Radioactive Liquid Effluents Research Articles

Determination of Decay Tank Capacity and Minimum Storage Time for Liquid Radioactive Effluent in Nuclear Medicine Departments Based on Environmental Requirements

Determination of Decay Tank Capacity and Minimum Storage Time for Liquid Radioactive Effluent in Nuclear Medicine Departments Based on Environmental Requirements

Strength evaluation of support frame of the low level liquid effluent pipes of nuclear plant

Disposal of low level radioactive liquid effluents of nuclear plant into the sea is carried out through carbon steel pipes inside a guard pipe. The treated effluents are discharged inside the sea beyond breaker zone through two disposal pipes. The disposal pipe discharges the treated effluent to two numbers of 1 m diameter pipe which goes down till sea bed. This guard pipe should be supported on top of the jetty against wind, wave, and dead loads. This study is intended to cover the design adequacy and material selection of the liquid discharge piping support configuration. Material and coating requirements of the support frame is chosen after careful evaluation of the prevailing saline environment and economic consideration. Elastomeric bearing provides necessary support & firm adherence of the support arrangement to the erected pile cap. Elastomeric bearing are easy to install and offer low maintenance. The support frame is checked for dead load, pipe load, live load, wind load, wave ¤t loads, and seismic loads. Wave and current loads were calculated based on Morison semi empirical equation. Linear elastic structural analysis of the support frame was carried out. This article provides insight to the materials used in supporting this pipelines and the basis behind the material selection. After material selection, strength evaluation of the structure is elaborately carried out for all the loads. Indian standards and API recommended practice were adopted for the strength assessment of the structure. Support arrangement is formed using trusses at different levels, spandrel girders and hangers connecting the spandrel girders. All the stresses for structural arrangement for liquid discharge pipe support are checked and found within acceptable limits.

Estimation of aquatic dilution factor for liquid effluent discharges from Kaiga Generating Station using tritium as a tracer

Very low-level radioactive liquid effluents much below the approved discharge limits are discharged to Kadra reservoir after dilution with condenser cooling water at Kaiga Generating Station. In addition, further dilution occurs from the discharge point to the nearest public utilization point in Kadra reservoir (Hartuga). The dilution obtained in Kadra reservoir is estimated by simultaneous measurements of tritium at the discharge point and at Hartuga. The on-site measurement indicates a dilution factor of 12.2 ± 2.4 for tritium. A dilution factor of 13.3 ± 3.4 was obtained using aquatic dispersion model.

Dilution or containment: controlling radioactive liquid effluents from the 1950s to the 1980s on Marcoule plant, France

Nuclear facilities constantly produce radioactive liquid effluents, as a part of their ordinary operation. However, controlling a circulating fluid is problematic. This paper focuses on Marcoule plant in Southern France, on the Rhône River, to document how radioactive effluent containment evolved from the 1950s to the early 1980s. The 1950s and 1960s were a time of experimentation: engineers intended to find a way to dilute radioactive wastes into the river, following a traditional conception inherited from the 19th century. However, due to multiple factors (scientific research and a better understanding of radiation effects, raising protests against nuclear wastes, first legislation on radioactive effluents), both guiding concepts and technical solutions evolved to a stricter containment and control of radioactive effluents.

Revision of discharge limit of gross beta activity to the aquatic environment based on public dose estimation

