ML20035E049

Summary of 920521 Joint ACRS Subcommittee Meeting on Matls & Metallurgy & Advanced Reactor Designs in San Francisco,Ca Re Application of High Temp Structural Matls in Advanced Liquid Metal Reactor

ML20035E049
Person / Time
Issue date:	07/16/1992
From:	Shewmon P, Ward D Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2822, NUDOCS 9304140215
Download: ML20035E049 (58)

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CERTlHED CERTIFIED BY P. SHEWMON 7/16/92 D. WARD 6/26/92 l

l SL7S.ARY/ MINUTES OF THE JOINT ACRS SUBCOMMITTEE MEETING ON MATERIALS AND METALLURGY AND ADVANCED REACTOR DESIGNS MAY 21, 1992 SAN FRANCISCO, CALIFORNIA l

PURPOSE The ACRS Subcommittees on Materials and Metallurgy and Advanced Reactor Designs held a joint meeting at the Holiday Inn-Financial j

District, San Francisco, Calif ornia on May 21, 1992. The purpose of this meeting was to discuss the application of high temperature structural materials in the Advanced Liquid Metal Reactor (ALMR).

A copy of the meeting agenda and selected slides from the presentations are attached. The meeting began at 8:30 am and adjourned at 4:25 pm and was held entirely in open session. No written comments or requests for time to make oral statements were received f rom members of the public. The attendees were as follows:

ATTENDEES ACRS NRC STAFF i

P.

Shewmon, Co-Chairman S. Sands, NRR D. Ward, Co-Chairman l

I.

Catton, Member GE-SAN JOSE i

l W.

Kerr, Member l

T.

Kress, Merber G. Gyorey l

H.

Lewis, Merber D. Hardy l

C. Wylie, Member C. Cockey 1

F.

Snow, Consultant L. Salerno I

D.

Houston, Cognizant Staff M.

Patel Engineer W. Kwant l

DOE ORNL N. Grossman J. White i

S.

El-Saifwany J. Corum l

C. Brinkman P.

Shen F.

Huang D.

Little l

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l Mat & Met /ARD Minutes May 21, 1992 DISCUSSION l

In his opening remarks, Dr. Shewmon discussed some of the areas of concern with the metal-cooled reactors. Specifically, he noted that these reactors operate at high enough temperatures so that creep can become a problem. He also noted that with liquid metal coolant, i

one could expect fatigue from oscillations in the flow.

He indicated that these design problems are not found in the LWRs which operate at somewhat lower temperatures.

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Introduction Dr. G. Gyorey, Manager of Safety and Licensing at GE for the ALMR, presented an overview of the ALMR Program and noted that the GE design is in the conceptual stage and the NRC review is in the pre-application stage. He indicated that the near-term goal was to receive a Staff SER in November 1992.

I Overview of Reactor Desian Mr.

W.

Kwant (GE) presented a description of the reactor system, focusing on the main features and the arrangement. The key features l

included: Power Block arrangement for a plant, Main Power System, Reactor Module, Seismic Isolation System, Reactor Vessel Auxiliary 7

Cooling System (RVACS),and Instrumentation.

The main metallic components (structural materials) were tM containment vessel, reactor vessel, upper internal system (UIS), EM pumps. intermediate t

heat exchangers (IHX), core support and closure head. He also discussed the reactor temperatures and pressures (both for normal operation and f or transients), structural radiation exposures, and the expected duty cycles for design basis and beyond design basis events. In closing, he presented a summary of the world operating experience for sodium cooled fast reactors.

Dr. Shewmon asked about heat removal from the reactor in case the steam generators fail. Mr. Kwant indicated that a second decay heat removal system, an air cooled jacket around the steam generator, would remove the heat.

i Dr. Catton asked about cooling of the vessel in the case of a core i

melt. Mr. Kwant indicated that GE had analyzed this case using certain assumptions about natural circulation and that the structural materials in the primary boundary maintain their integrity.

He further indicated that the analysis has been submitted in an amendment to the ALMR PSID.

In response to-other questions, GE personnel indicated that: the volume of the containment dome is about 40,000 cu. ft.; there are monitoring devices for fuel rod leaNrs and containment vessel leakage; the core is very heterogenous; and the control material is boron carbide.

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o-f Mat & Met /ARD Minutes May 21, 1992 1

ASME Code Aeolication in ALMRs Dr. J.

Corum (ORNL) discussed the ASME High Temperature Code, its l'

background and related development programs.

