ISSN 0040�6015, Thermal Engineering, 2014, Vol. 61, No. 2, pp. 123–132. © Pleiades Publishing, Inc., 2014.Original Russian Text © V.P. Povarov, A.B. Tereshchenko, Yu.N. Kravchenko, I.V. Pozychanyuk, L.I. Gorobtsov, E.I. Golubev, V.I. Bykov, V.V. Likhanskii, I.A. Evdokimov,V.G. Zborovskii, A.A. Sorokin, V.D. Kanyukova, T.N. Aliev, 2014, published in Teploenergetika.

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A complex approach to the failed fuel elementdetection (FFED) at nuclear power plants (NPPs)with VVER reactors has been in the process of devel�opment in Russia since the 2000s. The FFED meth�ods applicable at the shut�down and running reactorwere modernized within the framework of the complexapproach. Modernization was carried out for the pur�pose of increasing the FFED sensitivity and reliability.The upgraded FFED methods were tested and imple�mented for the first time at the Novovoronezh NPP.

The search for failed elements of fuel assemblies(FAs) in the core of a running reactor is the first stageof fuel leakage testing. The information on the param�eters of leaky fuel elements may be obtained promptlyby processing the current data on the activity of radio�active fission products (RFPs) in the primary coolantusing calculation aids.

The modernized method of FFED at a runningreactor was implemented as an automated computer�ized expert system. The first expert system was devel�oped for the Novovoronezh NPP unit no. 5 based on

the certified [1] RTOP�CA code.1 The mechanistic

1 The RTOP�CA and RTOP�KGO codes used to develop mod�ernized fuel leakage testing methods were created with financialsupport from OAO TVEL.

RTOP�CA code is designed for predicting the RFPactivity in the primary coolant and modeling the dete�rioration of characteristics of faulty fuel elements. Thefuel state analysis in the expert system is carried outusing the data from the in�core monitoring system(ICMS) equipment.

Owing to the use of FFED at a running reactor itbecomes possible to reduce the uncertainty of estimat�ing the parameters of faulty fuel elements and the riskof missing a leaky FA at the subsequent fuel leakagechecking procedures. If no equipment for on�lineFFED is installed at the fuel�handling machine mast,the estimation of core defectiveness during the reactoroperation remains the only way to lower the durationand optimize the cost of the cask FFED performedafter the campaign completion.

The cask FFED is the final stage of fuel leakagetesting. A method with pressure cycling in the casks ofthe faulty assembly detection system (FADS) [2–4]was developed in order to meet the new and morestringent FFED reliability and informativenessrequirements.

Reliable methods of monitoring and predicting thebehavior of leak�tight fuel during the power plant unitoperation are needed in order to ensure the safety ofthe nuclear fuel operation in the current severe operat�

Developing and Applying Modern Methods of Leakage Monitoring and State Estimation of Fuel at the Novovoronezh

Nuclear Power PlantV. P. Povarova, A. B. Tereshchenkoa, Yu. N. Kravchenkoa, I. V. Pozychanyuka, L. I. Gorobtsova,

E. I. Golubeva, V. I. Bykova, V. V. Likhanskiib, I. A. Evdokimovb, V. G. Zborovskiib, A. A. Sorokinb,V. D. Kanyukovab, and T. N. Alievb

a Novovoronezh Nuclear Power Plant, Branch of Rosenergoatom Concern, Industrial’naya zona Yuzhnaya 1,Novovoronezh, Voronezh Oblast, 396071 Russia

b Troitsk Institute for Innovation and Fusion Research (TRINITI State Research Center of the Russian Federation), Troitsk, Moscow oblast, 142190 Russia

Abstract—The results of developing and implementing the modernized fuel leakage monitoring methods atthe shut�down and running reactor of the Novovoronezh nuclear power plant (NPP) are presented. An auto�mated computerized expert system integrated with an in�core monitoring system (ICMS) and installed at theNovovoronezh NPP unit no. 5 is described. If leaky fuel elements appear in the core, the system allows oneto perform on�line assessment of the parameters of leaky fuel assemblies (FAs). The computer expert systemunits designed for optimizing the operating regimes and enhancing the fuel usage efficiency at the Novovor�onezh NPP unit no. 5 are now being developed.

