JP3492144B2 - Operating method of steam generator for pressurized water reactor - Google Patents
Operating method of steam generator for pressurized water reactorInfo
- Publication number
- JP3492144B2 JP3492144B2 JP10195897A JP10195897A JP3492144B2 JP 3492144 B2 JP3492144 B2 JP 3492144B2 JP 10195897 A JP10195897 A JP 10195897A JP 10195897 A JP10195897 A JP 10195897A JP 3492144 B2 JP3492144 B2 JP 3492144B2
- Authority
- JP
- Japan
- Prior art keywords
- steam generator
- molar ratio
- heat transfer
- water
- tube
- 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 - Lifetime
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
Landscapes
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Testing Resistance To Weather, Investigating Materials By Mechanical Methods (AREA)
Description
【発明の詳細な説明】
【０００１】
【発明の属する技術分野】本発明は、加圧水型原子炉用
蒸気発生器のように胴側液体を加熱蒸発せしめるシェル
・アンド・チューブ熱交換器型蒸気発生器の運転方法に
関する。
【０００２】
【従来の技術】加圧水型原子炉用蒸気発生器は、一種の
シェル・アンド・チューブ形熱交換器であり、細く且つ
薄肉の伝熱管内を高温の原子炉冷却材が貫流し、胴側流
体である給水を加熱し、蒸気を発生せしめる。このよう
に、伝熱管内を放射能を帯びる可能性のある原子炉冷却
材が流れるので、伝熱管の損傷は放射性物質の放出に繋
がるから、その損傷防止には一方ならぬ注意が払われて
いる。これを図８に示す代表的な加圧水型原子炉用蒸気
発生器について説明する。図において、蒸気発生器１の
胴３の下部に管板５が一体的に接合されていて、その下
部に原子炉冷却材の入口水室７及び出口水室９が画成さ
れている。１本のみ示された複数の逆Ｕ字形伝熱管１１
の両端は、管板５の穴内に挿着され、更に鉛直方向に間
隔を置いた複数の管支持板１３により伝熱管１１は横方
向に支持されている。そして原子炉から供給された高温
の冷却材は、入口水室７を経て伝熱管１１内に流入して
貫流し、その際後述するように熱交換により熱を失って
低温になり出口水室９に至る。そして、そこから原子炉
に戻る。一方、給水リング１５から蒸気発生器１内に流
入した給水は、包囲管１７と胴３との間を下向きに流
れ、次いで管板５の上を流れ、しかる後伝熱管１１に沿
って上向きに流れる。この際、前述の原子炉冷却材と熱
交換をし、一部は蒸気となる。その加熱される給水が上
向きに流れるに際し、管支持板１３を貫通し、そして汽
水分離ベーン１９を通って分離された蒸気が流出する。
【０００３】伝熱管１１の数は蒸気発生器１の規模によ
って異なるが、約４０００本内外であり、図９に示すよ
うに管支持板１３の支持穴に挿通されて横方向に支持さ
れるが、その構造上支持穴の内面と伝熱管１１の外面と
の間に間隔が０．２mm以下の狭い隙間（以下クレビスと
称する。）１４が発生することは避けられない。蒸気発
生器１内の給水は、そこで蒸発されるので含有不純物は
一般に濃縮される傾向があるから、蒸気発生器１内に供
給される給水の水質は厳しく管理されているけれども、
前記クレビス１４内では、その構造上蒸気発生器１内の
給水中のイオン性不純物が１０3〜１０6倍程度に濃縮
し、蒸気発生器１の内部給水の不純物濃度が高い場合に
は強アルカリ性或いは強酸性雰囲気となる可能性があ
る。そしてその傾向が著しい場合には、伝熱管１１に割
れ、変形或いは減肉といった腐食損傷が発生し易い。典
型的な腐食損傷の例が図１０に示されている。図１０
（ａ）は減肉Ａ，同（ｂ）は割れＢ、同（ｃ）は変形Ｃ
をそれぞれ概念的に示している。このような腐食損傷の
うち、伝熱管の割れ（Intergranular Attack ）損傷の
抑制を目的として、その主な発生原因の１つである遊離
アルカリの蒸気発生器１内への持ち込みを零にするとい
う考えに基づき、ナトリウムＮａに対して過剰の塩素Ｃ
ｌを存在させることにより遊離アルカリを管理するＮａ
/Ｃｌモル比管理をベースにした運転を行っている。加
圧水型原子炉の給水・蒸気循環系即ち２次系についての
前記モル比管理の考え方が図１１に示されている。そし
てそのＮａ/Ｃｌモル比管理は、復水脱塩装置出口及び
蒸気発生器内給水について規定されていて、管理値とし
ては分析精度等の裕度を見込み、０．７以下としてい
る。
【０００４】
【発明が解決しようとする課題】上述したように、加圧
水型原子炉の蒸気発生器の給水の運転時水質管理は、Na
/Clモル比を指標として行ってきた。又、詳述していな
いが Na/(Cl+SO4）モル比を指標とする管理も行ってき
た。しかしながら、最近の研究によれば、次のような問
題があることが判明し、給水の水質の現状にそぐわなく
なっている。
（１）曾ては、Na及びClが蒸気発生器１内の給水中の不
純物の大半を占め、伝熱管と支持板の間のクレビスの環
境は、これらのバランスで決まっていた。しかしなが
ら、給水の水処理法の高度化が進み、含有されるNa及び
Clの量が大幅に減少し、それ以外の不純物（Ca，Mg，Ｋ
或いはSO4）と同等の濃度レベルになってきた。従っ
て、これらの不純物がクレビスの環境に及ぼす影響が大
きくなり、Na／Clモル比がその環境の状態を必ずしも正
確に表さなくなってきた。従って、本発明は、加圧水型
原子炉用蒸気発生器の伝熱管と支持板との間のクレビス
の環境を表す新たな指標を使用して、クレビス腐食の発
生を抑制した状態で加圧水型原子炉用蒸気発生器を運転
