JPS5810678A - Fuel assembly for heavy water moderated pressure tube type reactor - Google Patents
Fuel assembly for heavy water moderated pressure tube type reactorInfo
- Publication number
- JPS5810678A JPS5810678A JP56108641A JP10864181A JPS5810678A JP S5810678 A JPS5810678 A JP S5810678A JP 56108641 A JP56108641 A JP 56108641A JP 10864181 A JP10864181 A JP 10864181A JP S5810678 A JPS5810678 A JP S5810678A
- Authority
- JP
- Japan
- Prior art keywords
- fuel
- fuel assembly
- pressure tube
- heavy water
- absorbing material
- 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.)
- Granted
Links
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
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Devices And Processes Conducted In The Presence Of Fluids And Solid Particles (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】
本発明は重水減速圧力管型原子炉用燃料集合体に係シ、
特に重水減速沸騰水冷却圧力管票原子炉の炉心核特性を
改善すると同時に燃料の燃焼特性を改善するのに好適な
燃料集合体に関するものである。DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a fuel assembly for a heavy water moderation pressure tube type nuclear reactor.
In particular, the present invention relates to a fuel assembly suitable for improving the core characteristics of a heavy water-moderated boiling water-cooled pressure tube nuclear reactor and at the same time improving the combustion characteristics of the fuel.
従来のこの種燃料集合体は、細径棒状の燃料棒を同心円
状に多数配列し、これらを束ねたクラスタ型燃料集合体
が一般的であった。この従来の燃料集合体を用いた重水
減速圧力管型原子炉の模式図を第1図に示す。この型の
原子炉の一般的な特性として、原子炉の起動時、特に冷
却材ボイドが発生し始める高温待機からタービン併入ま
での低出力領域において、冷却材ボイド係数があまシ正
の値を示すと、中性子束および蒸気ドラム水位が変動し
ゃすくなシ、炉の制御が困難になるということがある。Conventional fuel assemblies of this type have generally been cluster-type fuel assemblies in which a large number of small-diameter rod-shaped fuel rods are arranged concentrically and bundled together. FIG. 1 shows a schematic diagram of a heavy water-moderated pressure tube nuclear reactor using this conventional fuel assembly. A general characteristic of this type of reactor is that the coolant void coefficient has a fairly positive value during reactor startup, especially in the low power range from high-temperature standby where coolant voids begin to occur to turbine entry. If this occurs, the neutron flux and steam drum water level may fluctuate, making it difficult to control the reactor.
以下その原因を説明する。第1図において、低出力時に
は給水管1から比較的冷たい水が蒸気ドラム2へ送られ
、これが下降管3および入口管4會通って炉心5内の圧
力管6内の冷却材中へ入って行くため、冷たい水にょシ
冷却材中のボイドがつぶれ、蒸気ドラム2の水位が下降
する。そして冷却材ボイド係数が正の大きな値のときは
、この傾向を中性子束の変化を通して加速する結果とな
シ、原子炉がスクラムしやすい状態となる。冷却材ボイ
ド係数は、プルトニウム・ウラン混合燃料を用いた場合
は比較釣魚でアシ、上記問題は生じないが、ウラン燃料
を用いた場合は、冷却材ボイド係数が正となプ、上記問
題を生じゃすくなる。The cause will be explained below. In FIG. 1, at low power, relatively cold water is sent from the water supply pipe 1 to the steam drum 2, passes through the downcomer pipe 3 and the inlet pipe 4, and enters the coolant in the pressure pipe 6 in the reactor core 5. As the water cools, the voids in the coolant collapse and the water level in the steam drum 2 falls. When the coolant void coefficient has a large positive value, this tendency is accelerated through changes in the neutron flux, and the reactor becomes susceptible to scram. The coolant void coefficient is positive when using plutonium-uranium mixed fuel, and the above problem does not occur, but when uranium fuel is used, the coolant void coefficient is positive, causing the above problem. It's getting better.
最近、冷却材ボイド係数を改善する方法として、特開昭
55−135784号で提案されているように、燃料物
質中あるいは燃料棒外側に燃料棒に並行させて可燃性毒
物である中性子吸収材を燃料棒の長手方向全領域に装荷
する方法が知られている。Recently, as a method to improve the coolant void coefficient, as proposed in Japanese Patent Application Laid-Open No. 55-135784, a neutron absorbing material, which is a burnable poison, is added in the fuel material or on the outside of the fuel rods in parallel with the fuel rods. A method is known in which the entire length of the fuel rod is loaded.
