JPS62168091A - Nuclear reactor - Google Patents
Nuclear reactorInfo
- Publication number
- JPS62168091A JPS62168091A JP61007931A JP793186A JPS62168091A JP S62168091 A JPS62168091 A JP S62168091A JP 61007931 A JP61007931 A JP 61007931A JP 793186 A JP793186 A JP 793186A JP S62168091 A JPS62168091 A JP S62168091A
- Authority
- JP
- Japan
- Prior art keywords
- fuel
- core
- water
- reactor
- conversion
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Pending
Links
- 239000000446 fuel Substances 0.000 claims description 110
- 238000006243 chemical reaction Methods 0.000 claims description 30
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 23
- 239000000463 material Substances 0.000 claims description 22
- 239000003758 nuclear fuel Substances 0.000 claims description 2
- 229910052751 metal Inorganic materials 0.000 description 14
- 239000002184 metal Substances 0.000 description 14
- 238000010586 diagram Methods 0.000 description 13
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 12
- 238000005253 cladding Methods 0.000 description 11
- OKTJSMMVPCPJKN-UHFFFAOYSA-N Carbon Chemical compound [C] OKTJSMMVPCPJKN-UHFFFAOYSA-N 0.000 description 8
- 229910052770 Uranium Inorganic materials 0.000 description 8
- 239000002826 coolant Substances 0.000 description 8
- 238000010521 absorption reaction Methods 0.000 description 7
- 238000002485 combustion reaction Methods 0.000 description 7
- 229910002804 graphite Inorganic materials 0.000 description 7
- 239000010439 graphite Substances 0.000 description 7
- 229910052778 Plutonium Inorganic materials 0.000 description 6
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 description 6
- 230000005496 eutectics Effects 0.000 description 5
- OYEHPCDNVJXUIW-FTXFMUIASA-N 239Pu Chemical compound [239Pu] OYEHPCDNVJXUIW-FTXFMUIASA-N 0.000 description 4
- 229920000049 Carbon (fiber) Polymers 0.000 description 4
- 239000004917 carbon fiber Substances 0.000 description 4
- VNWKTOKETHGBQD-UHFFFAOYSA-N methane Chemical compound C VNWKTOKETHGBQD-UHFFFAOYSA-N 0.000 description 4
- WZECUPJJEIXUKY-UHFFFAOYSA-N [O-2].[O-2].[O-2].[U+6] Chemical compound [O-2].[O-2].[O-2].[U+6] WZECUPJJEIXUKY-UHFFFAOYSA-N 0.000 description 3
- 238000002844 melting Methods 0.000 description 3
- 230000008018 melting Effects 0.000 description 3
- 238000000034 method Methods 0.000 description 3
- 239000010935 stainless steel Substances 0.000 description 3
- 229910001220 stainless steel Inorganic materials 0.000 description 3
- 229910000439 uranium oxide Inorganic materials 0.000 description 3
- JFALSRSLKYAFGM-OIOBTWANSA-N uranium-235 Chemical compound [235U] JFALSRSLKYAFGM-OIOBTWANSA-N 0.000 description 3
- DGAQECJNVWCQMB-PUAWFVPOSA-M Ilexoside XXIX Chemical group C[C@@H]1CC[C@@]2(CC[C@@]3(C(=CC[C@H]4[C@]3(CC[C@@H]5[C@@]4(CC[C@@H](C5(C)C)OS(=O)(=O)[O-])C)C)[C@@H]2[C@]1(C)O)C)C(=O)O[C@H]6[C@@H]([C@H]([C@@H]([C@H](O6)CO)O)O)O.[Na+] DGAQECJNVWCQMB-PUAWFVPOSA-M 0.000 description 2
- 239000011248 coating agent Substances 0.000 description 2
- 238000000576 coating method Methods 0.000 description 2
- 238000001816 cooling Methods 0.000 description 2
- 239000000498 cooling water Substances 0.000 description 2
- 230000004927 fusion Effects 0.000 description 2
- 229910052739 hydrogen Inorganic materials 0.000 description 2
- 239000001257 hydrogen Substances 0.000 description 2
- 239000008188 pellet Substances 0.000 description 2
- 229910052708 sodium Inorganic materials 0.000 description 2
- 239000011734 sodium Substances 0.000 description 2
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 description 1
