JP2005207819A - Boiling water reactor - Google Patents

Boiling water reactor Download PDF

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JP2005207819A
JP2005207819A JP2004013262A JP2004013262A JP2005207819A JP 2005207819 A JP2005207819 A JP 2005207819A JP 2004013262 A JP2004013262 A JP 2004013262A JP 2004013262 A JP2004013262 A JP 2004013262A JP 2005207819 A JP2005207819 A JP 2005207819A
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JP4467995B2 (en
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Toshihisa Shirakawa
白川利久
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    • 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
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Abstract

<P>PROBLEM TO BE SOLVED: To efficiently make spent MOX nuclear fuel from a pull-thermal reactor burn in a boiling water reactor, without impairing safety. <P>SOLUTION: In a low-pressure overheating steam boiling water reactor, the steam in the upper part inside a shroud is overheated by setting the pressure of the coolant inside a pressure vessel, containing a core consisting of a depleted MOX nuclear fuel assembly which is characterized in that plutonium and neptunium from the pull-thermal reactor are used as nuclear fuel and in that depleted MOX nuclear fuel rods 131, consisting of a stainless heat-resistant cladding tube 141 where ceramic fiber 1412 of alumina or silicon carbide is loaded into a perforated heat-resistant channel box and a heat-resistant control rod made by loading sintered pellets, such as boron carbide (B<SB>4</SB>C), into the heat-resistant cladding tube 141 at 70 atmospheric pressure or lower, and the pressure of the coolant inside the shroud at 40 at the atmospheric pressure or lower. <P>COPYRIGHT: (C)2005,JPO&NCIPI

Description

本発明は、沸騰水型原子炉並びに炉心に装荷せる核燃料集合体に関する。 The present invention relates to a boiling water reactor and a nuclear fuel assembly loaded in a core.

