Text

Chapter 12 to Final Hazards Summary Rept for Big Rock Point, Safety Analysis
	ML20030A356
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/14/1961
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
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	NUDOCS 8101090382
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{{#Wiki_filter:f O 1 i SECTION 12 l SAFETY ANALYSIS ( 12.1 SAFEGUARDS CONSIDERATIONS The various safety features of the Big Rock Point plant have been indicated in the course ~of the descriptive information in t-the earlier sections of this report. The general safeguards criteria that guided the design of the plant were considered i in three broaa categories: radiological control during nor-mal operation; design features for preventing accidents; and features for mitigating the effects of accidents in the unlikely event they occur. These criteria are summarized in the following paragraphs. 12.1.1 Safety During Normal Operation All phases of normal operation of the plant, including dis-posal of radioactive waste materials is such that these operations do not result in exposure of any persons on or off the plant premises to rauiation in excess of the permiss-ible limits. Normal operational radiological control is providea by shielding, monitoring, and by equipment sys-tems for hanuling ana aisposint; of radioactive materials anu wastes. 12.1. 2 Safeguards Against Accidents The design features provided in the Big Rock Point plant that are important in preventing serious nuclear accidents incluue the following examples: d 12.1.2.1 The design of the nuclear reacttar is such that the reactor tends to shut itself down upon a potentially dangerous in-crease in its power generation rate---that is, an increase in fuel temperature or steam void volume tends to limit the extent of a power excursion. t 12.1.2.2 Two separate and independent neutron poison mechanisms are provided to assure shutdown of the reactor. The prin-cipal mechanism consists of a set of control rods for auto-matic fast shutdown in response to potentially dangerous levels of certain of the critical process variables. The second mechanism is a liquid poison system which provides an emergency source of negative reactivity for reactor shut-down in the event complete shutdown cannot be obtained with the control rod mechanism. This liquid poison system is manually controlled from the control room. 12.1.2.3 Three separate and independent systems for cooling the reactor are provided to avoid melting of fuel. The primary full power heat removal system is the turbine and main conaenser. The second system uses an emergency condenser yo goMA ,v,-.

i f Section 12 Page 2 i to cool the reactor following scram in the event the reactor is isolated from the turbine and main condenser. The third system introduces cooling water through a spray ring above the core to prevent significant impairment of fuel cladding integrity following scram and loss of all other (. coolant measures. 12.1.2.4 A reactor safety system is provided with sensing devices to detect and prevent potentially dangerous operating conditions from impairing the integrity of the fuel cladding. These devices are designed so that their malfunction will shut down the reactor rather than cause or permit a power rise.

12. 1. 3 Containment Should all other safety provisions fail, any radioactive materials which might be released in the event of an accident are contained within the provided high integrity, essentially vapor-tight steel enclosure.

12,2 CONCLUSIONS

12. 2. 1 On the basis of consideration of the various safeguards features provided in the Big Rock Point plant design and of the analysis of accidents discussed in the subsequent para-graphs of this Section, it is concluded that this plant is designed to adequately protect the health and safety of the public from the hazards of reactor operation, even under accident conditions.

12.2.2 The aiscussion that follows sh6uld not be taken as any indication that such accidents are routinely expected to occur, or that public exposure of the severity described is expected to result from any credible accident during operation of the Big Rock Point reactor, or would be i tolerated if it were humanly possible to predict its occur-Rather, it provides an indication of the upper limit rence. I of public exposure that may result under various com-binations of extremely pessimistic assumptions which have only been hypothesized, and which are not expected to occur I concurrently auring operation of the Big Rock Point reactor. i 12.3 DYNAMIC STABILITY OF THE REACTOR 12.3.1 An inherent characteristic in low pressure (less than 500 psi) i experimental boiling water reactors is the possibility of power oscillations due essentially to the mutual feedback relation-ship between hydrodynamic flow oscillations and the effect of voids on reactor power. These oscillations, which have occurred in some experimental reactors and have been observed and publicized, are in no way indicative of similar pbenomena

Section 12 Page 3 in high pressure (1000 psi) reactors. Extensive analysis and good experimental evidence to date clearly show that high pressure, central station, boiling water reactors are stable and easy to handle. l

12. 3. 2 This is borne out by extensive studies performed by General Electric Company utilizing information developed in its own facilities and in those of the United States Atomic Energy i

Commis sion. The studies have shown that it is possible to design tailing water reactors having both a high degree of stability and a desirably rapid response to changing system demands.

12. 3. 3 The principal criteria governing the design of the Big Rock Point plant which insure stable operation include: high system pressure, forced circulation, a conservative void reactivity coefficient, a long fuel heat transfer time constant, and use of a pressure regulator to control steam pressure. In addition, j

the design of the control and safety system provides sufficient instrumentation to shut the reactor down before any significant impairment of fuel cladding would result. 12.3.4 Analytical studies of the direct cycle,high power density, boil-ing water reactor for the Big Rock Point plant indicate that operation should bc stable. Ivpical calculations are reporte i in CEAle-3193 System dynamic analyses have been made showing satisfactory dynamic response. The basic computer model used for the study was proven by Dresden and VBWR tests. A new two phase hydraulic model which has been checked in out-of-pile experiments, has also shown the Big Rock Point core to be dynamically stable. 12.3.5 The analytical model conceived for the stability analysis includes analytical representatiom of reactor kinetics, fuel rod dynamic heat transfer, hydrodynamic flow characteristics including two-phase effects and thermodynamic effects due to load and subcooling changes. These analytical representa-tions were derived from the basic defining differential equations of physical phenomena and were, in turn, transformed into transfer functions. The stability investigation was accomplished by performing a frequency-phase shift analysis (Bode analysis) on the analytical model. The power conditions at which sta-bility was investigated were 50%j 100%,and 125% rated power for the 157 Mwt core. 12,3.6 The hydrodynamic portion of the analytical model was the distinguishing feature of the stability analysis. It is based upon the physical concepts of momentum interchange, con-servation of energy, and continuity of mass and not upon steam void versus quality correlations, assamed ratios or differences between steam velocity and water velocity in the two-phase flow region. In the past, 'this model has produced predictions which correspond well with natural circulation two-phase flow loop experiments. t

t { Section 12 Page 4 (. 12.3.7 The results of the Big Rock Point reactor stability analyses are summarized below, instability being indicated by zero phase and gain margin. ( Reactor Conaition Phase Margin, degrees Gain Margin. decibels ( 15.7 Mwt core at .50% rated power 66 15 ( 157 Mwt core at 100% rated power 68 14 ( 157 Mwt core at 125% rated power 87 11

12. 3. o The analytical model, as it was used in this stability analysis, has been applied recently to two previous operating conditions of the Dresden Nuclear Power-Plant, a) rated power and, b) 18% average core void test. In both cases it was found that

( the model predictions correlated well with the observed conditions (both cases were stable). The analytical predictions made for this Dresden correlation were obtained by analogue computer techniques. 12.3.9 The 157 MWt core is seen to be stable at all power levels inves tigated. 12.3.10 All of the results obtained in this study were obtained by assum-g ) ing that the change of reactivity due to 100% change in voids for the given power operating condi. tion is $4. 50. This is conser-vatively hi h for all cases considered, and in particular for t partial rated power conditions. ( 12.4 ANALYSES OF OFF-STANDARD CONDITIONS "Off-standa rd conditions" for purposes of the present discus-( sion are considered to be those abnormal conditions arising in the course of regular plant operation in.the power range in which reactor control is held by the operator. Normal system ( disturbances have been discussed earlier in connection with the dynamic stability of the reactor. The following events are the more representative of conditions which might be expected to occur within this category. 12.4.1 Steam Bypass Valve Testing ( 12.4.1.1 The main steam bypass valve will be tested periodically dur-ing normal plant operation to ensure that the valve will open in the event of an electrical load rejection. An adequate test will involve a momentary opening of the valve. The test will be initiated remotely from the control room. .(.

l S:cticn 12 Page 5 Rev 1 (3/23/62) 12.4.1.2 The system controlling the reactor and turbine-generator in the Big Rock Point plant is one making the turbine a slave to the reactor. As described in Section 5.6, the bypass valve regulates system pressure by passing the correct amount of steam without causing a pr essure dis-turbance greater than about 10 psi within the reactor under conditions of an unscheduled electrical load change. The test, therefore, will not result in any significant disturbance in the reactor system, but w'11 cause a momentary change in the output of the turbine-generator as a result of the change it causes in the turbine control valve setting. Pertinent data and tabulated results are given in Section 12. I1.1.

