NG-17-0111, Duane Arnold Energy Center, Revision 24 to Updated Final Safety Analysis Report, Chapter 15, Accident Analyses
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Initial Power
Feedwater Temperature at Vessel Inlet
°°Decay Heat Model
Initial Dome Pressure
Total drywell free volumeInitial drywell pressure Initial drywell temperature °Initial drywell relative humidity Number of downcomers Submergence of downcomers
Initial suppression pool volume Initial suppression pool temperature Initial wetwell free airspace volume
Initial wetwell airspace pressure Initial wetwell airspace temperature °Initial wetwell airspace relative humidity
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°
Heat exchanger K-value (per HX)
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°
°
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°Heat exchanger initiation time Service water temperature Drywell spray initiation time
Wetwell spray initiation time Drywell spray flow rate (1 RHR pump) Wetwell spray flow rate (1 RHR pump)
Number of pumps
Number of pumps
Feedwater system liquid and metal masses Amount of hot inventory available for injection Corresponding enthalpy vs mass
Acceptable effective bypass leakage area (A/K) Suppression chamber pressure at which operator will be alerted to the existence of a bypass leakage path Operator action time Available mass for vessel makeup Water temperature °
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This Section of the UFSAR contains the event descriptions, methods of analysis, assumptions, and analytical results of that subset of plant events classified as Transients (sometimes referred to as Abnormal Operating Transients (AOTs) or Abnormal
Operational Occurrences (AOOs)) (See Section 15.0.2). The plant response to each Transient will be discussed in terms of the impact on the fission product barriers
specifically, the fuel cladding and reactor coolant pressure boundary. Because these events generally do not lead to fuel failures or direct challenges to either the primary or secondary containments, those fission product barriers are not evaluated for Transients. Also, the methods used, and assumptions made, in the individual event analyses are specifically adjusted to provide conservative results for the specific event, it is recognized that between each Transient, there may not be full coherence between the various evaluations performed. For example, different initial conditions/values may be
used in the evaluation of the fuel and that used in the analysis of the reactor coolant pressure boundary for the Main Steamline Isolation Valve Closure Transient. Or, different computer codes may be better suited to one type of event over another. For example, a slow moving transient may be better evaluated using a series of calculations with a steady state model than using a dynamic model that is better suited to fast moving events. Thus, each event description will describe the methods, inputs and assumptions
used and will highlight any uniqueness in that evaluation. 15.1.1 TRANSIENTS RESULTING IN A REACTOR VESSEL WATER TEMPERATURE DECREASE
Events that result directly in a reactor vessel water temperature decrease are those that either increase the flow of cold water to the vessel or reduce the temperature of water being delivered to the vessel. The three events that result in the most severe transients in
this category are the following:
- 1. Feedwater controller failure - maximum demand.
- 2. Loss of feedwater heating.
- 3. Inadvertent HPCI actuation.
15.1.1.1 Feedwater Controller Failure - Maximum Demand
Description of Event
a) Initiator: A postulated failure in the feedwater control logic creates a demand in the feedwater flow to the maximum runout value of both feed pumps.
UFSAR/DAEC - 1 15.1-2 Revision 23 - 5/15 b) Sequence of Events: The plant is operating at 100% power and 105% core flow, when there is a maximum demand signal generated by the feedwater level control system causing
the feedwater regulating valves to go to full open and feedwater flow increases to 115% of rated flow (pump runout condition). Because there is an initial mismatch between steamflow and feedwater flow, the inlet subcooling to the core increases
because the feedwater flow is not sufficiently heated by the feedwater heaters.
This causes reactor power to increase due to the collapse of the voids in the core. Also, the mismatch causes the reactor water level to increase to the High Level trip setpoint (Level 8), which trips both feedwater pumps and the main turbine. (Note: at this point, the transient response becomes essentially a turbine trip with bypass from slightly higher than rated power.) The turbine trip signal to the EHC system causes the turbine stop valves begin to close. Upon reaching the 90% open (nominal) point, as sensed by valve position switches, a reactor trip signal (Scram) is initiated, along with an end-of-cycle recirculation pump trip (EOC-RPT). Control rods begin to insert and the recirculation pumps begin to coast
down, both of which help turn around the power increase generated by the collapsing of the voids in the core from the pressure increase due to the loss of a steam path with the closing of the stop valves. Because the turbine bypass valves do not have the capacity to accommodate the initial steam flow, reactor pressure increases and SRVs lift to relieve the pressure. This will arm the Low-Low Set
logic and control the SRVs opening and clos ing setpoints. Control rods are fully inserted to terminate the power increase. Long-term response (beyond the explicit analyzed period): Turbine bypass valves will control the pressure. HPCI and/or RCIC initiate to maintain reactor vessel level, as needed. Operators take manual control and guide the plant to a cold
shutdown condition.
c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: Feedwater and Main Turbine trip (Level 8 trip), RPS trip (TSV closure), EOC-RPT (TSV closure), Control Rod Scram, Recirculation Pump Trip, SRVs open/close. Long-term: Low-low set logic armed (SRV open and High Pressure Scram), Turbine bypass valves open to control reactor pressure, HPCI/RCIC initiations (low-low RPV level). The Operators intervene only after the initial transient is over and guide the plant
to a stable condition. UFSAR/DAEC - 1 15.1-3 Revision 23 - 5/15 Event Category & Acceptance Criteria This is an Anticipated Operational Occurrence - a single transient of moderate-to-infrequent frequency (feedwater controller failure to maximum runout flow conditions) with no other equipment failures or Operator errors. Fuel SAFDLs (Section 15.0.4) shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance). b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). No unique inputs for this analysis. c) Key Assumptions: Plant is initially at rated thermal power and 105% core flow. TSVs close in a linear ramp over 0.1 seconds. No Operator Actions are assumed during the initial transient response. Results a) Barrier Performance and Comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. See the current cycle's Supplemental Reload Licensing Report (SRLR) for actual values. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. See the current cycle's SRLR for
actual values. Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities:
This event has been analyzed assuming no high level trip of the Main Turbine and Feedwater pumps (Ref. 15.0-39). This analysis demonstrates that the results of
this event are not sensitive to the Level 8 setpoint, provided that the initial power UFSAR/DAEC - 1 15.1-4 Revision 23 - 5/15 level increase due to the increase in subcooling has stabilized prior to the turbine trip.
This event is very sensitive to the vessel pressure change. Changes in plant equipment characteristics, e.g., valve stroke times, setpoint changes, steamline lengths and volumes, etc., that will cause the pressurization rate and/or peak vessel pressure to increase will have a negative impact on the event results. As discussed in Section 15.0.9, the results of this transient are sensitive to the initial power level, as the impact of the total runout flow is more pronounced at lower powers (i.e., incremental increase in feedwater flowrate). The fuel thermal limits are adjusted to account for this by the use of the MCPRp and MAPFACp/LHGRFACp multipliers from the ARTS (APRM/RBM/Technical Specification) program. To support certain equipment being out of service during the operating cycle, (Ref. FRED form in Section 15.7) additional analysis of this event is performed assuming that equipment is not Operable. The results of these equipment out-of-
service conditions are found in the SRLR for the current cycle. c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion
a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an Abnormal Operating Transient.
b) Known Conservatisms/Margins: Allowance for 2% core thermal over-power (accounted for in the GEMINI methods). End-of-cycle core conditions are assumed. Conservative control rod scram times are used. Bounding TSV closure time. Bounding SRV opening setpoints (+3% tolerance to nominal settings)
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a Limiting Event and is re-analyzed as part of the reload
analysis for each operating cycle.
15.1.1.2 Loss of Feedwater Heating UFSAR/DAEC - 1 15.1-5 Revision 23 - 5/15 Description of Event a) Initiator: A number of various failure modes can lead to a loss of feedwater heating. For the purposes of this evaluation, we do not specify the exact failure mode, but only
that there is a loss of heating that results in a 100 °F reduction in feedwater temperature. b) Sequence of Events (NOT a time line): There is a loss of feedwater heating that results in a slow decrease in feedwater temperature. This increases the inlet subcooling to the core, which in turn, cause the power level to increase due to the collapse of the voids from the colder water.
The power level increases until a new steady state condition is achieved when the increase in steam flow to the turbine equilibrates to the new power level.
c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): It is assumed that the Operator notices the indication of increasing reactor power, steam flow, etc. and takes control of the event to lower the power back to within the licensed loadline/thermal power level by lowering recirculation flow and/or inserting control rods. Event Category & Acceptance Criteria
This is an Anticipated Operational Occurrence - a single transient of moderate frequency (loss of feedwater heating) with no other equipment failures or
Operator errors.
Fuel SAFDLs (Section 15.0.4) shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits.
Methods a) Calculation Tools & Computer Codes: The primary code used to perform this analysis is the GE 3-D Core Simulator (PANACEA). (See Table 15.0-2 for complete listing, code versions and NRC
acceptance). b) Inputs (Reference common list in 15.0, and/or include event-specific items): UFSAR/DAEC - 1 15.1-6 Revision 23 - 5/15 This event is analyzed at 100% power/100% flow (Note: no 2% allowance for overpower is used in this analysis). Cycle-specific core loading (FRED form). No other unique inputs are used. c) Key Assumptions: There is no scram signal generated by this event. Results a) Comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. See the current cycle's Supplemental Reload Licensing Report (SRLR) for actual values.
Reactor Pressure Boundary Performance: Reactor vessel pressure remains
essentially unchanged in this event due to it being a very slow transient.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities: The results of this event are mildly sensitive to the magnitude of the feedwater temperature change and the results are most limiting at rated
conditions. This event is analyzed at BOC, MOC and EOR conditions (see 15.7). c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an
Anticipated Operational Occurrence.
b) Known Conservatisms/Margins: The assumed 100 °F change in feedwater temperature is bounding for any single failure within the Feedwater system. UFSAR/DAEC - 1 15.1-7 Revision 23 - 5/15 c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is a Limiting Event and is evaluated as part of the cycle-specific reload
analysis. 15.1.1.3 Inadvertent HPCI Actuation
Description of Event a) Initiator:
For the purposes of this evaluation, we do not specify the exact cause of the HPCI actuation; only that there is an inadvertent injection to the vessel from the HPCI system. b) Sequence of Events (NOT a time line):
There is an inadvertent initiation of the HPCI system that injects colder water to the reactor vessel, via the Feedwater System. The Feedwater level control instrumentation compensates for the increase in level due to the additional inventory from the HPCI injection by reducing the feedwater flow. However, there is a decrease in feedwater temperature. Similar to the Loss-of-Feedwater Heating event, this temperature decrease increases the inlet subcooling to the
core, which in turn, causes the power level to increase due to the collapse of the voids from the colder water. The power level increases until a new steady state condition is achieved when the increase in steam flow to the turbine equilibrates
to the new power level. c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): The Feedwater level control system reacts to the increasing vessel level due to the HPCI injection and reduces feedwater flow to compensate. It is assumed that the Operator notices the indication of the HPCI initiation and takes
action to secure the injection. They also react to the increasing reactor power, steam flow, etc. and take control of the event to lower the power back to within the licensed loadline/thermal power level by lowering recirculation fl ow and/or inserting control rods. UFSAR/DAEC - 1 15.1-8 Revision 23 - 5/15 Event Category & Acceptance Criteria
This is an Anticipated Operational Occurrence - a single transient of moderate frequency (inadvertent HPCI injection) with no other equipment failures or
Operator errors. Fuel SAFDLs (Section 15.0.4) shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits.
Methods a) Calculation Tools & Computer Codes: Primary Code - REDY, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items) See OPL-3 (Table 15.0-3) HPCI injects at 3000 gpm. c) Key Assumptions: Plant is initially at 102% of rated thermal power and rated core flow. There is no scram signal generated by this event. Results a) Comparison to Acceptance Criteria:
Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. See the current cycle's Supplemental Reload Licensing Report (SRLR) for confirmation.
Reactor Pressure Boundary Performance: Reactor vessel pressure remains
essentially unchanged in this event due to it being a very slow transient. Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities: The results of this event are mildly sensitive to the magnitude of the HPCI injection flowrate (i.e., feedwater temperature change) and the results are most limiting at rated conditions. This event is analyzed at BOC, MOC and EOR
conditions (see 15.7).
UFSAR/DAEC - 1 15.1-9 Revision 23 - 5/15 c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an
Anticipated Operational Occurrence.
b) Known Conservatisms/Margins: Allowance for 2% core thermal over-power.
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is a non-limiting event and is only evaluated as part of the cycle-specific reload analysis to confirm that it remains bounded by the Loss-of-Feedwater
Heating event.
UFSAR/DAEC - 1 15.1-10 Revision 23 - 5/15 15.1.2 TRANSIENTS RESULTING IN A NUCLEAR SYSTEM PRESSURE INCREASE Events that result directly in significant nuclear system pressure increases are those that result in a sudden reduction of steam flow while the reactor is operating at power. Plant systems have been surveyed to identify event within each system that could result in the rapid reduction of steam flow. The survey revealed the following events:
- 1. Generator load rejection (turbine control valve fast closure).
