LD-83-019, Forwards Amended CESSAR-F Section 15.3.3, Single Reactor Coolant Pump Rotor Seizure W/Loss of Offsite Power, in Response to NRC 820115 Request & SER Comfirmatory Item 16
| ML20071E380 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 03/10/1983 |
| From: | Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| ISSUANCES-F, LD-83-019, LD-83-19, NUDOCS 8303150209 | |
| Download: ML20071E380 (29) | |
Text
s C-E P wer Sy:tems Tel. 203/688-1911 Combustion Engineenng. Inc. Telex. 99297 1000 Prospect Hill Road Windsor. Connecticut Ot5095 H POWER SYSTEMS March 10,1983 Docket No.: STN 50-470F LD-83-019 Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
CESSAR-F SER Confirmatory Item 16
Reference:
Letter, R. L. Tedesco to A. E. Scherer, dated January 15, 1982
Dear Mr. Eisenhut:
The reference requested that Combustion Engineering (C-E) perform an analysis of a locked reactor coolant pump shaft for System 80" assuming a turbine trip with coincident loss of offsite power and the worst single failure of a safety system active component. It was also stated that the low population zone dose acceptance criteria should be three hundred (300) REM, consistent with 10 CFR 100 guidelines.
The new analysis was performed and informally submitted to the staff reviewer last year. We have been subsequently informed that the informal submittal is acceptable and have been requested to formally submit the material on the docket. Enclosed please find, therefore, the amended CESSAR-F Section 15.3.3,
" Single Reactor Coolant Pump Rotor Seizure With Loss of Offsite Power". The enclosure will be incorporated into CESSAR-F via the next amendment. The revised section shows that the CESSAR-F locked rotor analysis is conservative with respect to the acceptance criteria. As a result, we believe that this submittal should resolve Confirmatory Item 16 of the CESSAR-F SER.
If I can be of any further assistance in this matter, please contact either myself or Mr. G. A. Davis of my staff at (203) 688-1911, extension 2803.
Very truly yours, COMBUSTION E INE ING, INC.,
an a -
A. E erer Director Nuclear Licensing r303150209 830310
'PDR ADOCK 05000 E
AES:las Enclosure
,r .
TABLE OF CONTENTS (Continued)
Chapter 15 Section Subject Page No. I 15.3.3.2 Sequence of Events and System Operations 15.3-7 15.3.3.3 Analysis of Effects and Consequences 15.3-11 15.3.3.4 Conclusions 15.3-15 15.3.4 REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF 0FFSITE POWER 15.3-16 15.3.4.1 Identification of Event and Causes 15.3-16 15.3.4.2 Sequence of Events and System Operations 15.3-16 15.3.4.3 Analysis of Effects and Consequences 15.3-16 15.3.4.4 Conclusions 15.3-17 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4-1 15.4.1 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER CONDITION 15.4-1 15.4.1.1 Identification of Event and Causes 15.4-1 15.4.1.2 Sequence of Events and System Operations 15.4-1 15.4.1.3 Analysis of Effects and Consequences 15.4-2 15.4.1.4 Conclusions 15.4-3 15.4.2 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL AT POWER 15.4-4 15.4.2.1 Identification of Event and Causes 15.4-4 15.4.2.2 Sequence of Events and System Operations 15.4-4 15.4.2.3 Analysis of Effects and Consequences 15.4-5 15.4.2.4 Conclusions 15.4-6 15.4.3 SINGLE FULL LENGTH CONTROL ELEMENT ASSEliBLY DROP 15.4-7 15.4.3.1 Identification of Event and Causes 15.4-7 15.4.3.2 Sequence of Events and System Operations 15.4-7
- 15.4.3.3 Analysis of Effects and Consequences 15.4-7 j
15.3.2 FLOW CONTROLLER MALFUNCTION CAUSING FLOW C0ASTDOWN This event is categorized as a Boiling Water Reactor event in SRP 15.3.2 and, therefore, will not be analyzed.
! 15.3.3 SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF 0FFSITE POWER 15.3.3.1 Identification of Event and Causes A single reactor coolant pump rotor seizure can be caused by seizure of the upper or lower thrust-journal bearings. Loss of offsite power subsequent to- l turbine / generator trip may be caused by a complete loss of the external electrical grid triggered by the turbine / generator trip. The loss of offsite power causes a loss of power to the start-up transformers which prevents the plant electrical loads from being transferred to them from the unit auxiliary transformers. Therefore, the onsite loads will lose power and the plant will experience a simultaneous loss of feedwater flow, condenser inoperability, and l a coastdown of all reactor coolant pumps. Approximately 12 seconds after the loss of offsite power occurs the diesel generators start providing power to the two plant 4.16 kV safety buses. No credit is taken for restoration of offsite power prior to initiation of shutdown cooling.
i For decreasing reactor coolant flow events, the major parameter of concern is the minimum hot channel DNBR. This parameter establishes whether a fuel design limit has been violated and, thus, whether fuel damage could be anticipated. Those fac, tors which cause a decrease in local DNBR are:
- a. increasing coolant temperature
- b. decreasing coolant pressure-
- c. increasin effects) g local heat flux-(including radial and axial power distribution
- d. decreasing coolant flow For the single reactor coolant pump rotor seizure event, the minimum DNBR occurs during the first one to four seconds of the transient, and the reactor is tripped by the CPCs on the approach to the DNBR limit. Therefore, any single failure which would result in a lower DNBR during the transient would have to affect at least one of the above parameters during the first one to four seconds of the event.
