ML20195J566
| ML20195J566 | |
| Person / Time | |
|---|---|
| Issue date: | 06/04/1980 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| Shared Package | |
| ML20195J563 | List: |
| References | |
| NUREG-0435, NUREG-0435-V02-N02, NUREG-435, NUREG-435-V2-N2, NUDOCS 8812020275 | |
| Download: ML20195J566 (131) | |
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Data As Of: O 04-80 0FFICE OF NUCLEAR l
REGULATORY RESEARCH i
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STATUS
SUMMARY
REPORT l
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RESEARCH RESULTS UTILIZATION l
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b IABLE OF CONTENTS PAGE NO.
DESCRIPTION 1
1.0 INTRODilCTION 2
2.0
SUMMARY
OF IMPACT OF RESEARCH INFORMATION LETTERS (RIL'5)DN THE REGULATORY PROCESS 6
2.1 LIST OF ISSUED RAL*5 12 2.2 POTENTIAL APPLICABILITY OF RESEARCH RESULTS IN THE REC-ULATORY PROCESS 114 3.0 PROJECTED NEAR TERM RESEARCH INFDPMATION LETTERS G
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1.0 INTRODUCTION
THIS REPORT DN "RESEARCH RESULTS UTILIZATION" PROVIDES STATUS AND CONTROL INFORMATION CONCERNING THE UTILIZATION OF RESEARCH RESULTS IN THE REGULATORY POLICIES AND PRACTICES OF THE NRC.
REiEAPCH INEEPMATION LETTEEi -- (RIL'5) ARE PREPARED BY RES TO TRAN5MIT RESEARCH RESULTS 10 NRC USEP OFFICE 5 UPON COMPLETION OF A SUBSTANTIAL, COHERENT AND REASONABLY COMPLETE BODY OF EXPERIMENTAL AND/OR ANALYTICAL RESEARCH WORK.
A RIL MAY COVER MATERIAL DEVELOPED FROM MORE THAN ONE RESEARCH PROJECT.
THE USER / PROGRAM OFFICE (5) IN THE HRC REVIEW THE INFORMATION CCHTAINED IR THE RIL AND CONSIDER ITS U!ILIZATION IN THE REGULATORY PROCESS. SECTION 2.0 0F THIS REPORT LISTS THE *IL*5 ISSUED TO DATE. TOGETHER WITH AN IDENTIFICATION OF THE RESEARCH PROGRAM MANAGER AWD THE RESEARCH DECISION UNIT WHICH GENERATED THE RIL.
THE POTENTIAL APPLICABILITY OF EACH RIL TO THE REGULATORY PROCF55 IS AL50 IDENTIFIED HERE, AND COMMENTS FROM THE COGNIZANT RES AND USER OFFICE STAFF ARE SUMMARIZED WHICH RELATE TO THE EXPECTED IMPACT OF THE REPORTED RIL'S ON THE REGULATORY Pc0CE55.
WHERE DEEMED APPROPRIATE BY MUTUAL AGREEMENT BETWEEN RES AND OTHER It8VOLVED PROGRAM OFFICES, A POSITION PAPER MAY BE PREPARED TO INFORM THE COMMISSION ABOUT THE POTENTIAL APPLICATION CF RESEARCH RESULTS. BRIEFINGS MAY BE CONDUCTED FOR THE COMMISSION, THE ADVISORY CCMMITTEE ON REACTOR SAFEGUARDS (ACRS) OR OTHER APPROPRIATE GROUPS.
PRESS RELEASES MAY ALSD BE MADE. AS APPROPRIATE.
THESE EVENTS ARE SCHEDULED AND TRACKED IN THIS SECTION.
A LISTING OF ALL RIL'S THAT MAY BE GENERATED IN THE NEAR FUTURE (2 DR 3 YEARS) IS PRESENTED IN SECTION 3.0.
THE SUBJECT. TITLE. IMPACT, TARGET DATE, AND RESEARCH REVIEW GROUP NUMBER, TITLE AND CHAIRMAN ARE SHOWN FOR EACH RIL.
RES 15 RELPONSIBLE FOR DISTRIBUTING EACH NEW RIL TO COGNIZANT OFFICES AS THEY ARE ISSUED.
RES AND OTHER COGNIZANT PROGRAM OFFICES ARE OBLIGATED TO COLLECT. REVIEW, AND FORWARD APPROPRIATE INFORMATION RELATING TO RILS AND ASSOCIATED FOLLOW-UP ACTIONS TO MPA.
MPA Is RESPONSIBLE FOR REFLECTING NEW INFORMATION IN THIS PUBLICATION AND PRODUCTION AND DISTRIBUTION OF THIS PUBLICATION ON A QUARTERLY SCHEDULE.
ALL COMMENTS SHOULD BE FORWARDED IN WRITING T0:
DIRECTOR. MPA MAIL STOP 12711 MNBB AT!N2 D. BIDLE -
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2.0 1HDDaEI DE IDEASI DE E1111 95 IHE RESHLAIREI EE2EE11 INCREASED TECH BASIS STANDARD STANDARD LICENSEE UNDERSTAhDING LICENSING REGU-REG.
TECH STD.
FORMAT 4 REVIEW INSPEC.
REPORIING RILS OF PHENDMENOM REVIEW LATION GUIDES SPECS DEVELOP.
CONTENT PLAN PROGRAM REQUIREMENT 1
YES 2
SCME POSSIBLE YES POSS.
3 YES 4
YES YES 5
YES 6
SOME 7
YES 88 YE5 POSSIBLE POSS.
POSS.
POSS.
9e YES POSSIBLE POSS.
POSS.
POSS.
10 YE5 YE5 11 POSS.
POSSIBLE 12 YES YES 13 YES YES YES YE5 YES YES 14 YES YES YE5 YES YES 15 YES YES YES 16 YES YES 17 YES YE5 18 POSS.
POSSIBLE 19 POSS.
20 YES YES YES YES YES 21 YE5 POSSIBLE 23 POSSIBLE POSS.
POSS.
POSS.
POSS.
24 POSSIBLE POSS.
POSS.
25 SOME 26 YES YES 27
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i 28 YES YES 29 30 YES YE5 31 32 YES YES YES 33 YE5 YES YE5 34 YES YES 35 36 YES YE5 37 POSS.
38 39 40 41 YES YES YES YES 42 43 44 YES YES 45 46 YES YES 47 4C YES PD55TBLE POSS.
YES 49 YE5 50 51 52 53 54 55 56 578 588
- 05 60 YES POSSIBLE POSS.
YES 61 62 YES YES POSS.
YES YES 638 64 YES YES POSS.
YES 65 YES YES POSS.
YES YE5 66 YE5 POSSIBLE POSS.
YES 678 68 YES YES YES YES 69 YES YES POSS.
YES YES 70 YES POSSIBLE POSS.
YES 71 YES POSSIBLE POSS.
YES 72 YEE YES POSS.
YES YES 73 YE5 YES YES YES 74=
ND NO NO YES 75 YES YES YES YES 76 YES YES YES POSS.
POSS.
POSS.
YES 77 YE5 YES YES YE5 78 79 YE5 YES YES POSS.
POSS.
POSS.
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83
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YE5 YE5 YES
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i 888 395 90s
- - CURRENTLY BEING REVIEMED BY USER OFFICE (5) 8 - AWAITING COMMISSION APPROVAL ON PROPOSED ACTION PL AN FOR CHANGING ECC5 RULE 2.1 LIST OF IS5UED PIL*5 RESEARCH INFORMATICM LETTEQS ISSUED BY CES TO DATE ARE LISTED BELOW IN TABLE 2.1 TOGETHER WITH THE DATE OF 155ut. ASSOCIATED RESEARCH PROGRAM ELEPENT AND P20GR.*.M MANAGER.
DATE RIL NO.
PAGE NO, 155UED RESEARCH INFORMtTION LETTER TITLE RES DECISION UNIT PROJECT MANAGER 1
13 03/19/74 ORNL V-5 INTERMEDIATE VESSEL TEST RESULT 5 PRIMARY SYS INTEG.
C. SERPAN 2
14 05/20/74 SEISM 0 TECTONIC MAP OF THE EASTLRN UNITED SEISMIC.
J. HARBOUR STATES ENGINEERING &
SITE SAFETY 3
15 08/07/74 DRNL V-7 INTERMEDIATE VE5SEL TEST RESULTS PRIMARY 515 INTEG.
C. SERPAN 4
16 09/10/76 fiAP SHOWING RECENCY OF FAULTING IN COASTAL SEISMIC.
J. HARBOUR SOUTHERN CALIFORMIA ENGINEERING &
SITE SAFETY 5
17 06/28/76 CONFIRMATORY PRESSURE 7E55EL TEST UNDER PRIMARY SYS INTEG.
C. SERPAN PNEUMATIC LOADING 6
18 10/12/76 A CRITIQUE OF THE BOARD-HALL MODEL FOR THERMAL FAST BREEDER R. WRIGHT DETONATIONS IN THE UO2-NA SYSTEM REACTCRS 7
19 08/25/76 THE SIMMER CORE FOR ANALYSIS OF HYPOTHETICAL FAST BREEDER R. CURTIS CORE DISRUPTIVE ACCIDENTS IN LMFBR'S REACTOR 5 8
20 01/31/77 DECAY HEAT DATA APPLICABLE TO LOCA EVALUATION FUEL BEHAVIOR R. DISALVO 9
25 03/14/77 HIGH TEMPERATURE OXIDATION OF ZIRCALOY FUEL FUEL BEHAVIOR M. PICKLESIMER CLADDING IN STEAM to 22 02/25/77 PRE 55URE VESSEL FAILURE PROBABILITY PREDICTION RISK ASSESSMENT /
W. VESELY (OCTAVIA CODE)
PRIMARY SYS INTEG.
11 23 09/15/77 IEEE NUCLEAR RELIABILITY DATA MANUAL RISK ASSESSMENT J. JOHNSON 12 24 06/16/77 MODIFICATIONS TO PRESSURE VESSEL FATLURE RISK ASSESSMENT /
W. VE5ELY PROBABILITY PREDICTION (OCTAVIA CODE)
PRIMARY SYS INTEG.
13 25 11/18/77 RESIDUAL STRESSE5 IN WELDS PRIMARY SYS INTEG.
C. SERPAN 14 26 11/09/77 PHYSICAL SEPARATION CRITERIA FOR ELECTRICAL SYSTEMS ENG.
R.
FEIT CABLE TRAYS (HORIZONTAL OPEN SPACE CONFIG.)
15 28 12/01/77 CHARACTERIZATION OF BWR FEEDWATER N0ZZLE PRIMARY SYS INTEG.
C. SERPAN CORNER CRACK 5 16 29 12/01/77 WARM PRESTRE55ING PRIMARY SYS INTEG.
C. SERPAN 17 30 05/05/78 PDWER BUEST FACILITY (PBF) SINGLE ROD-POWER FUEL BEHAVIOR R. VAN HOUTEN COOLING MISMATCH (PCM) TEST RESULTS 18 31 11/09/77 FPANTIC COMPUTER CODE RI5K ASSESSMENT F. GOLDBERG >
5 DATE RIL NO, gaiE_E2 ISSUED PESEARCH INFORMATION LETTER TITLE RES DECISION UNIT PROJECT MANAGER
'9 32 01/31/78 GO METHODOLOGY ASSES 5 MENT RISK ASSESSMENT J. PITTMAN 20 33 01/24/78 A STUDY OF PHYSICAL PROTECTION EQUIPMENT SAFEGUARDS E. RICHARD 21 35 03/24/78 CRITICAL REVIEW OF SODIUM HYDROXIDE AEROSOL RISK ASSESSMENT M.
CULLINGFORD T0XICITY 23 37 04/10/78
- EASI" ADVERSARY SEQUENCE EVALUATION MODEL SAFEGUARDS R. ROBINSON (CGMPUTER GRAPHICS VERSION) 24 38 04/10/78 "FE5EM" ADVER5ARY SEQUENCE EVALUATION SAFEGUARDS R. ROBINSON MODEL 25 39 03/21/78 FRAP-53 FUEL BEHAVIOR G. MARINO 26 41 04/27/78 THE IMPACT OF OFFSHORE NUCLEAR GENERATING FUEL CYCLE SAFETY D. BARNA STATION 5 DN RECREATIONAL BEHAVIOR AT ADJACENT AND ENVIRONMENTAL COASTAL SITES.
EFFECTS 27 42 06/02/78
- BEACON / MOD 2" CODE DEVELOPMENT
- 5. FABIC 28 43 05/09/78 "MELT / CONCRETE INTERACTION 5" FUEL BEHAVIOR R. DISALVO 29 45 06/07/78
- FUEL ROD ANALYSIS COMPUTER CODES FRAP-T3" FUEL BEHAVIOR H. SCOTT 30 47 06/28/78 PHASE I FINAL REPORT, "BARRIER PENETRATION SAFEGUARDS R. ZIMMERMAN DATA BASE"; 0F STUDY, "ASSISTANCE-PHYSICAL PROTECTION ASSESSMENT 5" 31 48 07/10/78 ASSAY OF STANDARD REFERENCE MATERIAL (SRM)950 SAFEGUARDS R. SHEPARD 32 49 08/03/78 IMPROVEMENTS IN THE AEROSOL BEHAVIOR CODE FOR FAST BREEDER J. LARKINS RADIOLOGICAL ASSESSMENTS OF LMFBR'S REACTORS 33 50 08/03/78 PLUTONIUM ACCIDENT CONTAINER PROGRAM RESEARCH, FUEL CYCLE SAFETY W. LAHS DESIGN AND DEVELOPMENT.
AND ENVIRONMENTAL EFFECTS 34 51 08/03/78 NUCLEAR DECAY DATA FOR RADIONUCLIDES OCCURRING FUEL CYCLE SAFETY J. FOULKE IN ROUTINE RELEASES FROM NUCLEAR FUEL CYCLE AND ENVIRONMENTAL FACILITIES EFFECTS 35 52 09/15/78 SFACTOR* A COMPUTER CODE FOR CALCULATING DOSE FUEL CYCLE SAFETY J. FOULKE EQUIVALENT TO A TARGET ORGAN PER MICROCURIE -
AND ENVIRONMENTAL DAY RESIDENCE OF A RADIONUCLIDE IN A SOURCE EFFECTS ORGAN 36 53 09/27/78 EVALUATION OF GENERAL ATOMIC CODEss OXIDE-3, FAST BREEDER J. LARKINS 50RS, TAP, AND RECA.
REACTORS 37 55 09/28/78 LOFT REACTOR SAFETY PROGRAM RESEARCH RESULT 5 LOFT G. MCPHERSON THROUGH OCTOBER 1,
1978 -
DATE RIL NO.
PAGE NO.
ISSUED RESEARCN INFDPMATION LETTER TITLE
- ES DECISION UNIT PROJECT "AMAGER 38 S7 10/13/78 RESULTS OF THE INITIAL SERIES OF ACPR EXPERI-FAST BREEDER R. WRIGHT MENTS ON PRCMPT-BURST ENERGETICS WITH FRESH REACTORS OXIDE FUEL 39 58 11/27/78 REL AP-4/ MOD 6 CODE DEVELOPMENT 5.
FABIC 40 59 12/15/78 THE COMPUTER CODE BRENDA -A COMPUTER PROGRAM FAST BRFEDER P. WOOD FOR THE DYNAMIC SIMULATION FOR A LIQUID METAL REACTORS FAST BREEDER REACTOR PLANT 41 60 12/19/78 LABORATORY TESTING PROCEDURES TO DETERMINE THE SEISMIC.
N.
STEUER CYCLIC STRENGTH OF SOILS ENGINEERING &
SITE SAFETY 42 61 12/20/78 CRITICAL EXPERIMENT PROGRAM FOR NEUTRONICS CODE FAST BREEDER P. WOOD VERIFICATION REACTORS 43 63 01/10/79 SUPER SYSTEM CODE. A COMPUTER PROGRAM FOR FAST BREEDER P. WOOD DYNAMIC SIMULATION OF LMFBR POWER PLANTS REACTORS l
l 44 64 01/04/79 RADIATION DOSE TO CONSTRUCTION WORKERS AT FUEL CYCLE SAFETY J.
FOULKE OPERATING NUCLEAR POWER PLANT SITES AND ENVIRONMENTAL EFFECTS 45 65 02/11/79 THE CONCEPT COMPUTER CODE & CAPITAL COSTS FUEL CYCLE SAFETY D. BARNA FOR BOILING WATER REACTOR PLANTS AND ENVIRONMENTAL EFFECTS 46 66 02/s2/79 THE EFFECTIVENESS OF CABLE TRAY COATING SYSTEMS R.
FEIT MATERIALS 1 BARRIERS IN RET ARDING THE ENGINEERING COMBUSTION OF CABLE TRAYS SUBJECTED TO EXPOSURE FIRES 8 IN PREVENTING i
PROPAGATION BETWEEN CABLE TRAYS l
(HORIZONTAL OPEN SPACE CONFIGURATION) 47 67 03/19/79 INREM II: A COMPUTER IMPLEMENTATION FUEL CYCLE SAFETY J.
FOULKE OF RECENT MODELS FOR ESTIMATING THE AND ENVIRONMENTAL DOSE EQUIVALENT TO ORGANS OF MAN FROM EFFECTS AN INHALED 02 INGESTED RADIONUCLIDE 43 69 04/03/79 A TECTONIC OVERVIEW OF THE CENTRAL
- SEISMIC, N.
STEUER MID20NTINENT ENGINEERING &
SITE SAFETY 49 70 04/04/79 IN VITRO DISSOLUTION OF URANIUM FUEL CYCLE SAFETY J.
FOULKE PRODUCT SAMPLES FROM FOUR URANIUM AND ENVIRONMENTAL HILLS EFFECTS SO 71 04/06/79 CRITICALITY SAFETY GUIDANCE FUEL CYCLE SAFETY D.
SOLBERG AND ENVIRONMENTAL EFFECTS S'
72 04/22/79 THE CONCEPT COMPUTER CODE AND CAPITAL FUEL CYCLE SAFETY D. BARNA COSTS FOR PRESSURIZED WATER REACTOR AND ENVIRONMENTAL PLANTS EFFECTS r t
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DATE RIL NO.
PAGE NO.
ISSUED RESEA8CH INFORMATION LETTER TITLE RES DECISION UNIT PROJECT MANT.GER l
52 73 04/23/79 EARTHQUAKE INTENSITY SCALE SEISMIC.
R. BRAZEE
.i ENGINEERING 1 SITE SAFETY 53 74 05/16/79 LEBRIS-BED CCOLABILITY LIMITS. RESULTS FUEL CYCLE SAFETY R. WRIGHT FROM IN-CORE TEST 5 D-1 D-2 AND D-3 AND ENVIRONMENTAL EFFECTS 54 75 05/15/79 THE SET EQUATION TRANSFORMATION SYSTEM RISK ASSESSMENT W. VE5ELEY l
55 77 05/29/79 THE CONCEPT COMPUTER CODE AND CAPITAL FUEL CYCLE SAFETY D. BARNA t
COST FOR HIGH AND LOW SULFUR COAL AND ENVIRONMENTAL PLANTS - 1200 MWE EFFECTS 56 78 07/25/79 EFFECTS OF NUCLEAR POWER PLANTS ON FUEL CYCLE SAFETY C. PRICHARD COMMUNITY GROWTH AND RESIDENTIAL AND ENVIRONMENTAL PROPERTY VALUES EFFECTS 57 79 08/10/79 SMALL SCALE ECC BYPASS RESEARCH SYSTEMS ENGINEERING A. SERKIZ RESULTS I
1 58 80 05/29/79 COMPARISON OF SIMULATION MODELS FUEL CYCLE SAFETY P. REED USED IN ASSESSING THE EFFECTS AND ENVIRONMENTAL 4
j OF POWER PLANT INDUCED MORTALITY EFFECTS ON FISH POPULATIONS 59 81 09/21/79 TRANSIENT FUEL ROD BEHAVIOR FUEL BEHAVIOR G. MARINO CODE: FRAP-74 j
60 82 10/12/79 SEISMICITY AND TECTONIC SEISMIC.
N.
STEUER RELATIONSHIPS OF THE NEMAHA ENGINEERING &
UPLIFT IN OKLAH3MA SITE SAFETY 61 83 10/11/79 MOLTEN SODIUM INTERACTION FAST BREEDER T. WALKER WITH BASALT CONCRETE REACTORS 62 84 11/01/79 NEW MADRID SEISM 3TECTCNIC STUDY SEISMIC.
M. STEUER ENGINEERING &
SITE SAFETY 63 85 11/01/79 LOFT REACTOR SAFETY PROGRAM LOFT G. MCPHERSOP RESEARCH RESULTS FROM NUCLEAR LOSS-OF-COOLANT EXPERIMENTS
.l L2-2 AND L2-3 64 86 11/05/79 A REVISED AND AUGMENTED LIST OF SEI%MIC.
M. STEUER EARTHQUAKE INTENSITIES FGR ENGINEERING &
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KANSAS. 1867-1977 SITE SAFETY 4
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DATE RIL CXL_
PAGE NO.
155tMED RE9E40CH ICFORMATJC;* LETTER TITLE RES DECISION UNIT PROJECT MAN 4GE3t 65 88 11/05/79 RECONNAISSANCE BEPT&CK GEOLOGIC MAP OF
- SEISMIC, N. STEUER MfRLBOROUGH Olff.DRANGLE, MA AND RECONNAISSANCE ENGINEERING &
BEDROCK CCULOGIC MAP OF 5HREWSBURG QUAD-SITE SAFETY P a If,i g, M4 66 89 11/05/79 A STUDY OF THi REGIONAL TECTONICS AND
- SEISMIC, M. STEUER SEISMICITY OF EASTERN KANSAS -
SUMMARY
OF ENGINEERING &
PROJECT ACTIVITIES AND RESULTS TO THE END SITE SAFETY OF THE SECOND YEAR OR SEPTEMBER 30, 1978 67 90 11/06/79 REFLOODING CF SIMUL ATED PER CORES AT LOW SYSTEMS ENGINEERING L. THOMPSON FLOW RATES 68 11 11/11/79 STRUCTURAL INTEGRITY OF WELD REPAIRED PRIMARY SYSTEMS M. VAGINS PRESSURE VESSELS INTEGRITY 69 92 11/19/79 AM INTEGRATED GEOPHYSICAL AND GE0 LOGICAL
- SEISMIC, N. STEUER STUDY OF THE TECTONIC FRAMEWORK OF THE ENGINEERING S 38TH PARALLEL LINEAMENT IN THE VICINITY OF SITE SAFETY ITS INTER 5ECTION WITH THE EXTENSION OF THE NEW MADRID FAULT ZONE 70 13 11/19/79 SEISMICITY AND TECTONIC RELATION 5 HIPS OF
- SEISMIC, N. STEUER THE MEMAHA UPLIFT I t* OK, PART II. JAN. 1979 ENGINEERING S SITE SAFETY 71 94 11/19/79 REGIcwAL TECTONICS AND SEISMICITY OF SEISMIC N. STEUER EASTERN NEBRA5KA A;AU4L REPORT - JUNE 1, ENGINEERING S 1977-MAY 30, 1978 SITE SAFETY 72 15 11/16/79 NEW ENGLAND SEISM 0 TECTONIC STUDY ACTIVITIES
- SEISMIC, N. STEUER DURING FISCAL YEARS 1977 AND 1978 ENGINEERING &
SITE 5AFETY 73
?6 11/16/79 IN VIRD COUNTING AT SELECTED URANIUM MILLS FUEL CYCLE SAFETY J. FOULKE AND ENVIRONMENTAL EFFECTS 74 17 11/16/79 STEADY-STATE FUEL ROD BEHAVIOR CODE:
CODE DEVELOPMENT G. MARING FRAPCON-1 75 98 11/27/79 INVENTORY, DETECTION, AND CATALOG OF
- SEISMIC, M. STEUER OKLAHOMA EARTHQUAKES AND EARTHQUAKE MAP OF ENGINEERING &
OKLAHOMA, MAP GM-19 SITE SAFETY 76 99 12/28/79 ANNEALING OF IRRADIATED REACTOF *RE55URE PRIMARY SYSTEMS C. SERPAN VE5SELS INTEGRITY 77 100 12/28/79 ORIGIN OF SURFACE LINEAMENTS IN NEMAHA
- SEISMIC, M. STEUER COUNTY, KANSAS EPGINEERING &
SITE SAFETY i f
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DATE RIL NO.
PAGE N9.
15 5 U E D __
PE5EARCH INFCPMATION LETTER TITLE RES DECISIOM UNIT _
PROJECT MANAGER 78 101 12/28/79 VERTICAL LOADS IN MARK I CONTAINMENT TORUS CODE DEVELOPMENT R. CUDLIN 79 102 12/28/79 EVALUATION OF SEISMIC QUALIFICATION TESTS SEISMIC.
B. BRCWZIN FOR NUCLEAR POWER PLANT EQUIPMENT ENGINEERING &
SITE SAFETY 80 103 01/95/80 DETEFMINING EFFECTIVENESS OF ALARA DESIGN FUEL CYCLE SAFETY J. FOULKE AND OPERATIDMAL FEATURES AND ENVIRONMENTAL EFFECTS 81 104 02/28/80 IRRADI ATED FUEL DI5RUPTION UNDER LOF ACCIDENT FAST BREEDER R. WRIGHT CONDITION 52 RESULTS OF ACPR TEST SERIES REACTORS FD-1 AND THE FI5GA5 CODE 82 105 02/29/80 THREE MILE ISLAND TELEPHONE SURVEY 2 FUEL CYCLE SAFETY C. PRICHARD PRELIMINARY REP 0RT ON PROCEDURES AND AND ENVIRONMENTAE FINDING 5. AND THE SOCIAL AND ECONOMIC EFFECTS l
EFFECTS OF THE ACCIDENT AT THREE MILE ISLAND 2 FINDINGS TO DATE.
83 106 03/24/80 STEAM GENERATOR TUBE INTEGRITY PRIMARY SYSTEMS C*.
SERPAN INTEGRITY r
84 107 03/24/83 STUDf 0F LIEQUEFACTION RESULTING FROM SEISMIC.
R. BRAZEE EARTHQUAKE CF FEBRUARY 4 1976 NEAR LAKE ENGINEERING S AMATI'LAN. GUATEMALA SITE SAFETY 85 108 03/24/80 AN INTEGRATED GEOPHYSICAL AND GEOLOGICAL SEISMIC.
N. STEUER STUDY OF THE TECTONIC FRAMEWORK OF 1HE 38TH ENGINEERING S PARALLEL LINEAMENT - ANNUAL REPORT FT 1979 SITE SAFETY 86 109 04/04/80 CAS SCINTILLATION PROPORTIONAL COUNTER FOR FUEL CYCLE SAFETY J. FOULKE MEASURI14G PLUTONIiV1 IN HUMANS AND THE AND ENVIRONMENTAL ENVIRONMENT EFFECTS 87 110 04/24/80 ECONOMETRIC MODEL FOR THE DI5 AGGREGATION OF FUEL CYCLE SAFETY C. PRICHARD STATE-LEVEL ELECTRICITY DEMAND FORECASTS TO AND ENVIRONMENTAL THE SERVICE AREA EFFECTS 88 111 04/25/30 DESIGN CRITERIA FOR CLOSELY-SPACED N0ZZLES PRIMARY SYSTEMS C. SERPAN IN PRESSURE VESSELS INTEGRITY 89 112 05/11/80 STRUCTURAL AND MECHANICAL COMPONENT TEST SEISMIC.
J. O'BRIEN TECHNIQUES ENGINEERING &
SITE SAFETY 90 113 05/22/E0 RELAP-4/ MOD 6 ASSESSMENT CODE DEVELOPMENT W. LYON U-A 9i HLnd u
2.2 POTENTIst APPLICAPILITY Cr RESEAPCM RESULTS IN THE MEGULAT07Y PPOCESS CCMMENTS ARE OrrEAED FRCM RES AND FROM COGNIZANY USER OFFICES ON THE NATURE OF THE REPORTED RESULTS AND THEIR POTENTIAL (OR ACTUAL) APPLICABILITY TO ANY PART OF THE REGULATORY PRCCE55. INCLUDING 2 A) TECHNICAL DATA SUPPORTING LICENSING REVIEWS OR REGULATORY JUDGMENT, 3) EVALUATION CODES, C) INFORMATIDM APPLICABLE TO REGULATORY GUIDES OR STANDARDS, AND D) IhFCEMATION SUPPORTING JUDGMENTS REGARDING REGULATORY POLICY.
. P a
5 l
PROGR A9 OF FIEE COM L*ts ON P31ENfIAL UTILIZATION 09 VALUE OF *ESEARCH RESULTS IN THE PEGULATORY PROCESS PIL 5:
1 PATE_1559EDE 03/19/74 RF% DECISION UNIls PRIMARY SYSTEMS INTEGRITY j
RIL TITLE-ORNL V-5 INTERMEDIATE VE5SEL TEST RESULT 5 (ORHL H55T PROGRAM) j
$PON59 RING OFFICE (5)
RES EES:
1-20 VE5SEL INTEGRITY RESEARCH PROJECT MGR C. SERPAN P11_[Q"I51.5ENils EXPERI"LNTAL EVIDENCE THAT A "$4FE" FAILURE MODE (LEAK-BEFORE-BREAK) FOR REACTOR PRESSURE VE5SELS MAY LX WA5 REPORTED FOR THE FIRST TIME.
TESTS ON THE V-5 INTERMEDIATE VESSEL AT ORNL DEMONSTRATED THAT A i
LEAK OCCURRED IMSTEAD OF A FRACTURE BREAK 'JHEN THE VESSEL WAS HEATED TO 190 F AND PRESSURIZED TO 26,600 PSI.
t FURTHER ANALYTICAL STUDY TO GENERALIZE TNIS RESULT 15 UNDERWAY.
USER DISCUSSION POSITIDH COMMISSION ACRS PRESS i
0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULT 5 i
E051_?ll_8G31VIJ111 PEl1[W HELD COMPLETED HELD HELD ISSUED IMPLEMENTED NRR i
OFFICE RESPdH51 ELL.........
!.R R/ S D UNSCHED UN5CHED UNSCHED UNSCHED UNSCHED N/A SCHEDULED COMPLETION DATE..
l ACTUAL COMPLETION CATE.....
08/21/78
-~
N/A lip LCOP""E Nil p 09109/77. IA/20'78.
B.
CPIM DL5(P10E AttLIEA.11La to PEGutAf0pv Pe.r 32 THE RESULTS OF THIS TEST ADDED TD THE STAFF's UNDERSTANDING CF FRACTURES 1
ORIGINA3ING Al FLAWS IN FUZ2LE CORNLks OF ilEAVY SECTION STEEL VE5SELS. IF LEAK-BEFORE-BREAK COULD BE DEMONSTRATED UNDER j
ALL REASONABLY CONCEIVABLE CONDITIONS AND CIRCUMSTANCES. IT WOULD DEMONSTRATE THAT CURRENT LICENSING POSITIONS ARE VERY CONSERVATIVE.
R$?P]BE_IEPALT_QE PE5y(T1: WHILE ADDING TO OUR UNDERSTANDING OF VE5SEL FAILURE MODES. THERE HAS BEEN NO DtFINITIVE IMPACI ON tICENSING AT THIS TIME.
(QMN[WT5/PEMAPE):
THE OPINIONS EXPRESSED IN THE RIL REGARDING LEAK-BEFORE-BREAK ARE ENCOURAGING BUT NEED j
FURTHER SUBSIANIIATION FEFORE CURRENT LICENSING PD51TIONS CAN BE RELAXED.
iA_GC";fN TipH _ 09/13/78. P.
PANDALL I
IHIS PkOGRAM HA5 GLNLRATED SUBSTANTIAL DATA USED IN PPEDICTING IRRADitTION EMBRITTLEMENT AND MARGINS i
l TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT. CONSIDdRING MATERIAL PROPERTIES AND LOADINGS.
THIS PROGRAM HAS PROVIDEG INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF i
rn653U;I *."CSEEL =avrpf aa 5 DESCRIBED IN APPENDIX G To to CFR 50 AND CONTRIBUTED DATA USED IN THE A5ME CODE WHICH WAS INCORPORATED IMau arrinGIX G.
IT n;~ ALOO P207IOCO PAE7 GF inE DATA BASE USED IN REGULA10RY GUIDE 1.99 CONCERNING THE EFFECT OF COPPER InPURITIES ON SENSITIVITY OF STEEL TO IRRADIATION AND 15 EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS 4
10 REGULATCRY GUILE 1.99.
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P90G8 A9 QF f f CE ACM92*i4_RM P.0 T EN T I A L UVILIZATICO OR VALUE OF RESEARCH RE:.15 IN THE REGULATORY PROCES3 QXL **
2 DaTE ISSUED
- 05/22/74 RES DECISION UNLL SEISMIC, ENGINEERING & Sil6 SAttfY t
QJL TITLE: SEI5M3 TECTONIC MAP OF THE EASTERN UNITED STATES
$PON50 RING OFFICE (5)s RES R122 3-2 GE030GY & SEISMIC CHARACTERISTICS RESEARCH PROJECT MGR J. HARBOUR P{1_{Q55{NTir RIL 2 REPORTS A COMPILATION OF EARTHQUAKE FAULT DATA FOR THE EASTERN UNITED STATES. THESE DATA ARE USED BY LICENSE APPLICANTS IN PREPARATION OF MATERIAL FOR PRELIMINARY SAFETY ANALYSIS REPORTS.
1 USER DISCUSSION POSITION COMMISSION ACR5 PRESS OFFICE MEETING PAPER 3RIEFING BRIEFING RELEASE RESULT 5 l
ER$T FIL ACIIV111Il PEVIEW HELD COMPLETED HELD HELD J*5UED IMPLENENTED OttICE RESPONSIBLE......... NRR/5D MRR SCHEDULED COMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UN5CHED UNSCHED 1974 ACTUAL COMPLETION DATE.....
09/27/77 1974 NPR CEM!ENTS ON 0?(27/77. R. DENI5fr Pf1EBIBE APPLICATION TJ PigyLAlpFY PROCf31* THIS RESEARCH DEVELOPED LIMITED SEISMOTECTONIC PROVINCES MAPPING OF THE EASIERN UNITED STATES AIMF.P AT IMPLIMENTING APPENDIX A REQUIREMENTS FOR DETERMINING SEISMIC DESIGN FOR NUCLEAR FACILITIES.
SEliRllEJMPACT CE_PE}VLl}s THE RESE4RCH ADDED VERY LITTLE TO OUR KNOWLEDGE OF EARTHQUAKE PROCESSES IN THE LASIERN UNIILD SIAILS WHICH COULD BE USED TO IMPLEMENT APPENDIX A REQUIREMENTS.
i SQE5131}ff Effff33 CURRENT EARTHQUAKE MONITORING AND TECTONIC STUDIES SHOULD BE SYNTHESIZED TO DEVELOP AN APPLICABLE SLISM0 TECTONIC MAP.
I E C955fMT5 ON 09fl]/78, G.
PIVENg4_PK THIS STUDY IS PRL5fMILY USED A5 GUIDANCE IN ASSESSING TECTONIC PROVINCES AND SEISMICITY IN LICENSING CASE REtIEMS IN THE EASTERN UNITED STATES. THIS STUDY, A5 WELL A5 DNGOING l
STUDIES, WILL EVEATUALLY BE USED 70 0FFER GUIDANCE ON A REGIONAL BASIS THROUGH REGULATORY GUIDES.
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P90CDA" CFFICE Ct""tNf5 ON FOTENilst UTILIZATION CR VALUE OF RESEARCH RESULT 5 IN THE REGULATORY PROCESS PIL st 3
DaTE 155UED: 08/07/76 PLS DECISIOM UNIT: PRIMARY SYSTEMS INTEGRITY Pit TITLE:
CPNL V-7 INTERMEDIATE VESSEL TEST RESULTS (ORNL HSST PROGRAM)
SPON50 PING OFFICE (5):
RES PES:
1-20 VE5SEL INTEGRITY RESEAPCH PROJECT MGR C. SERFAN gE1_SQ55E$T}: ADDITIONAL EXPERIMENTAL EVIDENCE 15 DEMONSTRATED THAT A "SAFE" FAILURE MODE FOR REACTOR PRESSURE VLSSLLS FAY EXIST, NAMELY. "LEAK-BEFORE-BREAK".
THESE RESULTS SUPPORT EARLIER RESULTS REPORTED IN RIL 81.
IN THIA TEST. THE FLOW WAS 53-INCHES LONG AND 5-INCNES DEEP IN THE 6-INCH THICK PRE 55URE VESSEL WALL.
THE VESSEL WAS ABLE TO SUSTAIN TWICE THE DESIGN LOAD FRIOR TO PENETRATION OF THE FLOW THROUGH THE REMAININC THIN LIGAMENT OF VESSEL MATERIAL.
THIS RESEARCH CONTRIBUTE
- TO OUR UNDERSTANDING OF VESSEL FAILURE MODE.
THIS INCREASED UNDER5TANDING SUPPCRT5 THE MER STAFF IN IT5 REVIEW EFFORTS.
USER DISCUSSION POSITION COMMIS$ TON ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULT 5 E051_'lL_ACT1Y111E2 REY 1LW FELD COMPLETED HELD HELD ISSUED IMPLEMENTED OfflCE RESPCN51BLL.........
hkR/LD NRR SCHEDULED COMPLETION DATE., --
IIN5CHED UNSCHED UNSCHED UMSCHED UNSCHED N/A ACTUAL COMPLETION DATE.....
09/09/77 N/A IL'PmCfr EU15JU1LCILZ7 _M*EES:
2 P!5 Chi $t_!PtL1 CATION TQ_ELLVtat0?Y PDOCES3r THE TEST RESULTS HAVE BEEN USEFUL IN ESTIMATING THE MARGIN OF SAFETY. IN ILvMS OF FLAW 512E. f0R A LAkGE FLAW IN A HEAVY SECTION STEEL VE5SEL TESTED AT UPPER SHELF TEMPERATURES. WHILE CRACK ADVANCEMENT THROUGH THE FEMAINING WALL LIGAMENT WAS IN THE STABLE TEARING MODE. FRCDUCING A LEAK. AND PRESSURE TO CAUSE LEAKAGE WA5 MORE THAN TWICE THE YE55EL DESIGN PRES $URE, THESE RESULTS MAY HAVE BEEN INDUCED BY THE SPECIAE EFFECTS OF THE TEST GEOMETRY.
PE51RlCE_J5fACl_Qf PE5VLT3: WHILE ADDING TO OUR UNDERST ANDING OF VESSEL FAILURE MODES. THERE ilA5 BEEN NO DEFINIIIvt IMPACT ON LICLNSING AT THIS TIME.
C9EMEN15EEfEAEf3:
THE OPINIONS EXPRESSED IN THE RIL REGARDING LEAK-BEFORE-EREAK ARE ENCOURAGING BUT NEED FURTHER Sba5TANIIAIION BEFORE CURRENT LICENSING POSITIONS CAN BE RELAXED.
S D C 2"_C N T S_I"*_UL13 df, P. Pawpalt INIS PROGRAM HA5 GLhtRATED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PPE55UPE VESSELS WHEN FLAWS ARE PRESENT. CONSIDERING MATERIAL PROPERTIES AND LOADING 5.
THIS PROGRAM HAS PPOWIDED INPJT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRE 55URE vtSSEL MATERIALS DESCEIBED IN APPENDIX G T3 IS CFR 50 AND CONTRIBUTED DATA USED IN THE ASME CUDE WHICH WAS INCORPORATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE DATA BASE USED IN REGULATORY GUIDE 1.99 CON"ERNING THE EFFECT OF COPPER IMPURITIES ON SENSITIVITY OF STEEL TO IFPADIATION AND 15 EXPECTED TO. ' VIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS TO REGULATORY GUIDE 1.99. _ - _ _ _ _ _ _ _
, ~ _ - - _ -
PPOGRA" OUICE C0"ENTS O f POT Emil AL UTILIZATIO 10R VALUE OF LESEfRCH RESULTS IN TME PEGULATORY PROCESS h
4 DATF ISSUED 09/10/74 RES DECISION UNIT 8 TEI5MIC. ENGINEERING 1 SITE SAFETY RIL TITLFS MAP SHOWING RECENCY OF FAULTING IN COASTAL SOUTHERN CALIFORh!A SPON50 PING OFFICE (5)
RES 52 3-2 GE0 LOGY 8 SEISMIC CHARACTERISTICS RESEARCH PROJECT MGR*
J. HARBOUR P L.CQTENY52 RIL 4 COVERED FAULTING IN COASTAL SOUTHERN CALIFORNIA AND 15 2EFERRED TO BY NRR EHEN CONSIDERING l
i
$LISMIC SAFLTY QUESTIONS REGARDING PLANTS IN THE AREA COVERED.
USER DISCUSSION POSITION COMMISSION ACR5 PRESS 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS J
E95L5f!L ACTiv!J1Q PLyl[y HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... MER/5D NRR SCHEDULED COMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 1975 ACTUAL COMPLETION DATE.....
09/27/77 1975 NE_q9""lt i A PP L I C AIJ ONEMILONJ/ZJ/ 77,__ R. DIN 15E Digi 10 k.t WLAlpff PPOCE53: TO BE USED IN SELECTING SITES FOR NUCLEAR FACILITIES AND i
FOR REGIONAL INPUI 10 STAFF FLVIEW OF NUCLEAR FACILITY APPLICATIONS.
Pl5CP_IB E If"P A C T_O F P E5tJ11$ r WORK PROVIDED NEW KNOWLEDGE OF THE DISTRIBUTION OF ACTIVE FAULTS IN CCASIAL CALIFORNIA. HA5 PROVIDED INPUT TO CUR REVIEW OF DIABLO CANYON AND OTHER SITES.
$_0M"ENTS/PEMAPF5s WORK HA5 LARGELY BEEN COMPLETED. ND EXTENSION IS ANTICIPATED.
SJLCCTr wNT.N y g ge,ze, s. PIvgNyaRK THE OBJECT OF THIS SIUDY WAS 10 TRODUCE DATA DESCRISING THE RELATION 5MIP BETWEEN EARTHQUAKE i
l MAGNITUDES AND DIMENSIONS OF FAULT DISPLACEMENT. THIS STUDY, ALONG W!TH OTHERS. 15 USED f
l A5 GUIDANCE IN ASSESSING
- CAPABLE FAULT 5" IN LICENSING REVIEWS. RIL s4 PROVIDED A MAP WHICH UDPATED INFORMATION ON FAULTING IN COASTAL SOUTHERN CALIFORNIA.
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P u D G e> M Cf f 1C E CD"3EN15 ON POTENTIAL UTILIZATION 09 V4LUE OF RESE*RCH RESULTS IN THE PtGULATORY PROCESS EIL st 5
DS.TE 155UED2 06/28/76 FE5 DECISION UNIT: PRIMARY SYSTEMS INTEGRITY PIL TITLE: CONFIRMATCRY PRESSURE VE5SEL lEST, UNDER PNEUMATIC LOADING (ORNL H551 PROGRAM)
SPON509.ING OrFICE(5):
RES PPS:
1-20 WESSEL INTEGRITY RESETRCH PROJECT MGRr C. SERFAN
- E y Cr"LWT58 RIL'S 1.
38 5 PEPORTED THE RESULTS FROM THE HEAVY SECTION STEEL TECHNOLOGY PROGRAM WHICH SHOWED THAs 1HE ANALYTICAL METHODS FOR PREDICTING FLAW INITIATION AND CRACK ARREST IN REACTOR PRES 5URE VESSELS HAVE BEEN WEtt V ALIDATED. THESE VALIDAT[D ANALYTICAL METHODS (LINEAR ELASTIC FRACTURE MECHANICS AND ELASTIC-PLASTIC FRACTURE MECHANICS) ALLCW A PREDICTION OF THOSE CONDITIONS UNDER WHICH FLAWS IN PRES 5URE VESSEL STEELS CAN CAUSE FAILURE OF THE VESSEL.
THIS PROVIDES THE NRR STAFF WITH A NEW TECHNIQUE FOR SETTING 5AFE LIMITS FOR NORMAL OPERATION AND FOR ABNORMAL AND ACCIDENT SITUATIONS TO REDUCE THE LIKELIHOOD OF PRESSURE VESSEL FAILURE.
I USER DISCUSSION POSITION COMMISSION ACR5 PRESS l
OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS l
EDSL5'11_8311Vlflli FIVlEW NELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLt......... N 6, R / S D NRR SCHEDULED CCMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UN5CHED UNSCHED UNSCHED L
ACTUAL COMPL[ TION DATE.....
12/16/77 1976 j
hPL C9""EN E C" 12['[9 f' / 7 7.
J.
FWlGH'8 M5J ulPldtt3L a
0N_T y lpyLgigy_geggg3$2 THE INTERMEDIATE TEST VESSELS. ITV-7 AND ITV-7A UNDER SUSTAINED L U ADING ULMONS TF A T LD THAT BOTH THL WESSELS RESPONDED TO PNEUMATIC 10ADING ESSENTIALLY AS THEF HAD TO HYDRAULIC l
LOADING.
EARLIER EXPERIENCE OF A RAPID CRACK EXTENSION IN THE GAS PIPE LINE WAS INHIBITING THE ACRS i
COMMITTEE ABOUT THE RESULT 5 OF ITV-TEST UNDER HYDRAULIC LOADING.
HENCE. THE PNEUMATIC LOAD TEST WA5 SUGGESTED.
DESCP.JEE IPPACT_OF *ESM1132 !MESE TESTS SHOW THAT THE VE5SELS UNDER SUSTAIN 2D LCADING BEHAVED SIMIL ARLY Yo l
HYLRAULIC LOADING. AND THF FESULT5 ARE APPLICABLE TO THE EVALUATION OF THE BEHAVIOR OF REACTOR PRES 5URE VE55Eis UNDER SUSTAINED LOAD.
THE TWO VE5SEL TESTS THAT RUPTURED. WITH5T00D PRESSURE 2.15 TO 2.75 TIMES DESIGN PRESSURE.
l THE TEST PRESSURES WERE ABOVE ASME B 8 P V CODE ALLOWABLE FOR FAULTED CONDITICHS.
i C OM9 EN T S/ 8 EM A P.r.5 : TESTS DEMONSTRATE THE LEAK WITHOUT BURST.
1R_CDENIMT5_DN_?3fal'2f; P. 'AEDALLI t
THIS PROGRAM HA5 GLNLRaftD SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBk!TTLEMENT AND MARGINS I
TO FAILURE OF REACTOR PRES 5URE VE5SELS WHEN FLAWS ARE PRESENT. CONSIDERING MATERIAL PROPERTIES I
AND LOADINGS. THIS PROGRAM HAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF L
PRES 5URE VESSEL MATERI ALS DESCRIBED IN APPENDIX G TO to CFR 50 AND CONTRIBUTED DATA USED IN THE ASME CODE WMICH WAS INCORPORATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE DATA B ASE *15ED IN REGUL ATORY CUIDE 1.99 C0% ERMING THE EFFECT OF COPPER IMPURITIES ON SENSITIVITY j
OF STEEL TO IRRADIATION AND 15 EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS TO REGULATORY GUIDE 1.99.
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POOfRAM OFFICc COMMENVS ON POTENIAAA UTIL22ATION OR VALUR_OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL s:
8 DATE ISSUED: 01/31/77 RES DECISION UNIT: FUEL BEHAVIUR l
RIL TITLE: DECAY HEAT DATA APPLICABLE TO LOCA EVALUATION SPONSCRING OFFICE (S)2 RES RRQ:
1-11 DECAY HEAT RESEARCH PROJECT MGR:
R. DISALVD RES COMMENTS: RECENT CALCULATIONS AND EXPERIMENTS WERE PERFORMED TO DETERMINE A BEST ESTIMATE OF THE VALUE OF THE RESIDUAL (DECAY) HEAT GENERATED IN NUCLEAR FUEL AFTER SHUTDOWN. THE VALUE CHOSEN FOR RESIDUAL HEAT RATE PLAYS A VERY IMPORTANT PART IN PREDICTING THE PERFORMANCE OF ECC SYSTEMS. RESULTS INDICATE THAT THE DECAY HEAT RATE CURRENTLY USED HAS A 2r% MARGIN OVER THE NEWLY DETERMINED VALUE.
THIS INDICATES A SIGNIFICANT CONSERVATISM IN CURRENT LICENSING BASES AND THEY ARE NOW BEING REEXAMINED.
RIL'S 8 & 9 ARE BOTH UNDER REVIEW BY HRR FOR PDSSIBLE INCLUSION IN A MCDIFICATION TO THE EMERGENCY CORE COOLING SYSTEMS (ECCS) RULE PROVIDED IN APPENDIX K OF to CFR 59.
THESE kIL'S REPORTED ON COMPLETED RESEARCH PROJECTS IN FISSION PRODUCT DECAY HEAT AND ZIRCALOY OXIDATION WHICH RESULTED IN DATA BASES AND C'RRELATIONS WHICH INDICA 1E THAT THOSE IN CURRENT USE IN THE SCCS RULE ARE HIGHLY CONSERVATIVE.
A CCMMISSION PAPER WAS PREPARED (SECY 77-368) ON JULY 1,
1977, PRESENT STAFF PROGRESS ON A PROPOSED ACTION PLAN REGARDING POSSIBLE MODIFICATION TO THE ECCS RULE.
THE COMMISSION PAPER (SECY 78-26) PRESENTING THE PROPOSED ACTION PLAN WAS PREPAkED ON JANUARY 18, 1978.
USER DISCUSSION POSIT 1cN COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELE/SE RESULTS POST RIL ACTIVTffES REVIEW HELD COMPLETED EEL D Ell D ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR,SD NRR NRR,SD,RES HRR,SD,RES NRR SCHEGULED COMPLETION DATE.. --
UNSCHED DEFERRED 9/0/77 1978 ACTUAL COMPLETION DATE.....
08/21/78 9/10/77 7/26/77 9/9/77 NPR COMMENTS ON 09/19/77, 08/21/78, D.
RQSSe DESCRIBE APPLICATION TO REGULATORY PROCESS:
1 NEW DATA IS BEING CONSIDERED AS PARTIAL BASIS FOR MODIFICATION OF PRESENT ECCS RULE (10 CFR S0.46 AND APPENDIX K).
PROPOSED OPTIONS FOR RULE CHANGE INCLUDE ACCEPTABILITY OF NEW DECAY HEAT DATA IN ECCS LICENSING CALCULATIONS.
2.
NEW DATA WILL BE INCORPORATED IN "BEST ESTIMA*E" ANALYSIS CODES.
THESE CCDES ARE INTENDED TO BE USED IN PROBABILISTIC ASSESSMENT OF SAFETY MARGINS.
pESCRIBE IMPACT OF RESULTS:
1 NEW DATA CONFIRMS MARGINS ALLOTTED IN PRESENT ECCS RULE ARE CONSERVATIVE.
2.
IF ACCEPIED FOR USE IN LOCA LICENSING CALCULATIONS, VENDORS COULD USE ADDITIONAL MARGIN IN DESIGN.
(I.E.,
INCREASE PEAK KW/FT, ETC.).
SCMMENTS/ REMARKS: THE COMMISSION APPROVED THE PROPOSED ACTION PLAN TO MODIFY APPENDIX K (SECY 78-26).
NRR AND RES ARE PREPARING A REQUEST TO SD TO INITIATE RULEMAKING TO MODIFY APPENDIX K.
SD COMMENTS ON 09/01/77, V.
PANCIFRA:
THESE SIUDIES PROVIDE INFORMATION NECESSARY TO ASSESS THE DEGREE OF CONSERVATISM OF TdE DECAY HEAT ASSUMFTIONS IN 10 CFR SO APPENDIX K 'ECCS EVALUATION MODELS).
t
.m.
m
_.-m i
t i
l 4
Py'. GRAM UH_f;;J CJN"ENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS j
RILs:
7 DATE ISSUED: 03/14/77 RES DECISION UNIT: FUEL BEHAVIOR f
.RIL Till - HlCH TEMPER ~TURE OXIDATION OF ZIRCALOY FUEL CLADDING IN STEAM SPONSO3 4". CCFICE($12 RES EES:
i'-S ZIRCALOY CLADDING RESEARCH PROJECT NGR M. L. PICKLESIMER l
RES C0F-EMTS: THE REPORTED RESULTS IMPROVE OUR UNDERSTAHDING OF THE BEHAVIOR OF ZIRCALOY FUEL CLADDING IN AN ENVIROM ENT REPRESENTED BY A LOSS OF COOL ANT ACCIDENT (LOCA).
THEY INDICATE THAT TMcRE WOULD BE SIGNIFICANTLY LESS i
4 DEPTH 05 EMBRITTLEHEhT IN THE FUEL CL ADDIMG WALL, CALCULATED FOR ANY POSTULATED LOCA t.ND MORE WALL MATERI AL WILL BE 1
LEFT WH1 H IS CAPABLE OF SUSTAINIhG LOADS LATER IN THE SEQUENCE OF SUCH AN ACCIDENT. THIS INFORMATION TOGETHER I
WITH DAT. ON 1HE RATE OF GRCWTH OF OXIDE AND OXYGEN - STABILIZED LAYERS IN CLADDING, PROVIDES A MORE SCIENTIFIC BASE FOR ESTA6LISHiNG FUEL CLAD EMBRITTLEMENT CRITERIA FOR EMERGENCY CORE COOLING SYSTEM ACCEPTfNCE. THESE RESULTS
[
CONFIRM THAT THERE IS A DEGREE OF CONSERVATISM IN THE EVALUATIGN MUDEL SEING USED BY THE REGULATORY STAFF FOR
]
CALCULATING THE OXIDATION OF ZIRCALOY DURING A POSTLLATED LOCA.
THIS IaFORMATION IS BEING CONSIDERED AS A PARTIAL BASIS FOR MODIFICATION OF THE PRESENT ECCS RULE (10CFR SQ, APPEMDIX KI.
RIL'S 8 1 9 ARE BOTH UNDER CONSIDERATIDH BY NRR FOR POSSIBLE INCLUSI0M IN A MODIFICATION TO THE EMERGENCY CORE COOLING SYSTEMS (ECCS)
RULE PROVIDED IN APPENDIX K OF to CFR 50.
THESE RIL'S R(PORTED ON COMPLETED RESEARCH PROJECTS IN FISSION PRODUCT DECAY HEA( AND ZIRCA'_SY OXIDATION WHICH RESULTED IN DAT A c ASES AMD CORRELATIONS WHICH INDICATE THAT THOSE IN CURRENT USE IN THE ECCS RULE ARE HIGHLY CONSERYATIVE. A COMMICSION PAPER WAS PREPARED (SECY 77-368)
ON JULY 1,
1977, PRESENT STAFF PROGRESS ON A PROPOSED ACTTON PLAN REGARDiHG POSSIBLE MODIFICATION TO THE ECCS i
RULE. AND THE COMMISSION P??ER (SECY 34-26) PRESENTIhG THE PROPOSED ACTION PLAN WAS PREPARED ON
]
JANUARY 18, 1978.
RIL 59 HAS BEEN REVIEWED BY NRR, NMSS, IE, AND SD.
THIS MATERIAL WAS INCLUDED IN THE 1
BRIEFING OF THE ACRS FULL COMMITTEE BY NRR IN SEPTEMBER, 1977, CONCERNING A POSSIBLE REVISION OF APPENDIX.4 to CFR 50.
THE ACRS DISCOURAGED A REVISION OF THE ECCS RULE AT THIS TIME.
i i
USER DISCUSSION POSITION C0f MISSION ACFS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE
'RESULTS POST RIL A CTIVI T_LES REVIEH HELD
[RMPLETED HELD HELD ISSUED IMPLEMENTED 0FFICE RESPONSIBLE......... NRR/SD NRR NRR SCHEDULED COMPLETION DATE.. --
UNSCHED DEFERRED 9/0/77 UNSCHED UNSCHED 1978 ACTUAL COMPLETION DATE.....
09/28/77 9/10/77 1978 I
NRR COMMENTS OK 01/21/78 08/21/78.
D.
RQSS:
Q3 SCRIBE APPLICATION TO REGULATORY PR0 CESS:
1.
NEW DATA MAY BE CONSIDERED AS PARTIAL BASIS FOR
}
MODIFICATIONS OF PRESENT ECCS RULE (10 CFR 50, APPENDIX K).
j 2.
NEW DATA MAY BE USED IN "BEST ESTIMATE" CODES.
THESE CODES ARE INTENDED TO BE USED IN t
PROBABILISTIC ASSESSMENTS OF SAFETY MARGINS.
l DESCRIBE IMPACT OF RESULTS:
1 NEW DATA CONFIRMS MARGIN ALLOTTED IN PRESENT ECCS RULE IS l
CONSERVATIVE.
]
2.
IF ACCEPTED FOR USE IN LOCA LICENSING CALCULATIONS, VENDORS COULD USE ADDITIONAL NARGIN IN DESIGN CI.E.,
INCREASE PEAK KW/FT).
l COMMENTS / REMARKS: THE COMMISSION APPROVED THE PROPOSED ACTION PLAN TO MODIFY APPENDIX K (SECY 78-26).
NRR AND RES ARE PREPARING A REQUEST TO SD TO INITI ATE RULEMAKING TO MODIFY APPENDIX K.
SD COPMENTS ON 09/13/78, V.
PANCIERA:
THIS PROJECT PROVIDES DATA ON THE AMOUNT OF METAL-WATER REACTION AND HYDROGEN GAS GENERATION FOLLOWING A LOCA AND AIDS IN THE DEVELOPMENT OF REVISIGHS TO REGULATORY GUIDE 1.7 AND IN ASSESSING THE CONSERVATISM OF ASSUMPTIONS IN 10 CFR 50 APPENDIX K.
i l t
PROGRAM OFFICE COMMENTS ON POTENTIAt UilAIZATION OR VALUE OF RESEARCN RESULTS IN THE REGULATORY PROCESS RIL **
10 DATE 1554f2: 02/25/77 RES DECISION UNIT: RISK ASSESSMENT / PRIMARY SYSTEMS INTEGRITY RIt TITLE: PRESSURE VESSEL FAILURE PR08 ABILITY PREDICTIDH (OCTAVIA CODE)
SPONSORING OFFICE (S): RES REQ 1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR W. VESELY EIS COPNENTS: OCTAVIA IS A COMPUTER CODE FOR PREDICTING FAILURE PROBABILITIES IN REACTGR PRESSURE VESSELS AS A FUNCTION OF TEMPERATURES AND PRESSURES THAT MIGHT OCCUR DURING STARTUP AND SHUTDOWN OPERATIONS. EFFECTS OF MATERIAL PROPERTIES AND OPERATING AGE WERE ALSO INCLUDED. THE RESULTS INDICATED THAT EXISTING SAFETY MARGINS COULD BE SIGNIFICANTLY REDUCED WITH CONTINUED AGING (RESULTING IN RADIATION EMBRITTLEMENT) 0F THE REACTCR VESSEL, IF THE RATE OF OCCURRENCE OF OVERPRESSURE EVENTS CONTINUES TO BE SIMILAR TO THAT PREVIOUSLY OLSERVED. FURTHER IMPROVEMENTS IN THE CODE ARE REPORTED ON IN RIL 12.
