ML19209D230
| ML19209D230 | |
| Person / Time | |
|---|---|
| Issue date: | 06/30/1979 |
| From: | Bennett G NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| NUREG-0581, NUREG-581, NUDOCS 7910220136 | |
| Download: ML19209D230 (81) | |
Text
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NU REG-0581
SUMMARY
OF NRC LWR SAFETY RESEARCH PROGRAMS ON FUEL BEHAVIOR, METALLURGY / MATERIALS AND OPERATIONAL SAFETY me@S%j='
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IJOTICE This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agenc/ thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liabtlity or responsibility for any third party's use, or the results of such use, of any information, apparatus product or crocess disclosed in this report, or represents that its um o'f such third party would not infringe privately owned rights.
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NU REG-0581 R1,R3,RS,RP,RV
SUMMARY
OF NRC LWR SAFETY RESEARCH PROGRAMS ON FUEL BEHAVIOR, METALLURGY / MATERIALS AND OPERATIONAL SAFETY G. L. Bennett Manuscript Completed: June 1979 Date Published: September 1979 Division of Reactor Safety Research Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, D.C. 20555 l169 107
SUMMARY
OF NRC LWR SAFETY RESEARCH PROGRAMS ON FUEL BEHAVICR, METALLURGY / MATERIALS, AND OPERATIONAL SAFETY Gary L. Bennett, Chief Research Support Branch U.S. Nuclear Regulatory Commission Executive Summary This paper summarizes the results of NRC-sponsored research into fuel behavior, metallurgy and materials, and operational safety. The NRC light water reactor safety research program is part of the NRC regulatory program for ensuring the safety of nuclear power plants which are owned and operated by electric utility organizations.
The NRC LWR safety research program focuses principally c,.1 the major subsystems or f unctions of a commercial LWR: thermalhydraulics, fuel behavior, metallurgy and materials, and operational safety. Computer codes are developed to analyze these various subsystems or f unctions. The thermalhydraulics research and associated computer codes are not discussed in this paper.
FUEL BEHAVICR The purpose of the NRC fuel behavior research program is to provide a detailed understanding of the response of r.uclear fuel assemblies to postulated off-normal or accident conditions. This information is then used to develop physical models which ar* incorporated into fuel analysis codes. The fuel codes are tested against the results of integrated in pile te ;s.
The understanding of the release and transport of fission products from damaged fuel rods and the study of the behavior of molten fuel are included in this research program.
The Power Burst Facility (PBF) tests at the Idaho National Engineering Laboratory are sponsored by NRC. To date, nuclear LOCA blowdown tests, power-cooling mismatch tests and reactivity-initiated accident tests have been run in PBF.
Since the cladding is the first arrier to the release of fission products, NRC has funded a number of studies of this important component which have shown that the cladding will contribute less beat, less hydrogen and more strength thaq predicted by conservatise LCCA licensing models.
Fission product release measurements made at ORNL indicate that the amount of cesium and iodine escaping f rem a defected PWR fuel rod during a LOCA with successful ECCS operations will be one to two orders of magnitude less than the gap release afsumptio s for these species used in licensing evaluations.
In the area of fuel code development, the fourth version of the transient code, FRAP-T4, has been released by INEL and undergone independent code assessment. The new steady-state code, rRAPCON-1 which is based on models from FRAP-5 at INEL and GAPCON-THERMAL at PNL, has been completed and is now undergoing independent assessment.
METALLURGY AND MATERIALS The objective of the hRC research into metallurgy and materials is to provide independent confirmation of the safe design of reactor vessels and piping and, if required, to establish ways for reducing the failure probabi-lities. The NRC research activites in this acea are divided into the following groups: (1) fracture mechanics; (2) irradiation embrittlement, stress corrosion, and crack growth; and (3) nondestructise examination.
Some of the principal NRC confirmatory tests on f racture mechanics have been conducted on both small and intermediate scale pressure vessels at ORNL. These tests have shown that for flaws less than one-half of the wall thickness in depth, a pressure of nearly three times the design value must be applied to initiate rapid fracture.
Since operational ef fects such as irradiation embrittlement, stress corrosion and crack growth may degrade the load bearing capability of structural components, hRC sponsors a number of research programs to quantify these effects. The results to date show that irradiation embrittlement can ba " annealed" out with heat treating.
In the area of nondestructive examination, NRC-sponsored research has upgraded the ultrasonic testing techni-ques which form the basis for in-service inspections of the primary system.
I169 108 1
11 0FEWATICNAL SAFETY NRC sponsors a broad category of research termed " reactor operational safety" which is research aimed at providing direct assistance to NRC officials concerned with the operational and operational-safsty aspects of nuclear power plants. The topics currently addressed include qualification testing evaluation, fire protection, human factors, and noise diagnostics.
The NRC qualification-testing evaluation program is foc4 sed on obtaining the data needed to answer certain questions about the testing of saf ety-class equipment to assess the performJnce during and af ter postulated accident conditions. The specific questions considered are given in this paper. A new facility ;s being developed at Sandia Laboratories for more sophisticated testing.
The NRC fire protection research program emphasizes the collection of confirmatory data needed in support of current design standards and regulatory guides for fire protection and control in LWR nuclear power plants.
Both small-scale and full-scale tests have been run at Sandia laboratories and Underwriters Laboratories to study fire-retardant coatings, shields, sprinklers, and fire prcpagation.
The NRC human factors research program is concerned with assessing the role of human errors in reactor opera-tional safety. It includes specific studies in support of human error investigations and the development of associated training programs, the study of safety related operator actions, and a continuing review of the application of ergonomics in the design of nuclear p9wer plants.
The NRC noise-diagnostics research program at ORNL has supported licensing activites by the use of noise diagnos-tics techniques in independent assessments of core-barrel motion in operating PWRs and in-core instrument tube vit' rations in cperating BWRs of the BWR-4 type.
Recently, ORNL researchers have been analyzing the power oscillations in the Fort St. Vrain power plant as well as assisting in the assessment of the status of Three Mile Island Unit 2.
CONCLUSION In general, the NRC LWR safety-research program has greatly expanded the data base on such areas as f racture mechanics, fuel behavior and operational s+tety.
This information is being used 5y the NRC licensing staff in support of decisions on reactor safety.
e 1169 109
SUMMARY
OF NRC LWR 5AFETY RESEARCH PROGRAMS ON FUEL BEHAVIOR, METALLURGY / MATERIALS, AND OPERATIONAL SAFETY Gary L. Bennett, Chief Research Support Branch U.S. Nuclear Regulatory Commission INTRODUCTION The NRC's confirmatory safety research program on light water reactors (LWRs) is structured to provide additional and/or independent information to gain confidence that the margins of safety identified in the licensing review are well defined and quantified.' The principal areas of NRC research in the field of LWR safety are structured approximately along the classic lines of defense-in-depth, which in this paper will include:
F uel Rehavior - Fuel rod behavior in postulated ar..idents and associated f ailure limits; Metalluegy/ Materials - Safety d* sign and protection of the integrity of the reactor pressure vessel and piping; Operational Safety - Operational safety aspects of nuclear power plant operation.
This paper complements a companion paper2 on the NRC LWR safety research programs related to thermal-hydraulics and computer Code development, in this summary paper only a brief overview of some very involved subjects Can be given. A more detailed presentation is given in Reference 1.
FUEL BEHAVICR RESEARCH The NRC research into the behavior of LWR fuel assemblies is directed at providing a detailed understanding of the response of these fuel assemblies to abnormal or accident conditions. The importance of this research is clear since the first barrier to the release of radioactivity is the fuel cladding. The understanding of fuel behavior ultimately is expressed in terms cf confirmed analytical models whica are incorporated into publicly available computer codes.
NRC sponsors research covering the fuel, gap conductance, the clad 1ing, integral f uel rods and rod-to-rod interactions. These research studies encompass both cut of-reactor and in-reactor experiment % the latter principally carried out in the Power Burst f acility (PBF) at the Idaho National Engineering Laboratory and in the Halden experimental boiling water reactor in Norway. These research studies spin a range of environments from normal cperation to abnormal or accident conditions.
The loss-of-coolant accident (LOCA) initiated by the rupture of a large primary-coolant pipe has been selected as a design-basis accident h.r evaluating many of the safety features of LWR power plants. Other postulated accident sequences that af f ect fuel-element t,ehavior include the " power-cooling mismatch" (PCM) in which the boundary of the reactor-coolant system remains intact but there is an imbalance between the heat being generated by the fuel and the heat removed by the coolant. A PCM would result from a loss of coolant or an everpower transient.
A reactivity-initiated accident (RI A) could result from such causes as control rod ejection or an anticipated transient without scram (ATWS).
The condition of the fuel element at the initiation of the accident could affect the course of the accident.
The principal initial parameters that must be known for the analysis of a transient are ti.c stored heat" and decay heat in the fuel, the gas pressure within the cladding, the extent of contact between the fuel and the cladding, and p-ior cladding strains. These parameters are interrelated and depend un a number of properties, such as therma! conductivity, thermal expansion, cracking and restructuring of the f uel, the initial fuel-to-cladding gap width, fission gas release, and cladding creep. It is therefoie nece nary to understand the fuel design and to evaluate, either by analysis or experiment, the above-mentioned parameters. Trends that affect these parameters include prepressurization of the fuel rods, improved stabilization of pellet density during irradiation, and changes in fuel-rod diateters instituted by all reactor vendors.
"The results of a recent investigation to determine the current state of the art of fuel temperature, gap conductance, and stored energy calculations may be found in Stored Energy Calculation: The State of The Art by M. E. Cunningham, et al., FNL-2581 (May 1978).
I169 110
2 The emergency core cooling systems (ECCS) are the principal safety features installed to maintain the integrity and long-term coolability of the fuel during a LOCA. The ECCS Acceptance Criteria are intended to ensure the 3
effectiveness of the ECC5 if it should ever be needed in maintaining the structural integrity of the cladding.
Two of the criteria supply direct guidance for planning the f uel behavior program:
" Peak Claddin emperature. The calculated maximum fuel element cladding temperature shall not exceed 77JD"T 7 T477
" Maximum Cladding Oxidation. The calculated total oxidation of the cladding shP 1 nowhere exceed 0.17 times the total cladding thickness before oxidation.
The following sections give a short summary of some of the principal achievements to date in the NRC fuel behav'r,r research program.
Decay Heat The acceptance criteria for LWR emergency core cooling systems require that the decay heat applicable to a 3
LOCA evaluation be calculated using the 1973 pivposed American Nuclear Society Standard ANS 5.1.4 The uncertainty in those data was judged to approach 15%, especially at cooling times V less than 100 seconds. The NRC has prescribed that a conservative value of 1.2 times the 1973 value be used in evaluating the ef f ectiveness of ECCS. More recent calculations 5 and measurements
'7 of decay heat in irradiated uranium-235 have demonstrated 6
that the 1973 standard is itself conservative during the first seconds after shutdown. F urther.noie, the uncertainties in the new data are nominally less than 5% at short cooling times and decrease as tle cooling time increases. Figure I shows a comparison of these measured data and the summation calculation. On the basis of these data, a new standard has been developed and is being proposed to the American Nuclear Society Standards Steering Committee and then to the ANSI Board of Standards Review.
Zircaloy Omidetton The oxidation of lircaloy in steam is an important phenomenon in acc;uent analysis sinco (1) hydre en is generated by the reaction between Zirconium and the steam, (2) the heat of reaction must be removed to preve... Overheating the cladding, and (3) the oxygen consumed forms two brittle layers that reduce the wall thickness capable of carrying teasi k st resses. The two ta ittle layers are zirconium oxide and an oxygen-stabilized alpha phase.
The oxygen aiso dissolves in the remaining beta phase and causes it to be embrittled. For Zircaloy oxidation in steam, th> Baker-Just rate-constant equation" 's current?y used f or conservative evaluation of postulated accidents. Objections have been raised to the conservatism of this equation. It does not agree with experi-rental data in the temperature range of interest. The oxidation M Zircaloy has therefore ueen the subject of several investigations. From data reported by Cathcart.S (see Figure 2) the rate constant, 62/2, at 1477A (2200 F) is only 58% of that of the Baker-Just equation, and thus only 167, of the oxidation predicted by the Baker-Just equation is actually observed. As a result, calculated peak cladding temperatures during a given postulated LOCA are estimated to be approximately 56K (100*F ) lower with the new rate equation than with the more conservative Baker-Just equation.80 of the rate of oxygen diffusion in beta phase lircaloy has indicated that the rate is A new determination 88 approximately half that previously reported.12 (See Figure 3)
Mechanical Properties of Zircaloy Containing 0xygen The mechanical properties of Zircaloy have been determined as functions of on gen distribution and content, 83 strain rate, biaxial stressing, microstructure, teatt.re, and temperature over the range between 423 and 1700K (300 and 2600 F).
The strength and ductility of Zircaloy cladding at any temperature are strongly dependent on such factors and are important in producing and controlling cladding deformat.icn during postulated LOCA and PCM events. The effects of quenching stresses on the properties of oxidized Zircaloy tubing have been determined.
Failure " maps" for f racture of the cladding by thermal shock were developed
- relative to the maximum oxidation 8
temperature and various time-dependent oxidation parameters, e.g., equivalent-cladding reacted to ferm Zr0,
2 fractional thickness of transformed beta layer, fractional saturation of the beta phase by oxygen, and thickness of beta phase with less than a specified critical oxygen content. The principal results are (1) if the cladjing is cooled rapidly (about 100 K/s) through the beta-to-alpha phase transformation, the thermal shock f ailure boundary corresponds to about 20% of the wall thickness in equivalent oxidation for oxidation temperatures greater than about 1650 K, (2) if the cladding is cooled slowly through the phase transformation, the thermal shock failure boundary corresponds to 28% equivalent wall thickness oxidized to Zr0, and (3) the best correla-2 tion of thermal shock failure boundary with parameters related to the degree of oxidation of the cladding we that of thickness of the beta phase layer having less than 0.9 and 1.0 wt% oxygen for slow and f ast-cooled cladding respectively. If the beta phase laver having less than this oxygen level was 0.1 mm in thickness, or more, the cladding did not f ail irrespective of will thickness, oxidation temperature, and total oxygen content.
In situ pendulum load impact tests at room temperature showed that the thermal shock boundary corre..ponded to approximately 0.03 J impact energy absorbed. As illustrate. in Figures 4 and 5, data were plotted as time of oxidaticn versus reciprocal temperature and evaluated as to (1) failed on thermal shock ouenching, (2) failed at 0.03 J impact, (3) survived 0.03 J but f ailed 0.3 J impact, and (4) survived 0.3 J impact lond. These results allow a quantitive statement of energy absorbed by f racture, and a quantitative f ailure map to be drawn with any desired degree of conservatism. A finite element model has been developed for crack growth in oxidized (ircaloy during thermal shock conditions, using mechanical properties measured for homogeneous specimens of I169 117
235U + n,%, CEC AY PP'ER ENDF/8-IV CALCUL ATION G,g iO'3 sec
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References:
Baker-Just:ANL-6548 WPI:EPRI 249-1 ORNL:0RNL/TM-5248 Kawasaki:JAERI-M 6181 Leistikow: private communication Heidrick: unpublished FIGURE 2.
ARRHENIUS PLOT OF THE TOTAL OXYGEN RATE manuscript CONSTANT FOR THE STEAM:
ZIRCALOY-4 REACTION c
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DIFFUSIVITY OF OXYGEN IN BETA-ZIRCALOY fl4
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REFERENCE:
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6 ISOTHERMAL OXIDATION TEMPERATURE (*C) 1400 1300 1200 l 100 1000 l
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IMPACT CAPALILITY OF OXIDIZED ZIRCALOY-4 CLADDING' (REF: NUREG/CR-0201) 1169 115~,
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IMPACT CAPABILITY OF OXIDIZED ZIRCALOY-4 CLADDING (REF: NUREG/CR-0201)
8 varied oxygen contents. The data suggest that the present ECCS embrittlement criteria are conservative both to failure by thermal shock and failure by impact loads likely tc be encountered in disassembly of the core.
Mechanical Properties of Irradiated Zircaloy A satisfactory data base has been developed for the important mechanical properties of fuel cladding, e.g.,
yield stress, ultimate stress, uniform and total elongations, reduction in ares and thickness at fracture, burst stresses, burst strains, stress rupture, ccllapse pressure, creep rates (both internal and external pressurization), and low-cycle f atigue over a range of environmental conditions. With respect to irradiation e'fects, information obtained from in-reactor tests and from post-irradiation tests is available for most of the properties of interest. For irradiated materials the scatter is generally greater (see, for example, Fig. 6), and since tests are more expensive, there are fewer data points for statistical evaluation. Until recently, most information was obtained at or below the operating temperatures of BWRs.
Irradiation results in a reduction in ductility, a reduction in impact strength, and an increase in strength at and below normal reactor operating temperatures. The yield strength of annealed Zircaloy-2 doubles and remains essentially constant above a neutron fluence of approximately 5 x 10 0 n/cm - The yield strength and ultimate 2
2 strength are effectively identical.15 The consensus is that strength changes tend to saturate between 1021 and 2
19 and 102o n/cm2 (>l MeV).
- The 10 2 n/cm2 (>l MeV) but that the etfect on ductility taturates between 3 x 10 strain at which plastic instability sets in is reduced with increasing irradiation.
A significant body of data on the as-irradiated mechanical properties and on the kinetics of irradiation-damage annealing has been obta.wdl7 ca one lot of Zircaloy tubing removed from spent commercial PWR fuel with a burnup greater than 30,0C0 mwd /MT U0. The prnperties examined were uniaxial tensile strengths and elongatiors 2
in tubing at temperatures f rom 300 to 975K (80 to 1290 F) and burst tests at 645K (700 F).
Specimens were tested in the as received condition, af ter isothermal annealing at temperatures of up to 975K (1290 F) and af ter annealing by trans knt heating to temperatures of up to 1275K (1830 F) at rates from 0.5 to 25K/sec.
Similar tests are now underway on a second lot of cladding from spent commerical fuel with a burnup of less than 20,000 mwd /MT 00 Burst tests have been conducted during transient heating to temperatures of up to about 1275K (1830 4 ) 2at several heating rates. The kinetics of irradiation-damage annealing appear to vary with the evaluatfon method. Yield, tensile, and burst strengths can be fully recovered at seme temperatures, while elongation decreases significantly below that observed in the as-received condition, with all tests conducted at 645K (700 F).
Thus there is a " strain aging" or " aging" phenomenon that af fects elongation but not strength properties. More data are needed for cladding at lower burnup, so that the " saturation exposure" for the several properties can be determined.
Transient heating burst tests have shown that the properties at but st temperatures of 980K (1300 F) and higher are essentially the same in both in adiated and unirradiated Zircaloy tut !ag (See Figure 6).
Cladding Deformation During Burst Testing One of the key items in analyzing a postulated LOCA is knowing the condition of the core, i.e., what is the deformation and extent of flow blockage of the coolant channels of a fuel assembly. Experiments are being performed with electrically heated rods to give flattened temperature gradients comparable to those in PWRs and BWRs, with internal pressures from 100 to 1800 psi and heating rates of up to 28 C/sec (50 F/sec). Single rods and clusters of 16 and 64 rods with typical PWR grid spacings have been or will be studied. The rods are approximately 2 meters (6.5 feet) long with a heated length of 0.92 meter (3 feet), and the grid spacings are about 0.61 meter (2 feet). The data from the single-roo testsia (over 40 have been completed-see Table 1) will be compared with the data from the cluster tests, +;ith the goal of having a preliminary correlation of rod-to rod interactions, scaling factors, flow blockage, heating rate, initial and burst pressures, and burst strain; by 1980 (See Fig. 7).
The correlation should allow the prediction of multirod performance f rom single-rod tests and should greatly decrease the cost of evaluating various cluster ccnfigurations and cladding modification:
To confirm the burst test results from the electrically heated 0.9-m (3-ft) length rod. experiments are planned on full length rods (3.7 m or 12 f t. ) heated by nuclear power in the NRU test reactor in Chalk River, Ontario.
These tests will also be of interest in comparison to the full length electrically heated bundles in the REBEKA test series being conducted by KfK in the Federal Republic of Germany. In addition to deterC ning the burst behavior of fuel rods, a series of experiments on the thermal-hydraulic reflood behavior is a so a part of the proposed NRU program. The experiments will use a bunale of 32 rods of commercial length and enrichment and will provide well-characterized data on the thermal-hydraulic and deformation behavier in a nuclear environment representative of the heatings and reflood of a LOCA in an LWR.
Gap ConJuctance The thermal behavior of an LWR fuel rod is complex. Mechanisms postulated to influence the thermal behavior of an LWR fuel rod include:
o Changes in the dimensions of the fuel-to-cladding gap from pellet cracking and pellet relocation, ft.el densificatiun and swelling, thermal expansion, and cladding creepdown.
e Changes in the fuel thermal conductivity from pellet cracking (nonradial cracks) and restructuring.
e Changes in the composition of the gas in the gat or in fual cracks from impurity gas release, fission-gas release, and fill gas absorption.
I169 117
1200 1
H. B. Robinson D
W 28C/sec restrained axially 6 28 C/sec f
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O 20 40 60 80 100 12 0 14 0 160 Burst Hoop Stress,MN/m2 FIGURE 6.
