RS-17-126, Quad Cities Nuclear Power Station, Units 1 & 2, Revision 14 to Updated Final Safety Analysis Report, Chapter 15, Accident and Transient Analysis
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QUAD CITIES - UFSAR 15.0-1 Revision 14, October 2017 15.0 ACCIDENT AND TRANSIENT ANALYSES The evaluation of the safety of a nuclear power plant includes analyses of the response of
the plant to postulated disturbances in proce ss variables and to postulated malfunctions or failures of equipment. These safety analyses provide a significant contribution to the
design and operation of components and system s from the standpoint of public health and safety. [15.0-1] In previous chapters, the important structu res, systems, and components have been discussed. Chapter 4 describes the reacto r and its' analyzed operational conditions, including provisions for Maximum Extended Load Line Limit (MELLLA), Increased Core Flow (ICF), and Equipment Out-of-Service (E OOS). The following EOOS conditions are analyzed or evaluated for impact on thermal limits for Westinghouse reloads (Unit 2) and the results are reported for the applicable transient analyses: feedwater temperature reduction, TBV out of service, one SRV out of service, single loop operation, TCV slow closure, PLU out of service, pressure regulator out of service, one TCV stuck closed, one TSV stuck closed, and one MSIV out of service (a t 75% power). Additio nal details related to the combination of equipment OOS options can be found in cycle specific reload documentation. The thermal limits associated with implementation of each EOOS options are provided in the Core Operating Limits Report.
For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), supported EOOS conditions are found in the current cycle's CO LR and are supported by the applicable cycle-specific reload documents.
The EOOS conditions have been analyzed for the impact on fuel thermal limits and the design basis of the fuel. The transient analysis does not evaluate the effect of the EOOS condition on the design basis of the system.
EOOS Options, as provided in the cycle and unit specific Core Operating Limits Report, ar e only implemented for temporary conditions where equipment is operated in a degraded co ndition. In these instances, the operability process for degraded / non-conforming conditio ns is followed, with the EOOS fuel thermal limits being compensating actions. The E OOS options do not support permanent plant modifications, procedure revisions, or other permanent changes to the facility.
In this chapter, the effects of anticipated process disturbances and postulated component failures are examined to determine their cons equences and to evaluate the capability built into the plant to control or accommodate such failures and situations (or to identify the limitations of expected performance).
The situations analyzed include anticipated oper ational occurrences (e.g., a loss of electrical load), unexpected operational occurrences, and postulated accidents of low probability (e.g., the sudden loss of integrity of a major component). The analyses include an assessment of
the consequences of an assumed fission produ ct release that would result in potential hazards not exceeded by those from any accident considered credible.
15.0.1 Frequency Classification
The effects of various postulated anticipated operational occurrences (AOOs) and accident events are investigated for a variety of plant conditions. Some of the events have been categorized into three groups according to frequency of occurrence. The frequency classifications are as following.
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QUAD CITIES - UFSAR Revision 14, October 2017 15.0-1a A. Incidents of moderate frequency - thes e are incidents that may occur with a frequency greater than once per 20 years for a particular plant. This event is referred to as an "anticipated (e xpected) operational occurrence." B. Infrequent incidents - these are incidents that may occur during the life of the particular plant (spanning once in 20 years to once in 100 years). This event is
referred to as an "abnormal (unexp ected) operational occurrence."
C. Limiting faults - these are incide nts that are not expected to occur but are postulated because their consequences ma y result in the release of significant amounts of radioactive material. This event is referred to as a "design basis (postulated) accident."
15.0.2 Transients and Accidents Analyzed
The core-wide anticipated operational occurre nces (AOOs) were analyzed to support the extended power uprate (EPU) conditions (including the MELLLA domain) and the incorporation of the APRM rod block monitor technical specifications (ARTS) power and flow dependent limits improvement program. Th is included re-evaluating a broad set of the most limiting transient events at EPU conditions. The basis of the selection of the
transient events for re-analysis is documented in Reference 1. The transient events which are re-analyzed with power uprate condit ions from 2511 MWt to 2957 MWt core thermal power are documented in Reference 2.
The existing licensing bases for Unit 2 were reviewed by Westinghouse to determine the potentially limiting analyses that must be done on a cycle-specific basis or on a one-time basis to support the introduction of SVEA-96 Opti ma2 fuel. The basis for the selection of the limiting events is discussed in the Westinghouse reload licensing methodology basis
document for Quad Cities (Reference 5). A summ ary of the results of the events that are re-analyzed is documented in the cycle-specific re load licensing report. The parameters used for the transient analysis are documented in the OPL-W. For the reload specific AOO safety analyses performed by Westinghouse, a reactor scram occurs on other RPS trip signals prior to the low reactor vesse l water level (L3) being reached.
The existing licensing bases for Unit 1 were reviewed by AREVA to determine the potentially limiting analyses that need to be pe rformed on a cycle-specific basis or on a one-time basis to support the introduction of ATRIUM 10XM fuel. A summary of the results of the events that are re-analyzed is do cumented in the cycle-specific design Reload Safety Analysis Report (RSAR). The plant para meters used for the cycle-specific transient and safety analyses are documented in the plant parameters document. The AREVA overall reload licensing methodology is described in References 6 and 7.
15.0.2.1 Anticipated Operational Occurrences
Transients typically occur as a consequence of a single equipment failure or malfunction or single operator error. Such transients are evaluated in the sections listed below:
[15.0-3]
QUAD CITIES - UFSAR Revision 5, June 1999 15.0-2 Analysis Section A. Startup of idle recirculation loop at incorrect temperature 15.4.4 (Cold recirculation loop)
B. Single and multiple recirculation pump trips 15.3.1
C. Inadvertent opening of a safety valve, relief valve or safety 15.6.1 Relief valve
D. Recirculation flow controller failure (malfunction) - 15.4.5 Increasing flow
E. Recirculation flow controller failure (malfunction), 15.3.2 zero speed demand - decreasing flow
F. Inadvertent actuation of high pressure coolant injection 15.5.1 (HPCI) during power operation
G. Inadvertent closure of main steam line isolation valves 15.2.4 (MSIVs)
H. Loss of normal feedwater flow (Feedwater controller 15.2.7 Malfunction) - zero flow
I. Increase in feedwater flow (Feedwater controller malfunction) - 15.1.2 max. flow
J. Turbine trip with failure of bypass system 5.2.2.2.2, 15.2.3.1
K. Turbine trip with partial bypass - max. power 15.2.3.2
L. Turbine pressure regulator malfunction:
- 1. Increase in steam flow 15.1.3
- 2. Decrease in steam flow 15.2.1
M. Loss of main condenser vacuum 15.2.5
N. Load rejection with bypass (Loss of electrical load) 15.2.2.2
O. Loss of auxiliary power 8.3.1
P. Instrument air failure 9.3.1.1
Q. Failure of one diesel generator to start 8.3.1.6.4
R. Power bus voltage 8.3.1
QUAD CITIES - UFSAR Revision 14, October 2017 15.0-3 S. Load rejection without bypass 15.2.2.1
T. Decrease in feedwater temperature (loss of feedwater heating) 15.1.1
U. Rod withdrawal error 15.4.2
V. Thermal Hydraulic Instability 15.4.11
W. Loss of Stator Cooling 15.2.8
Some of the transients listed above are evaluat ed on a cycle specific basis. The nominal reactor operating pressure is approximately 1005 psig. Transient analyses typically use the nominal reactor operating pressure as an input to the analyses. Small deviations (5 to 10 psi) from the nominal pressure are not expe cted to change most of the transient analyses results. However, sensitivity studies for fast pressurization events (main turbine generator load rejection without bypass, turbine trip wi thout bypass, and feedwater controller failure) indicate that the delta-CPR may increase for lo wer initial pressures. Therefore, the fast pressurization events have considered a bo unding initial pressure based on a typical operating range to assure a conserva tive delta-CPR and operating limit.
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15.0.2.2 Design Basis Accidents
In order to evaluate the ability of the plant sa fety features to protect the public, a number of accidents are analyzed herein. These accide nts are of very low probability; however they are considered in order to include the far end of the operating spectrum of challenges to the safeguards and the containment system. The a ccidents in this chapter are discussed in the following sections:
[15.0-5] Analysis Section A. Control rod drop 15.4.10 B. Loss of coolant 15.6.2, 15.6.5 C. Main steam line break 15.6.4 D. One recirculation pump shaft seizure 15.3.3 E. Fuel handling accidents 15.7.2 F. Mislocated fuel assembly 15.4.7* G. Misoriented fuel assembly 15.4.8* H. Spent fuel cask drop 15.7.3
- Note: For Quad Cities Unit 1, the misl ocated and misoriented fuel assembly are characterized as infrequent events (inf requent incidents) in AREVA methodology. (Reference 6)
The analyses of design basis accidents provide expected maximum concentrations and
discharge rates of radioactive effluents, and ca lculated offsite doses for certain postulated events. Table 15.0-1 tabulates this information.
Further additional information is provided in the individual sections listed above relati ve to current analysis methods and results.
QUAD CITIES - UFSAR Revision 14, October 2017 15.0-3a 15.0.2.3 Radiological Assessments of Design Basis Accidents
[15.0-6] A chronology of different radiological assessme nts is given in UFSAR Section 15.6.5.5 for the loss-of-coolant accident.
The previous UFSAR licensing basis prior to extended power uprate utilized the TID-14844 methodology, which establishes the source term based on rated core thermal power. The power level used in the radiological assessment of design basis accidents is at 102% of the extended power uprate; i.e., 3016 MWt. Radiolog ical doses following the power uprate were developed by applying scaling factors to the pr evious doses. These scaling factors accounted for higher fuel burnup levels and updated fi ssion product inventories using the industry-accepted ORIGEN2 code, as discussed in Refe rence 2. Resultant impacts are discussed under the relevant sections.
In Reference 3, the NRC approved the use of Alternative Source Term (AST) for the evaluation of the onsite and offsite dose co nsequences for the following Design Basis Accidents: Loss of Coolant Accident (LOCA), Control Rod Drop Accident (CRDA), Fuel Handling Accident (FHA), and Main Steam Line Break (MSLB). The power level used in the radiological assessment of design basis accidents under AST is 102% of the extended power uprate thermal power limit (i.e., 2957MWt x 1.02 = 3016 MWt).
QUAD CITIES - UFSAR Revision 14, October 2017 15.0-4 The design basis accidents assessed in the UFSAR which have a radiological release that is proportional to the core radio nuclide inventory are the following:
A. Control Rod Drop (Section 15.4.10)
B. Loss-of-Coolant Accidents Resulting fr om Piping Breaks Inside Containment (Section 15.6.5)
C. Design Basis Fuel Handling Accidents In side Containment and Spent Fuel Storage Buildings (Section 15.7.2)
The radiological assessments of the above de sign basis accidents as described in the referenced UFSAR sections utilize the core ra dionuclide inventory for Westinghouse Optima2 fuel and AREVA ATRIUM 10XM fuel. Both We stinghouse Optima2 core inventory and AREVA ATRIUM 10XM core inventory have been ev aluated for a core average burnup of 39 GWD/ MTU.
The design basis accidents assessed in the UF SAR which do not have a radiological release that is proportional to the core ra dionuclide inventory are the following:
A. Main Steam Line Break (Section 15.6.4)
B. Instrument Line Break (Section 15.6.2)
C. Loss of Feedwater Flow (Section 15.2.7)
The specific activity of the primary coolant is lim ited by Technical Specification. In addition, there is no core uncovery and no perforations of the fuel during a main steam line break, instrument line break, or loss of feedwater flow.
Therefore, since only the coolant activity is released, the radiological dose calculations are independent of fuel type or design.
QUAD CITIES - UFSAR Revision 14, October 2017 15.0-5 15.0.2.4 References
- 1. "Licensing Topical Report Generic Guidelines for General Electric BWR Extended Power Uprate," NEDC-32424P-A, Appendix E, February 1999.
- 2. "Safety Analysis Report for Quad Cities 1 & 2 Extended Power Uprate," NEDC-32961P, Revision 2, August 2001.
- 3. Letter from M. Banerjee (U. S. NRC) to C.
Crane (Exelon Corporation), Dresden Nuclear Power Station, Units 2 and 3, and Quad Citi es Nuclear Power Station, Units 1 and 2 -
Issuance of Amendments Re: Adoption of Al ternative Source Term Methodology," dated September 11, 2006 [SER correction letter: D.
Collins (U. S. NRC) to C. Crane (Exelon Corporation), "Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Correction of Safety Evaluation for Amendment Dated September 11, 2006," dated September 28, 2006].
- 4. Deleted. 5. "Westinghouse BWR Reload Licensing Meth odology Basis for Exelon Generation Company Quad Cities Nuclear Power Stat ion Units 1 and 2," WCAP-16334-P, June 2005.
- 6. XN-NF-80-19(P)(A) Volume 4 Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.
- 7. ANP-3338P Revision 1, "Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Ex tended Power Uprate," AREVA, August 2015.
(Sheet 1 of 1)
Revision 7, January 2003 QUAD CITIES - UFSAR
Table 15.0-1
SUMMARY
OF MAXIMUM OFFSITE DOSES FROM POSTULATED ACCIDENTS (Original analysis, retained for historical purpose)
Maximum Total Offsite Exposure - Rads Accident Whole Body Thyroid Rod drop 6.2 x 10 4 curies noble gases 1.8 curies halogens released to
condenser 1.2 x 10-2 1.2 x 10
-3 Fuel loading 5.7 x 10 3 curies noble gases 3.5 x 10 3 curies halogens released to reactor water 5.9 x 10-3 4.1 x 10
-3 Steamline rupture 5.4 curies noble gases 116 curies (principally) halogens
released to reactor water 4.1 x 10-3 5.2 x 10
-1 Loss-of-coolant 5.2 x 10 5 curies noble gases 2.7 x 10 4 curies halogens airborne in primary containment at
30 minutes 5.3 x 10-4 1.3 x 10
-4 QUAD CITIES - UFSAR 15.1-1 Revision 11, October 2011 15.1 INCREASE IN HEAT REMOVAL BY THE REACTOR COOLANT SYSTEM Events described in this section that result in decreased feedwater temperature may also result in a core thermal hydraulic instabilit y transient. Refer to Section 15.4.11 for an overview of this event.
This section covers transients which involv e an unplanned increase in heat removal from the reactor due to conditions or events in th e reactor coolant system that are expected to occur with moderate frequency.
Excessive heat removal, i.e
., heat removal at a rate in excess of the heat generation rate in the co re, causes a decrease in moderator temperature which increases core reactivity and can lead to an increase in power level and a decrease in shutdown margin. The power level increase, if sufficient, would be terminated by a reactor scram. Any unplanned power level increase, however, has the potential to cause fuel damage or excessive reactor coolant system pressure, and warrants analysis if expected with moderate frequency.
[15.1-1]
The following design basis transients are covered in this section:
A. Feedwater system malfunctions that result in a decrease in final feedwater temperature;
B. Feedwater system malfunctions that result in an increase in feedwater flow; and
C. Steam pressure regulator malfunctions that result in increased steam flow.
These events, including the associated assumptions and conclusions, continue to be part of
the plant's licensing basis. The conclusions of these analyses are still valid; however, specific details contained in the descriptions and associated figures should be used only to understand the analysis and its conclusions. Re fer to the cycle reload licensing documents for cycle specific analyses performed.
For plant operation under extended power upra te (EPU) conditions, the limiting events (in terms of minimum critical power ratio (MCPR
)) for an increase in heat removal by the reactor coolant system were found to be the loss of feedwater heating (LFWH) and feedwater controller failure (FWCF). The inad vertent HPCI event would also decrease the core coolant temperature similar to LFWH. Re fer to Section 15.5 for discussion on analysis of the inadvertent HPCI event.
15.1.1 Decrease in Feedwater Temperature
Refer to the cycle reload licensing document s for the cycle-specific analysis performed.
Decrease in feedwater temperature due to loss of feedwater heating would result in core power increase due to the increase in core inle t subcooling and the reactivity effects of the corresponding increase in moderator density.
[15.1-2]
15.1.1.1 Identification of Causes and Frequency Classification
Feedwater heating can be lost in at least two ways:
- 1. Steam extraction line to heater is closed, or
- 2. Feedwater is bypassed around heater.
QUAD CITIES - UFSAR 15.1-2 Revision 14, October 2017 The first case would produce a gradual cooling of the feedwater. In the second case the feedwater would bypass the heater and the re duction of heating would occur during the stroke time of the bypass valve, a faster even
- t. In either case the reactor vessel would receive feedwater that is cooler than normal. The maximum number of feedwater heaters which can be tripped or bypassed by a single ev ent represents the most severe transient for analysis considerations. For Quad Cities, the sl ower event would result in a final feedwater temperature decrease of 145 o F. A loss of feedwater heating capability would cause an increase in core inlet subcooling. This wo uld increase core power due to the negative moderator temperature and moderator void reactivity coefficients. In automatic recirculation flow control mode some compensa tion of core power would be realized by modulation of core flow.
