ML19340A515
| ML19340A515 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 06/23/1961 |
| From: | Collett W, Owen E GENERAL ELECTRIC CO. |
| To: | |
| References | |
| 61-858, NUDOCS 8008070701 | |
| Download: ML19340A515 (26) | |
Text
'
r*~ f -l t
~
Conference Paper $~/oPl$b C,.
wa ffz as.s. A s f L A of j
9..u cis-y Jh4 i % y A M ial7150 ca W
.'s 5
~U CONTROL AND TRANSIENT PERFORMANCE OF THE DRESDEN Q
NUCLEAR POWER STATION
.s W. I. Con ett E. R. Owen K
Non= ember AIEE Associate Member AIEE General Electric Co.
General Electric Co.
A Conference Paper prepared for presentation at the AIEE Su=er General Meeting, Ithaca, New York; June 18-23, 1961. Paper also submitted for transactions approval and subsequent presentation at Pacific General l
Meeting, Salt Lake City, Utah; August 23-25, 1961
~~il!9N T0 "ETJLATO.RY CENTRAL FILE
.300 016 ACui.AiORY 30:.'(Ei ri ac.
Paper No.CP 61-838 8008070 7pp g
CONTROL AND 'IRANSIENT PERFORMANCE OF THE DRESDEN NUCLEAR POWER STATION W. I. Collett E. R. Owen Nonmember AIEE Associate Member AIEE General Electric Co.
General Electric Co.
Synopsis In putting the first large cccmercie.1 boiling vater reactor utility plant into service a thorough series of control and transient performance tests were conducted.
Conventional electrical plant sensors with appropriate conversion equipment greatly facilitated obtaining and recordinr; these data. Correlation of plant transient perfor-
=ance with prior analog computer analyses was good. Test data is provided for dual cycle control characteristics, load changes, turbine trip off, and other unusual transients.
The maneuverability and operability of this type of nuclear power plant is established as being very satisfactory for general utility applications. Stability characteristics of this large boiling water reactor are excellent.
INTRODUCTION The Dresden Nuclear Power Staticn when it vent into service in early 1960 was the first large commercial nuclear power plant, though other boiling vater reactors have been operated over the past decade. All of these have been of limited capacity, and of an experimental nature. The Dresden Nuclear Power Plant is the first large boiling water reactor and has a net electrical plant output capacity of 185 megawatts.
In addition, it also involves a unique concept known as the dual cycle control system. This concept makes use of changes in reactor sub-cooling or ecolant inlet temperature to make power changes and, by incorporation of a unique control system, allows load demands on the reactor system to be shared by both primary and seccndary steam systems without control rod motion. During the design stages of this reactc,r, extensive analog computer studies and analytical calculations were cade of plant transient performance and control. Detailed studies of plant transient performance were carried out to help size plant equipment, to anticipate actual capabilities, and to provide extensive information relative to evaluation of plant safety. As a result of this design analysis 'vork, considerable interest existed in obtaining confirmation of this work by means of actual plant transient tests. A series of rather extensive plant transient tests, with scme of a rather severe nature, vere performed on this plant. This paper reports the manner in which these tests were made, as well as the test results and interpretations of these test results.
In order to facilitate the obtaining cf test data, use was made, insofar as poseible of existing plant instrumentation. Electrical signals were picked off plant sensors where signal response was still satisfactory (e.g. before going to pneumatic conversion). These test signals were then fed to electrical conversien equipment and onto a multi-channel recorder. This transient recording equipment performed very well and proved to be an extremely valuable tool during this test program. The test data obtained did in fact compare very well to program performance and contributed to further insight into the behavior and understanding of a boiling water reactor system. These data in addition to confirming the Dresden design and pointing te some specific systen improvements vill provide reactor designers with the ability to explcit further boiling vater reactor designs for future utility applications.
DESCRIPTION OF PLANT A simplified diagram of the Dresden Nuclear Power Statica is shown in Figure 1.
This figure indicates the reactor pressure vessel and the primary circulation loop wherein a flow of about 7,000 lbs/second of water circulates through the risers, drum, devncomers, pumps, steam generators, and reactor vessel. Four parallel primary recirculation pumps
-2 and loops are used. Primary steam is teken off the steam dru e.t 1,000 psi, and fed through a pressure regulating ve.1ve to the high pressure inlet of the turbine. Primary pressure at the steam drum is cicsely regulated. Secondary steam from the four steam generators is passed through e secondary admissiccs valve to the ninth stage or secondary inlet of the turbine. The prinary system del' vers stee.m at 1,000 psi vhile the secondary steam generator pressure re.nges frem 1,000 psi st no lead devn to 500 psi at full load, in a predictable manner. A bypase relief valve is shown which can provide a primary steam dump to the cendenser when for cne reascn or another primary stee.m is not able to be taken directly to the turbine. A centinuous flew of 125% cf rated pri=ary steam flev may be passed to the condenser. Ccndensate is pu= ped frem the condenser throush a full flow demineralizer and feedvater heaters, after vnich it is delivered to the primary steam drum.
In addition, condensate is alsc rumped thrcugn secondary feedvater heaters to the secondary steam generaters. Rated prinary atte.m flov is 333 lbs/seccnd and rated secondary steam flow is 330 lbs/second. The reacter cere is rated at 626 MW(t). Plant efficiencies at full load have been measured at 23.L% net efficiency. When secondary steam flov is increased, the primary coclant te=perature reaching the reacter cere vill decrease or vice versa. This effect is the basic phenomenen utilized in dual cycle contrcl cf a Boiling Water Reactor (BWR).
5"?.A7Y STATE CCN'RC1 AFD 1U', CYC'E TERFORMANCZ As is well kncvn, cne vay cf achieving reactor contrcl is through the manipulation of neutron absorbing centrol reds within the reactor core. In boiling water reactors the presence of the strong void feedtack causes a tignt coupling or feedback between fuel rod heat flux and control rod pcsition. This is cest illuctrated oy the feedback loop in block diagram for= thet is stevn in Tigure F.
