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A consideration to reduce the potential hazard of I-129 in radioactive wastes
Progress in NuclearEnergy,Vol. 32, No. 314,pp. 525-530.1998 0 1997Publishedby Elsevier Science Ltd
Pergamon
Printed in Great Britain 0149-1970/98 $19.00 + 0.00
PII: SO149-1970(97)00063-2
A CONSIDERATION
TO REDUCE THE POTENTIAL
OF I-129 IN RADIOACTIVE
H.NAKAMURA
Y.INAGAKI
MSENOO
HAZARD
WASTES
S.TASHIRO
Radioactive Waste Management Center No.15.Mori Bldg. 2-8-10, Tranomon, Minato-ku Tokyo, Japan, 105
ABSTRACT Transmutation of I2 ’ I to ’3’ I by (n, 7 ) reaction and isotope dilution with stable iodine were discussed. The transmutation in LWR is calculated by supposing that targets irradiated for 25 years are substituted with new targets. The result showed that initial amount of ’’’ I will be reduced to 11.5%. In order to lower dose equivalent of general public than O.lmSv/y, 400 times isotope dilution of 1” I is required. The feasibility of the isotope dilution was discussed at dissolution process of spent fuel, conditioning process of the waste and disposal conditions. 0 1997 Published by Elsevier Science Ltd
INTRODUCTION The potential hazard of ’’’ I on mankind from disposal of reprocessing plant wastes has been estimated relatively high. This is due to its high mobility in certain geological environment. It can migrate easily away from underground repositories and, hence, eventually reach the biosphere. The environmental condition including geosphere can hardly be estimated for the long term comparable to the half life of ’” I. Therefore, reduction of ’* ’ I potential hazard would be desirable. One possible method may be reduction of 1’’ I as a source by transmuting it to 13’ I by (n, 7 1 reaction. As Iodine is eventually separated and recovered in the dissolver off-gas treatment system in reprocessing plant, it is possible to prepare the target material for the transmutation of “’ I by changing the chemical form without any more separations. R.J.M.Koning and coworkers irradiated PbI 2 , CeI 3 , NaI and others with thermal neutrons with a flux of 1.5X10 ’4 cm _ ’ s _ ’ . The results show that they are stable enough against the irradiation condition[l]. M.Hugon and coworkers calculated the transmutation rate of ’” I in core region of a fast reactor and showed that 22kg of lZ9 I could be transmuted with a transmutation half-life of around 44 years[2]. Isotope dilution with stable iodine is also a possible method to reduce maximum potential dose equivalent. The iodine content in a thyroid is limited. Therefore the maximum potential dose equivalent is determined by the specific activity of iodine. Once the required isotope dilution is practiced, the potential dose equivalent after the dilution will be determined by the 525
H. Nakamuraet al.
526
specific activity for any ingestion scenario. The sea water contains relatively large amo_Jnt of iodine. If proper siting for underground disposal is carried out, the efficient isotope dilution will be possible. Feasibility of the above two methods was discussed on supposing simple conditions.
1.
PRODUCTION
OF IODINE IN SPENT FUEL
Production rate of iodine in LWR and the specific activity was estimated by using ORIGEN and shown in Table 1. Table 1 Production of iodine and it specific radioactivity in spent fuel
Burn up Production rate of iodine
45
GWD/t
3.07x10 2 g/t 15X10 ’ Bq/t 503x10 ’ Bqlg Specific activity of 1’’ I Uranium needed for 1 GW LWR 20 t/Y Production rate of iodine for years 6 kg/y
As most all1 from the fuel dissolution process is collected on the absorber of the off-gas system, the absorbed I should be the only one target for the transmutation.
2.
