JPS6153676B2 - - Google Patents
Info
- Publication number
- JPS6153676B2 JPS6153676B2 JP53007511A JP751178A JPS6153676B2 JP S6153676 B2 JPS6153676 B2 JP S6153676B2 JP 53007511 A JP53007511 A JP 53007511A JP 751178 A JP751178 A JP 751178A JP S6153676 B2 JPS6153676 B2 JP S6153676B2
- Authority
- JP
- Japan
- Prior art keywords
- control
- core
- neutron
- scram
- circulation type
- 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
- 239000000126 substance Substances 0.000 claims description 30
- 239000000446 fuel Substances 0.000 claims description 26
- 238000009826 distribution Methods 0.000 claims description 11
- 230000000712 assembly Effects 0.000 claims description 10
- 238000000429 assembly Methods 0.000 claims description 10
- 230000004907 flux Effects 0.000 claims description 7
- 239000000243 solution Substances 0.000 description 22
- KGBXLFKZBHKPEV-UHFFFAOYSA-N boric acid Chemical compound OB(O)O KGBXLFKZBHKPEV-UHFFFAOYSA-N 0.000 description 10
- 239000004327 boric acid Substances 0.000 description 10
- 239000002826 coolant Substances 0.000 description 5
- 230000037431 insertion Effects 0.000 description 5
- 238000003780 insertion Methods 0.000 description 5
- 150000002500 ions Chemical class 0.000 description 5
- 239000000463 material Substances 0.000 description 5
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 5
- 230000009257 reactivity Effects 0.000 description 4
- 238000010992 reflux Methods 0.000 description 4
- 230000001172 regenerating effect Effects 0.000 description 4
- 238000009835 boiling Methods 0.000 description 3
- 238000006243 chemical reaction Methods 0.000 description 3
- 230000001276 controlling effect Effects 0.000 description 3
- ZOXJGFHDIHLPTG-UHFFFAOYSA-N Boron Chemical compound [B] ZOXJGFHDIHLPTG-UHFFFAOYSA-N 0.000 description 2
- NBIIXXVUZAFLBC-UHFFFAOYSA-N Phosphoric acid Chemical compound OP(O)(O)=O NBIIXXVUZAFLBC-UHFFFAOYSA-N 0.000 description 2
- QAOWNCQODCNURD-UHFFFAOYSA-N Sulfuric acid Chemical compound OS(O)(=O)=O QAOWNCQODCNURD-UHFFFAOYSA-N 0.000 description 2
- 239000007864 aqueous solution Substances 0.000 description 2
- 238000005253 cladding Methods 0.000 description 2
- 239000007788 liquid Substances 0.000 description 2
- 239000000203 mixture Substances 0.000 description 2
- 231100000614 poison Toxicity 0.000 description 2
- 239000000843 powder Substances 0.000 description 2
- NWUYHJFMYQTDRP-UHFFFAOYSA-N 1,2-bis(ethenyl)benzene;1-ethenyl-2-ethylbenzene;styrene Chemical compound C=CC1=CC=CC=C1.CCC1=CC=CC=C1C=C.C=CC1=CC=CC=C1C=C NWUYHJFMYQTDRP-UHFFFAOYSA-N 0.000 description 1
- 229910000147 aluminium phosphate Inorganic materials 0.000 description 1
- 150000001450 anions Chemical class 0.000 description 1
- 229910052796 boron Inorganic materials 0.000 description 1
- 238000002485 combustion reaction Methods 0.000 description 1
- 238000005336 cracking Methods 0.000 description 1
- 238000010586 diagram Methods 0.000 description 1
- 238000007599 discharging Methods 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 239000003792 electrolyte Substances 0.000 description 1
- 229910052735 hafnium Inorganic materials 0.000 description 1
- VBJZVLUMGGDVMO-UHFFFAOYSA-N hafnium atom Chemical compound [Hf] VBJZVLUMGGDVMO-UHFFFAOYSA-N 0.000 description 1
- 238000000265 homogenisation Methods 0.000 description 1
- 239000003456 ion exchange resin Substances 0.000 description 1
- 229920003303 ion-exchange polymer Polymers 0.000 description 1
- 230000007935 neutral effect Effects 0.000 description 1
- 239000008188 pellet Substances 0.000 description 1
- 239000002574 poison Substances 0.000 description 1
- 230000007096 poisonous effect Effects 0.000 description 1
- 230000001105 regulatory effect Effects 0.000 description 1
- 239000011347 resin Substances 0.000 description 1
- 229920005989 resin Polymers 0.000 description 1
- 238000000926 separation method Methods 0.000 description 1
- 229910052709 silver Inorganic materials 0.000 description 1
- 239000004332 silver Substances 0.000 description 1
- 229910001220 stainless steel Inorganic materials 0.000 description 1
- 239000010935 stainless steel Substances 0.000 description 1
- 239000002699 waste material Substances 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
- Y02E30/30—Nuclear fission reactors
Landscapes
- Monitoring And Testing Of Nuclear Reactors (AREA)
Description
【発明の詳細な説明】
〔発明の技術分野〕
本発明は、例えば、沸騰水型原子炉における炉
心出力を平坦化する中性子束制御装置に関する。DETAILED DESCRIPTION OF THE INVENTION [Technical Field of the Invention] The present invention relates to a neutron flux control device for flattening core output in, for example, a boiling water nuclear reactor.
