JP4198168B2 - Reactor vessel thermal load relaxation device - Google Patents
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Description
本発明は原子炉容器の冷却材液面近傍の熱応力緩和、原子炉容器温度成層化界面近傍の熱応力緩和等に利用できる原子炉容器の熱荷重緩和装置に関する。 The present invention relates to a thermal load relaxation device for a reactor vessel that can be used for thermal stress relaxation in the vicinity of the coolant level of the reactor vessel, thermal stress relaxation in the vicinity of the reactor vessel temperature stratification interface, and the like.
高速増殖炉の原子炉容器は、上端を１００℃以下に保つ必要があるコンクリート壁により支持され、炉心上部プレナムに５５０℃以上の高温冷却材を有することから、冷却材液面から上端支持部の間に鉛直方向の大きな温度勾配が生じる。特に、起動時においては昇温と液位上昇が同時進行することから勾配が厳しくなる。この結果、温度勾配の変曲点となる炉壁液面近傍部に原理的に高い熱応力が発生する。 The reactor vessel of the fast breeder reactor is supported by a concrete wall whose upper end needs to be kept at 100 ° C. or lower, and has a high temperature coolant of 550 ° C. or higher in the upper plenum of the core. A large vertical temperature gradient occurs between them. In particular, at the time of start-up, since the temperature rise and the liquid level rise proceed simultaneously, the gradient becomes severe. As a result, a high thermal stress is generated in principle in the vicinity of the furnace wall liquid surface, which is the inflection point of the temperature gradient.
これに対し、従来の熱荷重緩和方策は、液位制御装置による液位上昇の防止、炉壁冷却装置による炉壁の一様冷却、および薄肉化による曲げ応力の低減であった。また、冷却材液面の下方の原子炉容器壁から容器蓋直下まで原子炉容器と協同して断熱空間を形成するライナー部を設け、断熱空間内に遮熱材を充填して液面近傍の温度勾配を緩和するものも提案されている（特許文献１）。
このように、従来の熱荷重緩和方策は、液位制御装置による液位上昇の防止、炉壁冷却装置による炉壁の一様冷却、および薄肉化による曲げ応力の低減であり、液位制御装置と炉壁冷却装置は物量増加によるコスト高を招き、薄肉化は他の破損モードの制限から限界があった。また、上記特許文献１においても物量増加によるコスト高を招くことになる。
As described above, the conventional thermal load mitigation measures are prevention of liquid level rise by the liquid level control device, uniform cooling of the furnace wall by the furnace wall cooling device, and reduction of bending stress due to thinning. However, the furnace wall cooling device incurs high costs due to an increase in the amount of material, and the thinning has a limit due to limitations of other failure modes. Also in
本発明は、上記課題を解決しようとするもので、確実な動作を図るとともに、建設コストに有意な影響を与えずに、応力の発生要因である熱荷重自体を緩和し、原子炉の安全性の向上、プラントの経済性の向上を図ることを目的とする。
そのために本発明は、外側に空間を介在させてガードベッセルが設置された原子炉容器内の冷却材の液面近傍における前記原子炉容器の熱荷重を緩和する装置において、前記ガードベッセルを原子炉容器より良熱伝導性の材料で構成し、液面下部の炉壁からの輻射熱で前記ガードベッセルを加熱し、液面上部の炉壁を前記ガードベッセルからの輻射熱で加熱することを特徴とする。
また、本発明は、前記良熱伝導性の材料は高クロム系鋼であることを特徴とする。
The present invention is intended to solve the above-mentioned problems, and while ensuring reliable operation, the thermal load itself, which is the cause of stress, is alleviated without significantly affecting the construction cost, and the safety of the reactor The purpose is to improve the efficiency of the plant and the economy of the plant.
Therefore, the present invention provides an apparatus for relaxing the thermal load of the reactor vessel in the vicinity of the coolant level in the reactor vessel in which a guard vessel is installed with a space outside, and the guard vessel is installed in the reactor. It is composed of a material having better heat conductivity than the container, and the guard vessel is heated by radiant heat from the furnace wall at the lower part of the liquid level, and the furnace wall at the upper part of the liquid level is heated by radiant heat from the guard vessel. .
Further, the present invention is characterized in that the high thermal conductivity material is high chromium steel.