TRANSIENT BURST-TEST DATA FOR ARCHIVE AND SPENT-FUEL Q
CLADDING INCLUDING ORNL CURVE FOR UNIRRADIATED TUBING (REF:
" MECHANICAL PROPERTIES OF SPENT FUEL CLADDING" BY L. M. LOWRY, ET AL, SIXTH NRC WRSR INFORMATION MEETING,1978, NRC PDR)
TABLE lA. TEST CONDITIONS AND RESULTS FOR SINGLE R00 BURST TESTS IN STEAM *
.s Initial conditions Burst conditions Fuel -
FPSgg Maximum Burst Tube heated Time to b
Test simulator volume Temperature Pressure pressure Pressure Temperature TC No.
TC position relative strain length change burst No.
No. -
(cm )
(*C)
(kPa)
(kPa)
(kPa)
(*C) to tube burst
(%)
(%)
(sec) 3 C
PS-1 PROTO 2 -
26.3 351 6,450 7,040 6,360 893 92 At burst, 20' around 18 NA 20.0 PS-3 PROTO 1 30.0 334 6,520 6,860 5,580 873 213 27 cm below, 70* around 29 NA 20.0 PS-4 PROTO 1 29.3 343 6,440 6,780 5,860 871 201 2.5 cm below, 70* around 21 NA 21.0 PS-5 PROTO 1 29.5 343 6,410 6,760 5,720 882 201 At burst, 60* around 26 NA 21.5 b-8 2628005 30.6 349 6,470 6,810 6,000 843 213 14 cm below, 90* around 20 NA 21.5 PS-9 2828005 35.6 346 6,480 6,890 5,650 866 201 4.5 cm above, 60* around 25 NA 23.0 PS-10 2828005 35.9 352 6,440 6,830 6,000 901 83 6 cm above, 0* around 20
-0.69 11.2 PS-12 2828006 41.0 340 6,520 6,900 6,140 898 49 At burst, 30* around 18 NA 21.75 PS-14 2828006 41.5 337 6.450 6,830 5,820 883 83 6.4 cm above, 30* around 25
-0.69 22.65 PS-15 2828006 46.9 352 6,490 6,780 6.160 885 84 18.4 cm above, 70* around 17
-0.61 20.95 PS-17 2828005 37.5 340 13.270 13,880 12,130 778 86 40.3 cm above,140* around 25
-0.43 16.1 PS-18 2828007 45.0 350 800 862 772 1171 93 11.4 cm above, 60* around 24 0.39 42.0 PS-19 2828005 27.5 348 2,590 2,820 2,590 959 88 45.2 cm above, 50* around 28 0.0 27.45 SR-1 2828013A 46.6 347 850 910 800 1166 91 38.9 cm above, 30* around 25 1.04 31.6 SR-2 2828017 48.5 344 1,130 1,220 1,010 1082 91 17.1 cm above, 20* around 44 0.61 25.7 "R-3 2828027 46.1 346 1,770 1,900 1,720 1011 90 5.5 cm above. 0* around 43 0.48 22.4 SR-4 2828026 44.5 337 4,400 4,700 4,480 921 93 14.6 cm above, 40* around 17 0.18 20.65 SR-5 2828028/
45.9 345 10,120 10.483 9,520 810 91 8.3 cm,100* around 26
-1.13 18.0 SR-7 2828014A 45.9 338 15.110 15,530 14,440 736 86 20.3 cm below,130* around 20
-0.39 15.55 o
SR-8 2828010 49.3 336 1,420 1,520 1,230 1020 90 15.2 cm below, 70* around 43 0.61 25.15 SR-13 3838016 47.3 325 1,31 0 1,430 1,070 1079 89 9 cm below,10* around 79 0.65 24.65 SR-15 2828005 38.8 342 20,350 21,280 19,150 714 91 At burst,10* around 14 0.35 14.5 SR-17 2828010 44.9 344 1,31 0 1,410 1,060 1049 90 21 cm below, 40* around 53 0.78 25.25 SR-19 2828031 35.2 335 19,970 20,830 19.040 668 86 27.5 c:a oelow, 90* around 16 0.09 14.6 SR-20 2828031 33.4 332 1.290 1,410 1,060 1049 93 30 cm above, 110* around 55 0.61 25.1 SR-21 2828005 36.7 340 1,31 0 1,430 1,120 1023 83 32.3 cm below, 140* around 48 0.56 24.5 SR-22 2828031 33.8 332 1,130 1,230 890 1081 90 At burst, 4* around 50 0.69 27.2 SR-23 38380G5 35.4 336 1,120 1,230 960 1077 86 25.7 cm belcw, 130* around 35 0.74 25.7 SR-24 2828031 35.5 332 1,200 1,300 990 1057 91 34.4 cm above,120* around 67 0.87 26.9 SR-25 2828036 34.6 345 1,130 1,240 960 1092 83 2.5 cm below,135* around 78 0.87 26.5 SR-26 2828031 43.1 340 1,000 1,060 830 1130 87 20.4 cm below,100* around 34 1.13 29.9 SR-27 2828036 42.3 340 1,130 1,190 920 1084 86 At burst, 0* around 41 0.87.
26.9 SR-28 2828021A 48.0 335 8,930 9,400 8,400 835 89 6.2 cm below, 45* around 27
-0.69 19.3 SR-29 282028A 42.0 340 8,680 9.050 8.040 843 94 26 cm above, 90* around 27
-0.87 20.0
\\\\
fuel pin simulator volume measured at room temperature; includes pressure transducer and connecting tube.
a Maximum measured by any thermocouple at time of burst; thermocouple number and location indicating burst are listed.
CNot measured.
- "Significant Results from Single-Rod and Multirod Burst Tests in Steam With Transient Heatinn." by R. H. Chapman, Fifth NRC Water Reactor Safety Research Information Meeting, Nov. 7-10, 1977, available in NRC PDR.
TABLE IB. 1EST CONDITIO% AND RESULTS FOR SPECIAL SIN 1E ROD BURST TESTS Initial conditions Burst conditfons Burst Tube heated Time to Manfmum b
Test simulator volume Temperature Pressure pressure Pressure Temperature TC No.
TC positic, relative strain lensth change burst Fuel FPSgag No.
No.
(cmi)
(*C)
(kPa)
(kPa)
(6Pa)
(*C) to tube burst (1)
(1)
(sec)
" * ^
Tests conducted in an argon environment SR-9 2828005 38.2 335 6480 6890 6260 880 92 18 cm above.170* around 18
-0.52 22.1 SR-11 2828010 45.9 330 1400 1530 1270 1015 91 5 cm below. 0* sround 98 0.18 24.4 SR-14 2928005 38.4 334 1740 1900 1740 1004 93 14 cm above. 0* around 24 0.18 22.8 Large-volume tests conducted in a steam environmant 54-16 2828005 159.2 345 6500 6580 6420 880 93 22 cm above,135' around 15
-0.61 22.15 SR-18 2828005 154.6 344 2590 2630 2590 968 89 4 cm above.15' around 22 0.30 22.95
' Fuel pin simulator volume measured at room temperature; included pressure transducer and connecting tube.
Masimum measured by any therwocouple at time of burst; thermocoupie number and locatfon indicating burst temperature are listed.
RESULTS OF CREEP RUPTURE AND LOW HEATING RATE BURST TESTS OF FUEL PIN SIMULATORS WITH !NTERNAL HEATER 5 TABtE IC.
Burst condf tfons Initial conditions Fuel FPS gas
- Mastmum Heatfng Control Burst Tube heated Time to Test simulator volume Temperature Pressure pressure rate temperature Pressure Temperature" strain length change burst b
Mo.
No.
(cm3)
(*C)
(kPa)
(kPa)
(*C/sec)
(*C)
(kPa)
(*C)
(1)
(s)
(sec)
Creep rupture tests in steam _
SR-33 2828036 51.0 370 6285 6515 13 162 5690 762 23.4
-1.04 103 d
SR-34 2828031 51.6 336 6260 6540 12 762 5820 766 31.6
-0.87 49' d
SR-35 2828031 51.5 350 4830 5050 13 761 4470 775 29.0
-0.26 250' d
SR-36 2828036 51.2 330 4805 5040 li 761 4555 821 28.8
-0.52 162' d
Transient burst tests in steam 54-31 2828031 50.3 303 14410 14965 28 13560 760 23.1
-1.13 17.4 SR-38 2828036 51.1 340 14660 15265 29 13775 770 20.0
-0.69 15.5 SR-41 2828031 50.0 340 10510 10915 9
9765 757 27.4
-1.04 46.9 SR 42 2828036 49.6 314 10495 10900 10 9465 761 28.4
-1.39 47.1
$R-4 3 2828031 48.7 340 8465 8800 4
7620 773 29.0
-1,39 89.1 5R-44 2828036 49.7 338 7935 8250 5
7310 777 30.0
-0.78 82.5
' Fuel pin simulator volume measured at room temperature; includes pressure transducer and connecting tube.
beeperature maintained constant at control thermcouple during creep time by feedba-i control on power fr.put.
NanimA measured by any thermocouple at time of burst.
C From initial temperature of creep temperature.
' Hole time for creep at quest-steady-state temperature level.
- Taken from *Ef fect of Creep Time and Heating Rate on Deformation of Zircoloy-4 Tubes Tested in Steam with Internal Heatups" by R. H. Chap N
USNRC Repo.t MUREG-CR-0343 0
1200
+ PS TESTS IN STEAM 1100 X SR TESTS IN STEAM t SR TESTS IN ARGON 6 LARGE GAS VOLUME IN STEAM
_1000 W_
E x
T = 20 342P -
~ T = BURST TEMPERATURE (OC)
E' P = BURST PRESSURE (kPa) l
~
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a x x a) 800 h
- X 700 i
X i
600
~
0 2000 4000 6000 8000 10.000 12.000 14.000 16.000 18.000 20.000 BURST PRESSURE (kPa)
FIGURE 7.
BURST TEMPERATURE AS A FUNCTION OF BURST PRESSURE.
ru (REF: ORNL/NUREG/TM-135)
13 With pressed and sintered pellets there is usually an appreciable resistance to heat transfer between the pellet surface and the cladding. The interfacial resistance may be the result of a gas-filled gap or uranium dioxide in actual contact with the cladding. Data on fuel centerline temperature and rod internal pressure tend to support the contention that the thermal response of an LWR fuel rod is strongly influenced by stochas-tic pellet cracking and pellet fragment relocation mechanisms. As fuel burnup progresses, pellet cracking and relocation, pellet swel!ing, thermal expansion, and cladding creepdown combine to close the gap. The rate of gap closure has been shown to depend on such cperating variables as the rate of power increase, number of power cycles, and power level.
Several in-reactor experiments to deduce values for gap conductance have been made, and a number of investiga-tors have attempted to infer gap conductance from the examination of fuel rods that were irradiated for other purposes. Gap conductance is not measured directly but is derived from measurements of fuel and/or cladding temperatures. Most of the reliable experiments utilized small (1 200 pm) diametral gaps. There is very little well-characterized data for thermal reactor fuel with larger diametral gaps, especially in the 33-to 50-kW/m (10- to 15-kW/ft) operating power range. Experiments involving instrumented fuel assemblies (IFAs) and spon-sored by hRC have been reported in References 19 and 20.*
These experiments have been run in the Norwegian Halden Boiling Water Reactor under the technical management of EG&G Idaho and Battelle-Pacific Northwest Laboratory (PNL).
Reference 19 presents test data from the EG&G Idaho-Halden experiment IFA-429. The IFA 429 is an 18-rod test assembly designed to study fission gas release and fill gas absorption in prepressurized (2.58-MPa helium)
PWR-type fuel rods. This data report presents assembly power history and individual fuel rod power, temperature, pressure, and burnup data frcm June 1975 through June 1978. Reported fuel-rod heat ratings cover a range from 17 to 30 kW/m with measured fuel centerline temperatures of 1375 to 1475 K (2015 to 2195 F) for the highest power. Measured f uel-rod pressures showed no appreciable change during the period covered.
Results from FNL-Halden experiment IFA-431 were reported in Reference 20 and analyzed in subsequent reports.
For one of the rods in this experiment, the average gap conductance uncertainty over the range of measurement was i 19L The uncertainty in the gap-conductance measurement changed as a f unction of linear heat rating.
(See Fig. 8).
The experiment includes two fill gases (pure xenon and pure helium) and three pellet-to-cladding gaps. The absolute error in determining the temperature drop across the gap is less than 100 C for any of the combinations of gap diameter and fill gas used.
Power Burst Facility The Power Eurst Facility (PBF) is a water-cooled and water-moderated reactor, contained in an open-top steel vessel ar.d is used for in-reactor tests of fuel rods. (See Figure 9).
It is operated for the U.S. Department of Energy and the NRC by EG&G Idaho, Inc.
lae reactor core is designed for both steady-state and pulsed-mode operation. One to twenty-five test fuel elements with i. active length no greater than 91 cm are fitted into a test train together with the necessary test instrumentation. The assembled test train is then fitted into a pressurizable thick wall a metal cylinder 15.5 cm in diameter (the inpile tube or IPT). The IPT is mounted vertically and concentric to the vertical axis of the reactor core and the containing vessel.
The IPT has six to eight openings, permit +.ing the use of up to 100 pairs of instrumentation test leads. Typical test instrumentation includes inlet and/or exit flow meters (up to five per test); absolute-and differential-pressure transducers for monitoring fluid and fuel-element plenum pressures; surf ace and internal thermocoup'es for monitoring fuel, cladding, plenum, Ed coolant temperatures; ultrasonic thermometers; linear variable dif ferential transformers (deflection indicators); radiation-flux monitor wires and foils; and self powered neutron detectors. Suitable instrumentation, signal-conditioning equipment, and data-accumulation and data-reduction equipment and services are avai M le.
The PBF test program includes the following areds: (1) power-cooling mismatch, with both unirradiated and pre-irradiated fuel rods (16 tests); (2) LOCA with both unirradiated and preirradiated fuel rods (10 tests);
(3) flow blockage, with previously unirradiated fuel rods (3 tests); (4) reactivity-initiated accident, with both fresh and pre-irradiated fuel rods (12 tests); and (5) gap conductance (stored energy--7 tests).
Table 2 is a summary of the PBF tests conducted to date. The PBF test series may be described as follows:
Power-Cooling-Mismatch Tests. These tests study the critical-heat-flux (CHF) and post-CHF behavior of single rods (four at a time) and nine-rod clusters under a variety of power and cooling conditions, in which CHF is achieved either by increasing the fuel-rod power at a steady coolant flow, or by decreasing v
the coolant flow at a steady fuel-rod power, or by simultaneously decreasing the coolant flew and incress-ing the fuel rod power. (To date, only the specific combinations of final fuel-rod power level and final flow rate appear to be important.) These tests also study the effects of irradiation and burnup on the thermal mechanical properties of fuel-rod components (particularly claddings).
Coolant flow, stored energy, and test-termination temperatures and post-CHF cladding deformation are among the test variables measured.
=A report has recently been published summarizing data on mixed oxide fuel rods irradiated in the Halden reactor:
Fuel Rod Temperature and Pressure Response in Halden Reactor Experiment IFA-226 by P. E. MacDonald et al.,
NUREG/CR-0267 (August 1978).
1169 122
14
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GAP CONDUCTANCE UNCERTAINTY B0UNDS (REF:
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R BURST FACILITY (pggy
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TABLE 2 PROGRAMMATIC TESTS PERFORMED IN THE POWER BURST FACILITY
~
- of Total Type of Test Tests Rods Test Conditions Results Power Cooling Mismatch 8
17 DNB at PWR pressure & temperature Basis for failure prediction at high cladding temperatures; brittle failures predictable from ex-reactor criteria; almost complete oxidation at power to get failure PCM-Cluster 1
9 3 x 3 rod array Similar to single rod tests; 2 rods failed at power or during cooldown m
Loss of Coolant Accident 5
4 PWR blowdown from 67 kW/m Good blowdown heat transfer (20 kW/ft) caused low cladding temperatures LOFT-LOCA 3
12 Simulate LOFT L2 series (from No cladding collapse; tempera-53 kW/m) tures lower than predicted Gap Conductance 4
16 BWR rods with various gases & gaps Steady state & transient data for code development PCM - Irradiation Effects 6
21 PWR rods from Saxton Reactor Similar behavior to unirradiated tests; some fission-gas-induced swelling Reactivity Initiated Accident 6
12 Energy input up to 300 cal /gm Failure thresholds near prediction h
V
17 The objective ci the power-cooling mismatch test, PCM-Is was to simulate the worst possible PCM incident leading to fuel rod f ailure at power with molten fuel-coolant interaction (MFCI) when f ailure occurred. The critical heat flux was surpassed during a power ramp from 40 kW/m to 78.7 kW/m. The feel rod operated approximately 8 minutes af ter the onset of film boiling (DNB; tefore it failed. Af ter failure, the fuel rod power was main-tained at 78.7 kW/m for approximately 7 minutes, at which time the re=ctor was scrammed. Following the scram the central portion uf the fuel rod disintegrated. No violent MFCI (vapor explosion) occurred as a result of the PCM-1 fuel rod failure even though over 50% of the UO fuel at the hot-spt was molten at the time of rod 2
failure. The oxidation of the cladding at the time of failure was considerably in excess of present licensing criteria for rod failure.
The first 9 rod cluster power cooling-mismatch test, pCM-5, was conducted in May 1978. The test consisted of 5 PWR-type fuel rods held in a 3x3 cluster with spacing typical of a 15x15 PWR lattice. The objective of PCM-5 was to study the film boiling behavior of a central rod when surrounded by other rods also in film boiling.
Cladding temperatures for the center rod during the transient were in the p phase Zircaloy range (T > 1245 K).
The cornee rod, which was held in DNB for about 11 miautes, failed about 5.5 minutes after DN8 occurrence. The center rod entered film boiling 4 minutes after the corner rod and was subjected to film boiling fo approxi-mately 1 minutes. Seven of the fuel rods achieved stabilized film boiling, and four of the nine rods failed at power or during the rapid post-test shutdown. Two of tne rods did not experience DNB (see Figure ld).
Recently a review 22 has been completed of the available literature on the operation of nuclear fuel rods under film boiling or dryout conditions. The following material is taken directly from that review.
Test fuel rods f rom both pressurized water reactors and boiling water reactors were subjected in-reactor to a combined total of more than 170 power ramp cycles and more than 250 flow coastdown cycles in order to force the rods into film boiling or dryout. Only 13 of the 667 tests rods failed and these 13 failed in a manner consist-ent with an ex-reactor time-at-temperature brittle-ductile boundary curve. This relationship shows that long times at very high temperatures are required for failure.
Figure 11 is a plnt of the reactor test data discussed in sections 3 and 4 and summarized in Tables 1, 2 and 3 of Reference 22.
The cladding peak equivalent temperatures are plotted against the logarithm of the time-at-equivalent-temperature. The unfailed rods are shown as open circles, squares and triangles and the rods which were oxidized severely enough to fail are shown as filled circles, squares and triangles.
23 The studies of Chung, Garde and Kassner at Argonne National Laboratory have shown that the severe shock of quenching the hot oxidized 0.6 mm thick Zircaloy cladding is roughly equivalent to a 0.03 Joule impact at 300K (See Figure 5).
These studies have also shown that if oxidized cladding can withstand a 0.3 Joule impact at 300K, the cladJing is still tough enough to witratand both operating stresses when at working temperature and the mechanical stresses of handling when at room terperature. This is true despite the fact that the ANL test rods have effectively the same hydrogen pickup from the hydrogen released by the zirconiua-steam reaction as would be picked up after clad rupture in an accident where reduced coolant pressure occurs, e.g., a LOCA. This level of hydrcgen pickup is higher than would t,e expected for most abnormal operating transient accidents.
Therefore, the thermal shock curve and the 0.3 Joule at 300K impact curve represent brittle-ductile boundary curves between the elevated isothermal exposure temperatures required to produce brittle behavior of cladding and those isothermal esposure temperatures which will not cause brittle behavior either during or af ter a given exposure time In Figure 11 the boundary for brittle vs. ductile behavior for shicked, oxidized Zircaloy from the out of pile isothermal ti7e at-temperature studies of Chung, et al.23 is plotted against the reactor test data for f ailed and unfailed rods.
It is apparent from Figure 11 that, for any selected exposure time, the required equivalent temperature for in-reactor embrittlement of both f resh and pre-irradiated Zircaloy cladding is ef fectively the same as the required isothermal temperature for out-of pile embrittlement of that same cladding. Therefore, the equivalent temperature corcept for identifying probable emDrittlement Dy the non-isothermal accident time-temperature profile seems to be valid.
Based on the gcod agreement of the out-of pile data on unirradiated cladding and the in pile data on both fresh and pre-irradiated cladding it would appear that brittle-ductile boundary curves generated out-of pile can then be used with reasonable confidence to predict cladding oxidation embrittlement and failure which might be caused by reac'.or upset and accident conditions particularly where good models exist for predicting peak cladding temperatures a7d times-at-temperature.
Both the in aile and out-of pile test data show that cladding integrity is generally maintained despite exposure to very severe accident environments.
