[Prospects of systemic radioecology in solving innovative tasks of nuclear power engineering].

A need of systemic radioecological studies in the strategy developed by the atomic industry in Russia in the XXI century has been justified. The priorities in the radioecology of nuclear power engineering of natural safety associated with the development of the radiation-migration equivalence concept, comparative evaluation of innovative nuclear technologies and forecasting methods of various emergencies have been identified. Also described is an algorithm for the integrated solution of these tasks that includes elaboration of methodological approaches, methods and software allowing dose burdens to humans and biota to be estimated. The rationale of using radioecological risks for the analysis of uncertainties in the environmental contamination impacts,at different stages of the existing and innovative nuclear fuel cycles is shown.

Evaluation and measurements of radioactive air emission and off-site doses at SLAC.

SLAC, a high-energy (GeV) electron accelerator facility, performs experimental and theoretical research using high-energy electron and/or positron beams that can produce secondary neutron and gamma radiation when beam losses occur. Radioactive gas production (mainly C, N, O, Ar) and release is one of the environmental protection program issues. U.S. DOE Order 458.1 requires that 40 CFR 61 Subpart H's NESHAP requirements be followed. These regulations prescribe a total dose limit of 0.1 mSv y to the Maximally Exposed Individual (MEI) of the general public, a requirement for a continuous air monitoring system if a release point within a facility can cause > 1 × 10 mSv y to the MEI, and a requirement for periodic confirmatory measurements for minor sources which give releases that contribute ≤ 1 × 10 mSv y to the MEI. At SLAC, all air release points for current operations are evaluated to be minor sources. This paper describes SLAC's evaluation following NESHAP requirements; measurements using the Air Monitoring Station (AMS) as periodic confirmatory measurements are also discussed.

A best fit approach to estimating multiple diffuse source terms using ambient air monitoring data and an air dispersion model.

Lawrence Livermore National Laboratory uses CAP88-PC Version 1.0 modeling software to demonstrate compliance with the Code of Federal Regulations Title 40 Part 61 Subpart H (National Emission Standards for Emissions of Radionuclides Other Than Radon From Department of Energy Facilities). Annual air emissions from both well characterized stack sources and difficult to characterize diffuse sources must be assessed. This paper describes a process that uses a mathematical optimization routine to find a set of estimated diffuse source terms that together with the measured stack source terms provides a best fit of modeled air concentrations to measured air concentrations at available sampling locations. The estimated and measured source terms may then be used in subsequent CAP88-PC modeling to estimate dose at the off-site maximally exposed individual. LLNL has found this process to be an effective way to deal with the required assessment of diffuse sources that have otherwise been difficult to assess.

A comparison of dose results from the Clean Air Act Assessment Package-1988, personal computer (CAP88-PC), version 3 to previous versions.

Computer software packages approved by the U.S. Environmental Protection Agency (U.S. EPA), including CAP88-PC, are used by U.S. Department of Energy (U.S. DOE) sites to demonstrate compliance with the radionuclide air emission standard under the Clean Air Act. CAP88-PC version 3, was approved by the U.S. EPA in February 2006 for use by U.S. DOE facilities. Version 3 incorporates several major changes that have the potential to affect calculated doses relative to calculations using earlier versions. This analysis examined the types and magnitudes of changes to dose estimates for specific radionuclides calculated using the version 3 software compared with the previous versions. For parent radionuclides and for the total dose from radionuclide chains, total effective dose calculated with version 3 was compared to effective dose equivalent calculated with previous versions. Various comparisons were also performed to determine which of the updates in version 3 accounted for changes in overall dose estimates. CAP88-PC version 3 would produce substantially different results relative to previous versions of the code for a number of radionuclides, including some isotopes that may be present at U.S. DOE facilities, as well as those used for industrial and medical applications. In general, doses for many radionuclides were lower using version 3 but doses for a few key radionuclides increased relative to the previous versions.

Comparison of CAP88 PC Ver. 3.0 and MAXDOSE dose assessment models involving co-located stack releases at the Savannah River site.

The Savannah River National Laboratory's Environmental Dosimetry Group performs dosimetry assessments for Savannah River Site (SRS) radionuclide air emissions utilizing the Clean Air Act Assessment Package-1988 (CAP88) code (CAP88 PC Ver. 3.0) and the MAXDOSE-SR Ver. 2011 code, which is an SRS-specific version of the Nuclear Regulatory Commission's MAXIGASP code. CAP88 PC and MAXDOSE-SR are used at SRS for demonstrating compliance with Environmental Protection Agency dose standards for radionuclide emissions to the atmosphere and Department of Energy Order 458.1 dose standards, respectively. During a routine comparison of these two assessment models, it was discovered that CAP88 PC Ver. 3.0 was not producing the expected results when using multiple co-located stacks in a single run. Specifically, if the stack heights are considered separately, the results for several radionuclides (but not all) differ from the combined run [i.e., 1 + 2 does not equal (1+2)]. Additionally, when two or more stack heights are considered in a run, the results depend on the order of the selected stack heights. For example, for a two stack-height run of 0 meter and 61 m input produces different results from a 61 m and 0 m input run. This study presents a comparison of CAP88 PC Ver. 3.0 and MAXDOSE-SR Ver. 2011 based on SRS input data and on two-stack release scenarios. The selected radionuclides for this study included gases/vapors (H, C, Kr, and I) and particulates (Sr, Cs, Pu, and Am) commonly encountered at SRS.

