ML19345G657

Pulstar Pulse Tests, Interim Rept

ML19345G657
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Site:	University of Buffalo
Issue date:	10/09/1964
From:	Lumb R, Macphee J NEW YORK, STATE UNIV. OF, BUFFALO, NY
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WNY-017, WNY-17, NUDOCS 8104080582
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PULSTAR PULSE TESTS October 9, 1964 Report No. WNY-017 l

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b Interim Report i

i, PULSTAR PULSE TESTS

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R. F. Lumb I

J. Mac Phee*

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Western New York Nuclear Research Center, Inc.

October 9, 1964 j

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• Project Manager,. AMF Atomics, York, Pennsylvania.

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o WESTERN NEW YORK NUCLEAR RESE/JtCH CENTER, INC.

A Subsidiary of State University of New York Ralph F. Lumb Power Drive Director BuffWiev New York October 9,1964 Dr. Richard L. Doan, Director Division of Reactor Licensing United States Atomic Energy Coanission Washington 25, D. C.

SUBJECT:

Docket 50-57 Reference is made to Appendix A to License R-77, Technical Specifications. In accordance with Section P-le of these Technical Specifications, the pulse testing of our reactor has been temporarily terminated. A detailed report of test results is transmitted herewith.

Your early review and comment on the proposed continuation of the pulse testing program would be appreciated.

Sincerely, l

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Acknowledgement It is not possible to recognize all contributions to the test program and this report. However, the advice and assistance of the dedicated staff of

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AMF Atomics is deeply appre:iated.

We are also indebted to staff members at SPERT; their critical courent and advice has been est helpful.

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ERRATTA Interim Report - PUISTAR Pulse Tests October 9, 1964 Page 16 Change to page 18.

Page 18 Change to page 16.

Page 41 1st paragraph (number 5) change msee to psec in two places.

Figure 5 Change reference series IV to series VI.

Figure 9 Sicpe should be 2.91 x 10-5

,,c, Reference (1)

Change Water Reflecting Reacter to Water Reflected Reactor.

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TABLE OF CONTENTS Page Introduction and Summary A Objective 1

B Scope 2

Section 1 - Steady State Tests A Introduction 7

B Excess Reactivity 7

C Control Rod Worth 8

D Fuel Element Worth 9

E Pownr Distribution 9

F Void Coefficient 11 G Moderator Temperature Coefficient 11 H Forced Convection Power Run 13 I Natural Convection Run 14 J Discussion 15 K Conclusions 21 Section II - Pulse Tests A Introduction 23 8 Chronological Events 23 1.

Ramp Tests 23 2.

Pulse Tests 24 C Test Results 27 1.

Ramp Tests 27 2.

Fulse Tests 27 a-Pulse Rod Calibration 28 b-Power Output 29 c-Energy Release 30 d-Test Pin Temperature 31 e-Pressure 31 f-Clad Strain 32' g-Test Pin Dimensions 32 D Analysis of Data 32 1.

Prompt Neutron Lifetime 32 2.

Power Data

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Energy 35 4.

Burst Shape 36 5.

Fue1 Surface Temperature 37 6.

Pressure 38 7.

UO2 Temperature 39 8.

Doppler Coefficient 39 E Concissions 40 i

TABLE OF CONTENTS-Continued i

Page Section III - Proposed Tests A Introduction 42 i

B Proposed Tests 42 C Proposed Limitations 43 D Sunnary 45 Tables and Figures References t

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LIST OF TABLES AND FIGURES Table I

Pulse Data Table II Repetative Pulse Table III Data on input Reactivity Table IV Summary of Pellet Calibration Runs Figure 1 Pulse Test Core Figure 2 Temperature Coef ficient Curves Figure 3 Log-N Trace of Ramp Run Figure 4 24" Fulse Rod Calibration Curves Figure 5 Pulse Rod Position vs. Reciprocal Period Figure 6 Linear Plot of 5.3 msee Pulse Figure 7 Peak Power vs. Reciprocal Period Figure 8 Energy to Peak Power vs. Reciprocal Period Figure 9 Prompt Reactivity vs. Reciprocal Period (From Shim Rod Calibration)

Figure 10 Prompt Reactivity vs. Reciprocal Period (From Pulse Rod Calibration)

Figure 11 Peak Power vs. Reciprocal Period Figure 12 Test Pin Temperature vs. Time-5.3 msec Pulse iii

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INTRODUCTION AND SUED.ARY A OBJECTIVE On September 27, 1963, the Western New York Nuc1 car Research Center, Incorporated (WNYNRCI) requested authority to install a PULSTAR reactor.

Subsequently,the submittal dated May 28, 1964 requested authorization to (1) conduct a test program comprising steady state tests aticw power and transient tests involving pulses of an energy release up to 90 Mw-sec., (2) operate routinely in the steady state mode to power levels up to 2 megawatts, (3) operate routinely in the pulse mode at energy release pulses up to 40 Mw-sec.

On June 19, 1964, the WNYNRCI was authorized by the Commission to carry out the steady state program and to operate routinely at 2 meEawatts. Authorization to conduct a limited transient program was issued on July 17, 1964.

The Technical Specifications issued in conjunction with this authori-zation limited the transient test program to pulses of either 40 Mw-see or 5 maec and required that a written report be submitted to the Commission when either of these limits were reached.

The PULSTAR core was installed in June and achieved initial criticality on June 22, 1964. Following completion of the steady state tests, the pulse program was initiated on July 31, 1964. On September 28, 1964, the reactor was pulsed on a period of 5.2 msee to a power level of 1050 MW and a total energy release in the pulse of 23 Mw-sec. Follow-ing 9 repetative pulses at this power level and perica,' the transient pulse program was suspended in accordance with the Technical

.9pecifications. This report has been prepared in compliance with

Technical Specifications and with the cbjectives of docucenting the results of the test program and requesting authorization to continue the transient test program in accordance with the request to the Cocnission dated May 28,19%.

B SCOPE In keeping with the stated objectives, the scope of this report en-compasses a presentation and analysis of expericental data, with particular emphasis on the transient tests and the steady state tests which had a direct bearing on transient behavior; and the derivation of operating limits proposed to cover subsequent tests. Accordingly, the report has been divided into three sections covering steady state tests, transient tests, and proposed operating limits. Sections I and II covering steady state and transient tests respectively, are concerned primarily with the operating performance of the reactor, its bearing on the operational safety of the facility, and the relationship of the test results to ocr understanding of the basic cechanisms which govern core behavior. Operating limits proposed for the balance of the test program are the subject of Section III.

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SUMMARY

AND CONCLUSION During the period from June through September,19%, the PUI. STAR core was installed and successfully tested up to the operating limits contained in the Technical Specifications.. Steady state tests carried out in-cluded excess reactivity, fuel' element worth and control rod worth, and measurements for a variety-of core configurations. Based on these 2

tests, a 20 element symmetrical configuration was selected as a basic core for more detailed steady state tests and the subsequent transient tests.

The detailed steady state tests included measurement of power distri-bution, void coefficient, moderator temperature coefficient, calori-metric calibration of nuclear instruments, and validation of the presence of a negative Doppler coefficient of reactivity. In the case of some static parameters, which do not have a significant effect on transient behavior, deviations were observed between measured and calculated values. Of particular note in this regard is the observed excess reactivity which is significantly higher thaa predicted for a given core size. This deviation is attributed to the fact that the contribution of epithermal U-235 fissions, knewingly neglected in the core design calculations, is significant.

Also of particular note is the steady state test in which the reactor was run at a few hundred killowatts. In addition to providing a calorimetric calibration of nucicar inatruments, these tests have con-vincingly verified the existence of the powerful shutdown mechanism resulting from the Doppler effect. Since this initial power run, the reactor has been operated for short periods at steady state power levels -

up to 1 MW to check instrument calibration.

With regard to steady state operation, it can be concluded that the test program carried out has demonstrated that the PULSTAR core is 3

superior to its predecessor and that its perforcuince is understood.

The transient tests were initiated with the calibration of the pulse rod. This was followed by a series of ramp reactivity insertions with che control system in the steady state mode. These ramp tests, initiated with a high speed shis rod, first demonstrated the shutdown capability of the Doppler effect. The minimum period achieved was 2.5 seconds and the corresponding reactor peak power level required for complete quenching of the excursion was a modest 530 kilowatts.

On August 12, the first of 100 pulses, utilizing the pulse rod and with the control system in the transient mode, was initiated. The first two such pulses run duplicated the two preceding ramp induced excursions and provided a reference for extrapolation to more energetic pulses.

The pulse output was gradually increased in accordance with the conditions contained in the Technical Specifications until, on September 28, the reactor'was pulsed to 1050 W on a 5.2 maec period. Following nine repetitive pulses at this period, the program was suspended in accordance with the Technical Specifications.

' he maximum transient generated during thA test program was a peak T

power of-1150 m at a 5.16 maec period. The energy released in the

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pulse was 26.5_Mw-sec. This corresponds to an energy of 47.7 Mw-sec in the test pins. The maximum clad temperature of the test pins after this pulse was a modest 272*F.

