WE HAVE MADE A COMPARISON BETWEEN THE GE STRESS INTENSITY CURVE (FIG. 3-26 OF NEDE-21480) FOR THE ONE CASE WHERE A DIRECT COMPARISON CAN BE MADE NAMELY, AT AN A/T RATIO SLIGHTLY GREATER THAN 0.50, AS PROVIDED BY FIG. 7 OF THE PROGRESS REPORT. FOR THIS ONE CASE, THERE IS ALMOST EXACT AGREEMENT BETWEEN THE GE CURVE AND THE VPI TEST RESULTS. THE GE CURVE, WHEN CONVERTED TO NORMALIZED STRESS INTENSITY FACTORS, WILL ALSO PRODUCE A CURVE (AS A FUNCTION OF A/T RATIO) WHICH IS QUALITATIVELY SIMILAR TO THAT OF FIG. 7 OF VIP-E-76-25; IN THIS CASE, EXACT CORRESPONDENCE IS NOT TO BE EXPECTED BECAUSE OF THE MATERIAL DIFFERENCE IN THE DIAMETER-TOTHICKNESS RATIOS OF VESSELS INVOLVED. THE CLOSE AGREEMENT BETWEEN THE GE CURVE AND THE TEST RESULT FOR THE ONE CASE WHERE A VALID COMPARISON CAN BE MADE PROVIDES ASSURANCE THAT THE STRESS INTENSITY FACTORS USED BY GE IN THEIR EVALUATION ARE REASONABLE APPROXIMATIONS.

THIS PROGRAM HAS GENERATED SUBSTANTIAL DATA USED IN PREDICTING IRRADIATION EMBRITTLEMENT AND MARGINS TO FAILURE OF REACTOR PRESSURE VESSELS WHEN FLAWS ARE PRESENT, CONSIDERING MATERIAL PROPERTIES AND LOADINGS. THIS PROGRAM HAS PROVIDED INPUT TO NRC REQUIREMENTS FOR FRACTURE TOUGHNESS OF PRESSURE VESSEL MATERIALS DESCRIBED IN APPENDIX G TO 10 CFR 50 AND CONTRIBUTED DATA USED IN THE ASME CODE WHICH 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 REGULATORY GUIDE 1.99.

THE EFFECT OF COLD EMERGENCY CORE COOLING WATER ON HOT REACTOR PRESSURE VESSELS WAS CONSIDERED. THE RESULTING THERMAL SHOCK
COULD, UNDER "WORST CASE" CONDITIONS, 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 WOULD
HAVE BEEN PREDICTED. TO DESCRIBE THIS EFFECT, ONCE A CRACK IS LOADED WHILE THE MATERIAL IS VERY TOUGH, NO RAPID EXTENSION
WILL OCCUR.

BY LIMITING CONCERNS REGARDING THERMAL SHOCK TO REACTOR VESSELS TO THOSE TRANSIENTS THAT INVOLVE REPRESSURIZATION OF THE VESSEL, VENDOR ANALYSES AND NRC EVALUATIONS WOULD BE SIMPLIFIED. IT WOULD ALSO PROVIDE AT LEAST A PARTIAL ANSWER TO THE QUESTIONS POSED IN REG. GUIDE 1.2, "THERMAL SHOCK TO REACTOR VESSEL."

NRR HAS DISCUSSED THE TECHNICAL ASPECTS COVERED BY THIS RIL WITH RES PERSONNEL. RESEARCH REGARDING WARM PRESTRESSING IS STILL
UNDERWAY, ESPECIALLY ITS RELEVANCE TO CYLINDRICAL VESSEL WALLS. WHILE IT APPEARS TO BE A PROMISING PHENOMENON FOR LIMITING CRACK
EXTENSION DURING A THERMAL SHOCK, THE DATA ARE STILL INSUFFICIENT TO BE USED AS A BASIS FOR LICENSING DECISIONS. A DETAILED
TECHNICAL EVALUATION OF THE RESEARCH DESBRIBED IN THE RIL AND THE NRR ASSESSMENT OF ITS RANGE OF APPLICABILITY WILL BE COMPLETED
FOLLOWING RECEIPT OF MORE SUBSTANTIAL DATA.

