o OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPORATION NUCLEAR DIVISION ‘ L for the U.S. ATOMIC ENERGY COMMISSION ORNL- TM- 2997 COPY NO. - DATE - April, 1970 EXPERIMENTAL DYNAMIC ANALYSIS OF THE MSRE WITH 2337 FUEL R. C. Steffy, Jr. N\S\“ ABSTRACT Tests were performed on the Molten-Salt Reacter Experiment to deter- mine the system time response to step changes in reactivity, the neutron- flux-to-reactivity frequency response, and the outlet-temperature-to- power frequency response. The results of each of these were found to agree favorably with theoretical predictions. ' The time response tests were performed with the reactor operating at 1, 5, and 8 MW and substantiated the theoretical predictions that fol- lowing a reactivity perturbation the system would return to its original power level more rapidly at higher power levels than at lower power levels and was load-following at all significant power levels., A noisy flux sig- nal (caused by circulating voids) hampered detailed comparison of the experimental results and theoretical predictions. Neutron flux-to-reactivity frequency-response measurements were per- formed using periodic, pseudorandom binary and ternary sequences. This. type of test effectively prevented much of the random noise contamination of the neutron flux from entering the final analyses and gave results which contained little scatter. The results were in good agreement with the theoretical predictions and verified that for the MSRE, the degree of stability increases with power level, Outlet-temperature-to-power frequency-response measurements were com- pared with similar measurements made during operation with the 235U fuel and verified that the basic thermal properties of the reactor system were essentially the same as expected. Keywords: MSKRE, fused salts, reactors, operation, reactivity, testing, time response, frequency response, stability, pseudorandom binary sequences, Dseudorandom ternary sequences. NOTICE This document contains information of a preliminary nature and was prepared primarily for internal use at the Ogk Ridge National Laboratory. It is subject to revision or correction and therefore does not represent a final report. DISTRIBUTION OF THIS DOCUMENT IS UNLIMITED -—— LEGAL NOTICE R This report was prepared as an account of Government sponscred work. Neither the United States, nor the Commission, nor any person acting on behalf of the Commission: A. Makes any warranty or representation, expressed or implied, with respect to the accuracy, completensss, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of any information, apparatus, methed, or process disclosed in this report. As used in the above, '‘person acting on behalf of the Commission’ includes eny employee or contractor of the Commission, or employese of such contractor, to the extent that such employee or contracter of the Commission, or employee of such contractor prepares, disseminates, or provides access to, any information pursuant to his employment or contract with the Commission, or his employment with such contracter. CONTENTS ABSTRACT . v v v v v v v v e e e e e e e e e e e e e e e e e e INTRODUCTION . . v v v v v v e e e e e e e e e e e e e e e e e e TRANSIENT RESPONSE. . . . v v+ v v v v v v e e e v e e e e e e e FREQUENCY RESPONSE. . . v + v v+ v v v v v o 4 o o 0 o v o v« « .12 Neutron Flux to Reactivity . . . . . . . . . . .« . . . . . .12 Testing Procedure . . . . . « « v v v v 0 v e w e e e .. 12 Analysis Programs . . . . . . « « v 4 o+ 4 e e 4 4+ 4 o+ 13 Discussion. . . v v v v 4 e e e e e e e e e e e e e e .. L Outlet Temperature to Power. . . . . . . « « « « « « « « « . . 