Low-level liquid effluents generated from Regional Centre, Board of Radiation, and Isotope Technology (BRIT), Delhi, are generally discharged to the aquatic environment after suitable treatment (delay, decay, and dilution for short-lived radionuclides) conforming the regulatory compliance as authorized by Atomic Energy Regulatory Board, Mumbai. The Regional Center, BRIT-Delhi is generating the liquid effluents containing short-lived radionuclides 99Mo (T1/2−66.7 h) and 99mTc (T1/2−6.02 h) and also long-lived radionuclide 99Tc (T1/2−2.13 × 105 years) from the production of ready to use 99mTc-radiopharmaceuticals as human injectable product. This liquid waste generated during the washing of radioactive contaminated glassware is inorganic in nature and is stored in 02 number of sump tanks (capacity 50 m3 each) for delay and decay of short-lived radionuclides. The center has also planned to produce and supply 68Ga radiopharmaceuticals in future. There is a possibility that some of these radionuclides may reach the drinking water by various natural pathways. The presence of radionuclides in the drinking water above certain level may result in radiation dose to the public through the ingestion pathways. The observance of prescribed radionuclide concentration in waste water, total activity limits, and other basic safety requirements stipulated by the regulatory body help to minimize the public radiation dose. This article is an effort to derive the annual discharge limit for gross β activity at Regional Center, BRIT-Delhi to the aquatic environment of an inland site based on the drinking water standard limits prescribed by the World Health Organization. This article also discusses the effective dose received by the actual discharge of radioactive liquid effluent from the Regional Center, BRIT-Delhi.

Additional radiation dose due to atmospheric dispersion of tritium evaporated from a hypothetical reservoir

With regard to an inland nuclear power plant bordered by a reservoir, a major concern was that fresh water might be polluted and the human body might be radiation exposed due to the discharge of liquid radioactive effluents. In contrast to other radionuclides in the effluents, tritium has specific dispersion behavior in the aquatic environment such as emission into the air along with water evaporation. Further, the evaporated tritium in the air could go toward the territorial system where the wind blows. As a result, the person staying in the vicinity of the plant discharge point would be exposed with an additional radiation dose. In light of this characteristic, this study first introduced this new exposure pathway and investigated the additional radiation dose on the basis of a hypothetical reservoir. The results indicated that annual tritium evaporation fraction is approximately 2.5%, which is a comparable level with the radioactive decay factor. This would produce an additional radiation dose of 0.63 μSv/a to a person staying 50 m away from the plant discharge point for the case of 1 g/a tritium discharge. Tritium evaporation effects could be decreased through controlling the discharge depth. Thus, a preliminary suggestion to adopt a deep discharge instead of surface discharge was proposed from the ALARA (as low as reasonably achievable) criterion of radiation protection.

Numerical study on tritium dispersion in coastal waters: The case of Hangzhou Bay, China

Identification of tritium dispersion characteristics is critically important to recognize the environmental impact due to discharge of radioactive liquid effluents from nuclear industry. For the first time, this study simulated the dispersion behavior in the coastal waters of tritium discharged from Qinshan nuclear power base, where the sole heavy water reactor is located and the largest tritium amount was discharged in China. Both the flow field and tritium concentration distribution in Hangzhou Bay were displayed under normal and accidental discharge conditions considering the realistic water dynamic characteristics. Effects of river and tidal flow on tritium dispersion, effects of tritium evaporation factor, as well as public radiation dose were specially discussed. The results show that tidal flow is the major factor in deciding tritium dispersion for the Hangzhou Bay case. River flow could also affect tritium dispersion, particularly in the southwest region of the discharge point. Under normal discharge condition, the maximum tritium concentration in the vicinity of Qinshan nuclear power base is at 104 Bq/m3 level, which is an order of magnitude higher than the natural background tritium. Under accidental discharge condition, maximum tritium concentration decreases along with an increasing tritium dispersion time. Results furthermore indicate that a considerable tritium evaporation fraction appears for the Hangzhou Bay case. In spite of this, radiation dose involving ingestion and inhalation pathways associated with liquid tritium discharge is still much lower than the regulatory dose limit. This study is expected to recall attentions to tritium dispersion characteristics apart from providing fundamental simulation data for tritium distribution in the waters of Hangzhou Bay.

