bluuebunny/crossref_metadata_2025_split huggingface dataset:
Building the UK ABWR: Developments in Structural Integrity Safety Cases for Highest Reliability Components in the UK
Authors: Jelfs Tim , O’Neill James , Beveridge Angus 
Published: October 2018

Nuclear power plants contain certain components whose gross failure would lead to intolerable radiological consequences. In the UK, a common terminology used for such components is Very High Integrity (VHI). If it is not possible to engineer lines of protection for these components, a safety case must demonstrate to UK regulators that the probability of gross failure is demonstrably so low that it can be discounted. A previous paper [Ref. 1] has described, at a high level, how the structural integrity safety case for a nuclear new build project in the UK — the UK Advanced Boiling Water Reactor (UK ABWR) is being structured.
As described in [Ref. 1], the structural integrity safety case for the UK ABWR is based on the guidance provided by the UK Technical Advisory Group on Structural Integrity (TAGSI) and aims to demonstrate a multi-legged safety case with robust and independent legs giving confidence of defense in depth. Design to the internationally recognized ASME code [Refs. 2, 3, 4] is supplemented by a significant number of beyond code requirements such as supplementary inspection and inspection qualification, augmented material testing requirements, defect tolerance assessment to the well-established R6 procedure [Ref. 5], and demonstration that design and manufacturing processes have reduced risks to As Low as Reasonably Practicable (ALARP).
This paper provides an updated position of the progress made on the UK ABWR project. It also provides more specific details on the activities the future licensee, Horizon Nuclear Power, has performed in support of the demonstration that design and manufacturing processes have reduced risks to ALARP. This kind of additional work is vital to providing the UK regulator with confidence that the risk of failure of VHI components has been reduced to ALARP.
PTS on Embedded Rotated Cracks and Projection Rules
Authors: Stefanini L. , Blom F. J. 
Published: October 2018

The Pressurized Thermal Shock (PTS) is one of the transients of concern in a nuclear reactor. The phenomenon takes place when the down-comer and internal Reactor Pressure Vessel (RPV) surfaces are rapidly cooled down while the system is still under pressure.
This event is considered life limiting for the safety case where cracks are postulated and the RPV material embrittled by neutron radiation damage.
During this study, an elliptical embedded crack was postulated in a cylindrical geometry simulating the RPV. The system, initially at high temperature and under pressure load condition, was subject to a PTS transient. Thermal and structural analyses were carried out for various crack configurations. Particularly, besides the axial crack, skewed embedded axial cracks were modeled and analyzed (30°, 45° and 65°) for PTS phenomenon. Finite element analysis were carried out with ANSYS considering two variants: no-heat transmission and total heat transmission through the crack. Results showed that the highest stress intensity factor is obtained at an axial crack (0°). Moreover, it is shown that considering no-heat transmission through the crack is a conservative approach.
The ASME code approach to the skewed axial cracks characterization by projection is considered. It is found that the ASME approach is conservative, with an excess conservatism for largely skewed axial cracks. The same conclusions were reached when considering the projection rules from the R6 code.
The Follow Up on Profiled Wire Gasket Development and Testing With Case Studies
Authors: Currie Anthony 
Published: October 2018

The paper discusses the development of sealing technology within the organisation with particular emphasis on the gasket manufactured by profiled winding wire. This will be referred to as a PWG.
The paper is a continuation of the paper presented in Paris PVP2013-97050 and follow up in PVP2014-28948. It discusses further laboratory testing and briefly describes 4 related field service case studies.
The paper details tests which include: Thermal cycle, sealing performance, compression and recovery testing carried out in both laboratory and field conditions. The 4 related studies are based around the use on large flanges, the use in oxygen rich environment, the use in a thermal cyclic environment and on problematic flange corrosion.
Compact Crack Arrest Tests for the Validation of a Finite Element Material Model of the Reactor Pressure Vessel Steel of the Nuclear Power Plant KKG
Authors: Mayer Uwe , Mutz Alexander , Nicak Tomas 
Published: October 2018

The integrity of a reactor pressure vessel (RPV) has to be given for its operating time in accordance with the regulations. An assessment of the RPV against brittle failure needs to be conducted specially for one of the most severe loading cases. This is the loss-of-coolant accident (LOCA). Cold water is injected into the RPV at operating conditions. This thermal shock of the ferritic pressure vessel wall leads to loading conditions at the beltline area known as Pressurized Thermal Shock (PTS).
The assessment against brittle failure is based on a deterministic fracture analysis. Common parameters like stress intensity factors are employed to calculate the PTS event for an assumed (postulated) flaw. Subsequently the results of the fracture mechanics analysis are compared with material properties obtained from the irradiation surveillance program of the RPV to demonstrate the exclusion of brittle fracture initiation.
The validation considers the material data and the velocity of the crack growing into the specimen until it stops. The measured crack propagation velocity for tests performed according to ASTM E1221 [1] is compared to the result of the Finite Element (FE) simulation of a Compact Crack Arrest (CCA) test.
For five of eight tests performed at −60 °C crack propagation velocity values were determined ranging from 509 m/s to 694 m/s with an average value of 618 m/s.
Pneumatic Testing of Piping: Managing the Hazards for High Energy Tests
Authors: Arti Bader , Weyer Robert , Dang Thanh , Taagepera Jaan 
Published: October 2018