The veid rea:tivity in steady state just compensates the centrcl red ex:.ess reactiv y,1 d t':cs reactor pcver level vill vary with control rod position. The entire lo:p respctds very rapidly, so that if a control rod position adjustment takes ple:e, the initiel pcsitive excess reac-ivity forces the peutron flux. The fuel red heat flux transf erred :a the veter ce.uses an ncrease in boiling and void formation. When tne added vcid r eactivity again equels tne added contrcl red reacti-vity, the excess reactivity rearte= :ere er.d Ole pc-er level cf the reactor steadies at that point.
The entire respe S e to a 2nf -_ step in err:rci red reattivity settles cut rapidly within 1 to 2 seccnds. Evidence cf C" e i s 4e rn later in cennection with the stability discussion.
In additicn to this ze:hed Of red ::n:rci vhich 13 e'a acteristic of the means used to change power level en Erevien v ngle cycle peretien, the plan is also designed to perform on dual cycle con:rol. 7 is d' a1 cycit centrol concept invcives the phenomenon of sub-cooling :cntrol cf a 3RR pc.e level.
Thi s ; articular pher.cnenon is clearlf illustrated in Figure 3 Ea-a +"-a= sequential tire vien.
f the sane boiling varer reacter are shown.
The first represents a steady state BWR cperation with typical exial ;over distribution, n, and void volume, Rg.
The 2e:ctd de;ic;9 the efft:' cn the :cre void dis *ribution at the first instant as a step change cf eccler vtter (ar.:reased inlet su'c-ccoling) sweeps through the core. As soon as the rea::cr vcid is recu /d, the increased or positive reactivity causet reacter pcVer to increase. Wi * "' " a =acend cr so the reactor power and void volume establish a new steady state pcwer level as aheen in the third drawing, having the same void reactivity as the first drawing. As a 1s ter of note, a similar saquence of events vill also cecur with a ficy in:rease inctead cf the sub-cocling increase that was described above.
The Dresden Nuclear Fever Station utilizas the a~acve su:-cooling power control principal to provide a centrcl range over whicn autcrocic icad fcileving takes place. When an increase of load da-and is sensed by the turbine speed governor, this, ty a suitable linkage, causes the secondary cr turbina ednis= ices valve to cpen to meet the demand.
In so doing, however, the seccnde.ry steam generaterr vill, with their increased lead, cause the primary ccclant temperature :: drop. This decres.se in ecclant tenperature, when it is felt by the reactor ccre, vill increase reacter pcVer level to ecm;ensate fcr this demand.
. When operating at large primary steam flows the primary steam flow will also increase and some of this load demand will also be carried by the primary steam flow, as the primary admissions valve opens to control system pressure.
To show the direct effect of secondary steam flow on sub-cooling, the dual cycle control characteristics as calculated and measured for the Dresden Nuclear Power Station are shown in Figure 4.
These characteristics show essentially the stee,dy state dependance of reactor power and primary steam flow on seconde.ry steam flow, for fixed control rod scttings. Each solid, nearly horizontal line shows the control characteristic for one control rod pattern. The diagonal solid lines are lines of constant reactor power level.
The dashed, nearly horizontal lines are calculated control characteristics and compare favorably with actual measured characteriatics. Near rated power the automatic control range is seen to cover from about 55% to 100% of rated load.
In reality, the operation of dual cycle control as described above where void reactivity was assumed to remain constant is not quite so simple. Actually, as core power level increases, the average Uranium Dicxide fuel temperature increases and this introduces another reactivity change (the Doppler effect) thattends to reduce reactivity. This is a stabilizing effect and has important safety effects. It is a phencmenon associated with the U-238 in the lov enriched fuel. Thus (although following an inlet coolant temperature increase, the reactor power vill always increase to the point where void reactivity has brought reactor excess reactivity tack to zero, as described above) the change in core power vill cause an increase in fuel average temperature which in turn contributes a negative reactivity component. This effect of a fuel temperature increase is virtually the same as that if control rods were very slightly inserted subsequent to the sub-cooling change. The control characteristic linee have a lesser slope than they would otherwise have without this Doppler effect. In the Dresden Reactor the void reactivity at rated is greater than the change in Dcppler reactivity with change in core power occurring when secondary steam flow is varied through its full range and as a result, the reduction in the dual cycle control range due to the Doppler effect is only slight. At much lower primary steam flows this Doppler effect becomes more important (much smaller void reactivity) and the control characteristics slope scmewhat negative as seen in Figure 4.
A void reactivity of 0 5%dK/K and a Doppler coefficient of.9 x 10-54K/K/*F vere used in these calculations, althcugh the actual values are believed to be slightly larger, nov.
Another reactor characteristic that effects the control characteristic curves on the dual c cle control map is g which has a high neutron absorption cross sectice
/
reactor Xenon transient behavior. One cf the fiosion products of U.235 is Xenon or capture probability. When power level is changed in a reactor, tne Xenon concentration and distri-bution changes somewhat with time. This change is verv slow; over hours, rather than 1
seconds. Thus, it is not important in its influence on faat transients. However, where power level changes are made very slowly, the Xenon changes in the reactor vill have some effect.
If, for example, the seccndary steam vere reduced very sicvly, over an hour or so, from rated down to zero secondary fiev to the turbine, the Xenon would build up (due to the lower power not burning it out as fast as before) and would act as if control rods were inserted slightly during the slow secondary steam reduction. The net effect is to make the appe. rent control characteristic curve steeper (slightly higher positive slope) than the curve through rated pcver level nov is.
' Itis Xenon effect is actually fairly insignificant and is made even more se by the unique control scheme used in the Dresden Dual Cycle Turbine Bypass Control, This valve linkage and control system which incor-porates the turbine valving and the Dual Cycle Ccctrol concept into a working system control in shown in block diagram fcrm in Figure 5 As a turbine load demand is made through the speed governor, referring to Figure 5, the only valve that is moved im=ediately is the secondary centrol valve. The bypass valve is normaDy closed and the primary admissions valve is under contrcl of the pressure
. regulator, even though the pressure regulator and speed governor links are tied together at the primary admissions valve. When load is picked up on the secondary valve, steam flow at the secondary steam generators is increased and as explained above this causes an increase in primary flow which is controlled by the pressure regulator. This regulator causes the primary ad=1ssions valve to open slightly in order +o hold primary reactor pressure essentially constant. In so doing the linkage connection closes the secondary valve, slightly. The increased demand is initially carried by the secondary flow, but is shortly, partially transferred to the primary and valve linkage control system. Should a trip out signal occur, the speed governor vill close down its linkage. This will close both primary and secon.iary valves and force the pressure regulator to open the bypass valve, as shown in Figure 5 In this way, the dual cycle turbine contrcl system can meet both normal and emergency control or load change de= ands.
