}&./3?J- ORNL-5183 PROGRAMS •ll-TEMFEMTIIE US ClllEI lEJtCTM SAFETY STIIIES PRWRESS wrwi lawni 1,1974 • Iwe 31,1975 T. t. Col* J. P. Sondvrs *. R. Kosttn ORRL-51S3 Distribution Category OC-77 Contract Ho. H-7405-eng-26 HICB-TEKPERATURE GAS-COOLED REACTOR SAPETT STUDIES PROGRESS lEPORT FOR JANUARY 1, 1974, THROUGH JURE 30, 1975 Tt S. Cole and J. P. Sanders, Program Managers P. R. Kasten, Prograa Director Dole Published: July 1977 OAK RIDGE RATIONAL LABORATORY Oak Ridge, Tennessee 37830 operated by UWI01I CARBIDE CORPORATIOR for the EHERCY RESEARCH AMD DEVELOPHERT ADMIRISTRATIOR mr CiSlRIBUTICN OF THIS DOCUMENT IS UNUMlTEd 90UHBKD The Gas-Cooled Reactor Program being carried oat at the Oak Ridge national laboratory contribute to the development and evaluation of nign-tenpereture gas-cooled reactors (nTGRs) and of gas-cooled fast reactors (OCRs). The development programa that are •nonsored by the U.S. Energy Besearch and Development Administration (Etta) include the Iberian Utilisation Program (RGg fuel lecycle Development Program), the RGR base Technology Program, the HTCK Safety Programs, and the CCFt Progran. In addition to the Una progress, safety stadias and assess- •eats have aleo been perfornad at the request and under sponsorship of the Division of Systens and Safety and the Division of leactor Safety Research of the U.S. nuclear Regulatory Commission. These stadias con­ sisted of Independent, objective assessnents of the safety aspects of specific systens of power stations for which construction or operating licenses were being considered. This report documents the work perfomed under all phases of the RGB safety atodies at OWL. Major incentives for developing BTCBs are the prospects for economi­ cally attractive power production, improved fuel utilisation, the potential for obtaining low environs*ntal insect at a diversity of sites, the potential for high-tenperature direct-cycle and process heat applications, and the pertinence of the consonant technology to OCFBa. lecognition of these Incentives has led to the developaent, construction, and operation of several experimental and prototype nuclear power plants based on gas-cooled reactora that utilise various configurations of fuel, soderator, and gaseous coolant. More recently a number of construction-permit applications were considered within the U.S. for large central-station power plants based on a particular HTGR concept being offered commercially by General Atonic Company (GA). Much of the work for the Safety Prograa has concerned the study and assessment of the safety of these plants, such am the Summit Station for Delasrva Power and Light Company, and of their systems and components. Although plans for these power stations have since bean cancelled for economic reasons, the original commitment on the pert of the utility companies and GA represented a conviction that ill BLANK PAGE /£ iv 1Kb can be operated with a high degree of reliability and without ansae risk to the health sad safety off the public. The RCft safety stadias have as their objective to docoaeat the assessaeat aad/or development of technology that will inset* appropriate levels of safety wader postalatad