TRANSMUTATION
RATE OF ’’’ I
The transmutation half life of I ’’ I by 1” I(n, 7 ) ’ 3 ’ I reaction with thermal neutron is shown as follows. N = N o exp(- CJ4 t ) = N o exp(- 0.0864t) Where N o : number of ’* ’ I nuclide in target material N : number of ’’’ I nuclide in target after the irradiiation for t cr : cross section for thermal neutron (27 b: 2.7x10 _ ’3 cm ’ ) @ : neutron flux( 1x10 ’4 cm _ * s _ ’ in core region of 1 GW LWR) t : irradiation time( 1 y: 365x24~60~60 = 3.2x10 7 s )
’’’ I in the target after one year irradiation is as follows. The reduction ratio of N 1ye a r /N o = exp(-0.0864X1) = 0.9. The transmutation rate par evry one year is about 10 %. For one practical system, it is supposed that a target is irradiated for 25 years. and substituted with new one. The reduction ratio of ’” I after irradiation of 25 years is calculated as follows. N/N o = exp(-0.0864X25) = 0.115 The irradiation capacity is necessary for 150 kg iodine which is produced for 25 years The
Potential hazard of I-129 in radioactivewastes
ratio of the total amount of ’’’ I in a reactor(N II ) to annual production(N o ) is calculated as follows. N, Ma = exp(-0.0864t)dt = (1 - 0.115)/0.0846 = 10.5 If this system will be operated continuously for ever, initial amount of ’’’ I will be reduced to 11.5 %. When it is ceased,however, the amount of ’” I corresponding to the production in the LWR for about 10.5 years will remain in the reactor finally. When irradiation facilities are installed out of core region, larger capacity and longer irradiation period will be required.
3 .
ISOTOPE DILUTION IN FUEL REPROCESSING
1) Calculation of dose equivalent on supposing all iodine in environment being substituted by iodine with the same specific activity as iodine in spent fuel. Effective dose(whole body) coefficient of ’* ’ I ingestion is 1.1x10 _ 7 Sv Bq -’ based on ICRP Pub.68[3]. Intake of iodine by general public is 0.2 mg/d based on ICRP Pub.23[4]. Supposing that all iodine in environment are substituted by the iodine of the same specific activity as that in spent fuel, ingestion of ’’@I in a year is calculated to be 3.6x10 ’ Bq and 40 mSv/y by using the above values and the specific activity shown in Table 1. In order to lower the dose equivalent of general public less than O.lmSv/y, 400 times isotope dilution of ’* !aI is required. Once the dilution is carried out, dose equivalent with any ingestion scenario will not be over 10% of 1 mSv/y(whole body) recommended by ICRP as dose limit for general public. 2) Isotope dilution process Supposing 400 times isotope dilution is performed during fuel reprocessing, an addition of about 2.4 t of iodine is required for processing 20 t of spent fuel. One of possible methods is the addition to the dissolver and another is the addition to matrix of solid waste of the iodine absorber. The addition to the dissolver may result in efficient isotope dilution and higher recovery of
’” I by carrier effect. In this case the amount of iodine contained in off-gas may be too large for solid absorber and then absorbing process by alkaline solution should be used instead. The unfavorable effect of iodine addition on downstream separation steps must be examined. Since this method requires change of the plant design and re-examination of the process, the application will not be expected without new plant construction. The conventional process has been completed by long term research and development. No change of the main process should be kept and the isotope dilution process should be added to the waste management steps. Isotope dilution on waste conditioning process may depend on waste forms. Following conditioning process are considered. 0 Iodine is dissolved from the absorber which could be reused and the dissolved iodine solution will be solidified with cement. @ The absorber is solidified with cement directly. In the case @ the process may be simple and reliable for the isotope dilution because most of iodine may be I _ in the solution. In the case @ an iodine compound with low solubility should be mixed in the cement because if the added iodine leach away before
527
H. Nakamura et al.
528
leaching of 1* ’ I in absorber, The Ag doped absorbent
isotope dilution will not be expected. are used generally. The reduction effect of leachability
by dboping
Ag + compound is not reliable for geological disposal because AgI is unstable in reducing underground condition where AgI will decomposed to Ag metal and soluble iodine compounds, Therefore isotope dilution by adding iodine compounds to cement may be efective if iodine wastes are disposed of into reducing underground condition.