Keywords: activity of the radioactive fission products, inert radioactive gases, failed fuel element detection,failed fuel element detection with pressure cycling in casks of the faulty assemblies detection system, expertsystem, RTOP�CA code, thermomechanical stresses, fuel–cladding interaction

DOI: 10.1134/S0040601514020098

NUCLEAR POWERSTATIONS

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ing conditions. Computer systems designed to providesupport for fuel campaigns [5–10] and combining cal�culation codes, ICMSs of power plant units, and theFFED at a running reactor are developed to these endsabroad. Preliminary versions of modules for monitor�ing the leak�tight fuel state at the individual elementlevel are included into the developed expert system forthe VVER fuel.

A review of the developed modernized methodsand calculation aids of FFED at a running and shut�down reactor is given below. The examples of complexapplication of the modernized FFED methods at theNovovoronezh NPP units are presented. The currentversion of the expert system for supporting the VVERfuel use that is installed at the Novovoronezh NPP unitno. 5 is described. The possible lines of its develop�ment are discussed.

THE METHOD OF FFED AT A SHUT�DOWN REACTOR WITH PRESSURE CYCLING

IN THE FADS CASK

The standard cask water method of FFED at ashut�down reactor adopted in Russia for NPPs withthe VVER reactors is described in detail in [11, 12]. Ina standard FFED method, the radionuclides arereleased from under the cladding of a leaky fuel ele�ment into the FADS circulation circuit waters under asingle pressurization and subsequent depressurizationin the FADS cask. The statistic analysis of theobtained data on the RFP activity in the FADS circuitis carried out for a group of tested fuel elements, andthe FAs that have leaky fuel elements are determinedaccording to [11, 12].

The practice in performing FFED at the VVER�type reactors shows that the used standard method ofFFED at a shut�down reactor does not always make itpossible to reliably identify the faulty FAs. The prob�lem of identifying FAs with major defects is related tothe ambiguous nature of the relation between the fuelelement defect size and the integral radionuclide yield.

The modernized cask FFED method with pressurecycling was developed and is now implemented at

Russian NPPs with the VVER reactors [2, 3, 13]. Thismethod is tested using the available FADS equipment.Multiple cyclic pressurizations and depressurizationsin the FADS cask are performed in the process ofFFED. This promotes efficient mass transfer withinthe fuel element, and more RFPs are released throughthe defect in the leaky fuel element cladding. Conse�quently, the reliability of identifying leaky FAs isenhanced. Since the use of the method with pressurecycling does not require statistical processing of activ�ity measurements for a great number of assemblies, theduration of the FFED at a shut�down reactor is usu�ally reduced. The RFP release kinetics is altered underpressure cycling. This makes it possible to determinethe effective hydraulic diameter of the defect in theleaky fuel element cladding. The estimate of the defectsize may be taken into account when deciding whetherthe leaky fuel assemblies should be used further orremoved. The method with pressure cycling may alsobe combined with the standard FFED method to carryout a rapid FA hermeticity analysis. The monitoringregimes and the results of their application at theNovovoronezh NPP are detailed below. The FFEDmethod with pressure cycling is nowadays imple�mented at all Russian NPPs with VVER reactors andintroduced into the regulatory documents of the OAORosenergoatom Concern [11, 12].

Enhancing the reliability of identifying leaky FAsusing the method with pressure cycling. The pressurevariation scenario and the sampling points in an FAtest performed according to the method with pressurecycling [14] are shown in Fig. 1a.

The presence of a leaky fuel element (or a number ofsuch fuel elements) in the tested FA is revealed by thedetection of at least one of the following signs [11, 12]:

(i) a significantly more rapid (relative to the corro�sion nuclides) increase in the activity of referenceRFPs in consecutive samples;

(ii) the activity of at least one of the reference RFPsin the last obtained sample is significantly (at least twotimes) higher than the statistical 3σ criterion of reject�ing the leaky FAs according to the standard method;and

50

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0.1

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10 15 20 25 30 35 40Time, min Time, min

Pre

ssur

e, M

Pa

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Pre

ssur

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Pa

Fig. 1. Scenarios of pressure variation in the FADS cask (solid curve) and sampling instants (dots) in the process of (a) FFEDwith pressure cycling and (b) standard FFED combined with the method with pressure cycling.

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(iii) the detection of the specific activity of the133Xe gaseous fission product in the last sample.