Pyrogenic carbon for decontamination of low-level radioactive effluents: Simultaneous separation of 137Cs and 60Co

Radiocesium (137Cs) and radiocobalt (60Co) represent main dose-significant radionuclides in low-level radioactive liquid effluents released from commercial nuclear power plants (NPPs). Although pyrogenic carbon-based adsorbents (biochars, BC) have been studied for toxic metal removal, the reports of their use in the adsorption of radionuclides are rare. This article focuses on the competitive removal of Cs+ and Co2+ ions from radioactive solutions to waste biomass derived biochar and the multi-component isotherms. BC exhibited a good affinity to both Cs and Co, and in single-component systems the maximal adsorption capacity Qmaxexp reached 97.4 ± 5.0 (Cs+) and 74.5 ± 0.4 μmol g−1 (Co2+). 3D adsorption surfaces obtained by Sheindorf-Rebuhn-Steintuch (SRS) and extended Freundlich models indicate that the sorption of Cs+ ions from the Cs–Co binary system is significantly affected by Co2+ ions and during the adsorption competitive and interaction effects between Cs+ and Co2+ ions occur. Moreover, mono- and divalent cations present in the solution acted as competing ions, decreasing the Kd values. SEM-EDX maps, FTIR and XPS spectra confirmed that both biochar mineral and carbon fraction participated in Cs and Co immobilization. Beside interactions of Cs+ and Co2+ ions with O-containing moieties on the biochar surface (chemisorption), mineral components and porosity of BC (physisorption) were involved in both Cs and Co removal. Finally, the adsorption loading and fast kinetics for both Cs and Co indicate suitability of biochar to decontaminate low-level radioactive effluents contaminated by 137Cs and 60Co.

Treatment of spent ion exchange resins: Comparison between hydrothermal and acid leaching processes

Ion exchange resins (IERs) are used to treat radioactive liquid effluents from nuclear industry. Today, the management of spent IERs as nuclear waste is an issue. We propose here the use of hydrothermal and leaching treatments to transfer radioactivity from organic solid phase to aqueous phase, in order to connect to existing outlet systems. Co, Cs, Sr and Eu were chosen as α and βγ radionuclide surrogates. The efficiency of the treatment was assessed in terms of extraction yields and contaminant concentrations in wastewaters. The highest extraction efficiencies of 90–100 % were obtained in supercritical water in batch and semi-dynamic mode in the presence of oxidant. In subcritical conditions, yields adequate are lower but suitable with the disposal criteria for the residual solid waste. Complete extraction was also achieved by leaching with ethylenediaminetetraacetic acid solutions. The optimal choice between all these processes is therefore discussed with regards to the requirements of effluent treatment plant including post-treatment steps such as coprecipitation and advanced oxidation.

A priori risk-informed approach to ensure the compliance with discharge limits for liquid radioactive effluent to be discharged from APR 1400

ABSTRACT In order to develop a methodology to guarantee the conformance to operational discharge limits for liquid effluent from a NPP at an early stage of design, a risk-informed approach was proposed and its applicability was verified for APR 1400. Existing methodology to calculate risk-based detection limit for a single radionuclide was improved by incorporating a new model to derive more realistic pathway dose factors. A new simple expression was also proposed to adjust risk-based detection limits for multiple radionuclides mixture if necessary. In addition, a new procedure to warrant the compliance with discharge limits by controlling detection limits of only a few principal radionuclides was established in accordance with risk-informed concept. Through case studies for APR 1400 to be commissioned at a hypothetical site, it was shown that calculated pathway dose factors are more realistic for majority of radionuclides. It also turns out that neither present detection limits nor unadjusted risk-based detection limits can be justified when the radionuclide composition is unknown, however further adjustment of detection limits or increasing additional dilution factor resolves the problem. Finally, ten principal radionuclides were identified and shown to be enough for liquid effluent control at APR 1400 from a risk-informed point of view.

