Text

Summary of LWR Safety Research in the U.S.A.,Presented at the International Conference on Nuclear Power & ITS Fuel Cycle Held at Salzburg (Austria) on May 2-3,1977
	ML20049H970
Person / Time
Issue date:	06/30/1977
From:	Bennett G, Murley T, Tong L; NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
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References
	NUREG-0234, NUREG-234, PB-267-328, NUDOCS 8203110047
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SUMMARY

OFLWR SAbrIY RESEARCH INTHE USA Presented atthe International Conference on Nuclear Power and its Fuel Cgle i

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May 2-13,197/

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SUMMARY

OF LWR SAPtIY RESEARCH IN THE USA Prepared ty T.E.Murley / L S. Tong / G. L Bennett U.S. Nuclear Regulatory Comrrussion Pres mted at the IntemationalConference on Nuclear Power and its FuelCycle Sal. burg. Austria May 213,1977 sponsored ty the International Atomic Energy Agency I

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SUMMARY

OF LWR SAFE 1Y RESEARCH IN THE USA Table of Contents I

Title Page

1.0 INTRODUCTION

1 1.1 NRC LWR Safety Research Program Areas 1

1.2 EPRI Nuclear Power Research 2

1.2.1 Interrelations between NRC and EPRI 3

1.2.2 NRC/EPRI Joint Programs 3

2.0 CURRENT TECHNICAL STATUS AND FUTURE RESEARCH 4

2.1 Primary System Integrity 4

2.1.1 Fracture Mechanics and Welding 4

2.1.2 Irradiation Embrittlement Stress Corrosion and Crack Growth 5

2.1.3 Non-Destructive Examination 9

2.2 LOCA Heat Transfer and Hydraulics 9

2.2.1 Hydrodynamics of Two-Phase Flow during LOCA 10 2.2.2 Blowdown Heat Transfer 11

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2.2.3 Emergency Core Cooling 13 2.2.4 Integral Test and Code Verification 14 2.2.4.1 Semiscale Tests 14 2.2.4.2 LOFT Tests 14 2.2.5 Assessment of Claimed Benefit of Alternate ECCS 19 2.2.6 Containment Research 20 2.3 Fuel Rod Behavior 21 2.3.1 Basic Studies 22 2.3.1.1 Zircaloy 0xidation 21 2.3.1.2 Zircaloy Mechanical Properties 22 2.3.1.3 Pellet Properties 23 2.3.1.4 Fuel Rod Thernal Performance 24 2.3.1.5 Decay Heat 24 2.3.2 Fuel Rod Performance under Transients 24 2.3.3 Fuel Meltdown Research 27 2.4 Computer Code Development 29 2.4.1 Improvement of Existing Codes 29 2.4.2 Advance System Codes 30 i

2.4.3 Compenent Codes 31 2.5 Reactor Operational Safety 32 2.5.1 Fire Protection and Aging Evaluation 32 2.5.2 Human Engineering 33 ib l

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Table of Centents (Cont'd)

Paoe Title 35

3.0 CONCLUSION

S 36 4.0 EXPECl2D FUTURE RESULTS 36 4.1 Frimary System Integrity 4.2 LOCA Heat Transfer and Hydraulics 36

'7 4.3 Fuel Rod Behavior s

38 4.4 Computer Code Development 38 4.5 reactor Operational Safety 39

5.0 REFERENCES

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1 List of Tables j

i Title Pfg Table I Surinary of Test Results from Eight 6-in.

Thick Intermediate Test Vessels 6

Table II Sunsnary of LOFT Nonnuclear LOCA/ECCS Tests 16 Table III Cog arison of Measured, Estimated, and Calculated Fuel Rod Behavior 25 G

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Figure 1 Research Plan to Confinn the LWR LOCA Analysis Methods 50 Figure 2 Computer Code Verification Process 51 Figure 3 Reference Curve for Minimum Toughness 52 Figure 4 Effects of Residual Elements in Irradiated Steel 53 Figure 5 Compact Tension Specimen Test Results for ASTM A-533 Grade BC 1 Steel 54 Figure 6 Cold Leg Break in a PWR System 55 Figure 7 Measured and Predicted Core Inlet Mass Flow for Test S-05-2A 56 Figure 8 Penetration and End-of-Bypass Data (with Dimensionless Flows Based on W) 57 Figure 9 Effect of Water Subcaoling on Water Delivery Delay Time 58 Figure 10 Correlation of Lower Plenum Voiding Data from Various Steam / Water Experiments 59 Figure 11 Transient Critical Heat Flux Data 60 Figure 12 Correlation for Post-CHF Heat Transfer 61 Figure 13 Rate of Quench Front Propaga'..f on 62 Figure 14 Comparison of Reflux Calculation with FLECHT Data 63 l

Figure 15 Heat Flux as a Function of Superheat for Transition and Film Bailing at 60 psi 64 Figure 16 Semiscale Core Inlet Mass Flow Rate - Drag Disc and Turbine Water Composite 65 Figure 17 Semiscale Mass Flow Rate in intact loop Ccid 66 Leg l

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9 List of Figures (cont'd)

Title Page Figure 18 Semiscale Mass Flow Rate in Broken loop Cold Leg 67 Figure 19 Plan View of LOFT Test Assembly (Hot leg Break Ccnfiguration) 68 Figure pn LOFT Reactor Vessel and Internals 69 Figure 21 LOFT Test L1-1 Predicted Pressure vs Measured Pressure at the Outlet of the Reactor Vessel 70 Figure 22 Measured Liquid Level in LOFT Lower

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Plenum and Downcomer Following Hot Leg Break Measurement Location: Down-comer Instrument Stalk No, 1, Shown in Inserts 71 Figure 23 Comparison of Measured and Predicted Coolant Density in Lower Plenum of Reactor Vessel (Demonstrates Conservatism of Analytical Models in Prediction of Lower Plenum Sweep-Out 72 Figure 24 Comparison of Coolant Density Heasurements in Operating Loop Cold Leg from LOFT Tests L1-2 L1-3A and L1-3 (Demonstrates Excellent Reliability) 73 Figure 25 Comparison of Semiscale and LOFT Mass Flow per Unit System Volume in the Intact Loop Cold Leg 74 Figure 26 Comparison of Semiscale and LOF'. Mass Flow per Unit System Volume (Sumation from the Broken Leg) 75 Figure 27 Boiling Water Reactor Pressure Suppression Containment System 76 Figure 28 Schematic Layout of 1/5-Scale Pressure Suppression Experiment 77 V

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figure 29 Photograph of V64-Scale Model of Pressure Suppression Containment Experiment 78 Figure 30 Arrhenius Plot of the Total Oxygen Rate Constant for the Steam: Zircalcy-4 Reaction 79 Figure 31 Diffusivity of Oxygen in Beta Zircaloy-4 80 Figure 32 Total Circumferential Elongation as a Function of Burst T(

rature in Zircaloy-4 Tubes 81 Figure 33 Maximum Pressure vs Burst Temperature (Baseline Data Illustrating the Inverse Relatic.. ship Betwe sure and Temperature) 82 Figure 34 Rupture Strains During Transient Heating Burst Tests -of Zircaloy-4' 83 Tubes Figure 35 Comparison of Predicted and Measured Fuel Centeritric Temperatures at Various Burnups in Halden Reactor 84 Figure 36 Comparison of Decay Heat Standards and Data.

85 Figure 37 Measured Fuel Rod Power, Cladding Surface Temperature, FuelrTemperature, Axial Length, and Internal Pressure vs Time During PBF Test 86 Figure 38 Measured Concrete Surface Erosion Rates 87 Figure 39 Fourier Number for Development of the Haximum Temperature Difference Within the Layer vs Step Change in Rayleigh Number 88 s

figure 40 Comparison of COBRA-IV and Score Code at About 300 ms Af ter Blowdown 89 vi

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SUMMARY

OF LWR SAFETY RESEARCH IN THE USA T. E. Murley, L. S. Tong and G. L. Bennett U.S. Nuclear Regulatory Commission Washington, D. C.

20555 l.0 JNTRODUCTION The U.S. Nuclear Regulatory Commission has a vigorous water reactor safety research program which is r.anaged on a results-The oriented basis to provide information of use to NRC.

objective of this program "is to provide the basis and means for' reliable and credible analysis of the course of events in hypothetical accidents to nuclear reactors, and the estimated consequences of such events."1 1.1 NRC,'WR_ S_afety Research program Areas There are five principal areas of NRC research iri the field of water reactor safety:

1.

Safety design and protection cf the integrity of the reactor pressure vessel and piping. The researcn tepics covereo by vessel and piping incegrity studies are: (a) fracture mechanics (the elastic and plastic behavior-of graterials subjewted to cracks), (b) the possibility of stress corroeion cracking in tt.e piping, (c) the potential for radir. ion cmbrittlement of the vessel, (d) detection of flaws in welds, (e) detection of leaks in the vessel or pipes, and (f) the possible thermal shock of the cold ECC water impinging on the relatively hot reactor vessel.

2.

Thermal-hydraulic tests of'hypottetical accidents and the ef fectiveness of engir,eered safcquard features. This researen area, which receives the major share of the LWR safety research resources, is devoted primarily to studying the various aspects of the hypothetical loss-of-coolant accident (LOCA), which '.s one of 47 design basis accidents l

used by NRC to evaluate the safety margins of propsed

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t.O nuclear power plants. The research stucies in this area may be broadly grouped into (a) separate effects tests designed to study certain phases of a LOCA or ssiccted components and (b) systems effects tests designed to study the interrelationships of the various phases and

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components.

1 3.

Fuel rnd behavior in hypothetical accidents and associated failure limits. Research studies in th V area focus o,i the behavior of the fuel rod and its constituent parts during and af ter off-normal or accidcat conditions.

4.

Computer code developeent for accurately predicting the con-sequences of reactor accidents. hRC h spensoring the parallel development of advanced systems codes to assess the const:r.atism in the existing licensing computer i

codes. In addition NRC is sponsoring the development of componeat coc.cs to p mit a more detailed analysis of a given component under accident conditions, i

Figure 1 shows the research plan to confim the LWR LOCA analysis method. Figure 2 illustrates t!c process which will be used to verify the computer codes.

5.

Reactor operati 1 56fety. NRC recognizes that reactor operational sat., can be further improved through the results of.esearch programs, such as fire protection, human engineering, and aging evaluation of existing reactor safety components.

