Research ArticleCalculation of Radioactivity and Dose Rate of ActivatedCorrosion Products in Water-Cooled Fusion Reactor

Jingyu Zhang Lu Li Shuxiang He and Yixue Chen

School of Nuclear Science and Engineering North China Electric Power University Beijing 102206 China

Correspondence should be addressed to Jingyu Zhang poptnt163com

Received 27 May 2016 Accepted 19 July 2016

Academic Editor Massimo Zucchetti

Copyright copy 2016 Jingyu Zhang et al This is an open access article distributed under the Creative Commons Attribution Licensewhich permits unrestricted use distribution and reproduction in any medium provided the original work is properly cited

In water-cooled reactor the dominant radioactive source term under normal operation is activated corrosion products (ACPs)which have an important impact on reactor inspection and maintenance A three-node transport model of ACPs was introducedinto the new version of ACPs source term code CATE in this paper which makes CATE capable of theoretically simulating thevariation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions For code testing MITPWR coolant chemistry loop was simulated and the calculation results from CATE are close to the experimental results fromMITwhich means CATE is available and credible on ACPs analysis of water-cooled reactor Then ACPs in the blanket cooling loop ofwater-cooled fusion reactor ITER under constructionwere analyzed usingCATE and the results showed that themajor contributorsare the short-life nuclides especially Mn-56 At last a point kernel integration code ARShield was coupled with CATE and the doserate around ITER blanket cooling loop was calculated Results showed that after shutting down the reactor only for 8 days the doserate decreased nearly one order of magnitude which was caused by the rapid decay of the short-life ACPs

1 Introduction

In the cooling loop of water-cooled reactor the oxidation andcorrosion of the metal material by water are inevitable andpart of the corrosion productswill be activated under neutronirradiation and become radioactive which are called acti-vated corrosion products (ACPs) Some ACPs continuouslydecay and emit harmful gamma-rays even after shutdown ofthe reactor for a long time According to the surveillance datafrom French PWR plants more than 90 of integrated doseunder normal operation is due to ACPs [1] whichmeans vitalimpact of ACPs on radiation protection For the water-cooledfusion reactor it is also necessary to research the formationmechanism and transport process of ACPs and to predictthe variation and distribution of radioactivity and dose rateof ACPs which is important to radiation shielding designinspection and maintenance as well as accident analysis ofthe reactor

Because of the importance of ACPs the relevant researchwas started since 1960s in PWR plants and many codeshave been developed [2] for example CORA from EPRIPACTOLE from CEA CRUDTRAN from KAERI and

MIGA-RT fromBSA which were widely used in ACPs analy-sis of PWR plants and showed good effect For ACPs analysisof water-cooled fusion reactor especially the InternationalThermonuclear Experimental Reactor (ITER) under con-struction three codes were developed which are PACTITERfrom CEA TRACT from UKAEA and CATE from NCEPU(North China Electric Power University) by the author In theinitial version of CATE the model adopted is mainly basedon the empirical coefficient method [3] which is simple butnot universal for different operation condition In the latestversion of CATE a three-node transport model was adoptedwhich is based on the theory that the main driving force forcorrosion products transport is the temperature change ofthe coolant throughout the loop and the resulting change inmetal ion solubility in the coolant [4] making CATE capableof theoretically simulating the variation and the distributionof ACPs and have a larger scope of application

The basic theory and equations of the three-node trans-portmodel were described in the second section of this paperand ACPs in MIT PWR coolant chemistry loop and ITERblanket cooling loop were calculated using the code CATErespectively in the third section and the fourth sectionThen

Hindawi Publishing CorporationScience and Technology of Nuclear InstallationsVolume 2016 Article ID 6051834 6 pageshttpdxdoiorg10115520166051834

2 Science and Technology of Nuclear Installations

In-flux pipe

Out-flux pipe

Flow direction

Dissolution ofcorrosion products

Deposition ofcorrosion products

Activ

atio

n of

corr

osio

n pr

oduc

tsDissolution of

ACPsDeposition of

ACPs

Neu

tron

irrad

iatio

n

Coolant

Figure 1 Transport process of ACPs in the cooling loop based onthe three-node model

the dose rate around ITER blanket cooling loop caused byACPs was calculated using the point kernel integration codeARShield in the fifth section In the last section a compre-hensive comment was presented

2 Description of the Three-NodeTransport Model

In the model the cooling loop is divided into three nodeswhich are the in-flux pipe the coolant and the out-flux pipeThe corrosion products transport between the nodes is bymeans of dissolution and deposition Nickel ferrite is themain corrosion products in the cooling loop and it shows aninverse solubility curvewith temperature As the coolant tem-perature changes around the cooling loop the saturation sta-tus of the corrosion products in the coolant also changes Inthe region of out-flux (eg the heat-exchanger) as the coolanttemperature decreases the soluble species become under-saturated so that there is a driving force for the corrosionproducts on the pipe surface to dissolve into the coolant torestore a saturation concentration of corrosion products inthe coolant On the other hand in the region of in-flux (egthe core in PWR or the blanket in fusion reactor) where thecoolant temperature increases the soluble species exist in asupersaturated state so that there is a driving force for the sol-uble species to deposit on the pipe surfaceThrough this pro-cess the corrosion products are transported from the out-fluxpipe to the in-flux pipe

In the region of in-flux part of the corrosion productsabsorb neutrons and become ACPsThe ratio of ACPs to cor-rosion products on the pipe surface is larger than that in thecoolant so there is a net flow of ACPs from the pipe surface tothe coolant by isotope exchange In the region of out-flux theratio of ACPs to corrosion products in the coolant is largerthan that on the pipe surface so there is a net flow of ACPsfrom the coolant to the pipe surface Through this processACPs are transported from the in-flux pipe to the out-fluxpipe which is described in Figure 1

In brief the three-node transport model is based on thetheory that the main driving force for ACPs transport is thetemperature change of the coolant throughout the loop andthe resulting change in metal ion solubility in the coolant Sowhether it is PWR temperature range (280sim320∘C) or ITER

temperature range (140sim180∘C) the three-node transportmodel is all applicable

In CATE the corresponding equations are as follows

119889119868

1198981

119889119905

= 119891

1198981(1 minus 120572

1)CR1119860

1

+ 119896

1198981(

119868

1198982

119881

minus 119878

1198981)119860

1

119889119868

1198982

119889119905

= 119891

1198981120572

1CR1119860

1+ 119891

1198983120572

3CR3119860

3

minus 119896

1198981(

119868

1198982

119881

minus 119878

1198981)119860

1

+ 119896

1198983(119878

1198983minus

119868

1198982

119881

)119860

3minus 120596

119876

119881CVCS119868

1198982

119889119868

1198983

119889119905

= 119891

1198983(1 minus 120572

3)CR3119860

3

minus 119896

1198983(119878

1198983minus

119868

1198982

119881

)119860

3

119889119873

1015840

1198991

119889119905

= sum

119896

120594

119896rarr119899120582

119896119873

1015840

1198961

+sum

119898

120590

120574119898rarr119899120601119868

1198981

119873

119860

119860

119898

minus 119896

1198991(

119860

119899

119860

119898

119873

1015840

1198991

119868

1198981

119878

1198981minus

119873

1015840

1198992

119881

)119860

1

minus (120582

119899+ 120590

119886119899120601)119873

1015840

1198991

119889119873

1015840

1198992

119889119905

= sum

119896

120594

119896rarr119899120582

119896119873

1015840

1198962

+ 119877sum

119898

120590

120574119898rarr119899120601119868

1198982

119873

119860

119860

119898

+ 119896

1198991(

119860

119899

119860

119898

119873

1015840

1198991

119868

1198981

119878

1198981minus

119873

1015840

1198992

119881

)119860

1

minus 119896

1198993(

119873

1015840

1198992

119881

minus

119860

119899

119860

119898

119873

1015840

1198993

119868

1198983

119878

1198983)119860

3

minus (120582

119899+ 119877120590

119886119899120601)119873

1015840

1198992

minus 120596

119876

119881CVCS119873

1015840

1198992

119889119873

1015840

1198993

119889119905

= sum

119896

120594

119896rarr119899120582

119896119873

1015840

1198963

+ 119896

1198993(

119873

1015840

1198992

119881

minus

119860

119899

119860

119898

119873

1015840

1198993

119868

1198983

119878

1198983)119860

3

minus 120582

119899119873

1015840

1198993

(1)

where 119868119898119894

is the mass of nuclide119898 in corrosion products kgthe subscript 119894 = 1 refers to the zone of in-flux pipe 119894 = 2refers to the zone of bulk coolant and 119894 = 3 refers to thezone of out-flux pipe CR

119894is the corrosion rate of the base

metal kg(m2sdots) 120572119894is the ratio of release rate to corrosion

rate of the corrosion products 119891119898119894

is the mass fraction of

Science and Technology of Nuclear Installations 3

nuclide119898 in the basemetal 119860119894is the contact area between

coolant and pipe m2 119896119898119894

is the mass transfer coefficient ofnuclide 119898 ms 119881 is the coolant volume in the cooling loopm3 119878119898119894

is the solubility of nuclide 119898 in the coolant close tothe pipe surface kgm3 120596 is the collection efficiency of filterand resin in chemical and volume control system (CVCS) 119876 is the volume flow rate of coolant into CVCS m3s 119881CVCSis the coolant volume in CVCS m31198731015840

119899119894

is the atom numberof nuclide 119899 in ACPs atoms and the superscript 1015840 meansnuclide 119899 is radioactive 120594

119896rarr119899is the decay branching ratio

from nuclide 119896 to nuclide 119899 120582119896is the decay rate constant of

nuclide 119896 sminus1 120590120574119898rarr119899

is the radiative capture cross section ofconverting nuclide119898 to nuclide 119899 cm2 120601 is the neutron flux119899(cm2sdots) 119860

119898is the molar mass of nuclide 119898 kgmol119873

119860is

the Avogadro number 120590119886119899

is the absorption cross section ofnuclide 119899 cm2 119877 is the ratio of coolant residence time in thezone of in-flux pipe to that in the whole cooling loop

In the above model some assumptions are made includ-ing the following (1) the concentration of nuclides in thecoolant keeps the same along the cooling loop which isbecause it takes the coolant only several seconds to circlearound the cooling loop which means the homogenizationeffect of nuclides in the coolant is dominant (2) the influenceof activation and decay on the mass of corrosion products isneglected because the quantity of ACPs is very limited com-pared to that of corrosion products (3) the concentration ofnuclides on the pipe surface is equal to its solubility in theadjacent coolant

To solve the above differential equations the fourth-orderRunge-Kutta method is used which is fast and can controlthe error well Moreover an algorithm of adaptive time stepis adopted in CATE for determining the time step reasonablyand the calculation speed can be improved

3 Verification of CATE Code through MITPWR Coolant Chemistry Loop (PCCL)

31 Description of MIT PCCL MIT PCCL was chosen hereto test the code CATE which is a small scale loop operatedunder constant coolant chemistry in the MIT reactor toclosely simulate the primary circuit of a typical PWR powerplant The main operation data of the loop is presented inTable 1 [5]

