JPS6150279B2 - - Google Patents
Info
- Publication number
- JPS6150279B2 JPS6150279B2 JP54102342A JP10234279A JPS6150279B2 JP S6150279 B2 JPS6150279 B2 JP S6150279B2 JP 54102342 A JP54102342 A JP 54102342A JP 10234279 A JP10234279 A JP 10234279A JP S6150279 B2 JPS6150279 B2 JP S6150279B2
- Authority
- JP
- Japan
- Prior art keywords
- reactor
- water
- pressure
- pressure vessel
- startup
- 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
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 41
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 claims description 29
- 239000001301 oxygen Substances 0.000 claims description 29
- 229910052760 oxygen Inorganic materials 0.000 claims description 29
- 238000000034 method Methods 0.000 claims description 19
- 238000009835 boiling Methods 0.000 claims description 5
- 230000004907 flux Effects 0.000 claims description 4
- 238000010586 diagram Methods 0.000 description 10
- 238000007872 degassing Methods 0.000 description 8
- 230000001629 suppression Effects 0.000 description 5
- 238000002955 isolation Methods 0.000 description 4
- 239000013535 sea water Substances 0.000 description 4
- MHAJPDPJQMAIIY-UHFFFAOYSA-N Hydrogen peroxide Chemical compound OO MHAJPDPJQMAIIY-UHFFFAOYSA-N 0.000 description 2
- 230000009471 action Effects 0.000 description 2
- 229910000963 austenitic stainless steel Inorganic materials 0.000 description 2
- 230000000694 effects Effects 0.000 description 2
- 239000007789 gas Substances 0.000 description 2
- 238000010438 heat treatment Methods 0.000 description 2
- 239000000463 material Substances 0.000 description 2
- 230000035945 sensitivity Effects 0.000 description 2
- 238000009849 vacuum degassing Methods 0.000 description 2
- 206010070834 Sensitisation Diseases 0.000 description 1
- 208000027418 Wounds and injury Diseases 0.000 description 1
- 238000009833 condensation Methods 0.000 description 1
- 230000005494 condensation Effects 0.000 description 1
- 239000002826 coolant Substances 0.000 description 1
- 238000001816 cooling Methods 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
- 230000006378 damage Effects 0.000 description 1
- 238000000354 decomposition reaction Methods 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 230000006870 function Effects 0.000 description 1
- 208000014674 injury Diseases 0.000 description 1
- 238000012544 monitoring process Methods 0.000 description 1
- 230000008569 process Effects 0.000 description 1
- 230000005855 radiation Effects 0.000 description 1
- 238000003608 radiolysis reaction Methods 0.000 description 1
- 230000009467 reduction Effects 0.000 description 1
- 229920006395 saturated elastomer Polymers 0.000 description 1
- 230000008313 sensitization Effects 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
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Description
【発明の詳細な説明】
本発明は、原子炉の起動方法、特に、起動前に
原子炉圧力容器内を真空脱気しておいて起動する
方法に関するものである。DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a method for starting a nuclear reactor, and in particular to a method for starting a nuclear reactor after evacuating the inside of the reactor pressure vessel.
原子炉、特に、沸騰水型原子炉において、炉水
中の溶存酸素は、原子炉一次系を構成するオース
テナイトステンレス鋼の応力腐食割れ（SCC）
の発生を助長するため、その濃度を極力低減する
ことが重要である。SCCは、応力と材料の鋭敏
化と、腐食環境が重なりあう条件下で発生するの
で、原子炉起動時のように、圧力と温度の過渡変
化時には、特に、腐食環境を形成する要因である
溶存酸素を低減する対策が実施されている。 In nuclear reactors, especially boiling water reactors, dissolved oxygen in the reactor water can cause stress corrosion cracking (SCC) in the austenitic stainless steel that constitutes the reactor primary system.
It is important to reduce the concentration as much as possible in order to encourage the generation of SCC occurs under conditions where stress and material sensitization coexist with a corrosive environment. Measures are being taken to reduce oxygen.
