’1 , 3 4456 0356873 H AL LABOR# ROCMI IRDADV al¥ial SHA M 'r..lj'v'A';l - Contract No. W-T40O5-eng-26 Reactor Division ZERO-POWER PHYSICS EXPERIMENTS ON THE MOLTEN-SALT REACTOR EXPERTMENT B. E. Prince J. R. Engel S. J. Ball P. N. Haubenreich T. W. Kerlin FEBRUARY 1968 OAK RIDGE NATTIONAL LABORATORY Oak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the U. S. ATOMIC ENERGY COMMISSION ORNL-423%% MARTIN MARIETTA ENERGY SYSTEMS I|J|BRARIES 3 4456 0356873 4 { f ii% PREFACE This report is a revised and expanded version of an internal memo- randum titled "Preliminary Report on Results of MSRE Zero-Power Experi- " issued in the summer of 1965, immediately following the com- ments, pletion of these experiments. It contains a complete description of all the measurements made of the neutronic behavior of the reactor, before the reactor was operated with substantial nuclear heat generation. Many of the sections in the original internal memorandum have been carried over essentially intact. In addition to these results, however, several other experiments had been performed which could not be properly evalu- ated by the time of the preliminary report writing, and which were of necessity omitted or mentioned only briefly in that writing. Subse- quently, the priority given to analysis of data obtained from power operation of the reactor necessitated a delay in preparing a formal re- port on this work. Rather than issue a separate report pertaining only to these experiments, we feel that a coherent account of all the MSRE zero-power nuclear experiments, in a single report, might be interesting to a wider range of readers as well as provide useful retrospect in later stages of development of the molten-salt reactor program. Both the performance of the zero-power nuclear exXperiments and the writing of this report were a group effort, with the former involving a much larger group than the latter. General areas of responsibility, how- ever, were as follows: The rod calibration experiments and the "static" reactivity coefficient measurement program were planned and analyzed by J. R. Engel and B. E. Prince. Frequency response measurements and ex- periments involving the dynamic effects of temperature and pressure on reactivity were devised by 5. J. Ball and T. W. Kerlin. Flux noise measurements were made by D. N. Fry and D. P. Roux. P. N. Haubenreich supervised the initial critical experiment and coordinated the remainder of the activities. Many members of the reactor operations staff provided indispensable aid in carrying out all of these experiments. The Authors CONTENTS Preface L A L A A N N I I N A N I R N A A N R E R E N E N * & ¢ 8 2 2 & 9 Abstract ® & & 5 2 08 s 000 s ® 4 8 8 & 8 8 s 0 0SS RS L EE YT e s LI A B R R B I 8 o o o 2 s l L] IntrOduC tion LA LA 2 B N R N N N A N A N I A I I A T N R R T 2. Initial Critical EXperiment ..e..oeveeeeeseecsesoonsessennss 5. ContrOl-ROd Calibration o 4 6 o8 28 a L R R R R R Y R N I A N A N T R ] 5.1 General DesSCripPtion suiveeeeeeeeceesaceescenosscacocenss 3.2 Theoretical Guidelines ....ceeveeneas Peecesssssasenanas 3.5 Differential-Worth Measurement: Fuel Stationary ..... 3.4 Reactivity Equivalent of 225U Additions ..e..veee.o... 5.5 Rod-Shadowing EXperiments ..ue.eeeseseececesescenenses 3.6 Reactivity Effects of Fuel Circulation seeeeeeee.ooo.. 5.7 ROA-Drop EXPeriments seu.e.ieeeecevetecoceocennoneness 3.8 Comparison of Measurements with Theoretical Analysis OfcontrOl"'ROdworth ...li...-.l..I..ll........‘...... L. Temperature and Pressure Reactivity Effects «.eeeeeeseee... k.1 Isothermal Temperature Coefficient of Reactivity ..... L.2 Fuel Temperature Coefficient of Reactivity ee.eeveoe... L.,3 Effect of Pressure on ReacCtivVity teeieeececeeecaneenss 5. Dynamics TeStsS seeeecieerannans S eesssecsseeatcetaettaanrane 5¢1l Purpose OFf TeStS cvecrerrerareescatsecscannseeesacnnnens 5.2 Frequency-Response MeaSUrements +.veeeesocsosesvosesos 5.3 Pulse Tests wieeeereevrrrrscersecocscennoosasnsanssacsns 5.4 Pseudorandom Binary Sequence TeStS v.veeeo.. Ceeeeeaaa 5.5 Neutron Fluctuation Measurements ...eeee.. cseeracun .o 5.6 TranSient FlOW-Rate TeStS [ N N A R E N T 5.7 Conclusions from Dynamics TesStS ceeeeeereserssocceeese 6. General Conclusions (IR L B 2 B D B Y B BN I BN Y N RN BN RN B B A A I A A B A I B References .l.....l.l.....l...'...."..I....Il...l.l..'.l..l.Il iii 11 11 12 15 18 19 23 25 31 37 37 41 42 45 45 45 46 47 50 51 54 54 56 Fig. No. 10 11 12 13 1k 15 16 17 18 vii LIST OF FIGURES Title Source and Instrumentation in Initial Critical Experiment ...iiveveecan.. teeens teeneen Ceees e et e s Relation of Rod Position and Levels in Reactor Vessel * & & & & & » & & % 4 5 8 & 9 b e s s s ® & & 8 8 5 5 8 " A B 8 o 8 & & ¢ & & 9 @ > 9 Count Rate Ratios After First Four Additions of 2357, (Vessel full, rods at 51 in., source at 829 ft-9 in., chamber locations as in Fig. 1.) ..... Graphical Description of Control-Rod Calibration Experiments ...... et e s erestsarssat et asean e ccieean Differential Worth of Control Rod No. 1, Measured with Fuel Stationary. (Normalized to initial critical Z2°U 10ading.) veeeeeeenneenn. et .. Integral Worth of Control Rod No. 1 t.evivivvencenna. Effect of °°U Mass on Reactivity ........... ceeaean Change in Critical Position of Rod 1 as Shim Rods 2 and 3 are Inserted into COre ceveeerrireeceens ceess Lattice Arrangement of MSRE Control Rods and Sample Holder coveiveecenreonnes ce e ceerenncans csae Differential Worth of Control Rod 1, Measured with Fuel Circulating. (Normalized to initial critical 2350 10ading.) cievenronniinnn. Cecerrcecsereneracenn Results of Rod-Drop Experiments After %0 Capsule Additions .eeveass s eecsetticenteaconnnresna ceenaas Results of Rod-Drop Experiments After 65 Capsule AddIitions veeevvverenensones cer e ceicerersiserecnees Results of Rod-Drop Experiments After 87 Capsule Ad—d—itions ® & & o & & & & 0 & § 8 8 T S G O S .S P E O S S F O P S T e s . * 8 sensitivity of Rod-Drop Experiment to Changes in Magnitude of Reactivity Insertion .e..eecececerenecs Geometric Models of MSRE Core Used for Nuclear CalculationNsS eeeeseeenceccncannnoen st cescsneseane e Effect of Slow Changes in Core Temperature on the Reactivity eoveeeess Ceeeasereasrtsasr a0t ass s e .o Photograph of a Three-Dimensional Plot of the Reactivity Measurement Data ..... beeasesas e ceaenns . Conditions During Rapid Pressure Release While Circulating Helium Bubbles .ieiieeeiicectanrscecnnces Page 13 17 18 19 20 21 25 29 30 30 31 33 38 40 bty viii Fig. No. Title Page 19 Reactivity-Pressure Frequency Response with 2% to 3% Void Volume in Circulating Fuel. (Calcu- lated from pressure release experiment using Samulon's method with 0.2 min sampling in- te—r‘val ) ......... 4 % 9 0 5 2 9 & B L ] o & 8 8 9 ” & s & & 9 a ® & 8 8 * 5 8 2 45 20 Frequency Response of (6n/n )/(Ek/k ) at Zero Power; Fuel Stationary ..v.ieeereensereenneeeennnnas 48 o1 Frequency Response of (En/n )/ ( (ok/ky,) at Zero Power; Fuel Circulating vuueeiveeeeeennonnneennnn, .. 49 oD Pump Speed and Flow Startup Transients ..eeeeeeee... 52 2% Pump Speed and Flow Coastdown Transients ...... . 53 24 Control-Rod Response to Fuel-Pump Startup and COASTAOWIL 4 vvevenvnnsosnnsenenns C et teieecee e 54 ZERO-POWER PHYSICS EXPERIMENTS ON THE MOLTEN-SALT REACTOR EXPERIMENT B. E. Prince J. R. Engel o, J. Ball P. N. Haubenreich T. W. Kerlin ABSTRACT This report describes the techniques and results of a program of experiments designed to measure the important neutronic characteristics of the MSRE, under conditions of negligible nuclear heat generation. The program includes the initial critical °°U loading, the control-rod cali- bration (period~differential worth and rod drop-integral worth measurements), determinations of the reactivity loss due to fuel circulation, the "static" reactivity coef- ficients of excess #2°U concentration and isothermal core temperature, the fuel salt temperature reactivity coef- ficient, the pressure effects on reactivity, and a series of system dynamics tests (frequency response, transient flow, and neutron flux noise measurements). These measure- ments, carried out in June 1965, form much of the experi- mental baseline for current evaluation of the nuclear operation at full power. The report includes discussions of the comparisons of the measurement results with the corresponding neutronic characteristics calculated from theoretical models. 