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Is Rule 28 of the implementing provisions to the Regulation on the EQE (REE) already in force?Yes, amended Rule 28 entered into force on 13 February 2017. Candidates may request any previous period of professional activity as defined in Article 11(2) REE to be taken into account without limitation.
Registration is free of charge until further decision of the Supervisory Board.For candidates successfully enrolled for the European qualifying examination, registration is not necessary since their periods of professional activity are already known. General informationWhat should I take into account when choosing an examination centre?The European Qualifying Examination is held in examination centres at venues appointed by the EPO in Munich, The Hague and Berlin. At these centres, set standards such as room temperature between 19-21ºC, a reasonable table size, low noise level, good accessibility are adhered to. Other EQE venues (examination centres) are made available by the national central industrial property offices of member states to the European Patent Convention. It is recommended to these offices to follow the EPO standards.
However, it is ultimately their decision at which centre the examination takes place.Before choosing an examination centre, candidates are advised to read the comments made in the for the different centres concerned.During the enrolment procedure for the EQE, candidates may always opt to sit the examination at one of the EPO centres, subject to availability. Whilst the EPO makes every effort to ensure that prescribed standard conditions are upheld, it cannot guarantee that this will be the case in other examination centres. Can the required supporting evidence be certified by a professional representative, and if so, what form should the certification take?In accordance with Rule 1(3) IPREE it is possible to have supporting evidence certified by a professional representative (Art. This is usually the quickest and easiest way to have documents certified. The certification should be written on a copy of the document and must include:. a statement that the certified copy corresponds with the original. the date and place of certification.
the signature of the professional representative and his/her name in block capitals. How can I pay EQE-related fees?Fees relating to the EQE may only be paid by credit card or bank transfer. It is not possible to pay the fees by debiting an EPO deposit account.Candidates are strongly recommended to use a credit card as their method of payment as this will ensure swift payment.
Payments by bank transfer are considered to have been made on the date on which the amount of the transfer is actually entered in the bank account held by the EPO (see Article 7 of the Rules relating to Fees).Candidates are reminded that a bank transfer may take time and it is therefore not recommended as a payment method shortly before the relevant deadline. Late receipt of payment will lead to the application being rejected.
Will a printed version of the the exam paper also be available?Yes, you will receive the examination paper in the three official languages in paper form like any other candidate not using a computer. As some candidates like to take hand written notes during the examination, you will also receive some scrap paper (blank sheets of A4 paper). You can use the paper copy of the examination paper and the scrap paper to highlight and make notes. However, your answer paper will consist only of the computer- written electronic answer. Handwritten notes or parts of the printed examination paper will not form part of your answer.
.Beg, Z.M.; Ghosh, R.S.1987-01-01The vacuum building (VB) and pressure relief structures (PRS) are the unique features of multiple unit CANDU containments. In case of loss-of-coolant accident, the released radionuclides are drawn through the PRS into the subatmospheric VB, doused and contained without being released to the environment.
This paper describes the differences in design, configuration and layout of the VB and PRS from Pickering NGS A to Darlington NGS A due to new developments in design concepts and to requirements which have proceeded from the experience gained in both the design and operation of the nuclear stations. (orig.).1995-05-01This is a report on the completion of work relating to the assessment of the capability of Darlington NGS to cope with a large fire incident. This included an evaluation of an exercise scenario that would simulate a large fire incident and of their fire plans and procedures which became the subject of interim reports as part of the process of preparing for the fire fighting and rescue exercise. Finally the execution of fire plans by Darlington Nuclear Generating Station ( NGS), as demonstrated by their application of human and material resources during a simulated large fire, was observed.
1 tab., 1 fig.Morrison, J.F.1992-01-01This paper details the Motor Operated Valve (MOV) Predictive Maintenance program at Darlington Nuclear Generating Station. The program encompasses the use of diagnostics tooling in conjunction with more standard maintenance techniques, with the goal of improving performance of MOV's.
