ML19318B523
| ML19318B523 | |
| Person / Time | |
|---|---|
| Issue date: | 10/15/1979 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| NUREG-0435, NUREG-0435-V01-N04, NUREG-435, NUREG-435-V1-N4, NUREG-435-V1-NO4, NUDOCS 8006260424 | |
| Download: ML19318B523 (85) | |
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STATUS
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RESEARCF RESULTS UTILIZATION
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TABL E OF CONTENTS PAGE NO.
E 1
1.0 IHTRODUCTION 2
2.0 RESEARCH INFORMATION LETTERS (RIL'S) AND SCHEDULING POST RIL ACTIVITIES 4
2.1 LIST OF ISSUED RIt's 8
2.2 POTENTIAL APPLICABILITY OF RESEARCH RESULTS IN THE REGULATORY PROCESS 71 3.0 PROJECTED HEAR TERM RESEARCH INFORMATION LETTERS I
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1.0 INTRODUCTION
THI; REPCRT ON "RESEARCH RESULTS UTILIZATION" PROVIDES STATUS AND CONTROL INF0nMATION CONCERNING THE UTILIZATION OF RESEARCH RESULTS IN THE REGULATORY POLICIES AND PRACTICES OF THE NRC.
-RESEARCH INFORMATION LETTERS - (RIL'S)-ARE PREPARED BY RES TO TRANSMIT RESJARCH RESULTS TO NRC USER OFFICES UPON COMPLETION OF A SUBSTANTIAL, COHERENT AND REASONABLY COMPLETE BODY OF EXPERIMENTAL AND/OR ANALYTICAL RESEARCH WORK.
A RIL MAY COVER MATERIAL DEVELOPED FROM MORE-THAN ONE RESEARCH PROJECT. THE USER / PROGRAM OFFICE (S) IN THE HRC REVIEW THE INFORMATION CONTAINED IN THE RIL AND C0dSIDER ITS UTILIZATION IN THE REGULATORY PROCESS. SECTION 2.0 0F THIS REPORT LISTS THE RIL'S ISSUED TO DATE, TOGETHER WITH AN IDENTIFICATION OF THE RESEARCH PROGRAM MANAGER AND THE RESEARCH PROGRAM ELEMENT WHICH GENERATED THE RIL.
THE POTENTIAL APPLICABILITY OF EACH RIL TO THE REGULATORY PROCESS IS ALSO IDENTIFIED HERE, AND COMMENTS FROM THE COGNIZANT RES AND' USER OFFICE-STAFF ARE SUMMARIZED WHICH RELATE TO THE EXPECTED IMPACT OF THE REPORTED RIL'S ON THE REGULATORY PROCESS.
WHERE DEEMED APPROPRIATE BY MUTUAL AGREEMENT BETWEEN RES AND OTHER INVOLVED PROGRAM OFFICES, A POSITION PAPER MAY BE PREPARED TO INFORM THE COMMISSION ABOUT THE POTENTIAL APPLICATION OF RESEARCH RESULTS. BRIEFINGS MAY BE CONDUCTED FOR THE COMMISSION, ALSO BE MADE, AS APPROPRIATE. THESE EVENTS ARE SCHEDULED AND TRACKED IN THIS SECTION.
A LISTING OF ALL RIL'S THAT MAY BE GENERATED IN THE HEAR FUTURE (2 OR 3 YEARS) IS PRESENTED IN SECTION 3.0.
THE SUBJECT, TITLE, IMPACT, TARGET DATE, AND RESEARCH REVIEW GROUP NUMBER, TITLE AND CHAIRMAN ARE SHOWN FOR EACH RIL.
RES IS RESPONSIBLE FOR DISTRIBUTING EACH HEW RIL TO COGNIZANT OFFICES AS THEY ARE ISSUED.
RES AND OTHER COGNIZANT PROGRAM OFFICES ARE OBLIGATED TO COLLECT, REVIEW, AND FORWARD APPROPRIATE INFORMATION RELATING TO RILS AND ASSOCIATED FOLLOW-UP ACTIONS TO MPA.
MPA IS RESPONSIBLE FOR REFLECTING NEW INFORMATION IN THIS PUBLICATION AND PRODUCTIDH AND DISTRIBUTION OF THIS PUBLICATION ON A QUARTERLY SCHEDULE.
ALL COMMENTS SHOULD BE FORWARDED IN WRITING TO:
DIRECTOR, MPA MAIL STOP 12711 MNBB-ATTH:
D. BIDLE
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15 YES YES YES i
16 YES YES
'17 YES-YES 18 POSS.
.POSSIBLE 19 POSS.
1 20 YES YES YES YES YES "21 YES.
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A 28 YES VES 29 30 YES YES 31M 32 (ES YES YES 33 YES YES YES 34 YES YES 35M 36 YES YES 37M POSS.
38N 39M 40*
41 YES YES YES YES 42M 43M 44 YES YES 45 46 YES YES 47M 48M 49 YES 50M 51
~52M 53M 54M 55 56M N - CURRENTLY BEING REVIEWED BY USER OFFICE (S)
- - AWAITING COMMISSION APPROVAL ON PROPOSED ACTION PLAN FOR CNANGING ECCS RULE
2.1 LIST OF ISSUED RIt's RESEARCH INFORMATION LETTERS ISSUED BY RES TO DATE ARE LISTED BELOW IN TABLE 2.1, TOGETHER WITH THE DATE OF ISSUE, ASSOCI ATED RESEARC.i PROGRAM ELEMENT AND PROGRAM MANAGER.
DATE RIL NO.
PAGE NO.
ISSUED RESEAPCH INFORMATION LETTER TITLE RES PROGRAM ELEMENT PROJECT MANAGER 1
9 03/19/74 ORNL V-5 INTERMEDIATE VESSEL TEST RESULTS PRIMARY SYS INTEG.
C. SERPAN 2
10 05/20/74 SEISMOTECTONIC MAP OF THE EASTERN UNITED SITE SAFETY J. HARBOUR STATES 3
11 08/07/74 ORNL V-7 INTERMEDIATE VESSEL TEST RESULTS PRIMARY SYS INTEG.
C. SERPAN 4
12 09/10/74 MAP SHOWING RECENCY OF FAULTING IN COASTAL SITE SAFETY J. HARBOUR SOUTHERN CALIFORNIA 5
13 06/28/76 CONFIRMATORY PRESSURE VESSEL TEST UNDER PRIMARY SYS INTEG.
C.
SERPAN PNEUMATIC iOADING 6
14 10/12/16 A CRITICUE OF THE BOARD-HALL MODEL FOR THERMAL FAST BREEDER REACTOR R. WRIGHT DETONATIONS IN THE UO2-HA SYSTEM 7
15 08/25/76 THE SIMMER CORE FOR ANALYSIS OF HYPOTHETICAL FAST BREEDER REACTOR R. CURTIS CORE DISRUPTIVE ACCIDENTS IN LMFBR'S 8
16 01/31/77 DECAY HEAT DATA APPLICABLE TO LOCA EVALUATION FUEL BEHAVIOR R. DISALVO 9
17 03/14/77 HIGH TEMPERATURE OXIDATION OF ZIRCALOY FUEL FUEL BEHAVIOR M. PICKLESIMER CLADDING IN S T E "1 10 18 02/25/77 PRESSURE VESSEL FAILURE PROBABILITY PREDICTICH RISK ASSESSMENT W. VESELY (OCTAVIA CODE)
PRIMARY SYS INTEG.
11 19 09/15/77 IEEE HUCLEAR RELIABILITY DATA MANUAL RISK ASSESSMENT J. JOHNSCH 12 20 06/16/77 MODIFICATIONS TO PRESSURE VESSEL FAILURE RISK ASSESSMENT W. VESELY PROBABILITY PREDICTION (OCTAVIA CODE)
PRIMARY SYS
'.."TEG.
13 21 11/11/77 RESIDUAL STRESSES IN WELDS PRIMARY SYS INTEG.
C. SERPAN 14 22 11/09/77 PHYSICAL SEPARATION CRITERIA FOR ELECTRICAL SYSTEMS ENG.
R. FEIT CABLE TRAYS (HORIZONTAL OPEN SPACE CONFIG.)
15 24 12/01/77 CHARACTERIZATION OF BWR FEEDWATER N0ZZLE PRIMARY SYS INTEG.
C. SERPAN
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CORNER CRACKS 16 25 12/01/77 WARM PRESTRESSING PRIMARY SYS INTEG.
C. SERPAN 17 26 05/05/78 POWER BURST FACILITY (PBF) SINGLE ROD-POWER FUEL BEHAVIOR R. VAN HOUTEN COOLING tiiSMATCH (PCM) TEST RESULTS
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18 27 11/09/77 FRANTIC COMPUTER CODE RISK ASSESSMENT F. GOLDBERG O
DATE RIL NO.
PAGE NO.
ISSUED RESEARCH INFORMATION 2?ER TITLE RES PROGRAM ELEMENT PROJECT MANAGER 19 28 01/31/78 GO METHODOLOGY ASSESSMENT RISK ASSESSMENT J. PITTMAN 20 29 01/24/78 A STUDY OF PHYSICAL PROTECTION EQUIPMENT SAFEGUARDS E. RICHARD
-21 31 03/24/78 CRITICAL REVIEW OF SODIUM HYDROXIDE AEROSOL RISK ASSESSMENT M.
CULLINGFORD T0XICITY 23 32 04/10/78 "EASI" ADVERSARY SEQUENCE EVALUATION MODEL SAFEGUARDS R. ROBINSON (COMPUTER GRAPHICS VERSION) 24 33 04/10/78 "FESEM" ADVERSARY SEQUENCE EVALUATION SAFEGUARDS R. ROBINSON MODEL 25 34 03/21/78 FRAP-S3 FUEL BEHAVIOR G. MARINO 26 35 04/27/78 THE IMPACT OF OFFSHORE NUCLEAR GENERATING FUEL CYCLE SAFETY D. BARNA STATIONS ON RECREATIONAL BEHAVIOR AT ADJACENT AND ENVIRONMENTAL COASTAL SITES.
EFFECTS 27 36 06/02/78
" BEACON / MOD 2" CODE DEVELOPMENT S.
FABIC 28 37
-05/09/78
" MELT / CONCRETE INTERACTIONS" FAST BREEDER REACTOR R.
DISALVO 29 39 06/07/78
" FUEL ROD ANALYSIS COMPUTER CODE:
FRAP-T3" FUEL BEHAVIOR H. SCOTT 30 41 06/28/78 PHASE I FINAL REPORT, " BARRIER PENETRATION SAFEGUARDS R. ZIMMERMAN DATA BASE"; 0F STUDY, " ASSISTANCE-PHYSICAL PROTECTICH ASSESSMENTS" 31 42 07/10/78 ASSAY OF STANDARD REFERENCE MATERIAL (SRM)950 SAFEGUARDS R. SHEPARD 32 43 08/03/78 IMPROVEMENTS Is THE AEROSOL BEHAVIOR CODE FCR FAST BREEDER J.
LARKINS RADIOLOGICAL AV5ESSMENTS OF LMFER'S REACTOR 33 44 08/03/78 PLUTONIUM ACCIDENT CONTAINER PROGRAM RESEARCH, FUEL CYCLE SAiETY W.
LAHS DESIGN AND DEVELOPMENT.
AND ENVIRONMENTAL EFFECTS 34 45 08/03/78 NUCLEAR DECAY DATA FOR RADIONUCLIDES OCCURRING FUEL CYCLE SAFETY J. FOULKE IN ROUTINE RELEASES FROM NUCLEAR FUEL CYCLE AND ENVIRONMENTAL FACILITIES EFFECTS 35 46 09/15/78 SFACTOR: A COMPUTER CODE FOR CALCULATING DOSE FUEL CYCLE SAFETY J. FOULKE EQUIVALENT TO A TARCET ORGAN PER MICR0 CURIE -
AND ENVIRONMENTAL DAY RESIDENCE OF A RADIONUCLIDE IN A SOURCE EFFECTS ORGAN 36 47 09/27/78 EVALUATION OF GENERAL ATOMIC CODES:
OXIDE-3, FAST BREEDER J.
LARKINS SORS, TAP, AND RECA.
REACTORS 37 49 09/28/78 LOFT REACTOR SAFETY PROGRAM RESEARCH RESULTS LOFT G. MCPHERSON THROUGH OCTOBER 1,
1978
DATE RIL NO; PAGF NO.
ISSUED RESEARCH INFORMATION LETTER TITLE RES FROGRAM ELEMENT PROJECT MANAGER 38 50 10/13/78 RESULTS OF THE INITI AL SERIES OF ACPR EXPERI-FAST REACTOR R. WRIGHT MENTS ON PROMPT-BURST ENERGETICS WITH FRESH SAFETY RFSEARCH OXIDE FUEL 39 51 11/27/78 RELAP-4/ MOD 6 CODE DEVELOPMENT S. FABIC 40 52
'12/18/78 THE CCMPUTER CODE BRENDA - A~ COMPUTER PROGRAM FAST REACTOR P. WOOD FOR THE DYNAMIC SIMULATION FOR A LIQUID METAL SAFETY RESEARCH FAST BREEDER REACTOR PLANT 41 53 12/19/78 LABORATORY TESTING PROCEDURES TO DETERMINE THE ADVANCED REACTOR N. STEUER CYCLIC STRENGTH OF-SOILS SAFETY RESEARCH 42 54 12/20/78 CRITICAL EXPERIMENT PROGRAM FOR NEUTRONICS CODE ADVANCED REACTOR P. WOOD VERIFICATION SAFETY RESEARCH 43 55 01/10/79 SUPER SYSTEM CODE, A COMPUTER PROGRAM FOR FAST REACTOR P. WOOD DYNAMIC SIMULATION OF LMFBR POWER PLANTS SAFETY RESEARCH 44 56 01/04/79 RADIATION DOSE TO CONSTRUCTION WORKERS AT FUEL CYCLE SAFETY J. FOULKE OPERATING NUCLEAR POWER PLANT SITES AND ENVIRONMENTAL EFFECTS 45 57 02/11/79 THE CONCEPT COMPUTER CODE & CAPITAL COSTS FUEL CYCLE SAFETY D. BARNA FOR BOILING WATER REACTOR PLANTS AND ENVIRONMENTAL EFFECTS 46 58 02/12/79 THE EFFECTIVENESS OF CABLE TRAY COATING SYSTEMS R. FEIT MATERIALS & BARRIERS IN RETARDING THE ENGINEERING CCMBUSTION OF CABLE TRAYS SUBJECTED TO EXPOSURE FIRES & IN PREVENTING PROPAGATION BETWEEN CABLE TRAYS (HORIZONTAL OPEN SPACE CCNFIGURATION) 47 59 03/19/79 INREM II: A COMPUTER IMPLEMENTATInN FUEL CYCLE SAFETY J. FOULKE OF RECENT MODELS FOR ESTIMATING TH.
1 ENVIRONMENTAL DOSE EQUIVALENT TO ORGANS OF MAN FROM EFFECTS AN INHALED OR INGESTED RADIONUCLIDE 48 61 04/03/79 A TECTONIC OVERVIEW OF THE CENTRAL SITE SAFETY N. STEUER MIDCONTINENT 49 62 C4/04/79 IN VITRO DISSOLUTION OF URANIUM FUEL CYCLE SAFETY J. FOULKE PRODUCT SAMPLES FROM FOUR URANIUM AND ENVIRONMENTAL MILLS EFFECTS 50 63 04/06/79 CPITICALITY SAFETY GUIDANCE FUEL CYCLE SAFETY D. SDLBERG AND ENVIRONMENTAL l
EFFECTS l
l 51 64 04/12/79 THE CONCEPT COMPUTER CODE AND CAPITAL FUEL CYCLE SAFETY D. BARNA COSTS FOR PRESSURIZED WATER REACTOR AND ENVIRONMENTAL l
EFFECTS PLANTS l
52 65 04/23/79 EARTHQUAKE INTENSITY SCALE SITE SAFETY R. BRAZEE !
DAVE RIL NO.
PAGE NO.
ISSUED-RESEARCN INFORMaTICN LETTER TITLE RES PROGRAM ELEMENT. PROJECT MANAGER 53 66 05/16/79 DEBRIS-BED C00 LABILITY LIMITS. RESULTS FUEL CYCLE SAFETY R. MRIGHT FROM IN-CORE TESTS D-1, D-2 AND D-3 ND EN IRONMENTAL 154
-67 05/15/79 THE: SET EQUATION TRANSFORMATION SYSTEM RISK ASSESSMENT W. VESELEY 55 69
'05/29/79 THE CONCEPT CCMPUTER CODE AND CAPITAL FUEL CYCLE SAFETY D. BARNA COST FOR HIGH AND LOW SULFUR CDAL AND ENVIRONMENTAL PLANTS - 1200 MWE EFFECTS 56-70 07/25/79 EFFECTS OF NUCLEAR POWER PLANTS ON ENVIRONMENTAL C. PRICHARD COMMUNITY GROWTN,AND RESIDENTIAL EFFECTS RESEARCH PROPERTY VALUES
l
'2.2 POTEN'TIAL APPLICABftfTY OF RESEARCH RESULTS IN THE REGULATORY PROCESS COMMENTS ARE OFFERED FRCM RES AND FROM COGNIZANT USER OFFICES ON THE NATURE OF APPLICABILITY TO ANY PART THE REPORTED RESULTS'AND THEIR POTENTIAL COR ACTUAL) TECHNICAL DATA SUPPORTING LICENSING REVIEWS OF THE REGULATORY PROCESS, INCLUDING: A)
OR REGULATORY JUDGMENT, B) EVALUATION CODES. C) INFORMATION APPLICABLE TO.
- REGULATORY GUIDES OR STANDARDS. AND D) INFORMATION SUPPORTINS JUDGMENTS REGARDING REGULATORY' POLICY.
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1 PROGR AM OFFYCE COMr1ENTS ON POTENYI AL UVILIZAYVON OR MatuE OF RESEARCH RESULV5 VN THE REGULATORY PROCESS RIL 8:
1 DATE ISSUED: 03/19/74 RES PROGR AM EL EMENT: PRIMARY SYSTEM INTEGRITY RIL TITLE: ORNL V-5 INTERMEDIATE VESSEL TEST RESULTS (ORNL HSST PROGRAM)
SPONSORING OFFICE (S): RES ggh:
1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR:
C. SERPAN
. RES' COMMENTS: EXPERIMENTAL EVIDENCE THAT A " SAFE".AILURE MODE (LEAK-BEFORE-BREAK) FOR REACTOR PRESSURE VESSELS MAY EXIST, WAS REPORTED FOR THE FIRST TIME. TESTS ON THE V-5 INTERMEDIATE VESSEL AT ORNL DEMONSTRATED THAT A LEAK OCCURRED INSTEAD OF A FRACTURE BREAK WHEN THE VESSEL WAS 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 g_0ST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR/SD NRR C'
SCHEDULED COMPL ETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED-N/A ACTUAL COMPLETION DATE.....
08/21/78 N/A NRR COMMENTS ON 09/09/77. 08/21/78.
B.
GRIMES DESCRIBE _ APPLICATION 10 REGUIATORY PROCESS: THE RESULTS OF THIS TEST ADDED TO THE STAFF'S UNDERSTANDING OF FRACTURES ORIGINATING AT FLAWS IN NG4ZLE CORNERS OF HEAVY SECTION STEEL VESSELS.
IF LEAK-BEFORE-BREAK COULD BE DEMONSTRATED UNDEP ALL REASONABLY CONCEIVABLE CONDITIONS AND C7,RCUMSTANCES, IT WOULD DEMONSTRATE THAT CURRENT LICENSING POSITIONS ARE VERY CONSERVATIVE.
Df1SRIBE IMPACT OF RESULTS: WHILE ADDING TO 00R UNDERSTANDING OF VESSEL FAILURE MODES, THERE HAS BEEN NO DEFINITIVE IMPACT ON LICENSING AT THIS TIME.
EQMMENTS/ REMARK 3: THE OPINIONS EXPRESSED IN THE RIL REGARDING LEAK-BEFORE-BREAK ARE ENCOURAGING BUT NEED FURTHER SUBSTANTIATIDH BEFORE CURRENT LICENSING POSITIONS CAN BE RELAXED.
SD COMMENTS ON 09/13/78, P.
RANDALL THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED Ah PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES AND LOADINGS. THIS PROGRAM HAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUCHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G T0 to CFR 50 AND CONTRIBUTED DATA USED IN THE ASME CODE WHICH WAS INCORPORATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE DATA BASE USED IN REGULATORY GUIDE 1.99 CONCERNING THE EFFECT OF COPPER IMPURITIES ON SENSITIVITY OF STEEL TO IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS TO REGU!.ATORY GUIDE 1.99.
8
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J PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN TH PIL 1:
2 DATE ISSUED 05/20/74-RES PROGRAM ELEMENT: SITE SAFETY
- RIL TITLE: SEISMO TECTONIC MAP OF THE EASTERN UNITED STATES
~ RESEARCH PROJECT MGR:
J. HARBOUR
- SPONSORING OFFICE (5): RES gggt 3-2 GEOLOGY & SEISMIC. CHARACTERISTICS l
THESE DATA RIL 2 REPORTS A COMPILATION OF EARTHQUAKE FAULT DATA FOR YHE EASTERN UNITED STATES.
ARE USED BY. LICENSE APPLICANTS IN PREPARATION OF MATERIAL FOR PRELIMINARY. SAFETY ANAL RES~ COMMENTS:
USER DISCUSSI0H l POSITION COMMISSIDH ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEA3E RESULTS i
POST RIL ACTIVITIES REVIEW HELD
~
CCMPLETED HELD HELD ISSUE 2_
IMPLEMENTED 1
- RR -
l 0FFICE RESPONSIBLE......... NRk/SD UNSCHED
.UNSCHED UNSCHED UNSCHED UNSCEED 1974 1974 SCHEDULED COMPLETION DATE. --
-- ACTUAL COMPLETICH'DATE...... 09/27/77 l
j THIS RESEARCH DEVELOPED LIMITED SEISM 0 TECTONIC PROVINCES NRR COMMENTS ON C9/27/77.
R.
DENISE:
DESCRIBE APPLICATION TO REGULATORY PROCESS:
MAPPING OF THE EASTERN UNITED STATES AIMED AT IMPLEMENTING APPENDIX A REQUIREMENTS FO SEISMIC DESIGN FOR NUCLEAR FACILITIES.THE RESEARCH ADDED VERY LITTLE TO OUR KNOWLEDGE OF EARTHQUAK DESCRIBE IMPACT Of_RESULTS:
THE EASTERN UNITED STATES WHICH COULD BE USED TO IMPLEMENT APPENDIX A R
' COMMENTS / REMARKS:
l AH APPLICABLE SEISM 0 TECTONIC MAP, SD COMMENTS ON 09/13/78. G. RIVENBAEK THIS STUDY IS PRESENTLY USED AS GUIDANCE IN ASSESSING TECTONIC PROVINCES AND SEISMICITY THIS STUDY, AS WELL AS ONGOING IN LICENSING CASE REVIEWS IN THE EASTERN UNITED STATES.
STUDIES, WILL EVENTUALLY BE USED TO OFFER GUIDANCE ON A REGIONAL BASIS THROUGH REGULATORY GUIDES.
I W -
1 PROGRAM OFFICE COMNENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL s 3
DATE ISSUEDr 08/07/74 PES PROGPAM EL EMENT PRIMARY SYSTEM INTEGRITY RIL TITLEr ORNL V-7 INTERMEDIATE VESSEL TEST RESULTS (ORNL HSST PROGRAM)
SPONSORING OFFICEfS): RES gag:
1-20 VESSEL INTEGRITY-RESEARC9 PROJECT MGR:
C. SERPAN RES COMMENTSr ADDITIONAL EXPERIMENTAL EVIDENCE IS DE!10NSTRATED THAT A " SAFE" FAILURE MODE FOR REACTOR PRESSURE VESSELS MAY EXIST, NAMELY, " LEAK-BEFORE-BREAK".
THESE RESULTS SUPPORT EARLIER RESULTS REPORTED IN RIL 81.
IN THIS TEST, THE FLCW WAS 13-INCHES LONG AND 5-INCHES DEEP IN THE 6-INCH THICK PRESSURE VESSEL WALL.
THE VESSEL WAS ABLE TO SUSTAIN TWICE ThE DESIGN LOAD PRIOR TO PENETRATION OF THE FLOW THROUGH THE REMAINING THIN LIGAMENT OF VESSEL MATERIAL. THIS RESEARCH CONTRIBUTES TO OUR UNDERSTANDING OF VESSEL FAILURE MODE. THIS INCREASED UNDERSTANDING SUPPORTS THE NRR STAFF IN ITS REVIEW EFFORTS.
USER DISCUSSION FOSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD
, OFFICE RESPONSIBLE......... NRR/SD
[25PLETED FELD HELD ISSUED IMPLEMENTED NRR SCHEDULED COMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED N/A ACTUAL COMPLETION DATE..... 09/09/77 N/A NRR COMMENTS ON 09/09/77, D.
EISENHUT DESCRIBE APPLICATION TO REGULATORY PROCESS: THE TEST RESULTS HAVE BEEN USEFUL IN ESTIMATING THE MARGIN OF SAFETY, IN TERMS OF FLAW SIZE, FOR A LARGE FLAW IN A HEAVY SECTION STEEL VESSEL TESTED AT UPPER SHELF TEMPERATURES. WHILE CRACK ADVANCEMENT THROUGH THE REMAINING WALL LIGAMENT WAS IN THE STABLE TEARING MODE, PRODUCING A LEAK, AND PRESSURE TO CAUSE LEAKAGE WAS MORE THAN TWICE THE VESSEL DESIGN PRESSURE, EFFECTS OF THE TEST GEOMETRY.
THESE RESULTS MAY HAVE BEEN INDUCED BY THE SPECIAL RESCRLIE IMPACT OF RESULTS: WHILE ADDING TO OUR UNDERSTANDING OF VESSEL FAILURE MODES, THERE HAS BEEN N0 DEFINITIVE IMPACT ON LICENSING AT THIS TIME.
t COMMENTS / REMARKS:
THE OPINIONS EXPRESSED IN THE RIL REGARDING LEAK-BEFORE-BREAK ARE STILL SPECUL ATIVE.
SD COMMENTS ON 09/13/78, P.
RANDALL THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES AND LOADINGS.
THIS PROGRAM HAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERI ALS DESCRIBED IN APPENDIX G TO 10 CFR 50 AND CONTRIBUTED DATA USED IN THE ASME CODE WHICH WAS INCORPORATED INTO AFPENDIX 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 IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS TO REGULATORY GUIDE 1.99.
4 l
I PROGRAM OFFICE cot 1MENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEAPCH RESULTS IN INE REGULATORY PROCESS RIL 4:
- 4
.DATE ISSUED: 09/10/74 RES PROGRAM ELEMENT: SITE SAFETY RIL TITLE: MAP SHOWING RECENCY OF. FAULTING IN COASTAL SOUTHERN CALIFORNIA SPCNSORING OFFICECS): RES EEg: 3-2 GEOLOGY & SEISMIC CHARACTERISTICS RESEARCH PROJECT MGR:
J. HARBOUR l
RES COMMENTSi RIL 4' COVERED FAULTING IN COASTAL SOUTHERN CALIFORNIA AND IS REFERRED TO BY NRR WHEN C
. SEISMIC SAFETY QUESTIONS REGARDING PLANTS IN THE AREA COVERED.
USER DISCUSSION-POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER 2RIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVIT.LES REVIEW HELD COMPLETED FELD HELD ISSUED IMPLEMENTED NRR H
l OFFICE RESPONSIBLE......... NRR/SD SCHEDULED COMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED' UNSCHED 1975 ACTUAL COMPLETI0H DATE....
09/27/77 1975 NRR COMMENTS ON 09/27/77.
R.
DENLif T0-BE USED IN SELECTING SITES FOR HUCLEAR FACILITIES AND DESCRIBE APPLICATION TO REGULATORY PROCESS:
TO STAFF REVIEW OF huCLEAR FACILITY APPLICATIONS.
FOR REGIONAL INPUT WCRK PROVIDED NEW KNOWLEDGE OF THE DISTRIBUTION OF ACTIVE FAULTS ~4H l
DESCRIBE IMPACT OF RESULTS:HAS PROVIDED INPUT TO OUR REVIEW OF DIABLO CANYON AND OTHER SITES.
i COASTAL CALIF 0PNIA.
COMMENTS / REMARKS: WORK HAS LARGELY BEEN COMPLETED. NO EXTENSION IS ANTICIPATED.
RIVENPARK SD COMMENTS ON 09/f3/78.
G.
.THE OBJECT OF THIS STUDY WAS TO PRODUCE DATA DESCRIBING THE RELATIONSHIP BETWEEN EARTHQUAKE MAGNITUDES AND DIMENSICHS OF FAULT DISPLACEMENT. THIS STUDY, ALONG WITH OTHERS, IS USED AS GUIDANCE IN ASSESSING " CAPABLE FAULTS" IN LICENSING REVIEWS. RIL #4 PROVIDED A MAP WHICH UDPATED INFORMATION ON FAULTING IN COASTAL SOUTHERN CALIFORNIA.
i
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3-a PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS
~
RIL 8:
5 DATE ISSUED: 06/28/76 RES PROGRAM ELEMENT: PRIMARY SYSTEM INTEGRITY RIt' TITLE: CONFIRMATORY PRESSURE VESSEL _ TEST UNDER PNEUMATIC LOADING CORNL HSST PROGRAM)
SPONSORING OFFICE (S):
RES R3g:
1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR:
C. SERPAN RES COMMENTS: RIL'S 1, 3 1 5 REPORTED THE RESULTS FROM THE HEAVY SECTION STEEL TECHNOLOGY PROGRAM WHICH SHOWED THAT THE ANALYTICAL METHODS FOR PREDICTING FLAW INITIATION AND CRACK ARREST IN REACTOR PRESSURE VESSELS HAVE BEEN WELL VALIDATED.
THESE VALIDATED ANALYTICAL METHODS (LINEAR ELASTIC FRACTURE NECHANICS AND ELASTIC-PLASTIC FRACTURE MECHANICS)
ALLOW A PREDICTION OF THOSE CONDITIONS UNDER WHICH FLAWS IN PRESSURE VESSEL STEELS CAN CAUSE FAILURE OF THE VESSEL.
THIS PROVIDES THE NRR STAFF WITH A NEW TECHNIQUE FOR SETTING SAFE LIMITS FOR NORMAL OPERATION AND FOR ABNORMAL AND ACCIDENT SITUATIONS TO REDUCE THE LIKELIHOOD OF PRESSURE VESSEL FAILURE.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEM HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR/SD NRR SCHEDULED COMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
12/16/77 1976 NRR COMMENTS ON 12/16/77.
J.
KNIGHT:
DESCRIBE APPLICATION TO REGULATORY PROCESS:
THE INTERMEDIATE TEST VESSELS, ITV-7 AND ITV-7A UNDER SUSTAINED LOADING DEMONSTRATED THAT BOTH THE VESSELS RESPONDED TO PNEUMATIC LOADING ESSENTIALLY AS THEY HAD TO HYDRAULIC LOADING.
EARLIER EXPERIENCE OF A RAPID CRACK EXTENSION IN THE GAS PIPE LINE WAS INHIBITING THE ACRS COMMITTEE ABOUT THE RESULTS OF ITV-TEST UNDER HYDRAULIC LOADING. HENCE, THE PNEUMATIC LOAD TEST WAS SUGGESTED.
DESCRIBE IMPACT OF RESULTS: THESE TESTS SHOW THAT THE VESSELS UNDER SUSTAINED LOADING BEHAVED SIMIL/r.Y TO HYDRAULIC LOADING, AND THE RESULTS AkE APPLICABLE TO THE EVALUATION OF THE BEHAVIOR OF REACTOR PRESSU~6 VESSELS UNDER SUSTAINED LOAD.
THE TWO VESSEL TESTS THAT RUPTURED, WITHSTOOD PRESSURE 2.15 TO 2.75 TIMES DES: oN PRESSURE.
THE TEST PRESSURES WERE ABOVE ASME B 1 P V CODE ALLOWABLE FOR FAULTED CONDITIONS.
COMMENTS / REMARKS: TESTS DEMONSTRATE THE LEAK WITHOUT BURST.
SD COMMENTS ON 09/13/78, P.
RANDALL:
THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES AND LOADINGS. THIS PROGRAM HAS PROVIDED INPUT TO HRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR 50 AND CONTRIEUTED DATA USED IN THE ASME CODE WHICH WAS INCORPORATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE DATA BASE USED IN REGULATORY GUIDE 1.99 CONCERNING THE EFFECT OF COPPER IMPURITIES ON SENSITIVITY OF STEEL TO IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIGHS TO REGULATORY GUIDE 1.99.
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL e:
6 DATE ISSUED:
10/12/76 DES PROGRAM ELEMENT: FAST BREEDER REACTORS A CRITIQUE OF THE BOARD-HALL MODEL FOR THERMAL DETONATIONS IN THE UO2-NA SYSTEM RIL TITLE:
SPONSORING OFFICEfS): NRR RES:
2-6 ACCIDENT ENERGETICS RESEARCH PROJECT MGR:
R. WRICHT THIS CRITIQUE OF THE BOARD-HALL THEORY FOR THERMAL EXPLOSIONS IN A URANIUM CXIDE-SODIUM SYSTEM THE CRITIQUE HAS BEEN USED AS AN AID IN RES COMMENTS:
REINFORCED THE BELIEF THAT SUCH DETONATIONS ARE OF LOW PROBABILITY. ANSWERING ACRS C
{ WAS EMPLOYED IN SUPPORTING NRR'S POSITION IN THE EVALUATION OF THE PRELIMINARY SAFETY ANALYSIS REPORT FO CLINCH RIVER BREEDER REACTOR.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR NRR SCHEDULED COMPLETICH DATE.. --
UNSCHF9 UNSCHED UNSCHED UNSCHED UNSCHED 1977 1977 ACTUAL COMPLETION DATE..... 09/09/77 NRR CRMMENTS ON 09/09/77. W.
GAMMILL:
APPLICATION TO REGULATORY PROCEll: ESTIMATES OF THERMAL DETONATION ENERGIES ARE DIRECTLY RELEVANT IN PARTICULARLY KINETIC ENERGY RELEASE, SYSTEM RESCRIBE LICENSING DECISIONS RELATED TO CONTAINMENT REQUIREMENTS FOR LMFBR,THE RIL TRANSMITTED A CRITIQUE OF THE BOARD AND HALL (U DAMAGE AND RADIO ACTIVITY RELEASES.
THIS INDEPENDENT THEORETICAL ASSESSMENT IS USEFUL IN REINFORCING THERMAL EXPLOSIONS IN THE UO2-50DIUM SYSTEM.
THE CURRENT NRR BELIEF THAT SUCH DETONATIONS ARE OF LCW PROBABILITY AND THAT FURTHER EXPERIMENTAL EFFORT IS REQUIRED.
THE RESULTS OF THE CRITIQUE ARE MINIMAL IN THIS ON-GOING TECHNOLOGICAL EXPLORATION, RESCRIBE IMPACT OF RESULTS:
SUCH WORK, INDEPENDENT EXCEPT AS NOTED ABOVE IN REINFORCING THE IDENTIFIED PRECEPTIONS AND NEED FOR EXPERIMENTS.
SHOULD CONTINUE.
ASSESSMENTS OF IDENTIFIED AND POTENTIAL SIGNIFICANT SAFETY ISSUES,THE INITI AL EFFORT OF AN EXPERIMENTAL AND ANALY 1975 FOR BOARD COMMENTS / REMARKS:
EXPLOSIONS WAS UNDERTAKEN BY DPM AT THE REQUEST OF THE ACRS (ACRS REQUEST DATED JULY 9, NOTE DFM RESPONSE TO THIS REQUEST IN A LETTER FROM R. P. DENISE TO M.