Methodology for determination of alarm and warning set-points for radioactive effluent monitors in Korean pressurized water reactors

All radioactive gaseous and liquid effluents discharged from Korean nuclear power plants are monitored by effluent monitors to prevent effluent releases to the environment under uncontrolled conditions. This paper provides the methodology and parameters used in the calculation of alarm (high) and warning (low) set-points for gaseous and liquid effluent monitors in Korean pressurized water reactors. Alarm set-points are determined to assure compliance with the Korean regulatory limits of concentration of radioactive effluents. Even though warning set-points are not required by the regulatory body, Korean pressurized water reactors determine the warning set-points of effluent monitors not only to take an active management of effluent discharge but also to keep radiation doses to members of the public living around nuclear power plants as low as reasonably achievable.

Superposition of Potential Chemical Polluants and Radioisotopes and Their Influence Upon the Environment and Living Organisms

Following the activities developed in a nuclear facility result gaseous and liquid radioactive effluents and radioactive solid waste. All these waste contain radioactive isotopes which are potentially pollutants for the environment. In the same time chemicals are, also, pollutants. According to the legislation, discharging of chemicals and radioactive liquid and gaseous effluents into the environment, should meet the requirements of the unrestricted discharge. However, what happens when several pollutants superpose: only chemical pollutants, or only radioactive pollutants, or chemical and radioactive pollutants? Such problems have been solved in this paper on the basis of the principle of critical energy.

Radioactive effluents released from Korean nuclear power plants and the resulting radiation doses to members of the public

Abstract Korean nuclear power plants (NPPs) periodically evaluate the radioactive gaseous and liquid effluents released from power reactors to protect the public from radiation exposure. This paper provides a comprehensive overview of the release of radioactive effluents from Korean NPPs and the effects on the annual radiation doses to the public. The amounts of radioactive effluents released to the environment and the resulting radiation doses to members of the public living around NPPs were analyzed for the years 2011–2015 using the Korea Hydro & Nuclear Power Co., Ltd's annual summary reports of the assessment of radiological impact on the environment. The results show that tritium was the primary contributor to the activity in both gaseous and liquid effluents. The averages of effective doses to the public were approximately on the order of 10 −3 mSv or 10 −2 mSv. Therefore, even though Korean NPPs discharged some radioactive materials into the environment, all effluents were within the regulatory safety limits and the resulting doses were much less than the dose limits.

Treatment of radioactive liquid effluents by reverse osmosis membranes: From lab-scale to pilot-scale

The recent use of the reverse osmosis (RO) process at the damaged Fukushima-Daiichi nuclear power plant generated a growing interest in the application of this process for decontamination purposes. This study focused on the development of a robust RO process for decontamination of two kinds of liquid effluents: a contaminated groundwater after a nuclear disaster and a contaminated seawater during a nuclear accident. The SW30 HR membrane was selected among other in this study due to higher retentions (96% for Cs and 98% for Sr) in a true groundwater. Significant fouling and scaling phenomenon, attributed to calcium and strontium precipitation, were evidenced in this work: this underscored the importance of the lab scale experiment in the process. Validation of the separation performances on trace radionuclides concentration was performed with similar retention around 96% between surrogates Cs (inactive) and 137Cs (radioactive). The scale up to a 2.6 m2 spiral wound membrane led to equivalent retentions (around 96% for Cs and 99% for Sr) but lower flux values: this underlined that the hydrodynamic parameters (flowrate/cross-flow velocity) should be optimized. This methodology was also applied on the reconstituted seawater effluent: retentions were slightly lower than for the groundwater and the same hydrodynamic effects were observed on the pilot scale. Then, ageing of the membrane through irradiation experiments were performed. Results showed that the membrane active layer composition influenced the membrane resistance towards γ irradiation: the SW30 HR membrane performances (retention and permeability) were better than the Osmonics SE at 1 MGy. Finally, to supplement the scale up approach, the irradiation of a spiral wound membrane revealed a limited effect on the permeability and retention. This indicated that irradiation conditions need to be controlled for a further development of the process.