この方法は、圧力管6内の冷却材中にボイドがない場合
は熱中性子束が大きく減少するのく対し、ボイド発生と
ともに圧力管6内熱中性子束が増大する効果を利用し、
圧力管6内に中性子吸収材を置くことによってボイドに
対して負の反応度投入効果をもたらすことを応用したも
のである。This method utilizes the effect that thermal neutron flux inside pressure tube 6 increases as voids occur, whereas thermal neutron flux decreases greatly when there are no voids in the coolant inside pressure tube 6.
This is an application of the fact that placing a neutron absorbing material in the pressure tube 6 produces a negative reactivity injection effect on voids.
しかし、この方法には、可燃性毒物が消滅した段階では
、冷却材ボイド係数をより負にする効果がなくなるとい
う欠点がある。燃料の燃焼を通じて上記効果を維持させ
るためには、燃焼末期に至るまで中性子吸収材が残るよ
うに、中性子吸収材を濃く装荷することが考えられるが
、このようにすると、当然のこととして燃料の達成燃焼
度が低下し、よシ多くの核分裂性物質を消費するという
問題を生ずる。However, this method has the disadvantage that once the burnable poison has disappeared, the effect of making the coolant void coefficient more negative is lost. In order to maintain the above effect throughout the combustion of the fuel, it is conceivable to load the neutron absorber in a concentrated manner so that the neutron absorber remains until the final stage of combustion. The problem is that the achieved burnup is lower and more fissile material is consumed.
本発明は上記に鑑みてなされたもので、その目的とする
ところは、低出力時の冷却材ボイド反応度係数をさらに
負側にすることができ、同時に燃料の達成燃焼度をさら
に向上することができる重水減速圧力管m原子炉用燃料
集合体を提供することにある。The present invention has been made in view of the above, and its purpose is to make it possible to make the coolant void reactivity coefficient at low power even more negative, and at the same time to further improve the achieved burnup of the fuel. An object of the present invention is to provide a fuel assembly for a heavy water moderation pressure tube m nuclear reactor that can be used in a nuclear reactor.
本発明は重水減速沸騰水冷却圧力管型原子炉の冷却材ボ
イド率変化による炉心反応度変化が冷却材流路の沸騰開
始点以降の下流側で生じることに着目してなされたもの
で、沸騰開始点より下流側である燃料集合体の炉心有効
長の冷却材下流側2/3の領域の範囲内に中性子吸収材
を装荷したこと全特徴としている。The present invention was developed by focusing on the fact that changes in core reactivity due to changes in coolant void ratio in heavy water-moderated boiling water-cooled pressure tube reactors occur downstream of the point at which boiling starts in the coolant flow path. The main feature of this system is that the neutron absorbing material is loaded within the range of 2/3 of the coolant downstream of the effective length of the core of the fuel assembly, which is downstream from the starting point.
以下本発明を第2図、第3図に示した実施例および第4
図ないし第6図を用いて詳細に説明する。The present invention will be described below with reference to the embodiments shown in FIGS. 2 and 3, and the embodiments shown in FIGS.
This will be explained in detail using FIGS. 6 through 6.
第2図は本発明の燃料集合体の一実施例を示す中央部分
縦断面図である。第2図の燃料集合体は、同心円状に多
数の燃料棒が配置されたクラスタ型燃料集合体で燃料ベ
レット7と燃料ベレット7を包む被覆管8とよりなる細
径棒状の多数の燃料棒、被覆管8を束ねている燃料スペ
ーサ9、燃料棒の上部、下部をそれぞれ固定する上部タ
イプレート10、下部タイプレート11および中央部に
配設され、燃料スペーサ90間隔を保つためのスペーサ
タイロッド12とから構成しである。とζろで、第2図
に示す実施例では、スペーサタイロッド12の上部には
炭化ボロンの粉末よシなる中性子吸収材13を収納しで
ある。FIG. 2 is a longitudinal cross-sectional view of the center portion of an embodiment of the fuel assembly of the present invention. The fuel assembly shown in FIG. 2 is a cluster-type fuel assembly in which a large number of fuel rods are arranged concentrically, and includes a large number of small-diameter rod-shaped fuel rods consisting of a fuel pellet 7 and a cladding tube 8 surrounding the fuel pellet 7. A fuel spacer 9 that bundles the cladding tubes 8, an upper tie plate 10 and a lower tie plate 11 that fix the upper and lower parts of the fuel rods, respectively, and a spacer tie rod 12 that is arranged in the center and maintains the spacing between the fuel spacers 90. It consists of: In the embodiment shown in FIG. 2, a neutron absorbing material 13 such as boron carbide powder is housed in the upper part of the spacer tie rod 12.
