ESTIMATION OF THE GAMMA DOSE RATE FOR HOMOGENOUS WASTE CONTAINERS *

Transcription

1 ESTIMATION OF THE GAMMA DOSE RATE FOR HOMOGENOUS WASTE CONTAINERS * C.N. DULAMA 1, M. DULAMA 1, R. DOBRIN 1, M. PAVELESCU 2 1 Insttute for Nuclear Research, Pteşt, Romana, crstan.dulama@nuclear.ro, 2 Academy of Scentsts - Bucharest, Romana Receved September 14, 2010 Spent on exchange resns and lqud radoactve wastes are two of the man types of radoactve wastes generated n quas-homogenous batches by the waste streams of a nuclear nstallaton. Transport regulatons requre, for gamma emttng radoactve waste, that the dose rate n the proxmty of the contaner should stand below a certan threshold. Also, the condtonng technologes for such wastes are based on mmoblzaton of certan quanttes of waste n varous matrces. The dosage s performed accordng to the technology prescrptons and to meet the acceptance crtera for transport and fnal storage. Gamma emttng radonucldes lke: Mn-54, Co-60, Co-57, Sb-124, Sb-125, Zn-65, Cs-134 and Cs-137 are the man contrbutors to the dose rate produced by the regular radoactve waste from a nuclear nstallaton. A method was developed to optmze the quantty of waste per unt volume of condtonng waste form, based on a mathematcal model for dose rate estmaton. A computer program was desgned to mplement ths method and computed data were tested aganst measured values. The paper presents the mathematcal model used for dose rate assessment and the results obtaned n the valdaton phase of the method development. A reference test case was, also, consdered for a smple geometry, where an analytcal soluton was calculated and compared wth the modeled results. Key words: dose rate, on exchange resns, contaner, Monte Carlo, radoactve waste. 1. INTRODUCTION Radoactve waste handlng actvtes are hazardous as concern both contamnaton and external exposure. Therefore strct regulatons are appled for radaton protecton n ths feld. Dose rate n the vcnty of the waste contaner has to be kept below certan constrants throughout entre treatment and condtonng process. Also, durng transportaton and for the fnal storage, the dose rate should meet acceptance crtera stated by the safety regulatons. On the other hand, the waste management actvtes are very costly and thus the fnal product, whch s the waste form, should ncorporate as much actvty as possble to optmze these costs. * Paper presented at the 11 th Internatonal Balkan Workshop on Appled Physcs, July 7 9, 2010, Constanta, Romana. Rom. Journ. Phys., Vol. 56, Nos. 9 10, P , Bucharest, 2011

2 2 Estmaton of the gamma dose rate 1137 In case of quas-homogeneous waste streams, lke spent on exchangers and aqueous radoactve waste from nuclear nstallatons, the most frequent used technology for condtonng s embeddng n stable sold matrces (btumen for on exchange resns and concrete for lqud waste) [1]. The dosage s made gravmetrcally usng as bass the ntal radologcal characterzaton data. The estmaton of radoactve burden for the waste form s made conservatvely whch leads to a decreased effcency of the condtonng technology. Snce 1993 the Waste Treatment Faclty (STDR) from the Insttute for Nuclear Research, Ptest, operates a plot scale nstallaton for condtonng by btumnzaton of spent on exchange resns used n the prmary coolng system of the TRIGA reactor. These resns are n the form of anon exchanger (60% INR-78) and caton exchanger (40% INR-77) wth actvty concentraton between 10-3 and 10 3 C/m 3 beng treated and condtoned by usng a technology developed n STDR. The man gamma emttng contamnants are Mn-54, Co-60, Co-57, Sb-124, Sb-125, Zn-65, Cs-134 and Cs-137 [2]. The fnal product of the treatment and condtonng technology for spent on exchangers s ABBSB-1 (Metallc drum Concrete Btumnzed on exchangers set), whch s manufactured accordng to the nternal standard SF ICN 012/1994. Ths product s n the form of a cylndrcal concrete block havng embedded the contaner wth btumnzed on exchangers. The external concrete wall s the bologc sheldng of the waste form. The spent on exchangng resns are condtoned n btumen class D25/40 at temperature between 110 C and 125 ºC. The volume embeddng rato s 2/3 (on exchangers versus btumen). The mxture s poured n 80 lters metallc drums whch are then postoned wthn larger (220 l) contaners and embedded n concrete. The average densty of the mxture of btumen and on exchangers lays between 1.03 and 1.14 g/cm 3. The dmensons of the fnal product are: dameter = 600 ± 5 mm, heght = 905 ± 10 mm, mass = kg [2, 3]. In order to optmze the loadng of the radoactve waste forms, we developed a methodology and a computer program for assessng the dose rate n the vcnty of the waste contaner. 2. METHOD S DESCRIPTION The method s based on a Monte Carlo algorthm used to smulate the emsson of gamma radatons wthn the contaner space whch are preferentally emtted towards the measurng pont. The path lengths wthn the waste matrx and the contaner wall are calculated and the self-absorpton coeffcent s computed based on the matrx characterstcs. A buld-up factor s then determned by usng the Taylor approxmaton formula. α µ r α µ r µ = BE (, r) Ae Ae

3 1138 C.N. Dulama, M. Dulama, R. Dobrn, M. Pavelescu 3 where A 1, A 2, α 1, α 2 are parameters gven by Shults et al. n [4], and µ s the lnear attenuaton coeffcent for gamma energy E 0. The dose converson coeffcents for a lst of radonucldes are extracted from Nucldes2000 software [5] and are ready to be used n dose estmatons. The dose rate s computed as the average of dose rate for each pont selected from the waste contaner multpled wth the estmated nuclde actvty. A DCF DAd (, ) = F ( d) B N N = 1 where A s the total actvty, d s the dstance to the contaner, DCF s the dose converson factor, F ( d ) s the attenuaton factor for a pont wthn the contaner att and B s the buld-up factor. The attenuaton factor s a functon of the total dstance between the pont selected wthn the contaner (P) (see Fg. 1) and the dose rate estmaton pont (M): F att µ lat e ( d) = ( d + R+ z ) + x + y att where lat s the path length of the radaton wthn the waste contaner, R s the contaner radus and (x, y, z ) are the samplng pont coordnate Y P lat X x 2 +y 2 0 R d M Z Fg. 1 Schematc representaton of the computaton model.