4 .
ISOTOPE DILUTION
ON THE MIGRATION
IN NATURAL ENVIRONMENT 1) On leaching to ground water Isotope dilution effect was discussed
by supposing
following conditions.
The ground water is same as sea water in which concentration of iodine ‘a’ is 6.4x10 -’ I: cme3. R.D.Scheele and coworkers determined leachability of iodine for various waste forms[5]. The results shows that normalized leaching rate from cementation from of AgI is 1X10 -’ g 2X10 -3 g cm -’ y -’ Supposing cm -’ y -’ and the one of Ba(I0 3 ) 2 -cements is concentration of iodine in waste forms is lo%, leaching rate of iodine ‘b’is 1X10 -’ g cm ’ Y ~’ for AgI-cement and is 2X10 m4 g cm -’ y 1 for Ba(I0 3 ) 2 -cement. When the change of rate of ground water in leaching cell is ‘n’y _ ’ , surface area of waste form is s cm ’ and volume of leaching cell is ‘A cm 3 , the value of ‘nA/s’ express approximately ground water flux. Maekawa and coworkers analyzed the hydrology of Tcthoku district in Japan[G] and found that hydraulic gradient at 500 m depth is distributed mostly in 0.004 - 0.04. In this discussion lower value 0.004 was used since low ground water flux area may be selected for siting process. By supposing that hydraulic conductivity is 25X10 5 cm set ~’ and porosity of rock matrix is 5 %, ground water flux v is estimated to be 63cm y ’ Isotope dilution ratio I d is calculated as follows. AgI-cement: I d = anA/bs = av/b = 6.4X10 -’ X63/1X10 mu= 4.0X10 3 Ba(I0 3 ) 2 -cement: I d = 6.4X10 -’ X63/2X10 -4 = 20 Based on above calculation the dose equivalent for any ingestion scenario after leaching will be less than 0.01 mSv/y for AgI-cement and 2 m&/y for Ba(I0 3 ) 2 cement. If the ground water is not saline water, isotope dilution will not, be expected. Ba(I0 3 ) 2 will change to soluble material isotope dilution on leaching process.
in case of reducing
condition,
As Agl and
it is difficult to expect
2) Isotope dilution with natural iodine contained in coast sea water The dilution factor of discharged radioactive liquid waste with coastal sea water at Pacific side in northern Japan is assessed form for authorization of Rokkasho
to be 5.8x10 ~I0 (Bq/cm 3 )/(Bqls) shown in the applicat.ion Fuel Reprocessing Plant[7]. Isotope dilution rate by natural
iodine in coastal sea water is calculated iodine concentration in sea water.
to be 3.4x10 6 g/y by using above dilution
factor and
To calculate the discharge rate ,it was supposing that I ” I containing wastes reprocessing of the spent fuel of 30,000 tU of 50 LWR(lGW) for 30 y operation is disposed an underground silo 25 m diameter and 50 m depth.
from of in
Potentialhazardof I-129 in radioactive wastes Other assumptions are follows; a) Hydraulic conductivity in the cement grouted silo is same as rock matrix and iodine is dissolved homogeneously in pore water. b) Ground water flows up in the silo of‘50 m depth by Darcy flow rate of 3.15 cm/y which is calculated by using above parameters. c) The ground water flow into the sea without dispersion. From these assumptions discharge rate of iodine is calculated. (3.07X10 ’ )X(3.0X10 4 ) + (5000 13.15) = 5.8X10 ’ g/y Then the isotope dilution is 3.4x10 B + 5.8X10 3 = 5.9X10 ’ Once the ground water is discharged into the sea, dose equivalent will not be over 0.07 mSv/y for any ingestion scenario.
5.
