SCIENTIFIC AND TECHNICAL SUPPORTSCIENTIFIC AND TECHNICAL SUPPORT OF CHANNEL-TYPE REACTOR PLANT OF CHANNEL-TYPE REACTOR PLANT

OPERATIONOPERATION

Dragunov Y.GDragunov Y.G., ., PetrovPetrov А.А. А.А.

MNTKMNTK-2010-2010

ОТКРЫТОЕ АКЦИОНЕРНОЕ ОБЩЕСТВООТКРЫТОЕ АКЦИОНЕРНОЕ ОБЩЕСТВО«ОРДЕНА ЛЕНИНА НАУЧНО-ИССЛЕДОВАТЕЛЬСКИЙ И«ОРДЕНА ЛЕНИНА НАУЧНО-ИССЛЕДОВАТЕЛЬСКИЙ И КОНСТРУКТОРСКИЙ КОНСТРУКТОРСКИЙ

ИНСТИТУТ ЭНЕРГОТЕХНИКИ ИМЕНИ Н.А. ДОЛЛЕЖАЛЯ» ИНСТИТУТ ЭНЕРГОТЕХНИКИ ИМЕНИ Н.А. ДОЛЛЕЖАЛЯ» N.A.Dollezhal N.A.Dollezhal

Research and Development Institute of Power EngineeringResearch and Development Institute of Power Engineering

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MAIN PERFORMANCE INDICATORS OF RBMK NPPsMAIN PERFORMANCE INDICATORS OF RBMK NPPs ININ 2009 2009

Power generation – 75382,3 million KW/h (46,2% of the total output);

Capacity factor – 78,23%;Availability factor – 80,41%;Number of violations– 13 (in 2008 – 18);Number of scrams – 7 (in 2008 – 4).

Note: This period was characterised by modernisation and special system introduction at power units Kursk-4 and Leningrad-4 which was the reason for those units long-term shutdowns.

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MAIN ACTIVITIES AT RBMK POWER UNITScompleted in the second half of 2008 – beginning of 2010 with OAO

RDIPE specialists’ involvement

Completion of Kursk-4 and Leningrad-4 power units modernisation and reconstruction;

ISA development for Smolensk-1 and Leningrad-3 power units; Performance of work on Leningrad-3 power unit life time extension; Development of substantiation for Kursk-2, Leningrad-2 & 3 power

units operation at 105% power; Testing of Kursk-1 & 2 and Leningrad-2 power units at increased

power.

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Introduction of (IMCPS) and other special systems at Kursk-4 power unit was performed in recordingly short time – 250 days

Modernised main control room

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WORK ON REACTOR NEUTRONICS AND FUEL UTILIZATION EFFICIENCY IMPROVEMENT

In 2008-2009 core modernisation involving IICPS introduction was completed at Leningrad-3, Kursk-3 & 4 power units. Replacement of CPS regulators with cluster-type ones. Reactor neutronics calculations and experimental study were performed.Modernisation of the reactor cores led to reactor neutronics and nuclear safety improvement.

Changes in reactor neutronics at the rated power following core modernisation are demonstrated with an example of Kursk-4 power unit.

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Neutron-physical characteristics of Kursk-4 reactor

(the first value – as of March 2010 / the second value – prior to upgrading in July 2008)

1. Core efficiency – 3.6 βef / 2/4 βef

2. Core efficiency , taking into account a failure of one most efficient organ– 3.28 βef / 2.06 βef

3. Reactivity effect Эффект реактивности in case of CPSCC dewatering – 0.54 βef / 1.1 βef

4. FPR-CPS system efficiency – 11,3 βef / 11,4βef

5. Subcriticality of cooldown depoisoned reactor with withdrawn core regulating organs– 3.7% / 3.0%

6. Fuel average burn-up in the core – 14.76 MW·day/kg / 14.1 MW·day/kg

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Introduction of cluster regulating organs (CRO) at RBMK-1000 reactors

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89 91

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256232

2001 2002 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 Year

142

94

135

16 17

319

- From the programme of CRO introduction at RBMK-1000 reactors

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NPP Power unit No.

Number of CRO, pcs.(as of March 2010 /after complete transfer)

Kursk

1 73/1362 98/1363 146/1664 166/166 (transfer completed)

Leningrad

1 96/1332 122/1353 141/1654 50/165

Smolensk1 155/1662 144/1663 105/166

Total: 1296/1700

Number of CRO in modernized IMCPS system at RBMK-100 reactors

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ТВС 2%4%

ТВС 2,4%62%

ЭТВС 2,6%31%

ДП3%

ДП1%

ЭТВС 2,6%22%ЭТВС 2,8%

77%

TRANSFER RBMK-1000 POWER UNITS TO URANIUM-ERBIUM FUEL OF HIGHER ENRICHMENT AND CHANGE OF REACTOR CHARACTERISTICS

2001 год 2009 год

Unloaded fuel power generation growth at different NPPs

Change of summary number of additional absorbers and average power generation of the fuel