第１図は、従来の起動方法を説明する原子炉廻
りの系統の概要を示す系統図で、１は炉心、２は
原子炉圧力容器、３は制御棒駆動装置、４は主蒸
気ライン、５は主蒸気隔離弁、６は主蒸気ドレン
ライン、７は弁、８はタービン、９は主復水器、
１０は真空装置、１１は原子炉格納容器である。
原子炉圧力容器２内には、冷却材（炉水）が定常
水位に保たれている。この炉水には、空気中の酸
素が溶解し、溶存酸素が５〜9ppmと高濃度にな
る。従つて、このままの状態で原子炉を起動させ
ると、高い溶存酸素濃度の下で応力が加わるの
で、SCCの発生感受性を高めることから、起動
前に酸素を脱気する技術が用いられている。すな
わち、現在のシステムでは、タービン８の主復水
器９を真空装置１０で真空とするようになつてい
るので、主蒸気ライン４と、その隔離弁５の主蒸
気ドレンライン６を用いて、弁７を開閉操作し
て、原子炉圧力容器２内を真空にし、炉水中の溶
存酸素を真空脱気している。 Figure 1 is a system diagram showing an overview of the system around the nuclear reactor to explain the conventional startup method, where 1 is the core, 2 is the reactor pressure vessel, 3 is the control rod drive device, 4 is the main steam line, and 5 is the system diagram showing the outline of the system around the nuclear reactor. is the main steam isolation valve, 6 is the main steam drain line, 7 is the valve, 8 is the turbine, 9 is the main condenser,
10 is a vacuum device, and 11 is a reactor containment vessel.
Inside the reactor pressure vessel 2, coolant (reactor water) is maintained at a steady water level. Oxygen in the air dissolves in this reactor water, resulting in a high concentration of dissolved oxygen of 5 to 9 ppm. Therefore, if a nuclear reactor is started in this state, stress will be applied due to the high concentration of dissolved oxygen, which will increase the susceptibility to SCC occurrence, so technology is used to degas the oxygen before starting up. That is, in the current system, the main condenser 9 of the turbine 8 is evacuated by the vacuum device 10, so using the main steam line 4 and the main steam drain line 6 of its isolation valve 5, The valve 7 is opened and closed to create a vacuum inside the reactor pressure vessel 2, and dissolved oxygen in the reactor water is degassed.
しかし、このように起動前に真空脱気しておい
て、起動した場合でも、時間の経過とともに一時
的に溶存酸素濃度にピークがあらわれ、SCCの
感受性を高める結果となることが明らかとなつ
た。 However, it has become clear that even if the engine is vacuum degassed before startup and then started, a peak in dissolved oxygen concentration will appear temporarily over time, resulting in increased susceptibility to SCC. .
本発明は、このような起動時に発生する一時的
な高SCC感受性領域の発生を防止できる原子炉
の起動方法を提供することを目的とし、原子炉の
起動前に原子炉圧力容器内を真空脱気しておいて
起動する原子炉の起動方法において、起動後炉心
の中性子束密度が1012n／cm2以上になると、前記
原子炉圧力容器の内部圧力を調整して、間欠的に
減圧沸騰状態にし、炉水の溶存酸素濃度を
0.2ppm以下に保持しながら起動することを特徴
とするものである。 The purpose of the present invention is to provide a method for starting a nuclear reactor that can prevent the occurrence of a temporary high SCC sensitivity region that occurs during startup. In the method of starting up a nuclear reactor, when the neutron flux density of the reactor core reaches 10 12 n/cm 2 or more after startup, the internal pressure of the reactor pressure vessel is adjusted and the vacuum boiling is intermittently carried out. condition, and the dissolved oxygen concentration in the reactor water is
The feature is that it starts up while keeping it below 0.2ppm.