1. INTRODUCTION A program of zero-power nuclear experiments, including the initial critical experiment, was conducted on the Molten-Salt Reactor Experiment (MSRE) in June 1965. The purpose of this program was to establish the basic nuclear characteristics of the reactor system and provide a base- line for evaluation of the system performance in nuclear operation. A secondary purpose was to evaluate the calculational techniques and models used in predicting the properties of the MSRE. The initial critical experiment established the minimum critical concentration of #?U in the fuel under the simplest possible condition; that is, with core isothermal, fuel salt stationary, and control rods withdrawn to their upper limits. The remainder of the tests were de- signed to provide information about control-rod worths, various re- activity coefficients, and dynamic behavior of the system, all under zero-power conditions. With the initial critical concentration established, more 235U was added to the circulating loop in increments to permit the attainment of criticality with the salt circulating and with various control-rod con- figurations. Measurements were made of the differential worth of one control rod as a function of position, both with the fuel salt stationary and with it circulating. In addition, rod-drop experiments were per- formed to provide an independent determination of the integral worth of various control-rod configurations. Measurements of the critical control-rod configurations as a function of uranium concentration, both with and without fuel circulation, provided information about the 235U concentration coefficient of reactivity, the effect of circulation on reactivity, and control-rod shadowing effects. At several fixed =335U concentrations, the reactor system temperature was varied to provide data on the isothermal temperature coefficient of reactivity. Tests were also made in which the system overpressure was varied to evaluate the Pressure coefficient of reactivity. Several types of measurements were made to provide information about the reactor dynamics. These included the response of the system to single and pseudorandom sequences of re- activity pulses, the response to flow and temperature transients, and neutron flux noise data. sufficient excess uranium was added during this program to permit calibration of one control rod over its entire length of travel. This was expected to provide enough excess reactivity to compensate for all transient effects associated with full power operation. Since the principal independent variable in these experiments was the ?°U concentration in the fuel, the various tests were scheduled around the uranium additions. Thus, many of the experimental tests were interwoven chronologically‘to provide the required data, The results presented in this report deal with individual topics without regard for the actual chronology of the tests. In describing the results of the experiments, some reference to the reactor physics theoretical background is often an indispensable aid in interpretation., For the purpose of this report, we have limited this either to brief qualitative descriptions or to summaries of calculated core characteristics. Sources of details of the MSRE physics analysis are ref, 1 and various MSRE semiannual progress reports cited in the following sections. 2. INITIAL CRITICAIL, EXPERTMENT The purpose of thilis experiment was to provide a check on the calcu- lations of critical concentration under the simplest conditions, that is, with the core isothermal, the control rods fully withdrawn, and the fuel stationary. It also served to establish the basepoint from which the 2357 additions necessary to reach the operating concentration could be made with confidence. The fuel salt composition specified for power operation is 65LiF- 29,2BeFo-52rF,-0.8UF, (expressed as molar percentages). The total uranium content is considerably above the minimum required for criti- cality if highly enriched uranium were used, and was chosen for reasons of chemistry. With this total uranium content, theoretical calculations™ predicted that the reactor would be critical at 1200°F, rods cut, fuel stationary with 0.256 mole % 2°5UF, {0.795 mole % total UF,).Z Instead of using 52%—enriched uranium to make up the fuel salt, we decided to start with depleted uranium in the salt and add the required amount of 27U as highly enriched uranium (93% =°°U). This permitted preliminary operation with uranium in the salt before the beginning of nuclear operation and alsc facilitated the manufacture of most of the uranium-bearing salt. The salt was prepared in three lots: the carrier x A review of the basis of these calculations is included in Sec. %.8. sait, containing the beryllium, zirconium and most of the lithium fluorides; 73LiF-27UF, eutectic containing 150 kg of depleted uranium; and eutectic containing 90 kg of U in the highly enriched form. Thirty-five cans of carrier salt and two cans of eutectic containing the depleted uranium were blended as they were charged into a drain tank. This mixture of salt was then circulated for 10 days at 1200°F while the sampler-enricher was tested and 18 samples were analyzed to establish the initial composition. The critical experiment then consisted of adding enriched uranium in increments to bring the 235U concentration up to the critical point, Nuclear instrumentation for the experiment consisted of two fission chambers, two BFs chambers, and an 2%41Am-242Cm-Be source, located as shown in Fig. 1. The fuel salt itself also constituted a neutron source, due to reaction of alpha particles from 2°%U with beryllium and fluorine. The enriching salt was added in two ways: by transfer of molten salt from a heated can into a drain tank, and by lowering capsules of frozen salt into the pump bowl via the sampler-enricher. The latter method was limited to 85 g 225U per capsule, only 0.0012 of the expected critical loading. Therefore the bulk of the 23°U was added in four additions to the drain tank. After each addition the core was filled and count-rate data were obtained to monitor the increasing multiplication. The amount of 2°°U expected to make the reactor critical was calcu- lated to be 68.7 kg, using the volumetric concentration from the criti- cality calculations and the volume of salt in the fuel loop and drain tank. Before the addition of enriched uranium, count rates had been deter- mined with barren salt at several levels in the core. Then as the core was filled after each “°°U addition, the ratio of count rates at each level was used to monitor the multiplication. (Figure 2 shows elevations; count rates were determined with salt at 0.4, 0.6, 0.8, and 1.0 of the graphite matrix and with the vessel full.) Count-rate ratios with the vessel full after each of the four ma jor additions are shown in Fig. 3. Each addition, fill, and drain took between one and two days, so only four major additions had been planned. ORNL-DWG 65-7575 REACTOR INLET LINE 102 < SOURCE TUBE {-in.BF3 CHAMBER TUBE ~— REACTOR QUTLET LINE 100 34° REACTOR VESSEL 2~in. BF3 CHAMBER FISSION CHAMBER NO. 2 THERMAL SHIELD INSULATION FISSION CHAMBER NO. ¢ NUCL. INST. PENETRATION PLAN SOURCE TuBE T com—{ QUTLE . 1 / | —THERMAL SHIELD ~ \ |~ INSULATION 2-in. BF3 CHAMBER TOP OF GRAPHITE ELEV. 8331t 5% in. ELEV. B3O ft 8in. (CORE MIDPLANE) REACTOR VESSEL ELEV. 829 ft 9in. NUCL. INST. PENETRATION ELEV. 8281t 3Y2in BOTTOM OF GRAPHITE FISSION CHAMBERS ELEV 828ft QY% in. vy 1-in. BF3 CHAMBER BOTTOM ELEV. 828 f1 3in. ELEVATION Fig. 1. Source and Instrumentation in Initial Critical Experiment. ELEVATION (ft) ORNL-DWG 65-7573 836 836 I L REACTOR OUTLET 835 ft —3%in. - - - - 835 (- 835 (o B34ft—2.15in. | 834 — // 834 0.9 1o — 833f1=5%in. TOP _OF MOST GRAPHITE 833 f1—3 in- 51 — 1l UPPER ROD LIMIT 833 —— 48 os L 0.9 8 : 833 3 - < — 42 — :J- 1 o 3 08 - = Fo7l K u = 832 —— € = 36 | > 832 & £ 07 . - - o6 = £ u w 6 — =z il - @ B = 30f = O ~ 08 O - = O x = - T 831 —— w @ E 24 - 831 > 05 |- = 0 - = < O o o 5 s 0.5 a - o Ll W o 18 ' = - O o 3 c = W o4 - T oa - = 830 ——§ < 12 |- = 830 a - O b o [ Q O 0.3 C E 03} 5 6 - — 1 I z - DRIVEN ROD = u LOWER LIMIT 829 —— " | E 02 = 829 0.2 O [ § - 15 —g L. [T 0.4 SCRAMMED ROD LOWER LIMIT 0.1 | - o | 828ft—0Ygin. HORIZONTAL GRAPHITE 828 —— 0 828 oL 827 ft—7.15in. \ 827 —~— | 827 Fig. 2. Relation of Rod Position and Levels in Reactor Vessel. ORNL-DWG 65-7574 06 T R T “" B 2-in. BF 3 CHAMBER -in. BFy CHAMBER FISSION CHAMBER NO.1 FISSION CHAMBER NO.2 05 - — -— - — — - Q [ J [~ /i/ COUNT RATE RATIO (GCR, /CR) o w \ T \ ~ 40 44 48 52 56 60 MASS OF 239U IN TOTAL SALT CHARGE { kg) | / e Fig. 3. Count Rate Ratios After First Four Additions of ?3°U. (Ves- sel full, rods at 51 in., source at 829 ft-9 in., chamber locations as in Fig. 1.) After the third addition, with 6L.54 kg 235U in the salt, the projected critical loading was T70.0 + 0.5 kg £2°U. (The l-in. BFs chamber, located in the thermal shield, whose count rates extrapolated to a higher value was known to be strongly affected by neutrons coming directly from the source.) The fourth addition was intended to bring the loading to about 1 kg below the critical point. After 4.38 kg of 235U was added, the count rates showed the loading was within 0.8 kg of critical when the rods were withdrawn and circulation was stopped. Preliminary estimates of rod worth and circulation effect, based on changes in subcritical multiplication, were approximately the expected values. In the final stage, enriching capsules were added through the pump bowl to bring the loading up 85 g at a time. After each addition, circu- lation was stopped, the rods were withdrawn, and count rates were measured. With the reactor within 0.2% ak/k of critical, slight variations in temper- ature caused considerable changes in multiplication. (Variations in the voltage of the area power supply change the heater inputs slightly, re- quiring fine adjustments of the heater controls to keep the temperature precisely at a specified value.) After seven capsules, it appeared that after one more, the reactor could be made critical., The eighth was added, circulation was stopped, and the rods were carefully withdrawn. At ap- proximately 6:00 p.m., June 1, 1965, the reactor reached the critical point, with two rods at full withdrawal and the other inserted 0.0% of its worth. Criticality was verified by leveling the power at successively higher levels with the same rod position. The £°°U loading was 69.6 kg. During the approach to critical, a substantial internal source of