ML19254E966
| ML19254E966 | |
| Person / Time | |
|---|---|
| Issue date: | 10/31/1979 |
| From: | Levine S NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19254E962 | List: |
| References | |
| NUDOCS 7911050162 | |
| Download: ML19254E966 (36) | |
Text
.
MEMORANDUM FOR:
Haorld R. Denton, Director Office of Nuclear Reactor Regulation FROM:
Saul Levine, Direc tor Office of Nuclear Regulatory Research
SUBJECT:
RESEARCH INFORMATION LETTER - #
LOFT PEACTOR SAFETY PROGRAM RESEARCH RESULTS FROM NUCLEAR LOSS-0F-COOLANT EXPERIMENTS L2-2 AND L2-3
1.0 INTRODUCTION
This Rese.rch Information Letter transmits the significant results that have bt:en obtained from the LOFT Reactor Safety Research Program from October 1,1978 through June 1,1979.
During this time two nuclear loss-of-coolant experiments, LOCE L2-2 and LOCE L2-3, were conducted successfully.
ihe LOFT f acility (1) consists of a complete operational pressurized water reactor (PWR) designed to operate over the range of power densities of comnercial PWR's and to simulate loss-of-coolant accidents (LOCA) from pipe ruptures.
The LOFT system and core configurations are shown in Figures 1 and 2.
The LOFT Research Program (2, 3) has been developed to provide experimental information relevant to the licensing criteria for large commercial PWRs.
The major portion of this program is directed at an improved understanding of the LOCA and the performance of emergency core cooling systems using therrral-hydraulic, core physics, structural and fuel behavior data obtained through loss-of-coolant experiments (LOCEs).
This letter is based on data obtained from the first two nuclear LOCEs conducted in the LOFT f acility.
The two LOCEs are part of the L2 experiment series in which the effect' of a complete offset shear of a primary coolant pipe are studied.
LOCE L2-2 was conducted at a power density of 26.4 + 2 KW/m which is approximately 2/3 nominal for commercial PWRs.
LOCE L2-3 wcs conducted at PWR nominal power density
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of 39.0 + 3.0 KW/m.
A description of the system configuration and initial conditions for these e;;periments is contained in Table I.
The results of LOCEs L2-2 and L2-3 are related to LOCA behavior in large commercial PWRs through the predictive modeling techniques developed to analyze the LOFT LOCEs.
Specifically, the LOFT modeling techniques are applied to the ZION PWR; and predictions of the ZION PWR LOCA behavior, with L2-2 and L2-3 initial conditions, are compared to the results of LOFT L2-2 and L2-3.
In addition to obtaining predictions of coninercial PWR LOCA behavior, the volumetric scaling rationale used to scale LOFT to a commercial PWR (and to scale the Semiscale facility to LOFT) is evaluated alorg with verification of the results presented in a previous letter (4) on LOFT nonnuclear LGTEs.
The LOFT program objectives and scaling rationale are briefly described for reference. The experimental results of LOCEs L2-2 and L2-3 are presented with relevance to commercial PWRs and the licensing criteria through application of the analytical models.
Finally, the forthcoming LOFT experimental program is described relative to the results of LOCEs L2-2 and L2-3 and in view of the recent TMI accident.
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2.0
SUMMARY
The results of the LOFT nuclear experiments presented herein support the conservative intent of the applicable parts of the evaluation model requirements contained in the licensing criteria.
The results also confirm the findings of earlier LOFT nonnuclear experiments.
The results and conclusions presented herein are considered applicable to 10 CFR 50.46, subsctions (a) (1), (b) (1), (2), (4), (5), (c) and to Append ix K, parts 1-a-1, 2, 3, 4, 6, 7,1-C inclusive, and 1-d-1, 2, 3, 4, 5.
Review and evaluation of these paragraphs in the light of the thermal-hydraulic phenomena observed in the LOFT nuclear experiments should aid in the evaluation of commercial PWR safety.
The nuclear experiment results presented herein are applicable to considerations involving large area pipe breaks.
Future LOFT experiments will provide additional information on large breaks and also on small area pipe breaks. A new experiment sequence has been tentatively defined as a result of the nuclear experiment results obtained thus f ar and as a result of the recent TMI accident.
The results of the first two LOFT nuclear loss-of-coolant experiments, LOCE L2-2 and LOCE L2-3, which are described in detail in the following section, indicate that the core thermal response was much less severe than anticipated.
The fuel cladding temperature rise was terminated within the first 10 s of the transients by hydraulic phenomena in the primary coolant system.
This cessation of cladding temperature rise to values well below cladding damage thresholds takes place before the actuation of any of the ECC systems.
The hydraulic phenomena are sufficiently dominant to achieve the same effect over the entire range of initial power densities in PWR cores up to the nominal operational value of 39.4 kW/m.
The scaling rationale used to scale LOFT to a commercial PWR and to scale the Semiscale facility to LOFT has been shown to be valid as a result of comparisons of LOFT data with Semiscale counterpart experiment data and as a result of application of identical modeling techniques to a commercial PWR as was used for post-experiment
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analysis of LOFT results.
Comparisons of predicted comnercial PWR thermal-hydraulics with LOFT experimental data show that, for the same initial conditions, the thermal-hydraulics of the loss-of-coolant accident in the two systems are in excellent agreement in both magnitude and time sequencing.
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3.0 BACKGROUND
3.1 LOFT Program Objectives The specific LOFT program objectives are:
1.
Provide integral system experimental data to the U.S. Nuclear Regulatory Commission (NRC) and the nuclear industry for the assessment and development of analytical methods used to predict:
The transient thermal-hydraulic, mechanical, and nuclear response of the reactor system and primary system components under LOCA cond itions.
The capability of current ECCS designs to fulfill their intended func ti on.
The margin of conservatism inherent in the capability of current ECCS designs.
The effectiveness of alternate ECCS concepts.
2.
Investigate thresholds or unexpected phenomena that could affect the validity of the analytical models used to predict the thermal-hydraulic mechanical, and nuclear response of the reactor system.
