ML20155C023

Compilation of Reports from Research Supported by the Electrical,Materials and Mechanical Engineering Branch, Division of Engineering

ML20155C023
Person / Time
Issue date:	10/31/1998
From:	Cayetano Santos NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:
References
NUREG-1426, NUREG-1426-V03, NUREG-1426-V3, NUDOCS 9811020047
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NUREG-1426 Vol. 3 Compilation of Reports From Research Supported by the Electrical, Materials, and Mechanical Engineering Branch, Division of Engineering 1994 - 1998 U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research I

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NUREG-1426 Vol. 3 Compilation of Reports From Research Supported by the Electrical, Materials, and Mechanical Engineering Branch, Division of Engineering 1994 - 1998 Manuscript Completed: August 1998 Date Published: October 1998 Compiled by C. Santos, Jr.

Electrical, Materials, and Mechanical Engineering Branch Division of Engineering OITice of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 f* **%

\ ..... /

I ABSTRACT Since 1965, the Materials Engineering Branch, Division of Engineering, of the This report provides as complete a Nuclear Regulatory Comission's Office listing as practical of formal of Nuclear Regulatory Research, and its technical reports submitted to the NRC predecessors dating back to the Atomic by the investigators working on these Energy Commission (AEC), has sponsored research programs. This listing research programs concerning the includes topical, final and progress integrity of the primary system reports, and is segmented by topic pressure boundary of light water area. In many cases a report will reactors. The components of concern in cover several topics (such as in the these research programs have included case of progress reports of multi-the reactor pressure vessel (RPV), faceted programs), but is listed under steam generators, and the piping. only one topic. Therefore, in These research programs have covered a searching for reports on a specific broad range of topics, including topic, other related topic areas should fracture mechanics analysis and be checked also. The separate volumes experimental work for RPV and piping of this report cover the following applications, inspection method periods:

development and qualification, and evaluation of irradiation effects to Volume 1: 1965 - 1990 RPV steels. Volume 2: 1991 - 1993 Volume 3: 1994 - 1998 iii NUREG-1426

Table of Contents ABSTRACT . . .. iii Introduction .. .

1 Advanced Reactors . . . .

3 Annealing . . .

. 3 Correlations . . . . . . . 4 Degradation of Mechanical Components .

. 5 Dosimetry . . . .

22 Electrical Systems .. .

25 EAC and Fatigue . .

29 Fracture Mechanics Testing and Analysis . . .

36 Non Destructive Examination . . .

. . 52 Piping . . . .. . .

57 Pressure Vessel Steels ... . . . . . 75 Radiation Embrittlement . . . .. .. . 80 Steam Generator Tube Integrity .. . . . . 96 Thermal Aging . . .. . . . 103 Underwater Welding . .. . . . . 107 v NUREG-1426

l Compilation of Reports - 1994-1998 Introduction Since 1965, the Materials Engineering Research Branch, Division Branch, Division of Engineering, of the of Reactor Safety Nuclear Regulatory Comission's Office Research, U.S. Nuclear of Nuclear Regulatory Research, and its Regulatory Comission predecessors dating back to the Atomic Energy Commission (AEC), has sponsored 1981-1986 Materials Engineering research programs concerning the Branch, Division of integrity of the primary system Engineering Technology, pressure boundary of light water U.S. Nuclear Regulatory reactors. The components of concern in Comission these research programs have included the reactor pressure vessel (RPV), 1986-1993 Materials Engineering steam generators, and the piping. Branch, Division of These research programs have covered a Engineering, broad range of topics, including U.S. Nuclear Regulatory fracture mechanics analysis and Comission experimental work for RPV and piping applications, inspection method 1995-1998 Electrical, Materials, and development and qualification, and Mechanical Engineering evaluation of irradiation effects to Branch, RPV steels.1 Division of Engineering, U.S. Nuclear Regulatory The branch sponsoring these research Commission programs has had various names and affiliations over the years, including This report provides as complete a the following: listing as practical of formal technical reports submitted to the NRC 1965-1973 Reactor Vessels Branch, by the investigators working on these Division of Reactor research programs. This listing Development and includes topical, final and progress Technology, U.S. Atomic reports, and is segmented by topic Energy Commission area. In many cases a report will cover several topics (such as in the 1973-1975 Metallurgy and Materials case of progress reports of multi-Research Branch, Division 1 faceted programs), but is listed under of Reactor Safety only one topic. Therefore, in Research, U.S. Atomic searching for reports on a specific Energy Commission topic, other related topic areas should I

be checked also.

1975-1981 Metallurgy and Materials 1 NUREG-1426

Compilation of Reports - 1994-1998 The separate volumes of this report cover the following periods:

Volume 1: 1965 - 1990 Volume 2: 1991 - 1993 Volume 3: 1994 - 1998 i

l l

l NUREG-1426 2

Compilation of Reports: 1994-1998 Advanced Reactors storage and handling system. In addition, the use of cobalt-based alloys in these plants was reviewed.

Title:

Review of the proposed The selected materials for both materials of construction for the SBWR reactors were generally sound, and no and AP600 advanced reactors major selection errors were found. It Author (s)/ Editor (s): Diercks. D.R. : was apparent that considerable thought Shack. W.J. ; Chung, H.M. : Kassner, had been given to the materials T.F. (Argonne National Lab., IL (United selection process, making use of States)) lessons learned from previous LWR Soonsorino Oroanization: NRC: Nuclear experience. The review resulted in the Regulatory Comission. Washington, OC suggestion of alternate an possibly (United States) better materials choices in a number of Publication Date: Jun 1994 cases, and several potential problem Reoort Number (s): NUREG/CR 6223: areas have been cited.

ANL--94/13 Order Number: TI94013716 Abstract: Two advanced light water AnneaHng reactor (LWR) concepts, namely the General Electric Simplified Boiling Water Reactor (SBWR) and the

Title:

Marble Hill Annealing Westinghouse Advanced Passive 600 Mwe Demonstration Evaluation Reactor (AP600), were reviewed in bathor(s)/ Editor (s): C.B. Oland, 8.R.

detail by Argonne National Laboratory. Bass. J.W. Bryson. L.J. Ott, J. A. ,

The objectives of these reviews were to Crabtree (Oak Ridge National (a) evaluate proposed advanced-reactor Laboratory) designs and the materials of Soonsorino Oroanization: NRC:

construction for the safety systems, Washington DC (United States)

(b) identify all aging and Publication Date: February 1998 environmentally related degradation Reoort Number (s): NUREG/CR-6552 mechanisms for the materials of Abstract: During the sumer of 1996, construction, and (c) evaluate from the an unirradiated reactor pressurc vessel safety viewpoint the suitability of the at the abandoned Marble Hill nuclear proposed materials for the design power plant was annealed to demonstrate appi1 cation. Safety related systems that existing technology can he used to selected for review for these two LWRs thermally anneal reactor pressure i included (a) reactor pressure vessel, vessels at commercial pressurized water (b) control rod drive system and reactor nuclear power plants in the reactor internals. (c) coolant pressure United States. Instrumentation boundary, (d) engineered safety installed on the reactor pressure systems. (e) steam generators (AP600 vessel and interfacing plant components only) (f) turbines, and (g) fuel provided evidence that the 3 NUREG-1426 i

Compilation of Reports - 1994-1998 demonstration was successful. An (United States)); Odette. G.R. : Mader, independent evaluation of engineering E.V. (California Univ. , Santa Barbara, issues associated with the annealing CA (United States))

demonstration was conducted at the Oak Soonsorino Oraanization: NRC: Nuclear Ridge National Laboratory for the Regulatory Comission, Washington, DC Nuclear Regulatory Comission. (United States)

Temperature, strain, and displacement Publication Date: May 1995 data acquired during the annealing Reoort Number (s): NUREG/CR-6327:

demonstration were used to verify MCS--950302 thermal and structural analysis Order Number: TI95011539 results. Based on findings and Abstract: The reactor pressure vessel observations from the annealing (RPV) surrounding the core of a demonstration and results of thermal comercial nuclear power plant is and structural analysis, an subject to embrittlement due to instrumentation system was developcd exposure to high energy neutrons, The for use in assessing tnermal annealing effects of irradiation embrittlement at other nuclear power plants similar can be reduced by thermal annealing at to Marble Hill. The objective of the temperatures higher than the normal instrumentation system is to provide operating conditions. However, a means sufficient data for determining if the of quantitatively assessing the observed time and temperature prafile- effectiveness of annealing for satisfies or exceeds the required embrittlement recosery is needed. The thermal annealing conditions, and for objective of this work was to analyze verifying thermal and structural the pertinent data on this issue and analysis results. Development of the develop quantitative models for instrumentation system involved estimating the recovery in 30 ft-lb (41 consideration of technical requirements J) Charpy transition temperature and as well-as issues related to minimizing Charpy upper shelf energy due to ,

occupational exposure to radiation in annealing. Data were gathered from the  !

accordance with the "as low as is Test Reactor Embrittlement Data Base reasonably achievable" principle. and from various annealing reports. An i analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work.

Correlations Independent variables considered in the analysis included material chemistries.

annealing time and temperature,

Title:

Models for embrittlement irradiation time and temperature, recovery due to annealing of reactor fluence, and flux. To identify pressure vessel steels important variables and functional Author (s)/ Editor (s): Eason. E.D. :

forms for predicting embrittlement Wright, J.E. : Nelson. E.E. (Modeling recovery, advanced statistical and Computing Services Boulder, C0 NUREG-1426 4 4

Compilation of Reports: 1994-1998 techniques, including pattern Publication Date: Mar 1994 recognition and transformation Reoort Number (s): NUREG/CR 5314-Vol.5:

analysis, were applied together with EGG- 2562-Vol.5 current understanding of the mechanisms Order Number: TI94009041 governing embrittlement and recovery. Abstract: This report evaluates the Models were calibrated using available technical information and multivariable surface-fitting field experience related to management techniques. Several iterations of of aging damage to light water reactor model calibration, evaluation with metal containments. A generic aging respect to mechanistic and statisti9al management approach is suggested for considerations, and comparison with the the effective and comprehensive aging trends in hardness data produced management of metal containments to i correlation models for estimating ensure their safe operation. The major Charpy upper shelf energy and concern is corrosion of the embedded transition temperature after portion of the containment vessel and irradiation and annealing. This work detection of this damage. The provides a clear demonstration that (1) electromagnetic acoustic transducer and microhardness recovery is generally a half-cell potential measurement are very good surrogate for shift recovery, potential techniques to detect and (2) there is a high level of corrosion damage in the embedded consistency between the observed portion of the containment vessel.

annealing trends and fundamental models Other corrosion-related concerns of embrittlement and recovery include inspection of corrosion damage processes. on the inaccessible side of BWR Mark I and Mark II containment vessels and corrosion of the BWR Mark I torus and Degradation of Hechanical emergency core cooling system piping that penetrates the torus, and Components transgranular stress corrosion cracking of the penetration bellows.

Fatigue-related concerns include Title- Insights for aging management reduction in the fatigue life (a) of a of light water reactor components: vessel caused by roughness of the Metal containments corroded vessv surface and (b) of

Author (s)/ Editor (s)

Shah, V.N. - bellows be'.ause o' any physical damage.

Sinha, U.P. (EG and G Idaho, Inc . Maintenance of sur' ace coatings and Idaho Falls. ID (United States)): sealant at the meta - concrete Smith, S.K. (0gden Environmental and inter. face is the best protection Energy Services, Southfield, MI (United against corrosion of the vessel.

States))

Soonsorina Oroanization: NRC: Nuclear ,

Regulatory Commission. Washington, DC

Title:

The effects of age on nuclear (United States) 5 NUREG-1426 l

l l

l

-- -- . . ~_ .

Compilation of Reports - 1994 1998 power plant containment cooling systems unavailability analysis was performed Author (s)/ Editor (s): Lofaro, R. : to examine the potential effects of Subudhi, M. . Travis. R. : DiBiasio, A. aging by increasing failure rates for

Azarm A. (Brookhaven National Lab. , selected components. A commonly found Upton, NY (United States)): Davis, J. containment spray system design and a (Science Applications International comonly found fan cooler system design ,

l Corp., New York, NY (United States)) were modeled. Parametric failure rates Soonsorina Graanization: NRC: Nuclear for those components in each system Regulatory Commission, Washington. DC that could be subject to aging were (United States) accounted for in the model to simulate Publication Date: Apr 1994 the time- dependent effects of aging Reoort Number (s): NUREG/CR-5939: degradation, assuming no provisions are BNL-NUREG--52345 made to properly manage it. System Order Number: TI94011190 unavailability as a function of Abstract: A study was performed to increasing component failure rates was assess the effects of aging on the then calculated, performance and availability of containment cooling systems in US commercial nuclear power plants. This

Title:

Development and application of study is part of the Nuclear Plant degradation modeling to define Aging Research (NPAR) program sponsored maintenance practices by the US Nuclear Regulatory Author (s)/ Editor (s): Stock, D. :

Comission. The objectives of this Samanta. P. (Brookhaven National Lab.,

program are to provide an understanding Upton. NY (United States)): Vesely, W.

of the aging process and how it affects (Science Applications International plant safety so that it can be properly Corp., Dublin 0H (United States))

managed. This is one of a number of Soonsorina Orcanization: NRC: Nuclear studies performed under the NPAR Regulatory Commission, Washington, DC program which provide a technical basis (United States) for the identification and evaluation Publication Date: Jun 1994 of degradation caused by age. The Recort Number (s): NUREG/CR-5967:

effects of age were characterized for BNL-NUREG -52353 the containment cooling system by Order Number: TI94013791 reviewing and analyzing failure data Abstract: This report presents the from national databases, as well as development and application of plant-specific data. The predominant component degradation modeling to failure causes and aging mechanisms analyze degradation effects on were ident:fied, along with the reliability and to identify aspects of components that failed most frequently, maintenance practices that mitigate Current inspection, surveillance, and degradation and aging effects. Using monitoring practices were also continuous time Markov approaches, a examined. A containment cooling system component degradation model is NUREG-1426 6

Compilation of Reports: 1994 1998 discussed that includes information program sponsored by the US Nuclear about degradation and maintenance. The Regulatory Commission. The objectives component model commonly used in of this program are to provide an probabilistic risk assessments is a understanding of the aging process and simple case of this general model. The how it affects plant safety so that it parameters used in the general model can be properly managed. This is one have engineering interpretations and of a numbar of studies performed under can be estimated using data and the NPAR program which provide a engineering experience. The generation technical basis for the identification of equations for specific models, the and evaluation of degradation caused by solution of these equations, and a age. The failure data from national methodology for estimating the needed databases, as well as plut specific parameters are all discussed. data were reviewed and analyzed to Applications in this report show how understand the effects of aging on the these models can be used to RCIC system. This analysis identified quantitatively assess the benefits that important components that should are expected from maintaining a receive the highest priority in terms component, the effects of different of aging management. The aging maintenance efficiencies, the merits of characterization provided information different maintenance policies, and the on the effects of aging on component interaction of surveillance test failure frequency, failure modes, and intervals with maintenance practices. failures causes. Current inspection, surveillance, and monitoring practices were also reviewed.

Title- The effects of aging on BWR core isolation cooling systems

Title:

The effects of age on nuclear Author (s)/ Editor (s): Lee B.S. power plant containment cooling systems (Brookhaven National Lab., Upton, NY Author (s)/ Editor (s): Iofaro, R. :

(United States)) Subudhi. M. : Travis, R , D181asio, A.

Soonsorino Oraanization: NRC: Nuclear  : Azarm. A. (Brookhaven National Lab.,

Regulatory Commission Washington DC Upton, NY (United States)): Davis, J.

(United States) (Science Applications International Publication Date: Oct 1994 Corp., New York. NY (United States))

Reoort Number (s): NUREG/CR-6087: Soonsorino Oraanization: NRC: Nuclear BNL-NUREG--52390 Regulatory Comission. Washington, DC Order Number: T195002269 (United States)

Abstract: A study was performed to Publication Date: Apr 1994 assess the effects of aging on the Reoort Number (s): NUREG/CR-5939:

Reactor Core Isolation Cooling (RCIC) BNL-NUREG--52345 system in commercial Boiling Water Order Number: T194011190 Reactors (BWRs). This study is part of Abstract: A study was performed to the Nuclear Plant Aging Research (NPAR) assess the effects of aging on the 7 NUREG-1426 f =

[ 1 L

Compilation of Reports - 1994-1998 performance and availability of containment cooling systems in US

Title:

Valve actuator motor commercial nuclear power plants. This degradation study is part of the Nuclear Plant Author (s)/ Editor (s): Kueck, J.D. (0ak Aging Research (NPAR) program sponsored Ridge National Lab. , TN (United by the US Nuclear Regulatory States))

Commission. The objectives of this Soonsorino Orcanization: NRC: Nuclear program are to provide an understanding Regulatory Commission, Washington DC of the aging process and how it affects (United States) plant safety so that it can be properly Publication Date:- Dec 1994 managed. This is one of a number of Reoort Number (s): NUREG/CR-6205:

studies performed under the NPAR ORNL--6796 program which provide a technical basis Order Number: TI95004904 for the identification and evaluation Abstract: Valve actuator motor of degradation caused by age. The degradation and failure has been a effects of age were characterized for significant, but little studied, the containment cooling system by problem in the nuclear industry. This reviewing and analyzing failure data study provides a discussion of the from national databases, as well as primary failure mode --thermal plant-specific data. The predominant degradation-- and reviews the basis for failure causes and aging mechanisms the solution to thermal degradation --

were identified, along with the thermal protection. The study also components that failed most frequently. provides reviews of various industry Current inspection, surveillance, and data bases, discusses effects of other monitoring practices were also failure modes such as corrosion, and examined. A containment cooling system provides a review of other unavailability analysis was performed considerations the user should to examine the potential effects of entertain when assessing thermal aging by increasing failure rates for protection, selected components. A commonly found containment spray system design and a commonly found fan cooler system design

Title:

Aging and service wear of were modeled. Parametric failure rates spring-loaded pressure relief valves for those components in each system used in safety-related systems at that could be subject to aging were nuclear power plants accounted for in the model to simulate Author (s)/ Editor (s): Staunton, R.H. ;

the time- dependent effects of aging Cox, D.F. (Oak Ridge National Lab., TN degradation, assuming no provisions are (United States))

made to properly manage it. System soonsorina Oraanization: NRC: Nuclear unavailability as a function of Regulatory Comission, Washington, DC increasing component failure rates was (United States) then calculated. Publication Date: Mar 1995 NUREG-1426 8 4

Compilation of Reports: 1994-1998 Reoort Number (s): NUREG/CR-6192: Abstract: The purpose of high pressure ORNL--6791 injection systems is to maintain an Order Number: TI95009655 adequate coolant level in reactor Abstract: Spring-loaded pressure pressure vessels, so that the fuel relief valves (PRVS) are used in some cladding temperature does not exceed safety- related applications at 1,200[ degrees]C (2.200[ degrees]F), and nuclear power plants. In general, they to permit plant shutdown during a are used in systems where, during variety of design basis loss of-coolant accidents, pressures may rise to levels accidents. This report presents the where pressure safety relief is results of a study on aging performed required for protection of personnel, for high pressure injection systems of system piping, and components. This boiling water reactor plants in the report documents a study of PRV aging United States. The purpose of the and considers the severity and causes study was to identify and evaluate the of service wear and how it is effects of aging and the effectiveness discovered and corrected in various of testing and maintenance in detecting systems valve sizes, etc. Provided in and mitigating aging degradation.

this report are results of the Guidelines from the United States examination of the recorded failures Nuclear Regulatory Conmission's Nuclear ari identification of trends and Plant Aging Research Program were used relationships / correlations in the in performing the aging study. Review failures when all failure-related and analysis of the failures reported parameters are considered. Components in databases such as Nuclear Power that comprise a typical PRV, how those Experience. Licensee Event Reports, and components fail, when they fail, and the Nuclear Plant Reliability Data the current testing frequencies and System, along with plant-specific methods are also presented in detail. maintenance records databases, are included in this report to provide the information required to identify aging

Title:

Aging study of boiling water stressors, failure modes, and failure reactor high pressure injection systems causes. Several probabilistic risk Author (s)/ Editor (s): Conley 0.A. : assessments were reviewed to identify Edson, J.L. : Fineman, C.F. (Lockheed risk-significant components in high Idaho Technologies Co., Idaho Falls ID pressure injection systems. Testing.

(United States)) maintenance, specific safety issues, Soonsorino Oraanization: NRC: Nuclear and codes and standards are also Regulatory Commission. Washington, DC discussed.

(United States)

Publication Date: Mar 1995 j

Reoort Number (s): NUREG/CR-5462:

Title:

Effect of aging on the PWR INEL--94/0090 Chemical and Volume Control System Order Number: TI95009514 Author (s)/ Editor (s): Grove. E.J. .

9 NUREG-1426

Compilation of Reports - 1994-1998 Travis, R.J. : Aggarwal, S.K. obtain specific information on system (Brookhaven National Lab., Upton, NY inspection, surveillance, monitoring, (United States)) and inspection practices. The results Soonsorina Oraanization: NRC: Nuclear of these visits indicate that adequate Regulatory Commission Washington, DC system maintenance and inspection is (United States) being performed. In some instances, Publication Date: Jun 1995 the frequencies of inspection were Beoort Number (s): NUREG/CR-5954:

increase in response to repeated BNL-NUREG- 52410 failure events. A parametric study was Order Number: TI95014434 performed to assess the effect of Abstract: The PWR Chemical and Volume system aging on Core Damage Frequency Control System (CVCS) is designed to (COF). This study showed that as provide both safety and non-safety motor-operated valve (MOV) operating related functions. During normal plant failures increased, the contribution of operation it is used to control reactor the High Pressure Injection to CDF also coolant chemistry, and letdown and increased, charging flow. In many plants, the charging pumps also provide high pressure injection, emergency boration.

Title:

Aging of turbine drives for and RCP seal injection in emergency safety-related pumps in nuclear power situations. This study examines the plants design, materials, maintenance, Author (s)/ Editor (s): Cox, D.F, (Oak operation and actual degradation Ridge National Lab., TN (United experiences of the system and main States))

sub-components to assess the potential Spqnprina Oraanization: NRC; Nucle "

for age degradation. A detailed review Regulatory Commission, Washington, DC of the Nuclear Plant Reliability Data (United States)

System (NPRDS) and Licensee Event Publication Date: Jun 1995 Report (LER) databases for the Report Number (s): NUREG/CR-5857; 1988--1991 time period, together with a ORNL--6713 review of industry and NRC experience Order Number: TI95014753 and research, indicate that age- Abstract: This study was performed to related degradations and failures have examine the relationship between time-occurred. These failures had dependent degradation and current significant effects on plant operation, industry practices in the areas of including reactivity excursions, and maintenance, surveillance, and pressurizer level transients. The operation of steam turbine drives for majority of these component failures safety-related pumps. These pumps are resulted in leakage of reactor coolant located in the Auxiliary Feedwater outside the containment. A (AFW) system for pressurized- water representative plant of each PWR design reactor plants and in the Reactor Core (W, CE, and B and W) was visited to Isolation Cooling and High-Pressure NUREG-1426 10

l Compilation of Reports: 1994-1998 Coolant Injection systems for typical accelerometer systems. Volume 2 boiling-water reactor plants. This Author (s)/ Editor (s): Goodenow, T.C. ,

research has been conducted by Shipman, R.L. . Holland. H.M. (Epoch exanination of failure data in the Engineering Inc., Gaithersburg. MD Nuclear Plant Reliability Data System. (United States))

review of Licensee Event Reports, Soonsorino Orcanization: NRC: Nuclear discussion of problems with operating Regulatory Comnission Washington, DC plant personnel, and personal (United States) observation. The reported failure data Publication Date_;. Jun 1995 were reviewed to determine the cause of Reoort Number (s): NUREG/CR-6313-Vol.2 the event and the method of discovery. Order Number: TI95015065 Based on the research results, attempts Abstract- Epoch Engineering, have been made to determine the Incorporated (EEI) has completed a predictability of failures and possible series of vibration measurements preventive measures that may be comparing their newly-developed Robust implemented. Findings in a recent Laser Interferometer (RLI) with study of AFW systems indicate that the accelerometer-based instrumentation turbine drive is the single largest systems. EEI has successfully contributor to AFW system degradation, demonstrated, on several pieces of However, examination of the data shows commonplace machinery, that non-that the turbine itself is a reliable contact, line-of-sight measurements are piece of equipment with a good' service practical and yield results equal to record. Most of the problems or, in some cases, better than documented are the result of problems customary field implementations of with the turbine controls and the accelerometers. The demonstration mechanical overspeed trip mechanism; included analysis and comparison of these apparently stem from three major such phenomena as nonlinearity, causes which are discussed in the text, transverse sensitivity, harmonics, and Recent improvements in maintenance signal-to-noise ratio. Fast Fourier practices and procedures, combined with Transformations were performed on the a stabilization of the design, have led accelerometer and the laser system to improved performance resulting in a outputs to provide a comparison basis.

reliable safety-related component.

~

The RLI was demonstrated, within the However, these improvements have not limits o the task, to be a viable, been universally implemented, line-of-sight, non-contact alternative to accelerometer systems. Several different kinds of machinery were Tit 10 R,obust, accurate, and instrumented and. compared, including a non-contacting vibration measurement small pump, a gear-driven cement mixer, systems: Supplemental appendices a rotor kit, and two small fans. Known presenting comparison measurements of machinery vibration sources were the robust laser interferometer and verified and RLI system output file 11 NUREG-1426 i

w

Compilation of Reports - 1994 1998 formats were verified to be compatible such phenomena as nonlinearity, with comercial computer programs used transverse sensitivity, harmonics, and for vibration monitoring and trend signal-to-noise ratio. Fast Fourier analysis. The RLI was also observed to Transformations were performed on the be less subject to electromagnetic accelerometer and the laser system interference (EMI) and more capable at output.s to provide a comparison basis.

very low frequencies. This document. The RLI was demonstrated, within the Volume 2. provides the appendices to limits of the task, to be a viable.

this report. line-of-sight, non-contact alternative to accelerometer systems. Several different kinds of machinery were

Title:

Robust, accurate, and instrumented and compared including a non-contacting vibration measurement small pump, a gear-driven cement mixer, systems: Summary of comparison a rotor kit, and two small fans. Known measurements of the robust laser machinery vibration sources were interferometer and typical verified and RLI system output file accelerometer systems. Volume 1 formats were verified to be compatible Author (s)/ Editor (s): Goodenow. T.C. . with commercial computer programs used Shipman R.L. . Holland. H.M. (Epoch for vibration monitoring and trend Engineering. Inc. Gaithersburg MD analysis. The RLI was also observed to (United States)) be less subject to electromagnetic Soonsorina Oraanization: NRC: Nuclear interference (EMI) and more capable at Regulatory Comission. Washington, DC very low frequencies.

(United States)

Publication Date: Jun 1995 Reoort Number (s): NUREG/CR-6313-Vol.1

Title:

Detection of pump degradation Order Number: TI95015064 Author (s)/ Editor (s): Greene. R.H. :

Abstract: Epoch Engineering. Casada. D.A. . Ayers. C.W. (and others)

Incorporated (EEI) has completed a Soonsorino Oroanization: NRC: Nuclear series of vibration measurements Regulatory Commission. Washington, DC comparing their newly-developed Robust (United States)

Laser Interferometer (RLI) with Publication Date: Aug 1995 accelerometer-based instrumentation Reoort Number (s): NUREG/CR-6089 systems. EEI has successfully Order Number: TI95017245 demonstrated, on several pieces of Abstract: This Phase II Nuclear Plant comonplace machinery, that non- Aging Research study examines the contact line-of-sight measurements are methods of detecting pump degradation practical and yield results equal to that are currently employed in domestic or, in some cases, better than and overseas nuclear facilities. This customary field implementations of report evaluates the criteria mandated accelerometers. The demonstration by required pump testing at U.S.

included analysis and comparison of nuclear power plants and compares them NUREG-1426 12

1 I

Compilation of Reports: 1994 1998 to those features characteristic of hydraulically unstable operation for a state of-the art diagnostic programs particular pump and motor set. The and practices currently implemented by concept of using motor current or power other major industries. Since the fluctuations as an indicator of pump working condition of the pump driver is hydraulic load stability is presented.

crucial to pump operability, a brief review of new applications of motor diagnostics is provided that highlights

Title:

Fire modeling of the Heiss recent developments in this technology. Dampf Reaktor containment The routine collection and analysis of Author (s)/ Editor (s): Nicolette. V.F.

spectral data is superior to all other (Sandia National Labs.. Albuquerque. NM technologies in its ability to (United States)); Yang, K.T. (Notre accurately detect numerous types and Dame Univ., IN (United States))

causes of pump degradation. Existing Soonsorino Oroanization: NRC: Nuclear ASME Code testing criteria do not Regulatory Comission. Washington. DC require the evaluation of pump (United States) vibration spectra but instead overall Publication Date: Sep 1995 vibration amplitude. The mechanical Reoort Number (s): NUREG/CR 6017:

information discernible from vibration SAND--93-0528 amplitude analysis is limited and Order Number: T!96001168 several cases of pump failure were not Abstract: This report summarizes detected in their early stages by Sandia National Laboratories' vibration monitoring. Since spectral participation in the fire modeling analysis can provide a wealth of activities for the German Heiss Dampf pertinent information concerning the Reaktor (HDR) containment building, mechanical condition of rotating under the sponsorship of the United machinery, its incorporation into ASME States Nuclear Regulatory Commission, testing criteria could merit a The purpose of this report is twofold:

relaxation in the monthly-to-quarterly (1) to summarize Sandia's participation testing schedules that seek to verify in the HDR fire modeling efforts and >

and assure pump operability. Pump (2) to summarize the results of the drivers are not included in the current international fire modeling community battery of testing. Operational involved in modeling the HDR fire problems thought to be caused by pump tests. Additional comments on the degradation were found to be the result state of fire modeling and trends in of motor degradation. Recent advances the international fire modeling in nonintrusive monitoring techniques comunity are also included. It is have made motor diagnostics a viable noted that, although the trend technology for assessing motor internationally in fire modeling is operability. Motor current / power toward the development of the more analysis can detect rotor bar complex fire field models, each type of degradation and ascertain ranges of fire model has something to contribute 13 NUREG-1426

Compilation of Reports - 1994-1998 to the understanding of fires in closing correlation to include low-nuclear power plants. flow and low-pressure loads. The report also includes a general j approach, presented in step-by-step  ;

Title:

Gata valve and motor-operator format, for determining operating  ;

research findings margins for rising-stem valves (gate )

Author (s)/ Editor (s): Steele. R. Jr. : valves and globe valves) as well as DeWall . K.G. : Watkins. J.C. . Russell, quarter-turn valves (ball valves and M.J. : Bramwell, D. butterfly valves).

Soonsorino Oroanization: NRC; Nuclear Regulatory Comission. Washington, DC (United States)

Title:

A sumary of the Fire Testing Publication Date: Sep 1995 Program at the German HDR Test Facility Report Number (s): NUREG/CR-6100: Author (s)/ Editor (s): Nowlen, S.P.

INEL--94/0156 (Sandia National Labs., Albuquerque NM Order Number: TI96000305 (United States))

Abstract: This report provides an Soonsorino Oroanization: NRC; Nuclear update on the valve research being Regulatory Commission Washington, DC sponsored by the US Nuclear Regulatory (United States)

Commission (NRC) and conducted at the Publication Date: Nov 1995 Idaho National Engineering Laboratory Reoort Number (s): NUREG/CR-6173:

(INEL). The research addresses the SAND--94-1795 need to provide. assurance that Order Number: TI96002834 motor-operated valves can perform their Abstract: This report provides an intended safety function, usually to overview of the fire safety experiments open or close against specified (design performed under the sponsorship of the basis) flow and pressure loads. This German government in the containment report describes several important building of the decommissioned pilot developments: Two methods for nuclear power plant known as HDR. This estimating or bounding the design basis structure is a highly complex, stem factor (in rising-stem valves), multi-compartment, multi-level building using data from tests less severe than which has been used as the test bed for design basis tests; a new correlation a wide range of nuclear power plant for evaluating the opening responses of operation safety experiments. These gate valves and for predicting opening experiments have included numerous fire requirements; an extrapolation method tests. Test fire fuel sources have that uses the results of a best effort included gas burners, wood cribs, oil flow test to estimate the design basis pools, nozzle release oil fires, and closing requirements of a gate valve cable in cable trays. A wide range of that exhibits atypical responses (peak ventilation conditions including full force occurs before flow isolation): natural ventilation, full forced and the extension of the original INEL ventilation, and combined natural and NUREG-1426 14 l

Compilation of Reports: 1994 1993 forced ventilation have been evaluated, design basis tests; a new correlation During most of the tests, the fire for evaluating the opening responses of products mixed freely with the full gate valves and for predicting opening containment volume. Macro- scale requirements: an extrapolation method building circulation patterns which that uses the results of a best effort were very sensitive to such factors as flow' test to estimate the design basis ventilation configuration were observed closing requirements of a gate valve and characterized. Testing also that exhibits atypical responses (peak-included the evaluation of selective force occurs before flow isolation);

area pressurization schemes as a means and the extension of the original INEL of smoke control for emergency access closing correlation to include low-and evacuation stairwells. flow and low pressure loads. The report also includes a general approach, presented in step by-step Title- Gate valve and motor-operator format, for determining operating research findings margins for rising-stem valves (gate e.gt.tpr(s)/ Editor (s): Steele, R. Jr. , valves and globe valves) as well as DeWall, K.G. : Watkins, J.C. : Russell, quarter-turn valves (ball valves and M.J. : Bramwell, D. butterfly valves).