第２図は、従来の起動前に真空脱気しておく起
動方法を実施した場合における実機の水質調査結
果を示すもので、横軸には、時間、縦軸には、炉
水酸素濃度（炉水中の溶存酸素濃度）（ppm）、
炉水温度（℃）、炉圧（Kg／cm2ｇ）がとつてあ
り、T1，T2は、それぞれ、脱気時、起動時を示
し、Ａ，Ｂ，Ｃは、それぞれ、炉水酸素濃度、炉
水温度、炉圧を示している。真空脱気操作によ
り、炉水酸素濃度Ａは、脱気時T1においては、
0.2ppm以下にすることができる。しかし、起動
時T2からの時間経過に従つて炉水温度Ｂと、炉
圧Ｃは上昇し、その過程で、200℃以下では、起
動に伴い発生する放射線によつて、炉水が分解
し、水や過酸化水素、及び酸素が生成するため、
起動前に脱気しておいても一時的な炉水酸素濃度
のピークが表れることを示している。しかも、こ
の一時的なピーク濃度は、0.2ppmをこえ、高い
場合は、1ppmオーダとなることがあり、これ
が、起動時の応力に加わるため、SCCの感受性
を高める結果となることが明らかとなつた。すな
わち、従来の脱気運転法では、中高温度の放射線
分解生成物の一時的高濃度スパイクを抑制するこ
とができなかつた。なお、現在の材料試験の結果
では、炉水酸素濃度が0.2ppm以下では、SCCは
発生しにくいと言われている。 Figure 2 shows the results of an actual water quality survey when the conventional startup method of vacuum degassing before startup was implemented. The horizontal axis is time, and the vertical axis is reactor water oxygen concentration ( Dissolved oxygen concentration in reactor water) (ppm),
The reactor water temperature (°C) and reactor pressure (Kg/cm 2 g) are given, T 1 and T 2 indicate the time of degassing and startup, respectively, and A, B, and C indicate the reactor water temperature, respectively. Shows oxygen concentration, reactor water temperature, and reactor pressure. Due to the vacuum degassing operation, the reactor water oxygen concentration A at degassing time T 1 is as follows:
It can be reduced to 0.2ppm or less. However, as time passes from startup T2 , the reactor water temperature B and reactor pressure C rise, and in the process, at temperatures below 200°C, the reactor water decomposes due to the radiation generated during startup. , water, hydrogen peroxide, and oxygen are produced.
This shows that even if the reactor is degassed before startup, a temporary peak in the oxygen concentration of the reactor water appears. Moreover, this temporary peak concentration can exceed 0.2 ppm and, in high cases, reach the order of 1 ppm, and it has become clear that this adds to the stress at startup, resulting in increased susceptibility to SCC. Ta. That is, the conventional degassing operation method was unable to suppress temporary high concentration spikes of radiolysis products at medium and high temperatures. According to current material test results, it is said that SCC is unlikely to occur when the oxygen concentration in the reactor water is 0.2 ppm or less.
本発明は、このような従来技術の検討結果に基
づきなされたもので、原子炉起動時の原子炉圧力
容器内のような密ぺい系内で、核加熱、加温によ
る圧力上昇過程において、系内の圧力を一時的に
減圧状態にして、系内の蒸気の減圧沸騰条件をつ
くり、この作用により、炉水中の溶存酸素等の不
凝縮性ガスを脱気し、SCCの感受性を低減する
ものである。 The present invention has been made based on the results of studies on the prior art, and is based on the results of a study of the prior art. Temporarily reduces the pressure inside the reactor to create a reduced-pressure boiling condition for the steam within the system, and through this action degass non-condensable gases such as dissolved oxygen in the reactor water, reducing susceptibility to SCC. It is.
以下、実施例について説明する。 Examples will be described below.
