KR20010076547A - Nozzle configuration in S/G for Miol-loop operation enhancement - Google Patents

Nozzle configuration in S/G for Miol-loop operation enhancement Download PDF

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Publication number
KR20010076547A
KR20010076547A KR1020000003749A KR20000003749A KR20010076547A KR 20010076547 A KR20010076547 A KR 20010076547A KR 1020000003749 A KR1020000003749 A KR 1020000003749A KR 20000003749 A KR20000003749 A KR 20000003749A KR 20010076547 A KR20010076547 A KR 20010076547A
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South Korea
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water level
steam generator
nozzle
support structure
reactor
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KR1020000003749A
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Korean (ko)
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KR100385838B1 (en
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이상원
박영섭
양준석
이종범
김병섭
조성제
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이종훈
한국전력공사
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Priority to KR10-2000-0003749A priority Critical patent/KR100385838B1/en
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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D1/00—Details of nuclear power plant
    • 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
    • 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

Abstract

PURPOSE: A nozzle structure of a steam generator having a low water level operational reliability is provided to maintain a water level of a cooling system as high as possible and smoothly obviate problems of a dynamic support structure. CONSTITUTION: An angle of a nozzle(210) on an outlet side is 40° on the basis of a bottom surface of a tube sheet. By varying the angle of the nozzle(210) installed on the outlet side of a steam generator(200), a water level is maximized according to a requested low level operation. An interference between a suction line of a cooling pump and a vertical column of an RCP pump support structure is minimized while the minimum water level is increased. As width of the vertical column of the reactor cooling pump support structure is reduced by 4 inches, thickness is increased by 1.5 inches. Accordingly, strength of the reactor cooling pump support structure is not changed. In this condition, if a water level is decreased, a water level is maintained to be much higher than that in the conventional apparatus. In addition, nozzle dams can be easily installed on an inlet of the steam generator and an outlet of the steam generator. In order to install the nozzle dams, first the water level is reduced to a mid-loop level, manways which is disposed on both sides of the steam generator are opened and then the nozzle dams are installed.

Description

저수위 운전신뢰성을 갖는 원자로 증기발생기의 노즐구조{Nozzle configuration in S/G for Miol-loop operation enhancement}Nozzle configuration in S / G for Miol-loop operation enhancement

본 발명은 원자력 발전소의 원자로에 관한 것으로 더욱 구체적으로는 원자력 발전소의 핵연료 재장전이나 장치의 보수시 냉각수의 수위를 낮추어 저수위 운전에 용이하도록 한 저수위 운전신뢰성을 갖는 원자로 증기발생기의 노즐구조에 관한 것이다.The present invention relates to a reactor of a nuclear power plant, and more particularly, to a nozzle structure of a reactor steam generator having a low water level operation reliability that is easy to operate at a low water level by lowering the level of cooling water during nuclear fuel reloading or maintenance of a device. .

원자로 냉각재 계통의 기기 배치는 도 1에 도시한 바와 같이 원자로(100)를 중심으로 두 개의 증기발생기(200)와 4개의 냉각재펌프(300)가 서로 대칭하게 배치된다. 원자로 지지구조물은 원자로(100)가 수직 중심선에 위치하도록 지지하지만 다른 기기들의 지지구조물은 원자로 냉각재계통의 가열, 냉각시 열팽창 방향으로 움직일 수 있는 구조로 되어 있어 열팽창에 따른 기기 및 배관에 미치는 하중을 최소화하도록 설계하여야 한다.In the apparatus arrangement of the reactor coolant system, as illustrated in FIG. 1, two steam generators 200 and four coolant pumps 300 are disposed symmetrically with respect to the reactor 100. The reactor support structure supports the reactor 100 so as to be positioned at the vertical center line, but the support structure of other devices is configured to move in the thermal expansion direction during heating and cooling of the reactor coolant system. It should be designed to be minimal.