DISCUSSION
1) For transmutation of ’” I, it is possible to balance between production rate and transmutation rate if about 10 times of ’’’ I corresponding to the production in LWR Earone year is irradiated constantly by thermal neutron flux of 10 ’’ cm _ ’ s _ 1 . However, following two items should be examined. a) The self-shielding effect of target material on neutron distribution in reactor and its deteriorating effect on transmutation rate b) The concentration of I ’g I required for finally disposable waste form. The irradiation condition will depend on the requirement for the final concentration. When the irradiation time is too long, the feasibility may be decreased from not only technical problems but also social ones. 2) For isotope dilution in dissolution process and waste conditioning process, it may be possible to add necessary amount of iodine on both processes. If iodine is added to dissolution vessel, off-gas process should be changed and effects on the efficiency of down stream separation steps should be examined. The addition of iodine compound to solidification materials may be more feasible. However, it is necessary to study the stability of the material in the reducing underground condition where the leachability of iodine may increase. 3) Deep ground water is saline water somewhere at coast. These place is suitable to dispose of not only 1” I waste but other radioactive waste, because drinking water scenario may be not considered due to salinity and migration rate of nuclides may be very low, where flow rate of ground water may be very slow or stagnant. Isotope dilution will be expected by saline ground water in disposal sites when the waste is disposed in deep underground at coast. [l] R. J. M. Konings, et.al.,: Global 1995, Verailles in France Proc. p1631(1995)
[2] M. Hugon., et.al.,:Third Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation Cadrache in France p21 (1995) [3] ICRP Pub.68 (1994) (41 ICRP Pub.23 (1975) [5] R. D. Scheele, et.al.,: * Leach Resistance of Iodine Compounds in Portland
529
H. Nakamura et al.
530
Cement”
Geochemical
Behavior
of Radioactive
Waste,
Am. Chem.
p373-387(1984) [6] K.Maekawa, et.al.,:PNC TN7410 94-029(1994) [7] JNFL : Application form for reprocessing project authorization
Sot. 0097~6166/84/0246
in Rokkasho
7-S-131(1989)
Pergamon
Printed in Great Britain 0149-1970/98 $19.00 + 0.00
PII: SO149-1970(97)00063-2
A CONSIDERATION
TO REDUCE THE POTENTIAL
OF I-129 IN RADIOACTIVE
H.NAKAMURA
Y.INAGAKI
MSENOO
HAZARD
WASTES
S.TASHIRO
Radioactive Waste Management Center No.15.Mori Bldg. 2-8-10, Tranomon, Minato-ku Tokyo, Japan, 105
ABSTRACT Transmutation of I2 ’ I to ’3’ I by (n, 7 ) reaction and isotope dilution with stable iodine were discussed. The transmutation in LWR is calculated by supposing that targets irradiated for 25 years are substituted with new targets. The result showed that initial amount of ’’’ I will be reduced to 11.5%. In order to lower dose equivalent of general public than O.lmSv/y, 400 times isotope dilution of 1” I is required. The feasibility of the isotope dilution was discussed at dissolution process of spent fuel, conditioning process of the waste and disposal conditions. 0 1997 Published by Elsevier Science Ltd
INTRODUCTION The potential hazard of ’’’ I on mankind from disposal of reprocessing plant wastes has been estimated relatively high. This is due to its high mobility in certain geological environment. It can migrate easily away from underground repositories and, hence, eventually reach the biosphere. The environmental condition including geosphere can hardly be estimated for the long term comparable to the half life of ’” I. Therefore, reduction of ’* ’ I potential hazard would be desirable. One possible method may be reduction of 1’’ I as a source by transmuting it to 13’ I by (n, 7 1 reaction. As Iodine is eventually separated and recovered in the dissolver off-gas treatment system in reprocessing plant, it is possible to prepare the target material for the transmutation of “’ I by changing the chemical form without any more separations. R.J.M.Koning and coworkers irradiated PbI 2 , CeI 3 , NaI and others with thermal neutrons with a flux of 1.5X10 ’4 cm _ ’ s _ ’ . The results show that they are stable enough against the irradiation condition[l]. M.Hugon and coworkers calculated the transmutation rate of ’” I in core region of a fast reactor and showed that 22kg of lZ9 I could be transmuted with a transmutation half-life of around 44 years[2]. Isotope dilution with stable iodine is also a possible method to reduce maximum potential dose equivalent. The iodine content in a thyroid is limited. Therefore the maximum potential dose equivalent is determined by the specific activity of iodine. Once the required isotope dilution is practiced, the potential dose equivalent after the dilution will be determined by the 525