- The single failures that have been postulated are listed in Table 15.0-6.
~
Most of these failures affect the secondary system, and during the first one to four seconds they do not affect the primary system parameters which determine the DNBR. The only failures which could affect the RCS behavior
! during this interval are (1) a loss of normal AC, (2) a failure of the pressurizer level control system, (3) a failure of the pressurizer pressure -
control system, and (4) a failure of the reactor regulating system. The loss 1
15.3-6
of normal AC power, which is assumed, results in loss of power to the reactor coolant pumps, the condensate pumps, the circulating water pumps, the pressurizer pressure and level control systems, the reactor regulating system, 4 and the feedwater control system.
Loss of function of the condensate and circulating water pumps and the feedwater control system initially affect only the secondary system and, thus, do not affect DNBR in the first one to four seconds of the transient. Loss of
- . power to the reactor regulating system and pressurizer level and pressure control systems renders those systems inoperable. This inoperability will have no significant impact on DNBR during the first one to four seconds. Loss of power to the reactor coolant pumps is the only potentially significant
. failure with regard to DNBR which results from a loss of AC. However, as a result of a three second delay between the time of turbine trip and the time '
of loss of offsite power, there is no effect on minimum DNBR.
. Edilure of the pressurizer level control, pressure control, or. reactor j regulating systems cannot appreciably affect any of the four factors which l determine DNBR during the first one to four seconds of the event. Thus, none of the single failures listed in Table 15.0-6 will result in a more adverse
The' assumed loss of AC renders the steam bypass control system inoperable as a i result of the loss of the circulating water pumps. This results in the l secondary safety valves system (prior energy to operator being action)released to the and the atmosphere Atmospheric by the Dump secondary (ADV)
Valves I after operator action is assumed. Operator action is assumed at 1800 seconds t
into the transient. At this time the operator begins a controlled cooldown of the plant. '
- A single active failure of an ADV to close is assumed at 1800 seconds. The
- stuck open ADV causes the eventual dryout of the affected steam generator which results in all of the iodine contained in this steam generator being released to atmosphere. Thus, this failure in combination with the loss of offsite power maximizes the radiological consequences of the single reactor
- coolant pump rotor seizure event. None of the other single failures listed in
! -Table 15.0-6 in combination with a loss of AC will yield more severe radiological consequences.
t 15.3.3.2 Sequence of Events and System Operation l
Table 15.3.3-1 presents a chronological list and time of system actions which occur following the single reactor coolant pump rotor seizure event. Refer to Table 15.3.3-1 while reading this and the following section. The success l l paths referenced are those given on the Sequence of Events Diagram (SED),
Figure 15.3.3-1. This figure, together with Figure 15.0-1, which contains a i glossary of SED symbols and acronyms, may be used to trace the actuation and interaction of the systems used to mitigate the consequences of this event.
(>
15.3-7
The timings in Table 15.3.3-1 may be used to determine when, after event initiation, each action occurs.
The loss of offsite power is assumed to occur due to grid instability. A three second delay between the time of turbine trip and the time of loss of offsite power is conservatively assumed in the analysis, based on the discussion presented in the report submitted via Combustion Engineering Letter LD-82-040, A. E. Scherer to D. G. Eisenhut, dated March 31, 1982.
The sequence presented demonstrates that the operator can cool the plant down to cold shutdown during the event. If offsite power can be restored, then the operator may elect instead to stabilize the plant at a mode other than cold shutdown. All actions required to stabilize the plant and perform the required repairs are not described here.
The sequence of events and systems operations described below represents the way in which the plant was assumed to respond to the event initiator. Many plant responses are possible. However, certain responses are limiting with .
respect to the acceptance guidelines for this section. Of the limiting responses, the most likely one to be followed was selected.
Table 15.3.3-2 contains a matrix which describes the extent to which normally operating plant systems are assumed to function during the transient.
Table 15.3.3-3 contains a matrix which describes the extent to which safety systems are assumed to function during the transient.
The success paths in the sequence of events diagram, Figure 15.3.3-1, are as follows:
Reactivity Control:
Following seizure of a reactor coolant pump shaft, the core flow rapidly decreases to the value that would occur with only three reactor coolant pumps operating. The rapid reduction in primary coolant flow rate causes an increase in the average coolant temperature in the core, a corresponding reduction in the margin to DNB, and an increase in the primary system pressure. A low DNBR reactor trip is generated by the Core Protection Calculators. The trip conservatively assumes the largest possible delay time for the sensor delay, calculation period, CEDM dead time, and the CEDM coil -
delay time (see Chapter 7). The CEAs begin to drop into the core at 1.25 1
seconds inserting negative reactivity.