THESE RESULTS WERE USED BY NRR DURING THIS PAST YEAR IN REVIEWING THE PROBABILITY OF FAILURE OF REACTOR PRESSURE VESSELS DUE TO SUCH EVENTS AND TO GIVE GUIDANCE ON PLANT IMPROVEMENTS THAT MAY BE NEEDED. THEY MAY ALSO BECOME THE BASIS FOR REVISING REGULATORY GUIDES OR STANDARDS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFINC RELEASE RESULTS POST PIL ACTIVITIES REVIEW HELD OFFICE-RESP 0hSIBLE.........
NRR/SD COMPLF,1ED HELD HELD 11)UED IMPLEMENTED NRR SCHEDULED COMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 1977 ACTUAL COMPLETION DATE.....
09/09/77 1977 NPR COMMENTS ON 05/07/77.
B. GRIMES:
DESCRIBE APPLICATION TO REGULATORY PROCESS: THE OCTAVIA COMPUTER CODE HAS BEEN USED TO EVALUATE THE PROBABILITY OF REACTOR VESSEL FAILURE FRGM OVERPRESSURE TRANSIENTS WHICH CAN OCCUR DURING PWR OPERATION. THE RESULTS OF THE ANALYSES INDICATED THAT EXISTING SAFETY MARGINS COULD BE SIGNIFICANTLY REDUCED WITH CONTINUED NEUTRON IRRADIATION OF THE REACTOR VESSEL IF THE HISTORICAL FREQUENCY OF OVERPRESSURE EVENTS CONTINUED.
DESCRIBE IMPACT OF RESULTS: THE ANALYSES ENABLED NRR TO CONFIRM, IN A MORE RIGOROUS, QUANTITATIVE MANNER, INITIAL LICENSING DECISIONS TO REDUCE THE FREQUENCY AND MAXIMUM PRESSURE OF THE TRANSIENTS. THESE DECISIDHS RESULTED IN MODIFICATIONS TO OPERATING PROCEDURES FOR REDUCING TnE FREQUENCY OF TRANSIENTS AND THE INSTALLATION OF PHYSICAL DEVICES TO LIMIT PRESSURES IN OPERATING PLANTS TO THOSE SPECIFIED BY THE PRESSURE-TEMPERATURE LIMITATIONS IN THE TECHNICAL SPECIFICATIONS.
COMMENTS / REMARKS: SEE COMMENTS TO RIL 812.
SD COMNENTS ON 09/13/78, P.
RANDALL:
THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE CF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES AND LDADINGS.
THIS PROGRAM HAS PROVIDED INPUT TO HRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR 50 AND CONTRIBUTED DATA USED IN THE ASME CODE WHICH WAS INCORPORATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE DATA BASE USED IN REGULATORY GUIDE 1.99 CONCERNING '7E EFFECT OF COPPER IMPURITIES ON $~NSITIVITY OF STEEL TO IRRADI ATION AND IS EXPECTED TO PROVIDE. ",*JBSTANTI AL INPUT FOR FUTURE REVISICNS TO REGULATORY GUIDE 1.99.
J PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL s:
11 pATE ISSUED: 09/15/77 RES DECISION UNIT: RISK ASSESSMENT RIL TITLE: IEEE NUCLEAR REL3 ABILITY DATA MANUAL SPONSORING OFFICE (S): RES RE28 NONE RESEARCH PROJECT MGR:
J. JOHNSON REl_s0MMENTS: A FAILURE RATE DATA MANUAL WAS DEVELOPED WHICH CAN BE USED Ih RISK AND RELIABILITY ANALYSIS OF REACIOR SYSTEMS.
THE MANUAL CONTAIh3 FAILURE RATES AND FAILURE MODE INFORMATION FOR OVER t,0C0 ELECTRICAL, ELECTRONIC AHD SENSING COMPONENTS USED IN NUCLEAR POWER PLANTS.
A METHOD IS GIVEN FOR COLLECTING AND PRESENTING RELIABILITY DATA FOR GUANTITATIVE RELIABILITY AND AVAILABILITY EVALUATIONS OF SAFETY-RELATED NUCLEAR PLANT SYSTEMS.
UNCERTAINTY BOUNDS ARE ALSO GIVEN FOR EACH ESTIMATE OF A COMPONENT FAILURE RATE.
THIS WORK IS PART OF A CONTINUING EFFORT TO ESTABLISH AN INTERIM D,*.'A BASE FOR USE IN MEETING NRC NEEDS IN THE ELECTRICAL AND ELECTRONIC AREA UNTIL SIGNIFICANT OPERATING DATA ON COMPONENTS USED IN THE NUCLEAR INDUSTRY BECOME AVAILABLE USER MEElING PAPER BRIEFING BRIEFING RELEASE RESULTS DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE PDST RI.L ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR NRR SCHEDULED COMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 04/30/78 ACTUAL CGMPLETION DATE.....
10/2S/77 NPR COMMENTS QN_10/25/77.
R.
TEDESCO:
DESCRIBE APPLICATION TO REGULATORY PROCESS' THE SUBJECT RIL ANNOUNCED THE AVAILABILITY OF A FAILURE RATE DATA MANUAL.
THIS DATA SUPPLEMENIS O!HER SOURCES OF RELIABILITY PATA SUCH AS THE REACTOR SAFETY STUDY.
SUCH DATA ARE BEING USED IN RELIABILITY STUDIES THAT SUPPORT OR PROVIDE THE BASES FOR LICENSiHG REQUIREMENTS.
RE1LRIBE IMPACT OF RESULTS: THE FIRST STUDIES USING THIS DATA HAVE NOT YET BEEN COMPLETED AND THEREFORE THE IMPACT CANMOI YEI BE DETERMINED.
COMMENTS / REMARKS: THIS DATA MANUAL PROVIDES THE STRUCTURE FOR INCORPORATING NEW OR REVISED DATA AS IT BECOMES AVAILABLE.
THE MANUAL NOW CONTAINS UNDIFFERENTIAiED HARD AND SOFT DATA WHICH IS A SERIOUS IMPEDIMENT TO THE APPLICATION OF THIS DATA.
THEREFORE, CONTINUING WORK IS REQUIRED TO INCREASE THE CONTENT OF HARD DATA BY INCORPORATING THE DATA DEVELOPED THROUGH SUCH PROGRAMS AS NPRDS.
EA COMMENTS ON O S/ 2't/75, L.
AlRAM1M:
THE DATA IS DIFFICULT TO INTERPRET BECAUSE OF THE MANY AMBIGUITIES AND INCONSISTENCIES IN THE DATA MANUAL.
THE RELATION OF THE AGGREGA-D FAILURE RATES TO THE COMPONENT FAILURE RATES IS NOT MADE CLEAR AND THEY ARE OFTEN INCONSI..ENT.
IN PARTICULAR, THE USE OF GEOMETRIC AVERAGING IS NOT JUSTIFIED AND LEADS TO INCONSISTENT RESULTS. THE WAY IN WHICH HIGd AND LOW FAILURE RATES ADD UP IS NOT CONSISTENT WITH THEIR STATED INTERPRETATION AS 9STH AND STH PERCENTILES OF THE 4
FAILURE RATE DISTRIBUTION. THE EQUALITY OF THE HIGH VALUE TO THE MAX VALUE IN MANY TABLES IS ALSO IHCONSISTENT WITH THIS INTERPRETATION. THE FAI8.URE MODE TYPES AND GEFINITIONS GIVEN EXCLUDE SUDDEN PARTIAL FAILURES AND GRADUAL COMPLETE FAILURES, BOTH OF WHICH ARE EXPERIENCED IN PRACTICE.
b.. -
PPQERa9 OFFICE C7MMENTS ON 90TENilAL UVILIZATION OR VALUE OF RESEARCH REhMAVS IN VHE REGMLATORV PROCESS RIL s:
12 DATE ISSUED: 06/16/77 RES DECISION UNIT: RISK ASSESSMENT / PRIMARY SYSTEMS INTEGRITY RIL T!1LE4 MODIFICATIONS TO PRESSURE VESSEL FAILL'RE PROBA3ILITY PREDICTION (OCTAVIA CODE)
SPONS0eING OFFICE (S):
RES REQ: NONE RESEARCH PROJECT MGR:
W.
VESELY PEi_SCMMENTS: MODIFICATIONS IN THE OCT AVI A COMPUe'ER CODE REPORTED IN RIL 810 WERE MADE.
THESE MODIFICATIONS INCLUDE A CAPASILITY TO HANDLE RESIDUAL STRESS IN A REACTOR PRESSURE VESSEL WHICH CAN EITHER BE CONSTANT. OR VARY WITH FLAW SIZE; THE CODE USER CAN IMPOSE IN UPPER BOUND ON THE VESSEL TOUGHNESS AND THE CODE HAS THE CAPABILITY TO HANDLE UNCERTAINTIES IN THE TOUGHNESS. USING THE MODIFIED OCTAVIA CODE, THE MEDIAN FAILURE PROBABILITY FOR THE SURRY REACTOR PRESSURE VESSEL WAS CALCULATED TO BE S X 10-7 PER VE5SEL YEAR FOR AN OPERATING TEMPERATURE OF 119 DEGREES C AND THE LURRENT AGE OF APPROXIMATELY 2.5 YEARS. THE FAILURE PROBABILIiY INCREASES TO 3 X 10-5 PER VESSEL YEAR AFTER 40 YEARS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST PIL ACTIVITIES REVIEN HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR/SD NRR SCHEDULED COMPLETION DATE.. 03/16/77 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 1977 ACTUAL COMPLETION DATE.09/23/773/77 1977 NRR (CHNfMTS ON 09/23/77, B.
GRIMES:
REl_CRipE APPLICATION TO REGULATORY PROCE11:
(SEE RIL 810)
DElf91)E IMPACT 07 RESULTS:
(SEE RIL 816)
LOMMENTS/ REMARKS: THIS IS A MODIFICATION CF RESULTS TRANSFERRED BY RIL 810. 02-25-77.
THE ANALYSES USING THE OCTAVIA CODE ARE COMPLETE. THE GENERIC POSITIONS RECARDING OVERPRESSURIZATION PROTECTION SYSTEMS HAVE BEEN DEVELOPED AND ARE DOCUMENTED IN NUREG-0224. ' REACTOR VESSEL PRES 5URE TRANSIENT PROTECTION FOR PRESSURIZED WATER REACTORS.'
IMPLE-MENTATION OF THE POSITIONS ARE UNDERWAY.
SD COMMENTS ON 02/21/78.
P.
RANDALL - SEE COMMENTS FOR RIL 810.
THIS PROGRAM HAS GENERATED SUBSIANTIAL DATA USED IN P?EDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE FRESENT. CONSIDERING MATERIAL PROPERTIES AND LOADINGS. THIS PROGRAM HAS PROVIDED INPUT TO NRC RECUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CI? 50 AND CONTRIBUTED DATA USED IN I
THE ASME CODE WHICH WAS INCORPORATED INTO APPENDIX G.
IT H45 ALSO PROVIDED PART OF THE DATA BASE USED IN REGULATORY GUIDE 1.99 CONCERNING THE EFFECT OF COPPER IMPURITIES ON SENSITIVITY OF STEEL TO IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIA INPUT FOR FUTURE REVISIONS TO REGULATORY GUIDE 1.99.
l I
i,
t PROGRAM OFFICE CCP9ENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 8:
13 EATE ISSUED:
11/1*/77 RES DECISION UNIT: PRIMARY SYSTEMS INTEGRITY RIL TITLE: RESIDUAL STRESSES IN WELDS S*0NSORING OFFICE (S):
RES RES:
1-20 VESSEL INTE,GRITY RESEARCH PROJECT MGR:
C. SERPAN PES COMMENTS: A VERIFIED MODEL IS PRESENTED FOR PREDICTING RESIDUAL STRESSES RESULTING FROM THE WELDING OF PIPES, AND THE ESTIMATION OF RESIDUAL STRESSES PESULTING FROM WELD REPAIRS OF REACTOR PRESSURE VESSELS. THE MODEL CAN BE USED IN THE LICENSING PROCESS TO AID IN THE EVALUATION OF CRACKING THAT HAS OCCURRED IN GIRTH-BUTT WELDS IN PIPING.
IT SHOULD ALSO PROVE TO BE USEFUL IN ANY SAFETY EVALUATION OF PROPOSED REPAIRS BY WELD BUILDUP IN THE CORNER REGIONS OF PRESSURE VESSEL N0ZZLES AFTER CRACKS HAVE BEEN REMOVED, AND IN VESSEL WELD REPAIRS.
l USER DISCUSSION POSITION COMMISSIDH ACRS PRESS I
0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PQ11_RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE........ NRR/SD SCHEDULED COMPLETION ba'.E..
03/31/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.
. 08/14/78 NRR COMMENTS ON 03/14/75, J.
KNIGHl:
RESEf t_ APPL IC A T ION 10 PESQLA.J_0RY PROCESS: THE AXISOL CODE COULD BE USED AS AN AID IN EVALUATING RESIDUAL SIRESSES IN FLUED HEAD-PROCESS PIPE WELDS OF CONTAINMENT PENETRATION ASSEMBLIES AND GIRTH BUTT WELDS IN PIPING.
IT SHOULD ALSD PROVE USEFUL AS AN AID IN DEVLOPING THE NECESSARY DECISIONAL INFORMATION IN ANY SAFETY EVAuyATON OF PROPOSED WELD REPAIRS.
Ef1SR1Rf_J MP A C T OF RESQLT11 THE COMPUTER CODE AXISOL COULD EVENTUALLY BE USED AS A DESIGN TOOL BY BOTH GOVERNMENT AND INDUSIRY, 10 IMPROVE WELDING TECHNIQUES AND PROCEDURES. THE RESULTS OF THIS PROGRAM SHOULD BE BROUGH, TO THE ATTENTION OF VARICUS ASME GROUPS ENGAGED IN PREPARING CODES AND STANDARDS ON WELDED FABRICATION AND INSPECTION PROCEDURES. THOROUGH DISCUSSICH AND EVALUATION BY SUCH GROUPS AND CONSIDERABLE TRIAL USE BY INDUSTRY IS NECESSARY BEFORE FULL APPLICATION OF THE CODE IN THE LICENSING PROCESS WOULD BE APPROPRIATE.
THE INABILITY OF THE CODE TO TAKE INTO ACCOUNT THE EFFECTS OF POST-WELD HEAT TREATMENT, AT THIS TIME, IS A DRAWBACK IN UTILIZING THE CODE IN THE i
LICENSING PROCESS.
SD CQEnfNTS ON 09/13/73, H.
COBE1 IHIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES AND LOADINGS. THIS PROGRAM HAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR S0 AND CONTRIBUTED DATA USED IN THE ASME CODE WHICH WAS INCORPORATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE DATA BASE USED IN REGULATORY GUIDE 1.99 CONCERNING THE EFFECT OF COPPER IMPURITIES ON SENSITIVITY 0F STEEL TO IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS l
TO REGULATORY GUIDE 1.99.
I
PROGDA" 0FFXCE CCMMENTS ON PCTENTIAL UTILJZATION OR V At UE OF RESE ARCH QE%plV$_JN VHE REGULATORY PROCESS Ell _f:
14 DATE ISSUED:
11/09/77 RES DECISION UNIT: SYSTEMS ENGINEERING RIL TITLE: PHYSICAL SEPARATION CRITERIA FOR ELECTRICAL CABLE TRAYS (HORIZONTAL OPEN SPACE CONFIGURATION)
$PONSCRING OFFICE (S): RES ggg: 1-23 ELECTRICAL STDS &
RESEARCH PROJECT MGR:
R. FEIT FIRE PROTECTION RES CC93{NT1: THE ADEQUACY OF THE REQUIRED SPACING OF ELECTRICAL CABLE TRAYS AT NUCLEAR POWER PLANTS WAS EXAMINED TO PREVENT THE SPREAD OF CABLE FIRES. RESULTS INDICATE THAT CURRENTLY USED CRITERIA FOR CABLE TRAY SEP/ RATION APPEAR TO BE ADEQUATE FOP ELECTRICALLY-INITIATED FIRES. BUT THAT CHANGES MAY BE REQUIRED FOR FIRES DUE TO EXTERNAL IGNITION SOURCES.
THIS WORK IS APPLICABLE TO A VERIFICATION OF REGULATORY GUIDE 1.75, "PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS".
EXPOSURE FIRE TESTING EMPLOYING EXTERNAL FUEL SOURCES WAS CONDUCTED TO PROVIDE
- ATA FCR THE DEVELOPMENT OF CURRENT NRC STAFF POSITION AS DOCUMENTED IN THE APPENDIX A TO THE BRANCH TECHNICAL cOSITION APCSB 9.5-1, "GUIDELINc5 FOR FIRE PROTECTION FOR' NUCLEAR POWER PLANTS" AND IN THE DRAFT REGULATORY GUIDE 1.120. "FIRE PROTECTION GUIDELINES FOR NUCLEAR POWER PLANTS."
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RrSULTS PpST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OF FIC E R ESPCNSIB L E......... NRR,SD SCHECULED COMPLETION DATE.. 01/09/73 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
04/26/78 NPR COMMENTS OM 04/26/75, 03/21/73, R.
TEDESCO:
RESCRIBE APPLI(_ATION TO REGULATORY PRQQfSS: THE RIL PROVIDES RESULTS OF A COMPLETED PORTION OF THE NRC FIRE PROTECIION RESEARCH PROGRAM CONDUCIEP ST THE SANDIA LABORATORIES REGARDING THE EFFECTS OF CABLE TRAY SEPARATION ON THE PROPAGATION OF ELts =rCaLLY INITIATED AND EXPOSURE FIRES.
DEifRIBE IMPACT OF RESULTS: THE INTRODUCTORY PARAGRAPH OF RIL 314 STATES THAT THE RESULTS OF THE SANDIA PP0 GRAM INDICATE THAI "CURRENTLY USED CRITERIA FOR CABLE TRAY SEPARATION APPEAR TO BE ADEQUATE FOR ELECTRICALLY INITIATED FIRES BUT THAT CHANGES MAY BE REQUIRED FOR FIRES DUE TO EXTERNAL IGNITION SOURCES."
RES NOTED IN A CONVERSATION ON HOVEMBER 11, 1977 THAT THE SPECIFIC POINT RAISED IN THAT PARAGRAPH IS DIRECTED TO THE ADEQUACY OF SOLE RELI ANCE ON THE SEPARATION CRITERI A 0F REGUL ATORY GUIDE 1.75 FOR PROTECTION AGAINST EXPOSURE FIRES.
THE FIRE PROTECTION CRITERIA DEVELOPED BY THE STAFF SINCE THE BROWNS FERRY FIRE HAVc RECOGNIZED THAT RELIANCE SHOULD NOT BE PLACED SOLELY ON THE SEPARATION CRITERIA 0F REGULATORY GUIDE 1.75.
CUR STAFF REPORT, DATED NOVEMBER 9, 1977 OH "THE QUESTION OF WHE1HER THE PETITION OF THE UNION OF CONCERNED SCIEh!ISTS RAI$ES MATTERS THAT REQUIRE IMMEDIATE COMMISSION ACTION " INDICATES THE STAFF POSITION THAT THE IEEE-334 AND THE REGUL ATORY GUIDE 1.75 SEPARATION GUIDELINES, AND THE IEEE-383 FIRE RETARDANCY STANDARDS FOR SAFETY CABLES, BY THEMSELVES ARE NOT SUFFICIENT TO PROTECT AGAINST EXPOSURE FIRES.
CONSEQUENTLY WE REQUIRE ADDITIONAL MEASURES FOR FIRE PROTECTION SUCH AS: FIRE BARRIERS BETWEEN REDUNDANT DIVISION CABLE TRAYS; FIRE RETARDANT COATINGS ON CABLING; AUTOMATIC FIRE DETECTION SYSTEMS; AUTOMATIC FIRE EXTINGUISHING SYSTEMS; ADMINISTRATIVE PROCEDURES; AND TRAINING PROGRAMS. THIS POSITION HAS BEEN HELD BY THE STAFF SINCE THE BROWNS FERRY FIRE AND IS REFLECTED IN OUR STANDARD REVIEW PLAN SECTION 9.5.1 (BTP 9.5-1), REVISION.
REG. GUIDE 1.120 HAS BEEN ISSUED FOR COMMENT.
TH95. EXPOSURE FIRES ARF REQUIRED TO BE CONSIDERED IN THE FIRE PROTECTION PROGRAMS FOR BOTH OPERATING PLANTS AND PLANTS IN THE LICENSING PROCESS. IT SHOULD BE NOTED THAT THE RIL ACKNCWLEDGES THAT FURTHER TESTING WITH ELECTRICALLY INITIATED CABLE TRAY FIRES UNDER DIFFERENT CONDITIONS MAY RESULT IN FULLY DEVELOPED FIRES.
WE CONCLUDE, AS DOES RIL s?4, THAT THESE RESEARCH RESULTS CONFIRM THE NEED FOR PROTECTION MEASURES IN ADDITION TO THE SEPARATION CRITERIA. SINCE SUCH MEASURES ARE INCLUDED IN OUR PRESENT CRITERIA WE CONCLUDE THAT THE INFORMATION IN RIL 814 DOES NOT INDICATE THE NEED FOR CHANGES IN OUR FIRE PROTECTION GUIDELINES.
BUT THAT IT CONFIRMS THE NEED FOR DUR PLANS TO UPDATE REGULATORY GUIDE 1.75.
COMMENTS / REMARKS: NONE -
SD COMPINTS ON 09/13/78, G. RIVEMBAM l
THE TESIS HAVE CONFIRMED THE VALIDITY OF GUIDELINES IN REGULATORY GUIDE '.120 WHICH CALL FOR SEPARATING REDUNDANT SAFETY SYSTEM CABLING BY RATED FIRE BARRIERS. THE TESTS HAVE CONFIRMED THE VALIDITY OF SEPARATION CRITERIA SPECIFIED IN REGULATORY GUIDE 1.75 ONLY WITH RESPECT TO FIRES RESULTING FROM ELECTRICAL FAILURE WITHIN A CABLE TRAY BUNDLE.
HOWEVER, THEY SHOW THE BASIC DEFICIENCIES OF THE SEPARATICN REQUIREMENTS WITH RESPECT TO EXPOSURE FIRES COMMON TO MORE THAN ONE REDUNDANT SYSTEM CABLE RUN.
l 1 ___-___- _ -
PRRARAM AFFICE COMMENTS _DN POTENTXAL UTILX2ATXON OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 9:
15 DATE ISSUED:
12/01/77 RES DECISION UNIT: PRIMARY SYSTEMS INTEGRITY RIt TITLE: CHARACTERIZATION OF BWR FEEDWATER N0ZZLE CORNER CRACKS SPCNSORING OFFICE (5): RES ggg: 1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR C. SERPAN RES CCMMENTS: PRESSURE LOADING OF CRACKS IN THE INSIDE CORNER OF FEEDWATER INTAKE N0ZZLES FOR BOILING MATER REACTOR (BWR) PRESSURE VESSELS HAVE BEEN CHARACTERIZED. THE CHARACTERIZATION ESTABLISHED THE RELATIONSHIP BETWEEN STRESS-GENERATED PRESSURE AND MEASURABLE CRACK PARAMETERS. IN ORDER TO DETERMINE THE GROWTH OF THE CRACK AND ITS CRITICAL SIZE.
THESE RESULTS CAN BE USED TO CHECK THE CALCULATIONS.
BASED ON INTERNAL PRESSURE. FOR THE SAFETY ANALYSIS OF BWR FEEDWATER N0ZZLE CORNER CRACKS DURING SUBSEQUENT CGMBINED COOLING AND UNLOADING. IT HAS BEEN SHOWN THAT UNDER THE MOST SEVERE CONDITIONS.
THE CRACK CAN PENETRATE NO MORE THAN 1/3 0F THE PRESSURE VESSEL WALL.
THIS MEANS THAT THE VESSEL WILL ALWAYS BE CAPABLE OF RETAINING EMERGENCY COOLING WATER. THUS KEEPING THE CORE COOL AND PROVIDING FOR A SAFE SHUTDOWN.
THUS. VESSEL FAILURE IS NOT POSSIBLE FOLLOWING WARM PRESTRESSING UNDER CONDITIONS WHERE COLD EMERGENCY CORE COOLING WATER IS INJECTED INTO THE HOT PRESSURE VESSEL FOLLOWING A LOSS OF COOLANT ACCIDENT.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIE1 REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR/SD SCHEDULED COMPLETION DATE.. 03/31/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED-ACTUAL COMPLETION DATE.....
02/21/78 NPR COMMENTS ON 02/21/78. B. GRIMES:
DESCRIBE APPLICATION TO REGULATORY PROCESS: HRR CONCURS THAT THESE DATA CAN BE USED FOR PARTIAL VERIFICATION OF THE APPROXIMATE METHODS USED BY GE IN ESTIMATING THE STRESS INTENSITY FACTORS APPLICABLE TO FEEDWATER AND CRD N0ZZLE CRACKS IN BWR REACTOR VESSELS. THEY CAN ALSO BE USED FOR VERIFICATION OF MORE SOPHISTICATED METHODS OF EVALUATION (SUCH AS FINITE ELEMENT ANALYSIS) IF AND WHEN SUCH METHODS ARE DEVELOPED.
DESCRIBE IMPACT OF RESULTS: WE HAVE MADE A COMPARISON BETWEEN THE GE STRESS INTENSITY CURVE (FIG. 3-26 0F NECE-21480) FOR THE ONE CASE WHERE A DIRECT COMPARISON CAN BE MADE -- NAMELY. AT AN A/T RATIO SLIGHTLY GREATER THAN 0.50. AS PROVIDED BY FIG. 7 0F THE PROGRESS REPORT. FOR THIS ONE CASE. THERE IS ALMOST EXACT AGREEMENT BETWEEN THE GE CURVE AND THE VPI TEST RESULTS. THE GE CURVE. WHEN CONVERTED TO NORMALIZED STRESS INTENSITY FACTORS. WILL ALSO PRODUCE A CURVE (AS A FUNCTION OF A/T RATID) WHICH IS QUALITATIVELY SIMILAR TO THAT OF FIG. 7 0F VIP REPORT VIP-E-76-25; IN THIS CASE. EXACT CORRESPONDENCE IS NOT TO BE EXPECTED BECAUSE OF THE MATERIAL DIFFERENCE IN THE DIAMETER-TO-THICKNESS RATIOS OF VESSELS INVOLVED. THE CLOSE AGREEMEN! BETWEEN THE GE CURVE AND THE TEST RESULT FOR THE ONE CASE WHERE A VALID COMPARISON CAN BE MADE PROVIDES ASSURANCE THAT THE STRESS INTENSITY FACTORS USED BY GE IN THEIR EVALUATION ARE REASONABLE APPROXIMATIONS.
COMMENTS / REMARKS: WE ENCOURAGE THE COMPLETION OF THIS WORK AND PARTICULARLY THE DEVELOPMENT OF VERIFIED ANALYTICAL METHODS WHICH WILL PROVIDE AN ASSURED MEANS FOR FUTURE CALCULATION OF SIF'S FOR CRACKS IN COMPLEX GECMETRIES (SUCH AS THROUGH THE USE OF FINITE ELEMENT ANALYSIS).
SD COMMENTS ON C9/t3/78.
G.
RIVENBARK:
THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT. CONSIDERING MATERIAL PROPERTIES AND LOADINGS. THIS PROGRAM HAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR 50 AND CONTRIBUTED DATA USED IN THE ASME CODE WHICH WA3 INCORPORATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE DATA BASE USED IN REGULATORY GUIDE 1.99 CONCERNING THE EFFECT OF COPPER IMPURITIES CH SENSITIVITY l
OF STEEL TO IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS TO REGULATORY GUIDE 1.99.
l - -
~ _.
J 4
i PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS i
RIL #2 16 DATE ISSUED:
12/01/77 RES DECISION UNIT: PRIMARY SYSTEMS INTEGRITY RIL TITLE: WARM PRESTRESSING SPONSORING OFFICE (S):
RES ERS:
1-20 VESSEL INTEGRITY RESEARCH /ROJECT MGR:
C. SERPAN l
eel _SDEMEN11: THE EFFECT OF COLD EMERGENCY CORE COOLING WATER ON HOT REACTOR PRESSURE VESSELS WAS CONSIDERED. THE RESULTING THERMAL SHOCK COULD, UNDER "WORST CASE" CCNDITIONS, LEAD TO THE PREDICTION 1
THAT FLAWS IN THE STEEL PRESSURE VESSEL WOULD EXTEND.
RESULTS REPORTFD HERE PROVIDE A VERIFICATION OF
]
THE "WARM PRESTRESSING" EFFECT WHICH CAN PRECLUDE CRACK EXTENSION WHEN IT OTHERWISE WOULD HAVE BEEN PREDICTED. TO DESCRIBE THIS EFFECT, ONCE A CRACK IS LOADED WHILE THE MATERIAL IS VERY TOUGH, HO RAPID EXTENSION WILL OCCUR.
USE9 DISCUSSION POSITION COMMISSION ACRS PRESS j
OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE
- RESULTS EOST RIL ACTIVITIES REVIEy HELD COMPLETED HELD HELD ISSUED IMPLEM"MTED 0FFICE RESPONSIBLE......... NRR/SD
)'
SCHEDULED COMPLETION DATE.. 02/01/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED l
ACTUAL COMPLETION DATE.....
02/14/78 NRR COMMENTS ON 02/14/78, 08/21/78, B.
GRIMES:
p6 SCRIBE APPLICATION TO REGULATORY PROCESS: IF "WARM PRESTRESSING" AS DESCRIBED IN THE RIL IS OPERATIVE ON HIGHLY IRRADIATED REACTOR VESSEL STEELS, THIS MECHANISM COULD PROVIDE ADDITIONAL MARGIN IN THE RPV TO ACCOMMODATE i
THERMAL SHOCK ASSOCIATED WITH ECCS INJECTION DURING A LARGE LOCA.
}
RESCRIBE IMPACT OF RESULTS: BY LIMITING CONCERNS REGARDING THERMAL SHOCK TO REACTOR VESSELS TO THOSE TRANSIENTS THAT 4
INVOLVE REPRESSURIZATION OF THE VESSEL, VENDOR ANALYSES AND NRC EVALUATIONS WOULD BE SIMPLIFIED. IT WOULD ALSO PROVIDE AT LEAST A PARTIAL ANSWER TO THE QUESTIONS POSED IN REG. GUIDE 1.2, "THERMAL SHOCK TO REACTOR VESSEL."
COMMENTS /REMARKSt NRR HAS DISCUSSED THE TECHNICAL ASPECTS COVERED BY THIS RIL WITH RES PERSONNEL. RESEARCH REGARDING WARM PRESTRESSING IS STILL UNDERWAY, ESPECIALLY ITS RELEVANCE TO CYLINDRICAL VESSEL WALLS. WHILE IT APPEARS i
TO BE A PROMISING PHENOMENON FOR LIMITING CRACK EXTENSION DURING A THERMAL SHOCK. THE DATA ARE STILL INSUFFICIENT 1
TO BE USED AS A BASIS FOR LICENSING DECISIONS. A DETAILED TECHNICAL EVALUATION OF THE RESEARCH DESCRIBED IN THE RIL AND THE NRR ASSESSMENT OF ITS RANGE OF APPLICABILITY WILL BE COMPLETED FOLLOWING RECEIPT OF MORE SUBSTANTIAL DATA.
y
)
SD COMMENTS ON 09/13/73. G. RIVENBARK:
i THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS i
TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES AND LOADINGS. THIS PROGRAM HAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR 50 AND CONTRIBUTED DATA USED IN 4
THE ASME CODE WHICH WAS INCORPORATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE i
DATA EASE USED IN REGULATORY GUIDE 1.99 CONCERNING THE EFFECT OF COPPER IMPURITIES ON SENSITIVITY 0F STEEL TO IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS I
TO REGULATORY GUIDE 1.99.
i e
]
l l 1
~-
I PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH REEMLl%_JH THE REGULATORY PROCESS RIL s:
17 DATE ISSUED: 05/05/78 RES DECISION UNIT: FUEL BEHAVIOR j
- It TITLE: POWER BURST FACILITY (PBF) SINGLE ROD-POWER COOLING MISMATCH (PCM) TEST RESULTS SPONSORING OFFICE (S):
RES REQ:
1-10 PBF EXPERIMENTAL RESEARCH PROJECT MGR R. VAN HOUTEN PROGRAMS l
UES COMMENTS-.
COMPLETED RESEARCH IS REPORTED ON SINGLE FUEL ELEMENTS EXPOSED TO POWER-COOLING MISMATCH (PCM) CONDITIONS 1
IN INE POWER BURST FACILITY (PBF).
THE RESULTS ARE OFFERED FOR USE IN DETERMINING POSSIBLE CHANGES IN REQUIRED l
DEPARTURE-FROM-NUCLEATE-BOILING RATIOS (DMBR'S) FOR ALL COMMERCIAL POWER REACTORS WHICH USE ZIRCALOY-CLAD URANIUM j
DIOXIDE FUEL RODS.
THE RESEARCH RESULTS SHCW THAT ZIRCALOY FUEL ROD CLADDING HORMALLY DOES NOT FAIL EVEN WHEN PROLONGED FILM BOILING OCCURS AS A RESULT OF INADEQUATE COOLANT FLOW RATES.
THE CLADDING GENERALLY WILL NOT FAIL UNLESS IT BECOMES 50
)
HEAVILY OXIDIZED THAT IT IS BRITTLE AT ROCM TEMPERATURE. SUCH SEVERE ZIRCALOY OXIDATION WOULD REQUIRE HIGHER CLADDING TEMPERATURES THAN ARE CURRENTLY PREDICTED FOR ANY LIGHT WATER REACTOR ACCIDENTS WHICH RESULT IN A PCM, WHETHER RELATED TO A LOSS OF COOLANT FLOW OR TO AN INCREASE IN FUEL ROD POWER.
1 USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOST RTL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED
]
0FFICE RESPONSIBLE..........NRR/SD SCHEDULED COMPLETION DATE. 07/05/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED j
ACTUAL COMPLETION DATE....
08/22/78 3
i NDP COMMENTS ON 08/01/78.
D.
HOUSTON:
I QESCRIDE APPLICATION TO REGULATORY PROCE111 VARIOUS FUEL FAILURE MECHANISMS AND THE CONSEQ'JENCES OF FAILURES ARE l
EVALUATED IN THE SAFETY ANALYSIS OF TRANSIENTS AND ACCIDENTS. DEPARTURE FROM NUCLEATE BOILING (DMB) IS ASSUMED TO PRODUCE FUEL ROD FAILURE AND IS THE FAILURE CRITERION USED FOR MANY LICENSING ANALYSES. PELLET / CLADDING j
INTERACTION (PCI) CAN ALSO BE A FUEL FAILURE MECHANISM. PBF PROVIDES THE CAPABILITY FOR STUDYING FUEL BEHAVIOR AND FAILURE MECHANISMS UNDER TRANSIENT AND ACCIDENT CONDITONS.
)
DESCRIBE IMPACT OF RESQLT1: THE DEMONSTRATED ABILITY OF MOST FUEL RODS TO EXPERIENCE DMB WITHOUT FAILURE SHOWS THAT THE CURREN1 DNS CRITERION IS CONSERVATIVE. REQUESTS FOR LESS CONSERVATIVE FAILURE CRITERIA HAVE BEEN MADE BY i
THE INDUSTRY.
TURBINE TRIP WITHOUT BYPASS (TTWOB) AND STEAM LINE BREAK (SLB) ARE HEAR-LIMITING EVENTS IN WHICH l
DNB IS PREDICTED TO OCCUR MOMENTARILY YET FAILURE BY THIS MECHANISM MAY NOT OCCUR.
DEFINITION OF A LESS CONSERVATIVE FAILURE CRITERION FOR THESE EVENTS WOULD RELIEVE THESE LIMITING CONDIT0NS. BASED ON THESE PBF l
RESULTS, NRR WILL GIVE SERIOUS CONSIDERATION TO THESE VENDOR REQUESTS; HOWEVER, APPROVAL OF RELAXED FAILURE CRITERIA j
WILL BE CONTINGENT UPON THE EVALUATION OF OTHER NON-DNS FAILURE MECHANISMS.
l fft[ENTS/ REMARKS: THE SINGLE ROD PCM TEST RESULTS SHOW THAT THE CURRENT DNB FAILURE CRITERION IS CONSERVATIVE.
THE WIDE RANGE OF CLADDING TEMPERATURES DURING DNB WHEN TEST PARAMETERS ARE HEARLY THE SAME PREVENT A QUANTITATIVE ASSESSMENT OF THE MARGIN TO FAILURE. FURTHERMORE, THE PAWEL CLADDING EMBRITTLEMENT CRITERICH WOULD NEED ADDITIONAL REVIEW BEFORE A QUANTITATIVE MEASURE OF MARGIN COULD BE USED IN LICENSING MATTERS.
l THE PCM TEST SERIES WAS DESIGNED PRIMARILY TO STUDY THE EFFECTS OF DNB.
THERFORE, PELLET / CLADDING INTERACTION
(?CI) DATA FROM THESE EXPERIMENTS WERE PROBABLY COMPROMISED BY THE TEST CONDIT0NS. FOR EXAMPLE, FUEL EXPOSURE j
TIME AT PCMER, FUEL PRECONDITIONING AND CLADDING TERMPERATURE WERE NOT IDEAL FCR PCI STUDIES.
t CONCLUSIONS ON THE SUBJECT OF FUEL FAILURE PROPAGATION DRAWM FROM THE PCM SERIES MAY BE PREMATURE SINCE VENDOR FUEL ROD PRESSURE CRITERIA HAVE BEEN CHANGED. NEW DESIGN CRITERIA, RECENTLY APPROVED BY NRC, ALLOW i
FCR INTERNAL ROD PESSURE TO EXCEED THE EXTERNAL SYSTEM PRESSURE DURING NORMAL OPERATION. IN ADDITION. THE FUEL-1 ROD BEHAVIOR OF A SINGLE ROD IN A COLD SHROUD IS NOT TYPICAL OF FUEL RODS IN A MULTIPLE ARRAY.
THE FORTHCOMING BUNDLE TESTS IN PBF SHOULD GIVE MORE INFORMATON ABOUT FAILURE PROPAGATION.
SD COMMENTS, G.
RIVENEARK: NO RESPONSE RECEIVED.
i
! 1 e++--e-
_ _ - _ _ _ - _ _ = - ____ _, - -, - - -,. -..
-r
=w w
I PROGRAM OFFICE CCPMENTS ON POTENTIAL UTILIZATION OR vat 0E OF RESEARCH RESULT $ IN THE REGULATORY PROCESS PIL a:
18 DATE ISSUED:
11/09/77 RES DECISION UNIT: RISK ASSESSMENT PIL TITLE: FRANTIC COMPUTER CODE SPONSCRING OFFICE (S):
RES EES: NONE RESEARCH PROJECT MGR:
F. GOLDBERG PL1_LQ"MENT 54 THE FRANTIC CCMPUTER CODE IS USED TO CALCULATE THE UNAVAILABILITY OF ANY SYSTEM MODEL.
COMPREHENSIVE SURVEILLANCE TESTING EVALUATIONS FOR
- SYSTEM ARE POSSIBLE WITH THE INCORPORATION OF TEST DOWNTIMES, TEST INEFFICIENCIES, AND TEST-CAUSED FAILURES IN THE ANALYSIS OF SYSTEM MODELS.
THE FRANTIC CODE HAS POTENTIAL SIGNIFICANT APPLICATION IN EVALUATING TECHNICAL SPECIFICATIONS ON lESTING AND ALLOWED DOWMTIMES FOR REACTOR SAFETY SYSTEMS.
THE EVALUATIONS CAN BE OF A GENERIC NATURE, OR CAN BE APPLIED TO SPECIFIC PLANT SYSTEMS IF APPLICABLE DATA ARE AVAILABLE.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PQ1_T_PIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD OFFICE RESPONSIBLE......... NRR/SD ISSUED _
IMPLEMENTED NRR SCHEDULED COMPLETION DATE.. 01/09/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 02/01/78 ACTUAL COMPLETION DATE.....
12/06/77 01/15/78
!Lt?_C_CCENTS, NOVEMSER 1979, F.
GRLQ1ERS:
D11LL1DL_ APPLICATION TO DEGULATORY PR0(E11: THE SUBJECT RIL ANNOUNCED THE AVAILABILITY OF THE FRANTIC COMPUTER CODE.
THIS CCMPUTER CODE IS RECOGNIZED BY NRR AS AN IMPORTANT TOOL TO QUANTIFY SYSTEM UNAVAILABILITIES. THE FRANTIC CCMPUTER CODE IS BEING USED BY SCIENCE APPLICATIONS. INC., TO PERFORM SENSITIVITY STUDIES ON ALLOWABLE OUTAGE TIMES FOR ECCS COMPONFNTS AS A PORTION OF THE EFFORT UNDER CONTRACT NO. NRC-03-07-059.
Di$(R11E_1E? AC T OF RESULTS: THE SAI CONTRACT EFFORT HAS BEEN COMPLETED. THE FRANTIC CODE WAS UTILIZED TO DETERMINE THOSE FACIORS TO WHICH SYSTEM UNAVAILABILITY IS MOST SENSITIVE INCLUDING, FOR EXAMPLE. TEST DOWNTIME, REPAIR TIME, TEST EFFICIENCY, TEST OVERRIDE CAPABILITIES, POSSIBLE TEST CAUSED FAILURES AND TEST STAGGERING.
THE FRANTIC CODE HAS BEEN SHOWN TO BE POTENTIALLY USEFUL IN ESTABLISHING TECH. SPECS. FOR SPECIFIC SAFETY SYSTEM DESIGNS, HOWEVER, SAI SUGGESTED THAT ADDITIONAL, MORE COMPLETE AND COMPREHENSIVE ANALYSES BE PERFORMED IN THE FUTURE.
CQ5NfMTS/ REMAP 51: DOR EXPECTS TO UTILIZE THE FRANTIC COMPUTER CODE, EITHER DIRECTLY OR UNDER TECHNICAL ASSISTANCE CCNTRACTS. TO QUANTIFY SYSTEM UNAVAILABILITIES ON SPECIFIC OPERATING PLANTS AS 1HE NEED ARISES. THE CALCULATIONAL RESULTS MAY AID IN DETERMINING APPROPRIATE LICENSING ACTION.
NO POST RIL ACTIVITIES ARE PLANNED.
SD COMMENTS, G.
RIVENBARK: NO COMMENT. I
PROGQAM OFFICE COMMENis ON PortNf/AL UTILIZATION OR VALUE OF RESEARCH RE%ptTS IN THE REGul4?OT rKHCESS RIL s:
19 DaTE ISSUED: 01/31/78 RES DECISION UNIT: RISK ASSESSMCir PIL TITLE: GO METHODOLOGY ASSESSMENT SPONSOPING OFFICE (S):
RES EES: NONE RESEARCH PROJECT MGR J. PITTMAN RES COMMENi$a CO PROVIDES A METHOD FOR SYSTEM MODELING AND A COMPUTER CODE TO CALCULATE A PREDICTION OF SYSTEM RELIABILITr.
THE STUDY DEMONSTRATES THAT THIS METHOD PROVIDES EQUIVALENT RESULTS TO THOSE UBTAINED FROM FAULT TREE ANALYSIS, WHICH WAS USED IN THE REACTOR SAFETY STUDY.
THE MODEL RESEMBLES THE SYSTEM SCHEMATIC OR PIPING DIAGRAM W.4ICH REDUCES THE BURDEN OF MODELING ALL SYSTEM COMPCNENTS. GO HAS A POTENTIAL SIGNIFICANT USE AS A MEANS OF DETERMINING SYSTEM RELIABILITY OR AS A DIVERSE METHOD FOR VERIFYING SYSTEM ANALYSIS PERFORMED USING FAULT TREE OR SIMILAR MODELING TECHNIQUES.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER' BRIEFING BRIEFING RELEASE RESULTS POST 2]L ACTIVITIES R EVIE'd HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPGNSIBLE......... NRR/SD NMSS SCHEDULED COMPLETION DATE.. 03/31/78 UNSCHED UNSCHED UNSCHED UNSCHED I,"NSCH ED UNSCHED ACTUAL CcMPLETION DATE.....
EPR CCMMENTS. NOVEM3ER 1979, R.
TERE 1SD:
Ri}[PIBE APPLICATION 10 REQQ1ATORY PROC [il: THIS METHODOLOGY HAS NOT BEEN AND IS NOT PLANNED TO BE USED IN UPCOM REVIEWS. HGWEVER, EVEN IHOUGH THE METHODS HAVE NOT BEEN DIRECTLY APPLIED IN THE LICENSING PROCESS WE CONSIDER SAFEIY THEIR USE WOULD ADD TO THE GENERAL KNOWLEDGE OF THE HRC STAFF IN THE AREA 0F REACTOR SYSTEM RELI ABILITY EVALUATION.
THAT THERE HAS BEEN NO IMPACT OF THESE RESULTS TO DATE ON THE LICENSING PROCESS. THE EElGRLRE_IFP ACT OF TESUL TS:
OF RESOURCES TO THIS ACTIVITY HAS BEEN PRECLUDED DUE TO THE HEEP TO SERVICE HIGHER PRIORITY TASKS.
ASSIGSMENT C O MM EN T S / R EM A R r.S : NONE SD COMMENTS. G.
- _IVENB ARK: NO COMMENT.
NMSS CCMMENTS.
E.
HATTER: NO RESPONSE RECEIVED.
]
l 1
P&OGGAM OFFICE CC"MENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS QIt s: 23 DATE ISSUED: 01/24/78 RES DECISION UNIT: SAFEGUARDS RIL TITLE: A STUDY OF PHYSICAL PROTECTION EQUIPMENT SPONSORING OFFICE (S): IAE EED:
4-4 SAFEGUARDS E(UIPMENT EVAL.
RESEARCH PROJECT MGR:
E. RICHARD RES CO*MENTS: THE PURPOSE OF THIS STUDY WAS TO PROVIDE THE HRC INSPECTOR, LICENSING REVIEWER AND FIELD EVALUATOR WITH HEW AkD IMPROVED METHODS FOR EVALUATING PHYSICAL PROTECTION EQUIPMENT THAT IS IN USE OR PROPOSED FOR USE AT LICENSED NUCLEAR FACILITIES. THE FIVE MAJOR PRODUCTS OF THIS STUDY ARE:
1.
A CATALOG OF PHYSICAL PROTECTIGN EQUIPMcHT.
2.
A GUIDE FOR EVALUATION OF PHYSICAL PROTECTION EQUIPMENT.
3.
A BOOK OF REFERENCE MATERIALS (RELEVANT TO THE EQUIPMENT CATALOG AND THE EVALUATION GUIDE).
4 A SET OF GUIDELINES FOR DEVELOPING A METHODOLOGY 10 MEASURE LEVELS OF EFFECTIVENESS FOR A FIXED-SITE PHYSICAL PROTECTION SYSTEM.
5.
A
SUMMARY
REPORT, INCLUDING RECOMMENDATIONS FOR FURTHER WORK.
ALL OF THE ABOVE PRODUCTS HAVE BEEN DISTRIBUTED TO THE VARICUS HRC 2EGIONAL OFFICES AND ARE PRESENTLY BEING USED BY INSPECTORS AS BASIC REFERENCE DOCUMENTS FOR EVALUATING PHYSICAL PROTECTION EQUIPMENT INSTALLED AT LICENSED HUCLEAR FACILITIES. DATA FROM THESE DOCUMENTS WERE ALSO USED IN THE DEVELOPMENT OF A NEW NRC REGULATORY GUIDE ON INTERIOR INTRUSION DETECTION ALARM SYSTEMS BY THE CFFICE OF STANDARDS DEVELOPMENT. OTHER FEDERAL AGENCIES HAVE REQUESTED THE RESULTS OF THIS STUDY AS A MEANINGFUL COMPENDIUM OF AVAILABLE PHYSICAL PROTECTION EQUIPMENT AND EVALUATION EQUIPMENT TECHNIQUES. THE RESULTS OF THIS STUDY WILL BE USED IN PHASE II 0F THIS SAFEGUARDS RESEARCH PROGRAM AS A BASIS FOR EXPANDING AND IMPROVING THE DATA AVAILABLE TO NRC STAFF REGARDING THE CHARACTERISTICS AND EFFECTIVENESS OF CCMBINATIONS OF PHYSICAL PROTECTION EQUIPMENT, AND THEIR ASSOCIATED ADMINISTRATIVE AND OPERATIONAL PROCEDURES.
THE DIVISIGN OF DOCUMENT CONTROL HAS BEEN REQUESTED TO PRINT THESE REPORTS FOR DISTRIBUTION ONLY TO OTHER AGENCIES AND NRC STAFF.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOST_Rit ACTIVITIES R EV11W HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE REaPONSIBLE......... NRR/hMSS/ --
S D/IS E SCHEDULED CCMPLETION DATE.. 03/31/78 UNSCHED 0-iCHED UNSCHED UNSCHED UNSCHED UNSCHED l
ACTUAL CGMPLETION DATE.....
1 I4E CO**ENTS, C. WEIS: MO RESP 0 HSE RECFIVED.
i ip_00*"*E N T S ON__0 2/72 n $.
R.
J ONS :
IHIS PROGRAM CONDUCTLD BY IHL MITRE CORPORATION RESULTED IN A MULTI-VOLUME REPORT ON PHYSICAL PROTECTION HARDWARE INCLUDING DESCRIPTIOMS, PERFORMANCE CHARACTERISTICS AND MANUFACTURERS SPECIFICATIONS.
THIS WORK HAS MATERIALLY ASSISTED IN THE PREPARATION OF SEVERAL TECHNICAL REPORTS ON SPECIFIC TYPES OF HARDWARE AND HAS PERMITTED THE CANCELL ATION OF TWO PLANNED REPORTS ON ITEMS THAT WERE ADEQUATELY COVERED IN THE MITRE REPORT.
THE REPORT ALSO IS TO BE REFERENCED IN THE DESIGN GUIDANCE REPORT NOW BEING PREPARED Ih CONNECTION WITH THE FUEL CYCLE FACILITY SAFEGUARDS UPGRADE RULE.
NM35 CCMTENTS, B.
HATTER:
N/A - REPORTS WEPE REVIEWED. HMS$ PLANS TO REFERENCE PARTS OF THESE REPORTS IN A FORIHCCMING GUIDANCE PACKAGE TO LICENSEES.
THE UPGRADE RULE GUIDANCE DEVELOPMENT WORKING GROUP REFERRED TO THESE DOCUMENTS IN THEIR PREPARATION OF THE FIXED SITE PHYSICAL PROTECTION UPGRADE RULE GUIDANCE C0hPENDIUM. __
R CC_MENTS 00 A T
f,
- 197Q, F.
Parg t 5
EIAF APPLA N 10 PEG 11LATORV PROCE13: SEE COMMENT 3/REMACKS.
E CRIBE iMPAC F eiSULTS-SEE Co mENIS/ REMARKS.
Ef*MENTS/*EFApS : THE FIVE REPCRTS (NUREG-0270. 0271, 0272. 0273. 0274) FORM A THOROUGH, THOUGH NOT EXHAUSTIVE, CESCRIPIIGN OF PHYSICAL SECURITY HARDWARE AND SROVIDE SCME BASES FOR ITS EVALUATION. THESE DOCUMENTS PROVIDE THF BASIC FAMILIARIZATION WITH AND SOME BASELINE DATA FOR PHYSICAL SECURITY EQUIPMENT TO THE LICENSING REVIEWER.
REVIEW FOR CCMMENT INITIATED 08/22/78.. - -
PROGRAM OFFICE COMMENTS CN POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 8: 21 DATE ISSUED: 03/24/75 RES DECISION UNIT: RISK ASSESSMENT RIL TITLE:
CRITICAL REVIEW OF SODIUM HYDROXIDE AEROSOL T0XICITY SPONSOPING OFFICE (5): RES EE$: NONE RESEARCH PROJECT MGR:
M. CULLINGFORD RE$_[p""gNT$: THIS WORK CONSISTED PRIMARILY OF A REVIEW OF RELEVANT LITERATURE (WITH SOME PRELIMINARY SUPPORTIVE ANALYSIS) PERTAINING TO THE T0XICITY OF SODIUM HYDROXIDE (NADH).
ONE INSIGHT HAS BEEN Tl!Af SGDIUM IK THE HYDROXIDE FCRM, FOLLCWING AN INCIDENT INVOLVING SODIUM RELEASE, MAY NOT EXIST IN SUFFICIENT AMOUNTS TO WARRANT FURTHER ATTENTIGN.
IN ADDITION, THE CHEMICAL SPECIES THAT WOULD BE PRESENT IN APPRECIABLE QUANTITIES (NA2CO3) MAY NOT BE OF CCNCERN IN TERMS OF HEALTH EFFECTS.
THE PRINCIPAL FINDINGS WHICH SUBSTANTIATE THE ABOVE INSIGHTS ARE:
(A) FCR RELATIVE HUMIDITIES EXCEEDING 35%, IT APPEARS THAT HADH DROPLETS IN THE ATMOSPHERE WILL BE TRANSFORMED TO SODIUM CARBONATE DECAHYDRATE IN AESS THAN A MINUTE IF THE NADH AEROSOL CCNCENTRATION IS LESS THAN OR EQUAL TO ABOUT 1C0 NG/M3.
THIS TRANSFORMATION WILL TAKE LONGER IF THE REL ATIVE HUMIDITY IS LESS THAN 35%.
(B) THE ALKALINITY OF A SODIUM CARBONATE SOLUIION WILL BE SUBSTANTIALLY LESS THAN THAT OF A SODIUM HYDROXIDE SOLUTION OF THE SAME NGRMALITY; THUS, CARBONATE AEROSOLS WILL 3E LESS HAZARDOUS, PER SODIUM ATOM. THAN HYDROXIDE AEROSOLS.
(C) THE TRANSFORMATION FROM THE SODIUM HYDROXIDE TO SODIUM CARBONATE DECAHYDRATE INCREASES THE AERODYHAMIC DIAMETER OF THE AEROSOL BY APPROXIMATELY 40%.
THIS INCREASE IN DIAMETER SHIFTS SCME OF THE AEROSOL QUT OF THE RESPIRABLE RANGE AND THUS LOWERS THE RESPIRABLE FRACTION OF THE AEROSOL.
HYDROXIDE OR CARBONATE PARTICLES ENTERING THE UPPER RESPIRATORY TRACT WILL ABSORB WATER AND GROW SO THE RESPIRABLE FRACTION WILL DECREASE.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEF'ING RELEASE RESULTS PR1T_?It AQTIVITIES PEVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE.. 05/29/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
05/01/78 RP X QENTS ON 05/0t/78, W.
CAMMILL:
klis11BE APPL] CATION TO REGM1ATORY PROGE111 THE REPORT PROVIDES THE LATEST AVAILABLE INFORMATION ON SODIUM HYDROXIDE AEROSOL BEHAVIOR AND ITS T0XICOLOGY.
THIS INFORMATION IS HEEDED IN OUR REVIEW ON CONTROL ROOM HABITABILITY AND OFF-SITE CONSEQUENCES FOLLOWING A POSTULATED ACCIDENTIAL RELEASE OF SODIUM METAL AND SODIUM FIRE.
DESCRIBE IMPACT OF RE_S_ULTS:
THE REPORT FORMS A BASIS FOR CONSIDERING OTHER SPECIES LESS T0XIC AND MORE READILY FORMED IHAM SODIUM HfDROXIDE IN OUR REVIEW OF SODIUM HAZARDS.
COMMENTS / REMARKS: AUTHORS RECOMMENDATIONS FOR FURTHER WORK SHOULD BE CONSIDERED WITH RESPECT TO HRR'S HEEDS.
ElR_ COMMENTS, NOVEMB1R_9, 1979 J.
LONG/T. SPETS:
DESCRIBE APPLICATION TO REG 21ATORY PROCESS.
THE AUTHOR HAS SURVEYED THE T0XICITY OF FUMES THAT EMANUATE FROM A LARGE FIRE OF NON-RADI0 ACTIVE SODIUM. HE HAS SHOWN THAT THE INITIAL PRODUCTS SODIUM OXIDES ARE ALMOST IMMEDIATELY CONVERTED TO SODIUM HYDROXIDE IN THE AIMOSPHERE, AND WITHIN A FEW MINUTES TO THE MUCH LESS T0XIC SODIUM CARBONATE _ THERE IS AN APPLICATION OF THIS WORK IN THE ANALYSIS OF SECONDARY SODIUM FIRES IN LMFBRS.
DESGQXBE IPPACV 0F RE1H1TS: THE RE3ULTS SEEM TO CONFIRM THAT THE COMMONLY USED US CEILLING OF 2 MG/M3 FOR NQGH IS PROBABLY ACCtPTABLE FOR ACCIDENT CALCULATIONS. THERE IS INSUFFICIENT EVIDENCE HOMEVER TO PERMIT THE fHCREASE OF VHH3 LIMIT BASED ON THE TRANSITION TO HA2CO3, ALTHOUGH THIS TRANSITION SEEMS PROBABLE WITHIN A FEW MINUTES.
(E MENTS/ REM 1353: EXPERIMENTAL WORK ON THE KINETICS OF THE TRANSITION OF HA0H TO NA2CO3 WOULD BE REQUIRED IF IT IS DESIRED 10 TAKE ADVANTAGE OF THIS TRANSITION IN ASSESSING THE HAZARDS FROM THE FUMES OF A SODIUM FIRE.
