An experimental study of inherently safe characteristics of LMFBR systems is being conducted in EBR-II. Experiments have been conducted which show (1) capabilities of heat removal without the aid of active safety systems and (2) capabilities to mitigate a reduction in-core cooling flow or heat rejection to the balance of plant (BOP) without aid of active control or protection systems. Experiments involving loss of forced reactor cooling flow and loss of all heat rejection to the BOP are planned to be conducted. Purpose of this paper is to summarize the result of the test program by presenting representative test data

Betten, P. R.

Chang, L. K.

Feldman, E. E.

Mohr, D.

Planchon, H. P.

Singer, R. M.

UNT Digital Library

The EBR-II Inherent Shutdown and Heat RemovalTests - A Survey of Test ResultsbyH. P. Planchon, R. M. Singer, D. Mohr, E. E. Feldman,L. K. Chang and P. R. BettenArgonne National LaboratoryArgonne, Illinois 60439To be presented at theInternational Topical Meeting1 on Fast Reactor Safety% April 21-15, 1985irsc The submitted manuscript has been authoreda bv a contractor of the U. S. Governmentp under contract No. W-3M09-ENG-38.© Accordingly« the U. S. Government retains aO nonexclusive, royalty-free license to publishJJ or reproduce the published form of thisa contribution, or allow others to do so, foryj U. S. Government purposes.|5CONF-850410—6DE85 000094The EBR-II Inherent Shutdown and Heat RemovalTests - A Survey of Test ResultsbyH. P. Planchon, R. M. Singer, D. Mohr, E. E. Feldman,L. K. Chang and P. R. BettenArgonne National LaboratoryArgonne, Illinois 60439SummaryAn experimental study of inherently safe characteristics of LMFBR systemsis being conducted in the Experimental Breeder Reactor II (EBR-II). Experi-ments have been conducted which show 1) capabilities of heat removal withoutthe aid of active safety systems and 2) capabilities to mitigate a reductionin-core cooling flow or heat rejection to the balance of plant (BOP) withoutaid of active control or protection systems.Experiments involving loss of forced reactor cooling flow and loss of allheat rejection to the BOP are planned to be conducted. The purpose of thispaper is to summarize the result of the test program by presenting representa-tive test data.The test series, referred to as the ^Shutdown Heat Removal Jest (SHRT)Program, is being conducted on the Experimental Breeder Reactor II - a 62.5MWth sodium cooled breeder reactor integrated with plant heat transport steamgenerating and electrical generating system. A nominal 20 MW of electricpower is supplied for local and grid use.In addition to the direct demonstration of an LMFBR plant's "inherentlysafe" capabilities, a major objective of the SHRT program is to provide datafor computer code validation so that the results of the testing can be extrap-olated and applied to other LMFBR designs (as well as to other EBR-II operat-ing conditions). To support this objective an extensive data set is recordedfor each test by the plant digital computer based Data Acquisition System.Two special instrumented in-core subassemblies (INSATs) are being used to*Work supported by the U. S. Department of Energy under contractW-31-109-Eng-38.-2-measure in-core temperatures and flows. One INSAT is thermally and hydrauli-cally similar to a fuel driver assembly the other INSAT is similar to ablanket assembly. Data from plant instrumentation is recorded to characterizeall the major systems and components of the plant including the primary andsecondary sodium system and the steam generating system.The summary which follows gives a brief description of each type of testthat has been conducted to date along vith some representative data from eachtest type.Type A Tests - Protected Loss of Flow from PowerThe purpose of the Type A tests was to demonstrate and provide measure-ments of the transition to natural circulation cooling following a sudden lossof primary forced circulation (loss of electrical power to the main coolantpumps and auxiliary pump) while at