3000–5000 km2 to grow the necessary biomass. It will not be easy—or cheap—to acquire such large areas, particularly in densely populated areas where the energy will be most needed (Blix (1997)). Globally, electricity accounted for about 18% of the total final energy consumption in 2016. About 70% of the final energy consumption was in the form of fossil fuels. Bioenergy and waste accounted for 12%. As electricity consumption is expected to increase faster than total final energy consumption in the coming years, the share of electricity consumption is expected to rise. Electricity consumption will grow at a higher rate of about 2.5% per year up to 2030 and around 2% per year thereafter. The share of electricity in total final energy consumption will thus increase from 17.8% in 2016 to 21% by 2030 and to 26.6% by the middle of the century. Total electricity production grew by 2.6% in 2016 while the growth in nuclear electricity production was 2.1%. Among the various sources for electricity production, coal remained dominant despite the significant growth of natural gas-based generation. The share of electricity production from natural gas increased by 0.4% points to reach 22.8% of total electricity production. The contribution of hydropower and renewable energy sources continued to increase significantly, reaching 24.8% in 2016, while the share of nuclear electricity production remained at about 11% of the total electricity production (IAEA, 2017a). Reference Data Series No. 1 (RDS-1) is an annual publication of International Atomic Energy Commission. In 2017, its 37th edition was published containing estimates of energy, electricity and nuclear power trends up to 2050. Compared with the 2016 projections to 2030, the 2017 projections were reduced by 45 GW(e) by 2030 in both the high and the low case. These reductions reflect responses to the Fukushima Daiichi accident and other factors noted earlier. There are increasing uncertainties in these projections owing to the considerable number of reactors scheduled to be retired in some regions around 2030 and beyond (IAEA, 2017a,b). Nuclear power reactors in the world are an annual publication that presents the most recent data pertaining to reactor units in IAEA Member States. The 2018 Edition (IAEA, 2018b) provides a detailed comparison of various statistics up to and including 31.12.2017. The tables and figures contain the following information: (i) General statistics on nuclear reactors in IAEA Member States, (ii) technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned, (iii) performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in publication (IAEA, 2018b) are product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. This is also published as book and CD-ROM (IAEA, 2018c) which is the 49th edition of the IAEA’s series of annual reports on operating experience with nuclear power plants in Member States.
6.3 Nuclear safety
6.3 Nuclear safety Energy production as well as other human activities is always connected with risk taking. Radiation, and everything related to it, generates a fear not easily understood. This probably comes from the “invisibility” of the danger and relation to the bomb. Therefore the safety of nuclear power must be compared with the safety of alternative ways of generating electricity. The largest accidents in terms of casualties in the energy field are connected with the collapse of hydro dams. Some 2500 people perished, for example, in a single dam failure in Machu, India. There are also, as we know, severe accidents connected with the transport and storage of gas, the mining of coal and the shipping of oil. A gas pipeline explosion in Guadalajara in Mexico killed 200 people in 1992. Although one knows that the risk of incidents and accidents is not zero for any form of energy generation, including nuclear, one needs to be aware that most events are not very damaging. To help the nuclear power industry clarify the magnitude of events, the IAEA introduced the International Nuclear Event Scale (INES), which grades accidents from 1 to 7—much as seismologists grade earthquakes. It is hoped that this scale will help the media and public to realize that most incidents are of very minor significance and result in no threat to public health. It should also be remembered that most evolving technologies, whether boilers during the 19th century, airplanes in this century, or nuclear plants, entail some accidents from which lessons are learned. Both the Three Mile Island accident, from which only limited radioactivity escaped to the environment, and the Chernobyl disaster, have led to the introduction of new safety features in nuclear reactors, in plant operating procedures, and in regulations. The development of nuclear and radiation safety standards is a statutory responsibility of the International Atomic Energy Agency, IAEA, in Vienna, Austria. The IAEA Statute authorizes the Agency to establish standards of safety and to provide for the applications of these standards. Until now IAEA has developed and issued more than 200 standards of safety in the Agency’s Safety Series publications. They cover the fields of nuclear safety and radiation safety, including radioactive waste safety and radioactive material transport safety. The IAEA publications on this matter can be grouped into five categories: general safety, nuclear safety, radiation safety, waste safety and transport safety. Some of the most important publications are listed in the references under “safety series”. In recent years, legally binding international conventions have come to play a crucial role in improving nuclear, radiation and waste safety. The major international conventions related to safety that have been negotiated and adopted under the auspices of the IAEA are listed in Table 6.4 (from Flakus and Johnson, 1998). (1)
The Convention of Early Notification of a Nuclear Accident and the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency deal with aspects of emergency response and preparedness. Both of these Conventions—briefly referred to as the “Notification Convention”
357
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CHAPTER 6 The nuclear fuel cycle
Table 6.4 The global legal framework for nuclear, radiation and waste safety Entry into force Convention on the Physical Protection of Nuclear Material Convention on Early Notification of a Nuclear Accident
8 Feb. 1987
Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency Convention on Nuclear Safety
26 Feb. 1987
27 Oct. 1986
24 Oct. 1996
Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management
Vienna Convention on Civil Liability for Nuclear Damage
Protocol to Amend the Vienna Convention and Convention on Supplementary Compensation for Nuclear Damage
12 Nov. 1977
Development & Status In 1997, two States (Cuba and Lebanon) acceded to the Convention. As of May 1998, the Convention had 60 Parties In 1997, four States (Lebanon, Philippines, Myanmar and Singapore) agreed to be bound by the Convention. As of May 1998, the Convention had 80 Parties In 1997, three States (Lebanon, Philippines and Singapore) agreed to be bound by the Convention. As of May 1998, the Convention had 75 Parties In 1997, 10 States (Argentina, Austria, Belgium, Brazil, Germany, Greece, Luxembourg, Pakistan, Peru and Singapore) and in 1998 four States (Italy, Republic of Moldova, Portugal and Ukraine) agreed to be bound by the Convention. As of May 1998, the Convention had 46 Parties A Diplomatic Conference, held in Vienna in September 1997, adopted the Joint Convention which was opened for signature on 29 September 1997. As of 4 June 1998, the Convention had been signed by 33 States and ratified by three States (Canada, Hungary, Norway) In 1997, one State (Lebanon) ratified the Convention, and two States (Belarus, Israel) signed the Convention. The Convention had 29 Parties Both of these legal instruments were adopted on 12 September 1997 and opened for signature on 29 September 1997. As of 18 June 1998, the Protocol had been signed by 13 States (Argentina, Czech Republic, Hungary, Indonesia, Italy, Lebanon, Lithuania, Morocco, Peru, Philippines, Poland, Romania and Ukraine); and the Convention on Supplementary Compensation for Nuclear Damage had been signed by 13 States (Argentina, Australia, Czech Republic, Indonesia, Italy, Lebanon, Lithuania, Morocco, Peru, Philippines, Romania, Ukraine and United States)
After Flakus, F.-N. and Johnson, L.D., 1998. Binding agreements for nuclear safety: the global legal framework. IAEA Bulletin 40, 21
6.3 Nuclear safety
(2)
(3)
(4)
(5)
and the “Assistance Convention”—were adopted within a very short time span of only 5 months after the Chernobyl accident in 1986. The Notification Convention applies in the event of any accident involving facilities or activities of a State Party, or those under its jurisdiction or control, from which a release of radioactive material occurs or is likely to occur, and which has resulted or may result in an international transboundary release that could be of radiological safety significance for another State. A State Party involved in an accident covered by the Convention is obliged to immediately notify, directly or through the IAEA, those States which are or may be physically affected. To perform its functions under this Convention, the IAEA set up, at its headquarters in Vienna, an Emergency Response Center (ERC) for receiving, collating and rapidly transmitting relevant information. Close cooperation with the World Meteorological Organization (WMO) resulted in the use of WMO’s Global Telecommunication System (GTS) for rapid simultaneous transmission of voluminous meteorological and radiological data to national contact points. The Convention on Nuclear Safety was developed during the period 1992–94. It applies to land-based civil nuclear power plants and is the first international legal instrument that directly addresses the issue of safety of such plants. The Convention contains obligations for State Parties to take national measures with respect to safety matters—such as the legislative and regulatory framework, assessment and verification of safety, emergency preparedness and operation of nuclear power plants—and to report on the measures taken to implement each of the obligations under the Convention. The Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management was adopted at a Diplomatic Conference in Vienna in September 1997. The Convention is focused predominantly on specific activities rather than on substances. It applies with certain restrictions to: (i) the safety of spent fuel management, (ii) the safety of radioactive waste management, (iii) the safety of management of spent fuel or radioactive waste resulting from military or defence programmes if and when such materials are transferred permanently to and managed within exclusively civilian programmes. The Convention on the Physical Protection of Nuclear Material came into force in 1987. This Convention prescribes the levels at which nuclear material used for peaceful purposes is to be protected while in international nuclear transport, and requires each party to the Convention not to permit the export or import of such material unless it is satisfied that the nuclear material will be protected at those levels. At a Diplomatic Conference in September 1997, delegates from 80 States adopted the Protocol to Amend the 1963 Vienna Convention on Civil Liability for Nuclear Damage and also the Convention on Supplementary Compensation for Nuclear Damage. The Protocol sets the possible limit of the operator’s liability at an amount roughly equivalent to US $400 million and also contains an enhanced definition of nuclear damage which covers costs of reinstatement of any
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(6)
damaged environment and costs of preventive measures, extends the geographical scope of the Vienna Convention and extends the period during which claims may be made for loss of life and personal injury. In February 2015, the Vienna Declaration on Nuclear Safety was reached by contract parties of Convention on Nuclear Safety. It requests that a new NPPs are to be built with the objectives of preventing accidents and, should an accident occur, avoiding possible large releases of radionuclides to the environment. It also requests that safety of existing nuclear power plants is to be assessed and improved as appropriate.
IAEA holds a series of conferences on effective nuclear regulatory systems. These conferences bring together senior regulators in the areas of nuclear and radiation safety as well as nuclear security from around the world to discuss how to improve regulatory effectiveness to insure the protection of the public and the environment. The last conference in the series was held in Vienna, Austria 11–15 April 2018 (IAEA, 2017a,b). In order to insure the safe operation of nuclear power plants IAEA has published in 2018 following texts among others: commissioning guidelines for nuclear power plants, IAEA (2018d), maintenance optimization programme for nuclear power plants, IAEA (2018e), computer security of instrumentation and control systems at nuclear facilities, IAEA IAEA (2018f) and physical protection of nuclear material and nuclear facilities (implementation of INFCIRC/225/Revision 5), IAEA (2018g).
6.4 Releases of effluents Radioactive materials released to the environment are sources of exposure and potentially harmful. Such releases may be from different activities in the nuclear fuel cycle, mining operations or industrial users. Strict control measures must be employed to keep the resulting doses “as low as reasonably achievable”. This implies the implementation of protective and control measures and includes the setting of limits for radiation exposure. A limit is a value that must not be exceeded and the primary dose limits for individuals are set by the ICRP. These limits are related to individuals irrespective of the source. If an individual is likely to be exposed to other sources of radiation, sourcerelated limits must be set by a regulatory authority. These limits must be lower than the dose limit and are called the source upper bound. Authorized limits are limits specified by the regulating authority for a specific practice or source. In setting limits the authority must consider the requirements of radiation protection and individual dose limitation. The authorized limits will not exceed the upper bound. For practical reasons limits for releases of radioactive effluents to the environment are expressed as limits of releases over a specified period.
6.4 Releases of effluents
It is important to set reference levels for all activities. A reference level is not a limit but indicates a course of action like recording data, investigation or intervention. These levels are determined by radiation protection factors and the extent of the measures taken must be described in the operating procedures. In the case of new practices where reassessment may result in lower or higher release rates being acceptable, setting authorized limits can be difficult. There is often justification for a specific source or practice to be exempted from normal regulations. The regulating authority may exempt such sources or practices on the basis that the individual and collective doses are so low that they may be ignored. The individual dose limit is the starting point for calculating the upper bound. The dose upper bound will be less by the dose contributed by global and regional sources of exposure. The regulating authority may reserve a margin for future development of the activity or practice. This margin is set by specifying that a fraction, F, of the primary dose limit must not be exceeded. The maximum annual dose limit to the critical group is limited by Hlocal + Hregional + Hglobal F Hlimit
(6.1)
where Hlimit is the primary dose limit and the suffixes refer to the components of the total dose to the critical group. The source specific dose upper bound (HUB) for all the controlled sources of exposure is given by HUB ¼ F Hlimit Hregional Hglobal
(6.2)
The upper bound for annual release can then be derived from the dose upper bound by using the overall transfer factors (fj0 kl) where j represents population group, k represents release mode and l represents the radionuclide. If the dose commitment to the critical group j0 per unit release of a radionuclide is given by fj0 kl then the release upper bound, Rkl, is given by Rkl ¼
HUB fj0 kl
(6.3)
provided that no other radionuclides are released. In practice the situation will be more complex because more than one nuclide may be released and different modes of release will be developed. The total dose contribution to each population group due to a release Rkl is given by Jjkl ¼ fjkl Rkl :
(6.4)
If different release modes (k) are developed, the release upper bounds (Rkl) for each release mode is given by X
fj0 kl Rkl HUB
(6.5)
k
This is true only for the release of one radionuclide. If a mixture of nuclides is released that contributes to the exposure of group j0 , the release upper bound, Rkl, must satisfy the condition:
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XX k
fj0 kl Rkl HUB
(6.6)
l
The condition defines a set of values that constitute the release upper bound. For routine releases of radioactivity two main control options are considered: 1. storage of effluents to allow short-lived radionuclides to decay before release and 2. treatment of effluents to remove radionuclides before release. Within these categories a number of options may be available. The various possibilities must be identified and investigated. Considerations like operating and maintenance cost, the implications for the waste management programme as well as the individual and collective dose for the workers and the public must be taken into account. The first step in optimizing is to ensure that the releases anticipated with the control options meet the requirements of the source upperbound. Any control option that does not meet this requirement is not considered. The final element in implementing a system of dose limitation is to optimize radiological protection by selecting the control option for which radiation doses are “as low as reasonably achievable”. For the monitoring of effluent releases the samples collected in the vicinity of nuclear installations must be representative of land and water utilization as well as meteorological factors. Samples must be analysed for those nuclides which contribute most to public exposure. Air sampling is of special interest. Usually fixed monitoring instruments are used for continuous routine monitoring in the vicinity of the installation. If a limit has been exceeded the cause must be traced and corrective measures must be taken immediately. Two types of sampling monitors are in general use: air samples are used to assess the airborne contamination levels at selected points. In the case of particulate materials a volume of air is drawn through a filter paper on which the particulates are deposited. An alarm may be set on increase of activity. For gaseous materials carbon cartridges are used to trap the contaminating materials. Special devices are used for trapping iodine. “Stack” monitors for gaseous effluents give a rough estimate of the radioactivity of the effluent from a stack. The radioactive content of the samples can be assessed by using standard counting equipment. Sample measuring instruments are operated in contamination-free laboratories. For the monitoring of the released liquid effluents the following methods are used. Samples of effluents are collected by simple dipping devices and analysed before release. In the case of monitoring streams in the neighbourhood of installations, automatic samplers collecting samples over a 24-h period are used. Samples are analysed and records must be kept of results. The water effluent metre monitors water or coolants and may be connected to a rate metre, recorder or alarm system. On-site and off-site environmental monitoring at and near nuclear power plants, nuclear reactors and other fuel cycle activities are shown in Tables 6.5 and 6.6.
6.4 Releases of effluents
Table 6.5 On-site monitoring at nuclear power plants, nuclear reactors and uranium mill and/or fuel cycle facilities Sample type
Collection frequency
Analysis frequency
Airborne particulates Liquid effluents Drinking water
Continuous
Continuous readout
Continuous
Continuous readout
Semi-continuous (samples taken 3–6 h) Monthly
Quarterly on composites
Surface water Noble gases Groundwater
Continuous Quarterly
Quarterly on composites Continuous readout γ-Spectrometry on each batch sample; annual composite on other nuclides detected
Table 6.6 Off-site environmental monitoring near nuclear power plants, nuclear reactors and/or uranium mill and/or fuel cycle facilities Sample type
Collection frequency
Analysis frequency
Drinking water
Semicontinuous composite
γ-Spectrometry on each batch sample; annual composite on other nuclides detected
Milk
Weekly or biweekly at farms; monthly at dairy, γspectrometry weekly At harvest
Food crops Fish Shellfish Sediments
During fishing season or semiannually Semiannually Semiannually
γ-Spectrometry γ-Spectrometry γ-Spectrometry γ-Spectrometry
By way of illustration, we mention the case of the Sellafield reprocessing plant in Cumbria, UK, as discussed by Jones et al. (1995). For the last decade, the existence of a higher than average rate of childhood leukaemia in young people from the village of Seascale in Cumbria has led to speculation that radioactive discharges from the reprocessing plant at Sellafield may be a causative factor, even though the calculated doses are too small for the leukaemia risk observed in epidemiological studies (Stather et al., 1984, 1986). These estimates of historical doses from discharge from the plant have relied on calculations based on recorded discharges and conventional environmental models. This has left open the question of whether the recorded
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discharges, particularly in the earlier years of plant operation, might have been seriously underestimated. A major reassessment of historical discharges and doses has been carried out, prompted in large part by civil litigation instigated by a number of local families against British Nuclear Fuels plc, the operators of the Sellafield plant. The reassessment involved the development of the Sellafield Environmental Assessment Model (SEAM), which was used both to calculate doses and to build confidence in the discharge chronology from recorded measurements of environmental concentrations and current assessments of environmental inventories. The SEAM model has put together established models of atmospheric dispersion and deposition, terrestrial food chains, marine dispersion and concentration in marine biota and the sea-to-land transfer of radionuclides. Environmental measurements from a wide variety of sources have been compared against values calculated from the discharge chronology and the SEAM model in order both to validate the model and to build confidence in the discharge chronology. The established chronology for liquid discharges to the Irish Sea was confirmed by validation against historic environmental monitoring data and dated sediment cores (Kershaw et al., 1990). It was possible to establish good agreement, thus building confidence in the discharge chronology. The review of atmospheric discharges has indicated that earlier figures for emissions, particularly for particulates, in the earlier years of plant operation were significantly underestimated. Further, improved estimates were made of unmonitored emissions of plutonium from the site, which were known to have occurred in the 1950s and 1960s by utilizing the results of cumulative deposition measurements in soil cores. Despite the higher assessed discharges to atmosphere, calculated doses to members of the public in Seascale remain low and are insufficient to account for any excess of leukaemia. Furthermore, where specific measurements of radionuclide body contents of local residents are available (Stather et al., 1988) the model significantly overestimates body content.
6.5 Management of radioactive wastes Like all industries, the generation of electricity produces waste. Whatever fuel is used, the waste produced in generating electricity must be managed in ways that safeguard human health and minimize the impact on the environment. Concern about nuclear power is usually focused on the highly toxic and radioactive spent fuel and nuclear waste. What is characteristic of these, however, in addition to their toxicity and radioactivity, is that they are limited in volume, which facilitates waste disposal. This contrasts sharply with the waste disposal problem for fossil-fuelled plants. More specifically, a 1000-MWe coal plant with optimal pollution abatement equipment will annually emit into the atmosphere 900 tonnes of SO2, 4500 tonnes of NOx, 1300 tonnes of particulates and 6.5 million tonnes of CO2. Depending on the quality of the coal, up to 1 million tonnes of ashes containing hundreds of tonnes
6.5 Management of radioactive wastes
of toxic heavy metals (arsenic, cadmium, lead and mercury) will have to be disposed of. By contrast, a nuclear plant of 1000-MWe capacity produces annually some 35 tonnes of highly radioactive spent fuel. If the spent fuel is reprocessed, the volume of highly radioactive waste will be about 3 m3. The entire nuclear chain supporting this 1000-MWe plant, from mining through operation, will generate, in addition, some 200 m3 of intermediate-level waste (ILW) and some 500 m3 of low-level waste (LLW) of year. Most countries using nuclear electrical generation have programmes for safe disposal of the wastes. Technical alternatives for disposal of spent fuel and high-level wastes (HLWs) have been assessed by several countries and international organizations. Scientific consensus exists that geologic disposal using a system of natural and engineered barriers is the preferred method to be used. Unlike chemically hazardous industrial wastes, the much smaller volumes of spent fuel and HLW make containment and isolation a feasible disposal option, and their radiological hazard will decrease with time. The volume of highly radioactive waste (HLW) produced by the nuclear industry is small. The IAEA estimates that 370,000 metric tonnes of heavy metal (MTHM) in the form of used fuel have been discharged since the first nuclear power plants commenced operation. Of this, the agency estimates that 120,000 MTHM has been reprocessed, with the balance, 250,000 MTHM, in storage. The amounts of ILW, LLW and very low-level waste (VLLW) produced are greater in volume, but are much less radioactive. Given its lower inherent radioactivity, the majority of waste produced by nuclear power production and classified as LLW or VLLW have already been placed in disposal. The IAEA estimates that over 80% of all LLW and VLLW produced to date are in disposal. For ILW, the agency estimates that about 20% are in disposal, with the balance in storage. Numbers include both solid and liquid radioactive waste (Table 6.7). Generic studies of geologic disposal conducted by the Swedish KBS, the Commission of European Communities (CEC) and others have concluded that geologic disposal systems can achieve an acceptable level of safety to protect future generations from the radiological hazards associated with these wastes. In 1991 IAEA established the Radioactive Waste Safety Standards (RADWASS) Programme to develop a special series of safety documents specifically directed at Table 6.7 Nuclear waste inventory, IAEA estimates IAEA (2018a)
VLLW LLW ILW HLW
Solid radioactive waste in storage (m3)
Solid radioactive waste in disposal (m3)
2,356,000 56,811,000 6.713,000 2.808,000
7,906,000 60,035,000 8.735,000 68.000
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radioactive waste management. The purpose of the RADWASS Programme is to document existing international consensus in the approaches and methodologies for safe waste management and disposal; to create a mechanism for establishing consensus where it does not exist; and to provide Member States with a comprehensive series of internationally agreed documents to complement national standards and criteria. RADWASS has been organized in a hierarchical structure of four levels of safety documents. The top-level publication is a document of safety fundamentals which provides the basic safety objectives and fundamental principles to be followed in national waste management programmes. The lower levels include safety standards, safety guides and safety practice documents. The series has been structured in a logical and clear manner to reflect the systems approach to waste management. Radioactive waste is any material that contains, or is contaminated with, radionuclides at concentrations of radionuclides greater than the “exempted quantities” established by the regulatory body and for which no future use is foreseen. This is after a definition by IAEA. Five main activities produce such waste: • • • • •
uranium and thorium mining and milling; nuclear fuel cycle operations such as uranium conversion and enrichment, fuel fabrication and spent fuel reprocessing; operations of nuclear power stations; decontamination and decommissioning of nuclear facilities; institutional uses of isotopes.
The waste resulting from the above activities comes in various forms (i.e. gaseous, liquid, or solid). These wastes have different characteristics. For safety and technical reasons, the various forms of wastes are usually categorized by their levels of radioactivity, heat content and potential hazard. With regard to disposal the wastes are categorized as follows. LLWs contain a negligible amount of long-lived radionuclides. Produced by peaceful nuclear activities in industry, medicine, research, and by nuclear power operations, such wastes may include items such as packaged gloves, rags, glass, small tools, paper and filters which have been contaminated by radioactive material. Disposal in near-surface structures or shallow burial is practised widely. ILWs contain lower levels of radioactivity and heat content than HLWs, but they still must be shielded during handling and transport. Such wastes may include resins from reactor operations or solidified chemical sludges, as well as pieces of equipment or metal fragments. Commercial engineering processes are being used to treat and immobilize these wastes. Disposal options are similar to those for LLWs. HLWs arise from the reprocessing of spent fuel from nuclear power reactors to recover uranium and plutonium. These wastes contain transuranic elements, and fission products that are highly radioactive, heat-generating and long-lived. Liquid HLW is usually immobilized as a solid glass matrix and stored in interim storage facilities prior to final disposal and isolation in deep, stable, geologic formations,
6.5 Management of radioactive wastes
as currently planned by many national programmes. Spent nuclear fuel that is not reprocessed is also considered HLW. α-bearing wastes (also called transuranic, plutonium-contaminated material or αwastes) include wastes that are contaminated with enough long-lived, α-emitting nuclides to make near-surface disposal unacceptable. They arise principally from spent fuel reprocessing and mixed-oxide fuel fabrication. The wastes may be disposed of in a similar manner to HLW.
6.5.1 Storage and disposal Management and disposal of nuclear waste depends mainly on its type. For example, LLW and ILW are often treated (volume reduction) and/or conditioned (waste immobilization) prior to disposal. This area of LLW and ILW waste management, having been established and proven over past years, is considered to be quite mature in terms of technology development. As a result, several effective, safe and feasible treatment and conditioning options exist for these types of wastes. They include storage and decay, compaction and super compaction, incineration, chemical precipitation, evaporation, filtration and ion exchange; these may be followed by immobilization in materials like concrete, bitumen or polymers (Chan, 1992). The most common disposal methods for ILW and LLW involve disposal in shallow earthen or concrete lined trenches or in structures on the ground (commonly referred to as engineered surface facilities). Safe near-surface disposal of LLW has been practised in a number of countries for almost 30 years. The rationale behind near-surface disposal is that the isolation period for this type of waste is relatively limited (up to 300 years) and, therefore, the institutional or administrative control of the disposal site can be assured. HLW management and disposal is quite different. After its useful life, spent nuclear fuel is removed from the reactor. Once removed, it is usually placed into temporary on-site storage before it is either: •
•
placed in interim away-reactor storage (5–100 years), conditioned after a sufficient decay period and stored before its eventual final disposal in a geologic repository or reprocessed after additional away-from-reactor storage. The resulting liquid HLW, containing mostly fission products and a small proportion of the actinides, is then immobilized in a stable matrix (i.e. borosilicate glass), and would then be disposed of in a geologic repository.
Regardless of which option is chosen, there is broad scientific agreement that deep geologic disposal using a system of engineered and natural barriers to isolate these wastes is the preferred method for their disposal (Chan, 1992). According to the report by Oi (1998) at the end of 1997, more than 130,000 tonnes of spent fuel from power reactors were estimated to be stored worldwide containing about 1000 tonnes of plutonium. Another 170 tonnes of separated plutonium were in storage from civilian reprocessing operations, and about
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100 tonnes of excess plutonium from dismantled warheads no longer required for defence purposes were scheduled to be released from the military sector of Russia and United States. Plutonium represents a dual challenge because it is a valuable energy source and a matter of global concern because of its potential health hazards and possible use for the production of nuclear weapons.
6.5.2 Processing of used nuclear fuel Over the last 50 years the principal reason for processing of used nuclear fuel has been to recover unused plutonium and close the fuel cycle gaining some 25%– 30% more energy from the original uranium in the process. The world commercial (civil) reprocessing capacity is shown in Table 6.8. Spent fuel from light-water reactors contains about 1% of plutonium. According to Oi (1998) the IAEA estimates that in 1997 about 10,500 tonnes of spent fuel were discharged from nuclear power reactors worldwide; this amount contains about 75 tonnes of plutonium. It is estimated that the annual production figure will remain more or less the same until 2010. The cumulative amount of plutonium in spent fuel from nuclear power reactors worldwide is predicted to increase to about 1700 tonnes by 2010. It is estimated that about 3000 tonnes of spent fuel discharged from power reactors were reprocessed in 1997, which corresponds to about 30% of the total. About 24 tonnes of plutonium were separated in reprocessing plants and 9 tonnes of plutonium were used mainly as mixed uranium–plutonium oxide fuel (MOX) in lightwater reactors. The imbalance between the separation and use of plutonium had resulted in an accumulated inventory of separated civil plutonium of about 170 tonnes at the end of 1997. IAEA projections of plutonium inventories show that the rate of separation of civil plutonium and its rate of use will fall into balance in a few years. This is Table 6.8 Reprocessing facilities Country
Facility
Fuel type
2017 (tonnes/year)
France UK
La Hague Sellafield (THORP) Sellafield Ozersk (Mayak) Rokkasho Tokai Four plants
LWR LWR Magnox LWR LWR MOX PHWR
1700 600 1500 400 800a 40 330
Russia Japan India a
To start operation in 2021. After WNA, 2018. Processing of Used Nuclear Fuel. World Nuclear Association. http://world-nuclear. org/information-library/nuclear-fuel-cycle/fuel-recycling/processing-of-used-nuclear-fuel; OECD/NEA, 2017. Nuclear Energy Data 2017. Organization for Economic Co-operation and Development (OECD), OECD 2017. Nuclear Energy Agency (EA), NEA No. 7365.
