In-service Inspection Program Research Articles

Operating Experience of Snubbers in Indian Pressurized Heavy Water Reactors

In a Nuclear Power Plant (NPP), the piping layout needs to have flexibility to reduce the operating stresses and nozzle loads due to thermal expansion. At the same time, the piping and equipment need to be supported adequately to sustain the dynamic loads viz., seismic loads, sudden opening/closing of valves etc. In Indian Pressurized Heavy Water Reactors (PHWRs), Snubbers are provided on nuclear equipment & piping to serve both purposes i.e. allow flexibility for thermal movement during operation and at the same provide requisite rigidity to take care of dynamic loads. Snubbers are subjected to harsh environment during operation including high temperature, radiation, vibration and corrosive condition, which may adversely affect its performance. Various degradations viz., high drag force, shift in dynamic characteristics (locking characteristic), jamming etc. have been observed. Periodic testing of Snubbers is required to ensure the proper functioning during normal operating condition as well as during dynamic event. A dedicated In-Service Inspection (ISI) program for snubbers exists in Indian PHWRs to ensure healthiness/functionality of Snubbers. Efforts have been made to reduce numbers of Snubbers progressively in PHWR units. Numbers of snubbers have been significantly reduced in latest 700MWe PHWR Design. This paper covers the various degradations observed in snubbers, ISI program for snubbers, design modifications, Snubber testing, assessment for fitness for service, replacement of Snubbers in older units and optimization of snubbers.

STUDY ON IN-SERVICE INSPECTION PROGRAM AND METHOD FOR FUEL HANDLING SYSTEM OF RDE

One of the main programs that should be established on the designing process of the fuel handling system is to establish an in-service inspection program for maintaining the integrity of the system, structure and component during service lifetime. The most important role of in-service inspection is the nondestructive examination techniques. The objective of this study is to propose a preliminary program for examining the integrity of the fuel handling system during operation and determining the best method to confirm the defects. The proposed programs are described as follows, defining the operating environment of the fuel handling system, identifying the material characteristics during operation which indicates to promote the defects, and selecting the appropriate method of non-destructive examination and analysis technique for such kind of defects. The proposed in-service inspection program is expected to give significant additional value to the fuel handling system design of RDE.

ASME in-service inspection and in-service testing programs for Angra 1

The in-service inspection program of the Angra 1 plant is updated every 10 years, according to applicable standards - designer (American Project - are followed NRC requirements) and Cnen. NRC approves the use of ASME Section XI (In-service Inspection of Nuclear Power Plant Components). The object of in-service inspection of components in nuclear power plants is to provide a continuing assurance that they are safe. To provide this assurance for those components that are subject to the requirements of the ASME Boiler and Pressure Vessel Code, a set of rules has been formulated to provide assurance that the functional requirements of the components are available when required. The rules have been arranged to provide appropriate levels of assurance according to the importance of the component in its relationship to plant safety. The classifications that are established during design and manufacturing have been adopted to provide the levels of importance for the components. The types of components typically found in the various classifications have then been identified and rules formulated for each type. For each type of component in each classification, the functions have been considered and methods of inspecting, testing, or monitoring each component is specified. These rules include methods of determining the limits of acceptance of the results. Should it be necessary to take corrective action to repair various components, rules have been provided to establish acceptable methods of repair or replacement. Angra 1 started the Renewal License and Long-term Operation project and there are three important Aging Management Programs (AMP) that are based on ASME section XI. This article will discuss the ASME section XI subsections that are important for the License Renewal and Long-term Operation for Angra 1.

Development of risk-based inspection programme for BARC Research Reactors–Dhruva: experience and insights

As part of defence-in-depth philosophy an In-Service Inspection programme for research reactors at BARC has been implemented. The current ISI methodology is based on deterministic approach through the application of rules and criteria provided in ASME Section 11 and Section 5. For the research reactor Dhruva two campaigns of ISI have been completed. The risk-informed approach is widely employed for Nuclear Power Plants; however, there is no published literature on Risk-informed ISI for a research reactor. Development work on risk-informed ISI has been performed for Dhruva and a document has been prepared and submitted to safety committees for review and approval. The objective is to utilize the available Level 1 PSA and ISI insights to optimize the ISI programme. This paper, discusses the development approach and insights gained from this project and provides recommendation on initiating activities to encourage other facilities to consider implementation of RI-ISI for research reactors.

