ML20058J618
| ML20058J618 | |
| Person / Time | |
|---|---|
| Issue date: | 12/27/1989 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Stepnoy N UNION OF SOVIET SOCIALIST REPUBLICS |
| Shared Package | |
| ML20058J623 | List: |
| References | |
| JCCCNRS-WG-3, NUDOCS 9012020132 | |
| Download: ML20058J618 (85) | |
Text
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Dr. Nikolai N. Ponomarev-Stepnoy First Deputy Director Kurchatov Institute of Atomic Energy Kurchatov Square Moscow. 123182 U.S.S.R.
U.S. Correspondence No. 89-45.
Dear Dr. Ponomarev-Stepnoy:
All our working group leaders are in the process of reviewing the meeting schedules and developing action plans for our JCCCNRS interactions in 1990.
This is the first in a series of letters I will be sending to you to transmit action plans and draft agendas for these meetings.
We look forward to receiving similar comunications from your side.
The first action plan was developed by the Co-Leader of Norkingteroup W L.C. Shao, and it is intended for Dr. A.D. Amaev of the L'archatov Institute.
The action plan describes all the topics and issues to be discussed in our j
meetings in June 1990, plus it includes a description of the actions to be
{-
taken, and who has the primary responsibility. We would be 41eesed to receive your comments on this plan.
Working Group 3 has also developtd a draft agenda for the June meeting; it is enclosed for comment by Dr. Amaev.
Please do not hesitate to propose your own ideas for the agenda or to suggest any modifica-tions to our draft.
As is noted in the action plan on 3.1.2.(4), we have enclosed a copy of a sumary paper on irradiation effects on cladding for your information.
Actions concerning information and materials from the Gundremingen reactor test program are noted in the action plan under 3.1.1.(2) and in the Sumary of Actions as items 5 and 6.
We have conferred with our German partners on this issue and find that they generally are in agreement. However, they wish to have a formal request from the USSR for the reports and material. May I then ask you to write to-the address noted below asking for reports, and what i
material you might wish from the Gundremingen reactor.
Be assured that by I
~ this letter, you have the agreement from the USNRC for access to the reports end the material.
Your letter should be addressed to:
j Dr. Rer. Nat. H. J. Gehrhardt Der Bundesminister fuer Umwelt, Naturschuetz, i
-und Reaktorschicherheit Postfach 120629 by i
5300 Bonn-1
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Dr. Nikolai N. Ponomarev Stepnoy 2-A copy of your letter should also be sent to:
Herr Ettemeyer Technical Director Kernkraftwerk Gundremingen 8871 Gundremmingen Federal Republic of Germany Enclosed is a copy of a report by Kussmaul, foehl and Weissenberg of the MPA Stuttgart that was published in the ASTM STP 1011; this report on Gundremmingen is the latest article currently in print.
However, available for release, pending agreement by the Germans, are NUREG/CR 5201, an article by Hawthorne in ASTM STP 1046 Vol 2, and another article by Hawthorne from the 4th International Symposium on Environmental Degradation of Metals, held in August 1989 at Jekyll Island, Georgia.
i Sincerely, crinir.nl sigt.ed by James M. Taylor Executive Director for Operations
Enclosures:
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Action Plan for US-USSR JCCCNR$ Exchange Working Group No. 3:
Radiation Embrittlement of Pressure Vessel and Support Structure and Annealing the Vessel.
=
Respons'ibiitties and actions for items agreed to at the Octot,er 30-31, 1989 meeting in Rockville, Maryland.
3.1.1 Scientific papers to be delivered at the June 1990 meeting in the USSR;
{
(1) Summing up of annealing experience for VVER-440 reactor vessel and the methodology for determining radiation embrittlement of vessel metal after annealing.
This item should be presented by the USSR, because it deals with the )YER-440 reactors.
(2) Non-destructive (incibding surveillance) net;ods for monitoring L
metal chcracteristics of reactor vessels during oparation and af ter annealing.
Both US and USSR should make presentations on this suSject. This subject was discussed at the June 1989 meeting by the 'JS.
The US would expect to review the previous material, and will tdd information regarding its new efforts to write a Regulatory Guide on Annealing. This information is applicable to topic (1) above s well.
/ Tendency toward radiation embrittlement of YVER-1000 caterials teel alloyed with nickel-chromium 15x2NHFAA and its weldeo seems)
This is a subject for presentation by the USSR.
(4) Research on VVER-440 reactor vessel materials removed froni operating units.
This should be a subject for presentation by the USSR, but will be of high interest to us.
(5) On the nature of radiation damage to reactor vessel materials and related factors.
r There should be presentations from both the US and USSR on this item. With a new contract going into pl6ce with Prof. Bob Odette at the University of California, Santa Barbara, we would have new information and an excellent presentation from him on this subject.
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o (6) Theoretical and experimental research of the thermodynaries applied to the problem of thermal shock in the reactor vessel.
The US will present a sumary of work done to establish heat transfer coefficients for our thermal shock work, and will present a sunnary of work to define transients that could impact our vessels.
We would look forward to similar presentations from the USSR.
j (7) Elastic plastic analysis of fracture mecharics of the embrittled reactor vessel with the goal of ensuring its reliable operation. Results of research of reactor vessel models.
The US plans to present a sumary of its research on pressurized thermal shock model vessels., which includes elastic-plastic analysis and tests on Because of time constraints, the US would not expect to discuss its overall vessel fracture analysis program, but rather would offer that to the USSR for their next visit,from tie USSR, perhaps at Oak Ridge. The US then, looks forward to hearing in 1990, on their PTS program.
(8) Reactor vessel rupture probability, embrittlement effects and nuclear power plant operating and control procedures.
TheUSwillpresenttheresultsofitsIPTS(IntegratedPressurized Thermal Shoc k) program completed several years ago. Assuming that funds are identified in the current budget, the US will present -
plans for an updated evaluation of vessel rupture probability via a new program planned for this fiscal year.
We would look forward to a similar presentation from the USSR on their vessel rupture probability program.
3.1.2 Scientific and organizational matters of the conduct of joint research on problems of radiation metallurgy and increasing operational reliability of reactor vessels and supports.
(1) Suming up and analyzing the scientific and technical results obtained in the US and USSR on annealing irradiated materials and vessels; forming into practical recommendations which could be used for reactor vessel annealing.
The US is interested to pursue this idea. Concrete plans for implementation shculd be made at the June 1990 meeting in the USSR.
It is suggested that contributions to the 1990 meeting be written in a style that could allow them to be easily incorporated into such a document. We note that EPRI has indicated an interest to contribute to the annealing recomendations documents, and would like to attend the June 1990 meeting in the USSR. Because this set of recomendations has been a USSR initiative, it is requested that the USSR prepare an outline of the combined recom-mendations in advance of the meeting so that contributions could be better prepared, and agreement could be much easier at the meeting.
(2) Using US instruments, carrying out joint research programs in the US on materials cut directly from the vessel of a VVER reactor removed from operation. The programs would also examine other irradiated materials.
The US requests a proposal from the USSR concerning what and how much material they wished to provide to the US for examination.
While it will be relatively easy for the US to study small amounts of materials in microscope studies, study of metallurgical test specimens will be hampered by the US lack of suitable facilities for preparation of specimens from irradiated materials.
If the USSR prepared such specimens, the US certainly can test them. Thus, the US needs the proposal to learn more of the USSR intent in this area. The US believes that such exchange should be limited to VVER mateMal at firsth other materials could be included later pending successic! accomplishment of the initial cooperation.
If the USSR de:1res samples of Gundremingen meterial, that should be indicated, but it is cautioned that agreement of the Germans is also necessary. The US will explore this possibility in advance of a formal request by the USSR.
(3) Participation in joint research of neutron flux density on radiation damage of vessel materials and supports, including research conducted within the framework of international programs and in accordance with provisions of those programs.
The US-is interested to follow up on the flux rate information presented by the USSR in June 1989.
It is requested that the pertinent data be sumarized by the USSR in advance and sent to the US for our study.
What is needed is the metallurgical data, including transition temperature shift, steel chemistry and reference material trends, as well as the dosimetry data, including flux, fluence, and spectrum. A description of the physical arrange-ment is necessary, including surveillance capsule location, distance of capsule from core and vessel wall, or test reactor location and associated pertinent data.
If the data are received in enough time for evaluation in-the US, better progress could be made on defining additional cooperative steps.
(4) Research on radiation embrittlement of cladding materials, including chemical composition characteristics and other possible factors.
Influence of post-irradiation annealing on change in mechanical and corrosion characteristics of the materials.
The US has recently completed a major program on this topic and has enclosed a sumery paper; the final report will probably be pub-lished after the next meeting and can be sent on imediately.
TheUShasnotdoneanyannealingofcladding,andhavenoplansfor it. By providing the USSR with this advance sumery of our work, it is hoped that the USSR will have time to form a proposal for additional exchange by the time of the next meeting.
(5) Corrosive mechanical characteristics of base metal and the metal of the welded seem in coolant emironment under the effect of ionizing radiation of varying intensity.
The US has not done any corrosion fatigue work under irradiation, and has no plans for such work.
However, all of the US corrosion fatigue work will be presented at an IAEA meeting to be held in Moscow in May 1990. Because there will also be a meeting of the International Cooperative Group on Cyclic Crack Growth at the same time, the US believes that coordination via this group would be more effective than just US-USSR bilateral exchange or coordination.
t (6) Determination of the vessel lifetime, especially as influenced by more precise elastic-plastic fracture evaluation, research on thermal shock, and calculation of uncertainties in determining metal characteristics and the presence of defects.
This is virtually the entire subject of " aging" of reactor vessels.
Much of what could be covered here is included in other parts of the program in both sections 3.1.1 and 3.1.2 above.
It does, however, provide an excellent vehicle for a quick start on any new issue that comes up of mutual interest. The US believes that this item (6) would be most profitably brought up at the next meeting to determine if there are any new critical issues that should be undertaken.
The US wishes to mention a possible new item for discussion during the 1990 meeting in the USSR. The item stems from a meeting held at NRC in August 1989 by four Soviets who were attendees at the 1989 SMiRT Conference in California.
These were 1. V. Gorynin and B. T. Timofeev of the Central Research Institute of Structural Materials, Y. G. Dragunov of EDO Gidro press, and N. Makhutov of the USSR Academy of Science. The issue, proposed by Prof. Makhutov is for verification of stress analysis codes used to predict stresses in reactor vessels.
Implementation of the proposal would take advantage of the nearly 1000 measurements of stresses and temperatures on a VVER-1000 reactor vessel, and would require the US to make a stress analysis of the plant using measure.
ments and physical data of that plant supplied by the Soviets.
Correlation and reconciliation of the measurements and aredictions would result in the desired code verification.
If the USSR wisies to continue to pursue this idea, then they could so inform NRC prior to the meetings, and assure that L
the appropriate Soviet personnel were present to begin discussions on implementation.
1 i
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0 Sumary of Actions for Working Group 3 Action 1.
Develop and agree on agenda for June 1990 Shao 12/31/89 meeting.
US to develop and send by 12/31.
USSR to coment and return by 1/31/90.
Ameev 1/31/90 2.
Papers must be written and in the mail by Shao, all US 4/13/90 to opposite team coordinator.
Participants
- Amaev 3.
USSR should prepare outline of the Ameev 1/31/90 combined annealing procedure and 3.1.1.(1 3.1.2(1))
recomendations document and assure that papers given by both sides conform to this outline. Outline and proposed revised US i
paper titles in mail by January 31, 1990.
i l
4.
Proposal from USSR on materials and desired Amaev 4/13/90 studies from samples taken from decommissioned 3.1.2(2)
VVER-440 reactors, for study by the US.
Proposal should be in mail by April 13, 1990.
l 5.
Pending confirmation by US partners in the Shao Federal Republic of Germany, US will send reports of completed studies of Gundremingen 3.1.2(2) reactor material.
Information will be mailed, pending agreement, by January 31, 1990.
6.
Offer, or lack thereof, by the US of materials Shao from the Gundremmingen reactor for study by the USSR.
US should have statement in 3.1.2(2) mail by December 31, 1989.-
7.
Provision by the US of a sumary report on Shao 12/31/89 irradiated cladding; report can be mailed by December 31, 1989.
Proposal needed from USSR for further 3.1.2(4) coordinated work on irradiation effects Amaev 4/13/90 on cladding, in mail by 4/13/90.
3.1.2(4) 8.
Notification by US of its interest to discuss Shao 1/31/90 possible verification of stress analysis New item codes using Soviet VVER-1000 measurements and US codes.
Letter notification by the US should be in the mail by January 31, 1990, 9.
Provision of this entire Action Plan and Shao 12/31/89 Sumary of Actions to USSR Working Group Chairman by December 31, 1989.
Hawthorne, Odette, Cheverton, Serpan
P oposed Agenda for JCCCNRS Working Group 3 on Embrittlement and Annealing Moscow, June 25 29, 1990 Monday June 25, 1990 Horning I hour Introductions and welcome Annealing of VVER-440 and Methodology 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> USSR:
I hour US:
Summary of US research on annealing Afternoon Radiation embrittlement of VVER-1000 Materials 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> USSR Research on VVER Materials Removed from Service 1.5 hrs USSR:
0.5 hr US:
Update on Shippingport Shield Tank Materialt Tuesday June 26, 1990 Morning I hour US:
Update of US Radiation Embrittlenert Studies Nondestructive Methods of Vessel Monitoring 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> USSR:
I hour US:
Update of US Surveillance Data Afternoon Radiation Damage Mechanisms 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> USSR:
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> US:
Recent Results and Summary of Position Wednesday June 27, 1990 Morning 1.5 hrs Complete Radiation Damage Mechanisms Thermodynamics of Thermal Shock 1.5 hrs USSR:
I hour US:
Thermodynamics and Accident Scenarios for US PTS Afternoon Elastic-PlasticFractureMechanicsofVessels(PTS) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> USSR:
Thursday June 28, 1990 Morning 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> US:
PTS Analysis Methodology and Validation Afternoon Probability of Yessel failure 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> USSR:
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> US:
Integrated PTS Program: Risk of RPV Failures Friday, June 29, 1990 Morning and afternoon Discussion of new, proposed topics for cooperation and exchange.
Preparation of meeting sumary.
Agreement and signing of memorandum of meeting.
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i SMIRT POWER CONFERENCE SEMINAR NO. 2 4
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Monterey, California August 21 - 22,1989 l-I 9
SESSION 3 l
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t RFFECTS OF 1RRADIATION ON THE FRACTURE PROPERTIES OF RfAINLESS RTEEL VELD OYDIAY CIADDING'
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F. M. Haggag, W. R. Corwin.' and R. K. Hanstad Metals and Ceramics Division Oak Ridge National Labo story Oak Ridge Tennessee 37831 6151 ABSTRACT Stainless steel weld overlay cladding was fabricated using the submerged ore, single. wire, oscillating electrode, and the three wire, series.are methods.
Three layers of sladding were applied to a pressure vessel plate to provide odequate thickness for fabrication of. test specimens, and irradiations were i
conducted at temperatures and to fluenc 9 relevant to power reactor operation.
For the single. wire method, the first 1ayer was type 309, and the upper two layers were type 308 stainless steel. The type 309 was diluted considerably by cucessive setting of the base plate.
