ML20195J603
| ML20195J603 | |
| Person / Time | |
|---|---|
| Issue date: | 03/20/1980 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| Shared Package | |
| ML20195J600 | List: |
| References | |
| NUREG-0435, NUREG-0435-V02-N01, NUREG-435, NUREG-435-V2-N1, NUDOCS 8812020286 | |
| Download: ML20195J603 (123) | |
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= - CURRENTLY BEINO REVIEWED BY l'SER CFFICECS) 3 - AMAITING COMNISSION APPROVAL ON PROPOSED ACTION PLAN FOR CHANGING ECCS RULE ---
2.1 tTST Or TSSUED QIt's RESEARCH INFORMATION LETTERS ISSUED BY RES TO DATE ARE LISTED BELOW IN TABLE 2.1, TOGETHER WITH THE DATE OF ISSUE, A5SOCIATED RESEARCH PROGRAM ELEMENT AND PROGRAM MtNAGER.
DATE RIt M9.
P8cE NO.
IS$"En RESEARCH INFORMATION LETTER TITLE RES DECISION UNIT PROJECT MANAGrg 1
12 03/19/74 ORNL V-S INTERMEDIATE VESSEL TEST RESULTS PRIMARY SYS INTEG.
C. SERPAN 2
13 05/20/74 SEISM 0 TECTONIC MAP OF THE EASTERN UNITED
- SEISMIC, J. HARBOUR STATES ENGINEERING &
SITE SAFETY 3
14 03/07/74 ORNL V-7 INTERMEDIATE VESSEL TEST RESULTS PRIMARY SYS INTEG.
C. SERPAN 4
15 09/10/74 MAP SHOMING RECENCY OF FAULTING IN COASTAL
- SEISMIC, J. HARBOUR 1
SOUTHERN CALIFORNIA ENGINEERING 8 SITE SAFETY S
16 06/23/76 CONFIRMATORY PRESSURE VESSEL TEST UNDER PRIMARY SYS INTEG.
C. SERP?N PMEUMATIC LOADItJG 6
17 10/12/76 A CRITICUE OF THE BOARD-HALL MODEL FOR THERMAL FAST BREEDER R. WRIGHT DETONATIONS IN THE UO2-N A SYSTEN REACTORS 7
13 03/25/76 THE SIMMER CORE FOR ANALYSIS OF ftY.OTHETICAL FAST B2EEDER R. CURTIS CORE DISRUPTIVE ACCIDENTS IN LMF3R'S REACTORS 3
19 01/31/77 DECAY HEAT DATA APPLICABLE 10 LOCA EVALUATIDH FUEL BEHAVIOR R. DISALVO 9
20 O!/14/77 HIGN TEMPERATURE OXIDATION OF ZIRCALOY FUEL FUEL BEHAVIOR M. PICKLESIMER CLADDING IN STEAM to 21 02/25/77 PRESSURE VESSEL FAILURE PROSABILITY PREDICTION RISK ASSESSMENT /
W.
VESELY (OCTAVIA CODE)
PRIMtRY SYS INTEG.
11 22 09/15/77 IEEE NUCLEAR RELIABILITY DATA MANUAL RISK ASSESSMENT J. JOHNSON 12 23 06/16/77 MODIFICATIONS TO PRESSURE VESSEL FAILURE RI5K ASSESSMENT /
W. VESELY PR03 ABILITY PREDICTION (OCTAVIA CODF)
PRIMARY SYS INTEG.
13 24 11/11/77 RESIDUAL STRESSES IN WELDS PRIMARY SYS INTEG.
C. SERPAN 14 25 11/09/77 PNYSICAL SEPARATION CRITERIA FCR ELECTRICAL SYSTEMS ENG.
R. FEIT C A3tE TRAYS (HORIZONT AL OPEN SPACE CCHFIG.)
15 27 12/01/77 CHARACTERIZATION OF BWR FEEDWATER E0ZZLE PRIMARY SYS INTEG.
C. SERPAN CORNER CRACKS 16 23 12/01/77 WARM PKESTRESSING PRIMARY SYS INTEG.
C. SERFAN 17 29 05/05/78 POWER BURST FACILITY (PBF) SsNGLE ROD-POWER FUEL BEHAVIOR R. VAN HOUTEN COOLING MISMATCH (PCM) TEST RESULTS 13 30 11/09/77 FRANTIC COM? UTER CODE RISK ASSESSMENT F. GOLDBERG
-S-
. -=
DATE EJL NO.
PAGE NO.
ISSUED _
RESF4RCH INFCPMATION tETIER TITLE
_RES DECISITN bMIT PRCJECT NAMAGER 19 31 01/31/78 C0 METHOD 9 LOGY ASSES 3 MENT RISK ASSESSMENT J. PITTMAN 20 32 01/24/73 A STUDY OF PHYSICAL PROTECTION EQUIP!1ENT SAFEGUARDS E. FICHARD 21 34 03/24/78 CRITICAL REVIEW OF SODIUM HYDROXIDE AEROSOL RASK ASSESSMENT
.%. CULLINGFORO T0XICITY l
25 36 04/10/78 "EASI" ADVERSARY SEQUENCE EVALUATION MODEL SAFECUARDS R. ROBISSON
{
(COMPUTER GRAPHICS VERSION)
I 24 37 04/10/73 "FESEM" ADVERSARY SEQUENCE EVALUATION SAFEGUARDS R. ROBINSON l
MODEL i
25 33 03/21/73 FRAP-53 Furt SEHAVICR G. MARINO 26 40 04/27/75 THE IMPACT OF OFFSHORE NUCLEAR uENERATING FUEL CYCLE SAFETY D. BARHA ST ATIONS ON RECREATIONAL BEllAVIOR AT ADJACENT AND ENVIRONMENTAL COASTAL SITES.
EFFECTS 27 41 06/02/73 "3EACON/MDD 2" CODE DEVELOPMENT S. FABIC 28 42 05/09/73 "MELT / CONCRETE INTERACTIONS" FUEL BEHAVIGR R.
DISALVO 29 44 06/07/78 "FUEL ROD ANALYSIS COMPUTER CODE:
FRAP-T3" FUEL BEHAVIOR H. SCOTT 4
1 30 46 06/28/73 PHASE I FINAL REPORT, "BARRIER PENETRATION SAFEGUARDS R. ZIMMERMAN 0;TA BASE"; 0F STUDY. "ASSISTANCE-PH(SICAL PROTECTION ASSESSMENTS" 31 47 07/10/73 ASSAY OF STtHD.4RD REFERENCE MATERIAL (SRM)950 SAFEGUARDS R. SHE.*ARD 32 48 CS/03/75 IMPROVEMENTS IN THF AEROSOL BEHAVIOR CODE FC.i FAST BRFEDER J.
LARKINS RADIOLOGICAL ASSESS.sENTS OF LMFER'S REACTORS 33 49 03/03/73 PLUTONIUM ACCIDENT CONTAINER PROGRAM RESEARCH, FUEL CYCLE SAFETY W. LAHS DESIGN AND DEVELOPMENT.
AND ENVIRONMENTAL EFFECTS 34 50 C3/03/73 NUCLEAR DECAY DATA FOR RADIONUCLIDES OCCURRING FUEL CYCLE SAFETY J. FOULKE IN ROUTINE RELEASES FROP NUCLEAR FUEL CYCLE AND ENVIRONMENTAL FACILITIES EFFECTS 35 51 09/15/73 SFACTOR: A COMPUTER CODE FOR CALCULATING DOSE FUEL CYCLE SAFETY J. FOULKE EQUIVALENT TO A TARGET ORGAN PER MICR0 CURIE -
AND ENVIRONMENTAL j
DAY RESIDENCE OF A RADIDNUCLIDE IN A SOURCE EFFECTS ORGAN 36 52 09/27/73 EVALUATION OF GENERAL ATOMIC CODES:
OXIDE-3, FAST BREEDFR J.
LARKINS SORS. TAP, AND PECA.
REACTGRS 37 S4 09/23/78 LOFT REACTOR SAFETY PROGRAM RESEtRCH RESULTS LOFT G. MCPHERSON THROUGH OCTOBER 1,
1973..
"ATE RIL NO.
PtGE NO.
T SSt)?D RESEARCH TNFOPMATION LETTER TITtE RES DECISION UNIT PROJECT MANAGER 33 56 10/13/73 RESULTS OF THE INITIAL SERIES OF ACPR EXPERI-FAST DREEDER R. WRIGNT MENTS ON PROMPT-BURST ENERGETICS WITH FRESH REACTORS OXIDE FUEL 39 57 11/27/78 RELAP-4/ MOD 6 CODE DEVELOPMENT S. FABIC 40 53 12/1B/73 THE COMPUTER CODE BRENDA - A COMPUTER PROGRAM FAST BREEDE2 P. WOOD FOR THE DYNAMIC SII-ULATION FOR A LIQUID METAL REACTORS FAST BREEDER REACT.)R PLANT 41 59 12/19/78 L ABOR ATORY TESTIN3 PROCEDURES TO DETERMINE THE
- SEISMIC, N. STEUER CYCLIC STRENGTH OF SOILS ENGINEERING &
SITE SAFETY 42 60 12/23/78 CRITICAL EXPERIMENT PROGRAM FOR NEUTRONICS CODE FAST BREEDER P. WOOD VERIFICATION REACTOR?.
43 62 01/10/79 SUPER SYSTEM CCDE A COMPUTER PROGRAM FOR FAST BREEDER P. WOOD DYNAMIC SIMULATION OF LMFBR POWER PLANTS REACTORS 44 63 01/04/79 RADIATION DOSE TO CONSTRUCTION WORKERS AT FUEL CYCLE SAFETY J. FOULKE
}'
OPERATING NUCLEAR POWER PLANT SITES AND ENVIRONMENTAL EFFECTS 45 64 32,11/79 THE CONCEPT COMPUTLR CODE 3 CAPITAL COST 5 FUEL CYCLE SAFETY D. BARNA FOR BOILING WATER REACTOR PLANTS AND ENVIRONMENTAL EFFECTS 46 65 02/12/79 THE EFFECTIVENESS OF CABLE TRAY COATING SYSTEMS R. FEIT MATERIALS 3 BARRIERS IN RETARDING THE ENGINEERING COMBUSTION OF CABLE TRAYS SUBJECTED TO EXPOSURE FIRES 3 IN PREVENTING PROPAGATION BETWEEN CABLE TRAYS (HORIZONTAL OPEN SPACE CONFIGURATION) 47 67 D5/19e79 INREM II: A COMPUTER IMPLEMENTATION FUEL C"CLE SAFETY J.
FOULXE OF RECENT MODELS FOR ESTIMATING THE AND enc RONMENTAL DOSE EQUIVALENT TO ORGANS OF MAN FROM EFFECTS AN INNALED OR INGESTED RACIONUCLIDE 43 59 04/03/79 A TECTONIC OVERVIEW OF THE CENTRAL
- SEISMIC, N. STEUER MIDCONTINENT ENGINEERING 8 SITE SAFETY 49 70 04/04/79 IN VITRO DISSOLUTION OF URANIUM FUEL CYCLE SAFETY J. FOULKE PRODUCT SAMPLES FROM FOUR URANIUM AND ENVIRONMENTAL MILLS EFFECTS 50 71 04/06/79 CRITICALITY SAFETY GUIDANCE FUEL CYCLE SAFETY D. SC1DERG AND ENVIRONMENTAL EFFECTS 51 67 04722/79 THE CONCEPT COMPUTER CODE AND CAPITAL FUEL CYCLE SAFETY D. BARNA COSTS FOR PRESSURIZED WATER REACTOR AND ENVIRONMENTAL PLANTS EFFECTS -.
DATE RIL NO.
PAGE NO.
ISSUED RESEARCH INFORMATION LETTER TITLE
_RES DECTSION UNIT PROJECT MANAGER 52 73 04/23/79 EARTHQUAKE INTENSITY SCALE
- SEISMIC, R. BRAZEE ENGINEERING &
SITE SAFETY 53 74 05/16/79 DEBRIS-BED CO3 LABILITY LIMITS, RESULTS FUEL CYCLE SAFETY R. HRIGHT FROM IN-CORE TESTS D-1, D-2 AND D-3 AND ENVIRONMENTAL EFFECTS 54 75 C5/15/?9 THE SET EQUATION TRANSFORMATION SYSTEM RISK ASSESSMENT W. VESELEY 55 77 05/29/79 THE CONCEPT COMPUTER CODE AND CAPITAL FULL CYCLE SAFETY D, BARNA COST FOR HIGN ANO LOW SULFUR CDAL AND ENVIRONMENTAL PLANTS - 1200 M4E EFFECTS 56 75 07/25/79 EFFECTS OF NUCLEAR POWER PLANTS ON FUEL CYCLE SAFETY C. PRICHARD COM' UNITY GRCWTH AND RESIDENTIAL PROFERTY VALUES AND ENVIRONMENTAL EFFECTS 57 79 03/10/79 SMALL SCALE ECC BYPASS RESEARCH RESULTS SYSTEMS ENGINEERING A.
SERKIZ 53 30 03e29/79 COMP
..,0N OF SIMULATION MODELS FUEL CYCLE SAFETY P. REED USED IN ASSESSING TifE EFFECTS AND ENVIRCMMENTAL OF POWER PLANT INDUCED MORTALITY EFFECTS ON FISH POPULATIONS 59 31 09/21/79 TRANSIENT FUEL ROD BEHAVIOR FUEL BEHAVIOR G. MARINO l
CODE: FRAP-T4 60 32 10/12/79 SEISMICITY AND TECTONIC
- SEISMIC, N. STEUER RELATIONSHIPS OF THE NEMAHA ENGINEERING &
UPLIFT IN OXLA!!OMA SITE SAFE 7Y 61 83 10/11/79 MOLTEN SODIUM INTERACTION FAST BREEDER T. WALKER WITH EASALT CONCRETE REACTORS 62 84 13/01/79 NEW MADRID SEISM 0 TECTONIC STUDY
- SEISMIC, H.
STEUER ENGINEERING &
SITE SAFETY
- l 63 35 11/08/79 LOFT REACTOR SAFETY PROGRAM LOFT RESEARCH RESULTS FRCf! NUCLEAR G. MCPHERSON i
l LGSS-OF-COOLANT EXPERIMENTS L2-2 AND L2-3 64 86 11/05/79 A REVISED AND AUG1ENTED LIST OF
- SEISMIC, N.
STEUER EARTHQUAKE INTENSITIES FOR ENGINEERING &
KANSAS, 1867-1977 SITE SAFETY i
i
}.
DATE RVL N9.
PAGE NO.
IS Sl1E D RFSFARCH INFORMATTON LETTER TYTLE RES DECISION UNIT PROJECT MANAGER 65 83 11-05-79 RECCNNAISSANCE BEDROCK GEOLOGIC MAP OF
- SEISMIC, H. STEUER MARLBOROUS!! CUADRANGLE, MA AND RECOHt1AISSANCE ENGINEERING &
BEDROCK GEOLOGIC MAP OF SdRE,4SBURG QUAD-SITE SAFETY RASGLE, MA 66 89 11-06-79 A STUDY OF THE REGIONAL TECTONICS AND
- SEISMIC, H. STEUER SEISMICITY GF 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-C6-79 REFLOODING OF SIMULATED PWR CORES AT LOW SYSTEMS ENGINEERING L. THOMPSON FLOW RATES 63 91 11-11-79 STRUCTURAL INTEGRITY OF WELD REPAIRED PRIMARY SYSTEMS M. VAGINS PRESSURE VESSELS INTEGR!TY 69 92 11-19-79 AN INTEGRATED GE0 PHYSICAL AND GEOLOGICAL
- SEISMIC, N. STEUER STUDY OF THE TECTONIC FRAMERORK OF THE ENGINEERING 8 3STH PARALLEL LINEAMENT IN THE VICINITY OF SITE SAFETY ITS INTERSECTION MITH THE EXTENSION OF THE NEtJ MADRID FAULT ZONE 70 93 11-19-79 SEISMICITY AND TECTONIC RELATIONSHIPS OF
- SEISMIC, N. STEUER Tl!E NEMAHA UPLIFT IN OK, PART II, JAM. 1979 ENGINEERING 1 SITE SAFETY 71 94 11-19-79 REGIONAL TECTONICS AND SEISMICITY OF SEISMIC H.
STEUER EASTERN NESRASKA ANNUAL REPORT - JUNE 1, ENGINEERING 2 1977-MAY 30, 1978 SITE SAFETY 72 95 tt-16-79 NEW ENGLAND SEISM 0 TECTONIC STUDY ACTIVITIES
- SEISMIC, N. STEUER DURING FISCAL YEARS 1977 AND 1978 LNCINEERING 3 SITE SAFETY 73 96 11-16-79 IN VIRO COUNTING AT SELECTED URANIUM MILLS FUEL CYCLE SAFETY J. FOULKE AND ENVIRONMENTAL EFFECTS 74 97 11-16-79 STEADY-STATE FUEL RCD BEHAVIOR CODE:
CODE DEVELOPMENT G. MARINO FRAPCON-1 l
75 93 11-27-79 INVENICRY, DETECTICN, AND CATALOG OF
- SEISMIC, H. STEUER OKL A!!C:1A EARTidQUAKES AND EARTHQUAKE MAP OF ENGINEERING &
OKLAHOMA, MAP GM-19 SITE SAFETY 76 99 12-28-79 ANNEALING OF IRRADIATED REACTOR PRESSURE PRIMARY SYSTEMS C. SERPAN VESSELS INTEGRITY 77 100 12-2L-79 ORIGIN OF SURFACE LINEAMENTS IN HEMAHA
- SEISMIC, N. STEUER COUNTY, KANSAS ENGINEERING &
SITE SAFETY
-9
DATE EIL NO.
PAGE NO.
ISSUED RESEARCH INFORMATION L ETTER TITL E RES DECISION UNIT PROJECT MANAGER 73 101 12-28-79 VERTICAL LCADS IN MARK I CONTAINMENT TORUS CODE DEVELOPMENT R. CUDLIN 79 102 12-23-79 EVALUATION OF SEISMIC QUALIFICATION TESTS SEISMIC.
B. BRCMZIN FOR NUCLEAR POWER PLANT EQUIPMENT ENGINEERING &
SITE SAFETY 80 103 01-15-80 DETERMINING EFFECTIVENESS OF ALARA DESIGN FUEL CYCLE SAFETY J.
FOULKE AND CPERATIOli4L FEATURES AND ENVIRONMENTAL EFFECTS 81 104 02-28-30 IRRADIATED FUEL DISRUPTION UNDER LOF ACCIDENT FAST BREEDER R. WRIGHT CONDITIONS: RESULTS OF ACPR TEST SERIES REACTORS
{
FD-1 AND THE FISGAS CODE 32 105 02-29-80 THREE MILE ISLAND TELEPHONE SURVEY:
FUEL CYCLE SAFETY C. PRICHARD PRELIMINARY REPORT ON PROCEDURES AND AND ENVIRONMENTAL FINDINGS, AMD THE SOCIAL AND ECONOMIC EFFECTS EFFECTS OF THE ACCIDENT AT THREE MILE i
ISLAND: FINDINGS TO DATE.
4 e -..
2.2 POVENT?AL AP*LICS9ftTTV 0F RESEARCH RESULTS IN THE PEGUtATORY PRCCFM CCMdENTS ARE OFFERED FROM RES AND FR0f1 CGGNIZANT USER OFFICES ON THE HATURE OF THE REPORTED RESULTS AND THEIR POTElsTIAL 'OR ACTUAL) APPLICfBILITY TO A!iY PART OF fME REGULATORY PROCESS, INCLUDING:
A) TECHNICAL DATA SUPPORTIh3 LICENSING REVIEf45 02 RE4ULATORY JUDGMENT, B) EVALUATION CODES, C) INFORitATION APPLICABLE 10 REGULAT3RY GUIDES OR STANDARDS, AND D) INFORMA;t0N SUPPORTING JUDGMENTS REGARDING REGULATCRY POLICY.
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PRMMM O H I C E CO'"'E 4 T S ON POTENIIAL tI T IL IZ A T ION OR VALUE OF RESEAPCH PFSut15 IN IHE REGULATORY PROCESS PIL a:
1 p TE IS$yFn: 03/19/74 RFS OFCISION UNIT: PRIMARY SYSTEMS INTEGRITY RIL TITtf2 ORNL f-5 INTERMEDIATE VESSEL TEST RESULTS (ORNL HS$f PROGRAM) f SPON90 RING OFFICEfS): RES ER_G :
1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR:
C. SERPAN 4
RES CF**F.fiLP EXPERIMENTAL EVILENCE THAT A "SAFE" FAILURE MODE CLEAK-BEFORE-BREAK) FOR REACTCR PRESSURE VESStiLS MAY EXISI,
- tS REPGRTED FOR THE FIRST TIME.
TESTS ON THE V-S INTERMEDIATE VESSEL AT ORNL DEMONSTRATED THAT A LEAK CCCURRED INSTEAD OF A FRACTURE BREAK WHEN THE VESSEL MAS HEATED TO 190 F AND PRESSURIZED TO 26,600 PSI.
FURTHER ANALYTICAL STUDY TO GENERALIZE THIS RESULT IS UNDERWAY.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PQ$T_.P1(_.ACB VJILFJ EFVEM lit 1 D COMPIETED HELD HEtD ISSUED.
IMPLEENTED OFFICE RESPONSIBLE......... NRR/SD NRR SCHEDULED CO*PLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED N/A ACTUAL CC".PLETION DATE.....
03/21/73 N/A i
hES Cr""5N15_0N 09/P9/77& y/?t/78, S.
GRIMel O
PEM G'lLP.AhlC AS GJLILE CQATORY PROOFM:
THE RESULTS OF THIS TEST ADDEIP TO THE STAFF'S UNDERSTANDING OF FRACTURES ORIG *NAIING AI FLANS IN N0ZZLE CORT;ERS OF HEAVY SECTION STEEL VESSELS'.
IF LEAK-BEFORE-BREAK COULD BE DEM0hSTRATED UNDER ALL REASONADLY CONCEIVA3tE Cor:DITIONS AND CIRCUMSTANCES, IT WOULD DEMONSTRATE THAT CURRENT LICENSING POSITIONS ARE VERY CCNSERVATIVE.
MSCRIRF,_J"M C_T_OF_PD_U1JJ:
W!!ILE ADDING TO OUR UNDERSTANDING OF VESSEL FAILURE MODES, THERE HAS BEEN NO DtFINIIIVE IMPACI ON LICthSING AT THIS TIME.
4 1
CM-'*p T S / P[M A R K S : THE OPINIONS EXPRESSED IN THE RIL REGARDING LEAK-BEFORE-BREAX ARE ENCOURAGING BUT NEED ttRIHER SUSSIANTIATION BEFORE CURRENT LICENSING POSITIONS CAN BE RELAXED.
i 1Q_CQEE N TJ_T4 09/13/7 E P.
RANDA_t1 11415 PROCRAM HAS GEMtR AT ED 5"BST ANTI AL DATA USED IN PREDICTING IRRADIATION EMSRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS UHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES AND LOADINGS.
THIS PROGRAM NAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR SO AND CONTRIBUTED DATA USED IN THE ASME CODE UHICH WAS INCCRPORATED INTO APPEND 1X G.
IT HAS ALSO PROVIDED PART OF TPE DATA IASE USED IN REGULATORY CUIDE 3.99 CONCERNING THE EF.~ECT OF COPPER IMPURITIES ON SENSITIVITY OF STEEL TD IRRADIATION AND IS EXPECTED TO PROVIDE A SUSSIANTIAL IdPUT FOR FUTURE REVISIONS TO REGULATORY CUIDE 1.99.
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PROGR A*1 0F FICE CC:' TEN TS UN POTtNTIAt UTILIZATION OR VALUE OF RESEARCH LEstit ?S T Ct THE REGt#ATORV PROCESS
_RJES DECISTO'i UNIT: S EI SMI,., ENGI*tEERING A SITE SAFETY RTt 2:
2 DATE ISSUEn: C5/20/74 RIL TITLE: SEISMO TECTONIC MAP OF THE EASTERN UNITED SIATES
$PONSOPING orFICE(S): RES EPg: 3-0 GEOLOGY & SEISMIC CHARACTERISTICS TESEARCH PROJECT MGR:
J. HARBOUR 1
EliS_CD'"d" N TP RIL 2 REPCRTS A CCMPILATIGN OF EARTHQUAKE FAULT DATA FOR THE EASTERM UNITED STATES.
THESE DATA ARE USED BY LICENSE AFPLICANTS IN PREPARATICH OF MATERIAL FOR PRELIMINARY SAFETY ANALYSIS REPORTS.
USER DISCUSSIJN POSITION CO*1 MISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEl-ING RELEASF RESULTS ED$1_El W CTIVID E$
37V_1F'!
EL D COMPLETED El.p HELD ISSLc0
]MPI. F,M EN T ED OHICE RESPONSIBLE......... NR"/SD NRR SCHEDULED CD't?tETION PATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSC'tED 1974 ACTUAL CCMPLETICM DATE.....
09/27/77 1974 Hg3_C000ENil pN 31/7?/77. R.J FliSE:
MSE IPDf_P d14.LIf_T4 10 R E 0 0'dQ R_Y_PROCF,SS : THIS RESEARCH DEVELOPED LIMITED SEISMOTcCTCNIC PROVINCES ttAPPItiG UF THE EASTERN UNITED STATES AIMED AT IMPLEMENTING APPENDIX A REQUIREMENTS FOR DETERMINING SEIS:1IC DESIGN FOR N"CLEAR FACILITIES.
F QQRJBE I"PJCT OF PSSUjT1: THE RESEARCH ADDED VERY LIT" LE TO OUR KNOWLEDGE OF EARTilQUAKE PROCESSES IN THE EAsitRti UNITED SIAlt;, WHICH COULD BE USED TO IMPLEMLNI APPENDIX A REQUIREMENTS.
S C fMT523Q3g15: CURRENT EARTHQUAKE FIOSITORING AND TECTONIC STUDIES SHOULD BE SYNTHESIZED TO DEVELOP AN APPLICADLE SEISM 3 TECTONIC M/.P.
E CCCFNTS PN 0a/13/73.
G.
STEMPARK 4
IdIS STUDr 15 PRESENTLY USt.D AS GUIDANCE IN ASSESSING TECTONIC PROVINCES AND SEISMIC 1TY IN LICEriSING CASE REVIEMS I*f THE EASTERN UNITED STATES.
THI'i SIUDY. AS WELL AS ONGOING I
STUDIES. MILL EVENTUALLY 3E USED TO OFFER CUIDANCE ON A REG 1CMAL LASIS THROUGH REGUtATORY GUIDES.
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f'R U WJ1 0 F F I C E C 0""E N T S 0" P0 r dr i el U11LIZATIU9 OR VALUE UF RESEARCH RESULTS IN TrtE REGUL ATORY PROCESS
, ef t 2:
3 DATE ISst?Di 33/07/74 R_F S DFCISION UNIT: PRIMARY SYSTEMS INTEGRITY ETJ_T I T L E -
ORNL V-7 INTERMEDIAIE VESSEL TEST RESULTS (ORNL HSST PROGRAM)
SPONS07IV. QFFT;FfS): RES RPG:
1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR:
C. SERPAN PES _C0""ENTS:
ADDITIONAL EXPERIMENTAL EVIDENCE IS DEMONSTRATED THAT A "SAFE" FAILURE MODE FOR REACTOR PRESSURE VESSELS MAY EXIST, NAMELY. "LEAK-EEFORE-BREAK".
IllESE RESULTS SUPPORT EARLIER RESULTS REPORTED IN RIL S t.
IN THIS TEST. THE FLCi! MAS 13-INCHES iONG AND S-INCHES DEEP IN TliE 6-INCH THICK FRESSURE VESSEL WALL.
Tlid VESSEL MAS ABLE TO SUSTAIN TWICE THE DESIGN LOAD PRIOR TO PENETRATION Of THE FLOW THROUGH THE REMAINING THIN LIGAMENT OF VESSEL MATERIAL. T!!IS RESEARCll CONTRIEUTES TO OUR UNDERSTANDING OF VESSEL FAILURE MODE.
THIS INCREASED UNDERSTANDIhG SUPPORTS THE NRR STAFF IN ITS REVIEW EFFORTS USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EG$TJIL ACTIvlTI Q S QT_F y HEtD COMPLETFJ HELD HELD ISJ_Uf2_
IMPLEM(NTRQ U!FICL RESPC,SI3tt......... NR2/SD NRR SCHEDULED CO~PLETION DATE.. --
UtiSCHED UNSCHED UNSCHED UNSCHED UNSCHED N/A ACTUAL COMPLETION DATE.....
09/07/77 N/A L;RR_CC""[NTS ON 01/07/77.
B.
CRIMU :
PL5CRlFE_sPEL]CATJoN 10J tgt%ATEP H RACF.S$:
THE TEST RESULTS HAVE BEEN USEFUL IN ESTIMATING THE MARGIN 0~ SAFETY, Ili TERMS OF FLAW SIZE, FOR A LARGE FLAM IN A HEAVY SECTION STEEL VESSEL TESTED AT UPPER SHELF TEMPERATURES. WHILE CRACK ADVANCEMENT Tif RCUCH Tif E RE'1AINING WAL L LIGAMENT WAS IN THE STABLE TEARING MODE. PRODUCING A LEAK. AND PRESSURE TO CAUSE LEAKAGE WAS MORE INAM TWICE Tl!E VESSEL DESIGil PRESSURE, THESE RESULTS MAY HAVE BEEN INDUCED BY THE SPECIAL EFFECTS UF THE TEST GEDMETRY.
DF'tCIU 3 c J"PACT Oc B RU_1T$: WHILE ADDING TO OUR UNDERSTANDING OF VESSEL FAILURE MODES, THERE ilAS BEEN HO DfFINITIVE IM?ACT CN 11 CESSING AT THIS TIME.
_C0"f2qS/REFAc3KS:
THE OPINICNS EXPRESSED IN THE RIL REGARDING LEAK-BEFORE-BREAK ARE ENCOURAGING BUT NEED FURTHER SUBSIANIIAIION EEFORE CUR 7ENT LICENSING POSITIONS CAN BE RELAXED.
ED_C_0""ENTS ON_C9/13/73nP. R A NDitJ 1HIS PRCCRAM HAS SENERAlt.D SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION ^ EMBRITTLEMENT AND MARGINS TO FAILURE CF REAC'CR PRESSURE VESSELS UHEN FLAWS ARE PRESENT, CONSIDi! RING MATERI AL PRCPERTIES AND LOADIMGS.
THIS PROGRAM HAS PROVIDED IMPUT TO NRC RCquale M NTS FOR FRACTURE TOUGMNESS OF PRESSURE *.'ESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR 53 ND CONTRIBUTED DATA USED IN THE ASME CODE b!ICH UAS INCORPORATED INTO APPENDIX G.
IT HAS AL ' FROVIDED PART OF THE DATA BASE USED IN REGULATORY GUIDE 1.99 CONCERNING THE EFFECT OF !. 'PER IMPURITIES ON SENSITIVITY OF STEEL TO IRCADI AT!ON AND IS EXPECTED TO PROVIDE A SUBS 1 ANTIAL Ih, 'If FOR FUTURE REVISIONS TO REGULATORY GUIDE 1.99.
P90GE A*1 M FICE C0prgt5 c+r Forcnt ist yrytypAgion og yAtyp of RESEAMH RESUtiS IN THE REGULA10RY PROCESS RIL *:
4 D AT E ISSt'F9:
09/10/74 EES DECTSIPH UNTT SEISMIC, ENGINEERING & SITE SAFETY 31L TITLE:_ MAP SHOMING RECENCY OF FAULTING IN COASTAL SOUTHERN CALIFORNIA S PONS 0?INf; CFFICE(5): RES R*Q: 3-2 GEOLOGY & SEISMIC CHARACTERISTICS RESEARCH PROJECT MGR:
J. HAR3OUR S E Lf_pr_1LN_T_S r RIL 4 COVERED FAULTING IN C0ASTi.L SCUTHERN CALIFORNIA AND IS REFERRED TO BY NRR WHEN CONSIDERING SEISMIC SAFETY QUESTIONS REGARDING PLANIS IN THE AREA COVERED.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFIt:G BRIEFING RELEASE RESULTS P05LP1(_aQT LUT1FJ EFJEt; HEtD COMPIETED HELD HEtD ISStf ED IMPL E;*FNT f Q OFFICE RESPCNSIBLE......... ti2X/SD NRR SC:tEDULED COMPLETION DATE.
UNSCHED UNSCIIED UNSCHED UNSCHED UNSCHED 1975 ACTUAL COMPLETION DATE.....
09/27/77 1975 E:LS.X_OEt'11_0 *8_t9B 7 / 7 7, L I)[N_YSE P MGP'!E APPLIC QI_qN TO k'EGt!! A IO"Y PPACF,SS: TO BE USED IN SELECTING SITES FOR NUCLEAR FACILITIES AND FCR REGIONAL IN?tri 10 STAFF REVILM OF NUCLEAR FACILITY APPLICATIONS.
RQQPy 1 2 PSCT OF EFS9 QS: !!CRK PROVIDED NEW KtiOWLEDGE OF THE DISTRIBUTION OF ACTIVE FAULTS IN CUAhiAL CALIFORNIA. HAS PROVIDED INPUT TO OUR REVIEM OF DIABL3 CANYON AND OTHER SITES.
C0" ENTS/5'EMatrS: WORK HAS LARCELY BEEN COMPLETED. NO EXTEt4SION IS ANTICIPATED.
SD C0Z MTS_f_ W G f_Z$2_(._R Eft [nRK INE C3JLCT OF TH15 S!UDY WAS 10 FRODUCE DATA DESCRIBING THE RELATIONSHIP BETWEEN EARTHQUAKE MAGNITUDES AND DIMENSIONS OF FAULT DISPLACEMENT. TitIS STUDY, ALOMG WITH OTHERS. IS USED AS GUIDANCE IN ASSESSIt:G "CAPABLE FAULIS" IN LICENSING KEVIEWS. RIL "4 PROVIDED A NAP WHICH UCPATED INFORMATICN ON FAULTING IN COASTAL SOUTHERN CALIFORNIA..
PDCGPAM OFFICE CC%7ENTS 04 POTENTIAL UTILIZATION OR VALUE OF RESEePCH RESULTS IN THE REGULATORY PROCESS i
RTL 8 5
DATE ISSUED: 06/23/76 RES DECISION UNIT: PRIMARY SYSTEMS INTEGRITY t
RIt TTTLE: CONFIRMAT02Y PRESSURE VESSEL TEST UNDER PNEUMATIC LOADING G'RNL HSST PROGRAM)
SPONSORING OrrICE(S):
RES PRQ:
1-20 VESSEL INTEGRITY PESEAPCH PP0 JECT MGR:
C. SERPAN R13_C_005EN TS : RIL's 1,
32 5 REPCRTED THE RESULTS FROM THE HEAVY SECTION STEEL TECHNOLOGY PROGRAM WHICH SHOWED THAT IHE ANALYTICAL METHODS FOR PREDICTING FLAW INITIATIGN AND CRACK ARREST IN REACTOR PRESSURE VESSELS HAVE BEEN !!EL L VALIDAT ED.
THESE VALIDATED ANALYTICAL METHODS (LINEAR ELASTIC FRACTIIRE MECHANICS AND ELASTIC-PLASTIC FRACTURE MECHANICS) ALLCM A PREDICTION OF THDSE CONDITIONS UNDER WHICH FLAWS IN PRESSURE VESSEL STEELS CAN CAU FAILURE OF THE VESSEL.
THIS PPOVIDES THE NRR STAFF WITH A NEW TECHNIQUE FOR SETTING SAFE LIMITS FOR NORMAL OPERATION AND F0% ABHORMAL AND ACCIDENT SITUATIONS TO REDUCE THE LIKELIHOOD OF PRESSURE VESSEL FAILURE.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BEIEFING RELEASE RESULTS UST P TLA C_TlV UlE3 B FV_IFM HELD COMPL ETFD HELD HELD 0FFICE RESPONSIP.LE......... NRR/SD ISWED_
IMPLFEENTED MRR SCHEDULED COM?LETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COM?tETION DATE.....
12/16/77 1976
(
tjELC E ENT LON 12/16/77.
J, ENTGNY:
pl1Cf R APP 11CU TON 10 RGTR A TWY PROCF15: THE INTERMEDIATE TEST VESSELS, ITV-7 AND ITV-7A UNDER SUSTAINED l
LOADING CEMONSTRATED Id4T EDIH INE VESSELS RESPONDED TO PNEUMATIC LOAD'NG ESSENTIALLY AS THEY HAD TO HYDRAULIC LOADING.
EARLILR EXPERIENCE OF A RAPID CRACK EXTENSION IN THE GA! PIPE LINE WAS INHIBITING THE ACRS COM:11TTEE ASOUT THE RESULTS OF ITV-TEST UNDER HYDRAULIC LOADING. HEt:CE. THE PNEUMATIC LOAD TEST WAS SUGCESTED.
_DESCS13L_T"**CT OF RESQLTS: THESE TESTS SHOU THAT THE VESSELS UNDER SUSTAINED LOADING BEHAVED SIMILARLY TO i
HYDRAULIC LOADi!U. AND TNE RESULTS ARE APPLICABLE TO THE EVALUATION OF THE BEHAVIOR OF REACTOR PRESSURE VESSELS I
UNDER SUSTAIMED LDAD.
THE TWO VESSEL TESTS THAT PUPTURED, WITHSTOOD PRESSURE 2.15 TO 2.75 TIMES DESIGN PRESSURE.
THE TEST PRESSURES RERE ALOVE ASME B & P V CODE ALLONABLE FOR FAULTED CONDITIONS.
}
$?9MENTS/PEM*, Prs TESTS DEM3NSTRATE THE LEAK WITHOUT BURST.
50 COSEMTS ON 9 9/ ? 3 / 7.%_fd NI(Q IHIS PROGRMi HAS GENERATED SU;s1At:TIAL DATA USED IN PREDICTING IRRADIATION EM3RITTLEMENT AND MARGINS TO FtILURE OF REACTOR PRESSURE VESSELS UHEN FL AWS ARE PRESENT. CONSIDERING MATERIAL PROPERTIES AMD 104CINGS.
THIS PROGRAM HAS PROVIDED INPUT TO NRC REQUIREf1ENTS FOR FRACTURE TCUGHNESS OF PRESSURE VESSEL MATERIALS DESCRISED IN APPENDIX G TO 10 CFR SS AND CONTRIEUTED DATA USED IN THE ASME CCDE UNICH UAS INCORP99?.TED INTO APPENDIX G.
IT ifAS ALSO PROVIDED PART OF THE DATA 2ASE USED IN RECULATORY CUIDE 1.99 CONCERNING THE EFFECT OF COPPER IMPURITIES ON SENSITIVITY OF STEEL TO IIRADIATIG?4 AN) IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FCR FUTURE REVISIONS TO REGULATCRf CifIDE t.99.
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PRCM AM C F F I C E C("*tif '415 GN POTENTIAt UTILI7Afl0N OR VALUE OF RESEAPCH RESULTS IN T H REGULATORY PROCESS PIL 7:
6 DATE TSSUED:
10/12/76 PES DECISION UNIT: FAST BREEDER REACTORS
$_LL TTTLE: A CRITIQUE OF THE BOARD.'.ALL MODEL FOR THERMAL DETONATIONS IN THE UO2-NA SYSTEM EPONSrPT % OrFICE(S): NRR RPC:
2-6 ACCIDENT ENERGETICS RESEARCH PROJECY MGR:
R. WRIGHT FIS _C O**MFM T S : THIS CRITIQUE OF THE BOARD-IIALL THEORY FO't THERMAL EXPLOSIONS IN A URANIUM OXIDE-SODIUM SYSTEM REItJCRCED IHE BELIES THAT SUCH DETONATIONS ARE OF LO'.! PROBASILITY. THE CRITIQUE HAS 3EEN USED AS AN AID IN ANSMERItiG ACRS CONCERNS A30UT TPE POSSIBILITY OF ENERGETIC FUEL-COOLANT INTERACTIONS IN ADVANCED REACTORS AND MAS EMPLOYED IN SUPPORTING NRR'S POSITION IN TiiE EVALUATION OF THE PRELIMINARY SAFETY ANALYSI', REPORT FOR THE CLINCH RIVER EREEDER REACTOR.
USER DISCUSSION POSITION COMMISSICH ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS P_US.T_RlL A C T EI T I FJ EEVIEW HFLD CntfPl ET ED HiLD_
UfLD ISS'f ED Jf1PL EMEN T ED OFFICE RESYUNSIBLE......... N.?R NRR SCHEEULED C07.?tETION DATE.. --
U:4SCHED UNSCHED "NSCMED UNSCHED UNSCHED 1977 ACTU AL CCMPL ET IC'4 DATE..... 09/09/77 1977 LR_COTTS ON 0 9/ n s/_Z7_,_u,_G AP*MI L1:
i EUCf D E_.!1P_ llc A Y LQ1_IJL!tWf t A TO rY PR0f GS : ESTIMATES OF THERMAL DETONATION ENERGIES ARE DIRECTLY RELEVANT IN LICENSING DECISI0t:5 REL AIED 10 CCNT AIhMENT REQUIREMENTS FOR LMFBR, PARTICULARLY KINETIC ENERGY RELEASE SYSTEM DAMAGE At:D RADIO ACTIVITY RELEASES. THE RIL TRANSMITTED A CRITIQUE OF THE EDARD AND HALL (UKAEA) THEORY FOR Tf.EEMAL EXPLOSIONS IN THE UO2-50DIUM SYSTEM. THIS INDEPENDEllT THEORETICAL ASSESSMENT IS USEFUL IN REIHFORCING THE CURRENT NRR BELIEF THAT SUCH DETONATIONS ARE OF LOW PR03ASILITY AND THAT FURTHER EXPERIMENTAL EFFORT IS RECUIRED.
JS CUM _U*P_A CT_tP:._PMU1D:
THE RESULTS OF THE CRITIQUE ARE MINIMAL IN THIS ON-GUING TECHNOLOGICAL EXPLORATION, EXCEPT AS N0ltD A30VE Iti REINFORCING THE IDENTIFIED PRECEPTIONS AND NEED FOR EXPERIMENTS. SUCH MORK, INDEPENDENT
]
ASSESS **E! CTS OF 10ENTIFIED AND PDTEllTIAL SIGNIFICANT SAFETY ISSUES, SHOULD CONTINGE.
j C0;*JiF 'iT $(P,_'2*WS : THE INITIAL EFFORT OF AN EXPERIMENTAL A*iD ANALYTICAL EVALUATION OF THE BOARD AND HALL MODEL 0F 1HERMAL EXPLCSI0 tis UAS UtIDERTAKEN SY DPM AT Tite REQUEST OF THE ACRS ( ACRS REQL;EST DAiED JULY 9, 1975 FOR BOARD AND HALL REVIEua NOTE DPM RESPONSE TO THIS REQUEST IN A LETTER FROM R. P. DENISE TO M.
LIBARKIN, ACRS, DATED 3
OCTOBER 24, 1S7. THE PIIENOMEf:0N IS APPLICABLE IN TfiE ONGOING REVIEu 0F THE FFTF REACTOR.
1 NRR ftAS LEFr - COMTINUING CONTACT WITH RES STAFF ON THE RESEARCH PROGRAMS PEP.TAINING TO INVESTIGATIONS ON THE BOARD AND.mLL EFFECT, AND )HE GENERAL SU3 JECT OF TilERt1AL EXPLOSIONS, AND MAINTAINS COGNIZANCE OF SUCH WORK IN THE U.S. AND OTHER COUNTRIES.
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P P0 0^ t1. C F f M LC O*:y *(I S C*4 POIENT!4 UT1117E IGN OR V A i t:I 0F RESEARCH RESHLTS IN THE REGULATORY PROCESS PIL 8:
7 D4TE ISSUED: 03/25/76 RFS DFCTSTOM UNIT: FAST BREECER REACTORS VIL TITLE: THE SI MER CODE FOR ANALYSIS OF HYPOTHETICAL CORE DISRUPTIVE ACCIDENTS IN LMFBR'S iPONSnPTNG OrFTCE(Sl: NRR RRS:
2-14 SIMMER CODE RESEARCH PROJECT MGR:
R. CURTIS i
R F S_Sq"'R'QS : THE SIC ER CODE IS DESIGNED TO PREDICT THE MOTION OF FAST REACTOR CORES DURING POTENTIAL j
CGRE DISRU?IIVE ACCIDENTS. A TRIAL VERSION OF THIS CODE WAS MADE AVAIL ABLE TO NRC'S LICEt3 SING STAFF AND HAS SUPPC tTED NRC ERANCH POSITIONS ON THE ANALYSIS OF CORE DISRUPTIVE ACCIPENTS FOR CLINCH RIVER BREEDER REACTOR LICENSING CONSIDERATIONS.
SIf*MER IS RECOCNIZED AS AN IM?ORTANT FIRST STEP IN THE DEVELOPMENT OF A MECHANISTIC j
CODE CAPA3LE OF DESCRIBING PHENOMENA NEEDED TO ASSESS CORE DISRUPTIVE ACCIDENTS IN LIQUID METAL FAST BREEDER RfACTORS.
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USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS IqST._RIl ACIIVIT H 1 E FVI Etl HELD COMPL Q(Q HFLD HELD ISSHED JMPLEMENIED OFFICE RESP 0t4SISLE.........
NRR NRR SCf!EDULED CCMPLETION DATE.. --
UNSCHED UNSCHED UNSCf!EP IINSCHED UNSCHED 1976 ACTUAL CCMPLETION DATE..... 09/09/77 1976 NRR CO* TENT 3_PN 0 9 / 0 9 /_7_7, }t GAf* W U u
TEM P R U Plklu i[n'4 To M G'tL A!WY _ PPOCQS: THE SUBJECT RIL ANNOUNCED THE AVAIL ABILITY OF THE SIMMER-t CODC.
THE S IM** ER-t CODE IS REC 0CNIZLD Bir NRR AS AN IMPURTANI FIRST STEP IN THE DEVELOPMENT OF A MECHANISTIC CUDE CAPABLE OF EESCRIEING RECRITICALITY, TRANSITION PHASE PHENOMENA, AND WORK ENERGY PARTITION, ALL VITAL TO ASSESSING CORE l
DISRUPTIVE ACCIDENTS IN LMFBRS.
)
P15 CPMl'"' A G.Lf F RESUUS : THERE IS NO IMPACT SYNCE RESULTS USING THE SIMMER-1 CODE ARE NOT SUFFICIENTLY SUBSTANTIATED 10 DE USED DIRECTLY IN THE LICENSIt<G DECISIONS FOR LMFBR'S.
IT IS ANTICIPATED TifAT FURTHER CODE DEVELOPMENT. AND THE PERFORMANCE OF EXPERIMENTS UHICH CONFIRM THE COMPUTATIONAL MODELS, WILL YIELD A TOOL WHICP UILL BE USEFUL IN REACHING SICNIFICANT LICENSING DECISICHS.
i fC*T nth'REMASO: NRR VIEWS !HE SIMT;ER PROGRAM AS AN IMPORTANT LONG TERM EFFORT Tn DESCRIBE THERMAL AND MECHANICAL 1
CCRE DISRUPIION ACCIDENT S SUENCES. NRR ALSO SEES NEAR-TERM BENEFIT IN !! SIN 3 SIMMER TO BETTER UNDERSTAND KEY PHENOMENA. AND THE INTEGRATION OF THESE PHENOMENA INTO AN ACCCIDENT SEQUENCE. SIMMER ALSO HAS THE BENEFIT OF EEING LARGELY INDEPENDENT OF ERDA SPONSORED WORK.
NRR I*AS hEEN IN CONSTANT COMMUNICATION WITH ARSR AND HAD 1
INITIATED A SMALL Y.A.
EFFORT WHICH MALES USE OF SIMMER.
NRR STAFF HAVE ATTENDED SIM*tER W0r.KSHOPS AND 3RIEFINGS ON 9/sc,-35/76. 4/11/77 AND 7/2f-22/77.
NRR CORRESPONDENCE RELATED TO SIMMER AND SIMMER-RELATFD WORK CONDUCIED BY RSR:
1 TO SAUL LEVINE. RES FROM B. RUSCHE, NRR, "LMFBR SAFETY RESEARCh PP,0 GRAM PLAN". 3/15/77.
3 2.
TO L. 5. RUBENSTEIN F R ON. R.
P.
DEt'ISE, "DEGREE OF SUPPORT FOR RESEARCH PROGRAMS FOR ADVANCED REACTORS",
7/6/77 (CC. TO C.
KELBER).
3.
TO E. G. CASE FRCM R. P.
DLtilSE, "NRR COMENTS TO ACRS ON ARSR PROGRAM", 7/15/77.
4 j
4 TO SAUL LEVINE, RES, FROM E. CASE, NRR, "RES INFORMATION ON THE UNDERSTANDING OF CDA ENERGETICS IN LMFBR'S".
l l
1 i
~ er6c m W ice CC'"tNTS 09 P0,94rfAt UTILId flON OR VALUE Of DESEARCH MESULf5 Ir THi RA GLAroR7 PROCESS l
RTL 8:
3 D4TE ISS' FED: 01/31/77 4 8 CU NTON UNIT: FUEL BEHAVICR PIL TITLE: L EC *fe HEAT DATA AP?LICASLE TO LOCA EVALUATICH 5'PMS G NC OFFICEM):
RES ggq:
1-11 DECAY HEAT PESEARCH tEGJECT FMP:
R. DISALVO R E S__CCLN T3 : RECENT CALCUL ATIONS AND EXPERIMENTS WERE PERFORMED TO DETERMINE % BEST ESTIMATE OF THE VALUE OF THE
".s_S I DU A L (CECAY HEAT GENERATED IN tWCLEAR FUEL AFIER SHUTD01:N. THE VALUE CHOSEN FOR RESIDUA'. HEAT RATE PLAYS A VERY IMPCRTANT PART IN PREDICTIN3 THE PERFORMAMCE OF ECC SYSTEMS. RESULTS I.iirICATE THAT THE DECAY HEAT RATE CURRENTLY USED IIAS A 27% MARGIN CVER 1HE NEWLY DETE*'!!IhED VALUE.
THIS INDICATES A SIGNIFICANT CONSERVATISM IN CURRENT i
LICENSING 3ASES AND ' HEY ARE ::0M 3EING REEXAMINED.
RIL'S 3 1 7 ARE 20TH UNDER REVIEW BY NRR FOR POSSIBLE INCLUSION j
IN A M3DIFICATICN TO THE EN RCENCY CCKE COOLING SYSTEf15 (ECCS) RULE PROVIDED IN APPENDIX K OF 10 CFR 50.
THESE RIL'S REPORTED CS CDP.PtLTED RESEARCH PROJECTS IN FISSION PRODUCT DECAY HLAT AND ZIRCALOY CXI0ATION WHICH RESULTED IN DATA BASES AND CORRELATIONS WHICH INDICATE THAT THOSE IH CURRENT USE IN THE ECCS RULE ARE HIGHLY CONSERVATIVE.
A CC?: MISSION PAPER MAS PREPARED (SECY 77-363) ON JULY 1,
1977, PRESENT STAFF PROGRESS ON A PROPOSED ACTION PLAN REGARDING POSSIELE MODIFICATION TO THE ECCS RULE.
THE COMMISSION PAPER (SECY 7b-26) PRESENTING THE PROPOSED ACTION l
PL AN WAS PREPARED CN JANUAE1r IS, 1973.
i USER DISCUSSION POSITION CCPJ1ISSION ACRS PRESS i
0FFICE MEETING PAPER BRIiFING BRIEFING RELEASE RESULTS POST 'l k ag;'r1Lg5
?Evm HFtD CO*PtETED MELD PELD LSMTD_
LMPL ET1FNT FD I
0'-FICE R %PONSILLE..
t;RR.SD NRR NRR.SD.RES NRR SP,RES NRR l
SCHEDULED CC7:PLETION DaeE.