Preparation and Management of In-service Inspection Schedule Program in Nuclear Power Plants

In-service Inspection (ISI) schedule program for nuclear island equipment was of the important part of ISI program file system in nuclear power plants (NPPs), which acted as a connecting link. Based on ISI rules for the mechanical components of PWR nuclear islands (RSE-M) and the plant ISI general program, the preparation method for schedule program of CPR1000 units were presented. It focused on the analysis and discussion of the existing problems during the preparation, upgrade and management process and some targeted suggestions were put forward. It may lay a foundation for a further optimization of the ISI program file system and the smooth implementation of the ISI activities.

Impact of In-Service Inspection on the Reliability of Small Piping

This chapter describes the cracking of stainless steel piping under Inter-granular Stress Corrosion Cracking (IGSCC) conditions using probabilistic fracture mechanics that predict the impact of in-service inspection (ISI) programs on the reliability of specific nuclear piping systems that have failed in service. The IGSCC is characterized by a single damage parameter, which depends on residual stresses, environmental conditions, and the degree of sensitization. The Probability of Detection (POD) curves and the benefits of in-service inspection in order to reduce the probability of the leak for nuclear piping systems subjected to IGSCC were discussed. The results show that an effective ISI requires a suitable combination of crack detection and inspection schedule. An augmented inspection schedule is recurred for piping with fast-growing crack to ensure that the inspection is done before the cracks reach critical sizes and that the use of a better inspection procedure can be more effective than a tenfold increase in the number of inspections of inferior quality.

STRATEGIES FOR INSERVICE INSPECTION OF COMPACT HEAT EXCHANGERS IN HIGH TEMPERATURE REACTORS

The U.S. Department of Energy is sponsoring research to support the development of the compact heat exchanger (CHX) for use in high temperature advanced reactors. The project is being executed by an Integrated Research Project (IRP) and includes team members from the University of Wisconsin– Madison, University of Michigan, Georgia Institute of Technology, University of Idaho, North Carolina State University, Oregon State University, Electric Power Research Institute, MPR Associates, and heat exchanger manufacturers CompRex and Vacuum Process Engineering (References 1& 2). The objective of the research is to enable the use of compact heat exchanger designs in high temperature advanced reactor service in order to improve plant efficiency and economics. A necessary step for achieving this objective is to ensure that the ASME Boiler and Pressure Vessel Code, Section III (Reference 3) has rules for the construction of CHXs for nuclear service. However, construction rules alone are not sufficient to deploy a compact heat exchanger in an advanced reactor. ASME Section XI (Reference 4) Rules for the Inservice Inspection (ISI) of a heat exchanger in an operating nuclear reactor will be required as well. The failure mechanisms for compact heat exchanger are unique and are not well understood since they have limited operating experience. The IRP will develop an inservice inspection roadmap to guide IRP research that will enable an owner/operator of an advanced nuclear reactor to implement an appropriate inservice inspection program for a compact heat exchanger. Failure mechanisms and their associated probabilities are the key inputs to a successful inspection program. The inspection roadmap will make use of the soon to be published ASME Section XI, Division 2, Reliability and Integrity Management (RIM) methodology. RIM is a non-deterministic, technology neutral, approach to inservice inspection. A RIM based ISI program is developed by an owner considering the entire reactor technology safety case rather than on the specific component technology. Therefore, owners will need specific information regarding the performance, failure modes, reliability, and inspection options for compact heat exchangers in order to incorporate such a heat exchanger into the plant design. This paper outlines the IRP research as it relates to inputs needed by an owner for the eventual development of an ISI program that includes compact heat exchangers.

The influence of core power distribution on neutron flux density behind a pressure vessel of a VVER-1000 Mock Up in LR-0 reactor

The neutron flux distribution behind a reactor pressure vessel (RPV) is an important parameter that is monitored to determine neutron fluence in the RPV. Together with mechanical testing of surveillance specimens, these are the most important parts of in-service inspection programs that are essential for a realistic and reliable assessment of the RPV residual lifetime. The fast neutron fluence values are determined by a calculation. These calculation results are accompanied by measurements of induced activities of the activation foils placed in the capsules behind the RPV at selected locations, namely in azimuthal profile. In case of discrepancies between the measured and calculated activities of the activation foils placed behind the pressure vessel, it is difficult to determine the source of the deviation. During such analysis, there arises a question on the influence of power peaking near core boundary on neutron profile behind the RPV.This paper compares the calculated and measured increase of the neutron flux density distribution behind the reactor pressure vessel in the azimuthal profile that has arisen from the replacement of 164 fuel pins located close to reactor internals by pins with the higher enrichment. This work can be understood as the first step in the characterization of the effect of incorrectly calculated pin power or burn-up in the fuel assembly at the core boundary relative to the neutron flux distribution behind reactor pressure vessel. Based on a good agreement between the calculated and experimental values, it can be concluded that the mathematical model used to evaluate the power increase is correct.