Pneumatic pressure testing is used extensively during construction of LNG plants to avoid the problems that can be caused by water that may be left in a piping system (particularly in valves) if hydrotesting was performed. Due to the much greater compressibility of gas, there is significantly more energy (and thus risk) associated with a pneumatic test than with a hydrostatic test.
A major gas project safely conducted numerous pneumatic tests with stored energies of up to 6,675 MJ. Observing a commonly used limit of 270 MJ would have resulted in hundreds of additional closure welds.
This paper discusses practical aspects of performing pneumatic testing, the risk mitigations put in place and present two calculation methods that can be used to check whether exclusion zones for blast wave pressure are adequate for fragment throw.
FEM Contact Stress Analysis at the Bearing Surfaces in Bolted Joints Under External Loadings
Authors: Shirakawa Atsushi , Sawa Toshiyuki , Naruse Tomohiro 
Published: October 2018

The higher bolt preload is better for increasing the integrity of bolted joints. However, the bearing surface could be permanently deformed with too much high preload and the bolt preload decreases with permanent deformation. Thus, an upper limit of bolt preload per contact area at the bearing surfaces, which is called as the critical contact stress, must be examined for various materials of clamped parts. In the previous research, the critical contact stresses were obtained by compressing a cylindrical bar. However, the contact stress distribution of compressing a cylindrical bar is much different to that of an actual bolted joint. In addition, a lot of previous researches used the stress-strain curves under tensile loadings for analyzing the contact stress distributions at the bearing surfaces.
In this study, the Finite Element Method is used for contact stress analysis at the bearing surfaces in clamped bolted joints under external loading. Various materials for clamped part are chosen such as steel, aluminum alloy, cast iron and stainless steel. Firstly, the stress-strain curves for the above materials were measured under compressive loadings. The differences in the stress-strain curves under tensile and compressive loadings are evaluated with the above materials. Next, using actual bolts, the displacements at the bearing surfaces in cylindrical clamped parts consisting of the above materials were measured. Finally, elasto-plastic FEM contact stress analyses are carried out for analyzing the plastic displacements at the bearing surfaces under compressive loadings. The mean displacement at the bearing surfaces are compared with those obtained from the experiments for each material. The FEM results of the relationship between the contact stress and the plastic displacement at the bearing surface are in a fairy good agreement with the experimental results for each clamped material. In addition, the critical contact stress at the bearing surface is discussed for each material. Two proposals for the critical contact stress at the bearing surface are described and compared with the critical contact stress of VDI2230.
Technical Basis for Proposed Fifth Revision to ASME Code Case N-513
Authors: Cimock Dylan , Houston Eric J. , Cipolla Russell C. , McGill Robert O. 
Published: October 2018

Code Case N-513 provides evaluation rules and criteria for temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. The application of the Code Case is restricted to moderate energy, Class 2 and 3 systems, so that safety issues regarding short-term, degraded system operation are minimized. The first version of the Code Case was published in 1997. Since then, there have been four revisions to augment and clarify the evaluation requirements and acceptance criteria of the Code Case that have been published by ASME. The technical bases for the original version of the Code Case and the four revisions have been previously published [1, 2, and 3].
There is currently work underway to incorporate additional changes to the Code Case and this paper provides the technical basis for the changes proposed in a fifth revision. These changes include clarification for buried piping, investigation of various radii used in the Code Case, removal of the 0.1 limit on the flexibility characteristic for elbow flaw evaluation, and an update of the stress intensity factor parameters for circumferential through-wall flaws. In addition, a new flaw evaluation procedure is given for through-wall flaws in gate valve body ends. This procedure evaluates flaws in the end of the valve body as if in straight pipe. These changes and their technical bases are described in this paper. Clarifications and changes deemed editorial are not documented in this paper.
A Treatise on Weld Evaluation
Authors: Chandiramani Dipak K. , Narayana Sadasivan 
Published: October 2018