Prior to initiating or during manual cooldown the operator adjusts the boron concentration to insure that a proper negative reactivity shutdown margin is achieved. This is accomplished by using the HPSI pumps which also replace RCS volume shrinkage. The operator may also borate using the charging pumps by manually loading them on the diesel generators and then aligning them to the Refueling Water Tank (RWT), the source of borated water.
15.3-8
Reactor Heat Removal:
The reactor heat removal takes place by means of natural circulation in the reactor coolant system following the coastdown of the undamaged reactor coolant pumps. The steam generator provides primary to secondary heat transfer.
The Shutdown Cooling System (SCS) is manually actuated when the RCS temperature and pressure have been reduced to the shutdown cooling entry conditions of 350*F and 400 psia, respectively. This system provides sufficient cooling to bring the RCS to cold shutdown.
Secondary System Integrity:
The CEDM bus undervoltage relays, sensing the interruption of power on the CEDM power supply buses, generate a turbine trip signal. This results in closure of the turbine stop valves. The external grid which the plant is feeding is assumed to collapse as a result of turbine / generator trip. The l loss of offsite power causes a loss of power to the start-up transformers which prevents the plant electrical loads from being transferred to them from the unit auxiliary transformers. Therefore, the onsite loads will lose power and the plant experiences a simultaneous loss of feedwater flow, condenser inoperability, and a coastdown of all reactor coolant pumps. The pressure in 4
both steam generators increases resulting in the opening of the Main Steam Safety Valves (MSSVs), which prevents secondary pressure from exceeding safety limits. The MSSVs close when the secondary pressure drops. Water level in each of the steam generators begins decreasing immediately after the loss of main feedwater flow and an emergency fsedwater actuation signal is generated on low steam generator water level. The Emergency Feedwater Actuation System (EFAS) setpoint is first reached in the steam generator in the unaffected loop. This leads to the start-up of the emergency feedwater pumps. The primary source of the emergency feedwater is the condensate storage tank. The capacity of the storage tank is 300,000 gallons, which is sufficient feedwater to maintain the plant at hot standby for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The condensate storage tank is provided with an atmospheric vent to maintain atmospheric pressure inside the tank. 4The maximum condensate radioactivity concentration is 0.1 pCi/lbm (2.2 x 10' uCi/gm) dose equivalent I-131.
After 30 minutes the operator initiates cooldo,m of the RCS by using the atmospheric dump valves and the Auxiliary Feedwater System (AFWS). Once the dump valves are opened, one valve is assumed to remain stuck open. This results in the eventual generation of a Main Steam Isolation Signal (MSIS) on low steam generator pressure. Once the Main Steam Isolation Valves (MSIVs) are closed, this' prevents further blowdown of the unaffected steam generator.
Auxiliary feedwater is automatically terminated to the affected steam generator as a result of a high steam generator differential pressure signal.
The affected generator is then allowed to dry out. Cooldown is continued by the operator by utilizing the atmospheric dump valves of the unaffected steam generator together with the Auxiliary Feedwater System. This process 15.3-9
1 continues until shutdown cooling entry conditions are reached. The AFWS may be a separate system or may be one emergency feedwater pump designated as
" auxiliary" and intended for normal startup and shutdown of the plant. The operator may let the ESFAS regulate the feedwater flow by issuing and withdrawing EFAS-1 and/or EFAS-2 signals down to cold shutdown entry conditions. See applicant's FSAR for details of the Emergency and/or Auxiliary Feedwater Systems.
Primary System Integrity:
The pressurizer assists in the control of the RCS pressure and volume changes during the transient by compensating for the initial expansion of the RCS fluid. The combination of the loss of primary system heat sink (turbine stop valves close) with the reduction of reactor coolant flow results in an increase in RCS pressure.
During cooldown, the operator may control RCS pressure and pressurizer level by turning on the HPSI pumps and throttling the HPSI discharge valves to control the rate of change of RCS pressure. The operator may also control RCS pressure and pressurizer level via manual actuation and control of the charging pumps and related auxiliary spray. When the RCS pressure has been reduced to approximately 650 psia, the operator will vent or drain the safety injection tanks to reduce their pressure and will isolate them.
Restoration of AC Power:
A low voltage on the 4.16 kV safety buses generates an undervoltage signal which starts the diesel generators. The non-safety buses are automatically separated from .the safety buses and all loads are shed (except 4890 V load
! centers). After each diesel generator set has attained operating voltage and frequency, its output breaker closes connecting it to its safety bus. ESF equipment is then loaded in sequence on to this bus.