The basemetal of in-flux pipe is Zircaloy-4 and the corro-sion rate of thismaterial is nearly 0The basemetal of out-fluxpipe is mainly Inconel and the corrosion rate of it is 355119864 minus7 kgday

32 Results of ACPs Radioactivity in MIT PCCL The surveil-lance data from MIT PCCL for 42 days of full poweroperation was published and the radioactivity of ACPs in theloop is as shown in Table 2

Calculation results from CATE are close to the experi-mental results fromMIT whichmeans CATE is available andcredible on ACPs analysis of water-cooled nuclear reactorThe calculation results of mass of corrosion products in thein-flux pipe the coolant and the out-flux pipe are respec-tively 166119864 minus 6 kg 148119864 minus 9 kg and 396119864 minus 5 kg for CATE

Table 1 The operation data of MIT PCCL

Parameter ValueTemperature (∘C) 2739sim3156Mass flow rate (kgs) 684119864 minus 2

Flow velocity (ms) 283sim318Volume of coolant (m3) 384119864 minus 4

Surface area of in-flux pipe (m2) 260119864 minus 2

Surface area of out-flux pipe (m2) 970119864 minus 2

Mass flow rate in CVCS (kgs) 534119864 minus 5

Collection efficiency of CVCS 50Concentration of boron (ppm) 800Concentration of LiOH (ppm) 184Concentration of H

2

(cckg-H2

O) 25

Using these data we can calculate the specific radioactivity ofACPs in the in-flux pipe the coolant and the out-flux pipewhich is respectively 101119864 + 3GBqkg 233119864 + 2GBqkgand 689119864 + 1GBqkg for CATE The values are degressivewhich is consistent with the transport direction of ACPs inthe cooling loop and is rational

4 Calculation of ACPs Radioactivity in ITERBlanket Cooling Loop Using CATE Code

41 Description of ITER Blanket Cooling Loop The Interna-tionalThermonuclear Experimental Reactor (ITER) is underconstruction now In its design plan the primary coolingloops aremost water-cooled such as the blanket the divertorthe NB injector and the vacuum vessel So ITER can betreated as a representative of water-cooled fusion reactorChina is developing its own fusion reactor CFETR (ChinaFusion Engineering Test Reactor) whose design plan of heattransfer and radiation shielding partly refers to ITER soACPs analysis of ITER will benefit the work of source termcontrol for CFETR

The planned operation data of ITER blanket cooling loopis presented in Table 3 [6]

The base metal in in-flux pipe and out-flux pipe is thesame kind of stainless steel SS316 and the corrosion rate of itis as follows

SS316 CR = 20 times 10minus5 times 119905minus0614 kgsdotmminus2 sdotsminus1 (2)

The above equation and the element composition of SS316are quoted from [7] The relevant nuclear reaction data arequoted from the European Activation File EAF-2007 [8 9]and the cross section of activation reaction is collapsed from172 groups into 1 group using the corresponding neutronspectrum which is calculated with MCNP code [10] andbased on an ITER blanket module [11]The neutronics modeland neutron spectrum of the chosen blanket module can beseen in Figures 2 and 3

42 Results of ACPs Radioactivity in ITER Blanket CoolingLoop After 12 years of full power operation the mass ofcorrosion products and the radioactivity of ACPs in the loopare as shown in Table 4

4 Science and Technology of Nuclear Installations

Table 2 Radioactivity of ACPs in MIT PCCL after normal operation for 42 days

Data sourceRadioactivity of ACPs (GBq)

In-flux pipe Coolant Out-flux pipeCo-58 Co-60 Co-58 Co-60 Co-58 Co-60

MIT 179119864 minus 3 116119864 minus 4 365119864 minus 7 243119864 minus 8 271119864 minus 3 195119864 minus 4

CATE 151119864 minus 3 172119864 minus 4 309119864 minus 7 354119864 minus 8 244119864 minus 3 287119864 minus 4

Table 3 The planned operation data of ITER blanket cooling loop

Parameter ValueTemperature (∘C) 140sim174Mass flow rate (kgs) 851Average flow velocity (ms) 40Volume of coolant (m3) 3336Surface area of in-flux pipe (m2) 9914Surface area of out-flux pipe (m2) 13951Mass flow rate in CVCS (kgs) 255Collection efficiency of CVCS 50Concentration of H

2

(cckg-H2

O) 25

Figure 2 The neutronics model of the chosen blanket module(marked with a red ring)

From Table 4 we can see that mass of corrosion productsin the coolant is much lower than that on the pipe surfacewhich is due to the limitation of solubility of corrosion prod-ucts in the coolant It should be noticed that the corrosionproducts on the pipe surface include oxides and depositsTheoxides are compact and fixed while the deposits are loose andmobilizable and their mass is usually one order of magnitudelower than that of oxides So although the total mass ofcorrosion products on the pipe surface is tens of kg the actualmass of deposits is only in the magnitude of kg which isconsistent with the experience in safety analysis

We also can see that ACPs radioactivity on the pipesurface is much higher than that in the coolant which meansthe pipe surface is themain radioactive contamination regionand should be decontaminated regularly through waterchemistry method The specific radioactivity of ACPs in thein-flux pipe the coolant and the out-flux pipe is calculated as

Table 4 Calculation results of ACPs in ITER blanket cooling afternormal operation for 12 years from CATE

In-flux pipe Coolant Out-flux pipeMass of corrosionproducts (kg) 217119864 + 1 625119864 minus 3 266119864 + 1

Radioactivity of ACPs(GBq) 105119864 + 6 277119864 + 0 890119864 + 3

001 0

1 1 10 100

00101

110

100

Neutron energy (MeV)

1E minus 3

1E minus 4

1E minus 5

1E minus 6

1E minus 7

1E minus 8

1E minus 9

1E minus 10

1E

minus8

1E

minus7

1E

minus6

1E

minus5

1E

minus4

1E

minus3

Neu

tron

flux

(10

18

ncm

2middotsmiddot

MeV

)

Figure 3 The neutron spectrum of the chosen blanket module

484119864 + 4GBqkg 443119864 + 2GBqkg and 335119864 + 2GBqkgThe values are degressive which is consistent with the trans-port direction of ACPs in the cooling loop and is rational

From Table 5 we can see that the short-life nuclides(V-52 V-53 Cr-55 Mn-56 Co-58m Co-60m and Ni-57)are the major contributors to radioactivity in all the threeregions especially Mn-56 which alone contributes 6333of the radioactivity in the region of in-flux pipe and nearly80 of the radioactivity in the regions of coolant and out-fluxpipe That is very different from PWR in which the long-lifenuclides Co-58 and Co-60 are the main ACPs And when thereactor shuts down for a period of time (about several days)ACPs radioactivity will decrease obviously due to the rapiddecay of the short-life nuclides and then the right time forworkers to do maintenance comes

Science and Technology of Nuclear Installations 5

Table 5 The main ACPs after normal operation for 12 years fromCATE

Nuclide Half-life Radioactivity (GBq)In-flux pipe Coolant Out-flux pipe

V-49 330 d 404119864 + 2 507119864 minus 7 414119864 minus 6

V-52 374m 103119864 + 4 781119864 minus 3 269119864 + 0

V-53 16m 595119864 + 2 427119864 minus 4 668119864 minus 2

Cr-51 277 d 235119864 + 5 321119864 minus 1 848119864 + 2

Cr-55 350m 378119864 + 3 285119864 minus 3 933119864 minus 1

Mn-56 258 h 659119864 + 5 215119864 + 0 707119864 + 3

Fe-55 274 y 404119864 + 4 209119864 minus 1 767119864 + 2

Fe-59 445 d 479119864 + 3 108119864 minus 2 347119864 + 1

Co-57 272 d 167119864 + 4 198119864 minus 2 608119864 + 1

Co-58m 904 h 272119864 + 4 131119864 minus 2 579119864 minus 1

Co-58 709 d 379119864 + 4 367119864 minus 2 944119864 + 1

Co-60m 105m 294119864 + 3 134119864 minus 3 236119864 minus 3

Co-60 527 y 369119864 + 2 904119864 minus 4 322119864 + 0

Ni-57 356 h 157119864 + 3 624119864 minus 4 551119864 minus 2

Ni-63 100 y 571119864 + 2 741119864 minus 4 241119864 + 0

Total 104119864 + 6 277119864 + 0 889119864 + 3

Short-lived 706119864 + 5 217119864 + 0 707119864 + 3

5 Calculation of Dose Rate Caused by ACPsUsing ARShield Code

TheARShield code developed by NCEPU of China is appliedto dose rate calculation which is a new version of the pointkernel integration code QAD-CG developed by Los AlamosNational Laboratory ARShield breaks some restrictions ofQAD-CG such as complicatedmodeling complicated sourcesetting 3D fine mesh results statistics and large-scale com-puting efficiency and is proved to be reliable and efficient ondose rate calculation

The density of each radionuclide at chosen regionscalculated by CATE is introduced into ARShield and thenconverted to dose rate using point kernel integrationmethodwhich is as follows

119863 (119903)

= int

119881

119870119904 (119903

1015840

) 119861 (120583

1003816

1003816

1003816

1003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

1003816

119864) exp (minus120583 1003816100381610038161003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

1003816

) 119889119881

2120587

1003816

1003816

1003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

2

(3)

where 119903 is the point at which gamma dose rate is to becalculated 1199031015840 is the location of source in volume 119881 119881 is thevolume of source region 120583 is the total attenuation coefficientat energy 119864 |119903 rarr 119903

1015840

| is the distance between source point andpoint at which gamma intensity is to be calculated 119870 is theflux-to-dose conversion factor 119861 is the dose buildup factor

The geometry of hot leg pipe in ITER blanket coolingloop is adopted to represent the region of out-flux pipe whichhas an internal diameter of 0527m thickness of 002mand length of 33m The dose rate around the hot leg pipecalculated by ARShield is shown in Figure 4

We can see that after shutting down the reactor for 8 daysthe dose rate is much lower than that of normal operation for

0

1

2

3

4

5

6

7

8

9

10

Dos

e rat

e (m

Svh

)

Distance (cm)

minus1

minus200 0 200 400 600 800 1000 1200 1400 1600

Normal operationShutdown for 8 days

Figure 4 Variation of dose rate with the distance from the outersurface of the hot leg pipe

12 years which is because that contribution from the short-life ACPs decreases almost to zero after shutdown for 8 daysThe typical values of dose rate are as follows during normaloperation the dose rate at the outer surface of the pipe is880mSvh and the dose rate at a distance 1m away from theouter surface of the pipe is 252mSvh after shutdown for 8days the values drop to 105mSvh (contact) and 031mSvh(1m away) When these values are compared to the annualpermissible worker dose rate of 20mSvyear recommendedby the ICRP [12] the contact dose rate value after shutdownfor 8 days would allow approximately 19 h exposure per year