- 2. Turbine trip (turbine stop valve closure).
- 3. Closure of the main steam line isolation valves.
- 4. Failure of the turbine bypass valves to open when required.
- 5. Loss of main condenser vacuum.
- 6. Pressure regulator malfunction causing turbine control valves to close.
- 7. Loss of Offsite Power A consideration of Events 4-6 above shows that turbine bypass valve failure, loss of condenser vacuum, and pressure regulator malfunction are specific cases of the first
two event types. A failure of the turbine bypass valves to open when required is analyzed as the most severe form of a turbine or generator trip. A loss of condenser vacuum causes turbine stop valve closure and turbine bypass valve closure; thus, a loss of vacuum is a turbine trip without bypass. Pressure regulator malfunctions that result in turbine steam flow shutoff and a nuclear system pressure rise are mild forms of a
generator load rejection. Thus, all of the effects of these events are included in the effects described for the generator load rejection and turbine trip.
The loss-of-offsite power (LOOP) is a complex sequence of events that occurs when the plant loses all auxiliary power. A loss of auxiliary power is an event that deenergizes all buses that supply power to the unit auxiliary equipment, such as recirculation pumps, condensate pumps, and circulating water pumps. Such an event can result if faults or trips occur in the auxiliary power distribution system.
The reactor is subjected to a complex sequence of events when the plant loses all auxiliary power. Estimates of the responses of the various reactor systems (assuming a loss of the auxiliary transformer) provide the following sequence:
- 1. The recirculation pumps are tripped with normal coastdown times.
- 2. Independent main steam isolation valve closure and scram are initiated UFSAR/DAEC - 1 15.1-11 Revision 23 - 5/15 because of the loss of power to their respective solenoids, i.e., RPS M/G set trip.
- 3. Motor driven feedwater pumps are tripped.
An alternative transient results if there is a loss of all electrical connections to the grids external to the plant. The same sequence as above would be followed except that the reactor would also experience a generator load rejection and its associated scram at
the beginning of the transient. Consequently, this transient can be characterized as either
a closure of all MSIVs or generator load rejection event.
15.1.2.1 Generator Load Rejection (Turbine Control Valve Fast Closure)
A loss of generator load causes the turbine-generator to increase in speed. The turbine speed and acceleration protection systems and the power load unbalance circuitry
in the electrohydraulic controller quickly close the turbine control valves to shut off the steam supply to the turbine, thus avoiding excessive turbine overspeed. Several variations in the Load Rejection transients are possible according to the assumptions made concerning the initial power level and the turbine bypass system. These cases are
discussed individually.
15.1.2.1.1 Main Generator Load Rejection with Bypass Vales (LRWBP) - High Power Description of Event
a) Initiator: A non-mechanistically caused trip of the main generator. b) Sequence of Events: The plant is operating at 100% power, when the EHC system receives a trip
signal and the turbine control valves begin to close. Upon actuation of the Turbine
Control Valve (TCV) fast closure, as sensed by EHC oil pressure, a reactor trip signal (Scram) is initiated, along with an end-of-cycle recirculation pump trip (EOC-RPT). Control rods begin to insert and the recirculation pumps begin to
coast down. Both of which help turn around the power increase generated by the collapsing of the voids in the core from the pressure increase due to the partial loss of a steam path. Although the bypass valves open, their steam flow capacity
is not enough initially to control the pressure. So, reactor pressure increases and SRVs lift to relieve the pressure. Control rods are fully inserted to terminate the
power increase. Long-term response (beyond the explicit analyzed period): Low-low set logic armed (SRV open and High Pressure Scram), Turbine bypass valves open to control reactor pressure, Feedwater and Condensate pumps are used to maintain reactor level. Operators take manual control and guide the plant to a cold shutdown condition using normal shutdown procedures.
UFSAR/DAEC - 1 15.1-12 Revision 23 - 5/15 e) Single Failure/Operator Error (as applicable): None f) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: RPS trip (TCV fast closure), EOC-RPT (TCV fast closure), Control Rod Scram, Recirculation. Pump Trip, SRVs open/close.
Long-term: Low-low set logic armed (SRV open and High Pressure Scram), Turbine bypass valves open to control reactor pressure, Feedwater/Condensate pumps operate. The Operators
intervene only after the initial transient is over and guide the plant to a stable condition. Event Category & Acceptance Criteria This is an Anticipated Operational Occurrence - an expected operational transient of moderate frequency (load rejection) with no other equipment failures or
Operator errors.
Fuel SAFDLs (Section 15.0.4) shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). No unique inputs for this analysis. c) Key Assumptions: Plant is initially at rated thermal power and core flow. TCVs operate in two admission mode (3x1 mode - TCVs 1, 2 & 3 move together
and TCV 4 is the controlling valve.) TCV RPS trip is based upon an assumed response time of 0.03 secs from
beginning of TCV fast closure to the trip of the RPS relay. No Operator Actions are assumed during the initial transient response. UFSAR/DAEC - 1 15.1-13 Revision 23 - 5/15 Results d) Barrier Performance and comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. This is a non-limiting event and is not re-analyzed as part of the reload process. See Table 15.0-1 for comparison of event response to the bounding events. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. See Table 15.0-1 for comparison of
event response to the bounding events.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. e) Known Sensitivities: This event is moderately sensitive to the vessel pressure change. Changes in plant equipment characteristics, e.g., valve stroke times, setpoint changes, steamline lengths and volumes, etc., that will cause the pressurization rate and/or peak vessel pressure to increase will have a negative impact on the event results. c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion
d) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an Abnormal Operating Transient. e) Known Conservatisms/Margins: Allowance for 2% core thermal over-power is accounted for in the analysis methods. End-of-cycle core conditions are assumed. Conservative control rod scram times are used. Bounding TCV (faster) closure time.
Bounding SRV opening setpoints (+3% tolerance) f) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a non-Limiting Event and is not re-analyzed as part of the reload analysis for each operating cycle. See Table 15.0-1 for comparison of event
response to the bounding events.
UFSAR/DAEC - 1 15.1-14 Revision 23 - 5/15 15.1.2.1.2 Main Generator Load Rejection with Bypass Valve Failure (LRNBP) - High Power Description of Event
a) Initiator: A non-mechanistically caused trip of the main generator with failure of the bypass
valves to open.
b) Sequence of Events: The plant is operating at 100% power, when the EHC system receives a trip
signal and the turbine control valves begin to close. Upon actuation of Turbine
Control Valve (TCV) fast closure, as sensed by EHC oil pressure, a reactor trip signal (Scram) is initiated, along with an end-of-cycle recirculation pump trip (EOC- RPT). Control rods begin to insert and the recirculation pumps begin to
coast down. Both of which help turn around the power increase generated by the collapsing of the voids in the core from the pressure increase due to the loss of a steam path, due to the failure of the turbine bypass valves to open. Reactor
pressure increases and SRVs lift to relieve the pressure. Control rods are fully inserted to terminate the power increase.
Long-term response (beyond the explicit analyzed period): Low-low set logic armed (SRV open and High Pressure Scram) and cycles the SRVs to control reactor pressure, Feedwater and Condensate pumps eventually are no longer
available, as condenser inventory is depleted, HPCI and/or RCIC initiate to maintain reactor vessel level, as needed. Operators take manual control and guide
the plant to a cold shutdown condition.
c) Single Failure/Operator Error (as applicable): Failure of Turbine Bypass Valves to open upon demand.
d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: RPS trip (TCV fast closure), EOC-RPT (TCV fast closure), Control Rod Scram, Recirculation Pump Trip, SRVs open/close.
Long-term: Low-low set logic armed and controls reactor pressure, FW/Condensate pumps operate, HPCI/RCIC initiations (low-low RPV level). The Operators intervene only after the initial transient is over and guide the plant to a stable condition. UFSAR/DAEC - 1 15.1-15 Revision 23 - 5/15 Event Category & Acceptance Criteria This is an Abnormal Operating Transient - an expected operational transient of moderate frequency (load rejection), but with the additional single failure of the
bypass valves failing to open. Fuel SAFDLs (Section 15.0.4) shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). No unique inputs for this analysis.
c) Key Assumptions: Plant is initially at rated thermal power and 105% core flow. TCVs operate in two admission mode (3x1 mode - TCVs 1, 2 & 3 move together
and TCV 4 is the controlling valve.) TCV RPS trip is based upon an assumed response time of 0.03 secs from
beginning of TCV fast closure to the trip of the RPS relay. No Operator Actions are assumed during the initial transient response.
Results a) Barrier Performance and comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. See the current cycle's Supplemental Reload Licensing Report (SRLR) for actual values.
Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. See the current cycle's SRLR for
actual values.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed.
UFSAR/DAEC - 1 15.1-16 Revision 23 - 5/15 b) Known Sensitivities: This event is very sensitive to the vessel pressure change. Changes in plant equipment characteristics, e.g., valve stroke times, setpoint changes, steamline lengths and volumes, etc., that will cause the pressurization rate and/or peak vessel pressure to increase will have a negative impact on the event results. To support certain equipment being out of service during the operating cycle, (Ref. FRED form in Section 15.7) additional analysis of this event is performed assuming that equipment is not Operable. The results of these equipment out-of-
service conditions are found in the SRLR for the current cycle. c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion
a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an Abnormal Operating Transient. b) Known Conservatisms/Margins: Allowance for 2% core thermal over-power is accounted for in the analysis methods. End-of-cycle core conditions are assumed. Conservative control rod scram times are used. Bounding (faster) TCV closure time.
Bounding SRV opening setpoints (+3% tolerance)
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a Limiting Event and is re-analyzed as part of the reload
analysis for each operating cycle.
15.1.2.1.3 Generator Load Rejection from Low Power without Bypass
a) Initiator: A non-mechanistically caused trip of the main generator with failure of the bypass
valves to open.
b) Sequence of Events: Note: Under the ARTS program (APRM/RBM/Technical Specification), this event is analyzed as a series of four cases representing the combination of thermal UFSAR/DAEC - 1 15.1-17 Revision 23 - 5/15 power and core flow: 21.7% power (representing the lowest power level for monitoring of fuel thermal limits), 26% power (representing the bypass of the direct scram signal on the TCVs), 50% core flow and 105% core flow. The results
of these four cases help to define the MCPRp, LHGRFACp and MAPFACp limits in the Core Operating Limits Report (COLR). The limiting case of 26%
power/105% flow is discussed below. The plant is operating at 26% power at 105% rated core flow, when the EHC system receives a trip signal and the turbine control valves begin to close. Because this power level is below the bypass for the direct scram signal on the TCVs, no direct scram is generated. Reactor pressure increases due to the loss of a steam path, as a result of the failure of the turbine bypass valves to open. The pressure quickly reaches the scram setpoint and control rods begin to insert, which help turn around the power increase generated by the collapsing of the voids in the core from the pressure increase. Reactor pressure increases and SRVs lift to relieve the pressure. Control rods are fully inserted to terminate the power
increase. Long-term response (beyond the explicit analyzed period): Low-low set logic cycles the SRVs to control reactor pressure. Feedwater and Condensate pumps
eventually trip, as condenser inventory is depleted. HPCI and/or RCIC initiate to maintain reactor vessel level, as needed. Operators take manual control and guide the plant to a cold shutdown condition using normal plant shutdown procedures. c) Single Failure/Operator Error (as applicable): Failure of Turbine Bypass Valves to open upon demand. d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: RPS trip (Rx Dome Pressure), Control Rod Scram, Recirc. Pump Trip/coast down, SRVs open/close. Long-term: Low-low set logic trip, FW/Condensate pumps trip, as condenser inventory is depleted, HPCI/RCIC initiations (low-low RPV level). The Operators intervene only after the initial transient is over and guide the plant to a stable condition.
Event Category & Acceptance Criteria This is an Abnormal Operating Transient - an expected transient (load rejection)
with the additional single failure of the bypass valves failing to open. Fuel SAFDLs (Section 15.0.4) shall not be exceeded. RPV Pressure shall remain within ASME Upset limits. UFSAR/DAEC - 1 15.1-18 Revision 23 - 5/15 Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). No unique inputs for this analysis.
c) Key Assumptions: Plant is initially at 21.7% of rated and also evaluated at 26% thermal power and
both power levels are evaluated at both 105% and 50% core flowspace. TCVs operate in the two admission mode (3x1 mode - TCVs 1, 2 & 3 move
together and TCV 4 is the controlling valve. Because this event is at low power, TCVs 1, 2 & 3 will be partially open and TCV 4 will be closed. No Operator Actions are assumed during the initial transient response. Results a) Barrier Performance and comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. The results of this event are used to develop the ARTS off-rated limits. See the current cycle's Supplemental Reload Licensing Report (SRLR) for actual values. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. Because the peak pressure is directly proportional to the initial power level, the peak pressure from this event is
not evaluated. Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities: This event is very sensitive to the vessel pressure change. Changes in plant equipment characteristics, e.g., valve stroke times, setpoint changes, steamline lengths and volumes, etc., that will cause the pressurization rate and/or peak vessel pressure to increase will have a negative impact on the event results.