22 CONCIUSION. . . v v v v v vt e v e e e e e e e e e e e e e e e 25 LIST OF REFERENCES . « v v v v v v o v e e e e e e e e e e e e e w26 LEGAL NOTICE ————— This report was prepared as an accourt of Government sponsored work. Netther the Uniied States, nor the Commission, nor any person acting on behalf of the Commission: A. Makes any warranty or representation, expressed or implied, with respect to the aceu- racy, completeness, or usefulness of the {uformation contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately ovned vights; or B. Assumes any llabliltles witk respect to the use of, or fur damages resulting from the use of any information, apparatus, method, or process disclosed in this report, As used in the above, ‘‘person acting on behalf of the Commission’ includes any em- ployee or contracter of the Commissian, or employee of such contractor, to the extent that such employee or contracinr of the Commissiun, or employee of such contractor prepares. disseminates, or provides access to, any informativn puranant to his employment or contract with the Commisgion, or his employment with such contractor, g e DTG e BUTLION OF ThbsD oo e : EXPERIMENTAL DYNAMIC ANALYSIS OF THE MSRE WITH 23 FUEL R. C. Steffy, Jr. INTRODUCTION Several reports and articles (References 1 - 6) relating either to the theoretical or actual (or both) dynamic response of the Molten Salt Reactor Experiment have been published. However, none of these has re- ported in & concise form the dynamic response of the U-233 fueled MSEE. Reference 4 contains much of the frequency-response information reported herein, but it is presented in a lengthy context which is primarily con- cerned with comparing testing signals and techniques. The purpose of this report is to give a brief description of the observed dynamic re- sponse of the U-233 fueled MSRE, compare it with the theoretical and sug- gest possible reasons for differences when applicable, but to eschew any lengthy description of the testing technigues. TRANSTIENT RESPONSE A common method of describing the dynamic response of a stable sys- tem is to display the system response to a step change in an input vari- able. For a nuclear reactor, reactivity is usually the perturbed para- meter. This type description (i.e. description in the time domain) has the advantage of an intuitive appeal to people since we deal directly with time in day-to-day living. However, analysis of a system response in the time domain does have some disadvantages. Notably, if the system output of interest is contaminated with a large noise component, the part of the output resulting from a step input may be undiscernible from the part caused by the noise. The reason for making this point is the large difference in the neutron noise level between the Z°°U fuel loading and 2330 fuel loading of the MSRE. (The increase in noise level was due to a concomitant increase in circulating void fraction and was not an in- trinsic function of the fissile isotope.) An example of the uncontrolled neutron flux during high-power operation for each fuel is shown in Fig. 1, and the relationship between the flux noise and void fraction is readily observable., The void fraction estimates which are labeled on Fig. 1 were achieved by varying the fuel pump speed; however, the fuel pump was operated at full speed (~ 1180 rpm) for all of the dynamics tests reported here. During the initial approach to power with the #°7U fuel, time re- sponses of the neutron flux to a step change in reactivity were recorded and are shown in Figures 2,* 3, and 4 for the reactor at 1, 5, and 8 MWQ** respectively. Also shown in these figures are the theoretical predictions for step reactivity changes of the same magnitudes, The theoretical cal- culations were performed using the mathematical model and method described in Reference 2, The noisy flux signal hinders a comparison of the finer detail of the theoretical and experimental curves, but the noise was low enough that some features may be compared., In general, the theoretical and the experimental curves are in good agreement, For the 1-MW case (Figure 2), the initial flux peak was slightly nigher thar the theory predicted, then it oscillated below the initial level and later increased again with a second peak occurring after about 360 sec. The theoretical curves agree that the change in power should nave returned to a positive indication at this time but indicate that it should not have teen as large in magnitude as the observed behavior. The extent to which noise contaminatiorn forced the positive indication is not KNCwn, The noise contamination in the 5-MW case (Fig. 3) makes it diffi- cult to compare directly the experimental and theoretical results. They 3 The original plot of the response at 1 MW was made by a