The first LOFT experiment series, designated the L1 series, was nonnuclear in nature and served to (a), evaluate and verify the structural integrity of the LOFT system; (b), provide data for evaluation and development of analytical models used to predict the hydraulics in response to large pipe breaks; and (c), provide an operational base for the simulation of large pipe breaks in PWR systems at power conditions.
Several nuclear LOCE series have been defined, one of which is the L2 series, to provide integral system experimental data to meet the LOFT program objectives.
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based on the results of two of the L2 series expe iments, LOCE L2-2 and LOCE L2-3, which are the first fully integrated PWR systems loss-of-coolant experiments to be conducted in the world.
3.2 Scaling The LOFT system is designed to " scale" significant features of a four-loop LPWR and to reproducibly simulate typical system transient response to a LOCA. The scaling rationale (5) applied in LOFT makes extensive use of principles that have been applied in a wide ranga of experiments within and beyond the nuclear power industry.
The general scaling rules applied in LOFT are as follows:
Fuel linear heat generation rate is full scale.
The nuclear fuel design has the same 15 x 15 geometry as commercial reactor fuel.
Design variations from the commercial fuel are length (1.7 m),
conservative fuel densification characteristics, and fuel pins not pre-pressurized (specific nuclear experiments with pre-pressurized fuel are planned).
Core power is taken as the basis for scaling of component volumes, that is, LOFT Power LOFT Volume = LPWR Powerx LPWR Volume; Flow areas are scaled to provide similar mass fluxes; The break area-to-actual system volume ratio is set identical to the LPWR value under study; Initial conditions (pressure, temperature, mass flux) are set identical to the LPWR values.
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s LOFT is actually one of three systems that are related by this scaling rationale. The Semiscale f acility (6) is a scale model of LOFT using about the same scale ratios as were used in scaling LOFT to the commercial PWR. The major scaling parameters for LOFT, Semiscale, and the commercial PWR are summarized in Table II, The design and operation of the LOFT f acility ensure that all the significant phenomena occur in approximately the same magnitude and time sequence as would occur in a commercial PWR LOCA.
Assessment of the scaling rationale is accomplished by (a) comparison of LOFT experimental results with results of counterpart experiments conducted in the Semiscale facility and (b) applying the same modeling techniques to LOFT and the commerical PWRs, with LOFT LOCE initial conditions, and evaluating the comparisons.
3.3 LOFT Program Activities The LOFT Program consists of both experimental and analytical phases.
The experimental phase consists of the planning, preparation, and conduct of the LOCEs.
This phase culminates in the acquisition of data and analysis of results that fulfill the objectives of the LOFT Program.
Planning LOFT experiments includes evaluation of Semiscale counterpart experiments and experiments in which parametric variations are made.
The experimental data obtained from LOCEs L2-2 and L2-3 are contained in references 7 and 8, respectively.
Semiscale counterpart experiment data are reported in references 9,10, and 11.
Concurrent with the experimental program, sur ing analysis provides pre-experiment prediction and post-experiment analysis for the purpose of developing and refining code models and identifying areas for additional code development. Thermal-hydraulic analysis of LOCEs L2-2 and L2-3 was carried out principally with the RELAP4/ MOD 6 (12) and RIAPT4 (13) codes by EG&G Idaho and with the TRAC-P1 (14) and TRAC-P1A (15) codes by Los Alamos.
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3.4 LOFT Data Uncertainties In general, the uncertainties in the measured principal variables were as follows:
temperature 13 K 1.0%
p re ssu. e 10.03 MPa 2.2%
differential pressure 10.01 MPa 0.2%
3 density 10.03 Mg/m 3.75%
momentum flux 112.0 Mg/m.s2 20.0%
veloc ity 12.7 m/s 13.5%
Techniques and instruments are well developed for measurements of the first four variables and consequently these measurements are relatively accurate.
However, the fuel cladding temperature measurements can indicate up to 30 K low during transient conditions due to hydraulic influences on the cladding external thermocouple.
The last two variables, momentum flux and velocity, are difficult to measure in two-phase flow conditions. The uncertainties stated for these variables reflect this difficulty and represent the 16rgest uncertainties which occur during low quality fluid conditions, lf'
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4.0 RESULib The results obtained from LOCEs L2-2 and L2-3 represent a significant achievement in the NRC's reactor safety program.
After many years of planning, construction, and experimentation the first experiments have been completed wherein the integrated effects from a loss-of-coolant accident in a fully operational PWR have been evaluated.
The experimental data obtained from those first two experiments have many important implications on the early thinking (16) and beginnings of reactor safety research which formed the basis of the licensing criteria in effect today.
The data also have implications on the expectations of thermal-hydraulic phenomena in commercial systems subjected to similar loss-of-coolant ;ccidents and on the safety margins designed into the systems in accordance with the licensing criteria.
The experimental data, in conjunction with earlier experiments discussed in a previous letter (4), cover the entire operational range of comercial PWRs from hot standby conditions with the reactor shut down to nominal operating conditions with a power density of 39.4 KW/m. All of the experimental data thus far obtained is indicative of the effects resulting from the largest possible pipe break -- the double ended offset shear of a primary coolant pipe.
The results of LOFT LOCEs L2-2 and L2-3, and of LOFT nonnuclear LOCE L1-5, are presented in the succeeding sections with relevance to the scaling rationale and expected commercial PWR thermal-hydraulics and to the quantitative aspects of specific phenomena implicitly or explicitly addressed in the licensing criteria.
4.1 LOFT LOCE Thermal-Hydraulics The thermal-hydraulic transients in experiments L2-2 and L2-3 are quantitatively described in specific detail by the seouence of events given in Table III and by the sumary of phenomena results given in Table IV. The information in Tables III and IV along with the initial conditions defined in Table I provide a clear and detailei description
.of the loss-of-coolant phenomena resulting from a double ended offset shear in the cold leg of a PWR primary coolant piping loop.
