JP3068288B2 - Auxiliary cooling water system for nuclear power plants - Google Patents
Auxiliary cooling water system for nuclear power plantsInfo
- Publication number
- JP3068288B2 JP3068288B2 JP3307509A JP30750991A JP3068288B2 JP 3068288 B2 JP3068288 B2 JP 3068288B2 JP 3307509 A JP3307509 A JP 3307509A JP 30750991 A JP30750991 A JP 30750991A JP 3068288 B2 JP3068288 B2 JP 3068288B2
- Authority
- JP
- Japan
- Prior art keywords
- heat
- condensate
- auxiliary
- seawater
- heat exchanger
- 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
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Description
【０００１】[0001]
【産業上の利用分野】本発明は原子力発電プラントの補
機冷却水システムに関する。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to an auxiliary cooling water system for a nuclear power plant.
【０００２】[0002]
【従来の技術】原子力発電プラントでは電気を発生する
主システムである原子炉、蒸気タービンおよび発電機の
他にこれ等主システムを円滑に運転させるための種々の
補助機器類が設置されている。2. Description of the Related Art In a nuclear power plant, in addition to a main system for generating electricity, a reactor, a steam turbine, and a generator, various auxiliary devices for smoothly operating the main system are installed.
【０００３】補助機器類には各種のポンプ、ファン、空
気圧縮器、冷凍機などと、それらを駆動するための電動
機などの回転機類、水質維持用脱塩装置および潤滑油な
どのための熱交換器があり、これ等の機器の機能を維持
するためにプラント運転中、常に除熱を行なう必要があ
る。[0003] Auxiliary equipment includes various pumps, fans, air compressors, refrigerators, etc., and rotary machines such as electric motors for driving them, desalination devices for maintaining water quality, and heat for lubricating oil. There is an exchanger, and it is necessary to constantly remove heat during plant operation in order to maintain the functions of these devices.
【０００４】代表的な発熱量の多い補助機器および発熱
源には表１から表３に示すものがある。なお、引用する
数値は最近の 110万kw級ＢＷＲ原子力発電プラントのデ
ータである。更に参考として示すと同上プラントでは 1
10万kwの発電時、原子炉内で核反応により発生する熱は
2.85×109 kcal／Ｈであり、熱効率は約33％である。[0004] Tables 1 to 3 show typical auxiliary devices and heat sources having a large amount of heat generation. The figures quoted are from recent 1.1 million kW class BWR nuclear power plants. For reference, 1
When generating 100,000 kw, the heat generated by nuclear reactions in the reactor
2.85 × 10 9 kcal / H, and the thermal efficiency is about 33%.
【０００５】[0005]
【表１】 [Table 1]
【０００６】[0006]
【表２】 [Table 2]
【０００７】[0007]
【表３】 [Table 3]
【０００８】これ等を合計すると約５×107 kcal／Ｈと
なり、原子炉で発生した熱のうち約1.7％が蒸気タービ
ンに利用されないで消費され、この熱は以下で示す様に
プラント外へ海水に廃熱を与えて逃がすことになる。こ
の各種発熱源を冷却するシステムとして原子炉補機冷却
系（以下、ＲＣＷと記す）とタービン補機冷却系（以
下、ＴＣＷと記す）がある。ＲＣＷを図２に基き説明す
る。図中、黒塗の弁はプラントが通常運転中の場合、閉
まっており、ＲＣＷの冷却水が流れないことを示してい
る。When these are summed up, about 5 × 10 7 It becomes kcal / H, and about 1.7% of the heat generated in the nuclear reactor is consumed without being used for the steam turbine, and this heat gives off waste heat to seawater outside the plant as shown below. As systems for cooling these various heat sources, there are a reactor accessory cooling system (hereinafter, referred to as RCW) and a turbine accessory cooling system (hereinafter, referred to as TCW). The RCW will be described with reference to FIG. In the figure, the black valves are closed when the plant is in normal operation, indicating that the RCW cooling water does not flow.