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I The pulsing performance of the core is entirely explainable and gen-erally is more conservative than assu=ed in the safety analysis report.

Prior to the test program, the lack of precise knowledge of one of the core parameters, prompt neutron lifetice, lead to an underestimate of reactor power and energy for a given period. Wile this underestimate did not reflect an error in predicting the relation between energy release and input reactivity, it did result in a departure frca the control curves used to monitor reactor performance. Consequently, it was necessary to request authorization, which was granted on Septecher 22, 1964, to revise the control curves.

A comparison of the actual reactor performance with the predicted performance, revised in accordance with the observed prompt neutron lifetime, indicates that the basic behavior of the reactor in the pulsing mode is well understood and can be predicted with confidence.

Ia support of this finding it can be noted that least squares linear fits of both energy to peak power and peak povsr versus reciprocal period are within 107. and 207. respectively of the predicted value for a period of 5 maec. The standard deviation from the mean is i 16.57.

in the case of pcwer versus reciprocal period and i 11.67, in the case of energy versus reciprocal period. It can be concluded that the reactor transient performance is in accordance with our understanding in general, and in particular has paralleled the perfortaance of the SPERT I oxide core. The output of the PULSTAR core was considerably less than SPERT I with regard to total energy and energy release per gram of UO, for a given reactor period. In addition, and as expected, 2

the PULSTAR shutdown coefficient expressed in 4/th-sec is significantly larger than in SPERT.

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With respect to subsequent pulsing tests, it is proposed to gradually incesase reactor output up to a nominal power level of 10,000 MW.

The estimated period and total energy release corresponding to this power level are 1.92 msec and 72.5 Hw-sec respectively. The size of the steps staken in approaching this output would be determined as described in the Technical Specifications. In addition, it is proposed that a 95% con-fidence interval be used in evaluating whether or not a particular pulse is in control. We believe that such a system for monitoring core performance is now feasible in light of the experimental data collected to date.

In conclusion, we believe that the balance of the test program can be carried out with no jeopardy to the safety of either the general public or the operating staff.

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SECTION I STEADY STATE TESTS A.

INTRODUCTION This section covers a presentation and discussion of the steady state tests which were carried out or the PULSIAR core. However, the full extent of these tests is not reflected here, in that only those tests having a bearing on the subsequent transient test program are empha-sized. The pre-critical steady state tests involving measurement of rod blackness, rod drop time, magnet release time, and check out of the nuclear instruments were completed and core loading commenced on June 22, 1964.

Initial criticality was achieved on that same day.

During the following period to July 31, a series of steady state tests to measure excess reactivity, fuel worth, rod worth, power distribution, and void and moderator tearsrature coefficient, were carried out to-gether with power calibration runs.. In subsequent paragraphs nach of these tests are described together with a discussion of the tests re-suits and a summary.

B.

EXCESS REACTIVIIY The excess reactivity of over 20 different core configurations was measured during the course of the program.. As a result of these mea-surements, the 20-element symmetrical configuration shown in Figure 1 was selected as the reference core to be used in the transient program.

A total of six-test pins were installed in the reflector region of this core as also shown in Figure 1.

These test pins were located such that they would lead the core in power density by a factor of at least 1.2.

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The excess reactivity of this core was measured to be 5.55% aK/K in-cluding the worth of the test pins. A comparison of the measured versus calculated values of excess reactivity as a function of core size indicates the meas *2 red excess reactivity is significantly higher than anticipated.

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CONTROL ROD WORTH As shown in Figure 1, the control rods were arranged in an in-line con-figuration as compared to the staggered array assuced in 'he 30-element configuration used in the calculation of control rod worth. The worth of the control rods in the in-line configuration was measured for a variety of core configurations by the period and rod drop methods.

These measurements indicated rod worth varies between 0.9 and 5.0% de-pending on rod position and core' configuration. For the core configura-tion shown in Figure 1 the total worth of the control rods is greater than 12% as ceapared with approximately 27% calculated for the staggered array in a 30-element core. A direct comparison of these measured and calculated values is not possible because of the significant difference in the two configurations. For example, two of the control rods in the measured array are only partially covered by fuel; whereas, all-the control rods in the calculated array were covered by fuel. Never-theless, a comparison of individual rod worths would appear to indicate that actual rod worth is about 70% of the calculated value. Although an underestimate _of control rod worth can have serious consequences in a marginal design, the PULSTAR core has ample rod worth to insure an adequate shutdown margin. For example, in the core shown in Figure 1, 8

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rod worth is 3.9%.

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FUEL ELEMENT WORTH 1

The reactivity worth of a single fuel element was measured in a variety of positions and core configurations. These measurements showed excell-eat agreement with calculations and confirmed that the worth of a central element is approximately 2% AK/K as compared to a much higher worth of a corresponding element in an aluminum plate-type core. This lower fuel element worth is an important factor which contributes to the greater safety of this core as compared to its predecessor.

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POWER DISTRIBUTION i

Experiments were carried out to measure power distribution with the objective of determining the following power distribution ratios in the core:

i) Maximum to average in the core j

.ii) Maximum in test pin to maximum in core iii) Average in core to average in pellet i

Three types of experiments were run to determine these values: (1)

. activation of copper wires in the core; (2) activation of fresh fuel pins in the core and. test' pin assemblies; (3) calibration of the pellet I

by temperature measurements with the teactor at 150 Kw.

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Data from the copper activation runs indicates an overall maximum to average in the core of 2.82 and a maximum test pin to maximum core ratio of 1.86.

Because of tha large thermal flux peak in the reflector and resulting gradient at the edge of the core, it was decided to check core power distribution and the ratio of the maximum pin to maximum core ratio using fuel pin activation. A representativa number of fresh fuel pins were activated in the core, and their activity measured. The data from these runs were in good agreement with the data from the copper activation experiments and indicated a maximum test pin to maximum core ratio o f 1.80.

Neither the copper activations nor pin activation techniques were con-sidered sufficiently accurate to calibrate the pellet; therefore, a third method involving operation of the reactor at 150 Kw was used.

Following calorimetric calit ration of the nuclear instruments, the reactor was run at 150 Kw and the pellet capsule was lowered into its monitoring position and the pellet temperature was recorded. After approximately one minute th..' reactor was scrammed and the pellet re-moved from the core. The energy release in the pellet in watt seconds per gram over a given tLee interval was determined from the temperature trace. The corresponding energ7. release in the core in watt-se;onds per gram was determined from the time interval, reactor power, and the mass of 002 in the core. The results of this experiment indicated that the ratio of core average to pellet average specific energy density ic 0.453.

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F.

VOID COEFFICIENT The void coefficient of reactivity in the reference core was measured in six representative fuel positions. Based on these measurements, the average void coefficient of reactivity was measured to be 0.39%

AK per percent void as compared to a value of 0.9% AK per percent void calculated for a 30-element core. Although the low value of measured void coefficient was something of a surprise, it, nevertheless, is significantly higher than that measured in the SPERT I oxide core

(-0.2847. aK per percent void) and the aluminum plate-type core pre-viously operated.

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MODERATOR TEMPERATURE COEFFICIENT Considerable effort was expended in attempting to determine the cod-erator temperature coefficient of reactivity.

The bulk pool tempera-ture coefficient of reactivity was first measured by varying pool temperature over the range from 65 to 1300 F.

The results of this mea-surement was quite similar to aluminum plate-type cores in that the bulk coefficient was positive from 650 to 806 F, and negative above 0

60.F with the absolute value increasing with temperature. The second measurement carried out involved an attempt to measure the temperature coefficient of the reflector. This test was carried out with a flat tank approximately 3" x 18" x 24" placed adjacent to one core face. The temperature of the water in the tank was varied and reactivity change was measured as a function of the temperature of the water inside the tank. This test was carried out at a pool temperature of 108.5'o F and the. temperature inside the tank was varied between 620 F and 124 F.

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Analysis of this data indicated that tha reflector te=perature coeffi-cient of reactivity is positive. Further= ore, extrapolation of the value measured for one core face to five core faces using weighting factors based on relative reflector reactivity sorths results in a total reflector te=perature coefficient of reactivity of approxi=ately

+ 3.3 x 10-5 alt / F.

While this value is not too reliable, it is completely consistent wf.th measured reflector coefficients for other pool reactors (1). These results would tend to indicate that the bulk te=perature coefficient of reactivity is not at all representative of the moderator coefficient and that the moderator coefficient is act-ually negative over the temperature range concerned with in these tests; 0

namely 65 to 130 F.