The turbine control scheme also provides another unusual feature which greatly minimizes the effects of control rod position changes, Xenon poison transients and recir-culation flow changes on turbine output. With the main load limit fixed, a change in control rod position vill reduce primary steaming rate, cause the pressure regulator to close down slightly on primary flow, and by means of the connecting linkage lift the secondary control valve. The additional secondary steam flow and resultant reduction in core inlet temperature causes reactor power level to increase by a sufficient amount to essentially follow a constant power line on the dual cycle control map of Figure 4.
- Thus, gradual disturbances of primary core power will be compensated for to a considerable extent by means of the controller adjustmente of secondary steem flow and the resulting sub-cooling effect on reactor power.
A final note in connection with the steady state control of this reactor relates to the operation of the control reds and the detemiration and adjustment of core power distributions. Each of the 80 control rods is indicated with respect to position, and can be individually, arsi only individually, manually positioned. The decision as to what rods to move, and how far to move them, is based in part on the observation of in-core distri-bution. Because of outer core instruments inability to "see" very far into a core, and because. of the existance of a multiplicity of critical volumes in a large core, the outer core flux instrumentation does not provide a good indication of in-core power distribution.
This is characteristic of any large core regardless of type of reactor system application -
BWR, PWR, etc. For this reason, this instmmentation need is supplemented by in-core instrumentation. The Dresden Nuclear Reactor has 64 in-ccre ion chambers, actually located internal to the core during reactor operation. These in-core chambers are monitored and displayed for use of the reactor operator. By this means, the reactor operator can easily trim control rod positions from time to time, during power operation, to maintain a desired flatness of power distribution. 'Ibere is economic incentive for so doing. Power distri-bution changes or shifts during dual cycle power level variations have been shown to be fairly small.
TEST INS'IRUNENTATION To obtain the detailed transient test data derived from the Dresden Nuclear Power Station, it was necessary to design and install certain test instrumentation that could be used to record the various pertinent parameters. Because of the nature of the nuclear power plant, all the desired plant transducer signals are electrical, being either AC or DC. The response of these transducers was found to be satisfactory for the test transients expected. Thus, the nomal plant transducers were utilized exclusively for the transient signals. All flow, pressure, and level instrumentation of test interest utillzed moveable core transfor=er type transducers. Pressure and flux sensors had very rapid recponse.
The special equipment required was reduced to that which would be necessary to convert the AC signals to DC and perfom certain functions to =ake the signal recordable on an eight-channel recorder. The AC to DC conversion equip =ent consisted of a chopper demodulator which was very suitable for the low level AC signals coming from the moveable core transformer of the transducer.
The basic unit of the amplifying equipment was the
5 operational amplifier. The operational amplifier provided:
(1) high imput impedance to prevent loading of the p'lant transducers, (2) flexibility in obtaining the desired gain to make the signal recordable, and (3) a means of performing square root, su= ming, cveraging, etc. of several signals, when such was required.
The signals which were made available for recording during these transient tests were the following; Primary Drum Pressure Outer Core Neutron Flux Reactor Vessel Pressure In-Core Neutron Flux Secondary Steam Generater Pressure Primary Admissions Valve Position Primary Steam Flow Secondary Admissions Valve Position Secondary Steam Flow Turbine Bypass Valve Position Primary Drum Level Pressure Drop Across the Secondary Primary Feed Flow Steam Generators The neutron flux signals were recorded either separately as a single outer core instrument signal or as an average of all six outer core signals. For resolution purposes, the in-core signals were recorded individually from one or more instruments. Where core spatial phenomena were of interest, several individual in-core chamber signals were simultaneously recorded. There are no direct flow measuring orifices in the primary loop for recirculation flov measuring purposes. For transient signal purposes, the pressure drop across the secondary steam generator pri=ary tube side was recorded as a measure of recirculation flow variation. The pressure drops across these four steam generators were averaged and recorded as a single pressure drop. This signal was used only as a transient flow indicator rather than as an accurate flow measurement. These sensors exhibited a 1 eps resonance (Fig.13)
Figure 6 is a typical circuit for one signal, namely the pri=ary steam drum pressure.
The signal transducer converts pressure to an AC signal. This signal is converted in the chopper demodulator to DC and amplified by the operational amplifier. The amplifier output is then recorded.
Other circuits are similar but more complex where it is necessary to take the square root of a pressure drop signal used to measure steam flow. Also, there vare several steam lines, each with a steam flow sensor which required that these signals be summed to obtain the total flow. Further complication is experienced in correcting the four secondary steam flow signals with seccndary steam pressure because of the variation of the secondary pressure with load. The test equipment proved to be very flexible and useful in provicing the necessary functions for recording. In addition to the function of obtaining recordings of plant transient behavior, the continuous recording of these plant parameters was found to be extremely helpful during unusual plant operating conditions, especially because of the feature of simultaneous portrayal of several system parameters.
'IRANSIENT TEST RESULTS The transient behavior of the Dresden Nuclear Pover system was explored carefully, with rather small system or reactor disturbances at first. As an example, pressure regulator sat point changes were =ade beginning with very small changes at low power levels and working up to step changes of some 10 to 20 psi at substantial power levels.
Similarly, other tests such as load step demands, pump trip offs, etc. vere perfomed at low power 1svels.
Thece vare repeated at successively higher power levels as each test indicated that satisfactory response could be anticipated from subsequent tests. The ab 1 ty of the pressure regulator to control reactor system pressure was examined closely over the entire rrgulating pressure range of from 150 psi to more than 1,0c0 psi.
The Dresden turbine control system utilized a Position Restored Pressure Regulator having about 20 psi regulation. This regulator, designed by M. A. Eggenberger of General I:lectric Company, Large Steam Turbine Department,was able to control pressure to within 1 psi deviation at rated load over long periods of time, and even over mild transients.