different machine than the other two plots. This accounts for the difference in general appearance of the plots. *¥ Full power was taken as 8.0 MW during the data analysis and writing o this repcrt. Pig. 1. ORNL-DWG 69-5374R '—; PERCENT RCENT PLRCEN PLRCENT o 0 50 60 10 0 40 50 40 50 ™~ £ ] £ 1 0 1[ ~ { L) J = : — | J; 1 , { PERCENT RCENT PERCFN PERCENT 0 50 60 10 0 40 50 40 50 {180 rpm 160 rpm 1120rpm 1070 rpm <0.1 vol % 0.6 vol% O0.3vol% OJvol% i ) 235U 233U RR-8{00 CHART ({percent of 15 Mw) Sections of Nuclear Power Recorder Chart Contrasting 275U Fuel, Full Flow and Few Bubbles with 73 Fuel, Varying Flow and Bubble Fraction. Conditions in each case: T MW, 1210°F, 5 psig, 52 - 56% Fuel Pump Level. ORNL-DWG 70-2922 0.7 | | | 1 0.6 POWER LEVEL =1 Mw . Ejl — — — THEORETICAL ' j \;L EXPERIMENTAL 0.4 ,. . REACTIVITY INSERTED=0.0139 % 34 — 3 \ = 0.3 < I 0.2 0.1 N“_’J-L__‘- 5 S~ —n _iilpfmjfl:“:_m.— ~0.4 0 60 120 180 240 300 360 420 TIME AFTER REACTIVITY INSERTION (sec) Fig. 2. Response of the Neutron Flux to a Step Change in Reactivity of 0.0139% Bk/k with the Reactor Initially at 1 MW. AMw ORNL-DWG 70-2921 08 | | | | | POWER LEVEL =5 Mw 06 --— THEORETICAL ] ' —— EXPERIMENTAL REACTIVITY INSERTED=0.0190 % 8k/k 0.4 > /\N"’\ \/\W\/\A /\ O ----- s sm e === v v == -0.2 0 60 120 180 240 300 360 420 TIME AFTER REACTIVITY INSERTION (sec) Fig. 3. Response of the Neutron Flux to a Step Change in Reactivity of 0.0190% &k/k with the Reactor Initially at 5 MW, are 1in general agreement, but detailed comparison would be guess-work, The swells and rolls that occur after about 150 sec are almost surely not directly related tc the original reactivity input since the system set- tling time at 5 MW is about 150 sec, For the reactor operating at 8 MW, the flux response to a reactivity step of 0.0248% 8k/k is shown in Figure 4. The maximum power level was reached during the first second after the reactivity input. This rapid incregse was accompanied by a rapid increase in fuel temperature in the core, which, coupled with the negative temperature coefficient of reac- tivity, more than counter-balanced the step reactivity input, so the power level began to decrease. The temperature of the salt entering the core was constant during this interval, and when the power had decreased enough for the reactivity associated with the increased nuclear average tempera- ture to just cancel the step reactivity input, the power leveled for a brief time (from ~ 6 to ~ 17 sec after the reactivity input). About 17 sec after the reactivity increase, the hot fluid generated 1n the initial power increase completed its circuilt of the loop external to the core, and the negative temperature coefficient of the salt again reduced the reac- tivity so that the power level startec down again. At large times the reactor power returned to its initial level, and the step reactivity in- put was counter-balanced by an increase in the nuclear average temperature in the core. For the 5-MW case, a short plateau was probably present .50, but the noisy signal obscured its presence. At lower powers, how- ever, the slower system response prevented the reactor from reaching the peak of its first oscillaticn tefcre the fuel completed one circuit of the external fuel loop. The plateau therefore did not appear in the 1-MW case. An important characteristic of the MSRE dynamic response was that as the power decreased the reactor tecame both more sluggish (slower respond- ing) and more oscillatory; that is, at low powers the time required for oscillations to die out was much larger than at higher powers, and the fractional amplitude of the oscillations (A power/power) was larger. 11 CRNL-DWG 70-2923 1.4 ] . | T T POWER LEVEL = 8 Mw 1.2 -—- THEORETICAL L0 —— EXPERIMENTAL ' REACTIVITY INSERTED = 0.0248% 8k/k 0.8 2 i = 06 f‘\ <] I \ W, 0.4 ¥ 0.2 . 