Soonsorino Oroanization: NRC: Nuclear Regulatory Commission, Washington, DC (United States) Title; Aging assessment of surge Publication Date: Sep 1995 protective devices in nuclear power Reoort Number (s): NUREG/CR-6100: plants INEL--94/0156 Author (s)/ Editor (s): Davis, J.F. .

Order Number: TI96000305 Subudhi, M. (Brookhaven National Lab.,

Abstract: This report provides an Upton, NY (United States)); Carroll, update on the valve research being D.P. (Florida Univ. , Gainesville, FL sponsored by the US Nuclear Regulatory (United States))

Comission (NRC) and conducted at the Soonsorino Oroanization: NRC: Nuclear Idaho National Engineering Laboratory Regulatory Commission, Washington, DC (INEL). The research addresses the (United States) need to provide assurance that Publication Date: Jan 1996 motor-operated valves can perform their Reoort Number (s): NUREG/CR-6340:

intended safety function, usually to BNL-NUREG--S2463 open or close against specified (design Order Number: T196006194 basis) flow and pressure loads. This Abstract: An assessment was performed report describes several important to determine the effects of aging on developments: Two methods for the performance and availability of estimating or bounding the design basis surge protective devices (SPDs), used stem factor (in rising stem valves), in electrical power and control systems using data from tests less severe than in nuclear power plants. Although SPDs 15 NUREG-1426

Compilation of Reports - 1994-1993 have not been classified as Order Number: TI96009348 Abstract: Question for resolution of.

' ~

safety-related they are risk-important because they can minimize the Generic Safety Issue No. 24 is whether initiating event frequencies associated or not PWRs that currently rely on a-with loss of offsite power and reactor manual system W ECCS switchover to trips. Conversely, their failure due recirculation would be required to to age might cause some of those install an automatic system. Risk initiating events, e.g., through short estimates are obtained by reevaluating

- circuit failure modes, or by allowing the contributions to core damage deterioration of the safety-related frequencies (CDFs) associated with component (s) they are protecting from failures of manual and semiautomatic

overvoltages, perhaps preventing a switchover at a representative PWR, reactor trip. from an open circuit This study considers each separate failure mode. From the data evaluated instruction of the corresponding during 1980--1994, it was found that emergency operating procedures (E0Ps),

failures of surge arresters and the mechanism for each control, and the suppressers by short circuits were relation of each control to its neither a significant risk nor safety neighbors. Important contributions to concern, and there were no failures of C0F include human errors that result in L surge suppressers preventing a reactor completely coupled failure of both i trip. Simulations, using the trains and failure to enter the Electromagnetic Transients Program required E0P. It is found that (EMTP) were performed to determine the changeover to a semiautomatic system is adequacy of high voltage surge not justified on the basis of arresters. cost-benefit analysis: going from a manual to a semiautomatic system reduces the CDF by 1.7 [ times) 10[sup

Title:

Estimated net value and [minus]5]perreactor- year, but the uncertainty for automating ECCS probability that the net cost of the modification being less than 51, 000

'switchover at PWRs Author (s)/ Editor (s): Walsh, B. : per person-rem is about 20% without Brideau, J. : Comes, L. : Darby, J. ; license renewal. Scoping analyses.

Guttmann H. : Sciacca. F. : Souto. F. using optimist assumptions, were 4  : Thomas, W. : Zigler, G. (Science and performed for a changeover to a

, Engineering Associates, Inc., semiautomatic system with automatic Albuquerque, NM (United States)). actuation and to a fully automatic Soonsorina Oraanization: NRC: Nuclear system: in these cases the probability Regulatory Connission. Washington, DC of a net cost being less than (United States) $1.000/ person-rem is about 50% without Publication Date: Feb 1996 license renewal and over 95% with

. Reoort Number (s): NUREG/CR-6432: license renewal.

SEASF-OR--94-001-NUREG-1426 16 1

1

____________ m _ _ _ _ _ . - _ _- _. , . - - - - . . .--_____- . .__m_____ _ _ _ _ _ _ _ _ _ _ - , _ -

l Compilation of Reports: 1994-1998

Title:

Applications of reliability quantitative change in component degradation analysis unavailability when no maintenance is Author (s)/ Editor (s): Vesely. W.E. performed. Assessment of these impacts (Science Applications International are important since they measure the Corp.. Dublin. OH (United States)); reliability and risk impacts of Samanta. P.K. (Brookhaven National maintenance and can be fed back to the Lab., Upton, NY (United States)) maintenance program to improve its Soonsorino Oroanization; NRC: Nuclear effectiveness.

Regulatory Comission. Washingte n. DC (United States)

Publication Date: Feb 1996

Title:

Aging of safety class 1E Reoort Number (s): NUREG/CR 6415: transformers in safety systems of BNL-NUREG--52488 nuclear power plants Order Number: TI96006221 Author (s)/ Editor (s): Roberts. E.W. .

Abstract: Reliability degradation Edson, J.L. ; Udy. A.C. (Lockheed Idaho analysis is the analysis of the Technologies Co., Idaho Falls. 10 occurrences of degradations and the (United States))

times of maintenance to determine their Soonsorino Oroanization: NRC: Nuclear reliability and risk implications. A Regulatory Comission. Washington, DC program is presented for applying (United States) reliability degradation analyses to Publication Date: Feb 1996 maintenance data collected at nuclear Reoort Number (s): NUREG/CR-5753; power plants. As a specific part of INEL--95/0573 the program, time trending of Order Number: TI96006598 maintenance data is illustrated. Abstract: This report discusses aging Maintenance data on residual heat effects on safety-related power removal (RHR) pumps and service water transformers in nuclear power plants.

(SW) pumps at selected boiling water It also evaluates maintenance testing, reactor (BWR) plants are evaluated to and monitoring practices with respect show how trends in maintenance data, to their effectiveness in detecting and which generally do not involve mitigating the effects of agirig. The failures. can be used to understand study follows the US Nuclear Regulatory effectiveness of maintenance. These Commission's (NRC's) Nuclear trends also are translated to specific Plant-Aging Research approach. It impacts on pump unavailability and on investigates the materials used in core-damage frequency (assuming that transformer construction, identifies the trends in failure rate are the same stressors and aging mechanisms, as those observed for degradation presents operating and testing rate). The second application shows experience with aging effects, analyzes the use of reliability degradation transformer failure events reported in analysis to quantitatively evaluate the various databases, and evaluates effect of maintenance. i.e.. the maintenance practices. Databases 17 NUREG-1426 l

Compilation of Reports - 1994-1998 maintained by the nuclear industry were provided information on the effects of andlyzed to evaluate the effects of aging on component failure frequency, aging on the operation of nuclear power failure modes, and failure causes.

plants. Current inspection surveillance, and monitoring practices were also reviewed.

Title:

Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

Title:

Effects of aging and service Author (s)/ Editor (s): Lee, B.S. . wear on main steam isolation valves and Travis. R. ; Grove. E. . DiBiasio. A. valve operators Soonsorino Oroanization: NRC: Nuclear Author (s)/ Editor (s): Clark. R.L. (0ak Regulatory Commission. Washington, DC Ridge National Lab., TN (United (United States) States))

Publication Date: Mar 1996 Soonsorino Oroanization: NRC: Nuclear Reoort Nunber(s): NUREG/CR-6339: Regulatory Comission. Washington, DC BNL-NUREG- 52462 (United States)

Order Number: TI96008079 Publication Date: Mar 1996 l Abstract: A study was performed to Reoort Number (s): NUREG/CR-6246:

assess the effects of aging on the ORNL--6814 Containment Isolation (CI) functions of Order Number: TI96008272 Westinghouse Pressurized Water Reactors Abstract: In recent years main steam and General Electric Boiling Water isolation valve (MSIV operating l

Reactors. This study is part of the problems have resulted in significant l

! Nuclear Plant Aging Research (NPAR) operational transients (e.g. , spurious l program, sponsored by the U.S. Nuclear reactor trips, steam generator dry out.

Regulatory Comission. The objectives excessive valve seat leakage).

of this program are to provide an increased cost, and decreased plant understanding of the aging process and availability. A key ingredient to an how it affects plant safety so that it engineering-oriented reliability can be properly managed. This is one improvement effort is a thorough of a number of studies performed under understanding of relevant historical l

the NPAR program which provide a experience. A detailed review of technical basis for the identification historical failure data available and evaluation of degradation caused by through the Institute of Nuclear Power age. Failure data from two national Operation's Nuclear Plant Reliability databases. Nuclear Plant Reliability Data System has been conducted for D3ta System (NPRDS) and Licensee Event several types of MSIVs and valve Reports (LERs), as well as plant operators for both boiling- water specific data were reviewed and reactors and pressurized-water analyzed to understand the effects of reactors. The focus of this review is aging on the CI functions. This study on MSIV failures modes. cctuator NUREG-1426 18

( l l

Compilation of Reports: 1994-1998 failure modes, consequences of failure identify the degradation and aging on plant operations, method of failure mechanisms affecting various components detection, and major stressors of these large motors, the failure affecting both valves and valve modes that result, and their effects operators, upon the function of the motor. The effects of large motor failures upon the systems in which they are

Title:

Aging assessment of large operating, and on the plant as a whole, j electric motors in nuclear power plants were analyzed from failure reports in the databases. The effectiveness of Author (s)/ Editor (s): Villaran. M. . the industry's large motor maintenance l Subudhi M. (Brookhaven National Lab., programs was assessed based upon the

Upton. NY (United States)) failure reports in the databases and i Soonsorina Orcanization

NRC; Nuclear reviews of plant maintenance procedures Regulatory Commission Washington, DC and programs.The Nuclear Regulatory

, (United States) Comission'; (NRC's) Office of Nuclear l Publication Date: Mar 1996 Regulatory Research wrote this draft l Reoort Number (s): NUREG/CR-6336: report at the request of NRC's Office BNL-NUREG--52460 of Nuclear Reactor Regulation. This Order Number: TI96008243 report is to serve as a reference that Abstract: Large electric motors serve the NRC staff and the nuclear industry as the prime movers to drive high and its suppliers can use when writing capacity pumps. fans compressors, and and applying sampling programs for generators in a variety of nuclear comercial grade dedication.The RES plant systems. This study examined the staff reviewed the history, practices, stressors that cause degradation and and guidelines for comercial grade aging in large electric motors dedication in the nuclear industry to operating in various plant locations understand the particular needs for a and environments. The operating new sampling reference. Additionally, history of these machines in nuclear it analyzed various material standards plant service was studied by review and such as those in the American Society analysis of failure reports in the of Mechanical Engineers Boiler and NPRDS and LER databases. This was Pressure Vessel Code.Section II, and supplemented by a re"iew of motor those of the American Society for designs, and their rJclear and balance Testing and Materials, as well as of plant applications, in order to standard industrial steel-making characterize the failure mechanisms practices. As a result of this review

' hat cause degradation, aging. and and analysis, the staff identified failure in large electric motors. A important principles that must be generic failure modes and effects applied to ensure the integrity of the analysis for large squirrel cage dedication process for simple, metallic induction motors was performed to comercial grade items. This report 19 NUREG-1426

Compilation of Reports - 1994-1998 A;Me:fents rrtrtams rsusphatirtra abyntttlieir rated starting torque during tests at is suitable for field application. normal voltages and temperatures.

For all five motors (dc as well as ac), actual motor torque losses oue to voltage

Title:

Motor-0perated Valve (MOV) degradation were greater than the Actuator Motor and Gearbox Testing losses calculated by methods typically Author (s)/ Editor (s): K.Dewall. J.C. used for predicting motor torque at Watkins. D. Bramwell (Idaho National degraded voltage conditions.

Engineering Laboratory)

Shortlaritmrdlmarindtnkd rNR0r tests compared well witn stall torques in Washington DC (United States) dynamometer-type tests.

Publication Date: July 1997 FReodMdtuntferthe)ac hh$tfdMCR4F&dttEG-motor torque i 'ses due to elevated operating 96/0219 temperatures were equal to or lower Abstract: This report documents the than losses calculated by the typical results of valve research sponsored by predictive method: for the de motor, the U.S. Nuclear Regulatory Comission the actual losses were significantly (NRC) and conducted at the Idaho greater than the predictions.

National Engineering and Entironmental Laboratory (INEEL) The research For a'l three actuator gearboxes, the provides technical bases to the NRC in tual running efficiencies determined support of their effort regarding from testing were lower than the motor-operated valves (MOVs) in nuclear running efficiencies published by the power plants. Specifically, the manufacturer. In most instances, the research measured the capabilities of actual pullout efficiencies were lower typical valve actuators during than the published pullout operation at simulated design basis efficiencies.

loads and operating conditions. Using a test stand that simulates the stem load Operation of the gearbox at profiles a valve actuator would elevemperature did not affect the experience when closing a valve against operating efficiency.

flow and pressure, we tested five typical electric motors (four ac motors and one de motor) and three gearboxes

Title:

Component unavailability versus at conditions a motor might experience inservice test (IST) interval:

in a power plant, including such off- Evaluations of component aging effects normal conditions as operation at high with applications to check valves temperature and reduced voltage. We Author (s)/ Editor (s): Vesely. W.E.

also monitored the efficiency of the (Vesely. (4.E.). Dublin. OH (United actuator gearbox. The testing produced States)); Poole. A.B. (0ak Ridge the following results: National Lab., TN (United States))

Soonsorino Oroanization: NRC; Nuclear NUREG-1426 20

l l

Compilation of Reports: 1994-1998 Regulatory Conmission. Wasnington. DC sensitivity evaluations, summary tables i (United States) are constructed showing how optimal IST l

Publication Date: Jul 1997 interval ranges for check valves can

! Reoort Number (s): NUREG/CR-6508: vary relative to different aging ORNL--6909 behaviors which might exist. The Order Number: TI97007394 evaluations are also used to identify Abstract: Methods are presented for IST intervals for check valves which calculating component unavailabilities are robust to component aging effects.

When inservice test (IST) intervals are General insights on aging effects are changed and when component aging is also extracted. These sensitivity explicitly included. The methods studies and extracted results provide extend usual approaches for calculatiag useful information which can be unavailability and risk effects of supplemented or be updated with plant changing IST intervals which utilize specific information. The models and Probabilistic Risk Assessment (PRA) results can also be input to PRA.s to methods that do not explicitly include determine associated risk implications.

component aging. Different IST characteristics are handled inc kding ISTs which are followed by corrective

Title:

Results of Pressure Locking and maintenances which completely renew or Thermal Binding Tests of Gate Vavles partially renew the component. ISTs Author (s)/ Editor (s): K.G.DeWall. J.C.

which are not followed by maintenance Watkins. M.G. McKellar. D. Bramwell activities needed to renew the (Idaho National Engeering and I component are also handled. Any Environmental Laboratory) downtime associated with IST. including Soonsorino Oroanization: NRC:

the test downtime and the following Washington DC (United States) maintenance downtime, is included in Publication Date: May 1998 the unavailability evaluations. A Reoort Number (s): NUREG/CR-6611:

range of component aging behaviors is INEEL/ EXT-98/00161 studied including both linear and Abstract: The U.S. Nuclear Regulatory nonlinear aging behaviors. Based upon Conmission (NRC). Office of Nuclear evaluations completed to date, pooled Regulatory Research, is funding the failure data on check valves show Idaho National Engineering and I relatively small aging (e.g., less than Environmental Laboratory (INEEL) in 7% per year). However, data from some performing research investigating the plant systems could be evidence for performance of gate valves subjected to larger aging rates occurring in time pressure locking and thermal binding periods less than 5 years. The methods conditions. Pressure locking and are utilized in this report to carry thermal binding are phenomena that make out a range of sensitivity evaluations a closed gate valve difficult to open.

to evaluate aging effects for different Pressure locking can occur when possible applications. Based on the operating sequences or temperature l

21 NUREG-1425 l

l

l Compilation of Reports - 1994-1998 changes cause the pressure of the fluid Nuclear Science Center); Kam. F.B.K.

in the bonnet (and, in most gate (0ak Ridge National Lab., TN (United valves, between the discs) to'be higher States)): McGarry. E.D. ' National Inst.

than the pressure on the upstream and of Standards and Technology.

downstream sides of the disc assembly. Gaithersburg, MD (United States))

' Thermal binding can occur when thermal soonsorina Oraanization* NRC: Nuclear expansion / contraction effects cause the Regulatory Commission, Washington, DC disc to be squeezed between the valve (United States) l body seats. If the loads associated Publication Date: Jul 1994 with pressure locking or thermal Reoort Number (s): NUREG/CR-6206:

binding are very high, the actuator ORNL/TM -12693 might not have the capacity to open the Order Number: TI94015398 valve. We tested a flexible-wedge gate Abstract: Comparison of transport valve and a do'iole-disc gate valve calculations of the dosimeter under pressure locking and thermal activities with the experimental binding conditions. The results show measurements shows that the values that these valves are susceptible to obtained with ENDF/B-VI cross-pressure locking: however, they are not section data overestimate the measured '

significantly affected by thermal results for high-energy-threshold l _ binding. For the flexible-wedge gate reactions in the cavity by up to 41%.

l valve, pressure locking loads (in terms and thermal reactions by up to a factor of stem thrust) were higher than of 3.0. The transport calculations corresponding hydrostatic opening loads performed with the original SAILOR by a factor of 1.1 to 1.5. For the cross-section library (based on parallel disc gate valve, pressure ENOF/B-VI data) overestimate measured locking loads were higher by a factor threshold reactions by only 15% and the of 2.05 to 2.4. The results also show thermal reactions by about a factor of that seat leakage affects the bonnet 2.50. These results are inconsistent pressurization rate when the valve is with those obtained in earlier studies subjected to thermally induced pressure that compared transport calculations locking conditions. done with SAILOR vs ENOF/B-VI, which indicate that SAILOR tends to underestimate ' cavity dosimeter activities for threshold reactions, Dosimetry while the ENOF/B-VI values usually agree better with experimental results.

One factor that probably contributes to

Title:

Transport calculations of the rather large discrepancy between

- radiation exposure to vessel support the computed and measured activities is structures in the Trojan Reactor

- Author (s)/ Editor (s): Asgari. M. :

the core power distribution used in the transport calculations. Because of Williams. M.L. (Louisiana State Univ.. unavailability of plant-specific data.

Baton Rouge, LA (United States).

NUREG-1426 22 h

.- _ _ m _ _ . . - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . . _ . . _

l Compilation of Reports: 1994-1998 a generic power distribution provided pseudo-problem independent format and by Westinghouse was used. Since the then collapsed into the final calculated cavity flux levels appear to broad-group format. The fine-group be over-estimated, the results library which is designated estimated for the exposure to the - VITAMIN-86, contains 120 nuclides. The support structure should be BUGLE-93 47-neutron-group /20-gamma-ray-conservative. group library contains the same 120 nuclides processed as infinitely dilute ,

I and collapsed using a weighing spectrum

Title:

Production and testing of the typical of a concrete shield, VITAMIN B6 fine group and the BUGLE-93 Additionally. BUGLE-93 contains 105 broad- group neutron / photon nuclides processed with resonance cross-section libraries derived from self-shielding and weighted using ENOF/B-VI nuclear data spectra specific to BWR and PWR j Author (s)/ Editor (s): Ingersoll. D.T. . material compositions and reactor White. J.E. : Wright. R.Q. . Hunter, models. Several dosimetry response l H.T. : Slater. C.O. (Nuclear Regulatory functions and kerma factors for all 120 l Comission. Washington OC (United nuclides are also included with the States)); Greene. N.M. : Roussin. R.W. library. An extensive integral data (0ak Ridge National Lab., TN (United testing effort was performed to qualify States)); MacFarlane, R.E. (Los Alamos the new library. In general, results National Lab., NM (United States)) using the new data show significant soonsorino Oroanization: NRC: Nuclear improvements relative to earlier ENOF Regulatory Commission. Washington DC data.

(United States)

B:'olication Date- Jan 1995 Reoort Number (s): NUREG/CR-6214:

Title:

Pool critical assembly pressure ORNL--6795 vessel facility benchmark Order Number: TI95005715 Author (s)/ Editor (s): Remec. I. : Kam.

Abstract: A new'multigroup F.B.K. (0ak Ridge National Lab., TN cross-section library based on (United States))

l

. ENDF/B-VI data has been produced and soonsorina Oraanization: NRC: Nuclear l tested for light water reactor Regulatory Comission. Washington DC l shielding and reactor pressure vessel (United States) l dosimetry applications. The Publication Date: Jul 1997 I broad group library, which is Reoort Number (s): NUREG/CR-6454: l designated BUGLE-93, is intended to ORNL/TM--13205 )

replace the aging BUGLE-80 and SAILOR Order Number: TI97008288 '

libraries. The processing methodology Abstract: This pool critical assembly is consistent with ANSI /ANS 6.1.2. (PCA) pressure vessel wall facility since the ENOF data were first benchmark (PCA benchmark) is described processed into a fine-group, and analyzed in this report. Analysis 23 NUREG-1426 l

l 1

1 Compilation of Reports - 1994-1998 of the PCA benchmark can be used for Soonsorino Oroanization: NRC:

partial fulfillment of the requirements Washington DC (United States) for the qualification of the Publication Date: February 1998 methodology for pressure vessel neutron Reoort Number (s): NUREG/CR-6453:

fluence calculations, as required by ORNL/TM-13204 1

the US Nuclear Regulatory Commission Abstract: The H. B. Robinson Unit 2 regulatory guide DG-1053. Section 1 of Pressure Vessel Benchmark (HBR-2 this report describes the PCA benchmark benchmark) is described and analyzed in and provides all data necessary for the this report. Analysis of the HBR-2 benchmark analysis. The measured benchmark can be used as partial quantities, to be compared with the fulfillment of the requirements for the calculated values, are the equivalent qualification of the methodology for fission fluxes. In Section 2 the calculating neutron fluence in pressure analysis of the PCA benchmark is ve::sels, as required by the U.S.

described. Calculations with the Nuclear Regulatory Commission computer code DORT, based on the Regulatory Guide DG-1053. Calculational discrete-ordinates nethod, were and Dosimetry Methods for Determining performed for three ENDFIB-VI-based Pressure Vessel Neutron Fluence.

multigroup libraries: BUGLE-93.

SAILOR-95, and BUGLE-96. An excellent Section 1 of this report describes the agreement of the calculated (C) and BR-2 benchmark and provides all the measures (M) equivalent fission fluxes dimensions, material compositions, and was obtained. The arithmetic average neutron source data necessary for the C/M for all the dosimeters (total of analysis. The measured quantities, to l 31) was 0.93 [+-] 0.03 and 0.92 [+-] be compared with the calculated values, 0.03 for the SAILOR-95 and BUGLE-96 are the specific activities at the end

' libraries, respectively. The average of fuel cycle 9. The charac' eristic C/M ratio, obtained with the BUGLE-93 feature ofthe HBR-2 benchmark is that library, for the 28 measurements was it provides measurements on both sides 0.93 [+-] 0.03 (the neptunium of the pressure vessel: in the measurements in the water and air surveillance capsule attached to the regions were overpredicted and excluded thermal shield and in the reactor from the average). No systematic cavity.

decrease in the C/M ratios with increasing distance from the core was In Section 2. the analysis ofthe HBR-2 observed for any of the libraries used. benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were

Title:

H.B. Robinson-2 Pressure Vessel performed with three multigroup Benchmark libraries based on ENDF/B-VI: BUGLE-93, Author (s)/ Editor (s): I. Remec F.B.K. SAILOR-95 and BUGLE-96. The average Kam (Oak Ridge National Laboratory) ratio of the calculated-tomeasured NUREG-1426 24

Compilation of Reports: 1994-1998 specific activities (C/M) for the six induced currents (GICs) caused by the dosimeters in the surveillance capsule solar disturbances on the in-plant was 0.90 0.04 for all three electrical distribution system and libraries. The average C/Ms for the equipment in nuclear power stations.

cavity dosimeters (without neptunium The plant-specific electrical dosirneter) were 0.89 0.10. 0.91 distribution system for a typical 0.10 and 0.90 0.09 for the BUGLE 93, nuclear plant is modeled using the SAILOR-95 and BUGLE-96 libraries. Electromagnetic Transient Program l respectively. (EMTP). The computer model simulates online equipment and loads from the It is expected that the agreement ofthe station transformer in the switchyard q- calculations with the measurements. of the power station to the similar to the agreement obtained in safety-buses at 120 volts to which all this research, should typically be electronic devices are connected for observed when the discrete-ordinates plant monitoring. The analytical model method and ENDFIB-VI libraries are used of the plant's electrical distribution i

for the HBR-2 benchmark analysis. system is studied to identify the transient effects caused Dy the half-cycle saturation of the station

> Electrical Systems transformers due to GIC. This study j provides results of the voltage harmonics levels that have been noted

Title:

The effects of at various electrical buses inside the solar geomagnetically induced currents plant. The emergency circuits appear on electrical systems in nuclear power to be more susceptible to high stations harmonics due to the normally light Author (s)/ Editor (s h Subudhi . M. load conditions. In addition to (Brookhaven National Lab., Upton, NY steady-state analysis, this model was

(United States)); Carroll. D.P.- further analyzed simulating various

(Florida Univ. ~Gainesville. FL-(United plant transient conditions (e.g., loss

-States)); Kasturi.'S. (MOS. Inc., of load or large motor start-up)

Melville. NY (United States)) occurring during GIC events. Detail Soonsorina Oraanization: NRC: Nuclear models of the plant's protective Regulatory Commission. Washington. DC relaying system employed in bus

. (United States) transfer application were included in 4

Publication Date: Jan 1994 this model to study the effects of the harmonic distortion of the voltage Reoort Number (s): NUREG/CR-5990:

BNL-NUREG--52359 input, Potential harmonic effects on Order Number: TI94005979 the uniterruptable power system (UPS)

Abstract: This report presents the are qualitatively discussed as well.

results of a study to evaluate the

potential effects of geomagnetically 25 NUREG-1426 1

'T

,, - - ,. . . - - . - , ~ , . . . - , , , - . - -- -- . -, - - . = ~ -

Compilation of Reports - 1994-1998

Title:

Selected fault testing of electronic isolation devices used in

Title:

Summary of work completed under nuclear power plant operation the Environmental and Dynamic Equipment Author (s)/ Editor (s): Villaran, M. : Qualification research program (EDOP)

Hillman, K. : Taylor, J. : Lara. J. . Author (s)/ Editor (s): Steele, R. Jr. :

Wilhelm, W. (Brookhaven National Lab., Bramwell, D.L. . Watkins, J.C. :

Upton, NY (United States)) DeWall. K.G. (EG and G Idaho. Inc.,

Soonsorino Oroanization: NRC: Nuclear Idaho Falls. ID (United States))

Regulatory Commission. Washington DC Soonsorino Oroanization: NRC: Nuclear (United States) Regulatory Comission. Washington. DC Publication Date: May 1994 (United States)

Reoort Number (s): NUREG/CR-6086: Publication Date: Feb 1994 BNL-NUREG--52385 Reoort Number (s): NUREG/CR-5935:

Order Number: TI94012160 EGG--2686 Abstract: Electronic isciation devices Order Number: TI94007115 are used in nuclear power plants to Abstract: This report documents the provide electrical separation between results of the main projects undertaken safety and non-safety circuits and under the Environmental and Dynamic l

systems. Major fault testing in an Equipment Qualification Research earlier program indicated that some Program (EDOP) sponsored by the U.S.

! energy may pass through an isolation Nuclear Regulatory Commission (NRC) device when a fault at the maximum under FIN A6322. Lasting from fiscal credible potential is applied in the year 1983 to 1987, the program dealt l, ~ transverse mode to its output with environmental and dynamic terminals. During subsequent field (including seismic) equipment .

qualification testing of isolators, qualification issues for mechanical and concerns were raised that the worst electromechanical components and case fault, that is, the maximum systems used in nuclear power plants.

I credible fault (MCF), may not occur The research results have since been with a fault at the maximum credible used by both the NRC and industry. The l

potential, but rather at some lower program included seven major research potential. The present test program projects that addressed the following investigates whether problems can arise issues: (a) containment purge and vent

(

when fault levels up to the MCF valves performing under design basis potential are applied to the output loss of coolant accident loads, (b) terminals of an isolator. The fault containment piping penetra'. ions and I

energy passed through an isolated isolation valves performing ander device during a fault was measured to seismic loadings and design o bis and determine whether the levels are great severe accident containment will

! enough to potentially damage or degrade displacements, (c) shaft seals for performance of equipment on the input primary coolant pumps performing under l

l (Class 1E) side of the isolator. station blackout conditions. (d) l l

NUREG-1426 26 l

l l

l l Compilation of Reports: 1994-1998 l

electrical cabinet internals responding beyond 40 years. After subsequent to in-structure generated motion investigation, the NRC Staff concluded (rattling), and (e) in situ piping and that questions related to the i valves responding to seismic loadings. differences in E0 requirements between Another project investigating whether older and newer plants constitute a certain containment isolation valves potential generic issue which should be will close under design basis evaluated for backfit, independent of conditions was also started under this license renewal activities. E0 tasting program. This report includes eight of electric cables was performed by main section, each of which provides a Sandia National Laboratories (SNL) brief description of one of the under contract to the NRC in support of projects, a summary of the findings, license renewal activities. Results l and an overview of the application of showed that some of the environmentally the results. A bibliography lists the qualified cables either failed or journal articles, papers, and reports exhibited marginal insulation that document the research, resistance after a simulated plant life of 20 years during accident simulation.

This indicated that the E0 process for

Title:

Workshop on environmental some electric cables may be qualification of electric equipment non-conservative. These results raised Author (s)/ Editor (s): Lofaro, R. , questions regarding the E0 process Gunther W. : Villaran, M. . Lee, B.S. including the bases for conclusions

Taylor, J. (comps.) (Brookhaven about the qualified life of components National Lab., Upton, NY (United based upon artificial aging prior to States)) testing.

Sponsorino Oroanization: NRC: Nuclear Regulatory Commission. Washington OC (United States)

Title:

Literature review of Publication Date: May 1994 environmental qualification of Reoort Number (s): NUREG/CP-0135; safety-related electric cables: Sumary BNL-NUREG--52409: CONF-9311207-- of past work. Volume 1 Order Number: TI94012761 Author (s)/ Editor (s): Subudhi, M.

Abstract: Questions concerning the (Brookhaven National Lab. , Upton, NY Environmental Qualification (E0) of (United States))

electrical equipment used in commercial Soonsorino ONanization: NRC: Nuclear nuclear power plants have recently Regulatory Comission. Washington, DC become the subject of significant (United States) interest to the US Nuclear Regulatory Publication Date: Apr 1996 Comission (NRC). Initial questions ReDort Number (s): NUREG/CR-6384 Vol.1:

centered on whether compliance with the BNL-NUREG--52480-Vol.1 E0 requirements for older plants were Order Number: TI96009367 adequate to support plant operation Abstract: This report summarizes the 27 NUREG-1426

_ ___ _ _ _ _ _ . _ _ _ _ _ __ _ _ _ m .

Compilation of Reports - 1994-1998 findings from a review of published Regulatory Commission. Washington DC l documents dealing with research on the (United States) environmental qualification of Publication Date: Apr 1996 safety-related electric cables used in Recort Number (s): NUREG/CR-6384-Vol.2:

nuclear power plants. Simulations of BNL-NUREG--52480-Vol.2 accelerated aging and accident Order Number: TI96009368 conditions are important considerations Abstract: In support of the US NRC in qualifying the cables. Significant Environmental Qualification (EO) research in these two areas has been Research Program. a literature review performed in the US and abroad. The was performed to identify past relevant results from studies in France, work that could be used to help fully Germany, and Japan are described in or partially resolve issues of interest this report. In recent years, the related to the qualification of development of methods to monitor the low-voltage electric cable. A sumary condition of cables has received of the literature reviewed is special attention. Tests involving documented in Volume 1 of this report.

chemical and physical examination of In this. Volume 2 of the report, cable's insulation and jacket dossiers are presented which document materials, and electrical measurements the issues selected for investigation of the insulation properties of cables in this program, along with are discussed. Although there have recomendations for future work to been significant advances in many resolve the issues, when necessary.

areas there is no single method which The dossiers are based on an analysis can provide the necessary information of the literature reviewed, as well as about the condition of a cable expert opinions. This analysis currently in service. However. it is includes a critical review of the oossibh that further research may information available from past and identify a combination of several ongoing work in thirteen specific areas methods that can adequately related to E0. The analysis for each characterize the cable's condition. area focuses on one or more questions which must be answered to consider a particular issue resolved. Results of

Title:

Literature review of the analysis are presented, along with environmental qualification of recomendations for future work. The safety-related electric cables: analysis is documented in the form of a Literature analysis and apW.,m e . dossier for each of the areas analyzed.

l Volume 2 Author (s)/ Editor (s): Lofaro, R. .

Bowerman B. . Carbonaro. J.

Title:

Long term aging and (Brookhaven National Lab. Upton. NY loss-of-coolant accident (LOCA) testing (United States)) (and others) of electrical cables Soonsorino Orcanization: NRC: Nuclear Author (s)/ Editor (s): Nelson. C.F. .

l NUREG-1426 28

i Compilation of Reports: 1994-1998 Gauthier G. . Carlin, F. (and others) rate, and the amount of degradation Soonsorina Oraanization: NRC: Nuclear decreased as the dose rate was Regulatory Connission. Washington OC increased.