第3図は第2図のスペーサタイロッド12の一実施例を
示す縦断面図である。スペーサタイロッド12は、2本
のジルカロイ合金製の円管14を連結用端栓15にネジ
込み結合し、2本の円管14の上部、下部にはそれぞれ
タイプレート固定用端栓16が溶接してToシ、上部の
円管14内には炭化ボロン17よりなる中性子吸収材(
13)を中性子吸収材収納管18に収納した状態で納め
た構成としである。なお、収納管18内の炭化ボロン1
7は端@19によシ密封されており、かつ、収納管18
の中には炭化ボロン17から発生するガスt−九める空
間がばね20によって確保しである。このようにして、
中性子吸収材である炭化ボロン17t−収納しである部
分の長さは、炉心の有効長の約1/2としてhbsかつ
、軸方向上部である下流側に収納しである。そして下部
の円管14の内部には冷却材が流れる構造としてあり、
また1上、下部のそれぞれの円管14にはスペーサ保持
用の突起21が溶接しである。FIG. 3 is a longitudinal sectional view showing one embodiment of the spacer tie rod 12 of FIG. 2. FIG. The spacer tie rod 12 has two circular tubes 14 made of Zircaloy alloy screwed together to a connecting end plug 15, and tie plate fixing end plugs 16 are welded to the upper and lower parts of the two circular tubes 14, respectively. In the upper circular tube 14, there is a neutron absorbing material made of boron carbide 17 (
13) is housed in a neutron absorbing material storage tube 18. Note that the boron carbide 1 in the storage tube 18
7 is sealed by the end @19 and the storage tube 18
Inside, a space for gas generated from boron carbide 17 is secured by a spring 20. In this way,
The length of the part containing 17t of boron carbide, which is a neutron absorbing material, is about 1/2 of the effective length of the reactor core, and the length is about 1/2 of the effective length of the reactor core, and it is stored on the upper downstream side in the axial direction. There is a structure in which the coolant flows inside the lower circular pipe 14,
Furthermore, a projection 21 for holding a spacer is welded to each of the upper and lower circular tubes 14.
第4図は沸騰水冷却圧力管型原子炉における炉心出力を
冷却材の沸騰開始点位置との関係を示す線図で、沸騰開
始点位置は炉心の軸方向有効距離で示しである。第4図
によれば、軸方向有効距離1850■で冷却材が沸騰を
開始するのは、炉心出力が約20%のときとなる。すな
わち、炉心出力20%以下の低出力では、冷却材ボイド
の変化は、炉心軸方向上半分で生じ、冷却材ボイド変化
による反応度変化も炉心の上半分で生じることになる。FIG. 4 is a diagram showing the relationship between the core output and the boiling start point position of the coolant in a boiling water cooled pressure tube reactor, and the boiling start point position is shown in terms of the effective distance in the axial direction of the core. According to FIG. 4, the coolant starts boiling at an effective axial distance of 1850 cm when the core power is approximately 20%. That is, at low core power of 20% or less, changes in coolant voids occur in the upper half of the core in the axial direction, and changes in reactivity due to changes in coolant voids also occur in the upper half of the core.
したがって、炉心出力が約20%以下では、燃料集合体
の有効長全長にわたって中性子吸収材を装荷した場合と
、上記し九実流側のように有効長の上部半分に装荷した
場合とで、冷却材ボイド反応度係数をよシ負にする効果
は同じになる。Therefore, when the core power is about 20% or less, cooling is possible when the neutron absorber is loaded over the entire effective length of the fuel assembly, and when it is loaded on the upper half of the effective length as in the above-mentioned nine flow side. The effect of making the material void reactivity coefficient more negative is the same.