ADDITIONAL STUDIES OF THE T0XICITY OF NADH AT EARLY STAGES AND DURING THE TRANSITION MIGHT ALSO BE REQUIRED. roR EXAMPLE, EFFECTS OF PARTICLE SIZE AS WELL 45 FUME DENSITY MAY BE IMPORTANT.
m '
f i
[ PROGPAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORr PROCESS l
PIL 8:
23 DATE ISSUED: 04/10/73 RES DECISION UNIT: SAFEGUARDS PIL TITLE:
"EASI" ADVERSARY SEQUENCE EVALUATIDH MODEL (COMPUTER GRAPHICS VERSIONJ SPONSODING OFFICE (S): NMSS RRS:
4-1 EFFECTIVENESS RESEARCM PROJECT MGR:
R. ROBINSON EVALUATION RES CCMMENTS: RESEARCH HAS BEEN COMPLETED ON DEVELOPING A GRAPHICS DISPLAY VERSION OF A COMPUTER MODEL CALLED ESTIMATE OF ADVERSARY SEQUENCE INTERRUPTION (EASI), AND RESPONDS TO AN EXPRESSED NEED FOR EVALUATIVE METHODS FOR FIXED-SITE THEFT AND SABOTAGE PREVENTION SYSTEMS. DOCUMENTATION HAS ALREADY BEEN MADE AVAILABLE THROUGHOUT NRC CCNCERNING PROGRAFMABLE POCKET CALCULATCR VERSIONS OF THE EASI MODEL.
THE OBJECTIVE OF THE "EASI" METHOD IS TO PROVIDE A USABLE EVALUATION METHOD WHICH CAN SERVE AS EITHER A PHYSICAL PROTECTION SYSTEM DESIGN AID OR AS A DECISION AID IN THE LICENSING AND INSPECTION PROCESS. THE EASI GRAPHICS PROGRAM i
ALLOWS THE USER TO INPGT FACILITY AND ADVERSARY PATH ATTRIBUTES AT A COMPUTER GRAPHICS TERMINAL, AND OBTAIN AS OUTPUT A CRT "PERSPECTIVE VIEW" LINE PLOT.
THE METHOD CAN TREAT BOTH THEFT AND SABOTAGE OBJECTIVES BY THREATS OF INSIDERS, CUTSIDERS. AND COMBINATIONS OF EACH GROUP.
THE RESULTS OF THE EASI ANALYSIS ARE EXPRESSED IN TERMS OF THE PROBABILITY THAT THE PHYSICAL PROTECTION SYSTEM CAN RESPOND IN TIME TO INTERRUPT AN ADVERSAPY ALONG A PHYSICAL PATH (ACTION SEQUENCE). TO SUPPLEMENT THE EASI CALCULATICNS, EASI GRAPHICS PROVIDES THE ANALYST WITH A SELECTION OF SIX TWO-DIMENSIONAL AND EIGHT THREE-DIMENSIONAL PLOTS.
THESE PLOTS ALLOW THE USER TO EXAMINE THE SENSITIVITIES OF VARIOUS COMPONENTS ALONG THE ADVERSARY'S PATH Ar.D TO SIUDY THE EFFECT ON THE PROBABILITY OF INTERRUPTION OF VARYING THE PERFORMANCE OF THESE COMPONENTS.
If IS RECCMMENDED THAT THE EASI METHOD BE USEC BY NMSS AND OTHER OFFICES AS AN ANCILLARY AID IN DEVELOPING PERFORMANCE-ORIENTED REGULATIONS OR IN CARRYING OUT A COMPREHENSIVE EVALUATION PROGRAM.
l i
USER DIdCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EQ3T RIL AC11VIT111 REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPGhSa8LE......... HMSS l
SCHEDULED CCMPLETION DATE.. 06/10/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED I
ACTUAL COMPLETION DATE.....
NMSS COMMENTS.
B. HATTER: TEST APPLICATIONS ARE BEING CONDUCTED TO DETERMINE USER SUITABILITY.
MS$_C_QTIFJT5 uc. LEWLS: ThE RESULTS OF THESE EFFORTS HAVE BEEi; USED IN THE CURRENT DEVELOPMENT OF IMPROVED EVALUATIVE METHODS FOR FIXED SITE SAFEGUARDS.
NPR CCSF,NTS O_N_AlLQUST t,
- 1979, F.
P A G_A NQ :
DEM RIRf_A FELC A T I O N 10 REGULA10RY PROCESS: SEE COMMENTS / REMARKS.
pD LPR U MPACT OF RESULTP SEE COMMENIS/ REMARKS.
5050LN i $/ R Et1 A R K S :
INE KEY TO THE UTILITY OF EVALUATION MODELS IS THE RATIO OF THE EFFORT REQUIRED TO RUN THE MODEL TO IHE USEFULNESS OF THE RESULTS OBTAINED. ALTHOUGH SAFEGUARDS EVALUATION MODELS DEVELOPED TO DATE TEND TOWARDS INORDINATE INPUT COMPLEXITIES BY COMPARISON WITH OUTPUT UTILITY, EASI HAS BEEN A WELCOME EXCEPTION TO ThIS TREND.
CONSEQUENTLY, EASI HAS BEEN USEFUL AS AN AID TO LICENSING DECISION MAKING, AND HAS BEEN DISTRIBUTED BY NRR TO THE POWER REACTOR LICENSEES AS AN AID IN PHYSICAL SECURITY SYSTEM DESIGN. THE ADDITIONAL COMPUTER PROGRAMMING WHICH PERMITS THE USE OF COMPUTER GRAPHICS IN CONJUNCTICH WITH THE EASI CODE MAY BE USEFUL FOR PARAMETRIC DESIGN STUDIES, ALTHOUGH IT PROVIDES LITTLE ADDITIONAL UTILITY TO THE LICENSING REVIEWER. ADDITIONS AND FURTHER VARIATIONS OF THIS SIMPLE TOOL APPEAR TO PROVIDE DIMINISHING RETURNS. ;
1 P900aAM OFFICE COM*EN15 09 POTENTIAL UTILIZA110N OR WALUE OF RESEDRCH RFJULTS IN THE REGULATORY PROCESS RIL 8-24 DATE ISSUED: 04/10/78 RES DECISION UNIT: SAFEGUARDS RIL TITLE:
"FESEM" ADVERSARY SEQUENCE EVALUATION MODEL 1EEiSCPING OFFICE (S): NMSS ERS:
4-1 EFFECTIVENESS RESEARCH PROJECT MGR R. ROBINSON EVALUATION ff5_CQENENTS* RESEARCH HAS BEEN COMPLETED ON THE FORCIBLE ENTRY S4FEGUARDS EFFECTIVENESS MODEL (FESEM), IN RESPONSE 10 A NEED FOR EVALUATIVE METHODS FOR FIXED-SITE THEFT AND SABOTAGE PREVENTION SYSTEMS.
THE PURPOSE OF THIS STUDY WAS TO DEVELOP A METHODOLOGY FOR ANALYZING FIXED-SITE SECURITY SYSTEMS AS TO THEIR EFFECTIVENESS AGAINST A FORCIBLE ATTACK BY AN ADVERSARY INTENT ON CREATING AN ACT OF SABOTAGE OR THEFT.
THE MODEL PROVIDES A FRAMEWORK FOR PERFORMING INEXPENSIVE EXPERIMENTS RELATED TO FIXED-SITE SECURITY SYSTEMS. FOR TESTING ALTERNATIVE DECISIONS, AND FOR DETERMINING THE RELATIVE COST EFFECTIVENESS ASSOCIATED WITH THESE DECISIDH POLICIES.
THE RESULTS CF THE FESEM ANALYSIS INC.UDE ESTIMATES OF THE PROBABILITY OF SABOTAGE OR THEFT WINS (AND LOSSES) BASED CN ATTACK FCRCE SIZE. ATTACK MOBILITY. AND TYPE OF ATTACK; COLLECTED STATISTICS ASSOCIATED WITH EACH VARIABLE (E.G.,
NUM3ER OF WIHS BY DEFENDERS AND BY ATTACKERS FOR SUCCESSFUL SABDTAGE OR THEFT, ALARM TYPES FOR ALL RUNS. TIME REGUIRED FOR SUCCESSFUL SABOTAGE OR COMPLETION OF THEFT, ETC.).
THE PROGRAM IS CURRENTLY AVAILABLE FOR NRC USE VIA AN ACCESS CODE NUMBER TO SANDIA'S CCrPUTER. A TRAINING PROGRAM WAS GIVEN IN FEBRUARY 1978 TO INTERESTED NRC PERSONNEL AND POTENTIAL USERS, IT IS RECOMMENDED THAT THE FESEM MODEL BE USED BY NMSS AND OTHER OFFICES AS AN ANCILLARY AID IN FORMULATING REGULATORY REQUIREMENTS, LICENSING. INSPECTION AND OTHER MONITORING OPERATIONS.
USER DISCUSSION POSITION COMMISSICH ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS E257.RIL A C T Iv f T T_E5 RELLEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSILiE......... NMSS/NRR NRR SCHEDULED COMPLETION DATE.. 06/10/73 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL CCMPLETION DATE.....
NMSS CCMMENTS, B. MATTER - TEST APPLICATIONS ARE BEING CONDUCTED TO DETERMINE USER SUITABILITY.
NMSS CC*"*ENW__GJfWi$ - THE RESULTS OF THESE EFFORTS HAVE BEEN USED IN THE CURRENT DEVELOPMENT OF IMPROVE 9 EVALUATavE METHODS FOR FIXED SITE SAFEGUARDS.
NN GM"ED_CN AyDUST t,
- 1979, F.
P A G A N._Q :
_Dil.CX D4 *PPLICATI0N TO 8E00LATQLY_PJQSEM: SEE COMMENTS / REMARKS.
RE1CM1Bl_JyeaCT OF RESULTS: SEE COMMENIS/ REMARKS.
EREMENT5/EtMARKS: THE KEY TO THE UTILITY OF EVALUATION MODELS IS THE RATIO OF THE EFFORT REQUIRED TO RUN THE MODEL TO THE USEFULNESS OF THE RESULTS. ALTHOUGH THE FCSEM OUTPUT PPOVIDES INFORMATION WHICH WOULD BE USEFUL IN THE DESIGN AND EVALUATION OF PNYSICAL SECURITY SYSTEMS FOR POWER REACTORS. THE UTILITY OF SUCH INFORMATION (MUCH OF WHICH IS INTUITIVELY OBVIOUS TO A PHYSICAL SECURITY EXPERT) DOES NOT APPEAR TO WARRANT THE EXTENSIVE EFFORT REQUIRED FOR INPUT PREPARATION AND EXECUTION OF THIS MODEL.
IN ADDITION. THE ACCURACY OF THE COMPUTED RESULTS CANNOT BE VERIFIED AS L RESULT OF THE LIMITED APPLICABILITY OF MILITARY-TYPE ENGAGEMENT MODELS TO THE PROBLEM OF SABOTAGE OF A REACTOR FACILITY.
1p_GMNTS ON 09/13/73, G.
RIVENBAQ:
IHE RESULIS OF THE EFFORIS WILL BE OSED IN DEVELOPMENT OF STANDARDS FOR SAFEGUARDS SYSTEMS AS WELL AS BY LICENSEES IN DEVELOPING THEIR SYSTEMS AND BY NRC IN EVALUATING THOSE SYSTEMS. --
~.-
___PdGGu1M OFFICE CC-*ENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 8:
25 DATE ~_SSUED:
03/21/73 RES DECISION UNIT-FUEL BEHAVIOR RIL TITLE:
FRAP-53 SPONSCPING CFFICE(S): NRR E39:
1-12 FUEL CODE EJSEARCH PROJECT MGR:
G. MARINO DEVELOPMENT PE}_[prrE3TSi FRAP-53 IS A BEST-ESTIMATE COMPUTER CODE THAT CALCULATES THE THERMAL AND MECHANICAL RESPONSE CHARACIERISIICS OF A NUCLEAR FUEL ROD OPERATING UNDER STEADY-STATE POWER CONDITIONS. AND WAS DEVELOPED TO PROVIDE ACCURATE INITIAL VALUES OF FUEL-ROD F ARAMETERS FOR INPUT INTO TRANSIENT ANALYSIS CODES SUCH AS FRAP-T AND RELAP.
IT IS CAPASLE OF SUPPLYING THE HOT-STATE VALUES OF SUCH QUANTITIES AS:
1.
STCRED ENERGY 2.
RADIAL TEM?ERATURE DISTRIBUTIONS AT GIVEN AXIAL LOCATICNS 3.
TOTAL FISSION GAS RELEASE 4.
RCD INTERNAL GAS PRESSURE AND COMPOSITION 5.
CLAD DEFCRMATICN 6.
AMOUNT OF PELLET-CLAD INTERACTION (PCI) 7.
FUEL DEFORMATION (SWELLINC, DENSIFICATION. RELOCATION, AND THERMAL EXPANSION) 8.
FUEL-CLAD GAP SIZE AND GAP CONDUCTANCE 1.
CLAD-CORROSICN AND HYDRIDING.
ALL OF THESE CUANTITIES ARE STRONGLY DEPENDENT UPON THE OPERATING HISTORY OF THE ROD, AND EACH WILL HAVE A LARGE EFFECT CN THE PREDICTED AND MEASURED RESPONSE OF A FUEL ROD DURING A TRANSIENT. THE CODE. THEREFORE. HAS BEEN DESIGNED TO PROVIDE THESE AND OTHER QUANTITIES FOR ANY GIVEN POWER HISTORY AS INITIAL CONDITIONS TO THE TRANSIENT CODES.
THE VERIFICATION OF THE FRAP-53 CODE HAD TWO MAJOR OBJECTIVES:
(1) TO DETERMINE THE CODE PERFORMANCE IN PREDICTING THE AVAILABLE. QUALIFIED. EXPERIMENTAL DATA, AND (2) TO IDENTIFY THOSE AREAS THAT REQUIRE MORE SOPHISTICATED MODELING OR MORE EXPERIMENTAL DATA.
THE CODE PERFORMANCE AND DATA WERE ANALYZED USING STATISTICAL METHODS.
THUS.
ALL OF THE MAJOR RESPONSE VARIABLES ARE PRESENTED ALONG WITH THEIR CORRESPONDING STANDARD ERROR BOUNDS.
THE VERIFICATICN PROCEDURE USED INFORMATION FROM OVER 700 FUEL RODS CONTAINING A WIDE RANGE OF OPERATING AND DESIGN P A?. AM E T E R S.
USER DISCUSSION POSITION COFfMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EpST RIL_AfTIVITIES REV10W HELD COMPLETED HEL D HEL D ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR/SD SCPEDULED COMPLETION DATE.. 05/21/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETICN DATE.....
NEE _C_GrrLNTS ON 05/22/75, _D, 90551 E11CRIBE APPLICATION TO REGULATORY PROCESS: STEADY-STATE FUEL PERFORMANCE CODES ARE REVIEWED BY HRR AS PART OF LOCA AND OIHER ACCIDENT ANALYSIS. NRR USES AN INDEPENDENT NRC-DEVELOPED CODE FOR AUDIT PURPOSES IN THESE REVIEWS.
pnc3nf_LMLA1T_p F_RE$f(15 : NO DIRECT IMPACT.
NRR USES THE GAPCON SERIES OF CODES IN AUDIT WORK AS IT HAS DONE SINCE BEFORE FRAP-S WAS DEVELOPED. NRR HAS NO PLANS TO USE FRAP-S IN THIS CAPACITY.
(DMMENTS/ REMAP 51: THE FRAP-S CODE WAS DEVELOPED BY RES TO INITIALIZE VARIOUS ANALYSIS CODES. RES AND NRR HAVE RECOGNIZED THE DUPLICATE CAPCON AND FRAP-S CODE EFFORTS AND HAVE CONSOLIDATED THESE EFFORTS INTO A HYBRID CODE FRAPCON.
FRAPCON WILL CONTAIN SOME ELEMENTS FROM FRAP-53, BUT FRAP-S DEVELOPMENT WILL BE DISCONTINUED.
SD CQ*"EMV9 R. 08vEPpANK - THE ENGINEERING METHODOLOGY STANDARD 3 BRANCH PLANS TO USE THE RESEAOCH RESULTS IN 2IS CNGOING DEGRADED CORE COOLING TAS% AND IN ESTIMATING THE FISSION PRODUCT SOURCE TERM F00 ENVXRONMENTAL FNVELCPE STANDARDS ACTIVITY.
RESULTS P.AY ALSD BE USED IN THE ECCS RULE CHANGE EFFORT AND. IN THE LONG TERM, IN REGULATORY GUIDES RELATED TG CORE DESIGN. I,
D PROGRsM QEEACE CAPMENTS_DN POTENTAAA UTIL22ATION OR VALUE OF RE3EARCH RESULT 3 IN THE REGULATORY PROCESS I
i
{
RIL s: 26 DATE ISSUED: 04/27/78 RES DECISION UNIT: FbEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS
]
RIL TETLE:
THE IMPACT OF OFFSHORE NUCLEAR GENERATING STATIONS ON RECREATIONAL BEHAVIOR AT ADJACENT COASTAL SITES j
SPONSCRING UFFICE(5): NRR REG: 5-21 SOCIO-RESEARCH PROJECT MGR D. BARNA 3
ECONOMIC IMPACTS d
RES CCMMENTS: THESE RESULTS ARE OFFERED TO SUPPORT NRC COST-BENEFIT ANALYSTS WITH NEW AND IMPROVED INFORMATION FOR l
ASSESSING LIKELY IMPACTS OF NUCLEAR GENERATING STATIONS ON RECREATIONAL BEHAVIOR AT ADJACENT COASTAL SITES.
}
THE RESEARCH RESULTS INDICATE THAT (A) PROXIMITY OF A FLOATING NUCLEAR PLANT IS LESS IMPORTANT THAN OTHER BEACH j
ATTRIBUTES IN DETERMINING BEACH ATTRACTIVENESS; (B) PROBABLY NO MORE THAN CAND PERHAPS LESS THAN) 5% TO 10% OF CURRENT i
BEACH PATRONS WOULD AVOID A BEACH AFTER FNr SITING 3 MILES DIRECTLY OFFSHORE; AND (C) IMPACT OF AN FMP WOULD DECREASE i
EXPONENTIALLY AS DISTANCE AWAY INCREASED.
IN
SUMMARY
, THE PERCENTAGE REDUCTION IN TOURISM ATTRIBUTABLE TO SITING OF NUCLEAR POWER PLANTS OFFSHORE WOULD BE SMALL, EUT NOT NECESSARILY NEGLIGIBLE, AT POINTS CLOSE BY.
THE STABILITY OF THOSE IMPACTS OVER TIME, HOWEVER, DEPENDS UPON t
{
THE STABILITY OF CURRENT ATTITUDES TOWARD AND BELIEFS ABOUT NUCLEAR POWER AND ITS SAFETY.
k USER DISCUSSIGN POSITION COMMISSION ACRS PRESS
{
OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS j
EOST RIL ACTIVITIES FEVIEW HEt D COMPLETED HELD HELD ISSUED IMPLEMENTED i
0FFICE RESPONSIBLE......... NRR J
SCHEDULED COMPLETION DATE.. 05/29/78 UNSCHED UNSCHED UNSCHED UNSCHi3 UNSCHED UNSCHED l
ACTUAL COMPLETION DATE.....
04/27/73 4
(
NRR CCMMENTS ON 04/27/78.
M.
ERNST:
j DEScelBE APLICATION TO REGULATORY PROCESS: THIS STUDY IS BEING INCORPORATED INTO STAFF TESTIMONY ON FNP t-8.
THE IESTIMONY IS CONCERNED WIIH POIENTIAL AVOIDANCE OF BEACH RESORTS, BY TOURIST, DUE TO PERCEIVED RISK FROM FLDATING 4
l NUCLEAR PLANTS SITED IN THE VICINITY OFFSHORE. EXTRAPOLATION OF TOURIST BEHAVIOR IN THE VICINITY OF LAND BASED NUCLEAR i
PCMER PLANTS TO OFFSHORE PLANTS IS TENUQUS.
THE STUDY ALSO PROVIDED A BROADER GENERIC UNDERSTANDING OF RECREATIONAL 1
LEHAVIOR WHICH WILL STRENGTHEN OUR CAPABILITY TO HANDLE THIS ISSUE IN EIS'S GENERALLY.
I D11GRIBE IMPACT OF RESULTS: THIS STUDY HAS RESULTED IN STRONGER, MORE OBJECTIVE STAFF TESTIMONY CONCERNING TCURIST l
3 AVOIDAhCE OF BEACHES IN IHE VICINITY OF FNP'S.
THE ESTIMATES OF POTENTIAL AVOIDANCE HAS ALLOWED CALCULATION OF i
LIKELY ECONOMIC IMPACT TO A RANGE OF COASTAL ECONOMIES. SATISFACTORY DISPOSITION OF THIS CONTENTION GENERICALLY IN THE q
FNP HEARINGS SHOULD REDUCE OR ELIMINATE THIS CONCERN IN FUTURE LICENSING ACTIONS FOR SPECIFIC FNP APPLICATIONS.
)
[0MMENTS/ REMARKS: THIS STUDY MAS AN APPLICATION TO NUCLEAR SITUATIONS OF SOCIAL RESEARCH TOOLS DEVELOPED TO ANALYZE HUMAN BLHAVIOR RELATIVE TO RISK FROM NATURAL HAZARDS.
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P&OGRAN OFFICE CCPNENTS ON POIENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS PIL s: 27 DATE ISSUED: 06/02/73 RES DECISION UNIT: CODE DEVELOPMENT WIL TITLE: "BEACON / MOD 2" SPCNSORING OFFICE (S): RES RES:
1-15 CONTAINMENT CODE RESEARCH PROJECT MGR:
S. FABIC REl_EQEMLMT3: THIS RIL TRANSMITS THE BEACON / MOD 2 COMPUTER CODE MANUAL, DESCRIBES ITS FIELD OF AIPLIC# TION AND DISCUSSES THE CODE'S STRENGTHS AND LIMITATIONS. BEACON / MOD 2 IS AN ADVANCED, BEST ESTIMATE CODE INTENDED FOR EVALJATICN OF SHORT-TERM THERM 0HYURAULIC CONDITIONS WITHIN "DRY" (FULL PRESSURE) MULTICOMPARTMENT CCHTAINMENTS, OR WITHIN CERTAIN REGIONS OF THE
- PRESSURE SUPPRESSION" DRYWELL.
THIS RESEARCH WAS INITIATED TC PROVIDE IMPORTANT MODELING IMPROVEMENTS FOR BEST ESTIMATE ANALYSIS OF THESE CONTAINMENT SYSTEMS.
BEACCN/ MOD 2 0FFERS CONSIDERABLE ADVANTAGES OVER THE EXISTING CONTAINMENT CODES FOR BEST ESTIMATE EVALUATION OF HYDRAULIC LOADS IN MULTICGMPARTMENT PWR TYPE CONTAINNENTS.
IT IS PARTICULARLY SUITABLE FOR EVALUATION OF THE REACTOR CAVITY LOADS (FOR POSTUL ATED BREAKS BETHEEN THE REACTOR VESSEL AND THE BIOLOGICAL SHIELD) IN BOTH PWR AND BWR CONTAINMENTS.
THE CODE HAS ALSO SHOWN A CAPABILITY TO DESCRIBE THE EVOLUTION OF A TWO-PHASE (FLASHING) JET AND THE RESULTIHF PRESSURE LOADS ON THE IMPACTED BARRIER.
IT IS RECOGNIZED THAT THE CODE MU5i BE MORE EXTENSIVELY TESTED AGAINST EXPERIMENTAL DATA.
NEVERTHELESS. THE BEACON / MOD 2 IS RECOMMENDED FOR CALCULATIONS OF THE REACTOR CAVITY LOADS, FOR BOTH PWR AND BWR INSTALLATIONS, AND FOR EVALUATION OF JET IMPACT LGADS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PQ1T PIL aqTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE EESPONSIBLE......... hRR SCHEDULED CCMPLETION DATE.. 06/10/73 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCPED ACTUAL COMPLETION DATE.....
N R _(QMENTS, R.
TERE 1ED:
PE1ERIBF AP7(lqAilRN TO REGVtATORY FROCE15: THE BEACON CODE, AS PRESENTLY DEVELOPED, REPRESENTS A POTENTIAL TEST ESTIMATE COMPUILR PROGRAM FOR THE CALCULATION BY HRR OF SUBCOMPfRTMENT PRESSURE / TEMPERATURE RESPnNSES. OF PARTICULAR IMPORTANCE TO CUR LICENSING PROCESS IS THE ABILITY OF THIS CODE 10 EVALUATE THE REACTOR CAVITY PRESSURE TRANSIENT IN TWO-DIMENSIONS AS OPPOSED TO THE ONE-DIMENSIONAL COMPUTER PROGRAMS CURRENTLY IN USE IN THE LICENSING PROCESS. UPCN COMPLETION OF TEE PLANNED CODE VERIFICATION EFFORTS,* BEACON WILL BE USED TO BENCHMRK ALL LICENSING COMPUTER PROGRAMS TO ESTABLISH A MEASURE OF THE MARGINS, UHICH CURRENTLY EXIST IN THESE LICENSING CODES.
D15C2HE_lf*P_A CD F_RE1MLD : THE BEACON COMPUTER PROGRAM REPRESENTS A NEXT GENERATION COMPUTER PROGRAM FOR THE CALCULATION OF SUBCOMPARTMENT PRESSURE / TEMPERATURE TRANSIENT RESPONSES. FOR THE PRESENT, HOWEVER, AND PENDING SUCCE55FUL COMPLETICN OF THE VERIFICATION EFFORTS, NRR DOES NOT PLAN TO USE THIS CODE IN THE LICENSING PROCESS.
WHEN THE CCDE HAS BEEN EVALUATED AGAINST EXPERIMENTAL DATA AND COMPLETELY CHECRED OUT, IT WILL BE MADE A PART OF THE LICENSING PROCESS.
THE CODE WILL THEN BE USED AS A BEST ESTIMATE TOOL TO DETERMINE THE SAFETY MARGINS WHICH EXIST ik OUR CURRENT LICENSING REQUIREMENTS WILL DEPEND ON THE RESULTS OF THESE COMPARISONS.
(QtO LN T 3/ Pl5t_RKS : THE BEACON CODE HAS AN EXCELLENT POTENTIAL FOR BECOMING AN IMPORTANT LICENSING TOOL.
AS A RESULT. II 15 RECCMNENDED THAT THE CODE VERIFICATION EFFORT BE PERFORMED IN A TIMELY MANNER.
THIS VERIFICATI0H EFFORT SHOULD TAKE PRIORITY OVER ANY ADDITIONAL MODIFICATIONS TO THE CODE WHICH ARE AIMED AT EXPANDING ITS BASIC ANALYTICAL CAPABILITY..
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PDCCOAM OFFICE CGM 7ENTS CN POTENTIAL UTILIZATION Ok VALUE OF RESEARCH RESULTS IN THE REGUL ATORY F ROCESS j
l PIL s: 28 DATE ISSUED: 05/09/78 RES DECISION UNIT: FUEL BEHAVIOR
]
l RIL TITLE:
"MELT / CONCRETE INTERACTIONS
- j SPON$0 RING OFFICE (S): RES RRS:
1-13 FUEL MELT RESEARCH PROJECT MGR:
R.
DISALVO 1
}
REl_(Q*MENTS: THIS RIL DFSCRIBES THE INTER-1 CODE FOR CALCULATING THE EFFECTS OF INTERACTION BETWEEN MOLTEN MATERIALS AND CONCRETE AND THE EXPERIMENTAL DATA BASE FROM WHICH IT WAS DEVELOPED. THIS WORK HAS RESULTED IN AN IMPROVED l
MODEL BASED ON EXPERIMENTS WITH PROTOTYPICAL MATERIALS.
IN SEPARATE EFFECTS EXPERIMENTS. MON 3 LITHIC SPECIMENS OF CONCRETE WERE SUBJECTED TO CONTROLLED THERMAL FLUXES IN ORDER TO j
MEASURE RATES OF EROSION. EROSION IS LINEAR WITH TIME FOR A GIVEN HEAT FLUX, AFTER CORRECTION FOR THERMAL LOSSES 1
THROUGHOUT REFLECTICN AND RADIATION.
THE COMI;4 ANT MODE OF EROSION IS QUIESCENT MELTING OF THE CEMENT (I.E.,
THE BINDING j
MATERIAL), WITH NO DIFFEENCES OBSERVED BY VARYING THE COMPOSITION OF THE AGGREGATE MATERIAL. THESE DATA ARE i
NECESSARY TO INTERPRET EROSION RATES 08SERVCD IN INTEGRAL EXPERIMENTS.
AS A RESULT OF THE INTEGRAL EXPERIMENTS, IN WHCH PROTOTYPICAL MOLTEN MATERIALS CONTACT CONCRETE, THE FOLLOWING CONCLUSIONS WERE DRAWN:
5
- EROSION OF CONCRETE IS THERMALLY DOMINATED. HITH INSIGNIFICANT CONTRIBUTIONS FROM MECHANICAL AND CHEMICAL EFFECTS.
THE PRINCIPAL MECHANISM OF EROSION IS MELTING OF THE BINDING MATERIAL, WITH NO SIGNIFICANT QUALITATIVE DIFFERENCES I
CAUSED BY CHANGING THE COMPOSITION OF THE AGGREGATE.
- THE CCMPOSITION OF THE CCNCRE(E DETERMINES THE COMPOSITION AND MASSES OF GASES RELEASED AT THE INTERFACE OF THE MELT AND CONCRETE.
- TURBULENCE AND ESSENTIALLY ISOTHERMAL CONDITIONS ARE INDUCED IN THE MELT BY THE PASSAGE OF DECOMPOSITION GASES.
HYEROGEN AND CARBON MONOXIDE ARE AMONG THE GASES EVOLVED FROM THE SURFACE OF THE MELT AND THEY BURN UPON CONTACTING l
1 AIR.
THIS INDICATES THAT THE H2O AND CO2 RELEASED FROM THE DECCMPOSING CONCRETE ARE REDUCED CHEMICALLY, MOST LIKELY BY OXIDIZING THE METALLIC CONSTITUENTS OF THE MELT.
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THE EXPERIMENTS HAVE CULMINATED IN AN ANALYTICAL MODEL (INTER-1) 0F THE MELT / CONCRETE INTERACTION WHICH CAN HELP EXTEND THEIR RANGE OF APPLICABILITY.
WHILE DIRECT EXTRAPOLATION OF THE DATA TO PROTOTYPICAL CONDITIONS MUST ALWAYS BE MADE CAUTIOUSLY, ENDUGH CONFIDENCE
- 1AS BEEN DEVELOPED SO THAT NO FUNDAMENTAL DIFFERENCES IN BEHAVIOR ARE ANTICIPATED IN SCALING TO FULL-SIZE SYSTEMS RESERVATIONS EXIST REGARDING THE APPLICATION OF INTER-1 TO PREDICT SUCH VARIABLES AS THE TIME OF CONTAINMENT MELTTHROUGH OR OVERPRESSURIZATION. THE MODEL CAN BEST BE UTILIZED IN ITS CURRENT FORM TO ESTIMATE THE RELATIVE SIGNIFICANCE OF I
VARIATIONS IN PARAMETERS SUCH AS MATERIALS, PROPERTIES AND COMPOSITIONS, INTERFACE HEAT TRANSFER COEFFICIENTS, GEOMETRY, ETC.
THE PRIMARY SIGNIFICAMCE OF THE WORK DESCRIBED IS THE IMPROVED UNDERSTANDING OF PHYSICAL PHENOMENA. RES RECOGNIZES 1
THAT THE RESULTS ARE UNLIKELY TO HAVE SIGNIFICANT NEAR-TERM IMPACT ON CURRENT LICENSING PROCEDURES. IT SHOULD, HOWEVER,
' PROVIDE ACPITIONAL BACKGROUND INFORMATION USEFUL IN ANALYZING REGULATORY ISSUES INVOLVING ACCIDENTS BEYOND DESIGN BASIS EVENTS.
4 i
USER DISCUSSION POSITION COMMISSION ACR$
PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PQ3J PIL ACTIVITIES REYIfy HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR/SD J
)
- SCHEDULED CCMPLETION DATE.. 03/30/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL CGMPLETION DATE.....
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N74 COMMENTS. W. GrMMILL CESCRIBE APPLICATION TO REGULATORY PROCESS: CALCULATING THE EFFECTS OF INTERACTIONS BETWEEN MOLTEN MATERIALS AhD C0hCREIE IS DIRECILY APPLICABLE 10 MAKING LICENSING DECISIONS RELATED TO CONTAINMENT REQUIREMENTS FOR ADVANCED REACTORS (SUCH AS LMFBR'S) TO ENSURE COMPARABILITY WITH PRESENT DAY LWR NUCLEAR POWER PLANTS.
THIS RESEARCA PROGRAM HAS IMPROVED OUR ABILITY TO PREDIC) THE TIME AND MODE OF CONTAINMENT FAILURE BECAUSE OF IMPROVED UNDERSTANDING OF THE INTERACTIONS BETWEEN MOLTEN CORE MATERIALS AND CONCRETE, SUCH AS THE RATE AND EXTENT OF CONCRETE EROSION, QUANTITY AND TYPE OF GASES RELEASED, FISSION PRODUCTS RELEASE, AND THE EFFECTS OF CCNCRETE CHEMICAL COMPOSITICN. AS DEMONSTRATED IN NUREG-0440, "LIQUID PATHWAY GENERIC STUDY,"
DATED FEBRUARY 1978, THE RESULTS OF THIS RESEARCH PROGRAM ARE ALSO DIRECILY APPLICABLE TO EVALUATING QUESTIONS OF COMPARABILITY OF CONSEQUENCES OF POSTULATED CORE MELTDOWN EVENTS AT LAND-BASED AND FLOATING NUCLEAR PLANTS.
DEif?IBE IMPACT OF RESULTS: THE IMPACT OF RESULTS OF THIS RESEARC H PROGRAM ON CURRENT LICENSING PROCEDURES REMAINS TO BE EVALUATED.
THESE RESEARCH RESULTS HAVE PLAYED AN IMPORTANT ROLE IN NRR'S SAFETY EVALUATIDH OF THE CLINCH RIVER BREEDER REACTOR (CRBR) AND THE FAST FLUX YEST FACILITY CFFTF), PARTICULARLY IN REGARD TO EVALUATING THE CONSEQUENCES OF POSTULAT2D CORE MELTDOWN EVENTS AND ESTABLISHING APPROPRIATE CONTAINMENT REQUIRE-MENTS TO ENSURE COMPARABILITY WITH PRESENT DAY LWR PLANTS.
THESE RESEARCH RESULTS HAVE ALSO HAD AN IMPORTANT IMPACT ON THE STAFF'S CONCLUSIONS CONTAINED IN THE "LIQUID PATHWAY GENERIC STUDY," NUREG-0127. DATED FEBRUARY 1978.
THE STUDY FOUND THAT THE RISKS ASSOCIATED WITI.
RELEASES TO THE HYDROSPHERE AT A FLOA1ING NUCLEAR PLANT (FNP) ARE GREATER THAN THOSE AT A LAND BASED PLANT (LBP) FCR CORE MELT ACCIDENTS. THE STAFF THEN ASKED THE APPLTCANT TO MAKE DESIGN CHANGES IN THE PLANT TO MITIGATE THE CONSEQUENCES OF THIT, KIND OF ACCIDENT; SPECIPIC/ LLY, THE STAFF IN THE DRAFT ENVIRONMENTAL STATEMENT, PART III (NUREG-0127), MAY 1978, AND IN A SUBSEQCfMT LETTER TO OFFSHORE POWER SYSTEMS (OPS)
(R.P. BALLARD TO A.P. ZECHELLA. JULY 25, 1978) REQUESTED THAT THE CONCRETE PAD BENEATH THE REACTOR VESSEL BE REPLACED BY SCME MATERIAL THAT PR9VIDES INCREASED RESISTANCE TO A MELT THROUGH BY THE REACTOR CORE.
FINALLY, THE RESULTS OF THIS RESEARCH PROGRAM ARE IMPROVING OUR ABILIT" TO MAKE QUANTITATIVE RISK ASSESSMENTS ON EXISTING AND PROPOSED NEW TYPES OF REACTORS WHICH WILL ASSIST NRR 'a MAKING LICENSING DECISIONS IN REGARD TO DESIGN AND SITING.
COMMENTS /REMAR11: THE ADVANCED 2EACTORS BRANCH (NRR) TRANSMITTED COMMENTS ON RIL 828 TO THE FUEL BEHAVIOR RESEARCH BRANCH (RES) IN A MEMCRANDUM FROM T. P. SPEIS TO W. V. JOHNSTON, DATED OCTOBER 13, 1978.
RESOLUTION OF THESE CGMMENTS IS NOW IN PROCESS.
IN ADDITION, HRR PLANS TO WRITE A RESEARCH REQUEST TO RES TO EXPAND IHIS PROGRAM TO EXAMINE MELT INTERACTIONS WITH SACRIFICIAL MATERIALS IN CONNECTION WITH THE LICENSYNG REVIEW OF CPS'S APPLICATION TO MANUFACTURE 8 FLOATING HUCLEAR PLANTS.
ALTHOUGH THIS RESEARCH PROGRAM HAS CREATlY IMPROVED OUR UNDERSTANDING OF THE PHYSICAL PHENOMENA ASSOCIATED WITH MELT / CONCRETE INTERACTIONS. THIS WORK SHOULD CONTINUE WITH THE OBJECTIVE OF EXPERIMENTAL VERIFICATION OF THE CORE MELT / CONCRETE INTERACTION COMPUTER MODEL (INTER), SUCH THAT EXTRAPOLATION TO PROTOTYPICAL CONDITIONS CAN BE MADE TO ACCURATELY PREDICT SUCH VARIABLES AS T.4E TIME OF CONTAINMENT MELT-THROUGH OR OVERPRESSURIZATION.
NRR HAS BEEN IN CLOSE CONTACT WITH THE RES STAFF ON RESEARCH RELATED TO MATERIALS INTERACTIONS BETWEEN CORE MELT DEBRIS aND CONCRETE AND MAINIAINS COGNIZANCE OF SUCH WORK IN THE U.S. AND OTHER COUNTRIES.
SD COMMENTS. G. RIVENBARK: NO RESPONSE RECEIVED.
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P&OSPAM OFFICE CONNENTS 04 POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS l
RIL_s: 29 DaTE ISSUED: 06/07/78 RES DECISION UNIT: FUEL BEHAVIOR FIL TITLE:
"FUEL RCD ANALYSIS CGMPUTER CODE:
FRAP-T3" SPCNSORING OFrICE(S): RES RES:
1-12 FUEL CODE RESEARCH PROJECT MGR:
H. SCOTT DEVELOPMENT eel _[t:3fMT): THIS RIL TRANSMITS THE RESULTS OF COMPLETED RESEARCH TO PREPARE AND TEST THE THIRD MODIFICATION OF THE CCP#u!ER CODE FRAP-T (FUEL RCD ANALYSIS PROGRAM - TRANSIENT).
FRAP-T IS A FORTRAN IV COMPUTER CODE BEING DEVELOPED TO PREDICT THE TRANSIENT REPONSE OF A LkR FUEL ROD DURING POSTULATED ACCIDENTS SUCH AS LOSS-OF-COOLANT ACCIDENTS. POWER CCCLING MISMATCH ACCIDENTS. REACTIVITY INITIATED ACCIDENTS. OR INLET FLOW BLOCKAGE ACCIDENTS.
FRAP-T IS ALSO BEING DEVELCPED TO PERFORM THE CALCULATIONS NEEDED FOR PLANNING AND ANALYZING POWER BURST FACILITY AND LOSS OF FLUID TEST EXPERIMENTS.
IN FRAP-T3. THE COUPLED EFFECTS OF MECHANICAL. THERMAL. INTERNAL GAS AND MATERIAL PROPERTY RESPONSE ON THE BEHAVIOR OF THE FUEL RCD ARE C0dSIDERED. GIVEN APPFOPRIATE COOLANT CONDITION AND POWER HISTORIES. FRAP-T3 CAN CALCULATE ROD BEHAVIOR FOR A WIDE VARIETY OF OFF-NORMAL SITUATIONS AND POSTULATCD ACCIDENT CONDITIONS (E.G.,
BWR OR PWR POWER TRANSIENTS. FLOW COASTDCWN. LOAD LOSS CR C03LANT DEPRESSURIZA!ON).
IN THE CCNTEXT CF LWR SYSTEM TRANSIENTS FRAP 15 WELL SUITED TO TO BE USED AS A COMPONENT CODE TO DESCRIBE FINE DETAILS OF FUEL RCD BEHAVICR.
FURTHERM3RE. SENSITIVITY STUDIES WITH TRAP WILL FACILITATE DEFINITIDH OF THE SIMPLEST ACCEPTABLE FUEL DESCRIPTICN IN SYSTEMS CODES.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER RRIEFING BRIEFING RELEASE RESULTS Eq3T_Rlt ACTIVITIES KLviEW NELD COPPLETED HELD HELD ISS9ED IMPLEMENTED OFFICE RESPCNSIBLE......... NRR/SD SCHEDULED CGMPLETION DATE.. 05/30/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL CCMPLETICM DATE.....
J.
VQA EWED_E:
P15CElPE APPLICattpH T O R EGV LAlH3 Y PRQ{ Ell: TRANSIENT FUEL PERFORMANCE CODES ARE REVIEWED BY NRR AS PART OF LOCA AND OTHER ACCIDENI ANALYSES.
NRR USES INDEPENDENT NRC-DEVELOPED CODES FOR AUDIT PURPOSES IN THESE REVIEWS.
LE$CEJ1E IPP ACT OF P[ifl[T} :
FRAP-T3 IS ONE VERSION OF THE FRAP TRANSIENT FUEL PERFORMANCE CODE.
BECAUSE 1HIS IS CNLY AN INTERIM VERSION OF A BEST-ESTIMATE CDDE. IT HAS NOT BEEN ADAPTED FOR LICENSING APPLICATIONS.
IN THE FUTURE NRR PLANS TO USE THE WATER REACTGR ANALYSIS PACKAGE (WRAP) WHICH WILL INCLUDE A MORE CURRENT VERSICN OF FRAP-Y THAT INCCRPORATES CONSERVATIVE MODIFICATIONS. IN PRESENT AUDIT WORK. NRR USES THE WATER REACTCR EVALUATION K0 DEL (WREM) SERIES OF CODES WHICH DOES NOT INCLUDE FRAP-T3.
((NMENTS/EfMAER}: THE NEED FOR A MORE DETAILED TRANSIENT FUEL BEHAVIOR CODE IN LICENSING APPLICATIONS MAS LONG BEEN RECOGNIZED BY NRR.
IT IS EXPECTED THAT A MORE CURRENT WERSION OF THE FRAP-T CODE SERIES WILL BE ADAPTED FCR LICENSING APPLICATIONS. ALTHOUGH CONSERVATIVE MODIFICATIONS WILL BE REQUIRED. THIS CODE WILL CONTAIN MANY OF THE ELEMENTS NOW IN FRAP-T3..
11LS*"ENVS, G. Rf vENA:P(:
Bilfd!PE APPLACATIAN TO PECW131QQV PROC W : TRANSIENT FUEL PEOFOGMANCE CODE 3 ARE DEVIEWED B( NRQ AS PAQT CF LOCA AhD OTHER ACCIDEMI ANALYSEL.
REVIEWS.
NRR USES INDEPENDENT HRC-DEVELOPED CODES FOR AUDIT PURPOSES IN THESE CLSCPIBE,IPPACT Of_Pf3U113:
FRAP-T3 IS ONE VERSION OF THE FRAP TRANSIENT FUEL PERFORMANCE CODE.
BECAUSE IHIS IS GNLY AN INTERIM VERSICN OF A BEST-ESTIMATE CODE. IT HAS BEEN ADAPTED FOR LICENSING APPLICATIONS.
IN THE FUTURE, NRR PLANS TO USE THE WATER REACTOR ANALYSIS PACKAGE (WRAP) WHICH WILL INCLUDE A MORE CURRENT VERSION OF FRAP-T THAT INCORPORATES COMSERVATIVE MODIFICATIONS. IN PRESENT AUDIT WORK, NRR USES TPE HATER REACTOR EVALUAT:0N MJDEL (WREM) SERIES OF CODES WHICH 00ES NOT INCLUDE FRAP-Y3.
[Q"U1NTS/PEFARKS: THE NEED FOR A GORE DETAILED TRANSIENT FUEL BEHAVIOR CODE IN LICENSING APPLICATIONS h*S LCMG BEEN RECOGNIZED SY NRR.
IT IS EXPECTED THAT A MORE CURRENT VERSION OF THE FRAP-T CODE SERIES WILL BE ADAPTED FOR LICCNSING APPLICATIONS. ALTHOUGH CONSERVATIVE MODIFICATIONS WILL dE REQUIRED, THIS CODE WILL CONTAIN MANY OF THE ELEMENTS NOW IN FRAP-T3.
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P60Gaa7 0FFICE CC-MENTS ON POTENTIAL UTILIZATION OR VAlbE OF RESEARCH RESULTS IN THE REGULATORY,EnkCC$5 QIt s: 30 DATE ISSUED: 06/23/78 RES DECISION UNIT SAFEGUARDS PIL TITLE: PHASE I FINAL REPORT, "BARAle! PENETRATION DATA BASE":
OF STUDY, "ASSISTANCE-PHYSICAL PR0!ECTION ASSESSMENTS" 5* CMS 0* INS OFFICE (S): NRR R3$1 NONE RESEARCH PROJECT MGR:
J. MILLER Pf1_CO*?ENTS: THE REPORfED RESULTS TiOVIDED:
(1) A CLASSIFICATION OF BARRIERS IN TERMS OF THE PENETRATION TIME FOR SELECIED CCuMTERMEASURES WHICH AN ADVERSARY MIGHT USE TO OVERCOME THE BARRIER, AND (2) PROCEDURES TO BE FOLLOWED IN TESTING REACTOR SITES FOR CCMPLIANCE FOR to CFR 73.55. THE REACTOR SAFEGUARDS RC;ULATION.
USER DISCUSSION POSITION COMMISSION ACRS PRESS
'0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS E211_PlL_iCTIVITIES EEVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED q
CFFICE RESPGhSIBLE......... NRR
)
SCHEDULED COMPLETION DATE.. 08/28/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL CCMPLETION DATE.....
NPP S MENTS OM 03/)C/7 L W 1LitR1 i
OLiC# J E APPL IC A T ION _10 RE GMt A T ORY PRQf155: AS STATED IN THE RIL, THE BARRIER PGIETRATION DATA BASE IS AVAILABLE As A SUPPLEMEhIAL DATA SOURCE FOR USE IN THE EVALUATION OF LICENSEE FALILITY SAFEGUARD PROGRAMS.
THE DATA PROVIDES A STANDARDIZED BASE OF BARRIER DELAY TIMES AGAINST VARIOUS ADVERSARY COUNTERMEA3URES FRCM CURRENTLY AVAILABLE LITERATURE. THIS DATA WILL PROVIDE FOR MORE RELIABLE AND CONSISTENT EVALUATION OF SAFEGUARD PROGRAMS BY BOTH THE NRR STAFF AND LICENSEE STAFFS.
Eli(DjlLJMCT OF PESUlTE THIS COMPILATION OF DATA IS USEFUL FOR THE DESIGN OF PHYSICAL SECURITY SYSTEMS BY IHE LICEN5EES. As WELL AS IN THE EVALUATION OF THE EFFECTIVENESS OF SUCH SYSTEMS BY NRR.
ALTHOUGH THE RESULTS OF THIS PROGRAM WERE N3T AVAILASLE DURING THE DESIGN PHASE OF THE CURRENT UPGRADING OF PHYSICAL SECURITY AT NUCLEAR PCMER PLAN 15. THIS DATA BASE WILL PROVIDE USEFUL DATA FOR FUTURE APPLICATIONS.
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PPAGP 4M OF FIRE _6frMENVS OM PQiENTIAL 41/LIZAY10N OR vaLUE OF PE$EAPCH RESULT 3 IN THE REGdLATORY PROCESS RIL 8: 31 DATF ISSUED: 07/10/78 RES DECISION UNIT: SAFEGUARDS RIL TITLE:
"ASSAY OF STANDARD REFERENCE EATERIAL (SRM) 950E" SPONSORING OFFICE (S): SD SPG:
4-? MEASUREMENTS A RESEAPCH PROJECT MGR:
R. SHEPARD STANDARDS Pf1_QfE313T3: THYS RIL TRANSNITS THE RESUt?S OF A COMPLE!ED PHASE OF RESEARCH ON THE ASSAY DETERMINATON OF URAN!UM IN SIANDARD REFERENCE r.ATERIAL (SRM) 950B, IN RESPONSE TO A HEED TO IMPROVE Yb2 QUALITY OF MEASUREMENTS MADE ON SPECIAL NUCLEAR MATERIAL FOR CONTROL AND ACCOUNTING PURPOSES. THE PURPOSE OF THE WORK WAS TO DEVELOP AND CERTIFY A URANIUM OXIDE (U308) ASSAY STANDARD TO REPLACE THE VIRTUALLY DEPLETED SRM 950A, USED IN NOND5STRUCTIVE ASSAYS.
TuE RESEARCH RESLLTS INDICATE THAT THE NEWLY DEVELOPED SRM 950B CALIERATION STANDARL HAS A CERTIFIED VALUE OF 9?.97 +/- 0.02 FERCENT URANIUM OXIDE (U308).
THESE RESULTS ARE EXPECTED TO IMPROVE THE STANDARDIZATION ANP CALIBRATION CAPABILITY OF BOTH HRC FIELD INSPECTORS AND THE NUCLEAR INDUSTRY AS A WHOLE; THEY ARE EXPECTED TO HAVE A SIGNIFICANT NEAR-TERM IMPACT CN CURRENT SD GUIDES THAT WILL ADDRESS THE IMPLEMENTATION OF 10 CFR 70.57, LIC(NSEES MEASUREMENT CONTROL PLANS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES _
PEVIfH HEL D COMPLETED HELD HELD ISSUED IMPLEMENTED q
OFFICE RESPO!4SIBLE........ HMSS/NRR NRR SCHEDULED COMPLETION DATE.. 08/30/73 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSC.HED ACTUAL CCMPLETION DATE.....
HMSS COMMENTS, B.
HATTER - TEST APPLICATI0HS ARE BEING CONDUCTED TO DETERMINE USER SUITABILITY.
NPP COMMENTS, W. CAMMItt: NO RESPONSE RECEIVED.
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i PPDGRAM OFFICE CCMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCd R ES'J L T S IN THE RECULATORY PROCESS RIL : 32 DATE ISSUED: 08/03/78 Rts DECISION UNIT: FAST BRIEDER REACTORS RIt TITLE:
IMPROVEMENTS IN THE AEROSOL BEHAVIOR CODE FOR RADIOLOGICAL ASSESSMENTS OF LMFBRS.
SPONSORINS OFFICE (5): RES EPG:
2-7 AEROSOL MODELING RESEARCH PROJECT MANAGER:
J. LARKINS AND PROPERTIES Pii_gp55ENT1: THIS MEMDRANDUM 1RANSMITS THE RESULTS OF CGMPLETED RESEARCH ON THE MEASUREMENT OF SODIUM OXIDE AER050L PROPERTIES.
SODIUM DXIDE IS THE KEY AEROSOL CONSTITUENT IN POSTULATED SEVERE LMFBR ACCIDENTS.
FOR THE MOST St.ERE POSTULATED LMFBR ACCIDENT SCENARIOS (HCDA AND CORE MELT). SODIUM-0XIDE AEROSOL REPRESENTS THE HIGHEST AIRBORNE MASS CONCENTRATIONS IN THE CONTAINMEN) VESSEL AND IS EXPECTED TO DOMINATE AND GOVERM THE BEHAVIOR OF THE FUEL AND FISSICN PRODUCT AER050L. THEREFDRE. AS A FIRST STEP IN IMPROVING THE AEROSOL BEHAVIOR CODE, HAARM-2.
SEPARATE EFFECTS WORK WAS CARRIE7 OUT ON SODIUM-0XIDE AEROSOL.
THE RESULTS OF THESE SEPARATE EFFECTS MEASUREhENTS HAVE BEEN INCORP3 RATED INTO THE MODELS OF THE AEROSOL BEHAVIOR CO HAARM-2. AND TOGETHEP WITH SOME ADDITIONAL IMPROVEMENTS USED TO GENERATE A NEW VERSION CALLED HAARM-3.
THE IMPROVED MODELS IN HAARM-3 PROVIDE A MORE REALISTIC DESCRIPTION OF PARTICLE CHARACTERISTICS AND THEREBY ALLOW IMPROVED ESTIMATES OF SODIUM-0XIDE AEROSDL BEHAVIOR DURING A POSTULATED HCDA.
THE HAARM CODE IS USED BY NRR FOR LMFBR SITE RADIOLOGICAL CONSEQUENCE ASSESSMENT.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING
-BRIEFING RELEASE RESULTS E95T_PIL AqllVIIJf1 REVIEW HELD OFFICE RESPCNSIBLE......... NRR COMPLET.fD HELD HELD ISSUED IMPLEMENTED SCHEDULED COMPLETION DATE.. 09/15/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NPP_fD55ENTS, J.
K.
LONG/T. SPEIS:
DDiRJ_PE APPLEABDUQEGUL A TORY PPOCESS: CODE DESCRIBED IN THIS RIL, HAARM3. IS USED BY NRR IN CALCULATING THE EFFECIS OF FAST REACIOR ACCIDENIS.
11 IS BASED ON A PREVIOUS CODE HAARM2, NOW MODIFIED TO REFLECT THC J
RESULTS OF MILLIKAN CELL EXPERIMENTS AT BCL.
THE GENERAL EFFECT OF THE MODIFICATIONS IS CONFIRMED IN LARGE SCALE TESTS AT HEDL.
PElfEllf_15PfCT OF PESULTS: CONFIRMATION OF THE NEW CODE PERMITS A REDUCTION BY AS MUCH AS A FACTOR OF 2-8 1
IN IHE AMOUNT OF RADIDACTIVE AEROSOLS LEAKED FRCM AN LMFBR CONTAINMENY IN THE EVENT OF A LARGE ACCIDENT.
(RADI0 ACTIVE GASES ARE NOT CONSIDERED TO BE AFFECTED.)
fR$M{ Nil /REMAPK32 THE ATTENUATION OF RADIDACTIVE AIRBORNE MATERIALS B( AGGLOMERATION AND FALLOUT IS AN IMPORTANT FACTOR IN THE C8.LCULATION OF ACCIDENT CONSEQUENCES. EVALUATION AND VERIFICATION OF THIS NATURAL MECHANISM FOR REDUCING RADI0 ACTIVE EMISSIONS HAS SIGNIFICANCE COMPARABLE TO AN ENGINEERED SAFETY FEATURE.
I i _ --.
PROGRAM OFFICE COMMENTS ON POVENTIAL Uillf2ATION OR VALUE 05 RESEARCH QESULTS IN THE REGULAT0QY PROCESS RIL s: 35 DATE ISSUED: 08/03/78 RES DECISION UNIT: FUEL CYCLE SnFET( 3 ENVIRONMENTAL EFFECTS PIL TITLE: PiU'CHIUM ACCIDENT CONTAINER PROGRAM - RESEARCH, DESIGN AND DEVELGP.1ENT.
SPONSORING OFrICE(S) NMSS pp3: 5-11 PLUTONIUM CONTAINER RESEARCH PROJECT MGR W. LAHS CERTIFICATIGN R,J3__(f MME N T3 : RESULTS ARE REPORTED OM THE DESIGN, DEVELOPMENT AND TEST OF IHE PAT-t PLUTONIUM PACKAGE THAT MEETS THE ARC QU AL IFICATION CRITERIA f-UBLISHED IN NUREG-0360 "QUALIFICATION CRITERI A TO CERTIFY A PACKAC-E FOR AIR TRANSPORT OF PLUTONIUM."
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS
[MT RILAITlVIT_If3 REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEENTED OFFICE RESPONSIBLE......... hM.55 SCHEDULED COMPLETION DATE., 09/15/78 UNSCHED UNSCHED UNSCHEU UNSCHE3 UNSCHED UNSCHED ACTUAL CGMPLETION DATE.....
bfdi_CITbnRJH A P P E L L - ST A T E D WORK IS COMPLETE AND NO OTHER ACTIVITY IS EXPECTED. RESULTS WERE USED IN THE CLRTIFICAIION OF THE PAT-t PLUTONIUM PACKAGE TO C0tfGRESS, LICENSING THE PACKAGE FOR USE, AND PROVIDING PROTOTYPES OF THE PACKAGE TO DOE AND IAEA.
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i P&OGRAM OFFICE COMMENTS CN P07ENilat UTIL114 HON OR VALUE OF DESEARCH RESULTS IN THE REGULATORY PROCESS RIL #5 34 DATE ISSUED: 08/03/78 RES DECISION UNIT: FUEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS RIL TITLE: NUCLEAk DECAY DATA FOR RACIONUCLIDES DCCURRING IN ROUTINE RELEASES FROM NUCEEAR FUEL CYCLE FACILITIES.
1 SPONSOP.ING OFFICE (S): NRR EES:
5-24 RADI0 BIOLOGY AND DOSIMETRY RESEARCH PROJECT MGR:
J. FOULKE EE1_.50?~1ENI5 : THIS IS A TABotATTON OF NUCLEAR DECAY DATA FOR 240 RADIONUCLIDES WHICH MIGHT BE EXPECTED TO OCCUR IN ROUTINE dELEASES OF EFFLUENTS FRCM NUCLEAR FUEL CYCLE FACILITIES. THIS CAN BE USED BY NRR AS A ? ASIS FOR ESTIMATION OF RtCIATION EXPOSURE TO MAN.
USER DISCUSSION POSITION COMMISSIDH ACRS PRESS OFFICE MEETING PAPEP.
BRIEFING BRIEFING RELEASE RESULTS fSS1 Rll ACTIVIllEl EfVJLI HELD CC"P L E,.11Q HELD H EL D ISSU R JMPLEMENTED OFFICE RESPONSIBLE......... hMS$
)
5:HEDULED COMPLETION DATE.. 09/30/78 UNSCHED UNSCHED UNSCHED UNSCHED dNSCHED UNSCHED 1
ACTUAL COMPLETION DATE.....