12.4.2 Pressure Regulator Set-Point Changing Fast changes in the initial pressure regulator set point may cause a pressure and resultant flux transient within the I e-actor. With a sufficiently rapid change in set point, a flux transient would result, which could be large enough to scram the reactor at 1257. of rated power. The rate of change will be limited by operating procedures to a value that will not cause such a flux transient. Pertinent data and tabulated results are given in Section 12. I1.2. 12.4.3 Control Rod Withdr awal and Insertion 12.4.3.1 The continuous withdrawal of control rods at power will cause a flux peaking in the area from which the rods are withdrawn. Ther e will be no excursion however, due to the strong negative void coefficient of reactivity. The rate of increase of power in this case is a function of the r, ate of rod removal, the particular rods removed, and the control rod and void configuration at the time of the incident. In any event, the local power, and the overall steam formation will increase smoothly. The bypass valve will continuously increase its opening to receive the larger steam flow. Before the flow capacity of the bypass valve is r eached, the reactor will be safely shut down from the over-flux signal that will arise from the flux monitoring instrumentation. Pertinent data and tabulated results are given in Section 12.11.3. 12.4.3.2 During a similar continuous control rod insertion, the pressure will drop smoothly and slowly at first, then more rapidly. As the primary steam flow is reduced, the initial pressure regulator will direct the closure of the turbine steam admission valve in an attempt to maintain the set reactor pressure. If continued, { the reactor is shut down. 12.4.4 Failure to Replenish Cooling Water in Emergency Condenser ( Gradual evaporation of the shell-side water in the emergency condenser will occur during its operation. Protection against ( failure to replenish cooling water in the emergency condenser is afforded by an initial water supply sufficient to last for

\\ Section 12 Page 6 I-at least four hours and indefinitely with makeup cooling water supplied by a motor driven pump, which is automatically con-( trolled by level sensors. In addition, a low water level alarm is provided in the control room to initiate operator attention. If the operators do neglect to replenish the cooling water, the ( reactor pressure and temperature recorders in the control room will indicate that the system is gradually heating up. Thus, there are several indications of this situation available ( in the control room to the operators in such a situation and appropriate acdon is expected to be taken. ( 12.5 ANALYSES OF EQUIPMENT MALFUNCTION 12.5.1 Safety against equipment malfunction is an essential ingredient in the design of the Big Rock Point reactor. The main princi-ples employed are: (a) use of equipment with the highest practical reliability against failure; (b) fail-safe design,

i. e,any failure will cause action to go in a safe direction; and (c) safety through redundant devices.

i 12.5.2 The designs incorporating these principles have been summarized in the earlier sections of this report. The present paragraphs outline the results of analyses of plant safety if any of the various important equipment pieces or components should fail to function as intended. 12.5.3 Control Rod Drive Malfunction

12. 5. 3.1 Safety in the event of most types of control rod drive failures is inherently provided by the number of rod: available for con-trol, and the fact that in the hot operating condition many rods 5

at random could fail without im1 airing the ability to shut down. Even in the cold clean condition, the reactor may be shut down if one rod of maximum worth is withheld from the core. The various contingencies that may arise as a result of mal-i function of the control rod drive system are discussed in the following paragraphs. i

12. 5. 3. 2 Failure of Normal Drive Power

( Loss of normal control rod drive power to a single rod would not create a serious condition and the reactor may be safely shut down. The rod drive power is obtained from one of two (. full capacity pumps. The loss of a pump is annunciated and the remaining pump would be actuated. Loss of the entire source of normal drive power is expected only with loss of station ( power, which will scram the reactor. Such a loss of normal drive power would not affect the power for scram insertion of the rods which is obtained from a separate stored energy { source. k 4.w e Ja m

( I jl Saction 12 Page 7 ' i I 12.5.3.3 Failure of Emergency Drive Power Safety in the event of loss of the source of pressure for emergency scram is obtained by providing a separate and inde-pendent pressure source for the scram of each r.od.. Th.us, lo s s of one of these sources, a pressure accumulator, can only affect one rod, so that the reactor can be shut down even if the rod cannot be inserted.

12. 5. 4 Control Rod Failure 12 5. 4.1 Separation of Rod From Drive In the event of mechanical failure causing separation of a rod from its drive at the coupling, the rod can still be driven into the core provided no other failure occurs simultaneously in the rod and drive system which would produce an obstruction to rod movement. In case of such a stuck rod, the remaining 31 rods provide adequate margin for reactor shutdown. If several such rods were stuck, reactor shutdown may still be achieved by actuation of the liquid poison system.

Safety against mechanical separation of rod from drive is .provided by the design of the system as described in Section 4 of this report. In comparison with the original design (Dresden type) of control rod system, the Big Rock Point reactor will incorporate the design improvements including the elimination of requiring a drive shear pin, provision for a new semi-spherical collet type coupling between rod and drive, over-travel position provision for checking the integrity of the blade to drive coupling, and the use of a more gradual decel-eration buffer section which decreases the impact and decel-eration loading in the rod and drive system. These improve-ments significantly reduce the potential for an accidental separation of rod from drive.

12. 5. 4. 2 Rod Separation and Fallout l

12.5.4.2.1 Protection against a control rod drop accident is provided by the design and operation procedures described in Sections 4 and 11. However, an accident involving the drop out of a con-i trol rod blade when the reactor is critical has been analyzed. The postulated accident assumes that a control rod blade - sticks in the fully inserted position, that the drive is then fully withdrawn, and that the blade then works loose and drops i out when the reactor is at or near critical. I 12.5.4.2.2 The magnitude of the power excursion that would accompany this accident would depend on the reactivity worth of the blade which in turn depends on the reactor operating condition and the overall control rod pattern at the time of the accident. Maximum rod worths for normal rod patterns (i. e., patterns that do not abnormally pean the flux around any one rod) are ( calculated to be of the order of about 2% ok/k considering i' _i L

