ML17059A578

Non-proprietary Version of Final Rept Plant-Specific Charpy Shift Model for Nine Mile Point Unit 1.

ML17059A578
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Site:	Nine Mile Point
Issue date:	12/31/1994
From:	Michelle Manahan MPM RESEARCH & CONSULTING
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MPM-59401-NP, NUDOCS 9412290180
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Attachment 2 NIAGAI4L.MOHAWKPOWER CORPORATION LICENSE NO. DPR-63 DOCKET NO. 50-220 PLANT SPECIFIC CHARPY SHIFT MODEL FOR NIjME MILE POINT UNIT 1 MPM-59401-NP DECEMBER 1994 (NON-PROPRIETARY VERSION) 94l2290l80 94i220 PDR ADOCK 05000220 P PDR

Plant-Specific

, Charpy Shift Model foi

'ine Mile Point Vnit I December, 1994 CCopyright 1994 M.P. Manahan, Sr.

All nghts reserved Copyright 1994 Niagara Mohawk Power Corporation All rights reserved Report No. MPM-59401-NP

,...SERVING CLIENTNEEDS THROUGH ADVANCED TECHNOLOGY

l Report Number MPM-59401-NP Final Report entitled 7

Plant-Specific Charpy Shift Model for Nine Mile Point Unit 1 prepared for Niagara Mohawk Power Corporation Research & Development 300 Erie Boulevard West Syracuse, NY 13202 by Dr. M.P. Manahan MPM Research & Consulting 915 Pike Street, P.O. Box 840 Lemont, PA 16851-0840 December, 1994 NOTE: This report is the non-proprietary version of MPM Research & Consulting's Report No. MPM-59401.

Copyright 1994 M.P. Manahan, Sr.

All rights reserved Copyright 1994 Niagara Mohawk Power Corporation All rights reserved

TABLE OF CONTENTS Abstract Executive Summary ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.0 Introduction 2.0 Plant-Specific Database Development 3 2.1 Database Scrub 3.0 Analytical Model .....................

3.1 Defect Production................ 5 3.2 Chemical/Microstructural Vanables ... ~ ~ ~ ~ ~ ~ 5

~ ~ ~ ~ ~ ~ ~

3.3 Hardening Mechanisms ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 3.4 Summ uy ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 11 4.0 Charpy Shift Modelling 14 4.2 Database Sub-Division 14 4.3 Regression Analysis 15 5.0 Summary and Conclusions............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 35 6.0 References................ ~ ~ ~ ~ ~ ~ ~ ~ ~ 36 7.0 Nomenclature 39

~Aendicee Appendix A Procedures for Evaluation of the Power Reactor Embrittlement Data Base .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 40 Appendix B Important Chemical and Microstructural Features in RPV Neutron Damage Modelling for A302B and A302B Modified Steel 47

Abstract This report documents the development of a plant-specific Charpy shift model for Nine Mile Point Unit 1 (NMP-1)..The plant-specific model is physically based and incorporates the important microstructural damage mechanisms which are now known and well understood. At fluences below -2 x 10" n/cm'typical boiling water reactor (BWR) end-of-license (EOL) fluence), it is shown that there is no correlation of yield strength elevation or Charpy shift with bulk Cu content for the NMP-1 beltline materials. The analyses and data trends demonstrate that most BWRs operate below the fluence threshold for significant Cu precipitation. This results in a different functional form for the Charpy shift (AT3Q) model than currently used in Regulatory Guide 1.99 (Revision 2) (RG1.99(2)). The Nuclear Regulatory Commission (NRC) model was based primarily on high fluence pressurized water reactor (PWR) data and there were very few surveillance data available in the BWR fluence range when the NRC model was developed.

Depleted zone (cascade cores) damage is expected to be the primary damage component for BWRs. Since depleted zones are shearable defects, the Charpy shift has been shown to be proportional to the square root of fluence. The insignificant change in work hardening behavior exhibited by the tensile data confirm that shearable defects (mainly depleted zones) are the predominant microstructural feature resulting from neutron irradiation.