【０００９】図２において、符号１は海水ポンプで、こ
の海水ポンプ１は海水２を海水熱交換器３に流入する。
海水熱交換器３は非常用熱負荷４と通常時熱負荷５の入
口側にそれぞれ元弁６、７を介して接続している。これ
ら非常用熱負荷４と通常時熱負荷５の出口側は元弁６
ａ，７ａを介して補機冷却水ポンプ９により海水熱交換
器３に接続されている。なお、この系統は２系統設置さ
れ、これらの系統は系統連絡弁により接続されている。In FIG. 2, reference numeral 1 denotes a seawater pump. The seawater pump 1 flows seawater 2 into a seawater heat exchanger 3.
The seawater heat exchanger 3 is connected to the inlet side of the emergency heat load 4 and the normal heat load 5 via main valves 6 and 7, respectively. The outlet side of the emergency heat load 4 and the normal heat load 5 is connected to the main valve 6.
The auxiliary cooling water pump 9 is connected to the seawater heat exchanger 3 via a and 7a. In addition, two systems are installed, and these systems are connected by a system communication valve.
【００１０】ＲＣＷは上記とほぼ同じ構成のシステムを
２系統合体させている。上記各熱負荷は、通常時熱負荷
５としてまとめて記載している。各種の原子炉関係の通
常時負荷は２系統のシステムの通常時熱負荷５がほぼ等
しい負荷量となる様に分散させる。非常用負荷４はプラ
ント運転中、事故などでプランドが停止した時に原子炉
で発生する崩壊熱を除去したり、原子炉に緊急注水を行
なうポンプを冷却する負荷であり、 100％容量の設備を
２系統持っている。これは原子炉の安全性の信頼度を上
げるためであり、一方のシステムに異常があり、作動で
きない時でも残りの１システムで原子炉が安全に停止さ
せることができるように配慮している。The RCW is a two-system integration of a system having substantially the same configuration as described above. The above heat loads are collectively described as a normal heat load 5. The normal loads related to various reactors are distributed so that the normal heat loads 5 of the two systems have substantially the same load. The emergency load 4 is a load that removes decay heat generated in the reactor when the plant is stopped due to an accident during operation of the plant, or cools a pump that performs emergency water injection into the reactor. I have two systems. This is to increase the reliability of the safety of the reactor. Consideration is given so that the reactor can be safely shut down in the remaining one system even when one system has an abnormality and cannot operate.
【００１１】ＲＣＷはこれ等の負荷４、５を閉サイクル
の中に組み入れシステムを構成している。補機冷却水ポ
ンプ９で圧力を上げた閉サイクルの水は海水熱交換器３
で海水により除熱冷却され各負荷に向かう。通常運転中
は通常時熱負荷５のみを通り各負荷を冷却し自身の冷却
水は排熱をもらって温度が上昇し、補機冷却水ポンプ９
に戻る。The RCW constitutes a system incorporating these loads 4, 5 in a closed cycle. The closed cycle water whose pressure has been increased by the auxiliary cooling water pump 9 is supplied to the seawater heat exchanger 3.
The heat is cooled by the seawater and goes to each load. During normal operation, each load is cooled by passing only the heat load 5 during normal operation, and the cooling water of the cooling water itself receives waste heat and the temperature rises.
Return to
【００１２】非常時は非常用熱負荷元弁６が自動的に開
き、通常時熱負荷元弁７が自動的に閉じて非常用熱負荷
のみを冷却する。海水熱交換器３で閉サイクルの冷却水
を冷却するには、海水取水口の海水２を海水ポンプ１で
汲み上げ熱交換器に送り込んで閉サイクルの水を冷却し
て自身温度が上がりそれを海水放出口から海へ逃がし除
熱をする。In an emergency, the emergency heat load main valve 6 is automatically opened, and the normal heat load main valve 7 is automatically closed to cool only the emergency heat load. In order to cool the closed-cycle cooling water with the seawater heat exchanger 3, the seawater 2 at the seawater intake is pumped up by the seawater pump 1 and sent to the heat exchanger to cool the closed-cycle water so that the temperature rises and the seawater is cooled. Release heat from the outlet to the sea to remove heat.