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ЛАЭС САЭС КАЭС

912836

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392309

256 246205

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количество ДП количество кобальтовых ДП энерговыработка

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Change of requirements to RMBK -1000FA design with the introduction of new generation FA

RBMK-1000 FA design Standard FA New generation FA

Average fuel enrichment in the FA 2.8% 3.0%

Fuel burn-up fraction 30 MW∙day/kgU

(3380 MWday/FA)35 MW∙day/kgU

(4000 MWday/FA)

Designed lifetime 8 years 10 years

Relative number of FC failures per power unit per year, no more than -

(1÷2) 10∙ -5

New generation RBMK-1000 fuel assembly design features

Fuel enrichment radial shaping

Fuel assembly equipping with tailpiece-filter

Central fastening of fuel assemblies10

Fuel assembly design equipped with tailpiece-filter

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фильтр всплыл фильтр включен Изменение перепада на фильтре

Filtering element working position

Filtering element in “emerced” position

Dependence of pressure differential on coolant flow

rate in the working and “emerced” position of the

filtering element

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Perspective design of fuel assembly for a new generation RBMK-1000

Fuel pellets, enrichment 2.5%, with erbium content of 0.3% ( 935 mm long)

Fuel pellets, enrichment 3.2% with erbium content of 0.7% ( 2590 mm long)

Support grids ensuring the fuel assemble central fastening

Tailpiece-filter

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CALCULATION, ANALYTICAL AND EXPERIMENTAL WORKS FOR CALCULATION CODES UPGRADING

Development operative three-dimensional neutron-hydraulic code based on PC SADCO (introduction at Power Unit 2 of Leningrad NPP in 2010);

Development of PC and calculation models for 3D precision neutron-physical calculations for RBMK reactors by Monte-Carlo method;

Performing experimental research at TKR (fuel channel – rupture) test device (ENIC) of FC brittle rupture and possibility of dependant rupture of neighbouring channels(for U_STACK code verification);

At the PSB-RBMK test device (ENIC), a series of experiments is being performed to support RELAP5/Mod3.2 calculation code verification.

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Modeling RBMK FC brittle rupture at ТКR test device (joint research by ENIC, IPC MAE and RDIPE)

Aim: Obtaining information on the parameters of high-dynamic thermal -hydraulic and structural-mechanical processes in RBMK cladding during RC tube brittle rupture. Measuring results are intended for calculation codes verification and demonstration of the cladding behaviour and FC around the rupture in the conditions of incident with FC brittle rupture

In methodological experiments (TKR-F test device) and in full-scale experiment (TKR test device) following measurements were performed:

Thermal hydraulics Structural mechanics pressure in FC pressure fluctuation in FC temperature of medium in FC graphite cladding temperature temperature of FC emergency tube coolant flow rate in the tube of emergency FC

peripheral columns bricks movement Axial deformations of FC tubes around the rupture peripheral columns bricks accelerations

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Module of reactor cladding (MRC) of TKR test device

Experiment characteristics

Cooling channel parameters:• pressure – 8.0 МPf;• entrance temperature – 295°С;• exit temperature – 285°С;• graphite temperature – 280°С.

Emergency FC rupture occurred at the pressure of 7.97 МPa and temperature of 246°С.

• scale by leveling marks– 1:1;• number of columns – 45;• pressure in FC – to 10 MPa;• pressure under the casing – to 0.07 MPa;• temperature – to 300°С

Temperature and pressure in the emergency FC

Coolant flow rate in feeding pipeline

Modeling RBMK FC brittle rupture at ТКR test device (joint research by ENIC, IPC MAE and RDIPE)

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Research resultsRupture zone in MRC of TKR testing device Mode of pipe FC tube rupture during brittle rupture

modeling (TKR-F testing device)

Modeling RBMK FC brittle rupture at ТКR test device (joint research by ENIC, IPC MAE and RDIPE)

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Experiments at PSB-RBMK test device for thermo-hydraulic codes verification

Main actual parameters of PSB RBMK testing device:

• scale by leveling marks – 1:1 • loop number – 1 • model FA number – 4 • electric power ≈1300 КВт• coolant max. flow rate through the circuit – 67 kg/f• feedwater temperature – 155-170°С• max. pressure in separator – 10 MPa

1 - separator; 2 – process condensers; 3 – experimental channels; 4 - downcomer; 5, 6 – distributive group header (DGH); 7 - header; 8 – ECCS tanks; 9 - pumps; 10 – suction collector 17

Experiments Main results

Reactor residual heat removal during lengthy de-energizing of the plant auxiliaries, including the actuation and further non-closure of MPV

Two-phase flow in a complex circuit was modeled in natural circulation (NC) conditions, accompanied with low-frequency oscillations of flow rate. Conditions for NC failure, drying out experimental channels and channel walls and FA model temperature growth were reached.

Steam line rupture beyond the accident localization system rooms with power unit auxiliaries de-energizing

Natural circulation in a complex circuit in the conditions of sufficiently fast pressure decrease was modeled. The mode is characterized by oscillation of the whole circuit flow rate and surges of FA temperatures suppressed by ECCS model switching on.

