Approach to Derivation of Waste Acceptance Criteria for Mochovce Disposal Facility

2003 ◽  
Author(s):  
V. Hanusˇi´k ◽  
Z. Kusovska´ ◽  
J. Bala´zˇ ◽  
A. Mrsˇkova´
Keyword(s):  
Fission Products ◽  
Total Activity ◽  
Safety Analysis ◽  
Radioactive Wastes ◽  
Primary Circuit ◽  
Radiological Protection ◽  
Waste Streams ◽  
Acceptance Criteria ◽  
Near Surface ◽  
Dose Limits

In Slovakia, low and intermediate level radioactive wastes are disposed in a near-surface repository at Mochovce site. The repository, which was commissioned in September 2001, has a disposal capacity 22,320 m3. It is a vault-type concrete structure repository with reinforced concrete containers as the final waste packages. The Mochovce repository is designed to receive acceptable radioactive wastes from decommissioned A-1 power plant at Jaslovske´ Bohunice, operational waste from NPPs V-1 and V-2 at Jaslovske´ Bohunice site and NPP Mochovce, as well as institutional radioactive wastes. Generally, calculation endpoint of disposal facilities performance assessment is radiological impact on humans and environment. In that case, starting points of assessment are the waste activity concentrations and inventory activity. The acceptance of radioactive waste in Mochovce repository is one of the many elements that directly contribute to the safety of the disposal system. In Mochovce repository safety analysis, end points are both the concentration per package and total activity values. On the other hand, radiological protection criteria are the starting points of the calculation. This approach was developed and applied because the actual inventory that will be disposed of is highly uncertain. As a result of the accidents, the primary circuit was contaminated by fission products. Some auxiliary circuits and facilities were also contaminated. The complicated problem is the relatively high content of long-lived radionuclides (inclusive transuranic elements) in some waste streams. After two technological incidents at NPP A-1 uncertainties in waste inventory are large because of variability in the types of waste streams and variability in the quality and completeness of the waste characterization data. This paper presents the philosophy of safety analysis, development of scenarios, their modelling and approach that have been used to derive waste acceptance criteria, specifically limits of activity. The approach consists of the determination of radionuclides important for safety, the use of relevant safety scenarios, the setting of dose limits associated with scenarios, the calculation of activity limits and application of the simple summation rule. Finally, information is provided about short operation of the repository.

Problems of Radioactive Sodium Removal and Disposal From Decommissioning LMFR Equipment

1997 ◽  
Vol 506 ◽  
Author(s):  
Yu. Ye. Shtynda ◽  
V. I. Polyakov
Keyword(s):  
Radioactive Wastes ◽  
Primary Circuit ◽  
Minimum Volume ◽  
Safe Disposal ◽  
Radionuclide Sorption ◽  
Sodium Removal ◽  
Radioactive Sodium

ABSTRACTPreparing for safe disposal of the LMFR primary circuit equipment and sodium reprocessing for storage and burial with minimum volume of radioactive wastes resulted in testing of radionuclide sorption trapping, distillation and rinsing with water under vacuum used for safe sodium removal and decontamination of equipment.

scholarly journals A Safety Re-Evaluation of the AVR Pebble Bed Reactor Operation and Its Consequences for Future HTR Concepts

2008 ◽  
Author(s):  
Rainer Moormann
Keyword(s):  
Fission Product ◽  
End Of Life ◽  
Fission Products ◽  
Normal Operation ◽  
Primary Circuit ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Hot Gas ◽  
Design Basis ◽  
Gas Tight