The three. wire method used vario s aoabinations of types 308, 309, and 304 stainless steel wold wires, and produced o hi hly controlled weld chemistry, microstructure, and fracture propertie : in oli three layers of the weld.
Postirradiation test results of all cladding specimens show that, in the test temperature range from.125 to 188'C, the yield ytrength increased by 40 to 56, ductility insignificant 1y increased, while there was almost no change in i
ultimate tensile strength. All cladding exhibited ductile to brittle transition behavior during Charpy impact testing due to the dominance of delta ferri,te failures at low tesperatures. On the upper shelf, energy was redue'ed up to 504 due to irradiation exposure.
In addition, radiation damage resulted f.n 13 to 100*C shifts of the Charpy impact tre.nsition temperature at the 41.J 1evel, hrthermore, irradiation exposure of 2.2.5 ma thich compact specimens (0.57cs),
from the three. wire sladding to an average fluence of 2.41 x 10" meutrons/en8 I
i
(>l MeV), resulted in decreases in the initiation ductile fracture toughness,
- Rese ich sponsored by the Office of Nuclear Regulatory Research, U.S.
Muclear Regu'satory Commission, under Interagency Agreement DOE 1886 801198 with the U.S.
Department of Energy under contract DE.AC05 640R21400 with Martin Marietta Rnergy Systems, Inc.
' Engineering Technology Division.
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2 Jh. and the tearing modulus in the test temperature range from 125 to 288'C.
Thia'is in agreement with the reduction in both the CVN upper shelf energy and the lateral expansion.
IrrRODUCTION The ability of stainless ste61 cladding to improve the fracture behavior of an operating nuclear reacter pressure vessel, particularly during certain overcooling transients, any depend grea;:ty on the properties of the irradiated y
cladding.' Therefore, weld overlay eladding ifradiated at temperatures and to fluences relevant to power reactor operation was examined.
Two weld cladding procedures were chosen for the two phases of this study, namely, the single wire oscillating submerged arc and the three vire series are. The primary. differences between these procedures are in the heat input and the resulting amounts of base metal dilution of the stainless steel eladding. In the first phase, previously reported in doisil,a a Charpy V notch (CVN) impact and ta'nsile specimens from a three layer stainless steel ' weld overlay fabricated using the single wire 18 8
procedure were irradiated to 2 x 10 neutrons /ca (>1 HeV) at 288'C.
Cladding x
from the upper veldsent layers, typical of good quality pressure vessel cladding, exhibited very little irradiatien induced degradation. However, ductile to-brittle transition behavior, caused by. temperature dependent failure of the i
l1 residual : delta ferrite, was observed during impact testing.
In contrast, j
specimens'from the first voldoent layer, which'also exhibited transition type A
'b6havior, were. markedly embrittled..
The cause of the-embrittlement was determined - to be high radiation sensitivity of the atypical alcrestructure resulting from excessive (=50%) base metal dilution of the first veldsent layer.
-(Cood commercial singlo wire cladding normally contains 15-25% dilution of base 3
metal.)
'In the second phase,s s a coamercially produced three-wire serias arc
' stainless steel cladding was evaluated under similar irradiation (except for the high-fluence specimens) and testing conditions as in the first phase.
The results of tensile, CVN impact, and f(acture toughness tests,are reported here and compared with the propertisa of the unirradiated asterial.
3 MATnina t
PHASE 1 SINGLE VIR.E CIADDING The specimens were all taken from a single laboratory weldsent fabricated by the automated single vire oscillating submerged are procedure. The welding wires for both types of 308 and 309 stainless steel were 4 sua in diameter and were chosen to be representative of cladding formerly applied in industry. The 1
cladding was deposited on plates that were 114 as thick by 406 as wide by 914
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ma long to minimise distortion and to provide a'dequate heat sink.
The clad I
plates were then postweld heat treated (FWHT) at 621'C'for 40 h to represent commercial practice.
More details of welding procedure and parameters and
- cladding sierostructure are given in ref. 5.
The three layers of cladding were applied to provide adequate cladding thickness (=20 mm) to obtain test specimens.
This contrasts with typical commercial practice, in which a single layer of overlay approximately 5 as thick is applied by either multiple wire or strip-cladding submerged are procedures. The material compositions of each layer of weld metal are given in Table 1.
Subsequent metallographic examination showed 6
that the upper layer appeared typical of good quality light water reactor (LWR) stainless steel weld overlay -whereas the lower layer had incurred excessive
(
_ =50%) dilution as a result of base metal melting during welding.
=
To examine the effect of the varying microstructures, two sets of tensile and Charpy V notch specimens were carefully fabricated to be contained as fully
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as possible within either the upper two layers (nominally type 308 specimens) or the lower layer (nominally type 309 specimens)(Fig.1).
All specimens were fabricated with the specimen axis parallel to the welding direction. The Charpy specimens. wore notched on the surface parallel to and nearer the base metal in
. all cas'es. Ferrite numbers were measured on the finished Charpy specimens with a Ferrite Scope, which locally asasures the percentage of ferromagnetic material in the sample. The nominally type 308 specimens consistently had ferrite numbers of 2 to 6 (corresponding roughly to percentages of ferrite), as did the portion
' of nominally type 309 speci:nons composed of upper weld pass layers. The notched side of the nominally type 309 specimens closest to the base metal interface
_ _ _ _ _ _ __. ~ _ _ _
4 exhibited a wide range of ferrite numbers from 2 to greater than 30 (off scale),
reflecting the large amounts of the ferritic base metal melted by the first weld pass.
FEASE 2: TEREE WIRE CIADDIN3 1
The specimens were taken from commercially produced stainless steel cladding overlaid on a pressure vessel steel plate. The base material, HSST Plate 0128, was a 178 am thick (7 in.) plate of A533 grade S class 1 steel.
Three layers
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of cladding were applied to provide adequate thickness (=20 mm) for fabrication of the test specimens.
The thies. wire series arc procedure, developed' by Combustion Engineering, Inc. (CE),, Chattanooga, Tennessee, produced a highly controlled weld chemistry, microstructure, and fracture properties in all'three layers of the weld.
The three wire series arc weld overlay cladding procedure of this work was-representative of that used in older nuclear pressure vessels. This method has generally been replaced by strip cladding processes.
Various combinations of type.308, 309, and 304 stainless steel wires were used in the three layers of cladding. Table 2 provides the chemical compositions for each layer of the weld overlay.
The cladding was given an initial postweld heat treatment (FWHT) by CE of 593*C for 10 h, notably less than the PWHT of 621'C for 40 h typically given to-the cladding of a pressurized water reactor (PWR) during fabrication.
The cladding rec *eived this= milder PWHT st CE due to requirements of the clad-te am drogram with which this irradiation program material was concurrently f.
icated.- To bring the heat treatment of the cladding into the. range more typically given FWRs. additional FWHT was performed.
Calculations were made -
using the following tempering parameter (TP) (ref. 9):
1 TP = T(20 + log t).x 10-s,
where T = tamperature (K) and t = time (h). According to this method, the FWHT
_ given the cladding at CE was calculated to be equivalent to a time of 2.2 h at 621*C. Therefore, an additional 37.8 h of FWHT at 621'C wa: given the cladding at ORNL to approximate the typical 40-h period.
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I 5
i The delta ferrite content was monitored for all three layers of the cladding
_ with a Fischer Ferrite scope.
The ferrite numbers varied from 7.5 to 10 throughout the three layers of the cladding. The, microstructure of this cladding (Fig. 2) fieveals a distribution of delta ferrite 'in an austenitic matrix, quite typical of microstructures seen in good practice commercial weld overlay cladding l'
in reactor pressure vessels.a s IRRADIATION HISTORY PEASE 1: SING 1.E-VIRE CIADDING The specimens were irradiated by Materials Engineering Associates in the core of the 2 MW pool reac' tor (UBR) at the Nuclear Science and Technology Facility, Buffalo, New York. Two separate capsules were used, one each for the type 308 and 309 stainless steel specimens. The capsules were instrumented with thermocouples and dosimeters and were rotated 180' once during the irradiation for fluence balancing.
The capsule containing the type 308 specimens reached 1
8 an average fluence i i standard deviation of 2,09 x 10 ' neutrons /cm (>l MeV) i 10% during 679 h of irradiation. The capsule containing the type 309 specimens 2
8 reached an average fluence of 2.02 x 10 ' neutrons /cm (>l MeV) 156 in 508 h.
The fluences are for a calculated spectrum based on Fe, Ni, and Co dosimetry wires. Temperatures were maintained at 288 1 14'c except for the initial week of irradiation.
During that time, temperatures as low as 263'c were recorded for the type 308 specimens.
PHASE 2: THREE-WIRE CIADDING The CVN and tensile specimens were irradiated in two capsules (with target 1
8 fluences of 2 and 5 x 10 ' neutrons /cm
(>l MeV)) by Materials Engineering Associates (MEA) in the core of the 2 MW pool rocctor (UBR) at the Nuclear
. Science and Technology Facility, Buffalo, New York.
Each capsule contained 20 CVN and 6 miniature tensile (MT) speciment. and was instrumented with thermocouples and dosimeters. Each capsule was rotated 180* at least once during its irradiation exposure for side to side fluence balancing.
Irradiation
6 temperatures were maintained at 288 i 11'C.
The average fluence for the first capsule was 2.14 x 10 e neutrons /ca
(>l MeV) i 84 following 631 h of t
s irradiation.
The second capsule reached an average fluence of 5.56 x 10 ' neutrons /cas (>1 MeV) i 54 in 1605 h.
These fluences are for a 1
calculated spectrum based on Fe, Ni, and Co dosimetry wires. Eight 12.7 as. thick compact specimens (0.5TCS) were irradiated in a third capsule to an average fluence of 2.41 x 10 ' neutrons /en8 (>1 MeV) i 34 following 637 h of irradiation.
1 REStTLTS AND DISCUSSION
)
FEASE 1: SINGLE FIRE CIADDING 4
I Tensile ' testing was conducted at room temperature, 149'C, and 288'C.
Irradiation increased the yield strength of the type 309 specimens by 30 to 404, whereas the incrasse of the type 308 specimens was only 5 to 254. surprisingly, the total elongation and reduction in area of both mac5 rials increased during irradiation. The effect of irradiation on the Charpy impact properties of the type 308 weld metal representative of typical weld overlay cladding was relatively small (Fig. 3).
Only a very slight upward shift in transition temperature (15'C) and drop in upper shelf (<104) were observed.
The interpretation of the impact results of the nominally type 309 specimens is more complicated. Since the type 309 weld pass was not thick enough to obtain specimens composed entirely of type 309 weld metal, a portion of all specimens nominally called type 309 is indeed type 308.
Macrographs of the irradiated specimens fracture surfaces 'show that over the range of the full Charpy curve, the portion composed of type 309 weldsent remains bright and faceted.
The remainder of the fracture surface, composed of upper cladding layers of type 308 weld metal, exhibits the same behavior seen in fully type 308 specimens. In the f
nominally type 309 specimens, interpreting the Charpy impact curves danands that the dual fracture properties of the type 308 and 309 portions of the material be taken into consideration. Examination of the fracture surfaces showed clearly
'that the type _308 weld metal has a lower transition-temperature than does the type 309. Examining the impact data reveals a bimodal population related to the amount of the tougher type 308 weld metal present in the sample.
The more M
+--
i
~
7 rype 308 in the specimen, the lower the apparent transition temperature of the specimen. Hence, the unirradiated and irradiated specimens were categorized into low and high energy populations based on the percentage of type 308 weld metal measured visually on the fracture surface of each specimen. The most appropriate criteria, for separating the low energy populations were arbitrarily chosen to be less than 70 and 806 type 308 weld metal for the unitradiated and irradiated data sets, respectively, because these produced the most distinct difference between the data sets, c
The effect of irradiation on type 309 cladding was appreciable (Fig. 4).
Both energy populations experienced large drops in upper shelf energy of up to 50% and shif ts in transition temperatute of up to 100*C. The extensive toughness degradation seen in the type 309 material as compared with little in the type 308 is probably due to the higher fraction of ferritic phases in the type 309 resulting from the excessive base metal dilution and their intrinsically higher radiation sensitivity.
PHASE 2: T)DEE VIRE CIADDING f
i t
Unirradiated Results l
Tensile tests were conducted in the temperature range of -125 to 288'C.
4 The effect of specimen orientation on tensile properties was insignificant (Fig. 5). Hence, only MT specimens with their axes oriented ir. the longitudinal (rolling and welding) direction were irradiated at 288'c to the two fluence levels mentioned earlier.
The cladding exhibits an extremely rapid rise in tensile strength below about O'C as shown in Fig. 5.
This fistxe also shows that the ductility increases from high temperatures to a peak near O'C, then decreases at lower temperatures.
At O'C and above, the fracture mode of the tensile specimens is strain controlled and matrix dominated. As the strength increases with decreasing temperature, void coalescence requires greater strain resulting t
l in higher measured elongation. At temperatures below 0*C, however, the ferrite phase begins to dominate in a stress controlled manner and its propensity for cleavage failure leads to lower specimen total strain to failure and, thus, reduced total elongation.
~
8 Charpy impact specimens were machined in the L T, L S, T L, and TS orientations.
The L orientation in all the cladding work here represents the welding direction as well as the rolling direction of the base plate. The four specimen orientations were chosen to simulate' the possibilities of crack extension in the axial and circumferential orientations, both across (T orientation) and through (8 orientation) the cladding of a pressure vessel.
All three wire cladding specimens exhibited ductile to brittle transition behavior (.i=11ar to that of single wire cladding in Fig. 4 and refs. 2-5) during' impact testing, due to the dominance of delta ferrite failures at low temperatures. The test results also show relatively small variations of Charpy impact toughness in four orientations [ Fig. 6(b)).
Hence, CVN frradiated specimens were machined only from the cladding with their notches in the L S orientation 'since this ' orientation exhibited a typical transition temperature as well as a slightly lower upper shelf energy. The CVN data scatter in the four orientations was typical to that shown in Fig. 6(a).
The fracture appearance macroscopically did not change substantially from the upper to lower shelf as shown in Fig. 7.
These CVN specimens (unirradiated, LS orientation), whose test results are shown in Fig. 6(a), were further l
examined in the scanning electron microscope.
The specimen tested at 100'J absorbed 80 J and fractured in a fully ductile manner by microvoid coalescence.
The. spherical particles that initiated the dimples were readily visible on the frt.cture surface. In contrast, the specimen tested at -100'c absorbed only 20 J and fractured in a much more brittle. mode. The fractdre surface of this specimen contained areas of cleavage associated w'ith the ferrite phase (Fig. 8).
Also present were smooth regions believed to be associated with the ferrite austenite l
interfaces, indicating that fracture occurred by interphase separation.
Some isolated patches of dimples and their initiating particles were also present.
Metallographic studies and scanning electron microscopic examination demonstrated-that the fracture of stainless steel cladding is matrix controlled on the upper shelf and ferrite controlled at lower temperatures.s The unirradiated and irradiated 12.7-ass thick compact specimens (0.5TCS)
. wore machined in' the L S orientation; the L orientation here rept rants the rolling direction for the base metal (A533 grade B class 1 steel) as well.:
the welding direction for the three wire stainless steel weld overlay cladding.-
l l
~
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9 l'
Since the cladding thickness was approximate 1*f 20 se, the specimens were machined carefully such that their back surfaces, perpendicular to the crack plane, were parallel and close 'to the top surface r,f the cladding.