第３図は、一実施例を実施する原子炉の概略を
示すもので、１は炉心、２は原子炉圧力容器、３
は制御棒駆動装置、１１は原子炉格納容器、１２
は配管、１３は電動弁、１４は真空破壊弁、１５
は圧力抑制プール水、１６は炉圧計、１７は溶存
酸素計を示している。 FIG. 3 shows an outline of a nuclear reactor in which one embodiment is implemented, in which 1 is a reactor core, 2 is a reactor pressure vessel, and 3 is a reactor.
is a control rod drive device, 11 is a reactor containment vessel, 12
is piping, 13 is an electric valve, 14 is a vacuum breaker valve, 15
16 indicates the pressure suppression pool water, 16 indicates the furnace pressure gauge, and 17 indicates the dissolved oxygen meter.
この原子炉においては、原子炉を起動し、制御
棒駆動装置３により制御棒を引抜いて核加熱に入
ると、炉水は加熱され、蒸気が発生し、飽和蒸気
圧で上昇する。これは、炉圧計１６で確認でき
る。炉心の中性子束密度が1012〜1013n／cm2・sec
附近で、水の分解が始まり、炉水中の溶存酸素濃
度は高くなる。従つて、この中性子束密度と温度
および溶存酸素計１７を監視しながら、溶存酸素
濃度が0.2ppm以下になるように、蒸気を原子炉
圧力容器２の気相部につながる配管１２を介し
て、圧力抑制プール水１５に導き、凝縮させる。
圧力の加減は、電動弁１３で操作する。配管１２
の排気管には、蒸気凝縮にともなう真空破壊弁１
４を設けてある。電動弁１３は、故障時を考慮
し、直列に２個設置することが好ましい。配管の
口径は、プラントの大きさにより異なるが１〜10
インチで十分である。弁の操作は、炉昇温速度に
よつて異なるが、一回以上実施するのが好まし
い。 In this nuclear reactor, when the reactor is started and the control rods are pulled out by the control rod driving device 3 to begin nuclear heating, reactor water is heated, steam is generated, and the steam rises at saturated steam pressure. This can be confirmed with the furnace pressure gauge 16. The neutron flux density in the core is 10 12 to 10 13 n/cm 2 sec
Water begins to decompose nearby, and the dissolved oxygen concentration in the reactor water increases. Therefore, while monitoring the neutron flux density, temperature, and dissolved oxygen meter 17, steam is passed through the pipe 12 connected to the gas phase part of the reactor pressure vessel 2 so that the dissolved oxygen concentration is 0.2 ppm or less. It is led to pressure suppression pool water 15 and condensed.
The pressure is controlled by an electric valve 13. Piping 12
There is a vacuum break valve 1 in the exhaust pipe for steam condensation.
4 is provided. It is preferable to install two electric valves 13 in series in consideration of failure. The diameter of the piping varies depending on the size of the plant, but is between 1 and 10.
Inches are sufficient. The valve operation varies depending on the furnace temperature increase rate, but is preferably performed at least once.
本実施例によれば、起動時特有の中高温領域で
水の分解生成物として発生した溶存酸素を一時的
にSCC発生の感受性を高めることのない低濃度
に抑制できるので、原子炉一次系のオーステナイ
トステンレス鋼のSCCの発生を防止することが
できる。第４図は、プラントの試験結果を示すも
ので、横軸には、時間、縦軸には、炉水酸素濃度
（ppm）、炉水温度（℃）、炉圧（Kg／cm2ｇ）がと
つてあり、T1，T2は、それぞれ、脱気時、起動
時、ｔは減圧脱気時間を示し、Ｄ，Ｅ，Ｆは、そ
れぞれ、炉水酸素濃度、炉水温度、炉圧を示して
いる。すなわち、従来の起動前の脱気運転が終了
して起動してから、中央制御室等からの遠隔操作
により電動弁１３を開閉し、減圧脱気（時間ｔ）
を行なうと、炉水の酸素濃度Ｄは0.2ppmをこえ
ることなく、高濃度になることを防止でき、起動
時の一時的スパイクを抑制できる。 According to this embodiment, dissolved oxygen generated as a water decomposition product in the mid-to-high temperature region unique to startup can be temporarily suppressed to a low concentration that does not increase the susceptibility to SCC generation, so that the The occurrence of SCC in austenitic stainless steel can be prevented. Figure 4 shows the test results of the plant, where the horizontal axis shows time, and the vertical axis shows reactor water oxygen concentration (ppm), reactor water temperature (°C), and reactor pressure (Kg/cm 2 g). T 1 and T 2 are the degassing time and startup time, t is the depressurization degassing time, and D, E, and F are the reactor water oxygen concentration, reactor water temperature, and reactor pressure, respectively. It shows. That is, after the conventional degassing operation before startup is completed and the startup is started, the electric valve 13 is opened and closed by remote control from the central control room, etc., and the depressurization degassing (time t) is performed.
