ML20037A886
| ML20037A886 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 06/15/1971 |
| From: | Cherry M CHERRY, M.M./CHERRY, FLYNN & KANTER |
| To: | Murphy A Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 8007170805 | |
| Download: ML20037A886 (14) | |
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OC T NUMBER PROD. & uTR, EAC. (C M9,998 1
i McDEnxoTT. WILL & EMERY CD lli WEST MON ROC STREET f"
S CHICAGO, ILLINOIS 60603 S
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THis DOCUMENT CONTAINS l
P00R QUALITY PAGES
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Arthur 11. Murphy, Ecq., Chairman Atonic dafety and T,1consing Board 1
Colum'oin f1nivernity School of Law J
uox 38, 435 ticct u6th atreet Ilew York, New York 10027 Q
Re:
_AEC Docket tion. 50-309 an<' 50-330
Dear Mr. Chairman:
g Encloscal la a copy of the docuacnt list referre l to in Inter /cnors' ictter of June 10, 19'/1.
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Thia Idct in not intended to be exinustive s
of the documenter ** catorial which Intervonors may offer into evidence.
esfiectfully, 1
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)tyr I-Cherry M?tC/ cam Enclosurc ec:
Dr. Dnvid 3. liall Dr. Clark Goodman Iir. Stanicy T. Robinson, Jr.
All Councc1 of Record W*
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U Z1RCONIUM
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ru WAPD-TitGS-119 Zirconium as Material for the Reactor Core AECL-3375 On the Oxidation of Zirconium Alloys in Air and the
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Dimensional Changes Associated with Oxidation GEMP-669
'ihermal Conductivity and Electrical Resistivity of Zircaloy-4 Aht-6160 Examination of an Irradiated Prototype Eucl Elcmont for the Elk River Reactor ANL-6232 Studies of the liydrogen Damage Mochanica in the Corrosion of Zirconium AdL-6370 Corrosion Studios of Tornary Zirconium Alloys in High-Temperature Water and Secam Atw-7252 Studies in zirconium. oxidation BMI-1689 Studios in the UO -Zro2 System 2
iml-1803 Specific Heats and Heats of Transformation of Zircaloy-2 and Low-Nickel Zircaloy-2 DP-859 Irradiation of Tandem-Extruded Joints Botween Zircaloy and Stainless Steel GA-2229 Irradiation Effects on the Surface Reactions of Metals GA-2235 A Program of Rascarch on Mechanical Metallurgy as delated to Fuct-Element Fabrication GEAP-3739 Plastic Strain in Thin Fuct Element Cladding Due to UO2 Thermal Expansion GEAP-3933 Slow Cyclo strain Fatigue in Thin Wall Tubing GEAP-3999 Corrosion Mechanism of Zirconium and its Alloys -
Diffusion of Oxygen in Zirconium Dioxido nW-62347 Extrusion Characteristics of Uranium-Zirconium and Uranium-Carbon Alloys M&
Zirconium page 2 MW-oS465 nydriding in Purposely Defected, Zircaloy-Clad Fuel aods HW 67677 inc Physical Integrity and corrosion Resistance of the Zircaloy-2 Pressuro Tubos for the PRTR HW-o7949 REV Evaluation of Properties of Irradiated Zircaloy-2 Pressurc Tube from KER Loop 1 dW-68195 A Study of the Wear and Galling of Autoclaved Zirealoy-2 by various Materials dW-69679 Recovery and Recrystalli=ction of Zirconium and its Alloys, Part 2 Anncaling of Cold-Worked Zirconium HW-69680 Part 3 Annealing Effects in Zircaloy-2 and Zircaloy-3 HW-70151
'1hc Activation Energies for Creep of Zircaloy-2 nW-72002 Ultrasonic Testing of Zirealoy Sheath Tubing for ruel Eicments Hw-73398 Strength and Iietallurgical Proporties of the Zircoloy-2 Pressure Tubos for the PRTR HW-73511 iiinh Temperature oxidation of Zirconium and Zircaloy-2 in O>:ysen and Water Vapor HW-73693 REV Postirradiation Evaluation of Zircaloy-2 PRTR Pecssuro Tubes HW-74339 Impact Testing and Slow Notch-Bend Testing of Zircaloy-2 HW-74955 Effects of Cold Work and Neutron Irradiation on tho Tensile Properties of Zircaloy-2 dd-75052 Postirradiation Evaluation of Zircaloy-2 Prosauro Tubo from KEL 3, Preliminary Report nW-75267 Creep Properties of Zircalov-2 for Design Application HW-7oS62 RIV Role of the 0::idation Rate on the Hydriding of Zirconium Alloys in Gas Atmosphorcs Containing Hydrogen HW-76636 Neutron Irradiation and Cold Work Effects on Zircaloy-2 Corrosion and Hydroacn Pickup HW-30309 rno Ef fccus of Hot-Water Thermal Trcarments on the Cold Work Recovery of Zircaloy-2,