Based on knowledge of the important radiation damage mechanisms operating in the NMP-1 reactor pressure vessel (RPV) steel, criteria were established for defining the NMP-1 plant-specific data set from the larger NRC Power Reactor-Embrittlement Data Base (PR-EDB).

Application of these criteria to the PR-EDB resulted in a data set containing 37 power reactor surveillance data points in addition to the 3 from NMP-1. Regression analyses yielded an accurate linear model of h.T>> as a function of the square root of fluence. Application of the plant-specific model to NMP-1 will reduce the leakage/hydrostatic test temperature by -41'F.

This will reduce the in-service leak test duration by approximately eight hours for each future startup. In addition, outage scheduling flexibility will be increased as a result of the in-service leak tests being conducted below 212'F.

Executive Summary Reactor pressure vessel (RPV) materials undergo a transition in fracture behavior from brittle to ductile as the test temperature of the material is increased. Charpy V-notch tests are conducted in the nuclear industry to monitor changes in the fracture behavior during irradiation.

Neutron irradiation to fluences above -5 x 10" n/cm'auses an upward shift in the Charpy curve and in the ductile-to-brittle transition temperature (DBTT). The Nuclear Regulatory Commission (NRC) has developed a trend curve model and a calculative procedure for modelling the DBTT shift and this information is described in Regulatory Guide 1.99 (Revision 2) (RG1.99(2)). The nuclear industry tracks this shift through Charpy 30 ft-lb transition temperature measurements.

At the time the RG1.99(2) model was developed, there were few surveillance capsule test results available for fluences in the boiling water reactor (BWR) operating range. Use of a data base which consists predominantly of pressurized water reactor (PWR) data has resulted in an overly conservative material behavior modelling for the Nine Mile Point Unit 1 (NMP-1) beltline plates.

This report shows that with the large amount of data available today, a much more accurate plant-specific model can be developed for use in the BWR fluence range. The development of a Charpy shift (b,T3Q) model for a particular plant is anticipated in RG1.99(2), "To use the surveillance data from a specific plant instead of Regulatory Position 1, one must develop a relationship of ERTM,r to fluence for that plant.". Therefore, the work documented in this report was undertaken to develop a plant-specific Charpy 30 ft-lb transition temperature shift (h.T>>)

model which can be applied to the NMP-1 beltline plates.

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1.0 Introduction Ferritic pressure vessel materials undergo a transition in fracture behavior as a result of their body-centered-cubic (BCC) lattice structure. The Charpy V-notch test is used extensively in reactor pressure vessel (RPV) surveillance programs to characterize the effects of neutron fluence on the Charpy curve. Two key parameters which are monitored are the Charpy curve shift indexed at the 30 ft-lb level (b, T,~) and the drop in the upper shelf energy (hUSE). Current regulations use the ET3Q to determine the shift in the American Society of Mechanical Engineers (ASME) reference stress intensity factor (Kg curve. Therefore, pressure-temperature (P-T) operating curves are shifted to higher temperatures (reduced operating window) precisely in accordance with the Charpy curve shiA. It is essential that accurate trend curve models (b, T3p vs.

fluence) be used to ensure that the P-T curves are appropriately calculated. In the case of Nine Mile Point Unit 1 (NMP-1), accurate representation of the actual material behavior is essential to determine whether in-service leak testing above 212'F is necessary. If it can be shown that in-service leak testing below 212'F is justified with adequate margins of safety, then substantial savings in outage time can be realized in the future.