원자력 발전소는 일정기간 전출력 운전을 한 후 핵연료의 재장전을 위해 원자로 정지기간을 갖는다.Nuclear power plants have reactor shutdown periods for reloading nuclear fuel after a full power operation for a period of time.

그러나 원자로 정지기간 중에도 핵연료의 핵반응 특성상 일정량의 잔열이 계속해서 발생하게 되며, 이러한 잔열의 제거를 위해 일반적으로 잔열제거계통 또는 정지냉각계통이 운전을 하여 열제거기능을 수행하도록 한다.However, due to the nuclear reaction characteristics of the nuclear fuel during the reactor shutdown period, a certain amount of residual heat is continuously generated. In order to remove such residual heat, a residual heat removal system or a stop cooling system is generally operated to perform a heat removal function.

또한 증기발생기 U 튜브 검사 및 원자로 냉각재펌프 밀봉재(Seal)교체와 같은 기기의 점검 및 유지보수를 위하여 도 2 및 도 3에 도시된 바와 같이 저수위운전이 요구되며, 증기발생기의 보수나 점검을 위해서는 작업자가 내부에 들어가서 작업을 하여야 하므로 별도의 관로 차단을 위한 노즐댐(220)을 설치하여야 한다.In addition, low water level operation is required as shown in FIGS. 2 and 3 to inspect and maintain the equipment such as steam generator U tube inspection and reactor coolant pump seal replacement, and to maintain or inspect the steam generator. Since the work must go into the interior should be installed a nozzle dam 220 for blocking the separate pipe.

저수위운전은 원자로 냉각재 계통의 수위가 원자로 용기 플랜지로부터 3feet 이하인 상태에서의 운전을 의미하며, 이때 수위상층부는 비응축성 기체인 공기로 채워져 있게 된다.Low water level operation refers to the operation when the level of the reactor coolant system is 3 feet or less from the reactor vessel flange, where the water level layer is filled with air, which is a non-condensable gas.

특히 원자로냉각재계통의 수위가 고온관 상단 이하인 상태에서의 운전을 부분충수운전(Mid-loop Operation)이라고 하며, 이런 부분충수운전 중에는 정지냉각계통(Shutdown Cooling System) 또는 잔열제거계통(Residual HeatRemoval System)의 펌프 흡입관(Suction Line)에 다량의 공기가 유입되거나 펌프 전원이 상실되면 정지냉각펌프의 기능이 상실될 수가 있다.In particular, the operation in which the reactor coolant system level is below the top of the hot pipe is called a mid-loop operation, and during such a partial filling operation, a shutdown cooling system or a residual heat removal system is used. If a large amount of air enters the pump suction line or the pump power is lost, the function of the stop cooling pump may be lost.

만약 장기간동안 정지냉각계통의 기능이 회복되지 못하고 적절한 원자로냉각재 보충(Makeup)이 이루어지지 않으면 노심내의 물은 비등하고 최악의 경우 노심노출(Core Uncovery)이 발생할 수 있다. 정지냉각계통의 흡입관으로 공기가 유입될 가능성은 원자로 냉각재계통의 수위가 낮거나 정지냉각펌프의 유량이 증가할수록 커진다.If the stationary cooling system is not restored for a long time and proper reactor coolant replenishment is not achieved, the water in the core may boil and, in the worst case, core uncovery may occur. The likelihood of air entering the stationary cooling system's suction line increases as the reactor coolant system is low or the flow rate of the stationary cooling pump increases.