Projected eqaivalent times-at-temperatures (see Figure 12) f6r most BWR and PWR reactor transients ir, commercial reactors are f ar below the values of time-at-temperature required for brittle failure by oxidation which were observed in the reported tests. Therefore, it would appear that fuel rod cladding should not fail by oxidation embrittlement either during or af ter most of the PCM related DNB or dryout events identified to date.
LOCA Tests.. These tests will study fuel rod behavior, e.g., cladding deformation ar.J oxidation of single-rod (four at a time) assemblies under blowdown conditions. Parameters to be varied include irradiation i
history and cold internal pressures. Sixteen rod clusters will be tested under heatup conditions. Results will be correlated with those of out-of reactor tests.
1169 126
Fig. 10 Rods in Film Boiling During Test PCM-5
~~
0.480-m 0.580-m od od f =2 ( )l f=7 T=0 Rod 2 T = 2.0 0.780-m 0.680-m od Rod s;
T=4 T=7 Rod 4 H
od T=
od F
e
= = = =
1978,NRCPDR)
19 Comparison of in-Reactor Post-DNB Survival / Failure Data for Zircaloy Cladding with FBRB/ANL Zircaloy Ductile-Brittle Boundary Curve 2000K g
g A g y
y NN A A g
O N
1800K N
O N
N 6
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m N
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A PBF/PWR
$ CR/BWR Ductile: (No Rod Failure)
O 6 PBF/PWR 800K O cR/BWR Q SGHWR/BWR O GETR/BWR O Halden/BWR 1
10 100 1000 10,000 100,000 Time After DNB (Seconds)
FIGURE 11 1169 128'"
(
REFERENCE:
~
7 1800K 1600K Ductile Brittle 2 1400K
%,7
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h 1000K g
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FIGURE 12.
COMPARIS0N OF PREDICTED P,WR AND PWR UPSET CONDITION PEAK CLADDING TEMPERATURE AND TIME AFTER DNB WITH NRC FUEL BEHAVIOR RESEARCH BRANCH /
y ARGONNE NATIONAL LABORATDRY ZIRCALOY DUCTILE-BRITTLE BOUNDARY CURVE m
(REFERENCE : NUREG-0562)
21 Fuel behavior LOCA testing in the PBF was started in January 1978 when the world's first nuclear blowdown tests from PWR initial conditions, LOC-IIA LOC-llB and LOC-11C, were conducted sequentially with four separately shrouded PWR 15x15 design fuel rods. h Calculations of coolant behavior aenerally agreed well with the measured coolant behavior, but the calculated cladding surface temperatures were slightly greater than measured (see Figure 13). None of the four test fuel rods failed. Two of the fuel rods were essentially unpressurized and the other two fuel ods were prepressurized at ambient temperature to about 2.42 and 4.81 MPa.
The cladding of the unpressurized rods collapsed very slightly and the cladding of the pressurized rods ballooned very slightly.
The cladding deformation observed during the LOC-ll tests is very well calculated with available high tempera-ture Zircaluy plastic deformation models.
Flow-Blockage Tests. These tests will study fuel-rod behavior, e.g., cladding temperatures and geometric profiles of multiple-rod assemblies (25 rods) under flow blockages of 80 to 98%.
Reactivity-Initiated Accident Tests. These tests will Study the behavior of irradiated and unirradiated fuel rods under rod-drop and rod ejection conditions. Independent rod tests, cluster tests, and model development / evaluation tests will be performed. The effects of irradiation, cluster size, coolant flow, and initial power level will be studied.
A series of RIA tests was begun ir. PBF to determine the behavior of the fuel rods unde.apid power burst conditions that could be caused, for example, by a control rod ejection from the core or a power reactor. In a power burst test the PBF reactor is given a sudden increase in reactivity by the ejection of fast moving control rods from the core, which causes a power burst to be initiated. The first RIA test ever conducted at power reactor temperature, pressure and flow conditior.s, RIA-ST, was a scoping test performed to (1) identify the energy deposition failure threshold for BWR hot-startup conditions, (2) evaluate calorimetry techniques for RIA transient tests, and (3) determine if sizable pressure pulses would result from fuel failure in a water-filled system. The RIA-ST test was comprised of five single rod '.ests.
The first three te;ts addressed the determina-tion of the failure threshold and the tv'luation of calorinetry techniques, and the iinal two te n s were performed at high energy depositions to determine p. assure pulse mognitudes.
Prior to performance of the first RIA test, forty power burst tests were performed to determine the dynamic characteristics of the PBF core for use in RIA tests and to quality the core down to a burst period of 1.6 milliseconds with an associated peak power of 92 GW.
Figure 14 compares the calculated and measured cladding surface temperatures for one of the RIA tests. Figure 15 shows the influences of fuel burnup on failure during an RIA.
Cap Conductance (Stored Energy) Tests. These tests study the gap conductance and stored energy of irra-diated and unirradiated rods. Parameters measured include irradiation history, gap size, fill gas pressure, and pellet densities. Power oscillation (transfer function) and integral k dt gap-conductance measurement methods are being compared.2 In Reactor Tests at Other Facilities The Power Burst Facility is particularly useful for the in react r testing of LWR fuels under abnormal operating conditions. Other reactor safety test facilities also provide useful data. Accordingly, comple-mentary fuel testing pruirams are performed or planned at a number of other facilities, including:
The LOFT Facility at the Idaho National Engineering Laboratory Nuclear Safety Research Reactor (NSRR) in Japan The FR-2 reactor in the Federal Republic of Germany The Halden experimental boiling water reactor in Norway The ESSOR/ SARA test loop in Italy The PHEBUS reactor in France The NRU reactor in Canada The BR-2 reactor in Belgium Fuel Behavior Computer Codes Fuel behavior codes must analyze the thermal, mechanical, and internal gas response of fuel-rod components with the goal of predicting rod condition and integrity. Modeling of thermal behavior during normal and accident conditions must include the surface heat transfer, heat transfer across the fuel-to-cladding gap, the ther.al conductivity of fuel and cladding, the power generation distribution in the fuel, and the solution of the conduction equation. These aspects of the thermal calculations are listed approximately in order of their importance.
Modeling the mechanical response of fuel rods involve; consideration of fuel-cladding mechanical interactions (FCMI); cladding creep, ballooning, and failure; fuel thermal expansion, swelling, aensification, and creep.
The phenomena that are important in steady-state operation (creep, swelling, etc.) significantly af fect behavior during transients. The transient codes must therefore in come manner consider these pne< omena, either by direct calculation or by linkage to a steady-state code. These response phenomena are, of course, coupled to one another as well as to the thermal behavior factors. One very important parameter that is difficult to calculate because of this strong coupling is the fuel-to cladding gap width.
Modeling of the internal gas response is important for determining the loading that it applies to the cladding and for determining heat transfer across the fuel-to-cladding gap. The key modeling areas associated with these effects are axial gas flow, fission gas release, plenum gas temperature, and voids and vo.id temperature.
1169 130
Fuel Rod Cladding Surface Temperature l'
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PBF-LOCA TEST LOC-11C
(
REFERENCE:
Cladding Surface Temperature p
Rod 2-1 0.46 m - 180
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1
\\
l
\\s a
s's
=
o> 1000 -
O
]
mt S
j
-- Calculated g
Measured m
O O
4 J
C
-5 0
5 10 15 20 25 30 Time (s)
FIGURE 14.
COMPARISON OF MEASUREMENT WITH PREDICTION FOR REACTIVITY INITIATED ACCIDENT TEST RIA 1-2 CONDUCTED IN THE POWER BURST FACILITY
(
REFERENCE:
%)
~
400 u
I I
I o
3 FBF Rods MAPI Rods GEX Rods GEP Rods S
k Rod Failed 9 Rod Failed E Rod Failed A Rod Failed O Rod oid not Faii o Rod oid not Faii o Rod oid not Faii a Rod oid not Faii 8 300 E
2 l l7-g[.[. [ m. mm - - - w - ---- -((
(([-[-
E 7[
ailure Threshold -- Unirradiated Rods Q
l Q
~
200
~
g GH) 3 E
.g M
N 100 g
ix 3
0 I
I I
O 10,000 20,000 30,000 40,000 Fuel Burnup (mwd /t)
FIGURE 15.
INFLUENCE OF FUEL BURNUP DURING RIA TESTS a
m e
(
REFERENCE:
NUREG/CR-0269) t.r4
25 The hRC Fuel Behavior Research Branch is sponsoring the development of two fuel behavior codes FRAP-S (Fuel Rod Analysis Program - Steady State)2s and FRAP-T (Fuel Rod Analysis Program - Transient).2c Recently NRC decided to combine FRAP-S with the code GAPCON-THERMAL-327 to prnduce the combined steady state code FRAPCON. 28 The steady-state code was developed for use as a normal-operation analysis tool and as the generator of burnup-dependent initial conditions required for the FRAP-T (transient) code. FRAP-5/FRAPCON seeks to model all of the important phenomena irvolved in nonaccident situations during the life of LWR fuel rtis.
It iteratively calculates the interrelated effects of fuel and cladding temperature, rod internal pressure fuel and cladding elastic and plastic deformation, release of fission product gases, fuel swelling, cladding,.sth resulting from irradiation, cladding corrosion, and crud deposition, all as a function of time and specific power.
FRAP-53 included a n a ber of features such as failure prediction models, sophisticated fuel cladding interaction models, and a package of material properties.* Frequent and independent assessment studies are also performed and published as a guide to code users and code developers. One study 2s found that improvements incorporated since the previous version (incluryng a new treatment of fuel pellet relocation and related fuel thermal conducti-vity effects) have resulted in a more realistic description of fuel behavior under moderate operating conditions.
The thermal and mechanical model development generally relates to the response regimes associated with highest power rods as opposed to core-average rods. FRAP-53 now yields a standard fuel centerline temperature error of 198 K for unpressurized rods and 254 K for pressurized rods. (See Table 3).
These discrepancies approach the present experimental uncertainties, however, and are of the same magnitude as those experienced by the other codes. Figure 16 shows a comparison of FRAP-S centerline temperature predictions to measured data obtained from a Halden fuel rod. Future research work is aimed at reducing these experimental uncertainties and to create a corresponding improvement in modeling capability.
FRAP-T is being developed to calculate the temperature increases and the accompanying time-and temperature-dependent processes expected during postulated occurrences such as LOCA's, power-cooling mismatch accidents, reactivity-initiated accidents, and inlet flow blockage, as well as the processes occurring during normal operation. The code will eventually be general enough to treat expected asymmetries and all of the phenomena occurring up to and including fuel melt.
The models for FRAP-T have been developed primarily from fundamental formulations 50 that they will apply over a wide range of response conditions and will not be limited by the range of available data. FRAP-T predicts the time dependence of many coupled variables at an arbitrary number of axial positions for any transient power history. Calculated are the fuel-rod temperature distribution, gap conductance, internal pressure, cladding strain, time and location of cladding failure, cladding surface temperature (including surface heat transfer),
and coolant conditions (including temperature, enthalpy, and quality). The primary input data required are the following: power history, descriptions of the fuel 3d cold state, the time-dependent conditions of the coolant surrounding the rod, the axial power profile, and code running requirements, including the mesh size, t.ime step, and convergence criteria for pressures and temperatures. The results of either a steady-state or an earlier transient calculation may, of course, be stored on tape and read by FRAP-T to satisfy these input requirements. FRAP-S/FRAPCON cr FRAP-T may be used for this purpose, and transient coolant conditions 2
calculated " with RELAP-4 may be read from a tape. The output may optionally include plots of up to 20 vari-ables as a function of time.
Both FRAP-T and FRAPCON have and will continue to undergo rigorous independent assessment to determine how well they predict experimental results. A summary of the st,ndard model errors for FRAP-T4 is given in Table 4.
Data comparisons 30 for two TREAT LOCA simulation tests were performed using FRAP-T4 to evaluate the effect of relocation and ballon model changes on cladding temperature and rod internal pressure calculations (see Figure 17). Surface temperature response under single phase steam cooling conditions was well represented by the model. Relocation effects en gap volume redistribution contributed to the overprediction of pressure for the relatively large gap, small plenum rod. An adequate representation of measured pressure response was obtained for the larger plenum rod, which is more typical of power reactor void volume conditions. Incorporation of strain rate dependence in the burst model improved predictions for time and duration of cladding rupture.
Versions up through FRAP-53, FRAPC0N-1 and FRAP-T4 are currently available for use trem the National Energy Sof tware Center at Argonne Nationel Laboratory. FRAP-T538 is under development.
Fuel Meltdown As a result of the Reactor Safety Studyaz analysis of a core meltdown, uncertainties in some of the physical phenomena were reexamined more closely. A review of available experimental data concluded that in general a 33 sizable body of useful data exists, but the experimental conditions are usually such that the results are not always directly applicable to the case of a reactor meltdown. The report also suggested that sensitivity studies using state-of-the-art models of meltdown should be conducted to determine the importance of physical phenor!ena ir relation to the overall consequences of the postulated meltdown accident.
As a result, programs were instituted to examine fission-product behavior during meltdown, natural convection in molten pools, interactions between molten core materials and concrete, steam explosions, and the effects of these p'.enomena on meltdown probabilities and consequences. Close cooperation with a similarly oriented program "G. A. Reymann (ed.), MATPRO-Version 10, A Handbook of Materials Properties for Usa in the Analysis of Light Water Reactor Fuel Rod Behavior, USNRC Report IREE-NUREC-ll80, February 1978.
i i 169 134
TABLE 3 C
Standard Model Errors FRAP-S3 Output parameter Sample (rods / pts)
Standard error (P - Mg)2/n-1)0.5
[(
i i=1 Fuel center ten:perature 33/290 254 K (pressurized)
Fuel center temperature 64/511 198 K (unpressurized)
Released fission gas 176/176 18.8% generated gas Rod Internal pressure 28/309 0.66 MPa (unpressurized)
Rod internal pressure 20/349 1.34 MPa (pressurized)
Gap closure heat rating 77/77 13.4 kW/m (local)
Fuel axial thermal expansion 19/173 0.37% active length Fuel aulal permanent expansion 100/368 0.44*/o active length Cladding hoop permanent strain 170/393 0.59% cladding OD Cindding axial permanent strain 115/161 0.47% active length Cladding surface oxide layer 48/84 6.6 #
Cladding hydrogen concentration 38/53 39 ppm
(
REFERENCE:
" INDEPENDENT FRAP-T4 ASSESSMENT" BY D. R. COLEMAN PRESENTED AT SIXTH NRC WRSR INFORMATION MEETING, 1978, NRC PDR) w Q
.a t.J1
27 Linear Heat Load (W'em) 0 100 200 300 400 500 600 3600
~
2700 P-P~
f f
Ba nup 4300 mwd T r
f
?
/
L
~
/
N Denmental Data FR AP-S2 and FR AP-53 p
i
/
i i
/
E E
/
6 0
a
,/
m
/
/
/
1 I
i 1
1 1
0 0
0 3
6 9
12 15 18 21 Linear Heat Load @WJft)
FIGURE 16. FRAP-S CENTERLINE TEMPERATURE PREDICTIONS COMPARED TO HALDEN R0D H8B END 0F LIFE DATA (REF:
TFBP-TR-172, JANUARY 1977)
'fl69 136'
TABLE 4 Standard Model Errors FRAP-T4 Output parameter Sample (rods / pts)
Standard error (P - M )2/n-1) ~
[(
i i
i=1 CHF power at known flow 18/78 0.06 kW/CC channel CHF flow at known power 18/78 400 kg/s-m2 initial fuel center temperature 21/32 280 K at SCRAM g
Fuel thermal decay constant 21/32 4.5 s during SCRAM Equilibrium fuel center temperature 21/32 54 K during SCRAM Cladding burst temperature at 158/158 290 K known pressure Cladding burst pressure at known 64/64 3.4 MPa temperature Cladding permanent hoop strain 370/370 57% cladding OD
(
REFERENCE:
" INDEPENDENT FRAP-T4 ASSESSMENT" BY D. R. COLEMAN, PRESENTED AT
{
SIXTH NRC WRSR INFORMATION MEETING, 1978, NRC PDR) e N
Internal Pressure and Cladding Surface Temperature
~
Response for Treat Test FRF-2 1700 Rod 12 3.5 Data 1500 Cold lower TP
.0 Hot upper predictions x
2 x
o Internal T&H, o
e 9x
^
2 1300 f
2.5 s
m E
E 8,1100 20 E
5 g
e,e m
o b
900 x
1.5 8
e a
m x
x v
t x
j 3
700 e
o o o x
x n
o o
o 1.0 o
x a
P c
y x
'6 500 8
y 6
0.5 5
T 300
,.p px' 0
5 02 4
6 8 1012141618 202224262830323436
]
Transient time (s)
FIGURE 17 (REF:
" INDEPENDENT FRAP-T4 ASSESSMENT" D. R. COLEMAN, SIXTH NRC WRSR INFORMATION MEETING )
30 in the Federal Republic of Gerr'any has also been instituted.
k' f the key findings to date include the following:
Three series of experiments conducted at three separate laboratories indicate that elevated system pressures are effective in reducing the probability of vapor explosion.a4 A relatively long time is required to establish steady-state single phase convection in molten pools at high Rayleigh numbers.85 Concretes with high and low carbona'.e contents behave in a qualitatively similar manner on contact with molten core materials.36 Concrete penetration is thermally dominated (I cm/ min is a nominal value).
Steel in the melt will chemically reduce decomposition gases to carbon monoxide and hydrogen, which will then burn on contact with air.36 1he initial heat transfer is dominated by the effects of concrete decomposition gases which tend to keep the melt thermally well mixed.
Single phase natural convection is not a dominant heat-transfer mechanism during core-concrete inter-actions.33 A first generation mechanistic model of core-concrete interactions has been developed 37 and an improved model is now undergoing assessment testing.
Significant progress has been made in identifying and characterizing these phenomena, but several key issues remain to be resolved. These include:
Directional partitioning of heat flux during interactions between concrete and molten core materials.
Probabilities associated with steam explosions.
Necessary scale of steam explosion experiments.
Prediccion of containment failure modes.
Leakage of radioactivity from containment.
Fission-Product Release and Transport Accurate estimates of fission product inventory in a reactor core can be made f rom a knowledge of the operating history. However, the amount and form of radionuclides released during a postulated accident can only be inferred by extrapolation of limited experimental data aad application of simplified analytical models. Experi-mental investigations have been conducted primarily out of reactor on small samples (1 to 100 g) of relatively low burnup fuel (trace to 4 GWd/MTM) with rare instances of higher burnup (up 20 GWD/MTM). Current acactice is to make what are judged to be conservative assumptionsMeegarding fission product release from the f uel and transport through the various barriers without detailed consideration of the mechanisms of release and transport.
Greater emphasis has recently been placed on instrumenting commercial and test reactors so that releases of activity from the fuel can be monitored with greater reliability and accuracy. Out of-reactor experiments 40'43 using irradiated fuel under more controlled conditions are elucidating the mechanisms of release and transport.
In res. tor experiments monitoring releases of activity during normal operating conditions as functions of power and of defect size for rods intentionally made defective are under way in Japan and France. The develo mechanisticmodelstotracethepathoffissionproductsoncetheyleavethefuelisalsoinprogress.4gmentof Research to date on the behavior of airborne fission products has involved the use of scaled f acilities having up to about 1% of the volume of the containment in a typical presssurized water reactor. This reseatch has examined atmospheric conditions, airborne concentrations, physical and chemical states of various species, mechanisms of fission product removal from the airborne state, and thermal effects. Particular attention has been paid to iodine, and one outcome of these investigations has been the establishment of Regulatory Guides38 specifying the partition of available iodine in the containment into discrete physical forms. Elemental vapor and chemically active particulate forms of iodine are readily removed from the air by chemical sprays and filtering systems. Methyl iodine, other organic iodides, and possibly hypaiodous acid have been identified as persistent airborne species but are conservatively believed to make up no more than 4% of the total iodine released into the containment. Fission product and fuel aerosols are effectively removed from the containment space by agglomeration and gravitational settling as well as by engineered safety features. Analytical models have been developed to predict the airborne concentration of fission products as a function of time for a given set of input data, including fission product concentration, particle-size distribution, containment atmospheric conditions, and geometry.43 ORNL researchers have developed p iiminary empirical models for the release of cesium and iodine in steam under LOCA temperature conditions (cladding temperature less than 1477 K, 2200F.)44 The models assume that the rplease is(and dif fusional release (that dif fusing from the fuel rod af ter the plenum gas has vented).
the sum.of two components: burst release (that carried out with escaping plenum gas when the rod ruMures),
1169 iM
31 A measure of the precision of the model was obtained by comparing the model predictions with the experimental data. This is done graphically in Figure 18 for both cesium and iodine.
44 Comparisons of these best-estimate predictions for total cesium and iodine release with those used in the Reactor Safety St g a2 show that the best estimate values are lower by a factor of 200 and 60, respectively than those used',n the Reactor Safety Study.
More recent experiments 45 conducted at ORNL at higher temperat'!res (1200*C to 1600 C) than would be expected under terminated LOCA conditions indicate a greatly accelerated s41 ease of cesium, iodine, and noble gases occurs at temperatures greater than approximately 1500 to 1400*C. The results of these tests suggest that some mechanism other than diffusion from either the pretest gap or the UO2 matrix is controlling fission product release at these higher temperatures. Table 5 compares the observed fission product release for four tests conducted at temperatures ranging from 1200*C to 1610*C.
