ISSN 0040�6015, Thermal Engineering, 2014, Vol. 61, No. 2, pp. 73–85. © Pleiades Publishing, Inc., 2014.Original Russian Text © A.I. Fedorov, S.L. Vitkovskii, I.L. Vitkovskii, V.I. Fomenko, V.F. Loskutov, R.M. Topchiyan, M.P. Nikitenko, A.V. Zhurbenko, 2014, published in Teploenergetika.

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The Novovoronezh nuclear power plant (NVNPP)is the oldest one in Russia. The main stages in theNVNPP development history and the technical char�acteristics of its power units are listed in Tables 1 and 2.

MAIN STAGES IN THE DEVELOPMENTOF THE NOVOVORONEZH NUCLEAR

POWER PLANT

In the early 1957, a decision was made to constructan NPP equipped with a VVER�type nuclear reactor(the abbreviation for the Russian name of a water�cooled water�moderated power�generating reactor)on the Don River. The choice of this type of reactorpredetermined the whole subsequent development ofthe Novovoronezh NPP. The first group of construc�tion workers arrived in the Novaya Alenovka settle�ment, the place where the NVNPP and the Novovor�onezhskii settlement had to be constructed, as soon as

in May 1957. That was the moment from which thehistory of the Novovoronezh NPP started. The param�eters of the first power unit, the type and layout ofequipment, and the building structures were designedwith certain margin for insufficient knowledge.Although the 210�MW capacity the first power unithad is quite a small one from the present�day consid�erations, but at that time this was the largest nuclearpower unit not only in the Soviet Union, but also in theworld.

Units 1 and 2. In December 1964, the NVNPP Unit 1was brought to operate at 100% of its power capacity.Operating modes with increased power outputs equalto 240 and 280 MW were tried out in the course of Unit 1operation, and a few lines of improving the power unitdesign and technology were revealed, using which thereactor power output could be increased without mod�ifying its dimensions, and which were implemented in

Fifty Years of Safe Operation of the Novovoronezh Nuclear Power Plant

A. I. Fedorova, S. L. Vitkovskiia, I. L. Vitkovskiia, V. I. Fomenkoa, V. F. Loskutova,R. M. Topchiyanb, M. P. Nikitenkoc, and A. V. Zhurbenkod

a Novovoronezh Nuclear Power Plant, a Branch of the Rosenergoatom Concern, Novovoronezh,Voronezh oblast, 396072 Russia

b Atomenergoproekt, Bakuninskaya ul. 7, str. 1, Moscow, 105005 Russiac OKB Gidropress, ul. Ordzhonikidze 21, Podolsk, Moscow oblast, 142103 Russia

d Kurchatov Institute Research Center, pl. Akademika Kurchatova 1, Moscow, 123098 Russia

Abstract—Information on the experience gained from safe operation of the Novovoronezh nuclear powerplant from the moment its first power unit was commissioned and till now is presented. The following modi�fications and design solutions on improving the equipment and systems of the Novovoronezh nuclear powerplant units that were introduced during plant operation are of special importance:(i) further development of the designs of fuel rods and fuel assemblies;(ii) improvement of the control members used in the reactor control and protection system (the specific fea�tures of the control and protection system members used in Units 1–5, and modifications made in the com�position of control member groups in Units 3 and 5 for optimizing the power density distribution;(iii) further development of fuel charge patterns;(iv) investigations carried out in the “hot chamber” at the Novovoronezh NPP (rendering support to theintroduction of new kinds of fuel);(v) further development of nuclear fuel management systems and introduction of systems for handling leakyfuel assemblies of the VVER�440 reactor;(vi) improvement of systems for cooling the reactor cores of Units 3 and 4; and(vii) further development of metal diagnostic and examination methods in revealing various types of flaws,and experience gained from the use of new equipment.Information about the modernization and service life extension of the Novovoroneh Units 3–5 beyond the30�year design service life is presented together with substantiating the residual life of power unit elementsimportant to safety.

Keywords: Novovoronezh NPP power unit, operation, repair, modernization, service life extension

DOI: 10.1134/S0040601514020037

NUCLEAR POWERSTATIONS

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the projects of Units 2–4. In April 1970, the NVNPPUnit 2 was brought to operate at 100% of its nominalpower output. The thermal power of this unit wasincreased by almost a factor of 2 as compared withUnit 1 owing to the following measures:

(i) More uniform distribution of power density overthe reactor volume was achieved by changing the prin�ciple used to control the reactor reactivity margin forburnup from a mechanical to a liquid one.

(ii) Fuel rods were shifted to operate at higher aver�age heat loads.

(iii) The heat�transfer surface was extended bymodifying the design of fuel assemblies (FAs) andincreasing the number of fuel rods from 90 to 126.

(iv) The reactor was set to operate with a higherflow rate of cooling water passing through it.

A vapor�type reactor coolant circuit pressurizationsystem was applied in Unit 2 instead of the gas�typeone that was used in Unit 1, due to which better quality

of water chemistry and better operating conditions ofthe reactor coolant circuit equipment were obtained.The experience gained from the operation of Unit 1found its reflection in the new reactor plant design (thethermal shield fastening assembly was modified) andlayout of equipment. The reactor core refueling sys�tem, the reactor coolant treatment system, the controland protection system (CPS) drives cooling system,and many other systems were refined. The Unit 2 reac�tor plant (RP) served as a basis for making a shift toconstruction of serially produced power unitsequipped with VVER reactors. In August 1984 (thetime by which the term of commercial operation of theUnit 1 reactor pressure vessel had expired), this powerunit was shut down for carrying out works on its retro�fitting and modernization. However, life altered theinitial plans of such retrofitting of Unit 1 and then ofUnit 2 as well. In 1986, after the disaster at the Cher�nobyl NPP, the safety concept of NPPs in the SovietUnion was revised, works on modernizing Unit 1 were

Table 1. Main development stages of the Novovoronezh NPP

Development stages

Station number of power unit, type of reactor

No. 1, VVER�210

No. 2, VVER�365

No. 3, VVER�440

No. 4, VVER�440

No. 5, VVER�1000

Start of construction, year 1958 1964 1967 1967 1972

Power startup, month, year September, 1964 December, 1969 December, 1971 December, 1972 May, 1980