accideat coaditioaa that coald affect either on-site or off-site personnel. all tasks sad assessaeats for the safety stadias are coordinated closely with similar work at GA and at the Los Alanos Scientific Laboratory and Broakhavea Ratioaal Laboratory, as well aa with gas-cooled reactor development efforts in Europe. Slgaiflcaat efforts oa RGB. safety-related work are being carried oat at Da in west Geraany, and anch work was conducted and reported by the Dragon Project in England before closure of the Project in March 1976. Before its closme, HTGR safety inforaation was exchanged under the EkDn/Dragon HIC1 Agreeaent. CONTENTS FOREWORD iii SUMMIT xlii PART I. HTGR SAFETY STUDIES FOR CAS REACTOR SAFETY BRANCH OF THE DIVISION OF REACTOR RESEARCH AND DEVELOPMENT 1 1. SYSTEMS AND SAFETY ANALYSIS 3 1.1 PLANNING GUIDE FOR HTGR SAFETY AND SAFETY-RELATED RESEARCH AND DEVELOPMENT 4 1.2 HTGR CORE SUPPORT STRUCTURE 6 1.2.1 Procedure 9 1.2.2 Reliability Pilot Study ..... 10 1.3 CONSEQUENCES OF CORE SUPPORT STRUCTURE FAILURE IN HTGRs 13 1.3.1 Reactivity and Temperature Effects 13 1.3.2 Fission Product Effects Associated with Hypothetical Failure of Core Supports 18 1.3.2.1 Dose Effect of Increased Plant Release During Nonal Operation .... 19 1.3.2.2 Coolant Activity Increase 22 1.4 STEAM GENERATOR REVIEW 26 1.5 REFERENCES 30 2. FISSION PRODUCT TECHNOLOCi 33 2.1 ACCIDENT HAZARD POTENTIAL OF KEY NUCLIDES 33 2.2 IODINE ADSORPTIONS AND DESORPTION 37 2.3 ASSESSMENT OF CESIVM TRANSPORT PARAMETERS 39 2.4 REFERENCES 43 2. PRIMARY COOLANT TECHNOLOGY EFFECT OF STEAM CORROSION ON CORE POST STRENGTH LOSS: CASE OF LOW-LEVEL LONC-TERM IN LEAKAGE 45 v vi 3.1 OBJECTIVE Alt) METHOD 45 3.2 PREDICTED COOLANT IMPURITY COMPOSITIOHS 45 3.3 GBAPHITE STRENGTH LOSS DDE TO STEAM CORBOSIOII 49 3.4 PIEDICTED CORE POST BUBNDFFS AMD STRENGTH LOSS AT ERD OF REACTOR LIFE 51 3.5 REFERENCES 54 SEISMIC AND VIBRATION TECHNOLOGY PEELDflHARY ASSESSMENT OF THE GENERAL ATOMIC COMPANT CORE SEISMIC PROGRAM 57 4.1 SCOPE 57 4.2 PROBLEM DESCRIPTION 57 4.3 DESIGN CRITERIA 58 4.4 CORE ASSEMBLY 60 4.5 REACTOR VESSEL INTERNALS 62 4.6 REACTIVITY CONTROL SYSTEMS 64 4.7 EXPERIMENTAL PROGRAM 66 4.8 ANALYTICAL PROGRAM 72 4.9 SOMMART AND RECOMMENDATIONS 77 CONFINEMENT COMPONENTS — PCRV TECHNOLOGY ASSESSMENT STUDY 79 5.1 STATE-OF-ART INTWHATIOR REVIEW 79 5.1.1 Materials Characterization 79 5.1.2 General Design Philosophy . 80 5.1.3 Model Testing 80 5.1.4 New Materials Development 81 5.1.5 InstriaMntation 83 5.1.6 Liners ... .......... 83 5.2 REVIEW OF WORK IN PROGRESS , 84 5.2.1 PCRV Development in Germany, France, and England 84 vii 5.2.2 PCRV Concrete Materials Evaluation Studies 85 5.2.3 IIscussions at General Atomic Coapany ..... 87 5.2.4 Meeting of the ACRS High-Teaperature Gas-Cooled Reactor Subcommittee, Denver, Colorado, January 30-31, 1975 90 5.3 PCRV SAFETY ANALYSIS STUDIES 95 5.3.1 Analysis Method Selection and Refinement .... 95 5.3.2 Analysis of Ohbayashi-Guni Test Vessel 99 5.3.2.1 Description of Analytical Models ..... 99 5.3.2.2 Conclusions 104 5.4 REFERENCES 104 6. PRIMARY SYSTEM MATERIALS TECHNOLOGY 105 6.1 ASSESSMENT OF METALLIC MATERIALS TECHNOLOGY i3R HTGR PRIMARY SYSTEM AND CC3E SUPPORT STRUCTURES .... 