To support certain equipment being out of service during the operating cycle, additional analysis of this event is performed assuming that equipment is not UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 19Operable (Ref. FRED form in Section 15.7). The results of this equipment out-of-service condition are found in the SRLR for the current cycle. c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion
a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an Abnormal Operating Transient.
b) Known Conservatisms/Margins: End-of-cycle core conditions are assumed. Conservative control rod scram times are used.
Bounding SRV opening setpoints (+3% tolerance) Bounding (faster) TCV closure time. c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a Non-Limiting Event and is confirmed as part of the reload analysis for each operating cycle.
15.1.2.2 Turbine Trip (Turbine Stop Valve Closure)
A variety of turbine or nuclear system malfunctions can initiate a turbine trip.
Once initiated, all of the turbine stop valves achieve full closure within about 0.10 sec. Several variations in the turbine trip transients are possible according to the assumptions made concerning the initial power level and the turbine bypass system. These cases are
discussed individually.
15.1.2.2.1 Main Turbine with Bypass (TTWBP) - High Power
Description of Event
a) Initiator: A non-mechanistically caused trip of the main turbine. b) Sequence of Events: The plant is operating at 100% power, when the EHC system receives a trip
signal and the turbine stop valves begin to close. Upon reaching the 90% open (nominal) point, a reactor trip signal (Scram) is initiated, along with an end-of-cycle recirculation pump trip (EOC-RPT). Control rods begin to insert and the recirculation pumps begin to coast down. Both of which help turn around the power increase generated by the collapsing of the voids in the core from the 2014-003 UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 20pressure increase due to the partial loss of a steam path. Although the bypass valves open, their steam flow capacity is not enough initially to control the
pressure. So, reactor pressure increases and SRVs lift to relieve the pressure. Control rods are fully inserted to terminate the power increase. Long-term response (beyond the explicit analyzed period): Low-low set logic armed (SRV open and High Pressure Scram), Turbine bypass valves open to control reactor pressure, Feedwater and Condensate pumps are used to maintain reactor water level. Operators take manual control and guide the plant to a cold
shutdown condition.
c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: RPS trip (TSV closure), EOC-RPT (TSV closure), Control Rod Scram, Recirculation Pump Trip, SRVs open/close. Long-term: Low-low set logic armed, Turbine Bypass Valves open, Feedwater/Condensate pumps operate. The Operators intervene only after the
initial transient is over and guide the plant to a stable condition. Event Category & Acceptance Criteria This is an Anticipated Operational Occurrence - an expected operational transient of moderate frequency (turbine trip) with no other equipment failures or Operator
errors. Fuel SAFDLs (Section 15.0.4) shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. Methods
a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3. No unique inputs for this analysis.
c) Key Assumptions: Plant is initially at rated thermal power and core flow. TSVs close in a linear ramp over 0.1 seconds. No Operator Actions are assumed during the initial transient response.
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 21 Results a) Barrier Performance and comparison to Acceptance Criteria:
Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. See the current cycle's Supplemental Reload Licensing Report (SRLR) for actual values. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed.
b) Known Sensitivities: This event is very sensitive to the vessel pressure change. Changes in plant equipment characteristics, e.g., valve stroke times, setpoint changes, steamline lengths and volumes, etc., that will cause the pressurization rate and/or peak vessel pressure to increase will have a negative impact on the event results. To support certain equipment being out of service during the operating cycle, additional analysis of this event is performed assuming that equipment is not Operable. Specifically, EOC-RPT is assumed to not to function. The results of this equipment out-of-service condition are found in the SRLR for the current
cycle. c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion
a) Statement of Acceptability This event meets all the fission product barrier performance criteria for an
Anticipated Operational Occurrence.
b) Known Conservatisms/Margins: Allowance for 2% core thermal over-power is accounted for in the analysis methods. End-of-cycle core conditions are assumed. Conservative control rod scram times are used. Bounding (faster) TSV closure time.
Bounding SRV opening setpoints (+3% tolerance) 2014-003 UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 22 c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a Non-Limiting Event and is bounded by TTNBP event.
15.1.2.2.2 Main Turbine Trip with Bypass Valve Failure (TTNBP) - High Power
Description of Event
a) Initiator: A non-mechanistically caused trip of the main turbine with failure of the bypass
valves to open.
b) Sequence of Events: The plant is operating at 100% power, when the EHC system receives a trip
signal and the turbine stop valves begin to close. Upon reaching the 90% open (nominal) point, a reactor trip signal (Scram) is initiated, along with an end-of-cycle recirculation pump trip (EOC-RPT). Control rods begin to insert and the recirculation pumps begin to coast down. Both of which help turn around the power increase generated by the collapsing of the voids in the core from the pressure increase due to the loss of a steam path, due to the failure of the turbine
bypass valves to open. Reactor pressure increases and SRVs lift to relieve the pressure. Control rods are fully inserted to terminate the power increase. Long-term response (beyond the explicit analyzed period): Low-low set logic cycles the SRVs to control reactor pressure. Feedwater and Condensate pumps
eventually are no longer available, as condenser inventory is depleted. HPCI and/or RCIC initiate to maintain reactor vessel level, as needed. Operators take manual control and guide the plant to a cold shutdown condition. c) Single Failure/Operator Error (as applicable): Failure of Turbine Bypass Valves to open upon demand. d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: RPS trip (TSV cl osure), EOC-RPT (TSV closure), Control Rod Scram, Recirculation Pump Trip, SRVs open/close.
Long-term: Low-low set logic armed, FW/Condensate pumps operate, HPCI/RCIC initiations (low-low RPV level). The Operators intervene only after the initial transient
is over and guide the plant to a stable condition. 2014-003 UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 23 Event Category & Acceptance Criteria This is an Abnormal Operating Transient - an expected operational transient of moderate frequency (turbine trip), but with the additional single failure of the
bypass valves failing to open. Fuel SAFDLs (Section 15.0.4) shall not be exceeded. RPV Pressure shall remain within ASME Upset limits. Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). No unique inputs for this analysis.
c) Key Assumptions: Plant is initially at rated thermal power and 105% core flow. TSVs close in a linear ramp over 0.1 seconds. No Operator Actions are assumed during the initial transient response.
Results a) Barrier Performance and comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. See the current cycle's Supplemental Reload Licensing Report (SRLR) for actual values. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. See the current cycle's SRLR for
actual values. Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed.
b) Known Sensitivities: This event is very sensitive to the vessel pressure change. Changes in plant equipment characteristics, e.g., valve stroke times, setpoint changes, steamline lengths and volumes, etc., that will cause the pressurization rate and/or peak vessel pressure to increase will have a negative impact on the event results. UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 24 To support certain equipment being out of service during the operating cycle, additional analysis of this event is performed assuming that equipment is not Operable (Ref. FRED form in Section 15.7). The results of this equipment out-of-
service condition are found in the SRLR for the current cycle.
Note: a special case of the TTNBP was analyzed, which assumes the direct RPS signal off the TSV closure is failed, i.e., the RPS trip (Scram) comes off the
resulting high neutron flux (APRM trip) instead. The results of this special case confirmed that the MSIV closure with direct scram failure is the limiting event for
ASME vessel overpressure analysis. See MSIV-F event (Section 15.1.2.3.2). c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion
a) Statement of Acceptability This event meets all the fission product barrier performance criteria for an Abnormal Operating Transient. b) Known Conservatisms/Margins: Allowance for 2% core thermal over-power is accounted for in the analysis methods. End-of-cycle core conditions are assumed. Conservative control rod scram times are used. Bounding (faster) TSV closure time.
Bounding SRV opening setpoints (+3% tolerance)
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a Limiting Event and is re-analyzed as part of the reload
analysis for each operating cycle.
15.1.2.2.3 Main Turbine Trip with Bypass Valve Failure (TTNBP) - Low Power
Description of Event
a) Initiator: A non-mechanistically caused trip of the main turbine with failure of the bypass
valves to open. b) Sequence of Events:
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 25Note: Under the ARTS program (APRM/RBM/Technical Specification), this event is analyzed as a series of four cases representing the combination of thermal
power and core flow: 21.7% power (representing the lowest power level for monitoring of fuel thermal limits), 26% power (representing the bypass of the direct scram signal on the TSVs), 50% core flow and 105% core flow. The results
of these four cases help to define the MCPRp, LHGRFACp and MAPFACp limits in the Core Operating Limits Report (COLR). The limiting case of 26%
power/105% flow is discussed below. The plant is operating at 26% power at 105% rated core flow, when the EHC system receives a trip signal and the turbine stop valves begin to close. Because this power level is below the bypass for the direct scram signal on the TSVs, no direct scram is generated. Reactor pressure increases due to the loss of a steampath, as a result of the failure of the turbine bypass valves to open. The pressure quickly reaches the scram setpoint and control rods begin to insert, which help turn around the power increase generated by the collapsing of the voids in the core from the pressure increase. Reactor pressure increases and SRVs lift to relieve the pressure. Control rods are fully inserted to terminate the power
increase. Long-term response (beyond the explicit analyzed period): Low-low set logic cycles the SRVs to control reactor pressure. Feedwater and Condensate pumps
eventually are no longer available, as condenser inventory is depleted. HPCI and/or RCIC initiate to maintain reactor vessel level, as needed. Operators take manual control and guide the plant to a cold shutdown condition.
c) Single Failure/Operator Error (as applicable): Failure of Turbine Bypass Valves to open upon demand. d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: RPS trip (High Dome Pressure), Control Rod Scram, SRVs open/close. Long-term: Low-low set logic armed and controls pressure, FW/Condensate pumps operate, HPCI/RCIC initiations (l ow-low RPV level). The Operators intervene only after the initial transient is over and guide the plant to a stable
condition. Event Category & Acceptance Criteria This is an Abnormal Operating Transient - an expected transient (turbine trip)
with the additional single failure of the bypass valves failing to open. Fuel SAFDLs (Section 15.0.4) shall not be exceeded. RPV Pressure shall remain within ASME Upset limits. Methods UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 26 a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). No unique inputs for this analysis.
c) Key Assumptions: Plant is initially at 21.7% of rated and also evaluated at 26% thermal power and
both power levels are evaluated at both 105% and 50% core flow. TSVs close in a linear ramp over 0.1 seconds. No Operator Actions are assumed during the initial transient response Results a) Barrier Performance and comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. The results of this event are used to develop the ARTS off-rated limits. See the current cycle's Supplemental Reload Licensing Report (SRLR) for actual values. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. Because the peak pressure is directly proportional to the initial power level, the peak pressure from this event is
not evaluated. Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed.
b) Known Sensitivities: This event is very sensitive to the vessel pressure change. Changes in plant equipment characteristics, e.g., valve stroke times, setpoint changes, steamline lengths and volumes, etc., that will cause the pressurization rate and/or peak vessel pressure to increase will have a negative impact on the event results. To support certain equipment being out of service during the operating cycle, additional analysis of this event is performed assuming that equipment is not Operable (Ref. FRED form in Section 15.7). The results of this equipment out-of-
service condition are found in the SRLR for the current cycle.
c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 27 Conclusion
a) Statement of Acceptability This event meets all the fission product barrier performance criteria for an Abnormal Operating Transient.
b) Known Conservatisms/Margins: End-of-cycle core conditions are assumed. Conservative control rod scram times are used. Bounding TSV closure time.
Bounding SRV opening setpoints (+3% tolerance)
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a Non-Limiting Event and is confirmed as part of the reload analysis for each operating cycle.
15.1.2.3 Main Steam Line Isolation Valve Closure
Automatic circuitry or operator action can initiate the closure of the main steam line isolation valves. Position switches on the valves initiate a scram if valves in three or more main steam lines are less than 90% open and the mode switch is in RUN. However, reactor protection system logic does permit the test closure of one valve without initiating a scram from the position switches. These cases were investigated separately.
15.1.2.3.1 Closure of All Main Steam Line Isolation Valves (MSIV) - High Power
Description of Event a) Initiator: A spurious trip that causes all the MSIVs to rapidly close.
b) Sequence of Events: The plant is operating at 100% power, when a trip signal causes all the MSIVs to rapidly close. Upon reaching the 90% open (nominal) point, a reactor trip signal (Scram) is initiated. Control rods begin to insert. The scram is fast enough to turn
around the power increase generated by the collapsing of the voids in the core from the pressure increase due to the loss of a steam path, due to the closure of
the MSIVs. Reactor pressure increases and SRVs lift to relieve the pressure. The pressure reaches the ATWS-RPT setpoint, which trips the reactor recirculation pumps and they begin to coastdown.
Long-term response (beyond the explicit analyzed period): Low-low set logic cycles the SRVs to control reactor pressure, Feedwater and Condensate pumps 2014-003 UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 28 eventually are no longer available, as condenser inventory is depleted, HPCI and/or RCIC initiate to maintain reactor vessel level, as needed. Operators take manual control and guide the plant to a cold shutdown condition using normal
shutdown procedures.
c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: RPS trip (MSIV closure), Control Rod Scram, ATWS-RPT Recirculation Pump Trip (Reactor Pressure), SRVs open/close.