する方法を提供することを課題とする。
【０００５】
【課題を解決するための手段】如上の課題を解決するた
め、本発明によれば、加圧水型原子炉用蒸気発生器の運
転に際し、蒸気発生器内の伝熱管と管支持板との間のク
レビス内の雰囲気と高い相関を有する蒸発器内給水の全
カチオン／ＳＯ ４ モル比を検出し、全カチオン／ＳＯ4
モル比を測定し、全カチオン／ＳＯ４モル比を伝熱管と
管支持板との間で腐食割れが生じない範囲に保持するこ
とを特徴とする。
【０００６】
【発明の実施の形態】本発明の有効性を確認するため図
１に示すような試験装置を用いて蒸気発生器内のバルク
水のモル比と模擬クレビス部の雰囲気（ｐＨ）との関係
を測定した。先ず濃縮試験装置２０を説明するとオート
クレーブ２１は、試験水の入口２３と出口２５を備えて
いて、上部から伝熱管模擬体２７を内部に受けれてい
る。伝熱管模擬体２７の中には、原子炉冷却材に相当す
る加熱ヒータ２９が設けられ、伝熱管模擬体２７の下部
は管支持板模擬体３１により囲まれていて、両者の間に
クレビス模擬部３３を形成している。そして、そのクレ
ビス模擬部３３からサンプル液採取管３５がオートクレ
ーブ２１の外部まで延び、適時サンプル液を採取してそ
の組成を分析、測定できるようになっている。そして、
蒸気発生器１の給水を模擬する実機模擬水を入口２３か
らオートクレーブ２１の内部に供給し、その内部で実機
にバルク水に対応する器内水３７を形成し、そして余剰
の器内水３７は出口から流出する。このように、器内水
３７の所定の流れがある中で、加熱ヒータ２９により原
子炉冷却材の授熱量に相当する熱量を発生させてクレビ
ス模擬部３３内の器内水を加熱し、そのサンプル液を適
宜抽出してその組成を測定する。その測定組成から高温
気液平衡計算を用いてそのｐＨを算出する。
【０００７】以上の濃縮試験装置２０を用いて試験計測
を行い、バルク水の全カチオン（ΣＣ）／SO4モル比と
クレビス模擬部３３のｐＨとの関係（図２)、バルク水
の Na/Clモル比とクレビス模擬部３３のｐＨとの関係
（図３）、バルク水のNa/(Cl+SO4)モル比とクレビス模
擬部３３のｐＨとの関係（図４）、バルク水の全カチオ
ン（ΣＣ）/全アニオン（ΣA）モル比とクレビス模擬部
３３のｐＨ殿関係（図５）及びバルク水の(Na＋K）/(Cl
+SO4)モル比とクレビス模擬部３３のｐＨとの関係（図
６）を求めた。その結果を前述の括弧内に示した図面に
記載している。この結果を総括すると次のようになる。
図３から分かるようにバルク水のNa/Clモル比とクレビ
ス模擬部３３のｐＨとは相関を示していない。即ち従来
型の水質管理手法では、適切な運転ができない。図４か
ら分かるようにバルク水のNa/(Cl+SO4)モル比とクレビ
ス模擬部３３のｐＨとの間にも相関が見られない。又図
５と図６から分かるようにバルク水の全カチオン（Σ
Ｃ）/全アニオン（ΣA）モル比とクレビス模擬部３３の
ｐＨとの間及びバルク水の(Na＋K）/(Cl+SO4)モル比と
クレビス模擬部３３のｐＨとの間には比較的良い相関が
認められるもののモル比の小さい領域でばらついてい
る。これに対し、図２から分かるようにバルク水の全カ
チオン（ΣＣ）／SO4モル比とクレビス模擬部３３のｐ
Ｈの間には低ｐＨから高ｐＨまでの広い領域について良
い相関が認められる。以上のような試験結果から、運転
中に蒸気発生器のバルク水の水質からクレビス部の環境
を推定できることが分かる。
【０００８】以上のような相関から、蒸気発生器内のバ
ルク水の全カチオン（ΣＣ）／SO4モル比を測定するこ
とにより、通常アクセスできない蒸気発生器内の伝熱管
外面と支持板との間の狭い隙間内の環境乃至ｐＨを推定
でき、このｐＨが中性領域であるｐＨ５乃至８の間にあ
るようにして蒸気発生器を運転すれば、その腐食割れを
防止できる。尚、図７はオートクレーブ２１内バルク水
とクレビス模擬部３３内の濃縮模擬液との相関を示し、
実線はバルク水／濃縮模擬液の組成モル比１：１の線で
ある。
【０００９】
【発明の効果】以上説明したように、加圧水型原子炉用
の蒸気発生器の運転に際し、蒸気発生器内の伝熱管と管
支持板との間のクレビス内の雰囲気と高い相関を有する
蒸発器内給水の全カチオン／ＳＯ ４ モル比を検出し、全
カチオン／ＳＯ4モル比を測定し、全カチオン／ＳＯ４
モル比を伝熱管と管支持板との間で腐食割れが生じない
範囲に保持するように運転するので、伝熱管の損傷を大
きく抑制することができる。Description: BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a shell-and-tube heat exchanger type steam generator for heating and evaporating a body side liquid like a steam generator for a pressurized water reactor. Related to the operation method of the vessel. 2. Description of the Related Art A steam generator for a pressurized water reactor is a kind of shell-and-tube heat exchanger in which a high-temperature reactor coolant flows through a thin and thin-walled heat transfer tube. Heats the feedwater, which is the body-side fluid, to generate steam. In this way, since the reactor coolant that may take on radioactivity flows through the heat transfer tubes, damage to the heat transfer tubes leads to the release of radioactive material, so extreme care has been taken to prevent the damage. I have. This will be described with respect to a typical steam generator for a pressurized water reactor shown in FIG. In the figure, a tube sheet 5 is integrally joined to a lower portion of a body 3 of a steam generator 1, and an inlet water chamber 7 and an outlet water chamber 9 for a reactor coolant are defined below the tube sheet 5. A plurality of inverted U-shaped heat transfer tubes 11 shown only in one