neutrons was observed. The MSRE fuel mixture has an inherent source of neutrons produced by the interaction of alpha particles (primarily from 2347) with the beryllium and fluorine. Measurements were made with the reactor only slightly subcritical to evaluate the intensity of this source. Count-rate determinations with and without the external neutron source in place, under otherwise identical conditions, showed that the internal source supplied 0.0% to 0.0 as many neutrons to the core as the external source. The external source at that time had an absolute in- tensity of 1 X 10® neutrons/sec. However, because of the source location in the thermal shield (Fig. 1), some distance from the reactor vessel, only a small fraction of these neutrons are effective in reaching the core. If this fraction were 10%,* the effective external source con- tribution would be 1 X 10’ neutrons/sec, and thus the internal source strength would be in the range of 3 X 10° to 5 X 10° neutrons/sec. The calculated intensity of the internal source was within this range.l Predicted and observed 2°°U requirements for criticality are com- pared most logically on the basis of volumetric concentration. The required volumetric concentration of #°°U is nearly invariant with re- gard to the fuel salt density (unlike the required mass fraction, which *The offset location of the Am-Cm-Be source makes it difficult to calculate this fraction reliably. However, the value of 10% is com- patible with the results of diffusion-theory calculations.?t varies inversely with salt density) and depends not at all on system volume or total inventory. The observed 22°U concentrations, however, are obtained in the first instance on a weight basis, either from in- ventory records or from chemical analyses. These weight fractions must therefore be converted to volumetric concentrations by multiplying by the fuel salt density. The amounts of ©°°U and salt weighed into the system gave a 223U mass fraction of 1.4l + 0.005 wt % at the time of the initial criticality. The chemical analyses during the precritical operation and the zero-power experiments gave uranium mass fractions which were 0.985 of the "book" fractions. Applying this bias to the book fraction at criticality gave an "analytical™ #3°U mass fraction of 1.39% wt %. On a statistical basis, the uncertainty in the mass fractions obtained from chemical analyses is about +0.007 wt %. At the time of the zero-power experiments, we recognized that a small amount of dilution of the fuel salt should occur, due to residues of flush salt left in freeze valves and drain-tank heels when the fuel salt was charged, (During the initial fill operations, ’'LiF-BeF, flush salt was admitted to the fuel circulating system.) Experience with drain- flush-fill cycles obtained from MSRE operation subsequent to the zero- power experiments has indicated that the fuel salt would have been diluted by 20 + 10 kg of flush salt, If we assume that this amount of dilution occurred, the corrected value of the book mass fraction of Z3°U would be 1.408 + 0,007 wt %. The density of the fuel salt at 1200°F was determined after the uranium was added to the fuel drain tank, using pre-calibrated drain tank weigh cells and salt level probes within the tanks.” The average of four measurements was 145.1 1b/ft®, with a maximum deviation of 1.1 lb/ft3. These weigh cell measurements were in close agreement with an indirect determination of the density, inferred as follows. The density of the fuel carrier salt (65LiF-30BeF.-5ZrF,) was measured as the salt was charged to the fuel drain tank. This measured density, computed from externally measured weights and the volume between the level probes within the tanks was 140.6 1b/ft® at 1200°F. Addition of all the uranium added 10 during the zero-power experiments would be expected to increase the den- sity to about 145.9 1b/ft>. Concurrent with the zero-power experiments, laboratory glove-box measurements of the fuel salt density were made.,* These experiments gave an average density very slightly larger than the MSRE measurements, but the statistical uncertainty was sufficiently large that little additional information could be provided. TFor the calculations given below, we have used 145 = 1 1b/ft> as our best estimate of the density of the fuel salt at 1200°F, and with the uranium concentration at the time of initial ‘criticality. In comparing the observed and calculated critical concentration of 235U, a small temperature correction should be applied to the salt den- sity given above, since the core temperature at the time of criticality was 1181°F instead of 1200°F. Based on a fractional change in density of —1.2 X 107%/°F (see discussion in Sec. 4.1), the density at 1181°F would have been 145.3 + 1.0 lb/ft3. Finally, corrections must also be applied to the calculated critical concentration, both for the lower temperature and the fact that one rod was at 46.6 in., compared to the reference conditions of 1200°F and all three rods at maximum withdrawal, 51 in. These two effects nearly compensated for one another. The calculated 235U concentration for criticality at the reference conditions was 3%2.87 g/liter; corrected to the actual conditions, using measured values of the temperature coefficient of reactivity and the control-rod worth increment (see later sections), it is 32.77 g/liter. This "predicted"” value is compared with "observed" £°°U concentrations in Table 1. Con- centrations corresponding to both the book mass fraction, corrected for the flush salt dilution, and the analytical mass fraction, described above, are listed in Table 1. The predicted concentration was found to be in remarkably close agreement with the observed concentration cor- responding to the corrected boock value of the mass fraction, and to be very slightly higher than the concentration calculated from the analyti- cal mass fraction. 11 Table 1. Comparison of Critical 235U Concentrations (1181°F, pump off; 0.08% &k/k rod poisoning) 235 - 235U Mass Fraction Fuel Density U Concen = tration Predicted 32T Corrected book 1.408 + 0.007 145 + 1 32.8 + 0.3 Analytical 1.39% + 0,007 145 + 1 32.4 £ 0.3 3. CONTROL-ROD CALIBRATION 3.1 General Description The addition of Z2°U beyond the minimum critical loading had a two- fold objective: to end with enough excess reactivity to permit operation at full power and in the process to make measurements which could be analyzed to give control-rod worth and various reactivity coefficients. The final amount of 2°°U was to be enough to be critical at 1200°F with the fuel stationary and one rod fully inserted. The general method was to add 85 g 275U at a time through the sampler-enricher, after the addition determine the new critical rod position, and at longer inter- vals do other experiments. Following the initial critical experiment, another eight capsules were required before the reactor could be made critical at 1200°F with the fuel pump running. This was a consequence of the effective loss of delayed neutrons due to precursor decay in the part of the circulating system external to the core. Once this #2°U concentration had been reached, the critical position of the control rod to be calibrated (designated as the regulating rod) was measured after addition of each capsule, with the fuel pump running. At intervals of four capsules, period measurements to determine control-rod differential worth were also 12 made with the pump running, Then the pump was turned off, the critical rod position with the fuel stationary was determined, and period measure- ments were made with the fuel stationary. This went on until a total of 87 capsules had been added. Three times during the course of additions of 2°°U (after 30, 65, and 87 capsules) rod-drop effects were observed. The results of all these experiments are considered separately in the following sections. 3.2 Theoretical Guidelines oome useful theoretical guidelines in interpreting the control-rod experiments described in the sequel can be obtained by reference to the curves in Fig. 4., Bach curve is a qualitative graphical description of the change in the static reactivity”® as a function of regulating-rod position, with the other two rods withdrawn to their upper limits (position 0). The various curves represent different total loadings of 2357, increasing in the direction shown by the arrow, The static re- activity, p_, corresponding to each specific rod position and 235y loading is defined by the equation: P, = 7 > (1) where p is the actual number of neutrons emitted per fission, and p is the fictitious value for which the reactor with the specified rod position *Experimental measurements of the reactivity effects associated with substantial changes in core conditions, such as control-rod insertions, fuel additions, and temperature variations, require some care in inter- pretation. Thisg is particularly true here, where the results of using a mixture of techniques, such as static measurements by compensating re- activity effects, and dynamic measurements by period and rod-drop experi- ments, are to be interpreted on a consistent basis. We have used the static reactivity concept and scale as a basis for in integrated and unified interpretation of the measured reactivities. This was done by introducing normalization corrections, wherever necessary for consistency, in the manner described in this section, and also by avoiding instances where important differences between the static reactivity