沸騰水型原子炉は、核燃料で発生した熱を原子炉内に液体で入ってきた水に伝達し水を沸騰させ飽和蒸気を発生する。飽和蒸気はタービンに導かれ電気を発生する。
図１は従来の沸騰水型原子炉（１）の圧力容器（１０）内の概観図を示す（非特許文献１）。タービンで仕事を終えた水は、給水配管（１７）を通って圧力容器（１０）壁とシュラウド（１８）との間のシュラウド外水（１６）に混じり込む。水はポンプモータ（２４）により回転する冷却材循環ポンプ（２３）で加速されてシュラウド（１８）の下端から矢印方向に核燃料物質を内包する核燃料棒を束ねた従来の核燃料集合体（３０）に未飽和水が流入し、熱を吸収して液体の水の一部が飽和蒸気になる。液体である水と気体である飽和蒸気が共存して流れている二相流となって上部に流れる。二相流断面において飽和蒸気が占める割合をボイド率と呼んでいる。ボイド率は核燃料集合体（３０）の下部ではゼロであり、中程では約４０％の中ボイド率になっており、上部では約７０％の高ボイド率になっている。
核燃料集合体（３０）の上部からの飽和蒸気を非常に多く含有した点線矢印方向の二相流と漏洩材通路（２０）からの矢印方向の水とが混合領域（１９）で混合した二相流は気水分離器（１５）の中に入り旋回させられることにより、開き矢印方向に上昇する飽和蒸気と矢印方向に下に落ちる水に分離される。上昇した飽和蒸気は水分を若干含んでいるため蒸気乾燥器（１２）により、開き矢印方向に上昇する乾燥した飽和蒸気と矢印方向に下に落ちる水に分離される。乾燥した飽和蒸気は蒸気ドーム（１１）から、圧力容器（１０）壁と蒸気乾燥器胴部（１３）の間を通って飽和蒸気配管（１４）からタービンへ蒸気が出て行く。
蒸気乾燥器（１２）内での飽和蒸気は破線で示した。なお、二相流から分離した飽和蒸気の飽和蒸気温度は、運転圧力約７０気圧での飽和蒸気温度で約摂氏２８６度である。
原子炉出力の制御は、制御棒駆動機構により上下に動く制御棒（２２）により達成する。
図2は核燃料物質を内包する従来の核燃料集合体（３０）の概略斜視図である（特許文献1）。核燃料集合体（３０）は、多数本正方格子状に配列された核燃料物質を内封している円柱形状の核燃料棒（３１）と、それ等の上端及び下端を夫々支持する上側結合板（３２）及び下側結合板（３３）と、前記核燃料棒（３１）の高さ途中に位置して核燃料棒（３１）間の間隔を規制する数個のスペーサ（３４）と、これ等を４面で覆うチャンネルボックス（３５）とから構成される。
図3は従来の核燃料棒（３１）の概観図である。ジルカロイ製の被覆管（４１）と、この被覆管（４１）の上下開口端を気密閉塞する上部端栓（４２）及び下部端栓（４３）と、被覆管（４１）内に装填される多数個の核燃料ペレット（４４）と、スプリング（４５）とから構成されている。核燃料ペレット（４４）はMOXと呼ばれるウラニウム（U）とプルトニウム（Pu）の混合酸化物からなる。近年、余剰のプルトニウムをMOXにして燃焼させてしまおうとしている。所謂プルサーマル炉の核燃料として使われようとしている。
図４は核燃料物質を内包する従来の核燃料集合体（３０）と従来の制御棒（２２）とからなる炉心平面の部分図である。炉心の下から核燃料集合体（３０）に入ってきた水は、核燃料棒（３１）の間の主冷却材通路（３６）を通り上に流れる間に核燃料棒（３１）から熱を吸収して飽和蒸気になり、飽和蒸気と液体の水とが混在した二相流となって流れている。チャンネルボックス（３５）の間の漏洩材通路（２０）には漏洩冷却水が流れている。制御棒（２２）はチャンネルボックス（３５）の間の漏洩材通路（２０）の中を上下に動ける。制御棒（２２）は原子炉出力を制御するための中性子を吸収する性質の強い物質であるハフニウムの薄板をステンレスで補強した構造になっている。または炭化硼素の粉末が充填された多数本の管をステンレスで補強した構造になっている。制御棒（２２）は制御棒駆動機構によって上下に動く。
図5は制御棒（２２）が引き抜かれた運転状態における、従来の核燃料集合体（３０）からなる炉心平面の部分図である。大半の制御棒（２２）は原子炉の下に引き抜かれている。制御棒（２２）の抜けたあとは漏洩材通路（２０）の水だけとなる。
現在稼動中の沸騰水型原子炉は、高価な濃縮ウランや再処理費用が高いプルトニウムの節約のために、冷却水速度を速くしたり、核燃料棒（３１）の間の主冷却材通路（３６）を広げて水領域を広げたり、漏洩材通路（２０）領域を広げたりして飽和蒸気の割合をできるだけ減らして、中性子速度の遅い中性子を利用している。
核燃料であるウランやプルトニウムの中性子との反応は中性子の速度によって変わり、中性子の速度は減速材である水の量により変わる。ボイド率の違いは水の量の違いであるから核燃料と中性子との反応の結果生じる出力に変化をもたらし、逆に出力の変化はボイド率に違いをもたらす。核燃料は遅い中性子とは激しく反応するため、少ない量の核燃料でも大きな出力を得ることができる。したがって、核燃料棒（３１）の周りに水を十分配して更に漏洩材通路（２０）領域を広げて、核分裂で生じた高速中性子を減速材である水により減速させて中性子速度を遅くさせる。ボイド率が高くなると減速材である水が減る訳であるから、水を冷却材循環ポンプ（２３）で加速して、タービンへの飽和蒸気流量の数倍の水を炉心に循環させて相対的にボイド割合が少なくなるようにしている。飽和蒸気とならなかった水は、気水分離器（１５）と蒸気乾燥器（１２）により飽和蒸気と分離されてシュラウド外水（１６）と混じって再び冷却材循環ポンプ（２３）で加速されて核燃料集合体（３０）を冷却する。
：昭61-37591、「核燃料集合体」。 ：コロナ社、著者都甲「原子動力」１１７、１２０頁。 In boiling water reactors, heat generated from nuclear fuel is transferred to water that has entered the reactor as a liquid, boiling water to generate saturated steam. The saturated steam is guided to the turbine to generate electricity.
FIG. 1 shows an overview of a pressure vessel (10) of a conventional boiling water reactor (1) (Non-Patent Document 1). The water that has finished work in the turbine is mixed with the shroud water (16) between the pressure vessel (10) wall and the shroud (18) through the water supply pipe (17). The water is accelerated by a coolant circulation pump (23) rotated by a pump motor (24), and enters a conventional nuclear fuel assembly (30) in which nuclear fuel rods containing nuclear fuel material are bundled in the arrow direction from the lower end of the shroud (18). Unsaturated water flows in, absorbs heat, and part of the liquid water becomes saturated vapor. It flows into the upper part as a two-phase flow in which liquid water and gas saturated vapor coexist. The proportion of saturated steam in the two-phase flow section is called the void fraction. The void ratio is zero in the lower part of the nuclear fuel assembly (30), the middle void ratio is about 40% in the middle, and the high void ratio is about 70% in the upper part.
Two-phase flow in which the two-phase flow in the direction of the dotted arrow containing a large amount of saturated steam from the upper part of the nuclear fuel assembly (30) and the water in the direction of the arrow from the leakage material passage (20) are mixed in the mixing region (19). The stream enters the steam separator (15) and is swirled to separate into saturated steam rising in the direction of the open arrow and water falling down in the direction of the arrow. Since the rising saturated steam contains some moisture, it is separated by the steam dryer (12) into dry saturated steam rising in the direction of the open arrow and water falling in the direction of the arrow. The dried saturated steam passes from the steam dome (11) between the wall of the pressure vessel (10) and the steam dryer body (13), and the steam exits from the saturated steam pipe (14) to the turbine.
The saturated steam in the steam dryer (12) is indicated by a broken line. The saturated steam temperature of the saturated steam separated from the two-phase flow is about 286 degrees Celsius at the saturated steam temperature at an operating pressure of about 70 atm.
Control of reactor power is achieved by a control rod (22) that moves up and down by a control rod drive mechanism.
FIG. 2 is a schematic perspective view of a conventional nuclear fuel assembly (30) containing nuclear fuel material (Patent Document 1). The nuclear fuel assembly (30) includes a cylindrical nuclear fuel rod (31) enclosing a nuclear fuel material arranged in a square lattice, and an upper coupling plate (32) for supporting the upper end and the lower end thereof. ) And the lower coupling plate (33), several spacers (34) which are located in the middle of the height of the nuclear fuel rod (31) and regulate the interval between the nuclear fuel rods (31), and these are arranged on four surfaces And a channel box (35) covered with.
FIG. 3 is an overview of a conventional nuclear fuel rod (31). Zircaloy-coated tube (41), upper end plug (42) and lower end plug (43) for hermetically closing the upper and lower opening ends of the coated tube (41), and a large number of tubes loaded in the coated tube (41) It consists of a single nuclear fuel pellet (44) and a spring (45). The nuclear fuel pellet (44) consists of a mixed oxide of uranium (U) and plutonium (Pu) called MOX. In recent years, surplus plutonium has been converted to MOX and burned. It is going to be used as nuclear fuel for so-called pull thermal furnaces.
FIG. 4 is a partial plan view of a core plane including a conventional nuclear fuel assembly (30) containing nuclear fuel material and a conventional control rod (22). Water entering the nuclear fuel assembly (30) from below the core absorbs heat from the nuclear fuel rod (31) while flowing upward through the main coolant passage (36) between the nuclear fuel rods (31). It becomes saturated steam and flows as a two-phase flow in which saturated steam and liquid water are mixed. Leakage cooling water flows through the leaking material passageway (20) between the channel boxes (35). The control rod (22) can move up and down in the leaking material passageway (20) between the channel boxes (35). The control rod (22) has a structure in which a thin plate of hafnium, which is a strong material for absorbing neutrons for controlling the reactor power, is reinforced with stainless steel. Alternatively, it has a structure in which a number of tubes filled with boron carbide powder are reinforced with stainless steel. The control rod (22) is moved up and down by a control rod drive mechanism.
FIG. 5 is a partial plan view of a core plane comprising a conventional nuclear fuel assembly (30) in an operating state in which the control rod (22) is pulled out. Most of the control rods (22) are drawn under the reactor. After the control rod (22) comes out, only the water in the leakage material passage (20) is obtained.
Boiling water reactors currently in operation have increased cooling water speeds and main coolant passages (36) between nuclear fuel rods (31) in order to save expensive enriched uranium and plutonium with high reprocessing costs. ) Is widened to widen the water region, or the leakage material passage (20) region is widened to reduce the ratio of saturated vapor as much as possible, and neutrons with a slow neutron velocity are used.
The reaction of nuclear fuel uranium and plutonium with neutrons depends on the neutron velocity, and the neutron velocity depends on the amount of moderator water. The difference in the void ratio is the difference in the amount of water, which causes a change in the output resulting from the reaction between the nuclear fuel and the neutron, and conversely, the change in the output causes a difference in the void ratio. Since nuclear fuel reacts violently with slow neutrons, a large output can be obtained even with a small amount of nuclear fuel. Accordingly, water is sufficiently distributed around the nuclear fuel rod (31) to further widen the leakage material passage (20) region, and fast neutrons generated by nuclear fission are decelerated by water as a moderator, thereby slowing down the neutron velocity. As the void ratio increases, the moderator water decreases, so the water is accelerated by the coolant circulation pump (23), and water several times the saturated steam flow to the turbine is circulated through the core. The void ratio is reduced. The water which has not become saturated steam is separated from the saturated steam by the steam separator (15) and the steam dryer (12), mixed with the shroud water (16), and accelerated again by the coolant circulation pump (23). To cool the nuclear fuel assembly (30).
: Sho 61-37591, “Nuclear Fuel Assembly”. : Corona, author Toko “Atomic power” 117, 120 pages.