Problems encountered and solutions developed during the first two phases of this program are presented, along with proposed actions for the final trending phase of the program. This paper also touches on the preventive and corrective maintenance aspects of an overall MOV maintenance program. 6 refs., 6 tabs., 6 figs.Leung, V.; Crouse, B.1996-01-01The ability to improve the capabilities and reliability of digital control systems in nuclear power stations to meet changing plant and personnel requirements is a formidable challenge. Many of these systems have high quality assurance standards that must be met to ensure adequate nuclear safety.
Also many of these systems contain obsolete hardware along with software that is not easily transported to newer technology computer equipment. Combining modern technology upgrades into a system of obsolete hardware components is not an easy task.
Lastly, as users become more accustomed to using modern technology computer systems in other areas of the station (e.g. Information systems), their expectations of the capabilities of the plant systems increase. This paper will present three areas of the Darlington NGS fuel handling computer system that have been or are in the process of being upgraded to current technology components within the framework of an existing fuel handling control system. 3 figs.Leung, V; Crouse, B Ontario Hydro, Bowmanville (Canada). Darlington Nuclear Generating Station1997-12-31The ability to improve the capabilities and reliability of digital control systems in nuclear power stations to meet changing plant and personnel requirements is a formidable challenge. Many of these systems have high quality assurance standards that must be met to ensure adequate nuclear safety. Also many of these systems contain obsolete hardware along with software that is not easily transported to newer technology computer equipment.
Combining modern technology upgrades into a system of obsolete hardware components is not an easy task. Lastly, as users become more accustomed to using modern technology computer systems in other areas of the station (e.g.
Information systems), their expectations of the capabilities of the plant systems increase. This paper will present three areas of the Darlington NGS fuel handling computer system that have been or are in the process of being upgraded to current technology components within the framework of an existing fuel handling control system. 3 figs.Huterer, J.; Ha, E.C.; Brown, D.G.; Cheng, P.C.1985-01-01The paper describes the consequences of new design requirements for the Darlington vacuum building on its structural configuration, analytical and reinforcing steel layout.
Attention focuses on the ring girder where the juncture of dome and perimeter wall produces a complex post-tensioning layout, and attendant difficulties in design and construction. At the wall base, full fixity imposes large local stresses. Long-term, shrinkage and creep, and temperature effects become significant. A research program and in-house analytical procedure established time-dependent concrete behaviour and corresponding wall-sectional stresses.
The outcome is examined in terms of reinforcement, temperature controls, and wall liner requirements. (orig.).Paetzold, H.; Hera, V.; Schaumburg, G.1996-01-01Following incidents at Pickering, Wolsung and Bruce NGS, involving instability of bleed condenser relief valves, Darlington station decided to replace the spring loaded RV's by new pilot operated SEBIM tandem valves.
This paper is presenting the approach taken, the design and the testing of the new solution, as well as some of the computer modeling work performed in connection with this project. The SEBIM tandems, following successful testing in France, will be installed in Darlington Unit 2, this spring. The new valves can perform with absence of instability and prevent a LOCA incident due to their design, which includes a protection and a redundant valve in series. (author).Judah, J.; Goodchild, S.2013-01-01Fuel performance at the Darlington nuclear generating station has historically been excellent. Until recently, the majority of these few fuel defects have been attributed to fretting by heat transport system debris. The minority have been linked to manufacturing issues. Recently, Darlington has experienced an increase in the number of fuel defects.
Although the defect rate remains low with respect to industry standards, this defect experience is considered to be unacceptable given current industry expectations and the OPG zero defect policy. Nine fuel defects have been discharged since 2007 from the four Darlington reactors. This represents a fuel defect rate of just 0.35 defects per year per reactor. At the time of this writing three additional defects are suspected to be in core. Although a definitive defect cause has yet to be identified, these fuel performance issues appear to be due to the coincidental degradation of manufacturing and operational factors, thereby decreasing the margins to fuel failure due to fuelling power ramps. All of the confirmed defected bundles have been long bundles and all experienced a relatively high power ramp when shifted from Position 2 to Position 6. High bundle uranium masses and low internal clearances are thought to be significant contributing factors.
