CN102693763B - Control rod comprehensive test method - Google Patents
Control rod comprehensive test method Download PDFInfo
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- CN102693763B CN102693763B CN201110067628.XA CN201110067628A CN102693763B CN 102693763 B CN102693763 B CN 102693763B CN 201110067628 A CN201110067628 A CN 201110067628A CN 102693763 B CN102693763 B CN 102693763B
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- 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
The invention relates to a reactor control rod test method, specifically to a control rod comprehensive test method for optimizing arrangement of the control rod comprehensive test and minimizing a nuclear power station's main line overhauling time occupied by the test. During cold-state and heating periods of a circuit, a single control rod test, a control rod group test and a control rod dropping test are successively conducted. According to the test method provided by the invention, the single control rod test, the control rod group test and the control rod dropping test are successively completed during the cold-state and heating periods of the circuit. Responsive data recorded by an instrumentation and control system during the test period are completely within a safety range, and the control rod comprehensive test no longer occupies the overhauling time of the main line.
Description
Technical field
The present invention is a kind of reactor control rod test method, is specifically related to optimize the arrangement of reactor control rod compbined test, reduces and tests the control rod comprehensive testing method that takies the nuclear power station overhaul main line time.
Background technology
Power regulation functions when gulf, field nuclear power station control rod system is mainly used in the defencive functions such as emergency protection, the acceleration of realization response heap are protected in advance, a class is protected in advance, the pre-protection of two classes and normally moves.The position of control rod system by 121 bundle control rods monitors and realizes above-mentioned functions in the control function manually in groups, under automatic, semi-automatic, single excellent isotype.The various functions of control rod system must working stability during unit operation, it is reliable to control.The fault of control rod system likely causes unit to fall even shutdown of power.
Control rod compbined test be for be arranged on control rod drive mechanisms (CRD) on the assembly of reactor top with and upper strata control system (comprise Reactor trip executive system and in groups with single control system) integration test.It is one of condition precedent of initial criticality after major overhaul.
This test need to promote and lower slotting operation control rod, in Russia's regulation " power producer device nuclear safety specifies ", regulation must take technical measures to get rid of the possibility of simultaneously being introduced positive reactivity by two or more reactive control device, due to a circuit cools agent temperature, to reach a certain temperature limit (depending on a loop boric acid concentration) front, coolant temperature coefficient is for just, be that heating process can be introduced positive reactivity, also can introduce positive reactivity and promote in groups control rod, therefore gulf, field nuclear power station is forbidden promoting control rod during a loop heating in control rod compbined test.
Gulf, field nuclear power station is designed with 121 bundle control rods altogether, has used at present 103 bundles, is divided into 10 groups, and every group of control rod quantity 6-18 restraints not etc., and wherein 1-7 group belongs to shut-down rod group, and 8-10 group belongs to regulating rod group.Control rod compbined test has mainly comprised 3 parts: control rod list rod test (needing 27 hours), and control rod Group test (needing 24 hours), control rod falls rod test (needing 3 hours).
When the test of control rod list rod, operator need promotes separately every bundle control rod, check control rod list rod control model function, check that control rod automatically drops to mechanical support position, bottom when bottom spacing (control rod is put in 5 steps) and triggers excellent signal after rack power-off, check lifting and the lower plugging function of control rod under single excellent pattern, check the control rod automatic hold function spacing in upper and lower, checklist bundle control rod is in upper limit (control rod is put in 181 steps) free-falling function, check that when control rod drops, single control pattern is cancelled function automatically, when checklist rod drops, block to fall the enhanced feature after excellent signal, under checklist rod pattern, can select at most control rod 6 to restraint the function of control rod simultaneously, inspection control rod is controlled function on the spot at rack place.103 bundle control rods need to all carry out above test.
