N/ IGC-199

1998sIN0000259

Act iv i ty Report ofReactor Physics Division - 1 9 9 7

JLJL

Bhahba .prc P- c r ^ Centre

Trorobay,Mut3uj.4OOO85.

Edited ByOm Pal Singh

GOVERNMENT. OF INDIA. DEmRTMENT OF ATOMC ENERGY

INDIRA GANDHI CENTRE FOR ATOMIC RESEARCH KALFAKKAM

IGC - 1991998

GOVERNMENT OF INDIADEPARTMENT OF ATOMIC ENERGY

REACTOR PHYSICS DIVISIONACTIVITY REPORT - 1997

Edited byOm Pal Singh

jj " p n '•' \ " *"""

Librsrv ^ ' -f . - • ••:~]->n

Indira Gandhi Centre for Atomic ResearchKalpakkam 603 102, Tamil Nadu, India

FOREWORD

Reactor Physics Division carries out the physics related R&D work required for thedevelopment of Fast Breeder Reactors in India. The division is responsible for the nuclear designof the proposed 500 MWe Prototype Fast Breeder Reactor (PFBR), including the coreneutronics, radiation shielding and safety analysis. The division is also deeply involved inproviding the physics support, including necessary measurements, for the mixed carbide fuelledFast Breeder Test Reactor (FBTR) and the U-233 fuelled, light water moderated KAMIN1neutron source reactor.

During 1997, most of the important design parameters of PFBR were frozen and thestagewise submission of the Preliminary Safety Analysis Report to the PFBR Design SafetyCommittee started. In FBTR the outstanding achievement was the synchronisation of the turbogenerator to the Southern Electric Grid and token electricity production for the first time by a fastreactor in the country. The core size was expanded to 27 fuel subassemblies by adding one moreMark II fuel subassembly while a peak burnup of 39000 MWd/t was achieved in the core. Thehigh point of KAMINI operation was the achievement of full power of 30 kWt.

This report contains the summaries of the reactor physics activities conducted during the yeartowards the above achievements. The activities have been organised under the following topics :

• Nuclear Data Processing and Validation• PFBR and KAMINI Core Physics• FBTR Core Physics• Radioactivity and Shielding• Safety Analysis

Activities reported under Nuclear Data Processing and Validation include the study of theeffect of dilution grid size and energy dependence of dilution on self-shielding in the resolvedresonance region and work on an approximation to the Reich-Moore (RM) cross sectionrepresentation that helps in reducing the computing time.

PFBR core physics studies pertain to optimum design of absorber rods, optimisation ofrefuelling interval and criticality safety study of fuel sub-assemblies in storage yards.Contributions on physics aspects were prepared for the Preliminary Safety Analysis Report.KAMINI related activities pertain to the measurements of neutron flux at different locations andpower calibration. These experiments provide necessary input towards the design of shields forbeam tubes and to plan subsequent irradiation studies.

Activities for FBTR core physics have been the calibration of control rods for 26 and 27fuel subassembly cores. The observed power coefficient during earlier power operations wasreanalysed by improved corrections for temperature change. Reactivity loss due to burnup at25000 MWd/t and high power operation during October 1997 have been estimated. Powerdistribution in the 26 subassembly core and its dependence on the neutron source subassembly hasbeen calculated. A thermal neutron flux standard has been set up for accurate measurement ofneutron flux and for calibration of neutron detectors.

Radioactivity and shielding studies on FBTR have been on the comparison of 2-Dtransport calculations with measured flux and cover gas activity, calculation of fission productspectrum in FBTR Mark II fuel subassembly, calculation of concentration and activities ofcertain isotopes in FBTR subassembly after 50 GWd/t burnup. On PFBR, neutron flux andgamma dose rates at under-vessel detector location have been calculated, bulk andcomplementary shield design have been completed, dose to personnel in RCB followingdischarge pipe rupture and release of radioactivity into environment due to core disruptiveaccident have been calculated. The choice of hardfacing material for NSSS components hasbeen finalised based on induced activities, dose rates and man-rem expenditure and neededshielding. For conducting PFBR shielding mock up experiments at APSARA, the first phaseof measuring the neutron flux at the APSARA shielding corner is over. Gamma spectrum andfission product activities in the converter assembly for these experiments are calculated.

In safety analysis, using reliability analysis methodology, flow blockage probability inthe FBTR fuel subassembly was estimated and distribution of SCRAM signals to the twoshutdown systems of PFBR was finalised. A comprehensive report on the conceptual designof shutdown system of PFBR has been prepared. Analysis of severe accident in BN-800 typefast reactor with near zero sodium void coefficient has also been done. Computer codedevelopment work on sodium boiling, fuel freezing and efficient solution of exact heatconduction equations was carried out.

A list of publications authored or co-authored by members of the Division as well asseminars delivered are included at the end of the report.

I wish to express my appreciation to Dr. Om Pal Singh for his meticulous editing andto Smt. S. Rangamani and Smt. S. Janaki for their excellent typing of this report.

S.M. LeeHead, Reactor Physics Division

CONTENTS

Page No

1. Nuclear Data Processing and Validation 1-1

1.1 Effect of Fineness of Dilution Grid and Energy Dependence of Dilutionon Self-Shielding in the Resolved Resonance Region 1 - 1

1.2 Rational Approximations to Reich-Moore Cross Sections 1-21.3 Accurate Evaluation of Resonance Integrals 1-4

2. PFBR and KAMINI Core Physics 2 - 1

2.1 Absorber Rods in PFBR 2 - 12.2 Refuelling Interval for PFBR 2 - 32.3 Criticality Safety Study of PFBR Fuel Subassemblies with Borated

Steel Casings in Spent Fuel Storage Yards Clustered Configuration 2 - 52.4 Initial Measurements of Neutron Flux at Different Locations in

KAMINI Reactor 2 - 62.5 PWR Lattice Benchmark Problem Solutions 2 - 72.6 Design of a Magnetic Sensors Based on Covariance Analysis 2 - 9

3. FBTR Core Physics 3 - 1

3.1 Calibration of Control Rods for 26 Fuel Subassembly Core 3 - 13.2 Subassembly wise Power Distribution for 26 SA Mark I Core

and its Dependence on Source SA 3-13.3 Observed Power Coefficient of Reactivity for the 24 SA Mark I core 3 - 23.4 Reactivity Loss Rate due to Burnup in FBTR Small Carbide Core at

25000 MWd/t Burnup at High Power Operation during October 97 3 - 33.5 Calibration of Control Rods for 27 Subassembly Core 3 - 53.6 Measurement of Isothermal Temperature Coefficient with

27 Subassemblies 3 - 73.7 Thermal Neutron Flux Standard Facility 3 - 7

4. Radioactivity and Shielding 4 - 1

4.1 Comparison of 2-D Transport Calculations with MeasuredDetector Location Flux and Cover Gas Activity in FBTR 4 - 1

4.2 Fission Product Gamma Spectrum in FBTR Mark II CoreSubassembly 4 - 1

4.3 Concentrations and Activities of I, Ce, Ru and Pd Isotopes in FBTRSubassembly after 50 GWd/t Peak Burnup 4 - 3

4.4 Neutron Flux and Gamma Dose Rate at Under-Vessel Detectorin PFBR 4 - 3

iii

4.5 Bulk Shielding and Complementary Shield Design in PFBR ReactorAssembly 4 - 5

4.6 Choice of Hardfacing Material for PFBR-NSSS Components based onInduced Activities, Dose Rates, Man-Rem Expenditure and Shielding 4 - 6

4.7 Dose to Personnel in RCB Following Discharge Pipe Rupture ofPFBR 4 - 8

4.8 Release of Radioactivity into Environment under Hypothetical CoreDisruptive Accident Condition from PFBR 4 - 9

4.9 A Study of Fission Product Activity in Typical Fast Reactors 4 -114.10 Fission Product Gamma Spectrum and Heavier Actinides

Concentration in a PFBR Spent Fuel Subassembly 4 -134.11. Gamma Spectrum and Fission Product Activities in a Converter

Assembly for PFBR Bulk Shield Experiments at APSARA 4-144.12. Measurement of Neutron Flux at the Shielding Corner of APSARA 4-16

5. Safety Analysis 5 - 1

5.1 PSA on Flow Blockage of FBTR Fuel Subassembly 5 - 15.2 Conceptual Design of Reactor Shutdown System for PFBR 5 -25.3 Distribution of SCRAM Signals to Absorber Rods in PFBR

Based on Reliability Analysis 5 -45.4 Calculation of Temperature, Porosity and Plutonium Distribution

in PFBR Fuel Central Pin 5 - 75.5 Analysis of Severe Accidents in BN-800 Like Reactor 5 - 75.6 Numerical Solution of Transient Heat Conduction Equation with

Temperature Dependent Thermal Conductivity 5 - 95.7 Improved Sodium Boiling Model for ULOF Studies 5-105.8 Fuel Freezing in Coolant Channels 5-11

Appendix A : Publications A - 1

A. 1 Journals A - 1A.2 Conferences/Symposia A - 1A.3 Thesis A - 3A.4 IGC Reports A - 3A. 5 News Letter A - 3A. 6 Internal Reports A - 3

Appendix B : Seminars A - 6

B.I RPD Seminars A - 6B.2 Other Lectures A - 6

IV

CHAPTER 1

NUCLEAR DATA PROCESSING AND VALIDATION

1.1 Effect of Dilution Grid Size and Energy Dependence of Dilution on Self-Shieldingin the Resolved Resonance Region(V. Gopalakrishnan and S. Ganesan* (*Th. P. D., BARC))

Essential discussions and results of a study comparing the effective Self-ShieldingFactors (SSF) obtained by the conventional method that ignores the energy fluctuations of thedilution cross-section, with those obtained without ignoring these fluctuations, were given inthe previous Activity Report [1]. Effect of fineness of the dilution grid in the conventionalmethod was also presented. Multigroup data prepared by various methods from JENDL-2basic data was used to study the effects of the nuclides in the core of the fast criticalbenchmark assembly ZPR-6-7 on the effective SSF. For taking into account the energydependence of the dilution cross-section, the macroscopic total cross-section of the coremixture was prepared over a grid of about 42,000 energy points This was later foundinadequate to completely represent all the resonances of all the nuclides in the core. Hence aunion energy grid was prepared from the individual grids giving total cross sections of eachnuclide and the new grid rose to 1,32,000 energy points. Representative results of calculations[2] done with this refined grid are given in Table 1.1.1 and 1.1.2 for the same case-study.

Table 1.1.1: Comparison of SSF for Coarse and Fine Dilution Grids with EnergyDependent Dilutions for ZPR-6-7 Core at 300 K for TJ-238 Capture

Grp.No.

(g)

Dilution1617181920212223

Dilution1617181920212223

Energylimit

Eg(eV)

grid: 0, 10, 13.36E+32.04E+31.24E+37.50E+24.55E+22.76E+21.01E+22.26E+1

grid: 40, 50.3.36E+32.04E+31.24E+37.50E+24.55E+22.76E+21.01E+22.26E+1

Unshieldedcross-section(barn)

00, 1000, 10000,1.35E+O1.69E+02.76E+03.48E+03.42E+01.39E+13.05E+11.38E+2

, 60, 70, 150, 2001.35E+O1.69E+02.76E+03.48E+03.42E+01.39E+13.O5E+11.38E+2

Effectivedilution(barn)

1000001.68E+24.87E+14.67E+14.76E+16.27E+15.35E+15.94E+16.51E+1

1.76E+24.90E+14.70E+14.77E+16.46E+15.40E+16.03E+16.47E+1

SSFconventional

0.77880.46890.39180.33050.32590.11660.06780.0486

0.82390.51510.43080.36430.35550.12510.07160.0512

SSFE-dependent

dilutions

0.81810.51630.42490.37070.37030.12380.07010.0516

0.81810.51630.42490.37070.37030.12380.07010.0516

Deviation(%)

5.0510.118.45

12.1613.626.153.466.17

-0.700.24

-1.361.744.17

-1.08-2.020.87

1-1

Table 1.1.2: Comparison of SSFs for Coarse and Fine Dilution Grids with EnergyDependent Dilutions for ZPR-6-7 Core at 300 K for Pu-239 Fission

GroupNo.

(g)

Dilution2021222324

Dilution2021222324

Energylimit

Eg(eV)

grid: 0, 10,4.55E+22.76E+21.01E+22.26E+13.06E+0

Unshieldedcross-section(barn)

100, 1000, 10000,9.85E+01.84E+14.41E+17.76E+12.34E+1

grid: 200, 300, 400, 500, 600,4.55E+22.76E+21.01E+22.26E+13.06E+0

9.85E+01.84E+14.41E+17.76E+12.34E+1

Effectivedilution(barn)

1000004.68E+24.09E+24.05E+24.32E+24.14E+2

10004.80E+24.14E+24.11E+24.36E+24.19E+2

SSFconventional

0.86770.76940.62800.47210.8691

0.92480.84520.70820.54870.9066

SSFE dependent

dilutions

0.93170.84370.71100.58080.8527

0.93170.84370.71100.58080.8527

Deviation(%)

7.389.65

13.2223.03-1.89

0.75-0.180.395.86

-5.94

References

1. S. John Collins (Ed.), "Activity Report of Reactor Physics Division-1996", ReportIGC-190 (1997) pp 1-3, Indira Gandhi Centre for Atomic Research, Kalpakkam.

2. V. Gopalakrishnan and S. Ganesan, "Self-shielding and Energy Dependence ofDilution in the Resolved Resonance Region", Annals of Nuclear Energy (in print).

1.2 Rational Approximations to Reich-Moore Cross Sections(K. Devan and R. S. Keshavamurthy)

The commonly used formalisms in the ENDF/B format to represent neutronresonances are Single-Level Breit-Wigner (SLBW), Multilevel Breit-Wigner (MLBW), Adler-Adler (AA) and Reich-Moore (RM). RM formalism is the most accurate and is presently beingrecommended for the resonance analysis. It is mainly because it is unitary and can representcross section in the windows as well as in peaks, even for the most complicated multilevelinterference patterns.

The latest American evaluated file, ENDF/B-VI, uses RM formalism for all theimportant isotopes. In ENDF/B-VI, the entire resolved resonance energy region of an isotopeis divided into several energy ranges. To calculate the cross section at an energy point withinan energy range, the contribution to the collision matrix from all the resonances in that rangehave to be calculated. The cross section in RM formalism is expressed in terms of the inverseof a 'channel matrix', which is expressed in terms of P and Q functions given by

A,Z0.5 ( TnX

7,

1/2

Tcx) ( E* - E) / [(Ex - E)2 + 0.25 (TyX f)

1-2

one has to create an energy grid for cross section. A rational approximation [2] has beendeveloped for P and Q for non-fissile materials which are valid within a resonance as well asmany widths beyond. The approximations for P and Q for the m* resonance are:

61/2P = (En, -y * rnm)1/2 ([0.25 r n m r

ymE)2 +0.25

= (Em-y*rnm)1 /2([0.5yr2nm/{(Em-E)2+0.25(rYm)2}VEm]+

]+ £

82 Gk

k=0

Ck y*k=0

])

where y = (En, - E) / Fmn , is the energy away from the resonance measured in terms ofneutron width. The coefficients { Ck } and {Gk }are obtained using certain algebraicexpressions involving the resonance parameters. The coefficients computed once can be usedfor calculation of cross sections at any point around the resonance. This reduces computingtime if cross sections are needed over an energy grid.

Ni-58 from ENDF/B-VI has been chosen to illustrate the method. The comparisonbetween the exact and the approximate methods given in Tables 1.2.1. and 1.2.2 for a broadand a sharp resonance show that the method is as good as exact in reproducing RM crosssections within a resonance, and works with very good accuracy (0.1%) much beyond theresonance. Work is currently under way to use these approximations for the difficult problemof Doppler broadening of RM cross sections.

References

1. P. F. Rose and C. L. Dunford, "ENDF-6 Formats Manual", IAEA-NDS-76, 1990.2. K. Devan and R. S. Keshavamurthy, " Rational Approximations To Reich-Moore

Cross Sections", Proc. Twelfth National Symposium on Radiation Physics (NSRP-12), Jodhpur, 1998.

Table 1.2.1: Comparison of Cross Sections Calculated by RationalMethod and Exact Method around 15.306 keV (Em) Resonance Level

Approximation» 1.3324 keV)

yWithin

0.000.50

-0.501.00

-1.00Outside

2.00-2.004.00

-4.006.00

-6.00

Em-y*rn m

the Resonance:1.531E+41.464E+41.597E+41.397E+41.664E+4

the Resonance:1.264E+41.797E+49.976E+32.064E+47.312E+32.330E+4

Exact

1.708E+25.027E+11.223E+27.017E+07.297E+1

2.920E-13.967E+15.794E+02.299E+11.039E+11.740E+1

ElasticRationalApprox.

1.708E+25.027E+11.223E+27.017E+07.297E+1

2.920E-13.967E+15.794E+02.299E+11.040E+11.739E+1

Rel.Error(%)

0.0E+00.0E+00.0E+0O.OE+00.0E+O

3.1E-50.0E+01.0E-3

-4.3E-42.9E-2

-2.1E-2

Exact

1.280E-17.529E-25.598E-23.096E-22.311E-2

1.057E-27.591E-35.014E-32.629E-34.893E-31.586E-3

CaptureRationalApprox

1.280E-17.529E-25.598E-23.096E-22.311E-2

1.057E-27.591E-35.014E-32.629E-34.890E-31.587E-3

Rel.Error(%)

O.0E+00.0E+00.0E+00.0E+00.0E+0

0.0E+00.0E+0-2.8E-32.2E-3

-6.7E-25.6E-2

1-3

Table 1.2.2:Comparison of CrossMethod and Exact Method around

Sections Calculated by Rational Approximation285.38 keV (Em) Resonance Level ( Tnm= 4.000 eV)

yWithin

0.00.50.81.0

-1.0Outside

50.0-50.0100.0

-100.0200.0

-200.0

En,- yTnmthe Resonance:

2.85380E+52.85378E+52.85377E+52..85376E+52.85384E+5

the Resonance:2.85180E+52.85580E+52.84980E+52.85780E+52.84580E+52.86180E+5

Exact

3.736E+01.028E+02.893E+04.415E+08.933E+0

8.547E+08.470E+08.655E+08.373E+08.845E+08.216E+0

ElasticRationalApprox.