Bundle burnups at the time of the power ramps were low and these bundles were not identified by existing power ramp defect predictive tools. Our assessment has resulted in a number of recommendations which are designed to mitigate these adverse conditions by restoring the margins to power ramp failures. These recommendations impact broadly across a number of organizations including reactor physics, fuel design, fuel manufacturing, reactor design, inspections and PIE. (author).Liauw, W.K.; Liu, W.S.1997-01-01The main purpose of reactor system code development is to support plant operation and to assist in safety analysis.
To illustrate the operational support activities of the TUF code, the assessment of a simple case of abnormal load rejection event at Darlington NGS is described. The main assessment of this event examines the flow conditions at the steam generators and the possible impact on the turbine. This assessment demonstrates the TUF capability in the operational support analysis for CANDU reactors. (author).1989-03-01A CATHENA input model has been developed and documented for the heat transport system of the Darlington Nuclear Generating Station.
CATHENA, an advanced two-fluid thermalhydraulic computer code, has been designed for analysis of postulated loss-of-coolant accidents (LOCA) and upset conditions in the CANDU system. This report describes the Darlington input model (or idealization), and gives representative results for a simulation of a small break at an inlet header.Crane, R.H.1992-01-01The Darlington Nuclear Generating Station is a new station, consisting of four 935 Mw units, built by Ontario Hydro, on the north shore of Lake Ontario, approximately 50 miles east of Toronto. In May, 1987, the first of the four units of this station was approaching the point where Ontario Hydro would be requesting a license to load fuel, and then proceed to first criticality. At this point, however, the regulatory authority, the Atomic Energy Control Board (AECB) started to show increasing concerns related to the Trip Computer Software associated with Darlington's newly-designed computerized shutdown systems. The concerns centered around whether or not the safety reliability, reviewability, and maintainability of this software could be demonstrated by Ontario Hydro or the system designer, Atomic Energy of Canada Limited (AECL).
In order to back up the validity of their concerns, they hired a well-known consultant, who reviewed the code, and made recommendations concerning its design, implementation, and documentation. Considerable effort was required by Ontario Hydro and AECL in order to comply with those recommendations. This paper describes those efforts, outlines the difficulties encountered, and assesses the lessons learned from them.Seppala, D.; Malaugh, J.; Kiisel, E.; Kamler, F.1996-01-01From the initial steam generator (SG) inspections at Darlington Nuclear Generating Station (DNGS), the authors know that the sludge accumulations on the secondary side tubesheets have been minimal.
DNGS is a fairly new station but the experience at the older Ontario Hydro plants have shown that significant accumulations will happen. A pro-active strategy has been adopted for maintaining SGs that will minimize corrosion product accumulation and the potential for component degradation.
During the four year planned Unit maintenance outages, SGs will be inspected and waterlanced using a waterlance system designed and built by Babcock and Wilcox International. This automated state-of-the-art system also allows fully recorded inspections of the tubesheet/first half-lattice supports. Some of the key elements covered include results of the initial field application (May, 1995), system development and design, system qualification, cleaning performance, and lessons learned for future outages.1992-11-01Ontario Hydro operated Darlington in a safe manner in 1991. Ontario Hydro violated the Atomic Energy Control Regulations once and the physical security regulations three times in 1991. They failed to observe the Operating Licence conditions on ten occasions. The AECB did not find that the individual events had a significant impact on safety. There were no violations of the construction licence.
None of the station staff received a radiation dose in excess of the regulatory limit. Radioactive emissions from the station were far below the regulatory limit.
Special safety system performance was not fully satisfactory. Ontario Hydro failed to meet the unavailability targets for shutdown system one and the negative pressure containment system.
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Ontario Hydro reported seventeen incidents under conditions of the Operating and Construction licences. Units 1 and 2 remained shut down for most of 1991 because of unexplained fuel bundle damage in the reactor core. Ontario Hydro has decided to replace the main generator rotors because of cracks discovered on the rotor shaft. A fully modified rotor was installed on Unit 1. Ontario Hydro staff have a significant backlog of maintenance work.