Reaching of 100% power output, month, year

December, 1964 April, 1970 June, 1972 May, 1973 Febryary, 1981

Decommissioning, year 1984 1990 2001 (design)

2002 (design)

2010 (design)

Extended service life, year — — 2016 2017 2036

Table 2. Main technical data of the NVNPP power units

IndicatorPower unit

No. 1 No. 2 No. 3 No. 4 No. 5

Number of the reactor plant project V�1 V�3M V�179 V�179 V�187

Installed electrical capacity, MW 210 365 417 417 1000

Thermal power, MW 760 1320 1375 1375 3000

Number of circulation loops (RCPs, SGs) 6 8 6 6 4

Quantity/capacity of turbine generators, pcs./MW 3/70 5/75 2/220 2/220 2/500

Coolant flow rate through the reactor, m3/h 36500 48000 42630 42750 88900

Coolant working pressure, MPa 10.0 10.5 12.5 12.5 16.0

Maximal coolant temperature at the reactor inlet, °C 250 252 268 268 289

Average rise of coolant temperature, °C 19.1 25.8 29.1 30.1 29.5

Mass of uranium in the core, t – – 42.3 47.2 70.0

Quantity of fuel assemblies, pcs. 312 276 313 349 151

Quantity of reactor reactivity mechanical control members, pcs.

37 73 73 73 109

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stopped, and activities on transferring its systems andequipment in preserved state were commenced.Through�thickness metal samples (trepans) weretaken from the reactor pressure vessels of Units 1 and2 for experimentally studying the changes occurring inthe properties of metal under the effect of strong radi�ation fields. The results from these studies served as abasis of calculations aimed at extending the service lifeof the reactor pressure vessels of Units 3 and 4. Thework on decommissioning power units had the samelevel of complexity for the Novovoronezh NPP per�sonnel as power unit startup and extension of its ser�vice life. It fells to this plant’s lot that it was the firstone at which VVER�based power units were put inoperation; this plant was the first one at which theproblem of extending the service life of power unitswas solved; and the plant became a sort of test groundsat which the technology and methods for decommis�sioning power units are being elaborated.

Units 3 and 4. These power units are equipped withType V�179 reactor plants. Units 3 and 4 were com�missioned in 1971 and 1972, respectively. Each ofthem has an installed electrical capacity of 417 MW.The reactor plant main equipment has an assigned ser�vice life of 30 years. According to the project, Units 3and 4 had to be decommissioned in 2001 and 2002,respectively. The NVNPP power units equipped withfirst�generation VVER�440 reactors (V�179 reactorplants) were designed in the 1960s in accordance withthe general�industry safety regulations, standards, andcodes that were then in force in the Soviet Union. Thedesign solutions adopted for Unit 5 equipped with aVVER�1000 reactor (a V�187 reactor plant) related tosecond�generation reactors are to a larger extent inconformity with the main modern safety assuranceprinciples.

EXPERIENCE GAINED FROM OPERATIONOF UNITS 3 AND 4. MODIFICATIONS

INTRODUCED IN THE COURSE OF OPERATION

A large number of repair and modernization workswas carried out on these power units in the course oftheir operation, as a result of which considerableimprovements were achieved in the performance char�acteristics of the equipment.

The following works were carried out on the reac�tor, upper unit, and reactor internals:

(i) Repair has been done for the sealing surfaces ofthe reactor vessel main joints, during which the profileof grooves for nickel gaskets was changed, and themain joints of upper units have been repaired bygrooving them for removing indents from the gaskets,due to which the reliability of the reactor main jointhas been enhanced.

(ii) An additional fastening was provided betweenthe pressure vessel bottom and core barrel by installingdowels for reducing loads exerted on the reactor inter�nals under the effect of coolant flow.

(iii) Leak limiting devices were installed on theupper unit thermal monitoring (TM) sleeves withDnom = 100 mm for decreasing coolant leaks in case ofTM sleeve ruptures within the limits of a design�basisaccident.

The heat release channels were fitted with sensorsfor operatively measuring the neutron flux over thecore height, and hydrogen detectors were installedunder the reactor head. The flow section was increasedin the throttle orifices used in the Unit 4 reactor fuelassemblies (FAs) for optimizing the flowrates throughthe FAs and preventing surface boiling on fuel rods,and for reducing the corrosion product depositionintensity on the fuel rod surfaces. In order to ensuremore reliable sealing and reduce stresses in the fasten�ers of joints, the existing gaskets of seals in the jointsbetween the automatic control assembly (ACA) casingand drive and in the ACA air vents were replaced bythose made of expanded graphite (a new sealing mate�rial).

The steam generators (SGs) were fitted with threeadditional level measurement channels, and the num�ber of interlocks ensuring safer operation of the steamgenerator itself and of the steam turbine unit wasincreased. Additional sleeves were installed in thelower generatrix of steam generator shells for removingsludge accumulating during operation. The joints ofthe primary and secondary coolant circuits were fittedwith gaskets made of expanded graphite for achievingmore reliable sealing of the joints and reducing stressesin the fasteners. The SG blowdown pipelines on thesecondary coolant side were replaced by those made ofGrade 0Kh18N10T steel, due to which it became pos�sible to completely exclude off�scheduled shutdownsof the power unit due to through�the�wall flaws thatappeared in the pipes made of pearlitic steel.

The cast impellers that were used in the reactorcoolant pumps (RCPs) and which were prone to crackformation were replaced by stamped�and�weldedones. Additional elements were installed in the inde�pendent loop pipework, due to which it became possi�ble to place supplementary thermocouples. As a result,a larger number of measurement signals was obtained,which opened the possibility to organize a multichan�nel structure of interlocks.

The assembly for injecting medium into the pres�surizer was replaced by a multicascade injection sub�system, which made it possible to exclude failures ofthe system caused by failure of one active component(a single�cascade injection arrangement was initiallyused in the project).

The K�220�44 turbine unit steam reheating sys�tems were subjected to retrofitting in the turbine setsnos. 11 and 12 (TA�11 and TA�12):

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(i) The existing separators were replaced by theirmodernized (cyclone�type) versions.