3.736E+01.028E+02.893E+04.415E+08.933E+0

8.547E+08.470E+08.655E+08.373E+08.844E+08.217E+0

Rel.Error(%)

0.0E+00.0E+00.0E+00.0E+00.0E+0

0.0E+00.0E+0-1.1E-52.3E-5

-7.3E-31.2E-2

Exact

1.457E+01.298E+08.016E-14.967E-12.505E-1

1.045E-39.465E-49.638E-47.841E-41.056E-36.899E-4

CaptureRationalApprox

1.457E+01.298E+08.016E-14.967E-12.505E-1

1.045E-39.465E-49.638E-47.841E-41.056E-36.899E-4

Rel.Error(%)

0.0E+0O.OE+00.0E+00.0E+00.0E+0

0.0E+00.0E+0-1.8E-53.0E-51.0E-2

-7.8E-3

1.3 Accurate Evaluation of Resonance Integrals(R.S.Keshavamurthy and R.Harish)

Evaluation of resonance integrals is admittedly a difficult problem. One major reasonis the complex nature of the integrand of the resonance integral J:

dx

where \\f (x,0) is the usual Doppler broadening function. Accurate calculation of \\f itself is noteasy and there is no closed form expression available for its evaluation. An inequality for vj/ hasbeen developed which is very useful in the numerical evaluation of J. The J integral is divided

as,

XT/p

J= f dx+fXU

dxsL

The first integral Ii does not pose any problem for numerical evaluation. It may be noted that

and, therefore,

where N is the number of division of length, h that we make for the interval 0 to x, and ux9. It can easily be shown that,

1 -4

I, <P

-{Gp/2)2

2. Choosing p = 10'6, it is easily seen that the relative error is less than 3 x 10"6 even for adifficult (0,P) value of (0.02, 0.00032). With higher values of N, even better accuracies can beobtained. We have presented in Table 1.3.1, the values of the resonance integral J calculatedby numerical method (with relative error less than 3 x 10"6) along with the values fromstandard compilation [1]. The values of standard compilation are in error to the extent of1.5%.

Table 1.3.1 Comparison Of Exact Values Of J(0,(3) With Those Of StandardCompilation For 3 = 2s x 10L-5

e0.020.030.040.050.060.070.080.09

Exact210.6159.3134.3120.1111.3105.6101.698.78

Standard compilation208.6157.3132.4119.4110.8105.4101.598.75

References

1. F.T.Adler and L.W.Nordheim, "Tables for the Computation of Resonance Integrals",GA-377, General Atomics Division, General Dynamics Corporation, 1958.

2. R.Harish and R.S.Keshavamurthy, "Proc. Tenth National Symposium on RadiationPhysics (NSRP-10), IGCAR Kalpakkam, Aug. 17-20, 1993.

1-5

CHAPTER 2

PFBR AND KAMINI CORE PHYSICS

2.1 Absorber Rods in PFBR(S. J. Collins, P. Mohanakrishnan, T. M. John)

The control of the fission chain reaction in PFBR during normal operation and anyabnormal situation requiring reactor shutdown, is achieved through the insertion or withdrawal ofneutron absorber rods (AR),in the core. Based on the safety criteria [1], the reactor has beenprovided with two independent and diverse shut down systems (SDS), one of which is used forthe reactivity control and power regulation during the normal operation of the reactor and theother, along with the first system, for the rapid shutdown (SCRAM) of the reactor. The ARswhich constitute the first system are called the Control and Safety Rods (CSR) and those of thesecond system are called the Diverse Safety Rods (DSR). The following studies have been carriedout on the AR of PFBR.

Worth Requirement: The excess reactivity of PFBR core at beginning of life is estimatedto be 9700 pern. Out of this, 3700 pem is to be controlled by diluent subassemblies. With ashutdown margin requirement of 5000 pern, the reactivity to be controlled by AR become 11000pcm. Allowing for 15% uncertainty in computations and 10% allowance of AR burnup, the totalreactivity requirement of AR become 13750 pcm. The division of the reactivity worth requirementbetween DSR and CSR is 3750 pcm and 10000 pcm respectively. Calculations showed that 3DSR with 50% B-10 enrichment provide total worth of 4000 pcm. Worth of the 9 CSR with 63%B-10 enrichment is 10110 pcm [1,2]. These worths meet the requirements for safe shutdown ofthe reactor. The worth computation for different AR configurations is done by the 3D diffusiontheory code, TREDFR and using the neutron cross sections in four energy groups. Thehomogenisation of the absorber pins is done using the flux distributions from the cell calculationsemploying the code, COHENT.

Shadow and Antishadow Effect: The worth of an AR is affected by the presence ofother rods in its neighbourhood. Due to the strong absorbing nature of the rods there is fluxdepression in the vicinity. If another AR is inserted in the flux depression zone of the first rod, theworth of the second absorber will be reduced. This phenomena is called the shadowing effect. Areverse effect called antishadowing effect is seen when the two ARs are outside the domain ofinfluence of each other. Due to this effect the combined worth of more than one ARs is higherthan the sum of their individual worth. It is always safer to select the AR positions so that theywill have the maximum antishadowing when all the rods are inserted, ensuring the maximumnegative reactivity insertion during any SCRAM. Table 2.1.1 gives [2] the total worth of theabsorber rod system and the antishadowing effect for various combination of the ARs. The worthof the ARs are higher when they are individually pulled out. This is again due to theantishadowing effect. Table 2.1.2 provides the reactivity worth of the absorber rods for some suchconfigurations.

2-1

Absorber Rod Life: During the normal operation, CSR are inserted into the core by 43cm to compensate the excess reactivity of 3120 pcm at the beginning of a cycle and are graduallywithdrawn. At the end of the cycle, they are at about 13 cm (500 pcm reactivity ) inserted into thecore. Due to neutron capture, the B-10 content gets depleted and so the worth graduallydecreases. The neutronic life of the AR is taken as the time for 10 % decrease in their worth fromthe original value. Assuming an average insertion of 28 cm into the core for the absorbers, therate of neutron capture in the different segments of one AR is as given in Table 2.1.3.

Table 2.1.1: Absorber Rod InsertionCase description

Single DSR inAUDSRinSingle inner CSR inSingle outer CSR inOne inner & one outer nearby CSR inAll inner CSR inAll CSR inAll Rods inTwo DSR inEight CSR in (one inner CSR out)Eight CSR in (one outer CSR out)

Worth from the All Rods Out ConfigurationAK(pcm)12304191122373217054166996513796260482878590

AK/K(pcm)114640171139679159639921011014562245882668595

Sum of AK/Kfor AR (pcm)

3438

18183417749110929229263526812

Antishadowing

17%

-12%17%35%33%7%30%26%

Table 2.1.2 : Absorber Rod Withdrawal Worth (pcm) from the All Rods In ConfigurationCase descriptionSingle outer CSRSingle inner CSRTwo near by CSR (one inner and one outer)All DSR

AK1468128636183831

AK/K1549135738194044

Antishadowing

31%

Table 2.1.3: The Capture Rate (s' ) in an AR at Full PowerAbsorber Segment

28 cm inside the core

30 cm inside the blanketsegment above the blanketTotal

Inner ringAR2.9xlO17

9.5xlO16

2.6xlO16

4.1xlOn

fpd for 10%worth reduction685

2464152692059

Outer ringAR2.3xlO17

7.7xlO16

2.1xlO16

3.3xlO17

fpd for 10%worth reduction855

3048190102563

Even though the worth reduction of the total AR happens only after 2059 fpd (11 refuellingcycles), the worth reduction in the portion inside the core happens in 685 fpd (3.7 cycles).Corresponding numbers for the AR of the outer ring are 2563 fpd (13.8 cycles ) and 855 fpd (4.6cycles) respectively. Since the worth of the absorber portion inside the core is getting depletedmuch faster than in the average depletion rate, it is advisable to replace the AR of the CSR after 6

2-2

cycles of reactor operation. An absorber replacement scheme in which 3 absorbers of the CSR (1of the inner ring and 2 of the outer ring) replaced in every alternate cycles will satisfy thisrequirement. Such a scheme would also ensure that the AR do not reach their end of life at thesame time. Since the DSRs are always kept out of the core, there is almost no burnup for them.Hence their life is determined from other considerations.

References1. T.M. John, P. Mohanakrishnan, Om Pal Singh and S.M.Lee, "Absorber Rods", Report

PFBR/01113/DN/1029,Nov. 1997.2. J. Collins and P. Mohanakrishnan, "Reactivity Worth of Control and Safety Rods and

Antishadowing Effects", Report PFBR/01113/DN/1028, Aug. 1997.3. J. Collins, P. Mohanakrishnan, "Reactivity Worth of Diverse Safety Rods", Report

PFBR/01113/DN/1027, April 1997.

2.2 RefueUing Interval for PFBR(T.M. John)

The refuelling interval (RFI) is an important parameter in reactor design. It affects coreaverage discharge burnup, excess reactivity and consequently the control requirements, breedingratio, in-vessel fuel storage, reactor availability, maintenance planning, thermal cycling, andoutpile fuel inventory. So, the effect of RFI on various parameters has been studied [1,2].

Due to the difference in the flux level and the fissile enrichments for the fuel subassemblies(FSA) loaded at different locations in the core, their burnup rate is dependent on the positions inthe core. For the target peak burnup of 100,000 MWd/t for the fuel, the residence time of theFSA inside the core varies from 555 full power days (fpd) at the core centre to 836 fpd at the coreperiphery (see Fig. 2.2.1) where the FSA locations in one twelfth segment of the core is shownaccording to their positions in the hexagonal arrangement. Under the batch mode refuellingscheme all the subassemblies will not reach their target burnup limit, because of the distribution of

residence time of the FSA atdifferent positions in the core.Hence, there is a loss in the peakbumup achieved by the FSA. Fig.2.1.2 shows the average of theloss in the peak burnup per FSAthat occurs as a function of theRFI. The curve shows distinctminima in the burnup loss forcertain RFI. These minimacorrespond to those days forwhich the residence time of thepeak rated FSA in the corebecome an integral multiple of theRFI, because of the fact that for

Number of equivlaentFSA in the core

Fig.2.2.1: The Residence Time (fpd) of FSA in the corethose RFI many of the FSA close

2-3

to the core centre nearly reach their target bumup. For shorter RFI the minima do not show veryhigh gain in the loss of bumup as most of the FSA reach close to their target peak bumupirrespective of the RFI. Some of the prominent minima correspond to 62, 69, 91, 110, 138, 185and 277 fpd. Table 2.2.1 presents the dependence of the important parameters on the RFI.

40000

80 210 240Refuelling Interval, d

330 360

Fig. 2.2.2: Average Loss of Peak Burnup per FSA for Different RFI

Table 2.2.1: The Reactor Performance for ]Parameter

Number of fuel cycles for Peak rated FSAAverage discharge bumup (MWd/t)Number of fuel SA discharged per cycleReactor availability (%)[n-vessel locations for irradiated FSABumup reactivity swing per cycle (pcm)In-vessel Cooling time (d) *Fuel discharged per cycle from core (kg)Time available for reprocessing (d)Required fuel reprocessing capacity (kg/d)Number of fuel S A reprocessed per yearOutpile Pu inventory (t)Extra Reactor availability due to fuel handling (d)Revenue due to power generation (Rs, million)Loss in revenue due to reprocessing (Rs., million)Extra inpile Pu required (kg)Extra outpile Pu required (kg)Extra interest on the fuel cost (Rs, million)Net gain over 62 d interval (Rs., million)

Different RFIRefuelling Interval (fpd)

6297081018.479.755610302339337811.9686.391.5930000000

6987070620.580.0841116017210398711.9586.841.3311.1975.220.494-262-30.96105.69

9166950727.580.80551550224139411312.3489.131.7853.84241.983.041219224.48214.46

1105681753481.13681860271172413612.6791.531.8395.03316.895.492024631.92279.48

1384677514381.52432325169218017012.8292.711.8616.44405.596.713226836.00362.88

18536751257.881.93583100226293019015.4293.401.8767.87495.817.504928339.84448.47

27726569789.582.21904650337453833713.4696.932.9058.90560.7611.17861312167.76381.83

2-4

The main contribution to the dependence of reactor availability on the RFI comes from thefact that at every shutdown, three days time is required for the change of state of the reactor,which means that more the number of shutdowns per year, more days are lost for powerproduction. The invessel storage locations increase with RFI. The fluctuations from the linearbehaviour is due to the mismatch between the 180 days in vessel cooling time required and theactual residence time permitted by the selection of RFI. The outpile inventory depend on thenumber of cycles of reactor operation between the first fuel discharge and the time for thosedischarged fuel assemblies to return to the core after reprocessing and fabrication. It is estimatedthat a fixed period of 260 days will be required for the transportation of the assemblies betweendifferent service facilities and other contingencies. Higher outpile inventory leads to higher loss ofinterest from the hold up of plutonium. The economics of higher RFI is computed with respect tothat of 62 fpd, which is the shortest one selected for the study.

Short refuelling intervals are beneficial in terms of high average fuel burnup for a giventarget burnup and minimisation of excess reactivity, control requirements, total fissile inventory inthe cycle and reprocessing and fabrication plant capacities. These advantages have to be balancedagainst increased thermal cycling and loss of reactor availability due to increased downtimes andthe need to distribute the maintenance activity over several shorter periods. The presentoptimisation gives 185 fpd as the optimum refuelling interval from the economics point of view.

References1. S.Govindarajan and P.V.K. Menon, "Refuelling Interval for PFBR Oxide Core", Report

PFBR/01113/DN/ 1020/R-A, July 1990.2. T.M.John, "Refuelling Interval", PFBR/01113/DN/1O31, Dec. 1998.

2.3 Criticality Safety Study of PFBR Storage Bay with Borated Steel(P.T.Krishnakumar and P. Mohanakrishnan)

PFBR fUel subassemblies are stored in storage bays before and after irradiation. Due tolarge number of subassemblies that may be stored in the storage bay, it is important to assess thecriticality of such interacting array of fissile subassemblies under accident scenarios while handlingthese fuel subassemblies. The criticality calculations have been performed under accident scenarioof clustering of few blanket subassemblies under water flooded conditions and using the SETRand the WIMS cross-section library [1]. The calculations have also been done with fuelsubassemblies in borated steel casings [2]. Borated steel has been widely used as casing in the fuelstorage of light water reactors and in KNK-JJ fast reactor.

As an accident scenario, we considered clustering of two, three, four and five fuelsubassemblies and in each case there is water column of 18 cm thickness outside the clusteredsubassemblies. The Keff is calculated in each of the above cases. The K^s in each case enables toestimate the maximum number of subassemblies that can be clustered under handling incidentalconditions such that K ^ is less than the IAEA specification of 0.95 [3]. The results of thecriticality calculations in ID using transport code, ANISN for clustered configuration and cross-sections derived from WIMS and homogenised by lattice cell code IGC-SMAXY is tabulated inTable 2.3.1. As expected, it is seen from the table that the K ^ value with subassemblies in borated

2 - 5

Table 2.3.1 : Criticality of Storage Yard Cluster ConfigurationNo.

1.2.3.4.5.

No. of S A in cluster(eq. radius in cms.)one (7.08)Two (10.01)Three (12.26)Four (14.16)Five (15.83)

K infinity withoutborated steel0.69240.81400.88650.93680.9746

K infinity with 0.1%borated steel0.65360.78410.86220.91650.9572

steel casing is much less than that without borated steel. The value of Keff without borated steelcasing for a cluster of three subassemblies is 0.8865 and the corresponding value with 0.1%borated steel is 0.8622. Taking the uncertainty in WIMS cross-section set and calculation modelto be 10%, it is found that in accident scenario, only three subassemblies give Keff < 0.95 andanything more than three is unsafe as per the IAEA specification of 0.95.

References1. P.T.Krishnakumar," Criticality Safety of PFBR Blanket Subassemblies in Blanket Storage

Yards", RPD/CPOS/56 Sep 1992.2. P.T.Krishnakumar, "Criticality Safety Study of PFBR Fuel Subassemblies With Borated

Steel Casings in Handling Incident Conditions: Preliminary Studies", RG/RPD/CPS/4,Dec 1997.

3. IAEA Safety guide 50-SG-D10, "Fuel Handling and Storage Systems in Nuclear Powerplant".

2.4 Initial Measurements of Neutron Flux in KAMINI Reactor(D.K.Mohapatra, S.Sivakumar, C.P.Reddy and P.Mohanakrishnan)

Kalpakkam Mini Reactor (KAMINI) is a neutron source facility for neutron radiographyand neutron irradiation experiments. This has three horizontal beam tubes adjacent to the core inthe north, south and west horizontal directions and one vertical pneumatic fast transfer facilitywhich runs along the core reflector junction from the west beam tube to the end of the top axialreflector [1]. Measurements of the neutron flux at the ends of three horizontal beam tubes havebeen carried out at a power level corresponding to 1.7xlO'n A on the third range of Multi RangeDC (MRDC) channel and at a power level corresponding to 5xlO"10 A on MRDC channel, for thepneumatic rabbit and core cover plate locations [2]. The main purpose of performing thisexperiment has been to give a proper input towards the design of the shields for these tube endsand to plan subsequent irradiation studies. The measurements have been done using the foilactivation technique with gold, manganese-copper (80-20) alloy and indium foils. These activationexperiments are done with and without Cd envelope to evaluate the epithermal component of theflux. On an average, the foils were irradiated for about 1.5 to 2.1 h without Cd cover and forabout 3 h with Cd cover. The activity of the samples is measured by a Nal detector employing aMulti Channel Analyser. The results are given in Table-2.4.1. By extrapolating the measured fluxvalues linearly, one can estimate the neutron flux in different locations (see Table-2.4.2 for 5 kW

2 - 6

and 30 kW power). With the power calibration experiment it is found [3] that 1.85 x 10"10 Areading on third range of MRDC channel corresponds to 22 W. Thus the two power levels usedfor irradiation at the beam tubes and the pneumatic rabbit location are 20 W and 6 W respectively.

