Jul 23, 2014 - Radiological Characterization of the former U-extraction facility ..... program at each stage of remediation for the former Uranium ... in the FSU.
Sweden SSM – Ukrainian Ministry of Energy and Coal Industry Cooperation Project ENSURE-2
Hazard Characterization and Safety Assessment of the Building no.103 Alex Buchnea, Oleg Voitsekhovitch, Tatyana Lavriova, Ivan Kovalets, Alexander Khalchenkov, Sergey Todosienko
DRAFT 23 July 2014
EXECUTIVE SUMMARY July 2014 A general Safety Assessment Methodology has been developed within the framework of the ENSURE-2 project in preparation for remediation activities at the Pridneprovsky Chemical Plant (PChP), a uranium production legacy site, located in Dneprodzerzhinks, Ukraine. The objective of a State Remediation Plan, presently under development, is to address and eliminate as much as practically feasible, potential radiation hazards that may be associated with the PChP uranium production legacy site. The focus of this report, in particular, is related to the hazards characterization and safety assessment of one of the most contaminated buildings at this site named ―Building 103‖. This report is developed as a preliminary study to be extended and developed during further Feasibility Studies for the demolition of Building 103 in the framework of the National State Remediation Program. Characterization studies have been performed of the radioactivity in and around Building 103. The characterization studies included measurements of radionuclides in the dust, spills, and aerosols on the three levels of Building 103. In addition, extensive gamma scans were made in the various parts of the building. The building layout is shown below.
General layout of the ―Building 103‖ structural elements 1 2 3 4 5 6 7
Functional Areas Extraction facility, section 1 Administrative premises New melting facility Old melting facility Storage room for containers with extraction products Electricity transformation Storage of flammable materials
Length 42,0 14,0 19,7 24,0
Width 12,0 15,0 23,5 23,5
66,3
12,0
Height 17,200 14,700 20,000 12,800
Area,m2 504 210 463 564
Volume, m3 8669 3087 9259 7614
796
5459
Complex form
5,000
73
365
Complex form
3,600
50
180
The extraction facility is the most contaminated portion of the building on all three levels. The first two levels are contaminated mainly with uranium; whereas the upper level has similar concentrations of uranium and its decay products. The gamma dose rates increase with level and
are highest on Level 3. Spills of tank contents and contaminated dust are found in all extraction areas, with the largest spills of uranium rich residues on Level 1. Areas other than the extraction facility have substantially lower gamma radiation levels and dust concentrations. The ground surface in the area directly adjacent to Building 103, exhibits high gamma radiation fields. These are especially high in the area between Buildings 103 and 102 in the vicinity of two tanks mounted on platforms exterior to Building 103. These radiation fields constitute a hazard for the workers in Building 102 in which non-radiation related work is being conducted. Most tanks and piping within Building 103 have residues remaining. In addition, the windows of Building 103 are broken for the most part and the doors are not secure. This also constitutes a hazard for Building 102 workers from radon and aerosol releases from Building 103 during normal and abnormal conditions. In order to investigate the existing impacts of Building 103 on the workers of Building 102 and the occasional worker entering Building 103, several exposure scenarios were assessed. There were two routine exposure scenarios under normal conditions and three abnormal condition exposure scenarios. The routine exposure scenarios were: A Building 102 worker A worker conducting characterization work in Building 103 The abnormal condition exposure scenarios were: Building 102 exposure under strong wind conditions Exposure of building 103 and 102 workers after a tank spill Unauthorized access of a worker in Building 103 to remove equipment In addition, an hourly exposure rate from various exposure pathways were calculated for workers involved in preparation for decommissioning activities. These were calculated for workers in the extraction facility on Level 1 and Level 3 of Building 103. The SAFRAN model was used for the assessments. The results of the assessment of the two routine exposure scenarios and the three abnormal condition exposure scenarios are shown in the Table. Scenario
Normal Scenario 1 Normal Scenario 2 Inside Building 103 Abnormal Scenario 1 - Strong Wind Abnormal Scenario 2 - Tank Spill Abnormal Scenario 3 - Intrusion
Receptor
Calculated Dose (mSv/a) Normal Conditions Building 102 worker 1.6 Characterization 1.7 Worker D&D Preparation 1 µSv/h – 1 mSv/h Abnormal Conditions Building 102 Worker 0.041 PChP Worker (inside and outside) PChP Worker
Main Pathway(s)
Gamma, Radon Gamma Gamma Aerosols
0.1 to 1
Aerosols
1-2
Gamma
The calculated exposures are clearly above the dose limit for a member of the public, emphasizing the need for remediation. Calculation of the hourly exposure rates for workers involved in decommissioning preparation activities were between 1 μSv/h to 1 mSv/h, emphasizing the need for a well-executed remediation plan and a comprehensive radiation protection and waste management plan.
Table of Contents 1
INTRODUCTION ................................................................................................ VI
1.1 1.2 1.3 1.3.1 1.3.2 1.3.3 1.3.4
Background .......................................................................................................... 1-1 Scope and Structure of the Report ....................................................................... 1-1 Assessment Framework ....................................................................................... 1-3 Decontamination and Decommissioning ....................................................... 1-3 Safety Assessment ......................................................................................... 1-4 Elements of Remediation Strategy................................................................. 1-6 Description of facility, its immediate environment, and remediation activities ......................................................................................................... 1-7
2
ASSESSMENT CONTEXT ................................................................................ 2-1
2.1 2.2 2.2.1 2.2.2 2.2.3 2.3 2.4
General ................................................................................................................. 2-1 Regulatory Framework ........................................................................................ 2-2 General ........................................................................................................... 2-2 International Framework ................................................................................ 2-2 National Regulatory Framework.................................................................... 2-2 Policy and Criteria to be applied to Decontamination and Decommissioning Activities ................................................................................ 2-5 Technical Basis of the Assessment ...................................................................... 2-7
3
SYSTEM DESCRIPTION ................................................................................... 3-1
3.1 3.2 3.2.1 3.2.2 3.3 3.4 3.5 3.5.1 3.5.2 3.6
3.6.4
General ................................................................................................................. 3-1 Description of Uranium Production Legacy Site ................................................. 3-2 Historical context of radiological characterization ........................................ 3-2 Former Uranium extraction facilities ............................................................. 3-6 Demography of the Area and Human Behavioral Characteristics ..................... 3-10 Environmental Baseline Climatic Conditions .................................................... 3-11 Radiological Characterization of the areas surrounding Building 103 .............. 3-12 Radioactive contamination of the adjacent area .......................................... 3-12 Radiological conditions at the territory of the industrial site ....................... 3-16 Radiological Characterization of the former U-extraction facility ―Building 103‖ ................................................................................................... 3-19 General ......................................................................................................... 3-19 General layout of building and structural details ......................................... 3-20 Radiological Characterization the former U-extraction areas in the Building 103................................................................................................. 3-21 Description of the other areas of Building 103 ............................................ 3-31
4
ASSESSMENT METHODOLOGY .................................................................... 4-1
4.1 4.2 4.3 4.3.1 4.3.2
SAFRAN .............................................................................................................. 4-1 Application of SAFRAN to Existing Situation at Building 103 .......................... 4-1 Routine Exposure Scenarios for Existing Condition of Building 103 ................. 4-3 Routine Exposure Scenario 1 - Worker in Building 102 ............................... 4-3 Routine Exposure Scenario 2 - Worker Conducting Characterization Activities in Building 103 .............................................................................. 4-3 Workers Preparing for Building D&D Work................................................. 4-5 Abnormal Exposure Scenarios for Existing Conditions in Building 103 ............ 4-5 Selection of Postulated Initiating Events ....................................................... 4-5 Development of Abnormal Exposure Scenarios ............................................ 4-6
3.6.1 3.6.2 3.6.3
4.3.3 4.4 4.4.1 4.4.2
5
PRELIMINARY DOSE ASSESSMENT ............................................................ 5-1
5.1 5.2 5.3 5.4
5.5.2 5.5.3 5.5.4 5.6
Safety assessment context .................................................................................... 5-1 Estimated Dose to Building 102 Worker - Normal Conditions ........................... 5-1 Estimated Dose to Characterization Worker........................................................ 5-2 Estimate of hourly exposures of workers involved in different elements of future decommissioning preparation activities ................................................ 5-4 Estimated Dose to Building 102 Worker During Abnormal Scenarios 1 and 2 ..................................................................................................................... 5-6 Aerosol and Rn concentrations at Building 102 under accidental releases from Building 103 ............................................................................ 5-6 Atmospheric dispersion assessment methods ................................................ 5-8 Results of the dispersion calculations ............................................................ 5-9 Dose Calculations for Abnormal Scenarios 1 and 2 .................................... 5-12 Estimated Dose for Abnormal Scenario 3 - Intruder ......................................... 5-13
6
CONCLUSIONS.................................................................................................. 6-1
7
TASKS FOR COMPREHENSIVE ASSESSMENT AND FURTHER STUDIES ............................................................................................................. 7-3
8
REFERENCES .................................................................................................... 8-1
9
ANNEXES ........................................................................................................... 9-1
5.5 5.5.1
List of abbreviations ALARA ARPANSA - Australian Radiation Protection and Nuclear Safety Agency BSS – Basic Safety Standards CNSC - Canadian Nuclear Safety Commission CREM - Center for Radioecological Monitoring (City of Zhovty Vody) DMK - Dneprovsky Metallurgical Combine enterprise D&D decommissioning and dismantling (demolishing as a part of remediation) d.w. - dry weight FEPs – Features, Events and Process GEK – Geo-Eco-Consulting Company (Kiev) EIA – Environment Impact Assessment EIS – Environment Impact Statement IAEA – International Atomic Energy Agency ICRP – International Commission on Radiation Protection MPC – maximum permissible concentration (in drinking water) NORM – Naturally Occurring radioactive Materials NIPIPT – Research Institute of Industrial Technologies (City of Zhovty Vody) NRBU – 97 - Norms of radiation safety of the Ukraine (from 1997) NUBIP – National University of Bio-resources and Environmental applications PChP - Pridneprovsky Chemical Plant (State Enterprise) PHMP - Pridneprovsky Hydro-Metallurgical Plant (State Enterpise) RPP – Radiation protection Plan SA – Safety Assessment SE - State Enterprise SSR – Safety Series Reports (IAEA) TDS – total dissolved solids UHMI - Ukrainian Hydrometeorological Institute UNSCEAR - United Nations Scientific Committee on the Effects of Atomic Radiation
1 1.1
INTRODUCTION Background
Radiological Impact Assessment (RIA) is a tool, which can help in the development of Remediation strategy and the development of a Radiation Protection Plan (RPP) for demolishing the former Uranium extraction facilities at the site of the former Uranium production Facility Pridneprovsky Chemical Plant. Radiological Impact Assessment in the context of an RPP for demolishing former uranium production facilities at this site has to be considered as an element of the overall Environment Impact Assessment and Safety Assessment. The objective of the Remediation Plan is to address and eliminate as much as practically feasible, potential radiation hazards that may be associated with structural building demolition activities at the PChP uranium production legacy site, located in Dneprodzerzhinks, Ukraine. The focus of this report, in particular, is related to the one of the most contaminated buildings at this site named ―Building № 103‖. The operator of this facility is the State enterprise ―Barrier‖, which belongs to the Ministry of Energy and Coal Industry and has responsibility for safe management of the uranium production facilities remained at the territory of PChP and should also be responsible for the State Remediation Program implementation at the legacy site. The objective of this document is to demonstrate the application of the general Safety Assessment Methodology, developed within the framework of the ENSURE-2 project to the specific tasks such as hazards characterization and safety assessment of the of the former Uranium extraction facility. Proper assessment of the main hazardous elements and key exposure pathways based on the radiological surveys carried out recently in Ukraine will assist in establishing a safe work environment for SE‖ Barrier‖ employees and their sub-contractors, and provide guidance to operations personnel with respect to anticipating, recognizing, evaluating, controlling and eliminating radiation exposures in their work. Responsible Authorities and SE ―Barrier‖ as a supervised uranium legacy site operators, shall ensure that radiation exposures to its workers are maintained below regulatory limits and deliberate efforts are taken to further reduce exposures and releases as low as reasonably achievable (ALARA). Thus, the development of a Radiological Safety Assessment is an important stage of the Remediation Planning. The experience gained during the cooperative efforts of EC (SM) and UA experts in the preparation of this study can help in implementing an adequate radiological control program at each stage of remediation for the former Uranium extraction facilities and in particular Building №103 at the territory of Pridneprovsky Legacy Site. This report is developed as a preliminary study and should be extended and developed during further Feasibility Studies for the demolition of Building №103 in the framework of a National State Remediation Program. 1.2
Scope and Structure of the Report
The report assumes a basic understanding of radiation and the effects of exposure to ionizing radiation on human health. If more information is needed on these subjects, the interested reader is referred to safety standards and safety guides such as IAEA, 2005; IAEA, 2008; IAEA, 2009; IAEA Safety Reports Series №77 (2013) or other publications describing best international 1-1
practice, for example documents available from Gunar Uranium Legacy Mill Demolition Plan, where remediation actions were successfully implemented during recent years in Canada (2012). The main purpose of the report as outlined is to present the safety assessment for the Building №103 and associated facilities, in past one of the main radiochemical uranium extraction facilities in the FSU. This assessment is not exhaustive but will require supplementary more detailed studies. However, even at this time, project partners in cooperation with SE ―Barrier‖ were able to provide studies and analysis of this facility and surrounding areas sufficient to serve as an example of how assessments of such safety cases and for these types of facilities can be performed. The report is structured as follows:
Section 1 is an introduction. Section 2 provides the assessment context: an overview of the regulatory framework, some policies and radiation safety criteria, which to compare the results of safety assessment, and the technical framework for the safety assessment for the building facility. Section 3 provides a summary description of the system based on site characterization and monitoring data, of the former uranium extraction facility itself, including some key hazardous elements of the facility, as well as adjacent areas; Section 4 shows the assessment methodology including the scenario development and justification process followed, Section 5 shows the preliminary results of dose assessments for the various scenarios, including contaminant dispersion calculations, and Section 6 suggests some preliminary conclusions, and recommendations for further studies and assessments needed.