(ii) The baskets and distributing pipe were replacedby those made of erosion�resistant material.

(iii) The rotary�type level control valves used in thefirst� and second�stage reheat system hot wells werereplaced by sliding�vane control valves with bodiesmade of erosion�resistant material.

The high�pressure regeneration systems of K�220�44 turbine units were subjected to retrofitting by mod�ernizing the high�pressure heater (HPH) pipe sys�tems, in which the feed water supply arrangements tothe coils were changed, and the HPH design waschanged from the two�pass to a single�pass one. Therotary�type level control valves in the HPH�6 to HPH�8heaters with bodies made of Grade 25l steel werereplaced by sliding�vane control valves with bodiesmade of erosion�resistant material.

In view of erosion wear that was revealed on theinlet edges of the last�stage low�pressure rotor bladesof the K�220�44 turbine unit, the TA�11 low�pressurecylinder (LPC) was subjected to modernization,which involved replacement of the flow path and last�stage blades of the TA�12 low�pressure rotor (LPR)followed by strengthening of blade edges using theKol’chuga electric spark method. The turbine high�pressure cylinder (HPC) was also subjected to mod�ernization, during which an additional guide vane wasinstalled upstream of the high�pressure rotor (HPR)first stage. In addition, the condenser tube systemswere replaced by tube systems made of MNZh�5�1copper–nickel alloy.

Certain replacements were done in the low�pres�sure regeneration system:

(i) The tube systems of the LPH�4 and LPH�5 low�pressure heaters were replaced by tube systems madeof MNZh�5�1 copper–nickel alloy.

(ii) The tube systems of the LPH�1–LPH�5 heat�ers in turbine unit no. 11 were replaced by those madeof erosion�resistant material.

(iii) The NPND�4 pumps in TA�11 and TA�12were replaced by AKsV�125�140 pumps with casingsmade of Grade 20 steel.

(iv) The rotary level control valves with bodiesmade of Grade 25l steel used in the LPH�2–LPH�5heaters were replaced by disk�type control valves.

EXTENDING THE SERVICE LIFEOF UNITS 3 AND 4

The NVNPP Units 3 and 4 were the first ones of thecommercial power stations constructed according tothe projects and using the equipment produced byenterprises of the former Soviet Union that hadreached a 30�year service life in 2001 and 2002. Workson preparing these power units for extension of their

service life were commenced in 1995. In the course ofthese works, the main service life extension principleswere formulated, and the service life extension meth�odology was elaborated, which were then used as abasis of the set of regulatory documents developed bythe state regulatory authority in the field of usingatomic energy. To ensure the possibility of operationduring the extended service life, the following mod�ernization activities were developed and implementedin the period 1999–2002:

—organizing a train�wise structure of safety sys�tems;

—making improvements in monitoring and con�trol systems;

—enhancing the reliability and fire safety of equip�ment;

—reducing the amount of radioactive releases;—increasing the spectrum of design�basis acci�

dents; and—enhancing fire safety.The emergency power supply systems were fitted

with additional storage batteries and direct currentboards; the existing first�category reliable AC powersupply network was modernized; two additional die�sel–generators (DG�7 and DG�8) with a capacity of1600 kW each were installed; and the second�categoryreliable 6 kV power supply network was modernizedfor organizing two trains of safety systems (SSs).

Two sets of reactor emergency protection andmonitoring devices were organized in the control andmonitoring systems constructed using the modern setof NFME�7 neutron flux monitoring equipment; twosets of process parameter�based emergency protection(within the reactor control and protection system)were installed, the impulse reactor power controllerwas replaced by a 5SRV automatic power controller;two sets of the reactor power setback and limitingdevice ROM�2SRV were installed, as well as two uni�fied sets of hardware intended for implementing logiccontrol, protection, interlocks, and alarm devices fortwo trains of safety systems.

The reactor plant is fitted with an in�core instru�mentation system constructed using the SVRK�V179equipment.

The scope of works on modernizing the processsafety systems included modernization of the reactorcoolant circuit emergency makeup system for organiz�ing two SS trains; the sprinkler system was also mod�ernized for organizing two SS trains; fast�acting stop�and�cutoff valves were installed in the main steamlines; the lever�type safety valves on the steam genera�tors and on the pressurizer were replaced; the essentialservice cooling water system was modernized for orga�nizing two SS trains in each power unit, and the com�ponents and equipment of sealed premises were mod�

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ernized. The power units were fitted with additionalsystems for supplying feed water to the steam genera�tors, and measures for ensuring brittle strength of thereactor pressure vessels were implemented. With theabove�mentioned measures on modernizing Units 3and 4 implemented, it became possible to widen thespectrum of design�basis accidents from the reactorcoolant pipeline rupture with an equivalent diameterof 32 mm to that with an equivalent diameter of100 mm. In scenarios with beyond�design�basis loss�of�coolant accidents (LOCAs involving reactor cool�ant pipeline ruptures with Dnom = 200 mm and larger),technical and organizational measures were taken tolimiting the radiation effect of the NPP on the person�nel, population, and environment. Deviations of cate�gories 3 and 4 (according to the IAEA classification)from the requirements of regulatory documents onsafety were removed. According to the results of theperformed Level 1 probabilistic safety assessment, thecore damage frequency was decreased from 1.08 ×10–3 1/reactor�year to 3.44 × 10–5 1/year in Unit 3 andto 5.12× 10–5 1/reactor�year in Unit 4.

With the program of extending the service life ofthe NVNPP Units 3 and 4 implemented, an essentiallyhigher level of their safety was achieved, and safety ofthe power units was substantiated for another 15 yearsof their operation. The main conditions subject to

which continued operation of Units 3 and 4 is permit�ted in the extended period of their service life are con�tinuation of works on enhancing the current levels ofsafety and reliability, and management of the equip�ment remaining life. Since 2009, works have beencommenced on estimating the possibility of furthersafe operation of Units 3 and 4 beyond the extended15�year life.