For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), the fast event is a 74.4°F decrease in feedwater temperature over a time period of 58.6 seconds. The slow event supports a larger 145°F decrease in f eedwater temperature over a time period greater than 80 seconds. The fast 74.4 °F de crease in feedwater temperature scenario is analyzed using the COTRANSA2/XCOBRA/XCOBRA-T code package. The longer term event supporting a 145°F decrease in feedwater temperature is evaluated using the Reference 1 methodology.
For Westinghouse reload cores (Unit 2), the modeling approach for this event is outlined in the cycle specific reload analysis reports.
This incident is analyzed as having moderate frequency.
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15.1.1.2 Sequence of Events and System Operation
The following plant operating conditions and assumptions form the principal basis for which reactor behavior is analyzed during the loss of feedwater heating transient:
[15.1-4]
A. The plant is operating at full power; and
B. The plant is operating in the manual recirculation flow control mode. (Automatic Flow Control is no longer used.)
For this event power would increase at a very moderate rate, and the operator would be expected to insert control rods as necessary to stay within the analyzed power-to-flow region. If this were not done the core power could exceed the scram setpoint and a scram would then occur.
15.1.1.3 Barrier Performance
The fuel-specific minimum critical power rati o (MCPR) limiting condition for operation (LCO) is determined for each reload core bas ed on bounding events for the cycle. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit.
Refer to the cycle-specific documentation or Core Operating Limits Report (COLR) for detailed results of the current cycle transient analyses.
QUAD CITIES - UFSAR 15.1-2a Revision 14, October 2017 15.1.1.4 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore, a radiological consequence analysis has not been performed.
15.1.2 Increase in Feedwater Flow
Refer to the cycle reload licensing docume nts for cycle-specific analyses performed.
QUAD CITIES - UFSAR 15.1-3 Revision 8, October 2005 15.1.2.1 Identification of Causes and Frequency Classification This event is postulated on the basis of a sing le failure of a control device, specifically one which can directly cause an increase in coolant inventory by increasing feedwater flow. The most severe applicable event is a feedwate r controller failure during maximum flow demand. The feedwater controller is forced to it s upper limit at the beginning of the event.
[15.1-5]
This is considered an incident of moderate frequency.
[15.1-6]
15.1.2.2 Sequence of Events and System Operation
The operating conditions and assumptions considered in this analysis are as follows:
[15.1-7]
A. Feedwater controller fails during maximum flow demand;
B. Maximum feedwater pump runout occurs; and
C. The reactor is operating in the m anual recirculation flow control mode, which provides for the most severe transient.
A feedwater controller failure under these circumstances would produce the following sequence of events:
A. The reactor vessel receives an excess of feedwater flow;
B. This excess flow results in an increase in core subcooling, which results in a rise in core power, and an increase in reactor vessel water level; and
C. The rise in the reactor vessel water le vel eventually leads to a high water level turbine trip and a feedwater pump trip, and results in a reactor scram.
Under most conditions, no operator action would be required. Th e reactor would scram following the turbine trip on high water leve l and end the transient. The operator would verify that a feedwater pump trip had occurred to terminate the initiating condition. The analysis was initiated from a typical low powe r condition with reactor power and flow at various points along the APRM rod block and minimum pump speed lines, including ICF to
108% of rated flow. In the analysis of this event, operation in the manual recirculation flow control is considered. (Automatic recircula tion flow control is no longer used.)
15.1.2.3 Core and System Performance
An excess feedwater flow transient due to a maximum feedwater demand by the feedwater controller was evaluated for the initial core and is shown in Figures 15.1-1 and 15.1-2. The
low initial power level resulted in a more se vere steam/feed flow mismatch and reactor water level transient.
[15.1-8]
QUAD CITIES - UFSAR Revision 10, October 2009 15.1-4 The increase in feedwater flow due to a failure of the feedwater control system to maximum demand results in an increase in the water level and a decrease in the coolant temperature
at the core inlet. The increase in core inlet subcooling causes an increase in core power. As the feedwater flow continues at the maximum de mand, the water level will continue to rise and eventually will reach the high water leve l trip set point. The high water level trip causes the turbine stop valves to close to prevent damage to the turbine from excessive liquid inventory in the steam line. The turbine stop valve closure creates a compression
wave that travels to the core causing a void collapse and subsequent rapid power excursion.
The closure of the turbine stop valves initiates a reactor scram. As vessel pressure
increases, the turbine bypass valves and relief valves provide pressure relief. The core
power excursion is mitigated in part by the pressure relief, but the primary mechanisms for termination of the event are reactor scram and re voiding of the core. At power levels below 38.5% of rated, a reactor scram on turbine stop valve closure is bypassed. However, the reactor scram on high neutron flux and high pr essure are still available. This transient is analyzed at numerous statepoint s on a cycle-specific basis.
The FWCF event is also analyzed assuming turbine bypass valves are not available to provide a basis for this possible mode of op eration. The thermal limits and limitations associated with implementation of this EOOS option are provided in the Core Operating Limits Report.
15.1.2.4 Barrier Performance
The fuel-specific minimum critical power rati o (MCPR) limiting condition for operation (LCO) is determined for each reload core bas ed on bounding events for the cycle. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit.
Cycle-specific results are described in the relo ad licensing documents or the Core Operating Limits Report.
15.1.2.5 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore a radiological consequence analysis has not been performed.
15.1.3 Increase in Steam Flow
See the introduction to Section 15.1 for inform ation regarding use of details from this analysis description which may not be a pplicable to the current fuel cycle.
15.1.3.1 Identification of Causes
This event is postulated on the failure of the turbine pressure regulator in the valve-open
direction. The maximum control-plus-bypass valv e demand is limited by the control system to the EHC control system maximum combined fl ow limit (MCFL) setpoint (Section 7.7.4.2).
[15.1-9]
QUAD CITIES - UFSAR Revision 14, October 2017 15.1-5 15.1.3.2 Core and System Performance
Figures 15.1-3 and 15.1-4 show the results of an analysis of this malfunction at 2511 MWt.
Vessel and steam line pressures drop 100 psi in th e first 10 seconds. Core flux is decreased significantly as the pressure drop increases the moderator void fraction. When steam line
pressure decreases, closure of the main steam isolation valves is initiated by a Group I
isolation from steam line low pressure. A scram occurs when the isolation valves have
reached 10% closed (analytical limit). The depr essurization is stopped by the isolation and the reactor is shut down with pressure rising slowly. Pressure rise is limited by operation of the relief valves and the reactor core isolat ion cooling (RCIC) system, which is discussed in more detail in Section 5.4.
The sequence of events above continues to apply for operation under EPU conditions.
15.1.3.3 Barrier Performance
This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each reload core based on bound ing events for the cycle. The MCPR LCO is calculated to preclude violation of th e fuel cladding integrity safety limit.
15.1.3.4 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore a radiological consequence analysis has not been performed.
15.1.4 References
- 1. ANF-1358(P)(A) Revision 3, "The Loss of F eedwater heating Transient in Boiling Water Reactors," Framatome ANP, September 2005.
QUAD CITIES - UFSAR 15.2-1 Revision 14, October 2017 15.2 DECREASE IN HEAT REMOVAL BY THE REACTOR COOLANT SYSTEM The AREVA methodology is only applicable to Quad Cities Unit 1.
Some events described in this section have no t been reanalyzed for the current fuel cycle, because these events continue to be bounded by other events which are analyzed for the current fuel cycle. These events, including the associated assumptions and conclusions, continue to be part of the plant's licensing basis. The conclusions of these analyses are still
valid; however, specific details contained in the descriptions and associated figures should be used only to understand the analysis and its conclu sions. These specific details should not be used as sources of current fuel cycle design info rmation. Refer to the cycle-specific reload licensing documentation or the COLR, for detaile d results of current cycle transient analyses.
For operation at EPU conditions, the events re sulting in a decrease in heat removal by the reactor coolant system were analyzed. The events in this category are primarily represented in the EPU analysis guidelines by the turbine trip and load reject transient events with the assumed failure of the turbine steam bypass function. The feedwater controller failure (maximum demand) event also includes some aspects of this area, since it involves a turbine
trip (from high water level). Other pressuriza tion events analyzed include the MSIV closure with direct scram, load rejection with bypass, and a single MSIV closure. The loss of condenser vacuum is another type of turbine tri p with bypass and is bound by events without bypass operation. The loss of offsite AC power and loss of normal feedwater are similar
events. These events result in initial power decreases and are not limiting with respect to thermal limits.
For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), an evaluation was performed to determine the limiting events in this category. The results of the evaluation are presented in Reference 7. The evaluatio n determined that the potentially limiting events in this category are the load reje ction without bypass (LRNB), the turbine trip without bypass (TTNB) and loss of stator coo ling (LOSC). The other events are benign and/or the consequences are bound by those of another event. Only the potentially limiting events are evaluated on a cycle-specific basis.
15.2.1 Steam Pressure Regulator Malfunction
15.2.1.1 Identification of Causes
For this event, the turbine pressure regulato r is assumed to fail low (i.e., zero output).
[15.2-1]
15.2.1.2 Sequence of Events and System Operation If one of the three processors in the pressure controller failed low, the pressure controller would maintain control of the turbine valves with no change in pressure. If either one or two
of the three pressure transmitters providing in put to the pressure controller failed low, the turbine control valves would adjust to the pre ssure sensed by the functioning transmitter and a small change in pressure could occur.
If a second processor or the third transmitter fa iled low, the turbine control valves will close resulting in an increase in reactor pressure.
QUAD CITIES - UFSAR 15.2-1a Revision 14, October 2017 15.2.1.3 Core and System Performance If one of the three pressure transmitters and two of the three processors in the pressure controller remained functional, the transient would be similar to a pressure setpoint increase as shown in Section 4.3.2.3.4.4.
If a second processor or the third transmitter fa iled low, the turbine control valves will close resulting in an increase in reactor pressure leading to a reactor scram on high flux.
15.2.1.4 Barrier Performance
This transient is not analyzed for reload co res since the fuel-specific minimum critical power ratio (MCPR) limiting conditions for oper ation (LCO) is determined for each reload core based on other events which bound this event. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit.
[15.2-2]
The failure of the pressure controller is bo unded by the analysis performed for the power load unbalance out-of-service in the equipment out of service report (Westinghouse Reload Licensing report applicable to Unit 2).
For AREVA reload cores (starting with Quad Ci ties Unit 1 Cycle 25), the steam pressure regulator malfunction is considered a benign event (Reference 7).
15.2.1.5 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore, a radiological consequence analysis was not performed.
QUAD CITIES - UFSAR 15.2-2 Revision 14, October 2017 15.2.2 Load Rejection
A power/load imbalance system is provided wh ich senses the generator load and makes a comparison to the thermal power (as indicated by intermediate steam pressure). When a mismatch in excess of 40% occurs the power/load imbalance relay will energize the fast
acting solenoid valves on the turbine control valves. This results in a turbine control valve
fast closure and a subsequent reactor scram.
[15.2-3]
15.2.2.1 Load Rejection (Generator Trip) Wi thout Bypass (LRNB) (Evaluated Each Cycle in the Reload Licensing Documents)
15.2.2.1.1 Identification of Causes and Frequency Classification
The following plant operating conditions and assumptions form the principal bases for which the reactor transient is anal yzed during a load rejection.
A. The reactor and turbine generator are initially operating at full power when the load rejection occurs.
B. All of the plant control systems continue normal operation.
C. Auxiliary power is continuo usly supplied at rated frequency.
D. The reactor is operating in the manual flow control mode when load rejection occurs. (Automatic Flow Control is no longer used.)
E. The turbine bypass valve system is failed in the closed position.
The LRNB transient is classified as a modera te frequency event. MCPR limits are defined such that the MCPR Fuel Cladding Integrit y Safety Limit is not violated during the occurrence of this transient.
[15.2-4]
This transient is a potentially limiting event re quiring analysis on a cycle-specific basis to verify or establish operating limits. The resu lts are contained in the cycle-specific reload licensing documents or the COLR.
The LRNB event is also analyzed or evaluated for EOOS options with the thermal limits associated with implementation of each EOOS option are provided in the Core Operating Limits Report.
15.2.2.1.2 Sequence of Events and System Operation
Complete loss of the generator load prod uces the following sequence of events:
[15.2-5]
A. The power/load imbalance actuation step s the load reference signal to zero and closes the turbine control valves at the earliest possible time. The turbine
accelerates at a maximum rate until the valves start to close. The turbine
control valves will close at a rate of 0.150 seconds for the full valve stroke.
B. Reactor scram is initiated upon sensing control valve fast closure.
QUAD CITIES - UFSAR 15.2-3 Revision 14, October 2017 C. If the pressure rises to the pressure re lief setpoint, some or all of the relief valves open, discharging steam to the suppression pool.
D. If the pressure rises to >1250 psig for Unit 2 or >1200 psig for Unit 1, the trip of recirculation pump drive motors occurs.
For assessing consequences of this event on thermal margin, however, the re circulation pump trip is conservatively assumed not to occur.
[15.2-6]
GE has identified in a 10 CFR Part 21 letter t hat at lower reactor power levels (above Pbypass), the Power Load Unbalance (PLU) devi ce may not actuate and the turbine control system will initiate turbine control valve clos ure at normal speed, which would not generate a direct scram (Reference 3). This would occur if the PLU is calibrated to actuate at power levels above Pbypass.
15.2.2.1.3 Core and System Performance
Fast closure of the turbine control valves would be initiated whenever electrical grid disturbances occur which result in significant loss of load on the generator. The turbine
control valves are required to close as rapi dly as possible to prevent overspeed of the turbine generator rotor. The closing would cause a sudden reduction of steam flow which results in a nuclear system pressure increase.
The reactor would be scrammed by the fast closure of the turbine control valves.
[15.2-7]
The reactor core isolation cooling (RCIC) and shutdown cooling mode of the RHR system would be initiated to handle lo ng-term decay heat removal.
The MCPR would not exceed the MCPR fuel cla dding integrity safety limit as determined by cycle-specific reload analysis.
[15.2-8] For the situation that turbine control valve fa st closure does not occur above Pbypass but below the power/load unbalance setpoint, the anal ysis accounting for how the plant actually behaves has been performed for the applicable fuel types. This analysis credits the
generator protection logic, which would initia te a turbine trip within 0.625 seconds of load rejection resulting in a turbine stop valve position scram. Based on the methodology described in Reference 5 and applicable to Un it 2, the Westinghouse Reload Licensing Report concludes that the equipment-in-servi ce thermal limits from the Core Operating Limits Report (COLR) are bounding for this ev ent. For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), the cycle-sp ecific, power-dependent AREVA reload analyses of the LRNB conservatively assume the PLU device does not cause the turbine control valve fast closure at 50% core thermal power. These results are used in determining the normal equipment-in-service power-de pendent thermal limits for the COLR.
15.2.2.1.4 Barrier Performance
The fuel-specific MCPR LCO is determined for each reload core based on bounding events for the cycle. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit. Refer to the cycle-sp ecific documentation or the Core Operating Limits Report (COLR) for detailed result s of current cycle transient analyses.
15.2.2.1.5 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore, a radiological consequence analysis was not performed.
QUAD CITIES - UFSAR 15.2-4 Revision 14, October 2017 15.2.2.2 Load Rejection With Bypass 15.2.2.2.1 Identification of Cause and Frequency Classification
The cause and frequency classification of this transient are the same as that for load rejection without bypass discussed in Section 15.2.2.1.
15.2.2.2.2 Sequence of Events and System Operation
A loss of generator load causes the turbine generator to overspeed. The turbine speed and
acceleration protection systems quickly close the turbine control valves to avoid excessive turbine overspeed. The control valves are fully closed in about 0.15 second following the initiation of the event. Above the PLU load setting, a scram is initiated by sensing the
turbine generator load imbalance and sending an electrical signal to the fast acting solenoid. This results in control valve closure and reactor scram.
[15.2-9]
15.2.2.2.3 Core and System Performance
The transient response of the unit to a ge nerator trip from 2511 MWt is shown for the initial core in Figures 15.2-1 and 15.2-2. The pr essure rise causes core voids to collapse and neutron flux reaches approximately 145% before the scram terminates the transient. The increased core pressure and saturation temper ature momentarily stores heat in the fuel causing the dip in average surface heat flux. The average heat flux never exceeds the initial value before it decays following the re actor scram. Coupled with the slight increase in core flow, this produces no decrease in MCPR.
The transient response is similar for operation under EPU conditions.
15.2.2.2.4 Barrier Performance
This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each reload core based on othe r events, e.g., LRNB, which bound this event.
The MCPR LCO is calculated to preclude viol ation of the fuel cladding integrity safety limit. [15.2-10] For AREVA reload cores (starting with Quad Ci ties Unit 1 Cycle 25), this analysis was evaluated at power levels below Pbypass.
It was shown that with a final feedwater temperature of > 55°F, this event is non-limit ing relative to establishing operating limits as outlined in the cycle specific Reload Report.
15.2.2.2.5 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore, a radiological consequence analysis was not performed.
15.2.3 Turbine Trip As discussed in Section 5.2.2.2.2, the relief valves are sized based on analysis of a turbine trip with simultaneous reactor scram while assuming failure of the turbine bypass QUAD CITIES - UFSAR 15.2-5 Revision 14, October 2017 system. Furthermore, as discussed in Section 5.2.2.2.3, the safety valves are sized based on analysis of a main steam isolation valve closure event without a direct scram. The turbine trip analyses without bypass and with bypass, presented in Sections 15.2.3.1 and 15.2.3.2, respectively, assume a reactor scra m due to turbine trip (stop valve closure).