LIBARKIN, ACRS, DATED CF THERMAL AND HALL REVIEW).
OCTOBER 24, 1975.
THE PHENOMENON IS APPLICABLE IN THE ONGOING REVIEW OF THE FFTF REACTOR.
NRR HAS BEEN IN CONTINUING CONTACT WITH RES STAFF ON THE RESEARCH PROGRAMS PERTAINING TO INVESTIGATION BOARD AND HALL EFFECT, AND THE GENERAL SUBJECT OF THERMAL EXPLOSIONS, AND MAINTAINS COGNIZANCE OF SUCH WORK IN THE U.S. AND OTHER COUNTRIES.
i e.
=
PROGRAM OFFICE CONMENTS CH POTENTIAL UTIt22ATION CR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS l
RIL 82 7
DATE ISSUED: 08/25/76 RES PROGRAM ELEMENT: FAST BREEDER REACTORS RIt TITLE:
THE SIMMER CODE FOR ANALYSIS OF HYPOTHETICAL CORE DISRUPTIVE ACCIDENTS IN LMFBR'S SPONSORING OFFICE (S): HRR ggg:
2-14 SIMMER CODE RESEARCH PROJECT MGR:
R. CURTIS RES COMMENTS:
THE SIMMER CODE IS DESIGNED TO PREDICT THE MOTION OF FAST REACTOR CORES DURING POTENTIAL CORE DISRUPTIVE ACCIDENTS.
A TRIAL VERSICH OF THIS CODE WAS MADE AVAILABLE TO NRC'S LICENSING STAFF AND HAS SUPPORTED NRC BRANCH POSITIONS ON THE ANALYSIS OF CORE DISRUPTIVE ACCIDENTS FOR CLINCH RIVER BREEDER REAC LICENSING CONSIDERATIONS. SIMMER IS RECOGNIZED AS AN IMPORTANT FIRST STEP IN THE DEVELOPMENT OF A MECHANISTIC CODE CAPABLE OF DESCRIBING PHENOMENA NEEDED TO ASSESS CORE DISRUPTIVE ACCIDENTS IN LIQUID METAL FAST BREEDER REACTORS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPL EMENTED OFFICE RESPONSIBLE......... NRR NRR SCHEDULED COMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 1976 ACTUAL COMPLETION DATE.....
09/09/77 1976 NRR COMMENTS ON 09/09/77. W. GAMMILL:
THE SUBJECT RIL ANNOUNCED THE AVAILABILITY OF THE SIMMER-1 CODE.
THE EESCRIBE APPLICATION TO REGULATORY PROCES%:
SIMMER-1 CODE IS RECOGNIZED BY NRR AS AN IMPORTANT FIRST STEP DESCRIBING RECRITICALITY, TRAPSITION PHASE PHENOMENA, IN THE DEVELOPMENT OF A MECHANISTIC CODE CAPABLE OF DISRUPTIVE ACCIDENTS IN LMFBRS.
AND WORK ENERGY PARTITION, ALL VITAL TO ASSESSING CORE DESCRIBE IMPACT OF RESULTS:
SUBSTANTIATED TO BE USED DIRECTLY IN THE LICENSING DECISIONS FOR LMFBR'S.THERE IS no IMPACT SINC
~
IT IS ANTICIPATED THAT FURTHER CODE
' DEVELOPMENT, AND THE PERFORMANCE OF EXPERIMENTS WHICH CONFIRM THE COMPUTATIONAL MODELS, WILL YIELD A WILL BE USEFUL IN REACHING SIGNIFICANT LICENSING DECISIGHS.
TOOL WHICH
$_0MMENTS/RfMARKS:
NRR VIEWS THE SIMMER PROGRAM AS AN IMPORTANT LONG TERM EFFORT TO DESCRIBE THERMAL AND MECHANIC CORE DISRUPTION ACCIDENT SEQUENCES.
AND THE INTEGRATION OF THESE PHENOMENA INTO AN ACCCIDENT SEQUENCE.NRR ALSO SEES NE PHENOMENA, SIMMER ALSO HAS THE BENEFIT OF BEING LARGELY INDEPENDENT OF ERDA SPONSORED WORK.
NRR HAS BEEN IN CONSTANT COMMUNICATION WITH ARSR AND HAD INITIATED A SMALL T.A. EFFORT WHICH MAKES USE OF SIMMER. HRR STAFF HAVE ATTENDED SIMMER WORKSHOPS AND BRIEFINGS ON 9/14-15/76, 4/19/77 AND 7/21-22/77.
NRR CORRESPONDENCE RELATED TO SIMMER AND SIMMER-RELATED WORK CONDUCTED BY RSR:
1 TO SAUL LEVINE, RES FROM B. RUSCHE, NRR, "LMFBR SAFETY RESEARCH PROGRAM PLAN", 3/15/77.
2.
TO L. S.
RUBENSTEIN FROM R. P. DENISE, " DEGREE OF SUPPORT FOR RESEARCH PROGRAMS FOR ADVANCED REACTORS",
7/6/77 (CC.
TO C. KELBER).
3.
TO E. G. CASE FROM R. P. DENISE, ",NRR COMMENTS TO ACRS ON ARSR PROGRAM",
7/15/77.
4.
TO SAUL LEVINE, RES, FROM E. CASE NRR, "RES INFORMATION ON THE UNDERSTANDI._ 0F CDA ENERGETICS IN LMFBR'S".
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PROGRAM OFFICE COMMENTS ON POTENTIAL UTItIZATION 09 VALUE OF RESEARCH RESULTS IN THE REGULATO RIL 4:
10 DATE ISSUED:
02/25/77 RES PROGRA!1 ELEMENT: RISK ASSESSMENT / PRIMARY SYSTEMS INTEGRITY PRESSURE VESSEL FAILURE PROBABILITY PREDICTION (OCTAVIA CODE)
RIL TITLE:
SPONSORING OFCICE(S): RES ESS:
1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR:
W. VESELY IS A CCMPUTER CODE FOR PREDICTING FAILURE PROBABILITIES IN REACTCR PRESSURE VESSELS AS A EFFECTS OF
'RES COMMENTS: OCTAVIA FUNCTION OF TEMPERATURES AND PRESSURES THAT MIGHT OCCUR DURING STARTUP AND SHUTDOWN OPERATIO MATERIAL PROPERTIES AND OPERATING AGE WERE ALSO INCLUDED.
0F THE REACTOR VESSEL, COULD BE SIGNIFICANTLY REDUCED WITH CONTINUED AGING (RESULTING FURTHER IMPROVEMENTS IN THE CCDE ARE REPORTED ON IN RIL 12.
REVIEWING THE PROBABILITY OF FAILURE OF REACTOR PRESSURE VESSELS DUE TO SUCH E PLANT IMPROVEMENTS THAT MAY BE NEEDED.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RTL ACTIVITIES REVIEW HELD COMPLETED HELD HEL D ISSUE 2_
IMPLEMENTED OFFICE RESPCHSIBLE......... NRR/SD NRR SCHEDULED COMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 1977 1977 ACTUAL COMPLETI0H DATE....
09/09/77 THE OCTAVIA COMPUTER CODE HAS BEEN USED TO EVALUATE THE PROBABILITY OF NPR COMMENTS ON 09/09/77 _f GRIMES:
2 DESCRIBE APPLICATION TO i 3ULATORY PROCESS:
THE RESULTS OF THE REACTOR VESSEL FAILURE Fr iM OVERPRESSURE TRANSIENTS WHICH CAN OCCUR DURING PWR OPERATION.
ANALYSES INDICATED THAT L TSTING SAFETY MARGINS COULD BE SIGNIFICANTLY REDUCED WITH CONTIN IF THE HISTORICAL FREQUENCY OF OVERPRESSURE EVENTS CONTINUED.
DESCRIBE IMPACT OF RESULTS: THE ANALYSES ENABLED NRR TO CONFIRM, IN A MORE RIGOROUS, QUANTITATIVE MANNER, INITIAL OF THE REACTOR VESSEL TRANSIENTS. THESE DECISIONS RESULTED IN LICENSING DECISIONS 10 REGUCE THE FREQUENCY AND MAXIMUM PRESSURE OF THC MODIFICATIONS TO OPERATING PROCEDURES FOR REDUCING THE FREQUENCY OF TRAKdI!
TECHNICAL SPECIFICATIONS.
COMMENTS / REMARKS: NDHE SD COMMENTS ON 09/13/78.
P.
RANDALL:
TO FAI*"RE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIA THIS PROGRAM HAS PROVIDED INPUT TO HRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF 10 CFR 50 AND CONTRIBUTED DATA USED IN AND LOAgINGS.
PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO IT HAS ALSO PROVIDED PART OF THE THE ASME CODE WHICH WAS INCORPORATED INTO APPENDIX G.t.99 CONCERNING THE EFFECT OF COPPER IMPURITIES ON DATA BASE USED IN REGULATORY GUIDEOF STEEL TO IRRADIATION AND IS EXPECTED TO PROVIDE A SUBST
[
TO REGULATORY GUIDE 1.99.
x-v
-PROGRAM OFFICE COMMENTS ON POTENTIAL OVilYEAYION OR VALUE OF RESEARCH RESUtTS IN THE PEGULATORY PROCESS RIL 8:
11 DATE ISSUED: 09/15/77 RES PROGPAM ELEMENT: RISK ASSESSMENT RIL TITLE: IEEE NUCLEAR RELIABILITY DATA MANUAL SPONSORING OFFICE (S): RES Rgg: NONE RESEARCH PROJECT MGR:
J. JOHNSON RES COMMENTS: A FAILURE RATE DATA MANUAL WAS DEVELOPED WHICH CAN BE USED IN RISK AND RELIABILITY ANALYSIS OF REACTOR SYSTEMS.
fHE MANUAL CONTAINS FAILURE RATES AND FAILURE MODE INFORMATION FCR OVER 1,000 ELECTRICAL, ELECTRONIC AND SENSING COMPONENTS USED IN NUCLEAR POWER PLANTS.
A METHOD IS GIVEN FOR COLLECTING AND PRESENTING
- = ,
RELIABILITY DATA FOR QUANTITATIVE RELIABILITY AND AVAILABILITY EVALUATIONS OF SAFETY-RELATED NUCLEAR PLANT
- s, SYSTEMS. UNCERTAINTY BOUNDS ARE ALSO GIVEN FOR EACH ESTIMATE OF A CCMPONENT FAILURE RATE.
THIS WORK IS PART OF A CONTINUING EFFORT TO ESTABLISH AN INTERIN DATA BASE FOR USE IN MEETING NRC NEEDS IN THE ELECTRICAL AND ELECTRONIC AREA UNTIL SIGNIFICANT OPERATING DATA CH COMPONENTS USED IN THE NUCLEAR INDUSTRY BECOME AVAILABLE USER DISCUSSICH POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR NRR SCHEDULED COMPLETION DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 04/30/78 ACTUAL COMPLETION DATE.....
10/25/77 NRR COMMENTS ON 10/25/77, R.
TEDESCO:
DESCRIBE APPLICATION TO REGULATORY PROCESS: THE SUBJECT RIL ANNOUNCED THE AVAILABILITY OF A FAILURE RATE DATA MANUAL.
THIS DATA SUPPLEMENTS OTHER SOURCES OF RELIABILITY DATA SUCH AS THE REACTOR SAFETY STUDY.
SUCH DATA ARE BEING USED IN RELIABILITY STUDIES THAT SUPPORT OR PROVIDE THE BASES FOR LICENSING REQUIREMENTS.
DESCRIBE IMPACT OF RESULTS: THE FIRST STUDIES USING THIS DATA HAVE NOT YET BEEN COMPLETED AND THEREFORE THE IMPACT CANNOT YET BE DETERMINED.
COMMENTS /REMAEX1: THIS DATA MANUAL PROVIDES THE STRUCTURE FOR INCORPORATING NEW OR REVISED DATA AS IT BECOMES AVAILABLE. THE MANUAL NOW CONTAINS UNDIFFERENTIATED HARD AND SOFT DATA WHICH IS A SERIOUS IMPEDIMENT TO THE APPLICATION OF THIS DATA.
THEREFORE, CONTINUING WORK IS REQUIRED TO INCREASE THE CONTENT OF HARD DATA BY INCORPORATING THE DATA DEVELOPED THROUGH SUCH PROGRAMS AS HPRDS.
MPA COMMENTS ON 08/24/78.
L.
ABRAMSON:
THE DATA IS DIFFICULT TO INTERPRET BECAUSE OF THE MANY AMBIGUITIES AND INCONSISTENCIES IN THE DATA MANUAL. THE RELATION OF.THE AGGREGATED FAILURE RATES TO THE COMPONENT FAILURE RATES IS NOT MADE CLEAR AND THEY ARE OFTEN INCONSISTENT. IN PARTICULAR, THE USE OF GEOMETRIC AVERAGING IS NOT JUSTIFIED AND LEADS TO INCONSISTENT RESULTS. THE WAY IN WHICH HIGH AND LOW FAILURE RATES ADD UP IS NOT CONSISTENT WITH THEIR STATED INTERPRETATION AS 95TH AND STH PERCENTILES OF THE FAILURE RATE DISTRIBUTION. THE EQUALITY OF THE HIGH VALUE TO THE MAX VALUE IN MANY TABLES IS ALSO INCONSISTENT WITH THIS INTERPRETATION. THE FAILURE MODE TYPES AND DEFINITIONS GIVEN EXCLUDE SUDDEN PARTIAL FAILURES AND GRADUAL COMPLETE FAILURES, BOTH OF WHICH ARE EXPERIENCED IN PRACTICE.
. l
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZAT10t4 OP VALUE OF PESEAPCH RESULTS IN THE REGULATORY PROCESS
-RTL es 12 DATE ISSUED: 06/16/77 RES PROGRAM ELEMENT: RISK ASSESSMENT / PRIMARY SYSTEMS INTEGRITY
-RTL TITLE: MODIFICATIONS TO PRESSURE VESSEL FAILURE PROBABILITY PREDICTION (OCTAVIA CODE)
SPONSORING OFFICE (S): RES PEG: NONE RESEARCN PROJECT MGR8 W. VESELY RES COMMENTS: MODIFICATIONS IN THE OCTAVIA COMPUTER CODE REPORTED IN RIL #10 WERE MADE.
THESE MODIFICATIONS INCLUDE A CAPABILITY TO HANDLE RESIDUAL STRESS IN A REACTOR PRESSURE VESSEL WHICH CAN EITHER BE CONSTANT, OR VARY WITH FLAW SIZE; THE CODE USER CAN IMPOSE AN UPPER BOUND ON THE VE3SEL TOUGHNESS AND THE CODE HAS THE CAPABILITY TO HANDLE UNCERTAINTIES IN THE TOUGHNESS. USING THE MODIFIED OCTAVIA CODE, THE MEDIAN FAILURE PROBABILITY FOR THE SURRY REACTOR PRESSURE VESSEL WAS CALCULATED TO BE 5 X 10-7 PER VESSEL YEAR FOR AN OPERATING TEMPERATURE OF 110 DEGREES C AND THE CURRENT AGE OF APPROXIMATELY 2.5 YEARS.
THE FAILURE PROBABILITY INCREASES TO 3 X 10-5 PER VESSEL YEAR AFTER 40 YEARS.
USER DISCUSSICN POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTTVITIES REVIEW HEL D COMPLETED HELD HELD ISSUED IMPLEMENTED NRR OFFICE RESPONSIBLE......... NRR/SD
' SCHEDULED COMPLETION DATE.. 08/16/77 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 1977 ACTUAL COMPLETION DATE.09/23/773/77 1977 NRR COMMENTS ON 09/23/77, D.-EISENHUT:
Df1CRIBE APPLICATION TO REGULATORY PROCESS:
(SEE RIL 810) l
-DESCRIBE IMPACT OF RESULTS:
(SEE RIL 810) 1 CCMf1E N T S/ R EM A R K S :
THIS IS A MODIFICATION CF RESULTS TRANSFERRED BY RIL 810, 02-25-77.
SD COMMENTS ON 02/21/78, P.
RANDALL - SEE COMMENTS FOR RIL 810.
THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PRF.DICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES
'AND LOADINGS. THIS FROGRAM HAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR 53 AND CONTRIBUTED DATA USED IN THE ASME CODE WHICH UAS 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 IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS TO REGULATORY GUIDE 1.99.
v n
PROGRAM OFFICE CGMMENVS CH POTENTfAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 4:
13 DATE ISSUED:
11/11/77 RES PROGRAM ELEMENT: PRIMARY SYSTEMS INTEGRITY
' RIL TITL E: RESIDUAL STRESSES IN WELDS SPONSORING OFFICE (S): RES 333:
1-20 VESSEL INTEGRITY PESEARCH PROJECT FGR C. SERPAN RES COMMENTS: A VERIFIED MODEL IS PRESENTED FCR PREDICTING RESIDUAL STRESSES RESULTING FROM THE WELDING OF
- PIPES, AND THE ESTIMATION OF RESIDUAL STRESSES RESULTING FROM WELD REPAIRS OF REACTOR PRESSURE VESSELS.
's s MODEL CAN BE USED IN THE LICENSING PROCESS TO AID IN THE EVALUATION OF CRACKING THAT HAS OCCURRED IN GI THE
- W.
WELDS IN PIPING.
IN THE CORNER REGIONS OF PRESSURE VESSEL N0ZZLES AFTER CRACKS HAVE BEEN REMOVED, AND IN VE USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED
- OFFICE RESPONSIBLE......... NRR/SD SCHEDULED COMPLETION DATE.. 03/3f/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE..... 08/14/78 NRR COMMENTS ON 08/14/78, J.
KNIGHT:
DESCRIBE APPLICATION TO REGULATORY PROCESS:
THE AXISOL CODE COULD BE USED AS AN AID IN EVALUATING RESIDUAL STRESSES IN FLUED HEAD-PROCESS PIPE WELDS OF CONTAINMENT PENETRATION ASSEMBLIES AND GIRTH BUTT W SHOULD ALSO PROVE USEFUL AS AN AID IN DEVLOPING THE NECESSARY DECISIONAL INFORMATION IN ANY SAFETY E IT PROPOSED WELD REPAIRS.
DESCRIBE IMPACT OF RESULTS:
TO IMPROVE WELDING TECHNIQUES AND PROCEDURES.THE COMPUTER CCDE AXISOL COULD EVENTUA AND INDUSTRY, ATTENTION OF VARIOUS ASME GROUPS ENGAGED IN PREPARING CODES AND STANDARDS OH WELDED FA PROCEDURES..
BEFORE FULL APPLICATION OF THE CODE IN THE LICENSING PROCESS WOULD BE APPROPRI THE INABILITY OF THE CODE TO TAKE INTO ACCOUNT THE EFFECTS OF POST-WELD HEAT TREATMENT, AT THIS TIME, IS A DRAWBACK IN UTILIZING THE CODE IN THE LICENSING PROCESS.
SD COMMENTS ON 09/13/78. H. COBB:
THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES
(
AND LOADINGS.
THIS PROGRAM HAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO THE ASME CODE WHICH WAS INCORPORATED INTO APPENDIX G. 10 CFR 50 AND CONTRIBUTED DATA USED IN IT HAS ALSO PROVIDED PART OF THE DATA BASE USED IN REGULATORY GUIDE OF STEEL TO IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE RE TO REGULATORY GUIDE 1.99 A
i PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN ThE REGULATORY FROCESS RTL 4:
14 DATE ISSUED:
11/09/77 RES PROGRAM ELEMENT: SYSTEM ENGINEERING TITLE: PHYSICAL SEPARATION CRITERIA FOR ELECTRICAL CABLE TRAYS (HORIZONTAL OPEN SPACE CONFIGURATION) fRIL I SPONSORING OFFICE (S): RES RRG: 1-23 ELECTRICAL STDS 1 RESEARCH PROJECT MGR R. FEIT FIRE PROTECTION i
l PES CRMMENIS: THE ADEQUACY OF THE REQUIRED SPACING OF ELECTRICAL CABLE TRAYS AT NUCLEAR POWER PLANTS WAS
-EXAMINED TO PREVENT THE SPREAD OF CABLE FIRES. RESULTS INDICATE THAT CURRENTLY USED CRITERIA FOR CABLE TRAY
' SEPARATION APPEAR TO BE ADEQUATE FOR ELECTRICALLY-INITIATED FIRES, BUT THAT CHANGES MAY BE REQUIRED FOR FIRES DUE TO EXTERNAL IGNITIDH SOURCES. THIS WORK IS APPLICABLE TO A VERIFICATION OF REGULATORY GUIDE 1.75. " PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS".
EXPOSURE FIRE TESTING EMPLOYING EXTERNAL FUEL SOURCES WAS CONDUCTED TO PROVIDE DATA FOR THE DEVELOPMENT OF CURRENT NRC STAFF POSITION AS DOCUMENTED IN THE AFPENDIX A TO THE BRANCH TECHNICAL POSITION APCSB 9.5-1, " GUIDELINES FOR FIRE PROTECTION FOR NUCLEAR POWER PLANTS" AND IN THE DRAFT REGUL ATORY GUIDE 1.120, " FIRE PROTECTION GUIDELINES FOR NUCLEAR POWER PLANTS."
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EDST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR,SD SCHEDULED COMPLETION DATE.. 01/09/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
04/26/78 NRR COMMENTS OH 04/26/78. 08/2f/78.
R.
TEDESCO:
TO REGULATORY PROCESS: THE RIL PROVIDES RESULTS OF A COMPLETED PORTION OF THE NRC DESCRIBE APPLICATION FIRE PROTECTION RESEARCH PROGRAM CONDUCTED AT THE SANDI A LABORATORIES REGARDING THE EFFECTS OF CABLE TRAY SEPARATION ON THE PROPAGATION OF ELECTRICALLY INITIATED AND EXPOSURE FIRES.
DESCRIBE IMPACT OF RESULTS: THE INTRODUCTORY PARAGRAPH OF RIL #14 STATES THAT THE RESULTS OF THE SANDIA
" CURRENTLY USED CRITERIA FOR CAELE TRAY SEPARATION APPEAR TO BE ADEQUATE FOR PROGRAM INDICATE THAT ELECTRICALLY INITIATED FIRES BUT THAT CHANGES MAY BE REQUIRED FOR FIRES DUE TO EXTERNAL IGNITION SOURCES."
11, 1977 THAT THE SPECIFIC POINT RAISED IN THAT PARAGRAPH IS DIRECTED RES NOTED IN A CONVERSATION ON NOVEMBER TO THE ADEQUACY OF SOLE RELIANCE ON THE SEPARATION CRITERIA 0F REGULATORY GUIDE 1.75 FOR PROTECTION AGAINST EXPOSURE FIRES. THE FIRE PROTECTION CRITERIA DEVELOPED BY THE STAFF SINCE THE BROWNS FERRY FIRE HAVE RECOGNIZED THAT RELIANCE SHOULD HOT BE PLACED SOLELY ON THE SEPARATION CRITERIA 0F REGULATORY GUIDE 1.75.
DATED NOVEMBER 9, 1977 DN "THE QUESTION OF WHETHER THE PETITION OF THE UNION OF CONCERNED OUR STAFF REPORT, SCIENTISTS RAISES MATTERE. THAT REQUIRE IMMEDI ATE COMMISSION ACTION," INDICATES THE STAFF POSITION THAT THE IEEE-384 AND THE REGULATORY GUIDE 1.75 SEPARATI0H GUIDELINES, AND THE IEEE-383 FIRE RETARDANCY STANDARDS FOR SAFETY CABLES, BY THEMSELVES ARE NOT SUFFICIENT TO PROTECT AGAINST EXPOSURE FIRES. CONSEQUENTLY WE REQUIRE ADDITIONAL MEASURES FOR FIRE PROTECTION SUCH AS: FIRE BARRIERS BETWEEN REDUNDANT DIVISION CABLE TRAYS; FIRE RETARDANT CDATINGS ON CABLING; AUTOMATIC FIRE DETECTION SYSTEMS; AUTOMATIC FIRE EXTINGUISHING SYSTEMS; ADMINISTRATIVE PROCEDURES; AND TRAINING PROGRAMS. THIS POSITION HAS BEEN HELD BY THE STAFF SINCE THE BROWNS FERRY FIRE AND IS REFLECTED IN OUR STANDARD REVIEW PLAN SECTION 9.5.1 (BTP 9.5-1), REVISION.
REG. GUIDE f.120 HAS BEEN ISSUED FOR COMMENT.
THUS, EXPOSURE FIRES ARE REQUIRED TO BE CONSIDERED IN THE FIRE PROTECTION
~
PROGRAMS FOR BOTH OPERATING PLANTS AND PLANTS IN THE LICENSING PROCESS.
IT SHOULD BE NOTED THAT THE RIL ACKNOWLEDGES THAT.JRTHER TESTING WITH ELECTRICALLY INITIATED CABLE TRAY FIRES UNDER DIFFERENT CONDITIONS MAY RESULT IN FULLY DEVELOPED FIRES.
WE CONCLUDE, AS DOES RIL #14, THAT THESE RESEARCH RESULTS CONFIRM THE NEED FOR PROTECTION MEASURES IN TO THE SEPARATION CRITERIA. SINCE SUCH MEASURES ARE INCLUDED IN OUR PRESENT CRITERIA WE CONCLUDE ADDITION THAT THE INFORMATION IN RIL 814 DOES NOT INDICATE THE NEED FOR CHANGES IN OUR FIRE PROTECTION GUIDELINES, BUT THAT IT CONFIRMS THE NEED FOR OUR PLANS TO UPDATE REGULATORY GUIDE 1.75.
COMMENTS /REMAPKS: NONE.
~
___-,_-~___--
2D COMMENTS ON 09/13/78, G.
PIVEN9 ARK THE TESTS HAVE CONFIRMED THE VALIDITY OF GUIDELINES IN REGULATORY GUIDE 1.120 MHICH CALL FOR SEPARATING REDUNDANT SAFETY SYSTEM CABLING BY RATED FIRE BARRIERS.THE TESTS HAVE CONFIRMED THE VALIDITY OF SEPARATION CRITERIA SPECIFIED IN REGULATORY GUIDE 1.75 ONLY WITH RESPECT TO FIRES RESULTING FROM ELECTRICAL FAILURE WITHIN A CABLE TRAY BUNDLE.
HOWEVER, THEY SHCW THE BASIC DEFICIENCIES OF THE SEPARATION REGUIREMENTS WITH RESPECT TO EXPOSURE FIRES COMMON TO MORE THAN ONE REDUNDANT SYSTEM CABLE RUN.
l
gr-'
PROGRAM OFFICE CONMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RE50LTS IN THE REGULATORY PROCESS RIL 8:
15 DATE ISSUED:
12/01/77 RES PROGRAM ELEPENI: PRIMARY SYSTEMS INTEGRITY RYL TITLE: CHARACTERIZATION OF B P FEECWATER H0ZZLE CORNER CRACKS SPONSORING OFFICECS): RES RES: 1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR C. SERPAN RES COMMENTS: PRESSURE LOADING OF CRACKS IN THE INSIDE CORNER OF FEEDWATER INTAKE H0ZZLES FOR BOILING WATER REACTOR (BWR) PRESSURE VESSELS HAVE BEEN CHARACTERIZED. THE CHARACTERIZATION ESTABLISHED THE RELATIONSHIP BETWEEN STRESS-GENERATED PRESSURE AND MEASURABLE CRACK PARAMETERS, IN ORDER TO DETERMINE
. THE GROWTH OF-THE CRACK AND ITS CRITICAL SIZE.
THESE RESULTS CAN BE USED TO CHECK THE CALCULATICHS, BASED ON INTERNAL PRESSURE, FOR THE SAFETY ANALYSIS OF BWR FEECWATER H0ZZLE CORNER CRACKS DURING
- SUBSEQUENT COMBINED CCOLING AND UNLOADING. IT HAS BEEN SHOWN THAT UNDER THE MOST SEVERE CONDITIONS.
THE CRACK CAN PENETRATE NO MORE THAN 1/3 0F THE PLESSURE VESSEL WALL.
THIS MEANS THAT THE VESSEL WILL THUS KEEPING THE CORE COOL AND PROVIDING FOR
-ALWAYS BE CAPABLE OF RETAINING EMERGENCY COOLING WATER, VESSEL FAILURE IS NOT POSSIBLE FOLLOWING WARM PRESTRESSING UNDER CONDITIONS WHERE COLD A SAFE SHUTDOWN.
- THUS, EMERGENCY CORE COOLING WATER IS INJECTED INTO THE HOT PRESSURE VESSEL FOLLOWING A LOSS OF COOLANT ACCIDENT.
USER DISCUSSICH POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... HRR/SD SCHEDULED COMPLETION DATE.. 03/31/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETI0H DATE.....
02/21/78 NRR COMMENTS ON 02/21/78.
B.
GRIMES:
DESCRIBE APPLICATION TO REGULATORY PROCESS: NRR CONCURS THAT THESE DATA CAN BE USED FOR PARTIAL VERIFICATION OF THE APPROXIMATE METHODS USED BY GE IN ESTIMATING THE STRESS INTENSITY FACTORS APPLICABLE TO FEEDWATER AND CRD H0ZZLE CRACKS IN BWR REACTOR VESSELS. THEY CAN ALSO BE USED FOR VERIFICATION OF MORE SOPHISTICATED METHODS OF EVALUATION (SUCH AS FINITE ELEMENT AHALYSIS) IF AND WHEN SUCH METHODS ARE DEVELOPED.
RESCRIBE IMPACT OF RESULTS: WE HAVE MADE A COMPARISON BETWEEN THE GE STRESS INTENSITY CURVE (FIG. 3-26 0F NEDE-21480) FOR THE ONE CASE WHERE A DIRECT COMPARISON CAN BE MADE -- HAMELY, AT AN A/T RATIO SLIGHTLY GREATER 1HAN 0.50, AS PROVIDED BY FIG. 7 0F THE PROGRESS REPORT.
FOR THIS ONE CASE, THERE IS ALMOST EXACT AGREEMENT BETWEEN THE GE CURVE AND THE VPI TEST RESULTS. THE GE CURVE, WHEN CONVERTED TO HORMALIZED STRESS INTENSITY FACTORS, WILL ALSO FRODUCE A CURVE CAS A FUNCTION OF A/T RATIO) WHICH IS QUALITATIVELY SIMILAR TO THAT OF FIG. 7 0F VIP REPORT VIP-E-76-25; IN THIS CASE, EXACT CORRESPONDENCE IS NOT TO BE EXPECTED THE CLOSE BECAUSE OF THE MATERIAL DIFFERENCE IN THE DIAMETER-TO-THICKNESS RATIOS OF VESSELS INVOLVED.
AGREEME"r BETWEEN THE GE CURVE AND THE TEST RESULT FOR THE ONE CASE WHERE A VALID COMPARISON CAN BE MADE ASSURANCE THAT THE STRESS INTENSITY FACTORS USED BY GE IN THEIR EVALUATION ARE REASONABLE APPROXIMATIONS.
PROVIDEb COMMENTS / REMARKS: WE ENCOURAGE THE COMPLETION OF THIS WORK AND PARTICULARLY THE DEVELOPMENT OF VERIFIED i
ANALYTICAL METHODS WHICH WILL PROVIDE AN ASSURED MEANS FOR FUTURE CALCULATION OF SIF'S FOR CRACKS IN COMPLEX l
GEOMETRIES (SUCH AS THROUGH THE USE OF FINITE ELEMENT ANALYSIS).
"SD COMMENTS ON 09/13/78.
G.
RIVENBARK:
THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES i
AND LOADINGS. THIS PROGRAM HAS PROVIDED IHPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF l
PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G To to CFR 50 AND CONTRIBUTED DATA USED IH THE ASME CODE WHICH MAS INCORPORATED INTO APPENDIX G.
IT HAS ALSO PROVIDED PART OF THE DATA BASE USED IN REGULATORf GUIDE 1.99 CONCERHING THE EFFECT OF COPPER IMPURITIES ON SENSITIVITY OF STEEL TO IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIDHS TO REGULATORY GUIDE 1.99..
PROGRAM OFFICE CCM/ENTS ON FOTENTIAL UTILVZATION OR VALUE OF RESEARCF ~EISULTS IN THE REGULATORY PROCESS IRIL #2 16 DATE ISSUED:
12/01/77 PFS PROGRAM ELEMENT: PRIMARY SYSTEMS INTEGRITY RIL TITLE: -WARM PRESTRESSING SPONSORING OFFICE (S):
RES gig:
1-20 VESSEL INTEGRITY RESEARCH PROJECT MGR:
C. SERPAN RES COMMENTS:
THE EFFECT OF COLD EMERGENCY CORE COOLING WATER ON HOT REACTOR PRESSURE VESSELS WAS CONSIDERED.
THE RESULTING THERMAL SHOCK COULD, UNDER " WORST CASE" CCNDITIONS, LEAD TO THE PREDICTION THAT FLAWS IN THE STEEL PRESSURE VESSEL WOULD EXTEND. RESULTS REPORTED HERE PROVIDE A VERIFICATION OF THE " WARM PRESTRESSING" EFFECT WHICH CAN PRECLUDE CRACK EXTENSION WHEN IT OTHERWISE WCULD HAVE BEEN PREDICTED.
TO DESCRIBE THIS EFFECT, ONCE A CRACK IS LDADED WHILE THE MATERI AL IS VERY TOUGH, EXTENSION WILL OCCUR.
NO RAPID USER DISCUSSION POSITION COMMISSIDH ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS ER}T RIL ACTIVITIES P EVI EM HELD COMPLETED HELD HELD ISSUED IMPLEMENTED
-OFFICE RESPONSIBLE......... NRR/SD SCHEDULED COMPLETION DATE.. 02/01/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
02/14/78 NRR COMMENTS ON 02/14/78. 0A/21/78.
B.
GRIMES:
DESCRIBE APPLICATION TO REGULATORY PRAR115:
IF " WARM 'PRESTRESSING" AS DESCRIBED IN THE RIL IS OPERATIVE ON HIGHLY IRRADIATED REACTOR VESSEL STEELS, THERMAL SHOCK ASSOCIATED WITH ECCS INJECTION DURING A LARGE LOCA.THIS MECHANISM COULD PROVIDE DESCRIBE IMPACT OF RESULTS:
INVOLVE REPRESSURIZATION OF THE VESSEL, VENDOR ANALYSES AND NRC EVALUATIONS WOULD BE SIMPLIFIED.BY LI PROVIDE AT LEAST A PARTIAL ANSWER TO THE QUESTIONS POSED IN REG.
IT WOULD ALSO GUIDE 1.2, " THERMAL SHOCK TO REACTOR VESSEL."
CQMMENTS/ REMARKS:
REGARDING WARM PRESTRESSING IS STILL UNDERWAY,NRR HAS DISCUSSE3 THE TECHNICAL ASPECTS COVERED BY THIS RIL WITH RES PERSON RESEARCH TO BE A PROMISING PHENOMENON FOR LIMITING CRACK EXTENSION DURING A THERMAL SHOCK,ESPECIALLY ITS RELEVANCE TO CYLIN TO BE USED AS A BASIS FOR LICENSING DECISIONS.
THE DATA ARE STILL INSUFFICIENT AND THE HRR ASSESSMENT CF ITS RANGE OF APPLICABILITY WILL BE COMPLETED FOLLOWING R SD COMMENTS ON 09/13/78.
G.
RIVENBARK:
THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PREDICTING IkRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES AND LOADINGS.