When control rod Group test, operator's rotating operation knob enters manual group schema, check the selection function in groups under manual group schema, check plugging function under the selected excellent lifting of organizing, check the block position deixis of control rod group, check the selection of the excellent pattern of list under manual group schema and cancel function, check manually in groups the priority feature of controlling with single rod, check in manual lifting and the lower slotting priority feature under control model in groups and on the spot, check that two classes protect under (this signal should cause control rod to promote) in advance, the function that manually cannot promote in groups, check that a class protects (this signal should cause control rod lower inserting in groups) to have precedence over the function manually promoting in groups in advance.Operator's operation knob enters semi-automatic pattern, semi-automatic while promoting in groups, check the control rod that is less than N group do not have the middle upper limit (with on to be limited to signal identical) N group control rod does not rise when the signal of position, semi-automatic lower when slotting, check that N does not organize slotting in the time that the control rod that is greater than N group does not have lower limit position signal, what check control rod walks function from fold, checks that single control mode prioritization is in semi-automatic pattern.Check that the pre-protection of a class has precedence over semi-automatic pattern, inspection one class is protected the slotting under fold step of lower control rod in advance, and when pre-protection accelerated in inspection, after the 6th group, 6 fasces drop; Operator's turn knob enters fully automatic mode (APC pattern; this pattern only acts on 8,9,10 3 groups); check 8,9,10 groups and be automatically connected and be selected as working group; check lifting and lower the inserting of APC order; check fully automatic mode and and single control, manual priority in groups, the fully automatic mode under check protection signal is cancelled function automatically.
Control rod falls when test rod, need to rise to reactor roof to one group separately by manual mode, and other excellent group in lower limit, then carries out urgent guarded command, measures dropping the time of rod group in heap top.
Due to the requirement of test restrictive condition, gulf, field nuclear power station plan of major repair has arranged single rod of control rod system to test during being arranged in a loop heating, and overhaul main line is a loop heating, after a loop heating completes, carries out control rod Group test and falls rod test.Because carry out when hot in the loop that the Group test of control rod compbined test and Rod drop test are arranged in after having heated, the overhaul main line time that this part test takies power station in No. one, gulf, field nuclear power station and No. two unit overhauls first reaches 27 hours, this test method causes test to take overhaul main line overlong time, reduce the unit load factor, affected the economic benefit of unit.
Summary of the invention
The object of the present invention is to provide a kind of control rod comprehensive testing method that control rod compbined test no longer takies the overhaul main line time, shortens the power station overhaul duration of realizing.
Technical scheme of the present invention is as follows: control rod comprehensive testing method, and it comprises the steps,
1) meet under test condition precedent when a loop cold conditions, before a loop bulge test, carry out the excellent part test of list of control rod compbined test, the mainly rod position demonstration to control rod list fasces, selection, lifting, insert down, drop, verify with the steering logic such as the interlocking of guard signal, ensure that single rod during unit operation controls the integrality of function;
2) after a loop bulge test completes, extremely hot heating process starts control rod Group test, arch-rival is moving in groups, semi-automatic, steering logic in groups under fully automatic mode is verified, ensures that control rod during unit operation controls the integrality of function in groups; Carry out this part duration of test, must monitor reactor period variation by out-pile Nuclear measurement system monitor screen, in the time that the reactor period is less than 60 seconds, should stop immediately promoting control rod, find out and the out-pile Nuclear measurement system display cycle was greater than after 999 seconds until reason determine whether continue test by value on duty is long; In the time carrying out banked rods lifting, every lifting 10% core height need to stop promoting 60 seconds; In the time promoting control rod in groups, allow at most one group of control rod to rise to upper limit, and another group control rod rises to and is no more than 74% core height, strictly forbid simultaneously two groups and above control rod to rise to upper limit;
3) after four main pumps all start, start to carry out control rod and fall rod test, main checking is occurring that after guard signal, control rod can normally drop.
Remarkable result of the present invention is: according to test method provided by the present invention, in a loop cold conditions with completed successively the test of control rod list rod, control rod Group test, control rod between the period of heating and fall rod and test.The reactive data of duration of test instrument control system log (SYSLOG) are completely in safe range, control rod compbined test no longer takies the overhaul main line time, compared with former test method, control rod compbined test has been saved overhaul main line and has been reached 27 hours, reaches 2,400 ten thousand yuan in gulf, field nuclear power station year creation economic benefit.