Specifically, he discussed the following areas:

o Elevated temperature failure modes o Genesis of code case and supporting technology o Overview of Code Case N-47-29 o Supporting high-temperature structural design methodology development and validation.

He indicated that Code Case N-47 was the early design criteria document for the FFTF and that significant improvements have been developed over the past 10 years. Many of the improvements had been made to resolve NRC concerns as expressed in the evaluation of the Clinch River Breeder Reactor (CRBR).

)

Reactor Structural Evaluation

~

Dr. M.

Patel (GE) discussed the reactor structural evaluation. He addressed the duty cycle envelope (events, seismic load and l

temperature),

design performance (evaluation

basis, elevated temperature rules, creep damage, fatigue, flow-induced vibrations, containment performance with reactor vessel leak, and thermal striping) and environmental ef fects (thermal ageing, liquid sodium and irradiation). He indicated that the ALMR evaluation is based on the ASME Code Section III, Subsection NB and Code Case N-47, Nuclear Systems Materials Handbook, and FFTF and CRBR documents.

i The ASME code data was extrapolated for the evaluation of 60 years operation for 304SS and 316SS and for the short time operation of j

Cr-2:o above 1200F. He noted that programs are planned to develop the database to confirm the ASME Code extrapolations. In closing, he indicated that due to the relatively low operating temperatures, mild thermal transients and small seismic loads, the ALMR has large

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design margins to accommodate uncertainties in the extrapolations.

In response to a question by Dr. Snow, Dr. Patel indicated that ASME Section III, Class I design rules apply for all non-removable components, the IHX and the primary sodium boundary.

Dr.

Catton asked whether GE had considered the usual fluid I

structural type interaction, where there is feedback between the fluid and the structure.

Dr.

Patel indicated that they had considered the case of vortex shedding but had not considered the case that was of concern to Dr.

Catton.

He indicated that verification tests would be performed in this area.

Dr.

Shewmon and Dr.

Catton questioned the source of helium l

generated in stainless steel during irradiation.

Dr.

Shewmon indicated that the helium was more from nickel than boron, and that in a fast flux, the boron would quickly burnout. Dr. Little (WHC) seemed to think that most of the helium was from boron.

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e Mat & Met /ARD Minutes May 21, 1992 F.eactor In-Service Inspection Mr. Kwant (GE) discussed the requirements for in-service inspection that come from ASME Code Section XI, Division 3,

and how these requirements will be met for examination, testing and inspection of components and systems in the ALMR. For each structure / component, he indicated the examination method and the extent and f requency of examination.

Radiation Effects on Material Ductility Dr.

F.

Huang (WHC) indicated that the ASME codes do not consider the effects of irradiation damage. He discussed the effects of irradiation in terms of

swelling, in-reactor creep and embrittlement and how exposure limits are determined. He indicated that material surveillance studies validated the component design for FFTF.

Desian for 60 Year Life Dr. C. Brinkman (ORNL) discussed the structural materials problems associated with design allowables for a

60 year life.

To substantiate this design life for ALMRs, appropriate creep rates, ductilities and rupture strengths need to be developed. In response to a question about accelerated tests, he indicated that none of these low strain rate tests can be accelerated and design values are obtained from the extrapolation of current data.

Hich Temnerature Acolication of.Cr-Mo Steel Dr. Brinkman discussed the short term, high temperature application of Cr-Mo steel in the reactor vessel auxiliary coolant system (RVACS). ASME Code Case N-47 provides stress allowables up to a temperature of 1200F for this material. In the ALMR, this component could be exposed to temperatures up to 1250F for several days or up to 1300F for shorter periods.

Sodium Effects on Structural Materials Dr.

Brinkman summarized the experience over 40 years in using liquid sodium as a heat transfer media. He indicated that corrosion in liquid metal systems is complex and he discussed the variables that affect corrosion. He noted that low corrosion rates are observed for flowing sodium in stainless steel systems and that no design penalties are required for the use of austenitic stu nless steels in sodium when the chemistry of the cover gas and sodium is controlled (typically holding oxygen below 10 ppm).

In response to a question from Dr. Shewmon, Dr. Brinkman indicated that there were not a lot of long term world-wide tests being conducted on 304SS or 316SS. Most of the past tests have been run to 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

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t Mat & Met /ARD Minutes May 21, 1992 Thermal Stricino Data and Criteria Dr. Corum (ORNL) discussed the thermal striping process that occurs from the random fluctuations of hot and cold fluids near the metal component. He described the component thermal and stress analysis, fatigue evaluation and tests performed in support of Clinch River.

l He indicated that the following issues require resolution:

-( 1 )

crack initiation design guidance for 316SS, (2) design guidance for Alloy 718, (3) experimentally-based interaction rule for treating interspersed thermal transient loads and (4) thermal striping crack propagation / crack arrest assessment procedures.