Applicability of LR-0 mock-up results to VVER-1000 reactor pressure vessel issues

Evaluation of neutron fluence in a reactor pressure vessel (RPV) together with surveillance specimen programs for RPV materials are one of the most important parts of in-service inspection programs that are necessary for realistic and reliable assessment of RPV residual lifetime. This paper covers transport of neutrons through the RPV of a VVER-1000 nuclear reactor. This problem is of increased importance as it concerns issues around VVER NPP life extension. With regards to the construction (reduced thickness of the lateral reflector), this issue plays greater role in VVER reactors than in Western types of PWR reactors.RPV material degradation depends mainly on neutron flux and spectra. Both quantities can be calculated or measured. This paper compares MCNP calculations and measurements on zero-power experimental reactor LR-0 with TORT calculations for VVER-1000. The goal is to find a reasonable method for precise estimation of neutron fluence and attenuation factor through the RPV. The calculations were performed with MCNP stochastic code and TORT deterministic transport code. The measurements were performed in a VVER-1000 mock-up placed in reactor LR-0 (Research Center Řež).

Neutron deep penetration through reactor pressure vessel and biological concrete shield of VVER-1000 Mock-Up in LR-0 reactor

Evaluation of the neutron fluence values in a reactor pressure vessel together with surveillance specimens programs from reactor pressure vessel materials are among the most important parts of in-service inspection programs that are necessary for realistic and reliable assessment of the RPV’s residual lifetime. The neutron fluence values are determined by means of calculation. These calculation results are accompanied by measurements of induced activity of the activation foil placed in capsules behind the reactor pressure vessel at selected locations. These results can be used for neutron flux density adjustment in the corresponding parts of the reactor system. Such information is not complete, because for determination of the radiation damage, the flux on the inner side of reactor pressure vessel has to be known. If the flux attenuation ratio is known, then the neutron flux density on the outer side can be easily recalculated for the inner side. These attenuation factors, applicable to the power reactors, can be determined experimentally in the VVER-1000 Mock-Up placed in the LR-0 reactor. The comparison of experimental results with calculated ones increases the reliability of their estimations. This paper aims to evaluate the measured neutron flux densities at various VVER-1000 Mock-Up positions by means of comparison with the Monte Carlo simulation. The effect of the various nuclear data libraries on the calculation results is also studied.

A plan for safety and integrity of research reactor components

Safety and integrity of a research reactor that has been operated over 40 years requires frequent and thorough inspection of all the safety-related components of the facility. The need of increasing the safety is the need of improving the reliability of its systems. Diligent and extensive planning of in-service inspection (ISI) of all reactor components has been imposed for satisfying the most stringent safety requirements. The Safeguards Officer's responsibilities of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code ASME Code have been applied. These represent the most extensive and time-consuming part of ISI program, and identify the components subjected to inspection and testing, methods of component classification, inspection and testing techniques, acceptance/rejection criteria, and the responsibilities. The paper focuses on ISI planning requirements for welded systems such as vessels, piping, valve bodies, pump casings, and control rod-housing parts. The weld in integral attachments for piping, pumps, and valves are considered too. These are taken in consideration of safety class (1, 2, 3, etc.), reactor age, and weld type. The parts involve in the frequency of inspection, the examination requirements for each inspection, the examination method are included. Moreover the flaw evaluation, the plan of repair or replacement, and the qualification of nondestructive examination personnel are considered.

Statistical Sampling for In-Service Inspection of Underground Storage Tanks (PVP2011-57011)

Savannah River Remediation, LLC (SRR) is implementing a statistical sampling strategy for in-service inspection (ISI) of liquid waste (LW) tanks at the United States Department of Energy’s Savannah River Site (SRS) in Aiken, South Carolina. As a component of SRS’s corrosion control program (CCP), the ISI program assesses tank wall structural integrity through the use of ultrasonic testing (UT). The statistical strategy for ISI is based on the random sampling of a number of vertically-oriented unit areas, called “strips,” within each tank. The number of strips to inspect was determined so as to attain, over time, a high probability of observing at least one strip of the worst 5% in terms of pitting and corrosion across all tanks. The hypergeometric distribution was used in the probability estimation to determine the number of strips for inspection. In addition to the strip sampling strategy, a single strip within each tank was identified to serve as the baseline for a longitudinal assessment of the tank safe operational life. Statistical tolerance limits for pit depth and corrosion rates were calculated using existing data and will be updated as additional inspections are made. The statistical tolerance limits provide conservative estimates for use in tank life projections. The statistical sampling strategy enables the ISI program to develop individual profiles of LW tank wall structural integrity that collectively provide a high confidence in storage tank safety and integrity over operational lifetimes.