Welds are inspected by various techniques which include visual examination, surface examination and volumetric examination.
While the above techniques would qualify a weld to workmanship criteria, they would not necessarily be indicative of weld properties.
Preparation and qualification of welding procedures and testing of production welds are indicative that the weldment would probably provide a safe and satisfactory service life.
However, weldments have to operate at their design conditions which may include high temperatures and ASME Codes do not necessarily stipulate tests for verification of high temperature properties. In addition, defective welds are often repaired by removing the originally deposited weld metal and re-welding. The effects of double heat input are not necessarily evaluated.
In this paper, an insight is provided into the factors which provide assurance that weldments will perform satisfactorily in service and the combination of non-destructive evaluation methods which would enable effective detection of imperfections.
Non-destructive volumetric examination method for welds has traditionally be radiography. With the advent of automated data acquisition methods in Ultrasonics, like Time of Flight Diffraction and Phased Array Ultrasonic Testing, these methods are rapidly replacing radiographic methods for weld inspection.
Ultrasonic acceptance criteria in ASME Section VIII Div. 1, ASME Section VIII Div. 2 and ASME Section IX do not include evaluation of porosity as ultrasonic methods do not easily detect porosity.
The result of all this is that today we are accepting welders qualified using Ultrasonic examination as per ASME Section IX but on the job there is still the option of inspecting the weld using Radiography in which, excessive porosity can be a cause for weld repair.
Considering this and various other criteria, a comprehensive weld evaluation methodology is proposed taking advantage of the strengths of each inspection technique while welding technology used would ensure that welds have required properties at service temperatures.
A proposal is also made to improve the detectability of imperfections using modifications of existing Ultrasonic A-scan Techniques.
Rules for Flaw Interaction for Subsurface Flaws in Operating Pressurized Vessels: Technical Basis of Code Case N-877
Authors: Lacroix Valéry , Dulieu Pierre , Blasset Sebastien , Tiete Ralf , Li Yinsheng , Hasegawa Kunio , Bamford Warren , Udyawar Anees 
Published: October 2018

When multiple flaws are detected in pressure retaining components during inspection, the first step of evaluation consists of determining whether the flaws shall be combined into a single flaw or evaluated separately. This combination process is carried out in compliance with proximity rules given in the Fitness-for-Service (FFS) Codes. However, the specific criteria for the rules on combining multiple flaws into a single flaw are different among the FFS Codes.
In this context, revised and improved criteria have been developed, to more accurately characterize the interaction between multiple subsurface flaws in operating pressure vessels. This improved approach removes some of the conservatism in the existing ASME Code approach, which was developed in the 1970s based on two flaws interacting with each other.
This paper explains in detail the methodology used to derive improved flaw proximity rules through three-dimensional FEM and XFEM analyses. After the presentation of the calculations results and the improved criteria, the paper also highlights the multiple conservatisms of the methodology using several sensitivity analyses.
Assessment of the Fitness for Service of the Flamanville EPR Reactor Pressure Vessel Closure Head and Bottom Head Domes Containing a Segregation Zone Characterized by a High Carbon Content
Authors: Delvallée-Nunio Isabelle , Loiseau Olivier , Monhardt Daniel , Buiron Audrey , Dubois Franck 
Published: October 2018

To complete the manufacturing process qualification of the closure head and bottom head domes of the Flamanville EPR RPV in 16MND5 ferritic steel (similar to A508 cl3), AREVA NP performed in 2014 material testing on a sacrificial dome so-called UA Sup. These tests revealed that the Charpy impact energy at 0°C — in the positive segregation zone at ¼ thickness — was lower than 60 J, the minimum value required by the French regulation. Thereafter, this result was explained by the high carbon content close to 0.3 % — against 0.18 % in the ladle — due to the forging process. The positive segregation was insufficiently eliminated to avoid high carbon content in certain location.
As the level of Charpy impact energy is an indicator of the level of fracture toughness, a thorough material characterization of the segregation zone, as well as a brittle fracture analysis have been proposed and carried out by AREVA NP to demonstrate that the Flamanville EPR RPV domes are tough enough for a safe operation.
The principles of the demonstration approach were reviewed by ASN and IRSN in 2015 and the final safety case was submitted to the ASN and IRSN review in December 2016. The latter led to the conclusion that, even if the segregation zone fracture toughness is lowered in comparison to that of the acceptance test zone, this does not put into question the Flamanville EPR RPV fitness for service as long as adequate in-service inspection of the domes is periodically performed.
This paper aims at presenting the key elements of the IRSN assessment from part of which this conclusion was drawn, namely the material testing program, the description of the loadings in normal and accidental conditions, and the brittle fracture analysis.
NTD ASI, Section IV: Evaluation of Residual Lifetime of Components and Piping in WWER Type NPPs
Authors: Brumovsky Milan 
Published: October 2018

VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was initiated and co-ordinated by the Czech and was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for Nuclear Power Plants (NPPs) with WWER (Water-Water-Energetic-Reactor = PWR type) type reactors, as those codes were developed only for design and manufacture and were not changed since their second edition in 1989.
VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes.
Last upgrading and principal extending of this VERLIFE Procedure was realized within the 3-years IAEA project (in close co-operation with another project of the 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) that started in 2009 with final approval and editing in 2013.
As all versions of the VERLIFE procedure were coordinated by the Czech and first version was based on the Czech version of the NTD ASI, there have been simultaneously incorporated into the Czech NTD ASI (Normative Technical Documentation of the Czech Association of Mechanical Engineers) guidelines that are accepted by the Czech State Office for Nuclear Safety for the use in evaluation of Czech NPPs.
This document has several parts: Section IV – “Evaluation of Residual Lifetime of Components and Piping in WWER type NPPs” deals with the evaluation during NPP operation. Main part of the document is divided into four main parts:
- Evaluation of resistance of components and piping against non-ductile failure
- Evaluation of resistance of component and piping against fatigue damage,
- Evaluation of resistance of components and piping against corrosion-mechanical damage
- Evaluation of residual lifetime of components and piping with defects found during in-service inspections
Additionally, several appendices are included for detailed description of individual parts of evaluation, e.g.
- Determination of neutron fluences in reactor pressure vessel and internals
- Determination of degradation of materials during operation
- Requirements for evaluation of pressurized thermal shock regimes
- Evaluation of corrosion-errosion effects in piping
- Environmental fatigue evaluation
- Evaluation of reactor pressure vessel failure probability
Finally, the following appendices dealing with components integrity have been included:
- Lifetime of reactor pressure vessel internals
- Leak-before-break concept for WWER piping
- No-break-zone for WWER piping
The paper will describe structure and main principles of this Section IV.
Evaluation of Reactor Internals Integrity and Lifetime According to the NTD ASI
Authors: Brumovsky Milan 
Published: October 2018

Reactor internals are components that are no typical pressure boundary but they are nevertheless very important as they hold fuel elements and all reactor control system elements and thus must ensure their safe and reliable operation during the whole reactor life under all operating and even beyond bases regimes.
In principle, reactor internals can be replaced but their weight, quantity of very high activated material and cost such possibility practically excluded.
Thus, evaluation of the reactor internals condition and prediction of their behavior during the whole or even extended lifetime is of high importance. Reactor internals are subjected to very high neutron irradiation that could initiated not only stress corrosion (irradiation assisted) cracking but also large embrittlement and changes in dimensions (swelling and creep).
VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was initiated and co-ordinated by the Czech and was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for Nuclear Power Plants (NPPs) with WWER (Water-Water-Energetic-Reactor = PWR type) type reactors, as those codes were developed only for design and manufacture and were not changed since their second edition in 1989.
VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes.
Within the last upgrading and principal extending of this VERLIFE Procedure was developed within the 3-years IAEA project (in close co-operation with another project of the 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) that started in 2009 with final approval and editing in 2013”) a part dealing with the evaluation of reactor internals lifetime was elaborated..
This IAEA VERLIFE procedure for internals has been implemented into the existing Normative Technical Documentation (NTD) ASI (Czech Association of Mechanical Engineers), Section IV – Evaluation of Residual Lifetime of Components and Piping in WWER type NPPs.
Main damaging mechanisms that should be taken into account in reactor internals and the procedure are described in detail with necessary formulae for materials of internals:
- Radiation hardening
- Radiation embrittlement
- Radiation swelling
- Radiation creep
- Swelling under stress effect
- Swelling inducing embrittlement
- Irradiated assisted stress corrosion cracking
- Transformation austenite-ferrite
and also the method for evaluation of the resistance against non-ductile failure of postulated defect.
The paper will describe these main principles and also more detailed information on the procedure for evaluation of reactor internals will be given.
Interlaboratory Study for Small Punch Testing Preliminary Results
Authors: Brumovsky Milan , Kopriva Radim 
Published: October 2018

Small punch test specimens are widely used for a long time as they are simple to produce and requires only a small volume of material. This fact is advantageous especially for high activity materials but also for assessment of operational damage in components materials when component integrity and strength may not be affected. In the same time, no test standard exists and several different specimen types and test procedures have been developed in different place.
Thus, to unify this activity, considerable attention has been paid since 2012 to the standardization of small punch test technique within the American Society of Testing and Materials (ASTM). In 2016 a large InterLaboratory Study has been launched within the ASTM subcommittee E10.02 - Behavior and Use of Nuclear Structural Materials, involving 12 laboratories and 6 evaluated structural materials from the nuclear and non-nuclear power plant components.
Paper describes the current status of ASTM standardization, results of the InterLaboratory Study, first analysis of the results with respect to some important test parameters, lessons learned and open questions remaining to be solved for the successful completion of the standardization process.