(United States)

Publication Date: Oct 1996 Reoort Number (s): NUREG/CR-6202: EAC and Fatigue IPSN--94-03: SAND--94 0485 Order Number: TI97000454 Abstract: Experiments were performed

Title:

Environmentally assisted to assess the aging degradation and cracking in Light Water Reactors:

loss-of- coolant accident (LOCA) Semiannual report. April behavior of electrical cables subjected 1993--September 1993 to long-term aging exposures. Four Author (s)/ Editor (s): Chopra. 0.K. :

different cable types were tested in Chung, H.M. . Karlsen T. : Kassner, both the U.S. and France: (1) U.S. 2 T.F. , Michaud. W.F. . Ruther. W.E. .

conductor with ethylene propylene Sanecki, J.E. . Shack. W.J. . Soppet, rubber (EPR) insulation and a Hypalon W.K. (Argonne National Lab. IL (United jacket. (2) U.S. 3 conductor with States))

cross-linked polyethylene (XLPE)

Soonsorino Oraanization: NRC: Nuclear insulation and a Hypalon jacket. (3) Regulatory Comission, Washington, DC French 3 conductor with EPR insulation (United States) and a Hypalon jacket. (4) French Publication Date: Jun 1994 coaxial with polyethylene (PE)

Reoort Number (s):

insulation and a PE jacket. The data NUREG/CR-4667-Vol.17: ANL--94/16-Vol.17 represent up to 5 years of simultaneous aging where the cables were exposed to Order Number: TI94014862 identical aging radiation doses at Abstract: This report summarizes work either 40[ degrees]C or 70[ degrees]C:

performed by Argonne National however, the dose rate used for the Laboratory on fatigue and aging irradiation was varied over a environmentally assisted cracking (EAC) wide range (2-100 Gy/hr). Aging was in light water reactors (LWRS) during folicwed by exposure to simulated the six months from April 1993 to French LOCA conditions. Several September 1993. EAC and fatigue of mechanical, electrical, and piping, pressure vessels, and core physical-chemical condition monitoring components in LWRs are important techniques were used to investigate the concerns as extended reactor lifetimes degradation behavior of the cables. are envisaged. Topics that have been All the cables, except for the French investigated include (a) fatigue of PE cable, performed acceptably during low-alloy steel used in piping, steam l the aging and LOCA simulations. In generators, and reactor pressure l general, cable degradation at a given vessels: (b) EAC of cast stainless i

dose was highest for the lowest dose steels (S$s); and (c) radiation-induced 29 NUREG-1426

Compilation of Reports - 1994-1998 segregation and irradiation-assisted Publication Date: Mar 1995 l

stress corrosion cracking of Type 304 Reoort Number (s):

SS after accumulation of relatively NUREG/CR-4667-Vol.18: ANL--95/2-Vol.18 high fluence. Fatigue tests were Order Number: TI95009017 conducted on medium-sulfur-content Abstract: This report sucinarizes work A106 Gr B piping and A533-Gr B pressure performed by Argonne National vessel steels in simulated PWR water Laboratory (ANL) on fatigue and and in air. Additional crack growth environmentally assisted cracking (EAC) data were obtained on in light water reactors (LWRs) during fracture-mechanics specimens of cast the six months from October 1993 to l

l austenitic SSs in the as- received March 1994. EAC and fatigue of piping.

l and thermally aged conditions in pressure vessels, and core components simulated boiling-water reactor (BWR) in LWRs are important concerns in water at 289[ degree]C. The data were operating plants and as extended compared with predictions based on reactor lifetimes are envisaged.

I crack growth correlations for wrought Topics that have been investigated I austenitic SS in oxygenated water include (a) fatigue of low-alloy steel I

developed at ANL and rates in air from used in piping, stt.am generators, and Section 11 of the ASME Code. reactor pressure vessels. (b) EAC of Microchemical and microstructural wrought and cast austenitic stainless changes in high- and corrercial-purity steels (SSs), and (c) radiation-induced l

Type 304 SS specimens from segregation and irradiation-assisted i control-blade absorber tubes and a stress corrosion cracking (IASCC) of I

control-blade sheath from operating Type 304 SS after accumulation of BWRs were studied by Auger electron relatively high fluence. Fatigue tests spectroscopy and scanning electron have been conducted on A302-Gr B microscopy. low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated

Title:

Environmentally assisted pressurized water reactor water are cracking in light water reactors. valid for high-sulfur heats that show Semiannual report. October 1993--March environmentally enhanced fatigue crack 1994. Volume 18 growth rates. Additional crack growth Author (s)/ Editor (s): Chung. H.M. : data were obtained on Chopra. 0.K. : Erck. R.A. . Kassner, fracture-mechanics specimens of T.F. : Michaud. W.F. : Ruther. W.E. . austenitic SSs to investigate threshold Sanecki, J.E. : Shack. W.J. . Soppet. stress intensity factors for EAC in W.K. (Argonne National Lab., IL (United high-purity oxygenated water at States)) 289[ degrees]C. The data were compared Soonsorino Orcanization: NRC: Nuclear with predictions based on crack growth Regulatory Coninission. Washington. OC correlations for wrought austenitic SS (United States) in oxygenated water developed at ANL NUREG-1426 30

l l

l Compilation of Reports: 1994-1998 i

and rates in air from Section XI of the irradiation assisted stress corrosion ASME Code. Microchemical and cracking (IASCC) of Type 304 SS, microstructural changes in high and Fatigue tests have been conducted on l comercial-purity Type 304 SS specimens A106-Gr 8 and A533-Gr 8 steels in I

from control-blade absorber tubes and a oxygenated water to determine whether a j control-blade sheath from operating slow strain rate applied during boiling water reactors were studied by different portions of a tensile-loading Auger electron spectroscopy and cycle are equally effective in scanning electron microscopy to decreasing fatigue life. Crack growth determine whether trace impurity data were obtained on fracture-l elements, which are not specified in mechanics specimens of SSs and Alloy the ASTM specifications, may contribute 600 to investigate EAC in simulated to IASCC of solution-annealed boiling water reactor (BOOR) and materials. pressurized water reactor environments at 289'C. The data were compared with predictions from crack growth

Title:

Environmentally Assisted correlations developed at ANL for SSs Cracking in Light Water Reactors in water and from rates in air from Semiannual Report April 1994-September Section X1 of the ASME Code.

1994 Microchemical changes in high- and Author (s)/ Editor (s): 0. K. Chopra. H. comercial-purity Type 304 SS specimens M. Chung. O. J. Gavenda. E. E. Gruber. from control-blade absorber tubes and a A. G. Hins. T. H. Hughes. T. F. control-blade sheath from operating Kassner. W. E. Ruther, W. J. Shack, and BWRs were studied by Auger electron W. K. Soopet (Argonne National spectroscopy and scanning electron Laboratory) microscopy to determine whether trace Soonsorino Oraanization: NRC: impurity elements may contribute to Washington DC (United States) IASCC of these materials.

Publication Date: September 1995 Reoort Number (s): NUREG/CR-4667/ANL-95/2 Vol 19

Title:

Environmentally assisted Abstract: This report sumarizes work cracking in Light Water Reactors:

performed by Argonne National Semiannual report. October 1994--March Laboratory (ANL) on fatigue and 1995 Volume 20 environmentally assisted cracking (EAC) Author (s)/ Editor (s): Chung. H.M. .

in light water reactors from April to Chopra. 0.K. . Gavenda. D.J. ; Hins.

September 1994. Topics that have been A.G. . Kassner. T.F. . Ruther. W.E. :

investigated include (a) fatigue of Shack. W.J. . Soppet. W.K. (Argonne carbon and low-alloy steel used in National Lab., IL (United States))

piping and reactor pressure vessels. Soonsorina Oraanization: NRC: Nuclear (b) EAC of austenitic stainless steels Regulatory Comission. Washington. DC

($$s) and Alloy 600, and (c) (United States) t 31 NUREG-1426 i

Compilation of Reports - 1994-1998 Publication Date: Jan 1996 contribute to IASCC of these materials.

ReDort Number (s):

NUREG/CR-4667 Vol.20: ANL--95/41-Vol.20

Title:

Corrosion fatigue of alloys 600 Order Number: Tl96005922 and 690 in simulated LWR environments Abstract: This report summarizes work Author (s)/ Editor (s): Ruther. W.E. :

performed by Argonne National Soppett W.K. : Kassner. T.F. (Argonne Laboratory on fatigue and National Lab.. IL (United States))

environmentally assisted cracking (EAC) Soonsorino Oraanization:- NRC: Nuclear in light water reactors (LWRS) from Regulatory Comission. Washington. OC October 1994 to March 1995. Topics (United States) that have been investigated include (a) Publication Date: Apr 1996 fatigue of carbon and low-alloy steel Recort Number (s): NUREG/CR-6383:

i used in reactor piping and pressure ANL--9S/37 vessels. (b) EAC of Alloy 600 and 690.. Order Number: TI96008966 and (c) irradiation-assisted stress Abstract: Crack growth data were corrosion cracking (IASCC) of Type 304 obtained on fracture mechanics SS. Fatigue tests were conducted on specimens of Alloys 600 and 690 to ferritic steels in water with several investigate environmentally assisted dissolvedoxygen (00) concentrations to cracking (EAC) in simulated boiling determine whether a slow strain rate water reactor and pressurized water applied during different portions of a reactor environments at 289 and 320 C.

tensile loading c/cle are equally Preliminary information was obtained on effective in decreasing fatigue life, the effect of temperature. load ratio.

Tensile properties Ond microstructures stress intensity (K). and the of several heats of Alloy 600 and 690 dissolved-oxygen and -hydrogen were characterized for correlation with concentrations of the water on EAC.

EAC of the alloys in simelated LWR Specimens of Type 316NG and sensitized environments. Effects of 30 and Type 304 stainless steel (SS) were electrochemical potential on included in several of the experiments susceptibility to interg' anular to assess the behavior of these cracking of high- and conurcial-purity materials and Alloy 600 under the same Type 304 SS specimens from water chemistry and loading conditions.

control-blade absorber tubes and a The experimental data are compared with control-blade sheath irradiated in predictions from an Argonne National boiling water reactors were determined Laboratory (ANL) model for crack growth in slow-strain-rate-tensile tests at rates (CGRs) of S$s in water and the 289(degrees]C. Microchemical changes ASME Code Section 11 correlation for in the specimens were studied by Auger CGRs in air at the K[sub max) and

. electron spectroscopy and scanning load-ratio values in the various tests.

electron microscopy to determine The data for all of the materials were whether trace impurity elements may bounded by ANL model predictions and i

NUREG 1426 32 l

. . --- - -- ~ - . . . . .. . - . - . - . _ .._ ._ .

l Compilation of Reports: 1994-1998 the ASME Section 11 air line, Centigrade. Microchemical changes in the specimens were studied by Auger l electron spectroscopy and scanning

Title:

Environmentally assisted electron microscopy to determine cracking in light water reactors whether trace impurity elements may Author (s)/ Editor (s): Chopra, 0.K. , contribute to IASCC of these materials.

Chung. H.M. Gruber, E.E. (and others)

Soonsorira Oraanization: NRC: Nuclear Regulatory Comission. Washington, DC

Title:

Environmentally assisted (United States) cracking in light water reactors.

Publication Date: Jul 1996 Semiannual progress report. January j Reoort Number (s): 1996--June 1996 l NUREG/CR-4667-Vol.21: ANL--96/1-Vol.21 Author (s)/ Editor (s): Cnopra, 0.K. . 1 l Order Number: TI96013829 Chung. H.M. , Gruber, E.E. (and others)

Abstract: This report sumarizes work Soonsorino Oroanization: NRC: Nuclear performed by Argonne National Regulatory Comission. Washington, DC Laboratory on fatigue and (United States) environmentally assisted cracking (EAC) Publication Date: May 1997 in light water reactors (LWRs) from Reoort Number (s):

April 1995 to December 1995. Topics NUREG/CR-4667-Vol.22: ANL--97/9-Vol.22 that have been investigated include Order Number: TI97006378 fatigue of carbon and low-alloy steel Abstract: This report sumarizes work used in reactor piping and pressure performed by Argonne National vessels, EAC of Alloy 600 and 690 and Laboratory on fatigue and irradiation-assisted stress corrosion environmentally assisted cracking (EAC) cracking (IASCC) of Type 304 SS. in light water reactors from January Fatigue tests were conducted on 1996 to June 1996. Topics that have ferritic steels in water that contained been investigated include (a) fatigue various concentrations of dissolved of carbon, low alloy, and austenitic oxygen (00) to determine whether a slow stainless steels (SSs) used in reactor  ;

strair, rate applied during different piping and pressure vessels (b) l portions of a tensilc- N ding cycle are irradiation-assisted stress corrosion  !

equally effective in decreasing fatigue cracking of Type 304 SS, and (c) EAC of life. Crack- growth-rate tests were Alloys 600 and 690. Fatigue tests were conducted on compact-tension specimens conducted on ferritic and austenitic from several heats of Alloys 600 and S$s in water that contained various 690 in simulated LWR environments. concentrations of dissolved oxygen (00) i Effects of fluoride-ion contamination to determine whether a slow strain rate I ori susceptibility to intergranular applied during various portions of a )

crack;ng of high- and comercial- tensile-loading cycle are equally 1 purity Type 304 SS specimens from effective in decreasing fatigue life.

control-tensile tests at 288 degrees Slow-strain-rate tensile tests were 33 NUREG 1426 1

--. -- l

Compilation of Reports - 1994-1998 conducted in simulated boiling water solutions used in this research were reactor (BWR) water at 288[ degrees]C on prepared using pure chemical reagents SS specimens irradiated to a low to simulate the halogens and inhibitors fluence in the Halden reactor and the found in insulation extraction results were compared with similar data solutions. The results indicated that from a control-blade sheath and sodium silicate compounds that were neutron absorber tubes irradiated in higher in sodium were more effective

-BWRs to the same fluence level, for preventing chloride-induced ESCC in Crack-growth rate tests were conducted Type 304 austenitic stainless steel.

on compact-tension specimens from Potassium silicate (all-silicate several heats of Alloys 600 and 690 in inhibitor) was not as effective as air and high-purity low-D0 water. 83 sodium silicate. Limited testing with refs. 60 figs., 14 tabs. sodium hydroxide (all-sodium inhibitor) indicated that it may be effective as i

an inhibitor. Fluoride, bromide, and

Title:

Effects of fluoride and other iodide caused mimma ESCC which could halogen ions on the external stress be effectively inhibi e d by sodium corrosion cracking of Type 304 silicate. The addition of fluoride to austenttic stainless steel the chloride / sodium silicate systems at Author (s)/ Editor (s): Whorlow. K.M. : the threshold of ESCC appeared to have Hutto, F.B. Jr. (Tutco Scientific no synergistic effect on ESCC. The Corp., Grand Junction, C0 (United mass ratio of sodium + silicate (mg/kg)

States)) to chloride (mg/kg) at the lower end of Soonsorino Oraanization: NRC: Nuclear the NRC RG 1.36 Acceptability Curve was Regulatory Comission, Washington, DC not sufficient to prevent ESCC using (United-States) the methods of this research.

Publication Date: Jul 1997 Reoort Number (s): NUREG/CR-6539 Order Number: TI9700739S

Title:

Environmentally assisted Abstract: _ The drip procedure from the cracking in light water reactors.

Standard Test Method for Evaluating the Semiannual report July 1996 -December Influence of Thermal Insulation on 1996 External Stress Corrosion Cracking Author (s)/ Editor (s): Chopra, 0.K. :

! . Tendency of Austenitic Stainless Steel Chung H.M. : Gavenda. D.J. (Argonne (ASTM C 692-95a) was used to research National Lab., IL (United States)) (and the effect of halogens and inhibitors others) on the External Stress Corrosion soonsorino Oraanization: NRC: Nuclear Cracking (ESCC) of Type 304 stainless Regulatory Commission. Washington, DC steel as it applies to Nuclear (United States)

Regulatory Comission Regulatory Guide Publication Date: Oct 1997 1.36, Nonmetallic Thermal Insulation Reoort Number (s):

for. Austenitic Stainless Steel. The NUREG/CR 4667-Vol.23: ANL--97/10 NUREG-1426 34

_ _ _ _ _ _ _ _ - . ~ . _ _ _ _ _ _ . _ _ . _ _ _ _ _ - _ _ _ _ . _

l l

l Compilation of Reports: 1994-1998 Order Number: TI98000789 Environments on Fatigue Design Curves Abstract: This report sunmarizes work of Carbon and Low-Alloy Steels.

performed by Argonne National Author (s)/ Editor (s): 0. K. Chopra and Laboratory on fatigue and W. J. Shack (Argonne National environmentally assisted cracking (EAC) Laboratory) in light water reactors from July 1996 Soonsorino Oroanization: NRC: Nuclear to December 1996. Topics that have Regulatory Conmission, Washington DC been investigated include (a) fatigue (United States) of carbon, low-alloy, and austen1 tic Publication Date: March 1998 stainless steels (SSs) used in reactor Reoort Number (s): NUREG/CR 6583: ANL-piping and pressure vessels, (b) 97/18 irradiation-assisted stress corrosion Abstract The ASME Boiler and Pressure cracking of Type 304 SS, (c) EAC of Vessel Code provides rules for the Alloy 600, and (d) characterization of construction of nuclear power plant residual stresses in welds of boiling components. Figures 1-9.1 through 1-9.6 water reactor (BWR) core shrouds by of Appendix I to Section 111 of the numerical models. Fatigue tests were Code specify fatigue design curves for i conducted on ferritic and austenitic structural materials. While effects of I SSs in water that contained various reactor coolant environments are not concentrations of dissolved oxygen to explicitly addressed by the design determine whether a slow strain rate curves, test data indicate that the applied during various portions of a Code fatigue curves may not always be tensile-loading cycle are equally adequate in coolant environments. This effective in decreasing fatigue life. report summarizes work performed by Slow-strain-rate-tensile tests were Argonne National Laboratory on fatigue conducted in simulated BWR water at 288 of carbon and lowalloy steels in light C on SS specimens irradiated to a low water reactor (LWR) environments. The fluence in the Halden reactor and the existing fatigue S-N data have been results were compared with similar data evaluated to establish the effects of l from a control-blade sheath and various material and loading variables neutron-absorber tubes irradiated in such as steel type, dissolved oxygen l BWRs to the same fluence level. level, strain range, strain rate.

l Crack-growth-rate tests were conducted temperature, orientation, and sulfur l on compact- tension specimens from a content on the fatigue life of these i

( low carbon content heat of Alloy 600 in steels. Statistical models have been '

high purity oxygenated water at 289 C. developed for estimating the fatigue S-Residual stresses and stress intensity N curves as a function of material.

l factors were calculated for BWR core loading, and environmental variables.

shroud welds. The results have iaen used to estimate the probability of fatigue cracking of reactor components. The different l

Title:

Effects of LWR Coolant methods for incorporating the effects  !

\. l

! 35 NUREG-1426 )

i i

4 Compilation of Reports - 1994-1998 of LWR coolant environments on the ASFE fluence in the Halden reactor and the Code fatigue design curves are results were compared with similar data presented. from a control-blade sheath and neutronabsorber tubes irradiated in BWRs to the same fluence level.

Title:

Environmentally Assisted Crack-growth-rate tests were conducted Cracking in Light Water Reactors on compact-tension specimens from Semiannual Report January 1997 - June several heats of Alloys 600 and 690 in 1997 low-DO. simulated pressurized water Author (s)/ Editor (s): Chopra 0.K. . reactor envirornents.

Chung. H.M. . Gavenda. 0.J. (Argonne National Lab., IL (United States)) (and others) Fracture Mechanics Testing Soonsorino Oroanization: NRC:

Washington DC (United States) and Analysis Publication Date: April 1998 Reoort Number (s): NUREG/CR-4667. Vol

Title:

Biaxial loading and 24 Abstract: lais report sumarizes work shallow-flaw effects on crack-tip constraint and fracture toughness performed by Argonne National Laboratory on fatigue and Author (s)/ Editor (s): Bass B.R. :

environmentally assisted cracking (EAC) Bryson, J.W. . Theiss. T.J. . Rao. M.C.

in light water reactors from January (0ak Ridge National Lab., TN (United 1997 to June 1997. Topics that have States))

been investigated include (a) fatigue Soonsorino Oroanization: NRC: Nuclear j

of carbon, low-alloy, and austenitic Regulatory Comission. Washington DC stainless steels (SSs) used in reactor (United States) l Publication Date: Jan 1994 piping and pressure vessels (b)

Recort Number (s): NUREG/CR-6132; l

1rradiationassisted stress corrosion cracking of Types 304 and 304L SS, and ORNL/TM--12498 (c) EAC of AUoys 600 and 690. Fatigue Order Number: TI94007598 Abstract: A program to develop and tests were conducted on ferritic and evaluate fracture methodologies for the austenitic SSs in water that contained various concentrations of dissolved assessment of crack-tip constraint oxygen (00) to determine whether a slow effects on fracture toughness of strain rate applied during various reactor pressure vessel (RPV) steels portions of a tensile loading cycle is has been initiated in the Heavy-Section equally effective in decreasing fatigue Steel Technology (HSST) Program.

life. Slow strain-rate-tensile tests Crack-tip constraint is an issue that were conducted in simulated boiling significantly impacts fracture mechanics technologies employed in water reactor (BWR) water at 288*C on SS specimens irradiated to a low safety assessment procedures for comercially licensed nuclear RPVs.

NUREG-1426 36

Compilation of Reports: 1994-1998 The focus of studies descricet herein (Naval Academy Annapolis, MD (United is on the evaluation of two States)): Link, R.E. (Naval Surface stressed-based methodologies for Warfare Center Annapolis, MD (United quantifying crack-tip constraint (i .e. , States))

J-Q theory and a micromechanical soonsorino Orcanization: NRC; Nuclear scaling model based on critical Regulatory Comission. Washington, DC stressed volumes) through applications (United States) to experimental and fractographic data. !vblication Date: Mar 1994 Data were utilized from single edge Eeoort Number (s): NUREG/CR-6051 notch bend (SENB) specimens and HSST- Order Number- T194009593 developed cruciform beam specimens that Abstract: Constraint has been an were tested in HSST shallow-crack and important consideration in fracture biaxial testing programs. Results from mechanics from the earliest work that I applications indicate that both the J-Q was done to develop the 1974 version of methodology and the micromechanical the ASTM Standard E399. O'Dowd and scaling model can be used successfully Shih (1991) have proposed that the to interpret experimental data from the difference in crack tip stress fields shallow- and deep-crack SENB specimen can be quantified in terms of a field tests. When applied to the uniaxially quantity that they have call Q. The Q and biaxially loaded cruciform quantity is a function of J, the crack specimens, the two methodologies showed shape and size, the structural some promising features, but also geometry, mode of loading and on the raised several questions concerning the level of deformation and can only be l interpretation of constraint conditions calculated from a high resolution ,

in the specimen based on near-tip elastic-plastic computational analysis. l stress fields. Fractographic data A similar, simpler, but more  !

taken from the fracture surfaces of the controversial approach has been SENB and cruciform specimens are used suggested by Betegon and Hancock

'to assess the relevance of stress-based (1991), who use the non singular term fracture characterizations to of the elastic, crack singularity

conditions at cleavage initiation solution, called the T-Stress, as a

. sites. Unresolved issues identified measure of elastic-plastic crack tip from these analyses require resolution constraint. The objective of this work l as part of a validation process for is to develop some upper shelf, biaxial loading applications. This elastic-plastic experimental results to report is designated as HSST Report No. attempt to inve3tigate the 142. applicability of the 0 and T stress parameters to the correlation of upper shelf initiation toughness and J

Title:

Effects of tensile loading on resistance curves. The first objective upper shelf fracture toughness was to obtain upper shelf J resistance 3

Author (s)/ Editor (s)- Joyce, J.A. curves, J[sub Ic], and tearing 37 NUREG-1426 l

i

- . - .-. - ~ - _ .. - - . - - - - - - - - - -

Compilation of Reports - 1994 1998 l resistance results for a range of lower transition region, cleavage I applied constraint. The J-0 and J-T fracture often occurs under conditions stress loci were developed and compared of large-scale yielding but without with the expectations of the O'Dowd and prior ductile crack extension. The Shih and the Betegon and Hancock increased toughness develops when analyses. Constraint was varied by plastic zones formed at the crack tip changing the crack length and also by interact with nearby specimen surfaces changing the mode of loading from which relaxes crack-tip constraint bending to predominantly tensile. The (stress triaxiality). In the principle conclusions of this work are mid-to-upper transition region. small that J[sub Ic] does not appear to be amounts of ductile crack extension dependent on T stress or 0 while the (often < 1-2 m) routinely precede material tearing resistance is termination of the J-[ Delta]a curve by dependent on T stress and 0. with the brittle fracture. Lar9e-scale tearing modulus increasing as yielding. coupled with small amounts of constraint decreases. ductile tearing, magnifies the impact of small variations in microscale material _ properties on the macroscopic

Title:

Numerical modeling of ductile fracture toughness which contributes to tearing effects on cleavage fracture the large amount scatter observed in toughness measured J[sub c]-values. Previous Author (s)/ Editor (s): Dodds. R.H. Jr. : work by the authors described a Tang M. (Univ. of Illinois, Urbana micromechanics fracture model to (United States)): Anderson. T.L. (Texas correct measured J[sub c]-values for A M Univ.. College Station. TX (United the mechanistic effects of large-scale States)) yielding. This new work extends the Soonsorino Oraanization: NRC: 000: model to also include the influence of Nuclear Regulatory Commission, ductile crack extension prior to Washington, DC (United States); cleavage. The paper explores Department of Defense. Washington, DC development of the new model, provides

, (United States) necessary graphs and procedures for its

! Publication Date: May 1994 application and demonstrates the l

Reoort Number (s): NUREG/CR-6162: effects of the model on fracture data VILU ENG--93 2014 sets for two pressure vessel steels Order Numbe.t1 T194015146 (AS338 and A515).

Abstract: Experimental studies demonstrate a significant effect of specimen size, a/W ratio and prior

Title:

Crack-speed relations inferred' ductile tearing on cleavage fracture from large single-edge notched toughness values (J[sub c)) measured in specimens of a 533 B steel 4 the ductile-to brittle transition Author (s)/ Editor (s); Schwartz, C.W.

region of ferritic materials. In the (Maryland Univ., College Park, MD NUREG 1426 38

Compilation of Reports: 1994-1998 (United States). Dept of Civil Author (s)/ Editor (s): Pennell. W.E.

Engineering) (0ak Ridge National Lab., TN (United Soonsorino Oroanization: NRC: Nuclear States))

Regulatory Comission, Washington DC Soonsorino Oroanization: NRC:

(United States) Washington DC (United States)

Publication Date: Jul 1994 Publication Date: September 1994 ReDort Number (s): NUREG/CR 5861: Reoort Number (s):

ORNL/ SUS--79-7778/9 NUREG/CR-4219-Vol.10 No.1:

EderNumber: TI94016708 ORNL/TM--9593/V10-N1 Abstract: A relationship between Abstract: The Heavy-Section Steel instantaneous crack-tip velocity [ dot Technology (HSST) Program is conducted a] dynamic stress-intensity factor for the Nuclear Regulatory Comission K[sub I], and temperature T for A 533 8 by Oak Ridge National Laboratory steel is estimated using dynamic crack (ORNL). The program focus is on the position us time data measured in a development and validation of series of very large-scale crack-arrest technology for the assessment of tests. The corresponding dynamic fracture-prevention margins in stress intensity us time history and comercial nuclear reactor pressure '

the dynamic-arrest toughness for each versels. The HSST Program is organized test are obtained from generation-mode in 12 tasks: (1) program management, elastodynamic analyses based on cubic (2) fracture methodology and analysis, polynomial fits to elastodynamic (3) mawrial characterization and analytical predictions based on the properties. (4) special technical proposed [ dot a]-K[sub 1]-T relation assistance, (5) fracture analysis are within 7% of experimentally computer programs (6) cleavage-crack measured arrested crack lengths and initiation. (7) cladding evaluations.

within 50% of measured arrest times. (S) pressurized-thermal shock These predictions within 50% of technology. (9) analysis methods l measured arrest times. These validation. (10) fracture evaluation I predictions represent significant tests. (11) warm prestressing, and (12) l improvements over results obtained biax161 loading effects. The program I using previous preliminary estimates of tasks have been structured to place l the [ dot a]-K[sub I] T relation for A emphasis on the resolution fracture l 533 B steel. The influence of issues with near-term licensing i nonlinear material behavior on the significance. Resources to execute the results is also evaluated. research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close

Title:

Heavy-Section Steel Technology contact is maintained with the sister Program Semiannual progress report, Heavy-Section Steel Irradiation Program October 1992 -- March 1993. Volume 10, at ORNL and with related research No. 1 programs both in the United States and 39 NUREG-1426 l

1

Compilation of Reports - 1994-1998 abroad. This report provides an testing procedure, and the experimental overview of principal developments in results form three specimens. The each of the 12 program tasks from yield strength of the weld material was determined to be 36% higher than the October 1992 to March 1993.

yield strength of the base material.

An irradiation-induced increase in Title- Preliminary assessment of the yield strength of the weld material fracture behavior of weld material in could result in a yield stress that full- thickness clad beams exceeds the upper limit where code Author (s)/ Editor (s): Keeney. J.A. ; curves are valid. The high yield Bass. B.R. . McAfee. W.J. : Iskander, strength for prototypic weld material S.K. (0ak Ridge National Lab.. TN may have implications for RPV (United States)) structural integrity assessments.

Soonsorino Oroanization: NRC: Nuclear Analyses :.,f the test data are Regulatory Commission. Washington, DC discussed, including comparisons of (United States) measured displacements with Publication Date: Oct 1994 finite-element analysis results.

Reoort Number (s): .NUREG/CR-6228: applications of toughness estimation ORNL/TM -12735 techniques, and interpretations of Order Number:. TI95001501 constraint conditions implied by Abstract: This report describes a stress-based constraint methodologies, testing program that utilizes Metallurgical conditions in the region full-thickness clad beam specimens to of the cladding heat-affected zone are

~ quantify fracture toughness for shallow proposed as a possible explanation for

, cracks in material for which the lower-bound fracture toughness I

metallurgical conditions are prototypic measured with one of the shallow-crack of those found in reactor pressure clad beam specimens. Fracture vessels (RPVs). The beam specimens are toughness data from the three clad beam fabricated from a section of an RPV specimens are compared with other wall (removed from a canceled nuclear shallow- and deep-crack uniaxial beam plant) that includes weld, plate, and and cruciform data generated previously clad material. Metallurgical factors from A 533 Grade B plate material.

potentially influencing fracture toughness for shallow cracks in the beam specimens include material

Title:

Cleavage behaviors in nuclear gradients due to welding and cladding vessel steels applications, as well as material Author (s)/ Editor (s): Irwin. G.R. :

inhomogeneities in welded regions due Zhang. X.J. (Univ. of Maryland. College to reheating in multiple weld passes. Park. MD (United States). Dept. of ,

A summary of the testing program Mechanical Engineering) includes a description of the specimen Soonsorino Oroanization: NRC: Nuclear geometry, material properties, the Regulatory Commission. Washington, DC i

NUREG-1426 40

_ _ . ._- _ _ _ _ - _ _ _ -___---_______-__-_________a

l l

l l Compilation of Reports: 1994-1998 i (United States) initiation loads in the wide plate Publication Date: Nov 1994 tests using two constraint assessment Reoort Number (s): NUREG/CR-6262: methodologies developed over the past ORNL/Sub- 79-7778/11 five years: the J 0 toughness locus Order Number: TI95004174 approach and the toughness scaling Abstract: Cleavage behaviors of approach based on a local failure l nuclear vessel steels in the transition criterion for cleavage. Both temperature range are reviewed. approaches demonstrate a significant Viewpoints are presented to assist loss of constraint in the understanding of cleavage crack speed, elastic-plastic fields ahead of the cleavage initiation, cleavage arrest, crack in the wide-plate specimens and the sensitivity of fracture caused by the inherent negative I toughness to constraint and T-stress of the shallow notch SE(T) temperature. The importance of high configuration. Moreover, the 25mm wide local stress elevations by high strain machined notch required for specimen rate is emphasized. This report is fabrication is shown to further reduce designated as HSST Report No. 149. constraint by introducing a traction free surface very near the crack tip.

l Both of these factors combined to

Title:

Constraint effects on fracture reduce near- tip stresses by 10%

initiation loads in HSST wide-plate below those of the small-scale tests yielding SSY (T-0), fields. This Author (s)/ Editor (s): Dodds , R.H. Jr. reduction places fracture results for (Illinois Univ. . Urbana. IL (United the wide-plate specimens within the J-States)) 0 toughness locus defined by fracture Soonsorino Orcanization: NRC: Nuclear toughness tests on the A5338 material Regulatory Cocrnission, Washington, DC and within the constraint corrected (United States) J[sub c] values defined by the Publication Date: Dec 1994 toughness scaling methodology.

Recort Number (s): NUREG/CR-6259:

VILU-ENG--94-2009: ORNL/TM--12796 Order Number: TI95005273

Title:

Validity limits in J-resistance Abstract: During the period curve determination: An assessment of l

1984--1987, researchers of the the J[sub M] Parameter. Volume 1 Heavy-Section Steel Technology program Author (s)/ Editor (s): Shih, C.F. . Liu.

l at the Oak Ridge National Laboratory X.H. (Brown Univ., Providence, RI l_ performed a unique series of fracture (United States). Div. of Engineering) l mechanics tests using exceptionally Soonsorino Oroanization: NRC: Nuclear large, SE(T) specimens (a/W-0.2) Regulatory Cocmission Washington, DC fabricated from a reactor pressure (United States) vessel material, A5338 Class 1 steel. Publication Date: Feb 1995 This study re-examines fracture Reoort Number (s): NUREG/CR-6264-Vol.1:

41 NUREG-1426

Compilation of Reports - 1994-1998 BMI--2181-Vol.1 Order Nurg t1 TI95008150 Title, Validity limits in J resistance Abstract: Significant advances in curve determination: A computational elastic-plastic fracture became approach to ductile crack growth under i possible with the introduction of large scale yielding conditions. Volume Rice's path independent J-integral 2 which has two physical meanings. Author (s)/ Editor (s): Shih. C.F. : Xia.

