JPH02222893A - Nuclear power plant - Google Patents
Nuclear power plantInfo
- Publication number
- JPH02222893A JPH02222893A JP1041923A JP4192389A JPH02222893A JP H02222893 A JPH02222893 A JP H02222893A JP 1041923 A JP1041923 A JP 1041923A JP 4192389 A JP4192389 A JP 4192389A JP H02222893 A JPH02222893 A JP H02222893A
- Authority
- JP
- Japan
- Prior art keywords
- water
- reactor
- piping
- hydrogen
- pressure vessel
- 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.)
- Pending
Links
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims abstract description 53
- 239000001257 hydrogen Substances 0.000 claims abstract description 29
- 229910052739 hydrogen Inorganic materials 0.000 claims abstract description 29
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 claims abstract description 27
- 238000002347 injection Methods 0.000 claims abstract description 15
- 239000007924 injection Substances 0.000 claims abstract description 15
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 abstract description 29
- 239000001301 oxygen Substances 0.000 abstract description 29
- 229910052760 oxygen Inorganic materials 0.000 abstract description 29
- 239000000498 cooling water Substances 0.000 abstract description 16
- 238000007599 discharging Methods 0.000 abstract 1
- 239000000463 material Substances 0.000 description 13
- 238000005336 cracking Methods 0.000 description 12
- 238000005260 corrosion Methods 0.000 description 11
- 230000007797 corrosion Effects 0.000 description 11
- 239000002826 coolant Substances 0.000 description 4
- 238000000746 purification Methods 0.000 description 4
- 238000009835 boiling Methods 0.000 description 3
- 230000000694 effects Effects 0.000 description 3
- 238000000034 method Methods 0.000 description 3
- MHAJPDPJQMAIIY-UHFFFAOYSA-N Hydrogen peroxide Chemical compound OO MHAJPDPJQMAIIY-UHFFFAOYSA-N 0.000 description 2
- 238000010612 desalination reaction Methods 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 150000002431 hydrogen Chemical class 0.000 description 2
- 150000002500 ions Chemical class 0.000 description 2
- 229910001220 stainless steel Inorganic materials 0.000 description 2
- 239000010935 stainless steel Substances 0.000 description 2
- 239000003795 chemical substances by application Substances 0.000 description 1
- 239000000470 constituent Substances 0.000 description 1
- 230000001276 controlling effect Effects 0.000 description 1
- 230000005611 electricity Effects 0.000 description 1
- 230000007613 environmental effect Effects 0.000 description 1
- 239000012530 fluid Substances 0.000 description 1
- 238000005259 measurement Methods 0.000 description 1
- 239000000203 mixture Substances 0.000 description 1
- 230000001105 regulatory effect Effects 0.000 description 1
- 238000003756 stirring Methods 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
Landscapes
- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
Description
【発明の詳細な説明】
［発明の目的］
（産業上の利用分野）
本発明はたとえば沸騰水型原子力発電所における原子炉
構造材の応力腐食割れを抑制した原子力発電プラントに
関する。DETAILED DESCRIPTION OF THE INVENTION [Object of the Invention] (Industrial Application Field) The present invention relates to a nuclear power plant in which stress corrosion cracking of reactor structural materials in a boiling water nuclear power plant is suppressed, for example.
（従来の技術）
沸騰水型原子力発電プラントの構成を第３図を参照して
説明する。図中１は沸騰水型原子炉の原子炉圧力容器を
示しており、この原子炉圧力容器１内には炉心等の各種
の炉内機器（図示せず）が収容されている。この原子炉
圧力容器１内を流通する冷却材すなわち一次冷却水は炉
心で加熱され、高温高圧の蒸気となる。その圧力容器１
を出た蒸気は主蒸気管２を通流して高圧タービン３およ
び低圧タービン４に送られ、そのタービン３，４を回転
させ電気を起す。低圧タービン４を出た蒸気は復水器５
に送られて凝縮され、復水となる。この復水器５内の復
水は低圧復水ポンプ６によって復水脱塩装置７に送られ
、復水中のイオンが除去される。この復水脱塩装置７を
出た復水は給水ポンプ８によって給水加熱器９に送られ
て加熱され、給水配管１０から原子炉圧力容器１内に送
られる。(Prior Art) The configuration of a boiling water nuclear power plant will be explained with reference to FIG. In the figure, reference numeral 1 indicates a reactor pressure vessel of a boiling water reactor, and the reactor pressure vessel 1 accommodates various in-reactor equipment (not shown) such as a reactor core. The coolant, ie, the primary cooling water, flowing through the reactor pressure vessel 1 is heated in the reactor core and becomes high-temperature, high-pressure steam. The pressure vessel 1
The steam exiting the main steam pipe 2 is sent to a high pressure turbine 3 and a low pressure turbine 4, which rotate the turbines 3 and 4 to generate electricity. The steam leaving the low pressure turbine 4 is sent to the condenser 5
The water is sent to the water, where it is condensed and becomes condensate. The condensate in the condenser 5 is sent to a condensate desalination device 7 by a low-pressure condensate pump 6, where ions in the condensate are removed. The condensate that has exited the condensate desalination device 7 is sent to a feed water heater 9 by a feed water pump 8, heated, and sent into the reactor pressure vessel 1 through a water feed pipe 10.