6 Conclusions

In this paper a three-node transport model was introducedinto theACPs source termcodeCATEmakingCATE capableof theoretically simulating the variation and the distributionof ACPs in a water-cooled reactor and suitable formore oper-ating conditions MIT PCCL was chosen to test the new ver-sion ofCATE and the calculation results fromCATEare closeto the experimental results from MIT which means CATEis available and credible on ACPs analysis of water-cooledreactor Then the radioactivity and composition of ACPs inITERblanket cooling loopwere analyzed usingCATE and theresults showed that the major contributors are the short-lifenuclides (V-52 V-53 Cr-55 Mn-56 Co-58m Co-60m andNi-57) for ITER especiallyMn-56That is very different fromPWR in which the long-life nuclides Co-58 and Co-60 arethe main ACPs At last the dose rate around ITER blanketcooling loop caused by ACPs was calculated through cou-pling the code CATE with a point kernel integration codeARShield The results showed that after shutting down thereactor only for 8 days the dose rate can decrease nearly oneorder of magnitude compared to that of normal operationwhich is caused by the rapid decay of the short-life ACPs In

6 Science and Technology of Nuclear Installations

the future CATEwill be applied for ACPs analysis of Chinesewater-cooled fusion reactor CFETR as soon as its design planis completed

Competing Interests

The authors declare that they have no competing interests

Acknowledgments

This work was sponsored jointly by ldquoNational Special Projectfor Magnetic Confined Nuclear Fusion Energyrdquo with Grantno 2014GB119000 and the ldquoFundamental Research Funds forthe Central Universitiesrdquo with Grant no 2014QN26 in ChinaThe authors would like to express their gratitude for thesupport

References

[1] A Rocher J L Bretelle and M Berger ldquoImpact of main radi-ological pollutants on contamination risks (ALARA) optimiza-tion of physico chemical environment and retention technicsduring operation and shutdownrdquo in Proceedings of the EuropeanWorkshop on Occupational Exposure Management at NPPs(ISOE rsquo04) Session 2 EDF Lyon France March 2004

[2] MRafiqueNMMirza SMMirza andM J Iqbal ldquoReviewofcomputer codes for modeling corrosion product transport andactivity build-up in light water reactorsrdquoNukleonika vol 55 no3 pp 263ndash269 2010

[3] L Li J Zhang W Song Y Fu X Xu and Y Chen ldquoCATE acode for activated corrosion products evaluation of water-cooled fusion reactorrdquo Fusion Engineering and Design vol 100pp 340ndash344 2015

[4] IAEA ldquoModelling of transport of radioactive substances inthe primary circuit of water-cooled reactorsrdquo IAEA-TECDOC-1672 2012

[5] C B Lee Modeling of Corrosion Product Transport in PWRPrimary Coolant Nuclear Engineering Department MIT 1990

[6] SNisan LD Pace and J-C Robin ldquoEvaluation of the activatedcorrosion products for the ITER heat transfer systemsrdquo inFusion Technology pp 1763ndash1766 Elsevier 1996

[7] P J Karditsas ldquoActivation product transport using TRACTORE estimation of an ITER cooling looprdquo Fusion Engineeringand Design vol 45 no 2 pp 169ndash185 1999

[8] R A Forrest J Kopecky and J-Ch Sublet ldquoTheEuropeanActi-vation File EAF-2007 neutron-induced cross section libraryrdquoUKAEA FUS 535 2007

[9] R A Forrest ldquoThe European Activation File EAF-2007 decaydata libraryrdquo UKAEA FUS 537 2007

[10] J F Briesmeister MCNP a general Monte Carlo N-particletransport code version 4C Los Alamos National LaboratoryLA-13709-M 2000

[11] H Iida V Khripunov and L Petrizzi ldquoNuclear analysis reportrdquoITER Document NAG-201-01-06-17-FDR 2001

[12] ICRP ldquoRecommendations of the international commission onradiological protection publication 60rdquo Annals of the ICRP 21Pergamon Press Oxford UK 1991

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2 Science and Technology of Nuclear Installations

In-flux pipe

Out-flux pipe

Flow direction

Dissolution ofcorrosion products

Deposition ofcorrosion products

Activ

atio

n of

corr

osio

n pr

oduc

tsDissolution of

ACPsDeposition of

ACPs

Neu

tron

irrad

iatio

n

Coolant

Figure 1 Transport process of ACPs in the cooling loop based onthe three-node model

the dose rate around ITER blanket cooling loop caused byACPs was calculated using the point kernel integration codeARShield in the fifth section In the last section a compre-hensive comment was presented

2 Description of the Three-NodeTransport Model

In the model the cooling loop is divided into three nodeswhich are the in-flux pipe the coolant and the out-flux pipeThe corrosion products transport between the nodes is bymeans of dissolution and deposition Nickel ferrite is themain corrosion products in the cooling loop and it shows aninverse solubility curvewith temperature As the coolant tem-perature changes around the cooling loop the saturation sta-tus of the corrosion products in the coolant also changes Inthe region of out-flux (eg the heat-exchanger) as the coolanttemperature decreases the soluble species become under-saturated so that there is a driving force for the corrosionproducts on the pipe surface to dissolve into the coolant torestore a saturation concentration of corrosion products inthe coolant On the other hand in the region of in-flux (egthe core in PWR or the blanket in fusion reactor) where thecoolant temperature increases the soluble species exist in asupersaturated state so that there is a driving force for the sol-uble species to deposit on the pipe surfaceThrough this pro-cess the corrosion products are transported from the out-fluxpipe to the in-flux pipe

In the region of in-flux part of the corrosion productsabsorb neutrons and become ACPsThe ratio of ACPs to cor-rosion products on the pipe surface is larger than that in thecoolant so there is a net flow of ACPs from the pipe surface tothe coolant by isotope exchange In the region of out-flux theratio of ACPs to corrosion products in the coolant is largerthan that on the pipe surface so there is a net flow of ACPsfrom the coolant to the pipe surface Through this processACPs are transported from the in-flux pipe to the out-fluxpipe which is described in Figure 1

In brief the three-node transport model is based on thetheory that the main driving force for ACPs transport is thetemperature change of the coolant throughout the loop andthe resulting change in metal ion solubility in the coolant Sowhether it is PWR temperature range (280sim320∘C) or ITER

temperature range (140sim180∘C) the three-node transportmodel is all applicable

In CATE the corresponding equations are as follows

119889119868

1198981

119889119905

= 119891

1198981(1 minus 120572

1)CR1119860

1

+ 119896

1198981(

119868

1198982

119881

minus 119878

1198981)119860

1

119889119868

1198982

119889119905

= 119891

1198981120572

1CR1119860

1+ 119891

1198983120572

3CR3119860

3

minus 119896

1198981(

119868

1198982

119881

minus 119878

1198981)119860

1

+ 119896

1198983(119878

1198983minus

119868

1198982

119881

)119860

3minus 120596

119876

119881CVCS119868

1198982

119889119868

1198983

119889119905

= 119891

1198983(1 minus 120572

3)CR3119860

3

minus 119896

1198983(119878

1198983minus

119868

1198982

119881

)119860

3

119889119873

1015840

1198991

119889119905

= sum

119896

120594

119896rarr119899120582

119896119873

1015840

1198961

+sum

119898

120590

120574119898rarr119899120601119868

1198981

119873

119860

119860

119898

minus 119896

1198991(

119860

119899

119860

119898

119873

1015840

1198991

119868

1198981

119878

1198981minus

119873

1015840

1198992

119881

)119860

1

minus (120582

119899+ 120590

119886119899120601)119873

1015840

1198991

119889119873

1015840

1198992

119889119905

= sum

119896

120594

119896rarr119899120582

119896119873

1015840

1198962

+ 119877sum

119898

120590

120574119898rarr119899120601119868

1198982

119873

119860

119860

119898

+ 119896

1198991(

119860

119899

119860

119898

119873

1015840

1198991

119868

1198981

119878

1198981minus

119873

1015840

1198992

119881

)119860

1

minus 119896

1198993(

119873

1015840

1198992

119881

minus

119860

119899

119860

119898

119873

1015840

1198993

119868

1198983

119878

1198983)119860

3

minus (120582

119899+ 119877120590

119886119899120601)119873

1015840

1198992

minus 120596

119876

119881CVCS119873

1015840

1198992

119889119873

1015840

1198993

119889119905

= sum

119896

120594

119896rarr119899120582

119896119873

1015840

1198963

+ 119896

1198993(

119873

1015840

1198992

119881

minus

119860

119899

119860

119898

119873

1015840

1198993

119868

1198983

119878

1198983)119860

3

minus 120582

119899119873

1015840

1198993

(1)

where 119868119898119894

is the mass of nuclide119898 in corrosion products kgthe subscript 119894 = 1 refers to the zone of in-flux pipe 119894 = 2refers to the zone of bulk coolant and 119894 = 3 refers to thezone of out-flux pipe CR

119894is the corrosion rate of the base

metal kg(m2sdots) 120572119894is the ratio of release rate to corrosion

rate of the corrosion products 119891119898119894

is the mass fraction of

Science and Technology of Nuclear Installations 3

nuclide119898 in the basemetal 119860119894is the contact area between

coolant and pipe m2 119896119898119894

is the mass transfer coefficient ofnuclide 119898 ms 119881 is the coolant volume in the cooling loopm3 119878119898119894

is the solubility of nuclide 119898 in the coolant close tothe pipe surface kgm3 120596 is the collection efficiency of filterand resin in chemical and volume control system (CVCS) 119876 is the volume flow rate of coolant into CVCS m3s 119881CVCSis the coolant volume in CVCS m31198731015840

119899119894

is the atom numberof nuclide 119899 in ACPs atoms and the superscript 1015840 meansnuclide 119899 is radioactive 120594

119896rarr119899is the decay branching ratio

from nuclide 119896 to nuclide 119899 120582119896is the decay rate constant of

nuclide 119896 sminus1 120590120574119898rarr119899

is the radiative capture cross section ofconverting nuclide119898 to nuclide 119899 cm2 120601 is the neutron flux119899(cm2sdots) 119860

119898is the molar mass of nuclide 119898 kgmol119873

119860is

the Avogadro number 120590119886119899

is the absorption cross section ofnuclide 119899 cm2 119877 is the ratio of coolant residence time in thezone of in-flux pipe to that in the whole cooling loop

In the above model some assumptions are made includ-ing the following (1) the concentration of nuclides in thecoolant keeps the same along the cooling loop which isbecause it takes the coolant only several seconds to circlearound the cooling loop which means the homogenizationeffect of nuclides in the coolant is dominant (2) the influenceof activation and decay on the mass of corrosion products isneglected because the quantity of ACPs is very limited com-pared to that of corrosion products (3) the concentration ofnuclides on the pipe surface is equal to its solubility in theadjacent coolant