THIS PROGRAM HAS PROVIDED INPUT TO NRC REQUIREMENTS fdR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO THE ASME CODE WHICH WAS INCORPORATED INTO APPENDIX G. 10 CFR 50 AND CONTRIBJTED DATA USED IN IT HAS ALSO PROVIDED PART OF THE DATA BASE USED IN REGULATORY CUIDE 1.99 CONCERNING THE EFFECT OF COPPER IMPURITIES ON SENSITIVITY OF. STEEL TO IRRADIATION AND IS EXPECTED TO PROVIDE A SUBSTANTIAL INPUT FOR FUTURE REVISIONS TO REGULATORY GUIDE 1.99.
6 __
PROGkAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCN RESULTS VN THE REGULATORY PROCESS RIL 8:
17 DATE ISSUED: 05/05/78 RES PROGRAM ELEMENT: FUEL BEHAVIOR RIL TITLE: POWER BURST FACILITY (PBF) SINGLE ROD-POWER COOLING MISMATCH (PCM) TEST RESULTS SPONSORING OFFICE (S):
-RES ggg:
1-10 PBF EXPERIMENTAL RESEARCH PROJECT MGR:
R. VAN HOUTEN PROGRAMS RES COMMENTS: COMPLETED RESEARCH IS REPORTED ON SINGLE FUEL ELEMENTS EXPOSED TO POWER-COOLING MISMATCH (PCM) CONDITICHS IN THE POWER BURST FACILITY (PBF).
THE RESULTS ARE OFFERED FOR USE IN DETERMINING POSSIBLE CHANGES IN REQUIRED DEPARTURE-FROM-HUCLEATE-BGILING RATIOS (DMER'S) FOR ALL COMMERCIAL POWER REACTORS WHICH USE ZIRCALOY-CLAD URANIUM DIDXIDE FUEL RODS.
THE RESEARCH RESULTS SHOW THAT ZIRCALOY FUEL ROD CLADDING HORMALLY DOES NOT FAIL EVEN WHEN PROLONGED FILM BOILING
,CCCURS AS A RESULT OF INADEQUATE COOLANT FLOW RATES.
THE CLADDING GENERALLY WILL NOT FAIL UNLESS IT BECOMES SO HEAVILY OXIDIZED THAT IT IS BRITTLE AT ROOM TEMPERATURE. SUCH SEVERE ZIRCALOY OXIDATION WOULD REQUIRE HIGHER CLADDING TEMPERATURES THAN ARE CURRENTLY PREDICTED FOR ANY LIGHT WATER REACTOR ACCIDENTS WHICH RESULT IN A PCM.
WHETHER RELATED TO A LOSS OF COOLANT FLOW OR TO AN INCREASE IN FUEL ROD POWER.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE..........NRR/SD SCHEDULED COMPLETION DATE. 07/05/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE....
08/22/78 NRR COMMENTS ON 08/01/78.
D.
HOUSTON:
DESCRIBE APPLICATION TO REGULATORY PROCESS: VARIOUS FUEL FAILURE MECHANISMS AND THE CONSEQUENCES OF FAILURES ARE EVALUATED IN THE SAFETY ANALYSIS OF TRANSIENTS AND ACCIDENTS. DEPARTURE FROM HUCLEATE BOILING (DMB) IS ASSUMED TO PRODUCE FUEL ROD FAILURE AND IS THE FAILURE CRITERICH USED FOR MANY LIC5NSING ANALYSES. PELLET / CLADDING INTERACTION (PCI) CAN ALSO BE A FUEL FAILURE MECHANISM. PBF PROVIDES THE CAPABILITY FOR STUDYING FUEL BEHAVIOR AND FAILURE MECHANISMS UNDER TRANSIENT AND ACCIDENT CONDIT0HS.
DESCRIBE IMPACT OF RESULTS: THE DEMONSTRATED ABILITY OF MOST FUEL RODS TO EXPERIENCE DMB WITHOUT FAILURE SHOWS THAT THE CURRENT DNB CRITERI0H IS CONSERVATIVE. REQUESTS FOR LESS CONSERVATlVE FAILURE CRITERIA HAVE BEEN MADE BY THE INDUSTRY. TURBINE TRIP WITHOUT BYPASS (TTWOB) AND STEAM LINE BREAK (SLB) ARE HEAR-LIMITING EVENTS IN WHICH DNB IS PREDICTED TO OCCUR MOMENTARILY YET FAILURE BY THIS MECHANISM MAY HOT OCCUR.
DEFINITION OF A LESS CONSERVATIVE FAILURE CRITERION FOR THESE EVENTS WOULD RELIEVE THESE LIMITING CONDIT0HS. BASED ON THESE PBF RESULTS, NRR WILL GIVE SERIOUS CONSIDERATION TO THESE VENDOR REQUESTS HOWEVER, APPROVAL OF RELAXED FAILURE CRITERIA WILL BE CONTINGENT UPON THE EVALUATION OF OTHER HOM-DNB FAILURE MECHANISMS.
COMMENTS / REMARKS: THE SINGLE ROD PCM TEST RESULTS SHOW THAT THE CURRENT DNB FAILURE CRITERION IS CONSERVATIVE.
THE WIDE RANGE OF CLADDING TEMPERATURES DURING DNB WHEN TEST PARAMETERS ARE NEARLY THE SAME PREVENT A QUANTITATIVE A3SESSMENT OF THE MARGIN TO FAILURE. FURTHERMORE. THE PAWEL CLADDING EMBRITTLEMENT CRITERION WOULD HEED ADDITIONAL REVIEW BEFORE A QUANTITATIVE MEASURE OF MARGIN COULD BE USED IN LICENSING MATTERS.
THE PCM TEST SERIES WAS DESIGNED PRIMARILY TO STUDY THE EFFECTS OF DNB.
THERFORE. PELLET / CLADDING INTERACTION (PCI) DATA FROM THESE EXPERIMENTS WERE PROBABLY COMPROMISED BY THE TEST CONDITONS. FOR EXAMPLE, FUEL EXPOSURE TIME AT POWER. FUEL PRECONDITIONING AND CLADDING TERMPERATURE WERE NOT IDEAL FOR PCI STUDIES.
COH;,JSIONS ON THE SUBJECT OF FUEL FAILURE PROPAGATION DRAWN FROM THE PCM SERIES MAY BE PREMATURE SINCE VENDOR FUEL ROD PRESSURE CRITERIA HAVE BEEN CHANGED. HEW DESIGN CRITERIA. RECENTLY APPROVED BY HRC. ALLOW FOR INTERNAL ROD PESSURE TO EXCEED THE EXTERNAL SYSTEM PRESSURE DURING HORMAL OPERATIOH. IN ADDITION, THE FUEL ROD BEHAVIOR OF A SINGLE ROD IN A COLD SHROUD IS NOT TYPICAL OF FUEL RODS IN A MULTIPLE ARRAY. THE
- FORTHCOMING BUNDLE TESTS IN PBF SHOULD GIVE MORE INFORMATON ABOUT FAILURE PROPAGATION.
1p COFMENTS, G.
RIVENBARK: NO RESPONSE RECEIVED..
-m b
O
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 8:
18 DATE ISSUED:
11/09/77 RES PROGP AM EL EMENT: RISK ASSESSMENT RTL TITLE: FRANTIC COMPUTER CODE SPONSORING OFFICE (S): RES 1Rg HCNE RESEARCH PROJECT MGR F. GOLDBERG RES COMMENTS: THE FRANTIC COMPUTER CODE IS USED TO CALCULATE THE UNAVAILABILITY OF ANY SYSTEM MODEL. COMPREHENSIVE SURVEILLANCE TESTING EVALUATIONS FOR A SYSTEM ARE POSSIBLE WITH THE INCORPORATION OF TEST DOWNTIMES. TEST INEFFICIENCIES, AND TEST-CAUSED FAILURES IN THE ANALYSIS OF SYSTEM MODELS. THE FRANTIC CODE HAS POTENTIAL SIGNIFICANT APPLICATION IN EVALUATING TECHNICAL SPECIFICATIONS ON TESTING AND ALLOWED DOWNTIMES FOR REACTOR SAFETY SYSTEMS. THE EVALUATIONS CAN BE OF A GENERIC NATURE, OR CAN BE APPLIED TO SPECIFIC PL ANT SYSTEMS IF APPLICABLE DATA ARE AVAILABLE.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL'ACTIVITIE.1 REVIEW HELD OFFICE RESPONSIBLE......... NRR/SD
' f.0MPLETED HELD HELD ISSUED IMPLEMENTED NRR SCHEDULED COMPLETION DATE.
01/09/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED 02/01/78 ACTUAL CGMPLETION DATE.....
12/06/77 01/15/78 HRR COMMENTS ON I?/06/77, D.
EISENHUT DESCRIBE APPLICATION TO REGULATORY PROCESS: PREVIOUS COMMENTS FRCM NRR ARE PRESENTLY UNDER REVISION.
SD COMMENTS, G.
RIVENBARK: NO COMMENT. __. --
I
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE PEGULATOPY PROCESS RIL #:
19 J4TE ISSUED: 01/31/78 RES PROGRAM ELEMENT: RISK ASSESSMENT RIL TITLE: GO METHODOLOGY ASSESSMENT SPONSORING OFFICE (S): RES REG: NONE RESEARCH PROJECT MGR:
J. PITTMAN RIS CCFMENTS: GO PROVIDES A METHOD FOR SYSTEM MODELING AND A COMPUTER CODE TO CALCULATE A PREDICTION OF SYSTEM RELIABILITY. THE STUDY DEMONSTRATES THAT THIS NETHOD PROVIDES EQUIVALENT RESULTS TO THOSE OBTAINED FROM FAULT TREE ANALYSIS, WHICH WAS USED IN THE REACTOR SAFETY STUDY..
THE MODEL RESEMBLES THE SYSTEM SCHEMATIC OR PIPING DIAGRAM WHICH REDUCES THE BURDEN OF MODELING ALL SYSTEM COMPONENTS. GO HAS A POTENTIAL SIGNIFICANT USE AS A MEANS OF DETERMINING SYSTEM RELIABILITY OR AS A DIVERSE METHOD FOR VERIFYING SYSTEM ANALYSIS LERFORMED USING FAULT TREE OR SIMILAR MCDELING TECHNIQUES.
USER DISCUSSION POSITION COMMISSICN ACRS PRESS OFFICE MEETING PAPER ERIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITI(1 EfVlfW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... hRR/SD HMSS SCHEDULED C0f1PLETION DATE.. 03/31/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NRR COMMENTS ON 03/21/78.
D.
ROSS:
SD COMMENTS, G.
RTVENBARK: NO COMMENT.
Nh1SS COMMENTS, B.
HATTER: NO RESPONSE RECEIVED.
n
g PRC3PAM CFFICE COPMENTS CN POTENTIAt UTILIZATION OP VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 2: 20 DATE ISSUED: 01/24/78 RES PROGRAM ELEMFNT: SAFEGUARDS RIL TITLE: A STUDY OF PHYSICAL PROTECTION EQUIPMENT SPONSORING OFFICE (S):
IRE RRQ:
4-4 SAFEGUARDS EQUIPMENT EVAL.
RESEARCH PROJECT MGR:
E. RICHARD RES COMMENTS: THE PURPOSE OF THIS STUDY WAS TO PROVIDE THE NRC INSPECTOR, LICENSING REVIEWER AND FIELD EVALUATOR WlIH NEW AND IMPROVED METHODS FOR EVALUATING PHYSICAL PROTECTION EQUIPMENT THAT IS IN USE OR PROPOSED FOR USE AT LICENSED NUCLEAR FACILITIES. THE FIVE MAJOR PRODUCTS OF THIS STUDY ARE:
1 A CATALOG OF PHYSICAL PROTECTION EQUIPMENT.
2.
A GUIDE FOR EVALUATION OF PHYSICAL PROTECTION EQUIPMENT.
3.
A BOOK OF REFERENCE MATERIALS (RELEVANT TO THE EQUIPMENT CATALOG AND THE EVALUATION GUIDE).
4.
A SET OF GUIDELINES FOR DEVELOPING A METHODOLOGY TO MEASURE LEVELS OF EFFECTIVENESS FOR A FIXED-SITE PHYSICAL PROTECTION SYSTEM.
S.
A
SUMMARY
REPORT, INCLUDING RECOMMENDATIONS FOR FURTHER WORK.
ALL OF THE ABOVE PRODUCTS HAVE BEEN DISTRIBUTED TO THE VARIOUS NRC REGIONAL OFFICES AND ARE PRESENTLY BEING USED BY INSPECTORS AS BASIC REFERENCE DOCUMENTS FOR EVALUATING PHYSICAL PROTECTION EQUIPMENT INSTALLED AT LICENSED NUCLEAR FACILITIES. DATA FROM THESE DOCUMENTS WERE ALSO USED IN THE DEVELOPMENT OF A NEW NRC REGULATORY GUIDE ON INTERIOR INTRUSION DETECTION ALARM SYSTEMS BY THE 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 EQUIPMENT TECHNIQUES. THE RESULTS OF THIS STUDY WILL BE USED IN PHASE II 0F THIS SAFEGUARDS RESEARCH PROGRAM AS A BASIS FOR EXPANDING AND IMPROVING THE DATA AVAILABLE TO NRC STAFF REGARDING THE CHARACTERIS1ICS AND EFFECTIVENESS OF COMBINATIONS OF PHYSICAL PROTECTION EQUIPMENT, AND THEIR ASSOCIATED ADMINISTRATIVE AND OPERATIONAL PROCEDURES.
THE DIVISIDH OF DOCUMENT CONTROL HAS BEEN REQUESTED TO PRINT THESE REPORTS FOR DISTRIBUTION ONLY TO OTHER AGENCIES AND HRC STAFF.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HEL D ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR/NMSS/ --
SD/I3E SCHEDULED COMPLETION DATE.. 03/31/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE....
18E COMMENTS, G.
WEIS: NO RESPONSE RECEIVED.
SD COMMENTS ON 02/22/78.
R.
JONES:
THIS PROGRAM CONDUCTED BY THE MITRE CORPORATION RESULTED IN A MULTI-VOLUME REPORT ON PHYSICAL PROTECTION HARDWARE INCLUDING BFSCRIPTIONS, PERFORMANCE CHARACTERISTICS AND MANUFACTURERS SPECIFICATIONS.
THIS WORK HAS MATEhnamY 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 BEING PREPARED IN CONNECTION WITH THE FUEL CYCLE FACILITY SAFEGUARDS UPGRADE RULE.
HMSS COMMENTS, B.
HATTER: N/A - REPORTS WERE REVIEWED. HMSS PLANS TO REFERENCE PARTS OF THESE REPORTS IN A FORTHCCMING GUIDANCE PACKAGE TO LICENSEES.
THE UPGRADE RULE GUIDANCE DEVELOPMENT WORKING GROUP REFERRED TO THESE DOCUMENTS IN THEIR PREPARATION OF THE FIXED SITE PHYSICAL PkOTECTION UPGRADE RULE GUIDANCE COMPENDIUM.
!NRR COMMENTS ON AUGUST 1,
1979.
F.
PAGANO:
DESCRIBE APPLICATION TO REGULATORY P R 0_C ESS: SEE COMMENTS / REMARKS.
DESCRIBE IdPACT OF PESULTS: SEE COMMENTS / REMARKS.
CONMENTS/ REMARKS:
THE FIVE REPORTS (NUREG-0270, 0271, 0272, 0273, 0274) FORM A THOROUGH, THOUGH NOT EXHAUSTIVE, DESCRIPTION OF PHYSICAL SECURITY HARDWARE AND PROVIDE SOME BASES FOR ITS EVALUATICH. THESE DOCUMENTS PROVIDE THE BASIC FAMILIAP,IZATION WITH AND SOME BASELINE DATA FOR PHYSICAL SECURITY EQUIPf1ENT TO THE LICENSING REVIEWER.
REVIEW FOR C07". MENT INITIATED 08/22/78.
l l
l l
l
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v
^
PROGRAM OFFICE COMNENTS ON POTENTIat UTVLIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 8: 21 DATE ISSUED: 03/24/78 RES PROGRAM ELEMENT: RISK ASSESSMENT
- RTL TITLE:
CRITICAL. REVIEW OF SODIUM HYDROXIDE AEROSOL T0XICITY SPONSORING OFFICE (S): RES PED: NONE RESEARCH PROJECT MGR:
M. CULLINGFORD RES' COMMENTS: THIS WORK CONSISTED PRIMARILY OF A REVIEW OF RELEVANT LITERATUAE (WITH SOME PRELIMINARY SUPPORTIVE ANALYSIS) PERTAINING TO THE T0XICITY OF SODIUM HYDROXIDE (NADH).
ONE INSIGHT HAS BEEN THAT SODIUM IN THE HYDROXIDE FORM, FOLLOWING AN INCIDENT INVOLVING SODIUM RELEASE, MAY NOT EXIST IN SUFFICIENT' AMOUNTS TO WARRANT FURTHER ATTENTION. IN ADDITION, THE CHEMICAL SPECIES THAT.WOULD BE PRESENT IN APPRECIABLE QUANTITIES (NA2CO3) MAY NOT BE OF CONCERN IN TERMS OF HEALTH EFFECTS.
THE PRINCIPAL FINDINGS WHICH SUBSTANTIATE THE ABOVE INSIGHTS ARE:
(A) FOR RELATIVE HUMIDITIES EXCEEDING 35%, IT APPEARS THAT HA0H 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 FQUAL TO ABOUT 100 MG/M3.
THIS TRANSFORMATION WILL TAKE LONGER IF THE RELATIVE HUMIDITY IS LESS THAN 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 DECAHYDRATt INCREASES THE AERODYNAMIC DIAMETER OF THE AEROSOL BY APPROXIMATELY 40%.
THIS INCREASE IN DIAMETER SHIFTS ~SOME OF THE AEROSOL OUT OF THE RESPIRABLE RANGE AND THUS LOWERS THE RESPIRABLE FRACTION O' THE. AEROSOL. HYDROXIDE OR CARBONATE PARTICLES ENTERING THE UPPER RESPIRATORY TRACT WF' A3 SORB WATER AND GROW SO THE RESPIRABLE FRACTION WILL DECREASE.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL As T VIL ~ 3 REVIEW HELD COMPLETED HELD HEL D ISSUED IMPLEMENTED OFFICE RESPONSINc.......-..
NRR SCHEDULED COMPLETION DATEL. 05/29/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLFJIDH DATE 05/01/78 NRR COMMENTSJW 05/01/78, W.
GAMMILL:
DESCRIBE A/PL{ CATION TO REGULATORY PROCESS:
THE REPORT PROVIDES THE LATEST AVAILABLE INFORMATION ON SODIUM HYDROXIDE AEROSOL BEHAVIOR AND ITS T0XICOLOGY. THIS INFORMATION IS NEEDED IN OUR REVIEW ON CONTROL ROOM HABITABILITY AND OFF-SITE CONSE0Eq.cES FOLLOWING A POSTULATED ACCIDENTIAL RELEASE OF SODIUM METAL AND SODIUM FIRE.
DESCRIBE IMPAFi 0F RESULTS: THE REPORT FORMS A BASIS FOR CONSIDERING OTHER SPECIES LESS T0XIC AND MORE READILY FORMED THAN SODIUM HYDROXIDE IN OUR REVIEW OF SODIUM HAZARDS.
CpMMENTS/ REMARKS: AUTHORS RECOMMENDATIONS FOR FURTHER WORK SHOULD BE CONSIDERED WITH RESPECT TO NRR'S NEE! a. -
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS o
RIL #: 23 DATE ISSUED: 04/10/78 RES PROGRAM ELEMENT: SAFEGUARDS RIL TITLE:
"EASI" ADVERSARY SEQUENCE EVALUATION MODEL (COMPUTER GRAPHICS VERSION)
SPONSORING OFFICE (S): HMSS RRH 4-1 EFFECTIVENESS RESEARCH PROJECT MGR:
R.
ROBINSON EVALUATION
.RES COMMENTS: RESEARCH HAS BEEN COMPLETED ON DEVELOPING A GRAPHICS DISPLAY VERSION OF A COMPUTER MODEL CALLED ESTIMATE OF ADVERSARY SEQUENCE INTERRUPTION (EASI), AND RESPONDS TO AN EXPRESSED NEED FOR EVALUATIVE METHODS FOR FIXED-SITE THEFT AND SABOTAGE PREVEHTION SYSTEMS. DOCUMENTATION HAS ALREADY BEEN MADE AVAILABLE THROUGHOUT HRC CONCERNING PROGRAMMABLE POCKET CALCULATOR VERSIONS OF THE EASI MODEL.
THE OBJECTIVE OF THE "EASI" METHOD IS TO PROVIDE A USABLE EVALUATION METHOD WHICH CAN SERVE AS EITHER A PHYSICAL PROTECTION SYSTEM DESIGN AID OR AS A DECISION AID IN THE LICENSING AND INSPECTION PROCESS. THE EASI GRAPHICS PROGRAM AND OBTAIN AS OUTPUi ALLOWS THE USER TO INPUT FACILITY AND ADVERSARY PATH AYTRIBUTES AT A COMPUTER GRAPHICS TERMINAL, A CRT " PERSPECTIVE VIEW" LINE PLOT.
THE METHOD CAN TREAT BOTH THEFT AND SABOTAGE OBJECTIVES BY THREATS OF INSIDERS.
OUTSIDERS, AND COMBINATIONS CF EACH GROUP.
THE RESULTS OF THE EASI ANAL' ISIS ARE EXPRESSED IN TERMS OF THE PROBABILITY THAT THE PHYSICAL PROTECTION SYSTEM INTEEDUPT AN ADVERSARY ALONG A PHYSICAL PATH (ACTION SEQUENCE). TO SUPPLEMENT THE EASI CAN RESPOND IN TIME TO EASI GRAPHICS PkOVIDES THE ANALYST WITH A SELECTIDH OF SIX TWO-DIMENSIONAL AND EIGHT THREE-DIMENSIONAL CALCULATIONS, THESE PLOTS ALLOW THE USER TO EXAMINE THE SENSITIVITIES OF VARIOUS COMPONENTS ALONG THE ADVERSARY'S PATH PLOTS.
AND TO STUDY THE EFFECT ON THE PROBABILITY OF INTERRUPTION OF VARYING THE PERFORMANCE OF THESE COMPONENTS.
I IT IS RECOMMENDED THAT THE EASI METHOD BE USED 3Y HMSS AND OTHER OFFICES AS AN ANCILLARY AID IN DEVELOPING PERFORMANCE-ORIENTED REGULATIONS OR IN CARRYING OUT A COMPREHENSIVE EVALUATION PROGRAM, USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RTL ACTIVITI11 REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... HMSS SCHEDULEC COMPLETION DATE.. 06/10/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
HMSS COMMENTS, B.
HATTER: TEST APPLICATIONS ARE BEING CONDUCTED TO DETERMINE USER SUITABILITY.
HMSS COMMENTS.
G.
LEWIS: THE RESULTS OF THESE EFFORTS HAVE BEEN USED IN THE CURRENT DEVELOPMENT OF IMPROVED EV ALUATIVE METHODS FOR FIXED SITE S AFEGUARDS.
NRR CDMMENTS ON AUGUST 1.
1979.
F.
PAGANO:
pg1C M Ej PA ICATION TO REGULATORY PROCESS: SEE COMMENTS / REMARKS.
plSCRlBE IMPACT OF RESULTS: SEE COMMENTS / REMARKS.
S CQUMENTS/RENARKl: THE KEY TO THE UTILITY OF EVALUATION MODELS IS THE RATIO OF THE EFFORT REQUIRED TO RUN THE MODEL TO THE USEFULNESS OF THE RESULTS OBTAINED. ALTHOUGH SAFEGUARDS EVALUATION MODELS DEVELOPED TO DATE TEND TOWARDS INORDINATE INPUT COMPLEXITIES BY COMPARISON WITH OUTPUT UTILITY, EASI HAS BEEN A WELCOME EXCEPTION TO THIS TREND.
CONSEQUENTLY, EASI HAS BEEN USEFUL AS AN AID TO LICENSING DECISION MAKING, AND HAS BEEN DISTRIBUTED BY NRR TO THE POWER REACTOR LICENSEES AS AN AID IN PHYSICAL SECURITY SYSTEM DESIGN.
THE ADDITIONAL COMPUTER PROGRAMMING WHICH PERMITS THE USE OF COMPUTER GRAPHICS IN CONJUNCTION WITH THE EASI CODE MAY BE USEFUL FOR PARAMETRIC DESIGN
^
STUDIES, ALTHOUGH IT PROVIDES LITTLE ADDITIONAL UTILITY TO THE LICENSING REVIEWER.
ADDITIONS AND FURT3ER VARIATIONS
- OF THIS SIMPLE TOOL APPEAR TO PROVIDE DIMINISHING RETURNS. -
-- o-.-
m m
O PROGRAM OFFICE COMMENTS ON POTENTIAL UTYLIZATVON OR VALUE OF RESEARCH RESULTS VN THE REGULATORY PROCESS PIL #: 24 DATE ISSUED:
04/10/78 RES PROGRAM ELEMENT: SAFEGUARDS RIL TITLE:
"FESEM" ADVERSARY-SEQUENCE EVALUATION MODEL SPONSO'<ING OFFICE (S): HMSS RRg:
4-1 EFFECTIVENESS RESEARCH PROJECT MGR:
R. ROBINSON EVALUATION RE' COMMENTS: RESEARCH HAS BEEN COMPLETED ON THE FORCIBLE ENTRY SAFEGUARDS EFFECTIVENESS MODEL (FESEM), IN RESPONSE TO A NEED FOR EVALUATIVE METHODS FOR FIXED-SITE THEFT AND SABOTAGE PREVENTION' SYSTEMS.
THE PURPOSE OF THIS STUDY WAS TO DEVELOP A METHODOLOGY FOR ANALYZING FIXED-SITE SECURITY SYSTEMS AS TO.THEIR-EFFECTIVENESS AGAINST A FORCIBLE ATTACK BY AN ADVERSARY INTENT ON CREATING AN ACT OF SABOTAGE OR THEFT.
THE MODEL
'PROVIDES A FRAMEWORK FOR PERFORMING INEXPENSIVE EXPERIMENTS-RELATED TO FIXED-SITE SECURITY SYSTEMS, FOR TESTING ALTERNATIVE DECISI0HS, AND FOR DETERMINING THE RELATIVE COST EFFECTIVENESS ASSOCIATED WITH THESE DECISION POLICIES.
'THE RESULTS-OF THE FESEM ANALYSIS INCLUDE ESTIMATES OF THE PROBABILITY OF SABOTAGE OR THEFT WINS (AND LOSSES) BASED ON ATTACK FORCE SIZE, ATTACK MOBILITY, AND TYPE OF ATTACK; COLLECTED STATISTICS ASSOCIATED WITH EACH VARIABLE (E.G.,
HUMBER OF WINS BY DEFENDERS AND BY ATTACKERS FOR SUCCESSFUL SABOTAGE OR THEFT, ALARM TYPES FOR ALL RUNS, TIME REQUIRED FOR SUCCESSFUL SADOTAGE OR COMPLETION OF THEFT, ETC.).
THE PROGRAM IS CURRENTLY AVAILABLE FOR HRC USE VIA AN ACCESS CODE NUMBER TO SANDIA'S COMPUTER. A TRAINING PROGRAM WAS GIVEN IN FEBRUARY 1978 TO INTERESTED HRC PERSONNEL AND POTENTIAL USERS.
IT IS RECOMMENDED THAT THE FESEM MODEL BE USED BY NMSS AND OTHER OFFICES AS AN ANCILLARY AID IN FORMULATING REGULATORY REQUIREMENTS, LICENSING, INSPECTION AND OTHER MONITORING OPERATIONS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER
. BRIEFING BRIEFING RELEASE RESULTS POST RTL ACTIVITIES REVIEW HELD COMPL ETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... HMSS/NRR HRR SCHEDULED COMPLETION DATE.. 06/10/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL' COMPLETION DATE.....
NHSS COMMENTS, B.
HATTER - TEST APPLICATIONS ARE BEING CONDUCTED TO DETERMINE USER SUITABILITY.
MM3s COMMENTS, G.
LEWJS - THE RESULTS OF THESE EFFORTS HAVE BEEN USED IN THE CURRENT DEVELOPMENT OF IMPROVED EVALUATIVE METHODS FOR FIXED SITE SAFEGUARDS.
NRR COMMENTS ON AUGUST 1,
- 1979, F.
PAGANO:
DESCRIBE APPLICATION TO REGULATORY PROCESS: SEE COMMENTS / REMARKS.
.DdSCRIBE IMPACT OF RESULTS: SEE COMMENTS / REMARKS.
COMMENTS /R$ MARKS: THE KEY TO THE UTILITY OF EVALUATION MODOLS IS THE RATIO OF THE EFFORT REQUIRED TO RUN THE MODEL TO THE USEFULNESS OF THE RESULTS. ALTHOUGH THE FESEM OUTPUT PROVIDES INFORMATION WHICH WOULD BE USEFUL IN THE DESIGN AND EVALUATION OF PHYSICAL SECURITY SYSTEMS FOR POWER REACTORS. THE UTILITY OF SUCH INFORMATION (MUCH OF WHICH IS INTUITIVELY OBVIOUS TO A PHYSICAL SECURITY EXPERT) DOES NOT APPEAR TO WARRANT THE EXTENSIVE EFFORT REQUIRED FOR INPUT PREPARATION AND EXECUTION OF THIS MODEL.
IN ADDITION, THE ACCURACY OF THE COMPUTED RESULTS CANNOT BE VERIFIED AS A RESULT OF THE LIMITED APPLICABILITY OF MILITARY-TYPE ENGAGEMENT MODELS TO THE PROBLEM OF SABOTAGE OF A REACTOR FACILITY.
SD COMMENTS ON 09/13/78.
G.
RIVENBARK:
THE RESULTS OF THE EFFORTS WILL BE USED IN DEVELOPMENT OF STANDARDS FOR SAFEGUARDS SYSTEMS AS WELL AS BY LICENSEES IN DEVELOPING THEIR SYSTEMS AND BY HRC IN EVALUATING THOSE SYSTEMS.
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESUtVS YN THE REGUL4 TORY PROCESS RTL #3 '25
.DATE ISSUED: 03/21/78 RES PROGRAM ELEMENT:
FUEL BEHAVIOR RIL TITLE:
FRAP-S3 SPONSORING OFFICEfS): NRR g33:
1-12 FUEL CODE RESEARCH PROJECT MGR:
G. MARINO DEVELOPMENT FRAP-S3 IS A BEST-ESTIMATE COMPUTER CODE THAT CALCULATES THE THERMAL AND MECHANICAL RESPONSE RES COMMENTS AND WAS DEVELOP!' PROVIDE CHARACTERISTICS OF A NUCLEAR FUEL ROD OPERATING UNDER STEADY-STATE POWER CONDITIONS, RELAP.
ACCURATE INITIAL VALUES OF FUEL-ROD PARAMETERS FOR INPUT INTO TRANSIENT ANALYSIS CODES SUCH AS FRAP-T IT IS CAPABLE 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 DEFORMATION-6.
AMOUNT OF PELLET-CLAD INTERACTION (PCI) 7.
FUEL DEFORMATION (SWELLING, DENSIFICATION, RELOCATION, AND THERMAL EXPANSION) 8.
FUEL-CLAD GAP SIZE AND GAP CONDUCTANCE 9.
CLAD-CORROSION AND HYDRIDING.
AND EACH WILL HAVE A LARGE ALL OF THESE QUANTITIES ARE STRONGLY DEPENDENT UPON THE OPERATING HISTORY OF THE ROD, EFFECT ON THE PREDICTED AND MEASURED RESPONSE OF A-FUEL ROD DURING A TRANSIENT. THE CODE, THEREFORE, HAS BEEN DESIGNED TO PROVIDE THESE AND OTHER-QUANTITIES FOR ANY GIVEN POWER HISTORY AS INITIAL CONDITIONS TO THE TRANSIENT CODES.
THE VERIFICATION OF THE FRAP-S3 CODE HAD TWO MAJOR OBJECTIVES:
(1) TO DETERMINE THE CODE PERFORMANCE IN PREDICTING THE AVAILABLE, QUALIFIED, EXPERIMENTAL DATA, AND (2) TO IDENTIFY THOSE AREAS THAT REQUIRE MORE SOPHISTICATED THE CODE PERFORMANCE AND DATA WERE ANALYZED USING STATISTICAL METHODS.
THUS.
MODELING OR MORE EXPERIMENTAL DATA.
THE ALL OF THE MAJOR RESPONSE VARIABLES ARE PRESENTED ALONG WITH THEIR CORRESPONDING STANDARD ERROR BOUNDS.
VERIFICATION PROCEDURE USED INFORMATION FROM OVER 700 FUEL RODS CONTAINING A WIDE RANGE OF OPERATING AND DESIGN PARAMETERS.
USER DISCUSSION POSITIUN COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RE.
3NSIBLE.........
NRR/SD SCHEDULED L3MPLETION DATE.. 05/21/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NRR COMMENTS ON 05/22/78, D.
ROSS:
DESCRIBE APPLICATION TO REGULATORY PROCESS: STEADY-STATE FUEL PERFORMANCE CODES ARE REVIEWED BY HRR AS PART OF LOCA OTHER ACCIDENT ANALYSIS. NRR USES AN INDEPENDENT NRC-DEVELOPED CODE FOR AUDIT PURPOSES IN THESE REVIEWS.
HESCRIBE IMPACT OF Rf}MLTS: NO DIRECT IMPACT. NRR USES THE GAPCON SERIES OF CODES IN AUDIT WORK AS IT HAS DONE SINC BEFORE FRAP-S WAS DEVELOPED. NRR HAS NO PLANS TO USE FRAP-S IN THIS CAPACITY.
COMMENTS / REMARKS: THE FRAP-S CODE WAS DEVELOPED BY RES TO INITIALIZE VARIOUS ANALYSIS CODES. RES AND NRR HAVE RECOGNIZED THE DUPLICATE GAPCON AND FRAP-S CODE EFFORTS AND HAVE CONSOLIDATED THESE EFFORTS INTO A HYBRID CODE FRAPCON FRAPCON WILL CONTAIN SOME ELEMENTS FROM FRAP-S3, BUT FRAP-S DEVELOPMENT WILL BE DISCONTINUED.
CSD COMMENTS, R.
RIVERBANK - THE ENGINEERING METHODOLOGY STANDARDS BRANCH PLANS TO USE THE RESEARCH RESULTS IN ITS ONGOING DEGRADED CORE COOLING TASK AND IN ESTIMATING THE FISSION PRODUCT SOURCE TERM FOR ENVIRONMENTAL ENVELOPE 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. O-o
PROGR AM OFFICE _2274ENYS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE QEGUL4 TORY PROCESS RIL #: 26
.DATE ISSUED:
04/27/78 RES PROGR AM EL EMENT: FUEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS
.RIL TITLE:
THE IMPACT OF 0FFSHORE NUCLEAR GENERATING STATIONS ON RECREATIONAL BEHAVIOR AT ADJACENT COASTAL SITES SPONSORING OFFICE (s): HRR RPQ: 5-21 SOCIO-RESEARCH PROJECT MGR:
D. BARNA ECONOMIC IMPACTS RES COMMENTS: THESE RESULTS ARE OFFERED TO SUPPORT HRC COST-BENEFIT ANALYSTS WITH HEW AND IMPROVED INFORMATION FOR ASSESSING LIKELY IMPACTS OF. NUCLEAR GENERATING STATIONS ON RECREATIONAL BEHAVIOR AT ADJACENT COASTAL SITES.
THE RESEARCH RESULTS INDICATE THAT:
(A) PROXIMITY OF A FLOATING HUCLEAR 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' BEACH PATRONS WOULD AVOID A BEACH AFTER FNP SITING 3 MILES DIRECTLY OFFS 90RE; AND (C) IMPACT OF AN FNP WOULD DECREASE EXPONENTIALLY AS DISTANCE AWAY INCREASED.
