JPS64675B2 - - Google Patents
Info
- Publication number
- JPS64675B2 JPS64675B2 JP58096695A JP9669583A JPS64675B2 JP S64675 B2 JPS64675 B2 JP S64675B2 JP 58096695 A JP58096695 A JP 58096695A JP 9669583 A JP9669583 A JP 9669583A JP S64675 B2 JPS64675 B2 JP S64675B2
- Authority
- JP
- Japan
- Prior art keywords
- fuel
- nuclear
- nuclear fuel
- pellet
- region
- 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.)
- Expired
Links
- 239000003758 nuclear fuel Substances 0.000 claims description 70
- 239000008188 pellet Substances 0.000 claims description 60
- 238000005253 cladding Methods 0.000 claims description 17
- 239000013078 crystal Substances 0.000 claims description 16
- 229910044991 metal oxide Inorganic materials 0.000 claims description 9
- 150000004706 metal oxides Chemical class 0.000 claims description 9
- 239000000463 material Substances 0.000 claims description 4
- WZECUPJJEIXUKY-UHFFFAOYSA-N [O-2].[O-2].[O-2].[U+6] Chemical group [O-2].[O-2].[O-2].[U+6] WZECUPJJEIXUKY-UHFFFAOYSA-N 0.000 claims 1
- 229910000484 niobium oxide Inorganic materials 0.000 claims 1
- URLJKFSTXLNXLG-UHFFFAOYSA-N niobium(5+);oxygen(2-) Chemical group [O-2].[O-2].[O-2].[O-2].[O-2].[Nb+5].[Nb+5] URLJKFSTXLNXLG-UHFFFAOYSA-N 0.000 claims 1
- 229910000439 uranium oxide Inorganic materials 0.000 claims 1
- 239000000446 fuel Substances 0.000 description 45
- 239000007789 gas Substances 0.000 description 38
- 239000002245 particle Substances 0.000 description 10
- 238000010586 diagram Methods 0.000 description 9
- 239000000654 additive Substances 0.000 description 8
- 238000009792 diffusion process Methods 0.000 description 5
- 230000000694 effects Effects 0.000 description 5
- 238000009835 boiling Methods 0.000 description 4
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 4
- 229910010413 TiO 2 Inorganic materials 0.000 description 3
- FCTBKIHDJGHPPO-UHFFFAOYSA-N uranium dioxide Inorganic materials O=[U]=O FCTBKIHDJGHPPO-UHFFFAOYSA-N 0.000 description 3
- 238000002485 combustion reaction Methods 0.000 description 2
- 230000003247 decreasing effect Effects 0.000 description 2
- 230000004992 fission Effects 0.000 description 2
- OOAWCECZEHPMBX-UHFFFAOYSA-N oxygen(2-);uranium(4+) Chemical compound [O-2].[O-2].[U+4] OOAWCECZEHPMBX-UHFFFAOYSA-N 0.000 description 2
- ZCYVEMRRCGMTRW-UHFFFAOYSA-N 7553-56-2 Chemical compound [I] ZCYVEMRRCGMTRW-UHFFFAOYSA-N 0.000 description 1
- GWEVSGVZZGPLCZ-UHFFFAOYSA-N Titan oxide Chemical compound O=[Ti]=O GWEVSGVZZGPLCZ-UHFFFAOYSA-N 0.000 description 1
- WGLPBDUCMAPZCE-UHFFFAOYSA-N Trioxochromium Chemical compound O=[Cr](=O)=O WGLPBDUCMAPZCE-UHFFFAOYSA-N 0.000 description 1
- 238000010521 absorption reaction Methods 0.000 description 1
- 230000000996 additive effect Effects 0.000 description 1
- 239000011651 chromium Substances 0.000 description 1
- 229910000423 chromium oxide Inorganic materials 0.000 description 1
- 230000007797 corrosion Effects 0.000 description 1
- 238000005260 corrosion Methods 0.000 description 1
- 238000005336 cracking Methods 0.000 description 1
- 239000001307 helium Substances 0.000 description 1
- 229910052734 helium Inorganic materials 0.000 description 1
- SWQJXJOGLNCZEY-UHFFFAOYSA-N helium atom Chemical compound [He] SWQJXJOGLNCZEY-UHFFFAOYSA-N 0.000 description 1
- 230000003993 interaction Effects 0.000 description 1
- 229910052740 iodine Inorganic materials 0.000 description 1
- 239000011630 iodine Substances 0.000 description 1