증기발생기 U튜브를 점검하는 동안 공기유입으로 인한 잔열제거기능상실(Loss of Residual Heat Removal, loss of Decay Heat Removal, or Loss of Shutdown Cooling) 가능성을 줄이고, 점검시 작업자의 방사능 피폭을 감소시키고, 원자로용기와 증기발생기를 동시에 유지 보수하여 발전소 이용성을 향상시키기 위하여 증기발생기 입구 및 출구 노즐에 노즐댐(nozzle Dams)(220)을 설치한다. 노즐댐(220)을 설치하기 위해서는 먼저 수위를 고온관 중심수위(Mid-loop Level)까지 낮추고 양쪽 증기발생기 일차측에 있는 보수용 작업자 출입구(Manway)를 모두 연 후 노즐댐을 설치한다.Reduce the likelihood of loss of residual heat removal, loss of Decay Heat Removal, or Loss of Shutdown Cooling during inspection of the steam generator U-tube, reduce the worker's radiation exposure during inspection, Nozzle Dams 220 are installed at the steam generator inlet and outlet nozzles to maintain the vessel and the steam generator at the same time to improve the utility of the power plant. In order to install the nozzle dam 220, first, the water level is lowered to the mid-loop level of the hot pipe, and both of the maintenance worker entrances on both sides of the steam generator are opened, and then the nozzle dam is installed.

노즐댐의 설치가 끝나는 즉시 도 3과 같이 원자로 냉각재계통 수위를 원자로 용기 플랜지 근처까지 상승시켜 정지냉각계통의 흡입관에 공기유입으로 인한 잔열제거기능상실 가능성을 줄인다.Immediately after the installation of the nozzle dam, as shown in FIG. 3, the level of the reactor coolant system is raised to near the reactor vessel flange to reduce the possibility of loss of residual heat due to air inflow into the suction pipe of the stationary cooling system.

그러나 이러한 경우 원자로 냉각재 수위감소에 따라 정지냉각 펌프의 유효 흡입수두 감소로 인한 운전부담 및 펌프의 고장 가능성이 있으며, 원자로 냉각재계통의 가열, 냉각시 열팽창 방향으로 움직일 수 있는 구조로 되어 있는 점을 감안하여 열팽창에 따른 기기 및 배관에 미치는 하중을 최소화하도록 설계하여야 하는 어려움이 있다.However, in this case, there is a possibility of operating burden and pump failure due to the reduction of the effective suction head of the stationary cooling pump due to the decrease of the reactor coolant level, and it is considered that the structure can move in the direction of thermal expansion when heating and cooling the reactor coolant system. Therefore, there is a difficulty in designing to minimize the load on the equipment and piping due to thermal expansion.

이와 같은 문제점을 보완하기 위하여 안출된 본 발명에서 저수위운전시 원자로 냉각재 계통의 수위를 가능한 높게 유지하도록 하되 열팽창에 따른 역학적 지지구조의 문제를 원만히 해결하면서도 저수위 운전신뢰성을 갖는 원자로 증기발생기의 노즐구조를 제공하는데 그 목적이 있다.In order to solve the above problems, in the present invention, the nozzle structure of the reactor steam generator having low water level operation reliability while smoothly solving the problem of the mechanical support structure due to thermal expansion while maintaining the level of the reactor coolant system as high as possible during low water level operation is possible. The purpose is to provide.

도 1은 본 발명이 적용된 원전의 주요 구조물 배치상태 사시도.1 is a perspective view of the main structure arrangement state of the nuclear power plant to which the present invention is applied.

도 2는 본 발명이 적용된 원전의 저수위 운전 개략 단면도.Figure 2 is a schematic cross-sectional view of low water level operation of the nuclear power plant to which the present invention is applied.

도 3은 본 발명이 적용된 원전의 재장전 운전 개략 단면도.Figure 3 is a schematic cross-sectional operation of the nuclear power plant to which the present invention is applied.

도 4는 본 발명이 적용된 원전의 재장전 운전상태를 도시한 상세 단면도.Figure 4 is a detailed cross-sectional view showing a reload operation state of the nuclear power plant to which the present invention is applied.

도 5는 본 발명과 종래의 출구측 노즐 각도 비교도.Figure 5 is a view of comparing the present invention and the conventional outlet side nozzle angle.

도 6은 본 발명에 의한 댐프 설치부의 요부 확대 단면도.Figure 6 is an enlarged cross-sectional view of the main portion of the damper installation portion according to the present invention.