Validation test for CAP88 predictions of tritium dispersion at Los Alamos National Laboratory.

Gaussian plume models, such as CAP88, are used regularly for estimating downwind concentrations from stack emissions. At many facilities, the U.S. Environmental Protection Agency (U.S. EPA) requires that CAP88 be used to demonstrate compliance with air quality regulations for public protection from emissions of radionuclides. Gaussian plume models have the advantage of being relatively simple and their use pragmatic; however, these models are based on simplifying assumptions and generally they are not capable of incorporating dynamic meteorological conditions or complex topography. These limitations encourage validation tests to understand the capabilities and limitations of the model for the specific application. Los Alamos National Laboratory (LANL) has complex topography but is required to use CAP88 for compliance with the Clean Air Act Subpart H. The purpose of this study was to test the accuracy of the CAP88 predictions against ambient air measurements using released tritium as a tracer. Stack emissions of tritium from two LANL stacks were measured and the dispersion modeled with CAP88 using local meteorology. Ambient air measurements of tritium were made at various distances and directions from the stacks. Model predictions and ambient air measurements were compared over the course of a full year's data. Comparative results were consistent with other studies and showed the CAP88 predictions of downwind tritium concentrations were on average about three times higher than those measured, and the accuracy of the model predictions were generally more consistent for annual averages than for bi-weekly data.

Expert assessments of the cost of light water small modular reactors.

Analysts and decision makers frequently want estimates of the cost of technologies that have yet to be developed or deployed. Small modular reactors (SMRs), which could become part of a portfolio of carbon-free energy sources, are one such technology. Existing estimates of likely SMR costs rely on problematic top-down approaches or bottom-up assessments that are proprietary. When done properly, expert elicitations can complement these approaches. We developed detailed technical descriptions of two SMR designs and then conduced elicitation interviews in which we obtained probabilistic judgments from 16 experts who are involved in, or have access to, engineering-economic assessments of SMR projects. Here, we report estimates of the overnight cost and construction duration for five reactor-deployment scenarios that involve a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants vary by more than a factor of 2.5. Expert judgments about likely SMR costs display an even wider range. Median estimates for a 45 megawatts-electric (MWe) SMR range from $4,000 to $16,300/kWe and from $3,200 to $7,100/kWe for a 225-MWe SMR. Sources of disagreement are highlighted, exposing the thought processes of experts involved with SMR design. There was consensus that SMRs could be built and brought online about 2 y faster than large reactors. Experts identify more affordable unit cost, factory fabrication, and shorter construction schedules as factors that may make light water SMRs economically viable.

Regulatory experience in applying a radiological environmental protection framework for existing and planned nuclear facilities.

Frameworks and methods for the radiological protection of non-human biota have been evolving rapidly at the International Commission on Radiological Protection and through various European initiatives. The International Atomic Energy Agency has incorporated a requirement for environmental protection in the latest revision of its Basic Safety Standards. In Canada, the Canadian Nuclear Safety Commission has been legally obligated to prevent unreasonable risk to the environment since 2000. Licensees have therefore been meeting generic legal requirements to demonstrate adequate control of releases of radioactive substances for the protection of both people and biota for many years. In the USA, in addition to the generic requirements of the Environmental Protection Agency and the Nuclear Regulatory Commission, Department of Energy facilities have also had to comply with specific dose limits after a standard assessment methodology was finalised in 2002. Canadian regulators developed a similar framework for biota dose assessment through a regulatory assessment under the Canadian Environmental Protection Act in the late 1990s. Since then, this framework has been applied extensively to satisfy legal requirements under the Canadian Environmental Assessment Act and the Nuclear Safety and Control Act. After approximately a decade of experience in applying these methods, it is clear that simple methods are fit for purpose, and can be used for making regulatory decisions for existing and planned nuclear facilities.

[Radioecological approaches to ranking radiation dangerous objects].

The paper gives complex criteria for evaluating the hazard of radiation dangerous objects (RDO). The proposed criteria include the following indicators: the ratio of the cumulative activity of radioactive waste to a hazard factor (D value) or to the allowable level of i-radionuclide in the storage; the power of an effective gamma-radiation dose; the rate of radionuclide migration; the doses of human radiation. A scoring system for the hazard of RDO from the above indicators is given.

Sixth Warren K. Sinclair keynote address: The role of a strong regulator in safe and secure nuclear energy.