The Quality Improvement Program seemed to work well, resulting in some noticeable improvements. Three Shift Supervisors and four Control Operators were licensed this year. All planned emergency exercises and drills took place as scheduled. Ontario Hydro identified and are addressing several areas for improvement during the drills.
Except for a power supply interruption to some IAEA equipment, Ontario Hydro achieved all its safeguards goals at Darlington in 1991. The Tritium Removal Facility (TRF) operated intermittently during 1991. Ontario Hydro is proceeding with the design and planning of an annex to the TRF to replace the present temporary facilities.
(Author).Huterer, J.; Brown, D.G.; Yanchula, S.1985-01-01This paper describes the evolution of the internal structure from initial concept to final design. Fundamental changes to the original configuration were precipitated by the action of large seismic forces acting on a top-heavy configuration. Prestressing was eliminated in deference to high humidity.
Aspects of the elevated water tank's peripheral support beam are discussed vis-a-vis an adjacent slipforming operation, and practical construction limitations on steel placement. Also reviewed are the shortening of peripheral columns due to shrinkage and creep, and considerations of crack control for purposes of water-tightness. The authors justify the choice of stainless steel for fabrication of the siphon system's riser pipes. The foundation slab must resist the combined effects of vacuum pressure, hydrostatic uplift, and the seismic reactions of the internal structure and perimeter wall. The dependency of a key foundation component, the gallery roof slab, on the dome tendon layout is high-lighted; and aspects of its constructability are reviewed in light of congestion of vertical tendon anchorages, and of reinforcement. The design of the air-tight slab liner is reviewed, attention focusing on weld design under vacuum and accident temperature loads; on corrosion protection; and on the related construction access bulkhead - its ASME requirements and fabrication tolerances. (orig.).Banica, C.; Foster, M., E-mail: Constantin.Banica@OPG.com Ontario Power Generation, Darlington Nuclear, Bowmanville, Ontario (Canada)2013-07-01In-core neutron flux detectors are used for protective and safety functions in the Darlington NGS 'A' CANDU reactors.
This paper presents new observations regarding the aging of flux detectors, including response to fuelling, response to unit shutdown and indicators of detector noise. Comparisons of detector signals before and after replacement confirm previous assumptions about aging effects. (author).1975-05-01The proposed Darlington GS A project, consisting of four 850 MW CANDU-type reactors, is described. Construction and operation will cause environmental changes with regard to air, water, aquatic life, the site area, safety and noise, and the predicted changes are described. (E.C.B.).McAllindon, D.; Sloan, D.; Mayer, P.1997-01-01Resistance Temperature Detectors (RTDs) and their measurement circuit components have been a significant maintenance item at Darlington.
Analysis of the problems has shown that RTDs and electrical penetrations (EPs) have been the largest sources of faults. A failure mechanism in which the RTD electrical resistance shifts upward was identified as a major source of RTD failures, which prompted a detailed failure investigation by Ontario Hydro Technologies (OHT). The investigation concluded that the root cause failure mechanism is chlorine contamination of the platinum wire during manufacture which resulted in surface damage to the wire and flaking of the wire surface during operation. Electrical penetrations in Darlington are of a pre-built modular design with two crimps internal to the EP. Spurious resistance in poor quality crimps in the EPs lead to errors in resistance measurement. The problem led to a complex and costly job to insert new modules in existing spare EPs. (author).1985-09-01This multi-part report outlines the considerations and results of technical, economic and financial analyses, and analyses of the impact on Ontario's economy associated with continuation of Ontario Hydro's Darlington nuclear generation project.
In addition, it assesses and compares the effects of partial or complete cancellation of the project. The introduction to follow briefly outlines the main contents of each part of the report.Spence, C.G.1995-01-01Darlington NGD is a four unit CANDU 1000 nuclear generation station, located on the shores of Lake Ontario, approximately 45 miles east of Toronto. The CANDU design is a Pressurized Heavy Water Reactor (PHWR), each unit at Darlington is designed to produce 940MWe.