If this is done, the oxygen concentration D in the reactor water will not exceed 0.2 ppm, preventing it from becoming too high, and suppressing temporary spikes during startup.
すなわち、炉水が100℃から200℃の領域で遠隔
操作仕切弁を間欠的に開閉することにより原子炉
圧力容器内の圧力を減圧しながら、炉水昇温を計
ることにより、核加熱による昇圧と、その昇圧よ
り若干小さい圧力差まで、一時的に、瞬時に減圧
し、このような操作を数回綴り返し実施すること
により、炉水中の溶存酸素濃度を0.2ppm以下に
抑制することができる。 In other words, the pressure inside the reactor pressure vessel is reduced by intermittently opening and closing a remote-controlled gate valve when the reactor water is in the range of 100°C to 200°C, and by measuring the temperature rise of the reactor water, the pressure increase due to nuclear heating is measured. By temporarily and instantaneously reducing the pressure to a pressure difference slightly smaller than the pressure increase, and repeating this operation several times, it is possible to suppress the dissolved oxygen concentration in the reactor water to 0.2 ppm or less. .
この実施例によれば、炉内の減圧沸騰脱気作用
を効果的に実施させることができるので、炉水中
の溶存酸素濃度をSCC感受性を避けた領域にし
て運転することが可能となつた。 According to this example, since the reduced pressure boiling degassing action inside the reactor can be effectively carried out, it has become possible to operate the reactor with the dissolved oxygen concentration in the reactor water in a range that avoids SCC susceptibility.
なお、炉圧の調整は、信頼性の高い電動仕切弁
を設けることにより、操作運転が容易に実施でき
る。また、ここで使われる蒸気は、圧力抑制プー
ルで冷却凝縮されるので、原子炉格納容器１１内
の温度を上げることなく運転することができる。 Incidentally, the furnace pressure can be easily adjusted by providing a highly reliable electric gate valve. Moreover, since the steam used here is cooled and condensed in the pressure suppression pool, the reactor can be operated without increasing the temperature inside the reactor containment vessel 11.
第５図は、他の実施例を実施する原子炉の概略
を示すもので、第１図と同一部分には同一符号が
付してあり、１８は、タービンバイパスライン、
１９は、タービンバイパス弁、２０は、冷却用の
海水循環水である。この実施例が、前述の実施例
と異なる点は、起動後の蒸気を排気し、炉圧を減
少させる方法として、その蒸気をタービン主復水
器に導いている点で、具体的には、タービンバイ
パスライン１８が主復水器９の胴体空間部に連結
されている。すなわち、原子炉で発生した蒸気
は、主蒸気ライン４から主蒸気隔離弁５を介して
タービンバイパス配管１８に導かれ、タービンバ
イパス弁１９を開閉して炉圧を減圧する操作を行
なうことができる。なお、主復水器９に導かれた
蒸気は、海水循環水２０により冷却され、復水に
なる。この方法は、蒸気を海水循環水によつて冷
却できるので最も効率よく減圧を行なうことがで
きる。 FIG. 5 schematically shows a nuclear reactor implementing another embodiment, in which the same parts as in FIG. 1 are given the same reference numerals, and 18 is a turbine bypass line;
19 is a turbine bypass valve, and 20 is seawater circulation water for cooling. This embodiment differs from the previous embodiments in that the steam after startup is exhausted and the steam is guided to the turbine main condenser as a method of reducing furnace pressure. Specifically, A turbine bypass line 18 is connected to the body space of the main condenser 9. That is, the steam generated in the reactor is guided from the main steam line 4 to the turbine bypass piping 18 via the main steam isolation valve 5, and the turbine bypass valve 19 can be opened and closed to reduce the reactor pressure. . Note that the steam led to the main condenser 9 is cooled by the seawater circulation water 20 and becomes condensed water. In this method, the steam can be cooled by circulating seawater, so the pressure can be reduced most efficiently.