and the re- activity inferred from experiment can occur. The problems of reactivity measurement and interpretation have been quite thoroughly explored in the reactor physics literature. The discussions given in refs. 5, 6 and 7 are particularly relevant to the present work. 13 ORNL-DWG 67-12322 FINAL 23%U LoaDiNG INCREASING 23%y INITIAL 235U LOADING Fd — — — ——— — ——— — — —— —— — — —— — — FRACTION OF INSERTION OF REGULATING ROD (CIRCULATION STOPRED) Fig. 4. Graphical Description of Control-Rod Calibration Experi- ments. and material composition, and with the fuel stationary, would be just critical. An equivalent expression is: p. = — , (2) where ke is the effective multiplication constant of the reactor. Since ke (or equivalently, pS) is the quantity normally calculated in reactor 14 physics analysis programs, it is convenient to attempt to interpret the experimental measurements of reactivity on a basis consistent with the theoretical analysis (Sec. 3.8). One may observe from ¥Fig. 4 that, if the reactivity equivalent of the 23°U addition is known, a direct means of calibration of the reac- tivity worth of the regulating rod is provided, simply by relating the critical position of the rod (solid points shown as examples in Fig. 4) to the 225U loading. Alternatively, calibration of the rod by independent experiments provides an empirical determination of the reactivity worth of the additional #7°U, or the concentration coefficient of reactivity. This latter approach was chosen, and the experiments specifically aimed at determining rod worth were the stable-period measurements and the rod- drop experiments. In Fig. 4 a typical measurement of the stable period corresponds to a motion from the critical position upward and to the left along the short segment marked (p). The measured change in reactivity along the vertical axis, pp, is divided by the increment in rod motion, and this sensitivity, or differential worth, is ascribed to the mean position. A typical rod-drop experiment is indicated in Fig. L4 by the segment marked (d), extending from the initial critical position into the subcritical region. The purpose of this experiment is to measure the negative reactivity inserted by the drop, marked Pq* One other characteristic of some importance is indicated in Fig. L. Because the reactivity worth of the rods is affected by the 225U concen- tration in the core, one finds that Py, < ,psol’ or equivalently, that the curves representing different fuel loadings are not exactly parallel, Although the 275U loading was continually being increased during the course of these experiments, it is useful for purposes of consistency to interpret the combined reactivity measurements, over the whole range of rod movement, on the basis of a single mass of “°°U., Theoretical calcu- lations of the rod worths, summarized later, were used to determine the effects of the 235U concentration on total worth, and these corrections were used as an aid in normalizing the experimental reactivity measure- ments to a single 27U loading. 15 3.3 Differential-Worth Measurement: Fuel Stationary Pericd measurements were generally made in pairs. The rod being calibrated was first adjusted to make the reactor critical at about 10 w. Then it was pulled a prescribed distance and held there until the power had increased by about two decades. The rod was then inserted to bring the power back to 10 w and the measurement was repeated at a some- what shorter stable period. Two fission chambers driving log-count-rate meters and a two-pen recorder were used to measure the period. The stable period was determined by averaging the slopes of the two curves (which usually agreed within about 2%). Periods observed were generally in the range of 30 to 150 sec. For the measurements with the pump off, the standard inhour re- lation® was used to calculate the reactivity increment corresponding to w, the observed stable inverse period, viz., i o= WA+ ) —— (3) The decay constants, Ai’ and the effective delay fractions, Bi’ used in these calculations, are listed in the second and fourth columns of Table 2. These delay fractions contain approximate corrections for the in- creased importance of delayed neutrons because of thelir emission at lower energies relative to the prompt fission neutrons in the MSRE.® The neutron generation time, A, was 2.6 X 10™% sec for the initial critical loading, obtained from theoretical analysis. When applied to the analysis of period-rod sensitivity measurements, Eq. (3) is quite insensitive to neutron generation time. Prior to pulling the rod for each period measurement, the attempt was made to hold the power level at 10 w for at least % min, in an effort to. help insure initial equilibrium of the delayed neutron precursors. Generally, however, it was difficult to prevent a slight initial drift in the power level (as observed on a linear recorder), and corrections were therefore introduced for this initial period. The difference between the reactivity during the stable transient and the initial reactivity, 16 Table 2. Delayed Neutron Fractions in the MSRE 10%* x Delay Fraction (n/n) Decay Constant Group - . (sec 1) Effective Actual (static fuel) 1 0.0124 2.11 2.23 2 0.0305 14.02 14.57 3 0.111% 12.54 13.07 Y 0.3013 25.28 26.28 5 1.140 740 7.66 6 3.010 2.70 2.80 both computed from Eq. (3), was divided by the rod movement to obtain the rod sensitivity at the mean position. The differential-worth measurements made with the fuel pump off are plotted in Fig. 5. As discussed above, theoretical corrections have been applied to these measurements to put them all on the basis of one 235U concentration, arbitrarily chosen as the initial critical concen- tration at the beginning of the rod-calibration experiments. Theoretical calculations described in Sec. 3.8 indicated that the static reactivity worth of a single rod is reduced by nearly 9% of its total worth for the total addition of 2°°U made during the course of these experiments. The approximate correction factors which were applied to the rod sensitivity measurements summarized in Fig. 5 increase linearly with 222U concen- tration, up to 1.087 for the measurements made near the final concentra- tion (corresponding to the points between 1 and 2 in. withdrawal). Some imprecision (scatter) is evident in the data shown in Fig. 5, because the differential worth is the ratio of the increment in re- activity to the increment in rod withdrawal, each of which is subject to experimental error. The root-mean-square deviation of the data points of Fig. 5 from the curve is about 2.8 X 107%*% &k /k/in., or ~0.7% of the mean differential worth; the maximum deviation of a single point was 8.7%. TFor the short term type of experiments described in this report, 17 ORNL-DWG 65-10292 0.07 | I [ ) 0.06 o ~3 7{' 0.05 o7 0.04 @ e, 0.03 . - . B W - / . 0.02 //4 N 0.01} DIFFERENTIAL WORTH [(% 84/4)/in] 0 4 8 12 {6 20 24 28 3z 36 40 44 48 52 ROD POSITION {in. withdrawn) Fig. 5. Differential Worth of Control Rod No. 1, Measured with Fuel Stationary. (Normalized to initial critical “2°U loading.) the precision of determining the rod position was about £0.01 in. Probably the most important source of imprecision in the differential worth was in the measurement of reactor period. As described above, only the con- ventional reactor instrumentation was used in recording this data. De- termination of the period in each measurement involved laying a straight- edge along the pen line record of the log n chart and reading the time interval graphically along the horizontal scale which corresponded to a change of several decades in the neutron level. Since these charts, to- gether with the pen speeds, are subject to variations, this was a prob- able source of error in the rod sensitivity measurements. Figure 6 shows a curve of the magnitude of rod reactivity vs position, which is the integral of the differential worth curve in Fig. 5. An in- tegral worth curve is also shown which is normalized to the final £2°U concentration, This latter is simply the first curve reduced by a factor of 1.087. 18 ORNL—DWG 65—-10293 (65.25 Kg '8 — ~- 1 1 1 CRITICAL LOADING ::::: - — FINAL LOADING 235 REACTIVITY WORTH (% 84/k) ro 1.71 Kg U IN LoorP) T 1.0 — -— —t -— ogr— { —o{ 4 LA —t 4 b1 ] ] - — 06 — - — -+ .4 |— — —— — 0.2 |— - — - 0 0 4 8 12 16 20 24 28 32 3% 40 44 4B 52 ROD POSITION (in. withdrawn) Fig. 6. Integral Worth of Control Rod Neo. 1. 3.4 Reactivity Equivalent of 225U Additions Following the initial achievement of criticality with the fuel pump rumning, the effect of each capsule addition on the critical position of the control rod was measured. The critical position of the control rod was measured with the pump off after every fourth capsule. Critical rod positions at each 22U level were then converted to reactivity by using Fig. 6 and linearly interpolating between the initial and final loadings to correct for the °°U concentration effect on the total rod worth. The results are shown in Fig. 7. In this figure, the ordinate is the total excess static reactivity which would result from withdrawing the rod from its critical position at that 22U loading to the upper limit of rod travel, The separation between the two curves reflects the net reduction in reactivity due to emission of delayed neutrons in the external piping and heat exchanger (see later discussion). 