power. Nine tests of this type were con-ducted. The parameters that were varied from test to test were the initialpower and initial flow conditions, and the secondary loop pump mode of opera-tion. The most severe test (SHRT 17) involved a simultaneous loss of electri-cal power to the main coolant pumps, the auxiliary pump, the secondary looppump and a reactor scram from 100% power and flow. Figure 1 shows tempera-tures (predicted and measured) indicative of the response of EBR-II to theSHRT 17 loss of forced circulation. Both the pretest predictions and themeasured temperatures correspond to coolant temperature at the top of theactive core. The predictions were made with the NATDEMO and HOTCHAN codesprior to conducting the test for which the maximum, nominal and minimum curvesdenote the HOTCHAN predictions for XX09 with positive uncertainties, withoutuncertainties and with negative uncertainties, which are labeled "MAX", "NOM",and "MIN", respectively, in the figure. The temperature measurements shown inFig. 1 are raw data from TTC 31 (the hottest thermocouple) and were takenduring testing for monitoring purposes. It should be noted that TTC 31 is oneof 13 "top-of-core" (coolant) thermocouples which are all located 0.83 in.below the actual top of fuel. However, because of other compensating effects(i.e., offset of ^ +8°F and center of bundle location), the TTC 31 readingscan be directly compared with the true top-of-core predictions without intro-ducing significant error.The test was initiated by a reactor scram and pump trips at time zero.As shown in the figure the temperature initially decreases rapidly in response-3-to the reactor scram and rapid power decrease. As the pumps coast down to astop (at about 50 sec), forced flow decreases end the decay power heats up thesodium in the channel. The temperature rise causes an increase of the thermalhead which induces sufficient natural circulation flow to terminate the risein temperature (at about 70 sec). At times beyond the temperature peak, aquasi-equilibrium is approached in which the gradually decreasing rate of heatgeneration in the core is almost matched by a gradually decreasing rate ofheat removal by the coolant. This is indicated by the decreasing flow andtemperature in the XX09 subassembly.As shown in the figure there is excellent agreement between the predictedand measured data. Post-test analysis is necessary to quantify effects ofactual test conditions (such as power history) and actual instrumentation off-sets on the results. However, subject to further confirmation in the post-test analysis, the data indicates 1) that natural circulation is a safe andreliable method of decay heat removal and 2) that the phenomena governing thetransition to natural circulation have been identified and are adequatelyunderstood and modeled.Type B Tests - Delayed Loss of Flow from Shutdown; Isothermal ConditionsThe goal of the Type B tests was to provide measurements of the develop-ment of thermal heads and natural circulation flows sufficient to cool thecore even from an isothermal initial condition. Nine Type B tests were con-ducted from shutdown conditions from which the decay heat load ranged fromvery small (SHRT 1 followed a lengthy shutdown) to about 1.6% power for SHRT2. All of the tests were performed from near isothermal (100% primary flow)conditions, except SHRT 2 which was initiated by tripping the auxiliary pumpwhile the primary pumps were off. The representative data for SHRT 2, whichis included in Fig. 2, is consistent with earlier EBR-II test results andindicates 1) adequate natural circulation cooling following a loss of forcedcirculation during shutdown, and 2) good agreement between measured andpretest predictions.Type C Tests - Reactivity Feedback TestsThe objective of the Type C tests was to demonstrate and measure the EBR-II reactor power and core temperature response to 1) perturbations in primarycoolant flow at a constant inlet temperature and 2) perturbations in primarycoolant temperature at the reactor