1.2 EFRI Nuclear Power Research The Electric Power Research Institute (EPRI) has a large research program addressing many phases of nuclea. power. An I

overview of the EPRI water reactor safety program was given by W. B. Loewenstein in Nuclear Safety, Vol. 16, No. 6 (November-

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December 1975).

The principal areas of investigation are:

- Water Reactor Safety

- Plutonium Recycle

- Primary System Integrity

- Reliability and Diagnostics l

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- fuel Performan;e and fuel Cycle

- Earthquakes and Tornadoes

- Advanced Nuclear Systems

- Operating Benchcarkers

- In-Service Inspection

- Plant Chemistry In addition, research is funded for general aspects of electric power generation and utilization such as environmental impact, enerav demano and conservation, and modeling energy requirements.

This work is only partially applicable to nuclear power.

1.2.1 Interrelations between NRC and EPRI Because NRC-and EPRI-Tunded nuclear research is closely inter-related, close cuoperation and comunication is in the best interest of both organizations to improve research efficiency Quarterly and minimize unnecessary duplication of effort.

NRC and EPRI research management have been meetings t'etweca fomation of the hRC. At the held since somectme predatir3 NRC-sponsored Fourth Water Reactor Safety Research Information 28, 1976 was Meeting, one af terncon session on Septe.scer devoted to presentations by EPRI contractors of their research EPRI reports are publicly available and copies are results.

sent to many of hRC's contractors.

1.2.2 NRC/EPP.1 Joint Programs Two research programs are being or will be jointly funded with The BWR Blowdown /ECC Program being performed by GE and CPRI.

scheduled to be completed in 1931 is jointly sponsored by NRC, Day to day management of the prosram is accom-EPill, and GE.

plished by a program management group consisting of one member from each of the sponsoring groups.

Prior to termiration of the PWR-FLECHT systems effects testing sequence at Westinghouse in 1974. EPRI had provided funds for NRC and EPRI have agreed in principle to an the program.

extension of the PWR-FLECHT program which would include systems The program is sct'eduled for completion in effects tests.

FLECHT provides unique reflood data because of the 1981.

full-length electrical heater rod simulation, full height steam generators and two laop operation.

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2.0 CURRENT TECHNICAL STATUS AND FUTURE RESEARC_H 2.1 Primary System Integrity The objective of research in Primary System Integrity is to confirm the safe design of reactor vessels and piping and to establish ways for reducing the failure probabilities, if required. Research programs are categorized into the following g oups:

Fracture Mechanics Icradiation Embrittlement, Stress Corrosion and Crack Groath Non-destructive Examination 2.1.1 Fracture Mechics and Welding Verification of the fracture analysis niethodologies has been advanced by the results of pressure vessel tests conducted under the auspices of PVP.C and AEC in the early 1960's.2-13 More recent rcults have been obtained in the early 1970's from boi;h small and intermediate scale pressure vessel tests conducted by Oak Ridge National Laboratory for AEC and NRC.I

The resulv of the ORNL tests, sumarized in Table I, show that for flaws less than half of the wall thickness in depth, a pressure of nearly three times design pressure must be applied to initiate rapid fracture. For one vessei having a flaw almost 90% - the wall thickness in depth, a pressure of 2.2 times design pressure was required to drive the crack through the wall. The results, which were predict 94 from both analytical methods and from small scale models, showed that the design basis safety margin against fracture of reactor pressure vessels is well founded.15 Other important tests have also been conducted to verify fracture analysis methodology. First, the analysis methodology for the effects of thermal shock has permitted correct prediction t.f the results of four tests on 21-inch diameter, 6-inch thick l

steel cylinders.16 The material was degraded to simulate irradiated steel, and each cylinder had a different flaw placcd in it. The cold simulated ECCS water (either 40'F or

-10*F for the simulant) flowing at a high rate, through the flawed cylinders initially at 550*F, caused crack extension

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and arrest in three tests, but no crack extension in the other, all as predicted. Second, the technique for evaluating the effect of sustained pne w tic loading upon a through-wall

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'Ihe U.S, Nuclear Regulatory Conr.ilesion water reactor safety research program consistP of five basic research areas: integrtty of vessel and pipings thermal-hydraulic tes;,

fuel rod behavior, code development and verification, and reactor operational safety.

Results from the vessel and piping integrity rese. arch have demonstrated the high safety margins in scaled ve.ssels and the analytical procedures for calculating vessel beha.vior under pressura. Die thermal-hydraulic tests have covered the various phases of a hypo-thetical loss-of-coolant accident (IDCA) and activation of the emergency core cooling system (ECCS). Tl!cse tests hav led to the developast of engineering correlations to describe the phenomena to further quantify the safety margins in cocnercial nuclear power plants. This report specifically presents as well as compares and evaluar.es between selected experimental data and analytical predictions from the initial tests in LOFT and E*fISCALE.

ispesqxxxx -.xiuex mautxioupsxx x adar axas mu six x Die fuel ischavior recearch has provided valuabic information en decay heat, cladding oxidation, fuel rod behavior, and fuel melting. Both the decay heat and the cladding oxidation have been shown to be lower than assumed in the licensing evaluations. The fuel behavior and thermo-hydraulic tesearch is being integrated into comptiter codes to be used to provide additional quantification of reactor behavoir under of f-normal conditions. Die reactor operational safety research is just beginning: it addresses fire protection, component aging, and hunin engineering.

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crack in a pressure vessel has been successful in predicting test results. In this case, pneumatic pressure was maintained against an elliptically shaped through-wall crack without further extension of the crack under the sustained load.

Suf ficient onalytical, expeiimental and theoretical progress has been made in crack arrest in the past several years, so that the two crack arrest theories which have been proposed, for which a great deal of confirmatory evidence is already in hand.17 converge to a unified basis. Two dimension &l, static and dynamic analyses are being developed for predicting material behavior related to fast fracture and track arrest from test A consistent specimens and related to analyses of components.

body of experimental data on crack arrest toughness has also been developed, with the result that much confidence can be placed on definition of a crack arrest toughness parameter.17 and standard test $1ec' mens a'4 testing methods ar,e now being written. A reference curve for mirimum toughness, called KIR*

has been developed from the data shown in Figure 3 and was incorporated in the ASME Boiler and pressure Vessel Code (Section !!!, Appendix C), in 1973. This curve is a representative lower bound of the lowest toughness values, dynamic and arrest, that have been measured.

For loads which impose clastic-plastic s'.resses in structures, the present analysis method using lis. ear elastic fracture mechanics (LEFM) is too conservative but useful engineering solutions are being obtained. The goal is fnr a valid criterion <

for elastic-plastic fracture.nechanics evaluations to be available by the late 1970's. Verification of the KgR curve for unirradiated datale has been completed and the verification for irradiated steels is underway, and an " irradiated KIR" curve should be available by the late 1970's. Analytical methods for evaluating crack arrest characteristics are under active developer.at and should be complete Ly 1978.

2.1.2 Irradiation Embrittlenent, Stress Corrosion and Crack Growth Factors which can sigaf ficantly reduce neutron irrad.ation embrittlement 'in pressurc vessel steel and welds were determined in the late 1960's and early 1970's.19 The limiting of resioual elements, mainly Cu and P, from the composition can reduce embrittlement in reactor vessel steel in nuclear radictica service. The influence of Cu, in particular, is illustrated in Figure 4.

The growing body of data on this subject (see reference 20 for an excellent review) has even resulted in the ability to quantify the effects, as shown by the curves in NRC Regulatory Guide 1.991 published in 1976.

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information on radiation embrittlement became available by a

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SUMMARY

OF TEST RESULTS FROM EICHT 6-IN.-THICK INTERMEDIATE TEST VESSELS l

f 60MINAL TEST FLAW DIMENSIONS FRACTURE FRACTURE VESSEL TEMPERATURE DEPTH LEhG1H'-

FLAW PRESSURE STRAlg NO.

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(in.)

LOCATION (ksi)

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l V-1 130 2.56 8.25 BASE METAL (o)b 28.8 0.92 h

V-2 32 2.53 8.30 BASE METAL (o) 27.9 0.19 V-3 130 2.11 8.50 WELD METAL (o) 31.0 1.47

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V-4 75 3.00 8.25 WELD METAL (i)d 26.5 0.17 c

75 3.10 8.10 BASE METAL (o) 26.5 0.17 V-6 190 1.87 5.25 WELDMETAL(oh 31.9 2.0 8

190 1.34 5.20 BASE METAL (i) 31.9 2.0 190 1.94 5.30 WELD HETAL (1) 31.9 2.0 h

V-5 190 1.20 3.75 BASE METAL (1)I 26.69 0.25 V-7 196 5.30 18.0 BASE METAL (o) 21.49 0.12 V-9 75 1.20 3.75 BASE METAL (1)I 26.9 1.05 a0UTSIDE CIRCUMFERENTIAL STRAIN ON CENTER LINE OF VESSEL REMOTE FROM FLAW.

b(o): OUTSIDE SURFACE. (1): It310E SURFACE.

CCONTAINED TWO FLAWS dFLAW WHERE FRACTURE OCCURRED.

' CONTAINED THREE FLAWS.

IN0ZZLE CORNER FLAW.

9 LEAK-BEFORE-BREAK.

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early 1975 on static and dynamic fracture toughness obtained fresi irradiated 4-inch thick specimens 22,23 which show, as seen in Figure 523 that irradiation-cordition toughness rises rapidly with temperatures in the transitiGn region and the absolute level of fracture toughness is quite high; although not shown hese, the magnitude of the vpward shift in transition temperature by th.se large specimens is quite well predicted by smaller surveillance type specimens.2; The reduction of radiation embrittlement by post-irradiation heat treatment has been shown feasible in many experimer.tal studies (see p. 197 of reference 20). Irradiation programs and fracture toughness recovery are both underway to provide the necessary data for improving the confidence in embrittlement safety analyses.

The ultimate solution for elimination of intergranular stress corrosio'. cracking (!GSCC) in austenttic stainless steels has not yet been found, but some of the price factors causing it are known.d.-2c Thus, it becomes important to defir.e all the factors and their interrelationships and to extend these findings to field application. A primary reason for the susceptibility of stainless steel to intergranular stress corrosion cracking is sensitization--the depletion of Cr from solid solution at the grain boundaries as it cochines with C to form chromium carbides.