The history of nuclear regulation is briefly reviewed to underscore the early recognition that independence of the regulator was essential in achieving and maintaining public credibility. The current licensing process is reviewed along with the status of applications. Challenges faced by both the NRC and the industry are reviewed, such as new construction techniques involving modular construction, digital controls replacing analog circuitry, globalization of the entire supply chain, and increased security requirements. The vital area of safety culture is discussed in some detail, and its importance is emphasized.

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described.

U.S. Department of Energy facilities needed to advance nuclear power.

This talk is based upon a November 2008 report by the U.S. Department of Energy (DOE) Nuclear Energy Advisory Committee (NEAC). The report has two parts, a policy section and a technology section. Here extensive material from the Technical Subcommittee section of the NEAC report is used.

Licensing a new industrial irradiator.

After nearly three decades of medical product sterilization, 3M launched a major new project to build and license an irradiator facility. 3M Corporate Health Physics was responsible for the licensing aspect of this project. The licensing process consisted of six amendments, over 30 submissions to the U.S. Nuclear Regulatory Commission (U.S. NRC) and four U.S. NRC site visits. It took approximately 22 months to complete. The six license amendments are reviewed and several of the submissions are discussed. These include 3M's response to the U.S. NRC's interest in the shielding calculations used for the bioshield, the development of a protocol of radiation safety system test methods, and an analysis to show that a dropped cask during loading operations would not fall on sealed sources. A number of lessons were learned during the course of licensing the new irradiator. Among these were the importance of understanding the U.S. NRC license reviewer's perspective, the need to thoroughly review the irradiator manufacturer's licensing package during project negotiations, the benefits of leaving the Health Physics Office and meeting with the non-health physicists involved in the project, and the necessity of maintaining the solid relationships that already existed with the site Radiation Safety Officer and Sterilization Engineer.

[Ecological estimation of a condition of protective barriers of storage at long-term localization of radioactive waste].

Data on change of a condition of natural and artificial protective barriers during long-term storage of the Russian Open Society in storehouses (see text) type are resulted.

Depleted zinc: Properties, application, production.

The addition of ZnO, depleted in the Zn-64 isotope, to the water of boiling water nuclear reactors lessens the accumulation of Co-60 on the reactor interior surfaces, reduces radioactive wastes and increases the reactor service-life because of the inhibitory action of zinc on inter-granular stress corrosion cracking. To the same effect depleted zinc in the form of acetate dihydrate is used in pressurized water reactors. Gas centrifuge isotope separation method is applied for production of depleted zinc on the industrial scale. More than 20 years of depleted zinc application history demonstrates its benefits for reduction of NPP personnel radiation exposure and combating construction materials corrosion.

Radiological criteria for the remediation of sites for spent fuel and radioactive waste storage in the Russian Northwest.

In the 1960s, two technical bases of the Northern Fleet were created in Northwest Russia, at Andreeva Bay in the Kola Peninsula and Gremikha village on the coast of the Barents Sea. They maintained nuclear submarines, performing receipt and storage of radioactive waste and spent nuclear fuel, and are now designated sites of temporary storage (STSs). An analysis of the radiation situation at these sites demonstrates that substantial long-term remediation work will be required after the removal of the waste and spent nuclear fuel. Regulatory guidance is under development to support this work. Having in mind modern approaches to guaranteeing radiation safety, the primary regulatory focus is on a justification of dose constraints for determining acceptable residual contamination which might lead to exposure to workers and the public. For these sites, four principal options for remediation have been considered-renovation, conversion, conservation and liquidation. This paper describes a system of recommended dose constraints and derived control levels formulated for each option. The unconditional guarantee of long-term radioecological protection provides the basis for criteria development. Non-exceedance of these dose constraints and control levels implies compliance with radiological protection objectives related to the residual contamination. Dose reduction below proposed dose constraint values must also be carried out according to the optimisation principle. The developed criteria relate to the condition of the facilities and the STS areas after the termination of remediation activities. The proposed criteria for renovation, conversion, conservation and liquidation are entirely within the dose limits adopted in Russia for the management of man-made radiation sources, and are consistent with ICRP recommendations and national practice in other countries. The proposed criteria for STS remediation and new industrial (non-radiation-hazardous) facilities and buildings on the remedied sites had, until now, no analogues in the Russian system of regulation of radiation-hygienic standardisation. The proposals made here may serve as a basis for corresponding standards at other sites.

Statistical criteria to set alarm levels for continuous measurements of ground contamination.

In the course of the decommissioning of the ASTRA research reactor at the site of the Austrian Research Centers at Seibersdorf, the operator and licensee, Nuclear Engineering Seibersdorf, conducted an extensive site survey and characterization to demonstrate compliance with regulatory site release criteria. This survey included radiological characterization of approximately 400,000 m(2) of open land on the Austrian Research Centers premises. Part of this survey was conducted using a mobile large-area gas proportional counter, continuously recording measurements while it was moved at a speed of 0.5 ms(-1). In order to set reasonable investigation levels, two alarm levels based on statistical considerations were developed. This paper describes the derivation of these alarm levels and the operational experience gained by detector deployment in the field.