Plutonium inventory (t)
6.5 Management of radioactive wastes
150
100
50
0 1995 1997 1999 2001 2003 2005 2007 2009 2011 2013 2015
Year FIG. 6.3 Projected worldwide civil inventories of separated plutonium. Solid line: minimum Pu inventory, no Pu market; dashed line: end of year Pu inventory, no Pu market; dotted line: end of year Pu inventory with Pu market; dashed-dotted line: minimum Pu inventory with Pu market.
due to an enhanced capacity of MOX fuel production which will amount to 360 tonnes of heavy metal per year in 2000. Beyond this period, the inventory is expected to decrease modestly and level off at around 130 tonnes. Despite the efforts to reduce the current inventories of separated civil plutonium, the worldwide inventories still remain at a substantial level, as shown in Fig. 6.3. In addition to the amounts of civil plutonium, plutonium is being released from dismantled warheads. Under the START-I and -II Treaties, many thousands of US and Russian nuclear warheads are slated to be retired within the next decade. As a result, at least 50 tonnes of plutonium from each side are expected to be removed from military programmes. Oi (1998) points out the problem, which is what to do with plutonium either in a separated form or contained in spent fuel. A number of issues arise because of plutonium’s potential use as an energy source and for the production of nuclear weapons. Presently, plutonium is used in light-water reactors as MOX fuel and also in small amounts for the development of fast-breeder reactors. Currently 22 power reactors in five countries (France, Germany, Switzerland, Belgium and Japan) are loaded with MOX fuel and this number is expected to rise to between 36 and 48 by 2000. The use of MOX reduces the inventory of separated plutonium and is regarded as an interim measure before plutonium’s possible full-scale use in fast reactors later in the next century. It is known that multiple recycling in light-water reactors degrades plutonium, which in turn limits the number of times it can be recycled to two or three. Such degraded plutonium can, however, be used as fuel in fast reactors. Without such reactors, spent MOX fuels will still end up in a final depository or in storage facilities (Oi, 1998). Recently much attention has been given to the accelerator-driven systems, burning in inert matrices, and the use of thorium to burn plutonium. The concept of a
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closed nuclear fuel cycle was traditionally considered as transmutation (burning) of only plutonium and recycled uranium, with minor actinides (neptunium, americium, curium) destined for final geological disposal. But as time goes on, a new understanding is emerging: reduction of the quantity of actinides would ease requirements for final repositories and make them relatively less expensive. Neutron transmutation of long-lived radioactive minor actinides by the fission process—which entails producing energy and simultaneously turning them into shorter-lived nuclides—is being intensely analysed by the technical community. Also being proposed is the neutron transmutation of selected long-lived fission products. Several possibilities for the transmutation of long-lived nuclides by nuclear reactions have been suggested. In the beginning, the best choice appeared to be the use of nuclear reactors. However, recently there has been renewed interest in what are called accelerator-driven systems (ADS), a technology that seems to show good promise. ADS would produce large amounts of electrical energy while simultaneously destroying the plutonium. This appears to offer a better solution to the plutonium problem than multimillennium storage. The use of accelerators for nuclear energy applications is not a new idea and was proposed as early as the late 1940s by E. Lawrence, inventor of the cyclotron. In the 1950s he promoted the development of a Materials Test Accelerator at Livermore to produce intense neutron fluxes for plutonium production. The Canadian Chalk-River Laboratory began intensive studies of accelerator-based systems to breed nuclear fuel for heavy-water reactors. Scientists at Brookhaven National Laboratory also actively promoted accelerator-based options in the late 1970s and early 1980s. For the last 5 years, scientists at Low Alamos National Laboratory have been reevaluating the accelerator-based technology in the light of new advances in technology and the world energy perspective (Bowman et al., 1992). When 1.6 GeV protons strike a large radius target consisting of heavy nuclei such as lead, approximately 55 neutrons are generated per proton. The energy deposition for this process is about 30 MeV of proton energy per neutron compared with about 200 MeV of fission energy deposited per useful neutron from a sustained chain reaction in fissile material such as 235U. The heat per unit volume which must be handled for a given neutron production rate is therefore considerably smaller for the spallation source than for the reactor. The reactor has the further disadvantage that in nearly all designs the fuel is fixed in position with coolant flowing past. Heat deposition in the fuel is therefore limited by the conductivity of the fuel and by the heat capacity and conductivity of the coolant. For the accelerator-driven neutron source the target is itself a flowing liquid heavy metal. The heat load on the target is therefore limited only by the thermal properties of the liquid metal and by the rate at which it flows. This accelerator target is therefore capable of generating a much higher density (and therefore flux) of neutrons than a reactor with fixed fuel because of the much greater power density capability of the flowing target and the much lower energy deposition in the target per neutron produced. The neutron–production–transmutation system considered by Bowman et al., 1992 consists of an accelerator for the proton beam, a flowing heavy metal proton
6.5 Management of radioactive wastes
target for neutron production, and a surrounding blanket containing primarily heavy water (D2O) for moderating the neutrons into the thermal range. The neutron flux may be further enhanced by neutrons from actinide fission in the blanket. The actinide material is transported through the blanket as a molten salt mixed with the carrier salt LiF-BeF2. Heat from the fission process is deposited in the molten salt and carried away by the salt at an exit temperature of 720°C which makes possible electric power generation at a high thermal-to-electric efficiency. A continuous flow system is essential because of the high burn-up rates of fissile material. For example, the lifetime of 239Pu in a thermal flux of 1016 n/cm2-s is only about 1 day so that use of fuel assemblies along the lines of standard practice for reactors is impractical. In addition there must be chemistry facilities for removing stable or short-lived fission products and returning radioactive waste to the blanket. The accelerator, target/blanket, electric power extraction and chemistry facilities are shown schematically in Fig. 6.4. The expected efficiency for conversion of thermal to electric power is 44% and the busbar efficiency of the accelerator is 45%. A fraction of the electrical power is fed Accelerator Heavy-water moderator
Electric power Energy extraction Waste feed
Neutron flux
Advanced chemical separations Stable and short-lived products
Molten salt loop
Liquid lead target
FIG. 6.4 The proton beam strikes the lead target generating neutrons which are moderated in the surrounding heavy water blanket. Molten salt carrying fissile material for heat generation and electric power production circulates in the heavy water blanket through double-walled pipes. Some of this power drives the accelerator. Nuclear waste including that produced in the molten salt is also circulated through the blanket in a separate loop and transmuted to stable and short-lived nuclides which are extracted and stored. After Bowman, C.D., et al., 1992. Nuclear energy generation and waste transmutation using an accelerator driven intense thermal neutron source. Los Alamos Report LAUR-91-260 and Nuclear Instruments and Methods A320, 336.
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Fission
Spallation target Protons 1 GeV
Spallation residues
Window
Neutrons Transmutation FIG. 6.5 Schematic representation of a hybrid plant. An intense beam of proton from an accelerator hits a massive target (Pb, Pb–Bi, W, Hg, …) and produces fast neutrons. In a surrounding blanket, the neutrons are used to drive a subcritical medium.
back to power the accelerator which operates at an energy of 1.6 GeV and produces neutrons in a Pb target. The beam power deposited in the Pb and the thermal power deposited in the D2O blanket are not converted to electrical power although energy recovery loops on these systems could boost the overall system efficiency. The material to be transmuted can be fed either into the molten salt carrier as molten salt or into the heavy water as dissolved salt depending on the application (Bowman et al., 1992). Recently there has been increased interest in the idea of accelerator-driven reactors, see, for example, Takizuka (1994), Van Tuyle et al. (1993), Carminati et al. (1993), Rubbia et al. (1995). In such systems (Fig. 6.5), spallation reactions induced by a high-intensity beam (10–250 mA) of GeV protons on a heavy target produce an intense neutron flux. These neutrons, after being more or less moderated, are used to drive a subcritical blanket. The extra neutrons provided by the accelerator allow the maintenance of the chain reaction while burning the long-lived nuclear waste. The plant generates electricity, part of which is used to supply the accelerator. Besides the more favourable neutron economy, additional advantages of accelerator-driven systems are safety and versatility. Obviously the operation of the blanket in a subcritical state is a major safety advantage. It could for instance allow the introduction of a large amount of Pu or minor actinides which is difficult in classical reactors because of control problems due to the smaller fraction of delayed neutrons. Accelerator-driven systems are also more flexible than reactors since the intensity of the accelerator can be adjusted to counteract the growth of
6.5 Management of radioactive wastes
poisonous isotopes or when adding elements to be transmuted. Their main drawbacks are their complexity and the technological progress they imply for the accelerator, the target-blanket and the interface between them (Boudard et al., 1998). Recently, it has been proposed to construct a demonstrator facility of significant power of the order of 100 MW (thermal) on a 10-year time schedule as a regional European facility.
6.5.3 Smart use of nuclear waste Use of nuclear energy as based on current technology has some drawbacks expressed as worries about reactor accidents, the potential for diversions of nuclear fuel into nuclear weapons, the management of long-lived radioactive waste and the depletion of economical uranium reserves. The proposed new fuel cycle would combine two innovations, pyrometallurgical processing and fast neutron reactors for burning that fuel. With this approach radioactivity of nuclear waste would drop to safe levels in a few hundred years (Hannum et al., 2005). In about 3 years of use nuclear fuel in reactors needs to be changed because of radiation-related degradation and the depletion of 235U. Not to mention that at this time 239Pu in the reactor, generated from 238U capture of thermal neutrons, contributes significantly to the reactor power. Neutrons are thermalised by reactor cooling water. At that time reactor has generated lots of “spent fuel” consisting of three classes of materials. It is generally accepted to divide spent fuel material into three components. First component is made of fission products (5% of used fuel)—this is the true waste—the content are the products of the fission process in which the energy was liberated, so by analogy this could be called “the ash”. This mix of elements is highly radioactive for some years after it is generated. After some time (10 years) its activity is dominated by two isotopes: 137Cs and 90Sr. This component is not highly radioactive; it could be separated from the rest of spent fuel and stored in dedicated facility for future use. Second component is made of unfissioned uranium, mainly 238U (94%), and this material resembles natural uranium. This component is only mildly radioactive and can be separated from the rest of spent fuel and stored in dedicated facility for future use. The third component is the one which contains all the problems; it contains all of transuranic elements (elements heavier than uranium). This component of spent fuel is made mainly of the mix of plutonium isotopes and americium. Although this component represents only 1% of spent fuel it generates almost all the problems we have with spent fuel today. The half-lives of these isotopes range up to tens of thousands years. An idealistic approach to the solution of this problem was to use produced plutonium in the fast-neutron reactors, so-called breeders because of producing more plutonium than consuming. However, the same plutonium could be used to make bombs and lead to uncontrolled proliferation of nuclear weapons. There is, however, an alternative recycling process which does not involve pure plutonium at any stage. This fast-neutron reactor technology minimizes the risk that spent fuel from energy production would be used for bomb manufacture. In addition,
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CHAPTER 6 The nuclear fuel cycle
Table 6.9 Transmutation probabilities (%) Isotope
Thermal neutrons
Fast neutrons
137
3 7 63 1 75 1 1 75 1 1 78
27 70 85 55 87 53 21 94 23 10 94
Np Pu 239 Pu 240 241 242 241 242m 243 242 243 238
After WNA, 2018. Processing of Used Nuclear Fuel. World Nuclear Association. http://world-nuclear.org/information-library/nuclear-fuel-cycle/fuel-recycling/ processing-of-used-nuclear-fuel.
fast reactors can extract more energy from fuel than thermal reactors because of 238U, 239 Pu and other transuranic isotopes having greater fission cross section for fast than thermal neutrons. Here, we need to introduce a concept of pyroprocessing process in which a mix of transuranic elements instead of pure plutonium is extracted from the used fuel. The procedure is based on electroplating; it uses metallic electrode to collect dissolved spent metallic fuel in a chemical bath. In the next step the accumulated material is scraped from the electrode, melted and casted into ingots to be refabricated as a fuel rod. The pyroprocess collects almost all of transuranic elements from the used fuel together with uranium and fission products. This combination is unsuitable for weapon production or the use in thermal reactors. This will be implemented in the fourth-generation fast neutron reactors in the late 2020s. Additionally, fast reactors make only 1% of spent fuel material compared to thermal reactors. As a result, the waste management would be greatly simplified. Transmutation of one radionuclide into another can be achieved by neutron bombardment in a nuclear reactor or accelerator. As known, the objective of transmutation is to change actinides into fission products and long-lived fission products into significantly shorter-lived nuclides. The goal is to have wastes which become radiologically innocuous in only a few hundred years. In comparison fast neutrons are more efficient than thermal neutrons, see Table 6.9 after WNA (2018).
6.6 Research reactors Nuclear reactors have supported research in many different fields and have contributed to discoveries in many scientific disciplines. Altogether there are about 180 research reactors in operation in the world (IAEA, 1996). Recently, IAEA has
6.6 Research reactors
published its Nuclear Energy Series no. NP-T-5.8 volume containing a CD-ROM containing 30 research reactor profiles, which provide technical descriptions and specific features for utilization. In addition, IAEA keeps Research Reactor Data Base, IAEA-RRDB (2018), the Research Reactor Database (RRDB) provides extensive information on research reactors all over the world. It is a combination of two individual databases: one that provides profiles of research reactors (both operational and shut down), and the other that compiles fuel information. The research reactor section of the database contains technical specifications and utilization information to assist potential users of research reactors. The fuel cycle section of the database is not publically accessible, but general reports on fuel information are compiled from the database. Both databases are populated and updated by officially nominated facility data providers (FDPs) who are typically located at the reactor. The present situation is shown in Table 6.10. Research reactors have a very wide variety of uses, including neutron scattering (in which beams of thermal neutrons are scattered by the atoms in a sample, revealing its structure, magnetic state and atomic binding energies); neutron activation analysis; radiography; irradiation testing of materials and production of radioisotopes for medical, research and industrial use. These capabilities are applied by researchers in many fields, ranging from archaeology to materials science and from fusion research to environmental science. Few generalizations can be made about the applications for research reactors or about their users. Research reactors themselves tend to have a very different set of safety-related parameters from power reactors. Some are helpful differences like simplicity, relatively low power and low-temperature coolant. Other differences, especially the need for a high-power density core, pose challenges not faced in a power reactor. These challenges can be met through thoughtful design solutions. A research reactor’s power is usually in the range 0–100 MW thermal. The fission product inventory and the stored energy in research reactors are smaller than in power reactors. However, some of the research reactors have large power density
Table 6.10 Status of 840 research reactors around the globe Status Planned Under construction Operational Temporary shutdown Extended shutdown Permanent shutdown Under decommissioning Decommissioned
Developed countries
Developing countries
All countries
2 4 140 8 5 43 62
11 6 85 5 9 14 4
13 10 225 13 14 57 66
413
29
442
375
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CHAPTER 6 The nuclear fuel cycle
Control rods
Water
Reactor Core
FIG. 6.6 Basic H2O-cooled swimming-pool reactor. After West, C.D., 1997. Research reactors: an overview. Nuclear News, October 1997, p. 50.
(>5000 kW thermal per kg of fuel). The typical research reactor is of the swimming pool type, as shown in Fig. 6.6. In many of these reactors the core is made up of so-called materials testing reactor-type fuel elements which are aluminium-clad, curved plates of fuel arranged in long rectangular boxes arranged between grid plates to form the core. Several positions in the grid are not occupied by fuel elements, but by control rods, beryllium reflectors or experimental capsules. Cooling may be by natural convection of the pool water, although this is augmented, for operation at higher power, by pumping pool water through the core. More powerful research reactors, of which the international Institut LaueLangevin (ILL) facility at Grenoble, France, and the high flux isotope reactor (HFIR) at Oak Ridge, Tenn., are well-known examples, have tanks that are full pressure vessels—for example, the coolant inlet pressure at HFIR is nominally 470 psi, and at ILL it is 200 psi. Again, aluminium-clad fuel plates are used, the fuel meat being a layer, about 50 mils thick, of U3O8 particles mixed with powdered aluminium for enhanced thermal conductivity, the layer being clad with aluminium plates about 10 mL thick. In these two reactors, the fuel elements are annular, with curved (involute) plates fitting into axial grooves down two concentric cylinders.
6.7 Advanced nuclear power plants
6.7 Advanced nuclear power plants New generations of nuclear power plants have been or are being developed, building on this background of success and applying lessons learned from the experience of operating plants. Advanced designs currently under development comprise three basic types: • • •
water-cooled reactors, using water as coolant and moderator, fast reactors, using liquid metal, e.g., sodium, as coolant, gas-cooled reactors, using gas, for example, helium, as coolant and graphite as moderator.
Global developments in this field have been summarized by Juhn et al. (1997). We present here their findings for some countries. United States Important programmes in development of ALWRs were initiated in the mid-1980s in the United States. In 1984, the Electric Power Research Institute (EPRI), in cooperation with the US Department of Energy (DOE) initiated a programme to develop utility requirements for ALWRs to guide their design and development. Utility requirements were established for large boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) having power ratings of 1200–1300 MWe, and for mid-sized BWRs and PWRs having power ratings of about 600 MWe. In 1986, the US DOE, in cooperation with EPRI and reactor design organizations, initiated a design certification programme for evolutionary plants based on a new licensing process, followed in 1990 by a design certification programme for mid-size plants with passive safety systems. The new licensing process allows nuclear plant designers to submit their designs to the US Nuclear Regulatory Commission (NRC) for design certification. Once a design is certified, the standardized units will be commercially offered, and a utility can order a plant, confident that generic design and safety issues have been resolved. The licensing process will allow the power company to request a combined licence to build and operate a new plant, and as long as the plant is built to preapproved specifications, the company can start up the plant when construction is complete, assuming no new safety issues have emerged. Four advanced reactor designs developed in the United States have been submitted to the NRC for certification under the US DOE ALWR programme. Two large evolutionary plants—the System 80+ of ABB-Combustion Engineering and the ABWR of General Electric—received Final Design Approval in 1994 and Design Certification in May 1997. The 600-MWe AP-600 of Westinghouse is under NRC review and a Final Design Approval is expected by March 1998. Up to mid-1996, the 600-MWe simplified BWR developed by General Electric was also under review, but then the company stopped work on the 600-MWe version and shifted its emphasis to a unit with larger output. The first-of-a-kind engineering programme (FOAKE, the detailed design needed to verify cost and the construction schedule)
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CHAPTER 6 The nuclear fuel cycle
authorized by the 1992 Energy Policy Act was completed for the ABWR in September 1996, and similar work on the AP-600 has also been done. The power company in Taiwan, China, recently selected General Electric’s ABWR design for two new units slated for operation in 2004. NASA is developing nuclear reactors for spacecraft propulsion and for a planetary power source; the goal is to have both types available for mission to Mars probably in 2030s. The agency is advancing a reactor technology that uses low-enriched uranium (LEU) containing less than 20% of 235U for propulsion, its planetary power source, known as kilopower, utilizes 90% or more enriched 235U. Kilopower reactor might compete with solar as power source for human settlement on Mars in spite of long lunar night period of half of month (Kramer, 2017). There is strong opposition to within some organizations since the 1 kW of kilopower reactor uses about 30 kg of weapon grade uranium, which is more than enough to produce a nuclear device. NASA states that planed 10 kW reactor will require only 50 kg of weapon-grade uranium. United States has set a mid of 2030s as goal for human travel to Mars. Although the HEU fuelled reactor is simple to build and test, the LEU reactor would eliminate concerns of fissile material failing into the wrong hands in the case of launch failure. France and Germany In Europe Framatome and Siemens have established a joint company, Nuclear Power International, which is developing a new advanced reactor, the European pressurized-water reactor (EPWR), a 1500-MWe plant with enhanced safety features. The basic design will be completed in mid-1997, and the design will be reviewed jointly by the French and German safety authorities. This procedure will provide strong motivation for the practical harmonization of the safety requirements of two major countries, which could later be enlarged on a broader basis. Siemens is also, together with German utilities, engaged in the development of an advanced BWR design, the SWR-1000, which will incorporate a number of passive safety features, for initiation of safety functions, for residual heat removal, and for containment heat removal. Sweden and Finland In Sweden, ABB Atom, with involvement of the utility Teollisuuden Voima Oy (TVO) of Finland, is developing the BWR-90 as an upgraded version of the BWRs operating in both countries. Republic of Korea In the Republic of Korea, an effort started in 1992 to develop an advanced design known as the Korean Next Generation Reactor (KNGR), a 4000-MWth PWR design. The basic design is currently being developed by the Korea Electric Power Corporation (KEPCO) with the support of the Korean nuclear industry. The goal is to complete a detailed standard design by 2000. Russian federation In the Russian Federation, design work is under way on the evolutionary V-392, an upgraded version of the VVER-1000, and another design version is being developed
6.7 Advanced nuclear power plants
in cooperation with the Finnish company Imatran Voima Oy (IVO). Also being developed is a mid-sized plant, the VVER-640 (V-407), an evolutionary design which incorporates passive safety systems, and the VPBER-600, which is a more innovative, integral design. Construction of the first unit of the VVER-640 is planned to start at Sosnovy Bor in 1997. Construction of two 1000-MWe VVERs is being discussed with the People’s Republic of China. Japan In Japan, the Ministry of Trade and Industry is conducting an “LWR Technology Sophistication” programme focusing on development of future LWRs and including requirements and design objectives. A large, evolutionary 1350-MWe advanced PWR is being developed by Japanese utilities together with nuclear vendors, with construction of a twin unit being planned at the Tsuruga site. In addition, an advanced BWR Improvement and Evolution study was started in 1991. It involves development of a reference 1500-MWe BWR that reflects the accumulated experience in operation and maintenance of BWRs. Development programmes for a Japanese Simplified BWR (JSBWR) and PWR (JSPWR) projects which involve vendors and utilities are also in progress. The Japan Atomic Energy Research Institute (JAERI) has been investigating conceptual designs of advanced water-cooled reactors with emphasis on passive safety systems. These are the JAERI Passive Safety Reactor (JPSR) and the System-Integrated PWR (SPWR). China In China, the Nuclear Power Institute (Chengdu) is developing the AC-600 advanced PWR, which incorporates passive safety systems for heat removal. In all of these countries, the advanced LWRs under development incorporate significant design simplifications, increased margins, and various technical and operational procedure improvements. These include better fuel performance and higher burn-up, a better man–machine interface using computers and improved information displays, greater plants standardization, improved constructability and maintainability and better operator qualification and simulator training. Canada The continuing design and development programme for heavy-water cooled reactors (HWRs) in Canada is primarily aimed at reduction of plant costs and at an evolutionary enhancement of plant performance and safety. Two new 715-MWe CANDU-6 units with improvements over earlier versions of this model are under construction in Quinshan, China. Up-front basic engineering continues on the 935-MWe CANDU-9 reactor, a single unit adaptation of reactor units operating in Darlington, Canada. The two-year licensability review by the Canadian Nuclear Safety Commission was completed in Japan 1997, and found that the CANDU-9 meets the country’s licensing requirements. Further studies are being carried out for advanced versions of these reactor models to incorporate further evolutionary improvements and to increase the output of the larger reactor up to 1300 MWe.
379
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CHAPTER 6 The nuclear fuel cycle
India Also under development is an advanced 500-MWe HWR in India, and construction of such units is planned. This HWR design takes advantage of experience feedback from the 220-MWe HWR plants of indigenous design operating in India. On the other hand, liquid metal-cooled fast reactors (LM-FRs), or breeders, have been under development for many years. With breeding capability, fast reactors can extract up to 60 times as much energy from uranium as can thermal reactors. The successful design, construction and operation of such plants in several countries, notably France and the Russian Federation, has provided more than 200 reactor-years of experience on which to base further improvements. In the future, fast reactors may also be used to burn plutonium and other long-lived transuranic radioisotopes, allowing isolation time for high-level radioactive waste to be reduced. Significant activities are occurring in the development of high-temperature gascooled reactors (HTGRs), particularly with regard to the utilization of the gas-cooled reactor to achieve high efficiency in the generation of electricity and in process heat applications. Technological advances in component design and processes—coupled with the international capability to fabricate, test and procure the components— provide an excellent opportunity for achieving HTGR commercialization. United Kingdom, Germany and United States Gas-cooled reactors have been in operation for many years. In the United Kingdom, nuclear electricity is mostly generated in CO2-cooled Magnox and advanced gascooled reactors (AGRs). Other countries also have pursued development of hightemperature reactors (HTGRs) with helium as coolant, and graphite as moderator. The 13-MWe AVR reactor has been successfully operated for 21 years in Germany demonstrating application of HTGR technology for electric power production. Other helium-cooled, graphite-moderated reactors have included the 300-MWe thorium high-temperature reactor in Germany, and the 40-MWe peach bottom and 330-MWe Fort St. Brain plants in the United States. South Africa In South Africa, the large national utility, Eskom, which has an installed generation capacity of about 38,000 MWe, is in the process of performing a technical and economic evaluation of a helium-cooled pebble bed module reactor. It would be directly coupled to a gas turbine power conversion system for consideration in increasing the capacity of the utility’s electrical system. China and Japan In China and Japan, test reactors are under construction which will have the capability of achieving core outlet temperatures of 950°C for the evaluation of nuclear process heat applications. Construction of China’s high temperature reactor (HTR-10) at the Institute of Nuclear Energy Technology (INET) continues with initial criticality anticipated for 1999. This pebble-bed reactor of 10 MW will be utilized to test and demonstrate the technology and safety features of the HTGR. Development of the HTGR by INET is being undertaken to evaluate a wide range
6.8 Nuclear fusion
of applications. They include electricity generation, steam and district heat production, combined steam and gas turbine cycle operation and the generation of process heat for methane reforming. The HTR-10 is the first HTGR to be licensed and constructed in China ( Juhn et al., 1997). Many countries have started the participation in IAEA programme for the development of small and medium-sized modular reactors (SMRs technology). The driving force in development of such reactors includes:
replacing the ageing fossil fuel power plants, enhancing safety performance through inherent and passive safety features, offering better economic affordability, suitability for nonelectric applications. options for remote areas, synergic energy systems that combine nuclear and renewable energy sources.
SMRs stands for small nuclear reactor and are defined as advanced nuclear reactors that produce equivalent electric power up to 300 MW(e) and are designed to be built in factories and transported to utilities for needed installation. For water-cooled SMRs modularity is achieved by integrating major components of the reactor coolant system inside the reactor pressure vessel (IAEA, 2016). Dozens of advanced SMR designs are under development or near deployment. The three reactors in SME category leading the group are one in Argentina (CAREM25—an industrial prototype integral PWR), one in the Russian federation (KLT-405 a barge mounted floating power unit) and one in China (HTR-PM, an industrial demonstration plant of high-temperature pebble-bed gas cooled reactor). The development of SMRs comes with several different concepts, coolants, neutron spectrum, deployment location and application (IAEA, 2016). The advanced SMRs currently under development in the United States represent a variety of sizes, technology options and deployment scenarios. These advanced reactors, envisioned to vary in size from a couple of MW up to hundreds of MW, can be used for power generation, process heat, desalination or other industrial uses. SMRs can employ light water as a coolant or other nonlight water coolants like gas, liquid metal or molten salt. As an example, here we mention SMRs called “U-Batteries” being developed by Urenco Ltd. (Tirone, 2017) able to generate 10 MW of power or heat. This SMR is being developed for small towns and industries in areas beyond the reach of large nuclear plants. According to its prospectus, central to the U-Battery design is its so-called TRISO fuel, a three-layered sphere with uranium kernel that can withstand temperatures as high as 1800°C.
6.8 Nuclear fusion A central issue for economic growth, prosperity and the quality of life in the industrialized world is the availability of secure, sustainable and financially competitive
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CHAPTER 6 The nuclear fuel cycle
50
Gt -OIL EQUIVALENT
382
40 high 30 20 low
10 0 1850
1900
1950
2000
2050
2100
YEAR
FIG. 6.7 Global energy use in gigatonne oil equivalent (Gtoe).
sources of energy. Given the expected growth in energy demand in the future, even with vigorous measures for energy savings, use will need to be made of all potential energy sources. The WEC projects growth in energy demand of anywhere between 50% and 300% over the next five decades, depending on environmental and economic factors (see Fig. 6.7). Strategic considerations favour the development of energy sources that offer greater sustainability and have less impact on health and the environment. Nuclear fusion, for which the fuel source is virtually limitless in quantity, could in the long term be an important option in this energy mix. There are several approaches to the problem of nuclear fusion. The most promising is definitely magnetic confinement fusion (MCF). In the course of the last 50 years research on magnetically confined plasmas has brought MCF to the threshold of net power production and has revealed much of the physics underlying the complex behaviour of hot plasmas immersed in a magnetic field. The focus of contemporary fusion research is the deuterium–tritium reaction: 2
H +3 H!4 Heð3:5 MeVÞ + nð14:1 MeVÞ
(6.7)
which is the fusion reaction with the largest cross-section at the temperatures which are likely to be achieved in laboratory experiments (several 108 K). A total of 80% of the reaction energy appears as the kinetic energy of the neutron, which would be absorbed in the structure of a power plant and provides most of the energy for steam generation. The α-particle would be trapped in the plasma where its energy would heat the plasma and maintain the conditions required for fusion reactions to occur. Since tritium is a radioactive gas and a high flux of 14.1 MeV neutrons would induce significant radioactivity in the structure surrounding the plasma, current experiments on magnetically confined plasma are usually carried out in hydrogen, so that no neutrons are produced, or in deuterium, for which the neutron production rate is almost two orders of magnitude lower than in a deuterium–tritium mixture.