inlet for a constant flow condition. These-4-tests are stepping-stones to the unprotected loss-of-flow and loss-of-heatsink tests to be run in future parts of the SHRT program because they willvalidate the NATDEMO reactivity feedback modeling.Eight Type C tests were completed. The first four tests were initiatedby perturbing primary flow or inlet temperature from an initial 25% power andflow steady state condition. Two subsequent tests were initiated from 50%power and flow and the final two tests were initiated from 70% power and flow.The primary flow perturbations were initiated by manually ramping primarypump speed while being careful to keep the flow balanced between the pumps.The sodium temperature at the reactor inlet was maintained as constant as pos-sible by controlling heat rejection via the IHX to the secondary loop withvarying secondary loop flow.Representative raw data for SHRT 25 are shown in Figs. 3 through 6, inwhich primary flow was perturbed from an initial 70% flow condition. Figure 3shows the measured core flow through the instrumented subassembly XX09.Figure 4 shows a representative in-core temperature response as measured onTTC 31. The excess reactivity, calculated online by the EBR-II plant DAS isshown in Fig. 5, and the measured power response is shown in Fig. 6. The dataindicate that the inherent response of the plant tends to keep power and flowmatched, thus keeping core temperatures within the operating range. Detailedanalysis of this data will be conducted to quantify reactivity feedback co-efficients and confirm the dynamic modeling in the NATDEMO model.Representative data for SHRT 26 &re shown in Figs. 7 through 10. The re-actor inlet temperature was perturbed by manually changing the secondary loopflow as shown in Fig. 7 while primary flow was held fixed. The perturbationsin secondary loop flow altered the IHX heat rejection resulting in a change inIHX primary outlet temperature and consequent change in reactor inlet tempera-ture. The inlet temperature response is shown in Fig. 8. Reactor inlet tem-perature variation causes the reactor power to change (Fig. 9) due to reac-tivity feedback, and consequently affects temperature in the reactor. Figure10 gives the measured XX09 temperature reading from TTC 31. Note that theincrease in inlet temperature causes a significant decrease in power and coreoutlet temperature. The Type C tests indicate a ^ 2% power reduction occursfor each 1°F increase in reactor inlet temperature. This negative temperaturecoefficient is larger than has been typically taken credit for in conservativeEBR-II pretest predictions.Fig. 1PREDICTED AND MEASURED TEMPERATURES OF SHRT 17LUI3OOI200 [-I I 00 I-I000gj 900800 I-700 I-6000 XX09 MEASUREMENTNOM XX09 COOLANTMAX XX09 COOLANTMIN XX09 COOLANTMAX HOT DRIVER CLAD0 40 80 I20TIME INTO TRANSIENT, sI60Fig. 2PREDICTED AND MEASURED TEMPERATURES OF SHRT 21000-0 XX09 Measurement1 Horn XX09 Coolant2 Max XX09 Coolant3 Min XX09 Coolant4 Max Hot Driver Clad. . .750--100 100 200TIME, s300 400/ •• , '• Fig. 3MEASURED FLOW IN INSTRUMENTED SUBASSEMBLY XX09 OF SHRT 255550£ 45C93 403530 ,01:—1• —'<-_I I I I I I I I I I20 40 60 80 100 120TIME, minCORE TEMPERATURE, °Fo•nOOamm•vm>3Dmm>(0mDXXo(0o•nCOenFig. 5ON-LINE CALCULATED EXCESS REACTIVITY2.0COOOQ:COCOLJoXUJ-2.0-3 .0 rTIME, min656055a:LJ504540Fig. 6MEASURED POWER OF SHRT 25i i i i i i0 20 40 60 80 100 120TIME, minFig. 7MEASURED SECONDARY FLOW OF SHRT 26250200|5OIOO50 ifILj 1 1— MQBmi i n i i i0 20 40 60 80 100 120TIME, minFig. 8MEASURED REACTOR INLET TEMPERATURE OF SHRT 26680670oI 660I -a.u 650Ixi640630 VTi;\\\120 40 60 80 100 120TIME, minFig. 9MEASURED REACTOR POWER OF SHRT 260 20 40 60 80 100 120TIME, minTEMPERATURE, °F03Olo(0oo(0OloooooOlooomCDoooro:iii/(s ,•'•rtm>cmoOOO>m•om59mo

EBR-II inherent shutdown and heat removal tests: a survey of test results

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EBR-II inherent shutdown and heat removal tests: a survey of test results

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