Are inserted into the holes of the tube sheet 5, and the heat transfer tubes 11 are supported in the lateral direction by a plurality of tube support plates 13 spaced apart in the vertical direction. The high-temperature coolant supplied from the nuclear reactor flows into the heat transfer tube 11 through the inlet water chamber 7 and flows therethrough. At that time, heat is lost due to heat exchange and becomes low temperature, and the outlet water chamber 9 Leads to. And then return to the reactor. On the other hand, the feedwater flowing into the steam generator 1 from the feedwater ring 15 flows downward between the surrounding pipe 17 and the body 3, then flows on the tube sheet 5, and then upwards along the heat transfer pipe 11. Flows. At this time, heat exchange is performed with the above-mentioned reactor coolant, and a part thereof becomes steam. As the heated feedwater flows upwardly, the separated steam flows through the tube support plate 13 and through the steam separation vane 19. The number of the heat transfer tubes 11 varies depending on the scale of the steam generator 1, but is about 4000 inside and outside, and as shown in FIG. 9, the heat transfer tubes 11 are inserted in the support holes of the tube support plate 13 and supported laterally. Due to its structure, it is inevitable that a narrow gap (hereinafter referred to as clevis) 14 having a distance of 0.2 mm or less between the inner surface of the support hole and the outer surface of the heat transfer tube 11 is generated. Since the feedwater in the steam generator 1 is evaporated there, the contained impurities generally tend to be concentrated. Therefore, although the quality of the feedwater supplied to the steam generator 1 is strictly controlled,
In the clevis 14, the ionic impurities in the feed water in the steam generator 1 are concentrated to about 10 3 to 10 6 times due to its structure, and when the impurity concentration of the feed water in the steam generator 1 is high, the alkali is strongly alkaline. Alternatively, there is a possibility that the atmosphere becomes a strongly acidic atmosphere. If the tendency is remarkable, corrosion damage such as cracking, deformation or thinning of the heat transfer tube 11 is likely to occur. An example of a typical corrosion damage is shown in FIG. FIG.
(A) is thickness reduction A, (b) is crack B, and (c) is deformation C