The method with pressure cycling was first appliedat the Novovoronezh NPP units nos. 3 and 4 with theVVER�400 reactors in the years 2004–2006. Since thefuel leakage level was then increased and the primarycircuit was contaminated with reference RFPs, a greatnumber of “suspicious” FAs that could contain leakyfuel elements were detected at planned maintenance(PM) at the NPP units nos. 3 and 4. Such FAs weresubjected to the FFED procedure with pressurecycling at which all leaky FAs exhibited positive kinet�ics of the reference RFPs activity in consequent sam�ples and a significantly increased (relative to the sam�ples from the standard FFED performed earlier)activity of these RFPs in the last samples. When leak�tight FAs were tested, the activity of reference RFPs inthe FADS coolant remained low and exhibited no sig�nificant alterations in the process of cycling. Thus, thefact that the FFED method with pressure cycling hashigher sensitivity and selectivity than the standardFFED method was confirmed experimentally.

Estimating the effective hydraulic diameter of thedefect in a fuel element. The rate of RFP release from aleaky fuel element depends on the defect hydraulicresistance. The RTOP�KGO calculation code [4] wasdeveloped and certified for the purpose of solving thedirect problem of determining the RFP release from aleaky fuel element under the cask FFED depending onthe defect size, its position, fuel element depletion,and other parameters. The fuel element condition(geometric characteristics and the activity of accumu�lated RFPs) at the moment of PM is calculated usingthe RTOP�CA code [1]. The inverse problem (theproblem of reconstructing the effective hydraulicdiameter of the defect based on the measured releasekinetics of the standard RFPs) may be solved usingvarious methods.

The first method consists in performing calcula�tions using the RTOP�KGO code and choosing thedefect parameters that produce the best match withthe measured data. This method was applied in thecask FFED experiments with pressure cycling thatwere carried out at the Novovoronezh NPP units withVVER�400 reactors [3]. Eleven fuel assemblies weretested according to the method with pressure cyclingafter the reactor shutdown. The kinetics of release ofvarious radionuclides (131I, 134Cs, 136Cs, 137Cs, and54Mn, which is an indicator of surface contaminationwashing�off from the FA construction elements) intothe FADS coolant was measured for these assemblies.Two FAs were subjected to the FFED procedure withpressure cycling after being stored in a pond for half ayear. Only the kinetics of release of 134Cs, 137Cs, and54Mn into the FADS coolant was measured for thispair. The measurement results were normalized to theradionuclide activity value in the first obtained sample.

A series of calculations were carried out using theRTOP�KGO code for various sizes and height posi�tions of the defect. It was assumed that the defect waslocated in the upper or lower parts of the fuel element(this is the typical case for leaky VVER fuel elements).The measured activities of monitored radionuclides inthe FADS cask and the estimated defect sizes for twostudied FAs are shown in Fig. 2. The obtained resultsconfirm the possibility of determining the effectivehydraulic diameter of the defect using the FFEDmethod with pressure cycling. It can be seen that themethod remains efficient even when applied to FAsthat were stored in a pond for a long time.

The second method of estimating the effectivehydraulic diameter of the defect consists in plottingthe maps of kinetic dependences in advance using theRTOP�CA and RTOP�KGO codes [4, 13]. In order todo that, a set of normalized curves of the kinetics ofrelease of reference RFPs into the FADS cask is calcu�lated depending on the defect size. The calculationsare performed with account for the uncertainty ofuncontrolled variables (such as the defect position orthe initial fuel–cladding gap at the fuel element pro�duction). In order to estimate the defect size, theresults of measuring the RFP activity in the FADScoolant are matched with the curves at the corre�sponding maps, and the curve that best fits the experi�mental data is selected. The data on the activity ofeach monitored reference RFP in the FADS cask thatare to be processed using the maps of kinetic depen�dences are normalized with respect to the RFP activi�ties at the start and at the end of pressure cycling [13].

The calculated maps of kinetic dependences for thestandard and upgraded fuel of the VVER�1000 andVVER�440 reactors and for the AES�2006 project fuel[13–15] are now available. The examples of such mapsfor the TVS�2 and TVSA fuel elements that were usedfor 1 year are shown in Fig. 3. The method of calculat�ing the maps of kinetic dependences and estimatingthe defect size is described in more detail in [13].