One way to preclude IGSCC in service is to assure that the matertal is not sensitized and therefore not susceptible to intergranular stress corrosion cracking. An electrochemical test is under development that will permit detectior of sensitized material in the field. Already this development has produced such correlations as to permit use of the test for materials qualification tests.27 Although stress corrosion cracking cannot be fully controlled in service, it nevertheless is not comidered to be an urgent safety consideration.28 Because stress corrcsion cracking creates such operational costs, however, much research is underway to effect positive solutions. Corrosion of steam generator tubing care lead to leaking or failure of the tubes.

The operational cost loss is so great that much industry and gne-~~wal research is sonderway to eliminate it or at least mitigate the effects.

NRC has recently establie.hed a steam generator tube integrity program at Battelle Pacific Northwect Laboratories (PNL) to experimentally evaluate the safety margins of degraded FWR

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> i steam generator tubes under normal operating and postulated accident conditions. The program will also consider the i

effects of denting of the steam generator tubes. In all cases the test tubes will be representative of tubes presently installed in PWR steam generators.

To accomplish these objectives, PNL will begin by crnducting a thorough pretest inspection of the steam generator tubes which are to be tested. This pretest inspection is dcsigned to fully characterize ti;e tubes and to detect ar.y pre-service defects. A number of the Lubes will then have defects artif t-cially induced in them by mechanical ar.d chemical means. The creation of the defects will be carefully controlled to make them similar to those defects found in PWR steam generators.

After the defects have been created in the tubes, the tubes will undergo a standard eddy current inspection to enable the develooment of a correlation between the inspection results and the test results. Haxt. Fill will perform a series of burst and collapse tests on both baseline and the artificially defected steau generator tubes using the environmental conditions associated with normal ani postulated accident conditions.

The margins of safcty for leakage, burst, and collapse pressure will then be developed as a function of flaw geometry, tube size and test conditions.

Based on the actual test data, empirical models will be developed to relate the failure pressures and leakage rates to the tube degradation characteristics. The goal is to have predictive modeis which can be used to determine the burst and collapse pressures of tubes based on oper.iting conditions and the measured results of defects from an actual str.am generator field inspection. In this manner, the margins of safety under normal operating ard postulated accident conditions may be determined. This information will be useful in deciding the extent of tube plugging to be allowed. Finally, criteria will be developed for the evaluation of the lifetime remaining in flawed tubes.

Crack growth caused primarily by cyclic stresses in the primary system is addesssed in detail in the ASME Boiler and Pressure Vessel Coc'e,Section XI. An effort is currer41y underway to define those specific loading parameters which cause greatest crack growth. At present, conservative solutions are derived for crack growth questions; thus, reactor safety is not compromised with respect to crack growth.

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,a L I ti 2.1.3 Nor.-Destructive Examination The in-service inspections forming the basis for ASME code evaluations are perfonoed primarily by ultrasonic testing (UT). Thus. UT techniques are being upgraded to yield far better data on flaw characterization. A digital synthetic array processing procedure for improved lateral and longitudinal resolution of ultrasor.ic images is in late stages of development with excellent progress being recorded.2%30 5 team gercrator tubing is inspected by eddy current (EC) techniques. A new appendix (Appendix IV) for eddy current examinat ion was incorporated into Section XI of the ASME Boiler and Pressure Vessel Code on Inservice Inspection during 1976. Improvements in eddy current instrumentation and charac-terization of signals are being made primarily using multiparam-eter and multifrequen y approaches.31.32 The techniques are of nigh importcnce because EC inspection is rapid and is the currently accepted method for inservice inspection of steam or.erator tubing.

Laboratory studies are underway to use acoustic emission (AE) for monitoring of stress corrosion cracking, and initial favorable results have been attained.38 AE is also being used to monitor for cracking during welding. In this pr.1cedure.

begun in 1971, transducers are mounted on the component, and as uelding proceeds, any cracking that develops as the weld bead cools is detected by the system so that the bad weld can be removed and the cause of the weld defect corrected. A weld monitor has been perfarTning well in service in a non-nuclear shop since 1974, and extensive development is underway in nuclear welding shops; the results to date are encour'ging.3

2.2 LOCA Heat Transfer and Hydraulics The objective of the LOCA heat transfer and hydraulics research is to provide additional quantitiative information on the thermal-hydraulic aspccts of the hypothetical LOCA. This research encompasses the following groups:

Hydrodynamics of Two-Phase Flow during the LOCA Blowdown Heat Transfer Emergency Core Cooling Integral Test and Code Verification Assessment of Alternate ECCS.

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! A LOCA can be postulated for both a PWR and a bWR. Pcause the postulated PWR LOCA is more complex than the BWR.0CA, the PWR thermal hydraulic studies will be discussed first especially since some of them are applicable #:,e both cases. A cold-leg break in a PWR is shown in Fig.e 6.

The right-hand side shows the assumed broken loop while the left-hand side of the figure represents the unbroken loops (which would be three loops for a four-loop PWR).

2.2.1 Hydrodynamics of Two-Phase Flow during LOCA The hypothetical LOCA in a water reactor involves flow discharge at the break and coolant depressurization inside the system.

These transient effects add more complications to the already very complex two-phase flow and boiling heat transfer.

A sample 35 of calculated core flow during a postulated LOCA and data from a Semiscale test are shown in Figure 7.

The calculations are based on the RELAP-4 computer code.36 The average penetration rate of the emergency core coolant (ECC) water in the downcomer seems dependent on the annulus circumference instead the annulus gap as shown in Figure 8, c andJfin are calculated using the annulus circum-where J7)as the characteristic length.37 ference It should be noted that these results are based on 1/15-scale testing data. The delay in penetration was found to be greatly reduced by the incr ase in subcooling of ECC water, as shown in Figure 9.

This delay will also be reduced in a gap larger than 2 inches as indicated in the recent LOFT test No. L1-2 for ECC bypass.

A best-estimate correlation for ECC bypass was reported by Block, et. al. 38 based on the test data from 1/15-scale models.

The Creare flooding correlation, based on time average values, is:

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p is in psia f 16. for Cresre tiat plate geometry b = 130, for all cylindrical geometries It should be noted that a large size annulus should have a beneficial effect on the water penetration based on an expected tendency of a tilt flow pattern in a large annulus witn water going down on the intact loop side and steam coming up on the broken lood side. Further study will be conducted at BCL in the 1/15-and 2/15-scale models.

When the reverse core steam flow comes through the lower plenum, '.t entrains some of its liquid and thus lowers the water lesel in the lower plenum. The criti:al water level (belot which no liquid entrainment is expected) can be correlated by a b.ber Number approach as shown in Figure 10.

2.2.2 Blowdown Heat Transfer To pradict LOCA behavior, phenomenological heat transfer correlations are required. Both the ext:, ting boiling heat transfer correlations and those under development are discussed in the following paragraphs. Development and verification of these correlations will be based on data obtained from various d

blowdown tests.

During FY-1976, five blowdown heat transfer tests were performed in the (PWR) Thermal Hydraulic Test Facility (THTF). The THTF data, which are based on 12-ft-long heated rods, compare favorably with the Semiscale data, which are based on 5.5-ft-long heated rods (see also Section 2.2.4).

The BWR blowdown heat transfer data obtained in the Two Loop Test Aoparatus (TLTA) compare favorably with calculations made with the conservative evaluation model. The comparison of TLTA heat transfer data with existing published heat transfer correlations will be performed in a verification program at INEL.

To incorporate transient CHF and dryout during blowdown. Hsu 3

and Beckner 9 have suggested a correlation for predicting the heat flux at the location of sudden wall temperature rise for a wide range of local void. Their correlation is 1/2 9 CHF ~ 9 steam =

[ 1.76 (0.96 - Ti) ]

(2) 9" W3 at X=0 l

1 l

l '

i l

where 9" stean = (T,,33 - Tsat).h0fttus-Boelter i

4 is the CHF predicted by W-3 (40) for zero qualitv..

q"W3 a t X=0 6

3

= 0.e579 x 106 (1.04 + 0.1480,'10 ) (2.lo2-0.530p/10 ),

{0.2664 + 0.8307 exp(-3.1510e))

3 is the (axially) local void fraction averaged over the flow G is in ib/ft hr, p is in psi, De is in inches.

2 cross section.

The corrparison of Equation (2) with existing data is shown in j

Figure 11.

Future work with microscopic observatien and measurement on the boiling surface at the threshold of transition flow will enhance understanding of the mechanism of CHF.

The theory of spontaneous nucleation on a heated surface as a cause of transient CHF is being studied by Henry at ANL.42 j

After CHF is exceeded, the surface beat transfer decays as the wall superheat increases and/or the local quality increases.

This decay in heat transfer is caused by a reduction o" wet area or a reduction of the chance of the droplets rewetting.42,43 The period of transition boiling is usually short, and particularly so at a high heat flux where the rate of wall temperatures rise is high. When the wall is completely dry, tne heat transfer mechanism becomes a film boiling with a steady, high wall temperature. Further heat transfer decay is gradual, as the local quality increases and the inverted annulas flor pattern converts to a dispersed flow.

beingmadeofthemodifiedTong-Youngcorrelation.***g5In treating the re Based on c rrently available bu1dle film boiling data, the following combined correlations can be used as a best estinate for analyzing transient film boiling:

"Dougall-Rohsenow correlation 46 can be used for the entire two-phase flow range, except for that P > 900 l

psia and G < 0.1 x 107 lb/hr ft2 where Groeneveld 5.7 47 should be used."

correlation l

l l

l 1

l I

I i d L

t The reason for using a combination of correlations is that the L

Dougall-Rohsenow correlation was based on a dispersed flow model without taking non-equilibrium into consideration, in the low flow region, non-equilibrium is strong with a large C

steam superheat, in which case the Groeneveld correlation with I

non-equilibrium should be used.

In a dispersed flow, the superheating of vapor takes away the heat which would otherwise be used to evaporate the droplets l

if thermodynamic equilibrium were. maintained. Thus, the j

actual quality is lower than the equilibrium quality. To account for the non-equilibrium effect in quality, recently l

4 f

Chen'.e has empirically determined the ratio of the actcal 3

quality to the equiU brium quality.

i i

If the wall superheat and void fraction are reduced, the probability of liquid contact increases. Chen found that the liquid contact probability decreases exponentially with increasing a's, while the transient heat flux during liquid contact, i

q' C, can be expressed by a complicated function to account for both transient conductive neating of liquid and subsequent evaporation. The final fom of Chen's correlation is

-A(aT - AT,)

+ q",

(

o" total = q"LC (1-X) e 3

3 is the heat flux to the vapor, ATso is the wall where q"l immediately downstream of point of CHF, and A is a superhea decry coef ficient. The compaiison of Chen's calculated heat flux dgainst experimental data is shown in Figure 12.