[15.2-11]
15.2.3.1 Turbine Trip Without Bypass (TTNB) (Evaluated Each Cycle in the Reload Licensing Documents)
15.2.3.1.1 Identification of Causes and Frequency Classification
A variety of turbine or nuclear system malfunctions will initiate a turbine trip (see
Section 10.2.2).
[15.2-12]
This event is classified as a moderate frequency event.
15.2.3.1.2 Sequence of Events and Systems Operations
The sequence of events for a turbine trip woul d be similar to those for a generator load rejection. Position switches at the stop valv es would sense the valve closure and provide a reactor scram signal. If the pressure were to rise to the pressure relief setpoints the relief valves would open and discharge steam to the suppression pool.
15.2.3.1.3 Core and System Performance
The turbine stop valves would close as rapidly as possible. The closing would cause a
sudden reduction of steam flow which results in a nuclear system pressure increase. The reactor would be scrammed by the closure of the turbine stop valves.
[15.2-13]
The reactor core isolation cooling (RCIC) and shutdown cooling mode of the RHR system would be initiated to handle lo ng-term decay heat removal.
The MCPR limit would not exceed the fuel claddi ng integrity safety limit as determined by cycle-specific reload analysis.
15.2.3.1.4 Barrier Performance
The fuel-specific MCPR LCO is determined for each reload core based on bounding events for the cycle. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit. Refer to the cycle-sp ecific documentation or the Core Operating Limits Report (COLR) for detailed result s of current cycle transient analyses.
[15.2-14]
.
QUAD CITIES - UFSAR Revision 14, October 2017 15.2-6 15.2.3.1.5 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore, a radiological consequence analysis was not performed
15.2.3.2 Turbine Trip with Bypass
15.2.3.2.1 Identification of Causes and Frequency Classification
A turbine stop valve closure can be initiated by a variety of turbine or reactor system malfunctions (see section 10.2).
[15.2-15]
This event is classified as a moderate frequency event.
[15.2-16]
15.2.3.2.2 Sequence of Events and System Operation
The sudden closure of the stop valves would cause a rapid pressurization of the steam line
and reactor vessel with resultant void colla pse and power increase. The reactor would scram immediately from position switches mount ed on the stop valves (turbine trip scram).
Closure of the stop valves would also im mediately initiate bypass valve opening.
[15.2-17]
15.2.3.2.3 Core and System Performance
The resulting transient from 2511 MWt is shown in Figures 15.2-3 and 15.2-4 (based on
initial reload core). The bypass valves woul d limit the peak pressure rise at the relief valves to 1105 psig, 10 psi below the lowest re lief valve setpoint of 1115 psig (analytical limit). Vessel pressure would peak at 1106 psig. A neutron flux peak of about 245% occurs about 0.5 second after the trip.
Results for EPU conditions are similar to non-EPU and are bound by the Turbine Trip
Without Bypass Event.
15.2.3.2.4 Barrier Performance
This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each reload core based on othe r events, e.g., TTNB, which bound this event.
The MCPR LCO is calculated to preclude viol ation of the fuel cladding integrity safety limit. [15.2-18]
15.2.3.2.5 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore, a radiological consequence analysis was not performed.
QUAD CITIES - UFSAR Revision 14, October 2017 15.2-7 15.2.4 Inadvertent Closure of Main Steam Isolation Valves A full closure of all main steam isolation va lves (MSIVs) without direct (from position switch) scram and with no credit taken for the relief valves is used to evaluate the required capacity of the main steam safety valves. This analysis is included in Section 5.2.2.2.3.
[15.2-19]
A full closure of all MSIVs with direct scram and relief valve operation is described in this section. This event is classified as a moderate frequency event.
15.2.4.1 Identification of Causes
The inadvertent closure of the MSIVs may be caused by operator error.
[15.2-20]
15.2.4.2 Sequence of Events and System Operation
A MSIV closure can occur in 3 seconds. Rea ctor scram is initiated when the valves reach 10% closed.
15.2.4.3 Core and System Performance
The transient response to inadvertent closure of these valves from 2511 MWt is shown in
Figures 15.2-5 and 15.2-6 (initial core). No safety problems would be encountered. No significant neutron flux or surface heat flux peaks would be encountered since the first 10%
of valve stroke would not reduce valve flow area, and therefore MCPR would not go below the MCPR Safety Limit. The relief valves wo uld open to remove excess stored heat. The peak pressure at the safety valves would re ach only 1144 psig, well below the lowest safety valve setpoint of 1240 psig. The reactor core isolation cooling (RCIC) and shutdown cooling mode of the RHR system would be initiate d to handle long-term decay heat removal.
[15.2-21]
15.2.4.4 Barrier Performance
This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each reload core based on othe r events which bound this event. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit.
[15.2-22]
15.2.4.5 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore, a radiological consequence analysis was not performed.
QUAD CITIES - UFSAR Revision 14, October 2017 15.2-8 15.2.5 Loss of Condenser Vacuum
15.2.5.1 Identification of Causes
The main condenser vacuum is assumed to be suddenly lost while the unit is operating at
rated thermal power. This event is cla ssified as a moderate frequency event.
[15.2-23]
15.2.5.2 Sequence of Events and System Operation
The following would occur due to the loss of condenser vacuum: Alarm at 24 in.Hg vacuum Scram at 20 in.Hg vacuum Turbine stop valve closure at 20 in.Hg vacuum Turbine bypass valve closure at 7 in.Hg vacuum The worst case would occur if the loss of vacuum were instantaneous. In this event the
transient would become identical to the tu rbine trip with bypass failure discussed in Section 5.2.2.2. The relief valves would open to prevent safety valve operation.
15.2.5.3 Core and System Performance
The majority of the stored heat would be re moved by the relief valves and the RCIC system using the suppression pool as a heat sink (see Section 5.4). Slower losses of condenser
vacuum would produce less severe transients because the scram would precede the stop valve closure and some bypass flow to the main condenser would remove stored heat.
15.2.5.4 Barrier Performance
This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each reload core based on othe r events which bound this event. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit. The loss of condenser vacuum is bounded by Turbine Trip Without Bypass Event.
[15.2-24]
15.2.5.5 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore, a radiological consequence analysis was not performed.
QUAD CITIES - UFSAR Revision 11, October 2011 15.2-9 15.2.6 Loss of Offsite AC Power The onsite power systems provide power to vital loads in the event of a loss of auxiliary power from offsite sources. The consequences of a lo ss of offsite ac power are addressed in Section 8.3. This event is bounded by the load reject no bypass or turbine trip no bypass events.
15.2.7 Loss of Normal Feedwater Flow
15.2.7.1 Identification of Causes
A loss of feedwater transient response is a ssumed to occur due to a feedwater controller malfunction demanding closure of the feedwater control valves.
[15.2-25]
15.2.7.2 Sequence of Events and System Operation
With an initial power level of rated power, feed water control valves are assumed to close at their maximum rate. The unit response to simultaneous tripping of all feedwater pumps
would be very similar to the transient analyz ed. The reactor water level would decrease rapidly due to the mismatch between the steam flow out of the vessel and the shut-off
feedwater flow. Low water level scram wo uld occur after about 7.4 seconds.
The recirculation flow controlle r would reduce to minimum speed demand when the feedwater flow dropped below 2 x 10 6 lbs/hr. This interlock would pr otect the recirculation drive pumps from steady-state NPSH problems.
15.2.7.3 Core and System Performance
The transient response to this event is shown in Figures 15.2-7 and 15.2-8 (initial core at 2511 Mwt).
Based on the initial reload core, the decrease in moderator subcooling would slightly decrease the neutron flux until a scram occurred and comple tely shut down the reactor. Vessel steam flow would closely follow the decay of fuel surf ace heat flux. Analysis of the transient was discontinued at 16 seconds because the model was not programmed to handle the situation when core inlet subcooling becomes negative, i.e.
saturation at core inlet. Subsequent events would be a complete recirculation pump driv e motor trip and main steam isolation valve closure, both occurring when the water level dr ops to the low-low level setpoint. The time when this would occur, predicted from the est ablished rate of level decrease, is about 33.5 seconds. Pressure would rise following the is olation, and eventually actuation of the RCIC system and the RHR system (shutdown coolin g mode) would handle the long-term shutdown heat removal. This again would be less se vere than the turbine or generator trips.
Water inventory loss from 16 seconds until 36.5 se conds, the time the isolation valves would be closed, was conservatively estimated to be less than 550 ft 3 of saturated water. (At 16 seconds, vessel steam flow would be 45% of rate
- d. For extreme conservatism, this rate was considered to exist until 36.5 seconds.)
QUAD CITIES - UFSAR Revision 13, October 2015 15.2-10 Accounting for the conservative inventor y loss after 16 seconds and assuming the recirculation pumps would trip, an estimate of the final water level was made. In this analysis, all steam existing as carry-under and as voids in the core, upper plenum, standpipes, and separators at 16 seconds was allowed to condense. The volume of water delivered to the scram discharge volume was co nsidered to be removed from the vessel. Even neglecting the inventory makeup from the RCIC system, the calculations showed that more than 5 feet of water would remain above the core.
The loss of normal feedwater flow is bounded by more limiting events which are performed on a cycle-specific basis.
15.2.7.4 Barrier Performance
No thermal limits would be violated since the transient would be less severe than the turbine or generator trips. The fuel-specific MC PR LCO is determined for each reload core based on bounding events for the cycle. The MC PR LCO is calculated to preclude violation of the fuel cladding integrity safety limit. As noted above and in the introduction to Section 15.2, this event is not reanalyzed for reload cores because its' results are bounded by turbine or generator trips. This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each re load core based on other events which bound this event.
[15.2-26]
15.2.7.5 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore a radiological consequence analysis was not performed.
15.2.8 Loss of Stator Cooling To protect the generator from overheating from the consequences of a loss of stator cooling, protection logic initiates an automatic runback of the turbine. The Loss of Stator Cooling (LOSC) event is characterized by a very slow turbine control valve (TCV) closure. The slow valve closure does not scram the reactor directly.
As the steam flow capacity of the turbine is reduced turbine bypass valves open to con trol pressure. If the reactor power exceeds the turbine plus bypass capacity, an increase in re actor pressure will occur. Without sufficient turbine plus bypass capacity, the event is te rminated by an automatic reactor protection system (RPS) actuation on high reactor pressu re or high neutron flux (high APRM scram).
The analysis models the system response incl uding the runback rate and stopping point of the TCV closure and an error band is applie d to these parameters to account for uncertainties in the system response. The LO SC event was found to be bounded by other limiting events (e.g. Load Rejection Without Bypass and Feed Water Controller Failure).
QUAD CITIES - UFSAR Revision 14, October 2017 15.2-10a For AREVA reload cores (starting with Quad Ci ties Unit 1 Cycle 25), the LOSC event is a potentially limiting event at certain power levels and is analyzed on a cycle-specific basis (Reference 7). Cycle specific results c an be found in the applicable reload safety analysis report.
QUAD CITIES - UFSAR Revision 14, October 2017 15.2-11 15.2.8.1 Identification of Causes and Frequency Classification A loss of stator cooling signal can be genera ted in a number of ways, including low stator water flow, low stator water pressure, or a high stator water temperature signal. This event is classified as a moderate frequency event.
15.2.8.2 Sequence of Events and System Operation A loss of stator cooling would produc e the following sequence of events:
A. Load set begins to run back when a loss of stator cooling condition is sensed. The TCVs start to close slowly when the load set drops below actual load.
B. The bypass valves open to keep the steam flow and pressure constant.
C. A reduction of feedwater temperature occurs in response to the loss of turbine steam flow. D. Reactor pressure begins to increase when the steam production is greater than the combined capacity of the TCV and turbine bypass valves.
E. The TCVs continue to close until they reac h the stopping point defined within the Digital Electro-Hydraulic Control system.
F. Reactor scram initiates upon sensin g high pressure or high flux.
15.2.8.3 Core and System Performance
The turbine runback and very slow closure of the TCVs would be initiated whenever there is a loss of stator cooling. The runback of T CVs would cause a relatively slow increase in reactor pressure, once the combined capacity of the TCV and turbine bypass is less than the steam production in the reactor vessel. The increase in coolant pressure causes a
subsequent increase in reactor power. The rea ctor would scram on high pressure or possibly high flux.
15.2.8.4 Barrier Performance
As described in Reference 6 for Westinghouse relo ad core (Unit 2), the loss of stator cooling is not considered to be one of the limiting even ts for the fuel cycle. This LOSC transient is checked but typically not analyzed for reload co res, since the fuel specific operating limit minimum critical power ratio (MCPR) is determin ed for each reload core based on events that are more limiting than the LOSC event.
For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), the LOSC event is a potentially limiting event at certain power levels. See the cycle-specific reload safety analysis report (RSAR) for details on the AREVA Transient analysis results. The operating limit MCPR is established to preclude violation of the fuel cladding integrity safety limit.
QUAD CITIES - UFSAR 15.2-12 Revision 14, October 2017 15.2.8.5 Radiological Consequences Since the fuel cladding integrity safety lim it would not be violated, a radiological consequence analysis was not performed.
15.2.9 References
- 1. Deleted
- 2. Deleted
- 3. SC04-15, "Turbine Control System Impact in Transient Analyses," 10 CFR Part 21 Communication, October 31, 2004.
- 4. Deleted
- 5. "Westinghouse BWR Reload Licensing Meth odology Basis for Exelon Generation Company Quad Cities Nuclear Power St ation Units 1 and 2," WCAP-16334-P, June 2005. 6. "Evaluation of Loss of Stator Cooling for Dresden and Quad Cities," NF-BEX-12-139, Revision 1, January 30, 2014.
- 7. ANP-3565P Rev 0, "Quad Cities Unit 1 Cycl e 25 Reload Safety Analysis," AREVA Inc., February 2017.
QUAD CITIES - UFSAR 15.3-1 Revision 14, October 2017 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE The AREVA methodology is only applicable to Quad Cities Unit 1.
Events described in this section that result in re duced core flow rates may also result in a core thermal hydraulic instability transient. Refer to Section 15.4.11 for an overview of this event.
The AOO events in this category are not limit ing for any GE BWR. These events are not reevaluated for reloads and are not required to be because they are not limiting (Reference 5).
The decrease in core flow causes a decrease in reactor power and thermal limits are not challenged. However, the SLO pump seizure a ccident was analyzed under Reference 6 for the introduction of GE14 and EPU (2957 MWt) co nditions. For Quad Cities Unit 1 AREVA reloads, AREVA concludes that the SLO pump seiz ure accident needs to be analyzed when the event is evaluated to protect the AOO acceptance criteria (Reference 8).
This section describes events which cause a de crease in reactor coolant system flow rates, except for anticipated transients without scra m (ATWS). The ATWS mitigation features, which include a double recirculation pump trip, are discussed in Section 15.8. The
recirculation flow control system is described in Section 7.7.3.1.
Because they continue to be bounded by other events which are analyzed for the current fuel cycle, some events described in this section hav e not been reanalyzed for the current fuel cycle.
These events, including the associated assumptions and conclusions, continue to be part of the
plant's licensing basis. The conclusions of th ese analyses are still valid; however, specific details contained in the descriptions and associat ed figures should be used only to understand the analysis and its conclusions. These specific details should not be used as sources of current
fuel cycle design information. Refer to the cy cle reload licensing documents for cycle-specific analyses performed.
15.3.1 Single and Multiple Recirculation Pump Trips
The transient responses of the plant to the trip of one and of both recirculation pumps due to
trip of the adjustable speed drives (ASDs) while operating at full power have been analyzed.
No reactor scram is assumed due to these trans ients. However, a simultaneous trip of both drives implies a loss of auxiliary power, which wo uld subsequently result in reactor scram.
[15.3-1]
Extensive tests and analyses were conducted during the original design of the reactor coolant system to evaluate the performance characteri stics of the jet pumps and the recirculation system, particularly with respect to pump desi gn requirements and the effect of the pumping system on hydraulic and nuclear stability. Th ese analyses included the evaluations which are in Section 5.4.1.3 and also included the evaluation of recirculation pump malfunctions which
are discussed in the following subsections.
[15.3-2]
If one of the recirculation pumps were to fail, flow through half of the jet pumps would decrease and the jet pumps would cease to function. Reci rculation flow and then core flow would decay to a value lower than rated. In this case, fl ow would reverse through the 10 idle jet pump diffusers and the other 10 jet pumps would continue to function. The core flow reduction would result in less core pressure drop and the active jet pump flow ratio would increase. The driving flow in the active loop would remain esse ntially constant since the loop hydraulic characteristics would not change.
[15.3-3]
QUAD CITIES - UFSAR 15.3-1a Revision 14, October 2017 Calculations for typical BWRs (see APED-5460
[1]) show that the 10 jet pumps would provide nearly 150% of their normally rated flow at the lower core pressure drop. Therefore, the total flow injected by the jet pump system would be 75% of rated. About 22% of rated flow would bypass the core through the idle diffusers, hence the core flow would be about 53% of rated.