N/A
!L_8L.fD5"EU3 i R -
H-VD1 EE38 pm REMPP LLC Alj oM TO REGUl.ATORY PROCESS: DOCUMENT PRESENTS BASIC NUCLEAR DECAY DATA FOR DEVELOPMENT OF LSIIMATES OF R A D I O8.0GI C A L LOSE.
NRR HAS BEEN EMPLOYING THESE DATA AND IN FUTURE UPDATING OF THE DDSE
)
ASSESSMENT METHODOLOGY WILL USE THIS DOCUMENT AS THE PRIMARY REFE2ENCE.
i Ef t f RJIL., IMP A CT Q F_E E}U t T.} : THE DOCUMENT PROVIDES THE BASIC NUCLEAR DECAY DATA IN THE FORM NECCSSARY FOR l
RADIOLOGICAL ASSESSMENIS. THE DOCUMENT PROVIDES TME STAFF WITH A REFERENCE DOCUMENT 10 ENSURE STAFF'S 1
UNIFCRM USAGE OF DATA BASIC TO ITS EFFORTS.
l
[0M*ENTS/8EMARri A MACHINE READABLE FILE IS AVAILAaLE FROM K.
ECKERMAN, NRR/DSE/RAB.
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~~ PROGRAM Off1CE CCMMENTS dN F0fENVIAL 41AARZAil0N OR VALUE OF PESEAPCH RESUL_T_S IN THE REGUlaiORY FGOCESS FIL 8:
35 DATE ISSUED: 09/IS/7A PES DECISION UNIT: FUEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS PIL TITLES SFACTOR: A COMPUTER CGDE FOR CALCULATING LOSE EQUIVALENT TO A 1ARGET ORGAN PER MICR0 CURIE - DAY RESIDENCE OF A RADIONUCLIDE IN A SOURCE ORGAN SPONSORING OFFICE (S): NRR ERS: S-24 RADI0 BIOLOGY 3 RESEARCH PROJECT MGR:
J. FOULKE DOSIMETRY 911_(Q5ZLNT3: TPE $ FACTOR CCMPUTER CODE CALCULATES S, THE AVFRACE DOSE EQUIVALENT TO EACH OF A SPECIFIED LIST OF IARGET ORGANS PER MICROCURIE-DAY RESIDENCE OF A RADIONUCLIDE IN SPECIFIED SOURCE ORGANS.
THE SFACTOR CODE COMPUTES COMPONENTS OF THE DOSE EQUIVALENT FROM ALFHA PARTICLES, ELECTRONS. GAMMA RAYS, FISSION FRAGMENTS, AND HEUTRONS.
5-FACTORS CAN BE COMPUTED FOR ANY RADIONUCLIDE FOR WHICH DECAY DATA ARE AVAILABLE.
USER DISCUSSION POSITION COMMISSION ACR$
PRESS OFFICE MEETING PAPER.
BRIEFING ERIEFING RELEASE RESULTS PO,$1_g1L ACTIVITIES PEVIEW HELD COMPLETED HELD H~LD ISSUED IMPLEMENTED GFFICE RESPONSIBLE......... NRR UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED SCHEDULED COMPLETI0H DATE.,
ACTUAL COMPLETION DATE.....
MP COMENTS, NEV1M!EP 9, 1979uK. ECKERMAN:
pli PIBE APPLICATIQN YO REGULATORY PPCC111: DOCUMENT PRESENTS COMPUTER CODE FOR ESTIMATING DOSE EQUIVALENT TO SERIOUS IARGET ORGANS PER MICR0 CURIE-DAY RESIDENCE OF A HUCLIDE IN A SOURCE ORGAN.
DESCPI1E_JMPACT OF PESULTS: THE $ FACTOR CODE RESULTS CAN BE USED TO ESTIMATE ORGAN DOSES GIVEN THE ORGAN BURDEN UF A RADIONUCLIDE.
IN ADDITION. IF THE SFACTGR RESULTS ARE EMPLOYED WITH METABOLIC INFORMATION THE DOSE PER UNIT ACTIVITY INGESTED OR INH ALED CAN BE ESTIMATED.
SQU5ENTS/PEMAPK$2 A LISTING OF THE CODE (REVISED TO If;CLUDE ALPHA DOSE TO END0 STEAL CELLS AND BONE MARROW)
IS AVAILABLE FR0!! K. ECKERMAN, NRR/DSE/RAB.
l -
P&GG&AM OFFICE CCMME%f5 ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS DIL er 36 DATE ISSUED:
9/27/78 PES DECISION UN[Ti FAST BREEDER REAC70RS EJ1 TITLE: EVALUATION OF GENERAL ATOMIC CODES:
OXIDE-3, SORS, TAP, AND RECA SPONSORING OFFICE (S)2 NRR EES: 2-12 GAS COOLED REACTOR RESEARCH PROJECT MGR2 J. LARKINS RE1_C0tNENf3: THIS RESEARCH EVALUATED THE APPLICABILITY AND UTILITY OF THE CODES FOR THE ARSR/ GAS COOLED REACIOR SAFETY PROGRAM.
THE OBJECTIVES OF THE EVALUATIONS ALSO INCLUDED AN ASSESSMENT OF THE MODELS AND NUMERICS USED IN THE CODES AND TO NOTE UNDER WHAT CONDITIONS OR FOR WHAT SCENARIOS THE CODES WERE USEFUL. THE APPLICABILITY AND Uf!LITY OF THE GAC CODES FOR RSR-HTGR SAFETY PROGRAMS WAS FOUND TO BE VERY LIMITED. THE OXIDE-3 CODE APPEARS TO HANDLE OXIDATION OF GRAPHITE BY MOSITURE APPROPRIATELY UNDER NORMAL OPERATING CONDITIONS. THE CODE WILL REQUIRE FURTHER EXPERIMENTAL VERIFICATION BEFORE THE LIMITS OF ACCURACY CAN BE ESTABLISHED. THE SORS CODE IS THE FORM PRESENTED FOR OUR EVALUATION APPEARS TO HAVE SOME SERIOUS DEFICIENCIES IN THE MODELS WHICH.' LACES DOUBTS ON THE ANALYSIS PERFORMED WITH THE CODE.
THE TAP AND RECA CODES CIVE GOOD AGREEMENT WITH OTHER ANALYTICAL TOOLS AND APPEAR TO BE USEFUL AHD APPROPRIATE FOR THEIR DESIGNED APPLICATIONS. QUANTIFICATION OF THE ACCURACY OF THE CODES WILL REQUIRE FURTHER CCMPARISON WITH OPERATING REACTOR CONDITIONS. A VENDOR VERIFICATION PROGRAM FOR THE CODES IS RECOMMENDED.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EQ1T_gIl_ASTIVITIES
- EVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... HRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
N_P,5LCOMME N T S.
R.
L.
TEDESCO SORS -
DfifEJ3E APPLICATLEN TO #EGVL ATORY PPDCESS2 THE RIL DESCRIBES WG2K PERFORMED BY BNL TO EVALUATE SORS, A CODE DEVELOPMENT BY GENERAL AIDMIC FOR THE ANALfSIS OF FISSION PRODUCT RELEASE FROM HTGR CORES UNDER TRANSIENT CONDITIONS INVOLVING CORE HEATUPS. A TOPICAL REPORT GA-Al2462, WAS SUBMITTED TO NRR FOR REVIEW WITH THE INTENTION THAT, AFTER BEING ACCEPTED, TdE CODE COULD BE USED AND REFERENCED FOR SAFETY ANALYSES RELATED TO LICENSING OF COMMERCIAL HIGRS.
DESCEJJE IMPACT.Cf_PESULTS: CONFIRMATORY.
NRR'S REVIEW OF THE SORS REPORT AND THE MAXIMUM HYPOTHETICAL FISSIDH PRODUCT RELEASE (MHFPR) MODELS AND DATA HAD PROCEEDED FAR ENDUGH AT THE TIME OF CURTAILMENT OF HTGR LICENSING ACTIVITY (1976) TO PERMIT SOME CONCLUSION 5 REGARDING THE UTILITY GF THE CODE.
THESE CONCLUSIONS WERE PRESENTED IN THE REACTOR FUELS INPUT TO THE CASSAR ISER (DECEMBER 15, 1976).
THE STATEMENTS IN RIL 836 SORS ARE IN GENERAL AGREEMENT WITH, AND THUS CONFIRMATORY OF, NRR'S ASSESSMENT IN THE GASSAR ISER.
CDr5ENTS/PLMAFE$8 AS INDICATED IN THE GASSAR ISER. FOR LICENSING PURPOSES NRR ADVOCATES THE ASSUMPTION OF INSIANIANEOUS RELEASE OF FISSION PRC;UCTS FROM FAILED PARTICLES TO THE PRIMARY COOLANT.
IN EFFECT. THIS ASSUMPTION RENDERS MoST THE SORSG MODELS, BECAUSE SOR5G MODELS TIME-DEPENDENT FISSION PRODUCT RELEASE AND TRANSPORT.
THIS ASSUMPTICH WAS JUSTIFIED IN THE GASSAR ISER PRIMARILY ON THE BASIS OF A POOR DATA BASE FOR FISSION PRODUCT RELEASE AND TRANSPORT IN HTGR FUEL MATERIALS. THE WORK REPORTED IN RIL B36 PROVIDES ADDITIONAL SUPPORT."0R THE VALIDI1Y OF THESE LICENSING ASSUMPTIDHS. IN VIEW OF THE LACK OF HTGR LICENSING ACTIVITY, HOWEVER.
NRR ENVISIONS NO FURTHER HEAR-TERM UTILIZATION OF THE SORS INFORMATION IN THE REGULATORY PROCESS.
CXICE THE RiL DESCRIDES WORK PERFORMED BY BNL TO EVALUATE OXIDE-3, A CODE DE$(EJ1f APPLIC ATION TO REGULATORY PPDCf112 LEWELOPED Bf GENERAL ATOMIC FOR THE ANALYSIS OF HTGR STEAM OR AIR INGRESS ACCIDEN'S.
A TOPICAL CEPORT, GA-A12493, WAS SUBMITTED FOR REVIEW TO NRR WITH THE INTENTICH THAT, AFTER BEING ACCEPTED, THE CODE COULD BE USED FOR SAFETY ANALYSES RELATED TO LICENSING OF COMMERCIAL HIGR*5.NPR'S REVIEW OF THE OXIDE-3 REPORT HAD PROCEEDED THROUGH CONFIRMATORY.
DE1SPIEE IMPACT OF PE1MLT32 AT THAT POINT SUFFICIENT REVIEW HAD ROUND QUESTTON ST AGE BEFORE CURT AILMENT OF GAS-REACTOR LICENSING ACTIVITY.THE LICENSING EVALUATION BEEN CONDUCTED TO ALLOW A PARTIAL EVALUATION OF 1HE CODE.
FUELS INPUT 70 THE GASSAR ISER (DECMPER 15, 1976).
THE RESULTS OF THE ASSESSMENT REPORTED IN RIL 336 ARE ESSENTIALLY THE NRR ASSESSMENT OF THE OXIDE-3 CODE IN 1976.
THE SAME AS, AND THEREFORE CONFIRM,THE COMPARISDN AND APPARENT GOOD AGREEMENT OF OXIDE-3 WITH THE GOPTWO CODE (REPORTED IN CO* MEN 15fp15Aff}s VIZ.
DOES BOT PROVIDE CONCLUSIVE EVIDENCE OF THE ABILITY OF OXIDE-3 YO PERFORM ITS STATED FUNCTION, TRANSIENT ANALYSIS OF STEAM AND AIR INGRES3 EVENTS, BECAUSF. IN CONTRAST TO OXIDE-3, GOPTWO IS A 4TEADY-STATE KIL s36)
SINCE THE TWO CODES WERE DESIGNED FOR DIFFERENT FUNLf!ONS, MEANINGFUL COMPARISDN OF THE C DES COULD CODE.
IN VIEW OF THE BE ACCOMPLISHED ONLY OVER A RELATIVELY NARROW RANGE OF EVENT CONDITIONS AND ASSUMPTIONS.
LACK OF HTGR LICENSING ACTIVITY, NRR ENVISIONS NO NEAR-TERM UTILIZATIDN OF THE INFORMATION ON OXIDE-3 To THE REGUL ATORY PROCESS, BEYOND THE PARTI AL CONFIRMATION IT PROVIDES, AS DESCRIBED HERE, i
- AP AND RECA -
NRR IS PRESENTLY SCHEDULED TO REVIEW THE RECA-3 CODE IN FY 1979.
pf1SEJEE APPLIC3JJ04 TO PEGULATORY PRCCfijsREPORT ORNL-HUREG/TM-178, "EVALUATION OF THE GENERAL ATOMIC CODES TAP AND
[
IHE RESUL15 AND CONCLUSIONS OF THE ORNL 7
ANALYSIS" WILL BE FACTORED INTO THIS REVIEW.
RECA FOR ~ATGR ACCIDENT THE ORNL REPORT IDENTIFIED A NUMBER OF CONCERHS WHICH NRR WILL LOOK INTO DURING DESCPI!E I"P ACT OF PE541T12 THE RECA 3 CODE.
NRR IS PRESENTLY NOT SCHEDULED TO REVIEW THE TAP CODE.
THE ORNL 2
THE SCHEDULED REVIEW OF BE STRONGLY RELIED UPON REGARDING THE INTERFACE BETWEEN THE TAP AND RECA 3 CODES.
REVIEW WILL SSUMENTS/PEFAPES: NONE l
}
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PROGRAM OFFICE CCM9Ehis 04 P O T EN T I A L UTILIZATION OR WALUE OF RESEhE6? RlSutiS IN THE REGUL ATORY PROCESS RIL #2 37 DATE ISSUED: 09/29/78 RES DECISION UNIT: LOFT RIL TITLE: LOFT REACTOR SAFETY PROGRAM RESEARCH RESULTS THROUGH OCTOBER 1,
1978 SPCNSORING OFFICE (S): NRR PPj:
1-1 LOFT RESEARCH PROJECT MGR8 G. MCPHERSON PEl_(DM5ENIS-THE LOFT kESEARCH PROGRAM HAS BEEN DEVELOPED TO PROVIDE EXPERIMENTAL INFORMATION RELEVANT TO THE LICENSING CRITERI A FOR LARGE COMMERICAL PWR'S.
THE MAJOR PORTION OF THIS PROGRAM IS DIRECTED AT AN IMPROVED UNDERSTAHDING OF THE LOSS-OF-COOL ANT ACCIDENT (LOCA) AND THE PERFORMANCE OF EMERGENCY CORE COOLING SYSTEMS USING THERMAL-HYDRAutIC, CORE PHYSICS, STRUCTURAL AND FUEL BEHAVIOR DATA OBTAINED THROUGH A SERIES OF LOSS-OF-COOLANT EXPERIMENTS. THIS RIL IS BASED ON DATA OBTAINED FROM THE FIRST SERIES OF EXPERIMENTS, Lt. WHICH WAS PERFORMED IN THE ABSENCE OF NUCLEAR POWER.
IN THE FINAL EXPERINENT OF TPIS SERIES, Lt-5, THE CORE WAS IN PLACE, BUT IN A SHUTDOWN CONDITION.
CONSECUENTLY, THE RESULTS DERIVES FROM THESE INVESTIGATIONS ARE APPLICABLE ONLY TO THE THERMAL-HYDRAULIC AND STRUCTURAL PHENOMENA ASSOCIATED WITH THE LOCA WITH THE EMERGtNCY CORE COOLING (ECC) INJECTION.
IN GENERAL THE RESULTS SUPPORT THE CONSERVATIVE INTENT OF THOSE PORTIONS OF THE EVALUATION MODEL REQUIREMENTS CONTAINED IN THE LICENSING CRITERIA WHICH WERE INVESTIGATED IN THE L1 SERIES. IN PARTICULAR, THE TIME DELAY IN THE DELIVERY OF EMERGENCY CORE COOLANT TO THE LOWER PLENUM DUE TO THE EFFECT OF CONTACT WITH THE HOT METAL SURFACES (THC HOT WALL EFFECT) WAS FOUND TO BE SMALL (0.5 TO 1.0 S).
BASED ON THE RELATIVE SURFACE AREA TO VOLUME RATIO OF THE DOWNCOMER, THE HOT WALL EFFECT IN A LPWR SHOULD BE LESS THAN IN LOFT.
THE RESULTS OF THE LOFT Lt SERIES ABOVE ARE RECOMMENDED FOR USE BY hRR IN ITS INTERPRETATION AND APPLICATION OF LOCA ECCS EVALUATION MODEL CRITERIA AND RELATED CODES.
ALTHOUGH THE DATA ARE LIMITED TO NONNUCLEAR BLOWDOWN CONDITIONS. THE PREDICTIONS TO WHICH DATA CGMPARISDNS HAVE BEEN MADE HAVE ASSUMED APPROPRIATE INITIAL CONDITIONS AND THEREFORE THE CONCLUSIONS ARE BELIEVED 19 BE VALID.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOST RIL ACTIVlllEl EEy1EM HELD QMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NKR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL C0MPLETION DATE.....
&LC DME NT11_IA W A R Y 1980, R, DENISE:
Df.i RIBE _APPll?A M N_TO PEGG A N Y PRQ_Cf13: THE LOFT TEST FACILITY PROVIDES EXPERIMENTAL DATA FOR THE EVALUATION OF CUMPUIER PRUGRAMS USED TO ANALYZE THE CONSEQUENCES OF POENTIAL LOSS OF COOLANT ACCIDENTS. THE TEST FACILITY IS St*FFICIENTLY LARGE AND REPRESENTATIVE OF A L ARGE PWR TO TEST THE BASIS PHYSICAL PHENOMENA 0F A PWR LOCA.
THE COMPUTER PROGRAMS ARE USED TO PROVIDE THE EXTRAPOLATION FROM THE TEST CONDITIDMS TO FULL SCALE REACTORS.
DESKR13E_InPACT Or REipl11: THE RESULTS OF THE NON-NUCLEAR BLOWDOWN TESTS HAVE BEEN USED TO EVALUATE REACTOR FUEL SUPPLIER AND NRC DEVELOPED COMPUTER PROGRAMS. THESE EVALUATIONS HAVE PROVIDED VALUABLE INFORMATION FOR EVALUATING VARIOUS MODELS IN THE CCMPUTER PROGRAM, AND HAVE INDICATED CERTAIN AREAS WHERE THE MODELS COULD BE IMPROVED. IN GENERAL, THE COMPUTER PROGRAMS USED TO PERFORM EVALUATION MODEL CALCULATIONS HAVE BEEN SHOWN TO BE SATISFACTORY FOR THE PURPOSE.
EXPERIMENTAL RESULTS FROM THIS PROGRAM DEMONSTRATING ASYMMETRICAL FLOW IN THE DOWNCOMER DURING ECC INJECTION, PRIMAFr COOLANT RETENTION IN THE LUWER PLENUM, NEAR ISENTROPIC GAS EXPANSION IN THE ACCUMUL ATORS DURING ECC INJECTION, AND PRL-50RIZER DISCHARGE EFFECTS IN THE FIRST FEW SECONDS BY BLOWDOWN HAVE NOT YET BEEN FULLY IMPLEMENTED IN COMPUTER MODELS.
SUCH CHANGES WOULD FROVIDE IMPROVED 'BEST ESTIMATE' SIMULATIONS OF THE LOCA TRANSIENTS IN SUB-S EQ'J LH T TESTS..
CQ""ENT5eREMAPK3 THE LOCA COMPUTER PROGRAMS HAVE BEEN EVALUATED AGAIM5T VA2IOUS SPECIAL AFFECTS AND INTECS:ATED TESTS RESULIS.
TH IPRIMA2Y NEED FC2 CODE EVALUATION 15 COMPA2ISON KITH A FULLY INTEGRATED TESTS INCORPORATING NUCLEA2 FUEL.
THE CONTINUATION OF THE NUCLEAR PORTION OF THE LOFT FROGRAM SHOULD PROVIDE MUCH OF THE NEEDED EXPERIMENTAL DATA FOR CODE VERIFICATION. PROMPT IMPLEMENTATION AND CHECKDUT OF MODEL CHANGE 5 MADE TO INTEGRATE INFORMATION LEARNED FROM PREVICUS TESTS ARE REQUIRED TO DERIYE THE FULL BENEFIT OF INFORMATION LEARNED IN EACH STAGE OF THE EXPERIMENTAL PROGRAM.
IT IS N3T APPARENT T!!AT SUCH AN EFFORT IS BEING MADE FROM THE
SUMMARY
PRESENTED IN THE RIL.
SILSDfnENTS. 5 _H_lifM&N - THE ENGINEERING METHODOLOGY STANDARDS BRANCH WILL USE THESE RESULTS IN THE PHASE II ECCS kutt CHANGE EFiORT. L l
PP0GR AM OFFICE CCM9ENis ON POIENTIAL UTILIZATION OR VALUE OF 8ESEARCH RESULTS IN THE REGULATORY PROCESS 1
RIL #2 38 DATE ISSUED:
10/13/78 RES DECISION UNIT: FAST BREEDER REACTORS PIL TITLE: RESULT OF THE INITIAL SERIES OF ACPR EXPERIMENTS ON PROMPT-BURST ENERGETICS WITH FRESH OXIDE FUEL SPONSOPING OFFICE (5): NRR EES:
2-6 ACCIDENT ENERGETICE E15EARCH PROJECT MGR R. WRIGHT PEi_fGrMENTS-THE RESULTS OF THESE ACPR TESTS ON PROMPT-BURST ENERGETICS GIVE NRR A MUCH IMPROVED DATA BASE FGR ASSESSING THE WORK PDTENTIAL AND THE RESULTING THREAT TO THE INTEGRITY OF THE PRIMARY SYSTEM AND EVENTUALLY THERMAL INTERACTION BETWEEN MOLTEN OXIDE FUEL AND SODIUM HAS BEEN DEFINITELY DEMONSTRATED. THE OBSERVED a
CONVERSION OF FUEL THERMAL ENERGY INTO WORK, HOWEVER. WAS LESS THAN ONE PERCENT.
THE SHOCK-TRIGGERED DELAYED I
FUEL-SODIUM INTERACTIONS OBSERVED IN THESE EXPERIMENTS RAISE QUESTIONS ABOUT THE POSSIBLE OCCURRENCE OF A LARGE SCALE PROPAGATING FUEL-COOLANT INTEkACTION UNDER CORE MELTDOWN CONDITIONS. AS PROPOSED BY BOARD.
THE CURRENT EXPERIMENTAL RESULTS DO NOT, HOWCVER, GIVE INFORMATION ON THE EXTENT (MASS INVOLVEMENT) 0F SUCH A PROPAGATING INTERACTION OR ON THE WORK POTENTIAL OF SUCH AN INTERACTION. IT IS RECOMMENDED THAT CONSIDERATION BE GIVEN BY NRR TO THESE UNCERTAINTIES AND TO THE CURRENT ABSENCE OF VERIFIED MECHANISTIC MODELS OF FUEL-COOLANT INTERACTIONS WHEN ASSESSING ACCIDENT WORK POTENTIAL.
THE EXPAND FRESH FUEL FAILURE CODE HAS BEEN VERIFIED FOR THE RANGE OF CONDITIONS COVERED BY THESE EXPERIMENTS.
AND IS THE BEST ACCIDENT ANALYSIS CODE AVAILABLE FOR THESE CONDITIONS. THE FRE-FAILURE IN-CLAD AXIAL FUEL MOTION PREDICTED BY EXPAND HAS BEEN VERIFIED BY POST-TEST EXAMINATION IN THESE EXPERIMENTS. THIS FUEL MOTION MAY HAVE REACTIVITY EFFECTS THAT ARE SIGNIFICANT IN ASSESSING THE ACCIDENT WORK POTENTIAL AND THE THREAT TO THE INTEGRITY 0F THE PRIMARY SYSTEM AND EVENTUALLY THE CONTAINMENT. EXPAND IS RECOMMENDED TO NRR AS THE BEST AVAILABLE TOOL FOR SAFETY ASSESSMENT FOR FRESH OXIDE FUEL IN THESE AREAS.
USER DISCUSSION POSITION COMM*SSION ACRS PRESS GFFICC MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS tQSl__RIL ACTIVITIES FEVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... hRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACT8?AL COMPL ETION DAT E.....
t!PP E rM[NTS ON JULY 19, 1979.
J. MEYER:
PfliEIDE_APPil(allen TO PEGULAIDRY PPDCill: THERE IS NO DIRECT APPLICATION OF THE RESULTS OF THE INITIAL SERIES OF PBE EXPLRIMENTS TO THE REGULATORY PROCESS.
THE EXPERIMENTS ARE FIRST GENERATION AND ARE DIFFICULT TO INTEPRET. NOTHING CONCLUSIVE CAN BE SAID FROM THESE EXPERIMENTS REGARDING SHOCK-PRESSURE TRIGGERING OF SIGNIFICANT FUEL COOLANT INTERACTIONS.
i dei (FitE_ IMP A C T OF RESU113: EXPERIMENTAL RESULTS LEADING TO A BETTER UNDERSTANDING OF FCI PHENOMENA AND PROPAGATION MECHANISMS WILL HAVE SIGNIFICANT IMPACT ON UNDERSTANDING OF LOW PROBABILITY HIGH CONSEQUENCE ACCIDENTS AND THEREBY ON LICENSING.
THESE EXPERIMENTS, IN AND OF THEMSELVES. CONTRIBUTE LITTLE BUT DO OFFER A STARTING POINT IN THE DIRECTION OF PROVIDING UNDERSTANDING OF FCI PHENOMENA.
4 COMMENTS /PEMAFKS: NONE _.
PROGRAM OFF'CE CCMMENTS JN POTENTIAL UTILIZATION GR VALUE OF RESEARCH RESULTS IN THE REGULAT04Y Pw0 CESS
~
RIL 8: 39 DATE ISSUED:
11/27/78 RES DECISION UNIT: CODE DEVE_0PMENT l
PIL IITLE RELAP-4/ MOD 6 SPONSCRING OrFICE(S): HRR Regr 1-16 REFERENCE SYSTEM RESEARCH PROJECT MGRt S. FABIC CODE PE3_fpr3{NTS: MOD 6 PETAINi Ali Or IME CAPABILITY OF PRIOR VERSIONS AND. IN ADDITION, CDHTAINS NEW BEST ESTIMATE (BE) BLOWDOWN HEAT TRANSFER MODELS, BE REFLOOD CAPABILITY 3 AND OTHER MODIFICATIONS.
RELAP-4/ MOD 6 WAS DEVELOPED TO FROVICE CAPABILITY FCR A PWR STATISTICAL LOCA STUDY AND FOR SENSITIVITY STUDIES TO ASSESS DATA PERTAINING TO LOCA RULE CHANGES AND TO IMPROVE BLOWDOWN AND REFLOOD UNDERSTANDING AND PROVIDE QUANTITATIVE
- PRAISALS OF ECCS.
MOD 6 IS AN EXTENSION OF PRIOR LOCA CODE CA? ABILITY TO ALLOW MODELING OF LWR AND EXPERIMENTAL FACILITY REFLOOD PHENCMENA. IN ADDI1 ION TO CALCULATION OF THE BLOWDOWN PHASE OF THE EVENT. RES IS APPLYING THE CODE TO THE UNCERTAINTY STUDY REQUESTED BY NRR. AS WELL AS TO INTERPRETATION OF TEST RESULTS FROM LOCA FACILITIES.
}
THE RELAP-4/MDD 6 CODE IS RECOMMENDED FOR THE BEST ESTIMATE CALCULATIONS OF BLOWDOWN AND REFLOOD.
j RELAP-4/ MOD 6 IS NGT RECOMMENDED FCR REFILL ANALYSES BECAUSE OF THE DIFFICULTIES ASSOCIATED WITH THE NON-EQUILIBRIUM PHEONOMENA WHICH ARE NOT MODELED. DOWNCOMER MODELING IS ALSO A PROBLEM, AND THE CODE DOES NOT MODEL NITROGEN FLOW i
i IF THE ACCUMULATCR SHOULD EMPTY DURING THE ECC BYPASS AND REFILL PHASE OF THE EVENT.
As WITH MOD 5, THE CODE IS y
VERY SLOW RUNNING DURING REFILL, GIVING CALCULATED RESULTS WHICH ARE NOT SATISFACTORY. THIS CODE IS NOT RECOMMENDED
]
FOR STEAM GENERATOR TUBE BREAK INVESTIGATIONS, ALTHOUGH USER GUIDELINES ARE BEING DEVELOPED TO ATTEMPT LIMITED INVESTIGATIONS WITH MOD 6.
1 1
l MOD 6 HAS BEEN SENT TO THE ARCONNE CODE CENTER, FRANCE (NEA FOR EUROPEAN DISTRIBUTION), ITALY, NRC, ORNL AND sat:DIA. IT IS IN USE ON A NUMBER OF NRC-FUNDED PROGRAMS INCLUDING ANALYSIS OF SEMISCALE MODE 3 LOFT, AND PKL, 9
AND PLANS ARE BEING IMPLEMENTED FGR ITS APPLICATION 10 STANDARD PROBLEMS. THE FOREIGN AND DOMESTIC RECIPIENTS i
0F REL AP-4/ MOD 6 WILL BE ADVISED OF THE CODE ASSESSMENT.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POSI_Rll_ACH VUIES MVlLH HELD COMPLETED KLD HELD ISSUED IDE F"ENTED OFFICE RESPONSIBLE......... hRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED l
ACTUAL C0"PLETION DATE.....
4 ILLCFEN.l$t_LAfW ?Y M3 0, _R.
D{NM 3 DL1G Rifd_A P PL I ? All1W 10 PlWLA T DvY PRORi$ s RELAP4/ MOD 6 IS DESIGNED TO PERFORM A 'BEST ESTIMATE' CALCULATION OF PWR LOCA'S.
THE CODE IS USLD TO PERFORM PRE-AND POST-TEST ANALYSES OF LOCA EXPERIMENTS TO CONFIRM THE ADEQUACY OF THERMO AND HYDRODYNAMIC MODELS USED TO COMPUTE LOCA TRANSIENTS. ON THE BASIS OF OUTPUT COMPARISONS BETWEEN 'BEST ESTIMATE' I
AND 'EVALUTION MODEL' CODE CALCUL.TIONS, THE MARGINS OF CONSERVATISM IN 'EVALUATIOh MODEL' CALCULATIONS CAN BE ASSESSED.
Dr.5IRIDE_lMPACT OE_RESULTS: THE RELAP4/ MOD 6 CJDE DEVELOPMENT IS AN EXTENSIDH OF EARLIF' DEVELOPMENT EFFORTS ON j
klLAP4/ MODS.
MANY IMPROVEMENTS CONFIRMED IN MOD 6 WOULD BE USEFUL AND APPLICABLE FOR 'E,ALUATION MODEL' CALCULATIONS, l
BUT CANNOT BE USED FDP LICENSING AUDIT STUDIES UNTIL MOD 6 IS FURTHER MODIFIED TO INCLUDE ' EVALUATION MODEL' FEATURES.
THE PRIMARY IMPACT OF THE MOD 6 DEVELOPMENT IS IN PROVIDNG A 'BEST ESTIMATE
- B ASE FOR ASSESSING THE CONSERVATISM OF I
THE PRESENT ' EVALUATION MODEL' IN MODS TO BE USED (N AUDIT STUDIES OF VENDOA ANALISES FOR PWR'S, AND IN PROVIDING AN ANALYTICAL UNDERSTANDING OF VARIOUS EXPERIMENTAL PROGRAMS.
CSEdEHT3fpEMAPfjr THE PRIMARY AREA 0F IMPROVEMENTS IN MOD 6 DVER THE MLDS VERSION IS IN THE REFLOOD PHASE OF THE PWR LOCA WHERE MORE DETAILED THERMO AND HYDRODYNAMIC MODELS ARE EMPLOYED TO COMPUTE THE ADVANCE OF FLUID UP THROUGH THE CORE IN THE REFLOOD TRANSIENT.
COMPARISONS WITH TEST RE5ULTS, SHOW GOOD HYDRODYNAMIC AGREEMENT, BUT ONLY BROADLY ACCEPTABLE THERMODYNAMIC AGREEMENT, BOTH OF WHICH ARE BETTER THAN CAN BE OBTAINED FROM MODS.
l
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PP0GR AM OF FIC E coven 15 ON FOT EN TI AL U TILIZA TION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 8:
40 DATE ISSUEDs 12/18/78 RES DECISION UNIT: FAST BREEDER REACTORS QIL TIT 11: THE CCMPUTER CODE BRENDA - A COMPUTER PROGRAM F0k THE DYNAMIC SIMULATION OF A LIQUID METAL FAST PRELDER REACTOR PLANT 1
SPON509ING OFFICE (5): NRR E2Q:
2-13 FAST REACT 0d PESEARCH PROJECT MGR P. WOOD SYSTEMS CODE AND ACCIDENT ANALYSIS P F $_ CJP"IINl$ 2 BRENDA IS A FAST PUNNING SYSTEM CODE INTENDED TO PROVIDE NRC WITH THE CAPABILITY OF DOING QUICK PARAMETRIC SURVEYS AND SCOPING STUDIES OF MORMAL OPERATING A5 WELL AS ACCIDENTAL 1RAN51ENTS IN LMFBR PLANTS.
THE CODEL FROVIDE NRC WITH AN INDEPENDENTLY DERIVED TOOL FOR SAFETY ASSES 5 MENT.
IT IS CURRENTLY BEING MODIFIED TO MAKE IT APPLICABLE TO PREOPERATICHAL AND STARTUP TESTS IN FFTF.
i USER DISCUSSION P05ITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS P.Q3]_PJLLC.THUICE ELV1Ly HELD CCMPLETED HEt D HELD ISSUED IMPLEME*+JJ OFFICE RESPONSIBLE......... NKR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
Iff 8LJGMLN_.lhMVlEIIPd3d17 9 n J._.fEYE8 :
K1C R IB LA"ELI C Alik%.__ID_fL GU L A I O P Y P 9 0_CE$3 SINCE THERE IS NO LMFBR LICENSING ACTIVITY AT THE PRESENT TIME, THERE IS NO IMMEDIA'fL APPLICAIION OF THIS COMPUTER CODE 10 THE REGUL ATORY PROCESS. HOWEVER, IF THE LMFBR LICENSING AGAIN BECOMES ACTIVE, THIS FROGRAM HAS P0(ENTIAL APPLICATION IF THE CLAIMS REGARDING ITS UTILITY ARE, IN FACT, REALIZED.
i LIMPlBE IMPAQT OF REW1TS: THIS RIL ONLY DESCRIBES THE CCMPUTER CODE ITSELF (TOGETHER WITH A COMPARISON BETWEEN BRENDA l
AND 55C) AND 1HUS IHERE ARE NO 'RESULTS' WHICH NRR CAN EVALUATE AND ASSESS. THE IMPACT OF THIS CODE WILL CHLY BE CONTRIBUTORY IF: A) IT TURNS OUT TO BE FAST-RUNNING FOR IMPORTANT APPLICATIONS; B) IT ADEGUATELY TRACKS SYSTEM BEHAVICR AS MEASURED AGAINST A MORE DET AILED. MECHANISTIC SYSTEMS CODE SUCH A5 55(,; AND C) THE DETAILED SYSTEMS CODE ON WHICH IT 15 BASED HAS BEEN PROPERLY VERIFIED.
))
CCM9ENTS/PEMARK5: NONE t
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PQQGP AM Of flCE CC"9ENTS ON POVENflAL UTXLJZATION OR V A L U E O F R ES E A R CM R E5y kik_AN THE REQMLAf0RV PROCESS PIL er 41 DATE ISSUED 12/19/78 PES DECISION UNIT: SEISMIC. ENGINEERING & SITE SAFETY RIL TITLE: LABORATORY TESTING PROCEDURES TO DETERMINE THE CYCLIC STRENGTH OF SDILS SPON50 PING OFFICE (5): SD RPL2 3-1 NRC/ STATE TEGIONAL RESEARCH PROJECT MGR:
H. STEUER ELATH SCIENCES PES CorrfMI): WICELY VARYING RESULT 5 WERE BEING OBTAINED IN TESTING SOILS FOR LIQUEFACTION POTENTIAL (CYCLIC SIRENGTH OF SOILS) BECAUSE STANDARD TEST PROCEDURES DID NOT EXIST.
HENCE. THE SUBJECT STUDY WAS CONDUCTED IN RESPONSE TO THIS FACT. AND BECAUSE STANDARD PROCEDURES ARE NEEDED FOR NUCLEAR POWER PLANT SITE INVESTIGATIONS. THE RESEARCH PROGRAM WAS CONDUCTED IN CONJUNCTION WITH DEVELOPMENT OF ASTM PERFORMANCE SPECIFICATIONS.
MAJCR S0lt MECHANICS LAB 02ATORIES THRCUGHOUT THE WORLD WERE CONTACTED TO DETERMINE THEIR 1ESTING METHODS AND TO EVALUATE THE CYCLIC TRIAXIAL EQUIPMENT IN USE.
THIS INFORMATION PROVIDED A BASIS TO DEVELOP THE TEST PROCEDURES FOR STRESS-CONTROLLED CYCLIC TRIAXIAL STRENGTH TESTS PRESENTED AS A PERFORnANCE SPECIFICATION 50 THAT A GEOTECHNICAL TESTING LABORATORY CAM: 1) ENSURE THAT 175 TEST EQUIPMENT AND PROCED'JRES MEET REQUIRED STANDARDS. AND 2) CHECK THAT RESULTS AGREE REASONABLY WITH RESULTS OBTAINED BY OTHER LABORATORIES.
RESEARCH RESULT 5 PRESENT RECOMMENDED PROCEDURES FOR LIQUEFACTION POTENTIAL TESTING BY USE OF STATE-OF-THE-ART TESTING TECHNIQUES A5 A PERFORMANCE SPECIFICATION.
IT 15 RECOMMENDED THAT NUREG-0031 BE USED A5 GUIDELINES BY THE OFFICE OF STANDARDS DEVELOPMENT IN DEVELOPMENT OF APPLICABLE REGULATIONS AND REGULATORY GUIDES. BY THE OFFICE OF NUCLEAR REACTOR REGULATION TO ASSIST IN THE REVIEW OF NUCLEAR POWER PLANT OPERATING APPLICATICMS. AND BY THE APPLICANT A5 A STANDARD IN THE ASSESSMENT OF LIQUEFACTION POT [NTIAL OF FOUNDATION SOILS AT NUCLEAR POWCR PLANT SITES.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS P_Dil PlL_ACTIVIllii REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED UfflCL RESPONSIBLE......... SD SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
iP_CCEE[NJS, R. M.!N9(21 - THE INFORMATION IN THIS REPORT WAS USED AND REFE"ENCED IN REGULATORY GUIDE 1.133.
"tABORATORY INWL5IIGATIONS OF SOILS FCR ENGINEERING ANALYSIS AND DESIGN OF NUCLEAR PCWER PL ANT 5".
IT WAS ALSO USED IN THE DEVELOPMENT OF A DRAFT REGULATORY GUIDE ON. "PROCEDURES AND CRITERIA FOR ASSESSING
)
SOIL LIQUEFACTION POTENTIAL AT NUCLEAR FACILITY SITES." WHICH WILL BE PUBLISHED FOR COMMENT IN THE HEAR FUTURE.
COPIES OF THE REPCRT WERE WIDELY DISTRIBUTED TO MEMBERS OF THE ASTM. D-18 COMMITTEE ON SOIL AND ROCK ENGINEERING AND WILL BE USED A5 ONE OF THE BASES FOR DEVELOPING A NATIONALLY ACCEPTABLE DYNAMIC TRIAXIAL TESTING STANDARD FOR SDILS.
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PROGDAM OF&lCE COMMEMf5 ON POTE%fIAL UTILIZATION OR VALUE OF PESEARCH RESULIS IN THE REGULATORY PROCESS PLL_ss 42 EATE ISSUED:
12/20/78 PES DECISION UNIT: FAST BREEDER REACTORS pit TITLE: CRITICAL EXPF.RIMENT PROGRAM FCR NEUTRONICS CODE VERIFICATION
]
SPON50 PING OFFICE (5)2 NRR gen:
2-13 FAST REACTOR RESEARCH PROJECT FGRs P. WOOD SYSTEMS CODE AND ACCIDENT ANALYSIS l
l PEi_fD55EMIS THIS PRCGRAM OF DISTORTED GEOMETRY CRITICAL EXPERIPENTS WAS PLANNED AND CARRIED OUT TO PROVIDE BLNCHMAkKS FOR THE VALIDATION OF THE NEUTRONICS PART OF CODES USED IN SAFETY ANALYSIS,.
A5 SIMMER.
A SECOND i
OBJECTIVE 15 TO VALIDATE THE VIM MONTE CARLD CODE FOR USE AS A SECONDARY STANDARD FOR VALIDATION OF OTHER NEUTRONICS METHODS.
MELTDOWN CONFIGURATIONS IN LMFERS CAN BE EXPECT [D TO HAVE REGIONS WITH HIGH FUEL CONCENTRATIONS GIVING EXTREME SPECTRAL l
CHANGES AND L ARGE REGIONS OF VOID GIVING RISE TO 1 ARGE STREAMING PATHS.
NEUTRONICS METHODS OTHER THAN MONTE CARLD HAVE DIrFICULTY IN CALCULATING THESE CONFIGURATIONS ACCURATELY.*
A SERIES OF EXPERIMENTS WAS NEEDED TO DETERMINE THE IM.'ORTANCE OF THESE DIFFICUL11ES AND TO PROVIDE A BASIS FOR IMPROVING THE ACCURACY AND RELIABILITY OF ACCIDENT ANALYSIS I
METHODS.
THIS PROGRAM HAS DEMONSTRATED THAT DIFFUSION THEORY HEUTRONICS CALCULATIONS %DULD UNDERPREDICT RAMP RATES WHICH MIGHT OCCUR IN A MELTDOWN AND COULD BE NON-CONSERVATIVE.
THE RESULT 5 OF THE PROGRAM LEAD TO THE CONCLUSION THAT DIFFUSION THEORY CALCULATIONS LEAD TO NONCONSERVATIVE ESTIMATION OF REACTIVITY GOING FROM THE REFERENCE TO THE SLUMPED CONFIGURATIONS. THE SIGNIFICANCE OF THE RAMP RATE AT PROMPT CRITICAL IN AN HCDA CALCULATIONS HAS BEEN POINTED OUT BY THE NRR STAFF IN NUREG-0122(1).
DATA REDUCTION AND PREPARATION OF THE FINAL REPORT WILL BE COSPLETED IN FY 1979.
A VIM MONTE CARLO CALCULATION ON i
CONFIGURATION
- SODIUM VOIDED TEST 20NE WILL BE MADE AND THE SN CASES NOT SHOWN ON FIGURE 2 OF 2PR-TM-327 WILL BE COMPLETED.
ANALYSIS TO RESOLVE CkD55 SECTION DIFFICULTIES THAT CAUSE DIFFERENCES BETWEEN EXPERIMENT AND VIN MONTE CARLD EIGENVALUES IS NEEDED. WHEN AND IF THE CROSS SECTION DIFFICULTIES ARE RESOLVED, IT WOULD BE DESIRABLE TO PREPARE SECONDARY BENCHMARE VIM MONTE CARLD CALCUL ATIONS USING A HOMOGENIZED MODEL OF THE EXPERIMENT AL CONFIGURATION.
f USER DISCUSSION POSITION COMMISSION ACR5 PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULT 5 i
EQ57_PIL,_ACT1v1TJf1 PfVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED i
0FFICE FESPONSIBLE......... hRR SCHEDULED COMPLETION DATE.,
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCNED l
ACTUAL COMPLETION DATE.....
I
!!RP fDPMEMI.52_NWL*TER 9 1979. J J EVES:
m AL52HLARLEADEN 10 PLWL A.10PY PNGCESS: BECAUSE THERE IS NO LMFBR BEING LICENSED. THERE IS NO IMMEDIATE AF PLICAIIGH 30 IHE REGULAIORY PROCESS.
HOWEVER, SINCE CORE DI5RUPTION ACCIDENT ANALYSIS WILL PLAY AN IMPORTANT PART IN FUTURE LMFER LICENSING. THERE 15 A REGUL ATORY MEED FOR RES EXPERIMENTAL PROGRAMS DIRECTED TO ACCIDENT i
ANALYSIS-CODE VERIFICATION.
THIS PROGRAM IS ONLY A VERY SMALL PART OF THE TOTAL CODE VERIFICATION PROGRAM AND ITS UTILITY AND APPLICATION WILL ONLY BE REALIZED WHEN LARGER (AND MORE IMPORTANT) PARTS OF THE VERIFICATION PROGRAM (E.G.,
SIMMER THERMAL MYDRAULICS) ARE COMPLETED.
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l PPOGW A*f 0F F IC E CC9L% f 5 0'* FD f EM f1 A L UT_ILI2ATION OR VALUE OF RESEARCH RESULTS IN THE REGUL ATORY PRQCESS i
f RIL er 43 DATE ISSUED: 01/10/79 RES. DECISION UNIT FAST BREEDER REACTORS RIL TITLES SUPER SYSTEM CODE. A COMPUTE 2 PROGRAM FOR DYNAMIC SIMULATION GF LMFBR POWER PLANTS 4
$PDN50 RING OFFICE (5)* NRR P1Q:
2-13 FAST REACTOR EE}EARCH PROJECT MQEt P. WOOD l
SYSTEMS CODE AND I
ACCIDENT ANALYSIS i
P,E5_COTEND s THE SUPE 8 SYSTEMS CODE (55C-L) 15 SPECIFICALLY DIRECTED TO THE ANALYSIS OF THE ADEQUACY W
OF NAIURAL CIRCULATION IN SODIUM-COOLED REACTORS TO PREVENT CLAD MELTING.
THE CODE ALSO HAS THE CAPABILITY j
TO ANALYZE NDPMAL OPERATING TRANSIENTS AND LESS SEVFRE ACCIDENTS THAT DO NOT BREACH THE INTEGRITY OF THE SYSTEM g
CP FUEL.
THE CODE IS OPERATIONAL ON THE BNL-CDC-7600 COMPUTER AND C AN BE OPERATED THROUGH THE NRC TERMINALS AT THE PHILLIPS AND WILLSTE BUILDINGS.
l USER DISCUSSION POSITION C0"MISSION ACRS PRESS OFFICE MEETING PAPER BRIEF.NG BRIEFING RELEASE RESULT 5 E95LUL_e cilv_UH5 EEYlIH HELD COMPLETED HELD EQD ISSUED IMPLEMENTED 0FFICE kL5PONSIBLE........, NRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
I fi3LP CDM* fell _W E53E11 -
a DLMRJ11_ATJ C9119MD PIDEL AlDRY PREfi)2 THEPE IS NO IMMEDIATE APPLICATION OF SSC TO THE REGULATORY PROCESS (518 _
1HERE IS NO LMF PR LICLNSING ACTIVIIY) EXCEPT FOR THE SSC VERIFICATION ACTIVITIES ASSOCIATED WITH THE FFTF NATURAL CIRCULATION TESTS.
IF LMFBR LICENSING AGAIN BECOMES ACTIVE, IT WILL BE IMPORTANT TO HAVE AN INDEPENDENT, VERIFIED, THERMAL-HYDRAULICS SYSTEMS LCDE IN PLACE.
i DESCED E_J5P_Afl_0LP ESLLLl$ 8 PRESENT IMPAC1 IS MINIMAL BECAUSE OF LACIC OF LMFBR LICENSING ACTIVITY. THE IMPACT OF THE VLkIIICAIION ACTIVITIES ASSOCIATED WITH FFTF NATURAL CIRCULATION TESTS MAY BE IMPORTANT DEPENDING ON TEST GUTCOME.
i fD_MMIN1}fs][CAE O AT THE PRESENT TIME IANUS AND DEMD ARE THE TWO CODES USLO FOR THE THERMAL HYDRAULIC ANALYSIS OF FFIF AND CkBR.
BOTH CODES NAVE BEEN DEVELOPED BY WESTINGHOUSE. DEVELOPMENT OF SSC-L IS A SPECIFIC CONFIRMATORY j
RESEARCH EFTORT WITH APPLICATION TO FFTF.
THE (P) AND (5) VERSIONS OF SSC HAVE FUTURE APPLICATION FOLLOWING THE DEVELGPMENT OF THE (1) VER510N.
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PROGDJ9 OFFICE CO*rE413 DM POTENVAGL uf!LIZATAON 04 U AtuE OF RESEARCH RESULT 3 IN THE REGgtATORY PROCESS PIL U 44 DLTE ISSUED: 01/04/79 PES DECISION UNIT: FUEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS RIL TITLE: RADIATION DOSE TO C0KSTRUCTION WORKFRS AT OPERATING NUCLEAR POWER PLANT SITES SPONSOPING OFFICE (S): HRR Pgg:
5-23 OCCUPATIONAL RESEARCH PROJECT MGR:
J. FOULKE EXPOSURE AND PROTECTION t
i Pf3_COE"EMTS: THE STUDY PROVIDES A DATA BASE WHICH ALLOWS A REALISTIC ASSESSMENT OF THE RADIOLOGICAL IMPACI ON CONSTRUCTICH WORKERS OF PRCPOSED MULTI-UNIT NUCLEAR POWER PLANTS.
MEASUREMENTS OF PERSONNEL EXPOSURE OF SEVERAL HUNLRED CONSTRUCTION WORKERS WERE CONDUCTED AT EACH OF THE FOUR SITES WHERE NEW FACILITIES WERE UNDER CONSTRUCTION NEXT TO OPERATING NUCLEAR POWER PLANTS.
FOR MOST WORKER GROUPS. THE AVERAGE DOSE EQUIVALENT PATES WERE MUCM LLSS THAN to MREM / MONTH GREAIER THAN OFF-SITE CONTROLS. CORPriATIONS BETWEEN CPER ATING PL ANT POWER LEVELS AND DOSIMETER READINGS WERE GENERALLY POOR INDICATING THAT VARI ATIONS IN OTHER RADIATION SOURCES HAD GREATER EFFECT ON THE MEASUREMENTS.
U5ER DISCUSSION POSITION COMMISSION ACRS FRESS OFFICE MEETING PAPER BRIEFING BRIEFIPS RELEASE RESULTS I
PQST Pll_!GTlvli1ES PEVIEH HELD COMPLETED HELD HELS ISSUED IMPLEMENTED Of fICE R ESPONSIBL L.........
NRR SCHEDULED COMPLETIDH DATE..
ACTUAL COMPLETION DATE.....
5 bER_CDM*fMTS, C..S.
H1NSDN -
AS PART OF ITS LICENSIhG REVIEW PROCESS, THE HRC MUST ASSESS DDiP13E_A11LIC AJi9N_ULPEGVI ATORY PPOR$$:
THE ENVIRONMENIAL IMPACI OF THE CONSTRUCTION AND OPERATION OF NUCLEAR POWER PLANTS. ASSESSMENT OF RADIATION a
DOSES TO CONSTRUCTION WORKERS AT OPERATING NUCLEAR POWER PLANT SITES IS PART OF THIS REVIEW. SINCE A FIRM DATA BASE FOR RADIATION DOSES TO CONSTRUCTION WORKERS WAS NOT AVAILABLE, ENVIRONMENTAL IMPACT STATEMENTS g
l COULD ONLY PF0 VIDE ROUGH ESTIMATES FOR THE DOSES TO CONSTRUC1 ION WORKERS. THE RESULTS OF THIS STUDf i"
PROVIDE RADIATION EXPOSURE DATA SUCH THAT A REASONABLE ENVIRONMENTAL IMPACT STATEMENT CAN BE MADE.
UNTIL THE PUBLICATION OF THIS REPORT, THERE HAS BEEN LITTLE DATA WHICH THE DESCPR f_]MP&C7_DE_'f}f d$:
STAFF COULD USL IN ASSESSING THE APPLICANT'S ESTIMATE OF CONSTRUCTION WORKER DOSES DURING NUCLEAR POWER PLANT CONSTRUCTION. THE RESULTS OF THIS STU9Y SHOW THAT MOST CONSTRUCTION WORKERS AT OPERATING NUCLEAR PCWER PLANTS RECEIVED LESS THAN to MILLIREEMS PER MONTH; AND NO WORKER'S ESTIMATED DOSE EXCEEDED S00 MILLIREMS PER YEAR.
THESE RESULTS INDICATE THAT CONSTRUCTION WORKERS AT OPERATING REACTOR SITES ARE NOT LIKELY TO RECEIVE SIGNIFICANT RADIATION DOSES. AND THEREFORE THAT INDIVIDUAL MONITORING WILL NOT BE REQUIRED.
HOWEVER, IT MAY SE APPROPRIATE 10 PLACE DOSIMETERS AT POINTS WHERE THE HIGHEST EXPOSURE RATES ARE TO Bt EXPECTED. TO ASSURE THAT UNUSUALLY HIGH EXPOSURE RATES DO NOT GO UNDETECTED.
~
PR96D AM GF FICE CEMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE PEGULAT0*Y PROCESS RIL Os 45 DATE ISSUED:
02/11/7*
RES DECISION UNIT: FUEL CYCLE SAFETv 8 ENVIRONMENTAL EFFICTS RIL TITLE: THE CONCEPT COMPUTER CODE ANL CAPITAL COSTS FOR BOILING WATER REACTOR PLANTS SPON50 RING OFFICE (5): NRR RES: 5-21 SOCIO-ECONOMIC RESEARCH PROJECT MGRE D. BARNA IMPACTS REi_f?tNESIS : THIS RIL TRANSMITS THE RESULT 5 OF COMPLETED RESEARCH UPDATING AND EXPANDING THE CONCEPT COMPUTER CODE FOR FORECASTING CAPITAL COSTS OF BOILING WATER REACTOR PLANTS.
IN 1971 THE ATCMIC ENERGY COMMISSION uJTHORIZED POWER PLANT INVESTMENT COST STUDIES. WHICH CULMINATED IN THE WASH-1230 REPORTS PUBLISHED IN 1972.
THEIR PURPOSE WA5 TO FACILITATE POLICY AND ECONOMIC DECISIONS ABOUT ELECTRIC GENERATION FACILITIES IN THE PUR81C AND PRIVATE SECTOR 5.
NATIONAL PRIORITIES ON ENERGY THE REGULATORY ENVIRONMENT AND THE COST OF i ABOR. E00!* MENT AND MATERIAL HAVE CHANGED SIGNIFICANTLY. THESE CHANGES DICTATED THE NECESSITY OF UPDATING THIS SERIES OF SiOSIES. AND EXPANDING THE SCOPE TO CONSIDER THE FUEL CYCLE AND THE TOTAL GENERATING COST.
AS A RESULT. A PROGRsM TO STUDY, REASSESS AND PRODUCE A NEW SET OF UPDATED REPORTS WAS AUTHORIZED AND UNDERTAKEN.
THE STUDIES IN THESE SERIES HAVE A UNIFORM SET OF ECONOMIC AND TEC:lNICAL CRITERIA AND A UNIFORM ACCOUNTING SYSILM.
THE INVESTMENT COST ESTIMATES IN THESE SERIES ARE DEVELOPED FOR REFERENCE PLANTS CONSTRUCTED AT A NYPOTHETICAL SITE.
THE ESTIMATED TOTAL EASE CONSTRUCTION COST FOR THE 1190 MWt BkR REFERENCE DESIGN IS 0582.748,510 CF S490/KW BASED ON JULY 1.
1976 PRICES.
THE TOTAL BASE CONSTRUCT!ON COST FOR THE BWR POWER PLANT (Itti MWE NET CUTPUT) REFERENCE T4 WASH-1230 WAS APPROXIMATELY $213,000.000 OR $201/KW, BASED UPON PRICES EFFECTIVE JANUARY 19?1 THUS. THE 1977 STUDY INDICATES APPROXIMATELY A 143 PERCENT INCREASE IN THE COST QF THE PLANT IN TERMS OF C/KW.
TPE YOTAL DIRECT CRAFT LABOR COST OF APPROXIMATELY S139.500.000 CORRESPONDS TO AN AVEPACE HOURtY RATE OF $12.29.
APPEDXIMATELY 11.350.000 CRAFT LABOR MANHOURS AVERAGE ABOUT 9.5 MANHOURS /KW.
THESE COMPARE TO AVERAGES OF $8.E4/H0i!R AND 6.3 MANHOURS /KW RESPLCTIVELY FOR THE EARLIER DESIGN REPORTED IN WASH-1230.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PDST.RIL ACTIVIT111 EEVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RL5PONSIBLF..........
NRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETIUM DATE.....
fiE U C OEN15 u D d NVIR. TEQH -
8 m
MSGtEALPllqARN _TQ_f tQyL ATORY PRO {E$$: THE PRINCIPAL APPLICATION OF THE INVESTMENT COST DATA IS TO UPDATE iHE CONCLPI C0KFUILR CODE.
IN TURN THE CONCEPT CODE IS USED TO ESTIMATE CAPITAL COST FOR DIFFERENT STZE PLANTS, DIFFERENT REGIONS OF THE COUNTY, DIFFERENT SCHEDULE LENGTH. DIFFERENT COMPLETION DATES.
DIFFERENT ESCALATION AND INTEREST RATES.
IN ADDITION TO UPDATING CONCEPT THE DATA IS USED AS A GENERAL REFERENCE FOR SUCif THINGS AS. COST OF COMPONENTS. QUANTITIES OF MATERIALS USED, TYPE AND QUANTITY OF LABOR. ETC.
N SHIFJP ACT OF_RESjn TS : DURING FY 1977 44 REQUESTS WERE MADE FOR CONCEPT CODE RUNS AND 32 REQUESTS WLkE MADE IN 1913.
MOSI REQUESTS INVOLVED RUNS FOR SEVERAL SIZE 5 0F COAL PLANTS AND ONE OR TW3 NUCLEAR UNITS FOR DIFFERENT ESCALATION AND INTEREST RATES.
THE USE OF THE CONCEPT CODE PERMITTED THE STAFF TO PERFCRM THESE ANat.YSES MORE EFFICIENTLY AND CUICKLY THAN IF THE CONCEPT CODE WERE NOT AVAILABLE.
COMMENTS /REMA*KS DDE IS FUNDING THE LATEST UPDATE.._- __-.