第６図は、さらに他の実施例を実施する原子炉
の概略を示すもので、第３図及び第５図と同一の
部分には同一の符号が付してあり、２１は、主蒸
気リリーフ弁である。この実施例が第３図の系統
を用いる実施例と異なる点は、起動後の蒸気を排
気して炉圧を減少させる方法として、蒸気を既設
の主蒸気リリーフ弁２１を用いて圧力抑制プーー
ル１５に導いている点である。すなわち、主蒸気
リリーフ弁１５は、手動遠隔操作により中央制御
室から開閉操作が可能であり、起動時に任意に炉
圧を減圧せしめることが可能となる。 FIG. 6 schematically shows a nuclear reactor implementing another embodiment, in which the same parts as in FIGS. 3 and 5 are given the same reference numerals, and 21 is the main steam relief It is a valve. The difference between this embodiment and the embodiment using the system shown in FIG. 3 is that in order to reduce the furnace pressure by exhausting the steam after startup, the steam is transferred to the pressure suppression pool 15 using the existing main steam relief valve 21. This is the point that leads to this. That is, the main steam relief valve 15 can be opened and closed from the central control room by manual remote control, and the furnace pressure can be arbitrarily reduced at startup.
この実施例は、減圧機能の点では、前述の実施
例の場合と同様な効果が得られるが、現在使われ
ている主蒸気リリーフ弁の構造では、主蒸気リリ
ーフ弁の開閉により弁シート面を傷つける可能性
があるため、多用する場合には、第３図の系統を
用いる実施例が使用される。 This embodiment has the same effect as the previous embodiment in terms of pressure reduction function, but with the structure of the main steam relief valve currently in use, the valve seat surface is reduced by opening and closing the main steam relief valve. Since there is a possibility of injury, the embodiment using the system shown in FIG. 3 is used when it is used frequently.
以上の如く、本発明の原子炉の起動方法は、起
動時に発生する一時的な高SCC感受性領域の発
生を防止可能としたもので、産業上の効果の大な
るものである。 As described above, the nuclear reactor startup method of the present invention makes it possible to prevent the occurrence of a temporary high SCC sensitivity region that occurs during startup, and has great industrial effects.