EXPERIENCE GAINED FROM THE OPERATION OF UNIT 5. MODIFICATIONS INTRODUCED

IN THE COURSE OF OPERATION

Unit 5 is the pilot one in the series of second�gen�eration VVER�1000 reactors (a V�187 reactor plant).The technical solutions implemented in this powerunit served as a basis for designing, construction, andoperation of power units equipped with V�302 and V�320reactor plants commissioned at NPPs in Russia,Ukraine, Czech Republic, and Bulgaria. The generalview of Unit 5 commissioned on September 25, 1980is shown in Fig. 1. Among the organizations that par�ticipated in the development of this power unit wereAtomteploelektroproekt, OKB Gidropress, and Kur�chatov Institute of Atomic Energy. The assigned ser�vice life of the power unit main equipment is 30 years,which expired in 2010.

Fig. 1. General view of the NVNPP Unit 5.

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The scope of works carried out on the reactor upperunit in 2004–2005 included replacement of 109 CPSsleeves, which had to be done due to the occurrence ofcracks in their lower part along the welding boundarybetween ZIO�8 surfacing and Grade 20 steel caused bya joint effect of welding stresses resulted from makingand repairing the surfacing at the manufacturer’s, andthe operating loads.

The PGV�1000 steam generators were replaced byPGV�1000M steam generators in 1988–1989 becausecracks were revealed in the SG headers. Through�thickness flaws in weld joint no. 111 were revealed forthe first time in SG�1 in 1998, and subsequently simi�lar flaws were recorded in SG�3 in 2001, then onceagain in SG�1 in 2005, and then in SG�2 and SG�4 in2007. A technology for removing flaws was elaborated,and the appropriate repair works were carried out. Theflaws that appeared in weld joint no. 111 were attrib�uted to stress corrosion cracking.

Erosion damage that appeared in the steam reheatsystems of the K�500�60/1500 turbine units led to theneed to replace the SPP�1000 separators (film�typeseparators of improved design were installed); theclusters of heating sections were replaced by thosemade of stainless steels; the dispenser pipes werereplaced by those made of stainless steel; cyclone�typeseparators were installed on the first�extraction steamline to the SPP�1000 first stage; phase separators wereplaced in the lines removing separated moisture fromthe film separators and from the first�extractioncyclone separator. In view of poor reliability of reheatsteam cutoff valves, gate valves were installed [stop andcontrol ones connected in series downstream of eachmoisture separator/reheater (MSR)]. To avoid non�uniform distribution of steam among individual sepa�ration packs and erosion of the separator shell, a flat�tening device (made in the form of a perforated stain�less steel barrel) was placed inside the separator.

In the low�pressure regeneration systems, the tubesystems of LPH�4 and LPH�3 heaters and of theLPH�4 drain cooler were replaced by tubes made ofGrade 08Kh14MF nickel�free stainless steel.

Two separate hydraulic control lines and a dedi�cated system for switching these lines that were usedinitially in the automatic control and protection sys�tem of the K�500�60/1500 turbine units failed toensure bumpless switchover from the main turbinecontrol system to the backup hydraulic control system.In the course of planned preventive repair in 2004,works on modernizing the system were carried out,during which the switching devices were removed inboth turbine units.

In view of the fact that distillate leaks occurredfrom the stator windings, and that the fastenings oftheir end�face parts lost their tightness, the turbinegenerator stators were replaced by those with a mod�

ernized winding. The occurrence of distillate leaksfrom the bus ducts and field winding, which caused adrop of insulation resistance led to the need to replacethe water�cooled rotors by rotors with direct hydrogen(gas) cooling of the winding produced by Elektrosila.The above�mentioned works accomplished, the num�ber of failures occurring in the Unit 5 turbine genera�tors dropped significantly. The storage batteries usedin Unit 5 were of nonseismic�resistant design, due towhich leaks of electrolyte often appeared over the stor�age cells vessels. In view of this, these storage batterieswere replaced by seismic�resistant VARTA storagebatteries (produced in Germany) installed on seismic�resistant racks. The uninterruptible power supply units(UPSs) for the first�group loads produced in Tallinnhad a number of failures and were replaced by UPSsproduced by PO Elektropreobrazovatel’ (the city ofOrenburg).

In view of poor reliability during operation (a lowpulling force was developed for lifting the CPS controlmembers, rubbing of the armature occurred in thedrive bearings, the electric motor suffered from open�circuit faults of its windings, and many short�circuitfaults occurred in the power connectors), the design ofthe linear step shifter’s drive was modernized in thethermal automatic control equipment. As a result ofthe performed modernization, the graphite bearingswere replaced by so�called termar ones, the CPSpower supply board was shifted to operate at 36 DC Vvoltage, the Type LShP drives were replaced by TypeLShP�M drives, and the CPS cable connectors werereplaced by ring ones. The insertion speed of CPScontrol members after actuation of the third�kindemergency protection and power limiting deviceturned to be insufficient in case of disconnection ofone or two reactor coolant pumps, one or two turbine�driven feedwater pumps, and one or two turbine sets orturbine generators. Such operating conditions led tooperation of the first�kind emergency protection. Theproject of the accelerated reactor unloading systemwas put in use.

Unique design and engineering solutions were putin use in process systems: a system for returninghydrogen to the reactor coolant circuit, due to whichmuch less quantity of ammonia had to be metered intothe reactor coolant system, and much less amount ofgas and aerosol emissions was achieved; and a technol�ogy for reprocessing liquid radioactive wastes using thedeep evaporation method, due to which it becamepossible to achieve an essentially smaller volume ofliquid radioactive wastes stored in power unit auxiliaryfacilities.