To solve the above differential equations the fourth-orderRunge-Kutta method is used which is fast and can controlthe error well Moreover an algorithm of adaptive time stepis adopted in CATE for determining the time step reasonablyand the calculation speed can be improved

3 Verification of CATE Code through MITPWR Coolant Chemistry Loop (PCCL)

31 Description of MIT PCCL MIT PCCL was chosen hereto test the code CATE which is a small scale loop operatedunder constant coolant chemistry in the MIT reactor toclosely simulate the primary circuit of a typical PWR powerplant The main operation data of the loop is presented inTable 1 [5]

The basemetal of in-flux pipe is Zircaloy-4 and the corro-sion rate of thismaterial is nearly 0The basemetal of out-fluxpipe is mainly Inconel and the corrosion rate of it is 355119864 minus7 kgday

32 Results of ACPs Radioactivity in MIT PCCL The surveil-lance data from MIT PCCL for 42 days of full poweroperation was published and the radioactivity of ACPs in theloop is as shown in Table 2

Calculation results from CATE are close to the experi-mental results fromMIT whichmeans CATE is available andcredible on ACPs analysis of water-cooled nuclear reactorThe calculation results of mass of corrosion products in thein-flux pipe the coolant and the out-flux pipe are respec-tively 166119864 minus 6 kg 148119864 minus 9 kg and 396119864 minus 5 kg for CATE

Table 1 The operation data of MIT PCCL

Parameter ValueTemperature (∘C) 2739sim3156Mass flow rate (kgs) 684119864 minus 2

Flow velocity (ms) 283sim318Volume of coolant (m3) 384119864 minus 4

Surface area of in-flux pipe (m2) 260119864 minus 2

Surface area of out-flux pipe (m2) 970119864 minus 2

Mass flow rate in CVCS (kgs) 534119864 minus 5

Collection efficiency of CVCS 50Concentration of boron (ppm) 800Concentration of LiOH (ppm) 184Concentration of H

2

(cckg-H2

O) 25

Using these data we can calculate the specific radioactivity ofACPs in the in-flux pipe the coolant and the out-flux pipewhich is respectively 101119864 + 3GBqkg 233119864 + 2GBqkgand 689119864 + 1GBqkg for CATE The values are degressivewhich is consistent with the transport direction of ACPs inthe cooling loop and is rational

4 Calculation of ACPs Radioactivity in ITERBlanket Cooling Loop Using CATE Code

41 Description of ITER Blanket Cooling Loop The Interna-tionalThermonuclear Experimental Reactor (ITER) is underconstruction now In its design plan the primary coolingloops aremost water-cooled such as the blanket the divertorthe NB injector and the vacuum vessel So ITER can betreated as a representative of water-cooled fusion reactorChina is developing its own fusion reactor CFETR (ChinaFusion Engineering Test Reactor) whose design plan of heattransfer and radiation shielding partly refers to ITER soACPs analysis of ITER will benefit the work of source termcontrol for CFETR

The planned operation data of ITER blanket cooling loopis presented in Table 3 [6]

The base metal in in-flux pipe and out-flux pipe is thesame kind of stainless steel SS316 and the corrosion rate of itis as follows

SS316 CR = 20 times 10minus5 times 119905minus0614 kgsdotmminus2 sdotsminus1 (2)

The above equation and the element composition of SS316are quoted from [7] The relevant nuclear reaction data arequoted from the European Activation File EAF-2007 [8 9]and the cross section of activation reaction is collapsed from172 groups into 1 group using the corresponding neutronspectrum which is calculated with MCNP code [10] andbased on an ITER blanket module [11]The neutronics modeland neutron spectrum of the chosen blanket module can beseen in Figures 2 and 3

42 Results of ACPs Radioactivity in ITER Blanket CoolingLoop After 12 years of full power operation the mass ofcorrosion products and the radioactivity of ACPs in the loopare as shown in Table 4

4 Science and Technology of Nuclear Installations

Table 2 Radioactivity of ACPs in MIT PCCL after normal operation for 42 days

Data sourceRadioactivity of ACPs (GBq)

In-flux pipe Coolant Out-flux pipeCo-58 Co-60 Co-58 Co-60 Co-58 Co-60

MIT 179119864 minus 3 116119864 minus 4 365119864 minus 7 243119864 minus 8 271119864 minus 3 195119864 minus 4

CATE 151119864 minus 3 172119864 minus 4 309119864 minus 7 354119864 minus 8 244119864 minus 3 287119864 minus 4

Table 3 The planned operation data of ITER blanket cooling loop

Parameter ValueTemperature (∘C) 140sim174Mass flow rate (kgs) 851Average flow velocity (ms) 40Volume of coolant (m3) 3336Surface area of in-flux pipe (m2) 9914Surface area of out-flux pipe (m2) 13951Mass flow rate in CVCS (kgs) 255Collection efficiency of CVCS 50Concentration of H

2

(cckg-H2

O) 25

Figure 2 The neutronics model of the chosen blanket module(marked with a red ring)

From Table 4 we can see that mass of corrosion productsin the coolant is much lower than that on the pipe surfacewhich is due to the limitation of solubility of corrosion prod-ucts in the coolant It should be noticed that the corrosionproducts on the pipe surface include oxides and depositsTheoxides are compact and fixed while the deposits are loose andmobilizable and their mass is usually one order of magnitudelower than that of oxides So although the total mass ofcorrosion products on the pipe surface is tens of kg the actualmass of deposits is only in the magnitude of kg which isconsistent with the experience in safety analysis

We also can see that ACPs radioactivity on the pipesurface is much higher than that in the coolant which meansthe pipe surface is themain radioactive contamination regionand should be decontaminated regularly through waterchemistry method The specific radioactivity of ACPs in thein-flux pipe the coolant and the out-flux pipe is calculated as

Table 4 Calculation results of ACPs in ITER blanket cooling afternormal operation for 12 years from CATE

In-flux pipe Coolant Out-flux pipeMass of corrosionproducts (kg) 217119864 + 1 625119864 minus 3 266119864 + 1

Radioactivity of ACPs(GBq) 105119864 + 6 277119864 + 0 890119864 + 3

001 0

1 1 10 100

00101

110

100

Neutron energy (MeV)

1E minus 3

1E minus 4

1E minus 5

1E minus 6

1E minus 7

1E minus 8

1E minus 9

1E minus 10

1E

minus8

1E

minus7

1E

minus6

1E

minus5

1E

minus4

1E

minus3

Neu

tron

flux

(10

18

ncm

2middotsmiddot

MeV

)

Figure 3 The neutron spectrum of the chosen blanket module

484119864 + 4GBqkg 443119864 + 2GBqkg and 335119864 + 2GBqkgThe values are degressive which is consistent with the trans-port direction of ACPs in the cooling loop and is rational

From Table 5 we can see that the short-life nuclides(V-52 V-53 Cr-55 Mn-56 Co-58m Co-60m and Ni-57)are the major contributors to radioactivity in all the threeregions especially Mn-56 which alone contributes 6333of the radioactivity in the region of in-flux pipe and nearly80 of the radioactivity in the regions of coolant and out-fluxpipe That is very different from PWR in which the long-lifenuclides Co-58 and Co-60 are the main ACPs And when thereactor shuts down for a period of time (about several days)ACPs radioactivity will decrease obviously due to the rapiddecay of the short-life nuclides and then the right time forworkers to do maintenance comes

Science and Technology of Nuclear Installations 5

Table 5 The main ACPs after normal operation for 12 years fromCATE

Nuclide Half-life Radioactivity (GBq)In-flux pipe Coolant Out-flux pipe

V-49 330 d 404119864 + 2 507119864 minus 7 414119864 minus 6

V-52 374m 103119864 + 4 781119864 minus 3 269119864 + 0

V-53 16m 595119864 + 2 427119864 minus 4 668119864 minus 2

Cr-51 277 d 235119864 + 5 321119864 minus 1 848119864 + 2

Cr-55 350m 378119864 + 3 285119864 minus 3 933119864 minus 1

Mn-56 258 h 659119864 + 5 215119864 + 0 707119864 + 3

Fe-55 274 y 404119864 + 4 209119864 minus 1 767119864 + 2

Fe-59 445 d 479119864 + 3 108119864 minus 2 347119864 + 1

Co-57 272 d 167119864 + 4 198119864 minus 2 608119864 + 1

Co-58m 904 h 272119864 + 4 131119864 minus 2 579119864 minus 1

Co-58 709 d 379119864 + 4 367119864 minus 2 944119864 + 1

Co-60m 105m 294119864 + 3 134119864 minus 3 236119864 minus 3

Co-60 527 y 369119864 + 2 904119864 minus 4 322119864 + 0

Ni-57 356 h 157119864 + 3 624119864 minus 4 551119864 minus 2

Ni-63 100 y 571119864 + 2 741119864 minus 4 241119864 + 0

Total 104119864 + 6 277119864 + 0 889119864 + 3

Short-lived 706119864 + 5 217119864 + 0 707119864 + 3

5 Calculation of Dose Rate Caused by ACPsUsing ARShield Code

TheARShield code developed by NCEPU of China is appliedto dose rate calculation which is a new version of the pointkernel integration code QAD-CG developed by Los AlamosNational Laboratory ARShield breaks some restrictions ofQAD-CG such as complicatedmodeling complicated sourcesetting 3D fine mesh results statistics and large-scale com-puting efficiency and is proved to be reliable and efficient ondose rate calculation

The density of each radionuclide at chosen regionscalculated by CATE is introduced into ARShield and thenconverted to dose rate using point kernel integrationmethodwhich is as follows

119863 (119903)

= int

119881

119870119904 (119903

1015840

) 119861 (120583

1003816

1003816

1003816

1003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

1003816

119864) exp (minus120583 1003816100381610038161003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

1003816

) 119889119881

2120587

1003816

1003816

1003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

2

(3)

where 119903 is the point at which gamma dose rate is to becalculated 1199031015840 is the location of source in volume 119881 119881 is thevolume of source region 120583 is the total attenuation coefficientat energy 119864 |119903 rarr 119903

1015840

| is the distance between source point andpoint at which gamma intensity is to be calculated 119870 is theflux-to-dose conversion factor 119861 is the dose buildup factor

The geometry of hot leg pipe in ITER blanket coolingloop is adopted to represent the region of out-flux pipe whichhas an internal diameter of 0527m thickness of 002mand length of 33m The dose rate around the hot leg pipecalculated by ARShield is shown in Figure 4

We can see that after shutting down the reactor for 8 daysthe dose rate is much lower than that of normal operation for

0

1

2

3

4

5

6

7

8

9

10

Dos

e rat

e (m

Svh

)

Distance (cm)

minus1

minus200 0 200 400 600 800 1000 1200 1400 1600

Normal operationShutdown for 8 days

Figure 4 Variation of dose rate with the distance from the outersurface of the hot leg pipe