従来、この種の中性子束制御は、通常運転時、
例えば、B4Cの粉末による中性子吸収物質を充填
した多数のポイズンチユーブを、断面が十字形に
配列し、さらに、これをステンレス鋼製のシース
で整形した制御棒を使用している。しかも、この
制御棒は、炉心に４本一組の燃料集合体に対し、
一本の割合で装填されており、この制御棒は炉心
の下部から挿入されるようになつている。
Conventionally, this type of neutron flux control has been applied during normal operation.
For example, a control rod is used in which a large number of poison tubes filled with a neutron-absorbing substance made of B 4 C powder are arranged in a cruciform cross section and shaped with a stainless steel sheath. Moreover, these control rods are used for each set of four fuel assemblies in the reactor core.
This control rod is inserted from the bottom of the reactor core.
一方、この種の原子炉における炉心出力の半径
方向分布及び軸方向の各分布は、深挿入制御棒と
浅挿入制御棒を組合せて、いわゆる1/4又は1/8対
称の制御棒挿入のパターンを作つて制御してい
る。この制御棒挿入のパターンは制御棒周囲の燃
料集合体の燃焼度を均一化するために、一定期間
経過後、深挿入と浅挿入の各制御棒の位置が交換
される。 On the other hand, the radial and axial distributions of core power in this type of reactor are based on a so-called 1/4 or 1/8 symmetrical control rod insertion pattern, which combines deep insertion control rods and shallow insertion control rods. is created and controlled. In this control rod insertion pattern, in order to equalize the burnup of the fuel assembly around the control rods, the positions of the deeply inserted and shallowly inserted control rods are exchanged after a certain period of time.
このように、制御棒操作において、深挿入され
ている制御棒の周囲の燃料集合体と浅挿入されて
いる制御棒の周囲の燃料集合体の軸方向出力分布
曲線Ａ，Ｂは、第１図ａ，ｂに示されるように大
きく変化し、他方、半径方向の出力分布曲線Ｃ
は、第１図Ｃに示されるように、不均一なものと
なり、熱的制限（限界）に近い燃料集合体と熱的
制限に充分余裕のある燃料集合体が混在する結果
となり、炉心の出力密度を低下させるおそれがあ
る。 In this way, during control rod operation, the axial power distribution curves A and B of the fuel assemblies around the deeply inserted control rods and the fuel assemblies around the shallowly inserted control rods are as shown in Figure 1. a,b vary greatly, while the radial power distribution curve C
As shown in Figure 1C, the output becomes non-uniform, resulting in a mixture of fuel assemblies that are close to the thermal limit (limit) and fuel assemblies that have sufficient margin for the thermal limit, resulting in a decrease in the core output. There is a risk of reducing the density.
又、制御棒引抜操作時及び制御棒挿入パターン
交換時には、燃料集合体における燃料ペレツトの
熱膨張による被覆管の機械的干渉作用（PCMI）
つまり、被覆管の歪みや亀裂による損傷を防止す
るために、制御棒の引抜き速度の制限をして制御
棒運転作業（PCIOMR）が行なわれている。 In addition, during control rod withdrawal operations and control rod insertion pattern exchanges, mechanical interference (PCMI) of the cladding due to thermal expansion of fuel pellets in the fuel assembly occurs.