i S2ction 12 Page 8 t-Rev 1 (3/23/62) fuel cycling aspects. The average rod worths are on the order of 1.5%8k/k under these conditions. The highest possible rod worths occur with rods withdrawn in an ab-normal pattern contrary to operating procedures causing the flux to peak around one rod. The maximum worths are calculated to be about 4.2% Ak/k for the hot standby con-dition and 3.9%dk/k for the cold reactor condition. r 12.5.4.2.3 Conservatively assuming that the Doppler effect is the only / ',, a negative reactivity effect acting to limit the excursion, the drop out of a maximum worth rod with a normal core flux distribution in the hot standby condition would result in a nuclear excursion that would increase the fuel temperature ,f ~ at the hottest point to about 4100 F. No fuel center melting or fuel rod cladding rapture would be expected to occur during the burst. Pertinent data and tabulated results are given in i Section 12.12.1. 12.5.4.2.4 A ca.lculation made to determine the rod worth to reach the rupture damage threshold indicates that rod wor th would have to be about 2.5%Ak/k for the hottest point to reach an estimated 8000 F rupture temperature. Rod worths this high would be difficult to achieve during routine operation of the reactor. 12.5.4.2.5 A rod drop out with the highest possible worth (4.2%dk/k) could r esult in extensive fuel damage. However, this has little probability of occurrence because of the compounded errors and failur es that would have to occur. If a drop out accident were to occur coincident with the abnormal flux distortion necessary to give this high worth, conservative calculations indicate that about 3% of the total fuel in the cor e would reach temper atures of 8000 F or above, and the rods containing this fuel (about 8% of the total number of rem in the core) could rupture. Damage to a lesser degree than gross rupture would probably occur to some of the other fuel rods because about 11% of the fuel in the cor e would exceed the 5000 F melting temperature of UO. It 2 is believed that these calculated effects exceed those that could actually occur because other negative reactivity effects, such as fuel rod expansion, moderator expansion and moder - 1 ator void formation, would act to limit the excursion sooner than predicted by these calculations which considered the Doppler effect from fuel heating as the only shutdown mech-i anism. The energy release resulting from an accident in-volving the 4.2%dk/k highest possible rod worth would not f be expected to be enough to endanger the reactor vessel or other primary system components, and fission products released from the fuel would be contained in the primary system. Pertinent data and tabulated results are given in Section 12.12.2 ( 12.5.4.2.6 The following design features and operating procedures show the improbability of a rod drop accident:

t S2ction 12 Page 9 Rev 1 (3/23/62) (a).The control blades have been carefully designed to minimize the possibility of sticking in the core. The 5/16-inch thick blades travel in a 21/32-inch gap between fuel channels and the blades are equipped with rollers which ride on the channel walls when it, contact is made. (b) The new coupling and other control rod drive improve-ments (see Section 4.4.1) significantly reduce the probability of an accidental separation of the red from the drive. Couplings of this design are currently I undergoing extensive tests under simulated reactor conditions and at conditions more extreme than those expected to be encountered in service in the reactor. (c) The design provision for testing coupling integrity by bringing the drive to the overtravel position provides an easy method for verifying rod coupling prior to reactor startup when rod following might not be checked by observing the response of the in-core monitors and other nuclear instrumentation when a drive is moved. (d) Operating procedures require rod following verification by the overtravel test method or by observing the response of the reactor instrumentation after every scram for each rod that was inserted on the scram. During reactor operation rod following will be verified by observing reactor instrument response with drive movement. Whenever reactor multiplication is above k = 0.995, a drive is not to be retractell further than an amount corres-ponding to A k = 0.01 without verification that the rod is fol-lowing the drive. If following of any rod cannot be veri-fied, the drive will be fully inserted and left in this position until the malfunction can be found and corrected during a reactor shutdown period. 4 (e) Operating procedures require that control rod movements i follow preplanned patterns designed to minimize the reactivity worth of individual rods. Thus, extensive fuel damage would not be expected even if this accident =I were to occur. 12.5.5 Loss of Recirculation Pumps -t 12.5.5.1 In the event of loss of the recirculation pumps, the power level will drop and settle out at a reduced level dependent on the I natural circulation flow rate. Computer analyses were made at the 277, overpower condition for both the 157 Mw - 1050 psia core configuration and the 240 Mw - 1500 psia core configuration. g The analysis considered the loss of both pumps, as tne loss of a single pump will always result in a less severe transient with (- respect to the burnout margin.

4 S2ctirn 12 Page 10 Rev 1 (3/23/62) 12.5.5.2 The analysis indicates that with a loss of pumps, the reactor conditions will level out at about 40[ power and 40$ flow fol-lowing loss at 1050 psia operation, and about 3$ power and 4070 flow at 1500 psia. The minimum burnout ratio reached during the transient is about 1. I and 1.15 for operation at 1050 psia and 1500 psia, respectively. Pertinent data and tabulated results are given in Section 12.12.3. 12.5.6 Loss of Electric Load 12.5.6.1 A sudden loss of electric load will cause a partial closure of the turbine valves. The bypass steam valve will open automatically to hold reactor system pressure constant by providing a path to the condenser for the turbine rejected steam. Thus, there will be no significant flux transient from the load change. The control rods would be manually repositioned to hold reduced power. Pertinent data and tabulated results are given in Section 12.12.4. 12.5.6.2 Analytical studies of this transient in support of design of the bypass valve have been carried out to assure that bypass valve action can, in fact, regulate system pressure to avoid severe pressure and flux transients within the reactor system. 12.5.6.3 Main steam shut off by stop valve closure under complete load rejection has been analyzed for 240 Mwt and 1500 psia assuming the bypass valve does not operate. The stop valve closes in 0. 5 seconds. The flux rises rapidly and it was assumed that flux scram was initiated when the flux reached 115 percent (for complete load rejection at 157 Mwt and flux scram at 125%, the pressure and flux transients are corres-pondingly less sever e). The resulting calculated flux peak was 125 percent. In case the flux scram failed to function, the reactor will be scrammed by high reactor pressure at 50 psi above normal pressure. The emergency condenser would be brought into service by the very high reactor pres-sure signal at 100 psi above normal pressure and would limit the pressure rise to avoid lifting the steam drum relief valves. Pertinent data and tabulated results are given in Section 12.12.5. 12.5.7 Closure of Steam Line Backup Isolation Valve The steam line that transports steam to the turbine, contains an automatically-initiated motor-operated isolation valve located between the steam drum and reactor containment vessel. If this valve was closed simultaneously with power operation, all steam flow would be cut off and reactor pres- 'I sure would rise. To avoid such a pressure increase, the safety circuit is set to scram the reactor on partial closure of this valve. This valve closes in about 40 seconds which i allows sufficient time for initiation of the scram before the reactor pressure can rise significantly. A signal from very high reactor pressure initiates diversion of steam flow from ( the steam drum to the emergency condenser. If a scram does not occur due to failure of the initiating signal (valve closing),

[

a high flux scram will subsequently occur, as reactor flux will l

i Section 12 Page 11 Rev 1 (3/23/62) rise simultaneously to the neutron flux scram level. If this in turn fails, another signal is provided by a pressure scram when the high pressure scram level is r eached. If none of these devices operate, the steam safety valves on the steam drum would operate to limit pressure rise. The size of these valves is large enough to pass the steam generated in such a case with 1 the reactor operating initially at r ated power. 12.5.8 Coincident Steam Shut Off With Failure to Scram 12.5.8.1 This is a hypothetical accident which sets the design of the safety relief valves located on the steam drum. The improba-bility of having such an accident can be realized by considering the number of coincidentil events necessary in order for safety valve lifting to occur: 12.5.8.1.1 Main steam flow would have to shut off completely. This would require either that a full load turbine trip occur simultaneously with failure of bypass to open or an abnormal condition causes the main steam line backup isolation valve to shut. 12.5.8.1.2 Failur e to scram. The abnormal condition which caused the backup isolation valve to shut also would initiate a scram. Also, both the high flux and high reactor pressur e scrams would have to fail in order for an automatic scram not to occur.

12. 5. B.1. 3 In addition, the oper ator would be alerted by multiple alar ms and loss of load indications and would be able to initiate manual ser am.