The Nuclear Regulatory Commission (NRC) updated Regulatory Guide 1.99 and issued Revision 2 in May, 1988 (RG1.99(2)) [RG199]. Revision 3, which is expected to parallel the Revision 2 work in terms of technical approach and content, is currently being developed. The Revision 2 work involved several changes including: the separate treatment of welds and plates; the addition of Ni as a model variable; the removal of P from the Revision 1 model; and the inclusion of guidance for calculating neutron attenuation through the vessel wall based on a displacement per atom (dpa) basis. The final model adopted was based heavily on the work of two NRC contractors (Odette and Gutherie) [Ra84]. The Revision 2 model is based exclusively on the assumption that only hardening mechanisms (particularly Cu precipitation) contribute to the embrittlement trend [Od83] (hence the removal of the P term since P is a surface active element).

Research conducted over the past decade (and particularly over the past 5 years) has brought the physical basis for the NRC model into question [IGRDM4, IGRDM5, Ig92]. It is currently believed that spherical microvoids rarely form in vessel materials [ESER94, Au94, Ba92] (Odette's hardening theory postulates significant microvoid number densities [Ep84]). An energy minimization model [ESER94] shows that in iron it is more likely that vacancy clusters will collapse to loops or exist as a loosely connected collection of individual vacancies. Recent Atom Probe Field Ion Microscopy (APFIM) studies have shown the process of Cu precipitation to be much more complex than originally envisioned [Mi88a, Mi88b, Au94]. Several APFIM workers have reported "clouds" of solute atoms which include Ni, Mn, and Si, and these clouds occasionally are associated with Cu. Miller [Mi88a] has also reported P-rich regions and the precipitation of small rod-shaped spherical Mo,C carbides in the ferrite matrix. The irradiation induced carbides are expected since there is a significant Mo concentration and Mo,C is more stable than the Fe,C produced during final heat treatment. However, little is known at present about the extent to which Mo,C contributes to the total hardening.

With regard to non-hardening embrittlement, the British have recently demonstrated large Charpy shiAs for tempered high P laboratory heats and verified the mechanism to be non-hardening embrittlement which results from transport of P to prior austenite grains Pvic94].

McElroy [Mc94] has also discussed grain boundary embrittlement in higher P Russian steels and MAGNOX welds. However, it is important to bear in mind that intergranular (IG) fracture has not been reported in the U.S. steels. This is most likely because the concentration of surface active elements at boundaries has not reached a critical level for fluences in the low 10" n/cm'ange.

The U.S. light water reactor (LWR) surveillance programs should continue to examine Charpy fracture surfaces to ensure that IG fracture is not a problem.

Finally, the addition of a large amount of data to the NRC's Power Reactor-Embrittlement Database (PR-EDB) has shown that further sub-division of the data beyond that of plate and weld categories is needed. This point is more fully discussed later in, this report. Further discussion concerning the physical mechanisms of neutron damage in RPV steels is also provided in the report sections which follow.

In light of these considerations, and the fact that the RG1.99(2) was developed using a data set with few boiling water reactor (BWR) fluence range data, a prudent approach is to develop a plant-specific trend curve for the NMP-1 beltline plates. Therefore, the focus of this report is strictly on the modified A302B (A302M) material. The model is referred to as "plant-specific" because the database was subdivided to a level which yields a data set which is representative of the A302M material in the NMP-1 beltline region.

2.0 Plant-S ecific Database Develo ment The NRC's PR-EDB [PREDB94] was used as the primary source of data for model development. The PR-EDB is a collection of data from surveillance programs of commercial nuclear reactors (primarily U.S. reactors). The PR-EDB is one database contained within the NRC's Embrittlement Data Base (EDB) which al'so includes data from test reactor irradiations.

While many useful insights can be gained from analysis of test reactor data, the current modelling eFort focused solely on commercial reactor data since the goal is to'produce a model which can be applied directly to NMP-1. The use of test reactor data would not, in general, be appropriate since there are widely varying flux, temperature, and neutron spectra in test reactor irradiations.