Closina Remarks Mr. Kwant (GE) summarized the presentations as indicating that the GE reactor design, with its combination of low temperatures and l

other factors, has a significant safety margin and has the capability for accident prevention and mitigation.

He also indicated that the design has the ability to contain a molten core and the energetics from hypothetical core disruptive accidents. He noted the three main areas where further development was needed:

(1) for 60 year life, (2) thermal striping criteria, and (3) extension of the ASME Code for slightly higher transient temperatures.

Dr. Shewmon thanked the group of presenters for a very interesting and informative day. He indicated that another meeting would be l

held in the near future.

i FUTURE ACRS ACTION The joint Subcommittees will schedule further meetings to review this matter as the staff's review of the ALMR submittals progress.

ACTIONS. AGREEMENTS AND COMMITMENTS No items of this nature were specifically identified during the meeting.

DOCUMENTS There were no specific documents provided for review at this meeting. The natters discussed at this meeting pertain to the i

licensing review of the GE ALMR. The GE ALMR design has been submitted previously in the following document:

o General Electric / Nuclear Systems Technology Operation (DOE Contract), GEFR-00793, " PRISM Preliminary Safety Information Document," Volumes I through VI, November 1986 Footnote: Any reference above to Cr-Mo steel is for a composition of 224 Cr-1 Mo steel.

Mat & Met /ARD Minutes May 21, 1992 NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 2120 L Street, NW, Washington, DC 20006, (202) 634-3273 or can be purchased from Ann Riley and Associates, LTD.,

1612 K Street, NW, Suite 300, Washington, DC 20006, (202) 293-3950.

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i JOINT ACRS SUBCOMMITTEE MEETING MATERIALS AND METALLURGY / ADVANCED REACTOR DESIGNS MAY 21, 1992 SAN FRANCISCO, CALIFORNIA OF THE HOLIDAY INN FINANCIAL DISTRICT l

750 KEARNY STREET I

SAN FRANCISCO, CALIFORNIA l

- TENTATIVE AGENDA -

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1 Accrox. Time I.

Chairman's Opening Statements and 8:30-8:40 a.m.

Comments - Shewmon II.

GE Presentation 1.

Introduction 8:40-8:55 a.m.

2. ALMR Reactor System 8:55-9:40 a.m.

Design e

e Materials t

Temperatura, Flux / Fluence, e

Environment (sodium / air) 4

. Duty CycJes e

Beyond Dessign Basis Events e

1

• * *

• BREAK * * *

• 9:40-9:55 a.m.

i i

3. ASME Boiler and Pressure Vessel Code 9:55-10:40 a.m.

Design Rules, Analysis Methods and Test Support i

4. ASME Code Application to ALMR Design 10:40-11:40 a.m.

Failure Modes, Design Rules and e

Limits j

Service Levels and Damage Levels Code Material Requirements /

e Qualification 1

Designer Responsibility e

t

• * *

• LUNCH * * *

• l 11:40-12:40 p.m.

5. AIMR Structural Evaluation 12:40-1:40 p.m.

Design Criteria - Structural l

Limits, Functional Limits, Environmental Effects Duty Cycle Envelopes e

Analyses Methods / Qualification e

of Methods Service Level Design Temperature e

Limits i

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t

- TENTATIVE AGENDA (cont'd.)

j i

ADorox. Time-60-Year Life, Sodium Effects e

(including impurities),

Radiation Effects, 1200*F/1500*

Extension for BDB Events t

6. ALMR In-Service Inspection 1:40-2:25 p.m.

ASME Code Requirements e

ALMR Approach e

e Code Status

• * *

• BREAK * * *

• 2:25-2:40 p.m.

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7.

Technology Development 2:40-3:40 p.m.

Program Overview f

e 60-Year Life Thermal Striping e

Short-Term High Temperature Exposure e

e Sodium Effects (including impurities) e Radiation Effects.

8.

Summary and Conclusion 3:40-4:10 p.m.

III.

Subcommittee Discussion 4:10-4:30 p.m.

Subcommittee Report to Full ACRS e

Consultants Comments Adjournment e

4:30 p.m.