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Zirconium Pago 3 uw-dOSb7 Crack Propacation Tests on Normal and Hydrided Z1.rcaloy-2 kcactor Pressure Tubing HW-82102 Effects of Ucutron Irradiation on the Flow and Fracture Jchavior of Zircoloy-2 HW-82631 Fracture Studies of Zircaloy-2 int-d3164 Poscirradiation Evaluation of Zircaloy-2 PRTR Pressure Tubes IA-1389 drittle echavior of Zircaloy in an Emergency Core Cooling Environment KAPL-2111 Mcchanical Proporties of Zircaloy-2 Weld Metal KAPL-2110 Mechanical Proporties of Zircaloy-2 Kt2L-2149 Hydrogen Absorption by Zirconium-2 at. % Tin-2 at. %
hiooium Alloy During Corrosion KAPL-2203 Corrosion of Zircaloy in Crevices Under Nucicate Boiling Conditions KAPL-2221 Effect of Hydrogen on the Strain Fatigue Proporties of Zirenloy-2 Wald Ectal sil-1233 Interdif fusion in Zircaloy-2 Clad U_2 w/o Zr Fuct Materials and its Effect Upon Corrosion Bohavior NMI-1243 Deformation Modes of Zirconium at 770K, 3000K, and 10750K omsL-3039 novicu and Corrolation of In-Pile Zircaloy-2 Corrosion Data and a Model for the Effect of Irradiation ORNL-3398 Tac Effect of Stress State on High-Temperaturo Low-Cycle Fatigue OKNL-3514 Mechanical Cladding-Fuel Interactions During Thermal Cycling of r.ctal-Clad Fuel Elements ORSL-TA-2347 Failure Modos of Zircaloy-Clad Fuel Rods e
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e-STRUCTURAL STEELS 8hWu 678 Quarterly Progress Report Irradiation Effects on Reactor Structural Marcrials Aug, Sep, Oct, 1967 AEC-tr-5219 Corrosion of Reactor Materials:
a Collection of Articles AEC-tr-7192 Properties of Heat-Resistant Stects Following High Temperaturo Roactor Irradiation AGL-6389 Design Criteria for Stect in Nuclear Reactors ANL-6701 Analytical and Experimental Investigation of a bucicar acactor Support Structuro AHL-6d6d Annual Report for 1963 Metallurgy Division ANL-7162 Stress Analysis of a Reactor Core Support Structuro Consisting of Two Interconnected Multiregion Platos ANL-7266 Hydrogen Embrittlement in Irradiated Steels BMI-1813 Investigation of the Initiation and Extent of Ductile Pipo Rupturc, Progress Report for January-June, 1967 BMI-1817 July-September, 1967 BMI-1828 October-December, 1967 aM1-1834 Effcces of Larac Fast Fluences on Mechanical Properties of Type 347 Stainicos Steel and Aluminua eM1-1836 Investigation of... Pipe Rupture January-March, 196'8 aM1-1373 Ditto BMI-1866 Ditto 3M1-1876 Ditto, Phase 11 adl-l'860 Ditto, Phase II, Pros. Rep. January-March, 1970 2MI-1887 April-Juno, 1970
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.2 Structural Stoclc page 2 da.,L-109 3 Calculated Dsance Functions for Determining Irradiction Effectiveness DP-579 v.cchanical Proporties of irradiated Stainicss stools DP-531 Lcchcnical Proporties of Irradiated Plain Carbon and Alloy Seccis DP-lod 9 A Mechanica for Stress Corrosion Cracking of Stainicas Stcol in Reactor Systems DP-1199 Repair of a I,uclear Roactor Vessel A Proc am of Basic Research on Mcchanical Properties GA-3585 r