The commissioning phase of the station was completed in the spring of 1993. (author).Carruthers, E.V.; Chow, H.C.1997-01-01Immediately after refuelling of a channel, the fresh bundles are free of fission products.
Xenon-135, the most notable of the saturating fission products, builds up to an equilibrium level in about 30 h. The channel power of the refuelled channel would therefore initially peak and then drop to a steady-state level. The RFSP code can track saturating-fission-product transients and power transients. The Fully INstrumented CHannels (FINCHs) in Darlington NGS provides channel power data on the refuelling power transients.
In this paper, such data has been used to identify the physical evidence of the fission-product transient effect on channel power, and to validate RFSP fission-product-driver calculation results.1975-08-01This report summarizes Ontario Hydro's existing aquatic environmental programs, presents results of these investigations, and outlines plans and activities for expanded aquatic environment studies including the evaluation of alternative cooling systems. This report outlines specific considerations regarding possible alternative cooling arrangements for the Darlington station. It concludes with a recommendation that a study be initiated to examine the potential benefits of using the heated discharge water in a warm water recreational centre.
(author).Liauw, W.K.; Liu, W.S.; Leung, R.K.; Phillips, B.S.1995-01-01Presented here is the TUF simulation of the initial transient of the Class IV power failure event that occurred on November 25, 1993 at Darlington Unit 4. The important physical parameters and models that relate to this event are discussed.
The agreements between the code predictions and the plant data on the thermal-hydraulics and controller responses demonstrate the code reliability for plant operational support. 4 refs., 1 tab., 12 figs.Banica, C. Ontario Power Generation, Darlington Nuclear, Bowmanville, Ontario (Canada); Slovak, R. Ontario Power Generation, IMandCS, Pickering, Ontario (Canada)2011-07-01In-core neutron flux detectors are used for protective and safety functions in the Darlington CANDU reactors. This paper presents observations to date regarding aging of detectors, including recent measurements of prompt fractions and lead cable behaviour during a reactor power rundown. Linear models have been used to estimate and predict the prompt fraction evolution in time using independent variables such as the integrated neutron flux at the detector location, the length of the detector lead cable and the residual current at near-zero flux. (author).Long, T Ontario Hydro, Toronto, ON (Canada); Davey, E C Atomic Energy of Canada Ltd., Chalk River, ON (Canada)1997-09-01The Darlington Nuclear Generating Station (DNGS) is located approximately 40 kilometers east of Toronto, Ontario on the coast of Lake Ontario.
The station consists of four 935 MW(e) pressurized heavy water CANDU type units with a nominal power output of 850 MW(e) per unit. The station was designed and is operated by Ontario Hydro and provides electricity to meet the commercial, industrial and residential needs for 3 million people.
Units 1 and 2 began commercial operation in 1990, followed by Unit 3 in 1991 and Unit 4 in 1992. Since commissioning in 1991, the station has continually achieved annual production of greater than 80% of capacity. At Darlington, as in most other industrial enterprises, the plant annunciation systems play a key role in supporting operations staff in supervising and controlling plant operations to achieve both safety and production objectives.
This paper will summarize the information needs of operations staff for annunciation of changing plant conditions, describe the operational experience with current plant annunciation systems, discuss areas for annunciation improvement, and outline some of the initiatives being taken to improve plant annunciation in the future. 8 refs, 2 figs, 1 tab.Long, T.; Davey, E.C.1997-01-01The Darlington Nuclear Generating Station (DNGS) is located approximately 40 kilometers east of Toronto, Ontario on the coast of Lake Ontario.
The station consists of four 935 MW(e) pressurized heavy water CANDU type units with a nominal power output of 850 MW(e) per unit. The station was designed and is operated by Ontario Hydro and provides electricity to meet the commercial, industrial and residential needs for 3 million people. Units 1 and 2 began commercial operation in 1990, followed by Unit 3 in 1991 and Unit 4 in 1992.