【００１３】系統連絡弁８は定期点検時２つのシステム
の片方のポンプ熱交換器を点検していて使用不可能の時
に残っている健全側の閉サイクル冷却水を点検側の通常
用熱負荷に送りプラント停止中でも運転される通常時負
荷（使用済燃料プール熱交換器、空気圧縮器、廃棄物処
理設備各種機器など）を冷却除熱するためにある。The system communication valve 8 checks one of the pump heat exchangers of the two systems at the time of the periodic inspection, and converts the remaining closed cycle cooling water on the healthy side into the normal heat load on the inspected side when the pump heat exchanger is not usable. This is for cooling and removing heat of normal loads (spent fuel pool heat exchangers, air compressors, various types of waste treatment equipment, etc.) that are operated even when the feed plant is stopped.
【００１４】ＴＣＷはＲＣＷより構成がシンプルであ
る。プラントの非常時にはタービンは自動的に停止して
しまうためＴＣＷにはＲＣＷにある非常用熱負荷はな
く、通常時熱負荷のみである。従ってシステムは 100％
容量の単一システムであり、海水ポンプ１を１セット海
水熱交換器３を１セット補機冷却水ポンプ９をセット配
管で結ぶだけである。The TCW has a simpler structure than the RCW. In a plant emergency, the turbine automatically shuts down, so the TCW has no emergency heat load in the RCW, only the normal heat load. So the system is 100%
It is a single system with a capacity, and only one set of seawater pump 1 is connected to one set of seawater heat exchanger 3 and one set of auxiliary cooling water pump 9 is connected by set piping.
【００１５】[0015]
【発明が解決しようとする課題】従来の技術では原子力
発電所内の補助機器を冷却あるいは除熱するため、前述
した様な設備を設けているが、これ等補助機器から取出
した熱は海水に廃棄している。この熱源は元々原子炉で
発生した熱なのである。僅かとはいえ原子炉で発生した
熱が直接タービンを駆動することに消費されずに廃棄さ
れるとなれば、熱の無駄使いであり、環境を悪化させる
ことになる。In the prior art, the above-mentioned equipment is provided for cooling or removing heat of auxiliary equipment in a nuclear power plant, but the heat extracted from these auxiliary equipment is discarded in seawater. doing. This heat source is the heat originally generated in the nuclear reactor. If the heat generated in the nuclear reactor is discarded without being consumed for directly driving the turbine, it is a waste of heat and deteriorates the environment.
【００１６】したがって、これ等の補助機器から海水へ
廃棄される熱はプラントの熱効率を低下させることに繋
がり、また、別の視点から見れば環境に放出する廃熱が
増加する課題にもなる。[0016] Therefore, the heat discarded from these auxiliary devices into seawater leads to a decrease in the thermal efficiency of the plant, and from another point of view, increases the waste heat released to the environment.
【００１７】本発明は上記課題を解決するためになされ
たもので、プラント運転中の廃熱を減少してプラントの
熱効率を高め、経済性を向上させるとともに環境問題を
軽減することができる原子力発電プラントの補機冷却水
システムを提供することにある。SUMMARY OF THE INVENTION The present invention has been made to solve the above-mentioned problems, and it is an object of the present invention to reduce waste heat during operation of a plant, increase the thermal efficiency of the plant, improve economic efficiency, and reduce environmental problems. An object of the present invention is to provide an auxiliary cooling water system for a plant.