- OYEHPCDNVJXUIW-AHCXROLUSA-N Plutonium-240 Chemical compound [240Pu] OYEHPCDNVJXUIW-AHCXROLUSA-N 0.000 description 1
- RTAQQCXQSZGOHL-UHFFFAOYSA-N Titanium Chemical compound [Ti] RTAQQCXQSZGOHL-UHFFFAOYSA-N 0.000 description 1
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 description 1
- 230000002159 abnormal effect Effects 0.000 description 1
- 229910045601 alloy Inorganic materials 0.000 description 1
- 239000000956 alloy Substances 0.000 description 1
- 229910052799 carbon Inorganic materials 0.000 description 1
- 238000003763 carbonization Methods 0.000 description 1
- 239000002131 composite material Substances 0.000 description 1
- 238000012790 confirmation Methods 0.000 description 1
- 230000003247 decreasing effect Effects 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 239000007789 gas Substances 0.000 description 1
- PCHJSUWPFVWCPO-UHFFFAOYSA-N gold Chemical compound [Au] PCHJSUWPFVWCPO-UHFFFAOYSA-N 0.000 description 1
- 239000010931 gold Substances 0.000 description 1
- 229910052737 gold Inorganic materials 0.000 description 1
- 150000002431 hydrogen Chemical class 0.000 description 1
- 238000004519 manufacturing process Methods 0.000 description 1
- 239000000155 melt Substances 0.000 description 1
- 238000013508 migration Methods 0.000 description 1
- 230000005012 migration Effects 0.000 description 1
- 230000003647 oxidation Effects 0.000 description 1
- 238000007254 oxidation reaction Methods 0.000 description 1
- 230000005855 radiation Effects 0.000 description 1
- 230000009257 reactivity Effects 0.000 description 1
- 229910052719 titanium Inorganic materials 0.000 description 1
- 239000010936 titanium Substances 0.000 description 1
- 230000001052 transient effect Effects 0.000 description 1
- 229910052726 zirconium Inorganic materials 0.000 description 1
Classifications
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Landscapes
- Analysing Materials By The Use Of Radiation (AREA)
- Treatment Of Water By Oxidation Or Reduction (AREA)
Abstract
(57)【要約】本公報は電子出願前の出願データであるた
め要約のデータは記録されません。(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.
Description
【発明の詳細な説明】
〔発明の利用分野〕
本発明は、軽水減速型原子炉に係り、特に中性子平均エ
ネルギを高くして親核燃料物質より核分裂性物質への転
換量を大きくする原子炉に関する。[Detailed Description of the Invention] [Field of Application of the Invention] The present invention relates to a light water-moderated nuclear reactor, and particularly to a nuclear reactor that increases the average neutron energy to increase the amount of conversion from nuclear-friendly fuel material to fissile material. .
軽水減速型原子炉（以下、軽水炉と記す）で生成される
プルトニウムの有効利用法として、親核燃料物質である
ウラン２３８から核分裂性物質（プルトニウム２３９）
への転換を高くし、装荷したプルトニウムの減少を小さ
くしようとする高転換型軽水炉（以下高転換炉と記す）
が提案されている＊ Ｎｕｃｌ、、Ｔｅｃｈｎｏｌ、、
旦９ 、２１２（１９８２年）における０１ｄｅｋｏｐ
らによる”Ｇｅｎｅｒａｌ ｆｅａｔｕｒｅ ｏｆａｄ
ｖａｎｃｅｄ ｐｒｅｓｓｓｕｒｉｚｅｄ ｗａｔｅｒ
ｒｅａｃｔｏｒｓ ｔｓｉｔｈｉｍｐｒｏｖｅｄ ｆ
ｕｅｌｕｔｉｌｉｚａｔｉｏｎ”と題する文献で高転換
炉の例が発表されている。第２図に現行ＰＷＲと比較し
て高転換炉の燃料集合体の外観と燃料棒配列を示す。親
核燃料物質であるウラン２３８から核分裂性物質である
プルトニウム２３９への転換を多くするためには、中性
子の平均エネルギーを高くし、中性子のウラン２３８へ
の共鳴吸収を大きくする必要がある。このためには、燃
料に対する減速材、すなわち水の体積比を大きくシ、中
性子の水による減速を小さくする必要があり、第２図に
示すように高転換炉では、この水対燃料対積比を現行の
約２．０から０．５まで下げた稠密六方格子系の燃料配
置としているにのように、水対燃料体積比を小さくする
と冷却材の流路断面積が小さくなり、炉心での圧力損失
の増大をまねく。As an effective use of plutonium produced in light water-moderated nuclear reactors (hereinafter referred to as light water reactors), it is possible to convert the pro-nuclear fuel material uranium-238 into fissile material (plutonium-239).
A high-conversion light water reactor (hereinafter referred to as a high-conversion reactor) that aims to increase the conversion to plutonium and reduce the loss of loaded plutonium.
has been proposed * Nucl, , Technol, .