. If.s peri'omance was outstanding. Its gain corresponded to that for good system stability based on analog computer studies. Two other regulators were also employed and were of the zero pressure error or integrating (reset) type. These vere used as backup regulators and possessed a lower gain with a resuitant slight increase in hunting, but nevertheless are suitable for this service.
To make pressure system transient tests it is necessary to:
(a) open or close a low capacity steam dump line (a bypass valve' or, (b) to raise or lower the system set 3.oint rapidly. Actually, both methods were performed at times. In some cases, even at substantial power levels, pressure regulator set point adjustments were made by manually, rapidly moving Radiation levels due to some N-]he regulator handvheels at the front end of the turbine.
in the primary steam from the reactor vere lov enough to permit easy access to the front standard for such tests. Neutron irradiation of the oxygen in the water of the reactor produces some N-16 and a little of this 7 second half-life isotope is carried J
over with the steam.
I The balance of transient tests were all made from the control room by means of remote control and instrumentation. Pressure regulator set point adjustments from the control roca can only be made at a rate of 1 psi /second by design, but at the front standard a regulator handvheel can be moved fast enough to change the set point at about 20 times that rated'.
The following paragraphs describe the more significant system transient tests and are discussed and interpreted to give a good insight and understanding into EWR system behavior.
The general question of BWR stability received a lot of attention during the early design phase of Dresden and several significant testa vere performed to establish the relative degree of its stability. The nature of these tests and the conclusions drawn from them are discussed in the final section.
Dual Cycle Transiett Performance A number of small load rejections and load pickups were made on the Dresden system to investigste overall system stability with respect to such disturbances. These were accom-plished by means of remote manual adjustments of the speed governor. A typical load dis-turbance transient is shcun in F1 ure 7, wherein 25 MW electrical was r idly dropped from 5
the system over a few seconds. The effect of this load rejectica on pruary drum pressure, secondary steam generator pressure, and reacter Gux or power is indicated in this figure.
The transient data shown in Figure 7 in solid lines is a reproduction of the actual trace obtataed on the 8 channel recorder. In addition to the data shown in the reproduction,.
the variation of in-core flux, and primary steam flow vas also recorded during this and other transients. It is apparent then, following this fairly rapid reduction in cecondary steam flow, that the secondary steam generator pressure rises en approximately a time constant of about 23 seconds. As socn as the change in temperature of the primary water leaving the steam generator reaches the reactor core, a gradual reduction in core power It is evident frem the recording of Figure 7, that there is about an 8 second occurs.
delay batveen the occurence of the change in pressure in the secondary steam generator and in its corresponding effect on reactor core pcVer. This time interval corresponds to a transport delay time plus the plenum mixing time constant effect of the lower plenum of the reactor vessel. A similar smooth transient occurs on load pick up.
A more significant system transient under dual cycle operation is the secondary steam flow trip off transient recorded in Figure 8.
In this particlar transient recording the reactor was initially operating at 127 MWe, when, at a certain time, the secondary steam flow to the turbine was rapidly cut off by closure of the secondary stop valves. After about two minutes, when the system had thoroughly settled, the stop valves were reset and the operator began increasing secondary steam ficv to the turbine. Again, as in Figure 7, following the rapid cut off in secondary steam flow, the secondary steam generator pressure L
7 rises smoothly on roughly a 23 second time constant. After the transport time the reactor flux rapidly decreases in a nearly exponential fashion to a steady lover value.
When recovery of the secondary steam flow was begun reactor pcver increased gradually as seen in Figure 8 following the secondary steam and power de=and.
During this entire transient, reactor pressure, controlled by the position restored pressure regulator, did not depart by more than 2 psi of its initial setting. This transient is typical of the very stable load following and transient performance obtained under dual cycle operating conditions. Plant operation was smooth, responsive, and predictable even under these large and rapid load change conditions.
Turbine Trip Off Transient One of the most severe, unusual operating conditions to be met by any power plant is l
that of turbine or generator trip off where the electrical load or turbine load is com-pletely separated from the steam producing device. On the Dresden system, both generator trip off and turbine trip off tests were performed. Generator trip tests were performed at half load, and turbine trip tests at 150 MWe. In Figure 9 a recording is given which indicates the primary drum pressure behavior, secondary steam generator behavior, primary steam flow variatione and outer core flux transient changes during a turbine trip off test near full load. In the turbine trip test the operator closed the turbine stop valves and the primary and secondary admission valves by initiating the master trip. The closing of these pri=ary valves opens the bypass valves through the nor=al control linkage discussed earlier. During the period of time that the primary admission valve is closing and the bypass valve is beginning to open, the reactor pressure rises very rapidly by about 20 psi. Almost immediately, within 1 second, this pressure turns around and is controlled by the bypass valve. The effect of this can be seen in the primary steam flow.
This is a recording of the total primary steam flow going either to the turbine or the bypass valve. The steam flev is seen to be initially cut to a relatively low value for about a half second snd then rapidly recovers to more than rated, in order to bring down reactor system pressure. Peak steam flow is limited by the bypass valve to 125% of rated.
It is interesting to compare the transient recording with analog computer studies performed with a lumped thermodynamic system, 160 MWe. The analog pressure rise is not as rapid as the recorded pressure rise during the transient. The reason for the rapid initial pressure rise in the actual primary stee.m system is believed to be the limited steam volume availability during the first instant folleving isolation. This is explained by the fact that when the primary admission valve begins to close, and before the bypass valves are opened, the steam volume that is immedictely available to absorb this pressure rise consists only of that volume of steam in the steam lines and in the pri=ary steam drum. The rest of the steam volume in the system, namely that in the risers and in the reactor vessel, cannot be immediately used to absorb the presture rise until the pressure disturbance is transmitted through the steam-vater mixture of the risers to the reactor vessel. In the lumped system used for sc=e analog studies, it is assu=ed that all steam volume and saturated steam-vater volu=e is lu= ped together and is instantly availab'.e to absorb this pressure rise. Hence, the lover calculated initial rate of rise of pressure.