This lower than normal core flow rate would result in more core coolant void formation. Core power would drop to and stabilize at about 70% of rated. If both drive pumps were tripped, natural circulation would provide approx imately 30% of rated core flow.
QUAD CITIES - UFSAR 15.3-2 Revision 11, October 2011 The one-pump trip transient would be less seve re than the complete loss of pumping power.
Even in the two-pump trip case the MCPR wo uld be greater than the MCPR Safety Limit.
A gradual power decrease would be the only result. Therefore, there are no safety implications for either a one-pump trip or two-pu mp trip, provided that either the reactor is not operated in the region of potential therma l-hydraulic instability or the OPRM system is fully functional, as discussed in Section 4.4.3.
1.10. After a pump trip, power could only be raised by a flow increase or by rod motion.
Protection against exce ssive power generation at any given flow is provided by the rod block interlocks of the APRM and RBM (see Chapter 7). Loss of a driving pump would not result in unstable operation of the remaining
pumps.
Subsequent to the original design evaluations , a series of tests performed at the General Electric Company (GE) Moss Landing Test Facility verified previous performance predictions. Throughout these tests, bas ic performance data were collected under conditions duplicating, in all important resp ects, the temperatures, pressures, and flow rates to be encountered by the recirculation sy stem. Concurrent with performing the tests at Moss Landing, the loop in which the tests were being performed was analytically
modeled. Agreement between the analytical mo del results and the actual test results was quite good. Recirculation system analyses were obtained and are presented in
Sections 15.4.5, 15.3.2, 15.4.4, and in the following subsections.
[15.3-4]
For all reloads, single and multiple recircula tion pump trips are bounded by more limiting events which are performed on a cycle-specific basis.
15.3.1.1 Trip of Both Drive Motors (Historica l for initial core with M-G sets, including Figures) The two-drive-motor trip transient analysis pr ovides an evaluation of the thermal margins associated with a core flow rate decrease co mmensurate with the rotating inertia of the recirculation drive equipment. The decrease in flow would cause additional void formation in the core which would decrease reactor power. The time constant of the fuel would cause the surface heat flux decay to lag behind th e flow decay. The mismatch between reactor thermal power and recirculation flow would br ing about a decrease in the minimum critical power ratio (MCPR).
[15.3-5]
During a flow coastdown from a two-drive-mo tor trip, the MCPR was determined to be greater than the MCPR Safety Limit. This sh ows that adequate inertia has been provided in the recirculation drive equipment. The uni t's response to the two-drive-motor trip is shown for initial core on Figures 15.3-1 and 15.3-2.
This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each reload core based on othe r events which bound this event. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit.
[15.3-6]
15.3.1.2 Trip of One Drive Motor (Historical for initial core with M-G sets, including Figures) The results of this transient would be less seve re than the trip of both drive motors or the stall of one pump. Therefore, the thermal ma rgins during this transient would be greater than either of those cases. For Quad Cities, flow would increase through the active loop jet pump diffusers, and would finally provide about 72% of the original total jet pump diffuser flow. This flow would be split in the lower plenum with about 60% going through the core and the remainder providing reverse flow through the jet pump diffusers of the QUAD CITIES - UFSAR Revision 11, October 2011 15.3-3 tripped loop. A small amount of forward flow would still be induced in the tripped drive loop due to the static pressure difference be tween the downcomer and the jet pump throat.
The unit's response to the one-drive-motor trip is shown for initial core on Figures 15.3-3 and 15.3-4.
[15.3-7]
This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each reload core based on othe r events which bound this event. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit.
[15.3-8]
15.3.1.3 Trip of One Recirculation Pump Motor
The transient responses of the plant to the trip of one or both recirculation pumps while operating at full power have been addressed in the plant licensing basis. No reactor scram is assumed due to these transients. The tr ansient analysis provides an evaluation of the thermal margins associated with a core flow rate decrease commensurate with the rotating inertia of the recirculation pump. The decrease in flow causes additional void formation in the core which decreases reactor power. The ti me constant of the fuel causes the surface heat flux decay to lag behind the flow decay.
The mismatch between reactor thermal power and recirculation flow brings about a decreas e in the minimum critical power ratio (MCPR).
The one pump transient is less severe than the complete loss of pumping power. The thermal margins would be greater than the se izure or stall of one pump, which was the most limiting transient analyzed at the time of the original license. Subsequent analyses (NEDO-10958-A
[4]) showed that the pump seizure even t was no longer a limiting transient.
[15.3-9]
This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each reload core based on othe r events which bound this event. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit.
[15.3-10]
15.3.1.4 Trip of Two Recirculation Pump Motors
The trip of two recirculation pump mo tors is discussed in Section 15.8.
15.3.2 Recirculation Flow Controller Malfunctions
The equipment associated with the variable sp eed recirculation pump motors is designed with the basic objective that any failure should maintain the operating pump speed.
However, the potential for failure in either direction (zero speed demand or full speed demand) does exist. These failures have been analyzed and are discussed here and in
Section 15.4.5, respectively.
[15.3-11]
[Start of HISTORICAL INFORMATION for initia l core with M-G sets, including Figures]
A failure in a M-G set speed controller could c ause the scoop tube positioner for the fluid coupler to move at its maximum speed in the direction of decreasing pump speed and flow.
In the transient, the failed speed controller woul d move its positioner to near zero coupling at a maximum rate of about 10% per second.
The resulting transient would be similar to a one-pump trip.
QUAD CITIES - UFSAR Revision 11, October 2011 15.3-3a The recirculation flow and thermal power de cay would be less severe than would result from tripping one of the recirculation drive motors. Therefore, the MCPR during this
transient would be higher than for a one-pump trip, that is, the MCPR would always be greater than the MCPR Safety Limit. The unit' s response to this transient is shown for initial core on Figures 15.3-5 and 15.3-6.
[15.3-12]
[End of HISTORICAL INFORMATION]
A failure in an ASD speed signal would cause the ASD output to either go to zero frequency, thus shutting off the motor voltage the same as any other trip, or would attempt to increase frequency until the programmed limit is reache
- d. Additionally, if the programmed limit function failed and the motor continued to acce lerate, the over frequency protective relays would immediately trip the ASD feed and shut down the motor. The failure results in an increase in pump speed, which has been analyzed at 2.5% rated rpm/sec for a maximum run-up rate at normal operation (2 pump op eration) and 42% rated rpm/sec for a maximum run-up rate for single pump operation.
The recirculation flow and thermal power deca y resulting from the output going to zero frequency would be the same as analyzed in Section 15.3.1 for a trip of the recirculation pump ASD.
QUAD CITIES - UFSAR 15.3-4 Revision 11, October 2011 This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each reload core based on ot her events which bound this event. Flow dependent MCPR limits are established to su pport operation at off-rated core flow conditions. The limits are based on the CPR c hanges experienced by the fuel during slow flow excursions. The slow flow excursion even t assumes a failure of the recirculation flow control system such that the core flow incre ases slowly to the maximum flow physically attainable by the equipment. An uncontrolled increase in flow creates the potential for a significant increase in core power. The primar y function of the flow-dependent MCPR limit is to protect against cladding overheat during transients initiated from less than rated flow conditions. Refer to the cycle reload licen sing documents for cycle-specific analyses performed.
[15.3-13]
[15.3-14]
15.3.3 Recirculation Pump Shaft Seizure
This accident is assumed to occur as a consequence of an unspecified instantaneous stoppage of one recirculation pump shaft while the reactor is operating at full power.
[15.3-15]
The pump seizure event would be a very mild acci dent in relation to other accidents such as a loss-of coolant accident (LOCA). In both a LOCA and a pump seizure, the recirculation
driving loop flow would be lost extremely rapi dly. Differences between the consequences of a LOCA and a pump seizure include:
A. In the case of the seizure, stoppage of the pump would occur; while for the LOCA, the severance of the line would have a similar, but more rapid and severe, influence.
B. Following a pump seizure event, fl ow would continue, water level would be maintained, and the core would remain submerged, which would provide a
continuous core cooling mechanism. Ho wever, for the LOCA, complete flow stoppage would occur and the water leve l would decrease due to loss of coolant, eventually uncovering the reactor core and subsequently overheating the fuel rod cladding.
C. For the pump seizure accident, reacto r pressure does not significantly decrease, whereas complete depressurizati on occurs for the LOCA.
The increased temperature of the cladding and reduced reactor pressure for the LOCA would combine to yield a much more severe stress and potential for cladding perforation than for the pump seizure. Therefore, the pote ntial effects of the hypothetical pump seizure accident are bounded by the effects of a LO CA and specific analyses of the pump seizure accident are not required.
QUAD CITIES - UFSAR 15.3-5 Revision 11, October 2011 15.3.4 Recirculation Pump Shaft Break
The recirculation pump flow reduction events which have been analyzed are the
trip of one drive motor (Section 15.3.1.2), tri p of both drive motors (Section 15.3.1.1), trip of one recirculation pump motor (Secti on 15.3.1.3) and, recirculation pump shaft seizure (Section 15.3.3). The recirculation pump shaft break event is a limiting fault. The
consequences of this event are bounded by other flow reduction events (Reference 6).
15.3.5 Jet Pump Malfunction
The effects of a malfunction of a single jet pump have been analyzed. For the purpose of
analysis, one of the jet pump nozzles is assume d to be plugged while the plant is operating at full power. Two effects would be observed
- flow through the blocked jet pump would reverse, bypassing some core flow and, th e remaining 19 jet pumps would operate at a slightly higher flow ratio due to the altered hydraulic characteristics. The net effect would be a reduction in core flow to approximately 98% of rated (power approximately 99% of rated). Since the reactor power decrease woul d be essentially analogous to a flow control decrease, the margin from thermal limits wo uld not be reduced. The flow and power reduction would be sensed by the monitoring equipment. Loss of a jet pump would not result in unstable operation of the remaining pumps.
[15.3-16]
This transient is not analyzed for reload cores since the fuel-specific MCPR LCO is determined for each reload core based on othe r events which bound this event. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit.
[15.3-17]
15.3.6 Transients During Single Loop Operation
[Start of HISTORICAL INFORMATION]
In Section 3.1 of NEDO-24807
[2], GE has summarized their review of abnormal operating transients during single loop op eration (SLO). In response to NRC questions raised during their review of Cooper Station's request fo r SLO Technical Specifications, GE completed specific analyses of numerous plant transie nts initiated during SLO. The results demonstrate the applicability of the analyses for Quad Cities Units 1 and 2.
[15.3-18]
The three most important aspects of plant transients and SLO are:
A. They are initiated from less than rated power due to the lower core flow.
B. The safety limit MCPR and the lim iting condition for operation (LCO) MCPR; i.e., operating limit MCPR, are increased. During SLO, the uncertainties of some
of the core parameters increase (e.g, core flow, TIP readings). Therefore, the MCPR safety limit is increased to ensure that 99.9% of the fuel rods do not
experience boiling transition during an AOO.
Further discussion is provided in Section 4.4.4.2.5.
C. The average power range monitor (APR M) scram, APRM rod block, and rod block monitor (RBM) flow-biased setpoints are ad justed to preserve the relationship normally existent between the setpoints and operating points in the power/flow map during two-loop operation.
QUAD CITIES - UFSAR 15.3-6 Revision 14, October 2017 The highest power attainable during SLO is less than rated two-loop thermal power because core flow is reduced and rod patterns will be changed. Reductions in reactor power cause plant transients to become less severe.
Abnormal plant transients initiated from SLO are conservatively bounded by two-loop analyses, provided that the above mentioned adjustments are made to the safety limit MC PR and operating limit MCPR, APRM scram, APRM rod block, and RBM setpoints.
A recirculation pump seizure is identified by GESTAR
[3] as a design basis event. The consequence of this accident during single loop versus two loop operation has been addressed by GE in GESTAR. This accident is defined as the instantaneous stoppage of one recirculation pump shaft while the rea ctor is operating at full power. GE has considered this accident to be mild by co mparison to a LOCA, but analyzed the event for SLO for EPU conditions (2957 MWt) at Quad Citi es. As a result of this analysis it is required that a SLO OLMCPR minimum be implemented based on the value of the
minimum DLO OLMCPR. These values can be found in Reference 6.
These events are not analyzed for reload co res since the fuel-specific MCPR LCO is determined for each reload core based on othe r events which bound this event. The MCPR LCO is calculated to preclude violation of the fuel cladding integrity safety limit.
[15.3-19]
[End of HISTORICAL INFORMATION]
For Westinghouse methodology, it has been demonstrated that the decrease in MCPR for the transients initiated from SLO are conser vatively bounded by the DLO analysis. The safety limit for SLO is adjusted to include the increase in some of the core parameters uncertainties as specified in the Technical Spec ifications. See Reference 7 for more details.
In addition, the maximum average linear heat generation rate (MAPLHGR) is adjusted for SLO as determined from the LOCA analysis.
For Quad Cities Unit 1 AREVA reloads, AREVA co ncludes that the licensing basis events in SLO do not need to be evaluated on a cycl e specific basis except the SLO pump seizure (Reference 8). The results of the SLO pump seizure event are potentially limiting at off-rated conditions in single loop operation. SL O is only supported for power levels and core flow as outlined in the current cycle's COLR per Reference 8. In addition, the MAPLHGR is adjusted for SLO as determined from the LOCA analysis. The cycle specific results are presented in the cycle-specific reload safety analysis report.
For reactor power operation with SLO, both the fuel cladding integrity safety limit MCPR and the operating limit MCPR are increased by an amount specified in the Technical Specifications and Core Operating Limits Report, respectively. This increase is to account for increased uncertainties in core flow and traversing incore probe (TIP) instrumentation readings for single loop operations.
QUAD CITIES - UFSAR Revision 14, October 2017 15.3-7 15.3.7 References
- 1. "Design and Performance of GE BW R Jet Pumps," General Electric Company, September 1968, APED 5460.
- 2. "Dresden Nuclear Power Station Units 2 and 3 and Quad Cities Nuclear Power Station Units 1 and 2, Single-Loop Operation," General Electric Company, December 1980, NEDO-24807.
- 3. "General Electric Standard Application for Reactor Fuel," (GESTAR II), General Electric Company, June 2000, NEDE-24011-P-A-14.
- 4. "General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," General El ectric Company, January 1977, NEDO 10958-A.
- 5. "Safety Analysis Report for Quad Cities 1 & 2 Extended Power Uprate," NEDC-32961P Revision 2, August 2001.
- 6. "Dresden and Quad Cities Extended Power Uprate, Task T0900: Transient Analysis,"
GE-NE-A22-00103-10-01 Revision 0, October 2000.
- 7. "Westinghouse BWR Reload Licensing Meth odology Basis for Exelon Generation Company Quad Cities Nuclear Power St ation Units 1 and 2," WCAP-16334-P, June 2005. 8. ANP-3565P Revision 0, "Quad Cities Unit 1 Cycle 25 Reload Safety Analysis," AREVA Inc., February 2017.
QUAD CITIES - UFSAR Revision 14, October 2017 15.4-12b 1. The reactivity excursion shall not exceed the licensing limit of a radially averaged rod enthalpy of 280 cal/gm at any axial location within the assembly.
- 2. The number of fuel rods predicted to reac h the assumed fuel failure threshold and associated parameters, such as the amount of fuel reaching melting conditions, does not exceed the inputs used in the radiological evaluation.
- 3. The maximum pressure during any point of the accident shall be less than the value that will cause stresses to exceed ASME "Service Limit C". (Note: The CRDA event cannot challenge this pressure limit, so cycle-specific verification is not required.)
The AREVA analysis is typically performed conser vatively assuming the reactor is at hot zero power conditions (isothermal temperatur e, non-voided, and xenon free). Additional conservatism is incorporated by assuming t hat adiabatic conditions remain during the power excursion (i.e. no dire ct moderator heating is credite d during the analysis), and that the reactor remains at hot zero powe r conditions for the analyzed control rod withdrawal sequence. The cycle specific portion of the CRDA analysis includes the determination of limiting rod withdrawals based upon calculated high rod worths and analyzing these withdrawals with MICROBURN-B2. The results of these analyzed withdrawals are used with the generic parameterized analysis to determine the maximum deposited enthalpy. These enthalpies are used to verify that items 1) and 2) above are met. For the purposes of item 2) above, any rod that exceeds a threshold of 170 cal/gm is assumed to experience failure and the total number of failures must remain within the assumptions used in the dose assessment (as described in FSAR Section 15.4.10.5.4). The AREVA CRDA cycle- specific analysis details and results are reported in the cycle-specific reload safety analysis report.
Input Parameters and Initial Conditions At the time of the control rod drop accident, th e core is assumed to be at an operating cycle point which results in the highest worth of th e dropped control rod. The core is also assumed to contain no xenon, to be in a hot-standby condition, and to have the control rods in sequence A and be near critical. The assumption to remove xenon, which competes well for neutron absorptions, increases the fractional absorptions, or worth, of the control rods.