Table 5.
Comparison of Results of Fission Product Release Tests Test a
Test Temperature period Percent of total inventory released No.
("C)
(min) 85 I34 129 kr Cs g
b c
HBU-ll 1200 10 1.3 0.012 0.018 d
HT-1 sl300 10 1.07 + $0.5 0.112 0.165 d
HT-2 sl445 7
5.0 + sl.0 4.82 2.35 d
HT-3 sl610 3
8.3 + sl.0 8.27 e
' Percentage release values are based on total inventory in the 12-in. (10.5-va) segment length. Percentage releases based on the heated length of 6 to 6.5 in.
(15.2 to 16.5 cm) will be approximately double the listed values.
DTotal test time was 27 min; release values were adjusted for a 10-min period.
c lncludes 85Kr released when this segment was used in a previous test at 900 C, test HBU-7.
dEstimated release during cladding expansion.
' Data acquisition and processing not yet ccmplete.
METALLURGY AND MATERIALS RESEARCH The NRC metallurgy and materials research program is concerned with the integrity of the primary-system pressure boundary in light-water reactors (LWRs). It is an experimental and analytical program designed to upgrade the bases for design, fabrication, operation, and inspection criteria, as well as for the analytical procedures required to evaluate performance under normal, upset, accident, and faulted conditions. Thus, a primary goal is to improve the definition of failure pro 5 abilities and failure modes, and to establish ways by which the failu n probabilities can be reduced if this is considered necessary.
The prin.ory system integrity research program consists of three major areas of research: (1) fracture and structural mechanics, (2) operational effects, and (3) flaw detection and evaluation.
The f racture and structural mechanics work encompasses (1) reactor vessel and piping-system per formance under pressure and thermal loading; (2) crack. initiation, propagation, and arrest (including static and dynamic studies and the use of irradiated specimens); and (3) response to operational and postulated conditions. In particular, the work on vessel response to postulated conditions includes thermal shock and steam-line-break accident conditions to assess the ef fects of abnormal pressures and thermal shock following the injection of cold emergency core cooling (ECC) water after various postulated loss-of-coolant accidents (LOCA's) or steam-line breaks.
The operational effects work is directed at obtaining data on (1) irradiation embrittlement, (2) annealing and re-irradiation, (3) residual element effects, (4) cycle crack growth, (5) steam generator tube integrity, (6) intergranular sensitization and stress-corrosion cracking and (7) neutron dosimetry.
The flaw detection and evaluation work covers (1) improved u.ltrasonic characterization of flaws, (2) acoustic emission studies of flaw growth in piping and pressure vessels and of flaws produced during welding, (3) im-proved eddy-current inspection of steam generator tubing, and (4) advanced nondestructive examination techniques.
Fracture and Structural Mechanics Vessel / Piping Performance and Response Fracture toughness and crack arrest in LWR vessel and piping materials have been studied extensively by r!ny organizations over the years. Much of the data was obtained with relatively small laboratory specimens, ' inch 1169 V40.
32 l IODINE CESluu TYPE OF EXPERIMENT
/
O 6
IMPLANT, PREDRILLED
/
V y
ihePL ANT, RUPTURED
/
[
803 -6 A
Csl IMPLANT, RUPTURED
/
o 4
Low BURNUP, PREDRILLED j
c m
HIGH BURNUP, PREDRILLED
/g 0
B N:GH BURNUP, RUPTURED A
b
/
l e
4 i
a
/
/
/
C
j 10
/
w e
/
/
2 jg S
/
o
/
y
/
m 10-'
/g o
/
/
/
10*2 l
l
/
10-3 10~3 10-2 io-i i
10 (02 3
10 10*
CALCULATED MASS RELEASED ( g)
FIGURE 18 COMPARIS0N OF CALCULATED AND OBSERVED CESIUM AND IODINE RELEASE FROM DEFECTED LWR FUEL RODS INT 0 STEAM
(
REFERENCE:
\\
1169 141
,k
33 or less in thickness." LU However, some very useful correlaticns have been made between small-specimen data and data obtained with 8-to 12-inch-thick compact specimens and 12-inch-thick dynamic tear specimens.M
Pipe-rupture studies have been conducted in the ductile regime, using piping generally less than 2 inches thick.
Research in this area has resulted in the development and validation of linear elastic fracture mechanics as a method for analyzing pressure vessels and pipin with cracks having sufficient constraint, under elastic stresses, and at or below the transition temperature. " 'S The criterion based on this method is embodied in A5ME Boiler and Pressure Vessel Code Secticn III, Appendix G, and has been validated for slow-load f racture toughness with up to 12-inch specimens (steel plate 02 f rom the Heavy Section Steel Technology Prngram) and f or rapid-load fracture toughness with up to 8-inch specimens. Criteria for fracture-safe operations under elastic plastic and fully plastic conditions are now of interest. For example, the arrest of a crack initiating in a local brittle region may be dependent on the increased toughness of the surrounding material. To quantify the transition from brittleness to increasing toughness, additional information is needed on material responses (e.g., possible crack initiation, propagation, and arrest) under appropriate test conditions of temperature, stress, and radiation-induced changes, and their gradients. Wesearch is under way at seweral laboratories and ss universities to develop an elastic plastic fracture-analysis criterion based on the J-R Curve and on tearing instability concepts," as well as to predict the elastic plastic stress state at the tip of a crack by means of a three-dimensional finite-element computer code. These techniques are being studied carefully from a theoretical standpoint for both pressure vessels and piping. The results will then be validated by experiments on a spectrum of test configurations, ranging from small to large specimens and to vessels. Unirradiated materials must be used to validate the criterion, but it must subsequently be validated with irradiated or simulated-irradiated materials as weII.
walls and in nozzle regions, under a series of temperature and stress conditions."y sized flaws placed in Pressure tests of 6-inch-thick pressure vessels have been carrieo out with carefull (See Table 6).
The flaw regions were formed either f rom brittle electron-beam welds to produce a " natural" crack as a result of a dynamic crack pop-in or were f atigue precracked. The tests were oesigned to produce very long ( ~ 12 inches) and deep (one-half the thickness) cracks for evaluating elastic and elastic plastic loading criteria and metho-dologies. The results will aid in establishing the relationship between the fracture-toughness and crack-stability criteria and the details of the ultimate fracture, and hence will aid in establishing the design margin of safety against failure.
Four thermal shock tests have t.een perf ormed (see Table 7) on cracked steel cylinders with degraded toughness properties somewhat simulatirg irradiation embrittlement. The results confirmed predictions of crack noniniti-ation under severe thermal shock loadings as well as crack initiation and estimated crack-arrest positior.s. It has been shown that linear elastic f racture mechanics does characterize the thermal shock methodology and that operational pressure-temperature transients can be accurately analyzed. " The beneficial effects of warm prestressing are currently being studied in more detail because this type of treatment is expected to severely limit crack initiation, and thereby penetration, in the thickness direction. "
Sustained-load testing of a vessel has bees, conducted to investigate the extent of fracture that would occur if a reactor vessel were to fail in a fully pressurized made. Specifically, the investigition considered the progress and extent of crack propanation that could result f rom the stored energy available in a high-temperature system (essentially loading resulting from system blowdown). The test showed that even though a through-the wall crack occurred at ductile shelf toughness temperatures, no further propagation occurred as a result of the sustained load."
Crack Arrest Crack initiation is governed by slow-load fracture toughness or rapid-load fracture toughness, whereas crack arrest is presently defined by the limits of the K curve (See Figure 19). Crack arrest has been actively studied in recent years M L The methodology for prbicting when, where, and under what conditions a running crack will stop has now been established f or specimen geometries ? and experimental validation is under way.
Standard test metheds and specimen geor.etries were submit ted to ASIM Ccmmittee E-24 in March 19D to be con-sidered as a tentative standard. An international cooperative testing program has been organized with 29 participants. The objective was to test the applicability of the proposed methods for measuring crack-arrest toughness in the range of practical interest and to define the clarifications and refinemerts that are f ound to be necessary. Two-dimensicaal, dynamic, finito element analyses are developed for cylinder geometries that will be applicable to test cylinders as well as reactor vessels. They have been used to analyze the thermal shock experiments and will be usej to analyze reactor vessels subjected to ECCS operation following a LOCA.
Thick specimens will be tested to determine the relationship between crack-arrest toughness measurement capacity and specimen thickness. An irradiation has been completed and specimen testing is ready to begin.
Theory and methodology are t'eing developed f or the analysis of crack arrest in reacter pressure vessels for situations such as thermal shock (resulting f rom ECCS operation) and embrittled region crack pop-in.
The analysis procedure will permit piediction not only of whether or not a crack will be arrested but also of the conditions for crack arrest (crack size in a given wall thickness, etc. ) in terms of the nature of the event, the initial operating conditions, the structural configuration and the initial crack configuration, and the appropriate toughness criterion. The relation between rapid-load fracture toughness and change in crack length as a function of temperature and irradiation is being used as input to the structu.al analysis of crack. crest.
The analysis considers the influence of the kinetic energy of the structure (during crack extension) on crack 1169 142-
TABLE 6
SUMMARY
OF TEST RESULTS FROM EIGHT 6-IN.-THICK INTERMEDIATE TEST VESSELS
(
REFERENCE:
ORNL/TM-5090)
NOMINAL TEST FLAW DIMENSIONS FRACTURE FRACTURE VESSEL TEMPERATURE DEPTH LENGTH FLAW PRESSURE STRAlg N0.
( F)
(in.)
(in.)
LOCATION (ksi)
(~)
'!- l 130 2.56 8.25 BASE METAL (o)b 28.8 0.92 V-2 32 2.53 8.?^
BASE METAL (o) 27.9 0.19 V-3 130 2.11 8.50 WELD METAL (o) 31.0 1.47 c
V-4 75 3.00 8.25 WELD METAL (i)d 26.5 0.17 75 3.10 8.10 BASE METAL (o) 26.5 0.17 V-6 190 1.87 5.25 WELD METAL (o)d 31.9 2.0 8
190 1.34 5.20 BASE METAL (i) 31.9 2.0 190 1.94 5.30 WELD METAL (i) 31.9 z.G V-5 190 1.20 3.75 BASE METAL (i)f 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 SASE METAL (i)f 26.9 1.05 a
0UTSICE CIRCUMFERENTIAL STRAIN ON CENTER LINE OF VESSEL REMOTE FROM FLAW.
b(o):
OUTSIDESURFACE,(1):
INSIDE SURFACE.
cCONTAINED TWO FLAWS d
FLAW WHERE FRACTURE OCCURRED.
' CONTAINED THREE FLAWS.
fN0ZZLE CORNER FLAW.
V -
TABLE 7.
TEST CONDITIONS FOR TSE-1. TSE-2, TSE-3, AND TSE-4 Test Conditions TSE-1 TSE-2 TSE-3 ISE-4 Test specinen TSV-1 TSV-2 TSV-1 ISV-2 Test specimen dimensions, m (in.)
00 0.53 (21) 0.53 (21) 0.53 (21) 0.53 (21) 10 0.24 (9.5) 0.24 (9.5) 0.24 (9.5) 0.24 (9.5)
Length 0.91 (36) 0.91 (36) 0.91 (36) 0.91 ' M)
Test specimen material A508, class 2 A508, class 2 A508, class 2 A508, class 2 Heat treatment Quench only from Quench only from Quench only from Quency only from 871*C (1600*F) 871'C (1600*F) 871'C (1600*F) 871*C (1600*F)
Flaw Long axial crack, Semicircular axial crack, Long axial crack, Long axial crack, a = 11 m (0.42 in.)
a = 19 m (0. 75 in. )
a = 11 m (0.42 in.)
a = 11 m (C.42 n.1 Temperatures, 'C ('F)
Wall (initial) 288 (550) 289(552) 291(555) 291 (555)
Sink (initial) 4 (40)
-23 (-10)
-23 (-10)
-25 (-13)
Sink (final) 7 (45)
-15 (4.5)
-15 (4.5)
-19 (-2)
Coolant Water 40 wt 1 methyl alcohol.
40 wt I methyl alcohol.
40 wt 1 methyl alcohol, 60 wt % water 60 wt 1 water 60 mt % water y
3 Coolant flow rate, m /hr (gpm) 59 (260) 114 (500)
+114 (500) sl14 (500)
Coolant pressure in test section, kPa (psi) 1520 (220) 917 (133) 965 (140) 1000 (145)
Back-pressure orifice diameter, m (in.)
25.43 (1.001) 43.18 (1.700) 43.18 (1.100) 43.18 (1.700) 3 3
3 K
s2800 (s500) s5700 (s10 )
s5700 (s10 )
s5700 (s10 )
Heat tran fer poeff{clent W m' (Btu hr*
ft'
'F'
)
(K /Kgc) 0.74 1.33 (0 = 75*)
7.13 1.29 g
Time of occurrence' of (Kg/KIc) = 1, min b
4 L7 Time of occurrence
- of (Kg/K!c) max, min 4
4 4
5 Duration of experiment, min 30 30 30 30 a Calculated tines based on measured temperature distributions. For TSE-3, the effect of core holes is not included.
artes along crack front.
RQ[RENCE: NUREG/CR-0107, October 1978 A
(ksi diT) (Pa 8) 6 220 - 242 a 10 I
I I
I I
I I
I I
l l
l l
l l
1 6
O Shabt>its (WC AP-7623) 200 - 220 = 10 O Riptinq and Crosley. HSST 180 - 198 x 10 5th asinual initormation Q
6 meeting.1971 paper No. 9 160 - 176 m 10
$ Un..'at>lisher! W data 6
E Materials Research Latioratory O
6 O
140 - 154 = 10 arrest stata.1972 HSST inf ormation m eting p$
no -ia2 io 6
a go j
o i
100 - 110 = 10" l
O l
w O
6 80 - 88 m 10 a$
O i
60 - 66 x 10 >
b
[
40 - 44 i O'.
o o
l I
I I
I I
I I
I I
I 20 - 22 iO'>
("C)
-89
-67
-44
-22 0
22 44 61 89 lil i
I I
I I
I I
I I
I
("F)
-160
-120 80
-40 0
40 80 120 160 200 TEMPEHATURE REL ATIVE TO NDT Q
FIGURE 19.
REFERENCE CURVE FOR MINIMUM TOUGHNESS
(
REFERENCE:
WRC BULLETIN 175 AND NUREG-0234)
37 arrest by evaluating the " path dependency" of the process, including multiple events of onset of crack exten-sion and arrest. The initial analysis will be based on liaear elastic material behavior. The need for developing an elastic plastic analysis will be evaluated from (and based on) the elastic analysis results. The crack arrest specimens and test method and analysis must lead to the adoption of standards by such organiza-tions as ASTM and ASME. 1he specimens and test methods must be evaluated for adequacy in reactor surveillance.
Validation is required for the initiation of a running rack iri *,imulated embrittled material in a pressure vessel configuration. The objective is to confirm the capability of the crack-arrest methodology to predict the arrest of a running crack before it reaches a critical size.
Operational Effects Irradiation Effects The effects of neutron bombardment on reacto-vessel beltline structural materials include an upward shift in the reference nil-ductility transition temperature by several hundred degrees Fahrenheit, a reduction in ductile shelf-level energy absorption strength, and a reduction in tensile ductility. All of these factors combine to reduce f racture toughness and the potential for crack arrest. Because of the problems inherent in research on irradiated materials--including space limitations in reactors, shipping-cask sizes, and radiation limits for hot cells--most research on irradiation effects on vessel materials has been conducted with test specimens 1 inch or less in thickness. However, initial correlations have been made between results (see Fig. 20) from small specimens and those from 2-and 4-inch thick compact specimens tested in 1975.s2 Systematic studies of post-irradiation heat treatment at temperatures above the operating temperature of irradiated steel have shown that a significant portion of the pre-irradiation toughness can be recovered in this way to extend the useful life of reactor vessels with renewed fracture-toughness capability.46 (See Figure 21)
Neutron-induced embrittlement in ferritic pressure vessel steels has been studied extensively.46'" It has been shown that embrittlement can be significantly reduced simply by complying with the draf t ASTM recommenda-tion for upper limits of 0.10 wt% copper and 0.012 wt% phosphorus in the chemical composition of the steel.
Furthermore, a mechanis.n by which copper af fects neutron embrittlement in steels has been proposed.64 (See Figure 22)
Cyclic Crack Growth Rate It is recognized that small flaws, material defects, and inhomogeneities will always exist to some extent in materials to be used in nuclear service. Although such irregularities will initially be below the established limits that would require repairs, they can grow as a result of cyclic loading during normal operation. The potential for cyclic crack growth in reactor structural materials should be experimentally assessed to gain confidence that flaws cannot grow to a " critical" size. Therefore, NRC is sponsoring research to extend the data ba9 and to improve the understanding of cyclic crack growth in reactor structural materials, especially for toe environment and cyclic loading rates that represent realistic reactor operating service.
Some crack growth rate data have been established for irradiated materials under reactor service. Much of the work on unirradiated steels was conducted at relatively rapid cyclic frequencies. More recently, however, slower cyclic rates (of I cpm and less) have shown significant increases in crack growth rates f rom cycling in a water environment." Because this slow cycling more nearly approaches the realistic service performance of an operating reactor, and using the results of an extensive study to determine the most realistic test param-eters, current research is emphasizing the use of a 1 cpm loading time for R ratios of 0.2 and 0.7 at pressure and temperature in water duplicating the chemistry of PWR water in tha development of data on cyclic crack growth rates.
Steam-Generator Tube Integrity Tubing for PWR steam generators is subject to wastage, cracking, and denting at support plate locations.
Denting is particularly insidious because it precludes meaningful eddy-current inspection. The large accumu-lated strain in the dented region causes primary-side intergranular stress-corrosion cracking, which can remain undetected until leakage. Large safety margins are established for steam generator tubing, so that large in-service degration (40 to 60%) of the tube wall can be tolerated.
Tubing typical of several major designs of PWR steam generators has been tested in both the burst and the collapse modes. In the first phase of the study, electric-discharge-machined (simulated) cracks with depths ranging from about 25% of the wall thickness to through-wall and areas of wastage ranging in depth from 25 to 90% of wall thickness were burst and collapse tested under simulated reactor conditions to determine their ef fect on both burst and collapse pressure. Leak rates from various size flaws were also determined. The second phase of tha study will involve similar testing of tubing containing defects induced by chemical corro-sive means.
Flawed tubing removed from operating steam generators will be tested in the future when available.
Margins of safety against burst and collapse were established from the test results, and an empirical predictive equation for steam generator tube burst pressure as a function of tube and flaw dimensions was developed.68 Stress-Corrosion Cracking Stress-assisted intergranular corrosion cracking in the BWR coolant environment continues to occur in seamless small-and intermediate-diameter austenitic steel piping. The primary factors causing this phenomenon are known. They include oxygen in the coolant, high stress, and sensitization of the stainless steel. The exact combination of factors that actually produces cracking has not yet been conclusively established.
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(
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(
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NUREG-0234) n 1169 149'
41 Research is under way in three primary areas of interest for intergranular stress-corrosion cracking (SCC). A field test is nearly complete for detecting and measuring the degree of sensitization and susceptibility to SCC of welded stainless steel piping.67 Laboratory studies are being extended to field confirmation. Adoption of the test as a standard acceptance and inspection procedure is anticipated.
The residual stress level resulting from welding in piping has been studied to develop an analytical method for predicting such residual stress levels on the basis of f abrication parameters." Evaluation of the analytical procedure was conducted on a variety of sizes and types of nuclear grade weldments.
Tha SCC characteristics of steam generator tube material (Inconel 600) are being investigated as a function of stress, strain, temperature, metallurgical condition, water chemistry, and other important parameters. fhe aim is to establish a predictive capability for stress-corrosion cracking in steam generator tubing during the service life of a steam generator.
, Neutron Dosimetry Neutron dosimetry--the measurement of the neutrons causing embrittlement of pressure vessel steel--is needed to accurately correlate steel embrittlement from surveillance irradiations with embrittlement in the power reactor vessel wall to assess safe reactor life. At present, mechanical property measurements are more accurate than neutron dosimetry. Laboratories and LWR vendors use ASIM procedures for neutron dosimetry in test and power reactor surveillance, but each uses its own set of experiments to verify that calculati as and extrapolations are conststent with measurements. Furthermore, an energy threshold of E > 1 MeV is currently being used as the criterion for neutron flux and fluence, even though research data" show that neutrons with energies between 0.1 and 1 MeV can contribute as much as 40% of the embrittIcment attributed to neutrons with energies higher than 1 MeV.
The neutron flux and the spectrum of neutrons by energy level are being both calculcted and measured in a wide variety of test and power reactors. The objective is to confirm procedures for calculating and extrapolating the neutron flux in reactor surveillance irradiations, for the experimental irradiations, as many as 20 different flux-monitor materials will be included to cover the entire neutron energy spectrum. In selected instances, spectrometry will be used to establish the spectrum in the important energy range between 0.1 and 1.0 MeV.