1-2
1.3
Assessment Framework
1.3.1 Decontamination and Decommissioning The IAEA Glossary characterizes safety assessment (SA) as an assessment of all aspects of practice that are relevant to protection and safety for an authorized facility, including siting, design and operation of the facility. In application to remediation options analyses Safety assessment should demonstrate radiological/environment and economically efficient way for implementation of the “decontamination” and “decommissioning” (D&D) of the hazardous facilities, which also can include legacy of the uranium ore mining and milling. ―Decontamination‖ is an activity which refers to the removal or reduction of radioactive and/or other hazardous contamination from facilities, including structural and non-structural materials and equipment. The decontamination activity can take place at any point in the decommissioning (or demolition). Generally a pre-cursor to decontamination is the characterization of the radioactive hazards both before and after the decontamination process to determine the risks associated with the level of contamination. The objective is to reduce radiation risk and exposure to a level that is protective of public health and safety, worker health and safety, and the environment. Decontamination technologies include chemical, electrochemical, and thermal processes as well as mechanical cleaning, washing, and other techniques. Decontamination methods may include the use of remote techniques that reduce the risk of worker exposure, in situ decontamination methods that reduce the generation of secondary wastes or reduce the requirement for waste handling and processing, and methods for decontaminating inaccessible areas. For example, decontamination can be a stand-alone operation conducted at a facility that is in operation and will remain so after the decontamination is completed. It can also be an operation closely associated with, and often preceding, decommissioning. It should be noted that some definitions of decommissioning include decontamination. It should also be noted that, in addition to contamination by radiological material, there is frequently contamination by chemicals or other hazardous materials that also must be dealt with, usually in concert with the radiological material. Though this document is focused on radiological decontamination, maintaining awareness of non-radiological, hazardous contaminants that are coextensive with the radiological contaminants is extremely important, especially since removing both types of contaminant in a single waste stream may result in mixed waste (MW), giving rise to serious complications in subsequent management and disposition. Decontamination is often followed by decommissioning. ―Decommissioning‖ is a process and refers broadly to actions taken at the end of the life of a facility to retire it from service. The objectives are to enable reuse or safe disposition of facilities and equipment. For radioactive contaminated facilities, the decommissioning process generally incorporates some or all of the following activities: deactivation and safe management of radioactive and other wastes, plant dismantling, demolition, and site remediation. Following successful decommissioning, residual contamination may require monitoring, institutional controls, and maintenance. Depending on the situation, the site may be released for appropriate alternative used. It should also be noted that the term ―D&D‖ (decommissioning & decontamination) is widely used in the literature to refer to a number of combinations of this terms associated with the general 1-3
decommissioning process—decommissioning, deactivation, decontamination, demolition, dismantlement, disposition—and used in many considered cases studied for decommissioning and demolition experience and it is used in application to the overall process of decommissioning. The typical decommissioning framework, which describes general steps for Decommissioning and Demolition Planning according recommendation in [ ] are the following:
Prepare decommissioning (demolition) scope document Develop Characterization Plan, Conduct Characterization of the facility Conduct Risk Assessment and Safety Analyses Inform Regulator, Stakeholders, Public Evaluate Decommissioning Alternative and Propose Response Final decision selection, Prepare Decommissioning (demolition) Plan Conduct Preparedness Actions and Review Conduct Decommissioning (demolition actions) Decontamination of the site Close-out Reporting Long-term Surveillance
During all these actions steps the site specific surveillance, monitoring and maintenance have to be conducted. As it was recommended in [ ] and many other studied case studies Risk Assessment and Safety Analyses as similar terms to the Integrated Safety Assessment (SA) and Environment Impact Assessment (EIA) are considered as the key elements of the assessment in justification of the remediation alternatives. There still no special methodology for SA and EIA developed for remediation of the former uranium facility and in particular for demolition of Buildings and facilities used for Uranium Production. However, Safety Assessment methodology for other type of facilities (IAEA Publication SRS № 77 (2013) or in particular SA methodology developed for ―Decommissioning of Facilities Using Radioactive Materials‖ (IAEA Safety Guide –WS-G-5.2 (2008) can be applied also for considering process with demolition of the highly contaminated building with former Uranium extraction facilities and equipment. 1.3.2
Safety Assessment
As one of the basic regulatory requirements, a graded approach to any safety assessment shell be used in determining the scope and level of details of the safety assessment carried out for any particular facility and activity, consistent with the magnitude of the possible radiation risks arising from the facility and activity. The inherent nature of the framework is systematic and structured, and provides for the continual improvement of the assessment through an iterative process. It consists of seven interrelated elements. The extent to which each element is incorporated into the assessment depends on the purpose, scope, and focus of the assessment. With some modification the IAEA ISAM (Integrated Safety Assessment methodology ), which was mainly developed for the safety assessment of near surface radioactive waste disposal facilities (IAEA, 2004b) can be also considered as a model for application in this particular assessment. 1-4
Small conceptual variations were introduced to make it more suitable for mining and mineral processing operations in general. However, the focus of this report is on the radiological impact assessment of a contaminated legacy site. While the framework is suitable for the assessment of these sites, the decision making framework for their remediation still needs to improve. The first element in the Figure 1.1 is the definition of the assessment framework (context), which is the purpose of this report. The purpose of this section is to describe the more practical terms the remaining elements of the assessment framework. Reference to the ISAM methodology and how each element is applied will be made as appropriate. A more detailed description of each element is provided in IAEA (2004b) and also Publications IAEA (2008) and IAEA (2013). Definition of Assessment Context • • •
Regulatory Framework Technical Basis of Assessment Assessment Philosophy
System Description
Process Level Modeling • • •
• • • •
Air Quality (PM2.5, PM10, TSP, Radon) Groundwater Flow and Mass Transport Surface Water Hydrology
Operational Processes and Facility Infrastructure Environmental Baseline Conditions Demographical and Human Behaviour Radiological Baseline and Conditions
Scenario Development and Justification FEPs Analysis
Scenario Definition
Exposure Conditions Development Source-Pathway-Receptor Analysis
Exposure Condition Definition
System Level Model Development Conceptual Model • • •
Define Safety Functions Description Interaction Matrix
Computer Model Implement in Appropriate Software
Mathematical Model
Consequence Analysis Source Term Analysis
Exposure Pathway Analysis
Dose Assessment Uncertainty and Sensitivity Analysis
Review and Modify
Yes
Integration of Safety Arguments • • • •
Interpretation of assessment results in terms of assessment context Evaluate reasonable level of assurance Management of uncertainties Application of management system
Rejection Make Decision
Effective to Improve Assessment Components
No
Acceptance Make Decision
Yes AquiSim Consulting (Pty) Ltd
Safety Adequately Demonstrated?
No
Figure 1.1 - Schematic illustration of the safety assessment framework, within which the safety assessment will be performed 1-5
The conceptually illustrated framework on Figure 1.1 is applicable for assessment of the demolition and decontamination scenarios of the Building 103 with some simplification of the assessment scenarios. This report will mainly consider the NO Action scenarios, in particular, possible existing impacts of the radioactive sources located inside of the Building 103 on personnel entering the building and on the adjacent areas. 1.3.3 Elements of Remediation Strategy The IAEA document TRS № 442 «Remediation of Sites with Mixed Contamination of Radioactive and Other Hazardous Substances» provides insight into the decision making process during the EIA and safety assessment in remedial actions (Figure 1.2).
Figure 1.2 - Phases of decision making process in remediation strategy and technology planning according IAEA TRS-42 1-6
As can be seen there are numerous other factors involved in this process, which may lead to successful implementation of the remediation actions. These principles and actions have to be taken into account when developing Safety cases for the decommissioning of the high contaminated Building 103. In this particular study only a limited set of scenarios and impact assessments are considered. These include the assessment of impacts of the hazardous materials accumulated in the Building 103 on the workers, who potentially will be involved in preparatory D&D activities (e.g. characterization) and also to the workers of enterprises at the adjacent areas in case of NO ACTION strategy options. Specific safety assessment and environment impact analyses for each activity to be planned in the decontamination and decommissioning action plan will be carried out later as a separate study.
1.3.4 Description of facility, its immediate environment, and remediation activities Contaminated legacy sites or even some particular facilities (for example, former Uranium extraction facility as a Building 103 at PChP) can be very complex, consisting of various interrelated features, processes and activities that have to be considered in an integrated manner with prevailing environmental site conditions. Furthermore, quantifying the potential radiological impact to human beings requires knowledge about their behavioural characteristics. The contaminated site, the prevailing environmental conditions, and the nearby human behavioural conditions are collectively viewed as an integrated system. The purpose of the system description in the assessment framework is thus to describe these main components of the system and its associated subcomponents in sufficient detail for the purpose of the safety assessment. The context of the assessment and regulatory requirements are described in the Chapter 2. A description of the facility and its surrounding environment is given in the Chapter 3. Safety assessment should consider the overall impact of the facility as well as the impacts of its key structural elements taking into account the specific physical, chemical, and radiological characteristics (including equipment and material contained or dispersed) of each structural elements of the facility. The remedial options have to be considered for each particular element of the facility as well.
1-7
2 2.1
ASSESSMENT CONTEXT General
The first step in the assessment framework illustrated in Помилка! Джерело посилання не знайдено.1.1 is the definition of the assessment context, which in simple terms define the basis or context within which the safety assessment will be conducted. The assessment context serves as a communication tool and provides the means by which stakeholders or the target audience are informed of what is included or excluded from the assessment and justification for the choices made. Viewed from this perspective, the context document defines the boundary conditions within which the assessment will be performed. This includes the regulatory framework that applied to the assessment, the technical basis of the assessment (e.g. purpose, scope and focus of the assessment), spatial and temporal boundaries, the assessment endpoints, and the assessment philosophy that will be applied. Site characterization and monitoring data serves a basis for exposure pathways and (SA) safety assessment (analyses) and environment impact assessment (EIA). An environment impact assessment (EIA) evaluates a number of options to remediate (D&D) of the hazardous facility or contaminated territories. EIA includes consideration of a ―NO ACTION‖ alternative. In case of remediation (decommissioning & decontamination), the no-action alternative is considered as a baseline option with which other strategies to be considered can be compared. Safety assessment in the case of environment impact assessment (EIA) may help choose an optimal sequence of remediation steps such that the exposure to workers and population as well as affected environment within each step of the remediation can be minimized. Safety assessment can assist in selecting the optimal options between several environmentally similar choices allowing minimization of the occupational exposure during and after remediation options. This report is limited to a consideration of the ―NO ACTION‖ alternative. Section 2.2 discusses the regulatory framework, in particular, the existing framework within Ukraine. Remediation criteria to be used in the present D&D activities are given in Section 2.3.
2-1
2.2
Regulatory Framework
2.2.1 General The regulatory framework for the assessment is defined by a combination of national legislation, and regulations, requirements and guidance defined in terms of the legislation. Depending on the maturity of the national regulatory framework, guidance may be sought from international organisations concerned with radiation protection and the management of radioactive waste. Finally, principles, requirements and guidance for the assessment of contaminated legacy sites as applied in other countries may complement the regulatory framework. The purpose of this section is to define the regulatory framework as basis for the safety assessment. Section 2.2.1 provides an overview of the general international framework for radiation protection and radioactive waste management, including criteria on doses and time frames for uranium residues used in some countries. Section 2.2.2 is devoted to an overview of national regulatory framework for the management of radioactive waste from mining and milling of ore. 2.2.2 International Framework The international radiation protection framework for the nuclear, medical, and mining industries is well established and recognised. Organisations that play a key role in this regard include (IAEA, 2004a) the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR), the International Commission on Radiological Protection (ICRP), and the International Atomic Energy Agency (IAEA). Specific requirements and guidance to be considered as basis for regulatory provisions of the sufficient planning and implementation of the remediation works are the following: duties and responsibilities for the legacy site management, monitoring and remediation; safety and remediation criteria (safety and environment objectives); requirements and guidance for remediation planning, radiation protection plans, waste management plans, safety assessment and safety management, site characterization and monitoring, project review, completion reports and inspections; staff qualification, licensing process, stakeholders involvement, inspections and long term surveillance and legacy site management reporting among others. All these issues are considered in a separate part of the ENSURE reporting materials, developed in the frame of WP-1 of the project. 2.2.3 National Regulatory Framework The legislative and regulatory situation concerning uranium legacy sites is still under development in the Ukraine. Currently, a mix of old (Soviet period) and new regulations and legislation exist that have not yet been fully harmonised. The review of the Ukrainian regulation has been carried out in the ENSURE-I Project in 2008. Though a number of draft regulations have been prepared since the previous review, the situation has not changed significantly and, in principle, the ENSURE-I Project review still applies to the current situation in the Ukraine. The main law related to uranium mining is the Law about ―uranium ore mining and milling‖ (1998). In accordance with this Law, the designs for decommissioning, re-profiling and 2-2
rehabilitation of uranium facilities should be justified with regards to their possible secondary economical use. Guarantees must be provided for radiation safety, possible decontamination of former facilities and territories and the proper covering and storage of radioactive wastes and residues of uranium ores. The general radiation protection limits established in Ukraine are quite similar to the international safety standards and have been established now in the two key documents on radiation protection: Norms of radiation safety of the Ukraine NRBU-97 (1997) with amendments related to radioactive waste management issued in 2000 and Main Sanitary Rules OSPU-2005 (2005). Chapters 16 and 17 of OSPU-2005 consider radiation safety rules in the case of operations with technologically (artificially) enhanced NORM as a source of exposure. These rules are applicable to all steps in the mining of uranium and thorium. Workers in such operations are considered to be employed in enterprises dealing with radiation technologies. Such workers are considered to be ―Category A‖ (according to NRBU-97). The rule 17.3 (OSPU-2005) also considers other workers who are not directly involved in radiation technology enterprises but who can be potentially exposed to NORM. Such workers are considered to be ―Category B‖ (according to NRBU-97). The Ukrainian radiation protection framework is generally consistent with new IAEA BSS (Basic Safety Standards, 2011) requirements (Appendix IV), ICRP Publication 64 (Protection from Potential Exposure: A Conceptual Framework, 1993) and ICRP Publication 76 (Protection from Potential exposures: Application to Selected Radiation Sources, 1997). At the moment, Regulatory Authorities in the field of radiation protection (State Inspectorate for Nuclear and Radiation Safety) have elaborated plans for modifying many regulatory norms and national standards to be in compliance to the recommendation of ICRP-101 and 103 and also the recently established IAEA New Basic Safety Standards. However till now the regulatory framework existing in Ukraine for the artificially enhanced NORM sources of human exposure do not provide detailed specifications for problems related to past uranium ore mining/processing activities and their consequences. In particular there still are no clear requirements to help in establish remediation and design criteria for tailing compounds and decontamination of the uranium legacy sites. The requirements for remediation planning and review are also missing in the regulatory framework and there are no clear recommendations on the contents of the safety assessment and environment impact assessment or specific requirements for site characterization, inventory studies, monitoring and surveillance. Clear requirements for uranium legacy site management as well as for licensing of the site operator for uranium production legacy site management and remediation projects implementation are also required. During the recent several years, within the framework of the State Remediation Program, which was developed in 2008 and partly implemented during period 2009-2012, there were several attempts to fulfil the gaps in regulatory requirements and safety criteria and remediation requirements. The first attempts of national experts were to modify and improve remediation requirements for former uranium facilities, which were established in Ukraine during the Soviet period and are still active. Regulatory Norms, such as SP-LKP-91 (―State sanitary-ecological norms on radiological safety concerning decommissioning, temporary closing-down and change of activity profile of 2-3
facilities for production and reprocessing of uranium ores‖) were developed. The new version of the slightly modified document SP-LKP-91 was developed by SE ―Barrier‖ in 2004; however, the modified version of the Sanitary Roles….was not approved by Regulatory Body as it was not really consistent with internationally accepted practice. During 2005, experts from Institute of Hygiene and Medical ecology (Los, Pavlenko, Soroka, et al, 2005), at the request of Regulatory Body and State Enterprise Barrier, developed requirements and recommendations for reference levels and safety criteria for residues and contamination of the environment to be applied to remediation activities planned for SE ―Pridneprovsky Chemical Plant‖. On a basis of this document, the Ministry of Health produced the following document ―Guidance on regulations regarding remediation of former Uranium production facilities at the Pridneprovsky legacy site‖ (―Методичні вказівки щодо радіаційно-гігієнічного регламентування робіт на уранових об’єктах колишнього Придніпровського хімічного заводу‖. МВ 6.6.1.2.6.-136-2007). This document provides mainly dose limits and reference levels for workers and the public in the Uranium legacy site affected areas as well as giving recommendations for establishing site specific monitoring activities at the former ―Pridneprovsky Chemical Plant‖ legacy site territory and surrounding areas. According to this document (MB 6.6. 1.2.6.-136-2007), the following dose limits are currently applicable to the PChP Site:
20 mSv.a-1 for Category A Workers (staff from the Barrier Enterprise); 5 mSv.a-1 for Category B Workers, i.e. for any other Enterprises staff who are working in this territory but not involved in waste management, decontamination or monitoring processes; and 1 mSv.a-1 for a Public.
Reference levels for inhalation, water consumption, and criteria for re-use of materials and scrap metals were also established in Annex 2 of this document. Expert review of this document, indicated that, in many parts, recommendations are very prescriptive and basically reflect old radiation protection approaches that are not in compliance with more modern radiation protection regulations (such as NRBU-97 and OSPU-2005). Many other very important requirements and criteria are simply missing from this document thereby making the recommendations insufficient for practical use. Several other documents were developed during recent years:
“Exemptions and clearance of radioactive materials for practice” (Порядок звільнення радіоактивних матеріалів від регулюючого контролю у рамках практичної діяльності (НП 306.4.159-2010. (Держатомрегулювання, Київ 2010); “Requirements and safety conditions for decommissioning and remediation of the former uranium facilities” (Вимоги щодо забезпечення безпеки виводу із експлуатації і приведення до безпечного стану об’єктів колишніх уранових виробництв». Draft, 2011) “Recommendations for compliance with radiation safety for the period of implementation of rehabilitation measures in the former industrial site of the Pridneprovsky Chemical Plant” (Рекомендації щодо дотримання радіаційної безпеки 2-4
на період здійснення реабілітаційних заходів на території колишнього проммайданчика ВО «Придніпровський хімічний завод». Draft 2012). These documents were prepared taking into account the best international practice and are based on the recommendations of ICRP Publications №101, №103 and №104 and also relevant IAEA documents. Review of these documents was one of deliverables of the ENSURE-2 project. The modified version of the document on ―Requirements and safety conditions for decommissioning and remediation of the former uranium facilities‖ (―Requirements…‖) was drafted in 2012 with supporting funding and expertise from ENSURE-2 project, but has not been officially approved yet. The ―Requirements…‖ document foresees that, in an early phase of decommissioning planning, an Inter-Agency Commission is to be established, which includes authorized representatives of Regulatory Authorities, Operator, local authorities and other stakeholders. The Commission considers, evaluates and approves the proposed remediation strategy, actions to ensure radiation safety and radiation protection and establishes criteria to evaluate the effectiveness of the remediation measures and controls their implementation. According to the ―Requirements…‖ document, the Conceptual Design for remediation of the uranium facility should include a Safety Assessment (SA) evaluating expected radiation risks for personnel and public (as the basis for evaluating the effectiveness of the applied remediation measures). The SA scenarios should consider the current condition at the site, as well as conditions during implementation and after completion of the remediation measures. The risk assessment procedures should be applied also to potential abnormal or accident cases including extreme natural conditions (weather, hydrological or geotechnical). Assessments have to be performed for the critical group of the population or for reference persons. The optimal remediation solution should be based on an optimization considering radiation safety criteria to be achieved and also the economic cost and social benefits of the remedial actions. It is expected that the ―Requirements…‖ regulatory document will be enacted in the Ukraine in 2013-2014. One other relevant and important regulatory document, which is currently under development, is entitled ―Recommendations for compliance with radiation safety for the period of implementation of rehabilitation measures in the former industrial site of the Pridneprovsky Chemical Plant‖. This document is in its early stage of drafting and review by national and international experts and authorities. Section 3 of this document contains a set of remediation objectives such as end-state criteria to be achieved as a result of remediation, reference levels for exposure and exemption levels for materials and secondary use. Remediation criteria proposed in this document were based on the analyses of experience gained from number of projects funded by EC FARE Programs during 1995-2004 in Czech Republic, Bulgaria, Slovenia and Romania. 2.3
Policy and Criteria to be applied to Decontamination and Decommissioning Activities
The proposed State policy and proposals for regulatory requirements and remediation objectives as a part of the State Remediation Programs of Ukraine are summarized below. The strategic aims of the Remediation Program are the following:
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The PChP U-production legacy site must be safe for workers at this territory and also for the public and the surrounding environment as well as for industrial use and further development in conjunction with expected economical PChP site re-vitalization in the future. The concept of the currently executed State Remediation Program (2010-2014) is based on the policy: Areas of contaminated lands and areas occupied by uranium residues have to be significantly reduced or put into a safe state on an economically reasonable basis:
Impacts and risks from the identified hazards have to be optimally minimized in compliance with the established radiation safety and remediation objectives; End-state remediation criteria (radiological, non-radiological) must be established and approved first by an Inter-Agency Commission established by the Government; Emission barriers (tailing covers) and engineering protective facilities have to be stable for a long-term period of at least 200 years (in case of temporary actions for at least 50 years); The waste management should be carried-out according to best international practice. The Legacy Site and Remediation Program should be operated and managed by specialized companies, using a multi-disciplinary partnership mechanism. The consolidated Action Plan for Long-term Remediation should be developed in close cooperation between Regulatory Bodies and the Ministry of Fuel and Coal Industry; The existing specialized state enterprise ―Barrier‖, should be reorganized, according to its primary functional tasks. The primary task of this enterprise is management and implementation of the State Remediation Program; The strategy for the long-term land use during the post-remediation period should be developed, taking into account Public opinion; Information systems and data records keeping should be developed taking into account best international practice.