The futbre work on post-CHF heat transfer should be concentrated on the measurement and correlation of the noncquilibrium and slip effects. These studies are being carried out by Griffith of MIT. Chen at Lehigh and Groeneveld at AECL.

f 2.2.3 Emergency Core Cooling Ti.e reflood heat transfer data obtained from the NRC-sponsored FLECHT program" are currently used for predicting the clad temperature before it is quenched. The more recent low flooding rate FLECHT data appear to be consistent with the older FLECHT data obtained at higher flooding rates. The rate of quench front propagation (rewetting) can be also estimated for FLECHT data or from Figure 13. A more proper approach is to (ormulate a phenomenological model to treat the whole reflood channel as en integrated system with different elevations being cooled by various modes of heat transfer. One attecpt being made in this direction is exemplified by the REFLUX program developed by HIT 0 in which all modes of flow boiling are involved.

5

V

.. _ - REFLUX was able to predict FLECHT and Semiscale with moderate j

success (peak temperatures were credicted to within about 10%.

but the quench time could b1 overpredicted). One example of I

50 is shown in Figure 14. This effect is I

a REFLUX prediction in the right direction; howeser, further refinement is needed and

}

expected. Also, the FLECHT facility will be modified to model I

the upper plenum more realistically.

l Since heat transfer is of prime importance to reflood, the l

heat transfer in the individual models needs to be examined.

1 l

The best-estimate heat tr6nsfer correlation for transition i

boiling developed from low void FLECHT forced reflood data S l

4 SI as shown in Figure 15.

was reported by Hsu 2.2.4 Integral Test and Code Verification I

2.2.4.1 Semiscale Tests Twenty-six tests were completed in Semiscale in FY-1976.

i Extensive comparisons have been made between test data obtained in the Semiscale Mod-l facility and prediccions made with the RELAp-4 system code. The focal point of attention in these s 2-se. is the prediction of core flow data / code comparisons which is the single parameter most strongly influencing the core thermal response. Experimentally, the core flow is highly sensitive to botn the flow leaving the system through the broken loop and the flow supply from the intact loop. The close agreement between the measured and the predicted core flows in Figure 1655 is indicative of the precision of RELAP-4 in predicting the loop hydraulic behavior. Figures 17 and 1855 which compare measured and predicted parameters in the intact and broken cold legs give further examples. In general the predictive values fall well within the error bands of the experimental data.

From tests in Semiscale, researchers have discovered the importance of the local mixing effect of unheated rods (analogous to a control rod thimble) on the delay of transient CHF during blowdown. The presence of three unheated rods changes the time delay of CHF from 0.5 sec to 3 sec.

2.2.4.2 LOFT Tests The LOFT program is unique to LWR safety research because it is the only complete reactor system dedicated to LOCA/ECCS research. LOFT is designed to provide experimental data relevant to the events which would occur during a postulated

~~

m 1

i ;

These data are then used to vertry and improve the LOCA.

computer codes which predict the response of a nuclear plant to a hypothetical LOCA.

Figure 19 is a schematic of LOFT and Figure 20 shows a cutaway Like Seiriscale, LOFT may be thought of of the reactor vessel.

as 1-1/2-loop system, i.e., it has one complete operating loop to simulate three of the operating loops in a four-loop PWR and a blowdown loop" or "l/2 loop" to simulate the loop that is postulated to break in a hypothetical LOCA.

LOFT contains all of the essential components of 5 large PWR, The first core is 5.5-ft long and 2 ft although scaled down.

in diameter (about 1/70 the volume of a large PWR core).

This core contains 13')0 fuel rods. However, the power-to-volume Furthermore.

radio in LOFT and a large PWR are approximately equal.

the subvolumes, such as the inlet plenum, core region, outlet plenum, outlet piping, steam generator and inlet piping are designed to have relative volumes approximately equal to those l

of a large PWR, The blowdiwn loop includes orifices to simulate various break sizes, and it contains a steam generator simulator and pump The simulator to model the effects of these components.

capability exists to simulate either hot leg or cold leg breaks. Quick opening vah es in tha blowdown loop can be opened in from 10 to 50 msec to simulate the initiation of the The escaping coolant is then collected in a suppression break.

tank which can model the various expected PWR containment backpressure transients.

The LOFT ECCS model tU.se systems in a PWR which iteject water into the cold legs with additional ccpability to inject into the hot legs, upper plenum and lowc plenum. The ECC is supplied by either of two high pressure injection pumps, by e.ther of two pressurized accumula'ces, and by either of two low pressure injection pumps. The general features of the LOFT facility and the LOFT experimental program are discussed in Reference 99. An article by G. D. McPherson now in printing in Nuclear Safety provides a good overview of the tests completed to date.

Four r..innuclear LOCA/ECCS tests have already been comficted in LOFT and two more nonnuclear tes" are scheduled. A summary of these four tests is given in Table 11 and they are disc used in the following paragraphs. All tests began at a primary coolant temperature cf 540 F.

- _~_

.= -

TABLE II

SUMMARY

OF LOFT NOHNUCLEAR LOCA/ECCS TESTS INEL REPORT NUMBER SYSTEM TE3T EREAK B2EAK PRES 50:iE ECC FOR EXPERIMENTAL DESIGNATION SIZE TYPE (psisJ,

INJECTION DATA REPCRT L1-1 1/2 Full Break Area Hot leg 1322 Cold Leg TREE-NUREG-1025 L1-2 Full Break Area Cold leg 2250 Cold leg TREE-NUREG-1026 (delayed)

L1-3 Full Break Area Cold leg 2250 Lower TREE-NUREG-1065 Plenum

,e TREE-NUREG-1027 L1-3A Full Break Area Cold leg 2250 Lower Plenum l

e I

1 m ena

g

_ _... l Test L1-1 e

This was the first LOCA/ECCS test to be run in LOFT, and for i

facility checkout purposes, it was deliberately designed to be mild. A hot leg break was chosen so that the flow would not i

teverse in the reactor vessel. This experimental configuration coupled with a 50%-of-full-break-area opening moderated the hydraulic forces.

Figure 21 shows good agrec-*nt between the predicted and experimental pressures in the operating loop not leg. Also shown are the Semiscale data, all in all a good match.

5 In addition to this agreement between predictions and expertraents, the L1-1 test showed:

The measured..yoraulic loads were less than expected.

ECC bypass was less than predicted.

The sweep-out of water from the lower plenum was icwer than predicted.

These last two conclusions can be seen in Figure 22 which is a composite of liquid level measurements taken in the downcomer and lower plenum. The data indicate that after the duwncomer empties at 10 sec, the lower plenum inventory remains essentially constant until the ECC begins to ' arrive ac 40 sec. Thus, we would expect less bypass and earlier r flood than predicted by conservative calculations.

Test L1-2 Tnis test was a full-sized cold leg break run from full system p* essure; thus, one would expect the most severe hydraulic loads. In addition, the test was run with delayed ECC injection to permit a measurement of the hot wall delay without interference from any reversed primary coolant flow up the downcomer. The important conclusions were:

The hot wall delay is much shorter than predicted.

Sweep-out of water from the lower plenum for a full sized break was less than predicted (see Figure 23).

Loads in the suppression tank were much lower than predicted.

(This should be of interest in BWR safety calculations.)

y w-

_. _. _ _ Tests L1-3 and L1-3A The third LOFT LOCA/ECCS test was originally designated L1-3, 6

but it had to be repeated as test L1-3A because of a procedural error in the first attempt. Test L1-3 was successful in all ways except that, while ECC was injected from the high pressure and low pressure injection systems as planned, the accumulator injection did not occur. The data are still useful and will be published. The repeat test, L1-3A was identical to L1-3 except that ECC was injected from all three systems. L1-3A simulated a full area break in the cold leg with blowdown initiation from 2250 psig. ECC was injected at normal times directly into the lower plenum to check for possible lower plenum voiding caused by liquid entrainment in the steam flowing from the core.

The important conclusions from this test are:

There was no sweepout of the ECC water injected into the lower plenum.

There was excellent repeatability of behavior during transtr.nts initiated at the same conditions (see Figure 24).

As in all the LOFT tests experiments, to confirm the understanding of the integral scale effects, a simulation of the LOFT L1-3A test wat. conducted in the Semiscale Mod-l as test %-01-3, under near identical initial conditions. Preliminary evaluation of the data from LOFT test L1-3A and Semiscale test S-01-3 (overlayed in Figures 25 and 26) indicate that LOFT and Semiscale response is quite similar with the exception of the downcomer response (in which two-dimensional effects would be expected to prevail in LOFT). However, primary system pressures, intact loop flow and broken loop flow all behave similarly between Semiscale and LOFT which indicates that the preservation of power-to-volume ratio is a good scaling criterion for the blowdown of both systems. Additional tests and analyses will be required to fully substantiate the conclusion that we understand how to scale complex LOCA phenomena.

In conclusien, the important points to note on the first three LOFT nonnuclear tests are:

Scaling effects between Semiscale and LOFT seem to be understood.

~

w

q l l 1

RELAP-4 predictions, in general, compared well with the i

data and the reaso*s for any significant discrepancies 4

are understood.

ECC was delivered rapidly to the core inlet regions in all cases. There were no appreciable effects from ECC bypass, hot wall delay or entrainment.

Repeatability in the LOFT system is ex:ellent.

LOFT suppression tank data indicate there may be less dynamic load on a BWR torus than currently conservatively estimated.

}

2.2.5 Assessment of Claimed Benefit of Alternate ECCS

.1RC is engaged in an analytical / experimental program to confirm the margins of safety inherent in alternate ECCS concerts proposed by reactor vendors.

One of the alternate concepts includes consideration of simulta-neous cold leg and upper head injection.57 Accumulator injection in the upper head comences early in blowdown and is intended to accelerate system depressurization leading to an earlier start of relooding than with current systems.

Westinghouse [W) has conducted experimental investigations on the effectiveness of the upper head injection (UHI) concept in their G-1 facility 58 while the Japanese conducted UHI tests in their ROSA-Il facility.59 Evaluation of the data from W and ROSA-II is underway to determine the benefit of the UHI system.

In addition, NRC is sponsoring a modification to the Semiscale facility to test UHI concepts beginning in April 1978.