- QHMGFQBUOCYLDT-RNFRBKRXSA-N n-(diaminomethylidene)-2-[(2r,5r)-2,5-dimethyl-2,5-dihydropyrrol-1-yl]acetamide Chemical compound C[C@@H]1C=C[C@@H](C)N1CC(=O)N=C(N)N QHMGFQBUOCYLDT-RNFRBKRXSA-N 0.000 description 1
- ZKATWMILCYLAPD-UHFFFAOYSA-N niobium pentoxide Inorganic materials O=[Nb](=O)O[Nb](=O)=O ZKATWMILCYLAPD-UHFFFAOYSA-N 0.000 description 1
- 238000005245 sintering Methods 0.000 description 1
- OGIDPMRJRNCKJF-UHFFFAOYSA-N titanium oxide Inorganic materials [Ti]=O OGIDPMRJRNCKJF-UHFFFAOYSA-N 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
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Solid Fuels And Fuel-Associated Substances (AREA)
- Processes Of Treating Macromolecular Substances (AREA)
Description
〔発明の利用分野〕
本発明は、原子炉燃料棒、特に核分裂性物質を
含み円柱状の焼結体よりなる核燃料ペレツトが充
填されている原子炉燃料棒に関するものである。
〔発明の背景〕
原子炉燃料棒に充填される燃料を沸騰水型原子
炉で用いる沸騰水型燃料棒によつて説明する。第
１図はその要部切欠き正面図で、１は円筒状の被
覆管、２は円筒状の被覆管１内に充填されている
二酸化ウラン（UO2）を焼結した核燃料ペレツ
ト、３は核燃料ペレツト２を支持するためのプレ
ナムスプリング、４は核燃料ペレツト２と被覆管
１との間のギヤツプ空間で、その中にはギヤツプ
空間４の熱伝達の向上のためにヘリウム（He）
ガスが充填されている。５及び６は被覆管１の
上、下端を密封する端栓を示している。
原子炉燃料棒は、通常原子炉内で４年間燃焼す
るが、燃焼するのに従つて核分裂の結果生じた気
体状核分裂生成物（以下、FPガスと称する）が、
核燃料ペレツト２からギヤツプ空間４に徐々に放
出される。
この核燃料ペレツト２からのFPガス放出量が
増加すると、燃料棒内のガス圧が上昇するほか、
核燃料ペレツト２と被覆管１との間のギヤツプ空
間４の熱伝達が悪化し、燃料温度が上昇する。そ
の結果、さらにFPガス放出の増加が生じ、燃料
の破損確率が増加する可能性がある。従つて、破
損確率を小さく抑え、燃料の健全性を維持してい
くには、核燃料ペレツト２からのFPガス放出を
できるだけ低減する必要がある。
また、核燃料ペレツト２は熱膨張するほか、ペ
レツト内部に蓄積されるFPガスの増加によりス
ウエリングするため、被覆管１と接触するように
なる。その結果、被覆管１に応力や歪みが生じる
ようになるが、核燃料ペレツト２がクリープ変形
し、応力を緩和させる効果を持つため、一定出力
時においては被覆管１に生じる応力および歪みは
安全な範囲に抑えられている。
しかし、燃料の出力を急激に上昇させると、燃
料温度が急上昇しFPガスが急激に放出される可
能性があるほか、核燃料ペレツト２の熱膨張によ
り核燃料ペレツト２と被覆管１との間に機械的相
互作用が生じる。この時、核燃料ペレツト２のク
リープ速さが被覆管１に生じる応力の増加の速さ
についていけなくなると、被覆管１に大きな応力
や歪みが生じる可能性がある。
そこで、従来は、燃料の出力や出力上昇速度に
制限を設け、燃料の健全性を維持した運転が行な
われてきた。しかし、この運転上の制限は、プラ
ント利用率を向上させるという面での障害となつ
て来た。
〔発明の目的〕
本発明は、これらの問題点を除去し、燃料の健
全性を維持しながら、かつ出力および出力上昇度
の制限を緩和し、効率の良い原子炉の運転を実現
可能とする原子炉燃焼棒を提供することを目的と
するものである。
〔発明の概要〕
本発明は、円筒状の被覆管内に核分裂性物質を
含み円柱状の焼結体よりなる原子炉用の核燃料ペ
レツトが充填されている原子炉燃料棒において、
前記核燃料ペレツトに結晶粒径調整用の金属酸化
物を添加し、該核燃料ペレツトの径方向の内側領
域の結晶粒径が径方向外側領域の結晶粒径より大
きくなつていることを特徴とするものである。
本発明は、原子炉運転時における核燃料ペレツ
トからのFPガス放出を低減することにより、燃
料棒内のガス圧の上昇を防止すると同時に、核燃
料ペレツト内のクリープ速さを増加させることに
より、被覆管に生じる応力を低減し、健全性余裕
を増加させた原子炉燃料棒の提供が可能な点に着
目してなされたものである。
核燃料ペレツト中で生成されたFPガスは、拡
散により結晶粒界へと移動し、結晶粒界に蓄積さ
れる。結晶粒界でのFPガス原子密度が高くなり、
FPガスが互いに連結するようになると、核燃料
ペレツトからのFPガス放出が始まる。
核燃料ペレツトから放出されるFPガス放出量
の全生成量に対する割合は、FPガス原子の拡散
定数をＤ、結晶粒の直径をａとすると、
[Field of Application of the Invention] The present invention relates to a nuclear reactor fuel rod, and particularly to a nuclear reactor fuel rod filled with nuclear fuel pellets containing fissile material and consisting of a cylindrical sintered body. [Background of the Invention] The fuel filled in nuclear reactor fuel rods will be explained in terms of boiling water fuel rods used in boiling water reactors. Figure 1 is a cutaway front view of the main parts, where 1 is a cylindrical cladding tube, 2 is a nuclear fuel pellet made by sintering uranium dioxide (UO 2 ) filled in the cylindrical cladding tube 1, and 3 is a nuclear fuel pellet filled with uranium dioxide (UO 2 ). A plenum spring 4 for supporting the nuclear fuel pellet 2 is a gap space between the nuclear fuel pellet 2 and the cladding tube 1, in which helium (He) is inserted to improve heat transfer in the gap space 4.
Filled with gas. 5 and 6 indicate end plugs that seal the upper and lower ends of the cladding tube 1. Reactor fuel rods usually burn for four years in a nuclear reactor, and as they burn, gaseous fission products (hereinafter referred to as FP gas) produced as a result of nuclear fission are released.