First, the J-integral is equivalent to L (Brown Univ., Providence. RI (United the energy release rate associated with States). Div. of Engineering):

a virtual crack advance. Secondly, J Hutchinson J.W. (Harvard Univ..

can be regarded as the strength of the Cambridge MA (United States). Div. of stress and strain singularity near a Applied Sciences) stationary crack tip. As a result of Soonsorina Oraanization: NRC: Nuclear several experimental studies, the Regulatory Connission. Washington. DC i J-integral is generally accepted as a (United States) valid parameter to characterize a Publication Date: Feb 1995 material's resistance to the onset of Reoort Number (s): NUREG/CR-6264-Vol.2:

crack growth under large-scale BMI--2181-Vol.2 yielding. Driven by simplicity and the Order Number: TI95008149 practical benefits that could be Abstract: In this report. Volume 2.

derived from a geometry and Mode I crack initiation and growth size-independent material resistance under plane strain cenditions in tough curve for large amounts of crack metals are computed using an growth J[sub M]. a modified J elastic / plastic continuum model which parameter was introduced. Initial accounts for void growth and results using J[sub M) were encouraging coalescence ahead of the crack tip, but subsequent studies did not support The material parameters include the the earlier results. The present stress-strain properties, along with computational study presented in Volume the parameters characterizing the 1 of this report investigates several spacing and volume fraction of voids in forms of this parameter, how they are material elements lying in the plane of derived and the validity of these the crack. For a given set of these parameters for small and large amounts parameters and a specific specimen, or of crack growth. It is concluded that component, subject to a specific neither J nor J[sub M] (nor any single loading, relationships among load, i parameter) can satisfactorily capture load-line displacement and crack the full range of near-tip fracture advance can be computed with no states. A discussion on the range of restrictions on the extent of plastic validity of J[sub M] is given in Volume deformation. Similarly, there is no

2. This work is relevant for assessing limit on crack advance, except that it structural integrity of nuclear must take place on the symetry plane pressure vessels and piping. ahead of the initial crack. Suitably NUREG-1426 42

Compilation of Reports: 1994-1998 defined measures of crack tip loading fracture toughness of reactor pressure intensity, such as those based on the vessel (RPV) steels have been completed J-integral, can also be computed, by the Heavy-Section Steel Technology thereby directly generating crack (HSST) Program. Objectives were to growth resistance curves. In this investigate effect of biaxial loading report, the model is applied to five on fracture toughness, quantify this specimen geometries which are known to effect through existing stress based.

give rise to significantly different dual-parameter, fracture toughness crack tip constraints and crack growth correlations, or propose and verify resistance behaviors. Computed results alternate correlations. A cruciform are compared with sets of experimental beam specimen with 2 D. shallow.

data for two tough steels for four of through-thickness flaw and a special the specimen types. Details of the loading fixture was designed and load, displacement and crack growth fabricated. Tests were performed using histories are accurately reproduced. biaxial loading ratios of 0:1 even when extensive crack growth takes (uniaxial). 0.6:1. and 1:1 place under conditions of fully plastic (equi-biaxial). Critical yielding. A description of material fracture-toughness values were resistance to crack initiation and calculated for each test. Biaxial subsequent growth is essential for loading of 0.6:1 resulted in a assessing structural integrity such as reduction in the lower bound fracture nuclear pressure vessels and piping. toughness of [approximately]l2% as compared to that from the uniaxial tests. The biaxial loading of 1:1

Title:

Biaxial loading effects on yielded two subsets of toughness fracture toughness of reactor pressure values: one agreed well with the vessel steel uniaxial data, while one was reduced by Author (s)/ Editor (s)- McAfee. W.J. : [approximately]43% when compared to the l Bass. B.R. : Bryson, J.W. Jr. . uniaxial data. Results were evaluated Pennell. W.E. (0ak Ridge National Lab., using J 0 theory and Dodds-Anderson TN (United States)) (D-A) micromechanical scaling model.

Soonsorino Oroanization: NRC: Nuclear The D-A model predicted no biaxial l Regulatory Connission. Washington, DC effect, while the J-Q method gave (United States) inconclusive results. When applied to Publication Date: Mar 1995 the 1:1 biaxial data, these constraint Reoort Number (s): NUREG/CR-6273: methodologies failed to predict the ORNL/TM--12866 observed reduction in fracture Order Number: TI95008768 toughness obtained in one experiment.

Abstract: The preliminary phases of a A strain- based constraint program to develop and evaluate methodology that considers the fracture methodologies for assessing relationship between applied biaxial crack-tip constraint effects on load, the plastic zone width in the i

l 43 NUREG-1426 i

Compilation of Reports - 1994-1998 crack plane, and fracture toughnest was increasing J0 R curve. This explains formulated and applied successfully to why the cracks eventually turn to the the data. Evaluation of this dual- fusion line in the pipe experiments. A parameter strain-based model led to the method of incorporating these results conclusion that it has the capability would be to use the weld metal J.R of representing fracture behavior of curve up to the fusion-line RPV steels in the transition region. steady-state J value. These results including the effects of out-of-plane may be more important to LBB analyses loading on fracture toughness. This than the ASME flaw evaluation r6 port is designated as HSST Report No. procedures, since there is more crack 150. growth with through-wall cracks in LBB analyses than for surface cracks in pipe flaw evaluations.

Title:

Stainless steel submerged arc weld fusion line toughness Author (s)/ Editor (s): Rosenfield A.R.

Title:

Heavy-Section Steel Technology

Held P.R. . Wilkowski, G.M. Program Semiannual progress report.

(Battelle. Columbus OH (United April-- September 1993. Volume 10.

States)) No. 2 Soonsorina Oraanization: NRC: Nuclear Author (s)/ Editor (s): Pennell, W.E.

Regulatory Comission. Washington DC (Oak Ridge National Lab., TN (United (United States) States))

Publication Date: Apr 1995 Soonsorina Orcanization: NRC: Nuclear Reoort Number (s): NUREG/CR-6251: Regulatory Comission. Washington DC BMI--2180 (United States) l Order Number: TI95010950 Publication Date: May 1995 l Abstract: This effort evaluated the Recort Number (s):

fracture toughness of austenitic steel NUREG/CR-4219-Vol.10 No.2:

submerged arc weld (SAW) fusion lines. ORNL/TM--9593/V10-N2 The incentive was to explain why cracks Order Number: T195012816 grow into the fusion line in many pipe Abstract: The Heavy-Section Steel tests conducted with cracks initially Technology (HSST) Program is conducted centered in SAWS. The concern was that for the Nuclear Regulatory Comission the fusion line may have a lower by Oak Ridge National Laboratory toughness than the SAW. It was found (ORNL). The program focuses on the that the fusion line. Ji. was greater development and validation of i

than the SAW toughness but much less technology for the assessment of

! than the base metal. Of greater fracture prevention margins in importance may be that the crack growth comercial nuclear reactor pressure resistance (J0-R) of the fusion line vessels. The HSST Program is organized appeared to reach a steady-state value. in 12 tasks: Program management, while the SAW had a continually fracture methodology and analysis.

NUREG 1426 44 1

i I

l Compilation of Reports: 1994-1998

{

material characterizations and Order Number: TI96002335 {

properties, special technical Abstract: The ASTM Standard Test assistance, fracture analysis computer Method for Plane-Strain Fracture l programs, cleavage-crack initiation. Toughness of metallic Materials cladding evaluations. (E399 90) restricts test specimen .

pressurized thermal shock technology, dimensions to insure the measurement of I analysis methods validation, fracture highly constrained fracture toughness  !

evaluation tests, warm prestressing, values (K[sub Ic]). These requirements and biaxial loading effects on fracture insure small-scale yielding (SSY) toughness. The program tasks have been conditions at fracture, and thereby the structured to emphasize the resolution validity of linear elastic fracture fracture issues with near-term mechanics. Recently, Dodds and licensing significance. Resources to Anderson have proposed a less execute the research tasks are drawn restrictive size requirement for from ORNL with subcontract support from cleavage fracture toughness measured in universities and other research terms of the J-integral (J[sub c]), as laboratories. Close contact is given by a, b. B [ge] 200 J[sub maintained with the sister c]/[ sigma][sub 0]. The size Heavy-Section Steel Irradiation Program requirement proposed by Dodds and at ORNL and with related research Anderson increases the applicability of programs both in the United States and fracture toughness experiments by abroad. This report provide s an expanding the range of conditions over overview of principal developments in which fracture toughness data meeting each of the 12 program tasks from April SSY conditions can be reliably

-- September 1993, measured. This investigation compares the proposed size requirement with that of ASTM Standard Test Method E399 and.

Title:

Size and deformation limits to by comparison with published maintain constraint in K[sub Ic] and experimental data for various alloys, J[sub c) testing of bend specimens provides validation of the new Author (s)/ Editor (s): Koppenhoefer,- requirements.

K.C. . Dodds, R.H. Jr. (Illinois Univ. ,

Urbana IL (United States). Dept. of Civil Engineering)

Title:

Heavy section steel technology Soonsorino Oroanization: NRC: 000: program: Semiannual progress report, Nuclear Regulatory Comission. October 1993--March 1994. Volume 11.

Washington, DC (United States): No. 1 Department of Defense. Washington, DC Author (s)/ Editor (s): Pennell, W.E.

(United States) (Oak Ridge National Lab., TN (United Publication Date: Oct 1995 States))

Reoort Number (s): NUREG/CR-6191: Soonsorino Oroanization: NRC: Nuclear VILU ENG--94 2002 Regulatory Commission, Washington, DC 45 NUREG-1426

_ ~ -_ ._

Compilation of Reports - 1994-1998 (United States)

Publication Date: Nov 1995

Title:

Application of fracture Reoort Number (s): toughness scaling models to the NUREG/CR-4219-Vol.11-No.1; ductile-to- brittle transition ORNL/TM--9593/V11-N1 A# hor (s)/ Editor (s): Link, R.E. (Naval Order Number: TI96003353 Surface Warfare Center Annapolis, MD Abstract: The Heavy-Section Steel (United States)); Joyce, J.A. (Naval Technology (HSST) Program is conducted Academy, Annapolis. MD (United States))

for the US Nuclear Regulatory Soonsorino Oroanization: NRC: Nuclear Comission (NRC) by Oak Ridge National Regulatory Commission. Washington, DC Laboratory (ORNL). The Program focus (United States) is on the development and validation of Publication Date: Jan 1996 technology for the assessment Of Reoort Number (s): NUREG/CR-6279 fracture-prevention margins in Order Number: TI96005815 commercial nuclear reactor pressure Abstract: An experimental vessels. The HSST Program is organized investigation of fracture toughness in in seven tasks: (1) program management the ductile- brittle transition (2) constraint effects analytical range was conducted. A large number of development and validation. (3) ASTM A533, Grade B steel, bend and evaluation of cladding effects. (4) tension specimens with varying crack ductile to cleavage fracture mode lengths were tested throughout the conversion, (5) fracture analysis transition region. Cleavage fracture methods development and applications, toughness scaling models were utilized (6) material Property data and test to correct the data for the loss of methods, and (7) integration of results constraint in short crack specimens and into a state-of-the-art methodology. tension geometries. The toughness The program tasM have been structured -scaling models were effective in to place emphasis on the resolution reducing the scatter in the data, but fracture issues with near-term tended to over-correct the results for licensing significance. Resources to the short crack bend specimens. A execute the research tasks are drawn proposed ASTM Test Practice for from ORNL with ~!bcontract support from Fracture Toughness in the Transition universities and other research Range, which employs a master curve laboratories. Close contact is concept, was applied to the results.

maintained with the sister Heavy- The proposed master curve over Section Steel Irradiation Program at predicted the fracture toughness in the ORN. and with related research programs mid-transition and a modified master both in the United States and abroad. curve was developed that more This report provides an overview of accurately modeled the transition principal developments in each of the behavior of the material. Finally, the seven program tasks from October modified master curve and the fracture 1993--March 1994. toughness scaling models were combined NUREG-1426 46

.~ . . .. - - - - - _ _ . - - - -

Compilation of Reports: 1994-1998 to predict the as- measured fracture methods development and applications.

toughness of the short crack bend and (6) material property data and test the tension specimens. It was shown methods, and (7) integration of that when the scaling models over results. The program tasks have been correct the data for loss of structured to place emphasis on the constraint. they can also lead to resolution fracture issues with non conservative estimates of the near-term licensing significance.

increase in toughness for low Resources to execute the research tasks i constraint geometries. are drawn from ORNL with subcontract support from universities and other research laboratories. Close tentact

Title:

Heavy-Section Steel Technology is maintained with the sister Program: Semiannual progress report for Heavy-Section Steel Irradiation (HSSI)

April--September 1994. Volume 11. Program at ORNL and with related Number 2 research programs both in the US and Author (s)/ Editor (s): Pennell, W.E. abroad. This report provides an (Oak Ridge National Lab., TN (United overview of principal developments in States)) each of the seven program tasks from Soonsorina Oraanization: NRC: Nuclear April 1994 to September 1994.

Regulatory Commission, Washington, DC (United States)

Publication Date: Apr 1996

Title:

Numerical investigation of 3-D Reoort Number (s): ccostraint effects on brittle fracture i NUREG/CR-4219-Vol.11-No.2: in SE(B) and C(T) specimens ORNL/TM--9593/V11 N2 Author (s)/ Editor (s): Nevalainen M.

Order Number: TI96010062 (Valtion Teknillinen Tutkimuskeskus, Abstract: The Heavy-Section Steel Espoo (Finland)); Dodds, R.H. Jr. i Technology (HSST) Program is conducted (Illinois Univ., Urbana IL (United l for the Nuclear Regulatory Commission States). Dept. of Civil Engineering)

(NRC) by Oak Ridge National Laboratory soonsorino Oraanization: NRC: Nuclear (ORNL) The program focus is on the Regulatory Conmission, Washington, DC development and validation of (United States) technology for the assessment of Publication Date: Jul 1996 fracture-prevention margins in Reoort Number (s): NUREG/CR-6317; commercial nuclear reactor pressure UILU-ENG--95-2001  ;

vessels. The HSST Program is organized Order Number: TI96013827  ;

in seven tasks: (1) program management. Abstract: This investigation employs (2) constraint effects analytical 3-D nonlinear finite element analyses development and validation, (3) to conduct an extensive parametric evaluation of cladding effects. (4) evaluation of crack front seess ductile-to-cleavage fracture-mode triaxiality for deep notch SE(B) and conversion (5) fracture analysis C(T) specimens and shallow notch SE(B)

~

47 NUREG-1426

I l

Compilation of Reports - 1994-1998 specimens, with and without side interaction of impact loading, grooves. Crack front conditions are inelastic material deformation and rate characterized in terms of J-Q sensitivity with the goal of improving trajectories and the constraint scaling the interpretation of ductile fracture model for cleavage fracture toughness toughness values measured under dynamic proposed previously by Dodds and loading. Three-dimensional, nonlinear Anderson. The 3-D computational dynamic analyses are performed for results imply that a significantly less SE(B) fracture specimens (a/W - 0.5.

strict size / deformation limit, relative 0.15. 0.0725) subjected to impact to the limits indicated by previous loading. Loading rates obtained in plane-strain computations, is needed to conventional drop tower tests (impact maintain small- scale yielding load-line velocities of [ approx]6 conditions at fracture by a stress- m/sec) are applied in the analyses.

controlled, cleavage mechanism in deep Strains at key locations on the notch SE(B) and C(T) specimens. specimens and the support reactions Additional new results made available (applied load) are extracted from the from the 3-D analyses also include analyses to assess the accuracy c' revised [ eta]-plastic factors for use static formulas commonly used to in experimental studies to convert estimate applied J values. Inertial measured work quantities to thickness effects on the applied J are quantified average and maxPt.m (local) J-values by examining the acceleration component over the crack front. of J evaluated through a domain integral procedure.

Title:

Strain rate and inertial effects on impact loaded single-edge

Title:

Heavy-section steel technology l

notch bend specimens program. Semiannual progress report Author (s)/ Editor (s): Vargas. P.M. ; October 1994--March 1995 Dodds. R.H. Jr. Author (s)/ Editor (s): Pennell W.E.

Soonsorino Orcanization: NRC; Nuclear Sggnsorina Oraanization: NRC; Nuclear l Regulatory Commission. Washington, DC Regulatory Comission. Washington, DC j (United States) (United States) l Publication Date: Jun 1996 Publication Date: Jul 1996 Reoort Number (s): NUREG/CR-6375; Report Number (s):

i VILU-ENG--94-2018 NUREG,CR-4219-Vol.12-No.1; Order Number; TI96013209 ORNL/Ti!--9593 f

Abstract: When the severity of impact Order Number: TI96013666 l

loads is sufficient to produce large Abstract: The Heavy-Section Steel l inelastic deformations, the assessment Technology (HSST) Program is conducted of crack-tip conditions must include for the Nuclear Regulatory Commission the effects of plasticity, strain rate (NRC) by Oak Ridge National Laboratory and inertia. This work examines the ('RNL). The program focus is on the NUREG-1426 48 l

l l

l

l I

Compilation of Reports: 1994-1998 development and validation of Regulatory Commission. Washington. OC technology for the assessment of (United States) fracture-prevention margins in Publication Date: Nov 1996 conmercial nuclecr reactor pressure ReDort Number (s): NUREG/CR-6460:

vessels. The HSST Program is organized ORNL/TM--13207 in seven tasks: (1) program management Order Number: TI97001357 (2) constraint effects analytical Abstract: A summary of Phase II of the development and validation. (3) Project for FALSIRE is presented.

evaluation of cladding effects. (4) FALSIRE was created by the Fracture ductile-to-cleavage fracture-mode Assessment Group (FAG) of the conversion, (5) fracture analysis OECD/NEA's Committee on the Safety of methods development and applications. Nuclear Installations (CNSI) Principal (6) material property data and test Working Group No. 3. FALSIRE I in 1988 methods, and (7) integration of assessed fracture methods through '

results. The program tasks have been interpretive analyses of 6 large- ,

structured to place emphasis on the scale fracture experiments in reactor resolution of fracture issues with pressure vessel (RPV) steels under near-term licensing significance. pressurized- thermal-shock (PTS)

Resources to execute the research tasks loading. In FALSIRE II, experiments are drawn from ORNL with subcontract examined cleavage fracture in RPV support from universities and other steels for a wide range of materials, I research laboratories. Close contact crack geometries, and constraint and is maintained with the sister loading conditions. The cracks were Heavy-Section Steel Irradiation Program relatively shallow, in the transition at ORNL and with related research temperature region. Included were programs both in the United States and cracks showing either unstable abroad. This report provides an extension or two stages of extensions overview of principal developments in under transient thermal and mechanical each of the seven program tasks from loads. Crack initiation was also October 1994-March 1995, investigated in connection with clad surfaces and with biaxial load. Within FALSIRE II, comparative assessments

Title:

CSNI Project for Fracture were performed for 7 reference fracture Analyses of Large-Scale International experiments based on 45 analyses Reference Experiments (FALSIRE II) received from 22 organizations Author (s)/ Editor (s): Bass. B.R. : representing 12 countries. Temperature Pugh C.E. . Keeney, J. (Dak Ridge distributions in thermal shock loaded National Lab., TN (United States)); samples were approximated with high Schulz, H. : Sievers, J. (Gesellschaft accuracy and small scatter bands.

fue- Anlagen- und Reaktorsicherheit Structural response was predicted i (GRS) mbH, Koeln (Gemany)) reasonably well; discrepancies could Soonsorina Oroanization: NRC: Nuclear usually be traced to the assumed

)

49 NUREG-1426

- -,--m . - - - . - u,,

i Compilation of Reports - 1994-1998 material models and approximated. Resources to execute the research tasks material properties. Almost all are drawn from ORNL with subcontract participants elected to use the finite support from universities and other element method. research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation Program

Title:

Heavy-section steel technology at ORNL and with related research program. Semiannual progress report, programs both in the United States and April-- September 1995 Vol. 12, No. abroad. This report provides an 2 overview of principal developments in Author (s)/ Editor (s): Pennell. W.E. each of the seven program tasks from Soonsorino Orcanization: NRC; Nuclear April 1995 to September 1995.

Regulatory Commission, Washington, DC (United States)

Publication Date: Jan 1997

Title:

Ductile fracture toughness of Report Number (s): modified A 302 Grade B Plate materials.

NUREG/CR-4219-Vol.12-No.2: data analysis. Volume 1 ORNL/TM--9593/V12 N2 Author (s)/ Editor (s): McCabe, D.E. .

Order Number: TI97002993 Manneschmidt, E.T. ; Swain, R.L.

Abstract: The Heavy-Section Steel Soonsorino Oroanization: NRC; Nuclear j Technology (HSST) Program is conducted Regulatory Commission. Washingtor, DC for the Nuclear Regulatory Commission (United States) by Oak Ridge National Laboratory Publication Date: Jan 1997 l (ORNL). The program focus is on the Reoort Number (s): NUREG/CR-6426-Vol.1; development and validation of ORNL--6892-Vol.1 technology for the assessment of Order Number: TI97004262 fracture-prevention margins in Abstract: The goal of this work was to comercial nuclear reactor pressure develop ductile fracture toughness data vessels. The HSST Program is organized in the form of J-R curves for modified in seven tasks: (1) program management. A302 grade B plate materials typical of (2) constraint effects analytical those used in reactor pressure vessels.

development and validation. (3) A previous experimental study on one evaluation of cladding effects, (4) heat of A302 grade B plate showed ductile-to-cleavage fracture-mode decreasing J-R curves with increased conversion (5) fracture analysis specimen thickness. This methods development and applications, characteristic has not been observed in (6) material property data and test tests made on recent production methods, and (7) integration of materials of A533 grade B and A508 results. The program tasks have been class 2 pressure vessel steels. It was structured to place emphasis on the unknown if the departure from norm for I resolution of fracture issues with the material was a generic near-term licensing significance. characteristic for all heats of A302 NUREG-1426 50

Compilation of Reports: 1994-1998 grade B steels or unique to that

Title:

Ductile fracture toughness of particular plate. Seven heats of modified A 302 grade B plate materials.

modified A302 grade B steel and one Volume 2 heat of vintage A533 grade B steel were Author (s)/ Editor (s): McCabe. D.E. .

tested for chemical content, tensile Manneschmidt, E.T. . Swain. R.L.

properties, Charpy transition Soonsorina Orcanization: NRC: Nuclear temperature curves, drop-weight Regulatory Commission, Washington, DC nil-ductility transition (NDT) (United States) temperature, and J- R curves. Publication Date: Feb 1997 Tensile tests were made in the three Reoort Number (s): NUREG/CR-6426-Vol.2:

principal orientations and at four ORNL--6892/V2 temperatures, ranging from room Order Number: TI97003758 temperature to 550F. Charpy V-notch Abstract: The objective of this work transition temperature curves were was to develop ductile fracture obtained in longitudinal, transverse, toughness data in the form of J-R and short transverse orientations. J-R curves for modified A 302 grade B plate curves were made using four specimen materials typical of those used in sizes (1/2T, IT, 2T and 4T). The fabricating reactor pressure vessels.

fracture mechanics-based evaluation A previous experimental study at method covered three test orientations Materials Engineering Associates (MEA) and three test temperatures (80, 400, on one particular heat of A 302 grade B and 550F). However, the coverage of plate showed decreasing J-R curves with these variables was contingent upon the increased specimen thickness. This amount of material provided. characteristic has not been observed in Drop-weight NDT temperature was numerous tests made on the more recent determined for the T-L orientation production mi.terials of A 533 grade B only. None of the heats of modified and A 508 class 2 pressure vessel A302 grade B showed size effects of any steels. It was unknown if the  ;

consequence on the J-R curve behavior. departure from norm for the MEA Crack orientation effects were present, material was a generic characteristic but none were severe enough to be for all heats of A 302 grade B steels

)

reported as atypical. A test or just unique to that one particular temperature increase from 180 to 550F plate. Seven heats of modified A 302 produced the usual loss in J R curve grade B steel and one heat of vintage A fracture toughness. Generic J-R curves 533 grade B steel were provided to this and curve fits were generated to project by the General Electric Company represent each heat of material. This of San Jose, California. All plates volume deals with the evaluation of were tested for chemical content. l data and the discussion of technical tensile properties, Charpy transition l findings. 8 refs. ,18 figs. , 8 tabs. temperature curves. drop-weight nil-ductility transition (NDT) temperature, and J-R curves. Tensile 51 NUREG-1426

Compilation of Reports - 1994-1998 tests were made in the toree principal Technology (HSST) Program is conducted orientations and at four temperatures. for the US Nuclear Regulatory ranging from room temperature to Commission (NRC) by the Oak Ridge 550[ degrees]F (288[ degrees]C). Charpy National Laboratory (0RNL). The V-notch transition temperature curves program focus is on the development and were obtained in longitudinal. validation of technology for the transverse, and short transverse assessment of fracture-prevention orientations. J-R curves were made margins in commercial nuclear reactor using four specimen sizes (1/2T, IT, pressure vessels. The HSST Program is 2T and 4T). None of the seven heats organized in seven tasks: (1) program of modified A 302 grade showed size management (2) constraint effects effects of any consequence on the J R analytical development and validation, curve behavior. Crack orientation (3) evaluation of cladding effects. (4) effects were present, but none were ductile to cleavage fracture mode severe enough to be reported as conversion. (5) fracture analysis atypical. A test temperature increase methods development and applications, from 180 to 550[ degrees]F (82 to (6) material property data and test 288(degrees]C) produced the usual loss methods, and (7) integration of results in J-R curve fracture toughness. into a state-of-the-art methodology.

Generic J-R curves and mathematical The program tasks have been structured curve fits to the same were generated to place emphasis on the resolution to represent each heat of material. fracture issues with near-term This volume is a compilation of all licensing significance. Resources to data developed. execute the research tasks are drawn from ORNL with subcontract support from universities and other research

Title:

Heavy-section steel technology laboratories. Close contact is program: Semiannual progress report for maintained with the sister October 1995--March 1996. Volume 13, Heavy-Section Steel Irradiation Program Number 1 at ORNL with related research programs Author (s)/ Editor (s): Pennell, W.E. both in the US and abroad. This report (Dak Ridge National Lab., TN (United provides an overview of principal States)) developments in each of the seven Soonsorino Oraanization: NRC: Nuclear program tasks from October 1995--

Regulatory Conmission Washington, DC March 1996.

(United States)

Publication Date: Sep 1997 Reoort Number (s): Non Destructive Examination NUREG/CR-4219-Vol.13 No.1:

ORNL/TM--9593/V13 N1 Order Number: TI98000428

Title:

Evaluation of computer-based Abstract: The Heavy-Section Steel ultrasonic inservice inspection systems NUREG-1426 52 1 v

Compilation of Reports: 1994-1998 l Author (s)/ Editor (s): Harris, R.V. Jr.

Title:

Feasibility of developing

Angel, L.J. : Doctor. S.R. . Park, risk-based rankings of pressure W.R. : Schuster, G.J. : Taylor, T.T. boundary systems for inservice ,

(Pacific Northwest Lab., Richland, WA inspection l (United States)) Author (s)/ Editor (s): Vo, T.V. : Smith.

Soonsorina Oraanization: NRC: Nuclear B.W. . Simonen, F.A. : Gore, B.F.  ;

Regulatory Connission Washington. DC Soonsorina Orcanization: NRC: Nuclear (United States) Regulatory Commission Washington DC l Publication Date: Mar 1994 (United States)

. -Reoort Number (s): NUREG/CR-5985: Publication Date: Aug 1994 PNL--8919 Report Number (s): NUREG/CR-6151:

Order Number: TI94010311 PNL--8912 1 Abstract: This report presents the Order Number: TI94018068 principles, practices, terminology, and Abstract: The goals of the Evaluation technology of computer-based ultrasonic and Improvement of Non-destructive testing for inservice inspection Examination Reliability for the (UT/ISI) of nuclear power plants, with In-service Inspection of Light Water extensive use of drawings, diagrams, Reactors Program sponsored by the and LTT iwages. The presentation is Nuclear Regulatory Commission at

. technical but assumes limited specific Pacific Northwest Laboratory (PNL) are knowledge of ultrasonics or computers, to (1) assess current ISI te hniques The report is divided into 9 sections and requirements for all pressure

co'/ering conventional LTT, boundary systems and components. (2) ccmputer-based LTT, and evaluation determine if improvements to the methodology. Conventional LTT topics requirements are needed, and (3) if include coordinate axes, scanning, necessary, develop recommendations for instrument operation, RF and video revising the applicable ASME Codes and signals, and A , B , and C-scans. regulatory requirements. In evaluating Computer-based topics include sampling, approaches that could be used to digitization, signal analysis, image provide a technical basis for improved presentation, SAFI, ultrasonic inservice inspection plans, PNL has

, holography, transducer arrays, and data developed and applied a method that interpretation. An evaluation uses results of probabilistic risk methodology for computer-based LTT/ISI assessment (PRA) to establish piping systems is presented, including system ISI requirements. In the PNL

questions, detailed procedures, and program, the feasibility of generic ISI test block designs. Brief evaluations requirements is being addressed in two of several computer-based LTT/ISI phases. Phase I involves identifying systems are given

supplementory and prioritizing the systems most volumes will provide detailed relevant to plant safety. The results evaluations of selected systems. of these evaluations will be later

consolidated into requirements for

53 NUREG-1426

Compilation of Reports - 1994-1998

) comprehensive inservice inspection of detection, digitizing, imaging, data nuclear power plant components that interpretation, operator interaction, will be developed in Phase II. This data handling, and record-keeping. It report presents Phase I evaluations for includes a general description, a eight selected plants and attempts to review checklist, and detailed results compare these PRA-based inspection of all tests performed, priorities with current ASME Section XI requirements for Class 1, 2 and 3 systems. These results show that there

Title:

A pilot application of are generic insights that can be risk based methods to establish extrapolated from the selected plants in-service inspection priorities for to specific classes of light water nuclear components at Surry Unit I reactors, Nuclear Power Station Author (s)/ Editor (s): Vo, T. : Gore, B.

. Simonen, F. . Doctor S. (Pacific .

Title- Review of P-scan computer-based Northwest Lab., Richland, WA (United ultrasonic inservice inspection system. States))

Supplement 1 Soonsorina Organization: NRC: Nuclear

' Author (s)/ Editor (s): Harris, R.V. Jr. Regulatory Commission, Washington, DC

Angel, L.J. (Pacific Northwest Lab., (United States)

.Richland, WA (United States)) Publication Date: Aug 1994 Soonsorina Oroanization: NRC: Nuclear Reoort Number (s): NUREG/CR-6181:

Regulatory Commission. Washington, DC PNL--9020 (United States) Order Number: T194018066 Publication Date: Dec 1995 Abstract: As part of the Reoort Number (s):

Nondestructive Evaluation Reliability NUREG/CR-5985-Suppl.1: Program sponsored by tne US Nuclear PNL--8919-Suppl.1 Regulatory Comission, the Pacific  !

Order Number: TI96008319 Northwest Laboratory is developing a Abstract: This Supplement reviews the method that uses risk-based approaches P-scan system, a computer-based to establish in-service inspection ultrasonic system used for inservice plans for nuclear power plant inspection of piping and other components. This method uses components in nuclear power plants, probabilistic risk assessment (PRA)

The Supplement was prepared using the results and Failure Modes and Effects methodology described in detail in Analysis (FEMA) techniques to identify Appendix A of NUREG/CR-5985, and is and prioritize the most risk-important based on one month of using the system systems and components for inspection.

in a laboratory. This Supplement The Surry Nuclear Power Station Unit I describes and characterizes: computer was selected for pilot applications of system, ultrasonic components, and this method. The specific systems mechanical components; scanning, addressed in this report are the NUREG-1426 54 ,

t Compilation of Reports: 1994-1998

! reactor pressure vessel, the reactor to provide significant enhancements to I coolant, the low pressure injection, the inspection of materials used in US 3

and the auxiliary feedwater. The nuclear power plants. This report ,

j results provide a risk-based ranking of provides guidelines for the l

components within these systems and implementation of SAFT-UT technology l relate the target risk to target and shows the results from its l l failure probability values for application. An overview of the l

- individual components. These results development of SAFT-VT is provided so I will be used to guide the development that the reader may become familiar l

! ~ of improved inspection plans for with the technology. Then the basic nuclear power plants. To develop fundamentals are presented with an

! inspection plans, the acceptable level extensive list of references. A

i. of risk from structural failure for comprehensive operating procedure, important systems and components will which is used in conjunction with the be apportioned as a small fraction SAFT-UT field system developed by (i .e., 5%) of the total PRA-estimated Pacific Northwest Laboratory (PNL),

j risk for core damage. This process provides the recipe for both SAFT data will determine target (acceptable) risk acquisition and analysis. The l and target failure probability va'ues specification for the hardware j for individual components. Insk:ction implementation is provided for the requirements will be set at levels to SAFT-VT system along with a description

assure that acceptable failure of the subsequent developments and i i probabilistics are maintained. improvements. One development of I l technical interest is the SAFT real

time processor. Performance of the

Title:

Real-time 3-D SAFT-UT system real-time processor is impressive and i j evaluation and validation Title comparison is made of this dedicated l j Augmentation: Synthetic Aperture parallel processor to a conventional Focusing Technique for Ultrasonic computer and to the newer high-speed l Testing computer architectures designed for Author (s)/ Editor (s): Doctor, S.R. : image processing. Descriptions of Schuster, G.J. : Reid L.D. : Hall, other improvements, including a robotic T.E. (Pacific Northwest National Lab., scanner 'are provided. Laboratory d

Richland WA (United States)) parametric and application studies, j Soonsorina Orcanization: NRC: Nuclear performed by PNL and not previously l Regulatory Commission, Washington, DC reported, are discussed followed by a (United States) section on field application work in Publication Date: Sep 1996 which SAFT was used during inservice e Reoort Number (s): NUREG/CR 6344: inspections of operating reactors.

