ML15223A947
| ML15223A947 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 02/28/2003 |
| From: | NRC/OGC |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| RAS 28151, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR | |
| Download: ML15223A947 (11) | |
Text
NRC000172 Submitted: August 10, 2015 Pf21 Materials Reliability Program:
Second International Conference on Fatigue of Reactor Components July 29-August 1, 2002 Snowbird Ski and Conference Center Snowbird, Utah Techn *cal Report
Materials Reliability Program:
Second International Conference o Fatigue of Reactor Components (MRP-84)
July 29-August 1, 2002 Snowbird Ski and Conference Center Snowbird, Utah 1003536 Also Referenced as OECD/NEA/CSNl/R (2003) 2 Proceedings, February 2003 Cosponsors Organisation for Economic Co-Operation and Development (0 CD)
Nuclear Energy Agency/Committee on the Safety of Nuclear Installations (NEA/CSNI)
U.S. Nuclear Regulatory Commission EPRI Project Manager S. Rosinski EPRI
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REPORT
SUMMARY
This proceedings contains information presented at the Second International Con~ ence on Fatigue of Reactor Components held 31 J uly-1 August, 2002, in Snowbird, Utah. his secon conference, again sponsored by EPRI, the Organisation for Economic Co-Operatio and Development (OECD ), Nuclear Energy Agency/Committee on the Safety of Nucl ar Installations (NEA/CSNI), and the U.S. Nuclear Regulatory Commission (NRC), rovided a forum for the technical discussion of fatigue issues that affect the integrity and ope ation of Ii ht water reactor components.
Background
Fatigue is a primary degradation mechanism affecting nuclear power plant compo ents worldwide. The effective management of fatigue is important to the continued safe operation f plant components during present operation and as plants consider long-term operat on. In 20 0, the EPRI Materials Reliability Program (MRP) organized the first conference in th s series to bring together international experts to discuss significant fatigue issues affecting n clear plan operations, share common experiences, and identify outstanding technical issues. he breadt of international efforts underway in this area dictated that future conferences be held report research results and discuss relevant issues. The second conference in this series, a ain organ zed by the EPRI MRP, was held 29-31 July in Snowbird, Utah. Additionally, a one-da workshop on flaw growth in austenitic and nickel-based materials was held on I August. Cospo sorship of the conference was provided by the OECD NENCSNI, and the U.S. NRC.
Objectives
- To provide a forum for the technical discussion of fatigue issues that affect the *ntegrity a d operation of light water reactor components
- To share common experiences regarding fatigue of reactor components in orde to ensure continued safe operation
- To identify common areas of interest to foster future international research/coll boration activities Approach The conference was organized in a series of technical presentation and group discu sion sessi©ns focused on the following fatigue-related topics:
- Environmental fatigue
- Fatigue monitoring/evaluation v
- Codes and standards
- General fatigue issues The conference was structured to directly benefit utility and plant managers, as we l as syste ,
materials, structural integrity, licensing, and maintenance/repair engineers. The dis ussion/pa el sessions were structured to foster open discussion among participants in relevant f tigue issu Results Approximately 60 fatigue experts, representing 8 countries, participated in the sec conference. Strong representation was made by nuclear operators, vendors, regular ry agenci1s, research and development organizations, and other experts. Following the technica presentations, a summary panel session was held to discuss key technical issues id ntified du ing the conference. As a result of these discussions, conclusions were developed with t e consen us 1
of the conference participants. Based on the degree of technical exchange that occ ed and t e quality of information provided during the conference, the participants recommend d that another conference on this topic be held in 18-24 months.
EPRI Perspective Fatigue management is an important aspect of the continued safe operation of plan compone ts.
Periodic discussion of fatigue-related issues in an international forum allows the s aring of common experiences and fosters international collaboration in the resolution of fat gue issues Future conferences in this series will continue to be a major forum for the discussi n of plant component fatigue issues.
Keywords Thermal fatigue Environmental fatigue Fatigue management Life-cycle management VI
CONFERENCE
SUMMARY
2"d International Conference on Fatigue of Reactor Components 29-31 July, 2002 Snowbird Ski & Conference Center Snowbird, Utah Following the technical presentations, a panel session was held to discuss key tech ical issue ~
identified during the conference. As a result of these discussions, the following co cl us ions j ere
- ~::ldF:::~:nsensus of the conference participants.