The main hazards in order of risk estimated priorities for remediation (decontamination, demolition) are the following:
former Uranium extraction facilities (Buildings 103 and 104), tailings dumps at the territory of PChP such as (―Zapadnoe‖, ―Central Yar‖, ―Dneprovskoe‖), remaining elements of Uranium production, storage and transportation; the contaminated territories and affected environment at the areas adjacent to the former Umilling and extractions facilities).
Radiological criteria to be achieved:
Individual dose exposure for personnel of any enterprises at the site after its remediation to be less than 1 mSv/y. The dose constraints for public living in surrounding areas to be less than 0.3 mSv/y above local background exposure. Unrestricted land use if the specific activity of the soil is less than 0.2 Bq/g of Ra-226 in equilibrium with all nuclides of the Uranium-Radium decay chain. For restricted use, e.g. as
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2.4
grassland or forest or for industrial purposes, a maximum specific Ra-226 activity of 1 Bq/g is recommended. The gamma dose rate at the territory (averaged is below 0.5 µSv/h at the industrial site and less than 0.3 µSv/h (Background +0.2) at the vicinities around the tailing and Uranium Legacy facilities Buildings with a ―non-removable‖ surface contamination less 0,5 Bq/cm2 ( natural uranium) can be used commercially or industrially. In construction debris from demolition work with less 0.2 Bq/g Ra-226 can be releases without restriction (scrap collection or recycle). At specific activities between 0.2 Bq/g and 1.0 Bq/g dumping onto contaminated areas follows for which unrestricted release is in any case not planned (heaps and tailings pond). At specific activities more than 1.0 Bq/g, dumping is to be studied from a radiation protection viewpoint. Scrap with a surface contamination less 0,5 Bq/cm2 (after cleaning) can be released for smelting. Radionuclides limitation to be established (control levels) established for all other environment pathways (surface alpha-beta contamination, dust, injections and direct gamma radiation etc. for specific areas and facilities for period of remediation), which will be derived from established overall limit for the individual dose. Special criteria to be established for surface and groundwater (consultation and assistance are needed to establish realistic remediation criteria according to ALARA) Design criteria and objectives to be developed for each specific project and approved by the regulatory body. The design requirements for safe demolition of the contaminated buildings and dismantling of the large and highly contaminated former uranium facilities/equipment should be developed. Technical Basis of the Assessment
The proposed action is the decontamination and demolition of Building 103, known as the Radiochеmical Extraction Facility (see description in the Chapter 3). The scope of the proposed action involves the following:
decontamination or removal of fixed radiological contamination within the building prior to demolition; the demolition of all interior tanks, pipelines, mechanical, electrical, and architectural components; the open-air demolition and removal of the physical structure, including the concrete foundations, sidewalk and asphalt surfaces adjacent to the facility; removal of contaminated soil in the vicinity of the building; and transportation of waste to approved disposal facilities.
All activities would be performed in accordance with design, which is currently under preparation contracted by SE ―Barrier‖. The assessment carried out in this document will be considering as a preliminary assessment of the initial safety condition of the radiation sources associated with former U-extraction facility Building 103 at the territory of former SE ―Pridneprovsky Chemical 2-7
Plant‖. The initial assessment will include the scenario ―NO Action‖ as a basis for further assessment and also the impact on the workers of the enterprises, which located in close vicinity to the Building 103. Section 3 describes the type and location of fixed and loose contamination, radon and suspended particulate concentrations and gamma radiation levels. The main exposure pathways which have to be considered in the assessment are direct gamma radiation and inhalation due to Rn-222 and highly contaminated dust covered all surfaces of the building. The basis of the safety assessment is the evaluation of doses to affected and potentially affected people. These are evaluated taking appropriate account of uncertainties, missing data, and missing knowledge. The intent is to provide a set of analyses that illustrate the trade-offs between risks to several different classes of affected people. Doses to workers entering the premises for characterization activities and other (non-radiation, or Type B) workers engaged at different enterprises in the area will be evaluated. Doses would be evaluated both for existing condition in the building and also under possible accident conditions. In the future assessment, the possible impacts from the Building 103 and planned remediation activities to the public, living today in Dnieprodzerchynsk (infants, children, adults) engaged in normal daily activities will be performed as well. The above endpoints will be calculated for the no-action alternative.
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3 3.1
SYSTEM DESCRIPTION General
The industrial complex for hydrometallurgical and chemical production at the former ―Pridneprovsky Chemical Plant‖ was a multifunctional complex enterprise, occupying more than 150 buildings and industrial facilities located on both the ―southern‖ and ―northern‖ parts of the industrial site. The majority of contaminated sites and remaining former facilities as well as all uranium tailings dumps are located in the ―southern‖ part of the industrial site. The Uranium extraction facilities were located at several main processing buildings and adjacent supplementary production facilities in this territory. The main former uranium production facilities at the territory of PChP Uranium Legacy Site are the Buildings 103, 104, 2b and some others. The current conditions in these buildings have been studied during recent years and indicate high contamination by spilled uranium ore containing materials, production residues (yellow cake) and other spilled NORM materials and radioactive dusts creating a risk from dispersion by air pathways by distributing the buildings’ radioactive pollution to the surrounding territories. The former Uranium extraction facility ―Building 103‖ is the most highly contaminated building. A general overview of the buildings, the remaining structures and contaminated equipment are given in the Annex 1 to this Report. The gamma dose rates near some tanks, which still contain remaining residue materials of radiochemical extraction in this building was found in range between 10 to 1200 µSv/hr. The present remediation plan is to demolish the walls, slabs, tanks and other structures. Dismantled equipment and all metal structural items will be categorized and sorted by different categories of radioactive waste or industrial wastes which may be re-cycled. Other residues will be disposed in the tailing bodies or in special storage facilities. The strategy for waste management is still not clear. The building foundation will be left in situ and covered with non-radioactive debris and local soils. All surrounding areas will be decontaminated. However, till now there is still no final decision regarding remediation objectives to be achieved upon clean-up of the areas adjacent to Building 103. The results of preliminary site characterization data and first radiological assessment of the former Uranium extraction facility and affected areas are described in this report.
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3.2
Description of Uranium Production Legacy Site
3.2.1 Historical context of radiological characterization Uranium mining was intensively developed in Ukraine from the end of the 1940’s to the beginning of the 1990’s. The State Industrial Enterprise Pridneprovsky Chemical Plant (PChP) was one of the largest metallurgical facilities in the former Soviet Union, and processed uranium ore from 1948 until 1991. During that time, uranium extraction was carried out using raw ore products delivered from Ukraine, Central Asia, Germany and the Czech Republic. In addition to imported ores, the PChP processed uranium-bearing sludge obtained from smelting of iron ores from the Uranium Mines of Ukraine. In the early 1990’s, due to the disintegration of the former Soviet Union and consequently their uranium industry, the PChP was split into several separate enterprises and processing of the uranium ore was terminated. During operations, nine tailing impoundments were created containing about 42 million tonnes of uranium extraction residues, with a total activity of 3.2 x 1015 Bq (IAEA, 2002). Significant amounts of highly contaminated equipment and scrap metal are still in the buildings of the former uranium extraction facilities, located in the territory of the industrial zone of Dnieprodzerzhinsk (Figure 3.1). Other residues were disposed about 14 km to the south-east of the site at the Suhachevskoe (cell-1 and cell-2) tailing dumps. This uranium production legacy site has not been properly decommissioned since operations ceased in 1991. The tailing dumps, contaminated buildings and pipelines are still present and create potential risks of radiation exposure for personnel and people living in the surrounding areas. In 2009, about 20 other enterprises were in operation at this territory. Most of the enterprises make use of this area, but are not related to the former uranium processing activities. However, their workplaces are close to the highly contaminated tailings, or buildings that were used for ore milling and extraction operations, and so workers may be exposed to radiation. The regulatory requirements and specific obligations for enterprises at this site are not well developed and require urgent improvement. Production was located in several shops and pilot production shops connected to one another by a technological sludge line. Tailings material resulting from uranium ore processing and primary ore concentration have been stored on site and in the nearest specially prepared tailings dumps on the territory within the industrial zone. This resulted in formation of nine facilities, i.e. 7 tailings ponds with the residues of uranium production and two storage sites for uranium raw materials (with a total area of 2.68 million m2). The distinctive feature of this former uranium legacy site and its uranium tailings is that it is also located inside industrial zone of the densely populated city of Dnieprodzerzhinsk. The residential area is located close to the industrial zone, only 1-2 km from the nearest tailings dump. Therefore, the state of the former uranium production site, its facilities and possible options for remediation are of great concern to the local population.
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It is important to note that the residues are located within the territory of the industrial zone of the city, inhabited by about 400 thousand people, as well as about 14 km to the South-east of the city. The tailings are also located close to the bank of the Dnieper river.
Figure 3.1 -The PChP Uranium production legacy site , where: 1 – Tailing Dneprovskoe (―D‖), 2 – Taling ―Zapadnoe‖, 3 – Tailing ― Yugo-Vostochnoe‖, 4 – Tailing ―Centralny Yar‖, Tailing ―Lazo‖, 6 – Southern sector, where are the most contaminated former, U-extraction facilities, and remained high contaminated territory are located and 7 - is Northern sector of PhCP, where most of existing enterprises are still in operation at this legacy site area, 8 is location of Former U-extraction facility Building №103 and adjacent area. Residues resulting from uranium production are located in engineering structures (tailings) of different types that were established without special anti-filtration screens in difficult landscape and geological conditions (often on relatively steep slopes with potential landslides). Some tailings are covered by perennial deciduous trees, destroying the cover, while other tailings don't have any cover at all. Among these tailings there are (i) dry and impounded tailings, (ii) those requiring removal of upper contaminated layer of an old cover with impurities of ore materials and pulp materials, (iii) those where cover requires only repair; and (iv) tailings, where the hydraulic conditions and condition of the dikes are in need of urgent repair. In addition to the tailings, a number of highly contaminated buildings containing elements of the former U-milling and U-concentrates production remain in this territory. During recent decades since U-production was stopped, many buildings and operational facilities, which were not contaminated, were transformed into other industrial productions, not linked with 3-3
the former U-production. Some of the U-production legacy buildings were decontaminated and are in use for different types of production by different operators. At present some of these facilities are operated by private companies. These include many buildings and other facilities that were used in past as a one integrated complex of the multi-purpose industrial production. These were rented to the private companies by the local state administration for long-term use, including land-use for their industrial production. However, in a number of cases, the buildings and other facilities are located in the areas, which have been significantly contaminated by the past U-production activities. These circumstances make the future development of the site uncertain, because a significant part of the territory has relatively high external gamma dose rates (Figure 3.2). Up to 40% of the southern sector area of the former PChP legacy site has gamma dose rate higher 0,4 µSv/h.
Figure 3.2 - Gamma-dose rate distribution over the site at the Northern (upper) and Southern sectors (down) of the former PChP Uranium production Legacy Site. The different colour indicated buildings at the site and belonging to the different operators According to the available monitoring data, exposure from gamma radiation in some parts of the site exceeds 10 µSv h-1. The background radiation varies in the range 0.05-0.15 µSv h-1. The most contaminated part of the U-production legacy is the southern sector of the site. Areas with gamma radiation dose rates in the range between 0.5 µSv/h and 10 µSv/h cover about 40% of this territory. Hot spots with the highest gamma dose rates are observed in some former uranium extraction facilities, where the gamma dose rate is 1000 µSv/h and higher. The nature of the relatively high level of contamination in this territory results from the dispersion of contaminated material, spillage of uranium residues and remaining elements of the former Uranium production. Some areas (for instance around sedimentation pond 220 and 230) were contaminated as result of the wind re-suspension of the accumulated dried sludge residues from these ponds to the surrounding area.
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Several buildings at the site have high radioactive contamination, showing high gamma dose rates and also high concentrations of dust containing alpha emitting radionuclides dispersed throughout the premises. Significant amounts of equipment used for U-extraction process in past still remain in the former U-extraction facilities. These premises and remaining facilities and equipment are highly contaminated with radionuclides of U-Th series. Until present time, the contaminated equipment and former shops workshop premises in some buildings have not been decommissioned according to the requirements, norms and standards existing in Ukraine. Only limited remedial actions have been performed on the site of the Pridneprovskiy Chemical Plant. The total number of enterprises operating in this site is about 30 with up to about three thousand employees (according to official data). The activities of these enterprises are not related to the utilization of nuclear technologies, but their employees are subject to risk of ionizing radiation from contaminated materials and facilities from former uranium ore processing at the site. According to assessments performed by the local administration, only between six hundred to about 1 thousand workers work in this territory at the present time, because of the poor economic situation. Most of the work places are related to the 3-5 main industrial producers at this territory. The largest enterprises on the site are the ―Dniprovsky mineral fertilizer plant‖, which produces phosphate fertilizers and reprocesses phosphor gypsum. The ―Zirconium‖ enterprise (produces metal zirconium and hafnium). Their working places are mainly located at the relatively clean northern sector of the legacy site. In addition, the enterprise ―Smoly‖ (Resins), which produces ion-exchange resins is located in the highly contaminated area of the southern sector of the site and its main production facility is located very close to the one of the most contaminated building of the former U-extraction, Building 103. Other enterprises are relatively small, but their influence and role in the future development of this industrial site might be substantial. Only some of these enterprises are operated by the town administration and Site Operator, i.e. ―Zirconium‖, ―Smoly‖, Barrier and specialized enterprises for physical protection "VITCh-38‖ (state owned). Other enterprises are privately owned. The special State enterprise ―Barrier‖ SE was created at this site under the financial support and control of the Ministry of Fuel and Energy of Ukraine in 2000 in order to:
safely manage this site provide radioactive residues hazards assessment, to establish monitoring and bring this site into an environmentally safe state
This would serve as a basis for further development and implementation of the State Programme to transform this legacy territory affected by the former Uranium production as now established in the State Ecological Programme on Transformation of Uranium Facilities of Pridneprovskiy Chemical Plant into Safe State for period from 2010 till 2014 (Decree No. 1029 of 30.09.2009).
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3.2.2 Former Uranium extraction facilities Hydrometallurgical processing of the Uranium ores included the following stages:
Delivering and storing of the raw materials ( Buildings 46, 1,2) Pre-concentration of the ores and preliminary crushing (Buildings 3,) Milling (6 wet mills, Buildings 4-5) Leaching, mainly by acids ( Buildings, 6, 2b) Precipitation and filtration of the Uranium ores slugs (Buildings 6, 104, 120, Separation of Ra-Th and rare metals fractions from radiochemical solutions (Building 104) Extraction of the Uranium oxidized forms from the solution ( Buildings 2b, 103) Concentration (centrifugation) and crystallization of the solution in to the ―yellow cake‖ forms (Building 103).