This approach was necessary since the nominal total width o'! sach specimen (1.25W) of 31.75 mm was greater than the full thickness of the cladding (20 mm), thereby assuring t'aat the fatigue procrack and its subsequent test estension resided fully 1a the cladding and that only a portion of the specimen, contiaining the loading, holes, was machined from the pressure vessel steel base metal.
The specimens were tested according to ASTM standard E 813 87 (ref.10) using a couputer controlled unloading compliance technique in the test temperature range of -75 to.288'C.
The J integral calculations for deformation J integral (as reconnended in ASTM Standard E 813 8'i) and for the modified J integral were performed t sing equations similar to those given in refs.11 and 12, respectively. The fracuure toughness test results of both unirradiated and irradiated specimens are presented and discussed below.
Effect of Irradiation on Tensile Properties The yield strength of three wire stainless steel cladding increased due to radiation exposure. The effects were greater at room temperature and below (Fig.
1 9); for example, at the fluence of 2 x 10 ' neutrons /cm8
(>l MeV) the yield strength increased by 9, 20, and 284 at test temperatures of 288'C, room 2
temperature, and -125'C, respectively. At the higher fluence level of 5 x 10 '
neutrons /cm8 (>l MeV), the yield strength increased by 6,16, and 344 at the test temperatures of 288'C, room temperature, and -125'C, respectively.
Hence, it can be seen that virtually all the radiation hardening occurred at the lower fluence; increasing the fluence by a factor of 2.5 did not result in further radiation hardening. The effects of irradiation on the ultimate strength were insignificant [ Fig. 10(a)).
The small increase in ductility (see Fig.10(b)]
due to irradiation at 288'C could be due to thermal aging and/or radiation exposure.
To separate these effects, a few tensile specimens will be aged at 288'c br 631 and 1605 h and their test results will be' compared to those of irrraiated specimens. It is believed that the effect of irradiation on ductility as measured by total elongation is similar to the effect of temperature as l
m m,
)
10 explained earlier in the discussion of Fig. 5, i.e.,
irradiation hardening increases the flow stress by strengthening the aastenite and ferrite phases in the stainless steel weld cladding. The ferrite phase is much more sensitive than the austenite phase to both irradiation hardening and low test temperatures.
_ Again, at ambient temperature and above, the increased flow stress results in i
higher strains associated with void coalescence and leads to higher total strain j
c to failure and thus hi her total elongation. Furthermore, at temperatures below 8
ambient, the ferrite phase, which is hardened by both irradiation and temperature, experiences localised cleavage failures.
Those cleavage events reduce the total specimen strain to failure and, thus, reduce the total elongation.
Effect of Irradiation on charpy Impact Properties Irradiation of the three wire stainless steel cladding specimens at 288'c to fluence levels of 2 and 5 x 10 ' neutrons /ca8 (>l HeV') resulted in decreases 2
of the CVN upper shelf energy by 15 and 20% and increases of the 41 J transition temperature by 13 and 28'C, respectively (Fig.11(a)). Figure 11(b) shows that increasing irradiation from 2 to 5 x 10 ' neutrons /ca8 1
further degraded the three wire stainless.' teel cladding. Irradiation also degraded the CVN Interal expansion significantly (Fig.12). The upper shelf lateral expansion was reduced by 43 and 414 at the low and. high fluences, respectively.
Furthermore, the 0.38-am_ (0.015 in.) transition temperature shifts were 41 and 46*C fo'r the low 7
and:high=fluences, respectively.
Lateral expansion values' in thI lower. shelf
. region were also substantially degraded by. irradiation.
Table 3 also provides the hyperbolic tangent curve fit results for the unirradiated and irradiated CVN test results. These results are in general agreement with those for the single-wire cladding produced with good welding practice.3 s Effect of Irradiation on Ductile Fracture Toughness and Testing Modulus Results of the unirradiated and irradiated C.5TCS fracture toughness specimens fabricated from three wire series-arc stainless steel cladding are suimmarized in Table 4.
Table 4 and Figs. 13 and 14 also show that irradiation l-
I 11 exposure'to an average fluence of 2.41 x 1018 8
neutrons /cm (>l MeV) resulted in decreases in both the initiation ductile fracture toughness, Jr, and the tearing modulus at test temperatures of 75'C, room temperature,120*C, and 288'C. This is consistent with the reductions in both the CVN upper shelf energy and lateral' expansion discussed above.
However, the percent reduction in initiation toughness of the 0.5TCS specimens at hish temperatures (e.g., at 288'C) is greater than that of the CVN impact energy but closer to the percent reduction of the CVN lateral expansion.
Table 4 and Fig.13 show that the initiation toughness, Jte, increased from high temperature to a peak (at about ambient temperature for unirradiated specimens and about 50*C for irradiated specimens) and then decreased at low temperatures similar to the ductile behavior shown earlier in Fig. 5.
Table 4 also shows, as expected, that the tearing modulus calculated according to re'f.12 (modified J integral calculation) wa's always higher than that calculated according to ref. 11 (deformation J-integral approach) for both unirradiated and irradiated materials. Unirradiated specimen A13F was not included in Figs.14 and 15 since it was rot side grooved; its results of higher (as compared to the 204 side grooved specimens) J, and tearing 2
modulus were also expected.
An example of the unirradiated and irradiated J integral vs crack extension (J R) curves for specimens tested at 120'C is shown l
in Fig. 15. The Jr, values for the two irradiated specimens tested at -75'c were significantly lower than those for unirradiated specimens (see Table 4).
8 The low value of Jr (23 kJ/m ) for the irradiated specimen tested at 288'c (see Table 4 and Fig. 13) is considerably less than the lowest.'r e value, 2
43.1 kJ/m, observed for the low upper shelf welds tested in the HSST Second and Third Irradiation Series.18 It is also substantially lower than the lowest J,
8 value, 83.3 kJ/a, obtained for the A533 grade B class 1 plate (HSST 02) in the HSST Fourth Irradiation Series.1*
CONCLUSIONS AND DESCRIPTION OF FUTETRE VORK The irradiation effects on the Charpy upper shelf impact and transition temperature of good quality single-wire stainless steel cladding were very small.
The tensile strength and ductility were improved slightly by irradiation.
l
12 Results from the highly diluted type 309 weld metal showed appreciable radiation-induced' degradation of notch impact toughness, even though both the tensile strength and ductility were improved slightly by irradiation. Although this is a single case of single wire cladding; for known cases where welding has produced abnormal cladding with excessive dilution in' operating reactors, the radiation effects on notch impact toughness may be cause for concern.
The effects of neutron irradiation on three wire stainless steel weld ladding, prototypical of commercial light water reactor materials, were c
evaluated at a wide range of test temperatures for conditions similar to those at the end of life of a pressurised water reactor. The yield strength of this cladding increased with irradiation exposure; the increase rate was appreciably higher at low temperaturn (room temperature and below). However, the effects of irradiation on the ultimate tensile strength and ductility (unifora and total elongation) were insignificant.
All the unirradiated and irradiated three wire cladding specimens exhibited ductile to brittle transition behavior similar to that s'oen previously for the single wire cladding.
Again, this was also attributed to the dominance of failure of delta ferrite at lower temperatur ss.
The upper shelf energy was reduced by 15 and 204, while the upper she'.f lateral expansion was re/.uced 43 and 414, at 2.14 and 5.56 x 10 ' neutrons /ca (>l MeV), respectively.
The 2
8 41-J transition temperature shifts were 13 and 28'c for the low and hir) levels of fluence, respectively.
Irradiated 0.57CS specimens tested from 75 to 288'c showed consistent decreases in both ductile initiation fracture toughness and tearing modulus i'n qualitative agreement with observed decreases in Charpy impact energy and lateral expansion.
Extremely low resistance to ductile. crack initiation was observed at the_ test tamperature of 288'c for the irradiated cladding. Considering those assults, the ability of the.cd uless steel cladding-to enhance the structural integrity of irradiated pressure vesaels clad with similar material should be investigated further.
It must be stressed that the results presented and discussed in this paper are derived from only two stainless steel cladding materials; hence, no conclusions can be drawn for different material chemistries and/or welding procedures.
1
- - + - -
.-n.-,
. =_
13 Stainless steel cladding from the decommissioned West German boiling water reactor at Gundreamingen will be examined using subsite specimen techniques to etapare to our reactor data. The subsize specimens will be machined from the rscently acquired four trepans cut from the decoimaissioned reactor.
4 se d
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+
1.
l l
1 l-j
14 ACKNOWLEDCMENTS The authors gratefully acknowledge the personnel of Materials Engineering Associates, particularly.
J. R. Hawthorne, for capsule fabrication and irradiation.
The fractography work of D. J. Alazander is highly appreciated.
We acknowledge T. N. Jones,
R. L. Swain, and E. T. Manneschmidt for their experimental assistance; and J. L. Bishop for preparing the manuscript. We also acknowledge the support of our technical monitor, Michael E. Mayfield, the Materials Engineering Branch Chief, Charles Z. Serpan, Jr., and the U.S. Nuclear Regulatory Commission.
j e
9 e
15 REFERENCES 1.
W. R. Corvin, Assessment of Radiation Elfacts Relating to Reactor Pressure Fessel Cladding, NUREC/CR 3671 (ORNL 6047), Martin Marietta Energy Systems, Inc., Oak Ridge Natl. Lab., July 1984.
2.
W. R. Corwin, R. C. Berggren, and R. K. Nanstad, Charpy Toughness and TensLie Properties of a Neutron Irradiated Stainless Steel Submerged Arc Wald Cladding Overlay, NUREG/CR 3927 (ORNL/TM 9309), Martin Marietta Energy Systems, Inc., Oak Ridge Natl. Lab., September 1984 3.
W. R. Corwin, R. C. Berggren, and R. K. Nanstad, " Fracture Properties of a Neutron Irradiated Stainless Steel Submerged Arc Weld C1' adding Overlay,'
pp. 26-47 in Proceedings of the U.S. Nuclear Regulatory Coamission 14elfth Water Reactor Safety Research Information Meeting, held at Galchersburg, Maryland, October 22 26, 1984, NUREG/CP 0058 Vol. 4. January 1985.
4 W.- R. Corwin, R. C. Berggren, and R. K. Nanstad, "Charpy Toughness and Tensile Properties of a Neutron Irradiat.ed Stainless Steel Submerged Arc Wald Cladding Overlay," pp. 951-71 in Effects of Radiation on Materials, ASIM STP 870, proceedings of the Twelfth International Symposius, Williamsburg, Va.,
F. A. Garner and J. S. Perrin, Eds., American Society for Testing and Materials, Philadelphia,1985.
4 5.
W. R. Cor sin, R. G. Berggren, and R. K. Hanstad, " Fracture Behavior of a Neutron-Irradir.ted Stainless Steel Submerged Arc Weld Cladding Overlay," Nucl.
Eng. Des-89, 119-221 (1985).
6.
F. M. Haggag, W. R. Corvin, P. J. Alexander, and R. K. Nanstad, " Effects of' Irradiation on Strength and Toaghness of Commercial LWR Vessel Cladding,"
. pp.177-93 in Proceedings of the U.S. Nuclest Reguincory Commission Fifteenth Water Reactor Safety Informat!.on Meeting, held at C'sichersburg, Maryland, October 26 29, 1987, NUREG/CP f,091. Vol. 2, February 1988.
~< - - -
_ - - -. ~.
16 7
F. M. Hagges and S. K. Iskander, 'Results of Irradiated Cladding Tests.
and 41ad Picte Experiments," pp. 355 69 in Proceedings of the U.S. Nuclear Regula ary Conunission Sixteenth Vater Reactor Safety Infonmstion Meeting, he Ga.thersburg, Maryland, October 24 27, '1988, NUREC/CP 0097, Vol.
2, at March 1989 8.
F. M. Ma5 gag, W. R. Corwin, D. J. Alexander, and R. K. Nanstad, ' Tensile and Charpy Impact Behavior of an Irradiated Three Vire Series Arc Stainless Steel Cladding," to be published in proceedings of ASTM 14th International Symposium on Effects of radiation on Materials, Andover, Mass., June 27 30, 1988.
9.
R. W. Swindeman, R. K. Nanstad, J. F. King, and' W. J. Stelsman,'Effect of Tempering on the Strength and Toughness of 21/4 Cr 1 Mo Steel Weldsents, ORNL/TM 9307, Martin Marietta Energy Systems, Inc., Oak Ridge Natl. lab., Octob 1984 10.
ASTM E 813 87, " Standard Test Method for Jr. A Measure of Fracture Toughness," pp. 686-700 in Annual Book of ASTM Standards, Vol. 03.01, American j
Society for Testing and Materials, Philadelphia, 1988.
11._ ASTM E 1152 87, ' Standard Test Method for Determining J-R Curves,"
pp. 800 810 in Annual Book of ASTM Standards, Vol. 03.01, American Society for
' Testing and Materials, Philadelphia, 1988,
)
p 12.
H. A. Ernst, " Materials Resistance and Instability Beyond J-Controlled
-Crack Crowth," pp. 191 213 in Klastic-Plastic Fracture, Vol. I, STP 803, C. F. ' Shih and J. P. Gudas, Eds., American Society for. Testing and Materials Philadelphia, 1983.
~_
J
e 17 13.
A. L. Hiser, F. J. Imss,'and B. H. Menke, J.R Curve Characterizacion of Irradiated Iow Upper Shelf Welds, NUREC/CR-3506 (MEA.2028), Materials Engineering Associates Inc., April 1984 14.
J. J. McGowan, R. K. Hanstad, and K. A. Thoms. Characterisation of Irradiated Current Practice Welds and A533 Grade B Class 1 Place for Nuclear Pressure Fessel Service NUREC/CR 4880, Vol. 1 (ORNI. 6484/V1), Martin Marietta Energy systems, Inc., Oak Ridge Natl. Lab., July 1988.
E e
Tcble 1.
Ch::icc1 stepssitten of singlo. wire stainless steel vold cled svaricy used in Phase 1 ef the Seventh Irredicticn Serios Content.a we g layer C
Cr Ni Mo Mn Si Co Cu V
Al Ti P
S Lower 0.145 13.46 6.90 0.47 1.47 0.56 0.066 0.14 0.02 0.014 <0.005 0.018 0.0)
Middle 0.081 18.52 8.81 0.27 1.47. 0.70 0.092 0.10 0.04 0.010 <0.005 0.021 0.01 Upper 0.065 20.01 9.36 0.21 1.49 0.76 0.100 0.09 0.04 0.16 0.006 0.022 0.0) 88alance Fe, with Nb < 0.01; Ta < 0.01; As < 0.03; and B < 0.001 for all layers.
Table 2.
Chemical composition of the three wire stainless steel weld clad overlay used in Phase 2 of the Seventh Irradiation Series Content,a vt g Layer C
Cr Ni Mo Mn
-Si Co Cu-V P
S Lower 0.052 19.75 9.75 0.18 1.59 0.63 0.03 0.07 0.03 0.016 0.014 Middle 0.049 19.38 9.18 0.23 1.28 0.78 0.07 0.36 0.06 0.023 0.017 Upper 0.049 19.34 9.04 0.23 1.34 0.82 0.06 0.39 0,06 0.023 0.017 aBalance Fe, with Nb < 0.01; Ti < 0.01; and N < 0.057 for all 1.syers.