Table 2.4.1: Measured Fluxes (n/cmVs)LocationPneumatic rabbitCore cover plateWest beam tube inner endWest beam tube outer endNorth beam tube outer endSouth beam tube outer end

Power (W)6.06.06.020.020.020.0

Average flux2.64xlO8

5.71xlO7

3.33xl08

1.12xlOs

3.07xl04

8.5xlO4

Flux above 0.625 eV---2.17xlO4

1.09xl04

3.2x104

Table 2.4.2 : Estimated Fluxes(n/cni2/s)Locations

Pneumatic rabbit irradiation locationCore cover plate irradiation locationWest beam tube outer endNorth beam tube outer end

5kWtTotal2.2x10"4.8xlO10

2.8xlO7

7.8xlO6

Above 0.625 eV--

5.4x106

2.7xlO6

30kWtTotal1.32xlO12

2.85x10"1.68xlO8

4.61xlO7

Above 0.625 eV--

3.26x107

1.64xlO7

South beam tube outer end 2.12xlO7 8.1xlO6 1.28xl08 4.85xlO7

References1. C.S.Pasupathy et al., "KAMINI REACTOR System Manual", Aug. 1996.2. D.K.Mohapatra et al., Internal Report RG/RPD/CPS-1, Sept. 1997.3. DVS Ramakrishna et al., "Power Calibration for KAMINI Reactor" BARC Note, 1997.

2.5 PWR Lattice Benchmark Problem Solutions(Mohanakrishnan and S.M.Lee)

The ANS-19 standards subcommittee of Reactor Physics Division and ComputationalBenchmark Problem Committee of Mathematics and Computations Division have sponsoredPWR uranium fuelled near critical core benchmarks. They essentially consisted of one uniformsquare lattice core (type A), a core with water holes in some lattice positions (type B) and a corewith absorber rods and water holes (type C). The benchmark has been analysed using the neutroncross sections of the WTMS 69 group data library of early 1970's [1], The two changes areimplemented in this library. That is, use of fission spectrum recommended in the 1986 version ofWTMS [2]) library which is equivalent to a Maxwellian of energy (kT) =1.37 MeV and forhydrogen bound in water, a* is obtained by a weighted column sum correction of total crosssection as given in Ref.[2]. The weighting function is derived from current spectra obtained by BN

calculations in water[3]. Computer code MURLI [4] based on collision probability method is usedto get a spectrum typical of the PWR lattice (type-A). This spectra is used to condense the 69group cross section library to a 27 group library. Computer code IGC-SMAXY [5] is used toperform the cell calculations of fuel cell, in type-A, fuel cell and water cell in type-B and fuel cell,

2-7

water cell and pyrex cell in type-C. The water cell and pyrex cell are modelled using the conceptof super cell where a homogenised fuel cell region surrounds the heterogeneous cell to behomogenised. Homogenised and condensed 6 group cross sections are obtained for cells in eachtype. These are directly used in core calculations without performing assembly calculations.

Core calculations are performed by computer code CEMESH [6] in 2D. This code solvesmultigroup diffusion equation by centre mesh finite difference scheme. A 2 x 2 mesh is consideredin each cell, resulting in a total of 94 x 94 meshes in quarter core. Axial leakage is accounted byusing buckling of 0.00037 cm'2. The effective multiplication factor prediction appeared to showthe need to use transport theory for core calculations especially in type-C (Table 2.5.1). So thecore calculations are repeated with SN computer code TRITAC [7] by imposing zero leakageboundary in the axial direction. The axial leakage is then accounted by buckling corrections. Forany assembly type, identical cross sections are used for both diffusion and transport calculations.Effects of increasing SN order as well as effect of using 4 x 4 meshes per cell were also studied.These K6ff are also quoted in Table 2.5.1. The diffusion theory calculations predict the lOr quitewell in a band of-0.86% to +0.09% of measured value. The best results are obtained by usingfiner spatial meshes in S4 transport calculations, i.e., -0.63% to +0.07% of measured value. Themeasured and calculated power distributions are compared in Fig. 2.5.1. Diffusion theorypredictions are reasonably good. It is found that the improvement in Keff predictions shown bytransport calculations is not clearly reflected in power distribution. This may be due tocancellation of errors in diffusion theory predictions. More insight into the possible areas ofimprovement will be possible when the result of all other groups participating in this benchmarkactivity are compared in October 1998.

AssemblyType ATypeBTypeC

Table 2.5.1 : Effective MultiplicationDiffusion

theory0.99900.99980.9921

S4:2x2 mesh per cell0.99981.00160.9942

Factors of PWR BenchmarksTransport theory

Ss:2x2 mesh per cell0.99971.00170.9936

S4:4x4 mesh per cell0.99991.00140.9944

Measured effective multiplication factor = 1.0007

References1. C.J. Taubman "The WIMS Nuclear Data Library", AEEW-M, 1324 (1975).2. M.J. Halsall and C.J. Taubman "The 1986 WIMS Nuclear Data Library", AEEW-R-2133

(1986).3. R.D.S. Yadav and Vinod Kumar et al., "Generation of Transport Cross Section of

Hydrogen using BN Approximation", Paper presented as BARC-IGCAR discussionmeeting on Reactor Physics Methods, BARC, Sept. (1990).

4. H.C. Huria "A Multigroup Integral Transport Theory Code for Thermal ReactorLattices", Atomkernenergie 31., 87(1978).

5. P. Mohanakrishnan, "LWR Assembly code SMAXY and its IGCAR Version",RG/RPD/CPOS/21, IGCAR, October (1990)

6. V. Jagannathan et al., The "PHANTOM-CEMESH Code System for PHWR DesignComputations" BARC-1444 (1988).

7. M. Bando et al., "3-D Transport Calculation Methods for Eigen-Value Problems UsingDiffusion Synthetic Accelaration", Jr. of Nucl. Sci. and Tech. 22 p.841 1985.

2-8

WaterHole

LegendDiffusionTransportMeasured

1.1511.1441.1481.0601.0481.036

1.0421.0351.0270.9540.9380.945PyrexRod

1.0321.0251.0450.9870.9761.0010.9000.8840.9010.9110.8970.914

1.0351.0301.0570.9910.9810.9820.8990.8860.9000.8730.8570.854PyrexRod

1.0411.0391.0470.9600.9480.962PyrexRod

0.9110.8980.9330.9660.9600.9701.0691.0761.071

1.0801.0791.0881.0471.0431.0700.9870.9781.0011.0451.0431.0491.0991.1061.0971.1431.1561.1401.1771.1961.164

1.1171.1201.1241.1101.1121.1051.1031.1071.0871.1211.1241.0881.1491.1591.1381.1751.1901.1951.1971.2181.1991.2121.2351.206

Fig. 2.5.1 Comparison of Pin Power Distribution (Assembly Type C)

2.6 Design of a Magnetic Sensor based on Covariance Analysis(P.T.Krishnakumar)

Proton beams have potential advantage in radiotherapy. It is essential that any machinechosen for therapy application should be simple to operate, reliable and cost effective.Accelerators producing such beams have immense use in micro dosimetry. In the presentinvestigation we considered the resonance reaction of 27Al(p,y)28Si to calibrate a NuclearMagnetic Resonance (NMR) flux meter which controlled the magnetic induction in a Van de graffaccelerator. The resonance of the above reaction were used to determine the relation between theproton energy and the NMR frequency. The study revealed that by proper design of ion beamsteerer and reduction of magnetic hysteresis, one can design the magnetic sensor with leastsystematic uncertainty. The details of the above study can be seen in Ref. 1

1.ReferencesP.T.Krishnakumar, "Design Of Magnetic Sensor For Process Engineering With LeastCalibration Uncertainty", Proc. Workshop on Automation in Process Engineering andmanufacture, Kalpakkam, Nov. 1997.

2-9

3.1

CHAPTER 3

FBTR CORE PHYSICS

Calibration of Control Rods for 26 Fuel Subassembly Core(A. John Arul)

Reactivity worth of control rods change when the core configuration is altered. So afterloading the 26 th fuel subassembly (SA), the control rods were calibrated on 9th Nov. 96 usingthe rod raising and lowering method. Inverse kinetics is used to derive the reactivity as a functionof position from the recorded power profile. The inlet sodium temperature was at 186° C. Thereactor power was changed between 10 kWt and 50 kWt. The total worth of all the six rods is6700 pcm [1]. This is 72 pcm more than the worth of control rods with 25 SA core [2].

The control rods were recalibrated on 4 th Dec. 96 following the movement of 26 th SAfrom location 03-18 to 03-10. The measured worths after shuffling and before shuffling are givenin Table 3.1.1. The total worth remains unchanged as expected.

Table 3.1.1: Integral Worth of Control RodsControl Rod

ABCDEF

Integral Worth (pcm)Before Reshuffling

1288.61090.01091.71049.71046.71135.7

After Reshuffling1086.21063.81119.31269.11068.11094.6

1.

2.

3.2

References:Sivakumar, A. John Arul, C.P. Reddy and V. Sathiamoorthy, "Calibration of Control Rodsfor 26 Fuel Subassembly Core," RPD/LMS/FBTR/01100/CR/043.C.P. Reddy, A. John Arul, S. Sivakumar and V. Sathiamoorthy, "Calibration of ControlRods for 25 Fuel Subassembly Core", RPD/CPOS/FBTR/01100/CR/034.

Power Distribution in 26 Subassembly Mark I Core(A. John Arul and C.P. Reddy)

When the 26 SA core was operated on power, the AT across Mark II fuel SA at 03-18location was found to be 92.2° C instead of the expected value of 59.6° C [1]. Subsequently theMark JJ SA was shifted to 03-10 from 03-18 location. The AT observed at 03-10 is 84.5°C [2].The power generated in each of the SAs in the core was calculated for three cases, i.e., 26 SAcore with and without source, 26 SA core with Mark II SA at 03-18 location and 26 SA corewith Mark II SA at 03-10 location. Calculations were performed with 2D diffusion code, 2DHEX

3 - 1

[3]. It was found from the results of the calculation that the power of the SA closest to the source(in the third ring) increases by 2 % due the source. The power of Mark II SA decreases by 2.1 %when it is moved to 03-10. After the movement, the fuel SAs close to the location 03-10 show1.5 - 2 % increase in power whereas the fuel SAs close to 03-18 exhibit 1.5 % decrease in power.Fig. 3.2.1 gives the fuel SA configuration while Fig. 3.2.2 gives the corresponding power profiles.This study explains the change in SA powers caused by Mark II SA because of its movementwithin the core and showed that the anomalous AT was due to causes other than subassemblypower.

AfterFH

FSA(03/10)-> 03/09

FSA(03/18)-> 03/10

03/07 02/06 02/0? 02/OB

04/08 03,06 02,06 01,5i 01*4 S o5 3 04,16

04,07 0*05 0*04 ,52 00,55 5S,0 0314 04,19

04,05 03,04 02/03 01/01 01/06 02/11 03/15 04I2C

03,03 02/02 02/01 02/12 03/16

04/04 03,02 03,01

a yyyy - Value After FH

xxxx - Value Before FH

Fig.3.2.2. Core Power Before and After Fuel Handling

1.

2.

3.

3.3

M-IFSA M-ILGFSA M-ll FSA SOURCE

Fig.3.2.1. Core Configuration

ReferencesS. Varadarajan et, al, "Analysis of Core and Reactor Vessel Temperatures with CCPM at80 mm", 1996.S. Varadarajan and G. Srinivasan, "Investigation of the Core Temperature Analysis ofMark JJ SA in the 26th SA Core, "ROD/FBTR/S-RS-02/31400/SAR-015, 1996.A. John Arul and C.P. Reddy, "Subassembly-wise Power Distribution for 26 SA Mark ICore and its Dependence on Source SA", RG/RPD/LMS-7.

Observed Power Coefficient Of Reactivity For The 24 Subassembly Mark I Core(A. John Arul and C.P. Reddy)

After loading the 24 th fuel SA,FBTR was operated between Dec. 93 and Jan 95. Themaximum operating power was 10.5 MWt. The power coefficient of reactivity was analysed [1]based on the data acquired during power raising and lowering operations. Since the power raisingand lowering operations were done within few hours, the burn up reactivity loss was neglected.The observed reactivity change [2] was corrected for temperature change using temperaturecoefficient measurement done on Sept.90. The reactivity loss with change in power is fitted to athird degree polynomial (Fig.3.3.1a). The power coefficient of reactivity is obtained as a functionof power as shown Fig.3.3.1b. The average over the four curves is given in the Table 3.3.1.

3 - 2

Table 3.3.1: Power Coefficient of Reactivity as:Power MWtPower coef.

175

263

353

445

539

i Function of Power, (]635

733

833

pcm/MWt)935

1039

6000Fig.3.3.1a Reactivity vs Power

4000

oQ.

100

50 -

Fig.3.3.1b Power Coefficient vs Power

1.

2.

ReferencesA. John Arul and C.P. Reddy, "Observed Power Coefficient of Reactivity for the 24 SAMark I core", RG/RPD/LMS-4, 1996.C.P. Reddy et. al, "Consolidated Control Rod Worth Curve for 24 FSA Mark I Core",RPD/CPOS/FBTR/0100/CR/027, 1994.

3.4 Reactivity Loss Rate Due To Burnup in FBTR Small Carbide Core At 25000MWd/t Burnup and at High Power Operation(A. John Arul, C.P. Reddy and A.G. Rafi Ahmed)

From October 85 to May 96, FBTR had been operated with 23, 24 and 25 fuel SAgenerating 1432 MWd of energy [1]. The total reactivity loss observed during this period fromexcess reactivity and subassembly worth measurements is 3017 pcm. Therefore, the averagereactivity loss rate for this period is 2.1 + 0.4 pcm/MWd.

3-3

FBTR has been operated subsequently with 26 and 27 fuel SA. With 27 fUel SA, it wasoperated at 11.5 MWt during October '97 for about 22 days. Using the critical heights observedat the beginning of this phase of operation and at the end of this phase of operation, the observedreactivity loss is 321 pcm. The bumup during this period is 237 MWd. Therefore, the averagebumup loss rate is 1.35 pcm/MWd. A quadratic fit to the observed data [2] gives an averageburnup loss rate of 1.35 ±0.1 pcm/MWd for the fifth irradiation campaign. The reactivity lossrate decreases to 1.2 pcm/MWd (end of campaign) from a maximum value of 1.5 pcm/MWd inthe beginning of the campaign.

£oQ.

>—**0)

"Strto

o

>

act

DC

c

1.8

1.6

1.4

1.2

v 300 mm

\

355 mm \V\

\ \\ ^\ \\ \\ \

~ \ \\ \\ \\ \\ \ 405 mm

\\ 405 mm

-

1 1 1 1 1 1 1 1 1 I 1 L- 1

- 24- 25-• 27

SA CORESA CORESA CORE

365 mm

\

\

\

\\\405 mm

500 1000 1500 2000 2500

Burnup ( MWd )

Fig.3.4.1. Reactivity Loss Rate as a Function of Burnup

The reactivity loss rate calculated using 2D diffusion and perturbation codes is 1.2pcm/MWd [3]. This calculation accounts for the reactivity loss due to the destruction of fissilematerials only. The reactivity loss due to the axial swelling of the fuel pins is calculated [4] using

3-4

data obtained from PIE and 2D perturbation code NEWPERT. The model uses five radial zones.The axial swelling effect has two components. One is the density effect which is negative and theother is boundary movement effect at constant density which is positive. The reactivity changedue to increase in the core height (boundary) depends on the position of the control rods, i.e.,when control rods are at the core mid plane the boundary effect is less compared to the case whenthe control rods or either fully out or fully down. Therefore the reactivity loss rate is expected todecrease as control rods move away from the core mid plane. The calculated value of this effect isin the range of 0.05 to 0.06 pcm/MWd. The calculated reactivity loss rate including the axialswelling of fuel pins is 1.54 pcm/MWd [4]. The observed reactivity loss rate as shown in Fig.3.4.1 has a decreasing trend with burnup and movement of control rods out of the core. Thisvariation is observed to be around 0.3 to 0.6 pcm/MWd.

For 27 fuel SA core operation, the observed average value of reactivity loss is less thanthe calculated value by about 0.2 pcm. This indicates that the swelling component may have comedown with burn up.

As expected from the calculations, the reactivity loss rate is higher with control rods in thecore mid plane compared to the case with control rods out. The variation of reactivity loss ratewithin a power campaign is about 0.6 to 0.3 pcm/MWd. Observations for 24, 25 and 27 FSAcore, consistently exhibit this kind of behaviour [2,4]. Calculations with improved cross sectionsets and 3D models are planned to improve the prediction capability.

References1. CP. Reddy, S. Sivakumar and S. Varadarajan, "Burnup Status of FBTR Fuel SA",

RG/RPD/LMS/1, Sept. 1996.2. A.John Arul and A.G. Rafi Ahmed, "Observation of Burnup Reactivity Loss at High

Power Operation of FBTR During October 1997", RPD/LMS/FBTR/01100/CR/048.3. A. John Arul, C.P. Reddy and S.M. Lee, "Reactivity Loss Due to Burn up in FBTR Small

Carbide Core", RPD/CPOS/66, May 1992.4. A. John Arul and C.P. Reddy, "Reactivity Loss Rate due to Burnup in FBTR Small

Carbide Core at 25000 MWd/t Burnup", RG/RPD/LMS-10, July 1997.