6.8 Nuclear fusion
To achieve the conditions necessary for “ignition”, where the α-particle power produced by fusion reactions exactly balances the heat loss due to transport processes, the plasma must be heated to a temperature of approximately 108 K at a particle density in the region of 1010 ions per cubic metre, while maintaining an energy replacement time of about 5 s. There are three different toroidal confinement configurations, each of them being a potential route to a possible fusion power plant: 1. Tokamak uses a strong toroidal field of several Tesla produced by a set of discrete coils. 2. The reversed field pinch (RFP) is a closely related configuration, since the plasma formation and ohmic heating are essentially identical to the tokamak. However, in the tokamak the average poloidal field is limited by stability requirements to approximately an order of magnitude smaller than the toroidal field, whereas the two are of similar magnitude in the RFP, both being typically less than 1 T. 3. The third class of toroidal confinement devices is the stellarator which differs in an essential way from tokamaks and RFPs in that the helical fields are created entirely by coils external to the plasma, with no net toroidal current following within the plasma. In order for plasma to achieve ignition the product of plasma density, energyconfinement time and ion temperature must reach a value of 5 1021 m3 skeV. The increasing scale of magnetic confinement experiments, together with the accompanying improvements in the understanding of the physics of magnetoplasmas, has raised the values attained experimentally by seven orders of magnitude since 1955 and has brought the field to the present point, where the largest experiments are within a factor of 5 of the required value. On this basis it can be expected that the parameters of the ITER tokamak are adequate to ensure that ignition will be achieved (Wesson, 1997). According to Campbell (1998), three principal conclusions can be drawn about the present status of MCF. First, there is now substantial, though still incomplete, understanding of plasma behaviour in the principal toroidal confinement configurations, and there is a much deeper appreciation of the complexity of the physics of high-temperature magnetoplasmas. Second, new opportunities for further improvement in plasma performance are opening with the advent of a new generation of large stellarators such as LHD, the development of “advanced tokamak scenarios”, which may offer a viable route to steady-state tokamak operation, and realization of a variety of new tools for enhancing plasma performance in RFPs. Finally, given the production of over 10 MW of DT fusion power in TFTR and 16 MW in JET, plasma performance in tokamaks has advanced to the point where the construction of a DT-burning plasma experiment such as ITER would be a timely next step. An alternative to magnetic confinement is so-called inertial confinement fusion (ICF). The basic idea is to ignite and burn a few milligrams of deuterium-tritium fuel by means of high-power laser or ion beam pulses. Two large laser facilities are presently
383
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CHAPTER 6 The nuclear fuel cycle
under construction which should demonstrate within the next 5–10 years the feasibility of single microexplosions. These are the National Ignition Facility (NIF) in Livermore, USA and the Laser MegaJoule (LMJ) in Bordeaux, France. In contrast to MCF, inertial confinement involves no magnetic fields to contain the fuel, but relies exclusively on mass inertia. In ICF fusion burn occurs in highly compressed deuterium–tritium fuel, heated to an ignition temperature of 108 K. In the standard scheme compression and heating is achieved by spherical implosion of small capsules containing the fuel. A short pulse of radiation (laser, ion beam or X-ray radiation) is used to ablate the outer layer of the capsule and to implode the inner part, driven by the ablation pressure like a spherical rocket. The energy yield of the ignited capsule (up to some 100 MJ) can be contained in a reactor vessel. For energy production the scheme implies pulsed operation with a few microexplosions per second. Presently, there are two paths to achieving uniform irradiation. First, the direct drive approach, where a large number of overlapping beams is shone directly on the fusion capsule, and second, the indirect drive approach, where one converts the beam energy into X-rays which then drive the capsule implosion. At present, direct drive is thought to be possible only with lasers. The scientific and technological basis has been developed to ignite and burn microfusion targets by means of MJ laser pulses. In scaled experiments, implosions with high convergence ratio and neutron yields have been achieved, showing close agreement between experiment and multidimensional simulations. The crucial problem of symmetry and stability is approached along two lines, direct drive using laser smoothing techniques and indirect drive using gas-filled hohlraums (Lindl, 1995). Let us describe in some detail the European Union fusion programme (Bruhns, 1998). The starting point of this programme could be considered the creation of the European Atomic Energy Community (Euroatom) in 1957. Today, all EU member states have institutions actively participating in the fusion programme—all states except Greece participate through “association” contracts. The Community’s own Joint Research Centre (JRC), which has institutes in various locations, also undertakes work for the programme. Switzerland is fully associated to the programme (as Sweden was before it had become an EU Member State). Associations were established in Finland (1995) and in Austria (1996) after enlargement of the Union took in these countries. The associations are the backbone of the fusion programme. They operate a number of fusion devices in their laboratories (see Table 6.11). Most of these fusion devices have been built along the tokamak principle, but there are also stellarators and RFPs. And there are a number of facilities for technological development such as large superconducting-magnet-testing facilities. At the end of the 1970s it was decided to build, under the name of the JET Joint Undertaking, a fusion device (a tokamak) of much larger size than any fusion experiment existing at the time, JET, the Joint European Torus, located at Abingdon in the United Kingdom, began operation in 1983 and has become the flagship of the whole EU fusion programme.
Table 6.11 EU fusion devices
Name
Location
Objective
JET
Abingdon, UK
TORE SUPRA
Cadarache, F
TEXTOR 96
€lich, D Ju
ASDEXUpgrade FTU
Garching, D
TCV COMPASS-D
Lausanne, CH Culham, UK
ISTTOK MAST
Lisbon, P Culham, UK
Integrated high performance operation, DT operation Long-pulse operation in next step relevant conditions Plasma/wall interaction, heating and exhaust, pumped divertor Poloidal divertor, plasma purity control in reactor Confinement at high density and high current, high power wall studies High-beta studies and disruption control High-beta and MHD stability studies, poloidal divertor-type shaped plasma MHD activity transport Tight aspect ratio tokamak physics with hot plasmas, disruption avoidance
Power/ radius (MW/m)
Major/ minor radius (m)
Current (MA)
Magnetic Field (T)
2.96/1.25
7
3.45
2.96/1.25
1.7
4.5
9.3
88-
1.75/0.48
<0.8
3.0
4.6
83-
1.65/0.50
<1.6
3.9
0.93/0.3 J
<1.6
8.0
8.5
90-
0.88–0.24 0.56/0.21
<1.2 0.35
1.4 2.1
5.1 4.5
9289-
0.46/0.09 >0.6/>0.5
0.091 >1
0.5 0.5
Operational from
Tokamaks
Frascati, I
83-
91-
9398-
Stellarators WENDESTEIN 7-X WENDESTEIN 7-AS TJ-II
Greifswald, D Garching, D Madrid, E
Modulator cryogenic coil system to study fully optimized HELIAS in hot plasmas Modulator coils, plasma behaviour in an optimized magnetic configuration Flexible stellarator, helical magnetic axis for confinement and high-beta studies
5.5/0.55
3.0
3–7
2006-
2.0/0.20
2.5
2.4
90-
1.5/0.20
1.2
2.9
97-
Largest RFP device, to study the reactor prospects for this concept Medium size RFP; shell, stability and transport studies
2.0/0.48
<2.0
91-
2.0/0.18
0.5
94-
Reversed field pinches RFX
Padova, I
EXTRAP-T2
Stockholm, S
After Bruhns, H., 1998. The EU fusion programme. Europhysics News, November/December, p. 206.
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Around the same time, the Next European Torus team (NET) was established and given the task of enhancing the programme’s activities on safety and the environment, concentrating on the preparation (in particular the engineering and technological side) of the next-step experiment beyond JET. The NET team has become the pivotal point for initiating and coordinating R&D in fusion technology, as well as for Europe’s contribution to the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER EDA) which was established in 1992 by the EU, Japan, Russia and the United States.
6.8.1 ITER In the fusion field research has been mainly focused on the magnetic confinement of extremely hot plasma of fusion fuel ions and electrons while the practical aspects of the scheme have often been neglected. Because of Russian tokamak success in the late 1960s fusion researchers worldwide dropped most other approaches and built tokamak experiments and designed power plants based on tokamak (Dean, 2013). In the late 1980s, the idea of building a large internationally funded and managed tokamak facility was born. More than 200 tokamaks around the world have paved the way to the ITER experiment. International effort on fusion, International Thermonuclear Experimental Reactor (ITER), was proposed by Gorbachev (General secretary of Soviet Union at that time) to US president Reagan in Geneva meeting held in November 1985 in Geneva. A year later an agreement was reached between Euroatom (EU-28 plus Switzerland), Japan Soviet Union and United States to jointly pursue the design of large international fusion facility. Conceptual design work began in 1998 and finished in July 2001 by approval of Member States (IAEA, 2001). The People’s Republic of China and the republic of Korea joined the Project in 2003. In December 2005 the delegations from China, European Union, Japan, the Republic of Korea, the Russian Federation and the United States of America met in Korea to complete their negotiations on an agreement on the joint implementation of the ITER international fusion energy project. In this meeting India was welcomed as a full party to the ITER venture, the delegation from India then participating fully in the discussions that followed. At that time the ITER Parties represented 80% of the global GDP and half of the world’s population. The aim of ITER is to demonstrate the scientific and technological feasibility of fusion energy by constructing a functional fusion power plant. The IAEA has been actively involved in the ITER project from its inception, providing its auspices and practical support, including publication of technical documents and ITER Newsletter. In December 2005 the ITER Joint Work Site in Cadarache, France, was inaugurated. The offices became operational in January 2006. The ITER Agreement was officially signed in Paris on 21 November 2006 by Ministers from seven ITER Members, this document established a legal international entity to be responsible for the building, operating and decommissioning the Project. Construction of ITER is expected to take about 10 years, and the reactor will then operate for a further 20 years.
6.8 Nuclear fusion
ITER’s mission is to demonstrate the feasibility of fusion power and to prove that the tokamak type of magnetic confinement can achieve the following (Freidberg, 2007): produce 10 times more thermal energy from fusion power (Q ¼ 10) than is supplied by auxiliary heating for period of few minutes, produce a steady-state plasma with a Q > 5 in a long pulse for about 8 min, develop the advanced technologies and processes needed for a fusion power plant, verify the rates of tritium breeding from lithium, find the life-time of vessel’s walls from the neutron bombardement and heating and erosion. Since Europe is the host of this experiment its financing is of the project represents almost half of the total price. The EU body to manage the European contribution to ITER is called “Fusion for Energy” (F4E) has about 450 employees in three locations: Barcelona (Spain), Cadarache (France) and Garching (Germany). Since its establishment F4E has invested in Europe’s economy about 4 billion EUR (EC News, 2018). The ITER project is now under construction in Saint-Paul-lez-Durance, France. It aims to demonstrate sustained fusion power output of 500 MW thermal. In December 2017 ITER celebrated an important milestone having reached 50% completion of the total construction work needed for the first operation stage, so-called First Plasma. ITER has been discussed in many books and scientific, technical and popular papers and presentation. One of the books to be recommended is the book by Horton and Benkadda (2015) which is dedicated to the ITER tokamak now under construction in France which aims to investigate the feasibility of fusion power. Another book of interest is book by Horton and Benkadda (2015) on ITER Physics discussing ITER machine architecture and objective, magnetohydrodynamic description of equilibrium and heating of thermal plasma, operational regimes, steady-state operation, plasma diagnostic and other aspects of d + t reaction including numerous references to scientific papers in all aspect of tokamak physics, Some authors (i.e. Hirsch, 2017) question the purpose of this project by asking why ITER continues to be built; he proposes a dramatic reorganization of fusion research with an accent on p + 11B reaction which does not produce neutrons. Arguing that the absence of neutrons would largely eliminate the risks due to radioactivity and thereby dramatically enhance economics, regulatory simplicity and public acceptance. Cowley (2017) has replied to concerns brought up by Hirsch, 2017 by stating that ITER success is not assured at this point of time so it is rather early to state that tokamaks fail against the Electrical Power Research Institute (EPRI) criteria of the economic viability of electricity production. Success has always seemed just a few decades away from ITER because of design and management problems which led to long delays and enormous increase in cost. It is expected that the burning plasma sustained for 6 or 7 minutes could be achieved 2035 at the earliest for a price of around 25 billion Euros. We finish this part of text
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with the title of the article in The New York Times written by Fountain (2017): “A dream of clean energy at a very high price”. The book by Horton and Benkadda (2015) is dedicated to the ITER tokamak under construction in France which aims to investigate the feasibility of fusion power.
6.9 Nuclear batteries The goal of nuclear battery is to convert energy from radioactive decay into electricity. This can be done by following options: (i) direct charge collection, (ii) indirect (scintillation), (iii) betavoltaic, (iv) thermoelectric, (v) thermionic and (vi) thermophotovoltaic. Nuclear batteries have been investigated and used because of the potential of nuclear battery for longer shelf-life and higher energy density when compared with other modes of energy storage. The performance of nuclear batteries is a function of the radioisotope used as a fuel, radiation transport properties and energy conversion transducers (Prelas et al., 2014). Table 6.12 lists the most often considered isotopes and their characteristic parameters. A useful short review and preview of nuclear battery technology is presented by Park (2017) which was submitted as coursework for PH241 at Stanford University in winter 2017. The extensive description of the physics of nuclear battery operation is given in the book by Prelas, Boraas, and De La Torre Aguilar (2016b). It provides a comprehensive background that allows readers to understand all past and future developments in the field. The book has a chapter on potential applications for nuclear batteries (Prelas, Boraas, De La Torre Aguilar, Seelig, et al., 2016a). Radioisotope heater units, RHU; several types are available. NASA’s RHU has heat output of 1 W, fuel loading is 33.6 Ci—2.7 g of 238Pu in oxide form, the unit weight is 39.69 g, unit size is ɸ ¼ 2.54 cm 3.56 cm. The unit is rugged and very Table 6.12 Radioisotopes often considered as a battery fuel
Isotope 3
H Ni 90 Sr/90Y 147 Pm 210 Po 238 Pu 242 Cm 244 Cm 63
Average energy (keV) 5.7 17 200/930 5300 5500 5810
Half-life/ year 12 100 29/2 days 2.6234 0.38 88 161.4 days 18
Specific activity (Ci/g)
Specific power (W/g)
Power density (W/cm3)
9700 57 140
0.33 0.0067 0.98
4500 17
140 0.56
81
2.8
— 0.056 2.5 1.8 1210 11 882 38
6.9 Nuclear batteries
reliable. Larger units, called radioisotope thermoelectric generators (RTGs), have been used in many NASA missions. Usually, ceramic 238Pu radioisotope was used to provide heat with subsequent electricity production by thermoelectric effect. Such a unit does not have any moving parts it contains about 2.7 kg of fuel (133 kCi) producing power of 276 W. Dimensions of the unit were D ¼ 42 cm, L ¼ 114 cm, total weight-56 kg, lifetime over 20 years. More than 40 units have been flown by United States. Since the energy conversion mechanisms vary significantly between different nuclear types, therefore RTG is usually considered a performance standard for all nuclear battery types. NASA and the DOE have developed a new generation of such power systems that could be used for a variety of space missions. The newest RTG, called a multimission radioisotope thermoelectric generator (MMRTG), has been designed to operate on Mars and in the vacuum of space. The MMRTG has a flexible modular design capable of meeting the needs of a wide variety of missions, as it generates electrical power in smaller increments than previous generations of RTGs, about 110 W at launch. The design goals for the MMRTG included optimizing power levels over a minimum lifetime of 14 years and ensuring a high degree of safety. The MMRTG contains a total of 4.8 kg of plutonium dioxide (including 238Pu) that initially provides approximately 2000 W of thermal power and 110 W of electrical power when exposed to deep space environments. The thermoelectric materials (PbSnTe, TAGS and PbTe) have demonstrated extended lifetime and satisfactory performance capabilities. The MMRTG generator is about 64 cm in diameter, 66 cm tall and weighs about 45 kg, for more information about NASA’s use of Radioisotope Power Systems, see: rps. nasa.gov (NASA, 2013). Let us mention that the Curiosity Rover exploring planet Mars (big as a large car) is also using MMRTP, as an energy source. It is designed to run at least one Martian year, which is almost two Earth years. However, 238Pu is in a short supply, according to some reports there is only enough left to make three more nuclear batteries! The solution may-be in the navy system called stored chemical energy power systems (SCEPS) developed in the 1980s by engineers at Pennsylvania State University. SCEPS harnesses the chemical reaction of two energy-dense reactants that remain stored and separated until needed. In torpedoes, the system commonly holds its energy in reserve as a solid block of lithium and a tank of the inert gas sulphur hexafluoride. When triggered, the combustion reaction of the two materials creates heat that turns the weapon’s steam turbine to produce thousands of kilowatts of power. Nuclear battery has been used in pacemakers. Fuel was 3 Ci of 238Pu producing <1 mW power and delivering 100 mrem/year to the patient. Tracking the used Pu was difficult and that made regulators concerned so the nuclear battery in pacemakers was replaced by Li battery having 10 years life. Another attempt was made with 147Pm beta sources combined with custom designed Si n/p cells to produce nuclear batteries for heart pacemakers. The power sources were referred to as Betacel batteries. The Betacel was used to power cardiac pacemakers in over 100 patients during the 1970s (Olsen et al., 2017), but due to gamma radiation concerns; however, the lithium battery was able to capture the pacemaker market.
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Tritium has been used in construction of nuclear batteries for many different applications. For the technology used see https://citylabs.net/technology-overview/. The company City labs, Inc. product is the long-life (20 + years) Nanotritium™ battery for microelectronics. The battery is resistant to varying temperatures and other environmental conditions due to the robustness of the tritium decay process. The battery also serves as a vital component for the security and defence of electronics. As confirmed in independent testing, the City Labs battery is resistant to extreme temperature variance ( 55°C to +150°C), as well as extreme vibration and altitude, due to the robust architecture of City Labs’ proprietary technology and the use of tritium (Olsen et al., 2017). Another manufacturer of long-lived nuclear batteries powered by hydrogen isotopes for military applications is Widetronix (Bourzac, 2009). Their batteries could potentially power electrical circuits that protect military planes and missiles from tampering by destroying information stored in the systems, or by sending out a warning signal to a military centre. The batteries are expected to last for 25 years. The company is also working with medical-device makers to develop batteries that could last decades for implantable medical devices. It should be mentioned that Defense Science Board identified “nuclear batteries” as an essential technology for the US military in the 21st century (Weintz, 2014). As the most promising candidate betavoltaic batteries have been identified; from the outside the future betavoltaic batteries could look identical to present day military batteries. The most promising candidates today for betavoltaic batteries include 90 Sr, 63Ni and 3H. All three radionuclides emit beta radiation with almost no γ radiation, and they have long life times. While the strontium and nickel-based batteries are still mostly experimental, tritium-based batteries are available on the market. Also, The Army Research Laboratory has developed prototype nuclear batteries powered by tritium matching the Army’s existing BA-5590 battery pack in size and using the same connector. It can last for 13 years but the problem of disposal hazards remains (see also Seffers, 2013). An interesting attempt to use nuclear waste to generate electricity in a nuclearpowered battery has been recently reported. Namely, one of the largest nuclear waste products by volume is radionuclide 14C produced by irradiation of graphite blocks used in graphite-moderated reactors. The radionuclide 14C could theoretically be extracted and compressed into radioactive diamonds. These diamonds would then be incased in a regular, nonradioactive diamond, to provide shielding (Connor et al., 2016). The theoretical 14C diamond battery would have a lower energy density than chemical-based batteries like the lithium-ion batteries that power laptops and phones, and the alkaline batteries like AA batteries used in small electronics gadgets. A typical AA battery stores 13,000 J and is exhausted after about 24 h, while a 14C diamond battery would produce only 15 J per day but have a half-life of almost 6000 years (Bormashov et al., 2015). Thus far, research groups in Russia and Canada have produced prototype radioactive diamond batteries from 63Ni (Conca, 2016). Although diamond batteries are by no means a solution to the issue of nuclear waste
6.9 Nuclear batteries
recycling and storage, they offer a channel to put some of the waste to productive use as safe, durable, energy sources to eventually power exploration and medical electronics. Bormashov et al. (2018) recently reported a fabrication of betavoltaic battery prototype consisting of 200 single conversion cells based on Schottky barrier diamond diodes which have been vertically stacked with 24% 63Ni radioactive isotope. The maximum electrical output power of about 0.93 μW was obtained in total volume of 5 5 3.5 mm3. They used the ion-beam assisted lift-off technique to obtain conversion cells of minimal thickness comparable with the characteristic penetration length of beta-particles emitted by 63Ni isotope. The obtained value of 15 μm was limited by the mechanical strength of produced structures and process reliability. The fabricated prototype provided the output power density of about 10 μW/cm3, which is the best known value for nuclear batteries based on 63Ni radioisotope. Moreover, the long half-life of 63Ni isotope gives the battery specific energy of about 3300 mWh/g, an order of magnitude higher than the typical value of commercial chemical. In conclusion radioisotopes provide a high-energy density power source suitable for many applications. They are outstanding for small-scale power. According to Prelas et al. (2014) the physics of nuclear batteries does not support the objectives of miniaturization, high efficiency and high-power density, rather the physics considerations imply that nuclear batteries should be of moderate size and limited power density. In addition the supply of radioisotopes is limited and cannot support largescale commercialization. M€ oller and Wegener (2016) have presented the mathematical formalism and fundamental considerations for the use of ion accelerators for isotope production. A focus is put on the production of nuclear power sources to substitute 238Pu-based batteries. 20 MeV proton bombardment of bismuth produced α-emitting polonium isotopes with an energy efficiency of up to 0.031%. Some hours were required to produce a 1 W power source of the 2.9 year half-life α emitter 208Po. Apparently, the accelerator approach offers more flexibility for tailoring of nuclear products and less waste. One possible next-generation nuclear battery technology, as proposed by Kim and Kwon (2014), could be Aqueous Nuclear Battery, which is also known as water-based nuclear battery. It uses liquid medium for radiolysis, absorbing the kinetic energy of beta particles which is lost in betavoltaic cells. In their design using nanoporous titanium dioxide semiconductors coated in platinum, a high efficiency of 53.88% was reached at a potential of 0.9 V. Using an aqueous solution for radiolytic energy conversion results in higher energy level and lower temperature than using a solid-state material does. We are not discussed the controversial subjects of thorium powered cars and thorium plasma battery technology. The internet is full of websites discussing pro and cons, conspiracy theories, etc., instead, we caution the reader not to be eager to believe statements found on scientifically valueless websites.
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References Blix, H., 1997. Nuclear Energy in the 21st Century. Nuclear News, p. 34. September. Bormashov, V., Troschiev, S., Volkov, A., Tarelkin, S., Korostylev, E., Golovanov, A., Kuznetsov, M., Teteruk, D., Kornilov, N., Terentiev, S., Buga, S., Blank, V., 2015. Development of nuclear microbattery prototype based on Schottky barrier diamond diodes. Phys. Status Solidi A 212, 2539–2547. Bormashov, V.S., Troschiev, S.Y., Tarelkin, S.A., Volkov, A.P., Teteruk, D.V., Golovanov, A.V., Kuznatzov, M.S., Kornilov, N.V., Terentiev, S.A., Blank, V.D., 2018. High power density nuclear battery prototype based on diamond Schottky diodes. Diamond Relat. Mater. 84, 41–47. Boudard, A., Leroy, S., Volant, C., 1998. Spallation studies for nuclear waste transmutation. Nucl. Phys. News 8, 18. Bourzac, K., 2009. A 25-Year Battery: Long-Lived Nuclear Batteries Powered by Hydrogen Isotopes Are in Testing for Military Applications. November 17, 2009, MIT Technology Review. https://www.technologyreview.com/topic/sustainable-energy/. Bowman, C.D., et al., 1992. Nuclear Energy Generation and Waste Transmutation Using an Accelerator Driven Intense Thermal Neutron Source. Los Alamos Report LAUR-91-260 and Nuclear Instruments and Methods A320. p. 336. Bruhns, H., 1998. The EU Fusion Programme. Europhysics News, p. 206. November/ December. Campbell, D., 1998. Magnetic Confinement. Fusion. Europhysics News, p. 196. November/ December. Carminati, F., Klapisch, R., Revo´l, J.P., Roche, Ch., Rubio´, J.A., Rubia, C., 1993. An energy amplifier for cleaner and inexhaustible nuclear energy production driven by a particle beam accelerator. Preprint CERN/AT/93-47/ET (1993). Chan, C.Y., 1992. Radioactive Waste Management: An International Perspective. 3 IAEA Bulletin, p. 7. City Labs, Inc Corporate Office, 301 Civic Court: Homestead, FL 33030, USA. Conca, J., 2016. Radioactive Diamond Batteries: Making Good Use of Nuclear Waste. Forbes. 9 Dec 9, 2016. Connor, D.T., Martin, P.G., Scott, T.B., 2016. Airborne radiation mapping: overview and application of current and future aerial systems. Int. J. Remote Sens. 37, 5953–5987. Cowley, S., Reply to Hirsch, R. L. 2017. Necessary and sufficient conditions for practical fusion power. Phys. Today 70, 13–14. Dean, S.O., 2013. Search for the Ultimate Energy Source: A History of the U.S. Fusion Energy Program. Springer. Fells, I., 1998. The Need for Energy. Europhysics News, p. 193. November/December. Flakus, F.-N., Johnson, L.D., 1998. Binding agreements for nuclear safety: the global legal framework. IAEA Bull. 40, 21. Fountain, H., 2017. A Dream of Clean Energy at a Very High Price. The New york Times, p. 2017. March 27. Freidberg, J.P., 2007. Plasma Physics and Fusion Energy. Cambridge University Press, Vienna. Hannum, W.H., Marsh, G.E., Stanford, G.S., 2005. Smarter Use of Nuclear Waste. 2005. Scientific American. December issue. Hirsch, R.L., 2017. Necessary and sufficient conditions for practical fusion power. Phys. Today 70, 11–13.
References
Horton, C.W., Benkadda, S., 2015. ITER Physics. World Scientific Publishing Co. IAEA, 2001. Summary of the ITER Final Design Report. International Thermonuclear Experimental Reactor (ITER)-Engineering Design Activities (EDA) Documentation Series No. 22. International Atomic Energy Agency, Vienna. 2001. IAEA, 2016. Nuclear Power Reactors in the World. Reference Data Series No. 2, 2016 Edition. International Atomic Energy Agency, Vienna. 2016. IAEA, 2017a. Energy, Electricity and Nuclear Power Estimates for the Period up to 2050. Reference Data Series No. 1, 2017 Edition. International Atomic Energy Agency, Vienna 2017. IAEA, 2017b. Research Reactors for the Development of Materials and Fuels for Innovative Nuclear Energy Systems. IAEA Nuclear Energy Series No. NP-T-5.8. International Atomic Energy Agency, IAEA, Vienna. 2017. IAEA, 2018a. Status and Trends in Spent Fuel and Radioactive Waste Management. IAEA Nuclear Energy Series No. NW-T-1.14. IAEA, Vienna. 2018. IAEA, 2018b. Nuclear Power Reactors in the World. 2018 Edition. Reference Data Series No. International Atomic Energy Agency, Vienna. 2018. IAEA, 2018c. Operating Experience With Nuclear Power Stations in Member States, 2018 Edition. STI/PUB/1828. International Atomic Energy Agency, Vienna. 2018. IAEA, 2018d. Commissioning Guidelines for Nuclear Power Plants. IAEA Nuclear Energy Series No. NP-T-2.10. International Atomic Energy Agency, Vienna. 2018. IAEA, 2018e. Maintenance Optimization Programme for Nuclear Power Plants. IAEA Nuclear Energy Series No. NP-T-3.8. International Atomic Energy Agency, Vienna. 2018. IAEA, 2018f. Computer security of instrumentation and control Systems at Nuclear Facilities, technical guidance. IAEA Nuclear Security Series No. 33-T. International Atomic Energy Agency, Vienna. 2018. IAEA, 2018g. Physical Protection of Nuclear Material and Nuclear Facilities (implementation of INFCIRC/225/revision 5). IAEA Nuclear Security Series No. 27-G. International Atomic Energy Agency, Vienna. 2018. Jones, S.R., Williams, S.M., Smith, A.D., Gray, J., 1995. Review of Discharge History and Population Doses From the Sellafield Reprocessing Plant in Cumbria, UK: The Sellafield Environmental Assessment Model (SEAM), Reported at International Symposium on Environmental Impact of Radioactive Releases. Vienna, 8–12.05.1995, IAEA-SM-339/II. Juhn, P.-E., Kupitz, J., Cleveland, J., 1997. Advanced nuclear power plants—high lights of global development. IAEA Bull. 39, 13. Kershaw, P.J., Woodhead, D.S., Malcolm, S.J., Allington, D.J., Lovett, M.B., 1990. A sediment history of Sellafield discharges. J. Environ. Radioact. 12, 201–241. Kim, B.H., Kwon, J.W., 2014. Plasmon-assisted radiolytic energy conversion in aqueous solutions. Sci. Rep. 4, 5249 (9 p.). Kramer, D., 2017. NASA sees a future with nuclear power. Phys. Today 26–29 (December). Lindl, J., 1995. Development of the indirect-drive approach to inertial confinement fusion and the target physics basis for ignition and gain. Phys. Plasmas 2, 3933. M€oller, S., Wegener, T., 2016. Production of nuclear sources and nuclear batteries by proton irradiation. arXiv. 1608.05199v2 [Physics.Ins-Det], August 2016. NASA, 2013. Multi-Mission Radioisotope Thermoelectric Generator (MMRTG). NASA Facts. www.nasa.gov. October 2013. News, E.C., 2018. Europe’s Investment in the ITER Fusion Project: Mastering the Power of the Sun and the Stars. European Commission NEWS, Energy, Brussels. 13 April 2018.
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International Atomic Energy Agency Nuclear Research Reactors in the World, Dec. 1996 ed., Reference Data Series No. 3, Vienna, 1996. Oi, N., 1998. Plutonium challenges—changing dimensions of global cooperation. IAEA Bull. 40, 12. Olsen, L., Serralta, D., Cabauy, P., 2017. Betavoltaic Batteries: A historical Review. http:// citylabs.net/wp-content/uploads/2017/08/BetavoltaicHistory.pdf. Park, J., 2017. Review and Preview of Nuclear Battery Technology. Submitted as Coursework for PH241. Stanford University. Winter 2017. Prelas, M.A., Weaver, C.L., Watermann, M.L., Lukosi, E.D., Schott, R.J., Wisniewski, D.A., 2014. A review of nuclear batteries. Prog. Nucl. Energy 75, 117–148. Prelas, M., Boraas, M., De La Torre Aguilar, F., Seelig, J.D., Tchakoua Tchouaso, M., Wisniewski, D., 2016a. Potential Applications for Nuclear Batteries. In: Nuclear Batteries and Radioisotopes. Lecture Notes in Energy, vol. 56. Springer, Cham. Prelas, M., Boraas, M., De La Torre Aguilar, F. 2016b. Nuclear Batteries and Radioisotopes. (Lecture Notes in Energy). first ed. 2016 ed.Springer, Cham. C. Rubbia et al., Conceptual design of a fast neutron operated high power energy amplifier, Preprint CERN/AT/95-44 (ET) (1995). Seffers, G.I., 2013. Radioisotope Research May Revolutionize Battlefield Batteries. SIGNAL December 1. AFCEA International, Fairfax, VA Semenov, B.A., Oi, N., 1993. Nuclear fuel cycles: adjusting to new realities. IAEA Bull. 3, 2. Stather, J.W., Wrixon, A.D., Simmonds, J.R., 1984. The Risks of Leukaemia and Other Cancers in Seascale From Radiation Exposure, NRPB-R171. HMSO. Stather, J.W., Dionian, J., Brown, J., Fell, T.P., Muirhead, C.R., 1986. The Risks of Leukaemia and Other Cancers in Seascale From Radiation Exposure: Addendum to Report R171. NRPB-R171 Addendum. HMSO. Stather, J.W., Clarke, R.H., Duncan, K.P., 1988. The risk of childhood leukaemia near nuclear establishments. NRPB-R215. HMSO. Takizuka, T., 1994. Proceedings of the 8th Journees SATURNE, Saclay, May 5–6. Tirone, J., 2017. This Company is Designing Nuclear “Batteries” for Towns and Industry. Source Urenco Ltd, Stoge Poges, England. Van Tuyle, G.J., et al., 1993. Nucl. Technol. 101, 1. Weintz, S., 2014. The US Military is Working on Nuclear Batteries. https://medium.com/ war-is-boring/powering-the-militarys-future-batteries-nuclear-style-472975d7de8. Wesson, J.A., 1997. Tokamaks, second ed. Oxford University Press. WNA, 2018. Processing of Used Nuclear Fuel. World Nuclear Association. http://worldnuclear.org/information-library/nuclear-fuel-cycle/fuel-recycling/processing-of-usednuclear-fuel.