UNSCHED DEFERRED 9/0/77 1973 J
ACTUAL CDP.PLETION DATE.....
Cai2'/73 9/1C/77 7/26/77 9/9/77 f
N?R CC'"RNTS_0?L_01/19/7 7,_0 * /11/ 7 ?.,_ D. 9055 M }(P I 'L,_Af f1L C A LLO'4 TO E ULATORY PROCEM:
1
Pr.0PuiES OPTIONS FOR RULE CHANGE INCLUDE ACCEPTABILITY OF HEW DECAY i
HEAT DATA IN ECCS LICEt: SING CALCUL ATIONS.
]
2.
NEM LATA UILL BE INCORPORATED IN "DEST ESTIMATE" ANALYSIS CODES.
THESE CODES APE INTENDED TO BE USED IN PR03ABILISTIC ASSESS;;ENT OF SAFETY MARGINS.
pr$CRIJ. lit
- PACT OF PEsti! T5:
1 NEM DATA CONFIRPS MARGINS ALLOTTED IN PRESENT ECCS RULE ARE CCHSERVATIVE.
IF ACCEPIED FO% USS IN LOCA LICENSING CALCULATIONS, VENDORS COULD USE ADDITIONAL MARGIN IN DESIGN.
(I.E.,
INCREASE d
PEAK KC/FI, ETC.).
THE CO. 11SSION APPROVED THE PRCPCSED ACTION PLAN TO MODIFY APPENDIX K (SECY 73-26).
NP.R AND 7
CC"RNTS/vEM RK_3:
ELS ARE PRLPARING A ~:EQUEST TO SD TO INITIATE TULEMAKING TO MODIFY APPENDIX K.
1(LCG1"iNTS ON e?/01/7 L V.
PANCIf9A:
IHLSE STUDIES PROVIDE INfCRIAIION NECESSARY TO ASSESS THE DEGREE OF CONSERVATISM OF THE DECAY HEAT ASSUMPTIONS il IN 10 CFR 50 APPENDlX K (ECCS EVALUATION MCDE;.5).
l i
1 i
i
---n -.
,----r-
+- -
_. _.~. - -.- - - -....-- - - - _
. ~. ~.
I i
PROP AM P? flr E LC;"Em s ON P0ftNTIAL 0 ILI7ATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS i
l Plin 9
DATE IS$???D:
03/1'*/77 RJS DFCISICH UNIT: FUEL BEHAVIOR 1
1 D'l TITLE: HIGil TEttPtsATURE OXIDATION 01-ZIRCALOY FUEL CLADDING IN STEAM SPOPSORIMG OFFICE (S):
RES EM: t-5 ZIRCALOY CLADDING RESEARCH PA0 JECT MCR:
M. L. PICKLESIMER REMC*MTJ: THE REPORTED RESULTS IMPROVE 002 UNDERSTANDING OF THE BEHAVIOR OF ZIRCALOY FUEL CLADDING XN AN i M IkCNtEN! RETRESEMTED SY A LOSS OF COOLANT ACCIDENT (LCCA).
THEY INDICATE THAT THERE WOULD BE SIGNIFICANTLY LESS DEPTH CF Eri!RITTLE:ENT IN THE FUEL CLADDING MALL, C ALCUL ATED FOR ANY POSTUL ATED LOCA AND MORE WALL MATERI AL WILL BE L EFT W!:ICH IS CAPABLE OF SUSTAINING LOADS LATER IN THE SEQUENCE OF SUCH AN ACCIDENT.
TitIS INFORMATION TOGETHER MITN DAT A 0'4 THE R ATE OF GROWTH C* i?XIDE AND OXYGEli - STABILIZED LAYERS IN CLADDING, PROVIDES A MORE SCIENTIFIC B ASE F02 ESTA3tISHING FUEL CLA4 EMERITILEI!ENT C2ITERIA FOR EMERGENCY CORE COOLING SYSTEM ACCEPTANCE. THESE RESULTS CONFIRM THAT TNERE IS A DECREE OF CONSERVATISM I!: THE EVAtllATION MODEL BEING USED BY THF n uULAsu C STAFF FCR
(
CALCUL ATING IttE OXIDATION OF ZIRCALOY
- U2ING A POSTULATED LOCA.
THIS INFORMATION IS STING CONSIDERED AS A d
PARTIAL BASIS FOR MODIFICATICN OF THE PRESE!!T ECCS RULE (10CFR 50. APPENDIX K).
RG'S & 3 9 JRE BOTH UNDER CONSICER?.IICN BY MRO FOR POSSIBLE 1::CLUSION IN A F10DIFICATION TO THE EMEF.GENCY C0kE COOLING SYSTEMS (ECCS)
RULE PROVIDED IN APPENDIX K cf 13 CFR 50 THESE RIt's REPORTED ON COMPLtTED RESEARCH PROJECTS IN FISSION i
PRCDUCT DECAY HEAT AND ZIRCALOY OXIDATION WHICH RESULTED IN DATA BASES AND CORREL ATIONS UHICH INDICATE THAT l
TI:0SE IN CURhNT USE IN THE ECCS RULE ARE HIGHLY CONSERVATIVE. A CC:1ISSION PAPER 1:AS PREPARED (SECY 77-368) j ON JULY 1,
1977, PRESENT STAFF PROGRESS ON A PROPUSED ACTICH PLAN RECARDING POSSIBLE MODIFICATION TC THE ECCS RULE. AND THE CCM11SSICN PAPER (SECY 73-26) PRESLNTING THE PROPOSED ACTION PLAN WAS PREPARED CN Jat:UARY tt, 1973.
RIL 31 HAS BEEN REVIEMED BY tlRR, NMSS, IE, t.ND SD.
THIS MATERIAL WAS INCLUDED Ira THE 1
CIEFING OF THE ACRS FULL PC-: tI T TEE LY t:RR IN SEPTEMBER, 1977 CONCERNING A POSSILLE REVISION OF APPENDIX K 10 CFR 53.
It!E ACRS DISCOURAGED A REVISION OF THE ECCS RULE AT THIS TI!!E.
1 USER DISCUSSION POSITION COMMISSION ACRS PRESS 1
0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS P._Q,STJILJC_ TIVI T_If}
g5Vyy HFLD COMP L E T ET)
M( fi HELD TSSUED If1PL EMM T EQ j
0FFICE KE5PCNSIBLE......... NZR/SD NER NRR 3
SCHECULED CG'iPLETION DATE.. --
UNSCilED DEFERRED 9/0/77 UNSCHED UNSCHED 1978 ACTUAL CO*1PLETICM DATE..... 07/23/77 9/10/77 1973
}
N?. U UU N E3_SIL U 'Il d N J 161 6 3:_ A 3 0SS:
j D OC 71 'i A PP_UCLU QN_T Q__E t GUL3 T 0lY_ P_R0g S S :
1.
NEW DATA MAY BE C0itSIDERED AS PARTIAL B ASIS FOR 1
MODIFICATIONS OF PEESENI ECCS RULE (10 CFR So, APPENDIX K!.
1 2.
NEW DATA MAY CE USED IM "BEST ESTIMATE" CODES.
THESE CODES spE IHitNDED TO BE USED IN i
PRC3.*,BILISTIC ASSESS?*ENTS OF SAFETY T1ARGINS.
i pM Ce[$ LT-PJCT OF STESUtTS:
1 MEM DATA CONTIRMS ttARGIN ALLOTTED IN PRESENT ECCS R*1LE IS i
CCNSERVAllvC.
2.
IF ACCEPTED FCR USE IN LCCA LICENSING C;.LCULATIONS, VENDORS COULD USE ADDITIONAL MARGIN IN DESIG't (I.E.,
ANCREASE PEAK W/FT).
q
_ G F FMT M EfAr%3: THE COMMISSION APPROVED THE PROPOSED ACTICN PLAN TO MODIFY APPENDIX K (SECY 78-26).
]
j SRR AND st15 ARE PREPARING A EEQUEST TO SD TO INITI ATE RULEMAKING TO T10DIFY APPENDIX K.
1 I
1.0_C Q* l"* NTJ 0 4 09/tF73, V.
PANCIFDA:
THIS PROJECI PROVIDES DATA ON THE AMOUNT OF METAL-WATER REACTION AND HYDROGEN GAS GENERATION FOLLOAING A 1
LOCA AND AIDS IN THE DEVELOPMENT OF REVISIONS TO REGULATGRY GUIDE 1.7 AND IN ASSESSING THE CONSERVATISM OF A55U;;PTIONS IN tc CFR SO APPENDIX K.
i 1
1 :
- -- ___._., =--
F e 99 t *1 p f I c s-Crytasy y s en en i ty r I A L UIILIZATION 09 VALUE OF RESEARCH RESULTS IN THE REGULA10RV PROCESS PTL s:
13 UATE ISSUED: 02/25/77 RES DECISTON UNIT: RISK ASSESSMENT / PRIMARY SYSTEMS INTEGRITY Pit TITLE: PRESSURE VESSEL FAILURE PROBABILITY PREDICTION (OCTAVIA CODE)
1-2C VESSEL INTEGRITY RESEARCH PROJECT MGR:
W. VESELY P_ELCf MLTj: GCTAVIA IS A COMPUTER CODE FOR PREDICTING FAILURE PROBABILITIES IN REACT 0et PRESSURE VESSELS AS A FC.CIION OF TEMPERATURES AND PRESSURES THAT MIGHT CCCUR CURING STARTUP AND SHUTDOWN OPERATIONS. EFFECTS OF MATERIAL PROPERTIES ANS CPERATING AGE MERE ALSO INCLUDED. THE RESULTS INDICATED THAT EXISTING SAFETY MARGINS CCULD BE SIGNIFICANTLY REDUCED WITH CONTIHUED AGING (RESULTING IN RADIATION EMBRITTLEMENT) 0F THE REACTOR VESSEL.
3 THE RATE CF OCCURRENCE CF OVERPRESSURE EVENTS CONTINUES TO BE SIMILAR TO THAT PREVIOUSLY OBSERVED. FURTHER IMPROVE *~ENTS IN THE CCDE ARE REPORTED ON IN RIL 12.
THESE RESULTS WERE USED BY NRR DURING THIS PAST YEAR IN REVIEMING THE PRCB4BILITY OF FAILURE OF REACTOR PRESSURE VESSELS DUE TO SUCH EVENT $ AND TO GIVE GUIDANCE ON PLANT I"PROVENENTS THAT MAY BE NEEDED.
THEY MAY ALS3 SECOME THE BASIS FOR REVISING REGULATORY GUIDES OR STANDARDS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS
_SSUED IMPL(MENTED I
fjST P[L_S_QTJVITIES P M Ey HEt D Cn'1P L E T ED HFt D HELD CtFICE RESPGNSIBLE......... ^2R/SD NRR SCHECULED CPMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 1977 ACTUAL COMPLETION DATE.....
09/09/77 1977 YZR_GCT**5975 ON Q/07/77_,_B GPJ_MFS:
P ricmL> P131E*llfLilL%ML A T m _P20C ESS : THE OCTAVIA COMPUTER CODE HAS BEEN USED TO EVALUATE THE PROBABILITY OF RLACIGR VESSEL FAILURE FROM OVERPRESSURE IRANSIENTS WHICH CAN OCCUR DURING PMR OPERATION. THE RESULTS OF THE AHaLYSES INDICATED THAT EXISTING SAFETY MARGINS COULD BE SIGNIFICANTLY REDUCED WITH CONTINUED HEUTRCH IRRADIATION OF THE REACTOR VESSEL IF THE HISTORICAL FREQUENCY OF OVERPRESSURE EVENTS CONTINUED.
Pf.3 G I15F J 1ACT OF PESHtTS: THE At;ALYSES ENABLED NRR TO CONFIRM, IN A MORE RIGOROUS, QUANTITATIVE MANNER, INITIAL LICENSING CLCISIONS TO REDUCE THE FREQUENCY AND MAXIMUM PRESSURE OF THE TRANSIENTS. THESE DECISIONS RESULTED IN MODIFICATICNS TO OPERATING PROCEDURES FOR REDUCING THE FRECUENCY OF TRANSIENTS AND THE INSTALLATION OF PHYSICAL DEVICES 70 LIMIT PRESSURES IN OPERATING PLANIS TO THOSE SPECIFIED BY THE PRESSURE-TEMPERATURE LIMITATIONS IN THE TECHNICAL SPECIFICATIONS.
M ENTS/RE m 35: SEE CCMNEN!S TO RIL 812.
$f._CONIENTS_EN 09'13'73 d d A_ W A R 1 THIS PEOGRAM UAS GLNERAILD SUM s ANTI AL DAT A USED IN PREDICTING IRRADI ATION EMBRITTLEMENT AND MARGINS To FAILURE OF REACIDR PRESS"It VESSELS UHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES AND LOADINGS.
THIS PROGRAM HAS PROVIDED IMPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR 50 AND CONTRIBUTED DATA USED IN THE ASME CODE MHICN MAS INCORPORATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE DATA BASE USED IN REGUL AT02Y GUIDE 1.99 CONCERNING THE EFFECT OF COPPER IMPUIITIES ON SENSITIVITY OF STEEL TO IRRADIATIC:4 AND IS EXPECTED TO PROVIDE A SUdSTANTIAL INPUT FOR FUTURE REVISIONS TO REGULATORY GUIDE 1.99.. -
4 PRfG?tM Of F ICE CON-ENis CN POTENilat illItI/aftnN OR VAtUE OF RESEARCH PEST!LTS IN THE REGULATORY PROCESS RIL 22 tt PATE IS5UED: 09/15/77 PES DECTSTON UNIT: RISK ASSESSMENT pit TITLE: IEEE NUCLEAR RELIA 3ILITY DATA MANUAL SPONSOPING OFFTCE(S): RLS PEQ: NOME PESEARCH PROJECT MGR:
J. JOHNSON Ef1_CCE ENT_S!
A FAILURE RATE DATA MANUAL NAS DEVELOPED MHICH CAN BE USED IN RISK AND RELIABILITY ANALYSIS OF REACTOR SYSTEMS.
THE MANUAL CONTAINS FAILURE RATES AND FAILURE MODE INFORMATION FOR OVER 1.000 ELECTRICAL.
ELECTRCNIC AND SENSING Co.5PONENTS USED IN NUCLEAR PCMER PLANTS.
A METHOD IS GIVEN FOR COLLECTING AND PRESENTING l
RELIABILITY DATA FOR CUANTITATIVE RELIABILITY AND AVAILABILITY EVALUATIONS OF SAFETY-RELATED NUCLEAR PLANT SYSTEMS. UNCERTAINTY BOUNDS ARE ALSD GIVEN FO? EACH ESTIMATE OF A COMPONENT FAILURE RATE.
THIS WORK IS PART OF A CONTINUING EFFORT TO ESTAELISH AN INTERIM DATA BASE FOR USE IN MEETING NRC NEEDS IN THE ELECTRICAL AND ELECTRONIC AREA UNTIL SIGNIFICANT CPERATING DATA ON COMPONENTS USED IN THE NUCLEAR INDUSTRY BECOME AVAILAELE l
l USER DISCUSSION POSITICN COMMISSION ACRS PRESS
)
0FFICE MEETING PAPER BRIEFIliG ERIEFING RELEASE RESULTS
.P 01L_R U _AS T T_V_I UIS PJVIEN HELD CAMPLFTED jiEL D HEL D
.LS$Ufp Ir*P L Et1EN T ED OFFICL RE5PeuSIBLE.......s. NRR URR SCHEDULED CCMPLETION DATE.
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED C4/30/78 ACTUAL COMPLETION OATE..... 13/25e77 El(Cf t_"E*ali_0N 10/23 g7 g g [D5$C3:
D l
PsiqEILE (PPLifAIJON TO PFC21Aff7Y PRDI(11: THE SU3 JECT RIL ANNOUNCED THE AVAILABILITY OF A FAILURE RATE DATA MANUAL.
i THIS LATA SUPPLLMENis OTHER SCURCES OF RELIABILITY DATA SUCH AS THE REACTOR SAFETY STUDY.
SUCH DATA ARE 4
BEING USED IN RELIAEILITY STUDIES T.9AT SUPPORT OR PROVIDE THE BASES FOR LICENSING REQUIREMENTS.
I D11CR[lf_JUPA$T OF PE$tgli: THE FIRST STUDIES USING THIS DATA HAVE NOT YET BEEN CCMPLETED AND THEREFORE THE IMPACT a
CANNOT YET SE OLTLR"INED.
GRr51NT3/REMt?K1: TNIS DATA MANUAL PROVIDES THE STRUCTURE FOR INCORPORATING NEW OR REVISED DATA AS IT BECOMES AVAILABLE.
It!E MANU?.L NOM CONT AINS UNDIFFERENTI ATED HARD AND SOFT DATA MHICH IS A SERIOUS IMPEDIMENT TO THE APPLICATION OF THIS DATA.
THEXEFORE. CONTINUING WORK IS REQUIRED TO INCREASE THE CONTENT OF HARD DATA SY INCCRPCRATING THE DATA DEVELGPED THROUGH SUCH PROGRAMS AS NPRDS.
?"* A Cf"'~ CN T S O N_0 t Q / 7 8, 8 A 3 ? f 55014 :
THE DATA 15 DIFFICULT IU INTERPRET BECAUSE OF THE MANY AMBIGUITIES AND INCONSISTENCIES IN THE DATA MANUAL.
THE RELATION OF THE AGCREGATED FAILURE RATES TO THE COMP 0NEMI FAILURE RATES IS NOT MADE CLEAR AND THEY ARE OFTEM INCCNSISTENT.
IN PARTICULAR. THE USE,F GEOMETRIC AVERAGING IS NOT JUSTIFIED AND LEADS TO INCONSISTENT RESULTS. THE WAY IN ZilC't it'cIl AND LCW F AILURE RATES AUD UP IS NOT CCNSISTENT MITH THEIR STATED INTERPRETATION AS 95T!! AND 'eTil PERCENTILES OF THE FAILURE RAYE DISTRI53 TION.
THE EQUALITY OF THE ilIGH VALUE TO THE MAX VALUE IN MANY TA3LES IS ALSO INCCNSISTENT WITH THIS INTERPRETATION. THE FAILURE MODE TYPES AND DEFINITIONS GIVEN EXCLUDE SUDOEN PARTIAL FAILURES AND GRACUAL C"MPLETE FAILURES, BOTH OF WHICH ARE EXPERIENCED IN PRACTICE.
)
l i
I 1
I
PROGm CH ICE Cm1ENTS CM POTENIIAt UTILIZATION 09 VALUE OF PESE4tCH RESULTS TH THE REGULATORY PROCESS PIL #:
12 DATE ISSUED: 06/16/77 RES DECISTON UNIT: RISK ASSESSMENT / PRIMARY SYSTEMS INTEGRITY RIt TITLE: MODIFICATIONS TO PRESSURE VESSEL FAILURE PROBASILITY PREDICTION (OCTAVIA CODE)
SPONSMIM OFFICE (5): RES PPQ: NDHE RESEAPCH PROJECT MGR:
W. VESELY F D._Q U M T3: MODIFICATIONS IN THE OCTAVIA COMPUTER CODE REPORTED IN RIL 810 WERE MADE THESE MODIFICATIONS It<CLUSE A CAPABILITY 'O HANDLE RESIDUAL STRESS IN A REACTOR PRESSURE VESSEL WHICH CAN EITHER BE CONSTANT, OR VARY WITH FLAU SIZE THE COLE USER CAN IMPOSE AN UPPER BOUND ON THE VESSEL TOUG:iNESS AND THE CODE HAS THE CAPABILITY TO HANDLE UNCERT AINTIES IN THE TOUGHNESS. USING THE MODIFIED DCTAVIA CODE, 7:lE MEDIAN FAILURE PROBABILITY FOR THE SUIRY REACTOR PRESSURE VESSEL MAS CALCULATED TO BE S X 10-7 PER VESSEL YEAR FCR AN OPERATING TEMPERATURE OF 110 DEGREES C AND THE CURRENT AGE OF APPROXIMATELY 2.5 YEARS.
THE FAILURE PR0BABILITY INCREASES TO 3 X 10-5 PER VESSEL YEAR AFTER 40 YEARS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASF RESULTS PpS_T_RlyC11VIRFJ REVlty HELD COMPLETED HEL O HEtD ISSUED IMPL EM[N T Ff) 0FFICE RtSPONSIELE......... NRR/SD NRR SCHEDULED CCMPLEYION DATE.. 03/16/77 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 1977 ACTUAL CCMFLETICN CATE.09/23/773/77 1977 NILCODENJ.} r'N _0Jff_3/77 ddPE_($:
RtLCPJ 5UP_PLIC A TJMl9 FEGUL A LORl_ PRE (11: (SEE RIL 0101 D-.E E P l._JbE!CT OF RtSJ)1J h: (SLE RIL 210)
C r. p t q sf p f:A MS s IMIS IS A MODIFICATION OF RESULTS TRANSFERRED BY RIL 810, 02-25-77.
THE ANALYSES USING THE OCTAVIA CODE ARE CG::PLETE. THE GENERIC POSITIONS REGARDING OVERPRESSURIZATION PROTECTION SYSTEMS HAVE BCEM DEVELOPED AND ARE OCCUMENTED IN NUREG-0224,
- REACTOR VESSEL PRESSURE TRANSIENT PROTECTION FOR PRESSURIZED WATER REACTORS.'
IMPLE-MENTATION OF THE POSITICNS ARE UNDERWAY.
ECOMEN TLCN_0RL1/73, P. P ANDER - SEE COMMENTS FOR RIL 810.
IHIS PROCRArt HAS GENERATED SuSST ANTI AL DAT A USED IN PREDICTING IRRADI ATION EMBRITTLEr1ENT AND MARGINS TO FAILURE CF REACTOR PRESSURE VESSELS WHEN FL AWS ARE PRESENT, CONSIDERING MATERI AL PROPERTIES AND LOADINGS. THIS PROGRAM ttAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TC 10 CFR S0 AND CONTRILUTED DATA USED IN THE ASME CODE k"4ICH 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 OF STEEL TO IPRADIATIC;; AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS TO REGULATORY GUIDE 1.99. --
Pucyn CHICL CONSTS ON P0ffN11AL UTILI/ATIC;4_Of VALUE OF ulSEAPCH PESutIS IN THE REhULAICRY PPnCFSS PIL 1:
13 CSTE TStyrD:
11/11/77 E[2_pECTSinN UNIT: PRIMARY SYSTEMS INTEGRITY RTL TITLE: RESIOUAL STRESSES IN WELDS sPnNs?2Iwg crrICE(S): RES EEG:
1-20 VESSEL INTEGRITY PESEAPCH PPOJECT MGR:
C. SERPAN RES_QF""ryTs: A VEPIFIED MODEL IS PRESENTED FOR PREDICTING RESIDUAL STRESSES RESULTING FRCM THE WELDING OF PIPES, AND IHE ESTIMATICN OF RESIDUAL STRESSES RESULTING FROM WELD REPAIRS OF REACTCR PRESSURE VESSELS.
THE MODEL CAN BE USED IN THE LICENSING PROCESS TO AID IN THE EVALUATION OF CRACKING THAT HAS OCCURRED IN GIRTH-30TT WEtOS IN PIPING.
IT S:tGULD ALSG PROVE TO BE USEFUL IN ANY SAFETY EVALUATION OF PROPOSED REPAIRS BY WELD BUILDUP IN THE CORNER REGICNS CF PRESSURE VESSEL N0ZZLES AFTER CRACKS HAVE BEEN PEMOVED. AND IN VESSEL WELD REPAIRS.
USER DISCUSSION POSITION CCMNISSION ACR$
PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST PJt a CUVJ TIF$
PEVL[y HEt D
[0]>!(TfD fi[L D
[i[(D ISSUED IMPTFMENTCD CFFICt KLsPC,SIttt......... tas/SD SCHECULED CONPLETIJN D%TE.. 03/31/73 UMSCNED UNSCHED UNSCMED UNSCHED UNSCHED UNSCHED ACTUAL CCM?LETION DLTE.....
05/14/73 N" R_Q O'~~E*4 T_5 f M_0 v 14 / 7 9 J.
ryIcny:
C P!'t qteL_APPtIcattos In Efrut*IcvV P90 CESS: THE AXISOL CODE COULD BE USED AS AN AID IN EVALUATING RESIDUAL STRESSLS Its FLULD HLJ.D-FRCCLSS PIPE WLLDS UF CCNTAINMENT PENETRATION ASSEM3 LIES AND GIRTH BUTT WELDS IN PIPING.
IT S!:CULD ALSO PR3VE USEFUL AS AN AID IN DEVLCPING THE NECESSARY DECISIONAL INFORMATION IN ANY SAFETY EVALUATON OF PROPOSED WELD REPATR5.
T&51217E_ IMPACT qf_yESULJ31 THE COMPUTER CODE AXISOL COULD EVENTUALLY BE USED AS A DESIGN TCOL BY BCTH GOVERNMENT AND I DJSTP.Y. IC IMcRUVL WLLD'NG TECHNIQUES AND PRGCEDURES.
THE RESULTS OF THIS PROGRAM SHOULS DE BROUGHT TO THE ATTENTION OF VAPIOUS ASME GPCUPS ENCAOED IN PREPARING CODES AND STANDARDS OF WELDED FABRICATION AND INSPECTION FROCEDUEES.
THOROUGH DISCUSSION AND EV5LUATION BY SUCH CROUPS AND CONSIDERABLE TRIAL USE BY INDUSTRY IS HECESSARY SEFCRE FULL APPLICATICN OF THE CODE IN THE LICENSING PROCESS WOULD BE APPRCPRIATE. THE INABILITY OF THE CODE TO TAKE INTC #CCCUNT T.iE EFFECTS CF PUSI-WELD HEAT TREATMENT. AT THIS IIME, IS A DRAU3ACK IN UTILIZING THE CODE IN THE LICENSING PRCCESS.
E Cn~*FNTS C *i P '/13 n t,_H,_ff ? 31 IHIS PRUGRAM MAS GENLRAILD SUCSTANTIAL DATA USED IN PREDICTING IRRADIATION EM3RITTLEMENT AND MARGINS TD FAILUPE CF REACT 02 PRESSURE VESSELS 1: HEN FLAWS ARE PRESENT. CONSIDERING MATERIAL PROPERTIES AND LO* DINGS.
THiS PRD0 RAM HAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE YOUGHNESS OF PRESSURE VESSEL MA!ERIALS DESCRIBED IN APPENDIX G TO 10 CFR S3 AND CONTRIBUTED DATA USED IN THE ASME COD: CtICH 2AS INCORPCRATED INTO APPLNDIX G.
IT HAS ALSO PROVIDED PART OF THE OATA St.SE USEP IN RECULAICRY GUIDE 1.99 CONCERNING THE LFFECT OF COPPER IMPURITIES PN SENSITIVITY l
CF STEEL TC 12RfDIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISICNS l
T3 RESJLATC2Y GUIDE 1.59.
l l l i
l l
- - -. -. -~ ~ _.. -
l PPC3*AM OFFICL C7PMENTS C1 POTENTIAt UV T* IZr t inn OR VALUF OF RESEARCH RESULTS IN THE FE00LAVORY PROCEq),
RIL 22 14 PATE ISSUEQ:
11/09/77 RES DFCTSTON UN L1: SYSTEMS ENSINEERING I
FIL TITLE: PHYSICAL SEPAR. ION CRITERIA FOR ELECTRICAL CABLE TRAYS (HORIZONTAL OPEN SPACE CONFIGURATION) 4
$PCNS0?TNG OrFTCE(S): RES EXE: 1-23 ELECTRICAL STDS A RESEARCH PROJECT MOR:
R.
FEIT FIRE PROTECTION l
PfL CQ""LNT1: THE A0EOUACY OF THE REQUIRED SPACING OF ELECTRICAL CA3LE TRAYS AT NUCLEAR POWER PLANTS WAS i
LXAMINED 10 PREVENT THE SITEAD OF CABLE FIRES.
RESULTS INDICATE THAT CURRENTLY USED CRITERIA FOR CABLE TRAY i
SEPARATICM A?? EAR TO BE ADEQUATE F02 ELECTRICALLY-INITIATED FIRES, BUT THAT CHANGES MAY BE REQUIRED FOR FIRES DUE TO j
EXTE2NAL IGNITICM SOURCES.
THIS WORK IS APPLICABLE TO A VERIFICITION OF REGULATORY GUIDE 1.75. "PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS".
EXPOSURE FIRE TESTING EMPLOYING EXTERNAL FULL SOURCES WAS CONDUCTED TO PROVIDE DATA F0t THE DEVELCPMENT OF CURRENT hRC STAFF POSITION AS DOCUMENTED IN THE APPENDIX A TO THE BRANCH TECHNICAL i
POSITIO:8 APCSS 9.5-8 "GUICELINES FOR FIRE PROTECTION FOR NUCLEAR PCMER PLANTS" AND IN THE DRAFT REGULATORY GUIDE 1.120. "FIRE PROTECTION GUIDELINES FOR NUCLEAR POMER PLANTS."
USER DISCUSSION POSITICH CCMMISSION ACRS PRESS d
I 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RTL ACTILITLL1 SEV UM HFLD CNtPLETED ELL D Efl D ISSUED IMPLEMENTED l
OFFICE RESPCMSIBLE......... NR2.50 SCHEDULED COMPLETION DATE.. 31/05/73 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETICN CATE.....'J%/26/73 E T C0""E*fTS_0LJ4/?6/73. SM/?X73 d. TEDESCO:
j f f 5CZJ P t_8 EELJ S af t t4 10 RF091Aiody PRnCESS: THE RIL PROVIDES RESULTS OF A COMPLETED PORTION OF THE NRC Flat Pi]ItCTION RESEARCH PROGRA:t CONLucitD AT THE SANDIA LACORATORIES FEGARDING THE EFFECTS OF CABLE TRAY SEPARATION ON THE PROPAGATION OF ELECTRICALLY INITIATED AND EXPOSURE l'RES.
RES.CPlff_1EPSCT OF #ffitT3 THE IN!R3CUCTORY PARAGRAPH OF RIL E14 STATES THAT THE REdULTS OF THE SANDIA P.t0GR A:1 INDICAlt THAT "CURRENTLY USED CRITERIA FCR CABLE TRAY SEPARATICN APPEAR TO BE ADEQUATE FOR ELECTRICAILY INITIATED FIRES BUT THAT CHANGES MAY BE REQUIRED FOR FIRES DUE TO EXTERNAL IGNITION SOURCES."
RES NOTED IN A CONVERSATICN ON MOVEM3ER 11 1977 THAT THE SPECIFIC POINT RAISED IN THAT PARAGRAPH IS DIRECTED 1
TO THE ADECUACY OF SDLE RELIANCE CN THE SEPARATION CPITERIA 0F REGULA! CRY GUIDE 1.75 FOR PROTECTION AGAINST i
l uXPOSURE FIRES.
TME FIRE PROTECTION CRITERIA DEVELOPED BY THE STAFF SINCE THE BROMNS FERRY FIRE HAVE
)
RECOGNIZED THf.T RELIT.NCE SHOULD NOT BE PL ACED SOLELY ON THE SEPARATION CRITERI A 0F REGUL ATORY GUIDE I.75.
CUR STAFF REPCRT, DATED NOVEM3ER 9, 1977 DN "THE QUESTION OF WhETHER THE PETITION OF THE UNION OF CONCERNED
- 4 J
SCIENTISTS RAISES MATTERS THAT REQUIRE IMMEDIATE COMNISSION ACTICN," INDICATES THE SIAFF POSITION THAT THE I
IEEE-33% AND THE REGULATORY GUIDE t.75 SEPARATION GUIDELINES, AND THE IEEE-383 FIRE RETARDANCY STANDARDS FOR SAFETY CABLES. SY THEMSELVES ARE NOT SUFFICIENT TO PROTECT AGAINST EXPOSURE FIRES.
CONSEQUENTLY UE REQUIRE ADDITIONAL MEASURES ICR FIRE PROTECTION SUCH AS: FIRE BARRIERS BETWEEH REDUhDANT DIVISION CABLE i
l TRAYS; FIRE 2ETARDANT COATINGS ON CABLING: AUTOMATIC FIRE DETECTION SYSTEMS; tUTOMATIC FIRE EXTINGUISHING SYSTEMS; ADMINISTRATIVE PRCCEDURES; AND TRAINING PROGRAMS.
THIS POSITION HAS BEEN HELD BY THE STAFF SINCE THE SROMMS FERRY FIRE AND IS REFLECTED IN OUR ST ANDARD REVIEM PL AN SECTION 9.5.1 (BTP 9.5-1). REVISION.
REG. GUIDE t.123 HAS BEEN ISSUED FOR CONNENT.
THUS, EXPOSURE FIRES ARE REQUIRED TO BE CONSIDERED IN THE FIRE PROTECTION 1
PROGRAMS FOR EOTH CPERATING PLANTS AND PLANTS IN THE LICENSING PROCESS.
IT SHOULD BE NOTED THAT THE RIL l
ACKNCMLEDOES THAT FURTHEd TESTING WITH ELECTRICALLY INITIATED CABLE TRAY FIRES UNDER DIFFERENT CONDITIONS MAY i
RESuti IN FULLY DEVELOPED FIRES.
j ME CCNCLUCE, AS LCES RIL 814, THAT THESE RESEARCH RESULTS CONFIRM THE NEED FOR PROTECTION MEASURES IN l
ADDITION TD THE SEPARATION CRITERIA. SINCE SUCH MEASURES ARE INCLUDED IN OUR PRESENT CRITERIA WE CONCLUDE THAT THE INFORMATION IN RIL 314 DOES NOT INDICATE THE NEED FOR CHANGES IN OUR FIRE PROTECTION GUIDELINES,
)'
2UT THAT IT CCNFIRMS THE NEED FOR CUR PLANS TO UPDATE REGULATORY CUIDE 1.75.
CC-MEMTS/cEM2P(S: NCNE l
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13_CO M NTL oft _py/13/_7b _C. PIV{N]A{K 11?tl tSIS HAVE conFIR7.LD ICE VALIDITY OF GUIDELINES IN REGULATORY GUIDE 1.120 WHICH CALL FOR SEPARATIt40 REtUNDtMT SAFETY SYSTEM CABLING BY RATED FIRE BARRIERS. THE TESTS HAVE CONFIRMED THE VALIDITY OF SE.*ATATICM CRITERIA SPECIFIED IN REGULATORY GUIDE 1.75 ONLY MITH RESPECT TO FIRES RESULTING FROM ELECTRICAL FAILU**E t?I!HIM A CABLE TRAY BUNDt2. HOMEVER. THEY SHOW THE BASIC DEFICIENCIES OF THE SEPARATIC'; REQUIREZINTS *IITH RESPECT TO EXPOSURE FIRES CC:7.ON TO MORE THAM ONE REDUNDANT SYSTEM CAELE RUN.
i
4 (4W A, CFF1cE MraE%r s os P0 LENT I At UTILIZATION CR VALUE GF RFSEHC54 PESUtis F4 THE MEGULAVORY PROCESS l
ELL _8 :
15 p2TE IS$2[fz-12/01/77 PES DECTSTnN UNTT: PRIMMY %YSTEMS INTEGRITY RIL TITLE: CHARACTERIZATION CF tMR FEEDWATER N0ZZLE CORNER CRACKS a
$ PONS 0?INC. OFFICE (S): RES Egg 1-20 VESSEL INTEGRITY RESEWCH PROFCT MGP: C. SERPAN PF5_( F r*E yT _: PRESSURE LOADING OF CRACKS IN THE INSI0E CORNER OF FEEDS!ATER INT AKE N0ZZLES FOR BOILING IATEL REAC.GR (3:JR) PRESSURE VESSELS tIAVE BEEN CHARAt,TERIZED.
THE CHARACTERIZATIqN ESTABLISHED THE RELATIONSHIP 5ETWEEN STRESS-GEHERA.ED PRES $URE AND MEASURABLE CRACK PARAMETERS. IN ORDER TO DETERMINC j
THE GRCMTH OF THE CRACK AND 175 CRIIICAL SIZE.
THESE RESULTS CAN BE USED TO CHECK THE CALCULATIONS.
4 BASED ON INTERNAL P2ESSU2E. FUR THE SAFETY ANALYSIS OF BMR FEEDMATER N0ZZLE CORNER CRACKS DURING SUBSECUENT CCM3INED CCOLING AND UPLOADING.
IT HAS BEEN SHOMM THAT UNDER THE M ST SEVERE CCNDITICNS.
THE CRACK CAN PENETRATE N3 MORE THAN 1/3 0F THE PRESSURE VESSEL MALL THIS MEANS THAT THE VESSEL MILL ALMAYS BE CAPABLE OF RETAINING EMERGENCY CCCLING MATER, THUS KEEPING THE CORE COOL AND PROVIDING FOR A SAFE SHUTDC*JN.
THUS. VESSEL FAILURE IS NOT POSSIBLE FOLLCMING MARM PRESTRESSING UNDER CO.lDITICNS MHERE COLD EMERGENCY CURE COOLING MATER IS INJECTED INTO THE HOT PRESSURE VESSEL FOLLOMING A LOSS OF C00L A.4T ACCIDENT.
LfSER DISCUSSICN POSITION COMMISSICN ACRS PRESS OFFICE McETING PAPER BRIEFING BRIEFING RELEASE RESULTS J
Po$T_PE ACTIVR E1 RV_Jiy MFLD COMPLETFD HELD HELD IS'3U E D IMPLEMENTED GF F ICt REPCMILL E......... NMD 7
J SCHECULED COMPLETIGN DATE.. 0 3/ 3 t/73 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED I
ACTUAL COMPLETION DATE.....
02/22/73 Mt10*~iEMJ_ C* LIM '/ 7 kB,_fMMF5 :
M ~t EJ M._!EtL IC 3 t I@_!LQ3LA @Y PP1Q$%: NRR CONCURS THAT THESE DATA CAN BE USED FOC PARTIAL VERIFICATION C
OF TPE APPRCXIMAf t MailHCDS USED BY GE IN ESTIMATIt:G THE STRESS JNTENSITY FACTORS APPLICABLE TO FEECMATER AMD CRD N0ZZLE CRACAS IN =R REACTOR VESSELS. THEY CAN ALSO B: USED FOR VERIFICATION OF MORE SCPHISTICATED f*ETHODS CF EV ALU ATION ($"CH AS FINIIE ELEMENT ANALYSIS) IF AND MHEN SUCH METHODS ARE DEVELOPED.
PE1CRT E_IM* MT nF mtRT$: LfC HAVE MADE A COMPARISON BETMEEN THE GE STRESS INTENSITY CURVE (FIG. 3-26
~;
UF MELE-21430) F0% TPE ONE CASE 10 TERE A DIRECT CCMPARISON CAN BC MADE -- NAMELY. AT AN A/T RATIO SLIGHTLY CREATER THAN 3.53. AS PRCVIDED BY FIG. 7 0F THE PROGRESS REPORT.
FOR TMIa DNE CASE. THERE IS ALMOST EXACT AGREEMENT BETMEEN THE GE CURVE AND THE VPI TEST RESULTS. THE GE CURVE. MHEN CONVERTED TO NCRMALIZED j
i 1
STRESS INTENSITY F ACT0RS. IIILL ALSO PRGDUCE A CURVE F AS A FUNCTION OF A/T RATID) MHICH IS QUALITATIVELY SIMILAR TO THAT OF FIG. 7 0F VIP REPORT VIP-E-76-25; IN THIS CASE. EXACT CORRESPCNDENCE IS NOT TO BE EXPECTED BECA815E IV THE Mr.TERI AL DIFFERENCE IN THE DI AMETER-TO-THICKNESS RATIOS OF VESSELS INVOLVED. THE CLO-*
j AGREr'*3 T BETMEEN THE GE CURVE AND THE TES'. RESULT F OR THE ONE CASE MHIRE A VALID CC"PARI'.ON CAN BE MADF PROVIDES ASSURANCE THAT THE STRESS INTENSITY FACTORS USED BY GE IN THEIR EVALUATION ARE REASONASLF APPRCXIMATIONS.
ME ENCCUKAGE THE COMPLETION OF THIS WORK AND PART'CULARLY THE DEVELOPMENT OF VERIFIED f&""* E N U / 8P"13ES -
AN ALYTIC AL iGI? ICDS RtICH MILL PROVIDE AN ASSURED MEANS FOR FUTURE CALCULATION OF SIF 5 FOR CRACKS IN COMPLEX GECMETRIES (SUCH AS T! COUCH THE USE OF FINITE ELEMENT ANALYSIS).
'IvfM9 apt:
$2_C CCaf $iT3_Ct_o ?/1 MZ%._G.
THIS r*ROGRAll HAS GENLRAIED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMFMT AND MARGINS j
TO FAILURE OF REACT 02 PRESSURE VESSELS UHEN FLAWS ARE PRESENT. CONSIDERING MATERIAL PROPERTIES l
AND LOADINGS.
THIS PROGtut HAS PROVIDED INPUT TO NRC REQUIREMENIS FOR FRACTURE TOUGHMESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR 50 AND CONTRIBUTED DATA USED IN I
THE ASME CODE WHICH MAS INCORPORATED INTO APPENDIX G.
IT HAS ALSO PRO 4IDED PART OF THE l
DATA BASE USED IN REGULATCRY CUIDE 1.99 CONCERNING THE EFFECT OF COPPER IMPURITIES Or. SENSITIVITY OF STEEL TO IRRADIATION AND IS EXPECTED TG PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIGNS TO REGULATORY GUIDE 1.99. 3
)
I 2
~.
I I
i
)
i j
FF0 Guam erFICE CC "'ENTS 01 t'O l t N f l a t U T I L IZ A T IOt4 OR VALUE 6F RESEARCH RESULTS IN THE REGULATORY PROCE Pll 1:
16 rate ISSUig:
12/r: s/77 Rt$ DECISTON UNTTr PRIMARY SYSTEMS INTEGRI;Y RIL TITtEt MARM PRESTRE55ING i
j FCN5CPING crrTCE(5):
RES FPfG2 1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR C. SERPAN Ff1 C C IM Q2 THE EFFECT OF COLD tMERGENCY CORE COOLING WATER ON HOT REACTOR PRESSURE VE5SELS WAS COMSIDERED. THE RESUllING THERMAL SHOCK COULD. UNDER "WORST CASE" CONDITIONS. LEAD TO THE PREDICTION TH AT FL AWS IN THE ST EEL PRE 550RE VE5SEL WGULD ETTEND.
THE *UARM PRESTRE55ING" EFFECT EMICH CAN PRECLUDE CRACK EXTENSION WHENRESULT5 REPORTED HERE PROVID PREDICTED. T3 DESCRIBE THIS EFFECT. ONCE A CRACK IT OTHERWISE WOULD HAVE BEEi:
]
EXTENSICN WILL 00 CUR.
15 LOADED KNILE THE MATERIAL IS VERY TOUGH. NO RAPID USER DISCUSSICN POSITION COMMISSION ACR5 PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULT 5 a
ff 5.T f1L AqTJVJUJJ EEyJfy HELD COMPLETED HELD HELD ISSUED IMPLEMENTED Cif1CL Rf5PCN51BLt......... hR<<5D l
ECHLDULED COM;'LETION DATE.
32/ tat /73 UNSCHED UNSCHED UNSCHED UNSCHED UN5CHED UNSCHED j
ACTUAL COMPLETICN DATE....
02/t%/78
]
NN CCNNH_0N p2/ L4/18t_CRf2 V7h_}JJr1FJ2 MW IM F 2 Ulls A llh._f 2_Eif'tMle
- Y PN_Qt55: IF "WARM PRESTRE55INGC AS DESCRIBED IN THE RIL IS OPERATIVE ON HIGHLY IERADIATED REACTOR vtSSEL SIEELS. THIS MECHANISM COULD PROVIDE ADDITIONAL MARGIN IN THE RPV TO ACCOMMODATE 1
THERM 2L SHOCK A55CCIATED WITH ECC5 INJECTION DURING A LARGE LOCA.
T E1Gl31__l?'" A C1 0F #F54LI_$: SY LIMITING CONCERNS REGARDING THERMAL SHOCK TO REACTOR VE5SELS TO THOSE TR Inv0Lvt wiPRE55L312AIION OF THE VE55EL, WENDOR ANALYSES AND NRC EVALUATIONS WOULD BE SIMPLIFIED. IT WOULD ALSO t
PR0dIDE AT LEAST A PARTIAL ANSWER TO 7ttE QUESTIONS POSED IN REG. GUIDr 1.2. "THERMAL SHOCK TO REACTOR VE5SEL."
CC7"*"iT 5f P"t
- F15 t NRR HAS DISCUSSED THE TECHNICAL ASPECTS COVERED BY THIS RIL WITH RES PERSONNEL. RESEARCH REGt.E2ING WARM PRESTRE55ING IS STILL UNDERWAY. ESPECI ALLY ITS RELEVAMCE ~0 CYLINDitICAL VE5S j
TG BE A PROMISING PHENCMENON FOR LIMITIMG CRACK EXTENSION DURING A T9ERT*>L 5 HOCK. THE DATA ARE STILL IN5bFF TO BE USED A5 A BASIS FOR LICENSING DECISIONS. A DETAILEO TECHNICAL EVALUATION OF THE RESEARCH DESCRIBED I AND THE NRR ASSESSMENT OF ITS RANCS OF APPLICABILITY WILL BE COMPLETED FOLLOWING RECEIPT OF MORE Sd35TANTIAL DATA.
FLCC O NT1 CM_OJ/_tJ/7 k G d IVENMA M :
)
IHIS PROGRAt HAS GENIRAIED SUSSIANIIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGIN 5 TO FAILURE OF REACTCR PRES 5URE VE5SEL5 UNE;8 FL AWS ARE PRESENT. CONSIDERING MATERIAL PROPERTIES AND 10ADIMC5.
THIS PROGRAM MAS PROVIDED INPUT TO NOC RIQUIREMENTS FOR FRACTURE TOUGHNESS OF PEE 55URE VE5SEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR 50 AND CONTRIBUTED DATA USED IN THE AEME CLDE WHICH MAS INCORPC2ATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE j
DATA CASE U3ED IN REOULATORY GUICE 1.99 CONCERMING THE EFf ECT OF CCPPER IMPURITIES ON SENSITIVITY t
CF STEEL TO IRRA01ATION AND 15 EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS 1
TO REGULATORY GUIDE 1.99.
l s
P v 0 ?S A". G i f iC E C t.~:VNTS 91 PqltNf!AL UTILIZATION OR VALUE OF RESEARCH RESUL15 IN THE RE60cATORY PROCESS Fil T:
17 FATE ISSUfDr OS/C5/78 PES DECISION UNIT: FUEL EDIAVIOR PIL TITLE: POWE2 BURST FACILITY (PBF) SINGLE ROD-PCMER COOLING MISMATCH (PCM) TEST RESULTS
$PFNip*ING CFFIC[(S):
RES Egg:
1-10 PBF EXPERIMENTAL PESEARCH PROJECT MG3:
R. VAN HOUTEN PRGGRAMS F;1_(trygyT3: CCM?tETED RESEARCH IS REPCRTED ON SINGLE FUEL ELEMENTS EXPOSED TO PCMER-COOLING MISMATCH (PCM) CONDITIONS IN ftt rcutR BUEST FACILITY (PBF).
THE RESULTS ARE OFFERED FOR USE IN DETERMINING POSSIBLE CHANGES IK REQUIRED DEPARTURE-FRCM-NUCLEATE-BOILING RATICS (LMER'S) FOR ALL COMMERCIAL POWER REACTORS WHICH USE ZIRCALOY-CLAD URANIUM DIDXIDE FUEL RC05.
THE RESEARCH RESULTS SHOW THAT ZIRCALOY FUEL RCD CLADDING NCEMALLY DJES NOT FAIL EVEN WHEN PROLONGED FILM BOILING OCCURS uS A RESUL. OF INADEQUATE COOLANT FLOW RATES.
THE CLADDING GENERALLY WILL NOT FAIL UNLESS IT BECOMES SO HEAVILY CXIDIZED THAT IT IS SPITTLE AT ROOM TEMPERATURE. SUCH SEVERE ZIRCALOY OXICATICN WCULD REQUIRE HIGHER CLA0DIh5 TEMPERATURES THAN ARE CURRENTLY PREDICTED FOR ANY LIGHT WATER dEACTOR ACCIDENTS WHICH RESULT IN A PCM.
WHE!HER RELATED TO A LCSS OF CCOLANT FLCW OR TO AN INCREASE IN FUEL RCD POWER.
USER DISCUSSION POSITION COMMISSTON ACRS PRESS OFFICE MEETING PAPLR BRIEFING BRIEF'NG RELEASE RESULTS
[% T_R[L A CTIVJ TE$
- TylPJ iLF,L D CO*tPLFiED
!iELD MELD 11F1Q_
IMPLEMENTEQ CF FIC E / ESF CN S I 8 t f..........NAL'SD SCHECULED COMPLETION DAIE..C7/35/73 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COM?LETION DATE.....C3/22/73 Ii%(WNT3y _p/e t / 73,_DJM T rW :
VARICUS FUEL FAILURE MECHANISMS AND THE CONSEQUENCES OF FAILURES ARE T OC?lf W 11C HlO T L M M PTOU.f_ROCES$l tVALCATED IN IhE SAFETt ANALYSIS OF TRAhSIENTS AND ACCIDENTS. DEPARTURE FRGM NUCLEATE BOILING (DMB) IS ASSUMED TO PRODUCE FUEL ROD FAILURE AND IS THF FAILURE CRITERION USED FOR MANY LICENSING ANALYSES. PELLET /CLACDING INTERACTION (FCI) CAN ALSD EE A FUEL FAILURE MECHANISM. PBF PROVIDES THE CAPABILITY FOR STUDYING FdEL BEHAVIOR AND FAllt!RE MECMANISMS UNDEL T2ANSIENT AND ACCIDENT CONDIT0NS, DEiC?lil_UIPAsl_9[_8E5VLT}: THE CEMONSTRATED ABILITY OF MOST FUEL RODS TO EXPERIENCE DMB WITHOUT FAILURE SHCWS THAT THE CO2 RENT D::6 CRIItAIGN IS CCNSERVATIVE. REQUESTS FOR LESS CONSERVATIVE FAILURE CRITERIA HAVE Bt*H MADE BY THE INDUSTRY.
TURBINE TRIP WI!HOUT BYPASS (TTUGB) AND STEAM LINE BREAK (SLB) ARE NEAR-LIMITING EVEN.'S IN WHICH DNS IS PREDICTED TO OCCUR MSMENTARILY YEI FAILURE BY THIS MECHANISM MAY MOT OCCUR.
DEFINITION OF A LCSS C*.MSEPVATIVE FAILURE CRITEkION FCR THESE EVENTS WOULD RELIEVE THESE LIMITING CONDITONS. BASED ON THESE r?F RESULTS. Ntt UILL GIVE SERIOUS CONSIDERATION TO THESE VENDOR REQUESTS; HCWEVER, APPROVAL OF RELAXED FAILUEL CRITERIA WILL BE CCST!MGENT UPON THE EVALUATICN GF OTHER NON-DNS FAILURE MECHANISMS.
TME SIGGLE RCD PCM TEST RESULTS SHCW THAT THE CURRENT CH3 FAILURE CRITERION 15 CONSERVATIVE.