PROGG. AM OF FICE C0P-EN T S ON POTENTIAL UTILIZATION OR v4LUE OF RESEASCH RESULT 5 IN THE REGULA10Rf PROCESA
~
RIL 8: 46 DATE 155UEDr c2ft2fy9 RES DECISI0M UNITr SYSTEMS ENGINEERING PIL TITLE 2 EFFECTIVENESS OF CABLE TRAY C0ATING MATERIALS 8 BARRIERS IN RETARDING THE COMBUSTION OF CABLE TRAYS SUBJECTED TO EXPOSURE FIRES AND IN PREVENTING PROPAGATION SETWEEN CABLE TRAYS HORIZONTAL OPEN SPACE CONFIGURATION)
$POM50 PING OFFICE (5): SD, NRR RES: 1-23 ELECTRICAL %TANDARDS RESEARCH PROJECT MGE:
R. FEIT
& FIRE PROTECTION PE5 COMMENI$2 DATA 15 PRESENTED DN THE EFFECTIVENESS OF SIX FIRE RETARDANT C0ATING MATERIALS AND BARRIER DESIGN 5 IN HORIZCNTAL OPEN SPACE CONFIGURATIONS THAT ARE IN USE OR BEING CONSIDERED FOR NUCLEAR POWER PLANTS.
AN ACCEPT ABLE TEST METHODDLOGY WAS DEVELOPED BY WHICH PASSIVE FIRE PROTECTICM MEASURES CAN BE EV ALU ATED.
THE TESTS DEVELOPED CAN BE PERF0PMED BY SUPPLIERS AND PLANT OPERATORS TO JUSTIFY ALTERNATIVE FIRE RETARDANT COATINGS AND BARRIERS NOT LISTED BY NRC OR TO DEMONSTRATE THE EFFECTIVENESS OF THOSE MEASURES TESTED BY THE NRC IN SITUATIONS WHERE THE DESIGN BASIS FIRE 15 SIGNIFICANTLY DIFFERENT THAN THE TEST CASE FIRES.
USER DISCUSSION POSITION COMMISSION ACR5 PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOST P,IL ACTIVITIES PEVIEW HELD COMPLETED HELD HE1D 115QER_
IMPLEMENTED OFFICE RESPON5IBLE......... SD/NRR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
M55ENI) - THE EhGINEERING METHODOLOGY STANDARDS BRANCH WILL USE THE RESULTS TO SUPPORT THE STAFF FOSITION IN REGULATORY GUIDE 1.120 AND IN THE DEVELOPMENT OF A REGULATORY GUIDE ON FIRE STOPS.
HPPETENTS. NOVEMEER 1980. R.
FERM399_zE. SYLVE1TLE:
EtikE1!E_*ffilCATION 10 RLG91AIf_RY P P Q1L)3 :
IHIS LETTER COMPARES THE EFFECTIVENESS OF C0ATING TRAY COVERS, AND CABLE INSULIION MAIERIALS IN PREVENIING FIRE PROPACATION IN HORIl0NTAL CABLE TRAYS.
SUCH CONFIGURATIONS ARE PROPOSED BY LICENSEES TO RETARD FIPE PROP AG ATION.
THE RIL RECOMMENDS QA REQUIREMENTS FOR MINIMUM THREE HOURS OF COATING AND FOR SOLID COVERING OF CABLE BUNDLE, I.E.,
NO AIR PATHS THRDUGH THE BUNDLE.
DESLPI)E_JUPACT OF PESULTS2 MINIMAL. Ali RESULTS WITHIN THE PERFCRMANCE GOALS ASSUMED IN THE FIRE HAZARDS ANALYSIS.
QUALIIf ASSURANCE SHOULD BE ACCORDING TO THE LICENSEES QA PROGRAM.
i'
[EMUINTS/PE58Pf23: THE TEST RESULT 5 HAVE LIMITED APPLICATION IN EVALUATIONS BECAUSE NO VERTICAL CONFIGURATIONS WERE 1ESILD, CABLE 5 WERE DEENERGIZED, RANDOM FILLED PATTERN WAS USED, TEST CONFIGURATION WAS IN RELATIVELY OPEN AREA, AND EFFECT OF FIRE SIZE 5 DN FIRE PROPAGATION IS NOT KNOWN.
THIS PROGRAM DID NOT HAVE SPECIFIC GOALS STATED IN TERMS OF LICENSTNG CONCERNS.
, 1,
~~
PGCGSAM OFFICE COM"ENTS ON POTENTIAL UfILIZnTION OW VALUE OF RESEAPCN PESULT5 IN THE REGUL ATORY PROCESS pit s 47 DATE ISSUED 03/19/79 PES DECISION UNIT: FUEL CYCLE *AFETY & ENVIRONMENTAL EFFECTS pit TITLE: IHREM II: A COMPUTER IMPLEMENTATION OF RECENI MODELS FOR ESTIMATING THE DOSE EQUIVALENT TO ORGANS OF MAN FROM AN INHALED OR INGESTED RADIONGCLIDE f
SPON50 PING CFFICE(5): NRR Pggr 5-24 RADI0 BIOLOGY RESEARCH PROJECT MGR2 J. FOULKE
& DOSIMETRY Pli_[05MLNT}: TEE INREM II CODE AS CONT AINED IN NUREG/CR-0114 ALSO TRAN5MITTED IS VOLUME 1 05 A TABULATION l
OF INILRNAL RADIATION DOSE CONVERSION FACTORS OBTAINED USING THE INREM II CODE.
THIS TABULATION IS GIVEN IM "ESTIMATES OF INTERNAL DOSE EQUIVALENT TO 22 TARGET ORGANS FOR RADIONUCLIDES OCCURRING IN ROUTINE RELEASES FROM NUCLEAR FUEL-CYCLE FACILITIES." VOL. 1 N U R E G/C R- 015 0.
THE CODE COMPUTES REFERENCE ADULT HUMAN DOSE l
EQUIVALENTS FROM USER-5UPPLIED DOSIMETRIC AND METABOLIC INFORMATION.
IN PRINCIPLE. INREM II APPROACH 15 SIMILAR TO THE ONE USED IN WASH-1400.
ORGAN DOSE EQUIVALENT 15 COMPUTED AS THE SUM OF CONTRIBUTICNS FROM EACH SOURCE r
ORGAN WNERE RADI0 ACTIVITY IS ASSUMED PRESENT; CROSS-IRRADIATION EFFECT5 CAN BE ASSESSED. DOSE CONVERSION FACTOR 5 IN NUREC/CR-0150 AhD UPCOMING VOL. 2 ARE ILLUSTR ATIVE OF INREM II CODE USE ONLY.
THEY ARE NOT ENDOR5ED 60R ADOPTION SINCE SOME OF THE METABOLIC MODEL5 ARE SUBJECT TO CRITICISM WITH FURTHER DEVELOPMENT OF ACCEPIABLE METABCLIC MODELS BY ORNL, THE CODE WILL FROVIDE STATE-OF-THE-ART METHODOLOGY FOR DOSE CALCULATIONS.
I l
J5ER DISCUSSION POSITION CCMMISSION ACR$
PRESS DeTICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS I
E931_fil_ AX1]VJT111 R ELIlH HELD COM*LETED HELD NELD ISSUED IMPLEMENTED I
0FFICE RESPON51 ele........,
NRR 1
SCHEDULED COMPLETION DATE..
ACTUAL COMPLETICN DATE.....
1 P P._.00MN_L _9 N.LJ W E.13 - 1971 M _l U f_8"ALYSIS3 O f10 Pl.D.E A P P L ] { A T1QN]QR0RAJ Q P Y P R OC E S)
PRESENT ANALYSES OF RADIATION EXPO 5URE SUCH AS PREPARED FOR WASH-14CO.
GLSMO AND IN kEAC10R LICENSING ACTIONS HAVE NOT BEEN UNIFORM IN DOSE ESTIMATION METHODOLOGY. THE VARIOUS APPROACHES USED FOR THE ABOVE STUDIES AND ACTION 5 REFLECT VARID'J5 IN;EkPRETATIONS OF THE STATE-OF-THE-ART IN DOSE ESTIMATION METHODOLOGY, THE NEEDS OF PARTICULAR STUDIES AND AVAILABILITY OF DOCUMENTED MODELS. NRC DOES NOT HAVE A BROADLY APPLICABLE, DOCUMENTED, CRITICALLY REVIEWED, STATE-OF-THE-ART DDSE ESTIMATIDA METHODOLOGY AVAILABLE TO VARIOUS OFFICES FOR THEIR ASSESSMENT NEEDS.
THE CLOSEST 15 REGULATORY GUIDE 1.109. REVISION 1,
1977, WHICH REPRESENTS A CURRENTLY APPROPRIATE APPLICATION OF A REASONABLY UP-TO-DATE DOSE ASSESSMENT METHODOLOGY AS APPLICABLE TO NUCLEAR j
POWER PLANT [FFLUENTS.
THE OFFICE OF NUCLEAR REGULATORY RESEARCH RECOGNIZED TME NEEDS IN THIS AREA, AND HAS CEEN FUNDING RESEARCH AT THE HEALTH AND SAFETY RESEARCH DIVISION OF THE DAK RIDGE NATIONAL LABORATORY (ORNt).
THE I
RESEARCH DESCRIBED IN RIL 847 PARTIALLY FULFILLS THE NEED OF A BROADLY APPLICABLE. DOCUMENTED AND CLOSER TO THE STATE-OF-THE-ARY MODEL FOR CALCULAT104 0F DOSES TO VARIOUS ORGANS OF A BODY FROM INTERNALLY DEPOSITED RADIONUCLIDES.
HOWEVER. ITS APPLICATION TO REGULATORY PROCESSES HAS TO WAIT UNTIL SOME ADDITIONAL SUPPORTING RESEARCH IS COMPLETD 4
(SEE COMMENTS /REMARES) MOST OF WHICH ARE ON-GOING AT ORNL FUNDED BY RES AND SOME BENCH-MARK COMPARISONS ARE MADE.
DU (P.JI L p*PAfl J F P QtJLI}: THE RESULTS HAVE CONFIRMED OUR UNDERSTANDING OF INTERNAL DOSIMETRY MODELING, BUT HAVE NO DIRECT IMPACT ON LICENSING UNTIL AFTER SOME ADDITIONAL STUDY IS PERFORMED.
I L _ _ _.
CD55f M T5e P E"A 0%5 2 ICOP HAS AN ONC0!CG C3 TR ACT WITH HEALTH AMD SAFETY CE5EAICH LIVISION OF ORML TO DEVELOP CEV 1ED NUCLEAR DATA. 5-FACT 025, METABOLIC PA2AMETERS DO5IMETRIC M0CEL5, ETC. F02 USE IQ IC"P'S FORTHCCMING PEBLICATION 830.
IT 15 NOW KNOWN THAT THE ORNL InTECNAL D05IMET2Y MODEL FOR ICRP IS SOMEWHAT DIFFERECT FROM THAT F02 NRC (INREM-II)
AL1 HOUGH BOTH MODEL5 ARE BASED ON ICRP'S TASK GROUP LUNG MODEL (1966, 1972) AND EVE'S GI-TRACT MODEL (1966).
THE ORNL MODEL FOR ICRP IS MORE COMPLETE IN THAT If NAS THE BODY FLUIDS (BLOOD) REPRESENTED IN THE MODEL A5 AN ADDITIONAL COMPARTMENT WHICH 15 LACKING IN INREM-II.
INCLU5 ION OF A BLOOD COMPARTMENT IS APPROPRIATE CONCEPTUALLY A5 WELL A5 FROM VIEW POINT OF REALISTIC MODELING; IT IS LIKELY TO REDUCE THE ORGAN DDSE CONVERSION FACIOR5 PARTICULARLY FOR THE SHORTER LIVED RADIONUCLIDES DUE TO DECAY DURING PARTIAL HOLD-UP IN THE BLOOD.
BENCHMARKING INREM-II RESULT 5 WITH THE ICRP'S WILL BE PRUDENT BEFORE CUR USING THEM.
THE BEST VALUES OF METABOLIC PARAMETERS REQUIRED AS INPUTS TO RUN INREM-II ARE NOT YET AVAILABLE. THE NOMINAL VAtUES OF THESE PARAMETERS USED BY THE AUTH0R5 0F TWO VOLUMES (VOL. 2,IS IN PREPARATION) DF DOSE CONVERSION FACTOR TABULATIONS APE FOR THE PURPOSES OF ILLUSTRATING THE USE OF INREM-II, AND THE AUTHORS DO CAUTION AGAINST UNCRITICAL USE OF THESE TABIES IN RADIOLOGICAL APPLICATIONS BECAUSE THEY CONSIDEP. THESE RESULTS A5 PRELIMINARY. RES SHOULD BE REQUESTED TO INCLUDE THE TASK OF GENERATING AND THEN PERIODICALLY UPDATING A DATA LIBRARY OF BEST VALUES OF METABULIC PARAMETERS FOR USE IN INREM-II IN THEIR CONTINUING RE5ti-RCH PROGRAMS WITH CRNL.
A SET OF VALUES OF METABOLIC PARAMETERS IS BEING USED IN ICRP WORK AT ORNL.
THE ICF4 VALUES *RE NOT YET AVAILABLE TO USE BUT WHEN AVAILABLE. IT WILL BE PRUDENT TO COMPARE THESE VALUES WITH THE ti:fLIMINARY VALUE5 USED BY ORNL FOR NRC, AND MAKE APPROPRIATE REVISION IF NECESSARY.
THE NUCLEAR DATA DEVELOPED AND UTILIZED BY ORNL FOR MRC ARE NOT NECESSARILY THE SAME AS BEING DEVELOPED FOR ICRP.
THOUGH SUBSTANTIAL DIFFERENCES ARE NOT ANTICIPATED IN THIS AREA, IT WILL BE PRUDENT TO COMPARE THESE DATA AL50.
LASTLY, THE 5-FACTORS AND METABOLIC FARAMETERS FOR INDIVIDUALS OF AGE GROUPS JIFFERENT FROM ADULTS ARE STILL BEING DEVELOPED AT CRNL FOR NRC.
UNTIL THIS WORK IS COMPLETED IT IS NOT POSSIBLE TO GENERATE DOSE CONVERSION FACTORS FOR AN INDIVIDUAL OTHER THAN Ah ADULT.
A5 A MATTER OF INFORMATION, 05D HA5 A CGNTRACT WITH DRNL TO PROVIDE THE NRC WITH AS MUCH INFORMATIDM REGARDING THEIR RESEARCH FOR ICRP A5 WOULD BE PERMITTED BEFORE AND AFTER THE ICRP PUBLICATION 830. t l
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PROCR AM Of fICE CC'""EMIS ON POTEM l?L 'JilLIZAl10M OR VALUE OF RESEARCH RE$UL15 IN THE PEGULATORY PROCESS i
PIL 82 48 DATE 155UED2 C4/03/79 RES DECISION UNIT: SEI5MIC, ENGINEERING 8 SITE 5AFETY SAFETY RL TITLE A TECTONIC OVERVIEW OF THE CENTRAL MIDCONTINENT l
iP1NSCRING OFFICE (5)r SD, NRR F29: 3-2 GEOLOGY 8 SEISMIC RESEARCH PPOJECT MGR M. STEUER CHARACTERISTICS 4
PCL10EfMD:
"A TECTONIC CVERVIEW OF THE CENTRAL MIDCONTINENT", NUREG-0352. IS THE MOST UP-TO-DATE SYNTHESIS OF GEOLOGIC i
RNOWLLDGE OF THE EARTHS CPUST IN THE STUDY AREA.
IT CONT AINS THE MOST COMPLETE BIBLIOGRAPHY OF THE GE0 DYNAMICS CF THE AREA AND WILL AID IN NUCLEAR POWER PLANT LICENSING.
IT 15 RECOMMENDED THAT THE INFORMATION AND HYPOTHESES 4
BE USED AS INPUT 10 THE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA.
1 I
USEP.
DISCUSSION POSITION COMMISSION ACR5 PRESS 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS j
E957 PJLA_CLIVJHU EHEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED
]
0FFICL EE5PONSIBLE......... SD/hRR SCHEDULED CCMPLETION DATE..
)
ACTUAL COMPLETION D7TE.....
I 1p_(P T jf5 - NO COMMENT AT THIS TIME DUE TO PRELIMINARY NATURE OF THE WORK.
WE PLAN AN EXTENSIVE REVIEW I
bHLN IHE PROJECT 15 FURTHER ALONG.
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P R OGQ AM M F AC E (A5"D TS ON P0l[Nl}4L HIJ1I2ATIAN AR vALME OF QESEARCH RESULT 5 IN THE REGULATORY PROCESS I
PIL 82 49 DATE ISSUEDs 04/04/79 PES DECI5f0N UNIT: FUEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS PIL fITLE: IN VITRO DIS 5OLUTION OF URANIUM PRODUCT SAMPLES FROM FOUR URANIUM MILLS f
SPON50RINS OFFICE (5)r SD E%: 5-23 OCCUPATIONAL RESEARCH PROJECT MGRr J.
FOULKE EXPOSURE & PROTECTION k
PlLif*f"fN_g THE MEASUREMENT OF THE SOLUBILITY OF VARICUS FORMS OF YELLOWCAKE IN VITRD UTILIZED TWO SOLVENT SYSTEMS: A SIMULANT OF AN ULTRAFILTRATE OF BLOOD SERUM AND 0.1M HCL.
THE RESULTS OF THIS STUDY WILL BE USED TO DETERMINE WHETHER YELLOWCAFE SHOULD BE CONSIDERED A SOLUBLE COMPOUND WITH RESPECT TO 10 CFR 20.
THE DATA OBTAINED USING THE SERUM SIMUL ANT SHCW THAT 25 TD 64 PERCENT OF ALL SAMPLES TESTED DISSOLVED WITH HALF-TIMES LESS THAN 16 HOUR $.
IDEALLY, SOLUBILITY CLA55IFICATIONS SHOULD BE BASED ON THE DISSOLUTION HALF-TIMES'0F THE PARTICULAR FRCDUCT UNOER CONSIDERATION. THE DATA INDICATE THAT YELLOWCARE SHOULD BE TREATED A5 A CLA55 D 4
COMPOUND FOR THE PURPOSE OF EXPOSURE CONTROL.
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USER DISCUSSION POSITION COMMISSION ACRS PRESS s
OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS j
f 95_J P,lL _ A QILVLTlfi P QL[y HELD CCMPLETED HELD HELD ISSUED IMPLEMENTED PFFICE RESPONSIBLE......... $D i
SCHEDULED COMPLET!ON SATE..
l ACT8JAL COMPLETION DATE....
1 i
i L CQ T fMT5, L G M JPl - THE RESULTS OF THIS RESEARCH WILL BE USED IN FINALIZING REGULATORY GUIDE 8.22.
"$10A55AY AT URANIUM MILL 5."
N3 FURTHER RESE4RCH CN THIS TOPIC IS PRESENTLY CONTEMPLATED.
}
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- ~ -
P&FG8'AM Orf1CE C0"MEN 15 ON PotENTI*t uf1LIZaTION OR VALUE OF RESEARCH RESULTS IN THE REGUL ATORY PROCESS l
I FIL es 50 I A TE 155L'ED:
04/06/79 RES DECISION UNIT FUEL CYCLE SAFETY 8 ENVIRONMENTAL EFFECTS l
RIL TITLE: CRITICALITY SAFETY GUIDANCE I
SPCN50 PING UTFICEt5)s NM55 (77-9)
EP.,Q :
5-7 CRITICALITY 5AFETY RESEARCH PPOJECT MGR D. 50LSERG STUDIES l
P RCO*SD : SINCE 1957 THE NUCLEAR SAFETY GUIDE HAS PROVIDED USERS OF t.,510NABLE MATERIAL WI1H SIMPLE MEIHODS TO.155URE SYSTEM SutCRITICALITY.
RESEARCH INFORMATION LETTER 550 TRAN5MITS NUREG/CR-0095.
- NUCLEAR l
SAFETY GUIDE. TID-7016. REVISION 2."
THIS REVISION UTILIZE 5 IMPROVED CRITICALITY DATA AND CCMPUTATIONAL T
J TECHNIQUE 5 NOT AVAILABLE IN THE PREVICUS GUIDE ISSUED IN 1961 REVISION 2 SHOULD BE REGARDED A5 A i
SUPPLEMENT 10 PREVIOUS GUIDE 5 SINCE IT DOES NOT HAVE INTENTIONAL CONSERVATISM 5 INCLUDED AS IN PREVIOUS I
GU I L E's.
IN USING REVISION 2 0F THE CUIDE. THE USER MUST IMPOSE APPROPRIATE SAFETY FACTORS FOR HIS APPLICATION AND THE NRC STAFF MUST DETERMINE THAT THE SAFETY MARGINS PROPOSED BY THE USER OF REVISION 2 ARE ADEQUATE.
USER DISCUSSION POSITION COMMI55IDH ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST PIL AQ1]VJJ1[1 PEVIEW HELD C0f"PtETED HELD HELD ISSUED IMPLEMENTED 1
De RITE u t SPoN5I e t t.........
2 SCHEDULfD CC"PLETION DATE..
I ACTUAL COMPLETICN DATE.....
(
l tt31EMM 5 L%.P P tH _L_IE I Yt h 1 _51R D5052 THE RLVI5LD NUCLEAR SA GUIDL IS QUOTED TO THE LICENSEES WHENEVER THEIR AMENDMENT OR RENEWAL OF APPLICATIONS REIER TO THE 1961 VER$10M OF THE GUIDE AND FAIL TO RECOGNIZE NEW DEVELOPMENTS IN NEUTRON INTERACTION-ACTION.
]
CERTAIN CF WHICH ARE SUMMARIZED IN L
THE UPDATED GUILE.
THESE NEW DEVELOPMENTS. SOME OF MHICH HAVE RESULTED FROM J
NRC-SPONSORED RESEARCH, HAVE RENDERED QUESTIONABLE CERTAIN PAST PROCEDURES IN THE ANALYSES OF NEUTRON INTERACTICH.
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1 PROSRAM OFFICE ([M1[NTS (N PDfENTAat pilLXZ4VIGN OR UALUE OF PESEARCH PE5MLV5 IN THE REGULATDQY PROCESS P.It es 51 DATE 155t>;pr 04/12/79 RES DECISION UNIT: FUEL CYCLE SAFETY 8 ENVIRONMENTAL EFFECTS RIL TITLE: THE CONCEPT COMPUTER CODE AND CAPITAL COSTS FOR PRE 55UZIZED WATER REACTOR PL ANTS SPON50*ING OFFICEf5): NRR gggt 5-21 50CID-ECONOMIC RESEARCH PROJECT FGR*
D. BARNA IMPACTS Ef1_CCEMENT3 THIS RIL TRAN5MITS THE RESULT 5 OF COMPLETED RESEARCH UPDATING AND EXPANDING THE CONCEPT CcMPUTER CODE FOR FORECASTING CAPIT AL COSTS OF PRE 55bRIZED WATER REACTOR PL ANTS.
IN 1971 THE ATOMIC ENERGY COMMISSION AUTHORIZED POWER PLANT INVESTMENT COST STUDIES, WHICH CULMINATED IN THE WASH-1230 REPORTS PUBLI5hED IN 1972.
THEIR PURPOSE WAS TO FACILITATE POLICY AND ECONOMIC DECISIONS ABOUT ELECTRIC GENERATION FACILITIES IN THE PUBLIC AND PRIVATE SECTORS. NATIONAL PRIORITIES ON ENERGY, THE REGULATORY ENVIRONMENT AND THE COST OF LABOR, EQUIPMENT AND MATERIAL HAVE CHANCED SIGNIFICANTLY. THESE CHANGES l
DICTATED THE NECESSITY OF UPDATING THIS SERIES OF STUDIES, AND EXPANDING THE SCOPE TO CONSIDER THE 1
I FUEL CYCLE AND THE TOTAL GENERATING COST.
A5 A RESULT, A PROGRAM TO STUDY, REASSESS AND PRODUCE A NEW SET OF UPDATED REPORT 5 WAS AUTHORIZED AND UNDERTAKEN.
j THE STUDIES IN THESE SERIES MAVE A UNIFORM SET OF ECONOMIC AND TECHNICAL CRITERIA AND A UNIFORM ACCOUNTING SYSTEM.
THE INVESTMENT COST ESTIMATES IN THESE SERIES ARE DEVELOPED FOR REFERENCE PLANTS CONSTRUCTED AT A HYPOTHETICAo SITE.
THE ESTIMATED TOTAL BASE CONSTRUCTION COST FOR THE 1139 MWE PWM RPFERENCE DESIGN I5 t
i S568.831.011 GR $449/KW BASED ON JULY 1,
1976 PRICES.
THE TOTAL BASE CONSTRUCTION COST FOR THE PWR POWER l
PL ANT (1031 MWE NET DUTPUT) REFERENCE IN W45H-1230 WAS AP?ROXIMATER.Y $211,000,003 CE $205/KW, BASED UPON PRICES EFFECTIVE JANUARY 1971 THUS. THE 1977 STUDY INDICATES APPROXIMATELY A 143 PERCENT INCREASE IN THE COST OF THE PLANT IN TEFMS OF $/KW.
THE TOTAL DIRECT CRAFT LABOR COST OF APPROXIMATELY $133,109,000 CORRESPONDS i'
TO AN AVERAGE HDURLY RATE OF $12.30.
APPROXIMATELY 10.820,000 CRAFT LABOR MANHOURS AVERAGE ABOUT 9.5 MANHOURS /KW.
l THESE C0MPARE TO AVERAGES OF SS.86/ HOUR AND 6.0 MANHOURS /KW RESPECTIVELY FOR THE EARLIER DESIGN REPORT IN WASH-1230.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EDIT PIL_4STiv1TLL1 EEEJ1W HELD COMPLETED BELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... ARR SCHEDULLD COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
HE?__$W %N h _AL _INVP, TECH I
pt5CR131_AttLIS AIJ ON 19_?fGy1ATORY PP3rE13: THE PRINCIPAL APPLICATION OF THE INVESTMENT COST DATA 15 TO a
UPDATE THE C0hCLPI C6MPUIER. CODE.
IN TURN THE CONCEPT CODE 15 USED TO ESTIMATE CAPIT AL COST FOR DIFFERENT SIZE PLANTS, DIFFERENT REGIONS OF THE COUNTY, DIFFERENT SCHEDULE LENGTN, DIFFERENT COMPLETION DATES. DIFFERENT ESCALATION AND INTEREST RATES.
IN ADDITION TO UPDATING CONCEPT THE DATA 15 USED AS A GENLRAL REFERENCE FOR SUCH THINGS A5, CO5T OF CCMPONENTS, QUANTITIES OF MATERIALS USED, TYPE AND QUANTITY OF LABOR. ETC.
t REitEIBf_lMPACT Lr et$y(T3s DURING FY 1977 44 REQUEST 5 WERE MADE FOR CONCEPT CODE RUN5 AND 12 REQUESTS WERE i
MADL IN 1978.
M35T'FEGUESTS INVOLVED RUNS FOR SEVERAL SIZE 5 OF CDAL PLANTS AND ONE OR TWD NUCLEAR l
UNITS FOR DIFFERENT ESCALATION AND INTEREST RATES.
THE USE OF THE CONCEPT CODE PERMITTED THE STAFT 10 PERFORM THESE ANALYSES MORE EFFICIENTLY AND QUICELY THAN IF THE CONCEPT CODE WERE NOT AVAILABLE.
t COMPENT5/PEMAPES: DOE 15 FUNDING THE LATEST UPDATE.,
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PP.DCPAM OFFICE CD-9tN!5 ON POTENTIAL UfftIZATION OR VALUE OF RESEARCH RESULTS IN THE REGUL ATORY PROCESS PIL e: 52 DATE IS5UED: 04/23/79 LES DECISION UNITr SEISMIC. ENGINEERING 8 SITE SAFETY PIL 'ITLEs EARTH 4UAKE INTENSITY SCALE SPDM50 RING OffiCL($l2 SD, NRR
!.1G s 3-2 CEDLOCIC A SEI5iTIC RESEARCH PROJECT FGR*
CHARACTERISTICS l
EE3_C0""[NT35
'HIS ME*0RANDU" TRAN5MITS THE RESULTS OF COMPLETED RESEARCH ON THE SYSTEMATIC EVALUATION OF j
IHE MACROLEISMIC DATA FILE COMPILED FOR DvER HALF A CENTUnY BY THE U.S. COAST AND GEODETIC SURVEY AND ITS j
SUCCESSOR ORGANIZATIONS. A NEW SEISMIC INTEN5!TY SCALE WA5 F3RMED BY REVISION OF THE MODIFIED MERCALLI INTENSITY SCALE OF 1931 THE RESULTS HAdE BEEN PUBLISHED AS NOAA TECHNICAL ME.10RANDUM. EDS NG W O -4 1
"REEVALUATION OF THE M00!FIED MERCALLI INTENSITY SCALE FOR EARTHQUAKES, USING DISTANCE AS A DETERMINANT."
l THE PURPOSE OF THE STUDY WA5 TO RELATE EARTHOUAKE INTENSITY TD THE ENERGY RELEASED AND ITS ATTENUATION WITH DIST ANCE TO PROVIDE A UNIFORM 5 CAL E REL ATED MORE LOGICAL LY TO THE PHYSICAL PaRAMET;-T3 'F ACCELER*: ION, /EL0 CITY, DISPiACEMENT AND SPECTRAL CONTE.4T.
IT IS SUGGESTED THAT NRC MAKF *. FORMAL RECOMMENDATION T0 sHE U.S.G.S.
THAT A NEW INTENSITY SCALE BASED ON THIS OR A 5 MILAR AkALYSIS BE PRPMULGATED AS AN OFFICIAL STANDARD.
THE MrW SCALE SHCULD PERMIT A TRANSITION FROM THE MM INTENSITY SCALE AND M4KE FUs.L USE OF THE EXTENS.VT DATA BASE ALREADr AVAILABLE.
4 USER DI3CU55 ION POSITION CDMMISSION ACR5 PRESS 0FFICE MEETING PAPER BRIEFING BRIFFING RELEASE RESULTS i
I E05LPlL _ACT LVIIJIS ERIEW NELD GMPL E T ED HELD HELD ISSUED If*PLEMENTED l
OFFICE RESPONSIBLE......... SD/NRR SCHEDU.ED COMPLETION PATE..
I ACTUAL COMPLETION DATE.....
i
$JLCfMMENTS: NO PESPONSE RECEIVED.
tL*P_C0 % NI L _E0VEPPE U L 591 %
R.
JACK 503:
D L M e i ri A ttLI CAT MNAP LWLA10 *LP).0 C L15 s NONE AT THIS TIME; TCO PREhATURF.
- 'TSCRIPL_lPP,ACT OF PL*s 115s NONE Ar IHIS TIME.
u 00*1"t hJ3 d1 T d :
INIS INFORMATION 15 IMPORTANT FOR THE EVALUATION OF EARTHQUAKE IMIENSITIES AND SHOULD BE 1
tokWAkOED 10 1HE USG5 FOR CONSIDERATION IN ANY 2EEVALUATION OF THE INTENSITY SCALES.
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~ PPO6MAM Off1CE CCM9LNIS CM POTENilAL OI!LIZAT10N CR VALUE OF RESEARCH MEiULIS iN THE REGVLAf0RY PPOCE55 RIL tz $4 DATE ISSUED: 85/15/19 FE5 DECI11t.7_UML11 RISK ASSESSMENT PIL TITLE 8 THE SET ECUATIDM TFANSF05MATION SYSTEM
$ rpm 50 RING (FFICE($18 NRR/NM55 Rig NOME EESEARCH PROJECT MGR:
W. VE5ELEY C05FfMI): A5 A RESUL7 OF THE WORK PERFORMED IN THE SETS COMPUTER CODE PROJECT. AN IMPROVED VERSION g[*THE iLIS CODE HA5 BEEN DOCUMENTEP AND MADE AVAILABLE FOR USE BY NRC CONTRACTOR 5 AND PERSONNEL FOR cf PROJECTS REQUIRING AN EFFICIthT TOOL FOR THE AN ALYSIS OF COMPL EX SYST EMS.
THE MAJOR NESULT5 0F THE SETS PROGRAM HAVE BEEd:
(1) DEVEL0rMENT cF AN AUTOMATIC TREE DECOMPOSITICM ALCORITHM WHICH IS CURRENTLY BEING INC0wr0 RATED INTO THE STANDARD VEPSION OF SET 5s (2) DEVELO* MINT, IN FRELIMINARY FORM, OF BASIC MINIMAL CUT SET QUANYIFICATION PROCEDURES:
(3) DEVELOPMENT CF A STANDARDIZED VERSION OF TEE SETS CODE FOR THE CDC 6600 COMPUTER. AND INSTAttATION OF THIS VERSION OF THE CODE AT THE BR00rHAVEN NATIONAL LABORATORY COMPUTER CENTER FOR USE BY NRC PER50FNEL; AhD (4) PREPAFATION OF A SET 5 USER 5* MANUAL ORIENTED SPECIFICALLY FOR THE FAULT TREE ANALYST.
USER DISCUSSION POSITION COMMISSION Acts PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULT 5 Plt _ACIlY11111 EEVIEW HELD
[CMPLETED HELD HELD ISSUED IMPLEMENTED E91T_L RESP 0h51BLL.........
OFFIC NRR/NM55 SCHEDULE 3 COMPLETION DATE..
ACTUAL COMPLETION DATE.....
ggsLgpTENTS pH LY 27, 1979, 5..NANAU Q:
[ESCR jti_AJP _ RATION TO *LGyLaTORY FRSCESS: THE IMPROVED VERSION OF THE SET EQUATION TRANSFORMATION SYSTEM (5ETS) 15 A.OM/UILR CODE ICR EVALUA(ING L ARGE FAULT TREES.
THERE ARE CURRENTLY TWO ACTIVITIES WITHIN NRR THAT USE FAULT TREES A5 AN ANALYSIS METHO3.
TMESE TWO ACTIVITIES ARE CENERAL TASK NO. A-17 AND THE VIT AL AREA ANALYSIS. THE SETS CODE 'PROVIDES THE NRC ST AFF WITH THE CAPABILITY TO INDEPENDENTLY AUDIT THE RESULT 5 OF ANALYSES THAT ARE BASED ON FAUL': 1REES. ALTHOUGH IM THE CASE OF TASK NQ A-57. THE CONTRACTOR WHO DEVELOP 5 THE FAULT TREE ALSO *ISES SETS CODE TO A'(AtV2E THE FAULT TREE.
THE AVAIL ABILITY OF THE SETS CODE AL50 PROVIDES A 700L FOR FUTURE USE Bf THE NRC STAFF IN APPL)?NG FAULT TREE MLTHODS 70 tHE LICENSING PRCCESS.
T4f %ETS CODE 15 AN IMPORTANT FEATURE IN THE METHODOLOGY (COMPUTER CODE) USED TO IDENTIFY TYPE I VITAL AREAS (I.E.,
"... 1H055 AREAS WHEREIN SUCCESSFUL SABOTAGE CAN BE ACCOMPLISHED BY COMPROMISING OR DESTROYING TH VIT AL SYSTEMS OR "CMf*0NENTS LOCATED WITHIN AN AREA).
. TE I VITAL AREAS ARE THE M051 SENSITIVE SECURITY AREAS IN THE PL ANT AND REQUIRE THE HIGHEST LEVELS OF PHYSICAL ts0TECTIfm.
IDENTIFYING THESE AREEAS IS AN IFFORTAN! CONSIDERATION IN THE STAFF'% EVALUATION OF PHYSICAL SECbRITY t
AT A NUCLEAR PCWER PLANT.
THIS ANALYSI5 AND EWALUATION IS CURRENTLY SEING USED FOR ALL OPERATING PLANT REVIEWS AND MILL ALSO BE AM 01 GGT"5 REQUIREMENT IN OPER ATING LICENSE TEVIEWS.
p 4FCBE IPPACT OF PESULTS: THE SETS CODE HAS NO DIRECT IMPACT ONN THE LICENSING PROCESS BUT IN THE FUTUQE IT WILL 6ERMIY VHE hRC STAFF 10 ANALYZE LARGE SYSTEMS THAT MIGHT OTHERWISE HAVE TO BE ANALYZLD BY LESS EFFICIENT OR LESS EFFECTIVE METHGDS.
THE RESULTS OF THE MORE EFFECTIVE AND EFFICIENCY METHODS COULD LEAD TO EITHTR A RELAXATION OF EXISTING R(QUIREMENTS OR CGUID LEAD TO NEW REQUIREMENT! OR A TIGHTENING OF AN EXI!. TING REQUIREMENT.
ECMMEN15/ Rett n#K S
- THE SETS CODE IS ONE OF SEVERAL COMPUTER CODES THAT CAN BE USED TO EVALUATE FAULT TREES.
1NCLUDING EVALUATION FOR COMMON CAUSE FAILURES. NO ATTEMPT HAS BEEN MADE HERE TO 0FTERMINE THE RELATIVE EFFECTIVENESS OR EASE OF USING THESE VARIOUS CODES.
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_ P9000AM OFFICE CCMMENTS ON POTENTIAt U*;.2?ATION OR VALUE OF PESEARCH RESULTS '. - THE REGULATORY PROCESS EJL_a: SS DATE ISSUED: 05/29/79 RES DECISION UNIT: FUEL CYCLE SAFETY 4 ENVIRONMENTAL EFFECTS PIL TITLE: THE CCNCEPi COMPUTER CODE AND CAPITAL COST FOR HIGH AND LOW SULFU2 COAL PLANTS - 1200 MWE SPONSORING OFFICEfS): HRR ggg: S-25 SOCIO-ECONOMIC RESEARCH PROJECT MGR:
D. BARNA IMPACTS PES COMMENTS: THIS MEMORANDUM TRANSMI'.S THE RESULTS OF COMPLETED RESEA2CH UPDATING AND EXPANDING THE CONCEPT COMPUTER CCCE FOR FORECASTING CAPITAL COSTS OF HIGH AND LOW SULFUR COAL PLANTS - 1200 MWE.
THE ESTIMATED TOTAL BASE CONSTRUCTION COST FOR THE 1200 MWE (NOMINAL) HIGH SULFUR CDAL PLANT REFERENCd DESIGN IS $465,493,393 OR $378/KW BASED ON JULY 1,
1976 PRICES.
THE ESTIMATED TOTAL BASE CONSTRUCTION COST FOR THE 1200 MWE (NOMINAL) LOW SULFUR COAL PLANT REFERENCE DESIGN IS S402,825,229 OR $324/KW BASED ON JULY 1,
1976 PRICES.
THE TOTAL BASE CONSTRUCTION COST FOR THE COAL-CIRED POWER PLANT (1000 MWE NET CUTPUT) REFERENCE IN WASH-1230 WHICH DID NOT HAVE FLUE GAS DESULFURIZATION IS APPROXIMATELY $174,000,000 OR $174/KW, BASED UPON PRICES EFFECTIVE JANUARY 1971 THUS, THIS 1977 STUDY INDICATES APPROXIMATCLY A 87.9 PERCENT INCREASE IN THE COST OF THE PLANT IN TERMS OF $/KW.
THE STUDY AND ITS METHODOLOGIES HAVE BEEN REVIEWED EXTENSIVELY WHILE IN PROGRESS EY THE RES PROJECT MA?tAGER AND VARIOUS STAFF MEMBERS FROM NRR.
RES RECOMMENDS THAT THE UPDATED METHODOLOGY BE USED BY NRR FOR APPLICATION TO THE IDENTIFIED REGULATORY HEED (RR-NRR-76-6).
i USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS
__QS1 91(_A_CTIVITIES REVIEW M
COMPLETLD H_fj D HELD ISSUED IMPLEMENTED P
OFFICE RESPONSIBLE......... NRR SCHEDULID COMPLETION DATE..
ACTUAL CCMPLETIDM **TE.....
ER,_C OMEN T S
- TECH, EESCRIBE AF
.gi_TO REGULATfRY PROCESS: THE FRINCIPA APPLICATI0H OF 1HE INVESTMENT COST DATA IS TO UPDATE THE
-PT COMPUTER CODE.
IN 'eURN THE CONCEPT COLT IS USED TO ESTIMATE CAPITAL COST FOR DIFFERENT SIZE PLANT.
FFERENT REGIONS OF THE COUNTY, DIFFERENT SCHEDULE LEhGTH, DIFFERENT COMPLETION DATES. DIFFERENT ESCALATION INTEREST CATES.
IN ADDITION TO UPDATING CONCEPY THE DATA IS USED AS A GENERAL REFERENCE FOR SUCH THINGS i COST OF COMPONENTS, QUANTITIES OF MATERIALS USED. TYPE AND QUANTITY OF LABOR. ETC.
Rf5fgl1E IMtf4T OF RE}ULTS: DURING FY 1977 44 REQUESTS WERE MADE FOR CONCEPT CODE RUNS AND 32 REQUESTS WERE MADE IN 1975.
MOST REQUESTS INVOLVED Rrh5 FOR SEVERAL SIZES OF CDAL PLAhTS AND ONE OR TWO NUCLEAR UNITS FOR DIFFERENT ESCALATION AND INTEREST RATES.
THE USE OF THE CSNCEPT CODE PERMITTED THE STAFF TO PFRFORM THESE ANALYSIS MORE EFFICIENTLY AND QUICKLY THAN IF THE CONCEPT CODE WERE NOT AVAILABLE.
COMMENTS /PEMARKS:
DOE IS FUNDING THE LATEST UPDATE.
i _ - _ _ _ _ _ - -
PROGPAM OFFICE COMMENTS ON POTENTIAL UTILXZATION OR UALpE OF RESEAPCH RESULTS IN TITE REGULATOR 1 PROCESS RIL s: 56 DATE ISSt*ED:
37/25/79 RES DECISION UNI?1 FUEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS RIL TITLE: EFFECTS OF NUCLEAR POWER PLAMTS ON COMMUNITY GROWTH AND RESIDENTIAL PROPERTY VALUES SPCNSORING OFFICE (S): NRR gggr 5-21 SGCIO-ECONOMIC RESEARCH PROJECT MGRr C. PRICHARD IMPACTS I
Ri$ CCM9ENTS: AS PART OF A LONGER iE'f?RESEARCH PROJECT TO DEVE'.0P FORECASTING METHODS FOR ESTIMATING THE IMPACT OF NUCLEAR PCWER STATIONS ON LAND VALUSS F.dD COMMUNITY GROWTN, THE FENNSYLVANIA STATE UNIVERSITY INSTITUTE FOR LAND AND d
WATER CESOURCES CARRIED OUT A STUDY sf THE EFFECTS CF NUCLEAR POWER STATIONS ON LAND VALUES AND COMMUNITY GROUTH AT FOUR PREVIOUSLY-LICENSED STF.TIONS IN THE NORTHEAST. THE RESULTS OF THE STUDY DID NOT INDICATE THE PRE 3ENCE OF SIGNIFICANT ADVERSE EFFECTS ON LAND VALUES OR COMMUNITY GROWTH.
USER DISCUSSION POSITICM COMMISSION ACRS PRESS OFFICE MEETING PAPER B R I G-IN G BRIEFING RELEASE RESULTS 2p3T RTL ACTIVITIES REVIEW HELD COMPLETED HELD HELD J33UED IMPLEMENTEQ OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
NRR COMMENTS, W.
RUSSELL -
RESCRIBE APPLICATION T0_REEMLATORY PROCf13: AS PART OF THE COST BENEFIT ANALYSIS OF LICENSING APPLICATIONS, THE NRC IS REQUIRED TO ASSESS THE LIKELY SOCI0 ECONOMIC IPPACTS ASSOCIATED slITH CONSTRUCTION AND OPERATION OF HUCLEAR POWER STATIONS ON LOCAL COMRUNITIES AND THE SURROUNDING REGIONS.
THE IMPACT OF OPERATING STATIONS ON LAND USE.
PRCPERTY iALUES, LAND USE PLANNING AND POPULATION CHANGE ARE EFFECTS WhICH HAVE RECEIVED CONSIDERABLE ATTENTTON FROM INTERVENORS AND HEARING BOARDS.
THIS RESEARCH CONTRACT WAS INITIATED WITH THE INTENTION OF TESTING A METHODOLOSY FOR QUANTIFYING IHE IMPACT OF STATION SITING ON RESIDFNTIAL PROPERTY VALUES.
DESCRIPE IMP ACr OF RE_1QLTi: BECAUSE OF THE NON-RANDOM SELECTION OF THE STUDY SITE, GENERALIZABLE CONCLUSIONS CANNOT BE DRAWN: HOWEVER, FOR THE 4 SITES INVOLVED. THE RESULTS DO INDICATE THAT THE NUCLEAR STATIONS HAD NO MEASURABLE. ADVERSE IMPACT ON R2SIDENTIAL VALUES IN SURROUNDING AREAS.
BESIDES ADDING TO THE T.TAFF'S GENERAL KNOWLEDGE. THE STLPDY FROVIDED THE STAFF WITH A TESTED, DBJECTIVE. AND REl(LICABLE METHODOLOGY FOR RETROSPECTIVELY DIMENSIONING P."0PERTY VALUE Id* ACTS AT NUCLEAR STATIONS. WITH RESPECT T THIS POINT, THE STAFF INTENbS TO EVALUATE THE IMPACs OF THE ACCIDENT AT THREE MILE ISLAND ON PROPERTY VALL S USING A METHODOLOGY BASED ON THE PROCEDURE USED IN THIS CONTRACT.
THE METHODOLOGY AND CONCLUSIONS OF THIS RESEARCH INCREASED THE STAFF'S UN 4RSTANDING OF THE PROCESS OF PROPERTY VALUE IMPACT AT THE LOCAL LEVEL.
CONTINUATION OF RESEARCH ON PROFERTY VALUE IM ACTS SHOULD RESULTS IH GENER.itLY RELIABLE NETHODS FOR PREDICTING RESIDENTIAL VALUE IMPACTS IN SPECIFIC SITING CASES '
SCMMENTS/ REMARKS: NONE l
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___PPOGPAM OFFICE CO"MENTS 04 p0TENTIAL UTILIZATION OR VALUE OF RESEARCH RESUf!'7 1N TPE REGULATORY PROCESS PIL s: 57 DATE 11121fR: 03/10/79 RES DECISI0hjJNIT: SYSTEMS ENGINEERING SIL YITLE: SMALL SCALE ECC BYPASS RESEARCH RESULTS SPCNSCRING OFFICE (S): NRR ERS2 1-6 ECC BYPASS RESEARCH PROJECT MGR:
A. SERKIZ f
PES CCMMENT_S:
ANALYTICAL AND EXPERIMENTAL STUDIES OF STEAM-WATER MIXING EFFECTS ON THE PENETRATION OF C0LLING WATER i
IN SMALL M00ELS CF PWR VESSELS HAVE BEET 3 CONOUCTED OVER THE PAST FIVE YEARS AT BATTELLE COLUMSUS LAs0RATORIES AND CREARE. INC.
FISCAL YEAR 1979 MARKED THE COMPLETION OF THE BULK OF THIS W3RK AND A RESEARCH INFORMtTION LETTER SUMMARIZING THE Fit 4 DINGS WAS PREPARED.
THE PHEFCMENA INFLUENCING COOLING WATER PENETRATION Ih THE SMALL t/tS AND 2/15 SCAL E MODELS IS WELL UNDERSTOOD AND A TRANSIENT MODEL DESCRIBING C00LIhG WATER PENETRATION IN THESE SMALL SCALE TESTS HAVE BEEN CEVELCPED. SENSITIVITY STUDIES OF THE MODEL UNCERTAINTY (PRIMARY SCALING; WHEN USED TO OALCULA1E COOLING WATER PENETRATION IN A PWR HAVE BEEN PERFORMED AND COMPARISONS MADE WITH LOFT DATA.
THE SMA*.L SCALE TESTS SUPPORT THE NEED FOR ADDITIONAL TESTS AT LARGER SCALE. BUT PROVIDE EVIDENCE OF THE CONSERVATISM IN THE MODELS USED IN THE LICENSING PROCESS USER DISCUSSION PO!ITION COMMISSICN ACRS PRESS OFFICL MEETING PAPER BRIEFING BRIEFING RELEASE FESULTS EDIT RIL ACT1yITIES PEVIEW EELD COMPLETED HEL D HELD ISSUED _
IMPLEMENT;d CFFICE RESPet4SIELE......... NRR SCHEDULED COMPLETION DATE..
I ACT P: AL 20MPLETION DATE.....
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1 iRc0&GM OFFICE COMMENTS ON POTENiXAL UTILIZATION 04 v4L6E CF RESEQRCH RE$UAVS XH VHE REGillATQ31_PRCCESS I
RES DECISION UNIT-f FUEL CYCLE SAFETY 8 ENVIRDNF. ENTAL EFFECTS RIL 4: 53 DATE ISSUED:
C8/"
i RIt TITLE: CCMPARISON OF SIMULATION MODELS USED IN ASSESSIN3 THE EFFECTS OF POMER PLANT INDUCED MORTALITY ON FISH POPULATIONS j
SPONSORING 0FFICE(S): NRR SPE:
5-15 CHEMICAL INPACTS RESEARCH PROJECT qq&:
P. REED ON AQUATIC ENVIRONMENTS RES CCMMENTS: THIS RESEARCH EVALUATED THE EXISTING MATHEkATICAL MODELS FOR PREDICTING THE IMPACT OF NUCLEAR POWER PLANT OPERATION ON ECONOMICALLY IMPORTANT FISH SPECIES.
'HE MATHEMATICAL fTATEMENTS, THE EQUATIONS. AND UNDERLYING ASSUMPTIONS USED FOR ASSESSING PCWER PLANT INDUCED FISH MJRTALITY WERE COMPARED. VALUES OF PARAMETERS AND THE TECHNICAL DATA SOURCES USED TO OBTAIN THEM WERE INVESTIGATED. B2CAUSE MANY OF THE MODELS HAD DIFFERENT BASIC ASSUMPTInMS, PARAMETRIC VALUES, OR BOTH, AN INTERACTIVE (IFE-CYCLE MODEL SIMULATOR WAS DEVELOPED TO COMPARE THE PREDICTIGNS OF THE VARIOUS MODELS.
IT WAS DETERMINfy THAT NO PRESENTLY EXISTING MODEL CAN BE USED TO MAKE QUANTITATIVE IMFACT PREDICTIONS.
!i USER DISCUSSION POSITION COMMI3SION ACRS PRESS OFFICE MEETING PAF!ER BRIEFING BRIEFING RELEASE RESULTS E957 'It ACTIVITIES REVIEW HELD
[EpPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPCNSIBLE......... NRR SCHEDULED COMPLETION DATE..
-d ACTUAL COMPLETICM DATE.....
-f
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i PROGRAM OFFICE CCP"ENTS ON POTENTIAL UillIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL s: 59 DATE ISSUED: 07/21/79 RES DECISION UNIT: FUEL BEHAVIOR g.It TI TL E: TRANSIENT FUEL RCD BEHAVIOR CODE:
FRAP-T4 SPONSORING OFFICE (S):
NRR RES:
1-12 FUEL CODE RESEARCH PROJECT MGR G. MARIND DEVELOPMENT PEl_CCEMfNTS:
FRAP-T4 IS A BEST-ESTIMATE COMPUTER CODE THAT CALCULATES THE THERMAL AND MECHANICAL RESPCNSE OF A NUCLEAR FUEL RCD DURING NORMAL, OFF-NORMAL, AND TRANSIENT CONDITIONS. THE CODE NOW HAS THE CAPABILITY TO BE USEO IN THE ANALYSIS OF THE ENTIRE SEQUENCE OF A LOCA THAT WENT THROUGH REFLOOD. THE CODE HAS BEEN USED TO ANALYZE SUCH EVENTS FOR THE POWER 3URST FACILITY TEST PROGRAM AND THE LOSS-OF-COOLANT FLUID TEST PROGRAM AT THE IDAHO NATICNAL ENGINEERING LABORATORY.
IT HAS AL50 BEEN USED TO STUDY OTHER TRANSIENT EVENTS SUCH AS REACTIVITY-INITIATED ACCIDENTS, POWER-COOLING MISMATCH EVENTS, AND ANTICIPATED TRANSIENTS WITHOUT SCRAM.
THE REPORT j
SUMMARIZES THE CODE'S PERFORMANCE IN THE ABOVE AREAS VIA DATA COMPARISONS, CONTAINS A SU;! MARY TABLE WHICH j
SHOWS THE STANDAR3 ERROR BETWEEN DATA AND CALCULATED RESULTS, AND CONCLUDES WITH A SECTION ON USER RECOMMENDATIGNS BASED ON THE CODE ASSESSMENT RESULTS.
l USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS E0iT_R1(_AQTIVITIES REVIEW HEL D COMPLETED HEL D HELD ISSUED IMPLEMENTED 0FFICE RESPONSIBLE......... NRR
-~
SCHEDULED COMPLETION DATE..
ACTUAL COM?LETION DATE.....
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__ PPCGRAM OFFICE COPMEN15 04 POffETIAL UTILIZATION OR VALUE OF RESEARCH RE50k1A_JN IPE 8EGULATORY PR6CE35
- i E1L_s: 60 DatE ISS'f ED t
~0/12/79 RES DECISTON UNIT: SEISMIC, ENGINEERING 8 SITE SAFETY pit TITLE: SEISMICITY AND TECTONIC RELATIONSHIPS OF THE NEMAHA UPLIFT IN OKLAHOMA SPONSCPIkG OFFICEfS): NRR R39:
3-2 GEOLOGY 3 SEISMIC RESEARCH PROJECT MGR:
N. STEUER CHARACTERISTICS EE5_Sp*r13T1: GEOLOGIC AND SEISMot0GIC INVESTIGATIONS OF THE HEMAHA UPLIFT BEGAN CH OCTOBER t. 1976.
THE CE0 LOGICAL SIUJIES HAVE FOCUSED, THUS FAR, ON THE CONSTRUCTION OF A SERIES OF STRUCTURE-CONTOUR MAPS ON KEY STRATIGRAPHIC HCRIZONS: THE TOP OF THE OR00VICIAN VIOLA FORMATION, THE BASE OF THE PENNSYLVANIAN. AND THE TOP OF THE OSWEGO FORMATIC'J.
THE CCHTOUR-MAPPING PHASE OF THE PROGRAM IS APPROXIMATELY TWO THIRDS COMPLETED. THE INITIAL MAPPING PROGRAM REVEALS A COMPLEX FAULT PATTERNS AND GEOLOGIC HISTORY OF THE NEMAHA RIDGE.
IT APPEARS THAT THE UPLIFT AND ASSOCIATED FAULTS BEGAN IN EARLY PENNSYLVANIAN TIME AND THAT TECTONIC ACTIVITY CEASED IN MIDDLE PENNSYLVANIA TIME. AT LEAST IN CENTRAL OKLAHOMA.
A DISCUSSION OF BASEMENT ROCKS IN CENTRAL OKLAHOMA IS INCLUDED WITHIN THIS REPORT. THE SEISMOLOGICAL STUDIES HAVE CONCENTRATED ON THE INSTALLATION OF EIGHT SEISM 0 METERS IN SUCH A WAY AS TO INCLUDE DETAILED COVERAGE OF THE ENTRIE NEMAHA RIDGE IN OKLAHCMA AS WELL AS MOST OF THE REMAINING AREAS OF OKLAHOMA.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST Rit_A.CTIVITIES REV11H HELD COMPLETED HELD NELD ISSUED IMPLEMENTED OFFICE RESP 0hSIBLE.........
NRK SCHEDt: LED CCMPLEi10N DATE..
ACTUAL CCMPLETION DATE.....
!LP8Lf0PTENTS, NOVE*1BER 9,
1979:
LE30PlPE A7tICATIgN_TQ_ELQULATORY PROCESS: IT IS RECOMMENDED THAT THE INFORMATION CONTAINED IN NUREG/CR-0050 BE CONSIDERED BY THE OFFICE OF NUCLEAR REACIOR REGULATION AS INPUT INFORMATIOP TO THE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA.
DESCRU;E IFraCT OF PESULT3: THE AVAILABILITY OF THE RESEARCH HAS BROADENED OUR DATA BASE FOR THIS REGION BUT CURRENTLY HAS NO DIRECT IMPACT ON OUR LICENSING ACTIVITIES.
THE GEOLOGIC AND SEISMIC DATA BASE AVAILABLE IN THIS REPORT IS INSUFFICIENT TO COMPLETELY EVALUATE THE AREA EXCEPT ON A PRELIMINARY BASIS.
A SEISM 0 TECTONIC MODEL FOR
!hE NEMAHA UPLIFT IN OXLAHOMA MUST ALSO BE BASED IN PART ON CONCEPTS DEVELOPED FROM RESULTS OF THE STUDIES AND I
MUST CONSIDER VERTICAL AND LATERAL VARIATIONS IN COMPCSITION AND PHYSICAL PROPtRTIES OF THE NEMAHA UPLIFT.
CQ_M3ENTS/ REMARK $: THIS heREG CONSTITUTES PART OF A LARGE DATA GATHERING AND SYNTHESIS EFFORT FOR THE NEMAHA RIDGE AREA.
THE TOTAL IMPACT CANNOT BE ASSESSED UMTIL THE OVERALL PROCRAM IS COMPLETED AND SYNTHESIZED WITH SEISMIC MONITORING DATA.
THIS WILL TAKE SEVERAL '.SARS. f
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PROG &AM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS DIL
- 61 DATE ISSUED:
10/11/79 RES DECISION UNIT: FAST BREEDER REACTORS PIL TITLE: M3tTEN SODIUM INTERACTION WITH BASALT CONCRETE i
SPONSCRING CFFICE(S): HRR RRL:
2-4 CORE MELT AND RESEARCH PROJECT MGR:
T. WALKER i
CONTAINMENT INTEGRITY 4
Rf5 COMMENTS: LARGE SCALE MOLTEN SODIUM-BASALT CONCRETE TESTS WERE CARRIED OUT IN SUPPORT OF THE NRC STAFF m
1 POSITICM THAT VIGCRCUS SODIUM-CONCRETE REACTIONS COULD BE EXPECTED. BASALT CONCRETE PENETRATION RATES OF ABOUT I
2.5 CM. PER HOUR WERE OBTAINED.
THE REACTION CONTINUES UNTIL THE SODIUM IS CONSUMED CN EITHER BARE CONCRETE OR f
3ENEATH A DEFECTED STEEL LINER.
DEFECTS AS SMALL AS 0.6 X 15 CM. DO NOT PLUG BUT PERMIT ESSENTIALLY ALL OF TFE I
SODIUM TO REACT WITH THE CONCRETE WHILE THE REACTION PRODUCTS DEFORM THE LINER.
