Martin Hornáček, Vladimír Nečas

Transcription

1 THE USE OF CODES ISIPLAN 3D ALARA AND MATLAB FOR ASSESSMENT OF CONTRIBUTIONS FROM RADIATION SOURCES WITHIN THE DECOMMISSIONING OF NUCLEAR POWER PLANTS Martn Hornáček, ladmír Nečas Insttute of Nuclear and Physcal Engneerng, Faculty of Electrcal Engneerng and Informaton Technology, Slovak Unversty of Technology n Bratslava E-mal: martn.hornacek@stuba.sk Receved 04 May 2015; accepted 12 May 2015 Abstract The knowledge of radaton stuaton s one of the crucal requrements for the realsaton of dsmantlng actvtes n a safe and economc way. The calculaton of external exposure s (among the materal composton and geometrc dmensons) strongly dependent on the source term actvty and nuclde composton. However, the exact data regardng ths source term are often not avalable. Thus only approxmate results can be obtaned. From ths reason the methodology was developed whch reflects ths phenomenon. The calculated values (usng ISIPLAN 3D ALARA code) are related to 60 Co only. These values are then used for calculaton of the dose rates wth consdered actvty and nuclde composton (usng MATLAB code). The obtaned results are compared wth the "standard" results obtaned from drect enterng all the data to the ISIPLAN 3D ALARA code. The comparson of the results from these two approaches s presented and analysed. The results of the comparsonenable the use of ths methodology n the calculaton of external exposure. 1. Introducton The decommssonng of nuclear power plants (NPP) s the end part of ther lfe-cycle. In Slovaka, 1 NPP n Jaslovské Bohunce s currently n the second decommssonng stage wth the planned duraton between 2015 and 2025 [1]. Accordng ths document, the actvated and contamnated parts wll be cut n stu and the fragmented parts wll be ether stored or dsposed n Natonal radoactve waste repostory n Mochovce. The components to be cut can be dvded to actvated components (the actvty content s mostly due to neutron actvaton) reactor pressure vessel, reactor cavty, reactor cavty bottom, core basket and to contamnated components (the actvty content s a result of contamnaton of actvated corroson products and/or fsson products) pressurser, steam generator. In ths paper, steam generator s studed from the perspectve of calculaton of external exposure durng varous tasks connected wth ts dsmantlng and fnal dsposal of resultng radoactve waste. Ths component s a part of prmary crcut and conssts of the followng parts: Heat exchange tubes Collectors Steam generator casng and steam collector (non-contamnated part of secondary crcut). As was mentoned, the source of the actvty s the contamnaton of the nner walls of heat exchange tubes and collectors due to the flow of prmary crcut coolant. 238

2 In the followng chapter, the ssues connected wth the source term are dentfed and dscussed. 2. The ssues connected wth the source term The estmaton of the actvty content and nuclde nventory of actvated parts can be carred out usng for nstance MCNP software. The exact data regardng materal composton, neutron flux (spatal dstrbuton, power and hstory of operaton frequency and duraton of outages) must be known. Sgnfcantly dfferent stuaton s n case of contamnaton. The level of contamnaton s strongly dependent (among the parameters mentoned n case of actvated parts) on the creaton of oxdaton layer of the metallc surfaces, chemcal condtons durng the operaton and outages, temperature and of the process of coolant s flow, the geometry and dmensons of technologcal equpment (e.g. ppes, elbows, etc.) and on the chemcal propertes of radonucldes (e.g. solublty and adheson). From ths reason, the calculaton of contamnaton s ether mpossble or wth hgh uncertantes [2] Thus the estmaton of the contamnaton s often based on n-stu measurements. The measurements (carred out durng radologcal charactersaton before decommssonng), however, could not reflect the real state n each part of the technologcal equpment. From these measurements and samples analyss average values are obtaned whch can vary n local places. Drect enterng of these data to calculaton software can lead to approxmate results. Durng partal tasks, however, the calculatons should be repeated to acheve hgher accuracy. Moreover, n case of applcaton of decontamnaton methods secondary waste s generated (e.g. spent on exchange resns) and the amount, nuclde nventory and actvty content can vary case-by case and the exact data wll be known just n the tme of realsaton.these factors make the calculaton of external exposure dffcult and requre development of such calculaton methodology whch s more flexble than the drect enterng of approxmate data nto calculaton software. 3. Computer code ISIPLAN 3D ALARA For calculaton of external exposure the computer code ISIPLAN 3D ALARA was used. The calculaton prncple s as follows [3] The photon fluency rate at the dose pont near the volume source can be determned by consderng the volume source as consstng of a number of pont sources. By addng the contrbuton of every pont source to the dose at the dose pont the photon fluency rate at the dose pont s expressed as: b S. B. e d (m -2.s -1 ) (1) 2 4 where: S source strength per unt volume (n.s -1 ), ρ dstance from a pont source (m), B buldup coeffcent (-), b attenuaton effectveness coeffcent (-), volume (m 3 ). Each small source s called a kernel and the process of ntegraton, where the contrbuton to the dose of each pont s added up, s called "pont kernel" ntegraton. Based on the photon fluency rate at a pont t s possble to calculate the effectve dose rate dependng on the dose converson factors selected n the calculatons: E h. (Sv/s) (2) where: h dose converson coeffcent for photons of energy E (Sv per photon/m 2 ), fluency rate of the photons at energy E (m -2.s -1 ). 239