19 ORNL-DWG 65—10291 20 - e : : o FUEL NOT CIRCULATING~Y o FUEL CIRCULATING 10 s e e /i;#”flffl e ] . | 08 : - P - Sk/k () 65 66 &7 68 69 70 71 72 MASS OF 233U IN FUEL LOOP (kg) Fig. 7. Effect of 227U Mass on Reactivity. The 235U concentration coefficient of reactivity is given by the ratio of the change in reactivity to the fractional change in 2350 con- centration, or circulating mass, as a result of a small addition. This is the slope of the curves in Fig. 7 at any particular concentration, multiplied by that concentration. The value of (&k/k)/(3m/m) obtained from the experimental curves was 0.22%, which was very nearly independent of 2357 mass over the range shown in Fig. 7. The theoretically calcu- lated value of this quantity was 0.248 for the approach to the initial critical mass, and 0.234 for the average during the excess uranium additions. 3.5 Rod—Shadowing BExperiments During the course of additions of enriching capsules, three separate experiments were performed in which the change in the critical position of the regulating rod (rod No. 1) was recorded as the shim rods (rods Nos. 2 and 3) were inserted into the core, These experiments were per- formed with the pump running. They included observations of both the effect of inserting a single shim rod (rod No. 2) with rod No. 3 held fixed in the fully withdrawn position, and the effect of inserting rods 20 2 and 3 as a bank, that is, with their tips at identical elevations. Figure 8 shows the data obtained from these experiments. Each experiment was terminated at the 45-deg line, where the tips of all three control rods are at equal insertions. Some useful information concerning the reactivity worths of various shim and regulating-rod combinations can be obtained from the results shown in Fig. 8. To the critical position of the regulating rod at the start of each experiment with a given 225U loading, there corresponds an equivalent excess reactivity, relative to the reference conditions. This reactivity may be determined from the curves in Fig. 6. Then, since each curve in Fig. 8 represents control rod positions at conditions of criticality, 1t follows that the curve pairs corresponding to a given 2357 loading represent various shim-regulating-rod combinations which are equal in excess reactivity worth. In addition, these curves can be used to determine the reactivity worth corresponding to full insertion of the banks of two and three control rods, if use 1s made of an approximate device. If we assume that the shape of the banked worth curve is suf- ficiently close to that for the single regulating rod, shown in Fig. 6, ORNL-DWG 65-10655 8 | | [ I | ! I I i ! I ®* ROD 3 FIXED AT 51 in. ROD 1 COMPENSATING FOR INSERTION 12} - OF ROD 2 5 a RODS 2 AND 3 AT EQUAL POSITIONS. 5 © ROD 1 COMPENSATING FOR : o R £ INSERTION OF BOTH SHIM //// o 20 b—— RODS — — — [} £= 2 / 24— - —— - - —_t —] L ] [ ] 828 —— — af ——.le . . o el - % 32 / I. A A l. o) .. A'A* .l (@] A ) A [ w A ® 4} ® = 36 [—t . Al | A LI —_ g » A . A L4 3 A ) A L W ° A ° A > O 40 —_— S - e — - — = fs '3 A Y A L] I®) A - A e A .. - A [ ('% 44 —1 aA-p f\fi.q Qc. . — o Ne Aw A & a he Ale o W fa’ J 1}. Ale 48 - " - e e fe - / & 6794 kg €9.94 kg se 7474 kg @ 235 » 235 235 52 48 44 40 36 32 28 24 20 16 2 8 4 0] CRITICAL POSITION OF ROD 1 (inches withdrawn) Change in Critical Position of Rod 1 as Shim Rods 2 and 3 are Inserted info Core. Fig. 8. Change in Critical Position of Rod 1 as Shim Rods 2 and 3 Are Inserted inteo Core. 21 a simple ratio converts each of the three reactivity levels to the cor- responding reactivity with the rod bank fully inserted in the core. {This relative invariance of the shape of the worth vs position curves is sup- ported by theoretical calculations.) The conversion of the reactivity measurements in the manner described above is summarized in Table 3. In addition to the experiments described above, at each of the three 2357 levels, an experimental check was made to determine if there was any asymmetry in the control-rod worths, depending on the rod designated as the regulating rocd. The configuraticon of control rods and graphite sample holder is shown in Fig. 9. ©Since the three control rods are of identical design, they could only differ in relative poisoning effect by virtue of OHNL WG (4-8414 TYPICAL FUEL PASSAGE NOTE: STRINGERS NOS. 7, 60 AND &1 (FIVE) ARE REMOVABLE. ¢ e < 59 4 N ' 2= CONTROL ROD GUIDE TUBE \g 60 ?iij)g / g \\ //é GUIDE BAR ! - REACTOR CENTERLINE THREE GRAPHITE AND & = INOR-8 REMOVABLE N SAMPLE BASKETS e REACTOR CENTERLINE Fig. 9. Lattice Arrangement of MSRE Control Rocds and Sample Holder, Table 3. Worth of Control-Rod Banks Measured in Rod Shadowing Experiments Experiment Rod Group Total 235U Excess Re- Banked Critical Worth at Banked Critical Position Total Worth at , i o , o (s RHP GSHORE s vionaram) OB atmaL T Terien 1 1-2 67.94 0.780 %6 .k 0.1903 4,099 2 1-2 69,94 1.460 28.8 0.3673 3,975 3 1-2 71.71 2.095 23,3 0.5142 4,075 Average 4,050 1 1-2-% 67.94 0.780 39.0 0.1394 5.5% 2 1-2-3% 69.94 1.460 3%, % 0.2602 5.611 3 1-2-% 71.71 2,095 28 .4 0.3761 5.570 Average 5.592 "Normalized to initial critical loading in loop (65.25 kg 235U), zero point of reactivity with all three rods at 51 in, cc 23 their position with respect to the graphite sample holder. At a given 235U loading and core temperature, the critical positions of each of the three rods were measured and compared, with the other two rods held in the fully withdrawn position. The amount of asymmetry in rod worths ob- served in these experiments was negligible ( °F~t. 38 ORNL-DWG 65-8032R 1.8 K4 1.6 — - & e 67.86 kg 235U IN LOOP ..¢* 14— [ — - - — 1 ,.’ Il/. 12 — - - o - — l'. £ 10 [— ‘ A e - < e < 1 - z o8 ——- — — — _7! — ,/_. P— o * 71.71kg 235U IN Locgz,- e 0.6 {— : P = - — ./. ,',. o ~"69.85 kg 23%U IN LOOP o4l | - I N s -~ | o : -1 : 0.2 |— e | - o i | o | 1000 1050 1100 150 1200 1250 FUEL TEMPERATURE (°F) Fig. 16. Effect of Slow Changes in Core Temperature on the Reac- tivity. Calculations consistent with the models described in the preceding section gave a value of —8.1 X 107> °F~! for the total isothermal temper- ature coefficient of reactivity. The associated components for the fuel salt and graphite were —4.1 X 107> °F1 and —4.0 X 107> °F~1, respec- tively. Since the fuel coefficient is very nearly proportional to the value of the coefficient of thermal expansion of the salt, it is necessary to qualify the calculated ccefficients with this value. These calcu- lations were based on an expansion coefficient of —1.18 x 107% °pF~1, obtained from an early empirical correlation of temperature with density. This value was later found to be in good agreement with the expansion coefficient determined from observations of the change in salt level in the MSRE fuel pump bowl with loop temperature (two separate measurements gave —1.09 X 107* and —1.15 X 10™% °p~1),3 The experiment at 71.7 kg °°U shows a lower slope below about 1140°F. We do not believe that the temperature coefficient is lower in this range but that another phenomenon became significant during this part of the experiment, That is the appearance of an increasing amount of helium 39 bubbles in the circulating salt as the temperature was lowered. The evi- dence for this is discussed in the section on pressure effects. The effect, so far as the temperature experiment is concerned, was that the bubbles tended to reduce the amount of fuel salt in the core, compensating to some extent for the increase in density of the salt itself as the tem- perature was lowered. Thus the slope of the lower part of the curve can- not be interpreted as a temperature coefficient of reactivity in the usual sense, The influence of the core temperature on reactivity, as reflected in the change in critical position of the regulating rod, is also illustrated in Fig. 17. This is a photograph of a three-dimensional model™ showing the critical position of the regulating rod (vertical axis) as a function of the 275U concentration (horizontal axis, nearly in the plane of the page), and core temperature (depth axis, front-to-back of the model). The measured critical position of the regulating rod with fuel circulating at 1200°F vs the 2°°U concentration is represented by the relatively "dense" curve sloping downward to the right. FEach point represents the addition of one capsule of enriching salt. The critical position of the regulating rod with circulation stopped, measured after every fourth capsule, is the curve with the "sparse" points lying vertically below and in the same plane as the first curve., The separation between the two curves repre- sents the increment of rod insertion required to compensate for the loss in delayed neutrons due to circulation. The three experiments in which the core temperature was varied are shown in Fig. 17 as the segments sloping upward from the front-to-back of the model and crossing the upper curve of critical position vs 22°U concentration at 1200°F. The points at the extreme upper left of the model represent the data taken at the time of the initial critical experiment (when the temperature was ~1180°F and the rods were poisoning 0.08% &8k/k). *This model was constructed by J. A. Watts, as a visual aid for demonstrating to visitors the important parameters influencing the neutronic behavior of the reactor. The position of the beads in the model corresponds as closely as possible to the actual data points obtained during the rod calibration experiments. Fig. 17. Photograph of a Measurement Data. Three=Dimensional Plot of the Reactivity Photo 86401 O% 41 4,2 Fuel Temperature Coefficient of Reactivity An attempt was made to separate the fuel (rapid) and graphite (slug- gish) temperature coefficients by an experiment in which the coolant system was used to increase the fuel salt temperature rather abruptly. This was done by stopping the fuel pump, raising the temperature of the circulating cocolant salt and the stagnant fuel in the heat exchanger, then restarting the fuel pump to pass the hotter fuel salt through the core. The reactor power was controlled at 10 w by the flux servo.