Proceedings of the 18th International Conference on Nuclear Engineering ICONE18 May 17-21, 2010, Xi'an, China ICONE18- THE IMPLEMENTATION OF RADIOLOGICAL CHRACTERIZATION FOR REACTOR DECOMMISSIONING DENG Junxian, ZHAO Huasong, LI Xin, DENG Feng China Nuclear Power Engineering Co. Ltd, Beijing 100840, China ABSTRACT: The radiological characterization includes: the collection and review of historical file; the performing calculation of radionuclide inventory in the reactor; in situ measurement; sampling analyses; the review and evaluation of the data obtained; the comparison of calculated result with measured date etc. The special attention should put to the information from the key part in reactor for end state radiological characterization. The sampling from the hot spot should not be lost; the number of the sampling should reasonable base on reliable statistics. The radioactivity density for site release should comply with the guide, standard and regulation of international atomic energy agency and our country. INTRODUCTION The radiological characterization is to determine the type, the distribution and the radioactivity of the radionuclide in the parts of the decommissioning reactor by the information research, the in situ measurements, the sampling and analyses and the calculations. It is used to make the decision of the implementation option of the decommissioning and the measure of radiological protection for the workers. After the completion of the decommissioning it is used to make sure if the site can be released. The radiological characterization should go through whole decommissioning process. From the decommissioning option making to the completion of the actions,the implementation of the radiological characterization should be done properly. OBJECTIVE OF RADIOLOGICAL At the beginning stage of decommissioning design the purpose is to collect sufficient information to assess the radiological status, nature and extent in the reactor. As the progresses of decommissioning the objective is to gather more detail data used for the decision of the operation technology ; the decontamination process;the dismantling procedure and tools needed;the radiological protection measure for the workers;the option of the waste treatment; the method of environment protection;and the cost estimation. After the completion of the decommissioning the objective is monitoring of the residual activity levels in all parts of the site, to verify if the activity levels are reached the criteria for site release. PROCESS The characterization should follow the reasonable method and procedure. The characterization program comprises the review of historical information ; the implementation of calculation;the performance of in situ measurement; the preparation of sampling and analysis;the review 1 Copyright 2010 by ASME
and evaluation of the data obtained; the comparison of calculation results and measured data. Finally the credible result is submitted. Collection and Review of Historical Information The document of the equipment and component manufacturing is used for the calculation of the radioactivity in unachievable parts by determination the material contents. The document of the operation and accident treatment in reactor is used for the evaluation of the contamination spread extant and the occupation exposures of the workers incurred during accident treatment. The document after reactor shut down is used for determination of the transmission of the radioactive waste. To avoid the unnecessary repetition characterization work, the known information should be arranged as clear as possible.. Implementation of the Calculation for Radionuclide Inventory in Different Parts Based on the collection data of the material content and weight, the exposure doze the calculation is performed by use of proper and advanced computer code or other theoretical means to evaluate the inventory of radionuclide. Performance of In Situ Measurements The thorough monitoring is performed for mapping the radioactivity distribution; finding the hot spots; setting the location of the sampling by means of the advanced detection instruments. Sampling and Analysis To save the time and to reduce the cost and the occupational exposure of characterization, less sampling are taken from the optimized location by thorough monitoring. The sampling and analysis is the best way for characterization. The precise monitoring and analysis can be done in the laboratory to determine the type and the radioactivity of the radionuclide. The number of the sampling should be optimized. The shortcoming of the sampling is that the integrity of the equipment is destroyed, so the equipment to take sample should be prepared foe dismantling. Review and Evaluation of the Data Obtained During the characterization process, the data should be assessed and analyzed as early as possible. By the result of the analyses the characterization program should be adjusted in time, the content and item of characterization should be supplemented or simplified. Comparison of Calculated Results and Measured Data To validate the accuracy of the calculation and to increase the confidence in the application of computer codes for future decommissioning project, the result of calculation should be compared with the measurement data. The uncertainties may be introduced by the basic data used for the calculation of neutron flux, such as absorber section of the material, geometric simplify, material content, approximation in calculation modeling. ESTIMATION OF RADIONUCLIDE INVENTORY Following reactor shutdown and discharge of irradiated fuel, the residual radionuclide inventory falls into two categories: neutron activated materials and contaminated material. Neutron Activated Materials These materials are located near reactor core and have been irradiated by neutrons. Neutron activated components make important effect to radioactivity in nuclear facility. In general these parts are difficult to achieve making direct investigation, some time the remote operated tools have to be installed. The characterization mainly is making the approximation as precise as possible. The approximation of neutron activation have to make sure the neutron flux distribution in space and energy, the material content and formation, the operation history of nuclear facility and the time after final shut down. The main activation reaction can be consulted from related literature. 2 Copyright 2010 by ASME