The primary steam flov is transferred to the bypass flow faster in the analog case because the actual bypass valve cpening rate at the time the test was cade was not as fast 1
as desired. The system transient behavior was not greatly affected by this change, when theopeningratewasreadjustedfromktothedesired1-1/2 seconds. The lower recording in Figure 9 is a recording of the reacter pcver as cbtained from the average of the six outer core ion chamber s1 nals. The effect of the rapid transfer of steam to the bypass 5
system appears on the flux recording as a short duration flux pulse or " spike".
The flux transient obtained in an analog computer simulation of this transient did not indicate such a rapid rise and fall as was recorded in the actue.1 system. It is apparent from other recorded data that the greater rate of rise and fall of flux in the actual turbine trip off case is due to the effect of the drum pressure rise on the core recirculation flow. A
4
. corresponding recording of recirculation flow shows an initial " kick" or short period increase of recirculation flow for the first half second or so when the pressure rises rapidly in the steam drum. The pressure exerts a force on the downcomer water area which is a solid water leg directly connected to the core inlet. The force is sufficient to accelerate this flow system so that core flow vill increase until the corresponding pressure wave can also travel devn the riser, two phase mixture system and reach the other end of the solid water leg in the core. This latter is a relatively slow path, taking an estimated 0 5 second for transmission down the risers. After this time delay, the recirculation flov has fallen back to its initial value and slightly undershoots. The short time increase in recirculation flow helps to increase core power more rapidly than the pressure rise itself would described earlier in connection with effects of sub-cooling and flow changes on core power.
Flow measurements in Figure 7 vere made by means of the pressure drop across the secondary steam generators. The short duration flux rise as indicated here en turbine trip off has a negligible effect en fuel temperature. In general, the system behavior under a trubine trip off operation settles fast, and follows rather closely to analog 4
predictions. The transient effect of recirculation flow was not anticipated in initial analog studies. It is an interesting but relatively insignificant effect.
Si=ultaneous Reactor Isolaticn and 3 cram In the event that the turbine oil supply is lost, turbine stop and bypass valves would close and completely isolate the reactor from the main heat sink or condenser. If this oil supply is lost the reactor is autcmatically scrammed so that reactor heat generation is reduced to a icv value within a few seconds. Once a scram (rapid reactor shutdown) signal occure, control reds are run in very rapidly and neutrcn flux levels are reduced in a fraction of a second to a very icv value. However, the stored or sensible heat of the uranium cxide fuel la significant and continues to deliver a heat flux to coolant water. This heat flux decays on a time constant of the order of 10 seconds. A parameter of particular interest in cennecticn with this reactor isolation and scram transient was the peak pressure rise that vould occur. If the pressure rise is greater than 50 psi, the emergency condenser vill be brought on and, if progressively higher than that, safety valves would be lifted at a ; <asure rise greater than 175 psi.
The pressure rises folleving this simultaneous isolation and scram occurs because of the stored heat in the fuel and the isolation of the system from any direct heat sink. Because secondary steam generators are utilized, the coolant reaching the core hat about 51.h BTU /lb sub-ccoling at rated load. The steam generntoro continue to act as a heat sink as their operating pressure is raised frem rated operating conditions of 500 psi to a no lead pressure of nearly 1,COC psi following cut off of secondary steam.
Of course, the capacity of this heat sink reduces as pressure apprcached 1,000 psi. The transient pressure peak is reached within the first few seconds after isolation. The effect of the steam generators heat absorption canes into play t. yond that time, while their heat absorbing capacity exceeds the decay heat cf the reactor.
The transient recording of a full load reactor isolaticn and scram test conducted at about 190 MWe is sLevn in Figure 10.
The solid lines incicate the actual system transient performance. Again, the dashed lines indicate analog computer runs that vere performed prior to plant operation. It is apparent from the drun pressure recording that the total pressure rise during this transient is only 23 psi. This pressure rise occurs within about 3 seconds. The corresponding primary valve position eignal is shown in the lower trace, and indicates closure of the pri=ary valves vit'11n 1 second. Bypass valvee did not lift during this transient because their oil nupply had been cut off by the same function that caused isolation and scram. The effect cf the scram rod motion on reactor neutron flux is seen in the third recording of the trace. The control rod scram motion causes a rapid flux reduction within 0 5 seccnds. The reactor flux falls to essentially zero in less than 1 second ef time folleving the scram signal.
The reason for the relatively lcw pressure rise under this full load reactor operation test is because of the large amount of sub-cooling in the dual cycle system at rated load. As soon as the heat flux drops fras rated level to below 58 5% of rated there is no longer enough heat to brin 6 the sub-cceled water completely up to saturation.
Thus, no more steam is formed and there is no further pressure rise in this system. This is illustrated in Figure 11 where the nearly exponential decay of heat flux following an assumed instantaneous cut off of neutron flux is plotted. This corresponds to an 8 second time constant which fits the actual more complicated heat flux decay expression most closely over the first several seconds. In a very simple manner, using this exponential heat flux decay and considering the heat required for sub-cooling, one can determine the number of equivalent full pcVer seconds of steam flov available during this heat flux decay period. It is seen from the figure that the decay heat going to steam formation (shaded area) is equal to 1.8 full steam flow seconds equivalent.
(Boiling ceases when heat flux falls to point A) Since under isolation conditions beginning at rated conditions, one full power second of steam flev yields a pressure rise of 18 psi, it is possible to approximate the pressure rise that vill occur in the actual system on an isolation and scram. The estimated pressure rise by this means is conservative due to neglect of liquid saturation enthalpy change with pressure, and neglect of steam condensatica on metal and liquid surfaces.
It is noted that the analog computer run indicated a significantly higher pressure rise than that measured. However, a study plotted in Figure 10 of the analog computer transient of outer core flux shows the reason for this.
In this particular analog transient for conservative purposes, it was assumed that no reactor scram occurred following isolation until the neutron flux rose to its scram level. The neutron flux rise, which is caused by the rise in pressure, the pressure effect on void volume and sub-cooling, reaches the scram level of about 125%. At that time s: ram motion rapidly reduces the neutron flux to zero. Under this condition the reactor was actually producing pcver for about 2 seconds after the isolation condition existed. On the above basis of the approximate equivalent full steam flov seconds analysis, this should acccunt for roughly another 36 psi rise over that for the conditions where a: ram occurs simultaneously with isolation. Thus, the analog run and the measured transients appear to be quite compatible, censidering the difference in flux scram conditions.