Ruptures of collectors and feed pipelines (LC, DGH, downcomer), including partial ruptures of the DGH and modes with imposing of ECCS valves and pumps failure

Data on pressure dynamics in the circulation circuit and on separator level in the conditions of large, medium and small leaks were obtained. Fast-acting ECCS and long-term ECCS operation was modeled. Processes of FC heating, rewetting and cooling were modeled.

Experiments at PSB-RBMK test device for thermo-hydraulic codes verification

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TECHNICAL PROBLEMS OF REACTOR CORE OPERATION AT THE FINAL STAGE OF OPERATION

Exhaust movement of telescopic connection of chains (ТСC) (the largest scopes of works at LNPP-1,2 KuNPP-1; SNPP-1) Possible bending of FC cells and CPS channels (at all reactors after 35 years of operation). FC elongation (most actual for LNPP-1,2 KuNPP-1; SNPP-1, where bellow compensators of old design are installed. Less actual for other power units where only a part of compensators may be of such type). FC internal diameter increase (all power units after 20 years of operation of the second set FC).Causes:• axial radiation-thermal deformation of graphite bricks;• radiation-thermal stress accumulation in graphite bricks leading to their cracking and, as a consequence, bending of graphite columns with FC and CPS channels;• axial and diametrical deformation of FC causing the exhaust of lower bellow compensator movement, deterioration of heat removal from FAs and their vibration level increase.

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Technical measures aimed at providing operability of reactor core elements during the operational period from 35 to 45 years

1. Monitoring of graphite cladding condition, including the margin of TCC movement, bending of graphite columns, FCs and CPS channels.

2. Timely preventive elimination of the deviations detected (restoration of TCC movement margin, maintaining CPS actuator operability, replacement of bellow compensators and FC with internal diameters exceeding critical values.

3. Performing R&D works for improving FC and graphite cladding behavior forecasting; measuring quality and conditions ; reducing labour intensity and dose rates during critical parameters monitoring; specifying calculation methods and limit values for critical parameters; developing new technologies of reconstructive maintenance.

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STATUS OF THE PROBLEM OF RD300 WELDED JOINTS CRACKING UNDER IGSCC MECHANISM

1. The number of welded joints (WJ) is constantly growing due to new WJ after repairs. At all RBMK-1000 power units, in the period from 1998 to 2010, the number of WJs grew for 2865 pcs. (~20%).

2. The number of defected WJs is not decreasing. The percentage of the defected WJs from the number of those inspected:

LNPP (1st generation): 3,3 – 4,5%LNPP (2nd generation): 8,3 – 14,0%KuNPP: 3,9 – 4,7%SNPP: 1,6 – 3,2%

3. WJ inspection problems that cannot be solved for a number of years:• Lack of methods and equipment for inspecting the WJs inaccessible for UT (~3%

of the total number of WJs);• Lack of certified UT methods for automated inspection of WJs with one-sided

access (about 30% of the total number of WJs);• Unsatisfactory detectability with all the methods used of axial cracks, located

across WJs, and cracks in WJ cast metal.

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Proposals for solving the problem of IGSCC cracking of RD300 welded joints

1. Finalize technological processes of compensating measures for IGSCC preventing (high temperature thermal treatment, redistribution of residual stresses by way of mechanical weld squeezing, repair by building-up welding, upgraded welding, etc.) and repair technologies by the results of their implementation and experience of application.

2. Arrange centralized administrative and technical management of the solution of the problem of RD300 welded joints cracking.

3. Consistently, taking into consideration the determined priorities, perform “Programme of works on the completion of solving the problem of RD300 welded joints of austenite pipelines at RBMK-1000”.

4. Perform the monitoring of actual effect of the technologies introduced, for determining the possibilities to decrease in-service inspection scope and periodicity.

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Decommissioning of Beloyarsk NPP Unit 1 and 2

Elimination of safety deficits

during SNF storage in CPs

1,2

SNF preparing for shipping from Beloyarsk NPP

Preparing Power Units 1 & 2 for

decommissioning

- safety case justification for SNF storage in CPs;

- developing and introduction of neutron and gamma scanning of casings with SNF

- removal of long-sized articles from reactor vaults (technology, equipment);

- Design of support systems for cutting assemblies into fuel and non-fuel parts;

- Safety justification at the stages of SNF removal from power units

- developing a system of monitoring graphite cladding with fuel spills;

- creation and upgrading of 3D database for decommissioning

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MAIN TASKS

1. Develop and implement R&D comprehensive programme, which results will permit to improve the methods of assessment of the reactor unit critical elements residual resource at the final operation stage.

2. Using upgraded methods, develop an operation programme for each power unit permitting to provide optimal technical and economic indicators, forecast necessary scope of in-service inspection and restorative maintenance in order to ensure safety and operability of reactor core elements at each stages of additional operation of all reactor plants.

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