4 Table 3.
Charpy impact test results for stainless steel three wire series arc cladding Transition temperature Energy Lateral
- Fluence, criterion (J)-
expansion s
.Orientationa n/ca
(.C)
(mm)
(>l MeV)
Upper Lower 41 J 68 J 0.38 na shelf shelf Upper Lower LS 0
-41 6
-57 82 13 1.15 0.25 1
LS 2 x 10 '
-28 56
-16 70 9
0.65 0.017 1
LS 5 x 10 '
-13 11 68 12 0.679 0.012 LT 0
-28 11 88 14 TL 0
40 4
86 16 TS 0
-55 7
83 12
- With respect to the base metal where L is the rolling as well as the l
welding direction.
l
Tcblo 4
'Effect of irradiation on the initiation toughness and tearing modulus.of three wire stainless steel cladding M dified J (ref. 12)
. Deformation J (ref.11)
Test Specimen temperature
('C)
J' Tearing' Jte Tearing (gj,3).
modulus (kJ/a )
modulus 8
Onirradiated specimena A130
-75 118 82 117 64 H2
-75 144 67 137 49
\\
A158 20 169 319 165 270 A13D 20 135 265 134 209 A100 20 174 223 171 176 A10E 120 128 311 128 246 H5 120 118 289 119 229 H3.
120 120 288 120 232 i
A13F*
120 159 449 159 359 H6 200 88 293 90 240 H4 200 111 285' 111 231 A15D 288 78 316 77 267 A13C 288 68 192 66 170 H1 288 79 241 82 192 Irradiated anecimens A15F 75 78 50 78 40 A150-
-75 57 40 56 36 A13A:
30 145 218 144 177 A15C 50 128 182 124 146 A10F 120 97 207 94 175 A15A 288 23 223 25 191
- This specimen was not side-grooved while all other specimens in-table were side grooved 204. This specimen was not included in the
- figures of J 2e or tearing modulus vs test temperature.
l
LIST OF FIGURES Fig. 1.
imcation of the Charpy specimens in the single wire stainless steel cladding, nominally called type 308 and 309.
Fig. 2.
Microstructure of three wire stainless steel cladding weld overlay is typical of good quality commercial reactor pressure vessel cladding with delta ferrite in austenitic matrix. Ferrite nusber ranges from about 7.5 to 10 in the three-layer cladding of this study.
Fig. 3.
Effect of irradiation on the Charpy impact energy of type 308 stainless steel cladding.
Fig. 4 Effect of irradiation on the Charpy impact energy of high and low energy populations of the specimens of nominal type 309 cladding.
Fig. 5.
Effect of specimen orientation on unirradiated tensile properties of three wire stainless steel cladding.
(a) Yield strength vs temperature, (b) Ultimate strength vs temperature.
(c) Total elongation vs temperature.
l Fig. 6.
Unitradiated three wire stainless steel eladding shows Charpy impact transition behavior.
(a) CVN energy vs temperature. L S orientation.
(b) Effect of specimen orientation on CVN energy curves.
Fig. 7.
Fracture surfaces of unirradiated three wire stainless - steel cladding CVN specimens tested at Charpy upper and lower shelf temperatures.
Fig. 8.
Scanning electron microscopic examination demorstrates that fracture of stainless steel cladding is matrix controlled at the upper shelf and ferrite controlled at lower temperatures.
Fig. 9.
Effect of neutron irradiation at 288'c on the yield strength of three wire stainless steel cladding (yield strength vs test temperature).
Fig. 10.
Effect of irradiation on the ultimate strength and elongation of three-wire stainless steel cladding.
(a) Ultimate strength vs temperature.
(b) Total elongation.vs' temperature.
J 4
Fig. 11.-
Effect of irradiation on the Charpy impact toughness of three-wire stainless steel cladding.
(a) Charpy transition temperature and upper-shelf energy show increasing degradation 'with increasing neutron fluence.
(b) Increasing neutron fluence from 2 to 5 x 10 ' neutrons /ca8 (>l MeV) resulted-1 in further transition temperature shift.
Fig. 12. Effect of irradiation on the Charpy V notch lateral expansion of three wire stainless steel.ladding.
Fig. 13.
Effect of irraciation on Jr. initiation fracture toughnesa (modified J integral calculation).
Fig. 14 Effect of irradiation on tearing modulus (modified J-integral calculation).
Fig. 15..
Effect of irradiation on the J R (modified J-integral) curve for three wire stainless steel cladding tested at 120*C.
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Location of the Charpy specimens in the single wire stainless steel cladding, nominally called type 308 and 309, unc Pworovpesa L
1 i
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l l
L Fig. 2.
Microstructure of three wire stainless steel cladding weld overlay
, is typical of good quality commercial reactor pressure vessel cladding with delta ferrite in austenitic matrix. Ferrite number ranges from about 7.5 to 10
~
in the three layer cladding of this study.
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Effect et irradiation on the Charpy impact energy of type 308 stainless steel cladding.
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. low energy populations of the specimens of nominal type 309 cladding.
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Fig.;5.
Effect of specimen orientation on unirradiated tensile properties of three wire stainless steel cladding.
(a) Yield strength vs: temperature.
(b) Ultimate strength vs temperature.
(c) Total elongation vs temperatura.
1
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. Fig. 6.
Unirradiated three. wire stainless steel cladding shows Charpy impact transition behavior.
(a) CVN energy vs temperature L.S orientation.
(b) Effect of specimen orientation on CVN energy curves.
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Scanning electron microscopic examination demonstrates that fracture of stainless steel cladding is matrix controlled at the upper shelf
- and ferrite controlled at lower temperatures.
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Fig. 9.. Effect of neutron irradiation at 288'C on the yield strength of three. wire stainless steel cladding (yield strength vs test temperature).
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. ' Fig. 10. Effect of irradiation on the ultimate strength.and elongation-of three wire stainless steel cladding.
(a) Ultimate strength vs temperature.
-(b) Total elongation vs temperature.
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TEMPERATURE ('C) -
fig. 11. Effect of irradiation on the charpy impact-toushness of three-wire stainless steel'eladding.
(a) Charpy transition temperature and upper.
sbsif o energy.. show. increasing degradation with increasing neutron fluence.
(b) Increasing neutron fluence from 2 to 5 x 10 ' neutrons /ca8 (>l MeV) resulted 1
in further transition temperature'shifc.
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CRACK EXTENSION (In)
O 0.025-0.050 0.075 0.100 800 O UNIRRA01ATED (A10E) 4000 0 IRR A01ATED (A10F)
ATURE 120'C/2484 k
600 3000 a a
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.me y
x, s.u y. '. e.n.. :e :..n: * :.,n,'
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Karl Krissmaul,' Jergen Fbhl,' and Thomas Weissenberg' Assurance of the Pressure Vessel Integrity with j
Respect to Irradiation Embrittlement: Activities in the Federal Republic of Gerrnany 8 DTRENCD Kussmaul, K., F6hl.J., and Weiswnberg. T., *Anaerence of the Prnaute Vee.
set inirgrit) ohh Mnpect so Irradist6ea imbrittlement: Aetlehnes la the Federal Republic of f,
Germans.* Rodo snon Emtwittlement o! Nuclear Recesor l'renure s 'esselSwis An Intrvno.
j IsonalReuen (Third lWume). ASTM $71'1011. L L Stetle.14., Amencen boeiet) for Test.
int and histenals, Philadelphia,1989, pp. 3 26 I
AllST R ACI: $afety research oflight water reaclot preuvre vesult in the Federal Repuhle of Germany (FRO)is presentl focuwd on sahdation programs bawd on fracture mechanics 3
concepts which stud) irradiation embntilement. pressunted thermal shock, and cycht ersek growth under operating tenditions Crack initiation values denved from the /rturve could i
be conhrmed as tchable matenal propeny to enable transfershihty of results from small spee.
]
imen testirit to comples strvetures.
For sehdauen ofirradiation surveillance prac, , trepans were saken from a forged shell course of a commercial reactor preuvre sessel wall to etmpare matenal degradation of the 4
mall with that of the st rveillance specimens and to estabhsh propenics through the thickness k
of the wait The esperimenully determined amount of reduction in toughness of the vesse:
wall depends strongly on the onentation of the specimens with respect to the main forging direction of the hot formed ;natsnal. The degredation of the matenal with respect to transi.
tion temperature shift and upper shelf energ) drop in Charp) impact tests could not be pre, dieted conwnstively on the basis of the eJuning tr_tnd tunes in the U.S. Code of Federal T Register. On the other hand, the gradient of embntilement measured in shon distances from
)'
the inner to the outer surface of the vesul wallis much less and therefore less imponsnt than asusse' n the propowd revision of the U.S. Code.
I Art...e rnaienal representing the shell course from which the trepens were taken v as irra.
?
disted in test reacto.s in the United States and the United Kingdom The irradiated arehne malenal shoes a lomtr degradation than the veswl well of the commercial reactor.
i AIT WORDS: fracture mechanics, irradiation embntilement, p:"suriaed thermal shock,
' 'i cye'ic creeb growth, neutron esposure, bght water reactor, femlic attel, pressure vessel steel The previously reported investigations on validation of the integrity oflight water reae.
tor pressure vessels have been continued with enhanced efforts. The validation principic J
telies on venfication and validation ofcalculation codes and mechanistic fracture mechen.
ies. as well as on probabilistic evaluation of nondestructive enemination methods. A chal.
]I 4'
lensing inal on in service degradation of pressure vessel well material of a nuclear power
?*-
plant is underway. Trepens were taken from the Gundremmingen A (KRil A) reactor pres.
' Profeuor and director, division leader, and sewatinc assistant, sospective'y, Staathche Matenat.
prufungsans6 alt thiPAL Universitet Stutt$an, Pfafienwa;dnns 31,7000 Stuttosti 80, West Gerrnany.
3 b
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, ?.s..
.. d,, e y,.$.,,,,,{r',P 4Q + '.,
,.4 3
i i
4 hADIATsON EMBRITTLEMENT i
)
l 1
sure vessel, which was shut down in 1977 after ten yean of operation. Intntiptions with this material and unitradiated reference matenal are being performed within the neope of
{,'p" i i
a cooperation program between the Staatliche Materialprofunpanstalt, Univenitat Stutt.
=
i pn (M PA), U.S. Nuclear Regulatory Commission (USNRC). Matenals Engineenng Asso-
/
l i
cistes(MLA), United States;and Atomic Energy Re.carch Estab!:shment Harwell( AERE.
I
. 4--
I Harwell), United Kingdom.
j F IN To evaluate component failure behavior over a wide toughness range, materials hate 3
j J ], se been selected and treated to achieve propenies repre enting neutron embrittled material Y
1 state, including "wont case" consideration. These materials are available in large quanti-ties and enable testing oflarge specimens with complex stress 6 elds with respect to initia-
]
tion and arrest toughness even under pressurised thermal shock condition. The experi.
b impa 4
ments are aimed at providing a general vahdation of the computational methods and the fracture mechanies concepts on the basis of a modi 6ed / integral evaluation method and i
l to demonstrate the underlaying principle for transferability.
With regard to irradiation and corrosion assisted crack growth, a cyclic esperiment m t
i Olb j
carried out in the VAK power reactor (boiling water reactor) to apply all operational
/-
i e
parameters coincidentally. At this time, however,only prehminary results are available.
l Aust Safel) Assessment According to Present Approach l
{ g,.
The safety assessment of a reactor prusure companent is based on 6ve pnneiples, w hich i
are:(1) quality through production;(2) multiple parties testing;(3) wont case simulation in R & D programs;(4) continuous in. service monitonnf. and documentation and vali-ri,
dation orcodes; and ($) fracture mechanics and nondestructive examinations [I). The 6rst principle is supponed by the four other principles, which serve as independent redundan.
ri cies. This strategy provides a sufheiently large safety margin even if there is no guarantee for a 100% eficetiveness of each single redundancy.
The most common tests to determine matbial19 ugh <ess in a wide temperature range be made w are the Charpy % notch impact test and the drofswc: h: test. From these two tests the prediction bnitic ductile transition temperature RT o,is derived. *4 da the help of the esperimentally perature ()
evaluated fracture toughness nference curve K, a lowee hoJnd fricture toughness curve With this :
for brittle failure can be established (Fig.1)SSME Boiler and Pressure Vessel Code, Sec.
(USNRC R tion lit, App. O," Protection against nonductile failure," and Section XI, App. A 4000, radiation d,
- Material Propenies"; KTA 3201.2 (1980)"Komponenten des Pnm6rkreises von Leicht.
lens enptbc wasserreaktoren: Auslegung, Konstruktion und Berechnung"). In comparison with the cal.
USNRC Re culated load,ing situation of the component in terms of stress intensity, a quantitative nel material safet) margin with respect to temperature and load can be given on this basis, however, The mair only in the linear. elastic fracture mechanies regime. Worldwide research work has dem.
a wide rang onstrated the conservatism of this procedurt, Extensive testing in the FRG using compact Rev. 2 of U
-/
tension (CT) specimens up to $00 mm thickness has essentially con 6rmed the N curve 3
as a lower bound even (c? degraded materials [2].
,..,,, s.,,((
ln the elastic plastic regime the safety against ductile failure is judged from the Charpy Fracture M.
upper shelf energy only, which does not give quantitative fracture toughness propenies and The gisen
,,t thus no quantitative safety margin against ductile failure or stable crack growth. However, ins utigator the required minimum upper shelf energy of 68 J (50 ft lb) has been developed from prac.
toughnus p
' ' ' ' i tical esperience in conventional power plant systems and chemical plants. Due to the com.
esaluation i ples loading altuation dunng pressurized thermal shock, the necessity arose to provide thick compa quantitative data also for this regime in order to verify the minimum toughness require.
The J.inte ments and to de6ne lower bound values.
in the vicini For design purposes a prediction of the degree of material toughress degradation has to initiation lo;
1 l
KUS$ MAUL ET AL ON Pnt$$URC VCS$tL INTEGRITY ptions with i
the SCCPC OI f, t),
- em siuti.
scenns Asso.
3 1.,,,
RT,et
! wil( AERL-i g
p.tsh o
g L
i P.
USE y lr ' y 7f
- merds have i
'8.d material I
'Y
, t,,,
--E brge cuanti.
l.
J j
g
. t..,
,ect to m ii,s.
3 itWAW
. The experi-y itwPERAturtt1*cl hods and the o
e a method and f
[q, i
periment was i Sib -.(
- ' 88
'.i e i
11 cperational E //
f
~~
tr avaihble.
tu ~
g
/ N h.Atttl e, AUSE R T.,oi Rt oi u 3
- /. c.
nciples,which l
nac simulation L
gh -
ten and vah.
tiv m e it>1u,y) slll The first ent redundan.
TtG. l-Code procedure to desermine heror clostoc fracture onechonin empman a to guarantee perature range be made which has to be monitored during operation through surveillance programt. The i two tests the prediction is performed ora the basis of trend curves for both the shift in transition tem.
saperimentally perature (AT :) and the drop in Charpy upper shelf energy (AUSE) as indicated in Fig. I, saghness curve With this predicted material state an adjusted fracture toughness curve can be assessed snel Code. see.