In other words, in order to prevent damage due to distortion or cracking of the cladding, control rod operation operations (PCIOMR) are performed with limits on the control rod withdrawal speed.
このような運転制限は、約１週間から１月間の
日時を費すばかりでなく、原子炉の運転出力及び
出力分布の調整目的や非常時におけるスクラム用
としての機能を併せ持つているため、制御棒の反
応度値を大きく取れず、スクラム時の特性上、不
都合を生じるおそれがあつた。 Such operational restrictions not only consume time and time from approximately one week to one month, but also have the purpose of adjusting the operating output and power distribution of the reactor, as well as the function of scramming in an emergency, so control rod It was not possible to obtain a large reactivity value, and there was a risk of inconvenience due to the characteristics during scram.
本発明は、上述した事情に鑑みてなされたもの
であつて、炉心の出力分布を平坦化して、出力密
度の向上を図り、通常運転時の制御棒操作をする
ことなく、炉心出力を平坦化するようにすると共
に、炉心径方向出力分布を制御するようにしたこ
とを目的とする中性子制御装置を提供するもので
ある。
The present invention was made in view of the above-mentioned circumstances, and aims to improve the power density by flattening the power distribution of the core, thereby flattening the core power without operating control rods during normal operation. The object of the present invention is to provide a neutron control device which aims at controlling the power distribution in the radial direction of the reactor core.
本発明は、圧力容器の炉心に４本１組の各燃料
集合体を保持するようにした各燃料支持部材を設
け、この各燃料支持部材の中央部に中性子制御物
質溶液の環流形制御管を挿脱自在に設け、この環
流形制御管及び各燃料集合体を基本単位体として
構成し、さらに、この基本単位体を４単位として
ユニツト体を構成し、このユニツト体の中央部に
スクラム用制御棒を挿脱自在に設け、上記圧力容
器の炉心に中性子制御物質溶液の濃度をそれぞれ
異にした環流形制御管による各制御管群を上記炉
心の中心部から外方に向つて同心的に配設して炉
心径方向の出力を制御するように構成したもので
ある。
The present invention provides fuel support members each holding a set of four fuel assemblies in the core of a pressure vessel, and a circulation type control pipe for a neutron control substance solution in the center of each fuel support member. This circulation type control pipe and each fuel assembly are provided as a basic unit, and this basic unit is made into four units to form a unit, and a scram control is installed in the center of this unit. A rod is provided in a removable manner, and each group of control tubes consisting of circulation type control tubes each having a different concentration of neutron control substance solution is arranged concentrically outward from the center of the reactor core in the core of the pressure vessel. It is configured to control the output in the radial direction of the core.
以下、本発明を図示の一実施例について説明す
る。
Hereinafter, the present invention will be described with reference to an illustrated embodiment.
第２図乃至第６図において、符号１は、沸騰水
型原子炉における圧力容器であつて、この圧力容
器１の一側には、冷却材の供給口１ａが設けられ
ており、この圧力容器１の上部には、冷却材の吐
出口（図示されず）が付設されている。又、上記
圧力容器１の下部には炉心支持板２が水平に設け
られており、この炉心支持板２の上位に位置する
圧力容器１のシユラウドヘツド３には、上部格子
板４が設けられている。さらに、上記炉心支持板
２には、第２図及び第４図に示されるように、各
燃料支持部材（燃料支持金具ともいう）５が嵌装
されており、この各燃料支持部材５には、４本一
組の各燃料集合体６がそれぞれ装填されている。
さらに、又、この各燃料支持部材５の中央部に
は、中性子制御物質溶液を環流し得るようにした
環流形制御管７が着脱自在に挿入されており、こ
の環流形制御管７および燃料集合体６は基本単位
体を構成している。又、上記基本単位体は４
単位としてユニツト体を構成しており、このユ
ニツト体の中央部にはスクラム用制御棒８が挿
脱自在に設けられている（第４図参照）。 2 to 6, reference numeral 1 denotes a pressure vessel in a boiling water reactor, and one side of this pressure vessel 1 is provided with a coolant supply port 1a. A coolant discharge port (not shown) is attached to the upper part of 1. Further, a core support plate 2 is provided horizontally at the bottom of the pressure vessel 1, and an upper lattice plate 4 is provided at the shroud head 3 of the pressure vessel 1 located above the core support plate 2. . Furthermore, as shown in FIGS. 2 and 4, fuel support members (also referred to as fuel support fittings) 5 are fitted into the core support plate 2. , a set of four fuel assemblies 6 are respectively loaded.