The nuclear fuel pellets 2 are gradually released into the gap space 4. As the amount of FP gas released from nuclear fuel pellets 2 increases, the gas pressure inside the fuel rods increases, and
Heat transfer in the gap space 4 between the nuclear fuel pellets 2 and the cladding tube 1 deteriorates, and the fuel temperature increases. This may further result in increased FP gas emissions and an increased probability of fuel failure. Therefore, in order to keep the probability of damage low and maintain the integrity of the fuel, it is necessary to reduce the release of FP gas from the nuclear fuel pellets 2 as much as possible. In addition to thermal expansion, the nuclear fuel pellet 2 also swells due to an increase in the FP gas accumulated inside the pellet, so it comes into contact with the cladding tube 1. As a result, stress and distortion occur in the cladding tube 1, but since the nuclear fuel pellets 2 undergo creep deformation and have the effect of relieving the stress, the stress and distortion generated in the cladding tube 1 are safely reduced at a constant output. It is kept within a range. However, if the fuel output is suddenly increased, the fuel temperature may rise rapidly and FP gas may be rapidly released, and the thermal expansion of the nuclear fuel pellets 2 may cause mechanical damage between the nuclear fuel pellets 2 and the cladding tube 1. interaction occurs. At this time, if the creep rate of the nuclear fuel pellets 2 cannot keep up with the rate of increase in stress occurring in the cladding tube 1, large stress and distortion may occur in the cladding tube 1. Therefore, in the past, limits have been placed on the fuel output and the rate of increase in the output to maintain the integrity of the fuel. However, this operational limitation has become an obstacle in improving plant utilization. [Objective of the Invention] The present invention eliminates these problems, maintains the integrity of the fuel, relaxes restrictions on output and output increase rate, and enables efficient nuclear reactor operation. The purpose is to provide nuclear reactor combustion rods. [Summary of the Invention] The present invention provides a nuclear reactor fuel rod in which a cylindrical cladding tube is filled with nuclear fuel pellets for a nuclear reactor, which are made of a cylindrical sintered body containing a fissile material,
A metal oxide for adjusting the crystal grain size is added to the nuclear fuel pellet, and the crystal grain size in the radially inner region of the nuclear fuel pellet is larger than the crystal grain size in the radially outer region. It is. The present invention reduces the release of FP gas from nuclear fuel pellets during reactor operation, thereby preventing an increase in gas pressure within the fuel rods, and at the same time increasing the creep speed within the nuclear fuel pellets, thereby increasing the This was done with the focus on the ability to provide nuclear reactor fuel rods that reduce the stress generated in the reactor and increase the integrity margin. The FP gas generated in the nuclear fuel pellet moves to the grain boundaries by diffusion and is accumulated there. The FP gas atom density at grain boundaries increases,
Once the FP gases become interconnected, the nuclear fuel pellet begins to release the FP gases. The ratio of the amount of FP gas emitted from nuclear fuel pellets to the total amount produced is as follows, where D is the diffusion constant of FP gas atoms and a is the diameter of the crystal grain.