The version of the PR-EDB (Version 2) used in this study contains the following Charpy data:

252 capsules from 96 reactors 207 heat affected zone (HAZ) Charpy curves (98 diFerent HAZs) 227 weld Charpy curves (105 different welds) 524 base material Charpy curves (136 different base materials) 2.1 Database Scrub Since the goal of the present work is to develop a Charpy shift plate model for NMP-1, the first step was to delete weld and HAZ data from the PR-EDB files. The NMP-1 data were then corrected (the NMP-1 data in the current PR-EDB does not reflect the material mix-up resolution) and data for several plants (ex., Oyster Creek) were verified for accuracy in cases where surveillance reports were readily available at MPM Research & Consulting.

Inconsistencies, such as temperature and energy units, were then corrected in the PR-EDB files containing the plate data.

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3.0 Anal tical Model This section of the report reviews neutron damage mechanisms and provides the basis for the physically based model. The discussion is limited to damage mechanisms which are relevant to the A302M material. The objective is to develop the proper functional form for least-squares regression using the U.S. LWR PR-EDB.

3.1 Defect Production 3.2 Chemical/Microstructural Variables Steels for RPV plates, such as A302M and A533B, are 0.23C steels with about 1.35Mn, 0.5Ni and 0.5Mo (weight percents). In addition, their levels of impurities or tramp elements are generally 0.01-0.02P, 0.01-0.02S and 0.1-0.3Cu [Kh80]. These impurities can have dramatic effects on mechanical behavior depending in part on the processing of the steels.'o produce the heavy plate needed for RPVs, the steels are generally cast into large ingots which, after cooling, are hot rolled or forged into thick plates which are then re-austenitized, quenched, and tempered.

The plates are then welded into the vessel and the final structure is stress-relief annealed and furnace cooled.

3.3 Hardenin Mechanisms Extensive LWR, Liquid Metal Fast Breeder Reactor, and fusion reactor databases have established that exposure of all metals to fast neutron fluxes results in an increase in the yield strength of the material. In the case of ferritic steels irradiated to high fluences, the yield strength is observed to increase, the ultimate tensile strength (UTS) increases the same as the yield strength, or for some steels modestly, and the ductility (measured as total or uniform elongation in a tensile test or reduction of area) is reduced. Neutron irradiation increases the strength of a metal in two ways:

Source Hardenin - it increases the stress required to start a dislocation moving within the slip system.

Friction Hardenin - once the dislocation is moving, it will be impeded by obstacles close to or lying in the slip plane.

In BCC metals, the pre-existing matrix C atmospheres are very effective in pinning dislocations prior to the application of stress, and the depleted zones formed at LWR fluence levels would not be expected to significantly affect the source hardening. Therefore, we focus attention on friction hardening in the discussion which follows.

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3.4 Summa

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4.0 Char Shift Modellin Examination of the ET,O and h,USE trends with fluence and composition indicate that the A508 forgings should be modelled as a separate sub-division of the database. Accordingly, in the report sections which follow, the A302B, A302M, and A533B materials are grouped together for development of the NMP-1 material-specific model.

13

4.3 Re ression Anal sis

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5.0 Summar and Conclusions 34

6.0 References

[ASM] Metals Handbook, Ninth Edition, Volume 1, Properties and Selection: Irons and Steels.

[AU94] P. Auger, P. Pareige, M. Akamatsu, J-C. Van Duysen, "Microstructural Characterization of Atom Clusters in Irradiated Pressure Vessel Steels and Model Alloys", to be published in the Journal of Nuclear Materials.

[Ba92] A. Barbu, T.N. Le, N. Lorenzelli, F. Maury and C.H. de Novion, "Electron Irradiation Effects on Cu Precipitation in Iron-Based Dilute Alloys", Materials Science Forum Vols. 97-99, 1992, pp. 351-358.

[Do71] C.C. Dollins, Radiat. Eff., 11:33, 1971.

[Ep84] Perrin, J.F., Wullaert, R.A., Odette, G.R., and Lombrozo, P.M., "Physically Based Regression Correlations of Embrittlement Data From Reactor Pressure Vessel Surveillance Programs", Final Report to EPRI, January, 1984.