Further reading Arkhipov, V., 1997. Future nuclear energy systems: generating electricity burning wastes. IAEA Bull. 39, 30. IAEA Bulletin, 39 (1997) 13, International Atomic Energy Agency, Vienna, Austria. Safety Series No. 101, 1990. Operational Radiation Protection: A Guide to Optimization. Safety Series No. 105, 1990. The Regulatory Process for the Decommissioning of Nuclear Facilities.
Further reading
Safety Series No. 107, 1992. Radiation Safety of Gamma and Electron Irradiation Facilities. Safety Series No. 108, 1992. Design and Operation of Radioactive Waste Incineration Facilities. Safety Series No. 109, 1994. Intervention Criteria in a Nuclear or Radiation Emergency. Safety Series No. 110, 1993. The Safety of Nuclear Installations. Safety Series No. 111, 1995. The Principles of Radioactive Waste Management. Safety Series No. 111 -G-1.1, 1994. Classification of Radioactive Waste. Safety Series No. 111-G-3.1, 1994. Siting of Near Surface Disposal Facilities. Safety Series No. 111-G-4.1, 1994. Siting of Geological Disposal Facilities. Safety Series No. 111-S-1: Establishing a National System for Radioactive Waste Management (1995). Safety Series No. 112, 1994. Compliance Assurance for the Safe Transport of Radioactive Material. Safety Series No. 113, 1994. Quality Assurance for the Safe Transport of Radioactive Material. Safety Series No. 120, 1996. Radiation Protection and the Safety of Radiation Sources. Safety Series No. 35-G1, 1994. Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report. Safety Series No. 35-G2, 1994. Safety in the Utilization and Modification of Research. Safety Series No. 50-C-Q, 1996. Quality Assurance for Safety in Nuclear Power Plants and Other Nuclear Installations. Safety Series No. 50-C-S (Rev. 1), 1988. Code on the Safety of Nuclear Power Plants: Siting. Safety Series No. 50-SG-010, 1985. Core Management and Fuel Handling for Nuclear Power Plants. Safety Series No. 50-SG-012, 1994. Periodic Safety Review of Operational Nuclear Power Plants. Safety Series No. 50-SG-06, 1982. Preparedness of the Operating Organization (Licensee) for Emergencies at Nuclear Power Plants. Safety Series No. 50-SG-D1, 1979. Safety Functions and Component Classification for BWR, PWR and PTR. Safety Series No. 50-SG-D12, 1985. Design of the Reactor Containment Systems in Nuclear Power Plants. Safety Series No. 50-SG-D2 (Rev. 1), 1992. Fire Protection in Nuclear Power Plants. Safety Series No. 50-SG-D4, 1980. Protection against Internally Generated Missiles and Their Secondary Effects in Nuclear Power Plants. Safety Series No. 50-SG-D5 (Rev. 1), 1996. External Man-induced Events in Relation to Nuclear Power Plant Design. Safety Series No. 50-SG-D7, 1991. Emergency Power Systems at Nuclear Power Plants. Safety Series No. 50-SG-D9, 1985. Design Aspects of Radiation Protection for Nuclear Power Plants. Safety Series No. 50-SG-G4 (Rev: 1), 1996. Inspection and Enforcement by the Regulatory Body for Nuclear Power Plants. Safety Series No. 50-SG-G6, 1982. Preparedness of Public Authorities for Emergencies at Nuclear Power Plants. Safety Series No. 50-SG-G8, 1982. Licenses for Nuclear Power Plants: Content, Format and Legal Considerations. Safety Series No. 50-SG-G9, 1984. Regulations and Guides for Nuclear Power Plants.
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Safety Series No. 50-SG-O1 (rev. 1), 1991. Staffing of Nuclear Power Plants and the Recruitment, Training and Authorization of Operating Personnel. Safety Series No. 50-SG-O4, 1980. Commissioning Procedures for Nuclear Power Plants. Safety Series No. 50-SG-Q1, 1996. Establishing and Implementing a Quality Assurance Programme. Safety Series No. 50-SG-Q10, 1996. Quality Assurance in Design. Safety Series No. 50-SG-Q11, 1996. Quality Assurance in Construction. Safety Series No. 50-SG-Q12, 1996. Quality Assurance in Commissioning. Safety Series No. 50-SG-Q13, 1996. Quality Assurance in Operation. Safety Series No. 50-SG-Q14, 1996. Quality Assurance Decommissioning. Safety Series No. 50-SG-Q2, 1996. Non-conformance Control and Corrective Actions. Safety Series No. 50-SG-Q3, 1996. Document Control and Records. Safety Series No. 50-SG-Q4, 1996. Inspection and Testing for Acceptance. Safety Series No. 50-SG-Q5, 1996. Assessment of the Implementation of the Quality Assurance Programme. Safety Series No. 50-SG-Q6, 1996. Quality Assurance in the Procurement of Items and Services. Safety Series No. 50-SG-Q7, 1996. Quality Assurance in Manufacturing. Safety Series No. 50-SG-Q8, 1996. Quality Assurance in Research and Development. Safety Series No. 50-SG-Q9, 1996. Quality Assurance in Siting. Safety Series No. 50-SG-S1 (Rev. 1), 1991. Earthquakes and Associated Topics in Relation to Nuclear Power Plants Siting. Safety Series No. 50-SG-S8, 1986. Safety Aspects of the Foundations of Nuclear Power Plants. Safety Series No. 50-SG-S9, 1984. Site Survey for Nuclear Power Plants. Safety Series No. 79, 1986. Design of Radioactive Waste Management Systems at Nuclear Power Plants. Safety Series No. 90, 1989. The Application of the Principles for Limiting Releases of Radioactive Effluents in the Case of the Mining and Milling of Radioactive Ores. Safety Series No. 93: System of Reporting Unusual Events in Nuclear Power Plants, 1989. Safety Series No. 96, 1989. Guidance for Regulation of Underground Repositories for Disposal of Radioactive Wastes. Safety Series No. 98, 1989. On-Site Habitability in the Event of an Accident at a Nuclear Facility. Semenov, B.A., 1992. Disposal of spent fuel and high-level radioactive waste: building international consensus. IAEA Bull. 3, 2.
The nuclear fuel cycle
6
Chapter outline 6.1 6.2 6.3 6.4 6.5
Introduction .....................................................................................................329 The Status of nuclear power in the world...........................................................330 Nuclear safety..................................................................................................357 Releases of effluents ........................................................................................360 Management of radioactive wastes ...................................................................364 6.5.1 Storage and disposal.......................................................................367 6.5.2 Processing of used nuclear fuel........................................................368 6.5.3 Smart use of nuclear waste..............................................................373 6.6 Research reactors ............................................................................................374 6.7 Advanced nuclear power plants ........................................................................377 6.8 Nuclear fusion .................................................................................................381 6.8.1 ITER..............................................................................................386 6.9 Nuclear batteries .............................................................................................388 References .............................................................................................................392 Further reading .......................................................................................................394
6.1 Introduction This subject is a part of a broader subject which could be called “a human need for energy”. According to the World Energy Council (WEC), the demand for energy is expected to increase further. The reasons for this are clear: 50% of the countries of the world are in the process of industrialization with economies growing rapidly, some at 10% or more. This calls for an ever-increasing supply of coal, oil and now natural gas. China, for instance, which sustained a 10% growth rate in its market economy for over a decade, burns 1.2 Gt of coal per year (1996 figures) and expects to burn 1.4 Gt of coal per year after 2000. In addition, it imports oil, has an embryonic nuclear industry, and is seeking to diversify its renewable energy supplies beyond hydroelectric power. The global situation is made more challenging by the expected rapid increase in the world population. UN figures suggest that the current population of nearly 6 billion will
Radioactivity in the Environment. https://doi.org/10.1016/B978-0-444-64146-5.00006-9 # 2019 Elsevier B.V. All rights reserved.
329
330
CHAPTER 6 The nuclear fuel cycle
increase to over 10 billion by 2050; 80% of that population will be in developing countries where, to put matters into perspective, 50% of the population do not have an electricity supply connection (neither do they have safe drinking water) so their demand for the services provided by energy and an improved life-style is undeniable (Fells, 1998). There are three energy sources, all of which have enormous potential, which might give us confidence that the world’s energy needs will be met in the second half of the 21st century. These are nuclear energy, renewable energy and nuclear fusion. They have the advantage that they are also environmentally clean and do not emit large quantities of greenhouse gases. Nuclear fusion has enormous potential, but will be very difficult to achieve from an engineering point of view. It is prudent to rely, for the time being at any rate, on the growth in renewable and nuclear energy. They are not, incidentally, mutually exclusive but complementary, and it is very difficult to imagine any future scenario post-2050 without a large slice of energy from both sources. Nuclear power, in particular, plays an important role in controlling carbon-dioxide emissions. The nuclear future may well lie with the fast-breeder reactor which uses uranium some 60 times more efficiently than current fission reactors. Prototype breeder reactors are in operation in Russia and Japan and were so until recently in the United Kingdom and France. They will be required post-2030 or so if a major new nuclear programme is embarked on, as uranium resources are predicted to start running into short supply at about that time. The problem with the continuing growth in nuclear power is the public’s perception of its safety, particularly the safety of radioactive waste disposal. If the spent nuclear fuel is reprocessed, as happens in the United Kingdom and France, and the high-level radioactive waste classified and stored in metal cylinders in a dry rock store, there is general international consensus that it will be very safety contained. In addition, the method of reprocessing recovers plutonium and unused uranium, both of which can be recycled as fuel for fission reactors. The other method of dealing with spent nuclear fuel is to merely store it either in cooling ponds or directly in a dry store. This presents a less tidy legacy for the future but is the preferred method in some countries, fearful that reprocessing will lead to proliferation (Fells, 1998).
6.2 The status of nuclear power in the world The concept of a nuclear fuel cycle is an old one, almost dating back to the concept of controlled nuclear fission to generate electricity. At the time of the development of the first nuclear power plants, it was generally taken for granted that fuel from power reactors would be reprocessed and that the recovered uranium and plutonium would be recycled.
6.2 The status of nuclear power in the world
In those days, uranium ore was a scarce and expensive commodity and it was naturally assumed that economically available supplies would not meet the demands required by a widespread use of nuclear power. Consequently, the extraction of all the potential energy content of uranium-235 seemed to be essential. Such a complete exploitation of uranium resources requires reprocessing of the spent fuel and the extraction of plutonium for burning in specially designed “fast” reactors. The approach became more attractive with the concept of fast-breeder reactors, which could produce more fuel than they consumed. For such reasons, many countries during the 1960s attached high priority to the development of fast reactors, and it was anticipated that they would be widely deployed in the 1980s (Semenov and Oi, 1993). Until the early 1970s then, the nuclear fuel cycle was pictured as an orderly sequence of processes. It began with uranium mining, milling, and conversion, was followed by fuel enrichment, fuel fabrication and power generation and was finally completed by reprocessing, recycling of plutonium and uranium to fast reactors and final disposal of waste streams from reprocessing plants. In essence, closure of the fuel cycle meant the effective use of plutonium. Three different types of fuel cycle are commonly identified for nuclear power generation, depending on whether fuel is recycled and on the type of reactor used for electricity production. • •
•
The “once-through” fuel cycle: in this cycle, the spent fuel is not reprocessed but kept in storage until it is eventually disposed of as waste. The thermal reactor cycle: in this cycle, the spent fuel is reprocessed and the uranium and plutonium can be recycled in new fuel elements. It is also possible to recycle only the uranium and to store the plutonium, and vice versa. The fast-breeder reactor cycle: in this cycle, the spent fuel is similarly reprocessed and the uranium and plutonium fabricated into new fuel elements. However, they are recycled to fast-breeder reactors, in which there is a central core of uranium/plutonium fuel surrounded by a blanket of depleted uranium (uranium from which most of the uranium-235 atoms have been removed during the process of enrichment) or to burner reactors. This depleted uranium consists mostly of uranium-238 atoms, some of which are converted to plutonium during irradiation. By suitable operation, fast-breeder reactors thus can produce slightly more fuel than they consume, hence the name “breeder” (see Fig. 6.1).
The situation has changed dramatically during the last 20 years. No closed fuel cycle of the type originally envisaged to be operational in the 1980s exists today. Although the closure of the nuclear fuel cycle has been experimentally demonstrated in France, Japan, Russia and the United Kingdom, it has not been demonstrated yet on a commercial scale. Current thinking is divided into two schools. One believes that plutonium as an energy source has no economic value and spent fuel should be disposed of in a safe
331
332
CHAPTER 6 The nuclear fuel cycle
Interim spent fuel storage Reprocessing plant
Spent fuel Uranium fuel
Thermal reactor
Enriched UF6
Reprocessing plant Recovered uranium
235
U enrichment
Spent fuel UF6
Pu oxide
Fuel fabrication
U/Pu (core) Max. fuel
Depleted
238
U (blanket)
Spent fuel
Fast reactor
Thermal reactor
Uranium fuel
(A)
U
Plutonium oxide
Max. fuel fabrication
UF6
U3O8
238
Pu
Uranium fuel fabrication
(B) Spent fuel and waste disposal
From enrichment plant
(C) Waste disposal
Waste disposal
FIG. 6.1 Schematic representations of (A) the once-through cycle, (B) the thermal reactor cycle and (c) the fast reactor cycle. U3O8 ¼ yellowcake, UF6 ¼ uranium hexafluoride and MOX ¼ mixed oxide fuel (uranium/plutonium). After Semenov, B.A. and Oi, N., 1993. Nuclear fuel cycles: adjusting to new realities. IAEA Bulletin 3, 2.
Extraction
Enrichment
Fuel fabrication
Nuclear power plant
Reprocessing Spent fuel Mining
HLW Conditioning
Waste disposal
FIG. 6.2 A schematic representation of the nuclear fuel cycle.
way (the “once-through” option). The other essentially adheres to the traditional nuclear fuel cycle (closed cycle option as illustrated in Fig. 6.2). The difference of opinions stems mainly from the predictions of nuclear electricity growth and the predicted availability of economical supplies of uranium, although it is influenced by political and environmental issues as well. We will return to this matter again. The present situation on the use of nuclear energy for electricity production is illustrated in Table 6.1 which lists power stations
6.2 The status of nuclear power in the world
Table 6.1 Nuclear power status in the world at the end of 2016
Argentina Armenia Belarus Belgium Brazil Bulgaria Canada China Czech Republic Finland France Germany Hungary India Iran Japan Korea, Republic of Mexico Netherlands Pakistan Romania Russian Federation South Africa Slovak Republic Slovenia Spain Sweden Switzerland Taiwan, China United Kingdom
In operation
Under construction
Nuclear electricity production in 2016
No. of units
No. of units
TW-h
Total net MWE
3 1
1632 375
7 2 2 19 36 6
5913 1884 1926 13,554 31,384 3930
4 58 8 4 22 1 42 25
2764 63,130 10,799 1889 6240 915 39,752 23,077
2 1 4 2 35
1552 482 1005 1300 26,111
2 4
1860 1814
1 7 10 5 6
688 7121 9740 3333 5052
15
8918
Total net MWE
1
25
2
2218
1
1245
21 2
21,622 1824
1 1
1600 1630
5
2990
2 3
% of total
7.7 2.2
5.1 31.4
41.4 15.0 15.1 95.7 197.8 22.7
51.2 2.6 35.0 14.0 3.5 29.3
2653 4020
22.3 386.5 80.1 15.2 35.0 5.9 17.5 154.3
33.7 72.8 13.0 51.3 2.6 2.1 1.8 29.2
3 1 7
2343 650 5520
10.3 3.7 5.4 10.4 184.1
3.5 3.4 5.2 16.5 18.1
2
880
15.2 13.7
6.5 54.1
2600
5.4 56.1 60.6 20.3 30.5
35.2 21.2 40.1 34.4 12.1
65.1
20.4
2
Continued
333
334
CHAPTER 6 The nuclear fuel cycle
Table 6.1 Nuclear power status in the world at the end of 2016 Continued
Ukraine United Arab Emirates United States World total
In operation
Under construction
Nuclear electricity production in 2016
No. of units
Total net MWE
No. of units
TW-h
15
13,107
2 4
2070 5380
76.1
50.2
99
99,869
4
4468
804.9
19.6
448
391,116
61
61,264
2476.2
10.6
Total net MWE
% of total
After IAEA, 2017. Energy, Electricity and Nuclear Power Estimates for the Period up to 2050. Reference Data Series No. 1, 2017 Ed. International Atomic Energy Agency, Vienna, 2017.
Table 6.2 Nuclear power units by reactor type, worldwide, as of December 2015 Type code
Reactor type—full name
In operation
Shut down
Under construction
282
46
57
78 14
36 38
4 –
49
8
4
15
9
–
3
7
1
No. units PWR BWR GCR PHWR LWGR FBR
Pressurized light-water reactors Boiling light-water reactors Gas-cooled, graphite moderated reactors Pressurized heavy-water reactors; all types Graphite-moderated lightwater reactors Fast-breeder reactors
After IAEA, 2016. Nuclear Power Reactors in the World. Reference Data Series No. 2, 2016 Ed. International Atomic Energy Agency, Vienna, 2016.
in individual countries. This is the status as of December 2016 as reported by IAEA. All these stations can be grouped by reactor type as shown in Table 6.2. Once a year Nuclear News publishes the world list of nuclear power plants which are operable, under construction, or an order (for power 30 MWe and over). The latest world list of nuclear power plants can be found in March 2018 issue of Nuclear News. Data on nuclear plants worldwide are operable, under construction, or on order as of 31 December 2017. Information are provided on net MWe, reactor type,
6.2 The status of nuclear power in the world
reactor model, initial criticality, commercial start, reactor supplier and the maps showing the location of each plant site together with GPS coordinates. We list the world power stations, shown in Table 6.3, based on information gathered from many sources, mainly national authorities listed in the table. It is of interest to refer to the article by Blix (1997) who was director general of IAEA for many years until 1997. At that time, oil, gas and coal—the fossil fuels— provided nearly 85% of the commercial energy that the world used: close to 37% for oil, 25% for coal and more than 21% for gas, with nuclear power and hydro power providing around 7% each, and commercial renewables such as solar, wind and biomass nearly 2.5%. (Noncommercial uses of renewable energy were estimated to provide another 10% of world energy consumption.) In China, coal supplied 75% of energy consumption, oil about 17%, nuclear and hydro 5%, and gas 2%. According to Blix (1997), the so-called renewable sources total a little more than 2% of world commercial energy. The bulk of that total comes from geothermal Table 6.3 World list of nuclear power plants Country
Authority
Power station
Argentina
Commision Nacional de Energia Atomica (CNEA)
Atucha 1 (Lima, Buenos Aires) Embalse (Rio Tercero, Cordoba) Armenia 2 (Metsa, pr. Armenia) Doel 1 (Doel, East Flanders) Doel 2 (Doel, East Flanders) Doel 3 (Doel, East Flanders) Doel 4 (Doel, East Flanders) Tihange 1 (Huy, Liege) Tihange 2 (Huy, Liege) Tihange 3 (Huy, Liege) Angra 1 (Itaorna, Rio de Janeiro) Angra 2 (Itaorna, Rio de Janeiro) Kozloduy 5 (Kozloduy, Vratsa) Kozloduy 6 (Kozloduy, Vratsa)
Armenia
Ministry of nuclear power
Belgium
Electrabel
Brazil
Bulgaria
Fumas Centrais Electricas SA
National Electric Co.
Net MWe
Type
335
PHWR
600
PHWR
400
PWR
392
PWR
392
PWR
985
PWR
1001
PWR
962 1008 1015 626
PWR PWR PWR PWR
1270
PWR
953
PWR
953
PWR Continued
335
336
CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
Canada
New Brunswick Power Corp.Ontario Hydro
Point Lepreau (Bay of Fundy, N.B.) Pickering 1 (Pickering, Ont.) Pickering 4 (Pickering, Ont.) Pickering 5 (Pickering, Ont.) Pickering 6 (Pickering, Ont.) Pickering 7 (Pickering, Ont.) Pickering 8 (Pickering, Ont.) Bruce 3 (Tiverton, Ont.) Bruce 4 (Tiverton, Ont.) Bruce 5 (Tiverton, Ont.) Bruce 6 (Tiverton, Ont.) Bruce 7 (Tiverton, Ont.) Bruce 8 (Tiverton, Ont.) Darlington 1 (Newcastle Twp., Ont.) Darlington 2 (Newcastle Twp., Ont.) Darlington 3 (Newcastle Twp., Ont.) Darlington 4 (Newcastle Twp., Ont.) Point Lepreau Gentilly 2 (Becancour, Que.) Qinshan 1 (Haiyan, Zhejiang) Qinshan 2 (Haiyan, Zhejiang) Qinshan 3 (Haiyan, Zhejiang) Qinshan 4 (Haiyan, Zhejiang)
New Brunswick Power Hydro Quebec China
Ministry of Nuclear Industry
Net MWe
Type
640
PHWR
515
PHWR
515
PHWR
516
PHWR
516
PHWR
516
PHWR
516
PHWR
769 769 785 785 785 785 881
PHWR PHWR PHWR PHWR PHWR PHWR PHWR
881
PHWR
881
PHWR
881
PHWR
635 638
PHWR PHWR
300
PWR
600
PWR
600
PWR
700
PHWR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Authority
Guangdong Nuclear Power Joint Venture Co. Ltd. Lingao Nuclear Co.
Cuba
Czech Republic
Finland
Ministry of Basic Industries
Czech Power Board
Imatran Voima Oy (IVO)
Teollisuuden Voima Oy (TVO)
Power station Qinshan 5 (Haiyan, Zhejiang) Qinshan 6 (Haiyan, Zhejiang) Guangdong 1 (Shenzhen, Guangdong) Guangdong 2 (Shenzhen, Guangdong) Lingao 1 (Lingao, Guangdong) Lingao 2 (Lingao, Guangdong) Lingao 2 (Lingao, Guangdong) Juragua 1 (Cienfuegos, Cienfuegos) Juragua 2 (Cienfuegos, Cienfuegos) Dukovany 1 (Trebic, Jihomoravsky) Dukovany 2 (Trebic, Jihomoravsky) Dukovany 3 (Trebic, Jihomoravsky) Dukovany 4 (Trebic, Jihomoravsky) Temelin 1 (Temelin, Jihocesky) Temelin 2 (Temelin, Jihocesky) Loviisa 1 (Loviisa, UUsimaa) Loviisa 2 (Loviisa, UUsimaa) TVO 1 (Olkiluoto, Turku-Pori) TVO 2 (Olkiluoto, Turku-Pori)
Net MWe
Type
700
PHWR
650
PWR
944
PWR
944
PWR
985
PWR
985
PWR
1000
PWR
417
PWR
417
PWR
428
PWR
412
PWR
4471
PWR
470
PWR
963
PWR
963
PWR
488
PWR
488
PWR
840
BWR
860
BWR Continued
337
338
CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
France
Commissariat a l’Energie Atomique Centrale Nucleaire Europeene a Neutrons Rapides S.A. (NERSA) Electricite de France (EdF)
Phenix (Marcoule, Gard) Creys-Malville (Bouvesse, Isere) Chinon B1 (Chinon, Indre-et-Loire) Chinon B2 (Chinon, Indre-et-Loire) Chinon B3 (Chinon, Indre-et-Loire) Chinon B4 (Chinon, Indre-et-Loire) Saint-Laurent B1 (SaintLaurent-des-Eaux, Loir-et-Cher) Saint-Laurent B2 (SaintLaurent-des-Eaux, Loir-et-Cher) Bugey 2 (Loyettes, Ain) Bugey 3 (Loyettes, Ain) Bugey 4 (Loyettes, Ain) Bugey 5 (Loyettes, Ain) Fessenheim 1 (Fessenheim, Haut-Rhin) Fessenheim 2 (Fessenheim, Haut-Rhin) Dampierre 1 (Ouzouer, Loiret) Dampierre 2 (Ouzouer, Loiret) Dampierre 3 (Ouzouer, Loiret) Dampierre 4 (Ouzouer, Loiret) Gravelines B1 (Gravelines, Nord) Gravelines B2 (Gravelines, Nord) Gravelines B3 (Gravelines, Nord) Gravelines B4 (Gravelines, Nord) Gravelines B5 (Gravelines, Nord)
Net MWe 233 600
Type LMFBR
905 905
915
LMFBR PWR PWR PWR PWR PWR
915
PWR
910 910 880 880 880
PWR PWR PWR PWR PWR
880
PWR
890
PWR
890
PWR
890
PWR
890
PWR
915
PWR
915
PWR
915
PWR
915
PWR
915
PWR
905 905
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station Gravelines B6 (Gravelines, Nord) Tricastin 1 (Pierrelatte, Drome) Tricastin 2 (Pierrelatte, Drome) Tricastin 3 (Pierrelatte, Drome) Tricastin 4 (Pierrelatte, Drome) Blayais 1 (Blaye, Gironde) Blayais 2 (Blaye, Gironde) Blayais 3 (Blaye, Gironde) Blayais 4 (Blaye, Gironde) Paluel 1 (Veulettes, Seine-Maritime) Paluel 2 (Veulettes, Seine-Maritime) Paluel 3 (Veulettes, Seine-Maritime) Paluel 4 (Veulettes, Seine-Maritime) Cruas 1 (Cruas, Ardeche) Cruas 2 (Cruas, Ardeche) Cruas 3 (Cruas, Ardeche) Cruas 4 (Cruas, Ardeche) Saint-Alban 1 (Auberives, Isere) Saint-Alban 2 (Auberives, Isere) Flamanville 1 (Flamanville, Manche) Flamanville 2 (Flamanville, Manche) Cattenom 1 (Cattenom, Moselle) Cattenom 2 (Cattenom, Moselle)
Net MWe
Type
915
PWR
915
PWR
915
PWR
915
PWR
915
PWR
910
PWR
910
PWR
910
PWR
910
PWR
1330
PWR
1330
PWR
1330
PWR
1330
PWR
915 915 915 915 1335
PWR PWR PWR PWR PWR
1335
PWR
1330
PWR
1330
PWR
1300
PWR
1300
PWR Continued
339
340
CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Germany
Authority
Bayernwerk AG Gemeinschaftskernkraft werk Grohnde GmbH (KWG) Gemeinschaftskernkraftwerk Neckar (GKN) Kernkraftwerk Brokdorf GmbH (KBR) Kernkraftwerk Brunsbuettel GmbH (KKB)
Power station Cattenom 3 (Cattenom, Moselle) Cattenom 4 (Cattenom, Moselle) Belleville 1 (Belleville s/Loire, Cher) Belleville 2 (Belleville s/Loire, Cher) Nogent s/Seine 1 (Nogent s/Seine, Aube) Nogent s/Seine 2 (Nogent s/Seine, Aube) Penly 1 (Saint-Martinen-Campagne, SeineMaritime) Penly 1 (Saint-Martinen-Campagne, SeineMaritime) Golfech 1 (Valence, Tarn et Garonne) Golfech 2 (Valence, Tarn et Garonne) Chooz B1 (Chooz, Ardennes) Chooz B2 (Chooz, Ardennes) Civaux 1 (Civaux, Vienne) Civaux 2 (Civaux, Vienne) Grafenrheinfeld KKG (Gragenrheinfeld, Ba.) Grohnde (Emmerthal, Nied.) Neckar 1 (Neckarwestheim, B.-W.) Neckar 1 (Neckarwestheim, B.-W.) Brokdorf (Brokdorf, S.-H.) Brunsbuettel (Brunsbuettel, S.-H.)
Net MWe
Type
1300
PWR
1300
PWR
1310
PWR
1310
PWR
1310
PWR
1310
PWR
1330
PWR
1330
PWR
1310
PWR
1310
PWR
1455
PWR
1455
PWR
1450
PWR
1450
PWR
1275
PWR
1360
PWR
785
PWR
1310
PWR
1370
PWR
771
BWR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Net MWe
Type
Isar 1 (Essenbach, Ba.) Isar 2 (Essenbach, Ba.) Kruemmel (Geesthacht, S.-H.) Hochtief/Hammers/ Heitkamp/Holzmann
878 1400 1260
BWR PWR BWR
Emsland (Lingen, Nied.)
1329
PWR
Philippsburg 1 (Philippsburg, B.-W.) Philippsburg 2 (Philippsburg, B.-W.) Gundremmingen B (Gundremmingen, Ba.) Gundremmingen C (Gundremmingen, Ba.) Unterweser (Rodenkirchen, Nied.)
890
BWR
1392
PWR
1284
BWR
1288
BWR
1345
PWR
Biblis A (Biblis, Hessen)
1167
PWR
Biblis A (Biblis, Hessen) Paks 1 (Paks, Tolna) Paks 2 (Paks, Tolna) Paks 3 (Paks, Tolna) Paks 4 (Paks, Tolna) Tarapur 1 (Tarapur, Maharashtra) Tarapur 2 (Tarapur, Maharashtra) Tarapur 3 (Tarapur, Maharashtra) Tarapur 4 (Tarapur, Maharashtra) Rajasthan 1 (Kota, Rajasthan) Rajasthan 2 (Kota, Rajasthan) Rajasthan 3 (Kota, Rajasthan) Rajasthan 4 (Kota, Rajasthan) Rajasthan 5 (Kota, Rajasthan)
1240 470 473 473 473 150
PWR PWR PWR PWR PWR BWR
150
BWR
490
PWHR
490
PWHR
90
PWHR
187
PWHR
202
PWHR
202
PWHR
202
PWHR
Authority
Power station
Kernkraftwerk Isar (KKI) Kernkraftwerk
Kruemmel Gmbh (KKK) Kernkraftwerke LippeEms GmbH (KKE) Kernkraftwerk Philippsburg (KKP) Kernkraftwerk RWE-Bayemwerk GmbH (KRB) Kernkraftwerk Unterweser GmbH (KKU) RWE Energie Aktiengesellshaft Hungary
Hungarian Power Companies, Ltd.
India
Atomic Energy Commission, Department of Atomic Energy
Continued
341
342
CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Japan
Authority
Chubu Electric Power Co.