0 M‘l' -0.2 0 20 40 60 80 100 TIME AFTER REACTIVITY INSERTION (sec) Fig. 4. Response of the Neutron Flux to a Step Change in Reactivity of 0.248% 8k/k with the Reactor Initially at 8 MW. 12 FREQUENCY RESPONSE Neutron Flux to Reactivity Most of the effort in experimentally determining the dynamic response of the MSRE was expended in determining the neutron-flux-to-reactivity frequency response. One advantage of working in the frequency domain 1is that a periodic waveform may be continuously imposed on a system input (e.g. reactivity, through control rod movement) until several periods of data have been collected. All of the signal power of a periodic signal is concentrated at harmonic frequencies, and subsequent analysis at a harmonic frequency very efficiently eliminates most of the noise contami- nation which is usually dispersed over a wide frequency band. There are other advantages to wecrking in the frequency domain, but the more noisy flux signal with the 33 fuel loading makes this a salient advantage. Several step and pulse tests (aperiodic tests) were also attempted but these do rot have the signal energy concentrated at particular frequencies and tke system noise was large enough that the results contained too much scatter to bhe useful,. Testing Procedure "he periodic signals used in the frequency-response tests were either pseudorandonr binary or pseudorandom ternary sequences.? These are par- ticular series of square wave pulses that were chosen because they evenly distributed the signal power at the harmonic frequencies over a wide fre- guency range, which permitted determination cf the freguency response cver a wide spectrum with conly one test. The frequency range over which we obtained freguency-response results was from about 0.005 to 0.8 rad/sec, The lower limit was set by the length of one period of the test pattern and the high-frequency limit was determined by the time width of the square wave pulse of shortest duration which the standard equipment would adequately reproduce. The shortest basic pulse width used in these tests was 3.0 sec. The frequency range covered by these tests was essentially the range cver which thermal feedback effecis are important. 13 The on-line computer, a Bunker-Ramo 340, was programmed to generate the sequences by opening and closing a set of relays. Voltage was fed through the relays from an analog computer (Electronic Associates, Inc., Model TR-10). This voltage was used to determine the mcvement of the con- trol rods, which were forced either to follow the pseudorandom test pat- tern themselves or to cause the flux to follow the test pattern.® The control-rod position and the neutron flux were digitized and reccrded every 0.25 sec on maghetic tape., The data were retrieved from the tape and stored on punched cards which could then be processed with the anal- ysls programs to yleld the frequency-response information. Analysis Programs Before discussing each of the programs used to analyze the data, it is pertinent to note that in some instances the different analysis pro- grams yielded markedly different results when applied to the same data,. It is beyond the intent of this report to delve into the possible theo- retical explanations, but the interested reader may consult Reference 4 for a more complete treatise on the subject. FOURCO.® This code directly Fourier transformed the time records., The transformed output (flux) was then divided by the trans- formed input (rod position) to give the frequency response., This analysis was usually performed on the full data record, which would contain several periods of the same waveform, but occasionally was performed on individual periods of data with the several resulting answers then ensemble gveraged. This latter method is denoted FOURCO ENGEMBIE on the figures, CPSD.”?»©® This analysis method utilized a digital simulaticn of an analog filtering technigue for obtaining cross-power spectral density, CPSD, functions. This code calculated the power spectrum of the input signal and the cross-power spectrum of the input and output signals and divided the cross-power gpectrum by the input power spectrum to obtain the frequency response at each frequency of analysis. The key feature of this code is an adjustable filter width about the analysis frequency. CABS.7 The third calculational procedure was more involved, The auto-correlation functions of the input and output signals were calcu- lated and the cross-correlation function of the signals was calculated. 