EXTENSION OF THE UNIT 5 SERVICE LIFE

The results of efforts aimed at achieving more reli�able and stable operation of Unit 1 equipped with a

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VVER�1000 reactor created prerequisites for continu�ing its operation after the 30�year service life assignedby its project. The analysis of the extent to which Unit 5conforms to the existing codes and regulations thatwas carried out in 2004 revealed the unit’s main defi�ciencies in terms of safety that had to be removed forsubstantiating the possibility of power unit operationin an additional period of time. First, this was insuffi�cient physical separation of active, support, and con�trol safety systems, due to which the occurrence ofcommon�cause failures was not excluded. Second, theinitial project did not incorporate safety systems thenecessity of which is stipulated in the modern regula�tory documents (a system for emergency removal andmonitoring of hydrogen concentration in the sealedvolume, an industrial seismic protection system, andso on). And third, the initial designs of systems impor�tant to safety were not in conformity with the modernregulatory documents. In parallel with analyzing theextent to which the power unit was in conformity withthe requirements of modern regulatory documents,the power unit safety prior to implementation of themodernization activities was evaluated using the prob�abilistic assessment methods with determining themajor contributors in the core damage frequency(CDF), the value of which before the modernizationwas estimated at 2.24 × 10–4 1/(reactor�year). The dia�gram characterizing the relative contribution of themain groups of initiating events in the CDF is shownin Fig. 2. The results of the performed analyses servedas a basis for elaborating a set of measures on modern�izing the power unit for ensuring its safe and reliableoperation in the additional period of operation. In2009, a comprehensive examination of the power unitwas finished, which showed that the NVNPP Unit 5components, systems, civil constructions, buildings,and structures, and the power unit as a whole were insatisfactory state, that they were in compliance withthe requirements of the relevant operating, project,design, and regulatory documents, so that their furtheroperation is possible, and that extension of the powerunit service life is technically possible provided thatthe remaining life of the power unit equipment impor�tant to safety is duly substantiated. The remaining lifeof this equipment was analyzed; the list of componentsthat had worked out their service life and had to bereplaced was drawn from the results of that analysis,and it was determined that the reactor pressure vesselservice life can be extended to 56 years, and that thelife of the remaining nonreplaceable components canbe increased to 60 years. For substantiating the possi�bility of further safe operation of the power unit, aReport on In�Depth Safety Assessment of Unit 5 Tak�ing into Account Measures on Modernizing andReplacing the Equipment was developed, which in2009 was submitted to Rostekhnadzor within the set of

documents substantiating safety to get a license forpower unit operation beyond the 30 year service life.

The list of main activities on modernizing thepower unit performed during its outage for plannedpreventive maintenance in 2010–2011 included thefollowing ones:

—physical separation of the emergency powersupply systems involving full replacement of powercables, construction of a new standalone building ofthe standby diesel power station, and shifting of the 6and 0.4 kV power supply sections of the respectivetrain of the emergency power supply system to a newsafety building;

—full replacement and physical separation of theequipment and cable routes of control safety systems;

—full replacement of the CPS equipment withorganizing a two�set emergency protection system;

—development of an emergency boron injectionsystem for injecting concentrated solution of absorberat the nominal parameters of the reactor coolant cir�cuit;

—modernization of the high�pressure emergencycore cooling system with the possibility of connectingthe system pumps to the sump in the sealed volume;

—development of an emergency feedwater systemcomplying with the modern safety requirements;

—replacement of the pressurizer pilot�operatedsafety valves by new ones with the possibility of imple�menting the feed and bleed mode;

—commissioning of an emergency hydrogenremoval and hydrogen concentration monitoring sys�tem in the containment sealed volume for design� andbeyond�design�basis accidents;

Ruptures of secondarycoolant system

Transients(20%) Reactor coolant

leaks inside the

Primary�to�secondary leaks (3%)

Loss of offsitepower (8%)

Leaks outside the containment (3%)

Fig. 2. Relative contribution of the main groups of initiat�ing events in the core damage frequency.

pipes (39%) containment (27%)

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—introduction of an industrial seismic protectionsystem with the actuation setpoint corresponding to4 points on the MSK�64 scale;

—development of a train of the reliable power sup�ply system for normal�operation equipment importantto safety based on one of the former emergency powersupply system trains;

—development of an auxiliary feedwater systembased on the former emergency feedwater system;

—full replacement of the equipment and cableroutes of normal�operation control systems;

—replacement of the generators that had fullyworked out their service life by new ones;

—modernization of the main and emergency con�trol rooms with almost 100% replacement of theirequipment; and

—construction of full�scale training simulators ofthe main and emergency control rooms.

The following positive results in terms of safetylevel enhancement were obtained:

—Unit 5 was brought in conformity with the mod�ern requirements of Russian codes and regulations andIAEA standards.

—The CDF value was decreased from 6.9 ×10–4 to 2.9 × 10–5 1/(reactor�year), which is in confor�mity with the IAEA recommendations.

—According to the results of the internationalinsurance inspection carried out in 2011, Unit 5 com�plies with the world�wide insurance risk requirementsfor second�generation NPPs.

IMPROVEMENT OF USING THE NUCLEAR FUEL OF VVER REACTORS

The Novovoronezh nuclear power plant being thepilot NPP equipped with VVER reactors, tests ofnuclear fuel were carried out in the reactor cores ofUnits 1 and 4, aimed at improving the design andachieving more reliable operation of fuel rods and fuelassemblies (FAs) and their performance characteris�tics. At the stage of mastering the production and useof first�generation VVER�210 and VVER�365 reac�tors, the state of fuel under real conditions of its oper�ation was studied. The results of these activities consti�tuted the basis for developing the design of fuel rodsand FAs for the second�generation VVER�440 reactor.The results of reactor tests and post�reactor investiga�tions were taken into account in the design of fuel rodsfor a VVER�1000 reactor. The following technologicaland design improvements were implemented based onthe results of post�reactor studies:

—The content of moisture (which was notchecked at the initial stage) and fluoride in the fuel wasdecreased in order to achieve lower susceptibility of

fuel rods to loss of tightness caused by corrosion dam�age inflicted to their claddings.

—The fuel manufacturing technology waschanged for decreasing the fuel temperature and itsswelling: pellet fuel with a central hole is producedinstead of rod�shaped fuel.

—The optimal gap between the fuel and claddingwas selected by decreasing the tolerance for the fuel pelletand cladding diameters (it was brought to 0.03 mm).Fuel pellets began to be made with chamfers toexclude crumbling of fuel in a fuel rod and to weakenthe mechanical interaction of fuel and cladding (the“ratchet” effect), due to which the claddings becamemuch less prone to loss of tightness.