CO3CIN[1L? LNiill -
til RIDE Rah 0E OF CLADOING TEM /ERATURES DURING DNB WHEN TEST PARAMETERS ARE NEARLY THE SAME PREVENT A QUANTITATIVE ASSESSMIKT CF THE MARGaN TO FAILURE.
FURTHERMORE. THE PAWEL CLADDING EMBRITTLEMENT CRITERION WOULD NEED ADDIT C53; REVIEW BEFCRs t CUANTITATIVE MEASURE OF MARGIN COULD BE USED IN LICENSING MATTERS.
THE *CM TESI SERIES WAS DESICNED PRIMARILY TO STUDY THE EFFECTS OF DNB.
THERFORE PELLET / CLADDING INTERACTION (PCI, IAT* FRO?. IHEiE EXPERIMENTS WERE PROBABLY COMPRCMISED BY THE TEST CONDITONS. FOR EVANPLE, FUEL EXPOSURE IIME AT *tMEP., FUEL PRECCNDITIONING AND CLADDING TERMPERATURE WERE NOT IDEAL FOR PCI STUDIES.
CGMCLUSIONS CN THE SJ? JECT GF FUEL FnILURE PROPAGATION DRAMM FRCM THE PCM SERTES MAY BE PREMATURE SINCE NENDCR FUEL RG3 PCESSURE CRITERI A MAVE BEEN CHANGED. NEW DESIGN CRITERIA. RECENTLY APPROVED BY NRC. ALLOW FOR INTERNAL MCD PESSURE TO EXCEED THE EXTERNAL SYSTEM PRESSURE DURING NORMAL OPERATION. IN ADDITION, THE FUEL POD BEHAVIC2 0F A SIEGLE ROD IN A CCED SMROUD IS NOT TYPICAL OF FUEL RODS IN A MdLTIPLE ARRAY.
THE FO*tMCCMING tvNDLE TESTS IN PBF SHOULO GIVE MDRE INFCRMATCN ABCUT FAILURE PEOPAGATICN.
1 M M.fil$ d. PIRN3453: N3 RESPONSE RECEIVED. -
--_._.__._._.______...m._m_.
m._ __-
__-..__m-----.
i
_ f?1SQt.flff.jCI(M5. Qt htf MTM itT ILIZA110'4 Ou V ALUE OF RESEARCH RESUL T$ IN INE REGU1M ORY PROCESS
)
J h. 2:
la DATE ISSUED:
81/09/77 RES DECTSION UNIT: RISK ASSESSMENT i
LL TITLE: FKINTIC C07? UTER C0Z*
)
}.528PIMEf f tf(51: RES EPS: NONE EESEARCH PROJECT f*GR:
Rt*LCP S 5TS: THE TRANTIC CC"PUTER C3DE IS USED TC'0ALCULATE THE UNAVAILABILITY OF ANY SYSTEM MODEL.
COMPREHENSIVE SuRWEILLAXE YESTING JYAtt;ATIONE FCR A SYSTEM ARE POSSIBLE WITH THE INCORPORATION OF TEST DOWNTIME 5. TEST
.4 INLFFICIESCIES. A%3 TEST-CAU323 FAILURES IN THE ANALYSIS OF SYSTEM MODELS.
THE FRANTIC CODE HAS POTENTIAL 51C'4IFICANT APPLICATICM IN FVALUATING TECNNICAL SPECIFICATICNS ON TESTING AND ALLOWED D0%NTIMES FOR REACTOR SAFETY SYSTEi*S INE EV ALUATIONS CAN 3E OF A GENEstIC NATURE. OR CA.4 BE APPLIED TO SPECIFIC PLANT SYSTEMS IF APPLICABLE i
SLT% ARE AVtILASLE.
V USER DISCUSSION Poh! TION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING ERIEFING RELEASE RESULTS P_nT_ e t_LagT W pi$
gtyj t'4 ILUE CCP-Pt ET ED HELD HELD ISSUED MPL EMENTED UtFIfi US.5PON3tAtt........, htR/SD
+-
NRR i
SCHEDUL ED COMh ETION DATE.. 01/09/7.?
Ud5CHED UNSCHED UNSCHED UNSCHED UNSCHED 02/01/78 i
ACTUAL CO*PLETE0:4 DATE..... 12/36/77 01/15/73 ESM5M.hJ'WMEt_197 5 F. MfRfRG:
M.SiliMP M All31*8 LM?!LLAIMLfjeCJ$$: THE SUBJECT RIL ANNOUNCED THE AVAILABILITY OF THE FRANTIC CCMPUTER CODE.
I? TIS CC '.'U fit ML 15 RICOGt41 ZED BY KRR AS AN ITU*DR T AN T TOCL TO QUANTIFY SYSTEM UNAVAILABILITIES. THE FRANTIC
]
CG*TUTER COD" 15 EEI:'G USED BY fCIENCE APPLICATIONS. INC.. TO PERFOR't SENSITIVITY STUDIES ON ALLOWABLE OUTAGE TIMES 3
FOR ECC5 CG:UJ?:ENTS As A PCRTRrtOF THE EFFORT UNDER CONTRACT NO. NRC-03-07-059.
j Ot %MfEPXT OLRES!1L Y$: THE SAI CCNIRACT .FFORT HA5 BEEN CCMPLETED. THE FRANTIC CODE WAS UTILIZED TO DETERMINE g
IHOSE FACICS TTt:HICH 5) STIM UNAVAILABILITY IS MOST SENSITIVE INCLUDING. FOR EXAMPLE. TEST DOWNTIME. REPAIR TIME, E
j TEST ERFICIENCY, TEST OVERRIDE CAPABILITIES. POSSIPLE TEST CAUSED FAILURES AND TEST STAGGERING.
j THE F M CIP CODE HAS BEE'i S W I TO BE POTENTIALLY USEFUL IN ESTABLISHING TECH. SPECS. FOR SPECIFIC 5AFETY SYSTEM 1
DES I G':*e. MOC YER. SAI SUCGESTED THAT ADDITIO :AL. MORE COMPLETE AND CCMPREHENSTVE AN ALYSES BE PERFORMED IN THE FUTURE.
MJ.N T;. 'J.O AliS : C3R LIPLCTS TO UTILIZ.E THE FRANTIC COMFdTER CODE. EITHER DIRECTLY OR UNDER TECHNICAL ASSISTANCE l
CCM E ACf w 30 QUANTIFY SYMEM UMAVAILABILITIES O'i SPECIFIC OPERA!ING PLANTS AS THE HLED ARISES. THE CALCULATIONAL RESULTS Miv AID I.M DETERNINING AFPROPRIATE LICINSING ACTIOM.
I NO POST RIL ACTIVITIES ARE PLANNED.
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&ctC9 t7 C41Ct-CD + NTS (M PO f t t4TI At itilLIZtit04 GR VALUE OF REM MCd RESUtis IM THE RMUL ATORY PROCESS Fil 2:
19 DATE !$5UED: 21/31/73 RES DECTSION 'jNTT RISK ASSESSMENT
- It TITLE: G3 MEkE0'JCLcGY ASSESSMENT 1
i S*CM$r# INS OFFICE (S): RLS PR.G: NONE RESEA""CH PROJECT MGR:
J. PITIMAN
?(1_f C"*1NJ{1:
GC PROVICES A METHOD FCR SYSTEM MODELING PND A COMPUTER CODE TO CALCULATE A PREDICTION OF SYSTEM liLL I A *,I L I I t. TbE STUOY DLMOMSTRATE$.' HAT TitI5 METHOD PROVIDES EQUIVALENT RESULTS TO TH35E OSTAINED FROM FAULT i
IREE AN ALYSIS L"MICH :AS USED I'4 THE REACTOR SAFETY STUDY.
THE MODEL RESEMBLES THE SYSTEM SCHEMATIC CR PIPING i
DI AORAM L"iICH REDUCES THE SURDEN CF MODELING ALL SYSTEM COMPCNENTS. GO HAS A POTENTIAL SIGNIFICANT USE AS A i
MEANS CF DETEFMINIMO SYSTEM RELIABILITY CR AS A DIVERSE METHOD FOR VERIFYING SYSTEM ANALYSIS PERFORMED USING F Alf L T TREE CR SIMILAR MCDELING TECHNIQUES.
USER DISCUSSION FOSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING ERIEFING RELEASE RESULTS i
EPST31L_3CYJvlfm gtVE y H5tp CnMPLETED
- iEL D HELD IS$ttED IMPLEMENTE3 CFF ICE A LSPL%I5L E......... N2%/SD 1
NMSS j
SCHEDULED CCMh;!'cM DATE.. 33/3160 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL CCMPLETION DATE.....
=
1 ELCCEMIS, NUvfMP E*_'S7', _ U EP'JCO2 i
f1M3 Ilk _Pf11C A T IQUn F t r',* t !1GIf';2GE51: THIS METHODOLOGY HAS NOT REEN AND IS NOT PLAN'NED TO BE USED IN UPCOMING l
5AFLIY WLVILi" HCmv let. EVEN I ncGGet 2HE MFTHODS HAVE NOT EEEN DIRECTLY APPLIED IN THE LIM aiING PROCESS. WE CONSIDER i
THAT THEIR USE w3ULD A03 TO TE CEMERAL KNOWLEDCE OF THE NRC STAFF IN THE AREA 0F REACTOR SYSTEM RELIABILITY EVALUATION.
'l DESC'J Bi_YJ_*J ClJF_?WJ1T5 : THERE HAS BEEN NO IMPACT OF THESE RESULTS TO 2 ATE ON THE LICENSING PRCCESS. THE A5510CtM T OF EL5004Ct5 10 1HIS ACTIVITY HAS 3EEN PRECLUEED DUE TO THE NEED TO SERVICE HIGHER PRIORITY TASK 5.
j CCmEMT 5 /R EPS E(F $ 3 NOME 1
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MSS CL"*!% f 5, S. HATTER: N3..ESPCNSE RECEIVED.
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PROGRAM OFFICE COMMENis 01 PGtEN11AL UilLI2ATION OR VALUE OF RESEAPCH RESULT 5 IN THE REPJLATORY PROCESS RIL 8: 20 DATE I55UED! G1/24/73 PES DECISION UNIT: SAFEGUARDS RIL TITLE ~ A STUDY OF PHYSICAL PROTECTION EQUIPMENT 1ESEiQ'IN". OFFICE (5): IIE EES:
4-4 SAFEGUARDS EQUIPMENT EVAL. RESEARCH PROJECT MGR:
E. RICHARD EE5_CD5ML5Thz THE PURPOSE OF THIS STUDY WAS TO PROVIDE THE Ntc INSPECTOR, LICENSING REVIEWER AND FIELD EVALUATOR WIIH NLW AND IMPROVED METHODS FOR EVALUATING PHYSICAL PROTECTION EQUIPMENT THAT IS IN U55 OR PROPOSED FOR U5E AT LICENSED NUCLEAR FACILITIES. THE FIVE MAJOR PRODUCTS OF THIS STUDY ARE:
1 A CATALOG OF PHYSICAL PROTECTION EQUIPMENT.
2.
A GUIDE FOR EVALUATION OF PHYSICAL PROTECTION EQUIPMENT.
3.
A E00E OF REFERENCE MATERIAL 5 (RELEVANT TO THE EQUIPMENT CATALOG AND THE EVALUATION GUIDE).
4.
A SET OF GUIDELINES FOR DEVELOPING A METHODOLOGY TO MEASURE LEVELS OF EFFECTIVENESS FOR A FIXED-SITE PHYSICAL PROTECTIOd SYSTEM.
5.
A
SUMMARY
REPORT, INCLUDING RECOMMENDATIONS FCR FURTHER WORK.
ALL CF THE AEDVE PRODUCTS HAVE EEEN DISTRIBUTED TO THE VARIOUS NRC REGIONAL OFFICES AND ARE PRESENTLY BEING USED BY INSPECTORS AS BASIC REFERENCE DOCUMENTS FOR EVALUATING PHYSICAL PROTECT!GN EQUIPMENT INSTALLED AT LICENSED NUCLEAR FACILITIES. DATA FRCM THESE DOCUMENTS WERE AL50 USED IN THE DEVELOPMENT OF A NEW NRC REGULATORY CUIDE ON INTERIOR INTRUSION DETECTION ALARM SYSfENS BY THE OFFICE OF STANDARDS DEVELOPMENT. OTHER FEDERAL AGENCIES HAVE REQUESTED THE RESULTS OF THIS STUDY AS A MEANINGFUL COMPENDIUM OF AVAILABLE PHYSICAL PROTECTION EQUIPMENT AND EVALUATION EQdIPMENT TECHNIQUES. THE RESULTS OF THIS STUDY WILL BE USED IN PHASE II 0F THIS SAFEGUARDS RESEARCH PROGRAM A5 A B ASIS FOR EXPANDING AND IMPROVING THE DAT A AVAIL ABLE TO NRC STAFF REGARDING THE CHARACTERISTICS AND EFFECTIVENESS OF CCMBINATIONS OF PLYSICAL PROTECTION ZQUIPMENT, ANI THEIR ASSOCIATED ADMINISTRATIVE AND OPERATIONAL PROCEDURES.
THE DIVI 5?ON CF DOCUMENT CONTROL HAS BEEN REQUESTED TO PRINT THESE REPORTS FOR DISTRIBUTION ONLY TO OTHER AGENCIES AND NRC STAFF.
USER DISCUSSION POSITION CCMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS P05) PIL AC_11VJIJIl E(ElfW HELP
[9MPLFTED HEL D NELD ISSUED IMPLEMENTED OfEICL RESPONSISLL......... NkR/hM55/ --
5D/I4E SCHEDULED COMPLETION DATE.. C5/31/73 UN5CHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL CCMPLETION DATE.....
IIE CCMMENTS. G. WElis NO RESPONSE RECE1VED.
51LCO:MNTSJLR/ DOS R.
JDNC1:
1HIS PROGRAt1 CONDUCIED BY THE MITRE CORPORATION RESULTED IN A MULTI-VOLUME REPORT ON PHYSICAL PROTECTION HARDWARE IPCLUDING CESCRIPTIONS, PERFORMANCE CHARACTERISTICS AND MANUFACTURERS SPECIFICATIONS.
THIS W3PK HAS MATERIALLY ASSISTED IN THE PREPARATION OF SEVERAL TECHNICAL REPORTS ON SPECIFIC TYPES OF HARDWARE AND HAS PERMITTED THE CANCELLATION 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 EEING PREPARED IN CONNECTION WITH THE FUEL CYCLE FACILITY SAFEG'JARDS UPGRADE RULE.
EU}1_[gg51t!T11_S. MATTER: N/A - REPORTS WERE REVIEWED. NM55 PLANS TO REFERENCE PARTS Oc THESE REPORTS IN A FORIHCOMI.4G GUIDANCE PACKAGE TO LICENSEES. THE UPGRADE RULE GUIDANCE DEVELOPMENI WORAING GROUP REFERRED TO THESE DOCUMENTS IN THEIR PREPAPATION OF THE FIXED SITE PHYSICAL PROTECTION UPGRADE RULE GUIDANCE COMPENDIUM.
1, f979 F.
F' A C A N3 t LQCC""{'113 C'4
'J3T ELMM il".,J Q M E9 b f0PY PPCS$33: SEE COMMENTS /QEMARKS.
FtMlW3: SnE CG:"AEN15/REMAR.K3.
PhiCflRL J't*Si INE FIVE REPCRIS D UREG-0273, 0271, 0272, 0273, 0274) FORM A THCROUGH. THOUGH NOT EXHAUSTIVE,
$000Uil)?R Ud PD s TWESE DOCUMENTS PROVIDE TME DESCRIPTION OF PHYSICAL SECURITY HARDWARE AND PROVIDE SCMC BASES FOR ITS EVALUATION.
BASIC FAMILIAR!ZATICN WITH AND 50Mr BASELINE DATA FOR PHYSICAL SECURITY EQUIPMENT TO THE LICEM5ING REVIEMER.
REVIEM FOR CCM"ENT INITIATED C3/22/73.
- - ~-
PPoreAM OFFICE CcMMENTS ON FOIENTIAL UTIlll.ATION OR VALUE OF RESEARCH RESUtis IN THE *.EGULATORY PROCESS RTL 8: 21 DATE ISSUED: 03/24/78 RES DECISION UNITr RISK ASSES 5 MENT PIL TITtEs CRITIC 1 REVIEW OF SODIUM HYDROXIDE AEROSOL T0XiCITY SPONSCRING nFFICE(5): -RES EXD: NONE RESEAPCH PROJECT MGR*
M. COLLINGFORD l
Ef1_C0F!EM_T_5 : THIS ucRK CONSISTED PRIMARILY GF A REVIEW OF RELEVANT LITERATURE (WITH SOME PRELIMINARY SUPPORTIVE ANALYSIS) PERTAINING TO THE T0XICITY OF SODIUM HYDROXIDE (NADH).
ONE INSIGHT HAS BEEN THAT 5031UM IN TME HYDROXIDE FORM, FOLLOWING AN INCIDENT INVOLVING SCDIUM RELEASE. MAY NOT EXIST IN SUFFICIENT AMOUNTS TO WARRANT FURTHER ATTENTION. IN ADDITION, THE CHEMICAL SPECIES THAT WOULD BE PRESENT IN APPRECIABLE QUANTITIES (NA2CO3) FAY NOT BE OF CONCERN IN TERMS OF HEALTH EFFECTS.
THE PRINCIPAL FINDINGS WHICH SUBSTANTIATE THE A30VE INSIGHTS ARE:
/
(A) FOR RELATIVE HUMIDITIES EXCEEDING 35%. IT APPEARS THAT NADH DROPLETS IN THE ATMOSPHERE WILL BE TRANSFORMED TO SODIUM CARBONATE DECAHYDRATE IN LESS THAN A MINUTE IF THE NADH AEROSOL CONCENTRATION IS LESS THAN OR EQUAL TO ABOUT 100 MG/M3.
THIS TRANSFORMATION WILL TAKE LONGER IF THE RELATIVE HUMIDITY 15 LESS TH.aN 35%.
(B) THE ALKALINITY OF A SODIUM CARBONATE SOLUTION WILL BE SUBSTANTIALLY LESS THAN THAT OF A SODIUM HYDROXIDE SOLUTION OF THE SAME NORMALITY; THUS, CARBONATE AEROSOLS WILL BE LESS HAZARDOUS. PER SODIUM ATOM. THAN HYDROXIDE AEROSOLS.
(C) THE TRANSFORMATION FROM THE SODIUM HYDROXIDE TO SODIUM CARBONATE DECAHYDRATE INCREASES THE AERODYNAMIC DIAMETER OF THE AEROSOL BY APPROXIMATELY 40%.
THIS INCREASE IN DIAMETER SHIFIS SOME OF THE AER05DL DUT OF THE RESPIRABLE RANGE AND THUS LOWERS THE RESPIRABLE FRACTIDM OF THE AER050L. HYDROXIDE OR CARBONATE PARTICLES ENTERING THE UPPER RESPIRATORY TRACT WILL ABSORB WATER AND GROW 50 THE RESPIRA5LE FRACTIDM WILL DECREASE.
USER JISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS E0.5T RJi_AfilvilJE1 EEVIEW HEL D COMPLETED HEL D HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE.. 05/29/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCI;ED ACTUAL COMPLETICH DATE.....
05/01/73 tL*P.GOENi}_0NR01/73, W G A_=f f1( t m
MM3]$L AP_PUC AEQN_10EG')MICRY PPOCU$r THE REPORT PROVIDES THE LATEST AVAILABLE INFORMATION ON SODIUM HYDROXIDE AEROSOL BLHAVIOR AND 115 T0XICOLOGY.
THIS INFORMATION IS NEEDED IN OUR REVIEW ON CCNTROL ROOM HABITABILITY AND OFF-SITE CONSEQUENCES FOLLCWING A POSTULATED ACCIDENTIAL RELEASE OF SODIUM METAL AND 50DIUM FIRE.
PGPJ B EJ"P 4_CLO F_P11tJL T S : THE REPORT FORMS A BASIS FOR CONSIDERING OTHER SPECIES LESS T0XIC AND MORE READILY FORMED IHAN SODIUM HrDROXIDE IN OUR REVIEW OF SODIUM HAZARDS.
COMMENTS /REMAPES: AUTHORS RECOMMEMDATIONS FOR FURTHER WORK SHOULD BE CONSIDERED WITH RESPECT TO NRR'S NEEDS.
((PJLf9I"#ENl$, HOV_LMSEP 9, 1979, L _(pMG/Y. SPEIst D11CRI M_F PLICATION 103ML A191Y PRCCES$:
THE AUTHOR HAS SURVEYED T'fE T0XICITY OF FUMES THAT EMANUATE FROM A LARGC FIRE OF HON-RADIDACTIVE SODIUM.
HE HA5 SHOWN THAT THE INITIAL PRODUCTS SODIUM OXIDES ARE ALMOST IMMEDIATELY CONVERTED TO SODIUM HYDROXIDE IN THE ATMO5PHDtM. AND WITHIN A FEW MINUTES. TO THE MUCH LESS 70XIC SODIUM CARBONATE. THERE IS AN APPLICATION OF THIS WORK IO THE ANALYSIS OF SECONDARY SODIUM FIRES IN LMFERS. s
Di> P fjPACT OF.OE37 Q12 VHE CESULTS SEEM TO CONFICM THAT THE COMM30LY U3ED 'JS CZILLING OF 2 MG/M3 (OR NADH a 0BABLY ACCLPIAWLE FOR ACCEDENT CALCULAT2GitS. THERE ES INSUFFICIENT EVIDENCE HOWEVER TO PERMIT THE INCREASE OF YHI3 LIMIT BASED ON VHE TRANSITION TO NA2003, ALTHOUGH THIS TRANSITION SEEMS PRDBABLE WITHIN A FEW MINUTE 5.
E*"jEN_T 5/ P FMEP K S : EXPERIMENTAL MORK ON THE KINETICS OF THE TRANSITION OF NADH TO NA2CO3 WOU'3 BE REQUIRED IF IT 15 DESIRTD TO I AKE ADVANTAGF OF THIS TRANSITION IN ASSESSING THE HAZARDS FROM THE FUMES OF A 30DIUM FIRE.
ADDITIONAL STUDIES OF T:iE T0XICITY OF NADH AT EARLY STAGES AND LORING THE TRANSITION MIGHT A',50 BE REQUIRED. FOR EXAMPLE, EFFECTS OF PARTICLE SIZE A5 WELL A5 FUME DENSITY MAY BE IMPORTAN7.
1 5
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"kOG2AM 0FFICE CCn"ENTS ON POTENTIAL UTill2ATION OR VALUE OF RESEAPCH RESULTS IN THE REGULATORY PRCCFSS RIL #: 23 DATE ISSU12: 04/10/78 RES DECISION UNili SAFEGUARDS PIL TITLE:
"EASI" ADVERSARY SEQUENCE EVALUATION MODEL (COMPUTER GRAPHICS VERSION)
%PONSORING OFFICE (5): NMSS EES:
4-1 EFFECTIVENESS RESEARCH PROJECT MGR:
R. ROBINSON EVALUATION RES CCMMENTS: RESEARCH HAS BEEN CCMPLETED ON DEVELOPING A GRAPHICS DISPLAY VERSION OF A COMPUTER MODEL CALLED ESTIMATE OF ADVERSARY SEQUENCE INTERRUPTION (EASI), AND RESPONDS TO AN EXPRESSED HEED FOR EVALUATIVE METHODS FOR FIXED-SITE THEFT AND SABOTAGE PREVENTION SYSTEMS.
DOCUMENTATION HAS ALREADY BEEN MADE AwAILABLE THPOUGHOUT HRC CONCERNING PROGR AM:!ABLE POCKET CALCULATOR VERSICNS OF THE EASI MODEL.
THE OBJECTIVE OF THE *EAdi" METHOD IS TO PROVIDE A USABLE EVALUATION METHDD WHICH CAN SERVE AS EITHER A PHYSICAL PROTECTION SYSTEM DESIGN AID CR AS A DECISION AID IN THE LICENSING AND INSPECTION PROCESS. THE EASI GRAPHICS PROGRAM ALLOWS THE USER TG INPUT FACILITY AND ADVERSARY PATH ATTRIBUTES AT L CGMPUTER GRAPHICS TERh!NAL, AND OSTAIN AS OUTPUT A CRT "PERSPECTIVE VIEW" LINE PLOT.
THE METHOD CAN TREAT B0!H THEFT AND SAEDTAGE OBJECTIVES BY THREAYS OF INSIDERS, OUTSIDERS, AND COMBINATIDNS OF EACH GROUP.
l THE PESULTS OF THE EASI ANALYSIS ARE EXPRESSED IN TERMS OF THE PR03 ABILITY THAT THE PHYSICAL PROTECTION SYSTEM CAN RESPOND IN TIME TO INTERRUPT AN ADVERSARY ALONG A PHYSICAL PATH (ACTION SEQUENCE). TO SUPPLEMENT THE EASI CALCULATIONS, EASI GRAPHICS PRCVIDE! THE ANALYST WITH A SELECTION OF SIX TWO-DIMENSIONAL AND EIGHT THREE-DIMENSIONAL PLOTS.
THESE PLOTS ALLCM THE USER "O EXAMINE THE SENSITIVITIES OF VARICUS COMPONENTS ALONG THE ADVERSARY'S PATH AND TO STUDY THE IFFECT ON THE PROB;BILITY OF INTERRUPTION OF VARYING THE PERFORMANCE OF THESE CDMPONENTS.
IT IS RECOMMLNDED THAT THE EASI METHOD EE USED BY NMSS AND OTHER OFFICES AS AN ANCItiARY AfD IN DEVELGPING PERFORMANCE-ORIENTED REGULATIONS OR IN CARRYING OUT A COMPREHENSIVE EVALUATION PROGRAM.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETINS PAPER BRIEFING BRIEFING RELEASE RESULTS E0_S U l L ASllVlllEl RfV111]
HELD C0r"PL ET ED HELD HELD ISSDED IMPLEMENTED OFFICE RESPONSIELE......... f.a55 SCHEDULED CCMPLETION DATE.. 06/10/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL CCMPLETION DATE.....
EUSS COMMENTS, B. HATTER: TEST APPLICATIONS ARE BEING CONDUCTED TO DETERMINE USER SUITABILITY.
fiUSLC 0_'"7.NT32_qd f t!IS : THE RESULTS OF THESE EFFORTS HAVE BEEN USED IN THE CURRENT DEVELOPMENT OF IMPROVLD EVALUATIVE METHODS FOR FIXED SITE SAFEGUARDS.
.tiRP_CELNI$_QN A@UST 1,
1979 F. P A G A 'Q :
RSSRILfLf PPdG4110N 10 REG 9LATORY PROMil: SEE COMMENTS / REMARKS.
DL1CPIB LMPlCLQF PE W }: SEE CC.T1ENTS/ REMARKS.
CFLUt115Cf f dEE5$: THE AEY TO THE UTILITY OF EVALUATION MODELS IS THE RATIO OF THE EFFORT REQUIRED TO RUN THE MOLEL 10 THE USEFULNESS OF THE RESULTS DBTAINED. ALTHOUGH SAFEGUARDS EVALUATION MODELS DEVELOPED TO DATE TEND TOWARDS Ih0RDINATE INPUT COMPLEXITIES BY COMPARISON WITH OUTPUT UTILITY, EASI HAS BEEN A WELCOME EXCEPTION TO THIS TREND.
CONSEQUENTLY, EASI H35 BEEN USEFUL AS AN AID TO LICENSING DECISION MAKING. AND HAS BEEN DISTRIBUTED BY NRR TO THE PCMER REACTOR LICENSEES AS AN AID IN PHYSICAL SECURITY SYSTEr1 DESIGN. THE ADDITIONAL COMPUTER PROGRAMMING WHICH PERMITS THE USE OF CCMPUTER GRAPHICS IN CONJUNCTION WITH THE EASI CODE MAY EE USEFUL FOR PARAMETRIC DESIGN STUDIES. ALTHOUGH iT PROVIDES LITTLE ADDITIONAL UTILITY TO THE LICENSING RE"? EWER.
ADDITIONS AND FURTHER VARIATIONS OF THIS SIMPLE TOOL APPEAR TO PROVIDE DIMINISHING RETURNS..
__________m_
- -- _ - _.__.. _ _ ~ _ _ - _ - _ _
_ PROM 41 OF FIC E COFT1EN T S ON POTENilAl O'!tlZAfl01 OR VALUF OF RESEARCH RESULTS IN THE REGULATORV PROCELS PIL 8-24 DATE ISSUE 7: 04/10/78 RES DECISION UNIT: SAFEGUARDS PIL TITLE:
"TESEM" ADVERSARY SEQUENCE EV'tUATION f.0 DEL SPONSORING OFFICrt$y: HMSS py:
4-1 EFFECTIVENESS PESEAPCH PROJECT MOR:
R. ROBINSON EVALUATION P E 5_C C"$ T $ 2 FESEARTH HAS BEEN COMPLETED ON THE FORCIBLE ENTRY SAFEGUARDS EFFECTIVENESS MODEL (FESEM), IN RESPONSE TO A NEED TOR EVALUATIVE T1ETHODS FOR FIXED-SITE THEFT AND SABOTAGE PREVENTIDH SYSTEMS.
THE PURPOSE OF THIS STUDY ut.S TO DEVELOP A METHODGLOGY FOR ANALYZING FIXED-SITE SECURITY SYSTEMS AS TO THEIR EFFECTIVENESS AGAINST A FCRCIBLE ATTACK BY AN ADVERSARY INTENT ON CREATING AN ACT OF sat 0TAGE OR THEFT.
THE MODEL PROVIDES A FRAMEUDRK FOR PERFCR*11NG INEXPENSIVE EXPERIMENTS REL ATED TO FIXED-SITE SECURITY SYSTEMS, FOR TESTING ALTERNATIVE DECISIONS, AND FOR DETERMINING THE RELATIVE COST EFFECTIVENESS ASSOCIATED WITH THESE DECISION POLICIES.
THE RESULTS OF THE FESEM ANALYSIS INCLUDE ESTIMAIES OF THE PROBABILITY OF SADOTAGE OR THEFT WINS (AND LOSSES) BASED ON ATTACK FCRCE SITE. ATTACK M33ILITY, AND TYPE OF ATTACK; COLLECTED STATISTICS ASSGCIATED WITH EACH VARIABLE (E.G.,
NUMBER OF WINS BY DEFENDERS AND BY ATTACKERS FOR SUCCESSFUL SABOTAGE OR THEFT, ALARM TYPES FOR ALL RUNS, TIME REQUIRED FOR SUCCESSFUL SABOTAGE OR COMPLETION OF THEFT, ETC.).
THE PROGRAM IS CURRENTtY AVAILABLE FOR NRC USE VIA AN ACCESS CODE NUMBER TO SANDIA'S COMPUTER. A TRAINING PROGRAM WAS GIVEN IN FEBRUARY l'778 TO INTERESTED NRC PERSONNEL AND POTENTIAL USERS.
IT IS RECCT*NENDED THAT THE FESEM MODEL BE USED DY NMSS AND OTHER OFFICES AS AN ANCILLARY AID IN FORMULATING REGULATORf REQUIREMENTS, LICENSING. INSPECTION AND OTHER MONITORING OPERI.TIONS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOS T_ PJ Le C.T J V LT IM ELVJIJJ HEtD COPPtETED HELD HELD ISSUED IPPLEMENTED OFF)CE RLi,PONSIBLE......... M5S/NRR t,RR SCHEDULED COMPLETION DATE.. 06/10/78 UNSCHED UNSCHED t'NSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLE' 0N DATE.....
tim _S$ Cf"EN 13.
3.
HtTTER - TEST APPLICATICNS ARE BEING CONDUCTED TO DETERMINE USER SUITABILITY.
tyyS CO*VNTSJJf t13 - THE RESULTS OF THESE EFFORTS HAVE BEEN USED IN THE CURRENT DEVELOPMENT OF IMPROVLD EVALUATIVE METHODS FCR FIXED SITE SAFEGUARDS.
MPP C ? "*EN T $_ C N A000$T 1,_1_9 L9,_F, PAGAg :
DMC R I B E A PPllCAUp T o PE0'JL A LCl(_P";f)EM : SEE COMMENTS / REMARKS.
PQ C M Hi.J ff 6Cl_M._P1WLI h : S[E COM LNIS/ REMARKS.
C CfdRtil5 /f LFf_PJ $ : THE AEt TO THE UTILITY OF EVALUATION MODELS IS THE PATIO OF THE EFFORT REQUIRED TO RUN THE MODEL 10 THE USLFUL:RSS OF THE RESULTS. ALTrf0LGH THE FESEM OUTPUT PROVIDES INFORMATION M!sICH WOULD BE USEFUL IN THE DESIGN AND EVALUATION OF PHYSICAL SECURITt SYSTEMS FOR POWER REACIORS, THC UTILITY OF SUCH INFORMATICH (MUCH OF WHICH IS INI!!ITIVELY 03VIOUS TO A PHYSICAL SECUR*TY EXPERT) DOES NOT APPEAR TO WARRANT THE EXTENSIVE EFFORT REQUIRED FOR INPUT PREPARATION At:D EXECUTICN OF THIS r"3 DEL.
IN ADDITION, THE ACCURACY OF THE CCMPUTED RESULTS CANNOT BE VERIFIED AS A RESULT CF THE LIMITED APPLICABILITY OF MILITARY-TYPE ENGAGEMENT MODELS TO THE PRCBLEM OF SABOTAGE OF A REACTOR FACILITY.
4
'd M M FNll_R132/_13fJ h _G.
PIV[Ny Rf*
j IHE RESULIS OF IHE EFFORTS WILL LE USED IN DEVELOPMENT OF STANDARDS FOR SAFEGUARDS SYSTEMS AS MELL AS BY LICENSEES IN DEVELOPING THEIR SYSTEMS AND BY NRC IN EVALUATING THOSE SYSTEF.S.
PP0064M OFFICE CCNnLNTS ON POTENilit UTILIZATInN OR VALUE OF RESEAPCH RESULTS IN THE REGULATORY
- ROC 11}
PIL 3: 25 DATE ISSUED: 03/21/73 RES DECISION UNIT: FUEL BEHAVIOR RIL TITLE: FRAP-53 SPONSORING DFFICE(S): NRR RPS:
1-12 FUEL CODE RESEAPCH PROJECT MCk:
G. MARINO DEVELOPMENT REi_LO55FNT}TCS OF A NUCLEAR FUEL ROD OPERATING UNDER STEADY-STATE POWER CONDITIONS,FRAP-53 IS A EEST.iTIMAT CHARACILRISI AND WAS DEVELOPED TO PROVIDE ACCURATE INITIAt VALUES OF FUEL-ROD PARAMETERS FOR INPUi INTO TRANSIENT ANALYSIS CODES SUCH AS FRAP-T AND RELAP.
IT IS CAPAELE OF SUPPLYING THE HOT-STATE VALUES OF SUCH QUANTITIES AS:
1 STORED ENERGY 2.
RADIAL TEMPERATURE DISTRIBUTIONS AT GIVEN AXIAL LOCATIONS 3.
TOTAL FISSION GAS RELEASE 4.
ROD INTERNAL GAS PRESSURE AND COMPOSITION 5.
CLAD EEFORMATION 6.
AMOUNT OF PELLET-CLAD INTERACTION (PCI) 7.
FUEL DEFORMATION (SkELLING, DENSIFICATION, RELOCATION, AND THERMAL EXPANSION) 8.
FUEL-CLAD G1P SIZE AMD GAP CONDUCTANCE 9.
CLAD-CORROSION AND HYDRIDING.
ALL OF THESE QUANTITIES ARE STRONGLY DEPENDENT UPON THE OPERATING HISTORY OF THE ROD, AND EACH WILL HAVE A LARGE EFFECT ON THE PREDICTED AND MEASURED RESPONSE OF A FUEL ROD DURING A T R A NSI C.H T. THE CODE, THEREFORE, HAS BEEN DESIGNED 10 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 CCDE PERFORMANCE IN PREDICTING THE AVAILABLE, QUALIFIED, 2XPERIMENTAL DATA, AND (2) TO IDENTIFY THOSE AREAS THAT REQUIRE MORE SOPHISTICATED MODELING OR MORE EXPERIMENTAL DATA.
THE CGDE PERFORMANCE AND DATA WERE ANALYZED USING STATISTICAL METHGDS.
- THUS, ALL OF THE MAJOR RESPONSE V'RIABLES ARE PRESENTED ALONG WI!H THEIR CORRESPONDING STANDARD ERROR BOUNDS.
THE VERIFICATION PROCECURE USED INFORMATI0h FROM OVER 700 FUEL RODS CONTAINING A WIDE RANGE OF OPERATING AND DESIGN PARAMETERS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFINJ RELEASE RESULTS PDSTJ 1LACTiv]1EES Efflfy HELD COMPLETED AELD HELD ISSUED IMPLEMENTED OFFICE RESPON31LLE......... f.R P/ SD SCHEDULED CCMPLETION DATE.. C5/21/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NPR_C050F.NTS ON fj/R&(lpt_p _RQi$:
[4.kC PJ1L A fl L] $ A_ tlc.N _10_KL@l110RY_f P QCES S : STEADY-STATE FUEL PERFORMANCE CODES ARE REVIEWED BY NRR AS PART OF LOCA AND D1HER ACCIDENI ANALYSIS.
NRR USES AN INDEPENDENT NRC-DEVELOPED CODE FOR AUDIT PURPOSES IN THESE REVIEWS.
Ef5CRTIE IMPAfl_Q'F PEST 2]JS s NO DIRECT IMPACT.
NRR USES THE GAPCON SERIES OF CODES IN AUDIT WORK AS IT HAS DONE SINCE BLFORE FRAP-S MAS EEVELUPED. NRR HAS NO PLANS TO USE FRAP-S IN THIS CAPACITY.
CUWjE N TS(RfM?d1S_ : THE FRAP-S CODE WAS DEVELOPED BY RES TO INITIALIZE VARIOUS ANALYSIS CODES. RES AND NRR HAVE RLCOGNIZLD INE DUPLICATE GAPCGN AND FRAP-S CODE EFFORTS AND HAVE CONSOLIDATED THESE EFFORTS INTG A HYBRID CODE FRAPCON.
FRAPCON WILL CONTAIN SOME ELEMENTS FROM FRAP-53, BUT FRAP-S DEVELOPMENT WILL BE DISCCNTINUED.
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12_faMME*87L R. QIVEESSNQ - VHE ENGINEEDING METHODOLOGY STANDARD 3 BRANCH PLAN 5 T3 USE THE RE9EAOCH RESULTS IN 115 OtiG01060. OEGRALED CORE COOLING TASK AND IN ESTIMATING THE FISSION POODUCY SOURCE TSRM FOR ENVIRONMENTAL ENVELCPE STANDARDS ACTIVITY. RESULTS MAY ALSO BE USED IN THE ECCS RULE CHANGE EFFORT AND, IN THE LONG TERM, IN REGULATORY GUIDES RELATED TO CORE DESIGN.
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PROCOM OF F IC E COMEN TS ON P0 f ENIX AL UTIUlATION OR VALUE OF RESEARCH RESULTS IN THE RECWLATORY PROCE'd PIL 3: 26 DATE ISSUED: 04/27/73 RES DECISION UNIT: FUEL CYCLE SAFETY 3 ENVIRONMENTAL EFFECTS RIL TITLE:
THE IMPACT OF OFFSHORE NUCLEAR GENERATING STATIONS ON RECREATIONAL BEHAVIOR AT ADJACENT COASTAL SITES SPONSORING OFFICE (5): NRR EPfg: 5-21 SOCIC-RESEAPCH PROJECT NGR:
D. PARNA ECONOMIC IMPACTS EfS_CF**l1LS: THESE RESULTS ARE OFFERED TO SUPPORT NRC COST-BENEFIT ANALYSTS WITH NEW AND IMPROVED INFORMATION FOR ASSESSIhG LIKELY IMPACTS OF NUCLEAR GENERATING STATIONS ON RECREATIONAL BEHAVIOR AT ADJACENT COASTAL SITES.
THE RESEARCH RESULTS INDICATE THATz (A) PROXIMITY OF A FLOATING NUCLEAR PLANT IS LESS IMPORTANT THAN OTHER BEACH ATTRIBUTES IN DETERMINING BEACH ATTRACTIVENESS; (B) PROBABLY NO MORE THAN (AND PERHAPS LESS THAN) 5% TO 10% OF CURRENT l
BEACH PATRONS WOULD AVOID A BEACH AFTER FNP SITING 3 MILES DIRECTLY OFFSHORE; AND (C) IMPACT OF AN FNP WOULD DECREASE t
EXPONENTI ALLY AS DISTANCE AWAY INCREASED.
4 IN SU"."'ARY.
THE PERCENTAGE REDUCTION IN TOURISM ATTRIBUTABLE TO SITING OF NUCLEAR POWER PLANTS OFFSHORE WOULD BE SMALL~,
BUT NOT NECESSARILY NEGLIGIBLE, AT POINTS CLOSE BY.
THE STABILITY OF THOSE IMPACTS OVER TIME, HOWEVER, DEPENDS UPON THE STABILITY OF CURRENT ATTITUDES TDMARD AND BELIEFS ABOUT NUCLEAR POWER AND ITS SAFETY.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PUSLRlLACTlVJT1f1 EfviEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED j
OFIICE PESPCNSIBLL......... NRR SCHEDUL LD COMPLETION DATE.. 05/29/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED l
ACTUAL CCMPLETICH DATE.....
04/27/78 FP_(QyNT3_C[(Q4R7/75, M.
ER N3 T_:
Pf M P LIlllE A_U.ON TO PF.f1!LA T_Q3f PROCESS: THIS STUDY IS BEING INCURPORATED INTO STAFF TESTIMONY AN FNP 1-8.
THE I ES110:0NY I3 CONCEU4ED WITH POTEN!!AL AVOIDANCE OF BLACH RESCRTS, BY TOURIST, DUE TO PERCEIVED RISK FROM FLOATING NUCLEAR PLANTS SITED IN THE VICINITY OFFSHORE. EXTRAPOLATION OF TOURIST BEHAVICR IN THE VICINITY OF LAND BASED NUCLEAR POWER PLANTS TO OFFSHORE PLANTS IS TENUQUS.
THE STUDY ALSO PROVIDED A BROADER GENERIC UNDERSTANDING OF RECREATIONAL BEHAVIOR UNICH 84ILL STRENGTHEN OUR CAPABILITY TO HANDLE THIS ISSUE IN EIS'S GENERALLY.
PAS.Q1131_1MP ACUF PMQLTj THIS STUDY HAS RESULTED IN STRONGER, MORE OBJECTIVE STAFF TESTIMONY CONCERNING TOURIST AVOIDANCE OF BEACM S IN IHE VICINIT't OF FNP'S.
THE ESTIMATES OF POTENTIAL AVOIDANCE HAS ALLCWED CALCULATION OF
.lKELY ECONUSIC IMPACT TO A RANGE OF COASTAL ECONOMIES. SATISFACTORY DISPOSITION OF THIS CONTENTION GENERICALLY IN THE FNP HEAPINGS SHOULD REDUCE OR ELIMINATE THIS CONCERN IN FUTURE LICENSING ACTIGHS FOR SPECIFIC FNP APPLICATIONS.
_QQ'1". LITS M AM3: THIS STUD" WAS AN APPLICATION TO NUCLEAR SITUATIONS OF SOCIAL RESEARCH TOOLS DEVELOPED TO ANALYZE HUMAN BEHAVICR RELJ!IVE TD RISK FROM NATURAL HAZARDS.
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PP0GuA1 CFFICE COM9ENTS 01 PUTENIXAL UTILIZATICN tR VALUE OF PESEARCH FESULTS IN THE REGULQTORY PROCESS RIL #2 27 pATE ISSUED: 06/02/78 PES DECISION UNIT: CCDE DEVELOPMENT PTL TITLE "LEACON/ MOD 2" SPONSCRING OFFICEfS): RES EPS:
1-15 CONTAINMENT CODE RESEARCH PROJECT MGR:
- 5. FABIC PE1_Cft5fyTS: THIS RIL TRANSMITS THE BEACON / MOD 2 COMPUTER COPE MANUAL, DESCRIBIS ITS FIELD OF APPLICATION AND DISCUSSES THE CODE'S STRENGTHS AND LIMITATIGHS. BEACON / MOD 2 IS AN ADVANCED, BEST ESTIMATE CODE INTENDED FOR EVALUSTION OF SHORT-TERM THERM 0HYDRACLIC CONDITIONS MITHIN "DRY" (FULL PRESSURE) MULTICCMPARTMENT CONTAINMENT 5, OR WITHIN CERTAIN REGIONS OF THE "PRESSURE SUPPRESSION" DR(UELL.
THIS RESEARCH WAS INITIATED TO PROVIDE IMPORTANT MODELING IMPROVEMENTS FOR BEST ESTIMATE ANALYSIS OF THESE CONTAINMENT SYSTEMS. BEACON /MDD 2 CCF R5 CONSIDERABLE ADVANTAGES OVER THE EXISTING CONTAINMEFT CODES FOR BEST ESTIMATE EVALUATION OF HYDRAULIC LOADS IN MUL TICCMPARTMENT PWR TYPE CONTAINMENTS. IT IS PARTICULARLY SUITABLE FOR EVALUATION OF THE REACTOR CAVITY LOADS (FOR POSTULATED BREAKS BETWEEN THE REACTOR VESSEL AND THE BIOLOGICAL SHIELD) IN BOTH PWR AND BWR CONTAINMENTS.
THE CODE HAS ALSD SHOWN A CAPABILI!Y TO DESCRIBE THE EVOLUTION OF A TWD-PHASE (FLASHING) JET AND THE RESULTING PRES $URE LOADS ON ThE IMPACTED BARRIER.
IT IS RECOGNIZED THAT THE CODE MUST BE MORE EXTENSIVELY TESTED AGAINST EXPERIMENTAL DATA.
MEVERTHELESS THE BEACON / MOD 2 IS RECOMMENDED FOR CALCULATIONS OF THE REACTOR CAVITY LOADS, FOR LOTH PWR AND BWR INSTALLATIONS, AND FOR EVALUATION OF JET IMPACT LCADS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EPST_?JL ACTJVillfi FEVILM HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICL Rf500hSIBLE......... NPR SCHECULED COMPLETIOM DATE.. 06/10/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED A C T U.'.6 COMPLETION DATE.....
EEg_[pMME N T), P.
TEPESID:
REM PR E_fEPLlC[il0N T0_RF 91A_TSPY PPOCEM: THE BEACON CODE AS PRESENTLY DEVELOPED, REPRESENTS A POTENTIAL TEST LSTIMAlt COMPUILR PROGRAM FOR THE CALCULATION BY NRR OF SUBCOMPARTMENT PRESSURE / TEMPERATURE RESPONSES. OF PARTICUL AR IMPORTANCE TO 00R LICENSING PROCESS IS THE ABILITY OF THIS CODE TO EVALUATE THE REACTOR CAVITY PRESSURE TRANSIENT IN TWO-DIMENSIONS A5 OPPOSED TO THE ONE-DIMENSIONAL COMPUTER PROGRAMS CURRENTLY IN USE IN THE LICENSING PRGCESS. UPON COMPLETION OF THE PLANNED CODE VERIFICATION EFFORTS, BEACON WILL BE USED TO BENCHMRK ALL LICENSING CCMPUTER PROGRAMS TO ESTABLISH A MEASURE OF THE MARGINS, WHICH CURRENTLY EXIST IN THESE LICENSING CODES.
DEiCEIIF_IEEA GI_SELEE5U LIS: THE ZEACON COMPUTER PROGRAM REPRESENTS A NEXT GENERATION COMPUTER PROGRAM FOR THE CALCULAIIGN OF SU3CCMPARIMENT PRESSURE / TEMPERATURE TRANSIENT RESPONSES. FOR THE PRESENT, HouEVER, AND PENDInG SUCCESSFUL CCMPLET;0N OF THE VERIFICATION EFFORTS, NRR DOES NOT PLAN TO USE THIS CODE IN THE LICENSING PROCESS.
WHEN THE CODE HAS BEEN EVALUATED AGAINST EXPERIMENTAL DATA AND COMPLETELY CHECKED 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 IN CUR CURRENT LICENSING REQUIREMENTS WILL DEPEND ON THE RESULTS OF THESE CCMPARISONS.
f005LhilfELUiEf}: THE BEACON CODE HA5 AN EXCELLENT POTENTIAL FOR BECOMING AN IMPORTANT LICENSING TOOL.
AS A RESULI, II IS RECOMNENDED THAT THE CODE VERIFICATION EFFORT BE PERFORMED IN A TIMELY MANNER.
THIS VERIFICATION EFFORT SHOULD TAKE PRIORITY OVER ANY ADDITIONAL MODIFICATIONS TO THE CODE WHICH ARE AIMED AT EXPANDING ITS BASIC ANALYTICAL CAPABILITY.
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PPCSS19 3FFICE CO*MENis 04 POTENTIAL UTILIZATION OR VAtVE OF PESE ARC 4 PESULTS IN THE REGULATORY *ROCESS EIL l' 21.
DATE ISSUED' 05/09/78 pf1 3ECISI0tM FUEL BEHAVICR PIL TITLE:
"NELT/ CONCRETE !NTERACTIONS" 1PPNSORINS OrFICE(S): RES 333:
1-13 FUEL MELT PESEADCH PR6 JECT MGR:
R. DISALVO Ef3_C025fM]3: TH:S RIL DESCRIBES THE INTER-1 CODE FOR CALCULATING THE EFFECTS OF INTERACTION BETWEEN MOLTEN MAILRI ALS AhD CONCRETE AND THE EXPERIMENTAL DATA BASE FROM WHICH IT WAS DEVELOPED. THIS WORK HAS RESULTED IN AN IMPROVED MODEL BASED ON EXPERIMENTS WITh PLOT 0 TYPICAL MATERIALS.
IN SEPARATE EFFECTS EXPERIMENTS, MONCLITHIC SPECIMENS OF CONCRETE WERE SUBJECTED TO CONTROLLED THERMAL FLUXES IN ORDER TO MEASURE RA ES OF ERGSION.
EROSION IS LINEAR WITH TIME FOR A GIVEN HEAT FLUX. AFTER CORRECTION FOR THERMAL LOSSES THROUGHOUT REFLECTICM AND RADIATION. THE DOMINANT MODE OF EROSION IS QUIESCENT MELTING OF THE CEMENT (I.E.,
THE BINDING MATERIAL), WITH NO DIFFEENCES OBSERVED BY VARYING THE C09 POSITION OF THE AGGREGATE-MATERIAL. THESE DATA ARE NECESSARY TO INTE*PRET EROSION RATES OBSERVED IN INTEGRAL EXPERIMENTS.
AS A RESULT OF THE INTEGRAL EXPERIMENTS, IN MHCH PROTOTYPICAL MOLTEN MATERIALS CONTACT CONCRETE. THE FOLLOUING CONCLUSIONS WERE DRAWN:
EROSION OF CONCRETE IS THERMALLY DCMINATED. WITH INSIGNIFICANT CONTRIBUTIONS FROM MECHfMICAL AN7 CHEMICAL EFFECTS.
THE PRINCIPAL MECHANISM OF EROSION I% MELTING OF THE BINDING MATERIAL, WITH NO pIGNIFICANT QUALITATIVE DIFFERENCES CAUSED BY CHANGING THE CCMPOSITICN OF THE AGGREGATE.
THE COMPOSITION OF THE CONCRETE 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.
HYDROGEN AND CARBON MONOXIDE ARE AM0hG THE GASES EVOLVED FROM THE SURFACE OF THE MEL'. AND THEY BURN UPON CONTACTING AIR.