(Sheet 1 of 1)
Revision 14, October 2017 QUAD CITIES - UFSAR
Table 15.4-1 Summary of Doses Calculated at LPZ, Control Room, and EAB for TID and Siemens Source Terms Pre-Uprate Conditions (Historical Information)
Fuel Type Path LPZ Control Room EAB Dose (Rem) Dose (Rem) Dose (Rem) Thyroid Whole Body Beta Thyroid Whole Body Beta Thyroid Whole Body Beta TID-14844 MVP 0.147 3.99E-2 1.51E-2 3.08 9.38E-3 0.188 1.34 0.361 0.137 AOG 0 0.440 0.258 0 0.208 4.21 0 2.26 0.973 Gland 0.892 2.69E-2 9.79E-3 18.7 7.15E-3 0.125 8.09 0.244 8.88E-2 Total 1.04 0.507 0.282 21.8 0.224 4.53 9.43 2.86 1.20 Siemens 20 GWd/MTU MVP 0.130 2.98E-2 1.02E-2 2.73 5.93E-3 0.114 1.18 0.270 9.27E-2 AOG 0 0.285 0.152 0 0.123 2.38 0 1.59 0.619 Gland 0.789 2.02E-2 6.75E-3 16.5 4.82E-3 7.86E-2 7.15 0.184 6.11E-2 Total 0.919 .0335 0.169 19.2 0.134 2.57 8.33 2.04 0.773 Siemens 60 GWd/MTU MVP 0.136 2.70E-2 8.92-E-3 2.86 4.50E-3 9.29E-2 1.23 0.244 8.08E-2 AOG 0 0.226 0.126 0 8.80E-2 1.87 0 1.36 0.514 Gland 0.823 1.85E-2 5.94E-3 17.3 3.94E-3 6.56E-2 7.46 0.168 5.39E-2 Total 0.959 0.271 0.141 20.1 9.64E-2 2.03 8.69 1.77 0.649
Note: AXIDENT computer runs were made assuming all the release passing through one release path at a time. The results present ed in the table were weighted with the actual release fraction through each path.
This dose data does not apply to current fuel cycles.
(Sheet 1 of 1)
Revision 9, October 2007 QUAD CITIES - UFSAR
Table 15.4-1a Control Rod Drop Accident EAB, LPZ and Control Room Doses following EPU (Historical Information)
Location Organ Dose (Rem) Regulatory Dose Limit (Rem)
EAB Thyroid 12.1 75 Whole Body 3.41 6.25 LPZ Thyroid 1.33 75 Whole Body 0.60 6.25 Control Room Thyroid 28 30 Whole Body 0.27 5 Beta 5.35 30 (Sheet 1 of 1)
Revision 14, October 2017 QUAD CITIES - UFSAR
Table 15.4-1b Control Rod Drop Accident EAB, LPZ and Control Room Doses (Alternative Source Term) (1)
Scenario 1 (Main and Gland Seal Condenser Leakage)
Control Room 30 Days 0.903 5.0 EAB 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (max) 1.422 6.3 LPZ 30 Days 0.122 6.3 Scenario 2 (Gland Seal Condenser Leakage and SJAE Release)
Control Room 30 Days 0.735 5.0 EAB 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (max) 2.746 6.3 LPZ 30 Days 0.210 6.3 Notes: 1 - The radiological consequences are based on th e Westinghouse Optima2 core inventory described in Section 15.4.10.5.1. These consequences are bounding for the AREVA ATRIUM 10XM design.
QUAD CITIES - UFSAR 15.5-1 Revision 14, October 2017 15.5 INCREASE IN REACTOR COOLANT INVENTORY
[Start of HISTORICAL INFORMATION]
The AREVA methodology is only applicable to Quad Cities Unit 1.
GE has determined that the analysis of Inad vertent HPCI Startup (IHPCIS) event is not required for reload licensing if the Loss of Feedwater Heating (LFWH) event is shown to be bounding based on more limiting core inlet subcooling. It is implicitly assumed in this
methodology that the reactor water high level trip would no t occur following an IHPCIS.
Were the reactor water high leve l trip to occur, the IHPCIS event would be a pressurization (i.e., turbine trips on high water level) event superimposed on a subcooling event. In this case, it is likely that IHPCIS would not be bounded by LFWH, and may potentially be worse than the Feedwater Controller Failure (FWCF) event.
Starting with Quad Cities Unit 2 Cycle 18 relo ad analysis, GE has reviewed this event with respect to the HPCI trip and re actor water high level turbin e trip setpoints and concluded that a high water level turbine trip may not be avoided. Therefore, inadvertent HPCI calculations were performed in all of th e same operating domains (ICF, MELLL, FWTR, Bypass OOS) as performed for the FWCF event.
The results show that this event is bounded by the FWCF event. GE conclude d that the FWCF event with appropriate penalties, bound HPCI and no further OLMCPR adjustment was needed. This evaluation will be confirmed for the future reloads and the results will be documented in the supplemental reload analysis report.
[End of HISTORICAL INFORMATION]
Westinghouse has analyzed IHPCIS as a core-wide pressurization transient for all the operating domains licensed for Quad Cities. In this case, it is assumed that the HPCI pump trip would not occur following an IHPCI event where the reactor water level would reach the high water level trip setpoint. Then the reactor water high level would initiate a turbine trip and the IHPCI would become a pressurization event superimposed on a subcooling event like the FWCF. In this case , the IHPCI event could be potentially more limiting than the FWCF. The results are docume nted in the Westinghouse reload licensing report.
For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), the IHPCI analysis assumes injection of the cold HPCI flow into the feedwater sparger and credits operation of the feedwater control system.
At high power level, the feedwater control system decreases the severity of the event by potentially preventing the water level from reaching the Level 8 high level trip.
The results are documented in the reload safety analysis report.
15.5.1 Inadvertent Initiation of High Pressu re Coolant Injection During Power Operation 15.5.1.1 Identification of Causes and Frequency Classification
Inadvertent startup of the high pressure coolant injection (HPCI) system is postulated for this analysis, i.e., operator error. This transie nt disturbance is categorized as an incident of moderate frequency.
[15.5-1]
Inadvertent startup of the high pressure cool ant injection (HPCI) system requires multiple equipment failures or operator error.
QUAD CITIES - UFSAR 15.5-2 Revision 14, October 2017 MCPR operating limits are defined such that th e MCPR fuel cladding integrity safety limit is not violated during the occurrence of this transient.
15.5.1.2 Sequence of Events and Systems Operation
[15.5-2] [15.5-3]
For operation prior to EPU, various power-fl ow initial conditions were analyzed for the IHPCIS event. The 100% power, 100% recircu lation flow case represented the most limiting of the cases analyzed.
The EPU transient analysis in Reference 5 cont ains information regarding this event as analyzed by GE.
The sequence of events for the event analyzed by Westinghouse is similar to the FWCF
transient described in Section 15.1.2. The IHPCIS event also shows similar results
compared to the FWCF event. Both sets of analyses include cases at rated and off-rated power and flow conditions to cove r the expected range of operation
. See Westinghouse cycle-specific reload reports for details app licable to the current analysis of this event.
For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), the inadvertent initiation of the high pressure coolant inje ction system (IHPCI) is a potentially limiting event and is analyzed on a cycle-specific basis. The water injection causes an increase in the water level. In an effort to maintain the water level, the feedwater/level control system responds to the water level increase by decre asing the feedwater flow. At lower power levels, the decrease in feedwater flow may not be enough to offset the HPCI flow so the water level may increase until a high level turb ine trip occurs. The turbine trip results in a pressurization event similar to the pressuriza tion portion of a FWCF event. At power levels at which the high level set point is not reached, the power will increase due to the increase in subcooling and the core approaches a new pseudo steady-state condition. Since the high level HPCI turbine trip occurs at the same set point as the high level turbine trip, it does not help keep the turbine trip from o ccurring. The IHPCI results are documented in the AREVA cycle-specific reload safety analysis report.
15.5.1.3 Core and System Performance
[Start of HISTORICAL INFORMATION]
Prior to EPU, the analysis of this event was based on a number of assumptions. For the
initial core, it was assumed that at time zero, 5600 gal/min of 40°F water was admitted to the vessel and mixed with the much warmer f eedwater. The level control system, sensing additional water causing a level rise, cuts back on the feedwater flow entering the vessel.
The colder water increases the core inlet subcoo ling, and due to the negative void reactivity coefficient, an increase in core power results.
All analysis was done assuming the plant was initially on manual flow control; this result s in no recirculation flow decrease and thus permits the maximum power increase due to the subcooling change.
[15.5-3a]
The EPU transient analysis in Reference 5 cont ains information regarding this event as analyzed by GE.
[End of HISTORICAL INFORMATION]
For reload cores licensed with Westinghouse me thods, the results and assumptions for this event are provided in the cycle-specific reload reports. At the start of the event the same initial HPCI flow rate is assumed (5600 gal/min), delivered to the feedwater system at near QUAD CITIES - UFSAR Revision 14, October 2017 15.5-3 freezing conditions (32°F). No credit is taken for a reduction in feedwater flow in response to the increase in water level vessel inventory.
Instead, it is conservatively assumed that feedwater and HPCI continue to operate until th e time of high water level turbine trip and feedwater pump trip. As a result of this conservative modeling, the IHPCIS event can become a limiting pressurization transient similar to the FWCF event. Results are provided in the cycle-specific reload reports.
For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), the results for this event are provided in the cycle-specific relo ad safety analysis report. Operation of the feedwater/level control system is credited in the analysis. The control system will respond to the increase in inventory by decreasing the feedwater flow. At lower power levels, the available reduction in feedwater flow is not su fficient to offset the HPCI flow so the water level will increase until a high level turbine trip occurs.
15.5.1.4 Barrier Performance
This transient is analyzed for each cycle sinc e the MCPR LCO is calculated to preclude violation of the fuel cladding integrity sa fety limit. See Section 15.5 for details.
[15.5-4]
15.5.1.5 Radiological Consequences
The fuel cladding integrity safety limit woul d not be violated; therefore a radiological consequence analysis has not been performed.
15.5.2 References
- 1. "General Electric Standard Applicatio n for Reactor Fuel" (GESTAR II), June 2000, General Electric Co., NEDE-24011-PA-14, and U. S. Supplement, NEDE-24011-PA-14-US.
- 2. Deleted.
- 3. Deleted.
- 4. Deleted.
- 5. "Dresden and Quad Cities Extended Power Uprate, Task T0900: Transient Analysis,"
GE-NE-A22-00103-10-01 Revision 0, October 2000.
QUAD CITIES - UFSAR Revision 7, January 2003 15.6-1 15.6 DECREASE IN REACTOR COOLANT INVENTORY
This section covers postulated transients which involve an unplanned decrease in reactor coolant inventory. These events include inadve rtent opening of a safety valve, relief valve, or safety relief valve, failure of an instrume nt line carrying reactor coolant outside primary containment, main steam line break outside primary containment, and loss of coolant resulting from the failure of pipes within the reactor coolant pressure boundary.
Because they continue to be bounded by analys es for previous fuel cycles, the events and radiological consequences described in this section have not been reanalyzed for the current fuel cycle. The conclusions of these radiolog ical analyses are still valid; however, specific details contained in the descriptions and associ ated results and figures should be used only to understand the analysis and its conclusions.
These specific details should not be used as sources of current fuel cycle design information.
For many of the following events, radiologic al analyses were performed by both CECo and the AEC/NRC for initial licensing of the plant.
Since the initial licensing, Quad Cities has changed from 7x7 fuel arrays to 8x8 fuel arrays, 9x9 fuel arrays, and 10x10 fuel arrays. As described in this section, radiological evaluat ions have shown that the postulated releases for 8x8, 9x9, and 10x10 arrays would be less than those for 7x7 arrays.
15.6.1 Inadvertent Opening of a Safety Valve, Relief Valve, or Safety Relief This event was not a specific consideration in the original licensing of Quad Cities, and the following evaluation shows that it is not of sa fety significance. The following is based on the NRC approved evaluation of a similar BWR/
3 (Dresden Unit 2) performed during the Systematic Evaluation Program.
The inadvertent opening of a safety valve, re lief valve, or safety relief valve (SRV) would result in a decrease in reactor coolant invent ory and a decrease in reactor coolant system pressure.
[15.6-1]
If a SRV or relief valve failed open, it would di scharge to the suppression pool. The safety valves discharge directly to drywell atmosphere. Although a drywell high pressure reactor trip may occur if a safety valve failed open, the following analysis assumes a safety valve discharge would result in a sequence of events similar to a relief valve or SRV discharge.
15.6.1.1 Identification of Causes
The cause of an inadvertent opening of a safety valve, relief valve, or SRV is attributed to malfunction of the valve.
The inadvertent opening of a safety valve, relie f valve, or safety relief valve are bounded by limiting events specified in the cycle-specific reload licensing documents and/or the COLR.
QUAD CITIES - UFSAR Revision 4, April 1997 15.6-2 15.6.1.2 Sequence of Events and Systems Operation
The following sequence of events is assumed for this analysis.
The normal functioning of plant instrumentatio n and controls is assumed for this incident; specifically, the operation of the pressure regulator and vessel level control systems is assumed normal. On relief valve or SRV open ing, the pressure regulator would sense the pressure decrease and would cause the turbine con trol valves to partially close. No reactor trip occurs, and the reactor conditions stabilize at a power level near the initial power. The feedwater system would be used to make up the continuing loss of reactor coolant inventory.
If the pressure regulator fails to respond, the decrease in main steam line pressure would cause the main steam isolation valves (MSIVs) to close. This event is discussed in
Section 15.1.3.
If the feedwater system became unavailable due to a single failure or loss of offsite power, the HPCI system could provide makeup water.
HPCI would be automatically actuated on low-low water level.
If a relief valve or SRV opens and fails to recl ose, the torus would experience an increase in temperature. Closure of the MSIVs could not halt the blowdown since the relief valves and SRV are upstream of the MSIVs.
[15.6-2]
15.6.1.3 Core and System Performance
This event is not limiting from a core performance standpoint.
Inadvertent opening of a safety valve, relief va lve, or SRV would cause a negligible pressure reduction which could lead to partial closure of the turbine control valve by the pressure regulator. The net change in power level and coolant conditions within the fuel assemblies would be negligible, and operating thermal margins would be relatively unaffected.
Therefore, minimum critical power ratio (MCPR) would not change significantly.
Refer to Section 6.2.1.3 for suppression pool temperature and pressure response to an opened relief valve or SRV.
15.6.1.4 Barrier Performance
The NRC has concluded,[1] based on the evaluation of other similar plants,[2, 3] there would not be any fuel failure resulting from a stuck open safety valve, relief valve or SRV event since MCPR would not change significantly.
Therefore, the transient resulting from an inadvertently opened safety valve, relief valv e, or SRV would not have a significant effect on the reactor coolant pressure boundary and wo uld not violate the fuel clad integrity limit.
QUAD CITIES - UFSAR Revision 7, January 2003 15.6-3 15.6.1.5 Radiological Consequences
The consequences of this event would not result in fuel failure. Discharge of normal coolant activity to the suppression pool via SRV or relie f valve operation would result. This activity would be contained in the primary containmen
- t. Any discharges to the environment may be made under controlled release conditions.
During purging of the containment, the release would be in accordance with the established limits; therefore, this event, at the
worst, would only result in a small increase in the yearly integrated exposure level.
15.6.2 Break in Reactor Coolant Pre ssure Boundary Instrument Line Outside Containment A break of a reactor coolant pressure boundary instrument line outside primary
containment could result in the pressurization of secondary containment and the release of radioactive material to the environs. The response of the secondary containment to an instrument line break is discussed in Section 6.2.3.3. The following section describes an instrument line break and the potential radiological consequences.
See the introduction to Section 15.6 for inform ation regarding use of details from this analysis description which may not be ap plicable to the current fuel cycle.
The remainder of Section 15.6.2 discusses the original GE disposition of this event.
15.6.2.1 Identification of Causes
[15.6-3]
A postulated 1 inch reactor coolant pressu re boundary instrument line break has been analyzed.
15.6.2.2 Sequence of Events and Systems Operation
The break was assumed to occur outside the pr imary containment but upstream of the flow check valve in the line. A manually operated stop valve is located outside the containment wall upstream of the break. This valve was no t assumed to be closed until after the reactor was shutdown and depressurized. The reactor was considered to be shutdown manually by the operator upon detection of the break.
Routine surveillance on the part of the operator as given in the following items A through G has been a sufficient program for the periodic testing and examination of the valves in these small diameter instrument lines.
Such leaks could be detected by one or a combination of the following:
A. Comparison of readings among severa l instruments monitoring the same process variable such as reactor level, jet pump flow, steam flow, and steam pressure; QUAD CITIES - UFSAR Revision 8, October 2005 15.6-3a B. Annunciation of the failure of the affected control function, either high or low, in the control room;
C. Annunciation of a half-channel scram if the rupture occurred on a reactor protection system instrument line;
D. A general increase in the area radiat ion monitor readings th roughout the reactor building;
E. Audible noise from the leakage heard either inside the turbine building or outside the reactor building on a normal tour;
F. Unexplained increased in floor drain collector tank water level; and
G. Increases in area temperature mo nitor readings in the reactor building.
Calculations of doses due to the released radioactive materials included the following assumptions. A coolant activity consistent with a plant off-gas release rate of 100,000 micro-Curies/sec was assumed to be released to the environment. While the release occurs at the top of the reactor building, it was a ssumed that downwash occurs resulting in an effective release height of 0-me ters. No core uncovering occurs and no perforations occur, so only coolant activity is released. Iodine in the 30,000 pounds of water that flashed to steam was assumed to be transported with the steam.