원전 이종금속 맞대기용접부 PWSCC 균열건전성평가

가압경수로형 원전의 Alloy 600 원자로압력용기헤드 관통노즐 및 Alloy 82/182 이종금속 맞대기 용접부에서 일차수응력부식균열(PWSCC)이 보고된 이후 전 세계적으로 PWSCC 에 의한 용접부 파단을 예방하기 위해 강화검사를 적용하고 있다. 본 이종금속용접부에 대한 가동중검사에서 균열이 발견된 경우 건전성평가 결과가 도출되기까지 발전소가 정지 상태에 있게 됨에 따라 원전 이용율 저하가 발생할 수 있는데, 이를 예방하기 위해서는 균열건전성평가 관련 기술의 정립뿐만 아니라 신속하게 평가 결과를 도출할 수 있는 시스템의 구축이 필요하다. 본 연구에서는 이종금속 맞대기 용접부를 대상으로 진행하고 있는 PWSCC 균열건전성평가 기준 정립 및 전산 시스템 개발 결과를 제시하였다. 본 연구를 통해 이종금속 맞대기 용접부 PWSCC 균열건전성평가 기술이 정립되고 전산 시스템으로 구현되어 원자로압력용기 주변 이종금속 맞대기 용접부에서의 PWSCC 균열에 대한 기술적 건전성평가 수단을 확보하였다.

Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

Investigations of Mixing Zones subject to Thermal Fatigue

Further the through wall crack in an elbow of the Residual Heat Removal (RHR) system of Civaux 1 in 1998, a large investigation program has been completed on removed parts of pipes to better assess the factors responsible for thermal fatigue cracking in pipe components where hot and cold fluids mix together. This article presents the results of destructive investigations performed on numerous parts of the RHR system of EDF plants, on a nozzle of the primary system and on a nozzle of the secondary system and their contribution to understanding and assessing harmfulness of the phenomenon.The damage concerns only certain RHR mixing zones: the mixing tee at intersection between downstream the exchanger and the by-pass line and the mixing tee at location of intersection of two portions of minimum flow line. Deep cracks are always associated with high temperature differences and long times of operation. The deepest cracks are always located close to the welds, in areas showing geometrical singularities and minimal thickness. Crack networks in running parts concern only the very first millimeters of the inner wall.Other mixing zones have been investigated: destructive examinations of a charging line primary nozzle of the Chemical and Volume Control System (CVCS), of two tees between the main line and the minimum flow line of the RHR system and of one tee connecting the Feed Flow Control System (FFCS) to the Auxiliary Feedwater System (AFS) did not show any thermal fatigue crack.EDF has developed a method for identification of fluid mixing zones subject to thermal fatigue damage. This method is based on the assessment of a usage factor, using the knowledge acquired during a large R&D program. This program investigated the thermal loadings in mixing tees acquired through several thermohydraulic mockup tests and the austenitic steel behavior under high cycle thermo-mechanical loadings.The results of the destructive examinations have been used to strengthen the identification methodology of mixing zones potentially affected by thermal fatigue. Damage found in the destructive examinations are always correlated with usage factors a lot greater than 1.The methodology has been applied to all mixing zones of safety related circuits of plants operated by EDF. It has lead to identify the ones with a usage factor greater than unity and to an appropriate in-service inspection program based on volumic non-destructive examinations.

Case studies of risk-informed inservice inspection of nuclear piping systems

In the present paper, a probabilistic failure analysis is used to find failure probabilities of piping segments, and a probabilistic risk assessment model is employed to obtain risks to a nuclear power plant should these failures occur. The multiplication of the piping failure probability and the consequence for that particular failure results in the risk contribution of the pipe. The degrees of risk for different piping segments can then be ranked, and their results can be used as the basis for planning a risk-informed inservice inspection program. Numerical studies are offered with special emphases on: (1) the status and experience with RI-ISI applications in Taiwan; (2) the comparison of risk-rankings performed with three different methods developed in the US; (3) aspects of the probabilistic fracture mechanics calculation including the flaw size distributions and stress corrosion cracking model. The results indicate the proposed method can indeed be adopted for planning a cost effective inservice inspection program.