In order to verify that the coderator temperature coefficient is indeed more negative than the bulk coefficient, a third experiment was per-formed with the objective of obtaining a direct censure =ent of the cod-erator coefficient. This experimen; involved the installation of cover plates over the fuel boxes to prevent mixing of water inside the fuel boxes with pool water. In addition plugs were installed in all the open holes in the grid plate. Hot water was then forced into the plenum and the temperature chang,.nside a representative number of fuel boxes was measured with thermocouples located inside the fuel boxes. This experi-ment was carried out with an average temperature 'of approximately 70 F and indicated that the moderator temperature coefficient is approxi-mately -4.5 x 10-5 aK per degree F at a temperature of 70 F, thereby confirming the observation that-the positive sign o f the bulk coefficient 12-J

in this temperature range is due to the positive reflector coeffi-cient.

From the foregoing tests we conclude that tha coderator temperature coefficient is negative over the entire range, ambient to operating.

At nominal operating temperature the best estimate of moderator tempern-ture coefficient is -8.06 x 10-5 6gjoy, H.

FORCED CONVECTION POWER RUN Following the tests at essentially zero power, the reactor was run for a short period at power levels up to 300 Kw to calibrate the nuclear instruments by calormetric means and to validate the presence and sign of the Doppler coefficient. These runs were made with forced convection at a reduced flow rate. Subsequently short runs up to 1 MW were also made to check calibration. Instrumented test pins were installed in~

the reactor during these runs and fuel'aurface temperatures measured.

The.results at 1 MW with. ful1~ flow.. show that no difficulties should.

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be encountered at 2 MW from the standpoint of boiling. These tests.also demonstrated that the core' structural members are free from vibration under full flow conditions.

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I experiment, designed to separate the contributions to the reactivity

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defect at a steady state lower level,= was ' run._ At 300 Kw, a reactivity defect of 0.24% A K was measured. The contribution of moderator heat-

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temperature rise across a representative number of fuel assemblies and the moderator temperature coefficient of reactivity. A con-servative value of moderator temperature of reactivity was used to preclude overestimating the Doppler contribution to the reactivity defect. It was determined that the contribution of the moderator heating to the total reactivity defect amounted to less than 0.01% AK. Accordingly, the reactivity defect due to Doppler coef-ficient was determined' to be 0.23% AK at 200 Kw.

This confirmed the presence of a very large negative Doppler temperature coefficient of reactivity.

I NATURAL CONVECTION RUN The final steady state experimental run involved operating the reactor.._

at natural convection flow up to power levels of several hundred kilowatts. The purpose of this experiment was to determine a limit of reactivity addition, up to a max. mum of 0.5%A K, for the initial transient test. That is, before initiating pulses, it was considered desirable to demonstrate that the amount of reactivity to be added could be compensated -for by 'a relatively low steady state power level.- -

Accordingly, the reactor power level was raised in steps and reactivity defect as a function of power level was determined from the change in critical rod position. ~ These tests indicated that the power level required to compensate for n reactivity insertion of 0.5% A K is approximately 400 Kw with natural convection flow. From fuel surface temperature measurements made on the test pins during these runs the steady state capability of the core with natural convection cooling is conservatively estimated to be at least 750 Kw.

These tests also imply very strongly that the reactor will be very insensitive to reactivity _ changes when operated at full' power.

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.7 DISCUSSION In general, the steady state tests bear out the ability of the reactor to operate in the steady state mode at 2 MW.

Although some of the mea-sured values did not coincide with predictions, these variations are not significant and can be explained. Of particular impottance in this re-gard is the observed excess reactivity, void coefficient, and control rod worth.

It: is estimated that the actual control rod worths are approx-i imately 707. of the value which would be predicted by th'e calculational model (a direct comparison between measured and calculat'ed values is not possible because of the different control rod configurations used).

s While this will result in lower shutdown margins than anticipated, this i

reduction can be tolerated because of the extremely large ' shutdown mar-gin predicted. The over-estimate of control rod worth is attributed primarily to the fact that epithermal fissions in U-235 wery knowingly neglected in the calculational model used. The higher excess reactivity available in the core as compared to that calculated has, as a principal i

result, reduced core size. That is, fewer elements are requ' ired for a particular power' level and arrangment of experimental facilities than had been predicted. As a result, somewhat higher fluxes a$e available.

i The greater than anticipated excess reactivity is also attributed to the i

same omission in the calculational model, t

The measured average void coefficient of

.397. AK per percent void is smaller than the predicted value of -0.97. AK per percent void. [ Part of I

this discrepancy can be explained by the fact that, in the original, machine calculation, the reflector region was inadvertently voided.

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imply that the reflector coefficient is core nearly +5.5 x 10-5 goy and that the extrapolation of the reflector coefficient measurement for one core face is in error. This uncertainty was not resolved and, as a result, it was decided to use the upper =ost curve in Figure 2 in the experiment run at 300 Kw with forced convection to detemine the Doppler coefficient of reactivity in that this would result in a more conserva-tive, (ie. Iower estimate) of Doppler coefficient. In addition, the upper curve in Figure 2 is considered to represent the better esti= ate of moderator temperature coefficient of reactivity since the reactivity change produced by a change in moderator temperature is due primarily

- to density effects; therefore, the product of void coefficient and modulous of expansion for water should give a reasonable prediction of moderator temperature coefficient. The temperature coefficient measured for-the SPERT I oxide core (2) has also been plotted in Figure 2 for comparisen. As shown, the moderator temperature coefficient for PL1 STAR is larger than SPERT regardless of which estimate is used.

The steady state experiments run to determine power distribution were extremely important to the transient program in that these measurements fomed the Imsis for determining the maximum power = density in both the l

core and the test pins.. The detemination made by copper activation in-dicated that the ratio of the maximum to average poser density in core in the radial direction is'1.67. and that the hottest pin is located at -

~ the edge of the core. The maximum to average power density in the axial-direction determined in the same manner was 1.69, with the hottest spot l

occurring at the center of the effective core height and at the bottom o f 16 l

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are summarized in Figure 2 which is a plot of noderator temperature coefficient versus moderator temperature. The data on bulk coefficient is shown in Figure 2, fitted to a curve based on the void coefficient and the modulous of expansion for water, As shown, this bulk coefficient varies from a positive value of 2 x 10-5 aK/ F at 60 F to a negative 0

value of 5 x 10-5 at a temperature of 130.

The accond experiment involved the measurement of the reflector temper-ature coefficient and yielded a positive value of 3.3 x 10-5 AK/ F, not including the reflector above the core. This top reflector was not included as it responds to moderator temperature changes and, therefore, should be included along with the moderator effect. This positive reflector coefficient has been observed in other pool reactors (1) and represents a temperature rather than a density effect in that the reflector void coefficient of pool reactors is negative. The value of 3.3 x 10-5 for the reflector coefficient represents an extrapolation of the value measured for one core face, where the extrapolation was based on the relative reactivity worths of the side and bottom reflectors.

The measured bulk coefficient was corrected by the reflector coefficient

'and is also shown plotted in Figure 2.

In 'the third experiment, the temperature coefficient of.the moderator only was measured inside the core region at a temperature of about' 70.. The four such measurements mado are plotted in Figure 0

2 and are seen to coincide with a coefficient derived from the product of the void coefficient and the modulous of expansion -for water. Since this curve lies above the corrected bulk coefficient curve, this would.

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L A subsequent hand calculation yielded a value of

.567. UK per percent void which is in better agreement with the observed value. Since the void coefficient arises pri=arily from changes in absorpt>n in the resonance region, the omission of epither=al U-235 fissions in the calculational model undoubtedly contributes to this dif ference. Since little reliance was placed on the void coefficient in the safety analy-sis, this change in void coefficient does not have a significant effect on reactor safety. In addition, the very large Doppler coefficient observed provides a cuch core effective shutdown cechanism.

The power runs with both natural and forced convection indicate that the performance of the reactor in the steady state code is somewhat better than expected. As a result, it is esticated that the reactor can be operated at steady state power levels significantly greater than.2 MW with the present cooling system. This capability has not been fully analyzed since power level is presently limited to 2 MW by license.

The results of the experiments carried out to measure mcderator

~ temperature coefficient and power distribution had the greatest effect on the subsequent experiments run in conjunction with the transient program and, therefore, will be discussed in greater detail.

The measurement of moderator temperature coefficient involved three seperate experiments in which the bulk coefficient of the pool, the tamperature coefficient of the reflector and the temperature coeffient of the moderator, were measured. The results of these experiments 18

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i the core. That is, the power density at theca two points were essen-tially the same. Both of these determinations reflect the-large themal flux peak in the reflector.

Tue ratio of the maximum power density in the test pins to the maximum power density in the core was measured to be 1.86 by copper activation.

Some reservation existed with respect to using these results, since the copper activation techniques do not provide an accurate measure of total fissions, but only of thermal fissions. For this reason, power distribu-tion measurements using ft.el pin activation techni~ ues were also carrie_d q

out.

It was expected that these measureme.cs would indicate a somewhat higher power density in the center of the core as compared to the edge because of the harder spectrum at the center. This expectation was borne out by the fact that the ratio of the anximum to average power density in the axial direction was determined to be 1.92 by the pin activation techniques as compared to the value of 1.69 using the copper activation technique.