1
- 1. Stronger U.S. utility participation in OECD and NEA/CSNI is encouraged reg ding thermal fatigue technical issues.
- 2. Similar efforts are underway by several organizations worldwide to understand the fundamental mechanisms associated with thermal fatigue and predict compone t susceptibility. Data sharing and collaboration are encouraged for the benefit of II organizations.
- 3. International knowledge regarding the phenomena associated with thermal fati ue is progressing. Results of these studies should be incorporated into aging manage ent programs, including in-service inspection (ISi) programs.
- 4. Consideration should be given to the assessment and screening of Class 2 pi pin systems for thermal fatigue.
- 5. International data efforts indicate that a change in the high-cycle end of the the mal fatigu l
mean data curve in the appropriate design code may be warranted. Generation f additional 6
data beyond I 0 cycles is recommended. This may also warrant a revision to fa igue evaluation procedures.
- 6. Fatigue usage factor is not necessarily a good indicator of component degradati n.
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Environmental Fatigue I. The effect of flow rate on environmental fatigue for carbon/low-alloy steels ha been shown by several organizations. Additional data are needed to characterize the effect r stainless steel materials.
- 2. The effect of flow rate should be considered in environmental fatigue evaluati s.
- 3. International studies indicate that threshold conditions are necessary for enviro mental fatigue to occur.
- 4. Development of a new fatigue design (S-N) curve that incorporates environme ta! effects is not recommended.
- 5. For consideration of environmental effects, the present preferred approach is a application of an environmental factor, Fen*
- 6. Clarification of applicable environmental fatigue threshold parameters is neede , especially when the notion of "moderate" environmental effects is considered.
- 7. Fatigue analysis procedures (including design curves) should not be revised wi hout a thorough understanding of all relevant effects:
- a. Applicability of load-controlled data was questioned; strain-controlled data are preferred.
- b. Clarification of surface finish effects.
- 8. The measurement and reporting of water conductivity and electro-chemical pot ntial (ECP) associated with environmental fatigue tests in boiling water reactor (BWR) env*ronments are encouraged.
- 9. Further reconciliation between operating experience and laboratory/componen structural data is recommended .
I 0. A more detailed evaluation of temperature/strain relationship in transient analy is (for example, the modified rate approach applied in Japan) should be considered fo potential application.
Fatigue Monitoring/Evaluation I. Fatigue transient monitoring (both globally and locally) is an important tool fo fatigue aging management that should be implemented as early in plant life as practical.
- 2. Evaluation and assessment of data integrity are critical factors in the successful interpretation of fatigue transient monitoring results.
- 3. Advanced methods for material condition monitoring are being developed and how promise for the successful monitoring of fatigue. Further development is encouraged.
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- 4. An overall integrated approach is critical to successful fatigue management of elevant structures. Training of plant personnel is an important aspect of any integrated pproach.
- 5. Additional discussion of fatigue evaluation of welds is recommended in future onferences.
Codes & Standards (American Society of Mechanical Engineers [ASME] Sect on XI)
I . Improved in-service inspection (ISI) probability of detection (POD) is an impo ant aspect in reducing component inspection frequency in flaw tolerance analyses.
- 2. The crack aspect ratio of propagating flaw s has been shown through analytical tudies to vary as a function of transient.
- 3. Multiple crack initiations may also need to be considered in a flaw tolerance e aluation.
- 4. da/dN information is needed for austenitic stainless steels.
x
28 FATIGUE EVALUATION OF A BWR FEEDWATE NOZZLE USING AN ON-LINE FATIGUE MONIT RING SYSTEM D. Pando Universidad de Cantabria, Spain I. Gorrochategui Nuclenor, S. A., Spain G. Stevens Structural Integrity Associates, USA 28-1
FATIGUE EVALUATION OF A BWR FEEDWATER NOZZLE USING AN ON-LINE FA TIGUE MONITORING SYSTEM Dalinda Pando (/), li'iaki Gorrochategui (2) and Gary L. Stevens (3).
(I) Departamento de Ciencia e Jn genieria de/ Terreno y de las Materiales, Un versidad de Cantabria, Santander, Spain.
(2) Nuclenor, S.A., Santander, Spain.