As result of milling, the raw ore materials were finally converted into the sandy and silt fractions of milled materials with particle sizes from 5 to 0,074 mm, and were transported to the leaching and extraction facilities. Sulphuric and sometimes nitric acids were used in the leaching procedures. Precipitation and extraction procedures were carried out in the presently closed buildings (6 and 2b). The final production of Uranium concentrates was carried out in the facilities (Buildings 103 and 104). Usually the yellow cake is produced in chemical ammonium forms (NH4)2 U2O7) or other nitrate uranium chemical forms. The final stage of Uranium extraction from the ore raw materials used at the PChP was completed in Building 103 by purification of the yellow cake into oxide-dioxide forms such as U3O8. The extraction process at the PChP was based on ion-exchange adsorption of the slag waters and acid extraction procedures at the hydrometallurgical plants. The residue slags after the extraction procedure were neutralized by chalk and alkali and finally delivered by pipeline transport as tailings to the closest depressions available in the natural landscape of the industrial territory or in its vicinity. The procedures for Uranium ore extraction at the Pridneprovsky chemical plant were typical of technologies used in the former USSR. At this site, all types of processing of the uranium containing ores were applied in one complex (Figure 3.3), including milling and hydrometallurgical extraction, adsorption, and also radiochemical separation and purification of U-concentrate from radium and thorium. The products of the radiochemical processing of the blast-furnace slag, such as nitrate solution, were transferred to the neighboring plant ―AZOT‖ (Nitrogen) for further production of ammonium nitrate and nitrate-containing residues.
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Figure 3.3 – Layout of Building facilities at the southern part of the PChP legacy site with the indicated properties to the different enterprises. After leaching uranium from phosphate ores, the complex radiochemical solutions were purified from the iron, thorium and rare earth elements and the phosphate fertilizers were produced. Phosphogypsum residues produces at the end of the production line chain, were usually pumped to the wet tailings ―Sukhachevskoe‖ or were put on to the surface of tailings containing high concentrations of Ra-266, thorium and other radioactive substances (Dneprovskoe tailing). Pictures of remaining buildings and equipment from the former U-production facilities can be seen in Figure 3.4.
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Building 2b. U-extraction facility is property of the Mineral Fertilizer Plant (DZMU) Gamma-dose rate – from 0.3-0.7 µSv h to 2.3-2.8 µSv h (radiochemical extraction levels).
Building 104 Th separation and adsorption facility, pumps for radiochemical solutions, presses. Gamma-dose rates varies in range from 0.5-8.0 µSv h-1 to 25-75 µSv –h.
Building 6. Former Hydrometallurgical Plant is property of FERREXPO LLC. Gamma-dose rate 1.5-3.0 µSv h-1 Max. surface beta contamination 2300 Bq cm-2
Building 2в. Uranium solution re-mobilization facility Gamma-dose rate -- from 0.5-8.0 µSv h-1 to 14-18 µSv h-1
2e. Uranium precipitation tanks 0.3-0.8 µSv h-1
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Building-1. U-ore storage ( γ-dose rate 15-25 µSv h-1)
Building 4-5. Former U-ore Mill and separators near shop-6 (Gamma dose rate 2.0-3.0 µSv h-1) after decontamination
Figure 3.4 – The remaining equipment legacies used for U-production at PCP The residues after purification of the radiochemical solution were transferred in furnaces (Building 112) and then transported again to Building 6 for secondary acid leaching. The remaining solid residues also were moved to the tailings. Such residues contain high activities of Rn and Th as well as their daughter radionuclides and toxic metals as well. Among the large number of buildings, milling, radiochemical facilities, and tailings that were directly involved in the uranium production cycle complex (for instance, ore material transportation, temporary storage, wet and dry milling, pipe line for hydraulic transportation of the milled materials between different shops, radiochemical extractions/adsorption, separation of Ra and Th from the complex radiochemical solutions, yellow cake and U-dioxide production and residue and waste management at the site), five facilities have been identified as having the highest priority for remediation. Among these five, the most heavily contaminated are Buildings 103, 104, and 112. The five most contaminated buildings and brief characterization of their present state according to the ongoing inventory process assessment are given in the Table 3.1
Table 3.1 – Characterization of the most contaminated U-ore processing facilities, which are presently are in operation by PChP legacy site operator - SE “Barrier” №
№ Use in the U-Production in past
γ-dose rates, µSv h-1 and β, cpm
1
103
Extraction and calcinations of the uranium dioxide. Equipment for U-product calcinations has been demolished. Extraction facilities are in place and still highly contaminated. The building has many highly contaminated areas.
0,75– 300 50-27000
Remediation Concept Preliminary design has been made for demolition of this building, the metal equipment will be partly cleaned up and contaminated residues will
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be removed to the tailings.
2
3
104
112
Sorption and de-sorption of the U from the complex radiochemical solutions. Purification of the solutions from thorium, radium and other residues. The equipment is partly dismantled.
0,50 – 75 30 - 1500
Kiln, burning of the residues after purification of the radiochemical solutions. Kiln super structure (most contaminated) was dismantled
5,3 - 7,3 10 - 200
The fate of the facility is still not defined
The basement will be demolished.
4
120
Pumping station for sludge pulp, with sedimentation ponds. Precipitation of the rareearth elements and thorium from the complex radiochemical solutions. Storage of the rareearth elements and thorium concentrate.
0,3 - 1,2
The storage facility is currently in use for storage of the contaminated pipe line.
5
46
Former garage for thawing of the carriages with U-ore materials in the winter time, decontamination of carriages after reloading
0,3 - 5,2
To be decontaminated or demolished
3.3
Demography of the Area and Human Behavioral Characteristics
The Pridneprovsky Chemical Plant Site is situated in the outskirts of Dneprodzerginsk City, which is located in Dnepropetrovsk Region of the Ukraine (at 35 km distance from the Dnepropetrovsk City). The population of Dneprodzerginsk is about 250 thousand people; the area of the city is about 138 km². The distance from Building 103 to the nearest location of the inhabited area of Dneprodzerginsk City is about 1.5 km. A number of current industrial enterprises are situated immediately at the Pridneprovsky Chemical Plant Site. The most important ones are SE ―Zirconium‖, SE ―Smoly‖, SE Pridneprovsky Hydro-Metallurgical Plant, SE ―Barrier‖, etc. The total number of staff working at different enterprises at the PChP territory varies from 650 to several thousand people. Some of the enterprises are located in the immediate vicinity of Building 103 and in particular State enterprise ―Smoly‖ (less than 50 m).
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1
1
2
Figure 3.5 – General view on Building 103 [1], which was a main facility for U-extraction at PChP legacy site and neighbouring Building 102 [2], that is still in operation (State enterprise “Smoly”.) Other enterprises are located at the territory. The closest distance to the administrative buildings of several other enterprises and in particular to the fences of former sanitary zone, that was established in past for this enterprise are located just several hundred meters from this former Uextraction facility. Therefore the possible impacts of the radioactive materials located in Building 103 on the personnel working in the adjacent building must be investigated and remediation of Building 103 and the surrounding environment should be a high priority in the remediation of the PChP industrial site. Potential effects to the citizens of Dniperodzerzhinsk are expected to be very low even during the demolition of this building, even in accidental situations. 3.4
Environmental Baseline Climatic Conditions
The PChP site is situated in the Dnepropetrovsk Region, in the steppe zone of Ukraine. The climate is temperate-continental, and is characterized by hot (sometimes draughty) summer, and relatively cold winter (GEC, 2009). Mean air temperatures during a year are presented in Table 3.2. Table3.2. Mean monthly temperatures for the Dneprodzerginsk City Month
І
II
III
Temperature, °С
-5.4 -4.8 0.4
IV
V
VI
9.0
16.4 19.8
VII
VIII
IX
X
XI
XII
22.3
21.3
15.7
8.8
2,0
-3.1
Data on wind characteristics (such as direction and velocity) are presented in Tables 3.3-3.4. Frequency for the wind speed of 9 m/s and higher is 5%.
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Table 3.3 - Mean yearly frequencies of wind directions in Dneprodzeginsk City Wind direction
West
North East
East
South East
South
South West
West
North West
Calm
24
9
15
4
20
9
9
10
20
Frequency, %
Table 3.4 - Wind speed characteristics during a year (mean monthly values) Month
1
Wind speed, m/s
4.7
2
3
4
5
6
7
4.7 4.8 4.2 4.1 3.4 3.3
8
9
10
11
12
Year
3.1
3.0
3.7
4.2
4.4
4.0
Detailed information about wind distribution at the site are described in special studies carried out for justification of meteorological mast, which was installed in 2013 and will serve for verification of wind forecast and simulation of the dust dispersion over the site in case of building demolition scenarios. With regard to amount of atmospheric precipitation, the study site is situated in the zone of insufficient humidification (the yearly evaporation is higher than precipitation). The mean yearly precipitation is 477 mm. Data on mean monthly amount of precipitation are presented in Table 3.5. The maximum daily amount of precipitation is 82 mm. Table 3.5 - Monthly and mean yearly precipitation in Dneprodzerginsk City Month Precipitation
І
II
III
ІV
V
VI
VII
VIIІ
IX
X
XI
XII
Year
35
29
31
35
46
65
53
40
40
37
37
39
477
(mm) Additional climatic and geological data available and can be submitted upon request 3.5 3.5.1
Radiological Characterization of the areas surrounding Building 103 Radioactive contamination of the adjacent area
During implementation of the State Remediation Programs 2009-2013, there were many activities with focus on the radioactive and ecotoxicology site characterization, establishing site specific monitoring programs, preliminary safety assessment, first priority cleaning of the most contaminated facilities, reparation of the tailing covers. A number of Environment Impact Assessment (EIA) and Feasibility Studies have been completed at the site. The results of initial assessments included initial exposure pathways analyses and dose assessments for workers presently located at this territory. These results are described below and have to be considered as the preliminary assessment for prioritization of the further remediation 3-12
strategy, which is currently under development by Ukrainian experts and international experts. Much of this in taking place under the ENSURE-2 project, financially supported by SSM, Sweden. The brief description of these data, in particular those describing the adjacent areas around Building 103 is given in this section. The gamma dose survey data presented at the territory of former PChP were showed on Figure 3.2. As was shown on Figure 3.2, the most contaminated sector of the site is southern sector. The gamma dose rates were exceeded 0.5 µSv/hr on at least on 40% of the PChP territory of southern sector. In some areas around tailing compounds direct gamma dose are exceeding 10 µSv/hr. The areas around Building 103 are also significantly highly contaminated as the result of radionuclide dispersion of dust materials from the uranium ore milling and extraction procedures in the building in past. Transportation of radiochemical solutions and crashed ore materials also occurred mainly at the boundaries of the southern sector of the PChP Uranium legacy site. Local ―hot spots‖ are also caused by secondary contamination due to re-distribution of the contaminated substances to the adjacent areas from open sources, such as contaminated dusts in the contaminated buildings. Figure 3.6 demonstrates spatial gamma-dose rates around Building 103 due to deposition of U-production residues. These results can be used for future assessment of radiological risks to workers, implementing the clean-up and demolition of the Building 103 and the surrounding areas. Figure 3.6 shows that the highest gamma dose rates directly adjacent area to this building greatly exceed a proposed remediation objective of 0.5 µSv h-1. All areas around this building as well as the building itself and external tanks require remediation. The presence of these high gamma dose rates can pose significant risks for workers of the conventional enterprise in the neighboring Building №102, an operation producing ion-exchange resin materials.
Figure 3.6 – Gamma dose rate spatial distribution around former U-extraction facility (Building 103) and its nearby adjacent infrastructure facilities 3-13
Figure 3.7 - View on the most contaminated elements of the former U-extraction facility located at the northwestern outside corner of the Building 103
These high gamma dose rates, the highest of which were observed near the tanks, located at the northwest corner of the Building 103 were observed to be between 20 and 40 µSv h-1 measured 1 m above the ground and up to 250 µSv h-1 observed on the roof of the tanks (Figure 3.7) constitute the greatest risk outside Building 103. Thus, the safe dismantling and removal of these tanks is a high priority in the remediation of this facility. When considering possible negative impacts in the remediation of Building 103 to human health and the environment, special attention must be paid to asbestos materials. They are present in many forms, including in rigid building cladding panels, roof deck panels, insulation cover of processing tanks (see Figure 3.7 left) and as spray-on insulation in the roof and walls of buildings. The friable asbestos insulation is separating from the cladding, and in many instances has fallen to the walkways and floors below and ground adjacent to the buildings and in some locations the asbestos pipe insulation has also deteriorated and fallen. Asbestos materials must be removed prior to demolition activities in accordance with best international practice. The elevated gamma-dose rates and dispersed materials, which containing U-extraction residues with relatively high activity concentration of Ra-226, create also area with elevated emanation of Rn-222. The averaged activity concentrations of Rn-222 at the adjacent areas around Building103 are varying in the range from 90 to 200 Bq/m3. The present state of Building 103 creates risks to the surrounding areas. The windows at the former U-extraction facility are broken for the most part, creating conditions for dispersing of the spilled materials in Building 103 to surrounding areas by wind entrainment. This can result in negative impacts from inhalation exposures to workers in the adjacent buildings and surrounding areas. 3-14
The surrounding contaminated facilities and adjacent areas, for instance, boxes in a building 1 (see Figure 3.7 and photo on Figure 3.8) after decontamination can be used as a temporary storage facilities for containers and metallic constructions generated during decontamination and demolition of the Building 103.
Figure 3.8 - View on remained facilities 1a in its current status (Photo dated by March 2013)
The monitoring studies also show that the areas adjacent to Building 103 contain additional radiological risks due to large amount of dispersed materials originating from contaminants inside of the building. The results of aerosol contamination surveys at the different locations around Building 103 are presented in Table 3.6. The results in the Table 3.6 show the range of the of U-Th series radionuclide activities concentration in the aerosols collected at 1 m above the ground. The concentrations vary from one side to another and with the season and the meteorological conditions. Therefore, these results can be considered as very preliminary. However, even these limited set of available data demonstrate that ambient activity concentrations of U-238 and Ra-226 in the area adjacent to Building 103 are 3-20 times higher than the ambient activity of the aerosols collected at the meteorological station in Dneprodzerzhinsk, that is 12 km far of the former PChP U-Mill site.
3-15
Table 3.7 - Specific radionuclide activity concentrations in aerosols collected nearby former U-extraction Building 103 during recent years Location
Air Pump volume
Ambient aerosol activity concentration aerosol, [Bq·м-3]·10-3 U-238
Ra-226
+/-
(м3)
Pb-210
+/-
Pо-210*
Th-228
+/-
+/-
North west area 23.06.09
302
0,44
0,05
0,11
0,01
1,16
0,11
0,25
0,06
0,09
0,01
North west area 04.08.09
210
0,11
0,06
0,24
0,05
1,21
0,13
0,56
0,14
0,05
0,01
North east area 12.11.09
342
0,04
0,02
0,02 0,006
1,02
0,13
0,25
0,06
0,01
0,004
22-23.02.11
372
0,10
0,03
0,07
0,02
2,11
0,63
0,63
0,19
0,01
0,003
Southern area 05.10.2011
328
0,05
0,02
0,03 0,007
2,06
0,09
0,62
0,19
0,01
0,006
North west, 05.11.2012
192
0,08
0,03
0,04
0,01
0,94
0,28
0,28
0,08
0,004
0,001
0,02 0,003
0,50
0,03
0,15
0,04
0,02
0,005
Southern area
Local background
0,020 0,01
It is also thought that the relatively high ambient activity concentration of Pb-210 and Po-210 in Dnieprodzerzhinsk are affected by the emissions of the aerosol pollution from a number of metallurgical ore melting plants located in the town. 3.5.2
Radiological conditions at the territory of the industrial site
The monitoring studies carried out at the territory of PChP and surrounding areas, indicated that the main exposure pathways were: gamma radiation from significant contamination of the legacy site with the residues of Uranium production; relatively high Rn-222 ambient concentrations in the vicinities of tailings dumps and former uranium production facilities and also from dust inhalation. All these factors were taken into account to estimate the current radiological conditions at the site. The preliminary results from recent site characterization studies (Voitsekhovych et al. 2012) are discussed below. Radiation exposure estimates showed that external gamma dose exposure and 222Rn inhalation pathways provide the highest contribution to the total doses. The maximum annual dose rates may
3-16
exceed dose limits 20 mSv a year for personnel, the limit at which intervention is justifiable. Preliminary assessments are given on Figure 3.9 and Table 3.8. Table 3.8 - Ranges in preliminary dose estimates for workers at the legacy site Type of scenario and representative group of potentially impacted individuals
Dose estimated , Main factors of exposure
mSv a year min
max
Gamma irradiation Rn-222,
2.3 – 4.0
25-35
Workers of the enterprises, which occupy the former uranium production facilities, and which workplaces are situated at the basements and still contaminated premises
Rn-222 and aerosol inhalation
6.0-8.0
10.013.0
Workers, which may regularly work at the territory and near the tailing dumps having access to the tailing (for instance having access to the forest at the ―Centralny Yar‖ tailing)
Gamma irradiation
1.0-1.6
2.5-3.5
Workers, who are implementing remediation tasks, maintaining tailings cover, and other services that require staying at the contaminated sites at least 40% of the working time.