本発明によれば、ガードベッセル壁を原子炉容器より良熱伝導性の材料で構成し、液面下部の炉壁からの輻射熱でガードベッセルを加熱し、液面上部の炉壁をガードベッセルからの輻射熱で加熱するようにしたので、新たに付加する部材なしに冷却材液面近傍の熱荷重を緩和することができ、建設コストに影響を与えず、非接触かつ静的構造物であるため確実に動作させることができる。 According to the present invention, the guard vessel wall is made of a material having better thermal conductivity than the reactor vessel, the guard vessel is heated by radiant heat from the lower reactor wall, and the upper reactor wall is removed from the guard vessel. Because it is heated by radiant heat, it is possible to relieve the thermal load near the coolant level without adding a new member, and it is a non-contact and static structure that does not affect the construction cost. It can be operated reliably.
以下、本発明の実施の形態について説明する。
図１は原子炉容器液面近傍の応力低減を行う熱荷重緩和装置の実施の形態の例を示す図である。
原子炉容器１の壁外面には、冷却材が万一漏洩した場合にこれを受け止めるためのガードベッセル２が設置され、原子炉容器壁とガードベッセルとの間の約１５０ｍｍ幅のアニュラス空間３には、原子炉容器の保護のために不活性ガスが満たされ、ガードベッセルの外壁には断熱材８が設けられてコンクリート温度を上げないようにしている。冷却材液面９より上部の原子炉容器内壁には断熱材１０が設けられて高温の冷却材と断熱されているためこの部分の炉壁は低温状態にある。したがって、冷却材液面より下部の高温炉壁と冷却材液面より上部の低温炉壁では鉛直方向に温度分布が発生して熱応力の発生原因となる。本実施形態では、原子炉容器壁の外側に、冷却材液面の上下に亘る範囲に良熱伝導性の材料、例えば長時間安定な黒鉛等からなる熱伝導部材を配置したもので、この例では熱伝導部材として熱伝導板２０を非接触で設置して原子炉容器壁鉛直方向の伝熱を促進して熱応力を緩和することを特徴としている。
Embodiments of the present invention will be described below.
FIG. 1 is a diagram showing an example of an embodiment of a thermal load relaxation device that reduces stress in the vicinity of the reactor vessel liquid surface.
A guard vessel 2 is installed on the outer surface of the
熱伝導板２０は図示するように、ガードベッセル等の原子炉容器外の構造物から支持する。例えば、熱伝導板の寸法は厚さ３０ｍｍ程度、長さ１５００ｍｍ程度とし、設置位置は原子炉容器壁から板表面までの距離を６０ｍｍ程度、鉛直方向位置は液面上部の板長さより液面下部の板長さを長くし、例えば液面上部の板長さ５００ｍｍ、液面下部の板長さ１０００ｍｍ程度となるようにする。 As shown in the figure, the heat conduction plate 20 is supported from a structure outside the reactor vessel such as a guard vessel. For example, the dimensions of the heat conduction plate are about 30 mm thick and about 1500 mm long, the installation position is about 60 mm from the reactor vessel wall to the plate surface, and the vertical position is lower than the plate length above the liquid level. For example, the plate length at the upper part of the liquid level is set to about 500 mm and the plate length at the lower part of the liquid level is set to about 1000 mm.
図２は原子炉容器冷却材液面近傍の熱荷重緩和の原理説明図であり、図２（ａ）は熱伝導板無しの場合、図２（ｂ）は熱伝導板有りの場合の図である。
前述したように、高速増殖炉の原子炉容器は、コンクリート構造物に支持されることから上端を１００℃以下に保つ必要がある。起動時には内包する冷却材の温度が２００℃から５５０℃まで昇温するため、その過程で生じる鉛直方向の局所的温度勾配により炉壁に高い熱応力が発生する。すなわち、原子炉容器の起動時の温度分布を成り行きまかせにした場合（図２（ａ））、高温の冷却材に接した接液部と低温のガス空間部の昇温終了時の急激な温度勾配（温度の折れ曲がり部）が発生し、昇温終了時に液面近傍炉壁外面（図２（ａ）のＳ点）に最大応力が生ずる。これを緩和するため、図２（ｂ）に示すように、液面下部の高温炉壁からの輻射熱で熱伝導板を加熱し、熱伝導板から液面上部の低温炉壁を輻射加熱することにより、応力の要因となっている鉛直方向の温度勾配を小さくする。その結果、最大応力発生位置Ｓにおける温度勾配は、最大応力発生時刻Ｔにおいて滑らかになり、熱荷重が緩和される。本実施形態では、簡素な熱伝導板を付加するだけであるので安価な上、非接触かつ静的構造物であるため確実に動作することが特徴である。
FIG. 2 is a diagram for explaining the principle of thermal load relaxation in the vicinity of the reactor vessel coolant level. FIG. 2 (a) shows a case without a heat conduction plate, and FIG. 2 (b) shows a case with a heat conduction plate. is there.