The AVR pebble bed reactor (46 MWth) was operated 1967–1988 at coolant outlet temperatures up to 990°C. Also because of a lack of other experience the AVR operation is a basis for future HTRs. This paper deals with insufficiently published unresolved safety problems of AVR and of pebble bed HTRs. The AVR primary circuit is heavily contaminated with dust bound and mobile metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory. A re-evaluation of the AVR contamination is performed in order to quantify consequences for future HTRs: The AVR contamination was mainly caused by inadmissible high core temperatures, and not — as presumed in the past — by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot be equipped with instruments. The maximum core temperatures were more than 200 K higher than precalculated. Further, azimuthal temperature differences at the active core margin were observed, as unpredictable hot gas currents with temperatures > 1100°C. Despite of remarkable effort these problems are not yet understood. Having the black box character of the AVR core in mind it remains uncertain whether convincing explanations can be found without major experimental R&D. After detection of the inadmissible core temperatures, the AVR hot gas temperatures were strongly reduced for safety reasons. Metallic fission products diffuse in fuel kernel, coatings and graphite and their break through takes place in long term normal operation, if fission product specific temperature limits are exceeded. This is an unresolved weak point of HTRs in contrast to other reactors and is particularly problematic in pebble bed systems with their large dust content. Another disadvantage, responsible for the pronounced AVR contamination, lies in the fact that activity released from fuel elements is distributed in HTRs all over the coolant circuit surfaces and on graphitic dust and accumulates there. Consequences of AVR experience on future reactors are discussed. As long as pebble bed intrinsic reasons for the high AVR temperatures cannot be excluded they have to be conservatively considered in operation and design basis accidents. For an HTR of 400 MWth, 900°C hot gas temperature, modern fuel and 32 fpy the contaminations are expected to approach at least the same order as in AVR end of life. This creates major problems in design basis accidents, for maintenance and dismantling. Application of German dose criteria on advanced pebble bed reactors leads to the conclusion that a pebble bed HTR needs a gas tight containment even if inadmissible high temperatures as observed in AVR are not considered. However, a gas tight containment does not diminish the consequences of the primary circuit contamination on maintenance and dismantling. Thus complementary measures are discussed. A reduction of demands on future reactors (hot gas temperatures, fuel burn-up) is one option; another one is an elaborate R&D program for solution of unresolved problems related to operation and design basis accidents. These problems are listed in the paper.

Analysis of Waste Arisings From the Chernobyl Exclusion Zone

2001 ◽  
Author(s):  
Mark Y. Gerchikov ◽  
L. Mark ◽  
C. Dutton ◽  
Elizabeth J. Kennett ◽  
Dmitry A. Bugai ◽  
...  
Keyword(s):  
Decision Making ◽  
Nuclear Power ◽  
International Study ◽  
Chernobyl Nuclear Power Plant ◽  
Decision Making Process ◽  
Industrial Zone ◽  
Exclusion Zone ◽  
Waste Streams ◽  
Management Options ◽  
Near Surface

Abstract The paper reviews the findings of a recent international study to characterise the waste arising from the decommissioning of dumps in the Industrial Zone of the Chernobyl Nuclear Power Plant and the Exclusion Zone. Studied sites included the Industrial Zone outside the Sarcophagus, three engineered disposal sites (the so-called PZRO), non-engineered near surface trench dumps (PVLRO), contaminated soil and sites of ‘unauthorised’ disposal within the Exclusion Zone. The paper summarises the inventory of wastes, the management options, which have been considered for various dumps, and the resulting estimates of the volumes of waste streams, as well as the approach that was used in the decision-making process.

Estimation of Clearance Potential Index and Hazard Factors of CANDU Fuel Bundle and Its Validation Based on the Measurements of Radioisotopes Inventories From Pickering Reactor Fuel

2007 ◽  
Author(s):  
Alexandru Octavian Pavelescu ◽  
Dan Gabriel Cepraga ◽  
Konstantina Voukelatou ◽  
Renato Tinti
Keyword(s):  
Cross Sections ◽  
Spent Nuclear Fuel ◽  
Safety Management ◽  
Fission Products ◽  
Radioactive Wastes ◽  
Radioactive Material ◽  
Complex Activity ◽  
Potential Index ◽  
Fuel Bundle ◽  
Biological Hazard