Chugoku Electric Power Co., Inc.
Hokkaido Electric Power Co.
Hokuriku Electric Power Co.
Japan Atomic Power Co. Ltd.
Power station Madras 1 (Kalpakkam, Tamil Nadu) Madras 2 (Kalpakkam, Tamil Nadu) Narora 1 (Narora, Uttar Pradesh) Narora 2 (Narora, Uttar Pradesh) Kakrapar 1 (Kakrapar, Gujarat) Kakrapar 2 (Kakrapar, Gujarat) Kaiga 1 (Kaiga, Karnataka) Kaiga 2 (Kaiga, Karnataka) Kaiga 3 (Kaiga, Karnataka) Hamaoka 3 (Hamaokacho, Shizuoka) Hamaoka 4 (Hamaokacho, Shizuoka) Hamaoka 5 (Hamaokacho, Shizuoka) Shimane 1 (Kashimacho, Shimane) Shimane 2 (Kashimacho, Shimane) Tomari 1 (Tomari-mura, Hokkaido) Tomari 2 (Tomari-mura, Hokkaido) Tomari 3 (Tomari-mura, Hokkaido) Shika 1 (Shika-machi, Ishikawa) Shika 2 (Shika-machi, Ishikawa) Tokai 2 (Tokai Mura, Ibaraki) Tsuruga 1 (Tsuruga, Fukui) Tsuruga 2 (Tsuruga, Fukui)
Net MWe
Type
155
PHWR
155
PHWR
202
PHWR
202
PHWR
202
PHWR
202
PHWR
202
PHWR
202
PHWR
202
PHWR
1056
BWR
1092
AEWR
1380
BWR
439
BWR
790
BWR
550
PWR
550
PWR
912
PWR
505
BWR
1358
AEWR
1056
BWR
341
BWR
1115
PWR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
Kansai Electric Power Co., Inc.
Mihama 1 (Mihama-cho, Fukui) Mihama 2 (Mihama-cho, Fukui) Mihama 3 (Mihama-cho, Fukui) Takahama 1 (Takahama-cho, Fukui) Takahama 2 (Takahama-cho, Fukui) Takahama 3 (Takahama-cho, Fukui) Takahama 4 (Takahama-cho, Fukui) Ohi 1 (Ohi-cho, Fukui) Ohi 2 (Ohi-cho, Fukui) Ohi 3 (Ohi-cho, Fukui) Ohi 4 (Ohi-cho, Fukui) Genkai 1 (Genkai, Saga)
Kyushu Electric Power Co., Inc.
Development Corp. (PNC) Shikoku Electric Power Co.
Tohoku Electric Power Co., Inc.
Genkai 2 (Genkai, Saga) Genkai 3 (Genkai, Saga) Genkai 4 (Genkai, Saga) Sendai 1 (Sendai, Kagoshima) Sendai 2 (Sendai, Kagoshima) Monju FBR (Tsuruga, Fukui) Ikata 1 (Ikata-cho, Ehime) Ikata 2 (Ikata-cho, Ehime) Ikata 3 (Ikata-cho, Ehime) Higashidori 1 (Higashidori, Aomori) Onagawa 1 (Onagawa, Miyagi) Onagawa 2 (Onagawa, Miyagi) Onagawa 3 (Onagawa, Miyagi)
Net MWe
Type
320
PWR
470
PWR
780
PWR
780
PWR
780
PWR
830
PWR
830
PWR
1120 1120 1127 1127 529
PWR PWR PWR PWR PWR
529 1127 1127 846
PWR PWR PWR PWR
846
PWR
280
LMFBR
538
PWR
538
PWR
846
PWR
1100
BWR
498
BWR
796
BWR
796
BWR Continued
343
344
CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Kazakhstan
Korea
Net MWe
Type
Fukushima Daiichi 1 (Ohkuma, Fukushima) Fukushima Daiichi 2 (Ohkuma, Fukushima) Fukushima Daiichi 3 (Ohkuma, Fukushima) Fukushima Daiichi 4 (Ohkuma, Fukushima) Fukushima Daiichi 5 (Ohkuma, Fukushima) Fukushima Daiichi 6 (Ohkuma, Fukushima) Fukushima Daini 1 (Naraha, Fukushima) Fukushima Daini 2 (Naraha, Fukushima) Fukushima Daini 3 (Naraha, Fukushima) Fukushima Daini 4 (Naraha, Fukushima) Kashiwazaki Kariwa 1 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 2 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 3 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 4 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 5 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 6 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 7 (Kashiwazaki, Niigata) BN-350 (Aktau, Mangyshlak)
439
BWR
760
BWR
760
BWR
760
BWR
760
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1315
BWR
1315
BWR
135
LMFBR
Kori Kori Kori Kori
563 612 903 903
PWR PWR PWR PWR
Authority
Power station
Tokyo Electric Power Co.
Kazakh State Atomic Power Engineering and Industry Corp. (KATEP) Korea Electric Power Corp.
1 (Kori, 2 (Kori, 3 (Kori, 4 (Kori,
Kyongnam) Kyongnam) Kyongnam) Kyongnam)
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Mexico
The Netherlands
Pakistan
Authority
Comision Federal de Electricidad
NV Elektriciteits Produktiemaatschappij Zuid-Nederland (NV EPZ) Pakistan Atomic Energy Commission
Net MWe
Type
Wolsong 1 (Kyongju, Kyongbuk) Wolsong 2 (Kyongju, Kyongbuk) Wolsong 3 (Kyongju, Kyongbuk) Wolsong 4 (Kyongju, Kyongbuk) Yonggwang 1 (Yonggwang, Chonnam) Yonggwang 2 (Yonggwang, Chonnam) Yonggwang 3 (Yonggwang, Chonnam) Yonggwang 4 (Yonggwang, Chonnam) Yonggwang 5 (Yonggwang, Chonnam) Yonggwang 6 (Yonggwang, Chonnam) Ulchin 1 (Ulchin, Kyongbuk) Ulchin 2 (Ulchin, Kyongbuk) Ulchin 3 (Ulchin, Kyongbuk) Ulchin 4 (Ulchin, Kyongbuk) Ulchin 5 (Ulchin, Kyongbuk) Ulchin 6 (Ulchin, Kyongbuk) Laguna Verde 1 (Laguna Verde, Veracruz) Laguna Verde 2 (Laguna Verde, Veracruz) Borssele (Borssele, Zeeland)
629
PHWR
700
PHWR
700
PHWR
700
PHWR
900
PWR
900
PWR
950
PWR
950
PWR
950
PWR
950
PWR
920
PWR
920
PWR
950
PWR
950
PWR
950
PWR
950
PWR
800
BWR
800
BWR
452
PWR
Kanupp-1 (Karachi, Sind) Chasnupp (Mianwali, Punjab)
125
PHWR
300
PWR
Power station
Continued
345
346
CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
Romania
Romanian Electricity
Cernavoda 1 (Cernavoda, Constanta) Cernavoda 2 (Cernavoda, Constanta) Balakovo 1 (Balakovo, Saratov) Balakovo 2 (Balakovo, Saratov) Balakovo 3 (Balakovo, Saratov) Balakovo 4 (Balakovo, Saratov) Beloyarskiy 3 (BN-600) (Zarechnyy, Sverdlovsk) Bilibino-1 (Chukotka Autonomous Okrug) Bilibino-2 (Chukotka Autonomous Okrug) Bilibino-3 (Chukotka Autonomous Okrug) Bilibino-4 (Chukotka Autonomous Okrug) Kalinin 1 (Udomlya, Tver) Kalinin 2 (Udomlya, Tver) Kalinin 3 (Udomlya, Tver) Kola 1 (Polyarnyye Zori, Murmansk) Kola 2 (Polyarnyye Zori, Murmansk) Kola 3 (Polyarnyye Zori, Murmansk) Kola 4 (Polyarnyye Zori, Murmansk) Kursk 1 (Kurchatov, Kursk) Kursk 2 (Kurchatov, Kursk) Kursk 3 (Kurchatov, Kursk) Kursk 4 (Kurchatov, Kursk)
Authority (RENEL) Russia
Ministry of Atomic Power
Net MWe
Type
655
PHWR
650
PHWR
950
PWR
950
PWR
950
PWR
950
PWR
560
LMFBR
12
LWGR
12
LWGR
12
LWGR
12
LWGR
950 950 950 411
PWR PWR PWR PWR
411
PWR
411
PWR
411
PWR
925
LGR
925
LGR
925
LGR
925
LGR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Slovakia
Slovenia
Authority
Slovak Power Board
Nuklearna Elektrana Krsko
Power station Leningrad 1 (Sosnovyy Bor, St. Petersburg) Leningrad 2 (Sosnovyy Bor, St. Petersburg) Leningrad 3 (Sosnovyy Bor, St. Petersburg) Leningrad 4 (Sosnovyy Bor, St. Petersburg) Novovoronezhskiy 3 (Novovoronezhskiy, Voronezh) Novovoronezhskiy 4 (Novovoronezhskiy, Voronezh) Novovoronezhskiy 5 (Novovoronezhskiy, Voronezh) Smolensk 1 (Desnogorsk, Smolensk) Rostov-1 (Volgodonsk-1) Rostov-2 (Volgodonsk-2) Smolensk 2 (Desnogorsk, Smolensk) Smolensk 3 (Desnogorsk, Smolensk) South Urals 1 (Chelyabinsk, Chelyabinsk) South Urals 2 (Chelyabinsk, Chelyabinsk) Vk-50 (Dimitrovgrad, Ulyanovsk, RSFSR) Bohunice 3 (Trnava, Zapadoslovensky) Bohunice 4 (Trnava, Zapadoslovensky) Mochovce 1 (Mochovce, Zapadoslovensky) Mochovce 2 (Mochovce, Zapadoslovensky) Krsko (Krsko, Vrbina)
Net MWe
Type
925
LGR
925
LGR
925
LGR
925
LGR
385
PWR
385
PWR
950
PWR
925
LGR
950 950 925
PWR PWR LGR
925
LGR
750
LMFBR
750
LMFBR
50
PWR
408
PWR
408
PWR
420
PWR
420
PWR
676
PWR Continued
347
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
South Africa
ESKOM
Koeberg 1 (Melkbosstrand, Cape) Koeberg 2 (Melkbosstrand, Cape) Asco 1 (Asco, Tarragona) Asco 2 (Asco, Tarragona) Vandellos 2 (Vandellos, Tarragona)
Spain
Asociacion Nuclear Asco Endesa—Iberdrola S.A. Central Nuclear Vandellos II, A.I.E. Central de Trillo Central Nuclear de Almaraz Centrales Nucleares del Norte, SA Iberdrola SA
Sweden
OKG Aktiebolag
Vattenfall
Sydkraft AB
Trillo 1 (Trillo, Guadalajara) Almaraz 1 (Almaraz, Caceres) Almaraz 2 (Almaraz, Caceres) Santa Maria de Garona (Santa Maria de Garona, Burgos) Cofrentes (Cofrentes, Valencia) Oskarshamn 1 (Oskarshamn, Kalmar) Oskarshamn 2 (Oskarshamn, Kalmar) Oskarshamn 3 (Oskarshamn, Kalmar) Ringhals 1 (Varberg, Halland) Ringhals 2 (Varberg, Halland) Ringhals 3 (Varberg, Halland) Ringhals 4 (Varberg, Halland) Forsmark 1 (Forsmark, Uppsala) Forsmark 2 (Forsmark, Uppsala) Forsmark 3 (Forsmark, Uppsala)
Net MWe
Type
920
PWR
920
PWR
996
PWR
992
PWR
1045
PWR
1003
PWR
941
PWR
950
PWR
446
BWR
1063
BWR
445
BWR
605
BWR
1400
BWR
835
BWR
875
PWR
915
PWR
915
PWR
970
PWR
970
BWR
1155
BWR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
Switzerland
Bernische Kraftwerke AG (BKW) Kernkraftwerk GoesgenDaeniken AG Kernkraftwerk Leibstadt AG Nordostschweizerische Kraftwerk AG (NOK)
Muehleberg (Muehleberg, Bern) Goesgen (Daeniken, Solothurn) Leibstadt (Leibstadt, Aargau) Beznau-1 (Doettingen, Aargau) Beznau-2 (Doettingen, Aargau) Chinshan 1 (Chinshan, Taipei) Chinshan 2 (Chinshan, Taipei) Kuosheng 1 (Kuosheng, Wang-Li, Taipei) Kuosheng 2 (Kuosheng, Wang-Li, Taipei) Maanshan 1 (Herng Chuen) Maanshan 2 (Herng Chuen) Khmel’nitskiy 1 (Neteshin, Khmel’nitskiy) Khmel’nitskiy 2 (Neteshin, Khmel’nitskiyn Rovno 1 (Kuznetsovsk, Rovno) Rovno 2 (Kuznetsovsk, Rovno) Rovno 3 (Kuznetsovsk, Rovno) Rovno 4 (Kuznetsovsk, Rovno) South Ukraine 1 (Konstantinovka, Nikolaev) South Ukraine 2 (Konstantinovka, Nikolaev) South Ukraine 3 (Konstantinovka, Nikolaev)
Taiwan, China
Ukraine
Taiwan Power Co.
Energoatom
Net MWe
Type
355
BWR
970
PWR
1030
BWR
365
PWR
365
PWR
604
BWR
604
BWR
948
BWR
948
BWR
890
PWR
890
PWR
950
PWR
950
PWR
402
PWR
416
PWR
950
PWR
950
PWR
950
PWR
950
PWR
950
PWR
Continued
349
350
CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
United Kingdom
Authority
British Energy plc— Nuclear Electric plc
British Energy plc— Scottish Nuclear Ltd.
Power station Zaporozhye 1 (Energodar, Zaporozhye) Zaporozhye 2 (Energodar, Zaporozhye) Zaporozhye 3 (Energodar, Zaporozhye) Zaporozhye 4 (Energodar, Zaporozhye) Zaporozhye 5 (Energodar, Zaporozhye) Zaporozhye 6 (Energodar, Zaporozhye) Dungeness B1 (Lydd, Kent) Dungeness B2 (Lydd, Kent) Sizewell B (Sizewell, Suffolk) Hinkley Point B1 (Hinkley Point, Somerset) Hinkley Point B2 (Hinkley Point, Somerset) Hartlepool 1 (Hartlepool, Cleveland) Hartlepool 2 (Hartlepool, Cleveland) Heysham A1 (Heysham, Lancashire) Heysham A2 (Heysham, Lancashire) Heysham B1 (Heysham, Lancashire) Heysham B2 (Heysham, Lancashire) Hunterston B1 (Ayrshire, Strathclyde) Hunterston B2 (Ayrshire, Strathclyde) Torness unit A (Dunbar, East Lothian) Torness unit B (Dunbar, East Lothian)
Net MWe
Type
950
PWR
950
PWR
950
PWR
950
PWR
950
PWR
950
PWR
555
AGR
555
AGR
1188
PWR
585
AGR
585
AGR
575
AGR
575
AGR
550
AGR
550
AGR
625
AGR
625
AGR
575
AGR
575
AGR
625
AGR
625
AGR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
United States
Authority
Power station
Nuclear Decommissioning Authority
Wylfa-1 Wylfa-2 Oldbury-1 Oldbury-2 Palo Verde 1 (Wintersburg, AZ) Palo Verde 2 (Wintersburg, AZ) Palo Verde 3 (Wintersburg, AZ) Calvert Cliffs 1 (Lusby, MD) Calvert Cliffs 1 (Lusby, MD) Pilgrim (Plymouth, MA) Brunswick 1 (Southport, NC) Brunswick 2 (Southport, NC) Robinson 2 (Hartsville, SC) Perry 1 (North Perry, OH)
Arizona Public Service Co.
Baltimore Gas & Electric Co. Boston Edison Co. Carolina Power & Light Co.
The Cleveland Electric Illuminating Co. Commonwealth Edison Co.
Consolidated Edison Co. Consumers Energy Co.
Braidwood 1 (Braidwood, IL) Braidwood 2 (Braidwood, IL) Byron 1 (Byron, IL) Byron 1 (Byron, IL) Dresden 2 (Morris, IL) Dresden 3 (Morris, IL) LaSalle County 1 (Seneca, IL) LaSalle County 2 (Seneca, IL) Quad Cities 1 (Cordova, IL) Quad Cities 2 (Cordova, IL) Indian Point 2 (Buchana, NY) Palisades (South Haven, MI)
Net MWe
Type
490 490 217 217 1270
GCR GCR GCR GCR PWR
1270
PWR
1270
PWR
825
PWR
825
PWR
670 767
BWR BWR
754
BWR
683
PWR
1205
BWR
1120
PWR
1120
PWR
1105 1105 912 794 1078
PWR PWR BWR BWR BWR
1078
BWR
789
BWR
789
BWR
975
PWR
781
PWR Continued
351
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Authority
Power station
Detroit Edison Co.
Enrico Fermi 2 (Newport, MI) Catawba 1 (Clover, SC) Catawba 2 (Clover, SC) McGuire 1 (Cornelius, NC) McGuire 2 (Cornelius, NC) Oconee 1 (Seneca, SC) Oconee 2 (Seneca, SC) Oconee 3 (Seneca, SC) Beaver Valley 1 (Shippingport, PA) Beaver Valley 2 (Shippingport, PA) Arkansas Nuclear One-1 (Russellville, AR) Arkansas Nuclear One-2 (Russellville, AR) Grand Gulf (Port Gibson, MS) Waterford 3 (Taft, LA) River Bend (St. Francisville, LA) St. Lucie 1 (Hutchinson Island, FL) St. Lucie 2 (Hutchinson Island, FL) Turkey Point 3 (Florida City, FL) Turkey Point 4 (Florida City, FL) Crystal River 3 (Red Level, FL) Edwin I. Hatch 1 (Baxley, GA) Edwin I. Hatch 2 (Baxley, GA) Alvin W. Vogtle 1 (Waynesboro, GA) Alvin W. Vogtle 2 (Waynesboro, GA)
Duke Power Co.
Duquesne Light Co.
Entergy Operations, Inc.
Florida Power and Light Co.
Florida Nuclear Corp. Georgia Power Company
Net MWe
Type
1139
BWR
1129 1129 1129
PWR PWR PWR
1129
PWR
846 846 846 810
PWR PWR PWR PWR
833
PWR
836
PWR
8965
PWR
1173
BWR
1075 936
PWR BWR
839
PWR
839
PWR
666
PWR
666
PWR
868
PWR
810
BWR
820
BWR
1162
PWR
1162
PWR
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Net MWe
Type
Oyster Creek (Forked River, NJ) Three Mile Island 1 (Londonderry Twp., PA) South Texas Project 1 (Palacois, TX) South Texas Project 2 (Palacois, TX) Clinton (Clinton, IL) Donald C. Cook 1 (Bridgman, MI) Donald C. Cook 2 (Bridgman, MI) Duane Arnold (Palo, IA) Cooper (Brownville, NE)
619
BWR
786
PWR
1250
PWR
1250
PWR
930 1020
BWR PWR
11,080
PWR
538 764
BWR BWR
James A. Fitzpatrick (Scriba, NY) Indian Point 3 (Buchanan, NY) Nine Mile Point 1 (Scriba, NY) Nine Mile Point 2 (Scriba, NY) Seabrook (Seabrook, NH) Millstone 2 (Waterford, CT) Millstone 3 (Waterford, CT) Monticello (Monticello, MN) Prairie Island 1 (Red Wing, MN) Prairie Island 1 (Red Wing, MN) Fort Calhoun (NE)
780
BWR
965
PWR
610
BWR
1080
BWR
1150
PWR
875
PWR
1149
PWR
536
BWR
503
PWR
500
PWR
478
PWR
Diablo Canyon 1 (Avila Beach, CA) Diablo Canyon 2 (Avila Beach, CA)
1130
PWR
1160
PWR
Authority
Power station
GPU Nuclear Corp.
Houston Lighting & Power Co.
Illinois Power Co. Indiana/Michigan Power Co. IES Utilities, Ine. Nebraska Public Power District New York Power Authority Niagara Mohawk Power Corp.
North Atlantic Energy Service Corp. Northeast Utilities
Northern States Power Co.
Omaha Public Power District Pacific Gas & Electric Co.
Continued
353
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CHAPTER 6 The nuclear fuel cycle
Table 6.3 World list of nuclear power plants Continued Country
Net MWe
Type
Susquehanna 1 (Berwick, PA) Susquehanna 2 (Berwick, PA) Limerick 1 (Pottstown, PA) Limerick 2 (Pottstown, PA) Peach Bottom 2 (Delta, PA) Peach Bottom 3 (Delta, PA) Hope Creek (Salem, NJ) Salem 1 (Salem, NJ) Salem 2 (Salem, NJ) R.E. Ginna (Ontario, NY)
860
PWR
860
PWR
1055
BWR
1055
BWR
1159
BWR
1035
BWR
1031 1106 1106 470
BWR PWR PWR PWR
Virgil C. Summer (Parr, SC) San Onofre 2 (San Clemente, CA) San Onofre 3 (San Clemente, CA) Joseph M. Farley 1 (Dothan, AL) Joseph M. Farley 2 (Dothan, AL)
885
PWR
1070
PWR
1080
PWR
860
PWR
860
PWR
Browns Ferry 1 (Decatur, AL) Browns Ferry 2 (Decatur, AL) Browns Ferry 3 (Decatur, AL) Sequoyah 1 (SoddyDaisy, TN) Sequoyah 2 (SoddyDaisy, TN) Watts Bar 1 (Spring City, TN)
1065
BWR
1118
BWR
1118
BWR
1148
PWR
1148
PWR
1177
PWR
Authority
Power station
Pennsylvania Power & Light Co.
PECO Energy Co.
Public Service Electric & Gas Co. Rochester Gas & Electric Corp. South Carolina Electric & Gas Co. Southern California Edison Co. and San Diego Gas & Electric Co. Southern Nuclear Operating Co.
Tennessee Valley Authority
6.2 The status of nuclear power in the world
Table 6.3 World list of nuclear power plants Continued Country
Net MWe
Type
Comanche Peak 1 (Glen Rose, TX) Comanche Peak 2 (Glen Rose, TX) Davis-Besse (Oak Harbor, OH) Callaway (Fulton, MO) Vermont Yankee (Vermont, VT) North Anna 1 (Mineral, VA) North Anna 2 (Mineral, VA) Surry 1 (Gravel Neck, VA) Surry 2 (Gravel Neck, VA) WNP-2 (Richland, WA)
1150
PWR
1150
PWR
877
PWR
1171 504
PWR BWR
893
PWR
897
PWR
801 801 1157
PWR PWR BWR
Point Beach 1 (Two Rivers, WI) Point Beach 2 (Two Rivers, WI) Kewaunee (Carlton, WI)
485
PWR
485
PWR
503
PWR
Wolf Creek (Burlington, KS)
1160
PWR
Authority
Power station
Texas Utilities Electric Co.
Toledo Edison Co. Union Electric Co. Vermont Yankee Nuclear Power Corp. Virginia Power
Washington Public Power Supply System Wisconsin Electric Power Co.
Wisconsin Public Service Corp. Wolf Creek Nuclear Operating Corp.
installations, new wind and solar technologies and biomass plantations. This share could increase, but only to a limited extent. The estimate made by the WEC for new renewable supplies in the medium term is that with adequate support, the share of new renewable energy supplies, currently only 2%, could reach 5–8% of increased world energy supply by 2020. The argument is related to the energy density which is so variable. For example: • • • • •
1 kg 1 kg 1 kg 1 kg 1 kg
firewood produces about 1 kWh of electricity, of coal produces about 3 kWh of electricity, of oil produces about 4 kWh of electricity, of natural uranium produces about 50,000 kWh of electricity and of plutonium produces about 6,000,000 kWh.
The low energy density of the renewable sources means that if you want significant amounts of energy (electricity) from them, you must “harvest” them over large
355
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CHAPTER 6 The nuclear fuel cycle
areas—and this is expensive. It has been calculated that to achieve the electricity generating capacity of a 1000-MWe power plant, an area of 50–60 km2 would be needed to install solar cells or windmills, or an area of 3000–5000 km2 to grow the necessary biomass. It will not be easy—or cheap—to acquire such large areas, particularly in densely populated areas where the energy will be most needed (Blix (1997)). Globally, electricity accounted for about 18% of the total final energy consumption in 2016. About 70% of the final energy consumption was in the form of fossil fuels. Bioenergy and waste accounted for 12%. As electricity consumption is expected to increase faster than total final energy consumption in the coming years, the share of electricity consumption is expected to rise. Electricity consumption will grow at a higher rate of about 2.5% per year up to 2030 and around 2% per year thereafter. The share of electricity in total final energy consumption will thus increase from 17.8% in 2016 to 21% by 2030 and to 26.6% by the middle of the century. Total electricity production grew by 2.6% in 2016 while the growth in nuclear electricity production was 2.1%. Among the various sources for electricity production, coal remained dominant despite the significant growth of natural gas-based generation. The share of electricity production from natural gas increased by 0.4% points to reach 22.8% of total electricity production. The contribution of hydropower and renewable energy sources continued to increase significantly, reaching 24.8% in 2016, while the share of nuclear electricity production remained at about 11% of the total electricity production (IAEA, 2017a). Reference Data Series No. 1 (RDS-1) is an annual publication of International Atomic Energy Commission. In 2017, its 37th edition was published containing estimates of energy, electricity and nuclear power trends up to 2050. Compared with the 2016 projections to 2030, the 2017 projections were reduced by 45 GW(e) by 2030 in both the high and the low case. These reductions reflect responses to the Fukushima Daiichi accident and other factors noted earlier. There are increasing uncertainties in these projections owing to the considerable number of reactors scheduled to be retired in some regions around 2030 and beyond (IAEA, 2017a,b). Nuclear power reactors in the world are an annual publication that presents the most recent data pertaining to reactor units in IAEA Member States. The 2018 Edition (IAEA, 2018b) provides a detailed comparison of various statistics up to and including 31.12.2017. The tables and figures contain the following information: (i) General statistics on nuclear reactors in IAEA Member States, (ii) technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned, (iii) performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in publication (IAEA, 2018b) are product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. This is also published as book and CD-ROM (IAEA, 2018c) which is the 49th edition of the IAEA’s series of annual reports on operating experience with nuclear power plants in Member States.
6.3 Nuclear safety
6.3 Nuclear safety Energy production as well as other human activities is always connected with risk taking. Radiation, and everything related to it, generates a fear not easily understood. This probably comes from the “invisibility” of the danger and relation to the bomb. Therefore the safety of nuclear power must be compared with the safety of alternative ways of generating electricity. The largest accidents in terms of casualties in the energy field are connected with the collapse of hydro dams. Some 2500 people perished, for example, in a single dam failure in Machu, India. There are also, as we know, severe accidents connected with the transport and storage of gas, the mining of coal and the shipping of oil. A gas pipeline explosion in Guadalajara in Mexico killed 200 people in 1992. Although one knows that the risk of incidents and accidents is not zero for any form of energy generation, including nuclear, one needs to be aware that most events are not very damaging. To help the nuclear power industry clarify the magnitude of events, the IAEA introduced the International Nuclear Event Scale (INES), which grades accidents from 1 to 7—much as seismologists grade earthquakes. It is hoped that this scale will help the media and public to realize that most incidents are of very minor significance and result in no threat to public health. It should also be remembered that most evolving technologies, whether boilers during the 19th century, airplanes in this century, or nuclear plants, entail some accidents from which lessons are learned. Both the Three Mile Island accident, from which only limited radioactivity escaped to the environment, and the Chernobyl disaster, have led to the introduction of new safety features in nuclear reactors, in plant operating procedures, and in regulations. The development of nuclear and radiation safety standards is a statutory responsibility of the International Atomic Energy Agency, IAEA, in Vienna, Austria. The IAEA Statute authorizes the Agency to establish standards of safety and to provide for the applications of these standards. Until now IAEA has developed and issued more than 200 standards of safety in the Agency’s Safety Series publications. They cover the fields of nuclear safety and radiation safety, including radioactive waste safety and radioactive material transport safety. The IAEA publications on this matter can be grouped into five categories: general safety, nuclear safety, radiation safety, waste safety and transport safety. Some of the most important publications are listed in the references under “safety series”. In recent years, legally binding international conventions have come to play a crucial role in improving nuclear, radiation and waste safety. The major international conventions related to safety that have been negotiated and adopted under the auspices of the IAEA are listed in Table 6.4 (from Flakus and Johnson, 1998). (1)
The Convention of Early Notification of a Nuclear Accident and the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency deal with aspects of emergency response and preparedness. Both of these Conventions—briefly referred to as the “Notification Convention”
357
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CHAPTER 6 The nuclear fuel cycle
Table 6.4 The global legal framework for nuclear, radiation and waste safety Entry into force Convention on the Physical Protection of Nuclear Material Convention on Early Notification of a Nuclear Accident
8 Feb. 1987
Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency Convention on Nuclear Safety
26 Feb. 1987
27 Oct. 1986
24 Oct. 1996
Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management
Vienna Convention on Civil Liability for Nuclear Damage
Protocol to Amend the Vienna Convention and Convention on Supplementary Compensation for Nuclear Damage
12 Nov. 1977
Development & Status In 1997, two States (Cuba and Lebanon) acceded to the Convention. As of May 1998, the Convention had 60 Parties In 1997, four States (Lebanon, Philippines, Myanmar and Singapore) agreed to be bound by the Convention. As of May 1998, the Convention had 80 Parties In 1997, three States (Lebanon, Philippines and Singapore) agreed to be bound by the Convention. As of May 1998, the Convention had 75 Parties In 1997, 10 States (Argentina, Austria, Belgium, Brazil, Germany, Greece, Luxembourg, Pakistan, Peru and Singapore) and in 1998 four States (Italy, Republic of Moldova, Portugal and Ukraine) agreed to be bound by the Convention. As of May 1998, the Convention had 46 Parties A Diplomatic Conference, held in Vienna in September 1997, adopted the Joint Convention which was opened for signature on 29 September 1997. As of 4 June 1998, the Convention had been signed by 33 States and ratified by three States (Canada, Hungary, Norway) In 1997, one State (Lebanon) ratified the Convention, and two States (Belarus, Israel) signed the Convention. The Convention had 29 Parties Both of these legal instruments were adopted on 12 September 1997 and opened for signature on 29 September 1997. As of 18 June 1998, the Protocol had been signed by 13 States (Argentina, Czech Republic, Hungary, Indonesia, Italy, Lebanon, Lithuania, Morocco, Peru, Philippines, Poland, Romania and Ukraine); and the Convention on Supplementary Compensation for Nuclear Damage had been signed by 13 States (Argentina, Australia, Czech Republic, Indonesia, Italy, Lebanon, Lithuania, Morocco, Peru, Philippines, Romania, Ukraine and United States)
After Flakus, F.-N. and Johnson, L.D., 1998. Binding agreements for nuclear safety: the global legal framework. IAEA Bulletin 40, 21
6.3 Nuclear safety
(2)
(3)
(4)
(5)
and the “Assistance Convention”—were adopted within a very short time span of only 5 months after the Chernobyl accident in 1986. The Notification Convention applies in the event of any accident involving facilities or activities of a State Party, or those under its jurisdiction or control, from which a release of radioactive material occurs or is likely to occur, and which has resulted or may result in an international transboundary release that could be of radiological safety significance for another State. A State Party involved in an accident covered by the Convention is obliged to immediately notify, directly or through the IAEA, those States which are or may be physically affected. To perform its functions under this Convention, the IAEA set up, at its headquarters in Vienna, an Emergency Response Center (ERC) for receiving, collating and rapidly transmitting relevant information. Close cooperation with the World Meteorological Organization (WMO) resulted in the use of WMO’s Global Telecommunication System (GTS) for rapid simultaneous transmission of voluminous meteorological and radiological data to national contact points. The Convention on Nuclear Safety was developed during the period 1992–94. It applies to land-based civil nuclear power plants and is the first international legal instrument that directly addresses the issue of safety of such plants. The Convention contains obligations for State Parties to take national measures with respect to safety matters—such as the legislative and regulatory framework, assessment and verification of safety, emergency preparedness and operation of nuclear power plants—and to report on the measures taken to implement each of the obligations under the Convention. The Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management was adopted at a Diplomatic Conference in Vienna in September 1997. The Convention is focused predominantly on specific activities rather than on substances. It applies with certain restrictions to: (i) the safety of spent fuel management, (ii) the safety of radioactive waste management, (iii) the safety of management of spent fuel or radioactive waste resulting from military or defence programmes if and when such materials are transferred permanently to and managed within exclusively civilian programmes. The Convention on the Physical Protection of Nuclear Material came into force in 1987. This Convention prescribes the levels at which nuclear material used for peaceful purposes is to be protected while in international nuclear transport, and requires each party to the Convention not to permit the export or import of such material unless it is satisfied that the nuclear material will be protected at those levels. At a Diplomatic Conference in September 1997, delegates from 80 States adopted the Protocol to Amend the 1963 Vienna Convention on Civil Liability for Nuclear Damage and also the Convention on Supplementary Compensation for Nuclear Damage. The Protocol sets the possible limit of the operator’s liability at an amount roughly equivalent to US $400 million and also contains an enhanced definition of nuclear damage which covers costs of reinstatement of any
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(6)
damaged environment and costs of preventive measures, extends the geographical scope of the Vienna Convention and extends the period during which claims may be made for loss of life and personal injury. In February 2015, the Vienna Declaration on Nuclear Safety was reached by contract parties of Convention on Nuclear Safety. It requests that a new NPPs are to be built with the objectives of preventing accidents and, should an accident occur, avoiding possible large releases of radionuclides to the environment. It also requests that safety of existing nuclear power plants is to be assessed and improved as appropriate.