減速材である水が多いプルサーマル炉ではプルトニウムの燃焼効率が悪い。プルサーマル炉で燃焼させた場合の取り出しプルトニウムの組成は核分裂し難いプルトニウム２４０（240Pu）やプルトニウム２４２（242Pu）の割合が多くなり核分裂し難くなるため、プルトニウムを何度も再処理して使用を繰り返すのは困難である。劣化プルトニウム（劣化Pu）となる。したがって、プルサーマル炉における使用済みMOX核燃料から核分裂生成物を除去しただけの核燃料組成は劣化MOXと呼べる。
減速材である水が少ない程またプルトニウムの混合割合が高くなる程高速中性子割合が多くなりプルトニウムの核分裂効率がよくなるため、MOX燃料の燃焼効率が上がる。特に、240Puや242Puも核分裂に寄与できるようになる。しかし、冷却材でもある水が急に減少すると減速材である液体の水の割合が急に減少し中性子が減速されなくなるため高速中性子割合が急に多くなるためプルトニウムの核分裂が急に活発になり発熱が高まり出力が急上昇する。出力の急上昇により蒸気割合が急に多くなり高速中性子割合が更に多くなり出力が益々増加する。核燃料棒を破損させる恐れがある。
劣化Puを組成とした劣化MOX核燃料の燃焼効率を高めてかつ、冷却材でもある水が急に減少しても出力が益々増加することなく核燃料棒を健全に保つ沸騰水型原子炉としたい。 Plutonium combustion efficiency is poor in a pull thermal furnace with a lot of water as a moderator. The composition of the extracted plutonium when burned in a pluthermal furnace is high in the proportion of plutonium 240 ( 240 Pu) and plutonium 242 ( 242 Pu) that are difficult to fission, making it difficult to fission. Is difficult to repeat. Degraded plutonium (degraded Pu). Therefore, a nuclear fuel composition that only removes fission products from spent MOX nuclear fuel in a pull thermal reactor can be called degraded MOX.
The smaller the moderator water and the higher the mixing ratio of plutonium, the higher the rate of fast neutrons and the better the fission efficiency of plutonium, thus increasing the combustion efficiency of MOX fuel. In particular, 240 Pu and 242 Pu can also contribute to nuclear fission. However, if the water, which is also the coolant, suddenly decreases, the ratio of liquid water, which is the moderator, suddenly decreases, and the neutrons are not decelerated. Heat generation increases and output increases rapidly. Due to the rapid increase in power, the vapor rate increases rapidly, the rate of fast neutrons increases further, and the output increases further. There is a risk of damage to the nuclear fuel rod.
We want to improve the combustion efficiency of deteriorated MOX nuclear fuel with the composition of deteriorated Pu, and make it a boiling water reactor that keeps the nuclear fuel rods healthy without any increase in power even if the water, which is also a coolant, suddenly decreases.