—The fuel density was increased, and the pelletmicrostructure was optimized, including the grainsize, porosity, and open porosity.

—The initial pressure of helium in fuel rods wasincreased to atmospheric pressure, due to which thefuel temperature was decreased by approximately300°C.

—A changeover was made from sealing fuel rods bymeans of four�seam electron�beam welding to sealingthem by means of two�weld butt�contact welding andby using a new design of plugs. Based on the studyresults, the manufacturer was informed about the pos�sibility of cladding areas adjacent to the plugs becom�ing contaminated by the material of the weldingmachine’s copper journals followed by damaging theseareas during fuel rod operation.

—The fuel pole fixing member with a split bushingwas replaced by a spring one. When variable thermalelongations of the pellet stack occurred in a fuel pole,the split bushing shifted upward, due to which largegaps appeared between the fuel pellets in the next cycleand, as a consequence, local bursts of neutron flux.

—The milled upper grid with holes 5 mm in diameterfor the fuel rod upper tips was replaced by a cellular grid.As a result, with increasing the temperature and radia�tion, the axial gap was kept no less than 25 mm.

—A new design and manufacturing technology ofzirconium spacer grids were developed instead of gridsmade of stainless steel (three fuel assemblies that hadbeen in operation for one, two, and three years wereinvestigated), which made it possible to avoid parasiticcapturing of neutrons by structural material. The stiff�ness of the entire grid, the elastic properties of cells,the strength of cell contact welding, and cell hydrationareas were determined during these investigations.

—In order to decrease leaks between assemblies,the casing tube wrench size was increased to 145 mm,and the wall thickness was decreased to 1.5 mm.

—To exclude a burst of neutrons in surroundingassemblies from the CPS assembly docking unit,

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FIFTY YEARS OF SAFE OPERATION OF THE NOVOVORONEZH NUCLEAR POWER PLANT 81

hafnium plates were installed in the CPS FA upperpart.

—Treatment of casing tubes by autoclaving wasreplaced by the technology of subjecting them toforced oxidation in air. This made it possible to essen�tially shorten the casing tube manufacturing cyclewhile retaining the protective film on the tube surface.

—Treatment of fuel rods by autoclaving wasreplaced by the anodization technology.

—Fuel enrichment with 235U was increased from1.5% in VVER�210 reactors to 4% in VVER�440 reactors.

—The fuel operation cycle and its burnup wereincreased: a four�year fuel cycle was put in use aftercarrying out life tests of an assembly to the burnupequal to 50.5 (MW day)/kg U.

—Fuel enrichment profiling over the fuel assemblysection was applied.

—Materials on increasing the service life ofabsorbing extensions to 20 years against the initiallyassigned life equal to 3 years were substantiated, due towhich the amount of radioactive wastes was decreasedsignificantly.

—The service life of state CPS absorbing assem�blies (AAs) of a VVER�1000 reactor with boron car�bide operating in the automatic control mode wasincreased to two years as compared with their initiallyassigned service life equal to one year.

—In�pile and post�pile tests of the VVER�1000reactor CPS AAs with the use of dysprosium titanateas absorbing material were carried out, after whichthese devices with a service life of up to 10 years wereput in use in all VVER�1000 reactors.

Owing to the improvements made to the design ofassemblies and fuel loading arrangements, the averageburnup of the unloaded fuel was increased from thedesign 28.4 to 39–40 (MW day)/kg U in VVER�440reactors and from the design 40.0 to 46.0 (MW day)/kg Uin the VVER�1000 reactor. In the course of puttinguranium�gadolinium fuel in use in Unit 5, the operat�ing limits with respect to local parameters limiting thepermissible linear power density and relative heatrelease in fuel rods were developed and introduced forthe first time for VVER reactors. As a result, it becamepossible to weaken the conservatism of the designoperational limits by 4% and widen the possibilities ofselecting fuel charges according to the time of theiroperation and the amount of makeup fuel. Along withthe state procedures for checking the tightness of fuelrod claddings, means and procedures improving thequality of identifying FAs containing leaky fuel rodsare adapted, developed, pass experimental operation,and are introduced at the NVNPP. On the whole,works on development of methods for monitoring thetightness of fuel rod claddings in FAs are carried out intwo lines of achieving more efficient monitoring of

cladding tightness (CTM): on the operating and shut�down reactor.

The combined use of the developed procedures formonitoring specific activity of inert radioactive gasescontained in the coolant circulating in the reactorplant primary circuit and methods for monitoring thespecific activity of iodine radionuclides makes it pos�sible to achieve more reliable assessment of the currentstate of fuel with the reactor plant operating in apower�generating mode. An expert system for estimat�ing the level of defects in the reactor core is passingexperimental operation. Such assessments have to becarried out in scheduling reactor plant shutdowns andin planning transient modes of reactor plant operationfor reducing risks of loss of fuel rod cladding tightness.The tightness of fuel rod claddings in the shut downreactor is commonly checked using the water methodin the bottles of the defective assembly detection sys�tem by the activity of reference radionuclides 131I,134Cs, 137Cs. To achieve more accurate detection ofleaky fuel rods, the procedure of additionally using theshort�lived radionuclides 136Cs and 133Xe as referenceones has been put in use. At present, works on com�missioning a system for carrying out operative CTMusing the sipping method (the MP�1000 CTMS)directly during the refueling process are currentlyunderway in Unit 5. Rearrangement of CPS CMgroups in Unit 5 and control algorithms is necessaryfor improving the power unit dynamic characteristicsand for replacing the CPS CMs with half�lengthabsorbing rods by control members with full�lengthabsorber rods. The control algorithms are developedtaking into account the experience gained from theintroduction of improved control algorithms at theRostov, Khmelnitsk, Kalinin, and Tianwan NPPs. Theabove�mentioned improved algorithms are essentiallya set of modern methods for control of heat release inthe core of a VVER�1000 reactor. The followingparameters were selected as criteria for safe applica�tion of the algorithms:

(i) limitations on the positive reactivity insertionrate in withdrawing the control groups of CPS CMswith the working speed in arbitrary and statesequences;

(ii) the emergency protection efficiency;(iii) the value of positive reactivity when one CPS

CM is ejected; and(iv) the recriticality temperature.To achieve more uniform distribution of flowrates

through ACA assemblies and FAs in the cores of Units 3and 4, the flow sections of throttle orifices in the corecage were increased. The work on changing the throt�tle orifice flow sections in Unit 3 was carried out bychanging the throttle orifices or by twisting�in insertsinto the cells the diameter of throttle orifices in whichwas changed from 56 to 52 mm. The optimal size of

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FEDOROV et al.

throttle orifices in Unit 4 was obtained by boring themfrom 45 to 50 mm.