3.5 Calibration Of Control Rods For 27 Subassembly Core(S. Sivakumar, C. P. Reddy and A. John Arul)

The control rods have also been calibrated after loading 27th fuel subassembly and usinginverse kinetics calculations. Of the six rods, two opposite rods are selected. One of the twoselected rods is kept fully inserted and the other fully withdrawn. The remaining four rods arebanked to achieve criticality. The fully inserted rod is withdrawn so as to add a small positivereactivity and the reactor is put on positive period. After waiting for about two minutes the otherrod is inserted to add negative reactivity and the reactor goes on negative period. The process isrepeated until the initially withdrawn rod is fully inserted and the initially inserted rod is fullywithdrawn. During the process, power variation data is recorded at one second intervals using anindustrial PC and the control rod movements are recorded separately. This experiment is repeatedfor the other pairs also. Using the delayed neutron A and 3i values for FBTR Mark I core and the

3 - 5

&

power variation data recorded during the calibration, the inverse kinetics method is used todetermine the reactivity as a function of time. Using these reactivity values and the recordedcontrol rod movements the worths of control rods are obtained. The integral worths of the sixcontrol rods in the two experiments are given in Table 3.5.1 along with the previousmeasurements. A consolidated normalised integral worth curve is given in Fig.3.5.1 and thedetails are recorded in Ref. [1].

50 100 150 200 250 300

Position of Control Rod in mm

350 400 450

Fig. 3.5.1: S-Curve for Integral Worth

Table 3.5.1: Measured Control Rod Worths in pcm for Various ConfigurationsControl rod

ABCDEF

Total

23 S A Core11008881073110786210676097

24 S A Core1085.9892.51048.81092.71121.51152.86394.2

25 S A Core1107.31097.11107.51093.31089.11133.76628.0

26 S A Core1288.641090.041091.681049.741046.691135.696702.48

27 S A Core1195.91055.31073.51189.81016.31085.86616.6

Reference1. S. Sivakumar, A. John Arul and C. P. Reddy, "Calibration of Control Rods for 27 Fuel

Subassembly Core", Internal Report RPD/LMS/FBTR/01100/CR/046, 1997.

3 - 6

3.6 Measurement of Isothermal Temperature Coefficient for 27 Subassembly Core(S.Sivakumar and C.P.Reddy)

The isothermal temperature coefficient, defined as the ratio of change in reactivity tochange in sodium temperature, is to be measured whenever the core configuration changes. Forthe 27 subassembly core of FBTR, the measurement was carried out on 24 - 25/6/97. Themeasurements were carried out for the temperature range 180°C to 325°C. The experimentalprocedure is given in Ref.l. The results of the experiment are given in Table 3.6.1.

Table 3.6.1; Isothermal Temperature Coefficient (pcm/°C)Temperature range range(°C)

246.6-275.7275.7-300.2300.2-333.0333.0-284.1284.1-259.7259.7-249.5249.5-224.7224.7-179.6

Average

Coefficient-4.62-4.41-4.28-4.56-4.33-4.44-4.21-4.34-4.40

Reference1. S.Sivakumar and C.P.Reddy, "Measurement Of Isothermal Temperature Coefficient With

27 Subassemblies", RG/RPD/LMS-9, 1997.

3.7 Thermal Neutron Flux Standard Facility(S.Sivakumar, C.P.Reddy and V.Sathiamoorthy)

A thermal neutron flux standard has been set up at IGCAR for accurate measurement ofneutron flux and for calibration of neutron detectors. Neutron flux measurements can be carriedout using either neutron detectors or by foil activation methods. Use of a detector needsknowledge of its sensitivity and intrinsic efficiency which can be obtained by calibrating it in anAccurately known flux. On the other hand for the activation method, it is necessary to obtain selfshielding factor, flux density depression, geometric correction, etc., which can be obtained byirradiating the foils in an accurately known flux. Thus whether one employs detectors or foilactivation methods, there is a need for a standard flux facility. Therefore, such a facility has beenset up as IGCAR.

The facility is primarily for a thermal neutron field, and essentially consists of a stack ofreactor grade graphite blocks (Fig.3.7.1) which moderate high energy neutrons emitted byembedded Americium-Beryllium (Am-Be) sources. Eight Am-Be sources, each of strength l.lx106 n/s, are placed symmetrically in two planes at a distance of 30 cm from the centre of thegraphite stack. Am-Be sources have been used as their half-life is 433 y and there is little changein source strength over a period of use of several decades. Also, the sources have an additionaladvantage of soft gamma ray emission avoiding shielding requirements. Due to the symmetrical

3-7 DEPOHTCP.Y LT^T' VC:Y

C •Tr em.'-;.-. Klv. •: .;:i-.;:0 035.

arrangement of the sources, a spatially uniform flux is obtained in the central region of the facility,eliminating the need for accurate positioning of the foils in the facility. Foils upto a radius of 30mm can be irradiated in the pits provided in a removable graphite plug. By removing the graphiteplug, neutron detectors with dimensions upto 10 cm x 10 cm x 35 cm can be inserted into thecentre of the facility for calibration. This set up can also be used to standardise various types ofneutron dosimeters. The flux at the central region of the facility was calibrated by absolutecounting done at BARC, and has a value of 7.0 x 103 n/cm2/s. Main features of the set up aresummarised in Table 3.7.1.

Table 1 3.7.1: Main Features of Thermal Neutron Flux Standard Facility

Kind of sourceStrength per sourceNumber of sourcesModeratorOverall sizeThermal flux at the centreRatio of thermal to total flux at the centreExtension of spatially uniform flux

Am-Be (a, n)l . l x l0 6 n / s8Graphite (Nuclear Grade)152 cm x 152 cm x 152 cm7.0 x 103 n/cm2.s0.730 cm

Irradiation Cavity

Am-Be Neutron source

Fig. 3.7.1:Thermal Neutron Flux Standard

3 - 8

CHAPTER 4

RADIOACTIVITY AND SHIELDING

4.1 Comparison of 2D Transport Calculations with Measured Detector Location Fluxand Cover Gas Activity in FBTR(R.S. Keshavamurthy, R. Indira, C.P. Reddy and A.K. Jena)

Transport calculations [1] for FBTR Mark I core have been carried out using 2D transportcode, DOT-IIW. The calculation geometry extended upto the end of the borated concrete inradial direction and to the top of sodium in axial direction. The calculations are done in 12 energygroups. Cross sections are obtained by collapsing the DLC-2 100 group cross sections with 100group neutron fluxes obtained from radial and axial ID transport calculations. Thirty materialregions are considered with sodium above the core divided into three regions to take into accountthe change in spectrum with distance in sodium. The estimated count rates obtained from thesecalculations at detector location is 810 cps/W. The measured count rate is 200 ± 5 0 cps/W.Considering the attenuation factor of ~ 106 in total flux between the core centre and detectorlocation, this comparison is fairly good. The Mn equivalent flux at the detector location iscalculated to be 2.8 x 109 n/cm2/s and that above the top of sodium along the central radial line tobe ~ 4.5 x 109 n/cm2/s. There have been no measurements of these fluxes in FBTR. However thevalues are comparable to those measured in RAPSODIE.

Cover gas in FBTR consists of 20% of argon and 80% of helium. Argon gets activated by(n, y) reaction and A-41 decays with a half life of 1.8 h. Argon activity is measured by takingsamples from the cover gas. A comparison of the calculated argon activity is an index of accuracyof calculated neutron flux above the top of sodium. Our estimation of saturated argon activity, Ais 0.0645 Ci/m3, if cover gas is not drawn for sampling. If the sampling is done at a constant rateS, the A-41 activity in cover gas is given by A/(X + S/v) where X is the decay constant of A-41and v is the cover gas volume. At present sampling rate S cannot be precisely controlled andranges between 20 cm3/s and 150 cm3/s in FBTR. Therefore, the computed activity is in the rangeof 0.05 to 0.06 Ci/m3. The measured value of A-41 activity in FBTR is 0.0145 Ci/m3. This againis a fairly good agreement considering the magnitude of attenuation.

References1. R.S. Keshavamurthy, et.al., "Comparison of 2D Transport Calculations with

Measuared Detector Location Flux and Cover Gas Activity In FBTR", Proc.National Symposium on Radiation Physics (NSRP-12), Defence Research Lab.,Jodhpur, Jan. 28-30, 1998.

4.2 Fission Product Gamma Spectrum in FBTR Mark II Core Subassembly(D. Sunil Kumar, M.S. Sridharan and A.K. Jena)

The fission product y-spectrum in FBTR Mark II core central subassembly has beencalculated. The y-spectrum in the active part of the subassembly arises due to fission of Pu-239

4 - 1

during reactor operation. The axial blanket containing fertile thorium forms part of thesubassembly. Calculations show that for reactor power of 40 MWt, the average fission power inthe axial blanket due to fast fission in Th-232 is about 50 kW. So the contribution of blanket isconsidered but contribution of U-233 that is bred in the blanket is ignored because fission powerdue to this is small.

The calculations are carried out using a code system GANGA [1]. The code uses a dataset of exponential fits that are based on ENDF/B-TV files [2]. These fits give gamma sourcespectra per unit fission for each isotope at any time after a fission event. These fits are integratedto obtain integral gammas in a certain energy group, for the given composition and irradiationhistory of the fuel. Results of the y-source spectra in the subassembly are given in Table 4.2.1 forcertain representative cooling times. Details can be found in Ref. [3].

Table-4.2.1: y-Spectrum for Irradiated Fuely- Energy (MeV)

0.01-0.020.02-0.030.03-0.060.06-0.100.10-0.200:20-0.400.40-0.600.60-0.700.70-0.800.80-1.001.00-1.501.50-2.002.00-3.003.00-4.004.00-5.005.00-6.006.00-7.50

TOTAL

Subassembly after 50 GWd/t Bui nupAfter cooling times of

10 s.235E+15.408E+15.360E+16.445E+16.907E+16.188E+17.159E+17.809E+16.856E+16.118E+17149E+17

.585E+16

.469E+16

.108E+16

.419E+15

.509E+14

.464E+13

.108E+18

100 s.581E+14.236E+15.188E+16.148E+16.478E+16.935E+16.844E+16.438E+16622E+16.646E+16.559E+16.292E+16.166E+16.142E+15.618E+13.604E+12.407E+10.536E+17

I d.248E+14.194E+15.170E+16.780E+15.204E+16.301E+16.301E+16.179E+16.324E+16.816E+15.444E+15.127E+16.922E+14.193E+13.809E+12.906E+11.270E+10.184E+17

10 d.137E+14.117E+15.127E+16.278E+15.765E+15.573E+15.164E+16.303E+15.183E+16.335E+15.679E+14.774E+15.440E+14.878E+12.351E+12.427E+11.162E+10.801E+16

100 d.766E+12.113E+13.249E+15.511E+13.113E+15.270E+13.243E+15.601E+14.720E+15.329E+13.7OOE+13.664E+13.271E+13.159E+10.191E+09.896E+08.121E+08.141E+16

References1. M.S. Sridharan and K.P.N. Murthy, "Data on Spectral Gamma Emission from a Burnt

PFBR Subassembly", Internal Report PFBR/01115/DN/1000/R-A, 1986.2. DC. George et al., "Delayed Photon Sources for Shielding Applications", LA Scientific

Lab., Report LA - UR 80 - 3305, 1980.3. D. Sunil Kumar, M.S. Sridharan, and A.K. Jena, "Fission Product y-Spectrum in FBTR

Mark II Core Subassembly", Internal Report RPD / RSS /124,1997.

4-2

4.3 Concentrations and Activities Of I, Ce, Ru and Pd Isotopes in FBTR Subassemblyafter SO GWd/t Peak Burnup(MS. Sridharan)

Atomic concentrations and activities of the fission product isotopes of I, Ce, Ru and Pd inan FBTR subassembly after 50 GWd/t peak burnup have been computed [1] using the code,CHANDY [2]. Results for the I isotopes are given in Table 4.3.1.

Table 4.3.1: Concentration and ActivitiesFBTR Mark-I Core Subassembly after 50

of Iodine Isotopes inGWd/t Peak Burnup

IsotopeI 127I 128I 1291130mI 130I 131I 1321133mI 1331134mI 134I 1351136m'1136mI 136I 137I 138I 139I 140I 141

Concentration (atoms)0.10295E+220.23535E+140.22886E+220.18719E+150.28819E+170.54325E+210.83667E+190.90552E+150.99480E+200.55620E+170.41532E+190.27803E+200.13964E+170.23179E+170.76746E+160.12020E+170.68110E+150.21698E+150.28996E+140.26795E+13

Activity (Ci)0.00000E+000.29426E+000.86529E-040.66259E+010.12192E+020.4650E+050.19056E+050.18847E+040.24767E+050.48778E+040.243 73E+050.21528E+050.26157E+040.52314E+040.29951E+040.91348E+040.19420E+040.15573E+040.63158E+030.11154E+03

1.

2.

4.4

ReferencesM.S. Sridharan, 'Concentration and Activities of I, Ce, Ru and Pd in FBTR SA after50 GWd/t peak burnup", Letter No. RPD/RSS/LET/83, April 1997.M.S. Sridharan and K P N . Murthy, "An Assessment of the Fission-Product Decay Datafor Decay Heat Calculations in Fast Reactors", Internal Report RPD/RSSP/10, 1987.

Neutron Flux and Gamma Dose Rate at Under-Vessel Detector in PFBR(R.Indira)

Neutron flux and y-dose-rate at under-vessel detector have been calculated for PFBR. Thegamma dose rate at the detector location is primarily due to the capture gammas and activationgammas from the primary sodium. Neutron detectors are placed under vessel for the purpose of

4-3

neutron monitoring during start up and power operation. The detector has 5 cm of graphite oneither side for thermalization purposes. There is 20 cm of lead to absorb the sodium capture andactivation gammas and thus reduce the gamma dose rate at detector location. The shieldconfiguration is shown in Table 4.4.1.

Table 4.4.1: Lower Axial Shield ConfigurationS.No.1234567891011

RegionCorelCore 2Blanket 1Blanket 2Blanket 3PlenumS A FootGP (top)GPGP (bottom)gap

Position (cm)0 - 7070 - 8888 - 9898 - 108108 - 118118 - 192192 - 218218 - 223223 - 273273 - 278278 - 293

S.No.1213141516171819202122

RegionCSS (top)CSSCSS (bottom)SodiumMVAr/NSafety vesselNitrogen

ss/csLeadGraphite

Position (cm)293 - 296296 - 416416 - 419419 - 520520 - 524524 - 544544 - 546546 - 606606 - 617617 - 637637 - 642

Neutron photon transport calculations have been carried out using ID discrete ordinatestransport code, ANISN and DLC-37 (100 neutron and 21 y-group) coupled library with SgPiapproximation. The core region is treated as equivalent spheres to the exact cylindrical geometrywith two enrichment zones. From the exit of the core, actual dimensions are used for all thezones. The composition of materials in lower axial shield zones are given in Ref. 1.

The core exit total neutron flux has been normalised to that of 2D-RZ diffusion theorycalculations. The total neutron flux at core exit is 2.2 E+15 n/cm2/s, whereas 2D-RZ calculationsgive a total neutron flux of 2.9 E+15 n/cm2/s. The total neutron flux at detector location turns outto be 6.8 E+04 n/cm2/s. The Mn thermal equivalent flux is 4.0 E+04 n/cm2/s. The attenuationfactor, i.e., the ratio of thermal equivalent flux at detector location to core central flux (with 2D-R-Z calculations as reference value) is 5.0 E-12.

At shutdown conditions, the neutron flux at the detector location is due to the inherentneutron source in the core. This is different for the fresh core and the equilibrium core. The totalneutron flux at core centre for fresh core is 7.8 E+05 n/cm2/s which corresponds to a shutdownpower of 0.12 W. For an equilibrium core, the total neutron flux and power are higher by a factorof 10. Using this, the thermal neutron flux at the detector location at shutdown condition turnsout to be 3.9 E-06 n/cm2/s for fresh core and 3.9 E-05 n/cm2/s for equilibrium core.

The attenuation factor and fluxes of PFBR have been compared with SPXI and Phenix inTable 4.4.2. The values for SPX I and Phenix have been obtained by extrapolation from themeasured count rates. It can be seen that as compared to PFBR, distance from core centre tomain vessel in SPX-I is much higher. This leads to a larger attenuation through higher thickness ofsodium. This has been offset by the use of neutron guides that have been used in SPX I toenhance the flux by a factor of more than 100. In Phenix, the distance from core centre to main

4-4

vessel is lesser by 120 cm, as compared to PFBR. This leads to higher flux level at detectorlocation in Phenix.

Table 4.4.2 : Comparison Of Fluxes With SPX- 1 And PhenixReactor core centre flux (n/cm2/s)Full powerShutdown (fresh core)Shutdown (equilibrium core)Distance from core centre to main vessel (cm)Ratio of detector location flux to core centre flux

PFBR8.0E+157.8 E+057.8 E+065305E-12

SPX16.0 E+151.9 E+062.0 E+07900

5 E-09®

Phenix6.7 E+15--4101.5 E-09

@with neutron guide tube

It may be noted that SPX I is provided with Helium-3 neutron counters for low powermonitoring (1 W to 10 kW) and fission counters for intermediate and high power monitoring (10kW to 3000 MWt) located under the vessel. In Phenix, initially boron coated counters and laterHelium-3 filled counters were used for neutron monitoring. Neutron guides were not provided inPhenix.

References1. R. Indira, "Neutron Flux and Gamma Dose Rate at Under-Vessel Detector Location in

PFBR", RPD/RSS/127, 1997.