® TFor these experiments, use was made of a Bunker-Ramo 340 on-line digital computer and data logger, installed as part of the MSRE instrumentation.** The output of a thermocouple in the reactor vessel outlet, logged digit- ally at l/h-sec intervals, showed a brief increase of 5 to 6°F as the hot salt first passed. It then leveled at about 3.5°F for a few loop transit times before decreasing gradually. The noise in the analog-to- digital conversion (+1°F) limited the accuracy of the measurement, but by taking an average of 50 points during the level period after mixing and before the graphite temperature had time to change significantly, a value was obtained for the increment in fuel temperature. The reactivity change was obtained from the change in rod position using the rod cali- bration results, corrected for the decrease due to circulation, and ascribed to the fuel-temperature increase. The resulting coefficient was —(4.9 + 2,%) X 1075 °F~1, Predicted values of the fuel-temperature coef- ficient lie in this range. This experiment was later repeated in a slightly different mamner, with the intent of improving upon the first measurement. The reactor inlet and outlet temperature thermocouple signals were preamplified and filtered before digitizing, reducing the noise to about +0.1°F. This *For rapid changes in reactivity, small perturbations in the flux do occur, even when the regulating rod is flux servo controlled. However, calculated corrections for these effects were found to be insignificant for these experiments. **At the time of the zero-power experiments, the logger-computer was still being debugged and was frequently unavailable. Thus, only the con- ventional recording instrumentation was employed for most of the rod calibration experiments described in the preceding sections. 42 time the fuel was kept in circulation and the coolant loop was stagnant. After heating the stagnant coolant salt about 20°F hotter than the fuel salt, the coolant pump was started, introducing a hot slug of fuel into the heat exchanger and subsequently into the core., 1In this test, then, the change in reactivity was due entirely to the change in temperature, and the rod-induced reactivity required to keep the reactor critical was approximately equal and opposite to that due to the fuel and graphite temperature change., If the immediate effects could be attributed to the fuel temperature coefficient alone, a value of —(L4.7 + 0.7) x 1072 °pF~1 was obtained from this test. However, noise effects relative to the temperature changes involved and also special problems due to temperature measurement lag effects still prevented us from obtaining a good assess- ment of the uncertainty in this measurement. 4.3 Effect of Pressure on Reactivity The possibility exists for a pressure coefficient of reactivity in the MSRE, because undissolved helium can be entrained in the circulating fuel through the action of the fission product gas stripping device. A small fraction of the fuel pump discharge stream (about 50 gpm out of 1250 gpm) is diverted into a spray ring in the gas space in the pump bowl. The purpose of the spray, or stripper, is to provide contact so that 13°Xe in the salt can escape into the gas space, which is continuously purged. balt jetting from holes in the spray ring impinges on the surface of the liquid pool in the pump bowl with sufficient velocity to carry under considerable guantities of gas, and a small fraction of the submerged bubbles can be swept through the ports at the pump suction into the main circulating stream of fuel, A steady state is reached when the helium concentration in the main stream has increased to the point where loss of helium through the stripper flow equals the rate of injection. At steady state, the volume fraction of gas in the circulating stream varies around the loop proportionally to the inverse of the local pressure, which changes with elevation, velocity, and head losses. If, after steady state is established, the loop pressure is changed rapidly, the mass ratio of undissolved gas to liquid would be expected to remain 43 practically constant and the volume fraction of gas in the loop would decrease with increasing pressure. This would give rise to a positive pressure coefficient of reactivity. On the other hand, for very slow increases in pressure, the volume fraction of gas at the pump suction would tend to remain constant, and the volume of gas in the core would increase, because the ratio of absolute pressures between the core and purp suction is reduced. These slow changes could give rise to a small negative pressure coefficient of reactivity. We performed three tests to explore the effect on reactivity of changing the system overpressure. In each of the three tests the loop overpressure was slowly increased from the normal 5 psig to 15 psig and then guickly relieved, through a bypass valve, to a drain tank that had previously been vented to atmospheric pressure. The first two tests were carried out at normal system temperature with the normal operating level of salt in the fuel-pump bowl., No change in control-rod position was required to maintain criticality and no significant change in pump-bowl level was observed during either of the tests. These indicated that the pressure coefficient was negligibly small and that essentially no helium bubbles were circulating with the salt. Further evidence of the lack of circulating voids was obtained from a gamma-ray densitometer on the reactor inlet line; this instrument showed no change in mean salt density during the tests.™ The third test was performed at an abnormally low pump-bowl level which was obtained by lowering the operating temperature to 1050°F. Pigure 18 shows the pressure transient and the responses of the regu- lating control rod, densitometer, and fuel-pump level during the rapid pressure release. For these conditions, a change of one unit on the densitometer was equivalent to a change of about 0.15 vol % in the circulating void fraction. Independent evaluations of the void fraction from these three parameters all gave values between 2% and 3% by volume, The frequency response characteristics of the effects of pressure on *The use of this instrument is described in more detail in ref. 18, The reported sensitivity of the measurement is 0.076 vol %. 4, ORNL-DWG 65-8074R 65 I — : _ B, | - —. — - — — 55 S S A — - i — ] FUEL PUMP LEVEL (%) OVERPRESSURE (psig) DENSITOMETER CONTROL ROD POSITION (in.) 2€ TIME (min) Fig. 18. Conditions During Rapid Pressure Release While Circulating Hellum Bubbles. reactivity were calculated, using Samulon's method,l® from the pressure and rod motion. These results are shown in Fig. 19 along with the pre- dicted high-frequency response for a void fraction of 1.2%. Extrapo- lation to the observed curve gives a void fraction of 2% to 2 1/2%. The low- and high-frequency pressure coefficients were +0.0003 and +0.01lL (% Sk/k)/psi, respectively, for this particular condition. ORNL-DWG 65-8304A 07— HER - 1T — SRR = 1 =1 e - l% Fi = “j; I __T mel 1 gia| 't“l -1 - o 1 O O N ‘ [ A O 0 I | i i _—t Jr« ] L __T fl»,, - ‘ Ll e ’ T Li_ fi" \‘I J;» i l 7\l V } T . ./.Ji_L‘L- V 7fi7i ] h I A A % ety O e e R = TR | e T e TG S 5 511 Y o 17,/’ T HIGH-FREQUENCY VALUE LA L1 I e e ““@J_[j' | ’iL+ Cfififififlfifilfg>/ 1] EL& S 1T “L“ o S H LE' *:J jEs /"/E e == = _fi{ Sl i S S 54 S UD s - .+ S JN B £ S s a0 A S St i ey v - - % - - I ] URE R L T fil/% SEALH I M a1 B 11 A H'* L o co . s 1444 [ — s A .t L E— et il - i | o i T I 0.0001 0.004 Q.01 oA { 10 w. FREQUENCY (radians/min} Fig. 19. Reactivity-Pressure Frequency Response with 2 to 3% Void Volume in Circulating Fuel. (Calculated from pressure release experiment using Samulon's method with 0.2-min sampling interval.) 5, DYNAMICS TESTS 5.1 Purpose of Tests We performed a variety of dynamic tests during the operation at zero power. These tests were the start of an extensive program to evaluate experimentally the inherent nuclear stability of the MSRE at all power levels. The reactor system was analyzed on a theoretical basis,®C and the tests were designed not only to characterize the present system but also to evaluate the techniques and mathematical models used in the theoretical analysis. Results from the analysis of the zero- power tests are presented below. An extensive discussion of the sub- sequent at-power tests is given in ref, 21. 