As regards matters concerned with handling ofFAs, a criterion was developed for the first time in thenuclear industry, according to which an FA containingleaky fuel rods has to be placed into bottles if the fol�lowing signs of leaks are simultaneously detected inCTM samples (in the same assemblies):

(i) the specific activity levels of at least one of thereference isotopes (131I, 134Cs, and 137Cs) exceed 1.0 ×10–3 Ci/kg (the CTM was carried out according to thestate procedure) and

(ii) there are statistically significant (differing frombackground values) quantities of radionuclides 144Ceand 106Ru with confirmation that these fission prod�ucts belong to the fuel rods in the assembly beingchecked.

The developed project of the system for storingspent fuel assemblies (SFAs) of a VVER�440 reactorwith leaky fuel rods in bottles is intended for storingSFAs with leaky fuel rods the leakage extent of whichexceeded the bottle storage criterion. The systemensures zero exchange of water between the fuel poolwater medium and bottle water medium during nor�mal operation.

IMPROVEMENT OF THE WATER CHEMISTRY IN THE PRIMARY AND SECONDARY

COOLANT CIRCUITS FROM VVER�210 TO VVER�1000 REACTOR

The NVNPP Units 1 and 2 were experimental�commercial ones, and Units 3, 4, and 5 are the pilotones in their series. From the time of their commis�sioning, the water chemistry of the primary (reactor)and secondary coolant circuits was regulated only bythe project and design documentation, whereas stan�dards were developed in the course of operation. Asmore insight was gained with regard to the process, theneed to revise the normative documents regulating thewater chemistry quality indicators was revealed. Tech�nical solutions allowing improvements to be made inthe means for maintaining water chemistry wereadopted, and modern instruments helping to makechemical monitoring more reliable were put in use.

Austenitic stainless steel served in Unit 1 as struc�tural material contacting with reactor coolant (exceptfor some parts of the reactor coolant pumps andmakeup pumps). The reactor pressure vessel also had a20�mm�thick surfacing of stainless steel, due to whichneutral (correction�free) water chemistry could bemaintained. To suppress excess oxygen, small quanti�ties of ammonia and hydrazine�hydrate were meteredin the primary coolant circuit makeup water, and since1966, only hydrazine�hydrate was metered with main�taining its concentration in makeup water at a level of

100 ppb. The residual content of oxygen under suchconditions did not exceed 0.02 ppm.

The pressure vessels of the reactors and pressurizersin the NVNPP Units 2–4 do not have stainless steelsurfacing, due to which more stringent requirementsare imposed on the water chemistry parameters. Inaddition, starting from Unit 2, the reactivity controlmethod by means of adjusting boric acid concentra�tion began to be used, which made it possible toachieve better power density distribution over the coreand thereby create a margin for reactor thermal power.However, the presence of boric acid in the reactorcoolant led to the need to introduce certain changes inthe water chemistry requirements. The ammonia–potassium water chemistry was adopted, which has notunderwent any fundamental changes in Units 3–5 upto now. The range of standardized concentrations ofalkali metals as a function of current boric acid con�centration was changed considerably during thatperiod in order to reduce the corrosion rate of struc�tural materials and to minimize mass transfer. In Units 3and 4, various water chemistry regimes were elabo�rated in the standardized range of concentrations, andthe optimal level of concentrations was determinedexperimentally. Improvements were made in thearrangement for removing excessive alkalinity by addi�tionally charging cation�exchange resin to the mixed�bed filter in the SVO�1 radioactive water treatmentsystem, which made it possible to maintain the waterchemistry in a smoother manner. The curves in Fig. 3show the total molar concentration of alkali metals inthe reactor coolant of Units 3–5 vs. the current con�centration of boric acid. An arrangement for returninghydrogen into the reactor coolant was implemented inUnit 5. As a result, it became possible, along withmaintaining the hydrogen concentration in the cool�ant in a more stable manner, to keep the equilibriumconcentration of ammonia at a lower level and,accordingly, to achieve more efficient performance ofthe ion�exchange charge and a smaller amount ofradioactive wastes. A high concentration of ammoniain the preserving solutions and low transparency offuel pool water were a serious problem in Units 3 and4 during their being in cold state. Investigations aimedat addressing this problem were carried out, based onwhich the following solutions were adopted: the low�ammonia water chemistry is maintained in the shut�down mode, and the water of the refueling and spentfuel pools is purified in a UWF�120�4 filtration plant(Fig. 4) and in the SVO�4 system anion�exchange fil�ter charged with nuclear�grade acryl anion�exchangeresin Amberlite IRN 67R having high selectivity toabsorb finely dispersed particles.

The water chemistry of the secondary coolant cir�cuit underwent, as that of the primary coolant circuit,significant changes for 50 years of operation. From thetime of Unit 1 commissioning, correction�free water

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FIFTY YEARS OF SAFE OPERATION OF THE NOVOVORONEZH NUCLEAR POWER PLANT 83

chemistry was initially kept in this power unit. Lateron, corrective additions were introduced, such asammonia, lithium hydroxide, and hydrazine�hydrate,

the use of which made it possible to minimize the cor�rosion processes. Since 1993, the SG heat�transfersurfaces of Units 3–5 on the side of the secondarycoolant circuit are subjected to operations for chemi�cally washing them from deposits during planned pre�ventive repairs, which are carried out by means ofunique technologies with the use of complexons.