PNNL--10571

! Order Number: TI97001474 i Abstract: SAFT-UT technology is shown

Title:

A pilot application of j

i I. 55 NUREG-1426 4

h 3-d 4

,. -,c-n. -,, ,,. - , - . --,-n- . - - - , ~n~ , - - , , . . - , . . . , , - . - , , , . , .

Compilation of Reports - 1994-1998 risk-informed methods to establish addressed in this report are the inservice inspection priorities for auxiliary feedwater, the low pressure nuclear components at Surry Unit 1 injection, and the reactor coolant Nuclear Power Station. Revision 1 systems. The results provide a Author (s)/ Editor (s): Vo T.V. : Phan, risk-informed ranking of components H.K. : Gore. B.F. . Simonen F.A. ; within these systems.

Doctor. S.R. (Pacific Northwest National Lab., Richland. WA (United States))

Title:

An Evaluation of Human Factors soonsorino Orcanization: NRC: Nuclear Research for Ultrasonic Inservice Regulatory Commission, Washington. DC Inspection (United States) Author (s)/ Editor (s): 0.J. Pond, 0.T.

Publication Date: Feb 1997 Donohoo. R.V. Harris, Jr. (Pacific Reoort Number (s): NUREG/CR 6181-Rev.1: Northwest National Laboratory)

PNNL--9020-Rev.1 Soonsorino Oraanization: NRC:

Order Number: TI97004344 Washington DC (United States)

Abstract: As part of the Publication Date: March 1998  ;

Nondestructive Evaluation Reliad lity Reoort Number (s): NUREG/CR- 6605:  !

Program sponsored by the US Nuclear PNNL-11797 Regulatory Commission, the Pacific Abstract: This work was undertaken to l Northwest National Laboratory has determine if human factors research has developed risk-informed approaches for yielded information applicable to inservice inspection plans of nuclear upgrading requirements in ASME Boiler power plants. This method uses and Pressure Vessel Code Section X1, probabilistic risk assessment (PRA) improving methods and techniques in results to identify and prioritize the Section V, and/or suggesting relevant most risk-important components for research. A preference was established inspection. The Surry Nuclear Power for information and recommendations i Station Unit I was selected for pilot which have become accepted and standard application of this methodology. This practice.

report, which incorporates more recent plant-specific information and improved Manual Ultrasonic Testing / Inservice risk-informed methodology and tools, is Inspection (UT/ISI) is a complex task

-Revision 1 of the earlier report subject to influence by dozens of (NUREG/CR 6181). The methodology variables. This review frequently discussed in the original report is no revealed equivocal findings regarding longer current and a preferred affects of environmental variables as methodology is presented in this well as repeated indications that Revision. This report. NUREG/CR-6181, inspection performance may bo more, and Rev. 1, therefore supersedes the more reliably, influenced by tho earlier NUREG/CR-6181 published in workers' social environment, including l August 1994. The specific systems managerial practices, than by other NUREG-1426 56

l Compilation of Reports: 1994-1998 situational variables. Also of fracture analyses for circumferentially l significance are each inspector's cracked nuclear oiping with cracks l relevant knowledge, skills, and sizes typically found during in-service l abilities, and determination of these flaw evaluations, Progress is the l is seen as a necessary first step in through-wall-cracked pipe efforts upgrading requirements, methods, ant involved (1) verification of techniques as well as in focusing deformation plasticity under research in support of such programs. nonproportional loading, (2) evaluation i While understanding tho effects and of the effect of weld metal strength on  ;

mediating mechanisms of the variables various J-estimation schemes, and (3) l 1mpacting inspection performance is a development of new GE/EPRI functions, worthwhile pursuit for researchers, Surface-cracked pipe evaluations initial improvements in industrial involved (1) material characterization UT/IS performance may be achieved by of B W C- Mn-Mo submerged arc weld  !

implementing practices already known to metal, and (2) 3D finite-element mesh mitigate the effects of potentially refinement study. The toughness of the adverse conditions. bimetallic weld fusion line was evaluated and showed unusual fracture behavior based on the results of the Piping Charpy tests. The dynamic strain aging J-R tests confirmed the screening criterion developed earlier in the

Title:

Short cracks in piping and program. The results from this program piping wells to date necessitated several additional Author (s)/ Editor (s): Wilkowski, G.M. , efforts. These were ir H ated and have Brust, F. . Francini, R. (Battelle, been reported here. Presentation of Columbus, OH (United State.s)) (and the results from this program to the ASME Section XI Pipe Flaw Evaluation others)

Working Group is also summarized here.

Soonsorino Oraanization: NRC: Nucleat Regulatory Commission, Washington, DC (United States)

Publication Date: Mar 1994

Title:

Review of Elastic Stress and Reoort Number (s):

Fatigue-to-Failure Data for Branch NUREG/CR-4599-Vol.3-No.2:

Connections and Tees in Relation to ASME Design Criteria for Nucledr Power BMI--2173-Vol.3-No.2 Piping Systems Order Number: TI94008267 Author (s)/ Editor (s): E.C.Rodabaugh, Abstract: This is the sixth semiannual report of the US Nuclear Regulatory S.E. Moore, R.C. Gwaltney Commission's 4-year research program So nsorina Oraanization- NRC:

Short Cracks in Piping and Piping Washington DC (United States)

Welds which began in March 1990. The Publication Date: May 1994 objective is to verify and improve Reoort Number (s): NUREG/CR-57 NUREG-1426

Compilation of Reports - 1994-1998 5359:0RNL/TM-11152 amplitude cycles. The report proposes Abstract: This is the third in a addidonal analytical and experimental series of reports on the state-of-the work, art design guidance for piping system branch connections and tees provided by Section 111 of the ASME Boiler and

Title:

Evaluation and refinement of Pressure Vessel Code. The other reports leak-rate estimation models covert /i pnmary or limit-loads and Author (s)/ Editor (s): Paul. 0.0. ,

nozzle t'exibility. The principal /M ad, J. . Scott P.M. , Flanigan, objective of this report, as with the L.F. . Wilkowski, G.M. (Battelle.

others, was to identify and collect the Colunbus. OH (United States))

pertinent literature on the subject and Soonsorinc Oraanization: NRC: Nuclear to identify needed improvements in the Regulatory Comission. Washington, DC design methods and criteria of the Code (United States) based on the evaluation of the Publication Date: Jun 1994 available information. This report does Reoort Number (s): NUREG/CR-5128-Rev.1:

not propose changes in the design BMI--2164-Rev.1 pracedure of the Code. This report Order Number: TI94015006 discusses the evaluation of stresses in Abstract: Leak-rate estimation models branch connections and tees, are important elements in developing a correlation of these stresses with leak- beforebreak methodology in fatigue failures, and the Code rules piping integrity and safety analyses.

for protection against fatigue failure Existing thermalhydraulic and in design applications. Because of the crack-opening-area models used in extensive amount of available current leak-rate estimations have been information, the report was divided incorporated into a single computer into two parts. Part I discusses cyclic code for leak-rate estimation. The internal pressure loading and Part II code is called SQUIRT, wtlich stands for discusses moment loadings for the Seepage Quantification of Upsets In branch and run. The cyclic pressure Reactor Tubes. The SQUIRT program has loading fatigue parameters are mostly been validated by comparing its l based on leakage, whereas. if the thermalhydraulic predictions with the parameters were based on crack limited experimental data that have initiation, different and possibly been published on two-phase flow higher valves would be developed. The through slits and cracks, and by fatigue evaluation procedure, which comparing its crack-opening-area attempts to relate fatigue strength of predictions with data from the Degraded piping components to strain controlled, Piping Program. In addition, leak-rate polished bar and fatigue data appears experiments were conducted to obtain to be inaccurate on the conservative validation data for a circumferential side for high amplitude cycles and on fatigue crack in a carbon steel pipe the unconservative side for low girth weld.

l NUREG-1426 58 I I

l

Compilation of Reports: 1994-1998 under seismic loading (repeating

Title:

Validation of analysis methods dynamic loads) are being pursued for assessing flawed piping subjected separately within the NRC's to dynamic loading International Piping Integrity Research Author (s)/ Editor (s): Olson, R.J. . Group (IPIRG) program. This report Wolterman, R.L. . Wilkowski G.M. describes developmental and validation (Battelle, Columbus. OH (United efforts to predict crack stability States)); Kot. C.A. (Argonne National under water hammer loading, as well as Lab. , IL (United States)) comparisons using currently used Soonsorino Oroanization: NRC; Nuclear analysis procedures. Current fracture Regulatory Commission, Washington, DC analysis methods use the elastic stress (United States) analysis loads decoupled from the Publication Date: Aug 1994 fracture mechanics analysis, while Reoort Number (s): NUREG/CR-6234; state-of-the-art methods employ ANL--94/22; BMI--2178 nonlinear cracked-pipe time-history Order Number: TI94017485 finite element analyses. The results Abstract: Argonne National Laboratory showed that the current decoupled and Battelle have jointly conducted a methods were conservative in their research program for the USNRC to predictions, whereas the cracked pipe evaluate the ability of current finite element analyses were more engineering analysis methods and one accurate, yet slightly conservative.

state-of-the-art analysis method to The nonlinear time-history cracked-pipe predict the behavior of finite element analyses conducted in circumferentially surface-cracked pipe this progrn were also attractive in system water-hamer experiment. The that they were done on a small Apollo experimental data used in the DN5500 workstation, whereas other evaluation were from the HDR Test Group cracked-pipe dynamic analyses conducted E31 series conducted by the in Europe on the same experiments Kernforschungszentrum Karlsruhe (KfK) required the use of a CRAY2 in Germany. The incentive for this supercomputer, and were less accurate.

evaluation was that simplified engineering methods, as well as newer state-of the-art fracture analysis Title- Stability of cracked pipe under methods, have been typically validated inertial stresses only with static experimental data. Author (s)/ Editor (s): Scott, P. ,

Hence, these dynamic experiments were Wilson, M. , Olson, R. Marschall, C.

of high interest. High-rate dynamic , Schmidt, R. : Wilkowski, G.

loading can be classified as either (Battelle, Columbus, OH (United repeating, e.g. , seismic, or States))

nonrepeating, e.g. , water hammer. Soonsorino Oroanization: NRC; Nuclear Development of experimental data and Regulatory Commission, Washington, DC validation of cracked pipe analyses (United States) 59 NUREG-1426

Compilation of Reports - 1994-1998 Publication Date: Aug 1994 significantly decreased the TP304 Reoort Number (s): NUREG/CR-6233-Vol.1: stainless steel surface-cracked pipe BMI--2177-Vol.1 apparent toughness. The inertial Order Number: TI94018624 experiments tended to achieve complete Abstract: This report presents the failure within a few cycles after results of the pipe fracture reaching maximum load in these experiments, analyses, and material relatively small diameter pipe characterization efforts performed experiments. Hence, a load-controlled within Subtask 1.1 of the IPIRG fracture mechanics analysis may be more Program. The objective of Subtask 1.1 appropriate than a was to experimentally verify the displacement-controlled analysis for analysis methodologies for these tests. 1 circumferentially cracked pipe subjected primarily to inertial stresses. Eight cracked-pipe Title _:- Effect of dynamic strain aging experiments were conducted on 6-inch on the strength and toughness of nominal diameter TP304 and A106B pipe, nuclear ferritic piping at LWR The experimental procedure was temperatures developed using nonlinear time-history Author (s)/ Editor (s): Marschall, C.W. ,

finite element analyses which included Mohan, R. ; Krishnaswamy. P. ,

Wilkowski. G.M. (Battelle. Columbus, OH l the nonlinear behavior due to the I crack, The model did an excellent job (linited States))

of predicting the displacements. Soonsorina Oraanization: NRC: Nuclear l

forces, and times to maximum moment. Regulatory Commission, Washington, DC The comparison of the experimental (United States)

loads to the predicted loads by the Publication Date

Oct 1994 Net-Section-Collapse (NSC), Reoort Number (s): NUREG/CR-6226; 1

Dimensionless Plastic-Zone Parameter, BMI--2176 J-estimation schemes. R6. and ASME Order Number: TI95002451 Section XI in-service flaw assessment Abstract: This topical report is on criteria tended to underpredict the the phenomenon of dynamic strain aging measured bending moments except for the (DSA) in ferritic nuclear piping steels NSC analysis of the A106B pipe. The and its effect on fracture at LWR effects of flaw geometry and loading temperatures. The report was a history on toughness were evaluated by deliverable from the US NRC's program calculating the toughness from the pipe entitled Short Cracks in Piping and l tests and comparing these results to Piping Welds The objective of this

! C(l) values. These effects were found work was to predict the occurrence of i l to be variable. The surface-crack and evaluate the effects of ductile geometry tended to increase the crack instabilities, which occur toughness (relative to CM results), frequently in ferritic steel pipe whereas a negative load-ratio fracture tests at 288 C (550 F), and

(

NUREG-1426 60 l

l 1 l

l

Compilation of Reports: 1994-1998 are believed to be due to dynamic stain analyses, which assume a hypothetical aging. Numerous laboratory tests and flaw size, there is also interest in one numerical simulation of a C(T) test the integrity of actual leaking cracks with crack instabilities were corresponding to current leakage undertaken. detection requirements in NRC Regulatory Guide 1.45, or for assessing temporary repair of Class 2 and 3 pipes

Title:

Refinement and evaluation of that have leaks as are being evaluated crack-opening-area analyses for in ASME Section 11. This study was circumferential through-wall cracks in requested by the NRC to review, pipes evaluate, and refine current analytical Author (s)/ Editor (s): Rahman, S. , models for crack-opening-area analyses Brust, F. . Ghadiali N. : of pipes with circumferential Krishnaswamy, P. : W11kowski, G, through-wall cracks. Twenty-five pipe (Battelle, Columbus, OH (United experiments were analyzed to determine States)); Choi, Y.H. (Battelle, the accuracy of the predictive models.

. Columbus, OH (United States) Korea Several practical aspects of Inst. .of Nuclear Safety, Taejeon crack opening such as: crack-face (Korea, Republic of)); Moberg, F. , pressure, off center cracks, restraint Brickstad, B. (Battelle, Columbus, OH of pressure-induced bending, cracks in (United States) Swedish Plant thickness transition regions, weld Inspection Ltd., Stockholm (Sweden)) residual stresses, crack-morphology Soonsorino Oraanization: NRC: Nuclear models, and thermal-hydraulic analysis, Regulatory Commission Washington, DC were also investigated. 140 refs., 105 (United States) figs., l.1 tabs.

Publication Date; Apr 1995 Reoort Number (s): NUREG/CR-6300:

BMI--2184

Title:

Short cracks in piping and Order Number: TI95010503 piping welds. Seventh program report.

Abstract: Leak before-break (LBB) March 1993- December 1994. Volume 4, analyses for circumferentially cracked Numoer 1 pipes are currently being conducted in Author (s)/ Editor (s): Wilkowski, G.M. :

the nuclear industry to justify Ghadiali. N. . Rudland, D. :

elimination of pipe whip restraints and Krishnasway, P. : Rahman, S. . Scott, jet impingement shields which are P. (Battelle, Columbus, OH (United present because of the expected dynamic States))

effects from pipe rupture. The Soonsorino Oraanization: NRC: Nuclear application of the LBB methodology Regulatory Comission Washington, DC frequently requires calculation of leak (United States) rates. These leak rates depend on the Publication Date: Apr 1995 crack-opening area of a tnrough-wall Reoort Number (s):

crack in the pipe. In addition to LBB NUREG/CR-4599-Vol.4-No.1:

61 NUREG-1426

Compilation of Reports - 1994-1998 BMI--2173-Vol.4 No.1

Title:

Fracture evaluations of fusion Order Number: TI95010955 line cracks ... nuclear pipe bimetallic Abstract: This is the seventh progress welds report of the U.S. Nuclear Regulatory Author (s)/ Editor (s): Scott P. .

Comission's research program entitled Francini. R. . Rahman. S. : Rosenfield, s [open quotes]Short Cracks in Piping and A. . Wilkowski, G. (Battelle. Columbus.

Piping Welds [close quotes). The OH (United States))

program objective is to verify and Soonsorina Oraanization: NRC: Nuclear improve fracture analyses for Regulatory Cemission. Washington. DC circumferentially cracked (United States) large-diameter nuclear piping with Publication Date: Apr 1995 crack sizes typically used in Reoort Number (s): NUREG/CR-6297:

leak-before-break (LBB) analyses and BMI--2182 in- service flaw evaluations. All Order Number: TI95010502 work in the eight technical tasks have Abstract: In both BWRs and PWRs there been completed. Ten topical reports are many locations where carbon steel are scheduled to be published, pipe or components are joined to Progress only during the reporting stainless steel pipe or components with period. March 1993 - December 1994, not a bimetallic weld. The objective of covered in.the topical reports is the research described in this report presented in this report. Details was to assess the accuracy of current about the following efforts are covered fracture analyses for the case of a in this report: (1) Improvements to the crack along a carbon steel to two computer programs NRCPIPE and austenitic weld fusion line. To NRCPIPES to assess the failure behavior achieve the program objective, material of circumferential through-wall and property data and data from a surface-cracked pipe respectively; (2) large-diameter pipe fracture experiment Pipe material property database PIFRAC: were developed to assess current (3) Circumferentially cracked pipe analytical methods. The bimetallic database CIRCUMCK.WKI; (4) An welds evaluated in this program were assessment of the proposed ASME Section bimetallic welds obtained from a III design stress rule changes on pipe cancelled Combustion Engineering plant.

flaw tolerance; and (5) A pipe frac);ure The welds joined sections of the carbon experiment on a section cf pipe v.aved steel cold-leg piping system to from service degraded by stainless steel safe ends that were to microbiologically induced corrosion be welded to stainless steel pump (MIC) which contained a girth weld housings. The major conclusion drawn crack. Progress in the other tasks is as a result of these efforts was that not repeated here as it has been the fracture behavior of the bimetallic covered in great detail in the topical weld evaluated in this program could De reports. evaluated with reasonable accuracy using the strength and toughness l

NUREG-1426 62 l

l

Compilation of Reports: 1994-1998 properties of the carbon steel pipe this work. The first was that material in conjunction with virtually all ferritic nuclear pipes conventional elastic plastic fracture will have toughness anisotropy. The mechanics or limit-load analyses. This second was that the ratio of the may not be generally true for all normalized crack driving force (as a bimetallic welds, as discussed in this function of angle) to the normalized report. toughness (also as a function of the angle of crack growth) showed that there was an equal likelihood of cracks

Title:

Effects of toughness anisotropy growing at any angle between 25 and 65 and combined tension, torsion, and degrees. This agreed with the scatter bending loads on fracture behavior of of crack growth angles observed in pipe ferritic nuclear pipe tests. Third, for combined loads with Autho'r(s)/Edi tor (s): Mohan, R. . torsional stresses, an effective moment Marschall, C. . Krishnaswamy. P : allows pure bending analyses to be used Brust. F. . Ghadiali, N. . Wilkowski, up to crack initiation. Crack opening G. (Battelle. Columbus OH (United area under combined loads could also be States)) determined in this mariner.

Soonsorina Orcanization: NRC: Nuclear Regulatory Comission. Washington, DC (United States)

Title:

Probabilistic pipe fracture Publication Date: Apr 1995 evaluations for leak-rate-detection Reoort Number (s): NUREG/CR-6299 applications Order Number: TI95012202 Author (s)/ Editor (s): Aahman. S. .

Abstract: This topical report Ghadiali . N. . Paul . D. . Wilkowski G.

summarizes the work on angled crack (Battelle, Columbus. OH (United growth and combined loading effects States))

performed within the Nuclear Regulatory soonsorino Oraanization: NRC: Nuclear Commission's research program entitled Regulatory Commission. Washington, DC (open quotes]Short Cracks in Piping and (United States)

Piping Welds [close quotes]. The major Publication Date: Apr 1995 impetus for this work stemmed from the Reoort Number (s): NUREG/CR-6004:

observation that initial BMI--2174 circumferential cracks in carbon steel Order Number: TI95011013 pipes exhibited angular crack growth. Abstract: Regulatory Guide 1.45. (open This failure mode was little quotes] Reactor Coolant Pressure understood, and the effect of angled Boundary Leakage Detection crack growth from an initially Systems.[close quotes) was published by circumferential crack raised questions the U.S. Nuclear Regulatory Comission of how pipes under combined loading (NRC) in May 1973, and provides with torsional stresses would behave. guidance on leak detection methods and There were three major conclusions from system requirements for Light Water 63 NUREG-1426

Compilation of Reports - 1994-1998 Reactors. Additionally, leak detection Soonsorino Oroanization: NRC: Nuclear limits are specified in plant Technical Regulatory Commission, Washington, DC Specifications and are different for (United States)

Boiling Water Reactors (BWRs) and Publication Date: Apr 1995 Reoort Number (s): NUREG/CR-6235; Pressurized Water Reactors (PWRs).

These leak detection limits are also BMI--2179 used in leak-before-break evaluations Order Number: TI95010506 performed in accordance with Draft Abstract: This topical report Standard Review Plan, Section 3.6.3, sumarizes the work performed for the

[open quotes] Leak Before Break Nuclear Regulatory Comission's (NRC)

Evaluation Procedures [close quotes] research program entitled Short where a margin of 10 on the leak Cracks in Piping and Piping Welds '

detection limit is used in determining that specifically focuses on pipes with the crack size considered in subsequent short through-wall cracks. Previous fracture analyses, This study was NRC efforts, conducted under the requested by the NRC to: (1) evaluate Degraded Piping Program, focused on the conditional failure probability for understanding the fracture behavior of BWR and PWR piping for pipes that were larger cracks in piping and fundamental leaking at the allowable leak detection fracture mechanics developments limit, and (2) evaluate the margin of necessary for this technology. This 10 to determine if it was unnecessarily report gives details on: (1) material large. A probabilistic approach was property determinations, (2) pipe

. undertaken to conduct fracture fracture experiments, and (3) evaluations of circumferentially development, modification, and cracked pipes for leak-rate-detection validation of fracture analysis applications. Sixteen nuclear piping methods. The material property data-systems in BWR and PWR plants were required to analyze the experimental analyzed to evaluate conditional results are included. These data were failure probability and effects of also implemented into the NRC's PIFRAC crack morphology variability on the database. Three pipe experiments with current margins used in leak rate short through-wall cracks were detection for leak-before break, conducted on large diameter pipe.

Also, experiments were conducted on a l

l large-diameter uncracked pipe and a Title. Assessment of short pipe with a moderate-size through-wall through-wall circumferential cracks in crack. The analysis results reported pipes. Experiments and analysis: March here focus on simple predictive methods 1990-' December 1994 based on the J-Tearing theory as well Author (s)/ Editor (s): Brust, F.W. : as limit-load and ASME Section 11 Scott, P. : Rahman. S. (Battelle, analyses. Some of these methods were Columbus, OH (United States)) (and improved for short-crack-length others) predictions. The accuracy of the NUREG-1426 64

Compilation of Reports: 1994 1998 various methods was determined by limit-load approaches. (ii) design comparisons with experimental results criteria, and (iii) elastic-plastic from this and other programs. 69 fracture methods. These methods were refs. ,124 figs, 49 tabs. evaluated by comparing the analytical predictions with experimental data.

The results, using 44 pipe experiments

Title:

Fracture behavior of short from this and other programs, showed circumferential1y surface-cracked pipe that the SC.TNP1 and DDZP analyses were Author (s)/ Editor (s): Krishnaswamy, P. the most accurate in predicting maximum

Scott, P. . Mohan, R. (Battelle, load. New Z-factors were developed Columbus OH (United States)) (and using these methods. These are being others) considered for updating the ASME Soonsorino Orcanization: NRC: Nuclear Section XI criteria.

Regulatory Comission. Washington, DC (United States)

Publication Date: Nov 1995

Title:

The effect of cyclic and Report Number (s): NUREG/CR 6298: dynamic loads on carbon steel pipe BMI--2183 Author (s)/ Editor (s): Rudland. D.L. ,

( der Number: TI96004052 Scott, P.M. : Wilkowski , G.M.

tract: This topical report (Battelle, Columbus. OH (United marizes the work performed for the States))

clear Regulatory Comn11ssion's (NRC) Soonsorino Oraanization

NRC: Nuclear research program entitled Short Regulatory Commission, Washington, DC Cracks in Piping and Piping Welds (United States) that specifically focuses on pipes with Publication Date: Feb 1996 short, circumferential surface cracks. Reoort Number (s): NUREG/CR-6438:

The following details are provided in BMI--2188 this report: (i) material property Order Number: TI96007099 deteminations. (ii) pipo fracture Abstract:_ This report presents the experiments, (iii) development, results of four 152-m (6-inch) modification and validation of fracture diameter, unpressurized, analysis methods, and (iv) impact of circumferential through-wall-cracked, this work on the ASME Section XI Flaw dynamic pipe experiments fabricated Evaluation Procedures. The material from STS410 carbon steel pipe properties developed and used in the manufactured in Japan. For three of analysis of the experiments are these experiments, the through-wall included in this report and have been crack was in the base metal. The implemented into the NRC's PIFRAC displacement histories applied to these database. Six full-scale pipe experiments were a quasi-static experiments were conducted during this monotonic dynamic monotonic, and program. The analyses methods reported dynamic, cyclic (R - [minus]1) history.

here fall into three categories (i) The through-wall crack for the third 65 NUREG 1426 i

I Compilation of Reports - 1994-1998 experiment was in a tungsten-inert-gas one day international round-robin i' weld, fabricated in Japan, joining two workshops which were organized by

! -lengths of STS410 pipe. The Battelle in conjunction with the Second displacement history for this International Piping Integrity Research experiment was the same history applied Group (IPIRG- 2) Program. The

! to the dynamic, cyclic base metal objective of these workshops was to experiment. Tha test temperature for develop a consensus in handling i each experiment was 300 C (572 F). The difficult analytical problems in objective of these experiments was to leak-before-break and pipe flaw l compare a Japanese carbon steel pipe evaluations. The workshops, which were j

! material with US pipe material, to held August 5, 1993. March 4, 1994, and ascertain whether this Japanese steel October 21, 1994 at Columbus, Ohio, was as sensitive to dynamic and cyclic involved various technical effects as US carbon steel pipe. In presentations on the related research support of these pipe experiments, efforts by the IPIRG-2 member quasi-static and dynamic, tensile and organizations and solutions to several fracture toughness tests were round-robin problems. Following review conducted. An analysis effort was by the IPIRG-2 members, four sets of performed that involved comparing round-robin problems were developed.

experimental crack initiation and They involved: (1) evaluations of I maximum moments with predictions based fracture properties and pipe loads, (2) l on available fracture prediction crack-opening and leak-rate models, and calculating J-R curves for evaluations (3) dynamic analysis of the pipe experiments using the cracked pipes, and (4) evaluations of (eta]-factormethod, elbows. A total of 18 organizations from the United States, Japan, Korea, and Europe solved these round-robin

Title:

Sumary of results from the problems. The analysis techniques IPIRG-2 round-robin analyses employed by the participants included Author (s)/ Editor (s): Rahman, S. : both finite element and engineering Olson, R. . Rosenfield, A. : Wilkowski, methods. Based on the results from G. (Battelle, Columbus, OH (United these analyses, several important States)) observations were made concerning the l

Soonsorino Oroanization: NRC: Nuclear predictive capability of the current Regulatory Comission, Washington, DC fracture-mechanics and (United States) thermal-hydraulics models for their Publication Date: Feb 1996 applications in nuclear piping and Reoort Number (s): NUREG/CR-6337: piping welds.

BMI--2186 Order Number: TI96006050 Abstract: This report presents a

Title:

Design of the IPIRG 2 simulated sumary of the results from three seismic forcing function NUREG-1426 66

- - _ _ ~ _ _ _ - - _ . _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ .__-

l l

l Compilation of Reports: 1994-1998 l Author (s)/ Editor (s): Olson, R. : elbows. The analyses involved l

Scott, P. : Wilkowski, G. (Battelle, development of a GE/EPRI type Columbus, OH (United States)) J-estimation scheme which requires an Soonsorino Oroanization: NRC: Nuclear elastic and fully plastic contribution Regulatory Comission. Washington, DC to crack-driving force in terms of the (United States) J- integral parameter. The elastic Publication Date: Feb 1996 analyses require the development of Reoort Number (s): NUREG/CR-6439; F-function values to relate the J[sub BMI--2189 e] term to applied loads. Similarly, Order Number: TI96006110 the fully plastic analyses require the Abstract: A series of pipe system development of h-function; to relate experiments was conducted in IPIRG-2 the J[sub p) term to the applied loads.

that used a realistic seismic forcing The F- and h-functions were cetermined function. Because the seismic forcing from a matrix of finite element function was more complex than the analyses. To minimize the cost of the single-frequency increasing-amplitude analyses, three- dimensional ABAQUS sinusoidal forcing function used in the finite element analyses were compared IPIRG-1 pipe system experiments, to a simpler finite element technique considerable effort went into designing called the line-spring method. The the function. This report documents line-spring method provides a the design process for the seismic significant computational savings over forcing function used in the IPIRG-2 the full three-dimensional analysis.

pipe system experiments. The comparison showed excellent agreement between the line-spring and three- dimensional analysis. This

Title:

Development of a J-estimation experience was consistent with scheme for internal circumferential and comparisons with circumferential axial surface cracks in elbows surface-crack analyses in straight Author (s)/ Editor (s): Mohan, R. , pipes during the NRC's Short Cracks in Brust, F.W. : Ghadiali, N. . Wilkowski, Piping and Piping Welds program.

G.

Soonsorina Oraanization: NRC: Nuclear Regulatory Comission, Washington, DC

Title:

Deterministic and probabilistic (United States) evaluations for uncertainty in pipe Publication Date: Jun 1996 fracture parameters in Reoort Number (s): NUREG/CR-6445: leak-before-break and in-service flaw BMI--2193 evaluations Order Number: TI96012174 Author (s)/ Editor (s): Ghadiali, N. ,

Abstract _;. This report sumarizes Wilkowski, G. (Battelle. Columbus, OH efforts to develop elastic and (United States)); Rahman, S. (Univ. of elastic-plastic fracture mechanics Iowa. Iowa City. IA (United States)):

l analyses for internal surface cracks in Choi, Y.H. (Korea Inst. of Nuclear 1

1 67 NUREG-1426

_ _ _ . . __s___. _ .. _ - _ _ _ _ _ _ .. _ _ _ _ _ . _ _ _ . . _ _

l j

' Compilation of Reports - 1994-1998 Safety (KINS) Daeduk-danji Taejon Reoort Number (s)

NUREG/CR-6444:

(Korea, Republic of)) BMI-2192 Soonsorino Orcanization: NRC: Nuclear Order Number: TI97002497 Regulatory Commission, Washington, DC Abstract: This report presents the (United States) results from Task 2 of the Second i Publication Date: Jun 1996 International Piping Integrity Research Recort Number (s): NUREG/CR-6443: Group (IPIRG-2) program. The focus of )

BMI--2191 the Task 2 work was directed towards l

Order Number: TI96012372 furthering the understanding of the l Abstract: This report presents new fracture behavior of long-radius l l

results from deterministic and elbows . This was accomplished through l probabilistic analyses to evaluate the a combined analytical and experimental l significance of a number of technical program. J estimation schemes were developed for both axial and I l aspects that may affect LBB or in-service flaw evah.ations. The circumferential surface cracks in following summarizes the objectives and elbows. Large-scale, quasi-static and ,

results from both the deterministic and dynamic, pipe-system, elbow fracture  !

probabilistic studies. The reasons for experiments under combined pressure and l including each technical aspect being bending loads were performed on elbows i

. evaluated are given first, Then a containing an internal surface crack at

-table is given that summarizes the the extrados, In conjunction with the relative significance of each technical elbow experiments, material property aspect. In most cases there are both data were developed for the A106-90 deterministic and probabilistic carbon steel and WP304L stainless steel results. The deterministic analyses elbow materials investigated. A were conducted independently of the comparison of the experimental data l probabilistic analysis, which offered with the maximum stress predictions

! the opportunity to validate conclusions using existing straight pipe fracture from each of these studies. prediction analysis methods, and elbow fracture prediction methods developed in this program was performed. This

Title:

Fracture bahavior of analysis was directed at addressing the  !

circumferential1y surface-cracked concerns regarding the validity of

elbows. Technical report, October using analysis predictions developed 1993--March 1996 for straight pipe to predict the j

Author (s)/ Editor (s u Kilinski, T. : fracture stresses of cracked elbows.

Mohan, R. : Rudlanc , D. : Fleming, M. Finally, a simplified fitting flaw (and others) acceptance criteria incorporating ASME l Soonsorino Orcanization: NRC: Nuclear B2 stress indices and straight pipe.

Regulatory Commission.. Washington, DC circumferential- crack analysis was (United States) developed.