Furthermore, a recirculation type control pipe 7 that can recirculate the neutron control substance solution is detachably inserted into the center of each fuel support member 5, and the recirculation type control pipe 7 and the fuel assembly The body 6 constitutes a basic unit body. Also, the above basic unit is 4
It constitutes a unit body, and a scram control rod 8 is provided in the center of this unit body so as to be freely insertable and removable (see Fig. 4).
一方、上記環流形制御管７は、第５図に示され
るように、断面が十字形をしたシース体７ａ内に
多数の制御細管７ｂを列設し、上記シース体７ａ
と一体をなす下端部開口部７ｃに、例えばホウ酸
水のような中性子制御物質溶液を給排する供給口
７ｄ及び排出口７ｅが付設されている。 On the other hand, as shown in FIG. 5, the circulation type control tube 7 has a large number of control thin tubes 7b arranged in a sheath body 7a having a cross-shaped cross section.
A supply port 7d and a discharge port 7e for supplying and discharging a neutron control substance solution such as a boric acid solution are attached to the lower end opening 7c which is integral with the lower end opening 7c.
他方、上記圧力容器１内には、第３図の横断面
図で示されるように、中性子制御物質溶液の濃度
をそれぞれ異にした環流形制御管７による各制御
管群a，b，cが炉心の中心部から外方向に
向つて同心的に配設されており、これにより炉心
出力を平坦化して制御するようになつている。 On the other hand , inside the pressure vessel 1, as shown in the cross-sectional view in FIG . They are arranged concentrically outward from the center of the core, thereby flattening and controlling the core output.
即ち、上記各制御管群a，b，cの中性子
制御物質溶液の濃度は、炉心の各領域のＫ∞を平
坦化するために、b＞a＞c又はc＞a
b＞d（図示されず）の値になるようになつてい
る。従つて、これにより調節して炉心出力及び出
力分布を制御するようになつている。 That is, the concentration of the neutron control substance solution in each of the control tube groups a , b , and c is such that b > a > c or c > a in order to flatten K ∞ in each region of the reactor core.
b > d (not shown). Accordingly, this allows adjustment to control the core power and power distribution.
なお、上記各環流形制御管７は、第２図に示さ
れるように、制御管延長管７ｆを介して圧力容器
１の底部に延設されており、又、上記各スクラム
用制御棒８は、第２図に示されるように、各制御
棒駆動装置９によつて、スクラム時、炉心に挿入
されるようにして設けられている。 As shown in FIG. 2, each of the circulation type control pipes 7 is extended to the bottom of the pressure vessel 1 via a control pipe extension pipe 7f, and each of the scram control rods 8 is As shown in FIG. 2, each control rod drive device 9 is inserted into the reactor core during scram.
一方、上記各環流形制御管７内を環流する中性
子制御物質溶液は、例えばホウ酸水を使用する。
しかし、この種の水溶液は、一般に弱電解質又は
イオン移動度が小さいため、電気化学的に連続し
て濃度測定が困難である。 On the other hand, the neutron control substance solution that circulates in each of the circulation type control tubes 7 is, for example, boric acid water.
However, since this type of aqueous solution is generally a weak electrolyte or has low ion mobility, it is difficult to continuously measure the concentration electrochemically.
そこで、本発明は、ホウ酸水溶液内に、ホウ酸
濃度測定のために、陰イオン樹脂に対し、前記中
性子制御物質と類似の挙動を示し、かつ電離度又
はイオン移動度の大きなモニター物質を一定比添
加して混合する。この中性子制御物質のホウ酸に
対しては、例えばリン酸又は硫酸が望ましい。 Therefore, in order to measure boric acid concentration, the present invention provides a constant amount of a monitor substance that exhibits similar behavior to the neutron control substance and has a large degree of ionization or ion mobility with respect to an anion resin in a boric acid aqueous solution. Add and mix. For example, phosphoric acid or sulfuric acid is preferable for the neutron control substance boric acid.