01dekop in Dan9, 212 (1982)
“General feature ofad” by et al.
vanced presssurized water
reactors tsithimproved f
An example of a high conversion reactor has been published in a document entitled ``PWR. In order to increase the conversion from plutonium-239 to the fissile material plutonium-239, it is necessary to increase the average energy of neutrons and increase the resonance absorption of neutrons into uranium-238. In other words, it is necessary to increase the volume ratio of water and reduce the moderation of neutrons by water. When the water-to-fuel volume ratio is reduced, as in the case where the fuel is arranged in a dense hexagonal lattice system, the cross-sectional area of the coolant flow path becomes smaller, leading to an increase in pressure loss in the core.
第２図に示した例では、流路断面積が現行炉の約１／４
となるため炉心圧損が約４気圧と現行炉の４倍近くにな
り冷却水循環ポンプ動力の大幅な増大をもたらすことと
なって高転換炉の問題点の１つとなる。さらに、緊急時
に炉心を冷却する場合、冷却材流路が狭いため炉心外部
から冷却水が入りにくくなり緊急時炉心冷却の確認が大
きな課頭となっている。水対ウラン体積比を小さくする
ことにより、転換比（核分裂性物質が消滅する量に対す
るウラン２３８、プルトニウム２４０などの親核燃料物
質から核分裂性物質へ転換する量の比）は高くできるが
、冷却材流路断面積が狭くなり、上述のような伝熱・流
動上の問題が生じる。In the example shown in Figure 2, the flow passage cross-sectional area is approximately 1/4 of that of the current furnace.
As a result, the core pressure drop will be approximately 4 atm, nearly four times that of the current reactor, resulting in a significant increase in the power of the cooling water circulation pump, which is one of the problems of high conversion reactors. Furthermore, when cooling the core in an emergency, the narrow coolant flow path makes it difficult for cooling water to enter from outside the core, making confirmation of core cooling in an emergency a major issue. By reducing the water to uranium volume ratio, the conversion ratio (the ratio of the amount of nuclear-friendly fuel materials such as uranium-238 and plutonium-240 converted to fissile material to the amount of fissile material destroyed) can be increased, but the coolant The cross-sectional area of the flow path becomes narrower, causing problems in heat transfer and flow as described above.
本発明の目的は、水対燃料体積比を大幅に減少すること
なく、中性子の平均エネルギーを高くし親核燃料物質か
ら核分裂性物質への転換量を大きくできる軽水炉を提供
することにある。An object of the present invention is to provide a light water reactor that can increase the average energy of neutrons and increase the amount of conversion from nuclear-friendly fuel material to fissile material without significantly reducing the water-to-fuel volume ratio.
軽水炉では、中性子減速能が元素中量も大きい水素を含
む軽水が、減速材および冷却材に使用されている。した
がって、炉心中の中性子の平均エネルギーを高めるため
には、水素原子数密度対燃料原子数密度比（以下Ｈ／Ｕ
と記す）を大幅に減少させる必要がある。このＨ／Ｕを
小さくする手段として従来は、水対燃料の体積比（以下
Ｖ Ｍ ／ＶＦと記す）を小さくする方法がとられてい
た。In light water reactors, light water containing hydrogen, which has a high neutron moderating ability and a large elemental content, is used as a moderator and coolant. Therefore, in order to increase the average energy of neutrons in the reactor core, the hydrogen atomic number density to fuel atomic number density ratio (hereinafter referred to as H/U
) needs to be significantly reduced. Conventionally, a method of reducing the H/U has been to reduce the water-to-fuel volume ratio (hereinafter referred to as V M /VF).
第２図に示す例では、ＶＭ／ＶＦを現行の約１／４とす
ることでＨ／Ｕを減少させて中性子の平均エネルギーを
高くし、転換比を０．９ 以上としている。In the example shown in FIG. 2, by setting VM/VF to about 1/4 of the current value, H/U is decreased, the average energy of neutrons is increased, and the conversion ratio is made to be 0.9 or more.