*도면의 주요부분에 대한 부호의 설명** Description of the symbols for the main parts of the drawings *

100: 원자로용기 110: 고온관100: reactor vessel 110: high temperature tube

120: 저온관 200: 증기발생기120: low temperature pipe 200: steam generator

210: 출구측 노즐 220: 노즐댐210: outlet nozzle 220: nozzle dam

300: 펌프 400: 압력조절기300: pump 400: pressure regulator

상기의 목적을 달성하기 위하여 본 발명에서는 원자력발전소의 원자로에 핵연료 재장전작업과 동시에 증기발생기와 부수되는 시설의 점검 및 유지보수작업시 요구되는 저수위운전에 따른 수위의 극대화를 위한 증기발생기의 출구측 노즐의 각을 튜브시트 밑면을 기준으로 40도를 유지하도록 구성함을 특징으로 한다.In order to achieve the above object, in the present invention, at the outlet side of the steam generator for maximizing the water level according to the low water level operation required during the inspection and maintenance work of the steam generator and accompanying facilities at the same time as the nuclear fuel reloading operation of the nuclear power plant; Characterized in that the angle of the nozzle to maintain 40 degrees relative to the bottom of the tube sheet.

이하 첨부된 도면에 의거하여 본 발명을 상세히 설명한다.Hereinafter, the present invention will be described in detail with reference to the accompanying drawings.

도 1은 본 발명이 적용된 원전의 주요 구조물 배치상태 사시도를 도시한 것으로 중앙에 원자로용기(100)가 위치하며, 양측으로는 증기발생기(200)가 각각 설치되어 있고 고온관(110)과 저온관(120)에 의해 상기 원자로용기(100)와 폐쇄관로로 연결되어 있다.1 is a perspective view showing a main structure arrangement state of a nuclear power plant to which the present invention is applied, and a reactor vessel 100 is located at the center, and steam generators 200 are installed at both sides, respectively, and a high-temperature tube 110 and a low-temperature tube. The reactor vessel 100 is connected to the reactor vessel 100 by a 120.

상기 저온관(120)에는 각각 물을 순환시켜주기 위한 펌프(300)가 구비되며 일차 측 고온관(110)에는 별도로 폐쇄관로내의 압력을 조절하도록 하는 가압기인 압력조절장치(400)가 설치되어 있다.Each of the cold pipes 120 is provided with a pump 300 for circulating water, and the primary side hot pipe 110 is provided with a pressure regulator 400 which is a pressurizer for controlling pressure in a closed pipe separately. .

도 2는 본 발명이 적용된 원전의 저수위 운전 개략 단면도이고, 도 3은 본 발명이 적용된 원전의 재장전시 수위를 상승한 상태의 운전 개략 단면도를 도시한 것으로 이와 같은 원전의 가동중 원전의 핵연료 재장전시에는 원전의 운전을 정지시킴과 함께 증기발생기(200)의 점검이나 유지 보수 즉, 증기발생기 U 튜브(121) 검사 및 원자로 냉각재펌프 밀봉재(Seal)교체와 같은 기기의 점검 및 유지보수를 수행한다.Figure 2 is a schematic cross-sectional view of the low water level operation of the nuclear power plant to which the present invention is applied, Figure 3 is a schematic schematic cross-sectional view of the operation of the nuclear power plant when the recharge of the nuclear power plant to which the present invention is applied. In addition to stopping the operation of the nuclear power plant, the inspection and maintenance of the steam generator 200, that is, the inspection and maintenance of the device, such as inspection of the steam generator U tube 121 and replacement of the reactor coolant pump seal (Seal).

이와 같은 작업을 위해서는 작업자가 증기발생기(200)의 내부에 직접 들어가서 작업을 하여야 하므로 상기 증기발생기(200)과 연결되는 관로를 차단하기 위한 차단 노즐 댐(220)을 설치하여야 한다.In order to perform such a work, the worker must enter the steam generator 200 directly and work to install a blocking nozzle dam 220 for blocking a pipeline connected to the steam generator 200.