以下、実施例について説明する。
第３図は一実施例の核燃料ペレツトの構造を示
すもので、核燃料ペレツト２は、内層領域２ａと
外層領域２ｂとを有しており、内層領域２ａは
UO2に金属酸化物として酸化ニオブ（Ｖ）
（Nb2O5）、酸化クロム（）（Cr2O3）、酸化チタ
ン（）（TiO2）等が添加してあり、結晶粒径は
30μｍ以上になつているが、外層領域はUO2より
なり粒径は約10μｍになつている。
核燃料ペレツトからのFPガス放出率ｆは、FP
ガス原子の拡散定数をＤ、FPガス放出開始後の
時間をｔ、結晶粒の直径をａとすれば、次式で与
えられる。
拡散定数Ｄは、第２図に示したように燃料温度
が1200℃を越えると急激に増大するため、粒径が
同一の場合、FPガス放出率も燃料温度が1200℃
以上で急激に増加する。
従来の沸騰水型原子炉では、燃料の熱的および
機械的余裕を十分維持するように、燃料の線出力
密度がおよそ13.4kW／ft以下で運転されている。
この時の核燃料ペレツト内の温度分布を第４図に
示す。この図の横軸及び縦軸にはそれぞれ核燃料
ペレツト半径方向位置（相対値）及び燃料温度が
とつてある。燃料中心部では、燃料温度がおよそ
1600℃で、燃料外周部ではおよそ400℃となつて
いる。この温度分布は、核燃料ペレツト−被覆管
のギヤツプコンダクタンスが照射中において最も
低い時のもので、そのため、核燃料ペレツトの温
度が最も高い場合に対応している。第５図は線出
力13.4kW／ftまで出力上昇し、FPガス放出が生
じた場合のFPガス放出率分布を示すもので、こ
の図の横軸及び縦軸にそれぞれ核燃料ペレツト内
径方向位置（相対値）及びFPガス放出率（％）
がとつてあり、Ａは粒径10μｍの従来の燃料のFP
ガス放出率分布、Ｂは粒径43μｍの場合のFPガス
放出率分布、Ｃはこの実施例の核燃料ペレツトの
場合のFPガス放出率分布を示している。この図
から燃料温度が1200℃以上である領域、即ち、燃
料中心部からの距離ｒが、核燃料ペレツト半径Ｒ
の3/5より小さい（ｒ＜0.6R）領域では、FPガス
放出率が１％以上となつていることが分る。
一方、(1)式から分るように結晶粒径を大きくす
ると、FPガス放出率を低減できる。そこで核燃
料ペレツトの燃料温度1200℃以上となる領域にあ
らかじめNb2O5やTiO2等の金属酸化物を添加し
て結晶粒径を増加させておき、FPガス放出を低
減する。燃料温度が1200℃以下の比較的低温の部
分については、添加物なしの核燃料ペレツトとし
て、金属酸化物添加時に生じる燃料クリープ速さ
の低下を防止する。
Nb2O5を添加した場合の添加濃度と結晶粒径と
の関係を第６図に示した。この図で横軸及び縦軸
にはそれぞれNb2O5添加濃度（mol％）及び結晶
粒径（μｍ）がとつてある。この図から明らかな
ように、Nb2O5をおよそ0.5mol％添加した場合に
粒径は最大43μｍに達する。粒径を従来の10μｍ
から43μｍに増加すると、燃料温度が同一の場
合、FPガス放出率をおよそ1/4に低減できる。
FPガス放出率を低減するには、結晶粒径をでき
るだけ増加することが望ましく、第６図から分る
ようにNb2O5添加濃度を0.4〜0.6mol％の範囲に
すれば粒径35μｍ以上を得ることができ、その結
果FPガス放出率を従来の核燃料ペレツトの1/3以
下に抑えることができる。
結晶粒径の最大値43μｍを得るようにNb2O5を
0.5mol％添加した場合の核燃料ペレツトの温度
とクリープ速さの関係、および添加物なしのUO2
