ML20246G032
| ML20246G032 | |
| Person / Time | |
|---|---|
| Issue date: | 04/10/1989 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2630, NUDOCS 8905150173 | |
| Download: ML20246G032 (15) | |
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SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE 1
MEETING ON MATERIALS AND METALLURGY m.
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March 15-16, 1989 1
Columbus, Ohio 1
1 The ACR3 Subcommittee on Materials and Metallurgy met on March 15-16, 1989 at Battelle Columbus in Columbus, Ohio to review, at the request of G. Arlotto, Director, Div. of Engineering, RES, the degraded piping program conducted by Battelle Laboratory in Columbus, Ohio. Other items reviewed at this meeting were the environmental cracking, cast stainless steel aging and acoustic emission leak detection programs by Argonne National Laboratory and the NDE reliability, steam generator I
l and NDE material property measurement programs by Battelle Pacific Northwest Laboratory.
A site visit by the subcommittee was made to the Battelle West Jefferson degraded pipe test facility.
Notice of the meeting was published in the Federal Register on Febru-i ary 28, 1989, (Attachment A). The schedule of items covered in the meeting is in attachment B.
A list of handouts kept with the office-copy of the minutes is included in attachment C.
There were no written or oral statements received or presented from members of the public at the meeting.
E. G. Igne was Cognizant ACRS Staff member for the meeting.
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. MINUTES / MATERIAL & METALLURGY 2
March 15-16 Subcommittee Mtg.
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i Principal Attendees ACRS P. G. S.hewmon, Chairman l
C. Y. Michelson, Member i
J. Hutchinson, ACRS Consultant NRC G. Arlotto S. Lee l
J. Muscara, NRC j
M. Mayfield, NRC E. Serpan, NRC l
K. Wichman, NRC l
Others 1
J. Ahmad, Battelle Columbus Div. (BCD)
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C. Marshall, BCD B. Saffell, BCD R. Schmidt, BCD S. Doctor, PNL G. Wilkowski BCD W.
Shack, Argonne Nat'l Lab.
G. Kramer, BCD M. Lapides, EPRI Highlights 1.
W. Shack, ANL, presented a status report on the environmentally assessed cracking in LWR systems program. Historically, the program has emphasized intergranular stress corrosion cracking (IGSCC), but emphasis is currently shifting to stress corrosion cracking (SCC) in ferritic materials, fatigue and irradiation assessed stress corrosion cracking (IASCC).
He stated that more than 1000 instances of cracking have occurred in the past 10 years and to date, it is by far the most costly program associated with environmental degradation in LWR piping.
ANL studies have demonstrated large impact of very low levels of impurities in reacter coolant water on susceptibility to SCC; he identified sulfate as being particularly deleterious.
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March 15-16 Subcommittee Mtg.
J Conversely, ANL has demonstrated that even sensitized SS can show resistence to IGSCC in very high purity water. Studies also showed that replacement materials such as Types 316 NG and 347 NG are not immune to SCC in conventional BWR water chemistry. Under certain loading conditions crack growth rates are comparable to those in sensitized materials, i.e., increased margin is primarily associated with the greater difficulty of initiating cracks.
W. Shack stated that additional work on SCC of stainless steel is
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needed to better quantify effects of water chemistry on crack growth rates of sensitized SS to assess industry requests for inspection relief due to improved water chemistry in operating reactors. This work may impact relief on inspection requirements granted in NUREG-0313, applicability of leak-before-break design of piping constructed of such materials, and need for hydrogen water chemistry in plants with replacement piping.
Mitigation of SCC by organic substance has been observed. Some organic substances adsorb on oxide surfaces and retard oxygen reduction reaction, which is coupled with anodic dissolution at the crack tip for crack growth' by slip-dissolution. This obser-vation affords new means to mitigate SCC in instances where it is either difficult or not desire.ble to decrease the dissolved
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oxygen concentration. Likely decomposition products of these substances do not have an adverse effect on IGSCC of sensitized Type 304 SS. Aliphatic acid at low concentrations, less than 1 i
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March 15-16 Subcommittee Mtg.
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IV ppm, are not toxic, carcinogenic or flammable. There appear to be no unusual requirements associated.with storing or introducing these substances into the feedwater.
Irradiation effects on
.these organ'c substances have not yet been determined..
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In PWR primary systems and BWR vessel, ferritic steels are weld clad with SS or Inconel. SensitizedSSclad(Type 308)has i
generally performed well, although thermal fatigue cracking has l
I been observed.