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ADVANCED LIQUID METAL REACTOR (ALMR)

STRUCTURAL MATERIALS For Presentation to the I

ADVISORY COMMITTEE ON REA CTOR SAFEGUARDS MAY21,1992 San Francisco, California WK2 05-11-92-A-01

e Contents E.esenter Affiliation Ippig r

traducti:n Dr. Gcza L Gyorey GeneralElectric Company ractorSystem Description WalterKwant GeneralElectric Company SMEBoilerandPressure VesselCodo and Application to ALMRDesign Dr. James M. Corum Oak Ridge Nationallaboratory ractorStructuralEvaluation Dr. Mahadeo R. Patel General Electric Company ractorIn-Service laspection WalterKwant GeneralElectric Company rchnology Status andDevelopment Program Overview WalterKwant GeneralElectric Company ediation Effects on Material Ductility Dr. Frank H. Huang Westinghouse Hanford Company evelopment of Sixty YearDesign Allowobles and Mechanical Proporties Dr. Charles R. Brinkman Oak Ridge Nationallaboratory tort T:rm. High Temperatura Exposure of21/4 Cr-1Mo Steel Dr. Chsries R. Brinkman Oak Ridge Nationallaboratory sdium Effects on Structural Materials Dr. Charlos R. Brinkman Oak Ridge Nationallaboratory rermalStriping Data and Cdteria Dr. James M. Corum Dak Ridge Nationallaboratory immary and Conclusions WalterKwant GeneralElectric Company WK 2-05-11 92-A-02 i

Introduction by Dr. Geza L Gyorey, Manager Safety andLicensing GeneralElectric Company San Jose, California I

WK245-1192-A-03

' ALMR Regulatory Review

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US ALMR Program Overview:

O National Program, Sponsored by the DOE O Broad Participation by Industry and National Laboratories O in the Advanced Conceptual Design Phase O Compact Modular Reactor Plant Design Based on the GE PRISM Concept (Power Reactor - Innovative Small Module)

O Metal Fuel Cycle, Based on the Integral Fast Reactor (IFR) Concept Developed by Argonne National Laboratory 0 High level of passive safety characteristics O Near Term Goal is Preapplication Safety Evaluation Report; NRC Schedule is November 1992 O Long Term Goal is Standard Design Certification by about 2005 Continuing interaction with NRC is Essential l Gyotey 5/21/92

ALMR Regulatory Review Objectives of this Meeting:

O Continue ACRS review of the ALMR, leading to a Preapplication

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Safety Evaluation Report (PSER)

O Follow-up to the comprehen'sive ALMR update meeting with the ACRS on August 6,1991 0 Discuss ALMR structural materials. environment. and ASME Code applications O Recognize decades of experience with most of these materials in liquid metal cooled reactors worldwide Gyorey 5/2M2

Reactor System Description by Walter Kwant, Manager Reactor Engineering i

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GeneralElectric Company San Jose, California

ALMR Design Data e OverallPlant

-NumberofReactorsperPowerBlock Three

- NanberofPowerBlocks One/Two/or Three

- NetElectricalOutput 465/930/or 1395MWe

- NetStation EWiciency 329 %

- Turbine Throttle Conditions 955 psia /540*F(sat'd) e ReactorModule

- ThermalPower 411 MWt

- PrimarySodiumInlet/ Outlet Temp.

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Design Basis Events ASME Code Service level A Events (Normal Operation)

Startups, Shutdowns and Load Following Steady State Temp. and Flow Fluctuations Power Block Operation with One or Two Reactors Out of Service 60 Years of Power Operation at 85% A vailaiblity ASME Code Service level B Events (P>10)

Reactor Trips from Partial or Full Power Uncontrolled Rod lasertion or Withdrawal Loss ofPrimaryorlatermediate Pump ExpectedEvents e

Trip ofOne FeedwaterPump y,,,pgygg,g,gg,,,,,g,y,,

Steam GeneratorBlowdown A and 8 StructuralLimits Turbine Trip Maintain Design life e

Loss ofAllOffsite Power Fast Runbacks from Various Power Levels DBE Earthquake up to 0.15g Peak Ground Acceleration

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ASME Code Service level C Events (10 > P > 16 )

Iass of Flow in Primary or Intermediate System 1

Loss of Feedwater Through Various Equipment Failures Loss ofIntermediate Sodium Due to Rupture. Disc Failure Water Side Isolation of Steam Generator With Dump Valve Failure UncontrolledRod Withdrawal Steam Generator Tube Failure ($_3 Tubes)

SmallIHXLeak RVACS Only Decay Heat Removal Due to loss oflatermediate System

- Infrequent Unexpected Events - Total of 4 Per Plant Life

- Maintain Design Life WK2-4 29 N N

Design Basis Events (Cont.)