3 From the Eq. (1) and Eq. (2) t s obvous that the relaton of the dose rate from the actvty s lnear,.e. f the actvty ncreases (decreases) n-tmes, the dose rate wll ncrease (decrease) n-tmes (when the other parameters reman unchanged). One of the utltes of the ISIPLAN 3D ALARA code s that the contrbuton from each source to the dose rate n the pont can be shown whch s crucal for the creaton of calculaton methodology descrbed n the followng chapter. 4. Calculaton methodology combnng ISIPLAN 3D ALARA and MATLAB codes From the perspectve of external exposure, only those nucldes are mportant, durng whose decay γ or rtg photons are emtted. The number of dfferent energy groups of photons emtted durng decay of specfc radonuclde can be charactersed by gamma ray dose 1 constant Γ [ 2 C. m. kg ] [4] In practcal applcatons the unt of (msv/h)/mbqcan be used. In ths case, Γ represents the unshelded gamma ray dose-equvalent rate at 1 meter from a pont source wth gven actvty [5]. Usng ths nterpretaton of Γ, the contrbutons of the nucldes to the dose rate can be related to one reference nuclde. In the analyss, 60 Co as the reference nuclde was selected. Ths was done from two man reasons: 60 Co s so-called easy-tomeasure nuclde whch can be measured by γ-spectrometry[6]. The other reason s that ths radonuclde s one of the most abundant radonucldes wthn the 1 NPP nventory. Based on these assumptons, the calculaton methodology can be as follows: Creaton of calculaton model n ISIPLAN 3D ALARA code wth the actvty of each source 1 Bq of 60 Co and creaton of set of ponts where the dose rate wll be calculated (so-called trajectory) Processng of obtaned data (dose rates from each source) nput data for further calculatons n Eq. (3)depcted as f(scalar) Determnaton of so-called converson vector set as rato Γ / Γ Co (Γ represents gamma ray dose constant of nuclde ) n Eq. (3)depcted as converse (vector) ector of share of radonucldes (obtaned by n-stu measurements) n Eq. (3) depcted as share(vector) The total actvty of the source (obtaned by n-stu measurements) n Eq. (3) depcted as actvty (scalar). Thus the dose rate can be calculated n matrx processng software (e.g. MATLAB): dose _ rate sum( share* actvty.* converse )* f (3) The symbol.* means the multplcaton of an element n the n-th row of the frst vector wth the element n n-th row of the second vector. The man advantage of ths approach s that the varable parameters are vector of share of radonucldes and the total actvty of the relevant source. In case of change of these two parameters only calculaton n MATLAB usng the Eq. (3) s requred. The repeated enterng the data to ISIPLAN 3D ALARA (whch can be n case of complex models tmeconsumng) and repeated calculaton s then avoded. 4.1 Comparson of the calculaton methodologes To compare the results obtaned by calculaton after drect enterng the same source data to the ISIPLAN 3D ALARA code wth the methodology already descrbed, the dsposal of radoactve waste (placed n fbre-concrete contaners FCC) nto Natonal Radoactve Waste Repostory n Mochovce s studed. Ths stuaton was analysed n detal n[7] and n 17th Regonal Semnar on Radoactve Waste Dsposal (20-22 October 2014, Budapest). 240