IN
SUMMARY
, THE PERCENTAGE REDUCTION IN TOURISM ATTRIBUTABLE TO SITING OF HUCLEAR POWER PLANTS OFFSHORE WOULD BE SMALL, BUT NOT NECESSARILY HEGLIGIBLE, AT POINTS CLOSE BY.
THE STABILITY OF THOSE IMPACTS OVER TIME, HOWEVER, DEPENDS UPON THE STABILITY OF CURRENT ATTITUDES TOWARD AND BELIEFS ABOUT NUCLEAR POWER AND ITS SAFETY.
USER DISCUSSICH POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EDST RIL ACTIVITI_E1 REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR 1
SCHEDULED COMPLETION DATE.. 05/29/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL. COMPLETION DATE.....
04/27/78 NRR COMMENTS ON 04/27/78, M.
ERNST:
DESCRIBE APLICATION TO REGULATORY PROCESS: THIS STUDY IS BEING INCORPORATED INTO STAFF TESTIMONY ON FNP 1-8.
THE TESTIMONY IS CONCERNED WITH POTENTIAL AVOIDANCE OF BEACH RESORTS, BY TOURIST, DUE TO PERCEIVED RISK FROM FLOATING NUCLEAR PLANTS SITED IN THE VICINITY OFFSHORE.
EXTRAPOLATION OF TOURIST BEHAVIOR IN THE VICINITY OF LAND BASED NUCLEAR POWER PLANTS TO OFFSHORE PLANTS IS TENUOUS. THE STUDY ALSO PROVIDED A BROADER GENERIC UNDERSTANDING OF RECREATIONAL BEHAVIOR WHICH WILL STRENGTHEN OUR CAPABILITY TO HANDLE THIS ISSUE IN EIS'S GENERALLY.
DESCRIBE IMPACT OF RESULTS: THIS STUDY HAS RESULTED IN STRONGER, MORE OBJECTIVE STAFF TESTIMONY CONCERNING TOURIST AVOIDANCE OF BEACHES IN THE VICINITY OF FNP'S.
THE ESTIMATES OF POTENTIAL AVOIDANCE HAS ALLOWED CALCULATION OF LIKELY ECONOMIC IMPACT TO A RANGE OF COASTAL ECONOMIES. SATISFACTORY DISPOSITION OF THIS CONTENTION GENERICALLY IN THE FNP HEARINGS SHOULD REDUCE OP. ELIMINATE THIS CONCERN IN FUTURE LICENSING ACTIONS FOR SPECIFIC FHP APPLICATIONS.
COMMENTS / REMARKS: THIS STUDY WAS AN APPLICATION TO NUCLEAR SITUATIONS OF SOCIAL RESEARCH TOOLS DEVELOPED TO ANALYZE HUMAN BEHAVIOR RELATIVE TO RISK FROM HATURAL HAZARDS.
PROGR AM OFFICE C0tiMENTS ON POTENTIAL UTILIZATION OR VALUE 05 RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 4:
27.
DATE ISSUED: 05/02/78 RES PROGRAM ELEMENT: CODE DEVELOPMENT ~
RTL TITLE: " BEACON / MOD 2" SPONSORING OFFICE (S): RES ER$2 1-15 CONTAINMENT CODE RESEARCH PROJECT MGR:
S.
FABIC RFS CcMMENTS: THIS RIL TRANSMITS THE BEACON / MOD 2 COMPUTER CODE MANUAL, DESCRIBES ITS FIELD OF APPLICATION AND DIJ. CUSSES THE CODE'S-STRENGTHS.AND LIMITATIONS. BEACON / MOD 2 IS AN ADVANCED, BEST ESTIMATE CODE INTENDED FOR EVALUATION OF SHORT-TERM THERM 0 HYDRAULIC CONDITIONS WITHIN " DRY" (FULL PRESSURE) MULTICOMPARTMENT CONTAINMENTS, OR WITHIN CERTAIN REGIONS OF THE " PRESSURE SUPPRESSION" DRYWELL.
THIS'RESEARCH WAS INITIATED TO PROVIDE IMPORTANT MODELING IMPROVEMENTS FOR BEST ESTIMATE ANALYSIS OF THESE CONTAINMENT SYSTEMS. BEACON / MOD 2 0FFERS CONSIDERABLE ADVANTAGES OVER THE EXISTING CONTAINMENT CODES FOR BEST ESTIMATE EVALUATION OF HYDRAULIC LOADS IN MULTICOMPARTMENT 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 ALSO SHOWN A CAPABILITY TO DESCRIBE THE EVOLUTION OF A TWO-PHASE (FLASHING) JET AND THE RESULTING PRESSURE LOADS ON THE IMPACTED BARRIER.
IT IS RECOGNIZED THAT THE CODE MUST BE MORE EXTENSIVELY TESTED AGAINST EXPERIMENTAL DATA.
HEVERTHELESS. THE BEACON / MOD 2 IS RECOMMENDED FOR CALCULATIONS OF THE REACTOR CAVITY LOADS, FOR BOTH PWR AND BWR INSTALLATIONS, AND FOR EVALUATION OF JET IMPACT LOADS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS PQiT RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED C0f1PLETION DATE.. 06/10/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NRR COMMENTS, R.
TEDfSCQ:
DESCRIBE _ APPLICATION TO REGULATORY PROCESS: BEACON / MOD 2 IS AN ADVANCED COMPUTER CODE WHICH HAS BEEN DEVELOPED TG COMPUTE SHORT TERM PRESSURE TEMPERATURE RESPONSES FOR EITHER SUBCOMPARTMENTS OR MULTI-COMPARTMENTED DR( CONTAIMMENTS. IT IS A "BEST ESTIMATE" CODE DEVELOPED TO YIELD BETTER RESOLUTION FOR THE SHORT TERM CONTAINMENT RESPONSE.
D.ESCRIBE IMPACT OF RESULTS: WHEN VERIFIED WITH' EXPERIMENTAL RESULTS, BEACON / MOD 2 COULD BE USED IN THE LICENSING PROCESS AS A "BEST ESTIMATE" CODE.
THIS CODE WOULD BE USED IN PLACE OF ' COMPARE' WHEN MORE ACCURATE RESULTS ARE NEEDED. AT THE PRESENT TIME, HOWEVER, THE CODE CANNOT BE USED FOR LICENSING ACTIONS DUE TO THE LACK OF EXPERIMENTAL VERIFICATION.
COMMENTS / REMARKS: THE PRESENT VERSION REQUIRES A SIGNIFICANT AMOUNT OF USER INPUT TO ESTABLISH SUCH PARAMETER AS INTERPHASE MASS TRANSFER, MOMENTUM EXCHANGE AND HEAT TRANSFER BETWEEN COMPONENTS. NO GUIDANCF IS PROVIDED FOR THIS INPUT. COMPARISON WITH EXPERIMENTAL DATA IS HEEDED FOR THIS TYPE OF INFORMATION WHICH IS LACKING AT THE PRESENT TIME.
THEREFORE, TO ALLOW USE OF THE IMPROVED CAPABILITY OF THIS CODE, EXTENSIVE VERIFICATICH WITH EXPERIMENTAL DATA IS REQUIRED.
l.
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PROGRAM OFFICE C0FitENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 5: 28 DATE ISSUED: 05/09/78 RES PROGRAM ELEMENT:
FUEL CEhAVIOR RIL TITLE: " MELT / CONCRETE INTERACTIONS" SPONSORING OFFICE (S): RES ERG:
1-13 FUEL MELT RESEARCH PROJECT MGR:
R.
DISALVO RES COMMENTS:
THIS RIL DESCRIBES THE INTER-1 CODE FOR CALCULATING THE EFFECTS OF INTERACTION BETWEEN MOLTEN MATERIALS AND CONCRETE AND THE EXPERIMENTAL DATA BASE FROM WHICH IT WAS DEVELOPED. THIS WORK HAS RESULTED IN AN IMPROVED MODEL BASED ON EXPERIMENTS WITH PROTOTYPICAL MATERIALS.
IN SEPARATE EFFECTS EXPERIMENTS, MONOLITHIC SPECIMENS OF CONCRETE WERE SUBJECTED TO CONTROLLED THERMAL FLUXES IN ORDER TO MEASURE RATES OF EROSION. EROSION IS LINEAR WITH TIME FOR A GIVEN HEAT FLUX, AFTER CORRECTION FOR THERMAL LOSSES THROUGHOUT REFLECTION 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 COMPOSITION OF THE AGGREGATE MATERIAL. THESE DATA ARE NECESSARY TO INTERPRET EROSION RATES OBSERVED IN INTEGRAL EXPERIMENTS.
AS A RESULT OF THE INTEGRAL EXPERIMENTS, IN WHCH PROTOTYPICAL MOLTEN MATERIALS CONTACT CONCRETE, THE FOLLOWING CONCLUSIONS WERE DRAWN:
- EROSION OF CONCRETE IS THERMALLY DCMINATED, WITH INSIGNIFICANT CONTRIBUTIONS FROM MECHANICAL AND CHEMICAL EFFECTS.
- THE PRINCIPAL MECHANISM OF EROSION IS MELTING OF THE BINDING MATERIAL, WITH NO SIGNIFICANT QUALITATIVE DIFFERENCES CAUSED BY CHANGING THE COMPOSITION 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 AMONG THE GASES EVOLVED FROM THE SURFACE OF THE MELT 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 EXPERIMENTS HAVE CULMINATED IN AN ANALYTICAL MODEL (INTER-1) 0F THE MELT / CONCRETE INTERACTION WHICH CAN HELP EXTEND THEIR RANGE OF APPLICABILITY.
WHILE DIRECT EXTRAPOLATION OF THE DATA TO PROTOTYPICAL CONDITIONS MUST ALWAYS BE MADE CAUTIOUSLY, ENOUGH CONFIDENCE HAS BEEN DEVELOPED SO THAT NO FUNDAMENTAL DIFFERENCES IN BEHAVIOR ARE ANTICIPATED IN SCALING TO FULL-SIZE SYSTEMS RESERVATIONS EXIST REGARDING THE APPLICATION OF INTER-1 TO PREDICT SUCH VARIABLES AS THE TIME OF CONTAINMENT MELTTHROUGH OR OVERPRESSURIZATION. THE MODEL CAN BEST BE UTILIZED IN ITS CURRENT FORM TO ESTIMATE THE RELATIVE SIGNIFICANCE OF VARIATIONS IN PARAMETERS SUCH AS MATERIALS, PROPERTIES AND COMPOSITIONS, INTERFACE HEAT TRANSFER COEFFICIENTS, GEOMETRY, ETC.
THE PRIMARY SIGNIFICANCE OF THE WORK DESCRIBED IS THE IMPROVED UNDERSTANDING OF PHYSICAL PHENOMENA. RES RECOGNIZES THAT THE RESULTS ARE UNLIKELY TO HAVE SIGNIFICANT NEAR-TERM IMPACT ON CURRENT LICENSING PROCEDURES.
IT SHOULD, HOWEVER, PROVIDE ADDITIONAL BACKGROUND INFORMATION USEFUL IN ANALYZING REGULATORY ISSUES INVOLVING ACCIDENTS BEYOND DESIGN BASIS EVENTS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR/SD SCHEDULED CCMPLETION DATE.. 08/30/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED l ACTUAL COMPLETION DATE.....
NRR COMMENTS. U, GAMMILL:
DESCRTBE APPLICATION TO REGULATORY PROCESS: CALCULATING THE EFFECTS OF INTERACTIONS BETWEEN MOLTEN MATERIALS p
AND CONCRETE IS DIRECTLY APPL 2 CABLE TO MAKING LICENSING DECISIGHS RELATED TO CONTAINMENT REQUIREMENTS FOR
~ ADVANCED REACTORS (90CH AS L?tFBR'S) TO ENSURE COMPARADILIVY WITH PRESENT DAY LWR NUCLEAR POWER PLANTS.
-THIS RESEARCH PROGRAM hAS IMPROBED OUR ABILITY TO PREDICT THE TIME AND MODE OF CONTAINMENT FAILURE BECAUSE OF IMPROVED UNDERSTANDING OF THE INTERACTIONS BETWEEN MOLTEN CORE MATERIALS AND CONCRETE, SUCH AS THE RATE AND EXTENT OF CONCRETE EROSION, QUANTITY AND TYPE OF GASES RELEASED, FISSION PRODUCTS RELEASE, AND THE EFFECTS OF CONCRETE CHEMICAL COMFOSITION AS DEMONSTRATED IN NUREG-0440, " LIQUID PATHWAY GENERIC STUDY,"
DATED FEBRUARY 1978, THE RESULTS OF THIS RESEARCH PROGRAM ARE ALSO DIRECTLY APPLICAPLE TO EVALUATING QUESTIONS
.0F COMPARABILITY OF CONSEQUENCES OF POSTULATED CORE MELTDOWN EVENTS AT LAND-BASED AfD FLOATING NUCLEAR PLANTS.
DESERISE IMPACT OF RESULTS: -THE IMPACT OF RESULTS OF THIS RESEARC H PROGRAM ON CURRENT LICENSING PROCEDURES REMAINS TO BE EVALUATED. THESE RESEARCH RESULTS HAVE PLAYED AN IMPORTANT ROLE IN NRR'S SAFETY EVALUATION OF THE CLINCH RIVER BREEDER REACTOR (CRBR) AND THE FAST FLUX TEST FACILITY (FFTF), PARTICULARLY IN REGARD TO EVALUATING THE CONSEQUENCES OF POSTULATED CORE MELTDOWN EVENTS AND ESTABLISHING APPROPRIATE CONTAINMENT REQUIRE-MENTS TO ENSURE CCMPARABILITY WITH PRESENT DAY LWR PLANTS.
THESE RESEARCH RESULTS HAVE ALSO HAD AN IMPORTANT IMPACT ON THE STAFF'S CONCLUSIONS CONTAINED IN THE " LIQUID PATHWAY GENERIC STUDY," NUREG-0127. DATED FEBRUARY 1978.
THE STUDY FOUND THAT THE RISKS ASSOCIATED WITH RELEASES TO THE HYDR 0 SPHERE AT.A FLOATING NUCLEAR PLANT (FNP) ARE GREATER THAN THOSE AT A LAND BASED PLANT CLBP)-FOR CORE MELT ACCIDENTS. THE STAFF THEN ASKED THE APPLICANT TO MAKE DESIGN CHANGES IN THE PLANT TO MITIGATE THE CONSEQUENCES OF THIS KIND OF ACCIDENT; SPECIFICALLY, THE STAFF IN THE DRAFT ENVIRONMENTAL STATEMENT, PART III (NUREG-0127), MAY 1978. AND IN A SUBSEQUENT LETTER TO OFFSHORE POWER SYSTEMS (OPS)
(R.P. BALLARD TO A.P. ZECHELLA, JULY 25, 1978) REQUESTED THAT THE CONCRETE PAD BENEATH THE REACTOR VESSEL BE REPL ACED BY SOME MATERIAL THAT PROVIDES INCREASED RESISTANCE TO A MELT THROUGH BY THE REACTOR CORE.
FINALLY, THE RESULTS OF THIS RESEARCH PROGRAM ARE IMPROVING OUR ABILITY TO MAKE QUANTITATIVE RISK ASSESSMENTS ON EXISTING AND PROPOSED NEW TYPES OF REACTORS WHICH WILL ASSIST HRR IN MAKING LICENSING DECISIONS IN REGARD TO DESIGN AND SITING.
COMMENTS / REM A R K S : THE ADVANCED REACTORS ERANCH (NRR) TRANSMITTED COMMENTS ON RIL 828 TO THE FUEL BEHAVIOR RESEARCH BRANCH CRES) IN A MEMORANDUM 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 TO EXPAND THIS PROGRAM TO EXAMINE MELT INTERACTIONS WITH SACRIFICIAL MATERIALS IN CONNECTION WITH THE LICENSING REVIEW OF OPS'S APPLICATION TO MANUFACTURE 8 FLOATING NUCLEAR PLANTS.
ALTHOUGH THIS RESEARCH PROGRAM HAS GREATLY IMPROVED OUR UNDERSTANDING OF THE PHYSICAL PHENOMENA ASSOCIATED WITH MELT / CONCRETE INTERACTIONS, THIS WORK SHOULD CONTINUE WITH THE OBJECTIVE OF EXPERIMENTAL VERIFICATION OF THE CORE MELT / CONCRETE INTERACTION COMPUTER MODEL (INTER), SUCH THAT EXTRAPOLATION TO PROTOTYPICAL CONDITIONS CAN BE MADE TO ACCURATELY PREDICT SUCH VARI ABLES AS THE TIME OF CONTAINMENT MELT-THROUGH OR OVERPRESSURIZATION.
NRR HAS BEEN IN CLOSE CCNTACT WITH THE RES STAFF ON RESEARCH RELATED TO MATERIALS INTERACTIONS BETWEEN CORE MELT DFBRIS AND CONCRETE AND MAINTAINS COGNIZANCE OF SUCH WORK IN THE U.S. AND OTHER COUNTRIES.
SD COMMEtTS, G.
RIVENBARK: NO RESPONSE RECEIVED..
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PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL 3: 29 DATE ISSUED:
06/07/78 PES PROGRAM ELEMFNT:
FUEL BEHAVIOR RIL TITLE: " FUEL ROD ANALYSIS COMPUTER CODE:
FRAP-T3" SPONSORING OFFICE (S): RES P3G:
1-12 FUEL CODE RESEARCH PROJECT MGR:
H.
SCOTT DEVELOPMENT RES COMMENTS: THIS RIL TRANSMITS THE RESULTS OF COMPLETED RESEARCH TO PREPARE AND TEST THE THIRD MODIFICATION OF THE COMPUTER CODE FRAP-T (FUEL ROD ANALYSIS PROGRAM - TRANSIENT).
FRAP-T IS A FORTRAN IV COMPUTER CODE BEING DEVELOPED TO PREDICT THE TRANSIENT REPONSE OF A LWR FUEL ROD DURING POSTULATED ACCIDENTS SUCH AS LOSS-OF-CCOLANT ACCIDENTS, POWER COOLING MISMATCH ACCIDENTS, REACTIVITY INITIATED ACCIDENTS, OR INLET FLOW BLOCKAGE ACCIDENTS.
FRAP-T IS ALSO BEING DEVELOPED TO PERFORM THE CALCULATIONS NEEDED FOR PLANNING AND ANALYZING POWER BURST FACILITY AND LOSS OF FLUID TEST EXPERIMENTS.
IN FRAP-T3, THE COUPLED EFFECTS OF MECHANICAL, THERMAL, INTERNAL GAS AND MATERI AL PROPERTY RESPONSE ON THE BEHAVIOR OF THE FUEL ROD ARE CONSIDERED. GIVEN APPROPRIATE COOLANT CONDITION AND POWER HISTORIES, FRAP-T3 CAN CALCULATE ROD BEHAVIOR FOR A WIDE VARIETY OF OFF-NORMAL SITUATIONS AND POSTULATED ACCIDENT CONDITIONS (E.G.,
BWR OR PWR POWER TRANSIENTS, FLOW COASTDOWN, LOAD LOSS OR COOLANT DEPRESSURIZATON).
IN THE CONTEXT OF LWR SYSTEM TRANSIENTS FRAP IS WELL SUITED TO TO BE USED AS A COMPONENT CODE TO DESCRIBE FINE DETAILS OF FUEL ROD BEHAVIOR. FURTHERMORE, SENSITIVITY STUDIES WITH FRAP WILL FACILITATE DEFINITION OF THE SIMPLEST ACCEPTABLE FUEL DESCRIPTION IN SYSTEMS CODES.
USER DISCUSSION POSITION COMMISSION ACR5 PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD FELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR/SD SCHEDULED COMPLETICN DATE.. 08/30/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NRR COMMENTS ON 07/31/78, J.
V0GLEWEDE:
DESCRIBE APPLICATION TO REGULATORY PROCf15: TRANSIENT FUEL PERFORMANCE CODES ARE REVIEWED BY NRR AS PART OF LOCA AND OTHER ACCIDENT ANALYSES.
NRR USES INDEPENDENT NRC-DEVELOPED CODES FOR AUDIT PURPOSES IN THESE REVIEWS.
REipRIBE IMPACT OF RESUITS FRAP-T3 IS ONE VERSION OF THE FRAP TRANSIENT FUEL PERFORMANCE CODE.
BECAUSE THIS IS ONLY AN INTERIM VERSIDH OF A BEST-ESTIMATE CODE, IT HAS NOT BEEN ADAPTED FOR LICENSING APPLICATIONS.
IN THE FUTURE, NRR PLANS TO USE THE WATER REACTOR ANALYSIS PACKAGE (WRAP) WHICH WILL INCLUDE A MORE CURRENT VERSION OF FRAP-T THAT INCORPORATES CONSERVATIVE MODIFICATIONS.
IN PRESENT AUDIT WORK, NRR USES THE WATER REACTOR EVALUATION MODEL (WREM) SERIES OF CODES WHICH DOES NOT INCLUDE FRAP-T3.
COMt1ENTS/ REMARKS: THE HEED FOR A MORE DETAILED TRANSIENT FUEL BEHAVIOR CODE IN LICENSING APPLICATIONS HAS LONG BEEN RECOGNIZED BY NRR.
IT IS EXPLCTED THAT A MORE CURRENT VERSION OF THE FRAP-T CODE SERIES WILL BE ADAPTED FOR LICENSING APPLICATICNS. ALTHOUGH CONSERVATIVE MODIFICATIONS WILL BE REQUIRED, THIS CODE WILL CONTAIN MANY OF THE ELEMENTS NOW IN FRAP-T3.
1p COMMENTS, G.
RIV E!!B3_Rii:
RfSCRIBE APPLl$ATION TO REGULATORY PROCESS:
TRANSIENT FUEL PERFORMANCE CODES ARE REVIEWED BY NRR AS PART
- OF LOCA AND OlHER ACCIDENT ANALYSES.
NRR USES INDEPENDENT NRC-DEVELOPED CODES FOR AUDIT PURPOSES IN THESE REVIEWS.
QXSCRIBE IMPACT OF RESULTS:
FRAP-T3 IS ONE VERSION OF THE FRAP TRANSIENT FUEL PERFORMANCE CODE.
BECAUSE THIS IS ONLY AN INTERIM VERSION OF A BEST-ESTIMATE CODE, IT HAS BEEN ADAPTED FOR LICENSING APPLICATIONS..
IN'THE FUTURE, NRR' PLANS'T:0 USE~THE. WATER REACTCR ANALYSIS PACKAGE (WRAP) WHICH WILL INCLUDE A MORE CURQENT!
' VERSION OF FRAP-T THAT INCORPORATES CONSERVATTVE MODIFICATIONS. -IN PRESENT AUDTT WORK, NRR USES THE WATER REACTOR EVALUATICH MODEL (WREM) SERIES OF CODES WHICH DOES'NOT INCLUDE FRAP-T3.
[pMMENTS/REMAR11: THE NEED FOR-A MORE DETAILED' TRANSIENT. FUEL BEHRUIOR CODE.IH LICENSING APPLICATIONS
-HAS LONG BEEN RECOGNIZED.BY-NRR.
IT IS EXPECTED THAT-A MORE CURRENT = VERSION 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. ELEMENTS NOW IN FRAP-T3.
-J f
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PROGRAM OFFICE COMMENVS ON POTENTIAL UTILY2ATION OR 94 TOE OF RESEASCH QESULTS IN THE CEGULATORY PROCESS PIL 8: 30 DATE ISSUED:
06/28/78 RES FROGRAM ELEMENT:
SAFEGUARDS
~
RTL TITLE: PHASE I FINAL REPORT, " BARRIER PENETRATION DATA BASE" OF STUDY, " ASSISTANCE-PHYSICAL' PROTECTION ASSESSMENTS"
' SPONSORING OFFICE (S): NRR RRG: NDHE RESEARCH PROJECT MGR:
J. MILLER RFS COMMENTS: THE REPORTED RESULTS PROVIDED:
(1) A CLASSIFICATION OF BARRIERS IN TERMS OF THE PENETRATION TIME FOR SELECTED COUNTERMEASURES WHICH AN ADVERSARY MIGHT USE TO OVERCOME THE BARRIER, AND (2) PROCEDURES TO BE FOLLOWED IN TESTING REACTOR SITES FOR CGMPLIANCE FOR 10 CFR 73.5S, THE REACTOR SAFEGUARDS REGULATION.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EDIT RTL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE., 08/28/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
1[RR COMMENTS ON 08/30/78, J.
M I fdfR.:,
RE1GRIBE APPLICATION TO REGULATORY PROPC'":
AS STATED IN THE RIL, THE BARRIER PENETRATION DATA BASE IS
/VAILABLE AS A SUPPLEMENTAL DATA SOURCL FOR USE IN THE EVALUATION OF LICENSEE FACILITY SAFEGUARD PROGRAMS.
THE DATA PROVIDES A STANDARDIZED BASE OF BARRIER DELAY TIMES AGAINST VARIOUS ADVERSARY COUNTERMEASURES FROM CURRENTLY AVAILABLE LITERATURE. THIS DATA WILL PROVIDE FOR MORE RELIABLE AND CONSISTENT EVALUATION OF SAFEGUARD PROGRAMS BY BOTH THE NRR STAFF AND LICENSEE STAFFS.
DESCRIBE IMPACT OF RESULTS: THIS COMPILATION OF DATA IS USEFUL FOR THE DESIGN OF PHYSICAL SECURITY SYSTEMS BY THE LICENSEES, AS WELL AS IN THE EVALUATIDH OF THE EFFECTIVENESS OF SUCH SYSTEMS BY NRR.
ALTHOUGH THE RESULTS OF THIS PROGRAM WERE NOT AVAILABLE DURING THE DESIGN PHASE OF THE CURRENf UPGRADING OF PHYSICAL SECURITY AT NUCLEAR POWER PLANTS, THIS DATA BASE WILL PROVIDE USEFUL DATA FOR FUTURE APPLICATIONS.
PP0GR AM OFFIC(_COMMEN T S 0*4 POTE;4TIAL UTItIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS R LL. # : 31 DATE ISSUED: 07/10/78 RES PROGRAM ELEMENT: SAFEGUARDS RTL TITLE: " ASSAY OF STANDARD REFERENCE MATERIAL (SRM) 950B" SP0h30 RING OFFICE (S)? SD RRG 4-S MEASUREMENTS 4 RESEADCH PROJECT MGR R.
SHEPARD STANDARDS RES 00MMENIS: THIS RIL TRANSMITS THE RESULTS OF A COMPLETED PHASE OF RESEARCH 0 ' THE ASSAY DETERMINATON OF URANIUM IN STANDARD REFERENCE MATERIAL (SRM) 950B, IN RESPONSE TO A NEED TO IMPROVE THE QUALITY OF MEASUREMENTS MADE ON SPEC'.AL NUCLEAR MATERIAL FOR CONTROL AND ACCOUNTING PURPOSES. THE PURPOSE OF THE WORK WAS TO DEVELOP AND CERTIFY A URANIUM OXIDE (U303) ASSAY STANDARD TO REPLACE THE VIRTUtLLY DEPLETED SRM 950A, USED IN NONDESTRUCTIVE ASSAYS.
THE RESEARCH RESULTS INDICATE THAT THE NEWLY DEVELOPED SRM 950B CALIBRATION STANDARD HAS A CERTIFIED VALUE OF 99.97 +/- 0.32 PERCENT URANIUM OXIDE (U303).
THESE RESULTS ARE EXPECTED TO IMPROVE THE STANDARDIZATICH AND CALIBRATION CAPABILITY OF BOTH NRC FIELD INSPECTORS AND THE NUCLEAR INDUSTRY AS A WHOLE; THEY ARE EXPECTED TO HAVE A SIGNIFICANT NEAR-TERM IMPACT ON CURRENT SD GUIDES THAT WILL ADDRESS THE IMPLEMENTATION OF to CFR 70.57, LICENSEES MEASUREMENT CONTROL PLANS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOST RIL ACTIVITIES PEtifW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... Nf1SS/NRR NRR SCHEDULED COMPLETION DATE.. 08/30/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.,
NMSS COMMENTS, B.
HATTER - TEST APPLICATIONG ARE BEING CONDUCTED TO DETERMINE USER SUITABILITY.
NRR COMMENTS, W G AMf1IL L : NO RESPONSE RECEIVED. -
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL #2 32 pare ISSUED: 08/03/78 RES PROGRAM ELEMENT: FAST BREEDER REACTORS RIL TITLE:
IMPROVEMENTS IH THE AEROSOL BEHAVIOR CODE FOR RADIOLOGICAL ASSESSMENTS OF LMFARS.
SPONSORING OFFICE (S):
RES RRg:
2-7 AEROSOL MODELING RI}gARCH PROJECT MANAGER:
J.
LARKIHS AND PROPERTIES RES COMMENTS: THIS MEMORANDUM TRANSMITS THE RESULTS OF COMPLETED RESEARCH ON THE MEASUREMENT OF SODIUM OXIDE AEROSOL PROPERTIES. SODIUM OXIDE IS THE KEY AEROSOL CONSTITUENT IN POSTULATED SEVERE LMFBR ACCIDENTS.
FOR THE MOST SEVERE POSTUL ATED LMFER ACCIDENT SCENARIOS (HCDA AND CORE MELT), SODIUM-0XIDE AEROSOL REPRESENTS THE -
HIGHEST AIRBORNE MASS CONCENTRATIONS IN THE CONTAINMENT VESSEL AND IS EXPECTED TO DOMINATE AND GOVERN THE BEHAVIOR OF THE FUEL AND FISSION PRODUCT AEROSOL.
THEREFORE, AS A FIRST STEP IN IMPROVING THE AEROSOL BEHAVIOR CODE, HAARM-2, SEPARATE EFFECTS WORK WAS CARRIED OUT-OH SODILM-0XIDE AEROSOL.
THE RESULTS OF THESE SEPARATE EFFECTS MEASUREMENTS HAVE BEEH INCORPORATED INTO THE MODELS OF THE AEROSOL BEHAVIOR CODE HAARM-2, AND TOGETHER WITH SOME ADDITIONAL IMPROVEMENTS USED TO GENERATE A NEW VERSION CALLED HAARM-3.
THE IMPROVED MODELS IN HAARM-3 PROVIDE A MORE REALISTIC DESCRIPTION OF PARTICLE CHARACTERISTICS AND THEREBY ALLOW IMPROVED ESTIMATES OF SODIUM-0XIDE AEROSOL BEHAVIOR DURING A POSTULATED HCDA.
THE HAARM CODE IS USED BY HRR FOR LMFBR SITE RADIOLOGICAL CONSEQUENCE ASSESSMENT.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RTL ACTIVITIES REVIFH HELD COMPLETED HELD HELD ISSUED It1PL EMENT ED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE.
09/15/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NRR COMMENTS, J.
K.
LONG/T. SPEIS:
DESCRIBE APPLICATION TO REGULA10RY PROCESS: CODE DESCRIBED IN THIS RIL, HAARM3, I* USED BY NRR IN CALCL'L ATING THE EFFECTS OF FAST REACTOR ACCIDENTS.
IT IS BASED ON A PREVIOUS CODE HAARM2, NOW MODIFIED TO REFLECT THE RESULTS OF MILLIKAN CELL EXPERIMENTS AT BCL.
THE GENERAL EFFECT OF THE MODIFICATIONS IS CONFIRMED IN LARGE SCALE TESTS AT HEDL.
pESCRIBE IMPACT OF RESULTS: CONFIRMATION OF THE NEW CODE PERMITS A REDUCTION BY AS MUCH AS A FACTOR OF 2-8 IN THE AMOUNT OF RADI0 ACTIVE AEROSOLS LEAKED FROM AN LMFBR CONTAINMENT IN THE EVENT OF A LARGE ACCIDENT.
(RADIOACTIVE GASES ARE NOT CONSIDERED TO BE AFFECTED.)
SOMMENTS/ REMARKS: THE ATTENUATION OF RADI0 ACTIVE AIRBORNE MATERIALS BY AGGLOMERATION AND FALLOUT IS AN IMPORTANT FACTOR IN THE CALCULATION OF ACCIDENT CONSEQUENCES.
EVALUATION AND VERIFICATION OF THIS NATURAL MECHAHISM FOR REDUCING RADI0 ACTIVE EMISSIONS HAS SIGNIFICANCE COMPARABLE TO AN ENGINEERED SAFETY FEATURE.
PROGRAM OFFICE COPMENTS ON POTENTIAL UTILYZ4 TION 02 VAtVE OF RESEARCH RESULYS IN THE REGULATORY PROCESS RIL #:
33 DATE ISSUED:
08/03/78 RES PROGRAM ELEMENT:
FUEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS RTL TITLE: PLUTONIUM ACCIDENT CONTAINER PROGRAM - RESEARCH, DESIGN AND DEVELOPMENT.
SPONSORING OFFICE (S): NMSS PR3: 5-11 PLUTONIUM CONTAINER RESEARCH PROJECT MGR:
W.
LAHS CERTIFICATION RES COMMENTS: RESULTS ARE REPORTED ON THE DESIGH, DEVELOPMENT AND TEST OF THE PAT-1 PLUTONIUM PACKAGE TH/T MEETS THE NRC QUALIFICATION CRITERIA PUB ISHED IN NUREG-0360 " QUALIFICATION CRITERIA TO CERTIFY A PACKAGE FOR AIR TRANSPORT OF PLUTONIUM."
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RE..LTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NMSS SCHEDULED COMPLETION DATE.. 09/15/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
HMSS COMMENTS, R.
CHAPPElt - STATED WORK IS COMPLETE AND NO OTHER ACTIVITY IS EXPECTED.
RESULTS WERE USED IN THE CERTIFICATION OF THE PAT-1 PLUT'JNIUM PACKAGE TO CONGRESS, LICENSING THE PACK %GE FOR USE, AND PROVIDING PROTOTYPES OF THE PACKAGE TO DOE AND IAEA.
e PROGRCM-OFFICE COMMENTS ON POVENTY4L UTYLXZATICM OR vetVE OF RESEARCH RESULTS YN THE REGULATORY PROCESS RIL en 34 DATE ISSUFD: 08/03/78 RES PROGRAM ELEMENT:
FUEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS RIL TITLE: HUCLEAR DECAY DATA FOR RADIONUCLIDES OCCURRING IN ROUTINE RELEASES FROM HUCLEAR FUEL CYCLE FACILITIES.
' SPONSORING OFFICE (S): HRR 3RS:
5-24 RADI0 BIOLOGY AND DOSIMET.1Y 'RESEARCH PROJECT MGR J.'FOULKE RES COMMENTS: THIS IS A TAEULATION OF NUCLEAR DECAY DATA FOR 240 RADIONUCLIDES WHICH MIGHT BE EXPECTED TO OCCUR IN-ROUTINE RELEASES OF EFFLUENTS FROM HUCLEAR FUEL CYCLE FACILITIES. THIS CAN BE USED BY HRR AS A BASIS FOR ESTIMATION OF RADIATION EXPOSURE 10 MAN.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER
. BRIEFING BRIEFING RELEASE RESULTS
.PD}T RIL ACTIVITIES REVIEW HELD COMPLETED HEL D HEL D ISSUED IMPLEMENTED 2
OFFICE RESPCHSIBLE.........
HMSS SCHEDULED COMPLETION DATE.. 09/30/78 UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED
. ACTUAL COMPLETION DATE.....
N/A NRR COMMENTS, R.'H.
VOLLMER:
DESCRIBE APPLICATION TG REGULATORY PROCESS
' DOCUMENT PRESENTS BASIC NUCLEAR DECAY DATA FOR DEVELOPMENT OF ESTIMATES OF RADIOLOGICAL DOSE.