ASME Code Service level D Events (P < 10^ )

Safe Shutdown Earthquate up to - 0.3g for Licensing Basis

- 0.5g for Design Basis FeedwaterLine Rupture Steam Line Rupture Intermediate Sodium Line Rupture Rupture in High Pressure Primary Circuit Steam Gerrator Tube Rupture Primary 57',m Sodium Boundary Leak R!fACS (With Fouled Surfaces) Only Decay Heat Removal Due to loss ofIHTS

- Totalof1 Occurence Per Plant Life

- Maintain Safe Shutdown

- Maintain Pressure Boundary and Core Cooling W1C2-4 29 S2 01

Beyond Design Basis Events 4

ASME Code Service Level D Events (P < 10 ^)

A TWS Events; Unprotected loss of Flow, Loss of Normal Heat Sink, or All-Rod Transient Overpower Decay Heat Removal with Partially Blocked RVA CS Earthquake Greater Than 0.5g But less Than 1.0g Peak Ground Acceleration (MarginAssessment)

Hypothetical Core Disruptive Accident (HCDA)

For Safety Performance, Capability and Margin Assessments

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World Sodium Cooled Fast Reactor Operating Experience i

Year Sodium Opera-Reactor Reactor Opera-Thermal Inlet /0utlet Time flow Rate Materials Name Purpose tional Power (MWt)

Temff)

(Yrs)

(GPM)

Hotleg Coldleg jnANCE Rapsodia test 1%1 40 95V132 25 5.440 316SS 316SS

%enix prototype 1974 560 104W125 16 60.000 316SS 316SS

R

:ia demonstration 1905 3000 1001/143' 3

314.000 316SS 316SS

& PAN layo test 1978 100 937J6%

10 12000 304SS M4SS tonje prototype 1%3 114 M4rl47 85.000 304SS 304SS l

P JR test 1%3 12 66?J446 14 9.000 18M/ISS IWB/ISS CR prototype 1916 60 104W150 16 61.800 321SS 321SS L

GR-2 test 1%4 62.5 88Y100 28 10.000 304SS 304SS

'nrico Fermi test 1%S 200 80 % 50 7

21700 304SS 304SS FTF test 1900 400 937Ai80 12 43.600 316SS 316SS blMR prototype 471 905%40 46.000 316SS 304f316SS FSSR t0R-60 test 1%9 60 101.Mi26 23 5.400 ICr18Ni9SS ICr18N19SS RN-350 prototype 1973 1000 932572 19 79.000 Cr18Ni9SSI ICr18Ni9SS IN-600 prototype 1980 1410 1027/116 12 13l000 Cr18Ni9SS Cr18Ni9SS sERMANY TNK test 1972 56 9175 80 20 5.600 1.6710SS 1.6170SS WKJE46-SI-Ot

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t ASMEBoilerandPressure VesselCode and Application to the AlMR Design by Dr. James M. Corum, Section Head StructuralMechanics Engineering Technology Division Oak Ridge NationalLaboratory Oak Ridge, Tennessee WK2-05-11 92. A-04

SIGNIFICANT IMPROVEMENTS IN N-47 HAVE BEEN DEVELOPED'IN THE PAST 10 YEARS (SINCE CRBRP SER)

ADDRESSES ITEMS NOW IN CASE N-47-29:

NRC CONCERN o RECOMMENDED RESTRICTED MATERIAL SPECIFICATIONS X

FOR 304 AND 316 SS TO REDUCE PROPERTY SCATTER o WELDMENT STRENGTH REDUCTION FACTORS X

o NEW CREEP-FATIGUE EVALUATION RULES SIMPLIFIED ELASTIC RULES X

EQUIVALENT STRESS, o., FOR CREEP RUPTURE X

REDUCED K' FROM 0.9 TO 0.67 X

o NEW GENERALLY APPLICABLE SIMPLIFIED INELASTIC ANALYSIS METHOD FOR SATISFYING STRAIN LIMITS o RECLASSIFICATION OF PRESSURE-INDUCED BENDING STRESSES AS PRIMARY IN SIMPLIFIED RULES i