As described above, since the reactor vessel of the fast breeder reactor is supported by the concrete structure, it is necessary to keep the upper end at 100 ° C. or lower. Since the temperature of the coolant contained in the reactor rises from 200 ° C. to 550 ° C. at the time of startup, a high thermal stress is generated on the furnace wall due to a local temperature gradient in the vertical direction generated in the process. That is, when the temperature distribution at startup of the reactor vessel is randomized (FIG. 2 (a)), the rapid temperature at the end of temperature rise in the wetted part in contact with the high temperature coolant and the low temperature gas space part A gradient (bending portion of temperature) occurs, and a maximum stress is generated on the outer surface of the furnace wall near the liquid level (S point in FIG. 2A) at the end of the temperature rise. In order to alleviate this, as shown in FIG. 2B, the heat conduction plate is heated by radiant heat from the high temperature furnace wall at the lower part of the liquid surface, and the low temperature furnace wall at the upper part of the liquid surface is radiantly heated from the heat conduction plate. Thus, the temperature gradient in the vertical direction, which is a factor of stress, is reduced. As a result, the temperature gradient at the maximum stress generation position S becomes smooth at the maximum stress generation time T, and the thermal load is relaxed. In the present embodiment, since a simple heat conduction plate is only added, it is inexpensive and is characterized by a reliable operation because it is a non-contact and static structure.
図３は輻射による等価熱伝達係数と使用温度との関係を示す図である。
平行平面間の輻射伝熱の大きさは数１式で表される。
FIG. 3 is a diagram showing the relationship between the equivalent heat transfer coefficient due to radiation and the operating temperature.
The magnitude of the radiant heat transfer between the parallel planes is expressed by
ただし、ε1 、ε2 は材料に依存する射出率（受熱量に対する輻射熱量の比）で、ステンレス製の炉壁と、高クロム系鋼、例えば12Ｃｒ鋼では０．１以上、黒鉛の場合は０．８以上の値である。また、σはステファンボルツマン定数である。ステンレス製の炉壁と黒鉛の熱伝導板を想定し、熱伝導板の効果を小さめに見積もる射出率の最小値ε1 ＝０．１、ε2 ＝０．８を数１式に代入し、等価熱伝達係数ｈeqとＴ（℃）＝Ｔ1 ＝Ｔ2 （均熱状態）の関係を計算すると図３のようになる。図３において、横軸は炉壁、熱伝導板（黒鉛）の温度（Ｔ）、縦軸は等価熱伝達係数（ｈeq）で、炉壁および熱伝導板の温度が高温になるにつれ急激に熱伝達係数が増加し、６００℃近傍では輻射加熱によりガス強制循環熱伝達（５Ｗ／ｍ2 Ｋ程度）に近い熱伝達が行われることが予想できる。 However, ε 1 and ε 2 are the injection rate (ratio of the amount of radiant heat to the amount of heat received) depending on the material, 0.1 or more for stainless steel furnace walls and high chromium steels such as 12Cr steel, and in the case of graphite The value is 0.8 or more. Further, σ is a Stefan Boltzmann constant. Assuming a stainless steel furnace wall and a graphite heat conduction plate, substituting the minimum injection rate values ε 1 = 0.1 and ε 2 = 0.8 to estimate the effect of the heat conduction plate to be smaller, FIG. 3 shows the relationship between the equivalent heat transfer coefficient h eq and T (° C.) = T 1 = T 2 (soaking state). In FIG. 3, the horizontal axis is the temperature (T) of the furnace wall and the heat conduction plate (graphite), and the vertical axis is the equivalent heat transfer coefficient (h eq ). The heat transfer coefficient increases, and it can be expected that heat transfer close to gas forced circulation heat transfer (about 5 W / m 2 K) is performed by radiant heating in the vicinity of 600 ° C.