第5図はウラン燃料の場合の炉心出力と冷却材ボイド反
応度係数差との関係を示す線図で、約2%濃度の微濃縮
ウラン燃料を用い、炭化ボロン17の中の中性子吸収材
であるボロン同位元素(質量的10)の原子数密度を新
燃料時に約1×1611個/国易とし、燃焼サイクル末
期に対応する約130WD/T燃焼し木時点のものを示
しである。第5図において、1曲線は中性子吸収材を装
荷しない場合、6曲線は中性子吸収材を有効長全長に装
荷し九場合、C曲線は本発明の実施例のように中性子吸
収材を軸方向上半分に装荷した場基準(零)として示し
である。なお、この図の約130WD/Tの燃焼時点に
おける中性子吸収材(ボロン)の原子数密度は約2X1
0”・個/cm”である。Figure 5 is a diagram showing the relationship between the core output and the coolant void reactivity coefficient difference in the case of uranium fuel. The atomic number density of a certain boron isotope (mass 10) is approximately 1 x 1611 atoms/Kokuyi when the fuel is new, and the figure shows the value at the time of combustion of approximately 130 WD/T, which corresponds to the final stage of the combustion cycle. In FIG. 5, curve 1 is when the neutron absorber is not loaded, curve 6 is when the neutron absorber is loaded over the entire effective length, and curve C is when the neutron absorber is loaded in the axial direction as in the embodiment of the present invention. It is shown as a half-loaded field standard (zero). In addition, the atomic number density of the neutron absorber (boron) at the time of combustion of approximately 130WD/T in this figure is approximately 2X1
0".pieces/cm".
第5図よシ、中性子吸収材を装荷した方が装荷しない場
合よシも一4X10−’Δに/に/ボイド程度冷却材ボ
イド反応度係数を負にしていることがわかる。また、炉
心出力20%程度までは、中性子吸収材を有効長全長に
装荷した場合と軸方向上半分に装荷し九場合とで冷却材
ボイド反応度係数差が全く同一で、炉心出力が20%以
上であっても、それが約40%程度までは、両者の間の
差がわずかであることがわかる。なお、炉心出力が約4
0%以上では、冷却材ボイド反応度係数の出力係数に対
する重みが小さくなるため、冷却材ボイド反応度係数差
が多少正方向に向っても炉心の出力上昇制御上特に問題
になることはない。As shown in FIG. 5, it can be seen that loading the neutron absorber makes the coolant void reactivity coefficient negative by about 4X10-'Δ compared to not loading it. Furthermore, up to about 20% of the core power, the difference in the coolant void reactivity coefficient is exactly the same when the neutron absorber is loaded over the entire effective length and when it is loaded in the upper half in the axial direction, and the core power is reduced to 20%. Even if it is above, it can be seen that the difference between the two is slight up to about 40%. In addition, the core power is approximately 4
If it is 0% or more, the weight of the coolant void reactivity coefficient with respect to the output coefficient becomes small, so even if the coolant void reactivity coefficient difference goes in the positive direction somewhat, it does not pose a particular problem in controlling the increase in the power of the core.
次に、本発明の実施例のように中性子吸収材を軸方向上
半分に装荷するようにすると、有効長全長にわ九って装
荷した場合より有利になる点について説明する。Next, it will be explained that loading the neutron absorbing material in the upper half in the axial direction as in the embodiment of the present invention is more advantageous than loading it across the entire effective length.