Inconel 182 or 82 weld metal is used to mitigate I
thermal stresses at attachment ' points of austenitic components to ferritic vessel. They are. susceptible to IGSCC in BWR.
In some tests " starter" cracks in the clad have penetrated into the ferritic material. Currently ANL is performing fracture mechan-ics tests to determine threshold levels of the stress intensity factor as a function of environment material and loading history variables.
ANL is also working on fatigue problems in LWRs. Numerous instances of small-diameter lines have failed due to vibratory loads. Thermal fatigue associated with mixing of hot and cold fluids have also caused piping system failures. Cyclic themal stresses produce relatively shallow cracks but once initiated, cracks can penetrate deeper as a result of low-cycle loads.
IASCC of stainless steels was first observed in the late 1960s in SS fuel cladding. This mechanism of failure is associated with microstructural changes induced by neutron irradiation in l
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March 15-16 Subcommittee Mtg.
4 irradiation studies, both segregation of impurity elements and chromium depletion near grain boundaries are observed. Threshold fluence is greater than 5 x 10E20 n.cm-2; susceptibility in-creases strongly above 10 x E21 n.cm-2.
Chief concern is major support structures like BWR top guide. For the tops guide, end-of-line fluence is 1 x 10E22 n.cm-2.
Current structures now have 2 - 5 x 10E21 n.cm-2.
Objectives of planned work are 1) characterization of the microstructural features of reactor-failed components by AEM and SAM to identify critical metallurgical processes that control susceptibility, 2) determine
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the effect of corrosion potential, impurities and strain rate of IASCC of stsinless steel and 3) evaluate current proposed " fixes; high purity water and high hydrogen addition.
2.
W. Shack, ANL, presented a status report on aging of cast sti,inless steels being performed by O. Chopra and H. Chung at ANL.
Embrittlementofcaststainlesssteel(CSS)mayoccur because of thermal aging during operation. LBB design presumes adequate toughness of materials, which could be affected by this embrittlement. Both test data and proposed mechanisms suggest primary effect of aging is an increase in transition temperature.
Changes in upper shelf toughness are much more modest. Current assessment is that embrittlement will occur during reactor operation even without extended operation.
For compositions of steels used in US reactors the degree of embrittlement does not appear to comprcmise LLB in most cases. Preliminary correlations have been developed to predict kinetics and degree of I
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' March 15-16 Subcommittee Mtg.
embrittlement for specific compositions. Development and valida-tion of these correlations will be completed in FY91.
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L. 'Silack, ANL, presented a status report on evaluation of 3.
W advanced acoustic emission and current leak detection systems being performed by D. Kupperman of ANL. The objective of the program was to 1) assess leak detection technology and establish how well current methods meet R. G. 1.45 and 2) assess potential improvements in leak detection technology. Current capabilities for acoustic emission leak detection system are as listed below:
O A sensitivity of I opm with the sensor located at a distance of 15 feet from the back and with moderate acoustic background noise, and A one foot leak location accuracy.
Based on the ANL program results, acoustic leak detection methodology is recommended for inclusion in R. G. 1.45.
A GARD /ANL two channel advanced acoustic leak monitoring system is available for field monitoring.
4.
G. Wilkowski, Battelle Columbus, presented a summary l
report of the leakage round robin results from the ASME PVP
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l meeting of July 1987 in San Diego, CA. The participants were l
l Battlelle, Ontario Hydro-Canada, GRS, IHI/EPRI/R. Cloud Asso.,
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CEA, KWU, GE, Institut Fuer Reactor-bauelemente, and the USNRC.
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March 15-16 Subcommittee Mtg.
Analytical predictions were made for the following crack conditions:
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crack opening area predictions for circumferentially cracked pipe, leak rate predictions for given crack openings, and leak rates for cracked pipe under loads.
Some conclusions are as listed below:
Tharmal hydraulic models agree with each other to within about 50%.
In general, models will over-predict leakage rate for a fatigue crack in a girth weld.
Crack opening area models were found in good agreement for base metal cracked pipe in bending.
Poor agreement was found for weld metal cracked pipe in bending and for base metal tracked pipe with combined loading.
More data is needed to assess leaks of cracks in welds with combined loading, d
5.