The comparison of initial pressure transients over the first few seconds of the turbine trip test and of the full load reactor isolation test are rather interesting. This comparison is made in Figure 12.
In the case of the turbine trip, steam is transferred rapidly to the bypass system, and thus, after the first rapid initial pressure rise, the pressure recovers quickly under control cf the bypass valve as described earlier. Under the full load reacter isolatica test the pressure rise, as shown in Figure 12, is LLsost exactly the same as en the turbine trip test, for the first second or so.
In both transient cases, during the initial period, the primary steem flev has been cut off (ccmpletely, in the case of reactor isolation, and mementarily in the case of turbine trip before the bypass opens). The risers continue to deliver steam flew to the drum (at least until the effects of reactor scram are felt in the case of the isolation and scram test). Thus, the initial rise in drc= pressure for these two transiente vculd be expected to be very close, as indeed they are seen to be in Figure 12, for the first 1-1/2 second. For the turbine trip transient, the drum pressure quickly falls fc11cving this pressure initial rise and subsequent cpening of the bypass valve. In the case of reactor isolation and scram test, the drum pressure continues to rise at a diminishing rate after which it gradually drops because of the heat absorption of the seccadary steam generators as previously discussed.
The vessel pressure variatica during the first few seconds of this reactor isolation test is also shown in this figure. It indicates that nc pressure effect is felt in the vessel until about a half second after the pressure rise begins in the steam drum. The same vessel pressure del'.y effect was characteristic of the turbine trip test also. This time delay is apparently due to the transport time of the pressure wave through the riser steam-water mixture frcs the drum to the reactor core. The vessel pressure rise
i
-10 is not'nearly as steep as the drum pressure rise because by the time the pressure rise begins in the vessel, the entire steam and saturated water-steam mixture volumes are able to accen:modate the loss of energy sink and minimized the pressure rise. Thus, the initial reactor vessel pressure rate of rise is approximately the same as the dashed line shown in Figure 12, which was calculated for the lumped thermodynamic system under 100% power mismatch conditions. The higher rate of initial pressure rise in the drum is apparently due to the lower amount of steam volume (drum and steam lines) initially available to absorb this power mismatch en either temporary (turbine trip) or complete 2
isolation. The initial pressure effect on reactivity is determined by the pressure changes at the reactor, and this is seen to correspond to that of the lumped thermodynamic system except for a very short (1/2 second) time delay.
IVO-PUMP TRIP OFF 2RANSIENT In the Dresden system there are actually four recirculation pumps and loops. Normally, these loops are separated into two paire insofar as their electrical supply is concerned.
It is not highly unlikely that electrical supply might be los t suddenly to two of the recirculation pumps. For this reason transient performance under the abrupt loss of power to two pumps was investigated. Because of the high drum arrangement in the Dresden primary system there is a large natural circulation (density) driving head so that even if all four pumps are lost a substantial recirculation flov (of the order of 50% of rated) vill still occur. In this particular transient, shown in Figure 13, the plant was operating at 130 Mwe under steady state dual cycle conditions when two pumps were suddenly tripped off from their supply. The pumps were allowed to vindmill and no valves were adjusted on these two loops. The flow decay transient for the huge primary recirculation lecp is quite fast with the flow decay having roughly a time constant of about 17 seccads.
The total rotational stored energy in the pumps is significant but only 1/3 as great as the fluid inertial stored energy. The lower trace in Figure 13 is a recording of pressure drop across the steam generators. To approximate the flow changes, the square root of the fractional change shown in this figure must be taken.
Following the rapid flow decay, the ree.ctor flux or nuclear power level falls almost as rapidly as the flow decay. Scme undershcot in the fl.ux response appears as expected.
The gradual rise in flux folleving the initial settling of this undershoot is due to the effect of sub-cooling. Since under dual cycle cperation if reactor power and steaming rate initially drops, the control system described earlier causes the pressure regulator to close up on the pri=ary valve and at the same time lift the secondary valve to attempt to hold the load.
As the secondary valve opens and the steam flow is increased from the steam generators, the sub-cocling increase causes reactor power to increase to continue to meet load demand. The recording in Figure 13 aise shcws the decrease in primary steam flow as the pressure regulator closes up to hcid pressure under the decrease in core power immediately following the drop in flev. The stability of the drum pressure is apparent.
The maximum deviation during the transient is 3 pai from the initial pressure. There is a swell or surge in drum level of about 6 inches immediately after this tve-pump trip off.
REACTCR AND REACTCR SYSTEM STABILITY Early in the boiling water reac&r development, the performance of the Borax and Spert reactors produced considerable conc ern in their evidence of an unstable or periodic power oscillation behavior. The possibility cf such instability was visualized as stemming from the inter-relationships of reac. tor pressure, steam voids, void reactivity, nuclear power and heat flux as depicted in Figure 2.
Extensive acalog computer studies of tbese inter-relations were performed during Dresden design stages. L as found that for toe Dresden system, an unstable situatien could not exist even for large variations from design parameters. The reactor system appeared to be rather insensitive to void reactivity values,
4 with the fuel time constant, core hydraulics and operating pressure selected for the design. In addition, careful design attentien was directed towards insuring core 7
operation removed from any inherent hydraulic instabilities such as ficv pattern transition thresholds, major flow redistributions with slight power changes, etc.
During initial Dresden operation, tests were conducted in several areas of concern to ascertain the existence of stable performance. Tests were performed on the riser flow to investigate flow sharing between risers and flow stability in the risers.
Strain gauges gnd vibration pickups mounted on the risers confirmed the conclusions of some N-lb steam activity tests that flow through the riser system was extremely smooth and stable. Flux signals measured by both in-core and outer core ion chambers showed very low noise content, with about 1% RMS at rated power, and no pre-dominant frequency content. System pressure stability also confirmed the stable reactor behavior.
The Dresden reactor is one of the largest reactors in operation with respect to its nuclear characteristics. Thus, the test program was also concerned with determining the degree of reactor spatial stability, and the controllability of reactor power distributions.
Evidence obtained from examination of the many in-core instruments showed a high degree of spatial stability of power distributions. One interesting test performed by Mr. E. S.