図６は本発明の劣化MOX核燃料棒（１３１）の概観図である。主冷却材通路（３６）に過熱蒸気が流れる上部にはネプツニウム（Np）の酸化物である酸化ネプツニウム（NpO2）を５０重量％以上とし残りがプルサーマル炉での使用済み核燃料から核分裂生成物を除去しただけの劣化Puとウラニウム（U）の混合酸化物である劣化MOXからなるNp劣化MOX核燃料ペレット（１０１）を装荷する。蒸気割合が比較的高い中部には劣化MOXからなる劣化MOX核燃料ペレット（１０２）を装荷する。蒸気割合が低い下部には劣化Puの酸化物富化度が１１重量％を上限とした劣化Pu高富化度MOX核燃料ペレット（１０３）を装荷する。耐熱被覆管（１４１）はアルミナや炭化珪素のセラミック繊維（１４１２）を埋め込んだステンレス製の外側被覆管（１４１１）と内側被覆管（１４１３）の3層構造をしている。なお、内側被覆管（１４１３）についてはモリブデン薄膜とすると高温性能が向上する。
図７は本発明の劣化MOX核燃料集合体（１３０）と本発明の耐熱制御棒（１２２）とからなる炉心平面の部分図である。劣化MOX核燃料棒（１３１）を装荷せる劣化MOX核燃料集合体（１３０）は、穴あき耐熱チャンネルボックス（１３５）を耐熱制御棒（１２２）と反対方向に広げ漏洩材通路（２０）を狭くし減速材たる水の領域を狭め更に、核燃料棒間隙を０．１ｃｍ〜０．２ｃｍに稠密に配列して劣化MOX核燃料棒（１３１）を多数本装荷できるようにした。劣化MOX核燃料集合体（１３０）の概観斜視図は図２と変わるところは殆どないが、核燃料棒（３１）が劣化MOX核燃料棒（１３１）に代わり狭い間隙で多数本配列されている。チャンネルボックス（３５）はステンレス製の耐熱材とし漏洩材通路（２０）と主冷却材通路（３６）とが連結され冷却水または蒸気が行き来できるように穴が開けてある。耐熱制御棒（１２２）は炭化硼素（B4C）または硼素とユーロピウムの化合物（EuB6）または酸化ユーロピウム（Eu2O3）の焼結ペレットを耐熱被覆管（１４１）に装荷した。
図8は本発明の劣化MOX核燃料集合体（１３０）と耐熱制御棒（１２２）を装荷した低圧過熱蒸気沸騰水型原子炉（３０１）である。圧力容器（１０）は従来のままとする。圧力容器（１０）内の冷却材圧力を４０気圧程度とする。気水分離器（１５）と蒸気乾燥器（１２）とは取り除いた。その代わりに、シュラウド覆い（１１８）と多数本の硼素またはB4C入りステンレス線をステンレス製の網篭に入れた加熱用線束（２００）を設置した。
MOX核燃料集合体（１３０）上端から過熱蒸気単相流が点線開き矢印方向に流れ、漏洩材通路（２０）の上端からも過熱蒸気単相流が点線開き矢印方向に流れ、混合領域（１９）で合流して加熱用線束（２００）を通過し過熱蒸気内管（１１５）を通りタービンへ出て行く。
加熱用線束（２００）の硼素を含む多数本のステンレス線は中性子やガンマ線を吸収して発熱し、通過する過熱蒸気を更に加熱する。
圧力容器（１０）壁とシュラウド（１８）の間のシュラウド外水（１６）が蒸発した飽和蒸気は飽和蒸気外管（１１４）を通って圧力容器（１０）外に出て、圧力容器（１０）に入ってくる給水を加熱したりするために利用される。
炉心への水の流入量はタービンへ出て行く過熱蒸気流量と同じである貫流とする。
圧力容器（１０）壁とシュラウド（１８）の間のシュラウド外水（１６）は高温となるシュラウド（１８）と圧力容器（１０）の健全性を損なうことがないように冷却する働きをする。
なお、過熱蒸気内管（１１５）からタービンへ過熱蒸気が出て行く間に熱電半導体を介在させ１０％程度の発電をさせれば発電効率が向上する。
なお、劣化MOXの代わりに通常のMOXを使っても燃焼効率が損なわれることはない。 FIG. 6 is a schematic view of a deteriorated MOX nuclear fuel rod (131) of the present invention. Neptunium oxide (NpO 2 ), which is an oxide of neptunium (Np), is 50% by weight or more in the upper part where superheated steam flows in the main coolant passage (36), and the remainder contains fission products from spent nuclear fuel in the pull thermal furnace. The Np-degraded MOX nuclear fuel pellet (101) composed of the degraded MOX, which is a mixed oxide of the degraded Pu and uranium (U), just removed is loaded. A deteriorated MOX nuclear fuel pellet (102) made of deteriorated MOX is loaded in the middle portion where the steam ratio is relatively high. In the lower part where the steam ratio is low, deteriorated Pu enrichment MOX nuclear fuel pellets (103) with an upper limit of 11% by weight of oxide enrichment of deteriorated Pu are loaded. The heat-resistant coated tube (141) has a three-layer structure of a stainless outer coated tube (1411) and an inner coated tube (1413) embedded with ceramic fibers (1412) of alumina or silicon carbide. In addition, about an inner side cladding tube (1413), when it is a molybdenum thin film, high temperature performance will improve.
FIG. 7 is a partial plan view of the core composed of the deteriorated MOX nuclear fuel assembly (130) of the present invention and the heat-resistant control rod (122) of the present invention. The deteriorated MOX nuclear fuel assembly (130) to which the deteriorated MOX nuclear fuel rod (131) is loaded is decelerated by expanding the perforated heat resistant channel box (135) in the opposite direction to the heat resistant control rod (122) and narrowing the leakage material passage (20). The region of water as material was narrowed, and the nuclear fuel rod gaps were densely arranged in a range of 0.1 cm to 0.2 cm so that a large number of deteriorated MOX nuclear fuel rods (131) could be loaded. The perspective view of the deteriorated MOX nuclear fuel assembly (130) is almost the same as that in FIG. 2, but a large number of nuclear fuel rods (31) are arranged in a narrow gap instead of the deteriorated MOX nuclear fuel rod (131). The channel box (35) is made of a heat-resistant material made of stainless steel, and the leakage material passage (20) and the main coolant passage (36) are connected to each other so that cooling water or steam can pass back and forth. The heat-resistant control rod (122) was obtained by loading sintered pellets of boron carbide (B 4 C), a compound of boron and europium (EuB 6 ) or europium oxide (Eu 2 O 3 ) into the heat-resistant cladding tube (141).
FIG. 8 shows a low pressure superheated steam boiling water reactor (301) loaded with a deteriorated MOX nuclear fuel assembly (130) and a heat-resistant control rod (122) of the present invention. The pressure vessel (10) remains the same. The coolant pressure in the pressure vessel (10) is about 40 atmospheres. The steam separator (15) and the steam dryer (12) were removed. Instead, a heating wire bundle (200) in which a shroud cover (118) and a large number of boron or B 4 C stainless steel wires were placed in a stainless steel net was installed.
A superheated steam single-phase flow from the upper end of the MOX nuclear fuel assembly (130) flows in the direction of the dotted arrow, and a superheated steam single-phase flow also flows from the upper end of the leakage passage (20) in the direction of the dotted open arrow. And passes through the heating wire bundle (200) and passes through the superheated steam inner pipe (115) to the turbine.
A number of stainless steel wires containing boron in the heating wire bundle (200) generate heat by absorbing neutrons and gamma rays, and further heat the superheated steam passing therethrough.
Saturated steam in which the water outside the shroud (16) between the pressure vessel (10) wall and the shroud (18) has evaporated passes through the saturated steam outer pipe (114) and exits from the pressure vessel (10). ) Used to heat incoming water.
The amount of water flowing into the core is a through-flow that is the same as the superheated steam flow to the turbine.
The water outside the shroud (16) between the wall of the pressure vessel (10) and the shroud (18) serves to cool the shroud (18) and the pressure vessel (10) at high temperatures so as not to impair the soundness of the shroud (18).
It should be noted that the power generation efficiency is improved if a thermoelectric semiconductor is interposed between the superheated steam inner pipe (115) and the thermoelectric semiconductor while the superheated steam goes out to the turbine.
It should be noted that combustion efficiency is not impaired even if ordinary MOX is used instead of degraded MOX.