NRR HAS BEEN EMPLOYING THESE DATA AND IN FUTURE UPDATING OF THE DOSE ASSESSMENT METHODOLOGY WILL USE THIS-D00UMENT AS THE PRIMARY REFERENCE.
DESCRIBE IMPACT OF RESULTS: THE DOCUMENT PROVIDES THE BASIC HUCLEAR DECAY DATA IN THE FORM HECESSARY FOR RADIOLOGICAL. ASSESSMENTS. THE DOCUMENT PROVIDES THE STAFF WITH A REFERENCE DOCUMENT TO ENSURE STAFF'S UNIFORM USAGE OF DATA BASIC TO ITS EFFORTS.
COMMENTS / REMARKS: A MACHINE READABLE FILE IS AVAILABLE FROM K. ECKERMAN, HRR/DSE/RAB.
l F
i 4
45 -
-_ ~
PROGRAM OFFICE COM"ENTS ON POTENTIAL tlTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL #2 35
-DATE ISSUED:
09/15/78 RES PROGRAM ELEMENT:
FUEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS RTL TITLE: SFACTOR:
.A COMPUTER CODE FOR CALCULATING DOSE EQUIVALENT TO A TARGET ORGAN PER MICR0 CURIE - DAY RESIDENCE OF A RADIONUCLIDE IN A SOURCE ORGAN SPONSORING OFFICE (S): 'HRR RRG 5-24 RADI0 BIOLOGY &
.RESEARCH PROJECT MGR J. FOULKE DOSIMETRY RES COMMENTS: THE SFACTOR COMPUTER CODE CALCULATES S. THE AVERAGE DOSE EQUIVALENT 1TO EACH OF A SPECIFIED LIST OF
' TARGET ORGANS'PER MICR0 CURIE-DAY RESIDENCE OF A RADIONUCLIDE IN SPECIFIED SOURCE ORGAHS. THE SFACTOR CODE COMPUTES COMP 0HEHTS OF THE DOSE EQUIVALENT FROM ALPHA PARTICLES, ELECTRONS, GAMMA RAYS, FISSIDH FRAGMENTS, AND HEUTRONS.
S-FACTORS CAN BE COMPUTED FOR ANY RADIONUCLIDE FOR WHICH DECAY DATA ARE AVAILABLE.
t USER DISCUSSION POSITION COMMISSIGH ACRS PRESS OFFICE
-MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS
- POST RIL ACTIVITTES REVIEW HELD COMPLETED Bllp HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... HRR SCHEDULED COMPLETI0H DATE.. --
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED-UNSCHED
' ACTUAL COMPLETI0H DATE.....
HRR COMMENTS: HD RESPONSE RECEIVED.
1
)
1,
4
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILYZATION OR VALUE OF RESEARCH RESULYS IN THE REGULATORY PROCESS RIL #: 36 DATE ISSUED:
09/27/78 PES PROGRAM EL EMENT :
FAST BREEDER REACTORS RIL TITLE: EVALUATION OF GENERAL ATOMIC CODES:
OXIDE-3, SORS, TAP, AND RECA SPONSORING OFFICE (S): NRR RRG: 2-12 GAS COOLED REACTOR RESEARCH PROJECT MGR:
J. LARKINS RES COMMENTS: THIS RESEARCH EVALUATED THE APPLICABILITY AND UTILITY OF THE CODES FOR THE ARSR/ GAS COOLED REACTOR SAFETY PROGRAM.
THE OBJECTIVES OF THE EVALUATIONS ALSO INCLUDED AN ASSESSMENT OF THE MODELS AND HUMERICS USED IN THE CODES AND TO NOTE UNDER WHAT CONDITIONS OR FOR WHAT SCENARIOS THE CODES WERE USEFUL.
THE APPLICABILITY AND UTILITY OF THE GAC CODES FOR RSR-HTGR SAFETY PROGRAMS WAS FOUND TO BE VERY LIMITED.
THE OXIDE-3 CODE APPEARS TO HANDLE OXIDATION OF GRAPHITE BY MOSITURE APPROPRIATELY UNDER NORMAL OPERATING CONDITIONS. THE CODE WILL REQUIRE FURTHER EXPERIMENTAL VERIFICATION BEFORE THE LIMITS OF ACCURACY CAN BE ESTABLISHED.
THE SORS CODE IS THE FORM PRESENTED FOR OUR EVALUATION APPEARS TO HAVE SCME SERIOUS DEFICIENCIES IN THE MODELS WHICH PLACES DOUBTS ON THE ANALYSIS PERFORMED WITH THE CODE.
THE TAP AND RECA CODES GIVE GOOD AGREEMENT WITH OTHER ANALYTICAL TOOLS AND APPEAR TO BE USEFUL AND APPROPRIATE FOR THEIR DESIGNED APPLICATIONS. QUANTIFICATION OF THE ACCURACY OF THE CODES WILL REQUIRE FURTHER COMPARISON WITH OPERATING REACTOR CONDITICNS. A VENDOR VERIFICATION PROGRAM FOR THE CODES IS RECOMMENDED.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HEL D ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
URR COMMENTS, R.
L.
TEDESCO SORS -
DESCRIBE APPLICATION TO RFGULATORY PROCESS: THE RIL DESCRIBES WORK PERFORMED BY BNL TO EVALUATE SORS, A CODE DEVELOPMENT BY GENERAL ATOMIC FOR THE ANALYSIS OF FISSION PRODUCT RELEASE FROM HTGR CORES UNDER TRANSIENT CONDITIONS INVOLVING CORE HEATUPS.
A TOPICAL REPORT GA-A12462, WAS SUBMITTED TO NRR FOR REVIEW WITH THE INTENTION THAT, AFTER BEING ACCEPTED, THE CODE COULD BE USED AND REFERENCED FOR SAFETY ANALYSES RELATED TO LICENSING OF COMMERCIAL HTGRS.
RESCRIBE IMPACT OF RESULTS: CONFIRMATORY.
NRR'S REVIEW OF THE SORS REPORT AND THE MAXIMUM HYPOTHETICAL FISSION PRODUCT RELEASE (MHFPR) MODELS AND DATA HAD PROCEEDED FAR ENOUGH AT THE TIME OF CURTAILMENT OF HTGR LICENSING ACTIVITY (1976) TO PERMIT SOME CONCLUSIONS REGARDING THE UTILITY CF THE CODE.
THESE CONCLUSIONS WERE PRESENTED IN THE REACTOR FUELS INPUT TO THE GASSAR ISER (DECEMBER 1S, 1976).
THE STATEMENTS IN RIL #36 SORS ARE IN GENERAL AGREEMENT WITH, AND THUS CONFIRMATORY OF, NRR'S ASSESSMENT IN THE GASSAR ISER.
COMMENTS /RFfARKS: AS INDICATED IN THE GASSAR ISER, FOR LICENSING PURPOSES NRR ADVOCATES THE ASSUMPTION OF INSTANTANEOUS RELEASE OF FISSIDH PRODUCTS FROM FAILED PARTICLES TO THE PRIMARY COOLANT.
IN EFFECT. THIS ASSUMPTION RENDERS MOST THE SORSG MODELS, BECAUSE SORSG MODELS TIME-DEPENDENT FISSION PRODUCT RELEASE AND TRANSPORT.
THIS ASSUMPTION WAS JUSTIFIED IN THE GASSAR ISER PRIMARILY ON THE BASIS OF A POOR DATA BASE FOR FISSION PRODUCT RELEASE AND TRANSPORT IN HTGR FUEL MATERIALS.
THE WORK REPORTED IN Rlt #36 PROVIDES ADDITIONAL SUPPORT FOR THE VALIDITY OF THESE LICENSING ASSUMPTIONS.
IN VIEW OF THE LACK OF HTGR LICENSING ACTIVITY, HOWEVER, NRR ENVISIONS NO FURTHER NEAR-TERM UTILIZATION OF THE SORS INFORMATION IN THE REGULATORY PROCESS.
OXIDE l DESCRIBE _ APPLICATION TO REGULATORY PROCESS: THE RIL DESCRIBES WORK PERFORMED BY BNL TO EVALUATE OXIDE-3, A CODE DEVELOPED BY GENERAL ATOMIC FOR THE ANALYSIS OF HTGR STEAM OR AIR INCRESS ACCIDENTS. A TOPICAL REPORT, GA-A12493, WAS SUBMITTED FOR REVIEW TO NRR WITH THE INTENTION THAT, AFTER BEING ACCEPTED, THE CODE COULD BE USED FOR SAFETY ANALYSES RELATED TO LICENSING OF COMMERCIAL HTGR'S. -
4
- DESCRIBEIMPACTOFPE1ULTI: cCONFIRMATORY.. NRR'S QEVIEW OF THE' OXIDE-3 REPORT HAD PROCEEDED THROUGH THE 2HD ROUND QUESTION ST AGE: BEFORE CURTAILMENT OF GAS-REACTOR LICENSING ACTIV8TY. AT THAT POINT SUFFICIENT ~ REVIEW HAD' BEEN CGHDUCTED:TO ALLOW A PARTIAL EVALUATION OF THE CODE.
THE LICENSING EVALUATION IS PRESENTED IH THE REACTOR FUELS-INPUT TO'THE GASSAR-ISER-(DECMBER~15, 1976).
THE RESULTS OF THE ASSESSMENT REPORTED IN RIL #36 ARE ESSENTIALLY ~
THE SAME AS. AND ThEREFORE CONFIRM, THE NRR ASSESSMENT- 0F-TME OXIDE-3 CODE IN 1976.
4 COMMENTS /REMAPKS: THE COMPARISON AND APPARENT GOOD AGREEMENTlDF OXIDE-3 WITH THE GOPTWO CODE (REPORTED IN RIL e36) DOES NOT PROVIDE CONCLUSIVE EVIDENCE OF -THE ABILITY OF OXIDE-3 TO PERFORM ITS STATED FUNCTION, VIZ.
-TRANSIENT-ANALYSIS OF STEAM AND AIR INGRESS EVENTS, BECAUSE, IN CONTRAST TO DXIDE-3, GOPTWO IS A STEADY-STATE' CODE.
SINCE THE.Tuo CODES WERE: DESIGNED FOR DIFFERENT FUNCTIONS, MEANINGFUL COMPARISON OF THE CODES COULD
=
'SE ACCOMPLISHED OHLY OVER A RELATIVELY' HARROW RANGE OF EVENT CONDITIONS AND ASSUMPTIONS. IN VIEW OF THE LACK OF HTGR LICENSING. ACTIVITY, HRR ENVISIDHS NO HEAR-TERM UTILIZATION OF THE INFORMATION OH' OXIDE-3 TO
.THE. REGULATORY PROCESS, BEYOND THE PARTIAL ~ CONFIRMATION IT PROVIDES. AS DESCRIBED HERE.-
iTAP'AND RECA' '
.. IS PRESENTLY SCHEDULED TO REVIEW THE RECA-3 CODE IN FY 1979.
DESCRIBE APPLICATION TO REGULATORY PROCESS 2: HRR
'THE RESULTS AND CONCLUSIONS OF THE ORNL REPORT'ORNL-HUREG/TM-178, " EVALUATION OF THE GENERAL' ATOMIC CODES TAP AND RECA FOR HTGR ACCIDENT ANALYSIS" WILL BE FACTORED INTO THIS REVIEW.
DESCRIBE IMP ACT OF RESULTS: THE ORHL REPORT IDENTIFIED A HUMBER OF CONCERHS WHICH HRR WILL LOOK INTO DURING THE SCHEDULED REVIEW OF THE RECA:3 CODE. -HRR IS PRESENTLY NOT SCHEDULED TO REVIEW THE TAP CODE.
THE ORNL REVIEW WILL BE STROHGLY RELIED UP0H REGARDING THE INTERFACE BETWEEN THE TAP AND RECA 3 CODES.
COMMENTS / REMARKS: NOME U
e e
PROGRAM OFFICE COFMENTS ON POTENTIAL UTILIZATION 09 VALUE OF RESEARCH RESULTS IN THE REGUL4 TORY PPOCESS PIL s: 37 DATE ISSUED: 09/29/78 PES PROOPAM ELEMENT:
LOFT RTL TITLE: LOFT REACTOR SAFETY PROGRAM RESEARCH RESULTS THRCUGH OCTOBER 1,
1976 SPONSORING OFFICE (S): HRR ggg:
1-1 LOFT RESEARCH PROJECT tqE:
G. MCPHERSON RES COMMENTg: THE LOFT RESEARCH PR00 RAM HAS BEEN DEVELOPED TO PROVIDE EXPERIMENTAL INFORMATION RELEVANT TO THE LICENSING CRITERIA FOR LARGE COMMERICAL PWR'S.
THE MAJOR PORTION OF THIS PROGRAM IS DIRECTED AT AN IMPROVED UNDERSTANDING OF THE LOSS-OF-COOLANT ACCIDENT (LOCA) AND THE PERFORMANCE OF EMERGENCY CORE COOLING SYSTEMS USING THERMAL-HYDRAULIC, CORE PHYSICS, STRUCTURAL AND FUEL BEHAVIOR DATA OBTAINED THROUGH A SERIES OF LOSS-OF-COOLANT-EXPERIMENTS. THIS RIL IS BASED ON DATA OBTAINED FROM THE FIRST SERIES OF EXPERIMENTS, L1, WHICH WAS PERFORMED IN THE ABSENCE OF' NUCLEAR POWER.
IN THE FINAL EXPERIMENT OF THIS SERIES, L1-5, THE CORE WAS IN PLACE, BUT IN A SHUTDOWN CONDITION. CONSEQUENTLY, THE RESULTS DERIVED FROM THESE INVESTIGATIONS ARE APPLICABLE ONLY TO THE THERMAL-HYDRAULIC AND STRUCTURAL PHENOMENA ASSOCIATED WITH THE LOCA WITH THE. EMERGENCY CORE COOLING (ECC) INJECTION. IN GENERAL, THE RESULTS SUPPORT THE CONSERVATIVE INTENT OF THOSE PORTIONS OF THE EVALUATION MODEL REQUIREMENTS CONTAINED IN THE LICENSING CRITERIA WHICH WERE INVESTIGATED IN THE L1 SERIES.
IN PARTICULAR, THE TIME DELAY IN THE DELIVERY OF EMERGENCY CORE COOL ANT TO THE LOWER PLENUM DUE TO THE EFFECT OF CONTACT WITH THE HOT METAL SURFACES (THE HOT WALL EFFECT) WAS FOUND TO BE SMALL (0.5 T0.1.0 S).
BASED ON THE RELATIVE SURFACE AREA TO VOLUME RATIO OF THE-DOWNCOMER, THE HOT WALL EFFECT IN A LPWR SdDULD BE LESS THAN IN LOFT.
THE RESULTS OF THE LOFT L1 SERIEd ABOVE ARE-RECOMf1 ENDED FOR USE BY HRR IN ITS INTERPRETATION AND APPLICATION OF LOCA ECCS EVALUATION MODEL CRITERI A AND
.RELATED CODES.
ALTHOUGH THE DATA ARE LIMITED TO NONNUCLEAR BLOWDOWN CONDITIONS, THE PREDICTIONS TO WHICH DATA COMPARISONS HAVE BEEN MADE HAVE ASSUMED APPROPRIATE INITIAL CONDITIONS AND THEREFORE THE CONCLUSIONS ARE BELIEVED TO BE VALID.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOST RIL ACTIVITIES REVIEW HELD COMPLETED HELO HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE....
NRR COMMENTS, D.
HOUSTON: NO RESPONSE RECEIVED.
SD COMMENTS, K.
WICHMAN -
THE ENGINEERING METHODOLOGY STANDARDS BRANCH WILL USE THESE RESULTS IN THE PHASE II ECCS RULE CHANGE EFFORT.
PROGRAM OFFICE COTMENTS 09 FOTENTIAL UTILIZATION OR VALUE OF RES44RCH RESULTS YH THE PEGULATORY PROCESS RIL as 38 DATE ISSUED:
10/13/78 RES PROGRAM ELEMENT: FAST REACTOR SAFETY RESEARCH' RESULT CF THE INITIAL SERIES OF ACPR EXPERIMENTS ON PRCMPT-BURST EHERGETICS RTL TITLE:
WITH FRESH OXIDE FUEL SPONSORING OFFICE (S): HRR RRS:
2-6 ACCIDENT EHERGETICS RESEARCH PROJECT MGR:
R. WRIGHT' THE RESULTS OF THESE ACPR TESTS ON PROMPT-BURST ENERGETICS GIVE NRR A MUCH IMPROVED DATA BASE RES COMMENTS:
FOR ASSESSING THE WCRK POTENTIAL AND THE RESULTING THREAT TO THE INTEGRITY OF THE PRIMARY SYSTEM AND EVENTUALLY THERMAL INTERACTION BETWEEN MOLTEN OXIDE FUEL AND SODIUM HAS BEEN DEFINITELY DEMONSTRATED. THE OBSERVED HOWEVER. WAS LESS THAN ONE PERCENT.
THE SHOCK-TRIGGERED DELAYED CONVERSI0H OF FUEL THERMAL EHFRGY INTO WORK, FUEL-SODIUM INTERACTIONS UBSERVED IN THESE EXPERIMENTS RAISE QUESTIONS ABOUT THE POSSIBLE OCCURRENCE OF A-LARGE SCALE PROPAGATING FUEL-COOLANT INTERACTION UNDER CORE MELTDOWN CONDITIONS, AS PROPOSED BY BOARD.
THE CURRENT EXPERIMENTAL RESULTS DO NOT, HOWEVER, GIVE INFORMATION ON THE EXTENT (MASS INVOLVEMENT) 0F SUCH A PROPAGATING INTERACTION OR ON THE WORK POTENTIAL OF SUCH AN INTERACTIDH. IT IS RECOMMENDED THAT CONSIDERATI0H BE GIVEH BY NRR TO THESE UNCERTAINTIES AND TO THE CURRENT ABSENCE OF VERIFIED MECHANISTIC MODELS OF FUEL-COOLANT INTERACTIONS WH ASSESSING ACCIDENT WORK POTENTIAL.
'THE EXPAND FRESH FUEL FAILURE CODE HAS BEEN VERIFIED FOR THE RANGE OF CONDITIONS COVERED BY THESE EXPERIMENTS, AND IS THE BEST ACCIDENT ANALYSIS CODE AVAILABLE FOR THESE CONDITIONS. THE PRE-FAILURE IN-CLAD AXIAL FUEL MOTIDH PREDICTED BY EXPAND HAS BEEH VERIFIED BY POST-TEST EXAMINATION IN THESE EXPERIMENTS. THIS FUEL MOTIDH MAY HAVE REACTIVITY EFFECTS THAT ARE SIGNIFICANT IN ASSESSING THE ACCIDENT WORK POTENTIAL AND THE THREAT TO THE INTEGRITY OF THE PRIMARY SYSTEM AND EVENTUALLY THE CONTAIHMENT. EXPAND IS RECOMMENDED TO HRR AS THE BEST AVAIL ABLE TOOL
-FOR SAFETY ASSESSMENT FOR FRESH OXIDE FUEL IN THESE AREAS.
USER DISCUSSION POSITION COMMISSIDH ACRS PRE 3S OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST R1i ACTIVITIE1 REVIEW HELD COMPLETFD HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... hRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NRR COMMENTS ON JULY 19, 1979.
J. MEYER:
DESCRIBE APPLICATION TO REGULATQRY PROCESS:
THERE IS HD DIRECT APPLICATION OF THE RESULTS OF THE INITIAL SERIES OF TO THE REGULATORY PROCESS.
THE EXPERIMENTS ARE FIRST GENERATIDH AND ARE DIFFICULT TO INTEPRET. NOTHING PBE EXPERIMENTS CONCLUSIVE CAN BE SAID FROM THESE EXPERIMENTS REGARDING SHOCK-PRESSURE TRIGGERING OF SIGNIFICANT FUEL COOLANT INTERA pESCRIBE IMPACT OF RESUL.T1:
EXPERIMENTAL RESULTS LEADING TO A BETTER UNDERSTANDING OF FCI PHEHOMEHA AND PROPAGATION MECHANISMS WILL HAVE SIGHIFICANT IMPACT OH UNDERSTANDING OF LOW PROBABILITY HIGH CONSEQUENCE ACCIDENTS AND THEREBY OH LICENSING. THESE EXPERIMENTS, IN AND OF THEMSELVES, CONTRIBUTE LITTLE BUT DO OFFER A STARTING POINT IN THE DIRECTION OF PROVIDING UNDERSTANDING OF FCI PHEHOMENA.
COMMENTS / REMARKS: NDHE l
r
~~~ PROGRAM OFFICE CONMENTS ON POTENTYAL UVILTZAVION OP VALUE OF PESEARCH RESULTS YN THE REGULATORY PROCESS RIL #: 39 DATE TSS05D:
11/27/78 RES PP0 GRAM ELEMENT: CCDE DEVELOPMENT RIL TITLE:
RELAP-4/MC" 6 SPONSORING OFFICE (S): 6RR ggg 1-16 REFERENCE SYSTEM RESEARCH PROJECT MGR:
S. FABIC CODE BFS COMMENTS: MOD 6 RETAINS ALL OF THE CAPABILITY OF PRIOR VERSIONS AND, IN ADDITION, CONTAINS NEW BEST ESTIMATE (BE) BLOWDOWN HEAT TRANSFER MODELS, SE REFLOOD CAPABILITY, AND OTHER MGDIFICATIONS.
RELAP-4/ MOD 6 WAS DEVELOPED TO PROVIDE CAPABILITY FOR A PWR STATISTICAL LOCA STUDY AND FOR SENSITIVITY STUDIES TO ASSESS DATA PERTAINING TO LOCA RULE CHANGES AND TO IMPROVE BLOWDOWN AND REFLCOD UNDERSTANDING AND PROVIDE QUANTITATIVE APPRAISALS OF ECCS.
MOD 6 IS AN EXTENSIDH OF PRIOR LOCA CODE CAPABILITY TO ALLOW MODELING OF LWR AND EXPERIMENTAL FACILITY REFLOOD PHENOMENA, IN ADDITION TO CALCULATION OF THE BLOWDOWN PHASE OF THE EVENT.
RES IS APPLYING THE CODE TO THE UNCERTAINTY STUDY REQUESTED BY NRR, AS WELL AS TO INTERPRETATION OF TEST RESULTS FROM LOCA FACILITIES.
THE RELAP-4/ MOD 6 CODE IS RECOMMENDED FOR THE BEST ESTIMATE CALCULATIONS OF BLOWDOWN AND REFLOOD.
RELAP-4/ MOD 6 IS NOT REC 0" MENDED FOR REFILL ANALYSES BECAUSE OF THE DIFFICULTIES ASSOCIATED WITH THE NON-EQUILIBRIUM PHECHOMENA WHICH ARE NOT MODELED. DOWNCOMER MODELING IS ALSO A PROBLEM, AND THE CODE DOES NOT MODEL HITROGEN FLOW IF THE ACCUMULATOR SHOULD EMPTY DURING THE ECC BYPASS AND REFILL PHASE OF THE EVENT.
AS WITH MOD 5, THE CODE IS VERY SLOW RUNNING DURING REFILL, GIVING CALCULATED RESULTS WHICH ARE NOT SATISFACTORY.
THIS CODE IS NOT RECOMMENDED FOR STEAM GENERATOR TUBE BREAK INVESTIGATIONS, ALTHOUGH USER GUIDELINES ARE BEING DEVELOPED TO ATTEMPT LIMITED INVESTIGATIONS WITH MOD 6.
MOD 6 HAS BEEN SENT TO THE ARGONNE CODE CENTER, FRANCE (NEA FOR EUROPEAN DISTRIBUTION), ITALY. NRC, ORNL AND SANDIA.
IT IS IN USE ON A NUMBER OF HRC-FUNDED PROGRAMS INCLUDING ANALYSIS CF SEMISCALE MODE 3, LOFT, AND PKL, AND PLANS ARE BEING IMPLEMENTED FOR ITS APPLICATION TO STANDARD PROBLEMS. THE FOREIGN AND DOMESTIC RECIPIENTS OF RELAP-4/ MOD 6 WILL BE ADVISED OF THE CODE ASSESSMENT.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PtPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED LNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NRR COMMENTS: HD RESPONSE RECEIVED..
t '
1 PROGRAM OFFICE COM;1ENTS 09 POTE*4TIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL #: 40:
DATE ISSUED:
12/18/78 RES PROGRAM ELEMENT:
FAST REACTOR SAFETY RESEARCH
~ RTL TITLE: 'THE COMPUTER CODE'BRENDA - A COMPUTER PROGRAM FOR THE DYNAMIC SIMULATIDH OF A LIQUID METAL FAST BREEDER. REACTOR. PLANT SPONSORING OFFICE (S): HRR-EEG:
2-13 FAST REACTOR.
RESEARCH PROJECT MGR P. WOOD SYSTEMS CODE AND ACCIDENT ANALYSIS i
EES COMMENTS: BRENDA IS A FAST RUNNING SYSTEM CODE INTENDED TO PROVIDE NRC WITH THE CAPABILITY OF DOING QUICK PARAMETRIC SURVEYS AND SCOPING STUDIES OF NORMAL GPERATING AS WELL AS ACCIDENTAL TRANSIENTS IN LMFBR PLANTS. -THE CODES PROVIDE HRC WITH AH INDEPENDENTLY DERIVED TOOL FOR SAFETY ASSESSMENT.
IT IS CURRENTLY BEING MODIFIED TO MAKE IT APPLICABLE TO PREOPERATIONAL AND STARTUP TESTS IN FFTF.
USER DISCUSSION POSITION' COMMISSION ACRS PRESS OFFICE
- TIEETING PAPER BRIEFING BRIEFING RELEASE RESULTS
- POST RIL ACTIVITIES REVIEW NELD COMPLETED HELD HELD ISSUED IMPLEMENTED
-OFFICE RESPONSIBLE......... HRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NRR COMMENTS: HO RESPONSE -RECEIVED.
e PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEAPCH RESULTS IN YHE REGULATORY PROCESS RIL #2 41 DATE ISSUED:
12/19/73 RES PP0GRAF ELEMENT: ADVANCED REACTOR SAFETY RESEARCH RIL TITLE: LABORATORY TESTING PROCEDURES TO DETERMINE THE CYCLIC STRENGTH OF SOILS SPONSORING OFFICE (S): SD REG:
3-1 HRC/ STATE REGIONAL RESEARCH PROJECT MGR:
H. 'STEUER EARTH SCIENCES RES COMMENTS: WIDELY VARYING RESULTS WERE BEING OBTAINED IN TESTING SOILS FOR LIQUEFACTION POTENTIAL (CYCLIC STRENGTH OF SOILS) BECAUSE STANDARD TEST PROCEDURES DID HOT EXIST.
HENCE. THE SUBJECT STUDY WAS CONDUCTED IN RESPONSE TO THIS FACT, AND BECAUSE STANDARD PROCEDURES ARE HEEDED FOR FUCLEAR POWER PLANT SITE IHVESTIGATIONS. THE RESEARCH PROGRAM WAS CONDUCTED IN CONJUNCTION WITH DEVELOPMENT OF ASTM PERFORMANCE SPECIFICATIONS.
MAJOR SOIL MECHANICS LABORATORIES THROUGHOUT THE WORLD WERE CONTACTED TO DETERMINE THEIR TESTING METHODS AND TO EVALUATE THE CYCLIC TRIAXIAL EQUIPMENT IN USE.
THIS INFORMATION PROVIDED A BASIS TO DEVELOP THE TEST PROCEDURES FOR STRESS-CONTROLLED CYCLIC TRIAXIAL STRENGTH TESTS PRESENTED AS A PERFORMANCE SPECIFICATION SO THAT A GE0 TECHNICAL TESTING LABORATORY CAN
- 1) ENSURE THAT ITS (EST EQUIPMENT AND PROCEDURES MEET REQUIRED STANDARDS, AND 2) CHECK THAT RESULTS AGREE REASONABLY WITH RESULTS OBTAINED BY OTHER LABORATCRIES.
RESEARCH RESULTS PRESENT RECCMMENDED PROCEDURES FOR LIQUEFACTION POTENTIAL TESTING BY USE OF STATE-OF-THE-ART TESTING TECHNIQUES AS A PERFORMANCE SPECIFICATION.
.IT IS RECOMMENDED THAT NUREG-0031 BE USED AS GUIDELINES BY THE OFFICE OF STANDARDS DEVELOPMENT IN DEVELOPMENT OF APPLICABLE. REGULATIONS AND REGULATORY GUIDES, BY THE OFFICE OF NUCLEAR REACTOR REGULATION TO ASSIST IN THE REVIEW OF HUCLEAR POWER PLANT OPERATING APPLICATIONS, AND BY THE APPLICANT AS A STANDARD IN THE ASSESSMENT OF LIQUEFACTION POTENTIAL OF FOUNDATION SOILS AT NUCLEAR POWER PLANT SITES.
USER DISCUSSI0H POSITION COMMISSIGH ACRS PRESS' 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS
_' POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED
.FFICE RESPONSIBLE......... SD SCHEDULED. COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETI0H DATE.....
SD COMMENTS, R.
MINOGUE - THE INFORMATION IN THIS REPORT WAS USED AND REFERENCED IN REGULATORY GUIDE 1.138,
" LABORATORY INVESTIGATIONS OF SOILS FOR ENGINEERING ANALYSIS AND DESIGH OF NUCLEAR POWER PLANTS".
IT WAS ALSO USED IN THE DEVELOPMENT OF A DRAFT REGULATORY GUIDE ON, " PROCEDURES AND CRITERIA FOR ASSESSING SOIL LIQUEFACTION POTENTIAL AT NUCLEAR FACILITY SITES," WHICH WILL BE PUBLISHED FOR COMMENT IN THE NEAR FUTURE.
COPIES OF THE REPORT WERE WIDELY DISTRIBUTED TO MEMBERS OF THE ASTM, D-18 COMMITTEE ON SOIL AND ROCK ENGINEERING AND WILL BE USED AS ONE OF THE BASES FOR DEVELOPING A NATIONALLY ACCEPTABLE DYNAMIC TRIAXIAL TESTING STANDARD FOR SOILS.
.+
. ~.
- -. - ~
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATICH OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS PIL s:- 42 DATE ISSUEDr 12/20/78
'RES PROGRAM ELEMENT: ADVANCED REACTOR SAFETY RESEARCH RIL TITLE: CRITICAL EXPERIMENT PROGRAM FOR NEUTRONICS CODE VERIFICATION SPONSOR 1HG OFFICE (S): NRR PEG:
2-13 FAST REACTOR RESEARCH PROJECT MGR:
P. WOOD--
SYSTEMS CODE AND ACCIDENT AHALYSIS RES COMMENTS: THIS PROGRAM OF DISTORTED GEOMETRY CRITICAL EXPERIMENTS WAS PLANNED AND CARRIED OUT TO PROVIDE BENCHMARKS FOR THE VALIDATION OF THE HEUTRONICS PART OF CODES USED IN SAFETY ANALYSIS SUCH AS SIMMER. A SECOND OBJECTIVE IS TO VALIDATE THE VIM MONTE CARLD CODE FOR USE AS A SECONDARY STANDARD FOR 7ALIDATIDH OF OTHER HEUTRONICS-METHODS.
MELTDOWN CONFIGURATIONS IN LMFBRS CAN BE EXPECTED TO HAVE REGIDHS WITH HIGH FUEL CONCENTRATIONS GIVING EXTREME SPECTRAL CHANGES AND LARGE REGIONS OF VOID GIVING RISE.TO LARGE STREAMING PATHS.
HEUTRONICS METHODS OTHER TH5H MONTE CARLO HAVE DIFFICULTY IN CALCUL ATING THESE CONFIGURATIONS ACCURATELY. A SERIES OF EXPERIMENTS WAS HEEDED TO DETERMIHE THE IMPORTANCE OF THESE DIFFICULTIES AND TO PROVIDE A BASIS FOR IMPROVING THE ACCURACY AND RELIABILITY OF ACCIDENT ANALYSIS METHODS.
THIS PROGRAM HAS DEMONSTRATED THAT DIFFUSION THEORY HEUTRONICS CALCULATIONS WOULD UNDERPREDICT RAMP RATES WHICH MIGHT OCCUR IN A MELTDOWN AND COULD BE NDH-CONSERVATIVE.
THE RESULTS OF THE PROGRAM LEAD TO THE CONCLUSION THAT DIFFUSI0H THEORY CALCULATIONS LEAD TO H0HCONSERVATIVE ESTIMATION OF REACTIVITY GOING FROM THE REFERENCE TO THE SLUMPED CONFIGURATIONS. THE SIGNIFICANCE OF THE RAMP RATE AT PROMPT CRITICAL IN AH HCDA CALCULATIONS HAS BEEN POINTED OUT BY THE HRR STAFF IN NUREG-0122(1).
. DATA REDUCTION AND PREPARATION OF THE FINAL REPORT WILL BE COMPLETED IN FY 1979.
A VIM MONTE CARLO CALCULATION OH CONFIGURATIDH 2 - SODIUM VOIDED TEST ZONE WILL BE MADE AND THE SH CASES NOT SHOWH ON FIGURE 2 0F ZPR-TM-327 WILL BE COMPLETED.
ANALYSIS TO RESOLVE CROSS SECTION DIFFICULTIES THAT CAUSE DIFFERENCES BETWEEN EXPERIMENT AND VIM MONTE CARLD EIGENVALUES IS NEEDED. WHEN AND IF THE CROSS SECTION DIFFICULTIES ARE RESOLVED, IT WOULD BE DESIRABLE TO PREPARE SECONDARY BENCHMARK VIM MONTE CARLO CALCULATIONS USING A HOMOGENIZED MODEL OF THE EXPERIMENTAL CONFIGURATION.
USER DISCUSSI0H POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETI0H DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED
" ACTUAL CGMPLETICH DATE.....
NRR COMMFHTi: HD RESPONSE RECEIVED.
PROGRAM OFFICE COMhENTS ON POTENTIAL UYYLIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL #: 43 DATE ISSUED: '01/')/79 RES PROGR AM EL EMENT:
FAST REACTOR SAFETY RESEARCH RIt TITLE: SUPER SYSTEM CODE, A C?JMPUTER PROGRAM FOR DYNAMIC SIMJLATION OF LMFBR POWER PLANTS' SPONSORING OFFICE (S): HRR RRG 2-13 FAST REACTOR RESEARCH PROJECT MGR:
P. WOOD SYSTEMS CODE AND ACCIDENT ANALYSIS RES COMMENTS: THE SUPER SYSTEMS CODE (SSC-L) IS SPECIFICALLY DIRECTED TO THE ANALYSIS OF THE ADEQUACY OF NATURAL CIRCULATIDH IN SODIUM-COOLED REACTORS TO PREVENT CLAD NELTING.
THE CODE ALSO HAS THE CAPABILITY TO ANALYZE NORMAL OPERATING TRANSIENTS AND LESS SEVERE ACCIDENTS THAT DO NOT BREACH THE INTEGRITY OF THE SYSTEM OR FUEL.
THE CODE IS OPERATIONAL ON THE BNL-CDC-7600 COMPUTER AND CAN BE OPERATED THROUGH THE NRC TERMINALS AT THE PHILLIPS AND WILLSTE BUILDINGS.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
ERR COM[1 HTS, W.
RUSSELL -
DESCRIBE APPLICATION TO REGULATORY PROCESS: THERE IS NO IMMEDIATE APPLICATION OF SSC TO THE REGULATORY PROCESS (SINCE THERE IS NO.LMFBR LICENSING ACTIVITY) EXCEPT FOR THE SSC VERIFICATION ACTIVITIES ASSOCIATED WITH THE FFTF HATURAL CIRCULATION TESTS.