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The chronology of events shows very similar behavior between the two nuclear experiments and also very similar behavior between nuclear and nonnuclear experiments for those events not strictly associated with core power. The cessation of fuel rod cladding temperature rise and subsequent core-wide return of fuel rod cladding temperature to fluid saturation temperature within the first 10 s of the transients were the dominant events that influenced the similar sequence characteristics of the nuclear and nonnuclear experiments. Th is thermal phenomena is the result of primary coolant system hydraulic
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phenomena that are dominant and which control and limit the fuel cladding maximum temperatures to well below damage thresholds.
As the chronology shows, this phenomena occurs before any of the ECC systems actuate and is sufficiently dominant to overcome initial core power densities up to the nominal PWR operating value of 39.4 KW/m.
The hydraulic limitation of the fuel cladding thermal response is attr ibu ted to two f actors (17,18).
These f actors are the resumption of the positive (inlet te outlet) core flow pattern at approximately 2.5 seconds and the injection of a significant quantity of high density fluid into the core from the intact loop cold leg af ter the end of the subcooled break flow. Positive core flow resumed af ter the lower plenum fluid reached saturation, which decoupled the broken loop cold leg and core inlet mass flows.
The driving force of the intact loop (pumps operating with little degradation) then reestablished positive core flow. The early resumption of positive core flow significantly diminished the fuel cladding temperature rise throughout the core by 5 seconds.
The end of subcooled break flow in the broken loop cold leg caused saturated choking to occur, which reduced the broken loop cold leg mass flow such that the mass ejected was less than that supplied to the downcomer from the intact loop cold leg.
This condition lasted approximately 2 s as given in Table IV. The subsequent mass addition to the downcomer resulted in a higher density fluid which propagated into and through the core.
The aaded mass flow in the core (higher density fluid) was sufficient to cause the core-wide return of fuel cladding temperature to fluid saturation temperature.
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The factors causing the hydraulic limitation of the fuel cladding thermal response reveal an
'herent limitation in cnalytical models utilizing lumped parameters (volumes).
Asynmetric downcomer fluid flow phenomena must be permitted by the analytical models to allow for the hydraulics observed in the LOFT experiments.
Propagation of such phonomena as density and temperature waves in the downcomer as measured in the LOFT experiments reveals that in lumped parameter models nodalization becomes an important consideration in minimizing the limitation.
The influence of the primary coolant system hydraulics on the fuel cladding thermal response in the first 10 s of the transients resulted in a significant removal of the stored energy in the fuel.
The stored energy in the fuel as a consequence of the nuclear heat source was depleted by 65% in L2-2 and 64% in L2-3.
Af ter 10 s then, the demand for stored energy removal by the ECCS was not sufficiently varied among the experiments to cause significant differences in the phenomena and chronology of events.
The thermal response of the fuel cladding, therefore, produced the highest cladding temperature in the first 10 s of the transients.
However, because of the hydraulic influence, the thermal response in the first 10 s did not result in cladding damage thresholds being reached as shown in Figure 3 for LOCE L2-3 (19).
Recent work involving the removal of the central (hottest) fuel module and installation of a new fuel module in the LOFT core confirms this result. The core internals were clean and undamaged after two nuclear LOCEs (20).
Water chemistry samples have shown no fuel damage or leakage of fission products.
The performance of the ECCS was essentially the same in both nuclear and nonnuclear LOCEs.
The mCC bypass increased slightly with increasing core power density.
At the end of accumulator flow in the nonnuclear experiment '_1-5 the ECC bypass was 30% of the total ECC injec ted up to that time.
The ECC bypass increased to 32% in L2-2 (initial power 26.4 KW/m) and 36% (initial power 39.0 KW/m) in L2-3.
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The core reflood rate was essentially the same in all experiments principally because of the significant removal of stored energy in the fuel early in the transient.
The quenching of the fuel cladding is compared with the core reflood rate for L2-3 in Figure 4.
The cladding in fuel module 5 (Figure 2) did not quench in the hot region until several seconds af ter the flooding level had passed the region.
However, all ladding had quenched prior to complete reflood of the core.
All investigators who have analyzed the quench data (21) have reached the conclusion that the presence of the LOFT thermocouples did not significantly influence the cladding temperature response.
Experiments in the LOFT test support facility (LTSF), PBF, Columbia University, and the German REBEKA f acility have led to the conclusion that the LOFT external thermocouples do not significantly perturb the cladding surf ace temperature.
The LTSF results indicate that the external thermocouples agree with imbedded thermocouples to within 30 K.
Independently, analysis with the FRAP-T4 and other codes has shown through fuel rod stored energy correlations that the LOFT Mternal thermocouples are indicating cladding surf ace temperature and not a locally perturbed surface temperature from a fin effect.
Further experimentation in this area in these and other f acilities is being corried out to verify the thermocouple effects and to determine measurement uncertainties.
ECCS in conjunction with the hydraulic phenomena in the primary coolant system prevented complete depletion of fluid mass in the reac tor vessel during the transients.
Calculations of mass inventory in the reactor vessel as a function of time showed that reactor vessel fluid mass does not deplete to less than approximately 40% o# e ax imum at any time.
The lower plenum did not void at any time durm; the transient.
With the completion of LOCE L2-3, the entire range of initial conditions had l'een covered for determination of the pressure as a f unction of time resulting from the largest pipe breaks possible in PWR systems.
The system pressure data for LOCEs L1-5 (isothermal),
L2-2 (22.7 K core AT), and L2-3 (32.3 K core AT) are shown in a
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With reference to the chronology of events in Table III, subcooled blowdown lasts longest for the isothermal LOCE L1-5, 0.1 s, as expec ted. Thus, subcooled blowdown is less severe on the structural components as the core AT increases.
Also, the data out to 0.2 s were taken at a bandwidth of 1000 Hz to ensure measurement of all significant frequencies.