本発明では、Ｈ／Ｕを小さくする手段として燃料原子数
密度を大きくする方法を新に軽水炉に適用するものであ
る。燃料原子数密度を大きくする具体的方法として全属
燃料を使用した軽水炉炉心を本発明に提供する。全属燃
料の密度は１９．０ｇ ／ ｃｉと酸化物燃料の密度約
１０．４ｇ／ａＩ？と比べて約２倍となっている。した
がって、燃料の寸法形状が同一で、かつ水対燃料体積比
も同一とすると、全属燃料を使用した場合のＨ／Ｕは酸
化物燃料を使用した場合の−となる。このように、全属
燃料を軽水炉心に使用した場合、水対燃料体積比を同一
とすると、酸化物燃料を使用した場合より転換比をより
大きくでき、逆に同じ転換比を得るには水対燃料体積比
、すなわち冷却材流路断面積をより大きくすることが可
能となる。In the present invention, a method of increasing the fuel atomic number density as a means of reducing H/U is newly applied to light water reactors. The present invention provides a light water reactor core using all metal fuels as a specific method for increasing the fuel atomic number density. The density of all metal fuels is 19.0 g/ci and the density of oxide fuels is about 10.4 g/aI? It is approximately twice as large as the previous year. Therefore, assuming that the dimensions and shapes of the fuels are the same and the water-to-fuel volume ratio is also the same, H/U when all metal fuels are used is - when oxide fuels are used. Thus, when all-metal fuels are used in a light water reactor core, the conversion ratio can be larger than when using oxide fuels, given the same water-to-fuel volume ratio; conversely, to obtain the same conversion ratio, water It becomes possible to further increase the fuel volume ratio, that is, the cross-sectional area of the coolant flow path.
以下、本発明を実施例にて詳細に説明する。 Hereinafter, the present invention will be explained in detail with reference to Examples.
第１図は、本発明による炉心構成の例であり、炉心１は
、燃料棒３および中性子吸収棒４が六角格子状に配列さ
れた六角水平断面を有す燃料集合体２から構成されてい
る。燃料体３は、ウランとプルトニウムの合金燃料３１
ガステンレスの被覆管で覆われている。燃料３１と被覆
３３との間には、熱あるいは中性子照射による燃料の膨
張を吸収するためのギャップ３２を設けている。全属燃
料は中性子照射による膨張が酸化物燃料より大きいため
、本実施例では約３倍ギャップを広げ、実効的な燃料の
密度を理論密度の約８５％である１６ｇ／ａ＋？として
いる。Ａ−Ｅ Ｗａｉｔｅｒ等による“Ｆａｓｔ Ｂｒ
ｅｅｄｅｒ Ｒｅａｃｔｏｒｓ” 、 Ｐｅｒｇａｍｏ
ｎ Ｐｒｅｓｓ。FIG. 1 shows an example of the core configuration according to the present invention, and the core 1 is composed of a fuel assembly 2 having a hexagonal horizontal cross section in which fuel rods 3 and neutron absorption rods 4 are arranged in a hexagonal lattice. . The fuel body 3 is an alloy fuel 31 of uranium and plutonium.
Covered with gas stainless steel cladding. A gap 32 is provided between the fuel 31 and the coating 33 to absorb expansion of the fuel due to heat or neutron irradiation. Since the expansion of all-metal fuels due to neutron irradiation is greater than that of oxide fuels, in this example the gap is widened by about three times, and the effective density of the fuel is set to 16 g/a+?, which is about 85% of the theoretical density. It is said that “Fast Br” by A-E Waiter et al.
eeder Reactors”, Pergamo
nPress.
Ｐ２Ｏ３−４１６では、ナトリウム冷却高速増殖実験炉
（ＥＢＲ−ｎ）では燃料の実効密度が理論密度の約７５
％となるようにギャップを広げた場合、燃焼度１００Ｇ
Ｗｄ／を以上まで問題がなかったことが記されている。In P2O3-416, the effective density of the fuel is approximately 75% of the theoretical density in the sodium-cooled experimental fast breeder reactor (EBR-n).
If the gap is widened to %, the burnup will be 100G.
It is written that there were no problems up to Wd/.
第３図は、第２図に示した燃焼に伴う中性子増倍率の変
化を酸化物燃料凝使用した場合と比較して示している。FIG. 3 shows the change in the neutron multiplication factor accompanying the combustion shown in FIG. 2 in comparison with the case where oxide fuel is used condensed.
両者で燃料棒の寸法形状は等しいとしている。The size and shape of the fuel rods are assumed to be the same in both cases.
（燃料ペレットの径は、両者で等しくし、全属燃料ペレ
ットについてはギャップを広げた分、密度を１９ｇ／ｃ
ｄから１６ｇ／ａＩｌと下げている。）また、燃料とし
て、ウラン−２３５が０ 、２ ｗ ／ 。(The diameter of the fuel pellets is the same for both, and the density of the all-genus fuel pellets is 19g/c due to the widening of the gap.)
d to 16 g/aIl. ) Also, as a fuel, uranium-235 is 0.2 w/.