原子炉容器液面近傍熱応力の熱伝導板による熱荷重緩和の効果を検証するため、熱伝導板無し、熱伝導板有りの場合について、原子炉起動時に生じる最大応力を有限要素法を用いた数値実験により求めて比較した。 In order to verify the effect of thermal load relaxation by the heat conduction plate of the thermal stress near the reactor vessel liquid surface, the finite element method was used to determine the maximum stress generated at the time of reactor start-up for the case without the heat conduction plate and with the heat conduction plate. They were obtained by numerical experiments and compared.
ここでは熱伝導板として黒鉛または12Ｃｒ鋼を使用した場合の解析を行った。ただし、黒鉛は結晶構造により熱伝導率が大きく異なるため、熱伝導板の効果が過少評価となる小さめの熱伝導率を用いることとした。ステンレス製の炉壁物性値一覧を表１に示す。 Here, analysis was performed when graphite or 12Cr steel was used as the heat conduction plate. However, since the thermal conductivity of graphite varies greatly depending on the crystal structure, it was decided to use a smaller thermal conductivity that underestimates the effect of the thermal conductive plate. A list of stainless steel furnace wall physical properties is shown in Table 1.
また、熱伝導板物性値（黒鉛）を表２に、熱伝導板物性値（12Ｃｒ鋼）を表３にそれぞれ示す。 In addition, Table 2 shows the heat conduction plate physical properties (graphite) and Table 3 shows the heat conduction plate physical properties (12Cr steel).
また、射出率も小さめの値を仮定し、ステレス製の炉壁と12Ｃｒ鋼で０．１、黒鉛で０．８を表４のように使用した。 Assuming that the injection rate was a small value, 0.1 was used for the stainless steel wall and 12Cr steel and 0.8 for the graphite as shown in Table 4.
荷重条件は、原子炉起動時にかかる熱荷重を模擬して、内部ナトリウム温度を２００℃から６００℃まで上昇させて解析を行った。昇温速度は２００℃から４００℃までを毎時１５℃、４００℃から６００℃までを毎時２０℃とした。また、ナトリウム温度の上昇に伴う液面の上昇を考慮した。ナトリウム液面の上昇は２００℃から４００℃の上昇区間で８８０ｍｍ、以後４００℃から６００℃の区間で３５０ｍｍそれぞれ上昇することとした。解析モデルの作成には有限要素解析用メッシュ生成プログラムＦＥＭＡＰ ｖ７．１を用い、解析ツールには汎用非線形構造システムＦＩＮＡＳ ｖ１４を用いた。 The load conditions were analyzed by increasing the internal sodium temperature from 200 ° C. to 600 ° C. by simulating the thermal load applied at the time of reactor startup. The rate of temperature increase was 200 ° C. to 400 ° C. at 15 ° C./hour and 400 ° C. to 600 ° C. at 20 ° C./hour. In addition, the rise in liquid level accompanying the rise in sodium temperature was taken into account. The rise of the sodium liquid level was 880 mm in the rising section from 200 ° C. to 400 ° C., and then 350 mm in the section from 400 ° C. to 600 ° C. A finite element analysis mesh generation program FEMAP v7.1 was used to create the analysis model, and a general-purpose nonlinear structural system FINAS v14 was used as the analysis tool.
図１を基に作成した解析メッシュを図４に、熱伝導板部分を拡大した解析メッシュを図５に示す。また、解析に用いた要素の一覧を表５に示す。 FIG. 4 shows an analysis mesh created based on FIG. 1, and FIG. 5 shows an analysis mesh obtained by enlarging the heat conduction plate portion. Table 5 shows a list of elements used for the analysis.