Gamma irradiation
1.4-2.0
8.0-12.0
Rn-222, aerosol inhalation
Workers of the enterprises, which are situated at the legacy sites, working in the relatively clean premises inside of the buildings
Gammairradiation and aerosol inhalation
0.6-1.0
1.4-1.8
Workers who are working in the administrative premises and not within the contaminated areas of the legacy site
Inhalation of aerosols
0.1-0.3
0.3-0.5
Workers, who regularly visited the most contaminated buildings at the legacy site
Rn-222
Most of the present site workers have their main workplaces in the administrative buildings and in the non-contaminated industrial premises and cannot be significantly exposed. Their total annual doses are estimated to be in the range of 0.1-0.3 mSv a year and less. The maximum annual doses for workers, who may potentially work at the tailings dumps, and also for those who may work in or near the high contaminated buildings (former uranium production workshops) vary from 1 to 12 mSv a year, depending on their specific duties and time spent at the contaminated areas. The 3-17
highest dose rates 0.03-0.04 mSv per hour and higher would be obtained by those workers, who will be involved for clean-up and demolition of the most contaminated buildings and former Uextraction facilities. Strict radiation safety control and dose constraint measures will be required for these categories of workers.
Figure 3.9 - Equivalent potential dose for personnel assumed to be working for 1 hour at the different locations (territory, buildings and tailing dumps) of the Pridneprovsky Chemical Plant Uranium production Legacy site Assessment studies show that Building 103 is the most contaminated former U-extraction facility and it potentially results in the highest doses to personnel working at the territory of PChP Uranium legacy site. Therefore its decontamination and dismantling of the most contaminated equipment in the building are of highest priority in the remediation activities on the site. Therefore, within the frame of the ENSURE technical assistance project, Ukrainian experts have prepared a detailed description of the recently obtained monitoring and U-extraction facility hazards characterization data as an input for a preliminary safety assessment of the current situation. Another objective is to develop the assessment methodology necessary during the decontamination, demolition and clean-up of the area, which is currently occupied by these facilities. Detailed assessment of the radioactive contamination and radiological situation in-site of contaminated premises of the former U-extraction facility Building 103 are considered in the following chapter.
3-18
Radiological Characterization of the former U-extraction facility “Building 103”
3.6 3.6.1
General
During recent years, special studies were carried out by UA project partners, investigating the radioactive contamination in the former uranium extraction facility, Building 103. The following elements of the characterization are summarized:
Assessment of the status of the building structures Assessment and inventory of the remaining equipment in the former U-extraction facilities Assessment of the conditions in the various rooms (administrative, control technology rooms, industrial workshops, supplementary facilities and storage rooms for Uproductions) in Building 103 Gamma dose rate survey throughout the building Characterization of spillage and dispersed materials, radiological properties through gamma-spectrometric analytical studies, Collection of aerosol samples and their gamma and alpha spectroscopic analyses
These results are discussed in this section and the data will serve as a basis for further preliminary safety assessment of the current situation and also for verification of the model simulation results. In total, about 40 samples of deposited dust, about 30 samples of the dispersed and spilled materials, and several tens of aerosol samples outside and inside of the building were taken in the Building 103 during recent years. All these samples were studied in the analytical laboratory of Ukrainian Hydrometeorological Institute, where specific radionuclide activities of the collected samples were determined. The activity concentrations of these samples are given in the Tables below as typical mean values as well as maximal and minimal values. Detailed specific data, at identified specific locations are given also in the attachment. It was found that the activities of U-238 and Ra-226 in the dust covering the window sills ranged from 7 to 10 kBq/m2 and from 0.3 to 1.5 kBq/m2 on the tiled and painted walls of the shop. The surface dust contamination in the adjacent service rooms of the former U-extraction shops were found in range from 2 to 5 kBq/m2. In many places spilled materials of the white, red and bright yellow colors were identified. The highest activity concentrations of Uranium, in range from 350-600 kBq/m2, were found at several places on the floor of the Level-1. The maximum detected U activity was 1965 kBq/m2. Bright yellow spilled residues covered a limited area (0.3-0.6 m2). The surface density of the materials, in weight units of yellow cake product, was estimated to be 30-150 g/m2. Detailed information on the spatial distribution of the highly contaminated U-production residues is given in the attached tables (Annex-2). The characterization data obtained for the former U-extraction facility Building 103 are shown and discussed in the following section.
3-19
3.6.2
General layout of building and structural details
The general layout of Building 103 is presented on Figure 3.10. This shows the 7 functional areas. The most contaminated area of the former Uranium extraction Facility Building 103 is section 1 (Figure 3.10). General dimension of the various areas are shown in Table 3.9.
Figure 3.10 – General layout of the “Building 103” structural elements (see Table 3.9.)
Table 3.9 -Dimensions of the Main Functional Areas of “Building 103” Area, Functional Areas
Length
Width
Height
m2
Volume, m3
1
Extraction facility, section 1
42,0
12,0
17,200
504
8669
2
Administrative premises
14,0
15,0
14,700
210
3087
3
New melting facility
19,7
23,5
20,000
463
9259
4
Old melting facility
24,0
23,5
12,800
564
7614
5
Storage room for containers with extraction products
66,3
12,0
796
5459
6
Electricity transformation
Complex form
73
365
5,000
3-20
7
3.6.3
Storage of flammable materials
Complex form
3,600
50
180
Radiological Characterization the former U-extraction areas in the Building 103
A site-specific investigation of Building 103 included systematic sampling of spilled material, dusts, aerosols during 2011-2013 and also analytical determination carried out in the radiometric laboratory of the Ukrainian Hydrometeorological institute (Kiev). The majority of the analytical measurements have been done with the financial support of the ENSURE-2 project. Presently the premises are locked and the access to Building 103 is restricted. Several views of the different elements of the facility and conditions inside of the Building 103 are shown in Figures 3.11 and 3.12 and also in photos presented in Annex 2, reflecting the existing situation.
The spilled yellow cake materials on the floor of the 1 level of the facility
General view on the walls, the remains fragment of painting materials as a potential source for dispersion of the contaminated materials
3-21
U-extraction tanks, with remaining fractions of the complex radiochemical solutions and residues of the milled U-ore materials Figure 3.11 - Inside view of the U-extraction facility Building 103 (2013) The results of the investigation indicated ore materials are dispersed ubiquitously over the floor in the extraction areas of Building 103. The materials are highly contaminated with radionuclides of the U-Th series and a potential risk of inhalation exposure exists. The spilled materials are also dispersed in some of the neighboring rooms inside of the building. The high Radium-226 content in the spilled materials acts as a source of increased Rn-222 exhalation in the premises and consequently there is a relatively high Rn-222 ambient concentration inside the building (up to 500-1000 Bq/m3). However, the highest potential risks for workers, who will be involved in the remediation are expected to be from the external gamma radiation. Detailed information about gamma-dose rates inside of the three levels of Building 103 where the former U-extraction and processing took place are shown in Figures 3.13 to 3.15 below.
General view on the floor of the Level 1 at the section 1 ( U-extraction facility)
3-22
General view on the floor of the Level 2 and Level 3 at the section 1 ( U-extraction facility) Figure 3.12 - Spilled materials in the extraction facilities and piping of Building 103 with relatively high radioactive and chemical contamination Radiological Characterization of Level 1 of Building 103 Radiation fields in Figure 3.13 show that highest level of external gamma radiation corresponds with an area spilled ore extraction materials between Tanks № 94, № 59 and № 1, where gamma dose rates between 20 and 300 µSv h-1 were observed. A maximum dose rate in the range of 300-500 µSv h-1 was measured near Tank № 59 (Figure 3.13).
Figure 3.13 - Spatial Distribution of Gamma Dose Rates in the Extraction Area on Level 1 of the Building 103 The spilled U-production materials are the source of the dust deposition throughout the building. The radionuclide activities of surface dust, taken on the surface of window sills, walls and equipment in the extraction area on the first Level of Building 103 are shown in the Table 3.10. Specific radionuclide activities in the spilled materials in this area are given in Table 3.11. The individual data used in obtaining the averages are given in the Attachment. The mean values were from the samples taken from the whole of the extraction area. The maximum values correspond to the area adjacent to Tank № 59 and the minimum values correspond to the area around Tanks 28 (see Figure 3.13). Table 3.10 - Radionuclide activities in surface dust on Level 1, kBq·m-2 Specific radionuclides in Dust, kBq·m-2
Level 1 U-238
Ra-226
Pb-210
Po-210
Th-230
Th-228
3-23
mean
2,6-6,8
0,4-0,8
0,2-0,6
0,2-0,6
1,3-5,2
0,02-0,04
min
0,01
0,1
0,02
0,02
0,1
0,002
max
13,7
1,0-1,2
1,2
1,2
13,2-16,7
0,06
Table 3.11 - Radionuclide activities in residues spilled over the floor of the extraction area on Level 1, Bq·g-1 (dry wt) Level 1
Radionuclide activities in the spilled materials, Bq·g-1 (dry wt) U-238
Ra-226
Pb-210
Po-210
Th-230
Th-228
mean
40-320
6,5-45
3,0-8,5
3,0-8,5
25-60
0,3-0,8
min
10
0,5-1,5
0,5-1,5
0,5-1.5
2.0
0,02
max
3116
168
44
44
236
2-3
The results of the analysis of radionuclides activities in aerosols collected on filters at the different shop locations on Level 1 are given in the Table 3.12. Table 3.12 - Radionuclide activities in aerosols in the Uranium extraction area of Level 1 Location
Entrance to workshop
Air Pump volume (м3)
Ambient radionuclide activities in aerosols, (Bq·м-3)·10-3 U-238 +/-
Ra-226 +/-
Pb-210 +/-
Pо-210*
Th-228 +/-
224
0,83
0,25
0,07
0,02
0,84
0,25
0,25
0,08
0,04
0,01
Integrated from 3 locations, left 29.08.2012
130
73,4
22,0
6,97
2,09
0,61
0,18
0,18
0,06
0,01 0,004
Integrated from 3 locations, right 29.08.2012
167
30,7
9,2
1,81
0,54
1,81
0,54
0,54
0,16
0,01 0,003
Central part 04.12.2012
270
3,06
0,92
0,16
0,05
0,43
0,13
0,13
0,04
0,03 0,008
0,02
0,01
0,02 0,003
0,50
0,03
0,15
0,05
0,02 0,010
05.11.2011
Local background 2009-2010
3-24
PC В inhal*
3,0
0,7
5,0
3,0
0,4
PC А inhal*
200
60
400
300
30
*PC Binhal and PC Ainhal are Permissible reference activity concentrations established for Personnel in cohort B and A accordingly to Radiation Norm of Ukraine NRB-97 Local background level of aerosol contamination in inhabited areas and also reference levels for Group B workers (people who are working at the PChP area, but not related to the U-legacies) and also reference levels established for radiation workers ( Group A) are given in a table as well. The results given in Tables 3.10-3.12 and Figure 3.13 can be used in a preliminary dose assessment from direct gamma and inhalation exposures. The results presented in the tables above show, that at Level 1 of the former U-production facility, the Uranium is the dominant radionuclide in the spilled materials. The bright yellow spills were identified as residues from the yellow cake production. U activities in the dispersed yellow cake materials on the floors were measured to range up to 3000 Bq per gram, having a surface density up to 100 g/m2. These data suggest that the U-residues still remaining in the tanks, tubes, and apparatus, could be segregated and reutilized for secondary reprocessing. Radiological Characterization of Level 2 of Building 103 Gamma dose rates in the U-extraction area on Level 2 are presented in Figure 3.14.
Figure 3.14- Spatial Distribution of gamma dose rates in the extraction area on Level 2 of the Building 103 The data given in Figure 3.14, clearly indicates two high radiation field regions on Level 2, with gamma dose rates between 20 to 500 µSv h-1 around Tank 66A and gamma dose rates between 500-700 µSv h-1 beside Tanks №11/1 and №11/2.
3-25
Fewer spilled materials were found in the U-extraction shop on Level 2. Radionuclide activities in the dust on the open surfaces of the equipment, shelves, windows on Level 2 are given in Table 3.13.
Table 3.13 - Radionuclide activities in surface dust on Level 2, kBq·m-2 Specific radionuclides in dust, kBq·m-2
Level 2 U-238
Ra-226
Pb-210
Po-210
Th-230
Th-228
mean
0,4-0,8
0,1-0,8
0,06-0,08
0,06-0,08
0,33-0,42
0,01-0,03
min
0,2
0,05
0,03
0,03
0,17
0,005
max
2,7
1,2
0,25
0,25
3,87-26,7
0,05
The results of radionuclide activities in the spilled materials on the floor of the Level 2 are presented in Table 3.14. The results of the investigation show a relatively small quantity of radioactive materials dispersed over the floor area of the Level 2. However in the locations, where the spilled material was U-yellow cake, the U-activity values were comparable with those of the spilled materials collected on Level 1. The maximum U-activity in such materials was estimated to be about 23 kBq/g. The activities for Ra-226 were in the range of 80-120 Bq/g and those for Pb-210 and Th-230 were comparable. Table 3.14 - Radionuclide activities in residues spilled over the floor of the extraction area on Level 2, Bq·g-1 (dry wt) Specific radionuclide in the spilled materials, Bq·g-1 (dry wt)
Level 2
mean min max
U-238
Ra-226
Pb-210
Po-210
Th-230
Th-228
24.1-80.5
2.5-7.4
2.6-7.5
2.6-7.5
2.4-6.9
0.3-0.5
10.1
0.5
0.18
0.18
0.5
0.01
60-86
55-75
24-114
1,4
1000-2330
64-105
3-26
Radionuclide activities in aerosols under normal conditions were determined by using small volume aerosol pumps at different locations of Level 2. The results are given in Table 3.15.
Table 3.15 - Radionuclide activities in aerosols in the Uranium extraction area of Level 2
Location
Right and central parts 15.11.2012 Left side part 15.11.2012 Local background 2009-2010 рр. ДКВinhal
Air Ambient radionuclide activities in aerosols, Bq·м-3·10-3 Pump U-238 Ra-226 Pb-210 Pо-210* Th-228 volume +/+/+/+/(м3) 173
2,41
0,72
0,92
0,28
1,43
0,43
0,43
0,13 0,05
0,02
253
0,95
0,28
0,47
0,14
0,65
0,20
0,20
0,06 0,04
0,01
0,02
0,01
0,02
0,003
0,50
0,03
0,15
0,05 0,02
0,01
3,0
0,4
3,0
0,7
5,0
Radiological Characterization of Level 3 of Building 103 Figure 3.15 indicates that the gamma dose rates are high around all of the major Tanks in the extraction area of the Level 3 in the Building 103. This results because the 7 main Tanks, into which the liquids from the first cycle of radiochemical extraction were routed after the acid leaching procedure was applied to the milled ore materials, are still partly filled with the dried radiochemical solutions. At the time of facility shutdown, there was no emptying and cleaning of tanks and piping according to a proper decommissioning plan and no initial decontamination was implemented. As a result, the dried materials with high Radium-226 activities in equilibrium with other uranium decay chain members have high gamma radiation fields.
3-27
Figure 3.15- Spatial Distribution of Gamma Dose Rates in the Extraction Area on Level 3 of the Building 103 The averaged gamma dose rates at the distance of 3-5 m from the Tanks were measured between 20 and 100 µSv h-1. At a distance of 1 m, the gamma dose rates near Tanks 134/2 and 134/3 were between 300-500 µSv h-1. The highest gamma dose rate was measured near Tanks 130/1 and 130/3, where the gamma dose rates exceeded 1000 µSv h-1 at some locations. Radionuclide activities in dust deposited on surfaces on Level 3 are comparable with those on other levels and are reported in the Table 3.16. Table 3.16 - Radionuclide Activities in Surface Dust on Level 3, kBq·m-2 Specific radionuclides in Dust, kBq·m-2
Level 3 U-238
Ra-226
Pb-210
Po-210
Th-230
Th-228
mean
2.0-4.7
0.3-1.9
0.25-0.64
0.25-0.64
1.0-3.0
0.006-0.01
min
0.18
0.03
0.03
0.03
0.18
0.002
max
13.3
4.2 (35.1)
11.6
11.6
33.3
0.03
The results of radionuclide activities in the spilled materials on Level 3 are given in Table 3.17. The surface dust and the spilled residues are not firmly fixed on the surfaces and can be easily resuspended in case of mechanical disturbances or in case of strong winds causing increased ventilation inside of the building. The results of the radionuclide activities in aerosols determined at two locations with relatively low gamma dose rates (10-20 µSv h-1) are given in the Table 3.18.