THIS INDICATES THAT THE H2O AND CO2 RELEASED FROM THE DECOMPOSING CONCRETE ARE REDUCED CHEMICALLY, MOST LIKELY bY OXIDIZING THE METALLIC CONSTITUENTS OF THE MELT.
THE EXFERIMENTS HAVE CULMINATED IN AN ANALYTICAL MODEL (INTER-1) 0F THE MELT / CONCRETE INTERACTION WHICH CAN HELP EXTEND THEIR RANGE OF APPLICABILITY.
WHILE DIRECT EXTRAPOLATIDN OF THE DATA TO PROTOTYPICAL CONDITIONS MUST ALWAYS BE MADE CAUTIOUSLY, ENOUGH CONFIDENCE IIAS BEEN CEVELOPED SO THAT NO FUNDAMENTAL DIFFERENCES IN BEHAVIOR ARE ANTICIPATED IN SCALING TC FULL-SIZE SYSTEMS RESERVATIONS EXIST REGARDING THE APPLICATION OF INTER-1 TO PREDICT SUCH 4ARIABLES AS THE TIME OF CONTAINMENT MELTTHROUGH OR OVERPRESSURIZATION. THE MODEL CAN BCST BE UTILIZED IN ITS CURRENT FORM TO ESTIMATE THE RELATIVE SIGNIFICANCE 3F VARIATICNS IN PARAMETERS SUCH AS MATERIALS, PROPERTIES AND CCMPOSITIONS, INTERFACE HEAT TRANSFER COEFFICIENTS, GEG;tETRY, ETC.
'HE PRIMARY SIGNIFICANCE OF THE WORK DE3CPIBED IS THE IMPROVED UNDERSTANDING OF PHYSICAL PHENOMENA. RES RECOGNIZES THAT THE RESULTS ARE UNLIKELY TO HAVE SIGNIFICANT NEAR-TtRM IMPACT CH CURRENT LICENSING PROCEDURES. IT SHOULD, HONEVER, PROVIDE ADDITIONAL BACKGROUND INF0EMATION USEFUL IN ANALYZING REGULATORY ISSUES INVOLVING ACCIDENTS BEYOND DESIGN BASIS EVENTS.
USER DISCUSSION POSITIDM COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOS T_ P]_U C T [VJ U[3 D E V LEW HEi3 CON?tETED HEL D ELD jflQFJ_
IMPLEMENTED Of f ICE R ESI GNSIL L E......... NOR/SD SCHEDULED COMPLETION DATE.. G3/30/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION PATE.....
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Ef!(CCE"ET151 F. G8F5 lES CALCULATING THE EFFECTS OF n'NTERACTIONS BETWEEN MOLTEN MATERIALS h51PltE pf1]MT ps TO PECWMERY PROIE32 AND CONCElTE 15 DIRECILY APPLICABLE TO P.AKING L1 CENSING DECISIONS REL ATED TO CONTAINMENT REQUIREMENTS F02 ADVANCED REACTORS (SUCM AS LMFBR*5) TD ENSURE COMPARABILITY WITH PRE %ENT DAY LWR NUCLEAR POWER PLANTS.
THIS RESEARCH PROGRAM HAS IMPROVED OUR ABILITY TO PREDICT IHE TIME AND MODE OF CONT AINMENT FAILURE BECAUSE OF IMPROVED UNDERSTANDING CF THE INTERACTIONS BETWEEN MOLTEN CORE MATERIALS AND CONCRETE. SUCH AS THE RATE AND EXTENT OF CONCRETC EROSION. QUANTITY AND TYPL OF GASES RFLEASED, FISSION PRODUCTS RELEASE.
AND THE EFFECTS CF CONCRETE CHEMICAL COMPOSITION. AS DEMONSTRATED IN NJREG-0440,
- LIQUID PATHWAY GENERIC STUDY,"
DATED FELRUARY 1978, THE RESULTS OF THIS RESEARCH PROGRAM ARE ALSO DIRECTLY APPLICABLE TC EVALUATING QUESTIONS OF CC".?ARABILITY CF CONSEQUENCES OF 705TULATED CORE MELTDOM1 EVENTS AT LAND-BASED AND FLOATING KUCLEAR PLANTS.
Df1C?ll D EEA WYE Pf3 Mil}: THF IMPACT OF RESULTS OF THIS RESEARC H PRUGRAM ON CURRENT LICENSING PROCEDURES RLl!AINS TO BE LVALUALED.
THESE RESEARCH RESULYS HAVE PLAYED AN IMPORTA:ef ROLE IN NRR'S SAFETY EVALUATION OF THE CLINCH RIVER BREEDER REACTOR (CRFR) AND THE FAST FLUX TEST FACILITY (FFTF), PARTICULARLY IN REGARD TO EVALUATING THE CONSEQUENCES OF POSTUL ATED CORE MELTDOWN EVENTS AWD ESTABLISHING APPROPRI ATE CONTAINMENT REQUIRE-MENTS TO ENSURE CCMPARABILITY WITH PRESENT DAY LUR PLANTS.
THESE RESEARCH RESULTS HAVE AL50 HAD AN IMPORTANT IMPACT ON THE STAFF'S CONCLUSIONS CONTAINED IN THE "LIQUID PATHWAY GENERIC STUDY." NUREG-0127. DATED FEBRUARY 1778.
THE STUDY FOUND THAT THE RISKS ASSOCIATED WITH RELEASES TO THE HYDROSPHERE AT A FLDATING NUCLEAR PLANT (FNP) ARE GREATE1 THAN THOSE AT A LAND BASED FLANT (LBP) FOR CORE MELT ACCIDENTS. THE STAFF THEN ASKED 'HE APPLICANT TO MAKE DESIGN CHA.iGES IN THE PLANT TO MITIGATE THE CCNSEQUENCES OF THIS KIND OF ACCIDENT; SPECYFICALLY. THE STAFF IN THE DRAFT ENVIROKMENTAL f
STAILMENT, PART III (NUREG-0127), MAY 1978. AND IN A SUBSEQUENT LETTER TO OFF5HORE F0WER SYSTEMS (OPS)
(R.P. BALLARD TO A.P. ZECHELLA, JULY 25. 1978) REQUESTED THAT THE CONCRETE PAD BENEATH THE REACTOR VESSEL l
BE REPLACED LY SOME MATERIAL THAT PROVIDE 5 INCREASED RESISTANCE TO A MELT THROUGH BY THE REACTOR CORE.
FINALLY, THE RESULTS OF THIS RESEARCH PROGRAM mRE IMP.*0VING OUR ABILITY TO MAKE QUANTITATIVE RISK ASSESSMENTS ON FXISTING AND PRCPOSED NEW TYPES OF REACTORS WHICH WILL ASSIST HRR IN MAKING LICENSING DECISIONS IN REGARD TO DESIGN AND SITING.
[Q5EENTILPEM&PE1: THE AD(ANCED REACTORS BRANCH (NRR) TRANSMITTED COMMENTS ON RIL 828 TO THE FUEL BEHAVIOR kESEARCH BRANCN (RES) IN A MCMORANDUM FROM T. P. SPEIS TO W. v. JOHNSTON. DATED OCTOBER 13. 1978.
RESOLUTION OF THESE COMMENTS IS NOW IN PROCESS.
IN ADDITION, NRR PLANS TO WRITE A RESEARCH REQUEST TO RES 10 EXPAUD THIS PROGRAM TO EXAMINE MELT INTERACTIONS WITH SACRIFICIAL MATERIALS IN CONNEClION WITH THE LICENSING REVIEW TO MANUFACTORE 8 FLOATING NUCLEAR PL ANT 3.
OF OP5'S APPt.ICATICH ALTHOUCP THIS RESEARCH PROGRAM HAS GREATLY IMPROVED OUR UNDERSTANDING OF *HE PHYSICAL PHENOMENA ASSOCIATED WITH MELT / CONCRETE INTERACTIONS. THIS WORK SHOULD CONTINUE WITH THE OBJECTIVE CF E/PERIMENTAL VERIFICATION OF TCE CORE MELT / CONCRETE INTERACTICN COMPUTER MODEL (INTER). SUCH THAT EXTRAPOLATIch TO PROTOTYPICAL CONDITIONS CAN BE MADE TO ACCURATELY PREDICT SUCH VARIABLES 45 THE TIME OF CONTAINMENT MELT *THROUGH OR CVERPRESSURIZATION.
NRR HAS BEEN IN C105E CONTACT WITH THE RES STAFF ON RESEARCH RELATED TO MATERIALS INTERACTIONS BETWEEN CORE MELT DEBRIS AND CONCRETE AND MAINTAINS COGNIZANCE OF SUCH WORK IN THE U.S. AND OTHER COUNTRIFS.
l SD COMMENTS. G.
PIVENBfPK: NO RESPONSE RECEIVED.
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P&CGN M 06FICE COM 4Nf5 ON POIENTIAL UTILIZATION OR VALUE OF &ESErACH PE50LT5 IN THE REGULATORY PROCESS i
'PIL 5 29 DATE ISSUED: 06/07/73 RES DECTSION UNIT: FUEL SEHAVIOR 1
RIL TITLE:
"FUEL ROD ANALYSIS CCMPUTER CODE:
FRAP-T3"
$PCNS W NG OFFICEf5): RES PPS:
1-12 FUEL CODE PESEAPCM PPOJECT MGR:
H. SCOTT DEVELOPMENT PE U QMalNTJ: THIS MIL TRANSMITS THE RFSULTS OF CCMPLETED RESEARCH TO PREPARE AND TEST THE THIRD MODIFICATION OF THE C0hPUILR CCDE FRAP-T (FUEL ROD ANALYSIS PROGRAM - TRAN5IENT).
FRAP-T IS A FORTRArt IV CGMPUTER CODE BEING DEVELOPED 10 PREDICT THE TRANSIENT REPON5E OF A LWR FUEL RCD DURING POSTULATED ACCIDENTS SUCH AS LOSS-OF-COOLANT ACCIDENTS, POWER i
C00 lit 4G MISMAICH ACCIDEt4TS, REACTIVITY INITI ATED ACCIDENTS, OR INLET FLOW BLOCKAGE ACCIDENTS. / RAP-T IS ALSO BEING DEVELOPED TO PERFORM THE CALCULATION 5 NEEDED FOR PLANNING AND ANALYZING POWER BURST FACILITT AND LD55 0F FLUID TEST i
EXPERIMLNTS.
IN FRAP-T3, THE COUPLED EFFECTS OF MECHANICAL. THERMAL. INTERNAL CAS AND MATERIAL PROPERTY RESPONSE ON THE BEHAVIOR OF THE FUEL ROD ARE CONSIDERED. GIVEN APPRCPRIATE COOLANT CO:4DITION AND POWER HIST 02IES, FRAP-T3 CAN CALCUL ATE ROD BEi!AVIOR FOR A WIDE VARIE1Y DF CFF-NDRMAL SITUATIONS AND POSTULATED ACCIDENT CONDITIONS (E.G., SWR OR PWR PCME* TRANSIENTS, FLOW COASTDOWN. LOAD LOSS OR COOL ANT DEPRE55URIZATON).
l IN THE CONTEXT OF LLtR SYSTEM TRAN5IENTS FRAP 15 WELL SUITED TO TO BE USED AS A t9MPONENT CODE TO DESCRIBE FINE DETAILS
^
0F FUEL ROD BEHAVIOR. FURTHERM3RE, SENSITIVITY STUDIES WITH FRAP WILL FACILITATE DEFINITION OF THE SIMPLEST ACCEPTABLE
{
FUEL DESCRIPTION IN SYSTEMS CODES.
USER DISCUSSION POSITION CCMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELE45E RESULTS EP51.?l k &CTIVlTlfi PdVIEW HEL D CC'*PLETED HELD HELD ISSUED IMPLEMENTED OFFICL RL5PCNSICLE......... NRR/SD SCHEDULED COMPLETION DATE.
03/n/"
i;NSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL CCMPLETION DATE.....
lig3_COMarH L$_QN 0_/_}1/J3 d VnG{ WERE:
PtSCU B_!M@]AllG! T3_ff WLAJ3mTQCf31:
TRAN5IENT FUEL PERFORMANCE CODES ARE REVIEWED BY NRR AS PARI 0F LUCA 1.HD 01HTR ACCIDLNI ANALYSLS. NRA USES INDEPENDENT MRC-DEVELOPED CODES F0it AUDIT PURPOSES IN THESE REVIEWS.
T ES C P I g;flP[q T_O F._,,_P ESUL TS : rRAP-T3 IS ONE VERSION OF THE FRAP TRANSIENT FUEL PERFORMANCE CODE.
BECAUSE 1HIS 15 vfLY AN INILRIt1 VLRSION OF A BEST-ESTIITATE CODE, IT HAS NOT BEEN ADAPTED FOR LIC' NSING APPLICATIONS.
It4 THE FUTURE, NRR PLANS TO USE THE WAYER RLACTOR ANALYSIS PACKAGE (VRAP) WNICH WIll INCLUDE A MORE CURRENT J
VERSION OF FRAP-T THAT INCORPORATES CONSERVATIVE MODIFICATI0t:5. IN PRESENT AUDIT WDEA, NRR USES THE WATER REACTOR EVALUAfICN MODEL t uREr1) SERIEL OF COGES W1ICM DOES NOT INCLUDE FRAP-T3.
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_CO* *JN TJg !"W Q : THE NEED FOR A MORE DETAILLD TRAN5IEF.T FUEL BEHAVIOR CODE IN LICENSING APPLICATIONS HAS LutG Blir4 RECOGNIZED BY NRR.
IT IS EXPECTED THAT A MORE CURRENT VERSIGN OF THE FRAP-T CODE SERIES WILL BE ADAPTED FOR LICENSING APPLICATIONS. ALTHOUGH CONSERVATIVE MODIFICATIONS WILL BE REQUIRED, THIS CODE WILL CONTAIN MANY OF THE ELLMENT5 A0W IN FRAP-T3.
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5A_COMENILJ F IVfM4425 Pf.10El%5f t ED C A110N 10__EEfMLAlD3Y PROCE152 TRAM 51 EAT FUEL PERFORMANCE CODES ACE REVIEWED BY NRR AS PART CF LOCA *ND OTHER AC M LPi! ANALYSES. NRR USE5 INDEPEhDENT NRC-DEVELOFED CODES FOR AUDIT PU2 POSES IN THESE REVI Lt:5.
Pf 5C1J30_It1101._f f PE$U_L15 :
FRAP-T3 15 ONE VERSION OF THE FRAP TRANSIENT FUEL PERFORMANCE CODE.
BECAUSE 1HIS 15 0:41 Y Ato ItTiiRIM WERSION OF A EEST-ESTIMATE CODE. IT HAS BCE*f ADA?TED FOR LICD45ING APPLICATIONS.
IN THE FUTURE. HER PL ANS TO USE THE MATER REACTOR ANALYSIS PACKAGE (kTAPI MHICH WILL ItiCLUDE A MGRE CURRENT VERSI0d 0F FRAP-7 THAT INCCRPORATES CONSERVATIVE MODIFICATIONS. It! PRESENT AUDIT WORK. NRR USES THE MATER REACTOR EVALUATION f*0 DEL (L1t EM) SERIES OF CODES WHICH DOES NOT INCLUDE FRAP-T3.
00'JjlN DLPfM EM51 THE NEED FOR A MORE LEIAILED TRAt45IENT FUEL BEHAVIOR CODE IN LICENSING APPLICATIONS HAS Lot.G LLEJ RECOG'41 ZED BY N2R.
IT IS EXPECTED THAT A MORE CURRENT VERSION OF THE FRAP-T CODE SERIES WILL BE ADAPTED FCR LICENSING APPLICATIONS. ALTEDUOH CONSERVATIVE MODIFICATIONS WILL BE REQUIRED, THIS CODE WILL C;t4IAIN MANY CF THE ELEMENT 5 NCu Its FRAP-T3.
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PkOGWiM OfflCE CO*MLNTS ON POTENTIAL UfILIZATION CR VALUE OF PESEAPCH eESULTS IN'THE REGULATGRY PROCESS
~
RIL 3: 30 DA,E YS5U122 D6/28/73 EES DECISION UNITr SAFEGUARDS PIL TITLE-rHASE I f!NAL REPORT,
- BARRIER /ENETRATION DATA BASE":
OF STUDY, "ASSISTANCE-PHYSICAL PROTECTICM ASSESSMENT 5" l
SPChi9 PING M FICE(51 t;R R PPQ: NONE RESEARCs! PROJECT 5E:
J. MILLER i
P($,CQM"[NTS-THE REPORTED RESULTS PROVIDED2 (f) A CLASSIFICATION OF BARRIERS IN TERMS OF THE PENETRATION TIME FOR SELLCILU C3tXTEPMEA5tRES WMICH AN ADVERSARY MIGHT USE TO GVERCOME THE BARRIER, AND (2) PROCEDURES TO BE FOLLOWED IN I
TESTING REACT 0p. SITES FOR CCMPLIANCE FOR 10 CFR 73.55, THE REACTCR SAFEGUARDS REGULATION.
UIER DISCU5510M POSITION CCMMISSION ACR5 PRESS 4
OFFICE MEETIMG PAPER BRIEFING BRIEFING RELEASE RESULTS 1
f05_T_RI L AC U v LTJ E1 EE21L}(
HELD CWP: ETED NELD HELD 155UED IMPLEMENTED CfFICL RESPONSIttL......... NRR SCHECUltD CC".PLETICM DATE., 03/23/78 UNSCHED UNSCHED UN5CHED UNSCHED U'45CHED UNSCHED ACTUAL COMPLET10M DATE.....
F R_C T IUS Cl3_88'3D'7B d d lif.Rl D ES M11 LJ_f P1LC All 000_8 LM LAJ D3Lf? ACE 55 : A5 STATED IN THE RIL, THE BARRIER PENETRATION DATA BASE IS AVAILALLE A5 A SUPPLEMLNIAL DAIA 50J2CL f 0R USE IN THE EVALUATION OF LICENSEE FACILITY 5AFEGUARD PROGRAMS.
THE DATA FROVIDES A STANDARDIZED BASE OF BARRIER DELAY TIMES AGAINST VARICUS ADVERSARY CDUHTERMEAst;RES TROM OURRENTLY AVAIIASLE LI!EFAIURE. THIS DATA WILL PROVIDE FCR MORE RELIA 3LE AND CONSISTENT EVALUATION OF j
S AFEGUARD PROGRA*is BY COTH THE NRR STAFF AND LICEM5EE SiAFFS.
l TG! PIPC" IHL TIC [EACT OF P3tJLTS: THIS CCN7ILATION OF DATA IS USEFUL FOR THE DESIGN OF PHYSICAL SECURITY SYSTEMS M
h5ELS, A5 WELL A5 IN THE EVAL'JATION OF THE EFFECTIVENESS OF SUCH SYSTEMS BY NRR.
ALTHOUGH PHE RESULTS OF THIS PROSPld MERE NDT AVAILABLE DURING THE DESIGN PHASE OF THE CURRENT UPGRADING OF PHYSICAL SECURITY AT NUCLEAR PCMER PLsMis, THIS DATA BASE MILL PROVIDE USEFUL DATA FOR FUTURE APPLICATIONS.
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j P & G fdJ M O f F I C E C C'Z EN I 5 _CM PDfESTIAL UfftIZATION 0? VALUE OF RESEiQCH 8tE5uti$ IN THE REG'JL ATODY FPCCESS l
RIL 4 32 DATE ISSUES: C3/03/TS PES PECISION t'MIT_:
FAST BREEDER REACTCJts j
RIL TITLE:
IMPROVEMENTS IN THE AEROSOL SEHAVIOR CODE FOR RADIOLOGICAL ASSESSMENTS OF LMFBRS.
SPONSCol.% OFFICEf 5): RE5 P8ft:
2-7 AEROSCL MCDELING FESEs#CH PROJECT MANAGER:
J. LARKINS j
AND PROPERTIES j
RE5_Crja 3n:T5: THIS f"EMOPANDUM TRAM 5MITS THE RESULTS OF CCMPLETED RESEARCH PN THE MEASUREMENT OF SODIIM OXIDE AEROSOL j
PROPERIILS. SODIUM CXIDE 15 THE trey AEE050L CCNSTITUENT IN OSTULATED SEVERE LMTB't ACCIDENTS.
]
FOR THE MOST SEVERE POSTUL ATED Lf*FER ACCIDENT SCEf8 ara 05 (ffCDA f.ND CORE MELT). SODIUM-0XIDE AER050L i'EPRESENTS THE I
HIGHEST AIREORNE MASS CCNCENTP ATICNS IN THE CONTAItET 1rE55EL AND IS EXPECTED TO DOMIN ATE AND GOVERN THE BEHAVICR 0F THE FULL A'43 FISSION PRODUUT AEE050L.
THEREFORE. AS A FIRST STEP IN IMPROVING THE AEROSOL BEHAVIOR CCDC. HAARM-2.
SEPARATI EFFICTS WORK MAS CAERIED CUT ON SODIUM-0XIDE AEROSOL.
THE RESULTS OF ft:ESE SEPARATE EFFECTS MEASUREMENTS MAVE BEEN 1.4CJRPORATED TNTO THE MdDELS OF THE AEROSOL EEHAVIOR CODE i
HAARM-2. AND 10CETHLR DN SOME ADDITIONAL If*,PROVEMENTS USED TO GENERATE A NElJ VERSIGN CmLLED HAARM-3.
THE IMPROVED j
MODLLS IN HAARM-3 P;:Ov%,E A !"CRE REALISTIC DESCRIPTIO" 0F PARTICLE CHARACTERISTICS AND THEREBY ALLOW IMPROVED ESTIf1ATES OF SOLIUM-0XIDE AERO50L EE:fAVIOR DURING A POSTULATED HCDA.
THE HAARM CCDE IS USED BY NPR FOR LMFER SITE RADIOLOGICAL CONSEQUENCE ASSESSMENT.
i USER DISCUSSION POSITION COMMISSION
'ACRS PRESS j
OFFIrE MEETING PA*ER ERIEFI!IG ERIEFING RELEASF RESULTS Post __p1La C_T IVI TJn Pfylty f4EL D fcMPLETED HEL D HELD IS5UED If?L EMENT ED OtFICE kE5PCNSIBLE_........ NRR
~
~
SCHEDULED CCnftrTION DATE.. 01/15/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL 00M7tETICh CATE.....
IEP_C 0*~_*E NIS, _JJ 3 03 G/1,.52E15 :
DLSMlli_8 7L JLAIJ C'4_lgyr.g( A f 0Py,gegin$ r CDDE DESCRIBED IN TilIS RIL. HAARM3. 15 USED BY NRR IN CALCULATING IHE LiFLC15 0F FASI RE ACIC.t ACCIDEH15.
IT IS BASED ON A PREVIOUS CODE HAARiT2. NC1J f.0DIFIED TO REFLECT THE l'
RESULT 5 OF MILLIKAN CELL EXPERIMEtiTS AT BCL.
THE GENERf.L EFFECT OF THE MODIFICATIONS IS CONFICED IN LARGE
{
SCALE TESTS AT HEDL.
DMCRIM_I;T ACTXRESULTS CCNTIRMAT101 CF THE NUJ CODE PERMITS A REDUCTION BY A5 f%3CH AS A FACTCR OF 2-8
}
Its THE JUOUNI OF RADIDACTIVE AEROSCL5 LEAKED FROM AN LMFER CONTAINMENT IN THE EVENT OF A EARGE ACCIDENT.
(RADICACTIVE CASES ARE Mai CCN'IDERED TO BE AFFECTED.)
C 0i~'*L'fA tt i87/_RfMt?t5: THE ATTENJATICN CF RADICACTIVE AIRBORNE MATERIALS BY AGGLCMERATION AND FALLOUT IS AN 1
li",F o r:
iACICR IN THE CALCULATICM OF ACCIDENT COM$EOUENCES. EVALUATION AND VERIFICATION OF THIS NATURAL j
j MECHANI5tlIOR REJUCING RADICACTIVE EMI551GNS HAS SIGNIFICANCE CCMPARAELE TO AN ENGINEERED SAFETY FEATURE.
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P4094'1 off f CE C9mENTS 0*4 P0f E*fl AL _Uf1L12 ATIcN OP VALUE OF RESEAPfH 3PE50LT5 IN THE REGUL4I0RY Pl0CE55 P.IL s*
34 D4TE 755UQ: C8/03/78 PES PECTSION 11MIT2 FUEL CYCLE SAFETY A ENVIRONMENTAL EFFECTS RIL TITLE: KUClEAR DECAY DATA FOR RADIONUCLIDES OCCURRING IN ROUTINE RELEASES FROM NUCLEAR FUEL CYCLE FACILITIES.
SPO450EING OFFICE (5)s NRR B:
5-24 RADI0 BIOLOGY AND D35IMETRY PESE4RCH PROJECT P!GR:
J. FOULEE l
PE5_llt.E RELLASES OF EFFLUENTS FROM NUCLEAR FUEL CYCLE FACILITIESTHIS IS A TABULATION OF NU COTE'81):
ROU THIS CAN BE USED BY NRR A5 A BASIS FOR ESTIMATICN OF RADIATICN EXPO 5URE TO MAN.
USER DISCUSSION POSITION COMMI55IDM ACES PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULT 5 EQ5_T P.11 f tTIVJilE)
PEvllM l'fLD CnMPLETED HLLD HELD ISSUED MPL EriEN1 EQ CFFlfE RC PONSIBLL......... AfLS SCHEDUttD COMPLETION DATE.. 01/30/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE..... H/A D FENIL__E _U YELLUE3:
?ltL E L1rAITc_1 1oJ2 cut _ATCPY PP1QE352 DOCUMENT PRESENTS BASIC NUCLEAR EECAY DATA FOR DEVELOPMENT OF
".A105 0F EADIOLOGICAL DOSE.
hkR HAS BEEN EMPLOYING THESE DATA AN7 IN FUTURE UPDATING OF THE DOSE 155*:ENT f*ETHODOLOGY WILL USE THIS DOCUTIENT AS THE PRIMARY REFERENCE.
,1C311 U M?AC'TL5V.EnIS.9F RE"MLI1: THE DOCU:1ENT PROVIDES THE BASIC NUCLEAR DECAY DATA IN THE FORM NECESSARY FOR
.JICLOGiCAL A5 THE DOCUMENT PROVIDES THE STAFF WITH A REFERENCE DOCUMENT TO ENSURE STAFF'S I
dIFORM USAGE OF DATA BASIC TO ITS EFFORTS.
1 5.0*"iE N T 5/P EM A p r 5 r A MACHINE READABLE FILE IS AVAILABLE FROM K.
ECKERMAN, NRR/D5E/RAB.
l 1 _ -. - - - _, - -
P?Off<rti M FICE Cc".1L= f 5 fM P0 f En fla t o fILIZ 4 f f oM OR V AtuE Of trESEARCH PE50L f 5 IN THE REGut4TO U PROCESS RIL 48 35 IATE ISSUEDt 29/15/78 RES DECTSION UNITS FUEL CYCLE SAFETY & ENVIRCNMENTAL EFFECTS RTL TITLE: 5FACTORr A COMPUTER CDDE FOR CALCULATING CO5E EQUIVALENT TO A TARGET ORGAN PER MICR0 CURIE - DAY RESIOENCE OF A RADIONUCLIDE IN A SOURCE CRGAN SPO'450 PING CFFICE(5)
- NRR Pf?S: 5-24 RADI0 BIOLOGY &
PESESPCM PPO?ECT M';R*
J. FCULKE j
DOSIMETRY i
P E(C0'1*EM15 s THE 5 FACTOR CCMPUTER CODE CALCULATFS 5. THE AVERAGE DOSE EQUIVALENT TO EACH OF A SPECIFIED LIST OF TAR.El CkGANS PER MICR0 CURIE-DAY RESIDENCE CF A RADICNUCLIDE IN SPECIFIED 50LRCE ORGANS.
THE SFACTCR CODE COMPUTES CCMPONENTS OF THE DCSE EOUIVALENT IRO*1 ALPHA PARTICLES, ELECTRONS, GAMMA RAYS, FI5SION FRAGMENTS, AND NEUTRONS.
5-FACTORS CAN BE COMPUTED FCR ANY RADICNUCLIDE FCR WHICH DECAY DATA ARE AVAILABLE.
I t
USER DISCUSSION POSITION CDMMISSION ACRS PRESS OFFICE MEETIN3 PAPER BRIEFING ERIEFING RELEASE RESULTS E0if RJL[5Pb41 LE......... t.RREfVIEW HEL D CCTLETED
}{RD NFLD 155UE9 IM'L EME'4 T ED AfflVJIlfi CFFTEL V l
SCHEDULED COMPLETION DATE.. --
UNSCHEO UNSCHED UNSCHED Ud5CHED UNSCHED UNSCHED ACTUAL CGMPL ETIOr4 DATE.....
i N2P,CC S E E f MEr$f U d 179,_E. ECFEPM*N:
, d5080 fj>P(RAlJfp 10 91M(4]pjY_ P[Qfdi$s DOCUMENT FRESENTS COMPUTER CODE FOR ESTIMATING rJ5E EQUIVALENT TC l'
St210US AkGE3 ORGAhTP[R i:1C&OCURIL-DAY RESIDENCE OF A NUCLIDE IN A SOURCE CR",AN.
i Pf1C33 frEEP.A{J_f_)F P E$1f L T5: Tile 5 FACTOR CODE RESULT 5 CAN BE USED TC ESTIMATE CRCAN DCSES GIVEN THE CRCAN EURDEN J
OF A RADIbhUCLIDE.
IN ADDITICN, IF THE 5 FACTOR RESULT 5 ARE ETLOYED WITH METABOLIC INFORMATION THE DOSE PER 4
UNIT ACTIVITY INGESTED CR IMHALED CAN BE ESTIMATED.
COTE *!Tif9 E'11P55 s A LISTING OF THE CODE (REVISED T3 INCLUDE ALPHA D35E TO ENDOSEEAL CELLS AND BCNE HARRCW)
J IS AWAILABLE FRCM K.
ECKEliMAN. NRR/DSE/RAB.
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i d ? - Q fy 7.7 X b,rq F @ T_ip y rg IZAricer ca v t. L U L OF RE5 LARCH k !Sk'LT S I ?' I H E ple.';L a f C R Y PPCCESS EI L_t :
M r
- T F ! ' 'W_' ') : 09/27/73 Pts PrllSION t' NIT: rAST BREEDER REACTORS TIL T[1[1: EVALUATION CF GL:iE RJ L ATCMIC CCCES:
CXIDE-3. SCRS. TAP, AND RECA spr w pTNr. rr gep ',):
NRR p;r: 2-12 GAS COOLED REACTOR
- ESEARCH PPOJECT MCR:
J. LARKIN5 y
P r e _ c f& ty. T ', : THIS RE',EAFCH EV4tUATED THE APPLICA:ILITY AND UTILITY OF THE CODE 5 FOR THE ARSR/CAS COOLED O_ i '. i U M itiL1Y PkOC"AM.
THE 05JtCTIVEL OF THE EVILUATION5 AtS3 INCLUDED AN ASSESSMENT OF THE MODELS AND NUMERICS USED IN ILE CCDI S AND 10 t:0YE LNDER La4T CC*.LITICNS CR FCR WHA! SCENARIOS THE CCDLS WERE USEFUL.
THE APPLICABILITY
'.9 UTILITY OF THE GtC C00E5 F02 ESP-HIGR SAFETY IPO5 RAMS WAS FCUND TO SE VEPV P1NITED.
THE OXICE-3 CODE APPEARS TO w49LE CITCA!ICN Cf C?frHI!E SY M351T URE APPRLPRI A T EL Y UNDER NORNAL C P E R A T I r.t,.,noiTIONS.
THE COLE MILL RECUTRF FURTHER f ut PIM::4TAL VLRIr CATICN ! F F C i THE LIMITS CF ACCURACY CAN LE ESTABLISHED. THE SCR5 CODE IS THE FOPM PPESENTED FCR CUR L V A L U P. T I G N APPLA75 To b".VE IC"E SERICUS CEFICIENCIE3 IN THE MODELS kHICH PLACES DOUBIS CN THE ANALYSIS PERFORMED WITH THE CODE.
1ME TAP AMD FEC-CCIE5 GIVE COCD ACRELNLNT WITH OTHER ANALYTICAL TCOLS AND APPEAR TO EE USEFUL AND APPROPRIATE IOR THr47 PL5 f t::ED APPL IC A T IONS.
CUINTIfTCATICN OF THE ACCURACY OF THE CODES WILL REQUIRE FURTHER COMPARISCN WITH O P L E A f ! N r. - SCTOR CGNDITICN5.
A VENDOR VERIFICATION PROGRAT1 FOR THE CCDES I L R E C C'".'"E N D E D.
USFR DISCUSSION POSITION CCM. MISSION ACR5 FRE55
.CITICE
!".'. L T I N G PAPER FRIEFING LRIEFING RELEASE RESULTS P r, S T _ p ! ( _ t c 1 T V I f 1 F '.
y r gily gr3 p cometETg urin gun legg_
I r -' L EP"F N T E D useICL Etsec c1Ltt.......
m LCHEDUttD CC:"PILIION DATE.
UNSCHED UNSCHED UNSCHED UNSCHED UN*CHED UNSCHED AClutt CCMrtEIIcN Daft.
'f p p MN T 'o_pj, T g('d u v5 -
N ' e r J r r, r P LI 0 t T T r *; Yp c r r.U( A T r a y P e r r e4 : THE RIL LESCRIBES k'ORK PERFORMED BY ENL TO EVALUATE 50RS. A CODE L L'!it td...i!
t, ( Kad AL AIDMIC FLR IHL INAtYSIS OF FISSIO:4 PRODUCT RELET.5E I E0*1 HTGR CCRES UNDER TPANSIENT CUEllIONS INiOLVIG CTRL HEATU"5.
A TCPICAL REPORT GA-A12402. WAS SUI'MITTLD TO NER FOR REVIEM WITH THE INTENTION THAT, ATTER ELING ACCEPTED. THE CODE COULD EE USED AND REFERENCED FOR SAFETY ANALYSES RELATED TO LICT M ih0 Of CCMJ" P CI A L HTGRS.
pFveter JrPact or Pre /)(TS: CONFIrmtTCRY.
NRR'S REVIEM OF ThE 50RS FEPOPT AND THE MA%IMUM HYPOTHETICtt FISSION i k O L t,C f PttL C,L mr.6 f R J MODELS IND DAT A HAD PROCEEDED FAR [NOUGH AT THE TIME OF CURTAILMENT OF HTGR LICENSING ACTIVITY (1976) TO FL MIT ' CMC CONCLUSIONS REGaRDING THE UTILITY OF THE CODE.
THESE CONCLUSIONS MERE PKf5ENTLD IN THE 9E#CTCR FUEL 5 IMPUT TO THE CASSAR ISER TDECEM3CR 15. 1976).
THE STATEMENTS IN RIL 835 SORS ARL IN GENLPAL ACFLEMENT WITH. AND TH95 CONFIRMATORf 0F, NRR*5 A55E55 MENT IN THE GA55AR ISER.
C N ~! N T S /f ' *" P r '. : AS It;DICATED IN THE GALSAR ISER. FOR LICENSIta PUPPG5ES NRR ADVOCATES THE ASSUNPTION CF INSIANitALOUS E LEASE OF FI55!0N PROD:'C TS FROM FAIL ED PARTICLES TO THE PRIMARY CCOL ANT.
IN EFFECT. THIS A550!".PTICM EINDERS MOST THE SLR5G MDDL15. EECAUSE 50R$G MODEL' TIME-DEPENDENT FISSICH PRODUCT RELEASE AND TRAT, PORT.
THIS A'5UiPTICN WAS JUSTIFIED IN THE GASSAR ISLR PRIMARILY 0:4 THE BASIS OF A POCR DATA BASE FCR FIL51CN PRCDUCT RELEALE AND TRP.?;;PCRT IN HTOR FUEL MATERIALS.
THE ':0RK REPORTED IN RIL 836 PROVIDES ADDITIONAL SUPPGRT FCV ItiE VALIDITY OF THESE LICENSING ASSUM?TIONS.
IN VIFW OF THE LACK CF HTCR LICEN5ING ACTIVITY HOWEVER, MR ENVISICNS N0 IURTHER NEAR-TERM UTILIZATION GF THE SORS INFORMATION IN TFE REGULATORY PROCESS.
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P RO GD P1_ fMIC E C M 0 _.,5 C M P9f t**f f AL UTILIZaiteg n4 yatuE OF FESEARCH RESUl15 IM THE REGULAIF _
N.E$$
t 7
)
j WIL er 43 DR;E 155UID.
12/1E/78 FES DECISION T,11: FAST EEFEDER REACT 05.5 1
l PI?_ TIT 11 THE CC*PUTER CODE BRENDA - A CCP.PUTER PROGRAM FOR THE DYNAMIC SIMULATION OF A LIQUID j
METAL FAST EREELER REACTCR PLANT i.PQ'850eIM CUICE(5): WER EFQ:
2-13 FAST REACTCR RESEAP.CM PPDJECT PCRt P. WOOD SYSTEMS CODE AND ACCIDENT Ar.ALYSIS E5_ @be *15:
lW ExtNDA 15 A FAST RUNMIN5 SYSTEM CSDE INTENDED TO PRDVIDE NRC WITH THE CAPABILITY OF DOING 1
GUICK IILTRIC SURVEY 5 AND SCCPING STUDIES OF MORMAL CPERATING A5 WELL A5 ACCIDENT AL TRA 51ENTS IN LMFER PLANTS.
THE CODES PROVIDE FPC WITH AN INDEPENDENTLY DERIVED T00L FDR SAFEff ASSESSMENT. IT 15 CURRENTLY BEING MODIFIED 10 MAKE IT APPLICABLE TO PRECPERATIONAL AND STARTUP TESTS IN FFTF.
l 1
USER DISCUSSIGN POSITION CC"MISSION A t.P5 PRESS I'
CFFICE MEETIM5 PAPER ERIEFING ERIEFING RELEASE RESULT 5 E05I P,1LJCTIY3JJf1 P.Ey11B "ELD COMPLETED PE10 HEL9 ISSUED _
IMPLEMENTED 1
Of f !TL RLSPONSI BL E......... Het
]'
SCHEDULFD COMPLETICN DATE..
UM!CHED UM5CHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL CCPJLETICM DATE.....
l j
tLTR.l W L9_11,_W') VIM T d 979 J
% JYl'8 DI.hRl fRr CLIc A3.> r LuPRaicurocess: SInCE THE.e 15 No termR tICt=5ING ACTIVITY AT THE rREStar TIME. THERE r5 i
NO IMLDILIE tJ PL IC ON Of IH15 CCl*PUILR CODE TO THE etir,JLATORY PROCESS.
HCk'EYL R.
.F THE LMrER LICEICING ACAIN BECCMES l
j ACTIVE, THIS P80 GRAM AAS POTINTIAL APPLICATION IF THE CL AIMS REGARDING ITS UTILITY awe. IN FACT. REAL72ED.
{
Df.ZEIPL I'*E.A_Cl_R8tf3 ULT 3 THIS RIL DNLY DESCRIBE 5 THE CCr".PUTER CODE ITSELF (TOGETHER WITH A CCFJARISCM BETWEEN BRENDA i
j AND SSC) AWD THUS RieE ARE NO
THE IMPACT OF THIS CODE WILL CNLY BE j
j CONTRIBUTCRY IF2 A) IT TUEMS CUT TD BE FAST-RUMMINC FCR Ir70RTi.;T APPLICATI0M5; B) IT ADEQUATELY TRACKS SYSTEM BEHAVIOR AS MEASURED AGAINST A MORE DETAILED. MECHANISTIC SYSTEMS CODE 5 TU AS 55C; AND C) TPE DETAILED 575fEMS CcDE CM WHICH i
}
If IS BASED HAS BEEM PRGPERIY VERIFIED.
j CO* MENT 5/EEM7FF58 MCME i
L i
}
1
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l l
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I
~
DROGRAH_pFFICE RD mENTS ON POTENTAAL HTILIZATION G3 VALUE OF PESEARCH REgULTS JN THE REGULATORV.PROCESb__
PIL 3: 41 DATE ISSUED:
12/19/78 PES DECISION UNITr SEISMIC, ENGINEERING Z SITE SAFETY RTL TITL E: LABORATORY TESTING PROCEDURES TO DETERMINE THE CYCLIC STRENGTH OF SDILS SPONSORING OFFICE (S): SD EES:
3-1 NRC/ STATE REGIONAL RESEARCH FROJECT MGR:
N. STEUER i
EARTH SCIENCES I
WIDELY VARYING RESULTS WERE BEING OBTAINED IN TESTING SDILS FOR LIQUEFACTION POTENTIAL RES COMMENy :GTH OF SOILS)
(CYCLIC sinEN BECAUSE STANDARD TEST PROCEDURES DID XOT EXIST.
HENCE, THE SUBJECT STUDY MAS CONCUCTED IN RESPGESE TO THIS FACT, AND BECAUSE STANDARD PRCCEDURES ARE NEEDED FOR NUCLEAR P0"ER PLANT SITE INVESTIGATIONS. THE RESEARCH PROGRAM WAS CONDUCTED IN CONJUNCTION WITH DEVELOPMENT OF ASTM PERFORMANCE SPECIFICATIONS.
FAJOR SLIL MECHANICS LABORATORIES THROUGHOUT THE WORLD WERE CCHTACTED TO DETERMINE THEIR TESTING METHODS AND TO EVALUATE THE CYCLIC TRIAXIAL EQUIPMEN! IN USE.
THIS INFORMATION PROVIDED A BASIS TO DEVELOP ThE TEST PROCEDURES FOR STRESS-CONTROLLED CYCLIC TRIAXIAL STRENGTH TESTS PRESENTED AS A PERFORMANCE SPECIFICATION S0 THAT A GEOTECHNICAL TESTIl4G LABCRATORY CAN: 1) ENSURE THAT ITS TEST EQUIPMENT AND PROCEDURES MEET 2EQUIRED STANDARDS, AND E) CHECK THAT RESULTS AGREE REASONABLY WITH RESULTS OBTAINED BY OTHER LABORATORIES.
RESE.RCN RESdlTS PRESENT RECCMMENDED PROCEDURES FOR LIQUEFACTION POTENTIAL TESTING BY USE OF STATE-OF-THE-ART j
TESTING TFCHNIQUES As A PERFORMANCE SPECIFICATION.
I IT IS RECOMMFNDED THAT NUREG-0031 CE USED AS GUIDELINES BY THE OFFICF OF STANDARDS DEVELOPMENT IN DEVELOPMENT OF APPLICABLE
.."tATIONS AND REGULATORY GUIDES, BY THE OFFICE OF NUCLEAR REACTOR REGULATION TO ASSIST IN THE REVIEW OF NUCLEAR OL _k FL ANT 0?ERATING APPLICATIONS. AND BY THE APPLICANT AS A ST ANDARD IN THE ASSESSMENT OF LIQUEFACTION POTENTIAL OF <Csd7ATION S0lLS AT NUCLEAR POWER PLANT SITES.
i USER DISCUSSION POSITION COMMISSIDH ACRS PRESS 0FFICE MEETING PAPER BRIEFING BRIEFING-RELEASE RESULTS EOSl_RIL AITlyITIES REVIEW PELD COMPLETED HE8.D HELD ISSUED IMPLEMENTED OFFICE RE5?0NSISLE......... SD SCHEDULED C0ttPLETION DATE..
UNSCHED
?JNSCHED UNSCRED UNSCHED UNSCHED UNSCHED ACTUAL CCMPLETION DATE.....
SD CCENENIS._RmjDLMQQRE - THE INFORMATIEN IN THIS REPORT WAS USED AND REFERENCED IN REGULATORY GUIDE 1.138, i
"LABORATORY INVESTIGATIONS OF SOILS FOR ENGINEEP.ING ANALYSIS AHD DESIGN OF NUCLEAR POWER PLANTS".
IT WAS ALSO USED IN THE DEVELCPMENT OF A DRAFT REGULATORY GUIDE 01..
SROCEDURES AND CRITERIA FOR ASSESSIhd SOIL LIQUEFACTION POTENTIAL AT NUCLEAR FACILITY SITES," WHICH WItu BE PUBLISHED FOR COMMENT IN THE HEAR CCPIES OF T'E REPCRT WERE UIDELY DISTRIBUTED TO MEMBERS OF THE ASIM, D-18 COMMITTEE ON SOIL A
i FUTURE.
AND ROCK ENGINEERING AND WILL BE USED AS ONE OF THE BASES FOR DEVELOPING A NATIONALLY ACCEPTABLE
)
DYNAMIC TRIAXIAL TESTING STANDARD FCR SOILS.
l
! l l
l
t I
l i
PROGRatt OFFICE COMMENTS ON POTENTI AL UTILIZATION OR VALUF OF RESEARCH RESULTS IN THE REGULATORY PRQ, CESS
]
RIL 4: 42 DATE ISSUED:
12/20/73 RES DECISION UNIT: FAST BREEDER REACTORS
}
EIL TITLE: CRITICAL EXPERIMENT PROGRAM FOR NEUTRONICS CODE VERIFICATION f
SPONSORING OFFICECS): NRR REG:
2-13 FAST REACTOR RESEARCh PROJECT MGR:
P. WOOD SYSTEMS CODE AND l.
ACCIDENT ANALYSIS I
)
EES CCMMENIS: THIS PROGRAM OF DISTORTED GEOMETRY CRITICAL EXPERIMENTS WAS PLANNED AND CARRIED OUT TO PROVIDE
]
EEt&C% NARIS FOR THE VALIDATION OF THE MEUTRONICS PART OF CODES USED IN SAFETY ANALYSIS SUCH AS SIMMER. A SECOND CBJECTIVE IS TO VALIDATE THE VIM MONTE CARLO CODE FOR USE AS A SECONDARY STANDARD FOR VALIDATION OF OTHER NEUTRONICS METHODS.
q MELTDOWN CONFIGURATIchs IN LMFBRS CAN BE EXPECTED TO HAVE REGIONS WITH HIGH FUEL CONCENTRATIONS GIVING EXTREME SPECTR CHANGES AND LARGE REGIONS OF VOID CIVING RISE ID LARGE STREAMING PATHS.
NEUIRONICS METHODS OTHER THAN MONTE CARLO HAVE DIFFICULTY IN CALCULATING THESE CONFIGURATIONS ACCURATELY. A SERIES OF EXPERIMENTS WAS NEEDED TO DETERMINE THE j
IMPGRTANCE OF THESE DIFFICULTIES AND TO PROVIDE A BASIS FOR IMPROVING THE ACCURACY AND RELIABILITY OF ACCIDENT ANALYSIS i
METHODS. THIS PROGRAM HAS DEMONSTRATED THAT DIFFUSION THEORY NEUTRONICS CALCULATIONS WOULD UNDERPREDICT RAMP RATES WHICH MIGHT OCCUR IN A MELTDOWN AND COULD BE NON-CONSERVATIVE.
THE RESULTS OF THE PROGRAM LEAD TO THE CONCLUSION THAT DIFFUSION THEORY CALCULATIONS LEAD TO NONCONSERVATIVE ESTIMATION OF REACTIVITY GOIL3 FROM THE REFERENCE TO THE SLUMPED CCNFIGURATIONS. THE SIGNIFICANCE OF THE RAMP RATE AT PRCMPT CRITICAL IN AN HCCA CALCULATIONS HAS BEEN POINTED OUT BY THE NRR STAFF IN NUREG-0122(1).
t j
DATA REDUCTION AND PREPARATION OF THE FINAL REPORT WILL BE COMPLETED IN FY 1979.
A VIM MONTE CARLO CALCULATION ON CONFIGURATICN 2 - SODIUM VOIDED TEST ZONE WILL BE MADE AND THE SN CASES NOT SHOWN ON FIGURE 2 0F ZPR-TM-327 WILL BE COMPLETED.
ANALYSIS TO RESOLVE CROSS SECTION DIFFICULTIES THAT CAUSE DIFFERENCES BETWEEN EXPERIMENT AND VIM MONTE CARLO EIGENVALUES IS NEEDED. WHEN AND IF THE CROSS SECTION DIFFICULTIES ARE RESOLVED. IT WOULD BE DESIRABLE TO PREPARE i
SECONDARY BENCHMARK VIM MONTE CARLO CALCULATIONS USING A HOMOGENIZED MODEL OF iHE EXPERIMEMTAL CONFIGURATION.
)
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BPIEFING RELEASE RESULTS j
POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED j
OFFICE RESPONSILL E......... NRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
~MRR COM3ENIL _N_DVftBER 9.
1979 J. M EYF.P :
DESCRIBE APPLICAJJON TO R E_Q2L A TORY PROCESS: BECAUSE THERE IS NO LMFSR BEING LICENSED. THERE IS NO IMMEDIATE APPLICATION To THE REGULATORY PROCESS. HOWEVER. SINCE CORE DISRUPTION ACCIDENT ANALYSIS WILL PLAY AN IMPORTANT PART IN FUTURE LMFBR LICENSING. THERE IS A REGULATORY HEED FOR RES EXPERIMENTAL PROGRAMS DIRECTED TO ACCIDENT ANALYSIS-CODE VERIFICATION. THIS PROGRAM IS ONLY A VERY SM4ll PART OF THE TOTAL CODE VERIFICATION PROGRAM AND ITS UTILITY AND APPLICATI3N WILL CNLY BE TEAIIZED WHEN *ARGER (AND MORE IMPORTANT) PARTS OF THE VERIFICATION PROGRAM (E.G., SIMMER THERMA'. HYDRAULICS) -RE COMPLETED.
1
). _ _
N'R VIEMS THE IMPACT OF THESE CESULTS TO BE SIGNIFICANT; BUT OF SECONDARY OVERALL DES _CRIBE IMP ACT OF CEst!LT_S:
IMPCRTANCE. WHEN JUDGED RELATIVE TO THE IMPACT OF UNKNOMNS AND UNCERTAINTIES WHICH STILL EXIST IN THE THERMAL HYER AULICS AriALYSIS OF CCRE DISRUPTICN. A "SIMMER" TYPE ANALYSIS, WHICH WOULD TAKE INTO ACCOUNT THE NEUTRONIC-CALCULATION UNCERTAINTIES WHICH HAVE SURFACED AS A RESULT OF THIS STUDY. WOULD BE HELPFUL IN UNDERSTANDING THE OVERALL IMPACT. SIMILAR ASSESSMENTS USING SAS AND VCNUS WOULD ALSO BE HELPFUL.
CGT ENTS/ REMARKS: NCHE -
-=
~ _ -, _
l a
PCCGRAM OFFICE COMMENTS O'4 P9TENTIAL OTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REG.UL a
l RIL 8: 43 DATE ISSUFD: 01/10/79 RES DECISION. UNIT FAST BREEDER REACTORS 4
RIL TITLE:
SUPER SYSTEM CUDE A COMPUTER PROGRAM FOR DYNAMIC SIMULATION OF LMFBR POWER PLANTS SPONSCRING OFFICE (S): NRR RP&:
2-13 FAST REACTOR RESEARCH PROJECT MGR:
P. WOOD SYSTEMS CODE AND ACCIDENT ANALYSIS a
J RES CQI?JfN_TS: THE SUPER SYSTEMS CODE (SSC-L) IS SPECIFICALLY DIRECTED TO THE ANALYSIS OF THE ADEQ 4
0F NATURAL CIRCULATION IN SGDIUM-COOLED REACTORS TO PREVENT CLAD MELTING.
THE CCDE ALSO HAS THE CAPABIL!TY TO ANALYZE NORMAL OPERATIN', TRANSIENTS AND LESS SEVERE ACCIDENTS THAT DO NOT BREACH THE INTEGRITY OF THE SYSTEM DR FUEL.