Long-term: Low-low set logic arm and control pressure, FW/Condensate pumps
operate, HPCI/RCIC initiations (low-low RPV level). The Operators intervene
only after the initial transient is over and guide the plant to a stable condition. Event Category & Acceptance Criteria This is an Anticipated Operational Occurrence - an expected operational transient of moderate frequency (MSIV closure), with no other equipment failures or
Operator errors. Fuel SAFDLs (Section 15.0.4) shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance). b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). No unique inputs for this analysis.
c) Key Assumptions: Plant is initially at rated thermal power and core flow. MSIVs close in a linear ramp over 3.0 seconds. No Operator Actions are assumed during the initial transient response. Results a) Barrier Performance and comparison to Acceptance Criteria:
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 29Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. Because of the direct scram, there is no power increase as a result of this event. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. The pressure increase is much less
severe than other pressurization events, especially those with loss of bypass
capacity. Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed.
b) Known Sensitivities:
This event is only mildly sensitive to changes in plant parameters. Changes in plant equipment characteristics, such as slower scram speed/time, MSIV closure speed and MSIV scram setpoint, will cause the pressurization rate and/or peak
vessel pressure to increase, which will begin to balance the negative reactivity from the direct scram.
c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion a) Statement of Acceptability This event meets all the fission product barrier performance criteria for an Abnormal Operating Occurrence.
b) Known Conservatisms/Margins: Allowance for 2% core thermal over-power is accounted for in the analysis methods. End-of-cycle core conditions are assumed. Conservative control rod scram times are used. Bounding (faster) MSIV closure time.
Bounding SRV opening setpoints (+3% tolerance) c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 30This is considered to be a Non-Limiting Event and is not re-analyzed as part of the reload analysis for each operating cycle.
15.1.2.3.2 Closure of All Main Steam Line Isolation Valves with Direct Scram Failure (MSIVF) - High Power Description of Event
a) Initiator: A spurious trip that causes all the MSIVs to rapidly close. However, the direct scram signal from the MSIV position switches is assumed to fail. Note the MSIV closure with direct scram failure, which is a special case of the MSIV transient, is analyzed as the limiting event for ASME vessel overpressure analysis. Fuel barrier performance is not analyzed as part of this event. b) Sequence of Events: The plant is operating at an overpower condition (102% of rated), when a trip
signal causes all the MSIVs to rapidly close. The power increases quickly by the collapsing of the voids in the core from the pressure increase due to the loss of a steam path from the closure of the MSIVs. Upon reaching the APRM 120% high flux (nominal) trip point, a reactor trip signal (Scram) is initiated. Control rods begin to insert and terminate the power increase. The pressure reaches the ATWS-
RPT setpoint, which trips the reactor recirculation pumps and they begin to coastdown. Reactor pressure continues to increase and SRVs, and potentially the Spring Safety Valves (SSVs), lift to relieve the pressure. Long-term response (beyond the explicit analyzed period): Low-low set logic armed (SRV open and High Pressure Scram) and cycles the SRVs to control reactor pressure. Feedwater and Condensate pumps eventually are no longer
available, as condenser inventory is depleted. HPCI and/or RCIC initiate to maintain reactor vessel level, as needed. Operators take manual control and guide
the plant to a cold shutdown condition.
c) Single Failure/Operator Error (as applicable): MSIV direct scram from the position switches is assumed to fail. d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: RPS trip (APRM Flux - High), Control Rod Scram, ATWS-RPT Recirculation Pump Trip (Reactor Pressure), SRVs (and SSVs) open/close.
Long-term: Low-low set logic armed (SRV open and High Pressure Scram) and cycles
the SRVs to control reactor pressure, FW/Condensate pumps operate, HPCI/RCIC initiations (low-low RPV level). The Operators intervene only after the initial transient
is over and guide the plant to a stable condition.
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 31 Event Category & Acceptance Criteria This is an Abnormal Operating Transient - an expected operational transient of moderate frequency (MSIV closure), with the additional single failure of the MSIV direct scram signal. RPV Pressure shall remain within ASME Upset limits.
Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN. (see Table 15.0-2 for complete listing, code versions and
NRC acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). No unique inputs for this analysis.
c) Key Assumptions: Plant is initially at 102% thermal power and 105% core flow. Reactor Dome Pressure is initially at the corresponding value of 1055 psia. MSIVs close in a linear ramp over 3.0 seconds. MSIV direct position scram fails. No Operator Actions are assumed during the initial transient response.
Results a) Barrier Performance and comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and
APLHGR) are not analyzed for this event.
Reactor Pressure Boundary Performance: Reactor vessel pressure remains below the 1375 psig ASME acceptance limit. See the current cycle's Supplemental
Reload Licensing Report (SRLR) for results.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed.
b) Known Sensitivities: This event is only moderately sensitive to changes in plant parameters. Changes in plant equipment characteristics, such as slower scram speed/time, faster MSIV closure speed and APRM flux scram setpoint, will cause the pressurization rate
and/or peak vessel pressure to increase, which will begin to balance the negative reactivity from the scram.
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 32 c) Uncertainties in Results: Application of the ASME Upset limit as the acceptance criterion compensates for the uncertainties in the analysis methods and input values. Conclusion
a) Statement of Acceptability: This event meets the high vessel pressure performance criteria for an ASME
Upset event. b) Known Conservatisms/Margins: Allowance for 2% core thermal over-power. End-of-cycle core conditions are assumed. Conservative control rod scram times are used. Bounding (faster) MSIV closure time. Failure of the MSIV direct position scram.
Bounding SRV opening setpoints (+3% tolerance). Application of ASME Upset limits as an acceptance criteria, when the Code would allow the application of Emergency limits (pressure < 1500 psig).
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a special case for demonstrating compliance to the ASME vessel overpressure protection requirements and is analyzed as part of the reload
analysis for each operating cycle.
15.1.2.3.3 Closure of One Main Steam Line Isolation Valve (1 MSIV) - High Power Description of Event a) Initiator: A spurious trip (or Operator error) causes one of the MSIVs to rapidly close. However, because only one MSIV closes, the direct scram signal from its position
switch will not cause a trip, by design. b) Sequence of Events: The plant is operating at rated conditions (100% of rated power and core flow), when a trip signal (or Operator error) causes one MSIV to fast close. The remaining three steamlines cannot compensate for the closed steamline and vessel pressure increases. The power increases quickly from the collapsing of the voids in the core from the pressure increase. Upon reaching the APRM 120% high flux (nominal) trip point, a reactor trip signal (Scram) is initiated. Control rods begin to insert and terminate the power increase. The steamflow is rebalanced between the remaining steamlines. Feedwater and pressure control systems react to the UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 33dynamic changes. Reactor pressure continues to increase and SRVs lift to relieve the pressure. The turbine bypass valves also open to control the pressure. Long-term response (beyond the explicit analyzed period): Low-low set logic
cycles the SRVs to control reactor pressure when the decay heat load is high and
then the TCV and Bypass Valves control the pressure once decay heat load is decreased. Feedwater and Condensate pumps maintain reactor vessel level. Operators take manual control and guide the plant to a cold shutdown condition. c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: RPS trip (APRM Flux - High), Control Rod Scram, SRVs open/close. Long-term: Low-low set logic Low-low set logic armed (SRV open and High Pressure Scram) and cycles the SRVs to control reactor pressure. The Operators intervene only after the initial transient is over and guide the plant to a stable condition using normal shutdown procedures. Event Category & Acceptance Criteria This is an Anticipated Operational Occurence - an expected operational transient of moderate frequency (single MSIV closure), with no other equipment failures or
Operator errors. Fuel SAFDLs (Section 15.0.4) shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). No unique inputs for this analysis.
c) Key Assumptions: Plant is initially at rated thermal power and rated core flow. The MSIV closes in a linear ramp over 3.0 seconds. No Operator Actions are assumed during the initial transient response. Results a) Barrier Performance and comparison to Acceptance Criteria:
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 34Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. This is a non-limiting event as the change in fuel thermal limits is bounded by the other
pressurization events.
Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. The pressure increase is much less
severe than other pressurization events, especially those with loss of bypass
capacity.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities: This event is only moderately sensitive to changes in plant parameters. Changes in plant equipment characteristics, such as slower scram speed/time, MSIV closure speed and APRM flux scram setpoint, will cause the pressurization rate
and/or peak vessel pressure to increase, which will begin to balance the negative reactivity from the scram. c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion
a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an
Anticipated Operational Occurrence. b) Known Conservatisms/Margins: Allowance for 2% core thermal over-power is accounted for in the analysis methods. End-of-cycle core conditions are assumed. Conservative control rod scram times are used. Bounding (faster) MSIV closure time.
Bounding SRV opening setpoints (+3% tolerance). c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a Non-Limiting Event and is not re-analyzed as part of
the reload analysis for each operating cycle. 15.1.3 TRANSIENTS RESULTING IN A CORE COOLANT FLOW DECREASE
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 35 Events that result directly in a core coolant flow decrease are those that affect the reactor recirculation system. The following events result in the most significant
transients in this category:
- 1. Recirculation flow control failure - decreasing flow.
- 2. Trip of one recirculation pump.
- 3. Trip of two recirculation pumps.
15.1.3.1 Recirculation Flow Control Failure - Decreasing Flow
Several varieties of recirculation flow control malfunctions can cause a decrease in core coolant flow. A master controller could malfunction in such a way that a zero
speed signal is generated for both recirculation pumps. The recirculation flow control system is provided with a speed demand limiter that is set so that this situation cannot be more severe than the simultaneous tripping of both recirculation pumps. A simultaneous trip of both recirculation pumps is evaluated in Section 15.1.3.3. The master controller has been removed, thus, this is no longer a credible event at the DAEC.
The remaining recirculation flow controller malfunction is one in which the speed controller for one recirculation pump motor-generator (M-G) set fails in such a way that
the speed controller output signal changes in the direction of zero speed. This transient is similar to the trip of one recirculation pump (evaluated in Section 15.1.3.2). However, the pump speed reduction is slower than that resulting from the opening of a field breaker so that the event is bounded by the single recirculation pump trip.
15.1.3.2 Trip Of One Recirculation Pump
NOTE: the information in the following section is historical in nature and was not
updated as part of the Extended Power Uprate Project. It is being presented here to show basic plant response and parametric trends only. Description of Event
g) Initiator: A malfunction occurs that cause one of the main reactor recirculation pumps to trip (e.g., opening the motor-generator set generator field circuit breaker opens)
while the reactor is operating at rated power/flow conditions on the highest permissible loadline.
h) Sequence of Events (NOT a time line): Short term: A malfunction causes one of the main reactor recirculation pumps to trip with the reactor at rated power/core flow conditions. This event is assumed to UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 36 occur on the highest allowable loadline, causing the final power level to be maximized. There is a sudden, swell in water level due to increased voiding in the core. Thus, reactor power initially goes down. The level swell is small and does not reach the high level trip setpoint (Level 8). The level control system quickly compensates for the increase in water level by closing down on the feedwater
regulating valves. The reactor stabilizes at new steady, state conditions. This is a very mild transient on the fuel and vessel. The Operators take control of the plant and maintain the water level and pressure at the new conditions. The plant is
licensed to operate in Single Loop Operation.
Note: a special case of this event is analyzed in Section 15.3.4 - Thermal-Hydraulic Stability.
i) Single Failure/Operator Error (as applicable): None j) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): No trips or other actuations occur during this event. The Operators intervene only after the initial transient is over and maintain the plant in a stable
condition. Event Category & Acceptance Criteria: This is an Anticipated Operational Occurrence - an expected operational transient of moderate frequency (single recirculation pump trip), with no other equipment
failures or Operator errors. Fuel SAFDLs (Section 15.0-4) shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. Methods d) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance).
e) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (See Table 15.0-3) f) Key Assumptions: Plant is initially at rated thermal power and core flow on the highest allowable loadline. Initial vessel level swell does not reach the Level 8 trip point.
Results d) Comparison to Acceptance Criteria: UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 37 Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. Because of the initial level swell (void increase), there is no power increase above the initial
value as a result of this event. Steady State operation in Single Loop is allowed by Technical Specifications, provided that the appropriate adjustments are made in the fuel thermal limits, see the current cycle's Core Operating Limits Report (COLR). Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. The reactor pressure decreases
during this event.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed.
e) Known Sensitivities: The tripping of the M/G set drive motor breaker, instead of the generator field circuit breaker, would maintain the M/G set in the dynamic response, such that its
inertia would lessen the recirculation flow decrease and overall plant response.
f) Uncertainties in Results: The amount of water level swell is the key variable. If the swell reaches the high level trip point, then a turbine trip and feedwater pump would occur.