IF LMFBR LICENSING AGAIN BECOMES ACTIVE, IT WILL BE IMPORTANT TO HAVE AN INDEPENDENT, VERIFIED, THERMAL-HYDRAULICS SYSTEMS CODE IN PLACE.
pgSCRIBE IfiPACT OF RESULTS: PRESENT IMPACT IS MINIMAL BECAUSE OF LACK OF LMFBR LICENSING ACTIVITY. THE IMPACT OF THE VERIFICATION ACTIVITIES ASSOCIATED WITH FFTF NATURAL CIRCULATION TESTS MAY BE IMPORTANT DEPENDING ON TEST GUTCOME.
COMMENTS / REMARKS: AT THE PRESENT TIME IANUS AND DEMO ARE THE TWO CODES USED FOR THE THERMAL HYDRAULIC ANALYSIS OF FFTF AND CRBR.
BOTH CODES HAVE BEEN DEVELOPED BY WESTINGHOUSE. DEVELOPMENT OF SSC-L IS A SPECIFIC CONFIRMATORY RESEARCH EFFORT WITH APPLICATION TO FFTF.
THE (P) AND (S) VERSIONS OF SSC HAVE FUTURE APPLICATION FOLLOWING THE DEVELOPMENT OF THE (L) VERSION.
^ -
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS
'RIL 44 DATE ISSUED: 01/04/79 RES PROGRAM ELEMENT:
FUEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS RIL TITLE: RADIATION DOSE TO CONSTRUCTION WORKERS AT OPERATING HUCLEAR PCWER PLANT SITES SPONSORING OFFICF(S): HRR ggg:
5-23 OCCUPATIONAL RESEARCH PROJECT MGR J. FOULKE EXPOSURE AND PROTECTIDH RES COMMENTS: THE STUDY PROVIDES A DATA BASE WHICH ALLOWS A REALISTIC ASSESSMENT OF THE RADIOLOGICAL IMPACT ON CONSTRUCTION WORKERS OF PROPOSED MULTI-UNIT NUCLEAR POWER PLANTS. MEASUREMENTS OF PERSONNEL EXPOSURE OF SEVERAL HUNDRED CONSTRUCTION WORKERS WERE CONDUCTED AT EACH OF THE FOUR SITES WHERE NEW FACILITIES WERE UNDER CONSTRUCTION HEXT TO OPERATING NUCLEAR POWER PLANTS. FOR MOST WORKER GROUPS. THE AVERAGE DOSE EQUIVALENT RATES WERE-MUCH LESS THAH 10 MREM / MONTH GREATER THAH OFF-SITE CONTROLS. CORRELATIONS BETWEEH OPERATING PLANT P0MER LEVELS AND DOSIMETER READINGS WERE GENERALLY POOR INDICATING THAT VARIATIONS IN OTHER RADIATION SOURCES HAD GREATER EFFECT CH THE MEASUREMENTS.
USER DISCUSSI0H POSITION COMMISSICH ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD 1130ED IMPLEMENTED OFFICE RESPONSIBLE.........
NRR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
HRR COMMENTS.
C.
S.
HIHSON -
ELSCR11f APPLICATION TO REGULATORY PROCESS: AS PART OF ITS LICENSING REVIEW PROCESS, THE NRC MUST ASSESS THE ENVIRONMENTAL IMPACT OF THE CONSTRUCTION AND OPERATION OF HUCLEAR POWER PLANTS.
ASSESSMENT OF RADIATICH DOSES TO CONSTRUCTION WORKERS AT OPERATING HUCLEAR POWER PLANT SITES IS PART OF THIS REVIEW.
SINCE A FIRM DATA BASE FOR RADIATION DOSES TO CONSTRUCTION WORKERS WAS NOT AVAILABLE, ENVIRONMENTAL IMPACT STATEMENTS COULD ONLY PROVIDE ROUGH ESTIMATES FOR THE DOSES TO CONSTRUCTION WORKERS.
THE RESULTS OF THIS STUDY PROVIDE RADIATION EXPOSURE DATA SUCH THAT A REASONABLE ENVIRONMENTAL IMPACT STATEMENT CAN BE MADE.
DESCRIBE IMPACT OF RESULTS: UNTIL THE PU3LICATION OF THIS REPORT, THERE HAS BEEN LITTLE DATA WHICH THE STAFF COULD USE IN ASSESSING THE APPLICANT'S ESTIMATE OF CONSTRUCTION WORKER DOSES DURING HUCLEAR POWER PLANT CONSTRUCTIDH. THE RESULTS OF THIS STUDY SHOW THAT MOST COH3TRUCTIDH WORKERS AT OPERATING HUCLEAR POWER PLANTS RECEIVED LESS THAH 10 MILLIREEMS PER MONTH; AND HO WORKER'S ESTIMATED DOSE EXCEEDED 500 MILLIREMS PER YEAR.
THESE RESULTS INDICATE THAT CONSTRUCTION WORKERS AT OPERATING REACTOR SITES ARE HOT LIKELY TO RECEIVE SIGNIFICANT RADIATION DOSES, AND THEREFORE THAT INDIVIDUAL MOHITORING WILL NOT BE REQUIRED. HOWEVER, IT MAY BE APPROPRI ATE TO PLACE DOSIMETERS AT POINTS WHERE THE HIGHEST EXPOSURE RATES ARE TO BE EXPECTED, TO ASSURE THAT UNUSUALLY HIGH EXPOSURE RATES DO NOT GO UNDETECTED.
~
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PROGQAM OFFICE COMMENTS ON POTENTIAL UTILIZATION GR VALUE OF RESEARCH RESULTS IN THE REGULAT02Y PWOCESS PIL'#:
45
.DATE ISSUED:
02/11/79 RES PROGRAM ELEMENT:
FUEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS RIL TITLE: ' THE CONCEPT COMPUTER CODE AND CAPITAL COSTS FOR BOILING WATER REACTOR PLANTS SPONSORING OFFICE (S): HRR EPG: 5-21 SOCIO-ECONOMIC RESEARCH PROJECT MGR:
D. B,.RHA IMPACTS RES COMMENTS: THIS RIL TRANSMITS THE RESULTS OF COMPLETED RESEARCH UPDATING AND EXPANDING THE CONCEPT COMPUTER CODE FOR FORECASTING CAPITAL COSTS OF BOILING WATER REACTOR PLANTS.
IN 1971 THE ATOMIC ENERGY COMMISSIGH AUTHORIZED POWER PLANT INVESTMENT COST STUDIES, WHICH CULMINATED IN THE WAdH-1230 REPORTS PUBLISHED IN 1972.
THEIR PURPOSE WAS TO FACILITATE POLICY AND ECONOMIC DECISIONS ABOUT ELECTRIC GENERATION FACILITIES IN THE PUBLIC AND PRIVATE SECTORS. NATIONAL PRIORITIES ON ENERGY, THE REGULATORY ENVIRONMENT AND THE COST OF LABOR, EQUIPMENT AND MATERIAL HAVE CHANGED SIGNIFICANTLY. THESE CHANGES DICTATED 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 ESTIMATED TOTAL BASE CONSTRUCTION COST FOR THE 1190 MUE BWR REFEREHCE DESIGN IS 9582,748,330 OR $490/KW BASED ON JULY 1,
1976 PRICES.
THE TOTAL BASE CONSTRUCTION COST FOR THE BWR POWER PL ANT ( 1061 MWE NET GUTPUT) REFERENCE IN WASH-1230 WAS APPROXIMATELY $213,000,000 OR $201/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 $/KW.
THE TOTAL DIRECT CRAFT LABOR COST OF APPROXIMATELY $139,500,000 CORRESPONDS TO AN AVERAGE HOURLY RATE OF $12.29.
APPROXIMATELY 11,350,000 CRAFT LABOR MANHOURS AVERAGE ABOUT 9.5 MAHHOURS/KW.
THESE COMPARE TO AVERAGES OF $8.84/ HOUR AND 6.3 MANHOURS /KW RESPECTIVELY FOR THE EARLIER DESIGN REPORTED IN WASH-1230.
USER DISCUSSION POSITIDH COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS
- POST RIL ACTIV[ TIE 1
.EEVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETI0H DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETION DATE.....
NRR COMMENTS.
AD, ENVIR. TECH. -
DESCRIBE APPLICATION TO REGULATORY PROCES1: THE PRINCIPAL APPLICATION OF THE INVESTMENT COST DATA IS TO UPDATE THE CONCEPT COMPUTER CODE.
IN TURN THE CONCEPT CODE IS USED TO ESTIMATE CAPITAL COST FOR DIFFERENT SIZE PLANTS, DIFFERENT REGIONS OF THE COUNTY, DIFFERENT SCHEDULE LENGTH, DIFFERENT COMPLETION DATES, DIFFERENT ESCALATION AND INTEREST RATES.
IN ADDITION TO UPDATING CONCEPT THE DATA IS USED AS A GENERAL REFERENCE FOR SUCH THINGS AS, COST 0F COMPONENTS, QUANTITIES OF MATERIALS USED, TYPE AND QUANTITY OF LABOR, ETC.
DESCRIBE IMPACT OF RESULTS: 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 PLANTS AND ONE OR TWO HUCLEAR UNITS FOR DIFFERENT ESCALATION AND INTEREST RATES.
THE USE OF THE CONCEPT CODE PERMITTED THE STAFF TO PERFORM THESE ANALYSES MORE EFFICIENTLY AND QUICKLY THAN IF THE CONCEPT CODE WERE NOT AVAILABLE.
! COMMENTS / REMARKS: DOE IS FUNDING THE LATEST UPDATE. -
PROGRAM CFFICE CCMMENTS ON POTENTIAt UTILIZATION OR VALUE OF RESEARCH REdULTS IN THE REGULATORY PROCESS RIL #: 46 DATE ISSUED:
02/12/79 RES PROGRAM ELEMENT: SYSTEMS ENGINEERING RIL TITLE: EFFECTIVENESS OF CABLE TRAY COATING !1ATERI ALS & BARRIERS IN RETARDING THE COMBUSTION OF CABLE TRAYS SUBJECTEC TO EXPOSURE FIRES AND IN PREVENTING PROPAGATION BETWEEN CABLE TRAYS HORIZONTAL OPEN SPACE CONFIGURATICH)
SPONSORING OFFICEfS): SD, NRR RPG: 1-23 ELECTRICAL STANDARDS RESEARCH PROJECT PGR:
R.
FEIT
& FIRE PROTECTION RES COMMENTS: DATA IS PRESENTED ON THE EFFECTIVENESS OF SIX FIRE RETARDANT COATING MATERIALS AND BARRIER DESIGNS IN HORIZONTAL OPEN SPACE CONFIGURATIONS THAT ARE IN USE OR BEING CONSIDERED FOR NUCLEAR POWER PLANTS.
AN ACCEPT ABLE TEST METHODOLOGY WAS DEVELOPED BY WHICH PASSIVE FIRE PROTECTION MEASURES CAN BE EVALUATED.
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 HRC IN SITUATIONS WHERE THE DESIGN BASIS FIRE IS SIGNIFICANTLY DIFFERENT THAN THE TEST CASE FIRES.
USER DISCUSSION POSITION CCMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITLE1 REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... SD/NRR SCHEDULED COMPLETION DATE.
ACTUAL COMPLETION DATE.....
SD_COMr1ENTS - THE ENGINEERING METHODOLOGY STANDARDS 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. -
PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VaLUE OF RESEARCH RESUtVS IN THE REGULATORY PROCESS RIL 4: 47 DATE ISSUED: 03/19/79 RES PROGRAM ELEMENT: FPEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS RIL TITLE *- IHREM II: A COMPUTER IMPLEMENTATIDH OF kECENT MODELS FOR ESTIMATING THE DOSE EQUIVALENT TO ORGANS OF MAN FROM AN INHALED OR INGESTED RADIONUCLIDE SPONSORING OFFICF(S): HRR g33: 5-24 RADI0 BIOLOGY RESEARCH PROJECT MGR:
J. FOULKE
& DOSIMETRY
~
RES COMMENTS: THE IHREM II' CODE AS CONTAINED IN NUREG/CR-0114.
ALSD TRANSMITTED IS VOLUME 1 0F A TABULATION OF INTERNAL RADIATION DOSE CONVERSION FACTORS OBTAINED USING THE INREM II CODE.
THIS TABULATICH IS GIVEN IN
" ESTIMATES OF INTERNAL DOSE EQUIVALENT TO 22 TARGET ORGANS FOR RADIONUCLIDES OCCURRING IH ROUTINE RELEASES FROM HUCLEAR FUEL-CYCLE FACILITIES " VOL.
1, N U R EG/ C R- 015 0.
THE CODE COMPUTES REFERENCE ADULT HUMAN DOSE EQUIVALENTS FROM USER-SUPPLIED DOSIMETRIC AND METABOLIC INFORMATION. IN PRINCIPLE, IHREM II APPROACH IS SIMILAR TO THE ONE USED IN WASH-1400.
ORGAN DOSE EQUIVALENT IS COMPUTED AS THE SUM OF CONTRIBUTIONS FROM EACH SOURCE ORGAN WHERE RADI0 ACTIVITY IS ASSUMED PRESENT; CROSS-IRRADIATION EFFECTS CAN BE ASSESSED. DOSE CONVERSION FACTORS IN HUREG/CR-01SO AND UPCOMING /DL. 2 ARE ILLUSTRATIVE OF INREM II CODE USE ONLY.
THEY ARE NOT ENDORSED FOR ADOPTION SINCE SOME OF THE METABOLIC MODELS ARE SUBJECT TO CRITICISM WITH FURTHER DEVdLCPMENT OF ACCEPTABLE METABOLIC MODELS BY ORNL, THE CODE WILL PROVIDE STATE-OF-THE-ART METHODOLOGY FOR DOSE CALCULATIONS.
USER DISCUSSICH POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RTL A CTIVJ T Tf1 REVIEW HELD COMPLETED HEL D HELD ISSUED IMPLFMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETI0H DATE.....
NRR COMMENTS ON JUNE 13. 1979, AD. SITE ANALYSIS:
DESCRIBE APPLICATION TO REGULATORY PROCE113 PRESENT ANALYSES OF RADIATION EXPOSURE SUCH AS PREPARED FOR WASH-1400, GESMO AND IN REACTOR LICENSING ACTIONS HAVE NOT BEEN UNIFORM IN DOSE ESTIMATION METHODOLOGY. THE VARIOUS APPROACHES USED FOR THE ABOVE STUDIES AND ACTIONS REFLECT VARIOUS INTERPRETATIONS OF THE STATE-OF-THE-ART IN DOSE ESTIMATION METHODOLOGY, THE HEEDS OF PART10ULAR STUDIES AND AVAILABILITY OF DOCUMENTED MODELS.
NRC DOES NOT HAVE A BROADLY APPLICABLE, DOCUMENTED, CRITICALLY REVIEWED, STATE-OF-THE-ART DOSE ESTIMATION METHODOLOGY AVAILABLE TO VARIOUS OFFICES FOR THEIR ASSESSMENT HEEDS.
THE CLOSEST IS REGULATORY GUIDE 1.109, REVISIDH 1,
1977, WHICH REPRESENTS A CURRENTLY APPROPRIATE APPLICATION OF A REASONABLY UP-TO-DATE DOSE ASSESSMENT METHODOLOGY AS APPLICABLE TO NUCLEAR POWER PLANT EFFLUENTS. THE OFFICE OF NUCLEAR REGULATORY RESEARCH RECOGNIZED THE NEEDS IN THIS AREA, AUD HAS BEEN FUNDING RESEARCH AT THE HEALTH AND SAFETY RESEARCH DIVISION OF THE CAK RIDGE NATIONAL LABORATORY (ORNL).
THE RESEARCH DESCRIBED 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 VARIOUS ORGANS OF A BODY FROM INTERNALLY DEPOSITED RADICHUCLIDES.
HOWEVER, ITS APPLICATION TO REGULATORY PROCESSES HAS TO WAIT UNTIL SOMt ADDITIDHAL SUPPORTING RESEARCH IS COMPLETD (SEE COMMENTS / REMARKS) MOST OF WHICH ARE ON-GOING AT GRNL FUNDED BY RES AND SOME BENCH-MARK COMPARISONS ARE MADE.
EgSCRIBE IMPACT OF RESULT 1: THE RESULTS HAVE CONFIRMED OUR UNDERSTANDING OF INTERNAL DOSIMETRY MODELING, BUT HAVE NO DIRECT IMPACI ON LICENSING UNTIL AFTER SOME ADDITIONAL STUDY IS PERFORMED.
C0FMENTS/ REMARKS:
ICRP HAS AN ONGOING CONTRACT WITH HEALTH AND SAFETY RESEARCH DIVISION OF ORNL TO DEVELOP REVISED NUCLEAR DATA, S-FACTORS, METABOLIC PARAMETERS, DOSIMETRIC MODELS, ETC. FOR USE IN ICRP'S FORTHCOMING PUBLICATION 830.
IT IS NOW KNOWN THAT THE ORHL INTERNAL DOSIMETRY MODEL FOR ICRP IS SOMEWHAT DIFFERENT FROM THAT FOR HRC (INREM-II)
ALTHOUGH BOTH MODELS ARE BASED ON ICRP'S TASK GROUP LUNG MODEL (1966, 1972) AND EVE'S GI-TRACT MODEL (1966).
THE ORNL MODEL FOR ICRP IS MORE COMPLETE IN THAT IT HAS THE BODY FLUIDS (BLOOD) REPRESENTED IN THE MODEL AS AN ADDITIONAL COMPARTMENT WHICH IS LACKING IN INREM-II.
INCLUSION OF A BLOOD COMPARTMENT IS APPROPRIATE CONCEPTUALLY AS WELL AS FROM VIEW POINT OF REALISTIC MODELING; IT IS LIKELY TO REDUCE THE ORGAN DOSE CONVERSION FACTORS PARTICULARLY FOR THE SHORTER LIVED RADIONUCLIDES DUE TO DECAY DURING PARTIAL HOLD-UP IN THE BLOOD.
BENCHMARKING INREM-II RESULTS = _ _
WITH THE ICRP'S WILL OE PRUDENT BEFORE OUR USING THEM.
THE BEST VALUES OF METABOLIC PARAMETERS REQUIRED AS INPUTS TO RUN INREM-II ARE NOT YET AVAILABLE. THE NOMINAL VALUES OF THESE PARAMETERS USED BY THE AUTHORS OF TWO VOLUMES (VOL. 2 IS IN PREPARATION) 0F DOSE CONVERSICH FACTOR TABULATIONS ARE FOR THE PURPOSES OF ILLUSTRATING THE USE OF INREM-II, AND THE AUTHURS DO CAUTION AGAINST UNCRITICAL USE OF THESE T ABLES IN RADIOLOGICAL APPLICATIONS BECAUSE THEY CONSIDER THESE RESULTS AS PRELIMINARY. RES SHOULD BE REQUESTED TO INCLUDE THE TASK OF GEhERATING AND THEN PERIODICALLY UPDATING A DATA LIBRARY OF BEST VALUES OF METABOLIC PARAMETERS FOR USE IN INREM-II IN THEIR CONTINUING RESEARCH PROGRAMS WITH ORNL.
A SET OF VALUES OF METABOLIC PARAMETERS IS BEING USED IN ICRP WORK AT ORNL.
THE ICRP VALUES ARE NOT YET AVAILABLE TO USE, BUT WHEN AVAILABLE, IT WILL BE PRUDENT TO COMPARE THESE VALUES WITH THE PRELIMINARY VALUES USED BY ORNL FOR NRC, AND MAKE APPROPRIATE REVISION IF NECESSARY.
THE NUCLEAR DATA DEVELOPED AND UTILIZED BY ORNL FOR NRC ARE NOT NECESSARILY THE SAME AS BEING DEVELOPED FOR ICRP.
THOUGH SUBSTANTIAL DIFFERENCES ARE NOT ANTICIPATED IN THIS AREA, IT WILL BE PRUDENT TO COMPARE THESE DATA ALSD.
LASTLY, THE S-FACTORS AND METABOLIC PARAMETERS FOR INDIVIDUALS OF AGE GROUPS DIFFERENT FROM ADULTS ARE STILL BEING DEVELOPED AT t NL FOR HRC.
UNTIL THIS WORK IS COMPLETED IT IS NOT POSSIBLE TO GENERATE DOSE CONVERSION FACTORS FOR AN INDI.. DUAL OTHER THAN AN ADULT.
AS A MATTER OF IHFORMATION, OS", HAS A CONTRACT WITH ORNL TO PROVIDE THE NRC WITH AS MUCH INFORMATION REGARDING THEIR RESEARCH FOR ICRP AS WOULD BE PERMITTED BEFORE AND AFTER THE ICRP PUBLICATION #30.
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PROGRAM OFFICE CChMENVS ON POTENTIAL UTILYZATION OR vetUE CF RESEAltCH RESULTS IN THE QEGul4 TORY PROCESS
'RIL #4
.48
.DATE ISSUED: 04/03/79 RES PROGRAM ELEMENT: SITE SAFETY RIL TITLE: A TECTCHIC OVERVIEW OF THE CENTRAL MIDCONTINENT
' SPONSORING OFFICF(S): SD, HRR gegt 3-2 GEOLOGY 8 SEISMIC RESEARCH PROJECT MGR:
H. STEUER CHARACTERISTICS i
RCS COMMENTS:
"A TECTONIC OVERVIEW OF THE CENTRAL MIDCONTINENT", NUREG-0382, IS THE MOST UP-TO-Df. E SYNTHESIS OF GEOLOGIC-KNOWLEDGE OF,THE EARTHS 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 I5 RECOMMENDED THAT THE INFORMATI0H AND HYPOTHESES BE USED AS INPUT TO THE DEVELOPMENT OF A TECTONIC PROVINCE OR SEISMIC ZONING MAP OF THE EASTERN U,$. AND TO PROVIDE A BASIS AND GUIDE FOR ONGOING STUDIES IN THE AREA.
l USER DISCUSSICN POSITION COMMISSION ACRS PRESS 0FFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... SD/NRR SCHEDULED COMPLETION DATE..
, ACTUAL COMPLETION DATE.....
3' SD COMMENTS - NO COMMENT AT THIS TIME DUE TO PRELIMINARY HATURE OF THE WORK. WE PLAN AN EXTENSIVE REVIEW WHEN THE PROJECT IS FURTHER ALONG.
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r-PROGRAM OFFICE COMMENTS ON POTENTIAL UT I L IZ A T IOr4 OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS
' RIL #: '49 pATE {}SUEQ: 04/04/79 RES PROGRAM ELEMENT: FUEL CYCLE SAFETY AND' ENVIRONMENTAL EFFECTS RIL TITLE: IN VITRO DISSOLUTION OF URAHIUM PRODUCT SAMPLES FROM FOUR URANIUM MILLS SPONSORING OFFICE (S): SD
-RRA: 5-23 OCCdPATIONAL RESEARCH PROJECT MGR:
J. FOULKE EXPOSURE & PROTECTIDH KfS COMMENTS: THE MEASUREMENT OF THE SOLUBILITY OF VARIOUS FORMS OF YELLOWCAKE IN VITRO UTILIZED TWO SOLVENT SYSTEMS: A SIMULANT OF AH ULTRAFILTRATE OF BLOOD SERUM AND 0.tM HCL.
THE RESULTS OF THIS STUDY WILL BE USED TO DETERMINE WHETHER YELLOWCAKE SHOULD BE CONSIDERED A SOLUBLE COMPOUND WITH RESPECT T0 to CFR 20.
THE DATA OBTAINED USING THE SERUM SIMULANT SHOW THAT 25 TO 64 PERCENT OF ALL SAMPLES TESTED DISSOLVED WITH HALF-TIMES LESS THAH 16 HOURS.
IDEALLY, SOLUBILITY CLASSIFICATIONS SHOULD BE BASED ON THE DISSOLUTION 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 POSITIDH COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVTEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... SD SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
SD COMMENTS, S. MCGQIRE - THE RESULTS OF THIS RESEARCH WILL BE USED IN FINALIZING REGULATORY GUIDE 8.22, "BI0 ASSAY.AT URANIUM MILLS."
HQ FURTHER RESEARCH ON THIS TOPIC IS PRESENTLY CONTEMPLATED.
PROGRAM OFFICE CONNENTS ON FOTENTIAL UTitfkATION OR VALUE OF QESEARCH RESULTS IN THE REGilLATORY PROCES3 RTL-8: 50 DATE ISSUED: 64/06/79 RES PP0 GRAM ELEMENT: FUEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS 811 TITLE: CRITICALITY SAFETY GUIDANCE SPONSORING OFFICE (S): HMSS ERG: NONE RESEARCH PROJECT MGR D. SQLBERG
]UE) COMMENTS: SINCE 1957, THE NUCLEAR SAFETY GUIDE HAS PROVIDED USERS OF FISSIONABLE MATERIAL WITH SIMPLE
-ME1 MODS TO ASSURE SYSTEM SUBCRITICALITY. RESEARCH INFORMATION LETTER 850 TRANSMITS NUREG/CR-0095, " NUCLEAR
' SAFETY GUIDE, TID-7016, REVISION 2."
THIS REVISION UTILIZES IMPROVED CRITICALITY DATA AND COMPUTATIONAL-TECHNIQUES NOT AVAILABLE IN THE PREVIOUS GUIDE ISSUED IN 1961 REVISION 2 SHOULD BE REGARDED AS A SUPPLEMENT TO PREVIOUS GUIDES SINCE IT DOES NOT HAVE INTENTIONAL CONSERVATISMS INCLUDED AS IN PREVIOUS GUIDES.
IN USING REVISION 2 0F THE GUIDE, THE USER MUST IMPOSE APPROPRIATE SAFETY FACTORS FOR HIS APPLICATION AND THE NRC STAFF.MUST DETERMINE THAT THE SAFETY MARGINS PROPOSED BY THE USER OF REVISION 2 ARE ADEQUATE.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS
.EDST RTL ACTIVITIES REVIEN HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE.........
SCHEDULED COMPLETION DATE..
ACTUAL COMPLETICH DATE.....
HMSS COMMENTS: NO RESPONSE RECEIVED. -
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PROGRAM OFFICE COMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN THE REGULATORY PROCESS RIL #
51 DATE ISSUED: 04/12/79 RES PROGRAM ELEMENT: FUEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS RIL TITLE: THE CONCEPT COMPUTER CODE AND CAPITAL COSTS FOR PRESSUZIZED WATER REACTOR PLANTS SPONSORING OFFICE (S): HRR Rgg: S-21 SOCIO-ECONOMIC RESEARCH PROJECT PGR D. BARNA IMPACTS RES COMMENTS: THIS RIL TRANSMITS THE RESULTS OF COMPLETED RESEARCH UPDATING AND EXPANDING THE CONCEPT COMPUTER CODE FOR FORECASTING CAPITAL COSTS OF PRESSURIZED WATER REACTOR PL ANTS.
IN 1971 THE ATOMIC ENERGY COMMISSION AUTHORIZED POWER PLANT INVESTMENT COST STUDIES. WHICH CULMINATED IN THE WASH-1230 REPORTS PUBLISHED IN 1972.
THEIR PURPOSE WAS TO FACILITATE POLICY AND ECONOMIC DECISI0HS ABOUT ELECTRIC GENERATIqH FACILITIES IN THE PUBLIC AND PRIVATE SECTORS. HATIONAL PRIORITIES CH ENERGY, THE REGULATORY ENVIRONMEHT AND THE COST OF LABOR, EQUIPMENT AND MATERIAL HAVE CHANGED SIGNIFICANTLY. THESE CHANGES DICTATED THE HECESSITY 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 HEW SET OF UPDATED REPORTS WAS AUTHORIZED AHD UNDERTAKEN.
THE STUDIES IN THESE SERIES HAVE A UNIFORM SET OF ECONCMIC AND TECHNICAL CRITERIA AND A UNIFORM ACCOUNTING SYSTEM. THE INVESTMENT COST ESTIMATES IN THESE SEP.IES ARE DEVELOPED FOR REFERENCE PLANTS CONSTRUCTED AT A HYPOTHETICAL SITE.
THE ESTIMATED TOTAL BASE CONSTRUCTION COST FOR THE 1139 MWE PWR REFERENCE DESIGN IS
$568,831,011 OR $449/KW BASED ON JULY 1,
1976 PRICES. THE TOTAL BASE CONSTRUCTION COST FOR THE PWR POWER PLANT (1031 MWE HET OUTPUT) REFERENCE IN WASH-1230 WAS APPROXIMATELY $211,000,000 OR $205/KW, BASED UPGH PRICES EFFECTIVE JANUARY 1971.
THUS, THE 1977 STUDY INDICATES APPROXIMATELY A 143 PERCENT INCREASE IN THE COST OF THE PLANT IN TERMS OF $/KW.
THE TOTAL DIRECT CRAFT LABOR COST OF APPROXIMATELY $133,100,000 CORRESPONDS TO AH AVERAGE HOURLY RATE OF $12.30..APPROXIMATELY 10,820,000 CRAFT LABOR MANHOURS AVERAGE ABOUT 9.5 MAHHOURS/KW.
THESE COMPARE TL AVERAGES OF $8.86/ HOUR AND 6.0 MAHHOURS/KW RESPECTIVELY FOR THE EARLIER DESIGN REPORT IN WASH-1230.
USER DISCUSSION POSITION COMMISSIGH ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS E0}T RIL ACTIVITIF,1 REVIEW HELD COMPLETED HELD HEtD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED UNSCHED ACTUAL COMPLETI0H DATE.....
NRR COMMENTS, AD. ENVIR. TECH.
DESCRIBE APPLICATION TO REGULATORY PROCESS: THE PRINCIPAL APPLICATION OF THE INVESTMENT COST DATA IS TO UPDATE THE CONCEPT COMPUTER CODE.
IN TURN THE CONCEPT CODE IS USED TO ESTIMATE CAPITAL COST FOR DIFFERENT SIZE PLANTS, DIFFERENT REGIONS OF THE COUNTY, DIFFERENT SCHEDULE LENGTH, DIFFERENT COMPLETION DATES, DIFFERENT ESCALATION AND INTEREST RATES.
IN ADDITION TO UPDATING CONCEPT THE DATA IS USED AS A GENERAL REFERENCE FOR SUCH THINGS AS, COST OF COMP 0HENTS, QUANTITIES OF MATERIALS USED, TYPE AND QUANTITY OF LABOR, ETC.
DESCRIBE IMPACT OF RESULTS: DURING FY 1977 44 REQUESTS WERE MADE FOR CONCEPT CODE RUNS AhD 32 REQUESTS WERE MADE IH 1978.
MOST REOUESTS INVOLVED RUNS FOR SEVERAL SIZES OF COAL PLANTS AND ONE OR TWO NUCLEAR UNITS FOR DIFFERENT ESCALATION AND INTEREST RATES.
THE USE OF THE CONCEPT CODE PERMITTED THE STAFF TO PERFORM THESE ANALYSES MORE EFFICIENTLY AND QUICKLY THAH IF THE COHCEPT CODE WERE NOT AVAILABLE.
COMMENTS / REMARK 1: DOE IS FUNDING THE L ATEST UPDATE.
PROGR AN OFFICE CCPNEMTS C'4 POT ENTlat UTILIZ2TIO9 09 VaLUE OF RESEARCH WESULTS IN THE REGULATORY PROCESS RJ( #8 52 DATE ISSUED: 04/23/79 RES PROGRAM ELEMFNT: SITE SAFETY RIL TITLE:
EARTHQUAKE. INTENSITY SCALE SPONSORING OFFICE (S)* SD, HRR ERG: 3-2 GEOLOGIC 1 SEISMIC EfSEARCH PROJECT FGR:
CHARACTERISTICS RES COMMENTS:
THIS MEMORANDUM TRANSMITS THE RESULTS OF COMPLETED RESEARCH ON THE SYSTEMATIC EVALUATION OF lHE MACROSEISMIC. DATA FILE COMPILED FOR OVER HALF A CENTURY BY THE U.S. COAST AND GEODETIC SURVEY AND ITS
' SUCCESSOR.ORGANIZATICHS.
A NEW SEISMIC INTENSITY SCALE WAS FORMED BY REVISION OF THE MODIFIED MERCALLI INTENSITY-SCALE OF 1931.
THE RESULTS HAVE BEEN PUBLISHED AS NOAA TECHNIC 4L MEMORANDUM, EDS NGSDC 4.
-" REEVALUATION OF THE MODIFIED MERCALLI INTENSITY SCALE FOR EARTHQUAKES, USING DISTANCE AS A DETERMINANT."
THE PURPOSE OF THE STUDY WAS TO RELATE EARTHQUAKE INTENSITY TO THE ENERGY RELEASED AND ITS ATTENUATION
'WITH DISTANCE TO PROVIDE A UNIFORM SCALE RELATED MORE LOGICALLY TO THE PHYSICAL PARAMETERS OF ACCELERATION, VELOCITY, DISPLACEMENT AND SPECTRAL CONTENT. IT IS SUGGESTED THAT HRC MAKE A FORMAL RECOMMENDATIDH TO THE U.S.G.S.
THAT A NEW INTENSITY SCALE BASED DH THIS OR A SIMILAR ANALYSIS BE PROMULGATED AS AN OFFICIAL' STANDARD.
THE HEW SCALE SHOULD PERMIT A FULL USE OF THE EXTENSIVE DATA BASE ALREADY AVAILABLE. TRANSITION FROM THE MM INTENSITY SCALE AND MAKE USER DISCUSSI0H POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIE1 REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... SD/hRR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
SD COMMENTS: HD RESPONSE RECEIVED.
'NRR COMMENTS: NO RESPONSE RECEIVED.
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PROGRAM OFFICE CCMMENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RE5ULTS IN THE REGULATORY PROCESS Rit 4: 53 DATE ISSUFD 05/16/79 RES PROGRAM ELEMENT:
FUEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS RIL TITLE: DEBRIS-BED COOLABILITY LIMITS, RESULTS FROM IN-CORE TESTS D-1, D-2 AND D-3 9PONSORING OFFICE (S): HRR 333: 2-6 ACCIDENT ENERGETICS RESEARCH PROJECT MGR:
R. WRIGHT t DOSIMETRY RES COMMENTS: THIS RIL REPORTS RESULTS OF THE INITIAL THREE EXPERIMENTS IN THE ACPR TEST REACTOR ON THE COOLABILITY LIMITS OF EEDS OF PARTICULATE FUEL DEBRIS UNDER A POOL OF SODIUM.
A MODEL DESCRIBING THIS BEHAVIOR IS ALSO REPORTED. THIS WORK SUPPLIES KEY INFORMATION FOR THE ASSESSMENT OF THE RISK OF A CORE-MELT ACCIDENT IN A LMFBR.
IN THE EXPERIMENTS THE SPECIFIC BED POWER WAS MEASURED AT WHICH LOCAL DRYOUT OF THE SODIUM COOLANT OCCURRED IN THE BED.
LOCAL DRYOUT HAD BEEN THOUGHT TO BE THE COOLABILITY LIMIT OF A CORE DEBRIS BED, AN ASSUMPTION THAT HAS BEEN USED IN SAFETY EVALUATIONS. BY SUSTAIN OPERATION AT POWER UNDER CGMDITIONS OF LOCAL DRYOUT OF THE SODIUM IN THE BED, THESE EXPERIMENTS SHOWED THAT LOCAL DRYOUT IS NOT THE TRUE BED COOLABILITY LIMIT.
USER DISCUSSION POSITION COMMISSIDH ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EDST RIL ACTIVITlf,1 REVJEW
_H_E t D COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE REbPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
NRR C0f1MENTS: HD RESPONSE REL EIVED.
M.