残在する減損ウランに、核分裂性プルトニウム（プルト
ニウム−２３９，２４１）の重量百分率（以下富化度と
記す）が、全属燃料に対して５．５ｗ１０、酸化物燃料
に対して７ 、５ ｗ ／ ｏ のものを使用してい
る。さらに、水対燃料体積比は、全属燃料が０．８５．
酸化物燃料が０．５０である。In the remaining depleted uranium, the weight percentage (hereinafter referred to as enrichment) of fissile plutonium (Plutonium-239,241) is 5.5w10 for all metal fuels and 7.5w for oxide fuels. /o is used. Furthermore, the water to fuel volume ratio is 0.85 for all metal fuels.
Oxide fuel is 0.50.
第４図は、同一条件で燃焼に伴う積分転換比（燃焼度０
における核分裂性物質の原子数密度と燃焼度各点におけ
る核分裂性物質の原子数密度の比）の変化を、全属燃料
を使用した場合と酸化物燃料を使用した場合とで比較し
て示している。第１図と第３図に示されるように、燃料
取出し燃焼度を４５ＧＷ／ｄｔとしたとき酸化物燃料を
使用の炉心と比較して、全属燃料使用の炉心では水対燃
料体積比が約１．７倍、プルトニウム富化度が約７０％
とできる。水対燃料体積比を１．７倍とできるため、冷
却材流路断面積が７０％広がり、炉心圧損は約３５％に
低減でき、現行軽水炉心の圧損の１．４倍程度に抑える
ことが可能である。Figure 4 shows the integral conversion ratio (burnup 0) accompanying combustion under the same conditions.
The graph shows the changes in the ratio of the atomic number density of fissile material at each point of burnup between when all-genus fuel is used and when oxide fuel is used. There is. As shown in Figures 1 and 3, when the fuel removal burnup is 45 GW/dt, compared to a core using oxide fuel, a core using all metal fuels has a water-to-fuel volume ratio of approximately 1.7 times, plutonium enrichment approximately 70%
It can be done. Since the water-to-fuel volume ratio can be increased to 1.7 times, the cross-sectional area of the coolant flow path can be expanded by 70%, and the core pressure drop can be reduced to about 35%, which is about 1.4 times the pressure drop of the current light water core. It is possible.
第５図は第２の実施例を示す図である。本実施例では全
属燃料３１と被覆管３３との間にグラファイトのうすい
層を設けた燃料体を炉心に使用している。全属燃料は溶
融温度は１１５０℃であるが、７００℃〜８００℃で被
覆管と共融を起す。通常運転時の全属燃料の中心温度は
約５００℃で、燃料表面温度も４５０℃程度であるから
被覆管との共融の恐れは無い。しかし、異常時の炉心出
力過渡変化時に燃料表面温度が共融点に達する可能性も
ある。本実施例によれば、燃料と被覆管との間に耐熱性
の高いグラファイトの層が存在するため燃料表面温度が
共融点を越し、あるいは万一燃料が融溶した場合でも、
燃料と被覆管とが直接接することが無いため燃料の健全
性が確保される。グラファイトは熱中性子のミクロ吸収
断面積が約３ミリバーンと被覆材として使われるジルコ
ニウムのミクロ吸収断面積の約６０分の１であり、グラ
ファイトを入れることによる中性子吸収は無視できる。FIG. 5 is a diagram showing a second embodiment. In this embodiment, a fuel body in which a thin layer of graphite is provided between the all-metal fuel 31 and the cladding tube 33 is used in the reactor core. The melting temperature of all metal fuels is 1150°C, but eutectic melting occurs with the cladding at 700°C to 800°C. During normal operation, the center temperature of all fuels is about 500°C, and the fuel surface temperature is also about 450°C, so there is no fear of eutectic fusion with the cladding tube. However, there is a possibility that the fuel surface temperature may reach the eutectic point during transient changes in core power during abnormal conditions. According to this embodiment, since there is a highly heat-resistant graphite layer between the fuel and the cladding tube, even if the fuel surface temperature exceeds the eutectic point or the fuel melts, the
Since the fuel and the cladding do not come into direct contact, the integrity of the fuel is ensured. Graphite has a thermal neutron micro absorption cross section of about 3 milliburn, which is about 1/60th of the micro absorption cross section of zirconium used as a coating material, so the neutron absorption due to the inclusion of graphite can be ignored.
また、グラファイトの熱伝導率は７００℃で３５〜７０
Ｗ／ｍ−にであり金ＪｉＬ燃料と同程度であるのでグラ
ファイトを入れることによる燃料温度の上昇は問題にな
らない。In addition, the thermal conductivity of graphite is 35 to 70 at 700℃.
W/m-, which is about the same as that of gold JiL fuel, so the increase in fuel temperature due to the addition of graphite does not pose a problem.