of Reactor Materials Hw-00425 Somo Effects of heutron Radiation on the Mechanical Properties and Structural Characteristics of High-Purity Iron Id-lll2 Characterization of the Fracture Surf ace of the PM-2A Pressure Vessel IN-1398 Incipient Failure Deccction by Acoustic Emission A Developr.cnt and Sectus Report OxNL-2334 An Investigation of the Corrosion Resistance of 2razina Alloys for Austenitic Stainless Steel Fuct Elements for Service in 565oF Pressurl=ed Water CM;L-2972 Sf fect of Enviromacnt on the Creep Properties of Type 304 Stainless Seccl at Elevated Temperaturoc OREL-4512 Heavy-section Steel Technology Program Semiannual Progress Report for period ending August 31, 1969 ORNL-4590 February 28, 1970 oRNL-4315 February 29, 1968 ORAL-NSIC-15 The Integrity of Reactor Pressurc Vossols TlD-7625 Technical Papers of the Thirteenth Metallographic Group hocting TID-17632 A Survey of 21 neactor Vessels in Light of the Anticipated Effects of Neutron Irradiation on the drittic-Rupture Proportion of Materials of Construction TID-17887 The True Stress-Strain Proporties of Brittic Materials to Very High Temperatures l
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4 LOCA WAPD-T-2222 Assessment of Safety injection Stratosics During Loss-of-Coolant Accidents IITRI-578-P-21-39 Water Decompression Experiments and Analysis for Blowdown of Nuclear Reactors 4
Laboratory Simulations of Cladding-Steam Reactions Ant-7609 Following Loss-of-Coolant Accidents in Water-Cooled 1
Power Reactors IDO-17219 Loft Core Design Report 1D0-17258 Loft Engineered Safety Systems Investigations ILJ-17250C Semiscalo Slowdown and ECC l
1DO-172SdJ ORhL-Lo f t Fission Product -Support Progrcm ID0-17256.(
Loft Integral Test Program t
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IDO-17256F Fucl Heatup Simulation Tests - FilUST 1h-1321 holap3 - A Computer Program for Reactor Bloudoun Analysis 13-1383 Tughnical Assistance in Reactor Safety Analysis 16-1334 Semiscalo Blowdown and Emergency Core Cooling Project IN-1338 Review of Heat Transfer Coefficients for Condensing Steam in a Containment Building Followins a Loss-of-Coolant Accident UC-80 Semiscale Blowdown and Emcreency Core Cooling Project Tests 822 and 823 1N-1393 Test koport UC-30 1i-1404 Tests 803-820 1h-1423 Particle Siro Distributions from Fuct Rods Fragmented During Power Burst Tasts in the Capsulo Driver Coro i
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LOCA pano 2 hporimental Investigations of Reactor System 31owdown I;;-104u Subcooled-dlowdown Forces on Roaccor-System Components:
Calculational i othod and Zxperimental Confirmation 16-1354 An Experimoncal Investination of Top and S2Cdr III lii-1355 Eottom Flooding of a Huelcar Cundio Simulator l
I l
l A Computer Code for Static and Dynamic Structural J
STuM Ls-1362 Analysis and Studies Made Using the Codo Experimental Results of the Fuci Heatup Simulation Scries Tests (FdUST) - Emergoney Core Cooling Test 15-1390 1i4-1391 Loft Core Length Study l
/.n Evaluation of the Applicability of histing Data to the taalytical Description of a Nucicar-kcactor eML-1856 for Ott-Doc, 1968 Accident Qaarterly Progress Ecport Apr-Jun; 1969 BMI-1367 Jul-Sep, 1969 dMI-1871 Oct-Dec, 1969 BMI-lS77 Jan-Mar, 1970 UM1-1331 Apr-Jun, 1970 BMI-1385 O
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URANIUM ANL-6702 itcat Capacity Studios of Uranium and Uranium-Fissium Alloys ANL 6o83 Examination of Irradiated As-In-Cd Alloys AM.-7070 Hochanical Properties of Uranium Compounds liW-b 707 2 Fi.nal Report:
Irradiction of Zircaloy-2 Clad Uranium Rods in Lan-Filled Cap::ulos,
11W-69234 Irradiation Effects on Uranium Dioxido Molting 11U-69393 Swolling in Uranium, a Comparison of the Effects of Irradiation and Postirradiation Anncaling M.CO-933 Structural Changes Associated with High-Temperature Leformation of Uranium NLCO-990 Rolling Studies of Uranium Rods and Tubes 4-1778 Formation of Corrosion-Rosistant oxide Film on Uranium AEC-tr-4467 Thormal Cycling Equipment and Experimental Data on Uranium.
Studies on Uranium Fuel Element 1.