The description that follows was a response to a follow-up question concerning a postulated 1" instrument line break within secondary cont ainment, during initial licensing of Quad Cities Station. The AEC requested Quad Cities Station to specifica lly provide assurance that the integrity of secondary containment would be maintained and that the building filters (Standby Gas Treatment) would not be bypassed.
The instrument line break has been re-evaluated from the standpoint of expected pressure
buildup inside the secondary containment of a two-unit plant with the use of the SBGTS.
The second analysis described below (and in the following barrier performance section) is historical and is not currently relied upon fo r plant activities. The assumptions included the original proposed technical specificatio n coolant activity of 20 micro-Curies/cc total Iodine, with an isolation of RB ventilation and SBGT system auto-start resulting from a vent duct high radiation trip signal.
[Start of historical analysis]
The SBGTS is assumed to be initiated on high radiation in the ventilation system and
continuously remove 4000 cfm from the secondary containment.
The analysis assumed that over a 3-hour period 100,000 pounds of liquid is released to the
reactor building from which 30,000 pounds of steam is formed.
QUAD CITIES - UFSAR Revision 7, January 2003 15.6-4 Reactor building pressure would start to increase, thereby caus ing back pressure to be seen by the ventilation supply fan such that essent ially no air flows into the reactor building.
While the blowdown would occur over a 3-hour period, a zero pressure differential between the reactor building and atmosphere would be reached in 3000 seconds, as shown in Figure 6.2-37, after which time the differential pressu re would be negative and all release would be via the SBGTS and the main chimney. During the first 3000 seconds, 43,000 pounds of
mass would be discharged to the secondary containment of which 14,000 pounds would be flashed to steam. Of the steam mass, 6,500 pounds would be released unfiltered and 7,450
pounds would be released via the SBGTS.
[End of historical analysis]
15.6.2.3 Barrier Performance
No core uncovering would occur and no perforat ions would occur, so only coolant activity would be released.
15.6.2.4 Radiological Consequences
This section describes instrument line break ra diological dose calculations based on the above transient analyses which were performe d for the initial licensing of the plant.
QUAD CITIES - UFSAR Revision 8, October 2005 15.6-5 The iodine activity associated with the rele ased liquid was 0.04 micro-Curies/cc of I-131 and 0.3 micro-Curies/cc of I-133. No further release of iodine from the water was assumed. Very stable 1 meter/sec meteorological conditions were assumed since this is the worst case for an equivalent ground level release. Calculated lifetime thyroid dose for the duration of the release of 0.3 Rem is well below the referenc e doses of 10CFR100 and is in fact less than the annual dose permitted in 10CFR20.
The iodine activity associated with the releas ed liquid of 0.04 mciro-Curies/cc of I-131 and 0.3 micro-Curies/cc of I-133 equates to 0.121 I-131 dose equivalent. The Technical Specification shutdown Limiting Condition for Operation (L CO) limit for coolant activity of 4.0 micro-Curies/gram (1 gram = 1 cc of reactor coolant) has been used to determine dose consequences.
Dose versus coolant activity is a linear rela tionship, and for coolant activity of 4.0 micro-Curies/cc of I-131 dose equivalent, the differenc e is conservatively a multiple of 40. The calculated lifetime thyroid dose for the duration of the release is therefore 12.0 Rem, which is well within 10CFR100 limits. An evaluation conducted using available emergency
preparedness models confirmed that the orig inal dose consequences analysis remains conservative with respect to cu rrent industry specifications on allowable concentrations of reactor coolant activity for I-131 dose equivalent.
[Start of historical analysis]
The analysis described below is historical and is not currently relied upon for plant activities.
Radiological dose calculations assumed the iodine activity associated with the released liquid was 20 micro-curies/cc. Iodine in the 30,000 pounds of water that flashed to steam was assumed to be transported with the steam.
The radiological consequences are based on th e sum of the releases via normal inleakage paths (where the effective release height is assumed to be zero, and maximum building dilution effects of 1/3 with type F-1 m/sec meteorology are assumed) and releases via the SBGTS (where type B-1 m/sec meteorological conditions and 90% filter efficiency are assumed).
The maximum offsite exposure with complete ai r/steam mixing and with steam condensation is 0.02 rem to the thyroid at the site boundar
- y. The maximum offsite exposure with no mixing of steam and air and no steam condensation is 6 rem to the thyroid. These exposures
take into consideration the leakage via unfilter ed leakage paths as well as leakage through the SBGTS and the main chimney.
[End of historical analysis]
Therefore, in the event of failure of the 1-inch instrument line, even assuming the very
conservative assumptions outlined in the prec eding paragraphs, the resulting dose would be a fraction of 10 CFR 100 limits.
[15.6-5]
The specific activity of the primary coolant is lim ited by Technical Specification. In addition, there is no core uncovery and no perforations of the fuel during an instrument line break.
Therefore, since only the coolant activity is re leased, the radiological dose calculations are independent of fuel type or design.
The reactor coolant released mass and flashe d fraction for this analysis envelop those parameters for extended power uprate. Theref ore, the calculated doses are not impacted by EPU.
QUAD CITIES - UFSAR Revision 7, January 2003 15.6-5a 15.6.3 Steam Generator Tube Failure
This section is not applicable to Quad Cities Station.
15.6.4 Steam System Line Break Outside Containment
15.6.4.1 Identification of Causes and Frequency Classification
The postulated accident is a sudden, comple te severance of one main steam line outside containment with subsequent release of steam and water containing fission products to the pipe tunnel and the turbine building. This large flow of steam to the turbine building could
fail the blowout panels and lead to the formation of a large steam cloud.
[15.6-6]
QUAD CITIES - UFSAR Revision 7, January 2003 15.6-6 This event is classified as a limiting fault, i.
- e. an event that is not expected to occur but is postulated because the consequences may result in the release of significant amounts of radioactive material.
[15.6-7]
The steam system line break outside containm ent is not re-analyzed for reload cores.
15.6.4.2 Sequence of Events and Systems Operation
To evaluate the overall consequences of the po stulated severance of one of the four main steam lines, the sequence of events following the break was investigated in detail.
[15.6-8]
The initial conditions prior to the main steam line break were assumed to be:
Reactor Power 2511 MWt Reactor Pressure 1020 psia Reactor Water Level Normal
The sequence of events assumed is:
Event Time After Break (seconds) Main steam line break 0 Main turbine control valves closure initiation 0.2 MSIV closure initiation 0.5
Reactor trip/control rod insertion started 1.5 Feedwater flow shut off 5.0 MSIVs closed 10.5 15.6.4.3 Core and System Performance
15.6.4.3.1 Main Steam Isolation Valve Closure
The steam blowdown flow rate through both ends of the postulated break would cause an increase in steam flow in each of the four lines to the maximum value allowed by critical
flow considerations. Flow limiters (venturis) are sized in conjunction with isolation valve closure time so that core submergence is assu red during blowdown and after termination of the accident. Therefore, venturi design wo uld limit the maximum initial steam blowdown rate to 200% of rated steam flow. Rapid depre ssurization in the steam lines downstream of the flow limiters would initiate closure of the main turbine control valves within 0.2 second
after the accident. The increased pressure differential across the flow limiters would indicate the severance immediately and would initiate MSIV closure (all 8 valves) within 0.5 second after the accident. Multiple fl ow limiter pressure differential sensors are provided in the primary containment isolat ion system to accomplish this function.
[15.6-9]
QUAD CITIES - UFSAR Revision 4, April 1997 15.6-7 15.6.4.3.2 Reactor Core Shutdown
A reactor scram would be initiated by a position switch on each MSIV at approximately 10%
closure of the valve stem, as described in Section 7.2. Therefore, control rod insertion would begin within 1.5 seconds after the line break wi th a MSIV total elapsed closure time of 10.5 seconds (0.5 second detection plus 10 seconds clos ure). The MSIVs are designed to close against reactor operating pressure. In addition to the scram from MSIV closure, moderator voids generated in the core by depressurization c aused by excess flow leaving the vessel would contribute sufficient negative rea ctivity to reduce re actor power immediately. Finally, as an additional backup, reactor low wa ter level, which would occur la ter during the blowdown when the steam-water mixture density in the reactor ve ssel would be sufficiently low, would initiate a scram and isolate the reactor.
[15.6-10]
15.6.4.3.3 Feedwater Flow
Since the design basis is the simultaneous loss of normal ac power, the analysis has been done
assuming no feedwater flow.
[15.6-11]
If normal ac power is available, the feedwater flow control valve would initially open fully due to the increased steam flow from the reactor pressure vessel. Within 1 second after the accident, the indicated high water level in th e reactor vessel would initiate closure of the feedwater control valve. The feedwater flow would then decrease linearly to shut off at approximately 5 seconds after the accident. Follo wing closure of the MSIVs, the reactor vessel water level would drop due to collapsing steam vo ids, thereby actuating the feedwater system to return the vessel water level to normal.
15.6.4.3.4 Reactor Coolant Blowdown
The two distinct intervals of blowdown are va por blowdown before the steam-water mixture flows into the steam line, and steam-water mixtur e blowdown. The steam flow rate through the upstream side of the break would increase from the initial value of 680 lb/s in the line to 1360
lb/s (200% of initial) with critical flow occurring at the flow limiter in the steam line. The steam flow rate was calculated using an ideal nozzle model.
[4] The flow model predicting the behavior of the flow limiter has been substantiated by te sts conducted on a scale model over a variety of pressure, temperature, and moisture conditions.
The steam flow rate through the downstream side of the break would consist of essentially equal flow components from the other three unbroken lines. In the three unbroken lines, criti cal flow would occur at their flow limiters since it was conservatively assumed that no friction losses exist.
[15.6-12]
The steam flow rate in each of the three unbroken lines would increase from the initial value of
680 lb/s to 1360 lb/s (200% of initial). Total break flow is shown in Figure 15.6-1. The total
steam flow rate leaving the vessel would be appr oximately 5500 lb/s which would be in excess of the generation rate of 2700 lb/s.
The simultaneous initial depressurization in the vessel would be at a rate of approximately 40
psi/s, as shown on Figure 15.6-2, which would cause flashing of the moderator throughout the reactor. Steam bubbles generated within the syst em would cause the reacto r water level to rise at a rate determined by the difference between the QUAD CITIES - UFSAR Revision 4, April 1997 15.6-8 rate at which bubbles are formed and the rate at which they break the water surface.
Steam bubbles rise by buoyancy at an average velocity of 1 ft/s
[5,6]relative to liquid, eventually separating from the mixture surface.
An analytical model was used to predict the rate of reactor water level rise. In a portion of the range of interest (i.e., steam blowdown) this model has been shown to be in reasonable
agreement with level rise data obtained in a large vessel undergoing depressurization.
When the reactor water level floods the steam dryers and reaches the vessel steam nozzles, the blowdown would change from single phase steam to a steam-water mixture.
A steam-water blowdown, would begin at 5 seco nds after the break and would blow down at an average value of 12,000 lbm/s
[7] as shown on Figure 15.6-1. At 8 seconds critical flow would be established because the MSIVs would be nearly closed.
15.6.4.3.5 Steam-Water Mixture Impact Forces
The maximum differential pressure which coul d be generated by continuous water flow past the MSIVs is 850 psi, that is, reactor vessel pressure when the valve is almost closed.
This is below the differential pressure a cross the valve during hydrostatic testing.
[15.6-13]
The surge pressure from the steam-water mixt ure in the steam line has been evaluated for the turbine-generator (T-G) design case, assumi ng that the MSIV closes in 10 seconds and assuming "instantaneous" deceleration of two-p hase mixture at the valve. Line friction was ignored and a driving pressure of 850 psi was a ssumed. The reactor vessel pressure is less than 1000 psig due to vessel depressurization pr ior to mixed flow. By the time two-phase flow occurs in the steam line, the flow limiters would restrict flow. When flow is choked at the flow limiters, the pressure upstream of th e valve (downstream from the limiter) is only 260 psig. Therefore, the maximum surge pressure is 260 psig when the steam-water
mixture encounters the valve.
The resultant total transient differential pre ssure across the valve (520 psi) is well below the piping design pressure of 1250 psig. The line pressure downstream of the valve was
conservatively assumed to be zero when calculating the differential pressure.
In addition to the preceding MSIV eval uation, a test program was conducted to demonstrate the ability of the MSIVs to withstand transient forces. This test program is described in Section 6.2.4.3.
15.6.4.3.6 Effect of Main Steam Isolation Valve Closure Time
A parametric analysis was performe d to determine the effect of variable closing time of the MSIVs. The results are shown in Table 15.6-1.
[15.6-14]
It would be necessary to lose approximatel y 120,000 pounds of water and steam before the top of the core would be exposed. As shown in Table 15.6-1, even if the MSIVs were closed in 10.5 seconds, the core would not be uncover ed, even for the limiting condition of zero steam separation in the separators.
QUAD CITIES - UFSAR Revision 12, January 2013 15.6-9 15.6.4.3.7 Core Cooling
The assumption of the simultaneous loss of normal ac power supply with the postulated break of
one of the main steam lines would result in the coastdown of the recirculation drive pump flow as
well as the feedwater flow. For the initial main steam line break analysis, core inlet flow was
calculated using a computerized transient mode l which simulates the dynamics of the system including volume changes, heat addition, water leve l rise, and pump inertia. The initial analysis results of the core inlet flow transient are given on Figure 15.6-3. Approximately 5 seconds after the break, the core inlet plenum would begin to flash thereby causing an increase in the core inlet flow. [15.6-15]
For the main steam line break analysis performed for initial licensing a computerized transient model was used to evaluate the thermal hydraulic performance of the reactor core following this postulated break. The model includes as input: dimensions of the fuel assembly, thermal
properties of the coolant and fuel, and time-var ying pressure and flow, thereby allowing the evaluation of the condition of the fuel throughout the accident. The primary result of the initial thermal hydraulic performance analysis of the fuel bundle is the peak fuel clad temperatures.
The evaluation of the accident was based on th e reactor power being at 2511 MWt (corresponding to the steam flow rate for turbine design), pressure at approximately 1000 psig, and the combination of peaking factors resulting in a pe ak linear power density of 17.5 kW/ft for the 7x7 array such that the fuel is operating at its maximum warranted value. Furthermore, the fuel rod thermal performance was evaluated with a skewed cosine (to the top) axial power generation, (see Figure 15.6-4), causing an initial minimum critical heat flux ratio (MCHFR) close to the
Technical Specification limit then in effect (ci rca 1970). The calculated MCHFR throughout the initial transient analysis is plotted on Figure 15.6-5 and shows the MCHFR of 1.6 approximately two seconds after the postulated accident occurs.
[15.6-16]
15.6.4.4 Barrier Performance
The reduction in the vessel depressurization rate caused by the change from single-phase steam blowdown to steam-water mixture blowdown wo uld result in an increase in MCPR and MCHFR primarily due to the reduction in flow caused by pump coastdown. The MCPR and the MCHFR through the initial part of the accident is shown on Figures 15.6-5 and 15.6-6. Following closure
of the MSIVs the vessel would start to pressuri ze and the MCPR and MCHFR would continue to increase. The maximum MSIV closure time of 10.5 seconds would limit the total amount of
liquid and steam lost from the primary system to prevent the core from being uncovered.
Therefore, continuous cooling of the reactor core would be maintained throughout the transient and the subsequent initiation of the HPCI syst em, approximately 40 seconds after the accident occurs, would provide additional co ntinuous core cooling. The peak (fuel rod) clad temperature (PCT) would never exceed its operating value and the fuel rod cladding would remain intact since no perforations of the fuel rod cladding would occur.
[15.6-17]
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-10 15.6.4.5 Radiological Consequences
15.6.4.5.1 Application of Alternative Source Term Methodology Regulation 10 CFR 50.67, "Accident Source Te rm," provides a regulatory mechanism for power reactor licensees to voluntarily replace th e traditional accident source term used in design basis accident analyses (i.e., TID 14844, Reference 45) with an "Alternative Source Term" (AST). The methodology for this appr oach is provided in NRC Regulatory Guide 1.183 (Reference 46).
Accordingly, Quad Cities applied for the AST methodology for key De sign Basis Accidents (DBAs). In support of a full-scope implementation of AST as described in Reference 46, AST radiological consequence are performed fo r the four DBAs that result in offsite exposure consequences: Loss of Coolant Accide nt (LOCA), Main Steam Line Break (MSLB), Fuel Handling Accident (FHA), and Control Rod Drop Accident (CRDA).
The NRC approved AST for Quad Cities in Reference 47.
Fission Product Inventory No fuel damage is expected to result from a MSLB. Therefore, the activity available for release from the break is that present in th e reactor coolant and steam lines prior to the event. Two cases are evaluated. Case 1 is for full power operation with a maximum equilibrium coolant concentration of 0.2
µCi/gm dose equivalent I-131. Case 2 is for a maximum coolant concentration of 4.0
µCi/gm dose equivalent I-131, based on a pre-accident iodine spike. This source term is consistent with the guidance contained in Reference 46. The MSIVs are assumed to isolate in 5.5 seconds. The radiological analysis assumes a conservative mass release of 1.4E5 pounds of liquid consistent with Standard Review Plan Section 15.6.4 (Reference 48).