Radioactive Effluent Control:
Containment Isolation Actuation Signal (CIAS) isolates various systems to reduce or terminate radioactive releases. CIAS actuates primary, secondary, and containment isolation equipment. Other actions may be initiated by 80P I systems. See Applicant's FSAR for details.
Spent Fuel Heat Removal:
l Spent Fuel Pool (SFP) cooling is terminated on the loss of normal power to the
! ESF loads. Spent fuel heat removal is continuously accomplished by utilizing the heat capacity of the SFP water. Pool cooling is restored by manually loading the SFP cooling pumps onto the diesel generators and by aligning the SFP heat exchangers to receive essential cooling water.
15.3-10 i
15.3.3.3 Analysis of Effects and Consequences 15.3.3.3.1 Core and System Performance A. Mathematical Model The NSSS response to a single reactor coolant pump rotor seizure with loss of offsite power resulting from turbine trip was simulated using the CESEC-III computer program described in Reference 27 of Section 15.0. The initial DNBR was calculated using the TORC computer code (see Section 15.0.3) which uses the CE-1 CHF correlation described in Reference 19 of Section 15.0.
B. Input and Parameters and Initial Conditions The ranges of initial conditions considered are given in Section 15.0. Table '
15.3.3-4 gives the initial conditions used in this analysis. The rationale for selecting the values of the initial conditions which have a first order effect on the analysis follows. Using the highest core power maximizes the RCS heatup, which is the driving force of the secondary steam release. A high core inlet temperature was chosen to minimize the degree of subcooling in the core. This also results in higher steam generator pressures and, thus, quicker opening of the secondary safety valves, which is more adverse from a radiological standpoint. Tge reactor coolant flow was chosen to be its minimum value of 146.1 x 10 lbm/hr. This low flow was chosen in combination with the other conditions mentioned above since this will allow operation with a low radial peaking factor. .The use of a low radial peaking factor maximizes the amount of fuel pins which may experience DNB. The primary system pressure was chosen to be compatible with the other initial conditions. The most positive moderator. temperature coefficient and the minimum available scram CEA worth tend to maximize the heat flux after a reactor trip occurs, increasing the RCS heat-up. The operator initiation of plant cooldown at 30 minutes with the subsequent failure of an atmospheric dump valve to close maximizes the offsite doses.
During this event two sources of radioactivity contribute to the offsite doses: the initial activity in the steam generator and the activity
- associated with the assumed one gallon per minute steam generator tube leak.
- The initial secondary activity is assumed to be at 0.1 uCi/gm dose equivalent I-131. The initial activity assumed to be present in the reactor coolant leaking through the steam generator tubes is 4.6 uCi/gm (see Table 15.3-5).
C. Results The dynamic behavior of important NSSS parameters following a single reactor coolant pump rotor seizure with a loss of offsite power is presented on Figures 15.3.3-2 to 15.3.3-10. Table 15.3.3-1 summarizes the significant results of the event. Refer to Table 15.3.3-1 while reading this section.
l 1
15.3-11
1 The single reactor coolant pump rotor seizure event results in a flow coast l down in the affected loop, a consequent reduction in flow through the core, an l increase in the average coolant temperature in the core, a corresponding '
reduction in the margin to DNB, and an increase in the primary system pressure. A low DNBR reactor trip is generated by the core protection calculators. The reactor trip causes a turbine trip signal to occur. The CEAs begin to drop into the core at 1.25 seconds. At this time the turbine / generator trips. Three seconds later the loss of offsite power occurs. The remaining RCPs do not begin their normal coastdown until after the loss of offsite power. However, there is a slight decrease in RCP flow during the three seconds immediately after turbine trip and prior to the loss of offsite power due to decreasing pump speed caused by frequency degradation (approximately 1 Hertz /second) of the electrical grid. The loss of offsite power also causes a loss of main feedwater and condenser inoperability. The turbine trip with the SBCS and the condenser unavailable leads to a rapid buildup in secondary system pressure and temperature. This increase in pressure is shown in Figure 15.2.2-8. The opening of the MSSVs limits this pressure increase. The maximum secondary system pressure is 1347 psia which is less than 110% of design pressure.
The increasing temperature of the secondary system leads to a reduction of the primary to secondary heat transfer. Concurrently, the failed reactor coolant pump and the three reactor coolant pumps coasting down (Figure 25.3.3-7) result in RCS flow which further reduces the heat transfer capability of the RCS. This decrease in heat removal from the RCS leads to an increase in the core coolant temperatures as shown in Figure 15.3.3-5. The core coolant temperatures peak shortly after the time of reactor trip.
The increase in RCS temperature leads to an increase in RCS pressure, as shown in Figure 15.3.3-4, caused by the thermal expansion of the RCS fluid. The RCS pressure reaches a maximum value of 2387 psia at 4.2 seconds which is less than 110% of design pressure. After this time, the RCS pressure decreases rapidly due to the declining core heat flux (see Figure 25.5.5-3), in combination with the opening of the MSSVs. Opening of the MSSVs limits the peak temperature and pressure of the secondary system. The MSSVs cycle until the emergency feedwater begins entering the steam generators. Emergency feedwater begins entering the steam generator in the unaffected loop at 263 seconds, thus, enhancing the RCS cooldown and the subsequent reduction in pressure.