【００１８】[0018]
【課題を解決するための手段】本発明は主復水器で凝縮
された復水を復水ポンプにより復水脱塩塔に導き、この
復水脱塩塔で浄化された復水をバイパスしてプラント運
転時用補機熱交換器に導き熱交換して冷却し、この冷却
された復水を前記復水脱塩器の下流側に接続した高圧復
水ポンプにより給水加熱器に導いて加熱し原子炉圧力容
器内へ冷却材として給水する系統からなり、前記プラン
ト運転時用補機熱交換器は海水熱交換器により熱交換さ
れる補機冷却系の非常時熱負荷と通常時熱負荷と同等の
流路配管係を有するものであることを特徴とする。According to the present invention, the condensate condensed in the main condenser is guided to a condensate demineralization tower by a condensate pump, and the condensate purified by the condensate demineralization tower is bypassed. The condensate is guided to an auxiliary heat exchanger for plant operation, heat exchanged and cooled, and the cooled condensate is guided to a feed water heater by a high-pressure condensate pump connected downstream of the condensate desalter and heated. The auxiliary heat exchanger for operation of the plant comprises an emergency heat load and an ordinary heat load of the auxiliary cooling system which is heat-exchanged by the seawater heat exchanger. It is characterized by having the same flow path piping system as that of (1).
【００１９】[0019]
【作用】原子炉補機冷却システムとタービン補機冷却シ
ステムを除熱するために海水の代りに復水器で凝縮し、
原子炉に給水する給水復水システムの復水を用いて再補
機冷却システムの廃熱を吸収する。この復水を原子炉に
給水することにより原子炉出力を軽くするかまたは、発
生する蒸気量を増やすことができ、結果として廃熱を回
収することができる。[Function] In order to remove heat from the reactor accessory cooling system and the turbine accessory cooling system, it is condensed by a condenser instead of seawater,
The condensate of the water supply condensing system for supplying water to the reactor is used to absorb the waste heat of the re-accessory cooling system. By supplying the condensed water to the reactor, the reactor output can be reduced or the amount of generated steam can be increased, and as a result, waste heat can be recovered.
【００２０】すなわち、主復水器で凝縮した復水を海水
の代りに用い除熱した熱を原子炉給水系に回収する。こ
の結果、プラントの熱効率は 0.5〜 0.6％高めることが
できる。また、運転中に機器の冷却水漏洩で海水中に放
射能を放出する可能性をより少くできる。That is, the condensed water condensed in the main condenser is used instead of seawater, and the heat removed is recovered to the reactor water supply system. As a result, the thermal efficiency of the plant can be increased by 0.5 to 0.6%. In addition, the possibility of releasing radioactivity into seawater due to leakage of cooling water from equipment during operation can be reduced.
【００２１】[0021]
【実施例】図１を参照しながら本発明に係る原子力発電
プラントの補機冷却水システムの一実施例を説明する。DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of an auxiliary cooling water system for a nuclear power plant according to the present invention will be described with reference to FIG.
【００２２】図１において符号１は海水ポンプで、この
海水ポンプ１は海水２を汲み上げて海水熱交換器３内を
循環させる。海水熱交換器３は非常用熱負荷４、プラン
ト運転時用補機熱交換器10および通常時熱負荷５に非常
用熱負荷元弁６、通常時熱負荷元弁７および補機熱交入
口弁21と補機熱交バイパス弁22を介して接続しているプ
ラント運転時補機熱交換器10は補機熱交復水元弁19を介
して復水脱塩器16の下流側に接続して復水脱塩器16で浄
化された復水の熱を冷却する。In FIG. 1, reference numeral 1 denotes a seawater pump. The seawater pump 1 pumps seawater 2 and circulates it in a seawater heat exchanger 3. The seawater heat exchanger 3 includes an emergency heat load 4, an auxiliary heat exchanger 10 for plant operation and a normal heat load 5, an emergency heat load main valve 6, a normal heat load main valve 7, and an auxiliary equipment heat exchange. The auxiliary heat exchanger 10 for plant operation, which is connected to the valve 21 via the auxiliary heat exchange bypass valve 22, is connected to the downstream side of the condensate desalinator 16 via the auxiliary heat exchange water supply valve 19. Then, the heat of the condensate purified by the condensate desalter 16 is cooled.