Combustion Engineering is following the effort of Kraftwerk Union (KWU) of Germany in their investigation of the principles underlying simultaneous hot and cold side injection. The principle behind this combination of injection locations was to condense the reflood generated steam leadirig to a reduction l

in steam binding and an increase in the reflood rate.60 Tests at KWU were initiated in November 1976 and scoping tests have been conducted in the Semiscale Mod-l facility. These tests indicate an earlier core quench time than with cold leg injection only but the core quenched from the top rather than from the bottom.

Current Babcock and Wilco.. designs incorporate vent valves in the core barrel which allow comunication between the upper annulus and the upper plenum. The vent valves serve to reduce l

l

. the pressure in the upper plenum during reflood thereby enhancing the reflood rate. Confirmation of this design is planned during the Winter of 1977 in the Semiscale Mod-l facility.

The vent valve principle can be tested in the current LOFT g

design using the reflood assist bypass valves which connect the broken hot and cold legs.

To improve the basic understandings of LOCA hydraulics, the effect of ECC injection locations, such as lower plenum injection, is also included in the LOFf design. Current investigations 61 in the Semiscale Mod-l configuration have illustrated the advantages of lower plenum injection relative to cold leg injection. With the lower plenum injection, residual water loss from the lower plenum and the loss of accumulator water through the break were minimized. In effect, lower plenum injection eliminated the countercurrent flow processes attendant with the cold leg injection and resulted in an early initiation of core reflooding. A similar effect was noted in the LOFT nonnuclear test L1-3A.

2.2.6 Containment Research NRC is sponsoring work at the Lawrence Livermore Laboratory (LLL) to better quantify the response of a BWR Mark I suppression pool to dynamic loads during a postulated LOCA.

F. ferring to Figure 27, which schematically shows a BWR Mark I containment system, it can be seen that if a LOCA were ever to occur, air followed by steam would be injected into a toroidal wet wall half filled with water. Displacement of the water by the injected air / steam fluid would result in a downward load followed by an upward load on the. toroidal wet wall. This loading may be followed by oscillatory steam chugging near the downcomers.

Based on tests in a 1/64-scale model, LLL determined that it wculd be possible to represent the full 360" torus with a 90*

segment at 1/5 scale. Additior.al investigations indicate that a reasonable scaling function for the hydrodynamic load due to air venting is:

r 1

P l

P

=

p gl model p gL full scale t

g J

which means this function would have the same value in the model and full scale.

e 4

f Thus, hydrodynamic similarity is maintained when the above condition, as well as some others, is satisfied, in this function. L is the linear scale. There' ore pressures in the model ire reduced in proportion to L whe.1 o of the water is g

maintained Th.e 1/5-scale containment test facility is shown schematically in Figure 28 and the 1/f,4 scale model experiment is shown in figure 29. The principal work to be accomplished during the Spring of 1977 is to complete a comprehensive program of experimental measurements on the initial air transient of a hypothetical LOCA, 1.e., air will be vented into the toroidal wet well. Variations in the dry well overpressure, drywell pressurization rate, down-comer submergence, vent pipe flow asymmetry, anri enthalpy flux are included in the test program. In addition to other measurements, high speed motion pictures of the bubble motion and pool swell will be obtained. LLL will also develop scaling relationships between the 1/f scale model and a full scale torus so that the experimental data con be related to a full-size BWR t'. ark I containn(nt system.

2.3 Fuel Rod Behavior

{

i This research includes basic studies on the constituents of

{

the fuel rod (fuel, gap, and cladding), studies on integral e

fuel rods, and definitions of fuel failure limits and consequences.

l t

All of this information will be incorporated into computer j

codes which can be used to assess the safety margins in fuel rods exposed to hypothetical' reactor accident conditions.

2.3.1 3asic Studies l

In response to the NRC ECCS Acceptance Criteria,62 the basic 1

studies have focused on Zircaloy cladding oxidacion Zircaloy mechanical behavior (including deformation on bursting which could influence the ability to meet the coolable geometry criterion),UO2 pellet properties, gap conductance and decay heat. The last three items have a strong influence on the fuel behavior for such postulated accidents as the LOCA. A useful compilation of fuel rod material properties has recently i

been made, 3 i

2.3.1.1 Zi-caloy 0xidation Conside'able attention his been focused on the rate of oxidation of Zircaloy, in the presence of steam in order to calculate the hydrogen generation, the extent of cladding oxidation, and the heat generation from oxidation of the Ziracloy by steam during a hypothetical LOCA.

__ Figure 30 shows the reported results obtained in the three years of Zircaloy-steam reaction kinetics research. For comparison, the oxygen rate constant obtained frori the Baker-64-66 is also shows.

Just equation From Figure 30 it is evident that the oxidation rate at 1200*C (2192*F) is only one-half % two-thirds the rate obtained in the Baker-Just equat h. This in turn implies less heating and less hydrogen generation than predicted by the Baker-Just equation for Zircaloy steam reactions.

Since oxygen-contaminated Zircaloy may be brittle, it may not withstand either the forces calculated to nccur during a hypothetical LOCA or the themal shock caused by the quenching action of the cooler ECC water. The ability of a partially oxidized fuel rod cladding to withstand these stresses depends on the oxygen content of the unoxidized portion of the cladding.72 The oxygen penetration is usually calculated frori the diffusion coefficients obtained by Hallett et al.73 some years ago. New 74 work obtained by Cathcart and Perkins at the Oak Ridge National Laboratory is shown in Figure 31 along with the results of Mallett et al.,73 Debuigne,75 Schmidt et al.,76 and Dechamps et al.77 The diffusion coefficient of oxygen in beta Zircaloy is shown to be only half of the previous value. The activation energies are comparable. By comparison, the diffusion data for unalloyed zirconium 75 are still lower.

2.3.1.2 Zircaloy Mechanical Properties Information on the mechanical behavior of Zircaloy cladding is of importance in determining the ability of the cladding both to contain the fission products and to avoid appreciable flow blockage. Some recent results78.79 show that when tested under more realistic conditions. Zircaloy exhibits less deformation before failure than previous workeo indicated. Axial restraint, testing in steam and reduction in the internal gas volume will each reduce the clad ballooning.

l The results of Chapman et al.78 at ORNL are shown in Figure 32 in comparison with other existing data.81-83 The ORNL work, based on more realistic modeling of the internal gas volume, t

internal pressure, and external environment, indicates signifi-l cantly less circumferential strain at failure than previous tests conducted in inert atmosphere.

(

Two of the tests (marked A), which were run in argon instead of steam, indicate that a steam atmosphere greatly reduces strain in the high strain regions. Examination of the cladding l

l

.. 1 1

af ter the tests revealed that deform: tion before the onset of plastic instability was comparable for rods heated in steam or argon but that the steam limited the amount of strain to failure af ter onset of plastic 'nstability, suggesting that a surface effect may be governing.

The burst temperature vs. pressure of Chapman et al.78 agrees with the data of Hobson et al.'O shown in Figure 33. The tests showed that the large defarmation is very localized extending only a few rod dianeters at most. The location of the burst is very sensitive to the highest local temperature.

These observations appear to have an important bearing on the ballooning and subsequent flow blockage experienced by nuclear fuel rods; i.e., axial and circumferential temperature distri-butions will tend to concentrate significant ballooning in highly localized hot spots on one side of the rod.

Axial restraint during the deformation also reduces the circum-ferential strain as shown in Figure 34 developed from recent experiments conducted by Kassner et al.79 These biaxal burst tests were conducted in argon on Zircaloy as-received and thus show a higher strain than would be expected from tests run with steam. Kassner et al. have identified superplasticity as a source of the high circumferential strain causing excessive ballooning.

The results discussed above are all obtained on unirradiated Zircaloy cladding. An NRC sponsored program at Battelle Columbus Laboratoriese4 is conducting similar experiments on irradiated rods from commercial reactors to determine the influence of prior irradiation on the cladding response. The results to date obtained on PWR rods indicate that ductility, as determined by axial elongation, is the slowest of the measured properties to recover during a rapid temperature transient. The low ductility gives less ballooning.

2.3.1.3 Pellet Properties Programs are being conducted at ANL,es INEL and ORNL to investigate and model the transient fission product release from irradiated 002 pellets. The programs have the dual purpose of understanding the fission gas contribution to the total pressure inside the fuel rod during an accident and determindag the amount and chemical and physical nature of the fission products transported from a failed fuel rod.

e

~

)

p i

1 I as.,

Huch of the current work on pellet properties (e.g., densi-fication, cracking, restructuring, etc.) is sponsored by EPRI and the nuclear industry.86 A useful compilation of U02 physical and mechanical properties used in fuel code development is given in reference 63.

2,3.1.4 Fuel Rod Thermal Performance _

The stored energy in the fuel rod at the onset sf an accident has a large influence on the magnitude of a thermal transient.

Prediction of the stored energy necessitates knowledge of the pellet / cladding gap conductance. As fission gas is released, the composition of the gas in the gap changes from helium which has a high thermal conductivity, to a mixture containing helium, xenen and krypton, which has a lower thermal conductivity.

Comparisons of FRAP-5287 predictions with experimental fuel centerline temperature datasa 90 obtained with fuel assemblies tested in the Norwegian Halden reactor t various burnups to 15,000 MWD /MtU are shown in Figure 35.

The code contains an empirical gap conductance correlation assuming pellet cracking and relocation resulting from burnup.

The effects of burnup do not become noticeable in the experimental data until about 1,000 mwd /Htu of burnup and become constant after 10,000 mwd /Mtb.

E>periments to measure the gap conductance.of LWR design test rods have Deen performed in the Power Burst facility using a thernal oscillator technique.91-93 2.3.1.5 Decay Heat The NRC ECCS Acceptance Criterf ac2 require the decay heat to be calculated on the basis of an American Nuclear Society standard 9 with a 20~. uncertainty factor added. NRC-sponsored research is aimed at providing more accurate data 95,96 and analysis 97 on the decay heat. Figure 36 compares the latest 97 with the ANS standard. It is evident decay heat evaluation that the more recent (with average standard deviation of 5%)

work shows there will be less heat available to increase the fuel temperature during a hypothetical LOCA.

2.3.2 Fuel Rod Performance under Transients The preceding sections have described programs in which only portions of a fuel rod or of a possible accident sequence have been studied. The results of these programs are being utilized s

t

r.