第６図は第３の実施例を示す図である。本実施例では全
属燃料の表面をチッ化、あるいは炭化処理して表面にＵ
ＮあるいはＵＣのうすい層を形成させる。また、中心部
に中空部３６を設ける。FIG. 6 is a diagram showing a third embodiment. In this example, the surface of all metal fuels is treated with nitridation or carbonization to coat the surface with U.
Form a thin layer of N or UC. Further, a hollow portion 36 is provided in the center.
ＵＮ、ＵＣは共に融点が約２８００℃なので、万一全属
燃料が溶融して１１５０℃以上になっても被覆管と直接
接する可能性は小さくできる。さらに、中空部３６を設
けたため、熱あるいは中性子照射による膨張を吸収でき
、ギャップ３２の厚さは酸化燃料と同程度にうすくでき
、また、燃料最高温度も低下でき、燃料健全性の一層の
向上がはかれる。Both UN and UC have melting points of about 2800°C, so even if all the metal fuels melt and reach a temperature of 1150°C or higher, the possibility of them coming into direct contact with the cladding tube can be reduced. Furthermore, since the hollow part 36 is provided, expansion due to heat or neutron irradiation can be absorbed, the thickness of the gap 32 can be made as thin as that of the oxidized fuel, and the maximum fuel temperature can also be lowered, further improving fuel integrity. is measured.
第７図は第４の実施例であり１本発明を高転換・バーナ
ー型軽水炉心に適用した例である。高転換・バーナー型
軽水炉心（以下高転換・バーナー炉心と記す）５は、高
転換領域２１とバーナー領域５２の２領域から構成され
ている。高転換領域には、燃料体本数密度を高くして水
対燃料体積比を約０．９ とした燃料集合体６が装荷さ
れている。FIG. 7 shows a fourth embodiment, in which the present invention is applied to a high conversion burner type light water reactor core. The high conversion/burner type light water reactor core (hereinafter referred to as high conversion/burner core) 5 is composed of two regions: a high conversion region 21 and a burner region 52. The high conversion region is loaded with a fuel assembly 6 having a high fuel body number density and a water-to-fuel volume ratio of about 0.9.
燃料棒３には本発明を適用したウラン−２３５゜濃縮度
が４．５ｗ１０の金属ウランが使用されている。高転換
領域では水対燃料体積比か現行軽水炉の約１／２と低く
親核燃料物質から核分裂性物質への転換が大きく燃料体
３内には核分裂性物質プルトニウム−２３９が蓄積され
ていく。燃料寿命中期に集合体６を高転換領域か取出し
、燃料体本数密度を集合体６の約１／２と低くした燃料
集合体７を再構成してバーナー領域に装荷する。バーナ
ー領域では水対燃料体積比が約２．２ となり熱中性子
利用率が高くなり燃料棒３内の核分裂性物質が効率良く
燃される。燃料として金属ウランを使用した場合と酸化
物燃料を使用した場合の燃焼に伴う中性子増倍率の変化
の比較を第８図に、積分転換比の比較を第９図に示す。The fuel rods 3 are made of metallic uranium having a uranium-235 degree enrichment of 4.5w10 to which the present invention is applied. In the high conversion region, the water to fuel volume ratio is low, about 1/2 that of the current light water reactor, and the conversion from nuclear-friendly fuel material to fissile material is large, and the fissile material plutonium-239 is accumulated in the fuel body 3. At the middle of the fuel life, the assembly 6 is removed from the high conversion region, the fuel assembly 7 is reconstituted with a lower fuel assembly number density of about 1/2 that of the assembly 6, and loaded into the burner region. In the burner region, the water to fuel volume ratio is about 2.2, the thermal neutron utilization rate is high, and the fissile material in the fuel rods 3 is burned efficiently. FIG. 8 shows a comparison of changes in neutron multiplication factor due to combustion when metallic uranium is used as fuel and when oxide fuel is used, and FIG. 9 shows a comparison of integral conversion ratio.
ただし、金属ウラン使用と酸化ウラン使用の両ケースで
、水対燃料体積比は高転換領域、バーナー領域ともに等
しくしているが、ウラン−２３５の濃縮度は金属ウラン
では４．５ｗ１０．酸化ウランでは５．ＯＷ ／ Ｏと
している。第７図に示されるように金属ウラン使用の場
合４．５ｗ１０の濃縮度で５．Ｏｗ ／ ｏの酸化ウラ
ン使用の場合とほぼ同じ反応度変化を得られたのは、金
属ウラン使用により中性子の平均エネルギーが高くなり
、第８図に示されるように核分裂性物質生成の割合が酸
化ウラン使用時より大きくなるためである。However, in both cases of using metallic uranium and using uranium oxide, the water to fuel volume ratio is the same in both the high conversion region and the burner region, but the enrichment of uranium-235 is 4.5w10. 5 for uranium oxide. It is set as OW/O. As shown in Figure 7, when metallic uranium is used, the enrichment level is 4.5w10. The reason why we were able to obtain almost the same change in reactivity as when using uranium oxide in Ow/O is because the average energy of neutrons increases due to the use of metallic uranium, and as shown in Figure 8, the rate of fissile material production is lower than that of oxidation. This is because it becomes larger than when using uranium.