劣化MOXが今後大量に発生してくると予想されるため劣化MOXは資源と見なせる。また、今迄のウラン２３５（２３５U）濃縮核燃料の燃焼により発生した処分し難いNpも大量にあるため資源とみなせる。処分や管理に厄介であった物質であるから安価もしくは処分費が不要となるため核燃料費が大幅に軽減される。
高温過熱蒸気が低圧であるため、構造物の健全性を損なうことなく発電効率を高めるため発電コストを低減することができる。 Since it is expected that a large amount of deteriorated MOX will occur in the future, the deteriorated MOX can be regarded as a resource. Moreover, since there is a large amount of Np that is generated by the combustion of uranium 235 ( 235 U) enriched nuclear fuel so far and is difficult to dispose, it can be regarded as a resource. Since it is a substance that was troublesome to dispose of and manage, it is cheaper and no disposal costs are required, so nuclear fuel costs are greatly reduced.
Since the high-temperature superheated steam is at a low pressure, the power generation cost can be reduced in order to increase the power generation efficiency without impairing the soundness of the structure.

劣化Puを組成とした劣化MOX核燃料の燃焼効率を高めてかつ、冷却材でもある水が急に減少しても出力が益々増加することなく核燃料棒を健全に保つ沸騰水型原子炉が提供できた。 It is possible to provide a boiling water nuclear reactor that improves the combustion efficiency of degraded MOX nuclear fuel with the composition of degraded Pu, and keeps the nuclear fuel rods healthy without any increase in power even if the coolant water suddenly decreases. It was.