3-28
Table 3.17 - Radionuclide Activities in residues spilled over the floor of the extraction area on Level 3, Bq·g-1 (dry wt) Specific radionuclides in the spilled materials, Bq·g-1 (dry wt)
Level 3 U-238
Ra-226
Pb-210
Po-210
Th-230
Th-228
mean
12.1-24.5
13.5-17.4
5.6-7.7
5.6-7.7
9.7-24.0
0.2-0.3
min
5.5
4.3
1.3
1.3
2.3
0.02
max
50-75
60-180
28.5
28.5
98.4
0.05
A comparison of aerosol data from Level 2 with that from Level 1, indicates a 1-2 order magnitude higher concentration for U-238 and Ra-226 on Level 1 than that on Level 2. This can be explained by the fact that much more materials are dispersed on Level 1 than on Level 2. The activity concentrations in the aerosols collected on Level 3 are quite similar to those on Level 2; however, these were collected in relatively low gamma dose rate areas. It can be assumed that near the tanks the aerosol concentrations will be higher. The dominant exposure pathway to workers on Level 2 and Level 3 is expected to be external gamma radiation Table 3.18 - Radionuclide activities in aerosols in the Uranium extraction area of Level 3
Location
Air Ambient radionuclide activities in aerosols, [Bq·м-3]·10-3 Pump U-238 Ra-226 Pb-210 Pо-210* Th-228 volume +/+/+/+/(м3
Right side, Apparatus № 61, 04.12.2012
231
1.31
0.39
0.24
0.07
0.37 0.11 0.11
0.04
0.02 0.004
154
2.25
0.68
0.88
0.26
1.27 0.38 0.38
0.12
0.03
0.01
0.02
0.01
0.02
0.003
0.50 0.03 0.15
0.05
0.02
0.01
Left side, Apparatus № 58 04.12.2012 Local background 2009-2010 рр. ДКВ inhal
3.0
0.7
5.0
3.0
0.4
ДКА inhal
200
60
400
300
30
3-29
Since Building 103 windows are not sealed and in many places are broken or simply absent, it is expected that there is potential for increased radioactivity in aerosols due to the re-suspension of the dust and residues deposited on the surfaces of the building during strong winds and this may increase the importance inhalation pathway both inside and immediately outside the building The results of radon measurements on Levels 1 and 2 of Building 103 are presented in Table 3.19 and 3.20. These are estimated on a basis at least 7 surveys carried out during period 2009-2012 in the building 103. Two methods were used to monitor ambient concentrations of Rn-222 in the air of the contaminated premises (charcoal method with 6 days exposure and track detector methods with 30 days exposure).
Table 3.19 - Ambient radon activity concentration, Bq·m-3 on Level -1 and at different sectors of the premises Ambient activity concentration, Rn-222, Bq·m-3 Level 1
South-Western
North-Western
North-Eastern
South-Eastern
mean
280±55
290±90
265±60
465±65
min
125
165
190
155
max
925
1580
755
1575
Table 3.20 - Ambient activity concentration, Bq·m-3 on Level-2 and at different sectors of the premises Ambient activity concentration, Rn-222, Bq·m-3 Level 2
South-Western
North-Western
North-Eastern
South-Eastern
mean
290±70
275±95
345±50
330±60
min
60
110
50
115
max
975
615
405
665
Only a few measurements were carried out on Level 3 of Building 103 with a short period of averaging (6 days using charcoal methods). The values of Rn-222 ambient activity concentration near group of apparatus 134 were estimated in the range 1100-1570 Bq/m3. 3-30
These preliminary characterization data collected and analyzed in this report will be used as an input for preliminary exposure pathway analyses and preliminary dose assessment from the existing situation and is discussed in the next chapters.
3.6.4
Description of the other areas of Building 103
The areas of Building 103 containing the uranium extraction facilities (tanks with contaminated residues, extraction columns, spills of high contaminated materials, and hundreds of tons of contaminated pipelines containing uranium residues) represented the most contaminated portion of Building 103 on each of the three levels of the building. The areas other than the extraction areas of Building 103 (the areas containing control rooms, administrative and service rooms in the past), are characterized by aerosol activities that are 2-3 times less. Referring to layout of the Building 103 (see Figure 3.10), section 2 of the building was used as administrative building and is relatively clean. In other areas such as section 3 and 5, most of equipment and in particular large volume melting and other equipment have been already dismantled. In the past, section 5 was used as storage for containers with extraction products and is almost empty and the remaining contamination is low. The gamma dose rates in most of the premises are not high and vary in the range between 0,20-0,50 µSv/h. In some places spilled uranium containing materials were identified with U-concentrations from 225 Bq/g to 575 Bq/g. The Ra226 activity concentrations were found as relatively low from 15,5 to 40,0 Bq/g. The Rn-222 ambient concentration in sections 3-5 of Building 103 varied from 350 to 900 Bq/m3 (UHMI, 2013) Other rooms were found as relatively radiological clean and used for the storage of emptied containers and some containers filled with unknown materials which must be studied. The suitability of some of these areas of Building 103 for use as temporary storage of the decontamination wastes and residues, and other wastes generated during dismantling and demolition should be investigated.
3-31
a)
b)
Figure 3.16 Inside view on the section 2 (control room) and storage of containers in section 4
3-32
4
ASSESSMENT METHODOLOGY
4.1
SAFRAN
The SAFRAN model was used in the assessment of the impacts to workers on the PChP site directly or indirectly from the existing conditions in Building 103. The SAFRAN (Safety Assessment Framework) is a user-friendly software application that incorporates the methodologies developed in SADRWMS (Safety Assessment Driven Radioactive Waste Management Solutions) project. The SADRWMS Project was an international programme of work to examine international approaches to safety assessment in aspects of predisposal radioactive waste management, including waste conditioning and storage. In comparing international approaches to safety assessment in those areas, it developed a safety assessment framework and the SAFRAN software tool that implements international best practice in these areas. The SADRWMS project encompassed all types of radioactive waste including disused sources, small volumes, legacy and decommissioning waste, operational waste, and large volume NORM residues. SAFRAN was designed to address all pre-disposal waste management activities and has the following main functions:
to define facilities for storing or processing radioactive waste including their relevant design features. to define waste streams including all relevant radiological and non-radiological properties and their changes through the waste management activities. to define relevant requirements from the regulatory framework (criteria, endpoints, other requirements). to perform safety assessments for all steps of pre-disposal waste management. to perform calculations for quantitative analysis.
Although SAFRAN was not initially designed to include the remediation and demolition of contaminated buildings, several features within SAFRAN are useful in the present assessment: 4.2
to allow the methodical division of the Building 103 into component parts (Levels, rooms, areas), to allow the linkage of specific data to the various areas, to define normal exposure scenarios to identify postulated initiating events (PIE) for accidents, to identify accidental exposure scenarios, and to make dose assessments. Application of SAFRAN to Existing Situation at Building 103
In order to describe Building 103 within SAFRAN, the physical structure of the Building was considered as discussed in Section 3.6.2. Each of the three Levels was represented as a facility in
4-1
SAFRAN. Within each level, the 7 areas were represented as rooms. The different parts of the other sections were represented as areas. This is shown in Figure 4.1.
Figure 4.1 - Visual layout of SAFRAN interface adapted for safety assessment waste management for the Building 103
The various characterization data, mainly external gamma dose rates and aerosol concentrations, were assigned to each area. Once the Building and the relevant data were organized, exposure situations for normal activities can be defined. For accidents, PIEs could be selected from the list of PIEs (in external natural, external manmade, and internal man-made categories). Within the selected PIEs, exposure scenarios could be developed. Receptors impacted by each scenario were chosen as well as exposure pathways, to allow exposure calculations by the model.
4-2
4.3
Routine Exposure Scenarios for Existing Condition of Building 103
4.3.1
Routine Exposure Scenario 1 - Worker in Building 102
Workers in Building 102 are affected by the radioactivity in and around Building 103 via three exposure pathways:
Direct gamma exposure from the high radiation fields in the area between Building 102 and Building 103 (Figure 3.9), Inhalation of aerosols emitted from Building 103, and Inhalation of Radon emitted from Building 103
The following scenario is developed for the most exposed worker in Building 102:
4.3.2
The worker spends 1 hour/day, during breaks in the area between the buildings with an average gamma dose rate of between 2 and 5 µSv/h for a total time of 200 hours in a year. The same worker inhales air at a rate of 1.3 m3/h for 2000 hr/year or 2600 m3/a. The same worker inhales air at a rate of 2600 m3/a with an average radon concentration of 140 Bq/m3.
Routine Exposure Scenario 2 - Worker Conducting Characterization Activities in Building 103
Another normal exposure scenario under the existing conditions in Building 103 is that to a worker conducting measurements within Building 103. This is a probable scenario since more characterization work is planned for Building 103. The worker would be involved in making gamma dose rate measurements (for personal radiation safety), conducting aerosol sampling, conducting radon sampling, sampling residues from five tanks on each level for a total of 15 tanks. The exposure pathways during this scenario would be direct gamma exposure from the ambient radiation fields, inhalation exposure from aerosols and radon gas, and inadvertent ingestion of radionuclides from contact with spilled material and residues in tanks. The following scenario is developed for a worker conducting measurements during an 8 hour day within Building 103:
An average of 5 minutes at 1 m from each of the 15 tanks sampled (the tanks with the highest radiation fields are included among the 15 tanks). The average radiation fields from the 15 tanks are given in Table 4.1. A total of 20 minutes in the section of the extraction area on each Level setting up the aerosol and radon sampling equipment. The radiation fields assumed for each Level are given in Table 4.2.
4-3
A total of 2 hours on each floor in the extraction area with average radiation fields (shown in Table 4.2) conducting miscellaneous measurement activities. The total eight hours would be split up by level (2.67 hrs per level) spent in average aerosol and radon concentrations given in Table 4.3.
Table 4.1 Level
Level 1
Level 2
Level 3
Average radiation field at 1 m from tank (High Field Area) Tank Number 1 59 41 142 94 66A 66 11/2 11/1 14 130/3 130/2 130/1 134/3 134/2
Total Dose at 5 min/tank (µSv) Table 4.2 Level Level 1 Level 2 Level 3
Gamma Dose Rate (µSv/h) 50 500 300 100 50 300 300 500 300 50 1000 500 1000 500 500 496
Gamma Dose Rates in Extraction Room on Three Levels Tank Number Extraction Room - High Extraction Room -Average Extraction Room - High Extraction Room -Average Extraction Room - High Extraction Room -Average
Gamma Dose Rate (µSv/h) 100 10 20 10 300 50
It is assumed that these activities may be repeated an additional time during the same year by the same worker.
4-4
Table 4.3
Average aerosol and Rn concentrations in extraction room on each level Radionuclide Activities (Bq/m3 x 10-3)
Level
U-238
Ra-226
Pb-210
Po-210
Th-230
Th-228
Rn-222
Level 1
50
4.5
1.2
0.25
9.0
0.11
700000
Level 2
1.7
0.65
2.0
0.3
1.3
0.05
400000
Level 3
1.8
0.55
0.8
0.25
1.1
0.02
400000
4.3.3
Workers Preparing for Building D&D Work
Detailed scenarios of exposure and dose assessment for demolition workers were not considered in this particular study and will be developed as a part of the RPP (radiation protection plan) for the building decontamination and demolition taking into account the options and procedures to be developed. Management of the radioactive waste generated during decontamination and demolition of the tanks, pipelines and structures of the Building 103 must also be considered. To help develop the necessary assessment methodology, preliminary hourly dose estimates for the exposure of workers involved in preparing for the decontamination and dismantling of the contaminated tanks at the Building 103 using the present conditions in the building shown in Section 3. 4.4 4.4.1
Abnormal Exposure Scenarios for Existing Conditions in Building 103 Selection of Postulated Initiating Events
From the list of external natural PIEs within SAFRAN, only high winds and heavy precipitation were thought to be feasible to cause impacts to workers on the site. High winds were thought to re-suspend dust and residues on the inside surfaces of the building because of the open windows in the building and the resultant increased ventilation inside the building. This would result in increased aerosol emissions from the building. Heavy precipitation could result in water entering the building through the winds and roof. This was believed feasible since the building was old and had not been maintained for many years and many of the windows were either broken or not present. There were no relevant PIEs from the list of human-induced external events. From the list of human-induced internal events within SAFRAN, the falling were considered plausible:
4-5
4.4.2
an object falling on the material and causing re-suspension. This was believed to feasible since the tanks were the source of the spilled material and their supports are old and in several cases severely corroded or deteriorated. In addition, old piping could also fall on the spilled residues. a spill from a falling tank. Most tanks still contain substantial residues and have openings through which the residues are visible. Since they are old and deteriorating, one could fall and would then spill its contents. a collapse of portion of the building (e.g. wall, roof). This PIE is less likely and would require some external force to initiate it. human intrusion for the purposes of theft. This PIE is highly probable, since this has occurred several times in the past and although there is security restricting entrance into the site, there are no controls on entry into the building. Development of Abnormal Exposure Scenarios
The PIEs listed above were evaluated for probability of occurrence, possible consequences, and similarities of consequences and from this evaluation, the following scenarios were developed: Abnormal Scenario 1: A strong wind, 10 – 20 m/s from a south-easterly direction, leads to increased re-suspension of spilled materials and increased aerosol concentrations within Building 103, and increased releases and inhalation impacts on receptors in Building 102. This scenario bounds the PIE for a dropped object on the residues. The receptor for this scenario would be a worker in Building 102 inhaling the radionuclides emitted from Building 103. Abnormal Scenario 2: A tank containing residues falls down from its mount after a wall collapse and spills its contents causing a puff release from the building. The receptors from this accident are a worker engaged in characterization work inside the room (low probability) and a worker in Building 102. The exposure would be via inhalation for both workers as well as increased gamma exposure for the worker inside Building 103. Abnormal Scenario 3: A PChP worker enters the building unauthorized and goes to Level 3 and spends between one and two hours removing a piece of equipment in the vicinity of Tank 130/3. The main exposure pathway is from gamma radiation. Scenarios involving a partial building collapse were thought to be highly improbable and although the scenario involving leakage of water into the building following a high precipitation event was highly probable, the consequences were thought to be very small since the water would likely remain within the building and there would only be a redistribution of spilled materials.
4-6
5 5.1
PRELIMINARY DOSE ASSESSMENT Safety assessment context
The safety assessment of Building 103 was limited to assessing the impacts of the situation with Building 103 in its existing state and its possible impacts on workers on the site. Within the existing context, there is the likelihood of exposures to workers in Building 102, adjacent to Building 103, through external gamma radiation from the area between Building 103 and Building 102, and through the inhalation pathway. A scenario for the most exposed worker from Building 102 was developed in Section 4.3.1. In addition, doses to workers involved in on-going characterization activities within Building 103 are assessed according to the scenario developed in Section 4.3.2. Scenarios by which workers can be exposed during abnormal events were developed in Section 4.4.2 after considering a number of PIEs to determine possible bounding exposure scenarios. The doses expected from these scenarios were also included in this assessment. 5.2
Estimated Dose to Building 102 Worker - Normal Conditions
In order to estimate the worker exposure from direct gamma radiation, an average gamma dose rate of 3.5 µSv/h was assumed as indicated in Section 4.3.1. During the 200 hours exposure each year the worker will receive a dose of 0.7 mSv/a from direct gamma exposure. The exposure to the worker from aerosols were calculated using the highest measured outside concentrations at the north-west part of Building 103 as shown in Table 3.6. It was assumed these concentrations also existed inside Building 102. The concentrations used are shown in Table 5.1 as are the calculated inhalation doses. The inhalation dose is calculated by multiplying the radionuclide activity in Bq/m3 by the volume of air inhaled, 2500 m3/a by the inhalation Dose Conversion Factor (DCF) (Sv/Bq) of the radionuclide. The total inhalation dose from aerosols was calculated to be 0.02 mSv/a. Table 5.1
Measured aerosol concentrations outside Building 103 and calculated inhalation doses for Building 102 worker Radionuclide Activities (Bq/m3 x 10-3)
Area U-238
Ra-226
Pb-210
Po-210
Th-230
NW corner
0.44
0.11
1.2
0.3
0.1-0.2
Inhalation DCFs (Sv/Bq)
2.9E-06
3.5E-06
1.1E-06
3.3E-06
4.0E-05
Dose (µSv/a)
3.2
1.0
3.3
2.5
10 - 20
Total
20 - 30
5-1
The radon exposure to workers in Building 102 from Building 103 was calculated using the average ambient radon concentration measured outside the north-west of Building 103. Measured radon concentrations ranged from 83 to 205 Bq/m3, with an average of 140 Bq/m3. It was assumed that the same concentration was inside Building 103 and that the equilibrium factor between radon gas and its progeny was 0.4. The following equation was used to estimate the worker exposure: Dose (µSv/a)=222Rn conc.(Bq/m3) * Breathing rate (2500 m3/a) * DCFRn, where, DCFRn = 2.5 E-03 µSv/Bq for radon at an equilibrium factor of 0.4 with its progeny. The resulting exposure from radon and its progeny was calculated to be 875 µSv/a. Thus, the total dose calculated for the maximum exposed worker in Building 102 is 1.6 mSv/a. Of this, about 43% is from direct gamma exposure, 57% from radon exposure and < 1% from aerosol inhalation. This can be compared with the total doses estimated outside of the Building 103 as was shown in Figure 3.9. It was estimated that a total dose of 6-8 µSv/h with contribution of 222Rn exposure pathway as 15-20% and aerosols about 5-7% for people outside Building 103 from the measured radiological conditions. 5.3
Estimated Dose to Characterization Worker
The dose estimated for a characterization worker from direct gamma exposure could be calculated using the gamma dose rates in Tables 4.1 and 4.2 and multiplying by the times (h) spent in these radiation fields. The total external gamma dose was calculated to be 0.5 mSv from residue sampling, 0.14 mSv from sampler set-up, and 0.14 mSv from miscellaneous activities. Thus the total dose from external radiation to a characterization worker is calculated to be about 0.8 mSv per sampling regime. If there are two per year, the total dose to the worker from external radiation would be 1.6 mSv/a. The dose expected from inhalation of aerosols can be calculated in a manner similar to that for the Building 102 worker. Using the radionuclide activities given in Table 4.3 for each Level and the total volume of air inhaled in 2.67 hours at each level (3.8 E-04 m3/s x 3600 s/h x 2.67 h = 3.64 m3) and the DCFs given in Table 5.1, the doses in Table 5.2 were calculated. The total aerosol inhalation dose from one sampling regime was calculated to be 0.66 µSv or 1.3 µSv/a. Table 5.2 Level Level 1 Level 2 Level 3
Aerosol Inhalation Doses for Characterization Worker Calculated Aerosol Inhalation Doses (µSv) U-238 Ra-226 Pb-210 Po-210 Th-230 Total 0.5 0.06 0.005 0.003 0.016 0.58 0.018 0.008 0.008 0.004 0.007 0.045 0.019 0.007 0.003 0.003 0.003 0.036
5-2
In order to estimate the radon exposures to the Characterization worker, the radon concentrations in Table 4.3 were assumed at the various Levels in Building 103 and calculated doses are given in Table 5.3. Table 5.3
Calculated Radon and Radon Progeny Doses
Level
Radon Concentrations (Bq/m3)
Dose (µSv)
Level 1
700
6
Level 2
400
3.5
Level 3
400
3.5
The doses in Table 5.3 were calculated using the equation, Dose (µSv) = Radon conc.(Bq/m3)*Breathing rate (1.3 m3/h)* Hours on Level (2.67) * DCFRn, where, DCFRn = 2.5 E-03 µSv/Bq for radon at an equilibrium factor of 0.4 with its progeny.