The combined FFED method in the FADS cask. Thefuel element leakage in an FA may be evaluated on ayes/no basis by combining the standard FFEDmethod and the method with pressure cycling withoutadditional rinsing of the FAs. The corresponding pres�sure variation scenario in the FADS cask is shown inFig. 1b. The A sample is obtained after the completionof the standard FFED, and the B sample is obtainedafter the pressure cycling is over. It is assumed that themonitored FA contains leaky fuel elements if the fol�lowing two requirements are met:

(i) a significant (relative to the nuclides of corro�sion origin) increase in the specific activity of refer�ence RFPs that occurred in the interval between sam�ples A and B is detected and

(ii) the specific activity of at least one of the refer�ence RFPs in sample B is at least two times higher thanthe statistical (3σ) rejection criterion.

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The combined cask FFED method was firstapplied at the Novovoronezh NPP in 2006. The resultsof testing 14 FAs according to this method showed thatleaky fuel assemblies exhibited an order of magnitudehigher (relative to the standard FFED) RFP activity inthe FADS coolant after pressure cycling. Two leakyFAs did not reach the rejection criterion level whenthey were subjected to the standard FFED procedure.The RFP activity in leak�tight FA samples did notexhibit a marked excess over the activity kinetics ofcorrosion radionuclide 54Mn or did not reach the sta�tistical rejection criterion value even at sample B.

THE EXPERT SYSTEM FOR MAINTENANCE OF THE VVER FUEL

The upgraded method of FFED at the runningVVER�1000 reactor [16, 17] was developed based onthe certified RTOP�CA calculation code. The methodallows assessing the following parameters based on thedata on the activity of reference radionuclides in theprimary circuit coolant:

(i) the mass of fuel deposits in the core (based onthe activity of 134I);

(ii) the presence of leaky FAs in the core (based onthe ratio of activities of 131I and 134I and on the pres�ence of bursts of activity of long�living isotopes ofiodine, cesium, and xenon radionuclides);

Nor

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arb.

un

its

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Nor

mal

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spe

cifi

c ac

tivi

ty,

arb.

un

its

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Fig. 2. Activity release kinetics under the FFED with pres�sure cycling in the FADS cask at the Novovoronezh NPPfor (a) FA no. 1 (calculation is done for a 35�µm�sizeddefect) and (b) FA no. 2 that was stored in the cooling pondfor more than half a year (calculation is done for a 85�µm�sized defect). The specific activity of reference radionu�clides in the FFED samples taken at cycling (seven sam�ples are taken) is normalized to the specific activity of radi�onuclides in the sample taken 20 min after the cycling isstarted and is indicated at the ordinate axis. Radionuclides:1—131I, 2—134Cs, 3—137Cs, and 4—54Mn. Curve 5 cor�responds to the calculation results.

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5% uncertainty 15% uncertainty

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From 29 to 34 From 29 to 34

From 33 to 40 From 33 to 40

From 36 to 45 From 35 to 45

From 45 to 73 From 44 to 80

From 50 to 95 Fromт 48 to 100

From 56 to 105 From 51 to 120

Larger than 80 Larger than 67

Larger than 90 Larger than 80

Fig. 3. Normalized kinetic dependences for the activityrelease from a leaky fuel elements of TVS�2 and TVSA afterthe first year of operation. Curves 1–8 correspond to differenteffective hydraulic diameters of the fuel element defect.

04 8 12 16 20

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DEVELOPING AND APPLYING MODERN METHODS OF LEAKAGE MONITORING 127

(iii) the faulty fuel burnup (based on the ratio of

activities of 134Cs and 137Cs under the spike effect2);

(iv) the faulty fuel burnup (using a complementarytechnique based on the analysis of activity ratios ofgaseous fission products (85mKr, 88Kr, 135Xe) in a sta�tionary power regime of reactor operation);

(v) the number of leaky fuel elements (based on theactivity of iodine radionuclides); and

(vi) the size (small/large) of the defect (based onthe data on the activity of gaseous fission products andiodine radionuclides).

The use of the data on the activity of gaseous fissionproducts in the primary circuit coolant, which areobtained according to the method of inert radioactivegases (IRGs) monitoring in the coolant developed atthe Novovoronezh NPP in the upgraded method ofFFED at the running reactor, offers the followingadvantages as a complimentary method of estimatingthe faulty fuel burnup:

(1) The faulty fuel burnup may be estimated evenwhen the 134Cs and 137Cs spike effect is not detected(specifically, if a sufficient amount of cesium is not yetaccumulated in low�burnup fuel).