原子炉圧力容器１に供給された炉水は再循環系１１によ
って炉心部を通し加熱される。また、原子炉冷却材浄化
系１２によって水質を保持する。この炉水中の溶存酸素
を抑制するために水素供給源１３から水素注入装置を通
して復水配管、給水配管１０あるいは再循環系１１から
水素を注入する。水素が炉内に入ることにより炉水中の
溶存酸素濃度が抑制きれ、炉水に接する構成材料の応ツ
ノ腐食割れを抑制する。Reactor water supplied to the reactor pressure vessel 1 is heated by a recirculation system 11 through the reactor core. In addition, water quality is maintained by the reactor coolant purification system 12. In order to suppress dissolved oxygen in the reactor water, hydrogen is injected from a hydrogen supply source 13 through a hydrogen injection device, from condensate piping, water supply piping 10, or recirculation system 11. By allowing hydrogen to enter the reactor, the dissolved oxygen concentration in the reactor water can be suppressed, thereby suppressing corrosive cracking of constituent materials that come into contact with the reactor water.
一般に原子炉圧力容器１内に収容された炉内機器は５（
ＪＳ３０４．５ＵＳ３１６等のステンレス鋼材料で形成
されている。ところで、このステンレス鋼材料は使用条
件によって応力腐食割れを生じることがある。この応力
腐食割れは材料に生じる応力、材料の組成等、および材
料の周囲の環境によって影響され、従来から応力、材料
的な面からこの応力腐食割れを防止する対策が講じられ
ている。また、材料の置かれている環境を改善して応力
腐食割れを防止することもなされている。その手段は原
子炉圧力容器内を流通する冷却水すなわち一次冷却水中
の酸素濃度を規制して応力腐食割れを防止するものであ
る。この−次冷却水の酸素！！度を規制するには一次冷
却水中に水素を注入し、この水素と酸素とを結合させて
水に戻すことによって酸素濃度を制御することがなされ
ている。In general, the reactor equipment housed in the reactor pressure vessel 1 is 5 (
It is made of stainless steel material such as JS304.5US316. By the way, stress corrosion cracking may occur in this stainless steel material depending on the usage conditions. This stress corrosion cracking is influenced by the stress generated in the material, the composition of the material, and the environment surrounding the material, and measures have been taken to prevent this stress corrosion cracking from the stress and material aspects. Efforts have also been made to improve the environment in which materials are placed to prevent stress corrosion cracking. The method is to prevent stress corrosion cracking by regulating the oxygen concentration in the cooling water, that is, the primary cooling water, flowing within the reactor pressure vessel. Oxygen in this next cooling water! ! In order to regulate the oxygen concentration, hydrogen is injected into the primary cooling water, and the hydrogen and oxygen are combined and returned to water to control the oxygen concentration.
〈発明が解決しようとする課題）
原子炉構造材料の応力腐食割れは材料、応力および環境
の３要素が関連して発生する。環境因子として溶存酸素
濃度および種々のイオンの存在が関係している。このう
ちで、溶存酸素濃度を抑制するために起動時の脱気運転
またはたとえば特開昭５７−３０８６号公報に開示され
たような水素注入運転方法が知られている。これらの運
転を行った場合、炉内に大局的に流動する炉水に接する
部分には有効であるが、炉水の流動が殆んどない部分に
はその効果が及ばない課題がある。すなわち、原子炉圧
力容器１に接続し、通常弁で閉じられている配管および
先端がフランジまたはキャップなどで閉塞され、行き止
りの配管では十分に溶存酸素濃度が抑制されない箇所が
存在する課題がある。<Problems to be Solved by the Invention> Stress corrosion cracking of nuclear reactor structural materials occurs due to the relationship between three factors: material, stress, and environment. Environmental factors include dissolved oxygen concentration and the presence of various ions. Among these methods, a deaeration operation at startup or a hydrogen injection operation method as disclosed in JP-A No. 57-3086, for example, are known to suppress the dissolved oxygen concentration. When these operations are carried out, they are effective in areas that come into contact with reactor water that generally flows within the reactor, but there is a problem in that the effects do not extend to areas where there is almost no flow of reactor water. That is, there is a problem in that there are places where the piping connected to the reactor pressure vessel 1 and normally closed with a valve and its tip are blocked with flanges or caps, and the dissolved oxygen concentration is not sufficiently suppressed in dead-end piping. .