Pyblication Date: Dec 1996 l

{ NUREG-1426 68 l

l

~. . . - - _ _ -, _ . _ _ - _ - . _ - . _ _ _ _ _ _ _ _ - .._ -

l l

l Compilation of Reports: 1994-1998

Title:

The effects of cyclic and that the maximum moments decrease dynamic loading on the fracture slightly from cyclic toughness resistance of nuclear piping steels. degradation as the pipe diameter Technical report. October 1992--April increases. (3) Dynamic stress-strain 1996 and compact tension tests were Author (s)/ Editor (s); Rudland. D.L. . conducted to expand on the existing Brust. F. : Wilkowski, G.M. dynamic database. Results from dynamic Soonsorino Oroanization: NRC: Nuclear moment predictions suggest that the Regulatory Comisswn. Washington, DC dynamic compact tension J-R and the (United States) quasi-static stress-strain curves are Publication Date: Dec 1996 the appropriate material properties to Reoort Number (s): NUREG/CK-6440: use in making dynamic pipe moment BMI- 2190 predictions.

Order Number: TI97002498

, Abstract: This report presents the results of the material property

Title:

IPIRG-2 task 1 - pipe system evaluation efforts performed within experiments with circumferential cracks Task 3 of the IPIRG-2 Program. Several in straight-pipe locations. Final l related investigations were conducted. report. September 1991--November 1995 (1) Quasi-static, cyclic-load compact Author (s)/ Editor (s): Scott P. .

tension specimen experiments were Olson, R. : Marschall, C. : Rudland. D.

conducted using parameters similar to (and others) those used in IPIRG-1 experiments on Soonsorino Oroanization: NRC: Nuclear

! 6-inch nominal diameter Regulatory Comission, Washington DC through-wall-cracked pipes. These (United States) experiments were conducted on a TP304 Publicaticn Date: Feb 1997 i

base metal, an A106 Grade B base metal. ReDort Number (s): NUREG/CR-6389:

j and their respective submerged-arc BMI--2187 l welds. The results showed that when Order Number:_ TI97003757 l

using a constant cyclic displacement Abstract: This report presents the increment, the compact tension results from Task 1 of the Second experiments could predict the International Piping Integrity Research

through-wall cracked pipe crack Group (IPIRG-2) program. The IPIRG 2 initiation toughness, but a different program is an international group control procedure is needed to program managed by the US Nuclear reproduce the pipe cyclic crack growth Regulatory Comission (US NRC) and in the compact tension tests. (2) funded by a consortium of organizations Analyses conducted showed that for from 15 nations including

Bulgaria, 6 inch diameter pipe, the quasi-static. Canada, Czech Republic. France, monotonic J-R curve can be used in Hungary. Italy, Japan, Republic of making cyclic pipe moment predictions: Korea Lithuania Republic of China, however, sensitivity analyses suggest Slovak Republic. Sweden. Switzerland.

+

69 NUREG-1426 l

l

Compilation of Reports - 1994-1998 the United Kingdom, and the United Order Number: TI97004340 States. The objective of the program Abstract: In the IPIRG-1 program, the was to build on the results of the J-R curve calculated for a 16-inch IPIRG 1 and other related programs by nominal diameter, Schedule 100 TP304 extending the state- of-the-art in stainless steel (DP2-A8) pipe fracture technology through the surface-cracked pipe experiment development of data needed to verify (Experiment 1.3-3) was considerably engineering methods for assessing the lower than the quasi " atic, monotonic integrity of nuclear power plant piping J-R curve calculated T/om a C(T) systems that contain defects. The specimen (A8-12a). The results from IPIRG-2 program included five main several related investigations tasks: Task 1 - Pipe System Experiments conducted to determine the cause of the with Flaws in Straight Pipe and Welds observed toughness difference are: (1)

Task 2 - Fracture of Flawed Fittings chemical analyses on sections of Pipe Task 3 - Cyclic and Dynamic Load DP2-A8 from several surface cracked Effects on Fracture Toughness Task 4 - pipe and material property specimen Resolution of Issues From IPIRG-1 and fracture surfaces indicate that there Related Programs Task 5 - Information are two distinct heats of material Exchange Seminars and Workshops, and within Pipe DP2-A8 that differ in Program Management. The scope of this chemical composition; (2) SEN(T) report is to present the results from specimen experimental results indicate the experiments and analyses associated that the toughness of a surface-cracked with Task 1 (Pipe System Experiments specimen is highly dependent on the with Flaws in Straight Pipe and Welds). depth of the initial crack, in The rationale and objectives of this addition, the J-R curves from the task are discussed after a brief review SEN(T) specimens closely match the J-R of experimental data whicn existed curve from the surface-cracked pipe after the IPIRG- 1 program. experiment: (3) C(T) experimental results suggest that there is a large difference in the quasi- static,

Title:

Fracture toughness evaluations monotonic toughness between the two of TP304 stainless steel pipes heats of DP2-A8, as well as a toughness Author (s)/ Editor (s): Rudland, D.L. : degradation in the lower toughness heat Brust. F.W. ; Wilkowski, G.M. of material (DP2-A811) when loaded with (Battelle. Columbus, OH (United a dynamic, cyclic (R - [minus]0.3)

States)) loading history.

Smnsorina Oroanization: NRC: Nuclear Regulatory Commission Washington, DC (United States)

Title:

The Second International Piping Publication Date: Feb 1997 Integrity Research Group (IPIRG 2)

Reoort Number (s): NUREG/CR-6446; program. Final report, October BMI--2194 1991--April 1996 NUREG-1426 70

l Compilation of Reports: 1994 1998 Author (s)/ Editor (s): Hopper. A. : analyses, and analyst's group meetings Wilowski, G. . Scott. P. : Olson R. to provide a forum for nuclear piping l (and others) experts from around the world to Soonsorino Orcanization: NRC: Nuclear exchange information on the subject of Regulatory Comission. Washington, DC pipe f,acture technology. 17 refs..

(United States) 104 figs. 41 tabs.

Publication Date: Mar 1997 i Recort Number (s): NUREG/CR-6452:

BMI- 2195

Title:

Proceedings of the Seminar on Order Number: TI97004743 Leak Before Break in Reactor Piping and Abstract: The IPIRG-2 program was an Vessels

international group program managed by Author (s)/ Editor (s)

C. Faidy, l the US NRC and funded by organizations (Electricit'e de France). Ph. Gilles from 15 nations. The emphasis of the (Framatome)

IPIRG-2 program was the development of Soonsorina Oroanization: Electricit'e data to verify fracture analyses for de France. Framatome. Commissariat a cracked pipes and fittings subjected to l'Energie Atomique. European Comunity.

l_ dynamic / cyclic load histories typical DGXI-WGCS, Nuclear Electric, of seismic events. The scope included: International Atomic Energy Agency.

l (1) the study of more complex OECD-Nuclear Energy Agency. USNuclear I

dynamic / cyclic load histories, i.e.. Regulatory Commission. French Nuclear multi-frequency, variable amplitude. Energy Society simulated seismic excitations, than Publication Date: April 1997 those considered in the IPIRG-1 Reoort Number (s): NUREG/CP-0155 ,

program. (2) crack sizes more typical Abstract: The sixth in a series of of those considered in international Leak-Before-Break (LBB)

Leak-Before-Break (LBB) and in-service Seminars was held at Hotel Sofitel in flaw evaluations. (3) Lyon, France on October 9 through 11.

through-wall-cracked pipe experiments 1995. The semincr updated international which can be used to validate LBB-type policies and supporting research on fracture analyses. (4) cracks in and L38. The more than 210 attendees that l

around pipe fittings, such as elbows. Joined the meeting included and (5) laboratory specimen and representatives from regulatory separate effect pipe experiments to agencies, electric utility l provide better insight into the effects representatives, fabricators of nuclear i

of dynamic and cyclic load histories, power plants, research organizations.

Also undertaken were an uncertainty and academic institutions.

! analysis to identify the issues most The objective of the seminar was to

! important for LBB or in- service present the current state of the art in

! flaw evaluations, updating computer LBB methodology development, codes and databases, the development validation, and application in an i and conduct of a series of round robin international forum. With particular i

1 71 NUREG-1426 i

I l

-,-. . - - - - - --- - - , - - , , - . - - ,, ,.,a n . , ,- , ~ - - , . , -

Compilation of Reports - 1994-1998 emphasis on industrial applications and In support of these activities, better regulatory policies, the seminar estimates of the limits to the LBB provided an opportunity to compare approach should follow, as well as an approaches, experiences, and improvement in codifying methodologies, codifications developed by different countries.

The seminar was organized into four

Title:

International Piping Integrity topic areas: Research Group (IPIRG) Program. Final Status of LBB Applications report Technical Issues in LBB Methodology Author (s)/ Editor (s): Wilkowski, G. .

Complementary Requirements (Leak Schmidt, R. . Scott, P. (and others)

Detection and Inspection) Soonsorino Orcanization: NRC; Nuclear Regulatory Comission, Washington, DC LBB Assessment and Margins. (United States)

In addition to the formal sessions Publication Date: Jun 1997 where papers were presented by Reoort Number (s): NUREG/CR-6233-Vol.4 participants from France, Germany. Order Number: TI97006968 Japan. Korea, Belgium, the United Abstract: This is the final report of Kingdom, the Czech Republic. Finland, the International Piping Integrity Russia, Sweden Canada, the Research Group (IPIRG) Program. The (

Netherlands, and the United States, IPIRG Program was an international informal LBB poster sessions were group program managed by the U.S.

available outside the presentation Nuclear Regulatory Comission and hall. A keynote address (see Appendix funded by a consortium of organizations B) by Mr. J. Branchu. Head of the from nine nations: Canada, France, Primary Nuclear Components Division of Italy, Japan. Sweden. Switzerland, Framatome, was delivered at the LBB 95 Taiwan, the United Kingdom, and the Banquet and summarized the goals and United States. The program objective objectives of the seminar. As a result was to develop data needed to verify of this seminar, an improved engineering methods for assessing the understanding of LBB gained through integrity of circumferentially-cracked sharing of different viewpoints from nuclear power plant piping. The different countries, permits primary focus was an experimental task consideration of: that investigated the behavior of Simplified pipe support design and circumferentially flawed piping systems possible elimination of loss of- subjected to high rate loadings typical coolant- accident (LOCA) mechanical of seismic events. To accomplish these consequences for specific cases objectives a pipe system fabricated as Defense in-Depth type of applications an expansion loop with over 30 meters without support modifications of 16-inch diameter pipe and five long Sup;urt of safetj cases for plants radius elbows was constructed. Five desig1ed without the LOCA hypothesis. dynamic, cyclic, flawed piping NUREG-1426 72

Compilation of Reports: 1994-1998 experiments were conducted using this (550 F). The results indicated dynamic facility. This report: (1) provides loading at seismic strain rates background information on marginally increased the load-carrying leak-before-break and flaw evaluation capacity of austenitic steel. The procedures for piping. (2) sumarizes ferritic steel tested was sensitive to technical results of the program. (3) dynamic strain-aging. and consequently, gives a relatively detailed assessment its load carrying capacity decreased at of the results from the pipe fracture dynamic strain rates. Two parameters experiments and complementary analyses, were found to affect the apparent and (4) sumarizes advances in the ductile crack growth resistance during state-of-the-art of pipe fracture cyclic loading, load ratio (R) and technology resulting from the IPIRG incremental plastic displacement that program, occurs in a cycle. Cyclic (R - 0) loading had minimal effect on ductile tearing for both materials. However.

Title:

Stability of cracked pipe under fully reversed loading decreased the l seismic / dynamic displacement-controlled load-carrying capacity and toughness stresses. Subtask 1.2 final report for both materials. The incremental Author (s)/ Editor (s): Kramer. G. , plastic displacement can be as Veith, P. ; Marschall. C. (and others) important as the load ratio; however. l Soonsorino Oraanization: NRC; Nuclear it is harder to quantify from design Regulatory Comission. Washington DC stress reports. Large plastic (United States) displacements will minimize the effect Publication Date: Jun 1997 of negative load ratios.

Reoort Number (s): NUREG/CR-6233-Vol.2:

BMI--2177 Order Number: TI97006961

Title:

Crack stability in a Abstract: Results of representative piping system under displacement-controlled pipe fracture combined inertial and seismic / dynamic experiments, analyses, and material displacement-controlled stresses.

characterization efforts performed Subtask 1.3 final report within the International Piping Author (s)/ Editor (s): Scott P. :

Integrity Research Group IPIRG. Olson, R. Wilkowski . 0.G. .

Program Subtask 1.2 are discussed. Marschall. C. , Schmidt, R.

Effects of dynamic versus quasi-static Soonsorina Oroanization: NRC: Nuclear and monotonic versus cyclic loading Regulatory Commission, Washington DC were evaluated for ductile tearing of (United States) two materials, A106 Grade B ferritic Publication Date: Jun 1997 steel and TP304 austenitic steel- 822prt Number (s): NUREG/CR-6233-Vol.3:

Twelve through wall-cracked pipe BMI--2177 experiments were conducted on 6-inch Order Number: TI97006967 diameter Schedule 120 pipe at 288 C Abstract: This report presents the 73 NUREG-1426

Compilaticq of Reports - 1994-1998

Title:

State-of-the-Art Report on results from Subtask 1.3 of the International Piping Integrity Research Piping Fracture Mechanics Group (IPIRG) program. The objective Author (s)/ Editor (s): G.M. Wilkowski, of Subtask 1.3 is to deve hp data to R.J. 01s01. P.M. Scott (Battelle) assess analysis methodologies for Soonsorina Oraanization: NRC:

characterizing the fracture behavior of Washington DC (United States) circumferentially cracked pipe in a Publication Date: January 1993 representative piping system under Reoort Number (s)- NURcG/CR-6540; BMI-combined inertial and 2196 displacement-controlled stresses. A Abstract: This report is an in-depth unique experimental facility was sumary of the state-of-the-art in

~

designed and constructed. The piping nuclear piping fracture mechanics. It system evaluated is an expansion loop represents the culmination of 20 years with over 30 meters of 16-inch diameter of work done primarily in the U.S., but Schedule 100 pipe. The experimental also attempts to include important facility is equipped with special aspects from other international hardware to ensure system boundary efforts. Although the focus of this conditions could be appropriately work was for the nuclear industry, the modeled. The test matrix involved one technology is also applicable in many uncracked and five cracked dynamic cases to fossil plants, pipe-system experiments. The uncracked petrochemical / refinery plants, and the experiment was conducted to evaluate oil and gas industry. In compiling this piping system damping and natural detailed summary report, all of the

frequency characteristics. The equations and details of the analysis cracked-pipe experiments evaluated the procedure or experimental results are fracture behavior, pipe system not necessarily included. Rather the response, and stability characteristics report describes the important aspects l

of five different materials. All and limitations, tells the reader where cracked-pipe experiments were conducted he can go for further information, and at PWR conditions. Material more importantly, describes the characterization efforts provided accuracy of the models. Nevertheless.

tensile and fracture toughness the report still contains over 150 l

properties of the different pipe equations and over 400 references. The l

l naterials at various strain rates and main sections of this report describe:

temperatures. Results from all (1) the evolution of piping fracture pipe-system experiments and material mechanics history relative to the  ;

character 12ation efforts are presented. developments of the nuclear industry. I Results of fracture mechanics analyses. (2) technical developments in stress  !

l analyses, material property aspects, l dynamic finite element stress analyses.

and stability analyses are presented and fracture mechanics analyses. (3)  !

and compared with experimental results. unresolved issues and technically l evolving areas, and (4) a summary of NUREG-1426 74

~ _ _ _ . - - _ _ _ . _ _ _ _ . _ .. _ . _ . . _ _ _ . _ . _ . - _ - _ _ _ _

Compilation of Reports: 1994 1998 conclusions of major developments to temperature of 727[ degrees]C during the date. accident. Because of the significance of these results and their importance to the overall analysis of the TMI Pressure Vessel Steels accident a panel of three outside peer reviewers, Dr. Robert W. Bohl, Mr.

Richard G. Gaydos, and Mr. George F.

Vander voort, was formed to conduct an

Title:

Peer review of the Three Mile Island Unit 2 Vessel Investigation independent review of the metallurgical Project metallurgical examinations analyses, After e thorough review of Author (s)/ Editor (s): Bohl, R.W. : the previous analyses and examination

, Gaydos, R.G. . Vander Voort. G.F. , of photo micrographs and actual lower

! Diercks. D.R. (Argonne National Lab., head specimens. the panel determined IL (United States))- that the conclusions resulting from the l Soonsorino Orotnization: NRC; Nuclear INEL study were fundamentally correct.

Regulatory Comission. Washington, DC In p rticular, the panel reaffirmed 1 -(United States) that four lower head samples attained l Publication Date: Jul 1994 temperatures as high as 1100[ degrees]C,

, Recort Number (s): NUREG/CR-6183; ara perhaps as high as l

ANL--94/3 1150--1200[ degrees]Cinonecase.

Order Numb?r: TI94016723 during the accident. They concluded Abstract: Fifteen samples recovered that these samples subsequently cooled L from the lower head of the Three Mile at a rate of Island (TMI) Unit 2 nuclear reactor [ approx]50--125[ degrees]C/ min in tho l toperature range of pressure vessel were subjected to detailed metallurgical examinations by 600--400[ degrees]C, in good agreement the !daho National Engineering with the original analysis. The Laboratory (INEL), with supporting work reviewers also agreed that the carried out by Argonne National remainder of the lower head 2,mples had Laboratory (AN!) and several of the not exceeded the ferrite to-austenite European partic pants. These transformation temperature during the examinations determined that a portion accident and suggested several of the lower head, a so-called refinements and alternative procedures

elliptical hot spot measuring that could have been employed in the

[ approx]0.8 [ times) 1 m. reached original analysis, l_

temperatures as high as 1100[ degree 3]C during the accident and cooled from these temperatures at

Title:

Review of the proposed L -[ approx]10--100[ degrees]C/ min. The materials of construction for the SBWR remainder of the lower head was found and AP600 advanced reactors Author (s)/ Editor (s): Diercks, D.R. :

to have remained below the Shack W.J. . Chung, H.M. : Kassner, ferrite-toaustenite transformation i

75 NUREG-1426 i

l

l l

Compilation of Reports - 1994-1998 T.F. (Argonne National Lab. IL (United selection process. making use of States)) lessons learned from previous LWR Soonsorino Orcanization: NRC: Nuclear experience. The review resulted in the Regulatory Commission. Washington, DC suggestion of alternate an possibly (United States) better materials choices in a number of Publication Date: Jun 1994 cases, and several potential problem Reoort Number (s): NUREG/CR-6223: areas have been cited.

ANL--94/13 Order Number- TI94013716 Abstract: Two advanced ' water

Title:

Unirradiated material reactor (LWR) concepts, rm -y the properties of Midland weld WF-70 General Electric Simplified Boiling Author (s)/ Editor (s): McCabe. D.E. :

Water Reactor (SBWR) and the Nanstad, R.K. . Iskander. S.K. . Swain.

Westinghouse Advanced Passive 600 MWe R.L. (0ak Ridge National Lab. . TN Reactor (AP600), were reviewed in (United States))

detail by Argonne National Laboratory. Soonsorinc Orcanization: NRC: Nuclear The objectives of these reviews were to Regulatory Commission. Washington DC (a) evaluate proposed advanced-reactor (United States) designs and the materials of Publication Date: Oct 1994 construction for the safety systems, Reoort Number (s): NUREG/CR-6249; (b) identify all aging and ORNL/TM--12777 environmentally related degradation Order Number: TI95003010 mechanisms for the materials of Abstract: Weld metal designated construction and (c) evaluate from the WF-70. taken from the nozzle course and safety viewpoint the suitability of the beltline welds of the Midland Reactor, proposed materials for the design Unit 1. has been given a preliminary application. Safety-related systems evaluation using the conventional selected for review for these two LWRs Charpy V-notch (CVN) drop-weight included (a) reactor pressure vessel. (DWT). and chemical analyses. These (b) control rod drive system and tests indicated essentially identical

reactor internals. (c) coolant pressure fracture toughness at both locations, boundary. (d) engineered safety but there was a significant deference systems. (e) steam generators (AP600 in copper content, nominally 0.25%

only). (f) turbines. and (g) fuel versus 0.40%. Because the objective of storage and handlii0 ystem. In this study was to evaluate the addition, the use of cobalt-based before-and-after irradiation alloys in these plants was reviewed. properties, these are regarded as different materials. This report l The selected materials for both reactors were generally sound, and no sumarizes material characterization major selectMn errors were found. It results and presents the results of I

was apparent that considerable thought fracture mechanics tests on the had been given to ine materials unirradiated material to establish NUREG-1426 76 u

Compilation of Reports: 1994-1998 baseline transition tem m ture and J-R of edancements and improvements were curves. Tensile proper w , were also added to the original SCANS program to determined. Five experimental meet requirements unique to storage objectives to be accompluhed from the casks. CASKS is an easy-to-use system testing of irradiated materials were that calculates global response of identified. One of the more important storage casks to impact loads, pressure objectives is to improve the precision loads and thermal conditions. This.

of transition temperature shift and to provides reviewers with a tool for an identify any curve shape changes after independent check on analyses submitted irradiation, concentrating on utilizing by licensees. CASKS is based on data from small surveillance capsule microcomputers compatible with the size specimens. IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and

Title:

CASKS (Computer Analysis of output display programs. All data is l Storage casks): A microcomputer based entered through fill-in-the-blank input analysis system for storage cask design screens that contain descriptive data review. User's manual to Version 1b requests.

l (including rpogram reference)

Author (s)/ Editor (s): Chen, T.F. :

Gerhard, M.A. ; Trummer, D.J. .

Title:

Microstructural Johnson, G.L. : Mok, G.C. (Lawrence characterization of selected AEA/UCSB Livermore National Lab., CA (United model FeCuMn alloys States)) Author (s)/ Editor (s): Rice, P.M. ;

l Soonsorina Oraanization: NRC; Nuclear Stoller, R.E.

i Regulatory Commission, Washington, DC Soonsorina Oraanization: NRC: Nucleer l (United States) Regulatory Commission, Washington, DC Publication Date: Feb 1995 (United States)

Reoort Number (s): NUREG/CR-6242: Publication Date: Jun 1996 UCRL-ID--117418 Recort Number (s): NUREG/CR 6332:

l Order Number: TI95007829 ORNL/TM--12980 l Abstract: CASKS (Computer Analysis of Order Number: TI96012332 l Storage casks) is a microcomputer-based Abstract: A set of 22 model ferritic system of computer programs and alloys was purchased as part of a L databases developed at the Lawrence collaborative research program by the

! Livermore National Laboratory (LLNL) AEA Harwell Laboratory and the for evaluating safety analysis reports University of California at Santa on spent-fuel storage casks. The bulk Barbara. Nine of these alloys were j of the complete program and this user's selected by the Oak Ridge National l manual are based upon the SCANS Laboratory for use in a series of ion I (Shipping Cask Analysis System) program irradiation experiments investigating

( previously developed at LLNL. A number dispersed barrier hardening. These L

. 77 NUREG-1426 i

Compilation of Reports - 1994-1998 nine alloys contain varying amounts of of reactor pressure vessel (RPV) heads copper, manganese, titanium, carbon, and penetrations. The emphasis was to and nitrogen. The alloys have been allow a better understanding of RPV characterized by transmission electron material behavior, to provide guidance I microscopy in the as-received condition supporting reliability and adequate to provide a baseline for comparison performance, and to assist in defining with the irradiated specimens. A directions for further invesdgations.

description of the microstructural The intemational nature of the meeting observations is provided for future is illustrated by the fact that papers reference. This sumary focuses on the ere presented by researchers from 10 type and size distributions of the countries. There were technical experts i

precipitates present: grain size and present from other countries who dislocation measurements are also participated in discussions of the included. results presented. The IAEA issued a Waricing Material version of the meeting papers (IAEA IWG-LMNPP-95/1),

l

Title:

Proceedings of the IAEA and this present document incorporates l the final version of the papers as Specialists' Meeting on Cracking in LWR RPV Head Penetrations received from the authors. The final Author (s)/ Editor (s): C.E. Pugh, S.J. chapter includes conclusions and Raney recomendations.

Soonsorino Oroanization: NRC; Nuclear Regulatory Comission, Washington DC (United States) T.TA;. An Improved Correlation Publication Date: July 1996 Procedure for Subsize and Full-Size Reoort Number (s): NUREG/CP-0151: Charpy Impact Specimen Data ORNL/TM-13187 Author (s)/ Editor (s): M.A. Sokolov, i

Abstract: This report contains 17 0.J. Alexander (0ak Ridge National papers that were presented in four Laboratory) sessions at the IAEA Specialists' Soonsorino Oroanization: NRC:

meeting on Cracking in LkR RPV Head Washington DC (United States)

Penetrations held at ASTM Headquarters Publication Date: March 1997 in Philadelphia on May 2-3, 1995. The Reoort Number (s): NUREG/CR-6379; ORNL-l papers are compiled here in the order 6888 that presentations were made in the Abstract: The possibility r' o ng l subsize specimens to monitor we sessions, and they relate to  !

operational observations, inspection properties of reactor pressure vessel techniques, analytical modeling, and steels is receiving increasing  !

regulatory c0ntrol. The goal of the attention for light-water reactor plant l meeting was to allow international life extension. This potential results (

i experts to review experience in the from the possibility of cutting samples I field of ensuring adequate performance of small volume from the internal NUREG-1426 78

Compilation of Reports: 1994-1998 surface of the pressure vessel for

Title:

The Characterization of determination of the actual properties Vicker's Microhardness Indentations and of the operating pressure vessel. In Pile-Up Profiles as a Strain -

addition, plant life extension will Hardening Microprobe require supplemental data that cannot Author (s)/ Editor (s): C. Santos Jr. , ,

be provided by existing surveillance G.R. Odette, G.E.Lucas. B. Schroeter. '

programs, Testing of subsize specimens D. Klinginsmith T. Yamamoto j

manufactured from broken halves of Soonsorino Oraanization: NRC:

previously tested surveillance Charpy Washington DC (United States) l specimens offers an attractive means of Publication Date: April 1998 j l extending existing surveillance Reoort Number (s): NUREG-1629  ;

l programs. Using subsize Charpy V-notch- Abstract: Microhardness measurements

( type specimers requires the have long been used to examine strength establishment of a specimen geometry properties and changes in strength l l

that is adequate to obtain a ductile- properties in metals, for example, as '

l to-brittle . transition curve similar to induced by irradiation. Microhardness 1 that obtained from full size specimens, affords a relatively simple test that and the development of correlations for can be applied to very small volumes of j l transition temperature and upper-shelf material. Microhardness is nominally

! energy (USE) level between sub size and related to the flow stress of the full-size specimens. Five different material at a fixed level of plastic l

geometries of subsize specimens were strain. Further, the geometry of the selected for testing and evaluation. pile-up of material around the The specimens were made from several indentation is related to the types of pressure vessel steels with a strain hardening behavior of the l wide range of yield strengths, material; steeper pile-ups correspond l transition temperatures, and USES. The to smaller strain hardening rates. In l effects of specimen dimensions, this study the relationship between j including notch depth, angle, and pile-up profiles and strain hardening l radius, have been studied. The is examined using both experimental and

!- correlations of transition temperatures analytical methods. Vicker's determined from different types of microhardness tests have been performed I subsize specimens and the full-size on a variety of metal alloys including  !

specimens are presented. A new low alloy, high Cr and austenitic procedure for transforming data from stainless steels. The pile-up topology subsize specimens is developed. The around the indentations has been l transformed data are in good agreement quantified using confocal microscopy  !

with data from full-size specimens for techniques. In addition, the

materials that have USE levels less indentation and pile up geometry has than 200 J. been simulated using finite element method techniques. These results have been used to develop improved i

i j 79 NUREG-1426 l

Compilation of Reports - 1994-1998 quantification of the relationship data points fo; welo materials (105 between pile up geometry and the s'. rain different welds), 524 data points for hardening constitutive behavior of the base materials (136 different base test material. materials), including 297 plate data points (85 different plates), 119 forging data points (31) different forging) and 108 correlation monitor Radiation Embrittlement materials data points (3 different plates). The data files are given in dBASE format and can be accessed with

Title:

PR-EDB: Power Reactor any computer using the DOS operating Embrittlement Data Base. Version 2 system. User-friendly utility Author (s)/ Editor (s): Stallmann, F.W. :

programs are used to retrieve and Wang, J.A. : Kam. F.B.K. : Taylor, B.J. select specific data, manipulate data, (Oak Ridge National Lab., TN (United display data to the screen or printer.

States)) and to fit and plot Charpy impact data.

Soonsorino Oraanization: NRC: Nuclear The results of several studies ,

Regulatory Commission Washington DC investigated are presented in Appendix (United States) 0.

Publication Date: Jan 1994 Reoort Number (s): NUREG/CR-4816 Rev.2:

ORNL/TM--10328/R2

Title:

TR-E08: Test Reactor Order Number: TI94006457 Embrittlement Data Base, Version 1 Abstract: Investigations of regulatory Author (s)/ Editor (s): Stallmann. F.W. :

issues such as vessel integrity over Wang, J.A. : Kam, F.B.K. (0ak Ridge plant life, vessel failure. and National Lab., TN (United States))

sufficiency of current codes Standard Soonsorino Oraanization: NRC: Nuclear Review Plans (SRP's) and Guides for Regulatory Commission, Washington, DC license renewal can be greatly e t expedited by the use of a well designed computerized datc base. Also, such a Reoort Number (s): NUREG/CR 6076:

data base is essential for the NL/T 415 validation of embrittlement prediction models by researchers. The Power Abstract: The Test Reactor Reactor Embrittlement Data Base Embrittlement Data Base (TR-EDB) is a (PR-E08) is such a comprehensive collection of results from irradiation collection of data for US comercial in materials test reactors. It nuclear reactors. The current version complements the Power Reactor of the PR-EDB contains the Charpy test Embrittlement Data Base (PR EDB), whose data that were irradiated in 252 data are restricted to the results from capsules of 96 reactors and consists of the analysis of surveillance capsules 207 data points for heat-affected zone in commercial power reactors. The (HAZ) materials (98 different HAZ). 227 NUREG 1426 80 i

____._-.m___ _ _ _ . _ . _ _ - _ _ _ . _ _ . _ -_ _ _ _ _ __ _ .__ _ _

i-Compilation of Reports: 1994-1998 4

rationale behind their restriction was 0.31 wt % were commercially fabricated the assumption that the results of test in 220-m -thick plate. Crack-arrest reactor experiments may not be specimens fabricated from these welds applicable to power reactors and could, were irradiated at a nominal therefore, be challenged if such data temperature of 288[ degrees]C to an were included. For this very reason average fluence of 1.9 [ times) 10[sup

the embrittlement predictions in the 19] neutrons /cm[sup 2] (>l MeV). This Reg. Guide 1.99, Rev. 2, were based is the second report giving the results 1 exclusively on power reactor data. of the tests on irradiated duplex-type However, test reactor experiments are crack-arrest specimens. A previous 4

able to cover a much wider range of report gave results of tests on materials and irradiation conditions irradiated weld-embrittled type that are needed to explore more fully a specimens. Charpy V-notch (CVN) variety of models for the prediction of specimens irradiated in the same irradiation embrittlement. These data capsules as the crack-arrest specimens are also needed for the study of were also tested, and a 41-J transition effects of annealing for life extension temperature shift was determined from ,

j of reactor pressure vessels that are these specimens. (open I difficult to obtain from surveillance quotes]Mean[close quote] curves of the i

! capsule results. same form as the American Society of I

) Mechanical Engineers (ASME) K[sub la] ,

curve were fit to the data with only l Title- Crack-arrest tests on two the [open quotes] reference irradiated high-copper welds temperature [close quotes) as a Author (s)/ Editor (s): 1skander S.K. : parameter. The shift between the mean Corwin W.R. ; Nanstad. R.K. (0ak Ridge curves agrees well with the 41-J i National Lab., TN (United States)) transition temperature shift obtained Soonsorino Oraanization: NRC: Nuclear from the CVN specimen tests. Moreover, Regulatory Comission. Washington, DC the fcur data points resulting from (United States) tests on the duplex crack arrest j Publication Date: Mar 1994 specimens of the present study did not Reoort Number (s): NUREG/CR-6139: make a significant change to mean curve ORNL/TM--12513 fits to either the previously obtained Order Number: T194007831 data or all the data combined.