The ratio of se maximum to average power density in the overall core volume was determined to be 2.7 by the pin activation technique as som-pared to 2.82 by copper activation. This smaller ratio is due primarily to the smaller ratio of maximum to average power density in the radial direction of 1.41 determined by pin activation compared to the value of 1,67 determined by the copper techniques.

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This difference in distribution in the radial direction as determined by the two techniques was also in keeping with the estimate that the pin techniques would indicate a higher power density at the center of J

the core. In the case of the distribution in the radial direction,

'this increase in the power density in the center of the core had the e2fect of reducing the max to average ratio since the hottest pin is at the edge rather than the center of the core.

Measurements of the ratio of the maximum power density in the core yielded a value of 1.80 as compared to the value of 1.86 determined by the copper technique. This would indicate that the spectrum in the test pins and in the bottest pin in the core edge is essentially the same which is consistent with the large thermal flux peak in the re-flector. For' the sake of conservatisim, it was decided to use the more i

. pessimistic hot spot factors in analyzing the. trans ent program. Accord-ingly, a core hot spot factor of 2.82 was used representing the results from the copper data, even though the value of 2.7 obtained by pin activation was considered more realistic.

1 For estimating power density in the test pins, the ratio of 1.8 repre-senting.the maximum power density in the test pins to the maximum power density in the, core measured by pin activation was used. This is slightly lower than the value of 1.86 measured by copper activation and is, incidentally, considered more realistic.

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K CONCLUSIONS 1.

The steady state characteristics of the PULSTAR core were determined and show that the reactor is capable of steady operation up to power levels significantly in excess of 2 MW with the present cooling system.

2.

The excess reactivity of the 20 element core configuration, selected for detailed measurements and subsequent pulse tests, is 5.55%, and the total rod worth is in excess of 12% OK/K.

3.

The average void coefficient of reactivity for the 20 element core is -0.39% AK/% void and the moderator temperature coefficient varies

-5 from -4 x 10' OK/ F at 65 F to -10.5 x 10 6 / F at 130 F.

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these parameters are larger than those of the aluminum plate type core, formerly operated, and the SPERT I oxide core.

4.

The presence of a Doppler coeffient was detected and determined to be negative and at Icast as large as calculated.

l 5

The power distribution in the core is such that the ratio of the maximum to average power da-sity is 2.82 and the ratio of the maximum l

power density in the test pins to the maximen in the core is 1.8.

l l

6 The worth of a centrally located fuel element is approximately 2%

OK/K which is at least a factor of two lower than in the aluminum plate l

t>pa cute formerly operated.

l 21

~

7.

All of the data obtained from the steady state tests indicate that the PULSTAR core is safer than its predecessor.

8.

The steady state test results verified that the program of transient tests could consnente and be carried out in accordance with the Technical Specifications.

22

SECTION II PULSE ~ TESTS A INTRODUCTION This section describes the transient tests which were carried out during the period from August 4 through September 29, 1964. These tests comprised 14 ramp induced excursions with the reactor in a

. steady state mode and 100 step induced pulses with the reactor control system in the pulse mode. The description of these tests as set forth in the following paragraphs consists of a summary of chronological events, a presentation of the experimental data, an analysis of the experimental results and a summary of the conclusions drawn from the program to date.

l

-B CHRONOLOGICAL EVENTS 1 Ramp Tests On August 4,1964, following completion of the steady state tests, the first ramp excursion was induced with a shim rod i

with the reactor control system in a steady state mode. Number 5 shim rod was selected because, as shown in Figure 1, this rod is j

symmetrical with the pulse rod. The gears in rod drive #5 were changed to provide a rod drive. speed of 8" per minute as permitted i

by the Technical Specifications. The procedure followed during the ramp tests.was to insert the excursion rod (ie. #5) such that I

j the desired amount of reactivity to be inserted was held in this rod. With the other rods in their critical position and the t

reactor at a very low power, the excursion rod was withdrawn in i

f 4

in such a manner as to maintain the desired period. Ramp tests involving periods from 35.6 secs down to approximately 3.5 see were initiated in this manner. These tests comprise the first 8 ramps.

In the course of these tests it became apparent that the different-tal worth of rod #5 was inadequate to induce a ramp with a period of 2-1/2 sec (corresponding to the maximum reactivity limit permitted in the Technical Specifications.) Consequently, the gears in rod drive mechanism #3 were changed to provide a rod drive speed of 8" per minute. At this point, two rods, #3 and #5, had high speed gears installed. An attempt was next made to produce a 2.5 sec. period excursion operating rods #3 and #5 in gang. TI:ese unsuccessful attempts comprise ramps #9 and #10. This attempt failed because of inadequate differential worth of the rods. The necessary differ-ential rod worth was finally achieved by using rod #3 only and limiting its maximum withdrawn position to 63%. That is, excursions were initiated by pulling rod #3 a required amount to a final position of 63% withdrawn. The tests performed in this manner comprise ramp runs 11 through 14.

The final ramp run was carried out on August 8,1964.

2 Pulse Tests Following completion of the ramp tests and calibra-tion of the 12" pulse rod, the first of 100 step induced pulses

- was initiated on August 12, 1964. These so-called step induced pulses were initiated with a 12" pulse rod initially, which sub-sequently was replaced with a 24" pulse rod. The 100 pulses run were divided into six series. These series are summarized 2G

chronologically below.

(a) Series I This series consisted of pulses 1 through 38 and have in cccmon the fact that the data were reviewed by Ccemission repre-sentatives on August 29, 1964. The minimum period reached during this series was 13.7 msec. This series was also characterized by the use of a control curve calculated on the basis of a 20.6 msec prompt neutron lifetime.

(b) Series II Series II, comprising pulses 39 to 48, were carried out during the period August 28 to September 2,1964. This series was similar to Series I in that the same control curves were used.

Transients down to estimated period of 8.4 msec were run, although subsequent analysis indicated that the actual period was probably longer. These tests indicated that the observed performance of the reactor was very close to, or above the upper limits, of the control curves. Consequently, the third series of pulses described in the next paragraph was initiated.

(c) Series III This series consisted of repeat pulses 49 through

~

57 and was carried out during the period September 3 to September 10, 1964. These tests were distinguished from Series I and II in that an improved method for determining reactor period was used. In addition,- an attempt was made to measure input reactivity for each

_ pulse by carefully. balancing the shim rod bank, both before and af ter pulse rod insertion and prior to pulse rod ejection. Analysis of these results indi ated that the actual prompt neutron lifetime 2 5,.

i was 36 microsec rather than the 20.6 microsee value used in construct-ing the control curves. As a result, a request to revise the control l

1 curve was submitted to the Commission and subsequently approved.

(d) Series IV. Series IV tests comprised pulses 58 through 64 and 72 through 75, repeat pulses run with nine test pins installed in the core rather than six as in all other series. The objective of this series was to determine if the difference between calculated and observed promot neutron lifetime was due to acutrons born in the test pins. ~This series covered the periods from September 15 through September 17, 1964 and September 24 through September 25, 1964.

(e) Series V Series V,~ comprising pulses 66 through 71 and 76 through 82, were carried out during the. periods from September 22-othrough September-24,'1964, and September 25 through September 26, 1964-(repeat pulses.) This series'was characterized by the use of control curves based on a-pulse lifetime of 36 microsec. The series involved only, one step down in the reactor period to approximately,7 msec. As it 'became apparent-that the reactor I

performance was falling below, ie. in-a conservative direction, the lower limit of the control curves, a review of the earlier analysis, including the!. incorporation of data obtained' during this series, l-

lead to a revision ofi the measured.vslue of prompt neutron. lifetime.

Accordingly,'a second' set 'of ~ revised control curves was constructed i

' and used to monitor the series V1 pulses described below.

26' E

4

~.

(f) Series VI Series VI comprised pulses 84 through 100 and included 10 repeat pulses at the maximum output. This series, covering the period from September 26 through September 29, 1964, marked the attainment of a 5.16 msec pulse, approximating the maximum output permitted by the Technical Specifications.

C TEST RESULTS 1 Ramp Tests The minimum period achieved during tbc ramp tests was 2.5 see and corresponded to a reactivity insertxcn of slightly less than 0.5% AK.

The peak power achieved was 530 Kw.

A steady power of 370 Kw was achieved one minute after peak power, demonstrat-ing that the reactor can easily compensate for a 0.5% AK insertion at a modest steady state power level. The maximum test pin surface temperature observed during this run was 17;*F.

A copy of the Log-N chart trace taken during this run is shown in Figure s.

2 Pulse Tests This section will present the results of the pulse tests,'but will be limited to those pulses pertinent;to an analysis of the data. As such, the 8 pulses performed with nine test pins installed, 7 repetitive pulses at 5.2 msec,19 repetitive pulses at various outputs and 21' pulses, during which premature scramming, or instrument - failures occurred, are not included. The balance of 45 pulses are summarized in Table I, arranged in ascending order of reciprocal period.

t lh e

2 [!>,6eO

k.

w.

v.

s s

(a) Pulse Rod Calibration The 12" pulse rod, and the 24" pulse rod which replaced it, were calibrated aganist the shio rod bank and a single shim rod prior to their use in pulsing. The differential and integral calibration curves for the 24" long rod are shown in Figure 4.