[USNRC Regulatory Guide 1.99, Rev.1/1977,
- Effects of residual elements on predicted App. A 4(00, radiation damage to reactor vessel materials"; KTA 3203 (1964)*0berwachung der Strah.
lensersprodung von Werkstoffen des Reaktordruckbehalters son Leichwasserreaktoren";
ses von Leicht.
i en with the cal.
USNRC Regulatory Guide 1.99, proposed Rev. 2/1966," Radiation damage to reactor ves.
a quantitative sel materials").
csis. however, The main goal of all the irradiation programs is either to confirm these trend curves for work has dem.
a wide range of materials or to establish modified curves as is the case with the proposed
' 'y -
tsing c:mpact Rev. 2 of U.S. llegulatory Guide 1.99.
tthe K ycurve Fracture Slechanics Peoperties on, the Charpy apruperties and The gis en limitations in specimen slie for irradiation surveillance programs have forced swth. However, investigators, on the one hand, to proceed with the development of quantitative fracture sped from prac.
toughness properties f'om Charpy testing and, on the other hand, to focus on test and
- e. e-
>ue to the com.
evaluation techniques using small fracture mechanies specimens, for example,10 mm.
rose to provide thicL compact tension specimen CT 10 [J).
, chness require-
, The J. integral has been proven to be a reliable measure to describe the material behavior 1
in the vicinity of the crack tip. Especially, the stabic extension of the crack beyond the
- l padati n hauo initiation load can be quantified as a function of the /. integral. From the most commonly O.
J
....~,.1,...-
1
,.yMh'j,};).h,J.,.
V
- - - ~,. -..
,,.,,.g_,,.,,_._,m
- e*
6 nAoiATioN tueniTTi tutNT i
i used unloading comphance test technique to establish a crack resistance rune, crack ini.
1 1
tiation values are denved by means of different evaluation methods. It becomes evidei,'
I that the slope of the crack resistance cune (/ccune) depends on specimen siac and geom.
j etr> (Fig 2r Using the ASTM procedure or recommended modifications of that [4.3) crack initiation (/e., Ju) cannot be evaluated independently from specimen sire and 4
peometry, if the blunting line, however,is evaluated on the basis of the esperimentally determined stretch sone and the /rcurve is fitted by an adequate polynomial of higher E
degree, physical crack initiation values /, can be obtained [6,7) (Fig. 3). This /, value is J
l independent of site and geometry and represents a material charnetenstic property, in l,
f some cases the electrical potential drop method leads to initiation values as low or lower l
N l
than J,. whereas from the ASTM procedure values result which can be almost twice as high depending on material toughness and specimen geometry.
N h
i Transferabilit) of Frsctore Mechanics Results Tremendous efforts have been made to demonstrate the transferability of results for the prediction of component failure on the basis of results from small specimen testing. Large.
v scale specimen testing has teen intensified in the past at MPA, using, for exampic. large.
t 5: ale double edte notched tension specimens (DENT) of materials with different toughness levels. B) comparing different fracture toughness properties it becomes obvious that the crack initiation toughness /, can be used as transferability criterion for materials in the toughness range from 40 to 200 J Charpy upper shelf energy. When the espenmentally determined Ja alue of the DENT specimen reaches the level of /, obtained from CT. spec.
r imens testing. crack initiation occurs in the large scale specimen expenment; this can tw l
detected by acoustic emission and electncal potential drop measurements and can be con.
firmed by fractography. The results thus obtained show that /, determined with C7 spec.
imens is in good agreement with the initiation value of the DENT specimen (Fig. 4). which i...ew sooo W
I exp.
I l
tint &
'M
[
mo DD s;
3 lo,gM = 05 p,
N I
econ
.g 4 '
CT25. 20*.sg j
4
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oo to son,qq ac c. nr 6
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mo0-f, r,sh se,'N
. p K5 01 T = 65' C o
J,.'y1 j
co s,',
- ao oc as 40 6.0 so no
(
/
emi ertenson 6 o / mm MO. 4 l**'"!J
- no. 2-3reurves for d&rnt tromemes, matenal 22 MMcCr 37 similar to A SOS Cl 2, "snitral testedat upper shr(ftempseroture. USE a 90 /.
,e
.: - ? ;
4 4
a g,
g
. @ m;, g.3:gfe y e..,y.
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- t f' 1. ' ; i[ W.~" I '#.
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.i 4,,
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- (*
i 1.
-q KUS$ MAUL ET AL. ON PRES $URE VL$$[L INTCORITY 7
i K501(22 NiMoCr 37),1 65*C,20%s 9
. erack ini-
/
60 NMg g}l es evident q,, -
tsetng l
,ce,. Agy andpesm.
that [4.41 h/7/
offset hs'J e site and f
4*
/ l h
, hiersec69n W /$/ j
- !rknentally l
300.o el of higher i
f
[),
J, v:lue is
)
f, p
g ym,t;,n/7j goperty, in J,.
a sw of lower
!i N
j power low /f /
eicees high h
200.0--
r y
J Ao t
Mim mm
.'(
J.
69 0.066 i
d E" ;s d.
5 4
us 0.131 i
c0.0 Jn s UJ 0./ /b 7
euht for the 4-l Jg.,
vu v,y;>4 c ing. Large.
(
snple. large*
1 jh m toughness
}
0.0 j
gus that the
_o g
ersals in the crock extens'on 4 o / mm perimentally h
arn CT spee-nG. 3-D$erent methods to evaluate onutsatson sout ness on CTspmment
- this c
- n be
! can be con.
,oo.
ith CT spec-Q.4), which MS 07 (USE LOJi F
]
B e 600mm 2 We 200mm properties eersved
' ' * ' O.8 from Cf specimen no.
iesimg 1 80*C
)o.s e h3
,/
~
A r,
9 l-1~-
7_
s n
L i
e, 1
to =
1 l
3 determined in
=
l DEN 1 test l3 i 9C
'o eso aio ado ado ese
%lY l
f.70 nG. 4-Anlacetion q(initiation eelues from CT testung to penhct joiture oflarge com.
panents. matenal 22 NiMcCr 31 similar to A SM Cl 2, low upper shelf energy (USE) test
. A SM Cl2.
malenal.
- m..,,.-.
,..c.,
7 mavx.c.
,:,alhrm%.
- a '
~*
s
,<.3 VX'3 0
ftADIATION EMBRfTTLEMENT i
i l
l calculated load for the initiation point of the DENT specim To vahdat ;
1 ings demonstrates that the applied 2 D clastic plastic calculation of J(plain strain) f seillance Tes l quate frseture mechanics concept. structure and the used fracture toughness 3201, a chalh l l
and Ml'A Stu the ability of the matenals to )ield under the applied cons material (forg s
remnungen u DENT specimens occurred on a load at initiation which is typical for deeply notche for insestigati !
DENT apecimens due to the constraint. This behavior was observed for low as we nal'Ja near(
the remaining fracture surface was of a macroscopically brit
' ill be checke and survet!!ar The large amount of data from Charpy and fracturt toughness testing of a varie Shell course ir i materials has provided the basis for a correla tion in the Charpy upper shelf regime bei the pla nt-ani upper shelf energy and crack initiation /,(Fig. 5). This is especially useful for the aness.
t renesors in the Charpy specimens tre available, ment of fracture toughness properties from sur considered in :
hmiled at pres !
Yalidation of Suntillance Results Sun cillance programs to monitor the degradation of the reactor pressure vessel b matenal due to neutron irradiation are being performed in each commercial reactor, Irra.
distion takes place under elevated neutron flux with a lead factor varying from abo 10 or even l$ depending on the <esseldesign. This involves deviation from the pa to which the vessel wallis subjected concerning flus density-time for damage and reco K
cry and thus the damage equihbrium-and the neutron energy spectrum.v.
,g s000.0 24 N/mm 0 = f - MPA
'a 600.0
[b o=
- ASTM 7
f 4
h
/-
i
$ 600.0 5
/
B b
a
.0 7
?
UA 5
,00,
u_ s M
i.
...m NV 0.0 0.0 40.0 100.0 90.0 3
200.0 CHARPY UPPER SHELF ENERGY 4
22 NoMcCr 31somslar to A 308 Cl2 and20 MahfoNi33 semilar to A m
+"w-
- " ~ ' * ' ' '
~~~ ~~
~'~~~ ~
Q KUS$ MAUL [T AL ON PR[$$URE VES$tt INT [GRITY 9
'omparison of the To sahdate the suncillance practice according to ASTM Practwe for Conducting Sur.
aperimental find.
scallance Tests for Light Water Cooled Nuclear Power Reactor Vessels (E 1851 and KT A sin strain) for the
.th).%. a challenging inal was staned in cooperation with USNRC, ME A, AERE Harwell.
t basis for an ade.
and MPA Stuttgart. USNRC has managed and financed the remm al of trepans of the base material (forging) from the reactor pressure vessel (R PV) of the hoiling water reactor Gund.
1 1 l2,F) depend on remmingen unit A (KRR A)in Germany (Fig. 6) and has provided those trepans to MPA instabiht))of the for insestigation. In addition MPA has arranged the remmal of trepans from weld mate-e deeply notched rial of a near core and oficore circumferential weld. The reliability of suncillance practice low as well as for will be checked by comparir.g trepan matenal with the aircady caisting surveillance results able crack growth, and suncillance specimens still untested and with archive material (same heat as veswl ee.
shell course in question), which was stored at General Electric Co. (GEHthe designer of t3 of a veriety of the plant-and was provided b) USN RC, in addition, archive material is irradiated in test if regime between reactors in the United States and United Kingdom so that a wide utnario of methods is ful for the assess-considered in the validation program (Fig. 7). However, the basis for comparison is rather acre usually only hmited at present since in the surveillance program of KR B.A only Charpy speciment and i
0*
X re veswl belt line 2
,N cial resetor. irts.
t frtm about 2 to 4
l e
i vi the r.arameters
}
(
t20'
-. mage and recos.
h O
-d-I 3{X,Y,2 l
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si QR t
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i N m119180$
L VIG. 6 Mi'Giundremmingen A andlacetion oftrensru.
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A ual orienmiim y,,,,
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(' strum ferrfihal t {
t i
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I i
i
,7
- 1.. = 12.7 mm
' Anurs.npo i 3
I
(
Besiden the co I '
l l..E"'.l.",,
ofIhe aims of th 1
which will be in spenmens as
"' T ",',,7,.'".",."m", " '_
- j imens hase beer; l aw, e
l'\\G. 7-Cooperame protroon en trepon onrestigation Garndremmorigers 4 speament as pm full Charpy encri
\\
trepans locaged a for trepan C and J tension specimens (6 mm diameter)from iongitudinaldirection werc investipted Bec crties. For the m of the importance of the results,investi ations are underway to further confirm the 5
e ticit) of the archive material.
of service a fter ten years of operation in 1977. The ve shells, chemical composition (Table I) tensile properties (Table 2) [V), represen sure vessel technology in Germany of the early 1960s. Therefore the chemical compo "g$
an'd the toughness properties do not ca.actly meet today's requirements, but parable to the older generation of plants. From measurements of the water tempe and an assumed miting effer' in the downcomer, the s esselinner surface temperatu ing operation was estimat'J lo be in the range Of 264*Cincluding a temperature ri K due to yheating.
y
[
,,,====
,v T AltLE l-Che nkolcompositoon fut %) offrepan andorchow maternal 20 NohtoCr 26 similar to.4 308 Cl2.
tittati t Baw Materials,20 NiMcCr 26 Weld Material Archne Maserial GEli 11 Tevian O O
& ample from Veneel Head Flange
,I C
0.23
!Q 0.22 Si 0.23 0 05 i
0.22 Mn 0.71 0.17 P
0.11 0.013 1.45 S-0.013 0.012 0.009 O
l 0.012 Cr 0.38 0.005 0.38 s
Mo 0.65 0.10
, g '. '
Ni 0.62 0.75 0.51 0.7$
Cu 0.16 0.25 0.16 flesiti I 0.26 g 3g gj,,,,,j orchne matersal.
l.?'.
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KUS$ MAUL C AL ON PRES $URE VES$tL INTEGRITY 11 T AftLL 2-1rnssit'prmictrots riforchase mattmaltiLN 1 at rumun Imstuvature.
1 emeding ta.tII:4 l9)
Weld Ultimate
$trength,
$trength, Elongation lleduction of l
e tMPal e.(MPal A'. %
Area Z,4 A nal orientation' T 460 620 33 60 Orcumferential onentation' tmain *orkmg direction)
[
L 473 625 40 70
' l., = 12.7 mm. d., = $ 74 mm; //d = 2.2,
- According to sestel amis.
)
liesides the comparison ofirradiated material and the unirradiated archive material, one of the aims of this work is directed to the change in propenies through the wall thickness, which wi!) tv investigated by means of Charpy, tension specimens and compact tension specimens as wcll(Fig. 8). From several trepans (diameter 107 mm). Charp> % notch spec.
imens base teen removed using electric discharge machining (EDM) to obtain as many i
specimens as possible and to facilitate the handling of radioactive material. To establish I
full Charp> energy / temperature curses, longitudinal and transserse stecimens from two
}
trepans located adjacent to each other in axial direction were combined (for en example for trepan C and O, see Fig. 9). A similar procedure is necessary to measure tensile prop.
enics For the machining of CT.2$ and CT.10 compact tension specimens, a sufficient Bee:use authen.
l
&RCHIV! M11(Ri&L l
$IBI*I
=
iket) out 1
a forged uitiesiastivat tat uitst natutil ng pres +
gu,ttu (suttittitt position tic com.
peratutt C.
l n
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l y
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js It
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11611ll 1 (1 Il (I 10 FIG. $~larestigatiors to be performedand spectmen tygies utedhr charattensation ofthe archne material.
m p--
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s.
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.r t;c.
- r. <
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a 12 RADLATION EMBRITTLEMEN1
]
i
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)
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e tue*, t.:
4 titsetts 1146ftll 3
g te tM'(8t tis.l altlllits 9F fill (t d
i 9
l l
l j
q-----f.-----
i D
l fuel loading N't$,'i,l E',5ht,j ENDF/B V r
i i
library. The Measuremei used to denc l
$f {}
f tears, the at i h....
Howes er, tt d f; %. r ff f)
W
. ii..i i i
,t i.,
to tottel 4 teetts TlG. 9-Cutting scheme of trepans for dferent spmenen orpes number of trepans is available. From archive and trepan material a maalmum number of eleven la> cts of Charpy specimens through the thickness could be machined. The test results obtained at MEA [10), AERE Harwell', and MPA for the unitradiated archive material are in good agreement, although testing was performed according to ASTM (MEA, AERE Harwell) and DIN (MPA), respectively (Fig.10). In the transverse (weak) direction (T.I.), the upper shelf amounts to about 105 J (M PA),in the longitudinal (strong) direction (L T) to about 14$ J.
The absolute neutron esposure (fluence) was calculated with the two dimensional trans-pon code DOT 4.2 in combination with a (RJ) and (A,Z) model to obtain three dimen-sional information for the vessel well [/l). All operating and standstill periods were l
considered and also the bumup distribution of the fuel elements at the beginning of each 8 Perwnal communication from C A. English, AERE llerwell. United Kingdom," Compilation of Preliminar> Data from Charry impact Testing on KRB.A Archive in the Pre and Postirradiation Condition. 1986. unpublished article.