Accumulation of Sellafield-derived radiocarbon (14C) in Irish Sea and West of Scotland intertidal shells and sediments

The nuclear energy industry produces radioactive waste at various stages of the fuel cycle. In the United Kingdom, spent fuel is reprocessed at the Sellafield facility in Cumbria on the North West coast of England. Waste generated at the site comprises a wide range of radionuclides including radiocarbon (14C) which is disposed of in various forms including highly soluble inorganic carbon within the low level liquid radioactive effluent, via pipelines into the Irish Sea. This 14C is rapidly incorporated into the dissolved inorganic carbon (DIC) reservoir and marine calcifying organisms, e.g. molluscs, readily utilise DIC for shell formation. This study investigated a number of sites located in Irish Sea and West of Scotland intertidal zones. Results indicate 14C enrichment above ambient background levels in shell material at least as far as Port Appin, 265 km north of Sellafield. Of the commonly found species (blue mussel (Mytilus edulis), common cockle (Cerastoderma edule) and common periwinkle (Littorina littorea)), mussels were found to be the most highly enriched in 14C due to the surface environment they inhabit and their feeding behaviour. Whole mussel shell activities appear to have been decreasing in response to reduced discharge activities since the early 2000s but in contrast, there is evidence of continuing enrichment of the carbonate sediment component due to in-situ shell erosion, as well as indications of particle transport of fine 14C-enriched material close to Sellafield.

Current research and development activities on fission products and hydrogen risk after the accident at Fukushima Daiichi Nuclear Power Station

After the Fukushima Daiichi nuclear power plant (NPP) accident, new regulatory requirements were enforced in July 2013 and a backfit was required for all existing nuclear power plants. It is required to take measures to prevent severe accidents and mitigate their radiological consequences. The Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R) has been conducting numerical studies and experimental studies on relevant severe accident phenomena and countermeasures. This article highlights fission product (FP) release and hydrogen risk as two major areas. Relevant activities in the S/NRA/R are briefly introduced, as follows: 1. For FP release: Identifying the source terms and leak mechanisms is a key issue from the viewpoint of understanding the progression of accident phenomena and planning effective countermeasures that take into account vulnerabilities of containment under severe accident conditions. To resolve these issues, the activities focus on wet well venting, pool scrubbing, iodine chemistry (in-vessel and ex-vessel), containment failure mode, and treatment of radioactive liquid effluent. 2. For hydrogen risk: because of three incidents of hydrogen explosion in reactor buildings, a comprehensive reinforcement of the hydrogen risk management has been a high priority topic. Therefore, the activities in evaluation methods focus on hydrogen generation, hydrogen distribution, and hydrogen combustion.

대덕부지 원자력관련시설 운영에 따른 주민피폭선량 현황분석

This paper describes the results of assessment of radiological dose resulting from operation of the Daedeok nuclear facilities including the HANARO research reactor, which has been performed to assure whether or not to comply with the regulation standards of the radioactive effluents releases. Based on the meteorological data and the radiation source term, the maximum individual doses were evaluated from 2010 to 2012. The atmospheric dispersion and the deposition factors of gaseous effluents were calculated using the XOQDOQ computer code. ENDOS-G and ENDOS-L code systems were also used for maximum individual dose calculation from gaseous and liquid effluents, respectively. The results were compared with the regulation standards for the radioactive effluents presented by the Nuclear Safety and Security Commission (NSSC). The effective doses and the thyroid doses of the maximum individual were calculated at the maximum exposed point in the Daedeok site, and contributions of exposure pathways to the radiological doses resulting from gaseous and liquid radioactive effluents were evaluated at each facility of the Daedeok site. As a result, the maximum exposed age was analysed to be the child group, and the operation of HANARO research reactor had a major effect more than 90% on the individual doses. The main exposure pathways for gaseous radioactive effluent were from ingestion and inhalation. The effective doses and the thyroid doses were considerably influenced by tritium and iodine, respectively. The gaseous radioactive effluents contributed more than 90% on the total doses, whereas the contributions of the liquid radioactive effluents were relatively low. Consequently, the maximum individual dose due to radioactive effluents from the nuclear facilities within the Daedeok site were less than 3% of the regulation standard over 3 years; therefore, it can be concluded that radioactive effluents from the nuclear facilities were well managed, with the radiation-induced health detriment for residents around the site being negligible.