次に、中性子制御物質溶液の循環系は、第６図
に示されるように、中性子制御物質溶液の濃度を
異にした各制御管群a，b，cにそれぞれ独
立して供給されるようになつている。 Next, as shown in Figure 6, the circulation system for the neutron control substance solution is such that the neutron control substance solution is supplied independently to each of the control tube groups a , b , and c , each having a different concentration. It's summery.
即ち、上記循環系を制御群aについて説明す
ると、下記のようにして構成されている。第６図
において、符号７は、各環流形制御管であつて、
この各環流形制御管７は中性子制御物質溶液を循
環ポンプ１０によつて供給され、この環流液は圧
力調整タンク１１を有する非再生熱交換器１２へ
環流されるようになつている。又、上記中性子制
御物質溶液は再生熱交換器１３によつて再生され
るけれども、この再生熱交換器１３は非再生熱交
換器１４を通して浄化器１５、脱ホウ素イオン交
換器１６に接続されており、この脱ホウ素イオン
交換器１６は化学体積制御タンク１７を介して充
填ポンプ１８に連結している。従つて、上記中性
子制御物質溶液は常に一定の濃度で制御管群a
に供給されるようになつている。 That is, the above-mentioned circulatory system is configured as follows when the control group a is explained. In FIG. 6, reference numeral 7 indicates each circulation type control pipe,
Each reflux type control pipe 7 is supplied with a neutron control substance solution by a circulation pump 10, and the reflux liquid is refluxed to a non-regenerative heat exchanger 12 having a pressure regulating tank 11. Furthermore, although the neutron control substance solution is regenerated by the regenerative heat exchanger 13, this regenerative heat exchanger 13 is connected to a purifier 15 and a deboronizing ion exchanger 16 through a non-regenerative heat exchanger 14. , this deboronizing ion exchanger 16 is connected to a filling pump 18 via a chemical volume control tank 17 . Therefore, the above neutron control substance solution is always kept at a constant concentration in the control tube group a.
It is now being supplied to
なお、上記充填ポンプ１８には、純水タンク１
９及びホウ酸タンク２０が連結されており、上記
化学体積制御タンク１７には廃液タンク２１が接
続されている。 Note that the filling pump 18 includes a pure water tank 1.
9 and a boric acid tank 20 are connected, and a waste liquid tank 21 is connected to the chemical volume control tank 17.
なお、制御管群aの循環系は上述した構成を
もつものであるが、他の制御管群b，cの循環
系も同様の構成をしている。 Note that although the circulatory system of the control pipe group a has the above-mentioned configuration, the circulatory systems of the other control pipe groups b and c also have similar configurations.
従つて、本発明では、通常運転時、原則的には
スクラム用制御棒８は圧力容器１の炉心から全部
引抜かれた状態にある。又、原子炉が出力状態か
らすみやかに、高温停止する場合には、上記スク
ラム用制御棒８は全数、急速度で挿入される。さ
らに、冷態停止時には環流形制御管７内を環流す
るホウ酸濃度を高める。 Therefore, in the present invention, in principle, the scram control rods 8 are completely withdrawn from the core of the pressure vessel 1 during normal operation. Further, when the reactor is quickly brought to a high-temperature shutdown from an output state, all of the scram control rods 8 are inserted rapidly. Further, during the cold stop, the concentration of boric acid circulating in the circulation type control pipe 7 is increased.
又一方、冷態停止時から起動するには、先ず、
スクラム用制御棒８を全数引抜き、次いで、環流
形制御管７内のホウ酸濃度を調整して臨界にす
る。 On the other hand, to start from a cold stop, first,
All scram control rods 8 are pulled out, and then the concentration of boric acid in the circulation type control pipe 7 is adjusted to make it critical.
このように、上記環流形制御管７とスクラム用
制御棒８の機能的な分離によつて、スクラム時
に、環流形制御管７のもつ反応度価値に更に、ス
クラム用制御棒８の反応度が加わるので、スクラ
ム特性が良好となり、炉停止余裕が増大する。 In this way, due to the functional separation of the reflux type control pipe 7 and the scram control rod 8, the reactivity value of the reflux type control pipe 7 and the reactivity value of the scram control rod 8 are increased during scram. This improves the scram characteristics and increases the margin for reactor shutdown.