ペレツトの温度とクリープ速さの関係を第７図に
示した。この図の横軸及び縦軸にはそれぞれ燃料
温度（℃）及び燃料クリープ速さ（相対値）がと
つてあり、ＤはNb2O5を添加した粒径43μｍの場
合のクリープ速さ、Ｅは添加物なしの粒径10μｍ
の場合のクリープ速さを示しており、核燃料ペレ
ツトに10MN／m2の力が加わつた場合の値を示し
ている。この図から分るように、燃料温度1180℃
以上では、Nb2O5を添加した場合の方がクリープ
速さが大きくなつているが、1180℃以下では添加
物なしのUO2の方がクリープ速さが大きい。線出
力密度13.4kW／ftで運転された場合の核燃料ペ
レツト内の径方向、周方向、軸方向の応力の分布
を第８図に示す。この図の横軸、縦軸にはそれぞ
れ（核燃料ペレツト径方向距離）／（核燃料ペレ
ツト半径）及び応力（相対値）がとつてあり、Ｆ
は核燃料ペレツト内径方向応力分布、Ｇは核燃料
ペレツト内周方向応力分布、Ｈは核燃料ベレツト
内軸方向応力分布を示している。核燃料ペレツト
のクリープは径方向、周方向、軸方向の応力に差
がある場合にのみ生じる。第８図の規格化した径
方向距離が0.6以下の領域では、３つの応力は等
しく、クリープは生じない。径方向距離が0.6以
上の領域では、第２図より分るように、燃料温度
がおよそ1200℃以下となつている。また、第５図
のＡに示したように、燃料温度1200℃以下では、
もともとFPガス放出が十分少ない領域であるこ
とを考慮すると、Nb2O5を添加せず、燃料のクリ
ープ速さ低下を防止することが有効であることが
分る。
また、添加物を入れた場合、中性子吸収が増加
するため、効率のよい燃料の燃焼を維持するには
添加物は可能な限り最小限に抑えることが必要で
ある。従つて、本実施例では、燃料温度1200℃以
下となる領域では、金属酸化物を添加せずにUO2
をそのまま焼結した核燃料ペレツトとする。
このように燃料温度1200℃以上の領域にのみ、
Nb2O5を0.5mol％添加した場合のFPガス放出率
を示したのが第５図のＣである。高燃料温度の部
分の結晶粒径を増加させているため、FPガス放
出率は、従来燃料のおよそ1/4に抑えられている。
本実施例では、Nb2O5を添加した場合について
説明したが、TiO2等の添加物を用いた場合でも
同様の粒径増加の効果とクリープ速さ低下の影響
があり、燃料温度1200℃以上の領域にのみ添加す
ることにより燃料の性能を向上させることができ
る。
以上述べたように、本実施例によれば、核燃料
ペレツトのクリープ歪み速度を低下させることな
く、FPガス放出率を低下させることができる。
その結果、ギヤツプ空間内のヨウ素濃度が減少
し、被覆管に過大な応力が発生するのを防止でき
る。以上の効果により、燃料棒の応力腐食割れ発
生確率が低下し、燃料棒の健全性を向上させるこ
とができる。
〔発明の効果〕
本発明の原子炉燃料棒は、燃料の健全性を維持
しながら、かつ、出力および出力上昇度の制限を
緩和し、効率の良い原子炉の運転を実現可能とす
る原子炉燃料棒を提供するもので、産業上の効果
の大なるものである。
Examples will be described below. FIG. 3 shows the structure of a nuclear fuel pellet according to an embodiment. The nuclear fuel pellet 2 has an inner layer region 2a and an outer layer region 2b.