Thus the total radon inhalation dose is 13 µSv per sampling regimen with an annual dose of 26 µSv. The dose from the inadvertent ingestion of radioactive residues can be calculated assuming an ingestion rate of 0.1 g/day. If the ingested residue has an average activity as shown in Table 5.4, the doses are calculated by the following equation: Ingestion dose (µSv) = Radionuclide activity in residue (Bq/g) * ingestion rate (1 g) * DFCing, The DCFs and the calculated doses are also shown in Table 6.4.
Table 5.4
Calculated inadvertent ingestion exposures Typical Radionuclide Activities in Residue (Bq/g)
Area U-238
Ra-226
Pb-210
Po-210
Th-230
All Levels
100
30
5
5
60
Ingestion DCFs (Sv/Bq)
4.5E-08
2.8E-07
6.9E-07
1.2E-06
2.1E-07
Dose (µSv)
4.5
8.4
3.5
6
12.6
Total
35
5-3
The dose from inadvertent ingestion of residues is calculated to be 35 µSv per sampling regimen or 70 µSv/a. Thus, the total annual dose to a characterization worker is calculated to be 1.7 mSv from all pathways, with the external gamma exposure pathway contributing more than 94 % of the dose. 5.4
Estimate of hourly exposures of workers involved in different elements of future decommissioning preparation activities
Estimates of hourly exposures received by workers preparing for D&D activities were made for some hypothetical scenarios and were carried using the methodology and Ecolego-5 safety assessment tool developed within the framework of the Ensure-2 project. To provide a basis for exposure estimates in this assessment, the methods and equations for dose calculations recommended by German Federal Ministry for the Environment, Nature Conservation and Reactor Safety were used [3]. These have been implemented in the software package, Ecolego5[4]. Ecolego-5 provides equations for estimating radiation exposures from all pathways that are relevant at uranium mining and processing sites, namely:
External exposure caused by soil contamination for reference persons inside and outside buildings; Exposure through contaminated aerosols inside and outside buildings; Exposure from the inhalation of 222Rn and its short-lived progeny.
In the present study, dose calculations were only performed for the seven radionuclides: 234 U, 230Th, 226Ra, 210Po and 210Pb.
238
U,
The first D&D worker dose calculation assumes that before any activities for dismantling and demolition of the construction elements will be started, the spilled materials, debris and remained small fragments in the Building 103 will be collected in containers (see Figure 3.16 b) and removed from the shops. The extraction room in every level was divided into 3 categories (relatively clean, average, and high contamination) (see Figures 3.13-3.15 and Table 4.2). Dose assessment results showed on Table 4.2 were related to the averaged conditions of the surface contamination related to level 1 and 2. The next D&D worker dose calculations assumed conditions of the exposure impact to the workers, who will be involved in dismantling of the tanks and pipelines in the relatively clean areas after decontamination and dismantling activities for tanks in the high contaminated areas. The worst case dose calculation assumes exposure conditions for decommissioning workers near tank 59 on Level 1 and tank 103(3) on Level 3 during 1 hour of exposure. The results presented in Figure 5.1 a) show that potential radiation exposures in the different areas of the former uranium extraction shop at the Level 1 and 3 vary from place to place from values of about 1 μSv/h to 1 mSv/h (3 orders of magnitudes). Shown in Figure 5.1 b) are the number of 5-4
hours that can be worked in each area before the limit of 20 mSv/year of allowed exposure for personnel is exceeded. Figure 5.1 c) shows the contribution to the dose from each exposure pathway. It can be seen that in the areas on Level 1 with a relatively low gamma-dose rate, but high contaminated spilled U-production residues, inhalation exposure can contribute up to 70% of total dose, while workers who will be involved in dismantling activities near apparatus 59 on Level 1 (Worst), the external gamma irradiation will be the dominant contributor to dose. On Level 3 in most of the extraction shop areas, the main factor for exposure of workers will be gamma irradiation. The average duration for the safe work time on Level 3, without exceeding the yearly dose limit of 20 mSv/a, is estimated to be approximately 130 hours. The workers involved in the dismantling of the most contaminated tanks may exceed the annual dose limit of 20 mSv/a after only 20 hours of exposure. These estimates indicate very high potential risks for the workers who will be involved in the D&D activities in Building 103 emphasizing the importance of preparing a comprehensive radiation protection plan and providing risk mitigation actions.
Level 1 2,00E-04
2000
1,86E-04
0%
1,80E-04 1,60E-04
1600
Aerosols
1,40E-04 1,20E-04
1400
1116
1200
Rn
1,00E-04
1000
8,00E-05
800
6,00E-05
600
4,06E-05 4,00E-05 2,00E-05
1760
1800
External
492
400
1,79E-05
200
8,74E-07 0,00E+00
108
0
Scenario 1 (average) Averaged
Scenario 2 (minimum) Low
Scenario 2 (maximal) High
Scenario 3 (worst) Worst
Scenario 1
Averaged (average)
Scenario 2 Low (minimum)
a)
Scenario 2 High (maximal)
Scenario 3
Worst (worst)
b) 3% 1%
9%
14%
2%
31%
2%
64%
5%
84%
89%
Averaged
Low gamma dose
c)
High gamma dose
96%
Worst
5-5
Level 3 2000
1,00E-03
9,03E-04 9,00E-04
1800
8,00E-04
1600
7,00E-04
1400
6,00E-04
1200
5,00E-04
1000 800
4,00E-04
3,02E-04
600
3,00E-04 2,00E-04
1760
400
1,51E-04
132
200
1,00E-04
6,53E-07
Scenario 1
Averaged (average)
Scenario 2 Low (minimum)
66
22
0
0,00E+00
Scenario 2 High (maximal)
Scenario 1
Scenario 3 Worst (worst)
(average) Averaged
a) 0% 1%
Scenario 2 (minimum) Low
Scenario 2 (maximal) High
Scenario 3 (worst) Worst
b) 0% 1%
14%
0%
3%
83%
99%
Averaged
99%
Low
c)
High
100%
Worst
Figure 5.1.- Radiological exposure conditions, where a) total doses in Sv per hour; b) the work hour limits for workers not to exceed the 20 mSv/year limit c) contribution of the different exposure pathways the areas with averaged, low and high contaminated and also for the worst scenarios of the potential exposure.
5.5
Estimated Dose to Building 102 Worker During Abnormal Scenarios 1 and 2
5.5.1
Aerosol and Rn concentrations at Building 102 under accidental releases from Building 103
The sources of contamination inside building. Sources of contamination fall into 2 broad categories: 1) contamination of process equipment; 2) contamination of building materials, dust and spills.
5-6
Contamination of process equipment is presently not known. In the present situation, only emissions resulting from contamination of building materials and dust were considered. The release scenarios (1 and 2) were considered for accidental releases:
Strong wind; increased re-suspension and continuous release through the broken windows; Equipment spill: instantaneous release following accidental failure of a structural component and spill of tank(s) contents on floor.
Abnormal Scenario 1 - Strong Wind Measurements of concentrations in air of radon and uranium and other nuclides in aerosols are used as boundary conditions to CFD atmospheric dispersion model describing flow and dispersion in close vicinity of B103 (section 5.4.2). Radon concentration in air inside building is assumed distributed uniformly and equal to 500 Bq/m3 following Ecomonitor monitoring data (2011-2013). Concentration of radionuclides in aerosols is also taken by averaging measurements data from ECOMONITOR as about: 30 mBq/m3 (1st floor), 0.4 mBq/m3 (2nd and 3d floor) for U238; 2 mBq/m3 (1st floor), 0.3 mBq/m3 (2nd and 3rd floor) for Ra-226, 0.5 mBq/m3 (floors 1-3) for Pb-210; other radionuclides (Po-210) were not considered because of their relatively small fraction in aerosols found inside Building 103. The respirable fraction of aerosols had been conservatively taken to be 100 % following (DOE, 1994). It was assumed that all windows on the Northern wall of B103 are opened and it was assumed that in total windows cover 1/4 part of the wall. Recalculation of the results presented below for another fraction of opened windows is straight forward. The flux of radionuclides were parameterized using specific formulas taken from literature as a function of wind velocity close to the wall of Building 103. It should be noted that the aerosol concentration data given above reflect present conditions under normal wind conditions outside Building 103. Expert estimates of potential aerosol activity concentrations under strong winds (because of increased re-suspension of particulates) or during decommissioning activities inside Building 103 are up to one order of magnitude higher without sufficient dust suppression. The highest measured U-238 concentration on the first floor of Building 103 was 73.4 mBq/m3 (Table 3.12), thus for the purposes of this assessment, an aerosol concentration of 100 mBq/m3 U-238, 10 mBq/m3 Ra-226, and 3 mBq/m3 Pb-210 was assumed inside Level 1 of Building 103 during the high wind scenario. Abnormal Scenario 2 - Tank Spill General framework of source term estimations in case of demolitions and dismantling of contaminated facilities is presented in (DOE, 1994). That methodology was applied in (PNNL, 2011) for assessment of contamination following demolition of industrial buildings of the Plutonium Finishing Plant Their approach is expressed by formula: ST = MAR * DR * ARF * RF * LPF
(5.1)
5-7
where source term (ST) is the amount of respirable pollutants released to the atmosphere during the demolition; MAR is the material-at-risk which is equal to the amount of radioactivity available to be acted on by a given physical stress; DR is damage ratio which is equal to the fraction of the MAR actually impacted by the demolition conditions; ARF is airborne release fraction, i.e. the fraction of radioactivity suspended in air as an aerosol and thus available for transport due to a physical stress from a specific activity; RF is respirable fraction which is commonly assumed to include particles 10-μm aerodynamic equivalent diameter (AED) and less; leak path factor (LPF) is the fraction of the radionuclides in the aerosol transported through some confinement system and/or emission mitigation methods (e.g., misters or foggers). The monitoring data related to building No. 103 could be used for assessment of MAR in formula (5.1). While other factors (DR, ARF, RF and LPF) depend on specific demolition plan and they are specified in any case using empirical data presented in (DOE, 1994). The monitoring data in Section 3 of surface contamination and spills containing different nuclides (U-238, Ra-226, Pb-210, Po-210, Th-230, Th-228) in different parts of building 2 were used. With those data it should be possible to estimate an MAR for the scenario assumed. The spill scenario can correspond to the following events: falling of one or more tanks after collapse of a structural component and spill of the contents. Since the first floor is the most contaminated; i.e. has the most spills, it is assumed that the Level 1 extraction area is affected by the accident. .It is assumed that as a result of the structural collapse on Level 1, 8.5 Tonnes of material at an average concentration of 400 Bq/g (estimated from Tables 3.10 and 3.11) are spilled onto the floor. Thus total material-at-risk, MAR, is 3.4 E9 Bq. For powders ARF and RF could be taken from (DOE, 1994). According to data provided, conservative estimates for the ARF=0.01 and RF=0.2 can be made. Both the DR and LPF are assumed to be 1. After multiplication by the ARF and RF factors, a total source term of about ST=7E6 Bq results. According to measurements about 51% of the radionuclide mixture is U-238; 32% is Th-230; 12% is Ra-226; the rest of radionuclides are not considered. The release is considered to be instantaneous. 5.5.2
Atmospheric dispersion assessment methods
For the wind scenario, the Open Foam CFD model (http://www.openfoam.com) has been used to consider the impact of dust from Building 103 to the surrounding areas. For such estimations Gaussian type models were not used as inappropriate. The computational domain of the size 200 x 200 m in horizontal plane and extending up to 300 m vertically covered B103 and several nearby buildings including Building 102 (B102) and Building 104 with grid resolution of 1 m. Simulations had been performed for 8 wind directions of the ambient (unperturbed by buildings) wind and for wind velocities of 0.5, 1, 2.5, 4.5, 7.5, 12.5, 20 m/s; neutral conditions had been 5-8
assumed in all runs. The results of all simulations (i.e. 56 runs corresponding to different wind speeds and wind directions) had been averaged with weights obtained from frequencies of the corresponding wind conditions reported by Dneprodzerzhinsk meteorological station. For the spill scenario, we are most interested in concentrations at distances of about 50-1000 m from B103 and thus Gaussian puff models could be used for such assessments. We performed long integration of CALPUFF model using data of Dneprodzerjinsk meteorlogical station for the period 2005-2009 available each 3-hours and GFS final analysis data for upper-air conditions. Each 3 hours the same amount of radioactivity (as specified in previous section) is released in the form of PM10. During 3 hours all matter from the last release is taken out from the computational domain by wind. Thus actually we perform series of releases with different meteorological conditions. Thus obtained maximum concentrations well represent the worst case meteorological scenario. The average concentration obtained in such simulation could be represented as:
Cavr x, y
N
f C x, y i
i
i 1
N
fi 1; Ci 1 Ci C
(5.2)
i 1
where fi is probability of ground-level concentration C (x,y) to fall in a range: Ci C / 2 . Thus the field of averaged concentrations Cavr x, y could be used for risk assessment estimates following the accidental demolition scenario. 5.5.3
Results of the dispersion calculations
Abnormal Scenario 1 – Strong Wind Results obtained for the case of scenario involving wind are presented on Figures 5.2 and 5.3 and in the Table 5.5. Despite the asymmetric location of Building 103 with respect to center of Building 102, the yearly averaged concentration on the wall of Building 102 is fairly symmetric (Figure 5.2). This is the consequence of the very intensive mixing in street canyon. Also from Fig. 5.3 we see that significant concentration values are reached not only between Building 102 and Building 103, but also to the West and to the East sectors of Building 103. The corresponding values are presented in Table 5.5.
5-9
z, m
40
20 0.025 0.05 0 -150
-100
0.1
0.2
0.2 -50
0
50
100
150
x, m
Figure 5.2 Isolines of C/Cb (annually averaged) near the wall of Building 102
Figure 5.3 - Ground level annually averaged activity concentration field; isolines are shown for C/Cb=0.25 and C/Cb=0.5; maximum value is C/Cb=2.5 (in light color).