(2) If a single leaky FA or several leaky FAs close inburnup are present in the core, the joint use of twomethods of estimating the defective fuel burnup allowsone to narrow the estimate uncertainty range.

(3) If leaky FAs with long and short operational usetimes are present in the core, the joint use of two meth�ods of estimating the burnup makes it possible todetermine the defective fuel parameters more reliably.If FAs with different operational use times are presentin the core, an analysis of the activity of gaseous fissionproducts reveals the presence of leaky fuel elements inFAs with low fuel burnup, and an analysis of the dataon the activities of 134Cs and 137Cs under the spikeeffect allows identifying leaky FAs with high fuel bur�nup in the core.

The dependence of the radiation yield of fissionproducts on the fuel isotope composition [16, 18] isused to obtain an estimate of the defective FAs burnupbased on the activity of gaseous fission products. Theyield efficiency for krypton isotopes (85mKr, 88Kr)decreases with increasing fuel burnup much fasterthan the yield efficiency for xenon (135Xe). Therefore,the ratio of the measured specific activities of kryptonand xenon in the primary circuit coolant is a measureof the defective fuel burnup.

The defective fuel parameters are estimated duringthe reactor operation by comparing the current dataon the RFP activity in the coolant with a database cal�culated using the RTOP�CA code for the given fuel

2 The spike effect consists in a fivefold (or greater) increase in thespecific activity of 131I in the primary reactor circuit coolant associ�ated with the release of this radionuclide from leaky fuel elementsafter the emergency protection was triggered or the reactor powerwas altered by no less than 20% of its current value.

type and fuel cycle. The comparison of the 85mKr/135Xeand 88Kr/135Xe ratios of the measured specific activi�ties in the coolant allows assessing the defective fuelburnup. The number of leaky fuel elements in the coreis estimated in a similar way using the results of mea�suring the specific activities of iodine radionuclides(131I, 133I, and 135I) in the primary circuit coolant [17].

The first version of the expert system for FFED at arunning reactor was created in 2009 based on thedeveloped methods and was put into pilot operation atthe Novovoronezh NPP unit no. 5. The system wasintegrated with the unit ICMS. An updated version ofthe expert system was installed at the NovovoronezhNPP in early 2013 in connection with the moderniza�tion of the unit no. 5 ICMS.

The application of the expert system and the RTOP�CA code at the Novovoronezh NPP. The expert systemwas tested by comparing the predictive estimates of thedefective fuel parameters with the data obtainedthrough FFED at a shut�down reactor (see the table).

The table shows that the expert system determinedthe leaky FA parameters adequately in all the consid�ered cases. The spread of the measurement data on theRFP activity in the primary circuit broadened the esti�mate uncertainty interval in several cases, but theexpert system predictions did not contradict theresults of FFED at a shut�down reactor.

Direct calculations of the coolant activity were car�ried out for several campaigns with no faulty FAs in thecore (such campaigns are designated with in thetable) using the RTOP�CA code. These results wereused to expand the RTOP�CA code verificationmatrix. The calculations started with the determina�tion of the average mass of fuel deposits based on the134I radionuclide activity level. The values of theobtained mass of fuel deposits and the coolant clean�up rate were used to calculate the average activity lev�els of all radionuclides. An effective rate of the primarycircuit coolant clean�up (degassing) was chosen forgaseous fission products.

The measured and calculated activities of iodineradionuclides and gaseous fission products are shownin Fig. 4. A method with cryogenic condensation ofIRGs that are released from the sampled coolant in theprocess of degassing was used in the NovovoronezhNPP FFED laboratory to measure the activity of gas�eous fission products in the primary circuit coolant.This method makes it possible to obviate the loss of thereleased IRGs when preparing a sample. As a result,the measured specific activity is close to the absolutevalues of the specific activity of gaseous fission prod�ucts in the primary circuit coolant.

Figure 4 shows that the experimental and calcu�lated data on the activity of radioactive gases agreewith each other well. This close agreement confirmsthe adequacy of the cryogenic method and the calcu�lation methods implemented in the RTOP�CA code.

∨

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The prediction of RFP activity in the coolant in theevent that the use of leaky FAs in the core is continued.The modernized FFED methods allow estimating thesize of a defect in the leaky fuel element. The RFPactivity evolution in the primary circuit coolant maybe predicted after the defect size is estimated using theRTOP�CA code. The current version of the expert sys�tem is fitted with a software module used for predictingthe RFP activity in the coolant in the coming cam�paign. This module is utilized when there is a possibil�ity of further use of a known leaky FA, which did not

reach the rejection criterion in the cask FFED, in thefuel inventory of the coming campaign.