Leak Before Break application for Primary Coolant Loop and Surge Line of VVER-1000/320 Plant: TACIS Project R2.09/96

Activities were carried out within the frame of TACIS-96 Program—Project R2.09/96 named “LBB Applicability Review and Basic Implementation Engineering for Primary Coolant Loop & Surge Line of VVER-1000/320 Nuclear Power Plant”. Main objective of the Project was to perform a Leak Before Break assessment for the Main Coolant Pipes and Pressurizer Surge Line of the Reactor Coolant System of a selected “reference” Unit of a VVER-1000, type V-320 Nuclear Power Plant, which was assumed the Balakovo Unit 2. This has included the following: • definition of a “reference” methodology for primary piping systems with cladding; • verification of the Leak Before Break concept applicability to primary piping, in accordance with the “reference” methodology defined; • preparation of necessary materials data and validation files; • evaluation of effectiveness and reliability of the Leak Detection systems and the In-service Inspection Programs (methodologies and qualification); • preparation of a recommendations list to improve the existing information and diagnostic systems; • transfer of know–how from Consortium to Sub-Contractor, integrating the feedback of Western experience. One of the main highlights of the Project was the definition and execution of a comprehensive Material Testing Program in order to establish the actual tensile and fracture properties of the primary piping materials: base materials, weld metals, and “heat affected zones” were tested. Compact tests and standard specimens for J– R curves determination were used; special attention was paid to the dissimilar typical joints of the Surge Line pipes, being made by carbon steel base metal with stainless steel weld. Fracture properties as J– R curves were determined both at room and operating temperatures. The “reference” methodology defined has taken into account both Western and Russian Leak Before Break approaches. The analysis has shown that criteria for Leak Before Break concept applicability to the primary piping of Balakovo VVER-1000/320 Unit 2, presented in the reference methods developed in the frame of this Project, are met.

Risk-Informed Inservice Inspection Program for Ulchin Unit 4

Risk-Informed Inservice Inspection Program for Ulchin Unit 4

Database development and uncertainty treatment for estimating pipe failure rates and rupture frequencies

Estimates of failure rates for nuclear power plant piping systems are important inputs to Probabilistic Risk Assessments (PRA) and risk informed applications of PRA. Such estimates are needed for initiating event frequencies for Loss of Coolant Accidents and internal flooding events and for risk informed evaluations of piping system in-service inspection programs. A critical issue in the estimation of these parameters is the treatment of uncertainties, which can exceed an order of magnitude deviation from failure rate point estimates. Sources of uncertainty include failure data reporting issues, scarcity of data, poorly characterized component populations, and uncertainties about the physical characteristics of the failure mechanisms and root causes. A methodology for quantifying these uncertainties using a Bayes' uncertainty analysis method was developed for the EPRI risk informed in-service inspection program and significantly enhanced in subsequent applications. In parallel with these efforts, progress has been made in the development of pipe failure databases that contain the quantity and quality of information needed to support piping system reliability evaluations. Examples are used in this paper to identify technical issues with previous published estimates of pipe failure rates and the numerical impacts of these issues on the pipe failure rates and rupture frequencies are quantified.

ICONE11-36233 ISI Program in ETRR-2 Welded Parts in Form of Logic Flow Charts

A complete in-service inspection (ISI) program for nuclear research reactors requires several steps that have to be performed in sequence. These steps were described in many logic flow charts (LFC's). These logic flow charts might be separately used as a guide for future work in case of developing expert system software for research reactors. These charts included a general LFC describing the steps required to complete the ISI program. Details of this general LFC describing the steps required to complete the ISI program. Details of this general LFC are given in many modules. The next step of the LFC is to define the class of the component to be inspected. For each component class there are related modules that should be followed (examination requirements modules, examination methods modules, and flaw acceptance/rejection modules). If the component is accepted according to the requirements found in American Society of Mechanical Engineers (ASME) CODE section XI, the record and report requirement modules will be used to complete the ISI program. In the present work the inspection requirements for welded parts in the Second Egyptian Testing Research Reactor (ETRR-2) are considered. ETRR-2 is a pool- type reactor with an open water surface and variable core arrangement. Its core power is 22 MW_<th>, cooled and moderated by light water and with beryllium reflectors. It contains platetype fuel elements (MTR type, 19.7% enriched uranium) with aluminum clad. This reactor consists of 57 systems and around 200 subsystems, these systems contain many mechanical components such as; tanks, piper, valves, pumps, heat exchangers, cooling tower, air compressors and suppors.