F
%:qqff Q g.[hff.l.;y ?,,),
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i KUSSM AUL ET AL ON PRESSURE VESSEL INTEGNTY 13 2 00 LRB.A
,1 3
aktml MaltttAL LMIE !Isnevadenedl MPA 150 i
g,3
+
g E
hatWILL E
,r 2-c-P
) T.t
$100 7
v s.
MtA WPA 5
g, i
s 0
.t00 100 0
100 200 100 't LOO itMPERA1URE V10. lO-Comparison of Charry resultsfrom Air 4. Al.'IW llarmfil. and hIl'A Suttgart.
fuel. loading eyele. The fission spectrum was assumed to be of the Watt type according to ENDF/B Y for the whole core. The tiuelest cross sections were based on the ENDF/R IV i
l library. The neutron esposure was calculated for each trepan (for an esemple, see Fig. II).
Measurements of the Mn.$4 activity on chips taken from several trepans (11] were also used to determine the absolute neutron exposure. Even after a decay time of about nine years, the agreement between transport calculation and measurement was within il$%.
q However, the recent calculation leads to much lo ver neutron esposure than reported car.
q //
l///
/
au n Mar' #
//
[y'y"fgatilit WALL /,
i i
c a
s e number of ~
U
) ///// /
/'
i
';6 A h Q ////
sed. The test ng to ASTM 8, Ts,'Q s
tied archive
,gggpgg g averse (weak) k//7 g
ulinal(strong) s //
's k
j/
csi:nal trans.
N
8
/
. three.dimen.
k//
mm periods were i //
//
saningtfeach.
O 1
100 105 100 105 cm 200
- Compilation of g Agiat st31&It! FROM tett CL8lt! List Pomeradation Fio. I l-Fluence rolculation for trepens C. D. and G.
d
,s g.
v
, ( yp y y.
14 RADIAtl0N EMBRITTLEMENT lier [l/J. which is due to more advanced computer codes and especially to the more accu-rate modelling of the 3 D geometry b) the application oflarge computer capacil).
Charp) impact tests with trepan material show a very pronounced drop in upper shelf j
energ) (USE) and shift in transition temperature for the transveru spctiment (T.L)(for an esample, see Fig.12). Longitudinal specimens (L T) are less sensitive to neutron irra-distion (Fig.13). The low sensitiveness in upper shelf reduction was already indicated by the surveillance specimens irradiated to about i.19"em*8 (IJJ The fluence level of the suncillance specimens has to be reevaluated according to the present state of the art.
Changes in fluence as great as those found in the case of the vessel wall compared with carher results are not to be espected in this case since the evaluation of the specimens is supported by dosimetry measurements. Compared with specimens of the surveillance pro-gram esposed only to temperature but not to neutron inediation, these longitudinal sper.
I i
imens show a behavior very similar to the trrpen material at the low fluence level of about 2.4 10" em*8 (see Fig.13).
I' The lower part of the two Charpy curves is almost identical. This implies that part of I
ri the material change is caused by temperature effects only. From the compilation of Charp>
t lance results in layers through the thickness of the wall, the near surface quenching effect on I
transition temperature becomes obvious (Figs.14,15) It is estended oser the first 10 mm on both sides and is especially pronounced for the transition temperature of the transverse specimens. The upper shelf energy fur archive matenalis uniform through the thickness and does not eshibit any quenching effects (Figt.16,17).
The compilation of results shows the gradient of material degradation by neutron irra-distion aRecting transition temperature and upper shelf energy of both directions (T.L and L.T) (Figs.14-17).
The change in toughness of the transserse (T.L) specimens (Fig.16)is significant even j
at the generally low fluence level. However, the longitudinal specimens (L.T)(Fig.17) are i
l insensitive and become even tougher at the outer surface where the fluence is in the range of I + 10" em*8, which can be attnbuted to a Lind of recovery effect at irradiation tem.
perature and likely insufheient heat treatment of the original material. The USE measured at the outside of the vessel (Fig 16)gives no indication that the archive matenal was not corteetly identified.
1 The projection of the change of properties into the trend cunes of Regulatory Guide Fl<
pan n 200 Its A 11 I
laV[I !
lil
$g ettelVI
- 100
.b i
5 a
j
^
s 50 ft(hLa 6/t e/o
- t. io%4 i
. rec.tec e
isc ree am c sec IlNP(tAfDRI i
VIG.12-Chorry results obtornedfrom trepara in comparsson wuh orchne material (erample)for transwese spenmens.
g g g%.H
,a: ;;at.p;,
1,.(, y - a ;
- fji5,. 'p A " i M.], *I I.',' M. '!*[.h,N..,g' [/ h, 'e ;?.?. i,)(
I
" ' ..~*,'! $ 9 8 '.V.
.9,,
3.
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l q
i KUSSMAUL CT AL. ON PRESSURE VESSEL INTEGRITY 15 the more aceu-gg apacity.
4tl
- A L 1 I
I e in upper shelf I
II'Al l
mens (T L)(for ill
- e neutron irra-SIWilttet t.
- r--
-~=
v intamat sistpv g- ~
i'W' l' ty indicated by E
tett I
1H f.410'ce' net level of the WHtu6e:I
! 3:1c cf the art.
f5*Yll8 ',*'CI
. compared with g 50
/.
1 10 cm se specin. ens is f
arveillance pro.
I i
I 8
igitudinal spee.
700 1H t
1 00 te 300 't 600 elevelef about 7gypg,3;g,g i
its tha1 part of VIG. lS-Charley resultsfrom trepon and archnt mormal on cornporoson estth survrot.
e sti:n cf Chatpy l
lana results (longitudinalsperomens).
ching effect on the first 10 mm f the transverse
& the thickness lH
'y neutron im-(
", a i.l
- tions (T.L and
.P,,,.,:,n's. min enmet u,i e
'a t.i i.i s
[
5'
'issu sit
- ignificant even i
T)(Fig.17) are g,
t is in the range -
a
=
~-
tradiation tem.
fastest mausat
~T USE measured
.u nateral m nog e
to se a
to not e ut 113feett f00N CLAtDat IEfitFatt SuistoTY GUIdt VIG. l4 Trontition ormperature Ter i of transverst specomens (T.L)for orchne and tre.
pan materoof, through thoclness twopertoes.
i see.a t.:
f ausameneste ensact
.P :
e. u n'..'n.ia.n gn W8t4 81 SSWl 81Afitel i<
.\\__
/
l.
. y.
,,,.-(
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ne a
=
=
. - ne i
allaget Faun ttaggies siteFAtt t
- **'"UI fig. lS -Trensorion temurature Too a q(longitudonalspreemeu (L T;for ordin and trepen matereof. through shockness properties
,d e
"".l.am
- n. i.}.,. :... *... a... s:.j..;.
- n '
- [ f L'li
. J'. ;{(?,qq. 'QR 1,*s l J
4 16 RADIATION EMemTTLEMENT iu kee a 1.s
,,g,,,
,,,; g
- g. gg
../....
I e
Isthe 6, o
16 40 sit
{g 46ssestnestit etter6tt::!=
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'J 70 40 M
N IN me !?0 i
9314ett f atet CL Attel letteratt FIG. l 6 Vpper shelfenergy of transserse spmenens (TM(or orchive end tressen ortotes roof. Ihr wth thsclness propertses.
l e
i I
l.99, Rev. I shows that, on the basis of the chemical composition (Cu and P) and the calculated local Duence, the material behavior cannot be predicted conservatively, neither with respeel to the transition shift nor to the drop in upper shelf (Figs.18,19). More f
Fla.I sa timW recentl> proposed trend curses (Regulatory Guide 1.99, proposed Rev. 2), which take Cu and Niinto secount for the evaluation ofshift of transition temperature caused by neutron irradiation.are in the same way nonconservative;however.at a Cuence below I 10*em*'
both trend curves converge. The prediction of upper shelfdrop according to proposed Rev.
case less sens 2 remains unaffected because it is still based on the isolated influence of copper.
neutron empo An important issue is the attenuation ofirradiation damage through the wall (Fig. 20).
Relative to the inner surface (fertite/ cladding interface) transition temperature, the atten-research wor, !
annealing stu untion through the wall is covered conservatively by the Regulatory Guide curves Rev. I parameters.
and proposed Rev. 2 when the local fluence is used to determine the shift at any depth.
With respec Entremely conservative is the equation proposed in Rev 2, which assumes a very Ast gra-imens from ti dient through the wall. This is not in accordance with the measurements obtained from specimens. Sii Mhe trepans.
KRs.4,there Several questions about material behavior under service conditions in comparison with trend curves or irradiation experiments in test reactors have been raised, since results of irradiation in the United States at MEA [H]and United Kingdom at AERE4farwellindi.
Im 848 4 t.t 3
twent auttant 1 Ill 0
nums set g
=
s.
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64 et at W en ut 85fe8:( FWet CLA Min 8 Bittisti FIG. l1-Upper sketfenergytflongitudsnalsomneens(L T)forarchierendtorpen neare.
corparsYn]
rial. through thickness properties.
Ti'5
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KUS$ MAUL ET AL ON PRESSURE VCS$tL INTEGRITY 17 LH
~
]
p.G-g sl heg'6*ee t H see t.-
g g g y g;;
4, /p,--
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g=
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76
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l R"
l 4"
sm 8 62" FLUE NCE (C o lMeV) ely. ncithet Fl0 lE-Atrosured troonitoon Irmperature shift of the Gurndrernnun. ten wttel enormal l9). More
"" n"Hieroson unh predacted enlues accordon.e to l'.S Reendatorr Guode I.W, Rev. l.
ich take Cu by neutron j
10* em'8 I'
cate less sensitiveness of the archive material (identical chemical composition) even at a Posed Rev.
r.
I neutron esposure three times as high as the trepans (Fig. 21). The continuation of the ll (Fig. 20).
I research work, including fracture mechanics testing. metallurgical investigations, and t the atten-j annealing studies, will give r, ore insight in the prevailing mechanisms and responsible parameters.
eves Rev I l
With respect to the validation of surveillance results. it is necessary that transverse spec.
any depth.
imens from trepans with the high degradation are directly compared with surveillance cry flat gra-specimens. Since only longitudinal specimens were used in the surveillance program of sined from KRD4, the feasibility of reconstitution of specimens using electron beam (EB) weld tech.
arison with e,esuns of arwellindl.
l
. US *s * *
- 6 i 7
%' ' ~ g,.m.. -
-g
(
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k.o.m. 94 ore.et...en y
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=_
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FLutesci t o tensv V10. l9-Mrasured arpper shr(f energy drop of the Go,odermmungen reswl material in a neate compurnon wah perducint volun accordung to U.S Regulatory Guide in Arv 1.
I siss===
.dQ U. 10 by 10 mm'.
I were performed Charpy compound specimens were produced.
OD = M10 mm.
The temperature measutement in the specimen center was below irradiation tempera-A major issue lure so that annealing ofirradiation damage can be excluded. Obviously, the residual tour.hness levels stresses do not nNeet the Charpy results inadmissibly, as can be seen from the energ)/
(NKS I and 2), ri I temperature curve in comparison with conventional Charpy specimens (Fig. 22).
sible end of. life e q Tremendous cRe upper shelf Chat Validation of Fracture Mechannes Concepts under Estreme 14ading Sir nation combination of s The most comples loading situation in a component is acting during pressurised thermal treatment.
lirse progra)m was started norae years ago and is now bein$ co in addition to-generait a stress f.
computational methods are used to describe the time. dependent stress / strain field in a
!$9 3
etCNIY! MAf(61A1 114 Ill.ll s,
I l
g Mitt &Olil!0 l
i no 8#AtoEAth&N!LL'MRW!"% Z8 gt L.
b i!31ItatteI(4 9 Af'emi 5
/'
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~!
le i
[
~
ltties 4tt Afft 0
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- 200 15 0
100 fee te 't 600 1(NP[RATURI gg Flo 2ldomprison n(trrpen results with resulufrom imndsstion torriments (archin maternal) on test rencors (ML4 and Harwll) and enarraduatedarchin data.
3 U, '.
y?4&;'ej.Ryjy
{y fif.'
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xussmut ET AL. oN Patssunt visstL witoRITY 19 i
m no.a I
l l
aEtivt lutilist Stl1 1 g
g 150.V LepitillsE rit 6 l
!!, t.,
p- - j- - *- r - -
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ise.v temmeo
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a
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C~'IT"I lit.wite 4.m
.iec e
ist m
Iso *t too itwitatutt
' *5Il rlG. 22-Results ofromtwstor Cloorpy spernmens compared mills rumsenelemolpremlure, Ina vc 3 G)C component like large specimen under PTS conditions, The fracture mechanies analysis is performed on the basis of K, and J lij,16). For the venheation of the calculation. tests were performed applying PTS conditions to thick walled hollow cylinders (ID = 400 mm.
10 mm'. -
OD = 800 mm. length = 1100 mm, circumferential crack)(Fig. 23).
A major issue of the esperimentalinvestigations was the use of materials with diRerent tempers.
toughness tesels to cover a wide range of materials representing optimited RPV stects residual (NKS I and 21. minimum requirements (N KS 3), and lower bound state according to pot.
e energy /
sible end.of hfe degradation including worst case consideration (NKS 4 and 6)(Fig. 24).
Tremendous eRons had to be undertaken to provide these materials. matching in both upper shelf Charpy energy and transition temperature. This could only be achieved by i
combination of selected chemical composition, melting and forging procedure, and heat treatment.
$ thermal in addition to the internal pressure, a tension load of up to 100 MN was applied to dients. A
. generate a stress field similar to the real component. During the cool down phase, as shown
~
advanced 6 eld in a
[ -.-
M penotion Preswe conteor -l voNe j
I i
g~3, 5svoy%
g j
- DeAe
[
1'y Reelstence i,ater W
' Start % Leap ll 100 st g,,,,
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l I
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+
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g,,,
T10. 25--Test sperimen and loopfor F15 simulerion.
i s
+
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f er. u.p.:.f. :. -
y I
20 RADIAtoN tuen TntutNT 1 50 1(elitt Peorittits 41 20 's 3
f L S ORIENTAtl0N L.0tittlAfl0g
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8ts)
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,g ung,gi gg
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400 650 IN
~
ell 3 560
?!0 3
,, t..s <r,,
et$4
$00 0 00 50
'st16 1100 1100 I?MeV84 i
i g
- 200 I
I 200
't 4 00 FEM,PERAtURE 9* hast atl3 utS4 ets6 i
st$2 HG. 24=hf0fereolwedfor PTSinrevogotton rongongpom *as stwofred"w "emeu rose" nt; conderson i
- espa, for example in the temperature / time curve (Fig. 25), starting at 300*C(inner surface) the l
internal pressure and the asialload remained constant. The / integral of the specimen was l
using h the ductile computed on line during the test based upon the instantanecus test conditions (tempera-j ture, pressure, tension load)[/ 7].
dow n. inel for the assessment oferack extension in the PTS expenment,it is assumed that the creek cated in Fi ettension characteristics of CT specimens can be applied to the component as well(Fig.
ldctermine
- 26) This indeed is the case because of the fact that the constraint calculated for the CT g,g,j,;, y
.pecimen used to determine the /, curve is alruost identical with the constraint in the PTS g
specimen. The crael extension derived from this comparison is in good agreement with the crael extension determined by ultrasonic testing and fractographical investigation (see, for example. Fig. 27). This confirms the reliability of the / calculation of the component i
and the fracture mechanics concept. The NKS 4 test with an upper shelf energy of about 70 J is presently being evaluated. Other tests will concentrate on a very low upper shelf energ) of abopt 40 J, b4 On the basis of a validated fracture mechanics concept, the safety of a reactor pressure 5
vessel can be assessed quantitatively also for PTS as demonstrated in a parametric study g3 for the nuclear power plant Obrigheim in Germany [18) v During a PTS transient the maximum load occurs when the temperature of the material I
at the nack tip is still elevated, that is, in the upper shelf regime of the Charpy energy /
temperature curve, if crack initiation takes place,it will be by a ductile mechanism causing k
crack tip blunting and subsequent stable crack extension depending on the applied energy 3
(/ integral). After reaching maximum stress intensity or / integral, respectively, the tran, sient load decreases and the crack tip region is not forced to respond with plastic defor-mation under tension anymore. Therefore, decreasing temperature in combination with decreasing stress intensity cannot cause the specimen to fall even when the load path is intercepting the K. failure curve or failure band, respectively [19,N). In simulation tests MG. *
- W
- i y., - dtl&fg2 O
f'4 f
f
-a,.f.# w.g.,...c w
- - - - - - - - ~ - - - - - - - -
- ~ ~ - - -
Mi yN.