PROGRAM OFFICE CpMMEH75 ON POTENTiel UTIL124 TION OR VALUE OF RESEARCH RESULTS U4 THE REGut4 TORY PROCESS RIL #
S4 DATE ISSUED: 05/15/79 RES PROGDAM ELEMENT: RISK ASSESSMENT RIL TITLE: THE SET EQUATION TRANSFORMATION SYSTEM SPONSORING OFFICEfS): NRP/NMSS RR2 NONE RESEARCH PROJECT MGR*
W. VESELEY RES COMMENTS: AS A RESULT OF THE WORK PERFORMED IN THE SETS COMPUTER CODE PROJECT, AN IMPROVED VERSIDH OF THE SETS 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.
THE MAJOR RESULTS OF THE SETS PROGRAM HAVE BEEN:
(1) DEVELOPMENT OF AN AUTOMATIC TREE DECOMPOSITION ALGORITHM WHICH IS CURRENTLY BEING INCORPOR'.TED INTO THE STANDARD VERSION OF SETS; (2) DEVELOPMENT, IN PRELIMINARY FORM, OF BASIC MINIMAL CUT SET QUANTIFICATION PROCEDURES; (3) DEVELOPMENT OF A STANDARDIZED VERSION OF THE SETS CODE FOR THE CDC 6600 COMPUTER, AND INSTALLATION OF THIS VERSION OF THE CODE AT THE BROOKHAVEN NATIONAL LASORATORY COMPUTER CENTER FOR USE BY NRC PERSONNEL; AND (4) PREPARATION OF A SETS USERS' MANUAL ORIENTED SPECIFICALLY FOR THE FAULT TRr.. ANALYST.
USER DISCUSSION POSITION COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS EOST RIL ACTIVlTLES REVIEW H EL D COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... MRR/NMSS SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.
ERR COMMENTS ON JULY 27, 1979 S.
HAHAUF3:
RESCRIBE APPLICATION 10 REGULATORY PROCESir THE IMPROVED VERSION OF THE SET EQUATION TRANSFORMATION SYSTEM (SETS)
IS A COMPUTER CODE FOR EVALUATING LARGE FAULT TREES.
THERE ARE CURRENTLY TWO ACTIVITIES WITHIN NRR THAT USE FAULT TREES AS AN ANALYSIS METHOD.
THESE TWO ACTIVITIES ARE GENERAL TASK 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 WHO DEVELOPS THE FAULT TREE ALSO USES SETS CODE TO ANALYZE THE FAULT TREE.
THE AVAILABILITY OF THE SETS CODE ALSO PROVIDES A TOOL FOR FUTURE USE BY THE NRC STAFF IN APPLYIHG FAULT TREC METHODS TO THE LICENSING PROCESS.
THE SETS CODE IS AN IMPORTANT FEATURE IN THE METHODOLOGY (COMPUTER CODE) USED TO IDENTIFY TYPE I VITAL AREAS (I.E.,
"....THOSE AREAS WHEREIN SUCCESSFUL SABOTAGE CAN BE ACCOMPLISHED BY COMPROMISING OR DESTROYING TH VITAL SYSTEMS OR COMPONENTS LOCATED WITHIH AN AREA).
TYPE I VITAL AREAS ARE THE MOST SENSITIVE SECURITY AREAS IN THE PLANT AND REQUIRE THE HIGHEST LEVELS OF PHYSICAL PROTECTION.
IDENTIFYING THESE AREEAS IS AN IMPORTANT CONSIDERATION IN THE STAFF'S EVALUATION OF PHYSICAL SECURITY AT A NUCLEAR POWER PLANT.
THIS ANALYSIS AND EVALUATION IS CURRENTLY BEING USED FOR ALL OPERATING PLANT REVIEWS AND WILL ALSO BE AN ONGOING REQUIREMENT IN OPERATING LICENSE REVIEWS.
DESCRIBE IMPACT OF RESULTS: THE SETS CODE HAS NO DIRECT IMPACT ONN THE LICENSING PROCESS BUT IN THE FUTURE IT WILL
- FERMIT THE NRC STAFF TO ANALYZE LARGE SYSTEMS THAT MIGHT OTHERWISE HAVE TO BE ANALYZED SY LESS EFFICIENT OR LESS EFFECTIVE METHODS.
THE RESULTS OF THE MORE EFFECTIVE AND EFFICIENCY METHODS COULD LEAD TO EITHER A RELAXATION OF EXISTING REQUIREMENTS
COMMENVS/PEMARg}: THE SETS CODE IS 0*tE OF SEVERAL COMPUTER CODES THAY CAN EE USED YO EUALUATE FAULT TREESo INCLUDING EUAlt oTION FOR COMMON CAUSE FAILURES. NO ATTEMPT HAS BEEN MADE HERE TO DETERMINE THE RELATIVE EFFECTIVENESS t EASE OF USING THESE VARIOUS CODES.
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PROGRAM OFFICE COMMF.NTS ON POTENTIAL UTILIZATION OR VALUE OF RESEasCN RESULTS IN THE REGULATORY PROCESS RIL #:
55 DATE ISSUED: 05/29/79 RES PROGRAM ELEMENT:
FUEL CYCLE SAFETY AND ENVIRONMENTAL EFFECTS RIL TITLE: THE CONCEPT COMPUTER CODE AND CAPITAL COST FOR HIGH AND LOW SULFUR COAL PLANTS - 1200 MWE SPONSCRING OFFICF(S): NRR ggg: 5-21 SOCIO-ECONOMIC RESEARCH PROJECT MGR:
D. BARNA IMPACTS RES COMMENTS: THIS MEMORANCUM TRANSMITS THE RESULTS OF COMPLETED RESEARCH UPDATING AND EXPANDING THE CONCEPT COMPUTER CODE FOR FORECASTING CAPITAL COSTS OF.HIGH AND LOW SULFUR COAL PLANTS - 1200 MWE.
THE ESTIMATED TOTAL BASE CONSTRUCTION COST FOR THE 1200 MWE (NOMINAL) HIGH SULFUR COAL PLANT REFERENCE DESIGN IS $465.498,393 OR $378/KW BASED ON JULY 1,
1976 PRICES. THE ESTIMATED TOTAL BASE CONSTRUCTION COST FOR THE 1200 MWE (NOMINAL) LOW SULFUR COAL PLANT REFERENCE DESIGN IS $402,825,229 OR $324/KW BASED ON JULY 1,
1976 PRICES.
THE TOTAL BASE CONSTRUCTION COST FOR THE COAL-FIRED POWER PLANT (1000 MWE NET DUTPUT) REFERENCE IN WASH-1230 WHICH DID NOT HAVE FLUE GAS DESULFURIZATION IS APPROXIMATELY $174,000,000 OR $174/KW, BASED UPON PRICES EFFECTIVE JANUARY 1971.
THUS, THIS 1977 STUDY INDICATES APPROXIMATELY A 87.9 PERCENT INCREASE IN THE COST OF THE PLANT IN TERMS OF 3/KW.
THE STUDY AND ITS METHODOLOGIES HAVE BEEN REVIEWED EXTENSIVELY WHILE IN PROGRESS BY THE RES PROJECT MANAGER AND VARIOUS STAFF MEMBERS FROM NRR.
RES RECOMMENDS THAT THE UPDATED METHODOLOGY BE USED BY HRR FOR APPLICATION TO THE IDENTIFIED REGULATORY HEED (RR-NRR-76-6).
USER DISCUSSION POSITIDH COMMISSION ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACIIVITIES REVIEW HELD COMPLETED HELD HELD ISSUED IMPLEMENTED OFFICE RESPONSIBLE......... NRR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETION DATE.....
NRR COMUENTS, AD, ENVIR. TECH.
EfSCRIBE APPLICATION TO REGULATORY PROCESS: THE PRINCIPAL APPLICATION OF THE INVESTMENT COST DATA IS TO UPDATE THE CONCEPT CGhPUTER CODE.
IN TURN THE CONCEPT CODE IS USED TO ESTIMATE CAPITAL COST FOR DIFFERENT SIZE PLANTS, DIFFERENT REGICHS OF THE COUNTY, DIFFERENT SCHEDULE LENGTH, DIFFERENT COMPLETION DATES, DIFFERENT ESCALATION AND INTEREST RATES.
IN ADDITION TO UPDATING CONCEPT THE DATA IS USED AS A GENERAL REFERENCE FOR SUCH THINGS AS, COST OF COMPONENTS, QUANTITIES OF MATERIALS USED, TYPE AND QUANTITY OF LABOR, ETC.
DESCRIBE IMPACT OF_EfiULTS: DURING FY 1977 44 REQUESTS WERE MADE FOR CONCEPT CODE RUNS AND 32 REQUESTS WERE MADE IN 1978.
MOST REQUESTS INVOLVED PUNS FOR SEVERAL SIZES OF COAL PLANTS AND ONE OR TWO NUCLEAR UNITS FOR DIFFERENT ESCALATION AND INTEREST RATES.
THE USE OF THE CONCEPT CODE PERMITTED THE STAFF TO PERFORM THESE.,NALYSIS MORE EFFICIENTLY AND QUICKLY THAN IF THE CONCEPT CODE WERE NOT AVAILABLE.
COMMENTS / REMARKS: DOE IS FUNDING THE LATEST UPDATE.
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PROGR AM OFFICE CormENTS ON POTENTIAL UTILIZATION OR VALUE OF RESEARCH RESULTS IN TFE REGULATORi-PROCESS R U. # : 56 DATE ISSUED: 07/25/79 RES Po0 GRAM ELEMENT: ENVIRONMENTAL EFFECTS RESEARCH
'PIL TITLE:
EFFECTS OF NUCLEAR POWER PLANTS ON CCMMUNITY GRGWTH AND RESIDENTIAL PROPERTY VALUES SPONSORING OFFICE (S): NRR RRg 5-21 SOCIDECONOMIC RESEARCH PROJECT MGR:
C. PRICHARD IMPACTS RES COMMENTS: AS PART OF A LONwER TERM RESEARCH PROJECT TO DEVELOP FORECASTING METHODS FOR ESTIMATING THE IMPACT OF NUCLEAR POWER STATIONS OH LAND VALUES AND COMMUNITY GROWTH, THE PENHSYLVANIA STATE UNIVERSITY INSTITUTE FOR LAND AND W4TER RESOURCES CAP.RIED OUT A STUDY OF THE EFFECTS OF HUCLEAR POWER STATIONS ON LAND VALUES AHD COMMUNITY GROWTH AT FOUR PREVIOUSLY-LICENSED STATIONS IH THE NORTHEAST. THE RESULTS OF THE STUDY DID HOT INDICATE THE PRESENCE OF SIGNIFICANT ADVERSE EFFECTS ON LAND VALUES OR COMMUNITY GROWTH.
USER DISCUSSION POSITION COMMISSIDH ACRS PRESS OFFICE MEETING PAPER BRIEFING BRIEFING RELEASE RESULTS POST RIL ACTIVITIES REVIEW HELD COMP L ET ED_
HELD HELD ISSUED IMPLEMENTED OFFICE RESP 0HSISLE.........
NRR SCHEDULED COMPLETION DATE..
ACTUAL COMPLETI0H DATE.....
NRR COMMENTS. W. RUSS1([ -
DESCRIBE APPLICATION TO REGULATORY PROCES}: AS PART OF THE COST BENEFIT ANALYSIS OF LICEHSING APPLICATIONS,
'"E NRC IS REQUIRED TO ASSESS THE LIKELY SOCIDECONOMIC IMPACTS ASSOCIATED WITH CONSTRUCTION AND OPERATIDH OF HUCLEA.
POWER STATIONS ON LOCAL COMMUNITIES AND THE SURROUNDING REGIGHS. THE IMPACT OF OPERATING STATIONS OH LAND USE, PROPERTY VALUES, LAND U5E PLANNING AND FGPUL ATION CHANGE ARE EFFECTS WHICH HAVE RECEIVED CONSIDERABLE ATTENTION FR0h INTERVENORS AND HEARIho BOARDS.
THIS RESEARCH CONTRACT WAS INITIATED WITH THE INTENTION OF TESTING A METHODOLOGY FOR QUANTIFYING THE IMPACT OF STATION SITING ON RESIDENTIAL PROPERTY VALUES.
' DESCRIBE _ldP ACT OF RESULTS: BECAUSE OF THE NON-RANE0M SELECTION OF THE STUDY SITE, GENERALIZABLE CONCLUSIDHS CANNOT BE DRAWN; HOWEVER, FOR THE 4 SITES INVOLVED. THE RESULTS DO INDICATE THAT THF NUCLEAR STATIONS HAD NO MEASURABLE, ADVERSE IMPACT ON RESIDENTIAL VALUES IN SURROUNDING AREAS.
BESIDES APLxNG TO THE STAFF'S GENERAL KNOWLEDGE. THE STUDY PROVIDED THE STAFF WITH A TESTED, OBJECTIVE. AND REPLICABLE METHODOLOGY FG? RETROSPECTIVELY DIMENSIONING PROPERTY VALUE IMPACTO AT HUCLEAR STATIONS. WITH RESPECT TO THIS POINT THE STAFF INTENDS TO EVALUATE THE IMPACT OF THE ACCIDENT AT THREE MILE ISLAND ON PROPERTY VALUES USING A METHODOLOGY b4 SED ON THE PROCEDURE USED IN THIS CONTRACT.
THE METHODOLOGY AND CONCLUSIONS OF THIS RESEARCH INCREASED THE STAFF'S UNDERSTANDING OF THE PROCESS OF PROPERTY VALUE IMPACT AT THE LOCAL LEVEL.
CONTINUATIDH OF RESEARCH ON PROPERTY VALUE IMPACTS SHOULD RESULTS IN GENERALLY RELIABLE METHODS FOR PREDICTING RESIDENTIAL VALUE IMPACTS IN SPECIFIC SITING CASES.
COMMENTS / REMARKS: NONE
3.0 'PQOJECTED NEAR-TERM RESEARCH INF0PNATION LETTER & (TARGEY DATES ARE FISCAL YEARS)
TEMP.
NUMB.
SUBJECT / TITLE / IMPACT TARGET DATE RRG # AND NAME RRG CHAIRMAN T-2 MARC STRUCTURAL ANALYSIS PROGRAMS.
QTR 4, 79 5-10 TRANSPORTATION SAFETY W. LAHS INSTRUCTIONS ON THE CAPABILITIES AND USE STRUCTURAL ANALYSIS OF THE MARC PROCRAM HAVE SEEN AND WILL AGAIN BE PRESENTED IN A 5 DAY COURSE TAUGHT BY THE MARC ANALYSIS CORP. STAFF.
T-17 BUILDING WAKE DIFFUSION MODELS. COMPARES 0TR 4, 79 6-6 LWR RISK ASSESSMENT J. MURPHY FIELD AND WIND TUNNEL MODELS TO VERIFY LICENSING DATA BASE.
T RADIATION (SOURCE TERM) SIMULATOR TESTS.
QTR 4, 79 1-25 QUALIFICATION, TESTING, R. FEIT EVALUATION T-26 IE APPLICATION OF WASH-1400 NE'HODOLOGY QTR 4, 79 6-6 LWR RISK ASSESSMENT J. MURPHY
'T-30 LOW FLOOD RATE CORRELATION. MAY REMOVE QTR 4, 79 1-5 REFL'<0D HEAT TRANSFER L. THOMPSON THE CURRENT RESTRICTION IN APPENDIX K OF STEAM CCOLING BELOW 1 INCH /SEC AND REPLACE WITH CORRELATION.
T-31 PHYSICAL SEPARATION CRITERIA (HORIZ.
QTR 4, 79 1-23 ELECTRICAL STDS & FIRE PROT. R. FEIT-CLOSED SPACE).
CONFIRMS PORTIONS OF
. REG. GUIDE 1.75.
T-33 CODE RECOMMENDATIONS FOR MULTIPLE INTER-QTR 4, 79 1-20 VESSEL INTEGRITY P. ALBRECHT ACTING N0ZZLES. PROVIDES NEW DESIGN RULES FOR SUCH AS S*FETY INJECTION N0ZZLES.
T-34 ENVIRONMENTAL IODINE SPECIES BEHAVIOR.
QTR 4, 79 5-16 AQUATIC RADIONUCLIDES, P. REED
. STUDY THE PHYSICAL & BIOLOGICAL TRANSPORT RADI0 ECOLOGY OF CHEt1ICAL FORMS TO RADI0 IODINES RELEASED TO THE ENVIRONMENT FROM AN OPERATING NUCLEAR STATION. DETERMINE THE INFLUENCE OF WET DEPOSITION (RAIN OR DEU) FOR METEOROLOGICAL MODELS & THE IODINE-AIR-GRASS-MILK PATdWAYS. PERFORM LABORATORY TESTS TO DETERMINE IF METHYIODIDE IS DEPOSITED ON GRASS UNDER WET DEPOSITION 1
CONDITIONS. DETERMINE ENVIRONMENTAL
~
PATHWAYS OF TRITIUM AND CARBON-14 RELEASED FRCM NUCLEAR STATIONS.
s.
M Y
i a'
T-35 MOHITORING ON RADI0 IODINE FROM CONTAINMENT.QTR 4, 79 5-19 TERRESTRIAL RADICECOLOGY P. REED.
ACCIDENTS. TO DETERMINE-THE PRACTICALITY 0F USING AVAILABLE CIVIL DEFENSE INSTRU-MENTS TO ASSESS PUBLIC HEALTH IMPACTS OF AN ACCIDENTAL RELEASE OF RADIOI0 DINE FROM HUCLEAR STATIONS. TO ASSESS INSTRUf!ENTA-TION CAPABILITIES (PARTICLE COLLECTION EFFICIENCY, PARTICULATE CONTRISUTIONS, AND INSTRUMENT RELIABILITY) BY EVALUATING A PORTABLE FIELD RADI0 IODINE COLLECTION SYSTEM AND A DETECTION SYSTEM USING A CDV-700 GM SURVEY INSTRUMENT.
T-50 WELD REPAIR INTEGRITY. PROVIDES BACK-QTR 4,'79 1-20 VESSEL INTEGRITY C. SERPAN GROUND TO PERMIT GR DEHY VESSEL WELD REPAIRS.
T-56 VERIFY FRAP T4.
ADDS TO FRAP T3; A)
QTR 4, 79 1-12 FUEL CODE DEVELOPMENT G. MARINO PRELIMINARY CAPABILITY FOR MULTI-ROD GEOMETRY AND B) CAPABILITY TO EVALUATE STATISTICAL VARIATIONS IN MODELS.
T-58 TRANSIENT ECC BYPASS CORRELATION. MAY QTR 4, 79 1-6 ECC BYPASS A.
SERKIZ
. CHANGE APPENDIX K FROM USE OF "END OF BYPASS" MODEL AND NO ECC PRIOR TO E0B, TO HEW CORRELATION.
T-98
' LOFT L2-2 AND L2-3 EXPERIMENTS QTR 4, 79 1-1 LOFT G. MCPHCRSCH T-113 ANNEALING TEST.
PROVIDES BACKGROUND TO QTR 4, 79 P. ALBRECHT l
EVALUATE ANNEALING PROPOSALS TO DEFINC l
RULES AND LIMITS FOR ANNEALING.
1 T-114 STEAM GENERATOR TUBE INTEGRITY.
QTR 4, 79 1-20 VESSEL INTEGRITY J. MUSCARA l
T-146 A REPORT ON QUALIFICATION TEST CRITERIA.
QTR 4, 79 3-7 SECONDARY CONTAIHMENT B. BROWZIH ANALYSIS CORRELATI0H OF VARIOUS TESTING.
T-188 SODIUM CONCRETE INTERACTION TESTS FOR QTR 4, 79 2-3 PRIMARY SYSTEMS INTEGRITY T. WALKFR FFTF CONTAIHMENT MARGINS. A GENERIC STUDY UF THE SODIUM-CONCRETE REACTION AND THE EVALUATION OF CONTAINMENT REQUIREMENTS FOR FFTF.
INTERMEDIATE AND LARGE SCALE TESTS OF BASALT AND MAGNETITE j
CONCRETE REACTION WITH METALLIC SODIUM l
AT VARIOUS TEMPERATURES.
l "T-191 PROMPT-BURST ENERGETICS EXPERIMENTS QTR 4, 79 2-6 ACCIDENT ENERGETICS R. WRIGHT l
WITH CARBIDE FUEL, ACRR TESTS SG-1, 3
SG-2, AND SG-3.
T-192 IRRADIATED FUEL DISRUPTION UNDER LOFT QTR 4, 79 2-6 ACCIDENT ENERGETICS R. WRIGHT ACCIDENT CONDITIONS, ACRR TEST SERIES FD-1.27-196 DYHANIC SIMULATION OF WASTE / ROCK PROCESS QTR 4, 79 6-5 HIGH LEVEL WASTE ISOLATION M.
CULLINGFORD (GEOLOGICAL FEEDBACK MECHANISM MODELING) _ _ _ _ _
T-4 EMERGENCY PLANNING QTR 1,
00 G-1 EMERGENCY PLANNING R. BLOND T-5 INSTITUTIONAL RADWASTE CHARACTERIZATIONS. QTR 1, 80 NONE C. BARTLETT RIL WILL BE A REPORT WHICH WILL PRESENT DATA ON SOURCE CHARACTERISTICS AND VOLUMES-0F LOW LEVEL WASTE GENERATED IN THE MEDICAL ACADEMIC USES OF RADIGISOTCPES. DATA WILL BE USED IN ESTABLISHING STANDARDS FOR PACKAGING AND DISPOSAL OF LOW LEVEL WASTES.
T-16 SIMMER-II RELEASED. REASSESSMENT OF WORK QTR 1, 80 2-14 SIMMER CODE T. WALKER ENERGY TO BE ACCOMMODATED IN HDCA..
T-27 COBRA-TF-AHALYSIS.
PROVILES NRR WITH QTR 1, 80 1-14 ADVANCED SYSTEM CODE S. FABIC INDEPENDENT EVALUATION OF UHI PERFORMANCE DURING LOCA.
1-29 CONSEQUENCE MODEL (INDIVIDUAL SITE QTR 1,
80 6-2 CONSEQUENCE MODELING R. BLOND ASSESSMENT, FINAL MODEL UPDATE, CRAC USERS MANUAL, AND COMPARISON TO SAFETY REVIEW METHODS IN 10 CFR 100)
T-44 INDEPENDENT ASSESSMENT RELAP4/ MOD 6.
QTR 1, 80 1-17 CODE APPRAISAL-W. LYON ASSESS PWR BE ANALYSIS THROUGH REFLOOD.
T-47 VERIFY FRAPCON - PROVIDES NRR WITH BE QTR 1, 80 1-12 FUEL CODE DEVELOPMENT G. MARINO i
AND EM CAPABILITY TO ANALYZE STEADY STATE FUEL BEHAVIOR.
T-73' VERIFY HEAT TRANSFER CORRELATION.
CON-QTR 1,
80 1-17 CODE APPRAISAL L. S. TONG FIDENCE IN THE CORRELATION IS INCREASED THROUGH COMPARISON WITH WORLD WIDE TEST DATA.
T-79 REACTOR PRESSURE VESSEL SURVEILLANCE QTR 1,
80 1-20 VESSEL INTEGRITY C.
SERPAN PROCEDURE. PROVIDES BACKGROUND TO ACCURATELY EVALUATE SURVEILLANCE EMBRITTLEMENT AND FLUENCE SAFE LIFETIMES.
T-101 1/5 SCALE TORUS TEST RESULTS (REISSUE QTR 1,
80 R. CUDLIN OF RIL TO INCLUDE ERROR ANALYSIS).
T-103 PWR LL'A P1A RELEASED. PROVIDES NRR QTR 1,
80 1-14 ADVANCED CODE & CODE L. SHOTKIN WITH ALfANCED PWR C00 LAND SYSTEMS CODE TO ASSESSMENT ESTABLISH MARGINS OF SAFETY.
T-116 BU'LDING WAKE DIFFUSION. WIND TUNNEL QTR 1,
80 3-4 SEVERE STORMS R.
ABBEY MODEL STUDIES.
T-117 BUILDING WAKE DIFFUSION. FIELD TEST QTR 1, 80 3-4 SEVERE STORMS R. ABBEY STUDIES.
T-127 NEW ENGLAND SEISMOLOGY REPOR1 (FY 77 QTR 1, 80 3-1 NRC/ STATE REGIONAL EARTH N.
STEUER I
CR-0051).
SCIENCES OT-128 NEW ENGLAND !EISMOLOGY REPORT (FY 78 QTR 1,
80 3-1 HRC/ STATE REGIONAL EARTH H.
STEUER CR 0939)
SCIENCES
, t i
- +
ET-129
SUMMARY
OF INTERIM RESULTS. PRELIMINARY QTR 1, 80 3-1 NRC/ STATE REGIONAL EARTH N. STLUER TECTONIC PROVINCE BOUNDARIES, REGIONAL SCIENCES GEOLOGY, SEISM 3 TECTONICS. TO ACCELERATE LICENSING.
T-130 INTERIM REPORT OF STUDY NEMAHA RESULTS.
QTR 1,
80 3-1 NRC/ STATE REGIONAL EARTH N. STEUER STUDY REGIONAL GEOLOGY, SEISM 0 TECTONICS.
SCIENCES TO ACCELERATE LICENSING.
T-131-NORTHEAST MARLBOROUGH AND SHREWSBURY QUAD QTR 1,
80 3-1 NRC/ STATE REGIDHAL EARTH H. STEUER SCIENCES T-13S NEW MADRID SEISM 0 TECTONIC STUDIES QTR 1,
80 3-1 NRC/ STATE REGIONAL EARTH N. STEUER SCIENCES CR-0379/0450
'T-136 OKLAHOMA FY 77 ANNUAL REPORT CR-0050 QTR 1,
80 3-1 NRC/ STATE REGIONAL EARTH N. STEUER SCIENCES T-137 PURDUE 38TH 11 REPORT CR-0449 QTR 1, 80 3-1 NRC/ STATE REGIONAL EARTH N. STEUER SCIENCES T-138 KANSAS FY 78 ANNUAL REPORT CR-0666 QTR 1,
80 3-1 NRC/ STATE REGIONAL EARTH N. STEUER SCIENCES T-139 OKLAHOMA FY 78 ANNUAL II CR-0785 QTR 1, 30 3-1 NRC/ STATE REGIONAL EARTH H. STEUER SCIENCES T-140 NEBRASKA FY 78 ANNUAL II CR-0876 QTR 1,'80 3-1 NRC/ STATE REGIONAL EARTH N. STEUER SCIENCES T-141 KANSAS EARTHQUAKE ANNUAL CR-0294 QTR 1,
30 3-1 NRC/ STATE REGIONAL EARTH H. STEUER SCIENCES T-142 ORIGIN SRFC. LINS. NEMAHA CR-0321 QTR 1,
80 3-1 NRC/ STATE REGIONAL EARTH N. STEUER I
SCIENCES T-143 OKLAHOMA EARTHQUAKE MAP GM-19 QTR 1,
80 3-1 NRC/ STATE REGIONAL EARTH N. STEUER SCIENCES T-170 NUMERICAL TECHNIQUES FOR DETESMINIt:G QTR 1,
80 DYNAMIC TESTING METHODS OF J. O'BRIEN FREQUENCIES, MODE SHAPES AND DAMPING OPERATING REACTORS VALVES USING LOW LEVEL TNPUT.
T-198 RSS METHODOLOGY APPLICATION PROGRAM QTR 1,
80 6-6 LWR RISK ASSESSMENT M. CUNNIP GNAM T-201 ZERO POWER TESTS COMPARED WITH PRIMARY QTR 1,
80 1-1 LOFT G. MCPHERSON PUMPS RUNNING.
T-206 INSPECTION MODULE STUDY DH RSS QTR 1,
83 6-6 LWR RISK ASSESSMENT J. PITTMAN T-215 ISEM ADVERSARY SEQUENCE EVALUATION MODEL QTR 1, 30 4-1 EFFECTIVENESS EVALUATION R. ROBINSON T-218 ORINCON TRANSPORTATION NETWORK MODEL QTR 1,
80 4-1 EFFECTIVENESS EVALUATION R. ROBISON
.T-221 FIXED-SITE NEUTRALIZATION MODEL (FSNM)
QTR 1, 80 4-1 EFFECTIVENESS EVALUATION R. ROBINSON
T-224 SOURCE AM9USH SIMULAV20H MODEL & USER OTR.1, 30 4-5 MEASUREMENTS 8 STANDARDS R. R093NSON MANUAL DOCUMENTATION T-229 SAFEGUARDS NETWORK ANALYSIS PROCEDURE QTR 1,
80 4-1 EFFECTIVT.riSS EVALUATION R. ROBINSON-(SNAP)
T-235 ATTRIBUTES OF THE INSIDER ADVERSARY QTR 1,
80 4-2 THREAT ASSESSMENT T-239 ' RESPONSE OF RAM PACKAGES TO PUNCTURE QTR 1, 80-5-10 TRANSPORTATION SAFETY W. LAHS ENVIRONMENTS STRUCTURAL ANALYSIS T-240 PARTICLE LEAK STUDIES QTR 1; 80 5-12 TRANSPORTATION SAFETY W. LAHS PROGRAM T-245 'IN-VIRO MEASUREMENTS OF MILL WORKERS QTR 1,
80 5-23 OCCUPATIONAL' EXPOSURE J. FOULKE PROTECTION T-246 SURFACE. PROPERTIES OF ENCAPSULANTS QTR 1, 80 T-247 WASTE-ROCK INTERACTIONS QTR 1, 80 T-248-' SOLUBILITY OF URANINITE QTR 1,
30 T-252 ENGINEERING EVALUATION OF LOW LEVEL WASTE QTR 1,
80 DISPOSAL SITES T-255 EXREM COMPUTER CODE QTR 1, 80 5-24 RADIOEIOLOGY & DOSIMETRY J. FOULKE T-256 DESIGN OBJECTIVES FOR LWR QTR 1,
80 5-23 OCCUPATIONAL EXPOSURE J. FOULKE PROTECTION T-20 GENERIC REVIEW OF THE IHPACT OF CONSERVA-QTR 2, 80 5-21 SOCI0 ECONOMIC IMPACTS C. PRICHARD TION, RATE STRUCTURE & LOAD MANAGEMENT ALTERNATIVES ON THE NEED FOR POWER.-
PREPARE A REFERENCE DOCUMENT WHICH REVIEWS 1 SYNTHESIZES RECENT EVIDENCE 3 EVALUATES 4
FUTURE PROSPECTS REGAPDING THE IMPACT OF ALTERNATIVE ENERGY CONSERVATION PROGP.AMS, RATE STRUCTURE 'ESIGN3, LOAD MANAGEMENT STRATEGIES,
'0-GENERATION SYSTEMS ON THE NEED FOR.
R.
'T-42 TRANSIENT CHF CORRELATION. NO CHANGE QTR 2, 80 1-3 PWR-BDHT HSU/ THOMPSON 34 APPENDIX K BUT MAY REPLACE CURRENT 1-4 ADVANCED SYSTEM CODE HSU/ THOMPSON CORRELATION.
T-54 WRAP EM/PWR - INCORPORATES INTEGRATED QTR 2, 80 NONE L. SHOTKIH
~
AND AUTOMATED IMPROVEMENTS INTO A LICENSING CODE PACKAGE.
T-57 BEACON MOD 3.
INTEGRATES SEVERAL QTR 2, 80 1-15 CONTAINMENT CODE R. CUDLIN PACKAGES IN BEACON INCLUDING PSS POOL DYNAMICS, ICE CONDENSER, AND h3 CONCEN-
.TRATION ANALYSIS.
"T-68 FSV DESIGN BASIS TRANSIENTS.
DEPENDS QTR 2, 30 2-12 GAS COOLED REACTOR R. SCHAMBERGER UPON THE REaVLTS OETAINED.
} -
T-80
,IMPROUED LOD 3E PARTS PJNYTORING. PROV2 DES QTR 2, 00 1-28 HOISE SURVEILLANCE AND W.
FARMER INPUT TO REG. GUIDE R QUIREMENT CHANGE AND DIAGNOSTICS TO LICENSING REVIEW.
T-87 VALIDATE ACOUSTICS EMISSIDH FLAW CORRELA> QTR 2, 80 1-21 NON-DESTRUCTIVE EXAMINATION J. MUSCARA TION.
VALIDATES THE DETECTION, IDENTIFICATI0H 8 ELIMINATION OF FLAWS DURING WELDING - LATER EVALUATIDH NOT NEEDED.
T COMMIX CODE (TWO-PHASE).
REASSESSMENT QTR 2, 80 2-13 FAST REACTOR SYSTEMS CODES R. CURTIS
'0F C00 LABILITY OF SUBASSEMBLY UNDER AND ACCIDENT ANALYSIS NATURAL CONVECTION.
T-99 EFFECT OF STEAM GENERATOR TUBE QTR 2, 80 W.
LANNING RUPTURE T-124 PRELIMINARY STUDY REPORT. PRELIMINARY QTR 2, 80 3-1 HRC/ STATE REGIONAL EARTH H. STEUER TECTONIC PROVINCE BCUNDARIES, REGIONAL SCIENCES GEOLOGY, SEISM 0 TECTONICS TO ACCELERATE LICENSING.
T-147 A COMPREHENSIVE REPORT WHICH WILL QTR 2, 80 3-7 SECONDARY CONTAINMENT B. BROWZIN RECOMMEND CRITERIA FOR HRR LICENSING STRUCTURAL POSITIONS ON SEISMIC SHEAR TRANSFER BASED ON MEDIUM SCALE TESTING.
T-153 A SERIES OF FORCE-TIME HISTORIES CHAR-QTR 2, 80 J. COSTELLO ACTERIZING AUTOMOBILE IMPACTS AT DIFFERENT
' VELOCITIES WILL BE PRESENTED TO ACCOUNT FOR TORNADO RISK REGIONS AND VEHICLE ORIENTATIONS AT IMPACT.
ENVELOPING FORCE-TIME HISTORIEa WHICH CAN BE USED FOR DESIGN AGAINST TORNADO MISSILE EFFECTS IN EACH TORNADO REGION UILL BE RECOMMENDED.
T-154 AN ASSESSMENT AND DESCRIPTION OF AVAILABLE QTR 2, 80 3-9 SEISMIC SAFETY MARGINS J.
COSTELLO METHODS UNDER DEVELOFMENT TO ANALYZE THE RESEARCH PROGRAM SDIL STRUCTUPE INTERACTION PHENOMENON WILL BE DEVELOPED BY A HUMBER OF EXPERTS. THE ASSESSMENT WILL INCLUDE THE RANGE OF APPLICABILITY, THE ADVANTAGES AND THE DIS-ADVANTAGES OF EACH TECHNIQUE, WAVE PASSAGE EFFECTS, 3D CHARACTER OF THE PROBLEM, NDHLINEAR BEHAVIOR OF THE SOLID AND LIMITATIONS OF AVAILABLE COMPUTER PROGRAMS.
"T-156 SUBSYSTEM RESPONSE REVIEW REPORTS QTR 2, 80 G.
BAGCHI ASSESSING THE STATE-OF-THE-ART OF SUB-SYSTEM RESPONSE DETERf1INATION, ACCURACY, AND UNCERTAINTIES.
T-171 STRUCTURAL AND MECHANICAL COMPONENT QTR 2, 80 DYNAMIC TESTING METHODS OF J. O'BRIEN TEST TECHNIQUES.
OPERATING REACTORS
T-179 DESCRIPTION OF A DESIGN CONCEPT FOR OTR 2, 80 MECHINERAL ENGENEERING J. BURNS CALCULATZHG PROBABILZVY OF RADIDACTIVE RELEASE, CORE MELT, SAFETY SYSTEM, STRUCTURAL AND CCMPONENT FAILURE, PROBABILITIES FROM A SET OF NUCLEAR PLANT SEISMIC RESPONSES.