第９図は第５の実施例を示す図であり、被覆管３２とし
て複合材料であるカーボンファイバーから作成された材
料を使朋した例である。カーボンファイバーは中性子吸
収も約３ミリバーンと小さく、耐熱性、耐放射線性も優
れている。また、機械的強度についても、常温での引張
り強さが、約３００ ｋｇ／ ｍｍ２とステンレスの約
５０ ｋｇ／ ｒｅ”と６倍程度高い。また、熱伝導度
もステンレスと同程度である。冷却材との共存性につい
ては、冷却材がナトリウムの場合には炭素のナトリウム
への移行の問題があるが、水の場合には共存性は極めて
よい。以上のカーボンファイバーの特徴から、本実施例
により、全属燃料と被覆管の共融の開運が解消され、か
つ、金属材被覆管を使用した場合より中性子経済が向上
した炉心が提供できる。FIG. 9 is a diagram showing a fifth embodiment, and is an example in which a material made from carbon fiber, which is a composite material, is used as the cladding tube 32. Carbon fiber has a low neutron absorption of about 3 milliburn, and has excellent heat resistance and radiation resistance. In addition, regarding mechanical strength, the tensile strength at room temperature is approximately 300 kg/mm2, which is approximately 6 times higher than that of stainless steel, which is approximately 50 kg/re''.Thermal conductivity is also comparable to that of stainless steel. Regarding coexistence with carbon fiber, when the coolant is sodium, there is a problem of carbon migration to sodium, but when the coolant is water, coexistence is extremely good.From the above characteristics of carbon fiber, this example This eliminates the problem of eutectic fusion between all-metal fuels and cladding, and provides a core with improved neutron economy compared to when metal cladding is used.
本発明によれば、酸化燃料を使用した場合に比べ同一の
水対燃料体積比では中性子の平均エネルギーを高くでき
、親核燃料物質から核分裂性物質への転換量を大きくで
きる。According to the present invention, the average energy of neutrons can be increased at the same water-to-fuel volume ratio compared to when oxidized fuel is used, and the amount of conversion from nuclear-friendly fuel material to fissile material can be increased.
第１図は本発明の一実施例である炉心と燃料集合体およ
び燃料体構成図、第２図は現行炉心と高転換炉の燃料集
合体と燃料体配列を示す説明図、第３図は第１図に示し
た実施例における燃料の燃焼に伴う中性子増倍率の変化
を酸化物燃料使用時と比較した線図、第４図は第１図に
示した実施例における燃料の燃焼に伴う積分転換比の変
化を酸化物燃料使用時と比較とだ線図、第５図、第６図
はそれぞれ他の実施例の炉心と燃料体集合体および燃料
棒構成図、第７図は本発明を高転換・バーナー炉心に実
施したときの炉心および燃料集合体構成図、第８図は第
７図に示した実施例における燃料の燃焼に伴う中性子増
倍率の変化を酸化物燃料使用時と比較した線図、第９図
は第７図に示した実施例における燃料の燃焼に伴う積分
転換比の変化を酸化物燃料使用時と比較した線図、第１
０図は他の実施例の炉心と燃料集合体および燃料体構成
図である。
１・・・炉心、２・・・燃料集合体、３・・・燃料体、
３１・・・全属燃料、３２・・・ギャップ、３３・・・
被覆管、４・・・中性子吸収棒、３４・・・グラファイ
ト層、３５・・・チツ化つラン層、３６・・・中空部、
５・・・高転換・バーナー炉心、７・・・バーナー領域
の燃料集合体。Fig. 1 is a configuration diagram of a reactor core, fuel assembly, and fuel assembly according to an embodiment of the present invention, Fig. 2 is an explanatory diagram showing the arrangement of the current reactor core, fuel assembly, and fuel assembly of a high conversion reactor, and Fig. 3 is Figure 1 is a diagram comparing the change in neutron multiplication factor due to fuel combustion in the example shown in Figure 1 with that when oxide fuel is used. Figure 4 is a diagram comparing the change in neutron multiplication factor due to fuel combustion in the example shown in Figure 1. A dashed line diagram comparing the change in conversion ratio with that when using oxide fuel, Figures 5 and 6 are diagrams of the reactor core, fuel assembly, and fuel rod configuration of other embodiments, respectively, and Figure 7 is a diagram showing the structure of the core, fuel assembly, and fuel rods of other embodiments. Figure 8 is a diagram of the reactor core and fuel assembly configuration when implemented in a high conversion/burner core, and compares the change in neutron multiplication factor due to fuel combustion in the example shown in Figure 7 with that when oxide fuel is used. 9 is a diagram comparing the change in integral conversion ratio due to fuel combustion in the example shown in FIG. 7 with that when oxide fuel is used.