(3) Structural Integrity Associates, Denver, CO, USA.
Abstract In 1994, the safe end of the feedwater nozzle of the Santa Marfa de Garofia Nuclear Power Plant reactor pressure vessel was replaced. Si this is the most critical location of the vessel with respect to the fatigue, it was decided to monitor the fatigue usage of this nozzle thro the installation of a fatigue monitoring system.
Thi s paper summarizes the fatigue usage results obtained from the fati ue monitoring system, from its original installation up to present-day, an a comparison of the results to the design basis fatigue estimates. First a comparison is made between the calculations included in the stress rep rt of the nozzle and the calculations performed by the fatigue monitoring system in order to demonstrate the conservative methodology emplo ed in the fatigue monitoring system. Then, the comparison is extended to actual plant transients. The conclusion is reached that the design ba is fatigue evaluation of the vessel includes a high degree of conservati m compared to actual fatigue usage accumulation.
1.- INTRODUCTION The Santa Marfa de Garofia Nuclear Power Plant (SMG) is a third-generaf on Boiling Water Reactor (BWR-3) designed by General Electric (GE) in the late 1960s The plant started commercial operation in 1971 , its rated thermal power is 1,38 1 MWt, a d its rated electrical power 460 MWe.
One of the most critical components of the plant is the Reactor Pressure Vessel RPV). The SMG RPV is a 188" diameter by 726" height cylinder with a hemispherical b ttom head and a removable hemispherical top head. The vessel was fabricated by the Rotterdam Dockyard Company (ROM) from low alloy steel (LAS) material. The design p essure and temperature are 1,250 psi and 575°F, respectively. Rated normal full powe operating pressure is 1,000 psi with an associated operating temperature of 545°F.
The systems of the plant that need to interchange fluid with the RPV, eithe in normal operation or for safety purposes, are connected to the RPV through nozzles t different 28-2
per-event fatigue usage is multiplied by the 2,600 occurrences specified in the d sign basis, more than 95 % of the total fatigue usage (U = 2.037) is obtained. The main re son for the low number of occurrences for this event is because the operators at SMG have been trained to avoid its occurrence.
6.- CONCLUSIONS The higher estimation of the fatigue usage calculated for design basis transients with FPro, compared to the fatigue usage reported in the design basis stress report for th monitored safe end location, demonstrates that the methodology used by the fatigue system installed for the SMG RPV FDW nozzles is conservative. The main justify this conservatism are: (i) the method of calculating and combining th individual stresses at the location of interest, (ii) the conservative grouping or "binning" f the stress cycles into discrete ranges, and (iii) the application of a conservative reduction ASME fatigue curve.
The fatigue usage computed by FPro for actual transients during a period of 7 75 years is significantly lower than the fatigue usage calculated for the equivalent d sign basis 1
transients and than the allowable value of 1.0 specified in the ASME Code. T o reasons that contribute to this reduction are: (i) the sequence of occurrence of the tran ients is not the "worst case" sequence typically postulated in traditional design basis eval ation, and (ii) the observed transients, although very similar in shape to the design ones, ar in general significantly less severe than the design basis definitions specified in the DS.
The most important factor that justifies the installation of the SMG fatigue monitoring system is the resulting projected number of system cycles. The number of cycl s projected based on FPro evaluation is in all cases significantly lower than the numb of cycles specified in the DS. This difference is especially significant for event #16, hich is the primary contributor to the total design basis fatigue usage of the monitored com 7.- REFERENCES
[I] "Nuclenor Reactor Pressure Vessel - Analysis of Nozzles", RDM-IGE 1104, Rev.
0, The Rotterdam Dockyard Co., 1968.
[2] "Technical Description Manual for the General Electric Fatigue onitoring System", General Electric Nuclear Energy, GE-NE-523-47-0592, Rev . 0, 1992.
[3] "Detailed Analysis of Feedwater Nozzle, Design Report, Equipos Nuc eares S.A.,
AR-4401 , Rev. 0, 1994.
[4] "Reactor Vessel - Feedwater Nozzle", Certified Design Specificatio , General Electric Nuclear Energy, 25A548 I, Rev. 0, 1993.
[5] ASME Boiler and Pressure Vessel Code, Section ill, 1986 Edition.
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