12.5.8.2 Failure of the bypass valve during loss of load coincident with failur e to scram is an incredible combination of events. Failur e to scram with closur e of the backup isolation coincident with failure of flux, pressur e and manual scr am is also an incredible combination of events. 12.5.8,3 In order to cause peak center fuel temperatures above melting it is necessary to postulate the coincident failur es indicated above (including incredible failur es) to the point where manual action scrams the reactor. Such a hypothetical case would result in lifting the safety relief valves and in peak center fuel temperatures r eaching about 150 percent of rated maximum. l The relief valves provide sufficient capacity to limit the peak pressure below that allowable by Code. Pertinent data and tabulated results are given in Section 12.12.6. 12.5.9 Loss of Condenser Vacuum i If the condenser vacuum is lost, the niain aactor heat sink is lost. If no action wer e taken by the automatic system protection circuit, condenser pressure and temperatur e would increase to a point where the turbine would be damaged. To eliminate this possibility, vacuum-sensing devices are used to transmit a scram signal to the reactor and to trip the turbine upon loss of 4 l (

4 S2ctisn 12 Page 12 Rev 1 (3/23/62) vacuum. An independent signal will initiate closure of the bypass valve to prevent turbine damage. When condenser vacuum decreases to about 13 inches Hg Absolute pr essure, the reactor is scrammed. If the cause of the incident was loss of circulating flow due to power failure, the reactor will aboreceive automatic shutdown signals from the entire safety system which is designed to be normally energized. Thus, loss of condenser vacuum should have initially pro-duced a reactor shutdown. If this does not occur, closure of the stop and bypass valves will produce a reactor scram through the mechanism of high r eactor pressure, and very high reactor pressure signal will initiate cooling via the emergency condenser. Additional information is given in Section 12.12.7. 12.5.10 Failure of Reactor Safety System Because of the failsaie design of the reactor safety system, significant malfunction will cause an immediate insertion of control rods and reactor shutdown. Also because of the failsafe design and the number of sensors pr ovided and variables monitor ed, failure of a sensor will not impair the ability to transmit a scram signal or effect a scram. 12.5.1I Fuel Cladding Failur e In the event of a fuel cladding failure, protection is provided by the off-gas holdup system, its x adiation monitor and valve isolation pr ovisions. The system is designed for manual closure of the gas holder isolation valve at a noble gas con-centr ation that could be reasonably expected to deliver the maximum permissible offsite dose if the operation were to a continue unabated for a year. Automati: shutdown will occur at a greater gas release rate in time to keep per sonnel exposur es within acceptable limits. 12.5.12 Loss of Feedwater Heaters Sudden loss of all the feedwater heaters would cause an immediate but smooth rise in flux. It is expected that the i flux will reach the ser am value within a few minutes with no significant overshoot or damage to the fuel. i 12.5.13 Loss of Feedwater Loss of feedwater will result in gradual lowering of the water '( level in the steam drum and if continued, will automatically initiate x eactor shutdown. Adequate r cactor cooling would then be available through the main condenser. t 12.5.14 Reactor System Ruptures The most severe reactor system rupture with res;sct to the resulta'nt effects on the environment is r epresented by the definition and analysis of the " maximum credible accident" i

t I-Section 12 Page 13 given in Section 13 of this report. Less severe system ruptures have been considered and analyzed. Results of these analyses are given in the following paragraphs,

12. 5.15 through 12. 5.17.

( 12.5.15 Reactor Idser Rupture

12. 5.15.1 In order to determine the effects on the core from the " worst credible" hydraulic condition caused by flow and pressure transients,a riser rupture accident was analyzed. For purposes of this study it was assumed that one of the six steam risers (the'se are 14" diameter) suffered a complete instantaneous circumferential break with coolant issuing i

from both sides of the break. Both of the cores (157 hiwt and 240 hiwt) were considered. Basic assumptions for the analysis included: (a) the steam and water in the system becomes a uniform, homogeneous mixture for the duration of the transient, (b) the recirculation pumps continue to run, (c) steam flow to the turbine-main condenser and feedwater flow to the drum are ignored, and (d) there is no physical damage to the steam crum internals which might obstruct the outflow of steam and water. 12,5.15,2 Calculations indicated results which are summarized as follows: (a) The maximum effect on the physical stability of the l l core was found to exist with the 157 hiwt core loading. This results from a smaller core flow cross-sectional area resulting in a higher core flow velocity than in the case of the 240 hiwt core. (b) The reactor safety system will initiate a scram immedi-ately after the break as a result of signals from low 2 water level in the drum and low water level in the reactor, which also will initiate closure of the steam line backup isolation valve and all other automatically operated open enclosure penetrations. (c) The maximum lift force on a fuel bundle was.400 pounds, which is less than the bundle weight of 420 pounds, thus, I fuel will not be lifted out of the core grid. 12.5.15.3 The radiological effects of such a riser break would be less severe than those for the " maximum credible accident," because the pressure in the containment vessel would bc comparatively lower, and because water remains for cooling the core during a relatively lonber time period. 12.5.16 Rupture of hiain Steam Line 12.5.16.1 The automatically actuated backup isolation valve and the sys-tem pressure controlled valves (i.e., turbine control and bypass g

Section I? Page 11 valves) would act to protect against the radiological effects of a main steam line rupture. In the event such a rupture occurred in the portion of line inside the containment vessel, both isolation valve s (backup and turbine control) would act to confine the released water and steam to within the con-tainment vessel. In the event the steam line ruptured outside the containment vessel, the backup isolation valve would act to limit the total release of steam-water and fission products which may be included.

12. 5.16, 2 The raalological effects on the plant environment of a steam line rupture accident occurring inside or outside the containment vessel would'be less severe than those effects resulting from the " maximum credible accident. " In comparing the effects between the two rupture locations, the "outside" break would be the more severe, since the radioactive materials would be free to move into the tur-bine building and from there, a portion would be free to flow through the turbine building ventilation louvers to the outside. However, the total amount of fission products released to the outsir.e wuld be below those released during the "maxin.um credi' le accident. "

u 12.5.17 Emergency Condenser Tube Failure

12. 5.17. 1 Failure of a tube in one of the condenser tube bundles would not interfere with tne ability of the remaining tube bundle to cool the reactor. Any failed tube bundle can be isolated by manual actuation (from the control room) of the motor operated valves.

12. 5.17, 2 Tube failure would be uetected by the radiation monitor located on the exhaust vent.

12. o ANALYSIS OF OPERATOR ERRORS 12.6.1 Safety in the event of operator errors is provided primarily by the reactor safety system, and special features,(i. e.,

interlocks) are provided to prevent startup if the reactor safety system has not been properly activated. Safety against operator errors is also provided by careful selection and thorough training of the operating staff, working conditions conducive to attention to duties, systematically planned oper-ating and maintenance procedures, and by special design measures for control of certain operations. For example, with respect to the latter point, only one control rod can be withdrawn at a time and at a limited rate. 12.6.2 The incidents or conditions arising wholly or in part from operator errors which are analyzed to establish the limits or response times of the reactor safety system.or other features to limit the consequences if such errors or other similar untoward incidents do occur,are discussed in the following paragraphs,12. 7 through 12. 9. 4