12 years which is because that contribution from the short-life ACPs decreases almost to zero after shutdown for 8 daysThe typical values of dose rate are as follows during normaloperation the dose rate at the outer surface of the pipe is880mSvh and the dose rate at a distance 1m away from theouter surface of the pipe is 252mSvh after shutdown for 8days the values drop to 105mSvh (contact) and 031mSvh(1m away) When these values are compared to the annualpermissible worker dose rate of 20mSvyear recommendedby the ICRP [12] the contact dose rate value after shutdownfor 8 days would allow approximately 19 h exposure per year

6 Conclusions

In this paper a three-node transport model was introducedinto theACPs source termcodeCATEmakingCATE capableof theoretically simulating the variation and the distributionof ACPs in a water-cooled reactor and suitable formore oper-ating conditions MIT PCCL was chosen to test the new ver-sion ofCATE and the calculation results fromCATEare closeto the experimental results from MIT which means CATEis available and credible on ACPs analysis of water-cooledreactor Then the radioactivity and composition of ACPs inITERblanket cooling loopwere analyzed usingCATE and theresults showed that the major contributors are the short-lifenuclides (V-52 V-53 Cr-55 Mn-56 Co-58m Co-60m andNi-57) for ITER especiallyMn-56That is very different fromPWR in which the long-life nuclides Co-58 and Co-60 arethe main ACPs At last the dose rate around ITER blanketcooling loop caused by ACPs was calculated through cou-pling the code CATE with a point kernel integration codeARShield The results showed that after shutting down thereactor only for 8 days the dose rate can decrease nearly oneorder of magnitude compared to that of normal operationwhich is caused by the rapid decay of the short-life ACPs In

6 Science and Technology of Nuclear Installations

the future CATEwill be applied for ACPs analysis of Chinesewater-cooled fusion reactor CFETR as soon as its design planis completed

Competing Interests

The authors declare that they have no competing interests

Acknowledgments

This work was sponsored jointly by ldquoNational Special Projectfor Magnetic Confined Nuclear Fusion Energyrdquo with Grantno 2014GB119000 and the ldquoFundamental Research Funds forthe Central Universitiesrdquo with Grant no 2014QN26 in ChinaThe authors would like to express their gratitude for thesupport

References

[1] A Rocher J L Bretelle and M Berger ldquoImpact of main radi-ological pollutants on contamination risks (ALARA) optimiza-tion of physico chemical environment and retention technicsduring operation and shutdownrdquo in Proceedings of the EuropeanWorkshop on Occupational Exposure Management at NPPs(ISOE rsquo04) Session 2 EDF Lyon France March 2004

[2] MRafiqueNMMirza SMMirza andM J Iqbal ldquoReviewofcomputer codes for modeling corrosion product transport andactivity build-up in light water reactorsrdquoNukleonika vol 55 no3 pp 263ndash269 2010

[3] L Li J Zhang W Song Y Fu X Xu and Y Chen ldquoCATE acode for activated corrosion products evaluation of water-cooled fusion reactorrdquo Fusion Engineering and Design vol 100pp 340ndash344 2015

[4] IAEA ldquoModelling of transport of radioactive substances inthe primary circuit of water-cooled reactorsrdquo IAEA-TECDOC-1672 2012

[5] C B Lee Modeling of Corrosion Product Transport in PWRPrimary Coolant Nuclear Engineering Department MIT 1990

[6] SNisan LD Pace and J-C Robin ldquoEvaluation of the activatedcorrosion products for the ITER heat transfer systemsrdquo inFusion Technology pp 1763ndash1766 Elsevier 1996

[7] P J Karditsas ldquoActivation product transport using TRACTORE estimation of an ITER cooling looprdquo Fusion Engineeringand Design vol 45 no 2 pp 169ndash185 1999

[8] R A Forrest J Kopecky and J-Ch Sublet ldquoTheEuropeanActi-vation File EAF-2007 neutron-induced cross section libraryrdquoUKAEA FUS 535 2007

[9] R A Forrest ldquoThe European Activation File EAF-2007 decaydata libraryrdquo UKAEA FUS 537 2007

[10] J F Briesmeister MCNP a general Monte Carlo N-particletransport code version 4C Los Alamos National LaboratoryLA-13709-M 2000

[11] H Iida V Khripunov and L Petrizzi ldquoNuclear analysis reportrdquoITER Document NAG-201-01-06-17-FDR 2001

[12] ICRP ldquoRecommendations of the international commission onradiological protection publication 60rdquo Annals of the ICRP 21Pergamon Press Oxford UK 1991

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Science and Technology of Nuclear Installations 3

nuclide119898 in the basemetal 119860119894is the contact area between

coolant and pipe m2 119896119898119894

is the mass transfer coefficient ofnuclide 119898 ms 119881 is the coolant volume in the cooling loopm3 119878119898119894

is the solubility of nuclide 119898 in the coolant close tothe pipe surface kgm3 120596 is the collection efficiency of filterand resin in chemical and volume control system (CVCS) 119876 is the volume flow rate of coolant into CVCS m3s 119881CVCSis the coolant volume in CVCS m31198731015840

119899119894

is the atom numberof nuclide 119899 in ACPs atoms and the superscript 1015840 meansnuclide 119899 is radioactive 120594

119896rarr119899is the decay branching ratio

from nuclide 119896 to nuclide 119899 120582119896is the decay rate constant of

nuclide 119896 sminus1 120590120574119898rarr119899

is the radiative capture cross section ofconverting nuclide119898 to nuclide 119899 cm2 120601 is the neutron flux119899(cm2sdots) 119860

119898is the molar mass of nuclide 119898 kgmol119873

119860is

the Avogadro number 120590119886119899

is the absorption cross section ofnuclide 119899 cm2 119877 is the ratio of coolant residence time in thezone of in-flux pipe to that in the whole cooling loop

In the above model some assumptions are made includ-ing the following (1) the concentration of nuclides in thecoolant keeps the same along the cooling loop which isbecause it takes the coolant only several seconds to circlearound the cooling loop which means the homogenizationeffect of nuclides in the coolant is dominant (2) the influenceof activation and decay on the mass of corrosion products isneglected because the quantity of ACPs is very limited com-pared to that of corrosion products (3) the concentration ofnuclides on the pipe surface is equal to its solubility in theadjacent coolant

To solve the above differential equations the fourth-orderRunge-Kutta method is used which is fast and can controlthe error well Moreover an algorithm of adaptive time stepis adopted in CATE for determining the time step reasonablyand the calculation speed can be improved

3 Verification of CATE Code through MITPWR Coolant Chemistry Loop (PCCL)

31 Description of MIT PCCL MIT PCCL was chosen hereto test the code CATE which is a small scale loop operatedunder constant coolant chemistry in the MIT reactor toclosely simulate the primary circuit of a typical PWR powerplant The main operation data of the loop is presented inTable 1 [5]

The basemetal of in-flux pipe is Zircaloy-4 and the corro-sion rate of thismaterial is nearly 0The basemetal of out-fluxpipe is mainly Inconel and the corrosion rate of it is 355119864 minus7 kgday

32 Results of ACPs Radioactivity in MIT PCCL The surveil-lance data from MIT PCCL for 42 days of full poweroperation was published and the radioactivity of ACPs in theloop is as shown in Table 2

Calculation results from CATE are close to the experi-mental results fromMIT whichmeans CATE is available andcredible on ACPs analysis of water-cooled nuclear reactorThe calculation results of mass of corrosion products in thein-flux pipe the coolant and the out-flux pipe are respec-tively 166119864 minus 6 kg 148119864 minus 9 kg and 396119864 minus 5 kg for CATE

Table 1 The operation data of MIT PCCL

Parameter ValueTemperature (∘C) 2739sim3156Mass flow rate (kgs) 684119864 minus 2

Flow velocity (ms) 283sim318Volume of coolant (m3) 384119864 minus 4

Surface area of in-flux pipe (m2) 260119864 minus 2

Surface area of out-flux pipe (m2) 970119864 minus 2

Mass flow rate in CVCS (kgs) 534119864 minus 5

Collection efficiency of CVCS 50Concentration of boron (ppm) 800Concentration of LiOH (ppm) 184Concentration of H

2

(cckg-H2

O) 25

Using these data we can calculate the specific radioactivity ofACPs in the in-flux pipe the coolant and the out-flux pipewhich is respectively 101119864 + 3GBqkg 233119864 + 2GBqkgand 689119864 + 1GBqkg for CATE The values are degressivewhich is consistent with the transport direction of ACPs inthe cooling loop and is rational

4 Calculation of ACPs Radioactivity in ITERBlanket Cooling Loop Using CATE Code

41 Description of ITER Blanket Cooling Loop The Interna-tionalThermonuclear Experimental Reactor (ITER) is underconstruction now In its design plan the primary coolingloops aremost water-cooled such as the blanket the divertorthe NB injector and the vacuum vessel So ITER can betreated as a representative of water-cooled fusion reactorChina is developing its own fusion reactor CFETR (ChinaFusion Engineering Test Reactor) whose design plan of heattransfer and radiation shielding partly refers to ITER soACPs analysis of ITER will benefit the work of source termcontrol for CFETR

The planned operation data of ITER blanket cooling loopis presented in Table 3 [6]

The base metal in in-flux pipe and out-flux pipe is thesame kind of stainless steel SS316 and the corrosion rate of itis as follows

SS316 CR = 20 times 10minus5 times 119905minus0614 kgsdotmminus2 sdotsminus1 (2)

The above equation and the element composition of SS316are quoted from [7] The relevant nuclear reaction data arequoted from the European Activation File EAF-2007 [8 9]and the cross section of activation reaction is collapsed from172 groups into 1 group using the corresponding neutronspectrum which is calculated with MCNP code [10] andbased on an ITER blanket module [11]The neutronics modeland neutron spectrum of the chosen blanket module can beseen in Figures 2 and 3

42 Results of ACPs Radioactivity in ITER Blanket CoolingLoop After 12 years of full power operation the mass ofcorrosion products and the radioactivity of ACPs in the loopare as shown in Table 4

4 Science and Technology of Nuclear Installations

Table 2 Radioactivity of ACPs in MIT PCCL after normal operation for 42 days

Data sourceRadioactivity of ACPs (GBq)

In-flux pipe Coolant Out-flux pipeCo-58 Co-60 Co-58 Co-60 Co-58 Co-60

MIT 179119864 minus 3 116119864 minus 4 365119864 minus 7 243119864 minus 8 271119864 minus 3 195119864 minus 4

CATE 151119864 minus 3 172119864 minus 4 309119864 minus 7 354119864 minus 8 244119864 minus 3 287119864 minus 4

Table 3 The planned operation data of ITER blanket cooling loop

Parameter ValueTemperature (∘C) 140sim174Mass flow rate (kgs) 851Average flow velocity (ms) 40Volume of coolant (m3) 3336Surface area of in-flux pipe (m2) 9914Surface area of out-flux pipe (m2) 13951Mass flow rate in CVCS (kgs) 255Collection efficiency of CVCS 50Concentration of H

2

(cckg-H2

O) 25

Figure 2 The neutronics model of the chosen blanket module(marked with a red ring)