本発明は上記課題を解決するためになされたもので、原
子炉圧力容器に接続し、高温になる箇所で炉水の流動が
なく溶存酸素濃度が炉水と同等にならない箇所の溶存酸
素濃度の低減をはかり、応力腐食割れを抑制する原子力
発電プラントを提供することにある。The present invention has been made to solve the above problem, and it is possible to reduce the dissolved oxygen concentration at a point connected to the reactor pressure vessel where the reactor water does not flow and the dissolved oxygen concentration does not reach the same level as the reactor water at a high temperature point. The object of the present invention is to provide a nuclear power plant that reduces stress corrosion cracking and suppresses stress corrosion cracking.
［発明の構成］
（課題を解決するための手段）
本発明は軽水炉型原子炉圧力容器に端部が閉塞されて、
該圧力容器内の炉水が停滞する配管部に小口径ラインを
接続し、この小口径ラインに脱気水または水素を溶解し
た水を注入する脱気水源または水素注入装置を設けてな
ることを特徴とする。[Structure of the Invention] (Means for Solving the Problems) The present invention provides a light water reactor type nuclear reactor pressure vessel with an end closed,
A small-diameter line is connected to the piping section where reactor water stagnates in the pressure vessel, and a deaerated water source or hydrogen injection device is installed to inject deaerated water or hydrogen-dissolved water into this small-diameter line. Features.
（作 用）
原子炉圧力容器内の炉水に接しているが、流動が殆んど
生じない停滞配管に小口径配管を接続し、この小口径配
管を通して脱気水または水素を溶解した水を注入すると
ぞの配管内の炉水の溶存酸素Ｉ！度が低減できる。この
溶存酸素濃度を低減することによって原子炉構造材の応
力腐食割れを防止することができる。(Function) A small-diameter pipe is connected to the stagnation pipe, which is in contact with reactor water in the reactor pressure vessel but has almost no flow, and degassed water or water with dissolved hydrogen is introduced through this small-diameter pipe. Dissolved oxygen in the reactor water in the piping where it is injected! degree can be reduced. By reducing this dissolved oxygen concentration, stress corrosion cracking of reactor structural materials can be prevented.
（実施例）
第１図を参照しながら本発明に係る原子力発電プラント
の一実施例を説明する。なお、第１図において第３図と
同一部分には同一符号で示し、重複する部分の説明を省
略する。(Example) An example of a nuclear power plant according to the present invention will be described with reference to FIG. In FIG. 1, the same parts as in FIG. 3 are indicated by the same reference numerals, and the explanation of the overlapping parts will be omitted.
すなわち、本発明が従来例と異なる部分は原子炉圧力容
器１の側面および下面に接続された停滞配管部１５．１
６にバルブＶ１．Ｖ２ 、Ｖ３を接続し、脱気水タンク
１７に接続された小口径配管１８にポンプ１９を接続し
、このポンプ１９の吐出側に分岐管１８ａ 、 １８ｂ
、 １８ｃを接続して、これら分岐管１８ａ。That is, the difference between the present invention and the conventional example is the stagnation piping section 15.1 connected to the side and bottom surfaces of the reactor pressure vessel 1.
6 with valve V1. V2 and V3 are connected, a pump 19 is connected to a small diameter pipe 18 connected to a deaerated water tank 17, and branch pipes 18a and 18b are connected to the discharge side of this pump 19.
, 18c are connected to these branch pipes 18a.
１８ｂ 、 １８ｃ ヲソｔｌｔＬ前記ｚニル７Ｖｔ
、 Ｖ２ 。18b, 18c woso tltL said z nil 7Vt
, V2.
Ｖ３の出口側に接、続したことにある。小口径配管１８
の内径は停滞配管部１５ｊ６ａの内径よりも小さく、脱
気水を注入した際に停滞配管部１５．１６内の炉水に攪
拌を与えて流動状態を生じさせるに十分な寸法を有ｊノ
でいる。停滞配管部１５．１６は原子炉圧力容器１に接
続し、通常弁で閉じられている配管または先端がフラン
ジないしはキャップなどで閉塞されて行き止まり状態に
なっているものを指している。This is because it is connected to the exit side of V3. Small diameter piping 18
The inner diameter is smaller than the inner diameter of the stagnation piping section 15.16, and has a size sufficient to stir the reactor water in the stagnation piping section 15.16 to create a fluid state when deaerated water is injected. There is. The stagnation piping section 15, 16 refers to a piping connected to the reactor pressure vessel 1, which is normally closed with a valve, or a piping whose end is closed with a flange or a cap, resulting in a dead end state.