Analysis of the subcooled blowdown pressure data reveals negligible energy content at frequencies above 40 Hz. The pressure data indicate that at high temperature and pressures where water density is only 2/3 maximum the attenuation of high frequencj pressure waves is very high. The LOFT system was de. signed using WHAM 6 code subcooled blowdown predictions as forcing functions for structural analysis codes.
The requirement; imposed on the WHAM 6 calculations were an isothermal system condition, a very rapid 1.0 ms break opening time, and no pressure wave attenuation.
As
. result the 1.0FT system has been shown to be str ucturally sound by measurements of strain and accelerat'an.
Commercial PWR systems have been designed with similar conservatisms; thus, a flaw that ultimately results in a pipe break in a commercial PWR is not expected to result in failure of other components nor alteration of core geometry.
4.2 Commercial PWR Thermal-Hydraulics The past months since LOCE L2-2 was conducted have resulted in several advances in computer codes and system models. The hydraulics had not been :;redicted correctly for LOCE L2-2. principally because of a lack of knowledge for defining the values of certain input parameters such as the critical flow transition quality and multipliers for the Henry-FdJske and homogeneous equilibrium models.
Post-experiment analyses have shown much improvement ar.d now best estimate predictions of system hydraulics agree very well nith L2-2 and L2-3 hydraulics.
Also, improvements in core thermal response predictions are being made with current analyses showing predictions of early cladding rewet.
The distinguishing feature of all licensing code predictions for LOFT LOCEs is that they predicted much higher cladding temperatures than were reached in the experiments.
Thus, all such predictions or calculations for commercial PWRs in the licensing process cont ained 33 3,n?
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large margins of safety for core fuel integrity (for large pipe b reak s). Therefore, all such calculations have been in keeping with the intent of the licensing criteria.
The post-experiment best estimate analysis of LOFT L2-2 and L2-3 has shown very good agreement with the experimental data as stated p revious ly.
The analysis has been done with the RELAP4/M006 code.
This code and the LOFT modeling techniques have been applied to the ZION commercial PWR.
Predictions of system thermal-hydraulics were made using LOFT L2-2 and L2-3 initial conditions.
Mass flowrate results were scaled on a system volume basis to LOFT for comparison.
Comparisons of the most significant thermal-hydraulic phenomena in LOFT and ZION are shown in Figure 6 for L2-2 initial conditions and in Figure 7 for L2-3 initial conditions.
The comparisons show a high degree of similarity.
The only area where LOFT data and ZION predictions differ appreciably is in the core thermal response.
The prediction shows that the cladding in the ZION core does not return to fluid saturation conditions within the first 10 s of the transient with L2-3 in itial conditions.
However, turnover does occur and the maximum cladding temperature is still reached within the first 10 s.
The LOFT core mass flux is one-half that in a commercial PWR in order to obtain the same fluid temperature rise through the core.
The mass flow / system volume in the LOFT intact loop is equal to that in the commercial PWR during initial conditions.
In a LOFT loss-of-coolant transient the mass flow / system volune is preserved during
.ie early part of the transient whereas in a commercial PWR one-fourth of the mass flow is lost in the broken loop.
The mass flow / system volume is one-third higher in LOFT during the early part of the transient.
Thus, the question arises...are or are not the LOFT experiment results a conservative indication of commercial PWR core thermal response?
The post-experiment analysis of the LOFT experiments with the RELAP4/M006 code showed that the predictions were conservative in that higher cladding temperatures were calculated and the core-wide return to fluid saturation temperature was not as rapid nor did the cladding in the hottest region return to fluid saturation temperature.
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Therefore, the comparisons shown in Figures 6 and 7 indicate that the LOFT results are a conservative indication of commercial PWR core thermal response.
Another view of the comparisons shown in Figures 6 and 7 is that they indicate that the LOFT thermal-hydraulics are quantitatively well in agreement with the thermal-hydraulics of a much larger PWR system.
Th is is, in essence, a verification of the scaling rationale and exemplifies the importance of the LOFT f acility in providing much needed information on reactor safety issues.
A further nore detailed investigation into one aspect of the scaling rationale was also done. The ZION steam generators were " redesigned" as defined in Table V to change the physical properties for haat transfer. The largest change was to use the LOFT steam generator tube de sign. The results of the changes to the ZION steam generaters on the system thermal hydraulics were negligible.
The hydraulics were essentially unchanged and the core thermal response showed a decrease of about 50 K for the largest change involving the LOFT steam generator tube design.
Thus, the steam generator shows only a minor influence on system thermal-hydraulics for large pipe break cases.
Scaled research facilities such as LOFT do not need to consider detailed aspects of steam generator scaling for large pipe break loss-of-coolant investigations. The details of the ZION calculations are being prepared for publication.
4.3 LOFT-Semiscale Comparisons The Semiscale counterpart experiments (9,10) to LOFT L2-2 and L2-3 show the same basic hydraulic phenomena on approximately the same timing sequence.
The various overrides to the scaling rationale imposed on the Semiscale design such as preservation of core and downcomer lengths appear-to produce cnly magnitude variations in the phenomena which can be classified as second or higher order type in fluences.
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comparisons.
The Semiscale facility has been and will continue to be valuable in the planning of the LOFT experiments and in the study of thermal-hydraulic phenomena in PWR systems.
Preservation of the LOFT core length in the Semiscale f acility resulted in the Semiscale core and downcomer being predominantly one-dimensional compared to the LOFT core and downcomer.
This led, in turn, to a decrease in the asynmetric behavior of the hydraulic phenomena in the downcomer.
In consequence, although the system hydraulic phenomena were fundamentally the same as in LOFT, there were some differences in magnitudes and timing sequence.
These small differences, when accounted for in the modelling of the Semiscale counterpart experiments showed that the thermal response of the Semiscale heater rod core was very sensitive to the system hydrau lic s.
This is consistent with the high sensitivity exhibited by the LOFT core thermal response to the system hydraulics.
The Semiscale heater rod surf ace did not show the rapid return to fluid saturation temperature within 10s as did the LOFT fuel cl adding.