T-183
SUMMARY
EVALUATION OF DYNAMIC QTR 2, 80 3-11 PL ANT COMPONENT AND D. REIFF EXCITATION TESTS OF A NUCLEAR VALVE.
EQUIPMENT BEHAVIOR T-197 HUMAN ERROR RATE ANALYTIS QTR 2, 80 6-4 RISK ASSESSMENT FOR M. CULLINGFORD FUEL PROCESSING T-207 RISK ASSESSMENT OF HIGH LEVEL WASTE QTR 2, 80 6-5 HIGH LEVEL WASTE ISOLATION M.
CULLINGFORD ISOLATION IN BEDDED S ALT T-216 SAFEGUARDS COMPCNENT DATA BASE MANUAL QTR 2, 30 4-1 EFFECTIVENESS EVALUATION R. ROBINSON T-217 COPS LLEA ROUTE DISTRIBUTION'MODEL QTR 2, 80 4-5 MEASUREMENTS S STANDARDS R. ROBINSON T-222 SAFE 1 SEMI-AUTOMATED SYSTEM EVALUATION QTR 2, 80 4-1 EFFECTIVENESS EVALUATION R. ROBINSON METHODOLOGY T-228 SACEM SMALL ARMS CASUALTY EVALUATION QTR 2, 80 4-1 EFFECTIVENESS EVALUATICH R. ROBINSON MODEL T-232 INSPECTION METHOD 5 FOR PHYSICAL QTR 2, 80 4-3 SAFEGUARDS. INFO. SYSTEMS E. RICHARD PROTECTION T-241 MODELING HORMAL SHOCK AND VIBRATION QTR 2, 80 5-10 TRANSPORTATION SAFETY W. LAHS ENVIRONMENT STRUCTURAL ANALYSIS T-249 REPOSITORY OPTION ASSESSMENT QTR 2, 80 T-250 THERMAL CONDUCTIVITY ROCK-iIINERAL QTR 2, 80 T-251 GLASS CERAMIC RADWASTE CONTAINER QTR 2, 80 EVALUATION T-254 EVALUATICH OF THORIUM CONTENT OF HUMAN QTR 2, 80 5-23 OCCUPATIONAL EXPOSURE J. FOULKE TISSUES PROTECTION T-265 VISUAL IMPACT OF ALTERNATIVE CLOSED QTR 2, 20 5-21 SOCIDECONOMIC IMPACTS C. PRICHARD CYCLE COOLING SYSTEMS
.T-266 SOCIDECCHOMIC IMPACT OF TMI QTR 2, 80 5-21 SOCIDECONOMIC IMPACTS C. PRICHARD I
T-15 ASSESSMENT OF AGRICULTURAL LAND QTR 3, 80 5-21 SOCI0 ECONOMIC IMPACTS C. PRICHARD T-2:
WRAP EM/BWR - INCORPORATES INTEGRATED QTR 3, 80 NONE L. SHOTKIN AND AUTCMATED IMPROVEMENTS INTO A LICENSING CODE PACKAGE.
T-38 REWET CORRELATI0H. MAY REMOVE APPENDIX K QTR 3, 80 1-3 PWR-BDHT HSU/ THOMPSON RESTRICTIONS ON ' RETURN OF HUCLEATE 1-4 BWR-BDHT HSU/ THOMPSON BOILING' WITH A CORRELATION. --
-m m. _. _ _. -_
m s
T-43
-TRAC PD2.
IMPROVED P1A.
QTR 3, 80 1-14 ADVANCED SYSTEM CODE L. SHOTKIN ADVANCED EWR AND PWR COOLANT SYSTEMS CODE TO ESTABLISH MARGINS OF SAFETY.
1-46 SOURCE TERM CORRELATION. CONFIRMS QTR.3, 80 1-13 FUEL MELT.
R. SHERRY CONSERVATISM CF REG. GUIDE ASSUMPTIONS
-FOR ACCIDENT ANALYSIS.
T-48 8X8 TRANSIENT CHF TESTS COMPLETE. SUPPORT QTR 3, 80 1-4 BWR-BDHT W. BECKHER EXISTING CORRELATION OR RECOMMEND A NEW CORRELATION.
T-70 EMBRITTLEMENT CRITERIA..MAY REPLACE QTR 3, 80 1-8 ZIRCALOY CLADDING M. PICKLESIMER EXISTING CRITERIA IN APPENDIX K.
T-76 UPDATE KIR CURVE-DUCTILE SHELF TOUGHNESS. OTR 3, 80 1-20 VESSEL INTEGRITY M.
VAGINS PROVIDES BASIS FOR HEAT UP C00LDOWN CURVES JH OPERATING PLANTS.
T-82 EXTINGUISHING SYSTEM EVALUATION QTR 3, 80 1-23 ELECTRICAL STANDARDS R. FEIT AND FIRE PROTECTION T-84
' AGING MODEL QTR 3, 80 1-25 QUALIFICATI0H TESTING R. FEIT EVALUATION T-90 CONTAINMENT PRESSURE PULSE ANALYSIS QTR 3, 80 2-12 GAS COOLED REACTOR R. SCHAMBERGER METHODS. CONCLUSIONS MAY EFFECT THE LICENSING POSITION CN DEPRESSURIZATION ACCIDENTS.
T-102 PELE-IC ISSUE 3 FOR BWR CONTAIHMENTS.
QTR 3, 30 1-15 CONTAINMENT CODE R.
CUDLIN CALCULATION OF DYHAMIC RESPONSE OF CONTAINMENT STRUCTURES WITH A COUPLED FLUID / STRUCTURE PROGRAM.
T-104 SOLA-FLX.
PROVIDES NRR WITH'A CODE TO QTR 3, 80 1-14 ADVANCED CODE L. SHOTKIH CONDUCT INDEPENDENT AUDIT OF VENDOR BLOWDOWN LOAD CALCULATIONS.
T-106 FAST RUNNING TRAC (PF1).
QTR 3, 80 1-14 ADVANCED CODE L. SHOTKIN i
T-107 TRAC PIA SENSITIVITY STUDY.
QTR 3, 80 1-14 ADVANCED CODE L. SHOTKIN T-108 CLADDING CREEPDOWN. IH-PILE CREEPDOWN QTR 3, 80 1-8 ZIRCALOY CLADDING M. PICKLESIMER OF ZIRCALOY CLADDING UNDER COMPRESSIVE LOADS.
l T-111 MELT CONCRETE INTERACTIONS. UPDATE TO QTR 3, 80 1-13 FUEL MELT R. SHERRY RIL 28.
DESCRIPTION OF CORCON CODE.
T-121 ASSESSMENT AND EXPANSICH OF STRONG QTR 3, 80 3-2 GEOLOGY & SEISMIC R. BRAZEE GROUND MOTION DATA.
BASIC INPUT TO CHARACTERISTICS SEISMIC RISK ASSESSMENT.
-.. - -.. ~ -
.T-122 RELATIVE EFFECT OF VMO BETWEEN QTR 3, GO 3-2 GEOLOGY t SEIBMZC-R. BRAZEE EASTERN AND WESTERN SIVES. VD CONFIRM CHARACTERISTICS METHODS OR PROVIDE BASIS FOR ALTERATION OF THE SPECTRAL CURVES AS APPLIED TO DIFFERENT REGIONS.
T-149 A REPORT FOR ORIENTATION FOR LICENSING QTR 3, 80 3-7 SECONDARY CONTAINMENT B. BROWZIN OH THE EFFECT OF SEISMIC DESIGN LEVEL STRUCTURAL ON THE HPP COST.
.T-150 INTERIM REPORT ON UAVE HEIGHTS.
QTR 3, 80 3-7 SECONDARY CONTAINMENT B. BROWZIN MEASUREMENT METHOD AT PROSABLE MAXIMUM STRUC! URAL HURRICANES (PMH).
T-157 SOIL STRUCTURE INTERACTION (SSI) REVIEW QTR 3, 80 G. BAGCHI REPORTS ASSESSING THE STATE-OF-THE-ART OF SSI ANALYSIS METHODOLOGY, ACCURACY, UNCERTAINTIES, AND ITEMIZING BENCHMARK PROPLEMS.
.T-158-SENSITIVITY STUDY INVESTIGATING THE QTR 3, 80 G. BAGCHI EFFECTS OF UNCERTAINTIES ON SUBSYSTEM RESPCNSE.
T-162 RESPONSE OF HUCL EAR POWER PLANT STRUCTURES QTR 3, 80 3-9 SEISMIC SAFETY MARGINS G. BAGCHI TO THREE IHPUT CJMP0HEHTS.
RESEARCH PROGRAM T-163 EFFECT OF STRUCTURAL DAMPING ON NUCLEAR QTR 3, 80 3-9 SEISMIC SAFETY MARGINS G. BAGCHI POWER PLANT STRUCTURES. THIS IS ANOTHER RESEARCH PROGRAM STUDY BY EXPANDING LLL/ DOR SEISMIC CONSERVATISM PROGRAM TO TYPICAL NUCLEAR POWER PLANT STRUCTURES (ZION STATION).
T-164 GENERAL STRUCTURAL BUILDING RESPONSE QTR 3, 80 3-9 SEISMIC SAFETY MARGIHS G. BACCHI ANALYSIS REVIEW WITH SPECIAL EMPHASIS RESEARCH PROGRAM ON DAMPING AND NONLINEARITY.
T-166 MATHEMATICAL MODEL FOR PREDICTON OF QTR 3, 80 3-7 SECONDARY CONTAINMENT B. BROWZIH NEAR SHORE TSUNAMI FOR ESTABLISHING A STRUCTURAL LICENSING METHOD.
T-168 SRP METHODOLOGY COMPARISDNS. THE QTR 3, 80 CURRENTLY USED (SRP) METHODOLOGY WILL BE J. COSTELLO COMPARED TO OTHERS IN THE SOIL-STRUCTURE
-INTERACTION REVIEW.
SYSTEMATIC DIFFERENCES WILL BE EXPLORED AND THEIR SIGNIFICANCE ASSESSED.
T-173 DOCUMENTATION OF STATE OF AVAILABLE QTR 3, 80 MECHANICAL ENGINEERING J. BURNS FRAGILITIES RELATED INFORMATION.
T-174 DOCUMENTATION OF FAILURE MODES FOR QTR 3, 80 MECHANICAL ENGINEERING J. BURHS COMPONENTS AND STRUCTURES FOR A REPRESENTATIVE PWR PLANT.
l l.
2 T-189 POSITRON ANNIHILATION - EVALUATION QTR 3, 80 2-4 CORE MELT AND CONTAINMENT T. WALKER AS AN NDE PROCECURE. EVALUATION OF INTEGRITY ELECTRCNS AT DEFECT SITES BY POSITRONS AND CORRELATION WITH CREEP-FATIGUE DAMAGE.
.T-193 FUEL FRAGMENTATION TESTS WITH MOLTEN QTR 3, 80 2-2 POST ACCIDENT HEAT REMOVAL R. WRIGHT RESULTS OF LARGE SPECIMEN TESTING OH SEISMIC SHEAR TRANSFER.
FRAGMENTATION AND THE FUEL-000LANT INTER-ACTION POTENTIAL FRull THE CONTACT OF THE MOLTEN FUEL METAL THERMITE WITH SODIUM.
T-194 DEBRIS BED COOLABILITY AT HIGH DECAY-HEAT QTR 3, 80 2-2 POST ACCIDENTAL HEAT R. WRIGHT POWERS, ACRR TEST D-4.
REMOVAL T-202 THREE FULL' POWER TESTS WITH QTR 3, 80 1-1 LOFT
- 0. MCPHERSON (1) CONTINUOUS DEPRESSURIZATION, (2) PRESSURE HANGUP, AND
'(3) REPRESSURIZATION.
T-219 STANDARDIZED LWR-GENERIC FAULT TREE QTR 3, 80 4-1 EFFECTIVENESS EVALUATION R. ROBINSON CHARACTERIZATION CODE T-223 BWR FACILITY CHARACTERIZATION CODE QTR 3, 80 4-1 EFFECTIVENESS EVALUATION R. ROBINSON T-225 PWR FACILITY CHARACTERIZATION CODE QTR 3, 80 4-1 EFFECTIVENESS EVALUATION R. ROBINSON T-227 ADVERSARY PATH SELECTION METHODS QIR 3, 80 4-1 EFFECTIVENESS EVALUATION R. ROBIHSON T-230 SABRES II CONVOY ENGAGEMENT MODEL QTR 3, 80 4-5 MEASUREMENTS 1 STANDARDS R. ROBINSON T-231 EARS COMMUNICATICHS MODEL QTR 3, 80 4-5 MEASUREMENTS 8 STANDARDS R. ROBINSON T-233 GENERIC METHOD FOR MATERIAL ACCOUNTING QTR 3, 80 4-2 THREAT ASSESSMENT R. SHEPARD METHODOLOGY j
T-234 GENERIC COST-EFFECTIVENESS MODEL QTR 3, 80 4-2 THREAT ASSESSt1ENT R.
SHEPARD l
T-237 CRITICALITY SAFETY METHODS - SOLID ANGLE QTR 3, 80 5-7 CRITICALITY SAFETY STUDIES D. SOLBERG 8 SURFACE DENSITY
.T-242 DISEQUILIBRIUM OF URANIUM ORE DAUGHTERS QTR 3, 80 5-23 OCCUPATIONAL EXPOSURE J. FOULKE IN ORE DUST PROTECTION T-243 MEASUREMENT TECHNOLOGY FOR PLUTONIUM QTR 3, 80 5-23 OCCUPATIONAL EXPOSURE J. FOULKE PROTECTION
~T-244 ASSESSMENT OF RESPIRATORY PROTECTION QTR 3, 80 5-23 OCCUPATIONAL EXPOSURE J. FOULKE SYSTEMS PROTECTION T-258 PATliOGENIC AMOEBAE IN CLOSED CYCLE QTR 3, 80 5-18 ECOLOGICAL IMPACT FROM J. FOULKE COOLING TOWERS REACTOR 8 FUEL CYCLE FACILITIES
,T-263 BNL EMERGENCY AIR SAMPLER FOR RADI0 IODINES QTR 3, 80 5-17 ENVIRONMENTAL MONITORING P. REED
T-105 K-FIX (3D).'TWO PHASE FLOW TH IDEALIZED QTR 4, 80 1-14 ADVANCED CODE L. SHOTKIN GE0 METRICS.
T-110 FISSION PRODUCT BEHAVIOR WITHIN LWR QTR 4, 80 1-9 FISSION PRODUCT RELEASE R.. SHERRY PRIMARY SYSTEMS UNDER ACCIDENT CONDITIONS.
CONTROLLED LOCA AND MELTDOWN DESCRIPTION OF TRAP-MELT CODE.
T-115 EVALUTE METHOD TO CALCULATE COASTAL QTR 4, 80 3-8 ATMOSPHERIC TRANSPORT 1 R. ABBEY DISPERSION. MAY IMPACT ON LICENSING DEPOSITION EVALUATION OF CLOSE-END ACCIDENTAL RELEASE CONCENTRATION.
T-123 IMPROVED BAYESIAN METHODS FOR SITE QTR 4, 80 3-2 GEOLOGY AND SEISMIC R. BRAZEE DEPENDENT SPECTRA. A HAZARD CODE WHICH CHARACTERISTICS WILL ALLOW THE INCLUSION OF EXPERT JUDGMENT AT ALL STAGES TO PRODUCE STATISTICAL ESTIMATES OF EARTHQUAKE HAZARD.
T-132' INTERIM REPORT STUDY RESULTS. PRELIMINARY QTR 4, 80 3-1 NRC/ STATE REGIONAL EARTH N.
STEUER TECTONIC PROVINCE BOUNDARIES, REGIONAL SCIENCES GEOLOGY, SEISMOTECTONICS. TO ACCELERATE LICENSING.
T-144 MODELING TORNADO DYNAMICS.
QTR 4, 80 3-4 SEVERE STORMS R.
ABBEY T-148 C0t1PRdHENSIVE REPORT ON QUALIFICATION QTR 4, 80 3-7 SECONDARY CONTAINMENT B. BROWZIH TEST NEEDED FOR REGULATORY GUIDES VERIF.
STRUCTURAL T-159 SENSITIVITY STUDY ON NONLINEAR ANALYSIS QTR 4, 80 G.
BAGCHI 0F A KEY SAFETY SYSTEM.
T-160 SENSITIVITY STUDY COMPARING LINEAR QTR 4, 80 G. BAGCHI FINITE ELEMENT ANALYSIS WITH CONTINUUM ANALYSIS APPROACH FOR ZION NUCLEAR POWER STATION.
T-161 COMPARATIVE ANALYSIS OF ZION REACTOR QTR 4, 80 G. BAGCHI CONTAINITENT BUILDING USING LINEAR FINITE ELEMENT ANALYSIS CONTINUUM ANALYSIS APPROACH, AND NONLINEAR FINITE ELEMENT ANALYSIS.
T-165 IDENTIFY MODELING VARIABLES, SYSTEMATIC QTR 4, 80 3-9 SEISMIC SAFETY MARGINS C. BURGER OR RANDOM, IN THE MAJOR STRUCTURE RESPONSE RESEARCH PROGRAM ANALYSIS. RANK IN TERMS OF THF'R EFFECT ON THE UNCERTAINTY OF STRUCTIF.AL RESPONSE.
T-167 MATHEMATICAL MODEL FOR PREDICTIDH OF FAR QTR 4, 80 3-7 SECONDARY CONTAINMENT B. BROWZIN OFF-SHORE TSUNAMI FOR ESTABLISHING A STRUCTURAL LICENSING METHOD.
T-175 FRAGILITY CURVES FOR COMPONEHTS AND QTR 4, 80 MECHANICAL ENGINEERING J. BURNS STRUCTURES FOR A REPRESENTATIVE PWR
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PLANT. -
T-264 REVIEU OF CONSERVATICN, LCAD MANAGEMENT, QTR 3, 30 5-21 50Cf0 ECONOMIC IMPACTS C. PRICHARD RATE RESTRUCTIVE AND C0 GENERATION
. T-12 CRACK ARREST. BASIS FOR SEVERE QTR 4, 80 1-23 VESSEL INTEGRITY M. VAGINS ACCIDENT ANALYSIS ARREST EVALUATION.
T 18
. FISSION GAS RELEASE MCDEL VERIFIED. ADDS QTR 4, 80 1-9 FISSION PRODUCT RELEASE G. MARIN0 TRANSIENT FISSION GAS RELEASE QUANTI-
.FICATION TO FRAP-T ENHANCING ACCIDENT CALCULATIONS.
T-32 FIELD VALIDATE ELECTRCCHEMICAL TEST.
QTR 4, 80 1-22 CORROSION J. MUSCARA UPDATES REG. GUIDE TO PRECLUDE SENSITIZED STAINLESS STEEL IN SERVICES.
T-45 MECHANICAL PROPERTIES REPORT. PROVIDES QTR 4, 80 1-8 ZIRCALOY CLADDING M. PICKLESIMER INDEPENDEhTLY VERIFIED IRRADIATED ZIRCALOY-DATA FOR USE IN LICENSING CALCULATIONS.
T-55 LOCA THERMAL SHOCK AND STEAM LINE BREAK QTR 4, 80 1-20 VESSEL INTEGRITY M. VAGINS ANALYSIS. PROVIDES A LICENSING PROCEDURE FOR THERMAL SHOCK ANALYSIS INCLUDING EFFECTS OF WARM PRESCREEHING. MAY RESULT IH TECH. SPEC.' CHANGES.
T-67.
17 X 17 P.EFLOOD HEAT TRANSFER CORRELA-QTR 4, 80 1-5 REFLOOD HEAT TRANSFER E. DAVIDSCH TION VERIFIED. MAY REQUIRE NEW SECTION OF APPENDIX K TC ADDRESS THIS GEOMETRY.
T-89 PIPING CCMPONENTS UNDER CREEP AND QTR 4, 80 2-3 PRIMARY SYSTEM INTEGRITY T. J. WALKER PLASTIC DEFORMATICH FINITE ELEMENT ANALYSIS OF PIPING COMPONENTS AND COMPARISDN WITH EXPERIMENTAL RESULTS.
T-92 8 X 8 BUNDLE TE3TS.
CONFIRMS ABILITY QTR 4, 80 1-8 ZIRCALOY CLADDING M. PICKLESIMER TO LIMIT PROPAGATION AND MAINTAIN C00LABLE GEOMETRY.
T-93 CYCLIC CRACK GROUTH RATE EVALUATION.
QTR 4, 30 1-20 VESSEL INTEGRITY P. ALBRECHT HEW CURVE FOR ASME XI CRACK. GROWTH RATE.
T-94 RECOMMEND OPTIMIZED PIPING SYSTEM QTR 4, 80 1-20 VESSEL INTEGRITY P. ALBRECHT DESIGN. GUIDELINES FOR SAFER PIPING SYSTEMS.
T-96 SAFETY RELATED OPERATOR ACTING CRITERIA.
QTR 4, 80 1-24 HUMAN ENGINEERING W. FARMER
~
MAY IMPACT ENGINEERED SAFETY SYSTEMS i
T-97 URANIUM OXIDE AEROSOL PROPERTIES AND QTR 4, 80 2-9 LARGE AEROSOL TRANSPORT M. SILBERBERG i
NSPP TEST AEROSOL CODE VERIFICATION.
TESTS DATA USED TO VERIFY INFORMATION IN HAARM-3.
T-100 ALTERNATE ECCS SYSTEMS QTR 4, 80 NONE W. LANNING 4
0
=
T-176 ESTA9LISHMENV 0F A QUASI-DELPHI QTR 4, 90 MECHAN 8 CAL ENG!NEER8HG J. BURNS PROCEDURE FOR MANDLING EXPECT CPINION ON COMPONENT FRAGILITIES.
T-177 RESULTS AND USE OF A PPOGRESSIVE FAILURE QTR 4, 80 MECHANICAL ENGINEERING J. BURNS ANALYSIS ON A BUILDING (S) 0F A REPRE-SENTATIVE PWR PLANT.
T-178 DOCUMENTATION OF EXPERT OPINION ON QTR 4, 80 MECHANICAL ENGINEERING J. BURNS COMPONENT FRAGILITY.
T-181 DESCRIPTICH OF COMPUTATICHAL METPODOLOGIES QTR 4, 30 MECHANICAL ENGINEERING J. BURNS FOR PREDICTICH OF CORE MELT PROBABILITIES IN A NUCLEAR POWER PLANT DUE TO SEISMIC EVENTS.
T-182 PROBABILITY CF LOCA INDUCED BY QTR 4, 80 3-9 SEISMIC SAFETY MARGINS J. O'BRIEN EARTHQUAKES.
RESEARCH PROGRAM T-203 FULL POWER TEST HITH PRESSURE RELEASE QTR 4, 80 1-1 LOFT G. MCPHERSON VALVE STUCK OPEN.
T-204 FULL POWER TEST WITH STEAM LINE BREAK QTR 4, 80 1-1 LOFT G. MCPHERSON T-211 RISK ASSESSMENT OF SPENT FUEL IN BEDDED QTR 4, 09 6-5 HIGH LEVEL WASTE ISOLATION M.
CULLINGFORD SALT T-214 PLANT DATA ANALYSIS.
QTR 4, 80 W. LESELY T-220 SAFEGUARDS ENGINEERING AND ANALYSIS DATA QTR 4, 80 4-1 EFFECTIVENESS EVALUATION R. ROBINSON BASE (SEAD)
T-226 SKIRMISH /AMSUSH BOARD GAMES QTR 4, 80 4-5 MEASUREMENTS & STANDARDS R. ROBINSON T-238 GUIDANCE FOR ADMINISTRATIVE CONTROL OF QTR 4, 80 5-7 CRITICALITY SAFETY STUDIES D. SQLBERG CRITICALITY SAFETY T-253 BURIAL GRCUND SITE SURVEY - KENTUCKY QTR 4, SO T-257 PEDIATRIC PHANTOMS QTR 4, 80 5-24 RADIOSIOLOGY 8 DOSIMETRY J. FOULKE T-238 GUIDANCE FOR ADMINISTRATIVE CONTROL OF QTR 4, 80 5-7 CRITICALITY SAFETY STUDIES D. SOLBERG T-260 EFFECTS OF AIRE 0RNE CONTAMINANTS ON QTR 4, 80 5-6 FUEL. CYCLE FACILITIES D. SOLBERG ACTIVATED CHARCOAL EFFLUENT CONTROL T-261 ASBESTOS IN COOLING TOWER WATERS QTR 4, 30 5-15 CHEMICAL IMPACTS ON P. REED
~
AQUATIC ENVIRONMENT T-262 MODEL TO FREDICT CHLORINE CONCENTRATIONS QTR 4, 80 5-14 PHYSICAL TRANSPORT SURFACE P.
REED IN POWER STATION DISCHARGER HATERS T-6 REPORT ON WASTE MANAGEMENT RISK QTR 1,
81 6-5 HIGH LEVEL WASTE ISOLATION M. CULLINGFORD ASSESSMENT METHODOLOGY DEVELOPMENT
~
,T-28 CORE MELTDOWN SENSITIVITY STUDY QTR 1, 31 6-6 LWR RISK ASSESSMENT M. CUNNINGHAM-T-60 COMPREHENSIVE REPORTC WHICH WILL QTR 1, 81 3-7 SECONDARY CONTAINMENT
- 5. BR0GZIN.
SUMMARIZE RESULTS OF LARGE SPECIMEN STRUCTURAL TESTING.ON SEISMIC SHEAR TRANSFER.
.T-69 UNI-ECCS TESTS. PROVIDES NRR BASIS FOR QTR 1, 81 1-2 SEMISCALE W. LANNING UHI LICENSING DECISIONi AND MAY REQUIRE NEW SECTION IN APPENDI) W,
T-74 RELAP 4 MOD 7.
USER CONVENIENT-QTR 1, 81 1-16 REFERENCE SYSTEM CODE W. LYON PWR LOCA CODE T-75 PROBABILITY AND CONSEQUENCES OF STEAM QTR 1, 81 1-13 FUEL MELT W. JOHNSTON EXPLOSIONS. POSSIBLY WILL PROVIDE A QUANTITATIVE PREDICTIGN OF STEAM l
EXPLOSIONS MOSTLY APPLICABLE TO FLOATING PLANTS.
T-77 HYDROELASTIC' EFFECTS STUDIES. FOR PSS QTR 1, 81 1-15 CONTAINMENT CODE L. SHOTKIN (PRESSURE SUPPRESSION SYSTEMS) CONTAIN-MENTS T-33 DETECTION SYSTEM EVALUATION QTR 1,
81 1-23 ELECTRICAL STANDARDDS R. FEIT AND FIRE PROTECTION T-85 DATA BASE.
INCORPORATES REACTOR DESIGN QTR 1, 81 NONE S. FABIC AND OPERATING DATA BASES IN RELAP5, T-91 FRAGILITY DESCRIPTIONS FOR COMP 0HENTS AND QTR 1, 81 3-10 PLANT STRUCTURES J. RICHARDSON STRUCTURES. DOCUMENTATION OF AN OVERALL FRAGILITY DESCRIPTION DEVELOPMENT METHODOLOGY.
T-118 ULTIMATE HEAT SINK VERIFICATION EXP.
QTR 1,
81 3-8 ATMOSPHERIC TRANSPORT AND R. ABBEY DEPOSITION T-119 SOUTHEAST SEISMIC NETWORK. HISTORY OF QTR 1,
81 3-2 GEOLOGY AND SEISMIC R. BRAZEE INSTALLATION OPERATION RESEARCH.
2-1/2 CHARACTERISTICS YEAR SEISMIC ACTIVITY
SUMMARY
AND RESUME.
T-151 A COMPREHENSIVE REPORT WHICH WILL QTR 1,
81 3-7 SECONDARY CONTAINMENT B. BROWZIN RECOMMEND CRITERIA FOR NRR LICENSING SIRUCTURAL POSITIONS ON PUNCHING SHEAR WITH CCM3INED BIAXIAL TENSION.
T-152 COMPREHENSIVE REPORTS WHICH SUMMARIZE QTR 1, 31 3-7 SECONDA%f CONTAINMENT B. BROWZIN ANALYTICAL INVESTIGATIONS OF TEST RESULTS STRUCTURAL ON LARGE AND MEDIUM SCALE SPECIMENS.
T-155 SENSITIVITY STUDY OF SDIL-STRUCTURE QTR 1,
81 G. BAGCHI INTERACTION PHENOMENON FOR SOIL-STRUCTURE INTERACTION, SOIL PROPERTIES AND SOIL CONFIGURATION, HAVE PASSAGE AND AZIMUTH EFFECTS FOR ZION NUCLEAR POWER STATION USING CONTINUUM ANALYSIS APPROACH..
l T-180 OPERATING SEISMIC SAFETY ANALYSES QTR 1,
81 MECHANICAL ENGINEERING J.
BURNS CODE - SEISMIM.
T-184 ACCURACY OF FINITE ELEMENT METHOD AND QTR 1,
81 STRUCTURAL ENGINEERING C. BURGER LUMPED MASS METHOD IN SEISMIC ANALYSIS.
T-185 STUDY OF SEISMIC ANALYSIS MODELING QTR 1,
81 STRUCTURAL ENGINEERING C. BURGER TECHNIQUE IN VERTICAL DIRECTION. THE STATE-OF-THE-ART IN VERTICAL MODELING TECHNIQUE IS EITHER TO ASSUME RIGID FLOOR OR TO SIMULATE THE FLOOR FLEXIBILITY BY ONE CR SEVERAL SINGLE DDR SYSTEMS. THIS STUDY IS TO EXAMINE THE ACCURACY OF THIS ASSUMPTION AND TO EXPLORE BETTER ALTERNATIVES.
T-186 RESPONSE COMBINATION METHODOLOGY.
QTR 1,
31 3-9 SEISMIC SAFETY MARGINS J. O'BRIEN RESEARCH PROGRAM T-195 EXTENDED DRY-CUT IN SODIUM COOLED DEBRIS QTR 1, 81 2-2 POST ACCIDENT HEAT REMOVAL R. WRIGHT BEDS, ACRR TEST D-5.
T-200 ANALYSIS & PREDICTION OF MAJOR FLOODS QTR 1,
81 6-3 LIMITING CONDITIONS FOR W. VESELY OPERATIONS T-205 FULL POWER DECL BREAK WITH OFF-SITE QTR 1,
81 1-1 LOFT G. MCPHERSON POWER.
EARLY CORE QUENCH SEEN IN L2-3 IS NOT EXPECTED TO OCCUR.
T-208 RELIABILITY DATA MANUAL QTR 1,
81 NDHE J.
JOHNSON T-259 NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR QTR 1, 81 5-23 OCCUPATIONAL EXPOSURE J. FOULKE SITES PROTECTION T-61 ATWS TRAC VERSICH. PROVIDES ADVANCED QTR 2, 81 1-14 ADVANCED SYSTEM CODE H. ZUBER CAPABILITY TO ANALYZE ATWS IN LWR'S.
T-112 IODINE TRANSPORT BEHAVIOR WITHIN PWR QTR 2, 81 1-9 FISSION PRODUCT RELEASE R. SHERRY STEAM GENERATOR AND THE SECCNDARY SYSTEM UNDER SGTR ACCIDENT CONDIT0NS.
T-120 ANNA, OHID SEISMIC NETWORK.
INSTALLATION QTR 2, 81 3-2 GEOLOGY AND SEISMIC R. BRAZEE OPERATION AND RESEARCH HISTORY. SEISMIC CHARACTERISTICS ACTIVITY AND RESEARCH
SUMMARY
T-187 MODELING TECHNIOUES FOR NONLINEAR QTR 2, 81 3-11 PLANT COMP 0HENT AND D. REIFF SYSTEMS.
EQUIPMENT BEHAVIOR T-209 HAZARDS TO NUCLEAR POWER PLANTS QTR 2, 81 6-6 LWR RISK ASSESSMENT K. MURPHY T-125 RECENT VERTICAL CRUSTAL MOVEMENTS IN QTR 3, 31 3-2 GEOLOGY AND SEISMIC J. HARBOUR THE EASTERN U.S. WILL FURNISH IMPORTANT CHARACTERISTICS CORROBORATIVE INFURMATION ON FAULT MOVEMENTS, SEISMIC ACTIVITY AND INTRAPLATE EARTHQUAKE MECHANISMS.
T-210 FLOOD RISK SYSTEMS ANALYSIS QTR 3, 31 NONE
- 5. STURGES T-236 EXP. EVALUATION OF SYSTEMS COMPONENTS QTR 3, 81 5-6 FUEL CYCLE FACILITIES D. SQLBERG DURING LARGE PRES 3URE PULSES EFFLUENT CONTROL T-65 POST CHF HEAT TRANSFER CORRELATION QTR 4, 81 1-3 PWR-BCHT MSU/ THOMPSON VERIFICATION COMPLETE. SUPPORTS OR 1-4 BWR-BDHT HSU/ THOMPSON REPLACED CURRENT CORRELATIONS.
T-126 FINAL REPORT OF NEW ENGLAND SEISMD-QTR 4, 31 3-1 NRC/ STATE REGIONAL EARTH N. STEUER TECTONIC STUDY. TECTONIC PROVINCE SCIENCES BCUNDARIES DELINEATED, POSSIBLE EARTH-QUAKE SOURCE AREAS IDENTIFIED.
GED-TECHNICAL MAPS COMPLETED.
T-133 NEW MADRID / UPPER MISSISSIPPI VALLEY QTR 4, 81 3-1 HRC/ STATE REGIONAL EARTH H. STEUER SEISMOTECTONIC STUDY FINAL REPORT.
SCIENCES TECTONIC PROVINCE BOUNDARIES DELINEATED, PROSAELE EARTHQUAKE SOURCE AREAS IDENTIFIED GE0 TECHNICAL MAPS COMPLETED.
T-134 NEMAHA/MIDCONTINENT GRAVITY ANOMALY STUDY QTR 4, 81 3-1 NRC/ STATE REGIONAL EARTH N. STEUER FINAL REPORT.
TECTONIC PROVINCE BOUNDARIES SCIENCES 1.
DELINEATED. PROBABLE EARTHQUAKE SOURCE AREAS IDENTIFIED. GE0 TECHNICAL AFD SEISMIC MAPS COMPLETED.
T-145 LABORATORY SIMULATION OF TORNADO WIND-QTR 4, 81 3-3 ENVIRONMENTAL STRUCTURAL R. ABBEY LOADS.
DESIGN T-86 INCORPORATE IMPROVED UT IN CODE.
IMPROVED QTR 1, 82 1-21 NON-DESTRUCTIVE EXAMINATION J. MUSCARA CONFIDENCE IN EVALUATIDH OF FLCW SIG-HIFICANCE FROM ACCURATE FLOW SHAPE AND SIZE.
T-212 SELECTED ALTERNATIVES FOR MANAGEMENT OF QTR 1, 82 NONE J. CURRY RADI0 ACTIVE WASTE GASES T-72 VERIFY TRAC-2.
GIVES CONFIDENCE THROUGH QTR 2, 82 1-17 CODE APPRAISAL S. FABIC COMPARISON WITH LOFT, GERMAN AND JAPANESE TEST DATA.
T-109 LOCA (SINGLE ROD).
QTR 2, 82 1-10 PBF M. PICKLESIMER T-213 FIRE RISK SYSTEMS ANALYSIS QTR 3, 82 NONE S. STURGES W __
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DISTRIBUTION
- Office of the Executive Director for Operations 5
Office of Inspection & Enforcement 15 Office of. Nuclear Material Safety & Safeguards 25
- Office of Nuclear. Reactor Regulation 40
.. Office of Nuclear Regulatory Research 150 Office of Standards Development 25
- Office of Management'& Program Analysis -
25 l
1 l
.