From Table 4 we can see that mass of corrosion productsin the coolant is much lower than that on the pipe surfacewhich is due to the limitation of solubility of corrosion prod-ucts in the coolant It should be noticed that the corrosionproducts on the pipe surface include oxides and depositsTheoxides are compact and fixed while the deposits are loose andmobilizable and their mass is usually one order of magnitudelower than that of oxides So although the total mass ofcorrosion products on the pipe surface is tens of kg the actualmass of deposits is only in the magnitude of kg which isconsistent with the experience in safety analysis

We also can see that ACPs radioactivity on the pipesurface is much higher than that in the coolant which meansthe pipe surface is themain radioactive contamination regionand should be decontaminated regularly through waterchemistry method The specific radioactivity of ACPs in thein-flux pipe the coolant and the out-flux pipe is calculated as

Table 4 Calculation results of ACPs in ITER blanket cooling afternormal operation for 12 years from CATE

In-flux pipe Coolant Out-flux pipeMass of corrosionproducts (kg) 217119864 + 1 625119864 minus 3 266119864 + 1

Radioactivity of ACPs(GBq) 105119864 + 6 277119864 + 0 890119864 + 3

001 0

1 1 10 100

00101

110

100

Neutron energy (MeV)

1E minus 3

1E minus 4

1E minus 5

1E minus 6

1E minus 7

1E minus 8

1E minus 9

1E minus 10

1E

minus8

1E

minus7

1E

minus6

1E

minus5

1E

minus4

1E

minus3

Neu

tron

flux

(10

18

ncm

2middotsmiddot

MeV

)

Figure 3 The neutron spectrum of the chosen blanket module

484119864 + 4GBqkg 443119864 + 2GBqkg and 335119864 + 2GBqkgThe values are degressive which is consistent with the trans-port direction of ACPs in the cooling loop and is rational

From Table 5 we can see that the short-life nuclides(V-52 V-53 Cr-55 Mn-56 Co-58m Co-60m and Ni-57)are the major contributors to radioactivity in all the threeregions especially Mn-56 which alone contributes 6333of the radioactivity in the region of in-flux pipe and nearly80 of the radioactivity in the regions of coolant and out-fluxpipe That is very different from PWR in which the long-lifenuclides Co-58 and Co-60 are the main ACPs And when thereactor shuts down for a period of time (about several days)ACPs radioactivity will decrease obviously due to the rapiddecay of the short-life nuclides and then the right time forworkers to do maintenance comes

Science and Technology of Nuclear Installations 5

Table 5 The main ACPs after normal operation for 12 years fromCATE

Nuclide Half-life Radioactivity (GBq)In-flux pipe Coolant Out-flux pipe

V-49 330 d 404119864 + 2 507119864 minus 7 414119864 minus 6

V-52 374m 103119864 + 4 781119864 minus 3 269119864 + 0

V-53 16m 595119864 + 2 427119864 minus 4 668119864 minus 2

Cr-51 277 d 235119864 + 5 321119864 minus 1 848119864 + 2

Cr-55 350m 378119864 + 3 285119864 minus 3 933119864 minus 1

Mn-56 258 h 659119864 + 5 215119864 + 0 707119864 + 3

Fe-55 274 y 404119864 + 4 209119864 minus 1 767119864 + 2

Fe-59 445 d 479119864 + 3 108119864 minus 2 347119864 + 1

Co-57 272 d 167119864 + 4 198119864 minus 2 608119864 + 1

Co-58m 904 h 272119864 + 4 131119864 minus 2 579119864 minus 1

Co-58 709 d 379119864 + 4 367119864 minus 2 944119864 + 1

Co-60m 105m 294119864 + 3 134119864 minus 3 236119864 minus 3

Co-60 527 y 369119864 + 2 904119864 minus 4 322119864 + 0

Ni-57 356 h 157119864 + 3 624119864 minus 4 551119864 minus 2

Ni-63 100 y 571119864 + 2 741119864 minus 4 241119864 + 0

Total 104119864 + 6 277119864 + 0 889119864 + 3

Short-lived 706119864 + 5 217119864 + 0 707119864 + 3

5 Calculation of Dose Rate Caused by ACPsUsing ARShield Code

TheARShield code developed by NCEPU of China is appliedto dose rate calculation which is a new version of the pointkernel integration code QAD-CG developed by Los AlamosNational Laboratory ARShield breaks some restrictions ofQAD-CG such as complicatedmodeling complicated sourcesetting 3D fine mesh results statistics and large-scale com-puting efficiency and is proved to be reliable and efficient ondose rate calculation

The density of each radionuclide at chosen regionscalculated by CATE is introduced into ARShield and thenconverted to dose rate using point kernel integrationmethodwhich is as follows

119863 (119903)

= int

119881

119870119904 (119903

1015840

) 119861 (120583

1003816

1003816

1003816

1003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

1003816

119864) exp (minus120583 1003816100381610038161003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

1003816

) 119889119881

2120587

1003816

1003816

1003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

2

(3)

where 119903 is the point at which gamma dose rate is to becalculated 1199031015840 is the location of source in volume 119881 119881 is thevolume of source region 120583 is the total attenuation coefficientat energy 119864 |119903 rarr 119903

1015840

| is the distance between source point andpoint at which gamma intensity is to be calculated 119870 is theflux-to-dose conversion factor 119861 is the dose buildup factor

The geometry of hot leg pipe in ITER blanket coolingloop is adopted to represent the region of out-flux pipe whichhas an internal diameter of 0527m thickness of 002mand length of 33m The dose rate around the hot leg pipecalculated by ARShield is shown in Figure 4

We can see that after shutting down the reactor for 8 daysthe dose rate is much lower than that of normal operation for

0

1

2

3

4

5

6

7

8

9

10

Dos

e rat

e (m

Svh

)

Distance (cm)

minus1

minus200 0 200 400 600 800 1000 1200 1400 1600

Normal operationShutdown for 8 days

Figure 4 Variation of dose rate with the distance from the outersurface of the hot leg pipe

12 years which is because that contribution from the short-life ACPs decreases almost to zero after shutdown for 8 daysThe typical values of dose rate are as follows during normaloperation the dose rate at the outer surface of the pipe is880mSvh and the dose rate at a distance 1m away from theouter surface of the pipe is 252mSvh after shutdown for 8days the values drop to 105mSvh (contact) and 031mSvh(1m away) When these values are compared to the annualpermissible worker dose rate of 20mSvyear recommendedby the ICRP [12] the contact dose rate value after shutdownfor 8 days would allow approximately 19 h exposure per year

6 Conclusions

In this paper a three-node transport model was introducedinto theACPs source termcodeCATEmakingCATE capableof theoretically simulating the variation and the distributionof ACPs in a water-cooled reactor and suitable formore oper-ating conditions MIT PCCL was chosen to test the new ver-sion ofCATE and the calculation results fromCATEare closeto the experimental results from MIT which means CATEis available and credible on ACPs analysis of water-cooledreactor Then the radioactivity and composition of ACPs inITERblanket cooling loopwere analyzed usingCATE and theresults showed that the major contributors are the short-lifenuclides (V-52 V-53 Cr-55 Mn-56 Co-58m Co-60m andNi-57) for ITER especiallyMn-56That is very different fromPWR in which the long-life nuclides Co-58 and Co-60 arethe main ACPs At last the dose rate around ITER blanketcooling loop caused by ACPs was calculated through cou-pling the code CATE with a point kernel integration codeARShield The results showed that after shutting down thereactor only for 8 days the dose rate can decrease nearly oneorder of magnitude compared to that of normal operationwhich is caused by the rapid decay of the short-life ACPs In

6 Science and Technology of Nuclear Installations

the future CATEwill be applied for ACPs analysis of Chinesewater-cooled fusion reactor CFETR as soon as its design planis completed

Competing Interests

The authors declare that they have no competing interests

Acknowledgments

This work was sponsored jointly by ldquoNational Special Projectfor Magnetic Confined Nuclear Fusion Energyrdquo with Grantno 2014GB119000 and the ldquoFundamental Research Funds forthe Central Universitiesrdquo with Grant no 2014QN26 in ChinaThe authors would like to express their gratitude for thesupport

References

[1] A Rocher J L Bretelle and M Berger ldquoImpact of main radi-ological pollutants on contamination risks (ALARA) optimiza-tion of physico chemical environment and retention technicsduring operation and shutdownrdquo in Proceedings of the EuropeanWorkshop on Occupational Exposure Management at NPPs(ISOE rsquo04) Session 2 EDF Lyon France March 2004

[2] MRafiqueNMMirza SMMirza andM J Iqbal ldquoReviewofcomputer codes for modeling corrosion product transport andactivity build-up in light water reactorsrdquoNukleonika vol 55 no3 pp 263ndash269 2010

[3] L Li J Zhang W Song Y Fu X Xu and Y Chen ldquoCATE acode for activated corrosion products evaluation of water-cooled fusion reactorrdquo Fusion Engineering and Design vol 100pp 340ndash344 2015

[4] IAEA ldquoModelling of transport of radioactive substances inthe primary circuit of water-cooled reactorsrdquo IAEA-TECDOC-1672 2012

[5] C B Lee Modeling of Corrosion Product Transport in PWRPrimary Coolant Nuclear Engineering Department MIT 1990

[6] SNisan LD Pace and J-C Robin ldquoEvaluation of the activatedcorrosion products for the ITER heat transfer systemsrdquo inFusion Technology pp 1763ndash1766 Elsevier 1996

[7] P J Karditsas ldquoActivation product transport using TRACTORE estimation of an ITER cooling looprdquo Fusion Engineeringand Design vol 45 no 2 pp 169ndash185 1999

[8] R A Forrest J Kopecky and J-Ch Sublet ldquoTheEuropeanActi-vation File EAF-2007 neutron-induced cross section libraryrdquoUKAEA FUS 535 2007

[9] R A Forrest ldquoThe European Activation File EAF-2007 decaydata libraryrdquo UKAEA FUS 537 2007

[10] J F Briesmeister MCNP a general Monte Carlo N-particletransport code version 4C Los Alamos National LaboratoryLA-13709-M 2000

[11] H Iida V Khripunov and L Petrizzi ldquoNuclear analysis reportrdquoITER Document NAG-201-01-06-17-FDR 2001

[12] ICRP ldquoRecommendations of the international commission onradiological protection publication 60rdquo Annals of the ICRP 21Pergamon Press Oxford UK 1991

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

4 Science and Technology of Nuclear Installations

Table 2 Radioactivity of ACPs in MIT PCCL after normal operation for 42 days

Data sourceRadioactivity of ACPs (GBq)

In-flux pipe Coolant Out-flux pipeCo-58 Co-60 Co-58 Co-60 Co-58 Co-60

MIT 179119864 minus 3 116119864 minus 4 365119864 minus 7 243119864 minus 8 271119864 minus 3 195119864 minus 4

CATE 151119864 minus 3 172119864 minus 4 309119864 minus 7 354119864 minus 8 244119864 minus 3 287119864 minus 4