図６は解析の結果得られた熱伝導板の有無による発生応力の違いを示す図で、横軸は発生応力Ｓｎ（ＭＰａ）、縦軸は軸方向座標（ｍｍ）であり、熱伝導板による応力緩和方策の無いケース、黒鉛熱伝導板を施したケース、12Ｃｒ鋼熱伝導板を施したケースの３パターンの解析結果を示している。
解析結果は強度設計の指標として使用される炉壁外表面の応力強さ範囲（Ｓｎ）を算出し示した。計算結果より熱伝導板の射出率と熱伝導率の違いによりＳｎが変動することが分かる。熱伝導板として射出の良好な黒鉛を用いた解析から得られた結果は、熱伝導板を考慮しないケースに比べ、Ｓｎの最大値が約５９０ＭＰａから約４３０ＭＰａへと約２７％減少している。また、熱伝導板として12Ｃｒ鋼を用いた解析結果では、Ｓｎの最大値が約５００ＭＰａと約１５％減少することが確認された。このことから、熱伝導板を使用した簡素な設備により、炉壁に生じる熱応力を有意に緩和させられることが検証できた。
FIG. 6 is a diagram showing the difference in generated stress depending on the presence or absence of a heat conducting plate obtained as a result of analysis. The horizontal axis represents the generated stress Sn (MPa), the vertical axis represents the axial coordinate (mm), and the heat conducting plate The analysis results of three patterns of a case without a stress relaxation measure, a case with a graphite heat conduction plate, and a case with a 12Cr steel heat conduction plate are shown.
The analysis results calculated and indicated the stress intensity range (Sn) of the outer wall of the furnace wall used as an index for strength design. From the calculation results, it can be seen that Sn varies depending on the difference between the injection rate and the thermal conductivity of the heat conductive plate. The result obtained from the analysis using graphite with good injection as the heat conduction plate shows that the maximum value of Sn is reduced by about 27% from about 590 MPa to about 430 MPa as compared with the case not considering the heat conduction plate. In addition, in the analysis result using 12Cr steel as the heat conduction plate, it was confirmed that the maximum value of Sn was reduced by about 15% to about 500 MPa. From this, it was verified that the thermal stress generated in the furnace wall can be significantly relieved by simple equipment using a heat conduction plate.
次に新たに部材を付加することなく材料の変更のみで原子炉容器の熱荷重を緩和する例を説明する。
図７は原子炉容器液面近傍の応力低減を行う熱荷重緩和装置の実施の形態の他の例を示す図である。
通常、ガードベッセルは原子炉容器と同じ材料で作られるが、本実施形態ではガードベッセル２の材料を原子炉容器より熱伝導性が良好な材料に変更してガードベッセル２を熱伝導部材とし、ガードベッセルが炉壁と輻射によって熱的に結合して原子炉容器壁鉛直方向の伝熱を促進し、液面近傍の熱応力を緩和することを特徴としており、原子炉容器の構成は熱伝導板がない点を除いて図１の場合と同じである。
Next, an example will be described in which the thermal load of the reactor vessel is alleviated only by changing the material without adding a new member.
FIG. 7 is a view showing another example of the embodiment of the thermal load relaxation device for reducing the stress in the vicinity of the reactor vessel liquid surface.
Normally, the guard vessel is made of the same material as the reactor vessel. However, in this embodiment, the material of the guard vessel 2 is changed to a material having better thermal conductivity than the reactor vessel, and the guard vessel 2 is used as a heat conduction member. The guard vessel is thermally coupled to the reactor wall by radiation and promotes heat transfer in the vertical direction of the reactor vessel wall, reducing the thermal stress near the liquid level. It is the same as the case of FIG. 1 except that there is no plate.
図８は本実施形態の原子炉容器冷却材液面近傍の熱荷重緩和の原理説明図であり、図８（ａ）はガードベッセル無しの場合、図８（ｂ）は良熱伝導材ガードベッセル有りの場合の図である。
図２で説明したと同様に、起動時には内包する冷却材の温度が２００℃から５５０℃まで昇温するため、その過程で生じる鉛直方向の局所的温度勾配により炉壁に高い熱応力が発生し（図８（ａ））、高温の冷却材に接した接液部と低温のガス空間部の昇温終了時の急激な温度勾配（温度の折れ曲がり部）によって、昇温終了時に液面近傍炉壁外面（図８（ａ）のＳ点）に最大応力が生ずる。これを緩和するため、図８（ｂ）に示すように、液面下部の高温炉壁からの輻射熱で良熱伝導材ガードベッセルを加熱し、良熱伝導材ガードベッセルから液面上部の低温炉壁を輻射加熱することにより、応力の要因となっている鉛直方向の温度勾配を小さくする。その結果、最大応力発生位置Ｓにおける温度勾配は、最大応力発生時刻Ｔにおいて滑らかになり、熱荷重が緩和される。本実施形態では、新たに付加する部材は一切なく、建設コストに影響を与えず、非接触かつ静的構造物であるため確実に動作することが特徴である。
FIG. 8 is a diagram for explaining the principle of thermal load relaxation in the vicinity of the reactor vessel coolant level according to this embodiment. FIG. 8A shows a case without a guard vessel, and FIG. 8B shows a good heat conduction material guard vessel. It is a figure in the case of existence.