G. Wilkowski, R. Schmidt, and P. Scott, all from Battelle Columbus, presented briefings on the International l
PipingIntegrityResearchGroup(IPIRG) Program. This program basically extends the degraded pipe program that will account for j
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, ' March 15-16 Subcommittee Mtg.
dynamic or cyclic load interactions, cracks in fittings and torsional loads. The program w il develop international consensus on technical basis for LBB and in-service flow ac5ptancecriteria. The cost of the program will be shared by the participating countries involved with the US NRC being-responsible for up to 25% of the total cost. Countries involved are the US, Canada, UK, Japan, France, Switzerland, Sweden, Taiwan and; hopefully Italy coming on board this year.
A major task at the present time is the construction of a test facility of representative piping systems under combined dynamic stresses. This facility is under construction at the West Jefferson facility of Battelle Columbus. Some dynamic material characteristic tests are being run to obtain base line data.
6.
G. Wilkowski, Battelle Columbus, presented highlights of general developments from the recently completed NRC degraded piping program. The program objective was to verify and improve fracture mechanic analyses for nuclear piping. The scope was to assess various piping materials at nuclear power plant operating conditions,288C(550F).
Pipe sizes tested varied from 4 to 42 inches with a circumferential crack configuration. Quasistatic bending and pressure loads were applied to the pipe during testings. Data obtained from the tests were used for LBB and in-f.
service flow assessment analyses. The four year program was I
completed in January 1989 at a cost of approximately 6 million
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.. March 15-16 Subcommittee Mtg.
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dollars.
G. Wilkowski emphasized that the state-of-the-art at the start of the degraded program were as follows:
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Net-section-collapse analysis (limit load) was f
developed and verified on small diameter 304 SS.
GE/EPRI J-Estimation EPFM analysis was developed.
NRC-LBB analysis method was being developed.
j Limited pipe material and pipe fracture data was I
available.
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C. Marshall, Battelle Columbus, presented highlights of the material characterization efforts of the degraded piping program.
The objective of this test was to provide material 4
characterization data for pipes subjected to full-scale fracture tests. These were chemical composition, tensile stress-strain curves, charpy V-notch transition curves and crack-growth resistance, J.
These additional tasks were performed, 1) prediction of large-crack-growth J-R curves, 2) development of SETtest,3)studyofpipeanisotropyeffectsand4) j participation of round robins to determine Tensile tests, J-R curves calculations and use of d-c EP to maintain crack growth.
1 From the large crack growth J-R curves from small specimens, Battellestudies1)revealedpossiblesizeeffectsonJ-Rcurves, especiallyinweld-metaltests,2)developedmethodsfor extrapolating-Rcurvesand3)comparedusefulnessinJD and J, for extrapolating J-R curves.
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Interesting ~ findings from the material characterization studies are as follows:
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Flux welds are significantly less tough than
-inert-gas welds.
Unusually low toughness along the fusion line was found.
Many carbon steel pipes were susceptible to dynamic strain aging 1.e., tensile strength at 550 F was greater than at RT and J and dJ/da at R
550 F were less than at RT.
Many carbon steel pipes exhibited crack instabilities at 550 F.
8.
G. Kramer, Battelle Columbus, discussed the full-scale pipe fracture experiments conducted in the degraded piping program.
Two major objectives of the tests were to obtain fracture data on circumferential1y cracked piping at 550 F, and evaluate limit-load analyses using net-section-collapse. Various crack configurations i.e., through-wall crack, surface crack and complex crack, and combined loads tests were run. Sunnary of the net-section collapse test results are as listed below:
Initiation to maximum load margins are larger for through-wall crack than for surface crack.
Separate screening criterion were developed for through-wall crack and surface crack. The i
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j criterion-shows the limit of net-section collapse analysis.
Unified sample screening criterion were developed for both through-wall and surface cracks that gives 95% lower bound failure loads.
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J. Ahmad, Battelle Columbus, discussed fracture mechanics analyses in the degraded piping program. The objective was to q
provide NRC with simple engineering models for assessing the integrity of nuclear power plant pipes containing cracks using EPFM techniques. The bottom line deliverable of the project was a computer code program NRCPIPE. Material property data and initial flow and pipe geometry are fed into NRCPIPE and LBB assessments i.e., pipe fracture and deformation behavior prediction are the output of the program. The program is based on a finite element model.
10.
G. Wilkowski, Battelle Columbus, discussed the significance of the degraded program results. Some of the state-of-the-art developments are listed below:
Verified and developed new analyses for LBB through-wall crack stability evaluation /.
Developed finite length surface crack pipe EPFM engineering analysis.
Developed energy balance method to predict start of instability and crack arrest.
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March 15-16 Subcommittee Mtg.
Cracked pipe element concept was developed for dynamic pipe system analyses.