As indicated in Figure 4, the total worth of the 24" pulse rod is 2.567. AK/K. In accordance with the Technical Specifi-cations, the worth of the pulse rod was both mechanically and electrically limited to 2.28%AK (

). $3.00). Also plotted on the integral curve in Figure 4 are some typical values of pulse rod worth obtained from the corresponding reactor period and obtained from the corresponding movement of the shim rods. The-a data are in reasonable agreement with the value obtained by integrating the diff rential worth curve; however, some error is to be expected as a result of the integration process and the fact that the pulse rod integral worth is expected-to vary, to a small degree, with shim rod position. This is illustrated in Figure 5 which is a plot of pulse rod ready. position versus the corresponding reciprocal period induced by repeated ejection of the pulse rod

'from this position. The points in Figure 5 taken from Series I and II show considerable scatter and reflect the fact that the shim rod bank was not perfectly' balanced with the pulse rod inserted.

On the other hand, the points in Figure 5 taken from Series III, V and _ VI show less scatter which is attributed to the fact that the shim rod bank was exactly balanced each time. The difference -in the straight lines fitted to these two sets of points is attributed to the effect of shim rod imbalance on pulse rod worth.

2g

(b) Power Output The maximum power output of 1150 MW was achieved on pulse #92. A typ cal plot of power versus time is shown in i

Figure 6 for pulse #100. As indicated in this figure, and discussed further on in the report, the pulse power is quite symmetrical. A plot of peak power versus reciprocs1 period is shown in Figure 7 for those pulses summarized in Table I.

For the most energetic pulse, data from both high level channels has been plotted. These data indicate a behavior consistant with the performance of the SPF.RT I oxide core. However, these data also indicate that the power level achieved in PULSTAR,.for a given period, is a factor of approximately 2.5 lower than the corresponding power achieved in SPERT. This difference can also be expressed in terms of period; namely, for a given power level, a period approximately half as long is required in PULSTAR as compared to SPERT.

2 Much of the scatter of the data in Figure 7 is attributed to errors in period measurements; although, as shown, some scatter is due to drif t in instrument calibration. This is evidenced by the disagree-ment in the readings of the two high level channels. Reproduci-I bility of pulses with respect to power Icvel was within about 10%

l as shown in Table II which summarizes the nine repetitive pulses l'

perforced at approximately 1000 MW.

Some of the variation in power l

from one pulse to the next is obviously due to -variation in period l

resulting from the fact that the pulse rod position varied slightly from one pulse to the next.. The analysis of the power data plotted in Figure 7 is presented in the subsequent paragraph entitled Analysis.

29

(c) Enerav Release A maximum energy to real'.t.cuer of 12.6 Mw-sec was achieved in pulse #92. The ratio of the total energy in the pulse to the energy at peak power was determined to be 2.1 based on a numerical integration of the pulse shape shown typically in Fi;ure 6.

This factor is identical to the value observed in the SPERT experiments and has been used throughout the analysis of PULSTAR data. Thus, the total energy release achieved amounted to 26.5 Mw-sec while the corre-sponding equivalent energy release achieved in the test pins was 47.7 Mu-sec. A plot of energy to peak power versus reciprocal period is shown in Figure 8 for those pulses sumarized in Table I.

These data are analyzed in a subsequent sub-section entitled Analysis. Figure 8 illustrates that the actual behavior of energy versus reciprocal period followed the anticipated behavior in the range of periods from i second to about 25 maec. That is, energy release reached a maxi-mum between a.5 and 1 second period, and then decreased to a minimum of 1 Mw-sec at a period of about 70 maec. The data plotted in Figure 8 are from the output of the power level integrator in the high level channel #1, which was recorded ch the Sanborn oscillograph.

The accuracy of the energy measurements was periodically checked with the pellet" capsule. The data from pulses run with the pellet capsule installed is given in Table II.

Because the thermocouple used to measure pellet temperature did not have a short enough time constant to enable measurements of pellet temperature at peak power to be made,

'it was only possible to compara measurements of total energy in the pulse and tail. As shown in Table II, reasonable agreement was achieved between the various methods.

30

(d) Test Pin Temperatures Two test pins, equipped with thernocouples for measuring surface temperature, were installed in the core through-out the pulsing tests. Measurements of test pin surface temperature 0

never exceeded 300 throughout all pulses and showed remarkably little variation with energy release. The maximum test pin temp'erature observed in each pulse is summarized in Table I.

These results are particularly encouraging in view of the fact that the test pins led the core hot sp;t by a factor of 1.8 and, therefore, experienced a maximum equiva-lent total energy release in the pulse of 47.7 Mu-sec.

(e) Pressure Two pressure transducers were installed adjacent to the core throughout all the pulse tests. One transducer was located approximately 10" above the active region of the hottest element. The second transdueer was located at the side of the core a half inch away from the fuel box. The maximum pressure indicated was 1.5 psi at the side of the core. The indicated pressure increased gradually with energy release as shown by the data summarized in Table 1.

The pressure transient at the side of the core resembled integrated energy, both in shape and time, giving rise to the suspicion that the pressure measure-ment is not a true indication and that the actual pressure is con-siderably lower. The pressure transient above the core had the shape of a ramp and did not reach a peak value until several seconds af ter peak power. Again, there is a suspicion that this information does not present a true picture of pressure or, at worst, that it represents the change in absolute pressure arising from thermal convection.

4 However, even if this pressure information is taken at face value, it can be concluded that the preatures encountered in pulses to date are quite small.

31

(f) Clad Strain Two test pins equipped with strain gauges were installed in the core during all pulses. The inforaation obtained from the strain gauges indicated little variation with temperature and energy. Further-more, the small variation observed showed no correlation with either of these variables. At Best, it can be concluded that the strain gauges indicated little or no change from the mildest pulse to the most energetic and the value of maintaining strain gauges in the core appears questionable.

(g) Test Pin Dimensions The dimensions of the static test pins $nstalled throughout the test were checked periodically in accordance with the conditions in the Technical Specifications. No change in test pin dimensions was observed within the accuracy of the measuring device which is estimated to be within i 1 mil.

D ANALYSIS OF DATA 1 Prompt Neutron Lifetire Following the first indications that the actual prompt neutron lifetime differed significantly from the calculated value of 20.6 microseconds, extrece care was exercised in obtaining critical rod positions, with the pulse rod fully withdrawn and with the pulse rod in its ready position, prior to each pulse. The data on rod positions and corresponding reactivity change obtained from rod calibration curves are sunnarized in Table III. These data consist of two sets of reactivity data; one obtained from shim rod bank posi-tions and one obtained from pulse rod positions. These two sets of data are shown plotted versus reciprocal period in Figures 9 and 10.

For the shim rod data shown in Figure 9 a least squares fit of the data 32

I.

results in a prompt lifetime of 29.1 microseconds 1 2.0 microseconds where the error repsesents the standard deviation frcm the mean. A correspond-ing fit of the pulse rod data shown in Figure 10 yicids a value for the prompt neutron lifetime of 30.5 microseconds 2 4.7 microseconds. The Y intercept of the best fit curve for the pulse rod data shown in

~

Figure 10 is attributed to the error in the pulse rod integral worth curve. This error was probably generated in the process of integrating the differential curve. Although the lifetime derived from shim rod data is c.casidered to be more reliable, the average value of 29.9 micro-4 seconds was used in predictions since both experimentally determined i:

values are in agreement within the indicated error.

1 To date, the ressor. for the discrepancy between the measured and calcu-lated values of prompt neutron lifetime has not been determined.

Analysis and tests have indicated, however, that the discrepancy is not due to neutrons born in either the test pins which are located I

approximately an inch and a half from the ccre, or the thermal column i

nosepiece which'is approximately 5" from the core. Furthermore, it is strongly_' suspected that the relatively long lifetime observed, as e

compared to the calculated value, arises from the delay in neutrons j

thermalized in the reflector. This effect was not considered in the i

L c alculations, based on experience with other water moderated, water

[

reflected cores. However. the.relatively'large reactivity worth of f-the reflector in the 20 element core tested, coupled with a relatively

(:

l-

- short. lifetie. of core neutrons,: could combine to result in a significant l-l'

. lengthening.of the effective prompt neutron lifetime.

33.

i l

In conclusion, the dif ferent lifetime observed is not considered to have any bearing on the safety of the pulse tests. While it might be argued that thelonger lifetime constitutes improved safety in that longer periods result for a given reactivity insertion, it is our opinion that the significant parameter is the shutdown coefficient which is independent of prompt neutron lifetime. That is, the only affect small lifetime variations have on core performance in the puls-ing mode is to vary the relationship between peak power and input reactivity. This is relatively unimportant compared to the relation-ship between energy release and input reactivity, uhich is invariant with prompt neutron lifetime.