ITEMS PENDING:

o HOT ACCEPTANCE TEST TO WEED OUT WEAK HEATS OF X

i 304 AND 316 SS o REDUCED ALLOWABLE STRESSES ACCOUNTING FOn EFFECTS OF LONG-TERM PRIOR SERVICE M

HTSD DEVEL.OPMENT PROGRAM APPROACH mnae7)c memx emume ANA W B38 CM]71M]A E3DCIDUMIB EI7MDD8 EXPLORATORY EXPLORATORY MATERIAL MATERIAL DEFORMATION FAILURE TESTS TESTS M

M I f i f FORMULATION FORMULATION DEVELOPMENT OF ANALYSIS OF FAILURE OF SIMPLIFIED METHODS CRITERIA PROCEDURES M

M M

1 1 f 1 I 1 f 1

DESIGN COMPARISONS PROPOSED WITH RIGOROUS GUIDELINES CODE METHOD DOCUMENT RULES (F95T)

RESULTS 1 f LIDATED CONFIRMATORY SIMPLIFIED STRUCTURAL PROCEDURES TESTS 1 f VALIDATED.

j RIGOROUS DESIGN METHODOLOGY Y -

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IN CONCLUSION...

e MATERIALS AND STRUCTURES TECHNOLOGY WIDELY RECOGNIZED AS ONE OF MOST SIGNIFICANT ACCOMPLISHMENTS OF U.S. LMR i

PROGRAM j

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CASE N-47 IS WELL DEVELOPED AND USED WORLDWIDE e

SEVERAL MODIFICATIONS OF PAST 10 YEARS ADDRESSED NRC CONCERNS EXPRESSED IN CRBRP SER e

FEW REMAINING UNCERTAINTIES ARE COVERED-1 BY LARGE MARGINS IN N-47 AND CONSERVATIVE-OPERATING CONDITIONS OF ALMR

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e N-47 ALLOWABLES MUST BE EXTENDED TO 60 YEARS FOR ALMR i

O liaactor StructuralEvaluation by Dr. Mahadeo R. Patel, Principal Engineer StructuralMechanics GeneralElectric Company San Jose, California WK2-0511-92-A-05

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L REACTOR STRUCTURAL EVALUATION o

Duty Cycle Envelope o

Design Performance o

Environmental Effects m

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Design Performance Evaluation Basis Reactor Structures Containment Vessel High-Cycle Fatigue Loads k

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Environmental Effects Thermal Aging Liquid Sodium Irradiation Mill' 5-92 21

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The ALMR Design Evaluation is Based on the ASME Code Section lit, o

l Subsection NB and Code Case N47, Nuclear Systems Materials Handbook, and FFTFand CRBR Documents, and Extrapolation of the ASME Code Data to:

i 60 Years Operation for SS304 and SS316.

Short Time Operation of 21/4 Cr-1 Mo Above 1200 F.

Program is Planned to Develop the Database Required to Confirm the o

ASME Code Extrapolations in the Design Evaluation.

The Relatively Low Operating Temperatures, Mild Thermal Transients, o

and Small Seismic Loads Provide the ALMR with Large Design Margins to Accommodate Uncertainties in the Data Extrapolations.

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Reactorin-Service Inspection by Walter Kwant, Manager Reactor Engineering I

GeneralElectric Company San Jose, California a

WK2 05-1192 A-06

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VME Scction XI, 0ivision 3 Inspection Requirernents for liquid Metal Cooled Reactors Examination Extent andFrequency Structura/ Component Method Examination

~33% of All Welds VisualInspection for cter V:ssel Welds Presence of Sodium Every3 to 4 Years Continuous Monitoring Continuous Monitoring for Sodium leakage of Entire Vessel Continuous Monitoring Monitoring forRadio-ctorCoverStructural ulSe:I Welds active Gas Leakage of All Welds Continuous Monitoring Monitoring forRadio-eterCoverNon-Welded als active Gas leakage ofAllSeals Continuous Monitoring of Monitoring forRadio-trolRedDrivesand herAbove-Cover active Gas Leakage allAbove-Cover pchanisms Contain-Mechanisms p CoverGas 100% ofAll Structures VisualInspection pony Weldedhternal openenen - Core for PositionalRelation-Approximately Every pport Someture, ships.

3to 4 Years revCucts, Plenums d ThermalBarriers VisualInspection 25% of All Wolds nainment Vessel Welds forGeneralMechanical Approximately andStructuralConditions Every 10 Years WW2-4-30-s?-04

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Development of Sixty Year Design Allowables andMechanicalProperties by Dr. Charles R. Brinkman, Group Leader MechanicalProperties Engineering Technology Division

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L LONG TERM TIME DEPENDENT STRESS ALLOWABLES IN ASME CODE CASE N-47 WERE DEVELOPED FROM THE LOWEST OF THE FOLLOWING MECHANICAL PROPERTIES 67% of the minimum stress to cause rupture in time t, 80% of the minimum stress to cause onset of tertiary creep in time t, and 100% of the minimum stress to cause accumulation of 1% total strain in time t.