Since commissioning in 1991, the station has continually achieved annual production of greater than 80% of capacity. At Darlington, as in most other industrial enterprises, the plant annunciation systems play a key role in supporting operations staff in supervising and controlling plant operations to achieve both safety and production objectives. This paper will summarize the information needs of operations staff for annunciation of changing plant conditions, describe the operational experience with current plant annunciation systems, discuss areas for annunciation improvement, and outline some of the initiatives being taken to improve plant annunciation in the future. 8 refs, 2 figs, 1 tab.Lee, J.H.S.; Knystautas, R.1989-03-01The present report reviews the Darlington Safety Report (DSR) which has been used as basis for decisions regarding intentional ignition in the Darlington reactor vault. The validity of the assumptions in the DSR regarding mixing of contents is assessed and possible hydrogen release scenarios, specific to the Darlington reactor vault, are examined. The combustion analysis in the DSR vent code calculations are reviewed in the light of existing state of the art information on high speed turbulent flames and transition to detonation. Limitations of the vent code, in this context, are identified and improvements recommended.Plourde, J.; Parmar, R.2006-01-01In 2004, the Darlington Nuclear (DN) Plant of Ontario Power Generation (OPG) undertook a project, in partnership with Nuclear Safety Solutions (NSS) Limited, to develop Risk-Informed Asset Management (RIAM) and Generation Risk Assessment (GRA) models.
The models are intended to optimize plant decision-making. The objective of this paper is to present the scope of the project, the methodology employed, the results and the potential applications. DN has recognized the strategic importance of RIAM in the plant decision-making process and has begun its implementation.
The required work was split into three phases. Phase 1 involved industry benchmarking, along with collection and review of the industry literature such as EPRI publications and other relevant papers. Based on the review, a description of the requirements to produce a prototype RIAM model was developed. Phase 2 consisted of the development of prototype RIAM and GRA models. Phase 3, currently underway, consists of the work required to translate the prototype models into an operational decision-making tool.
RIAM and GRA are relatively new concepts hence the related methodology and the tools are still evolving. NSS has tailored the available methodology to suit the needs of the DN plant. Draft EPRI guides on GRA/RIAM were used in developing DN specific methodology. The details are provided in the paper.
At DN, RIAM is expected to support business decisions by facilitating the assessment of risks associated with projects, programs and business case alternatives. These applications are further discussed in the paper. (author).Norsworthy, A G; Ditschun, A Atomic Energy of Canada Ltd., Mississauga, ON (Canada)1996-12-31As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle.
Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. 4 refs., 2 tabs., 13 figs.Norsworthy, A.G.; Ditschun, A.1995-01-01As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube.
The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning.
This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. 4 refs., 2 tabs., 13 figs.Skelton, P.H.; Sie, T.1996-01-01Darlington NGD requires eight fuelling machine heads to fuel the four 932 MW reactors. Six heads are used on the three fuelling machine trolleys for normal fuelling operations. A further two heads are required to allow for maintenance and to provide for such reactor face activities as PIPE and CIGAR. Seven heads were successfully delivered to site from the head supplier. During acceptance testing, stalls on the charge tube screw assembly of the eighth and final head prevented its delivery to site.
Replacement of the charge tube screw with a spare screw did not alleviate the problem. An in depth series of tests were undertaken at site, at the supplier and at the screw sub-supplier to determine the root cause of the problem. These tests included taking torque measurements under different operating conditions and using different components to assess the effects of the changes on torque levels.
An assessment of the effects of changing chemical conditions (particularly crud levels) was also made. To ensure that the results of the testing were well understood, additional torque testing was also completed on a head and screw assembly at site that was known to work well. Based on all of the above series of tests, a recommendation was made to re-machine the charge tube screw(s). The original charge tube screw from Head eight was subsequently returned to the sub-supplier for re-work. Follow-up torque measurements and acceptance testing showed that the screw rework was effective and that Head eight could be successfully delivered to site.
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This paper focuses on the results of the head/screw test program. Results of the acceptance testing are also discussed. 2 refs., 4 figs.Skelton, P H; Sie, T Ontario Hydro, Bowmanville (Canada). Darlington Nuclear Generating Station; Pilgrim, J Canadian General Electric Co.