2 Power Data A best fit of the peak power versus reciprocal period data in the region of alpha greater than 100 seconds-1, shown in Figure 7, indicates a slope of 2.41 10.44, the error representing the standard deviation from the mean. The corresponding value obtained from the SPERT data (2) was 2.35 (no error indicated.) The above error for the PULSTAR data can also be expressed as an error of i 16.5% in power.

Thus, the 95% confidence interval for these data is t 33%. Based on our experience, it is estimated that this error arises from significant errors in both period measurement and in power level measurement. Much of the error in power level measurement is attributed to drift in instrument calibration as evidenced by the disagreement between the two high level readings given in Table I.

While some improvement is expected in this respect, it is not considered feasible to estimate this improvement in absolute terms. Interestingly, the 95% confidence interval corresponds closely to the original error ectimate of i 35% based on calculational uncertainty.

34

A comparison of the least squares fit of the power data and the predicted performance based on the measured prcmpt neutron lifetime of 29.8 micro-seconds is shown in Figure 11.

This comparison shows that the best fit is within 20% of the predicted value. This is within the expected error of 30% in the predicted value, based on the 10% error in measured lifetime.

3 Energy A best fit of the energy data, shown in Figure 8, was determined by the least squares method for reciprocal periods greater than 100 seconds-1 The best fit yielded a value for the slope of energy to peak power versus reciprocal period of 1.26 1.32, the error representing the standard deviation. The corresponding value for the SPERT I oxide core was 1.36 (no error given) (2). The above error for PULSTAR also cau be expressed as an error of i 11.6% in energy. As in the case of the power data, this error arises from errors in both period and energy measurenent. The major source of error in the energy measurement is estimated to arise from the error in determining the time at which peak power occurs.

That is, energy to peak power is determined from a Sanborn Oscillograph trace with the time of peak power being determined from the corresponding power trace. The peak in the power trace is flat enough to induce some uncertainty in the determination of time at peak power.

I A calculation of energy release to peak power was made using the measured prompt neutron lifetime value of 29.8 microseconds. These data are also plotted in Figure 8 and are within 10% of the best fit which represents a considerable improvement over the error of i' 35% originally estimated in the calculations.

35

I The maximum specific energy release to peak power achieved in the core during the pulse tests was 258 W-sec/gm. This energy release corresponds to the maximum energy achieved in pulse #92 and is based on a maximum measured ratio of peak to average power in the core of 2.82.

The maximum specific energy achieved in the SPERT I oxide core for a 3.2 msec period pulse was estimated to be 634 W-sec/gm based on an effective co_e loading of 425 kg and a maximum to average power density of 2.5.

The specific energy release in SPERT I core, corresponding to a 5.2 maec pulse, has been estimated to be 328 W-sec/gm which is 27% higher than the value in PULSTAR for the same period. This difference is due primarily to the stronger Doppler coefficient in PULSTAR, and the difference in hot spot factors. That is, for the same hot spot factor, the difference in maximum specific energy would be 30% representing the difference in Doppler coefficient.

In summary, the data on energy release is consiscant with anticipated behavior and represents a performance more conservative than SPERT I.

t 4 Burst Shape The burst shape parameter, defined as the ratio of reciprocal period times energy (to peak power) to peak power, was determined from a best fit of a linear plot of peak power versus energy to peak power, by the least squares method. The value thus obtained was I

1.92 as compared to a value of 2.14 for the SPERT I core (2.) Thus, for PULSTAR the following is derived:

l 36

l En = 1.94 P,T where:

E

= energy to peak power in Mw-see o

P,

= peak power in W T

= period in seconds 5 Fuel Surface Temperature The maxicum test pin surface temperatures observed were considerably lower than predicted by transient temperature calculations performed during the design phase. However, these calcu-lations assumed a pessimistic value (ie. Iow values) for the contact resistance between clad and pellet. Therefore, the low ialue of fuel surf ace temperature observed in the test program is attributed primarily to a larger than assumed.value of contact resistance. Another indication that this explanation is correct can be derived from the plot of test pin surface temperature versus time for pulse #97 shown in Figure 12.

As shown in this figure, the time required for the surface temperature rise to decay to one half its maxiaum value is the order of 20 seconds, whereas a corresponding decay time the order of I second was predicted in the design calculations. This large difference could be explained by a large difference in the contact resistance between pellet and clad.

The observed behavior. in test pin surface temperature also implies much lower transient heat fluxes compared to calculated values and correspond-ing reductions in the value of cladd'.ng thermal stress. Therefore, it is concluded that the predictions regarding clad surface temperature and clad thermal stress were extremely conservative.

37

~ _..

l a

A i

6 Pressure A review of the pressure daca strorgly suggests that no i

measurable pressure changes were observed throughout the test program.

l This finding is based on the shape. and doration of the pressure i

transient, and the fact that no variations in the water level at the i

top of the pool were observed. With regard to the first of these reasons, the shape of the pressure transient, at the side of the core,

. resembles the integral of the power trace, both in shape and timing.

The maximum value of this trace corresponded to 1.5 psi and was reached several seconds after peak power. This shape strongly suggests

.that the observed signal is due to radiation effects in the pressure transducer.rather than a pressure signal. In addition, had such a pressure change occurred over such a long time interval, we believe that a corresponding variation in pool surface level would have followed.

Since no such variation occurred, it is concluded that this pressure J

signal is fallacious. The shape of the pressure transient from the pressure transducer located 10" above the core was a ramp which also l

reached a' maximum several seconds after peak power. The same arguments i

for discounting this as a pressure signs 1 apply, although the conditions suggest that the signal represents temperature effects, rather than i

radiation effects, in the pressare transducer.

J' These observations were discussed with 'Mr. Clyde Toole of the SPERT

' staff. He indicated that both radiation and temperature effects in

' pressure transducers have been observed at SPERT, and agreed that it is J-quite possible that the signale-observed could be explained on the above basic.

38.

While the evidence at hand does not constitute positive proof, we be% fs it strongly indicates that no measurable pressure variations occu.:. ed.

Proof of this hypothesis could probably be obtained by isolating the pressure transducer from the pool water in order to eliminate pressure as a variable. This particular experiment was not attempted because of restrictions ita the Technical Specifications.

7 UO Temperature The available data make it possible to estimate the 2

maximum UO temperature which occurred in the most energetic pulse in 2

which the total energy release in the pulse was 26.5 Mw-sec. Using the maximum measured hot spot factor of 2.82 and the UO heat capacity data 2

used in all predictions, the maximum UO temperature in core has been 2

~

0 ca1culated conservatively to be 1510 F.

The corresponding temperature in the test pins has been calculated to be 2500 F.

For comparison, the maximum UO temperature in the SFERT I core was calculated by us to be 1860 F for a 5.2 msec pulse and 3300 F for a 3.2 maec pulse. These data again illustrate the larger shutdown mechanism in PULSTAR as compared to the SPERT core.

8 Doppler Coefficient The Doppler coef ficient of teactivity in AK/ F was calculated using data from both the steady state forced convection power runa and the data from the pulse tests. The results of these calculations indicate.the Doppler coefficient of reactivity is within 107. of the calculated value based on a square root temperature dependence of the resonance integral, which was used in constructing the control curves. The uncertainty in the values derived from the measured data is due primarily to uncertainies in UO2 temperature 39

arising from uncertainties in UO thermal conductivity and pellet clad 2

contact resistance in the case of steady state data, and effective core heat capacity in the case of the pulse data.

A more reliable calculation of the Doppler coefficient can be made in terms of 4/Hw-sec. An analysis of all the pulses and an analysis of the most energetic pulse yield e Doppler coefficient of 64/Mw-sec. For coaparison, Spano (3) quotes a value of 2.2 to 2.9 4/Mw-see for the SPERT core over the range in period from 3.2 to 6.7 maec. These results are.in keeping with the relatively conservative behavior of PULSTAR with respect to the SPERT I performance.

E CONCLUSIONS 1 The ability of the PULSTAR core to operate safety in the pulse mode at power levels up to 1150 MW, total energy releases of up to 26.5 Mw-sce and a maximum period of 5.16 asec has been demonstrated.

2 Test pins installed in the reflector region operated at 1.8 times the maximum specific power in the core and experienced equivalent energy releases of up to 47.7 Mw-sec.

3 Test pin surface temperatures did not exceed 290*F and the pins did not experience any dimensional changes.

4 No significant pressure surges outside the core were observed s.

the test program.

40

=-_ _ _

._______.__________.___.._._...m i

+.

c j

i -.

l 5 The prompt lifetime of the PULSTAR core has been measured as 29.8 meec ccupared to a calculated value of 20.6 msec. This represents the only l

sf yificant difference between observed and predicted performance.

6 The basic behavior of the core was in accordance with predictions and closely paralleled the behavior of the SPERT I oxide core.