GA-2691 A Program of Roscarch on Mcchanical Metallurgy as Rotated to Fuct-I:.lcment Fabrication T1D-11295 i<ucicar Fucis and Materials Development Program Summary for 1965 OML-4440 Fuels and Materials Developnkent Program Quarterly Progress Report for period ending Juno 30, 1969 oaNL-4480 for period ending Soprember 30, 1969 Oght-4520 for period ending December 31, 1969 olut-4560 for period endins March 31, 1970 O
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NSIC DOCUMENTS OlaL-hSIC-23 Potential Metal-k'acer Reactions in Light 'n'acer-Cooled Power Reactors ORI?L hSIC 22 Missile Generation and Protection in Light-Water-Cooled Power Reactor Plants 014sL-hSIC-25 Air Cleaning as an Engineered safety Feature in Light-L'ater-Cooled Power Reactors 01GL L'31C 25 Testing of Containment Systems Used with Light-Water-Cooled Power Reactors 01hsL-nSIC-27 Review of Methods of Mitigating Spread of Radioactivity from a Failed Containment System CahL-hSIC-15 Tne Integrity of Reacccr Pressure vessels Od.NL-hSIC-23 Earthquakes and huclcar Power Plant Design CinNL.-KS IC-29 Protection Instrumentation Systems in Light-Wator-Coolod Power Reactor Plants e
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l l-SAFETY RESEARCH REPORTS BhsL-1315-1 Nuclear Safoty Quarterly Report, Nov, Doc,1969, Jan,70 uhM.-1315-2 Feb, March, April, 1970 otein-3319 nuclear Safocy Program Seminnnual Progross Report for porlod onding Junc 30, 1962 oluu.-3401 for period ending December 31, 1962 oeL-3483 for period ending June 30, 1963 l
canL-3547 for period ending December 31, 1963 01hsL-3091 for period ending June 30, 1964 0:uiL-3776 for period ending December 31, 1964 ou t-4071 for period ending December 31, 1966 02aL-4228 for period ending December 31, 1967 O!GL-4374 for period ending December 31, 1968 0!uu.-4511 for period ending December 31, 1969 O
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MISC 2LLAtGOUS GENER.U 1GFERENCES AEC-tr-4440 Procedure for Reactor Plant Site Selection Based on nuclear Aspects LA-4316 Tne Application of Risk Allocation to Reactor Siting and Design LA-4449 A Risk Analysis of the Onega West Reactor 01ciL-liUE-ll The Siting of Nuclear Reactors as Related to Urban ricat Supply OlML-NSIC-64 Abnormal Reactor Operating E>:periences 1966-68 OiGL-NSIC-69 Safety-Rotated occurrences in Nuclear Facilitics as Reported in 1967 and 1968 ID0-17252 A Digient Computer Code for Predicting the Pressuro-Temperaturc liistory within' a Pressurc-Suppression Containment Voa.ict in Response to a Loss-of-Coolant Acc:. dent IN1330 Analysis of Fault Trees by i<inctic Troo Tacory TlD-2SS37 Operating History - U. S. Nuclear Poucr Roaccors heteorology and Atomic Encray 1968 e
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. s TWO PHASE FLOW AND OSCILLATION CHEC 156 Water injection Test Simulating F.odcrator Flooding of monus Superheater Acccmbly NASA TN D-3553 Stability of Intermixing of High-Velocity Vapor with its subcooled Liquid in Cocurrent Streams of Subcooled Nucleato Boiling of Experimental Study /4-Inch-Diameter Tubes at Low NASA water Flowing in 1 Pressures WlJD-T-1824 Flou Patterns in High Prec:ure Two-Phaso (Steam-Water) Flow with Heat Addition 0
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-.c FUEL ROD FAiLUKE EXPERD12NTS PERForo:ED AT O.R.li.L.
OldL-m.2919 01mL huclear safety : cscarch and Development Prot; ram dimonthly ; cport for January-February 1970 OldqL-TM-2984 OEL...di. monthly Report for ?f.e.rch-April 1970 OAL-n-3061 OAiL...Simonthly Report for.F.ay-June 1970 ciGL-r.:-25bd OImL.. 3imonthly Report for March-Aprii 1969 ORht-TM-2548 Failuro Modes of Zircaloy-Clad Fuct Rods OiGL-4635 Final Report on the first Fuel Rod Failure Transient Test of a Zircaloy-Clad Fuct Rod Cluster in Treat UC-80 A Metallurgical Ivaluation of Simulated SLR Emerncncy Core Coolina Tcsts OA.L-T6-2029 oluL huclear safety Research and Development Program cimonthly Report for isovember-Docember 1969 i
u&d-100d Physico-Chemical Studies of Clad UO2 UndU Reactor Accident Conditions 0.UL-TA-2850 Design, Equipment, and Program for Fuct Rod Durst Experiments i
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