THE CODE IS OPERATIONAL ON THE SNL-CDC-7600 COMPUTER AND CAN BE OPERATED THROUGH THE NRC i
j THE PHILLIPS AND WILLSTE BUILDINGS.
l 7
USER DISCUSSION POSITYON COMMISSION ACRS PRESS OFFICE HEETING PAPER BRICFING BRIEFING RELEASE RESULTS P_QST RIL ACT ULU.LFJ EEVIEW HELD COMPL~TED HELD HELD i
0FFICE RESPONSIBLE......... mR
)
SCHEDULED COMPLETION DATE..
UNSCHED
-- ~_
ISSUED _
ACTUAL COMPLETION DATE.....
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ELCMENTS W-RJ1SSELL -
l QUCME_FPL IC A T ION TO REGULATORY PROCE15: THERE IS NG IMMEDIATE APPLICATION OF SSC TO THERE IS NO LMF3R LICENSING ACilVITY)
CIRCULATION TESTS.
EXCEPT FOR THE SSC VERIFICATION ACTIVITIES ASSOCIATED WITH THE FFTF NATURAL i
THERMAL-HYDRAULICS SYSTfMS CODE IN PLACE.IF LMF3R LICENSING AGAIN BECOMES ACTIVE, IT WILL BE IMPORTANT TO HAVE AN IdDEPE
]
DEEEDE. IMPACT OF RESULTS:
PRESENT Ir1 PACT IS MINIMAL BECAUSE OF LACK OF LMFBR LICENSING ACTIVITY.
i VERIFICAIICH ACTIVITIES ASSOCIATED WITH FFTF NATURAL CIRCULATION TESTS MAY BE IMPORTANT DE THE IMPACT OF THE i
s_0121ENTS/REMAE:
AT THE PRESENT TIME IANUS AND DEMO ARE THE TWO CODES USED FOR THE THEP, MAL HYDRAULIC ANALYSIS 0F FFTF AND CRBR.
BOTH CODES HAVE BEEN DEVELOPED BY WESTINGHOUSE. DEVELOPMENT OF SSC-L IS A SPECIFIC C RESEARCH EFFORT WITH APPLICATION TO FFTF.
THE (P) AND (S) j THE DEVELOPMENT OF THE (L) VERSION.
VERSIONS OF SSC HAVE FUTURE APPLICATION FOLLOWING I
i 3
1 h
4 J
j,
PROGPA9 CEFICE COMMENTS ON PGTENTIAL UTILIZATION OR VAtVE OF RESEARCH RESULTS IN THE NEGULATORY PROCESS FIL 3: 44 DATE ISSUED: 01/04/79 RES PECISION UNIT: FUEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS PIL TITLE: RmDIATION DOSE TO CUNSTRUCTION WORKERS AT OPERATING HUCLEAR PCWER PLANT SITES SPONSORING OFFICE (S): NRR RRG:
S-23 OCCUPATIONAL RESEARCH PROJECT MGR:
J. FOULKE EXPOSURE AND PROTECTION RES CQ35ENIS: TNE STUDY PROVIDES A DATA BASE WHICH ALLOWS A REALISTIC ASSESSMENT OF THE RADIOLOGICAL Ir1 PACT CN CONSTRUCTION WORKERS OF PROPOSED MULTI-UNIT NUCLEAR POWER PLANTS. MEASUREMENTS OF PERSONNEL EXPOSURE OF SEVERAL HUNORED CONSTRUCTION WORKERS WERE CCNDUCTED AT EACH OF THE FOUR SITES WHrRE NEW FOR MOST WORKER GROUPS. THE FACILITIES WERE UNDER CONSTRUCTIDH 6 EXT TO OPERATING NUCLEAR POWER PLANTS.
AVERAGE DOSE ECUIVALENT RATES WERE MUCH LESS THAN to MREM / MONTH GREATER THAN OFF-SITE CONTROLS.
CORRELATIONS BETHEEN CPERATING PLANT POWER LEVELS AND DOSIMETER READINGS WERE GENERALLY POOR INDICATING THAT VARIATIONS IN OTHER RADIATION SOURCES HAD GREATER EFFECT ON THE MEASUREMENTS.
USER DISCUSSION POSITIJN COMMISSION ACRS PRESS OFFICE M:ETING PAPER BRIEFING BRIEFING RELEAJE RESULTS ERST 'It ACTIVIIIEi ELV11H HELD COMPLETED EEt D HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE.........
NRR SCHEDULED CCMPLETION DATE..
ACTUAL CCMPLETION DATE.....
.NRLCCEIN_TL_L__L HiNSHM -
AS PART OF ITS LICENSING REVIEW PROCESS, THE NRC MUST ASSESS E151R RE_ Af P L IC A T I ON_10_EfGML A_TERY PRQCf$$ 2 THE ENVIRONMENTAL IMPACT OF THE CONSTRUCTION AND OPERATION OF NUCLEAR POWER PLANTS. ASSESSMENT OF RADIATICN DOSES TO CONSTRUCTICS WORKERS AT OPERATING NUCLEAR POWER PLANT SITES IS PART OF THIS REVIEW. SINCE A FIRM DATA SASE FOR RADIATION DOSES TO CONSTRUCTICN WORKERS WAS NOT AVAILABLE, ENVIRONMENTAL IMPACT STATEMENTS COULD ONLY PROVIDE ROUGH ESTIMATES FOR THE DOSES TO CONSTRUCTION WORKERS. THE RESULTS OF THIS STUDY PROVIDE RADI ATION EXPOSURE DATA SUCH THAT A REASONABLE ENVIRONMENT AL IMPACT STATEMENT CAN BE MADE.
UNTIL THE PUBLICATION OF THIS REPORT. THERE HAS BEEN LITTLE DATA WHICH THE DfiqRIEE IMPACT OF RESQLTS:
STAFF COULD USE IN ASSESSING THE APPLICANT'S ESTIMATE OF CONSTRUCTION WORKER DOSES DURING NUCLEAR POWER PLANT CONSTPUCTION. THE RESULTS OF THIS STUDY SHOW THAT MOST CONSIRUCTION WORKERS AT GPERATING NUCLEAR POWER PLANTS RECEIV2D LESS THAN 10 MILLIREEMS PER MONTH; AND HO WORKER'S ESTIMATED DOSE EXCEEDED 500 MILLIREMS PER YEAR.
THESE RESULTS INDICATE THAT CONSTRUCTICH WORKERS AT OPERATIhG REACTOR SITES ARE NOT LIKELY TO RECEIVE SIGNIFICANT RADIATION DOSES, AND THEREFORE THAT INDIVIDUAL MONITORING WILL NOT BE REQUIRED. HOWEVER, IT MAY BE APPROPRIATE TO PLACE DOSIMETERS AT POINTS WHERE THE HIGHEST EXPOSURE RATES ARE TO BE EXPECTED,
'O ASSURE THAT UNUSUALLY HIGH EXPOSURE RATES DO NOT GO UNDETECTED. E
i 1
i 1
]
PRCCRAM OFFICE CCMMENTS ON POTENTIAL UilL[ZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS 5
]
RIL 4: 45 DATE ISSUED: 02/11/79 RES DECISION UNIT: FUEL CYCLE SAFETY 1 ENVIRONMENTAL EFFECTS PTL TITLE: THE CONCEPT COMPUTER CODE AND CAPITAL COSTS FOR BOILING WATER REACTOR PLANTS l
_SPOMSORING OFFICE (S): HRR EES: 5-21 SOCIO-ECONOMIC RESEARCh PROJECT MGR:
D. BARNA~
It1 PACTS 1
{
EES COMMENTS: THis RIL TRANSMITS THE RESULTS OF COMPLETED RESEARCH UPDATING AND EXPANDING THE COHCEPT i
COMPUIER CODE FOR FORECASTING CAPITAL COSTS DF BOILING WATER REACTOR PLANTS. IN 1971 THE ATOMIC ENERGY i
COMNISSION AUTHORIZED POWER PLANT INVESTMENT COST STUDIES, WHICH CULMINATED IN THE WASH-1230 REPORTS PUBLISHED IN 1972.
THEIR PURPOSE WAS TO FACILITATE POLICY AND ECONOMIC DECISIONS ABOUT ELECTRIC i
GENERATION FACILITIES IN THE PUBLIC AND PRIVATE SECTCRS. NATIONAL PRIORITIES ON ENERGY, THE REGULATORY ENVIRCNMENT AND THE COST OF LABCR. EQUIPMENT AND MATERIAL HAVE CHANGED SIGNIFICANTLY. THESE CHANGES DICTATED THd NECESSITY OF UPDATING THIS SERIES OF STUDIES. AND EXPANDING THE SCOPE TO CONSIDER THE FUEL CYCLE AND THE i
1 TOTAL GENERATING COST.
AS A RESULT. A PROGRAM TO STUDY. REASSESS AND PRODUCE A NEW SET OF UPDATED REPORTS WAS t
AUTHORIZED AND UNDERTAKEN.
l THE STUDIES 7N 1HESE SERIES HAVE 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 4 HYPDTHETICAL SITE.
THE ESTIMATED TOTAL BASE CONSTRUCTION COST FOR THE 1190 EWE BWR REFERENCE DESIGN.IS i
5582.748.330 OR S493/KW BASED CN JULY 1 1976 PRICES. THE TOTAL BASE CONSTRUCTIDM COST FOR THE BWR PCWER PLANT (1061 MWE NET OUTPUT) REFERENCE IN WASH-1230 WAS APPROXIMATELY $213.000.000 OR $201/KW. BASED UPON PRICES
+
I EFFECTIVE JANUARY 1971.
THUS. THE 1977 STUDY 'NDICATES APPROXIMATELY A 143 PERCENT INCREASE IN THE COST OF 1
THE PL ANT IN TERMS OF S/KW.
THE TOTAL DIRECT CRAFT LABOR COST OF APPROXIMATELY S139.500.000 CORRESPONDS TO AN
)
AVERAGE HOURLY RATE OF $12.29.
AP?ROXIMATELY 11.350.000 CRAFf LAar_ MANHOURS AVERAGE ABOUT 9.5 MANHOURS /KW.
THESE COMPARE TO AVERAGES OF $8.84/r00R AhD 6.3 MANHOURS /KW RESPEcTIVELY FOR THE EARLIER DESIGN REPORTED IN WASH-1230.
4 USER DISCUSSION POSITION COMMISSION ACRS PRESS 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS 4
ER11_RIL AIllVJlll}
PEVIEN HEL D COMPLETED HEL D HELD ISSUED IMPLEMENTED i
OF FICE PESPONSIBL E......... NRR l
SCHEDULED COMPLETION DATE..
_UNSCHED UNSCHED UNSCHED UNSCHFD UNSCHED UNSCHED ACTUAL C3MPLETIGN DATE.....
!!1R_ XOTMEN_L% A Rz_.fN VIR. TECH -
m DESCRIBE APPL ICAT_{0N 10_REcttL ATORJ__P_RACf31: THE PRINCIPAL APPLICATION OF THE INVESTMENT COST DATA IS TO UPDATE THE CONCEPT COMPUIER CODE.
IN TURN THE CONCEPT CODE IS USED TO ESTIMATE CAPITAL COST FOR DIFFERENT SIZE PL ANTS. DIFFERENT REGIONS OF THE COUNTY. DIFFERENT SCHEDULE LENGTH. DIFFERENT COMPLETION DATES.
DIFFERENT ESCALATION END INTEREST RATES.
IN ADDITION TO UPDATING CONCEPT THE DAT* IS USED AS A GEHERAL j
REFERENCE FOR SUCH THINGS AS. COST OF COMPONENTS, QUANTITIES OF MATERIALS USED. TYPE AND QUANTITY OF t
LABOR. ETC.
DsiCPIPE IMPACT OF RESQLT1: DURING FY 1977 44 REQUESTS WERE MADE FOR CONCEPT CODE RUNS AND 32 REQUESTS 4
j MERE f*ADE IN 1973.
fl05I REQUESTS INVOLVED RUNS FOR SEVERAL SIZES OF COAL PLANTS AND ONE OR TWO NUCLEAR i
UNITS FOR DIFFERENT ESCALATION AND INTEREST RATES.
THE USE OF THE CONCEPT CODE PERMIfTED THE STAFF TO 1
PERFCRM THESE ANALYSES MORE EFFICIENTLY AND QUICKLY THAN IF THE CONCEPT CODE WERE NO A"AILABLE.
l COMMENTS / REMARKS: DOE IS FUNDING THE L ATEST UPDATE.
i n
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a Pv10 RAM OFFICE CON.MENIS OM POTENTIAL 41ALJZ AilDN OR V AL9E OF RC%E ARCH RESyL T3 IN THE REGULATORY PROCESS RIt l' 46 DATE ISSUED: 02/12/79 RES DECISION UNIT: SYSTEMS EPGINEERING RIL TITLE: EFFECTIVENESS OF CABLE TRAY COATING MATERIALS & BARRIERS IN RETARDING THE COMSUSTION OF CABLE TRA15 SUBJECTED TO EXPOSURE FIRES AND IN PREVENTING PROPAGATION BETWEEN CABLE TRAYS HORIZONTAL OPEN SPACE CONFIGURATION)
S?ONSORIN". CFFICE(S): SD, NRR RRL: 1-25 ELECTRICAL STANDARDS RESEARCH PROJECT MGR:
R. FEIT
& FIRE PROTECTION RES_JfC1ESIS : DATA IS / RESENTED ON THE EFFECTIVENESS OF SIX FIRE RETARDANT COATING MATERIALS 'ND BARRIER DESISNS IN HCRIZONTAL OPEN SPACE CONFIGURATIONS THAT ARE IN USE OR BEING CONSIDERED FOR HUCLE
'0WER PLANTS.
AM ACCEFTABLE TEST METHODOLOGY WAS DEVELOPED BY WHICH PASSIVE FIRE PROTECTION MEASURES CAN BE ALUATED.
t 1
THE TESTS DEVELOPED CAN BE PERFORMED 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 SITUATIGNS WaEP.E THE DESIGN BASIS FIRE IS SIGNIFICANTLY DIFFERENT THAN THE TEST CASE FIRES.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS ESSLPlL_ACTIVIT fi EEVIEy HELD C0f1PLETED HELD HELD 1121Ep,_
IMPL EMEtiTED l
OFFICE RESP 0MSILLE.........
SD/NRR SCHEDULED CCMPLETION DATE..
ACTUAL CGMPLETICN DATE.....
SD COMMENTS - THE ENGINEERING METHODOLOGY STANDARES BRANCH WILL USE THE RESULTS TO SUPPORT THE STAFF POSITION IN REGULATORY GUIDE 1.120 AND IN THE DEVELOPMENT OF A REGULATORY GUIDE ON FIRE STOPS.
tLRPmC E ENTS, NOVEMBER 1980, R.
FERGM50IlLE. SYLVESIIR:
213fR131 AFPLICATION_10 RE0uipTORY PROCESS: THIS LETTER COMPARES THE EFFECTIVENESS OF COATING TRAY COVERS, AND C ABLE INSUL TIO:4 UAIERI ALS IN PREVENIING FIRE PROPAGATION IN HORIZONTAL CABLE TRAYS.
SUCH CONFIGURATIONS ARE PROPOSED BY LICENSEE 5 TO RETARD FIRE PROPAGATIOH. THE RIL RECOMMENDS QA REQUIREMENTS FOR MINIMUM THREE HOURS OF COATING AND FOR SOLID COVERING OF CAELE BUNDLE, I.E.,
d0 AIR PATHS THROUGH THE BUNDLE.
HEM 9_I1Lif' PACT OF_RESULTS: MINIMAL, ALL RESULTS WITHIN THE PERFORMANCE GOALS ASSUMED IN THE FIRE HAZARDS ANALYSIS.
QUALITY ASSURANCE SH3ULD BE ACCORDING TO THE LICENSEES QA PROGRAM.
SROUIS11LREU&Ril: THE TEST RESULTS HAVE LIMITED APPLICATION IN EVALUATIONS BECAUSE NO VERTICAL CONFIGURATIONS WERE IESIED, CALLES WERE DEENERGIZED, RANDOM FILLED PATTERN WAS USED, TEST CONFIGURATION WAS IN RELATIVELY OPEN AREA, i
AND EFFECT OF FIRE SIZES ON FIRE PROPAGATION IS NOT KNOWN.
THIS PROGRAM DID HOT HAVE SPECIFIC GOALS STATED IN TERMS OF LICENSING CONCERNS.
l j
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PROG 247 0FFICE COMN tN{$ CN POTENilAl_ UTILIZATION OR VALUE OF PESEARCH RESULTS IN THE REGULATORY PROCESS RIL s: 47 DATE ISSUED: 03/17/79 RES DECISION UNIT: FUEt c' OLE SAFETY & ENVIRONMENTAL EFFECTS PIL TITLE: INREM II: A COMPUTER IMPLEMENTATIO. sF RECENT MODELS FOR ESTIMATING THE DOSE EQUIVALENT TO CRGANS OF MAN FROM AN INHALED OR INGESTED RADIONUCLIDE SPONSCDING OFFICE (5): NRR ggg: 5-24 RAD 10 BIOLOGY RESEARCH PROJECT MGR:
J.
FOULKE
- DCSIt1ETRY R ES_C05j{N{$ : THE INREM II CODE AS CONTAINED IN NUREG/CR-9114.
ALSO TRANSMITTED IS VOLUME 1 0F A TABULA 1 ION OF INTERNAL RADIATION DOSE CONVERSION FACTORS OBTAINED USING THE INREM II CODE.
THIS TABULATION IS GIVEN IN "ESTIMATES OF INTERNAL DOSE EQUIVALENT TO 22 TARGET ORGANS FOR RADIONUCLIDES OCCURRING IN ROUTINE RELEASES FRCM NUC6 EAR FUEL-CYCiE FACILITIES." VOL.
1, NUREC/CR-0150.
THE CODE COMPUTES REFERENCE ADULT HUMAN DOSE EQUIVALENTS FROM USER-SUPPLIED DOSIMETRIC AND METABOLIC INFORMATION.
IN PRINCIPLE, INREM II APPROACH IS SIMILAR TO THE GTE USED IN MASH-1400.
ORGAN DOSE EQUIVALENT IS COMPUTED AS THE SUM OF CONTRIBUTIONS FROM EACH SOURCE CRGAN MHERE RADI0 ACTIVITY IS ASSUMED PRESENT; CROSS-IRRADIATION EFFECTS CAN BE ASSESSED. DOSE CONVERSION FACTORS IN NUREG/CR-0150 AND UFCCMING VOL. 2 ARE ILLUSTRATIVE OF INREM II CODE USE ONLY.
THEY ARE NOT ENDORSED FOR ADOPTION SINCE SOME Oa THE METABCLIC MODELS ARE SUBJECT TO CRITICISM WITH FURTHER DEVELOPMENT OF ACCEPTABLE MLTAB0LIC MODELS BY ORNL, THE CODE WILL PROVIDE STATE-OF-THE-ART METHODOLOGY FOR DOSE CALCULATIONS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING ERIEFING RELEASE RESULTS EOST RIL AC.TIVITIES MV F
t,M HFLD COMPLETED HELD HELD ISSUEQ_
IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHECULED COMPLETION DATE..
ACTUAL CGMPLETION DATE.....
fiRR_C OM ENIS_CN__PR E 15, 1979, Aih_51TE_AN3 LYSIS:
D L1C R I B E A P P L I.C3110 M_lR_R LGU LALO R Y PRO 1ESS: PRESENT ANALYSES OF RADIATION EXPOSURE SUCH AS PREPARED FOR WASH-1400, GESMO AND IN REACTOR LICENSING ACTIONS HAVE NOT SEEN UNIFORM IN DOSE ESTIMATION METHODOLOGY. THE VARIOUS APPROACHE USED FOR THE ASOVE STUDIES AND ACTIONS REFLECT VARIOUS INTERPRETATIONS OF THE STATE-OF-THE-ART IN DOSE ESTIMATION METH000LCGY, THE NEEDS OF PARTICULAR STUDIES AND AVAILABILITY OF DOCUMENTED MODELS.
NRC DOES NOT HAVE A BROADLY
/PPLICABLE. DOCUMENTED, CRITICALLY REVIEWED, STATE-OF-THE-ART DOSE ESTIMATION METHODOLOGY AVAILABLE TO VARIOUS
.? RICES FCR THEIR ASSESSMENT NEEDS.
THE CLOSEST IS REGULATORY GUIDE t.109. REVISION 1,
1977. WHICH REPRESENTS A Cbm 'D4T L Y AP?RCPRIATE APPLICATION OF A REASONABLY UP-TO-DATE DOSE ASSESSMENT METHODOLOGY AS APPLICABLE TO NUCLEAR POWEw PLANT EFFLUENTS.
THE OFFICE OF NUCLEAR REGULATORY RESEARCH RECOGNIZED THE NEEDS IN THIS AREA, AND HAS BEEN FUNDIA RESEARCH AT THE HEALTH AND SAFETY RESEARCH DIVISION OF THE CAK RIDGE NATIONAL LABORATORY (ORNL).
THE RESEARCH CESCRIBED IN RIL :47 PARTIALLY FULFILLS THE NEED OF A BROADLY APPLICABLE. DOCUMENTED AND CLOSER TO THE STATE-OF-THE-ART MODEL FOR CALCULATION OF DOSES TO VARICUS ORGANS OF A BODY FROM INTERNALLY DEPOSITED RADICHUCLID HOMEVER. ITS APPLICATION TO REGULATORY PROCESSES HAS TO WAIT UNTIL SOME ADDITIONAL SUPPORTING RE5EARCH IS COMPLE (SEE CCMMENTS/ REMARKS) MOST OF WHICH ARE ON-GGING AT ORNL FUhDED dY RES AND SOME BENCH-f1 ARK COMPARISONS ARE MADE.
DESCRlEE_10P_*Cl_QF_P_Fdt G :
THE RESULTS DAVE CONFIRMED OUR UNDERSTANDING OF INTERNAL DOSIMETRY MODELING, BUT HAVE NO DIRLCI IMPACT ON LICENSING UNTIL AFTER SOME ADDITIONAL STUDY IS PERFORMED. -
C_055EMT1/2EMAPrd: ICRP HAS AN ONGOING CONTRACT WITH HEALTH ASU SAFETY RESEARCH DIVISION OF ORNL TO DEVELOP REVISED NUCLEAR DATA, S-FACTORS. METABOLIC PARAMETERS. DOSIMETRIC MODELS, ETC. FOR USE IN ICRP'S FORTHCOMING PUBLICATION 333.
IT IS NCM KNC*.!N THAT THE CRNL INTERNAL DOSIMETRY MODEL FOR ICRP IS SOMEWHAT DIFFERENT FROM THAT FOR NRC (INREM-II)
ALTHOUGH EDTH MODELS ARE EASED CN ICRP'S TASK GROUP LUNG MODEL (1966, 1972) AND EVE'S GI-TRACT MODEL (1966).
THE ORN*.
MODEL FOR ICRP IS MORE CCM? TETE IN THAT IT HAS THE BODY FLUIDS (BLOOD) REPRESENTED IN THE MODEL AS AN ADDITIONAL CGMPARTMENT WHICH IS L ACKING IN INRE 1-II.
INCLUSION OF A BLOOD COMPARTMENT IS APPROPRIATE CONCEPTUALLY AS WELL AS FROM VIEW POINT OF REALISTIC MODELIN0; IT IS LIKELY TO REDUCE THE ORGAN DOSE CONVERSION FACTORS PARTICULARLY FOR THE SHORTER LIVED RADIONUCLIDES DUE TO DECAY CURING PARTIAL HOLD-UP IN THE BLOOD.
BENCHMARKING INREM-II RESULTS WITH THE ICRP'S WILL BE FRUDENT BEFORE OUR USING THEM.
THE 2EST VALUES OF METABOLIC PARAMETERS REQUIRED AS INPUTS TO RUN INREM-II ARE NOT YET AVAILABLE. ThE NOMINAL VALUES CF THESE PARAMETERS USED BY THE AUTHORS OF TWO VOLUMES (VOL. 2 IS IN PREPARATION) 0F DOSE CONVERSION FACTOR TABULATICNS ARE FOR THE PURPCSES OF ILLUSTRATING THE USE OF INREM-II. AND THE AUTHORS DO CAUTION AGAINST UNCRITICAL USE CF THESE TASLES IN RADICLOGICAL APPLICATIONS BECAUSE THEY CONSIDER THESE RESULTS AS PRELIMINARY. RES SPOULD 3E REQUESTED TO INCLUDE THE TASK OF GENERATING AND THEN PERIODICALLY UPDATING A DATA LIBRARY OF BEST VALUES OF METAB0LIC PARAMETERS FCR USE IN INREM-II IN THEIR CONTINUING RESEARCH PROGRAMS WITH ORNL.
A SET OF VALUES OF METABOLIC PARAMETERS IS BEING USED IN ICRP WCRK AT ORNL.
THE ICRP VALUES ARE NOT YET AVAILABLE TO USE, BUT WHEN AVAILA3LE. IT WILL BE PRUDENT TO COMPARE THESE VALUES WITH THE PRELIMINARY VALUES USED BY ORNL FOR HRC, AND MAKE APPROPRIATE REVISION IF NECESSARY.
THE NUCLEAR DATA DEVELOPED AND UTILIZED BY ORNL FOR NRC ARE NOT HECESSARILY THE SAME AS BEING DEVELOPED FOR ICRP.
THOUCH SUBSTANTIAL DIFFERENCES ARE NOT ANTICIPATED IN THIS AREA. IT WILL BE PRUDENT TO COMPARE THFSE DATA ALSO.
LASTLY. THE S-FACTORS AND METABOLIC PARAMETFRS FOR INDIVIDUALS OF AGE GROUPS DIFFERENT FROM ADULTS ARE STILL BEING DEVELCPED AT CONL FCR NRC.
UNTIL THIS WORK IS COMPLETED IT IS NOT POSSIBLE TO GENERATE DOSE CONVERSION FACTORS FOR AN INDIVIDUAL OTHER THAN AN ACULT.
AS A MATTER CF INFCRMATICN. OSD HAS A CONTRACI WITH ORNL TO PROVIDE THE NRC WITH AS MLCH INFORMATION REGARDING THEIR RESEARCH FOR ICRP AS WOULD BE PERMITTED BEFORE AND AFTER THE ICRP PUBLICATION 830.
l l
l PROG 9AM OFFICE CONNENTS CN PoiENTIAL UTILI2ATION GR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS l
RIL 3: 48 DATE ISSUED: 04/03/79 RES DECISION UNIT: SEISMIC, ENGINEERING & SITE SAFETY SAFETY l
RIL TITLE: A TECTONIC OVERVIEW OF THE CENTRAL MIDCONTINENT
} PONS 0?INO OFFICE (S): SD, NRR EES: 3-2 GEOLOGY 2 SEISMIC RESEARCH PROJECT MGR*
N. STEUER l
CHARACTERISTICS 1
RES_fDt31N152 "A TECTONIC OVERVIEW OF THE CENTRAL MIDCONTINENT", NUREG-0332. IS THE MOST UP-TO-DATE SYNTHESIS OF GEOLOGIC KNOWLEDGE OF THE EARIHS CRUST IN THE STUDY AREA.
IT CONTAINS THE MOST COMPLETE BIBLIOGRAPHY OF THE GE0 DYNAMICS OF THE AREA AND WILL AID IN NUCLEAR POWER PLANT LICENSING. IT IS RECOMME;4DED THAT THE INFORMATIGN AND HYPOTHESES EE USED AS INPUT TO THE DEVELCPMENT OF A TECTONIC PROVINCE GR SEISMIC ZONING MAP OF THE EASTERH U.S. AND TO PEDVIDE A BASIS AND GUIDE FOR ONGDING STUDIES IN THE AREA.
I USER DISCUSSION POSITION COMMISSION ACRS PRESS i
0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS l
EpST RIL AC11VITIES REVIFy HEt3 f.0MPLETED HELD HEL D ISSUED IMPLEMENTED 0FFICE RESFONSILLE......... SD/NRR SCHEDULED COMPLETION DATE..
ACTUAL CGMPLETIGN DATE.....
l SD CDMMENT$ - NO CGMMENT AT THIS IIME DUE TO PRELIMINARY NATURE OF THE WORK. WE PLAN AN EXTENSIVE REVIEW MMEN IhE PROJECT IS FURTHER ALONG.
l I -
PROGRsM OFFICE CCrWENTS 04 POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS I
4 RIL 4: 49 DATE ISSUED: 04/04/79 RES DECTSION UNITt FUEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS 1
PIL TITLE: IN VITRO DISSOLUTION OF URANIUM PRODUCT SAMPLES FROM FOUR URANIUM MILLS SPONSORING OFFICE (5): SD M: 5-23 DCCUPATIONAL RESEARCH PROJECT MGR:
J. FOULKE EX?OSURE 3 PROTECTION RES ComENT_1: THE MEASUREMENT OF THE SOLUSILITY OF VARICUS FORMS OF YELLOWCAKE IN VITRO UTILIZED TWO SOLVENT SYSTEMS: A J
SIMULANT OF AN ULTRAFILTRATE OF ELOOD SERUM AND 0.1M PCL.
THE RESULTS OF THIS STUDY WILL BE USED TO DETERMINE WHETHER YELLOWCAKE SHOULD BE CONSIDERED A SOLUBLE COMPOUND WITH RESPECT TO 10 CFR 20.
THE DATA OBTAINED USING THE SERUM SIMUL ANT SHOM THAT 25 TO 64 PERCENT OF ALL SAMPLES TESTED DISSOLVED WITH HALF-TIMES LESS THAN 16 HOURS.
IDEALLY. SOLUBILITY CLASSIFICATIONS SHOULD BE BASED ON THE DISSDLUTION HALF-TIMES OF THE PARTICULAR PRODUCT UNDER CONSIDERATION. THE DATA INDICATE THAT YELLOWCAKE SHOULD BE TREATED AS A CLASS D COMPOUND FOR THE PURPOSE OF EXPOSURE CONTROL.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEN HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... SD SCHEDULED COMPLETION DATE..
ACTUAL COMPLLIION DATE.....
SD COMMENTS. S. MCGUIRE - THE RESULTS OF THIS RESEARCH WILL BE USED IN FINALIZING REGULATORY GUIDE 8.22, "BI0 ASSAY Al URANIU.'1 MILLS." NC FURTHER RESEARCH ON THIS TOPIC IS PRESENTLY CONTEMPLATED.
. l
PROGRAM OFFICE CC*MEHIS 0% POTENIIAL UIILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL g: 53 DATE ISSUED: 04/06/79 RES CECISION UNIT: FUEL CYCLE SAFETY Z ENVIRONMENTAL EFFECTS RIL TITLE: CRITICALITY SAFETY GUIDANCE SPONSOPING OFFICE (S): NMSS (77-9)
REG:
5-7 CRITICALITY SAFETY RESEARCH PROJECT MOR:
D. SQLBERG i
STUDIES E15_CQ33ENT5: SINCE 1957. THE NUCLEAR SAFETY GUIDE HAS PROVIDED USERS CF FISSIONABLE MATERIAL WITH SIMPLE j
MEIHODS TO ASSURE SYSTEM SUBCRITICALITY.
RESEARCH INFORMATION LETTER 850 TRANSMITS NUREG/CR-0095. "NUCLEAR SAFETY CUIDE. TID-7016. REVISION 2."
THIS REVISION UTILIZES IMPROVED CRITICALITY DATA AWD COMPUTATIONAL TECHNICJ' ES NOT AVAIL ABLE IN THE PREVIOUS GUIDE ISSUED IN 1961.
REVISICN 2 SHOULD BE REGARDED AS A SUPPL EMENT TO PREVIOUS GUIDES SINCE IT DOES NOT HAVE INTENTIONAL C0i!SERVATIS!!S INCLUDED AS IN PREVIOUS GUIDES. IN USING REVISION 2 0F THE GUI3E. THE USER MUST IMPOSE APPROPRIATE SAFETY FACTJRS FOR HIS APPLICATICH AND THE NRC STAFF MUST DETERMINE THAT THE SAFETY MARGINS PROPOSED BY THE USER OF REVISION 2 ARE ADEQUATE.
l USER DISCUSSION POSITION COMh1SSION ACRS PRESS 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS 1
PR$T RIL_ACTLVITIES REVIEW HELD COMPLETED HELD 0FFICE RESPONSIELE.........
UELD ISSUED IMPLEMENTED SCHEDULED COMPLETION DATE..
ACTUAL CGMPLETION DATE.....
l E553 CO L~*ENl b_ M CH 3.
1930_R, $TEVf tiSAN:
1 H r. REVISED huCLEAR SAFEIY GUIDE IS QUDIED TO THE LICENSEES WHENEVER THEIR AMENDMENT OR RENEWAL OF APPLICATIONS REFER TO THE '965 VERSION OF THE GUIDE AND FAIL TO RECOGNIZE NEW DEVELOPMENTS IN NEUTRON INTERACTION-ACTION.
CERTAIN OF MHICH ARE SUMMARIZED iN THE UPDATED GUIDE.
THESE NC1 DEVELOPMENTS, SOME OF WHICH HAVE RESULTED FROM NRC-SPONSORED RESEARCH. HAVE RENDERED CUESTIGNABLE CERTAIN PAST PROCEDURES IN THE ANALYSES OF NEUTRON INTERA 1
a i
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PROGRAM OFFICE COMNENTS ON POTENil#L UTIt124 TION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL.s: 51 DATE ISSUED: 04/12/79 RES DECTSION UNIT: FUEL CYCLE SAFETY A ENVIRONMENTAL EFFECTS RIL TITLE: THE CONCEPT CCMPUTER CODE AND CAPITAL COSTS FOR PnC;;*J2!'** "_'.TER REACTOR PLANTS 4
SPONSORING OFFICE (S): NRR PPS: 5-21 SOCIO-ECONOMIC RESEARCH PROJECT MGR:
D. BARNA IMPACTS i
RE1_CQ7NENTS: THIS RIL TRANSMITS THE RESULTS OF COMPLETED RESEARCH UPDATING AND EXPANDING THE CONCEPT CGMPUTER CODE FOR FORECASTING CAPITAL COSTS OF PRESSURIZED WATER REACTOR PLAhTS. IN 1971 THE ATOMIC ENERGY COMMISSION AUTHORIZED PCWER PLANT INVESTMENT COST STUDIES, WHICH CULMINATED IN THE WASH-1230 i
REPCRTS PUBLISHED IN 1972.
THEIR PURPOSE WAS TO FACILITATE POLICY AND ECONOMIC DECISIONS ABOUT ELECTRIC GENERATICH FACILITIES IN THE PU2LIC AND PRIVATE SECTORS. NATIONAL PRIORITIES ON ENERGY, THE REGULATORY l
ENVIRONMENT AND THE COST OF LABOR. EQUIPMENT AND MATERIAL HAVE CHANGED SIGNIFICAMILY. THE?E CHANGES DICT ATED THE NECESSITY OF UPDATING THIS SERIES OF STUDIES. AND EXPANDING THE SCOPE TO CONSIDER THE FUEL CYCLE AND THE TOTAL GENERATING COST.
AS A RESULT. A PROGRAM 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 TECHNICAL CRITERIA AND A UNIFORM ACCOUNTING SYSTEM. THE INVESTMENT COST ESTIMATES IN THESE SERIES ARE DEVELOPED FOR REFERENCE PLANTS CONSTRUCTED AT A HYPOTHETICAL SITE.
THE ESTIMATEC TOTAL BASE CONSTRUCTION COST FOR THE 1139 ltJE PWR REFERENCE DESIGN IS
$568.831.011 OR $449/KW BASED ON JULY t.
1976 PRICES. THE TOTAL BASE CONSTRUCTISH COST FOR THE PWR POWER PLANT (1031 MME NET GUTPUT) REFERENCE IN WASH-1250 WAS APPROXIMATELY S2tt.000.000 OR S205/KW, BASED UPON PRICES EFFECTIVE JANUARY 1971 THUS. THE 1977 STUDY INDICATES APPROXIMATELY A 143 PERCENT INCREASE IN THE COST OF THE PLANT IN TERMS OF S/KW.
THE TOTAL DIRECT CRAFT LABOR COST OF APPROXIMATELY S133.100.000 CORRESPONDS TO AN AVERAGE HOURLY RATE OF $12.30.
APPROXIMATELY 10.820.000 CRAFT LABOR MANHOURS AVERAGE ABOUT 9.5 MANHOURS /KW.
THESE COMPARE TO AVERAGES OF $8.86/ HOUR AND 6.0 MANHOURS /KW RESPECTIVELY FOR THE EARLIER DESIGN REPORT IN WASH-1230.
1
}
USER DISCUSSION
?OSITION COMMISSION ACRi PRESS 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS l
E011_R1L_8CTIVITIES EEYIIH HELD COMPLETED HELD HELD ___
ISSUED IMPLEMENTED 0FFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
l ER_l0EE'8TS. #DdNVJR TECL J
EESCRI9E APPLIC ALION TQ_REGul3TERY PRAC133: THE PRINCIPAL APPLICATION OF THE INVESTMENT Cb5T DATA IS TO UPDATE THE CONCEPT COMPUTER CODE.
IN IURN THE CONCEPT CODE IS USED TO ESTIMATE CAPITAL COST FCP.
l DIFFERENT SIZE PLANTS. DIFFERENT REGICNS OF THE COUNTY. DIFFERENT SCHEDULE LENGTH. DIFFERENT COMPLETION l
DATES. DIFFERENT ESCALATION AND INTEREST RATES.
IN ADDITION TO L*PDATING CONCEPT THE DATA IS USED AS A GENERAL REFERENCE FOR SUCH THINGS AS, COST OF COMPONENTS. QUANTITIES OF MATERIALS USED. TYPE AND QUANTITY i
OF LABOR. ETC.
DISCRIBE IMPACT OF RESUlli: DURING FY 1977 44 REQUESTS WERE MADE FOR CONCEPT CODE RUNS AND 32 REQUESTS WERE MADE IN 1978.
MOST REQUESTS INVOLVED RUNS FOR SEVERAL SIZES OF COAL PL ANTS AND ONE OR TWO NUCLEAR UNITS FOR DIFFERENT c5 CAL.*.Y!ON AND INTEREST RATES.
THE USE OF THE CONCEPT CODE PERMITTED THE STAFF TO PERFORM THESE ANALYSES MORE EFFICIENTLi AND CUICKLY THAN IF THE CONCEPT CODE WERE NOT AVAILABLE.
COMMENTS /REMapgS DDE IS FUNDING THE LATEST UPDATE.
. 1 l
i4- ~
PRGMA!1 OFFICE CO TENf 5 CN POTENIIAL UTILIZATION OR VALUE OF RESET.ECH L ULTS IN THE REGULATORY PROCESS RIL 5 52 DATE ISSUQ:
07./2 979 RES DECISION UNIT: SEISMIC. ENGINEERING 3 SITE SAFETY PIL TITLE: EARTHQUAKE INTENSITY SCALE SPONSORING OFFICE (5): SD, NRR RAQ: 3-2 GEOLOGIC 1 SEISMIC RESEARCH PROJECT MGR:
CHARACTERISTICS F ES_LE*'EN TS : THIS MEMORANDUM TRANSMITS THE RESULTS OF COMPLETED RESEARCH ON THE SYSTEMATIC EVALUATION OF
[
THE MACROSETSMIC CATA FILE COMPILED FOR OVER HALF A CENTURY BY THE U.S. COAST AND GEODETIC SURVEY AND ITS 1
SUCCESSOR ORGANIZATIONS. A NEM SEISMIC INTENSITY SCALE WAS FORMED BY REVISION OF THE MODIFIED MERCALLI j
INTENSITY SCALE OF 1931 THE RESULTS HAVE BEEN PUBLISHED AS NOAA TECHNICAL MEMORANDUM. EDS NOSDC 4.
"REEVALUATIDH OF THE MCDIFIED MERCALLI INTENSITY SCALE FOR EARTHQUAKES. USING DISTANCE AS A DETERMINANT."
THE PURPOSE OF THE STUDY WAS TO RELATE EARTHCUAKE INTENSITY TO THE ENERGY RELEASED AND ITS ATTENUATION WITH DISTANCE TO PROVIDE A UNIFCRM SCALE RELATED MORE LOGICALLY TO THE PHYSICAL PARAMETERS OF ACCELERATICN. VELOCITY. DISPLACEMENT AND SPECTRAL CONTENT.
IT IS SUGGESTED THAT NRC MAKE A FORMAL RECCMMENDATION TO THE U.S.G.S.
THAT A NEW INTENSITY SCALE SASED ON THIS OR A SIMILAR ANALYSIS BE PROMULGATED AS AN OFFICIAL STANDARD.
THE NEW SCALE SHOULD PERMIT A TRANSITION FROM THE MM INTENSITY SCALE AND MAKE FULL USE OF THE EXTENSIVE DATA BASE ALREADY JVAILABLE.
t USER DISCUSSION POSITION COMMISSION ACRS PRESS 2
0FFICE MEETING PAPER 3RIEFING BRIE 71NG RELELSE RESULTS ITST RIL _Aq11yLTLES REyLfy hEL D COMPLETED HELD HELD ISSUED _
JJiPLEMENTED OFFICE RESPONSIBLE........
SD/NxR SCHEDULED CCMPLETION DATE..
1 ACTUAL COMPLETION DATE.....
SS COMMENTS: NO RESPONSE RECEIVED.
- 4 i
E2 R_CK"NNlb_t!0VD*aE5LE 1979 a JACESDN:
P_E3_C? lqu P?_LI C t U C N To REGutAiaR1 P1ggE11: NONE AT THIS TIME; TOO PREMATURE.
Elif?1S1_IUf1CT Of R L$u1J 1: liOME AT IHIS TIME.
CCMd1Ni$rgE11?L5: THI5 I t:F O RMA T I O N IS IMPORTANT FOR THE EVALUATION OF EARTHQUAKE INTENSITIES AND SHOULD BE FCAMARDED TO IHE USGS FOR CONSIDERATION IN ANY REEVALUATION OF THE INTENSIT7 SCALES.
4 1
1 i.. -
PROGR Art OFFIC E CO. vet *IS OM P0I LN TI A L UIILIZ A T ICN c5t V Ryf_pF (;tESEAECH REffLIS IN't'tE REGULATORY PROCESS RIL 8: 53 DATE ISSUED: 05/16/79 RES DECISION UNIT: FUEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS RIt TITLE: DEBRIS-BED COOLABILITY LIMITS, RESULTS FROM IN-CCRE TESTS D-8, D-2 AND D-3 SPONSCRINS OFFICE (S): NRR P&: 2-6 ACCIDENT ENEFGETICS RESEARCH IROJECT MGR:
R. WRIGHT
& DOSIMETRY Ef5_C W ENTS: THIS RIL REPORTS RESULTS OF THE INITIAL THREE EXPERIM5NTS IN THE ACPR TEST REACTOR ON INE COCLABILITY LIMITS OF BEDS OF PARTICULATE FUEL DEBRIS UNDER A PCCL OF SODIUM. A MODEL DESCRIBING THIS mEHAVICR IS ALSO REPORTED. THIS WORK SUPPLIES KEY INFORMATION FOR THE ASSESSMENT OF THE CISK OF A CORE-MELT ACCIDENT IN A LMTER.
IN THE EXPERIMENTS THE SPECIFIC FED POUER WAS MEASURED AT WHICH LOCAL DRYOUT OF THE S0DIUM COOLANT UCCURRED IN THE BED.
LOCAL ORY' JUT HAD BEEN THOUGHT TO BE THE COCLABILITY LIMIT CF A CCRE DEB 2IS BED, AN ASSUMPTION THAT HA3 BEEN USED IN SAFETY EVALUATIONS. SY SUSTAIN OPERATION AT POWER UNDER CD:IDITIONS OF LOCAL DRYOUT OF TH'i SODIUM IN THE BED. THESE FXPERIMENTS SHOWED THAT LCCAL DRYOUT IS NOT THE TRUE BED COOLABILITY LIMIT.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOST RIL ACJ R ITIES REVlfM HEL O CO*1P L ET ED HELD HELD ISSUED IMPLEMENTED OF F ICE R ESPONSIBL E......... hkR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
F.38L[ E ENTS. NOVf9fR 9._L979 J. !"EY EM :
F_01.1G Dt_ A P.fL I C A i ! O N_] pRift J L A_T 0 3 Y P ROCfS$ 2 SINCE THERE 15 NO LMFBR LICENSING ACTIVITY AT THE PRESENT TIME, THERE f
IS NO Iri;tEDIATE APPLICA ICH OF IHESE EXPERIMENTAL RESULTS TO THE REGULATORY PROCESS. HOWEVER, IF THE LMFBR LICENSING AGAIN BECC::ES SCTIVE, THIS PROGRAM HAS APPLICATION TO POST ACCIDENT HEAT REMOVAL F0"LOMING A CORE DISRUPTION ACCIDENT, IN PARTICULAR TO AN ASSESSMENT WHF.THER A DEBRIS-BED IS C00LABLE IN PLACE.
Df3E B_E IMP _ A_CT_1F_R13?jlTji: PRESENT IMPACT IS MINIMAL BECAUSE OF LACK OF LMFBR LICENSING ACTIVITY. HOWEVER, IN A GENERIC SENSE. THE RLSULIS ARE IMPORTANT. THEY SUGGEST MORE HEAT CAN BE REMOVED FROM DEBRIS BEDS THAN PREVI0d5LY THOUGHT. LOCAL DRYOUT IS NOT NECESSARILY LIMITING. THESE ARE THE FIRST EXPERIMENTS EMPLOYING FISSICH HEATING IN DEBRIS BEDS.
PAST EXPERIt1ENTS HAD TO SIMULATE FISilCN HEATING AND WERE NON PROTOTYPIC.
CQGE'iT. S/EEM A PK$ : THE GOALS OF THE FUTURE WORK ARE WORTWWHILE BUT MAY BE BETTER ACHIEVED WITH A SERIES OF EMALL SiUDIES USING SIMULANTS. THE PURPOSE SHOULD BE TO UNDERSTAND THE PHYSICAL PRCCESSES FROM WELL INSTRUMENTED EX-CSSEL EXPERIMENTS. THE M3RE COMPLEX IN-CORE EXPERIP.ENTS SHOULD BE FOR CONFIRMATION: THE INITIAL TESTS MAY HAVE AL&~CY ACHIEVED THIS COAL.
ATTEMPTING TO GAIN DETAILED INFORMATION FROM IN-PILE EXPERIMENTS MAY NOT BE COST EFFECI CC. -
~
. _.- ~
i PROGRtM OFFICE COM*ENis CN POTENTIAL UTILIZA(ION OR VALUE OF RESEARCH RESULTS IN THE REGULAIORY PROCESS RIL 8: 54 2AJf ISSUED 05/15/74 RES DECISION UNIT RItK ASSESSMENT eft TITLf THE SET EQUATION TRANSFORFATION SYSTEM SPONSORINO OF{1[{131: NRR/NNSS Egg: NONE RESEARCH PROJECT MGR:
W. VESELEY RES CQ33E$fi: AS A RESULT OF THE WCRK PERFORMED IN TdE SETS COMPUTER CODE PROJECT. AN IMPROVED VERSION OF THE Siis CODE HAS BEEN DOCUMENTED AND MADE AVAILABLE FOR USE BY NRC CONTRACTORS AND PERSONNEL FOR PROJECTS REQUIRING AN EFFICIENT TOOL FOR THE ANALYSIS OF COMPLEX SYSTEMS.
i IHE SAJOR RESULTS OF THE SETS PROGRAM HAVE BEEN:
(fs DEVELOPMENT OF AN AUTCMATIC TREE DECOMPOSITION ALGORITHM WHICH IS CURREMILY BEING INCORPORATED INTO THE STANDARD VERSION OF SETS; 1
- 42) DEVELOPMENT, IN PRELIMINARY FCRM, OF BASIC MINIMAL CUT SET QUANTIFICATION PROCEPURES; l
(3) DEVELOPMENT OF A STANDARDIZED VERSION OF THE SETS CEDE FOR THE CDC 660E COMPUTER, AND lhSTALLATION OF THIS VERSION OF THE CODE AT THE BROOKHAVEN NATIONAL LABORATORY CCMPUTER CENTER FOR USE BY NRC PERSONNEL: AWD (4) PREPARATI01 0F A SETS USERS" MANUAL ORIENTED SPECIFICALLY FOR THE FAULT TRFE ANALYST.
USER DISCUSSION POSITION COMMISSION ACRS PRESS 0FFICE 4EETING PAPER BRIEFING BRIEFING RELEASE RESULTS ED}LRJLA_CTIVIILFi EEVE u MELD
[0MPLETED HELD HEL D ISSUED ITIPL EMENTED OFFICE RESPONSIBLE......... hRR/hMSS SCHEDULED CCMPL2 TION CATE..
ACTUAL COMPLETION DATE.....
ER_(D*"*EN'.i_QN_)(ILY 27, 1979, S HANAyER:
DLSCRIBEdPPJJOU_Cy 10 Rta!LA[04Y PR_t tCf15 : THE IMPROVED VERSION OF THE SET EQUATION TRANSFORMATION SYSTEM (SETS)
IS A CCnPUTER CODE FOR EVALUAIING LARGE FAULT TREES.
THERE ARE CURRENTLY TWO AC'IVITIES WITHIN NRR THAT USE FAULT IREES AS AN ANALYSIS METHOD.
THESE TWO ACTIVITIES ARE GENERAL YASK NO. A-17 AND THE VITAL AREA ANALYSIS. THE SETS CODE PROVIDES THE NRC STAFF WITH THE CAPABILITY TO INDEPENDENTLY AUDIT THE RESULTS OF ANALYSES THAT ARE BASED ON FAULT TREES, ALTHOUGH IN THE CASE OF TASK NO A-17. THE CONTRACTOR WM9 DEVELOPS THE FAULT TREE ALSO USES SETS CODE TO ANALYZE THE FAULT TEEE.
THE AVAILABILITY OF THE SETE CODE ALSO."ROJIDES A TOOL FOR FUTURE USE BY THE NRC STAFF IN APPLYING FAULT TREE METH3CS TO THE LICENSING PROCESS.
THE SEY! CODE IS AN IMPORTANT FEATURE IN THE METHUDOLOGY (COMPUTER CODE) USED TO IDENTIFY TYPE I VITAL AREAS (I.E.,
"....THOSE AREAS WHEREIN *,UCCESSFUL SA30TAGE CAN PE ACCOMPLISHED BY COMPRGMISING OR DESTROYING TH VITAL SYSTEMS OR ConPONENTS LDCATED MITHIN AN AREA).
TYPE I VITAL AREAS ARE THE MOST SENSITIVE SECURITY AREAS IN THE PL ANT AND RCQUIRE THE HIGHEST LEVELS OF PHYSICAL PROTECTION. IDENTIFYING THESE AREEAS IS AN IMPORTANT CONSIDERATION IN THE STTFF's EVALUATION OF PHYSICAL SECURITY l
AT A NUCLEAR PO' ER PLANT.
THIS AN A' YSIS AND EVALUATION IS CURRENTLY BEING USED FOR ALL OPERATING PL ANT REVIEWS J
AND KILL ALSO BE AN CNGGING REQUIREMENT IN OPERATING LICENSE REVIEMS.
I
. I i
-,.--w.-.,.-
- --. -.,,,,,--~ _
THE SETS CODE HAS NO DIRECT IMPACT ONH THE LICEH3ING PROCESS BUT ID THE FUTURE IT WILL DL5f111LJW ACT QF PJ.5ULTS:INE NRC SI AFF 10 ANALYZE LARGE SYSTEMS THAT MICHT OTHERWISE HAVE TO BE ANALYZE PEEttII EFFECTIVE METHODS.