Table 3 The planned operation data of ITER blanket cooling loop

Parameter ValueTemperature (∘C) 140sim174Mass flow rate (kgs) 851Average flow velocity (ms) 40Volume of coolant (m3) 3336Surface area of in-flux pipe (m2) 9914Surface area of out-flux pipe (m2) 13951Mass flow rate in CVCS (kgs) 255Collection efficiency of CVCS 50Concentration of H

2

(cckg-H2

O) 25

Figure 2 The neutronics model of the chosen blanket module(marked with a red ring)

From Table 4 we can see that mass of corrosion productsin the coolant is much lower than that on the pipe surfacewhich is due to the limitation of solubility of corrosion prod-ucts in the coolant It should be noticed that the corrosionproducts on the pipe surface include oxides and depositsTheoxides are compact and fixed while the deposits are loose andmobilizable and their mass is usually one order of magnitudelower than that of oxides So although the total mass ofcorrosion products on the pipe surface is tens of kg the actualmass of deposits is only in the magnitude of kg which isconsistent with the experience in safety analysis

We also can see that ACPs radioactivity on the pipesurface is much higher than that in the coolant which meansthe pipe surface is themain radioactive contamination regionand should be decontaminated regularly through waterchemistry method The specific radioactivity of ACPs in thein-flux pipe the coolant and the out-flux pipe is calculated as

Table 4 Calculation results of ACPs in ITER blanket cooling afternormal operation for 12 years from CATE

In-flux pipe Coolant Out-flux pipeMass of corrosionproducts (kg) 217119864 + 1 625119864 minus 3 266119864 + 1

Radioactivity of ACPs(GBq) 105119864 + 6 277119864 + 0 890119864 + 3

001 0

1 1 10 100

00101

110

100

Neutron energy (MeV)

1E minus 3

1E minus 4

1E minus 5

1E minus 6

1E minus 7

1E minus 8

1E minus 9

1E minus 10

1E

minus8

1E

minus7

1E

minus6

1E

minus5

1E

minus4

1E

minus3

Neu

tron

flux

(10

18

ncm

2middotsmiddot

MeV

)

Figure 3 The neutron spectrum of the chosen blanket module

484119864 + 4GBqkg 443119864 + 2GBqkg and 335119864 + 2GBqkgThe values are degressive which is consistent with the trans-port direction of ACPs in the cooling loop and is rational

From Table 5 we can see that the short-life nuclides(V-52 V-53 Cr-55 Mn-56 Co-58m Co-60m and Ni-57)are the major contributors to radioactivity in all the threeregions especially Mn-56 which alone contributes 6333of the radioactivity in the region of in-flux pipe and nearly80 of the radioactivity in the regions of coolant and out-fluxpipe That is very different from PWR in which the long-lifenuclides Co-58 and Co-60 are the main ACPs And when thereactor shuts down for a period of time (about several days)ACPs radioactivity will decrease obviously due to the rapiddecay of the short-life nuclides and then the right time forworkers to do maintenance comes

Science and Technology of Nuclear Installations 5

Table 5 The main ACPs after normal operation for 12 years fromCATE

Nuclide Half-life Radioactivity (GBq)In-flux pipe Coolant Out-flux pipe

V-49 330 d 404119864 + 2 507119864 minus 7 414119864 minus 6

V-52 374m 103119864 + 4 781119864 minus 3 269119864 + 0

V-53 16m 595119864 + 2 427119864 minus 4 668119864 minus 2

Cr-51 277 d 235119864 + 5 321119864 minus 1 848119864 + 2

Cr-55 350m 378119864 + 3 285119864 minus 3 933119864 minus 1

Mn-56 258 h 659119864 + 5 215119864 + 0 707119864 + 3

Fe-55 274 y 404119864 + 4 209119864 minus 1 767119864 + 2

Fe-59 445 d 479119864 + 3 108119864 minus 2 347119864 + 1

Co-57 272 d 167119864 + 4 198119864 minus 2 608119864 + 1

Co-58m 904 h 272119864 + 4 131119864 minus 2 579119864 minus 1

Co-58 709 d 379119864 + 4 367119864 minus 2 944119864 + 1

Co-60m 105m 294119864 + 3 134119864 minus 3 236119864 minus 3

Co-60 527 y 369119864 + 2 904119864 minus 4 322119864 + 0

Ni-57 356 h 157119864 + 3 624119864 minus 4 551119864 minus 2

Ni-63 100 y 571119864 + 2 741119864 minus 4 241119864 + 0

Total 104119864 + 6 277119864 + 0 889119864 + 3

Short-lived 706119864 + 5 217119864 + 0 707119864 + 3

5 Calculation of Dose Rate Caused by ACPsUsing ARShield Code

TheARShield code developed by NCEPU of China is appliedto dose rate calculation which is a new version of the pointkernel integration code QAD-CG developed by Los AlamosNational Laboratory ARShield breaks some restrictions ofQAD-CG such as complicatedmodeling complicated sourcesetting 3D fine mesh results statistics and large-scale com-puting efficiency and is proved to be reliable and efficient ondose rate calculation

The density of each radionuclide at chosen regionscalculated by CATE is introduced into ARShield and thenconverted to dose rate using point kernel integrationmethodwhich is as follows

119863 (119903)

= int

119881

119870119904 (119903

1015840

) 119861 (120583

1003816

1003816

1003816

1003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

1003816

119864) exp (minus120583 1003816100381610038161003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

1003816

) 119889119881

2120587

1003816

1003816

1003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

2

(3)

where 119903 is the point at which gamma dose rate is to becalculated 1199031015840 is the location of source in volume 119881 119881 is thevolume of source region 120583 is the total attenuation coefficientat energy 119864 |119903 rarr 119903

1015840

| is the distance between source point andpoint at which gamma intensity is to be calculated 119870 is theflux-to-dose conversion factor 119861 is the dose buildup factor

The geometry of hot leg pipe in ITER blanket coolingloop is adopted to represent the region of out-flux pipe whichhas an internal diameter of 0527m thickness of 002mand length of 33m The dose rate around the hot leg pipecalculated by ARShield is shown in Figure 4

We can see that after shutting down the reactor for 8 daysthe dose rate is much lower than that of normal operation for

0

1

2

3

4

5

6

7

8

9

10

Dos

e rat

e (m

Svh

)

Distance (cm)

minus1

minus200 0 200 400 600 800 1000 1200 1400 1600

Normal operationShutdown for 8 days

Figure 4 Variation of dose rate with the distance from the outersurface of the hot leg pipe

12 years which is because that contribution from the short-life ACPs decreases almost to zero after shutdown for 8 daysThe typical values of dose rate are as follows during normaloperation the dose rate at the outer surface of the pipe is880mSvh and the dose rate at a distance 1m away from theouter surface of the pipe is 252mSvh after shutdown for 8days the values drop to 105mSvh (contact) and 031mSvh(1m away) When these values are compared to the annualpermissible worker dose rate of 20mSvyear recommendedby the ICRP [12] the contact dose rate value after shutdownfor 8 days would allow approximately 19 h exposure per year

6 Conclusions

In this paper a three-node transport model was introducedinto theACPs source termcodeCATEmakingCATE capableof theoretically simulating the variation and the distributionof ACPs in a water-cooled reactor and suitable formore oper-ating conditions MIT PCCL was chosen to test the new ver-sion ofCATE and the calculation results fromCATEare closeto the experimental results from MIT which means CATEis available and credible on ACPs analysis of water-cooledreactor Then the radioactivity and composition of ACPs inITERblanket cooling loopwere analyzed usingCATE and theresults showed that the major contributors are the short-lifenuclides (V-52 V-53 Cr-55 Mn-56 Co-58m Co-60m andNi-57) for ITER especiallyMn-56That is very different fromPWR in which the long-life nuclides Co-58 and Co-60 arethe main ACPs At last the dose rate around ITER blanketcooling loop caused by ACPs was calculated through cou-pling the code CATE with a point kernel integration codeARShield The results showed that after shutting down thereactor only for 8 days the dose rate can decrease nearly oneorder of magnitude compared to that of normal operationwhich is caused by the rapid decay of the short-life ACPs In

6 Science and Technology of Nuclear Installations

the future CATEwill be applied for ACPs analysis of Chinesewater-cooled fusion reactor CFETR as soon as its design planis completed

Competing Interests

The authors declare that they have no competing interests

Acknowledgments

This work was sponsored jointly by ldquoNational Special Projectfor Magnetic Confined Nuclear Fusion Energyrdquo with Grantno 2014GB119000 and the ldquoFundamental Research Funds forthe Central Universitiesrdquo with Grant no 2014QN26 in ChinaThe authors would like to express their gratitude for thesupport

References

[1] A Rocher J L Bretelle and M Berger ldquoImpact of main radi-ological pollutants on contamination risks (ALARA) optimiza-tion of physico chemical environment and retention technicsduring operation and shutdownrdquo in Proceedings of the EuropeanWorkshop on Occupational Exposure Management at NPPs(ISOE rsquo04) Session 2 EDF Lyon France March 2004

[2] MRafiqueNMMirza SMMirza andM J Iqbal ldquoReviewofcomputer codes for modeling corrosion product transport andactivity build-up in light water reactorsrdquoNukleonika vol 55 no3 pp 263ndash269 2010

[3] L Li J Zhang W Song Y Fu X Xu and Y Chen ldquoCATE acode for activated corrosion products evaluation of water-cooled fusion reactorrdquo Fusion Engineering and Design vol 100pp 340ndash344 2015

[4] IAEA ldquoModelling of transport of radioactive substances inthe primary circuit of water-cooled reactorsrdquo IAEA-TECDOC-1672 2012

[5] C B Lee Modeling of Corrosion Product Transport in PWRPrimary Coolant Nuclear Engineering Department MIT 1990

[6] SNisan LD Pace and J-C Robin ldquoEvaluation of the activatedcorrosion products for the ITER heat transfer systemsrdquo inFusion Technology pp 1763ndash1766 Elsevier 1996

[7] P J Karditsas ldquoActivation product transport using TRACTORE estimation of an ITER cooling looprdquo Fusion Engineeringand Design vol 45 no 2 pp 169ndash185 1999

[8] R A Forrest J Kopecky and J-Ch Sublet ldquoTheEuropeanActi-vation File EAF-2007 neutron-induced cross section libraryrdquoUKAEA FUS 535 2007

[9] R A Forrest ldquoThe European Activation File EAF-2007 decaydata libraryrdquo UKAEA FUS 537 2007

[10] J F Briesmeister MCNP a general Monte Carlo N-particletransport code version 4C Los Alamos National LaboratoryLA-13709-M 2000

[11] H Iida V Khripunov and L Petrizzi ldquoNuclear analysis reportrdquoITER Document NAG-201-01-06-17-FDR 2001

[12] ICRP ldquoRecommendations of the international commission onradiological protection publication 60rdquo Annals of the ICRP 21Pergamon Press Oxford UK 1991