Are conceptually shown. Among such corrosion damages, in order to suppress the damage of the heat transfer tube (Intergranular Attack), the idea of bringing free alkali, which is one of the main causes, into the steam generator 1 to be zero is considered. Excess chlorine C based on sodium
1 to control free alkali by the presence of
The operation is based on the / Cl molar ratio control. FIG. 11 shows the concept of the molar ratio management for the feed water / steam circulation system, that is, the secondary system of the pressurized water reactor. The Na / Cl molar ratio control is defined for the condensate demineralizer outlet and the water supply in the steam generator, and the control value is set to 0.7 or less in view of the margin of analysis accuracy and the like. [0004] As described above, water quality management during operation of feed water of a steam generator of a pressurized water reactor is controlled by Na
/ Cl molar ratio was used as an index. Although not described in detail, management has also been performed using the Na / (Cl + SO 4 ) molar ratio as an index. However, according to recent research, the following problems have been found, and the quality of the water supply has become inconsistent with the current situation. (1) Formerly, Na and Cl accounted for most of the impurities in the feedwater in the steam generator 1, and the environment of the clevis between the heat transfer tube and the support plate was determined by these balances. However, the sophistication of the water treatment method for feedwater has progressed, and the contained Na and
The amount of Cl is greatly reduced, and other impurities (Ca, Mg, K
Alternatively, the concentration level has become equivalent to that of SO 4 ). Accordingly, the influence of these impurities on the environment of the clevis has increased, and the Na / Cl molar ratio has not always accurately represented the state of the environment. Accordingly, the present invention provides a pressurized water reactor in a state in which the occurrence of clevis corrosion is suppressed, using a new index indicating the environment of the clevis between the heat transfer tube and the support plate of the steam generator for the pressurized water reactor. An object of the present invention is to provide a method for operating a steam generator. [0005] In order to solve the above problems, according to the present invention, a heat transfer tube and a tube support plate in a steam generator for operation of a steam generator for a pressurized water reactor are operated. In between
Total water supply in the evaporator that has a high correlation with the atmosphere in the levis
The cation / SO 4 molar ratio was detected and the total cation / SO 4
The molar ratio was measured, and the total cation / SO 4 molar ratio was determined as the heat transfer tube.