Abstract: The objective of the l Heavy-Section Steel Irradiation Program '

- Sixth Irradiation Series is to

Title:

Heavy-Section Steel Irradiation determine the effect of neutron Program. Volume 2. No. 1: Semiannual

! . irradiation on the shift and shape of progress report, October 1990--March

[ the lower-bound curve to crad-arrest toughness data. Two submerged- arc 1991 Author (s)/ Editor (s): Corwin. W.R. (0ak

welds with copper contents of 0.23 and Ridge National Lab., TN (United f

i 81 NUREG-1426 i

1 Compilation of Reports - 1994-1998 i States)) of irradiation, (8) in- service aced Soonsorina Oraanization: NRC: Nuclear material evaluations, and (9)

Regulatory Commission. Washington, DC correlation monitor materials. During (United States) this period, additional analyses on the Publication Date: Jul 1994 effects of precleavage stable ductile Recort Number (s): tearing on the toughness of high copper NUREG/CR-5591-Vol.2-No.1: welds 72W and 73W demonstrated that the i ORNL/TM -11568 Vol.2-No.1 size effects observed in the transition Order Number: TI94016626 region are not due to substantial l Abstract: -Maintaining the integrity of differences in ductile tearing the reactor pressure vessel (RPV) in a behavior. Possible modifications to light-water-cooled nuclear power plant irradiated duplex crack-arrest l 1s crucial in preventing and specimens were examined to increase the

-controlling severe accidents that have likelihood of their successful testing.

the potential for major contamination Characterization of a second batch of release. The RPV is the only key 72W and 73W welds was begun and results safety-related component of the plant of the Charpy V-notch testing is for which a duplicate or redundant provided. A review of literature on  ;

i backup system does not exist. It is 'he annealing response of reactor

! therefore imperative to understand and pressure vessel steels was initiated.

be able to predict the capabilities and limitations of the integrity inherent in the RPV. For this reason, the

Title:

Heavy-Section Steel Irradiation Heavy-Section Steel Irradiation (HSSI) Program Program has been established with its Author (s)/ Editor (s): Corwin, W.R. (0ak primary goal to provide a thorough, Ridge National Lab., TN (United quantitative assessment of the effects States))

of neutron irradiation on the material Soonsorino Oraanization: NRC: Nuclear

behavior, and in particular the Regulatory Comission, Washington DC fracture toughness properties, of (United States) typical pressure-vessel steels as they Publication Date

Oct 1994 relate to light-water reactor pressure. Reoort Number (s):

vessel integrity. The HSSI Program NUREG/CR-5591-Vol.2-No.2:

is arranged into nine tasks: (1) ORNL/TM--1?568-Vol.2-No.2 program management, (2) K[sub ic] curve Order Number: TI95001892 l shift in high copper welds, (3) K[sub Abstract:. Goal is to provide a l ia) curve shift in high-copper welds, thorough, quantitative assessment of i (4) irradiation effects on cladding, the effects of neutron irradiation on ,

the material behavior, and in l (5) K[sub ic] and K[sub ia) curve shifts in low upper-shelf (LUS) weld, particular the fracture toughness l (6) irradiation effects in a commercial properties, of typical pressure vessel l

LUS weld. (7) microstructural analysis stools as they relate to light-NUREG-1426 82 ,

L i -. - . . . - ._. - __ __ . _ _ _ _

Compilation of Reports: 1994-1998 water reactor pressure-vessel that act to impede dislocation motion, integrity. Effects of specimen size, Radiation-induced point defect clusters material chemistry, product form and (PDC) and radiation-enhanced, copper-microstructure, irradiation fluence, rich precipitates (CRP) provide two flux, temperature and spectrum, and plausible sources of this matrix post-irradiation annealing are being hardening. These PDC can be of either examined on a wide range of fracture interstitial or vacancy type and could properties. The HSSI Program is into exist in either two- or three-10 tasks: (1) program management, (2) dimensional morphologies, e.g., small K[sub Ic] curve shift in high-copper loops, voids, or stacking fault welds, (3) K[sub la] curve shift in tetrahedra. The formation and evolution high-copper welds (4) irradiation of PDC are primarily determined by the ,

, effects on cladding. (5) K[sub Ic] and displacement damage rate and K[sub la] curve shifts in low irradiation temperature. There is upper-shelf welds, (6) irradiation experimental evidence that the the effects in a comercial low upper-sheer distributions of these clusters are weld, (7) microstructural analysis of also influenced by impurities such as irradiation effects. (8) in-service copper. A theoretical model has been aged material evaluations (9) developed to investigate the relative correlation monitor materials, and (10) importance ofPDC and CRP in RPV special technical assistance. This embrittlement. The model includes a report provides an overview of the detailed description of the interstial activities within each of these tasks cluster population: vacancy clustering from April to September 1991. and copper precipitation are treated in a more approximate fashion. The model has been used to examine a broad range

Title:

A Comparison of the Relative of irradiation and material parameters.

Importance of Copper Precipitates and The results indicate that there are Point Defect Clusters in Reactor temperature and displacement rate Pressure Vessel Embrittlement regimes wherein either CRP or PDC can Author (s)/ Editor (s): R.E. Stoller (0ak dominate the material's response to Ridge National Laboratory) irradiation. Both interstitial and Soonsorina Oraanization: NRC: vacancy-type defects contribute to the Washington DC (United States) PDC component. wth their relative Publication Date: December 1994 importance determined by the specific Reoort Number (sh NUREG/CR-6231: irradiation conditions. The varying ORNL/TM 6811 dependencies of the CRP and PDC on Abstract: The embrittlement of temperature and displacement rate irradiated reactor pressurevessel (RPV, indicate that simple data steels is believed to arise primarily extrapolations could lead to poor from the hardening of the material due predictions of RPV embrittlement.

to the formation of extended defects 83 NUREG 1426

Compilation of Reports - 1994-1998

Title:

Heavy-Section Seteel assistance. This report provides an Irradiation Program: Volume 3. Progress overview of the activities within each report. October 1991--September 1992 of these tasks from October 1991 to Author (s)/ Editor (s): Corwin. W.R. (0ak September 1992.

Ridge National Lab., TN (United States))

Soonsorina Oraanization: NRC: Nuclear

Title:

Heavy-section steel irradiation Regulatory Comission. Washington. DC program. Volume 4. No. 2. Semiannual (United States) progress report. April 1993--September Publication Date: Feb 1995 1993 Reoort Number (s): NUREG/CR-5591-Vol.3: Author (s)/ Editor (s): C. win W.R. (0ak ORNL/TM--11568-Vol.3 Ridge National Lab., TN (United On@rNumber: TI95007768 States))

Abstract: The primary goal- of the Soonsorino Orcanization: NRC: Nuclear Heavy-Section Steel Irradiatir Program Regulatory Comission. Washington. DC is to provide a thorough, qub.citative (United States) assessment of the effects of neutron Publication Date: Mar 1995 1rradiation on the material behavior. Reoort Number (s):

and in particular the fracture NUREG/CR-5591-Vol.4-No.2:

toughness properties, of typical ORNL/TM--11568/V4-N2 pressure vessel steels as they relate Order Number: T195009673 to light-water reactor pressure- Abstract: Maintaining the integrity of vessel integrity. Effects of specimen the reactor pressure vessel (RPV) in a f light-water-cooled nuclear power plant l size, material chemistry, product form and microstructure, irradiation is crucial in preventing and fluence, flux. temperature and controlling severe accidents which have spectrum, and postirradiation annealing the potential for major contamination are being examined on a wide range of release. The RPV is the only key fracture pronerties. The HSSI Program safety-related component of the plant is arranged into 10 tasks: (1) program for which a duplicate or redundant management, (2) K[sub Ic] curve shift backup system does not exist. In in high-copper welds. (3) K[sub la) particular. it is vital to fully curve shift in high- copper welds, understand the degree of (4) irradiation effects on cladding, irradiation-induced degradation of the (5) K[sub Ic] and K[sub Ia) curve RPV's fracture resistance which occurs shifts in low upper-shelf welds. (6) during service, since without that irradiation ffects in a comercial low radiation damage, it 's virtually upper-shelf weld. (7) microstructural impossible to postulate a realistic analysis of irradiation effects. (8) scenario that would result in RPV service aged material failure. For this reason, the in-evaluations. (9) correlation monitor Heavy-Section Steel Irradiation (HSSI) materials, and (10) special technical Program has been established to provide NUREG-1426 84

Compilation of Reports: 1994-1998 a quantitative assessment of the Soonsorinc Orcanization: NRC: Nuclear effects of neutron 1rradiation on the Regulatory Commission Washington, DC material behavior and. in particular, (United States) the fracture toughness properties of Publication Dat t Apr 1995 typical pressure vessel steels. Reoort Nunber(s):

Effects of soecimen size: material NUREG/CR-5591-Vol.5-No.1:

chemistry; product form and ORNL/TM--11568/V5-N1 microstructure: irradiation fluence, Order Nurrber: TI95010954 flux, temperature, and spectrum: and Abstract- Maintaining the integrity of postirradiation annealing are being the reactor pressure vessel (RPV) in a examined on a wide range of fracture light-water-cooled nuclear power plant properties. The HSSI Program is is crucial in preventing and arranged into 14 tasks: (1) program controlling severe accidents that have management. (2) fracture toughness the potential for major contamination (K[sub ic]) curve shift in high-copper release. The RPV is the only component welds, (3) crack-arrest toughness in the primary pressure boundary for (K[sub la]) curve shift in high- which, if it should rupture, the copper welds, (4) irradiation effects engineering safety systems cannot on cladding, (5) K[sub lc] and K[sub assure protection from core damage. It la] curve shifts in low upper-shelf is therefore imperative to understand (LUS) welds, (6) annealing effects in and be able to predict the capabilities LUS welds, (7) irradiation effects in a and limitations of the integrity comercial LUS weld, (8) inherent in the RPV. In particular, ft microstructural analysis of irradiation is vital to fully understand the degree effects, (9) in-service aged material of irradiation-induced degradation of evaluations (10) correlation monitor the RPV's fracture resistance that materials, (11) special technical occurs during service. The assistance, (12) Japan Power Heavy Section Steel (HSS) Irradiation Development Reactor steel examination. Program has been established: its (13) technical assistance for Joint primary goal is to provide a thorough, Coordinating Comittee on Civilian quantitative assessment of the effects Nuclear Reactor Safety (JCCCNRS) of neutron irradiation on the material Working Groups 3 and 12 and (14) behavior, and in particular the additional requirements for materials. fracture toughness properties of typical pressure-vessel steels, as they relate to light-water RPV integrity.

Title:

Heavy-section steel irradiation The program incit. des the direct program. Semiannual progress report, continuation of irradiation studies September 1993--March 1994 previously conducted within the HSS Author (s)/ Editor (s): Corwin, W.R. (Oak Technology Program augmented by Ridge National Lab. , TN (United enhanced axaminations of the States)) accompanying microstructural changes.

85 NUREG-1426

Compilation of Reports - 1994 1993 During this period, the report on the irradiation on the material behavior duplex-type crack-arrest specimen tests and the fracture toughness properties from Phase 11 of the K[sub la] program of typical pressure-vessel steels as was issued, and final preparations for they relate to light- water RPV testing the large, irradiated integrity. Effects of specimen size:

crack-arrest specimens from the Italian material chemistry; product form and Comittee for Research and Development microstructure: 1rradiation fluence.

of Nuclear Energy and Alternative flux, temperature, and spectrum: and Energies were completed. Tests on postirradiation annealing are being undersize Charpy V notch (CVN) energy examined on a wide range of fracture specimens in the irradiated and properties . The HSSI Program is annealed weld 73W were completed. The arranged into 14 tasks: (1) program results are described in detail in a management. (2) fracture toughness draft NUREG report. In addition. the curve shift in high-copper weldments ORNL investigation of the embrittlement (Series 5 and 6). (3) K[sub ic] and of the High Flux Isotope RPV indicated K[sub la] curve shifts in low that an unusually large ratio of the upper-shelf (LUS) welds (Series 8). (4) high-energy gama- ray flux to irradiation effects in a comercial LUS fast-neutron flux is most likely weld (Series 10). (5) irradiation responsible for the apparently effects on weld heat-affected zone and accelerated embrittlement. plate materials (Series 11). (6) annealing effects in LUS welds (Series 9), (7) microstructural and

Title:

_ Heavy-Section Steel Irradiation microfracture analysis of irradiation Program. Volume 5. No. 2. Progress effects. (8) in-service irradiated and report. April 1994--September 1994. aged material evaluations. (9) Japan Power Development Reactor (JPOR) steel Author (s)/ Editor (s): Corwin. W.R. (0ak Ridge National Lab. . TN (United examination, (10) fracture toughness curve shift method. (11) special States))

Soonsorina Oraanization: NRC: Nuclear technical assistance, (12) technical Regulatory Comission. Washington. DC assistance for Joint Coordinating (United States) Comittee on Civilian Nuclear Reactor Publication Date Jul 1995 Safety (JCCCNRS) Working Groups 3 and Reoort Number (s): 12. (13) correlation monitor materials.

h"JREG/CR 5591-Vol .5 No.2: and (14) test reactor coordination.

CRNL/TM--11568-Vnl.5-No.2 Progress on each task is reported.

Order Number: T195016159 Abstract: The Heavy Section Steel Irradiation (HSSI) Program has been

Title:

Heavy-section stael irradiation established with its primary goal to program. Progress report. October provide a thorough. quantitative 1994-- March 1995 assessment of the effects of neutron Author (s)/ Editor (s): Corwin. W.R. (0ak NUREG-1426 86

Compilation of Reports: 1994-1998 Ridge National Lab., TN (United Safety Issue No. 15. (GSI-15).

States)) " Radiation Effects On Reactor Pressure Publication Date: Oct 1995 Vessel Supports." GSI-15 was Re00rt Number (s): established to evaluate the potential NUREG/CR-5591-Vol.6-No.1 for low-temperature, low-flux level Order Nuriber: DE96002233 neutron irradiation to embrittle Abstract: This task was established to reactor pressure vessel (RPV) supports supply and coordinate irradiation to the point of compromising plant services needed by NRC contractors safety. An evaluation of surveillance other than ORNL. These services samples from the high flux isotope include the design and assembly of reactor (HFIR) at the Oak Ridge irradiation capsules as well as National Laboratory (ORNL) had arranging for their exposure, suggested that some materials used for disassembly, and return of specimens. RPV supports in pressurized-water During this period, the final design of reactors could exhibit higher than the facility and specimen baskets was expected embrittlement rates. However, determined through an iterative process further tests designed to evaluate the involving the designers and thermal applicability of the HFIR data to analysts. The resulting design should reactor RPV supports under operating permit the irradiation of all test conditions led to the conclusion that specimens to within 5[ degrees]C of RPV supports could be evaluated using their desired temperature. Detailing traditional methods. It was found that of all parts is ongoing and should be the unique HFIR radiation environment completed during the next reporting allowed the gama radiation to period. Procurement of the facility contribute significantly to the will also be initiated during the next embrittlement. The shielding provided review period. by the thick steel RPV shell ensures that degradation of RPV supports from gama irradiation is improbable or

Title:

Radiation Effects of Reactor minimal.

Pressure Vessel Supports Author (s)/ Editor (s): R.E. Johnson, and The findings reported herein were used.

R.E. Lipinski in part, as the basis for technical Soonsorino OrcanizatiCn:. NRC: resolution of the issue.

Washington DC (United States)

Publication Date: May 1996 Reoort Number (s): NUREG-1509

Title:

Analysis of the irradiation Abstract: The purpose of this report is data for A30?B and A533B correlation to present the findings from the work monitor materials done in accordance with the Task Action Author (s)/ Editor (s): Wang, J.A.

Plan developed to resolve the Nuclear Soonsorino Orcanization: NRC: Nuclear Regulatory Comission (NRC) Generic Regulatory Comission. Washington. DC 87 NUREG-1426

Compilation of Reports - 1994-1998 (United States) analysis, irradiation environments.

Publication Date: Apr 1996 fluence evaluation and inhomogeneous Reoort Nunber(s): NUREG/CR 6413: material properties. Thus in order to ORNL/TM--13133 improve the prediction model control Order hmber: TI96010415 of the above-mentioned error sources Abstract: The results of Charpy needs to be improved. In general the V-notch impact tests for A302B and embrittlement behavior of both the A5333-1 Correlation Monitor Materials A302B and A533B-1 plate materials is l (CMM) listed in the surveillance power similar. There is evidence for a reactor data base (PR-EDB) and material fluence-rate effect in the CMM data test reactor data base (TR-EDB) are irradiated in test reactors: thus its analyzed. The shift of the transition implication on power reactor temperature at 30 ft-lb (T[sub 30]) is surveillance programs deserves special considered as the primary measure of attention.

radiation embrittlement in this report.

The hyperbolic tangent fitting model and uncertainty of the fitting

Title:

Heavy Section Steel Irradiation parameters for Charpy impact tests are Program Progress Report for April -

presented in this report. Fu the September 1995 surveillance CMM data the transition Author (s)/ Editor (s): W.R. Corwin temperature shifts at 30 ft-lb Soorsorina Oroanization: NRC:

([ Delta]T(sub 30]) generally follow the Washington DC (United States) predictions provided by Revision 2 of Publication Date: August 1996 Regulatory Guide 1.99 (R.G. 1.99). EeportNumber(s) NUREG/CR-5591 Vol 6 Difference in capsule temperatures is a No 2/0RNL/TM-11568/V6&N2 likely explanation for large deviations Abstract Maintaining the integrity of from R.G. 1.99 predictions. the reactor pressure vessel (RPV) in a Deviations from the R.G. 1.99 light-water-cooled nuclear power plant predictions are correlated to similar is crucial in preventing and deviations for the accompanying controlling severe accidents which have materials in the same capsules, but the potential for major contamination large random fluctuations prevent release. The RPV is the only key precise quantitative determination. safety-related component of the plant Significant scatter is noted in the for which a duplicate or redundant surveillance data, some of which may be backup system does not exist. It is attributed to variations from one therefore imperative to understand and specimen set to another, or inherent in be able to predict the c6pabilities and Charpy V notch testing. The major limitations of the integrity inherent contributions to the uncertainty of the in the RPV. In particular, it is vital R.G. 1.99 prediction model, and the to fully understand the degree of overall data scatter are from irradiation-induced degradation of the mechanical test results, chemical RPV's fracture resistance which occurs l

NUREu-1426 88

Compilation of Reports: 1994-1998 during Service, since without that Committee on Civilian Nuclear Reactor radiation damage. H is virtually Safety (JCCCNRS) Working Groups 3 and impossible to postulate a realistic 12. (13) correlation monitor materials, scenario that would result in RPV and (14) test reactor coordination.

failure.

During this period, results of testing For this reason, the Heavy-Section the Italian crack-arrest specimens were Steel Irradiation (HSSI) Program has analyzed and a draft NUREG report been established with its primary goal prepared. A test plan was developed for to provide a thorough quantitative irradiation of HSSI weld 73W to a high assessment of the effects of neutron fluence [5 x 10'9 neutrons /cm2 (> 1 1rradiation on the material behavior MeV)] to determine whether the KJC and, in particular, the fracture curve shape change observed in the toughness properties of typical Fifth Series is exacerbated. ihe pressure-vessel steels as they relate fabrication of the third of the three to light-water RPV integrity. Effects trial LUS scoping welds to identify of specimen size: material chemistry: possible materials for studies on KIC product form and microstructure: shifts in LUS materials was completed.

irradiation fluence, flux, temperature. Data from fracture mechanics testing of and spectrum: and postirradiation specimens of the irradiated LUS Midland annealing are being examined on a wide Weld WF-70 from both scoping capsules range of fracture properties. The HSSI and the first large capsule (exposed to Program is arranged into 14 tasks: (1) 0.5 and 1.0 x 1019 neutrons /cm2 (~1 program management. (2) fracture MeV) respectively ] was completed and toughness curve shift in high-copper the results reported. Precracked Charpy weldments (Series 5 and 6), (3) KIC and specimens of beltline weld were also Kl. curve shifts in low upper-shelf tested in the unirradiated and (LUS) welds (Series 8). (4) irradiation irradiated condHions and showed an effects in a corrmercial LUS weld irradiation-induced fracture toughness (Series 10). (5) irradiation effects on shift very close to that indicated by weld heat-affected zone (HAZ) and plate the compact specimens. The second large materials (Series 11). (6) cnnealing capsule that was shipped to Oak Ridge effects in LUS welds (Series 9), (7) National Laboratory (ORNL) has been microstructural and microfracture disassembled, and the specimens are analysis of irradiation effects. (8) awaiting testing in the hot cell.

in-service irradiated and aged material Arrangements were made with Yankee evaluations. (9) Japan Power Atomic Electric Company for the Development Reactor (JPDR) steel procurement of two A 302 grade B examination. (10) fracture toughness plates identified for examination of curve shift method. (11) special the effects of neutron irradiation on technical assistance. (12) technical the fracture toughness of the HAZ of assistance for Joint Coordination welds of plate materials typical of 89 NUREG-1426

Compilation of Reports - 1994-1993 those used in fabricating older RPVs. shown that, on average, the fracture Microstructural characterizations of toughness shifts generally exceeded the long-term (- 100.000 h) thermally aged. Charpy 41-J shifts by about 20. 50, and neutron-irradiated, and annealed 8% for plates, forgings, and welds, surveillance materials were completed, respectively. Overall. the fracture An atom-probe field-ion microscopy toughness shifts exceed the CVN shifts characterization of a simple thermally by about 14%. similar to the results aged model alloy was performed to reported previously from analysis of investigate its suitability as a model HSSI Program data. Evaluation of the for comercial RPV steels. The validity precracked cylindrical tensile specimen of the low-load 'nanoindentation" continued with a report produced for technique to monitor strength changes ORNL by SRI International and AEA was established. A comparison of model Technology. Harwell United Kingdom, predictions and available data seemed regarding their test results. A to confirm that the lead factors detailed plan was developed for removal comonly employed in comercial of material from the pressurized-water surveillance programs should have RPV the Pressure Vessel Research negligible impact on the validity of User's Facility located at the Oak the data obtained. Modification of the Ridge Gaseous Diffusion Plant (K-25 computer numerically controlled site). The remaining correlation machining center for irradiated monitor materials were moved from the materiais continued with the completion storage area at the Y-12 Plant and of the drawings, new cables and table, placed into the HSSI storage facility machine enclosure, fittings, and a at the ORNL site. Two blocks of floor tub for installation inside the Heavy-Section Steel Technology (HSST) hot cell. Tensile and Charpy V-notch plate 03 were sent to the Hanjung impact tests of type 308 stainless America Corp. for use as correlation steel weld metals aced at 343*C for up monitor materials in Units 3 and 4 of to 50.000 h showed that aging had the Ulchin Nuclear Power Plant ir little effect on the tensile Korea. Most of the engineering drawings properties, but did result in for the irradiation facility and embrittiement as shown by the impact specimen baskets for University of testing. The baseline testing for the California. Santa Barbara, irradiations cross comparison of the effects of the were completed, and procurement and different tups used in U.S. and fabrication of selected portions of the Japanese Charpy impact machines was facility were initiated.

performed to provide a basis for understanding any differences that might later arise from jointly testing

Title:

Heavy Section Steel Irradiation the JPOR materials. Available fracture Program Semiannual Progress Report for toughness databases were analyzed for October 1995 March 1996 plates, forgings, and welds, and H was Author (s)/ Editor (s): W.R. Corwin NUREG-1426 90

Compilation of Reports: 1994 1998 i

Soonsorino Oraanization: NRC: annealing are being examined on a wide Washington DC (United States) range of fracture properties. The HSSI Publication Date: April 1997 Program is arranged into 14 tasks: (1)

Reoort Number (s): NUREG/CR-5591 Vol7 program management (2) fracture No 1/0RNL/TM-11568V7&N1 toughness curve shift in high-copper Abstract: Maintaining the integrity of weldments (Series 5 and 6), (3) KIc and the reactor pressure vessel (RPV) in a K., curve shifts in low upper shelf light water-cooled nuclear power plant (LUS) welds (Series 8), (4) irradiation is crucial in preventing and effects in a commercial LUS weld controlling severe accidents which have (Series 10), (5) irradiation effects on the potential for major contamination weld heat-affected zone (HAZ) and plate release. The RPV is the only key materials (Series 11), (6) annealing safety related component of the plant effects in LUS welds (Series 9), (7) for which a duplicate or redundant microstructural and microfracture backup system does not exist. It is analysis of irradiation effects (8) therefore imperative to understand and in-service irradiated and aged material be able to predict the capabilities and evaluations, (9) Japan Power limitations of the integrity inherent Development Reactor (JPDR) steel in the RPV, In particular, it is vital examination, (10) fracture toughness to fully understand the degree of curve shift method, (11) special irradiation-induced degradation of the technical assistance (12) technical RPV's fracture resistance which occurs assistance for Joint Coordinabng during service, since without that Cocmittee on Civilian Nuclear Reactor radiation damage, it is virtually Safety (JCCCNRS) Working Groups 3 and impossible to postulate a realistic 12, (13) correlation monitor materials, scenario that would result in RPV and (14) test reactor coordination, failure.

During this period, a draft NUREG For this reason, the Heavy Section report was prepared describing the Steel Irradiation (HSSI) Program has results of testing the irradiated been established with its primary goal Italian crack arrest specimens The to provide a thorough, quantitative fabrication of the third of the three assessment of the effects of neutron trial LUS scoping welds was completed.

1rradiation on the material behavior Charpy V-notch (CVN) specimens from the and, in particular, the fracture trial weld which was fabricated with toughness properties of typical HSSI weld 73W weld wire and Linde 80 pressure-vessel steels as they relate flux to identify possible materials for to light-water RPV integrity. Effects studies on K,e shifts in LUS materials of specimen size: material chemistry: have been tested showing a relatively product form and microstructure, small reduction in upper-shelf energy irradiation fluence, flux temperature, from 136 J for the original weld and spectrum: and postirradiation (fabricated with Linde 124 flux) to 121 91 NUREG-1426

Compilation of Reports - 1994 1998 J for the biaxial weld made with Linde simulations were used to develop 80 flux. Data from fracture mechanics effective defect production cross testing of specimens of the irradiated sections for relevant reactor neutron LUS Midland Weld WF 70 from both spest a. Modification of the computer scoping capsules and both large numerically controlled (CNC) machining capsules [ exposed to 0.5 and 1.0 x 10-9 center continued with ail drawings neutrons /cm2 (> 1 MeV). respectively completed and new cables and table, were analyzed. An A 302 grade B plate machine enclosure. fittings, and a was procured from Yankee Atomic floor tub for installation inside the Electric Company for examination of the hot cell procured. Tensile and CVN effects of neutron irradiation on the impact tests of type 308 stainless fracture toughness of the HAZ of welds steel weld metals aged at 343'C for up of plate materials typical of those to 50.000 h showed that aging had used in fabricating older RPVs. little effect on the tensile properties Detailed planning was performed and but did result in embrittlement as work begun to examine grain boundary shown by the impact testing. Planning segregation of phosphorous and was initiated at Oak Ridge National resultant intergranular fracture of Laboratory (0RNL) for the machining of ,

steel heat treated to give large, prior the JPOR vessel trepans, and recent austenite grains such as would be found studies conducted in Japan have shown in the HAZ. Tne annealing and testing that the through-wall attenuation is of specimens irradiated within capsule somewhat greater than would be 10.06 was completed and planning of the predicted by the attenuation formula in specimen complement for the first Regu/atory Gufde 1.99. As part of the reirradiation capsule begun. Two evaluation of the database of Charpy irradiation, annealing, and impact and fracture toughness data for reirradiation facilities: data RPV steels, instrumented CVN and acquisition and control dynamic precracked CVN tests were instrumentation; and the associated analyzed for potential use in reusable temperature verification estimating various toughness capsule have been fabricated and parameters. The end of unstable crack assembled. The analysis of solute propagation indicated by the effects in ion-irradiated model alloys load-displacement record was compared was largely completed, indicating a to the drop weight strong effect of copper and a strong nil-ductility-transition temperature copper-manganese interaction. The (NOT) and crack-arrest toughness tests effect of the interstitial solutes showing results that are encouraging nitrogen and carbon was more modest. A with regard to the potential use of the detailed comparison of neutron flux and instrumented CVN test record to provide spectral effects on tensile properties a reasonable estimate of the NOT at 50 to 60'C was :ompleted. The temperature and the crack-arrest results of molecular dynamics cascade toughness for RPV steels. Preparations NUREG-1426 92

Compilation of Reports: 1994-1998 continued for the characterization of Order Number: TI97005912 the beltline weld from a Abstract: A capsule containing Charpy pressurized-water reactor pressure V notch (CVN) and mini-tensile vessel, the Pressure Vessel Research specimens was irradiated at User's Facility (PVRUF). located at the [approximately) 30[ degrees]C Oak Ridge K-25 plant site. Material ([approximately] 85[ degrees]F) in the will be removed for experimental cavity of a corrercial nuclear power projects within the HSSI and plant to a fluence of 1 x 10[sup 16]

Heavy-Section Steel Technology (HSST) neutrons /cm[sup 2] (> 1MeV). The programs at ORNL. as well as the capsule included six CVN impact Pacific Northwest National Laboratory specimens of archival High Flux Isotope Nondestructive Evaluation Program. The Reactor A212 grade 8 ferritic ateel and CVN and round tensile specimens of two five CVN impact specimens of a Russian weld metals irradiated in HSSI well-studied A36 structural steel.

capsule 10.06 were returned to ORNL This irradiation was part of the where the capsule was disassembled and ongoing study of neutro1-induced damage preparations made for specimens to be effects at the low temperature and flux tested in both the irradiated and experienced by reactor supports. The thermally annealed conditions. The plant operators shut down the plant remainder of the engineering drawings before the planned exposure was for the irradiation facility and reached. The exposure of these specimen baskets for University of specimens produced no significant California. Santa Barbara, irradiatiens irradiation-induced embrittlement. Of were completed, and procurement and interest were the data on unirradiated fabrication of selected portions of the specimens in the L-T orientation facility were continued. machined from a single plate of A36 structural steel, which is the same specification for the structural steel

Title:

Results of charpy V-notch used in some reactor supports. The impact testing of structural steel average CVN energy of five unirradiated specimens irradiated at specimens obtained from one region of

[approximately]30[ degrees]C to 1 x the plate and tested at room 10[sup 16] neutrons /cm[sup 2] in a temperature was [approximately] 99 J.

comercial reactor cavity while the energy of 11 unirradiated Author (s)/ Editor (s): Iskander. S.K. . specimens from other locations of the Stoller. R.E. same plate was 43 J. a difference of Soonsorino Oroanization: NRC: Nuclear [approximately) 220%. The CVN impact Regulatory Connission. Washington DC energies for all 18 specimens ranged (United States) from a low of 32 J to a high of 111 J.

Publication Date: Apr 1997 Moreover, it appears that the ReDort Number (s): NUREG/CR 6399: University of Kansas CVN impact energy ORNL--6886 data of the unirradiated specimens at 93 NUREG 1426

Compilation of Reports - 1994-1993 l

the 100 J level are shifted toward (CVN) specimens were aged at various higher temperatures by about 20 K. The temperatures and tested to examine the results were an example of the extent reason for overrecovery of upper shelf of scatter possible in CVN impact energy that has been observed.

testing. Generic values for the CVN Molecular dynamics cascade simulations impact energy of A36 should be used were extended to 40 kev and have with caution in critical applications. provided information representative of most of the fast neutron spectrum.

Investigations of the correlation

Title:

Heavy section steel irradiation between micros'.ructural changes and program. Progress report. April 1996-- hardness changes in irradiated model September 1996 alloys was also completed. Preliminary Author (s)/ Editor (s): Corwin. W.R. planning for test specimen machining Soonsorino Oroanization: NRC; Nuclear for the Japan Power Development Reactor Regulatory Comission. Washington. DC was completed. A database of Charpy (United States) impact and fracture toughness data for Publication Date: Sep 1997 RPV materials that have been tested in Reoort Number (sh the unirradiated and irradiated NUREG/CR-5591-Vol.7-No.2; conditions is being assembled and ORNL/TM -11568-Vol.7 No.2 analyzed. Weld metal appears to have Order Number: T198000039 similar CVN and fracture toughness Abstract: The Heavy-Section Steel transition temperature shifts, whereas Irradiation Program was established to the fracture toughness shifts are quantitatively assess the effects of greater than CVN shifts for base neutron irradiation on the material metals. Oraft subcontractor reports on behavior of typical reactor pressure precracked cylindrical tensile vessel (RPV) steels. During this specimens were completed, reviewed. and period, fracture mechanics testing of are being revised. Testing on specimens of the irradiated low upper precracked CVN specimens, both quasi-shelf (LUS) weld were completed and static and dynamic, was evaluated.

analyses performed. Heat treatment of Additionally, testing of compact

five RPV plate materials was initiated specimens was initiated as an to examine t hosphorus segregation experimental comparison of constraint effects on tie fracture toughness of limitations. 16 figs., 2 tabs.

the heat aff?cted zone of welds.