-- ~ m -m-,

1 TABLE !!

COMPARISON OF MEASURED, ESTIMATED, AND CALCULATED FUEL ROD BEHAVIOR FRAP-T1 Predictions Using These CHF and Post-CHF Correlations

Maximum Estimated W-3 W-3 B&W-2 B&W-2 Measured From Fuel W-3 and and B&W-2 and and During Rod Posttest and Groeneveld Groeneveld and Greeneveld Groeneveld Parameter Test Condition Tong-Young 5.7 5.9 Teng-Young 5.7 5.9 Cladding surface 1,530 2,200 1.875 2,150 2,525 1,850 2,175 2,520 temperature at 25 inches (*F)

Maximum cladding 2,560 2,300-2.450 2.900 2,100 2,350 2,800 surface temperature 1

Fuel centerline 4,155 Less than 3,800 4,250 4,550 3,800 4,275 4,525 2 (elting temperature at 00 m

5,144) 29 inches (*F)

E Maximum fuel Less than 5,600 5,950 6.300 5,200 5,800 6,100.

centerline temperature U0 melting 2

(*F)

Fuel roa internal 1,770 Less tnan 1,630 '

1,630 1,650 1,630 1,630 1,630 l

pressure (psig) 2,080 Axial length change 90 Not possible 220 315 510 l10 230 285 after CitF (mils) j f

The following references provide more information on the indicated correlations:

B&W Ref. 100 Ref.101 W-3 Groeneveld 5.7 and 5.9 - Ref. 102 Tong-Young - Ref. 103 e m=ee

=-a-

- * - =

==n..,,.,

.,-w=

_____..___.s

(

~ -.

~

i

- in the development of the fuel modeling codes FRAP-T9s and i

FRAP-5.87 The verification of the codes is accomplished through comparison with the integral tests conducted in-Facility (PBF)gle or multiple fuel rods.

reactor on sin The l'ower Burst

" at INEL is an important source of this information.

I As of Jaruary 28, 1977 twelve experiments have been conducted utilizing a total of 38 highly instrunented fuel rods.

j Sixteen of the rods were previously irradiated to burnups of 6

i approxin.a tely 16,00014Wd/litU. Three types of experiments have been run in PBF: gap conductance (see Section 2.3.1.5), flow coastdown and power ramp. Additional tests are planned to include reactivity-initiated accidents (RIA) and LOCA simulations.

Approximately one-third of the basic test series has been I

completed.

I t

The flow coastdown or power ramp experiments are intended to show what might happen to a Nel rod in which the heat flux from the rod exceeds the ability of the coolant to remove the heat. This results in the red passing thermal hydraulically into the flim boiling region from nucleate boiling. The ti.:e-dependent fuel rod power, measured cladding surface tempe ature, axial length and fuel centerline te.nperature in response to such an experiment are shown in Figure 37. A compariscn of i

the predicted and measured fuel rod paraneters is give in Table III for several thermal hydraulic correlations. The code predictions using the combination of the W-3301 and Groeneveld 5.7107 correlations were found in this first test to be closest to measured values and nave been used for subsequent pretest predictions. Additional test results may be found in references 104-109.

The results of these tests have been summarized in a recent paper by Quapp and 14cCardell of INEL 30 I

1 "The tests to date have been conducted at rod powers of 500.to 800 W/cm and nave resulted in film boiling periods ranging from a few seconds up to more than ten minutes. Cladding surface temperatures experienced have been as high at 1400*C.

Fuel temperatures have exceeded the 002 melting temperature.

Extensive oxidation of the zircaloy surface has resulted from metal water reaction. Additionally, following cladding collapse onto the fuel pellets in the areas of high temperature, an internal reaction between the U0p and the zircaloy has contributed to additional cladding embrittlement. The combined effects of internal and external cladding attack have resulted in failure of the fuel rods in several of the tests following reactor

27 -

4 1'

I shutdown. The highly embrittled cladding fails a few minutes following shutdown apparently as a result of the small forces i

induced by the flowing coolant."

Quapp and McCardell'io point out the significant PBF result 4

that "none of the tests have failed at power in spite of the presence of molten fuel and severe mechanical interaction.

The nature of the failures in these tests is such that failure propagation or severe fuel coolant interactions are considered unlikely during an abnormal event ir. a power reactor resulting in conditions similar to those of the current test."

f, 2.3.3 Fuel Meltdown Research s

In add' tion to the PBF fuel damage studies. NRC is also sponsoring research on phenomena associated with hypothetical fuel meltdown accidents. A good background review report has been prepared for NRC by Sandia Laboratories.ll! The basis for these studies may be found in the Reac_ tor _ Safety Study _.112 "The only way that poteettally large amounts of radioactivity could be released is by melting the fuel in the reactor core....To melt the fuel requires a failure of the cooling system (such as a failurt: of all the ECCS after a LOCA) or the occurrence of a heat imbalance that would allow the fuel to heat up to its melting point, aboJt 5000*F." Using probabilistic techniques it is estimated that the total probability of melting the core

')

is about one in 20.000 per reactor per year. The Reactor Safety Study observes:

"*t is significant that in some 200 reactor years of commercic' operation of reactors of the type considered in the report there have been no fuel melting accidents."

NRC sponsors several programs to address aspects of the hypothetical core meltdown accident and the release of fissiori products.

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i Responsible laboratory Program Fission Product Release from Oak Ridge National Laboratory LWR Fuel

}

Sandia Laboratories Holten fuel Interactions Natural Convection in Melton Ohio State University j

Pools Sar.dia Laboratories l

Steam Explosion Phenomena f

Transient fuel Response and Argonne National Laboratory Fission Product Release I

Vapor Explosion Triggering Argonne National Laboratory Battelle Columbus Laboratories Firsion Product Transport Analysis Battelle Columbus Laboratories Analysis of the Physical Events Associated with Degraded Reactor Accidents Exemplary is the workil3-Il5 oroemding at Sandia Laboratories to identify the chemical and physical processes which occur when molten core materials contact concrete, specifically the gas production rate and the penetration rate of a melt. To examine the thermal aspect of the decomposition of concrete, Sandia has subjected specirrcns to heat fluxes onione surface of 28-122 W/cm in the Radiant Heat Flux Facility and 123-180 in the 2MW Plasmajet Facility.!!5 Both basalt and W/cm2 limestone aggregates are used yielding concrete with 6 weicht percent (4 w/o H 0, I w/o CO ) and 20 weight percent (

2 51:0,16 w/o C0 )2 volatiles, respectively. Preliminary..

vations 2

and conclusions include:

The principal thermal erosion mechanism is quiescent melting of the concrete matrix (primar;1y silica) with little spallation. Thermal shock effects do not appear to be important.

The rates of surface crcsion (typically I cm/rr.in) appear to be directly dependent on surface heat flux for a given type of concrete (Figure 38).

In tests employing similar heat flux, no signficiant or unexplainabic differences in crosion rates were observed between radiant and plasmajet heating.

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Differential thermal analysis and thermogravimetric analysis

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have been us:d to identify up to four distinct decomposition i

reactions in the 100-1200*C region involving dehydration or i

decarboxylat bn.

At Ohio State,116'll7 an experimental study of the transient I

response of a horizontal fluid layer subjected to a step change in internal energy generation has been conducted to determine the time scale for the development and decay of natural convection. For both cases, the time required for the development of the final steady state is determined by measuring the temperature response of the fluid with a thermccouple i

probe. The time required for the development of the maximum temperature difference in a horizontal layer with internal gereration is correlated with the Rayleigh number by (Figure 39).

Fo = 11.577 Ra-0.213 The time required for the complete decay of the maximum temperature difference of steady convection at a given Rayleigh rumber when internal energy penetration is suddenly stopped is given by Fo = 11.956 Ra-0.215,

2 In both of these equations, Fo is the Fourier number for tne layer, and Ra is the Rayleigh number. The equations will find oeneral applicatica in analyzing the post-accident heat removal

{PAHR)situationinnuclearpowerreactors.

2.4 Computer Code Development j

This research includes improvement of existing codes, development of advanced systems codes and development of componant codes.

The fuel modeling codes are discussed in Section 2.J. and the code verification effort is discussed in Section 2.2.4 and illustrated in Figure 2.

2.4.1

_ Improvement of Existing Codes Top priority is being given to the improvement of the cresent intermediate level system code, RELAP-4.106,119 RELAP-4 exists in two principal versions: a "best estimate" model which is used to describe realistically the accident phenomena and the

" evaluation model" which incorporates the conservative assumptions required for licer. sing analyses. RELAP-4/ Hod 5,119 which contain an improved description of the blowdown phase of a hypothetical PWR LOCA, was released in Summer 1976 and it 1

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predicts Semiscale and LOFT results with reasonably good accuracy as discussed in Section 2.2.4.

RELAP-4/ Mod 6, which extends the RELAP improvements to the reflood phase of a postulated PWR LOCA, will be released to the Argonne Code Center (ACC) by Summer 1977. The BWR reflood analysis capability will be covered by RELAP-4/flod 7 which is to be released to ACC during the Winter 1977/1978.

2.4.2 Advanced System Codes, Because of the complexity of the mathematical description of the two-phase (steam / water) flow and heat transfer and other physical processes uccurring a postulated accident, NRC is examining alternative approaches to ar.tve at a satisfactory description of the overall system behavior during an accident.

These different approaches include using alternate model forou'.ations or solution procedures or both.

20 Brookhaven National Laboratory is develcping the TH0R code which is an advanced ccmputer code for predicting the accident-induced, one-dimensional thermal hydraulic transients in LWRs including (i) the capability to account for thermodynamic non-equilibrium betwecn phases. (2) unequal phase v'elocities, (3) axial variations in power generation, (4) initial steady state conditions, and (5) components. This code is intended to serve as the next generation licensing code. By Fall 1976, BNL completed the TH0R system self 'nitalization, break flow models consistent with those errployed for intarior regions, and numerical coupling of tte distributed and the lumped partmeter regions. During 1576 a parallel path effort was chosen to insure that the next ecneration licensing code will be timely and will sa';sfy licerising requirements. Hence, an initial effort is being sponsored at INEL on the development of the RELAP-5 code to complement BNL*: effort on THOR.

LASL is developing the TF.AC advanced system codel21 to describe LOCA phenomena. TRAC will differ from THOR in two aspects:

(1) TRAC will utilize the most copropriate techniques for solving the basic equations of non-equilibrium two-phase fluid dynamics, dependent on the degree of steam / water coupling in various system components, and (2) TRAC will errploy multi-dimensional descriptions of these reactov components which require it (e.g., plenums, downcomer, and the :cee). In FY 1976 LASL corrpleted the TRAC (1) calculationa 6 strategy (data transfer, and storage organized and coded in modular fashion),

(2) Nplicit solution technique for multi-dimensional views of the whole reactor vesset, (3) drift flux non-equilibrium

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31 -

module with wall heat transfer, and (4) two-dimensional transient simulation of ECC penetration and bypass in a PWR downcomer, including condensation effects, e

Work is also undcrway on improved containment analysis techniques.