しかして、上記構成の原子力発電プラントにおいて°、
原子炉を停止常態から起動させるには、まず−次冷却水
を循環させるとともに復水器５内を真空引して一次冷却
水を脱気し、原子炉停止中に−次冷却水中に溶解した酸
素等の濃度を所定の値以下に下げる。つぎに原子炉の運
転を開始するが、原子炉の運転によって一次冷却水の一
部は放射線分解されて酸素と水素を発生し、発生した酸
素は溶存酸素や過酸化水素の形で一次冷却水中に存在す
る。この−次冷却水中の溶存酸素は測定装置で測定され
、その測定結果から最適の水素注入量が算出され水素注
入量が制御される。この水素注入量を制御することによ
って溶存酸素ａ度を所定の範囲に維持し、原子炉圧力容
器１内の流動する一次冷却水と接触している原子炉構造
材の応力腐食割れを防止することができる。However, in a nuclear power plant with the above configuration,
In order to start the reactor from a normal shutdown state, firstly, the primary cooling water is circulated and the inside of the condenser 5 is evacuated to degas the primary cooling water, and the primary cooling water is dissolved in the secondary cooling water during the reactor shutdown. Lower the concentration of oxygen, etc. below a predetermined value. Next, the reactor starts operating, but as the reactor operates, a portion of the primary cooling water is radiolyzed and generates oxygen and hydrogen, and the generated oxygen is dissolved in the primary cooling water in the form of dissolved oxygen and hydrogen peroxide. exists in Dissolved oxygen in this secondary cooling water is measured by a measuring device, and the optimum hydrogen injection amount is calculated from the measurement result and the hydrogen injection amount is controlled. By controlling this amount of hydrogen injection, dissolved oxygen degree A is maintained within a predetermined range, and stress corrosion cracking of reactor structural materials in contact with flowing primary cooling water in the reactor pressure vessel 1 is prevented. Can be done.
ここで、前述したように原子炉圧力容器１に接続し、弁
１５ａで閉じられている配管部１５および先端がフラン
ジまたはキャップなどで行き止まりの配管部１６内の酸
素低減効果が及んでいない。そこで、脱気タンク１７か
ら小口径配管１８および駆動用ポンプを通して配管部１
５．１６内に溶存酸素を低減した水を供給することによ
って配管部１５．１６内の炉水の溶存酸素濃度を低減す
ることができる。Here, as described above, the oxygen reduction effect in the piping part 15 connected to the reactor pressure vessel 1 and closed by the valve 15a and the piping part 16 whose tip is a dead end due to a flange or cap, etc., is not achieved. Therefore, the piping part 1 is passed from the deaeration tank 17 to the small diameter piping 18 and the driving pump.
The dissolved oxygen concentration of the reactor water in the piping section 15.16 can be reduced by supplying water with reduced dissolved oxygen in the pipe section 15.16.
なお、脱気水を得るため、水源を脱気タンク１γに代え
復水または給水系配管から導くことができる。In addition, in order to obtain degassed water, the water source can be led from condensate water or water supply system piping instead of the deaeration tank 1γ.
第２図は本発明の他の実施例を示したもので、第１図と
同一部分には同一符号を付して重複する部分の説明を省
略する。FIG. 2 shows another embodiment of the present invention, in which the same parts as those in FIG.
すなわち、この実施例では小口径配管１８に水素注入装
置２１を接続し、原子炉圧力容器１内への水素供給ライ
ンとして利用するとともに停滞配管部１５、１６の溶存
酸素濃度を低減するものである。まず、原子炉圧ツノ容
器１から配管２０で取り出した炉水中に水素注入装置２
１から注入しポンプ１９で小口径配管１８を通し配管部
１５．ｔ６部から原子炉圧力容器１に戻す。この場合、
水素注入量は必要量仝」をこの水素注入装置２１から注
入する場合もあり、また給水系等信からの水素注入と分
担して注入することも可能である。原子炉圧力容器１か
ら取り出す配管２０は、別に原子炉圧力容器１から取り
出した配管例えば原子炉冷却材浄化系配管から分岐する
ことも可能である。That is, in this embodiment, a hydrogen injection device 21 is connected to the small-diameter pipe 18 and used as a hydrogen supply line into the reactor pressure vessel 1, and at the same time, the dissolved oxygen concentration in the stagnation pipe parts 15 and 16 is reduced. . First, the hydrogen injection device 2
Inject from the pipe section 15.1 through the small diameter pipe 18 using the pump 19. Return to the reactor pressure vessel 1 from the t6 section. in this case,
The required amount of hydrogen may be injected from this hydrogen injection device 21, or it is also possible to inject the hydrogen in a shared manner with the hydrogen injection from the water supply system, etc. The pipe 20 taken out from the reactor pressure vessel 1 can also be branched from another pipe taken out from the reactor pressure vessel 1, for example, a reactor coolant purification system pipe.