Se'cti:n 12 Pege 15 Rev 1 (3/23/62) s 12.7 STARTUP ACCIDENT 12.7.I Control rod withdrawal procedures, instrumentation, and safety devices are designed to provide contr ol of the reactor during control rod withdrawal to criticality, and on to low and rated power operation. If, during startup, the operator should deviate from procedure and withdraw an off-standard sequence of con-trol rods so that the rate of power rise is too rapid for operator action, the perlod monitor scrams the reactor. The control rod withdrawal interlocks (see Section 7.7.1) would insure that the neutron flux instrumentstion is set down to its most sensitive range in order to start up the reactor. With this instrumentation set properly on its most sensitive range, the reactor would be scrammed before power could increase appreciably as a result of continued, rapid, withdrawal of control rods. For such an accident to be serious, one must assume failure of these multi-ple devices in addition to the errors made. 12.7.2 The worst credible startup accident for the Big Rock Point reactor also includes a particular rod withdrawal sequence, that is, removal of control rods in local groupings, which maximizes the rate of power rise. Normally, procedures prescribe a more uniformly distributed pattern of control rods so that flux :s not spiked in one location. With a distributed flux shape, the strongest rod is reduced in worth. l 12.7.3 No fuel damage will occur if safety circuit trips function as designed. Even under the pessimistic assumptions of this accident, no center fuel melting is expected to occur. 12.7.4 The effects of this conservatively chosen accident are thought to encompass all other credible startup-type accidents which one might hypothesize for this reactor. Results of this analysis are summarized below, and other pertinent data and further results are given in Section 12.13 for both the hot and cold startup accidents. 12.7.5 Accident Conditions 12.7.5.1 Initial power level is eleven decades below z ated power, l 12.7.5.2 The moderator is at 100 F. 12.7.5.3 A conservative Doppler coefficient of -0.8 x 10 ok/k/ F ( is used. 12.7.5.4 No credit is taken for the negative void coefficient of reactivity. 'l ^ Calculations show that the " worst" contr ol rod can be achieved 12.7.5.5 by peaking the flux at the center of the core. For the cold reac-tor, with a clumped rod withdrawal at the center of the core, 3 the worst rod is estimated to have a ok of 0.039. 12.7.5.6 Subsequent withdrawal of the worst rod at its maximum rate (three Tnches per second) causes the excursion.

Section 12 Pa,e 16 12.7.s.7 Both the scram from short period, and neutron flu:. scram at 125% of each scale are assumed to fail. The latter failure also implies that the low level setdown permistive switch on control rou withurawal failed to operate. Scram is assumed to occur at 125% of rated neutron flux.

12. 7. 5. 8 The cola shutaown margin of the control rods is taken as

0. 05 Lk/x.

12. 7. 5. 9 The scram uelay time and control roa insertion speeds are as follows:

Rod Position Time (seconas) (% inse rted) 0 (scram signal) 0

0. 2 0 (rod motion starts)

O. 5 3

0. 6 10 1.0 50

2. 5 90

3. 0 100

12. 7. S.10 Fuel time constant

is 6. 3 plus seconds.

12.7.6 Calculateu Accident Effects

12. '. 6.1 Unuer the conditions described, the maximum rate of reactivity aduition is about 0. 46% Z.k/ k/ second. Total reactivity addition was 3. 9% tk/.. The minimum period created was aoout 10 milli s e conas.

12. 7. 6. 2 As the fuel temperature rises as a result of the power excursion, the prompt Doppler coefficient of the fuel contrioutes nebative reactivity, thus tenuing to reuuce the magnitu e of the continuin,,

rou witharawal reactivity aduition. The Doppler effect causes the neutron flux ta pean before any control roos have oeen inserted. The flux pean is 350 times rr. tea, and occurs about 0,10 secont.s af ter ratem power is reacheu.

12. 't. 6. 3 The acciuent occurs over a sufficiently short periou of time so that no voia formation has resulteu. The pressure increase effects of the startup accioent a re less than 50 psi.

I

12. s. 6. 4 Control rom insertion begins 0. 2 seconds after the high flun trip, with insertion speeds in accorcance with the scheuule above. The neutron flux ecays as rapiuly as it increased.

Fuel time constant is mefinea as the time requireu to transfer 63% of the heat generateu at any instant in the center of a fuel rou, to the coolant.
e

h Section 12 Page 17

12. 7. 6. 5 The resultant average fuel temperature is about 960*F.

12.7.6.6 The maximum fuel center temperature at the core hot spot is about 3900*F. 12.7.7. Conclusion 12.7.7.1 It is not expected that the startup accident will ever occur as strong procedural and mechanical safeguards are provided to prevent it. The acciuent, as discussed above, results from a com-bination of pessimistic assumptions of operator error and coinciuent equipment failure. The principal measures pro-viaed to avoid such an incident are:

12. 7. 7. 2 Procedures a) Startup procedures which prescribe distributed rod with-urawal patterns. That is, control rods will not be with-arawn in the configuration assumed for this incident.

b) Procedures limiting continuous withdrawal of a control rou. c) Required observation of count rate and periou instrumen-tation uuring startup.

12. 't. 7. 3 Mechanical Devices a) Source level flux counter reau-out.

b) Periou scram circuitry. c) High sensitivity flux scram circuitry. a) Startup interlock on flux monitor scram settings.

12. o COLD WATER ACCIDENT 12.8.1 Because a boiling water reactor holds considerable reactivity in steam voias during boiling operation, any means which might quicaly reauce these voios in the core coula precipitate a t

nuclear excursion. t

12. 8. 2 The worst " cold water accident" considered in the des'ign of the Big Rock Point plant involves a rapid introduction of cold water into the reactor from a cold coolant loop. Normally, this

( accident is not expected to-occur as the reactor is operated with both circulation loops in service. It is assumed for this analysis that one loop is out of service during startup from a cold con-dition and that the reactor is brought up to 50 percent of 240 Mwt power level. It is then assumed that the operator violates pro-cedure in the act of placing the second loop in service. 1 P

Section 12 Page 18 - 12.8.3 The introduction of cold water does not cause a violent excursion but leads to an essentially linear rate of power increase over the period at which the admission: valve is opened. The rise in power can be corrected by an alert operator before any damage occurs to the core. However, it is assumed that the operator failed to note the increase in power level and thus took no corrective action, and that the high neutron flux scram instrumen-3 .tation fails to operate. Shutdown is affected by a high pressure scram. 12.8.4-Accident Conditions p

12. 8. '4.'1

'One recirculating pump is operating, the other is shut down and its discharge valve is closed. 12.8.4.2 Col'd water, at 100*F, is assumed to exist in one coolant loop. 12.8.4.3 The reactor is bding started up from a cold condition, and at the time of the incident is. at rated pressure, and 50 percent of rated powe r (240 Mwt). .12.8.4.4 The operator violates procedure by starting the second pump without first opening the small bypass valve.and allowing 60 seconds for the cold recirculation loop to warm up. Instead, he starts the pump, then immediately initiates opening of the discharge valve. 12.8.4.5 No credit is taken for heating of water in the cold recirculation loop by natural circulation and heat transmission through the ~ pipes prior to the incident. This is a highly conservative as-sumption as considerable heating will take place by natural circulation through the 4" cross-tie line connecting the recir-culation pump suction lines. The principal purpose in making this assumption was for design in order to establish the maxi-mum opening rate for the discharge valve. As indicated earlier in this report (Section 5. 4.1. 5) this cross-tie line provides cir-culation through all four downcomers even if one loop is out of se rvice. This feature will preclude ever having water whose temperature is 100*F at most, below the temperature of the water in the reactor, available for a cold-water accident.

12. 8. 4. 6 It is further assumed that the operator fails to note the increase in powe r level and thus taken no corrective action.