ネプツニウム（Np）は核分裂作用の低い物質であるが、高速中性子とはプルトニウム２３９（２３９Pu）程度の核分裂をする。NpO２からなる核燃料棒が７０気圧での飽和蒸気１００％の中にある場合には無限増倍係数（ｋ∞）が１．２にもなる。本発明の劣化MOX核燃料棒（１３１）において、過熱蒸気が流れる上部のNp劣化MOX核燃料ペレット（１０１）ではｋ∞を１．０以上にすることができる。通常運転時が過熱蒸気であるため万一冷却材流量が減少しても蒸気割合の増加は少なく反応度の急上昇は殆どなく安全性に問題がない。
蒸気割合が高く熱除去が悪くかつ中性子束も高い中部では、劣化MOX核燃料ペレット（１０２）を装荷することにより出力が高くなるのを抑制して核燃料棒の健全性を図った。核燃料の大半はウラン２３８（２３８U）であるため万一冷却材流量が減少して蒸気割合が増加しても反応度は減少するため安全性に問題がない。
蒸気割合が低く熱除去が良くかつ中性子束も低い下部では劣化Pu高富化度MOX核燃料ペレット（１０３）を装荷することにより出力を高くするようにした。万一冷却材流量が減少して蒸気割合が増加しても劣化Pu酸化物の富化度が１１重量％以下なら反応度が急激に上昇することはないため安全性に問題がない。
耐熱被覆管（１４１）はセラミック繊維で補強されているため、万一核燃料棒が高温になっても酸化物核燃料ペレットが崩れ落ちないように保つことができる。
劣化MOX核燃料集合体（１３０）は劣化MOX核燃料棒（１３１）を稠密に配列して多数本にしたため単位長さ当たりの出力が抑制され、かつ被覆管の高温強度が増しているため高温過熱蒸気に対して耐久性がある。
万一耐熱被覆管（１４１）が破損し劣化MOX核燃料棒（１３１）が崩れ主冷却材通路（３６）が狭まったとしても穴あきチャンネルボックス（１３５）の穴を通って漏洩冷却材通路（２０）から冷却材が供給されるため劣化MOX核燃料集合体（１３０）の大きな破損には至らない。
耐熱制御棒（１２２）のB４C等の焼結ペレットの融点は摂氏２０００度程度であるためセラミック繊維で補強された耐熱被覆管（１４１）により形状を保ち中性子吸収機能を長時間保つことができる。
劣化MOXは崩壊熱が比較的高いため、核分裂生成物の崩壊熱と相俟って所要の出力の２０％程度を担っている。したがって、核分裂による出力は８０％程度で済むため出力分布は平坦化され局所的核燃料棒の破損は少なくなる。
摂氏６００度程度の高温過熱蒸気も可能である本発明の原子炉（３０１）の発電効率は４０％にもなるため、核分裂による出力は更に低くてもよくなる。
本発明の原子炉（３０１）の圧力は約４０気圧であるため、過熱蒸気が高温であっても構造物や配管の強度上の問題は軽減される。
主蒸気隔離弁（MSIV）の緊急閉鎖作動を緩慢閉鎖にし、かつバイパス弁緊急作動により蒸気１００％をタービン復水器にバイパスすれば炉心を１時間程度健全に保ち続けることができる。１時間程度冷却が続けば事象への対応も進み核燃料棒の大きな破損が免れる。
なお、Npが資源的に不足した場合は、アメリシウムや劣化Pu低富化度MOXを利用すればよい。 Neptunium (Np) is a substance with low fission action, but fast neutrons fission about plutonium 239 ( 239 Pu). When a nuclear fuel rod made of NpO 2 is in 100% saturated steam at 70 atm, the infinite multiplication factor (k ∞ ) is 1.2. In the deteriorated MOX nuclear fuel rod (131) of the present invention, k ∞ can be 1.0 or more in the upper Np deteriorated MOX nuclear fuel pellet (101) through which superheated steam flows. Since it is superheated steam during normal operation, even if the coolant flow rate decreases, the increase in the steam ratio is small and there is almost no sudden increase in reactivity, and there is no problem in safety.
In the middle part where the steam ratio is high, the heat removal is poor, and the neutron flux is high, loading the deteriorated MOX nuclear fuel pellet (102) prevents the output from becoming high, and the nuclear fuel rod is sound. Since most of the nuclear fuel is uranium 238 ( 238 U), even if the coolant flow rate decreases and the steam ratio increases, the reactivity decreases, so there is no safety problem.
In the lower part where the steam ratio is low and heat removal is good and the neutron flux is also low, the degraded Pu high enrichment MOX nuclear fuel pellet (103) is loaded to increase the output. Even if the coolant flow rate decreases and the steam ratio increases, if the enrichment degree of the deteriorated Pu oxide is 11% by weight or less, the reactivity does not increase rapidly, so there is no problem in safety.
Since the heat-resistant cladding tube (141) is reinforced with ceramic fibers, the oxide nuclear fuel pellets can be kept from collapsing even if the nuclear fuel rods become hot.
The deteriorated MOX nuclear fuel assembly (130) is made up of a large number of deteriorated MOX nuclear fuel rods (131) arranged densely, so the output per unit length is suppressed, and the high temperature strength of the cladding tube is increased, so the high temperature superheated steam Durable against.
Even if the heat-resistant cladding tube (141) breaks and the deteriorated MOX nuclear fuel rod (131) collapses and the main coolant passage (36) narrows, the leaked coolant passage (20 ) Will not cause major damage to the deteriorated MOX nuclear fuel assembly (130).
Since the melting point of sintered pellets such as B 4 C of the heat-resistant control rod (122) is about 2000 degrees Celsius, the heat-resistant cladding tube (141) reinforced with ceramic fibers can maintain the shape and maintain the neutron absorption function for a long time. it can.
Degraded MOX has a relatively high decay heat, so it takes about 20% of the required output in combination with the decay heat of fission products. Therefore, since the output by fission is about 80%, the output distribution is flattened and the damage of the local nuclear fuel rod is reduced.
The power generation efficiency of the nuclear reactor (301) of the present invention, which is capable of high-temperature superheated steam of about 600 degrees Celsius, is as high as 40%, so that the power generated by nuclear fission may be even lower.
Since the pressure of the nuclear reactor (301) of the present invention is about 40 atmospheres, the problem of strength of structures and piping is reduced even when the superheated steam is at a high temperature.
If the emergency closing operation of the main steam isolation valve (MSIV) is slowly closed and 100% of the steam is bypassed to the turbine condenser by the emergency operation of the bypass valve, the core can be kept healthy for about 1 hour. If cooling continues for about an hour, the response to the event will proceed and major damage to the nuclear fuel rod will be avoided.
If Np is insufficient in terms of resources, americium or degraded Pu low enrichment MOX may be used.