4.5 Bulk Shielding and Complementary Shield Design in PFBR Reactor Assembly(R. Indira and A.K. Jena)

The bulk shielding inside the reactor vessel and the complementary shielding are designedon the basis of the dose criteria specified for various zones, which in turn depend upon theradiation source, accessibility and occupancy of the area of interest. One of the main AERBcriteria which determines shield design is that the dose rate must be less than 1 uSv/h (0.1 mR/h)in the freely accessible zone.

The core shielding of PFBR mainly serves the purpose of limiting the secondary sodiumactivation to permissible level. The dose rate at a distance of 0.5 m from the centerline of thesecondary sodium pipes is limited to 1 uSv/h, as the steam generator building is a freely accessiblezone. Correspondingly, the secondary sodium activation is limited to 3.2 Bq/cm3. This is achievedby providing radial shields around the core and integrated shields in the upper axial region of thefuel subassemblies. Details of the conceptual design of core shielding are given in Ref. 1.

The roof slab shielding and the reactor vault shielding are designed to shield against theprimary sodium activation gammas which is 16 mCi/cm3 in PFBR. As the reactor control buildingis not a freely accessible zone, the dose rate in the cells around the reactor vault (concrete) andthe dose rate on the top platform above the roof slab are specified as 10 jiSv/h [2]. The regionbetween the top platform and the roof slab has only controlled access and the dose rate in thisregion is specified as 100 uSv/h.

4-5

Based on these specifications, the thickness of reactor vault concrete has been worked outto be 185 cm of ordinary concrete of density 2300 kg/m3. The roofslab is 109 cm thick concreteof density 4300 kg/m3. As there are many sources of radiation on the top of roof slab, such ascover gas activity due to cover gas leak into RCB, the shielding for various streaming paths in theroofslab are worked out with 25 uSv/h as the dose rate criterion. Details of the various streamingpaths, sodium and cover-gas lines emerging above the roof-slab and the complementary shielddesign for these are discussed in Ref. 3.

References1. P. Pudiyavinayagam, R. Indira et.al, "Conceptual Design of Core Shielding",

PFBR/31100/DN/1023 Rev.A, 1997.2. Minutes of the meeting of RPD, NSD and SR&HPG, December 1997.3. S. Ragupathy, R.Indira, A.K. Jena et.al, "Conceptual Design of Shielding for

Reactor Assembly" PFBR , 1997.

4.6 Choice of Hardfacing Material for PFBR Components Based On Induced Activities,Dose rates, Man-Rem Expenditure and Shielding(Rlndira)

In PFBR, many components are presently hardfaced with either Stellite-6 or Stellite-12 asthe hard facing material. As these are cobalt based alloys, induced activity and dose rate due toCo-60 results in difficulties in maintenance, decommissioning and necessitates use of larger sizesand weights in handling flasks. Hence possible replacement of stellite with nickel based hardfacingalloys such as Colmonoy-6 (Col-6) and Triballoy-700 (T-700) has been studied.

Induced activity computations are carried out considering the same amount of Stellite,Col-6 and T-700 for the different components to be hardfaced. The calculations are preliminary innature and do not take into account the exact geometry. The nominal compositions of Stellite-6,Stellite-12, Col-6 and T-700 and the various activation reactions, activation products, their halflives and gamma energies for the materials in these materials are given in Ref.l. Only theactivation reactions, with the activation product of half-life greater than one hour are considered.The activities have been computed using thermal equivalent flux for (n,y) reactions and fissionequivalent flux for (n,p), (n,a) and (n,2n) reactions. The activity and dose-rate computations havebeen carried out for 0% Co-60 and 0.25% Co-60 in Col-6 and T-700. In PFBR specifications foraustenitic stainless steels, a maximum of 0.25 % cobalt has been assumed for T-700 and Col-6.

For the grid plate components, absorber rod drive mechanisms (CSRDM & DSRDM),failed fuel identification module (FFIM) and primary sodium pump (PSP) saturation activity iscalculated, as these components are expected to be irradiated for 20 years. A cooling time of 5 yis considered in the case of grid plate components, as these are expected to be handled only fordecommissioning purposes. In the case of CSRDM, DSRDM, FFIM & PSP a cooling time of 2 dhas been considered. For absorber rods (CSR & DSR), an irradiation time of 2 y and cooling timeof 2 d is considered. The dose-rate from the hardfacing materials and also from the structuralmaterial (austenitic stainless steel) for various NSSS components are compared in Table 4.6.1.From induced activity and dose rate considerations, stellite can be replaced by either Col-6 or

4 - 6

T-700. The actual choice should be based on the temperature and other engineeringconsiderations.

Table 4.6.1

Components

Grid Plate

Antirotation lugsGP(PCD: 13200)IHX seal flangesin Inner Vessel(PCD 13200)CSRDM (Guidebush & SleeveDSRDM (Bush)FFIM(Components)PSP Bearings(PCD :9760)DSR (Bush)CSR Bottom

CoolantPassageTubeGuide-cum-LabyrinthGuide

: Dose Rates from

Elevation(mm)185001800017500

23500

2566024650

2320026450

18500

2212518500

19250

21900

SS

3.5 x2.5 x4.4

2.9 x

9.0 x3.5

18 x4.7 x

9.0 x

4.4 x5.0 x

1.7x

l . l x

Austenitic Stainless Steel (with 0.25 % 4Hardfacing Materials

106

105

lO"1

lO"2

IO+1

IO-1

10'2

101

104

105

104

Dose-rateStellites

7 5x2.5 x6

3

38

2

Ox

7x

l xl x

l x5.0 x

1

42

6

2

5

6xl x

3x

7 x

10y

10s

103

102

102

101

103

101

102

106

10*

104

( R/h) fromCol-6

0% Co3.3 x4.32.1 x

6.3 x

5.4 x6.5 x

2.0 x3.8 x

1.4 x

8.19.0 x

3.4 x

l . l x

103

io-4

io-5

io-5

lO"1

lO'1

lO"3

lO"4

102

105

101

0.25% Co2.8 x9.2 x2.2 x

1.4

1.13.0 x

8.82.0 x

5.7 x

9.89.0 x

2.7 x

1.0 x

103

105

101

101

10"1

io-3

103

io-6

102

Co) and from

T-7000 % Co221

4

38

24

1

51

3

1

2 x94 x

3 x

6 x5x

4 x8x

6 x

5Ox

5x

3x

103

10"4

io-*

io-5

10"3

lO"1

io-3

lO"4

103

105

101

0.25% Co3.0xl.Ox2.3 x

1.6

1.44.0 x

l.Ox2.0 x

6.7 x

9.59.2 x

2.7 x

1.2 x

107

106

101

10"1

101

10"1

io-3

103

10"

102

Shield computations have been carried out assuming 1.3 MeV gammas from Co-60. Thetenth layer thickness of lead is around 3.5 cm for 1.3 MeV gammas. This means, to reduce thedose by a factor of 10, lead shield of 3.5 cm is required. The basic stainless steel component itselfbecomes radioactive and requires shielding. An approximate estimate of the shield requirementassuming a permissible dose-rate of 0.1 mSv/h has been given in Table 4.6.2.

Reference1. A. K. Bhaduri, R. Indira et al, "Selection of Hardfacing Material for NSSS Components of

PFBR" IGC/HTF/97. 02, 1997.

4-7

Table 4.6.2 : Shielding of PFBR Components Hardfaced with Different MaterialsPFBR Components

Grid Plate

Anti-rotation lugs onGP (PCD: 13200)IHX seal flanges in IV(PCD : 13200)CSRDM(Guide bush& Sleeve)DSRDM(Bush)FTTM (Components)PSP Bearings (PCD:9760)DSR (Bush)CSR Bottom

CoolantPassageTubeGuide-cum-LabyrinthGuide

Lead Shield Thickness (cm) when Hardfaced withElevation(mm)185001800017500

235002566024650

232002645018500

2212518500

19250

21900

SS

25217

2-7

89-

918

32

10

Stellites

383215

121210

1495

1325

38

18

Col-60% Co25217

2-7

89-

918

32

9

0.25%Co282510

447

89-

918

32

10

T-7000% Co25217

2-7

89-

918

32

9

0.25%Co282510

447

89-

918

32

1.2 xlO2

4.7 Dose to Personnel in RCB Following Discharge Pipe Rupture of PFBR(R.S. Keshavamurthy)

In the two loop design of PFBR, discharge pipe rupture can cause coolant flow to reduceto less than 25% in a short time resulting in the failure of a number of fuel pins [1]. This results ina puff release of fission gases into cover gas. All fission gases except the very short lived ones areexpected to come to cover gas and leak into RCB through the leakage paths at the top of reactorvessel in this incident. This results in a sudden increase in the activity level of RCB. The quantityof dose received by the personnel exposed to this increase in activity in RCB is estimated in thisstudy. The time development of activity is properly taken into account. It is also assumed that thereactor is shutdown immediately after the incident.

All the 11 nuclides with half lives greater than 3 minutes are considered in the calculation.The number of fuel pins that may fail is treated as a parameter due to lack of definite data. Dosesare calculated for failure of 0.1%, 1%, 5% and 10% of the fuel pins. At the time of the incident,the reactor may already be operating with a few failed fuel pins. Therefore, cover gas flush ratescould be anywhere upto 30 m3/h. Estimates are made for sweep rates of 10 m3/h and 30 m3/h. Theresults for the dose are expressed in Figure 4.7.1 as a function of time of stay in RCB for various

4 - 8

pin failure conditions. Table 4.7.1 gives the time during which the dose goes upto 0.1 Sv(l 0 Rem)under different sweep rate (S) and pin failure conditions. It is calculated that the time integrateddose received in RCB does not exceed 0.1 Sv if the number of pin failures is of the order 0.1%.Failure of 1% pins will result in a dose of 0.1 Sv in 4.1 h at a cover gas sweep rate of 10 m3/h.For pin failures 5% or higher, the dose can exceed 0.1 Sv in less than an hour irrespective of thecover gas flush rate. The details of the study are given in Ref [2].

Table 4.7.1: Time Taken for a Dose of 0.1 Sv to Personnel in RCBNumber of fuel pin failures

1%5%10%

Time taken to impart 0.1 SvS = 10 m3/h S = 30 m3/h

4.1 h0.61 h0.36 h

< 0.1 Sv for all times0.65 h0.37 h

References1. N. Kasinathan, "Plant Dynamic Studies for Conceptual Design",

PFBR/66040/DN/1100, 1997.2. R.S. Keshavamurthy, "Dose to Personnel in RCB Following Discharge Pipe

Rupture Incident in the Two Loop Conceptual design of PFBR due to Fuel PinFailures", RPD/RSS/130, 1997.

4.8 Release of Radioactivity into Environment under Hypothetical Core DisruptiveAccident Condition from PFBR(R.S. Keshavamurthy)

In the case of extremely unlikely core disruptive accident, melting and vaporization of coreinventory may cause transport of radioactive material to the cover gas and the sodium slug mayimpact the reactor vessel head, causing damage and leaks and thus providing a pathway for theescape of radioactive material into the Reactor Containment Building (RCB). A part of this mayfurther leak out to the environment. The objective of this study is to estimate source term ofradioactivity into environment under such a severe accident condition.

The fractions for radioactive materials expelled from core into RCB are taken from theOECD Expert Committee [1]. The evolution of concentration aerosols of a particular size in RCBhas to be obtained from a non-linear equation with many experimentally determined parametersput in [2]. Since the code which solves this equation is not available, recourse is taken tosimplifying the equation by assuming an average radius for the aerosol but taking into account thegravitational sedimentation along with wall plating and thermophoresis. With this approach, theradioactivity carried by the aerosols into environment over a period T has been obtained using,

A =1 100R

where A represents the radioactivity of initial number of aerosols, L the leak rate, and R theremoval rate which is a sum of removal rates due to gravitational sedimentation, wall plating andthermophoresis.DL represents the reduction factor due to aerosol deposition in leakage paths.

4-9

Table 4.8.1: Removal Rates (R) for Radiologically Important AerosolsAerosol Type

Na2ORbTeSr

PuO2

CeBaRu

pm (kg/m3)2270153062402600116006800360012400

R (s"1)4.80E-043.24E-041.32E-035.50E-042.45E-051.44E-037.61E-042.62E-03

Table 4.8.2 : Release of Radioactivity in Environment (RGB leak rate: 0.1% v/h)Isotope

1-131Cs-137Rb-88Ru-103Sr-90

Ce-141Te-131mBa-140Kr-85mKr-87Kr-88Kr-85

Xe-133Xe-135Pu-239

Core inventory(Bq)

1.25E187.10E164.44E172.04E182.47E161.79E181.35E171.64E171.92E173.49E174.22E173.96E152.15E182.24E18

4.51E15(1948 kg)

Release intoenvironment (Bq)

6.49E131.84E143.O8E138.51E111.41E113.77E124.44E116.66E114.14E157.51E159.14E158.55E134.66E164.85E164.51E10(19.5 g)

Fraction of coreinventory5.21E-52.59E-36.94E-54.22E-75.57E-62.12E-63.36E-64.00E-62.2E-22.2E-22.2E-22.2E-22.2E-22.2E-29.96E-6

Based on some experiments, this factor is taken as 0.1 in the present estimation. Table4.8.1 gives the removal rates calculated for various aerosol type assuming that the average radiusaerosol is 10 microns. The results for the release of radioactivity into environment is given inTable 4.8.2. The estimate for release fraction are comparable to that estimated for EFR [3] underidentical leak rate condition for the most important isotope 1-131.

References1. OECD Group of Experts Report, "State of the Art Report on Nuclear Aerosols in

Reactor Safety", NEA, OECD, 1979.2. R.S. Keshavamurthy, "Release of Radioactivity into Environment under Hypothetical

Core Disruptive Accident from PFBR", PFBR/01115/DN/1018, 1997.3. Carluec et al, "Radiological Release Analysis for the EFR project", 1995 TCM on

Evaluation of Radioactive Materials Release and Sodium Fires in FBR, O-arai, Japan,IWGFR/92, 1996.

4-10

4.9 A Study on Fission Product Activity in Typical Fast Reactors(M.S. Sridharan, A.K. Jena and S.M. Lee)

The half-lives of fission products (FP) produced in a nuclear reactor range from a fractionof a second to millions of years. The knowledge of the radioactivity from FP is necessary toensure adequate radiation protection during fuel handling/transportation/processing/storage/disposal of spent fuel wastes. It is also necessary from the point of environmental concerns aroundnuclear reactors and waste disposal-sites. A study is made on the FP activities and the results arepresented for PFBR and FBTR Mark I and Mark TI cores. For determining the activity estimates,the code system CHANDY [1] has been used. The associated data library contains data on 569FP in 112 mass chains, based on the compilations of Tobias [2] and Crouch [3]. This data libraryincludes the decay schemes of individual FP, half-lives, branching ratios, details of parent nuclides,FP yield data of Crouch [3] for fast fission, daughter nuclides and the emitted radiation.

Typical fuel composition of the fast reactor fuels that have been considered in theseestimations are given in Table 4.9.1. The fuels are mixed oxides or carbides of Uranium andPlutonium. Estimates of FP activities in the central region fuels after a typical irradiation of 1 yearare presented in Table 4.9.2 as a function of cooling times. The activities decrease with coolingtime as expected. The study indicates that as the percentage of plutonium content decreases, theactivity increases for short cooling times; decreases for long cooling times and remains roughlythe same for intermediate times. The changes are, however, small. The variation arises as a resultof the relative fast fission yields in U and Pu contents of the fuel. Estimates as a function ofburnup and cooling times are given in Table 4.9.3. The estimates are influenced by variation in the

Table-4.9.1 : Typical Fuel Composition of Fast ReactorsReactor

PFBR(oxide)FBTR Mark 1(carbide)FBTR Mark n(carbide)

Fuel Mass(kg/SA)50.63

3.27

3.12

NormalisedFuel Mass*6.33

8.18

5.21

U235/U238

0.25/99.75

0.7/99.3

0.7/99.3

UO2/PUO2

79.3/20.7

30/70

45/55

Pu238/Pu239/Pu240/Pu241/Pu 242-/68.79/24.6/5.26/1.350.017/93.36/6.21/0.394/0.023-/68.79/24.6/5.26/1.35

Table 4.9.2 : Fission Product Activity in Central Region Fuel after 1 y Irradiation

Cooling Time0s1 minlhr.I d10 d100 d1000 d

Activity* (Ci)PFBR4.49E63.32E61.74E69.42E54.71E51.42E51.45E4

FBTR-Mark I4.37E63.29E61.73E69.39E54.67E51.41E51.46E4

FBTR Mark II4.40E63.30E61.73E69.40E54.68E51.41E51.46E5

*see comments at the bottom of Table 4.9.3

4 - 1 1

irradiation days among the reactors for a given burnup, besides the cooling time and the relativefast-fission yields of uranium and plutonium contents of the fuel. The above estimates have anuncertainty margin of + 10% and have been normalised to a fuel mass that gives a total of3.12E16 fission/s which is approximately equal to 1 MW fission power in the central region fuels.The details of the study are given in Ref. [4],

Table 4.9.3 : Fission Product Activity in Central Region Fuels as a(Burn = Actual Burnup in GWd/t.)Reactor

PFBR

FBTRMark I

Mark II

CoolingTime0 s1 minlh .I d10 d100 d1000 d0 s

1 minlh .I d10 d100 d1000 d0s1 minlh .I d10 d100 d1000 d

Function of Burnup.

Activity *(Ci)Burn = 13.91E62.74E61.16E63.98E57.44E46.73E32.84E23.83E6

2.76E61.19E64.25E58.44E47.85E33.39E23.78E62.68E61.12E63.63E56.17E45.34E32.27E2

104.28E63.12E61.53E67.41E52.88E55.18E42.74E34.20E6

3.12E61.55E67.62E53.06E55.80E43.24E34.16E63.O6E61.49E67.O8E52.60E54.31E42.20E3

254.40E63.23E61.65E68.56E53.90E59.32E46.43E34.31E6

3.23E61.66E68.73E54.05E51.01E57.61E34.28E63.19E61.62E68.28E53.64E58.13E45.24E3

504.47E63.30E61.72E69.22E54.52E51.28E51.17E44.37E6

3.29E61.72E69.33E54.61E51.37E51.36E44.36E63.26E61.69E69.01E54.31E51.16E59.71E3

754.50E63.33E61.75E69.51E54.80E51.49E51.61E4

1004.51E63.35E61.76E69.70E54.98E51.64E51.98E4

*Normalised to a fuel mass which has an operating fission power of 1 MWt (i.e., 6.33 kg. inPFBR, 8.18 kg. in FBTR-Mark I and 5.21 kg. in FBTR-Mark II respectively).