Ltd., Toronto, ON (Canada)1997-12-31Darlington NGD requires eight fuelling machine heads to fuel the four 932 MW reactors. Six heads are used on the three fuelling machine trolleys for normal fuelling operations. A further two heads are required to allow for maintenance and to provide for such reactor face activities as PIPE and CIGAR. Seven heads were successfully delivered to site from the head supplier.
During acceptance testing, stalls on the charge tube screw assembly of the eighth and final head prevented its delivery to site. Replacement of the charge tube screw with a spare screw did not alleviate the problem. An in depth series of tests were undertaken at site, at the supplier and at the screw sub-supplier to determine the root cause of the problem. These tests included taking torque measurements under different operating conditions and using different components to assess the effects of the changes on torque levels.
An assessment of the effects of changing chemical conditions (particularly crud levels) was also made. To ensure that the results of the testing were well understood, additional torque testing was also completed on a head and screw assembly at site that was known to work well. Based on all of the above series of tests, a recommendation was made to re-machine the charge tube screw(s). The original charge tube screw from Head eight was subsequently returned to the sub-supplier for re-work. Follow-up torque measurements and acceptance testing showed that the screw rework was effective and that Head eight could be successfully delivered to site. This paper focuses on the results of the head/screw test program.
Results of the acceptance testing are also discussed. 2 refs., 4 figs.Suryanarayan, S.; Husain, A.; Williams, D.2010-01-01Darlington Nuclear Generating Station (DNGS) has accumulated over 48 drums of chemistry laboratory waste arising from analysis of heavy water (D 2 O). Several organic, including Arsenazo III, and inorganic contaminants present in these drums results in high total organic carbon (TOC) and conductivity. These drums have not been processed due to uncertainties related to clean-up of Arsenazo III contaminated heavy water.
This paper provides details of chemical characterization as well as bench scale studies performed to demonstrate the feasibility of treating the downgraded D 2 O to the stringent target specifications of.Irvine, H.S.; Bennett, E.J.; Talbot, K.H.1986-10-01Being able to technically and economically replace the most radioactive components (excluding the nuclear fuel) in operating reactors will help to ensure the ongoing acceptance of nuclear power as a viable energy source for the future. Ontario Hydro is well along the path to meeting the above objective for its CANDU-PHW reactors. Following the failure of a Zircaloy-II pressure tube in unit 2 of Pickering NGS A in August, 1983, Ontario Hydro has embarked on a program to replace all Zircaloy-II pressure tubes in units 1 and 2 at Pickering. This program integrates the in-house research, design, construction, and operating skills of a large utility (Ontario Hydro) with the skills of a national nuclear organization (Atomic Energy of Canada Limited) and the private engineering sector of the Canadian nuclear industry. The paper describes the background to the pressure tube failure in Pickering unit 2 and to the efforts incurred in understanding the failure mechanism and how similar failures are not expected for the zirconium-niobium pressure tube material used in all other large CANDU-PHW units after units 1 and 2 of Pickering NGS A. The tooling developed for the pressure tube replacement program is described as well as the organization to undertake the program in an operating nuclear station. The retubing of units 1 and 2 at Pickering NGS A is nearing a successful completion and shows the benefits of being able to integrate the various skills required for this success.
Pressure tube replacement in a CANDU-PHW reactor is equivalent to replacement of the reactor vessel in a LWR. The fact that this replacement can be done economically and with acceptable radiation dose to workers augurs well for the continued viability of the use of nuclear energy for the benefit of mankind. (author).Washington, A. II Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab.
(SRNL); Peters, T. Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)2014-03-03This report summarizes the results of the extraction, scrub, and strip testing for the September 2013 sampling of the Next Generation Solvent ( NGS) Blended solvent from the Modular Caustic Side-Solvent Extraction Unit (MCU) Solvent Hold Tank.