Core Inventory Release Fractions No fuel damage is expected to result from a MSLB. Therefore, the activity available for release from the break is that present in th e reactor coolant and steam lines prior to the event. Atmospheric Dispersion Factors The Exclusion Area Boundary (EAB) and Lo w Population Zone (LPZ) atmospheric dispersion factors (/Q) were determined using the guidance in NRC Regulatory Guide 1.5 (Reference 49), specifically:
/Q = 0.0133/yµ, where y = horizontal standard deviation of the plume (meters)
µ = wind velocity (meters/second)
EAB /Q = 8.64E-04 sec/m 3 LPZ /Q = 8.69E-05 sec/m 3
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-10a Atmospheric dispersion factors were not calculated for the control room dose assessment. The plume was modeled as a hemispherical volume, the dimensions of which are determined based on the portion of the liquid reactor coolant release that flashed to steam. The activity of the plume is based on the total mass of water released from the break. This assumption is conservative because it considers the maximum release of fission products.
Release Paths The entire radioactivity in the released c oolant is assumed to be released to the atmosphere instantaneously as a ground level release. No credit is taken for holdup, plateout, or dilution within facility buildings.
In addition, no credit is taken for control room isolation.
Dose Consequences Dose consequences were determined usin g simplified models consistent with the requirements of Reference 46. Offsite do se consequences (i.e., EAB and LPZ) were determined based on the following factors: radi oactivity releases (in Curies), dispersion factors, breathing rates, and dose conversion s factors. The main control room doses are determined somewhat differently. Steam cloud concentrations are used rather than /Q multiplied by a curie release rate. No credit is taken for main control room filtration.
Dose conversion factors were obtained from Federal Guidance Reports 11 and 12 (References 50 and 51).
15.6.4.5.2 Acceptance Criteria The AST acceptance criteria for postulated majo r credible accident scenarios are provided by 10 CFR 50.67 and Regulatory Guide 1.183. For the main control room, adequate radiation protection is provided to perm it access to and occupancy under accident conditions without personnel rece iving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the du ration of the accident. The AST acceptance criteria for an individual located at any point on the boundary of the exclusion area (the Exclusion Area Boundary or EAB) are provided by 10 CFR 50.67 as 25 rem TEDE for any 2-hour period following the onset of the post ulated fission product release. The AST acceptance criteria for an individual located at any point on the outer boundary of the low population zone (LPZ) are provided by 10 CFR 50.67 as 25 rem TEDE during the entire period of passage of the radioactive cloud re sulting from the postulated fission product release. Regulatory Guide 1.183 applies the fo llowing additional limits to events with a higher probability of occurrence including the MSLB. For the MSLB event (BWR), in the case of an accident assuming fuel damage (i.e
., pre-incident iodine spike) doses at the EAB and LPZ should not exceed 25 rem TEDE. For MSLB accidents assuming normal equilibrium iodine activity, doses should not exceed 2.5 rem TEDE.
15.6.4.5.3 Computer Models Dose consequences were determined usin g simplified models consistent with the requirements of Reference 46.
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-10b 15.6.4.5.4 Key Plant Assumptions Mass of liquid water released: 140,000 lb (con servative relative to expected release) Flashing fraction: 40% Mass of steam in cloud: 56,000 lb CREV mitigation: Not credited Reactor coolant activity: Case 1: 0.2
µCi/gm; Case 2: 4.0
µCi/gm 15.6.4.5.5 Results The reanalysis of the MSLB accident event using AST methodology is documented in calculation QDC-0000-N-1266 (Reference 52). Th e results are summarized in Table 15.6-2a. Section 15.6.4.5.6 describes the radiological a ssessment performed prior to the adoption of an updated accident source term in accordanc e with 10 CFR 50.67, "Accident Source Term."
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-10c 15.6.4.5.6 Historical Information The following describes main steam line break radiological dose analyses performed for the initial licensing of the plant, i.e., 7x7 fuel arrays.
The predominant activity in the discharge d coolant would be N-16, which would be significantly reduced by decay due to its sho rt half-life (about 7 seconds). If the reactor contained fuel with cladding leaks, the wa ter released through the break would contain some fission products.
[15.6-18]
During 1964 the Dresden Unit 1 reactor was oper ated with a significant number of cladding leaks. Analysis of reactor water samples in dicated the following yearly average fission product content:
Activity Fission Product (µCi/cc) I-131 0.025 I-133 0.1 Other halogens 0.25 Other fission products 0.25
With a separate reactor cleanup system for ea ch unit, it is estimated that the maximum coolant activity would be approximately 2.3
µCi/cc at 100,000 µCi/s release rate, and would have the following fission product contents:
Activity Fission Product (µCi/cc) I-131 0.067 I-132 0.38 I-133 0.40 I-134 0.53 I-135 0.49 Other halogens 0.14 Other fission products 0.28
Measurements of halogen concentrations in the Dresden Unit 1 reactor water and condensate showed that the steam to water hal ogen concentration ratio was in the range of 3 x 10-5 to 1 x 10
-5. The only halogens carried out through the break would therefore be those absorbed in the water. Thus, 116 curies, including 3.4 curies of I-131, 19.1 curies of
I-132, 20.2 curies of I-133, 27.0 curies of I-134, and 24.9 curies of I-135 would be carried out through the break.
Based on operating experience, the above fi ssion product concentrations in the reactor coolant would occur when the off-gas emissi on was at about 100,000 µCi/s measured after 30 minutes decay in the off-gas system. The noble gas activity discharged from the break, assuming a 10.5 seconds MSIV closure time, woul d be 5.4 Ci (calculated for 2 minutes decay time).
QUAD CITIES - UFSAR Revision 4, April 1997 15.6-11 Superheated steam would discharge from the break and exhaust into the turbine building.
At atmospheric pressure the total coolant discharged is assumed to separate to 55,000 pounds of steam and 45,000 pounds of water. The steam would travel upward through the
turbine building, partially condensing on walls and equipment in the building. Pressure in
the turbine building would be released through special sections of siding which have been
installed in both the reactor building and turbine building. No damage to building structural members would occur. A realisti c value for the water-steam mixture which would be lost during a main steam line break accident is 85,000 pounds. The value of
100,000 pounds stated previously is used to calc ulate radiological exposures as a result of the main steam line break accident, the 15,000 pounds additional representing additional
margin of conservatism.
Depending upon the release rate from the turb ine building, the steam cloud could rise to a height of 4600 feet or more in a 2-mp h wind, or 920 feet in a 10-mph wind.
[8] Due to the uncertainty in predicting the actual steam re lease rate the conservative assumption is made that the release occurs at the top of th e turbine building with no additional rise.
Experiments
[9, 10] on halogen partitions indicate that 99 - 99.99% of the halogens would remain absorbed in the water which does not ev aporate. Nevertheless, for this analysis, it was assumed that all of the halogens contained in the low quality steam-water mixture
were released to the environs. The use of these conservative assumptions results in
calculated radiological exposures which are at least four orders of magnitude greater than the actual case, however, even with such a cons ervative approach the radiological doses, as shown in Table 15.6-2 are well below the guidelines set forth in 10 CFR 100.
The break of a main steam line outside both the drywell and the reactor building was also
evaluated by the AEC
[27] during the initial licensing of the plant. For the analysis the MSIVs were assumed to start to close within 0.5 seconds after the steam line break was sensed and fully closed in a maximum time of 5 seconds, the limit established in the
Technical Specifications. During the 5 seco nd closure time, approximately 30,000 pounds of primary coolant would be lost through the brea
- k. The iodine concentration resulting from the total fission product inventory of the pr imary coolant system was assumed to be 20
µCi/ml. In calculating the consequences of the steam line break accident, the AEC followed Safety Guide 5 entitled "Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors." Based on these assumptions the AEC determined that the potentia l offsite 2-hour dose at the site boundary would be 22 rem to the thyroid and less than 1 rem whole body.
[15.6-19]
Subsequent to and separate from the prec eding radiological dose analyses, NRC staff calculations reported showed the resultant radiol ogical dose at the site boundary to be less than 30 rem to the thyroid. This dose was calculated on the basis of the radioiodine
concentration limit of 5 µCi of I-131 dose e quivalent per gram of water, atmospheric diffusion from an elevated release at 30 meters under fumigation conditions for Pasquill
Type F, 1 meter per second wind speed, and a steam line isolation valve closure time of
5 seconds.
[15.6-20]
The Technical Specification limitations on th e specific activity of the primary coolant ensure that the two hour thyroid and whole body doses resulting from a main steam line failure outside the containment during stea dy state operation w ill not exceed small fractions of the dose guidelines of 10 CFR 100. Therefore, the radiological dose
consequences are not increased by fuel type or design changes.
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-11a The radiological consequences for a main steam line break outside of containment are not
increased by the extended power uprate bec ause the mass and energy releases remain the same and the activity released is the reacto r coolant activity at Technical Specification concentrations which are not impacted by EPU.
[END OF HISTORICAL INFORMATION]
QUAD CITIES - UFSAR Revision 14, October 2017 15.6-12 15.6.5 Loss-of-Coolant Accidents Resulting from Piping Breaks Inside Containment A LOCA would result in the heating and pressuri zation of containment, a challenge of the emergency core cooling systems, (ECCS), and the potential release of radioactive material to the environs. The response of the containment to a LOCA is discussed in Section 6.2.1.3.2.
The fuel thermal response and ECCS performance are described in Section 6.3.3, and
additional evaluations referenced in section 4.2.
The following section describes the potential radiological consequences due to a LOCA. De tailed containment and fuel thermal responses for LOCAs have been reanalyzed since original licensing, using updated models, codes, and assumptions. The radiological consequences in the following discussion, are conservative with respect to fuel cladding perforations and sou rce terms. The original UFSAR licensing basis prior to Extended Power Uprate utilized the TID 14844 methodology, which establishes source term based on rated core thermal power.
The impact of Extended Power Uprate on the radiological consequences of a LOCA is addressed at the end of Section 15.6.5.5.
See the introduction to Section 15.6 for informatio n regarding use of details from this analysis description which may not be applic able to the current fuel cycle.
15.6.5.1 Identification of Causes and Frequency Classification
The full range of LOCAs has been analyzed, from a small rupture where the makeup flow is greater than the coolant loss rate, to a highly improbable circumferential recirculation line break. The analysis shows that the circumferen tial recirculation line break would result in the maximum fuel temperature and containment pressure.
[15.6-21]
This event is classified as a limiting fault, i.
- e. an event that is not expected to occur but is postulated because the consequences may result in the release of significant amounts of radioactive material.
[15.6-22]
15.6.5.2 Sequence of Events and Systems Operation
The HPCI, residual heat removal (RHR), and co re spray systems would act to cool the core following the accident.
[15.6-23]
For breaks in small liquid lines up to about 0.12 ft 2 in area, HPCI can supply sufficient coolant to depressurize the vessel and cool the core, de pending only on the core spray system and the LPCI mode of RHR for long-term cooling. For breaks in liquid lines between 0.12 ft 2 and 0.2 ft 2 in area, the depressurizing function of the HPCI and the coolant makeup function of either the core spray subsystem or the LPCI mode of RHR would act in conjunction to provide effective core cooling. In the event of a LO CA without HPCI capability (i.e., if the normal feedwater and HPCI are assumed to be unavail able), the ADS would cause the reactor vessel blowdown to occur in a time interval sufficientl y short to permit operation of the core spray system and LPCI mode of RHR to assure adequate core cooling.
[15.6-24]
QUAD CITIES - UFSAR Revision 7, January 2003 15.6-13 For breaks in liquid line larger than about 0.2 ft 2 in area, where no depressurization assistance is required, such as the design basis recirculation line break described previously, the core spray subsystem in combination with the LPCI mode of the RHR system would be capable of cooling the core independently of the HPCI or ADS.
The coolant lost through the rupture would be co ndensed in the pressure suppression pool, thus reducing primary containment pressure. En ergy would be removed from the pressure suppression pool by the containment cooling system.
[15.6-25]
15.6.5.3 Core and System Performance
The methodology to analyze the consequences of a LOCA depends upon the particular break size and the location being evaluated. The analyses for LOCAs originally were performed using calculational models and techniques different fr om the models and techniques currently used.
Since the original analysis, additional information and results of tests related to the performance of the ECCS systems also have become available.
[15.6-26]
Section 6.3.3 and additional evaluations refere nced in section 4.2 discuss the fuel thermal response, the ECCS performance, and the current analysis models.
Section 6.2.1.3.2 discusses the containment and coolant blowdown responses.
For 10 CFR 50 Appendix K analyses, the changing thermal and hydraulic phenomena that are associated with a design basis LOCA may be des cribed in five phases: (a) temperature changes and heat removal during reactor blowdown with associated flow coastdown, (b) achievement of critical heat flow at any point on the fuel rod clad ding and associated temperature rise of fuel and clad material, (c) lower plenum flashing caus ing a temporary resurgence of core flow, (d) temperature rise of fuel and cladding with di minished cooling and complete depressurization, and (e) temperature changes and heat removal during ECCS operation.
The first phase of the LOCA is the short-term bl owdown period during which energy is removed from the core by coolant passing through the co re and exiting through the postulated break, causing the reactor coolant system pressure to decre ase rapidly. Initial conditions are nearly the same as during normal operation and nucleate boiling continues undisturbed.
A short time later, the core flow and system pressure decrease sufficiently that nucleate boiling cannot be sustained. The short-term blowdown phase ends when the coolant flow through the core reaches the inlet of the jet pump.
In the third phase of the LOCA, lower plenum flashing may occur. This is a flow phenomenon
during the blowdown wherein a sudden transient in crease in the core flow begins a few seconds after the core flow has decayed to near zero.
The increase in core flow results when the liquid level in the vessel drops below the recirculation line suction nozzles causing the flow out the
break to change from a liquid phase to a steam p hase and increasing the rate of depressurization of the system.
This rapid depressurization results in a rapidly changing thermodynamic state of the fluid in the primary system. Because the fluid in the lower plenum beneath the core was initially in the subcooled state, it does not c hange thermodynamic state during ea rly blowdown as does the rest of the fluid system; however, when the system pressure QUAD CITIES - UFSAR Revision 14, October 2017 15.6-14 decreases to the level where this fluid flashes to steam a large increase in steam flow through the core results. This period of the LOCA is called "lower plenum flashing". Calculation of flows, temperatures and pressures during this phase depends on the knowledge of the flashing process, the effect of flow maldistribution, th e resistance to flow of a two-phase mixture through the core and jet pump diffusers, and the rate of blowdown through the break.
Following the period of lower plenum flashing, heat generation, produced by the radioactive decay of the fission products, and thermal radiatio n among the fuel rod causes the core to heat up.
Although the loss of water level or the increase in drywell pressure resulting from a pipe break is sensed immediately and the ECCS is signaled to start, the actual injection of water by the low pressure systems does not occur for about 30 se conds. This time is required for the diesel generators to start and accept load, the re actor pressure to fall below the ECCS pump discharge pressure and the ECCS pumps to achi eve full flow. Water is injected into the reactor through both the LPCI system and the core spray system.
15.6.5.4 Barrier Performance
[START HISTORICAL INFORMATION]
The discussion in this subsection represents evaluations performed for initial licensing.
Results of extensive LOCA experimental programs since 1974 have demonstrated the conservatism which the LOCA models have with respect to modeling the vessel inventory, inventory distribution, and core heat transfer.
A new thermal-hydraulic model (SAFER) and a new fuel rod thermal-mechanical model (GESTR-L OCA) have been developed to provide more realistic calculations for LOCA analyses.
Analysis of LOCA conditions using the SAFER-GESTR
[12] codes results in lower values; hence, the following discussion of dose calculations are considered conservative.
[15.6-27]
The following calculations are for 7x7 arrays.
The original UFSAR licensing basis prior to Extended Power Uprate utilized the TID 14844 me thodology, which establishes source term based on rated core thermal power. The radiolog ical release is unchanged from 7x7 to 8x8 to 9x9 to 10x10 fuel assemblies. The impact of Extended Power Uprate on radiological consequences is discussed at the end of relevant sections.
Section 15.6.5.4.1 presents historical informatio
- n. The radiological consequences following a LOCA have been reassessed using a new accident source term in accordance with 10 CFR 50.67, "Accident Source Term." 10 CFR 50.67 provides a regulatory mechanism for power reactor licensees to voluntarily replace the tradi tional accident source term used in design basis accident analyses (i.e., TID 14844, Reference
- 45) with an Alternative Source Term (AST).
Since the publication of TID 14844, significant advances have been made in understanding the timing, magnitude and chemical form of fissi on product releases from severe accidents.
The methodology for this approach is provided in NRC Regulatory Guide 1.183 (Reference 46).
The radiological assessment for AST is provided in Section 15.6.5.5.