1h These were then Fcurier transformed to obtain the input, output, and cross-peower spectra. The input power-spectrum was then divided into the crcoss-power spectrum to obtain the frequency response. Discussion With the fuel stationary, the frequency response of the zero-power MSRE was essentially the same as that of any stationary-fuel, zero-power, £33J-fueled reactor. The measured frequency response with the fuel not circulating is shown in Figure 5. The magnitude ratio, Bn/NO-Sk, is seen Tc be in general agreement with the theory, but the phase angle is rot in particularly good agreement. At the higher frequencies for tests at all power levels, the magnitude ratio and the phase angle were lower than the theoretical. This is thought to have been caused by the control rod not adequately following the test pattern yet giving the indication that it was. The indicators, which are physically located with the drive assembly, accurately display the action of the rod-drive motors; however, the flexibility of the control rod makes it doubtful that the tip of the rod, which 1s about 17 ft from the drive assembly, reprcduces the high frequency component of the rod-drive movement. The results of a typical zero-power test with the fuel circulating are shown in Figure €. The shape of the magnitude ratic curve is in ex- cellent agreement with the theoretical curve, but the results have been normalized by multipiying each experimental value by 1.75. The phase angle data was in better agreement with the theoretical predictions than wag the case for the non-circulating data, but there is scatter in the results. The need to normalize some results and not to normalize others is 2ls0 considered to be caused by poor contrel rod indication.® The nor- malization was not Power dependent since some data did and some did not need normalization at each power level, and the normalization factors, when they were required, were different for different tests. As we mentioned in the intrecduction, several different testing tech- niques were used in obtaining the experimental results. An example of 8n 15 ORNL -DWG 69-12050 FUEL STATIONARY POWER LEVEL - ZERO g [ s} ‘o > ANALYSIS METHODS s CPSD a FOURCO ENSEMBLE — THEORY / o ,””. ® —30 - © _‘/ o 7 /4’1" "2 fifi"- 7 1 441182 % ‘fi "B 2 5 0% 2 5 10 2 FREQUENCY (rad/sec) Fig. 7. Neutron Flux-to-Reactivity Frequency Response of the 2335-Fueled MSRE at 1 MW. - 19 ORNL—-DWG 69—-12245 10% PHASE (deg) POWER LEVEL — 5Mw ANALYSIS METHODS v FOURCO, CABS (EACH GAVE SAME RESULTS) — THEORY 30 7N % o N ) /‘\ %W?% "-n-.....‘\.. -30 Y o 7 -60 ~ W 10~3 102 10" 1 FREQUENCY (rad/sec) Fig. 8. Neutron Flux-to-Reactivity Frequency Response of the 233j-Fueled MSRE at 5 MW. 20 4 ORNL -DWG 69-12246 POWER LEVEL =8Mw ANALYSIS METHODS o FOURCO, CABS, CPSD (EACH GAVE SAME RESULTS) - THEORY x | mao o s3] Z i o < a 0 G N ik%%: % %%5 \N\ n —-30 1073 2 5 4072 2 5 107! 2 5 100 FREQUENCY (rad/sec) Fig. 9. DNeutron Flux-to-Reactivity Frequency Response of the 233-Fueled MSRE at 8 MW. 21 to the core one pericd later and, because of the negative temperature coef- ficient of reactivity, produced a reactivity feedback effect that partially canceled the external perturbation, The dip i1s obviously present in the experimental results as well as in the theoretical curves; however, the dip in the experimental data 1s not as pronounced as the theory predicts, Since the magnitude of the dip has been shown® to be a function of the amount of salt mixing which occurs as the fuel circulates around the loop, this difference between the experimental and theoretical implies that not enough mixing was assumed in the theoretical model. Additional work with the theoretical model has shown that 1f the salt transport in the piping is represented by a series of 2-sec first-order lags (well-stirred tanks with mean holdup times of 2 sec) rather than the pure delays that were assumed in the earlier work, the dip in the experimental and theoretical responses are in good agreement. Below about 0.5 rad/sec, the magnitude ratic decreases as the power is increased. This substantiates the observation drawn from the time response plots; the degree of stability for the MSRE increases with power level, The lower magnitude ratic at the higher power levels over the frequency range in which thermal effects are