References1. K P N . Murthy and R. Shankar Singh, "Fission-product Decay Power after Fast Neutron

Fissions in U-235, U-238 and Pu-239 - Calculations and Comparison with Experiments,"IGC AR Report RRC-15, 1976.

2. A. Tobias, "Data for the Calculation of Gamma Radiation Spectra and Beta Heatingfrom Fission Products" Rev.3, Report RD/B/M2669, CEGB, Berkeley Nuclear Lab, 1973.

3. E.A.C. Crouch "Atomic & Nuclear Data Tables", 19, pp 417-532, 1977.4. M.S.Sridharan, A.K. Jena and S.M. Lee, "A Study on Fission Product Activity in Typical

Fast Reactors", Proc. Nat. Symp. onRad. Phy. (NSRP-12), Jodhpur, Jan. 28-30, 1998.

4-12

4.9 A Study on Fission Product Activity in Typical Fast Reactors(M.S. Sridharan, A.K. Jena and S.M. Lee)

The half-lives of fission products (FP) produced in a nuclear reactor range from a fractionof a second to millions of years. The knowledge of the radioactivity from FP is necessary toensure adequate radiation protection during fuel handling/transportation/processing/storage/disposal of spent fuel wastes. It is also necessary from the point of environmental concerns aroundnuclear reactors and waste disposal-sites. A study is made on the FP activities and the results arepresented for PFBR and FBTR Mark I and Mark II cores. For determining the activity estimates,the code system CHANDY [1] has been used. The associated data library contains data on 569FP in 112 mass chains, based on the compilations of Tobias [2] and Crouch [3]. This data libraryincludes the decay schemes of individual FP, half-lives, branching ratios, details of parent nuclides,FP yield data of Crouch [3] for fast fission, daughter nuclides and the emitted radiation.

Typical fuel composition of the fast reactor fuels that have been considered in theseestimations are given in Table 4.9.1. The fuels are mixed oxides or carbides of Uranium andPlutonium. Estimates of FP activities in the central region fuels after a typical irradiation of 1 yearare presented in Table 4.9.2 as a function of cooling times. The activities decrease with coolingtime as expected. The study indicates that as the percentage of plutonium content decreases, theactivity increases for short cooling times; decreases for long cooling times and remains roughlythe same for intermediate times. The changes are, however, small. The variation arises as a resultof the relative fast fission yields in U and Pu contents of the fuel. Estimates as a function ofburnup and cooling times are given in Table 4.9.3. The estimates are influenced by variation in the

Table-4.9.1: Typical Fuel Composition of Fast ReactorsReactor

PFBR(oxide)FBTR Mark 1(carbide)FBTR Mark H(carbide)

Fuel Mass(kg/SA)50.63

3.27

3.12

NormalisedFuel Mass*6.33

8.18

5.21

U235/U238

0.25/99.75

0.7/99.3

0.7/99.3

UO2/PUO2

79.3/20.7

30/70

45/55

Pu238/Pu239/Pu240/Pu241/Pu 242768.79/24.6/5.26/1.350.017/93.36/6.21/0.394/0.023-/68.79/24.6/5.26/1.35

Table 4.9.2 : Fission Product Activity in Central Region Fuel after 1 y Irradiation

Cooling Time0s1 minlhr.I d10 d100 d1000 d

Activity* (Ci)PFBR4.49E63.32E61.74E69.42E54.71E51.42E51.45E4

FBTR-Mark I4.37E63.29E61.73E69.39E54.67E51.41E51.46E4

FBTR Mark II4.40E63.30E61.73E69.40E54.68E51.41E51.46E5

*see comments at the bottom of Table 4.9.3

4-11

irradiation days among the reactors for a given burnup, besides the cooling time and the relativefast-fission yields of uranium and plutonium contents of the fuel. The above estimates have anuncertainty margin of + 10% and have been normalised to a fuel mass that gives a total of3.12E16 fission/s which is approximately equal to 1 MW fission power in the central region fuels.The details of the study are given in Ref. [4].

Table 4.9.3 : Fission Product Activity in Central Region Fuels as a(Burn = Actual Burnup in GWd/t.)Reactor

PFBR

FBTRMark I

Mark II

CoolingTimeOs1 minlh .I d10 d100 d1000 d0 s

1 minlh .I d10 d100 d1000 d0s1 minlh .I d10 d100 d1000 d

Function of Burnup.

Activity *(Ci)Burn= 13.91E62.74E61.16E63.98E57.44E46.73E32.84E23.83E6

2.76E61.19E64.25E58.44E47.85E33.39E23.78E62.68E61.12E63.63E56.17E45.34E32.27E2

104.28E63.12E61.53E67.41E52.88E55.18E42.74E34.20E6

3.12E61.55E67.62E53.06E55.80E43.24E34.16E63.06E61.49E67.08E52.60E54.31E42.20E3

254.40E63.23E61.65E68.56E53.90E59.32E46.43E34.31E6

3.23E61.66E68.73E54.05E51.01E57.61E34.28E63.19E61.62E68.28E53.64E58.13E45.24E3

504.47E63.30E61.72E69.22E54.52E51.28E51.17E44.37E6

3.29E61.72E69.33E54.61E51.37E51.36E44.36E63.26E61.69E69.01E54.31E51.16E59.71E3

754.50E63.33E61.75E69.51E54.80E51.49E51.61E4

1004.51E63.35E61.76E69.70E54.98E51.64E51.98E4

*Normalised to a fuel mass which has an operating fission power of 1 MWt (i.e., 6.33 kg. inPFBR, 8.18 kg. in FBTR-Mark I and 5.21 kg. in FBTR-Mark II respectively).

References1. K.P.N. Murthy and R. Shankar Singh, "Fission-product Decay Power after Fast Neutron

Fissions in U-235, U-238 and Pu-239 - Calculations and Comparison with Experiments,"IGCAR Report RRC-15, 1976.

2. A. Tobias, "Data for the Calculation of Gamma Radiation Spectra and Beta Heatingfrom Fission Products" Rev.3, Report RD/B/M2669, CEGB, Berkeley Nuclear Lab, 1973.

3. E.A.C. Crouch "Atomic & Nuclear Data Tables", 19, pp 417-532, 1977.4. M.S.Sridharan, A.K. Jena and S.M. Lee, "A Study on Fission Product Activity in Typical

Fast Reactors", Proc. Nat. Symp. onRad. Phy. (NSRP-12), Jodhpur, Jan. 28-30, 1998.

4 - 1 2

Table 4.11.1: Gamma Spectrum in Converter Assembly(Cooling time=180 d)

Energy (MeV) y/s0.01-0.02 .121E+070.02-0.03 .772E+060.03-0.06 .341E+090.06-0.10 .175E+080.10-0.20 .301E+090.20-0.40 .127E+070.40-0.60 .201E+090.60-0.70 .862E+080.70-0.80 .304E+100.80-1.00 .116E+071.00-1.50 .677E+071.50-2.00 .239E+072.00-3.00 .907E+073.00-4.00 .264E+044.00-5.00 .376E+025.00-6.00 .200E+016.00-7.50 459E+00

Total .401E+10

Table 4.11.2 ; Activities of Major Fission Product IsotopesIsotopeSr 89Sr90Y 90Y 91Zr 95Nb95Rul03

Rhl03mRu 106RhlO6Csl37

Bal37mCel41Cel44Prl44Pml47

Concentration (Atoms)0.37906E+160.58048E+170.14888E+140.71009E+160.95699E+160.94450E+160.13847E+160.13914E+130.35548E+160.33858E+100.68480E+170.10403E+110.12583E+160.38009E+170.16038E+130.22009E+17Total Curies

Activity (Ci)0.16269E-010.12090E-020.12093E-020.26313E-010.32430E-010.58258E-010.76296E-020.76213E-020.20898E-020.20898E-020.13492E-020.12696E-020.83915E-020.28976E-010.28977E-010.49864E-020.23116E+00

4-15

4.12 Measurement of Neutron Flux at the Shielding Corner of APSARA(A.K. Jena, R. Indira, D. Sunil Kumar, H.R. Dravid *, S.L. Mehta* and S.Sankaranarayanan* (*RSMD, BARC))

It is proposed to carry out PFBR shielding mockup experiments at APSARA. As a firstphase of the experiment, the flux at shielding corner has been measured after installation of ahollow aluminum box in a position between the core and shield corner. The box is used to reducethe absorption of neutrons by water. Mn-Cu,Au and In foils were irradiated at shielding corner for1 h with reactor operating at power of 400 W and removed after 5 h of cooling. The counting wasdone using Ge detector setup. The neutron flux measured along the core centre line are given inTable 4.12.1 and are compared with the calculated values in Table 4.12.2. The calculations havebeen done by ID transport code and WIMS neutron data set [2]. It may be observed from thetables that the calculated value is well comparable with measured value in fast flux region butcalculations overpredict the flux by a factor of two in the thermal region. The measured flux isless by a factor of 15 than the desired thermal flux of 1010n/cm2/s at 40 kW reactor power. Effortsare being made to change the aluminium box to get the desired flux at shielding corner.

Table 4.12.1: Measured Neutron flux (n/cm2/s

a.

b.

c.

Measurement

The manganese equivalent flux using Mn55(n,y)Mn56 reactionThermal neutron flux using gold foils(Cadmium ratio method) using Aul97(n, y)Aul98 reactionFission equivalent neutron flux (>1.0 MeV)using Inll5(n, n')Inll5m reaction

At aluminum panel

(1.00 +0.05) E+6

(1.30 +0.05)E+6

(7.63 ± 0.80)E+6

At 23.7 cm away fromaluminum panel(0.8O±0.05)E+6

(1.14±0.05)E+6

(5.41±0.5)E+6

1.

2.

Table 4.12.2; Comparison of Calculated and measured Flux

Flux .>1 MeVMn eq. flux

Calculated7.06 E62.32 E6

Measured7.63 E 61.00 E6

References:H.K. Dravid, A.K. Jena and S. Sankaranarayanan, "Preliminary Result of theMeasurement of Neutron Flux at Shielding Corner', RSMD/RPNES/A-11/6043/97, 1997A.K. Jena et al., "PFBR Bulk Shielding Experiments at APSARA - Measurement ofNeutron Flux Level at Shielding Corner", RPD/RSS/125, Aug. 1997.

4-16

CHAPTER 5

SAFETY ANALYSIS

5.1 PSA On Flow Blockage Of FBTR Fuel Subassembly(A. John Aral and Om Pal Singh)

The observed temperature rise across FBTR Mark II Fuel Subassembly (FSA) ismore than that expected by 30 to 40 % [1]. It has been confirmed to be due to ameasurement problem due to the thermocouples and the core cover plate having stuck at araised position. Subsequent analysis [2], has shown that flow blockages in Mark II FSAcan be detected, inspite of a higher value of minimum detectable flow blockage level. Inview of this situation, SARCOP recommended Probabilistic Safety Assessment (PSA) forflow blockage. Thus a study has been carried out to calculate the flow blockage frequencyfor FBTR. In addition, Shutdown System (SDS) failure probability of FBTR is alsocalculated to find the probability of unprotected flow blockage incident.

In order to calculate the flow blockage probability, three alternatives are available.That is, PSA using fault tree set up with events leading to flow blockage, accidentprecursor based analysis and parameter (blockage rate) estimation based on observedevents. The third alternative is used to find the flow blockage probability. Four flowblockage incidents have been reported in fast reactors. The first is in Enrico Fermi reactor,due to an external piece of metal in the flow path. This incident led to the emphasis on thedesign aspects like radial entry of flow through multiple hole in FSA, provision ofreactivity meter, surveillance of FSA outlet temperature and Delayed Neutron Detection(DND). The second incident which can be considered a precursor to flow blockageoccurred in KNK-II reactor due to partial accumulation of structural material corrosionproducts in flow path. It was detected by temperature rise in FSA. This incidentemphasizes the proper on-line monitoring of purity of sodium and change to stainless steelmaterial from ferritic steel which was used in KNK-II. The third incident was in SPXreactor when a rubber packing vessel during air test got left at the inlet of a subassemblyand was detected by FSA outlet temperature rise during power raising. The fourth incidentwhich can be considered again as a precursor to flow blockage is reported in PFR, due tooil leakage into the coolant from the primary pump and the subsequent deposition ofcarbon particles on the fuel pins and sub-assembly inlet filters resulting in detectablereduction of the coolant flow rate.

Based on these incidents reported in fast reactors, the frequency of flow blockageis calculated as 1.5 x 10'2/ry. By using the Duane's method [3] of test-fix-test-fix process(improvement in reliability due to improved design measures), it is found that the flowblockage frequency in fast reactors is 5.4 x 10"3/ry. If out of 4 events, the KNK-II andPFR event are taken as precursors with 10 % weight instead of 100 %, then the resultingflow blockage frequency is 3.7 x 10"3/ry. This can be considered conservative because forFBTR the reported type incidents are unlikely to occur owing to the following designmeasures.

5-1

• Coolant radial entry in FSA is provided so that the probability of blockage due toexternal piece like in Fermi reactor is reduced.

• Stainless is used as a structural material for the primary circuit unlike ferritic steel inKNK-H. Further, the sodium purity is monitored continuously and so the KNK-II typeand PFR type incidents are unlikely to occur.

• Elastomer is not used in FBTR.• Reactivity meter is provided to detect the reactivity changes that may result due to

boiling of coolant as a result of flow blockage.• Thermocouples are provided at the outlet of each FSA to monitor the temperature

changes though the readings are degraded.• There is automatic trip on DND signal to detect fuel pin failure as a result of flow

blockage

Based on the failure rates observed in FBTR, the probability to fail to detect flowblockage through thermocouple signals is 1.5xlO"3/y and through DND is 8.7xl(T7y.Considering failures of scram logic, CRDM and absorber rods, the probability thatshutdown system will fail on demand is 6 x 10"5. Thus the probability that flow blockagewill take place and shutdown system will not respond on demand is 6 x 10'5 x 5.4 x 10"3 =3.2 x 10"7/ry. The details of this study are given in Ref.4.

References1. S. Varadarajan and G. Srinivasan, "Analysis of Core Temperature Data for 27 SA

Core upto 9 MWt", ROD/FBTR/S-RS-02/31400/SAR-026 April 97.2. S. Clement Ravichander et. al, "Fuel Subassembly Cooling Anomaly Detection

Capability in FBTR Small Carbide Core", FBTR/NSD/31411/DN/June 97.3. E. E. Lewis, "Introduction to Reliability Engineering," John Wiley and Sons, New

York, 1987.4. A. John Aral and Om Pal Singh, " PSA on Flow Blockage of FBTR Fuel

Subassembly", RPD/SAS-93, Dec. 1997

5.2 Conceptual Design Of Reactor Shutdown System For PFBR(Om Pal Singh, S. Govindarajan*, G. Vaidyanathan* and G. Muralikrishna*)*Design and Technology Group

The basic purpose of reactor shutdown system (SDS) is to quench the fission chainreaction and bring and maintain the reactor in a desired subcritical state when required.The reactor SDS consists of sensors to monitor plant operating parameters and signalprocessing, logic system to trigger the reactor shutdown, drive mechanisms to insert theabsorber rods in the reactor core and absorber rods to shutdown the reactor. The designrequirements of SDS are:

• It should be highly reliable (probability of failure <10"6/ry).• The delay in shutdown due to detector response, signal processing and release of

absorber rods by electromagnet should be small, the speed of shutdown should befast and reactivity worths of absorber rods should be appropriate so that all the

5 - 2

design basis events are terminated fast enough without any safety parametercrossing their design limit.

• The reactivity worths of absorber rods should be such that when all the rods are inthe core, the sufficient shutdown margin (SDM) is available for protection againsterrors during refuelling. The withdrawal speed of absorber rods should not belarge.

• The SCRAM signals should be connected to absorber rods optimally from thepoint of view of safety and economics.

• The system should be capable to shutdown the reactor even in seismic conditions.

A detailed study has been carried out on the design of SDS keeping the aboverequirements in view. The following are the conclusions of the study

1. Two independent, redundant and diverse systems, right from sensors to theabsorber rods meet the reliability requirements of the SDS. The first set ofabsorber rods is termed as CSR and is for reactor control and shutdown. Thesecond set is termed as DSR which is only for reactor shutdown.

2. Based on the reduction in the probability of spurious scram and failure on demand,a 2/3 voting for neutronic and flow sensors, and a 2/2 voting for thermocouplesensors is adequate.

3. Safety parameters have been identified based on the analysis of transient overpower and transient under cooling incidents such that each event is protected bytwo diverse parameters except the total instantaneous blockage and gasentertainment.

4. The design should be such that if SCRAM is not effective, LOR is initiated and ifLOR is not effective SCRAM is triggered. Manual SCRAM should be possible forboth the system.

5. Absorber rods worth needed is 11,000 pern and SDM is 5,000 pcm. It has beenfound most appropriate to have 9 CSR and 3 DSR with their respective worths as8000 pcm and 3000 pcm respectively. Each set of rods can bring the reactor tocold critical state under 1 rod stuck criteria.

6. To improve reliability, CSR and DSR be clubbed in two groups for SCRAMpurpose; group 1 containing 4 CSR + 2 DSR or 5 CSR + 1 DSR and group 2containing 4 CSR + 2 DSR. Neutronic including DND parameters shall triggerreactor SCRAM by both the groups of rods. Flow and power/flow parameters willtrigger reactor SCRAM by group 1 of rods and pump drive current, pump speeddifference and temperature parameters shall trigger SCRAM by group 2 of rods.