図９は本発明の超低圧過熱蒸気沸騰水型原子炉（４０１）の概観図である。シュラウド（１８）の内側の圧力を５気圧程度にした高温過熱蒸気を蒸気発生器に送り、原子炉への戻り蒸気を７気圧程度にコンプレッサーで昇圧し戻り蒸気管（４１７）から温度の下がった過熱蒸気を原子炉に戻せば、圧力容器（１０）の内側の圧力を７気圧程度に抑制することができる。シュラウド（１８）と圧力容器（１０）との間の低温過熱蒸気は低温過熱蒸気外管（４１４）を通ってタービンへ行く。タービンからの水は戻り蒸気管（４１７）に合流して原子炉に戻る。安全性が高まる。 FIG. 9 is an overview of the ultra-low pressure superheated steam boiling water reactor (401) of the present invention. High-temperature superheated steam with the pressure inside the shroud (18) set to about 5 atmospheres was sent to the steam generator, and the return steam to the reactor was pressurized to about 7 atmospheres by the compressor and the temperature dropped from the return steam pipe (417). If superheated steam is returned to the nuclear reactor, the pressure inside the pressure vessel (10) can be suppressed to about 7 atmospheres. The cold superheated steam between the shroud (18) and the pressure vessel (10) goes to the turbine through the cold superheated steam outer pipe (414). Water from the turbine joins the return steam pipe (417) and returns to the reactor. Increased safety.

MOX燃料を熱中性子利用の従来と同じ核燃料集合体（３０）に充填して燃焼させるプルサーマル炉が計画されている。本発明によりプルサーマル炉からの取り出し核燃料を再利用し続けられる可能性が高まったためプルサーマル炉の開発が促進され、並びに利用の進んだ従来型沸騰水型原子炉を本発明の原子炉への改造が進展する。 A pull thermal furnace is planned in which MOX fuel is filled in the same nuclear fuel assembly (30) that uses thermal neutrons and burned. The possibility of continuing to reuse the nuclear fuel removed from the pull thermal reactor has been promoted by the present invention, so that the development of the pull thermal reactor has been promoted, and the conventional boiling water reactor that has advanced use can be converted to the nuclear reactor of the present invention. Progress.

従来の沸騰水型原子炉（１）の圧力容器（１０）内の概観図。1 is a schematic view of a pressure vessel (10) of a conventional boiling water reactor (1). 従来の沸騰水型原子炉（１）の圧力容器（１０）内に装荷せる従来の核燃料集合体（３０）の概観斜視図。1 is a perspective view of a conventional nuclear fuel assembly (30) that can be loaded into a pressure vessel (10) of a conventional boiling water reactor (1). 従来の核燃料棒（３１）の概観図。Overview of a conventional nuclear fuel rod (31). 従来の核燃料集合体（３０）と制御棒（２２）とからなる炉心平面の部分図。FIG. 3 is a partial plan view of a reactor core composed of a conventional nuclear fuel assembly (30) and a control rod (22). 制御棒（２２）が引き抜かれた運転時における、従来の核燃料集合体（３０）からなる炉心平面の部分図。FIG. 6 is a partial plan view of a core plane composed of a conventional nuclear fuel assembly (30) during operation in which a control rod (22) is pulled out. 本発明の劣化MOX核燃料棒（１３１）の概観図。1 is an overview of a deteriorated MOX nuclear fuel rod (131) of the present invention. 本発明の劣化MOX核燃料集合体（１３０）と耐熱制御棒（１２２）とからなる炉心平面の部分図。FIG. 3 is a partial plan view of a core plane including a deteriorated MOX nuclear fuel assembly (130) and a heat-resistant control rod (122) according to the present invention. 本発明の低圧過熱蒸気沸騰水型原子炉（３０１）の圧力容器（１０）内の概観図。1 is a schematic view of the inside of a pressure vessel (10) of a low-pressure superheated steam boiling water reactor (301) according to the present invention. 本発明の超低圧過熱蒸気沸騰水型原子炉（４０１）の圧力容器（１０）内の概観図。1 is a schematic view of a pressure vessel (10) of an ultra-low pressure superheated steam boiling water reactor (401) according to the present invention.