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OF DATA NEEDS SEVENTEEN LONG TERM CREEP-RUPTURE TESTS ARE ONGOING ON TYPES 304 AND 316 STAINLESS, AND SHOULD BE CONTINUED UNTIL TEST TIMES OF APPROXIMATELY 150,000H ARE ACHIEVED.

LONG TERM CREEP-RUPTURE TESTS ON APPROPRIATE WELD -METALS FOR TYPES 304 AND 316 STAINLESS STEELS SHOULD BE INITIATED TO DEFINE SIXTY YEAR WELDMENT STRENGTH REDUCTION FACTORS.

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by Dr. James M. Corum, Section Head StructuralMechanics Engineering Technology Division I

Oak Ridge NationalLaboratory Oak Ridge, Tennessee WK2-05-11-92-A-09

t ASSESSMENI OF THERMAL STRIPING REQUIRES l

SEVERAL STEPS l

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- CR ACK PROPACATION EVALUATION

- FAILURE ANALYSIS I

i THERMAL STRIPING DEVELOPMENT NEEDS:

PROVIDE CONSENSUS CRACK INITIATION DESIGN e

GUIDANCE FOR 316 SS BASED ON EXISTING-i INFORMATION r

1 DEVELOP DESIGN GUIDANCE FOR ALLOY 718 e

l DEVELOP EXPERIMENTALLY-BASED INTERACTION e

RULE FOR TREATING INTERSPERSED THERMAL TRANSIENT LOADINGS l

DEVELOP THERMAL STRIPING CRACK PROPAGA-e TION / CRACK ARREST ASSESSMENT PROCEDURES (BASED ON EPRl/CRIEPl/NE HIGH-TEMPERATURE FLAW ASSESSMENT GUIDE FOR 304 SS) ll 4

Sodium Effects on Structural Materials 4

by Dr. Charles R. Brinkman, Group leader MechanicalProperties Engineering Technology Division Oak Ridge NationalLaboratory Oak Ridge, Tennessee WK245-1192 A-11 l

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VARIABLES AFFECTING CORROSION OF MATERIALS IN SODIUM COMPOSITION AND THERMO-MECHANICAL PROCESSING HISTORY

-TEMPERATURE-OXYGEN CONTENT OF SODIUM MATERIAL POSITION IN RELATION TO HEAT INPUT ZONE SODIUM AXIAL HEATING RATE MECHANICAL STRESS OF MATERIAL SODIUM VELOCITY-LOOP TEMPERATURE DIFFERENTIAL, DT EXPOSURE TIME DISSIMILAR MATERIALS IN CONTACT WITH SODIUM e

. CARBON CONTENT OF MATERIALS AND SODIUM

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EXPERIENCE WITH AUSTENITIC MATERIALS IN SODIUM COLD TRAPS AND PLUGGING METERS GIVE OXYGEN LEVELS TYPICALLY BELOW 10 ppm o

THE INFLUENCE OF CORROSION, CARBURlZATION, DECARBURIZATION, FORMATION OF o

FERRITIC SUB-SURFACE LAYERS AND SENSITIZATION ARE NOT IMPORTANT FOR COMPONENTS WITH THICKNESS GREATER THAN 2 mm THE EFFECT OF DECARBURIZING SODIUM ON THE CREEP-RUPTURE BEHAVIOR OF TYPE 304 o

STAINLESS STEEL CAN BE IGNORED FOR THICK-WALLED COMPONENTS. HOWEVER, IT HAS TO BE TAKEN INTO ACCOUNT FOR THIN WALLED COMPONENTS. THE CARBON POTENTIAL OF THE SODIUM SHOULD BE EVALUATED FOR THIS CASE.

SINCE CRACKING IN SERVICE IS OFTEN ASSOCIATED WITH WELDMENTS, HIGH STANDARD o

OF WELDING PRACTICE MUST BE ADOPTED. WHEREVER POSSIBLE STRESS RELIEF HEAT TREATMENTS SHOULD BE USED AND HARD MICROSTRUCTURES AVOlDED.