IAEA holds a series of conferences on effective nuclear regulatory systems. These conferences bring together senior regulators in the areas of nuclear and radiation safety as well as nuclear security from around the world to discuss how to improve regulatory effectiveness to insure the protection of the public and the environment. The last conference in the series was held in Vienna, Austria 11–15 April 2018 (IAEA, 2017a,b). In order to insure the safe operation of nuclear power plants IAEA has published in 2018 following texts among others: commissioning guidelines for nuclear power plants, IAEA (2018d), maintenance optimization programme for nuclear power plants, IAEA (2018e), computer security of instrumentation and control systems at nuclear facilities, IAEA IAEA (2018f) and physical protection of nuclear material and nuclear facilities (implementation of INFCIRC/225/Revision 5), IAEA (2018g).
6.4 Releases of effluents Radioactive materials released to the environment are sources of exposure and potentially harmful. Such releases may be from different activities in the nuclear fuel cycle, mining operations or industrial users. Strict control measures must be employed to keep the resulting doses “as low as reasonably achievable”. This implies the implementation of protective and control measures and includes the setting of limits for radiation exposure. A limit is a value that must not be exceeded and the primary dose limits for individuals are set by the ICRP. These limits are related to individuals irrespective of the source. If an individual is likely to be exposed to other sources of radiation, sourcerelated limits must be set by a regulatory authority. These limits must be lower than the dose limit and are called the source upper bound. Authorized limits are limits specified by the regulating authority for a specific practice or source. In setting limits the authority must consider the requirements of radiation protection and individual dose limitation. The authorized limits will not exceed the upper bound. For practical reasons limits for releases of radioactive effluents to the environment are expressed as limits of releases over a specified period.
6.4 Releases of effluents
It is important to set reference levels for all activities. A reference level is not a limit but indicates a course of action like recording data, investigation or intervention. These levels are determined by radiation protection factors and the extent of the measures taken must be described in the operating procedures. In the case of new practices where reassessment may result in lower or higher release rates being acceptable, setting authorized limits can be difficult. There is often justification for a specific source or practice to be exempted from normal regulations. The regulating authority may exempt such sources or practices on the basis that the individual and collective doses are so low that they may be ignored. The individual dose limit is the starting point for calculating the upper bound. The dose upper bound will be less by the dose contributed by global and regional sources of exposure. The regulating authority may reserve a margin for future development of the activity or practice. This margin is set by specifying that a fraction, F, of the primary dose limit must not be exceeded. The maximum annual dose limit to the critical group is limited by Hlocal + Hregional + Hglobal F Hlimit
(6.1)
where Hlimit is the primary dose limit and the suffixes refer to the components of the total dose to the critical group. The source specific dose upper bound (HUB) for all the controlled sources of exposure is given by HUB ¼ F Hlimit Hregional Hglobal
(6.2)
The upper bound for annual release can then be derived from the dose upper bound by using the overall transfer factors (fj0 kl) where j represents population group, k represents release mode and l represents the radionuclide. If the dose commitment to the critical group j0 per unit release of a radionuclide is given by fj0 kl then the release upper bound, Rkl, is given by Rkl ¼
HUB fj0 kl
(6.3)
provided that no other radionuclides are released. In practice the situation will be more complex because more than one nuclide may be released and different modes of release will be developed. The total dose contribution to each population group due to a release Rkl is given by Jjkl ¼ fjkl Rkl :
(6.4)
If different release modes (k) are developed, the release upper bounds (Rkl) for each release mode is given by X
fj0 kl Rkl HUB
(6.5)
k
This is true only for the release of one radionuclide. If a mixture of nuclides is released that contributes to the exposure of group j0 , the release upper bound, Rkl, must satisfy the condition:
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XX k
fj0 kl Rkl HUB
(6.6)
l
The condition defines a set of values that constitute the release upper bound. For routine releases of radioactivity two main control options are considered: 1. storage of effluents to allow short-lived radionuclides to decay before release and 2. treatment of effluents to remove radionuclides before release. Within these categories a number of options may be available. The various possibilities must be identified and investigated. Considerations like operating and maintenance cost, the implications for the waste management programme as well as the individual and collective dose for the workers and the public must be taken into account. The first step in optimizing is to ensure that the releases anticipated with the control options meet the requirements of the source upperbound. Any control option that does not meet this requirement is not considered. The final element in implementing a system of dose limitation is to optimize radiological protection by selecting the control option for which radiation doses are “as low as reasonably achievable”. For the monitoring of effluent releases the samples collected in the vicinity of nuclear installations must be representative of land and water utilization as well as meteorological factors. Samples must be analysed for those nuclides which contribute most to public exposure. Air sampling is of special interest. Usually fixed monitoring instruments are used for continuous routine monitoring in the vicinity of the installation. If a limit has been exceeded the cause must be traced and corrective measures must be taken immediately. Two types of sampling monitors are in general use: air samples are used to assess the airborne contamination levels at selected points. In the case of particulate materials a volume of air is drawn through a filter paper on which the particulates are deposited. An alarm may be set on increase of activity. For gaseous materials carbon cartridges are used to trap the contaminating materials. Special devices are used for trapping iodine. “Stack” monitors for gaseous effluents give a rough estimate of the radioactivity of the effluent from a stack. The radioactive content of the samples can be assessed by using standard counting equipment. Sample measuring instruments are operated in contamination-free laboratories. For the monitoring of the released liquid effluents the following methods are used. Samples of effluents are collected by simple dipping devices and analysed before release. In the case of monitoring streams in the neighbourhood of installations, automatic samplers collecting samples over a 24-h period are used. Samples are analysed and records must be kept of results. The water effluent metre monitors water or coolants and may be connected to a rate metre, recorder or alarm system. On-site and off-site environmental monitoring at and near nuclear power plants, nuclear reactors and other fuel cycle activities are shown in Tables 6.5 and 6.6.
6.4 Releases of effluents
Table 6.5 On-site monitoring at nuclear power plants, nuclear reactors and uranium mill and/or fuel cycle facilities Sample type
Collection frequency
Analysis frequency
Airborne particulates Liquid effluents Drinking water
Continuous
Continuous readout
Continuous
Continuous readout
Semi-continuous (samples taken 3–6 h) Monthly
Quarterly on composites
Surface water Noble gases Groundwater
Continuous Quarterly
Quarterly on composites Continuous readout γ-Spectrometry on each batch sample; annual composite on other nuclides detected
Table 6.6 Off-site environmental monitoring near nuclear power plants, nuclear reactors and/or uranium mill and/or fuel cycle facilities Sample type
Collection frequency
Analysis frequency
Drinking water
Semicontinuous composite
γ-Spectrometry on each batch sample; annual composite on other nuclides detected
Milk
Weekly or biweekly at farms; monthly at dairy, γspectrometry weekly At harvest
Food crops Fish Shellfish Sediments
During fishing season or semiannually Semiannually Semiannually
γ-Spectrometry γ-Spectrometry γ-Spectrometry γ-Spectrometry
By way of illustration, we mention the case of the Sellafield reprocessing plant in Cumbria, UK, as discussed by Jones et al. (1995). For the last decade, the existence of a higher than average rate of childhood leukaemia in young people from the village of Seascale in Cumbria has led to speculation that radioactive discharges from the reprocessing plant at Sellafield may be a causative factor, even though the calculated doses are too small for the leukaemia risk observed in epidemiological studies (Stather et al., 1984, 1986). These estimates of historical doses from discharge from the plant have relied on calculations based on recorded discharges and conventional environmental models. This has left open the question of whether the recorded
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discharges, particularly in the earlier years of plant operation, might have been seriously underestimated. A major reassessment of historical discharges and doses has been carried out, prompted in large part by civil litigation instigated by a number of local families against British Nuclear Fuels plc, the operators of the Sellafield plant. The reassessment involved the development of the Sellafield Environmental Assessment Model (SEAM), which was used both to calculate doses and to build confidence in the discharge chronology from recorded measurements of environmental concentrations and current assessments of environmental inventories. The SEAM model has put together established models of atmospheric dispersion and deposition, terrestrial food chains, marine dispersion and concentration in marine biota and the sea-to-land transfer of radionuclides. Environmental measurements from a wide variety of sources have been compared against values calculated from the discharge chronology and the SEAM model in order both to validate the model and to build confidence in the discharge chronology. The established chronology for liquid discharges to the Irish Sea was confirmed by validation against historic environmental monitoring data and dated sediment cores (Kershaw et al., 1990). It was possible to establish good agreement, thus building confidence in the discharge chronology. The review of atmospheric discharges has indicated that earlier figures for emissions, particularly for particulates, in the earlier years of plant operation were significantly underestimated. Further, improved estimates were made of unmonitored emissions of plutonium from the site, which were known to have occurred in the 1950s and 1960s by utilizing the results of cumulative deposition measurements in soil cores. Despite the higher assessed discharges to atmosphere, calculated doses to members of the public in Seascale remain low and are insufficient to account for any excess of leukaemia. Furthermore, where specific measurements of radionuclide body contents of local residents are available (Stather et al., 1988) the model significantly overestimates body content.
6.5 Management of radioactive wastes Like all industries, the generation of electricity produces waste. Whatever fuel is used, the waste produced in generating electricity must be managed in ways that safeguard human health and minimize the impact on the environment. Concern about nuclear power is usually focused on the highly toxic and radioactive spent fuel and nuclear waste. What is characteristic of these, however, in addition to their toxicity and radioactivity, is that they are limited in volume, which facilitates waste disposal. This contrasts sharply with the waste disposal problem for fossil-fuelled plants. More specifically, a 1000-MWe coal plant with optimal pollution abatement equipment will annually emit into the atmosphere 900 tonnes of SO2, 4500 tonnes of NOx, 1300 tonnes of particulates and 6.5 million tonnes of CO2. Depending on the quality of the coal, up to 1 million tonnes of ashes containing hundreds of tonnes
6.5 Management of radioactive wastes
of toxic heavy metals (arsenic, cadmium, lead and mercury) will have to be disposed of. By contrast, a nuclear plant of 1000-MWe capacity produces annually some 35 tonnes of highly radioactive spent fuel. If the spent fuel is reprocessed, the volume of highly radioactive waste will be about 3 m3. The entire nuclear chain supporting this 1000-MWe plant, from mining through operation, will generate, in addition, some 200 m3 of intermediate-level waste (ILW) and some 500 m3 of low-level waste (LLW) of year. Most countries using nuclear electrical generation have programmes for safe disposal of the wastes. Technical alternatives for disposal of spent fuel and high-level wastes (HLWs) have been assessed by several countries and international organizations. Scientific consensus exists that geologic disposal using a system of natural and engineered barriers is the preferred method to be used. Unlike chemically hazardous industrial wastes, the much smaller volumes of spent fuel and HLW make containment and isolation a feasible disposal option, and their radiological hazard will decrease with time. The volume of highly radioactive waste (HLW) produced by the nuclear industry is small. The IAEA estimates that 370,000 metric tonnes of heavy metal (MTHM) in the form of used fuel have been discharged since the first nuclear power plants commenced operation. Of this, the agency estimates that 120,000 MTHM has been reprocessed, with the balance, 250,000 MTHM, in storage. The amounts of ILW, LLW and very low-level waste (VLLW) produced are greater in volume, but are much less radioactive. Given its lower inherent radioactivity, the majority of waste produced by nuclear power production and classified as LLW or VLLW have already been placed in disposal. The IAEA estimates that over 80% of all LLW and VLLW produced to date are in disposal. For ILW, the agency estimates that about 20% are in disposal, with the balance in storage. Numbers include both solid and liquid radioactive waste (Table 6.7). Generic studies of geologic disposal conducted by the Swedish KBS, the Commission of European Communities (CEC) and others have concluded that geologic disposal systems can achieve an acceptable level of safety to protect future generations from the radiological hazards associated with these wastes. In 1991 IAEA established the Radioactive Waste Safety Standards (RADWASS) Programme to develop a special series of safety documents specifically directed at Table 6.7 Nuclear waste inventory, IAEA estimates IAEA (2018a)
VLLW LLW ILW HLW
Solid radioactive waste in storage (m3)
Solid radioactive waste in disposal (m3)
2,356,000 56,811,000 6.713,000 2.808,000
7,906,000 60,035,000 8.735,000 68.000
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radioactive waste management. The purpose of the RADWASS Programme is to document existing international consensus in the approaches and methodologies for safe waste management and disposal; to create a mechanism for establishing consensus where it does not exist; and to provide Member States with a comprehensive series of internationally agreed documents to complement national standards and criteria. RADWASS has been organized in a hierarchical structure of four levels of safety documents. The top-level publication is a document of safety fundamentals which provides the basic safety objectives and fundamental principles to be followed in national waste management programmes. The lower levels include safety standards, safety guides and safety practice documents. The series has been structured in a logical and clear manner to reflect the systems approach to waste management. Radioactive waste is any material that contains, or is contaminated with, radionuclides at concentrations of radionuclides greater than the “exempted quantities” established by the regulatory body and for which no future use is foreseen. This is after a definition by IAEA. Five main activities produce such waste: • • • • •
uranium and thorium mining and milling; nuclear fuel cycle operations such as uranium conversion and enrichment, fuel fabrication and spent fuel reprocessing; operations of nuclear power stations; decontamination and decommissioning of nuclear facilities; institutional uses of isotopes.
The waste resulting from the above activities comes in various forms (i.e. gaseous, liquid, or solid). These wastes have different characteristics. For safety and technical reasons, the various forms of wastes are usually categorized by their levels of radioactivity, heat content and potential hazard. With regard to disposal the wastes are categorized as follows. LLWs contain a negligible amount of long-lived radionuclides. Produced by peaceful nuclear activities in industry, medicine, research, and by nuclear power operations, such wastes may include items such as packaged gloves, rags, glass, small tools, paper and filters which have been contaminated by radioactive material. Disposal in near-surface structures or shallow burial is practised widely. ILWs contain lower levels of radioactivity and heat content than HLWs, but they still must be shielded during handling and transport. Such wastes may include resins from reactor operations or solidified chemical sludges, as well as pieces of equipment or metal fragments. Commercial engineering processes are being used to treat and immobilize these wastes. Disposal options are similar to those for LLWs. HLWs arise from the reprocessing of spent fuel from nuclear power reactors to recover uranium and plutonium. These wastes contain transuranic elements, and fission products that are highly radioactive, heat-generating and long-lived. Liquid HLW is usually immobilized as a solid glass matrix and stored in interim storage facilities prior to final disposal and isolation in deep, stable, geologic formations,
6.5 Management of radioactive wastes
as currently planned by many national programmes. Spent nuclear fuel that is not reprocessed is also considered HLW. α-bearing wastes (also called transuranic, plutonium-contaminated material or αwastes) include wastes that are contaminated with enough long-lived, α-emitting nuclides to make near-surface disposal unacceptable. They arise principally from spent fuel reprocessing and mixed-oxide fuel fabrication. The wastes may be disposed of in a similar manner to HLW.
6.5.1 Storage and disposal Management and disposal of nuclear waste depends mainly on its type. For example, LLW and ILW are often treated (volume reduction) and/or conditioned (waste immobilization) prior to disposal. This area of LLW and ILW waste management, having been established and proven over past years, is considered to be quite mature in terms of technology development. As a result, several effective, safe and feasible treatment and conditioning options exist for these types of wastes. They include storage and decay, compaction and super compaction, incineration, chemical precipitation, evaporation, filtration and ion exchange; these may be followed by immobilization in materials like concrete, bitumen or polymers (Chan, 1992). The most common disposal methods for ILW and LLW involve disposal in shallow earthen or concrete lined trenches or in structures on the ground (commonly referred to as engineered surface facilities). Safe near-surface disposal of LLW has been practised in a number of countries for almost 30 years. The rationale behind near-surface disposal is that the isolation period for this type of waste is relatively limited (up to 300 years) and, therefore, the institutional or administrative control of the disposal site can be assured. HLW management and disposal is quite different. After its useful life, spent nuclear fuel is removed from the reactor. Once removed, it is usually placed into temporary on-site storage before it is either: •
•
placed in interim away-reactor storage (5–100 years), conditioned after a sufficient decay period and stored before its eventual final disposal in a geologic repository or reprocessed after additional away-from-reactor storage. The resulting liquid HLW, containing mostly fission products and a small proportion of the actinides, is then immobilized in a stable matrix (i.e. borosilicate glass), and would then be disposed of in a geologic repository.
Regardless of which option is chosen, there is broad scientific agreement that deep geologic disposal using a system of engineered and natural barriers to isolate these wastes is the preferred method for their disposal (Chan, 1992). According to the report by Oi (1998) at the end of 1997, more than 130,000 tonnes of spent fuel from power reactors were estimated to be stored worldwide containing about 1000 tonnes of plutonium. Another 170 tonnes of separated plutonium were in storage from civilian reprocessing operations, and about
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100 tonnes of excess plutonium from dismantled warheads no longer required for defence purposes were scheduled to be released from the military sector of Russia and United States. Plutonium represents a dual challenge because it is a valuable energy source and a matter of global concern because of its potential health hazards and possible use for the production of nuclear weapons.
6.5.2 Processing of used nuclear fuel Over the last 50 years the principal reason for processing of used nuclear fuel has been to recover unused plutonium and close the fuel cycle gaining some 25%– 30% more energy from the original uranium in the process. The world commercial (civil) reprocessing capacity is shown in Table 6.8. Spent fuel from light-water reactors contains about 1% of plutonium. According to Oi (1998) the IAEA estimates that in 1997 about 10,500 tonnes of spent fuel were discharged from nuclear power reactors worldwide; this amount contains about 75 tonnes of plutonium. It is estimated that the annual production figure will remain more or less the same until 2010. The cumulative amount of plutonium in spent fuel from nuclear power reactors worldwide is predicted to increase to about 1700 tonnes by 2010. It is estimated that about 3000 tonnes of spent fuel discharged from power reactors were reprocessed in 1997, which corresponds to about 30% of the total. About 24 tonnes of plutonium were separated in reprocessing plants and 9 tonnes of plutonium were used mainly as mixed uranium–plutonium oxide fuel (MOX) in lightwater reactors. The imbalance between the separation and use of plutonium had resulted in an accumulated inventory of separated civil plutonium of about 170 tonnes at the end of 1997. IAEA projections of plutonium inventories show that the rate of separation of civil plutonium and its rate of use will fall into balance in a few years. This is Table 6.8 Reprocessing facilities Country
Facility
Fuel type
2017 (tonnes/year)
France UK
La Hague Sellafield (THORP) Sellafield Ozersk (Mayak) Rokkasho Tokai Four plants
LWR LWR Magnox LWR LWR MOX PHWR
1700 600 1500 400 800a 40 330
Russia Japan India a
To start operation in 2021. After WNA, 2018. Processing of Used Nuclear Fuel. World Nuclear Association. http://world-nuclear. org/information-library/nuclear-fuel-cycle/fuel-recycling/processing-of-used-nuclear-fuel; OECD/NEA, 2017. Nuclear Energy Data 2017. Organization for Economic Co-operation and Development (OECD), OECD 2017. Nuclear Energy Agency (EA), NEA No. 7365.
Plutonium inventory (t)
6.5 Management of radioactive wastes
150
100
50
0 1995 1997 1999 2001 2003 2005 2007 2009 2011 2013 2015
Year FIG. 6.3 Projected worldwide civil inventories of separated plutonium. Solid line: minimum Pu inventory, no Pu market; dashed line: end of year Pu inventory, no Pu market; dotted line: end of year Pu inventory with Pu market; dashed-dotted line: minimum Pu inventory with Pu market.
due to an enhanced capacity of MOX fuel production which will amount to 360 tonnes of heavy metal per year in 2000. Beyond this period, the inventory is expected to decrease modestly and level off at around 130 tonnes. Despite the efforts to reduce the current inventories of separated civil plutonium, the worldwide inventories still remain at a substantial level, as shown in Fig. 6.3. In addition to the amounts of civil plutonium, plutonium is being released from dismantled warheads. Under the START-I and -II Treaties, many thousands of US and Russian nuclear warheads are slated to be retired within the next decade. As a result, at least 50 tonnes of plutonium from each side are expected to be removed from military programmes. Oi (1998) points out the problem, which is what to do with plutonium either in a separated form or contained in spent fuel. A number of issues arise because of plutonium’s potential use as an energy source and for the production of nuclear weapons. Presently, plutonium is used in light-water reactors as MOX fuel and also in small amounts for the development of fast-breeder reactors. Currently 22 power reactors in five countries (France, Germany, Switzerland, Belgium and Japan) are loaded with MOX fuel and this number is expected to rise to between 36 and 48 by 2000. The use of MOX reduces the inventory of separated plutonium and is regarded as an interim measure before plutonium’s possible full-scale use in fast reactors later in the next century. It is known that multiple recycling in light-water reactors degrades plutonium, which in turn limits the number of times it can be recycled to two or three. Such degraded plutonium can, however, be used as fuel in fast reactors. Without such reactors, spent MOX fuels will still end up in a final depository or in storage facilities (Oi, 1998). Recently much attention has been given to the accelerator-driven systems, burning in inert matrices, and the use of thorium to burn plutonium. The concept of a
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closed nuclear fuel cycle was traditionally considered as transmutation (burning) of only plutonium and recycled uranium, with minor actinides (neptunium, americium, curium) destined for final geological disposal. But as time goes on, a new understanding is emerging: reduction of the quantity of actinides would ease requirements for final repositories and make them relatively less expensive. Neutron transmutation of long-lived radioactive minor actinides by the fission process—which entails producing energy and simultaneously turning them into shorter-lived nuclides—is being intensely analysed by the technical community. Also being proposed is the neutron transmutation of selected long-lived fission products. Several possibilities for the transmutation of long-lived nuclides by nuclear reactions have been suggested. In the beginning, the best choice appeared to be the use of nuclear reactors. However, recently there has been renewed interest in what are called accelerator-driven systems (ADS), a technology that seems to show good promise. ADS would produce large amounts of electrical energy while simultaneously destroying the plutonium. This appears to offer a better solution to the plutonium problem than multimillennium storage. The use of accelerators for nuclear energy applications is not a new idea and was proposed as early as the late 1940s by E. Lawrence, inventor of the cyclotron. In the 1950s he promoted the development of a Materials Test Accelerator at Livermore to produce intense neutron fluxes for plutonium production. The Canadian Chalk-River Laboratory began intensive studies of accelerator-based systems to breed nuclear fuel for heavy-water reactors. Scientists at Brookhaven National Laboratory also actively promoted accelerator-based options in the late 1970s and early 1980s. For the last 5 years, scientists at Low Alamos National Laboratory have been reevaluating the accelerator-based technology in the light of new advances in technology and the world energy perspective (Bowman et al., 1992). When 1.6 GeV protons strike a large radius target consisting of heavy nuclei such as lead, approximately 55 neutrons are generated per proton. The energy deposition for this process is about 30 MeV of proton energy per neutron compared with about 200 MeV of fission energy deposited per useful neutron from a sustained chain reaction in fissile material such as 235U. The heat per unit volume which must be handled for a given neutron production rate is therefore considerably smaller for the spallation source than for the reactor. The reactor has the further disadvantage that in nearly all designs the fuel is fixed in position with coolant flowing past. Heat deposition in the fuel is therefore limited by the conductivity of the fuel and by the heat capacity and conductivity of the coolant. For the accelerator-driven neutron source the target is itself a flowing liquid heavy metal. The heat load on the target is therefore limited only by the thermal properties of the liquid metal and by the rate at which it flows. This accelerator target is therefore capable of generating a much higher density (and therefore flux) of neutrons than a reactor with fixed fuel because of the much greater power density capability of the flowing target and the much lower energy deposition in the target per neutron produced. The neutron–production–transmutation system considered by Bowman et al., 1992 consists of an accelerator for the proton beam, a flowing heavy metal proton
6.5 Management of radioactive wastes
target for neutron production, and a surrounding blanket containing primarily heavy water (D2O) for moderating the neutrons into the thermal range. The neutron flux may be further enhanced by neutrons from actinide fission in the blanket. The actinide material is transported through the blanket as a molten salt mixed with the carrier salt LiF-BeF2. Heat from the fission process is deposited in the molten salt and carried away by the salt at an exit temperature of 720°C which makes possible electric power generation at a high thermal-to-electric efficiency. A continuous flow system is essential because of the high burn-up rates of fissile material. For example, the lifetime of 239Pu in a thermal flux of 1016 n/cm2-s is only about 1 day so that use of fuel assemblies along the lines of standard practice for reactors is impractical. In addition there must be chemistry facilities for removing stable or short-lived fission products and returning radioactive waste to the blanket. The accelerator, target/blanket, electric power extraction and chemistry facilities are shown schematically in Fig. 6.4. The expected efficiency for conversion of thermal to electric power is 44% and the busbar efficiency of the accelerator is 45%. A fraction of the electrical power is fed Accelerator Heavy-water moderator
Electric power Energy extraction Waste feed
Neutron flux
Advanced chemical separations Stable and short-lived products
Molten salt loop
Liquid lead target
FIG. 6.4 The proton beam strikes the lead target generating neutrons which are moderated in the surrounding heavy water blanket. Molten salt carrying fissile material for heat generation and electric power production circulates in the heavy water blanket through double-walled pipes. Some of this power drives the accelerator. Nuclear waste including that produced in the molten salt is also circulated through the blanket in a separate loop and transmuted to stable and short-lived nuclides which are extracted and stored. After Bowman, C.D., et al., 1992. Nuclear energy generation and waste transmutation using an accelerator driven intense thermal neutron source. Los Alamos Report LAUR-91-260 and Nuclear Instruments and Methods A320, 336.
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Fission
Spallation target Protons 1 GeV
Spallation residues
Window
Neutrons Transmutation FIG. 6.5 Schematic representation of a hybrid plant. An intense beam of proton from an accelerator hits a massive target (Pb, Pb–Bi, W, Hg, …) and produces fast neutrons. In a surrounding blanket, the neutrons are used to drive a subcritical medium.
back to power the accelerator which operates at an energy of 1.6 GeV and produces neutrons in a Pb target. The beam power deposited in the Pb and the thermal power deposited in the D2O blanket are not converted to electrical power although energy recovery loops on these systems could boost the overall system efficiency. The material to be transmuted can be fed either into the molten salt carrier as molten salt or into the heavy water as dissolved salt depending on the application (Bowman et al., 1992). Recently there has been increased interest in the idea of accelerator-driven reactors, see, for example, Takizuka (1994), Van Tuyle et al. (1993), Carminati et al. (1993), Rubbia et al. (1995). In such systems (Fig. 6.5), spallation reactions induced by a high-intensity beam (10–250 mA) of GeV protons on a heavy target produce an intense neutron flux. These neutrons, after being more or less moderated, are used to drive a subcritical blanket. The extra neutrons provided by the accelerator allow the maintenance of the chain reaction while burning the long-lived nuclear waste. The plant generates electricity, part of which is used to supply the accelerator. Besides the more favourable neutron economy, additional advantages of accelerator-driven systems are safety and versatility. Obviously the operation of the blanket in a subcritical state is a major safety advantage. It could for instance allow the introduction of a large amount of Pu or minor actinides which is difficult in classical reactors because of control problems due to the smaller fraction of delayed neutrons. Accelerator-driven systems are also more flexible than reactors since the intensity of the accelerator can be adjusted to counteract the growth of
6.5 Management of radioactive wastes
poisonous isotopes or when adding elements to be transmuted. Their main drawbacks are their complexity and the technological progress they imply for the accelerator, the target-blanket and the interface between them (Boudard et al., 1998). Recently, it has been proposed to construct a demonstrator facility of significant power of the order of 100 MW (thermal) on a 10-year time schedule as a regional European facility.