5 - 3

To improve the reliability of SDS further, if optical link is provided between thetwo systems, then grouping of rods is not necessary.

7. Delay in release of absorber rods is specified as 100 ms. Absorber rod drop time isspecified < 1 s. The withdrawal speeds of CSR and DSR are 2 mm/s and 4 mm/srespectively.

8. A few passive devices have been identified for development for risk minimization.

The details of the study are given in Refs. [1] to [3].

References1. Om Pal Singh et al, "Reactor Shutdown System for PFBR", Proc. Workshop on

Reactor Shutdown System, IGCAR, March 4-6, 1997.2. S. Govindarajan et al, "Conceptual Design of Reactor Shutdown System",

PFBR/31000/TF/1014, May 1997.3. S. M. Lee, "Safety Considerations in Selection of Absorber Rod Reactivity Worth

in PFBR", Proc. Workshop on Reactor Shutdown System, IGCAR, Kalpakkam,March 4-6, 1997.

5.3 Distribution Of Scram Signals To Absorber Rods In PFBR Based OnReliability Analysis(A. John Aral and Om Pal Singh)

As any other nuclear reactor, PFBR is provided with suitable instrumentation formonitoring its operating characteristics so that if any anomaly develops, the reactor isshutdown. The SDS should be designed such that the probability of failure < 10"6 /ry [1].The most important plant monitoring signals are neutronic, temperature, flow and DNDsignals. A choice is made to select the safety parameters by analyzing the various events,so that reactor is protected against all the Design Basis Events (DBE) by at least twodiverse signals. The SDS is provided with two types of absorber rods, i.e., Control andSafety Rods (CSR) and Diverse Safety Rods (DSR). The SCRAM signals correspondingto different safety parameters are to be connected to CSR and DSR in such a way that itmeets the required reliability targets while simultaneously satisfying the design constraints.For instance, one obvious way is to duplicate all the safety channels and connect themwith CSR and DSR. However, duplication of all the channels may not be necessary toachieve the required reliability and it would also add to the cost of the plant. Therefore, astudy has been carried out, based on reliability analysis, to evolve a scheme of distributingthe SCRAM signals to CSR and DSR.

Several options of distributing the SCRAM signals to CSR and DSR have beenstudied to find the scheme that gives highest reliability.

Option 1: All the safety signals connected to all the rods (only CSR). This optionalthough employs redundancy, it lacks features such as diversity (of rods) and

5-4

independence which help to mitigate the Common Mode Failure (CMF). Hence due toCMF, the failure probability/d is 10"5.

Option 2: This option has two sets of diverse absorber rods (CSR & DSR) and all thesafety signals are connected to both sets of rods. In this case, redundancy and diversity ofrods and signals is incorporated but there is no independence of signals. The probability offailure of SDS decreases to 2.6 x 10"* per demand.

Option 3: In this option, out of the two sets of selected safety signals, one is connected toCSR and the other to DSR to form two independent subsystems. However there is nodiversity of rods and signals in each subsystem though redundancy in rods and signals isthere. The probability of failure/d of SDS is 2.8 x 10"6 per demand. This option appears tobe slightly worse than the option 2. This is due to the fact that signals are isolated andconnected to each type of rods. Any gain in reliability due to the independence of twosubsystems is not accounted for, as it requires full CMF analysis of detailed design.

Option 4: In this option neutronic and DND signals are duplicated to improve thereliability of sensors and electronics. The probability of failure of SDS decreases to 6.4 x10"7 per demand. There is no grouping of rods within a sub-system and hence there is nodiversity of shutdown rods and drives in each subsystem.

Option 5: In this case the rods are grouped in each sub-system in addition to the featuresof option 4 (see Fig. 5.3.1). Two schemes of grouping are considered. That is, 4 CSR + 2DSR in groupl and 4 CSR + 2 DSR in group2 or 4 CSR + 2 DSR in groupl and 5 CSR +1 DSR group2. The failure probability/d of SDS with groupl is 6xl0~8. For group2, thisvalue increases by approximately a factor two and is 10"7 (see values given within bracketsofFig.5.3.1).

Option five is finally considered most appropriate because only in this option themandatory requirement of failure probability of SDS is met and failure probability of SDSdecreases by an order of magnitude as compared to option 4.

It is suggested that neutronic signals, flow signals and pump drive current can beconnected to group 1 of rods. Neutronic signals, temperature signals and pump speedsignals can be connected to group 2 of rods. The option of connecting the two systems byoptical link also has been studied. It has been found [2] that optical link at best can lead toa decrease in the combined system failure probability by a factor 2.

References:1. Status of Liquid Metal Cooled Fast Breeder Reactors, Technical Report Series

No.246, pp 431, IAEA, Vienna, 1985.2. A. John Arul and Om Pal Singh, "Distribution of SCRAM Signals to Absorber

Rods in PFBR Based on Reliability Analysis", RPD/SAS-90, Aug. 1997.

5-5

Neutronic Flow

CSR DSR

a) 4 2b) 4 2

Neutronic Temp. DND

in rffl ri)

.Voting] [Voting

111I I

CSR DSR4 2

5 1

6x10-8

6x1 (r8 (io7)

SDS2 13X1Q-6 (10'5)

.2 5,1)

FOR CSR & DSR FORS&E

GROUP 1DUPLICATED

CHANNELSIGNAL SET 1 ( N E U T n o N I C or DND)

Fig.5.3.1Option 5:Selected Scram Parameters to Each Group of Rods

and Drives with Neutronic and DND Channels Duplicated.

5-6

5.4 Calculation of Temperature, Porosity and Plutonium Distribution in PFBRFuel Central Pin(S. Ponpandi)

In a fuel with temperature gradient, restructuring of fuel takes place. This is soparticularly for oxide fuel where temperature gradient is rather large. The migration ofpores in the direction of temperature gradient in the high temperature region is caused bythe transfer of matter through solid state diffusion and/or the vapor phase in the pores. Itultimately leads to the compaction of this region and the development of central hole. Alsothe fuel components like plutonium and oxygen are redistributed over the radius of thefuel element. This in turn affects the temperature distribution in the fuel. An attempt hasbeen made to calculate the temperature and porosity distribution in a fuel pellet of PFBRcentral fuel pin.

The temperature field is calculated iteratively when the porosity and plutonium areredistributed. Initially uniform porosity is assumed over the pellet. The one dimensionalheat conduction equation is solved to calculate the temperature profile by Galerkinweighted residual finite element method and assuming the thermal conductivity to befunction of temperature and porosity. Once the temperatures are known, the porosity isrecalculated by using the porosity equation. These porosities are substituted in the thermalconductivity and the temperatures are recalculated. This process is repeated untilconverged values of porosity and temperature field are obtained. Finally the concentrationof plutonium redistribution is calculated for the converged values of porosity andtemperature field. The plutonium distribution is shown in Fig.5.4.1. The pore migrationwhich is towards the hot region is shown in Fig.5.4.2. The study is carried out based onthe formulations given in Refs. 1 to 3 and the details of the results are given in Ref. 4.

References:1. K. Lastmann, "The OXIFRED Model for the Redistribution of Oxygen in Non-

stoichiometric U-Pu Oxides", Nucl. Mater., 150, No.l, 10, 1987.2. W. Lackey, F. Homan and A, Olsen, "Porosity and Actinide Redistribution during

Irradiation of (U, Pu) O", Nucl. Tech. 16, 120, 1972.3. Donald R. Olander, "Fundamental Aspects of Nuclear Reactor Fuel Elements"

Technical information centre, ERDA, 1977.4. S. Ponpandi, "Calculation of Temperature, Porosity and Plutonium Distribution in

PFBR Fuel Pin" RPD/SAS-95, Dec. 1997.

5.5 Analysis Of Severe Accidents In BN-800 Like Reactor(R. Harish and Om Pal Singh)

After Chernobyl accident, there have been renewed attempts to reduce the sodiumvoid coefficient of reactivity in large sized fast reactors. In BN-800 Russian reactor, thisis achieved by replacing the upper axial blanket by sodium plenum so that the neutronleakage is increased and the net sodium void coefficient is near zero. To evaluate theimpact of this design change on the consequences of severe accidents like unprotected loss

5-7

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7/ i

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.—

\

1HtM 1

_• ', •

• r--1— •

|

1 1 1 1

1 • • —

?=-?*

JUTg Of.

V

1 1 1 1

_ _

•

Illll.

L J I

—"

—

—

- f

H

till

a *

iiil

y

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ill illl

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-

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iii 1

i t I t

li II

—

I I I I

1 S S S S I 1 S S S ?Fuel Pin Radius (r/R)

Fig. 5.4.1: Pu Concentration Distribution Fig. 5.4.2: Porosity Distribution

of flow (ULOF), transient over power (UTOP) and UTOP followed by ULOF, acomparative study was initiated jointly by IAEA and European Commission in 1995; theparticipating countries being France, Germany, India, Italy, Japan and Russia and U.K.The initial fuel pin characterization for an irradiated core, common neutronic and transientcalculations upto onset of boiling have been completed by 1996. In this year, the resultshave been obtained for post boiling and post fiiel pin failure stage for the three severeaccidents mentioned above. At IGCAR, the code PINCHTRAN which is a modifiedversion of the code PREDIS is being used to analyse the BN-800 reactor for severeaccidents. The improvements, that is, a more accurate method of calculating fiiel meltfraction, exact heat conduction model of heat transfer between fuel and clad, refinedcalculations of temperature rise in sodium plenum and B4C shield, more appropriatecalculations of pressure drop over the core, sodium plenum and the shield and a newdamage parameter based .criteria for fuel pin failure have been incorporated in the code.The important outcome of the studies on the three severe accident analyses are givenbelow:

ULOF : ULOF with flow halving time of 5.5 s is analysed. All the codes predict onset ofboiling. Thus, the boiling can not be prevented by negative reactivity feedback fromsodium plenum. Indian code predicts slow boiling propagation axially downward and sothe fuel pin does not fail even after a few tens of second after the onset of boiling.

S S 5 8 S ! S § 3 ! !Fuel Pin Radius (r/R)

5-8

However, other codes predict clad dryout within a few seconds after the onset of boiling.Russian studies show an oscillatory sodium boiling behavior if the apparent absorberexpansion feedback is accounted along with the spacer pad expansion feedback. After claddryout and subsequent fuel pin failure, clad relocation takes place in the upper part of thecore and hexcan melting occurs due to radiative heat transfer from unclad fuel. It is ageneral consensus from all the participants that the initiating phase is followed bytransition phase which is not yet analysed.

UTOP : UTOP with a reactivity addition rate of 0.05 $/s and 0.5 $/s is analysed. A 50%fuel melt fraction and damage parameter based criteria are used for fuel pin failure. It isobserved that for both the types of TOP incidents, fuel pin fails in the upper part of thecore and the subsequent fuel coolant interaction initiated fuel sweepout shutsdown thereactor. Subsequent to this, a detailed transition phase analysis is needed. Russian resultsare also similar, though, the prediction of time of fuel pin failure is early as compared toour results. German studies where failure strain exceeds 2.4 %, prediction of time offailure is later than Indian and Russian predictions. Fuel relocation initiates a rapidshutdown. Partial blockage formation in the failed subassembly results in a configurationthat cannot be cooled in place on a long time scale. The incident leads to progressivemelting and the issue of recriticality is to be studied. The Japanese and French studies alsoindicate that fuel pin integrity can not be maintained and transition phase studies arerequired.

UTOP/ULOF : UTOP followed by ULOF scenario is similar to the ULOF scenario. Thedetails of the results are given in Refs. [1] and [2].

References1. Papers presented in 6th Consultancy Meeting on "IAEA/EC Comparative

Calculations of the BN-800 Reactor with Near Zero Void Coefficient ofReactivity, Brussels, Belgium, June 28-30, 1997

2. Paper presented in 7th Consultancy Meeting on IAEA/EC ComparativeCalculations of the BN-800 Reactor with near Zero Void Coefficient ofReactivity^, Vienna, Austria, Dec. 11-12, 1997.

5.6 Numerical Solution Of Transient Heat Conduction Equation WithTemperature Dependent Thermal Conductivity(T. Sathiyasheela and R. Harish)

To understand and predict the transient reactor power and temperature of thereactor core materials against transient over power and undercooling incidents, the reactorpoint kinetics equations are solved simultaneously along with the heat transfer equationsapplicable to the coolant channels. In the inhouse developed computer code, PREDISwhich is developed for this purpose, lumped model of heat transfer between fuel and cladand coolant, is used. However, for accurate predictions, the model of heat transferinvolving exact heat conduction equation is to be used. Thus, a methodology is developedto solve the heat conduction equation for fuel pellet and clad. For proper validation of the

5 - 9

code, first the method to solve the heat conduction equation with thermal conductivitybeing independent of temperature, is developed. It is because one can obtain analyticalsolutions in this case and hence the numerical method can be tested against analyticalsolution. Then a method to solve the heat conduction equation with thermal conductivitybeing dependent on temperature is developed. For the first method, explicit finitedifference scheme is used. Transient temperature calculations are done for the case whenpower increases linearly upto 15 times and then remaining constant at that level. Thencorrespondingly the temperature rise is attained and ultimately the new steady state isreached. The correctness of the transient calculations have been checked by comparing theresults of the new steady state with the one obtained using analytical expressions. For thesecond case when thermal conductivity is a function of temperature, the implicit method(Crank Mckolson method) of finite differencing is used. In this case analytical expressionscan not be derived. So it has been checked qualitatively by comparing the results withconstant thermal conductivity data. By taking the advantage of implicit method, time stepcan be increased up to 0.005 s. The program developed is being incorporated in the code,PREDIS and PINCHTRAN. The details of this study are given in Ref. [1].

Reference1 T. Sathiyasheela and R. Harish, " Numerical Solution of Transient Heat

Conduction Equation, with Temperature Dependent Thermal Conductivity",RPD/SAS-92, Nov. 1997.

5.7 Improved Sodium Boiling Model For ULOF Studies(G. S. Srinivasan)

In any large LMFBR core, voiding of sodium coolant leads to positive voidcoefficient of reactivity. Hence, it is of utmost importance to have a precise sodium boilingmodel and estimate void fractions accurately for accurate feedback reactivity calculations.

The sodium boiling model which is existing in the code PREDIS is based oncalculations, which take into account only the energy conservation equation and LockhartMartinelli correlation for determining the void fraction ( a ) and vapor quality ( x ). Itdoes not solve the momentum conservation equation for getting pressure values over thechannel. Hence, pressure buildup due to boiling, flow reversal and slug ejection have notbeen characterized. An attempt has been made to study void propagation underunprotected loss of flow (ULOF) conditions to improve the old model, by performingmore accurate calculations for mass and energy conservation equations and by includingmomentum conservation equation. The routine thus developed has been incorporated intoPREDIS code, which is used to carry out pre-disassembly phase calculations for severeaccidents like ULOF.

In the present model, initially the mass and energy coupled conservation equationsare solved numerically to get first estimates of a . The coupled equations are implicit ina and x . The numerical scheme involves propagation along spatial meshes as well assimultaneous calculations using adjacent time step values. The Lockhart-Martinelli

5-10

correlation between a and x is used to get unknown x values in the adjacent time stepsbased on calculated values of a. The difference equation is solved iteratively to getaccurate values of a at a particular time step. The convergence of a is determined byreaching a stage wherein the successive values of x fall within a specific convergence limit.Having obtained optimum values for a , the momentum conservation equation is solvednumerically to get new pressure values at various meshes. The whole cycle of operationsis repeated at the subsequent time steps to get void propagation and pressure buildup withtime.

The two-phase momentum conservation equation has also been separately studiedto check the relative contribution of various terms on pressure gradient (temporalacceleration, spatial acceleration, gravity force and friction) as boiling initiates andpropagates. With reduction of flow rate, the contribution from friction term to pressurereduces. Once boiling initiates, the spatial acceleration term starts contributing to thepressure values. As the value of a crosses 0.95, the contribution from this term increasesenormously, thus accounting for the pressure buildup from sodium vapor. The other termsare negligible. Conditions for flow reversal and slug ejection occur at a very high values ofa > 0.95. To model the motion of the slugs, simple Runge Kutta method is used to solvefor the velocity of ejection in either direction.

The code has been tried on BN 800 benchmark data and PFBR data. One couldobserve good void propagation for PFBR but not in BN-800 where the void coefficient isclose to zero. The conditions of flow reversal and slug ejection are not observed in eitherof the cases. The results and formulations are being rechecked further.

5.8 Fuel Freezing in Coolant Channels(S. Marimuthu and Om Pal Singh)

The study of freezing molten fuel following core meltdown accident in fastreactors is important from the point of view of knowing the rate of fuel relocation, finalfuel relocation site, plugging of flow paths and the extent of plugging, and finally whetherthe post accident heat removal is possible. Therefore, a study was taken up to develop themethodology of calculating the time and rate of freezing of molten fuel flowing throughcylindrical channels, the walls of which are below the freezing temperature of the moltenfuel. Initially a simple case in which molten fuel at its freezing point is filled in a channelwith radius of 12.7 mm, is taken up to find the time of freezing of fuel and its freezingrate. Considering the heat transfer from molten fuel to solid fuel crust and then to stainlesssteel wall and finally to the coolant, a formalism is reported [1] that gives the rate ofcooling of molten fuel and solid crust formation. Based on the reported formulation a codehas been developed and checked against the CRBR data. The total time of freezing iscalculated as 16.04 s. The freezing time calculated by Gassor and Kazimi [1] is 16.5 s. Therate of solid crust formation and the radius of the molten channel as a function of time areshown in Figs. 5.8.1 and 5.8.2. The details are given in Ref. [2]. The formulation is to beextended further for a flowing molten fuel and finally a specific study for PFBR whenseven subassemblies are in molten form and the fuel is flowing downward through lower

5-11

12

10

I*I-

2.5

4 6 8 10 12 14 16

Cooling Time (s)

Fig. 5.8.1: Radius of Fuel Channel

11.5

OJK

•-5

2 4 6 8 10

Cooling Time (s)

Fig. 5.8.2: Rate of Cooling

12 14 16

axial blanket is to be taken up to find the rate of molten fuel freezing and the time taken bythe fuel and the extent of fuel to reach the core catcher.