SILICEOUS FIREBRICK INSULATING f
L AYERS BENEATH THE LINERS ALSO REACT READILY WITH THE SCSIUM.
IF THE SIDE WALLS ARE EXCLUDED, THE TOTAL CONCRETE PENETRATION IS ABDUT SO: OF THE INITIAL SODIUM POOL DEPTH.
4 USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASF RESULTS 4
Pp1T_FIL ACTIVITIES PfviEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED d
0FFICE RESPCNSIBLE......... hRR SCHEDULED COMPLETION DATE..
I ACTUAL COMPLETION DATE.....
i NDR C955fNTS, NOVEM1ER 9,
1979:
EllSP.lAE APPLICA11QN TO REGM1AlpRY PROCEll: AS INDICATED IN DR. LEVINE'S LETTER OF OCTOBER 11, 1979, NRR REQUESTED
{
(CASE IO LEVINE LETIER DATED 3/7/19) IHAT THE CONFIRMATORY RESEARCH PROGRAM INVESTIGATE THE INHERENT RETENTION j
CAPABILITY FOR SCDIUM AND SCME CORE DEBRIS BE REDIRECTED TOWARDS FFTF RATHER THAN CRBR FOR WHICH LICENSING ACTIVITIES HAD BEEN SUSPENDED.
THERE IS NO LICENSING ACTIVITY DH CRBR AND ITS FUTURE IS UNCERTAIN. THE FFTF IS ESSENTIALLY COMPLETE, THEREFORE. THE INTEREST IN THIS EFFORT FOR LMFBRS IS GENERIC. IN HATURE.
THE MOST l
DIRECT APPLICATION OF THE REMAINDER OF THIS EFFORT MAY BE TOWARDS THE PROTUTYPE LARGE BREEDER STUDY.
PE1ERi&f_JMPACT OF RE1QLTS: THE RESULTS OF THIS WORK STRONGLY INFLUENCED THE ADVICE AND RECOMMENDATIONS GIVEN 1
TO THE FFif PROJECT REGARDING THE ADEQUACY OF EXISTING FFTF CONTAINMENT MARGINS AS REPORTED IN THE STAFF'S SAFEIY EVALUATION REPORT NUREG-0358. SUPPLEMENT 1 DATED MAY 1979.
THE RECOMMENDATION MADE BY THE STAFF AND SUPPORTED BY ACRS INCLUDED MEANS FOR MONITORING THE CONTAINMENT ATMOSPHERE FOR RADIATICN, TEMPERATURE. PRESSURE, OXYGEN, AND THE INCCRPORATICN OF HYDROGEN IGNITERS AS WELL AS MEANS TO SCRUB / FILTER THE CONTAINMENT ATMOSPHERE IN THE EVENT OF MELTDCWN TO REDUCE THE RADIOLOGICAL RELEASES.
f
{QC]fN114ELMARK$: FUTURE GENERIC STUDIES OF OTHER CONCRETE AGGRAGATES, CELL CONFIGURATIONS (CYLINDRICAL, RECTANGULAR)
SHOULD BE CCNDUCTED AS WELL AS TESTS WITH AND WITHOUT FLAWED LINERS.
(FLAWS SIMULATING WELD CRACKS SHOULD BE j
CCNSIDERED.) DIFFERENT TYPES OF FIREBRICK (E.G.,
COMMERCIAL MGO, AL203, ZRC2) SHOULD BE EXAMINED IN TERMS OF COMPATIBILITY WITH SODIUM AND ABILITY TO PROTECT THE CONCRETE.
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PkOGOAM OAFICE CCPMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGQATORV PROCESS PIL s: 62 DATE TSSUED:
11/01/79 RES DECISTON UNIT: SEISMIC. ENGINEERING 3 SITE SAFETY PIL TITLE: NEW MADRID SEISM 0 TECTONIC STUDY SPONSCDING OFFICE (S):
3-2 GEOLOGY 8 S(ISMIC EESEARCH PROJECT MGR:
N. STEUER CHARACTERISTICS PES COMMENTS: AN IMPORTANT GOAL OF THE RESEARCH PROGRAM IS TO PRODULE USEFUL SEISM 0 TECTONIC AND SEISMIC ZONING MAPS FOR THE STUDY AREA.
FISCAL YEAR 1978 WAS THE SECOND YEAR OF A FIVE-YEAR PROGRAM.
RESULTS OF AER0 MAGNETIC SURVEYS FUNDED IN FY 77 WERE INTEGRATED WITH PREVIOUSLY EXISTING DATA IN ADJACENT AREAS.
EXTENSIVE GRAVITY SURVEYS WERE MADE IN KENTUCKY AND INDIANA NEAR THE INTERSECTION OF THE 33TH PARALLEt LINEAMENT AND THE NGRTHEASTERN EXTENSION OF THE NEW MADRID SETSMIC ZONE.
THE STATIONS WERE GRAVIMETRICALLY CONSTRUCTED. AN INTERESTING RELATIVE POSITIVE ANOMALY CCCURS PARALLEL TO THE WABASH VALLEY FAULT SYSTEM.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PQ3T *It AQTIVITIE$
REVIEW HELD COMPLETED HELD HEL D ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR.SD SCHEDULED CCMPLETION DATE..
ACTUAL CCMPLETION DATE.....
NPR QSCEN_l$,_tLQVEMR ER 15,
- 1979, R.
JACKSON:
piLER1QE APPLICATION To REGULA10EY_PP0 CESS: THE GOAL OF THIS RESEARCH BY SEVERAL STATE GEOLOGICAL SURVEYS AND UNIVERSITIES IS 10 DEFINE THE SIRUCTURAL SETTING AND TECTONIC HISTORY OF THE REGION AROUND NEW MADRID. MISSOURI IN ORDER TO PROVIDE THE BASES FOR A MORE REALISTIC APPRAISAL OF THE EARTHQUAKE RISKS IN THE SITING OF NUCLEAR FACILITIES IN THE NORTH AMERICAN MID CONTINENT.
IT IS RECOMMENDED THAT THE INFORMATION CONTAINED IN NUREG'S 0739 AND 0450 BE USED AS A BASIS FOR CONTINUING RESEARCH, AS INPUT TO THE EVALI!ATION OF SEISMIC RISK IN THE REGION WITHIN AND AROUND THE MISSISSIPPI EMBAYMENT, AND AS A CONTRIBUTION TO OUR UNDERSTANDING OF INTERPLATE TECTONICS IN GENERAL.
RES C_Ei B_E_15P A C T OF R ESU LT): THE RESEARCH EFFORT THUS FAR HAS INCREASED OUR CURRENT DATA BASE AND OUR UNDERSTANDING OF EARIHQUAKE AND FAULT PHENOMENA IN THE MISSI3SIPPI EMBAYMENT REGION, BUT AS YET NO DIRECT IMPACT ON LICENSING.
0PMENTS/EENAPf3:
NUREG'S 0739 AND 0450 SUMMARIZE THE STUDIES AND RESULTS OF THE FIRST TWO YEARS OF A FIVE YEAR PROGRAM, THEREFORE, IT IS T"*
..mY TO ASSESS THE IMPACT ON NUCLEAR POWER PLANT LICENSING EXCEPT IN A VERY PRELIMINARY WAY.
THE TOTAL I
'*r'
.J. BE ASSESSED UNTIL THE OVERALL PROGRAM IS COMPLETED AND SYNTHESIZED WITH SEISMIC MONITORIbG DATA.
- ROGRAM OFFICE CC*MENTS 04 POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN 1HE REGULATORY PROCESS RIL 8:
63 DATE ISSUED:
11/01/79 RES DECISION. UNIT: LOFT PIL TITLE: LOFT REACTOR SAFETY PROGRAM RESEARCH RESULTS FROM NUCLEAR LOSS-OF-COOLANT EXPERIMENTS L2-2 AND L2-3 SPONSORING OFFICE (S):
NRR RRG:
1-1 LOFT RESEARCH PROJECT MGR:
G. MCPHERSON PES CO*MENTS:
THE 2ND LOFT RIL SUMMARIZES THE RESU' TS OF THE FIRST TWO NUCLEAR LOCE'S PERFORMED AT THE LOFT THE T' 0 TESTS ARE PART OF THE L2 * *RGE COLD 8 EG BREAK EXPERIMENTS.
L2-2 WAS CGNDUCTED OF FACILITY IN FY1979.
4 A LINEAR HEAT GENERATION RATE OF S KW/FT AND L2-3 WAS AT 12 KW/FT.
BOTH TESTS, WHICH ASSUMED AVAILABILITY OF OFF-SITE POWER, DISPLAYED A DOUBLE REVERSAL OF CORE FLOW DURING DEPRESSURIZATION. THE RETURN TO POSITIVE CORE FLCW WAS SUFFICIENT TO QUENCH THE CORE PRIOR TO THE INITIATION OF EMERGENCY CORE COOLANT INJECTION. IMPROVEMENTS ON COMPUTER CODE MODELS AND NODALIZATION ARE IDENTIFIED WHICH PERMIT A GOOD PREDICTION OF THE OBSERVED PHENCMENA.
Ali 0F THE BEST-ESTIMATE CODES USED TO PREDICT L2-3 PREDICTED GENERALLY HIGHER CLADDING TEMPERATURES THAN WERE MEASURED.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS C01T_Ril ACTIVITIES PEVIEW HELD COMPLETED HEtD HELD ISSUED IMPLEMENTiQ OFFICE RESPONSIBLE......... NRR SCHEDULED CCMPLETION DATE..
ACTUAL C09PLETION DATE.....
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I PROGDA9 CFFICE_ COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESET *CH RESULTS IN T H E R EGU L A T O RY PR QRLiL_
RIL 8: 64 DATE ISSUED:
11/05/79 RES DECISION UNIT: SEISMIC, ENGINEERING 2 SITE SAFETY PIL TITLE: A REVISED AND AUGMENTED LIST OF EARTHQUAKE INTENSIl!ES FOR KANSAS, 1867-1977 SPONSORING OFFTCE(S):
HRR, SD RRQ:
3-2 GEOLOGY 4 SEISMIC RESEARCH PROJECT MGR:
N.
STEUER CHARACTERISTICS PES CC**ENTS:
THE PURPOSE OF THIS RESEARCH IS TO GAIN A BETTER UNDERSTANDING OF THE SOURCES OF EARTHQUAKES IHA! HAVE OCCURRED IN THE REGION AS AN AID TO DEVELOPING A MORE RATIONAL EVAL *JATION OF EARTHQUAKE RISK AS IT APPLIES TO THE SITING AND DESIGN OF NUCLEAR FACILITIES.
TWENTY-FIVE EARTHQUAKES WHOSE EPICENTERS WERE WITHIN THE BORDERS OF KANSAS HAVE BEEN REPORTED DURING THE PAST 110 YEARS.
BECAUSE OF THE CRITICAL NATURE OF EARTHQUAKE INFORMATION IN ESTIMATION OF SEISMIC RISK, IT IS IMPORTANT THAT THE DATE, LOCATION, AND SIZE OF EACH EARTHQUAKE BE DETERMINED AS ACCURATELY AS POSSIBLE.
THE INVESTIGATION INCLUDED A REVIEW OF THE REFERENCES CITED FOR KANSAS EARTHQUAKES BY AUTHORS OF PREVIOUSLY PUBLISHED STATE, REGIONAL, AND NATIONAL EARTHQUAKE LISTINGS.
IN ADDITION, OLD NEWSPAPER FILES, MICROFILMS, AND OTHER RECORDS AT THE UNIVERSITY OF KANSAS AND THE KANSAS STATE HISTORICAL SOCIETY WERE SEARCHED FOR REPORTS WHICH MAY HAVE BEEN PREVIOUSLY OVERLC0KED OR NOT RECORDED.
THIS REPORT INCLUDES A COMPLETE LIST OF ALL FELT REPORTS COMPILED DURING THIS STUDY.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEEIING PAPER BRIEFING BRIEFING RELEASE RESULTS RQST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSULD IMPLEMENTED OFFICE RESPONSIBLE......... NRR,SD SCHEDULED COMPLETION DATE.
ACTUAL COMPLETION DATE.....
!*2Rl 0 EENTS, DE[fMi[R S,
- 1979, J.
KNIGHT:
RL1SR15.E APPLICATION TO REGULATORY PROCESS: THE COAL OF THiS RESEARCH WAS TO REVIEW THE REPORTS OF ALL EARTHQUAKES WHOSE EPICENTERS WERE WIIHIN THE BOUNDARIES OF KANSAS.
THIS RESEARCH IS PART OF A COOPERATIVE GEOLOGIC, SEISMIC, AND GEOPHYSICAL RESEARCH PROGRAM BY SEVERAL STATE GEOLOGICAL SURVEYS THAT IS SEEKING TO DEFINE THE STRUCTURAL SETTING AND TECTONIC HISTCRY OF THE NEMAHA UPLIFT AND THE MIDCONTINENT GRAVITY ANOMALY IN ORDER TO PROVIDE THE BASES FOR A MORE REALISTIC APPRAISAL OF THE EARTHQUAKE RISKS IN THE SITING OF HUCLEAR FACILITIES IN THE NORTH AMERICAN MID CONTINENT.
IT IS RECOMMENDED THAT THE INFORMATION CONTAINED IN NUREG/CR-0294 BE USED AS A BASIS FOR CONTINUING RESEARCH AND AS INPUT TO THE EVALUATION OF SEISMIC RISK IN THE REGION WITHIN AND AROUND THE NEMAHA UPLIFT.
REig&LBE IMPACT OF PESULT_L:
THE RESEARCH EFFORT THUS FAR HAS INCREASED OUR CURRENT DATA BASE AND OUR UNDERSTANDING OF EARIHQUAKE PHENOMENA IN f HE VICINITY OF THE NEMAHA UPLIFT, RUT AS YET HAS HAD HO DIRECT IMPACT ON LICENSING.
NUREG/CR-0294 HAS BEEN REFERENCED BY AN INTERVENOR IN THE WOLF CREEK HLARING. THE INTERVENOR REQUESTED THE STAFF TO REASSESS THE SAFE SHUTDCWN EARTHQUAKE (SSE) AT THE WOLF CREEK SITE IN LIGHT OF THE NEW INFORMATI0H PRESENTED IN THIS NUREG.
THE STAFF REVIEWED THE REPORT AND FOUND THAT THE SSE AT THE WOLF CREEK SITE WAS STILL ADEQUATELY CONSERVATIVE.
[qrMENTS/PEMAPK5:
NUREG/CR-0294 COMPRISES A PART OF A FIVE-YEAR DETAILED STUDY OF THE SOURCES OF SEISMICITY IN THE NLMAHA UPLIFT AREA.
THIS IS AN INTERIN TOPICAL REPORT PRESENTING RESULTS OF WORK COMPLETED IN PHASE I; THEREF0F.2, IT IS TOO EARLY TO ASSESS THE IMPCT ON NUCLEAR POWER PLANT LICENSING EY2EPT IN A VERY PRELIMINARY WAY.
THE T.OTAL IMPACT f l
CANNOT BE ASSESSED UNTIL THE OVERALL PROGRAM IS COMPLETED AND SYNTHESIZED WITH SEISMIC MONITORING DATA.
THESE PRELIMINARY RESULTS ARE BEING CONSIDERED IN THE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S.
THE STAFF DOE! DISAGREE WIT:1 ONE OF THE RESUL TS STATED IN THE NUREG - THE 1867 MAhMATTAM EARTHQUAKE. THE ASSIGNMENT OF INTENSITY VII-VIII IS BASED UPON AN 1877 REPORT OF LIQUEFACTION ON A FARM ON THE FLOODPL AIN OF THE KANSAS GEOLOGICAL SURVEY.
THAT CBSERVATION WAS ASSIGNED INTENSITY VIII AND PLACED CLOSE TO THE EPICENTER BY THE KANSAS l
GEOLOGICAL SURVEY. MUCH RECENT WORK HAS SHOWN THAT LIQUEFACTION IS EXTREMELY DEPENDENT UPON LOCAL SITE CONDITIONS AND MAY OCCUR IN ISOSEISMAL AREAS THAT MAY OTHERWISE BE ASSOCIATED WITH INTENSITIES LESS THAN VIII (AS LOW AS VI FOR EXAMPLE).
THE STAFF AGREES WITH THE STANDARD REFERENCES WHICH LIST THIS EARTHQUAKE AS AN INTENSITY VII (MM).
THE STAFF ALSC FINDS THE EPICENTRAL LCCATION OF THE 1867 MANHATIAN EARTHQUAKE QUESTIONABLE. ACCORDING TO THE MANSAS GECLOGICAL SURVEY, SHAKING AND BUILDING DAMAGE EQUIVA*.ENT TO INTENSITY VII OCCURRED DVER AN AREA AT LEAST 200 KM ACROSS. PINPOINTING THE EPICENTER WITHIN THAT AREA MAY BE BEYOND THE RESOLVING POWER OF THE PRESENT DATA.
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PROGkAM OFFICE CCPMENT! ON PorENTIAL UTILIZATION 03 vatVE OF RESEARCH RESULTS IN VHE REGJ1LOTORY PROCESS PIL er 65 DATE ISSUED:
11/05/79 RES DECISION UNIT: SEISMIC. ENGINEERING & SITE SAFETY 1
j PIL T(LLE RECONNAISSANCE BEDROCK GEOLOGIC MAP OF MARLBOROUGH QUADRANGLE, MA AND RECONNAISSANCE EEDROCK GEOLOGIC MAP OF SHREWSBURY QUADRANGLE, MA SPONSOPING OFFICE (S1:
3-1 NRC/ STATE REdIONAL RESEARCH PROJECT MGR:
N. STEUER EARTH SCIENCES PE1_qpFMENTS: THE MAPS TRANSMITTED BY THIS RIL ARE PRODLCTS OF THE NEW ENGLAND SEISM 0 TECTONIC STUDY WHICH IS A PR',oR EM OF INVESTIGATICNS TO BETTER UNDERSTAND THE MANIFESTATIONS AND CAUSES OF SEISMICITY IN NEW ENGLAND AND ADJACENT AREAS TO ASSESS THE SEISMIC HAZARD TO PROSPECTIVE NUCLEAR POWER PLANTS IN THE REGION. RTS RECOMMENDS
- 1) THAT SD AND NRR CONSIDER THE MAPS AS INPUT TO THE DEVELOPMENT GF # TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA.
USER DISCUSSION POSITION COMMISSION ACR$
PRESS OFFICE MEETING PAPER BRIEFING BRIEFINC RELEASE RESULTS EQST FIL AQTIVITIES REV11W HELD COMPLETED HELD HELD ISSUED IMPLEMENTED CFFICE RESFONSIBLt......... NRR.SD SCHEDULED COMPLETION DATE..
ACTUAL COMPLETICN DATE....
NPR C053ENT). JANUARY 9, 19 ?. 0,
J.
KNIGHT:
DESCRIBE APPLICATION TO REGULATORY PROC E}S : THESE TWO GEOLOGIC MAPS ARE PRODUCTS OF THE NEW ENGLAND SEISM 0 TECTONIC STUDY. WHICH IS A PROGRAM OF INVESTIGATIONS TO ASSESS THE SEISMIC HAZARD OF THE REGION.
THEY CONTAIN THE KIND OF DETAIL THAT IS NEEDED IN AREAS OF COMPLEX MAJOR FAULTING AND/OR RELATIVELY HIGH SEISMICITY.
THEREFORE, IT IS l
RECOMMENDED THAT THE MAPS BE CONSIDERED AS INPUT INFORMATION TO THE DEVELOPMENT OF A SEISM 0 TECTONIC PROVINCE MAP J
AND AS A BASIS FOR CONTINUED STUDY IN THE AREA.
pfire 'J E IMPACT OF RESULTS: THE DATA ON THE MAPS HAVE NO DIRECT IMPACT ON LICENSING ACTIVITIES. BUT HAVE ADDED TO 1
IHE STAFF *S Gtt4ERAL KNCWLEDGE REGARDING THE MOST GEOLOGICALLY COMPLEX AREA IN NEW ENGLAND.
THEY CONFIRM THAT THERE IS A MAJOR STRUCTURAL BOUNDARY SEPARATING SOUTHEASTERN MSSACHUSETTS FROM THE REST OF THE NEW ENGLAND PIEDMONT PROVINCE.
NO EVIGENCE OF RECLNT MOVEMENT ON THE FAULTS WAS FOUND.
THE BENEFITS OF INVESTIGATIVE EFFORTS SUCH AS THIS WILL NOT BE FULLY REALIZED UNTIL OTHER CCM5fMT3/ REMARKS:
GEOLOGICALLY COMPLEX AND/OR RELATIVELY HIGH SEISMIC AREAS ARE STUDIED AND ALL THE INFORMATION SYNTHESIZED AND INTER-PRETED AT THAT TIME THE PRODUCTS WILL CONTRIBUTE TO THE CONSTRUCTION OF A SEISMOTECTONIC PROVINCE MAP.
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P&OGRAM O'FICE CCPNENis ON POTENTIAL UTILIZATION OR VALUE OF RESEADCH RESULTS IN THE REGULATCRY PROCESS RIL 8:
66 DATE ISSUED:
11-06-79 RES DECISION UNIT: SEISMIC, ENGINEERING 4 SITE SAFETY pit TIILE: A STUDY OF THE REGIONAL TECTONICS AND SEISMICITY OF EASTERN KANSAS -
SUMMARY
OF PROJECT ACTIVITIES AND RESULTS TO THE END OF THE SECOND YEAR OR SEPTEMBER 30, 1978 SPONSCDING OFFICE (S):
NRR, SD RE3: 3-1 NRC/ STATE REGIONAL RESEARCH PROJECT MGR:
N.
STEUER EARTH SCIENCES RE3_forrENT3: THIS RIL TRANSMITS RESULTS TO SEPTEMBER 30, 1978 0F THE STUDY CONDUCTED OF THE EARTH SCIENCE PARAMETERS OF THE NEMAHA UPLIFT AND THE MIDCONTINENT GRAVITY ANOMALY GEOLOGIC STRUCTURES. THE INFORMATION GAINED IS OF VITAL IMPORTANCE IN THE SITING AND LICENSING OF NUCLEAR POWER PL ANTS.
THIS RIL PRESENTS PROJECT WORK COMPLETED IN PHASES I AND II 0F A 3-PHASE PROJECT AND PRESENTS 1) EXISTING DATA SYNTHESIS, AND 23 ACQUISITION OF NEW DATA. SEISMIC NETWORK INSTALLATION AND GPERATION.
RES RECOMMENDS THAT SD AND NRR CONSIDER THIS AS INPUT TO THE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA.
DOCUMENT ISSUED: NUREG/CR-0666 (ALSO SEE RIL 870. 11/19/,79 AND NUREG/CR-0375).
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS P_0S T_Rll aC11VITIES RMF W HELD COMPLETER HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE.........
NRR.SD SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
P'S"EN T S, DEC[M ER 15, 1979 J.
KNTCHT:
j kt E(R IAL_A P P L I C A T I ON TO_PLEylAIORY PROCESS:
THE PURPOSE OF THIS KANSAS "ESEARCH IS TO GAIN A BETTER UNDERSTANDING OF THE SOURCES OF EARTHQUAKES THAT HAVE OCCURRED IN THE REGION.
THE KANSAS EFFORT IS A PART OF A REGIONAL RESEARCH PROGRAM WNICH ALSO INCLUDES THE STATE GEOLOGICAL SURVEYS OF OKLAHOMA, NEBRASKA AND IDWA.
THIS RESEARCH PROGRAM IS TO SERVE e
AS AN AID TO THE NRC IN DEVELOPING A RATIONAL EVALUATION OF EARTHQUAKE RISK AS IT APPLIES TO THE SITING, DESIGN AND REEVALUATION OF NUCLEAR FACILITIES LOCATED WITHIN THE STUDY REGION.
OESCPIBE IMPACT OF RESULTS: THIS RESEARCH EFFORT HAS INCREASED OUR GEOLOGICAL, GEOPHYSICAL AND SEISMOLOGICAL DATA BASE OF EASTERN KANSAS, BOTH THROUGH THE ACQUISITION AND GENERATION OF NEW INFORMATION AS WELL AS THROUGH A COMPILATION AND EVALUATION OF F REVIOUSLY EXISTING DATA.
THIS KANSAS RESEARCH EFFORT, IN PARTICULAR, HAS STRENGTHENED OUR CONFIDENCE IN THE PREVIOUS GEOSCIENCES DCCISIONS MADE IN CONJUNCTION WITH THE SITING AND LICENSING OF THE WOLF CREEK SITE IN COFFEY COUNTY, KANSAS.
FUTURE SITING AF' LICENSING DECISIONS FOR EATERN KANSAS WILL BE EXPEDITED nS A RESULT OF THIS RESEARCH EFFORT.
THE RESULTS OF THE reb?lNAL RESEARCH EFFORT, WHEN COMPLETED. WILL PROVE VALUABLE IN THE DEVELOPMENT OF A TECTONIC PROVINCE / SEISMIC ZONING MAP OF THE EASTERN UNITED STATES.
CONNENTS/PEMaRK3: ALTHOUGH ITS CONCLUSIONS ARE PRELIMINARY (IT CONSTITUTES ONLY A TWO-YEAR SEGMENT OF A FIVE-YEAR PROGRAM INVOLVING SEVERAL STATES). THIS KANSAS RESEARCH REPORT (NUREG/CR-0666) IS A WELL THOUGHT OUT AND EXECUTED EFFORT CONTAINING CONSIDERABLE VALUABLE DATA THAT HAS NOT ONLY INCREASED OUR KNOWLEDGE. ON A BROAD SCALE, OF MID-CONTINENT TECTONICS BUT HAS, MORE DIRECTLY, PROVIDED ADDITIONAL SPECIFIC INFORMATION FOR EASTERN KtNSAS.
THIS NEW INFORMATION HAS NOT CAUSED LICENSING DELAYS.
FUTURE LICENSING ACTION, INVOLVING NOT ONLY WOLF CREEK BUT OTHER NUCLEAR FACILITIFS THAT MAY BE LOCATED WITHIN ~HE REGION SHOULD BE EXPEDITED AS A RESULT OF THESE STUDIES. WE STRONGLY RECOMMEND CONTINUATION OF THE KANSAS RESEARCH PROGRAM.
SEISM 0 LOGICALLY, HOWEVER, THE NRR STAFF QUESTI0HS THE KANSAS GEOLOGICAL SURVEY'S (KGS) RATIONALE FUR THE RELOCATION OF TWO EVENTS (1867 AND 1*06) FEELING THAT SUCH RELOCATION MAY BE BEYOND THE RESOLVING POWER OF THE PRESENT DATA.
ADDITIONALLY, THE ASSIGNMENT OF AN INTENSITY VII-VIII OR VIII TO THE 1867 EARTHQUAKE BY THE KGS ON THE BASIS OF A SINGLE OCCURRENCE OF LIQUEFACTION REPORTED 10 YEARS AFTER THE EARTHQUAKE IS QUESTIONABLE. THE STAFF AGREES WITH THE STANDARD REFERENCES LISTING THIS 1867 EVENT AS AN INTENSITY VII (MM).,
PROG 3AM OFFICE COMMENTS ON POTENTIAL UTILIZATICH OR VALUE OF RESEARCH RESULTS IN THE REGULATORV PROCES3
- It er 67 DATE ISSUED: 11-C6-79 RES DECISION UNIT:
S'4 5TEMS ENGINEERING RIL TITLE PEFLOODING OF SIMULATED PWR CORES
.*T LOW FLOW RATFS S*0NSORING OFFICEfS): NRR (77-rz)
ERS:
t-S REFLOOD HEAT RESEARCH PROJECT MGR:
L. THOMPSON TRANSFER i
RES CO* MENT 1: THIS RIL DESCRIBES THE COOLING OF ELECTRICALLY HEATED RODS DURING BOTTOM FLOODING EXPERIMENT AI CONSTANT INLET FLOODING RATES.
THE INFORMATION PRESENTED IS CONSIDERED APPLICABLE TO THE EVALUATION OF EMERGENCY COOLING SYSTEM PERFORMAhCE IN PRESSURIZED WATER REACTORS.
THE RESULTS PRESENTED IN THE RIL ARE RECOMMENDED FOR CONSIDERATION JN THE APPLICATION AND APPRAISAL OF EVALUATION MODELS FOR REFLOOD HEAT TRANSFER.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS COST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTEg OFFICE RESPONSIELE......... hRR l
SCHECULED CCMPLETION DATE..
ACTUAL CCMPLETION DATE.....
I i
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1
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I epcSasM O F FIC E C0"MEN f 5 ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS I
i PIL s: 63 DaTE ISSUED:
11-11-79 RES DECISION UNIT: PRIMARY SYSTEMS INTEGRITY PIL TITLE: STRUCTURAL INTEGRITY OF WELD REPAIRED PRESSURE VESSELS SPCNS0 PINS OFFICE (S):
NRR RRQ:
1-20 VESSEL INTEGRITY PESEARCH PROJECT MGR:
M. VAGINS Di$ CO*iiNT}:
THIS RIL DESCRISES THE RESULTS OF A TEST PROGRAM TO DETERMINE THE ADEQUACY OF A PROCEDURE WHICH EMPLOYS 1
THE NALF-BLAD WELD REPAIR TECtlNICUE.
THE RESULTS OF THE STUDIES REVEAL THAT THE ASME SECi10N XI WELD REPAIR TECHNIQUE WILL PRCDUCE A REPAIRED STRUCTURE HAVING A SUFFICIENTLY HIGH LEVEi 0F RESIDUAL STRESSES THAT THE STRUCTURE COULD HAVE AN INSUFFICIENT MARGIN OF SAFETY AGAINST FRACTURE. ADDITIONAL RESEARCH IS NEEDED TO PERFECT MODIFICATIONS TO THE PRESENT PROCEDURE 50 THAT RESIDUAL STRESSES ARE MINIMIZED OR ELIMINATED IN THE REPAIRED STRUCTURE. RES NOTES THAT SUCH A PROGRAM j
IS ALREADY UNDERWAY UNDER EPRI SPONSORSHIP; RES IS FOLLOWING THIS WORK.
l USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PEST RIL ACTIVLTLE1 REV113 HELD COMPLETED HEL D HELD ISSUED IMPLEMENTED 0FFICE RESPGNSIBLE........
NRR I
SCHEDULED CCMPLETION DATE..
ACTUAL CGM?LETION DATE.....
NPP ST5EfMTS, DECL53ER **,
'979, t.
SHAD:
E E CIR APP {lGATg N__19_R1WLATOW PF0Q113: THE TESTS DESCRIBED IN THE RIL ACCOMPLISHED THEIR PURPOSE, WHICH WAS TO 1
MEASURE THE FRACTURE BEHAVIOR OF HALF-BLAD WELD REPAIRS IN A PRESSURE VESSEL AT TEMPERATURES NEAR RT/NDT AND ON THE UPPER SHiLF.
SCUND, CRACK-FREE WELDS WERE PRODUCED, WHICH BEHAVED AS WELL AS BASE METAL WHEN THE VESSEL WAS PRESSURIZED AT THE UrPER-SHEL F TEMPERATURE.
HOWEVER. THE TECHNIQUE LEFT AESIDUAL TENSILE STRESSES OF YIELD STRENGTH MAGNITUDE h0RMAL TO THE WELD IN THE MATERIAL ADJACENT TO THE WELD. WHEN A LARGE FLAW WAS PLACED IN THE RESIDUAL STRESS REGION AND j
THE VESSEL WAS PRESSURIZED AT THE RT/NDT TEMPERATURE. THE CRACK POPPED IN ABOUT 2 INCHES AT A PRESSURE OF 0.4 TIMES DESIGN AN" POPPED THROUGH THE WALL AT ABOUT DESIGN PRESSURE. TO EXPLAIN THIS FRACTURE BEHAVIOR. THE EFFECTS OF RESIDUAL j
STRESS MUST BE CONSIDERED ADDITIVE TO THE EFFECTS OF PRESSURE WHEN THE TEMPERATURE IS IN THE TRANSITION REGION.
THE TESTS DESCRIBED IN THE RIL 00 NOT ADDRESS FRACTURE BEHAVIOR WHEN THE UPPER-SHELF ENERGY HAS BEEN REDUCED BY IRRADIATION.
ALTHOUGH SECTION SI 0F THE ASME BSPV CODE PERMITS IT, NO HALF-BEAD WELD REPAIRS OF SIGNIFICANT DEPTH WITHOUT POST-WELD STRESS RELIEF HAVE BEEN PERMITTED IN THE REACTOR C00LANI PRESSURE BOUNDARY.
HOWEVER. PERMISSION TO DO SO WILL PROBABLY j
BE REQUESTED IN THE FUTURE.
WE CONCLUDE THAT HALF-BEAD WELD REPAIR WITHOUT STRESS RELIEF CANNOT BE PERMITTED WITHOUT A FRACTURE ANALYSIS THAT MEETS THE REQUIREMENTS OF SECTION XI SUPPLEMENTED AS FOLLOWS:
A) THE EFFECT OF A POSTULATED FLAW IN THE REGION OF HIGH RESIDUAL STRESS MUST BE EVALUATED.
THE ASSUMED SIZE OF SUCH FLAW MUST BE JUSTIFIED BY CCNSIDERATION OF THE POST WELD NDE PROCEDURES USED AND THE POSSIBILITY OF RECURRENCE OF THE CONDITIONS (HAT CAUSED THE ORIGINAL FLAW (I.E.,
IF THE FLAW WAS SERVICE-INDUCED, ANY MCASURES TAKEN TO REDUCE FATIGUE USAGE OR OTHER CAUSES a
OF CRACKING SHOULD BE CONSIDERED). B) THE EFFECTS OF RESIDUAL STRESS, AS REPORTED IN THIS RIL. MUST BE ADDED TO THE EcFECTS GF PRESSURE AND OTHER LOADS IN CALCULATING K1 FOR THE POSTULATED FLAW.
FRACTURE ANALYSIS AT UPPER-SHELF j
TEMPERATURES WILL REQUIRE CONSIDERATION ON A CASE-BY-CASE BASIS.
I Ef5 C RJ.Bl_J FP A C_T O F PESULTS:
THERE IS A POTENTIAL CONFLICT BETWEEN REGULATORY PRACTICE AND CURRENT PROVISIONS OF THE ASML CODE. SECTION SI. PARAGRAPH IWB-4320.
THE CONCLUSIONS OF THIS RIL WILL BE TRANSMITTED TO THE RESPONSIBLE '0DE BODY l
WITH THE SUGGESTION THAT THE REFERENCED HATERIAL DESCRIBING THE TEST PROGRAM JUST COMPLETED BE CONSIDERED. 04k GOAL WILL i
SE TO OBTAIN AN ADDITIONAL REGUIREMENT IN PARAGRAPH IWB-4320 THAT CALLS FOR THE FRACTURE ANALYSIS DESCRIBED AbOVE.
f95EINT3fPjMAf33: SGM5 EDITORIAL CORRECTIONS AND SUGGESTED CHANGES IN WORDING, WHICH DO NOT AFFECT THE CONCLUSIONS, HAVE a
BEEN FORWARDED TO THE ORIGINATOR OF THE RIL.
l l 1
I P&OODAM OFFICE COMMENTS a4 80TENTIAt UTILIZATION OR VAtOE OF RESEARCH RFluhTS XN THE REGULATORY PROCESS PIL s: 69 DATE ISSUED:
11-19-79 RES DCCISION UNIT: SEISMIC. ENGINEERING & SITE SAFETY DIt TITLE: AN INTEGRATED GEOPHYSICAL AND GEOLOGICAt STUDY OF THE TECTONIC FRAMEWORK OF THE 38TH PARALLEL LINEAMENT IN ihE VICINITY OF ITS INTERSECT;0N WITH THE EXTENSION OF THE NEW MADRID FAULT ZONE SPCMSOPING OFFICE (S):
3-1 NRC/ STATE REGIONAL RESEARCH PROJECT MGR:
N. STEUER EARTH SCIENCES EES CO[*fMT):
THIS RIL IS AN INTERIM PEPORT REFLECTING INFORMATION AVAILABLE AS OF 1978.
THIS INFORMATION RELATES TO A STUDr DESIGNED TO DEFINE THE STRUCTURAL SETTING AND TECTONIC HISTORY OF THE AREA TO REALISTICAt.LY EVALUATE EARTHQUAKE RISKS IN THE SITING OF NUCLEAR POWER PLANTS.
WHILE THESE IhTERIM RESULTS ARE NOT DEFINITIVE, R.S RECOMMENDS THAT THE CUERENT PRACTICE OF EXTENDING THE NEW MADRID 1811-1812 EARTHQUAKES NORTH OF THE ROUGH CREEK FAULT ZONE (38TH PARALLEL LINEAMENT) BE CONTINUED UNTIL ADDITIONAL DATA BEING DEVELOPED INDICATE THAT THIS PRPCTICE SHOULD DE CHANGED. WE ALSO RECO* MEND THAT THE INFORMATION IN NUREG/CR-0449 BE CONSIDERED BY SD AND NRR AS INPUT TO THE DEVELOPMENT OF A TECTONIC PROVINCE CR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVI9E A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA.
DOCUMENT ISSUED:
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS P01T__RIL_A C TJ v_IT_LL$
R F.V_L[y HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR,$D SCHEDULED COMPLETION LATE..
AC*UAL COMPLETI0?4 DATE.....
ER_(p**ENYS, J*NUA_RY 21, 1930, J.
K N ICJi_T :
PLEUEL A'T1[C ALLON TQ_fi[@LA19RY ppCESS: THIS STUDY IS A PART OF THE 'NEW MADRID SEISM 0 TECTONIC STUDY' WHICH IS A CGORDINATED PROGRAM OF CE0 LOGICAL, GEOPHYSICAL, AND SEISMOLOGICAL INVESTIGATIONS OF THE AREA WITHIN A 200 MILE RADIUS CF NEW MALRID, MISSOURI.
THE PURPOSE OF THE RESEARCH IS 10 DEFINE THE STRUCTURAL SETTING AND TECTONIC HISTORY OF THE AREA TO FACILITATE EVALUATION OF EARTHQUAKE RISK IN THE SITING OF HUCLEAR FACILITIES. A COAL OF THIS PROGRAM IS THE
/RODUCTICN OF SEISM 0 TECTONIC AND SEISMIC ZONING MAPS OF THE STUDY AREA.
INTERPRETATION OF THE DATA IS AT A PRELIMINARY STAGF.
THE RESULTS TO DATE HAVE ADDED TO THE GENERAL KNOWLEDGE OF THE STAFF EVEN THOUGH IT IS TOO EAELY IN THE PROGRAM TO MAKE DIRECT APPLICATION TO THE LICENSING PROCESS.
Efif1 TEE _IEPACT Of p[1MLT3: WHILE THE INTERIM RESULTS ARE NOT DEFINITIVE, THEY TEND TO CONFIRM THE CONSERVATISM INHERENT IN INE CURRtNT PRACTICE OF EXTENDING THE NEW MADRID 1811-1812 EARTHQUAKES, NORTH OF THE 38TH PARALLEL LINEAMENT IN LICENSING PROCdDURES.
COMPENTS/DEMARKS: NOME i _.
l i
Pa93pa9 0FFICE CC"MENTS ON POTENTIAL UTILIZATION OR VAtVE OF RESEARCH RESULTS IN THE REGULATORY PROCESS
- It e: 70 PATE I$$UED:
11-19-79 RES DECISION UNITi SEISMIC, ENGINEERING S SITE SAFETY DIL TITLE: SEISMICITY AND TECTONIC RELATIONSHIPS OF THE HEMAHA UPLIFT IN OK, PART II, JAN. 1979 EPONSCPING OFFICE (S):
3-1 NRC/ STATE REGIONAL RESEARCH PROJECT MGR:
N.
STEUER EARTH SCIENCES PES _(Q* MENT}:
THIS RESEARCH IS BEING CONDUCTED TO STUDY THE EARTH SCIENCE PARAMETERS OF THE NEMAHA UPLIFT AND THE MIDCONTINLNT GRAVITY ANOMALY.
THIS RIL FORWARDS THE REPORT WHICH PRESENTS RESULTS OF PHASE II 0F A 3-PHASE REPORT RES RECCMMENDS THAT THE INFORMATICN CONTAINED IN NUREG/CR-0375 BE CONSIDERED BY SD AND NRR AS INPUT TO THE DEVELOPMENT OF A TECTONIC PROVINCE GR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AHD GUIDE FOR ONGOING STUDIES IN THE AREA.
DOCUMENT ISSUED: NUREG/CR-0375 (ALSO SEE RIL 466, 11/16/79 AND NUREG/CR-0666).
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PAS T_P_LL A CT IVI T I F1 RFVIIW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... hRR,SD SCHEDCLED COMPiETION DATE..
ACTUAL COMPLETION DATE.....
1p_CQ"M[NTS, MARCH 19, 1980, R. F_._MJ3t)1:
50 RECOGNIZES IHAT THE NUREG tar 0RIS REPRESENT ONLY A PORTION OF A FIVE YEAR PROGRAM.
NONETHELESS. THEY SHOULD PROVIDE USEFUL INPUT FOR NRC STANDARDS DEVELOPMENT PARTICULARLY IN REGARD TO A POSSIBLE REVISION OF APPENDIX A IO CFR PART 100, AND THE DEVELOPMENT OF TECTONIC PROVINCE OR SEISMIC ZONING MAPS OF THE EASTERN UNITED STATES.
tLPR COP"*ENTS, DECLMajR 12, 1979, J.
KNIGHT:
dim 81fl1_W L I C A D kN_.LOJLELL A T O RY PROXEll: SITING IN OKLAHOMA, KANSAS AND NEBRASKA HAS BEEN MADE MORE DIFFICULT BY THE CONTROVERSY OVER THE ASSOCIATION OF SEISMICITY WITH THE HEMAHA RIDGE. WITH EXCEPTION OF NEW EARTHQUAKE DATA, THE DATA DESCRIBED IN THE REPORT IS NOT DIRECTLY APPLICABLE TO THE REGULATORY PROCESS.
PHASE III SHOULD PROVIDE THE USFABLE DATA.
plS1ELLE IMPACT _OF RESULT 5:
THE PROJECT HAS NOT BEEN COMPLETED, NOR FINAL CONCLUSIONS DRAWN.
THE GEOLOGIC AND SEISMIC DATA BASE HAVE BELN BROADENED. PRELIMINARY SEISMIC DATA SUGGESTS A ZONE OF EARTHQUAKES BEGINNING NEAR EL RENO STRIKING NORTHEAST AND CUTTING DIAGONALLY ACROSS THE NEMAHA STRUCTURE, A CONCENTRATION NEAR WILSON AND SEVERAL EPICENTERS PARALLEL THE NORTHERN FRONT OF OUACHITAS.
AL50. THE EARTHQUAKE TO DATE SUPPORTS STAFF CONCLUSION THAT THE LEVEL OF SEISMICITY IS LOW FOR NORTHEAST OKLAHOMA.
THE LARGEST MAGNITUDE RECORDED WITHYN 10-20 KM OF THE BLACK FOX SITE IS 1.4 THE LCW LEVEL CF SEISMICITY ALSO APPLIES TO WESTERN OKLAHOMA AND POSSIBLY THE PANHANDLE. WHEN COMPLETE, THE GEOLOGIC AND SEISMIC DATA ARE IN, HOPEFULLY, THF STAFF WILL BE ABLE TO CONCLUSIVELY DETERMINE IF THE NEMAHA UPLIFT IS A TECTONIC STRUCTURE ANC RELATE THE SEISMICI)Y TO THE CAUSITIVE STRUCTURE, THEREBY, MORE CONCLUSIVELY DEFINING THE SSE FOR NEW PLANTS IN !HE AREA.
CCMMENTS/PE"SPKS:
NONE __
i PROGRAM OFFICE CC"MENTS ON POTENTIAL UIILIZATION OR VALUE OF RESEARCH RESULTS IN VHE REGULATORY PROCESS l
ell _s : 71 DATE ISSUED:
11-19-79 RES DECISTON UNIT: SEISMIC, EKGINEERING & SITE SAFETY PIL TTTLE: REGIONAL TECTONICS AND SEISMICITY OF EASTERN NEBRASKA ANN;..a F1? ORT JUNE 1 1977-MAY 30, 1973 SPONSOPING OFFICE (S):
3-1 HRC/ STATE REGIONAL RESEARCH PROJECT MGR:
H.
STEUER EARTH SCIENCES fl5 CC"51NTS: THE PURPOSE OF THIS RESEARCH IS TO STUDY THE EARTH SCIENCE PARAMETERS OF THE NEMAHA UPLIFT AND THE MIECONTINLNT GRAVITY ANOMALY.
KNOWLEDGE OF THESE GEOLOGICAL STRUCTURES IS NEEDED TO DETERMINE WHETHER OR NOT THEY ARE LCCALIZERS OF EARTHQUAKES.
'.HE INFORMATIOC CAINED IS Or IMPORTANCE IN THE SITING AND LICENSING OF NUCLEAR POWER PLANTS.
THIS INTERIM REPORT PRESENTS AND INTERPRETS INFORMATION OBTAINED BETWEEN JUNE 1 1977 TO MAY 30, 1973.
RES RECCMMEh0S THAT THE INFORMATION CONTAINED IN NUREG/CR-067s BE CONSIDERED BY SD AND NRR AS INPUT TO THE DEVELOPMENT CF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN O.5. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EQST PIL ACTIVITIES REVlfW HEL D COMPLETEu HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR.SD SCHEDULED COMPLETION DATE..
I ACTUAL CCMPLETION DATE.....
\\
M_GE""LN T S, M A P C H tb 1930nR. MINE VE:
59 RECOGNIZES THAT THE NUREG REPORIS REPRESENT ONLY A PORTION OF A FIVE YEAR PROGRAM. NONETHELESS. THEY SHOULD l
PROVIDE USEFUL INPUT FOR NRC STANDARDS DEVELOPMENT PARTICULARLY IN REGARD TO A POSSIBLE REVISION OF APPENDIX A TO CFR PART 100. AND THE DEVELOPMENT OF TECTONIC PROVINCE OR SEISMIC ZONING MAPS OF The EASTERN UNITED STATES.
2 Now (f'* MEN TSJEH _1?c 15 0, J.
KNIGHT:
1 pli_C RJ B E A P P L I C A ' LON_T O R (MM A T Q P Y PROCE13: THE PURPOSE OF THIS RESEARCH IN THE CENTRAL STABLE REGION IS AN ATTEMPT TO CETERMINE THE REASchS FCR. AND SOURCE OF, LARGER EARTHQUAKES IN THE E4 STERN UNITED STATES.
THIS NEBRASKA' EFFORT CONSTITUTES A PORTION OF A COOPERATIVE GEOLOGIC. SEISMIC, AND GEOPHYSICAL EFFORT OF THE STATE GEOLOGICAL SURVEYS OF OKLAHGMA, KANSAS. NEBRASKA, IOWA. AND MINNESOTA TO STUDY THE EARTH SCIENCE PARAMETERS OF THE NEMAHA UFLIFT AND i
THE MIDCONTINENT GRAVITY ANOMALY.
THIS REPORT (NUREG/CR-0376) PRIMARILY ADDRESSES THE GEOLOGY, STRUCTURE. TECTONICS, J
AND SEISMICITY OF SOUTHEASTERN NEBRASKA IN THE VICINITY OF THE HUMBOLDT FAULT ZONE (A GEOLOGIC STRUCTURE HEAR THE NEMAHA UPLIFT).
THIS NEBRASKA RESEARCH PROGRAM, IN CONJUNCTION WITH OTHER REGIONAL EFFORTS, IS TO SERVE AS AN AID TO d
THE NRC IN DEVFLOPING A RATIONAL EVALUATION OF EARTHQUAKE RISK AS IT APPLIES TO THE SITING, DESIGN AND REEVALUATION OF l
NUCLEAR FACILITIES LOCATED WITHIN THE STUDY REGION.
1 EE1ERIJf_JMPACT OF PESy(TS: THIS RESEARCH EFFORT HAS INCREASED OUR KNOWLEDGE OF THE GEOLOGY, STRUCTURE. TECTONICS AND SEISMICITY OF EASTERN NEBRASKA (VICINITY OF THE NEMAHA UPLIFT), EUT HAS NO DIRECT IMPACT ON LICENSING.
C0"MENTS/REMAPK$:
NUREG/CR-0376 CONSTITUTES CHLY A PORTION OF A MAJOR REGIONAL EFFORT DIRECTED TOWARD DETERMINATION u
0F THE SOURCES OF SEISMICITY IN THE MIDCONTINENT AREA IN THE VICINITY OF THE MIDCONTINENT GRAVITY ANOMALY AND THE i
d NEMAHA UPLIFT.
THIS NUREG. COMBINED WITH ThE OTHER RESEARCH REPORTS, HAS THE POTENTIAL FOR HELPING TO RESOLVE ONE i
0F THE MAJOR QUESTIONS PERPETUALLY CONFRONTING THE NUCLEAR REGULATORY COMMISSION WHEN ASSESSING NUCLEAR POWER PLANT APPLICATIONS IN THE M.DCGNTINENT AREA - NAMELY, WHAT IS THE SOURCC (CR MOST LIKELY SOURCE) 0F SEISMICITY OCCURRING IN THE VICINITY OF 1HE NEMAHA UPLIFT AND MIDCONTINENT GRAVITY ANOMALY.
SINCE THIS REGIONAL STUDY IS QUITE COMPREHENSIVE, PERIPHERAL STUDIES. APPARENTLY NOT DIRECTLY-RELATED TO EITHER THE NEMAHA UPLIFT OR THE MIDCONTINENT GRAVITY ANOMALY.
1 J
HAVE BEEN CONDUCTED.
IN THE CASE OF THIS NUREG, THE REPORT ENTITLED ' RELATION OF EARTHQUAKE EPICENTERS TO GLACIATION,'
IS SOMEWHAT PERIPHERAL TO THE MAIN PURPOSE OF THE OVERALL RESEARCH EFFORT BUT DOES ADDRESS A GENERIC ISSUE OF SOME j
RELEVANCE.
1 l
_ ~
I l
P&0028M OFFICE COMMENTS ON POTENTIAL UTILIZATION OV VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS PIL s: 72 rate ISSUED:
11-16-79 RES DECISION UNIT: SEISMIC. ENGINEERING S SITE SAFETY i
PIL TITLE NEW ENGLAND SEISM 0 TECTONIC STUDY ACTIVITIES DURING FISCAL YEARS 1977 AND 1978 SPONSO9 INS OFFICE (S):
3-1 NRC/ STATE REGIONAL RESEARCH PROJECT MGR*
H. STEUER EARTH SCIENCES i
P11_[fFMENTS:
THE N.'M ENGLAND SEISM 0 TECTONIC STUDY IS A 5-YEAR PROGRAM TO STUDY THE GEOLOGY AND SEISMICITY OF NEW ENGLAND AND CONTIGUQUS AREAS TO ASSESS THE POTENTIAL SEISMIC HAZARD TO PROSPECTIVE NUCLEAR POWER PLANT SITES IN THE REGION.
j PPELIMINARY RESULTS THUS FAR DOCUMENT THE PRGNINENCE OF FAULTING IN THE REGION AND DEMONSTRATE THE EFFECTIVENESS OF REMOTE-SENSING. MAGNETIC-LINEAMENT AN' GRAVITY-LINEAMENT ANALYSES TO REVEAL FAULTS IN THE REGION.
THE REPORTS FORWARDED WITH RIL DESCRIBE THE STUDY FRCM JULY 1 1977 TO JUNE 30, 1973.
RES RECCMNENDS THAT THE INFOPMATION CONTAINED IN NUREG/CR-0031 AND NUREC/CR-0930 BE CONSIDERED BY SD AMD NRR AS INPUT TO THE DLVELOPMENT OF A TECTCNIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA.
1 DOCUMENTS ISSUED: NUREG/CR-0051; NUREG/CR-0930.
4 USER DISCUSSION POSITION COMMISSION ACES PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST P.LL_ KT11LilEl 6JV_U9 HELD COMPLETED HELD HEL D ISSUED IMPLEMENTED OF F ICE R LSF ONSIB L E......... MRR.SD SCHEDULED COMPLFTION DATE..
]
ACTUAL COMPLETILN DATE.
I E3R._QCE RNT} d ANtycY ltu _ M,
J.
KNIGHT) t u
1 RiiEEIJE APPt IC Ai1[N !U (5QULATORY PROSE: THE STUDY IS A COOPERATIVE EFFORT AMONG FEDERAL AGENCIES, UNIVERSITIES, AND SIATE GEOLOGICAL SURVEYS IN A PROGRAM TO STUDY THE GEOLOGY AND SEISMOLOGY OF NEW ENGLAND, WITH THE INTENT TO PROVIt'E A STRONG BASIS FOR THE ASSESSMENT OF THE SEISMIC HAZARD IN THE REGION IN ACCORDANCE WITH NRC REGULATIONS. THE SUBJECT NUREG'S ARE A
SUMMARY
OF THE RESULTS OF THE FIRST 2 YEARS ACTIVITY OF A S YEAR PROGRAM. THE RESULTS HAVE ADDED GREATLY TO THE CENERAL KNOWLEDSE OF THE STAFF. QESSRI.Pl_JMP A C T OE_PE1ULLS: THE RESULTS HAVE THUS FAR HAD !!3 DIRECT IMPACT ON LICENSING DECISIONS. THEY TEND. HCWLVER. TO CONFIRM PAST CONCLUSIONS THAT CURRENT SEISMICITY, DETECTED BY INSTRUMENTS. IS CONCENTRATED IN THE SAME GENERAL AREAS AS HISTORIC EARTHQUAKES RECORDED DURING THE PAST SE'!ERAL HUNDRED YEARS; add THERE IS NO DEMONSTRAT A3'.E f CCRRELATION BETWEEN SEISMICITY AND MAPPED FAULTS. CIFliMIlff EElfi$ : AN ADDITIONAL BENEFIT OF THIS PROGRAM IS THAT IT HAS DEPPMSTRATED THE EFFECTIVENESS OF IDENTIFYING FAULIS USING REMOTE SENSING TECHNIQUES. IT IS RECOMMENDED THAT THE INFOAMATIV4 CONTAINED IN NUREG/CR-0939 AND j NUREG/CR-3031 BE USED AS INPUT TO THE CONSTRUCTICN OF A SEISM 0 TECTONIC MAP AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING i STUDIES. I i I i f i
l PPOGDtM OFFICE COMMENTS ON POTEMTIAL UTILIZATION OP vt.t 9E OF RESEARCH RESRLTS XN VHE RFAULATORV PROCES3 pit e: 73 DATE TSSUED: 11-16-79 RES DECISION UNIT: FUEL CYCLE SAFE!Y & ENVIRONMENTAE EFFECTS PIL TIfLE: IN VIVO COUNTING AT SELECTED URANIUM MILLS SPON50 PING OFFICh(S): SD (79-5) RPS: 5-23 OCCUPATIONAL RESEARCH PROJECT MGR: J. FOULKE EXPOSURE & PROTECTION 911_ C qrrE5 TS
- THIS RESEARCH PROJECT PROVIDED MEASUREMENTS OF THE INTERNAL DEPOSITION OF URANIUM IN THE LUNGS AND RADIUM IN THE $FELETCN OF URANIUM MILL WORKERS.