6.5.3 Smart use of nuclear waste Use of nuclear energy as based on current technology has some drawbacks expressed as worries about reactor accidents, the potential for diversions of nuclear fuel into nuclear weapons, the management of long-lived radioactive waste and the depletion of economical uranium reserves. The proposed new fuel cycle would combine two innovations, pyrometallurgical processing and fast neutron reactors for burning that fuel. With this approach radioactivity of nuclear waste would drop to safe levels in a few hundred years (Hannum et al., 2005). In about 3 years of use nuclear fuel in reactors needs to be changed because of radiation-related degradation and the depletion of 235U. Not to mention that at this time 239Pu in the reactor, generated from 238U capture of thermal neutrons, contributes significantly to the reactor power. Neutrons are thermalised by reactor cooling water. At that time reactor has generated lots of “spent fuel” consisting of three classes of materials. It is generally accepted to divide spent fuel material into three components. First component is made of fission products (5% of used fuel)—this is the true waste—the content are the products of the fission process in which the energy was liberated, so by analogy this could be called “the ash”. This mix of elements is highly radioactive for some years after it is generated. After some time (10 years) its activity is dominated by two isotopes: 137Cs and 90Sr. This component is not highly radioactive; it could be separated from the rest of spent fuel and stored in dedicated facility for future use. Second component is made of unfissioned uranium, mainly 238U (94%), and this material resembles natural uranium. This component is only mildly radioactive and can be separated from the rest of spent fuel and stored in dedicated facility for future use. The third component is the one which contains all the problems; it contains all of transuranic elements (elements heavier than uranium). This component of spent fuel is made mainly of the mix of plutonium isotopes and americium. Although this component represents only 1% of spent fuel it generates almost all the problems we have with spent fuel today. The half-lives of these isotopes range up to tens of thousands years. An idealistic approach to the solution of this problem was to use produced plutonium in the fast-neutron reactors, so-called breeders because of producing more plutonium than consuming. However, the same plutonium could be used to make bombs and lead to uncontrolled proliferation of nuclear weapons. There is, however, an alternative recycling process which does not involve pure plutonium at any stage. This fast-neutron reactor technology minimizes the risk that spent fuel from energy production would be used for bomb manufacture. In addition,
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CHAPTER 6 The nuclear fuel cycle
Table 6.9 Transmutation probabilities (%) Isotope
Thermal neutrons
Fast neutrons
137
3 7 63 1 75 1 1 75 1 1 78
27 70 85 55 87 53 21 94 23 10 94
Np Pu 239 Pu 240 241 242 241 242m 243 242 243 238
After WNA, 2018. Processing of Used Nuclear Fuel. World Nuclear Association. http://world-nuclear.org/information-library/nuclear-fuel-cycle/fuel-recycling/ processing-of-used-nuclear-fuel.
fast reactors can extract more energy from fuel than thermal reactors because of 238U, 239 Pu and other transuranic isotopes having greater fission cross section for fast than thermal neutrons. Here, we need to introduce a concept of pyroprocessing process in which a mix of transuranic elements instead of pure plutonium is extracted from the used fuel. The procedure is based on electroplating; it uses metallic electrode to collect dissolved spent metallic fuel in a chemical bath. In the next step the accumulated material is scraped from the electrode, melted and casted into ingots to be refabricated as a fuel rod. The pyroprocess collects almost all of transuranic elements from the used fuel together with uranium and fission products. This combination is unsuitable for weapon production or the use in thermal reactors. This will be implemented in the fourth-generation fast neutron reactors in the late 2020s. Additionally, fast reactors make only 1% of spent fuel material compared to thermal reactors. As a result, the waste management would be greatly simplified. Transmutation of one radionuclide into another can be achieved by neutron bombardment in a nuclear reactor or accelerator. As known, the objective of transmutation is to change actinides into fission products and long-lived fission products into significantly shorter-lived nuclides. The goal is to have wastes which become radiologically innocuous in only a few hundred years. In comparison fast neutrons are more efficient than thermal neutrons, see Table 6.9 after WNA (2018).
6.6 Research reactors Nuclear reactors have supported research in many different fields and have contributed to discoveries in many scientific disciplines. Altogether there are about 180 research reactors in operation in the world (IAEA, 1996). Recently, IAEA has
6.6 Research reactors
published its Nuclear Energy Series no. NP-T-5.8 volume containing a CD-ROM containing 30 research reactor profiles, which provide technical descriptions and specific features for utilization. In addition, IAEA keeps Research Reactor Data Base, IAEA-RRDB (2018), the Research Reactor Database (RRDB) provides extensive information on research reactors all over the world. It is a combination of two individual databases: one that provides profiles of research reactors (both operational and shut down), and the other that compiles fuel information. The research reactor section of the database contains technical specifications and utilization information to assist potential users of research reactors. The fuel cycle section of the database is not publically accessible, but general reports on fuel information are compiled from the database. Both databases are populated and updated by officially nominated facility data providers (FDPs) who are typically located at the reactor. The present situation is shown in Table 6.10. Research reactors have a very wide variety of uses, including neutron scattering (in which beams of thermal neutrons are scattered by the atoms in a sample, revealing its structure, magnetic state and atomic binding energies); neutron activation analysis; radiography; irradiation testing of materials and production of radioisotopes for medical, research and industrial use. These capabilities are applied by researchers in many fields, ranging from archaeology to materials science and from fusion research to environmental science. Few generalizations can be made about the applications for research reactors or about their users. Research reactors themselves tend to have a very different set of safety-related parameters from power reactors. Some are helpful differences like simplicity, relatively low power and low-temperature coolant. Other differences, especially the need for a high-power density core, pose challenges not faced in a power reactor. These challenges can be met through thoughtful design solutions. A research reactor’s power is usually in the range 0–100 MW thermal. The fission product inventory and the stored energy in research reactors are smaller than in power reactors. However, some of the research reactors have large power density
Table 6.10 Status of 840 research reactors around the globe Status Planned Under construction Operational Temporary shutdown Extended shutdown Permanent shutdown Under decommissioning Decommissioned
Developed countries
Developing countries
All countries
2 4 140 8 5 43 62
11 6 85 5 9 14 4
13 10 225 13 14 57 66
413
29
442
375
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CHAPTER 6 The nuclear fuel cycle
Control rods
Water
Reactor Core
FIG. 6.6 Basic H2O-cooled swimming-pool reactor. After West, C.D., 1997. Research reactors: an overview. Nuclear News, October 1997, p. 50.
(>5000 kW thermal per kg of fuel). The typical research reactor is of the swimming pool type, as shown in Fig. 6.6. In many of these reactors the core is made up of so-called materials testing reactor-type fuel elements which are aluminium-clad, curved plates of fuel arranged in long rectangular boxes arranged between grid plates to form the core. Several positions in the grid are not occupied by fuel elements, but by control rods, beryllium reflectors or experimental capsules. Cooling may be by natural convection of the pool water, although this is augmented, for operation at higher power, by pumping pool water through the core. More powerful research reactors, of which the international Institut LaueLangevin (ILL) facility at Grenoble, France, and the high flux isotope reactor (HFIR) at Oak Ridge, Tenn., are well-known examples, have tanks that are full pressure vessels—for example, the coolant inlet pressure at HFIR is nominally 470 psi, and at ILL it is 200 psi. Again, aluminium-clad fuel plates are used, the fuel meat being a layer, about 50 mils thick, of U3O8 particles mixed with powdered aluminium for enhanced thermal conductivity, the layer being clad with aluminium plates about 10 mL thick. In these two reactors, the fuel elements are annular, with curved (involute) plates fitting into axial grooves down two concentric cylinders.
6.7 Advanced nuclear power plants
6.7 Advanced nuclear power plants New generations of nuclear power plants have been or are being developed, building on this background of success and applying lessons learned from the experience of operating plants. Advanced designs currently under development comprise three basic types: • • •
water-cooled reactors, using water as coolant and moderator, fast reactors, using liquid metal, e.g., sodium, as coolant, gas-cooled reactors, using gas, for example, helium, as coolant and graphite as moderator.
Global developments in this field have been summarized by Juhn et al. (1997). We present here their findings for some countries. United States Important programmes in development of ALWRs were initiated in the mid-1980s in the United States. In 1984, the Electric Power Research Institute (EPRI), in cooperation with the US Department of Energy (DOE) initiated a programme to develop utility requirements for ALWRs to guide their design and development. Utility requirements were established for large boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) having power ratings of 1200–1300 MWe, and for mid-sized BWRs and PWRs having power ratings of about 600 MWe. In 1986, the US DOE, in cooperation with EPRI and reactor design organizations, initiated a design certification programme for evolutionary plants based on a new licensing process, followed in 1990 by a design certification programme for mid-size plants with passive safety systems. The new licensing process allows nuclear plant designers to submit their designs to the US Nuclear Regulatory Commission (NRC) for design certification. Once a design is certified, the standardized units will be commercially offered, and a utility can order a plant, confident that generic design and safety issues have been resolved. The licensing process will allow the power company to request a combined licence to build and operate a new plant, and as long as the plant is built to preapproved specifications, the company can start up the plant when construction is complete, assuming no new safety issues have emerged. Four advanced reactor designs developed in the United States have been submitted to the NRC for certification under the US DOE ALWR programme. Two large evolutionary plants—the System 80+ of ABB-Combustion Engineering and the ABWR of General Electric—received Final Design Approval in 1994 and Design Certification in May 1997. The 600-MWe AP-600 of Westinghouse is under NRC review and a Final Design Approval is expected by March 1998. Up to mid-1996, the 600-MWe simplified BWR developed by General Electric was also under review, but then the company stopped work on the 600-MWe version and shifted its emphasis to a unit with larger output. The first-of-a-kind engineering programme (FOAKE, the detailed design needed to verify cost and the construction schedule)
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CHAPTER 6 The nuclear fuel cycle
authorized by the 1992 Energy Policy Act was completed for the ABWR in September 1996, and similar work on the AP-600 has also been done. The power company in Taiwan, China, recently selected General Electric’s ABWR design for two new units slated for operation in 2004. NASA is developing nuclear reactors for spacecraft propulsion and for a planetary power source; the goal is to have both types available for mission to Mars probably in 2030s. The agency is advancing a reactor technology that uses low-enriched uranium (LEU) containing less than 20% of 235U for propulsion, its planetary power source, known as kilopower, utilizes 90% or more enriched 235U. Kilopower reactor might compete with solar as power source for human settlement on Mars in spite of long lunar night period of half of month (Kramer, 2017). There is strong opposition to within some organizations since the 1 kW of kilopower reactor uses about 30 kg of weapon grade uranium, which is more than enough to produce a nuclear device. NASA states that planed 10 kW reactor will require only 50 kg of weapon-grade uranium. United States has set a mid of 2030s as goal for human travel to Mars. Although the HEU fuelled reactor is simple to build and test, the LEU reactor would eliminate concerns of fissile material failing into the wrong hands in the case of launch failure. France and Germany In Europe Framatome and Siemens have established a joint company, Nuclear Power International, which is developing a new advanced reactor, the European pressurized-water reactor (EPWR), a 1500-MWe plant with enhanced safety features. The basic design will be completed in mid-1997, and the design will be reviewed jointly by the French and German safety authorities. This procedure will provide strong motivation for the practical harmonization of the safety requirements of two major countries, which could later be enlarged on a broader basis. Siemens is also, together with German utilities, engaged in the development of an advanced BWR design, the SWR-1000, which will incorporate a number of passive safety features, for initiation of safety functions, for residual heat removal, and for containment heat removal. Sweden and Finland In Sweden, ABB Atom, with involvement of the utility Teollisuuden Voima Oy (TVO) of Finland, is developing the BWR-90 as an upgraded version of the BWRs operating in both countries. Republic of Korea In the Republic of Korea, an effort started in 1992 to develop an advanced design known as the Korean Next Generation Reactor (KNGR), a 4000-MWth PWR design. The basic design is currently being developed by the Korea Electric Power Corporation (KEPCO) with the support of the Korean nuclear industry. The goal is to complete a detailed standard design by 2000. Russian federation In the Russian Federation, design work is under way on the evolutionary V-392, an upgraded version of the VVER-1000, and another design version is being developed
6.7 Advanced nuclear power plants
in cooperation with the Finnish company Imatran Voima Oy (IVO). Also being developed is a mid-sized plant, the VVER-640 (V-407), an evolutionary design which incorporates passive safety systems, and the VPBER-600, which is a more innovative, integral design. Construction of the first unit of the VVER-640 is planned to start at Sosnovy Bor in 1997. Construction of two 1000-MWe VVERs is being discussed with the People’s Republic of China. Japan In Japan, the Ministry of Trade and Industry is conducting an “LWR Technology Sophistication” programme focusing on development of future LWRs and including requirements and design objectives. A large, evolutionary 1350-MWe advanced PWR is being developed by Japanese utilities together with nuclear vendors, with construction of a twin unit being planned at the Tsuruga site. In addition, an advanced BWR Improvement and Evolution study was started in 1991. It involves development of a reference 1500-MWe BWR that reflects the accumulated experience in operation and maintenance of BWRs. Development programmes for a Japanese Simplified BWR (JSBWR) and PWR (JSPWR) projects which involve vendors and utilities are also in progress. The Japan Atomic Energy Research Institute (JAERI) has been investigating conceptual designs of advanced water-cooled reactors with emphasis on passive safety systems. These are the JAERI Passive Safety Reactor (JPSR) and the System-Integrated PWR (SPWR). China In China, the Nuclear Power Institute (Chengdu) is developing the AC-600 advanced PWR, which incorporates passive safety systems for heat removal. In all of these countries, the advanced LWRs under development incorporate significant design simplifications, increased margins, and various technical and operational procedure improvements. These include better fuel performance and higher burn-up, a better man–machine interface using computers and improved information displays, greater plants standardization, improved constructability and maintainability and better operator qualification and simulator training. Canada The continuing design and development programme for heavy-water cooled reactors (HWRs) in Canada is primarily aimed at reduction of plant costs and at an evolutionary enhancement of plant performance and safety. Two new 715-MWe CANDU-6 units with improvements over earlier versions of this model are under construction in Quinshan, China. Up-front basic engineering continues on the 935-MWe CANDU-9 reactor, a single unit adaptation of reactor units operating in Darlington, Canada. The two-year licensability review by the Canadian Nuclear Safety Commission was completed in Japan 1997, and found that the CANDU-9 meets the country’s licensing requirements. Further studies are being carried out for advanced versions of these reactor models to incorporate further evolutionary improvements and to increase the output of the larger reactor up to 1300 MWe.
379
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CHAPTER 6 The nuclear fuel cycle
India Also under development is an advanced 500-MWe HWR in India, and construction of such units is planned. This HWR design takes advantage of experience feedback from the 220-MWe HWR plants of indigenous design operating in India. On the other hand, liquid metal-cooled fast reactors (LM-FRs), or breeders, have been under development for many years. With breeding capability, fast reactors can extract up to 60 times as much energy from uranium as can thermal reactors. The successful design, construction and operation of such plants in several countries, notably France and the Russian Federation, has provided more than 200 reactor-years of experience on which to base further improvements. In the future, fast reactors may also be used to burn plutonium and other long-lived transuranic radioisotopes, allowing isolation time for high-level radioactive waste to be reduced. Significant activities are occurring in the development of high-temperature gascooled reactors (HTGRs), particularly with regard to the utilization of the gas-cooled reactor to achieve high efficiency in the generation of electricity and in process heat applications. Technological advances in component design and processes—coupled with the international capability to fabricate, test and procure the components— provide an excellent opportunity for achieving HTGR commercialization. United Kingdom, Germany and United States Gas-cooled reactors have been in operation for many years. In the United Kingdom, nuclear electricity is mostly generated in CO2-cooled Magnox and advanced gascooled reactors (AGRs). Other countries also have pursued development of hightemperature reactors (HTGRs) with helium as coolant, and graphite as moderator. The 13-MWe AVR reactor has been successfully operated for 21 years in Germany demonstrating application of HTGR technology for electric power production. Other helium-cooled, graphite-moderated reactors have included the 300-MWe thorium high-temperature reactor in Germany, and the 40-MWe peach bottom and 330-MWe Fort St. Brain plants in the United States. South Africa In South Africa, the large national utility, Eskom, which has an installed generation capacity of about 38,000 MWe, is in the process of performing a technical and economic evaluation of a helium-cooled pebble bed module reactor. It would be directly coupled to a gas turbine power conversion system for consideration in increasing the capacity of the utility’s electrical system. China and Japan In China and Japan, test reactors are under construction which will have the capability of achieving core outlet temperatures of 950°C for the evaluation of nuclear process heat applications. Construction of China’s high temperature reactor (HTR-10) at the Institute of Nuclear Energy Technology (INET) continues with initial criticality anticipated for 1999. This pebble-bed reactor of 10 MW will be utilized to test and demonstrate the technology and safety features of the HTGR. Development of the HTGR by INET is being undertaken to evaluate a wide range
6.8 Nuclear fusion
of applications. They include electricity generation, steam and district heat production, combined steam and gas turbine cycle operation and the generation of process heat for methane reforming. The HTR-10 is the first HTGR to be licensed and constructed in China ( Juhn et al., 1997). Many countries have started the participation in IAEA programme for the development of small and medium-sized modular reactors (SMRs technology). The driving force in development of such reactors includes:
replacing the ageing fossil fuel power plants, enhancing safety performance through inherent and passive safety features, offering better economic affordability, suitability for nonelectric applications. options for remote areas, synergic energy systems that combine nuclear and renewable energy sources.
SMRs stands for small nuclear reactor and are defined as advanced nuclear reactors that produce equivalent electric power up to 300 MW(e) and are designed to be built in factories and transported to utilities for needed installation. For water-cooled SMRs modularity is achieved by integrating major components of the reactor coolant system inside the reactor pressure vessel (IAEA, 2016). Dozens of advanced SMR designs are under development or near deployment. The three reactors in SME category leading the group are one in Argentina (CAREM25—an industrial prototype integral PWR), one in the Russian federation (KLT-405 a barge mounted floating power unit) and one in China (HTR-PM, an industrial demonstration plant of high-temperature pebble-bed gas cooled reactor). The development of SMRs comes with several different concepts, coolants, neutron spectrum, deployment location and application (IAEA, 2016). The advanced SMRs currently under development in the United States represent a variety of sizes, technology options and deployment scenarios. These advanced reactors, envisioned to vary in size from a couple of MW up to hundreds of MW, can be used for power generation, process heat, desalination or other industrial uses. SMRs can employ light water as a coolant or other nonlight water coolants like gas, liquid metal or molten salt. As an example, here we mention SMRs called “U-Batteries” being developed by Urenco Ltd. (Tirone, 2017) able to generate 10 MW of power or heat. This SMR is being developed for small towns and industries in areas beyond the reach of large nuclear plants. According to its prospectus, central to the U-Battery design is its so-called TRISO fuel, a three-layered sphere with uranium kernel that can withstand temperatures as high as 1800°C.
6.8 Nuclear fusion A central issue for economic growth, prosperity and the quality of life in the industrialized world is the availability of secure, sustainable and financially competitive
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CHAPTER 6 The nuclear fuel cycle
50
Gt -OIL EQUIVALENT
382
40 high 30 20 low
10 0 1850
1900
1950
2000
2050
2100
YEAR
FIG. 6.7 Global energy use in gigatonne oil equivalent (Gtoe).
sources of energy. Given the expected growth in energy demand in the future, even with vigorous measures for energy savings, use will need to be made of all potential energy sources. The WEC projects growth in energy demand of anywhere between 50% and 300% over the next five decades, depending on environmental and economic factors (see Fig. 6.7). Strategic considerations favour the development of energy sources that offer greater sustainability and have less impact on health and the environment. Nuclear fusion, for which the fuel source is virtually limitless in quantity, could in the long term be an important option in this energy mix. There are several approaches to the problem of nuclear fusion. The most promising is definitely magnetic confinement fusion (MCF). In the course of the last 50 years research on magnetically confined plasmas has brought MCF to the threshold of net power production and has revealed much of the physics underlying the complex behaviour of hot plasmas immersed in a magnetic field. The focus of contemporary fusion research is the deuterium–tritium reaction: 2
H +3 H!4 Heð3:5 MeVÞ + nð14:1 MeVÞ
(6.7)
which is the fusion reaction with the largest cross-section at the temperatures which are likely to be achieved in laboratory experiments (several 108 K). A total of 80% of the reaction energy appears as the kinetic energy of the neutron, which would be absorbed in the structure of a power plant and provides most of the energy for steam generation. The α-particle would be trapped in the plasma where its energy would heat the plasma and maintain the conditions required for fusion reactions to occur. Since tritium is a radioactive gas and a high flux of 14.1 MeV neutrons would induce significant radioactivity in the structure surrounding the plasma, current experiments on magnetically confined plasma are usually carried out in hydrogen, so that no neutrons are produced, or in deuterium, for which the neutron production rate is almost two orders of magnitude lower than in a deuterium–tritium mixture.
6.8 Nuclear fusion
To achieve the conditions necessary for “ignition”, where the α-particle power produced by fusion reactions exactly balances the heat loss due to transport processes, the plasma must be heated to a temperature of approximately 108 K at a particle density in the region of 1010 ions per cubic metre, while maintaining an energy replacement time of about 5 s. There are three different toroidal confinement configurations, each of them being a potential route to a possible fusion power plant: 1. Tokamak uses a strong toroidal field of several Tesla produced by a set of discrete coils. 2. The reversed field pinch (RFP) is a closely related configuration, since the plasma formation and ohmic heating are essentially identical to the tokamak. However, in the tokamak the average poloidal field is limited by stability requirements to approximately an order of magnitude smaller than the toroidal field, whereas the two are of similar magnitude in the RFP, both being typically less than 1 T. 3. The third class of toroidal confinement devices is the stellarator which differs in an essential way from tokamaks and RFPs in that the helical fields are created entirely by coils external to the plasma, with no net toroidal current following within the plasma. In order for plasma to achieve ignition the product of plasma density, energyconfinement time and ion temperature must reach a value of 5 1021 m3 skeV. The increasing scale of magnetic confinement experiments, together with the accompanying improvements in the understanding of the physics of magnetoplasmas, has raised the values attained experimentally by seven orders of magnitude since 1955 and has brought the field to the present point, where the largest experiments are within a factor of 5 of the required value. On this basis it can be expected that the parameters of the ITER tokamak are adequate to ensure that ignition will be achieved (Wesson, 1997). According to Campbell (1998), three principal conclusions can be drawn about the present status of MCF. First, there is now substantial, though still incomplete, understanding of plasma behaviour in the principal toroidal confinement configurations, and there is a much deeper appreciation of the complexity of the physics of high-temperature magnetoplasmas. Second, new opportunities for further improvement in plasma performance are opening with the advent of a new generation of large stellarators such as LHD, the development of “advanced tokamak scenarios”, which may offer a viable route to steady-state tokamak operation, and realization of a variety of new tools for enhancing plasma performance in RFPs. Finally, given the production of over 10 MW of DT fusion power in TFTR and 16 MW in JET, plasma performance in tokamaks has advanced to the point where the construction of a DT-burning plasma experiment such as ITER would be a timely next step. An alternative to magnetic confinement is so-called inertial confinement fusion (ICF). The basic idea is to ignite and burn a few milligrams of deuterium-tritium fuel by means of high-power laser or ion beam pulses. Two large laser facilities are presently
383
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CHAPTER 6 The nuclear fuel cycle
under construction which should demonstrate within the next 5–10 years the feasibility of single microexplosions. These are the National Ignition Facility (NIF) in Livermore, USA and the Laser MegaJoule (LMJ) in Bordeaux, France. In contrast to MCF, inertial confinement involves no magnetic fields to contain the fuel, but relies exclusively on mass inertia. In ICF fusion burn occurs in highly compressed deuterium–tritium fuel, heated to an ignition temperature of 108 K. In the standard scheme compression and heating is achieved by spherical implosion of small capsules containing the fuel. A short pulse of radiation (laser, ion beam or X-ray radiation) is used to ablate the outer layer of the capsule and to implode the inner part, driven by the ablation pressure like a spherical rocket. The energy yield of the ignited capsule (up to some 100 MJ) can be contained in a reactor vessel. For energy production the scheme implies pulsed operation with a few microexplosions per second. Presently, there are two paths to achieving uniform irradiation. First, the direct drive approach, where a large number of overlapping beams is shone directly on the fusion capsule, and second, the indirect drive approach, where one converts the beam energy into X-rays which then drive the capsule implosion. At present, direct drive is thought to be possible only with lasers. The scientific and technological basis has been developed to ignite and burn microfusion targets by means of MJ laser pulses. In scaled experiments, implosions with high convergence ratio and neutron yields have been achieved, showing close agreement between experiment and multidimensional simulations. The crucial problem of symmetry and stability is approached along two lines, direct drive using laser smoothing techniques and indirect drive using gas-filled hohlraums (Lindl, 1995). Let us describe in some detail the European Union fusion programme (Bruhns, 1998). The starting point of this programme could be considered the creation of the European Atomic Energy Community (Euroatom) in 1957. Today, all EU member states have institutions actively participating in the fusion programme—all states except Greece participate through “association” contracts. The Community’s own Joint Research Centre (JRC), which has institutes in various locations, also undertakes work for the programme. Switzerland is fully associated to the programme (as Sweden was before it had become an EU Member State). Associations were established in Finland (1995) and in Austria (1996) after enlargement of the Union took in these countries. The associations are the backbone of the fusion programme. They operate a number of fusion devices in their laboratories (see Table 6.11). Most of these fusion devices have been built along the tokamak principle, but there are also stellarators and RFPs. And there are a number of facilities for technological development such as large superconducting-magnet-testing facilities. At the end of the 1970s it was decided to build, under the name of the JET Joint Undertaking, a fusion device (a tokamak) of much larger size than any fusion experiment existing at the time, JET, the Joint European Torus, located at Abingdon in the United Kingdom, began operation in 1983 and has become the flagship of the whole EU fusion programme.
Table 6.11 EU fusion devices
Name
Location
Objective
JET
Abingdon, UK
TORE SUPRA
Cadarache, F
TEXTOR 96
€lich, D Ju
ASDEXUpgrade FTU
Garching, D
TCV COMPASS-D
Lausanne, CH Culham, UK
ISTTOK MAST
Lisbon, P Culham, UK
Integrated high performance operation, DT operation Long-pulse operation in next step relevant conditions Plasma/wall interaction, heating and exhaust, pumped divertor Poloidal divertor, plasma purity control in reactor Confinement at high density and high current, high power wall studies High-beta studies and disruption control High-beta and MHD stability studies, poloidal divertor-type shaped plasma MHD activity transport Tight aspect ratio tokamak physics with hot plasmas, disruption avoidance
Power/ radius (MW/m)
Major/ minor radius (m)
Current (MA)
Magnetic Field (T)
2.96/1.25
7
3.45
2.96/1.25
1.7
4.5
9.3
88-
1.75/0.48
<0.8
3.0
4.6
83-
1.65/0.50
<1.6
3.9
0.93/0.3 J
<1.6
8.0
8.5
90-
0.88–0.24 0.56/0.21
<1.2 0.35
1.4 2.1
5.1 4.5
9289-
0.46/0.09 >0.6/>0.5
0.091 >1
0.5 0.5
Operational from
Tokamaks
Frascati, I
83-
91-
9398-
Stellarators WENDESTEIN 7-X WENDESTEIN 7-AS TJ-II
Greifswald, D Garching, D Madrid, E
Modulator cryogenic coil system to study fully optimized HELIAS in hot plasmas Modulator coils, plasma behaviour in an optimized magnetic configuration Flexible stellarator, helical magnetic axis for confinement and high-beta studies
5.5/0.55
3.0
3–7
2006-
2.0/0.20
2.5
2.4
90-
1.5/0.20
1.2
2.9
97-
Largest RFP device, to study the reactor prospects for this concept Medium size RFP; shell, stability and transport studies
2.0/0.48
<2.0
91-
2.0/0.18
0.5
94-
Reversed field pinches RFX
Padova, I
EXTRAP-T2
Stockholm, S
After Bruhns, H., 1998. The EU fusion programme. Europhysics News, November/December, p. 206.
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Around the same time, the Next European Torus team (NET) was established and given the task of enhancing the programme’s activities on safety and the environment, concentrating on the preparation (in particular the engineering and technological side) of the next-step experiment beyond JET. The NET team has become the pivotal point for initiating and coordinating R&D in fusion technology, as well as for Europe’s contribution to the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER EDA) which was established in 1992 by the EU, Japan, Russia and the United States.