This work will be correlated with the BWR pressure suppression testing underway at LLL. The first version of the BEACON containment system codel38 will be released in 1978.

2.4.3 Component Codes A number of component codes are under development through NRC programs. The purpose of the component code development is to model in greater detail the behavior of the various single cornponents of a reactor. In this way, the simplifications made in developing a systems ccde can be assessed in terms of their degree of accuracy. The following component codes are under develcpment:

1 SCORE 22 - this is a " core" component code which is used for the multi-dimensional description of core flow under the assumption of a homogeneous, thermal equilibrium two-phase mixture. The code will be documented during 1977 and released to the Argonne Code Center (ACC).

l COBRA 23 - this is another " core" component cc,de which can also model a significant assembly or even a single (hot) channel. The latest version is COBRA-DF which includes the drift flux and full non-equilibrium in axial and lateral flow components. During 1977 this code will be applied so'.ely to modeling the multi-dimensional aspects of UHI behavior in the entire PWR vessel.

A comparison of the results from the COBRA-4 and SCORE codes is shown in Figure 40. Both codes give similar core flow patte:rns at s 300 msec after blowdown. The importance of this j

agreement is two-fold:

1.

An agreement between two completely independent codes indicates a sign of validity of the code predictions even without data verifications.

2.

At early blowdown, the flow velocity in the void region at the middle of the core is almost lateral, because of the difference in subcoolings between neighboring channels.

This lateral velocity strongly affects the delay of transient CHF and the post-CHF heat transfer. A good code can explain phenomena which occur so fast in a

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, hypothetical LOCA that thet can hardly be measured in an experiment.

- This advanced component code, which is a derivative K-TIF_124 of the KACHINA code,125 is being applied to detailed, multi-dimensional modeling of PWR downcomer flow during the ECC injection period.

K-FIX126 - This advanced component code, which is another derivative of the KACHINA code, is being used to explore the numerical simulation of the details of the transport processes across steam / water interfaces. The first version has been released to ACC. K-FIK will also figure in INEL's development of the BEACON advanced contaiment code.

Both K-FIX and K-TIF will be continually updated. These codes represent NRC's frontier of knowledge in the numerical simulation of multi-dimensional, transient two-phase flow and they use the most sophisticated models available to NRC. The advanced systems code TRAC, profits from these detailed analyses since it retains modeling of only those effects which the detailed component codes show to be important.

50LA-FLX121 - This is a hydro-elastic code which permits a more realistic description of the blowdown-induced loads on PWR vessel internals.

SOLA-DF, SOLA-PL00_P_127-129 - These two codes are experimental predecessors of TRAC. They were released to ACC in 1977.

2.5 Reactor Operational Safety _

NRC is expanding its research into reactor operational safety matters, specifically fire protection, aging evaluation and human engineering. This research is an integral part of the overall NRC reactor safety research program.

2.5.1 Fire Protection and Aging Evaluation At present the fire protection program is focused on the evaluation of the effectiveness of cable tray separation in preventing the spread of a cable fire to redundant trays.

l Scieening tests were conducted to determine which of the currently utilized generic cable types are most susceptible to the spread of a cable tray fire, and two full-scale cable tray tests completed. The results of these tests are being evaluated and a report will be issued shortly. Cable insulation, cable jacket and coating materials are continually being developed l

by manufacturers and new products introduced. The performance n

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5 of these materials wnen used in Class 1 equipment and, systems as well as the change in the properties of those materials Performance evaluation of the with age need to be evaluated.

materials in use today is primarily by separate effects testing oriented towards single component evaluation and where aging 3

is considered, only accelerated aging methods are being utilized.

The future NRC effort will be aimed at verifying performance of safety class materials and equipment in systems typically found 1.a power plants and under the conditions that they will Tests encounter during their respective design basis events.

are being designed now for the verification of all aspects of NRC Regulatory Guide 1.75130 including conduits, fire breaks, fire barriers and penetration fire stops. Future plans include program in the following areas:

Evaluation of the effectiveness of coating materials.

1.

2.

Evaluation of aged materials.

l Development of small scale cable fire tests to predict 3.

total system performance.

Evaluation of the vulnerability of safety class equipment 4.

to non-electrically initiated fires.

5.

Evaluation of fire and smoke detection systems.

Evaluation of the vulnerability to fire of Class IF 6.

equipment other than cable.

The qualification test program is to provide technical information for improved oging qualification testing of safety class Areas to be addressed include improvement of aging equipment.

models, nuclear source definition, synergisms, the performance indicators to be monitored during qualification testing, l

failure definition, allowable thermal and nuclear radiation j

flux gradients, test sample preparation and quality control, l

mounting and corrections to test samples, chemical and steam t

l flow rates, impingement effects and vibration. Research programs are currently underway at Sandia Laboratories.

f 2.5.2 Human Engineering The risk assessment analysis as presented in Report WASH-1400 :2 identified the role that human interaction and intervention 1

play in the unavailability of safety systems ar.d components in This conclusion was based in part on a nuclear power plants.

preliminary human factors analysis of a typical pWR nuclear 1

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_ t12 of power plant control room performed earlier by Swain i

Sandia. He observed three general categories of human factors problems. (1) human engineering design in deficiencies in control rooms (2) shortcomings in training, and (3) poor format for written operating instructions.

Shortcomings in human engineering were also listed in the major rccommendations in the reportl31 to the American Physical i~

Society by the study group on light water reactor safety as follows: " Human engineering of reactor controls, which might significantly reduce the chance of operator errors, should be improved. We also encourage the automation of more control functions and increased operator training with simulators, especielly in the accident-simulation mode."

Safety research programs by the Nuclear Regulatory Commission and EPRI directed at providing answers for these concerns are in progress. The NRC safety research program goal's for Human -

Engineering are to (1) reduce the potential for human error by identifying human factors improvements in operator training, the design of control consoles and optimum use of automation of controls for safety systems; (2) develop an actuarial human factors data base to enable more accurate risk assessment and (3) provide the technology base for developing guides and standards for controls and control. room designs.

Efforts to identify human errors and develop a statistical data base for human reliability have been initiatea by NRC through the Licensee Event Reports (LER), whose format has been modified to categorize human errors. A contract has also been writt n with Sandia Laboratories for the preparation of a handbook of human error rates in nuclear power plants. Improved operator training in order to reduce human errors is being studied by both NRC and EPRI. The first phase of an EPRI contract with the General Physics Corporation for a "Perfonnance Measurement System for Training Simulatory" is scheduled to be finished in May 1977. A " Human Factors Review of Nuclear Power Plant Control Room Design" by Lockheed Missile & Space Systems Division for EPRI was completed in December,1976.

"An Analysis of Control Room Displays and Operator Performance" by Acrospace Corporation under an NRC contract was finished in February 1977. Industrial guides and standards are being drafted for safcty related operator actions (ANSI No. 660),

for the design of display and control facilities (IEEE P. 566) and for the design of control rooms (IEEE P. 567).

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3.0 CONCLUSION

S The water reactor safety research efforts in the primary system integrity area have (1) validated the conservatism of vessels; (2) pressure vessel design by testing on nine intermediate the reactor developed the remedy for curing radiation damage by using thermal annealing techniques; and (3) improved the nondestructive inspection techr.iques used for flaw detection by digitizing the signals emanating from acoustic emission and ultrasonic testing. Future research in this area is directed i

toward providing additional verification of NRC analytical j

techniques and further improvements in NRC inspection techniques.

NRC is expanding its reactor operational safety research, specifically concentrating on fire protection and human engineering research. The objective of this research is to further reduce the already low risk associated with the opera.tional aspects of commercial reactors.

The experimental and analytical research programs which address the fuel behavior and thermal-hydraulic behavior of reactor plants have confirmed, either quantitatively or qualitatively, a number of conservatisms used in current licensing assumptions.

As examples, the decay heat, the Zircaloy oxidation rate and the ECC bypass rate have Deen found to be lower than assumed while the post-CHF heat transfer rate appears to be higher than assumed. Based on the experimental data from this research we have developed more realistic correlations to predict these phenomena. In addition, this research has also greatly improved the understanding of the hypothetical LOCA through comparisons of RELAP-4 computer code predictions with the test data from such facilities as Semiscale and LOFT. To further improve both the physical understanding of LOCA behavior and predic-tive capabilities, NRC is developing more sophisticated analytical fonnulations of the computer codes with the goal of having operational versions in December 1977.

In general, the water reactor safety research program has greatly expanded the data base in such areas as fracture mechanics, transient boiling heat transfer and two-phase flow analyses. Special credit is due the individual investigators for their scientific and engineering achievements in these areas.

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-- I, 4.0 EXPECTED FUTURE RESULTS Based upon our current projections, a number of interesting results are expected within the next two to three years. Some highlights of these expected results are described in the following sections (listed according to the five principal areas of NRC research in the field of water reactor safety research).

4.1 Primary System Integrity In the area of fracture mechanics, NRC researchers expect to confirm the integrity of repair welds, and establish a correlation between low ductile shelf energy and fracture toughness for irradiated, pressure vessel weld metal. Moreover, a data bank of crack arrest data from irradiated steel specimens will be developed, and the methodology for analyzing cyclic crack growth will be validated.

In the area of operational effects, the failure criteria and safety margins for steam generator tube burst and collapse will be established, the elastic-plastic analysis methodology for irradiated steels will be validated, and the met'iodology for analysis of large steam line treaks will be developed and validation testing begun.

As part of the nondestructive examination (NDE) pro 5 ram, the feasibility of alternate NOE methods will be established, the means for in-process acoustic emission monitoring of welding will be validated for the ASME code, a mechanics-acoustic emission model for LWR defect significance will be verified, and an improved ultrasonic testing data processing / evaluation technique will be validated against the ASME code evaluation requirements.

4.2 LOCA Heat Transfer and Hydraulics Over the next three years the PWR blowdown heat transfer (BDHT) tests on the first two electrically heated 49-rod bundles will be completed. These tests will provide invaluable transient CHF and post-CHF heat transfer information or. full-length rods. In the PWR reflood area, FLECHT tests on a 15x15 electrically heated bundle with a skewed power profile at low reflood rates (forced convection) will be completed. These tests will be followed by a series of FLECHf systems effects tests in which the influence of the steam generator will be further quantified.