[ESER94] Manahan, M.P., Cuddy, L.J., "The Physical Basis for Upper Shelf Energy Drop in Irradiated Reactor Pressure Vessel Steels", Final Report to ESEERCO, March, 1994.

[Fa89] Fabry, et al, "Improvement of LWR Pressure Vessel Steel Embrittlement Surveillance:

1984-1986 Progress Report on Belgian Activities in Cooperation with USNRC and Other REED Programs", Reactor Dosime Methods A lications and Standardization ASTM STP 1001, American Society for Testing and Materials, Philadelphia, PA, 1989, pp. 17-37.

[G171] Gladman, T., Holmes, B., and McIvor, L.D., "Effect of Second Phase Particles on the Mechanical Properties of Steel", p. 78, Iron and Steel Institute, London, 1971.

Pg92] Igata, N. and Kayano, H., "Ductility and Hardening of Neutron-Irradiated Fe-Cr and Fe-Cr-Ni Steels", Effects of Radiation on Materials: 15th International Symposium, ASTM STP 1125, R.E. Stoller, A.S. Kumar, and D.S. Gelles, Eds., American Society for Testing and Materials, Philadelphia, 1992, pp. 1243-1255.

PGRDM4] "International Group on Radiation Damage Mechanisms in Pressure Vessel Steels",

IGRDM-IV, November 16-20, 1992, Fontainbleau, France.

PGRDM5] "International Group on Radiation Damage Mechanisms in Pressure Vessel Steels",

IGRDM-IV, May 2-6, 1994, Santa Barbara, CA.

35

~'

[Kh80] J.N. Khass, A.J. Giannuzzi, D.A. Hughes, "Radiation Effects in Boiling Water Reactor Pressure Vessel Steels", Journal of Engineering Materials and Technology, Vol. 102, April 1980, pp. 177-185.

[Li94] E.P. Lippincott, "Westinghouse Surveillance Capsule Neutron Fluence Reevaluation",

Westinghouse Electric Corporation, Report No. WCAP-14044, April, 1994.

[Lu85] G.E. Lucas, G.R. Odette, P.M. Lombrozo, and J.W. Sheckherd, "Effects of Composition, Microstructure, and Temperature on Irradiation Hardening of Pressure Vessel Steels", Effects of Radiation on Materials: TwelAh International Symposium, ASTM STP 870, F.A. Garner and J.S. Perrin, Eds., American Society for Testing and Materials, Philadelphia, 1985, pp. 900-930.

[Ma60] M.J. Makin and F.J. Minter, Acta Met., 8:691 (1960).

[Ma62] M.J. Makin and T.H. Blewitt, Acta Met., 10:241 (1962).

[Ma92] Manahan, M.P., "Upper Shelf Energy Drop Trend Curve Modelling", Niagara Mohawk Power Corporation, NMPC Project No. 03-9425, Report No. MPM-1292315, November 30, 1992.

McElroy, R., "Presentation on Temper Embrittlement", presented at the January, 1994 ASTM E10 meeting, San Francisco, CA.

[Mi88a] M.K. Miller and M.G. Burke, "Microstructural Characterization of Irradiated PWR Steels Using the Atom Probe Field-Ion Microscope", Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, G.J. Theus and J.R. Weeks, Eds., The Metallurgical Society, 1988.

[Mi88b] M.K. Miller, D.T. Hoelzer, F. Ebrahimi, J.R. Hawthorne and M.G. Burke, "Microstructural Characterization of Irradiated Fe-Cu-Ni-P Model Steels",

Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, G.J. Theus and J.R. Weeks, Eds., The Metallurgical Society, 1988.

[0d83] G.R. Odette, "On the Dominant Mechanism of Irradiation Embrittlement of Reactor Pressure Vessel Steels", Pergamon Press Ltd., 1983.