因に、既に提案されているこの種の中性束制御
装置では、制御棒が出力制御機能と停止機能を併
せもつているため、出力制御上は制御棒駆動時の
隣接燃料集合体の燃料棒の線出力密度の増加率を
抑えるから、制御棒の反応度価値を大きくするに
も制限があつた。又、制御棒が通常運転中も炉心
に挿入されている関係上、ホウ素のnd反応が進
みB4C管のHe圧の上昇やB4C粉末の体積変化を考
慮せねばならなかつた。 Incidentally, in this type of neutral flux control device that has already been proposed, the control rods have both an output control function and a stop function, so in terms of output control, when the control rods are driven, the fuel rods of adjacent fuel assemblies There was also a limit to increasing the reactivity value of control rods because the rate of increase in linear power density was suppressed. In addition, since the control rods are inserted into the reactor core even during normal operation, it was necessary to take into account the increase in He pressure in the B 4 C tube and the change in the volume of the B 4 C powder due to the progress of the nd reaction of boron.
本発明では、スクラム用制御棒８は、通常運転
時は、炉心に挿入されていないので、停止機能の
みを考慮して、ホウ素の粉末の密度を上げたり、
B4Cのペレツト化を行なうことができる。又、ガ
スプレナムも小さくできる。さらに、制御棒価値
を高める目的で熱外中性子等を吸収する銀、ハフ
ニウム等の毒物をも混合できる。 In the present invention, since the scram control rods 8 are not inserted into the reactor core during normal operation, the density of the boron powder is increased by considering only the shutdown function.
B 4 C can be pelletized. Also, the gas plenum can be made smaller. Furthermore, poisonous substances such as silver and hafnium, which absorb epithermal neutrons, etc., can be mixed in to increase the value of the control rod.
通常運転時には、スクラム用制御棒８は全部引
抜かれているから、炉心出力制御は、制御管７内
を流れるホウ酸溶液の濃度を調整し、燃焼が進む
ことによるＫ∞の低下等ゆるやかな反応度変化に
対応する。又、冷却材の炉心流量、又は冷却材の
炉心入口温度の調整はホウ酸溶液の濃度によつて
も行なわれる。 During normal operation, all of the scram control rods 8 are withdrawn, so the core power control is performed by adjusting the concentration of the boric acid solution flowing inside the control tube 7, resulting in gradual reactions such as a decrease in K ∞ as combustion progresses. Responds to temperature changes. Further, the core flow rate of the coolant or the core inlet temperature of the coolant is also adjusted by the concentration of the boric acid solution.
他方、炉心半径方向の出力分布は、第３図に示
されるように、同心的に配設された各制御管群
a，b，cによつて燃料バンドルのＫ∞を略同
一にるようになつている。従つて、これにより燃
料バンドルの最適な燃焼度を得られ、しかもバン
ドルピーキングの低下、炉心出力密度の向上、炉
心径方向の出力分布の平坦化及び炉心の均質を行
なうことができる。 On the other hand, as shown in Figure 3, the power distribution in the radial direction of the core is
K ∞ of the fuel bundle is made to be approximately the same by a , b , and c . Therefore, this makes it possible to obtain the optimum burnup of the fuel bundle, and also to reduce bundle peaking, improve the core power density, flatten the power distribution in the radial direction of the core, and make the core homogeneous.
従つて、本発明は、これまで沸騰水型原子炉で
行なわれてきた制御棒パターンの変換作業時、前
もつて行なわれる制御棒パターンリーチ計算が不
要となり、運転が簡略化されて容易になる。又本
発明は既に提案されているこの種の装置に較べ
て、炉心の平坦化が均質なものとなるので、中性
子モニター数を少なくして中性子計装を簡単にす
ることができるようになる。 Therefore, the present invention eliminates the need for control rod pattern reach calculations that are previously performed during control rod pattern conversion work that has been performed in boiling water reactors, simplifying and facilitating operation. . Furthermore, compared to devices of this type that have already been proposed, the present invention enables more uniform flattening of the core, making it possible to reduce the number of neutron monitors and simplify neutron instrumentation.