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I LOCA thermalhydraulic tests relating specifically to ECC bypass will be run in a 2/15 scale facility. The low pressure g

tests are scheduled to be completed in 1978 with high pressure tests following in 1980.

BWR BDHT tests will be completed on an 8x8 electrically heated bundle and test data will be available on BWR blowdown /ECC interaction effects, and countercurrent flo, limited phenomena.

Semiscale will complete a series of systems tests to provide confirmatory information on the upper head injection (alternate ECCS) concept. Following this work Semiscale will be modified to include two complete loops and a 12-ft core.

By the end of 1978 it is planned that the remaining two nonnuclear LOCA/ECCS tests will be completed in the LOFT facility and preparations will be well underway for the first nuclear test to be conducted in 1979.

4.3 Fuel Rod Behavior The correlations describing Zircaloy oxidation kinetics have been established and attention will now be focused on Zircaloy properties leading to such accomplishments as the establishment of an embrittlement criteria, completion of single rod and initial mulitrod bu st tests with the issuance of a preliminary correlation on rod-on-rod interactions, completion of in-pile stress rupture tests on two lots of Zircaloy, and completion of in-pile untaxial low cycle fatigue tests.

^

A pellet-clad interaction (PCI) correlation will be developed from the Halden irradiation tes'.s.

The contact conductance experiments will be completed and a gap conductance model verified. The physical and mechanical property correlations will be confirmed for plutonium recycle fuel. The 2350 and 23SPu decay heat work will be completed and the decay heat calculation for plutonium recycle fuel will be confirmed.

A sizable number of power-cooling mismatch, flow blockage, LOCA, irradiation effects, and RIA tests will be completed in PBF and/or ESSOR thereby providing confirmatory systems-type data on fuel clusters.

In the area of fuel codes, both FRAp-T (Mod-5) and FRAP-S (Mod-4) will be verified using PBF and Halden data. This will be followed by improved RIA, PCI and multiple-rod models for FRAP. The fission product transport models will be integrated to describe a hypothetical core meltdown.

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1 Several interesting results will be forthcoming from the fuel meltdown and fission product transport research. The studies of heat transfer in hemispherical fluid layers will be completed.

The upper limits will be established for the temperature and pressure needed to initiate explosigns between liquids. The vapor explosions experiments will be completed and integral experiments will be conducted to verify the models of molten core-concrete interactions.

4.4 Computer Code Development The schedules for the release of the various computer colr>

were outlined in Section 2.4 and will not be repeated here.

The overall goal is to have a complete, verified LOCA advanced systems code by 1981.

4.5 Reactor Operational Safety HRC is vigorously pursuing a fire protection research program designed to assess the requirements for cable tray separation and to evaluate the IEEE-383 cable fire test. This program will include tests of cable systems in enclosed areas, evaluation of cable aging characteristics and development of fire protection guidelines for cables, barriers, stops and detectors.

The guidel;nes for qualification testing of equipment for LOW conditions will be verified. Included in this effort is a series of long term aging tests which will be conducted to permit development of an improved aging model.

rriteria for safety related operator actions in nuclear control rcoms will be developed and the initial evaluation of advanced concepts for the display and automatic control of safety functions will be completed. Assessments in the area of noise diagnostics to be completed in this period include (1) the confirmation of alternate noise diagnostic techniques for detecting general core r

l internals vibration, (2) the assessment and experimental verification of the sensitivity of loose parts mcnitoring system techniques, and (3) the initial assessment of automated safety system on-line surveillance and diagnostic techniques.

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5.0 REFERENCES

U.S. Nuclear Regulatory) Commission," Rear. tor Safety Resea.ch Program, 2

[1]

NUREG-75/058 (June 1975.

[2]

Puzak, P.

P.,

Pellini, W.

S., Standard Method for NRL Drop-Weight Test, Naval Research Laboratory Rep. NRL 5831 (August 21,1962).

[3]

Pellini, W.

S., et al., Review of Concepts and Status of Procedures for Fracture-Safe _ Design of Complex Welded Structures Involving Metals of Low to Ultra-High Streagth Levels, Naval Research Labora-i tory Rep. NRL 6300 (June 19 3 ).

Loss, F.

J., Pellini, W. S., Jou ling of Fracture Mechanics and C

[4]

Transition Temperature Approaches to Fracture-Safe Design, Naval Research Laboratory Rep. HRL 6913 (April 14, 1969).

[5]

Irwin, G.

R., " Fracture," Handbuch der Physik, Vol. 6, 551-590 (1958).

[6]

Fracture Toughness Testing and Its Applications, ASTM STP 381, Chicago (1964).

[7]

Swed10w, J. L. (ed.), The Surface Crack: Physical Problems and Com-putatational__ Solutions, ASME, New York (1972).

[8]

Rice, J.

R., "A Path Independent Integral and the Approximate Analysis of Strain Concentration by Notches and Cracks," Journal of Applied Mechanics, Trans. ASME, Vol. 90 Series E No. 2, 379-386 (June 1968).

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A., "Recent Developments to JIC Testing,"

Scientific Paper 76-lE7-JINTF-P3, Westinghouse Research Labs.,

Pittsburgh, Pa (May 5, 1976).

[10]

Witt, F. J., Mager, A Procedure for Determining Bounding Values of FracturR Toughness KIC at any Temperatu_r_e,, ORNL-TM-3894 (October e

1972).

l

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Burdekin, F. M., Dawes, H. G., " Practical Use of Lineat Elastic and i

Yielding Fracture Mechanics with Particular Reference to Pressure Vessels," Practical Application of Fracture Mechanics to Prr.ssure Vessel Technology, Institute of Mechanical Engineers, London, 28-37 (1971).

[12]

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Monahan, J. M., Walker, T.

J., PM-2A Reactor Vessel Test - Descrip-tion of Testing and Failure Conditions WAPD-TM-640, Bettis Atomic Power Laboratory (January 1967).

[14]

Derby, R. W., et al., Test of 6-Inch Thick Pressure Vessels, Series 1: Intermediate Test Vessels. Vols. I and 2. ORhL-4895, (February 1974).

[15]

Merkle, J. G., Whitman, G. D., Bryan, R.

H., An Evaluation of the l

HSST Program Intermediate Pressure Vessel Tests in Terms of Light-Water-Reactor iressure Vessel Safety ORNL-TM-5090, (November 1975).

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l

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. Chicago, Illinois.

Tune 19/6 (to be published).

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Server, W. L., Shekherd, J. W., Wulleart, R. A., " Fracture Toughness Data for Ferritic Nuclear Pressure Vessel Materials," EPRI NP-il9, t

Effects Technology Incorporated, Santa Barbara, California (April 1976).

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Hawthorne, J.

R., " Demonstration of Improved Radiation Embrittlement Resistance of A5338 Steel through Control of Selected Residual Elements," ASTM STP 484, Am. Soc. Test. Mats.,96-127,(1970).

[20]

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E., Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steelg, IAEA, Technical Report Series 163 (1975).

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P., " Stress Corrosion Cracking of Type 304 Stainless Steel in High Purity Heavy Water," 2nd. International Congress on l

Metallic Corrosien, New York (March 1963).

[25]

Berry, W. E., Corrosion in Nuclear Applications, John Wiley & Sons, Inc., New York (1971).

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Latanision, R. M., Staehle, R. W., " Stress Corrosion Cracking of Iron-Nickel-Chromium Alloys," Proceedings, International Conference on Fundamental Aspects of Stress Corrosion Cracking, National Association of Corrosion Engineers, Houston, Texas, Ohio State University (September 1967).

=.-,-----.;----.-..-

n.-

[

_ _ _ _ _ l i

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i

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Fairchild, R. C., " Improved Ultra-i sonic Nondestructive Testing of Pressure Vessels," Yearly Progress Report August 1,1974 - July 31,1975. U.S. Nuclear Regulatory Commission Report NUREG-0007-1, (January 1976).

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E., Zebroski, E.

L., Hutton, P.

H., " Detection of Stress Ccrrosion Cracking in Stainless Steel Piping by Acoustic Emission,' (to be published in Processing of Third Conference on Periodic Ir:spection of Pressurized Components, London, England, September 1176.)

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~

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[37]

Crowley, C.J. and Block, J.A., ECC Delivery Study - Experimental Results and Discussion, Creare Technical Note TN-217 (October 1975)

[38]

Block, J.A. et. al., Analysis of ECC Delivery, Creare TN-231 (April 1976).

l l

l

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_ z _ m _. _. _

-e -

. t i

Hsu, Y. Y. and Beckner. W., " Blowdown CHF," under preparation for

[39]

publication.

Tong L.S., " Heat Transfer in Water-cooled Nuclear Reactors," Nuc.

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Henry, R. E., and Leung, J.

C., " Spontaneously Generated Heat flux Crises" AN5 Transaction, Vol. 23, p. 283. Toronto Meeting (1976).

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[42]

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Iloeje, O. C., Rchsenow, W. M. and Griffith, P., "Three-Step Model ASME of Dispersed Flow Heat Transfer (Fost-CHF Vertical Flow)."

[43]

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Tong, L. S. and Young, J.

D., "A Phenomenological Transition and Film Bolling Heat Transfer Correlation." Paper B3l9, Vol. 4, Proc.

[44]

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[47]

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t 1 h l

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~

~~-

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E., Hann, C. R., Lanning, D.

D., Panisko, F.

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~

the Thermal Behavior of an Oxide Fuel Rod, USNRC Report BNWL-1898 (November 1975).

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T., MacDonald, P. E. and Quapp, W.

J., USNRC-0 ECD Halden Project Fuel Behavior Test Frogram - Experimental Data Report for Test Assemblies IFA-226 and IFA-239 USNRC Report ANCR-1270 (December 1975).

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1

\\

i

[103]

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{

[107]

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D., and Romers, N. C., SOLA - A Numerical Solution Algorithm for Transient Fluid Flows, LA-5852 (1975).

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[129]

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No.1 (Sumer 1975).

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FIGURE 30. ARRHENIUS PLOT OF Ti!E TOTAL CXYCEl: RATE Kawasaki Ref. 69 CONSTANT FOR THE STEN 1: ZIRCALOY-4 REACT!0?!

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Text

SUMMARY

SUMMARY

SUMMARY

1.0 INTRODUCTION

3.0 CONCLUSION

5.0 REFERENCES

SUMMARY

SUMMARY

SUMMARY

3.0 CONCLUSION

5.0 REFERENCES

Dearten,

Dearien,

.-

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