5.2 Frequency-Response Measurements A series of tests was run to determine the frequency response of neutron level to reactivity perturbations. These experiments included pulse tests, pseudorandom binary reactivity-perturbation tests, and measurements of the inherent noise in the flux signal. Tests were run 46 with the fuel pump on and with it off. Noise measurements were also made during a special run with a low pump-bowl level where there were entrained bubbles in the core. The frequency response is a convenient measure of the dynamic characteristics of a reactor system. Classically the frequency response 1s obtained by disturbing the reactor with a sinusoidal reactivity per- turbation, and observing the resulting sinusoidal neutron-level variations. The magnitude ratio is defined as the ratio of the amplitude of the out- put sinusoid to that of the input sinusoid. The phase angle is defined as the phase difference between the output sinusoid and the input sinus- oid. Other procedures, such as those described in this report, can be used to yield the same results as the classical method but with less ex- perimental effort. The zero-power frequency response tests serve to check the theoret- ical, zero-power frequency-response predictions, but they do not furnish direct information on the stability of the power producing reactor. The zero-power tests, however, do serve as an indirect, partial check on the at-power predictions because the dynamic behavior at power is simply the zero-povwer case with the addition of reactivity feedback from the system resulting from power-induced temperature changes. Thus, the verification of the zero-power kinetics predictions lends some support to the pre- dictions regarding power operation. 5.3 Pulse Tests In the pulse tests a control rod was withdrawn 1/2 in., held there for 3.5 or 7 sec., then returned to its original position. The precise rod positioning was done using a special analog computer timing circuit. The rod was located such that 1/2-in. travel caused a change in re- activity of about 0.03%. The rod position signal and flux signhal were recorded digitally at 0.25 sec intervals using the BR-3L0 on-line digital computer, The frequency-response characteristics are generally deter- mined from pulse-response tests by obtaining, numerically, the Fourier transforms of the input and output signals. One reguirement, however, is that the Fourier integrals rmust be closed; that is, both input and 47 output signals must eventually return to their initial values. The zero- power reactor, however, is an "integrating" system, since a temporary reactivity perturbation can cause the flux to level out at a new value. To circumvent this problem, the flux output signal was filtered with a 50-sec time-constant, high-pass digital filter., The output of this filter, which eventually returns to its initial value, was then Fourier-transformed, and the resulting transform corrected for the filter response. "Practice" tests on a simulated zero-power reactor indicated that reasonably accurate results could be expected from these tests only if the fuel-temperature drift during a test could be held essentially to zero, and if the control rod could be repositioned to its initial loca- tion within an accuracy of better than 0.01 in. The results of the pulse tests, shown in Figs. 20 and 21 for the cases of the fuel stationary and circulating, are in adequate agreement with the theoretical frequency response curves. This agreement provides indirect evidence that the stringent accuracy requirements described above were actually achieved in these tests. 5.4 Pseudorandom Binary Sequence Tests The pseudorandom binary sequence (PRBS) is a specially selected series of positive and negative rod movements which repeats itself after a certain number of basic pulse or bit times. The principle advantage of a PRBS in frequency response testing is that its frequency spectrum typically consists of a large number of harmonics of approximately equal size.® This means that the response may be analyzed at many frequencies generated in a single test, and the signal-to-noise ratio at these known frequencies is typically very good. A PRBS is characterized by the number of bits in the sequence and the bit duration., The bit time 1s defined as the mininum possible pulse duration in the sequence. All pulses in a PRBS are minimm width or integral rmultiples thereof. Numerous sequences may be generated, but they are restricted to certain specific numbers of bits. For a sequence of Z bits and a bit duration at At sec, the PRBS pericd is ZAt sec. The lowest harmonic radian frequency W, and the spacing between harmonics Aw 48 ORNL-DWG 66-10708A 104 < ZERO POWER 6 AN )y 7 sec PULSE, TEST NO. 1 . l N 7 sec PULSE, TEST NO. 2 A > A 35 sec PULSE, TEST NO. 3 & < 35 sec PULSE, TESTNO. 4 o 103 AN NOISE ANALYSIS v \‘ 6 h A NG A s R 2% gn/n v ? Sk/k, . ?}? 2 ST, o 3 34 n 80 Yy ?-V.\i 102 i 6 ! R 2 A% 10' 0001 0.01 0.1 10 10 100 w, FREQUENCY (radians/sec) ORNL-DWG 65-8909A 20 0 ik Ll = 2 l f'o g < AM”' Ny | ~ | T e o - J ,,Jl,éf'L . [+] E o A ? opn a‘ - 4] 5 40 | et lpln) L 0 Pe ! pt - s A7 4« a i , ZERO POWER -60 ot A1 7 sec PULSE , TEST NO. | @ il //’/ 7 sec PULSE , TEST NO. 2 4 | ||| , 1(/ 3.5 sec PULSE ,TEST NO. 3 & L1 - 1 A= PULSE ,TEST NO. 4 O 11 80“ " :/r“ 3.5 sec PU ,TEST NO I [ -100 l ‘ l t ‘ | 0.001 0.01 0.1 1.0 10 w, FREQUENCY ( radians /sec) Fig. 20. Frequency Response of (&n/ng)/(8k/kg) at Zero Power; Fuel stationary. 49 19 BIT PRBS—INDIRECT ANALYSIS @ 1ot ORNL-DWG 66-10286 RN ZERO POWER 5 LAY 19 BIT PRBS DIRECT ANALYSIS @ 19 BIT PRBS INDIRECT ANALYSIS & \\\ 63 BIT PRBS DIRECT ANALYSIS & 2 - 63 BIT PRBS INDIRECT ANALYSIS ¥ 5 q 7 sec. PULSE, TEST NO. | v 103 7 sec, PULSE, TEST NO. 2 ] "Eui ~ NOISE ANALYSIS o] 5 A4+ SM“Q Sk 102 ot o 5 (9] o\ 2 10' 0.00t 0005 001 002 005 O1 02 05 10 20 50 100 200 500 1000 w, FREQUENCY (radions/sec) ORNL—DWG &5-89074 20 i T T T T ! AR ZERQ POWER f 11;! | i ;H} 10 + 19 BIT PRBS—DIRECT ANALYSIS A - r__“47T+YT! : ‘|yfi ° |63 BIT PRBS—DIRECT ANALYSIS ¥ | J | ‘l —-10 163 BIT PRBS—INDIRECT ANALYSIS ¥ SR —20 | 7 sec. PULSE, TEST NO. | o ‘\fifl 7 sec PULSE, TEST NO. 2 A N -30 |— P T T T NG t . ! ; \ A | - | ! PHASE (deg} & B o O B fiA R ) —-60 - dehe J7/<; . ?‘ a/ o | L o 70 — 1 oww — | wl w0 | AT e —g0 4=} — { i -100 0.001 0.1 OR 1 10 w, FREQUENCY (rodians/sec) l'ig. 2L. Frequency Response of (®&n/ng)/(dk/kg) at Zero Power; Fuel Circulating. are given by ) :Am:g—rf— (15) o) yAA Sequences of 19 and 63 bits were used in the zero-power tests with bit times of 6.58 and 3.%5 sec, respectively. The 19-bit sequence was generated manually, while the 63-bit test sequence, which was run after the on-line computer was operational, was generated automatically by a speclal shift-register algorithm. In both cases, analog computer timer circuits were used to control the rod-drive motor insert and withdraw "on" times. The rod motion about the average position was adjusted to give a reactivity perturbation of about +0.015%. As with the pulse tests, the rod-position and neutron-level signals were recorded digitally at 0.25- sec intervals using the MSRE computer. The frequency response was ob- tained by two different methods. The direct method used a digital filtering technique to obtain the power spectrum of the input and the cross-power spectrum of the input and output. The frequency response at some frequency is the ratio of the component of the cross-power spectrum at that frequency to the component of the input power spectrum at that frequency. The indirect method involved calculation of correlation functions and subsequent numerical Fourier transformation. Both methods gave essentially the same results. A description of the indirect method is given in ref. 23, and the direct method is discussed in ref. 21. The results of the PRBS tests for the circulating fuel case are compared in Fig. 21 with the results from both the pulse-tests and the predicted curves. The agreement is reasonably good, considering that the stringent conditions of no temperature drift and extremely accurate rod positioning, described in Sec. 5.3 for the pulse tests, had also to be imposed on the PRBS tests. 5.5 Neutron Fluctuation Measurements Flux-noise measurements were made by Roux and Fry=% at three dif- ferent zero-power conditions: fuel-salt stationary, normal fuel-salt circulation, and circulating fuel with entrained bubbles. The 51 fluctuations were recorded on an analog band-pass filter power-spectral- density (PSD) analyzer. One run was also digitized and analyzed by auto- correlation and digital filter techniques for comparison. All methods gave essentially equivalent results. Results of the PSD analysis for stationary and circulating fuel are shown in Figs., 20 and 21. These points were obtained by taking the square root of the measured PSD after subtracting the asymptotic high-frequency value of PSD due to noise in the neutron detection process.=® This computation results in an ampli- tude-ratio vs frequency curve if one assumes that the input reactivity noise is white, that is, has a constant PSD in the frequency range of interest. Since the magnitude of the input PSD is unknown, only the relative magnitude of the noise results are available; thus the curves were arbitrarily normalized to the theoretical results at ¢ radians/sec. Based on the good correspondence of noise response and theoretical re- sponse, the assumption of white noise input appears valid. The results of the noise analysis of the case where bubbles are entrained in the fuel salt showed a substantial increase in PSD in the frequency range of 1 to 10 fadians/sec over the no-bubble case, indicating an increase in the input reactivity fluctuation. Previous experiments with the MSRE core hydraulic mockup indicated that random, hydraulically induced pressure fluctuations in the core would probably cause a signif- icant modulation of the core void fraction, thus causing reactivity fluctuations., Hence, additional flux noise in this frequency range was expected, although it was not possible to predict the "shape" or charac- teristics of this spectrum. 