important says, in effect, that for the same change in reactivity the fractional power (A power/power) change will be less at higher power. The frequency-response curves shown in this document display the MSRE 's frequency response at several power levels. Of course, several tests were performed at several different power levels, but in order to keep the presentation as straightforward as possible, we chose to show the results from representative tests. Table 1 summarizes the freguency-= response tests performed with the £33 fuel loading and indicates the scope of the testing program which included 28 different tests of approxi- mately one-hour duration each. Other experimental results for the &3] fuel loading are given in References 4 and 5. Complete results of theo- retical dynamic analyses are given in References 2, 5, and 6. Note that some tests were performed shortly after the start of operation with 2°-U fuel, and others were performed near the end of operation with 27U fuel. There were no indications that the response of the reactor had changed with operating time, 22 Table 1 Information Related to Frequency-Response Testing of ©2°U-Fueled MSEE Integrated No., of Testing Power Power Tests Dates (MW-hrs) Level Performed 10/15/68 0 100 W 1 11/7-8/68 0 50 W 6 1/16/69 86 1 MW 3 1/20/69 435 5 MW 1 2/3/69 2,390 8 MW 1 2/17/69 L, 080 5 MW 1 2/20/69 L, L90 8 MW 3 3/11/69 7,220 10 kW 1 L/2L /69 14,000 8 MW 2 5/26/69* 19,500 8 MW 9 .X. These tests were performed for M. R. Buckner and T. W. Kerlin of the University of Tennessee as part of a gradugte studles program. Outlet Temperature to Fower During the neutron-flux-to-reactivity frequency-response tests which were conducted at significant power levels, the response of a thermocouple (TE-100-1A) on the outlet pipe was also recorded. The data records then included power (or more specifically, neutron flux) and outlet temperature during a time in which the power was varied in a periodic waveform. Hence, the outlet-temperature-to-power freguency response could be determined at the same harmonic frequencies as the neutron-flux-to-reactivity frequency 23 response. The results of this determination could then be compared with the results of theoretical predictions.® The outlet-temperature-to-power frequency-response results.from a test conducted during operation with 25U fuel as well as two tests per- formed during operation with 277U fuel are shown in Figure 10. The experi- mental results of all three tests are essentially the same. This should be expected since the temperature response to a given change in power is a function of the thermal properties of the system, and these were changed very little with the change in fissionable material. Three theoretical magnitude ratio plots are alsc shown in Figure 10. Curve 1 is the as-calculated curve and curves 2 and 3 are this same curve multiplied by 0.5 and 0.1, respectively. Normalization of the theoretical by multiplying by 0.5 forces agreement with the experimental results at low frequencies and multiplying by 0.l forces agreement at high frequencies. The reason for the discrepancies between the experimental and theoretical have not been explained leaving this as an area open for more analysis. It is of interest to note that in some experimental work!'® performed by S. J. Ball and T. W. Kerlin in which they attempted to determine the re- sponse of outlet-temperature-to-inlet-~temperature perturbations, they too found a larger degree of attenuation than had been theoretically predicted. The phase angle plots shown in Figure 10 are in good agreement if the theoretical thermocouple response to a power perturbation is delayed by 0.7 sec more than was assumed in the original calculation. (A pure delay gives a phase shift that changes linearly with frequency.) The theoretical response of the thermocouple was represented by a l-sec pure delay plus a 5-sec first-order lag. This was based on calculations per- formed by 8. J. Ball.'