12.8.4.7 The high Icogl flux scram is assumed to fail, with the scram finally being initiated by high reactor pressure. t 12,8.5 Calculations t 12.8.5.1 The rates of reactivity addition from starting one cold recir-culation loop with the othe r in service are uniquely set by the design value for valve movement which is 5 in/ min. 4

12. 8. 5. 2 Under accident conditions postulated, transients resulting from various cold-water admission rates were analyzed by anal,og

. compute r. The power peak associated with the 5 in/ min valve

53 ction.12 Page 19 travel time was 190 percent of rated power. The maximum hot - spot fuel center temperature reached about 180 percent of rated maximum. The results indicate that some fuel center melting may occur, however, core damage with subsequent release of fission products is not expected to occur. 12.8.6 Conclusions It is not likely for any cold-water accident to occur because of the strong procedural and mechanical safeguards provided to avoid such an accident. The principal measures available are; 12.8.6.1 Procedure s The procedural failures necessary to permit such an incident are: The operator must fail to wait about 60 seconds to allow the a. coolant loop to warm up by natural circulation through the bypass valve before starting the recirculati,on pump. b. The operator must fail to observe the temperature of the water in the coolant loop which would show that it was not up to temperature. The operator must fail to observe the power rise caused by c. the error, and thus take no corrective action. 12.8.6.2 Mechanical Devices a. An interlock is provided to prevent: i. Starting the recirculation pump unless its discharge valve is closed. e ii. Starting the recirculation pump unless the bypass is open. b. The 4" cross-tic line provided, which connects the recir - culation pump suction lines inherently limits the total reac-tivity available from subcooled water in the non-operating loop. If the accident did occur, power level would rise sufficiently c. slowly that either the high neutron flux or the high pressure scram functions, or both, w ould provide adequate reactor protection. The fairly slow, nearly linear rise in flux level is due to the self-limiting, negative feedback effect of reac-tor void reactivity. d. In the event radioactive gases have been released from the fuel, the off-gas radiation monitors would effect automatic closure of the isolation valve in the off-gas line. 12.9 FUEL LOADING ACCIDENT 12.9.1 As indicated in Section 11, refueling will be done according to the following sequential steps:

Secti:n 12 Page 20 12.9.1.1 Immediately preceding the removal of a fuel bundle from the ~ core, an appropriate, strongest, adjacent rod is withdrawn and multiplication count observed to assure the existence of subcriticality. 12.9.1.2 .Following determination of the existence of suberiticality, all rods are fully inserted. Power for motion to the refueling cranes which are positioned over the reactor at the time is blocked by interlock whenever any rod i's withdrawn from its fully inserted position. f-12.9.1.3 The cid fuel bundle is removed and replaced as scheduled. 12.9.1.4 After the new bundle is in place, the multiplication count is again observed to establish the effect of the new bundle accord-ing to the above first step. This step may be simultaneous with, or followed by, the above first step. 12.9.2 By postulating several possible, but improbable errors, one can estimate the severity of a refueling accident as follows:

12. 9. 2.1 If_ the above pr ocedure is followed and he loading change has been planned correctly so that keff = 0.99 with the adjacent blade withdrawn after the change is completed, there is no excursion; the core is subcritical during the entire process.

12. 9. 2. 2 If keff = 0. 99 with the adjacent blade withdrawn before the loading change is made, the core would be subcritical after the change only if the blade remains inserted. Withdrawal of the blade would produce super-criticality terminated immediately by a low level flux scram. After the change had been made, the configuration would have constituted a violation of the shutdown margin t

requirement. 4 12.9.2.3 If k ff.= 1. 0 prior to the loading change with all blades inserted (violating the shutdown margin requirement), then the core would be supercritical anytime a blade is withdrawn during a loading change involving the replacement of a single fuel bundle. Assuming no blade is withdrawn (a second violation), the resulting keff after the replacement is completed would be less than 1. 01,

12. 9. 2. 4 If keff = 1. 0 during the replacement with all blades inserted and an empty position in the lattice (water hole), then keff >1. 0 before the replacement with a full loading. This is an impossible situation unless more than one bundle had been removed daring the change.

i' However, if this violation did, in fact, occur and a fresh fuel element was inserted, then the resulting keff = 1. 012, with all 4 blades in. I Withdrawal of the adjacent blade during insertion of the fuel bundle is not considered possible in view of the interlocks. This refueling i accident involving an excess of 1. 2% okeff resulted from failure to satisfy shutdown margin at several times during the refueling process, violation of the blade withdrawal check procedure, and ( l

Section 12 Page 21 addition of more than one bundle between checks. Paragraph

12. 9. 3 below shows that this wo rst refueling accident would result in a power excursion.

12.9.3 LoaGng a fuel bundle into the core at the maximum hoist speed at the elevation of bundle insertion into the core of 100 inches per minute under conditions given by above item 12. 9. 2. 4 produces j a power transient which turns around at a low level by the Doppler effect. Assuming that the fuel bundle inserted all of its excess reactivity in 50% of its travel, results in a reactivity insertion i rate of about 0. 06% Ak/k per second. The result of such a transient is a small rise in fuel temperature, with the peak fuel center temperature below 2000*F. In addition, if the operator responds and stops crane motion, the transient is terminated at a still lower level. 12.9.4 In the event of a loading error as depicted by item 12. 9. 3 above, but at the instant the fuel bundle begins its travel into the core, the hoist cable breaks or other failure occurs in a sequence that neatly guides the free falling bundle into the open fuel space, the resultant transient is more severe. Under such a drop, assuming that buoyancy is acting to reduce the accel.eration forces, the time involved for insertion of half the length of a fuel bundle is about 0. 45 seconds. Assuming that all of the available excess reactivity is inserted in half its length, the rate of reactivity insertion is about 2. 7% ok/k per second. The resultant power transient is turned by the Doppler effect after reaching about 1400 times rated power. The peak fuel center temperature is about 7400*F, consequently, such a transient results in some center fuel melting (about 7%). However, the fuel cladding does not reach perforation tempe ratures nor are fuel vapor pressures sufficient to rupture the cladding. Thus, it is not expected that significant amounts of fission products would be released. 4 12.10 SAFETY AGAINST FIRES AND OTHER GENERAL HAZARDS 12.10.1 Fires Like most conventional power plants, the Big Rock Point plant will employ some flammable materials: fuel oils (for the diesel-driven emergancy generator, emergency fire pump) lubricating oils, transformer and circuit breaker oils, and t hydrogen (for generator cooling). There is also hydrogen in the condenser air-ejector off-gases. The fire protection measures follow conventional practice. ( 12.10. 2 Floods { There is no danger of flooding at the plant site. Fu rthe r information relative to the flooding potential may be obtained by reference to the consultant's report on geology and hydrol-t ogy for the site contained in Volume Two. I

Section 12 Page 22 Rev 1 (3/23/62) 12.10.3 Earthquakes 1 The plant site is located in an area of low earthquake intensity, and plant construction is in accordance with the conventional design practice for nuclear facilities in such zones. Further information on the seismology of the plant area may be found in Volume Two, containing the consultant's report on seismology. 12.10.4 Weather and Miscellaneous Loads All plant structures are designed to withstand maximum wind and other potential loadings in accordance with standard codes and normal engineering practice. 12.-11 TABULATIONS OF OFF-STANDARD CONDITION ANALYSES 12.I1.I Bypass Valve Testing 12.I1.1.1 Conditions The bypass valve is opened momentarily by a reduction in the bypass valve controller set point. 12.I1.1.2 Analysis 1 (a) Reactor at full power 157 Mwt,1050 psia. (b) Assume steam flow through bypass valve to vary linearly with valve position. 12.I1.1.3 Re sults (a) The following tabulation shows results of bypass valve 1 opening to 5% and 10%, and then reclosing the bypass valve in 0. 5 and 1.0 seconds. Flux Maximum Bypass Valve Bypass Valve Neutron Pressure Peak Fuel Minimum Position Change, Change Rate,