It is characterized in that it is maintained within a range in which corrosion cracking does not occur between the tube support plate and the tube support plate . DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS In order to confirm the effectiveness of the present invention, the molar ratio of bulk water in a steam generator, the atmosphere (pH) of a simulated clevis part, and the like were measured using a test apparatus as shown in FIG. Was measured. First, the concentration test apparatus 20 will be described. The autoclave 21 has an inlet 23 and an outlet 25 for test water, and receives a heat transfer tube simulation body 27 from above. A heater 29 corresponding to a reactor coolant is provided in the heat transfer tube simulation body 27, and a lower portion of the heat transfer tube simulation body 27 is surrounded by a tube support plate simulation body 31, and a clevis simulation is provided therebetween. The part 33 is formed. Then, a sample liquid collection tube 35 extends from the clevis simulator 33 to the outside of the autoclave 21 so that the sample liquid can be collected at appropriate times to analyze and measure its composition. And
Simulated water for simulating the water supply of the steam generator 1 is supplied from the inlet 23 to the inside of the autoclave 21, inside of which the inside water 37 corresponding to the bulk water is formed in the actual machine, and the surplus inside water 37 is removed. Spill from exit. As described above, in the presence of the predetermined flow of the water 37 in the chamber, the heater 29 generates heat corresponding to the heat transfer amount of the reactor coolant to heat the water in the clevis simulator 33, The sample liquid is appropriately extracted and its composition is measured. The pH is calculated from the measured composition using a high-temperature vapor-liquid equilibrium calculation. [0007] Test measurement was performed using the above concentration test apparatus 20, and the relationship between the total cation (ΣC) / SO 4 molar ratio of bulk water and the pH of the clevis simulator 33 (FIG. 2), Relationship between Cl molar ratio and pH of clevis simulator 33 (FIG. 3), relationship between Na / (Cl + SO 4 ) molar ratio of bulk water and pH of clevis simulator 33 (FIG. 4), total of bulk water Relationship between the cation (ΔC) / total anion (ΔA) molar ratio and the pH of the clevis simulator 33 (FIG. 5) and the (Na + K) / (Cl
+ SO 4 ) The relationship between the molar ratio and the pH of the clevis simulator 33 (FIG. 6) was determined. The results are described in the figures shown in parentheses above. The results are summarized as follows.