! This represents a good estimate, but could be in error by 0.7 sec for this particular thermocouple depending on its particular response characteristics and physical contact with the pipe. Another possible source of error is the estimate of the location of the thermocouple on the pipe. The experimentally-measured outlet-temperature-to-power frequency response verified that the basic thermal properties of the MSRE were es- sentially unchanged by the change in fuel loading. The disagreement (8T/3n) (°F/Mw) PHASE ANGLE (deg) Fig. 2k ORNL -DWG 70-3423 100 50 o FROM TEST ON 23%U—FUELED SYSTEM ® FROM TEST ON 233U~FUELED SYSTEM s FROM TEST ON 233U~FUELED SYSTEM 20 10 0.5 - (D THEORETICAL @ THEORETICAL x 0.5 ® THEORETICAL x 0.1 0.2 ~60 S — %" “ba ™ o \r\\~ '“’%a? N -120 -180 %q,: — THEORETICAL \ —— THEORETICAL WITH o —-240 ADDITIONAL O.70sec TIME LAG L] 1073 2 5 4072 2 5 107! 2 5 100 FREQUENCY (rad/sec) Sl ] 10. Outlet Temperature-to-Power Frequency Response of the MSRE with the Reactor at 8 MW. 25 between the theoretical and experimental magnitude ratio determinations makes 1t meaningless to draw any conclusions about the mixing effects in the circulating system. CONCIUS ION The dynamic response of the £3Fj-fueled MSRE was analyzed by three different methods, each of which had deficiencies but each of which added information. The transient response of the neutron flux to a step change in reactivity at various power levels verified that the general response of the system was as anticipated, but the noisy flux signal made detailed comparison of the theoretical and experimental results difficult. The shape of the experimentally-determined neutron flux-to-reactivity fre- quency-respense curves was in excellent agreement with the theoretical curves over most of the frequency range which was realizable with the in=- stalled hardware. There were problems assoclated with finding a test method which would give good results, and erroneous control rod position indications necessitated normalization of some experimental results. The outlet temperature-to-power frequency-response determination did not agree well with theory but did show that the basic thermal properties of the MSRE were essentially unchanged by the change from 235 to 27 fuel. At high powers, the MSRE is a highly damped system. It returns to its original power level rapidly with no undershoot or wallowing. At low power levels, the uncontrolled reactor tends to be sluggish and slow in returning to its original power level. With the reactor at 1 MW, it was observed that over LOO sec was required for the flux level to stabi- lize after a step change in reactivity. In summary, the MSRE was stable at all power levels and the stability increased with power as predicted. 1. 10. 11, 26 LIST OF REFERENCES S. J. Ball and T, W. Kerlin, Stability Analysis of the Molten-Salt Reactor Experiment, USAEC Report ORNL-TM-1070, Oak Ridge National Laboratory, (December 1965). R. C. Steffy, Jr., and P. J. Wood, Theoretical Dynamic Analysis of the MSRE with U-233 Fuel, USAEC Report CRNL-TM-25T71, Oak Ridge National Laboratory (July 1969). T. W. Kerlin and S. J. Ball, Experimental Dynamic Analysis of the Molten-Salt Reactor Experiment, USAEC Report ORNL-TM-1647, Oak Ridge National Laboratory, (October 1966). R. C. Steffy, Jr., Frequency-Response Testing of the Molten-Salt Reactor Experiment, USAEC Report ORNL-TM-2823, Oak Ridge National Laboratory (March 1970). MSR Program Semiann. Progr. Rept., Feb. 28, 1969, USAEC Report ORNL-4396, Osk Ridge National Laboratory. MSR Program Semiann. Progr. Rept., Aug. 31, 1968, USAEC Report ORNI-L3LL, Oak Ridge National Laboratory, pp. 46 - 52. S. J. Ball, A Digital Filtering Technigue for Efficient Fourier Transform Calculations, USAEC Report ORNL-TM-1662, Oak Ridge National Laboratory, (July 1967). S. J. Ball, Instrumentatlon and Control Systems Division Annual Pro- gress Report, September 1, 1965, USAEC Report ORNL-3875, pp. 126-127, Oak Ridge National Iaboratory (September 1965). T. W. Kerlin and J. L. Lucius, CABS — A Fortran Computer Program for Calculating Correlation Functions, Power Spectra, and the Frequency Response from Experimental Data, USAEC Report ORNL-TM-1663, Oak Ridge National Laboratory, (September 1966). MSR Program Semiann. Progr. Rept., Feb. 28, 1966, USAEC Report ORNL-3936, Osk Ridge National Laboratory. S. J. Ball, Personal Communication to R. C. Steffy, Jr., July 2L, 1968. 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