During the first few seconds of the transient, the combination of decreasing flow rate, and increasing RCS temperatures results in a decrease in the fuel pins' DNBR. The transient minimum DNBR of 0.967 occurs at 1.4 seconds as indicated in Table 15.3.3-1. Figure 15.3.3-9 shows the variation of the minimum DNBR with time. The negative CEA reactivity inserted after reactor trip causes a rapid power and heat flux decrease which causes DNBR to increase again. For this event no more than 0.85 percent of the fuel pins are l calculated to experience DNB. All fuel pins which experience DNB are I conservatively assumed to fail.
(
15.3-12 l
The offsite doses for this event result from steam released through the Main Steam Safety Valves (MSSVs) and Atmospheric Dump Valves (ADVs).
At 30 minutes, the operator is assumed to use the ADVs to begin cooldown. At this time, one atmospheric dump valve is assumed to stick open. This leads to the generation of an MSIS and eventual termination of emergency feedwater flow 4
to the affected steam generator. Once the affected steam generator has blown dry, the operator may continue a controlled cooldown via the atmospheric dump -
valves of the unaffected steam generator. Additionally, the operator will be feeding auxiliary feedwater flow to the unaffected steam generator. Table 15.3.3-1 shows the integrated steam release from the MSSVs and the ADVs. The radiological release produced by the transient results in a 277 rem two hour thyroid-inhalation dose at the exclusion area boundary. The two hour thyroid inhalation dose at the exclusion area boundary is shown in Table 15.3.3-7.
15.3.3.3.2 Radiological Consequences A. Physical Model To evaluate the consequences of the single reactor coolant pump rotor seizure with a loss of offsite power event, it is assumed that the condenser is not av.a11able for the entirety of the transient. For the first thirty minutes of -
the event, the cooldown is performed via the main steam safety valves.
Afterwards, an atmospheric dump valve is assumed to stick open once the
, valve (s) are actuated by the operator in an attempt to initiate a controlled cooldown. After MSIS occurs and the affected generator has blown dry, the operator may then proceed to initiate a controlled cooldown via the atmospheric dump valves of the unaffected steam generator and the auxiliary feedwater system.
B. Assumptions, Parameters, and Calculational Methods The major assumptions, parameters, and calculational methods used to evalur.te I
the radiological consequences of the single reactor coolant pump rotor seizure are presented in Tables 15.3.3-5 and 15.3.3-6. Additional clarification is ~
provided as follows:
- 1. The Reactor Coolant System (RCS) equilibrium activity is based on long tenn operation at 108% of the ultimate core power level of 3800 MWt (3800 MWt x 1.08 = 4100 MWt) with 1% failed fuel. Refer to Table 11.1.1-2 for the isotopic distribution of RCS activity.
The RCS activity was calculated to determine the total amount of activity transmitted into the secondary system during the duration of the accident due to a 1 gal / min primary to secondary leak. The primary to secondary leakage of 1 gal / min (technical specification limit) is assumed to
. continue to the steam generators for the entire event. The fluid density 3
is assumed to be constant at its initial value of 45 lbm/ft . The activity in the fuel clad gap is 10% of the iodines and 10% of the noble f
(>
- 15.3-13
l gases accumulated in the fuel at the end of core life, assuming continuous fuel power operation. All of the activity in the fuel gap for fuel rods that are calculated to experience DNB is assumed to be unifonnly mixed with the reactor coolant. This assumption is consistent with Regulatory Guide 1.77.
- 2. The steam generator equilibrium activity is assumed to be 0.1 pCi/gm dose l equivalent Iodine-131 (I-131) prior to the accident. This is the technical specification limit for steam generator activity.
- 3. Offsite power is not available. - At 1800 seconds the operator attempts to take control of the plant using the atmospheric dump valves. One atmospheric dump valve is assumed to stick open at this time for the remainder of the transient.
i
- 4. Credit is assumed for emergency feedwater flow. For the fluid leaked t from primary to secondary, iodine is assumed to released to the atmosphere with a partition coefficient of 1.0.
- 5. No credit for radioactive decay in transit to dose point is assumed.
- 6. The atmospheric dispersion factors used in this analysis, which are based on meteorological conditions assumed present during the course of the accident, are calculated according to the model described in subsection 2.3.4. The 5% level X/Q presented in Table 2.3-1 was used.
i 7. The mathematical model used to analyze the activity released during the l course of the accident is described in Section 15.0 s t 8. Table 15.3.3-6 presents the integrated mass release from the secondary safety valves and the total primary to secondary leakage.