符号の説明Explanation of symbols

１は従来の沸騰水型原子炉概観図。
１０は圧力容器。
１１は蒸気ドーム。
１２は蒸気乾燥器。
１３は蒸気乾燥器胴部。
１４は飽和蒸気配管。
１５は気水分離器。
１６はシュラウド外水。
１７は給水配管。
１８はシュラウド。
１９は混合領域。
２０は漏洩材通路。
２２は制御棒。
２３は冷却材循環ポンプ。
２４はポンプモータ。
３０は核燃料集合体。
３１は核燃料棒。
３２は上側結合板。
３３は下側結合板。
３４はスペーサ。
３５はチャンネルボックス。
３６は主冷却材通路。
４１は被覆管。
４２は上部端栓。
４３は下部端栓。
４４は核燃料ペレット。
４５はスプリング。
１０１はNｐ劣化MOX核燃料ペレット。
１０２は劣化MOX核燃料ペレット。
１０３は劣化Pu高富化度MOX核燃料ペレット。
１１４は飽和蒸気外管。
１１５は過熱蒸気内管。
１１８はシュラウド覆い。
１２２は耐熱制御棒。
１３０は劣化MOX核燃料集合体。
１３１は劣化MOX核燃料棒。
１３５は穴あき耐熱チャンネルボックス。
１４１は耐熱被覆管。
３０１は本発明の低圧過熱蒸気沸騰水型原子炉概観図。
４０１は本発明の超低圧過熱蒸気沸騰水型原子炉概観図。
４１４は低温過熱蒸気外管。
４１７は戻り蒸気管。
１４１１は外側被覆管。
１４１２はセラミック繊維。
１４１３は内側被覆管。
1 is an overview of a conventional boiling water reactor.
10 is a pressure vessel.
11 is a steam dome.
12 is a steam dryer.
13 is a steam dryer trunk.
14 is a saturated steam pipe.
15 is a steam separator.
16 is water outside the shroud.
17 is a water supply pipe.
18 is a shroud.
19 is a mixing area.
20 is a leakage material passage.
22 is a control rod.
23 is a coolant circulation pump.
24 is a pump motor.
30 is a nuclear fuel assembly.
31 is a nuclear fuel rod.
32 is an upper coupling plate.
33 is a lower coupling plate.
34 is a spacer.
35 is a channel box.
36 is a main coolant passage.
41 is a cladding tube.
42 is an upper end plug.
43 is a lower end plug.
44 is a nuclear fuel pellet.
45 is a spring.
101 is an Np-degraded MOX nuclear fuel pellet.
102 is a deteriorated MOX nuclear fuel pellet.
103 is a deteriorated Pu enrichment MOX nuclear fuel pellet.
114 is a saturated steam outer pipe.
115 is a superheated steam inner pipe.
118 is a shroud cover.
122 is a heat-resistant control rod.
130 is a deteriorated MOX nuclear fuel assembly.
131 is a deteriorated MOX nuclear fuel rod.
135 is a perforated heat resistant channel box.
141 is a heat-resistant coated tube.
301 is an overview of the low-pressure superheated steam boiling water reactor of the present invention.
401 is an overview of the ultra-low pressure superheated steam boiling water reactor of the present invention.
Reference numeral 414 denotes a low-temperature superheated steam outer pipe.
417 is a return steam pipe.
Reference numeral 1411 denotes an outer cladding tube.
1412 is a ceramic fiber.
1413 is an inner cladding tube.

Claims (2)

上部にはネプツニウム（Np）の酸化物である酸化ネプツニウム（NpO2）を５０重量％以上とし残りがプルサーマル炉での使用済みMOX核燃料から核分裂生成物を除去しただけの劣化MOXからなるNp劣化MOX核燃料ペレット（１０１）を装荷し、中部には劣化MOXからなる劣化MOX核燃料ペレット（１０２）を装荷し、下部には劣化Puの酸化物富化度が１１重量％を上限とした劣化Pu高富化度MOX核燃料ペレット（１０３）を装荷した、アルミナまたは炭化珪素のセラミック繊維（１４１２）を埋め込んだステンレス製の外側被覆管（１４１１）と内側被覆管（１４１３）の3層構造をしている耐熱被覆管（１４１）からなる劣化MOX核燃料棒（１３１）の間隙を０．１ｃｍ〜０．２ｃｍに多数本束ね、耐熱制御棒（１２２）と反対方向の漏洩材通路（２０）を狭くして広げたる穴あき耐熱チャンネルボックス（１３５）に装荷したことを特徴とする劣化MOX核燃料集合体（１３０）。 The upper part is Np-degraded MOX consisting of depleted MOX that only removes fission products from spent MOX nuclear fuel in the pluthermal furnace, with neptunium oxide (NpO 2 ), which is an oxide of neptunium (Np), at 50% by weight or more. Loaded with nuclear fuel pellets (101), loaded with deteriorated MOX nuclear fuel pellets (102) consisting of deteriorated MOX in the middle, enriched with deteriorated Pu up to 11% by weight of oxide enrichment of deteriorated Pu in the lower part Heat-resistant coating with a three-layer structure of an outer cladding tube (1411) and an inner cladding tube (1413) made of stainless steel embedded with ceramic fibers (1412) of alumina or silicon carbide, loaded with a high degree MOX nuclear fuel pellet (103) A large number of gaps of deteriorated MOX nuclear fuel rods (131) consisting of tubes (141) are bundled to 0.1 cm to 0.2 cm, and the leakage material passage (20) in the opposite direction to the heat-resistant control rod (122) is narrowed. To spread upcoming perforated degradation MOX nuclear fuel assembly, characterized in that the loaded heat channel box (135) (130). 請求項１における劣化MOX核燃料集合体（１３０）と炭化硼素（B4C）または硼素とユーロピウムの化合物（EuB6）または酸化ユーロピウム（Eu2O3）の焼結ペレットを耐熱被覆管（１４１）に装荷した耐熱制御棒（１２２）からなる炉心を内蔵せるシュラウド（１８）内上部を過熱蒸気とし、圧力容器（１０）内の冷却材圧力を７０気圧以下としたことを特徴とする低圧過熱蒸気沸騰水型原子炉（３０１）。
The deteriorated MOX nuclear fuel assembly (130) according to claim 1 and sintered pellets of boron carbide (B 4 C), a compound of boron and europium (EuB 6 ) or europium oxide (Eu 2 O 3 ) are heat-resistant cladding (141). The low pressure superheated steam is characterized in that the upper part in the shroud (18) containing the core composed of the heat-resistant control rod (122) loaded on the superheated steam is superheated steam, and the coolant pressure in the pressure vessel (10) is 70 atm or less. Boiling water reactor (301).
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JP2012237574A (en) * 2011-05-10 2012-12-06 Yuji Uenohara Cladding tube and nuclear reactor
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