105 6.2 ASSESSMENT OF MATERIALS IN HTGR REACTIVITY CONTROL AND SHIELDING COMPONENTS 109 6.3 ASSESSMENT OF CORE SUPPORT POST GRAPHITE Ill 6.4 REFERENCES 113 7. SAFETY INSTRUMENTATION 115 7.1 REACTOR TRIP SYSTEM 115 7.1.1 Logic and Actuation 115 7.1.2 Reactor Trip Inputs 119 7.2 CORE AUXILIARY COOLING SYSTEM (CACS) 123 7.3 STEAM GENERATOR ISOLATION AND DUMP SYSTEM 127 7.4 CAHE (HELIUM) ISOLATION 128 7.5 ORIFICE CONTROL SYSTEM 129 7.6 MOISTURE DETECTION IN HTGRs: PRESENT STATUS AND DEVELOPMENT NEEDS 130 viii 7.6.1 Requireaents 130 7.6.2 Operating Experience in Detecting Steam Generator Leaks in HTGRs 130 7.6.3 Comparison Among Methods of Hygrcaetry 133 7.6.3.1 Electrolytic Hygrometer . 133 7.6.3.2 Optical Dew-Point Hygroaeter 133 7.6.3.3 Infrared Detectors 134 7.6.3.4 Microwave Methods 134 7.6.3.5 Thermal Conductivity Sensor 135 7.6.3.6 Nuclear Magnetic Resonance (NMR) .... 135 7.6.3.7 Helium Afterglow Monitor 136 7.6.3.8 Other Methods 136 7.6.'* Summary and Recommendations for Developmental Work 137 7.7 BRIEF SUMMARY OF CONCLUSIONS AND RECOMMENDATIONS .... 138 7.7.1 Reactor Trip System 138 7.7.2 Core Auxiliary Cooling System 140 7.7.3 Steam Generator Isolation and Dump Systen .... 140 7.7.4 Core Auxiliary Heat Exchanger Isolation System 140 7.7.5 Moisture Detection in HTGRs 140 7.8 REFERENCES 141 ?ART II. HTGR Si*r£TY STUDIES FOR THE GAS-COOLED REACTORS BRANCH OF THE DIVISION OF REACTOR LICENSING, USNRC 143 8. ANALYSIS OF LOSS OF FORCED CONVECTIVF. COOLING ACCIDENT USING THE HEATUP CODE 145 8.1 DEVELOPMENT OF THE HEATUP COMPUTER CODE 145 8.2 THERMAL ANALYSIS OF LOFC ACCIDENT INVOLVING SUMMIT POWER STATION 146 ix 8.3 THERMAL ANALYSIS OF LOFC ACCIDENT INVOLVING FULTON GENERATING STATION 148 8.3.1 Results of Computations 148 8.3.2 Thermal Analysis Involving Age Dependence in Fuel 149 8.4 REFERENCES 151 DEVELOPMENT AND USE OF THE COUPLED CONDUCTION-CONVECTION MODEL FOR CORE THERMAL ANALYSIS 153 9.1 DEVELOPMENT OF THE COUPLED CO! OUCTION-CONVECTION MODEL 153 9.2 THERMAL ANALYSIS INVOLVING BLOCKED COOLANT CHANNEL AWAY FROM EDGE OF FUEL ELEMENT 154 9.3 THERMAL ANALYSIS INVOLVING BLOCKED COOLANT CHANNEL AT THE EDGE OF A FUEL ELEMENT 15? 9.4 RESULTS OF PARAMETRIC STUDIES IN THE HIGHEST POWERED REFUELING REGION 158 9.5 TEMPERATURE DISTRIBUTION IN THE HIGHEST-POWERED REFUELING REGION FOR A TOTALLY CLOSED ORIFICE CONDITION 159 9.6 THERMAL CALCULATIONS INVOLVING A RAPID DEPRESSURIZATION ACCIDENT FOR THE FORT ST. VRAIN HTGR 159 9.7 REFERENCES 161 . DEVELOPMENT OF THE MULTICHANNEL CONDUCTION-CONVECTION PROGRAM, HEXEREI 163 10.1 HEXEREI CODE 163 10.2 FLOW DISTRIBUTION ROUTINES 165 10.3 REFERENCES 166 . COOLING SYSTEM PERFORMANCE AFTER SHUTDOWN 167 11.1 COMPUTER PROGRAMS FOR CALCULATING CAHE PERFORMANCE 167 11.2 TRANSPORT PROPERTIES FOR CACHE 16S 11.3 CAHE HEAT TRANSFER PERFORMANCE CALCULATIONS 173 11.4 REFERENCES 174 X EVALUATION OF CIRCULATOR AND MOTOR FOR THE CACS 175 12.1 REFERENCES 176 DEVELOPMENT OF THE CORE AUXILIARY COOLING SYSTEM CALCULATION, ACREROHS 177 13.1 SUBROUTINE HELIX 178 13.2 SUBROUTINE TUBE 183 13.3 SUBROUTINE DELTPR 184 13.4 SUBROUTINE AEOLUS 185 13.5 INVESTIGATION OF DBDA FOR FULTON REACTOR 188 13.6 FORT ST.