H. Nakamuraet al.
526
specific activity for any ingestion scenario. The sea water contains relatively large amo_Jnt of iodine. If proper siting for underground disposal is carried out, the efficient isotope dilution will be possible. Feasibility of the above two methods was discussed on supposing simple conditions.
1.
PRODUCTION
OF IODINE IN SPENT FUEL
Production rate of iodine in LWR and the specific activity was estimated by using ORIGEN and shown in Table 1. Table 1 Production of iodine and it specific radioactivity in spent fuel
Burn up Production rate of iodine
45
GWD/t
3.07x10 2 g/t 15X10 ’ Bq/t 503x10 ’ Bqlg Specific activity of 1’’ I Uranium needed for 1 GW LWR 20 t/Y Production rate of iodine for years 6 kg/y
As most all1 from the fuel dissolution process is collected on the absorber of the off-gas system, the absorbed I should be the only one target for the transmutation.
2.
TRANSMUTATION
RATE OF ’’’ I
The transmutation half life of I ’’ I by 1” I(n, 7 ) ’ 3 ’ I reaction with thermal neutron is shown as follows. N = N o exp(- CJ4 t ) = N o exp(- 0.0864t) Where N o : number of ’* ’ I nuclide in target material N : number of ’’’ I nuclide in target after the irradiiation for t cr : cross section for thermal neutron (27 b: 2.7x10 _ ’3 cm ’ ) @ : neutron flux( 1x10 ’4 cm _ * s _ ’ in core region of 1 GW LWR) t : irradiation time( 1 y: 365x24~60~60 = 3.2x10 7 s )
’’’ I in the target after one year irradiation is as follows. The reduction ratio of N 1ye a r /N o = exp(-0.0864X1) = 0.9. The transmutation rate par evry one year is about 10 %. For one practical system, it is supposed that a target is irradiated for 25 years. and substituted with new one. The reduction ratio of ’” I after irradiation of 25 years is calculated as follows. N/N o = exp(-0.0864X25) = 0.115 The irradiation capacity is necessary for 150 kg iodine which is produced for 25 years The
Potential hazard of I-129 in radioactivewastes
ratio of the total amount of ’’’ I in a reactor(N II ) to annual production(N o ) is calculated as follows. N, Ma = exp(-0.0864t)dt = (1 - 0.115)/0.0846 = 10.5 If this system will be operated continuously for ever, initial amount of ’’’ I will be reduced to 11.5 %. When it is ceased,however, the amount of ’” I corresponding to the production in the LWR for about 10.5 years will remain in the reactor finally. When irradiation facilities are installed out of core region, larger capacity and longer irradiation period will be required.
3 .