The RFP activity in the coolant is calculated withaccount of the planned position of the leaky FA in thecore. The data on the energy release distribution in anFA determined for the coming campaign using thePERMAK software for neutron�physical calculationsor the ICMS software are taken into account.

The problem of predicting the RFP activity in theprimary circuit coolant in the coming campaign is notthe only one that arises when a leaky FA is detected.

Results of application of the updated FFED method at the Novovoronezh NPP Unit 5

Campaign number Interval

Results of application of the method of FFED at a running reactor

Results of FFED at a shut�down

reactor

Direct RTOP�CA calculationsCs spike* GFPs** overall**

18 12/1999–09/2000 – Year 3 or 4 One year 3 or 4 FA One year 4 FA –

19 12/2000–10/2001 – Year 4 One year 4 FA The same –

23 08/2005–07/2006 36–38 MW day/kg U Year 4 The same '' –

24 09/2006–07/2007 41–46 MW day/kg U Year 3 One year 3 FA One year 3 FA, 42 MW day/kg U

–

25 08/2007–07/2008

No leaky FAs

Cask FFED was not conducted

26 08/2008–07/2009

27 10/2009–09/2010

28 10/2011–07/2012

* Results of assessing the defective fuel burnup using the Cs spike data.** Results of assessing the burnup using the data on the gaseous fission products activity.

*** The final prediction on the defective fuel parameters in the core.

∨

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ivit

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10–6

Fig. 4. Results of measuring the activity of (a) iodine radionuclides and (b) radioactive fission gases in the primary coolant circuitin the starting period of campaign no. 27 at the Novovoronezh NPP unit no. 5. Dashed lines indicate the background activitylevels calculated using the RTOP�CA code. Radionuclides: 1—131I, 2—132I, 3—133I, 4—134I, 5—135I, 6—133Xe, 7—135Xe,8—85Kr, 9—87Kr, and 10—88Kr.

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DEVELOPING AND APPLYING MODERN METHODS OF LEAKAGE MONITORING 129

Another important problem consists in predicting thevariation of the RFP activity level in the coolant in thecourse of the current campaign. Its importance stemsfrom the fact that large secondary defects may formafter the cladding integrity of a fuel element is com�promised. If water gets inside a fuel element, hydrogenstarts to accumulate in it due to the oxidation of fuelpellets and the inner cladding surface in a vapor�con�taining atmosphere. When the hydrogen concentra�tion reaches critical levels, the embrittlement andhydrogenation of the fuel element cladding occur. Theembrittled areas may crack and collapse, thus poten�tially leading to a loss of sizable cladding fragments.The formation of large secondary defects increases therisk of a sharp increase in the RFP activity in the cool�ant and contamination of the primary circuit due tothe release of fission products and the fine�dispersedfuel composition from the fuel element. Therefore,reliable criteria that would indicate the conditions andthe rate at which secondary defects may form areneeded.

It is known from foreign and domestic experiencethat the secondary defect development time is signifi�cantly dependent on the operating conditions [19]. Atlow power, a leaky FA may be operated for 2–3 yearsand maintain the RFP release from the faulty fuel ele�ment at a moderate level. At high power, the time tothe development of a secondary defect in the FA and asharp increase in the RFP activity in the coolant mayamount to several days.

Studies aimed at determining the parametricdependences that describe the formation of secondarydefect in the VVER FAs were conducted at the TroitskInstitute for Innovation and Fusion Research (in col�laboration with the scientists from the Advanced Sci�entific Research Institute for Inorganic Materials).The analysis showed that the development of a sec�ondary defect in a cladding was a process with severalstages. These include the accumulation of a criticalhydrogen concentration in the fuel element, the dis�ruption of the protective properties of the oxide film atthe inner cladding surface, and the penetration of amassive hydride to a significant depth into the clad�ding. At moderate power, most of the time to thedevelopment of a secondary defect is spent at the stageof accumulation of hydrogen beneath the cladding.The dependences of the rates of hydrogen accumula�tion in leaky fuel elements and the volumes of accu�mulated hydrogen on the generated power, fuel bur�nup, and the size and height position of the defect weredetermined using the RTOP�CA code.