Confirmation procedures will use a simulated pressure vessel wall wherein neutron-flux monitors and mechanical property specimens can be irradiated for comparison with the pretest calculations. Primary mechanical property characterizations will be in fracture-mechanics terms.
Flaw Detec? ion and Evaluation In-Service Ultrasonic Inspection Inspection of nuclear reactor components by ultrasonic techniques is required both prior to service and during shutdown for periodic in-service inspections.Section XI of the ASME Boiler and Pressure Vessel Code, " Rules for In-Service Inspection of Nuclear Power Plant Components," defines inspection criteria and allowable flaw In the sizes, based on linear elastic fracture mechanics, for various locations within reactor components.
present inspection procedure, the pulse echo amplitude and search unit position are evaluated as a basis for flaw detection and sizing. Although ultrasonic testing is the presently accepted and most useful volumetric inspection technique, its reliability for flaw detection and sizing (using the Code procedure) is questionable Significant advances have been made recently in the signal processing of pulse-echo data and often inadequate.
to form a synthetic aperture focused image of high resolution for greatly enhanced flaw characterization.70-73 The upgrading of ultrasonic inspection is focusing on developing more aspects of the information resulting from Sensitivity of the results a pulse-echo test, including phase, frequency, amplitude, and search unit position.
to the specific operator, a specific calibration test, or a specific transducer is also being reduced or elimin-ated. Greater detection sensitivity is being developed, and flaws are being characterized with much improved resolution. Means for the storage and ready rctrieval of the information for meaningful reevaluation are also being developed. The importance of ultrasonic inspection records is expected to increase, with reference being made to past records for ccmparison with current information. For such comparisons to be most accurate, flaw and sensor locations must be accurately determined and recorded, and the inspection results must be made inde-pendent of changing transducers and electronics properties. The difficulty of ultrasonically inspecting austen-itic stainless steel base metal, weld deposits, and the interface is being minimized by processing the data to greatly decrease electronic and grain-boundary scattering noise and to obtain focused images independert of signal amplitude above a detection level. These procedures greatly increase the sensitivity and resolation of ultrasonic inspection and flaw evaluation in stainless steel. The laboratory test procedures are being vali-dated on realistic plate and piping samples, in addition to being adapted for typical in service inspection procedures.
Flaw Detection By Acoustic Emission The technique Continuous on-line surveillance represents a goal because feasibility is yet to be demonstrated.
to be employed is acoustic emission, and while important advances in instrumentation have Deen recently realized, acoustic emission data analysis and extrapolation to real structures still require development and final proof testing in operating nuclear systems.74 Furthermore, it is noted that acoustic emission data on crack growth may need validation by ultrasonic testing during a shutdown period inspection.
1169 150
42 Continuous acoustic emmission inspection during welding, for the detection of microcracking during weld solidi-fication ad cooling, has been established in nonnuclear applicationsM and has been carried forward in the nuclear application to demonstrate and characterize detection of cracks and other types of rejectable flaws. "
Acoustic emission is being developed for on-line flaw monitoring. Acoustic emission probes eventually will be placed on vessels and piping to monitor signals emitted during operation; improved nondestructive examination methods would then be used during shutdown periods for further characterization of the flaws detected. At present, relationships are being dra.n between acoustic emission signals and mechanical property effects obtaine1 during the testing of fracture and fatigue-type specimens of both plate and weldments. Acoustic emission signals from other sources that may be present during reactor monitoring are also being evaluated. A labor-atory program will then be conducted or, fully characterized natural flaws to validate both the detection and the quantification abilities of the techniques developed (See Figure 23). The techniques will then be extended to the full range of conditions prevalent in reactor operation, and procedures will be established for dis-tinguishing among acoustic emission signals from various sources such as flaws, strained regions, reactor operations, etc. With such a baseline library available, t will be possible to begin actual structure and i
reactor vessel monitoring so tnat signal detection and quantification will be much more meaningful.
Improved Eddy-Current inspection for Steam-Generator Tubing The presently used ASME Code eddy-current inspection techniques are fast, but they can produce unreliable inspection results because of the many independent variables that affect the signals. For example, the detec-tion of flaws in a tube dented region surrot'ded by corrosion products and the steam generator support plate is extremely difficult. Existing mathematical models will be used 'o develop computer programs for designing optimum probes, instrumentation, and techniques for improved edd,-current inspection of steam generator tubing.
Improved eddy-current techniques will be deve?oped ', separate the ef fects of diameter variations, probe wobble, tube supports, and conductivity variations from defect-size, defect-depth, and wall-thickness variations.
Mathematical models and computer codes for eddy-current tests will be used to computer-design optimized probes, instrumentation, and techniques for multi-frequency, multiproperty examinations. The program will develop at least two optimized designs: one for the ganeral inspection of steam generator tubing and another for special conditions such as denting. Optimized probes and instrumentation will subsequently be built, evaiuated in the laboratory, and finally validated by in-service inspections of steam generator tubing.
Advanced T a hniques Numerous new techniques for nondestructive examination are being developed. Such techniques are continually reviewed and, if seen to be especially promising, are funded in carefully controlled assessment studies. A program was recently initiated to study the feasibility of internal friction monitoring techniques for the prediction of incipient intergranular stress-corrosion cracking in welded stainless steel BWR piping.
OPERATIONAL SAFETV RESEARCH NRC sponsors a category of research termed " reactor operational safety," that is, research aimed at providing direct assistante to NRC officials concerned with the operational and operational safety aspects of nuclear power plants.
The NRC requires a defense-in-d(pth philosophy to ensure the safety of nuclear power plants. Essentiaily, this means that three levels of safety are incorporated: (1) the plant is designed and f abricated for maxima safety, (2) protective systems are provided to monitor and correct abnormal conditions, and (3) engineerti safety features are installed to mitigate the consequences of accidents.*
Criteria for the defense-in-depth concept are presented in Chapter I of Title 10 (" Energy") of the U.S Code of Federal Regulations, in particular, Appendix A (" General Design Criteria for Nuclear Power Plants") and Appendix B (" Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants") of Part 50 of Chapter 1.
The Operational Safety Research programs are related primarily to research in support of NRC imple-mentation of the criteria contained in these two appendices and the related NRC standards, guides, and branch technical positions. The topics currently addressed include qualification-testing evalution, fire protection, human factors, and noise diagnostics.
A bibliography of NRC-sponsored reports on operational safety research is included at the end of this report.
Qualification Testing Evaluation The qualification-testing evaluation program is forused e obtaining the data needed to answer certain questions about the testing of Class IE safety related** eg;ipment to assess performance during and after postulated R
An excellent short discussion of the defense-in-depth concept is contained in U.S. Nuclear Regulatory Commission Annual Report 1976.77 an Safety classification of the electrical equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal or are otherwise essential"in preventing significant releases of radioactive material to the environment.
I169 151
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SUMMATION AE EVENT COUNT \\S. ;0D, SPECIMEN 1-2A-6B, 650 F
(
REFERENCE:
"AC0USTIC EMISSION-MATERIAL PROPERTY RELATIONSHIPS FOR CONTINUOUS MONITORING 0F REACTORS" BY P. H. HUTTON, ET AL, SIXTH NRC WRSR INFORMATION MEETING, 1978,NRCPDR) 1169 152
44 accident tonditions. The near-term qualification tests program is being cor. ducted to answer questions about assessing conformance with IEEE Std 323-1974,78 " Standard for Qualifying Class IE Equipment for Nuclear Power Generatiy St ations," and NRC Regulatory Guide 1.89,75 " Qualification of Class IE Equipment for Nuclear Power Plants," which endorses IEEE Std 323-1974 with certain changes. The end products are Jata, criteria, and procedures that will enable the applicant and the NRC to ascertain that an acceptable qualification-testing program has been conducted.
The specific questions considered include aging; nuclear source term definition; synergisms; the performance indicators that must be monitored; f ailure definition; allowable thermal and nuclear-radiation-flux gradients; test sample preparation, quality control, mounting, and connections; chemical and steam ficw rates; degree of mixing required; degree of impingement; and vibration. The first thrPe will be discussed further in the following sections.
Aging Considerations of aging in the qualification test program are important because of the potential for scme aging mechanism, not d.stected through routine periodic testing, to create a weakened condition in a safety-related component. Such a weakened condition could result in common-mode failures in redundant safety-related equipment subjected to overstress conditicos resulting from an accident condition such as a LOCA. Qualification testing is difficult because the weakened conditions must be simulated by exposing the component to low levels of stress over long periods of time. Any practical aging qualification test must therefore be based on an accelerated-aging methndology. At present, ther.aal aging is simulated by using the Arrhenius equation as a basis for performing accelerated aging. The Arrhenius equation, which is used to extrapolate from high thermal stress applied for a short time to lower strtss applied for a longer time, is based on the assumptions that the chemical reaction rate is dependent only on temperature and that over the range of consideration it is constant.
For many safety-related materials in use today, there is some doubt as to the validity of using this assumption to extrapolate fr;m times of less than I year to times up to 40 years. The NRC research program is centered around this issue, and its goal is to develop and prove techniques that will adequately simulate long periods of aging at low levels of stress for currently used materials.
The current research effort on aging consists of six tasks as described in the succeeding sections:
Task 1 - Single-Environment Aging Tests Single environment aging tests are being conducted to obtain data on the separate effects of radiation, temperature, and humidity."o The tests at present are limited to polymeric electrical cable materials utilized with safety-related systems but will be expanded to include oth?r safety-related materials at a later date. The testing is based on the assumption that tne important failure mode of electrical cable will be mechanical damage of the insulating or jacket material caused by embrittlement. Data obtained as part of 'he state-of-the-art report and scoping tests have confirmed this assumption. The parameter used as a mea'ure of this damage is relative elongation of the insulating and jacket material with the conductor removed. From these tests, accelerated damage at the higher stress levels (acceleration functions) in a single environment have been obtained, using a test oeriod of about 1 year. Elongation is being used as a relative damage indicator to evaluate the aging-simulation techniques ano for comparison with naturally aged cable samples.
Task 2 - Combined-Environment Aging Tests Combined environment aging tests will be conducted to obtain data on the synergistic aging effect of temperature and radiation. Synergism with other aging parameters will be evaluated later in the program.
As in the case of the single-environment aging tests, relatively low stress levels and long test cyles will be used.
Temperatures from 90 to 150 C and radiation dose rates fear 103 to 105 rad /hr are planned for this testing.
Preliminary indicators shcw there is a synergistic ef fect with some materials when radiation and temperature stress are applied. The test methodology for accelerated aging testing for combin,.J stress has been developed and is currently being extended and verified with other materials.
Figure 24 summarizes the method by which single environment aging is normally carried out. Gencrdlly the experimenter overstresses the environment to accelerate the aging. For ccmbineo environments, rillen -"
prcposed accelerating matched sets of environments (e.g., temperature and radiation as shown in figure 25) to see if the same predicted result is obtained. Figure 26 shows aging data for chloroprene jacketing material including the separate aging etfects of radiation and temperature plus the combined aging effect.
Using these data, figure 27 shows how the single and combined environment studies can be analyzed according to the formalism of the proposed method. In effect, one is building on act.ual data, including synergistic effects, to piedict the damage to the material under various radiation and temperature conditions. In addition, the matched set approach can be used to accelercte by a chosen factor any ambient aging conditions Task 3 - Rate Effects Tests to determine rate effects are undeivay. Of particelar concern are the rate effects associated with oxygen diffusion and with radiation. These tests will determine the minimum accelerated aging period from which extrapolation to the required material life can be made.
l" 1169 151
45 CONVENTION AL ACCELERATED AGING Single environment thermal aging (const. accel. )
1
' spredicted T
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46 COMBINED ENVIRONMENT AGING Use environment: T,R u
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(
REFERENCE:
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~1169 157 TIME REQUIRED FOR e/e0 = 0.5
(
REFERENCE:
SAND 78-1907A)
49 Task 4 - Damage Indicators A study is in progress to identify alternative damage indicators that could be utilized in addition to the material elongation criterion currently used as the reference aging-damage indicator for electrical cable.
Although this study is at present limited in scope to the selection of a relative damage indicator to confirm the aging-simulation techniques for electrical cable insulation, the general question of damage criteria will have to be addressed in conjunction witn LOCA qualification testing of other safety related equipment. The current qualification-testing standards require that tSe component under test continue to carry out its prescribed function following the LOCA. It is usually not practical to demonstrate this in dny test that can be designed and used. A more basic measure of damage will have to be developed and used for important safety-related equipment.
Task 5 - Comparison Study Naturally aged samples are being collected so that the aging-simulation techniques being developed in tasks I through 4 can be checked with naturally aged material. Some comparisons have been completed showing that the accelerated aging methodology developed to date does simulate the long term natural aging.
fask 6-Requalification Tests As a backup to the aging-simulation techniques being developed in tasks 1 through 4, an alternative method is being evaluated in which resistance to aging degradation for short periods of time would be assessed and requalifiCation tests used to extend the acceptable lifetime in short time increments through the use of duplicate sacrificial samples.
Nuclear Source Term Definition This work covers the calculation of the nuclear source terms for the accident assumptions made in Regulatory Guide 1.89.
Progress to date has consisted of analysis to determine the time relationships following a LOCA for radiation doses, dose rates, energy spectra, and particle types. These data show that current industry practice with regard to radiation simulation testing may differ significantly in terms of dose rate, spectra, and particle type from that implied by Regulatory Guide 1.89.
The ongoing work in this area is aimed at deter-mining the importance of these differences in terms of potential damage to safety-related equipment.na.c, The current effort consists of three tasks.
Task 1 - Source-Term Calculations Additional source-term calculations will be made on the basis of Regulatory Guide 1.89 assumpticns, taking into account new codes and test data developed in other programs. Also per S rmed will be calculations based on the proposed revision to Regulatory Guide 1.89, which allows for reduced release assumptions for certain classes of safety-related equipment. In addition, source-term calculations will be made with best estimate LOCA release assumptions as required. Figures 28 and 29 show some of the calculations which have been made. Note that the " gap release" values (LCCA source term) are much lower than these derived from Regulatory Guide 1.89.
Task 2 - Evaluation of Radiation Simulators An evaluation is being made of the adequacy of radiation simulators currently used to duplicate the hypo-thetical environment following the radioactive release postulated in Regulstory Guide 1.89.
(See Figure 30 as an example). An initial assessment has been made by comparing the dose rates 5nd energy spectra resulting from the conservative accident assumptions in Regulatory Guide f.89 with the dose rates and energy spectra obtainable from practical simulators. The final simulator evaluation will take into account the practical significance of these differnces as evaluated in Task 3 of the LOCA testing. Cobalt-60, cesium-137, spent-fuel elements, and beta particle machines will be included in this assessment.
Task 3 - Assessment of Radiation Effects The damage to safety-related equipment materials will be determined as a ' unction of the gamma and beta dose rates. A determination will also be made of how closely the dose-rate profiles resulting from the Regulatory Guide 1.89 assumptions must be simulated during qualification testing. Materials studies will be conducted using available radiation damage data, and additional experimental data will be obtained as needed. Also of concern is the damage of safety-related equipment materials by beta particles as a function of depth-dose profiles. Materials studies will be conducted and supplemented by experimentation if needed.
Assessment of Test Methodologies This work, which was initiated in fiscal year 1975, covers the assessment of LOCA and MSLB (main steam line break) testing procedures, including synergisms amng the stresses applied. Progress to date has bee'n the development of test equipment and methods to comp e sequea.lal and simultaneous tests on identical test samples of safety-related material.86 The curre. effort consists of four tasks.
Task 1 - LOCA Qualification Tests At Sandia Laboratories, LOCA qualification tests have been conducted (1) sequentially, as. recommended in IEEE Std 323-1974, with radiation exposure preceding the steam and chemical spray environme
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GAMMA-RAY ENERGY RELEASE RATE FOR "BEST-ESTIMATE" SOURCE SHOWING THE CONTRIBUTIONS OF THE CONSTITUENTS, AND COMPARED TO THE TOTAL REGULATORY GUIDE 1.89 SOURCE RESULT.
(
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BETA ENERGY RELEASE RATE FOR "8EST-ESTIPATE" SOURCE SHOWING THE CONTRIBUTIONS OF THE CON-STITUENTS, AND COMPARED TO THE TOTAL REGULATORY GUIDE 1.89 SOURCE RESULT.
(
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CALCULATIONS OF RELATIVE DEPTii-DOSE PROFILES IN A TYPICAL ELECTRICAL POWER CABLE FOR COBALT-60 COMPARED TO THE BEST ESTIMATE BETA AND
'O GAMMA-RAY PLATE-0VT SOURCES AT COOLING TIMES OF ONE MINUTE AND FOUR DAYS AFTER A LOCA
(
REFERENCE:
"BEST-ESTIMATE LOCA RADIATION SIGNATURE" IRT 0056-005)
53 simJltaneously, with radiation and steam and chemical environments imposed together. The same experi-mental test chamber and identical test samples are used in both cases. The specimens from both tests are subjected to a qualitative comparison of performance, using on-line measurements and post-test evaluation.
Tests have been conducted on electrical caDles, paints, connector assemblies, cable splices, and lubricants.
The radiation source utilized was cobalt-60, and the usable test chamber size was approximately 10 by 20 by 51 cm (4 by 8 by 20 in.) The test profile was a composite of the PWR/BWR profile recommended in the appendix to IEEE Std 323-1974. The current series of tests was completed in 1977, and a report on syner-gistic ef fects in LOCA qualification testing has been prepared." No functional synergisms rere found in these preliminary tests.
Task 2 - Generic Test Data Generic test data have been obtained through a contract with a commerical testing laboratory. These data have been used by Sandia Laboratories in its evaluation of the synergistic effects associated with LOCA testing. The commerical testing laboratory has been used in an advisory ca M city for the design of a new test f acility at Sandia Laboratories and will continue to be used in formulating a long-term test program at Sandia.
Task 3 - New Test Facility Sandia has recently comp!eted construction of a new test f acility (see Figure 31) which will be able to accommodate larger and more diverse Class IE test items. The actual test chamber is 52 cm 10 by 150 cm high. The test facility is designed to allow a wide selection of radiation dose rates with better control of dose profiles. Typical coolant steam environments for postulated accidents can be applied simultan-eously with the postulated accident radiation environment.
Task 4 -Test Plan A test plan will be prepared for the new test facility by Sandia Laboratories, defining the components and materials, the types of tests to be conducted, and the performance parameters to be monitored and evaluated.
Th( basis for this test plan is an evaluation of the LOCA sensitivity of safety-related equipment. The design data on safety-related equipment have been obtained by subcontract with a reactor plant design organization using a pressurized water reactor currently under design. The specific data obtained include a list of safety-related equipment along with the functional requirements for normal service and for accident conditions, plant location, and normal-service and accident environmental conditions. In addition to the items that are currently t,eing tested (electrical cable, cable conaectors, and splices) it is anticipated that the test plan will include such items as electric motor and solenoid valve operators, limit switches, electronic components, junction boxes, pressure transmitters, neutron detectors, electrical signal cable for instruments, resistance temperature detectors, and materials and components utilized in hydrogen recombiners, decay heat removal equipment, and containment penetrations. Specific tests will probably include aging, thermal radiation LOCA, and main steam-line break.
In order to ensure that all qualification tests are conducted on materials and components that meet current NRC requirements for quality ssurance, a comprehensive review of all potential test specimens will be conducted before conducting any methodology tests Test specimens will be chosen from generic categories identified by a survey of manufacturers supplying the material or component to be tested. A pretest review of the specific test item will be made to identify factors that could impact the methodology assess-ment. When test specimens have been judged to have been designed, fabricated, and procured in accordance with NRC quality assurance requirements and this judgment has been confirmed by a pretest inspection and/or test, methodology-oriented qualification tests will be conducted for the purpose of confirming existing test procedures and to provide data that can be used to modify these procedures where required.
Fire Protection Research The fire protection research program is based w.
- search areas identified in NUREG-0050" and by review of current design standards and guidelines. The program is aimed at providing confirmation data relating to the operation of Class I system under the conditions they would encounter during the appropriate design-basis fires.
The following specific program elements are included in the fire protection research program:
1.
Obtain data on the effectiveness of cable-tray separation criteria in ensuring the functional integrity of redundant safety systems.
2.
Obtain data on the effectiveness of conduits, fire barriers, and penetration firestops.
3.
Obtain data on the effe aeness of coating materials.
4.
Obtain data on the fire retardancy of aged materials.
5.
Cotain data on IEEE Std 383-1974 and development of improved small-scale cable-system qualification tests.
6.
Obtain data on the effectiveness of safety-related equipment (other than cable) when subjected to exposure fires.
~1169 162 l
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Obtain data on fire detection system performance.
8.
Obtain data on the effectiveness of water and other fire extinguishing agents.
The first program element is based on work at Sandia Laboratories started in 1975 to confirm cable-tray-separation design practice. The scope of this effort was expanded, and program elements 2 through 6 added to cover the effectiveness of additional cable-tray configuratior. and components and an evaluation of separation criteria utilized for other safety-related equipment. Program elements 7 and 8 cover specific research areas required for licensing decisions and will be implemented in fiscal year 1979 and 1980. These elements will also be used in support of Regulatory Guide 1.120," " Fire Protaction Guidelines f or Nuclear Power Plants."