Initial results show that all five materials exhibited very large prior

Title:

Results of Crack-Arrest Tests austenite grain sizes as a consequence on Irradiated A508 Class 3 Steel I of the initial heat treatment. Author (s)/ Editor (s): S.K. Iskander, Irradiated and annealed specimens of P.P. Milella, A. Pini LUS weld material were tested and Soonsorino Oroanization: NRC:

analyzed. Four sets of Charpy V-notch Washington DC (United States)

NUREG-1426 94

Compilation of Reports: 1994-1998 Publication Date: February 1998 data points that appeared to lie close Reoort Number (s): NUREG/CR 6447 to or lower than the American Society Abstract: Ten crack-arrest toughness of Mechanical Engineers K. curve to values for irradiated specimens of A positions that seemed more reasonable 508 class 3 forging steel have been with respect to the remaining data. A obtained. The tests were performed special fixture was designed, according to the American Society for fabricated, and successfully used in Testing and Materials (ASTM) Standard the testing. For reasons explained in Test Method for Determining the text. Special blocks to receive the Plane-Strain Crack-Arrest Fracture Oak Ridge National Laboratory clip gage Toughness. Kla, of Ferritic Steels. E were designed, and 1221-88. None of these values are greater-than-standard crack-mouth strictly " valid. in all five ASTM E opening displacements measured were 1221-88 validity criteria. However, accounted for.

they are useful when compared to unirradiated crack-arrest specimen toughness values since they show the

Title:

Heavy-Section Steel Irradiation small (averaging approximately 10*C) Program Semiannual Progress Report for shifts in the mean and lower-bound October 1996 - March 1997 crack-arrest toughness curves. This Author (s)/ Editor (s): T.M. Rosseel confirms that a low copper content in Soonsorino Orcanization: NRC:

ASTM A 508 class 3 forging material can Washington DC (United States) be expected to result in small shifts Publication Date: February 1998 of the transition toughness curve, The Reoort Number (s): NUREG/CR 5591 Vo18 shifts due to neutron irradiation of No 1/0RNL/TM 11568Vo18 No 1 the lower bound and mean toughness Abstract: Maintaining the integrity of curves are approximately the same as the reactor pressure vessel (RPV) in a the Charpy V notch (CVN) 41-J light-water-cooled nuclear power plant temperature shift. The nine is crucial in preventing and crack-arrest specimens were irradiated controlling severe accidents that have at temperatures varying from 243 to the potential for major contamination 280*C, and to a fluence varying from release. Because the RPV is the only 1.7 to 2.7 x 1019 neutrons /cm2 (> 1 key safety-related component of the MeV). The test results were plant for which a redundant backup

" normalized" to reference values that system does not exist, it is imperative correspond to those of CVN specimens to fully understand the degree of irradiated at 284*C to a fluence of 3.2 irradiation-induced degradation of the x 1019 neutrons /cm2 (> 1 MeV) in the RPV's fracture resistance that occurs same capsule as the crack-arrest during service. For this reason, the specimens. This adjustment resulted in Heavy-Section Steel Irradiation (HSSI) a shift to lower temperatures of all Program has been established. Its the data, and in particular moved two primary goal is to provide a thorough, l

95 NUREG-1426

Compilation of Reports - 1994 1998 quantitative assessment of the effects Reoort Number (s): NUREG/CR-5161-Vol.2:

of neutron irradiation on the material PNL- 6462-Vol.2 behavior and, in particular, the Order Number: TI94007939 fracture toughness properties of Abstract: This report summarizes the typical pressure-vessel steels as they results of three previous studies to relate to light-water RPV integrity, evaluate and compare the effectiveness '

Effects of specimen size: material of sampling plans for steam generator chemistry: product form and tube inspections. An analytical microstructure: irradiation fluence, evaluation and Monte Carlo sirrulation flux, temperature, and spectrum: and techniques were the methods used to postirradiation annealing are being evaluate sampling plan performance. To examined on a wide range of fracture test the performance of candidate properties. The HSSI Program is sampling plans under a variety of arranged into seven tasks: (1) program conditions, ranges of inspection system management (2) irradiation effects in reliability were considered along with engineering materials, (3) annealing, different distributions of tube (4) microstructural analysis of degradation. Results from the eddy radiation effects (5) in service current reliability studies performed

' irradiated and aged material with the retired-from-service Surry 2A evaluations. (6) fracture toughness steam generator were utilized to guide curve shift method, (7) special the selection of appropriate technical assistance, and (8) foreign probability of detection and flaw

! research interactions. The work is sizing models for use in the analysis, performed by the Oak Ridge National Different distributions of tube Laboratory, degradation were selected to span the range of conditions that might exist in l operating steam generators. The l principal means of evaluating sampling Steam Generator Tube performance was to determine the Integrity effectiveness of the sampling plan for detecting and plugging defective tubes.

A sumary of key results from the eddy

Title:

Evaluation of sampling plans current reliability studies is for in service inspection of steam presented. The analytical and Monte generator tubes Carlo simulation analyses are discussed Author (s)/ Editor (s): Kurtz, R.J. : along with a synopsis of key results Heasier, P.G. : Baird. 0.B. (Pacific and conclusions.

Northwest Lab., Richland, WA (United States))

Soonsorino Oroanization: NRC: Nuclear

Title:

Applications of a new magnetic Regulatory Comission, Washington, DC monitoring technique to in situ L (United States) evaluation of fatigue damage in ferrous Publication Date: Feb 1994 NUREG-1426 96 i

1

Compilation of Reports: 1994-1998 components critical changes in Barkhausen Author (s)/ Editor (s): J11es. D.C. . emissions, coercivity and hysteresis Biner. S.B. ; Govindaraju M.R. . Chen. loss occurred iri the last ten to twenty Z.J. (Iowa State Univ. of Science and percent of fatigue life.

Technology. Ames. IA (United States).

Center for Nondestructive Evaluation)

Soonsorino Orcanization: NRC: Nuclear

Title:

Piping inspection round robin Regulatory Comission, Washington, DC Author (s)/ Editor (s): Heasier. P.G. .

(United States) Doctor S.R. (Pacific Northwest Publication Date: Jun 1994 National Lab. Richland. WA (United Reoort Number (s)- NUREG/GR 0013 States))

Order Number: TI94015174 Soonsorino Orcanization: NRC: Nuclear Abstract- This project consisted of Regulatory Commission. Washingtro. DC research into the use of magnetic (United States) inspection methods for the estimation Publication Date: Apr 1996 of fatigue life of nuclear pressure ReDort Number (s): NUREG/CR-5068:

vessel steel. Estimating the PNNL--10475 mechanical and magnetic properties of Order Number: TI96009923 ferromagnetic materials are closely Abstract: The piping inspection round interrelated, therefore, measurements robin was conducted in 1981 at the of magnetic properties could be used to Pacific Northwest National Laboratory monitor the evolution of fatigue damage (PNNL) to quantify the capability of in specimens subjected to cyclic ultrasonics for inservice inspection loading. Results have shown that is and to address some aspects of possible to monitor the fatigue damage reliability for this type of nondestructively by magnetic nondestructive evaluation (NDE). The techniques. For example, in round robin measured the crack load-controlled high- cycle fatigue detection capabilities of seven field tests, it has been found that the inspection teams who employed plastic strain and coercivity procedures that met or exceeded the accumulate logarithmically during the 1977 edition through the 1978 addenda fatigue process. Thus a quantitative of the American Society of Mechanical relationship between coercivity and the Engineers (ASME) Section 11 Code number of fatigue cycles could be requirements. Three different types of established based on two empirical materials were employed in the study coefficients, which can be detennined (cast stainless steel, clad ferritic, from the test conditions and material and wrought stainless steel), and two properties, Also it was found that different types of flaws were implanted prediction of the onset of fatigue into the specimens (intergranular failure in steels was possible under stress corrosion cracks (IGSCCs) and certain conditions. In thermal fatigue cracks (TFCs)). When strain controlled low cycle fatigue, considering near-side inspection.

97 NUREG-1426

Compilation of Reports - 1994-1998 far-side inspection, and false call detection portion of the performance rate, the overall performance was found demonstration test is given.

to be best in clad ferritic, less effective in wrought stainless steel and the worst in cast stainless steel.

Title:

Data analysis for steam Depth sizing performance showed little generator tubing samples correlation with the true crack depths. Author (s)/ Editor (s): Dodd, C.V. 3 Soonsorino Oroanization: NRC: Nuclear Regulatory Comission. Washington, DC

Title:

Performance demonstration tests (United States) for eddy current inspection of steam Publication Date: Jul 1996 generator tubing Recort Number (s): NUREG/CR 6455 Author (s)/ Editor (s): Kurtz, R.J. Order Number: T196013809 Heasier, P.G. : Anderson. C.M. Abstract: The objective of the Soons3rino Oraanization: NRC: Nuclear Improved Eddy-Current ISI for Steam Regulatory Commission. Washington. DC Generators program is to upgrade and (United States) validate eddy-current inspections.

Publication Date: May 1996 including probes, instrumentation, and Reoort Number (s): NUREG/CR 6227: data processing techniques for PNNt.--9433 inservice inspection of new. used, and Order Number: T196011419 repaired steam generator tubes: to Abstract: This report describes the improve defect detection, methodology and results for development classification and characterization as of performance demonstration tests for affected by diameter and thickness eddy current (ET) inspection of steam variations, denting. probe wobble, tube generator tubes. Statistical test sheet, tube supports, copper and sludge design principles were used to develop deposits, even when defect types and the performance demonstration tests. Other variables occur in combination:

Thresholds on ET system inspection to transfer this advanced technology to performance were selected to ensure NRC's mobile NDE laboratory and staff.

that field inspection systems would This report provides a description of have a high probability of detecting the application of advanced and and correctly sizing tube eddy-current neural network analysis degradation. The technical basis for methods for the detection and the ET system performance thresholds is evaluation of comon steam generator presented in detail. Statistical test tubing flaws including axial and design calculations for probability of circumferential outer diameter stress-detection and flaw sizing tests are corrosion cracking and intergranular described. A recommended performance attack. The report describes the demonstration test based on the design training of the neural networks on calculations is presented. A computer tubing samples with known defects and program for grading the probability of the subsequent evaluation results for NUREG 1426 98 4

Compilation of Reports: 1994-1998 unknown samples. Evaluations were done Jitle- Evaluation and field validation in the presence of artifacts. Computer of Eddy-Current array probes for steam programs are given in the appendix. generator tube inspection Author (s)/ Editor (s): Dodd, C.V. .

Pate. J.R.

Title:

Computer programs for the sconsorino Oraanization: NRC: Nuclear acquisition and analysis of Regulatory Comission. Washington. DC eddy current array probe data (United States)

Author (s)/ Editor (s)- Pate, J.R. . Publication Date: Jul 1996 Dodd, C.V. Reoort Number (s): NUREG/CR 6357 Soonsorino Orcanization: NRC: Nuclear QrderNu-ber.;. TI96013374 Regulatory Comission. Washington, DC Abstract: The objective of the (United States) Improved Eddy-Current ISI for Steam Publication Date: Jul 1996 Generator Tubing program is to upgrade ReDort Nurber(s): NUREG/CR-6163: and validate eddy-current inspections.

ORNL/TM--13212 including probes, instrumentation, and Order Number: TI96013016 data processing techniques for Abstract: Objective of the Improved inservice inspection of new used, and Eddy-Curent ISI (in-service inspection) repaired steam generator tubes: to for Steam Generators Tubing program is improve defect detection, to upgrade and validate eddy-current classification, and characterization as inspections, including probes, affected by diameter and thickness instrumentation, and data processing variations, denting probe wobble. tube techniques for ISI of new. used, and sheet. tube supports, copper and sludge repaired steam generator tubes: to deposits even when defect types and improve defect detection, other variables occur in combination:

classification and characterization as to transfer this advanced technology to affected by diameter and thickness NRC's mobile NDE laboratory and staff.

Variations, denting, probe wobble, tube This report describes the design of sheet tube supports, copper and sludge specialized high speed 16-coil deposits, even when defect types and eddy-current array probes. Both other variables occur in combination: pancake and reflection coils are to transfer this advanced technology to considered. Test results from NRC's mobile NDE laboratory and staff. inspections using the probes in working This report documents computer programs steam generators are given. Computer that were developed for acquisition of programs developed for probe eddy-current data from specially calculations are also supplied.

designed 16 coil array probes.

Complete code as well as instructions for use are provided.

Title:

Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants 99 NUREG-1426 l

l' ~

Compilation of Reports - 1994-1998 Author (s)/ Editor (s): Otercks D.R. rupture. The model has now been (Argonne National Lab., il (United implemented in a compute code, called States)) STAC (Statistical Analysis of Cracks).

Publication Date: Feb 1997 This paper is aimed to give a brief Reoort Number (s): NUREG/CP-0154: outline of the model to facilitate the ANL--96/14: NEA/CNRA/R--(96)1: understanding of the possibilities and CONF 9510423 - limitations associated with the model.

Order Number: Tl97004312 Abstract: Steam generator (SG) tubes in pressurized water reactor plants are

Title:

Steam generator tube integrity exposed to various types of degradation program. Semiannual report. August processes, among which stress corrosion 1995-- March 1996 cracking in particular has been Author (s)/ Editor (s): Oiercks, D.R. ;

observed. To be able to evaluate the Bakhtiari, S. : C',opra.

. 0.K. (and safety importance of such cracking of others)

SG-tubes one has to have a good and Soonsorina Orcalization: NRC: Nuclear empirically founded knowledge about the Regulatory Commission, Washington, DC scope and the size of the cracks as (United States) well as the rate of their continuous Publication Date: Apr 1997 growth. The basis of experience is to Reoort Number (s): NUREG/CR-6511 Vol.1:

a large extent constituted of the ANL- 96/17 annually performed SG-inspections and Order Number: Tl97005975 crack sizing procedures. On the basis Abstract: This report sumarizes work of this experience one can estimate the performed by Argonne National distribution of existing crack lengths. Laboratory on the Steam Generator Tube and modify this distribution with Integrity Program from the inception of regard to maintenance (plugging) and that program in August 1995 through the predicted rate of crack March 1996. The program is divided propagation. Finally, one can into five tasks, namely (1) Asses ment  ;

calculate the rupture probability of of Inspection Reliability (2) Resurch SG tubes as a function of a given on ISI (in service- inspection) critical crack length. On account of Technology (3) Research on Degradation the Swedish Nuclear Power Inspectorate Modes and Integrity, (4) Development of an introductory study has been Methodology and Technical Requirements performed in order to get a survey of for Current and Emerging Regulatory what has been done elsewhere in this issues, and (5) Program Management.

field. The study resulted in a Under Task 1, progress is reported on proposal of a computerizable model to the preparation of and evaluation of be able to estimate the distribution of nondestructive evaluation (NDE) true cracks, to modify this techniques for inspecting a mock up distribution due to the crack growth steam generator for round robin and to compute the probability of tube testing, the development of better ways NUREG-1426 100

Compilation of Reports: 1994-1998 to correlate burst pressure and leak (United States) rate with eddy current (EC) signals, Publication Date: February 1993 the inspection of sleeved tubes, Repgrt Number (s): NUREG/CR-6511 Vol2, workshop and training activities, and ANL-97/3 the evaluation of emerging NDE ,Ajtq1r_a 1 , This report summarizes work technology. Under Task 2 results are performed by Argonne National reported on closed-form solutions and Laboratory on the Steam Generator Tube finite element electromagnetic modeling Integrity Program from the inception of of EC probe response for various probe the program in August 1995 through designs and flaw characteristics. September 1996. The program is divided Under Task 3 facilities are being into flve tasks: ( 1) Assessment of designed and built for the production Inspection Reliability, (2) Research on of cracked tubes under aggressive and ISI (in-service-inspection) Technology, near-prototypical conditions and for (3) Research on Degradation Modes and the testing of flawed and unflawed Integrity (4) Tube Removals from Steam tubes under normal operating, accident, Generators, and (5) Program Management.

and severe accident conditions. In Under Task 1, progress is reported on addition, crack behavior and stability the preparation of facilities and are being modeled to provide guidance evaluation of nondestructive evaluation on test facility design, to develop an techniques for inspecting a mock-up improved understanding of the expected steam generator for round-robin rupture behavior of tubes with testing. the development of better circumferential cracks, and to predict correlations between failure pressure the behavior of flawed and unflawed and leak rate with eddy current (EC) tubes under severe accident conditions. signals, the inspection of sleeved Task 4 is concerned with the cracking tubes, workshop and training and failure of tubes that have been activities, and the evaluation of repaired by sleeving, and with a review emerging NDE technology, Under Task 2, of literature on this subject. results are reported on closed-form solutions and finite-element electromagnetic modeling of EC probe

Title:

Steam Generator Tube Integrity responses for various probe designs and Program: flaw characteristics. Under Task 3.

Annual Report August 1995-September facilities are being designed and built 1996 for the production of cracked tubes Author (s)/ Editor (s): , R. Diercks, S. under aggressive and near-prototypical Bakhtiari, K. E Kasza, D. S. conditions and for the testing of Kupperman, S. Ma,lumdar, J. Y. Park, flawed and unflawed tubes under normal and W. J. Shack (Argonne National operating, accident, and severe Laboratory (United States)) accident conditions. In addition, crack Soonsorino Oroanization: NRC: Nuclear behavior and stability are being Regulatory Commission, Washington DC modeled to nrovide guidance for test 101 NUREG 1426

e Compilation of Reports - 1994-1998 facility design, develop an improved The failure temperatures predicted by understanding of the expected failure the model for two temperature and behavior of tubes with circumferential pressure histories, calculated for cracks, and predict the behavior of severe accidents initiated by a station flawed and unflawed tubes under severe blackout, agrec .ery well with tests accident conditions. Task 4 is performed on both flawed and unflawed concerned with the acquisition of tubes specimens.

, and tube sections from retired steam l generators for use in the other research tasks. Progress on the Title The Role of Time-Dependent acquisition of tubes from the Salem and Deformation in Intergranular Crack McGuire 1 nuclear plants is reported. Initiation of Alloy 600 Steam Generator Tubing Material

. Author (s)/ Editor (s): G.S. Was, K.

Title:

Fai?ure Behavior of Internally Lian Pressurized Flawed and Unflawed Steam $ponsorino Oroanization: NRC:

Generator Tubing at High Temperatures - Washington DC (United States)

Experiments anti Comparison with Model Publication Date: March 1998 Predictions Reoort Number ( 1;. NUREG/GR-0016

. Author (s)/Editer(s): S. Majumdar. Abstract: Intergranular stress W.J. Shack, D.R. Diercks. K. Mruk, J. corrosion cracking (IGSCC) of two i Franklin, and L. Knoblich commercial alloy 600 conditions (600LT,

, Soonsorino Oroanization: NRC: 600HT) and controlled purity Washington DC (United States) Ni-18Cr-9Pe alloys (CDMA, CDTT) were Publication Date: March 1998 investigated using constant extension Reoort Number (s): liUREG/CR-6575 rate tensile (CERT) tests in primary Abstraq1: This report summarizes water (0.01M LiOH + 0.01M H3803) with 1 experimental work performed at Argonne bar hydrogen overpressure at 360 C and l

NP.ional Laboratory on the failure of 320*C. Heat treatments produced two internally pressurized steam generator types of microstructures in both tubing at high temperatures (-700'C). A commcrcial and controlled-purity model was developed for predicting alloys: one dominated by grain boundary failure of flawed and unflawed steam carbides (600HT and CDTT) and one generator tubes under internal pressure dominated by intragranular carbides and temperature histories postulated te (600LT and CDMA). CERT tests were occur during severe accidents. The conducted over a range of strain rates model was validated by failure tests on and at two t aperatures with -

specimens with part-through-wall axial interruptions at specific strains to and circumferential flaws of various determine the crack depth leng'.t and depths, conducted under distributions. Results show that in all

! various constant and ramped internal samples. IG3CC was the dominant failure pressure and temperature conditions. mode. For both the commercial alloy and i

i NUREG-1426 102  ;

l l l l

l i a

Compilation of Reports: 1994-1998 the controlled-purity alloys, the characterization of thermally aged cast microstructure with grain boundary stainless steels carbides showed delayed crack Author (s)/ Editor (s): Michaud. W.F. .

initiation and shallower crack depths Toben. P.T. . Soppet. W.K. . Chopra.

than did the intragranular carbide 0.K. (Argonne National Lab., IL (United microstructure under all experimental Sta%s))

, c W itions. This data indicates that a Soonsorino Oroanization: NRC: Nuclear grain boundary carbide microstructure Regulatory Comission. Washington OC is more resistant to IGSCC than an (United States) intragranular carbide microstructure. Publication Date: Feb 1994 Observations support both the f-lm Reoort Number (s): NURE3/CR-6142; rupture / slip dissolution mechanism and ANL--93/35 enhanced localized plasticity. The Order Number: TI94007118 advantage of these results over Abstract: The effect of thermal aging l previous studies is that the different on tensile properties of cast stainless l carbide distributions were obtained in steels during service in light water I the same cormercial alloy using reactors has been evaluated. Tensile different heat treatments, and in the data for several experimental and j l other case in nearly 16entical commercial heats of cast stainless I l controlled-purity alloys. Therefore. steels are presented. Thermal aging observations of the effects of carbide increases the tensile strength of these distribution on IGSCC can more confi- steels. The high C Mo-bearing CF-8M dently be attributed to the carbide steel, are more susceptible to thermal distribution alone rather than other aging than the Mo- free CF-3 or CF-8 potentially signif1 cant differences in steels. A procedure and correlations microstructure or composition. Crack are presented for predicting the change growth rates increased with increasing in tensile flow and yield stresses and strain rate according to a power law engineering stress- vs.-strain curve

-rclation with a strain rate exponent of cast stainless steel as a function ,

between 0.4 and 0.64. However, crack of time and temperature of service.

growth rate measured in m/ unit strain The tensile properties of aged cast decreased with increasing strain rate stainless .eteel are estimated from indicating an effect of either the known material information, i .e. .

environment or creep. The. temperature chemical composition and the initial dependence of the crack growth ratc was tensile strength of the steel. The consistent with the literature. correlations described v this report may be used for assessing thermal l embrittlemen'. of cast stainless steel Therma'l Aging components.

l l

Title:

Tensile-property

Title:

Estimation of fracture I 103 NUREG-1426 i

1 i

i i

Compilation of Reports - 1994-1998 toughness of cast stainless steels chemical composition. The initial during thermal aging in LWR impact energy of the unaged steel is systems-revision 1 required for these estimations.

Author (s)/ Editor (s): Chopra. 0.K. Initial tensile flow stress is needed Soonsorino Orcanization: NRC; Nuclear for estimating the flow stress of the Regulatory Comission. Washington, DC aged material . The fracture toughness (United States) J-R curve for the material is then Publication Date: Aug 1994 obtained by correlating room-Reoort Number (s): NUREG/CR-4513-Rev.1; temperature Charpy-impact energy with ANL--93/22-Rev.1 fracture toughness parameters. The

,0rder Number: TI94018628 values of J[sub IC] are determined from Abstract: This report presents a the estimated J-R curve and flow revision of the procedure and stress. A common [open correlations presented earlier in quotes] predicted lower-bound [close NUREG/CR-4513. ANL-90/42 (June 1991) quotes] J-R curve for cast stainless for predicting the change in mechanical steels of unknown chemical composition properties of cast stainless steel is also defined for a given grade of components due to thermal aging during steel, range of ferrite content, and '

service in light water reactors at temperature. Examples of estimating 280-330[ degrees]C (535- mechanical properties of cast stainless 625[ degrees]F). The correlations steel components during reactor service presented in this report are based on are presented.

an expanded data base and have been optimized with mechanical-property data on cast stainless steels aged up to

Title:

Mechanical properties of l [ approx]58,000 h at 290-350(degrees]C thermally aged cast stainless steels (554- 633[ degrees]F). The fracture from Shippingport reactor components toughness J-R curve, tensile stress, Author (s)/ Editor (s): Chopra, 0.K. :

and Charpy- impact energy of aged Shack W.J. (Argonne National Lab., Il cast stainless steels are estimated (United States))

from known material information. Soonsorino Oroanization: NRC: Nuclear Mechanical properties of a specific Regulatory Comission, Washington DC

! cast stainless steel are estimated from (United States) the extent and kinetics of thermal Publication Date: Apr 1995 embrittlement. Embrittlement of cast Reoort Number (s): NUREG/CR-6275:

stainless steels is characterized in ANL--94/37 terms of room-temperature Charpy- Order Num,bstri TI95012084 impact energy. Charpy-impact energy as Abstract: Thermal embrittlement of a function of time and temperature of static-cast CF-8 stainless steel reactor service is estimated from the components from the decomissioned i

I kinetics of thermal embrittlement. Shippingport reactor has been which are also determined from the characterized. Cast stainless steel I

NUREG-1426 104 i ,

l Compilation of Reports: 1994-1998 materials were obtained from four 15 y and the KRB reactor pump cover I cold-leg check valves, three hot-leg plate (CF-8) after [approximately] 8 y main shutoff valves, and two pump of service.

volutes. The actual time-at-temperature for the materials was

[approximately]13 y at

Title:

Effects of thermal aging on

[approximately]281 C (538 F) for the fracture toughness and Charpy-impact hot-leg components and strength of stainless steel pipe welds

[approximately]264 C (507 F) for the Author (s)/ Editor (s): Gavenda. D.J. .

cold- leg components. Baseline Michaud, W.F. : Galvin, T.M. : Burke, mechanical properties for as-cast W.F. ; Chopra, 0 K.- (Argonne National material were determined from tests on Lab., IL (United States))

either recovery-annealed material, Soonsorino Oroanization: NRC: Nuclear i.e., annealed for 1 h at 550 C and Regulatory Commission, Washington, DC then water quenched, or material from (United States) the cooler region of the component. Publication Date- May 1996 The Shippingport materials show modest Reoort Number (s): NUREG/CR-6428; decreases in fracture toughness and ANL--95/47 Charpy-impact properties and a small Order Number: TI96011018 increase in tensile strength because of Abstract: Degradation of fracture relatively low service temperatures and toughness. tensile, and Charpy-impact ferrite content of the steel. The properties of Type 304 and 304/308 SS procedure and correlations developed at pipe welds due to thermal aging was Argonne National Laboratory for studied at room temperature and 290 C.

estimating mechanical properties of Thermal aging of SS welds results in cast stainless steels predict accurate moderate decreases in charpy-impact l cr slightly lower values for strength and fracture toughness.

l Charpy-impact energy, tensile flow Upper-shelf energy decreased by 50-80 stress, fracture toughness J-R curve, J/cm[sup 2]. Decrease in fracture and J[sub IC] of the materials. The toughness J-R curve or J[sub IC) is kinetics of thermal embrittlement and relatively small. Thermal aging had no degree of embrittlement at saturation, or little effect on tensile strength of 1.e., the minimum impact energy the welds. Fracture properties of SS achieved after long-term aging, were welds are controlled by the established from materials that were distribution and morphology of

' aged further in the laboratory. The second-phase particles. Failure occurs results were consistent with the by formation and growth of microvoids estimates, The correlations near hard inclusions; such processes successfully predicted the mechanical are relatively insensitive to thermal properties of the Ringhals 2 reactor aging. The ferrite phase has little or

hot and crossover-leg elbows (CF-8M no effect on fracture properties of the j steel) after service of [approximately] welds. Differences in fracture 1

l l 105 NUREG-1426 1

i e

r -, ,- , , , , . - -,,c-_-,, - . , , ., , , , - - - , . . . ~ , , - - - - - - , . . . . - .

i l

Compilation of Reports - 1994-1998

. resistance of the welds arise from the initiation fracture toughness (J,.)

[ differences in the density and . size of much more man did thermal aging alone, inclusions. Mechanical-property data Irradiation slightly decreased the I

from the present study are consistent tearing modulus, but no reduction was with results from other investigations, caused by themal aging alone. Other The existing data have been used to results from tens 11e. CVN. and fracture 4,

establish minimum expected fracture toughness specimens showed that the  !

! properties for SS welds. effects of thermal aging at 288 or 3 r

343'C for 20.000 h each were very small l and similar to those at 288*C for 1605 l

Title:

Effects of Thermal Aging and h. The effects of long-term thermal j Neutron irradiation on the Mechanical exposure time (50.000 h and greater) at 1 Properties of Three-Wire Stainless 288 C will be investigated as the l Steel Weld Overlay Cladding specimens become available in 1996 and Author (s)/ Editor (s): F.M. Haggag. R.K. beyond. i

. Nanstad )

i Soonsorino Orcanization: NRC Washington DC (United States)

Title:

Influence of Long-Term Thermal 1 i Publication Date: May 1997 Aging on the Microstructural Evolution Reoort Number (s): NUREG/CR- of Nuclear Reactor Pressure Vessel 6363/0RNL/TM-13047 Materials Abstract Thermal aging of three-wire Author (s)/ Editor (s): P. Pareige, K.F.

' series-arc stainless steel weld overlay Russell, R.E. Stoller. M.K. Miller (Oak

cladding at 288*C for 1605 h resulted Ridge National Laboratory) in an appreciable decrease (16%) in the Soonsorino Oroanization

NRC: '

Charny V-notch (CVN) upper-shelf energy Washington DC (United States)

(USE), but the effect'on the 41- Publication Date: March 1998 transition temperature shift was very Reoort Number (s): NUREG/CR-small (3*C). The combined effect of 6537:0RNL/TM-13406 aging and neutron irradiation at 288'C Abstract: Atom probe field ion to a fluence of 5 x 10-9 neutrons /cm2 microscopy (APFIM) investigations of  ;

-(> 1 Me\/) was a 22% reduction in the the microstructure of unaged (as- l USE and a 29'C'shif t in the 41- fabricated) and long-term thermally transition temperature. The effect of aged (-100.000 h at 280*C) surveillance thermal aging on tensile properties was materials from commercial reactor i very small. However, the combined pressure vessel steels were performed.

effect of irradiation and aging was an This combination of materials and increase in the yield strength (6 to conditions permitted the investigation

34% at test temperatures from 288 to of potential thermal-aging effects.

-125*C)-but no apparent change in This microstructural study focused on ultimate tensile strength or total the quantification of the compositions elongation. Neutron irradiation reduced of the matrix and carbides. The APFIM NUREG-1426 106 a

Compilation of Reports: 1994-1998 results indicate that there was no are subjected to high neutron fluences.  ;

significant microstructural evolution This assessment was requested because after a 1cag-term thermal exposure in of the recent increased level of  ;

weld.' plate. or forging materials. The activity in the commercial nuclear  ;

matrix depletion of copper that was industry to address generic issues  !

observed in weld materials was concerning the reactor vessel and l consistent with the copper internals, especially those issues )

concentration in the matrix after the related to repair options. This '

i stress-re!1ef heat treatment. The literature review revealed a l compositions of cementite carbides aged preponderance of general information ,

for 100.000 h were compared with the about underwater welding technology, as Thermocalc" prediction. The APFIM a result of the active research in

' comparisons of materials under these this field sponsored by the U. S. Navy conditions are consistent with the and offshore oil and gas industry l measured change in mechanical concerns. However, the literature )

properties such as the Charpy search yielded only a limited amount of transition temperature. information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no Underwater Welding information at all concerning underwater welding experiences in high-fluence environments.

Title:

Underwater Welding of Highly Irradiated Boiling Water Reactor In- Research reported by the staff of the Vessel Components U. S. Department of Energy (DOE)

Author (s)/ Editor (sh L. Lund Savannah River Site and researchers Soonsorino Oraanization: NRC; from the 00E fusion reactor program Washington DC (United States) proved more fruitful. This research documented relevant experience Publication Date: November 1997 Recort Number (sh ' NUREG-1616 concerning welding of stainless steel Abstract: In February 1997, the U. S. materials in air environments exposed Nuclear ' Regulatory Comission (NRC). to high neutron fluences. It also Office of Nuclear Regulatory Research addressed problems with welding highly (RES), initiated a literature review to irradiated materials, and primarily assess the state of underwater welding attributed those problems to

, technology. In particular, the helium-induced cracking in the objective of this literature review was material . (Helium is produced from the

! neutron irradiation of boron, an l to evaluate the viability of underwater welding in vessel components of boiling impurity, and nickel.) The researchers i water reactor (BWR) in-vessel found that the amount of helium-induced components, especially those components cracking could be controlled. or even

}

eliminated, by reducing the heat input j fabricated from stainless steels that I.

107 NUREG-1426 I

l t

l

Compilation of Reports - 1994-1998 into the weld and applying a compressive stress perpendicular to the weld path, l

l i

i 1

l NURPG-1426 108

_ -.. _m _ _ _ _ .m _ . ~ _ _ . _ ..m.

I NRC FCRM s3s ua NUCLEAR REGULATORY COMM,ss40N 1. REPORT NUheER QM (Aemgaed by Mtc. Add Vel, Supp., Rev EE BIBLIOGRAPHIC DATA SHEET **d**"*""""*""

i NUREG-1426 l 2.TtTLE ANo suet TLE Vol. 3 Compilation of Reports From Research Supported by the Electrical, Materials, and Mechanical Er.gineering Branch, Division of Engineering 3. DATE RFPORT PUBUSHED uoerrs YEAR l

October 1998

4. FIN OR GRANT NUheER

5. AUTHOR (S) 6. TYPE OF REPORT C. G. Santos, Jr.

Technical

7. PERCO COVERED (hcA, eve cens) l s. PERFORhENG ORGANIZAflON . NAAE AND ADORESS (irNRC, povsfe orvasm of' ice a Aegm u & wcmar Megeery Carmum end meeng ***ess; recreech.

pove nere one menne estou)

Division of Engineering Offic) of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission j Washington, DC 20555-0001

9. SPONSORING ORGANIZATON . NAME AND ACORESS (FNRC, ene 'Some as abovet tcatech, pave NRc ovem carce or Region, u & weiser Aegetry carmum end moung e6sess) l Same as above.

l

10. SUPPLEAENTARY NOTES l

l l 11. A8sTRACT(200 nove or ases)

Since 1965, the Materials Engineering Branch, Division of Engineering, of the NvJear Regulatory Commission's Office of Nuclear Regulatory Research, and its predecessors dating back to the Atomic Energy Commission (AEC), have sponsored research programs conceming the integnty of the primary system pressure boundary of kght-water reactors. The components of concem in these research programs have included the reactor pressure vessel (RPV), steam generators, and the piping. These research programs have covered a broad range of topics, induding fracture mechanics analysis and experimental work for RPV and piping apphcations, inspection method development and qualification, and evaluation of irradiation effects on RPV steels.

Thb report provides as complets; a listing as practical of formal technical reports submitted to the NRC by the investigators working on these research programs. Th,s listing indades topical, final, and progress reports and is divided by topic area. In many cases, a report will cover several topics (such as in the case Of progress reports of multi-faceted programs) but is Ested under only one topic.

Therefore, in searching for reports on a specific trec, other related topic areas should be checked also. The separate volumes of this report cover the fotowing periods:

Volume 1: 1965-1990 Volume 2: 1991 - 1993 Volume 3: 1994 1998

, 12. KEY WORDS/DESCRWrTORS (Usf wads orpvmee me war samtnie arware m eseang Niepay n AvAucunatArwar unErnned reactor pressure vesseis, piping, fracture mechanics, non-destructive M SECURrWCLASS#lCAM examination, radiation embrittlement, dosimetry, environmentally assisted cracking, fatigue, steam generators, annealing, research reports, f** F***)

degradation of mechanical components, thermal aging, electrical systems, unclassified

. pressure vessel steeis, underwater welding, advanced reactors (rn. ne,ao unclassified

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