THE IN VIRO COUNTING WAS CONDUCTED AT THE HINE MILL SITES. NO WORKER HAD MORE THAN THE MAXIMUM PERMISSIBLE ORGAN BURDEN. RES RECOMMENOS THAT RESULTS PRETENTED IN NUREG/CR-0341 BE USED BY SD IN DETERMINING THE VALUE OF TAKING IN VIRO NEASUREMENTS AT MILL SITES. DOCUMENT ISSUED: NUREC/CR-C341 USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS Ep5 T_ PJi_ Af,II V I T I ES R E V I E'4 HELD COMPLETED HEL D HELD ISSUED. IMPLEMENTED GFFICE RESPONSIBLE......... $D SCHECULED COMPLETION DATE.. ACTUAL CCMPLETION DATE..... Sp__Crm NTS, M A Y_U, 1930, S. MCGUJKl: p t i_C H L&E A P P LLC A T. "" 10 PLGU1ATOR( PPOCJSS: THE RESEARCH PERFORMED ON IN VIVO COUNTING OF URANIUM MILL WORKERS BEARS DIRECILY ON BIDASSAY REGUIREMENTS AI UkANIUM MILLS. NRC'S POLICY ON BI0 ASSAY AT URANIUM MILLS IS CONTAINED IN REGULATORY GUIDE 8.22. 'BIDASSAY AT URANIUM MILLS." EFj{yJfE IFPACT OF P($ULTS: THE RESEARCH DEMONSTPATED THAT URANIUM CONCENTRATIONS IN THE LUNGS OF URANIUM HILL WORKERS WERE NOT HIGHER THAN EXPECTED. THE RESEARCH ALSO DEMONSTRATED THAT COMMERCI ALLY AVAIL ABLE IN VIVO COUNTING FOR MATURAL URANIUM USING MOBILE TRUCK MOUNTED COUNTERS IS NOT SUFFICIENTLY SENSITIVE TO DETECT URANIUM IN MILL WORKERS. SUCH IN VIVO COUNTING IS NOT A USEFUL TOOK AT URANIUM MILLS. CCP"ENTS/PEMaPrS: THE RESEARCH PERFORMED WAS QUITE USEFUL. - -
a Pa0GeAM GF6fCE CC**ENTS ON POTENTIat orILIZAi!ON CR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS DIL 8: 74 PtTE ISSUED: 11-16-79 RES DECISION UNIT: CODE DEVELOPMENT DIt TITLE: STEADY-STATE FUEL RCD BEHAVIOR CODE: FRAPCON-1 SPONSOPING OFCICEfS): hRR, SD Egg: 1-12 FUEL CODE RESEARCH PROJECT MGR: G. MARINO DEVELOPMENT RES_gC--ENTS: THIS RIL TRANSMITS THE DESCRIPTION AND ASSESSMENT DOCUMENTATION OF THE LATEST VERSION OF THE STEADY-STATE FULL kOD BEHAVIOR CODE - FRAPCON-1. FRAPCON-1 IS A FORTRAN IV COMPUTER MODEL WHICH CONSIDERS THE COUPLED EFFECTS OF FUEL AND CLADDING DEFORMATION TEMPERATURE. AND INTERNAL GAS PRESSURE ON THE OVERALL RESPONSE CHARACTERISTICS OF A FUEL ROD OPERATING UNDER NORMAL CONDITIONS. THE CODE IS USED: 1) AS A BE CODE TO INITIALIZE THE CURRENT RES BEST ESTIMATE TRANSIENT CODE; 23 AS A STAND-ALONE, BEST ESTIMATE STEADY-STATE CODE; OR 3) AS A LICENSING TOOL WITH APPROPRIATE EM MODELS SUPFLIED BY NRR. USER DISCUSSION FOSITION CGMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST PI! AqlJyT11El bey 11N HEL D COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE kESPONSIBLE......... hRR.SD SCHEDULED CC"PLETION DATE.. ACTUAL COMPLETION DATE..... L.
l Pv0G&aM OFFICE CCPMEmiS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULAT04Y PROCESS RIL 1: 75 DATE ISSUED: 11-27-79 RES DECISION UNIT: SEISMIC. ENGINEERING & SITE SAFETY pit TITLES INVENTORY, DETECTION, AND CAT ALOG OF OKL AHOMA EARTHQ' JAKES AND EARTHQUAKE MAP OF OKL AHOMA, MAP GM-19 SPONSOPING OFFICE (5): NRR, SD RED: 3-1 NRC/ STATE REGIONAL RESEARCH PROJECT MGR: N. STEUER EARTH SCIENCES PL1_CCTEENT): THIS RESEARCH WAS A COOPERATIVE GEOLOGIC, SEISMIC AND GEOPHvtYcAL EFFORT TO STUDY THE EARTH SCIENCE PARAMETERS OF THE NEMAHA UPLIFT AND THE MIDCONTINENT GRAVITY ANOMALY. A KNOWLt?GE OF THESE GEOLOGIC STRUCTURES IS OF VITAL IMPORTANCE IN THE SITING AND LICENSING OF NUCLEAR PCWER PLANTS. FROJECT WDRK IS SEPARATED INTO THREE PHASES. THIS RIL WITH ITS ENCLOSURES PRESENTS RESULTS OF WORK COMPLETED IN PHASE I. RES RECCNMENDS THAT THE INFORMATION IN THE INVENTORY BE CONSIDERED BY SD AND NRR AS INPUT TO THE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES. DOCUMENTS ISSUED: MAP GM-19; REPORT: INVENTORY, DETECTICH AND CATALOG OF OKLAHOMA EARTHQUAKES. USER DISCUSSION POSITION COMMISSIGH ACF.; PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PpST PIL ACTIVITIES PEVIIM HELD COMPLETED HELD HELD ISSUED IMPLEMENTED urFICE RESPCNSIBLt......... NRR.SD -s SCHEDULED CCMPLETION DATE.. ACTUAL COMPLETIGN DATE..... 19.COMNE5IS, MAPCH 19, 1980, R, M{NQgul: SD RECOGNIZES THAT THE NUREG REPORIS REPRESENT ONLY A PORTIDH OF A FIVE YEAR PROGRAM. NONETHELESS. THEY SHOULD PROVIDE OS_FUL INPUT FOR NRC STANDARDS DEVELOPMENT PARTICULARLY IN REGARD TO A POSSIBLE REVISION OF APPENDIX A TO LFR PART 100, AND THE DEVELOPMENT OF TECTONIC PROVINCE OR SEISMIC ZON NG MAPS OF THE EASTERN UNITED STATES. NDP CI TENT $, JANUARY 28, 193AL, M N (QH T : pE1 cpi 1E APPLIGAR ON TO_P1GulAIORY PROCLSJ: THIS RESEARCH IS PART OF A COOPERATIVE GEOLOGIC, SEISMIC AND GEOPHYSICAL Eff0RI 0F THE STATE GEOLOGICAL SURVEYS OF OKLAHOMA, KANSAS, NEBRASKA, IDWA, AND MINNESOTA TO STUDY THE EARTH SCIENCE PARAMETERS OF THE MIDCONTINENT GRAVITY ANOMALY. THE OVERALL GOAL OF THIS PROGRAM IS TO ESTABLISH A STRONG BASIS FOR DETERMINING THE SEISMIC RISK FOR NUCLEAR POWER FT.CILITTES IN THESE SECTIONS OF THE CENTRAL STABLE REGION. IT IS TOO EARLY IN THE PROGRAM FOR THE DATA TO HAVE DIRECT APPLICATI0H TO THE LICENSING PROCESS. HOWEVER, WHEN ALL INFORMATION HAS BEEN INTERPRETED AND SYNTHESIZED, IT WILL BE USED TO DEVELOP A SEISM 0 TECTONIC PROVINCE OR SEISMIC ZONING MAP. Q E5_CSIRf_JFP A C T OF RE.SyLT$: THE RESULTS OF THIS RESEARCH HAVE NOT HAD A DIRECT IMPACT ON LICENSING ACTIVITIES. HGWEVER, THEY HAVE ADDED 10 OUR GENERAL KNOWLEDGE SEGARDING THE TECTONIC SETTING OF THE REGION IN WHICH OKLAHOMA LIES, AND HAVE PROVIDED A COMPLETE CATALOGUE AND EARTHQUAKE MAP OF OKLAHOMA. (Q[NINTS/PEMARK5: IT IS RECOMMENDED THAT THE DATA IN ' INVENTORY, DETECTION AND CATALOG CF OKLAHOMA EARTHQUAKES AND LARTHCUAKES MAP GM-19' BE CONSIDERED BY THE OFFICE OF STANDARDS DEVELOPMENT AND THE OFFICE OF NUCLEAR REACTO AND TO REGULATION AS INPUT TO THE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. PMOVIDE A BASIS AND CUIDE FOR ONGOING STUDIES IN THE AREA. f
~~'Pe0GoaM OF F IC E CCP"EN T S ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS i pit s: 76 DaTE ISSUED: 12-28-79 RES DECISION UNIT: PRIMARY SYSTEMS INTEGRITY PIL TITLE: ANNEALING OF IRRADIATED REACTOR PRESSURE VESSELS SPONSOPING OFFICE (S): NRR, SD ESS: 1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR: C. SERPAN l RECL-E N_TJ : THE RESEARCH PROJECT REVIEWED THE INFORMATION DEVELOPED DVER THE PAST 15 YEARS ON THE USE OF POST-IRRADIAIION HEAT TREATMENT (ANNEALING) TO RECOVER THE PRE-IRRADIATION PROPECTIES OF REACTOR VESSELS OF COMMERCIAL NUCLEAR POLER PLANTS FOR COMMERCIAL SAFE OPERATION. THE INFORMATION IS TO BE VIEWED AS PROVIDING A BACKGROUND FOR INTERPRETING l CURRENT AND FUTURE RESEARCH ACTIVITIES AND POTENTIAL LICENSING APPLICATIONS OF VESSEL STEEL ANNEALING. DOCUMENT ISSUED: NUREG/CR-0436. USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS P_01T_ R I L ACTIVITH S ELVL(H HELD COMPLETED HELD HELD ISSUED IMPLEMENTED GFFICE RESPCNSIBLE......... NRR,SD SCHECULED CCMPLETION DATE.. ACTUAL COMPLETION DATE..... SD CH'1"1575, r* A y 19, tog P. PgAg: AS AN INTERIM PROGRESS REPORT ON THE FEASIBILITY STUDIES SPONSORED BY THE NRC AND OTHERS THE RIL IS A VERY SATISFACTORY REPORT. CONCLU;.ICMS AND RECOMMENDATIONS HAVE NOT YET BEEN MADE, HENCE NO COMMENTS REGARDING APPLICATION OF THE FINDING,5 ARE IN ORDER AT THIS TIME. a l 1 I ( ) 1 l
PbOGa*M CFFICE C0tNENi$ OM 90iENTIAL UTILIZATION OR VAtCE OF RESEAPCH RESULTS IN THE GEGULATORY PROCESS PIL s: 77 pATf TSSUED: 12-23-79 PES DECISION UNIT: SEISMCI. ENGINEERING 8 SITE SAFETY PIL TITLE: ORIGIN OF SURFACE LINEAMENTS IN NEMAHA COUNTY, KAN5AS SPONSSPING OFFICEfS): NRR, SD ERG: 3-1 NRC/ STATE REGIONAL RESEARCH PROJECT MGR: N. STEUER EARTH SCIENCES fEi_Cp7prNTS THE PURPOSE OF THIS RESEARCH WAS TO GAIN A BETTER UNDERSTANDING OF THE SOURCES OF EARTHQUAKES THAT HAVE OCCUkRED AS AN AID TO DEVELOPING A MORE RATIONAL EVALUATION OF EARTHQUAKE RISK AS IT APPLIES TO THE SITING AND DESIGN OF NUCLEAR FACILITIES. RES RECCMMENDS THAT INFORMATICN IN NUREG/CR-ct21 BE CONSIDERED BY SD AND NRR AS INPUT TO THE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA. DOCUMENT IS$dED: HUREG/CR-C321. USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEEITNG PAPER BRIEFING BRIEFING RELEASE RESULTS E0ll Ell _A(11EIIII$ SillfW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE PL5PONSIBLL......... NRR,SD SCHEDULED CCMPLETION DATE.. ACTUAL CGMPLETION DATE.... $f_C M MT_S,JiPCH ti, 140dmMINOCUf: SD R ECGG'4IZLS THAT IHt huREG REPORIS REPRESENT ONLY A PORTION OF A FIVE YEAR PROGRAM. NONETHELESS, THEY SHOULD PROVIDE USEFUL INPUT FOR NRC STANDARDS DEVELOPMENT PARTICULARLY IN REGARD TO A POSSIBLE REVISION OF APPENDIX A TO CFR PART 1C0. AND THE DEVELOPMENT OF TECTONIC PROVINCE OR SEISMIC ZONING MAPS OF THE EASTERN UNITED STATES. NEP CO[NT % A P P [LJ, 19?%f) F N I C.M : PilEEJSt APPLICATION t_0_Mf3QL112EY P 90(f _SS : THE STATE GEOLOGICAL SURVEYS OF KANSAS. OKLAHOMA, NEBRASKA, AND IOWA ARE CCNDUCTING A S-YLAR GE4 LOGICAL, SEISMOLOGICAL AND GEOPHYSICAL STUDY OF THE REGIONAL 1ECTONICS AND SEISMICITY OF THE NEMAHA UPLIFT AND THE MIDCONT.NENT CRAVITY ANOMALY AND OTHER STRUCTURES IN THE REGICH. THE GOAL OF THIS STUDY IS TO ESTABLISH A BETTER UNDERSTANDING OF EARTHQUAKE SOURCE MECHANISMS TO DETERMINE THE SEISMIC RISK FOR NUCLEAR POMER FACILITIES IN THIS PART OF THE CENTRAL STABLE REGION. THE DATA DERIVED FROM THE STUDY WILL PROVIDE DATA IN THE DEVELOPMENT OF A SEISMJTECTONIC PROVINCE MAP. EESCPf3E !PPACT or RFSQtT3: THE RESULTS OF THIS STUDY HAVE NOT HAD A DIRECT IMPACT ON NUCLEAR POWER PLANT LICENSING A C I I 'v. I I E S. THEY ICENIIFY THE LINEAMENTS IN NEMAHA COUNTY AND APPEAR TO DEMONSTRATE THAT THE LINEAMENTS ARE NOT RELATED TU THE !EISMICITY IN THE AREA, WHICH ADDS TO CUR CENERAL KNOWLEDGE OF THE GEOL OGIC TOOLS TI.AT MAY BE EMPLOYED IN ESTABLISHING THE TECTONIC SETTING OF THE CENTRAL STABLE REGION. C0"ME%Y5/ PEP':EES: THE STUDY OF THE ' ORIGIN OF SURFACE LINEAMENTS IN NEMAHA COUNTY, KANSAS' CONSTITUTES ONLY A PERIPHERAL SIUDY AND DOES NOT DIRECTLY ADDRESS THE CUESTION OF EARTHCUAKE SOURCE MECHANISMS WHICH IS IMPORTANT 70 AN UNDERSTANDING OF THE SEISMIC RISK IN THE REGION. THE STUDY IS PERIPHERAL TO THE MAIN PURPOSE OF THE OVERALL RESEARCH EFFORT BUT DOES ADDRESS A GENERIC ISSUE OF SOME RELEVANCE. IT IS RECOMMENDED THAT THE DATA CONTAINED IN NUREG/CR-0521 3E USED FOR Or. 0ING STUDIES IN THE AREA WHICH WILL EVENTUALLY LEAD TO THE DEVELOPMENT OF A SE!SMOTECTCNIC PROVINCE MAP AND A BETTER UNDLRSTANDING OF THE SEISMIC RISK FOR THE SITING OF NUCLEAR FACILITIES. - 100 -
PROSJAM OFFICE CCP9ENf 5 ON POTENTIAL U11LIZ4 TION OR VALUE OF RESEARCH RESULTS IN T':E REGUL410RY PROCESS RIL 8 78 DaTE TSSUED: 12-23-79 EES DECTSTON UNIT: CODE DEVELOPPENT pit TI LLg: VERTICAL LOADS IN MARK I CONTAINMENT TORUS SPONSODING CFFICE(S): hPR (76-13) EES: RESEARCH PROJECT MGRt R. CUDLIN Pf3_Sp 1E$T3: THE RESEARCH WAS PERFORMED TO QUANTITATIVELY EVALUATE THE HYDRODYNAMIC LOAD IN THE MARK I TYPE CGNTAINMENT IN i;tt EVENT CF A LOCA. THE TESTS CCNDUCTED PROVIDE CONFIRMATION OF THE AESOLUTE VALUES AND SENSITIVITIES OF THE AIR VENTING LOADS MEASURED IN OTHER EXPERIMENTS. THE TESTS ALSO PROVIDE EVIDENCE OF A 3-DIMENSIONAL CHARACTER OF THE UPLOAD h0T 3EFORE RECOGNIZED. THE DATA IN THE SCALING LAW EXPERIMENT VERIFIED TNAT SMALL SCALE TEST DATA CAN BE SCALED UP TO PROTOTYPICAL PLANT SIZE. THE t/5 SCALE TORUS TEST RESULTS SUPPLEMENTED BY THE VERIFIED SCALING LAW PROVIDE AN APPROPRIATL YARDSTICK FOR LICENSING ASSESSMENT OF THE MARK I CONTAINMENT CONCERNING AIR VENTING LOADS. USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST Ell _ACTlVJTlf3 ?(VICW Hit D CtCPLETED HEL D HEtD ISSUED IMPLEMENT (p 0FFICE RESPONSIBLE......... Ns* SCHEDULED CCMPLETICN DATE.. ACTUAL CCMPLETION DATE....
- LE_SfCiNlb l_A'CH 87
'957 C. OPTEES: pfMju % t A P P_Uf a i ! ON 10 v'M A ig3Y FOCf15: TEST RESULTS FROM THE LLL 1/5 SCALE MARK I POOL SWELL TESTS AND THE MIT POOL SWELL SCALING STUDIc5 WERE CGh?ARLD TO SIMILAR TESTS AND STUDIES PERFORMED BY THE MARK I OWNERS G20UPS. RES TEST f DATA WERE USED TO ICENTIFY THE SIGNIFICANT PARAMETERS AND TRENDS FOR THE PRESSURE LOADS ON THE TORUS. PESI?IPE IM' eel _EE_?[?JLTS: THE MIT SCALING STUDIES GENERALLY CONFIRMED THE SCALING RELATIONSHIPS USED FOR BOTH INE LLL A N D M A e.e. I OwiLRS CROUP PCOL SWELL TESTS, AND IDENTIFIED THE SIGNIFICANT PARAMETERS THAT AFFECT SCALING. THE LLL TEST RESULTS IDENTIF.ED AN UNCERTAINTY IN THE NET UPWARD LOAD ON THF TORUS BETWEEN THE 2D AND 3D SECTORS. BUT CONFIRMED THE NET DCWNWifD LPADS AND THE 'ASIC PHENOMENA ASSOCIATED WITH POOL SWELL. THE LLL TEST DATA WERE USED TO QUANTIFY A MAPGIN FOR THE UACERTAINTY IN THE UPWARD TORUS LDADS. COE"13TS/PFM4 PES: hRR AND ITS CONSULTANTS DID NOT UNILATERALLY AGREE WITH THE CONCLUSIONS DRAWN BY LLL. CONSEQUENTLY. THL LLL IEST DATA WERE ADJUSTED IN ORDER TO DEVELOP THE APPROPRIATE ttARGIK FOR THE NET UPWARD LOADS ON THE TORUS. - 101 -
-_m__ .m_.- _____.-m-r20casM OSFICE CernEstS ON POTENTIAt UTILIZATION 0* VALUE OF QESEARCH RESULTS IN THE REGut4 TORY PROCESS pit.*: 79 P4TE TSSUED: 12-23-79 RES DECISION UNIT: SEISMIC. ENGINEERING & SITE SAFLTY I RTL TITLEr EVALUATICN CF SEISMIC QUALIFICATION TESTS FOR NUCLEAR PCWER PLANT. EQUIPMENT l SPnN5CPINS OFFICE (S): NRR PR2: RESEARCi, PROJECT MGR: C. BROWZIN h RE}_CQT5I3T$: THE OBJECTIVES OF THIS TEST WERE TO SUBJECT A TYPICAL ELECTRICAL CABINET S'ECIMEN TO A SERIES OF DIFFERENT CLMRtNILY ACCEPTABLE SEISMIC CUALIFICATION TESTS. TO ACQUIRE THEREFROM DYNAMIC RESPCNSE DATA AND TO PROVIDE A BASIS FOR CGMPARISCN OF THE TESTS' EFFECTIVENESS. R E' CONCLUDES THAT A REASSESSMENT OF THE TEST RESPDNSE SPECTRUM (TRS) ENVELOPING ] A REOUIRED RESPONSE SPECTRUM (RPS) NEEDS 10 sE AUGMENTED TO ASSURE PROPER DISTRIBUTION OF ENERGY WITH FREQUENCY DURING A i CUALIFICATION TEST. RESONANCE SEARCHES SHOULD BE CONDUCTED FOR BOTH SIMULATOR-MOUNTED AND FLOOR MOUNTED CONFIGURATIONS i FGR ITEMS WWERE Dfh'MIC CCUPLING WITH THE SIMUL ATOR TU3E IS EXPECTED. A NEW PARAMETER. THE DAMAGE SEVENTY FACTOR. HAS BEEN DEVELOPED FOR COMPARING SEVENTY OF SEISMIC QUALIFICATION TESTS. DSF MAf MARE IT PDSSISLE TO UPGRADE EQUIPMENT SUPPORT TO HIGHER SEISMIC EXCITATION. DCCUMENT ISSUED: NUREG/ CR-0 34 5. 1 UAER DISCUSSION POSITION COMMISSION ACRS PRESS OFriCE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PPS LRlL A C_U VI UM .W HELD Cf?PLETED HELD NELD ISSUED IMPLEMENTtQ OFFICE RESPCNSIBLE......... SCFEDULED CCMPLETICM DATE. ACTUAL COMPLETION DATE. 2' _COMllt F_ff M !EY U T-MllTSON 4 D. QSEld QT: CL1M '1_.A MLMS 31c n to.. _ (o mCM} : THE RLSEARCH RESULTS HAVE: A. FA0VIDED AN INDEPENDENT CuMFIkMATICN OF THE FOLLOWING LICEN!ING POSITIONS: (1) THE SINGLE FREQUENCY TEST INPUT MAY BE SEVERE e0R '.?RIFYING THE STRUCTURAL INTEGRITY OF PASSIVE EQUIPMENT AND SUPPORTS. BUT MAY BE INADEQUATE FOR VERIFYING i THE OPERABIf..'Y OF ACTIVE EQUIPMENT. (2) TEST ITEMS SHOULD $1MULATE THE ACTUAL SERVICE MOUNTING DURING THE TEST. AND DYNAMIC CCUPLING WITH THE FIXTURE SHOULD BE AVOIDED. i 8. IDENTIFIED AREAS OF FUTURE RESEARCH FOR POTENTIAL ASSISTANCE IN THE LICEN3ING PROCESS: (t) T;lE DAMAGE SEVERITY FACTOR. IF FURTHER DEVELOPED. MAY BE USEFUL TO ASSESS THE RELATIVE DAMAGE THAT CAN BE INFLICTED BY EARTHQUAKE l TRANSIENTS OR TEST IhPUTS TO STRUCTURAt CCMPONENTS. CURREhTLY THE DSF IS NOT USEFUL IN ASSESSING FUNCTIONABILITY t BUT ADDITIONAL WCRK IS WARRANTED IN THIS AREA. (2) EXPLICIT GUIDANCE SHOULD BE DEVELOPED TO HANDLE GENERIC TESTING i l CONCERNS RLLATIVE TO: (A) ENERGY CONTENT VERSUS FREQUENCY DISTRIBUTION. AND (B) PROPER VALUE OF THE ZPA LEVEL IN THE j TEST INPUi. j Cf 591T 51_J M?! Q1_RE_Pf1MLLS : THE RESEARCH RESULTS ARE ESSENTIALLY CONFIRMATIVE IN NATURE. AREAS IDENTIFIED HAVE ALRLAD( kl.;IVED SIAFF ATTENTION EVEN FRIOR TO THE BEGINNING OF THIS RESEARCH PROGRAM. CONSEQUENTLY THE RESEARCH 4 RESULTS F#*E ND IMPACT ON REGULATORY REQUIREMENTS. HOWEVER. FUTURE EFFORTS SY RES CR THE IEEE STANDARD COMMITTEE ON 3 FFFINING c'RRENT CRITERIA MAY HAVE IMPACT. 4 CCTC517$f3-0?P$$2 THIS RESEARCH PROGRAM HAS HAD RELATIVELY LIMITED SCOPE AND RESOURCd5 AND CONSEQUENTLY HAS NOT I PROVIDLD b.RLCTLY USABLE NEW FROCEDURES OR METHODS OF QUALIFILATION OF EQUIPMENT. HCWEVER, IT DID SERVE A USEFUL PURPOSE IN IDENTIFYING THE POTENTIAL AREAS NEEDED FOR FURTHER INVESTIGATION WHICH SHOULD BE CONTINUED. BASED ON THE RESEARCH RESULTS. NP.R HAS TAEEN THE FOLLOWING ACTIONS:
- 1) THE IEEE STANDARDS COMMITTEE RESPONSIBLE FOR DEVELOPING i
EQUIPMENT SEISMIC QUALIFICATICN GUIDANCE HAS BEEN INFORMED OF THE RESEARCH RESULTS. INVEST!"ATION BY THE COMMITTEE I TO DETLEMINE IF ANY POSSISLE REFINEMENT OF CURRENT CRITERIA IS NECESSARY IS UNDERWAY. 2) Fb?'HER RESEARCH CN THE DAMAGE l SCVERITY FACTOR AND THE CRITERIA TO IMPROVE THE FREQUENCY VERSUS ENERGY DISTRIBUTION IN THE TEST INPUT SPECIFICATION 3 CONTINUES TO BE A RECO NENDATION OF NRR FOR CONSIDERATION BY THE CFFICE OF NUCLEAR REGULATORY RESEARCH. 3) REVIEWERS 9 HAVE BEEN INSTRUCTED OF THE CONTINUED NEED TO MAINTAIN A CAREFUL REVIEW OF ALL TEST INPUT FUNCTIONS USED OR PROPOSED FOR 8 ECUIPMENT QUALIFICATION. i 1 - 102 - l
^ 'O z p { j- - p ; g e, :_ e c 713 e o r V9 L orIti g !r O V:L tf E OF W 8? t A EO+ FE%Lilf* J THF REGULAinRY PROC [$$ .e c!t 8: 3 Eave {1<.__7 :
- t-15-3 PFs
"[ QSTnN Niit FUEL CYCLE SAFETY & EtWIRCN'iENTAL EFFECTS U;_?!YLS: DCTEP"IN!*. tifiCTIV'NE55 0F ALARA DESIGN AND CPEn.TICNAL FEATURE 5 e r v2I =: 5 errier(s3: sgr ( 7,, _ t ; ) peg-5-25 GCCUPATICNAL PFS$ ACCH P20]ECT M'R: J. FOULKE EXPOSLRE a rROTECTICN srs e""" N7%: THE PU'CLE CF THIS PGGVM WAS TO IGENT!fY AND CUANTIFY THE EXPOSURE FECUCTION POTENTIAL OF THE DESIGN A *. ; u tRATILNAL GUIDELINES GIVEN IN RECULATCRY CUIDE 3.5 AND 10 ASSESS THE COSTS INVOLVED IN IMPLEMENTING THEM. A THREE ??>T PETHJ ';CGY WAS CEV!LC"ED TC ASSESS CCCMPATIONAL EXPCSURE USAGE AT LIGH! W4TrR REACTORS. TO CETERMINE QUANTIIATIVELY ice POIENTIAL FCF RADIA!!GN EXPCSURE REOLCTION, AND TO EVALUATE THE JUSTIFICATICNS FOR MAKING IMPROVEMENTS. SECL-ENT 155ULD: REG /C7-0445. USER DISCUSSION POSITION COPMISSICS ACR5 PRESS CfflCE MELTING PAPER ERIEFING 2.F I E F I NG FELEASE RE50LTS r:+T r i r, AeT:"ITTF' 271M MS [nWt ETF D HFG Hr(3 Js$ ng_ JgFgN Tfy a f ' I '. L *LLPsSSIfLL. N s *? SC"f;ULED CCPLETICN CATE. ACTUAL CC"PLETION DATE. sa. e r N 7T s. p e t ty.. 99*7 M ratcro: 6 sesv s A n, I c a r v r.. i 23v'.af 'f FIPCFS%: ThE PURPO5F CF THIS RESEARCH 58A5 10 DEVELOP 4 METHOD TO QUANTITATIVELY sAtJAIL THt L",E f A :. t S S A N.' t t r L C I ! v t % L T, OF THE DLSICN AND CPERATICNAL CUIEdLINES GIVEN IN REGULATORY GUIDJ 3.3. h;R L E. L 5 THESE CUICELINt$ 70 t.V A L U A T E fUCLEAR PCW[R PtANT APPLICATIGMS. THE RESEARCH PESULTED IN THE DEVELOPMENT OF A ME % %0100f fCP CE.E'"INING TEL iffECTIVLNE55 0F ALARA DESIGN AND OPERATIONAL FEATURf5 FOR LWR'S ON A PLAN 1 SY PL NT 3:515.
- putasj nc?
e. rum Ts: THE RESULTS CONFIkM THAT RLCLtAIURY GUIDE 3.3 DOES NOT ADDRESS THE SIGNIFICANT METHODS OF t t ? t: ; L s t *tDJ I I L 't. INLY DiF2N51 RETE THAT IT IS f E A S I E '. " TO RENK EXPC5 CRC REDUCTICN PROJECTS dY PRIORITY ON AN IN5!/IC)AL PLANT BASIS, GIVEN THE >RCPLR INPUT I' A T A. THE RLSULTS SPCU THAT ALARA PRIORITIES VARY FRCM PLANT TO Pt sNT AND CLPENO ON SUCH THING 5 A5 D]5E PAIES. EXF05URE RECORDS AND COST CF ACTION. THERE IS NO INTENT THAT THE ELSULTS LF 7815 R E L E A ':CH SE APPLIED 10 THE LICENSING ROCE55 ALTHOUGH THE RESULTS MAY BE USEFUL TO IdDIVIDUAL PLANTS. THE*E RESULTS. HLWEVER, HAVE ADCED TO Ttil ST AFF'S GENERAL UNDLRSTAND!NG OF THE EFFECTIVENESS OF ALARA DESIGN AND C FL P L 'ICN A*, FCA!$ES. - ~ ;NTucp*.ev5: T H E '*E T HSD010 GY 15 NDY CAPALLE OF PF0VIDING #N OVERALL NON-PLANT SPECIFIC RANKIN,OF THE FEATURES Is iUtri;-y CJ I t' E 3. 3. TcE RESEARCH SHOWED THAT THE CATA NECESSARY FOR MAKING A COST-BENEFIT ANALYSIS OF ALARA ( f f ATLM 5 L E5 N0! CURRENTLY EXISI IN t.: F IhDUSTRY. LVEN If !HIS DATA WERE AVAILALLE, THE M3 DEL DEVELOPED Bf UNI
- 4'.9 St CNLY N L OF PANY M30ELS (E.G.,
NESP-310, AN NESP-Ot?) AVAILABLE TO PERFCRf1 !*:CH A COST-BENEFIT ANtLYSIS. - 103 -
P&O3&AM OFFICE CO-tMTi_pN *0TENIKAL UV tIZATAON OR VALUE OF RESEARCH RES4LTS IN THE REn#LAVOR7 PR0yESS ~ LIL J: 81 DATE I S~,U E D : 02-28-80 RES DECISION UNIT FAST BREEDER REACTORS PIL TITLE! IRRADIATED FUEL DISRUPTION UNDER LOF ACCIDENT CCNDITIONS: RESULTS OF ACPR TEST SERIES FD-; AND THE FISGAS CODE S?ONSOPING OFFICEfsl: HRR PED: 2-6 ACCIDENT ENERGETICS RESEARCH PROJEc? MGR: R. WRIGHT RES_$0*((N13: SUMMARIZED IN THIS RIL ARE RESULTS OF THE FUEL DISRUPTION-1 (FD-1) SERIES OF IN-REACTOR EXPERIMENTS OM THE SWELLING AND DISRUPTION OF IRRADIATED FUEL UNDER THE CONDITIONS OF AN UNPROTECTED (N0-SCRAM) LOSS-OF-FLOW (LOF) ACCIDENT IN AM IMFBR. ALSD PRESENTED IS ANALYSIS WITH THE FISGAS IRRADIATED-FUEL FISSIDH-GAS-BEHAVIOR CODE THAT WAS DEVELOPED FFo" !.9ESE RESULTS. IN THE CRBR PSAR, THE APPLICANT INVGAED AN ASSUMED FISSION-GAS-DRIVEN FUEL DISPERSAL AND SWEEP CUT 'O ACHIEVE A NOF-ENERGETIC TFRMINATION OF THE UNPROTECTED LOF ACCIDENT. THIS ASSUMPTION WAS REJECTED BY THE NRR STAFF CN THE BASIS THAT NO BACK-UP OATA WERE AVAILABLE. THE FD-1 SERIES OF EXPERIMENTS AND ANALYSIS WERE UNDERTAKEN TO RESOLVE THIS ISSUE, A5D THE RESULTS SUPPORT THE STAFF POSITION. NO EVIDENCE FOR THE HYPOTHESIZED PRE-MELTING FUEL DISPERS A'. WAS SEEN. BUT SUBSTANTIAL RAPID SWELLING OF THE TEST FUEL DID OCCUR ON THE APPROACH 10 MELTING. I DOCUMENT 3 ISSUED: NUREGeCR-0914; HUR EG/ C R-1124. USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS tpST PIL_ACTIVJ;,. REVIEu EELD CO*PLETED HfLD PELD IS3UED IMPLEMENTED OFFICL kESPONSIBLE......... NRR ~~ SCHEDULED COMPLETION DATE.. ACTUAL COMFLCTION DATE..... - 104 - l
PROGWAM CFFICE COMMENTS 04 POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN ThE REGULATORY PROCESS RIL s: 82 DATF ISSUED: 02-29-80 RES DECISION UNIT: FUEL CYCLE SAFETY 3 ENVIRONMENTAL EFFECTS o RTt TITLE: SOCIAL AND ECONOMIC EFFECTS OF TMI ACCIDENT SPONSOPING OFFICE (5): HRR.19-12) EED: 5-21 SOCID-ECONOMIC RESEA?CH PROJECT MGR: C. PRICHARD IMPACTS PES COMMFNT;: THIS STUDY DEALS WITH THE SOCIAL AND (CONOMIC EFFECTS OF THE ACCIDENT AT THREE MILE ISLAND DURING THE f*ksi 6 MONIHS FOLLOWING THE ACCIDENT. A VARIETY OF DATA SOURCES WERE UTILIZED INCLUDING PUBLISHED DOCUMENTS AND STATISTICS, HOUSEHOLD SUkVEYS, WEWSPAPER F11ES. INTERVIEWS, AND OTHEP RESEapCH AEDUT THE ACCIDENT. THE FINDINGS CAN BE GROUPED INTO EFFECTStON (1) THE REGIONAL ECONOMY, (2) INSTITUTIONS. AND (3) INDIVIDULLS. DIRECT ECONOMIC EFFECTS DURING THE EMEAGENCY PERIOD FOLLOWTNG THE ACCIDENT WERE INTERRUPTED LOCAL PROE?CTION AND REDUCED LOCAL INCOME AND EMPLOYMENT. LOSSES WERE CONSPICUOUS DURING THE FIRST WEEK OF APRIL BUT SUBSEQUEhiLY VERY MINOR. THERE IS NO EVIDENCE OF ANY CONTINUING INTERRUPTION OF ACTIVITY BECAUSE OF THE ACCIDENT. HOWEVER, THERE IS CONCERN WITHIN THE BUSINESS COMMUNITY ABOUT THE EFFECT OF THE ACCIDENT ON THE CONTINUED GROWTH AND DEVELOPMENT OF THE AREA. MAJOR INSTITUTIONAL EFFECTS WERE A STRAIN ON THE EMERGENCY PREPAREDNESS NETWORK IN THE AREA AND AN INCREASED FOCUS ON THE ISSUE OF THE TMI PL ANT BY THE LOCAL POPULACE. MAJOR EFFECTS ON INDIVIDUALS WERE THE EVACU ATION ITSELF AND INCREASED STRESS DURING THE ACCIDENT PERIOD. FOR MOST PEOPLE, THE EFFECTS OF INE ACCIDENT WERE SHORT-LIVED, BUT FOR OTHERS, THE ACCIDENT HAS CAUSED A MORE PERMANENT CHANGE IN THEIR DAY-TO-DAY ACTIVITIES. IT WILL BE USEFUL TO YOUR STAFF IN LICENSING HEARINGS. DOCUMENTS ISSUED: NUREG/CR-1093; NUREG/CR-12tS. USER DISCUSSION POSITION COMMISSION ACRS PRESS OrFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS Pp3T RTL_ACJ1y1TLLS LEVIEW HELD $0MPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE.. ACTUAL COMPLETION CATE..... !L3_f OP"3N_lh A PP[L_,b 1930, M. ( A LJfLA N : OliC21M_$ PM C A T I O N 10_FE ULATORY P R QM,$_$ 2 AS PART OF THE COST-BENEFIT ANALYSIS OF LICENSING APPLICATIONS, THE NRC IS RLQUIRED TO ASSESS THE LIKELY SOC 10 ECONOMIC IMPACTS ASSOCIATED WITH THE CONSTRUCTION A.1D OPERATION OF NUCLEAR POWER STATIONS ON LOCAL COMMUNITIES AND THE SURROUNDING REGIONS. THE SUBJECT REPORTS ARE AN OUTGROWTH OF CONTRACT NO. NRC-D4-78-193, ENTITLED ' POST LICENSING STUDIES OF THE SOCIDECONOMIC IMPACT OF NUCLEAR POWER STATION SITING.' AS DESIGNED, THE SONTRACT REQUIRED ANALY'SES AT 14 SITES, CNE OF WHICH WAS THREE NILE ISLAND. THE METHODOLOGY A: CONCLUSIONS OF THE SUBJECT RESEARCH REPORTS INCREASED THE STAFF'S UNDERSTANDING OF THE BEHAVIORAL RESPONSES AND COST OF ACCIDE*.75 AT NUCL EAR POWER PL ANTS. P E1 EPI EE iUP A C.l_p F PESUL T S - THE RESEARCH SERVED AS PRIMARY VLHICLES FOR DISCLOSING INFORMATION ON THE ACCIDENT EFFtCIS 10 THL PUBLIC. THE FINDINGS OF BOTH STUDIES WILL ALSO BE USEFUL IN DELINEATING THE GENERIC SOCI0 ECONOMIC EFFECTS OF CLASS 9 ACCIDENTS AT REACTORS. C 0F.M EN T S / P EM a 'ES - NOME - 105 -
PROG 0A1 0FFICE COM*ENTS CM PO!ENVIAL UTILIZATIUN OR estuE OF RESEARCN RESULIS IN THE REGULAVOQY PROCESS PIL es 33 DATE ISSUED 03-24-80 RES DECISION UNIT: PRIMARY SYSTEMS INTEGRITY 4 RIL TITLE: STEAM GENERATOR TUBE INTEGRITY SPONSQRING OFFICE (S): NRR (76-03') El$2 1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR: C. SERPAN R E.1_CQ55E NI5 s THIS RESEARCH INFORMATION LETTER TRANSMITS RESULTS OF COMPLETED RESEARCH DEALING WITH THE INTEGRITY OF ARTIFICALLY DEFECTED INCONEL 600 TU5ING TYPICAL OF THAT FOUND IN PRESENT SERVICE IN PRESSURIZED STEAM GENERATORS. t MACHINED DEFECTS IN THE FORMS OF SLOTS. ELLIPTICAL WASTAGE. ELLIPTICAL WASTAGE PLUS THROUGH-WALL EDM JLOTS AND UNIFORM TNINNING WERE CONT AINED IN TYPICAL PWR STEAM GENERATOR TUBING. BURST TESTS CHOWED THAT THE TWO PRIMARY FACTORS GOVLRNING THE BURST PRESSUPES FOR THE DEFECTED TUSES IN THIS PROGRAM ARE THE DEFECT DEPTH AND LENGTH; THIS i LATTER FACCTOR IS SIGNIFICANT BECAUSE CURRENT REGULATIONS 10 NOT REQUIRE ASSESSMENT OF DEFECT LENGTH TO JUDGE TUBE INTEGRITY. TUBE COLLAPSE CCCURRED AT PRESSURES CONSIDERABLY HIGHER THAN COULD OCCUR UNDER MOST CREDIBLE ACCIDEl.T CONDITIONS. THE SINGLE FREQUENCY EDDY CURRENT TUBE INSPECTION METHOD IS CAPABLE OF SIGNIFICANT INACCURACIES IN DEFECT 6 MEASUREMENT AND HAS A LOW PROBABILITY OF DETECTING SMALL VOLUME DEFECTS IN STRAIGHT SECTION TUBES UNDER OPTIMUM TEST CONDITIONS. NEVERTHELESS, BECAUSE OF THE LARGE MARGIN OF SAFETY BUILD INTO THE CHOICE OF STEAM GENERATOR TUBING WALL THICKNESS DIMENSIONS AND THE INHERENT TOUGHNESS OF INOONEL 600 MATERIAL, PRESENT INSPECTION AND PiUGGING CRITERIA APPEAR TO BE CONSERVATIVE. t DOCUMENT ISSUED: NUREG/CR-07tB U$tR DISCU!SION POSITION COMMISSION ACRS PRESS 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS [p}T RIL ACTIv(TIL} EEVl[H PEL D COMPLETED HELD FELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... hRR SCHEDULED COMPLETION DATE.. ACTUAL COMPLETION DATE..... N_9 PE* tin T S, A_ER_Ilm23, 19 8 0, B. LIAW: REM M f._Mf RGATION TQ_P1M141pKY PRQQElit THIS PROGRAM DEVELOPED DATA FOR NRR TO EVALUATE STEAM GENERATOR TUBE PLUGGING CRIILRIA AND EODf-CURRENT TESTING PERF0PMANCE ON A GENERIC AND PLANT SPECIFIC BASIS. THE RESULTS HAVE BEEN USED IN THE EVALUATION OF MANY OPERATING PLANTS WITH DEGRADED STEAM GENERATORS. THE PROGRAM HAS CONFIRMED THE CCH$ERVATISM OF CURRENT TUBE PLUGGING CRITERIA AND RESULTS ARE BEING USED AS INPUT TO UNRESOLVED SAFETY ISSUES A-3, A-4 AND A-5, REGAPDIEC STEAM GENERATOR TUBE INTEGRITY. AS PART OF THE UNRESOLVED SAFETY ISSUES, THESE RESULTS MAY BE USED TO DEVELOP MORE DESCRIMIN ATING TUBE PLUGGING CRITERI A AND REVISE REGUL ATORY GUIDE t.21 015 cpl h MPAQT OFjf$tM ): THE RESULTS OF THIS PROGRAM HAVE CONFIRMED THE CONSERVATISM OF THE CURRENT CRITERIA FOR PLUGGING SIEAM TENERATCR IUBES AND ARE BEING USED TO DEVELOP MORE DESCRIMINATING TUBE PLUGGING CkiTERIA IN CONNECTIDH WITH THE A-3, A-4. AND A-5 ACTIVITIES. ?pETH Q1CP[MLPr$: THIS PROGRt.M HAS PRO 7IDED A SIGNIFICANT CONTRIBUTION IN EVAL *JATING THE EXTENSIVE DEGRADATION PROBLLMS Af f ECTING OPERATING STEAM GENERATORS. k - 106 -
1 P90GDAM OFFICE COMNENTS ON PGIENTIAL Ui!LIZAI10N GM VALUE OF RESEARCH RESULIS IM THE REGULATORY PROCESS RIL 9: 64 DATF ISSULD: 03-24-80 RES DECISICN UNIT 2 SEISMIC, ENGINEERING & SITE SAFETY PIL ', I T L E 2 STUDY OF LIQUEFACTION RESULTING FROM EARTHQUAKE OF FEBRUARY 4, 1976 NEAR LAKE AMATITLAN, GUATEMALA SPONSOR ~NG OFFICEts): NRR (79-07) RES: 3-2 GEOLDGY AND SEISMIC RESEARCH PROJECT MGR2 R. BRAZEE r pgyg 0MMLNT_S: THIS RIL TRANSMITS RESULTS OF A STUDY OF SOIL CHARACTERISTICS EHICH RESULTED IN EXTFNSIVE SUBSISTENCE DUE j r 10 LICUEFACIION ALONG THE NORTHEAST SHORE OF LAKE AMATITLAN, GUATEMALA, DURING THE EARTHQUAKE OF FEBRUARY 4, 1976. THE RESutTS OF THE INVESTIGATION PROVIDE CASE HISTORY IN WHICH FIELD DATA DN SDIL CHARACTERISTICS IN AN EARTHQUAKE-l LIQUEFIED ZONE AND A NCNLIQUEFIED ZONE CA.' BE CORREL ATED WITH FIELD PERFORMANCE, SUPPL EMENTING CASE STUDIES FOR PREDICTING PROBABLE BEHAVIOR AT DTHER SITES. THE RESULTS ALSO TEND TO CORROBORATE CURRENTLY-USED PROCEDURES FOR 4 EVALUATING LIQUEFACTION POTENTIAL, DEPENDING ON THE DEGREE TO WHICH THE IN SITU PROPERTIES OF THE SOIL ARE REPRESENTED BY THE ' UNDISTURBED' SAMPLES EXTRACTED FROM T.iE DEPOSIT. THIS STUDY HAS PARTICULAR PERTINENCE T3 SITE EVALUATION AND DETERMINATION OF SEISMIC HAZARD. IT SHOULD BE REFERENCED WHENEVER THE DECISIONS ARE REQUIRES UNDER 10 CFR, P'RT 100, APPENDIX A. SECTION V, PARAGRAPH D(V) CONCERNING UNSTABLE SOILS. USED WITH APPROPRIATE JCDGEMENT, THE5E WESULTS WILL AUGMENT THE PRFSENTLY AVAILABLE DATA BASE e RELATING TO EARTHQUAKE INDUCED LIQUEFACTION AND WILL IMPROVE OUR PREDICTIVC CAPABILITY IN THIS AREA. USER DISCUSSION POSITION COMMISSION ACRS PRESS i 1 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS IT1T_PJL _AS11VJ TlQ EEVlf.W HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBtL......... hkR i SCHEDULED COMPLETION DATE.. ACTUAL COMPLETION DATE.... $D_.W;MMTL_.MAY L Mh_& EN_QQU12 2 THE RIL DOCUMLNIS A CASE Hl$10kr OF EARTHQUAKE INDUCED LIQUEFACTION. THE STUDY COMPARE 5 ACTUAL SOIL BEHAVIOR WITH TMAi PREDICTED BY EMPIRICAL CURVES. THE CURVES USED ESTIMATE LIQUEFACTION PDTENTIAL FRGM THE CYCLIC STRESS RATIO AND F#CM THE STANDAPD PENETRATION RFSISTANCE. IT WAS PARTICULARLY IMPORTANT THAT D4TA FROM WITHIN AND ADJACENT TO THE p LT1JZFACTION ZONES WAS COMPARED. THE EMPIRICAL CURVES MATCHED THE DBSERVED BEHAVIOR VERY WELL. THIS IS ENCOURAGING r 1 BUT ME ARE CONCERNED WITH THE UNIQUENESS OF THE SOIL AND GEOLOGY OF THE SITE. f ) i?8 CINIC SEDIMENTS HAVE PARTICLE SHAPES, DENSITIES. STRENGTHS, AND SIZE DISTRIBUTIONS THAT ARE NOT REPRESENTATIVE OF Ola1 TIC SEUIMENTS AS A WHOLE. ALSD. THE PHYSICAL ENVIRONMENT OF THE SITE IS NOT REPRESENTATIVE OF CONDITIONS !?rI.AL OF THE U.S. IT IS A DELTA IN A FRESH WATER LAKE AND THE UNDERLYING STRATA ARE THOUGHT TO HAVE COMPLEX FACIES RELATIONSHIPS. THE STUDY CONFIRMS AT THE EXTREME END OF TdE EMPIRICAL CURVES CURRENTLY USED AN4 LYSES FOR DETERMINING SOIL LIQUEFACTION i POTENTIAL. BECAUSE OF THE UNIQUENESS OF THE SITE. 1HE STUDv DOES NOT APPLY TO SITING OF NUCLEAR POMER PLANTS IN THE l UNITED STATES. BECAUSE OF THE UNIO"fEMESS OF THE SITE CONDITIONS THIS TYPE OF STUDY MIGHT BETTER HAVE BEEN FUNDED BY NSF. t i i i i ~ l - 107 - i l.-- . ~. _ - - -., -, _ -
? PROGDAM OFFICE COMMEN75 04 POTENilat UTXLIZATION Ok Vai'F OF #ESEARCH EESULT$_XN TdE REGULAVOQY PRQq&i) J RIL s: 85 DATE ISSUED: 03-24-80 PES DECISION U$ll:. SEISMIC, ENGINEERING A SITE SAFETY RIL TITLE: AN INTEGRATED GEOPHYSICAL AND GEOLOGICAL STUDY OF THE TECTONIC TRAMEMORK OF THE 38TH PARALLEL LINEAMENT - ANNUAL REPORT FT 1979 SPONSORING OFFICE (S): NRR, SD EEG: 3-1 NRC/ STATE REQIONA. Rfl.EARCH PROJ_EJT MGR3 N. STEUER EARTH SCIENCES RL1_R0"?ENTS: THIS STUDY IS A PART OF THE 'NEW MADRID SEISM 0 TECTONIC 51UDY' WNICH IS A CODRDINATED PROGRAM OF GLOLOGICAL, GEOPHYSICAL, AND SEI5M0 LOGICAL INVESTIGATIONS OF THE AREA WIIHIN A 200-MILE RADIUS OF NEW MhDRID, MISSUURI. THE STUDY IS DESIGNED TO DEFINE THE STRUCTURAL SETTING AND TECTONIC HISTORY OF THE AREA IN ORDER TO REALISTICALLY EVALUATE EARTHQUAKE RISKS IN THE SITING OF NUCLEAR FACILITIES. AN IMPORT ANT GOAL D.~ THE RESEARCH PROGRAM IS TO PRODUCE USEFUL SEISM 0 TECTONIC AND SEISMIC ZONING MAPS FOR THE STUDY AREA. THE PRINCIPAL PROGRESS IN THIS INTEGRATE 9 STUDY PROGRAM HAS BEEN IN ACQUIRING AND SYNTHESIZING CRITICAL MAGNETIC, GRAVITY, AND GEOLOGIC DATA; CONDUCTING CRUSTAL SEISMIC INVESTIGATIONS; AND IN INTERPRETATION OF THE AVAILABLE DATA. WHILE THESE INTERIM RESULTS ARE **0T DEFINITIVE, WE RECOMMEND ' HAT THE CURRENT PRACTICE OF EXTENb1NG THE NEW MADRID 1811-1812 EskTHQUAKES NORTH OF THE POUGN CREEK FAULT ZONE (38TH PARALLEL LINEAMENT) BE CONTINUED. IT IS ALSO RECCMMENDED THAT THE INFORMATION IN NUREG/CR-1014 BE CONSIDERED BY THE OFFICE OF STANDARDS DEVELOPMENT AND THE OFFICF OF NUCLEAR REACTOR F.EGULATION AS INPUT TO THE DEVELCPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA. DOCUMENT ISSUED: NUREG/CR-1014 USER DISCUSSION PGSITION COMMISSION ACRS PRESS CFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS Pp31 eIL ACTIVITIES EEElly HELD COMPLETED HELD HELD J13UED IMPLEMENTED OFFICE RESPONSIBLt......... NRR.SD SCHEDULED CCMPLETION DATE.. ACTUAL COMPLETION DATE..... - 108 - i
P & O'_a c M LaFICE COM*5_NTS CN POTENTIst UTitIZATION 09 vaLUE OF RESEADCH RESULTS IN THE REGULATORY PROCESS PIL e: 56 DATE ISSUED: C4-04-80 PES DECISION U:3IT FUEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS PIL TITLE: GAS SCINTILLATION PROPOR'IONAL COUNTER FOR MEASURING PLUTONIUM IN HUMANS AND THE ENVIRONMENT I SPONSOPING OFFICE (5): IE (76-3) EES: NONE RESEARCH PROJECT MGR: J. FOULKE P(1_(QS2[NTS: GAS SCINTILLATION PROPORTIONAL COUNTERS (GSPC) WERE ORIGINALLY CEVELOPED FOR USE IN SPACLCRAFT FOR X-RAY ASIRONOMY STUDIES. THIS RIL TRANSMI*S TdE RESULTS OF COMPLETED RESEARCH ON THE USE OF A GSPC FOR MEASURING PLUTONIUM IN HUMANS AND THE ESVIRONMENT. AFTER TESTING SEVERAL TYPES OF COUNT ERS. AN ELECTRON-FOCUSING PROTOTYPE COUNTER WAS BUILT WHICH HAS EXCELLENT RESOLUTION AND IS UNIFURMLY RESPONSIVE OVER THE ENTIRE VOLUME OF THE COUNTER. THE LONG-TERM RELIABILITY OF THE COUNTER WAS CHECKED FOR ONE YEAR AND FOUND TO BE EXCELLENT. THE COUNTER HAS BEEN CALIBRATED FOR MEASURING THE LUNG CONTENT OF 238 PU, 23S PU, AND 241 AM. THE COUNTING EFFICIE'4CY OF THE COUNTER IS GOOD BUT THE BACKGROUND OF THE COUNTER IS T00 HIGH. A SECOND LOW BACKGR9UND PROTOTYPE IS UNDER CONSTRUCTION IN WHICH ULTRA PURE MATERIALS WILL BE USED. UPON CGMPLETION, THIS COUNTER SMOULD PROVIDE MUCH GREATER SENSITIVITY FOR IN VIVO MEASURING PLUTONIUM IN THE LUNG. DOCUMENT ISSUED: NUREG/CR-1107, 11/79 i USER DISCUSSION POSITION COMMISSION ACRS PRESS j OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS i PD1L_PJ(__AQ[lylll[} EEVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICL RLSPONSIBLt......... IE 1 SCHEDULED C0fiPLETION DATE.. ACTUAL COMPLETION DATE.... JE E E T T_3 naPPU O. 1930. J. SyI_Eif K : THE SUBJECT RLSLARCH REPORT OUILINES IHE DEVELOPMENT OF A COUNTER FOR MEASURING PLUTONIUM IN VIVO IN HUMANS AT AN IMPROVED LEVEL OF EFFICIENCY. INCREASED EFFICIENCY WOULD PERMIT DETECTION AND CUANTIFICATION OF SMALLER QUANTITIES OF PLUTONIUM AND THUS EE USEFUL FOR EVALUATION OF LOW LEVELS OF HUMAN DEPOSITION AND ENVIRONMENTAL CONTAMINATION. IT IS OUR UNDERSTANDING THAT THE DOE HAS UNDER REVIEW A PROPOSAL FOR CONTINUING DEVELOPMENT OF THIS COUNTER. IF i THE DOE DOES NOT FUND THIS PROJECT FOR THE NEXT YEAR. WE RECCMMEND THAT NRC CONSIDER FUNDING RESEARCH INTO THE FIELD L APPLICATIONS OF 1HE COUNTER. i 1 l l - 109 - i i --__,--.m m.-
I P40Gb39 0$ FICE CCMMENT S CH #0IENTIAL UTILIZa110N OR VALUE OF #E5EAECH RESULis AM THE WE60L ATO&f PROCESS RIL s: 87 DATE ISSUED 94-24-30 RES DECISION UNIT: FUEL CYCLE SAFETY & ENVIROHNENTAL EFFECTS RIL TITLE: ECONOMETRIC MODEL FOR THE DISACCREGATION OF STATE-LEVEL ELE;TRICITY DEMAND FCRECAST5 TO THE SERVICF. AREA SPON50 PING OFFICE (5): MRR EPS: 5-21 SOCI0 ECONOMIC RESEARCH PROJECT MGRt C. PRICHARD IMPACTS EE5_CRr5ENI}: THIS RIL TRANSMITS THE RESULTS OF COMPLETED RESEARCH TO DEVELOP A MODELING CAPABILITY FOR INDEPLNDLNI ASSESSMENT OF NEED FCR POWER ESTIMATES FOR UTILITY SERVICE AREAS AS REQUIRED BY HEPA IN THE LICENSING PROCESS FOR NUCLEAR PCMER STAI!ONS. THE SLED (STATE-LEVEL ELECTRICITY DEMAND) MODEL. DEVELO*ED BY OAK RIDGE NATIONAL LABORATORIES, WAS USED AS THE BASE FOR OBTAINING PROJECTIONS OF STATE LEVEL ELECTRICITY DEMAN9. THE SLED MODEL IS A THREE SECTOR (RESIDENTIAL, C5.. MERCI A L. INDUSTRIAL) MODELINC SYSTEM IN WHICH DEMAND FOR ELECTRICITY IS DEFINED AS A FUNCTION OF ELECTRICITY PRICE, PRICES OF ALTERNATIVE FUELS, INCOME. NUMBER OF ELECTRICITY CUSTOMERS. AND HEATING AND COOLING DEGREE DAYS IN THE CASE OF THE RESIDENTIAL SECTCR; ELECTRICITY PRICE, PRICES OF ALTERNATIVE FUEL $, INCOME, PCPULATION, AND HEATING AND COOLING DEGREE DAYS IN THE CASE OF THE COMMERCIAL SECTOR; AND VAJ1E ADDED IN MANUFACTURING. THE PRICE OF ELECTRICITY, AND PRICES OF ALsERNATIVE FUELS IN THE CASE OF THE INDUSTRIAL SECTOR. THE MODEL HAS SEEN USED TO PRODUCE FORECAST 5 FOR 48 STATES THROUGH THE YEAR 2J00. IT HAS RECEIVED FAVORABLE ACADEMIC REVIEW, AND HAS PERFORMED RELATIVELY WELL IN LIMITED OUT-OF-5 AMPLE PERIOD FORECASTING. THE MODEL MAS ESTIMATED AND USED TO FORECAST ELECTRICITY DEMAND FOR SIX SERVICE AREAst CON 53LIDATED EDISON, CENTRAL HUDSON GA5 AND ELECTRIC (NEW YORK). COMMONWEALTH EDISON (ILLINDIS). SAN DIEGO GA5 AND ELECTRIC (CfLIFORNIA). CAROLINA POWER AND LIGHT (NORTH CAROLINA), AND DETROIT EDISON (MICHIGAN' IN FIVE STATES. THE ESTIMATIDH PERIOD W15 1900-1974 FOR MOST SERVICE AREAS WITH ANNUAL DATA BEING USED. AS A RESULT OF THIS STUDY, AN EFFECTIVE METHOD OF DI5 AGGREGATING STATE-LEVEL ELECTRICITY DEMAND FORECASTS TO UTILITY SERVICE AREA FORECASTS HA5 BEEN DEVELCPED. THIS CAPABILITY WILL ENABLE HRC TO MAKE AN INDEPENDENT EVALUATION OF OTHER FORECASTS OF UTILITY SERVICE AREA ELECTRICITY DEMAND. WE RECOMMEND THAT YOUR STAFF USE THIS METHOD A5 PART OF ITS ASSESSMENT OF THE NEED FOR PDWER REGIREMENTS CALLED FOR BY NEPA AS PART OF THE LICENSING PROCESS. DOCUMENT ISSUED: NUREG/CR-tt47, 2/80 USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS P051_RIL 8_CI1EJIIIS EEVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED ofFICL RESPONSIBLE......... NkR SCHEDULED COMPLETION DATE.. ACTU COMPLETION DATE..... - 110 -
l i PROGRA9 OFFICE CDP-emf 5 ON PoiENTIAL UTILIZATION OR w ALUE 05 PESEARCH RESULTS IN THE REGULATORY PROCE!_5 PIL ps 88 DATE ISSUED: 04-25-83 PES DECISION UNIT 2 PRIMARY SYSTEMS INTEGRITY RIL TITLE: DESIGN CRITERIA FOR CLOSELY-SPACED N9ZZLES IN PRESSURE VE5SELS SPON50 PING OFFICL*5): NPR EE$s 1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR* C. SERPAN P E1_ cpm"_f MI3 2 THE INFORMATION IN THIS RIL DEALS WITH THE DEVELOPMENT OF MORE ACCURATE RULES FOR THE DESIGN OF N0ZZLES IN PRES 5URE VE55ELS AND BRANCH CONNECTIONS IN PIPING FOR NUCLEtR POWER REACTOR $ THAN ARE PRESENTLY AVAILABLE IN THE A5ME BOILER AND PRESSURE VESSEL CODE. SECTION III. THE PRESENT RULES WERE DEVELOPED OVER A LONG I PERIOD OF TIME, BASED MAINLY ON EXPERIENCE AND DEMONSTRATED TRIAL AND EREDR PROCEDURES IN NONNUCLEAR APPLICATIONS. THE ADVENT OF MODERN FINITE ELEMENT COMPUTER CODE CALCULATIONAL PROCEDURES HA5 ENABLED US TO DEVELOP A NEW RATIONAL BASIS FCR THE DESIGN OF SUCH CCMPONENTS. THESE CALCULATIONAL PROCEDURES, CJUPLED WITH THE RELATIVELY MODEST EXPERIMENTAL DATA AVAILABLE. HA5 AL50 ALLOWED US TO DETERMINE THE DEGREES OF CONSERVATISM (CR UNCONSERVATISM) INHERENT IN THE PRESENT RULES, AS APPLIED TC THE ACTUAL STRESS LEVELS INDUCED IN THE STRUCTURES UNDER REVIEW. THIS WORK WA5 DONE OVER A 3-YEAR PERIOD AND HAS RESUL iD IN T4E DEVELOPMENT OF PROPOSED NEW DESIGN RULES WHICH HAVE BEEN SUBMITTED TO THE ASME FOR INCLUSION (OR REPsACEMENT) CF THE CURRENT APPLICABLE PARAGRAPHS IN SECTICN III. IN GECERAL. THIS WORK HAS SHOWN THAT THE PRESENTLY APPLIED DESIGN RULES HAVE, IN THE MOST PART, BEEN OVERCONSERVATIVE. I.E., REQUIRING MORE MATERIAL THAN NEEDED TO MAINTAIN REQUIRED STRESS LEVELS. FURT3tt, IN THE FEW CASES WHERE THE PRESENT DESIGN RULES LEAD TO SLIGHTLY UNCONSERVATIVE DESIGN, THE RESULTANT MODEST DtcREASE IN THE ALREADY SIGNIFICANT I SAFETf FACTORS DOES NOT COMPROMISE THE SAFETY OF THE STRUCTURES IN OUESTION. THUS, NO REEXAMINATION OF MODIFICATION OF EXISTING STRUCTURES IS CALLED FOR. USER DISCUSSION POSITION COMMISSION ACR5 PRESS OFFICE MEETING PAPER 3RIEFING BRIEFING RELEA5E RESULTS EDil_PIL_ACIJ VJll Ci EEVIIM H_r L D COMPLETED HELD HELD _ 155'JED IMPLEMENTED OFFICE kE5PONSIBLL......... NRR SCHEDULED COMPLETION DATE.. ACTUAL CCMPLETION DATE..... L I 1 - Ill - ,--,m----. ,,.-,_m_.- -, ~ - - - -., - m-7 --. e
. ~.. ~. - - PPOGRAM OF(R_CE CC N ENYS AN P6?ENTIAL HifLAZAVRON OR UALHE OF QESEARCH RESMLTS IN THE REGutaT0*Y PROCESS RIL en 81 DATE 1559EDs 95-11-80 PES DECISION UNITt SEISMIC, ENGINEERING & SITE SAFETY i PIL TI7tEs STRUCTURAL AND MECHANICAL COMPONENT TEST TECHNIQUES SPONSCRING OrrICE(5): t# R R (77-18) PAQ: DYNAMIC TESTING RESEAPCH PROJFCT MOR: J. O'BRIEN METHODS OF OPER4 TING REACTORC PF'gC_C P'*M*Tj : THIS RIL TR ANSMITS THE RESULTS OF STUDIES L:HICH EXPLORE AND INVESTIGATE THE FEASIBILITY. COSTS, BLNEFII5. RLLIABILITY. LIMITATIONS AMD POTENTI AL PL ANT DEGRADATION ASSOCI ATED WITH CONFIRMATORY IN SITU DYNAMIC TESTING UTILIZING VARIOUS MEANS OF EXCITING VIBRATICH5 IN SAFETY-RELATED STRUCTURES AND MECHANICAL EQUIPMENT. 2 USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS fMT_PlLACTlVJ TIES PEVIEW HEL D CC9PLETED !!1Lp HELD ISSUED IMPLEMENTED 0FFICE RL5PONSILLE......... N PA ~~ SCHEDULED COM.PLETION DATE..