THE RESULTS OF THE MORE EFFECTIVE AND EFFICIENCY METHODS CCULD LEAD TO EITHER A RELAXATION CF EXISTING REQUIREMENTS TICHTENING OF AN EXISiING REQUIREMENT.
OR COULD LEAD TO NEW REQUIREMENTS CR A CCf"EtiTS/P FMLP 11: THE SETS CODE IS ONE OF SEVERAL COMPUTER CODES THAT CAN BE USED TO EVALUATE FAULT TREEL INC*UDItO EVALUATION FCR CCM. ION CAUSE FAILURES. K3 ATTEnPT HAS BEFN MADE HERE TO DETERMINE TH2 RELATIVE EFFECTIVENESS OR EASE CF USING THESE VARIuuS CODES.
P 7 0 91'1 0FFICE CCm1ENTS CH POIE..TI?,t UT ILI/ATION OR V ALUE OF RESEAPCd "tEdutIS IN THE REGULATORY PROCESS ell 8: 55 DATE IS%UED: 05/29/79 RES DECISION UNIT: FUEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS RIL TITLE: THE CCNCEPT COM1 UTER CCD2 AND CAPIiAL COST FOR HIGH AND LCW SULFUR COAL PLANTS - 120 0 t"AE SPONSORING OFFICE (S): NRR g,RL : 5-21 SOCI3-ECONOMIC RESEARCH PRCJECT MGR:
D. BARNA IMPACT 5 P/$_G *E*T5:
THIS MEMORANDUM TRAESMITS THE RESULTS OF COMPLETED RESEARCH UPDATING AN3 EXPANDING THE CONCEPT CCMPUTUt CL,DE FCP. FCRECASTING CAPITAL COSTS OF HIGH AND LCW SULFUR COAL PL ANTS - 1200 MWE.
THE ESTIMATED i
i TGTAL BASE CONSTRUCTICN COST FCR THE 1200 MWE f NGMINAL ) HIGH SULFUR COAL PLANT REFERENCE DESIGN IS $465,498,393 CR $373/KM BASED CN 'ULY 1,
1976 PRICES. THE ESTIMATED TOTAL BASE CONSTRUCTION COST FOR THE 1200 MWE (NCMINAL) LOW SULFUR CDAL PLANT REFERENCE DESIGN IS 6402,825,229 OR $324/dW BASED ON JULY 1,
1976 PRICES.
THE TOTAL B ASE CONSTRUCTION CCST FOR IHE COAL-FIRED PC!!ER PL ANT (1000 MWE NET GUTPU O REFERENCE IN WASH-1230 4
WHICH DID N3T HAVE FLUE GAS DESULFURIZATION IS APPROXIMATELY $174,000,000 CR f174/KW, BASED UPON PRICES EFFECTIVE JANUARY 1971.
THUS. THIS 1977 STUDY INDICATES APPROXIMATELY A 87.9 PERCENT INCREASE IN THE COST OF THE PLANT IN TERMS OF $/Ku.
THE STUDY AND ITS METH000LOGIES HAVE BEEN REVIEWED EXTENSIVELY WHILE IN PRCGRESS BY THE RES PROJECT MANAGER AND VARIOUS 57AFF MEMBERS FRCM NRR.
RLS RECDMMENDS THAT THE UPDATED METHODOLOGY BE USE9 BY NRR FOR APPLICATION TO THE IDENTIf IED REGULATORY NEED (RR-NRR-76 -C).
USER DISCUSSION POSITION C0f tMISSION ACRS PRESS CFFICE MEETING PAPCR BRIEFING BRIEFING RELEASE RESULTS PDST RIl_AQH YITIES EEYlfM HELD SONPLETED HEtD HELD ISSUED _
.INPLfMENTED OFFICE RESP 0xsIBLE......... NRR SCHEDULED CCMPLETICN DATE..
ACTUAL CCMPLETION DATE.....
BLR GTE N T $, 3 %-.EM1R. TECHa DE5_MJ B_E APPL IC AIQ*4_lq_R LG_L!1ATORY PROCfM: THE PRINCIPAL APPLICATION OF THE INVESTMENT COST DATA IS TO UPDAIE TriE CONCEPT CC:*2 UTER CODE.
IN It.oM THE CONCEPT CODE IS USED TO ESTIMATE CAPITAL COST FOR DIFFERENT SIZE PLANT $, DIFFERENT REGIONS OF TFE COUN*Y, DIFFERENT SCHEDULE LENGTH, DIFFERENT CCMPLETION DATES, DIFFERENT ESCALATION AND INTEREST RATES.
IN ADDITION 0 UPDATINC CONCEPi THE DAT1. IS USED AS A GEP:ERAL REFE".ENCE FJR l
SUCH THINGS AS, CCST OF CCMSGNENTS, QUANTITIES CC Mt.TERIALS USED, TYPE AND QUANTITY OF LABOR, ETC.
KSQ9JEE 15PACY OF PE$ 1TJ: DURIMS FY 1977 44 REQUESTS WERE MADE FOR CONCEr*T CODE RUNS AND 32 REQUESTS WERE MADE IN 1918.
MOST REQUESTS INVOLVED RUNS FOR dEVERM SIZE 3 0F COAL PLANTS AND ONE OR TWO NUCLEAR i
I UNITS FOR DIFFERENT ESCALATION AND INTEREST RAiES.
THE USE OF THE CONCEri CODE PERMITTED THE STAFF TO PERFCRM THESE ANALYSIS MCRE EFFICIENTLY AND QUICKLY.THAN IF THE CONCEPT CODE WERE t:0T AVAILA3LE.
I I
COPENTS/ REMARKS: DGE IS FUNDING THE 'ATEST UPDATE.
2 i
4 a
4 l i
PNGo*M OF Fif E CCvtNTS CN PO!ENTIAL uiILIZATICH CR VALUE OF R ESEP ?tCH RESULIS fN THE REf,ULATOPY PROCE*S RIl t: 56 DATE ISSUED: 07/25/79 PFS DFC*SION tih!T FUEL CYCLE SAFETY A ENVIRONMENTAL EFFECTS RIL T I T L _E : EFFECTS OF NUCLEAR ?CMER PLANTS CN COMMUNITY GROWTH AND RESIDENTIAL PROPERTY VALUES SPONSCPING OF QCEtSI: NRR 33p,:
5-21 SOCIO-ECONOMIC RESEARCH PROJECT MGR:
C. PRICHARD IMPACT',
ELES CO*""EtiLS:
AS PART OF A LONGER TERM RESEARCH PROJECT TO DEVELOP FORECASTING METHODS FOR ESTIMATING THE IMPACT OF NUCLEAR PCRER STATIONS CN LAN3 VALUES AND COMMUNITY GROWTH, THE PENNSYLVANIA STATE UNIVERSITY INSTITUTE FOR LAND AND WATER RESOURCES CARRIED OUT A STUDY OF THE EFFECTS OF NbCLEA8! POWER STATIONS ON LAND VALUES AND COMMUNITY GROWTH AT FOUR P.tEVIOUSLY-LICENSED STATIONS IN THE SCRTNEAST.
THE RESULTS OF THE STU.Y DID NOT INDICATE THE PRESENCF. OF SIGSIFICANT ADVERSE EFFECTS CN L AhD VALUES OR COMMUri!!Y GROWTH.
USER DISCUSSION POSITI0tl COMMISSIGN ACJS PRESS OFFICE MrETING
. APER BRIEFING BRIEFING RELEASE RESULTS EOST RTL ACTIVIHES ETVIE4 HELD COMPLETED HELD HEED ISSUED IMPLEMENTED OFFICE RESFCSSIBLE......... MRR SCHEDUL ED COMPB ETIGN DATE..
ACTUAL CCM?LETICN DATE.....
NM COIJNlls_iL_EUS$1LL -
DESC RIS E _ APP _LIC A TION 10_ELQ'lL ATJPY PRQCESS: AS PA2T OF THE COST BENEFIT ANALYSIS OF LICENSING APPLICATIONS, THE NRL a5 REGUIRED 10 ASSESS THE LIKELY SCCI0 ECONOMIC Ih? ACTS ASSOCI ATED WITH CONSTRUCTION AND OPERATION OF NJCLEAR PCMER STATIONS ON LOCAL CCMMUNITIES AND THE CURROUNDING REGIONS.
THE IMPACT OF OPERATING STATIONS ON LAND USE, PROPERTY VALUES, L AND USE PL ANNING ND POPUL ATION CHANGE ARE EFFECTS WHICH ltAVE RECEIVED CONSIDERABLE ATTENTION FROM INTERVENORS AND HEARING E0ARDS.
THIS RESEARCH CONTRACT WAS INITIATED WITH THE INTENTION OF TESTING A METHODOLOGY FOR QUANTIFYING THE IMPACT OF STATIbn SITING Od RESIDENTIAL PROPERTY VAIUES.
DE1ClI1E It'P ACT OF PE1MLLS: BECAUSE OF TriE MON-R ANDOM SELECTION OF THE STUDY SITE, CENERALIZABLE CONCLUSIONS C A S t*01 BE DRate.: HO.:tVER, F02 THE 4 SITES INVOLVED, THE RESULTS DO INDICATE THAT THE NUCLEAR STATIONS HAD NO MEASURAELE, ADVERSE IMPACT ON RESIDENTYAL VALUES IN SURROUNDING AREAS.
BESIDES ADDING TO THE STAFF'S GENERAL ENOWLEDGE. THE STUDY PROVIDED THE STAFF WITH A TESTED. OBJECTIVE, AND REPLICABLE METHODOLOGY FOR RETROSPECTIVELY DIMENSICNING PROPERTY VALUE IMPACTS AT NUCLEAR STATIONS. WITH RESPECT TO THIS POINT, THE STAFF INTENDS TO EVALUATE THE IMPAC1 CF THE ACCIDENT AT THREE MILE ISLAND ON PROPERTY VALUES USING A METHODOLOGY BASED ON THE PROCEDURE USED IN THIS CONTRACT.
THE T1ETHODOLOGY AND CONCLUSIONS OF THIS RESEARCH INCREASED THE STAFF'S UNDERSTANDING OF THE PROCESS OF PROPERTY VALUE IMPACT AT THE LOCAL LEVEL.
CONTINUATION OF RESEARCH CH PROPERTY VALUE IMPACTS SHOULD RESULTS IN GENERALLY RELIABLE METH005 FOR PREDICTING RESICENTIAL VALUE IMPACTS IN SPECIFIC SITING CASES.
CO*NENTS/ REM PKS: NONE l 1
...=
PROGR A*10FFICE CCEENTS ON POTENTIAL UTILIZAfl0N OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PRGCESS RIL 4: 57 DtTE ISSUEQ: 03/10/79 RES DECISION UNIT SYSTEMS ENGINEERING RIt TITLE: SMALL SCALE ECC BYPASS RESEARCH RESULTS SPONSCRING OFFICE (S): NRR Elfa:
1-6 ECC BYPASS RESEARCH PROJECT MGR:
A. SERKIZ SE1_CC"MrN D: ANALYTICAL AND EXPERIMENTAL STUDIES OF STEAM-WATER MIXING EFFECTS ON THE PENETRATION OF COLLING WATER IN SMALL MODELS OF PWR VESSELS HAVE BEEN CONDUCIED OVER THE PAST FIVE YEARS AT BATTELLE COLUMBUS LABORATORIES AND i
CREARE. INC.
FISCAL YEAR 1979 MARKED THE CD:!PLETION OF THE DULK OF THIS WORK AND A RESEARCH INFORMATION LETTER SUNTARIZING THE FINDINGS WAS PREPARED.
THE PHENCMENA INFLUENCING COOLING WATER PENETRATION IN THE SMALL 1/15 AND 2/15 SCALE MODELS IS WELL UNDERSTOOD AND A TRANSIENT MODEL SESCRIBING COOLING WATER PENETRATION IN THESE SMALL SCALE -
TESTS HAVE BEEN DEVEL0 RED.
SENSITIVITY STUDIES OF THE MODEL UNCERTAINTY (PRIMARY SCALING) WHEN USED TO CALCULATE COOLING MATER PENETRATION IN A PWR HAVE BEEN PERFORMED AND CCMPARISONS MADE WI(H LOFT DATA.
THE SMALL SCALE TESTS SUPPORT THE NEED FOR ADOITIONAL TESTS AT LARGER SCALE, BUT PROVIDE EVIDENCE OF THE CONSERVATISM IN THE MODELS USED I
IN THE LICENSING PROCESS 4
USER DISCUSSICH POSITION C00tISSION ACRS PRESS OFFICE MELTING PAPER BRIEFING BRIEFING RELEASE RESULTS P_p1T_Rll_A_CllVIT El REVID4 EELD C_0MPLETED HELD itELD ISSUED IMPLEMENTED l
OFFICE RESr0:4SILLE......... HAR SCHEDULED CD"./LETION DATE..
ACTUAL CC PLETION DATE.....
4 4
i
P90GRA7 Os5 ICE COMMENVS ON POVENTLAL UTXtiEATION OR VALUE OF OESEARCH RESULTS IN THE REGULATORY PPDCESS RIL 8: $3 DATE ISSUED:
CE/29/7')
RES DECISION UNIT: FUEL CYCLE SAFETT A ENV2RONMENTAL EFFECTS RIL TITLE: COMPARISON OF SIMULATION MODELS USED IN ASSESSING THE EFFECTS OF POWER PLANT INDUCED MORTALITY ON FISH POPULATIONS SPCNSC7ING OFFICER,11: NRR ElQ:
5-15 CHEMICAL IMPACTS RESEARCH_ PROJECT MGR:
P. REED ON AQUATIC ENVIRONMENTS PEl_EEEEETl$: THIS RESEARCH EVALUATED THE EXISTING MATHEMATICAL MODELS FOR PREDICTING THE IM?4CT OF HUCLEAR POWER PLANT OPERATION ON ECONCMICALLY IMPORTANT FISH SPECIES. THE MATHEMATICAL STATEMENTS. THE EQUATIONS, AND LHDERLYING ASSUMPTIONS USED FOR ASSESSING PCWER PL ANT INDUCED FISH MORTALITY WERE COMPAPED. VALUES OF PARAMETERS AND THE TECHNICAL CATA SOURCES USED IC C&TAIN ! HEM WERE INVESTIGATED. BECAUSE MANY OF THE MODELS HAD DIFFERENT BASIC ASSUMPTIONS, PARAMETRIC VALUES, CR EDTH, AN INTERACTIVE LIFE-CYCLE MODEL SIMULATOR WAS DEVELOPED TO COMPARE THE PREDICTIONS 1F THE VIRICUS MODELS.
IT WAS DETERMINED THAT NO PRESENTLY EXISTING MODEL CAN BE USED TO MAKE QUANTITATIVE IM?ACT PREDICTIDAS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS Ep37_FLA CTIVI LLE1 PEvlE4 HEiD COMPLETED HELD HELD ISSUED IMPLEMENTED LFFICE RESPCNSIBLE......... hKR SCHEDULED CCM?LETICM DATE..
ACTUAL CCMPLETION DATE.....
t l
m m
P90GRA*1 CFFICE COF?EN f 5 09 POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN T HE REGUL A TORY,fSCCESS i
l RIL et 59 DATE ISSUED: 09/214 9 RES CECISION UN((r, FUEL BEHAVIOR RIL TITLE: TRANSIENT FUEL R3D BEHAVICR CODE:
FRAP-T4 i
SPONS0*TN", OFFICE N : NRR RP&:
1-12 FUEL CODE RESEARCH PROJECT MGR:
G. MARINO DEVELOPMENT 1
i RES C E iSTJ:
FRAP-Y4 IS A BEST-ESTIMATE COMPUTER CODE THAT CAtTULATES THE THERMAL AND MECHANICAL RCSPONSE OF A hUCLEAR FULL RCD DURING NORMAL. OFF-N3RMAL, AND TRANSIENT CONDITIb;*L THE CODE HOW HAS THE CAPABILITY To BE USED 3
IN TeiE ANALYSIL OF THE ENTIRE SEQUENCE OF A LOCA THAT WENT THROUGH RULOOD. THE CODE HAS BEEN uSED TO ANALYZE SUCH EVENTS FOR THE PCMER BURST FACILITY TEST PROGRAM AND THE LOSS-OF-COOLANT FLUID TEST PROGRAM AT THE IDAHO NATIONAL ENGINEERINC LABCRATORY.
IT HAS ALSO BEEN USED TO STUDY OTHER 124NSIENT EVENTS SUCH AL P.EACTIVITY-
{
INITIATED ACCIDENTS, PCMER-CCOLING MISMATCH EVENTS. AND ANTICIPATED TRANSIENIS.!ITH0di SCRAM.
THE REPORT SUSARIZES THE CODE'S PERFORMANCE IN THE ABOVE AREAS VI A DATA COMPARISONS. CONTAINS A
SUMMARY
TABLE WHICH SHCMS THE STANDARD ERROR 3ETMEEN DATA AND CALCULATED RESULTS, AHD CONCLUDES WITH A SECTION ON USER RECGT*Eh3ATIONS BASED ON THE CODE ASSESSMENT RESULTS.
1 USER DISCUSSION POSITION COMMISSION ACRS PRESS i
0FFICE MEETING PAPER BRIEFING ERIEFING RELEASE RESULTS l
P_DS T__ MIL ACTIv_IIJf1 EEVIEW HELD CMPLETED HFLD HELD ISSUED IMPLEMENTED i
0FFICE RESPONSIELE......... hRR SCHEDULED CCMPLETION DATE.,
ACTUAL CCMPLETION DATE.....
4 i
l l 1
&ROG#fM OfrICE CtrMENf5 C4 PGTENTIAt UTItIZATIOM nP VALUE OF RESEARCH QESULTS IN THE REGptATORY PROCESS RIL 8 60 DATF ISSUED:
10/12/79 RES DECISTON UNIT: SEISMIC, ENGINEERING & SITE SAFETY DIL TITLEe SEISMICITY AND TECTONIC RELATIONSHIPS OF THE NEMAHA UPLIFT IN OKLAHOMA 5FONSSPINS OFFICEfS): NRR REQ:
3-2 GEOLCGY a SEISMIC RESEARCH PROJECT MGR H. STEUER CHARACTERISTICS Pf5_C055sNTS: CE0 LOGIC AND SEISM 0 LOGIC INVESTIGATIONS OF THE NEMAHA UPLIFT BEGAN ON OCTOBER 1,
1976.
THE GEOLOGICAL SIUDIES HAVE FOCUSED, THUS FAR. ON THE CONSTRUCTION OF A SERIES OF STRUCTURE-CONTOUR MAPS CH KEY STRATIGRAPHIC HOFIZCNS: THE TCP OF THE ORDOV'.CIAN VIOLA FORMATION. THE BASE OF THE PENNSYLVANIAN, AND THE TOP OF THE OSWEGO FCkMATICN.
THE CCNTOdR-MAPPING PHASE OF THE PROGRAM IS APPROXIMATELY TWO THIRDS COMPLETED. THE INITIAL MAPPING PROGRAM REVEAL 5 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 PENHSYLVANIA TIME, AT LEAST IN CENTRAL CKLAPLiA.
A DISCUSSION OF RASEMENT ROCKS IN CENTRAL OKL HOMA IS INCLUDED WITHIN THIS REPORT.
THE SEISMOLOGICAL STUDIES HAYE CONCENTRATED ON THE INSTALLATION OF EIGHT SEISM 0 METERS IN SUCH A WAY AS TO INCLUDE DETAILED COVERAGE OF THE ENTRIE NEMAHA RIDGE IN CKLANOMA AS WELL AS.MOST OF THE REMAINING AREAS OF OKLANCMA.
USER DISCUSSION POSITION COMMISSION ACR$
PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS E.01_T_R T L. a C T IV LT] El EF.V1fB HFtD LOMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHCOULED CCMPLETICN DATE..
ACTUAL CCMPLET70N DATE.....
filR._Cf'"jEH1L_1401_EMER 9,
1979:
ED ER15t E _LLCAll R TO RE utATOPY PRSCf33: IT IS RECOMMENDED THAT THE INFORMATION CONTAINED IN NUREG/CR-0050 BE CCNSIDERED LY IHE OFFICE OF KUCLEAR REACIOR REGULATION AS INPUT IMFORMATION TO THE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZCNING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND CUIDE FOR ONGOING STUDIES IN THE AREA.
DESCRIBE IPEACT O F_RES')1TS : THE AVAILA3ILITY OF T:tE RESEARCH HAS BROADENED 002 DATA BASE FOR THIS REGION BUT CURRENTRY EAS h3 DIFECI InPACT CN OUR LICENSING ACTIVITIES. THE GEOLOGIC AND SEISMIC DATA BASE AVAILABLE IN THIS MEPCRT IS INSUFFICIENT TO CCMPLETELY EVALUATE THE AREA EXCEPT ON A PRELIMINARY EASIS.
A SEISUOTECTONIC MODEL FOR (HE 6EMAHA UPLIFT IN CKLAHOMA MUST ALSO BE BASED IN PART ON CONCEPTS DEVELOPED FR0il RESULTS OF THE STUDIES AND MUST CCNSIDER VERTICAL AND L ATERAL VARI ATIONS IN COMPOSITION AND PHYSICAL PROPERTIES OF THE NEMAHA UPLIFT.
C255LN11/ DEN 3RKS: THIS MUREG CCNSTITUTES PART OF A LARGE DATA GATHERING AND SYNTHESIS EFFORT FOR THE NEMAHA RIDGE AREA.
THE TOTAL IMPACT CANNOT BE ASSESSED UNTIL THE OVERALL PROGRAM IS COMPLETED AND SYNTHESIZED WITH SEISMIC MONITORING DATA. 'THIS WILL TAKE SEVERAL YEARS.
t k
n -
PM G> c1 0 F F I C E C D"V N T S O N POTENTIAL U f f t f 2
- T ig4 CR VALUE CF RESEAPCH RESULis IN THE REGULATORY PROCESS RIL 9 61 D4TE ISSUED:
1;/11/79 RES DECISION UNIT: FAST BREEDER REACTORS RIL TITLE: f*0LTEM SODIUM INTERACTIf'N WITH B ASALT CONCRETE SPONS0eING OFFICEfS): NRR Eers:
2-4 CORE MELT AND RESEARCH PROJECT MGR:
T. WALKE2 CONTAINMENT INTEGRITY
?Ei_f7NNW LARGE SCALE MOLTEN SODIUM-BASALT CONCRETE TESTS WERE CARRIED OUT IN SUPPORT OF THE NRC STAFF POSITION Te4AT VIGOROUS SODIUM-CONCRETE REACTI0fts COULD BE EXPECTED.
BASALT CONCRETE PENETRATION RATES OF ABOUT 2.5 CM. Ptd HOUR MERE OBTAINED.
THE REACTION CONTINUES UNTIL THE SJDIUM IS CONSUMED ON EITHER BARE CONCRETE OR BENEATH A CEEECTED STEEL LINER.
CEFECTS AS SMALL AS 0.6 X IS CM. DO NOT PLUG BUT PERMIT ESSENTIALLY ALL OF THE SODISM TO KEACT WITH THE CONCRETE WHILE THE REACTION PRODUCTS DEFORM THE LINER.
SILICEOUS FIREBRICF INSULATING LAYERS BENEATH THE LINERS ALSO REACT READILY WITH THE SODIUM.
IF THE SIDE WALLS ARE EXCLUDED, THE TOTAL CONCRETE PENETRATION IS ABOUT S n CF THE INITIAL SODIUM POOL DEPTH.
USER DISCUSSION POSITION CCMMISSION A C P.S PRESS OFFICE MEETING PAPER BRIEFING B2IEFING RELEASE RESUL15 E01TELaCTJVJTJ E1
- EvifW HELD COMPLETED HEL D
'< E L D ISSUED IMPLEMENTED OFFICE RESPONSICLE......... NRR SCHEDULED COMPLETION DATE..
ACTUAL COMPLEIICM DATE.....
iip 8Lif""Ew TiWVlM t ER,_L19 79 :
Rfm13 E A PPL IC3lLON 10 REGQLATORY PROCESS: AS INDICATED IN DR. LEVINE'S LETTER OF OCTOBER 11, 1979 NRR REQUESTEC (CASE 10 LEVINE LEI 1ER CAtED 3/7/19) THAT THE CONFIRMATORY RESEARCH PROGRAM INVESTIGATE THE INHERENT RETENTION CAPABILITY r0R 50DIU-1 AND SOME CORE DEBRIO BE REDIRECTED TCMARDS FFit' RATHER THAN CRBR FOR WHICH LICENSING ACTIVITIES HAD EEEN SUSPENDED.
THERE IS NO LICENSING ACTIVITY ON CRBR AND ITS FUTURE IS UNCERTAIN.
THE FFTF IS ESSENTIALLY COMPLETE. THEREFORE. THE INTEREST IN THIS EFFORT FOR LMFBRS IS GENERIC, IN NATURE.
THE MOST DIRECT APPLICATION OF THE REMAIN 0ER CF THIS EFFORT MAY BE TOWARDS TrE PROTOTYPE LARGE BREEDER STUDY.
PElC 53 BJL_l?2 A C T_0E_RE1ULTi t THE RESULTS OF THIS WORK STRONCLY INFLUENCED THE ADVICE AND RECOMMENDATIONS GIVEN 10 THE fFif PROJECT REGARDING THE ADEGUACY OF EXISTING FFTF CONTAINMENT MARGINS AS REPORTED IN THE STAFF'S SAFETY EVALUATION REPORT NUREG-0353. SUPPLEMENT
- t. DATED MAY 1979.
THE RECOMMENDATION MADE BY TH5 STAFF AND SUPPORTED BY ACRS INCLUCED MEANS FOR MONITDP'NG THE CONTAINMENT AIMOSPHERE FOR RADIATICN. TEMPERATURE, PRESSURE, OXYGEN, AND THE INCORPCRATION OF HYDROGEN NITERS AS WELL AS MEANS TO SCRUB / FILTER THE CONTAINMENT ATMOSPHERE IN THE EVENT CF MELTDOWN TO REDUCE THE RADIGLOGICAr RELEASES.
f/4"ENT1/EE"AH 3 FUTURE GENERIC STUDIES OF OTHER CONCRETE AGGRAGATES. CELL CONFIGURATIONS (CYLINDRICAL. RECTANGULAR)
SHOULD BE CONDUCTED AS WELL AS TESTS WITH ANC WITHOUT FLAMED LINERS.
(FLAWS SIMULATING WELD CRACKS SHOULD BE CONSIDERED.) DIFFERENT TYPES OF FIREBRICK (E.G.,
COMMEPCIAL MGO, AL203, ZR02) SHOULD BE EXAMINED IN TERMS OF COMPATIBILITY WITH SCDIUM AND ABILITY TO PROTECT THE CONCRETE. _ _ _ _. _ _ _ _ _ _ _ _.. -. _ _ _ _ _ _ ___
PROGR AM OFFICE CorEFf 5 ON P01ENf1AL UTILIZxTION 09 VALUE OF RESEA*iCH R ESu t T S IN THE LEGULATO:zY PROCESS RIL #: 62 UATE ISSUER: 11/01/79 RES DECISION HNITr SEISMIC. ENGINEERING & SITE SAFETY RIL TITLE: NEW MADRID SEISM 3TECTCMIC STUDY SPONSORING OFFICE (5):
HRR. SD ERO 3-2 GEOLOGY & SEISMIC RESEARCH PROJECY MGR:
N. STEUER CHARACTERISTICS PEl_.C M LNT_3:
AN IMPCRTANT GOAL OF THE RESEARCH PROGRAM IS TO PRODUCE USEFUL SEISM 0 TECTONIC AND SEISMIC ZCMING MAPS FOR THE STUDY AREA.
FISCAL YEAR 1973 WAS THE SECOMD YEAR OF A FIVE-YEAR PROGRAM. RESULTS OF AER0 MAGNETIC SURVEYS FUNDED IN FY 77 WERE INTEGRATED WITN PREVIOUSLY EXISTING DATA IN ADJACENT AREAS.
EXTENSIVE GRAVITY SURVEYS WERE MADE IN KENTUCKY AND INDI ANA NEAR THE INTER 5ECTICM CF Tt:E 3&TH PARALLEL LINEAMENT Ar4D THE NORTHEASTERN EXTENSION OF THE NEW MADRID SEISMIC ZONE.
THE STATICMS WERE GRAVIMETRICALLY CONSTRUCTED. AN INTERESTING REL ATIVE POSITIVE ANOMALY OCCURS PARALLEL TO THE WA2ASN VALLEY FAULT SYSTEM.
USER DISCUSSION POSITION COMMISSION ACRS PRESS i
CFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS i
E.pST pit _aCTlv_UJE1 E.EllEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OfflCE RESPONSIBLE......... NRR,SD l
SCHEDULED CCMPLETION DATE..
1 ACTUAL COMPLETICM DATE.....
W28L.C.pMN T3, hoygMf Ee_,,13 1979.
R.
JACK $pH:
REM ilf_E_AP.^tICATICN IQ_QQqLa193r F9QSD.3: THE GOAL OF THIS RESEARCH BY SEVERAL STATE GEOLOGICAL SURVEYS AND UNIVERSIIIES IS 10 DEF*NE IHc SIRUCYURAL SETTING AND TECTONIC HISTORY OF THE REGION AROUND NEW MADRID. MISSOURI IN ORDER TO PROVIDE THE BASES FG2 A MORE REALISTIC APPRAISAL CF THE EARTHQUAKE RISKS IN THE SITING OF NUCLEAR FACILITIES IN THE NORTH AMERICAN MID C0dTINENT.
IT IS RECCEMENDED THAT THE INFORMATION CCNTAINED IN NUREG'S 0739 AND 0450 BE USED AS A B ASIS FCR CONTINUING RESEARCH. AS INPUT TO THE EVALUATION OF SEISMIC RISK IN THE REGION WITHIN AND AROUND THE MISSISSIPPI EMBAYMENT, AND AS & CONTRIBUTION TO OUR UNDERSTANDING CF INTERPLATE TECTONICS IN GENERAL.
DEls1 LEE IPP CT OF R ESVL!_S : THE RESEARCH EFFORT THUS FAR HAS INCREASED OUR CURRENT DATA BASE AND QUR UNDERSTANDING J
CF EARIHOUAKE AND FAULT PHENOMENA IN THE MISSISSIPPI EMBAYMENT REGION. BUT AS YET NO DIRECT IMPACT GN LICENSING.
[0 Y ENTig r*apr$:
NURES'S 0739 AND 0450 SUMMARIZE THE STUDIES AND RESULTS OF THE FIRST TWO YEARS OF A FIVE YEAR FECGRAM. THEREFCRE, IT IS 700 EARLY TO ASSESS THE IMPACT ON NUCLEAR POWER PLANT LICENSING EXCEPT IN A VERY PRELIMINARY i
WA (.
THE TOTAL IMPACT CANN3T BE ASSESSED UNTIL THE OVERALL PROGRAM IS COMPLETED AND SYNTHESIZED WITH SEISMIC j
MONITORING DATA.
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fFQGRAM OFFICE C0*MENTS ON POTENTIAL UTILIZATI0tt OR VALUE OF RESEARCH RESULis IN THE REGULATORY PROCEiS PIL 6: 6S DATE ISSUED:
11/01/79 RES DECISION UNIT: LOFT l
PIL TITLE: LCFT REACTCR SAFETY PROGRAM RESEARCH RESULTS FROM NUCLEAR LOSS-OF-COOL ANT EXPERIMENTS L2-2 AND L2-3 l
SPONSORING OFFICE (S): KRR RES:
1-1 LOFT RESEARCH PROJECT MGR G. MCPHERSCM FES CC5MSNT%: THE 2ND LOFT RIL SUM 4ARIZES THE RESULTS OF THE FIRST TWO NUCLEAR LOCE'S PERFORMED AT THE LOFT FACILIIY IN F~t 19 7 9.
THE TWO TESTS ARE PART OF THE L2 LARGE COLD LEC EREAK EXPERIMENTS.
L2-2 WAS CONDUCTED OF A LINEAR HEAT GENERATION PATE OF 3 KW/FT AND L2-3 WAS AT 12 KW/FT.
EDTH TESTS, WHICH ASSUMED AVAILABILITY OF OFF-SITE PCWER. DISPLAYED A DOU3LE REVERSAL OF CORE FLCW DLRING DEPRESSURIZATION. THE RETURN TO POSITIVE CORE FLCW WAS SdFFICIENT TO GUENCH THE CORE PRIOR TO THE INITIATION CF EMERGENCY CORE COOLANT INJECTION. IMPROVEMENTS ON CCMPUTER C00E MODELS AND N00A*IZATION ARE IDENTIFIEO WHICH PERMIT A C00D PREDICTION OF THE OBSERVED PHENCMENA.
ALL OF THE BEST-ESTIMATE CODES USED TO PREDICT L2-3 PREDICTED GENERALLY HIGHER CLADDING TEMPERATURES THAN WERE MEASURED.
i USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPCR BRIEFING BRIEFING RELEASE RESULTS EqST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HEL D ISSUED
. IMPLEMENTED OFFICE RESPCNSIBLE......... MRR SCHEDULED COMPLETICN DATE..
ACTUAL CJMPLETION DATE.....
5 4
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Pw3CR AM 0FFICE CCMMENTS ON POTEMTIAL UTILIZATION OR VALUE OF PESEARCH RESULTS IN THE REGULATORY Pn0CES$
RIL 9: 64 DATE ISSUED:
11/05/79
.RES DECISION UNIT: SEISMIC. ENGINEERING & SITE SAFETY UL TITLE A REVISED AND AUGMENTED LIST OF EARTHQUAKE INTENSITIL "~ - KANS AS. 1367-1977 SPONSORINS OFFICE (S):
5-2 GEOLOGr & aEISMIC RESEAPCM PROJECT MOR:
N. STEUER CHARACTERISTICS EES COMMENTS: THE PURPOSE OF THIS RESEARCH IS TO GAIN A BETTER UNDERSTANDING OF THE SOURCES OF EARTHQUAKES THAI HAVE CCCURRED IN THE REGICN AS AN AID TO DEVELOPING A MORE RATICHAL EVALUATION OF EARTHQUAKE RISK AS IT APPLIES TO THE SITING AND CESIGN OF NUCLEAR FACILITIES.
TWENTY-FIVE EARTH 1UAKES WHOSE EPICENTERS WERE WITHIN THE BORDERS OF KANSAS HAVE BEEN REPORTED DURING THE PAST ttC YEARS.
BECAUSE OF THE CRITICAL NATURE OF EARTHQUAKE INFORMATION IN ESTIMATION OF %EISMIC RISK. IT IS IMPORTANT THAT 1HE 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. REGICSAL. AND NATIONAL EARTHQUAKE LISTINGS. IN ADDITION. OLD NEWSPAPER FILES. MICROFILMS, AN3 OTHER REC 9RDS AT THE UNIVERSITY CF KANSAS AND THE KANSAS STATE HISTORICAL SOCIETY WERE SEARCHED FOR REPCRTS WHICM MAY HAVE BEEN PREVIOUSLY DVERLOOKED OR NOT RECORDED.
THIS REPORT INCLUDES A C07.PLETE LIST OF ALL FELT REPORTS COMPILED DURING THIS STUDY.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS UliT_?It aCTlyJLili erVIEW HELD COMPLETED HELD HELD 755UED IMPLEMENTED OFFICE RESPONSIBLE......... NRR.SD SCHEDULED CCMPLETION DATE..
ACTUAL CCMPLETION DATE.....
+-
td U D*"Egl$,_ R Cf1MER 5.
1979. J,_{ N_LG_H T :
RSCRIBE aPr uc3_T10N TO PF99(A_LOyY_PROCEli THE GOAL OF THIS RESEARCH WAS TO REVIN D4E REPORTS OF JLL EARTH 4UAKES WHOSE EPICEMitRS WERE WIIMIN THE BOUNDARIES OF KANSAS.
THIS RESEARCH IS PART OF A CorPEftTIFE GEOLOGIC. SEISMIC, AND CECPHYSICAL RESEARCH PROGRAM BY SEVERAL STATE GEOLOGICAL SURVEYS THAT IS SEEKING TO DEPIF2 THE STRUCTURAL SETTING AND TECTCNIC HISTORY OF 1HE NEMAHA UPLIFT AND THE MIDCONTIMENT GRO!ITY ANOMALY IN ORDE9 3G ?ROVIDE THE BASES FOR A MORE REALISTIC AFvRAISAL OF THE EARTHQUAKE RISKS IN THE SITING OF NUCLEAR FACILTIIES IN THE NMW NECAN MID CONTINENT.
IT IS RECO:~".t.nDED THAT THE INFORMATION CONTAINED IN NUREG/CR-0294 BE USED AS A B ASIS FOR CONd WING RESEARCH AND AS INPUT TC TAE EVALUATION OF SEISMIC RISK IN THE REGION WITHIN AND ARQUND THE NEMAHA UPLIFT.
ESC #13 E_JTP_A C_T_p FlFJ QGS : THE RESEARCH EFFORT THUS FAR HAS INCREASED OUR CURRENT DAT A BASE AND OUR UNDERSTANDING OF EAAis1 % KE PHEM MENA IN IHE VICINITY OF THE NEMAHA UPLIFT. BUT AS YET HAS HAD NO DIRECT TMPACT ON LICENSING.
NUREC/CR-5294 MAS BEEN REFERENCED BY AN INTERVENOR IN THE WOLF CREEK HEARING.
THE INTERVENDR REQUESTED THE ST AFF TO REASSESS THF SAFE SMU!D0%N EARTH'IUAKE (SSE) AT THE WOLF CREEK SITE IN LIGHT OF 1HE NEW INFORMATION PRESENTED IN THIS MUREG.
THE STAFF REVIEWED THE REPORT AND FOUND THAT THE SSE AT THE WOLF CREEK SITE WAS STILL ADEQUATELY CONP.ERVATIVE.
SC E NTSeR9uPJ_$: NUREG/CR-0294 COMPRISES A PARf 0F A FIVE-YEAR CETAILED STUDY OF THE SOURCES OF SEISMICITY IN THE NLMAMA UeLIFT AREA.
THIS IS AN INTERIM TCPICAL REPORT PRESENTING RESULTS OF WCRK COMPLETED IN PHASE I: THEREFORE.
IT IS T03 EARLY TO ASSESS THE IMPCT ON NUCLEAR PO K R PLf.NT LICENSING EXCEPT IN A VERY PRELIMINARY WAY.
THE TOTAL IMPACT.
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CANNOT SE ASSESSED UNVIL THE OVERALL P00 GRAM IS CGMPLETED AND SYNTHE5IZED WITE SEISMIC MOSITCRING DATA. THESE
(
PEELIMIMARY RESULYS ARE BEING CataIDERED IN THE DEVELOPMENT OF A TECTONI" PRCVINCE 02 SEISMIC ZONING MAP OF THE EASTERN U.S.
THE STAFF DGE5 DISAGREE WITH CNE CF THE RESULTS STATED IN THE NUREG - THE 1867 MANHATTAN EARTHQUAKE. THE ASSIGNMENT OF INTENSITY VII-VIII 15 BASED UPON AN 1L77 REPORT OF LIQUEFACTION ON A FARM ON THE FLOODPLAIN OF THE KAMSAS GEOLOGICAL SURVEY.
THAT C3SERVATION MAS ASSIGNEC I'4 TENSITY VIII AND PLACED CLOSE TO THE EPICENTER BY THE KANSAS GEOLCGICAL SURVEY.
MUCH RECENT MORK HAS 530' N THAT LIQUEFACTION 15 EXTREMELY DEPENDENT UPON LOCAL SITE CONDITIONS AND J
MAY 6CCUR IN ISOSEISMAL AREA 5 THAT MAY OTVERMISE BE ASSOCI ATED WITH INTENSITIES LESS THAH VIII ( AS LCM A5 VI FOR EXAM?tE).
THE STAFF AGREES MITH THE STANDARD REFERENCES WHICH LIST THIS EARTHQUAKE AS AN INTENSITY VII (MM).
THE STAFF ALSO FINOS THE EPICENTRAL LOCATION OF THE 1867 MANHATT AN EARTHQUAKE QUESTIONABLE. ACCORDING TO THE KANSAS
{
CEOLOGICAL SURVEY. SHAKING AND EUILDING DAMAGE EQUIVALENT 10 INTENSITY 'oII OCCURRED DVER AN AREA AT LEAST 200 KM i
ACROSS. PINPOINTING THE EPICENTER MITHIN THAT AREA MAY SE BEYOND THE RESGLVING POWER OF THE PRESENT DATA.
- RA -
~.. _. - --
PRgG7AM 0FFICE COMMENIS ON POTENTIAL UTILI2ATION OR VAtyE OF RESEARCH QESULTS IN THE REGULATORY PROCES5 RIL 8: 65 D.4TE ISV).E.D:
11/05/79 RES DECISION UNIT! 5EISMIC, ENGINEERING 1 SITE SAFETY RIL TITLE: RECONNAISSANCE BEDRCCK CE0 LOGIC MAP OF MARLBOROUGH QUADRANGLE, MA AND RECONNAISSANCE BEDROCK GEOLOGIC MAP GF SHREWSBURY QUADRANGLE, MA
- PONSORING OFFICE (S):
HRR, SD EPft:
3-1 NRC/ STATE REGIONAL RESEARCH PROJECT MGR:
N. STEUER EARTH SCIENCES RES_COU'3EfjTS: THE HAPS TRANSMITTED BY THIS RIL ARE PRODUCTS OF THE NEW ENGi AND SEISM 0 TECTONIC STUDY WHICH IS A PROGRAM OF INVESTIGATIONS TO BETTER UNDERSTAND TME MANIFESTATIONS AND CAUSES OF SEISMICITY IN NEW ENGLAND AND ADJACENT APEAS TO ASSESS THE SEISMIC HAZARD TO PROSPECTIVE NUCLEAR PDF R PLANTS IN THE REGION. RES REC 0Z'.iNDS
USER DISCUSSION POSITION COMMISSIDH ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEA%E RESULTS EDSI_RLL_ACTIYUlfi M ILEH HELD COM?tETED HELD HELD ISSUtD IMPLEMENTED OFFICE RESPONSIBLE......... NRR,SD SCl!EDULED COMPLETION DATE..
ACTUAL COMPLETI0N DATE.....
H3P_C9""(M13,_ LAM [ARY 9, 19$b J Pl!G T,:
uDE1CE1EJf P_ U Gu_I_LON_ I(LPJ,9ULA T QRY PRDCES3: THESE TWO GEOLOGIC MAPS ARE PRODUCTS OF THE NEW ENGLAND SEISM 0 TECTONIC STUDY, WHICH IS A PROGRAM CF INVESTIGATIDHS TO ASSESS THE SEISMIC IMZARD OF THE REGION.
THEY CONTAIN THE KIND OF DETAIL THAT IS NEEDED IN AREAS OF CCMPLEX MAJOR FAULTING AND/OR RELATIVELY HIGH SEISMICITY. THEREFORE, IT IS RECOMMENDED THAT THE MAPS BE CONSIDERED AS INPUT INFORMATION TO THE DEVELOPMENT OF A SEISMOTECTCNIC PROVINCF MAP AND AS A BASIS FOR CONTINUED STUDY IN THE AREA.
DfJCPLPjLlMPACT OF PESULTS: THE DATA ON THE MAPS HAVE NO DIRECT IMPACT ON LICENSING ACTIVITIES, BUT HAVE ADDED TO THE STAFF'S GENERAL KNCWLEDGE REGARDING THE MOST GEGLOGICALLY COMPLEX AREA IN NEW ENGLAND.
THEY CONFIRM TH'.T THERE IS A MAJOR STRUCTURAL EDUNDARY SEPARATING SOUTHEASTERN MSSACHUSETTS FROM THE RLST OF ThE NEW ENGLAND PIED.%NT PROVINCE.
NO EVIDENCE OF RECENT MOVEMENT ON THE FAULTS WAS FOUND.
S0ZM_E N T3fffM A_RKS : THE BENEFITS OF If:VESTIGATIVE r.FFORTS SUCH AS THIS WILL NOT BE FULLY REALIZED UNTIL OTHER GEOLOGICALLY COMPLEX ANC/02 RELATIVELY HIGH SEISMIC AREAS ARE STUDIED AND ALL THE INFORMATI0:4 SYNTHESIZED AND INTER-i
'8.IED AT T H AT TIME THE PRODUCTS WILL CONTRIBUTE TO THE CONSTRUCTION OF A SEISM 0 TECTONIC PROVIA l MAP.
l - -- -
PROG;U1 05FICE COMMENTS ON POTEN?IAL UTILIlATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS PIL t'
66 DATE ISSUED:
11-06-79 RES DEOISION UNIT: SEISMIC. ENGINEERING Z SITE SAFETY RIL TITLE: A STUDY OF THE REGIONAL TECTONICS AND SEISMICITY OF EASTERN KANSAS -
SUMMARY
OF PROJECT ACTIVITIES AND RESblTS TO THE END OF THE SECOND YEAR OR SEPTEMBER 30, 1978 SP9NSCPIN*, OFFICE (5):
NRR, SD PEQ: 3-1 NRC/ STATE REGIONAL RESEARCH PROJECT MSE:
N. STEUER EARTH SCIENCES fl2_C05"ENTS: THIS RIL TRANSMITS RESULTS TO SEPTEM3ER 30, 1973 0F THE STUDY CONDUCTED OF THE EARTH SCIENCE PARAMETERS OF THE NEMAHA UPLIFT AND IHE MIDCONTINENT GRAVITY ANOMALY GEOLOGIL STRUCTURES. THE INFORMATION GAINED IS OF VITAL IMPORTANCE IN THE SITING AND LICENSING OF NU"LIAR POWER PLANTS.
THIS RIL PRESENTS PROJECT WGRK COMPLETED IN PHASES I AND II 0F A 3-PHASE PROJECT AND PRESENTS 1) EXISTING DATA SYNTHESIS, AND 2) ACQUISITION OF NEW DATA, SEISMIC NETWORK INSTALIATION AND GPERATION.
RES RECCMMENDS THAT SD AND NRR CONSIDER THIS 45 INPUT TO THE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND A BJSIS AND GUIDE FOR DHGOING STUDIES IN THE AREA.
COCUMENT ISSUED:
NUREG/.-0666 ( ALSO SEE RIL 870, 11/19/79 AND dUREG/CR-0375).
USER DISCUSSION POSITION COMMISSION ACRS PRESS GFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST PIL_ACTIVlllfi
- F EV11M HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIbtE......... hRR,CD SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
11PLC05'iENTS, 01CLMEP 1} d 979, J.
M M IAH T_:
PLSLRIBk6tD1C Allp 10 PJMMTMLMC133: THE P"JRPOSE OF THIS KANSAS RESEARCH IS TO GAIN A BETTER UNDERSTANDING OF 1HE SOURCES Of EARINQUAKES IHAI HAVE UCCURRED IN TAE REGION.
THE KANSAS EFFORT IS A PART OF A REGICNAL RESEARCH PROGRAM WHICH ALSO INCLUDES THE STATE GEOLOGICAL SURVEYS 0: OKLAHOMA, NEBRASKA AND IOWA.
THIS RESEARCH PRGGRAM IS TO SERVE AS AN AID TO THE NRC IN DEVELOPING A RATIONAL EVALJATICH OF EARTHQUAKE RISK AS IT APPLIES TO THE SITING, DESIGN AND REEVALUATION OF NUCLEAR FACILITIES LOCATED WITHIN THE STUDY RLGION.
RE3CPJ E_Ui? ACUF RESULTS: THIS RESEARCH EFFCRT HAS INCREASED OUR GEOLOGICAL, GECPHYSICAL AND SEISMOLOGICAL DATA BASE OF EASTERN KANSAS, BOTH THROUGH THE ACQUISITION AND GENERATICH OF NEW IMFCRMATION AS WELL AS THROUGH A COMPILATION AND EVALUATION OF PREVIOUSLY EXISTING DATA.
THIS KANSAS RESEARCH EFFORT, IN PARTICULAR, HAS STRENGTHENED OUR CONFIDENCE IN THE PREVIOUS GEOSCIENCES DECISIONS MADE IN CONJUNCTIDH WITH THE SITING AND LICENSING OF THE WOLF CREEK SITE IN COFFEY CCUNTY, KANSAS.
FUTURE SITING AND LICENSING DECISIONS FOR EATERN KANSAS WILL BE EXPEDITED AS A RESULT OF THIS RESEARCH EFFORT.
THE RESULTS OF THE REGIONAL RESEARCH EFFORT, WHEN COMPLETED, WILL PROVE VALUABLE IN THE DEVELOPMENT OF A TECTONIC PROVINCE / SEISMIC ZONING MAP OF THE EASTERN UNITED STATES.
CCpjEi.T S/ R EMA RK3 : ALTHOUGH ITS CONCLUSIONS ARE PRELIMINARY (IT CONSTITUTES ONLY A TWO-YEAR SEGMENT OF A FIVE-YEAR PROGRAM Inv0LVING SEVERAL STATES). THIS KANSAS RESEARCH REPORT (NUREG/CR-0666) IS A MELL THOUGHT DUT AND EXECUTED EFFORT CONT AINING CONSIDERABLE VALUABLE DATA THAT HAS NOT ONLY INCREASED OUR KNOWLEDGE, DN A BROAD SCALE, OF MID-CONTINENT TECTONICS BUT HAS, MORE DIRECTLY, PROVIDED ADDITIONAL SPECIFIC INFORMATION FOR EASTERN KANSAS.
THIS NEW INFORMATION HAS NOT CAUSED LICENSING DELAYS.
FUTURE LICENSING ACTION, INVOLVING NOT ONLY WOLF CREEK BUT OTHER NUCLEAR FACILITIES THAT MAY BE LOCATED WITHIN THE REGICN 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 QUESTIONS THE KANSAS GEOLOGICAL SURVEY'S (KGS) RATIONALE FOR THE RELOCATION OF TWO EVENTS (1867 AND 1906) FEELINt THAT SUCH RELOCATIDH NAY Br i.EYOND THE RESblVING POWER OF THE PRESENT DATA.
ADDITIONALLY, THE ASSIGNMENT OF AN INTENSITY VII-VIII OR VIII TO iwE 1867 EARTHQUAKE BY THE KGS ON THE BASIS OF A SINGLE OCCURRENCE OF LIQUEFACTION REPORTED 10 YEARS AFTER THE EARTHQUAKE IS QUESTIONASLE. THE STAFF AGREES WITH THE STAhDARD REFERFHCES LISTING THIS 1367 EVENT AS AN INTENSITY VII (MM).
88 -
PROGRAN OFFICE COMFENTS GN POVfMIAAA OfillZATION QR VALUE OF RESEARCH PE3ULTS IN THE RM ULATORY PROCESS RIL s, 67 DATE IS%UED: 11-06-79 RES DECISION UNIT SYSSEMS ENGINEERING RIt TITLE: REFLOODING OF SIMULATED PWR CORES AT LOW FLCM RATES l
SPO*150 RING OFFICE (S): NRR (77-02)
R3fq:
1-5 frFLOOD HEAT RESEARCH PROJECT MGR:
L. THCMPSON TRANS.ER l
THIS RIL DESCRIBES THE COOLING OF ELECTRICALLY HEATED RODS DURING BOTTCM FLOODING EXPERIMENTS CONDUCTED l
EE}_COMENTS:
THE INFORt1ATION PRESENTED IS C0tiSIDERED APPLICABLE TO THE EWALUATION OF EMERGENCY l
AT cot 4ST ANI INLET FLOODING RATES.