Instead, the heater rod surf aces showed a thermal response similar to the LOFT experiment predictions. The prescribed power history for the Semiscale heater rods in the counterpart experiments was determined from the Semiscale system hydraulics and the predictions of the LOFT fuel cladding thermal response.
Using the actual hydraulics measured in the LOFT experiments with same heater rod power profiles resulted in a predicted Semiscale core thermal response that was in excellent agreement with the measured LOFT core thermal response.
The Semiscale-LOFT analyses have shown that Semiscale can simulate LOFT results and that core thermal response is highly sensitive to hydraulic phenomena.
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5.0 RECOMMENDAT IONS The results of the LOFT experiments described in this letter are recommended for use by NRR in its interpretation and application of LOCA ECCS evaluation model criteria and related codes.
The results are applicable in those considerations involving large breaks in the primary piping coolant loops of PWRs.
A task is currently underway to translate the LOFT experiment results into recommendations for licensing criteria revisions. This task could result in a more complete understanding of the LOFT results and provide the proper methodology for NRR and the industry to implement new knowledge of reac tor safety.
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6.0 FUTURE PROGRAM The results of the first two nuclear LOFT experiments have altered the importance of the remaining three experiments in the L2 large break series. The importance of the remaining three experiments is in the evaluation of the core thermal response dependencies on increased power density (52.5 KW/m), loss of offsite powe* (primary pump coastdown at LOCA initiation and delay of HPIS and LPIS), and prepressurized fuel. Ar.31ysis of the results of the first two nuclear experiments, Semiscale experiments, and code predictions has led to the expected results of LOFT experiments L2-4, L2-5, and L2-6 as shown in Table VI. Differences from L2-3 results are not expected to be significant except in the case of L2-5 where the cladding thermal response is expected to reach a maximum at a much later time.
However, the maximum cladding temperature is not expected to be significantly larger than that in L2-3.
The experiments remain important in + hat definition of those parametric dependencies can provide additional knowledge of the range of safe operation of PWRs.
With the advent of the TMI accident, the importance of LOFT experime.ntation shif ted to the small break experiments.
The existing testing sequence was revised as a consequence of the shif t in importance and also as a consequence of the large break results.
A new expericent sequence has tentatively been defined as shown in Table VII which moves up the small break experiments and includes the remaining L2 large break experims:nts in nonsequential order.
The experiment sequence shown in Table VII begins early in FY-80 and extends to approximately mid FY-83.
Information on small area pipe breaks should be available early in 1980.
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7.0 C0 ORDINATION CONTACT For coordination of any further evaluation of these results and for discussion and future experiments, contact Dr. G. Donald McPherson LOFT Program Manager, RES, Telephone 427-4437.
Saul Levine, Director Office of Nuclear Regulatory Research 3qii
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Enclosures:
1.
Figure 1 LOFT System Configuration 2.
Figure 2 LOFT Core Configuration 3.
Table I System Configuration and Initial Conditions for Nuclear LOCEs L2-2 and L2-3 4.
Table II LOFT - Semiscale - LPWR Scaling Parameters 5.
Table III Chronology of Events for Nuclear LOCEs L2-2 and L2-3 With Nonnuclear LOCE L1-5 Comparative Values 6.
Table IV Summary of LOFT Nuclear LOCE Results 7.
Figure 3 Temperature-Pressure History of Fuel Rod Experiencing Peak Cladding Temperature During LOCE L2-3 Compared With Modes of Cladding Deformation (Ref.19) 8.
Figure 4 Fuel Cladding Quench in the LOFT Core Compared With the Reflood Rate in LOCE L2-3 9.
Figure 5 Pressure as a Function of Time for LOCEs L1-5, L2-2, and L2-3 10.
Figure 6 LOFT Data and Zion Prediction Comparisons for LOCE L2-2 Initial Conditions 11.
Figure 7 LOFT Data and Zion Prediction Comparison for LOCE L2-3 Initial Conditions 12.
Table V ZION Steam Generator Parametric Variations for Thermal-Hydraulic Eff ec ts Analysis 13.
Table VI Expected Results of Remaining L2 LOCEs 14.
Table VII Tentative LOFT Experiment Sequence Beginning in FY-80
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o @..,
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. 3....p 3,
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,~................~,......w.......<
~ p.n.=.. 5.~,.;.~a.~.C.: k".:.:..T.V =.M..w" V.;~ : " ; ;.* ; ; :. ; 7./
7 7 ~ ~'
.......o
. n
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- 3.,. 3.
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43.
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,... 3..... i p,..... p...........,3.. 4./-
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'*,M.,
Intacl loop
- k k
intact 100p hOf leg
,,)
Cold leg tPsEL-A-6901 Figure 2.
LOFT Core Configuration T
9fe
'b )
r J
TABLE I SYSTEM CONFIGURATION AND INITIAL CONDITIONS FOR NUCLEAR LOCES L2-2 and L2-3 Parameter LOCE L2-2 LOCE L2-3 Pipe brean:
Locatitri cold leg cold leg Size 200%
200%
Opening time (ms) 17 17.2 Primary system pump operation Powered to TO + 200 s Power to To + 200 s Broken 1000 pump simulator
- Operating pump K = 9.95 Operating pump K = 9.95 Intact loop resistance Low resistance K = 131.7 Low resistanu K = 131.7 ECCSS HPIS, LP!S and HPIS, LPIS, and accumulator accumulator ECC injection location Intact loop cold leg Intact loop cold leg ECC actuation mode:
Accumulator Pressure Pressure l
LPIS Pressure-level Pressure-level i
HPIS Pressure-level Pressure-level Steam generator secondary:
Pressure (MPa) 6.35 + 0.08 6.18 + 0.08 Flow rate (kg/s) 12.76 + 0.40 19.510.4 Primary system:
Pressure (MPa) 15.64 + 0.03 15.06 + 0.03 Temperature (K):
Hot leg 580.4 + 3 592.9 + 1.8 Cold leg 557.7 7 3 560.7 7 1.8 Core power (MW) 24.9 + 1.0 36.0 + 1.0 MLHGR (kW/m) 26.4 7 2 39.0 7 3.0 Mass flow (kg/s) 194.2 + 16 199 + 6.3 Beration (ppm) 838 + T 679I4 ECCS accumulator:
Pressure (MPa) 4.11 + 0.05 4.18 + 0.05 Temperature (K) 300.8'+ 3 307.8 + 3 Boration (ppm) 32'". + 19 3281 + 17 3
Injected voluge (m )
1.68 7 0.03 1.71 7 0.03 Gas volume (m )
1.05I0.03 0.9610.03 Darcy K f actor based on 0.01C r.' f eow area.