6.8.1 ITER In the fusion field research has been mainly focused on the magnetic confinement of extremely hot plasma of fusion fuel ions and electrons while the practical aspects of the scheme have often been neglected. Because of Russian tokamak success in the late 1960s fusion researchers worldwide dropped most other approaches and built tokamak experiments and designed power plants based on tokamak (Dean, 2013). In the late 1980s, the idea of building a large internationally funded and managed tokamak facility was born. More than 200 tokamaks around the world have paved the way to the ITER experiment. International effort on fusion, International Thermonuclear Experimental Reactor (ITER), was proposed by Gorbachev (General secretary of Soviet Union at that time) to US president Reagan in Geneva meeting held in November 1985 in Geneva. A year later an agreement was reached between Euroatom (EU-28 plus Switzerland), Japan Soviet Union and United States to jointly pursue the design of large international fusion facility. Conceptual design work began in 1998 and finished in July 2001 by approval of Member States (IAEA, 2001). The People’s Republic of China and the republic of Korea joined the Project in 2003. In December 2005 the delegations from China, European Union, Japan, the Republic of Korea, the Russian Federation and the United States of America met in Korea to complete their negotiations on an agreement on the joint implementation of the ITER international fusion energy project. In this meeting India was welcomed as a full party to the ITER venture, the delegation from India then participating fully in the discussions that followed. At that time the ITER Parties represented 80% of the global GDP and half of the world’s population. The aim of ITER is to demonstrate the scientific and technological feasibility of fusion energy by constructing a functional fusion power plant. The IAEA has been actively involved in the ITER project from its inception, providing its auspices and practical support, including publication of technical documents and ITER Newsletter. In December 2005 the ITER Joint Work Site in Cadarache, France, was inaugurated. The offices became operational in January 2006. The ITER Agreement was officially signed in Paris on 21 November 2006 by Ministers from seven ITER Members, this document established a legal international entity to be responsible for the building, operating and decommissioning the Project. Construction of ITER is expected to take about 10 years, and the reactor will then operate for a further 20 years.
6.8 Nuclear fusion
ITER’s mission is to demonstrate the feasibility of fusion power and to prove that the tokamak type of magnetic confinement can achieve the following (Freidberg, 2007): produce 10 times more thermal energy from fusion power (Q ¼ 10) than is supplied by auxiliary heating for period of few minutes, produce a steady-state plasma with a Q > 5 in a long pulse for about 8 min, develop the advanced technologies and processes needed for a fusion power plant, verify the rates of tritium breeding from lithium, find the life-time of vessel’s walls from the neutron bombardement and heating and erosion. Since Europe is the host of this experiment its financing is of the project represents almost half of the total price. The EU body to manage the European contribution to ITER is called “Fusion for Energy” (F4E) has about 450 employees in three locations: Barcelona (Spain), Cadarache (France) and Garching (Germany). Since its establishment F4E has invested in Europe’s economy about 4 billion EUR (EC News, 2018). The ITER project is now under construction in Saint-Paul-lez-Durance, France. It aims to demonstrate sustained fusion power output of 500 MW thermal. In December 2017 ITER celebrated an important milestone having reached 50% completion of the total construction work needed for the first operation stage, so-called First Plasma. ITER has been discussed in many books and scientific, technical and popular papers and presentation. One of the books to be recommended is the book by Horton and Benkadda (2015) which is dedicated to the ITER tokamak now under construction in France which aims to investigate the feasibility of fusion power. Another book of interest is book by Horton and Benkadda (2015) on ITER Physics discussing ITER machine architecture and objective, magnetohydrodynamic description of equilibrium and heating of thermal plasma, operational regimes, steady-state operation, plasma diagnostic and other aspects of d + t reaction including numerous references to scientific papers in all aspect of tokamak physics, Some authors (i.e. Hirsch, 2017) question the purpose of this project by asking why ITER continues to be built; he proposes a dramatic reorganization of fusion research with an accent on p + 11B reaction which does not produce neutrons. Arguing that the absence of neutrons would largely eliminate the risks due to radioactivity and thereby dramatically enhance economics, regulatory simplicity and public acceptance. Cowley (2017) has replied to concerns brought up by Hirsch, 2017 by stating that ITER success is not assured at this point of time so it is rather early to state that tokamaks fail against the Electrical Power Research Institute (EPRI) criteria of the economic viability of electricity production. Success has always seemed just a few decades away from ITER because of design and management problems which led to long delays and enormous increase in cost. It is expected that the burning plasma sustained for 6 or 7 minutes could be achieved 2035 at the earliest for a price of around 25 billion Euros. We finish this part of text
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with the title of the article in The New York Times written by Fountain (2017): “A dream of clean energy at a very high price”. The book by Horton and Benkadda (2015) is dedicated to the ITER tokamak under construction in France which aims to investigate the feasibility of fusion power.
6.9 Nuclear batteries The goal of nuclear battery is to convert energy from radioactive decay into electricity. This can be done by following options: (i) direct charge collection, (ii) indirect (scintillation), (iii) betavoltaic, (iv) thermoelectric, (v) thermionic and (vi) thermophotovoltaic. Nuclear batteries have been investigated and used because of the potential of nuclear battery for longer shelf-life and higher energy density when compared with other modes of energy storage. The performance of nuclear batteries is a function of the radioisotope used as a fuel, radiation transport properties and energy conversion transducers (Prelas et al., 2014). Table 6.12 lists the most often considered isotopes and their characteristic parameters. A useful short review and preview of nuclear battery technology is presented by Park (2017) which was submitted as coursework for PH241 at Stanford University in winter 2017. The extensive description of the physics of nuclear battery operation is given in the book by Prelas, Boraas, and De La Torre Aguilar (2016b). It provides a comprehensive background that allows readers to understand all past and future developments in the field. The book has a chapter on potential applications for nuclear batteries (Prelas, Boraas, De La Torre Aguilar, Seelig, et al., 2016a). Radioisotope heater units, RHU; several types are available. NASA’s RHU has heat output of 1 W, fuel loading is 33.6 Ci—2.7 g of 238Pu in oxide form, the unit weight is 39.69 g, unit size is ɸ ¼ 2.54 cm 3.56 cm. The unit is rugged and very Table 6.12 Radioisotopes often considered as a battery fuel
Isotope 3
H Ni 90 Sr/90Y 147 Pm 210 Po 238 Pu 242 Cm 244 Cm 63
Average energy (keV) 5.7 17 200/930 5300 5500 5810
Half-life/ year 12 100 29/2 days 2.6234 0.38 88 161.4 days 18
Specific activity (Ci/g)
Specific power (W/g)
Power density (W/cm3)
9700 57 140
0.33 0.0067 0.98
4500 17
140 0.56
81
2.8
— 0.056 2.5 1.8 1210 11 882 38
6.9 Nuclear batteries
reliable. Larger units, called radioisotope thermoelectric generators (RTGs), have been used in many NASA missions. Usually, ceramic 238Pu radioisotope was used to provide heat with subsequent electricity production by thermoelectric effect. Such a unit does not have any moving parts it contains about 2.7 kg of fuel (133 kCi) producing power of 276 W. Dimensions of the unit were D ¼ 42 cm, L ¼ 114 cm, total weight-56 kg, lifetime over 20 years. More than 40 units have been flown by United States. Since the energy conversion mechanisms vary significantly between different nuclear types, therefore RTG is usually considered a performance standard for all nuclear battery types. NASA and the DOE have developed a new generation of such power systems that could be used for a variety of space missions. The newest RTG, called a multimission radioisotope thermoelectric generator (MMRTG), has been designed to operate on Mars and in the vacuum of space. The MMRTG has a flexible modular design capable of meeting the needs of a wide variety of missions, as it generates electrical power in smaller increments than previous generations of RTGs, about 110 W at launch. The design goals for the MMRTG included optimizing power levels over a minimum lifetime of 14 years and ensuring a high degree of safety. The MMRTG contains a total of 4.8 kg of plutonium dioxide (including 238Pu) that initially provides approximately 2000 W of thermal power and 110 W of electrical power when exposed to deep space environments. The thermoelectric materials (PbSnTe, TAGS and PbTe) have demonstrated extended lifetime and satisfactory performance capabilities. The MMRTG generator is about 64 cm in diameter, 66 cm tall and weighs about 45 kg, for more information about NASA’s use of Radioisotope Power Systems, see: rps. nasa.gov (NASA, 2013). Let us mention that the Curiosity Rover exploring planet Mars (big as a large car) is also using MMRTP, as an energy source. It is designed to run at least one Martian year, which is almost two Earth years. However, 238Pu is in a short supply, according to some reports there is only enough left to make three more nuclear batteries! The solution may-be in the navy system called stored chemical energy power systems (SCEPS) developed in the 1980s by engineers at Pennsylvania State University. SCEPS harnesses the chemical reaction of two energy-dense reactants that remain stored and separated until needed. In torpedoes, the system commonly holds its energy in reserve as a solid block of lithium and a tank of the inert gas sulphur hexafluoride. When triggered, the combustion reaction of the two materials creates heat that turns the weapon’s steam turbine to produce thousands of kilowatts of power. Nuclear battery has been used in pacemakers. Fuel was 3 Ci of 238Pu producing <1 mW power and delivering 100 mrem/year to the patient. Tracking the used Pu was difficult and that made regulators concerned so the nuclear battery in pacemakers was replaced by Li battery having 10 years life. Another attempt was made with 147Pm beta sources combined with custom designed Si n/p cells to produce nuclear batteries for heart pacemakers. The power sources were referred to as Betacel batteries. The Betacel was used to power cardiac pacemakers in over 100 patients during the 1970s (Olsen et al., 2017), but due to gamma radiation concerns; however, the lithium battery was able to capture the pacemaker market.
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Tritium has been used in construction of nuclear batteries for many different applications. For the technology used see https://citylabs.net/technology-overview/. The company City labs, Inc. product is the long-life (20 + years) Nanotritium™ battery for microelectronics. The battery is resistant to varying temperatures and other environmental conditions due to the robustness of the tritium decay process. The battery also serves as a vital component for the security and defence of electronics. As confirmed in independent testing, the City Labs battery is resistant to extreme temperature variance ( 55°C to +150°C), as well as extreme vibration and altitude, due to the robust architecture of City Labs’ proprietary technology and the use of tritium (Olsen et al., 2017). Another manufacturer of long-lived nuclear batteries powered by hydrogen isotopes for military applications is Widetronix (Bourzac, 2009). Their batteries could potentially power electrical circuits that protect military planes and missiles from tampering by destroying information stored in the systems, or by sending out a warning signal to a military centre. The batteries are expected to last for 25 years. The company is also working with medical-device makers to develop batteries that could last decades for implantable medical devices. It should be mentioned that Defense Science Board identified “nuclear batteries” as an essential technology for the US military in the 21st century (Weintz, 2014). As the most promising candidate betavoltaic batteries have been identified; from the outside the future betavoltaic batteries could look identical to present day military batteries. The most promising candidates today for betavoltaic batteries include 90 Sr, 63Ni and 3H. All three radionuclides emit beta radiation with almost no γ radiation, and they have long life times. While the strontium and nickel-based batteries are still mostly experimental, tritium-based batteries are available on the market. Also, The Army Research Laboratory has developed prototype nuclear batteries powered by tritium matching the Army’s existing BA-5590 battery pack in size and using the same connector. It can last for 13 years but the problem of disposal hazards remains (see also Seffers, 2013). An interesting attempt to use nuclear waste to generate electricity in a nuclearpowered battery has been recently reported. Namely, one of the largest nuclear waste products by volume is radionuclide 14C produced by irradiation of graphite blocks used in graphite-moderated reactors. The radionuclide 14C could theoretically be extracted and compressed into radioactive diamonds. These diamonds would then be incased in a regular, nonradioactive diamond, to provide shielding (Connor et al., 2016). The theoretical 14C diamond battery would have a lower energy density than chemical-based batteries like the lithium-ion batteries that power laptops and phones, and the alkaline batteries like AA batteries used in small electronics gadgets. A typical AA battery stores 13,000 J and is exhausted after about 24 h, while a 14C diamond battery would produce only 15 J per day but have a half-life of almost 6000 years (Bormashov et al., 2015). Thus far, research groups in Russia and Canada have produced prototype radioactive diamond batteries from 63Ni (Conca, 2016). Although diamond batteries are by no means a solution to the issue of nuclear waste
6.9 Nuclear batteries
recycling and storage, they offer a channel to put some of the waste to productive use as safe, durable, energy sources to eventually power exploration and medical electronics. Bormashov et al. (2018) recently reported a fabrication of betavoltaic battery prototype consisting of 200 single conversion cells based on Schottky barrier diamond diodes which have been vertically stacked with 24% 63Ni radioactive isotope. The maximum electrical output power of about 0.93 μW was obtained in total volume of 5 5 3.5 mm3. They used the ion-beam assisted lift-off technique to obtain conversion cells of minimal thickness comparable with the characteristic penetration length of beta-particles emitted by 63Ni isotope. The obtained value of 15 μm was limited by the mechanical strength of produced structures and process reliability. The fabricated prototype provided the output power density of about 10 μW/cm3, which is the best known value for nuclear batteries based on 63Ni radioisotope. Moreover, the long half-life of 63Ni isotope gives the battery specific energy of about 3300 mWh/g, an order of magnitude higher than the typical value of commercial chemical. In conclusion radioisotopes provide a high-energy density power source suitable for many applications. They are outstanding for small-scale power. According to Prelas et al. (2014) the physics of nuclear batteries does not support the objectives of miniaturization, high efficiency and high-power density, rather the physics considerations imply that nuclear batteries should be of moderate size and limited power density. In addition the supply of radioisotopes is limited and cannot support largescale commercialization. M€ oller and Wegener (2016) have presented the mathematical formalism and fundamental considerations for the use of ion accelerators for isotope production. A focus is put on the production of nuclear power sources to substitute 238Pu-based batteries. 20 MeV proton bombardment of bismuth produced α-emitting polonium isotopes with an energy efficiency of up to 0.031%. Some hours were required to produce a 1 W power source of the 2.9 year half-life α emitter 208Po. Apparently, the accelerator approach offers more flexibility for tailoring of nuclear products and less waste. One possible next-generation nuclear battery technology, as proposed by Kim and Kwon (2014), could be Aqueous Nuclear Battery, which is also known as water-based nuclear battery. It uses liquid medium for radiolysis, absorbing the kinetic energy of beta particles which is lost in betavoltaic cells. In their design using nanoporous titanium dioxide semiconductors coated in platinum, a high efficiency of 53.88% was reached at a potential of 0.9 V. Using an aqueous solution for radiolytic energy conversion results in higher energy level and lower temperature than using a solid-state material does. We are not discussed the controversial subjects of thorium powered cars and thorium plasma battery technology. The internet is full of websites discussing pro and cons, conspiracy theories, etc., instead, we caution the reader not to be eager to believe statements found on scientifically valueless websites.
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References Blix, H., 1997. Nuclear Energy in the 21st Century. Nuclear News, p. 34. September. Bormashov, V., Troschiev, S., Volkov, A., Tarelkin, S., Korostylev, E., Golovanov, A., Kuznetsov, M., Teteruk, D., Kornilov, N., Terentiev, S., Buga, S., Blank, V., 2015. Development of nuclear microbattery prototype based on Schottky barrier diamond diodes. Phys. Status Solidi A 212, 2539–2547. Bormashov, V.S., Troschiev, S.Y., Tarelkin, S.A., Volkov, A.P., Teteruk, D.V., Golovanov, A.V., Kuznatzov, M.S., Kornilov, N.V., Terentiev, S.A., Blank, V.D., 2018. High power density nuclear battery prototype based on diamond Schottky diodes. Diamond Relat. Mater. 84, 41–47. Boudard, A., Leroy, S., Volant, C., 1998. Spallation studies for nuclear waste transmutation. Nucl. Phys. News 8, 18. Bourzac, K., 2009. A 25-Year Battery: Long-Lived Nuclear Batteries Powered by Hydrogen Isotopes Are in Testing for Military Applications. November 17, 2009, MIT Technology Review. https://www.technologyreview.com/topic/sustainable-energy/. Bowman, C.D., et al., 1992. Nuclear Energy Generation and Waste Transmutation Using an Accelerator Driven Intense Thermal Neutron Source. Los Alamos Report LAUR-91-260 and Nuclear Instruments and Methods A320. p. 336. Bruhns, H., 1998. The EU Fusion Programme. Europhysics News, p. 206. November/ December. Campbell, D., 1998. Magnetic Confinement. Fusion. Europhysics News, p. 196. November/ December. Carminati, F., Klapisch, R., Revo´l, J.P., Roche, Ch., Rubio´, J.A., Rubia, C., 1993. An energy amplifier for cleaner and inexhaustible nuclear energy production driven by a particle beam accelerator. Preprint CERN/AT/93-47/ET (1993). Chan, C.Y., 1992. Radioactive Waste Management: An International Perspective. 3 IAEA Bulletin, p. 7. City Labs, Inc Corporate Office, 301 Civic Court: Homestead, FL 33030, USA. Conca, J., 2016. Radioactive Diamond Batteries: Making Good Use of Nuclear Waste. Forbes. 9 Dec 9, 2016. Connor, D.T., Martin, P.G., Scott, T.B., 2016. Airborne radiation mapping: overview and application of current and future aerial systems. Int. J. Remote Sens. 37, 5953–5987. Cowley, S., Reply to Hirsch, R. L. 2017. Necessary and sufficient conditions for practical fusion power. Phys. Today 70, 13–14. Dean, S.O., 2013. Search for the Ultimate Energy Source: A History of the U.S. Fusion Energy Program. Springer. Fells, I., 1998. The Need for Energy. Europhysics News, p. 193. November/December. Flakus, F.-N., Johnson, L.D., 1998. Binding agreements for nuclear safety: the global legal framework. IAEA Bull. 40, 21. Fountain, H., 2017. A Dream of Clean Energy at a Very High Price. The New york Times, p. 2017. March 27. Freidberg, J.P., 2007. Plasma Physics and Fusion Energy. Cambridge University Press, Vienna. Hannum, W.H., Marsh, G.E., Stanford, G.S., 2005. Smarter Use of Nuclear Waste. 2005. Scientific American. December issue. Hirsch, R.L., 2017. Necessary and sufficient conditions for practical fusion power. Phys. Today 70, 11–13.
References
Horton, C.W., Benkadda, S., 2015. ITER Physics. World Scientific Publishing Co. IAEA, 2001. Summary of the ITER Final Design Report. International Thermonuclear Experimental Reactor (ITER)-Engineering Design Activities (EDA) Documentation Series No. 22. International Atomic Energy Agency, Vienna. 2001. IAEA, 2016. Nuclear Power Reactors in the World. Reference Data Series No. 2, 2016 Edition. International Atomic Energy Agency, Vienna. 2016. IAEA, 2017a. Energy, Electricity and Nuclear Power Estimates for the Period up to 2050. Reference Data Series No. 1, 2017 Edition. International Atomic Energy Agency, Vienna 2017. IAEA, 2017b. Research Reactors for the Development of Materials and Fuels for Innovative Nuclear Energy Systems. IAEA Nuclear Energy Series No. NP-T-5.8. International Atomic Energy Agency, IAEA, Vienna. 2017. IAEA, 2018a. Status and Trends in Spent Fuel and Radioactive Waste Management. IAEA Nuclear Energy Series No. NW-T-1.14. IAEA, Vienna. 2018. IAEA, 2018b. Nuclear Power Reactors in the World. 2018 Edition. Reference Data Series No. International Atomic Energy Agency, Vienna. 2018. IAEA, 2018c. Operating Experience With Nuclear Power Stations in Member States, 2018 Edition. STI/PUB/1828. International Atomic Energy Agency, Vienna. 2018. IAEA, 2018d. Commissioning Guidelines for Nuclear Power Plants. IAEA Nuclear Energy Series No. NP-T-2.10. International Atomic Energy Agency, Vienna. 2018. IAEA, 2018e. Maintenance Optimization Programme for Nuclear Power Plants. IAEA Nuclear Energy Series No. NP-T-3.8. International Atomic Energy Agency, Vienna. 2018. IAEA, 2018f. Computer security of instrumentation and control Systems at Nuclear Facilities, technical guidance. IAEA Nuclear Security Series No. 33-T. International Atomic Energy Agency, Vienna. 2018. IAEA, 2018g. Physical Protection of Nuclear Material and Nuclear Facilities (implementation of INFCIRC/225/revision 5). IAEA Nuclear Security Series No. 27-G. International Atomic Energy Agency, Vienna. 2018. Jones, S.R., Williams, S.M., Smith, A.D., Gray, J., 1995. Review of Discharge History and Population Doses From the Sellafield Reprocessing Plant in Cumbria, UK: The Sellafield Environmental Assessment Model (SEAM), Reported at International Symposium on Environmental Impact of Radioactive Releases. Vienna, 8–12.05.1995, IAEA-SM-339/II. Juhn, P.-E., Kupitz, J., Cleveland, J., 1997. Advanced nuclear power plants—high lights of global development. IAEA Bull. 39, 13. Kershaw, P.J., Woodhead, D.S., Malcolm, S.J., Allington, D.J., Lovett, M.B., 1990. A sediment history of Sellafield discharges. J. Environ. Radioact. 12, 201–241. Kim, B.H., Kwon, J.W., 2014. Plasmon-assisted radiolytic energy conversion in aqueous solutions. Sci. Rep. 4, 5249 (9 p.). Kramer, D., 2017. NASA sees a future with nuclear power. Phys. Today 26–29 (December). Lindl, J., 1995. Development of the indirect-drive approach to inertial confinement fusion and the target physics basis for ignition and gain. Phys. Plasmas 2, 3933. M€oller, S., Wegener, T., 2016. Production of nuclear sources and nuclear batteries by proton irradiation. arXiv. 1608.05199v2 [Physics.Ins-Det], August 2016. NASA, 2013. Multi-Mission Radioisotope Thermoelectric Generator (MMRTG). NASA Facts. www.nasa.gov. October 2013. News, E.C., 2018. Europe’s Investment in the ITER Fusion Project: Mastering the Power of the Sun and the Stars. European Commission NEWS, Energy, Brussels. 13 April 2018.
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International Atomic Energy Agency Nuclear Research Reactors in the World, Dec. 1996 ed., Reference Data Series No. 3, Vienna, 1996. Oi, N., 1998. Plutonium challenges—changing dimensions of global cooperation. IAEA Bull. 40, 12. Olsen, L., Serralta, D., Cabauy, P., 2017. Betavoltaic Batteries: A historical Review. http:// citylabs.net/wp-content/uploads/2017/08/BetavoltaicHistory.pdf. Park, J., 2017. Review and Preview of Nuclear Battery Technology. Submitted as Coursework for PH241. Stanford University. Winter 2017. Prelas, M.A., Weaver, C.L., Watermann, M.L., Lukosi, E.D., Schott, R.J., Wisniewski, D.A., 2014. A review of nuclear batteries. Prog. Nucl. Energy 75, 117–148. Prelas, M., Boraas, M., De La Torre Aguilar, F., Seelig, J.D., Tchakoua Tchouaso, M., Wisniewski, D., 2016a. Potential Applications for Nuclear Batteries. In: Nuclear Batteries and Radioisotopes. Lecture Notes in Energy, vol. 56. Springer, Cham. Prelas, M., Boraas, M., De La Torre Aguilar, F. 2016b. Nuclear Batteries and Radioisotopes. (Lecture Notes in Energy). first ed. 2016 ed.Springer, Cham. C. Rubbia et al., Conceptual design of a fast neutron operated high power energy amplifier, Preprint CERN/AT/95-44 (ET) (1995). Seffers, G.I., 2013. Radioisotope Research May Revolutionize Battlefield Batteries. SIGNAL December 1. AFCEA International, Fairfax, VA Semenov, B.A., Oi, N., 1993. Nuclear fuel cycles: adjusting to new realities. IAEA Bull. 3, 2. Stather, J.W., Wrixon, A.D., Simmonds, J.R., 1984. The Risks of Leukaemia and Other Cancers in Seascale From Radiation Exposure, NRPB-R171. HMSO. Stather, J.W., Dionian, J., Brown, J., Fell, T.P., Muirhead, C.R., 1986. The Risks of Leukaemia and Other Cancers in Seascale From Radiation Exposure: Addendum to Report R171. NRPB-R171 Addendum. HMSO. Stather, J.W., Clarke, R.H., Duncan, K.P., 1988. The risk of childhood leukaemia near nuclear establishments. NRPB-R215. HMSO. Takizuka, T., 1994. Proceedings of the 8th Journees SATURNE, Saclay, May 5–6. Tirone, J., 2017. This Company is Designing Nuclear “Batteries” for Towns and Industry. Source Urenco Ltd, Stoge Poges, England. Van Tuyle, G.J., et al., 1993. Nucl. Technol. 101, 1. Weintz, S., 2014. The US Military is Working on Nuclear Batteries. https://medium.com/ war-is-boring/powering-the-militarys-future-batteries-nuclear-style-472975d7de8. Wesson, J.A., 1997. Tokamaks, second ed. Oxford University Press. WNA, 2018. Processing of Used Nuclear Fuel. World Nuclear Association. http://worldnuclear.org/information-library/nuclear-fuel-cycle/fuel-recycling/processing-of-usednuclear-fuel.
Further reading Arkhipov, V., 1997. Future nuclear energy systems: generating electricity burning wastes. IAEA Bull. 39, 30. IAEA Bulletin, 39 (1997) 13, International Atomic Energy Agency, Vienna, Austria. Safety Series No. 101, 1990. Operational Radiation Protection: A Guide to Optimization. Safety Series No. 105, 1990. The Regulatory Process for the Decommissioning of Nuclear Facilities.
Further reading
Safety Series No. 107, 1992. Radiation Safety of Gamma and Electron Irradiation Facilities. Safety Series No. 108, 1992. Design and Operation of Radioactive Waste Incineration Facilities. Safety Series No. 109, 1994. Intervention Criteria in a Nuclear or Radiation Emergency. Safety Series No. 110, 1993. The Safety of Nuclear Installations. Safety Series No. 111, 1995. The Principles of Radioactive Waste Management. Safety Series No. 111 -G-1.1, 1994. Classification of Radioactive Waste. Safety Series No. 111-G-3.1, 1994. Siting of Near Surface Disposal Facilities. Safety Series No. 111-G-4.1, 1994. Siting of Geological Disposal Facilities. Safety Series No. 111-S-1: Establishing a National System for Radioactive Waste Management (1995). Safety Series No. 112, 1994. Compliance Assurance for the Safe Transport of Radioactive Material. Safety Series No. 113, 1994. Quality Assurance for the Safe Transport of Radioactive Material. Safety Series No. 120, 1996. Radiation Protection and the Safety of Radiation Sources. Safety Series No. 35-G1, 1994. Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report. Safety Series No. 35-G2, 1994. Safety in the Utilization and Modification of Research. Safety Series No. 50-C-Q, 1996. Quality Assurance for Safety in Nuclear Power Plants and Other Nuclear Installations. Safety Series No. 50-C-S (Rev. 1), 1988. Code on the Safety of Nuclear Power Plants: Siting. Safety Series No. 50-SG-010, 1985. Core Management and Fuel Handling for Nuclear Power Plants. Safety Series No. 50-SG-012, 1994. Periodic Safety Review of Operational Nuclear Power Plants. Safety Series No. 50-SG-06, 1982. Preparedness of the Operating Organization (Licensee) for Emergencies at Nuclear Power Plants. Safety Series No. 50-SG-D1, 1979. Safety Functions and Component Classification for BWR, PWR and PTR. Safety Series No. 50-SG-D12, 1985. Design of the Reactor Containment Systems in Nuclear Power Plants. Safety Series No. 50-SG-D2 (Rev. 1), 1992. Fire Protection in Nuclear Power Plants. Safety Series No. 50-SG-D4, 1980. Protection against Internally Generated Missiles and Their Secondary Effects in Nuclear Power Plants. Safety Series No. 50-SG-D5 (Rev. 1), 1996. External Man-induced Events in Relation to Nuclear Power Plant Design. Safety Series No. 50-SG-D7, 1991. Emergency Power Systems at Nuclear Power Plants. Safety Series No. 50-SG-D9, 1985. Design Aspects of Radiation Protection for Nuclear Power Plants. Safety Series No. 50-SG-G4 (Rev: 1), 1996. Inspection and Enforcement by the Regulatory Body for Nuclear Power Plants. Safety Series No. 50-SG-G6, 1982. Preparedness of Public Authorities for Emergencies at Nuclear Power Plants. Safety Series No. 50-SG-G8, 1982. Licenses for Nuclear Power Plants: Content, Format and Legal Considerations. Safety Series No. 50-SG-G9, 1984. Regulations and Guides for Nuclear Power Plants.
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Safety Series No. 50-SG-O1 (rev. 1), 1991. Staffing of Nuclear Power Plants and the Recruitment, Training and Authorization of Operating Personnel. Safety Series No. 50-SG-O4, 1980. Commissioning Procedures for Nuclear Power Plants. Safety Series No. 50-SG-Q1, 1996. Establishing and Implementing a Quality Assurance Programme. Safety Series No. 50-SG-Q10, 1996. Quality Assurance in Design. Safety Series No. 50-SG-Q11, 1996. Quality Assurance in Construction. Safety Series No. 50-SG-Q12, 1996. Quality Assurance in Commissioning. Safety Series No. 50-SG-Q13, 1996. Quality Assurance in Operation. Safety Series No. 50-SG-Q14, 1996. Quality Assurance Decommissioning. Safety Series No. 50-SG-Q2, 1996. Non-conformance Control and Corrective Actions. Safety Series No. 50-SG-Q3, 1996. Document Control and Records. Safety Series No. 50-SG-Q4, 1996. Inspection and Testing for Acceptance. Safety Series No. 50-SG-Q5, 1996. Assessment of the Implementation of the Quality Assurance Programme. Safety Series No. 50-SG-Q6, 1996. Quality Assurance in the Procurement of Items and Services. Safety Series No. 50-SG-Q7, 1996. Quality Assurance in Manufacturing. Safety Series No. 50-SG-Q8, 1996. Quality Assurance in Research and Development. Safety Series No. 50-SG-Q9, 1996. Quality Assurance in Siting. Safety Series No. 50-SG-S1 (Rev. 1), 1991. Earthquakes and Associated Topics in Relation to Nuclear Power Plants Siting. Safety Series No. 50-SG-S8, 1986. Safety Aspects of the Foundations of Nuclear Power Plants. Safety Series No. 50-SG-S9, 1984. Site Survey for Nuclear Power Plants. Safety Series No. 79, 1986. Design of Radioactive Waste Management Systems at Nuclear Power Plants. Safety Series No. 90, 1989. The Application of the Principles for Limiting Releases of Radioactive Effluents in the Case of the Mining and Milling of Radioactive Ores. Safety Series No. 93: System of Reporting Unusual Events in Nuclear Power Plants, 1989. Safety Series No. 96, 1989. Guidance for Regulation of Underground Repositories for Disposal of Radioactive Wastes. Safety Series No. 98, 1989. On-Site Habitability in the Event of an Accident at a Nuclear Facility. Semenov, B.A., 1992. Disposal of spent fuel and high-level radioactive waste: building international consensus. IAEA Bull. 3, 2.