COOLING SYSTEM PERFCRMANCE IN PRESSURIZED WATER REACTORS. THE RESULIS PRESENTED IN THE RIL ARE RECOMMENDED FOR CONSIDERATION IN THE APPLICATION AND APPRAISAL OF EVALUATION MODELS FOR REFLOOD HEAT TRANSFER.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESUL15 P_0ST_flLAClIYIJJIS EfVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMEMTED OFFICE RLSPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
PROGOAM OF FIC E C07 MEN T S Or4 PQTfNTIAL UTILIZATION OR VALUE OF PESEARCH RESULTS IN THE REGULATOkV Pk& CESS RIL #
68 DATE ISSftED:
11-11-79 RES DECISION UNIT: PRIMARY SYSTEMS INTEGRITY RIL TIT (E: STRUCTURAL INTEGRITY 0F WELD REPAIRED PRESSURE VESSELS 120NSDPING OrFICEfS): NRO P33:
1-20 VESSEL INTEGRITY EE1EARCH PROJECT MGR:
M. VAGINS Rf1_fCgggNT3: THIS RIL DESCRIBES THE RESULTS OF A TFST PROGRAM TO DETERMINE THE ADEQUACY OF A PROCEJURE WHICH EMPLOYS THE HALF-BEAD WELD REPAIR TECHNIQUE. THE RESULTS OF THE STUDIES REVEAL THAT THE ASME SECTION XI WELD REPAIR TECHNIQUE WILL PRODUCE A EEPAIRED STRUCTURE HAVING A SUFFICIENTLY HIGH LEVEL CF RESIDUAL STRESSES THAT THE STRUCTURE COULD HAVE AN INSUFFICIENT MARGIN OF SAFETY AGAINST FRACTURE. ADDITIONAL RESEARCH IS NEEDED TG PERFECT MODIFICATIONS TO THE PRESENT PROCEDURE S0 THAT RESIDUAL STRESSES ARE MINIMIZED OR ELIMINATED IN THE REPAIRED STRUCTURE. RES NOTES THAT SUCH A PROGRAM IS ALREADY UNDERMAY UNDER EPRI SPONSORSHIP: RES IS FOLLOWING THIS WORK USER DISCUSSION PGSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS fMT_PJL_A QHyJ TJ(1 DEHEW HELD COMPLETED HEL D HELD ISSUED IMPLMEtLYJJ OFFICE RESPCNSIBLE......... h2R SCHEDULED COMPLETION DATE.,
ACTUAL COMPLETION DATE.....
IER_CRM5LNlie._Pf CIUB ER 19.
1979 L. SHAQ:
D MC PJ BL i tP LI_C A IJ pf4_I O R B U LA10 R Y PR2[fli: THE TESTS DESCRIBED IN THE RIL ACCOMPLISdED THEIR PURPOSE, WHICH WAS TO MEASURE lHE FRACIURE BEHAVIOR OF HALF-BEAD WELD REPAIRS IN A PRESSURE VESSEL AT TEMPERATURES NEAR AT/NDT AND UN THE UPPER SHELF.
SOUND, CRAC%-FREE WELDS WE2E PRODUCED, UHICH BEHAVED AS WELL AS BASE METAL WHEN THE VESSEL MAS PRESSURIZED AT THE UPPER-SHELF TEMPERATURE.
HOWEVER. THE TEthNIQUE LEFT RESIDUAL TENSILE STRES%ES OF YIELD STRENG!M MACNITUDE NORMAL TO THE WELD IN THE MATERIAL ADJACENT TO IHE WELD. WHEN A LARCE FLAW WAS FLACED IN THE RESIDbAL SiRESS REGION AND THE VESSEL MAS PRESSURIZED AT THE RT/NDT TEMPERA.URE, THE CRACK POPPED IN ABOUT 2 INCHES AT A PRESSURE OF 0.4 TIMES DESIGN AND POPPED THROUGH THE WALL AI ABOUT DESIGN PRESSURE.
TO EXPLAIN THIS FRACTURE BEHAVICR, THE EFFECTS OF RESIDUAL STRESS MUST BE CONSIDERED ADDITIVE TO THE EFFECTS OF PRESSURE WHEN THE TEMPERATURE IS IN THE TRANSITION REGION.
THE TESTS DESCRIBED IN THE RIL DO HUT ADDRESS FRACTURE BEHAVIOR WHEN THE UPPER-SHELF ENERGY HAS BEEN REDUCED BY IT2ADIATION.
ALTHOUGH SECTION SI 0F THE ASME BIP7 CODE PERMITS II, NO HALF-BEAD WELD REPAIRS OF SIGNIFICANT DEPTH WITHOUT POST-WELD STRESS RELIEF HAVE BEEN PERMITTED IN THE REACTOR COOLANT PRESSURE BOUNDARY. HOWEVER, PERMISSION TO CO !O WILL PROBABLY BE REQUESTED IN THE FUTU?E.
WE CONCLUDE THAT HAtF-BEAD WELD REPAIR WITHCUT STRESS RELIEF CANNOT EE PEiMITTED WITHOUT l
A FRACTURE ANALYSIS THAT MEETS THE REQUIREMENTS OF SECTICN XI SUPPLEMENTED AS FOLLOWS: A) THE EFFECT OF A POSTULATED FLAW IN THE REGION OF HIGH RESIDUAL STRESS MUST BE EVt'"ATED.
THE ASSUMED SIZE OF SUCH FLAW MUST SE JUSTIFIED BY CONSIDERATION CF THE POS) WELD NDE PROCEDURES USED ANL etE POSSIBILITY OF RECURRENCE OF THE CONDITIONS THAT CAUSED i
THE ORIGINAL FLAW (I.E., IF THE FLAM UAS SERVICE-INDUCED, ANY MEASURES TAKEN TO RELUCE FATIGUE USAGE OR OTHER CAUSES OF CRACKING SHOULD BE CCNSIDERED).
B) THE EFFECTS OF RESIDUAL STRESS, AS REPORTED IN THIS RIL, MUST BE ALDED TO THE EFFECTS OF FRESSURE AND OTHER LO ADS IN CALCUL ATING Ks FOR THE FOSTUL ATED EL AW.
FRACTURE ANALYSIS AT UPPER-SHELF TEMPERATURES WILL REQUIRE CONSIDERATION ON A CASE-BY-CASE BASIS.
pr$q3BE_IIIP A C T_J F_R QQLD : THERE IS A POTENTI AI CONFLICT BETHEEN REGULATORY PRACTICE AND CURRENT PROVISIONS OF THE ASME COLE, SECTION SI, PARAGRAPH IW3-4320.
THE COMCLUSIONS OF THIS RIL WILL BE TRANSMITTED TO THE RESPONSIBLE CODE BODY WITH THE SUGGESTION THAT THE REFERENCED MATERIAL DESCRIBING THE TEST PROGRAM JUST COMPLETED BE CONSIDERED. OUR GOAL WILL BE TO OBTAIN AN ADDITICHAL REGUIREMENT IN PARAGRAPH IWB-4320 THAT CALLS FOR THE FRACTURE ANALYSIS DESCRIBED ABOVE.
CR531M114Pf2EP11: SOME EDITORIAL CORRECTICNS AND SUGGESTED CHANGES IN WC2 DING WHICH DO NOT AFFECT THE CONCLUSIONS, HAVE BLEN F0kMAkDED TO THE ORIGINATOR OF THE RIL. - - - - -
_. _ ~
r ROGR Ari 0F FIC E CCMEHis 0** P0 f EN 141 UfILIZATION OR _VALUE OF RESEAnCH kESULTS IN THE REGULATO&* PROCE$5 PIL 4 69 DATE ISSUED:
11-19-79 RES DrCISTON UNIT: SEISMIC, ENGINEERING & SITE SAFETY j
PIL TLT_LE: AN INTEGRATED GEOPHYSICAL AND GEOLOGICAL STUDY OF THE TECTONIC FRAMEWCRK OF THE 38TH PARALLEL LINEAMENT IN IHE VICINITY OF ITS INTERSECTION WITH THE EXTENSION OF THE HEW MADRID FAULT ZONE SPPNSORING OFFIEE(S):
3-1 NRC/ STATE REGIGMAL RESEAPCH PPOJECT MGR N. STEUER EARTH SCIENCES pet ???fM THIS RIL IS AM INTERIM REPORT REFLECTING INFORMATION AVAIL ABLE AS OF 1978.
THIS INFORMATION RELATES TO A STJM DES!uiED TO EEFIHE THE STRdCTU2AL SFTTING AND TECTCHIC HISTOP OF THE AREA TO REALISTICALLY EVALUATE EARTHQUAKE 1
RISKS IN THE SIT 14G OF AUdLEAR POWER PLANTS.
WHILE THESE INTERIM RESULTS ARE NOT DEFINITIVE, RES RECOMMENDS THAT THE CORRENT PRACTICE OF EXTEADIt4G THE NEW MADRID 18 81-1812 EART6 QUAKES NORTH OF THE ROUGH CREEK FAULT ZONE (38TH PARALLEL LINEAMENT) BE SONTINUED UNTIL ADDITIONAL DATA BEING DEVELOPED INDICATE THAT THIS PR ACTICE S1100LD BE CHANGED. WE ALSO RECOMMEND THAT THE INFORMATION IN NU2EG/CR-0449 BE CONSIDERED BY SD AND NRR AS INPUT TO THE DEVELOFMENT OF A TECTONIC PROVINCE OR SEIS'1IC 20hlNG MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA.
DOCUMENT ISSUED:
HUREG/CR-0449.
USER DISCUS $ICN POSITION COMMISSION ACPS PRESS OFFICE MEETING FAPER BRIEFING BRIEFING RELEASE RESULTS E%LE11JCJJV_JTIES REVIEW HELD
$0MPLETED HELD HELD ISSUED._
IMPL EMEN T,gg 0FFICE RLSPLNSI3tE......... Nak,SD SCHEDULED COMPLETION DATE..
ACTUAL COMPLETICH DATE.....
N18LC Qr#[N T '., Q((ff*MPj U 279, J. FH(Glil-DEW I bl6ttW 91fM_1.0_.LMLL A IRf PRg[g SITING IN OKLAHCMA, ICANSAS AND NEBRA%KA HAS BEEN MALE MORE DIFFICULT BY THE CONTRUWLRSY OVER IHE ASSOCIATI0H OF SEISMICITY WITH TI!E NEMAHA RIDGE. WITH EXCEPTION OF NEW EARTHQUAKE DATA, THE DATA DESCRIBED IN THE FEFORT IS NOT DIRECTLY APPLICABLE TO THE REGULATORY PROCESS. PHASE III SHOULD PROVIDE THE USEABLE DATA.
REMPRE IMERLCE_PEWLM: THE PROJECT HAS NOT BEEN COMPLETED, NCR FINAL CONCLbSIONS DRAWN.
THE GEOLOGIC Ato SEISMIC DAIA OASE HA/E BELN BROADENED. PRELIMINARY SEISMIC DATA SbaGESTS A ZONE OF EARTHQUAKES BEGINNING NEAR EL RENO STRIKING NORTHEAST AND CUTT 2NG DIAGONALLY ACROSS THE NEMAHA STRUCTURE, A CONCENTRATION NEAR WILSON AND SEVcRAL EPICENTERS PARALLEL THE NORTHERN FRCNT OF UUACHITAS.
ALSO, THE EAPTHQUAKE TO DATE SUPPDRTS STAFF CONCtUSION THAT THE LEVEL OF SEISMICITY IS LOW FOR NORTHEAST CKLAHOMA.
THE LARGEST MAGNITUDE RECORDED WITHIN 10-20 KM OF THE BLACK FOX SITE IS
.I 1.4.
THE LCW LEVEL OF SEISMICITY ALSD APPLIES TO WESTERN OKL AHOMA AND POSSIBL) THE PANHANDLE. WHEN COMPLETE, THE GEOLOGIC AND SEISMIC DATA ARE IN, HOPEFULLY, THE STAFF WILL BE ALLE TO CONCLUSIVELY DETERMINE IF THE NEMAHA CPLIFT IS A TECTCHIC STRUCTURE AND RELATE THE SEISMICITY TO THE CAUSITIVE STRUCTURE. THEREBY, MORE CONCLUSIVELY DEFINING THE SSE FOR NEW PLANTS IN THE AREA.
SQL* MEN T S. EM A P r S : NONE 1..
PROGkfM OFCICE CCMMENTS CN POTENTIAL UTILIZATION OR VALUE OF RESEAPCH RESULTS IM THE REGULATORY PROCESS PIL,9:
70 DATE ISSUFD:
11-19-79 FES DECTSION UNITY SEISMIC, ENGINEE2ING & SITE SAFETY RIL TITLEt SEISMICITY AND TECTCNIC RELATIONSHIPS OF THE NEMAHA UPLIF$ IN OK, PART II. JAN. 1979
$PONSORI'4G OFFICE (S):
HRR. SD EPfg t 3-1 NRC/ STATE REGIONAL RESEARCH PROJECT MGR N. STEUER EARTH SCIENCES l
E E S_CCP"H.!dI5 : THIS RESEARCH IS EEING CONDUCTED TO STUDY THE EARTH SCIENCE PARAMETERS OF THE NEMAHA UPLIFT AND THE ttI D00n fit 4EH1 GRAVITY AMOMALY.
THIS RIL FORMARDS THE REPORT WHICH PRESENTS RESULTS OF PHASE II 0F A 3-PHASE REPORT.
RES RECOM4;END", THAT THE It4 FORMATION CO:8TAINED IN NGREG/CR-0875 BE C3NSaDERED BY SD Al4D NRR AS INPUT TO THE DEVELOPMENT OF A TECf DNIC PL")VINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONOGING STUDIES IN THE AREA.
I'0 UMENT I'3 SUED: HUREGeCR-0875 (ALSO SEE RIL 366, 11*16/79 AND NUREG/CR-0666).
4 USER DISCUSSION POSITION COT 11SSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS f3SLEll_tCllV11]El PflifW HELD COraPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPON51 ALE.........
ARR.SD SCHEDULED COM/LETION DATE..
ACTUAL COMPLETICN DATE.....
1D._fSMUESLblMCH_ LL__1l h R. MINQfQ{t
$b EECOGNIZES ThAT IHE huREG REPORTS REPRESENT OMLY A PORTION OF A FIVE YEAR PROGRAM. NONETHELESS. THEY SHOULD 1
PROVIDE USEFUL IN7UT FOR NRC SIANDARDS DEVELOPMENT PARTICULARLY IN REGARD TO A POSSIBLE REVISION OF APPENDIX A i
TO CFR PART 800. AND THE DEVELOPMENT OF TECTONIC PROVINCE OR SEISMIC ZONING MAPS OF THE EASTERN UNITED STATES.
J i
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PROGRON OFFICE CCmENIS OV PCTENTI AL UIALIZATICH OR U4tuE OF RESEARCH RESUL15 IN THE REGULATOR't PROCE3S E.11 J s 71 DATE ISSUED
- 11-19-79 RES DECISION UNITY SEISMIC, ENGINEERING h SITE SAFETY RIL TITLES REGIONAL TECTONICS AND SEISMICITY OF EASTERN HEBRA5KA ANNUAL REPORT, JUPE 1,
1977-MAY 30, 1978 SPON50RIN'; CFFICE(5)
NRR, SD EPGs 3-1 HRC/ STATE REGIONAL RESEARCH PROJECT MOR; N. STEUER EARTH SCIENCES BE3_[p ggNTS: THE PURPOSE OF THIS RESEARCH IS TO STUDY THE EARTH SCIENCE PARAMETERS OF TME ND1AHA UPLIFT AND THE MIDCONIINENT GRAVITY ANOMALY.
KNOWLEDGE OF THESE GEOLOGICAL STRUCTURES 15 NEEDED TO DETERMINE MHETHER OR NOT THEY ARE LOCALIZERS OF EARTHQUAKES. THE INFORMATION GAINED IS OF IMPORTANCE IN THE SITING AND LICENSING OF NUCLEAR POWER PLANT!.
THIS INTERIM REPORT PRESCHTS AND INTERPRETS IMFORMATION OBTAINED BETWEEN TJNE 1,
1977 TD MAY 36, 1978.
RES RECOMMEND 5 THAT THE INFORMATION CONTAINED IN NUREG/CR-0376 BE CONSIDEREc BY SD AND NRR As INPUT TO TPE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZOMING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIL AND GUIDE FOR ONGOING STUDIES IN THE AREA.
USER DISCUSSION POSITION COMMISSION ACR$
prest CFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RCSULTS EQ1T_ELLJ CIIVIT_lfi R[ylfy HELD COMPLETED HERD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... HRR.5D SCHEDULED COMPLETION PATE..
ACTUAL CCMPLETICN DATE.....
ip_fEMENT_b MARCH 19, 19ER. MLMOGtJf s SD WECOGHIZtS THAT THE N'JREG REPORIS REPRESENT ONLY A PORTION OF A FIVE YEAR PROGRAM. NDHETHELESS. THEir SHOULD PROVIDE USEFUL INPUT FOR NRC STAEDARDS DEVEt0PMENT PARTICULARLY IN REGARD TO A POSSIBLE RF"ISION OF APPENDIX A TO CFR PART 100, AND THE DEVELOFMENT OF TECTCNIC PRGVINCE OR SEISMIC ZONING MAPS OF THE EASTERN UNITED STATES.
_m_.______<_
PRDGPAM OFFICE Co m ENTS ON POTLNTIAL UIILIZATION OR VALUE OF RESEARCH RESULTS TH THE REGULATORY PROCESS RIL 72 72 DATE ISSUED:
11-16-79 RES DECISION UNIT: SEISMIC, ENGINEERING S SITE SAFETY RIL TI TLI. : NEW ENGLAND SEISM 0 TECTONIC STUDY ACTIVITIES DURING FISCAL YEARS 1977 AND 1978 SPONSORING CFFICE(S)2 NRR. 53 ERfQs 3-1 NRC/ STATE REGIONAL RESEARCH PPOJECT MGR:
N. STEUER 4
EARTH SCIENCES i
ELS_CQTOENlS THE NEW ENGLAND SEISM 0 TECTONIC STUDY IS A 5-YEAR PROGRAM 10 STUDY THE GEOLOGY AND SEISMICITY OF NEW ENGLAND AND CONTIGUOUS AREAS TO ASSESS THE POTENTIAL SEISMIC HAZARD TO PROSPECTIVE NUCLEAR POWER PLANT SITES IN THE REGION.
PRELIMINARY RESULTS THUS FAR DOCUMENT THE PROMINENCE OF FAULTING IN THE REGION AND DEMONSTRATE THE EFFECTIVENESS OF REMOTE-SENSING. MAGNETIC-LINEAMENT AND GRAVITY-LINEAMENT ANALYSES TO REVEAL FAULTS IN THE REGION. THE REPORTS FORWARDED WITH RIL DESCRIBE THE STUDY FPCM JULY 1,
1977 TO JUNE 30, 1978.
RES RECOMMENDS THAT THE IN; 7'MATION CONTAINED IN NUREC/CR-A081 AND HUREG/CR-0930 BE CONSIDERED BY SD AND HRR AS INPUT TO THE DEVELOPMENT OF A TECiONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA.
DOCUMENTS ISSUED: NUREG/CR-0081; NUREG/CR-0930.
USER DISCCSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS ED51_P_IL_ ACTIVITIES SEVE y HELD COTTL ET ED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR,SD SCHEDULED COMPLETION DATE..
ACTU4L COMPLETION DATE.....
lid 3_COMMIN_1L JANQARY 11 193 b,_). KNJQHTt kt$iP,JAE ietLJ C Alj 0N 10 RE @ LATORf PROCE3_12 THE STUDY IS A COOPERATIVE EFFORT AMONG FEDERAL AGENCIES, UNIVERSITIES.
i AMD STATE GEOLOGICAL SURVEYS IN A PROGRAM TO STUDY THE GEOLOGY AND SEISMOLOGY OF NEW ENGLAND, WITH THE INTENT TO PROVIDE A STRONG BASIS FOR THE ASSESSMENT OF THE SEISMIC 4AZARD IN THE REGION IN ACCORDANCE WITH NRC REGULATIONS. THE SUBJECT NUREG'S ARE A
SUMMARY
OF THE RESULTS OF THE FIRST 2 YEARS ACTIVITY CF A 5 YEAR PROGRAM, THE RESULTS HAVE ADDED l
GREATLY TO THE GENF.RAL KNOWLEDGE OF THE STAFF.
I RWRIELIMPKT_QF PESUll31 THE RESULTS HAVE THUS FAR HAD NO DIRECT IMPACT ON LICENSING DECISIONS. THEY TEND, HOWEVER, TO CONFIRM PASI CONCLUSIO;45 THAT CURRENT SEISMICITY, DETECTED BY INSTRUMENTS, IS CONCENTRATED IN THE SAME GENERAL AREAS AS HISTORIC EARTHQUAKES RECORDED DURING THE PAST SEVERAL HUNDRED YEARS; AND THERE IS NO DEMONSTRATAELE CORRELATION BETWEEN SEISMICITY AND MAPPED FAULTS.
C OMt'E N T S L.R emf _P K$ : AN ADDITIONAL EENEFIT OF THIS PROGRAM IS THAT IT HAS DEMONSTRATED THE EFFECTIVENESS OF IDENTIFYING 1
F4UL35 USING REMOTE SENSING TECHNIGUES. IT IS RECCMMENDED THAT THE INFORMATION CONTAINED IN NUREG/CP-0939 AND NUREC/CR-0081 BE USED AS INPUT TO THE CONSTRUCTION OF A SEISMOTECs0NIC MAP AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES. i
-~ --
P&OGRPi gf FICE COMMENIS ON PoiENTI AL O!ILIEAVION OR VALUE OF RESEARCH RESUL15 IN THE REGULATORY PROCE55 l
RIL l' 73 DATE ISSUED:
11-16-79 RES DECISION UNIT:
.~UEL CYCLE SAFETY & ENVIRONMENTAL EFFECTS l
RIL TITLES IN VIVO C9UNTING AT SELECTED URANIUM MILLS SPONSORING OFFICE (5)?
SD (79-5)
- EL:
5-23 OCCUPATIONAL RESEARCH PROJECT MGR
.'. FOULKE l
EXPOSURE & PROTECTICN EE5 COM3EttT3 THIS RESEARCN rROJECT PROVIDED MEASUREMENTS OF THE INTERNAL DEPOSITION OF URANIUM IN THE LUNGS AND PADIUM IN IHE SKELETCN OF URANIUM MILL W3RKERS.
THE IN VIRO COUNTING MAS CONDUCTED AT THE NINE MILL SITES. NO WORKER HAD MORE THAN THE MAXIMUM PERMISSIBLE GRGAN BURDEN.
RES RECOMMENDS THAT RESULT 5 PRESE:4TED IN NUREGICR-53%1 BE USED BY SD IN DETERMINING THE VALUE OF TAKING IN VIRO MEASUREMENTS AT MILL SITES.
DOCUMENT ISSUED: NUR E G/ CR- 0341 l
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS P337_RLLA_CllHIIE1 FFJ1EH HELD GMPLETED HELD HEtp ISSUED _
IMPLEMENTED OFFICE kt5FONSIBLE......... SD SCHEDULED CCMPLETION DATE..
ACTUAL CCM?LETICN DAIE.....
m 1
PPOGetM OFFICE C05>LN1% ON 30fth11at UTILIZATION OR VALUE OF RESEARCH PESulTS IN THE REGUL ATORY PROCESS RIL 98 74 DATE ISSUED 11-16-79 RES DECISION UNITr CODE DEVELOPMENT l
RTL TITLE: STEADY-STATE FUEL ROD 3EH4VIOR CODE: FRAPCON-1 iPf450 pit lG OFFICE (512 NRR. SD P R_g 1-12 FUEL CCDE RESEARCH PPOJECT MGR G. MARINO DEVELaPMENT PE5_CStNINII: THIS RIL TRANSMITS THE DESCRIPTION AND ASSESSMENT DOCUMENTATION OF THE LATEST VERSION OF THE STEADY-STATE FUEL RCD LLHAVICR CODE - FRAPCON 1 TRAPCDN-1 IS A FORTRAN IV COMPUTER MODEL WHICH CONSIDERS THE COUPLED EFFECTS OF FUEL AND CL ADDII;G DEFORMATION TEMPER ATURE AN3 INTERNAL GAS PRESSURE ON TfiE OVERALL RESPONSE CHARACTERISTICS OF A FUEL ROD OPERATING UNDER HORMAL CONDITICNS.
THE CODE IS USEL: 1) AS A BE CODE TO INITIALIZE THE CURRENT RES BEST ESTIMATE TRANSIENT CODE: 2) AS A STAND-ALONE. BEST ESTIMATE STEADY-STATE CODE; OR 3) AS A LICENSING TOOL u!TH APPROPRIATE EM MODELS SUPPLIED BY NRR.
i USER DISCUSSION POSITION COMMISSION ACRS PRESS 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS E05LPllELEllH1 FIV_I.I.J HELD COMPLETED HELD HEL D ISSUED IMPLEMENTED OFFICE RESP 0tsSIBLE.........
hKR.SD SCHEDULED CCMPLETION DATE..
ACTUAL COMPLETION DATE.....
i 1
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PROGRAM OrFICE COPNENV3 ON POVENVIAL MTILIZATION 09 VALUE OF QESEARCH RESULTS IN THE QEGULATORY PROCESS PIL 88 75 DATE 155UED2 11-27-79 PES DECISION UNITr SEISMIC. ENGINEERING 2 SITE SAFETY PIL TITLES INVENTORY. DETECTION, AND CATALOG OF OKLAHOMA EARTHQUAKES AND EARTHQUAKE MAP OF OKLAHOMA, MAP GM-19 SPONSORING OrrICEf5):
3-1 NRC/5 TATE REGIONAL RESEARCH PROJECT MGRr N. STEUER EARTH SCIENCES fl3_CCKMENJ5: THIS RESEARCH WAS A COOPERATIVE GEOLOGIC. SEISMIC AND GEOPHYSICAL EFFORT TO STUDY THE EARTH SCIENCE PARAMFfLRS OF THE NEMAHA UPLIFT AND THE MIDCONTINENT GRAVITY ANOMALY.
A KNOWLEDGE OF THESE GEOLOGIC STRUCTURES IS OF VITAL IMPCRTANCE IN THE SITING AND LICENSING OF NUCLEAR POWER PLANTS.
PROJECT WORK IS SEPARATED INTO THREE PHASES. THIS RIL WITM ITS ENCLOSURES PRESENTS RESULT 5 OF WORK COMPLETED IN PHASE I.
RES RECOMMENDS THAT THE INFORMATION IN THE INVENTORY BE CONSIDERED BY SD AND NRR AS INPUT TO THE DEVELCPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTER 4 U.S. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES.
DOCUMENTS ISSUED: MAP GM-19; REPORT 2 INVENTORY, DETECTION AND CATALOG OF OKLAHOMA EARTHQUAKES.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS ISSUED _
IMPLEMENTED EDSI_Pll_A CIIVIT L(1 PLVLEy HELD C0t'PLFTED HELD QELD OFFICE RL5P0tiSIBLE......... SAR 5D SCHEDULED CCMPLETICH DATE..
ACTUAL COMPLETION DATE.....
EQM"_f MLS,. F43H 19, M3 0uR. MJN3 Wfs SD RECOGt4IZE5 1 HAT THE NUREG kEPORTS REPRESENT ONLY A PORTION OF A FIVE FEAR PROGRAM. HONETHELESS. THEY SHOULD PROVIDE USEFUL INPUT FOR NRC STANDf.RDS DEVELOPMENT PARTICULARLY IN REGARD TO A POSSIBLE REVISION OF APPENDIX A TO CFR PART 100. AND THE DEVELOPMENT OF TECTONIC PRGVINCE OR SEISMIC ZONING MAP 5 OF THE EASTERN UNITED STATES.
j _
l H00R AM OF FIR IO!NeNIS ON POTENTIAL UilLIZATION M VALUE OF RESEARCH RESULT 5 IN THE REGULATORY PROCESS 1
- RIL 8 76 MLr ISSUED:
12-23-79 RES DECISION UNIT 2 PRI".'ARY SYSTEMS INTEGRITY
.i RIL TITLE: ANNEALIMG OF IRPADIATED REACTOR PRESSURE VE5SELS SPON50 PING OFFICE (5): 14R R. SD ge3 1-20 VESSEL INTEGRITY RESEARCH PROJECT MSR C. SERPAN ffMMjf'LTft : THE RESE*RCH PROJECT REVIEWED THE INFORMATION DEVELOPED DVER THE PA5Y 15 YEARS ON THE USE OF POST-T I RR ADI A11014 HEAT TREATt;ENT (ANNEALING) TO RECOVIR THE PRE-IRRADI ATION PROPERTIES OF REACTOR VE5SELS OF COMMERCI AL NUCLEAR POMER PLANTS FOR CGmERCIAL SAFE OPERATION.
THE INFORMATION IS TO BE VIEWED AS PROVIDING A BALKGROUND FOR INTERPRETING CURREN1 AND FUTURE RESEAPCH ACTIVITIES AND POTENTIAL LICENSING APPLICATIONS OF VESSEL STEEL ANNEALING.
50CUMENT IS$UED: NUR E G/CP-34 f,6.
l USER DISCUSSION
- CSITION COMMISSION ACRS PRESS DFFICE MEETING PAPER BRIEFING ERIEFING RELEASE RESULTS E0iT_, UL4C11V,J TJ11 ESID4 HEt D COMPLETED HELD HELD ISSUED _
IMPLEMENTED OFFICE RL5IONSIJLE......... NN.5D SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
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PP4f,R Af1 M FICE CLf rENi s CM PalENTIAL UVILIZATIOM OR UAU!E OF RESEARCH RF M 15 IN THE REGULATORY PROCESS RIL 32 77 DATE ISSUED:
12-28-79 RES DECISION UNIT 2 SEISMCI, ENGINEERING & SITE 5AFETY RIL TITLEt ORIGIN OF SURFACE LINEAMENTS IN NENAHA COUNTY, KANSAS SPON50 PING OrFICE(5):
NRR, SD gggr 3-1 NRC/ STATE REGIONAL RESEARCH PROJECT MOR:
N. STEUER EARTH SCIENCES PI5_CD*"1ENT S : THE PURPOSE OF THIS RESEARCH WAS TO GAIN A BETTER ijNDERSTANDING OF THE SOURCES OF EARTHQUAKES THAT HAVE OCCURRLD AS AN AID TO DEVELOPING A F1 ORE RATIONAL EVALUATION OF EARTHQUAKE RISK A5 IT APPLIES TO THE SITING AND DESIGN OF HUCLEAR FACILITIES.
RES RECOMMEND 5 THAT INFORMATION IN NUREG/CR-0321 BE CONSIDERED BY SD AND NRR AS INPUT TO THE DEVELOPMENT OF A TECTONIC 1
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
USER DISCUSSION POSITION COMMISSION ACR5 PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS E95T_Rll_ACIlYLT.lE.1 EEVlIW HELD COMPLETED HELD HELD ISSU ED.
IMPLEMENTED OF FICE RESPONSIBL E......... NRR.5D SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
1D_fD55ENLL_UAlCH 19. '95.0 R. UIfGOVE
$D RLC0Gul2LS IHAI THE NUREG REPORIS REPRESENT ONLY A PORTION OF A FIVE YEAR PROGRAM. NONETHELESC THEY SHOULD PROVIDE USEFUL INPUT FOR NRC ST ANDARDS DEVELCPMENT PARTICUL ARLY IN REGARD TO A POSSIBLE REVISION CF APPENDIX A TO CFR PART 100. AND THE DEVELOPMENT OF TECTONIC PROVINCE DE SEISMIC ZONING MAPS OF THE EASTERM UNITED STATES.
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PR06P AM 0Ff ICE ComENTS ON POTENilAL OilLIZATION OR VALUE OF PESEARCH RESULT 5 IN THE REGULATORY PROCESS PIL gr 78 DATE ISSUED:
12-28-79 PES DECISION UNIT: CODE DEVELOPMENT PIL TITLE: VERTICAL LCAD5 IN MARK I CONTAINMENT TORUS SPONSORING OFFICE (5):
NRR (76-13)
PE:
RESEARCH FROJECT MGR R. CUDLIN PEyHEEbtNT0'E1ER$ r THE RESEARCH WA5 PERFORMED TO QUANTITATIVELY EVALUATE THE HYDRODYNAMIC LOAD IN THE MARK I TYPE CONTAINMENT IN OF A LOCA.
THE TESTS CONDUCTED PROVIDE CONFIRMATION OF THE ABSOLUTE VALUES AND SENSITIVITIES OF THE AIR VENTING LGADS MEASURED IN CTHER EXPERIMENTS. THE TESTS ALSO PROVIDE EVIDENCE OF A 3-DIMENSIONAL CHARACTER OF THE UPLOAD NOT BEFORE RECGGNIZED.
THE DATA IN THE SCALING L AW EXPERIMENT VERIFIED THAT SMALL SCALE TEST DATA CAN BE SCALED UP 70 OT'TYP~.at PLANT SIZE.
1 THE 1/5 SCALE TORUS TEST RESULTS SUPPLEMENTED BY THE VERIFIED SCALING L AW PROVIDE AN APPROPRI ATE YARDSTICK FOR LICENSING ASSESSMENT OF THE MARK I CONTAINMENT COHCERNING AIR VENTING LOADS.
USER DISCUSSION POSITION Co MISSION ACRS PRESS OFFICE MEETING PAPER ERIEFING BRIEFING RELEASE REST 3LTS Ej)3T FIL_ACTJU D U P EVI E14 HELD
$92tPL ETED HELD HELD ISSUED IM?tEMENTED OFFICE RESPONSIBLE......... HRR SCHEDULED COMPLETION DATE..
ACTUAL CCMPLETIDH DATE.....
1 I
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PG.05 RAM offlCE COMMENTS ON POTENilAL UffLIZATION 04 VALUE OF LESEAiCH PE50t15 IN THE REGULA?ORY PROCESS RIL #2 79 DAT' ISSUED 12-28-79 RES DECISION UNIT 2 SEISMIC. ENGINEERING A SITE SAFETY RIL TITLE EVALUATION OF SEISMIC QUALIFICATION TESTS FOR NUCLEAR POWER PLANT EQUIPMENT SPON50 PING OFFICEis)2 NRR RES2 RESEARCH P90 JECT MGR B. BROWZIM PI5_Cfd5ENTS: THE CBJECTIVE5 0F THIS TEST WERE TO SU3 JECT A TYPICAL ELECTRICAL CA3INET SPECIMEN TO A SERIES OF DIFFERENT CURkENTLY ACCEPTABLE SEISMIC QUALIFICATION TESTS. TO ACQUIRE THEREFROM DYNAMTC RESPONSE DATA AND TO PROVIDE A BASIS FOR COMPARISON OF THE TEST 5' EFFECTIVEMLSS. RES CONCLUDES THAT A REA5;E55 MENT OF THE TEST RESPONSE SPECTRUM (TRS) ENVELOPING A REQUIRED RESPONSE SPECTRUM (RR5s NEEDS TO BE AUGMENTED TO ASSURE PROPER DISTRIBUTION OF ENERGY WITH FREQUENCY DURING A QUALIFICATION TEST.
RESONANCE SEARCHES SHOULD BE CohDUCTED FOR BOTH SIMULATOR-MOUNTED AND FLOGR MOUNTED CONFIGURATIONS FOR ITEMS WHERE DYNAMIC COUPLING WITH THE SIMULATOR TUBE IS EXPECTED.
A NEW PARAMETER. THE DAMAGE 52VENTY FACTOR. HAS BEEN DEVELOPED FOR COMPARING SEVENTY OF SEISMIC QUALIFICATION TESTS.
DSF MAY MAKE IT POSSIBLE TD CPGRADE EQUIPMENT SUPPORT TO HIGHER SEISMIC EXCITATION.
j DOCUMENT ISSUEDs NUREG/CR-0345.
USER DISCUSSION POSITION COMMISSION ACR5 PRESS CFFICE MEETING PAPER BRIEFING BRIEFIXG RELEASE RESULTS Ep5LP1 LAC 11VITifi REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMFNTED OFFICE RESPONSILLE......... NkR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
HSP,_CnF ENTS. FEPPUSRI 15, 1980, Ruf4TISON 8 D.
EISEN_HU.T
- DESCRDE_MP_Lic A T IM TO Ric2LA1037 PROILM 2 THE RESEARCH RESULTS HAVEt A.
PROVILED AN I ADLPDdDLNT CONFIRMATION OF THE FOLLOWING LICENSING POSITIONSr (t) THE SINGLE FREQUENCY TEST INPUI MAY 3E SEVERE FOR WERIFYING THE STRUCTURAL INTEGRITY OF PASSIVE EQUIPMENT AND SUPPORTS. EUT MAY BE INADEQUATE FOR VERIFYING THE OPERABILITY OF ACTIVE EQUIPMENT.
(2) TEST ITEMS SHOULD SIMULATE THE ACTUAL SERVICE MOUNTING DURING THE TEST. AND DYNAMIC COUPLING WITH THE FIXTURE SNCULD BE AVOIDED.
B.
IDENTIFIED AREAS OF FUTURE RESEARCH FOR POTENTIAL ASSISTANCE IN THE LICENSING P20CE552 (1) THE DAMAGE SEVERITY FACTOR. IF FURTHER DEVELOPED, MAY BE USEFUL TO A55555 THE RELATIVE DAMAGE THAT CAN BE INFLICTED BY EARTHQUAKE TRANSIENTS OR TEST INPUTS TO STRUCTURAL COMPONENTS. CURRENTLY THE DSF 15 NOT USEFUL IN ASSESSING FUNCTIONABILITY BUT ADDIIIONAL WORK 15 WARRANTED IN THIS AREA.
(2) EXPLICIT GUIDANCE SHOULD BE DEVELCPED TO HANDLE GENERIC TESTING CONCERNS RELATIVE T02 (A) ENERGY CONTENT VERSUS FREQUENCY DISTRIBUTION. AND (B) PROPER VALUE OF THE ZPA LEVEL IN THE YEST INPUT.
D13Cflh1_JUP!sT_DE_Ef5ULTS: THE RESEARCH RESULTS ARE ESSENTIALLY CONFIRMATIVE IM NATURE. AREAS IDENTIFIED HAVE AiREADY RECEIVED SIAFF ATTENTION EVEN PRIOR TO THE BEGINNING OF THIS RESEARCH PP0 GRAM, CONSEQUENTLY THE RESEARCH RESULTS HAVE NO IM/ACT ON REGULATORY REQUIREMENTS. HOWEVER. FUTURE EFFORTS BY RES OR THE IEEd STANDARD COMMITTEE ON LiFINING CURRENT CRITERIA MAY HAVE IMPACT.
EUJOENTSLEEUAPK51 THIS RESEARCH PROGRAM HAi HAD RELATIVELY LIMITED SCOPE AND RESOURCES AND CONSEQUENTLY HAS HOI PROVIDdD DIRECTLY USALLE NEW PROCEDURES OR METHODS OF QUALIFICATION OF EQUIPMENT.
HCWEVER. IT DID SERVE A USEFUL PURPOSE IN IDENTIFYING THE POTENYIAL AREAS NEEDED FOR FURTHER INVESTI$ATION WHICH SHCULD BE CONTINUED. BASED ON THE EESEARCM RE!8tt TS. NRR HA5 TAKEN THE FOLLOWING ACTIONS:
- 1) THE IEEE STANDARDS COMMITTEE RES?ONSIBLE FOR DEVELOPING EQUIPMENT SEISMIC QUAlltaCATIOM CL8InANCE HAS BEEN INFORMED OF THE RESEARCH RESULTS. INVESTIGATIGN BY THE COMMITTEE TO DETERMINE IF ANY POSSIBLE REFINEMENT OF cuRREM! Cp!TERIA IS NECESSARY IS UNDERWAY. 23 FURTHER RESEARCH ON THE DAMAGE SEVERITY FACTOR AND THE CRITERIA TO IMPROVE THE FREQCENCY vtA505 ENrRGY DISTRIBUTION IN THE TEST INPUT SPECIFICATION THE OFFICE or *OCLFAR REGULATORY RESEARCH. 3) REVIEWERS J
CONTINUES TO BE A PECOMMENDATIDH OF NRR FCR CONSIDERATION BY HAVE BEEN INSTRUCTED OF THE CONTIt:UED NEED TO MAINTAIN A CAREFUL REVIEW OF ALL TEST InF;; rHNCTIONS USED OR PROPOSED FOR i
EQUIPMENT QUALIFICATION.
i
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PROGT.M CfFICL CO-ENT5 C4 F0fI 4TI AL UTIll2ATION OR VALUE OF RESE#RCH PESULTS IN THE REGULAfoey PROCESS RIL st 80 DATE ISSUED
SPONSOPING OFFICE (5): NRR (76-12)
REsis 5-23 OCCUPATIONAL RESEARCH PROJECT MGR2 J. FOULKE EXPOSURE & PROTECTION RE5J_QPl*jEN T3 : THE PURPOSE OF THIS PROGRAM WAS TO TDENTIFY AND QUANTIFY THE EXPOSURE REDUCTION POTENTIAL OF THE DESIGN AND OrLRAYIONAL GUIDELINES GIVEN IN REGULATORY SUIDE 8.8 AND TO ASSESS THE LOSTS INVOLVED IN IrtPLEMENTING THEM.
A THREE PA!!T T1ETHODOLOGY WAS DEVELOPED TO ASSESS DCCUPAT10t4AL EXPOSURE USAGE AT LIGHT WATER REACTORS. TO DETERMINE QUANTITATIVELY THE PGTENTIAL FOR RADIATION EXPOSURE REDUCTION, ANJ TO EVALUATE f t.E JUSTIFICATIONS FOR MAKING IMPROVEMENTS.
DOCUMENT ISSUED: NUR EG/CR-04%.
USER DISCUSSION POSITION COMMISSION ACR5 PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULT 5 E051._RILACILUU11 EEVI E'A HELD COP *PLETED HELD H EL D ISSUED IMPLEMENTED Of FICE RESI Dra%I BL E......... NRR SCHEDULED CGt'.PLETION DATE..
ACTUAL COMPLETION DATE.....
1 I
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- ____m,
m m PPOGP AN OFflCE COPMCHIS ON POIENilAt UTILIZATION OR vat 9E OF RESEAMCH MESULT5 IM THE REGULATORY Pw9CE55 RIL 9' 51 DATE ISSUED:
C2-28-80 RES DECTSION UNIT 2 FAST BREEDER REACTORS FIL TITLE: IRRADIATED FUEL DI5RUPTION UNDER LOF ACCIDENT CONDITIONS: RESULTS CF ACPR TEST SERIES FD-1 AND THE FI5 GAS COLE SPOM50 PINS OFFICE (5): NRR PFf}
- 2-6 1.CCIDENT ENERGETICS RESEAECH PROJECT MSR2 R. WRIGHT P
SUMMARIZED IN THIS RIL ARE RESULTS OF THE FUEL DISRUPTION-1 (FD-1) SERIES OF IN-REACTOR EXPERIMENTS ON li_f,Ir2iLN_15 :
THE SWELLIhG AND DISRUPTICN OF IRRADI ATED FULL UNDER THE CONDITIO;45 DF AH UNPROTECTED (N0-SCRAM) LOSS-OF-FLOW (LOF)
ACCIDENT IN AM LMTER.
ALSO PRESENTED 15 ANALYSIS WITH THE FI5 GAS IRRADIATED-FUEL FISSION-GAS-BEHAVICR CODE THAT HAS DEVELOPED FRCM THESE RESULTS.
IN THE CRBR PSAR, THE APPLICANT INVOKED AN ASSUMED FISSIGM-GAS-D2IVEN FUEL DISPERSAL AND 52EEP OUT TO ACHIEVE A MDN-ENERGETIC TERMINATION OF THE UNPACTECTED LOF ACCIDENT. THIS ASSUMPTION WAS REJECTED OY THE NRR STAFF ON THE EASIS THAT NO BACK-UP DATA WERE AVAILABLE. THE FD-1 SERIES OF EXPERIMENTS ANL-ANALYSIS WERE UNDERTAKEN TO RESQLVE THIS ISSUE, AND THE RESULT 5 SUPPORT THE STAFF POSITION. NO EVIDENCE FOR THE HYPOTHE5IZED PRE-MELTING FUEL DISPERSAL WAS SEEN, BUT SUBSTANTIAL RAPID SWELLING OF THE TEST FUEL DID OCCUR ON THE APPROACH TG MELTIhG.
DOCUMENTS 15!'JED NUREG/CR-0914; NUREG/CR-1124.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER LRIEFING BRIEFING RELEASE RESULT 5 f.011_ PIl_ASC_T_lBlHi PEVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE FESPONSIELE......... hRP.
SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
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i 1
PROM AM OFFICE ComEwis ON PoiENilot UTILIZATION OR VALUE 05 RESE8RCH REiSULTS IN THE REGULATORY PROCESS i
RIL 92 82 D ATE 255'JED4 C2-29-80 RES DECISION UNIT FUE8 lllYCLE SAFETY & ENVIRONMENTAL EFFECTS RIL TITLES SOCIAL AND ECONOMIC EFFECTS OF TMI ACCIDENT SPON5ORING orFICE(5)* NRR (79-fi')
E_Pd 5-21 SOCIO-ECONCMIC RESEARCH PROJECT MOR:
C. PRITCHARD i
IMPACTS PUqC? TJMr THIS STUDY DEALS WITH THE SOCIAL AND ECONOMIC EFFECTS OF THE ACCIDENT AT THREE (1ILE ISLAND D;JRING THE FIRSI 6 MONIHS FOLLOWING ThE ACCIDENT.
A VARIETY OF DATA SOURCE 5 WERE U11LIZF7 INCLUDING PUBLISHED DOCUMENTS AND STATISTICS, HOUSEHOLD SURVEYS, NEWSPAPER FILES, INTERVIEWS, AND OTHER RESEARCH AB0bi TFE ACCIDENT. THE FINDIP*GS CAN BE GROUPED INTO ErFECTS ON (1) THE REGIONAL ECONCMY, (2) INSTITUTIONS, AND (3) INDIVIDUALS. DIRECT ECONOMIC EFFECTS DURING THE iMERGENCY PERIOD FOLLOWING THE ACCILENT WERE INTEREUPTED LOCAL PRODUCTION AND REDUCED LOCAL INCOME A% EMPLOYMENT. LOSSES WERE CON 5*ICUOUS DURING THE FIRST WEEK OF APRIL SUT SUB5EQUENTLY VERY MINOR.
TWLRE IS NO EVIDENCE OF ANY CONTINUING INTERRUPTION OF ACTIVITY i
BECAUSE OF THE ACCIDENT.
HOWEVER, THERE IS CONCERN WITHIN THE BUSINESS COMMUNITY ABOUT THE EFFECT OF THE ACCIDENT ON THE CONTIt4Ue'D GROWTH AND DEVELOPMENT OF THE AREA.
MAJOR INSTITUTIONAL EFFECTS WERE A 5' RAIN ON THE EMERGENCY PREPAREEME55 HETWORK IN THE PREA AND AN INCREASED F'1CUS ON THE ISSUE OF THE Tri! PL ANT BY THE LOCAL POPUL ACE.
MAJOR EFFECTS CN I
INDIVIDUALS WERE THE EVACUATION ITSELF AHD INCREASED STRESS LURING THE ACCIDENT PERIOD. FOR MOST PEOPLE, THE EFFECTS OF THE ACCICENT WERE SHORT-LIVED, BUT FOR OTHERS. THE ACCIDENT HAS CAUSED A MORE PERMANENT CHANGE IN THEIR DAY-TO-DAY j
ACTIVITIES. I? WILL BE USEFUL TO YOUR STAFF IN LICENS:NG '4 EARINGS.
I j
DOCUMENTS ISSUED NUREG/CR-1093; NL?2EG/CR-1285.
i USER DISCUSSION POSITION COMMISSION ACRS PRESS l
OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULT 5 1
P35T_Rit. _ARTIVIUli PEVIEW PELD SO:*PLETED H EL D HELD ISSUED IMPLEMENTED OFFICE RELF0NSIBLE......... hRR l
SCHECULED CCMPLETICN DATE..
)
ACTUAL CcMPLETION DATE.....
1 t
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1 1
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- 104 -
1 I
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3.0 PP01ECTED MEAR-TEPM FESEARCH TtreiMATION LETTEPS (TARGET DATES ACE FISCAL YEA 25)
TEMP.
RRG S AND NAME/
NUNS.
SUBJECT /TTTtF/TMPACY TARGET DATE (DECISTON UNTT)
R9C CHATPFAN T-269 PUMP LER ANALYSIS QTR 1, 30 NONE (RISK ASSESSMENT)
T-35 MCflITCRING CF RADICIODIME FRCM CONTAINMENT QTR 2, 30 5-19 TERRESTRIAL RADICECOLCGY P. REED ACCIDENTS.
70 CETERMINE THE PRACTIC LITY (REACTOR ENVIRCKMENTAL) 0F USING AVAIL AELE CIVIL DEFENSE INSTRU -
MENTS T3 ASSESS PUBLIC HEALTH IMPACTS a2
- f4 ACCIDENTAL RELEASE OF RADICICDINE...A hDCLE*R STATICN5. TO ASSESS INSTRUMEMTA-TICN CAPA3ILITIES (PARTICLE CCLLECTICH EFFICIENCY. PARTICULATE CCNTRIEUTIGM5.
LMD INSTRUMENT RELIABILITY) BY EVALUATING A PORTA 3LE FILLD RADICIODINE CCLLECTICH SYSTEM Af;D A DETECTION SYSIEM USING A CDV-7c0 G:1 'URVEY THSTRUMENT.
T-114 STEAM GENERATOR TUEE INTEGRITY.
QTR 2, 30 1-20 VESSEL INTEGRITY J. MUSCARA (PRIMARY SYSTEMS INTEGRITY)
T-196 DYNAMIC SIMULATION CF 'LSTE/ ROCK PROCESS QTR 2, 80 6-5 HIGH LEVEL WASTE ISOLATION M. CULLINGFORD (GEOLOGICAL TEEDaACK MECHANISM MODELING)
(RISK ASSESSMENT)
T-228 SACEM SMALL ARM 5 CASUALTY EVALUATION QTR 2, 30 4-1 EFFECTIVENESS EVALUATIDM R. ROBINSON l
MODLL 15AFEGUARDS)
T-239 F ESPONSE OF RAM PACKAGES TO PUNCTURE GTR 2, 80 5-10 TRANSPORTATION SAFETY W. LAHS LNVIE04MENTS STRUCTURAL ANALYSIS (FUEL CYCLE 3 ENVIRONMENTAL RESEARCH)
T-24(
SURFACE FAOPERTIES OF ENCAPSULANTS QTR 2, 30 NONE K. KIM (MASTE MANAGEMENT)
T-247 WASTE-ROCK INTERACTIONS CTR 2, 80 NONE K. KIM i
(WASTE MANACLMENT)
T-248 SOLUBILITY OF URANINITE QTR 2, 30 NOME K. KIM (WALTE MANAGEMENT)
T-249 KIP 05ITCRY CPTION ASSESSMENT OTR 2, 30 NONE K. KIM i
(WASTE MANAGEMENT)
T-251 Gt'55 CERAMIC RADWASTE CONTAINER GTR 2, 30 NONE K. KiM EWALUATION (WASTE MANAGEMENT)
T-254 EVAlfjATICH OF THORIUM CONTENT OF HUMAN QTR 2, 30 5-23 OCCUPATIONAL EX.P05U2E J. FOULKE TISLUEi PROTECTICN (WASTE MANAGEMENT)
- 105 -
________m l
1 T-265 VISUAL IMPACT OF ALTERNATIVE CLOSED QTR 2, 80 5-21 50CIDECONOMIC If. PACTS C. PRICHARD CYCLE CDOLING SYSTEMS (REACTOR ENVIRONMENTAL) 1 T-266 50CIDECONOMIC IMPACT OF TMI QTR 2. 30 5-21 SOCIDECONCMIC ~;# ACTS C. PRICHARD l
(REACTOR ENVIRD*~ecMTAL)
T-270 FAULT TREE NAND300K QTR 2. 80 NONE i
(RI5K 'isE55 MEN 7)
{
T-272 CONTROL RCD LER ANALYSIS QTR 2, 80 WNE (RISK ASSESSMENT) i-273 VALVE LER ANALYSIS QTR 2. 30 NONE (RISK ASSESSMENT) e f
T-4 EMERGENCY PLANNING QTR 3, 30 6-1 EMERGENCY PLANNING R. ELOND (RISK ASSESSMENT)
Y-15 ASSESSMENT OF AGRICULTL' rat. LAND QTR 3, 80 5-21 SOCI0 ECONOMIC IMPACTS C. PRICHARD (REACTOR ENVIRCNMENTAL) 1 T-16 sit?ER-II RELEAu, REASSESSMENT OF WORK QTR 3. 80 2-14 SIMMER CODE T. WALKER ENERGY TO LF,CCOMMODATED IN HDCA.