- 9. Calculated secondary mass releases are presented in Table 15.3.3-5. l C. Identification of Uncertainties and Conservatisms in the Evaluation of the Results The uncertainties and conservatisms in the assumptions used to evaluate the radiological consequences of the single reactor coolant pump rotor seizure with a loss of offsite power are as follows:
- 1. The RCS equilibrium activity is based on 1% failed fuel, which is greater by a factor of two to eight than that normally observed in past PWR operation.
- 2. The steam generator equilibrium activity for the affected steam generator is assumed to be equal to the technical specification limit (0.1 uCi/gm dose equivalent I-131). This specific activity is greater by a factor of approximately 1300 than the normal expected steam generator activity (refer to Table 11.1.8-1).
i 15.3-14
- -a- -
, )
- 3. The primary to secondary leakage of 1 gal / min (technical specification limit) is conservative because operation with a 1 gal / min primary to secondary leak is not expected (the expected leakage rate is equal to 20 gal / day).
- 4. The meteorological conditions assumed to be present at the site during the course of the accident are based on 5% level X/Qs. Meteorological conditions will be less severe 95% of the time. This results in the poorest values of atmospheric dispersion calculated for the EAB or LPZ outer boundary. Furthermore, no credit has been taken for ths transit time required for activity to travel from the point of release to the EAB or LPZ outer boundary.
- 5. The assumption of no operator action for 1800 seconds (30 minutes) is a conservative assumption.
15.3.3.4 Conclusions The maximum RCS and steam generator pressures due to a single reactor coolant pump rotor seizure in combination with loss of offsite power following generator trip event remain less than 110% of their design values. Only a small fraction of the fuel pins experience DNB and are conservatively assumed to fail. The two hour thyroid dose is within 10CFR100 guidelines. .
l i
15.3-15
Table 15.3.3-1 l
(Sheet 1 of 2) l l
SEQUENCE OF EVENTS FOR THE SINGLE REACTOR COOLANT PUMP !
ROTOR SEIZURE WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP Total Setpoint Integrated Time or Safety Valve Success (sec.) Event Value Flow (1bm) Path 0.0 Seizure of a Single Reactor Coolant Pump 0.76 Low DNBR Trip Signal 1.19 Reactivity Generated, projected Control 1.25 CEAs Begin to Drop Reactivity Into the Core Control 1.25 Turbine Trip / Generator Trip 1.4 Minimum Transient DNBR 0.967 4.1 Main Steam Safety 1280 Secondary Valves Open, System Unaffected Loop, psia Integrity 4.2 Maximum RCS Pressure, 2387 psia 4.25 Loss of Offsite Power Occurs 4.5 Main Steam Safety Valves 1280 Secondary Open, Affected Loop, System psia Integrity l 6.8 Maximum Steam Generator 1347 3,492 l
Pressure, Unaffected t Loop, psia 7.4 Maximum Steam Generator 1340 5,451 Pressure, Affected Loop, psia
Table 15.3.3-1 (Continued)
(Sheet 2 of 2)
SEQUENCE OF EVENTS FOR THE SINGLE REACTOR COOLANT PUMP ROTOR SEIZdRE WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP Total Setpoint Integrated Time or Safety Valve Success (Sec.) Event Value Flow (lbm) Path 218 Low Water Level EFAS 20 85,679 Secondary Setpoint Reached in System the Steam Generator, '
Integrity Unaffected Loop, percent of wide range 263 Emergency Feedwater 119 91,407 Secondary Begins Entering Steam System Generator, Unaffected Integrity Loop,lbm/sec . 697 Low Water Level EFAS 20 115,189 Secondary Setpoint Reached in System the Steam Generator, Integrity Affected Loop, percent of wide range Emergency Feedwater 119 Begins Entering the Steam Generator, Affected Loop, lbm/sec 821 Steam Generator 1218 120,398 Secondary Safety Valves Close, System Affected and Unaf- Integrity fected Loop, psia 1800 Atmospheric Dump -100.0 120,398 Secondary Valves Opened to System Initiate Plant Integrity Cooldown, F/ hour.