【００２３】復水脱塩器16の上流側は復水ポンプ17およ
び主復水器が接続され、また下流側には高圧復水ポンプ
15、中低圧給水加熱器14、高圧給水ポンプ13、高圧給水
加熱器12および原子炉圧力容器11が順次接続されてい
る。通常時熱負荷５の上下両端側配管と、プラント運転
時用補機熱交換器10と復水脱塩器16との間を接続する配
管にはそれぞれ別系統同一システム20への配管が接続さ
れている。A condensing pump 17 and a main condenser are connected to the upstream side of the condensing desalter 16, and a high-pressure condensing pump is connected to the downstream side.
15, a medium / low pressure feed water heater 14, a high pressure feed water pump 13, a high pressure feed water heater 12, and a reactor pressure vessel 11 are sequentially connected. The pipes connecting the upper and lower ends of the normal heat load 5 and the auxiliary heat exchanger 10 for plant operation and the condensate demineralizer 16 are connected to separate systems and the same system 20, respectively. ing.
【００２４】図中海水ポンプ１、海水２、海水熱交換器
３、非常用熱負荷４、非常用熱負荷元弁６、通常時熱負
荷元弁７、補機冷却水ポンプ９は図２で説明した従来設
備と同等である。また、通常時熱負荷５にはタービン発
電気関係の熱負荷を含めたものとする。図４は２つのシ
ステムのうちの片方のシステムのみを示している。つぎ
に本実施例中で使用する原子炉給水系について説明す
る。In the figure, the seawater pump 1, seawater 2, seawater heat exchanger 3, emergency heat load 4, emergency heat load main valve 6, normal heat load main valve 7, and auxiliary equipment cooling water pump 9 are shown in FIG. It is equivalent to the conventional equipment described. The normal heat load 5 includes a heat load related to turbine power generation. FIG. 4 shows only one of the two systems. Next, the reactor water supply system used in this embodiment will be described.
【００２５】主復水器18は主タービン発電機を回して仕
事を終えた蒸気を冷却凝縮して復水にする設備である。
ここで、蒸気は海水により冷却され35〜40℃程度の凝縮
水（復水とも言う）になる。ここの圧力は、絶対圧力0.
05kg／cm2 ・ａ程度であり、これを復水ポンプ17を通し
て昇圧する。その後、復水脱塩器16を通して水質を整え
る。ここでの水温は約40℃程度である。The main condenser 18 is a facility for turning the main turbine generator to cool and condense the steam that has completed its work to condense it.
Here, the steam is cooled by seawater to form condensed water (also called condensate) at about 35 to 40 ° C. The pressure here is absolute pressure 0.
05kg / cm 2 The pressure is about a, and the pressure is increased through the condensate pump 17. Thereafter, the water quality is adjusted through the condensate desalter 16. The water temperature here is about 40 ° C.
【００２６】この水を高圧復水ポンプ15でさらに昇圧
し、中、低圧給水加熱器14を通して原子炉に給水する前
に予熱を行う。更に給水ポンプ13で原子炉圧力容器11内
から圧力を上げ高圧給水加熱器12を通して 210℃程度迄
予熱して原子炉圧力容器11内に注入し、炉心で加熱して
蒸気を発生させる。このような原子炉給水系では主復水
器18で約35〜40℃0.05kg／cm2 ・ａの復水となった給水
を約 210℃80kg／cm2 ・ａまでの温度と圧力に昇温・昇
圧する。本実施例では復水脱塩器16を流出した約40℃10
kg／cm2 ・ａ程度の復水を海水２の代りの除熱媒体とし
て用いることにある。The pressure of the water is further increased by a high-pressure condensate pump 15, and preheating is performed before water is supplied to the reactor through a medium- and low-pressure feedwater heater 14. Further, the pressure is increased from the inside of the reactor pressure vessel 11 by the feed water pump 13, preheated to about 210 ° C. through the high pressure feed water heater 12, injected into the reactor pressure vessel 11, and heated in the reactor core to generate steam. In such a reactor water supply system, the main condenser 18 has a temperature of about 35-40 ° C. and 0.05 kg / cm 2. ・ The supply water that has been condensed is about 210 ° C 80kg / cm 2 ・ Temperature and pressure are increased to the temperature and pressure up to a. In the present embodiment, approximately 40 ° C. 10
kg / cm 2 The use of condensate of degree a as a heat removal medium instead of seawater 2.