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

Science and Technology of Nuclear Installations 5

Table 5 The main ACPs after normal operation for 12 years fromCATE

Nuclide Half-life Radioactivity (GBq)In-flux pipe Coolant Out-flux pipe

V-49 330 d 404119864 + 2 507119864 minus 7 414119864 minus 6

V-52 374m 103119864 + 4 781119864 minus 3 269119864 + 0

V-53 16m 595119864 + 2 427119864 minus 4 668119864 minus 2

Cr-51 277 d 235119864 + 5 321119864 minus 1 848119864 + 2

Cr-55 350m 378119864 + 3 285119864 minus 3 933119864 minus 1

Mn-56 258 h 659119864 + 5 215119864 + 0 707119864 + 3

Fe-55 274 y 404119864 + 4 209119864 minus 1 767119864 + 2

Fe-59 445 d 479119864 + 3 108119864 minus 2 347119864 + 1

Co-57 272 d 167119864 + 4 198119864 minus 2 608119864 + 1

Co-58m 904 h 272119864 + 4 131119864 minus 2 579119864 minus 1

Co-58 709 d 379119864 + 4 367119864 minus 2 944119864 + 1

Co-60m 105m 294119864 + 3 134119864 minus 3 236119864 minus 3

Co-60 527 y 369119864 + 2 904119864 minus 4 322119864 + 0

Ni-57 356 h 157119864 + 3 624119864 minus 4 551119864 minus 2

Ni-63 100 y 571119864 + 2 741119864 minus 4 241119864 + 0

Total 104119864 + 6 277119864 + 0 889119864 + 3

Short-lived 706119864 + 5 217119864 + 0 707119864 + 3

5 Calculation of Dose Rate Caused by ACPsUsing ARShield Code

TheARShield code developed by NCEPU of China is appliedto dose rate calculation which is a new version of the pointkernel integration code QAD-CG developed by Los AlamosNational Laboratory ARShield breaks some restrictions ofQAD-CG such as complicatedmodeling complicated sourcesetting 3D fine mesh results statistics and large-scale com-puting efficiency and is proved to be reliable and efficient ondose rate calculation

The density of each radionuclide at chosen regionscalculated by CATE is introduced into ARShield and thenconverted to dose rate using point kernel integrationmethodwhich is as follows

119863 (119903)

= int

119881

119870119904 (119903

1015840

) 119861 (120583

1003816

1003816

1003816

1003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

1003816

119864) exp (minus120583 1003816100381610038161003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

1003816

) 119889119881

2120587

1003816

1003816

1003816

1003816

119903 997888rarr 119903

10158401003816

1003816

1003816

1003816

2

(3)

where 119903 is the point at which gamma dose rate is to becalculated 1199031015840 is the location of source in volume 119881 119881 is thevolume of source region 120583 is the total attenuation coefficientat energy 119864 |119903 rarr 119903

1015840

| is the distance between source point andpoint at which gamma intensity is to be calculated 119870 is theflux-to-dose conversion factor 119861 is the dose buildup factor

The geometry of hot leg pipe in ITER blanket coolingloop is adopted to represent the region of out-flux pipe whichhas an internal diameter of 0527m thickness of 002mand length of 33m The dose rate around the hot leg pipecalculated by ARShield is shown in Figure 4

We can see that after shutting down the reactor for 8 daysthe dose rate is much lower than that of normal operation for

0

1

2

3

4

5

6

7

8

9

10

Dos

e rat

e (m

Svh

)

Distance (cm)

minus1

minus200 0 200 400 600 800 1000 1200 1400 1600

Normal operationShutdown for 8 days

Figure 4 Variation of dose rate with the distance from the outersurface of the hot leg pipe

12 years which is because that contribution from the short-life ACPs decreases almost to zero after shutdown for 8 daysThe typical values of dose rate are as follows during normaloperation the dose rate at the outer surface of the pipe is880mSvh and the dose rate at a distance 1m away from theouter surface of the pipe is 252mSvh after shutdown for 8days the values drop to 105mSvh (contact) and 031mSvh(1m away) When these values are compared to the annualpermissible worker dose rate of 20mSvyear recommendedby the ICRP [12] the contact dose rate value after shutdownfor 8 days would allow approximately 19 h exposure per year

6 Conclusions

In this paper a three-node transport model was introducedinto theACPs source termcodeCATEmakingCATE capableof theoretically simulating the variation and the distributionof ACPs in a water-cooled reactor and suitable formore oper-ating conditions MIT PCCL was chosen to test the new ver-sion ofCATE and the calculation results fromCATEare closeto the experimental results from MIT which means CATEis available and credible on ACPs analysis of water-cooledreactor Then the radioactivity and composition of ACPs inITERblanket cooling loopwere analyzed usingCATE and theresults showed that the major contributors are the short-lifenuclides (V-52 V-53 Cr-55 Mn-56 Co-58m Co-60m andNi-57) for ITER especiallyMn-56That is very different fromPWR in which the long-life nuclides Co-58 and Co-60 arethe main ACPs At last the dose rate around ITER blanketcooling loop caused by ACPs was calculated through cou-pling the code CATE with a point kernel integration codeARShield The results showed that after shutting down thereactor only for 8 days the dose rate can decrease nearly oneorder of magnitude compared to that of normal operationwhich is caused by the rapid decay of the short-life ACPs In

6 Science and Technology of Nuclear Installations

the future CATEwill be applied for ACPs analysis of Chinesewater-cooled fusion reactor CFETR as soon as its design planis completed

Competing Interests

The authors declare that they have no competing interests

Acknowledgments

This work was sponsored jointly by ldquoNational Special Projectfor Magnetic Confined Nuclear Fusion Energyrdquo with Grantno 2014GB119000 and the ldquoFundamental Research Funds forthe Central Universitiesrdquo with Grant no 2014QN26 in ChinaThe authors would like to express their gratitude for thesupport

References

[1] A Rocher J L Bretelle and M Berger ldquoImpact of main radi-ological pollutants on contamination risks (ALARA) optimiza-tion of physico chemical environment and retention technicsduring operation and shutdownrdquo in Proceedings of the EuropeanWorkshop on Occupational Exposure Management at NPPs(ISOE rsquo04) Session 2 EDF Lyon France March 2004

[2] MRafiqueNMMirza SMMirza andM J Iqbal ldquoReviewofcomputer codes for modeling corrosion product transport andactivity build-up in light water reactorsrdquoNukleonika vol 55 no3 pp 263ndash269 2010

[3] L Li J Zhang W Song Y Fu X Xu and Y Chen ldquoCATE acode for activated corrosion products evaluation of water-cooled fusion reactorrdquo Fusion Engineering and Design vol 100pp 340ndash344 2015

[4] IAEA ldquoModelling of transport of radioactive substances inthe primary circuit of water-cooled reactorsrdquo IAEA-TECDOC-1672 2012

[5] C B Lee Modeling of Corrosion Product Transport in PWRPrimary Coolant Nuclear Engineering Department MIT 1990

[6] SNisan LD Pace and J-C Robin ldquoEvaluation of the activatedcorrosion products for the ITER heat transfer systemsrdquo inFusion Technology pp 1763ndash1766 Elsevier 1996

[7] P J Karditsas ldquoActivation product transport using TRACTORE estimation of an ITER cooling looprdquo Fusion Engineeringand Design vol 45 no 2 pp 169ndash185 1999

[8] R A Forrest J Kopecky and J-Ch Sublet ldquoTheEuropeanActi-vation File EAF-2007 neutron-induced cross section libraryrdquoUKAEA FUS 535 2007

[9] R A Forrest ldquoThe European Activation File EAF-2007 decaydata libraryrdquo UKAEA FUS 537 2007

[10] J F Briesmeister MCNP a general Monte Carlo N-particletransport code version 4C Los Alamos National LaboratoryLA-13709-M 2000

[11] H Iida V Khripunov and L Petrizzi ldquoNuclear analysis reportrdquoITER Document NAG-201-01-06-17-FDR 2001

[12] ICRP ldquoRecommendations of the international commission onradiological protection publication 60rdquo Annals of the ICRP 21Pergamon Press Oxford UK 1991

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

6 Science and Technology of Nuclear Installations

the future CATEwill be applied for ACPs analysis of Chinesewater-cooled fusion reactor CFETR as soon as its design planis completed

Competing Interests

The authors declare that they have no competing interests

Acknowledgments

This work was sponsored jointly by ldquoNational Special Projectfor Magnetic Confined Nuclear Fusion Energyrdquo with Grantno 2014GB119000 and the ldquoFundamental Research Funds forthe Central Universitiesrdquo with Grant no 2014QN26 in ChinaThe authors would like to express their gratitude for thesupport

References

[1] A Rocher J L Bretelle and M Berger ldquoImpact of main radi-ological pollutants on contamination risks (ALARA) optimiza-tion of physico chemical environment and retention technicsduring operation and shutdownrdquo in Proceedings of the EuropeanWorkshop on Occupational Exposure Management at NPPs(ISOE rsquo04) Session 2 EDF Lyon France March 2004

[2] MRafiqueNMMirza SMMirza andM J Iqbal ldquoReviewofcomputer codes for modeling corrosion product transport andactivity build-up in light water reactorsrdquoNukleonika vol 55 no3 pp 263ndash269 2010

[3] L Li J Zhang W Song Y Fu X Xu and Y Chen ldquoCATE acode for activated corrosion products evaluation of water-cooled fusion reactorrdquo Fusion Engineering and Design vol 100pp 340ndash344 2015

[4] IAEA ldquoModelling of transport of radioactive substances inthe primary circuit of water-cooled reactorsrdquo IAEA-TECDOC-1672 2012

[5] C B Lee Modeling of Corrosion Product Transport in PWRPrimary Coolant Nuclear Engineering Department MIT 1990

[6] SNisan LD Pace and J-C Robin ldquoEvaluation of the activatedcorrosion products for the ITER heat transfer systemsrdquo inFusion Technology pp 1763ndash1766 Elsevier 1996

[7] P J Karditsas ldquoActivation product transport using TRACTORE estimation of an ITER cooling looprdquo Fusion Engineeringand Design vol 45 no 2 pp 169ndash185 1999

[8] R A Forrest J Kopecky and J-Ch Sublet ldquoTheEuropeanActi-vation File EAF-2007 neutron-induced cross section libraryrdquoUKAEA FUS 535 2007

[9] R A Forrest ldquoThe European Activation File EAF-2007 decaydata libraryrdquo UKAEA FUS 537 2007

[10] J F Briesmeister MCNP a general Monte Carlo N-particletransport code version 4C Los Alamos National LaboratoryLA-13709-M 2000

[11] H Iida V Khripunov and L Petrizzi ldquoNuclear analysis reportrdquoITER Document NAG-201-01-06-17-FDR 2001

[12] ICRP ldquoRecommendations of the international commission onradiological protection publication 60rdquo Annals of the ICRP 21Pergamon Press Oxford UK 1991

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014