4 The comparson of obtaned results by drect enterng source data nto ISIPLAN code and obtaned usng Eq. (3) s depcted n Tab. 1, Tab. 2 and Tab. 3. The devaton was calculated as follows: where: ( D M D ) devaton *100 (4) D D M dose rate obtaned usng Eq. (3) (msv/h), D dose rate obtaned by drect enterng the source data nto ISIPLAN(mSv/h). Note: all the values depcted n the tables are rounded at two decmal places. Tab. 1.Calculated dose rates for truck drver. ISIPLAN and Devaton [%] ISIPLAN MATLAB Transport to the repostory 2.30E E E+00 Arrval to the repostory 2.30E E E+00 Unloadng the truck 2.30E E E+00 Tab. 2.Calculated dose rates for assstant. ISIPLAN Devaton [%] ISIPLAN andmatlab sual control of FCCs 5.15E E E+00 Unloadng the truck 2.25E E E+00 Dose rate measurements 2.60E E E+00 Lftng one FCC 2.00E E E-01 Transport of FCC to the box 3.20E E E-01 Dsposal of the FCC 5.30E E E+00 Coverng the box 1.75E E E-04 Tab. 3.Calculated dose rates for crane operator. ISIPLAN Devaton [%] ISIPLAN andmatlab sual control of FCCs 5.50E E E+00 Unloadng the truck 1.25E E E+00 Dose rate measurements 1.00E E E+00 Lftng one FCC 4.10E E E+00 Transport of FCC to the box 9.40E E E-01 Dsposal of the FCC 3.00E E E+00 Coverng the box 2.80E E E-07 From the depcted data t s obvous that the dose rates obtaned by calculaton methodology usng Eq. (3) are slghtly hgher than the results of drect entrance to ISIPLAN code. The dfferences can be explaned by the fact that output from ISIPLAN 241

5 code (the dose rates) s rounded at two decmal places. The other fact s that durng samplng the sources (chapter 3 pont kernel ntegraton) Monte Carlo method s used. That means that the dose rates can slghtly vary f the calculaton would be repeated. 5. Concluson The calculaton of external exposure durng dsmantlng of contamnated components s connected wth many uncertantes of the source term (nuclde composton and actvty) due to the non-homogeneous dstrbuton of contamnaton. From ths reason the presented calculaton methodology was developed whch reflects ths stuaton and enables flexblty when the source term s changed (as a result of n-stu measurements). Acknowledgement Ths project has been partally supported by the Slovak Grant Agency for Scence through grant EGA 1/0796/13 and by the Slovak Unversty of Technology n Bratslava wthn the bounds of project HORAR (2015). References: [1] Nuclear and Decommssonng Company: The Intent n terms of the Act No. 24/2006 Coll. on Envronmental Impacts Assessment and Alternatons and Amendments of certan Acts as amended The 2nd Stage of Decommssong of 1 NPP Jaslovské Bohunce. [onlne]. Avalable: < (n Slovak). [2] Internatonal Atomc Energy Agency. Radologcal Characterzaton of Shut Down Nuclear Reactors for Decommssonng Purposes: Techncal Reports Seres No enna: IAEA, ISBN X. [3] F. ermeersch (2005): Dose Assessment and ALARA Calculaton wth ISIPLAN 3D ALARA Plannng tool. Tranng Course. IDPBW Nuclear Studes, Boeretang: SCK.CEN, Belgum, 2005 [4] J. Lpka et al.: Nuclear Physcs and Techncs: Gude for Laboratory Exercses. Bratslava: FEI STU, ISBN (n Slovak). [5] D. Trubey, L. Unger: Specfc Gamma-Ray Dose Constants for Nucldes Important to Dosmetry and Radologcal Assessment. Oak Rdge Natonal Laboratory: ORNL/RSIC-45. [onlne]. Avalable: < [6] Internatonal Atomc Energy Agency. IAEA Nuclear Energy Seres No. NW-T Determnaton and Use of Scalng Factors for Waste Characterzaton n Nuclear Power Plants. enna: IAEA, ISBN [7] M. Hornáček,. Nečas: Calculaton of External Exposure durng Transport and Dsposal of Radoactve Waste arsen from Dsmantlng of Steam Generator. In: European Nuclear Conference ENC2014, May 2014, Marselle, France. ISBN