【００２７】図１中の補機冷却システムでは図２に対し
プラント運転時用補機熱交換器10とその運用に係わる補
機熱交復水元弁19、補機熱交入口弁21および補機熱交バ
イパス弁22が加わったものである。つぎにこのシステム
の運用を記述する。In the accessory cooling system in FIG. 1, the auxiliary equipment heat exchanger 10 for plant operation and the auxiliary equipment heat exchange water supply valve 19, the auxiliary equipment heat exchange inlet valve 21 and the auxiliary equipment The heat exchange bypass valve 22 is added. Next, the operation of this system will be described.
【００２８】プラント通常運転（発電）中のシステム運
用は次の様になる。すなわち、補機冷却水ポンプ９で昇
圧された閉サイクル冷却水は海水熱交換器３に流入す
る。この海水熱交換器３に関係する海水ポンプ１は停止
している。従って、海水熱交換器３を通っても閉サイク
ルで冷却水は除熱冷却されない。次いで、通常用熱負荷
元弁７を通る。この時非常用熱負荷元弁６は閉じてい
る。冷却水は補機熱交入口弁21を通りプラント運転時用
補機熱交換器10に入る。補機熱交バイパス弁22は閉じて
いる。The system operation during the normal operation (power generation) of the plant is as follows. That is, the closed cycle cooling water pressurized by the auxiliary cooling water pump 9 flows into the seawater heat exchanger 3. The seawater pump 1 related to the seawater heat exchanger 3 is stopped. Therefore, even if the cooling water passes through the seawater heat exchanger 3, the cooling water is not heat-removed and cooled in the closed cycle. Next, it passes through the normal heat load main valve 7. At this time, the emergency heat load main valve 6 is closed. The cooling water passes through the auxiliary equipment heat exchange inlet valve 21 and enters the auxiliary heat exchanger 10 for plant operation. The auxiliary equipment heat exchange bypass valve 22 is closed.
【００２９】前記熱交換器10で原子炉給水系の水により
冷却され通常時熱負荷５に閉サイクル冷却水が供給さ
れ、約40℃の冷却水で各負荷を冷却する。各負荷を除熱
して流出した閉サイクル冷却水は補機冷却水ポンプ９に
戻る。The closed-cycle cooling water is supplied to the normal heat load 5 by cooling with the water in the reactor water supply system in the heat exchanger 10, and each load is cooled by cooling water of about 40 ° C. The closed cycle cooling water flowing out after removing the heat of each load returns to the auxiliary equipment cooling water pump 9.
【００３０】一方、熱交換器10を冷却した原子炉給水系
の水は予熱され、補機熱交復水弁19（戻り側）を通って
原子炉給水系の合流する。この過程で各熱負荷から除熱
された熱は原子炉給水を予熱することに使われ回収され
る。従って、海水に熱を廃棄することにならない。On the other hand, the water in the reactor water supply system which has cooled the heat exchanger 10 is preheated and merges with the reactor water supply system through the auxiliary heat exchange water valve 19 (return side). The heat removed from each heat load during this process is used to preheat the reactor feedwater and is recovered. Therefore, no heat is wasted in seawater.