One Atmospheric Dump Valve Sticks Open
! 7200 Total Steam Release 1,128,293 to Atmosphere, lbm 1
i
TABLE 15.3.3-2 (shest 1 of 2)
DISPOSITION OF NORMALLY OPERATING SYSTEMS FOR THE SINGLE REACTOR COGLANT PUMP ROTOR SEIZURE WITH LOSS OF 0FFSITE POWER RESULTING FROM TURRINE TRIP
+
o%,.w
,Q
- A A o
. _ SYSTEM C Q,
- 1. Main Feedwater Control System /
- 2. Main Feedwater Pump Turbine Control System * /
- 3. Turbine-Generator Control System * /
- 5. Pressurizer Pressure Control System /
- 6. Pressurizer Level Control System /
- 7. ' Control Element Drive Mechanism Control System /
- 8. Reactor Regulating System /
- 9. Core Operating Limit Supervisory System /
- 10. Reactor Coolant Pumps / 1
- 11. Chemical and Volume Control system /
- 12. Secondary Chemistry Control System * /
- 13. Condenser Evacuation System * /
- 14. Turbine Gland Sealing System * /
- 15. Nuclear Cooling Water System * /
, 16. Turbine Cooling Water System * . /
l 17. Plant Cooling Water System * /
- 18. Condensate Storage Facilities * / j f I
- 19. Circulating Water System * /
- 20. Spent Fuel Pool Cooling and Clean-Up System * /
21 . Non-Class lE (Non-ESF) A.C. Power * /
- 22. Class lE (ESF) A.C. Power * /
- 23. Non-Class lE D.C. Power * /
- 24. Class lE D.C. Power * /
- Balance-of-Plant Systems
. TABLE 15.3.3-2 (Cont'd.) (Sheet 2of2)
DISPOSITION OF fl0RMALLY OPERATING SYSTEMS
. FOR THE SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF 0FFSITE POWER RESULTING FROM TURBINE TRIP Oc c ^
Q
- .pn h,+o +,
- Q SYSTEM i
- 23. Non-Class lE D.C. Power * /
- 24. Class lE D.C. Power * /
l l
Note:1 A Failure in this System is the event initiator oBalance-of-P1 ant Systems l
TABLE 15.3.3-3 UTILIZATION OF SAFETY SYSTEMS FOR THE SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF 0FFSITE POWER RESULTING FROM TURBINE TRIP A 0 c Q
^& eg QQ +%
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- 2. DNBR/LPD Calculator /
- 3. Engineered Safety Features Actuation Systems / /
- 4. Supplementary Protection System 1
- 5. Reactor Trip Switch Gear /
- 6. Main Steam Safety Valves * /
- 7. Primary Safety Valves
- 8. Main Steam Isolation System * / /
- 9. Emergency Feedwater System * / /
- 10. Safety Injection System /
- 11. Shutdown Cooling System / 2
- 12. Atmospheric Dump Valve System * / / 2,3
- 13. Containment Isolation System * / ,
- 14. Containment Spray System
- l 15. Iodine Removal System *
- 16. Containment Combustible Gas Control System
- I 17. Diesel Generators and Support Systems * / .
- 18. Component (Essential) Cooling Water System * /
- 19. Station Service Water System * /
i l
, Notes:
- 1. Safety grade back-up to a safety grade system.
- 2. Manually actuated dil. ing normal cooldown
- 3. One atmospheric dump valve is assumed to stick open.
- Balance-of-Plant Systems - l
Table 15.3.3-4 ASSUMED INITIAL CONDITIONS FOR THE SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFF5ITE POWER RESULTING FROM TURBINE TRIP Parameter Value Core Power Level, MWt 3876 Core Inlet Coolant Temperature, *F 580 Reactor Coolant System Pressure, psia 2257 Steam Generator Pressure, psia 1221 Core Mass Flow,106 lbm/hr 146.1
. Maximum Radial Power Peaking Factor 1.40 Maximum Axial Power Peak 1.47 Core Minimum DNBR 1.48 Doppler Coefficient Multiplier 0.85 CEA Worth on Trip, 10-2 3, .
-10.0 (most reactive CEA stuck) .
Moderator Temperature Coefficient 0.0 i
t
TABLE 15.3.3-5 (Sheet 1 of 3)
PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSE0VENCES OF A SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOS5 OF OFFSITE POWER RESULTING FROM TURBINE TRIP Parameters Value A. Data and Assumptions Used to Evaluate the Radioactive Source Term
- a. Power Level, Hwt 4200
- b. Burnup 2 year
- c. Percent of Fuel Calculated to 0.85 Experience DNB, %
- d. Reactor Coolant Activity 4.6 pCi/gm Before Event (based on 4100 MWt) Table 11.1.1-2
- e. Secondary System Activity Section 15.0.4 Before Event
- f. Primary System Liquid 525,600 Inventory, lbm .