Conclusion
g) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an
Anticipated Operational Occurrence.
h) Known Conservatisms/Margins: The off-rated power and flow multipliers for the fuel thermal limits (ARTS) are conservatively derived and provide margin to the actual operating limits at the
final steady state operating conditions.
i) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a Non-Limiting Event and is not re-analyzed as part of
the reload analysis for each operating cycle. However, Single Loop Operation is
re-validated as part of each reload analysis. 15.1.3.3 Trip Of Two Recirculation Pumps
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 38 This transient primarily evaluated the fuel thermal margin maintained by the rotating inertia of the recirculation system drive equipment. The inertia from the recirculation flow control system M-G sets is included because no single event can simultaneously open the generator field circuits of both M-G sets. This transient results if the power supply to both M-G sets is lost, the most-likely cause of which would be a
loss-of-offsite power (LOOP), which is discussed in Section 15.1.2. A special case of
this event is discussed in Section 15.3.4. UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 3915.1.4 TRANSIENTS REACTIVITY AND POWER DISTRIBUTION ANOMALIES
Events that result directly in rapid power increases and core reactivity are
included in this section. The following events result in a positive reactivity insertion:
- 1. Continuous rod withdrawal during power range operation.
- 2. Continuous rod withdrawal during reactor startup.
- 3. Control rod removal error during refueling.
- 4. Fuel assembly insertion error during refueling.
15.1.4.1 Rod Withdrawal Error At Power
Description of Event
k) Initiator: With the reactor operating at high power (rated power and core flow), the
Operator selects the highest worth, fully-inserted control rod and begins to
continuously withdraw it.
l) Sequence of Events: The reactor is operating at high power (> 85% of rated), when the Operator
selects a fully inserted control rod, with the highest rod worth, and begins to continuously withdraw it at the maximum withdrawal speed of 3.6 inches/sec. The local power in the adjacent fuel assemblies begins to increase. When the increase in power begins to approach the thermal limits for those bundles, the Local Power Range Monitors (LPRMs) will alarm. The Operator ignores these alarms and continues to withdraw the control rod. The Rod Block Monitor (RBM) also monitors the local power change and alarms when the change in power reaches the setpoint. The Operator ignores the alarm and continues to withdraw the control rod. At the 108% rod block setpoint (Analytical Limit), the RBM
generates a rod block to prevent further rod withdrawal before the local power increase can violate the fuel thermal limits.
m) Single Failure/Operator Error (as applicable): The Operator ignores the LPRM and RBM alarms and continues to withdraw the control rod. The most-responsive channel of the RBM is assumed to be inoperable. Random LPRM failures are assumed in the analysis.
n) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures) RBM rod block at the high power (>85% of rated) setpoint (>108%).
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 40 Event Category & Acceptance Criteria This is an Abnormal Operating Transient - an expected operational transient (continuous rod withdrawal), but with the Operator error (ignores the LPRM and RBM alarms) and the additional equipment failure of one channel of RBM. Fuel SAFDLs shall not be exceeded. Methods g) Calculation Tools & Computer Codes: The primary code used to perform this analysis is the GE 3-D Core Simulator (PANACEA), with input to another code (GROMT) that simulates the RBM system. (See Table 15.0-2 for complete listing, code versions and NRC
acceptance). h) Inputs (Reference common list in 15.0, and/or include event-specific items): Cycle-specific fuel bundle designs (FRED form - see Section 15.0.7). RBM setpoint (FRED form).
i) Key Assumptions: Reactor is operating at high power/flow. Error rod is assumed to be the highest worth control rod in the core. The fuel assemblies adjacent to the error rod are initially operating at the maximum allowable fuel thermal limits (Operating Limits). The "most responsive" channel of RBM is not available/operable during the
event. Results g) Comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. See the current cycle's Supplemental Reload Licensing Report (SRLR) for actual values.
Reactor Pressure Boundary Performance: Reactor vessel pressure is not
challenged by this event and is not explicitly analyzed. Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. h) Known Sensitivities: The results are somewhat sensitive to the initial core power; hence, the power-dependent setpoints for the RBM system.
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 41 i) Uncertainties in Results The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. The original generic RWE analysis performed for the APRM, RBM and Technical Specification (ARTS) program (Ref. 15.0-57), which determined the
power-dependent RBM setpoints, was done to 95%/95% confidence levels. Conclusion
j) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an Abnormal Operating Transient.
k) Known Conservatisms/Margins: The control rod pattern is manipulated in the analysis to generate the high worth rod and localized conditions of the adjacent fuel assemblies operating at the allowable thermal limits as an initial condition of the event. Maximum allowable control rod withdrawal speed is used. A random distribution of LPRM failures is assumed. l) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is a Limiting Event and is evaluated as part of the cycle-specific reload analysis. This evaluation is done to confirm the original generic RWE analysis done for the ARTS program.
15.1.4.2 Rod Withdrawal Error At Startup
Description of Event
a) Initiator: The reactor is critical and in the startup range when the Operator makes a
selection error (out-of-sequence rod) and continuously withdraws the control rod with the highest rod worth from the fully inserted position.
b) Sequence of Events: With the reactor critical and operating in the startup range, the Operator makes a
selection error of an out-of-sequence control rod and continuously withdraws the highest worth control rod in the core from the full-in position at the maximum withdrawal speed (3.6 inches/sec). The Rod Worth Minimizer (RWM) is not
functioning and the second Licensed Operat or does not catch the out-of-sequence control rod selection and withdrawal. The core power reaches the scram setpoint of the Intermediate Range Neutron Monitor (IRM), and the scram inserts the
control rods, including the error rod, and stops further increases in core power.
c) Single Failure/Operator Error (as applicable): UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 42The most responsive channel (nearest to the control rod) of the IRM system is assumed to fail/not operable (bypassed).
d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): The operable IRM channels trip and generates an RPS scram (dependent on IRM range, trip
setpoint of either 40/40 or 125/125 of scale (Analytical Limits)). Control Rods scram at Technical Specification insertion speed.
The second Licensed Operator does not catch th e out-of-sequence control rod selection and withdrawal. Event Category & Acceptance Criteria This is an Abnormal Operating Transient - an expected operational transient (rod withdrawal error), but with the additional single failure (most-responsive IRM channel fails).
Note: in reality, this is a highly unlikely event, as multiple equipment failures (RWM and IRM), coupled with multiple Operator errors (selection error and
second Licensed Operator error) have to occur for this event to happen.
Fuel SAFDLs shall not be exceeded. In particular, the peak fuel enthalpy shall be < 170 cal/gm. Methods a) Calculation Tools & Computer Codes: This is a generic analysis and is not done on a plant-specific basis (Ref. 15.0-2).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): This is a generic analysis and is not done on a plant-specific basis (Ref. 15.0 -2). c) Key Assumptions: Reactor is critical and in the startup range, below the low power setpoint of the RWM. Error rod is assumed to be the highest worth control rod in the core. The most responsive channel (nearest channel) of IRM is not available/operable
during the event. Results a) Comparison to Acceptance Criteria: Fuel Performance: Peak fuel enthalpy is well below the limit of 170 cal/gm.
Reactor Pressure Boundary Performance: Reactor vessel pressure is not
challenged by this event and is not explicitly analyzed.
UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 43Containment Performance: The containment is not challenged by this event and is not explicitly analyzed. b) Known Sensitivities The result of this event is primarily sensitive to the rod worth of the error rod.
c) Uncertainties in Results Use of conservative assumptions in the evaluation are intended to bound the uncertainties in the final results.
Conclusion a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an Abnormal Operating Transient.
b) Known Conservatisms/Margins: Maximum allowable control rod withdrawal speed is used. The error control rod is assumed to be fully withdrawn, when in fact, the scram will terminate the withdrawal after only partial withdrawal. The most-responsive IRM channel is not available (bypassed), which delays the scram. Banked Position Withdrawal Sequence (BPWS), which is programmed into the RWM limits control rod worth to well below the value assumed in this evaluation.
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents) This is a non-Limiting Event and is not evaluated as part of the cycle-specific
reload analysis.
15.1.4.3 Control Rod Removal Error During Refueling
The nuclear characteristics of the core ensure that the reactor is subcritical even in its most reactive condition with the most reactive control rod fully withdrawn during
refueling.
When the mode switch is in REFUEL, only one control rod can be withdrawn.
The selection of a second rod initiates a rod block, thereby preventing the withdrawal of more than one rod at a time. Therefore, the refueling interlocks prevent any condition UFSAR/DAEC - 1 15.1- Revision 23 - 5/15 44 that could lead to inadvertent criticality due to a control rod withdrawal error during refueling.
In addition, the design of the control rod, incorporating the velocity limiter, does not physically permit the upward removal of the control rod without the simultaneous or prior removal of the four adjacent fuel bundles, thus eliminating any hazardous condition.
15.1.4.4.1 Fuel Assembly Insertion Error During Refueling-Inadvertent Criticality
The core is designed such that it can be made subcritical under the most reactive
conditions with the strongest control rod fully withdrawn. The refueling shutdown margin is determined each cycle by using a 3-D, safety-related BWR simulator code as referenced in Table 15.0-2. Refueling shutdown margin is confirmed to meet Technical
Specification LCO 3.1.1. Therefore, any si ngle fuel bundle can be positioned in any available location without violating the shutdown criteria, providing all the control rods are fully inserted. The refueling interlocks require that all control rods must be fully inserted before a fuel bundle may be inserted into the core. Because of the above-mentioned constraints, there is no analysis required for this event.
15.1.4.4.2 Fuel Loading Error - Mislocated Bundle
Description of Event
a) Initiator: During the fuel reloading process, two bundles are misloaded into the core in the
opposite core locations (i.e., swapped).
b) Sequence of Events: During the fuel reloading process, two bundles are loaded into the core in the opposite core locations. These mislocated bundles are not discovered during the
core loading verification process and the reactor is started up and operates at rated
power and flow with the bundles in the wrong locations for the entire fuel cycle.
c) Single Failure/Operator Error (as applicable) The verification of the core loading does not catch the mislocated bundle error.
d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures) None. UFSAR/DAEC-1 15.1-45 Revision 23 - 5/15 Event Category & Acceptance Criteria This is an Abnormal Operating Transient - an expected operational transient (mislocated bundle), but with the Operator/Reactor Engineer error (fails to catch the mislocated bundles during the various core loading verifications.).
Fuel SAFDLs shall not be exceeded.
Methods a) Calculation Tools & Computer Codes The primary code used to perform this analysis is the GE 3-D Core Simulator (PANACEA). (See Table 15.0-2 for complete listing, code versions and NRC
acceptance). b) Inputs (Reference common list in 15.0, and/or include event-specific items) Cycle-specific fuel bundle designs (FRED form-see Section 15.0.7).
c) Key Assumptions A high power bundle is swapped with a low power bundle in a core cell that maximizes the power increase on the high power bundle. The core cell containing the mislocated high power bundle is not a location directly monitored by an LPRM string or a TIP monitor (i.e., separated by at least one fuel bundle from the detectors). Results a) Comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. Reactor Pressure Boundary Performance: Reactor vessel pressure is not
challenged by this event and is not explicitly analyzed. Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities The results of this event are dependent upon the mismatch of the bundle powers
and core loading locations involved. UFSAR/DAEC-1 15.1-46 Revision 23 - 5/15 c) Uncertainties in Results The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion a) Statement of Acceptability This event meets all the fission product barrier performance criteria for an Abnormal Operating Transient.
b) Known Conservatisms/Margins Use of TIP adaptive core monitoring methods would reduce the impact if the mislocated bundle were in a monitored location (highly likely), as the local power would be adjusted to maintain the fuel within thermal limits and there would be no impact from the misloading.
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents) This is a non-Limiting Event and is confirmed as part of the cycle-specific reload analysis.
15.1.4.4.3 Fuel Loading Error - Rotated Bundle Description of Event a) Initiator: During the fuel reloading process, a bundle is misloaded into the core in the
proper core location, but is rotated in orientation by either 90° or 180°, whichever
produces the worst result. b) Sequence of Events: During the fuel reloading process, a bundle is misloaded into the core in the
proper core location, but is rotated in orientation by either 90° or 180°, whichever produces the worst result. This misoriented bundle is not discovered during the
core loading verification process and the reactor is started up and operates the
entire cycle with the bundle in the wrong orientation for the entire fuel cycle.
c) Single Failure/Operator Error (as applicable): The verification of the core loading does not catch the error of the misoriented
bundle. d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): None. 2014-003 UFSAR/DAEC-1 15.1-47 Revision 23 - 5/15 Event Category & Acceptance Criteria This is an Abnormal Operating Transient - an expected operational transient (misoriented bundle), but with the Operator/Reactor Engineer error (fails to catch the misoriented bundle during the various core loading verifications.).
Fuel SAFDLs shall not be exceeded. Methods a) Calculation Tools & Computer Codes: The primary code used to perform this analysis is the GE 3-D Core Simulator (PANACEA). (See Table 15.0-2 for complete listing, code versions and NRC
acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): Cycle-specific fuel bundle designs (FRED form-see Section 15.0.7).
c) Key Assumptions: A high power bundle is rotated in a core cell that maximizes the power increase on the rotated bundle (i.e., maximizes the rotated R-factor on the bundle). To account for the fact that the misoriented bundle may be tilted and not seated
correctly, a bias of 0.02 CPR is added to the results to account for the variable water gap.