Criterion dependences with respect to the thresh�old conditions of secondary defect formation [20, 21]were determined based on the physical modeling ofthe processes in leaky fuel elements. The obtaineddependences relate the composition of a steam–hydrogen mixture in the leaky fuel element to thecladding temperature, the thickness of the oxide film

on its inner surface, and the irradiation intensity. Thestudies that include comparing the data on the FAoperation before and after the occurrence of a leak, thedata on postreactor examination of leaky FAs, and theresults of calculating the coolant activity and the evo�lution of the composition of a steam–hydrogen mix�ture in leaky fuel elements using the RTOP�CA codeare conducted in order to determine the numericalcoefficients of the parametric dependences.

Preliminary estimates of the length of time inter�vals needed to form penetration secondary defects inleaky fuel elements with various energy release levelsare already obtained. This allows one to plot a curve inpower–time coordinates that defines the probabilityand rate of secondary defect formation in a leaky fuelelement (an analog of a curve on the Locke diagram[19] for the BWR fuel). The obtained criterion depen�dences in combination with the RTOP�CA code makeit possible to determine the relations between the timeto the secondary defect formation and the FA opera�tion conditions. Specifically, short�term VVER unitpower dips may be regarded as a means of reducing therisk of the coolant activity increasing sharply. Thedepth, duration, and (if needed) frequency of thepower dips may be chosen with the use of the RTOP�CAcode and the secondary defect formation criteria thatare being developed.

The secondary defect formation criteria beingdeveloped are planned to be introduced into the expertsystem. This will make it possible to:

(i) give well�grounded predictions of the RFPactivity evolution in the coolant taking into accountthe possibility of secondary defect formation in a leakyfuel element and

(ii) set the guidelines for the NPP personnel onhow to minimize the risk of the radiation situation atthe nuclear unit turning for the worse after the leakageof fuel elements in the core is detected.

The development of inspection techniques for check�ing the thermomechanical operating state of the VVERfuel. The fuel state monitoring methods operating at alevel of a single fuel element in the core are now beinggradually introduced both in Russia and abroad inorder to enhance the power generation capabilities ofreactor facilities and increase the efficiency of fuel use.The assurance of radiation safety at power maneuversis one of the key aspects of fuel monitoring. It isimportant to conduct continuous monitoring of thefuel cladding strength margins. This computationalproblem is most adequately solved using modern inte�gral fuel codes that demand serious computing power.An alternative method of adequate on�line analysis offuel thermomechanics consists in the use of simplifiedmodels. If such models give a correct description ofthe main mechanisms of mechanic interaction of thefuel with the cladding, they may be adjusted withrespect to quantitative indicators in the most danger�ous parameter range through cross�verification with

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detailed fuel codes. The simplified models are compu�tational aides that may provide sufficient accuracy inan on�line regime with a reasonably low computa�tional cost. An example of successful application ofsimplified calculation methods in determining thestress dynamics in the fuel element cladding is pro�vided by the XEDOR model developed for the BWRand PWR fuel [22]. A method based on this model iscertified by the US regulatory bodies and is applied atcommercial units of US NPPs.

A similar technique was implemented in the cur�rent version of the expert system for the VVER fuel inthe form of a separate program block. The interactionof the expert system with the Novovoronezh NPP unitno. 5 ICMS is established in order to collect the inputdata needed for the calculation of fuel element ther�momechanics. The model used to estimate themechanical stresses in fuel element claddings is based

on several equations. The first equation describes thetime variation of the fuel–cladding gap and takes intoaccount the cladding irradiation creep and sintering,swelling, and relocation of the fuel due to pellet crack�ing. The second equation describes the stresses in thecladding caused by the contact interaction with thefuel. The third equation takes into account the accu�mulation of plastic deformations in the cladding in theprocess of interaction with the fuel kernel. The modelis described in more detail in [23].

Since the Novovoronezh NPP unit no. 5 is nowa�days not operated in a flexible mode, a model scenarioof the reactor power variation was prepared in order todemonstrate the capabilities of the method. The sce�nario included several nuclear unit power rises anddips occurring in the course of about 5 days. The char�acteristic power change time equaled about 3 h. Thevariation of local power in one of the core fuel ele�

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Fig. 5. Time variations of (a) the linear power (curve 1) and fuel burnup (curve 2), (b) the width of the gap between the fuel andthe cladding (curve 3), and (c) the circumferential tensile stresses in the cladding (curve 4) in one of the fuel elements in the core.