w
[ #
)*
.L s
M TA8LE II LOFT - SEMISCALE - LPWR SCALING PARAMETERS Semiscale LOFT LPWR Volmes 3
Total PCS (m )
0.23 7.80 347 Reactor Vessel (% of PCS) 37 34 38 Intact Loop (% of PCS) 44 47 51 Broken Loop (% of PCS) 19 19 11 Power (MW) 1.6 50 3400 Length of Active Core (m) 1.67 and 3.66 1.67 3.66 R atios 3
Volume / Power (m /MW) 0.14 0.16 0.10 Break Area /PCS Volume (m-1) 0.0026 0.0026 0.0026 PWR Volume /Volune 1530 44 1
TABLE !!!
CHRONOLOGY OF EVENTS FOR NUCLEAR LOCE L2-2 AND L2-3 WITH NONNUCLEAR LOCE L1-5 COMPARATIVE VALUES Time After LOCE initiation (s)
Event LOCE L2-3 LOCE L2-2 LOCE L1-5 LOCE initiated 0
0 0
Subcooled blowdown ended a 0.06 0.07 0.1 Reactor scram signal received 0.103 0.085 0.087 at control room Earliest departure of cladding temperature 0.96 1.0 25.6 from fluid saturation temperature (Tclad > Tsat)
Controlrodscompletelyingerted 1.683 1.725 1.85 Subcooled break flew ended 3.0 3.8 0.1 Maximum cladding temperature 4.95 5.8 steady state attained value at time O Earliest core-wide return of cladding 8.5 8.0 48 temperature to fluid saturation temperature HPIS injection initiated 14 12 13 Pressur12er emptied 14 15 14 Accumulater injection initiated 16 18 19 LPIS injection initiated 29 29 34 Lewer plenum filled with liquid 35 35 37 Saturated blowdown ended 40 44 47 Accumulator liquid flow ended 45 49 54 Core volume reflooded 55 55 59 a.
End of subcooled blowdown is defined as the occurrence of the first phase transition in the system other ttan at the pipe break location.
b.
End of subcooled break flow.is defined as the completion of subcooled fluid discharge from the break (het and Cold legs) in the broken 1000.
I
.,c
TABLE IV SUFt4ARY OF LOFT NUCLEAR LOCE RESULTS Experiment Results LOCE L2-3 LOCE L2-2 Times for cladding temperature to excerJ fluid saturation temperature (s) minimum in hot region 0.94 1.00 maximum in hot region 1.84 2.30 Peak cladding temperature (K) 914 + 3 789 + 3 Core reflood rate (m/s) 0.10 + 0.02 0.12 f 0.02 Minimum mass / volume in reactor vessel (kg/m3) 431 1 75 468 + 75 Accumula tor flow duration (s) 29 31 Maximum accumulator flow / system volume 3
(kg/s/m )
6.42 + 0.45 7.71 1 0.45 Accumulator polytropic gas constant 1.25 1 0.02 1.22 2 0.02 Cladding quench time / core reflood time
<1 for all
<1 for all measurements measurements ECC bypass at end of accumulator flow
(% of total ECC injected) 36 + 4 32 + 3 First 10 s of the transient Duration of primary pump pressure dif f eren tial (s)
O to 9 0 to 8 Mass flow rate / system volume Initial value intact loop cold leg 25.6 + 2.0 (kg/s/m )
24.9 + 2.0 3
Maximum value (t >0) broken loop 96.2 + 14.4 60.8 + 9.1 cold leg (kg/s/m3)
Time interval (5BLCL>EILCL)*
(s) to 3.65 to 3.60 Time interval (*lLCL>EBLC'. )*
(s) 3.65 to 5.71 3.60 to 6.16 Integral 6BLCL (kg/m )
323.6 1 22.7 254.7 + 17.8 3
Integral AILCL (kg/m )
231.6 1 16.2 215.0 1 15.0 3
Of fference in m integrals (kg/m )
92.0 + 27.9 39.7 + 23.3 3
Stored energy removed
(% of nuclear heat source energy) e 64 e 65 LHGR linear heat generation rate BLCL broken loop cold leg iLCL intact loop cold leg.
] 9 / '.
' _} v/'
s 1200 1100 Waisting E
e 1000 3
Collapse 9e E
900 2
Buckle 0
800 8
E LOCE L2-3 8
700 s
O 600 W
I I
I I
I 500 0
2 4
6 8
10 12 14 16 External pressure (MPa)
INEL-A-12 309 Figure 3.
Temperature-Pressure History of Fuel Rod Experiencing Peak Cladding Temperature During LOCE L2-3 Compared with Modes of Cladding Deformation (Ref.19).
l, 2 f s
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MMuei s
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't 25 i
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e.-,-......
A N
- m. m.e...
,m Figure 4.
Fuel Cladding Quench in the LOFT o
Core Compared with the Reflood Rate in LOCE L2-3.
en.o
- LOCE L1-5
~
- LOCE L2-2 87.5 g
i I
~
(
\\
J is.o N
f N NS e
NSN NN\\
~ LOCE L2-3 AVN 1 - (
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-W\\
M I I IIII
'A ]- -
-~
~
o.o o.colo o.oteo n e'.se l** **
tis atra surtunt iseconosi Figure 5.