As described with reference to FIG. 2, since the temperature of the coolant contained is raised from 200 ° C. to 550 ° C. at the time of start-up, a high thermal stress is generated on the furnace wall due to a local temperature gradient in the vertical direction generated in the process. (FIG. 8 (a)), a furnace near the liquid surface at the end of the temperature rise due to a rapid temperature gradient (temperature bent portion) at the end of the temperature rise in the wetted part in contact with the high temperature coolant and the low temperature gas space Maximum stress is generated on the outer wall surface (point S in FIG. 8A). In order to alleviate this, as shown in FIG. 8 (b), the good thermal conductive material guard vessel is heated by radiant heat from the high temperature furnace wall at the lower part of the liquid level, and the low temperature furnace at the upper part of the liquid level from the good thermal conductive material guard vessel. By radiantly heating the walls, the vertical temperature gradient that causes stress is reduced. As a result, the temperature gradient at the maximum stress generation position S becomes smooth at the maximum stress generation time T, and the thermal load is relaxed. The present embodiment is characterized in that there is no member to be newly added, the construction cost is not affected, and it operates reliably because it is a non-contact and static structure.
図９は輻射による等価熱伝達係数と使用温度との関係を示す図（図３に対応）で、横軸は炉壁、ガードベッセル（高クロム系鋼、例えば12Ｃｒ鋼）の温度（Ｔ）、縦軸は等価熱伝達係数（ｈeq）である。
本実施形態では原子炉容器壁と12Ｃｒ鋼製ガードベッセルの射出率を小さめに仮定し、数１式にε1 ＝０．１、ε2 ＝０．１を代入し、等価熱伝達係数ｈeqとＴ（℃）＝Ｔ1 ＝Ｔ2 （均熱状態）の関係を計算したものであり、炉壁およびガードベッセル（高クロム系鋼、例えば12Ｃｒ鋼）の温度が高温になるにつれ急激に熱伝達係数が増加し、６００℃近傍ではガードベッセルを通しての輻射加熱により良好な熱伝達が行われることが予想できる。
FIG. 9 is a diagram showing the relationship between the equivalent heat transfer coefficient due to radiation and the operating temperature (corresponding to FIG. 3), and the horizontal axis is the temperature (T) of the furnace wall, guard vessel (high chromium steel, for example, 12Cr steel), The vertical axis represents the equivalent heat transfer coefficient (h eq ).
In the present embodiment, the injection rate of the reactor vessel wall and the 12Cr steel guard vessel is assumed to be small, and ε 1 = 0.1 and ε 2 = 0.1 are substituted into
特別な熱応力緩和措置を施さない簡素な316FR 鋼製容器とガードベッセルからなる原子炉構造を対象とし、ガードベッセルによる輻射加熱を考慮しない場合、原子炉容器と同じ316FR 鋼製ガードベッセルが存在した場合、良熱伝導材である12Ｃｒ鋼製ガードベッセルが存在した場合について、原子炉起動時に６００℃まで昇温した場合の生じる最大応力の違いを数値実験により求めて比較した。 For a reactor structure consisting of a simple 316FR steel vessel and guard vessel without special thermal stress relaxation measures, the same 316FR steel guard vessel as the reactor vessel existed when radiation heating by the guard vessel was not considered. In this case, the difference in the maximum stress generated when the temperature was raised to 600 ° C. at the time of starting the reactor was compared and obtained by a numerical experiment in the case where a 12Cr steel guard vessel, which is a good heat conductive material, was present.