RIL TITLE: PBF SINGLE ROD-POWER COOLING MISMATCH (PCM) TEST RESULTS SPONSORING OFFICE(S) RES

COMPLETED RESEARCH IS REPORTED ON SINGLE FUEL ELEMENTS EXPOSED TO POWER-COOLING MISMATCH (PCM) CONDITIONS IN THE POWER BURST FACILITY (PBF). THE RESULTS ARE OFFERED FOR USE IN DETERMINING POSSIBLE CHANGES IN REQUIRED DEPARTURE-FROM-NUCLEATE-BOILING RATIOS (DMBR'S) FOR ALL COMMERCIAL POWER REACTORS WHICH USE ZIRCALOY-CLAD URANIUM DIOXIDE FUEL RODS.

VARIOUS FUEL FAILURE MECHANISMS AND THE CONSEQUENCES OF FAILURES ARE EVALUATED IN THE SAFETY ANALYSIS OF TRANSIENTS AND ACCIDENTS. DEPARTURE FROM NUCLEATE BOILING (DNB) IS ASSUMED TO PRODUCE FUEL ROD FAILURE AND IS THE FAILURE CRITERION USED FOR MANY LICENSING 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 CONDITIONS.

THE DEMONSTRATED ABILITY OF MOST FUEL RODS TO EXPERIENCE DNB WITHOUT FAILURE SHOWS THAT THE CURRENT DNB CRITERION IS CONSERVATIVE. REQUESTS FOR LESS CONSERVATIVE FAILURE CRITERIA HAVE BEEN MADE BY THE INDUSTRY. TURBINE TRIP WITHOUT BYPASS (TTHOB) AND STEAM LINE BREAK (SLB) ARE NEAR-LIMITING EVENTS IN WHICH DNB IS PREDICTED TO OCCUR MOMENTARILY YET FAILURE BY THIS MECHANISM MAY NOT OCCUR. DEFINITION OF A LESS CONSERVATIVE FAILURE CRITERION FOR THESE EVENTS WOULD RELIEVE THESE LIMITING CONDITIONS. THESE PBF RESULTS, HRR WILL GIVE SERIOUS CONSIDERATION TO THESE VENDOR REQUESTS; HOWEVER, APPROVAL OF RELAXED FAILURE CRITERIA WILL BE CONTINGENT UPON THE EVALUATION OF OTHER NON-DNB FAILURE MECHANISMS.

BASED ON

THE SINGLE ROD PCM TEST RESULTS SHOW THAT THE CURRENT DNB FAILURE CRITERION IS CONSERVATIVE. THE WIDE RANGE OF CLADDING TEMPERA-
TURES DURING DNB WHEN TEST PARAMETERS ARE NEARLY THE SAME PREVENT A QUANTITATIVE ASSESSMENT OF THE MARGIN TO FAILURE. FURTHERMORE,
THE PAWEL CLADDING EMBRITTLEMENT CRITERION WOULD NEED 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. THEREFORE, PELLET/CLADDING INTERACTION (PCI) DATA FROM
THESE EXPERIMENTS WERE PROBABLY COMPROMISED BY THE TEST CONDITIONS. FOR EXAMPLE, FUEL EXPOSURE TIME AT POWER, FUEL PRECONDITION-
ING AND CLADDING TEMPERATURE WERE NOT IDEAL FOR PCI STUDIES.

CONCLUSIONS ON THE SUBJECT OF FUEL FAILURE PROPAGATION DRAWN FROM THE PCM SERIES MAY BE PREMATURE SINCE VENDOR FUEL ROD PRESSURE
CRITERIA HAVE BEEN CHANGED. NEW DESIGN CRITERIA, RECENTLY APPROVED BY NRC, ALLOW FOR INTERNAL ROD PRESSURE TO EXCEED THE EXTERNAL
SYSTEM PRESSURE DURING NORMAL OPERATION. 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 INFORMATION ABOUT FAILURE PROPAGATION.