第１図は、従来の原子炉廻りの系統の概要を示
す系統図、第２図は、従来の原子炉の起動方法の
特性を示す線図、第３図は、本発明の原子炉の起
動方法の一実施例を実施する原子炉の概略系統
図、第４図は、本発明の原子炉の起動方法の特性
を示す線図、第５図は、本発明の原子炉の起動方
法の他の実施例を実施する原子炉廻りの系統の概
要を示す系統図、第６図は、同じく、他の実施例
を実施する原子炉の概略系統図である。
２……原子炉圧力容器、４……主蒸気ライン、
５……主蒸気隔離弁、８……タービン、９……主
復水器、１０……真空装置、１１……原子炉格納
容器、１２……配管、１３……電動弁、１５……
圧力抑制プール水、１６……炉圧計、１７……溶
存酸素計、１８……タービンバイパスライン、１
９……タービンバイパス弁、２０……海水循環
水、２１……主蒸気リリーフ弁。
Figure 1 is a system diagram showing an overview of the system around a conventional nuclear reactor, Figure 2 is a diagram showing the characteristics of a conventional nuclear reactor startup method, and Figure 3 is a diagram showing the characteristics of a conventional nuclear reactor startup method. A schematic system diagram of a nuclear reactor implementing one embodiment of the method, FIG. 4 is a diagram showing the characteristics of the nuclear reactor startup method of the present invention, and FIG. 5 is a diagram showing the characteristics of the nuclear reactor startup method of the present invention. FIG. 6 is a system diagram showing an overview of the system around a nuclear reactor in which this embodiment is implemented. Similarly, FIG. 6 is a schematic system diagram of a nuclear reactor in which another embodiment is implemented. 2... Reactor pressure vessel, 4... Main steam line,
5... Main steam isolation valve, 8... Turbine, 9... Main condenser, 10... Vacuum device, 11... Reactor containment vessel, 12... Piping, 13... Electric valve, 15...
Pressure suppression pool water, 16...Furnace pressure gauge, 17...Dissolved oxygen meter, 18...Turbine bypass line, 1
9...Turbine bypass valve, 20...Sea water circulation water, 21...Main steam relief valve.
Claims (1)
気しておいて起動する原子炉の起動方法におい
て、起動後炉心の中性子束密度が1012n／cm2以上
になると、前記原子炉圧力容器の内部圧力を調整
して、間欠的に減圧沸騰状態にし、炉水の溶存酸
素濃度を0.2ppm以下に保持しながら起動するこ
とを特徴とする原子炉の起動方法。 ２ 前記原子炉圧力容器の内部圧力の調整を、該
原子炉圧力容器の内側と外側とを接続する配管、
及び、該配管系において該原子炉圧力容器の内外
を隔離する弁を用いて行なう特許請求の範囲第１
項記載の原子炉の起動方法。 ３ 前記原子炉圧力容器の内側と外側とを接続す
る配管が、その端部においてタービン主復水器胴
体空間部に連結されている特許請求の範囲第２項
記載の原子炉の起動方法。[Claims] 1. A method for starting a nuclear reactor in which the reactor pressure vessel is vacuum degassed before starting the reactor, wherein the neutron flux density of the reactor core after startup is 10 12 n/cm 2 or more. A method for starting a nuclear reactor, characterized in that the internal pressure of the reactor pressure vessel is adjusted intermittently to bring it to a reduced pressure boiling state, and the reactor is started while maintaining the dissolved oxygen concentration of the reactor water at 0.2 ppm or less. . 2. Piping that connects the inside and outside of the reactor pressure vessel to adjust the internal pressure of the reactor pressure vessel;
and claim 1, which uses a valve in the piping system to isolate the inside and outside of the reactor pressure vessel.
How to start up a nuclear reactor as described in section. 3. The method for starting a nuclear reactor according to claim 2, wherein the piping connecting the inside and outside of the reactor pressure vessel is connected to the turbine main condenser body space at its end.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP10234279A JPS5626297A (en) | 1979-08-10 | 1979-08-10 | Nuclear reactor starttup method |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP10234279A JPS5626297A (en) | 1979-08-10 | 1979-08-10 | Nuclear reactor starttup method |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS5626297A JPS5626297A (en) | 1981-03-13 |
| JPS6150279B2 true JPS6150279B2 (en) | 1986-11-04 |
Family
ID=14324819
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP10234279A Granted JPS5626297A (en) | 1979-08-10 | 1979-08-10 | Nuclear reactor starttup method |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS5626297A (en) |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPH0766074B2 (en) * | 1985-07-31 | 1995-07-19 | 株式会社島津製作所 | Neutron shield |
-
1979
- 1979-08-10 JP JP10234279A patent/JPS5626297A/en active Granted
Also Published As
| Publication number | Publication date |
|---|---|
| JPS5626297A (en) | 1981-03-13 |
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