As can be seen from FIG. 3, there is no correlation between the Na / Cl molar ratio of the bulk water and the pH of the clevis simulator 33. That is, the conventional water quality management method cannot perform appropriate operation. As can be seen from FIG. 4, there is no correlation between the Na / (Cl + SO 4 ) molar ratio of the bulk water and the pH of the clevis simulator 33. As can be seen from FIGS. 5 and 6, all cations (バ ル ク
C) / the total anion (ΔA) molar ratio and the pH of the clevis mimic unit 33 and the (Na + K) / (Cl + SO 4 ) molar ratio of bulk water and the pH of the clevis mimic unit 33 are relatively high. Although a good correlation is observed, it varies in a region where the molar ratio is small. On the other hand, as can be seen from FIG. 2, the total cation (ΔC) / SO 4 molar ratio of the bulk water and the p
A good correlation is observed between H in a wide range from low pH to high pH. From the above test results, it can be seen that the environment of the clevis can be estimated from the quality of the bulk water of the steam generator during operation. From the above correlation, by measuring the total cation (ΔC) / SO 4 molar ratio of the bulk water in the steam generator, the outer surface of the heat transfer tube in the steam generator, which cannot be normally accessed, and the support plate are measured. The environment or pH in the narrow gap therebetween can be estimated, and if the steam generator is operated such that this pH is in the neutral range of pH 5 to 8, corrosion cracking of the steam generator can be prevented. FIG. 7 shows the correlation between the bulk water in the autoclave 21 and the concentration simulation liquid in the clevis simulation unit 33,
The solid line is a line having a composition molar ratio of bulk water / concentration simulation liquid of 1: 1. As described above, when the steam generator for the pressurized water reactor is operated, the heat transfer tube and the tube in the steam generator are used.
High correlation with the atmosphere in the clevis between the support plate
Detecting a total cation / SO 4 molar ratio of the evaporator in the water supply, to measure the total cation / SO 4 molar ratio, the total cation / SO 4
Corrosion cracking does not occur between heat transfer tube and tube support plate
Since the operation is performed so as to maintain the temperature in the range, damage to the heat transfer tube can be largely suppressed.
【図面の簡単な説明】
【図１】本発明の実施形態の効果を実証するための試験
装置を示す立断面図である。
【図２】前記試験装置による実験結果を示すグラフであ
る。
【図３】前記試験装置による実験結果を示すグラフであ
る。
【図４】前記試験装置による実験結果を示すグラフであ
る。
【図５】前記試験装置による実験結果を示すグラフであ
る。
【図６】前記試験装置による実験結果を示すグラフであ
る。
【図７】前記試験装置による実験結果を示すグラフであ
る。
【図８】本発明の方法が適用される蒸気発生器の一例を
示す立断面図である。
【図９】図８のIX部を拡大して示す部分断面図である。
【図１０】蒸気発生器における不具合現象を説明する概
念図である。
【図１１】従来の運転方法における水質管理の考え方を
示す概念図である。
【符号の説明】
１ 蒸気発生器
３ 胴
５ 管板
７ 入口水室
９ 出口水室
１１ 伝熱管
１３ 管支持板
１５ 給水リング
１７ 包囲管
１９ 汽水分離ベーン
２０ 濃縮試験装置
２１ オートクレーブ
２３ 入口
２５ 出口
２７ 伝熱管模擬体
２９ 加熱ヒータ
３１ 管支持板模擬体
３３ クレビス模擬部
３５ サンプル液採取管