[Od85] Odette, G.R., Lombrozo, P.M., and Wallaert, R.A., "Relationship Between Irradiation Hardening and Embrittlement of Pressure Vessel Steels", Effects of Radiation on Materials: TwelAh International Symposium, ASTM STP 870, American Society for Testing and Materials, Philadelphia, PA, 1985, pp. 840-860.

36

[0176] Olander, D., "Fundamental Aspects of Nuclear Reactor Fuel Elements", U.S.

Department of Commerce, National Technical Information Service, 1976.

[PREDB94] PR-EDB: Power Reactor Embrittlement Data Base, Version 2, NUREG/CR-4816.

[Ra 84] P.N. Randall, "Basis for Revision 2 of U.S. NRC Regulatory Guide 1.99", U.S.

Nuclear Regulatory Commission, 1984.

[Rg 199] U.S. Nuclear Regulatory Commission Regulatory Guide, Revision 2, May 1988.

[Ri51] Rineholt, J.A., and Harris, Jr., W.J., "Effect of Alloying Elements on Notch Toughness of Pearlitic Steels", Transactions of the American Society for Metals, Vol.

43, 1951, p. 1175-1214.

[Se58] A. Seeger, in Proceedings of the Second United Nations International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1958, vol. 6, p. 250, United Nations, New York, 1958.

[St85] Steel, L.E., Davies, L.M., Ingham, T., and Brumovsky, M., "Results of the International Atomic Energy Agency gAEA) Coordinated Research Programs on Irradiation Effects on Advanced Pressure Vessel Steels", Effects of Radiation on Materials: TwelAh International Symposium, ASTM STP 870, American Society for Testing and Materials, Philadelphia, PA, 1985, pp. 863-899.

[Ta89] Taboada, A., Randall, P.N., and Serpan, C.Z. Jr., "Overview of U.S. Research and Regulatory Activities on Neutron Irradiation Embrittlement of Pressure Vessel Steel",

Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels: An International Review hird Volume ASTM STP 1011, American Society for Testing and Materials, Philadelphia, PA, 1989, pp. 27-38.

37

7.0 Nomenclature A angstrom AA atomic absorption APFIM Atom Probe Field Ion Microscopy ASME American Society of Mechanical Engineers BCC body-centered cubic BWR boiling water reactor 6 an~ change in flow stress DBTT ductile brittle transition temperature DPA displacements per atom EDB Embrittlement Data Base EOL End-of-License HAZ heat affected zone ICPS inductively coupled plasma spectrometry IG Intergranular Ka, ASME reference stress intensity factor curve LWR Light Water Reactor NMP-1 Nine Mile Point Unit 1 NMPC Niagara Mohawk Power Corporation NRC Nuclear Regulatory Commission P-T Pressure-Temperature PR-EDB Power Reactor-Embrittlement Database PWR pressurized water reactor b,RTNDr neutron induced shift in ASME nil-ductility reference temperature RB RusselM3rown RG1.99(2) Regulatory Guide 1.99 (Revision 2)

RPV reactor pressure vessel RT room temperature ET)0 Charpy curve shift indexed at the 30 ft-ib b,USE drop in the upper shelf energy USE upper shelf energy UTS ultimate tensile strength XRF x-ray fluorescence 38

I' Appendices 39

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Appendix A Procedures For Evaluation of The Power Reactor Embrittlement Data Base Part I Data Base Evaluation 40

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Part H Data Anal sis for PR-EDB ersion 2 44

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Appendix B Im ortant Chemical and Microstructural Variables in RPV Neutron Dama e Modellin for A533B A302B and A302B Modified Steel 47

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RESEARCH & CONSULTING 915 Plkc Strcctt PO Box 840 Office I 814) 234-88G0 lxmont> PA 1G851%840 I'ax (814)234 ~ 0248 USA

Attachment 3 NIAGARAMOHAWKPOWER CORPORATION LICENSE NO. DPR-63 DOCKET NO. 50-220 WAIVER OF COPYRIGHT RESTRICTIONS

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