【００３１】プラントの非常時は自動的に次の様な運用
となるようにインターロックを設定する。非常時には海
水ポンプ１は自動起動する。従って、ここで海水熱交換
器３は除熱冷却が可能となる。次いで非常用熱負荷元弁
６が開き、通常用熱負荷元弁７が閉じる。この変化によ
り原子炉の崩壊熱は非常用熱負荷４を流れる閉サイクル
冷却水により海水熱交換器３に運ばれ海水２に廃棄され
る。In an emergency of the plant, an interlock is set so that the following operation is automatically performed. In an emergency, the seawater pump 1 starts automatically. Therefore, here, the seawater heat exchanger 3 can perform heat removal cooling. Next, the emergency heat load main valve 6 is opened, and the normal heat load main valve 7 is closed. Due to this change, the decay heat of the reactor is carried to the seawater heat exchanger 3 by the closed cycle cooling water flowing through the emergency heat load 4 and is discarded in the seawater 2.
【００３２】プラントが停止して定期点検に入る場合は
原子炉給水系も停止するし、また予熱も意味がないので
補機熱交バイパス弁22を開き、補機熱交入口弁21を閉
じ、同じ出口弁も閉じることによりプラント運転時用補
機熱交換器10の開放点検も可能となる。When the plant is shut down and the periodic inspection is started, the reactor water supply system is also stopped, and the preheating is not meaningful, so the auxiliary equipment heat exchange bypass valve 22 is opened, and the auxiliary equipment heat exchange inlet valve 21 is closed. By closing the same outlet valve, open inspection of the auxiliary heat exchanger 10 for plant operation is also possible.
【００３３】このシステムは、従来のＲＣＷと同様にほ
ぼ同一なシステムの系統の合体で構成され、閉サイクル
冷却水系の別系統同一システム20により結合される。原
子炉給水系も主系列から枝分れして補機熱交換器10に向
かう途中で２系統のシステムに流れを分割させる。This system is composed of a combination of almost the same system as the conventional RCW, and is connected by another same system 20 of the closed cycle cooling water system. The reactor water supply system is also branched from the main system and splits the flow into two systems on the way to the auxiliary heat exchanger 10.
【００３４】[0034]
【発明の効果】本発明によれば、プラント運転中はプラ
ント内補機から廃棄しなければならない熱を原子炉給水
を加熱するために使用することによって約 1.7％の熱が
回収される。その結果、約 0.5〜 0.6％程プラント熱効
率が向上する。すなわち経済性が向上し、その結果とし
て環境へ廃棄する熱が相対的に減少する。According to the present invention, approximately 1.7% of the heat is recovered by using the heat that must be discarded from the auxiliary equipment during the operation of the plant to heat the reactor feedwater. As a result, the plant thermal efficiency is improved by about 0.5 to 0.6%. That is, the economy is improved, and as a result, the amount of heat that is discarded into the environment is relatively reduced.
【００３５】また、プラント運転時用補機熱交換器の３
次側の冷却水がプラント運転中は原子炉給水となるた
め、各負荷の何れかで一次水の漏洩で放射能が洩れ且つ
熱交換器のチューブが破損しても放射能を環境に放出す
る恐れはない。Further, the auxiliary heat exchanger for plant operation 3
Since the cooling water on the secondary side becomes the reactor water supply during plant operation, even if the primary water leaks at any of the loads and the radioactivity leaks and the tubes of the heat exchanger break, the radioactivity is released to the environment There is no fear.
【図１】本発明に係る原子力発電プラントの補機冷却水
システムの一実施例を示す系統図。FIG. 1 is a system diagram showing an embodiment of an auxiliary cooling water system for a nuclear power plant according to the present invention.
【図２】従来の原子力発電プラントの補機冷却水システ
ムを示す系統図。FIG. 2 is a system diagram showing a conventional auxiliary cooling water system of a nuclear power plant.