３７ 器内水BRIEF DESCRIPTION OF THE DRAWINGS FIG. 1 is an elevational sectional view showing a test apparatus for demonstrating the effects of the embodiment of the present invention. FIG. 2 is a graph showing experimental results obtained by the test apparatus. FIG. 3 is a graph showing an experimental result by the test device. FIG. 4 is a graph showing an experimental result by the test apparatus. FIG. 5 is a graph showing experimental results obtained by the test apparatus. FIG. 6 is a graph showing an experimental result by the test device. FIG. 7 is a graph showing an experimental result by the test device. FIG. 8 is a sectional elevation view showing an example of a steam generator to which the method of the present invention is applied. FIG. 9 is an enlarged partial cross-sectional view showing a portion IX of FIG. 8; FIG. 10 is a conceptual diagram illustrating a malfunction phenomenon in the steam generator. FIG. 11 is a conceptual diagram showing a concept of water quality management in a conventional operation method. [Description of Signs] 1 Steam generator 3 Body 5 Tube plate 7 Inlet water chamber 9 Outlet water chamber 11 Heat transfer tube 13 Tube support plate 15 Water supply ring 17 Surrounding tube 19 Steam separation vane 20 Concentration test device 21 Autoclave 23 Inlet 25 Outlet 27 Simulated heat transfer tube 29 Heated heater 31 Simulated tube support plate 33 Simulated clevis 35 Sample liquid sampling tube 37 Internal water
───────────────────────────────────────────────────── フロントページの続き (72)発明者 門上 英 兵庫県神戸市兵庫区和田崎町一丁目１番 １号 三菱重工業株式会社 神戸造船所 内 (56)参考文献 特開 平４−24456（ＪＰ，Ａ) 特開 平５−346492（ＪＰ，Ａ) Ｓ． Ｗ． Ｌｕｒｉｅ， Ｊ． Ｗ． Ｋｌｉｓｉｅｗｉｃｚ，Ｎｕｃｌ ｅａｒ Ｓｔｅａｍ Ｇｅｎｅｒａｔｏ ｒ Ｔｕｂｅ Ｃｏｒｒｏｓｉｏｎ Ａ ｓｓｏｃｉａｔｅｄ ｗｉｔｈ Ｃｏｎ ｄｅｎｓａｔｅ Ｐｏｌｉｓｈｉｎｇ, Ｐａｐ Ａｍ Ｓｏｃ Ｍｅｃｈ Ｅｎ ｇ，米国，1982年，82−ＷＡ−ＨＴ− 79，６ｐ，ＪＳＴ Ｎｏ． Ａ0478ＢＡ Ｗ Ｍ． Ｈ． Ｌｉｅｔｚｋｅ， Ｗ. Ｌ． Ｍａｒｓｈａｌｌ，Ｓｏｄｉｕ ｍ−Ｓｕｌｆａｔｅ Ｓｏｌｕｂｉｌｉ ｔｉｅｓ ｉｎ Ｈｉｇｈ−Ｔｅｍｐｅ ｒａｔｕｒｅ（250−374℃） Ｓａｌｔ ａｎｄ Ａｃｉｄ Ｓｏｌｕｔｉｏｎ ｓ，ＵＳ ＤＯＥ Ｒｅｐ，米国，1983 年 ７月，ＥＰＲＩ−ＮＰ−3047，72 ｐ，ＪＳＴ Ｎｏ．Ｐ0998Ａ (58)調査した分野(Int.Cl.7，ＤＢ名) G21D 1/00 GDP G21D 3/08 GDP G01N 33/18 ──────────────────────────────────────────────────続 き Continuation of the front page (72) Inventor Eiji Monjo 1-1-1, Wadazaki-cho, Hyogo-ku, Kobe-shi, Hyogo Inside Mitsubishi Heavy Industries, Ltd. Kobe Shipyard (56) References JP-A-4-24456 JP, A) JP-A-5-346492 (JP, A) W. Lurie, J.M. W. Klisiewicz, Nucleear Steam Generator Tube Corrosion Associated with Condensate Polishing, Pap Am Soc Mech Eng, USA, 1982, 82-WA-H. A0478BA W M.A. H. Lietzke, WL. Marshall, Sodium-Sulfate Solutions in High-Temperature (250-374 ° C.) Salt and Acid Solutions, US DOE Rep, USA, July, 1983, EPJ-72, EPRI-N. P0998A (58) Field surveyed (Int. Cl. 7 , DB name) G21D 1/00 GDP G21D 3/08 GDP G01N 33/18
Claims (1)
際し、蒸気発生器内の伝熱管と管支持板との間のクレビ
ス内の雰囲気と高い相関を有する蒸発器内給水の全カチ
オン／ＳＯ ４ モル比を検出し、全カチオン／ＳＯ4モル
比を測定し、全カチオン／ＳＯ４モル比を伝熱管と管支
持板との間で腐食割れが生じない範囲に保持することを
特徴とする加圧水型原子炉用蒸気発生器の運転方法。(57) [Claims] [Claim 1] For operation of a steam generator for a pressurized water reactor
At the time, the cracks between the heat transfer tubes in the steam generator and the tube support plate
Of water in the evaporator, which has a high correlation with the atmosphere in the evaporator
Detecting an on / SO 4 molar ratio, and measures the total cation / SO 4 molar ratio, the total cation / SO 4 molar ratio and the heat transfer tube Kan支
A method for operating a steam generator for a pressurized water reactor, wherein the steam generator is maintained within a range where corrosion cracking does not occur between the steam generator and a holding plate .