Contaminated Materials Contamination arises from the activation and deposition of the corrosion and erosion products conveyed by the coolant, the dispersion of the fission products leaked from defected irradiated fuel, the leakages in the primary circuit, the processing and storage of radioactivity effluents and wastes, the working incidents in fuel handling. The airborne contamination may give rise to deposit of radioactive substances on walls, ceilings and in the ventilation system. Contamination is of two types, the loose contamination capable of being removed by simple mechanical means and the fixed contamination requiring more aggressive removal methods. Contamination generally accumulates on the equipment surfaces and does not penetrate very deeply except in concrete. Fission products and actinides may be present in the area effected by defected fuel leakages. Relative Importance of Radionuclide with Time The principal activation products present in shutdown reactor are Fe-55, Co-60, Ni-59, Ni-63, Ar-39 and Nb-94 (in steel): H-3, C-14, Ca-41, Fe-55, Co-60, Eu-152, Eu-154 (in Reinforce concrete) and H-3, C-14, Eu-152, Eu-154 (in graphite). In term of radiation levels, Co-60 is the most predominant radionuclide. In the first ten years after shutdown Fe-55 and Co-60 are the major part. In the following 50 years longer lived nickel, niobium and silver isotopes are to dominate. For graphite and concrete leaving the longer lived C-14, Ca-41, Eu-152, Eu-154 are to dominate. After decay periods of more than 100 years, sufficient activity from trace rare earth elements is present. After reactor shutdown for 10-20 years the most abundant radionuclide in contamination residues are H-3, Co-60, Fe-55, Cs-137. For 20-30 years the abundant radionuclide are Ni-63, Cs-137, Co-60, and Sr-90. Parameters Influencing the Radionuclide Inventory The radionuclide composition of the activated and contaminated material may vary in a wide range for different type of reactor. The radionuclide inventory is affected by many factors, such as reactor type, design, power level, neutron flux, operational parameters, operational period and period after shutdown, composition of construction materials, unplanned event. METHODS AND TECHNIQUES FOR The objective of characterization is to provide the condition of safe operation and acceptance for planning and implementation of decommissioning. The methods and techniques are representative calculation, in situ measurements, sampling and analysis. Calculation of Neutron Induced Activity The collected input data are reactor operational history time-power histograms, cross-section data set for given neutron spectra and temperatures, nuclear fuel characteristics, geometry and masses of the components subjected to the neutron irradiated, decay period after final shutdown. In general the neutron transport method is based on either one or two dimensional calculation. In order to accommodate a particular complex geometry, three dimensional modeling may be used to achieve the required accuracy. The calculation methods fall in two categories. The deterministic methods solve the transport equations by applying different mathematical approximations to the treatment of the spatial and energy variables. The stochastic methods employ Monte Carlo and other techniques. The appropriate codes have been developed and used in the nuclear industry for many years. In Situ Measurements Three kinds of in situ measurement may be used for characterization: dose rate measurement, radioactive contamination measurements and relative individual radionuclide activities by spectrometry. The methods of measurement should take into account the geometry, the surface conditions and the nature and 3 Copyright 2010 by ASME
extent of radioactive contaminations. Sampling and Analyses Accurate characterization requires that representative samples be taken from the material to be characterized. The spectrum of radiation from the sample is measured and from this the constituents and their activities are determined, the activity of the material concerned per unit weight can be deduced. The analysis generally requires the use of advanced instruments, such as germanium detectors with high purity, multichannel analyzers, spectroscopy or liquid scintillation systems. The samples are analyzed at reactor site by spectrometer, then comprehensive quantitative radiochemical analyses will be done in the laboratory out of the site. The radionuclide can be separated from the matrix, and then measured with special program. PRINCIPLE OF END STATE End state characterization should be performed after completion of the decommissioning engineering work, the removal of radioactive substance, the decontamination of the site. Special attention should paid to the information of the key area. Sampling can not be taker every where, but the spot (peak value of radioactivity) can not be lost, so it should be undertaken on a statistical sound basis. There are two approaches; (a) Stratified random sampling (random within a survey unit) When the item to be surveyed cannot easily be subdivided into regular grids, a survey unit may be defined as a room or part of a building. (b) Systematic sampling based on a grid system (random within a grid) The facility may be divided into a number of grids. The dimensions of each grid are taken such that the variation of radioactivity across the grid is small in comparison with the variation from grid to grid. Initially, the choice will be based on the information of the operation history and known contamination. The accepted radiological characterization should include detail of the residual activity and should be part of the final report to be submitted to the regulatory body at the conclusion of the decommissioning. GUIDES AND STANDARD USED FOR SITE RELEASE The radioactivity density for site release should comply with the guides of International Atomic Energy Agency and Chinese standard. There are IAEA Safety Standard No.RS-G-1.7 (2004) and No.RS-G-5.1 (2006), and Chinese standard GB18871-2002. QUALITY ASSURANCE IN RADIOLOGICAL Before the implementation of the characterization process, the quality assurance program should be established, to ensure that the gathered data are realistic and reliable, to demonstrate that the results of the calculations are reasonable and that the result of measurement and analyses are sufficiently accurate. The steps of quality assurance are as follows: Acquisition of codes standards for measurement and analyses; Selection of appropriate statistical approach for data collection; Verification that the qualifications of the personnel involved in the characterization; Definition of procedures for data acquisition, recording, evaluation and archiving, Definition of procedures for validation of computer codes; Definition of result interpretation. REFERENCES [1]IAEA, Radiological Characterization of Shut Down Nuclear Reactors for Decommissioning Purposes, Technical Reports Series No. 389, IAEA, Vienna, 1998. [2] IAEA, Monitoring for Compliance with Criteria for Unrestricted Release Related to Decommissioning of Nuclear Facilities, Technical Reports Series No.334, IAEA, Vienna, 1992. [3] IAEA, Applications of the Concepts of 4 Copyright 2010 by ASME
Exclusion, Exemption and Clearance, Safety Standard Series No.RS-G-1.7, IAEA, Vienna, 2004. [4] IAEA, Release of Sites from Regulatory Control on Termination of Practice, Safety Standard Series No.WS-G-5.1, IAEA, Vienna, 2006. [5] 国家质量技术监督局, 中华人民共和国国家标准 GB18871-2002 电离辐射防护与辐射源安全基本标准,2002 5 Copyright 2010 by ASME