Beckjord involved the oscillation of an off-center control rod, and the simultaneous recording of in-core ion cha=ber signals at several core locations. A typical piece of data from this test is shown in Figure 14 for a rod oscillation period of 15 seconds.
Of particular interest is the manner in which the flux responds to the control rod changes.
The response of the chamber nearest the control rod tip follows rod position (reactivity) closely, with no overshoot. This illustrates the tight void reactivity feedback loop behavior of Figure 2.
Also of interest is the fact that from this and other recordings it is apparent that the entire reactor responda in time phase spatially, with only attentuation frnm the point of disturbance to other core sensor points. No out of phase spatial behavior of the core was evident. At all times the reactor core behaved as a unit and in a very stable fashion. The one " notch" motion of Figure 14 is about 9 inches.
In order to investigate possible proximity to a stability conducted where primary steam flow was increased from 1.4 x 10+ghreshold, a test was lbo/hrtoabout2.x 10+
lbs/hr. This was achieved by dumping steam to the main condenser, in addition to supplying rated steam to the turbine. Pri=ary system pressure control and general system behavior appeared quite stable, even at this very high steam flow conditionf indicating no close proximity to any unstable mode. Pcver level and pcver distribution of the reactor core were able to be easily controlled by the reactor operator at all times, through individual control rod adjustments. Under dual cycle load control, reactor power distributions re=ain fairly constant. The stable behavior of the Dresden reactor and reactor control system provides for easy control of the plant. There was no evidence of spatial Xenon oscillations of an azimuthal variety. Some Xenen power distribution changes have occurred in an axial mode following a large dual cycle power change, but such are long tera (hours), and require little if any rod changes. The presence of steam void reactivity seems to provide a strong stabilizing effect on large core Xenon oscillation tendencies.
CONCLUSIONS A very large number of transient and stability tests have been conducted on the Dresden Nuclear Power Station. The test data discussed and portrayed here represents a small part of the total data obtained. The cencept of using existing plant primary censors in conjunction with spe cial conversion and recording equipment proved to be a very simple and effective method of providing the desired transient test data.
t
=
, The test results closely confirm the analog system predictions of the reactor system performance. The analog computer studies utilized a one node reactor model and a three node thermodynamic model to represent the reactor vessel, risers, and steam drum.
The ability of the dual cycle characteristic of the Dresden Station to be easily controlled and maneuvered to meet typical utility power plant load demands was confirmed by this extensive testing program. A fully automatic load control over the range of 55%
to 100% of rated power is obtainable without moving control rods. Load pickup rates in excess of 7% per minute were demonstrated during the tests. Plant design vill permit pickup rates considerably in excess of this on the dual cycle range. Load rejection capabilities were also demonstrated.
The tests performed established the extremely stable performance of the Dresden reactor and primary flow system, even at much higher than rated primary steam flows and core void conditions. In fact, the results of these tests clearly establish the smooth, stable, predictable and responsive behavior of the Dresden Boiling Water Reactor Plant, and its feasibility for utility power applications.
The assistance and cooperation of the many General Electric Company personnel and of the Commonwealth Edison Company personnel, who made this thorough test program possible, deserve considerable recognition.
a l 5
EMERGENCY RELIEF VALVE PRESSURE BY-8 ASS REGULATING RELIEF VALVE VALVE PRIMARY STEAM DRUM g
TUR8INE jM j i TURBINE
[
GOVERNOR CONDENSER
/%><-
n l'
=
1 I
\\
/
SECONDARY REACTOR 1 r
)
SECONDARY LEVEL f
CONTROL VALVE Jk
=
FEED WATER HEATERS PRIMARY LEVEL DEMIN.
CONTROL VALVE l
1 l
Fig. 1.
Dresder. dual cycle system diagram
I L
l AKR AKex REACTOR Ah A0 TO
- v FUEL THERM 0 DYNAMICS KINETICS m
MODEL AKv 4
1 1
VOID ay V010 REACTIVITY ACCUMULATION FROM FACTOR THERMODYNAMIC MODEL PRESSURE i
RECIRCULATION FLOW i
Fig. 2.
Boiling water reactor block i
diagram
cm m
C M
o o
cp N
- o g
o a
o o.
0 o
0*,o*
o I
cp 8:
,s o
fm o
o*
- o.o o
6 o,
Fig. 3 Errect or subcooling or flow increase in a boiling water reactor.
Upper dravings show axial distribution of steam voids, Rg, and power, n, corresponding to lower drawings sequential physical behavior of the reactor
i 4
i i
EXPERIMENTAL LOAD SWINGS
--- COFFI CODE CALCULATIONS l.6 10 0 %
RATED 1.4 t
"* * ~
1O MlN
==""" *,
1.2
^
E 80 %
e
\\
o 1.0 3
o I
g g
=-
~
a8 a
o h
> 0.6
\\60%
e 1
E O.4 \\
xx
~s 40%
\\
0.2
\\
20 %
l O O O2 0.4 0.6 0.8 1.0 1.2 SECONDARY STEAM FLOW (106 lb/HR)
Fig. 4.
Dual cycle operating charac-teristics of the Dresden plant.
m
_m.
J l
BACKUP PRESSURE REGUL ATOR PRIMARY PRESSURE REGR ATM PRESSUR LOAD SET WITH E MERGE NCY TREP ADJUST ABL E PRIM ARY STOP SPEED RELAY GOVERNOR e SPEED
& LOAO LlutT O
[
SPRIN G BYPASS SECONDARY S CONDAR p
RELAY 4,OAD LIMIT MTE LIMIT I
PRIM AR Y SECOND AR Y ADM'SSION ggy pagg IIE#"
VM.VE ADJOSTOR PRESSURE DF"dE S
'l TUR8iNE l
SECONDARY BYPASS
~ g pppigggy CONTROL VALVE STFAM VALVE DRIVES EMERGENCY ORIVES TRIP A
JI_ PRIMARY i
e SECONDARY
+ WATER STOP STEAM D NSER ya p K P. TURBINE STAG S EA N
A
' I-SECONDARY C
965 PSAA 1p PRIM A R Y SYPASS VALVES EXMAUST NECK STEAM II stop SPRAY TO VALVE TURStNE i
INLET 475-925 PSIA SECONDARY Fig. 5 Turbine control system block STEAM diagram.