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DEVELOPING AND APPLYING MODERN METHODS OF LEAKAGE MONITORING 131

ments under the set maneuvering scenario is shown inFig. 5a. The fuel burnup equals about 38 MW day/kg Uand varies little over the studied time interval (about5 days). Figures 5b and 5c present the results of calcu�lating the gap width variation and the circumferentialstresses in the cladding in one of the height regions ofthe fuel element.

The calculations for the model maneuvering sce�nario were carried out under the assumption that thestresses in the fuel element are already fully relaxedand the pellets exert no pressure on the cladding by theinitial time instant indicated in Fig. 5. At the initialtime, the compressive stresses caused by the differencein the coolant pressure and the gas pressure inside thefuel element act in the cladding. The first power dip(see Fig. 5) creates a small gas gap due to the reductionof fuel temperature and the outer diameter of pellets.When the power rises above the initial level, the gapdisappears, and tensile stresses start to act in the clad�ding. Plastic deformations start to accumulate in thecladding at this stage. Therefore, the gap forming atthe next power dip is wider than the first one (comparethe first and second maxima in Fig. 5b). When thepower rises again to a yet higher level, the gap “col�lapses” once again, and tensile stresses emerge in thecladding. Since the stress level is heightened, the stressrelaxation rate (the rate of stress reduction from a peakvalue) is also increased. The gap width and stress vari�ations may be traced in a similar fashion up to the endof the studied time interval.

The implemented method is undemanding of com�puting power. The analysis of individual fuel elementsin the entire core is done in a fraction of a second.Therefore, the developed method may be used in anon�line regime.

The possibility of calculating all key parametersthat determine the safety of fuel element operation(including the gaseous fission product yield and otherparameters) is planned to be implemented in theexpert system modules that are now being developed.This will make it possible to determine the allowed fueloperation conditions without the use of overly conser�vative limit curves on the basis of the actual perfor�mance history and the state of fuel elements. As aresult, the NPP personnel will be allowed to design thefuel inventory plans with more flexibility, and novelopportunities of increasing the energy production willarise.

CONCLUSIONS

(1) The modernized methods of FFED at a run�ning and shut�down reactor help enhance the infor�mativeness and reliability of the inspection of the FAfuel element cladding. Their use promotes radiationsafety at NPPs. The experiments carried out in recentyears at the Novovoronezh NPP made it possible totest and implement the updated FFED methods

within a short space of time. The results of these stud�ies provided the basis for implementing the said meth�ods at all Russian VVER nuclear units. The techniquefor measuring the activity of gaseous fission productsin the primary circuit coolant, which was developed atthe Novovoronezh NPP, expands opportunities forassessing the leaky fuel parameters during reactoroperation.

(2) The expert system for fuel maintenance makesit possible to process the data on RFP activity in theprimary circuit coolant automatically and receiveupdates on the leaky FAs state in the core. It is alsopossible to predict the RFP activity in the coolant inthe event that the operational use of leaky FAs, whichdid not reach the rejection criterion in the cask FFED,is continued. The integration of the expert system withthe ICMS allows one to narrow the uncertainty rangeof the assessed parameters and expands the range ofapplication of the calculation aides. The expert systemblocks are nowadays being updated in order to opti�mize the operating modes and enhance the fuel useefficiency of the VVER units.

REFERENCES

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2. V. V. Likhanskii, I. A. Evdokimov, N. M. Efremov, et al.,“Experimental ex�core studies on modeling the masstransfer processes for the development of methods offailed fuel element detection in the VVER reactors,”Vopr. At. Nauki Tekh., Ser.: Obespechenie Bezop. AES,No. 15, 145 (2006).

3. V. Likhanskii, E. Afanasieva, N. Efremov, et al.,“Development of defect size determination procedurein a cask of WWER defective assembly detection sys�tem,” in Proceedings of the 6th International Conferenceon WWER Fuel Performance: Modelling and Experimen�tal Support, Albena, Bulgaria, September 19–23, 2005,pp. 117–124.

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5. W. A. Boyd and R. W. Miller, “The BEACON on�linecore monitoring system: functional upgrades and appli�cations,” in Proceedings of a Specialists’ Meeting “In�Core Instrumentation and Core Assessment”, Mito�shi,Japan, October 14–17, 1996 (Nuclear Energy Agency,1997).

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Translated by D. Safin