さらに又、本発明は、環流形制御管７内を環流
する制御材物質溶液の濃度を測定する場合に、予
め、上記制御材物質溶液の中にイオン交換樹脂に
対し、制御材物質と類似の挙動を示し、かつ電離
度又はイオン移動度の大きいモニター物質を一定
比率で溶解し、これにより制御材物質溶液の濃度
を連続して測定することによつて、制御材物質の
濃度を間接的に検出することができる。 Furthermore, in the present invention, when measuring the concentration of the control material solution circulating in the circulation type control pipe 7, an ion exchange resin similar to the control material substance is added to the control material solution in advance. The concentration of the control material substance can be indirectly measured by dissolving the monitor substance that exhibits the behavior and having a high degree of ionization or ion mobility at a fixed ratio, and then continuously measuring the concentration of the control material substance solution. can be detected.
以上述べたように本発明によれば、環流形制御
管７とスクラム用制御棒８とを独立して設けてあ
るので、制御システムが簡素化され、原子炉の自
動制御が容易にすることができるばかりでなく、
スクラム用制御棒８はスクラム専用であるから、
この駆動機構が簡単となり、装置全体を小形にす
ることができる。さらに、本発明は各制御管群
a，b，cを炉心から外方に向つて同心的に配
設されているから、炉心径方向の出力を制御でき
ると共に、中性子制御物質溶液の濃度を調整制御
できるから炉心の均質化を図ることができる。特
に、本発明の中性子束制御装置は、中性子制御物
質溶液を環流形制御管の中に流す手段にしてある
ため、冷却材の中に直接中性子制御物質を溶かす
手段よりも、該溶液濃度制御のために取扱う溶液
量は少なくてすむばかりでなく、急速な溶液濃度
変更も実施しやすくなり、しかも中性子制御物質
溶液の濃度制御を行なう機器の容量も小さくする
ことができる。
As described above, according to the present invention, the circulation type control pipe 7 and the scram control rod 8 are provided independently, which simplifies the control system and facilitates automatic control of the reactor. Not only can you do it, but
Since the scram control rod 8 is dedicated to scram,
This drive mechanism becomes simple, and the entire device can be made smaller. Furthermore, the present invention provides for each control tube group to
Since a , b , and c are arranged concentrically outward from the core, it is possible to control the output in the radial direction of the core, and the concentration of the neutron control substance solution can be adjusted and controlled, thereby achieving homogenization of the core. be able to. In particular, since the neutron flux control device of the present invention uses a means for flowing a neutron control substance solution into a circulation type control tube, it is easier to control the concentration of the neutron control substance than by directly dissolving the neutron control substance in a coolant. Therefore, not only the amount of solution to be handled can be reduced, but also it becomes easier to change the solution concentration rapidly, and the capacity of the equipment for controlling the concentration of the neutron control substance solution can also be reduced.
第１図は、既に提案されている中性子束制御装
置における炉心出力分布の各グラフを示し、第２
図は、本発明による中性子束制御装置を組込んだ
原子炉の縦断面図、第３図は、本発明の要部を線
図的に示す横断面図、第４図は、本発明の主要部
のみを示す平面図、第５図は、本発明に使用され
る環流形制御管の斜視図、第６図は、本発明によ
る中性子制御物質循環系を示す線図である。
１……圧力容器、２……炉心支持板、４……上
記格子板、６……燃料集合体、７……環流形制御
管、８……スクラム用制御棒、a，b，c…
…制御管群。
Figure 1 shows each graph of the core power distribution in the neutron flux control device that has already been proposed.
The figure is a vertical cross-sectional view of a nuclear reactor incorporating the neutron flux control device according to the present invention, FIG. 3 is a cross-sectional view diagrammatically showing the main parts of the present invention, and FIG. FIG. 5 is a perspective view of a circulation type control tube used in the present invention, and FIG. 6 is a diagram showing a neutron control substance circulation system according to the present invention. DESCRIPTION OF SYMBOLS 1...Pressure vessel, 2...Core support plate, 4...The above-mentioned grid plate, 6...Fuel assembly, 7...Recirculation type control pipe, 8...Scram control rod, a , b , c ...
...Control tube group.