解析に使用した炉壁物性値は表１に示したもの、ガードベッセルの良熱伝導材の物性値は表２（12Ｃｒ鋼）に示したものであり、不確定性のある射出率については熱応力緩和の効果を小さめに見積もるため、ε1 ＝０．１、ε2 ＝０．１を使用した。荷重条件、解析モデルの作成、解析ツールは熱伝導板の場合と同じである。 The properties of the furnace wall used in the analysis are shown in Table 1. The properties of the good heat conduction material of the guard vessel are shown in Table 2 (12Cr steel). In order to estimate the effect of stress relaxation slightly, ε 1 = 0.1 and ε 2 = 0.1 were used. The load conditions, analysis model creation, and analysis tools are the same as for the heat conduction plate.
図７を基に作成した解析メッシュを図１０に、液面近傍部分を拡大した解析メッシュを図１１に示す。 An analysis mesh created based on FIG. 7 is shown in FIG. 10, and an analysis mesh obtained by enlarging the vicinity of the liquid surface is shown in FIG.
図１２は解析の結果得られた発生応力の違いを示す図（図６に対応）で、横軸は発生応力Ｓｎ（ＭＰａ）、縦軸は軸方向座標（ｍｍ）であり、ガードベッセルによる輻射加熱を考慮しないケース、ガードベッセルの材質を12Ｃｒ鋼としたケース、316FR 鋼としたケースの３パターンの解析結果を示している。
解析結果はそれぞれ設計における強度評価の指標となる応力強さ範囲（Ｓｎ）の炉壁外表面に沿った鉛直方向分布に従って示した。ガードベッセルとして12Ｃｒ鋼を用いた場合は、ガードベッセルによる輻射加熱を考慮しないケースに比べ、Ｓｎの最大値が約５９０ＭＰａから約５１９ＭＰａへと約１２％減少している。また、ガードベッセルの材質に316FR 鋼を用いた解析結果では、Ｓｎの最大値が約５６２ＭＰａと約４．７％減少することが分かった。このことから、ガードベッセルの材料を通常材から良熱伝導材に変更することによって、顕著な熱荷重緩和を達成できることが確認された。
FIG. 12 is a diagram showing the difference in generated stress obtained as a result of the analysis (corresponding to FIG. 6). The horizontal axis is the generated stress Sn (MPa), the vertical axis is the axial coordinate (mm), and the radiation by the guard vessel is shown. Three patterns of analysis results are shown: a case not considering heating, a case where the guard vessel material is 12Cr steel, and a case where the material is 316FR steel.
The analysis results are shown in accordance with the vertical distribution along the outer surface of the furnace wall in the stress intensity range (Sn), which is an index for strength evaluation in the design. When 12Cr steel is used as the guard vessel, the maximum value of Sn is reduced by about 12% from about 590 MPa to about 519 MPa, compared to the case where radiation heating by the guard vessel is not considered. The analysis results using 316FR steel as the material for the guard vessel showed that the maximum value of Sn was reduced by about 4.7% to about 562 MPa. From this, it was confirmed that significant thermal load relaxation can be achieved by changing the material of the guard vessel from a normal material to a good heat conductive material.
本発明によれば、新たに付加する部材なしに冷却材液面近傍の熱荷重を緩和することができ、建設コストに影響を与えず、非接触かつ静的構造物であるため確実に動作させることができるので、産業上の利用可能性は極めて大きい。 According to the present invention, it is possible to relieve the thermal load in the vicinity of the coolant liquid level without a member to be newly added, without affecting the construction cost, and since it is a non-contact and static structure, it can be operated reliably. Therefore, the industrial applicability is extremely large.
１…原子炉容器、２…ガードベッセル、３…アニュラス空間、８、１０…断熱材、９…冷却材液面、２０…熱伝導板。
DESCRIPTION OF
Claims (2)
前記ガードベッセルを原子炉容器より良熱伝導性の材料で構成し、液面下部の炉壁からの輻射熱で前記ガードベッセルを加熱し、液面上部の炉壁を前記ガードベッセルからの輻射熱で加熱することを特徴とする原子炉容器の熱荷重緩和装置。 In the apparatus for relaxing the thermal load of the reactor vessel in the vicinity of the liquid level of the coolant in the reactor vessel in which a guard vessel is installed with a space on the outside,
The guard vessel is made of a material having better thermal conductivity than the reactor vessel, the guard vessel is heated by radiant heat from the furnace wall at the lower part of the liquid level, and the furnace wall at the upper part of the liquid level is heated by the radiant heat from the guard vessel. A thermal load relaxation device for a nuclear reactor vessel, characterized in that:
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