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'ff 4
.,l.> '., '.
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w w.wn. m.i.e..
KU$SMAUL ET AL. ON PRESSURE VESSEL INTEGRITY 21 g
tRAN$lENI FOR Pts. It$15 g
e6Lt tuittstti telem i
flB 4
05160:ttapu 3 30 qwtri(6 ptiett 4
t?lan E
E 1-
- 200 g
Mme N
R W
1,0 x
(
, tem i
0 10 20
- 'a 30 flME a tene*
rl0. 2.G. Temperature tomtfunction for dallerent depth locaroon on the mall durant l'TS operom,'est.
.urface) the -
using $0-rnm thick CT specimens with 20% side groves, specimens were overstressed in the ductile regime (starting point) and then subjected to di#erent load paths during cool-
' U""P"
dow n. including constant load (I), partial (2), total (3), and steady (4) unloading e.s 3.ndi.
cated in Tit 28 (10). When the load path intercepted the K aestler band of the material suhet'stk
[ determined according to ASTM Test Method for Plane. Strain Fracture Toughness of
'g,*',Ih(
Metallic Materials (E 399)]. in none of the cases did failure occur. Only when the speci.
in the PTS nnent with 39%y gati:n (see, component W
1000 gy cf about
't PTS TEST NK$3 i
est track length
. Epper shelr g,ggg 6.
62.6 mm
- 8 00 i )g RFee b
g
{; *y' *',",, g _
orpresture t tric study a
ar ng teese (3) b e me see U
e ts. e,es.ne.
he material 5 200
^ ~-
g g/
/
rpy enersy/
i km csusing
/
w ggg 9
died energy ioniereine y, the tran-88 8H4
- ' I'**I
'4
- astle defor.
O Y**
g_
y g
1.tE cRicx o.3 g _ g g;,;ith 3, c o 1
.a a n ies,s n n u.i - i>. -..as i > ~ - r.In m. -
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.l
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(one cycle pc with about (
mens were reloaded at low temperature, for which bnttle failure has to be expected. did the specimens fail, but then only at significantly higher load than that applicable to those specimens which had not been overstressed in the upper shelf regime. Failure of the spec.
imens basically occurred at a level above the upper bound of the K. scatter band. The so.
called warm prestress eNect"is explained by means of a strip yield model which desenbes the plastic zone at the crack tip and the resulting residual compression forces when s
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T KUSSMAUL FT AL. ON PRESSURE VESSEL INTEGRITY23
[
unloaded. Supplementary to the validation of the / concept, even for superimposed i
mechanical and thermal stresses, credit can be taken from the " warm prestress cKect"in the safet).'..ilysis. This model has been adopted in the safety assessment of the Stade RPV
[
l20). In addition to the provision of the fracture mechanics assessment, the injection of L
l i
coohng water into the hot les during emergency core cooling (EEC)is an important mes-sure to mitigate the PTS situation in the RPV [2/]. which has teen realtred according to requirements of the German Reactor Safety Commission (RSK).
i j
In.Serilee Crack Crowth in a safety case assessment, a postulated crack in the RPV is assumed penetrating through the cladding at the vesselinner surface to that utter has access to the ferritic mate-3 rial. In combination with cyclic loading, temperature, neutrot, and a irradiation changes I
r in the materials as well as the Asw state have to be considered according to diferent mech-anisms in consequence of the postulated crack (Fig. 29). The simultaneous action of these i
diferent parameters can only be studied in the real environment of a power reactor.
Within the German FKS R & D program an experiment was initiated by the MPA and performed in cooperation with GKSS and the Versuchsstomkraftwerk Kahl(VAK) as teponed earlier [21.22). Compact tension specimens (CT 40 mm) were loaded cyclically L
(one cycle per minute)in an open loop in the VAK power reactor (BWR water chemistry with about 0.4 ppn4 oxygen) and the crack advancement determined by measuring (Se preted, did sie to those bf the spec.
nd. The so-s..t'..
- v
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- 00. N4MlWiWed $0er$ by Wi0nal condillCn3 On NneldYiel depadelion and cred ponth.
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24 RADLAT60N EMBRITTLEMENT crack opening displacement of each specimen. Preliminary evaluation of the measure.
Adnnledeme ments by G KSS give no indication of synergistic effects due to the combination of all seth.g I
l operating parameten. Final results will be reponed shonly l#).
I
. e a 'h dinb$-
Conclesion
)
cooperate.
Thants are d Ongoing R & D programs in the Federal Republic of Germany are focused on the vali.
Mi"I5 IO' D dation of the existing concepts for stress strain cateulation (computer codes), determina.
1 Technical Advi tion of reliable fracture mechanics material propenies,ibe time dependent material stale and encouragen as it degrades by operational parameters, and validation otNondestructive examination.
The authors Challenging trials in this content are the investigations direced to pressurited thermal Nuclear Regula y
shock (PTS) and especially to trepans removed from a commercici reactor pressure vessel
(
with respect to code verification.
/
The main results of the research work in the Federal Republic of Germany up to now ll References including evaluation ofinternatioral p ograms can be summarized as follows.
p) Kussmaul, k Nuclear Enn
- 1. J. integral and an assessment of erack initiation J,(which differs from the ASTM pro.
(2) Kussmaul.K cedure)is a reliable tool to evaluate onset of ersel ettension in complet geometries and of C'*(L'd C loading situations even for the combination of mechanical and thermal stresses as acting
,'he I 5
during PTS.
\\;
p) Kuismaul, K
- 2. The amount of stable crack catension depends on the Charpy V. notch upper shelf nation Techm energy (USE), the existing constraint, and the severity of the transient. For deep creeks on Recent Te (aX T) the growing crael does not penetrate through the wall even for low USE materials Madnd, Spai N
and severe transients. This is due to the gradient of temperature and the resulting stress
h,[,,'
decreasing in thickness direction, in case of shallow eracks, which can initiate in a brittle Nuclear Rest mode, funher investigations are under consideration with respect to crael arrest and PTS pl Schasibe.K.
behavior.
10th MPA Se
- 3. Investigations of trepans from the Gundremmingen reactor pressure vesselin com.
py(
bination with catensive neutron field calculation indicate a strong conservatism of the uoflen mmet Code (USNRC Regulatory Guide 1.99, proposed Rev. 2) presently in preparation with barkeit in dei i
respect to attenuation of material degradation through the thickness of the vessel wall.
%'est German
- 4. The materialof the Gundremmingen vesselexhibits directionality effects in the initial 17)
E and state and differences in sensitivity against neutron irradiation depending on specimen ori.
ASTM Journ, entation relative to the main working direction of the forging. Transverse specimens are p) koos, E., Eine more sensitive with respect to both upper shelf drop and transition temperature shift. The nungszusiand material degradation of these transverse specimens could not be predicted conservatively und verfusbs by the existing trend curves of the Code.
$"g'7taan. W
- 5. The irradiation response of the Gundremmingen reactor pressure vessel material dur.
- 19) Ha.ihorne, J l
ing operation is not in accordance with recently obtained results from irradiation experi.
MEA Report.
1 ments performed with the corresponding archive materialin test reactors. The caperiments ha m, M D,19; recently performed in the United States and United Kingdom show less degradation than UO) gs.
the specimens taken from the vessel wall. Raasons might be attributed to the neutron flux pfj pntinger,a '
level, which is about 100 times higher in the test reactor than at the RPV wall and differ.
CR a791, NJe ences in neuiron energy distribution.
l 112) Schroder, R..'
O. L und P de in the Federal Republic of Germany a safety strategy has been developed on a theoretical and experimental basis including plant specific surveillance programs for monitoring mate.
pjj hE%,,K" i
pr,et rgs.,
rial degradation through irradiation, in combination with operational changes, for enam.
149-W.
plc. reduction of design life time (DLT) fluence and mitigation of PTS by hot leg water ll() Ham 1horne,J.
injection, irradiation embrittlement is not considered an aspect that requires restrictions p siti n and
- for the light water reactor (LWR) plants.
p3;I[$
E mvn
-- - - ~ * * - + " ' - ' " * ' ' ' ' - ~ ' - - ' " ^ ^ " " - ' - ' - * * " ~ " " ~ ' ~ '
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i i
KUSSMAUL ET AL ON PRESSURE VESSEL INTEGRITY 2$
I i si the encasure.
I Acknowledgment 50"'I'U *#N"8 The authors wish to thank all cap:rts from research establishments and industry involved in this national and panly international R A D work for their willingness to i
cooperate.
l Thanks are due to the Federal Minister for Research and Technology (BMFT) and the I
Minister for Environment, Protection of Nature and Nuclear Safety (BMU) and their l
i Technical Advisory Committees for continuous scientific. Anancial and practical support de n.
and encouragement.
s
' * * ' ' " ' ;,'[
The authors acknowledge especially the cooperation of C. Z, Serpen from the U.S.
surized thermal i
Nuckar hgdata Gmmnsion.
pressure vessel t
{
eny up to now
[
References j
WS (1) Kussmaul, K.,
- German basis Safety Concept Rules Out Possibihty of Catastrophie Failure **
l Nucimr Enaineenng iniernanonal. December I984, pp. 4I 46.
l the ASTM pro.
(2) Kusimaul. Ka rohl, J., and Roos. E.,"Some Conclusions with Regard to the Safety Assessment n
I I
8cometries and of Crsched Components Drawn from the Research Program Integnty of Components (FKS) at fesses cs acting the Present State " 12th MPA Seminar on Safety and Reliabihty of Pressure Components.
Stasthche Materialpr9fungsanstalt. 9.-10 Oct.1986. Stungart, West Germany,1986.
lJ) Kussmaul, K., Fohl. J., and Roos. E.," Application or Advanced Matenal. Design and Compu.
1eh uppet shelf tahonlechnologies to Enhanced Safety Rehabihty and Availability,"I AEA Somahsts' Meeting
- ct deep cracks on Recent Trends in Development of Reactor Pressure Circuit Technology,25 28 Nov.1985,
, USE materials Madnd. Spain, l'rcssure l'essel and hinng. Vol. 25,1986, pp.183-215.
resulting strest
[i) Las. F. J. et al., *Siructuralintegrity of Water Resetor Pressure Boundo Componenis," Quar.
tiate in a brittle teri) Progress Repon (April-June 1979), NUREO/CR 094), NRC Memorandum Repon a( M, i
Nucitar Regulator) Commission, Washington, DC, September 1979, crrest and PTS l.') Schwalbe, K. H. and Heerens, J.,"Vorschlag rur Modshaierung der JeNorm ASTM E 813 8L" 10th M PA Seminar on Sicherheit und Verfugberkeit in der Anlagentechnik, Nuc/sr Enginerse,t
- 5el in com' j
ond Design. Vol. 87,1985,pp.101 107.
uvatism of the
[6] Roos. E., Eisele, U., Beyer, H., and Gillot, R., *Einordnung und Charskientierung von Werk.
stoffen mittels bruchmechanischer Parameter," 12th MPA Seminar on Sicherheit onel Verfug.
teparation with
'1 barkeit in der Anlagentechnik, Staathche Matenalprofungarnstalt 910 Oct.1986, Stuttgart, 1 vessel trall.
West Germany,1986, prepared for publication in Nuc/cor Engineering sad tksign.1987.
?
etsin the a..tial
[7] Roos, E. and Eisele U.,
- Remarks Concerning the Probl.m of Determination of Elastic Plastic j
m Fracture Meehanies by Means of J. integral Creek Re6 stance Curves" to be pubhshed in the n specimen ori' ASTM Journalof Tesnnt andErcluatoon.
t specimens are I
\\
[8] Roos. E., Eisele, U., Silcher, H., and Speeth, F.,"El alluss der Werksieffsahigkeit und des Span.
cture shift The nungsrustandes auf das Versagensverhalten von Ora 0 proben." 12th MPA Seminar on Sicherheit J consenstivel#
und Verfugbarkeit in der Anlagentechnik, Staatl the Matenstprofungsanstalt,9 10 Oct.1986, 1
Stungan, West Germany,1986, prepared for p' bhcation in Nuc/cor Engenunas and Design.
- 1987,
(
Ci material dur.
l9) Hachome, J. R., "Preirradiation Oualincatit n of Meterials identined as KRB.A Archive,"
adiation esperi.
MEA Repon.2095, NRC contract 04 84102,1985, Matenals Engineenns Associates, Inc, Lan.
l 1
.he riaperiments ham. M D,1985.
, agradatiin than
[10) Hamhorne, J. R.,"Through. Thickness Mechanical Propenies of KRB.A Archive Sicel Forging
- lie neutron flut OEB.* MEA repon 2159, Matenals Engmeenns Associates, Inc., Lanham, MD,1986.
lill Pnthaser,G.,* Neutron Spectrum Calculation forGundremmingen KRB.A Reactor,"NUREO/
wall and differ.
j CR 4791, Nuclear Regulatory Commission, Washiagion, DC, November 1986.
ll2) Schr6 der, R., *0ammaspektrometnache Messungen von Sohrsp6nen ave den Bohrhernen A. D, O, I. und P des RDB Oundremmingen A," Techn. Notis der KfA.Johch, IRW TN.24/86, Kem.
en a htical forschungsanlage Jhhch, Julich, West Germany,1986.
Enittring mate-(/J) Eickelpatch, N. and Seepolt, R., *Empenmentelle Ermittlung der Neutronendo.is des KRB Druckgethsses und deren betnebhche Bedeutung." Aro*Aernenert'e. Vol. 29,1977. No. 2. Pp.
ages, for esam.
149-l$2.