FIG. 0 is a configuration diagram of a reactor core, fuel assembly, and fuel assembly of another embodiment. 1... Core, 2... Fuel assembly, 3... Fuel body,
31...All genus fuels, 32...Gap, 33...
Cladding tube, 4... Neutron absorption rod, 34... Graphite layer, 35... Titanium layer, 36... Hollow part,
5... High conversion/burner core, 7... Fuel assembly in the burner area.
Claims (1)
速を小さくして中性子の平均エネルギーを熱中性子炉よ
り高くし、親核燃料物質からの核分裂性物質の転換を多
くする炉心において、被覆管内の燃料として全属燃料を
使用したことを特徴とする原子炉。1. In a reactor core that uses light water as a neutron moderator, reduces neutron moderation to make the average energy of neutrons higher than that of a thermal neutron reactor, and increases the conversion of fissile material from parent nuclear fuel material, A nuclear reactor characterized by using all-genus fuel as fuel.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61007931A JPS62168091A (en) | 1986-01-20 | 1986-01-20 | Nuclear reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61007931A JPS62168091A (en) | 1986-01-20 | 1986-01-20 | Nuclear reactor |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPS62168091A true JPS62168091A (en) | 1987-07-24 |
Family
ID=11679264
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP61007931A Pending JPS62168091A (en) | 1986-01-20 | 1986-01-20 | Nuclear reactor |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS62168091A (en) |
Cited By (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2002265196A (en) * | 2001-03-07 | 2002-09-18 | Tadano Ltd | Telescopic boom device |
| JP2004301831A (en) * | 2003-03-20 | 2004-10-28 | Hitachi Ltd | Boiling water reactor light water reactor core and fuel assembly |
| JP2016224068A (en) * | 2010-05-11 | 2016-12-28 | トリウム・パワー、インクＴｈｏｒｉｕｍ Ｐｏｗｅｒ，Ｉｎｃ． | Fuel assembly |
| US10170207B2 (en) | 2013-05-10 | 2019-01-01 | Thorium Power, Inc. | Fuel assembly |
| US10192644B2 (en) | 2010-05-11 | 2019-01-29 | Lightbridge Corporation | Fuel assembly |
-
1986
- 1986-01-20 JP JP61007931A patent/JPS62168091A/en active Pending
Cited By (11)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2002265196A (en) * | 2001-03-07 | 2002-09-18 | Tadano Ltd | Telescopic boom device |
| JP2004301831A (en) * | 2003-03-20 | 2004-10-28 | Hitachi Ltd | Boiling water reactor light water reactor core and fuel assembly |
| JP2016224068A (en) * | 2010-05-11 | 2016-12-28 | トリウム・パワー、インクＴｈｏｒｉｕｍ Ｐｏｗｅｒ，Ｉｎｃ． | Fuel assembly |
| US10037823B2 (en) | 2010-05-11 | 2018-07-31 | Thorium Power, Inc. | Fuel assembly |
| US10192644B2 (en) | 2010-05-11 | 2019-01-29 | Lightbridge Corporation | Fuel assembly |
| US10991473B2 (en) | 2010-05-11 | 2021-04-27 | Thorium Power, Inc. | Method of manufacturing a nuclear fuel assembly |
| US11195629B2 (en) | 2010-05-11 | 2021-12-07 | Thorium Power, Inc. | Fuel assembly |
| US11837371B2 (en) | 2010-05-11 | 2023-12-05 | Thorium Power, Inc. | Method of manufacturing a nuclear fuel assembly |
| US11862353B2 (en) | 2010-05-11 | 2024-01-02 | Thorium Power, Inc. | Fuel assembly |
| US10170207B2 (en) | 2013-05-10 | 2019-01-01 | Thorium Power, Inc. | Fuel assembly |
| US11211174B2 (en) | 2013-05-10 | 2021-12-28 | Thorium Power, Inc. | Fuel assembly |
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