Peak, Change,

Temp Change, Burnout %/Second Psi F Ratio i 5 5 100.6 -0.8 0 2.57 5 10 100.4 -0.6 0 2.57 10 10 100.8 -1.2 0 2.57 10 20 100.6 -1.0 0 2.57 (b) These results are shown by computer curves (Figures

12. I and 12.2) Pages 32 and 33 of this section.

l 1 i

Section 12 Page 23 / Rev 1 (3/23/62) 12.I1.2 Increasing Set Point on Pressure Regulator 12.I1.2.1 Conditions Increasing the set point of the initial pressure regulator causes t the turbine admission valve to close momentarily; this results ,9 in increasing the pressure of the system, and the turbine ad-mission valve then reopens to stabilize the pressure at the new set point. 12.11.2.2 Analysis (a) Maximum rate of changing the set point is 20 psi per second. (b) Reactor at full power 157 Mwt,1050 psia. 4 (c) Assume bypass valve does not operate. (d) Assume set point is increased 5,10, and 20 psi above the original setting. 12.11.2.3 Results (a) The following tabulation shows the results of such a set point change on the system: Set Point Rate Peak Core

Change, of Peak Maximum Peak Surface Exit Psi Above Change, Neutron Pressure Fuel Hea Steam Quality,(3)

Burnout Original Psi per F lux, Rise,

Temp, Flux, Setting Second Psi Ratio 5

20 106 6.6 103 103 101.4 2.5 10 20 112 12,6(4) 106 106 102.8 2.4 20 20 124(6) 24.2( } 111 112 105.8 2.3 Notes: (1) Peak fuel temperature = (T-%) (2474) + 546 = F. TUU-(2) Peak heat flux = ( )(352,000) = Btu /hr-ft. 300 '( (3) Steam quality =.(Qua tY )(4.9) = % c re exit steam quality. 0 (4) Bypass valve actually would have operated, which would result in lower peaks. (5) Bypass valve actually would have operated, which would result in no scram. (6) Without bypass valve operation, reactor would have scrammed from t high neutron flux and other peak values would have been lower. i (b) Thsse results are shown by computer curves (Figure 12.3) 4 Page 34 of this section. L

S 2 ction.12 - Page 24 Rev 1 (3/23/62) s 12.11.3 Rod Withdrawal at Power

12. I l'. 3.1 Conditions 4

The continuous withdrawal of a control rod when the reactor is operating at rated power would result in a power excursion which would be terminated by a scram initiated by the high neutron flux trip. To continuously withdraw a control rod an operator must actuate two controls simultaneously, since normal operation is designed to withdraw a control rod in single-notch steps. Controls do not allow ganged rod removals. 12.I1.3.2 Analysis (a) Reactor at full power 157 Mwt,1050 psia. (b) Reactivity insertion rates at 0.10, 0.25, 0.50, 0.67 and 1.00 dollars per second. These rates cover the range of possible reactivity values insertable by control i rod withdrawal at maximum drive speed. (c) Analog computer runs were made for the withdrawal of a $ 1.00 rod at the above rates and the results tabulated for the case of the reactor not scramming, and for the case of reactor scram from high neutron flux. 12.11.3.3 R esults (a) One dollar total reactivity, inserted; without reactor scram: Reactivity Insertion Rate, Cents Per Second 10 25 50 67 100 Peak Neutron Flux, Percent 123 138 155 167 200 Maximum Pressure Rise, Psi 4.4 5.3 5.6 5.6 5.6 Initial Peak Fuel Temperature, Percent 114 114 115 115 115 Initial Peak Surface Heat Flux, Percent 115 116 116 117 119 Initial Core Exit Steam Quality, Percent 119 119 120 121 123 Minimum Burnout Ratio 2.21 2.19 2.19 2.17 2.13 t (b) Reactor scram from high neutron flux at 1257 : Reactivity Insertion Rate, Cents Per Second 25 50 67 100 Maximum Net Reactivity Inserted, Cents 1I 24 32 42 Peak Neutron Flux, Percent 128 130 140 170 Maximum Pressure Rise, Psi 0.6 0.5 0.2 0.2 Peak Fuel Temperature, Percent 108 103 104 106 Peak Surface Heat Flux, Percent 110 106 108 112 Core Exit Steam Quality, Percent 111 106 106 110 Minimum Burnout Ratio 2.33 2.43 2.35 2.33 i (c) Th'ese results are shown by the computer curves (Figures 12.4 and 12. 5. ) Pages 35 and 36 of this section.

' t Section 12 Page 25 Rev 1 (3/23/62) i ~~ . j 12.12 TABULATION OF EQUIPMENT MALFUNCTION ANALYSES 12.12.I Fallout of Average Worth Rod 4 12.12.1.1 Conditions i (a) Assume rod fully inserted just prior to fallout. (b) Average worth rod for 4 out at rate given by gravity, retarded only.by buoyancy effect. ~(c) Fuel and moderator temperatures initially at 546 F. ( 12.12.1.2 Analysis (a) Initial reactor power at 10-rated. -(b) Average worth rod is 1.4% A k/k (~ $ 2. 00). (c) Maximum reactivity insertion rate is 6.5% A f/sec. (d) Negative reactivity effect of Doppler only. ') Scram on high neutron flux, but this has no effect on initial power peak. 1 12.12.1.3 Results ^ (a) Minimum reactor period, m sec 3 (b) Peak neutron flux, times rated 520 (c) Peak fuel temperature, F 4090 (d) Peak reactor power, Mwt 8 x 10 (e) Integrated energy, Mw-sec 1555 \\ (f) Power and fuel temperature curve (Figure 12.6 ) on Page 37 of this section. 12.12.2 Fallout of Maximum Worth Rod t 12.12.2.1 Conditions Same conditions as for average worth rod fallout above. i 4

t Saction 12 Page 26 Rev 1 (3/23/62) i ( 12.12.2.2 Analysis (a) Initial reactor power at 10-10 rated. 4 (b) Maximum worth rod is 4.2% Ak/k. ?' (c) Maximum reactivity insertion rate is 19.5% A (/sec. (d) Negative reactivity effect of Doppler only. (e) Scram on high neutron flux, but this has no effect on II initial power peak. 12.12.2.3 Results (a) Minimum reactor period, m sec 16 (b) Peak neutron flux, times rated - 6250 (c) Peak fuel temperature, F 13,900 6 (d) Reactor peak power, Mwt 1 x 10 (e) Integrated energy, Mw-sec 3310 (f) UO2 exceeding 5000 F (melting), Ib 2200 (g) UO2 exceeding 8000 F (clad rupture), Ib 529 (h) UO in ruP ured fuel rods, Ib 1550 t 2 (i) Power and fuel temperature curve (Figure 12.7 ) shown on Page 38 of this section. 12.12.3 Loss of Recirculation Pumps 12.12.3.1 Conditions The reactor was assumed to be operating at 127% power when 7 electric power to both pumps was lost. (These conditions are much more stringent than the 100% power conditions, and are not considered credible.) 12.12.3.2 Analysis Flow was assumed to fall off linearly at a rate such that flow reached one-half its final value at a time equal to the hydraulic time constant of the system (2.6 seconds).