본 발명은 도 4는 본 발명이 적용된 원전의 재장전 운전상태를 도시한 상세 단면도이다.4 is a detailed cross-sectional view showing a reload operation state of a nuclear power plant to which the present invention is applied.

도 5는 본 발명의 주요 부분인 증기발생기(200)의 입구측 고온관(110)과 출구측 노즐(210)을 도시한 단면도로서 음선으로 도시한 종래의 출구측 노즐을 설계 변경하여 실선으로 도시한 바와 같이 출구측 노즐(210)을 변경하여 도시한 것이다.5 is a cross-sectional view showing the inlet-side high-temperature tube 110 and the outlet-side nozzle 210 of the steam generator 200, which is a main part of the present invention, by changing the design of the conventional outlet-side nozzle shown in a negative line in a solid line. As shown, the outlet nozzle 210 is changed.

즉, 증기발생기(200)의 출구측 노즐(210)의 설치각도를 변경하여 줌으로서 요구되는 저수위운전에 따른 수위의 극대화한 것으로 출구측 노즐(210)의 각을 튜브시트 밑면을 기준으로 40도를 유지하도록 구성한다.That is, by maximizing the water level according to the low water level operation required by changing the installation angle of the outlet nozzle 210 of the steam generator 200, the angle of the outlet nozzle 210 is 40 degrees from the bottom of the tube sheet. Configure to keep.

이와 같이 함으로서 원자로 냉각재펌프 흡입관과 RCP 펌프 지지구조물의 수직기둥 사이의 간섭이 최소화 되도록 하면서도 최소 수위를 상승시켜 본 발명의 목적을 달성하도록 하는 것이다.By doing so, the interference between the reactor coolant pump suction pipe and the vertical column of the RCP pump support structure is minimized while the minimum water level is raised to achieve the object of the present invention.

즉 원자로냉각재펌프 지지구조물의 수직기둥의 폭을 4인치 정도 감소시킨 만큼 그에 대한 보완으로 두께가 1.5인치 정도 증가시켜 원자로냉각재펌프 지지구조물의 강성도를 변경시키지 않는 설계를 얻을 수 있다.That is, as the width of the vertical column of the reactor coolant pump support structure is reduced by about 4 inches, a design that does not change the stiffness of the reactor coolant pump support structure can be obtained by increasing the thickness by about 1.5 inches.

이와 같은 상태에서 수위를 낮추면 종래의 장치보다 훨씬 높은 수위를 유지할 수 있으며, 아울러 증기발생기 입구 및 출구 노즐에 노즐댐(nozzle Dams)(220)을 설치하기가 용이하다.By lowering the water level in this state, it is possible to maintain a much higher water level than the conventional apparatus, and it is easy to install nozzle dams 220 at the steam generator inlet and outlet nozzles.

상기 노즐댐(220)을 설치하기 위해서는 먼저 수위를 고온관 중심수위(Mid-loop Level)까지 낮추고 양쪽 증기발생기 일차측에 있는 보수용 작업자 출입구(Manway)를 모두 연 후 노즐댐(220)을 설치한다.In order to install the nozzle dam 220, the water level is first lowered to the mid-loop level of the high temperature pipe, and both of the maintenance worker entrances on both sides of the steam generator are opened, and then the nozzle dam 220 is installed. do.

도 6는 본 발명의 요부 확대 단면도로서 노즐댐(220)을 설치한 상태를 도시한 부분 단면도이다.6 is a partial cross-sectional view showing a state where the nozzle dam 220 is installed as an enlarged cross-sectional view of the main part of the present invention.