방사선안전관리 실무: (II) 배기중 및 배수중 배출관리기준의 적용

방사선안전관리 실무: (II) 배기중 및 배수중 배출관리기준의 적용

Inverse Modeling for Aquatic Source and Transport Parameters Identification and its Application to Fukushima Nuclear Accident

Inverse modeling technique based on nonlinear least square regression method (LSRM) is developed for the identification of aquatic source and transport parameters. Instantaneous line source release model in two-dimensional domain and continuous point source release model in three-dimensional domain are used for the purpose. Case studies have been carried out for both types of releases to illustrate their application. Error analysis has been carried out to identify the maximum error that can be tolerated in the input concentration data used in the inverse model and to specify the minimum number of sampling points to generate such input data. The LSRM is compared with the well-established correlation coefficient optimization method for instantaneous line source release model, and good comparison is observed between them. The LSRM is used to quantitatively estimate the releases of different radionuclides into the Pacific Ocean which has resulted due to the discharge of highly radioactive liquid effluent from the affected Daiichi Nuclear Power Station at Fukushima in Japan. The measured concentrations of these radionuclides in seawater samples collected from two sampling points near Fukushima are used for the estimation. The average release works out to be 1.09 × 1016 for 131I, 3.4 × 1015 Bq for 134Cs, and 3.57 × 1015 Bq for 137Cs. Very good agreement is observed between the releases estimated in this study and those estimated by other different agencies.

Estimation of radioactive leakages into the Pacific Ocean due to Fukushima nuclear accident

High concentrations of several radionuclides were reported in the sea near the Fukushima Daiichi Nuclear Power Station (FDNPS) in Japan due to the nuclear accident that occurred on 11 March 2011. The main source of these concentrations was leakage of highly radioactive liquid effluent from a pit in the turbine building near the intake canal of Unit-2 of FDNPS through a crack in the concrete wall. In the immediate vicinity of the plant, seawater concentrations reached 68 MBq m−3 for 134Cs and 137Cs, and exceeded 100 MBq m−3 for 131I in early April 2011. These concentrations began to fall as of 11 April 2011 and, at the end of April, had reached a value close to 0.1 MBq m−3 for 137Cs. Following the nuclear accident, the Tokyo Electric Power Company (TEPCO) had initiated intense monitoring of the environment including the Pacific Ocean. Seawater samples were collected and the concentrations of few radionuclides were measured on a wide spatial and temporal scale. In this study, the measured concentrations of different radionuclides near the south discharge canal of the FDNPS were used to estimate their leakages into the Pacific Ocean. The method is based on estimating the release rates that reproduce the concentration of radionuclides in seawater at a chosen location using a two-dimensional advection–dispersion model in an iterative manner. The radioactive leakages were estimated as 5.68 PBq for 131I, 2.24 PBq for 134Cs and 2.25 PBq for 137Cs. Leakages were also estimated for 99mTc, 136Cs, 140Ba and 140La and they range between 0.02 PBq (99mTc) and 0.53 PBq (140Ba). It was estimated that about 11.28 PBq of radioactivity in total was leaked into the Pacific Ocean from the damaged FDNPS. Out of this, 131I constitutes 50.3 %; 134Cs 20 %; 137Cs 20 %; 140Ba 4.6 %; 136Cs 2.6 %; 140La 2.3 % and 99mTc 0.2 % of the total radioactive leakage. Such quantitative estimates of radioactive leakages are essential prerequisites for short-term and local-scale as well as long-term and large-scale radiological impact assessment of the nuclear accident.