Program flement 1 - Cable-Tray Separation In support of some of the provisions of NRC Regulatory Guide 1.75" " Physical Independence of Electric Systems," tests were Nnducted at Sandia with varying separation distances to determine the minimum separ-ation necessary f or cables most susceptible to tire. Vertical separation distances from 152 cm (5 ft) down to 26.' cm (10.5 in) and horizontal separation distances f rom 91 cm (3 f t) down to 20 cm (8 in) were tested. For electrically initiated fires in a horizontal open-space configuration, it was determined that a fire will not propagate from the ignited tray to adjacent trays. These tests were conducted with fire retardant 12 gage single-conductor and 12 gage triplex wire, utilizing both uniform and random pattern cable packing.
Tests were also conducted with an experimental exposure (fuel) fire. The objective was to determine whether cable-tray separation alone is sufficient to prevent fire propagation between trays and between redundarat saf ety divisions if an exposure fire resulted in a f ully developed cable-tray fire.
The type and size of the worst-case exposure fire that must be considered for licensing are based on a fire-hazard analysis and will vary from plant to plant; they will also differ among different locations within the plant. Accordingly, no attempt was made to define a design-basis fire for the exposure-fire tests. Single-tray tests were conducted to find a reasonable set of conditions that would result in a fully developed cable-tray fire. The experimental exposure fire was then used in al; full-scale cable-tray exposure-fire tests. Propane burners were used to start an exposure fire in one tray, with a barrier placed between it and the tray above. When a fully developed fire was obtained in the first tray, the burners were turred off and the barrier was removed. This method allows experimental study of fire propa-gation from tray to tray under specific conditions and without the exposure fire effecting ths other cable trays.
A series of tests were conducted on arrays of cable trays, with both electrical and exposu/e-fire initiation.
An array of 14 closely spaced cable trays was used to simulate a single safety division. Simulated redundant safety divisions were separated by the required 152-cm (5-ft) vertical and 91-cm (3-ft) horizontal distance.
(See Figure 32). The results cf these tests were summarized in a Research Information Letter.S The principal conclusion was that a fully developed fire in the bottom cable tray of a stacked array may propagate to a redundant safety division without fire suppression systems (as expected). On the other hand, electrically initiated fires do not propagate because they do not result in a fully developed cable tray fire.
In order to determine the characteristics of a cat'le-tray fire in cable tunnels or in areas where structural walls are close enough to the tray to influence the fire, some of the tests were repeated to include the effect of walls and ceilings. The first tests were conducted with a 91-cm (3-ft) vertical and 30-cm (1-ft) horizontal separation. Other separation distances, representative fire loadings, and closed (dead-ended) tests may be included in the testing program. The preliminary indication is that there is a greater chance of fire propagation under these conditions than with a similar configuration in an open area.
In typical plant install dions, cable trays are oriented vertically at some locations and in others are oriented both vertically and horizontally. %ertical cable trays have been" and will be tested in both the open space ccrfiguratien and with walls and ceilings close enough to affect the fire.
Program Element 2-Ef fectiveness of Fire Shields Researchers at Sandia laboratories completed a series of tests using different fire shields:
ceramic wool blanket over ladder tray solid bottom tr y with no cover solid cover on ladder tray with no vents vented cover on solid bottom tray 2.54-cm (1-inch) fire barrier (thermal board) between trays using single and double tray configurations as well as electrical cable which passed the IEEE Std 383-1974 flame retardancy testv2 and cable which did not pass this test.
The results*3 of the Sandia research showed that all fire shield designs of fered some protecticn. None of the cable which passed the flame retardancy test in IEEE Std 383-1974 ignited. It is possible to ignite the cable which did not pass this flame retardancy test; however, no propagation was observed past the fire shields.
56 A
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I169 165' -
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57 Program Element 3 - Effectiveness of Fire-Petardant Coating Materials The objective of this program e'ement is to provide information on the effectiveness of fire-retardant coating m:terials when used in typical cable-tray installations A survey of coating materials available for use in table trays was initiated in August ;976. Generic t) pes were chosen for testing and evaluation in small-and large-scale cable systems tests. Small-scale tests on basic coating properties have been conducted by using six coatings and two cable types. Fu?l scale tests ware conducted using both single and double trap.
While the results" showed that all coatings offer a measure of additional protection, there was a wide range in the relative effectiveness of the different coatings tested. No propagation to the second tray was observed in any of two-tray tests in which cable that passed the IEEE 5td 383-1974 test was used.
(Propagation was observed in three tests involving cable which did not pass the IEEE Std 383-1974 test).
Overall, a good correlation was obtained bet-een small-scale and large-scale tests.
Program Element 4 - Fire Retardancy of Aged Materials The objective of program of element 4 is to provide infotmation on the fire retardancy of aged materials.
As presently planned, the program is aimed at cable and coating materials used in the first three program elements but may be expanded at a later date to include materials from other safety-related equipment.
Both accelerated and natural aging will M considered in the program. The results of an ongoing material-aging study at Sandia Laboratories wi's be applied. Some naturally aged cable and coating samples will be tested and the fire retardancy compa ed with that of cable aged by accelerated methods. The physical properties of cable material will be evaluated only to the estelt that they affect the fire retardancy of the material. Cracking and spalling of protective coating material applied to a cable tray and changes in the bond between coating and cable ant. "*m. coating and tray will be examined after simulated long-term aging.
The present plan calls for a survey of generic cables to identify the fire retardants in current use.
A screening test will be developed for determining the figu4e of merit for stability as a function of age and ambient temperatures. Small scale tests will be conducted on different generic cable insulations and fire-retardant additives to determine the worst-case combinations. Aging tests will be performed to determine aging-acceleratim f unctions using the methods being developed in the qualification-testing evaluation program at Sandia Laboratories. Concurrently, cables will be aged by currently used methods, and full-tray electrical initiation and/cr exposure-fire t+sts will be r.onducted.
Program Element 5 - Fire Tests for Cable Systems and Systems Components This program element was set up to ertablish how well defined and repeatable is the cable flame-retardancy test of IEEE Std 383-1974 and to what extent total cable-system performance can be predicted from this small-scale tests.
Certain test features not specificalh addressed or defired in the current ftandard were studied at Under-writers Laboratories to det rmine their significance, specifically:
Test cell size an6 configurat;on Air flow requirements Cable-tray design and orientation to the flame source Flame source energy rate Extent to which tray is filled with cable (percentage of volume)
Cable size and material Mixing cable sizes and materials Cable-tray orientation (vertical and/or horizontal)
As a result of this UL study" suggestions for revisions of IEEE Std 383-1974 were made with respect to (1) construction of cable trays. (2) test enclosure, (3) ventilation, (4) type, size and spacing of cable ties, (5) measurement of f uel and air rates, (6) flame temperature measurement, (7) initial ambient temper-ature, and (8) reporting of results.
Program Element 6 - Effects of Fires on Other Safet d elated Equipment The objective of this program element is to evaluate the ef fects of exposure fires on safety-related equipment.
C.ly safety-related equipment that is not totally isolated from redundant eauipment will be considered.
' scope of this program element will be established by other NRC studies on the probability of redundant eiy'related equipment being affected by the same exposure fire and the risk assessment for this event.
risk assessment will be made by first identifying those areas in the plant in hich it is difficult to stally isalate redundant safety related equipment (e.g., the cable spreading room and the comrol room) and then examining the ef fects of a plant-specific design-basis exposure fire on all safety relatto equiprent in that area.
Il69 166
58 Program Element 7-Performance of Fire-Dectection 5 stems This program element is concerned with testing the performance of fire-detection systems. It will provide data that can be used as the basis for a guide or standard for the design, installation, and utilization of systems for detecting fire and products of combustion. A survey and performance evaluation of currently used fire and smoke detection systems will be conducted. The evaluation will take into account constraints and requirements imposed by typical plant installations, including air flow and local air stratification.
Detection system sensitivities will be established and evaluated for adaquacy under conditions prevailing in typical design-basis fires; this should help determine detector locatlun and response-time requirements.
Development cf in place test methods will be considered for detection systems.
Program Element 8 - Effectiveness of Extinguishing Agents The objective of this program element is to test the effectiveness of water and other fire extinguishing agents and their potential damage to safety related equipment. An improved technical base will be provided for establishing criteria on when and where water and other fire extinguishing agents can be used. This effort is planned not as a complete study on fire extinguishing systems, but rather as a study of technical issues identified as requiring evaluatfor.
Current plans call for determining the minimum concentration and soaking time required for Halon and carbon,40xide to extinguish a fully developed cable-tray fire. Later testing will study the effective-ness of water as a means of fire suppression. The testing will be conducted in a facility large enough to simulate a full-scale cable fire and typical plant enclosure. Data on smoke and fire detection and actua-tion systems will also be obtained.
Noise Diagnostics Noise diagnostics refers to the stur' of the fluctuating portion of plant signals to gain an improved under-standing of function, especially sys' i dynamics. Typical plant signals include neutron flux, coolant pressure, temperat ees, acoustic pressure, vibration (acceleration, velocity, displacement), cnd coolant flow. Noise diagnostic techniques are attractive for system studies because they are nonperturbative to operations.
The noise diagnostic research program at the Oak Ridge National Laboratory [ conducted in conjunction with the NRC Of fices of Nuclear Regulation (NRR) and Standards Development (50)] has supported lice.. sing activities by the use of noise diagnostics techniques in independent assessments of core-barrel motion in operating pres-surl2ed water reactors and in-core instrument-tube vibrations in operating boiling water rcactors of the BWR-4 type. More recent noise diagnostic studies jointly supported by RES, NRR, and SD have been concerned with assessing the performance of existing loose parts-monitoring systems in operating reactors.
The-current RES prrfas includes methods development stu'iies, laboratory research on looseparts monitoring, and assessment of the use of noise diagnostics techniques to determine reactor stability. The noise diagnostics studies support the development of guides and standards such as Regulatory Guide 1.133,95 " Loose-Part Detection Program for the Primary System of light-Water-Cooled Reactorr," and proposed regulatory guides on core-barrel motion monitoring and pipe vibration. This research program st ) ports NRR ef forts as needed for independent maluation of the cause and correction of various vibration-in; aced malf unctions at operating nuclear pm.er plants.
Noise diagnostics has been used in the past to identify and monitor system malfunctions in a number of operating nucle 3r power plants; for example, it vss used by NRC to analyze and quantify excessive core-barrel motion in the Palisades nuclear power plant.**
(See Figure 33). The problem was studied with the aid of a Fourier analyzer, using onsite measurements ci signals from in-core and out of core neutron-flux chambers, primary coolant temperature variations, and out-of-vessel accelerometers to compute the power spectral density. Tech-nical specifications were later issued by NRC to provide guidance on monitoring for excessive co<e-barrel motion in operating PWRs. " More recently, noise analysis was used by NRC to assist in determining the cause, and monitoring the magnitude, of instrument-tube vibrations to EWR-4 cores.9a (See Figure 34).
Noise analysis is a powerful tool for diagnosing and identifying the source of malfunctions that cause the vibration of reactor internals and flow anomalies. The ultimate use of this technig.e would be in conjunction with an automated surveillance and diagnostic system to provide an alarm in the control room on the detection of an impending component failure or system malfunction.
Monitoring systems fc loose parts are being ins W ied in nuclear stations currently starting operation. They use signals from accelerometers and sonic detectors mounted on primary-system piping and the reactor pressure vessel to locate and iden'? ? hose metallic parts in the primary system. Loose parts could result in primary-system degradation by blocking fuel coolant channels, impairing valve functioning, or damaging pumps. Early detection and identification of the size and location of loose The performarce and use of existing systems were evaluated in 1977.garts are thus important to saf e ope ation.
Reactor khetics and stability measurements were made in 1978 at several BhRs, using standard transfer-function techniques and analytical calculations of reactor systems kine'ics. Several months are required to obtain stability results by this technique. In addition, the result is dependent on an exact knowledge of many reactor coefficients and the model employed. Noise diagnostics has been proposed by Japanese researchees as a means of measuring reactor stability directly.
Noise diagnostics has been applied to two recent operational incidents: the power oscillations at tL Fort St. Vrain nuclear power plant and the transient at Three Mile Island Unit II (TMI-2). In both cases, ORNL
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COMPARISON OF AN ABNORMAL SIGNATURE WITH A RANGE OF BASELINE SIGNATURES FOR PWRs.
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61 reseai chers were called in to evaluate plant signals in order to determine the condition of the plant. At TMI-2, ttw uRNL researchers were able to show in both the forced and natural circulation phases that the plant was ?r the condition stated and most importantly that there was n3 bulk boiling.
Hume
- tors Research The human factors research program is concerned with assessing the role of human errors in reactor operational safety. It includes spe:ific studies in support of human error investigations and the development of associated training programs by the NRC Of fice of Inspection and Enforcement, the study of safety related operator actions in support of the development of guides and standards by the NRC Of fice of Standards Development (SD), and a continuing review of the application of ergonomics in the design of nuclear power plants.
Human errors have been a significant cause of abnormal occurrences in nuclear power plants.t The contribution of human errors to the unavailability of safety systems and components was noted in the Reactor Safety Study.32 The latter stated that "an actuarial data base for human error rates in nuclear power plants does not exist" and "in general, the design of controls and displays and their arrangements on operator panels in the nuclear plants studied in this analysis deviate from human engineering standards specified for the design of man-machine 100 systems and accepted as standard practice for military systems." The report to the American Physical Society by the Study Group on Light-Water-Reactor Safety recommended that " 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, especially in the accident-simulation mode." Human errors and the design of control rooms were also identified as a major area of concern by former General Electric employees, in testimony before the Joint Committet
'n Atomic Energy.10' In an effort to determine whether improvements could be made, NRC contracted in 1976 with the Aerospace Corpora-tion for a study of control-room displays and operator performance; the final report o2 was issued in March 1977.
i This study identified a number of instances in which the design of control rooms in operating nuclear plants was not based on optimum human engineering principles. It did conclude that control-room designs and associated operating procedures, utilizing current training and licensing practices for nuclear operators, are sufficient to ensure safe operation. Improvements in control-room design, operator training, and operating practices to increase the margins of safety were recommended. Other recommendations for improved human engineering in control centers were presented.
Advancedcontrol-roomdesignsutilizinggraphicdisplays(CRTs)inconjunctionwithcomputersareheinggoposed 3 and for the next generation of nuclear plants. Typical examples are the General Electric Company's NUCLENET the advanced control room design developed by Combustion Engineering, Inc. These control rooms provide for more informative displays for system diagnosis and hence should contribute to safer and more efficient plant operation. Safety criteria and guides for assessing advanced control room designs will have to be prepared by NRC after the completion of a detailed technical review of the requirements for human engineering.
The NRC is a member of the Halden Program
- Group, which is conducting a number of studies on automatic process supervision ali control in nuclear power plants. The use of advanced control-room displays, computer operation, and diagnostic i is being studied and demonstrated experimentally at the Halden reactor f acility.
Training progro s in MORT (Management Oversight and Risk Tree), R505 (Reported Significant Observation Studies),
rist management, and accident investigation techniques have been given to DOE contractors and employees for several years. The NRC currently dces not have an ongoing progr w for training inspectors anr 'icensee reviewers in the principles of ergonomics. Many of the techniques and print i es presented in manuals uad training programs for the above-mentioned topics would, with minor changes, be applicable to nuclear power plant safety studies and investigations.
The NRC participates in the development of industrial guides and standards. An industrial standard (ANS 51.4, ANSI N660) has been draf ted and issurd for trial use in evaluating safety related operator actions.1" As noted in this standard, "there are now no generally accepted criteria for safety related operator actions."
The standard propcses criteria f or determining the time to be allowed for manually initiating the operation of safety systems. However, a firm technical base is lacking for the criteria proposed, and the need fcr research in this area is highlighted. Asaresult,NRCissponsoringamodestresearchgrogramtodeveloptheneeded l
and shows that it is possible technical base. The first phase of this research has recently been completed to develop information on the response time of rsactor operators to various operational incidents. Figure 35 shows the process used by ORNL researchers to obtain and evaluate information on operator response. For one event inadvertent safety injection, it was possible to obtain suf ficient data on specific operation actions to perform a graphical analysis using probability plotting. (See Figure 36). This very preliminary work suggests that the mean times to respond are within the range of values proposed in the draf t of ANSI N660 as released for trial use and comment. An expansion of this and related work is anticipated.
A general review of the role of human errors as reported in NRC licensee event reports and DOE abnormal occur-rence reports,is being made at INEL. This study is egected to categorize the role of human errors and identify potential future research that would contribute to reducing the potential for human errors in the operation of nuclear power plants.
t See the NRC Annual Reports for 1976,1977, and 1978.
An esperimental program conducted by the Organization for Economic Cooperation and Development (OECD)
,.at.the Halden BWR facility in Norway. (See also discussion of Halden on page 13.)
1169 170
62 Preliminary Search h f Events and Sites of Interest I f Site Work Consultant Support 7
I I l f Documentation of Literature Search Specific Events PAQ Survey Operator Interviews Azulysis of Results PAQ Survey PHASE - 1 PRODUCT Evaluation of Data Availability Outline Methods for Collection Preliminary Results for N660 Document Events Operator Estimates Non-Nuclear Data FIGURE 35.
WORK FLOW DIAGRAM ON SAFETY RELATED OPERATOR ACTION l,
1169 171
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64 CONCLUSION The NRC research programs in f uel behavior, metallurgy and materials, and operational safety have been developed in response to defined regulatory needs. These research programs have produced a considerable body of data of use to NRC, the public and industry in quantifying the safety margin of nuclear power plants. More information is expected in the future.
1169 173
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65 REFERENCES
- 1.
U.S. Nuclear Regulatory Commission, " Water Reactor Safety Research Program," NUREG-0006, February 1979.
Additional information on the USNRC LWR safety research program may be found in the following references:
L. S. Tong and G. L. Bennett, "NRC Water-Reactor Saf ety-Research Program," Nuclear Saf ety, 18, 1, January-February 1977. Available in public technical libraries.
G. L. Bennett, L. 5. Tong, and T. E. Murley, " Light Water Rcactor Safety Research in the United States,"
paper presented to the meeting of the International Atomic Energy Agency Advisory Group on Reactor Safety Research, February 14-18, 1977. Available for inspection and copying for a fee in the NRC Public Document Room.
T. E. Murley, L. 5. Tong and G. L. Bennett, " Summary of LWR Safety Research in the USA," USNRC Report NUREG-0234, May 1977.
T. E. Murley, L. S. Tong and G. L. Bennett, " Summary of LWR Safety Research in the United States of America,"
paper IAEA-CN-36/584, Nuclear Power and Its Fuel Cycle, International Atomic Energy Agency, Vienna,1977.
Available from IAEA.
" Summary of Current Reactor Safety Research," prepared by G. L. Bennett as Appendix D of " Plan for Research to Improve the Safety of Light-Water Nuclear Power Plants,' USNRC Report NUREG-0428, April 1978.
2.
Y. Y. Hsu and L. B. Thompson, " Summary of NRC LWR Safety Research Programs I: Thermal-Hydraulics and Computer Code Development," ANS Transactions, Vol. 32, page 90, June 1979.
3.
U.S. Atomic Energy Commission, " Acceptance Criteria for Emergency Core Cooling Systems f or Light-Water-Cooled Nuclear Power Reactors," Docket No. RM-50-1, December 1973.
4.
American Nuclear Society, ANS Standard 5.1, " Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors," Hinsdale, Illinois, October 1971 (Revised 1973).
5.
B. I. Spinrad, " Evaluation of Fission Product Af terbeat, Annual Report," July 1,1975-September 20, 1976,"
Oregon State University, USNRC Report NUREG-0018-4, January 1977.
M. A. Bjerke, et al., "A Review of Short-Term Fission-Product Decay Power," Nuclear Safety, Vol.18(5) pages 596-616, Octobe-1977.
2 6.
J. K. Dickens, et al., " Fission-Product Energy Release for Times Following Thermal-Neutron Fission of U
Between 2 and 14000 Seconds," USNRC Report ORNL/NUREG-14, Octaber 1977.
7.
J. L. Yarnell and P. J. Bendt, " Decay Heat from Products of U-235 Thermal Fission by Fast Response Boil-Of f Calorimetry," USNRC Repcrt LA-NUREG-6713, Septemtser 1977 8.
L. Baker and L. C. Just, " Studies of Metal-Water Reactions at High Temperatures; III. Experimental and Theoretical Studies of the Zirconium-Water Reaction," Argonne National Laboratory, ANL-6548, May 1962.
9.
J. V. Cathcart, et al., " Zirconium Metal-Water Oxidation Kinetics IV: Reaction Rate Studies," USNRC Report ORNL/NUREG-17, August 1977.
10.
S. Levine, Research Information Letter 9, "High Temperature Oxidation of Zircaloy Fuel Cladding in Steam,"
Memorandum to B. C. Rusche, Director CNRR/NRC, March 14, 1977. Available for inspection and copying for a fee in the NRC Public Document Room, 11.
R. A. Perkins, " Zirconium Metal-Water Oxidation Kinetics, II: Oxygen-18 Diffusion in p-Zircaloy," USNRC Report ORNL/NUREG/TM-19, July 1976.
12.