4 4 Estmaton of the gamma dose rate RESULTS For model valdaton a set of 23 waste drums contanng btumnzed spent on exchange resns was used. The waste forms were characterzed by n stu gamma ray spectrometry. The detecton system and calbraton method are extensvely descrbed n prevous works of the authors [6]. The effcency calbraton method was also evaluated by other authors [7, 8] and was found to be approprate for homogeneous waste contaners. The dose rate was measured wth a hand held devce at contact wth the sde of the waste drum. The value selected as representatve for the dose rate at contact was the average of readngs regstered at the half-heght on the sde of the waste drum. By usng the method descrbed above, the dose rate was estmated based on the spectrometrc measurement results. The Table 1 contans the mentoned data and the percent dscrepancy calculated. Table 1 The valdaton results for the dose rate estmaton method Waste Nuclde actvty (MBq) Dose rate (msv/h) drum # Mn-54 Co-58 Co-60 measured calculated % dscrepancy The assocated uncertantes were estmated for both drect measurement method and model based estmaton method and the results are plotted n Fg. 2.

5 1140 C.N. Dulama, M. Dulama, R. Dobrn, M. Pavelescu 5 To be noted that the hand held dose rate measurng devce used, s factory set to acheve a 15% relatve uncertanty n drect readng mode. Ths does not nclude the dose rate uncertanty assocated wth the dstance estmaton, or dose rate probe postonng Calculated dose rate (µsv/h) Measured dose rate (msv/h) Fg. 2 Calculated vs. measured dose rate wth ther assocated uncertantes. The large dscrepances recorded n some cases are manly due to the nternal nhomogenetes of the waste form whch can alter both the dose rate measurement results and spectrometrc results. To be noted that the man hypotess of the model s the homogenety of the waste form content and the unformty of the actvty dstrbuton. The computaton model was verfed also by comparng the results provded for extreme case of lnear source. In ths case an analytcal soluton for dose rate calculaton can be obtaned as: h 2 A DCF atg 2 d Ddh (, ) = h d where: A s the source actvty, h s the heght of the source, d s the dstance between the source and the dose estmaton pont and DCF s the dose converson factor.

6 6 Estmaton of the gamma dose rate 1141 The Table 2 shows the values obtaned for dose rate wth the Monte Carlo based method both for drum source and lnear source and the dose rate calculated for lnear source. The last column shows the percent dscrepancy between the values calculated wth the model and the analytcal formula for lnear source. For calculaton the followng data were used: drum radus 20 cm, drum heght 70 cm, drum wall thckness 1 mm, drum content densty 1 g/cm 3, lnear source length 70 cm, source actvty 1 MBq, source composton Co-60. Table 2 Comparson between model calculatons and analytcal calculatons for a lnear source Dstance Dose rate (µsv/h) % dscrepancy (m) MC Analytcal CONCLUSIONS The model calculatons were n good agreement wth the measured values of the dose rate for the set of radwaste consdered for valdaton. The test case based on lnear source showed a very good agreement of the dose rate provded by the model calculatons and the dose rate gave by the analytcal formula. The flexblty of the model mplemented n the computatonal tool, makes t a useful nstrument for the estmaton of the dose rate n the vcnty of a waste contaner based on the pror characterzaton of ts radoactve content. REFERENCES 1. M. Dulama, N. Deneanu, M. Pavelescu, L. Pasăre, Combned radoactve lqud waste treatment processes nvolvng norganc sorbents and mcro/ultrafltraton, Rom. Journ. Phys., Vol. 54, Nos. 9 10, , Bucharest, Bălăşou, M., Antonescu (Dulama), M., Tehnologa de tratare condţonare a schmbătorlor de on uzaţ de la reactorul TRIGA, ICN Pteşt, Popescu, I., Ansamblu Buto metalc - Beton - Deseur radoactve de joasa actvtate, SF SCN 012 /1994, act.3/2004.

7 1142 C.N. Dulama, M. Dulama, R. Dobrn, M. Pavelescu 7 4. J. Kenneth Shults and Rchard E. Faw, Radaton Sheldng, ISBN , Amercan Nuclear Socety, La Grange Park, IL, Nucldes 2000: An Electronc Chart of the Nucldes, Ver. 1.2, European Communtes, R.I. Dobrn, C.N. Dulama, Al.Toma - Shell Source Method n Radwaste Assay - Romanan Journal of Physcs, Vol. 49, Nos. 5 6, , M. Haralambe, L. Dnescu, O. Sma, P. Stoca, Study concernng the effcency calbraton of a drum waste assay system, Rom.Journ.Phys. Vol. 51, Nos. 1 2, 77 87, Bucharest, Magdalena Toma, Octavan Sma, Carmen Crstache, Felca Dragolc, Laurentu Done, Effcency calbraton studes for gamma spectrometrc systems: the nfluence of dfferent parameters, Rom. Journ. Phys., Vol. 53, Nos. 7 8, , Bucharest, 2008.