- 1, ACTUAL COMPLETION DATE.....
a v i 4 9 - 112 - i n-, ,.,-_._.-,_.~..-.-._-,n_=-
PR06&a5 OFFICE 50 " ENTS ON PO T E t:T l a t UTILIZATION OR vaLUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS PIL ss 90 DATE ISSUED: 05-22-80 RES DECISION UNIT: CODE DEVELOPMENT PIL TITLE: RELAP-4/M006 ASSESSMENT SPONSORINS OFFICE (S): NRR (77-5) RPG: 1-17 CODE APPRAISAL RESEARCH PROJECT MGR 14. LYON UMMN13: RESULTS OF INDEPENDENT ASSESSMENT OF THE RELAP-4/ MOD 6 CODE ARE FRESENTED, TOGETHER WITH A
SUMMARY
OF IHL PWR LARGE 3REAK LOCA UNCERTAINTY STUDY, LIMITED TO THE BLOWDCWN REGIME OF LOCA. THE CODE WAS FOUND TO HAVE ADECUATE TREATMENT OF THERMAL-HYDRAULICS DURING THE BLOWDOWN PERIOD OF LOCA. ITS PERFORMANCE WAS SPOTTY DURING THE REFLOOD PERIOD AND INADEQUATE DURING THE REFILL STAGE. THE CODE CR M9 DEL INPUT PARAMETERS FOUND TO HAVE THE STRONGEST EFFECT ON THE PEAK CLAD TEMPERATURE DURING THE BLOWDOWN STAGE OF LOCA COMPRISED PGWER PEAKING FACTORS. F8;EL Gt.P CONDUCTANCE. THERMAL CONDUCTIVITY OF FUEL PELLETS, AND THE INITIAL POWER. USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS I P_D $ 7 PIL_ASTIE T_IQ P F.VI D! HEtD COMPLETED-HEL D HELD ISSUED IhPLEMENTED OFFICE RESPON5IBLL......... NRR SCHEDULED COMPLETION DATE.. ACTUAL COMPLETION DATE..... ~~ I I - 113 - l l
3.0 PPDJECTED NEAR-TEPM RESEARCH INFO *MATION LETTERS (TARGET DATES ARE FISCAL YEARS) RRG S AND NAME/ TEMP. N tJ"S. 1HI)ECT/ TITLE / IMPACT TARGET DATE (DECISION UNIT) RRG CHAIRMAN T-4 ENLRGENCY PLANNING QTR 3, 30 6-1 EMERGENCY PLANNING R. BLOND (RISK ASSESSMENT) jp j T-23 RADIATICH E50"DCE LEE 4>- SIMUL ATER 4E5fS. QTR 3, 80 1-25 QUALIFICATION, TESTING, R. FEIT EWALUATION 4 (SYSTEMS ENGINEERING) T-29 CONSEQUENCE MODEL (INDIVIDUAL SITE QTR 3, 30 6-2 CONSEQUENCE MODELING R. BLOND ASSESSMENT, FINAL MODEL UPDATE. CRAC (RISK ASSESSMEN7) USERS MANUAL, AND COMPARISON TO SAFETY REVIEW METHODS IN 10 CFR 100) T-32 ELECTROCHEMICAL TEST FOR SENSITIZATICN. QTR 3, 80 1-22 CORROSION J. MUSCARA UPDATES REG. GUIDE TO PRECLUDE SENSITIZED (PRIMARY SYSTEMS INTEGRITY) STAINLESS STEEL IN SERVICES. T-34 ENVIRONMENTAL 10 DINE SPECIES BEHAVIOR. QTR 3, 80 5-16 AQUATIC RADIONUCLIDES, P. REED I STUDY THE PHYSICAL & BIOLOGICAL TRAMSPORT RADI0 ECOLOGY j OF CHEMICAL FORMS TO RADIOI0 DINE 3 RELEASED (REACTOR ENVIRONMENTAL) TO THE ENVIRONMENT FROM AN OPERATING 4 4 NUCLEAR STATION. DETERMINE THE INFLL6HCE OF WET DCP051 TION (RAIN OR DEW) FOR METEOR 0'.GICAL MODEL5 & THE IODINE-AIR-q CRASS-MILK PATHWAYS. PFRFORM LABORATORY TESTS TO DETERMINE IF MLTHYIODIDE 15 DEPOSITED ON GRASS UNDER WET DEPOSITICH l CONDITIONS. DETERMINE ENVIRONMENTAL i PATHWAYS OF TRITIUM AND CARBCN-54 RELEASED FROM NUCLEAR STATIONS. { r T-46 SOURCE TERM CORRELATION. CONFIRMS QTR 3, 80 1-13 FUEL MELT R. SHERRY CONSERVATISM OF REG. GUIDE ASSUMPTIONS (FUEL BEHAVIOR) FCR ACCIDENT ANALYSIS. / 21 T-84 AGING MODEL QTR is 80 1-25 QUALIFICATION TESTING R. FEIT EVALUATION (SYSTEMS ENGINEERING) { [ d'T 51 A COMPREHENSIVE REPDRT WHICH WILL QTR 3, 30 3-7 SECONDARY CONTAINMENT B. BROWZIH n 'N \\ RECOMMEND CRITERI A FOR NRR LICENSING STRUCTURAL 4 (Q POSITIONS ON PUNCHING SHEAR WTTH COMBINED (SEISMIC, ENGINEERING & BIAXIAL TENSION. SITE SAFETY l T-162 RESPONSE OF NUCLEAR POWER PLANT STRUCTURES QTR 3, 80 3-9 SEISMIC SAFETY MARGINS C. BURGER TO THREE INPUT COMPONENTS. RESEARCH PROGRAM (SEISMIC, ENGINEERING 1 i SITE SAFLfY) i i - 114 -
T-170 NUMERICAL TECHNICUES FOR DETERMINING QTR 3, 80 DYNAMIC TESTING METHODS OF J. C'5RIEN FREQUENCIES, MODE SJtPES AND DAMPING OPERATING REACTORS VALVES USING LOW LEVEL INPUT. (SEISMIC, ENGINEERING 8 SITE SAFETY) 7-179 CESCRIPTION OF A DESIGN CONCEPT FOR QTR 3, 83 MECHANICAL ENGINEERING J. BURNS CAL CUL A TING PF.0B ABIL I TY OF R ADIDACTIVE (SEISMIC. ENGINEERING 8 RELEASE, CORE MELT, SAFETY SYSTEM. SITE SAFETY) STRUCTURAL AND COMPCNENT FAILURE, / PROBAEILITIES FRCM A SET OF NUCLEAR / PiANT SEISMIC RESPONSES. d} b T-189 POSITRON ANNIHILATION - E V ".. U A T I O N QTR 3, 80 2-4 CORE MELT AND CONTAINMENT T. WALKER k,* AS AN NLE PROCEDURE. EVGUATION OF INTEGRITY ELECTP035 AT DEFECT SITES EY POSITRONS (FAST EREEDERS) Ep AND CORRELATION WITH CREEP-FATIGUE 4 DAMAGE. T-191 PROMPT-BORST ENERGETICS EXPERIMENTS QTR 3, 80 2-6 ACCIDENT ENERGETICS R. WRISHT WITH CAREIDE FUEL, ACRR TESTS SG-1, (FAST BREEDERS) S G-2, AND SG-3. i T-196 DYNAMIC SIMULATION OF WASTE / ROCK PROCESS QTR 3, 30 6 5 HIGH LEVEL WASTE ISOLATION M. CULLINGFORD (G' 0 LOGICAL FEEDB ACK MECHANISM MODELING) (RISK ASSESSMENT) T-2:S ORINCCN TRANSPORTATION METWORW. MODEL QTR 3, 80 4-1 EFFECTIVENESS EVALUATION R. R05INSON (SAFEGUARDS) ' h$N ' T-721 FIXED-SITE NEUIRALIZATION MODEL (FSNM) QTR 3. 80 4-1 EFFECTIVENESS EVALUATION
- 4. ROBINSON
("+ ' (SAFEGUARDS) \\ T-224 SOUPCE AMBUSH SIMULATION MODEL 3 USER QTR 3, 80 4-5 MEASUREMENTS 8 STANDARDS R. ROBINSON MANUAL DOCo.iENTATION (SAFEGUARDS) h.. -227 T ADVERSARY PATH SELECTION METHODS OTR 3, 80 4-1 EFFECTIVENESS EVALUATIDH R. ROSINSON f'Y ' T-228 SACEM SMALL ARMS CtSUALTY EVALUATION QTR 3, 80 4-1 EFFECTIVENESS EVALUATION R. ROBIFSON (SAFEGUARDS) MODEL (SAFEGUARDS) T-241 MODELINC NORMAL SHOCK AND VIBRATION QTR 3, 80 5-10 TRANSPORTATION SAFETY W. LAHS ENVIRONMENT STRUCTURAL ANALYSIS (FUEL CYCLE F ENVIRONMENTAL 5tESEARCH) T-243 MEASUP.EMENT TECHh0 LOGY FOR PLUTONIUM QTR 3, 80 5-23 OCCUPATIONAL EXPOSURE J. FOULKE PROTECTION 'g (FUEL CYCLE A ENVIRONMENTAL RESEARCH) T-246 SURFACE PROPERTIES OF ENCAPSULANTS QTR 3, 80 NONE K. KIM (MASTE MANAGEMENT) T-247 WASTE-ROCK INTERACTI9NS QTR 3, 80 NONE K. KIM (WASTE MANAGEMENT) 115 - _ _ _ _ _ _, _., ~ ______ _.
_ _ __=.____._.___.- -- i T-248 SOLUBI?,ITY OF URANINITE QTR 3, 80 NONE K. KIM (WASTE MANAGEMEN13 T-249 REPOSITORY CPTION ASSES $FENT QTR 3, 80 NONE K. KIM (WASTE MANAGEMENT) T-251 GLAS5 CERAMIC RADWASTE CONTAINER QTR 3, 80 NONE K. KIM EVALUATION (WASTE MANAGCMENT) T-254 EVALUATION OF THbRIUM CCNTENT OF HUMAN QTR 5, 30 5-23 OCCUPATIONA" EXPOSURE J. FOULKE TI55UE5 PROTECTION (WASTE MANAGEMENT) T-255 EXREM COMPUTER CODE QTR 3, 80 5-24 RADI0 BIOLOGY 8 DOSIMETRY J. FOULKE (RE9C104 ENVIRONMENTAL) T-258 PATHOGENIC AMOEBAE IN CLOSED CYCLE QTR 3, 80 5-18 ECOLOGICAL IMPACT FROM J. FOULKE COOLING TCWERS REACTOR 3 FUEL CYCLE FACILITIES (REACTOR ENVIRONMENTAL 3 T-263 EHL EMERGENCY AIF f*MPLER FOR RADIOI0 DINES QTR 3. 8E 5-17 ENVIRONMENTAL MONITORING P. REED (REACTOR ENVIRONMENTAL) T-254 REVIEW OF CONSERVATION, LO AD M AN AGEMENT, QTR 3, 80 5-21 SOCIDECONOMIC IMPACTS C. PRICHARD RATE RESTRUCTIVE AND C0 GENERATION (REACTOR ENVIRONMENTAL) T-265 VISUAL IMPACT OF ALTERNATIVE CLOSED QTR 3, 80 5-21 SOCIDECONOMIC IMPACTS C. PRICHARD CYCLE COOLING SYSTEMS (REACTOR ENVIRONMENTAL) T-267 ASSESSMENT OF ST ATE-OF-THE-ART OF MECH. QTR 3, 30 NONE J. BURNS SYS. AND COMP. (SEISMIC, ENGINEERING & SITE SAFETY) T-273 VALVE LER ANALYSIS QTR 3, 80 NONE (RISK ASSESSMENT) T-275 HUMAN FACTOR 5 ANALYSIS QTR 3, 30 NONE (RISK A55C55 MENT) T-276 FIRE DATA ANALYSIS QTR 3, 80 NONE (RISK ASSESSMENT) 7-277 PENETRATION LER ANALYSIS QTR 3, 80 NONE (RISK ASSESSMENT) 'T-283 SUS-SURFACE PROFILING OF THi BEATTY, QTR 3, 80 NDNE C. JUPITER NEVADA SHALLOW LAND NUCLEAR WASTE BURIAL (WASTE MANAGEMENT) SITE. ) A Tpf85 FORECASTING ELECT DEMANDS BY STATE / QTR 3, 80 NONE C. PRICHARD '( UTILITY SERVICE AREAS (REACTOR ENVIRONMrHTAL) T-286 NPS CONSTRUCTION-LABOR FORCE MIGRATIDH QTR 3, 80 NONE C. PRICHARD 8 RESIDENTIAL CHDICE (REACTOR ENVIRONMENTAL) - 116 -
1 T-12 CRACK APREST TEST METHODOLCGY. BASIS QTR 4, 40 1-20 VE5SEL INTEGRITY M. VAGINS FCR SEVEFE ACCTDENT ANALYSIS ARREST EVAlbATION. (PRIMARY SYSTEMS INTEGRITY) T-15 A55E55 HEN' UF AG2ICULTUR AL L AND QTR 4, 80 5-21 SOCIDECONCMIC IMPACTS C. PRICHARD (REACTOR ENVIRONMENTAL) T-25 WRAP EN/EWR - INCCRPCRATES INTEGRATED QTR 4 80 NONE L. SHOTKIN AND AUTCMATED IMPROVEMENT 5 IHfD A (CDDE DEVELOPMtNT) LICENSIhG CODE PACP: AGE. T-38 REWET CURRELATION. SUMMARIZE 5 CURRENT QTR 4 80 t-5 PWR-BDHT Y. HSU P/EDICTIVE 00ALITIES FGR ' RETURN OF 1-4 BWR-BDH! Y. HSU NUCLEATE BOILING'. (SYSTEM 5 ENGINEERING) T-42 TRANSIENT CHF CCRRELATION. NO CHANGE QTR 4 80 1-5 PWR-BDHT Y. H5U IN APPENDIX K BUT MAY REFLACE CURRENT 1-4 ADVANCED SYSTEM CODE Y. HSU CORRELATION. ($YST EMS ENGINEERING) T-4S MECHANICAL Pw0PERTIES REPorf. PROVIDES QTR 4, 30 1-8 ZIRCA10Y CLADDING M. PICKLE 5IMER INCEPENDENil.Y VERIFIED I"RADIATED (FUEL BEf' VIOR) ZIRCAt0Y DATA ruR USE IN LICENSING CALCULAT!GNS. f %4' O 12% nd)Lt/ EES TRANSIENI M A f* M OMPttif-SUPPORT / $l T-48 4EST QTR M AC 1-4 EWR-BDHT W. BMKNER EXISTING CCFRELATION CR RECCMP.END A (SYSTEMS ENGINEERING) I NEW CORRELATION. / fj i T-67 17 x 17 REFLOCD HEAT TRANSFER. QTR,F JES 1-5 REFLCOD HEAT TRANSFER E. DAVIDSON
- MPROVfn HEAI TRANSFER CCRRELATIONS
($YSTEMS ENGINEERING) (CR UNht0Cr.ED EUNDLES. 1 T-70 EMERITTLEMENT CRITERIA. MAY REPLACE QTR 4, 80 1-8 ZIRCALOY CLADDING M. PICKLESIMER EXISTINr. CRITERIA IN APPENDIX K. (FUEL BEHAVIOR) T-74 RELAP 4 MOD 7. USER CCMVENIENT QTR 4, 80 1-16 REFERENCE SYSTEM CODE F. ODAR PWR LOCA CCDE (CDDE DEVELOPMENT) T-76 CRACK ARREST. BASIS FCR SEVERE QTR 4, 80 NONE M. VAGIN5 ACCIDENT ANALYSIS ARREST EVALUATION. (PRIMARY SYSTEMS INTEGRITY) i T-32 EXTINGUISHING SYSTEM EVALUmTION QTR 4, 80 1-25 ELECTRICAL STANDARDS R. FEIT AND FIEE PROTECTION i ($YSTEMS ENGINEERING) 1-87 VALIDATE ACOU 5 TIC EMISSION FLAW QTR 4, 80 NONE J. hUSCARA CGRRELATICN. (PRIMARY SYSTEMS INTEGRITY) / $$ T-100 ALTERNATE ECC5 SYSTEMS QTR A'.)30 NDNE W. LYON (SYSTEMS EhGINEERING) 4 I ) T-10$ K-FIX (3D). TWO PHASE FLOW IN IDEALIZED QTR 4 30 1-14 ADVANCED CODE L. tHOTKIN GE0 METRICS. (CODE DEVELOPMENT) l - 117 -
~ i 1-119 FI5510N PFCDUCT QEHAb10R WITHIN LWR QTO 4, GO NONE R. SHEQQY PRIMARY SYSTEMS UWDEC ACCIDENT CCNDITIONS (FUEL BEM4UIOR) (C9NTROLLED LOCA AND MELTDOWN) CESCRIPTION CF TRAP-MELT CODE. T-111 MEtf CONCRETE INTERACTIONS. UPDATE 10 QTR 4, 80 1-13 FUEL MELT R. SHrRRY RIL 28. DESCRIPTION OF CCRCON CODE. (FUEL BEHAVICR) T-116 BUILDING Mar.E DIFFUSION. WIND TUhMEL QTR 4, 80 3-4 SEVERE STORMS R. AB3EY MODEL STUDIES. (SEISMIC, ENGINEERING a SITE SAFETY) T-117 BUILDING WAKE DIFFUSION. FIELD TEST QTR 4,-80 3-4 SEVERE STORMS R. AB3EY STUDIES. (SEISMIC, ENGINEERING & SITE SAFETY) + -+ T-121 A55E55MCHT AND EXPANSICN OF STRONG QTR 4, 80 3-2 GEOLOGY
- SEISMIC R. BRAZEE 2
CPOUND MJTION DATA. BASIC INPUT TO CHARACTERISTICS SEISMIC RISK ASSESSMENT. (SEISMIC, ENGINEERIFG & SITE SAFETY) T-144 MODELING TORNADO DYNAMICS. QTR 9, 86 NONE R. AFBEY (SEISMIP. ENGINEERIhG & SITE SAFETY) T '49 A REPORT FOR ORIENTATION FOR LICENSING QTR 4, 80 3-7 SECONDARY CSNTAINMENT B. B;0WZIN ON THE FFECT OF SEI5MIC DESIGN LEVEL STRUCYuRAL ON TbE NPP C05T. (SEISMIC, ENGINEERING 3 SITE SAFETY) T-163 EFFECT OF STRUCTURAL DAMPING ON NUCLEAR QTR 4, 80 3-9 SEISMIC 5AFETY MARGINS C. BURGER POWER Ps. ANT STRUCTURES. THIS IS At:0THER RESEARCH PROCRAM I STUDY BY EXPANDING LLL/ DOR SEI5MIC (SEISMIC, Et.GINEERING & CONSERVATISM PROGRAM TO TYPICAL NUCLEAR SITE SAFETY) POWER
- PLANT STRUCTURES (ZION STATION).
T-164 GENERAL STRUCTURAL BUILDING RESPONSE QTR 4, 80 3-9 SEISMIC 5AFETY MARGINS C. EURGER i ANALYSIS REVIEW WITH SPECIAL.EPJHA515 RESEARCH PROGRAM ON DAMPING AND NONLINEARITY; * (SEISMIC, ENGINEERING & SITE SAFETY) T-198 RSS METHODOLOGY APPLICSTION PROGRAM QTR 4, 83 6-6 LWR RISK ASSE5SMENT M. CUNNINGHAM (RISK A55E55;1ENT) T-207 RISK ASSESSMENT OF HIGH LEVEL WASTE Mi QTR 4, 6-5 HIGH LEVEL WmSTE ISOLATION M. *:ULLINGFORD (RISK ASSESSMENT) ISOLATION IN BEDDED SALT -213 15EM ADVERSARY SEQUENCE EVALUATION M3 DEL QTR 4, 80 4-1 EFFECTIVENESS EVALUATION R. 40 BIN 50N ( (SAFEGUARDS) J T-217 COPS LLEA ROUTE DISTRABUTION MODEL QTR 4, 80 4-5 MEASUREMENTS & STANDARD 5 R. ROBINSON (SAFEGUARDS) l T-219 STANDARDIZED LWR CENERIC FAULT TREE QTR 4, 80 4-1 EFFECTIVENESS. EVALUATION R. ROBINSON. CHARACTERIZATICN CODE (SAFEGUARDS) - 113 - i ,n-,- -,. -,.,-en -,.-n ,em- , - - r, v -n- ---n
~ 4 l i l T-222 S AFEGUARDS ENGINEERING AND ANALYSIS DAT A QTR 4, 80 4-1 EFEECTIVENE55 EVALUATION R. ROBINSON 1 BASE (SEAD) (SAFEGUARDS) T-222 5AFE f SEMI-AUTDMATED SYSTEM EVALUATION STR 4, 60 4-1 EFFECTIVENESS EVALUATION R. ROBINSON METHODOLOGY (SAFEGUARDS) T-223 BWR FACILITY CHARACTERIZATICN CODE QTR 4 80 4-1 EFFECTIVENESS EV LUATION R. RCBINSON (SAFEGUARDS) I T-225 PWR FACILITY CHARACTERI2ATION CDCE QTR 4, 30 4-1 EFFi?TIVENE55 EVauUATION R. ROBINSON I (S AFEGU. 7.DS) T-226 SKIRMI5H/ AMBUSH BOARD GAMES QTR 4, 80 4-5 MEASUREMENTS & STANDARDS R. ROBIN 50H l (SAFEGUARDS) T-229 5AFEGUAPDS NETWORK ANALYSIS PROCEDURE QTR 4, 30 4-1 EFFECTIVENESS EVALUATION R. ROBINSON l (5 NAP) (SAFEGUARDS) T-230 SABRES II CONVOY ENGAGEMENT MODEL 97* 4 80 4-5 MEASUREMENTS & STANDARDS R. ROBINSON i (SAFEGUARDS) n i T-231 EARS CCMMUNICATIONS MODEL QTR 4, 81 4-5 MEASUREMENTS 8 STANDARDS R. ROBIN 5ON [ ("AFEGUARDS) L T-235 ATTR*BUTES OF THE INSIDER ADVERSARY QTR 4. 80 4-2 THREAT ASSESSMENT R. SHEPARD (SAFEGUARDS) T-237 CRITICALITY SAFETY METHODS - SOLID ANGLE QTR 4, 8J 5-7 CRITICALITY SAFETY STUDIES D. SOLBERG 8 $bRFACE DENSITY (FUEL CYCLE & ENVIRONMENTAL RESEARCH) T-238 GUIDANCE POR ADMINISTRATIVE CONTROL OF QTR 4, 30 5-7 CRITICALITY 5AFETY STUDIES D. SOLBERG CPITICALITY SAFETY (FUEL CYCLE & ENVIRONMENTAL l i' RESEARCH) i i d T-240 ' ARTICLE LEAK STUDIES QTR 4, 80 5-12 TRANSPORTATION SAFETY W. LAH5 PROGRAM 1 (FUEL CYCLE 8 ENVIRONMENTAL RESEARCH) i T-244 ASSESSMENT OF RESPIRATORY PROTECTION QTR 4, 30 5-23 OCCUPATIONAL EXPO 5URE J. FOULKE j SYSTEMS PROTECTION I (FUEL CYCLE 4 ENVIRONMENTAL RESEARCH) t T-252 ENGINEERIWG EVALUATION OF LOW LEVEL QTR 4, 80 NONE E. HELD 1 WASTE DISPOSAL SITES (WASTE MANAGEMENT) T-253 BURIAL GROUND SITE SURVEY - KENTUCKY QTR 4, 80 NONE E. HELD l (WAST E MANAGEMENT) I l T-256 ALARA DESIGN OBJECTIVES FOR LWR QTR 4, 80 5-23 OCCUPATIONAL EXPOSURE J. FOULKE PROTECTION l (REACTOR ENVIRONMENTAL) t T-257 PEDIATRIC PHANTOMS QTR 4. 80 5-24 RADI0 BIOLOGY & DOSIMETRY J. FOULKE l (REAC10R ENVIRONMENTAt) - 119 - --v-w.-vv------p ~m w s-- -r m, w, +r-m.,.---.,-4 ,n.-e-w- ,,---e-_,, vr- - - -.- .w.,---m-w,e-w-
T-260 EFFECTS OF AIRD0RME CONTAMIMANTS ON QTR 4, 80 5-6 FUEL CYCLE FACILITIES
- 3. SOLBERG ACTIV TED CHARCCAL EFFLUENT CONTROL (REACTOR ENVIRONMENTAL)
T-251 A5BESTOS IN C00 LING TOWER WATERS QTR 4, 30 5-15 CHEMICAL IMPACTS ON ?. REED AQUATIC ENVIR0hMENT (REACTOR ENVIRONMENTAL) 7-262 MODEL 70 PREDICT CHLORINE CONCENTRATIONS QTR 4, 80 5-14 PHYSICAL TRANSPORT SURFACE 7 REED IN POWER STATION DISCHARGES WATERS (REACTdR ENVIRONMENTAL) T-26B IF31GHT5 FROM HDR SEISMIC TESTS QTR 4, 80 NONE J. O'BRIEN (SEISMIC, ENGINEERING 1 SITE SAFETY) T-271 COMMON CAUSE MODELING QIR 4, 80 NONE (RISK ASSESSMENT) T-278 FLOOD OCCURRENCE M9DELLING QTR 4, 80 NONE (RISK ASSESSMENT) T-279 SMALL BREAK TEST L3-1 AND L3-2 QTR 4, 30 1-1 LOFT (LOFT) T-288 RISK METHODOLOGY DEVELOPMENT FOR FUEL QTR 4, 80 6-4 RISK ASSESSMENT FOR M. CULLINGFORD PRCCESSING FUEL PROCESSING (RISK ASSESSMENT) T-6 REPORT ON WASTE MANAGEMENT RISK QTR t, at 6-5 HIGH LEVEL WASTE ISOLATInM M. CULLJNGFORD ASSESSMENT METHODSLOGY DEVELGFMENT T-;6 SIMMER-II RELEASED. REA5SESSMENT OF WORK Q!R t, 81 2-14 SIMMER CODE R. CURTIS ENERGY TO BE ACCOMMODATED IN NDCA. (F AST BREEDERS) T-28 CORE MELTDOWN SENSITIVITY STUDY QTR t, 81 6-6 LWR RISK ASSESSMENT M. CUMMINGHAM T-6F COMPREHENSIVE REPORTS WHICH WILL QTR t, 31 3-7 SECONDARY CONTAIPMENT
- 3. BROWZ1N SUMMARIZE RESULTS OF L ARGE SPECIMEN STRUCTURAL TESTING ON SEISMIC SHEAR TRAN5FER.
(SEISMIC. ENGINEERING S SITE SAFETY) T-75 PROBABILITY AND CONSEQUENCES OF STEAM QTR t, 81 1-13 FUEL MELT R. SHERRY ] EXPLO5 IONS. POSSIBLY WILL FROVIDE A (FUEL BEHAVIOR) QUANTITATIVE PREDICTION OF STEAM EXPLCSIONS MOSTLY APPLICABLE TO FLOATING PLANTS. T-83 DETECTION SYSTEM EVALUATION QTR 1, 88 t-23 ELECTRICAL STANDARDD5 R. FEIT AND FIRE PROTECTION I (SYSTEMS ENGINEERING) T-97 URANIUM OXIDE AEROSOL PROPERTIES AND QTR 1, St 2-9 LARGE AEROSOL TRANSPORT M. SILBERBERG 1 NSPP TEST AEROSOL CODE VERIFICATION. TESTS 1 DATA USED TO VERIFY INFORMATION IN HAARM-3. 4 I l - 120 -
~ T-115 EVALUTE METHOD TO CALCULATE COASTAL QTR 1, 8 5-8 ATMO5PHERIC TRAN5? ORT & R. ABBEY SISPERSION. MAY IMPACT CN LICEM5ING DEPOSITION i EVALUATION OF CLOSE-END ACCICENTAL (SEISMIC. ENGINEERING 1 RELEASE CCNCENTRATION. SITE SAFETY) i 4 T-sta ULTIMATE HEAT SINF. VERIFICATICN EXP. QTR 1, 81 3-8 ATMO5PHERIC TRANSPORT AND R. ABBEY I DEPOSITICN (SEISMIC. ENGINEERING & SITE SAFETY) T-119 SOUTHEAST SEISMIC NETWORK. HISTOP' 0F QTR 1 81 3-2 GEOLOGY AND SEISMIC R. BRAZEE INSTALLATION OPERATION RESEAPCH. 2-1/2 CHARACTERISTICS YE AR SEI5MIC ACTIVITY SUP. MARY AND RESUME. (SEISMIC. ENGINEERING & SITE SAFETY) T-832 INTERIM REPCRT STUDY EE5ULTS. PPELIMINARY QTR 1 81 3-1 NRC/ STATE REGIONAL EARTH N. STEUER TECTDNIC PP0t!NCE BOUNDARIES, REGIONAL SCIENCES a GEOLOGY, SE15MOTECTch!C5. TO ACCELERATE (SEISMIC. ENGINEERING 1 7 LIC[NSING. SITE 5AFETY) T-147 A CCMPREHENSIVE REPORT WNICH WILL QTR 1 81 3-7 SECONDARY CONTAINMENT
- 8. BROWZIM RECCMMEND CRITERIA FCR NRR LICENSING POSITIONS ON SEISMIC SHEAR TRANSFER STRUCTURAL (SEISMIC, ENGINEERING &
BASED ON MEDIUM SCALE TESTING. SITE SAFETY) T-153 A SERIES OF FCRCE-TIME HISTORIES CHAR-QTR 1 81 NONE J. COSTELLO ACTERIZING AUTCMOBILE aMPACTS 47 DIFFERENT (SEISMIC, ENGINEERING & j VELCCITIES WILL BC PRESENTED TO ACCOUNT SITE SAFETY) j FOR TORNADO RISK REGIONS AND WEHICLE CRIENTATIONS AT IMPACT. ENVELOPING FORCE- .l TIME HISTORIES WHICH CAN BE U%ED FOR DESIGN 1 AGAINST 70FNADD MISSILE EFFECTS IN EACH TOPNADO REGION WILL BE RECOMMENDED. 2 T-155 SENSITIVITY STUDY OF SOIL-STRUCTURE QTR 1, 31 3-9 SEISMIC SAFETY MARGINS J. COSTELLO INTERACTION PHLNOMENON FOR SOIL-RESEARCH PROGRAM STRUCTURE INTERACTION, 50!L PROPERTIES (SEISMIC. ENGINEERING S AND SOIL CONFIGURATION. WAVE PASSAGE SITE SAFETY) AND AZIMUTH EFFECTS FOR ZION NUCLEAR POWER STATION USING CONTIKUUM ANALYSIS APPROACH. T-157 SDIL STRUCTURE INTERACTION (SSI) REVIEW QTR 1, 81 3-9 SEISMIC SAFETY MARGINS J. COSTELLO l REPORTS ASSESSING THE STATE-OF-THE-ART RESEARCH PROGRAM I 0F 551 ANALYSIS MEfHODOLOGY, ACCURACY, (5EI5MIC, ENGINEERING & 3 j UNCERTAINTIES, AND IIEMIZING BENCHMARK SITE SAFETY) i PROBLEMS. i T-168 SRP MEiHODOLOGY COMPARISDNS. THE QTR 1, 31 3-9 SEISMIC SAFETY MARGINS J. COSTELLO l CURRENTLY USED (SRP) METHODOLOGY WILL BE RESEARCH PROGRAM l COMPARED TO OTHERS IN THE SOIL-STRUCTURE (SEISMIC. ENGINEERING & INTERACTION REVIEW. SYSTEMATIC DIFFERENCE 5 SITE SAFETY) WILL BE EXPLORED AND THEIR SIGNIFICANCE I ASSE5 SED. I - 121 - l
T-175 DOCUMENTATICW OF STATE OF AVAILt.BLE QTR 1, 81 MECHANICAL ENGINEERING J. BU2M5 FRAGILITIE5 RELATED INFORMATION. (SEISMIC. ENGINEERIhG 8 SITE SAFETY) T-13C OPERATING SEISMIC SAFETY ANALYSIS QTR 1, 81 MECHANICAL ENGINEEF:!NG J. BURN 5 CODE - SCISMIM. (SEISMIC, ENGINEERING S SITE SAFETY) T-185 STUDY Or SEISMIC ANALYS!$ MODELING QTR 1, 81 STRUCTURAL ENGINEERING C. BURGER TECrNI;UE IN VERTICAL DIRECTION. THE (SEISMIC. ENGINEERING 8 STATE-OF-THE-ERT IN VERTICAL MODELING SITE SAFETY) TECHN!QUE 15 EITHLR TO ASSUME RIGID FLOOR OR TO SIMULATE THE FLOOR FLEXIBILITY BY ONE OR SEVERAL SINGLE DDR SYSTEMS. THIS STUDY 15 in EXAMINE THE ACCURACY OF THIS ASSUNPTIOM AND TO EXPLORE BETTER ALTERNATIVES. T-195 FUEL FRAGMENTATION TESTS WITH NOLTEN QTR 1 81 2-2 POST ACCIDENT HEAT REMOVAL R. WRIGHT RESULTS OF LARGE SPECIMEN TESTING ON (FAST BREEDERS) SEISMIC 5 HEAR TRANSFER. FRAGNENTATICM AND THE FUEL-COOLANT INTER-ACTION POTENTIAL FROM THE CONTACT OF THE MOLTEN FUEL METAL THERMITE WITH SODIUM. i T-194 DEBRIS BED C00 LABILITY AT HIGH DECAY-HEAT QTR 1, 81 2-2 POST ACCIDENTAL HEAT R. WRIGHT I POWERS, ACRR TEST D-4. REMOVAL (FAST BREEDERS) T-197 HUMt4 ERRCR RATE ANALYSIS QTR 1 81 NONE M. CULLINGFORD (RISK ASSESSMENT) T 00 ANALYSIS 4 PREDICTION OF MAJOR FLOODS QTR 1 81 6-3 LIMITING CONDITIONS FOR W. VESELY OPERATIONS (RISK ASSESSMENT) T-208 RELIABILITY DATA MANUAL QTR 1, St NONE J. JOHNSON (RISK ASSESSMENT) T-211 RISK ASSESSMENT OF SPENT FUEL Ih BEDDED QTR 1, 81 6-5 HIGH LEVEL WASTE ISOLATION M. CULLINGFORD SALT (RISK ASSESSMENT) T-242 DISECUILIBRIUM OF URANIUM ORE DAUGHTERS QTR 1, 81 5-23 OCCUPATIONAL EXPOSURE J. FOULKE IN ORE DUST PROTECTION (FUEL CYCLE A ENVIRONMENTAL RESEARCH) T-iS9 NEUTRON DDSIMETRY AT COMMERCIAL NUCLEAR QTR 1, 81 5-23 OCCUPATIONAL EXT ~5URE J. FOULKE SITES PROTECTIDH T-289 ASSESSMENT OF BACIS THERMAL HYDRAULIC QTR 1, 81 .30 N E F. ODAR MODELS IN TRAC-P1A AT BNL (CODE DEVELOPMENT) T-290 ASSESSMENT OF BASIC THERMAL HYDRAULIC QTR 1, 81 NONE F. ODAR MODELS IN TRAC-P1A AT INEL ~ (CODE DEVELOPMENT) if(3d[T-2,i 0 A55E55nENT Or BASIC THERnAt HYDRAUtIC QTR i. Si McNE
- r. ODAR MODELS IN TRAC-P1A AT LASL (CODE DEVELOPMENT)
- 122 -
h f T-292 ASSES 5 MENT EE5ULTS OF COMPLIANCE UNDER QTR 1, 81 NOME R. SHEPA2D 1 PART 73 UPGRADE REGULATICNS USING 5AA (SAFEGUARDS) AND/CR SVAP. T-293 5ATEGUARCS FOR PROLIFERATION RESISTANCE QTR 1, 81 NONE R. ROBINSON FUEL CYCLES. (SAFEGUARDS) i T-294 LOCE L3-1 AND ZION RELATION 5 NIP EXTRA-QTR 1, 81 NONE G. MCPHERSON P0tATION OF SMALL BREAK WITH CONTINU0US (LOFT) I DEPRE55URIZATION TO LARGE PWR EEHAVIOR. T-296 SPENT FUEL CAST-FISSION PRODUCT RELEASE. QTR 1, 31 NONE W. LAHS I (SAFEGUARDS) i T-300 EXTRAPOLATION RESULT 5 FRCM SMALL BREAK QTR 1, 31 NONE C. MCPHERSON WITH CGNTINUOUS 6: PRES 5URIZATION L3-1 (LOFT) TO LAPCE PWR BEHAVIOR. T-18 FISSION GAS RELFASE MODEL VERIFIED. ADDS QTR 2, 81 1-9 FISSION FRODUCT RELEASE C. MARIND TRAN5IENT FISSION GAS RELEASE QUANTI-(FUEL BEHAVIOR) i FICATICN TO FRAP-T ENHANCING ACCIDENT j CALCULATIONS. l T-27 C"BRA-IF ANALY3IS. PROVIDES NRR WITH QTR 2, 81 NONE
- 5. FABIC INDEPENDENT EVALUATION OF UHI PERFORMANCE (CODE DEVELOPMENT) i DURING LOCA.
} T-35 MONITCRING OF RADI0 IODINE FROM CONTAINMENT QTR 2, 81 5-19 TERRESTRIAL RADI0 ECOLOGY P. REED f j ACCIDENTS. TO DETERMINE THE PRACTICALITY (REACTOR ENVIRONMENTAL) 0F USING AVAILABLE CIVIL DEFENSE INSTRU-I MENTS TO ASSESS PUBLIC HEALTH IMPACTS OF l AN ACCIDENTAL RELEASE OF RADICIODINE FROM i NUCLEAR STATIONS. TO ASSESS INSTRUMENTA-i TION CAPABILITIES (PARTICLE COLLECTION EFFICIENCY, PARTICULATE CONTRIBUTIONS, AND INSTRUMLNT REtIABILITY) BY EVALUATING A PORTABLE FIELD RADIDIODINE COLLECTION l SYSTEM AND A DETECTION SYSTEM USING A CDV-700 GM SURVEY INSTRUMENT, ) T-43 TRAC PD2 - IMPROVED P14. QTR 2, 81 1-14 ADVANCED SYSTEM CODE L. SHOTKIN 1 ADVAPCED BWR AND PWR COOLANT SYSTEMS CODE (CODE DEVELOPMENT) TO ESTABLISH MARGINS OF SAFETY. I ? I T-77 HYDR 0 ELASTIC EFFECTS STUDIES. FOR PSS QTR 2, 31 1-15 CONTAINFENT CODE L. SHOTKIN i (PRESSURE SUPPRESSION SYSTEMS) CONTAIM-(CODE DEVELOPMENT) l MENTS T-83 COMNIX CODE (TWD-PHASE). REASSESSMENT QTR 2, 81 2-13 FAST REACTOR SYSTEMS CODES P. WOOD OF COOLABILITY OF SLBASSEMBLY UNDER AND ACCIDENT ANALYSIS NATURAL CONVECTION. (FAST BREEDERS) L T-89 PIPING CDP.PONENTS UNDER CREEP AND QTR 2, 81 2-3 PRIMARY SYSTEM INTEGRITY T. J. WALKER I PLASTIC DEFORMATION FINITE ELEMENT ANALYSIS OF PIPING COMPCHENTS AND COMPARISDN WITH EXPERIMENTAL RESULTS. - 123 - i ._,___- - _ -._,._____,. -__ _ _ _ _ _ _ - -,.. - ~. ~ - - - -._
2m s-4-.-L--aa a hm 4 -m. m m_. T-01 FRAGILITY DESCRIPTICMS F00 COMPONENTS AND QTQ 2, 81 3-10 PLANT STRUCTURE 5 J. BURNS STRUCTURES. DOCUMENTATION OF AN GVECALL (SEISMIC. ENGINEERING & FPAGILITY DESCRIPTION DEVELOPMENT SITE SAFETY) ME1 DDOLOGY. T-102 PELE-IC ISSUED FOR BWR CCNTAINMENTS. QTR 2, 81 1-15 CONTAINMENT CODE
- 5. FABIC CALCULATION OF DYNAMIC RESPONSE OF (CODE DEVELOPMENT)
CONTAINMENT STPUCTURES WITH A COUPLED FLUID /5TRUCTURE PROGRAM. T-112 IODINE TRANSPORT BEHAVIOR WITHIN PWR QTR 2, 81 1-9 FISSION PRODUCT RELEASE R. SHERRY STEAM GENERATOR AND THE SECCNDARY SYSTEM (FUEL BEHAVIOR) UNDER SGTR ACCIDENT CONDITONS. T-129
SUMMARY
OF INTERIM RESULTS. PRELIMINARY QTR 2, 31 3-1 NRC/ STATE REGIONAL EARTH N. STEUER TECTONIC PPOVINCE BOUNDARIES, REGIONAL SCIENCES GEOLOGY, SEISM 0 TECTONICS. TO ACCELERATE (SEISMIC, ENGINEERING & LICENSING. SITE SAFETY) T-160 SENSITIVITY STUDY COMPARING LINEAR QTR 2, 81 3-9 SEISMIC SAFETY MARGINS J. COSTELLO FINITE ELEMENT ANALYSIS WITH CONTINUUM RESEARCH PROGRAM ANALYSIS APPRDACH FOR ZION NUCLEAR POWER (SEISMIC. ENGINEERING 5 STATION. SITE SAFETY) T-161 COMPARATIVE ANALYSIS OF ZION REACTOR QTR 2, 81 3-9 SEISMIC SAFETY MARGINS J. COSTELLO COHTAINMENT BUILDING USING LINEAR FINITE RESEARCH PROGRAM ELEMENT ANALYSIS CONIINUUM ANALYSIS (SEISMIC, ENGINEERING & APPROACH, AND NONLINEAR FINITE ELEMENT SITE SAFETY) ANALYS15. T-174 DOCUNENTATION OF FAILURE PODES FOR QTR 2, 81 MECHANICAL ENGINEERING J. DURNS COMPONENTS AND STRUCTURES FOR A (SEISMIC, ENGINEERING & REPRESENTATIVE PWR PLANT. SITE SAFETY) i l T-176 ESTABLISHMENT OF n CUA5I-DELPHI PROCEDURE QTR 2, 81 NONE J. BURNS FOR HANDLING EXPERT OPINION ON COMPONENT ($EISMIC, ENGINEERING & i FRAGILITIES. SITE SAFETY) ) i T-178 DOCUMENTATION OF EXPERT OPINION ON QTR 2, 81 NONE J. BURNS CCMP0hENT FRAGILITY. (SEISMIC, ENGINEERING E SITE SAFETY) T-182 PROBABILITY OF LOCA INDUCED BY QTR 2, si NONE J. O'BRIEN EARTNGUAKE5. (SEISMIC. ENGINEERING & SITE SAFETY) T-153
SUMMARY
EVALUATION OF DYn.i QTR 2, 31 3-11 PLANT COMPONENT AND D. REIFF EXCITATION TESTS OF A NUCLEAR VALVE. EQUIPMENT BEHAVIOR T-184 ACCURACY OF FINITE ELEMENT METHOD AND QTR 2, 81 STRUCTURAL ENGINEERING C. BURGER LUMPED MASS METHOD IN SEISMIC ANALYSIS. (SEISMIC. ENGINEERING & SITE SAFETY) T-186 RESPON5E COMBINATION METHODOLOGY. QTR 2, 81 3-9 SEISMIC SAFETY MARGINS J. O'BRIEN l RESEARCH PROGRAM (SEISMIC, ENGINEERING & t SITE SAFETY) l - 124 -
i T-197 M9DELING TECHNICUE5 F0Q NONLINEAR eTR 2, 81 3-11 PLANT COMPONENT AND D. GEIFF SYSTEMS. EQUIPMENT BEHAVICR T-202 THREE FULL POWER SMALL BREAK TESTS WITH QTR 2. 81 1-1 LOFT G. MCPHERSON (1) CONTINUOUS DEPRE55URIZATICN. (LOFT) (2) FRE55URE HANGvP. AND (3) REPRE55URIZATION, (4) LOSS A RETURN OF NATURAL CIRCULATION. (5) LONG-TERM HEAT REM 3 VAL. T-204 FULL PCWER TEST WITH LOSS OF STEAM QTR 2, 81 1-1 LOFT G. MCPHERSON 1 LOAD LINE BREAr.: (1) LOSS OF FEEDWATER, (LOFT) (2) LOSS OF PCS FLCW. 1-209 HAZARD 5 TO NUCLEAR POWER PLANTS QTR 2, 81 6-6 LWR RISK ASSESSMENT K. MURPHY (RISK ASSESSMENT) T-232 INSPECTION METHDDS F02 PHYSICAL QTR 2, 81 4-3 SAFEGUARDS INFO. SYSTEMS E. RICHARD 1 PROTECTICN T-298 FULL POWER TEST WITH 1055 0F FEEDWATER QTR 2, 81 NUNE G. MCPHERSON L6-5. (LOFT) T-125 RECENT VERTICAL CRUSTAL MOVEMENTS IN QTR 3. 81 3-2 GEOLOGY AND SEISMIC J. HARBOUR THE EASTERN U.S. WILL FURNISH IMPORTANT CHARACTERISTICS CORROEORATIVE INFDRMATION ON FAULT (SEISMIC. ENGINEERING A MOVEMENT 5, SEISMIC ACTIVITY AND INTRAPLATE SITE SAFETY) EARTHCUAKE MECHANISMS. T-201 SMALL BREAK AT FULL POWER WITH/WITHOUT QTR 3. 81 1-1 LOFT G. MCPHERSON PRIMARY PUMP 5. (LOFT) i T-205 FULL PCWER TEST WITH PRE 55URIZER RELIEF QTR 3, 81 NONE G. MCPHERSON VALVES STUCK OPEN. (LOFT) T-210 FLOOD RISK SYSTEMS ANALYSIS QTR 3, 81 NONE
- 5. STURGES (RISK ASSESSMENT)
A i T-236 EXP. EVALUATICN OF SYSTEMS COMPONENTS QTR 3, 81 5-6 FUEL CYCLE FACILITIES D. SOLBERG f DURING LARGE PRESSURE PULSES EFFLUENT CONTROL T-284 EVALUATION OF PULSED RADAR SYSTEMS FOR OTR 3, 81 NONE C. JUPITER 2 SUB-5URFACE PA0 FILING OF 5HALLCW LAND { BURIAL SITES. T-295 STRATIFIED DEBRIS-BED COOLABILITY, ACRR QTR 3, 81 2-2 POST ACf,IDENT HEAT R. WRIGHT TEST. D-6. REMOVAL (FAST BREEJERS) T-297 LOWER PLENUM SWEEPOUT. COMPLETES QTR 3. 81 NONE W. BECKNER ] TRANSIENT REFILL WORK AT SMALL SCALE. (5YSTEMS ENGINEERING) T-299 FULL POWER TEST WITH LOSS OF PCS FLOW QTR 3. 81 NONE G. MCPHERSON L6-2. (LOFT) l I - 125 - 1
T-301 LARGE PWt RENAVICR DURING SMALL BREAKS QTR 3, 8 5 NONE G. MCPHERSON LEADING TO CEPRE55URIZATION AND (LOFT) REPRESSURIZATICN L 3-2 AND L J-7. T-302 REVIEW OF SMALL BREAK PREDICTICM QTR 3, 81 HONE G. MCPHERSON CAPABILITIES. (LOFT) Y-65 POST CHF ME T TRANSFER CDPRELATION. QTR 4, St 1-3 PWR-BCHT Y. H5U SUFPCRIS OR REPLACES CURRCNT t-4 EWR-BDHT Y. HSU CORRE! ATION5. (5YSTEM) ENGINEERING) T-79 REACTOR PRES 5URE VE55Et SURVEILLANCE QTR 4, At NONE C. SERPAN PROCEDURE. (PRIMARY SYSTEMS INTEGRITY) T-80 IMPROVED LOOSE 7 ARTS MONITORING. QIR 4, 31 1-28 NOISE SURVEILLANCE AND W. FARMER PROVICES INPUT TD REG. CUIDE REQUIREMENT DIAGNOSTICS CHANGE AND TO LICENSING REVIEW. (5YSTEMS ENGINEERING) T-92 EX8 BUNDLE TESTS. CONFIfM ABILITY TO QTR 4, 81 NONE M. PICKLESIMER LIMIT PROPAGATION. (FUEL BEHAVIOR) T-93 CYCLIC CRACK CRCWTH EVALUATICM. QTR 4, 31 NONE M. VAGINS (PRIMARY SYSTEMS INTEGRITY) T-96 SAFETY RELATED OPERATOR ACTING CRITERIA. QTR 4, 81 1-24 HUMAN ENGINEERING W. FARMER MAY IMPACT ENGINEERED SAFETY SYSTEMS (SYSTEMS ENGINEERING) T-123 IMPROVED BAYESIAN METHOD 5 FCR SITE QTR 4, St MONE R. BRAZEE DEPENDENT SPECIRAt. (SEISMIC, ENCINEERING A SITE SAFETY) T-126 FINAL FEPORT OF NEW ENGLAND SEISMO-QTR 4, St 3-1 NRC/ STATE REGIONAL EARTH N. STEUER TFCTONIC STULY. TECTONIC PROVIhCE SCIENCES BOUNDARI[5 DELINEATED, POSSIPLE EARTH-(SEISMIC. ENGINEERING & QUAKE SOURCE AREAS IDENTIFIEG. GEC-SITE SAFETY) TECHNICAL MAPS CCMPLETED. T-133 NEW MACRIt/ UPPER MISSISSIPPI VALLEY QTR 4, 81 3-1 NRC/ STATE REGIONAL EARTH N. STEUER SE15MOTECt0*4IC STUDY FINAL REPORT. SPIENCES TECTCNIC PR041NCE BOUNDARIES DELINEATED. (SEISMIC, ENGINEERING a PR0t# ELE EARTHOUAKE SOURCE AREAS IDENTIFIED SITE SAFETY) GE0 TECHNICAL MAPS CCMPLETED. 1-134 NEMAHA/MIDCONTINENT GRAVITY ANCMALY STUDY QTR 4, 81 3-1 NRC/5 TATE REGTONAL EARTH N. STEUER FINAL REPORT. TECTONIC PROVINCE BOUNDARIES SCIENCES DFLINEATED. PR'0BABLE EARTHOUAKE SOURCE (SEISMIC. ENGINEERING & AREAS IDENTIFIED. GEDTECHNICAL AND SITE SAFETY) 5 FILMIC MAPS CCMPLETED. l l l l - 126 - l
T-145 LABCPATORY 51tfULATION OF TORNADO WIND-QTR 4, 81 3-3 ENVIRONMENTAL STRUCTURAL R. ABBEY LOADS. DESIGN (SEISMIC, ENGINEERING & SITE 5AFETY) T-190 GENERIC SODIUM CCNCRETE INTERATIGM. QTR 4, 81 NONE T. WALKER (FAST EREEDERS) T-195 EXTEhJED DRf-CUT IN SODIUM COOLED CEBRIS QTR 4, 81 2-2 POST ACCIDENT HEAT REMDVAL R. WRIGHT BEDS, ACRR TEST D-5. T-282 EVALUATICN OF NON-INTR 95?VE TECHNIQUES QTR 4, 81 NOME C. JUPITER f0R ROCK MASS CHARACTERIIATICN. T-86 INCCRPCRATE IMPROVED UT IN CODE. IMPROVE 3 QTR 1, 82 1-21 NON-DESTRUCTIVE EXAMINATION J. MUSCARA CONFICENCE IN EVALUATION DF FLCW $!G-(PRIMARY SYSTEMS INTEGRITY) HIFICANCE FRCM ACCCPATE FLCW SHAPE AND SIZE. T-115 EVALUATE METHOD TD COASTAL DISPERSION. QTR 1, 82 NONE R. ABBEY '5EISMIC, ENGINEERING & SITE SAFETY) T-119 5.E. SEILMIC NETWORK. QTR 1, 82 NONE R. BRAZEE (5ETSMIC, ENGINEERING 1 SITE JAFETY) T-212 SELECTED ALTERNATIVES FOR MANAGEMENT OF QTR 1, 82 NONE J. CURRY RADICACTIVE WASTE GASES T-287 BEHAVIOR OF LCFT FOLLCWING LOSS GF QTR 1,.82 1-1 LOFT G. MCP'HERSON FfEDWATER AND DELAYED SCRA9 L6-7 (LOFI) COMPLICATED BY A SMALL BREAK L3-3; LEADS TO PLANT REPRES5URIZATION. T-61 ATWS TRAC VERSION. PROVIDES ADVANCED QTR 2, 8? 1-14 ADVANCED SYSTEM CODE N. ZUBER CAPABILITY TO ANALYZE ATW5 IN LWR's. (CODE DEVELOPMENT) T-72 VERIFY TRAC-2. GIVES CONFIDEECE THROUGH QTR 2, 82 1-17 CODE APPRAISAL
- 5. FA31C COMPARISCN WITH LDFT, GERMAN AND JAPANESE (CDDE DEVELOPMENT)
TEST DATA. T-106 FAST FUNNING TRAC (PF1). QIR 2, 82 NONE L 5HOTKIN (CODE DEVELOPMENT) T-109 LOCA (SING' FDD). QTR 2, 82 1-10 PBF M. PICKLESIMER (FUEL BEHAVIOR) T-120 ANNA, CHID SEISMIC NETWCRK. INSTALLATION QTR 2, 82 3-2 GEDLOGY AND SEISMIC R. BRAZEE CPERATION AND FESE ARCH HISTORY. SFISMIC CHARACTERISTICS ACTIVITY AND RESEARCH
SUMMARY
(SEISMIC, ENGINEERING 3 SITE SAFETY) T-205 FULL PCWER CECL BREAK WITH OFF-SITE QTR 2, 82 1-1 LOFT G. MCPHEE50N PCMER. EARtY CCEE CUENCH SELN IN L2-3 (LOFT) 15 NOT EXPECTED TO OCCUR. - 127 -
f-250 THER?:7.L COMDUCTICITY COCK-MINEDAL QTQ 2, 82 NCNE C. JUPITER (WA5VE MANAGEMENT) T-213 FIRE RISK SYSTEMS ANALYSIS QTR 3, 82 NONE D. STURGIS i (RISK ASSESSMENT) T-55 LOCA THERMAL SHOCK. PROVIDES LICENSING QTR '+, 82 NONE M. VAGIN5 PROCrDURE FOR THEPMAL 5 HOCK ANALYSIS (PRIMARY SYSTEMS INTEGRITY) INCLUDIhG EFFECTS OF WARM PRESCREENING. P l i p: I - 128 -
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