References1. Ronald D. Gasser and Musjid S! Kazimi, "A Study of Post Accident Molten Fuel

Dovmward Streaming through the Axial Shield Structure in LMFBR" Nucl. Tech.,Vol.33, pp.248 (1977).

2. S. Marimuthu and Om Pal Singh, "Fuel Freezing in Coolant Channel" RPD/SAS-91(1997).

5 - 1 2

APPENDIX A

PUBLICATIONS

A.I Journals

1. K. Devan, V. Gopalakrishnan, P. Mohanakrishnan and M.S. Sridharan/'Preparation ofMultigroup Lumped Fission Product Cross-sections from ENDF/B-VI for FBRs", Annalsof Nuclear Energy, 25 , p 161, 1998.

2. S. John Collins, N. L. Mary, G. Radhakrishnan and A. Dhathathreyan," Studies on Spreadof Monolayer Derivative of Syrene - Maleic Anhydride Copolymer", Journal of ChemicalSociety, Faraday Trans., 93^4021, 1997.

3. M. Rachi, T. Yamamoto, A.K. Jena and T. Takeda, "Parametric Study on Fast Reactorswith Low Sodium Void Reactivity by the use of Zirconium Hydride Layer in InternalBlanket", Journal of Nuclear Science and Technology, 34 (2), 193-201, 1997.

A.2 Conferences/Symposia

1. T.M. John and J.E. Hoogenboom, "A Homogenisation Procedure for Central ElementsLocated in the Reflector of an HTR", Annual Meeting on Nuclear Technology, Achen,Germany, May 1997.

2. Placid Rodriguez, S.B. Bhoje and S.M. Lee, "Reactor Safety for Nuclear ResourceUtilisation in India", Int. Symposium on Nuclear Fuel Cycle and Reactor Strategies :Adjusting to New Realities, IAEA, Vienna, June 2-6, 1997.

3. P.T. Krishnakumar, "Chemometric Study of Match Factory Industrial Effluents",Conference on Water Pollution, Slovenia, June 1997.

4. R. Harish and Ora Pal Singh, "Analysis of UTOP and ULOF Accident in BN-800 LikeReactor", 6th Consultancy Meeting on IAEA/EC Comparative Calculations for SevereAccident in BN-800 Reactor, held in Brussels, Belgium, June 30 - July 3, 1997.

5. T.M. John, J.E. Hoogenboom and A.J.J. Bos, "The Monte Carlo Midway CouplingEstimator for Calculation of Deep Penetration Fluxes", Int. Conf. on MathematicalMethods & Supercomputing for Nucl. Appl., Saratoga, New York, USA, Oct. 1997.

6. R. Harish and Om Pal Singh, "Analysis of Unprotected Transient Over Power and Loss ofFlow Accidents in BN-800 Like Reactor", 7th Consultancy Meeting on IAEA/ECComparative Calculations for Severe Accident in BN-800 Reactor, held in Vienna,Dec. 11-12, 1997.

A - l

7. P.T. Krishnakumar,"Design of Magnetic Sensor for Process Engineering with LeastCalibration Uncertainty", Proc. Seminar on Automation in Process Engineering andManufacture (SAPREM 97), IGCAR Kalpakkam, Nov. 97.

8. G.S. Srinivasan and Om Pal Singh, "A Neural Network Methodology for Automation inPlant Condition Monitoring", Proc. Seminar on Automation in Process Engineering andManufacture (SA-PREM), IGCAR Kalpakkam, Nov. 26-28, 1997.

9. Om Pal Singh et al., "Reactor Shutdown System for PFBR", Proc. Workshop on ReactorShutdown System", IGCAR, Kalpakkam, March 4-6, 1997.

10. S.M. Lee, "Safety Considerations in Selection of Absorber Rod Reactivity Worth inPFBR", Proc. Workshop on Reactor Shutdown Systems, Kalpakkam, March 4-6, 1997.

11. S.M. Lee, "Developments in Reactor Physics in Support of the FBR Programme in India",Invited Talk, Proc. 12th National Symposium on Radiation Physics (NSRP-12), Jodhpur,Jan.27-29, 1998.

12. M.S. Sridharan, A.K. Jena and S.M. Lee, "A Study of Fission Product Activity in TypicalFast Reactors", Proc. 12th National Symposium on Radiation Physics (NSRP-12),Jodhpur, Jan.27-29, 1998.

13. V. Gopalakrishnan and S. Ganesan, "Influence of Energy Dependence of Dilution Cross-Section on Self Shielding in the Resolved Resonance Region", Proc. 12th NationalSymposium on Radiation Physics (NSRP-12), Jodhpur, Jan.27-29, 1998.

14. K. Devan and R.S. Keshavamurthy, "Rational Approximations to Reich-Moore CrossSections", Proc. 12th National Symposium on Radiation Physics (NSRP-12), Jodhpur,Jan.27-29, 1998.

15. R.S. Keshavamurthy, R. Indira, C.P. Reddy and A.K. Jena, "Comparison of 2D TransportCalculations with Measured Detector Location Flux and Cover Gas Activity in FBTR",Proc. 12th National Symposium on Radiation Physics (NSRP-12), Jodhpur, Jan.27-29,1998.

16. R.S. Keshavamurthy and R. Harish, "Accurate Evaluation of Resonance Integrals", Proc.12th National Symposium on Radiation Physics (NSRP-12), Jodhpur, Jan.27-29, 1998.

17. R. Indira, "Neutron Transport Studies for Calculation of Neutron Flux at Under-VesselDetector Location in PFBR", Proc. 12th National Symposium on Radiation Physics(NSRP-12), Jodhpur, Jan.27-29, 1998.

18. D.K. Mohapatra, S. Sivakumar, C.P. Reddy and P. Mohanakrishnan, "Initial Neutron FluxMeasurements in KAMINI Reactor," Proc. 12th National Symposium on RadiationPhysics (NSRP-12), Jodhpur, Jan.27-29, 1998.

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19. P.T. Krishnakumar, "Experimental Design of Neutron Cross-section Measurements withLeast Systematic Uncertainty", Proc. 12th National Symposium on Radiation Physics(NSRP-12), Jodhpur, Jan.27-29, 1998.

20. Om Pal Singh, "A Methodology for Sodium Temperature Reactivity CoefficientMeasurements in FBTR by Reactor Noise Analysis Method", Proc. 12th NationalSymposium on Radiation Physics (NSRP-12), Jodhpur, Jan.27-29, 1998.

21. G.S. Srinivasan, S. Ponpandi and Om Pal Singh, "Dynamic Power Coefficient ofReactivity Calculation for PFBR", Proc. 12th National Symposium on Radiation Physics(NSRP-12), Jodhpur, Jan.27-29, 1998.

22. R. Harish and Gm Pal Singh, "Analysis of Severe Accidents in Near Zero Void ReactivityFast Reactor", Proc. 12th National Symposium on Radiation Physics (NSRP-12),Jodhpur, Jan. 27-29, 1998.

A.3 Thesis

1. P.T. Krishnakumar," Study of Correlations in Neutron Cross-Section Measurements",Ph. D. Thesis, Indian Institute of Technology, Madras, 1997.

A.4 IGC Report

1. S. John Collins (Ed), "Activity Report of Reactor Physics Division - 1996", IGC-90,1997.

A.5 News Letter

1. D.K. Mohapatra and P. Mohanakrishnan, "A New Reactor for Radiation Physics Studiesand Radiation Applications", ISRP Newsletter, Oct. 1997.

A.6 Internal Reports

1. P. Mohanakrishnan and S.J. Collins, "Reactivity Loss with Burnup of FBTR 23 SA Core -Use of 3 D Burnup Code FARCOB", RG/RPD/CPS-2, Oct. 1997.

2. T.M. John, P. Mohanakrishnan, Om Pal Singh and S.M. Lee, "Absorber Rods",PFBR/01113/DN/1029,Nov. 1997.

3. A. John Arul and C. P. Reddy, "Subassembly-wise Power Distribution for 26 SA Mark-ICore and its Dependence on Source SA", RG/RPD/LMS-7, Feb. 1997.

4. C.P. and S. Sivakumar, "Operation of FBTR with 28 Fuel SA in the Core",RG/RPD/LMS-8, Feb. 1997.

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5. S. Sivakumar and C.P. Reddy, "Measurement of Isothermal Temperature Coefficient with27 Subassemblies", RG/RPD/LMS-9, July 1997.

6. A. John Aral and C.P. Reddy, "Reactivity Loss Rate due to Burnup in FBTR SmallCarbide Core at 25000 MWd/t Burnup", RG/RPD/LMS-10, July 1997.

7. S. Sivakumar and C.P. Reddy, "Power Coefficient of Reactivity for the 27 SA Core",RG/RPD/LMS-11, Dec. 1997.

8. S. Sivakumar, C.P. Reddy and V. Sathiamoorthy, "Calibration of Control Rods for 26 fuelSA Core", RPD/OMS/FBTR/Ol 100/CR-043, Jan. 1997.

9. S. Sivakumar, A.John Arul and C.P. Reddy,"Calibration of Control Rods for 26 Fuel SACore after Resuffling", RPD/LMS/FBTR/01100/CR/044, Feb. 1997.

10. S. Sivakumar, A. John Arul and C.P. Reddy, "Calibration of Control Rods for 27 FuelSubassembly Core", RPD/LMS/FBTR/01100/CR/046, Sept. 1997.

11. S. John Collins and P. Mohanakrishnan, "Reactivity Worth of Diverse Safety Rods",PFBR/01113/DN/10271, April 1997.

12. S. John Collins and P. Mohanakrishnan, "Reactivity Worth of Control and Safety Rodsand Antishadowing Effects", PFBR/01113/DN/1028, August 1997.

13. S. Govindarajan, T.M. John and P. Mohanakrishnan, "Active Core Height",PFBR/31110/DN/1038, Nov. 1997.

14. P.P. Vinayagam, S.J. Winston and P. Mohanakrishnan, "Size of Fuel Subassembly",PFBR/31100/DN/1021, Oct. 1997.

15. R.R. Ramanarayanan and P. Mohanakrishnan, "Report on Raising of KAMESfl Power to30 kW", ROD/KAMINJ7CR-04, Oct. 1997.

16. D.K.Mohapatra, S. Sivakumar, C.P. Reddy and P.Mohanakrishnan, "Initial Measurementsof Neutron Flux at Different Locations in the KAMDSfl", RG/RPD/CPS-1, Sep. 1997.

17. Om Pal Singh and T.M. John, "Homogeneous Versus Heterogeneous Core", PFBR/01113/DN/1030,Nov. 1997.

18. T.M. John, "Refuelling Interval", PFBR/01113/DN/1031, Dec. 1997.

19. S. J. Collins/'Parallel Computing Through Message Passing Interface", RG/RPD/CPS-3,Oct. 1997.

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20. P. Puthiyavinayagam and R. Indira, "Design of Core Shielding", PFBR/31100/DN/1023,Dec. 1997

21. Radhakrishnan G. and A.K. Jena, "Shield Requirements of PFBR Fresh Fuel SubassemblyTransfer Chamber, RPD/RSS/120, February 1997.

22. R. Indira, "Neutron Flux at Under Vessel Detector Location in PFBR", RPD/RSS/127,

23. A.K. Jena, etal. "Revised Proposal, PFBR Bulk Shielding", RPD/RSS/129, Dec. 1997

24. R.S. Keshavamurthy, "Dose to Personnel in RCB Following Discharge Pipes RuptureIncident in the Two Loop Conceptual Design of PFBR due to FPFs, RPD/RSS/130, Dec.1997.

25. A.K. Jena and R.S. Keshavamurthy, "Tritium Generation and Pathways in PFBR",PFBR/01115/DN/1017/R-A, Jan 1997.

26. R.S. Keshavamurthy, "Release of Radioactivity into Environment Under HCDA fromPFBR", PFBR/01115/DN/1018, Aug. 1997.

27. A.K. Jena and P. Mohanakrishnan, "PFBR Bulk Shielding Experiments at APSARANeutron Flux at Converter Assembly", RPD/RSS/122, July 1997.

28. R. Indira and A.K. Jena, "Analysis of the Effects of Using Aluminium Box in APSARAShielding Experiments", RPD/RSS/121, April 1997.(not released for distribution)

29. A.K. Jena and P.Mohanakrishnan, "Neutron Flux at Converter Assembly (PFBR BulkShielding Experiments at APSARA), RPD/RSS/122, April 1997.

30. A.K. Jena et al., "Converter Plate Assembly", RPD/RSS/123, March 1997

31. D. Sunil Kumar, M.S. Sridharan and A.K. Jena, "Fission Product Gamma SpectrumCalculation for PFBR Mark II Core Subassmbly", RPD/RSS/124, Dec. 1997.

32. A.K. Jena etal., "PFBR Bulk Shielding Experiments at APSARA - Measurements ofNeutron Flux at Shielding Corner", RPD/RSS/125, Aug. 1997.

33. R. Indira and A.K. Jena, "Activation of Aluminium Box and the Consequent Dose Rates :PFBR Shielding Mockup Experiments in APSARA, RPD/RSS/126, Aug. 1997.

34. A.K. Jena, "PFBR Bulk Shielding Experiments at APSARA - Shielding Models,RPD/RSS/128, Aug. 1997.

35. S. Raghupathy, R. Indira et al, "Conceptual Design of Shielding for Reactor Assembly"PFBR/3100/DN/1005/R-B, Dec. 1997.

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36. A. K. Bhaduri, R. Indira et al. "Selection of Hardfacing Material for NSSS Components ofPFBR" IGCAHTF/97/01, March 1997.

37. A. K. Bhaduri, R. Indira et al. "Final Report on Selection of Hardfacing Material forNSSS Components of PFBR" IGCAHTF/97/01, Dec 1997.

38. S. Govindarajan, Om Pal Singh, G. Vaidyanathan and G. Muralikrishna, "ConceptualDesign of Reactor Shutdown System", PFBR/31000/TF/1014, May 1997.

39. A. John Arul and Om Pal Singh, "Distribution of SCRAM Signals to PFBR Based onReliability Analysis", RPD/SAS-90, June 1997.

40. S. Marimuthu and Om Pal Singh, "Fuel Freezing in Coolant Channels", RPD/SA-91,1997.

41. T. Satiyasheela and R. Harish, "Numerical Solution of Transient Heat ConductionEquation with Temperature Dependent Thermal Conductivity", RPD/SAS-92, Nov.97.

42. A. John Arul and Om Pal Singh, "PSA on Flow Blockage of FBTR Fuel Subassembly",RPD/SAS-93, Dec. 1997.

43. S. Marimuthu, "Failure Rate Data on FBR Systems", RPD/SAS-94, Dec. 1997.

A - 6

B.I RPD Seminars

1. T. Satiyasheela

2. S. John Collins,

3. A. K. Babbar,

4. R. Harish

5. V. Jagannathan

6. A. K. Jena

7. S. John Collins

8. R. Krimse

B.2 Other Lectures

1. S.M.Lee

2. S.M. Lee

3. Om Pal Singh

APPENDIX B

SEMINARS

Numerical Solutions of Point Kinetics Equation (Jan. 8,1997).

Highlights of "Second Int. Workshop on Parallel Processing andSupercomputing Applications in Science and Engineering" held atICTP, Trieste, Italy during 9-27 Sept. 1996(Jan. 10,1997).

Probabilistic Safety Analysis (Jan. 27,1997).

Highlights of the IAEA Meeting on "IAEA/EC ComparativeCalculations of ULOF in BN-800 Reactor" held in Vienna, Austria,Dec.4-6, 1996(July 9,1997).

Physics of Nuclear Power Options(July 16,1997).

PFBR Bulk Shielding Experiments at APSARA(Oct. 15,1997)

Parallel Computing through Message Passing Interface(Nov. 12,1997)

The Next Generation Power Plants - International Trends inNuclear Safety(Dec. 22,1997).

"Nuclear Electricity in India", Junior Science Club, Kalpakkam(Jan. 31,1997)

"Relevance and Status of FBR Technology in India", Workshop forJournalists organised by DAE and Subhas Chandra Bose Instituteof Journalism & Communication, Bombay(June 30,1997)

"Nuclear Scenario in India",Anna Institute of Management, Chennai(Aug.20,1997).

A - 7

REACTOR PHYSICS DIVISION

Dr. S.M. LEE, HHEAD

GRST.S.

SAFETY ANALYSIS SECTION

Dr. Om Pal

.S. SrinivasanHarishPonpandi

SathiyasheelaMarimuthu

Singh, G

EECC

c

CORE PHYSICS SECTION

Dr. P. Mohanakrishnan, G

Dr. T.M. JohnDr. P.T. Krishna KumarD.KMohapatraS. John Collins

GECC

RADIATION SHIELDING SECTION

Dr. A. K. Jena, F

Dr. R. IndiraDr. R.S. KeshavamurthyRadhakrishnan G.M.S. SridharanD.Sunil Kumar

FEDDC

LABORATORY ANDMEASUREMENTS SECTION

Dr. C.P.Reddy, F

A. John AmiS. SivakumarA.G.Rafi AhmedV. SathiamoorthyS.Raghu Kumar

DDD

SBSB

NUCLEAR DATA SECTION

V.Gopalakrishnan, E

KDevan

OFFICE

S. RangamaniS. JanakiN. Subbulakshmi

Total strength 29

Sr.PASteno