Embodiment
Describe the present invention below in conjunction with specific embodiment:
The compbined test of control rod system can arrange as follows: in a loop cold conditions with carry out successively the test of control rod list rod, control rod Group test, control rod between the period of heating and fall rod and test.Under this scheme, the overhaul main line time is a loop heating, and control rod compbined test no longer takies overhaul main line.
Control rod compbined test must meet following condition precedent:
Unit completes and reloads, and a loop seals; One 25 DEG C of loop temperatures～280 DEG C; One circuit pressure 0.49MPa～15.7MPa; One loop boron concentration is not less than 16g/L; 0～4 operation of main circulation pump, wherein needs 4 main circulation pumps all to put into operation in the measurement Dropping of control rod time; All control rod drive mechanisms (CRD) are all installed to top assembly; Institute is dynamic to be all connected on control rod drive mechanisms (CRD) and reactor shaft cement wall from tape cable with instrument control; Control rod drive mechanisms (CRD) circulating cooling system puts into operation; Heap outer core surveys and 4 passages of safe instrument control service system all put into operation; On master-control room back-up disk without any reactor protection system and reactor restriction system alerting signal; Control rod system protection-protect in advance part compbined test to complete, all racks put into operation.Control rod system power supply part compbined test completes, and all racks put into operation.
Concrete operation step of the present invention is as follows:
1) meet under test condition precedent when a loop cold conditions, before a loop bulge test, carry out the excellent part test of list of control rod compbined test, the mainly rod position demonstration to control rod list fasces, selection, lifting, insert down, drop, verify with the steering logic such as the interlocking of guard signal, ensure that single rod during unit operation controls the integrality of function;
2) a loop bulge test (17.6Mpa, 120 DEG C) complete after, extremely hot heating process starts control rod Group test, arch-rival is moving in groups, semi-automatic, steering logic in groups under fully automatic mode is verified, ensures that control rod during unit operation controls the integrality of function in groups; Carry out this part duration of test, must monitor reactor period variation by out-pile Nuclear measurement system monitor screen, in the time that the reactor period is less than 60 seconds, should stop immediately promoting control rod, find out and the out-pile Nuclear measurement system display cycle was greater than after 999 seconds until reason determine whether continue test by value on duty is long; In the time carrying out banked rods lifting, every lifting 10% core height need to stop promoting 60 seconds; In the time promoting control rod in groups, allow at most one group of control rod to rise to upper limit, and another group control rod rises to and is no more than 74% core height, strictly forbid simultaneously two groups and above control rod to rise to upper limit;
3) after four main pumps all start, a loop, in the time of full flow state, starts to carry out control rod and falls rod test, and main checking is occurring that after guard signal, control rod can normally drop.
Use of the present invention must have strict foundation, and in order to ensure nuclear safety, the former restrictive condition of control rod compbined test is during a loop heating, to forbid promoting in groups that control rod, object are to prevent that pile reactivity is out of control.The relating to parameters such as reactor core reactivity and reactor core burnup, reactor capability, coolant temperature, boric acid concentration, control rod position.Reactor is reaching before critical conditions, cold conditions, heating and hot under, reactor core burnup and power are all indeclinable, therefore, reactivity is only relevant with control rod position to coolant temperature, reactor core boric acid concentration, therefore can calculate the reactivity of reactor.As calculated, in reactor-loop heating process, even if one group, two groups of control rods of Maximum Value are mentioned reactor core top, another group is mentioned 60% core height, reactive maximal value is still less than-16%, meets the reactive requirement that is less than-2% in heating process of technical specification book; Even if consider limiting case, all ten groups of control rods all propose reactor core, in heating process, rely on boron effect, and pile reactivity maximal value-7.93%, still meets the requirement of technical specification book.According to this result of calculation, lifting in groups and the scram operation of during a loop heating, carrying out control rod compbined test are safety, feasible.