Niobium(V) oxide as metal oxide in UO2
(Nb 2 O 5 ), chromium oxide ( ) (Cr 2 O 3 ), titanium oxide ( ) (TiO 2 ), etc. are added, and the crystal grain size is
The particle size is 30 μm or more, but the outer layer region is made of UO 2 and has a particle size of about 10 μm. The FP gas release rate f from nuclear fuel pellets is FP
If the diffusion constant of gas atoms is D, the time after the start of FP gas release is t, and the diameter of the crystal grain is a, then it is given by the following equation. As shown in Figure 2, the diffusion constant D increases rapidly when the fuel temperature exceeds 1200℃, so when the particle size is the same, the FP gas release rate also increases when the fuel temperature exceeds 1200℃.
Above that, it increases rapidly. Conventional boiling water reactors operate at fuel linear power densities of approximately 13.4 kW/ft or less to maintain sufficient fuel thermal and mechanical headroom.
The temperature distribution inside the nuclear fuel pellet at this time is shown in FIG. The horizontal and vertical axes of this figure represent the nuclear fuel pellet radial position (relative value) and fuel temperature, respectively. In the fuel center, the fuel temperature is approximately
The temperature is 1600℃, and the temperature at the outer periphery of the fuel is approximately 400℃. This temperature distribution corresponds to the lowest nuclear fuel pellet-cladding gap conductance during irradiation and therefore corresponds to the highest nuclear fuel pellet temperature. Figure 5 shows the FP gas release rate distribution when the linear output increases to 13.4kW/ft and FP gas is released. value) and FP gas release rate (%)
A is the FP of conventional fuel with a particle size of 10 μm.
Gas release rate distribution, B shows the FP gas release rate distribution in the case of a particle size of 43 μm, and C shows the FP gas release rate distribution in the case of the nuclear fuel pellet of this example. This figure shows that the area where the fuel temperature is 1200°C or higher, that is, the distance r from the fuel center, is the radius R of the nuclear fuel pellet.
It can be seen that in the region where r is smaller than 3/5 of R (r<0.6R), the FP gas release rate is 1% or more. On the other hand, as seen from equation (1), increasing the crystal grain size can reduce the FP gas release rate. Therefore, metal oxides such as Nb 2 O 5 and TiO 2 are added in advance to the region of the nuclear fuel pellet where the fuel temperature is 1200°C or higher to increase the crystal grain size, thereby reducing FP gas release. Regarding the relatively low temperature portion where the fuel temperature is 1200°C or less, nuclear fuel pellets without additives are used to prevent the decrease in fuel creep rate that occurs when metal oxides are added. FIG. 6 shows the relationship between the concentration of Nb 2 O 5 added and the crystal grain size. In this figure, the Nb 2 O 5 addition concentration (mol %) and crystal grain size (μm) are plotted on the horizontal and vertical axes, respectively. As is clear from this figure, when approximately 0.5 mol% of Nb 2 O 5 is added, the particle size reaches a maximum of 43 μm. Particle size reduced to conventional 10μm
When increasing from 43 μm to 43 μm, the FP gas release rate can be reduced to approximately 1/4 when the fuel temperature is the same.
In order to reduce the FP gas release rate, it is desirable to increase the crystal grain size as much as possible, and as shown in Figure 6, if the Nb 2 O 5 addition concentration is in the range of 0.4 to 0.6 mol%, the grain size can be increased to 35 μm or more. As a result, the FP gas release rate can be suppressed to less than 1/3 of that of conventional nuclear fuel pellets. Nb 2 O 5 was added to obtain a maximum grain size of 43 μm.