Results a) Comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. See the current cycle's Supplemental Reload Licensing Report (SRLR) for actual values. Reactor Pressure Boundary Performance: Reactor vessel pressure is not
challenged by this event and is not explicitly analyzed.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities: The results of this event are dependent upon the pin-to-pin peaking factor of the
bundle (rotated R-factor) and its core loading location.
c) Uncertainties in Results:
UFSAR/DAEC-1 15.1-48 Revision 23 - 5/15 The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. A bias of 0.02 CPR is added to the results to account for the uncertainty due to a variable water gap if the fuel assembly not seated correctly and is tilted toward the control rod (preferential direction). Conclusion a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an Abnormal Operating Transient.
b) Known Conservatisms/Margins: There are five separate visual indicati ons of proper bundle orientation. Operating experience has shown that these indications are readily visible during the fuel loading process. Thus, the actual probability of this event is quite low.
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents) This is a Limiting Event and is evaluated as part of the cycle-specific reload
analysis.
UFSAR/DAEC-1 15.1-49 Revision 23 - 5/15 15.1.5 TRANSIENTS RESULTING IN AN INCREASE IN CORE FLOW
Events that result in an increase in core coolant flow cause a decrease in core void
fraction and a corresponding increase in neutron flux and reactor power.
The following are the identified events in this category:
a) Startup of an Idle Recirculation Pump b) Recirculation Flow Controller Failure - Increasing Flow (Fast) c) Recirculation Flow Controller Failure - Slow Flow Runout 15.1.5.1 Startup Of An Idle Recirculation Pump
NOTE: the information in the following section is historical in nature and was not
updated as part of the Extended Power Uprate Project. It is being presented here to show basic plant response and parametric trends only. Description of Event
a) Initiator: The plant is initially operating in Single Loop Operation (SLO), when the Operator starts up the idle recirculation pump without pre-warming the coolant in
the loop, as required by procedures and Technical Specifications.
b) Sequence of Events (NOT a time line): Case a) The plant is initially operating in SLO at 68% power (pre-Uprate) and 48% core flow, when the Operator starts up the idle recirculation pump without pre-warming the loop. The pump discharge valve is opened. The resulting surge in core flow, with the accompanying decrease in inlet subcooling from the slug of colder water in the idle loop, collapses the
voids in the core and reactor power increases. However, the resulting
increase in core power/neutron flux is not great enough to cause an APRM flow-biased scram. Reactor vessel water level and pressure are only slightly affected by this event. Case b) The plant is initially operating in SLO at 55% power (pre-Uprate) and 38% core flow, when the Operator starts up the idle recirculation pump without pre-warming the loop. The pump discharge valve is opened. The resulting surge in core flow, with the accompanying decrease in inlet subcooling from the slug of colder water in the idle loop, collapses the
voids in the core and reactor power increases. In this case, the resulting
increase in neutron flux/core power is sufficient to reach the APRM flow-UFSAR/DAEC-1 15.1-50 Revision 23 - 5/15 biased scram setting and a reactor scram occurs. Again, reactor water level and pressure are only slightly affected by this transient. c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Case a) Feedwater and Pressure control systems respond to these changes and maintain vessel level and pressure within normal operating limits. Case b) RPS trip on reaching the APRM flow-biased scram trip setpoint and Control Rod Scram. Feedwater and Pressure control systems respond to these changes and maintain vessel level and pressure within normal operating limits. Event Category & Acceptance Criteria: This is an Anticipated Operational Occurrence - a single transient of moderate frequency (idle recirculation pump start with inadequate loop warmup) with no other equipment failures or Operator errors. Fuel SAFDLs shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. Methods a) Calculation Tools & Computer Codes: This is an historical event analysis. The computer code/version used to perform
this evaluation is no longer used. Any future re-analysis would be a change in methods.
b) Inputs (Reference common list in 15.0, and/or include event-specific items): This is an historical event analysis. The inputs used are no longer valid, due to
Extended Power Uprate and other plant changes since the original analysis was
conducted. c) Key Assumptions: Recirculation pump starts up in 8 seconds from closing the generator field breaker. The recirculation discharge valve stroke time is 30 secs. The temperature difference between the reactor and the coolant in the idle loop is
100°F. Results UFSAR/DAEC-1 15.1-51 Revision 23 - 5/15 a) Comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain well within their respective acceptance limits.
Reactor Pressure Boundary Performance: Reactor vessel pressure is not significantly affected by this event and is well within limits.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities: The results of this event are highly sensitive to the initial conditions. The two
cases analyzed represent the bounding results. Case a) represents the highest
power level achievable in SLO (pre-Uprate). At this higher power and core flow, the resulting core P is sufficiently high to cause part of the running recirculation loop flow to bypass the core and cause reverse flow in the idle loop jet pumps.
Thus, when the idle loop is started up, there is an initial resistance of this reverse flow to be overcome, which limits the impact of the surge of cooler water from the idle loop. However, in Case b), the power and flow are maximized to be just below the point where reverse flow occurs in the idle loop jet pumps. Thus, when
the idle loop is started up, the flow in the idle loop is in the forward direction, which causes a greater initial impact, as the surge of cooler water goes immediately into the core, causing a higher spike in neutron flux and resulting scram to be generated.
c) Uncertainties in Results: Use of conservative assumptions (e.g., T greater than allowed by Tech Specs) in the evaluation are intended to bound the uncertainties in the final results. Conclusion a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an
Anticipated Operational Occurrence. b) Known Conservatisms/Margins: This analysis used a change in coolant temperature of 100°F, whereas TS limit
this to 50°F. Per SIL No. 517, the licensing basis for this event was changed to allow the 50°F change in temperature as part of the ARTS program. c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is a non-Limiting Event and is not evaluated as part of the cycle-specific
reload analysis. UFSAR/DAEC-1 15.1-52 Revision 23 - 5/15
15.1.5.2 Recirculation Flow Controller Failure - Increasing Flow (Fast)
Description of Event
a) Initiator: A failure in the recirculation flow controller causes one recirculation pump to increase speed at maximum rate. b) Sequence of Events (NOT a time line): The plant is initially at 55.7% core thermal power and 39% of rated core flow - two pump minimum speed at the highest allowable core loadline. A failure within the recirculation flow control system causes one of the recirculation pumps to increase speed at the maximum rate. The pump runs out to the MG set scoop tube
lockup position. Core flow increases and initially reduces core voiding, leading to
an increase in neutron flux/reactor power. The increase is large enough to cause a reactor scram on high neutron flux. The control rods insert and terminate the event. Initially reactor level drops, due to the mismatch between steam flow and
feedwater flow. The feedwater control system reacts and increases feed flow to recover level. Reactor pressure is only mildly affected by this event and the pressure control system easily maintains pressure to the initial value.
c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): RPS trips (high neutron flux-fixed and Control Rod Scram. Feedwater Control system reacts to the steam flow-feed flow mismatch and corrects the decreasing water level. The Operators
intervene only after the initial transient is over and guide the plant to a stable condition.
Event Category & Acceptance Criteria: This is an Anticipated Operational Occurrence - an expected operational transient of moderate frequency (recirculation pump speed increase) with no other equipment failures or Operator errors.
Fuel SAFDLs shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. UFSAR/DAEC-1 15.1-53 Revision 23 - 5/15 Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete
listing, code versions and NRC acceptance). b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). c) Key Assumptions: The controller increases pump speed at the maximum rate of 25%/sec. MG set scoop tube positioner set at 102.5% speed (nominally 1710 RPM). Note: this evaluation is for EPU and was not re-performed for ICF, which permits core flow up to 105% of rated flow.
Results a) Comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. The calculated TOP (37.99%) exceeded the TOP limit of 28%. However, when the ARTS MAPFACp multiplier is applied for the off-rated condition, the adjusted TOP is within limits. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities: Small changes in the initial power/flow point have a negligible impact on the
results.
The increase in maximum pump flow to 105% for ICF does not impact these results as the scram terminates the event prior to the recirculation pump reaching
the 102.5% speed in the original analysis.
c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods.
UFSAR/DAEC-1 15.1-54 Revision 23 - 5/15 Conclusion a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an
Anticipated Operational Occurrence. b) Known Conservatisms/Margins: The recirculation flow controllers have a clamp (lockup) on rate of change at 40%/minute (0.67%/sec). End-of-cycle core conditions are assumed. Conservative control rod scram times are used.
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is a non-Limiting Event and is not evaluated as part of the cycle-specific
reload analysis. It was included in the analyses for Extended Power Uprate, as required by the NRC to confirm that it remained a non-limiting event after the
uprate. 15.1.5.3 Recirculation Flow Controller Failure - Slow Flow Runout Description of Event a) Initiator: A failure in the recirculation flow controller causes one (or both) recirculation pump(s) to increase speed at a slow rate. b) Sequence of Events (NOT a time line): The plant is initially at 55.7% core thermal power and 39% of rated core flow - two pump minimum speed at the highest allowable core loadline. A failure within the recirculation flow control system causes one (or both) of the recirculation pumps to increase speed at a slow rate. The pump runs out to the MG set scoop
tube lockup position. Core flow increases and initially reduces core voiding, leading to an increase in neutron flux/reactor power. The increase is slow enough such that the core conditions are in quasi-equilibrium. The feedwater control system maintains vessel water level. Reactor pressure is not affected by this event and the pressure control system easily maintains pressure at the initial value.
c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Feedwater Control system and Pressure Control System maintain vessel level and pressure constant, as this is a slow event. The Operators intervene only after the initial transient is over
and guide the plant to a stable condition. UFSAR/DAEC-1 15.1-55 Revision 23 - 5/15 Event Category & Acceptance Criteria: This is an Anticipated Operational Occurrence - an expected operational transient of moderate frequency (recirculation pump speed increase) with no other equipment failures or Operator errors.
Fuel SAFDLs shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. Methods a) Calculation Tools & Computer Codes: Primary Code - ISCOR (see Table 15.0-2 for complete listing, code versions and
NRC acceptance). b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (Table 15.0-3). Cycle-specific fuel bundle designs (FRED form-see Section 15.0.7). c) Key Assumptions: The rate of change in recirculation pump speed/core flow is slow enough that the
reactor is in a quasi-steady state condition throughout the event. MG set scoop tube positioner set at 102.5% speed (nominally 1710 RPM). Note: this evaluation is for EPU and was not re-performed for ICF, which permits core
flow up to 105% of rated flow. Results a) Comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. The slow recirculation flow increase event is the basis for the ARTS MCPR f limits. Because ARTS MCPRf curves exist for 107% of rated core flow, ICF implementation does not impact these results. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit.
Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities:
UFSAR/DAEC-1 15.1-56 Revision 23 - 5/15 Small changes in the initial power/flow point have a negligible impact on the results. c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods.
Conclusion a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an
Anticipated Operational Occurrence.
b) Known Conservatisms/Margins: The resulting MCPR f multiplier is based upon an MCPR Safety Limit of 1.08. End-of-cycle core conditions are assumed.
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is a Limiting Event, but is not evaluated as part of the cycle-specific reload
analysis. It was included in the analyses for Extended Power Uprate, as required by the NRC to confirm the MCPR f multiplier.
UFSAR/DAEC-1 15.1-57 Revision 23 - 5/15 15.1.6 TRANSIENTS RESULTING IN AN INCREASE IN REACTOR COOLANT INVENTORY
There are no events in this category analyzed in the original FSAR with the
possible exception of the feedwater controller failure, which is discussed in Section
15.1.1.1. UFSAR/DAEC-1 15.1-58 Revision 23 - 5/15 15.1.7 TRANSIENTS RESULTING IN A REACTOR VESSEL COOLANT INVENTORY DECREASE Transients that result directly in a decrease of reactor vessel coolant inventory are those that either restrict the normal flow of fluid into the vessel, increase the removal of fluid from the vessel, or are the result of the inadvertent opening of a relief or safety
valvereactor coolant pressure boundary. Events identified in this category are the
following:
- 1. Pressure regulator failure.
- 2. Inadvertent opening of a relief or safety valve.
- 3. Loss of feedwater flow.
- 4. Trip of One Feedwater Pump.
15.1.7.1 Pressure Regulator Failure - Open
NOTE: the information in the following section is historical in nature and was not updated as part of the Extended Power Uprate Project. It is being presented here to show basic plant response and parametric trends only. Description of Event a) Initiator: A failure of either the primary or back-up pressure regulator occurs at rated
conditions, sending a signal to the turbine control and turbine bypass valves to open to the maximum combined flow limit setpoint. b) Sequence of Events (NOT a time line): Short term: Either the primary or backup pressure regulator fails to the full open position. This causes the turbine control valves to open to full flow and turbine bypass valve to fully open, because the maximum combined flow limiter setpoint in the EHC system is set at 125% of rated steamflow. This sudden increase in steamflow causes the vessel pressure to drop. The decrease in pressure causes a sudden swell in water level due to
increased voiding in the core. Thus, reactor power initially goes down. The level swell is
large enough to reach the high level trip setpoint (Level 8), causing a turbine trip and feedwater pump trip. The turbine trip will initiate the RPS trip (scram) and EOC-RPT trip of the recirculation pumps upon turbine stop valve closure. However, the pressure
regulator failure causes the bypass valves to be fully open, so the resulting pressure increase from the turbine trip is mild. In addition, the turbine trip was initiated from less than the initial power level, which also reduces the impact of the turbine trip. Thus, only
one or two SRVs are needed to open to regulate the vessel pressure. The turbine inlet
pressure eventually drops below the low pressure setpoint and the MSIVs close on low UFSAR/DAEC-1 15.1-59 Revision 23 - 5/15 steamline pressure to preclude an unacceptable cooldown rate on the reactor pressure vessel. This essentially terminates the event.