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Table 5.5
Results of the simulated radionuclide activity concentrations (near Building 102 (B102) affected due to contaminated dust releases from the Building 103 (B103) the obtained for the case of wind scenario Description
Rn-222 Bq/m
Yearly average concentration, averaged on the wall of B102 Yearly average ground level concentration averaged on the wall of B102 Maximum concentration on the wall of B102 Yearly average ground level concentration between B102 and B103 Maximum concentration in 20 m of the East wall of B103 Maximum concentration in 20 m of the West wall of B103
3
U-238 Ra-226 Pb-210 mBq/m3 mBq/m3 mBq/m3
62.5
0.88
0.25
0.09
125
1.75
0.5
0.18
875
12.2
3.5
1.23
187 125 140
2.63 1.75 1.96
0.75 0.5 0.56
0.26 0.18 0.2
The maximum aerosol concentration on the wall of Building 102 from average aerosol concentrations in Building 103 are given in the third line of Table 5.5. Under high wind conditions, it is estimated that the maximum concentrations at Building 102 are 2.5 times higher than the concentrations inside Building 103 or 250 mBq/m3 U-238, 25 mBq/m3 Ra-226, and 8 mBq/m3 Pb-210. It is assumed that the same concentrations are inside Building 102 and last for 40 hours. Abnormal Scenario 2 – Tank Spill The computational domain is shown on Figure 5.2. The concentrations modeling results were compared with monitoring data and as results allowing validate model and estimate dispersion of the dust from Building 103 at distances until 1000 m. The dependence of maximum yearly C avr
max
concentrations ( C1h ) and of yearly averaged concentrations ( 1 year ) on distance from the source is presented in Table 5.6. Note that spatial distributions of both calculated parameters are max
C avr
somewhat asymmetric, therefore for both parameters C1h and 1 year the maximum value at a given distance is presented in Table 5.6. As it was discussed above (formula (5.2) the yearly averaged concentrations could be used for the assessment of collective dose that could be received following the tank spill.
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Table 5.6
Dependence of maximum hourly and yearly averaged concentrations on distance from Building 103 following accidental demolition scenario
3 Distance, Maximum hourly concentrations, Bq/m m U-238 Ra-226 Pb-210 Po-210 Th-230 0 28.6 6.72 6.72 6.72 17.9 50 21.6 5.10 5.10 5.10 13.6 100 15.1 3.56 3.56 3.56 9.47 300 6.32 1.49 1.49 1.49 3.97 500 2.86 0.67 0.67 0.67 1.79 800 1.25 0.29 0.29 0.29 0.79 1000 0.78 0.18 0.18 0.18 0.49
5.5.4
Yearly averaged concentrations, Bq/m3 U-238 0.040 0.030 0.024 0.007 0.002 0.001 0.0006
Ra-226 0.0096 0.007 0.006 0.002 0.0006 0.0003 0.0002
Pb-210 Po-210 Th-230 0.0096 0.0096 0.026 0.007 0.007 0.019 0.006 0.006 0.015 0.002 0.002 0.005 0.0006 0.0006 0.0015 0.0003 0.0003 0.0007 0.0002 0.0002 0.0004
Dose Calculations for Abnormal Scenarios 1 and 2
Abnormal Scenario 1 – Strong Wind Based on the concentrations estimated from section 5.5.3, the inhalation doses from the strong wind scenario for occupants of Building 102 based on the inhalation of 50 m3 of air are the following: Table 5.7
Doses from Inhalation of Aerosols in Building 102 from increased emissions from Building 103 Area
Radionuclide Activities (Bq/m3) U-238
Ra-226
Pb-210
Rn-222
NW corner
250
25
8
875000
Inhalation DCFs (Sv/Bq)
2.9E-06
3.5E-06
1.1E-06
2.5E-09
Dose (µSv)
36
4.4
0.44
110
Total
151
Thus, the total estimated increased aerosol inhalation dose to Building 102 occupants during strong wind conditions is 41 µSv. The radon dose is 110 µSv. Abnormal Scenario 2 – Tank Spill Inhalation doses following a tank spill depending on the location of the worker. Someone inside the building would receive the largest dose. Table 5.8 shows the calculated individual doses following a one hour exposure from a puff release following a tank spill. An average of 5-12
concentrations at 50 m and 100 m were used for Building 102 activities and an average of the concentrations at distances between 300 m and 1000 m were used for the areas around the buildings. Table 5.8
Doses from Inhalation of Aerosols at various distance following a tank spill in Building 103 Radionuclide Activities (Bq/m3)
Inhalation Dose (µSv) per Individual
Area U-238
Ra-226
Pb-210
Po-210
Th-230
Inhalation DCFs (Sv/Bq)
U-238
Ra-226
Pb-210
Po-210
Th-230
2.9E-06
3.5E-06
1.1E-06
3.3E-06
4.0E-05
Total
Inside Building
29
6.7
6.7
6.7
18
110
31
10
29
940
1100
Building 102
18
4.3
4.3
4.3
12
68
20
5
18
620
730
Area around Buildings 102 and 103
2.8
0.66
0.66
0.66
1.8
11
3
1
3
92
110
5.6
Estimated Dose for Abnormal Scenario 3 - Intruder
The dose to an intruder stealing equipment from Building 103 is expected to be almost entirely from the external gamma pathway. This scenario only calculates exposure during the time taken for the removal of the equipment. It doesn’t take into account the exposure received by the worker from residual radioactivity in the stolen equipment. This can potentially result in significantly more dose than was received during the removal. The gamma dose rates in the vicinity of Tank №130/3 have exceeded 1 mSv/h. It is assumed that the worker is in a radiation field of 1 mSv for a period of one to two hours. Thus the dose received from external gamma radiation will be between 1 and 2 mSv, depending on the time spent. Although exposures from aerosol and radon inhalation and inadvertent ingestion of residues are also exposure pathways, from the results of the exposures received by the characterization worker and the much shorter time spent in Building 103, they are minor contributors to the dose.
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6
CONCLUSIONS
The work summarized in this report has shown that significant doses (of the order of 1 mSv/a) can be received from personnel inside or in the vicinity of Building 103 in its present state. The estimated doses for normal and abnormal conditions are summarized in Table 6.1. Table 6.1 Scenario
Calculated Doses to PChP Personnel from Building 103 Receptor
Calculated Dose (mSv/a)
Main Pathway(s)
Normal Conditions Normal Scenario 1
Building 102 worker
1.6
Gamma, Radon
Normal Scenario 2
Characterization Worker
1.7
Gamma
Inside Building 103
D&D Preparation
1 µSv/h – 1 mSv/h
Gamma
Accident Conditions Abnormal Scenario 1 Strong Wind
Building 102 Worker
0.041
Aerosols
Abnormal Scenario 2 Tank Spill
PChP Worker (inside and outside)
0.1 to 1
Aerosols
Abnormal Scenario 3 Intrusion
PChP Worker
1-2
Gamma
The results show that high existing doses are mainly the result of very high activity concentration of the remaining materials in the tanks and spilled materials in the vicinity of tanks and premises of the former Uranium extraction facilities. It can be seen that under existing normal conditions, the chronic exposures to a Building 102 workers are similar to those expected from the occasional exposure by a characterization worker within Building 103. The intrusion scenario contributes the highest dose under abnormal conditions and in the case of a tank spill only the worker inside Building 103 would receive dose similar to that expected under normal conditions. In terms of preparation for decommissioning activities, preliminary radiological assessment for workers in the Building 103 at different levels of the building show that potential radiation exposures in the different areas of the former uranium extraction shop at the Level 1 and 3 vary from place to place from values of about 1 μSv/h to 1 mSv/h (3 orders of magnitudes).
6-1
It can be seen that in the areas on Level 1 with a relatively low gamma-dose rate, but high contaminated spilled U-production residues, inhalation exposure can contribute up to 70% of total dose (from aerosols and radon), while, for workers who will be involved in dismantling activities near apparatus 59 on Level 1 (Worst), the external gamma irradiation will be the dominant contributor to dose. The external gamma dose expose is a main factor of radiological impact in the all areas of Building 103. The contribution from inhalation exposure pathways on Level 1, varies from 70% in areas with large spills and relatively low external gamma radiation impacts only 10% near the tanks with high gamma dose rate. On Level 3, the main exposure factor is the high gamma radiation. Inhalation exposure contributes from 1 to 10% of the total 1 hour exposure to personal staying in the Building. Safe work times on Level 3 have been calculated to be between 20 and 130 hours, without exceeding the yearly dose limit of 20 mSv/a. These estimates indicate very high potential risks for the workers who will be involved in the D&D activities in Building 103 emphasizing the importance of preparing a comprehensive radiation protection plan and providing risk mitigation actions. In addition, a comprehensive waste management plan is needed.
6-2
7
TASKS FOR COMPREHENSIVE ASSESSMENT AND FURTHER STUDIES
The work in this report served to demonstrate an assessment methodology that can be applied to the decontamination and decommissioning of a contaminated building on the PChP site. Building 103 was used to demonstrate the methodology since it has been most extensively characterized and is designated as the highest priority building in need of remediation. The impacts of the building in its present state (the DO NOTHING) option were analysed for a number of scenarios, normal and accidental. In order to complete a comprehensive assessment, details must be known about the various activities and sequence of activities during the decontamination and the decommissioning of Building 103. In addition, a detailed radiation protection plan must be in place to detail the controls on the workers conducting the activities. Controls such as dust suppression, ventilation and shielding must also be in place. A comprehensive monitoring program to determine conditions inside and outside the building must also be present, as well as a dosimetry program. Action levels must also be established. The tools used in the present assessment can be used for the comprehensive assessment with minor adaptations. Safran is scheduled to be adapted to be more flexible in addressing the various activities within building 103 during its decommissioning. It will be adapted to allow the creation of more detailed areas with differing characteristics, the movement of materials out of an area with the result of changing characteristics for the area, and the inclusion of additional calculational tools. The possibility of linking to different data sets from various areas will be included. Additional characterization work is necessary to characterize the contents of the different tanks. A waste management strategy is needed as well as a location for the decontamination of equipment, either inside the building or at a centralized remote location.
7-3
8
REFERENCES
ARPANSA (2005), Code of Practice and Safety Guide: Radiation Protection and Radioactive Waste Management in Mining and Mineral Processing, Radiation Protection Series Publication No. 9, Australian Radiation Protection and Nuclear Safety Agency, Victoria, Australia. BNFL (2002), Drigg Post-closure Safety Case, British Nuclear Fuels plc, Sellafield, United Kingdom. CNSC (2006), ssessing the Long Term Safety of Radioactive Waste Management, Regulatory Guide G–320, Canadian Nuclear Safety Commission, Ottawa, Canada. Cranwell, R. M., R. W. Guzowski, J. E. Campbell, and N. R. Ortiz (1990), Risk Methodology for Geological Disposal of Radioactive Waste: Scenario Selection Procedures, Report No. NUREG/CR-1667. SAND80-1429, Sandia National Laboratories, Albuquerque, NM 87185. Galson, D. A., and P. N. Swift (1994), Scenario Development for the Waste Isolation Pilot Plant: Building Confidence in the Assessment, SAND94-0482, Sandia National Laboratories. IAEA (1996), International Safety Standards for Protection against Radiation and for the Safety of Radiation Sources, International Atomic Energy Agency Basic Safety Standards Report No. 115, International Atomic Energy Agency, Vienna. IAEA (2004a), Radiation, People and the Environment, IAEA/PI/A.75/ 04-00391, International Atomic Energy Agency, Vienna. IAEA (2004b), Safety Assessment Methodologies for Near Surface Disposal Facilities. Results of a Co-ordinated Research Project. Volume I: Review and Enhancement of Safety Assessment Approaches and Tools, IAEA-ISAM, International Atomic Energy Agency, Vienna. IAEA (2006), Fundamental Safety Principles Safety Standard Series No. SF-1, International Atomic Energy Agency, Vienna, Austria. IAEA (2011), Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards: General Safety Requirements, IAEA Safety Standards Series No. GSR Part 3 (Interim), International Atomic Energy Agency, Vienna, Austria. IAEA (2012), The safety case and safety assessment for the disposal of radioactive waste : specific safety guide, Safety Standard Series No. SSG-23, International Atomic Energy Agency, Vienna, Austria. ICRP (1991), 1990 Recommendations of the International Commission on Radiological Protection. Annals of the ICRP 21 (1-3), ICRP Publication 60, International Commission on Radiological Protection. 8-1
ICRP (1998), Radiation Protection Recommendations as Applied to the Disposal of Longlived Solid Radioactive Waste, ICRP Publication 81, International Commission on Radiological Protection, Pergamon Press, Oxford. ICRP (2000), Publication 82. Protection of the Public in Situations of Prolonged Radiation Exposure. The Application of the Commission's System of Radiological Protection to Controllable Radiation Exposure Due to Natural Sources and Long-Lived Radioactive Residues. Annals of the ICRP, First ed., Elsevier Science Ltd, Oxford. ICRP (2007), Publication 103. Recommendations of the ICRP - Annals of the International Commission on Radiological Protection (ICRP), Published for the ICRP by Elsevier Inc., Vienna. Kathren, R. L. (1998), NORM Sources and Their Origins, Applied Radiation and Isotopes, 49(3), 149-168. Kozak, M. W. (1994), Decision analysis for low-level radioactive waste disposal safety assessments, Radioactive Waste Management and Environmental Restoration, 18, 209–223. Martin, J. E. (2006), Physics for Radiation Protection: A Handbook. Second Edition, Completely Revised and Enlarge, Wiley-VCH, Weinheim. NCRP (2005), Performance Assessment of Near-Surface Facilities for Disposal of Low-Level Radioactive Waste, NCRP REPORT No. 152, National Council on Radiation Protection and Measurements, Bethesda. NEA (1997), Lessons learnt from ten performance assessment studies. Report of the NEA working group on integrated performance assessment of deep repositories, Nuclear Energy Agency, Organisation for Economic Co-operation and Development, Paris. Savage, D. (1995), The Scientific and Regulatory Basis for the Geological Disposal of Radioactive Waste, John Wiley & Sons. Skagius, K., A. Ström, and M. Wiborgh (1995), The Use of Interaction Matrices for Identification, Structuring and Ranking of FEPs in a Repository System: Application on the Far-field of a Deep Geological Repository for Spent Fuel, SKB Technical Report 95–22, Swedish Nuclear Fuel and Waste Management Company, Stockholm, Sweden. SSM (2008), Swedish Radiation Safety Authority Regulatory Code, SSMFS 2008:37, Swedish Radiation Safety Authority, Sweden.
References to Section 3
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CREM (2009), Carrying out estimation of technical and economic feasibility of secondary extraction of uranium and other rare-earth elements from materials of tailings sites. Experimental and technological tests. Final report (Yu.N.Soroka, A.I.Molchanov). Center for Radioecological Monitoring (CREM), Zhovty Vody, 2009 (in Russian). GEC (2009), Modeling predictions of radionuclide transport from tailings sites by water pathways with the aim of selection of technology of converting tailings sites to ecologicallysafe state. Report on contract No.1/09 with SE Baryer. Geo-Eco-Consulting, Kiev, 2009 (in Ukrainian). ECOMONITOR (2007), Development of the concept and structure of the State purposeful ecological program “Conservation, liquidation or conversion and transfer to ecologically safe conditions of the objects of the former industrial association Pridneprovsky Chemical Plant. Part 1. Analysis of the status of the problem and substantiation of main principles of the new program of measures for the period until 2030. Ecomonitor, Kiev, 2007 (in Ukrainian). ECOMONITOR (2008), Carrying out works in accordance with programs and procedures of radiation monitoring (O.Voitsekhovitch et al.). Report on contract with SE Baryer from 28.05.08. Ecomonitor, Kiev, 2008 (in Ukrainian). ECOMONITOR (2009), Carrying out works in accordance with programs and procedures of radiation monitoring. Completion of programs of radiation monitoring of uranium objects: at the Industrial Site of the PChP; in the zone of influence of Sukhachevskoe Tailings and Base C storage site. Assessment of their environmental impact. (O.Voitsekhovitch et al.). Report on contract with SE Baryer from 04.06.09. Ecomonitor, Kiev, 2009 (in Ukrainian). ECOMONITOR (2010), Carrying out observations of changes of environmental parameters at uranium sites and in their sanitary-protection zones, assessment of their environmental impact. (O.Voitsekhovitch et al.). Report on contract with SE Baryer No.43 from 24.11.10. Ecomonitor, Kiev, 2010 (in Ukrainian). ECOMONITOR (2012a) Pridneprovsky Chemical Plant uranium production legacy site characterization report. (Prepared for support of IAEA expert mission 20-24 February 2012). (by O.Voitsehovitch). Ecomonitor, 2012. ECOMONITOR (2012b) Techical note on site-specific hydrogeology characterization studies at Zapadnoe Tailings: hydraulic slug-tests and tracer tests on monitoring wells. ENSURE II Project, Task 2.3.5. Ecomonitor, 2012. NIPIPT (2001), assessment for radioactive waste management practices. Zapadnoe Tailings. Vol.1. Analysis of conditions of radioactive waste storage. NIPIPT Institute. Inv. No. A-15969. 2001 (in Russian). 8-3
NUBIP (2009). Study of physical and chemical forms and prediction of transformations of radionuclides of uranium – thorium series in tailings. Report on the contract no.32 with SE Baryer from 16.07.2009 (Sci. supervisor V. Kashparov). National University of Bio-resources and Environmental Management of Ukraine (NUBIP), 2009 (in Ukrainian). UHMI (2009), Analytical measurements of samples of material from tailings according to parameters of spectrum of priority alfa- and gamma- emitting radionuclides of U – Th series using methods of low-background semiconductor alfa- and gamma- spectroscopy. Final report on contract No.31 with SE Baryer (Sci. supervisor V.G.Laptev), UHMI Institute, 2009 (in Ukrainian).
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9
ANNEXES
9-1