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Developed in-service flow acceptance criteria for stainless steel and carbon steel piping.
Weld overlay repair pipe tests show large deformation of uncracked section.
Developed additional through-wall crack fracture estimation schemes with better accuracy.
NRCPIPE Code developed for LBB licensing evaluations.
Pipe experimental data used to assess leakage area predictions in other NRC programs.
11.
- 5. Doctor, PNL, presented a briefing on steam generator group project, NDE results and code activity. The project objectives were to determine the reliability and effectiveness of conventional and near-term field practice eddy-current techniques to detect and size tube degradation and to develop input for revision of ISI Reg. Guide 1.83 and Tube Plugging Reg.
Guide 1.121.
Some conclusions of this program are listed below:
Even with 100% inspection most teams that inspected the steam generator would detect and plug only 65% of the defective tubes.
It was stated that 40% systematic sequential j
sampling was almost as effective as 100%
inspection assuming some clustering of I
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.. March 15-16 Subcommittee Mtg.
degraded tubes.
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- Revised Reg. Guides 1.83 and 1.121 were prepared and reviewed by
' the NRC staff. The draft value-impact study of the revised a
regulatory guides have been completed and under review by the
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NRC.
PNL has participated in the ASME Working Group on eddy current examination for revision'to Appendix IV and development of performance demonstration criteria. These tasks are still in progress.
12.
S. Doctor, PNL, presented a briefing on the evaluation and improvement of NDE reliability for inservice inspection of LWRs.
The program objectives were to quantify the reliability of inservice inspection techniques for LWR primary system components through independent research and to establish means for obtaining improvements in the reliability of inservice inspections.
One of the major aspects of this program is the reliability of inspecting centrifuga11y cast stainless steel (CCSS) piping. The problem is that CSS is anisotropic and nonhomogeneous with extremely large grain size in the austenitic steel. This results in large variations as a function of location in sound velocity, attenuation, field skewing and field partitioning. During the program so far, 1) PNL has documented both material microstructure and sound field maps from a spectrum of CCSS samples, 2) PNL has conducted extensive studies on CCSS through roundrobintestandtechnologiessuchasSAFT,and3)a 1
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MINUTES / MATERIAL.& METALLURGY 14 March 15-16 Subcommittee Mtg.
A state-of-the-art topical report is being drafted and should be
, ready' for internal review on April 1 1989.
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' Extensive ASME Code Activities were discussed. The primary objective is to' upgrade ASME Code and NRC requirements using programmatic research results to improve the reliability of section XI inspe'ctions. Code Case N-409-2 was published. Tiiis code case includes statically designed performance demonstrations to qualify personnel, equipment, and procedure for UT/ISI of
. piping welds.
PNL is preparing NUREG/CR-4882 (Qualification Document) to document qualification activities. Appendix VII on NDE Personnel Training and Qualification was published in the Winter 1988 addenda (1/89) and Appendix VIII on UT/ISI Performance Demonstration was approved by the main committee on 2/89 and submitted to BNCS.
13.
S. Doctor. PNL, discussed a new development program plan for non-destrattive measurement (NDM) methods for measurement of materials properties and property changes due to aging. The i
objective of this program is to review the literature, current expertise and related activities, and prepare a plan for the development of engineering data base and validation of prototypic system for NDM material-properties and degradation due to aging.
S. Doctor-stated that material properties and property changes related to aging such as fracture toughness, hardness, fatigue and yield strength could be measured by nondestructive means I
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. Narch 15-16 Subcommittee Mtg.
i.e., magnetic hysteresis, eddy current, internal friction, holography interferometry, acoustic and acoustic emission. This was the consensus of the recently held workshop. Other workshop I
conliusionsarelistedbelow:
1 Sufficient samples exist that could be used -
to start screening techniques for effectiveness now.
Scme work on magnetic techniques has focused directly on degradation mechanisms and materials relevant to LWR bolts, shows great promise.
The Subcommittee questioned the basis of the NDM program and stated that it would like to be kept informed on its progress.
Subcommittee Action The Subcommittee Chairman will present a subcommittee report on the Degraded Pipe Program at the April 1989 meeting. No ACRS comments are planned by the Subcommittee Chairman.
i NOTE:
A transcript of the meeting is available at the NRC Public Document Room, Gelman Bldg. 2120 "L" Street NW.,
Washington, D.C. Telephone (202) 634-3383 or can be purchased from Heritage Reporting Corporation, 1220 L Street, N.W., Washington, D.C. 20005 Telephone (202) 628-4888.
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