Long-term response (beyond the explicit analyzed period): Low-low set logic cycles the SRVs to control reactor pressure. Feedwater and Condensate pumps eventually trip, as
condenser inventory is depleted. HPCI and/or RCIC initiate to maintain reactor vessel level, as needed. Operators take manual control and guide the plant to a cold shutdown condition using normal shutdown procedures.
c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Short term: Turbine and Feedwater pump trip on high vessel level - Level 8 RPS trip (TSV closure), Control Rod Scram, MSIV closure on low steamline pressure (< 850 psig in RUN - nominal), SRV open and close. Long-term: Low-low set logic trip, FW/Condensate pumps trip, HPCI/RCIC initiations (low-low RPV level). The Operators intervene only after the initial transient is over and guide the plant to a stable condition.
Event Category & Acceptance Criteria: This is an Anticipated Operational Occurrence - an expected operational transient of moderate frequency (pressure regulator failure), with no other equipment failures or
Operator errors. Fuel SAFDLs shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits. Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (Note: because this analysis is
historical, an earlier NRC-approved version of the code was used). b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (See Table 15.0-3) EHC Maximum Combined Flow Limiter setting of 130% of rated steamflow was used in
the actual analysis.
c) Key Assumptions: Plant is initially at rated thermal power and core flow. MSIVs close in a linear ramp over 3.0 seconds.
UFSAR/DAEC-1 15.1-60 Revision 23 - 5/15 Results a) Comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. Because of the initial level swell (void increase) and direct scram, there is no power increase
above the initial value as a result of this event. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. The pressure increase is much less
severe than other pressurization events, especially those with loss of bypass capacity. Also, the MSIV closure on low steamline pressure precludes an
unexceptable cooldown rate on the pressure vessel. Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed.
b) Known Sensitivities: The initial depressurization rate of the vessel (MCFL setting) sets the amount of
level swell experienced.
Sensitivity studies were performed as part of the resolution of GE SIL 502 (Ref. 15.0-58) that demonstrated that the event is not very sensitive to MSIV stroke time (10 second versus 5 seconds) and only mildly sensitive to MSIV isolation
pressure setpoint (800 psig versus 850 psig).
c) Uncertainties in Results: The plant performance is analyzed to 95%/95% confidence levels using GEMINI methods. Conclusion a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an
Anticipated Operational Occurrence. b) Known Conservatisms/Margins: End-of-cycle core conditions are assumed. Conservative control rod scram times are used. Maximum Technical Specification MSIV closure time.
c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): UFSAR/DAEC-1 15.1-61 Revision 23 - 5/15 This is considered to be a Non-Limiting Event and is not re-analyzed as part of the reload analysis for each operating cycle.
15.1.7.2 Inadvertent Opening Of A Safety/Relief Valve
NOTE: the information in the following section is historical in nature and was
not updated as part of the Extended Power Uprate Project. It is being presented here to show basic plant response and parametric trends only. Description of Event
a) Initiator: A malfunction occurs that cause one Safety/Relief Valve (S/RV) to open while the reactor is operating at rated power/flow
conditions. b) Sequence of Events (NOT a time line): Short term: A malfunction causes one S/RV to open with the reactor at rated power/core flow conditions. There is a sudden, short increase in steamflow, which
causes the vessel pressure to drop. The decrease in pressure causes a sudden swell
in water level due to increased voiding in the core. Thus, reactor power initially goes down. The level swell is small and does not reach the high level trip setpoint (Level 8). The pressure regulator quickly compensates for the drop in reactor
pressure by closing down on the turbine control valves. The reactor stabilizes at new steady, state conditions. This is a very mild transient on the fuel and vessel.
Longterm (beyond the explicit analyzed period): The Operators take control of the plant and attempt to close the open S/RV. If the valve can not be closed, the reactor is brought to a cold shutdown condition using normal operating procedures. Containment cooling is initiated to handle the steam flow to the suppression pool from the open S/RV. c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): No trips or other actuations occur during this event. The Operators intervene only after the initial transient is over and guide the plant to a stable
condition. UFSAR/DAEC-1 15.1-62 Revision 23 - 5/15 Event Category & Acceptance Criteria: This is an Anticipated Operational Occurrence - an expected operational transient of moderate frequency (open S/RV), with no other equipment failures or Operator errors. Fuel SAFDLs shall not be exceeded.
RPV Pressure shall remain within ASME Upset limits.
Methods a) Calculation Tools & Computer Codes: Primary Code - ODYN, using GEMINI methods. (see Table 15.0-2 for complete listing, code versions and NRC acceptance). b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (See Table 15.0-3)
c) Key Assumptions: Plant is initially at rated thermal power and core flow.
Initial vessel level swell does not reach the Level 8 trip point. Results
a) Comparison to Acceptance Criteria: Fuel Performance: Fuel thermal limits (MCPR, LHGR (MOP and TOP) and APLHGR) all remain within their respective acceptance limits. Because of the initial level swell (void increase), there is no power increase above the initial value as a result of this event. Reactor Pressure Boundary Performance: Reactor vessel pressure remains well below the 1375 psig ASME acceptance limit. The reactor pressure decreases
during this event. Containment Performance: The containment is not challenged by this event and is
not explicitly analyzed. b) Known Sensitivities: The initial depressurization rate of the vessel sets the amount of level swell experienced. c) Uncertainties in Results:
UFSAR/DAEC-1 15.1-63 Revision 23 - 5/15 The amount of level swell is the largest uncertainty and would have the most impact on the results, especially if the Level 8 trips were reached. Conclusion a) Statement of Acceptability: This event meets all the fission product barrier performance criteria for an Anticipated Operational Occurrence. b) Known Conservatisms/Margins: None c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is considered to be a Non-Limiting Event and is not re-analyzed as part of
the reload analysis for each operating cycle.
15.1.7.3 Loss Of Feedwater Flow
Description of Event
a) Initiator: A failure in the feedwater control system (or other cause) trips both Feedwater pumps with the reactor at rated thermal power/core flow
conditions. b) Sequence of Events (NOT a time line): Both Feedwater pumps trip and begin to coastdown. Reactor vessel level begins to drop with the loss of feedwater. Recirculation pumps begin to runback to minimum speed upon the loss of both feedwater pumps after 15 seconds. The recirculation pump runback helps initially moderate the level drop. A reactor scram occurs when level reaches Level 3. Level
continues to decrease to Level 2 and RCIC is initiated and starts to inject in about 30 seconds. Water level inside the
vessel will begin to recover when the injected flow exceeds the steamflow from vessel (decay heat). If water level in the downcomer region doesn't recover fast enough, the level may reach the Level 1 trips (Low Pressure ECCS, ADS and Group I isolation - MSIV closure). Low Pressure
ECCS will not actually inject, as the reactor pressure remains high. The Operators would monitor recovering level and inhibit ADS actuation before the 2 minute timer
expires. The MSIV closure would initially collapse the water level, but there is sufficient margin to maintain ample
core coverage. The S/RVs would cycle (Low-Low Set) to UFSAR/DAEC-1 15.1-64 Revision 23 - 5/15 maintain pressure after the vessel isolation. The Operators would take over and guide the plant to a stable condition
and eventually to cold shutdown. c) Single Failure/Operator Error (as applicable): The HPCI system is assumed to not operate.
d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Recirculation pump runback to minimum speed, RPS trip on low level (Level 3) , Control Rod Drive (Scram), RCIC initiation at low-low level (Level 2). If water level reaches low-low-low (Level 1), CS and LPCI initiation, ADS timer start, and
Group I isolation (MSIV closure). Operators will inhibit ADS actuation if water level reaches the Level 1 trip
Event Category & Acceptance Criteria: This is an Abnormal Operating Transient - an expected operational transient of moderate frequency (loss of feedwater flow), but with the additional single failure (HPCI fails to operate).
Because this is a unique analysis, to satisfy NUREG-0737, Item II.K.3.44, the acceptance safety criterion is that the RCIC system is able to maintain reactor vessel level above the Top of Active Fuel (TAF).
There is a second, operational, acceptance criterion of Level 1 trip avoidance. Methods a) Calculation Tools & Computer Codes: Primary Code - SAFER (see Table 15.0-2 for complete listing, code versions and
NRC acceptance).
b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-4 (see Table 15.0-4)
OPL-3 (See Table 15.0-3) Condensate Store Tank water temperature of 125°F.
c) Key Assumptions: Initial water level is the normal operating water level. The feedwater flow linearly ramps to zero in 5 seconds. Recirculation pumps trip, with a 5 second coastdown, at Level 3 (Scram) to simulate the actual pump runback to minimum speed. Results a) Comparison to Acceptance Criteria: UFSAR/DAEC-1 15.1-65 Revision 23 - 5/15 For the safety criterion, the minimum water level inside the core shroud (i.e., above the core) remains several feet above the Top of Active Fuel. However, there is some probability that the level outside the shroud (i.e., indicated level for vessel instrumentation) will not remain above the Level 1 trip point. Thus, conservatively, we assume that the operational acceptance criterion is not met. This is found to be acceptable, given the likelihood of the event occurring, plus, taking credit for Operator actions, would minimize the impact of not meeting the Level 1 criterion. b) Known Sensitivities: The results of this event are sensitive to the decay heat (initial power level) and the initial water level (vessel inventory), which is influenced by the steam dryer pressure drop.
c) Uncertainties in Results: The calculated water level is adjusted downward by 1 foot to account for known
biases between the code calculation and actual plant tests. Instrument uncertainties dictate whether water level reaches Level 1 (i.e., nominal
trip setpoint (NTSP) versus Spurious Trip Avoidance). Conclusion a) Statement of Acceptability: The safety criterion is satisfied. As discussed above, the operational criterion may
not be satisfied. b) Known Conservatisms/Margins: A 2% overpower allowance is added (initial power is 102% of rated). RCIC flowrate is assumed to only be 98% of rated. No credit for CRD flow to the vessel.
Decay Heat based upon ANS 5.1 (1979) + 10%. c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): This is a unique event, required by NUREG-0737 (TMI Action Items) and is not part of the normal reload transient analysis. It was re-
analyzed as part of Extended Power Uprate.
15.1.7.4 - Trip Of One Feedwater Water Pump
Description of Event
a) Initiator: UFSAR/DAEC-1 15.1-66 Revision 23 - 5/15 A failure in the feedwater control system (or other cause) trips one Feedwater pump, with the reactor at rated thermal power/core flow
conditions. b) Sequence of Events (NOT a time line): One Feedwater pump trips and begin to coastdown. Reactor vessel level begins to drop with the loss of feedwater. Recirculation pumps begin to runback to 45% speed upon the loss of one feedwater pump with vessel level at Level 4 (low alarm). The recirculation pump runback helps initially moderate
the level drop. Feedwater level control opens the Feedwater Regulating Valves (FRV) to attempt to compensate for the lowering level. Water level in the downcomer region doesn't recover fast enough and a reactor scram occurs when level reaches the Level 3 trip point. The scram also
contributes to the level decrease due to void collapse. The level most-likely will reach the Level 2 trips (HPCI and
RCIC initiation) and they will start to inject in about 30 seconds. Water level inside the vessel begins to recover
with this additional injected flow. The Operators would
take over and guide the plant to a stable condition and eventually to cold shutdown, using normal plant
procedures. c) Single Failure/Operator Error (as applicable): None d) Key Equipment Responses (trips/actuations) & Operator Actions (successes & failures): Recirculation pump runback to 45% speed (only one Feedpump running), RPS trip on low level (Level 3), Control Rod Drive (Scram), HPCI and
RCIC initiation at low-low level (Level 2). Event Category & Acceptance Criteria: This is an Anticipated Operational Transi ent - an expected operational transient of moderate frequency (single Feed pump trip), with no additional single failures or Operator errors.