Pressure as a Function of Time for LOCEs L1-5, L2-2, f 7 and L2-3 s
so
PDE DEK et'f *a fg t t's M
4 OC-PC-409 t DE-BL-10S'eLc-2 fasts a RELaP se 8.0 3 attap,
e?
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=
0.s g,,
r 3,- N ap,p*q[.
f s
s g.g O.S 0
S 10 IS 70
- 5 30 0
5 to 15 20
- S In T I paE AFTER #vpfvat ses f l ot AFTER AvPipt Iss IN!ACT LOOP COLD Ltl DEh517V 880Fim LX* COLD LEG DEtilTV s Fn-ac-10a'eLa-e ersfe sm"-et-ica' eta-r vrste i
900 500 e
e 1
f g
tSe e
eSe E
q 8
8
%%.! +,
t t
0
- NQ a
0 l d h
y 0
5 to 15 to FS 30 0
9 10 IS PJ 26 30 f f eet 4 Fife RWPfvet es*
T 1 *ef artre Ruefvwg aei 14 FACT LNP ColO LiG al455 FL0dATI sa'attu Loor CM LE' 455 fl0mimATE 1
900 x
na-
.l.L.e -4 1t 51
,4.I Eu. pa 5 e-030'ttf-t TEsfi L.
i
.00 g
'9 pWM 8
{
p-~# ;
8 a0e r
7,As,
,, s j
e
.. 0 0
9 60 IS 20 PS 30 0
S so IS 20 c5 is o,1v.c...
,,,ec.,,...v.,,<
DIFFIR!gC[ gf Td ER INTACT LCOP AMD ORCEN LOUP COLD LEG m15 FLDIAATES CLA30!bG ftpfRAftslE 14 THE PEAT P0eER Alfe:M Figure 6.
LOFT Data and Zion Prediction Com-parisons for LOCE L2-2 Initial Conditions.
f 7 ','
4n y.;
P00R DilGIN/L e9 1 DE -pC -
,9 8 L f-3 f( S i t I DE-96-109 IL f-3 1E li t
- #E t ap e 7 attap =
I e i ee 8
E,
- ,.#w* *}
w.
/
1.
l
/
o
=
}
ly.e o
i 3
'*e4. mag,x J 0;;..,./
4 A
e0
" " " ^ '
eo S e s0 0 IS 0
- e e 25 0 30 0 0 e S0 10 0 is a to o 25 0 30 o f i ng erfte auptvet ses f l eet artge nuptuaL ses 1mTACT LOOP COLD LIG EhilTV BaJttu LOC + COLO LEG N mSITY 190 I re PC-see ata-3 ftSta a re-et-sie sta-3 vgnfa
- 'I' # '
See
- 'I
~
50s I
[h, d r
a e eso hk I
(e s
%:# %, ft g A(fla 4
4 t
i I
E q
a
-dSO eo S e io o as e to e e9 e 33 e oo So io 0 iS o ao 0
- 5 0 30 o riac areta auptver...
v =t arren ooprver...
lt f 4C T LOOP COL D LIC 44% FL 0dA f f 880K14 LOOP C3LD LEG 455 FL0dAT[
, Lg.g ggg, e fr-Ss*-030'tLf-3 TESfl 3 agLap =
I #ELA' *
[e j
- Se i000 w
,e t
.+,
b eD e
- f $s gi *.
s
(*4,Y, QTg
/
}
./
- *yF,2
- 8 t
'd
~
4
,,0 i
o
- -~,,
a y
s s -s.
n
-300
'88 0 0 S0 10 0 e5 e 20 0 PS 0 30 0 0 0 S0 10.0 st 0 20 0
- 5 0 30 0 f l at arfan avetvet see
'8"E
'FTE" "UU"E DIFiltEEE BtTiftig luTACT LOOP AND CLADCitlG TE'tf 4Af tset In Te( PLAE taatle s00F COLD LEG MP"5 FL0 DATES Polett REGION Figure 7.
LOFT Data and Zion Prediction Comparisons for LOCE L2-3 Initial Conditions.
pf -1 j j re s
TABLE V ZION STEAM GENERATOR PAR AMETRIC VARIATIONS FOR THERMAL-HYDRAULIC EFFECTS ANALYSIS Steam Generefor Parametric Variation Physical Change Tube material changed from Inconel Thermal conductivity decreased:
to SS 316 at 478 K 10.6%
Tube geometry unchanged at $89 K 13.9%
Heat capacity increased slightly at 478 K negitgible at 589 K 1.3 %
Tube geometry changed:
Heat transfer area reduced:
ID increased from 19.7 mm to 22.2 mm Primary side 13.4%
00 increased from 22.7 mm to 25.7 mm Secondary side 13.2%
Number of tubes decreased f ree 3250 to 2430 Tube material unchanged (Inconel) l l
LOFT tube geometry used:
Heat transfer ares increased:
i 10.2 mm 10 primary side 92.8%
12.7 mm 00 secondary side 112.4%
Number of tubes 12050 Tube material unchanged (Incenel)
I j c) r /
j} 'l
o TABLE VI EXPECTED RESULTS OF REMAINING L2 LOCES Variations in initial conditions Experiment or system configuration relative Expected di'ferences in to LOCE L2-3 results relative to LOCE L2-3 L2-4 Power 33% greater Same hydrau lic phenomena.
Mass flow 25-30% greater. Core Similar fuel cladding tem-fluid temperature differential perature transient with a remains unchanged.
peak value of 1100 K occurring at 5 s and core wide return to fluid saturation by 9 s.
Subsequent clad temperatures lower than peak value during blowdown. Core wide cladding quench by ECC by 60 s (5 s later than L2-3).
L2-5 Pumps tripped at experiment Initial cladding temperature initiation. HPIS and LPIS response will be similar delayed. All initial condi-with the clad temperature tions are unchanged.
at 5 s possibly up to 30 K higher. However, there will not be a core wide return of cladding temperature to fluid saturation temperature in the hot region. The cladding temperature will reduce 100 K by 7 s followed by a gradual increase to the peak value of 950-1050 K by 35 s.
ECC quench of the cladding will be complete by 65 s.