(FAST BREEDERS)
T-20 GENER % REVIEW OF THE IMPACT OF CONSERVA-QTR 3. 80 5-21 SOCIDECONOMIC IMPACTS C. PRICHARD j
Tf5n. RATE STRUCTURE A LOAD MANAGEMENT (REACTOR ENVIRONMENTAL)
ALTERNATIVES ON THE NEED FOR POWER.
PREPARE A REFEFENCE DOCUMENT WHICH REVIEWS 4 SYNTHESIZE *. RECENT EVIDENCE & EVALUATES
]
FUTURE PROSPECTS REGARDING THE IMPACT OF i
ALTLEMATIVE ENEPGY CONSERVATION PROGRAMS, RATE STRUCTURE DESIGNS. LOAD MANAGEMENT i
STRATEGIES. AND Co-GENERATICN SYSTLf15 ON THE NEED FOR POWER.
T-23 RADIATION (SOURCE TERM) SIMULATOR TESTS.
QTR 3. 30 1-25 CUALIFICATION. TESTING.
R. FEIT EVALUATION (SYSTEMS ENGINEERING)
T-29 CONSEQUENCE MODEL (INDIVIDUAL SITE QTR 3. 80 6-2 CONSEQUENCE MODELING R. BLOND ASSESSMENT. FINAL MODEL UPDATE. C2AC (RISK ASSESSMENT)
USERS MANUAL. AND COMPARISDN TO SAFETY REVIEW METHODS IN TO CFR 100) 4 T-32 ELECTROCHEMICAL TEST FOR SENSITIZATION.
QTR 3, 80 1-22 CORR 05 ION J. MUSCARA UPDATES REG. GUIDE TO PRECLUDE SENSITIZED (PRIMARY !YSTEM5 INTEGRITY)
STAIhtE55 STEEL IN SERVICES.
T-33 CODE RECO*:MENDATIONS FOR MULTIPLE INTER-QTR 3. SS 1-20 VESSEL INTEGRITY M. VAGIN5 ACTING HDZZLES. PROVIDES NEW DESIGN (PRIMARY SYSTEMS INTEGRITY)
RULES FOR SUCH A5 SAFETY IN.fECTICM N0ZZLES.
- 1% -
T-34 ENVIRONMENTAL 10 DICE SPECIES EEHAVIOR.
QTR 3. 80 5-16 AQUATIC RADIONUCLIDES.
P. CEED STUDY THE PhYSICAi & BIOLOGICAL TRANSPCRT RADIDECOLOGY OF CHEMICAL FORf:5 10 RADIDIODINES RELEASED (REACTOR ENVIRONMENTAL) l TO THE ENVIRD;;;*ENT FROM AN OPER ATING NUCLEAR STATION. LETERMIME THE INFLUENCE 0F WET DEPOSITIGN (RAIN OR DEM) FOR METEOROLCGICAL MODELS 5 Ti.E ICDINE-AIR-GRASS-MILK PATH' JAYS.
PERFORM LABO2ATORY TESTS TO DETERMINE IF METHYIODICE IS DEPOSITED ON 0R455 UNLER WET DEPOSITION CONDITIONS.
DETEEMINE ENVI2ONMENTAL i
1 PATHUAYS OF TRITIUM Af*D CAPBON-14 RELEASED FROM NUCLEAR STATIONS.
1 T-44 INDEPENCFMT A55E55 MEN! RELAP4/ MOD 6.
OTR 3. 80 1-?7 CCDE APPRAISAL W. LYON ASSESS FLR BE ANALYSIS ThR39GH REFLCOD.
(CDDE DEVELOPMENT)
T-46 SOURCE TERM CORRELATION. CONFIRMS QTR 3. 80 1-13 FUEL MELT R. SHERRY CONSERVATISM OF REG. GUIDE ASSUMPTIONS (FUEL BEHAVIOR)
FOR ACCIDENT ANALYSIS.
T-57 BEACON MOD 3.
INTEGRATES SEVERAL QTR 3, 20 1-15 CONTAINMENT CODE
- 5. FABIC PACKAGES IN BEACGM INCLUDING P55 POOL (CODE DEVELOFMENT) i DYNAMICS, ICE CCNDEN5ER. AND N2 CONCEN-TRATION ANALYSIS.
i
)
T-34 AGING MODEL QTR 3. 80 1-25 QUALIFICATION TESTINC R. FEIT EVALUATION (SYSTEMS ENGINEERING) t f
T-102 PELE-IC ISSUED FOR B2R CONTAINMENT 5.
QTR 3. 30 1-15 CONTAINMENT CQDE
- 5. FABIC CALCULATION OF DYNAMIC RESPONSE OF (CODE DEVELOPMENT)
CONTAINMENT STRUCTURES WITH A CDUPLED FLUID / STRUCTURE FROGRAM.
T-136 OELAHOMA FY 77 ANNUAL REPCRT CR-C050
- TR 1. 80 3-1 NRC/ STATE REGIONAL EARTH N. STEUER SCIENCES (SEISMIC. ENGINEERING S SITE SAFETY)
T-147 A COMPREHENSIVE REPCRT WHICH WILL QTR 3. 80 3-7 SECONDARY CONTAINMENT B. Ek3WZIN RECCMMEhD CRITERIA FOR NRR LICENSING STRUCIURAE P051TICh5 CN $[ISMIC SHEAR TRANSFER (SEISMIC. ENGINEERING 1 BASED ON MEDIUM SCALE TESTIHr..
SITE SAFETY)
T-162 RESPONSE OF NUCLEAR POWER PLANT STRUCTURES QTR 3, SC 3-9 SEISMIC SAFETY MARGINS G. BAGCHI TO THREE INPUT COMPONENTS.
RESEARCH PROGRAM
~
(SEISMIC. LNGINEERING &
4 SITE SAFEfY) l T-170 NUMERICAL TECHNIQUES FOR DETERMINL1G QTR 3, 40 DYNAMIC TESTING METHODS OF J. O'BRIEN FRFCUENCIES. NOCE SHAPES AND DAM?ING OPERATING REACT 0F5 VALVES USING LOM LEVEL INPUT.
(SEISMIC. ENGINEERING &
SITE SAFETY) i
~
- 107 -
_ _ _ - _ _ _ ~ _ _, - - - - - _
_ - - _. _ _. -.. -. - _ _ _ _ _ _ _ ~ _. - _ _ _.., _ _ _ _ _ _ _ _ _ __
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I i
1 T-171 STRUCTU2AL AND MECHANICAL CCMPDNEnf QTR 3, 80 DYHAMIC TESTItG METHOD 5 OF J. O'BRIEN TEST TECHNIGUES.
SPERATING REACTOR 5 (SEISMIC, INGINEERING 8 i
SITE SAFETY) j T-179 DESCRIPTION OF A DESIGN CONCEPT FOR QTR 3, 80 MECHANICAL ENGINEERING J. BURN 5
+
CALCUL ATING PROEABILITY OF FADICACTIVE (SEISMIC, ENGINEERING &
RELEASE. CORE MELT, SAFETY SYSTEM.
SITE SAFETY)
STRUCTURAL AND ConPONENT FAILURE.
PROBABILITIES FR0ft A SET OF NUCLEAR PLANT SEILMIC RESPONSES.
4 2
l T-189 POSITRCN At;NIHILATION - EVALUATION QTR 3, SS 2-4 CORE MELT AND CONTAIKMENT T. WALKER i
AS AN NLE PPCCEDURE.
EVALUATION OF INTEGRITY j
ELECTRCh5 AT DEFECT SITES EY POSITRONS (FAST EREEDERS)
'i AND CORRELATION WITH CREEP-FATIGUE DAMAGE.
T-191 PRCMPT-EURST ENERGETICS EYrERIMENTS QTR 3, 30 2-6 ACCIDENT ENERGETICS R. WRIGHT WITH CARBICE FUEL, ACRR TESTS 55-1, (FAS BREEDERS)
SG-2, AND SG-3.
T-215 ISEM ADVERSARY SEQUENCE EVALUATION MODEL QTR 3, 80 4-1 EFFECTIVENESS EVALUATION R. ROBINSON (SAFEGUARDS) 4 T-218 CRINC4N TRANSPORTATIDH NETWORK MODEL QTR 3, 30 4-1 EFFECTIVENESS EVALUATION R. ROBINSON
.(SAFEGUARDS)
T-221 FIXED-SITE NEUTRALIZATION MODEL (F5KM)
QTR 3, 80 4-1 EFFECTIVENESS EVALUATION R. ROEINSON (SAFEGUARDS) 1 T-223 BWR FACILITY CHARACTERIZATION CODE QTR 3, 80 4-1 EFFECTIVENESS EviLUATION R. ROBINSON i
l (SAFEGUARDS)
T-224 SOURCE AMBUSH SIMULATION MODEL 1 USER QTR 3, 30 4-5 MEASUREMENTS 8 STANDARD 5 R. ROBINSON MANUAL DOCUMENTATION (SAFEGUARDS)
T-225 PWR FACILITY CHARACTERIZATION CODE CTR 3, SS 4-1 EFFECTIVEME55 EVALUATION R. ROBINSON (SAFEGUAR4%)
T-227 ADVERSARY PATH SELECTION METHODS OTR 3, 80 4-t EFFECis ?ENE55 EVALUATION R. ROBINSON (SAFEGUARDS)
T-225 SAFEGUARDS HETWORK ANALYSIS PROCEDURE SYR 3, 80 4-1 EFFECTIVENESS EVALUATION R. ROBINSON
($ NAP)
(SAFEGUARDS)
_T-233 CENERIC METHOD FOR MATERIAL ACCOUNTING QTR 3, 30 4-2 THREAT ASSESSMENT R. SHEPARD METHODOLOGY (SAFEGUARDS)
T-234 GENERIC COST-EFFECTIVENESS MODEL QTR 3, 30 4-2 THREAT ASSESSMENT R. SHEPARD j
(SAFEGUARDS)
J T-2*7 CRITICALITY SAFETY METHODS - LOLID ANGLF QTR 3. 80 5-7 CRITICALITY 5AFETY STUDIES D. SOLBERG
& SURFACE DENSITY (FUEL CYCLE & ENVIRCNMENTAL i
RESEARCH) 1 l
- Ina -
l
_ ___ __ _ _ _.___ _ _ _.___ _ -,=. - - _.. _ _ = _ _ - - _ - _ _ _ _ _
T-240 PARTICLE LEAK STUDIES QTR 3, 30 5-12 TRANSPG2TATIG3 5AFETY W. LAMS PROGRAM (FUEL CYCLE 2 ENVIRCKMENTAL RESEARCH)
T-241 MCDELING NDRMAL SH3CK AND WIBRATION QTR 3. 80 5-10 TRANSPORTATION SAFETY W. LAHS j
ENVIRONMENT STRUCTURAL ANALYSIS (FtlEL CYCLE & ENVIRONMEDTAL RESEARCH) 7-241 MEASUREMENT TECHNOLOGY FOR PLUTCNIUM QTR 3, 80 5-23 OCCUPATIONAL EXPOSURE J. FCULKE PROTECTION (FUEL CYCLE & ENVIRONMENTAL RESEARC8) 1 l
T-244 ASSE!5 MENT OF RE!?IRATORY PROTECTION QTR 3, 80 5-23 OCCUPATION AL EXPG'aURE J. FOULKE l
5(STEMS PROTECTION l
(FUEL CYCLE & ENVIRONMENTAL j
RESEARCH)
J j
T-255 EXREM COMPUTER CODE GTR 3, 30 5-24 RADI0 BIOLOGY & DOSIMETRY J. FOULKE (REACTOR ENVIRONMENTAL) i T-258 PATHGGEMIC AMOEBAE IN CLOSED CYCLE QTR 3, 30 5-18 ECOLOGICAL IMPACT FROM J. FOULKE COOLING TOWER 5 REAC*0R 4 FUEL CYCLE 4
i FACILITIES (REACTOR EMVIRONNENTAL)
T-263 BNL EMERGENCY AIR SiiPLER FOR RADICIODINES QTR 3. 80 5-17 ENVIRONMENTAL MONITORING P. REED (REACTOR ENVIRONMENTAL) i T-264 REVIEW OF 00NSERVATION. LOAD MANAGEMENT, QTR 3, 80 5-21 SOCIDECONOMIC IMPACT 5 C. PRICHARD RATE RESTEUCTIVE AND C0 GENERATION (REACTOR ENVIRONMENTAL) 7-26/ ASSESSMENT OF STATE-0F-THE-ART OF MECil.
QTR 3. 80 NONE J.
BURNS SYS. AND CGMP.
($EISMIC. ENGINEERING 4 SITE SAFETY) i T-274 DIESEL LER ANALYSIS OTR 3, f3 NONE (RISA ASSESSMENT)
T-273 HUMAJ FACTORS ANALYSIS QTR 3, 80 NONE (RIS% ASSESSMENT)
T-276 FIRE DATA ANsLYSIS QTR 3, 80 NONE (RISK ASSESSMENT)
)
- T-277 PENETRATION LER ANALYSIS QTR 3, 80 NONE l
(RISC ASSESSMENT) i T-283 SUS-SURFACE PROFILING OF THE BEATTY.
QTR 3, 80 NONE C. JUPITER NEVADA SHALLOW LAND NUCLEAR WA5'E BURIAL
~
(WASTE MANAGEMENT)
SITE.
T-285 FORECASTING ELECT D[MANDS BY STATE /
QTR 3, 80 NONE C. JUPITER
~
UTILITY SLRVICE AREAS (REACTOR ENVIRONMENTAL)
I j
- 109 -
I l
I T-286 HPS C0%5fPUCTION-LABOR FORCE MIGRATION QTR 3, 30 NONE C. JUPITER j
1 RESIDENTIAL CHOICE (REACTOR ENVIRONMENTAL)
T-12 CRACK AFFEST TEST METHODOLOGY. BAshi QTR 4, 80 1-20 VESSEL INTEGRITY M. VAGINS fCR SEVEPE ACCIDEN! ANALYSIS ARREST (PRIMARY SYSTEMS INTEGRITY)
EVALUATION.
T-18 FISSION CAS RELEASE MODEL VERIFIED. ADDS OTR 4, 30 1-9 FI5SION PROJUCT RELEASE G. MARIN0 TRANSIENT FI5SION GA5 RELEASE QUANTI-(FUEL BEHAVIOR)
FICATION TO FRAP-T ENHANCING ACCIDENT CALCULATICMS.
T-25 WRAP EM/EER - INCORPORATES INTEGRATED
' 0TR 4, 80 NONE L. SHOTKIN AND AUTOMATED IMPROVEMENTS INTO A (CODE DEVELOPMENT)
LICENSING CODE PACKAGE.
I T-38 REWET CORRFLATION. SUMMAPIIES CURRENT OTR 4, 83 1-3 PWR-BDHT Y. H5U S
FREDICTIVE CUALITIES FOR ' RETURN OF 1-4 EWR-BDHT Y. HSU NUCLEATE BOILING'.
(SYSTEMS ENGINEERING) i T-42 TRANSIENT CHF CORRELATION. NO CHANGE QTR 4, 40 1-3 PWR-BDHT Y. HSU i
IN APPENDTX K EUT MAY REPLACE CURRENT 1-4 ADVANCED SYSTEM CODE Y. H5U j
CORRELATION.
(SYSTEMS ENGINEERING)
I-45 MECHANICAL PROPERTIES REPORT.
PROVICES QTR 4, 30 1-8 ZIRCALOY CLADDING M. PICKLESIMER INDEPEHLENTLY VERIFIED IRRADIATED (FUEL SEHAVIOR)
ZIRCALDf DATA FOR USE IN LICENSING CALCULATIONS.
T-48 8X8 TEANSIENT CHF TESTS COMPLETE. SUPPORT OTR 4, 80 1-4 BWR-BDHT W. BEC"NER EXISTING CDPRELATION OR RECOMMEND A (SYSTEMS ENGINEERING)
NEW COPRELATION.
T-67 17 X 17 REFLOOD HEAT TRANSFER.
OTR 4, 30 1-5 REFLOOD HEAT TRANSFER E. DAVIDSON IMPROVED HEAT TRANSFER CORRELATIGNS (SYSTEMS ENGINEERING) i TOR UNBLOCKED BUNDLES.
T-70 EMBRITTLEMENT tRITERIA. MAY REPLACE QTR 4, 30 1-8 ZIRCALOY CLADDING M. PICKLESIMER 4
EXISTING CRITERIA IN APPENDIX K.
(FUEL BEHAVIOR)
T-74 RELAP 4 MOD 7.
USER CCNVENIENT OTR 4. 80 1-16 REFERENCE SYSTEM CODE W. LYON PWR LOCA C00E (CDDE DEVELOPMENT)
T-82 EXTINGUISHING SYSTEM EVALUATION QTR 4, 80 1-23 ELECTRICAL STANDARDS R. FEIT AND FIRE PROTECTION l
(SYSTEMS ENGINEERING)
"T-88 COMMIX CODE (TWO-PHASE).
REASSESSMENT OTR 4, 80 2-13 FAST REACTOR SYSTEMS CODES R.
CURTIS OF COOLABILITY OF SUSA55EMBLY UNDER AND ACCIDENT ANALYSIS
{'
NATURAL CONVECTION.
(FAST BREEDERS) 1
- 110 -
1 T-100 ALTECuaTE [CCS SYSTEMS QTR 4, St CONE W. LANNING (5YSTEM.' ECGINEEQICC)
T-105 K-FIX (3D). TW3 PbASE FLOW IN IDFALIZED QTR 4, 80 1-14 ADVAh*.ED CCDE L. SHOTKIN GEOMETRIC 5.
(CODE DEVELOPMENT) j T-107 IRAC PtA.
QTR 4, SS 1-14 ADVANCED MDE L. SHOTKIN (CODE DEVELOPMEM.'S 1
T-tC8 CLADDING CREEPDOWN. IM-PILE CREEPCOWN QTR 4, 80 t-8 ZIRCaLOY CLADDING M. PICKLESIMER OF ZIRCALOY CL ADDING UNDER 40MPRE55IVE (FUE*, BEHAVIOR)
LOADS.
T-It1 MELT CONCRETE INTERACTIONS. UPDATE TO QTR 4 Jo 1-13 TUEL MELT R. SHERRY RIL 28.
DESCRIPTIGN OF CDRCCN CODE.
(FUEL SEHAVIOR)
T-ft6 BUILDING MAKE DIFFUSICN. WIND TUNNEL QTR 4, 80 3-4 SEVERE ST3RMS R. ABBEY MODEL ST3 DIES.
(SEISMIC, ENGINEERIhG S SITE SAFETY)
+
T-st7 BUILDING WAKE DIFFU5104 FIELD TEST QTR 4, 80 3-4 SEVERE STORM 5 R. AF3EY STUDIES.
(SEISMIC. ENGINEERING S SITE SAFETYs T-121 ASSESSMENT AND EXPANSION OF STRONG QTR 4, 80 3-2 CEOLOGY $ SEISMIC R.
BRAZEE GROUND MOTION DATA.
BASIC INPUT TO CHARACTERISTICS SEI5MIC RISK ASSESSMENT.
(SEISMIC. ENGINEERING &
SITE SAFETY)
]
T-163 EFFECT OF STRUCTURAL DAMPING ON NUCLEAR QTR 4, 80 3-9 SEI5MIC 5AFETY MAPGINS C. BAGCHI POWER PLANT STRUCTURES.
TNIS 15 ANOTHER RESEARCH PROGRAM i
STUDY BY EXPANDING LLL/&DR SFISMIC (SEI5MIC. ENGINEERING I l
CONSERVATISM PROGRAM 'O TYP. CAL NUCLEAR SITE SAFETY) j POWER PLANT STRUCTURES (ZION STATION).
l T-164 CENERAL STRUCTURAL BUILDING RESPONSE QTR 4, 80 3-9 SEISMIC SAFETY MARGINS G. BAGCHI ANALYSIS REVIEW WITH SPECIAL EMPHASIS RESEARCH PROGRAM i
ON DA!; PING AND NONLINEARITY.
(SEISMIC, ENGINEERING S l
SITE SAFETY) 1 T-194 DEBRIS BED COOLABILITY AT H'GH DECAY-HFAT QTR 4, 30 2-2 POST ACCIDENTAL HEAT R. WRIGHT J
POWERS. ACRR TEST D-4 REMOVAL (FAST BREEDERS)
T-198 255 METHODOLOGY APPLICATION PROGRAM QTR 4 TO 6-6 LWR RISC ASSESSMENT M. CUNNINGHAM (RISK ASSESSMENT) i
~T-207 RISK ASSESSMENT OF HICH LEVEL MASTF QTR 4, 30 6-5 HIGH LEVEL Wr*TE ISCLATION M. CULLINGFCRD l
ISOLATION IN BECDED SALT (RISK ASSESSMENT)
T-217 COPS LLTA ROUTE DISTRIBUTION MODEL STR 4, 80 4-5 MEASUREMENTS A STANDARDS R. ROBINSCM (SAFEGUARDS)
T-2*9 ST ANDARDTZED LWR CENERIC FAULT TREE QTR 4, 80 4-1 EFFECTIVEMESS E?ALUATION R. R J51N50N CHARACTERIZATION CODE (SAFLGUARDS)
I
- 111 -
i
. - _. ~, _..
~. -
=
T-22C S AFEGUARD5 ENGINEERING AND ANALYSIS DATA QTR 4, 80 4-1 EFFECTIVENESS EVAIUATION R. R031N50N BASE (5EAD)
(SAFEGUARDS)
T-222 SAFE 1 SEMI-AutuMATED SYSTEM EVALUATICW QTR 4, 80 4-1 EFFECTIVENESS EVALUATION R. RCBIN$0N METHOPOLOGY
( S A FE G'f A2DS)
T-226 SKIdMI5H/AMBUSN BOARD CAMES QTt 4, 80 4-5 MEASUREMENTS & STANDARDS R. ROBINSON (SAFEGUARDS) 1 T-230 SAERES II CONVOY ENGLCEMENT MODEL QiR 4, 80 4-5 MEASUREMENTS & STANDARDS R. ROBINSON (SAFEGUARDS) 1 T-231 EARS COMMUNICATIONS MODEL QTR 4, 80 42 MEASUREMENTS S STANDARDS R. ROBINSON i
(SAFEGUARDS) 1 T-235 ATTRIBUTES OF THE INSIDER ADVERSARY QTR 4, 80 4-2 THREAT ASSESSMENT R. SHEPARD (SAFEGUARDS) 7-238 CUIDANCE FOR AOMINISTRATIVE CONTRDL OF QTR 4, 80 5-7 CRITICALITY SAFETY STUDIES D. SOLBERG CRITICALITY SATETY (FUEL CYCLE & ENVIROMMENTAL i
RESEARCh)
T-242 sI5EtUILIIRIUM OF URAMIUM ORE DAUGHTERS QTR 4, 80 5-23 OCCUPATIDNAL Ex.'05URE J. FOULKE i
IN CEE DUST PROTECTION
(
(FUEL CYCLE & ENVIROKSENTAL RESEARCH)
)
T-252 ENGINEERING EVALUATIDM OF LOW LEVEL QTR 4, 80 NOME E. HELD i
MASTE DISP 05At 511E5 (WASTE MANAGEMENT) l T-25T BURIAL GPOUND SIT 2 SURVEY - KENTUCKY QTR 4, 30 NONE E. HELD (WASTE MANAGEMENT)
I T-256 ALARA DESIGN OBJECTIVE 5 FCR IMR QTR 4, 80 5-23 OCCUPATIONAL EXPOSURE J. FOULKE j
PROTECTION (REACTCR ENVIRONMENTAL)
T-257 PEDIATRIC PHANTCMS QTR 4, 80 5-24 RADIORICLOGY Z DOSIMETRY J. FOULKE (REACTOR ENVIRONMENTAL)
T-260 EFFECTS OF AIRBORNE CONTAMINANTS CM QTR 4, 80 5-6 FUEL CYCLE FACILITIES D. SOLBERG j
ACTIVATED CHARCOAL EFFLUENT COMs20L j
(REACTOR ENVIEONMENTAL)
's T-261 ASBESTOS IN CCOLI?4G TCWER WATERS QTR 4, 80 5-15 CHEMICAL IMPACTS ON P. REED AQUATIC ENVIRONMENT (REACTOR ENVIRCNMENTAL)
T-262 MODEL TO PREDICT CHLORINE CONCENTRATICMS QTR 4, 80 5-14 PNYSICAL TRANSPORT SURFACE P REED IN POWER STATION DISCHAkGES WATERS j
(REACTOR ENVIROMMENTAL)
T-268 INSIGHT 5 FROM UDR SEI5MIC TESTS QTR 4, 80 NOME J. O'BRIEN i
(SEISMIC. ENGINEERING &
SITE SAFETY)
]
T-271 CCMMON CAUSL MODELING QTR 4, 80 NONE i
(RISK A55E55MENI)
- 117 -
l 1
i n
~,,-- - =,,
c-,...----~r.
w-~
1 T-279 SMALL 3REAK TEST L3-1 AND 13-2 QTR 4, 30 1-1 LOFT i
3 (LOFT)
T-280 APPLICATION cF SAI COMMc?!ICATICN5 MODEL MR 4, 80 NONE (SAFEGUARD 5) 1 T-281 CA5 SCINTILLATICN PPDPORTICWAL COUNTER QTR 4, 80 NONE J. FOULKE (FUEL CYCLE 8 ENVIRONMENTAL RESEtRCH) i T-286 FLOOD CCCURRENCE MODELLING QTR 4, 30 NCHE (RISK ASSESSMENT)
I j
T-t REPORT ON WASTE MANAGEMC4T RI5K QTR t, 81 6-5 HIGH LEVEL WASTE ISCLATION H. CULLINGFORD ASSESSMENT METHODOLD0Y DEVELCPMENT T-28 CCRE MELTDOW SENSITIVIT Y STUDY QTR t, 81 6-6 LMR RISK ASSESSMENT M. CUNMINCAAM T-43 TRAr '
MPPobED PTA.
4TR 1, 31 1-16 ADVANCED SYSTEM CODE L. SHOTKIN l
A Dv a.
' T AND PPR COOLANY SYSTEMS CODE (CODE DEVELOPMENT) j TO Es.
.aN MARGINS OF SAFETY.
T-60 CCMPRENe'95%YE RIPORTS W 4ICH WILL STR t, 31 3- ; SECONDARY CONTAINMENT B. EL 2IN SUP.P.APIZE FESULT5 DF LARGE SPECIMEN STRUCTURAI T EST ING ON SEI',111C SHEAR TRANSFER.
(SEIS91C, ENGINEERING &
SITE SAFETY) l T-75 t'ROBABILITY AND CONSECUEtCES OF STEAM QTR 1, 81 1-13 FUEL MEl.T W. JOHNSTON EXPLO5 IONS.
70551ELY WILL PROVIOE A (FCEL BEHAVIOR)
CUANTITATIVE PEEDICTION CF STEA:t EXPLO5 IONS MOSTLY APPLICABLE T0 l
ft0ATING PLANTS.
t T-77 HYDP0 ELASTIC EFFECTS STUDIES.
FCR P55 QTR 1,
St 1-15 CONTAINMENT CODE t.
SHOTKIN (rRf550RE SUPPRESSIGN SYSTEMS) CDNTAIN-(CODE DEVELOP".ENT) i MENTS I
j T-83 DETECTICM SYSTEM EVALUATICM QTR
- t. 81 1-23 ELECTRICAL STANDARDL5 R. FEIT AND FIRE PROTECTION (5YSTEMS ENGINEERING) l' T-85 DATA BASE.
IN.DRPORATF5 RFACTOR DESIGN QTR t, $1 NOME
- 5. FASIC AND GPERATING DATA BASE 5 IN RE1AP5.
(CODE DEVELOPMENT) j i
T-89 PIPING COMPONENTS UNCER CREEP AND QTR 1, 81 2-3 PRIMARY SYSTEM INTLGRITY T. J. MALKER l
PLASTIC DEFCFMATICid FINITE ELEMENT AN ALYSIS OF FIPING COMPONENTS AND
~
COMPARISON WITW EXPERIMENTAL RESULTS.
T-97 UP ANIt'M OXIDE AEROSOL PROPERTIES AND OTR 1, 31 2-9 LARGE AEROSOL TRANSPORT M. SILBER 8ERC A
1;5PP TEST AT*t050L CODE VERIFICATION.
TESTS DATA USED TO bERIFY INFCEMASIGN IN HAARM-3.
I I
i l
i 113 -
I
1 1
T-st5 EVALUTE.7ETHOD TO CALCULATE COASTAL STR 1, 31 3-8 ATM05PHE21C T2005FC0Y &
C. ABBEY DI5PEPSION, MAY IMPACT ON LICENSING DEPOSITION EVALUATICN OF CLOSE-EED ALCIDE4TAL (5EISMIC. ENGJNEERING S j
RELEASE C0hCENTRATION.
SITE SA?ETY)
T-118 ULTIMATE HEAT SINK VERIFICATION EyP.
QTR t, 81 3-8 ATM3 SPHERIC TRANSPORT AND R. ABBEY DEPOSITICN
(%E!SMIC. ENGINEERING &
i SITE SAFETY)
T-119 SOUTHEAST SEISMIC NETW3RE. NISTCRY CF QTR 1, 31 3-2 GEOLOGY AND SEISMIC R. BRAZEZ INSTALLATION OPERATION RESEAECH.
2-1/2 CHARACTERISTICS YEAR SEISMIC ACTIVITY tunMARY AND RESUME.
(SEISMIC, ENGINEERING &
SITE SAFETY)
T-132 ihTLRIM REPORT STUDY RESULTS. PPELIMINARY QTR 1, 31 3-1 NRC/ STATE REGIONAL CARTN N. STEUER l
i 17CTONic PROVINCE 33UNDARIES, EEGIONAL SCIENCES l
GEOLOGY, SEISMOTECTCNICS. TO ACCELER.!YE (SEISMIC, ENGINEERING 2 LICENSING.
SITE SAFETY) l l
T-148 COPPRLHENSIVE REPORT CN CUALTFICATION QTR t,
St 3-7 SECONDARY CONTAINMENT B, SROMZ1N TEST NELLED FDn RECULATORY GUIDES VERIF.
STRUCTURAL (SEI5MIC. ENGIN(ERING &
SITE 5AFETV)
T-14*
A RTPCRT TCR CRIENTATICN FCR LICE %5ING QTR t, 31 3-7 SECONDARY CONTAINMINT B. BROMZIN UN 1HE EFFECT CF SEISMIC DESIGN LEVEL ST40CTURAL ON THE NPP COST.
(SEISMIC. ENGINEERING 1 SITE SAFETY)
I t
Y '- 8 5 0 INTE51M *EPORT ON WAVE NEIGHTS.
STR t, 81 3-7 SECONDARY CONTAINMENT
- 3. BRCMIIN MEASUROME NT ME"t0D AT PROB ABL E MAXIMUM STRUCTURLL HURRICANES (PMH).
(SEISMIC, ENGINEERING S SIIE 5AFETY)
T-tSt A COMPREHEN5*VE REPORT WNICH WItt QTR 1, 31 3-7 SECONDARY CONTAINMENT
SITE SAFETY I
T-152 COMPREHENSIVE REPCRTS WHICN SUMMARIZE QT2 t, St 3-7 SECONDARY sJNTAIMMENT
- 3. SROMZIN j
ANAlfi1 CAL TNVESTIGATIONS OF TEST RESULT 5 STRUCTURAL ON LARGE AND MEDIUM SCALE SPECIMLNS.
(SEISM'C. ENGINEERING L f
SITE SAFEYY)
]
4 i
J 9
l
)
- 118 -
1 1
T-153 A SERIES OF FORCE-TIME HISTORIES CHAR-QTR 1, 31 NONE J. COSTELLO l
ACTERIZING AUTCMDBILE IMPACTS AT DIFFERENT (SEISMIC, ENGINEERING &
j VELOCITIES WILL BE PRESENTED TO ACCOUNT SITE SAFETY)
FOR TORNADO RISK REGIONS AND VEHICLE ORIENTATIONS AT IMPACT.
ENVELOPING FORCE-JTME HISTORIES WHICH CAN BE USED FOR DESIGN 4ad%T TORNADO. MISSILE EFFECTS IN EACH TsiNAL3 REGION WILL BE RECOMMENDED.
l T-154 AN ASSESSMENT AND DESCRIPTION OF AVAILABLE QTR 1, 81 3-9 SEISMIC SAFETY MARGINS J. COSTELLO METHODS UNDER DEVELOPMENT TO ANALYZE THE RESEARCH PROGRAM 3
SOIL STRUCTURE INTERACTION PUENOMENON WILL (SEISMIC. ENGINEERING A I
BE DEVELOPED BY A NUMBER OF EXPERTS. THE SITE SAFETY)
ASSESSMENT WILL INCLUDE THE RANGE OF APPLICABILITY, THE ADVANTAGES AND THE DIS-j ADVANTAGES OF EACH TECHNIQUE, WAVE PASSAGE 1
EFFECTS, 3D CHARACTER OF THE PROBLEM, NONLINEAR BEHAVIOR OF THE SOLID AND LIMITATIONS OF AVAILABLE CCMPUTER PROGRAMS.
)
T-tS5 SENSITIVITY STUDY OF SOIL-STRUCTURE QTR 1,
31 3-9 SEISMIC SAFETY MARGINS G. BAGCHI INTERACTION PHENOMENON FOR SOIL-RESEARCH PROGRAM STRUCTURE INTERACTION, SDIL PROPERTIES (SEISMIC. ENGINEERING 3 AND SOIL CONFIGURATION, WAVE PASSAGE SITE SAFETY)
AND AZIMUTH EFFECTS FOR ZION NUCLEAR POWER STATION USING CONTINUUM ANALYSIS APPRDACH.
j T-156 SUBSYSTEM RESPONSE REVIEu REPORTS QTR t,
El 3-9 SEISMIC SAFETY MARGINS G. BAGCHI ASSESSING THE STATE-OF-THE-ART OF SUB-RESEARCH PR9 GRAM SYSTEM RESPONSE DETERMINATION, ACCURACY, (SEISMIC. ENGINEFRING &
AND UNCERTAINTIES.
SITE SAFETY)
T-tS7 SOIL STRUCTURE INTEWACTION (SSI) REVIEW QTR 1,
31 3-9 SEISMIC SAFETY MARGINS G. BAGCHI REPORTS ASSESSING THE STATE-OF-THE-ART RESEARCH PROGRAM OF SSI ANALYSIS METHODOLOGY, ACCURACY, (SEISMIC. ENGINEERING 1 UNCERTAINTIES, AND ITEMIZING BENCHitARK SITE SAFETY)
PR03LEMS.
T-1:'
1?4SITIVTi t STUDY INVESTIGATING THE QTR 1,
81 3-9 SEISMIC SAFETY MARGINS G. BAGCHI cF FECTS OF UNCERTAINTIES ON SUBSYSTEM RESEARCH PROGRAM REv'2 ;C.
(SEISMIC, ENGINEERING A SITE SAFETY)
T-159 SFXSITIVITY STUDY ON NONLINEAR ANALYSIS QTR 1, 81 3-9 SEISitIC SAFETY MARGINS G. BAGCHI OF A KEY SAFETY SYSTEM.
RESEARCH PROGRAM (SEISMIC. ENGINEERING S SITE SAFETY)
!-15G SEMSITIVITY STUDY COM?ARING LINEAR QTR 1, 81 3-9 SEISMIC SAFETY MARGINS G. BAGCHI FINITE ELEMENT ANALYSIS WITH C097INUUN RESEARCH PROGRAM ANALYSIS APPROACH FOR ZION NUCLEAR PO%ER (SEISMIC. ENGINEERING &
STATION.
SITE SAFETY)
T-161 COMPARATIVE ANALYSIS OF ZION REACTOR QTR t, 81 3-9 SEISMIC SAFETY MARGINS G. BAGCHI CONTAINMENT BUILDING USING LINEAR FINITE RESEARCH PROCRAM ELFMENT ANALYSIS CONTINUUM ANALYSIS (SEISMIC. ENGINEERING &
APPROACH, AND NONLINEAR FINITE ELEMENT SITE SAFETY)
ANALYSIS.
- 115 -
~
n T-166 MATHEMATICAL MODEL FOR PREDICTCH OF QTR 1, 31 3-7 SECONDARY CONTAINMENT B. BROWZIH NEAR SHORE TSUNAMI FOR ESTABLISHING A STRUCTURAL LICENSING METHOD.
CSEISMIC. ENGINEERING &
SITE SAFETY)
~
T-568 SRP METHODOLOGY CCMPARISDNS. THE QTR t, 31 3-9 SEISMIC SAFETY MARGINS J. COSTELLO CURRENTLY USED (SRP) METHODOLOGY WILL BE RESEARCH PROGRAM COMPARED TO OTHERS IN THE Soll-STRJCTURE (SEISMIC. ENGINEERING 2 INTERACTION FEVIEW. SYSTEMATIC DIFFERENCES SITE SAFETY)
WILL BE EXPLORED AND THEIR SICNIFICANCE ASSESSED.
T-173 DOCUMENTATION OF STATE OF AVAILABLE QTR 1,
88 MECHANICAL ENGINEERING J. BURNS FRAGILITIES RELATED INFORMATION.
(SEISMIC. ENGINEERING &
SITE SAFETY)
T-180 OPERATING SEISMIC SAFETY ANALYSIS QTR 1, 81 MECHANICAL ENGINEERING J. BURNS CODE - SEISMIM.
(SEISMIC. ENGINEERING &
SITE SAFETY)
T-184 ACCURACY OF FINITE ELEMENT METHOD AND 1, 81 STRUCTURAL ENGINEERING C. BURGER LUMPED MASS MEIHOD IN SEISMIC ANALYSIS.
(SEISMIC. ENGINEERINb &
SITE SAFETY) j T-185 STUDY OF SEISMIC ANALYSIS MODELING QTR 1, 81 STRUCTURAL ENGINEERING C. BURGER TECHNIQUE IN VERTICf.L DIRECTION. THE (SEISMIC. ENGINEERING &
i STATE-OF-THE-ART IN VERTICAL MODELING SITE SAFETY) i TECHNIQUE IS EITHER TO ASSUME RIGID FLOOR OR TO SIMULATE THE FLOOR FLEXIBILITY BY CNE OR SEVERAL SINGLE DGR SYSTEMS. THIS STUDY IS TO EXAMINE THE ACCURACY OF THIS ASSUMPTION AND TO EXPLORE BETTER l
ALTERNATIVES.
l T-193 FUEL FRAGMENTATION TESTS WITH MOLTEN QTR 1,
81 2-2 POST ACCIDENT HEAT REMOVAL R. WRIGHT RESULTS OF LARGE SPECIMEN TESTING ON (FAST BREEDERS)
SEISMIC SHEAR TRANSFER.
FRAGMENTATION AND THE FUEL-COOLANT INTER-ACTION POTENTIAL FROM THE CONTACT OF THE MOLTEN FUEL METAL THERMITE WITH SODIUM.
l T-197 HUMAN ERROR RATE ANALYSIS QTR 1, 31 6-4 RISK ASSESSMENT FOR M.
CULLINGFORD FUEL PROCESSING (RISK ASSESSMENT) l T-200 ANALYSIS & PREEICTION OF MAJOR FLOODS QTR 1,
81 6-3 LIMITING CONDITIONS FOR W. VESELY OPERATIONS j
(RISK ASSESSMENT)
T-202 THREE FULL POWER TESTS WITH QTR 1,
81 1-1 LOFT G. MCPHERSON (1) CONTINUOUS DEPRESSURf2ATION, (LOFT)
(2) PRESSURE HANGUP, AND (3) REPRESSURIZATION, (4) LOSS & RETURN OF NATURAL CIRCUL.4 TION, (S) LONG-TERM HEAT REMOVAL.
116 -
l T-208 CELIAB5LITY DAVA MLNUAL QTR 1, 61 NONE.
J. JOHNSON (RISK ASSE3SMENT)
T-211 RISK ASSESSMENT OF SPENT FUEL IN BEDDED QIR 1, 81 6-S HIGH LEVEL WASTE ISOLATION M. CULLINGFORD SALT (RISK ASSESSMENT)
T-259 NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR QTR 1, 81' S-23 OCCUPATIONAL EXPOSURE J.
FOULKE SIT'S PROTECTION T-91 FRAGILITY DESCRIPTIONS FOR COMP 0HENTS AND QTR 2, 81 3-10 PLANT STRUCTURES J. BURNS STRUCTURES. DOCUMENTATION OF AN OVERALL (SEISMIC, ENGINEERING &
FRAGILITY DESCRIPTION DEVELOPMENT SITE SAFETY)
METHJDOLOGY.
T-112 IODINE TRANSPORT BEHAVIOR WITHIN PWR QTR 2, 81 1-9 FISSION PRODUCT RELEASE R. SHERRY STEAM GENERAIGR AND THE SECONDARY SYSTEM (FUEL BEHAVIOR)
UNDER SGTR ACCIDENT CONDITONS.
T-129
SUMMARY
OF INTERIM RESULTS. PRELIMINARY QTR 2, 81 3-1 MRC/ STATE REGIONAL EARTH N. STEUER TECTONIC PROVINCE BOUNDARIES, REGIONAL SCIENCES GEOLOGY, SEISM 0 TECTONICS. TO ACCELERATE (SEISMIC, ENGINEERING &
~
LICENSING.
SITE SAFETY)
T-174 DOCUMENTATION OF FAILURE MODES FOR QTR 2, 81 MECHANICAL ENGINEERING J. BURNS COMPONENT! AND STRUCTURES FOR A (SEISMIC, ENGINEERING &
REPRESENTATIVE PWR PLANT.
SITE SAFETY)
T-183
SUMMARY
EVALUATION OF OYNAMIC QTR 2, 81 3-11 P8_ ANT COMPONENT AND
- b. REIFF EXCITATION TESTS OF A NUCLEAR VALVE.
EQUIPMENT BEHAVIOR T-186 RESPONSE COMBINATION METHODOLOGY.
QTR 2, 11 3-9 SEISMIC SAFETY MARGINS J. O'BRIEN RESEARCH PROGRAM (SEISMIC, ENGINEERING &
SITE SAFETY)
T-187 MODELING TECHNIQUES FOR NONLINEAR QTR 2, 31 3-11 PLANT COMPONENT AND D. REIFF SYSTEMS.
EQUIPMENT BEHAVIOR T-204 FULs PCWER TEST WITH LOSS OF STEAM QTR 2, 81 1-1 LOFT G. MCPHERSON LOAD LINE BREAK 2 (1) LOSS OF FEEDWATER, (LOFT)
(2) LOSS OF PCS FLOW.
T-201 SMALL BREAK AT FULL POWER WITH/WITHOUT QTR 3, 31 1-1 LOFT G. MCPHERSON PRIMARY ? UMPS.
(LOFT)
T-209 HAZARDS TO NUCLEAR PCWER PLANTS QTR 2, 81 6-6 LWR RISK ASSESSMENT K. MURPHY (RISK ASSESSMENT)
T-232 INSPECTION METHODS FOR PHYSICAL QTR 2, 31 4-3 SAFEC* LARDS INFO. SYSTEMS E. RICHARD PROTECTION
- 117 -
i T-1.75 RECENT VERTICAL CRUSTAL MOVEMENTS IN QTR 3, 31 3-2 GEOLOGY AND SEISMIC J. HARBOU*.
THE EASTERN U.S WILL FURNISH Ir:PORTANT CHARACTERISTICS CORROBORATIVE INFCRMATION ON FAULT (SEISMIC, ENGINEERING 1 MOVEMENTS, SEISMIC ACTIVITY AND INTRAPLATE SITE SAFETY)
EARTHQUAKE MECHANISMS.
T-195 EXTENDED DRY-0UT IN SODIUM COOLED DEBRIS QTR 3, 81 2-2 POST ACCIDENT HEAT REMOVAL R. WRIGHT BEDS, ACRR TEST D-S.
T-210 FLOOD RISK SYSTEMS ANALYSIS QTR 3, 81 NONE S. STURGES (RISK ASSESSMENT)
T-236 EXP. EVALUATICH OF SYSTEMS COMPONENTS QTR 3, 81 S-6 FUEL CYCLE FACILITIES D.
SOLBERG DURING LARGE PRESSURE PULSES EFFLUENT CONTROL T-284 EVALUATION OF PULSED RADAR SYSTEMS FOR QTR 3, 31 NONE C. JUPITER SUB-SURFACE PROFILING OF SHALLCW LAND BURIAL SITFS.
T-61 ATWS TRAC VERSION.
PROVIDES ADVANCED QTR 4, 81 1-14 ADVANCED SYSTLM CODE H. ZUBER CAPABILITY TO ANALYZE ATWS IN LWR'S.
(CODE DEVELOPMENT)
T-65 POST CHF HEAT TRANSFER CORRELATION.
QTR 4, 81 1-3 PWR-BDHT Y. HSU SUPPORIS OR REPLACES CURRENT 1-4 BWR-BDHT Y.
HSU CORRELATIONS.
(SYSTEMS EhGINEERING)
T-80 IMPROVED LOOSE PARTS MONITORING.
QTR 4, 81 1-28 NOISE SURVEILLANCE AND W.
FARMER PROVIDES INPUT TO REG. GUIDE REQUIREMENT DIAGt10STICS CHANGE AND TO LICEhSING REVIEW.
(SYSTEMS ENGINEERING)
T-96 SAFETY RELATED OPERATOP ACTING CRITERIA.
QTR 4, 31 1-24 HUMAN ENGINEERING W.
FARMER MAY IMPACT ENGINEERED SAFEfY SYSTEMS (SYSTEMS ENGINEERING)
T-126 FINAL REPORT OF NEW ENGLAND SEISMO-QTR 4, 81 3-1 HRC/ STATE REGIONAL EARTH H.
STEl'ER TECTONIC STUDY.
TECTONIC PROVINCE SCIENCES BOUNDARIES DELINEATED, POSSIBLE EARTH-(SEISMIC. ENGINEERING 3 QUAKE SOURCE AREAS IDENTIFIED.
GEO-SITE SAFETY)
TECHNICAL MAPS COMPLETED.
T-133 NEW MALRID/ UPPER MISSISSIPPI VALLEY QTR 4, 31 3-1 NRC/ STATE REGIONAL EARTH H.
STEl'ER l
l SEISM 3 TECTONIC STUDY FINAL REPORT.
SCIENCES l
TECTCHIC PROVINCE BOUNDARIES DELINEATED, (SEISMIC, ENGINEERING 2 i
PROBABLE EARTilQUAKE 501RCE AREAS IDENTIFIED SITE SAFETY)
GE0 TECHNICAL MAPS COMP 6ETED.
l T-134 NEMAHA/MI9 CONTINENT GRAVITY ANOMALY STUDY QTR 4, 81 3-1 NRC/ STATE REGIONAL EARTH N.
STEUER FINAL REPORT.
TECTO4IC PRGVINCE BOUNDARIES SCIENCES
~
DELINEATED. PPCOABLE EARTHIUAKE SOURCE (SEISMIC, ENGINEERING &
AREAS IDENTIFIED. GE0 TECHNICAL AND SITE SAFETY)
SEISMIC MAPS CD:IPLETED.
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T-145 LABCRATORY SIMULATION OF TCRNADO WIND-QTR 4, 31 3-3 c.NVIRONMENTAL STRUCTURAL R.
ABBEY LO4DS.
DESIGN (SEISMIC. ENGINEERING &
SITE SAFETY)
T-282 EvitoATION OF HON-INTRUSIVE TECHNIQUES QTR 4, 81 NONE C. JUPITER FOR ROCK MASS CHARACTERIZATION.
T-287 LOSS OF FEEDWATER WITH DELAYED SCRAM QTR 4, 31 1 -1 LOFT G. MCPHERSON (LOFI)
T-86 INCORPORATE IMPROVED UT IN CODE.
IMPROVED QTR 1,
82 1-21 NON-DESTRUCTIVE EXAMINATION J. MUSCARA CONFIDENCE IN EVALUATION OF FLOW SIG-(PP.IMARY SYSTEMS INTEGRITY)
NIFICANCE FRCM ACLURAiE FLOW SHAPE AND SIZE.
T-212 SELECTED ALTERNATIVES FOR MANAGEMENT GF QTR 1,
22 NONE J. CURRY RADI0 ACTIVE WASTE GASES T-72 VERIFY TRAC-2.
GIVES CONFIDENCE THROUGH QTR 2. 82 1-17 CODE APPRAISAL S.
FABIC CCMPARISON WITH LOFT, GERMAN AND JAPANESE (LODE DEVELOPMENT)
TEST CATA.
T-109 LOCA (SINGLE ROD).
OTR 2, 82 1-10 PBF M.
FICKLESIMER (FUEL SEHAVICR) l T-12C ANNA. OHIO SEISMIC NETWORK.
INSTALLATION QTR 2, 82 3-2 GEOLOG( AND SEISMIC R. BRAZEE l
CPERATION AND RESEARCH HISTOR).
SEISMIC CHARACTERISTICS i
ACTIVITY AND RESEARCH
SUMMARY
(SEISMIC. ENGINEERING 1 l
S!YS SAFETY) l T-205 FUtl PCWER DECL BREAK WITH 07F-SITE QfR 2, 82 1-1 LCFT G.
MCPHERSON i
j POWER.
EARLY CORE QUENCH SEE 4 IN L2-3 (LOFT)
IS NOT EXPECTED TO OCCUR.
l T-250 THERMAL CONDUCTIVITY ROCK-MINERAL QTR 2, 82 NONE C. JUPITER l
(WASTE MANAGEMENT)
T-213 FIRE RISK SYSTEMS ANALYSIS QTR 3, 82 NONE D. STURGES (RISK ASSESSMENT)
T-55 LOCA THERMAL SHOCK.
PR0"IDES LICENSING QTR 4, 32 NONE M.
VAGINS l
PROCEDURE FOR THERMAt SFOCK ANALYSIS (PRIMARY SYSTEMS INTEGRITY)
INCLUDING EFFECTS OF WARM PRESCkEENING.
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DISTRIB31 TION I
l Office of the Commissioners 5
l l
Office of the Executive Directo for Operations 5
l l
Office of Inspection & Enforcement 15 Office of Nuclear Material Safety & Safeguards 25 l
l Office cf Nuclear Reactor Regulation 40 Office of Nuclear Regulatory Research 150 Office of Standards Development 25 Office of Management & Program Analysis 25 i
Office of Administration 2
Office of the Controller 2
l l
Office of Policy Evaluation 5
Office of State Programs 2
Advisory Committee on Reactor Safeguards 12 Atomic Safety & Licensing Board Panel 5
Atomic Safety & Licensing Appeal Panel 5
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