Claims (1)
1. control rod comprehensive testing method, is characterized in that: it comprises the steps,
1) meet under test condition precedent when a loop cold conditions, before a loop bulge test, carry out the excellent part test of list of control rod compbined test, rod position demonstration to control rod list fasces, selection, lifting, insert, drop, verify with the interlocked control logic of guard signal down, ensure that single rod during unit operation controls the integrality of function;
2) after a loop bulge test completes, extremely hot heating process starts control rod Group test, steering logic in groups manually in groups, under semi-automatic, fully automatic mode is verified, ensured that control rod during unit operation controls the integrality of function in groups; Carry out this part duration of test, must monitor reactor period variation by out-pile Nuclear measurement system monitor screen, in the time that the reactor period is less than 60 seconds, should stop immediately promoting control rod, find out and the out-pile Nuclear measurement system display cycle was greater than after 999 seconds until reason determine whether continue test by value on duty is long; In the time carrying out banked rods lifting, every lifting 10% core height need to stop promoting 60 seconds; In the time promoting control rod in groups, allow at most one group of control rod to rise to upper limit, and another group control rod rises to and is no more than 74% core height, strictly forbid simultaneously two groups and above control rod to rise to upper limit;
3) after four main pumps all start, start to carry out control rod and fall rod test, checking is occurring that after guard signal, control rod can normally drop.
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| CN201110067628.XA CN102693763B (en) | 2011-03-21 | 2011-03-21 | Control rod comprehensive test method |
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| CN201110067628.XA CN102693763B (en) | 2011-03-21 | 2011-03-21 | Control rod comprehensive test method |
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| CN102693763A CN102693763A (en) | 2012-09-26 |
| CN102693763B true CN102693763B (en) | 2014-07-23 |
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| CN103500524B (en) * | 2013-09-29 | 2016-01-06 | 中广核工程有限公司 | Nuclear power plant's Control rod drive line line cold test device and test method |
| CN107093475B (en) * | 2017-05-03 | 2019-03-05 | 中国核动力研究设计院 | A kind of linear motor type control rod drive mechanism falls stick Method Of Time Measurement |
| CN113128714A (en) * | 2021-04-20 | 2021-07-16 | 广东核电合营有限公司 | Nuclear power unit control rod-based test system and checking test method |
Citations (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US5001841A (en) * | 1986-09-22 | 1991-03-26 | Mitsubishi Jukogyo, K.K. | Apparatus for inspecting a profile of a rod |
| JPH09189792A (en) * | 1996-01-10 | 1997-07-22 | Nuclear Fuel Ind Ltd | Visual inspection device for control rod cluster assembly |
| CN1176706A (en) * | 1995-12-29 | 1998-03-18 | 法玛通公司 | Device and method for testing command bundle rods for nuclear reactors |
| US5889832A (en) * | 1995-12-26 | 1999-03-30 | Framatome | Nuclear reactor control cluster having removable rods |
-
2011
- 2011-03-21 CN CN201110067628.XA patent/CN102693763B/en active Active
Patent Citations (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US5001841A (en) * | 1986-09-22 | 1991-03-26 | Mitsubishi Jukogyo, K.K. | Apparatus for inspecting a profile of a rod |
| US5889832A (en) * | 1995-12-26 | 1999-03-30 | Framatome | Nuclear reactor control cluster having removable rods |
| CN1176706A (en) * | 1995-12-29 | 1998-03-18 | 法玛通公司 | Device and method for testing command bundle rods for nuclear reactors |
| JPH09189792A (en) * | 1996-01-10 | 1997-07-22 | Nuclear Fuel Ind Ltd | Visual inspection device for control rod cluster assembly |
Non-Patent Citations (5)
| Title |
|---|
| Office of Nuclear Regulatory Research.Initial Test Programs for Water-Cooled Nuclear Power Plants.《NRC Regulatory Guide 1.68 》.2007, * |
| 徐霞军等.核电厂控制棒综合试验安全限制研究.《电力安全技术》.2011,第13卷(第6期), |
| 核电厂控制棒综合试验安全限制研究;徐霞军等;《电力安全技术》;20110630;第13卷(第6期);第24-26页 * |
| 王继英等.秦山核电二期工程控制棒驱动线热态综合试验.《核动力工程》.2003,第24卷(第2期), |
| 秦山核电二期工程控制棒驱动线热态综合试验;王继英等;《核动力工程》;20030630;第24卷(第2期);第150-154页 * |
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