また取り出し配管２０はたとえば炉内計装系ハウジング
等の配管を適用することもできる。Further, as the take-out piping 20, for example, piping such as a housing for an in-furnace instrumentation system can be used.
なお、本発明は原子炉起動時、通常運転時および停止操
作時にも利用できる。Note that the present invention can also be used at the time of reactor startup, normal operation, and shutdown operation.
［発明の効果］
本発明によれば原子炉圧力容器に接続し、高温になる箇
所で流動がないため溶存酸素濃度が炉水と同等にならな
い箇所の溶存酸素濃度を低減することができる。[Effects of the Invention] According to the present invention, it is possible to reduce the dissolved oxygen concentration at a location connected to the reactor pressure vessel and where the dissolved oxygen concentration is not equal to that of reactor water because there is no flow at the location where the temperature becomes high.
第１図および第２図はそれぞれ本発明に係る原子力発電
プラントのそれぞれの実施例を線図的に示ず系統図、第
３図は従来の原子力発電プラントを線図的に示す系統図
である。
１・・・原子炉圧力容器
２・・・主蒸気配管
３・・・高圧タービン
４・・・低圧タービン
５・・・復水器
６・・・復水ポンプ
７・・・復水浄化系
８・・・給水ポンプ
９・・・給水加熱器
１０・・・給水配管
１１・・・再循環系
１２・・・原子炉冷却材浄化系
１３・・・水素供給源
１４・・・水素注入装置
１５．１６・・・停滞配管部
１７・・・脱気水タンク
１８・・・小口径配管
１９・・・ポンプ
２０・・・取り出し配管
２１・・・水素注入装置
（８７３３）代理人 弁理士 猪 股 祥 晃（ばか
１名）
第
図
第
図1 and 2 are system diagrams that do not diagrammatically show respective embodiments of a nuclear power plant according to the present invention, and FIG. 3 is a system diagram that diagrammatically shows a conventional nuclear power plant. . 1...Reactor pressure vessel 2...Main steam piping 3...High pressure turbine 4...Low pressure turbine 5...Condenser 6...Condensate pump 7...Condensate purification system 8 ... Water supply pump 9 ... Water supply heater 10 ... Water supply piping 11 ... Recirculation system 12 ... Reactor coolant purification system 13 ... Hydrogen supply source 14 ... Hydrogen injection device 15 .16...Stagnation piping section 17...Deaerated water tank 18...Small diameter piping 19...Pump 20...Takeout piping 21...Hydrogen injection device (8733) Agent Patent attorney Inomata Yoshiaki (baka)
1 person) Figure Figure
Claims (1)
内の炉水が停滞する配管部に小口径ラインを接続し、こ
の小口径ラインに脱気水または水素を溶解した水を注入
する脱気水源または水素注入装置を設けてなることを特
徴とする原子力発電プラント。A small-diameter line is connected to the piping section where the end of the light water reactor pressure vessel is blocked and the reactor water in the pressure vessel stagnates, and degassed water or water with dissolved hydrogen is injected into this small-diameter line. A nuclear power plant characterized by being equipped with a deaerated water source or a hydrogen injection device.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP1041923A JPH02222893A (en) | 1989-02-23 | 1989-02-23 | Nuclear power plant |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP1041923A JPH02222893A (en) | 1989-02-23 | 1989-02-23 | Nuclear power plant |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH02222893A true JPH02222893A (en) | 1990-09-05 |
Family
ID=12621768
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP1041923A Pending JPH02222893A (en) | 1989-02-23 | 1989-02-23 | Nuclear power plant |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH02222893A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2000058974A1 (en) * | 1999-03-26 | 2000-10-05 | Hitachi, Ltd. | Method of operating reactor |
-
1989
- 1989-02-23 JP JP1041923A patent/JPH02222893A/en active Pending
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2000058974A1 (en) * | 1999-03-26 | 2000-10-05 | Hitachi, Ltd. | Method of operating reactor |
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