MCU is in the process of transitioning from the BOBCalixC6 solvent to the NGS Blend solvent. As part of that transition, MCU has intentionally created a blended solvent to be processed using the Salt Batch program. This sample represents the first sample received from that blended solvent. There were two ESS tests performed where NGS blended solvent performance was assessed using either the Tank 21 material utilized in the Salt Batch 7 analyses or a simulant waste material used in the V-5/V-10 contactor testing. This report tabulates the temperature corrected cesium distribution, or D Cs values, step recovery percentage, and actual temperatures recorded during the experiment. This report also identifies the sample receipt date, preparation method, and analysis performed in the accumulation of the listed values.
The calculated extraction D Cs values using the Tank 21H material and simulant are 59.4 and 53.8, respectively. The DCs values for two scrub and three strip processes for the Tank 21 material are 4.58, 2.91, 0.00184, 0.0252, and 0.00575, respectively.
The D-values for two scrub and three strip processes for the simulant are 3.47, 2.18, 0.00468, 0.00057, and 0.00572, respectively. These values are similar to previous measurements of Salt Batch 7 feed with lab-prepared blended solvent. These numbers are considered compatible to allow simulant testing to be completed in place of actual waste due to the limited availability of feed material.Washington, A. II; Peters, T. B.2014-01-01This report summarizes the results of the extraction, scrub, and strip testing for the September 2013 sampling of the Next Generation Solvent ( NGS) Blended solvent from the Modular Caustic Side-Solvent Extraction Unit (MCU) Solvent Hold Tank. MCU is in the process of transitioning from the BOBCalixC6 solvent to the NGS Blend solvent. As part of that transition, MCU has intentionally created a blended solvent to be processed using the Salt Batch program.
This sample represents the first sample received from that blended solvent. There were two ESS tests performed where NGS blended solvent performance was assessed using either the Tank 21 material utilized in the Salt Batch 7 analyses or a simulant waste material used in the V-5/V-10 contactor testing. This report tabulates the temperature corrected cesium distribution, or DCs values, step recovery percentage, and actual temperatures recorded during the experiment. This report also identifies the sample receipt date, preparation method, and analysis performed in the accumulation of the listed values.
The calculated extraction DCs values using the Tank 21H material and simulant are 59.4 and 53.8, respectively. The DCs values for two scrub and three strip processes for the Tank 21 material are 4.58, 2.91, 0.00184, 0.0252, and 0.00575, respectively. The D-values for two scrub and three strip processes for the simulant are 3.47, 2.18, 0.00468, 0.00057, and 0.00572, respectively. These values are similar to previous measurements of Salt Batch 7 feed with lab-prepared blended solvent.
These numbers are considered compatible to allow simulant testing to be completed in place of actual waste due to the limited availability of feed material.CERN PhotoLab1983-01-01Final prototype for the LEP vacuum chamber, see 8305170 for more details. Here we see the strips of the NEG pump, providing 'distributed pumping'.
The strips are made from a Zr-Ti-Fe alloy. By passing an electrical current, they were heated to 700 deg C.Bartlett, A.J.; Lessard, P.A.1984-01-01This paper is a review of the problems and tradeoffs involved in cryogenic vacuum pump analysis, design and manufacture. Particular attention is paid to the several issues unique to cryopumps, e.g., radiation loading, adsorption of noncondensible gases, and regeneration. A general algorithm for cryopump design is also proposed. 12 references.CERN PhotoLab1971-01-01Some of the most important components of the vacuum system are shown. At the left, the rectangular box is a sputter-ion pump inside its bake-out oven. The assembly in the centre includes a sector valve, three roughing valves, a turbomolecular pump, a rotary backing pump and auxiliary equipment.
At the right, the small elbow houses a Bayard-.CERN PhotoLab1970-01-01A pressure of 5 x 10-11 Torr has been obtained repreatedly in this pilot section of the ISR vacuum system. The pilot section is 45 m long is pumped by 9 sputter-ion pumps pf 350 l/s pumping speed, and is baked out at 200 degrees C before each pump down.1983-01-01This is a cut-out of a LEP vacuum chamber for dipole magnets showing the beam channel and the pumping channel with the getter (NEG) strip and its insulating supports.
A water pipe connected to the cooling channel can also be seen at the back.The lead radiation shield lining is also shown.