2
__...__m_
.__.m f
i i
1 -
PRESSURE SENSOR i
ZERO R5 POTENTIOMETER CHOPPER "h
g j flLTER
]
{
{
{
I
\\
R2 R3 R4 EXCITATION Rt "M
"N.
^N.
RECORDER l
l
[
d i
1
- Ci
- C2 OPERAT ONAL
, g y, g
i CORE COUPLIN G AMPLIFIER TRANSFORMER TRANSFORM E R j
=
I I
I I
I I
I I
I l
- - PL ANT SENSOR -
A- - - - - DEMODULATOR --- - --
A--SCALER -
A-RECORDER O
Fig. 6.
Sensor signal conversion schematic diagram 1
i
DRUM PRESSURE
~
2 PSI "1IIllIliIllil1IlIlIlI SECONDARY PRESSURE 1000 PSI
' NA50 [RUN
~~
400 PSI IIIIIIIIIIIIII OUTER CORE FLUX 8 0 % ~ ~ ~ - ~ ~ ~ ~',
s
[
ANALOG RUN -
40 %
0 1IIIIIIIIIIIIIIIIIIII SECONDARY VALVE POSITION
~
ANALOG RUN 50 %
~~~~T 1.__________
O IIIII'IIIIIIIIII'I'
+25 SEC Fig. 7 Transient behavior during 25 We load rejection at full load O
DRUM PRESSURE 2 PSI Y
Z N
IIIIIillIIlIlllIll!Illlll SECONDARY PRESSURE IC 00 PSI Z
f IIIIIIIIIIIIIIIIIIIIIIIII 400 PSI OUTER CORE FLUX 50 %
lIIIIIIIIIllllIIIIIIlllIll O
SECONDARY VALVE POSITION SECONDARY TRIP I,000,000
-- RECOV ERY y
BEGAN O lIII'''''!"IIIIIIIII
+25 SEC Fig. 8.
Transient behavior during abrupt cut-off of secondary steam flow at 125 we load
..s DRUM PRESSURE 4 PSI y
r~~'%
%, ANALOG RUN p
~~,~~..
~~~~~~
/
- 978 PSI llIIfIliIIIIIlIIIIiiiiiiiiiil!IIIIIIIIIIII SECONDARY PRESSURE IOOO PSI
~,,AN ALO G RU N 400 PSI IIIIIII'IIIIIIIIIIIIIIIIIIIIIIIIIIIIIII PRIMARY STEAM FLOW
~
f-~~
ANALOG RUN 10 0 *4 m
/
%-~~
~
\\
(/
O ll'IIIIIIIIIll'IIIIIIIIIIII'lll'Ill
OUTER CORE FLUX ANALOG RUN 10 0 % 2#
~.
A-O 1'IIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIII'III
+
4--
I S E C.
Fig. 9 Transient behavior during turbine trip-Off at 150 L'e load
DRUM PRESSURE ANALOG RUN
~~~~~~~___
20 PSI
~
E p
985 PSI NOMINAL lllIIlIllIilIl!IilIlIlllIllIllillIIIllIIIIIIIIII DRUM LEVEL AUTOMATIC CONTROL O
IO" h;;;;;;;;;;;;;;;;;
,,i,,ii;ei,,e e i - i - - ' ' + ' ' '
OUTER CORE FLUX
~
- m I
10 0 *4 ANALOG RUN g
l 1
N O ji;i,,,ji,,,,,i,,,,i,,,,,i;,,i,,,,i,,,,,,,,,,,,,
PRIMARY VALVE POSITION 10 0 %
Oh;;;iij,,,,,,,,,,,,,,,,,,,,,,ij,ii,,,,,,ii,,,,,,
+ 1SEC.
Fig. 10.
Transient behavior during simultaneous reactor isolation and scram at full load.
HEAT FLUX 100%
FLOW SECONDS I
I I
)
BOILING 58.5%
A' LI N E
/
O 4
8 l2 TIME, SEC.
Fig. 11.
Reactor decay heat charac-teristic irranediately following scram, showing 1.8 full steam flow seconds equivalent to shaded boiling heat area
30 CALCULATED LUMPED SYSTEM 100 % MISMATCH 25 18.5 psi /sec INITIAL PRESSURE RATE OF RISE
</
gFULL LOAD REACTOR
- 20
/
f ISOLATION DRUM PRESSURE i
d TRACE-y kg/ (%v N
W
-~%
, VESSEL,s e
15 4
/,
-3 W
N s RESSURE P
/
s TRACE
/
f
's a_ 10 I
's
/
's
/
DRUM PRESSURE TRACE
's j
j RATED PRIM ARY STEAM 5
FLOW 130 MWe O
O I
2 3-4 5
Tl M E - S ECON DS Fig. 12.
Detailed comparison of primary system pressure transients for initial period of turbine trip and for simultan-eous reactor isolation and scram tran-sients
o DRUM PRESSURE 2 PSIt 7IIIIIIIfIIfIIIIIIIII h
PRIMARY STEAM FLOW 10 0 %
s_
O IIIIIIIIIIIIIIIIIIII OUTER CORE FLUX 2
50 %
g
-SUBCOOLING EFFECT O lIIIIIIIIIIIIIIIIIII!
100 %
q
~
4 O
IIIIIIIIIIIIIIIIIIII
+
4-5 SEC Fig. 13 Transient behavior during simultaneous trip-off of 2 recirculation pump motors at 150 We load
O e
p ROD E-7 I NOTCH MOVEMENT
~
Ef_
~
E z
_ o
~~
7 til ffff i
INCORE lON CHAMBER IOS C tv T N 1Itiiii!tfIIIII INCORE ION CHAMBER IO9C Z
~
__I IV lffIIff'!IfItfI INCORE lON CHAMBER 186 0 tv 1ItilIIIII111lf
+
4-5 SEC.
CORE PLAN
/~"T N n
I e x
[ 't-7 <!?1--\\
(____n;og._J t~~-1l6:$
%=V Fig. 14. Spatial transient behavior of i
Dresden reactor during 15 second period control rod oscillation at full load, t
.. - - - -