This paper is related to the clearance potential levels, ingestion and inhalation hazard factors of the spent nuclear fuel and radioactive wastes. This study required a complex activity that consisted of more steps such as: the acquisition, setting up, validation and application of procedures, codes and libraries. The paper reflects the validation stage of this study. Its objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from the Pickering CANDU reactor with the inventories predicted using a recent version of the SCALE 5\ORIGEN-ARP code coupled with the time dependent cross sections library for the CANDU 28 reactor (produced by the sequence SCALE4.4a\SAS2H and SCALE4.4a\ORIGEN-S). In this way, the procedures, the codes and the libraries for the characterization of radioactive material in terms of radioactive inventories, clearance, and biological hazard factors could be qualified and validated, in support of the safety management of the radioactive wastes.

UNCERTAINTY EVALUATION IN MEASUREMENT OF THE PERSONAL DOSE EQUIVALENT AT NINE INDIVIDUAL MONITORING SERVICES IN ASIA AND THE PACIFIC REGION

2020 ◽  
Vol 190 (2) ◽  
pp. 217-225
Author(s):  
Chadia Rizk ◽  
Panagiotis Askounis ◽  
H Burçin Okyar ◽  
John Konsoh Sangau ◽  
Samaneh Baradaran ◽  
...  
Keyword(s):  
Uncertainty Evaluation ◽  
Radiological Protection ◽  
International Electrotechnical Commission ◽  
Pacific Region ◽  
Dose Equivalent ◽  
Annual Dose ◽  
Uncertainty In Measurement ◽  
The Pacific ◽  
Dose Limits ◽  
Relevant Dose

Abstract This paper presents the results of the evaluation of the uncertainty in measurement of the personal dose equivalent, Hp(10), at nine individual monitoring services (IMSs) in Asia and the Pacific region. Different types of passive dosemeters were type-tested according to the International Electrotechnical Commission 62387 requirements. The uncertainty in measurement was calculated using the Guide to the Expression of Uncertainty in Measurement approach. Expanded uncertainties ranged between 24 and 86% (average = 38%) for Hp(10) values around 1 mSv and between 14 and 40% (average = 27%) for doses around the annual dose limit, Hp(10) = 20 mSv. The expanded uncertainties were lower than the 1.5 factor in either direction proposed by the International Commission on Radiological Protection for doses near the relevant dose limits. This indicates an acceptable level of uncertainty for all participating IMSs. Uncertainty evaluation will help the IMSs to acknowledge the accuracy of their measurements.

Synthesis of rare earth phosphates in molten LiCl–KCl eutectic: Application to preliminary treatment of chlorinated waste streams containing fission products

2008 ◽  
Vol 381 (3) ◽  
pp. 284-289 ◽  
Author(s):  
Damien Hudry ◽  
Isabelle Bardez ◽  
Aydar Rakhmatullin ◽  
Catherine Bessada ◽  
Florence Bart ◽  
...  
Keyword(s):  
Rare Earth ◽  
Fission Products ◽  
Preliminary Treatment ◽  
Waste Streams

Computational Study for Inventory Estimation of Se-79, Tc-99, Sn-126, and Cs-135 in High-Level Radioactive Wastes From Spent Nuclear Fuels of Light Water Reactors

2011 ◽  
Author(s):  
Keisuke Okumura ◽  
Shiho Asai ◽  
Yukiko Hanzawa ◽  
Tsutomu Okamoto ◽  
Hideya Suzuki ◽  
...  
Keyword(s):  
Parametric Study ◽  
Computational Study ◽  
Estimation Method ◽  
Fission Products ◽  
Radioactive Wastes ◽  
Light Water Reactors ◽  
Nuclear Fuels ◽  
Light Water ◽  
Water Reactors ◽  
High Level

Inventory estimation of long-lived fission products (LLFPs) in high-level radioactive wastes (HLW) from spent nuclear fuels of light water reactors is important for a safety assessment of their disposal. In order to develop an inventory estimation method of difficult-to-measure LLFPs (Se-79, Tc-99, Sn-126, and Cs-135), a parametric study was carried out by using a sophisticated burnup calculation code and data. In the parametric study, fuel specifications and irradiation conditions are changed in the conceivable range. The considered parameters are fuel assembly types (PWR / BWR), U-235 enrichment, moderator temperature, void fraction, power density, and so on. From the calculated results, we clarify the burnup characteristics of the target LLFPs and their possible ranges of generations. Finally, candidates of the key nuclide are proposed for the scaling factor method of HLW.