노즐댐(220)의 설치가 끝나는 즉시 원자로 냉각재계통 수위를 원자로 용기 플랜지 근처까지 상승시켜 정지냉각계통의 흡입관에 공기유입으로 인한 잔열제거기능상실 가능성을 줄인다.As soon as the installation of the nozzle dam 220 is completed, the level of the reactor coolant system is raised to the vicinity of the reactor vessel flange to reduce the possibility of the residual heat elimination function due to air inflow into the suction pipe of the stationary cooling system.

도 3은 부분충수운전에서 노즐댐을 설치한 후 수위를 상승시켰을 때의 배치도이며, 이 상태에서 증기발생기는 보수를 수행하고 원자로는 연료재장전 작업을 수행하도록 한다.3 is a layout view when the water level is increased after installing the nozzle dam in the partial filling operation, in this state, the steam generator performs the repair and the reactor to perform the fuel reloading operation.

이와 같이 하므로서 원자로 냉각재 주배관의 중심선이 싱글노즐댐 최하단의 간격을 8인치 이상으로 확보하는 것이 가능해졌으며, 이는 한국 표준형 원전에 비하여 1인치 이상 증가하여 잔열제거 계통의 기능상실 가능성을 현저하게 감소시킨다.In this way, the center line of the reactor coolant main pipe can be secured more than 8 inches apart from the bottom of the single nozzle dam, which increases by more than 1 inch compared to the Korean standard nuclear power plant, significantly reducing the possibility of malfunction of the residual heat removal system.

이상과 같은 본 발명은 원자력 발전소의 핵연료 재장전이나 보수시 냉각수의 수위를 낮추어 저수위 운전에 용이하도록 한 저수위 운전신뢰성을 갖도록 하는 것으로서 원전운전의 효율성을 높이며, 아울러 작업의 신뢰성을 갖도록 하는 것으로서 매우 유용한 가치를 갖는 기술인 것이다.The present invention as described above is to have a low water level operating reliability that lowers the level of the coolant during nuclear reloading or repair of nuclear power plants to facilitate low water level operation to increase the efficiency of nuclear power plant operation, and also very useful as a reliable operation. It is a technology that has value.

Claims (1)

원자력발전소의 원자로에 핵연료 재충전작업과 동시에 증기발생기(200)와 부수되는 시설의 점검 및 유지보수작업시 요구되는 저수위운전에 따른 수위의 극대화를 위한 증기발생기(200)의 출구측 노즐(210)의 각을 튜브시트 밑면을 기준으로 40도를 유지하도록 구성함을 특징으로 하는 저수위 운전신뢰성을 갖는 원자로 증기발생기의 노즐구조.The outlet nozzle 210 of the steam generator 200 for maximizing the water level according to the low water level operation required during the inspection and maintenance work of the steam generator 200 and accompanying facilities at the same time as the nuclear fuel recharging operation in the nuclear power plant reactor. The nozzle structure of the reactor steam generator having a low water level operating reliability, characterized in that the angle is maintained to 40 degrees relative to the bottom of the tube sheet.
KR10-2000-0003749A 2000-01-26 2000-01-26 Nozzle configuration in S/G for Mid-loop operation enhancement KR100385838B1 (en)

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Cited By (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
KR20020064046A (en) * 2001-01-31 2002-08-07 한국수력원자력 주식회사 an atomic power generator steam system one nozzle
KR100900730B1 (en) * 2007-11-09 2009-06-05 한국원자력연구원 Steam generator cassette structure for flow mixing in integrated reactor
KR20220090109A (en) * 2020-12-22 2022-06-29 한국수력원자력 주식회사 System for nuclear power plant removing mid-loop operation during nuclear reactor outage period

Cited By (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
KR20020064046A (en) * 2001-01-31 2002-08-07 한국수력원자력 주식회사 an atomic power generator steam system one nozzle
KR100900730B1 (en) * 2007-11-09 2009-06-05 한국원자력연구원 Steam generator cassette structure for flow mixing in integrated reactor
KR20220090109A (en) * 2020-12-22 2022-06-29 한국수력원자력 주식회사 System for nuclear power plant removing mid-loop operation during nuclear reactor outage period

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