- g. Steam Generator Inventory
- Liquid, lbm per steam generator 167,075
- Steam, lbm per steam generator 14,863 B. Data and Assumptions Used to Estimate Activity Released from the Secondary System
- a. Primary to Secondary Leak Rate, 1.0 (total) gpm
- b. Total Mass Release Through the 1,128,293 Main Steam Safety Valves and Atmospheric Dump Valves (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)
. l TABLE 14.3.3-5 (Continued)
(Sheet 2of3)
PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFF5ITE POWER RESULTING FROM TURBINE TRIP Parameters Value C. Reactor Coolant System Activity After Event, Ci Isotope I-131 1.039 +
I-132 1.519 +
I-133 2.097 +
I-134 2.265 +
I-135 1.954 +
Kr-85M 2.621(+4)
Kr-85 8.315 +
Kr-87 4.806 +
Kr-88 6.867 +
Xe-131M 7.320(+2)
Xe-133 2.105(+5 Xe-135 3.767 +
Xe-138 1.679 +
- d. Percent of Core Fission Products Refer to Section Assumed Release to Reactor Coolant 15.3.3.3.2B
- e. Iodine Partition Coefficient for the 100.0 Unaffected Steam Generator
- f. Iodine Partition Coefficient for the 1.0 Affected Steam Generator
- g. Credit for Radioactive Decay in No Transit to Dose Point
- h. Loss of Offsite Power Yes D. Dispersion Data
- 1. Distance to Exclusion Area Boundary, m 500
TABLE 15.3.3-6 SECONDARY SYSTEM MASS RELEASE TO THE ATMOSPHERE FOR THE SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOS5 OF OFF5ITE POWER RESULTING FROM TURBINE TRIP EVENT Time Integrated Safet Integrated Primary to (sec.) Valve Flow (lbm)y Secondary Leakage (gallons) 0.0 0.0 0.00 2.0 0.0 0.03 3.0 0.0 0.05 5.0 480 0.08 10.0 13,884 0.17 20.0 36,677 0.33 40.0 53,354 0.67 60.0 54,812 1.00 80.0 62,536 1.33 100.0 66,282 1.67 150.0 73,248 2.50 200.0 83,047 3.33 300.0 95,560 5.00 500.0 108,710 8.33 821.0* 120.398 13.68 1800.0** 120,398 30.00
~
- Main steam safety valves close. *
- Operator begins cooldown utilizing the atmospheric dump valves. One atmospheric dump valve is assumed to stick open.
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TABLE 15.3.3-7 RADIOLOGICAL CONSEOUENCES OF A POSTULATED SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITF LOSS OF OFF5ITE POWER RESULTING FROM TURBINE TRIP From Secondary System Result Steam Releases Exclusion Area Boundary Dose (0-2 hours), rem:
Thyroid 277.0 l
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C-E SINGLE RCP ROTOR SEIZURE WITH LOSS OF 0FFSITE Figure POWER RESULTING FROM TURBINE TRIP 15.3.3 -
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COOLANT PUMP ROTOR SElZURE WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP 15.3.3-1 A l '
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5 M C.E SEQUENCE OF EVENTS DIAGRAr1 FOR SINGLE REACTOR Roure COOLANT PUVP ROTOR SElZURE WITH LOSS OF OFFSITE POWER RESULTING FROr.1 TURBirdE TRIP 15.3.11C 4-e -
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j C-E SEQUENCE OF EVENTS DIAGRAM FOR SINGLE REACTOR figwe COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP IL33-11
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15.3.4 REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF 0FFSITE POWER 15.3.4.1 Identification of Event and Causes A single reactor coolant pump sheared shaft could be caused by mechanical failure of the pump shaft. This is assumed to result from a manufacturing defect in the shaft. Loss of offsite power following turbine / generator trip may be caused by a complete loss of the external electrical grid triggered by the turbine / generator trip.
15.3.4.2 Sequence of Events and Systems Operation The sequence of events and systems operations is similar to that for the reactor coolant pump rotor seizure event, Section 15.3.3. The difference is that for the shaft break event, the reactor is tripped on differential pressure across either steam generator, whereas, for the pump rotor seizure event, the reactor is tripped by the CPC on a low projected DNBR condition.
The flow coastdown for a Rotor Seizure (RS) event is faster than the coastdown for a Shaft Break (SB) event. For a shaft break, the rotor is still capable of rotating, thereby offering less resistance to flow during the rapid flow decrease. This results in a less severe coast down for the shaft break event than for the rotor seizure event. The SB trip time is 1.2 seconds; the RS trip time is 0.91 seconds. Despite the later trip time, the slower SB coastdown results in a higher minimum DNBR and less fuel failure for SB than for LR.
For both RS and SB, three seconds after turbine trip a Loss of Offsite Power (LOP) was assumed. Both RS and SB reach the same 3-pump asymptotic flow before their respective LOPS and do not result in decreasing DNBR after LOP.
The RS plus LOP minimum DNBR is lower and fuel failure higher than those for SB plus LOP.
15.3.4.3 Analysis of Effects and Consequences 15.3.4.3.1 Core and System Performance The analysis of effects and consequences for this event is similar to that for the reactor coolant pump rotor seizure event, Section 15.3.3. The SB coastdown is slower and trip is later than those of the rotor seizure event.
The SB plus LOP event produces a higher minimum DNBR and less radiological release than RS plus LOP.
15.3.4.3.2 Radiological Consequences The radiological consequences due to steam release from the secondary system would be less than the consequences of the RS event as described in Section 15.3.3. Thus, the two hour thyroid inhalation dose for the SB with loss of offsite power event is less than 277.0 rem.
15.3-16
l 15.3.4.4 Conclusion The conclusion from the SB event is that this event would be no more adverse than the RS event. For both events, the total number of fuel pins calculated in DNB, and which are conservatively assumed to fail, is less than 0.85%. The l resultant radiological consequences, which are given in Table 15.3.3-7, are within the guidelines of 10CFR100.
15.3-17
/