'y hit leg water
[l#) Hamhome, J. R. and Hiser, A. L. " Dose Rate Efeess on irradiation Embrittlement and Com.
g rWet;ons position and Temperature ENects on AnnealingeReitradiation Sensitivity," MEA report 2187, Matenals ens neenns Associates, Inc., Lanham, MD,1986.
i (13) Iskander, S. K., Sauter, A. W., and Fohl, J.,"Resetor Pressure Vessel Implientions of Embrittle.
t
.. a. -.. c.,,,,
s em
,_____-e
- - - - - - - - - - - - - - - - ' ^ - ^ ' '
F
-a i ;
9 26 RADIATION EMBRITTLEMENT i
ment to the Pressurised Thermal Shock $cenario.* in Radiatson Emhrittlement o/Nuc/me A'coc.
sor Pressure l'esselSteely An internationalReview(Swond l'olume). ASTAl N1r wv. L sinte, Ed., American Society for Tesiing and Maienals. Philadelphia,1966. pp. If0-116.
j (16) Jansk), J., F6hl, J.,5avier, A. W., and 16kander, S. K.,"Lnfluss der Werksto0Ahiskeit auf Sche.
i J
disung ourth Thermoschack,* 9th MPA seminar on Scherticit der druckruhrenden Umschhes.
aung s on leichtweseerreaktoren, Staathche Matenalprtfunpanstalt,13 14 Oct.1983. Stuttaan, 4
West Germany,1983.
(IT) Kussmaul, K. and Sauter A. W., *Anlagensimulation sur behandlung der Notkbhlproblematik,
l beim teichtwaaner Drvekbehalter," 10th MPA Seminar en sicherheit und Verfusbarkeit in der Anlageniechnik, staathche Matenalprbrungsanstalt,10 13 Oct.1984, Stuttsart. % est German),
Nuclear Engineering and Design. Vol. 81, I985. pp. 313 321.
[l$) Kunmaul, K., Sauter, A. W., Weber, U., and Bansch, R., *Rcthnerische Studie eur Thermo.
achockbelastung des Reaktordruckbehalters im Kernkraftwerk Obngheim," lith MPA Seminar on Sicherheit und Verfusbarkeit in der Anlagentechnik,1011 Oct.1985, Staathche Maienal.
prbfungsanstatt, Stutisert West Germany,1985.
l19) Loss. F. J., Oray, R. A., and Hamthorne, J. R. *$isni6cance of Warm Prutress to Crack initia, tion During Thermal Shock,* NRL/NUREG Repon 8165. Naval Research Laboratory. Wash.
ington, DC,1985.
l20) Hollstein, T, Blauci, J. G., Kiensier, R., and Nagel, G.,"Use of Load Path Dependent Material Fracture Toughness Values (WPS)in the Safety Analysis of RPV Stade,* Nuclear Engineersas and Design. Vol. 94,1986,pp.233-239.
(21) Kussmaul, K.,
- West German Research Program on irradiation E#eets on Reactor Pressure Ves.
sens "in Radsotic. EmbrstilementandSurveolianceofNuclear Rt : tor Pressure l'essels:
A n Inter.
nationalStudr. ASTM STP$19, L Steele, Ed., Amencen Society for Testing and Maienals, Phil.
i adelphia.1983, pp.16-28.
(JJJ Fohl, J., Leitz, C., and Anders, D., *lrradiation Experiments in the Testing Nuclear Power Plant V AK." in Efnts ofRadiation on Matertals (Eleventh Sympossum), ASTAISTP 762. H. R. Barger i
and J. S. Perrin, Eds., Amencan Society for Testing and Matenals. Philadelphia,1982 pp. 520-519.
[23) Efect of f rradiation, Corrosion, Temperai ere and Cyche leading
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L'0l, ou BIBLIOGRAPHIC DATA SHEET no.no is,,,,u,,a
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NRC TRANSLATION 2246
- i. tn ti aw suont it COi4PARISON OF Tile RESULTS OF CALCULATIONS BY 3
oArt airont rut.uswie SPECI ALISTS FROli Tile USA AND USSR OF Tile PR0k ABILITY
$'
- N 1
'taa OF FAILURE OF llDR-ilYPO CONTAINMENT DURING JANUARY 1990 ACCIDENTS INVOLVING T!!ERMAL S!!OCKS
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SC1'.'RAN COMPANY 1482 EAST VALLEY ROAD SANTA BARBARA, CALIFORMIA 93150 (805) 069-2413
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KURCilATOV INSTITUTE FOR ATOMIC ENERGY
- 10. LUPet tlet tal AfiY Noit $
- 11. Ah&1mAc1 uoo.,,as e, sul Soviet specialists calculated the probability of failure The of the llDR-ilYPO containment using the PVES program.
input data used by specialists from the USA are examined Data are along with the results of the calculations.
given on geometric dimensions of the llBR containment.
Data on mechanical and thermophysical properties of base metal and facing are presented.
The results of the calculations are discussed and compared, is...is
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WORKING GROUP 3.1
- RADIATION EMBRITTLEMENT
)
i l
OF THE HOUSING AND SUPPORT STRUCTURES 1
l i
i TOPIC:
3.1.4 VESSEL SURVEILLANCE PROGRAMS l
AND EMBRITTLEMENT REGULATIONS t
1 k
9:00 TO 10:15 WEDNESDAY -
JUNE 7,
1989 l
i j
SURVEILLANCE AND VESS'EL PROBABILITY OF FAILURE P.N. RANDALL AND H. W. WOODS, NRC i
OBJECTIVE:
TO DESCRIBE THE REGULATORY PROCESS USED
)
TO PREVENT FRACTURE OF THE REACTOR j
VESSEL BELTLINE AS A RESULT OF l
EMBRITTLEMENT BY NEUTRON RADIATION PNR1 i
f 6
m,.~..
-+
l OUTLINE -OF THE REGULATORY PROCESS A.
FOR NEW PLANTS, CONTROL MATERIAL COMPOSITION TO LIMIT SENSITIVITY TO EMBRITTLEMENT B.
MONITOR EMBRITTLEMENT BY A SURVEILLANCE PROGRAM C.
FOR NORMAL OPERATION - PROVIDE PRESSURE-TEMPERATURE LIMITS FOR HEATUP AND COOLDOWN i
I l
D.
FOR LOW TEMPERATURE OVERPRESSURE EVENTS -
PROVIDE LOW-SET-POINT RELIEF VALVES i
1 E.
FOR PRESSURIZED THERMAL SHOCK EVENTS -
i 4
LIMIT THE LEVEL OF EMBRITTLEMENT TO f
i LIMIT THE RISK
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' ASTM E-185 TEMPERATURE LIMITS THE MINIMUM NUMBER OF TEST SPECIMENS IS:
o MATERIAL CHARPY TENSION BASE METAL 12 3
WELD METAL 12 3
FRACTURE -TOUGHNESS SPECIMENS OF COMPACT, o
C(T), FORM ARE RECOMMENDED IF MATERIAL HAS LOW UPPER-SHELF TOUGHNESS ASTM E-1214 TEMPERATURE MONITORS o
ASTM E-844 o
DOSIMETRY PNR3
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' REQUIREMENTS -FOR CAPSULE WITHDRAWAL 5 CAPSULES ARE -REQUIRED IF THE PREDICTED o
0 END-OF-LIFE CHARPY SHIFT EXCEEDS 200 F
~
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FIRST 1.5 5 X 10 (1) 50 SECOND' 3
1/2 - E.O.L THIRD 6
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Hazardous to Vessel Integrity Rapid m
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Pump Seal Temperature, deg F
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AS EMBRITTLEMENT ENLARGES THE HAZARDOUS REGION, THE P-T LIMIT MUST BE MOVED UP THE TEMPERATURE SCALE Hazardous to Vessel integrity Rapid \\ _
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4ARTNor + Margin RT o7 = Initial RTNor w
ART is the. Measure of Radiation Embrittlement ug7 ART is the Charpy Shift, Measured at the 30 ft Ib Level yg7 O.
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ORNL-DWG 85-5356 ETD 220
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TEMPER ATURE RELATIVE TO NDT
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CHARPY--V NOTCH IMPACT ENERGY FOR THE POINT BEACH UNIT NO.1 PRESSURE VESSEL WELD METAL Temperature ( C)
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FROM POSITIOnl C 1.1 OF REVISION 2 TO REGULATORYL GUIDE 1.99 r
ARTyg7= M WMAWF CF = Chemistry Factor (Copper and Nickeli l
- Table I for Welds
- Table ll for Base Mefal (Plates and Forgings;'
i F = Fluence,10 n/cm2 (E > 1 MeV;L at the Tip of 18 the Crack f
4 I
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3 EFFECT OF COPPER AND NICKEL CONTENT ON THE REFERENCE TEMPERATURE, deg. F, FOR WELDS IRRADIATED TO 1x1018 2
n/m (E > 1 MeV) 4 i
Nickel, Wt-%
Copper,. Wt-%
0.20 0.60 1.00 0.05 49 68 68 O.10 65 122 135 0.15 84 146 191 0.20
-104 160 223 0.25 126 176 243 l
0.30 146 194 257 l
0.35 168 212 272 0.40 189 231 288 i
O
COMPARISON OF CHEMISTRY FACTORS FOR WELDS AND BASE MsTAL, GIVEN IN REVISION 2 TO REGULATORY GUIDE 1.99 i.2 1
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i l
RESEARCH 'RESULTS NOT --YET QUANTIFIED l
FOR' REGULATORY USE i
j EFFECTS fOF PHOSPHORUS (AND OTHER ELEMENTS) o 4
o IRRADIATION TEMPERATURE
~
o FLUENCE RATE s
o-NEUTRON ENERGY SPECTRUM o
REFINEMENT OF THE FLUE CE FUNCTION AT LOW FLUENCES o
STRESS RELIEF HEAT TREATMENT. TIME AND TEMPERATURE runs
i OUTLINE OF THE REGULATORY PROCESS' A.
FOR NEW PLANTS, CONTROL MATERIAL COMPOSITION YO LIMIT SENSITIVITY ' TO EMBRITTLEMENT I
l l-B.
MONITOR EMBRITTLEMENT BY A SURVEILLANCE PROGRAM i
l i
l C.
FOR NORMAL OPERATION - FROVIDE PRESSURE-TEMPERATURE LIMITS FOR HEATUP AND COOLDOWN D.
FOR LOW TEMPERATURE OVERPRESSURE EVENTS -
PROVIDE LOW-SET-POINT RELIEF VALVES E.
FOR PRESSURIZED THERMAL SHOCK EVENTS -
l LIMIT THE LEVEL OF EMBRITTLEMENT TO LIMIT THE RISK PNR2 i
O l
s
?
l l
Working Group 3.* Radiation Embrittlement of the Housing and Support Structures l
- andAnnealingoftheHousings
[
l Transmitted before the meetings.
i 3-1 Radiation Embrittlement of the Materials of VVER-440 Reactor Vessels from
?
Results of Control Sample Testing - A. D. Amayev, A. M. Kruykov, and M. A. Sokolov, Kurchatov Institut<< of Atomic Energy, Moscow.
(NRC Translation 2092) 3-2 Annealing Restoration of Mecharical Properties of Irradiated Materials of the VVER-440 Reactor Vessel - 4. D. Amayev, A. M. Kruykov, E. P. Ryazantsev, and M. A. Sokolov, Kurchatov Institute of Atomic Energy; and V. I. Badanin,
)
V. A. Ignatov, and V. A. Nikolaev, Central Scientific-Research Institute for Construction Materials, Leningrad.
(NRCTranslation2093) 3-3 Influence of Fast Neutron Flux Density on Radiation Embrittlement of VVER-440 Reactor Vessel Materials - A. D. Amayev, V. I. Vikhrov, A. M. Kryukov, and M. A. Sokolov, Kurchatov Institute of Atomic Energy, Moscow.
(NRC Transl6 tion 2094) 3-4 Scientific-Engineering Aspects Regarding the Process of Annealing of the-VVER-440 Reactor Vessels and the Creation of Equipment for this Purpose -
S. K. Morozov, All-Union Research Institute for Nuclear Power Plant Operations, Moscow.
(NRC Translation 2131) 3-5 Mechanisms Determining the Dependence of Radiation Embrittlement of the Vessel Steel on Content of Its L e ical Elements - V. A. Nikolayev and-V. V. Rybin, Prometey Complex Machineiy Institute.
(NRCTranslation2132) 3-6 Quantitative-Probabilistic Analysis od the Danger of Failure of Nuclear Reactor Structural Elements as a Mean; of Evaluating the Lifetime vi l
These Elements Under Conditions of In:omplete and indeterminate Initial Data on the' Condition of the Facilit) - A. A. Tutnov and V. V. Tkachev.
Kurchatov Institute of Atomic Energy Moscow.
(NRC Translation 2133)
-l 3-7 Calculation' of the Probability of Onset of Brittle Failure-of Vessels Under Pressure - A. A. Tutpov-and V. V. Tkachev, Kurchatov Institute of Atomic Ener'gy, Moscow.
(NRC Translation 2134)
=.
. Transmitted at the meetings.
3.8 Research on the Characteristics of the Weld Me al from the Reactor Vessel of Novovoronezh Nuclear Power Station, Unit 1, 'fter 20 Years of Operation, Step 1 - A. D. Amaev, Kurchatov Insti'ute of Atomic Energy, Moscow (NRC Translation 2152) 3.8-1 Study of Possible Modeling of Radiation Embrittlement of Steels for VVER Vessels - L. A. Vayner and Yu. I. Zvezdin, Soviet Atomic Energy, Vol.
66, No. 2, February 1989, pp. 86-88 (in Russian).
(NRC Translation 2223) 4
m 1, q w
3.8-2 Resistance to Brittle Failure of Austenite-Ferrite Steel Fused on to Steel 15Kh2MFA - V.-l. Badanin, V. A. Ignatov, V. A. Nikolayev, V. V.
Rybin and B. T.-Timofeyev, USSR Academy of Sciences, Moscow, Automatic Welding, 1989, No. 3, pp. 4-7 (in Russian).
(NRCTranslation2225) 3.8-3 Reduction of Fracture Toughness of a Low-Alloy Steel During Irradiation
- V. A. Nikolaev, V. F. Vinodurov, A. M. Morozov, V. S. Panteleev, and V. K. Shamardin, Soviet Atomic Energy, Vol. 62, No. 5, May 1987, pp. 400-
- 403, 3.8-4 Effect of Neutron Irradiation and Corrosive Medium on the Crack Resistance of the Water-Moderated Power Reactor Vessels - L. A. Vainer, Soviet Atomic Energy, Vol. 62, No. 5, May 1987, pp. 404-407.
3.8-5 Solidus Criteria for the Binding Energy of Vacancies with Impurity Atoms
- V. A. Nikolaev, Journal of Technical Physics, Vol. 56, No. 4,1986, pp.776-778-(inRussian).
(NRCTranslation'2194) 3.8-6 Effect of Defect Annealing of Hydrogen Absorption in Neutron Irradiated Iron Alloys - V. A. Nikolayev and S. V. Shapovalov, USSR Academy of Sciences, Moscow.
Physics of Metals and Metallography, Vol. 61, April 1986,pp.822-824(inRussian).
(NRC Translation 2224) 3.8-7' Damage Summation in Annealing and Repeated' Irradiation of Pressure-Vessel Steel - V. A. Nikolaev, V. I. Badanin, and A. M. Morozov, Soviet Atomic Energy, Vol. 57, No. 3. Sept.1984, pp. 603-606, 3.8-8 Effect of Chemical Composition and Annealing Conditions on the Radiation Embrittlement of the Metal of Low-Alloy Welded Seams - V. A. Nikolaev, A. M. Morozov, V. I. Badanin, A. S. Teshchenko, and R. P. Vinogradov, Soviet Atomic Energy, Vol. 57, No. 3, Sept. 1984, pp. 606-613, 3
9 5