L2-6 All conditions same as in L2-5 Thermal response is expected except pressurized fuel is used.
to be the same as in L2-5.
)2
TABLE VII TENTATIVE LOFT EXPERIMENT SEQUENCE BEGINNING IN FY-80 Experiment Description MLHGR (kW/m)
Mass Figw/ System Volume kg/s/mJ L3-1 Small cold leg break; break (!ow HPIS flow 52.5 61.4 L3-2 Small cold leg breik; break flow HPIS flow 52.5 61.4 L3-3 Small cold leg break; break flow = HPIS flow 52.5 61.4 L6-1 Operational transient; loss of steam load 39.4 61.4 L3-4 Small break; pressurizer power operated relief valve 52.5 61.4 L6-2 Operational transi ent; loss of PCS flow 39.4 61.4 L2-5 Double ended cold leg break; Appendix K assumptions 39.4 24.1 L6-3 Operational transient; excessive load increase 39.4 61.4 LS-1 Intermediate break (break location not yet specified) 39.4 24.1 L6-4 Operatianal transient; control rod withdrawal 39.4 61.4 L5-2 Intermediate break (break location not yet specified) 39.4 24.1 L6-5 Operational +ransient; loss of feedwater 39.4 61.4 L2-4 Double ended cold leg break; LPWR expected conditions 52.5 31.0 L6-6 Operational transient (not yet specified) 39.4 61.4 L2-6 Double ended cold leg break; pressurized fuel 39.4 24.1 L7-1 LOCA with SG tube rupture 39.4 24.1 L7-2 LOCA with SG tube ruptura 39.4 24.1 L3-5 Small break ( size and location not yet specified) 52.5 61.4 L3-6 Small break (size and location not yet spacified) 52.5 61.4 1 7 ' ',
/) [
REFERENCES 1.
D. L. Reeder, LOFT System and Test Description (5.5 Foot Nuclear Core 1 Loss-of-Coolant Experiments), TREE-NUREG-1208 (July 1978).
2.
G D. McPherson, The Purpose of the LOFT Program and Its Application to Licensing Activities, presented at the Institute for Reac tor Safety, Vienna, Austria, September 20, 1978.
3.
D. L. Reeder and V. T. Berta, The Loss-of-Fluid Test (LOFT)
Facility, presented at the 14th Intersociety Energy Conversion Engineering Conference, Boston, Mass., Aug 6-11, 1979.
4 Research Information Letter - #37 LOFT Reactor Safety Program Research Results Through October 1,1978.
5.
L. J. Ybarrondo, S. Fabic, P. Griffith, and G. D. McPherson, "Examiniation of LOFT Scaling," presented at the ASME Winter Annual Meeting, New York, New York (Nov. 17-22,1974).
6.
L. J. Ball, et al, Semiscale Program Description, TREE-NUREG-1210 (May 1978).
7.
M. McCormick-Barger, Experiment Data Report for LOFT Power Ascension Test L2-2, NUREG/CR-0492, TREE-1322, (February 1979).
8.
P. G. Prassinos, B. M. Galusha, and D. B. Engleman, Experiment Data Report for LOFT Power Ascension Experiment L2-3, NUREG/CR-0792, TREE-1326 (July 1979).
C.
M. L. Patton, Jr., B. L. Collins, and K. E. Sackett, Experiment Data Report for Semiscale MOD-1 Test S-06-2 (LOFT Counterpart Test), TREE-NUREG-1122 ( August 1977).
3 q i ',
< p, C,.3 c
10.
B. L. Collins, et al, Experiment Data Report for Semiscale M00-1 Test S-06-3 (LOFT Counterpart Test), NUREG/CR-0251, TREE-1123 (July 1978).
11.
R. L. Gillins, K. E. Sackett, and C. E. Coppin, Experiment Data Report for Semiscale M00-1 Test S-06-4 (LOFT Counterpart Test),
TREE-NUREG-1124 (December 1977 ).
12.
EG&G Idaho, Inc., RELAP4/ MOD 6 -- A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems -- User's Manual, CDAD-TF-003 (January 1978).
13.
L. J. Siefken et al, FRAP-T4 -- A Computer Code for the Transient Analysis of 0xide Fuel Rods, CDAP-TR-78-027 (July 1978).
14.
Los Alamos Scientific Laboratory, TRAC-P1:
An Advanced Best Estirnate Computing Program for PWR LOCA Analysis, LA-NUREG/CR-0063 (June 1978).
15.
Los Alamos Scientific Laboratory, TRAC-P1A:
An Advanced Best Estimate Computing Program f or PWR LOCA Analysis, Vol.1, LA-7777-MS, NUREG/CR-0665 (April 1979).
- 16. LOFT Engineered Safety Systems Investigations, 100-17258 (April 1959).
17.
J. H. Linebarger, D. L. Batt, and V. T. Berta, LOFT Isothermal and Nuclear Experiment Results, presented at the 14th Intersociety Energy Conversion Er J neering Conference, Boston, i
Mass., Aug 6-11, 1979.
18.
D. L. Reeder, Blowdown Hydraulic Influence on Core Thermal Response in LOFT Nuclear Experiment L2-3, presented at the ANS 1979 Winter Meeting, San Francisco, CA., Nov. 11-16, 1979.
19.
C. S. Olsen, Zircaloy Cladding Collapse Under Off-Normal Temperature and Pressure Conditions, TREE-NUREG-1239 (April 1978).
l2'$
4b
20.
M. L. Russel, LOFT Instrumented Fuel Design and Operating Experience, presented at the 14th Intersociety Energy Conversion Engineering Conference, Boston, Mass.. Aug 6-11, 1979.
21.
E. L. Tolman, D. A. Niebruegge, and P. G. Prassinos, Nuclear Fuel Rod Behavior During LOFT Experiment L2-2, International Colloquium On Irradiation Tests for Reactor Safety Programs, Petten, The Netherlands, June 25-28, 1979.
1 7 ' ',
' 4. 7
!