7 The specific performance of PULSTAR was more conservative, or safer, 1

than.the SPERT I oxide core with regard to Doppler coefficient, specific energy release, total energy release, and power output. A comparison l

of performance for. period of 5.2 maec is as follows:

SPERT I PULSTAR Doppler Coefficient

-2.2 to -2.9d/m-sec

-64/W-sec 328 W-sec/gm 258 W-sec/sm Max Specific Energy Release in Pulse

=

131 W-sec/sm 91 W-sec/gm Average Specific Energy Release in Pulse

=

=

55.7 W-sec.

26.5 W-sec

]-

Total Energy Release in Pulse Peak Power

=

2400 W 1150 W 8 The satisfactory performance of the test pins strongly implies that the ability of the core to operate at total energy releases up to 47.7 W-sec'has beenidemonstrated.

i-9 'All the experimental data obtained to date strongly suggest continuation

~

of'the pulse' test program in'accordance with the proposals contained in Section III of this report.

i 41

~,

_... ~.

_ =..

[

i I

SECTION III 1

PROPOSED TESTS A INTRODUCTION i

l This section describes the pulse tests proposed for the balance of the i

program together with recommended operating limits. The description of the proposed tests include estimated performance characteristics j

based on the performance observed to date. The proposed operating i.

limits reflect the experience gained to date and are designed to take advantage of the experimental data already in hand. Each of these topics are taken up in detail in subsequent sections and the section 1

is concluded with a summary.

i B PROPOSED TESTS In accordance with the request dated May 28, 1964, it is proposed to gradually raise the pulse output of the 20 element core to 10,000 mega-J watts in steps defined by the conditions given in the present Technical Specifications. This output would correspond to an energy release in the pulse of 72.5 megawatt-seconds and a reactor _ period of 1.9 milli-seconds. _ The present.ly installed pulse rod would be used and the re-activity input required would be equal to 2.37. AK/K. Based on the mart== observed hot spot factor of 2.82, the maximum UO2 temperature rise in the core which would be achieved in this pulse is estimated to be 3,640' F.

However, this maximum hot spot factor was measured at a shim rod bank position of approximately 50%,'i.e., with the pulse rod fully withdrawn and the reactor just critical. 'The position of the shim rod bank during this:maxPaum pulse is estimated to be 657, withdrawn.

42' I

~ ^

,.ar a

--rs,

,s y-.--

e

--+~4w.'

y-p

=

i 1

Therefore, a significant reduction in the axial hot spot factor of 1.92, measured with the shim rod bank at 507., is expected with the shim rod bank at 65'/..

Although the axial hot spot factor under the latter condition has not been measured, indication of its importance on U3 2

temperature can be illustrated. For example, if the axial hot s pt facter were to decrease from 1.92 to 1.7, the maximum UO2 temperature rise would decrease from 3,640 F to 3,260 F.

Consequently, we be-lieve that the estimated value of 3,6400 F is conservative and will actually be much less. However, we believe that the very low values of fuel surface temperature observed to date indicate that even this conservatively estimated UO temperature can be tolerated.

2 i

C PROPOSED L1HI?ATIONC Based on our experience, we believe that the present restrictions governing incremental increases in pulse output should be retained.

Addit,ionsilly, we propose that extrapolations beyond the following limits would not be made:

Maximum Power Output 10,000 Megawatts Maximum Energy Release in the Pulse 72.5 Hegawatt-seconds Minimum Period 1.9 Hilliseconds However, the absolute limits on energy release, power level, period, UO temperature and fuel pin surface temperature must be raised in 2

keeping with the proposed core output. Accordingly, the following limits, reflecting experimental error, are proposed:

'I 43'

~

t Power Level 12,000 Megawatts Energy Release 80 Ikgawatt-seconds Minimum Period 1.7 Milliseconds Haximum UO Temperature 2

in Core 3,700 F Haximum Test Pin Surface 0

Temperature 500 F With regard to monitoring core performance, we beleive that advantage should be taken of the experimental data obtained to date. Accordingly, control limits equal to a 95% confidence interval for energy to peak power and peak power as a function of reciprocal period are proposed.

That is, it is recommended that the determination of whether or not the next step in increasing output can be taken should be based on whether or not the deviation of energy and power output of the prior pulses is within the 95% confidence interval. If the deviations are within this limit, a new best fit curve would be determined and the next step taken. If the deviation in any given pulse exceeded the 95% confidence interval, the pulse would be repeated to determine if the deviation arose from a random variation. In the event that the average deviations of the original pulse and repetitive pulses were less than the 95% confidence interval, the next step would be taken

.after deriving a new best fit curve using all the data. If the ave-rage deviation exceeds the 95% confidence' interval, further steps would not be undertaken, and a report would be submitted to the Commission with reccennendations ~ for further action.

44

D SUlM\\RY It is proposed to Bradually raise the pulse output of the 20-element core to a maximum power level of 10,000 megawatts with a corresponding estimated period of 1.9 milliseconds and a total energy release of 72.5 megawatt-seconds. This recommendation reflects the confidence in core performance arising from the results of the test program to date. Further, it is proposed that the performance of the core during this escalation be monitored with control limits based on all experi-mental data available at the time. Ue believe that this escalation program is entirely within the capability of the PULSTAR core for safe operation and can be accomplished without jeopardizing the safety of either the general public or the operating staff.

45

_J

L_.i L..i 1_J l_i L.J

)

_, 4-i Poiro P ul t.e Pe t.i t i No. _

Serier.

(7.)

1 I

63.(

I 61.5 2

5 I

59.4 6

1 58.

8 I

57.o 9

I 55.4 10 1

54.0 17 I

54.1 18 I

53.5 28 I

53.@

24 I

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54 III St.F t

27 I

52.s 29 I

52.@

55 III 50.9

i 1

}.

i J L.;

L2 L_j LJ La LJ La L

La La L'

["

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TABLE I PULSE DATA tod Reciprocal Energy to Max. Test Max.

>n Period Period T:31 Peak Power P

Total Fnorcy Pin Temp.

Pressure mx hsec)

(Sec~l)

Ir. ut OK (MW)

(Mw-nec)

In Pulse (Mu-sec)

(OF)

(p n O Froa From LL llL#2 Core Core Equiv. in Pe r io :' Pulse Rod Test Pins 3

3 3450 0.29 0.452 0.51 0.315 0.5 1.05 1.9 217 0.25 6

2090 0.48 0.517 0.552 0.54 2.3 4.8 8.7 240 0.5 9

1350 0.74 0.566 0.566 0.82 3.1 6.5 11.7 235 0.5 1

870 1.15 0.619 0.655 1.08 2.57 5.4 9.7 230 0.3 4

520 1.92 0.663 0.663 1.35 2.74 5.8 10.4 230 0.5 9

260 3,83 0.712 0.760 2.3 2.0 4.2 7.6 249

)

209 4.79 0.726 0.795 3.0 1.56 3.3 5.9 252 0.5

)

139 7.2 0.743 0.820 3.7 1.27 2.7 4.9 253 0.6 7

71 14.1 0.780 0.825 6.56 1.11 2.3 4.1 263 0.6 57.9 17.3 0.792 0.846 7.05 1.17 2.5 4.5 262 0,3

)

37.6 26.6 0.819 0.855 13.9 1.34 2.8 5.0 271 0.33

)

37.1 27.0 0.0671* 0.14**

17.5 1.2 2.5 4.5 223 0.3 1

35.7 28.1 0.822 0.879 17.9 1.45 3.0 5.6 262 0.3 29.5 33.9 0.835 0.890 24.6 1.67 3.5 6.3 270 0.4 24.9 40.1 0.106*

0.169**

26.5 1.6 3.4 6.1 220 0.4

• Prompt SK from Shim Bank

• • Proept tJ from Pulse Rod Position w -.

u- -.-

L. -.

L.-

L. -

t_ s

,1 e-PLil $

~

Pulse POSI

_go.

Series

(.

82 V

43 7f V

4 48 II 47 80 V

43 69 V

44 i

76 V

44 j

77 V

47 84 VI 4tc 85 VI 41 86 VI 41 89 VI 42 90 VI 42 94 VI 41 91 VI 41.

92 VI 41; k

eine -

,g gbr

e r

1 T

Ld L._'s d

L4 k

a, LJ =

U L3 E.4

.i J la iJ iJ

-N

%m.-.

TABLE 1 - continued Rod ~

Reciprocal Energy to Max. Test M.i x,,

Lon 'g. g (Sec-1)

Input />K (1 04 )

(Mu-sec)

In Pulce ('tu-ccc)

__ ( F)

( r, a n-~

Period.

Period

- Penh Pouer P

Total Energy-Pin Tenn.