Because this is a unique analysis, performed as part of the Extended Power Uprate program, there are only operational, acceptance criteria applied. They are based upon trip avoidance - both Level 3 (Scram) and Level 2 (HPCI and RCIC initiation). Methods
a) Calculation Tools & Computer Codes: Primary Code - ODYN (see Table 15.0-2 for complete listing, code versions and
NRC acceptance). UFSAR/DAEC-1 15.1-67 Revision 23 - 5/15 b) Inputs (Reference common list in 15.0, and/or include event-specific items): OPL-3 (See Table 15.0-3) Feedwater Controller settings for a lead time constant of 0.5 and the lag time constant is 5.0, with proportional gain at 110% and Integral gain at 0.2 resets/min. Recirculation System controller settings: runback rate of 12%/sec, with proportional gain at 333% and Integral gain at 10 resets/min. c) Key Assumptions: Power and core flow are at rated conditions (100%). Initial water level is the normal operating water level. The tripped Feed pump linearly ramps to zero flow in 3 seconds. UFSAR/DAEC-1 15.1-68 Revision 23 - 5/15 Results a) Comparison to Acceptance Criteria: For the first operational criterion, the water level decreases to well below the Level 3 trip point and the Scram is not avoided. For the second criterion, there is some probability that the level will remain above the Level 2 nominal trip setpoint (NTSP), but the
Spurious Trip Avoidance (STA) setting is exceeded by several inches. Thus, conservatively, we assume that this operational acceptance criterion is not met. This is found to
be acceptable taking credit for Operator actions, would minimize the impact of not meeting the Level 2 criterion. b) Known Sensitivities: As part of this study, a sensitivity case was run, increasing the recirculation pump runback to 35% to determine if enough margin to Level 2 could be obtained to provide sufficient confidence that the trip could be avoided. While margin to the
Level 2 NTSP was gained, it was not sufficient to avoid the STA point. Noteworthy is the fact that the settings on the Feedwater controllers regulating the FRVs have little impact on the event results.
The change in initial power level due to EPU causes the results to be slightly more severe than at pre-EPU conditions. However, the acceptance criterion for the pre-EPU case was also not met. c) Uncertainties in Results: Instrument uncertainties dictate whether water level reaches Level 2 (i.e., nominal
trip setpoint (NTSP) versus Spurious Trip Avoidance).
Conclusion
a) Statement of Acceptability: As discussed above, the operational criterion for Level 3 (Scram) avoidance is not met and the Level 2 avoidance (HPCI and RCIC initiation) may not be satisfied. However, the Operators' ability to manage the event is not unacceptably compromised. b) Known Conservatisms/Margins: The setpoint methodology has inherent conservatism in it as the results provide for 95% probability at 95% confidence statistical limits. c) Limiting or Non-Limiting Event (Reload - transients; DBA - accidents): UFSAR/DAEC-1 15.1-69 Revision 23 - 5/15 This is a unique event, which was re-analyzed as part of Extended Power Uprate and is not part of the normal reload transient analysis.
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4lIO..2ClQ'Iif"S\L/"---------.i.I§It&rall.1IIo)DesignBasisAccident,RecirculationSuctionLineBreak,LPCIInjectionValveFailureDUANEARNOLDENERGYCENTERIESUTILITIES,INC.UPDATEDFINALSAFETYANALYSISREPORTWaterLevelInsideShroudvsTimeLargeBreakLOCAFigure15.2-1"Revision17-10/03 -i....a:::>>rI....a:..-IuarIua>a:0...400ucwa:300,TIMEAFTERBREAK(s)400DesignBasisAccident,RecirculationSuctionLineBreak,LPCIInjectionValveFailureDUANEARNOLDENERGYCENTERIESUTILITIES,INC.UPDATEDFINALSAFETYANALYSISREPORTReactorPressurevsTimeLargeBreakLOCAFigure15.2-2Revision17-10/03 1HIGH-POWEREDAXIALPLANEUNCOVEREDBEGINNINGOFSPRAYCOOLING1.1OISETOFBOILINGTRANSITION2&00,,ii"iii,,i,i,,,,,,,ii*'j,,i,,;i,*HIGH-POWEREDAXIALPLANEREFLOODED&&Icc:::ta1&00Ii&&I...e.::t!!1000aaC...u500'"..-:-2000-a'1IIJJ,IIIIIIIII,,IIIIIIIIII,IJJ,I24610204080100200400TIMEAFTERBREAK(s)DesignBasisAccident,RecirculationSuctionLineBreak,LPCIInjectionValveFailureDUANEARNOLDENERGYCENTERIESUTILITIES,INC.UPDATEDFINALSAFETYANALYSISREPORTPeakCladdingTemperaturevsTimeLargeBreakLOCAFigure15.2-3..Revision17-10/03 \IMARYCONTAINMENTIJ(4)/....UFLOW.1RESTRICTORSMAINISOLATIONVALVESLINESBYPASSHEADERSTOPVALVESIiONTROLVALVESBYPASSr:VALVES(9)TURBINEBYPASSSYSTEMDUANEARNOLDENERGYCENTERIESUTILITIES,INC.UPDATEDFINALSAFETYANALYSISREPORTBreakLocationforMainSteamLineBreakAccidentFigure15.2-7Revision17-10/03 .'....t"'tN....CIenlD.......--x)lV3ij8tron<llH!101:1SSVRlNV100:lDUANEARNOLDENERGYCENTERIESUTILITIES,INC.UPDATEDFINALSAFETYANALYSISREPORTMainSteamLineBreakAccident,MassofCoolantLostthroughBreakFigure15.2-8Revision17-10/03 0o0000<tI<)N000......,...,.--IXl-lg<;toIUcr::::>IU<tcrcr::::>IU(/lll.(/lIUIUcrl-ll.00*<DI<)00I<)(/lIUI-::::>0z<tNIU0""----'__-..L__-'-__.......__""-_-.JL-_--"'__-I..__.......___eo-G':",N'=0oH'U!r------------------------,DUANEARNOLDENERGYCENTERIESUTILITIES,INC.UPDATEDFINALSAFETYANALYSISREPORTSecondaryContainmentPressureTemperatureResponsetoInstrumentLineBreakFigure15.2-9Revision17-10/03
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Spent Fuel Pool (SFP) cooling:HPCI Equipment AreaRCIC Equipment Area
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Battery Room
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Criterion 12-Suppression of re actor power oscillations.
+
Growth Rate Overshoot Oscillation Period LPRM FailuresOscillation Contours LPRM Assignments
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UFSAR/DAEC-1 Table 15.3-1 Inputs to ATWS Analysis T15.3-1 Revision 23 - 5/15 Parameter Value Dome Pressure 1025 psig Rated Core flow 49.0 Mlbm/hr Minimum Core flow at Rated Power 48.5 Mlbm / 99% of rated Rated Power 1912 MWt Rated Steam Flow 8.35 Mlbm/hr Feedwater temperature 431°F Fuel Exposure 200 MWD/T (BOC) 14,800 MWD/T (EOC) Initial Suppression Pool Liquid Volume 58900 ft 3 Initial Suppression Pool Temperature 90 °F Nominal Closure Time of MSIV 4.0 sec Relief Valve System Capacity No. of Valves 59.6% NBR Steam Flow at 1080 psig 6 Relief Valve Opening Analytic Setpoint Range 1154/1162/1164/1172/1187/1196 psig Relief Valve Closing Setpoint 1107/1115/1117/1124/1138/1148 psig Relief Valve Time Delay On Opening Signal 0.2 sec Relief Valve Opening Duration 0.2 sec Relief Valve Closure Time Delay 0.2 sec Relief Valve Closure Duration 0.2 sec Safety Valve System Capacity No. of Valves 15.4% NBR Steam Flow at 1240 psig 2 Safety Valve Opening Analytic Setpoint 1277 psig Safety Valve Closing Setpoint 1226 psig Relief Valve Opening Duration 0.2 sec Relief Valve Closure Duration 0.2 sec Vessel Pressure Pump Trip Sensor Time Constant 0.0 sec Total SRV and SSV Capacity 73% NBR Steam Flow at 1080 psig UFSAR/DAEC-1 Table 15.3-1 Inputs to ATWS Analysis T15.3-2 Revision 23 - 5/15 Parameter Value Low-Low Set Setpoint Open 1025/1030 psig* Low-Low Set Setpoint Close 915/920 psig* Recirc Pump Trip Delay 0.175 sec SLCS Injection Location Lower Plenum Standpipe Number of SLCS Pumps 2 SLCS Injection Rate per Pump 26.2 gpm Nominal Boron-10 Enrichment 19.8 % Sodium Pentaborate Concentration 11.8 wt % Boron Injection Initiation Temperature (BIIT) 110 o F SLCS Liquid Transport Time 60 sec SLCS Liquid Solution Enthalpy 48.09 Btu/lbm RCIC Flow Rate 400 gpm Enthalpy of the RCIC Flow 68.04 Btu/lbm HPCI Flow Rate 3000 gpm Enthalpy of the HPCI Flow 68.04 Btu/lbm ATWS High Pressure Setpoint (Analytical Limit) 1168.6 psig Low Pressure Isolation Setpoint 850 psig Number of RHR Loops 2 RHR Service Water Temperature 85 o F RHR Heat Exchanger K-Factor per Loop in Containment Cooling Mode 142 Btu/sec-o F RHR Heat Exchanger K-Factor per Loop during the Loss of Offsite Power Event 135 Btu/sec-o F *Subsequent to the EPU analysis, the LLS NTSP values were revised as follows:
LLS SRV Opening min / max (NTSP) Closing min / max (NTSP) PSV-4401 1030 910 PSV-4407 1035 915
There is no impact on any of the transient or accident analysis results, as those analyses use the Analytical Limits, which were not revised. See footnote to Table 15.3-1 for evaluation of impact on the ATWS analyses. UFSAR/DAEC-1 T15.3-3 Revision 23 - 5/15 Table 15.3-2 Page 1 of 2 SBO Safe Shutdown Systems and Safety Actions
SCRAM Equipment
- 1. RPS logic (fail safe action)
- 2. CRD Drives
- 3. HCU solenoids
- 4. HCU accumulators
- 5. N2 supply (recovery only)
Pressure Relief Equipment
- 1. ADS including SRV
- 2. LLS including SRV
- 3. N2 accumulators
- 4. N2 supply (recovery only)
- 5. 125vdc Div I or II (valve controls ,sensors,logic,valve solenoids, low-low set logic)
Core Cooling Equipment
- 1. HPCI
- 2. 250vdc (valves, motors)
- 3. 125vdc Div II (valve controls, se nsors, logic, valve solenoids) 4. Condensate inventory an d condensate storage tank 5. Torus water (extended event only)
- 6. Auxiliary 480vac power (recovery only)
- 7. Emergency service water (recovery only) 8. RCIC
- 9. 125vdc Div I (valves, valve controls , sensors, logic, valve solenoids) 10. Low pressure systems (recovery only)
Primary Nuclear System Isolation Equipment
- 1. MSIV logic (fail safe action)
- 2. 125vdc Div I and II (valve controls , sensors, logic, valve solenoids)
Auxiliary AC Power Equipment (Recovery only)
- 1. Offsite & Onsite AC Generator System (battery charging, N2 supply, equipment cooling, area coo ling, decay heat removal) 2. 125vdc Div I and II (breaker control)
125vdc and 250vdc Power Equipment
- 1. 125vdc Div I & II (scram, LLS, HPCI , RCIC, PCIS, Auxiliary AC power -
valves, valve controls, sensors, logic, valve solenoids, breaker control, inverters) 2. 250vdc (valves, motors, inverters) UFSAR/DAEC-1 T15.3-4 Revision 23 - 5/15 Table 15.3-2 Page 2 of 2 SBO Safe Shutdown Systems and Safety Actions
Primary Containment Integrity Equipment
- 1. PCIS logic (fail safe action)
- 2. 125vdc Div I and II (valve controls , sensors, logic, valve solenoids)
Fuel Pool Cooling Equipment (Recovery only)
- 1. 480vac auxiliary power
Equipment Room Cooling Equipment (Recovery only) 1. 480vac auxiliary power
UFSAR/DAEC - 1 T15.3-5 Revision 23 - 5/15 Table 15.3 - 3 This Table has been deleted. 2012-020 UFSAR/DAEC - 1 T15.3-6 Revision 23 - 5/15 Table 15.3 - 4 Parameters for Hot Bundle Oscillation Magnitude Reload Review Evaluation Parameter Description Acceptance Criterion (Range) Base ValueC21 Value for Current Cycle Data Source Disposition (OK/Not OK) #LPRMs Number of installed LPRMs No change from base value 80 80 FRED Ok APRM assignment LPRM assignment to APRMs in 6 channels, etc. No APRM design change See EPU OPL-3 Same as C21 FRED Ok APRM trip @ NC Flow-biased APRM trip power level (nominal value) at natural circulation base value 65.4% rated power 65.4% rated power FRED OK Reactor Power @ NC Average power level on the rated licensing procedure flow-control line at natural circulation base value 47.2% rated power 47.2% rated power FRED Ok APRM trip @ 45% Rated Core Flow Flow-biased APRM trip power level (nominal value) at 45% Rated Core Flow base value 88.5% rated power 88.5% rated power FRED OK Reactor Power @ 45% Rated Core Flow Average power level on the rated licensing procedure flow-control line at 45% Rated Core Flow base value 60.2% rated power 60.2% rated power FRED Ok Tdelay Total delay time (APRM response time, RPS processing time, delay before startof control rod motion, plus time for 2 feet of control rod insertion) base value 885 msec 885 msec FRED Ok 2012-020 2012-020 2015-004}}