ISOTOPE DILUTION IN FUEL REPROCESSING
1) Calculation of dose equivalent on supposing all iodine in environment being substituted by iodine with the same specific activity as iodine in spent fuel. Effective dose(whole body) coefficient of ’* ’ I ingestion is 1.1x10 _ 7 Sv Bq -’ based on ICRP Pub.68[3]. Intake of iodine by general public is 0.2 mg/d based on ICRP Pub.23[4]. Supposing that all iodine in environment are substituted by the iodine of the same specific activity as that in spent fuel, ingestion of ’’@I in a year is calculated to be 3.6x10 ’ Bq and 40 mSv/y by using the above values and the specific activity shown in Table 1. In order to lower the dose equivalent of general public less than O.lmSv/y, 400 times isotope dilution of ’* !aI is required. Once the dilution is carried out, dose equivalent with any ingestion scenario will not be over 10% of 1 mSv/y(whole body) recommended by ICRP as dose limit for general public. 2) Isotope dilution process Supposing 400 times isotope dilution is performed during fuel reprocessing, an addition of about 2.4 t of iodine is required for processing 20 t of spent fuel. One of possible methods is the addition to the dissolver and another is the addition to matrix of solid waste of the iodine absorber. The addition to the dissolver may result in efficient isotope dilution and higher recovery of
’” I by carrier effect. In this case the amount of iodine contained in off-gas may be too large for solid absorber and then absorbing process by alkaline solution should be used instead. The unfavorable effect of iodine addition on downstream separation steps must be examined. Since this method requires change of the plant design and re-examination of the process, the application will not be expected without new plant construction. The conventional process has been completed by long term research and development. No change of the main process should be kept and the isotope dilution process should be added to the waste management steps. Isotope dilution on waste conditioning process may depend on waste forms. Following conditioning process are considered. 0 Iodine is dissolved from the absorber which could be reused and the dissolved iodine solution will be solidified with cement. @ The absorber is solidified with cement directly. In the case @ the process may be simple and reliable for the isotope dilution because most of iodine may be I _ in the solution. In the case @ an iodine compound with low solubility should be mixed in the cement because if the added iodine leach away before
527
H. Nakamura et al.
528
leaching of 1* ’ I in absorber, The Ag doped absorbent
isotope dilution will not be expected. are used generally. The reduction effect of leachability
by dboping
Ag + compound is not reliable for geological disposal because AgI is unstable in reducing underground condition where AgI will decomposed to Ag metal and soluble iodine compounds, Therefore isotope dilution by adding iodine compounds to cement may be efective if iodine wastes are disposed of into reducing underground condition.
4 .
ISOTOPE DILUTION
ON THE MIGRATION
IN NATURAL ENVIRONMENT 1) On leaching to ground water Isotope dilution effect was discussed
by supposing
following conditions.
The ground water is same as sea water in which concentration of iodine ‘a’ is 6.4x10 -’ I: cme3. R.D.Scheele and coworkers determined leachability of iodine for various waste forms[5]. The results shows that normalized leaching rate from cementation from of AgI is 1X10 -’ g 2X10 -3 g cm -’ y -’ Supposing cm -’ y -’ and the one of Ba(I0 3 ) 2 -cements is concentration of iodine in waste forms is lo%, leaching rate of iodine ‘b’is 1X10 -’ g cm ’ Y ~’ for AgI-cement and is 2X10 m4 g cm -’ y 1 for Ba(I0 3 ) 2 -cement. When the change of rate of ground water in leaching cell is ‘n’y _ ’ , surface area of waste form is s cm ’ and volume of leaching cell is ‘A cm 3 , the value of ‘nA/s’ express approximately ground water flux. Maekawa and coworkers analyzed the hydrology of Tcthoku district in Japan[G] and found that hydraulic gradient at 500 m depth is distributed mostly in 0.004 - 0.04. In this discussion lower value 0.004 was used since low ground water flux area may be selected for siting process. By supposing that hydraulic conductivity is 25X10 5 cm set ~’ and porosity of rock matrix is 5 %, ground water flux v is estimated to be 63cm y ’ Isotope dilution ratio I d is calculated as follows. AgI-cement: I d = anA/bs = av/b = 6.4X10 -’ X63/1X10 mu= 4.0X10 3 Ba(I0 3 ) 2 -cement: I d = 6.4X10 -’ X63/2X10 -4 = 20 Based on above calculation the dose equivalent for any ingestion scenario after leaching will be less than 0.01 mSv/y for AgI-cement and 2 m&/y for Ba(I0 3 ) 2 cement. If the ground water is not saline water, isotope dilution will not, be expected. Ba(I0 3 ) 2 will change to soluble material isotope dilution on leaching process.