M. W. Mallett, W. M. Albrecht, and P. R. Wilson, "The Diffusion of Oxygen in Alpha and Beta Zircaloy-2 and Zircaloy-3 at High Temperatures," J. Electrochem. Soc.,lg,181-184(1959). Available in public technical libraries.
13.
H. M. Chung, A. M. Garde, and T. F. Kassner, " Mechanical Properties of Zircaloy Containing 0xygen," Light-Water-Reactor Safety Research Program: Quarterly Progress Report, July-September 1976, USNRC Report ANL-76-121, December 1976.
14.
T. F. Kassner, et al., "Zircaloy Cladding Embrittlement, Recommended Criteria," paper presented at the USNRC Sixth Water Reactor Safety Research Information Meeting, November 6-9, 1978. Available for inspection and copying for a fee in the NRC Public Document Room.
WNote: Unless otherwise indicated, NRC-sponsored reports ("NUREG") are available for sale from the Na'.ional Technical Information Service, Springfield, Virginia 22161.
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66 15.
E. Irvin, " Effects of Irradiation and Environment on the Mechanical Properties and Hydrogen Pickup of Zircaloy," in Zirconiym and its Alloys, the Electrochemical Society, Inc.,1966.
16.
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17.
A. A. Bauer, et al., " Evaluating Strength and Ductility of Irradiated Zircaloy, Quarterly Progress Reports for April-June 1976" USNRC Report BMI-NUREC-1967; January-March 1977, BMI-NUREG-1971; April-June 1977, BMI-NUREG-1976.
18.
R. H. Chapman, "Multirod Burst Test Program, Quarterly Progress Report for October-December 1976", USNRC Report ORNL/NUREG/TM-95, April 1977.
19.
M. E. Waterman, " Corrected and Updated Data for IFA-429 From Beginning of Life Through Jur., 1
%" USNRC Report NUREG/CR-0478, December 1978.
20.
C. R. Hann, et al., " Data Report for the NRC/PNL Halden Assembly IFA-431," USNRC Report PNL-2494, April 1978.
21.
R. W. Garner and D. T. Sparks, "Results of Gap Conductance Tests In the Power Burst Facility," Proceedings of Topical Meeting on Thermal Reactor Safety - Sun Valley, Idaho, CONF-770708, August 1977.
22.
R. Var Houten, " Fuel Rod Failure as a Consequence of Nucleate Boiling or Dryout," USNRC Report NUREG-0562, June 1979.
23 H. M. Chung, A. M. Garde, and T. F. Kassner, " Mechanical Properties of Zircaloy Containing Oxygen," USNRC Report NUREG/CR-0201, ANL-78-49, LWR Safety Research Program Quarterly Progress Report for January-March 1978, July 1978.
24.
T. F. Cook, "An Evaluation of Fuel Rod Behavior During Test LOC-II," USNRC Report NUREG/CR-0590 (TREE-1328),
March 1979.
25.
J. A. Cearien, et al., "FRAP-SJ: A Computer Code for the Steady-State Analysis of 0xide Fuel Rods Analytical Models and Input Manual" (TFBP-TR-164, March 1978, Revision 2).
D. R. Coleman, et al., "FRAP-53: Model Assesment Report" USNRC Report NUREG/CR-0786, April 1979.
26.
L. J. Sief ken, et al., "FRAF T4: A Computer Code for the Transient Analysis of 0xide Fuel Rods,"
CDAP-TR-78-027, July 1978. Available for inspection and copying for a fee in the NRC Public Document Room.
27.
D. D. Lanning, C. L. Mohr, F. E. Panisko, and K. B. Stewart, "GAPCON-THERMAL-3 Code Description," USNRC Report PNL-2434, January 1978.
28.
G. A. Berna, et al., "F MPCCN-1: A Computer Code for the Steady State Analysis of 0xide Fuel Rods,"
C0AP-TR-78-032-RI, November 1918. Available for inspection and copying for a fee in the NRC Public Document Room.
29.
"RELAP4/M005, A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems," Vols. 1-III, USNRC Report ANCR-NUREG-1335, September 1976.
- 5. R. Fischer, et al., "RELAP4/M006: A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems, User's Manual," EG&G Report COAP TR 003, January 1978.
30.
D. R. Coleman, " Independent FRAP-T4 Assessment," paper presented at the NRC Sixth Water Reactor Safety Research Information Meeting, November 6-9, 1978. Available for inspection and copying for a fee in the NRC Public Document Room.
31.
L. J. Siefken, et al, "FRAP-TS: A Computer Code for the Transient Analysis of 0xide Fuel Rods," USNRC Report NUREG/CR-0840, June 1979.
32.
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33.
W. B. Murfin, ed., " Core-Meltdown Experimental Review;" Sandia Laboratories, SAN 074-0382, August 1975.
34.
R. E. Henry, " Light Water Rea: tor Safety Research Program: Quarterly Progress Report, July-September 1976" USNRC Report ANL-76-121, December 1976.
R. E. Henry, H. Hohmann, and H. Kottowski, "The Ef fect of Pressure on Nacl-H O Exr'7sions," paper presented 2
at the Fourth CSNI Specialist Meeting on Euel-Coolant Interactions in Nuclear Reactar Safety, Bournemouth, UK, April 2-5, 1979.
L. S. Nelson, L. D. Buxton, and W. B. Lenedick, " Light Water Reactor Safety Research Program Quarterly Report, Octcber-December 1977," Section 2. " Steam Explosion Phenomena," USNRC Report NUREG/CR-0307, June
[, 1978.
}
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67 35.
F. A. Kulacki and A. A. Emara, " Transient Natural Convection in an Internally Heated Fluid Layer," Ghio State University, USNRC Report NUREG-0078, August 1976.
36.
D. A. Dahlgren, " Light Water Reactor Safety Research Program, Quarterly Report, January-March 1976,"
Sandia Laboratories, SAND-76-0369, September 1976.
37.
W. B. Murfin, "A Preliminary Model of Core / Concrete Interactions," Sandia Laboratories, SAND-77-0"'s, August 1977.
38.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.3(1.4), " Assumptions Used for Es aluating the Potential Radiological Consequences of a loss of Coolant Accident for Boiling Water Reactors (Pressurized Water Reactors)," June 1974.
39.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors," May 1974.
40.
R. A. Lorenz, J. L. Collins, and O. L. virkland, " Quarterly Progress Report of Fission Product Release from LWR Fuel for the Period October 4 Ter 1976," USNRC hport ORNL/NUREG/IM-88, Marr.h 1977.
41.
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42.
J. A. Gieseke, et al., " Analysis of Fission Product Transport under Terminated LOCA Conditions," USNRC Report BMI-NUREG-1990, December 1977.
H. Jordan, J. A. Gieseke, and Paul Baybutt, " TRAP-MELT Users Manual," USNRC Report NUREG/CR-0632, February 1979.
43.
A. K. Postma, R R. Sherry, and P. S. Tam, " Technological Bases for Models of 9 ray Washout of Airborne Contaminants in Containment Vessels," USNRC Report NUREG/CR-0009, October 1978.
A. K. Postma and W. F. Pasedag, "A Review of Mathematical Models f or Predicting Spray Removal of fission Products in Reactor Containment Vessel," USAEC Report WASH-1329, June 1974.
44.
A. P. Malinauskas, et al., ', ~ <sion Product Source Terms for the LWR Loss-of-Coolant Accident," paper presented at the NRC Sixth - S actor Safety Research Information Meeting, November 6-9, 1978. Avail-able for inspection and copying for a fee in the NRC Public Document Room.
P. Malinauskas, et al., " Quarterly Progress Report on Fi- ;on Product Behavior in LWRs for the Perind 45.
A.
October-December 1978," USNRC Report NUREG/CR-0682, April 1979.
46.
L. E. Steele, Seutron Irradiation Embrittlement of Reactor Pressure Vessel Steels," Technical Report Series No. 163, International Atomic Energy Agency, Vienna, 1975 (STI/ DOC /10/163).
S. Pellini, et al., " Review of Concepts and Status of Procedures for Fracture-Safe Desig, of Complex 47.
W.Welded Structures Involving Metals of Low to Ultra-High Strength Levels," NRL Report 6300, Naval Research Laboratory, Washington, D.C., June 19ra.
48.
F. J. Loss and W. S. Pellini, " Coupling of fracture Mechanics and Transition Temperature Approaches to Fracture-Safe Design," NRL-6913, Naval Research Laboratory, Washington, D.C., April 14, 1969.
" Fracture Toughness Testing and Its Applications," American Society for Testing and Materials, Philadelphia, 49.
Sheckherd, and R. A. Wulleart, " Fracture Toughness Data for Ferritic Nuclear Pressure W. L. Server, J. W.
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Vessel Materials," EPRI NP-119, prepared for the Electric Power Research Insititute by Effects Technology, Inc., Santa Barbara, California, April 1976. Available from the Electric Power Research Institute, Research Reports Center, P.O. Box 10090, Palo Alto, Colifornia 94303.
F. J. Loss, " Dynamic Tear Test Investigations of the Fracture Toughness of Thick-Section Steel," NRL-7056, 51.
Naval Research Laboratory, Washington, D.C., May 1970.
W. O. Shabbits, " Dynamic Fracture Toughness Properties of Heavy Section t.533 Grade B, Class 1 Steel Plate,"
52.
WCAP-7623, Westinghouse Electric Corporation, Pittsburgh, PA, December 1970.
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F. J. Witt and T. R. Mager, "A Procedure for Determining Bounding Values on Fracture Toughness K at Any Ic 54.
Tempu ature," ORNL/TM-3894, October, 1972.
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P. C. Paris, " Fracture Mechanics in the Elastic-Plastic Regime," in Flaw Growth and Structure, American Society for Testing and Materials, Philadelphia, PA, ASIM STP 631, 1977, pp. 3-27.
56.
P. C. Paris, et al., "A Treatment of the Subject of Tearing Instability," USNRC Report NUREG-0311, August 1977.
57.
J. G. Merkle, G. D. Whitman, and R. H. Bryan, "An Evaluation of the HSST Program Intermediate Pressure Vessel Tests in Terms of Light-Water-Reactor Pressure Vesse! Safety," ORNL/IM 5090, November 1975.
58.
R. D. Cheverton, S. K. Iskander and S. E. Bolt, " Applicability of LEFM to the Analysis of PWR Vessels Under LOCA-ECC Thermal Shock Conditions," USNRC Report NUREG/CR-0107, October 1978.
59.
F. J. Loss, R. A. Gray, Jr., and J. R. Hawthorne, " Investigation of Warm Prestress for the Case of Small ST During a Reactor Loss of-Coolant Accident," NRL Report 8198, March 9, 1978.
60.
R. H. Bryan, et al., " Test of 6-in.-thick Pressure Vessels Series 3: Intermediate Test Vessel V-7A Under Sustained Loading," USNRC Report ORNL/NUREG-9, February 1978.
61.
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62.
J. A. Davidson, L. J. Ceschini, R. P. Shogan, and G. V. Rao, "The Irradiated Dynamic Fracture Toughness of ASTM A533, Grade B, Class 1 Steel Plate and Submerged Arc Weldment," Heavy Section Steel Technology Program, Technical Report No. 41, WCAP-8775, Westinghouse Electric Corporation, Pittsburgh, PA, October 1976.
63.
J. R. Hawthorne, " Demonstration of Improved Radiation Embrittlement Resistance of A533-8 Steel Through Control of Selected Residual Elements," in Proceedings of Symposium on Radiation Effects on Structural Alloys for Nuclear Reactor Applications, Niagara Falls, Canada, June 29-July 1, 1970, American Society for Testing and Materials, Philadelphia, PA, ASTM SIP 481, 1970, pp.96-127.
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F. A. Smidt and J. A. Sprague, " Property Changes Resulting from Impurity-Defect Interactions in Iron and Pressure Vessel Alloys," in Effects of Radiation on Substructure and Mechanical Properties of Metals and A11nys, American Society for Iesting and Materials, Philadelphia, PA, ASTM STP 529, 1973, p. 78.
65.
T. R. Hager, J. D. Landes, D. M. Moon, and V. J. McLaughlin, "The Ef fect of Low Frequencies on the f atigue Crack Growth Characteristics of A533 Grade B Class 1 Plate in an Environment of High-Temperature Primary Grade Nuclear Reactor Water," Technical Report No. 35, WCAP-8256, Westinghouse Electric Corporation, December 1973.
66.
J. M. Alzheimer, et al., " Steam Generator Tube Integrity Program Phase I Report " USNRC Report NUREG/CR-0718, PNL-2937, March 1979.
67.
W. L. Clarke and V. M. Romero, " Detection of Sensitization in Stainless Steel:
li. EPR Method for Nonde-structive Field Tests," USNRC Report GEAP-12697, February 1978.
68.
E. F. Rybicki, " Residual Stresses Due to Weld Repoi s, Cladding and Electron Beam delds and Effect of Residual Stresses on Fracture Behavior," Annual Report September 1,1977 to November 30, 1978, USNRC Report NUREC/CR-0059, BMI-2012, December 1978.
69.
C. Z. Serpan, Jr., " Damage-Function Analysis of Neutron-Induced Embritt-12-B Steel at 550F (288C)," in Ef fects of Radiation on Substructure and Mechanical Properties of Metals and Alloys, American Society for Testing and Materials, Philadelphia, PA, ASTM STP 529, 1973, pp.92-106.
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J. R. Frederick, J. A. Seydel, and R. C. Fairchild, " Improved Ultrasonic Nondestructive Testing of Pressure Vessels, Yearly Progress Report, August 1,1974-July 31,1975," USNRC Report NUREG-0007-1, January 1976.
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J. R. Frederick, R. C. Fairchild, and B. H. Anderson, " Improved Ultrasonic Testing of Pressure Vessels, Annual Progress Report, Augast 1, 1975-July 31, 1976." USNRC Report NUREG-C007-2, September 1977.
72.
J. R. Frederick et al., " Improved Ultrasonic Nondestructive Testing of Pressure Vessels," USNRC Report NUREG CR-0135, May 1978.
73.
J. R. Frederick et al., " Improved Ultrasonic Nondestructive Testing of Pressure Vessels," USNRC Report NUREG/CR-0581, January 1979.
74.
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Progress Repe-t No. O, March 1, 1972-March 1, 1973, Final Technical Report to Steering Committee, EEI Project RP 79 (Edison Electric Institute / Tennessee Valley Authority), Southwest Research Institute, May 21, 1973.
Available for purchase from the National Technical Information Service, Springfield, VA 22161.
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D. W. Piine and R. H. Clark, "Inspectior f Nuclear Reactor Welding by Acoustic Emission, First Interim Report," May 1975. Available in the NRC Public Document Room in file for Report NUREG-0006.
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80.
K. T. Gillen, E. A. Salazar, and C. W. Frank, Proposed Research on Class I Components to Test a General Approach to Accelerated Aging Unde-Combined Stress Environments, USNRC Report SAND 76-0715, April 197E 81.
K. T. Gillen, " Accelerated Aging in Comtined Stress Environments," Proceedings of Conference on Environmental Degradation of Engineering Materials, VPI Press, 1977 82.
K. T. Gillen, "A Method for Combined Environment Accelerated Aging," ENS /ANS Topical Meeting on Naclear Power Reactor Safety, Brussels, Belgium, Oct. 16-19, 1978.
83.
L. L. Bonzon, Radiation Signature Following the Hypothesized LOCA, USNRL Heport SAND 76-0740, September 1977, revised October 1977.
84.
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85.
L. L. Bonzon, N. A. Lurie, D. H. Houston, and J. A. Naber, Definition of loss-of-Coolant Accident Radiation Source: Summary and Conclusions, USNRC Report NUREG/CR-0183, May 1978.
86.
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87.
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88.
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91.
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92.
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93.
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94.
L. J. Przybyla and W. J. Christian, " Development and Verification of Fire Tests for Cable Systems and System Components," Quarterly Report 4, March 1 - May 31,1978, USRNC Report NUREG/CR-0346, September 1978.
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w-
\\\\69 179
71 BIBLIOGRAPHY OF NRC SPGN50 RED RFPORTS ON OPERATIONAL SAFETY RESEARCH FIRE PROTECTION STUDIES
" Analysis cf CaDie Tray Fires," L.W. Hunter, Sixth Water Reactor Safety Research Infrrmation Meeting of the USNRC, Gaithersburg, MD, November 1978.
" Applied Physics Laboratory Cable Fire Analysis," L. W. Hunter, USNRC Fire Protection Review Group, Albuqueraue, NM, September 1978.
Nable Tray Fires," L. W. Hunter, Johns Hopkins University Applied Physics Laboratory, May 1977.
" Cable Tray Fire Tests," L. J. Klamerus, SAND 78-18iOC, PES-IEEE 1979 Winter Power Meeting, New York, NY, February 1979.
" Cable Tray Fire Tests," L. J. Klamerus and R. H. Nilson, SAND 77-Il25C, Sandia, Albuquerque, NM, July 1977.
" Development and Verification of Fire Tests for Cable Systems and dystem Components," Quarterly Reports 2 and 3, for USNRC, UL, Inc., Northbrook, IL, Sept. 1977 - Feb. 1978
" Development and Verification of Fire Tests for Cable Systems and System Components," Quarterly Report 4 L. J. Przybyla and W. J. Christian, UL-USNC 75 Q4, NUREG/CR-0346, September 1978.
"The Feasibility of Modeling Horizontal Cable Tray Fire Tests by Klamerus," L. W. Hunter, Fif th Water Reactor Safety Research Information Meeting of the USNRC. Gaithersburg, MD, November 1977.
" Fires in Nuclear Power Plants," L. W. Huntec, APL Accomplishments for FY 1978, n d.
" Fire Protection Research Quartarly Progress Report October - December 1977," L. J. Klamerus, SAND 78-0477, Sandia Laboratories, Albuquerque, N.M., August 1918.
" Fire Frotection Research Quarterly Pr' gress Report January - March 1978," L. J. Klamerus, F. R. Krause, 5AND79-087L Sandia Laboratories, Albuquerq e, NM, May 1979.
" Fire fasearch on Grouped Electrical Cables," L. J. Klamerus, SAND 79-0031, Sixth Annual Energy Conference and Exhibition, WATTec, Knoxville, TN, February 1979.
"The Fire Resistance of Cable Penetration Seals. I: Theory for Radially Isothermal Cables," L. W. Hunter, FPP TR-40, Johns Hopkins University Applied Physics Laboratory, March 1979.
" Fire Spread f rom a Burning Horizontal Tray to Other Trays Directly Above," L. W. Hunter, NRC-77-005, Johns Hopkins University Applied Physics Labora ory, November 1977.
" Effects of Cable Spacing in a Horizontal Tray," L. W. Hunter, Johns Hopkins University Applied Physics Laboratory, February 1978.
"The Ignition of a Thin Cable in a Pre-existing Fire Plume," L. W. Hunter, Johns Hopkins University Applied Physics Laboratory, August 1977.
"Models of Cable Tray Fires," L. W. Hunter, Center fcr Fire Research, National Bureau of Standards, June 1978.
"Models of Horizontal E nectric Cables and Cable Trays Exposed to a Fire Plume," L. W. Hunter, NUREG/CR-0376, Johns aopkins University Applied Physics Laboratory, Septmoer 1979.
" Nuclear Power Plant Fire Protection Research," G. L. Cano, L. J. Klamerus, and E. A. Bernard, 5AND77-0389A, ANS Topical Meeting, Chattanooga, TN, August 1977.
"On the Mechanism of Upward Fire Spread Through a Stack of Horizontal Cable Trays," L. W. Hunter, Johns Hopkins University Applied Physics Laboratory, May 1978.
"A Preliminary Report on Fire Protection Research Program (Jaly 6,1977 Test)," L. J. Klamerus, 5AND77-1424, Sandia Laboratories, Albuquerque, NM, October 1977.
"A Preliminary Report on Fire Protection Research Program Fire Barriers and Fire Retardant Coatings Tests,"
L. J. Klamerus. NUREG/CR-0381 (5AND78-1456), Sandia Laboratories, Albuquerque, NM, September 1978.
Il69 180
72 "A P eliminary Report on Fire Protection Research Program Fire Barriers and Suppression (September 15, 1978),"
L. J. Klamerus, NUREG/CR-0596 (SAND 78-2238), Sandia Laboratories, Albuquerque, NM, December 1978.
"A Preliminary Report on Fire Protection Research Program Fire Retardant Coatings Tests (December 7,1977 -
January 31, 1978)," L. J. Klamerus, SAND 78-0518, Sandia Laboratories, Albuquerque, NM, March 1978.
" Quick Look Report on Fire Protection Research," L. J. Klamerus, Sandia laboratories, Albuquerque, NM, July 1976, August 1976, October 1976, November 1976, December 1976, February 1977, March 1977, July 1977 (eight reports).
" Quick Look Report on Full Scale Fire Test No.
1," L. J. Przyby)a and W. J. Christian, UL-USNC 83 QL-1, UL, Inc., Northbrook, IL, October 1978.
" Quick look Report on IEEE 383 Cable Experiments," L. J. Przybyla and W. J. Christian, UL-USNC 75 QL-1, UL, Inc., Northbrook, IL, June 1978.
"A Report on Fire Protection Research Program Corner Effects Tests," L. J. Klamerus, SAND 79-0914, Sandia Laboratories, Albuquerque, NM, May 1979.