S2ction 12 Page 27 Rev 1 (3/23/62) 12.12.3.3 Results a See flow, power and minimum burn-out ratio curves (Figures 12.8 and 12.9 ) for the 157 Mw thermal and 1050 psia and 240 Mw thermal and 1500 psia cores shown on Pages 39 and 40 of this section. 12.12.4 Load Rejection From 100% to 7% i-The steam bypass system for this plant is designed to transfer the steam directly to the main condenser on loss of electrical load without the reactor scramming. Under normal operating conditionc the plant electrical load is 7% of rated, therefore, normal load rejection will consist of 100% to 7%. 12.12.4.1 Conditions (a) Load rejection from 157 Mwt to 11 Mwt. (b) Turbine admission valve closes to 7% in 0.7 seconds. (c) Bypass valve opens in 0.7 seconds. 12.12.4.2 Analysis (a) Anticipation signal is used to initiate opening of bypass valve. Analysis includes failure of anticipation signal. (b) Reactor does not scram. 12.12.4.3 Results Maximum Maximum Minimum Load Rejection Peak Neutron Pressure Rise, Fuel Temp Burnout Percent Flux, Percent Psi Rise,0F Ratio With ( Anticipation 100 to 7 110 20 100 2.45 ( Without Anticipation 100 to 7 112 25 200 2.40 4 12.12.5 Full Load Rejection,100% to 0% For load rejection from 100% to 0% with bypass valve opening as designed, the reactor would p,bably scram from high neutron flux levels. However, in order to. determine the effect of more stringent conditions, it is assumed that the bypass valve does not open.

Section 12 Page 28 Rev 1 (3/23/62)- 12.12.5.1 Conditions (a) Reactor initially at 157 Mwt, 1050 psia. Load rejection from 100% to 0%. i 'b) Turbine stop valve closes in 0.7 second. Bypass valve does not open. 12.12.5.2 Analysis Three cues of reactor scram considered: neutron flux at 115%, neutron flux at 125% and reactor pressure at 50 psi ~ above initial pressure regulator setting. 12.12.5.3 Results Reactor scram at - - - - - - - -115% Flux 125% Flux 50 Psi Pressure Peak neutron flux, percent 118 128 150 Maximum pressure rise, psi 75 75 160 Maximum fuel temp rise, F 100-100 110 Minimum burnout ratio 2.5 2.5 2.3 12.12.6 Coincident Steam Shut Off With Failure to Scram This is a hypothetical accident which is used only to establish the size and settings of the safety relief valves. 12.12.6.1 Conditions (a) Reactor is operating at 240 Mwt,1500 psia. (b) Reactor does not scram from the multiple trips from high condenser pressure, high neutron flux, high reactor pressure or backup isolation valve closure, as the par-ticular accident sequence may produce. (c) The emergency condenser does not operate. s 12.12.6.2 Analysis Computer analyses were made to determine the safety relief valve settings. These runs are shown on (Figure 12.10) Page 41 of this section. In these runs it was assumed .a full load rejection occurred in 0.7 second and that u.c oypass 4 valve and all scrams failed. The setting of the first safety valve was varied from 155 psi to 250 psi. The remaining safety valves were assumed to operate proportionately over a range of 40 psi.

S2ction 12 Page 29 Rev 1 (3/23/62) 12.12.7 Loss of Condenser Vacuum 12.12.7.1 Conditions f In the event that the condenser vacuum pump or ai: ejectors fail, or the air ejector off-gas line is closed, the pressure in the main condenser will rise. The rate of pressure rise within the condenser does not determine the nature of the sub-sequent transient, which depends only on the characteristics of the turbine isolation valves. Thus the results of such a transient are exactly the same as those from 100% to 0% loss of load described above if credit is not taken for the anticipation-type signals provided by the condenser pressure trips. These trips operate as follows: Annunciate at 5 inches Hg absolute Reactor scram at 8 inches Hg absolute Turbine valve closure at 10 inches Hg absolute. 12.12.8 Failure of a Safety Relief Valve to Reseat This is an academic-type analysis which may be used only to determine the adequacy of feedwater supply. i 12.12.8.1 Conditions (a) A safety relief valve is fully opened. (b) Reactor is scrammed. (c) The valve does not reseat. 12.12.8.2 Analysis Assume that system pressure has reached its maximum possible value with steam flow through the relief valve also at its maximum possible rate. J: 12.12.8.3 Results ( Reactor pressure 1870 psia Steam flow through relief valve (rating plus 10%) 103 lb/sec Minimum feedwater pump capacity 135 lb/sec l w

(.

Sactien 12 Page 30 Rev 1 (3/23/62)

F 12.13 TABULATIONS OF. OPERATOR ERROR ANALYSES r

l12.13.1 Hot Startup.

i

~ 12.13.'I.1 Conditions d' (a) Maximum worth rod withdrawn at maximum possible rate. \\ (b) Rod withdrawal interlock circuit is assumed to fail. 'I (c) Fuel and moderator temperatures initially at 546 F. 12.13.1.2 Analysis (a) Initial reactor. power at 10-rated. (b) Maximum rod worth is 4.2% bk/k (~$ 8. 00) 'I (c) Maximum reactivity insertion rate is 0. 52%6 (2/second (~ $ 0.98 per second). (d) Negative reactivity of Doppler effect only. (e)' Scram from high neutron flux at 125% of rated. (f) Scrammed rods insert at 0.2 second after signal (scram Ok with time curve (Figure 12.11) given on Page 42 of this section. 12.13.1.3 Results (a) Minimum reactor period, m sec 12 (b) Peak neutron flux, times rated 100 (c) Peak fuel temperature, F 2030 ,( l (d) Pea > reactor power, Mwt 15,700 f(- (e) Integrated energy, Mw-sec 370 (f) Power and fuel temperature curve (Figure 12.12) shown

g~

on Page 43 of this section.

-(

12.13.2 Cold Startup On the basis of current values for Doppler coefficient and flux l( distribution factors, the previous analysis given in Section l 12.7 is very conservative. The following analysis is based on current values for Doppler coeificient. ,(

E Ssction 12 Page 31 Rev 1 (3/23/62) j' 12.13.2.1 ' Conditions (a) ~ Maximum worth rod withdrawn at maximum possible rate. (b) Rod withdrawal interlock circuit is assumed to fail. (c) Fuel and moderator temperatures initially at 680 F. r 12.13.2.2 Analysis (a) Initial reactor' power at 10-10 rated. o (b) Maximum rod worth is 3.97. ok/k. (c) Maximum reactivity insertion rate is 0.46fd [/sec. (d). Negative reactivity of Doppler effect only. This effect taken as a function of fuel temperature; initial coefficient is negative 1.47 x 10-5 hkoo/k o per F at 68 F. o (e) Scram from high neutron flux at 1257. rated. 12.13.2.3 Results (a) Minimum reactor period, m sec 13 (b) Peak neutron flux, times rated 102 (c) Peak fuel temperature, F, 1950 (d) Peak reactor power, Mwt 16,000 (e) Integratdd energy, Mw-sec 250 (f) Power and fuel temperature curve (Figure 12.13) shown l on Page 44 of this section. i i k t w

Section 12 Page 32 Figure 12.1 (3/23/62) y 7 ,I p ..,.~

p

.,t..~.5. 7.n-r+-}'- 'T** P"' * n r '%~. j .1 n i .. i,,,,J.,. d. L ..i, i i ._r x' n....(..,' ..M

p. ~ g.

l..

i-b.- 3.5 n .,A..,.. -.? eq - g Ly.. 1. .oL8 } .t P. ~] 4 .'j . p' e, 4 i i .c lt

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