in case of reducing
condition,
As Agl and
it is difficult to expect
2) Isotope dilution with natural iodine contained in coast sea water The dilution factor of discharged radioactive liquid waste with coastal sea water at Pacific side in northern Japan is assessed form for authorization of Rokkasho
to be 5.8x10 ~I0 (Bq/cm 3 )/(Bqls) shown in the applicat.ion Fuel Reprocessing Plant[7]. Isotope dilution rate by natural
iodine in coastal sea water is calculated iodine concentration in sea water.
to be 3.4x10 6 g/y by using above dilution
factor and
To calculate the discharge rate ,it was supposing that I ” I containing wastes reprocessing of the spent fuel of 30,000 tU of 50 LWR(lGW) for 30 y operation is disposed an underground silo 25 m diameter and 50 m depth.
from of in
Potentialhazardof I-129 in radioactive wastes Other assumptions are follows; a) Hydraulic conductivity in the cement grouted silo is same as rock matrix and iodine is dissolved homogeneously in pore water. b) Ground water flows up in the silo of‘50 m depth by Darcy flow rate of 3.15 cm/y which is calculated by using above parameters. c) The ground water flow into the sea without dispersion. From these assumptions discharge rate of iodine is calculated. (3.07X10 ’ )X(3.0X10 4 ) + (5000 13.15) = 5.8X10 ’ g/y Then the isotope dilution is 3.4x10 B + 5.8X10 3 = 5.9X10 ’ Once the ground water is discharged into the sea, dose equivalent will not be over 0.07 mSv/y for any ingestion scenario.
5.
DISCUSSION
1) For transmutation of ’” I, it is possible to balance between production rate and transmutation rate if about 10 times of ’’’ I corresponding to the production in LWR Earone year is irradiated constantly by thermal neutron flux of 10 ’’ cm _ ’ s _ 1 . However, following two items should be examined. a) The self-shielding effect of target material on neutron distribution in reactor and its deteriorating effect on transmutation rate b) The concentration of I ’g I required for finally disposable waste form. The irradiation condition will depend on the requirement for the final concentration. When the irradiation time is too long, the feasibility may be decreased from not only technical problems but also social ones. 2) For isotope dilution in dissolution process and waste conditioning process, it may be possible to add necessary amount of iodine on both processes. If iodine is added to dissolution vessel, off-gas process should be changed and effects on the efficiency of down stream separation steps should be examined. The addition of iodine compound to solidification materials may be more feasible. However, it is necessary to study the stability of the material in the reducing underground condition where the leachability of iodine may increase. 3) Deep ground water is saline water somewhere at coast. These place is suitable to dispose of not only 1” I waste but other radioactive waste, because drinking water scenario may be not considered due to salinity and migration rate of nuclides may be very low, where flow rate of ground water may be very slow or stagnant. Isotope dilution will be expected by saline ground water in disposal sites when the waste is disposed in deep underground at coast. [l] R. J. M. Konings, et.al.,: Global 1995, Verailles in France Proc. p1631(1995)
[2] M. Hugon., et.al.,:Third Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation Cadrache in France p21 (1995) [3] ICRP Pub.68 (1994) (41 ICRP Pub.23 (1975) [5] R. D. Scheele, et.al.,: * Leach Resistance of Iodine Compounds in Portland
529
H. Nakamura et al.
530
Cement”
Geochemical
Behavior
of Radioactive
Waste,
Am. Chem.
p373-387(1984) [6] K.Maekawa, et.al.,:PNC TN7410 94-029(1994) [7] JNFL : Application form for reprocessing project authorization
Sot. 0097~6166/84/0246
in Rokkasho
7-S-131(1989)