5.6 Transient Flow-Rate Tests Several transient flow-rate tests were run in order to: (1) obtain startup and coastdown characteristics for fuel- and coolant-pump speeds and for coolant-salt flow rate; (2) infer fuel-salt flow-rate coastdown characteristics from the results of (1), since there is no flowmeter in the fuel loop; and (3) determine the transient effects of fuel flow-rate changes on reactivity. 52 Figures 22 and 23% show the fuel-pump speed, cooclant-pump speed, and coolant-salt flow rate vs time for pump startup and coastdown. Data were taken with the on-line computer and with a Sanborn oscillograph. The output of a differential-pressure cell across the coolant-salt venturi was recorded directly on the oscillograph, and the square root of that signal was taken as flow rate. The lag in the response of the computer flow signal is due to the response characteristics of the EMF-to-current converter and the square-root converter between the differential-pressure transmitter and the computer input. It was hoped that the coolant-pump speed and coolant flow rate would coast down in unison so that the fuel flow-rate coastdown could be in- ferred directly from the fuel-pump speed coastdown curve. This was not the case, however, and attempts to infer the fuel-flow coastdown transient were abandoned. Reactivity effects of fuel flow-rate transients were measured by letting the flux servo controller hold the reactor critical during the transients; the reactivity added by the rod is then equal (and opposite) to the reactivity change due to the flow perturbations. Due to the absence of voids in the fuel loop during normal operation, this transient is due entirely to flow effects. Figure 24 shows the response of the CRNL—DWG 65-8S11A 120 100 i——-—- I R - T - i—ew —0—0=-0 FUEL o-0-00-8-8-0=8~ e-8-¢ PUMP /COOLANT WP e w SPEED PUMP ol 3 a0 e| ® SPEED _* G)’} O Pl < 7 / ¥ / _ o - E o [+] / / & > . COOLANT FLOW RATE wl o wl [« o LOGGER S ® OSCILLOGRAPH q 5 6 T 8 9 10 TIME (sec} Fig. 22. Pump Speed and Flow Startup Transients. PERCENT OF FULL SCALE 53 ORNL-DWG 65—-8942A T T T T [ T COOLANT SALT F%OW RATE | | — - : | | o LOGGER ‘ ® OSCILLOGRAPH ‘ “| - = | 1 I L* | | 20— """ _COOLANT PUMP SPEED ~' ‘ P ! ! " FUEL PUMP SPEED- e i o | | | 0 2 4 6 8 10 12 14 16 18 20 TIME (sec} Fig. 23. Pump Speed and Flow Coastdown Transients. ORNL-DWG &7-12323 20 So g 19 S . ° . FUEL PUMP STARTUP L © e * o'...o l.. ’g folele] .0....0..‘. L . '0.0 (o] S 18 —o * £ < o _ . [ ] = g o D o K e - * % o Qn = 16 Q0o L ] o S): %% 0000000 . 0%0 © FUEL PUMP COASTDOWN %65°%0% o 15 ——o0- L 00000 o° 00} 60400% ° . 02%000%60004 Te ® 14 0 10 20 30 40 50 60 70 TIME (sec) Fig., 24. Control-Rod Response to Fuel-Pump Startup and Coastdown. 54 control rod in attempting to maintain the reactor critical during fuel- pump startup and coastdown. The positive reactivity effect of the re- circulated precursors entering the core 13 sec after pump startup is clearly shown. 5.7 Conclusions from Dynamics Tests The two most significant conclusions obtained from the dynamic tests were: (1) these tests gave results which show good agreement with theo- retical predictions, giving increased confidence in the theoretical model and in the predictions for stable power operation, and (2) the selected testing procedures were, on the whole, quite satisfactory. Iater tests performed over the full range of operating power levels have provided additional evidence in support of these conclusions.2%t 6. GENERAL CONCLUSIONS This completes our account of the techniques and results of the program of zero-power physics measurements on the MSRE. Because the predicted neutronic characteristics were untested by experiment prior to the beginning of nuclear operation on June 1, 1965, these experiments provided an interesting test of the ability of then-available calcu- lational programs and neutron data for the nuclear design predictions of an unusual new system. The principal interest in the results of these measurements, however, stems from their applications in interpreting the neutronic behavior during operation of the reactor at power. The es- sential conclusions of this program can be summarized as follows: 1. The predicted minimum critical loading of 22U in the clean fuel salt agreed with the observed 27U loading, within the limits of uncertainty in the measured fuel-salt density at operating conditions. 2; In comparison to the calculated values of the remaining impor- tant functional neutronic characteristics of the reactor, the measured integral reactivity worths of the control rods were within 10%, the 2°°U concentration coefficient of reactivity was within 5%, and the temperature 55 coefficients of reactivity were within 20% of the calculated quantities. This provides confidence in pre-operational nuclear-safety studies which were based on these calculations. 3. The loss of reactivity due to the transport of delayed-neutron precursors by circulation at first apfieared to be smaller than would be expected from calculations. ©Subsequent analysis revealed the importance of including the top and bottom plenums of the reactor core in the theo- retical calculations of this effect. Although these are regions of low neutron importance, they are not "out" of the neutron flux, and the increased fuel-salt volume fraction and relatively long residence times in these regions both tend to enhance the reactivity contribution of delayed neutrons emitted in transit through these regions. k, At normal operating levels of salt in the fuel-pump bowl, and with the "'clean" fuel-salt conditions throughout the zero-power experi- ments, no evidence of circulating helium bubbles, entrained in the salt by the action of the pump-bowl spray-ring apparatus, was detected. In pressure transient tests at abnormally low pump-bowl levels, both the control-rod reactivity response and direct densitometer measurements indicated that 2% to 3% of voids by volume were in circulation with the fuel salt. 5. Dynamics tests at zero power, which were the initial measure- ments in a continuing program to assess the MSRE dynamic behavior at various stages of operation, gave no evidence to support any expectation of operational stability problems in the reactor system. lOI 11. 12. 13. 56 REFERENCES P. N. Haubenreich et al., MSRE Design and Operations Report: Part ITT. DNuclear Analysis, USAEC Report ORNL-TM-730, Oak Ridge National Laboratory, February 3%, 196k, MSR Program Semiann. Progr. Rept. Jan. 31, 1964, USAEC Report ORNL- %626, p. 54, Oak Ridge National Laboratory. MSR Program Semiann. Progr. Rept. Aug. 31, 1965, USAEC Report ORNL- 3872, pp. 30—31l, Oak Ridge National Laboratory. MSR Program Semiann. Progr. Rept. Aug. 31, 1965, USAEC Report ORNL- 3872, pp. 119121, Oak Ridge National Laboratory. A. F. Henry, "The Application of Reactor Kinetics to the Analysis of Experiments,™ Nucl. Sci. Fng., 3(1): 52—70 (January 1958). T. 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Blankenship 5%6. T. W. Kerlin R. Blumberg 57. S. 5. Kirslis E. G. Bohlmann 58. A. I. Krakoviak C. J. Borkowski 59. dJ. W. Krewson G. E. Boyd 60. R. C. Kryter R. B. Briggs 61. J. A. Lane R. H. Bryan 62. C. E. Larson J. M. Chandler 6%. R. B. Lindauer R. H. Chapman 6k. M. I. Lundin F. H. Clark 65. R. N. Lyon C. E. Clifford 66. H. G. MacPhnerson W. R. Cobb 67. R. E. MacPherson W. H. Cook 68. C. D. Martin L. T. Corbin 69. T. H. Mauney W. B. Cottrell 70. H. E. McCoy J. L. Crowley 71, H. C. McCurdy F. L. Culler, Jr. 2. H. . McDuffie D. G. Davis 73. C. K. McGlothlan R. J. DeBakker Th. A, J. Miller 5. J. Ditto >. R. L. Moore F. A. Doss 76. E. L. Nicholson W. P. BEatherly 77. L. C. Oakes J. R. Engel 78. A. M. Perry E. P. Epler 79. H. B. Piper D. E. Ferguson 80-84, B. E. Prince A. P, Fraas 85. G. L. Ragan D. N. Fry %. J. L. Redford J. H. Frye, Jr. 87. M. Richardson H. A. Friedman 88. R. C. Robertson C. H. Gabbard 89. J. C. Robinson R. B. Gallaher 90-114., M. W. Rosenthal W. R. Grimes 115. A. W. Savolainen A. G. Grindell 116. D. Scott, Jr. E. D. Gupton 117. J. H. Shaffer R. H. Guymon 118. E. G. Silver 60 119, M. J. Skinner 132. A, M, Weinberg 120, A. N. Smith 133, J. R. Weir 121, 0. L. Smith 134, K. W. West 122. P. G. Smith 135, M. E. Whatley 123, I. Spiewak 1%6. J. C. White 124. D. A. Sundberg 137. G. D. Whitman 125. R. C. Steffy, Jr. 138. Gale Young 126. H. H. Stone 1%39-141. Central Research Library 127. J. R. Tallackson 142-143, Y-12 Document Reference 128, R. E. Thoma Section 129. D. B. Trauger 144-209., Laboratory Records Department 130. C. 5. Walker 210. Laboratory Records Department, 131. B. H. Webster ORNL-RC External Distribution 2l11. C. B. Deering, USAEC, ORO, Oak Ridge 212-213, K. E. Elliott, USAEC, ORO, Oak Ridge 214, H. M. Roth, USAEC, ORO, Oak Ridge 215, J. W. Lewellyn, USAEC Division of Reactor Development and Technology, Washlngton, D. C. 216-22%3, T. W. McIntosh, USAEC, Division of Reactor Development and Technology, Washington, D. C. 224, E. E. Purvis, USAEC, Division of Reactor Development and Technology, Washington, D. C. 225. T. G. Schleiter, USAEC, Division of Reactor Development and Technology, Washington, D. C. 226, M. Shaw, USAEC, Division of Reactor Development and Technology, Washington, D. C. 227. Laboratory and University Division, USAEC, ORO, Oak Ridge 228-L48L. Given distribution as shown in TID-4500 under Reactor Technology category (25 copies — CFST1)