第6図は燃料取出時の燃料集合体中央部の中性子吸収材
(ボロン)の原子数密度と取出平均燃焼度差との関係線
図で、6曲線は中性子吸収材を有効長全長に装荷した場
合、6曲線は軸方向上半分に装荷した場合の特性曲線で
ある。第6図に示すように、中性子吸収材を有効長全長
に装荷した場合よシ本発明に係るように軸方向上半分に
装荷した方が取出平均燃焼度差が向上する0例えば、燃
料取出時の中性子吸収材の原子数密度が2X1(F@個
/cm ”の場合、中性子吸収材を有効長全長に装荷し
九場合は、燃焼度が約&5GWD/’r低下するのに対
し、軸方向上半分に装荷した場合のそれは、約1.60
WD/Tの低下にとどまシ、約20WD/Tだけ燃焼度
が向上する。Figure 6 is a relationship diagram between the atomic number density of the neutron absorbing material (boron) in the center of the fuel assembly at the time of fuel removal and the difference in average burnup of the fuel assembly. In this case, curve 6 is the characteristic curve when the upper half in the axial direction is loaded. As shown in Fig. 6, when the neutron absorber is loaded over the entire effective length, the difference in the average burn-up of the extraction is improved when the neutron absorber is loaded in the upper half in the axial direction as in the present invention. When the atomic number density of the neutron absorber is 2X1 (F@pieces/cm), when the neutron absorber is loaded over the entire effective length, the burnup decreases by about &5GWD/'r, whereas in the axial direction When loaded in the upper half, it is approximately 1.60
Although the WD/T only decreases, the burnup improves by about 20 WD/T.
上記し九ように、本発明の実施例によれば、原子炉起動
時の低出力時の冷却材ボイド反応度係数をさらに負側に
することができ、起動時の運転制御を容易にできる。ま
た、燃料の達成燃焼度を従来と同一の濃縮度でさらに向
上することができる。As described above, according to the embodiment of the present invention, the coolant void reactivity coefficient at low power during reactor startup can be made more negative, and operation control during startup can be facilitated. Further, the achieved burnup of the fuel can be further improved with the same enrichment as before.
なお、上記した実施例では、中性子吸収材として炭化ボ
ロンの粉末を用いたが、炭化ボロンの焼結ペレットある
いはほうけい酸ガラスを用いるようにしてもよく、同様
の効果を得ることができる。In the above embodiment, boron carbide powder was used as the neutron absorbing material, but sintered boron carbide pellets or borosilicate glass may be used, and similar effects can be obtained.
また、中性子吸収材である炭化ボロン17を燃料集合体
径方向中央のスペーサタイロッド12の上部円管14内
に収納するようにしたが、燃料棒の軸方向上部1/2の
燃料ペレット中に中性子吸収材を混入するようにしても
よく、ただし、この場合は中性子吸収材として酸化ガド
リニウムを用いる。このようにしても、上記した実施例
と同一の効果が得られる。まえ、実施例では中性子吸収
材を軸方向上半分に装荷し九が、下流側2/3の領域の
範囲内であれば、同様の効果がある。In addition, although boron carbide 17, which is a neutron absorbing material, is stored in the upper circular tube 14 of the spacer tie rod 12 at the radial center of the fuel assembly, neutron An absorbing material may be mixed, however, in this case, gadolinium oxide is used as the neutron absorbing material. Even in this case, the same effect as the above embodiment can be obtained. In the embodiment, the neutron absorbing material is loaded in the upper half in the axial direction, but the same effect can be obtained if the neutron absorbing material is loaded within the 2/3 region on the downstream side.
以上説明したように、本発明によれば、低出力時の冷却
材ボイド反応度係数上さらに負側にすることができ、原
子炉起動時の運転制御を容易にでき、かつ、燃料の達成
燃焼度をさらに向上することができるという効果がある
。As explained above, according to the present invention, the coolant void reactivity coefficient at low power can be made more negative, operation control at reactor startup can be easily controlled, and the achieved fuel combustion This has the effect of further improving the level of performance.
体の一実施例を示す中央部分縦断面図、第3図は第2図
のスペーサタイロッドの一実施例を示す縦断面図、第4
図は炉心出力と冷却材の沸騰開始点位置との関係を示す
線図、第5図は炉心出力と冷却材ボイド反応度係数の差
との関係を示す線図、第6図は中性子吸収材の原子数密
度と取出平均燃焼度差との関係線図でら°る。
7・・・燃料ペレツ)、8−・・被覆管、9・・・燃料
スペーサ、12・・・スペーサタイロッド、13−・・
中性子吸収材、14・・・円管、15・・・連結用端栓
、17・・・炭大戸・(上刃 (7#〕
pC出力(#+42FIG. 3 is a longitudinal sectional view of the central part showing one embodiment of the body; FIG. 3 is a longitudinal sectional view of the spacer tie rod of FIG. 2;
Figure 5 is a diagram showing the relationship between core power and coolant boiling start point position, Figure 5 is a diagram showing the relationship between core power and difference in coolant void reactivity coefficient, and Figure 6 is a diagram showing neutron absorbing material. A diagram of the relationship between the atomic number density and the average burnup difference is shown. 7... fuel pellets), 8-... cladding tube, 9... fuel spacer, 12... spacer tie rod, 13-...