Relationship between temperature and creep rate of nuclear fuel pellets when 0.5 mol% is added, and UO 2 without additives
Figure 7 shows the relationship between pellet temperature and creep rate. The horizontal and vertical axes of this figure show the fuel temperature (°C) and fuel creep rate (relative values), respectively, where D is the creep rate when Nb 2 O 5 is added and the particle size is 43 μm, and E is is particle size 10μm without additives
It shows the creep speed when a force of 10 MN/m 2 is applied to the nuclear fuel pellet. As you can see from this diagram, the fuel temperature is 1180℃
Above, the creep rate is higher when Nb 2 O 5 is added, but below 1180°C, the creep rate is higher for UO 2 without additives. Figure 8 shows the stress distribution in the radial direction, circumferential direction, and axial direction within the nuclear fuel pellet when operating at a linear power density of 13.4 kW/ft. The horizontal and vertical axes of this figure are (nuclear fuel pellet radial distance)/(nuclear fuel pellet radius) and stress (relative value), respectively.
indicates the stress distribution in the inner diameter direction of the nuclear fuel pellet, G indicates the stress distribution in the inner circumferential direction of the nuclear fuel pellet, and H indicates the stress distribution in the inner axial direction of the nuclear fuel pellet. Creep in nuclear fuel pellets occurs only when there are differences in radial, circumferential, and axial stresses. In the region where the normalized radial distance is 0.6 or less in FIG. 8, the three stresses are equal and no creep occurs. As can be seen from Figure 2, in the region where the radial distance is 0.6 or more, the fuel temperature is approximately 1200°C or less. Also, as shown in A of Figure 5, when the fuel temperature is below 1200℃,
Considering that this is a region where FP gas release is sufficiently low to begin with, it is found that it is effective to prevent the creep speed of the fuel from decreasing without adding Nb 2 O 5 . Also, since additives increase neutron absorption, it is necessary to minimize additives as much as possible to maintain efficient fuel combustion. Therefore, in this example, in the region where the fuel temperature is 1200°C or less, UO 2 is reduced without adding metal oxides.
is sintered as is into nuclear fuel pellets. In this way, only in areas where the fuel temperature is 1200℃ or higher,
C in FIG. 5 shows the FP gas release rate when 0.5 mol% of Nb 2 O 5 was added. By increasing the crystal grain size in the high fuel temperature area, the FP gas release rate is suppressed to approximately 1/4 of that of conventional fuels. In this example, the case where Nb 2 O 5 was added was explained, but even when additives such as TiO 2 are used, the same effect of increasing the particle size and decreasing the creep speed is obtained, and the fuel temperature is 1200°C. By adding it only to the above range, the performance of the fuel can be improved. As described above, according to this embodiment, the FP gas release rate can be reduced without reducing the creep strain rate of the nuclear fuel pellet.
As a result, the iodine concentration in the gap space is reduced, and excessive stress can be prevented from being generated in the cladding tube. Due to the above effects, the probability of stress corrosion cracking occurring in the fuel rod is reduced, and the soundness of the fuel rod can be improved. [Effects of the Invention] The nuclear reactor fuel rod of the present invention provides a nuclear reactor that maintains the integrity of the fuel, relaxes restrictions on output and output increase rate, and realizes efficient nuclear reactor operation. It provides fuel rods and has great industrial effects.
第１図は沸騰水型燃料棒の要部切欠き正面図、
第２図はFPガス原子の拡散定数の温度依存性を
示す特性線図、第３図は本発明の原子炉燃料棒の
一実施例の核燃料ペレツトの一実施例の説明図、
第４図は核燃料ペレツト内の温度分布を示す特性
線図、第５図は核燃料ペレツト内のFPガス放出
率分布を従来の場合との比較において示す特性線
図、第６図はNb2O5を添加した場合の添加濃度と
結晶粒径との関係を従来の場合との比較において
示す特性線図、第７図は燃料温度と燃料クリープ
速さとの関係を示す特性線図、第８図は核燃料ペ
レツト内の応力分布を示す特性線図である。
１……被覆管、２……核燃料ペレツト、２ａ…
…（核燃料ペレツトの）内層領域、２ｂ……（核
燃料ペレツトの）外層領域、３……プレナムスプ
リング、４……（核燃料ペレツト２と被覆管１と
の間の）ギヤツプ空間。
Figure 1 is a cutaway front view of the main parts of a boiling water fuel rod.