CONTROL OF ENVIRONMENT IN REACTOR TO HIGHEST STANDARD OF PURITY IS A KEY o

ELEMENT IN PREVENTING ENVIRONMENTAL ASSISTED CRACKING AND MUST ALWAYS BE A VERY HIGH PRIORITY TO REACTOR OPERATORS. FOR THIS REASON, STANDARDS OF, FOR INSTANCE, COVER GAS PURITY MUST BE STRICTLY MAINTAINED.

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OVER FOUR DECADES OF EXPERIENCE WITH LIQUID SO o

HEAT TRANSFER MEDIA CORROSION IN LIQUlO METAL SYSTEMS IS COMPLEX o

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- LOW STAINLESS STEEL CORROSION RATES WITH FLOWIN ABSENCE OF STAINLESS STEEL CORT OSION MECHANISMS IN e

SODIUM NO DESIGN PENALTY IS REQUIRED FOR USE OF AUSTENITI STAINLESS STEELS IN SOD!UM WHEN CHEMISTRY IN COVER e

AND SODIUM IS CONTROLLED

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Short Term, High Temperature Exposure of 21/4 Cr-1Mo Steel by Dr. Charles R. Brinkman, Group Leader MechanicalProperties Engineering Technology Division Oak Ridge NationalLaboratory Oak Ridge, Tennessee WK2-05-1192 A-10

SHORT TERM HIGH TEMPERATURE EXPOSURE The ASME Code Case N-47 provides stress allowables up to the following peak temperatures:

Alloy Temperature.

F( C) 2%Cr-1Mo Steel 1200 (649)

Type 304 Stainless 1500 (816)

Type 316 Stainless 1500 (816?

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OF DATA NEEDS SINCE 2%CR-1MO. STEEL IN PRISM COULD BE EXPOSED TO TEMPERATURES AS HIGH AS 677 C (1250 F) FOR SEVERAL DAYS, SHORT TERM TENSILE AND CREEP-RUPTURE STRENGTH VALUES NEED TO BE DETERMINED INORDER TO DEFINE STRESS-ALLOWABLES TO 732 C (1350 F).

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Radiation Effects on Material Ductility by Dr. Frank H. Huang, Principal Scientist Materials and Welding Westinghouse Hanford Company Richland, Washington WK2-05-11-92-A-12

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INTRODUCTION Neutron irradiation degrades the mechanical properties of LMR structures

• Establish test programs to monitor the tensile and fracture properties of LMR components fabricated from 304 and 316 stainless steel, SS welds, alloy 718 and 2.25Cr-1Mo e

Use displacements per atom (dpa) to measure irradiation effects on material properties Develop exposure limits in dpa for the design life of LMR components

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-e FFTF/CRBR Programs

- Tensile and fracture specimens were irradiated in EBR-il FFTF Surveillance Program

- Five Assemblies Contain Tensile and Fracture Specimens

- Fracture Toughness ' Surveillance

- Testing of materials from actual FFTF components irradiated in prototypic neutron environment Material property degradation by irradiation is evaluated in exposure units (dpa) based on atom displacements

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CONCLUSIONS Postirradiation test results establish exposure limits on the basis of residual total elongation.

Material surveillance studies verify safe operation and valid component design of the FFTF.

Similar surveillance programs will be instituted for ALMR

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Summary and Conclusions The Reactor Design lias Significant Safety Margins, Accident Prevention and Accident Accommodation Capabilities The ASME Code, (Section Ill Plus Code Case N-47) Provides Design Rules and Analysis M:thods that are Well Developed and Used Worldwide Tha ASME Code (Section XI, Division 3) Provides Requirements for In-Service Inspection; the Reactor Can Meet These Requirements; Major Portions of Section XI Division 3 of the C de are in Course ofPreparation Reactor Normal Operating Temperatures are Sufficiently low For Time-Independent Analysis R: actor Temperatures, loads and Time Durations During Transient and Accident Canditions Result in low Creep-Fatigue Damage WK2-05-11-92-02

ummary and Conclusions, Cont.

Tha ALMR Structural Materials are Used in Other Sodium Cooled Fast Reactors cad Have Extensive Operating Histories Tha Reactor Structures Have Significant Margin Against ASME Code Limits for Stress, Temperatures, Crcep and Fatigue Tha Effects of the Reactor Environment (Sodium and Radiation) on the Structural Mcterials are Well Understood and Accommodated by the Design Further Developments are Planned for 60 Years life Extention, Establishment of Th:rmal Striping Criteria and Development of Test Data for Short Term, High Tcmperature Exposure of 21/4 Cr-1Mo Stool - All are Part of the ALMR TcchnologyDevelopmentPlan WK2-05-11-92-03