Claims (1)
保持するようにした各燃料支持部材を設け、この
各燃料支持部材の中央部に中性子制御物質溶液の
環流形制御管を挿脱自在に設け、この環流形制御
管及び上記各燃料集合体を基本単位体として構成
し、さらに、この基本単位体を４単位としてユニ
ツト体に構成し、このユニツト体の中央部にスク
ラム用制御棒を挿脱し得るようにして設け、上記
圧力容器の炉心に中性子制御物質溶液の濃度をそ
れぞれ異にした上記環流形制御管による各制御管
群を上記炉心の中心部から外方に向つて同心的に
配設して炉心径方向の出力分布を制御するように
したことを特徴とする中性子束制御装置。1 Each fuel support member that holds a set of four fuel assemblies is provided in the core of the pressure vessel, and a circulation type control pipe for a neutron control substance solution can be freely inserted and removed in the center of each fuel support member. This circulation type control pipe and each of the fuel assemblies described above are configured as a basic unit, and this basic unit is further configured as a unit with four units, and a scram control rod is installed in the center of this unit. Each group of control tubes including the circulation type control tubes each having a different concentration of a neutron control substance solution is installed in the core of the pressure vessel so as to be removable and inserted concentrically outward from the center of the core. A neutron flux control device characterized in that it is arranged to control power distribution in a radial direction of a reactor core.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP751178A JPS54101088A (en) | 1978-01-26 | 1978-01-26 | Neutron flux control system |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP751178A JPS54101088A (en) | 1978-01-26 | 1978-01-26 | Neutron flux control system |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS54101088A JPS54101088A (en) | 1979-08-09 |
| JPS6153676B2 true JPS6153676B2 (en) | 1986-11-19 |
Family
ID=11667802
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP751178A Granted JPS54101088A (en) | 1978-01-26 | 1978-01-26 | Neutron flux control system |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS54101088A (en) |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS57141595A (en) * | 1981-02-25 | 1982-09-01 | Tokyo Shibaura Electric Co | Nuclear reactor control device |
Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5270298A (en) * | 1975-12-08 | 1977-06-11 | Hitachi Ltd | Nuclear reactor |
-
1978
- 1978-01-26 JP JP751178A patent/JPS54101088A/en active Granted
Patent Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5270298A (en) * | 1975-12-08 | 1977-06-11 | Hitachi Ltd | Nuclear reactor |
Also Published As
| Publication number | Publication date |
|---|---|
| JPS54101088A (en) | 1979-08-09 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| CN109509562B (en) | Reactor core structure, molten salt pebble bed reactor, fuel management method of molten salt pebble bed reactor and reactor core loading method | |
| US6925138B2 (en) | Reactor core and method for operating nuclear reactor | |
| CN111799000B (en) | Critical approaching method for pressurized water reactor primary reactor core without additional primary neutron source | |
| EP2421005B1 (en) | Nuclear reactor | |
| EP2316120A2 (en) | Mixed oxide fuel assembly | |
| GB964841A (en) | Nuclear reactors cooled by liquid metal | |
| WO2018074341A1 (en) | Fuel assembly and reactor core of boiling water nuclear reactor loaded with same | |
| US3257285A (en) | Nuclear power reactor having a high, prompt negative temperature coefficient of reactivity | |
| CN111508622A (en) | Reactor core and reactor | |
| JPH0511081A (en) | Fuel assembly and core | |
| EP2447952B1 (en) | Boiling water reactor core and fuel assembly thereof | |
| CN111007888A (en) | Dissolved oxygen control system in liquid metal coolant | |
| JPS6153676B2 (en) | ||
| JP2510561B2 (en) | Fuel assembly | |
| CN111554414B (en) | Reactor core | |
| CN211956233U (en) | Dissolved oxygen control system in liquid metal coolant | |
| JPH07167988A (en) | Boiling water type thermal neutron reactor and operating method thereof | |
| JPH06174874A (en) | Fuel assembly and reactor core | |
| JPS5927287A (en) | Fast breeder | |
| JP3603878B2 (en) | Operating method of boiling water thermal neutron reactor | |
| EP3422359B1 (en) | Fuel assembly and reactor core into which same is loaded | |
| Li et al. | Conceptual core design of HAPPY200 reactor | |
| JP3015487B2 (en) | Fuel assemblies and reactors | |
| JPH0467158B2 (en) | ||
| JPH0833465B2 (en) | Boiling water reactor and its operating method |