Fig. 2 is a characteristic diagram showing the temperature dependence of the diffusion constant of FP gas atoms, Fig. 3 is an explanatory diagram of an embodiment of a nuclear fuel pellet of an embodiment of the reactor fuel rod of the present invention,
Fig. 4 is a characteristic diagram showing the temperature distribution within the nuclear fuel pellet, Fig. 5 is a characteristic diagram showing the FP gas release rate distribution within the nuclear fuel pellet in comparison with the conventional case, and Fig. 6 is a characteristic diagram showing the temperature distribution within the nuclear fuel pellet . Figure 7 is a characteristic diagram showing the relationship between additive concentration and crystal grain size in comparison with the conventional case. Figure 7 is a characteristic diagram showing the relationship between fuel temperature and fuel creep rate. FIG. 3 is a characteristic diagram showing stress distribution within a nuclear fuel pellet. 1... Cladding tube, 2... Nuclear fuel pellet, 2a...
...Inner layer region (of the nuclear fuel pellet), 2b... Outer layer region (of the nuclear fuel pellet), 3... Plenum spring, 4... Gap space (between the nuclear fuel pellet 2 and the cladding tube 1).
Claims (1)
状の焼結体よりなる原子炉用の核燃料ペレツトが
充填されている原子炉燃料棒において、前記核燃
料ペレツトに結晶粒径調整用の金属酸化物を添加
し、該核燃料ペレツト径方向の内側領域の結晶粒
径が径方向外側領域の結晶粒径より大きくなつて
いることを特徴とする原子炉燃料棒。 ２ 前記核燃料ペレツトが、前記金属酸化物が添
加してある径方向内側領域と前記金属酸化物が添
加してない径方向外側領域とからなる特許請求の
範囲第１項記載の原子炉燃料棒。 ３ 前記核燃料ペレツトの核分裂性物質が酸化ウ
ランで、前記径方向内側領域が原子炉運転時に経
験する温度が1200℃以上になる領域であり、前記
金属酸化物が濃度0.4〜0.6モル％の酸化ニオブ
（Ｖ）であり、前記径方向外側領域が原子炉運転
時に経験する温度が1200℃より低い領域である特
許請求の範囲第２項記載の原子炉燃料棒。[Scope of Claims] 1. In a nuclear reactor fuel rod in which a cylindrical cladding tube is filled with nuclear fuel pellets for a nuclear reactor that are made of a cylindrical sintered body containing a fissile material, the nuclear fuel pellets have a crystal grain size. 1. A nuclear reactor fuel rod characterized in that a metal oxide for adjustment is added and the crystal grain size in the radially inner region of the nuclear fuel pellet is larger than the crystal grain size in the radially outer region. 2. The nuclear reactor fuel rod according to claim 1, wherein the nuclear fuel pellet comprises a radially inner region to which the metal oxide is added and a radially outer region to which the metal oxide is not added. 3. The fissile material of the nuclear fuel pellet is uranium oxide, the radially inner region is a region where the temperature experienced during reactor operation is 1200°C or higher, and the metal oxide is niobium oxide with a concentration of 0.4 to 0.6 mol%. (V), and the radially outer region is a region where the temperature experienced during reactor operation is lower than 1200° C.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP58096695A JPS59220677A (en) | 1983-05-30 | 1983-05-30 | Nuclear fuel pellet |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP58096695A JPS59220677A (en) | 1983-05-30 | 1983-05-30 | Nuclear fuel pellet |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS59220677A JPS59220677A (en) | 1984-12-12 |
| JPS64675B2 true JPS64675B2 (en) | 1989-01-09 |
Family
ID=14171903
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP58096695A Granted JPS59220677A (en) | 1983-05-30 | 1983-05-30 | Nuclear fuel pellet |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS59220677A (en) |
Families Citing this family (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS62280688A (en) * | 1986-05-30 | 1987-12-05 | 日本ニユクリア・フユエル株式会社 | Nuclear fuel pellet |
| JPS63117292A (en) * | 1986-11-04 | 1988-05-21 | 日本ニユクリア・フユエル株式会社 | Manufacture of nuclear fuel sintered body |
| GB9813696D0 (en) * | 1998-06-26 | 1998-08-26 | British Nuclear Fuels Plc | Fuel pellet |
Family Cites Families (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS58165085A (en) * | 1982-03-25 | 1983-09-30 | 日本核燃料開発株式会社 | Nuclear fuel element |
-
1983
- 1983-05-30 JP JP58096695A patent/JPS59220677A/en active Granted
Also Published As
| Publication number | Publication date |
|---|---|
| JPS59220677A (en) | 1984-12-12 |
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