Neutron absorbing material, 14...Circular tube, 15...Connection end plug, 17...Charcoal door (upper blade (7#)) pC output (#+42
Claims (1)
料集合体において、鋏燃料集舎体の炉心有効長の冷却材
下流側273の領域の範囲内に中性子吸収材を装荷して
なること1*黴とする重水減速圧力管型原子炉用燃料集
合体。 2、前記中性子吸収材が前記燃料棒に沿って配置してあ
シ、中性子吸収材として炭化はう素あるいはほう素を含
むガラス食用い、該中性子吸収材の周囲を被覆管で包ん
である特許請求O@s第1項記載の重水減速圧力管型原
子炉用燃料集合体。 3、前記中性子吸収材が燃料集合体の中央位置に装荷し
である特許請求の範囲第2項記載の重水減速圧力管型原
子炉用燃料集合体。 4、前記中性子吸収材は酸化ガドリJlLりムよ)なシ
前記燃料棒の燃料物質中に添加して多る特許請求の範囲
第1項記載の重水減速圧力管型原子炉用燃料集合体。[Claims] 1. In a fuel assembly for a heavy water moderated pressure tube type nuclear reactor consisting of a large number of fuel rods, neutron absorption occurs within the range of the coolant downstream side 273 of the core effective length of the scissor fuel assembly. A fuel assembly for a heavy water moderation pressure tube type nuclear reactor that is loaded with materials and made into mold. 2. A patent in which the neutron absorbing material is arranged along the fuel rod, a glass material containing boron carbide or boron is used as the neutron absorbing material, and the neutron absorbing material is surrounded by a cladding tube. A fuel assembly for a heavy water moderating pressure tube type nuclear reactor according to claim O@s. 3. The fuel assembly for a heavy water moderating pressure tube type nuclear reactor according to claim 2, wherein the neutron absorbing material is loaded at a central position of the fuel assembly. 4. The fuel assembly for a heavy water moderated pressure tube type nuclear reactor according to claim 1, wherein the neutron absorbing material is added to the fuel material of the fuel rod, such as oxidized neutron JIL rim.
Priority Applications (1)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
JP56108641A JPS5810678A (en) | 1981-07-10 | 1981-07-10 | Fuel assembly for heavy water moderated pressure tube type reactor |
Applications Claiming Priority (1)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
JP56108641A JPS5810678A (en) | 1981-07-10 | 1981-07-10 | Fuel assembly for heavy water moderated pressure tube type reactor |
Publications (2)
Publication Number | Publication Date |
---|---|
JPS5810678A true JPS5810678A (en) | 1983-01-21 |
JPS6331062B2 JPS6331062B2 (en) | 1988-06-22 |
Family
ID=14489940
Family Applications (1)
Application Number | Title | Priority Date | Filing Date |
---|---|---|---|
JP56108641A Granted JPS5810678A (en) | 1981-07-10 | 1981-07-10 | Fuel assembly for heavy water moderated pressure tube type reactor |
Country Status (1)
Country | Link |
---|---|
JP (1) | JPS5810678A (en) |
Citations (2)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
JPS5811163A (en) * | 1981-07-11 | 1983-01-21 | Nippon Telegr & Teleph Corp <Ntt> | Color recording by ink jet printer |
JPS5967060A (en) * | 1982-10-07 | 1984-04-16 | Ricoh Co Ltd | Color recorder |
-
1981
- 1981-07-10 JP JP56108641A patent/JPS5810678A/en active Granted
Patent Citations (2)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
JPS5811163A (en) * | 1981-07-11 | 1983-01-21 | Nippon Telegr & Teleph Corp <Ntt> | Color recording by ink jet printer |
JPS5967060A (en) * | 1982-10-07 | 1984-04-16 | Ricoh Co Ltd | Color recorder |
Also Published As
Publication number | Publication date |
---|---|
JPS6331062B2 (en) | 1988-06-22 |
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