１…海水ポンプ、２…海水、３…海水熱交換器、４…非
常用熱負荷、５…通常時熱負荷、６…非常用熱負荷元
弁、７…通常時熱負荷元弁、８…系統連絡弁、９…補機
冷却水ポンプ、10…プラント運転時用補機熱交換器、11
…原子炉圧力容器、12…高圧給水加熱器、13…高圧給水
ポンプ、14…中、低圧給水加熱器、15…高圧復水ポン
プ、16…復水脱塩器、17…復水ポンプ、18…主復水器、
19…補機熱交復水元弁、20…別系統同一システム、21…
補機熱交入口弁、22…補機熱交バイパス弁。DESCRIPTION OF SYMBOLS 1 ... Seawater pump, 2 ... Seawater, 3 ... Seawater heat exchanger, 4 ... Emergency heat load, 5 ... Normal heat load, 6 ... Emergency heat load main valve, 7 ... Normal heat load main valve, 8 ... System communication valve, 9: auxiliary equipment cooling water pump, 10: auxiliary equipment heat exchanger for plant operation, 11
... Reactor pressure vessel, 12 ... High pressure feed water heater, 13 ... High pressure feed water pump, 14 ... Medium and low pressure feed water heater, 15 ... High pressure condensate pump, 16 ... Condensate desalinator, 17 ... Condensate pump, 18 ... the main condenser,
19: Auxiliary heat exchange water main valve, 20: Separate system, 21 ...
Auxiliary heat exchange inlet valve, 22 ... Auxiliary heat exchange bypass valve.
───────────────────────────────────────────────────── フロントページの続き (58)調査した分野(Int.Cl.7，ＤＢ名) G21D 1/00 G21C 19/30 ──────────────────────────────────────────────────続 き Continued on front page (58) Field surveyed (Int. Cl. 7 , DB name) G21D 1/00 G21C 19/30
Claims (1)
により復水脱塩塔に導き、この復水脱塩塔で浄化された
復水をバイパスしてプラント運転時用補機熱交換器に導
き熱交換して冷却し、この冷却された復水を前記復水脱
塩器の下流側に接続した高圧復水ポンプにより給水加熱
器に導いて加熱し原子炉圧力容器内へ冷却材として給水
する系統からなり、前記プラント運転時用補機熱交換器
は海水熱交換器により熱交換される補機冷却系の非常時
熱負荷と通常時熱負荷と同等の流路配管系を有するもの
であることを特徴とする原子炉発電プラントの補機冷却
水システム。1. A condensate condensed in a main condenser is guided to a condensate demineralization tower by a condensate pump, and the condensate purified by the condensate demineralization tower is bypassed and an auxiliary machine for plant operation is provided. It is led to a heat exchanger to exchange heat and cool, and the cooled condensate is guided to a feed water heater by a high-pressure condensate pump connected to the downstream side of the condensate desalter and heated to enter the reactor pressure vessel. The auxiliary heat exchanger for plant operation comprises a system for supplying water as a coolant, and the auxiliary heat exchanger for operating the plant is a flow pipe system equivalent to the emergency heat load and the normal heat load of the auxiliary cooling system in which heat is exchanged by the seawater heat exchanger. Auxiliary cooling water system for a nuclear power plant, characterized by having:
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP3307509A JP3068288B2 (en) | 1991-11-22 | 1991-11-22 | Auxiliary cooling water system for nuclear power plants |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP3307509A JP3068288B2 (en) | 1991-11-22 | 1991-11-22 | Auxiliary cooling water system for nuclear power plants |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPH05142389A JPH05142389A (en) | 1993-06-08 |
| JP3068288B2 true JP3068288B2 (en) | 2000-07-24 |
Family
ID=17969936
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP3307509A Expired - Lifetime JP3068288B2 (en) | 1991-11-22 | 1991-11-22 | Auxiliary cooling water system for nuclear power plants |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JP3068288B2 (en) |
-
1991
- 1991-11-22 JP JP3307509A patent/JP3068288B2/en not_active Expired - Lifetime
Also Published As
| Publication number | Publication date |
|---|---|
| JPH05142389A (en) | 1993-06-08 |
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