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP10195897A JP3492144B2 (en) | 1997-04-18 | 1997-04-18 | Operating method of steam generator for pressurized water reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP10195897A JP3492144B2 (en) | 1997-04-18 | 1997-04-18 | Operating method of steam generator for pressurized water reactor |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPH10293194A JPH10293194A (en) | 1998-11-04 |
| JP3492144B2 true JP3492144B2 (en) | 2004-02-03 |
Family
ID=14314387
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| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP10195897A Expired - Lifetime JP3492144B2 (en) | 1997-04-18 | 1997-04-18 | Operating method of steam generator for pressurized water reactor |
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| Country | Link |
|---|---|
| JP (1) | JP3492144B2 (en) |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2011059091A1 (en) | 2009-11-16 | 2011-05-19 | 株式会社東芝 | Corrosion resistant structure and corrosion protection method in high-temperature water system |
| WO2012014894A1 (en) | 2010-07-27 | 2012-02-02 | 株式会社東芝 | Method for suppressing corrosion of plant, and plant |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US8877335B2 (en) | 2009-03-30 | 2014-11-04 | Kabushiki Kaisha Toshiba | Corrosion-resistant member and method of manufacturing the same |
-
1997
- 1997-04-18 JP JP10195897A patent/JP3492144B2/en not_active Expired - Lifetime
Non-Patent Citations (2)
| Title |
|---|
| Ｍ． Ｈ． Ｌｉｅｔｚｋｅ， Ｗ． Ｌ． Ｍａｒｓｈａｌｌ，Ｓｏｄｉｕｍ−Ｓｕｌｆａｔｅ Ｓｏｌｕｂｉｌｉｔｉｅｓ ｉｎ Ｈｉｇｈ−Ｔｅｍｐｅｒａｔｕｒｅ（２５０−３７４℃） Ｓａｌｔ ａｎｄ Ａｃｉｄ Ｓｏｌｕｔｉｏｎｓ，ＵＳ ＤＯＥ Ｒｅｐ，米国，１９８３年 ７月，ＥＰＲＩ−ＮＰ−３０４７，７２ｐ，ＪＳＴ Ｎｏ．Ｐ０９９８Ａ |
| Ｓ． Ｗ． Ｌｕｒｉｅ， Ｊ． Ｗ． Ｋｌｉｓｉｅｗｉｃｚ，Ｎｕｃｌｅａｒ Ｓｔｅａｍ Ｇｅｎｅｒａｔｏｒ Ｔｕｂｅ Ｃｏｒｒｏｓｉｏｎ Ａｓｓｏｃｉａｔｅｄ ｗｉｔｈ Ｃｏｎｄｅｎｓａｔｅ Ｐｏｌｉｓｈｉｎｇ，Ｐａｐ Ａｍ Ｓｏｃ Ｍｅｃｈ Ｅｎｇ，米国，１９８２年，８２−ＷＡ−ＨＴ−７９，６ｐ，ＪＳＴ Ｎｏ． Ａ０４７８ＢＡＷ |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2011059091A1 (en) | 2009-11-16 | 2011-05-19 | 株式会社東芝 | Corrosion resistant structure and corrosion protection method in high-temperature water system |
| WO2012014894A1 (en) | 2010-07-27 | 2012-02-02 | 株式会社東芝 | Method for suppressing corrosion of plant, and plant |
Also Published As
| Publication number | Publication date |
|---|---|
| JPH10293194A (en) | 1998-11-04 |
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