Pres:.w m.x From From IIL#1 Ilt92 Core-Core Equiv. in

~

Shin Pulse Rod Banh Tect Pins 167 8.55 117 0.345 0.418 291 240 5.64 11.8 21.2 253 1.0 25 8.5 117.5 0.322 0.385 286 390 5.6 11.8 21.2 289 1.0 49-8.4 119-0.321 295 5.9 12.4 22.3 262 0.6 ISO 8.02 125 0.366 0.424 354 5.8 12.2 22.0 266 1.0 91 7.2 139 0.390 0.456 369 360 6.1 12.8 23.0 270 0.9 75 7.2 139 0.391 0.465 395 430 6.4 13.4 24.1 268 00 7.2 139 0.391 0.45 379 450 6.7 14.1 25.4 264 52 6.89 145 0.407 0.478 423 330 6.4 13.4 24.1 264 1.0 82 6.23 160 0.445 0.514 533 440 7.2 15.1 27.2 245 1.2 01 5.9 170 0.492 0.560 669 120 7.4 15.5 27.9 244 1.0 24 5.78 173 0.511 0.574 745 262 1.2 94 5.4 185 0.54 0.609 806 269 1.3 D6 5.3

' 189 0.557 0.635 970 1010-10.96 23.0 41.4 269 1.5 76 5.2 192.5 0.575 0.645 1070 1030 11.86 24.9 44.8 262 1.5

'5 5.16 194 0.579 0.639 1101 1150 12.6 26.5 47.7 272 1.5 h,m. 9

i L.._,

Lu LJ L.-;

" ~

~

,, w Pulse Pulce Posit No.

Series Q

30 1

51.

31 1

51, 56 III 50, 34 I

50<

43 II 50<

32 I

50c 36 I

50.

35 I

50.

57 III 48, 37 I

50, 45 II 49.<

78 y

47.

52 III 47.

49 III 47, 81 V

46. <

Y iJ LJ.LU W

$~J ~

b bd L)

I, )

id b.'

wJ G ns TABLE I - continued Jtod Reciprocal Energy to Max. Test Max.

on Period Period Prompt Peak Power P

_ Total Energy Pin Temp.

Preccure

_nsec)_

(Sec~l)

Input SK OfU)

(

m.x

_Mw-sec)

In Pulse (Mu-sec)

(OF)

_ (pni)

(

From From LL llL#2 Core Core Equiv. In Shim Pulse Rod.

Test Pins Bank

1. 0 24.8 40.3 0.139~

32.

1.84 3.86 2

262 0.4

$0 20.7 48.3 0.160 38.1 2.0 4.2 7.6 271 0.5 32-19 52.6 0 145 0.206 62 2.8 5.9 10.6 232 0.4 79 18.4 54.4 0.176 47.5.

2.26 4.75 8.6 271 0.4 0

17.1:

58.5 0.208 86.6 2.4 5.0 9.0 1

16.5 60.6 0.191 64 2.5 5.25 9.4 275 0.3 0

15.3 65.4 0.203 86.1 2.76-5.8 10.4 271 0.5 0

14 71.5 0.208 89.3 2.8 5.9 10.6 280 0.5 9

13.8 72.5 0.216 0.256 150 3.75 7.88 14.2 244 0.5

.0.

13.7 73 0.208 105 3.18 6.68 12.0 252 0.4 11.7 85.5 0.257 151 3.8 8.0 14.4 267 0.4 HL#1 1

11.4 87.6 0.279 0.340 211 280 4.9 10.3 18.5 275 0.75

(

10.1 99 0.289 0.310 278 6.2 13.0 23.4 0.6 3

9.74 102.8 0.320 285 5.6 11.8 21.2 0.6 1

8.75 114 0.330 0.398 275 260 5.64 11.8 21.2 240 r-

TABLE II REPETATIVE PULSE Reciprocal Energy to Total Energy Pulse Rod Pos.

Period Period Penk Power Peak in Pulse No.

(%)

(msec.)

(sec-1)

(MW)

(MW-Sec.)

(MW-Sec.)

Digital High Level High Ievel Voltmeter Channel #2 Channel #1 91 41.66 5.2 192 1070 1030 11.86 24.9 92 41.75 5.16 194 1101 1150 12.6 26.5

. 93 41.99 5.2 192 976 980 11.52 24.2 94 41.96 5.3 189 970 1010 10.96 23.0 95 41.78 5.25 191 972 940 10.44 21.9 96 41.85 5.25 191 969 1010 11.24 23.6 97 41.95 973 98 42.07 1005 1010 11.6 24.4 99 42.10 5.3 189 992 980 11.96 25.1 100 42.26 5.3 189 969 975 11.6 24.4

J TABLE III MTA ON INPUT REACTIVITY SHIM ROD DATA PULSE ROD MTA Critica L Position (%)

A T (%)

AAP (%)

Ready AKT (7*)

AKp (%)

Pulse Pulse Pulse K

cr,1)

(Sec Rod In Rod Ouc Position (%)

49 102.8 47.51 1.08 00.32 52 99 57.93 52.46 1.049

.289 47.52 1.077 0.31 53 98 47.50 1.081 0.32 54 27 56.20 51.95

.8271

.0671 51.70

.90

.14 55 40.1 56.30 51.85

.866

.106 50.97

.929

.169 56 52.6 56.50 51.84

.905

.145 50.12

.966

.206 57 72.5 56.83 51.80

.976

.216 48.99 1.016

.256 66 135 57.00 51.14 1.14

.38 45.02 1.209

.449 67 139 57.12 51.21 1.149

.389 45.01 1.210

.45 68 139 56.91 51.03 1.147

.387 45.03 1.208

.448 69 139-56.95 51.05 1.15

.39 44.91 1.216

.456 71 139 56.94 51.04 1.15

.39 44.91 1.216

.456 76 139 56.88 50.98 1.151

.391 44.75 1.225

.465 77 139 56.84 50.94 1.151

.391 45.0 1.21

.45 78 87.6 56.29 50.99 1.039

.279 47.11 1.10

.34 -

l' TABLE III - continued SHIM ROD DATA PCI.SE ROD DATA Critical Position (%)

Pulse (I

Pulse Pulse AKT (%)

A p (%)

Ready AKT (1)

K A p (%)

K (Sec"I)'

Rod In Rod Out Position (7.)

79 117.5 56.52 50.99 1.082

.322 46.25 1.145

.385 80 725 J6.70 50.94 1.126

.366 45.50 1.184

.424 81 114 56.75 51.16 1.09

.33 46.01 1.158

.398 82 117 56.74 51.08 1.105

.345 45.67 1.178

.418 84 145 57.07 51.07 1.167

.407 44.52 1.238

.478 85 160 57.23 51.03 1.205

.445 43.82 1.274

.514 86-170 57.43 50.98 1.252

.492 43.01 1.32

.56 89 173 57.55 51.00 1.271

.511 42.84 1.334

.574 90 185 57.70 50.99 1.30

.54 42.24 1.369

.609 91 192.5 57.88 50.98 1.335

.575 41.66 1.405

.645 A

92 194 57.88 50.96 1.339

.579 41.75 1.399

.639 93 192 57.72 50.92 1.317

.557 41.99 1.386

.626 94 189 57.74 50.94 1.317

.557 41.96 1.395

.635 95 191 57.73 50.93 1.317

.557 41.78 1.399

.639 t

96 191 57.73 50.93 1.317

.557 41.85 1.394

.634,

4 TABLE III - continued SHIM ROD PATA PULSE ROD MTA Critical Position (%)

Pulse Pulse AKT(

0 P (%)

Ready AK II) 4P II)

Pulse E

cr,g)

T (See Rod In Rod Out Position (%)

99 189 58.19 51.40 1.308

.548 42.10 1.381

.621 100 189 58.20 51.40 1.310

.55 42.26 1.369

.609

TABLE IV

SUMMARY

OF PELLET CALIBRATION RUNS (Energy in Hw-sec)

Energy to Total Core Energy Pulne Enere,y to Peak Max. Pollet Temp.

Encrev from Pellet No.

H L#1

. Integral

• HL#1 Integral

• Digital HL#2 HL62 Voltneter 37 3.18 3.13 16.8 17.4 21.0 49 5.6 5.55 24.8 23.35 22.4 65 4.4 5.13 20.6 22.8 21.87 21.4 70 6.6 20.0 20.27 17.7 98 11.6 38.92 39.4 100 11.6 l'O.0 38.0 31.55 38.60 38.8

• Graphical integration',of log trace-considered to have largest error.

e k

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Available Excess: 5.587 AK/K Test Pins in Positions T A - Log-N Chamber B - Lin.-N Chamber P - Pulse Chambers S - Safety Chambers L

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o ~1 REFERENCES (1) The Effect of the Reflecter on the Temperature Coefficient of A Water Ccoled, Water Reflecting Reactor, by C. C. Geisler, et al, Paper 18-7, ANS Transactions, Volume I, No. 2, December 1958 (2) 'Self-Limiting Power Excursion Tests of a Water-Moderated Low-Enrichment UO Core in SPERTI,byA.H.Spano,etal,Ib016751, February 28, 1962 (3) Direct Measurement of the Dynamic Doppler Coefficient by Self-Limiting Power Excursion Tests, by A. H. Spano and W. K. Ergen, Paper 27-3, ANS Transactions, Volume 5, No. 1, June 1962 8 L ,}}