In 1995, the Unit 5 feed water system was retrofit�ted with organizing a salt compartment in the steamgenerators, due to which more efficient removal ofsalts with blowdown water was achieved and, accord�ingly, a smaller content of salts was obtained in the SGmain volume. Experimental commercial operationwith correction treatment of the secondary coolantcircuit working fluid with ethanolamine has been con�ducted in Unit 3 since 2010 and in Units 4 and 5 since2011. An analysis of conducting the secondary coolant

8.40

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Month

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water chemistry

Fig. 5. Variation with time of the pH values of feed waterand blowdown water with different water chemistries.(1) Feed water and (2) blowdown water. PPM is plannedpreventive maintenance.

Fig. 6. Flaw in a weld joint.

2010

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Fig. 4. Characteristics of the Unit 4 refueling pool waterpurification plant and results of its operation. (1) Waterclarity, (2) approximation of dependence (1) by a logarith�mic function, and (3) pressure difference across the purifi�cation plant.

0.2

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Fig. 3. Total molar concentration of alkali metals in thereactor coolant of Units 3 and 4 (a) and of Unit 5 (b) vs. thecurrent concentration of boric acid. (1) Range (maximaland minimal values) of standardized total molar concentra�tion of alkali metals before introduction of new norms forthe primary coolant circuit, (2) range (maximal and mini�mal values) of standardized total molar concentration ofalkali metals according to STO (Industry Standard)1.1.1.02.005.0621�2007 (a) and STP (Enterprise Standard)EO 00004�00 (b), and (3) optimal values of the total molarconcentration of alkali metals in accordance with STO1.1.1.02.005.0621�2007 (a) and STP EO 00004�00 (b).

3

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water chemistry involving experimental�commercialmetering of ethanolamine confirmed the expectedadvantages of this water chemistry as compared withthe all�volatile water chemistry. With ethanolaminemetered into the secondary coolant circuit, the рН25

values have become more uniform in the flows of thesecondary coolant system and are stably maintained inthe range 9.0–9.3 in the feed water and in the range9.2–9.5 in the SG blowdown water. Figure 5 showshow the pH values of feed water and blowdown watervary with time for different water chemistries. Until1996, the operations on making up and filling the pri�mary and secondary coolant circuits were carried outusing water that had passed two�stage demineraliza�tion. After the new standards on the quality of chemi�cally demineralized water had been put in force, a

need arose to construct an additional treatment stage,and in 1996 the third stage of chemical water treat�ment with mixed�bed filters was put in operation.Since 2005, a new chemical water treatment plantserving the needs of Units 3–5 has been in operation,which was designed in accordance with the Europeanstandard with a high degree of automation. An infor�mation�analytical system for displaying and process�ing the results of water chemistry monitoring at all pre�determined check points and for all monitored qualityindicators of water coolant and auxiliary water mediain different operating modes of the power units wasdeveloped and put in use for the primary and second�ary coolant circuits of these power units. At present,works are underway at the NVNPP on renewing thefleet of instruments and on introducing an automatedchemical monitoring system for the Unit 5 primaryand secondary coolant circuits. As a result, stablemaintenance of the most important characteristics ofthe primary and secondary coolant circuits has beenensured for 50 years of operation, and their levels havebeen brought maximally close to the optimal values.

IMPROVEMENT OF THE METAL DIAGNOSTIC AND EXAMINATION

METHODS

For achieving better quality of revealing flaws inmetal, work is being constantly carried out at theNVNPP for introducing modern means and new pro�cedures for examination of metal. In particular, newultrasonic examination instruments and systems, dig�ital radiographic systems, and automated installationsfor processing X�ray shadowgraphs, and new develop�ments in the field of eddy�current examination havebeen put in use. In 2007, automated ultrasonic exam�ination of the composite welded joints no. 23 connect�ing Dnom = 100 mm sleeves with the PGV�4M steamgenerator header was carried out for the first time inUnit 3. As a result, a flaw was timely revealed in thesteam generator, which could soon lead to the weldedjoint failure (Fig. 6). In order to perform more rapidexamination of metal, the NVNPP management initi�ated development of a digital radiographic system withthe use of a matrix detector; this work was entrusted to

(a)

(b)

(c)Fig. 7. Operation with the use of matrix detector. (a) Digi�tal radiographic system, (b) matrix detector, and (c) viewof weld joint on the display screen.

Fig. 8. Check sample of an SG heat�transfer tube with sup�port plates.

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FIFTY YEARS OF SAFE OPERATION OF THE NOVOVORONEZH NUCLEAR POWER PLANT 85

specialists of the Diagnostica�M company. A matrixdetector is used in this system instead of X�ray film orplate (Fig. 7). This instrument allows an image (an X�ray shadowgraph) to be obtained on the display screenwithin 30–40 s. The Novovoronezh NPP was selectedas the pilot facility for introducing a system for carry�ing out eddy�current examination of SG heat�transfertubes and shunt pipes between the SG headers. Exam�ination carried out with the use of such probe allowsthe revealed indications to be evaluated in a moretrustworthy manner. For example, the use of a matrixsensor in examining a sample with support plates(Fig. 8), it is possible to recognize not only pitlikeflaws of the tube appeared at the edge of supportplates, but also to see the outlines of the plates them�selves.

Thus, the Novovoronezh NPP equipped withVVER reactors played one of the key roles when thenuclear power industry of the Soviet Union and Russiawas in its making and development stages. The 50�yearexperience gained from operation of VVER reactors

has demonstrated that power units built around VVERreactors can operate for a long period of time and thatthis technology is a safe one. This experience alsoserves as practical evidence of the fact that the scien�tific�technical achievements and practical solutionsthat were adopted 50 years ago were correct. TheNVNPP power units served as facilities on which thefundamental solutions proposed by plant and equip�ment designers were elaborated and checked. Theexperience gained from the pilot power units wasaccumulated and generalized for being taken intoaccount in serially produced power units. TheNVNPP Units 3–5 were the first ones of RussianNPPs that passed the service life extension procedure.Owing to the performed works on substantiating theresidual life of power unit buildings, systems, andequipment, further safe operation for another 15 yearshas been secured for VVER�440 reactors (Units 3 and 4)and 26 years for a VVER�1000 reactor (Unit 5).

Translated by V. Filatov