" Status of the Fi'e Protection Research (FPR) Program," L. J. Klamerus, SAND 78-2025C, Sixth Water Reactor Safety Research Information Meeting, Gaithersburg, MD, November 1978.
"The Suppression of Cable Tray Fires," L. W. Hunter, Johm Hopkins University Applied Physics Laboratory, December 197f
" Transient Heat Transfer to an Insulated Cable," L. W. Hunter and S. Favin, Johns Hopkins University Applied Physics Laboratory, April 1978.
" Upward Fire Propagation in a Vertical Cable Tray. I: Fire Propagation Over Central, Thermally Thin Cables,"
L. W. Hunter, FPP TR-39, Johns Hopkins University Applied Physics Laboratory, March 1979.
QUALIFICATION TESTING EVALUATION
" Accelerated Aging in Combined Stress Environments," K. T. Gillen, SAND 77-0511A, Proceedinas of the International Conference on Environmental Degradation of Engineering Materials, Virginia Polytechnic Institute Press, October 1977.
" Accelerated Aging Studies of Electric Cable Material," K. T. Gillen and E. A. Salazar, SAND 77-1654C, USNRC Fifth Water Reactor Safety Research Information Meeting, November 1977.
" Adequacy of Radiation Sources for Qualification of Class lE Reactor Componerts," N. A. Lurie, J. A. Naber, and L. L. Bonzon. IRT 8167-0U3/ SAND 78-0161 A, Trans. ANS, Vol. 28, page 638, June 1978.
" Aging of Nuclear Power Plant Safety Cables," K. T. Gillen and E. A. Salazar, SAND 78-0344, International Topical Meeting on Nuclear Power Reactor Safety, Brussels, Belgium, October 1978.
" Apparatus and Procedures for Qualification of Class IE Electrical Cable," D. V. Paulson and S. P. Carfagno, FIRL Final Report F-C4598-2, Prepared for Sandia Laboratories, Albuquerque, NP, Octeer 1977.
"The Best-Estimate LOCA Radiation Signature: What It Means to Equipment Qualificat.on," L. L. Bonzon and N. A. Lurie, SAND 78-0349, International Topical Meeting on Nuclear Power Reactor Safety, Brussels, Belgium, October 1978.
" Calculations to Support Radiation Simulator Adequacy Assessments for Class 1 Equipment," N. A. Lurie, Draf t IRT 8167-010, for Sandia Laboratories, Albuquerque, NM, April 1978.
" Comparison of Polymer Flammabilities and a Determination of the Loss of Fire Retardant Additives with Aging,"
R. L. Clough and E. A. Salazar, SAND 79-0978C, Sandia Laboratories, Albuquerque, NM, April 1979.
" Computer Modeling of Polymer Radiation Chemistry," R. L. Clough, Fifth International Symposium on the Chemistry of the Organic Solid State, Brandeis University, Waltham, MA, June 1978.
"A Consolidated Program to Evaluate Class 1E Equipment Qualification Techniques," L. L. Bonzon, K. T. Gillen and E. A. Salazar, SAND 77-1075A, Trans. ANS, 27:684, November 1977.
" Definition of Loss-of-Coolant Accident Radiation Source," L. L. Bonzon, N. A. Lurie, C. M. Houston, and J. A. Naber, SAND 78-0050, Sandia Laboratories, Albuquerque, NM, February 1978.
" Definition of Loss-of-Coolant Accident Radiation Source: Summary and Conclusions," L. L. Bonzon, N. A. Lurie, D. H. Houston, and J. A. Naber. NUREG/CR-0183 (SAND 78-0091), Sandia Laboratories, Albuqueraue, NM, May 1978.
" Design Basis Event (DBE) Testing," L. L. Bonzon SAND 77-0150C, Technology Monograph, A Compilation of Qualification Considerations for Nuclear Power Generating Stations, Institute of Environmenta.1 Sciences,1977.
"An Experimental Investigation of Synergisms in Class lE Components,"
L. L. Bonzon, SAND 78-0346, International Topical Meeting op Nuclear Power Reactor Safety, Brussels, Belgium, October 1978.
4'
.l169 181
73 "An Experimental Investigation of Synergisms in Class 1 Components Subjected to LOCA Typetests," L. L. Bonzon, NUREG/CR-0275 (5AND78-0067), Sandia Laboratories, Albuquerque, NM, August 1978.
" Experimental Verification of a Combined Environment Accelerated Aging to Electrical Cable Material,"
K. T. Gillen, SAND 78-1907C, USNRC Sixth Water Reactor Safety Research Information Meeting, November 1978.
"The Hypothesized LOCA Radiation Signature and the Problem of Simulator Adequacy," N. A. Lurie and L. L. Bonzon.
SAND 78-0348/IRT 8167-005, International Topical Meeting on Nuclear Power Reactor Safety, Brussels, Belgium, October 1978.
" Implications of Sorption Effects of Accelerated Aging Studies," K. T. Gillen, SAN 079-0925C, Sandia Labora-tories Albuquerque, NM, April 1979.
"A Method for Accelerated Aging under Combined Environmental Stress Conditions," K. T. Gillen, SAND 79-0939C, Sandia Laboratories, Albuquerque, NM, April 1979.
"A Method for Combined Environment Accelerated Aging," K. T. Gillen, SAND 78-0501, International Topical Meeting on Nuclear Power Reactor Safety, Brussels, Belgium, October 1978.
" Methodology Assessment: An Overview of the Qualification Testing (QTE) Program," R. E. Luna and L. L. Bonzon, SAND 78-0342, International Topical Meeting on Nuclear Power Reactor Safety, October 1978.
"A Model for Combined Environment Accelerated Aging Applied to a Neoprene Cable Jacketing Material,"
K. T. Gillen and E. A. Salazar, SAND 78-0559, 1978 Conf (rence on Electrical Insulation and Dielectric Phenomena, Pocono Manor, PA, October 1978.
" Preliminary Data Report; Test IX, Simultaneous Mode: Cables, Splice Assemblies, and Electrical Insulation Samples," F. V. Thome, SAND 78-0718, Sandia Laboratories, blbuquerque, NM, April 1978.
"A Proposed Method for Combined Environment Accelerated Aging," K. T. Gillen, SAND 78-0501A, Sandia Labora-tories, Albuquerque, NM, October 1978.
" Proposed Research on Class 1 Components to Test a General Approach to Accelerated Aging Under Combined Stress Environments," K. T. Gillen, E. A. Salazar, and C. W. Frank, SAND 76-0715, Sandia laboratories, April 1977.
" Qualification Issues: The Rest of the Iceberg," L. L. Bonzon, R. E. Luna, and S. P. Carfagno, SAND 78-0350, Sandia Laboratories, Albuquerque, NM, n.d.
" Qualification of Class 1E Equipment: The Role of the Utility and Architect-Engineer," 5. G. Kasturi, G. T. Dowd, and L. L. Bonzon, SAND 78-0347, Sandia Labo.atories, Albuquerque, NM, n.d.
Qualification Testing Evaluation Program, Quarterly Report, October-December 1977, SAND 78-0341, Sandia Labora-tories, Albuquerque, NM, April 1978.
Qualification Testing Evaluation Program, Light Water Reactor Safety Research, Quarterly Report, January-March 1978, L. L. Bonzon, K. T. Gillen, and F. V. Thome SAND 78-0799, NUREG/CR-0276 Sandia Laboratories, August 1978.
Qualification Testing Evaluation Program, Quarterly Report, April-June 1978, L. L. Bonzon, K. T. Gillen, L. H. Jones, and E. A. Sala'ar, SAND 78-1452, NUREG/CR-0401, Sandia Laboratories, Albuquerque, MM, November 1978.
Qualification Testing Evaluation Program, Quarterly Report, Jdy-September 1978, L. L. Bonzon, K. i. Gillen, and D. h. Dugan, SAND 78-2254, NUREG/CR-0696, Sandia Laboratories, Albuquerque, NM, March 1979.
Qualification Testing Evaluation, Quarterly Report, October-December 1978, L. L. Bonzon, K. T. Gillen, and E. A. Salazar, SAND 79-0761, NUREG/CR-0813, Sandia Laboratories, Albuquerque, NM, June 1979.
" Radiation Signature Following the Hypo',hesized LOCA," L. L. Bonzon, SAND 76-0740, Sandia Laboratories, Albuquerque, NM, Revised October 1977.
" Single and Combined Environment Accelerated Aging of Electric Cable Insulation and Jacketing Materials,"
K. T. Gillen, and E. A. Salazar, SAND 78-0559A, 1978 Conference on Electrical Insulation and Dielectric Phenomena, Pocono Manor, PA, October 1978.
" Status of the Qualification Testing Evaluation (QTE) Program," L. L. Bonzon, SAND 78-1884C, USNRC Sixth Water Reactor Safety Research Information Meeting, November 1978.
"A Study of Strong Synergism in Polymer Degradation," R. L. Clough, K. T. Gillen, and E. A. Salazar, SAND 79-0924C, Sandia Laboratories, NM, April 1979.
" Synergistic Effects and Source Term Considerations Associated with Class lE LOCA Qualification Testing,"
L. L. Bonzon, SAND 77-1713C, USNRC Fif th Water Reactor Safety Research Information Meeting, Novemtar 1977.
VALVE STUDIES
_ Study of Safety Relief Valve Operation Under AIWS Conditions, E. 5. Hutmacher, B. J. Nesmith, J. B. Brukiewa, NUREG/CR-0687,(LIEC-TDR-78-19), March 1979.
1169 182-
74 HUMAN FACTOR STUDIES Criter a for Safety-Related Nuclear Plant ODerator Actions: A PrelSinary Assessrrent of Available Data, P. M. Haas and 1. F. Bott, NUREG/CR-0901 (ORNL/NUREG/lM-330), July 1979.
NOISE STUDIES
" Bandwidth Related Errors in the Interference of PWR Barrel Motion f rom Ex-Core Neutron Detection Signals,"
J. C. Robinson and R. C. Kryter, Trans. ANS, 24, pp. 413-415, (November 15-19, 1976).
" Calculation of Scale Factor for Interference of Pk'R Core Barrel Motion from Neutron Noise Spectral Density,"
J. C. Robinson, F. Shahrokhi, and R. C. Kryter, Nucl. Tech., 40, (1), pp. 35-46, (Aug. 1978).
Characteristics and Performance Experience of Loose-Part Monitoring Systems in U.S. Commercial Power Reactors, R. C. Kryter and C. W. Ricker, NUREG/CR-0524, December 1978.
" Characterization Studies of BWR-4 Neutron Noise Analysis Spectra," M. V. Mathis, C. M. Smith, D. N. Fry, and M. L. Dailey, Progress in Nuclear Energy, 1, (2-4), pp. 175-181,(1977).
" Determination of Core Barrel Motion from Neutron Noise Spectral Density Data-Scale Factor," J. C. Robinson and F. Shahrokhi, Trans. ANS, 23:458, (June 1976).
" Detection of Impacts of Instrument Tubes Against Channel Boxes in BWR-4s Using Neutron Noise Analysis,"
J. E. Mott, J. C. Robinson, D. N. Fry, and M. P. Brackin, Trans. American Nuclear Society, 23:465 (1976).
" Determination of Void Fraction Profile in a BWR Channel Using Neutron Noise Analysis," M. A. Atta, D. N. F ry,
and J. E. Mott, Nucl. Sci. and Eng. 66 (2), 264-268, (May 1978).
" Determination of Void-Fraction in BWRs Using Neutron Noise Analysis," M. A. Atta, J. E. Mott, D. N. Fry, T_rans. ANS, 23:466, (June 1976).
"The Effect of Multiple Boiling Chaqnels on Void Fluctuation Spectra in BWRs," F. J. Sweeney, Trans. ANS, 30:743-745, (November 1978).
Interference of Core Barrel Motion From Neutrnn Noise Special Density.
J. C. Robinson, F. Shahrokhi, and R. C. Kryter, USNRC Report ORNL/NUREG/IM-100, April 1977.
" Loose-Parts Monitoring: Present Status of Technology Its Implementation in U.S. Reactors," R. C. Kryter C. W. Ricker and J. E. Jones, Progress in Nucl. Energy, 1, (2-4), 667-671, (September 1977).
" Method for Detecting Bypass Coolant Boiling in BWRs," D. N. F ry, et al., Trans ANS, 30:511 (1978).
Noise Diagnostics for Safety Assessment, Quarterly Progress Report for January-March 1977, R. C. Kryter, U5NRC Report ORNL/NUREG.'IM-ll2, June 1977.
Noise Diagnostics for Safety Assessment, Quarterly Progress Report for April-June 1977, R. C. Kryter, USNRC Report ORNL/NUREG/TM-143, September 1977.
Noise Diagnostics for Saf ety Assessment, Quarterly Progress Report for July-September 197, R. C. Kryter and K. R.
Piety, USNRC Report ORNL/NUREG/IM-161, November 1977.
Noise Diagnostics for Safety Assessment, Quarterly Progress Report for October-December 1977, R. C. Kryter, et al., USNRC Report ORNL/NUREG/IM-176, February 1976.
Noise Diagnostics for Safety Assessment, Standards, and Regu.ation, Quarterly Progress Report for January-March 1978, D. N.
Fry, R. C. Kryter, et al., USNRC Report NUREG/CR-0145 (ORNL/NUREG/TM-207), Jule 1978.
Noise Diagnostics for Safety Assessment, Standards, and Regulation, Quarterly Progress Report for April-June 1978, D. N. Fry, R. C. Kryter, et al., USNRC Report NUREG/CR-0525, December 1978.
" Qualification of Core Barrel Motion Using Analytically Derived Scale Factor and Statistical Reactor Noise Descriptors," J. C. Robinson, F. Shahrokhi, and R. C. Kryter, Tech. Note in Nucl. Technology, 40, (1), 47-51, (August 1978).
"Second Specialists Meeting on Reactor Noise," R. S. Booth, Nuclear Safety, ], (4),111 B-1, (September-October 1978).
Summary of ORNL Investigation of In-Core Vibrations in BWR-45, D. N. F ry, R. C. Kryter, M. V. Mathis, J. E. Mott, and J. C. Robinson, USNRC Report ORNL/NUREG/TM-101, May 1977.
"Use of Neutron Noise to Detect BWR-4 In-Core Instrument Tube Vibrations and Itrpacting," D. N. Fry, et al.,
Nuclear Technology, g, (1), 42-54, (April 1979).
"U.S. Experience with In-Service Monitering of Core Barrel Motion in PWRs Using Ex-Core Neutron Detectors,"
R. C.
Kryter, J. C. Robinson, and J. A. Thie, Proceedings of the Internati,nal Conference on Vibration in Nuclear Plants, May 9-11, 1978, Keswick, England.
ll69 185 -
c
75 NUCLEAR SAFETY INFORMATION CENTER PUBLICATION 5*
" Analytical Techniques for Stress Analysis of tLe Nuclear Steam Supply System," ORNL/NUREG/NSIC-157.
" Annotated Bibliography on Common Mode / Common Cause Failure," ORNL/NUREG/NSIC-148.
" Annotated Bibliography of Licensee Event Reports in Boiling-Water Nuclear Power Plants as Reported in 1977,"
ORNL/NUREG/NSIC-149.
" Annotated Bibliograpny of Licensee Event Reports in Boiling-Water Nuclear Power Plants as Reported in 1978,"
ORNL/NUREG/NSIC-164.
" Annotated Bibliography of Licensee Event Reports in Pressurized-Water Nuclear Power Plants as Reported in 1977," ORNL/NUREG/NSIC-150.
" Annotated Bibliography of Licensee Event Reports in Pressurized-Water Nuclear Power Plants as Reported in 1978," ORNL/NUREG/NSIC-165.
" Annotated Bibliography on the Safeguards Against Proliferation of Nuclear Materials," ORNL/NUREG/N51C-160.
" Bibliography of Microfiched Foreign Reports Distributed Under the NRC Reactor Safety Research Foreign Technical Exchange Program 1975-77," ORNL/NUREG/NSIC-154.
" Bibliography of Reports on Research Sponsored by the Of fice of Nuclear Regulatory Research, January-June 1977, ORNL/NUREG/NSIC-143.
" Bibliography of Reports on Research Sponsored by the Office of Nuclear Regulatory Research, January-June 1978," ORNL/NUREG/N51C-155.
" Bibliography of Reports on Research Sponsored by the Office of Nuclear Regulatory Research, July-December 1977," ORNL/NUREG/NSIC-145.
" Bibliography oi Reports on Research Sponsored by the Office of Nuclear Regulatory Resea ch, July-December 1978,"
ORNL/NUREG/NSIC-158.
" Breeder Reactor Safety: Review of Current Issues and Bibliography of u terature," ORNL/NUREG/NSIC-151.
" Breeder Reactor Safety: Review of Current Issues and Bibliography of Literature,1978," CRNL/NUREG/NSIC-166.
"HIGR Safety: Review of Current Issues and Bibliography of Literature," ORNL/NUREG/NSIC-128.
" Indexed Bibliograpny of Foreign Reports deceived Under the NRC Reactor Safety Research rechnical Exchange Program Through 1978," ORNL/NUREG/NSIC-163.
"Index to NUCLEAR SAFETY, A Technical Progress Review by Chronology, Permuted by Title and Author, Vol. II (No.I) Through Vol. 18 (No. 6)," CRNL/NUREG/NSIC-147.
"Index to NUCLEAR SAFETY, A Technical Prngress Review by Chronology, Permuted by Title and Author, Vol. !!
(No.1) Through Vol. 19 (No. 6)," ORNL/NUREG/162.
" Reactor Operating Experiences, 1975-1977, U.S. Nuclear Regulatory Commission," ORNL/NUREG/NSIC-144.
" Reports Distributed Under the NRC Light-Water Reactor Safety Research Foreign Technkal Exchange Program, Vol. III (Januiry-June 1977)," ORNL/NUREG/NSIC-142.
" Reports Distributed Under the NRC Reactor Safety Research Foreign Technical Exchange Program, Vol. IV (July-December 1971)," ORNL/NUREG/NSIC-146.
" Reports Distributed Under the NRC Reactor Safety Research Foreign Technical Exchange Program, Vol. V, January-June 1978," ORNL/NUREG/NSIC-156.
" Reports Distributed Under the NRC Reactor Safety Research Foreign Technical Exchange Program, Vol VI, July-December 1976," ORNL/NUREG/NSIC-159.
ThislistconbsonlyrecentNSICpublications. A complete list may be obta' red from the Nuclear Safety Information Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37830.
I169 134
U" U.S. NUCLE AR REGUL ATORY COMMISSION (7 77)
BIBLIOGRAPHIC DATA SHEET NUREG-0581 4 TITLE AND SUBTIT LE (Add volume bio.. of.wpecproate) 2 (Leave bian ki Summary of NRC LWR Safety Research Programs on a recipient S ACCESSION NO' Fuel '>ehavior, Metallurgy & Materials, and Operational Safety
- 7. AU THOH(S)
- 5. D A TE RE PORT COMPLE TE D l vt AR Gary L. Bennett v os m 9 PE RFOP%8%G ORGANIZATION N AME AND MAILING ADDRESS (Include Iro Code /
DATE REPORT ISSUED Research Support Branch August 1979 Division of Reactor Safety Research g
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Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission a < tea,ossoni
- 12. SPONSORINf3 ORG ANIZ ATION N AYE AND M AILING ADD RE SS //nclude lip Code /
10 PROJECT:T ASK WORK UNIT NO Same as above 11 CONTRACT NO None 13 TYPE OF RE PORT PE RI O D COV E RE D //"clus ve dafF51 Co'iference paper 15 SUPPLEMEN T ARY NOTE S 14 (te n e n, m /
American Nuclear Society Annual Meeting, Atlanta, Georgia
- 16. At;ST R ACT (200 words or lessl The NRC light-wacer reactor safety-research program is part of the NRC regulatory program for ensuring the safety of nuclear power plants. This paper summarizes the results of NRC-sponsored research into fuel behavior, metallurgy and materials, and operational safety. The fuel behavior research program provides a detailed understanding of the response of nuclear fuel assemblies to postulated off-normal or accident conditions.
Fuel behavior research includes studies of basic fuel rod properties, in-reactor tests, computer code development, fission product release and fuel meltdown. The metallurgy and materials resea Oh program provides independent ccnfirmation of the safe design of reactor vessels and piping. This program includes studies on fracture mechanics, irradiation embrittlement, stress corrosion, crack growth, and nondestructive examination. The operational safety research provides direct assistance to NRC officials concerned with the operational and operational-safety aspects of nuclear nower plants. The topics currently being cidressed include qualification testing evaluation, fire protection, human factors, and noise diagnostics.
- 17. NE Y WORDS AND DOCUVE NT AN ALYSiS i la CE SC H iP T ORS Fuel Behavior Primary system integrity LWR safety research LOCA Vessel integrity Human factors Nondestructive examination Flaw detection Fire Protection Safety margin Qualification testing evaluation Operational safety Noise diagnostics Fuci behavior research 17b IDE N TIF if R$ OPE N E N CE D TE R MS 10 AV AIL ABILITY S T ATE VENT M SE C ou i T Y C L A SS ' T * >
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