Radiation biology and radiation protection

2012 ◽  
Vol 41 (3-4) ◽  
pp. 64-71 ◽  
Author(s):  
J.H. Hendry
Keyword(s):  
Biological Effects ◽  
Biological Response ◽  
Threshold Dose ◽  
Radiological Protection ◽  
Turnover Rates ◽  
Injury Threshold ◽  
Dose Limits ◽  
Threshold Doses ◽  
Tissue Reactions ◽  
Hereditary Effects

For protection purposes, the biological effects of radiation are separated into stochastic effects (cancer, hereditary effects) presumed to be unicellular in origin, and tissue reactions due to injury in populations of cells. The latter are deterministic effects, renamed ‘tissue reactions’ in the 2007 Recommendations of the International Commission on Radiological Protection because of the increasing evidence of the ability to modify responses after irradiation. Tissue reactions become manifest either early or late after doses above a threshold dose, which is the basis for recommended dose limits for avoiding such effects. Latency time before manifestation is related to cell turnover rates, and tissue proliferative and structural organisation. Threshold doses have been defined for practical purposes at 1% incidence of an effect. In general, threshold doses are lower for longer follow-up times because of the slow progression of injury before manifestation. Radiosensitive individuals in the population may contribute to low threshold doses, and in the future, threshold doses may be increased by the use of various biological response modifiers post irradiation for reducing injury. Threshold doses would be expected to be higher for fractionated or protracted doses, unless doses below the threshold dose only cause single-hit-type events that are not modified by repair/recovery phenomena, or if different mechanisms of injury are involved at low and high doses.

scholarly journals Study on Occupational Radiation Exposure at Different Hospitals in Nepal using ThermoluminescenceDosimetry

2020 ◽  
Vol 19 (1) ◽  
pp. 180-185
Author(s):  
Prabhash Acharya ◽  
Gita Chalise ◽  
Bipin Rijal ◽  
Hari Prasad Lamichhane ◽  
Buddha Ram Shah
Keyword(s):  
Radiation Exposure ◽  
Radiological Protection ◽  
Occupational Radiation Exposure ◽  
Dose Rates ◽  
Routine Work ◽  
Occupationally Exposed ◽  
Dose Limits ◽  
Radiation Workers ◽  
Set Up ◽  
Occupational Radiation

 The clinical efficacy of using ionizing radiation in diagnosis and treatment of diseases has been revolu­tionized, benefitting humankind and, at the same time, imposing deleterious health effects, if not han­dled carefully. Personnel dosimetry has emerged as an essential tool to monitor occupational radiation exposure. The present study intends to reveal and describe the scenario of occupationally exposed staff by assessing an individual’s dose in radiological facilities at different hospitals in Nepal. Thermo luminescent Dosimetry (TLD) has been used for assessing individual doses. Altogether eight hospitals were chosen from different locations of Nepal. TLD badges were worn by the radiation workers on their chest level while working with the radiological equipment for about three to ten months during their routine work. Later the badges were read out in the TLD badge reader system available at Nepal Academy of Science and Technology (NAST) to obtain the exposed doses. CaSO4: Dy was used as TL phosphor. Exposed doses for health personnel during their routine work were founding the range of (2.21 to 16.17) 0.01 mSv per year, which was low compared to the permissible dose limit of 20 mSv per year set up by International Commission on Radiological Protection(ICRP). Although the exposed dose rates in the monitored hospitals are below the recommended dose limits, it should be made as low as possible based on the ALARA principle.

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