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-14a 15.6.5.4.1 Historical Information Fission Product Release from the Fuel
Calculations performed for initial licensing show that about 8% of the fuel rods in the core might experience cladding perforation, based on a 1500°F perforation temperature; however, the conservative assumption is made that 15%
of the fuel rods would experience cladding perforation. The thermal analyses also show that none of the fuel would reach melting conditions. A maximum of 1% of the noble gas activity and 0.5% of the halogen activity
contained in a fuel rod is in the plenums and would be available for release if the cladding were to be perforated. Negligible solid or pa rticulate activity would be released from the perforated rods. The amount of the total rea ctor fission product inventory released from the fuel would be about 0.15% of the noble gases and about 0.075% of the halogens. The release
would occur as the cladding is perforated.
[15.6-28]
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-15 Fission Product Release to the Drywell
The fallout and plateout of fission products with in the reactor vessel and piping would reduce the amount of fission products available for tr ansport to the drywell. Of the halogens that would be released from the fuel, 5% are assume d to be instantly converted to organic halides, principally methyl iodide.
Because organic halogens are both less soluble in water and more difficult to filter than uncombined halogens, a conservatively large fracti on of halogens was assumed to be organic.
Fuel melting experiments
[13,14,15]
have shown that 0.1% - 3%
of the released halogens are organic. For the LOCA analysis, 5% of the halogen s released from the fuel are assumed to be organic. This assumption is conservative by a factor of 1.5 to 50.
All organic halogens are assumed to be released and not fallout or plateout. Of the remaining 95% (which are inorganic), 50% will be subject to plateout on metal surfaces. The fallout and plateout in the reactor vessel and piping assumed in the analysis is:
Fallout and Plateout Fission Product Group Percent Noble gases 0 Halogens, organic 0 Halogens, inorganic 50 The pressure suppression pool contains approximately 112,000 ft 3 of water for absorption of halogens. The containment air-to-water volume ratio is about 2.5. All the organic halogens are assumed to remain airborne; although at an air-to-water ratio of 2.5, about half would be expected to be absorbed in water.
[16] In Oak Ridge Reactor (ORR) in-pile UO 2 melting experiments, the condensation of the steam in the gas stream removed essentially all halogens
from the gas stream.
[17] The inorganic halogen partition factor according to Allen would be greater than 10 4.[18] Watson[19] also reports the partition facto r to be greater than 10. These experiments, including both steam condensation in vapor suppression systems and in air, correspond to the conditions accompanying a LOCA.
The initial blowdown through the suppression pool would be mostly air with the trailing phases of blowdown essentially being all st eam. Most fission product release would accompany the final steam release and would be efficiently scrubbed by the condensing steam.
Airborne inorganic halogen and solid fission pr oducts in the drywell wo uld be rapidly removed by the containment spray and steam condensation then mixed with water in the suppression pool. For the accident analysis, a partition factor of 100 for inorganic halogens is used.
Inorganic halogens are assumed to be re-evolved from the water as leakage from the containment reduces the inventory of airborne halog ens. The assumption of a high fraction of organic halogens with no absorption in water and the conservative water to air partition factor for inorganic halogens results in a conservative high fraction of halogens remaining airborne which could leak from the containment. The inventory of airborne fission products in the
drywell which could leak into the reactor building is shown in Table 15.6-3.
The following discussion further defines the assumptions and mechanisms concerning the
time-dependent airborne fission-product inventory (Table 15.6-3).
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-16 The fission product airborne activity in th e primary containment is dependent upon the fraction of contained activity released, plat eout, fallout, washout, and conversion effects appropriate to a given species or group of isotopes. For the LOCA, a maximum of 8% of the fuel rods would be perforated; however, it is conservatively assumed that 15% of the fuel rods are perforated.
Those rods experiencing clad damage are assume d to release 1% of their noble gas and 0.5%
of their iodine activity. Of the iodine activi ty released, 5% is assumed to be instantaneously converted to methyl iodide and the remainin g 95% is assumed to experience a reduction factor of two due to plateout. That fraction of inorganic iodine which escapes plateout is carried to the suppression pool where it is assumed to instantaneously form and maintain
an equilibrium condition with the air in the primary containment. Since the analysis
shows the total number of fuel perforations would occur within approximately 5 minutes after initiation of the accident, the actual mechanism which transports activity from the drywell to the wetwell is unimportant as lo ng as a transport mechanism exists. Less than 2% of the total assumed number of perforations occurs prior to approaching an equilibrium pressure condition between the wetwell and drywell. Therefore, less than 2% of the released activity would be driven to the suppre ssion pool in the initial pressure transient.
The differential pressure between the re actor vessel and drywell would also be approximately zero after 30 seconds. Therefor e, the only mechanisms available for fission product transport out of the reactor vessel wo uld be by diffusion or confinement and transport in the liquid reactor coolant. The iodine activity released to the drywell after the initial pressure transient would therefore be released primarily with the liquid discharging through the break in the recirculation line.
The core spray would be actuated in approxim ately 60 seconds, liquid level in the reactor vessel would be at the top of the jet pumps af ter 125 seconds, and the liquid level would be at the bottom of the core after 180 seconds. Al so, since the core is being sprayed from above by a two-phase mixture, fission product migr ation would be through the jet pumps to the drywell via the water.
Less than 4% of the assumed maximum number of damaged rods will experience cladding damage in the 125 seconds. It can therefore be concluded that the iodine activity would reach the suppression pool and that the mode of transport to the pool is not important.It is irrelevant whether or not the noble gas acti vity reaches the suppression pool as the decontamination effects of water for the noble ga ses is negligible. It is therefore assumed that all of the released noble gases are airborne in the primary containment.
Fission Product Release from Drywell to the Reactor Building
The primary containment leakage rates were calculated assuming that the primary containment leaks 0.5% of the contained free volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 62 psig; the turbulent rough passage equation
[20] was used for interpolation to lower pressures. The long-term primary containment pressure is shown in Section 6.2.1.3.2. The corresponding containment leakage for case d, shown in Figure 6.2-14 represents a highly faulted condition of one core spray, three RHR LPCI pu mps, and one heat exchanger all inoperable.
[15.6-29]
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-17 If fission products leak from the drywell, hi gh pressure or reactor building high radiation signals would isolate the affected zone of th e reactor building and start the standby gas treatment system. The SBGTS fan maintain s the reactor building below atmospheric pressure and discharges a volume equivalent to 100% of the building volume per day through high-efficiency filters and charcoal abs orbers to the 310-foot chimney. The analysis assumed all the noble gases and halogens released into the reactor building remain
airborne.
[15.6-30]
The airborne fission product inventory in the reactor building, which was evaluated considering the leakage from the drywell to th e reactor building, radioactive decay, fallout and plateout, and an air change rate of 100% of the building volume per day, is shown in
Table 15.6-4.
Fission Product Release from Re actor Building to Atmosphere The halogens which leak from the pressure suppression containment into the reactor building are exhausted by the SBGTS through a demister, air heater, roughing filter, charcoal filter, and a high efficiency filter.
The reactor building exhaust air is treated to reduce the humidity so that the filters will be effective for removal of organic halogens.
Tests on filter efficiencies have shown that inorganic halogens are removed by charcoal filters with efficiencies greater than 99.99%.
[21,22] Tests on filter efficiencies have also shown that organic halogens are removed by c harcoal filters at a relative humidity less than 30% with filter efficiencies from 99.9 - 99.9999%.
[22,23,24,25,26]
The charcoal filter on the EVESR at Vallecitos Atomic Power Laboratory retained organic halogens produced at power operation with a filter efficiency from 99.8 - 99.9% at a relative humidity of 10 -
15%. The latest experimental results were used as the basis of the original SBGTS design.
The system is designed to provide the necessary residence time in the filters. Thus, the analysis assumption of 99% filter efficiency fo r the removal of inorganic and organic halides in the SBGTS is conservative by approximately four orders of magnitude.
The compounding of conservative assumptions used in the LOCA analysis results in
calculated doses from halogens that are 20 to 1000 times higher than would actually be expected. Discharge rates from the LOCA to th e elevated release point are shown in Table 15.6-5.
[END OF HISTORICAL INFORMATION]
15.6.5.5 Radiological Consequences
15.6.5.5.1 Application of Alternative Source Term Methodology Regulation 10 CFR 50.67, "Accident Source Te rm," provides a regulatory mechanism for power reactor licensees to voluntarily replace th e traditional accident source term used in design basis accident analyses (i.e., TID 14844, Reference 45) with an "Alternative Source Term" (AST). The methodology for this appr oach is provided in NRC Regulatory Guide 1.183 (Reference 46).
QUAD CITIES - UFSAR Revision 14, October 2017 15.6-17a Accordingly, Quad Cities applied for the AST methodology for key De sign Basis Accidents (DBAs). In support of a full-scope implementation of AST as described in Reference 46, AST radiological consequence are performed fo r the four DBAs that result in offsite exposure consequences: Loss of Coolant Accide nt (LOCA), Main Steam Line Break (MSLB), Fuel Handling Accident (FHA), and Control Rod Drop Accident (CRDA).
The NRC approved AST for Quad Cities in Reference 47.
Fission Product Inventory The inventory of reactor core fission pr oducts is based on maximum full power operation at a power level of 3016 MWth, th e Extended Power Uprate (EPU) thermal power of 2957 MWth plus 2% to account fo r uncertainties in accordance with NRC Regulatory Guide 1.49 (Reference 53). The Westinghouse Optima2 fuel core inventory was analyzed using the RADTRAD computer co de at a core average exposure of 39 GWD/MTU for a 24-month cycle. The AREVA AT RIUM 10XM fuel core inventory was analyzed using the RADTRAD computer code at a core average exposure of 39 GWD/MTU and core average enrichments be tween 3.9 and 4.5 weight percent U-235 (Reference 61). The isotopic inventorie s for primary containment and secondary containment (reactor building) that are used for estimating the radiological consequences are provided in Tables 15.6-5a and 15.6-5, respectively. These tables are based on the Westinghouse Optima2 core inventory, which are bounding for the AREVA ATRIUM 10XM fuel design fo r radiological consequences.
Core Inventory Release Fractions The core inventory release fractions were determined using the guidance in NRC Regulatory Guide 1.183, Section 3.2 (Release Fractions).
Atmospheric Dispersion Factors The main control room and offsite dose /Q values were calculated using the methodology described in UFSAR Section 2.3.6. The atmospheric relative concentrations used for the LOCA analysis are as presented in Table 15.6-9.
Release Paths Timing of Release Phases:
The fission product inventory is assumed to be released in phases following a design basis LOCA. Regulatory Guide 1.183 specifies the onset and
duration of each sequential release phase. The onset of fuel gap release occurs in 2
minutes and is assumed to last 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The early in-vessel phase begins in 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
and lasts for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The inventory of each phase is released at a constant rate over the duration of the phase. Once dispersed in the primary containment, the release to
the environment is assumed to occur th rough the following three pathways.
The radioactive release from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of primary
containment. A containment leakage of 3%
volume per day is assumed, which is the sum of the primary-to-secondary leakage and leakage through the MSIVs. Leakage from primary containment is assumed to mix in 50% of the reactor building free volume which is treated by the standby gas filtration system. Reduction in containment QUAD CITIES - UFSAR Revision 14, October 2017 15.6-17b leakage after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is not credited.
MSIV Leakage:
The MSIVs are postulated to leak at a total design leak rate of 150 scfh (at 48 psig) as follows:
- 60 scfh through a steam line with a failed inboard MSIV; and
- the other three steam lines are assumed to leak at 60 scfh, 30 scfh, and 0 scfh respectively.
The radiological consequences from postulated MSIV leakage are combined with the consequences for the other fission product release paths.
Emergency Core Cooling System Leakage:
Systems that circulate suppression pool water outside primary containment are assumed to leak during their intended operation. This release source includes leakage through valv e packing glands, pump shaft seals, flanged connections, and other similar components. The analysis conservatively assumes leakage begins at the onset of the accide nt and to continue throughout the 30-day duration. The leakage from all components is 1 gpm into the reactor building (the
radiological assessment conservatively assumes 2 gpm).
Source Term Mitigation Reduction in Airborne Activity Inside Containment:
The activity of elemental iodine and aerosols released from the core into the dryw ell is reduced by deposition (i.e., plate-out) and settling in the drywell utilizing the natural deposition values identified in the
RADTRAD code. Iodine removal by suppressi on pool scrubbing is not credited because the bulk core activity is released to cont ainment well after the initial mass and energy release. Containment spray is also not cre dited. The Decontamination Factor (DF) for elemental iodine is consistent with SRP 6.5.2 and is limited to a DF 200.
Aerosol Deposition in Main Steam Lines:
Main steam line pipe deposition was modeled using the RADTRAD code with removal coeffi cients based on gravitational settling.
Two-node treatment is used for each steam lin e in which flow occurs. No credit is taken for holdup or plate-out in the main steam lines beyond the outboard MSIV. Only
horizontal sections of piping are credited.
Additionally, no credit is taken for holdup and plate-out in the main condenser. Main steam line deposition is based on using the
shortest line (i.e., most rapid transport) for the worst case line (i.e., the one with the assumed failed inboard isolation valve).
Ventilation Cleanup Systems:
Containment leakage into th e reactor building is collected by the standby gas treatment (SGT) System which exhausts the reactor building, via filters, and reduces releases. The SGT exhaust charcoal and HEPA filter efficiencies are assumed to be 80% for elemental iodine, 80%
for organic iodide, and 98% for particulate aerosols. The control room emergency ventila tion (CREV) system mitigates dose to the main control room operators. The CREV cha rcoal and HEPA filter efficiencies are assumed to be 99% for elemental iodine, 99%
for organic iodide, and 99% for particulate aerosols. The CREV is assumed to be actuat ed 40 minutes following a LOCA. Prior to initiation, unfiltered inleakage is assume d to be 60,000 cfm. Following actuation, unfiltered inleakage is assumed to be 400 cfm, which includes 10 cfm for normal
ingress/egress.
QUAD CITIES - UFSAR Revision 14, October 2017 15.6-17c Suppression Pool Post-LOCA pH Control:
The guidance in NRC Regulatory Guide 1.183 provides that the iodine species released to containment include 95% cesium iodide, 4.85% elemental iodine, and 0.15% iodine in organic forms. This assumption is valid only if the suppression pool water is maintained at a pH of 7.0 or higher to ensure
against the re-evolution of elemental iodine. The standby liquid control system is used
to control pH by injecting sodium pent aborate. Following a DBA LOCA, sodium pentaborate is injected into the RPV where it will be mixed with ECCS flow and migrate to the suppression pool. Credit for standby liquid injection is based on operation of one pump, manually initiated and injected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a DBA LOCA (assumed to be injected within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of an accident where there is indication of fuel
damage).
Dose Consequences As per NRC Regulatory Guide 1.183, Total E ffective Dose Equivalent (TEDE) offsite doses are determined as the sum of the co mmitted effective dose equivalent (CEDE) from inhalation and deep dose equivalent (DDE) from external exposure from radionuclides. Dose conversion factors ar e taken from Federal Guidance Report No. 11 (Reference 50) and Federal Guidance Report No. 12 (Reference 51). Control room dose consequences are determined assuming a buildup of contamination in the control room.
Included in this assessment is radiation shine from external sources (i.e., the
radioactive plume and secondary containmen t) and sources inside the main control room (due to the buildup of radionuclides).
No credit is taken for issuance of potassium iodide pills or respirators.
15.6.5.5.2 Acceptance Criteria
The AST acceptance criteria for postulated majo r credible accident scenarios are provided by 10 CFR 50.67 and Regulatory Guide 1.183. For the main control room, adequate
radiation protection is provided to perm it access to and occupancy under accident conditions without personnel rece iving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the du ration of the accident. The AST acceptance criteria for an individual located at any point on the boundary of the exclusion area (the Exclusion Area Boundary or EAB) are provided by 10 CFR 50.67 as 25 rem TEDE for any 2-hour period following the onset of the post ulated fission product release. The AST acceptance criteria for an individual located at any point on the outer boundary of the low population zone (LPZ) are provided by 10 CFR 50.67 as 25 rem TEDE during the entire
period of passage of the radioactive cloud re sulting from the postulated fission product release.
15.6.5.5.3 Computer Models
AST calculations for the DBA LOCA were prepar ed to simulate the radionuclide release, transport, removal, and dose estimates associat ed with the postulated accident scenario.
Source term calculations were determin ed using the ORIGEN2 computer code for Westinghouse Optima2 fuel and ORIGEN-S for AREVA ATRIUM 10XM fuel (References 54 and 62). The ORIGEN2 and ORIGEN-S computer codes were designed fo r reactor fuel cycle mass and radioactivity inventory calculatio ns. The ORIGEN2 and ORIGEN-S computer codes are widely recognized for calc ulating fission product inventories.
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-17d Offsite /Qs were calculated with the PAVAN computer code (Reference 56) using the guidance of NRC Regulatory Guide 1.145 (Reference 57); control room /Qs were calculated with the ARCON96 (Reference 58) and PAVAN computer codes consistent with the guidance of Regulatory Guide 1.194 (Referen ce 59). The PAVAN and ARCON96 codes are generally used to calculate relative concentra tions in plumes from nuclear power plants at offsite locations and control r oom air intakes, respectively.
The RADTRAD computer code (Reference 60) wa s used for determining dose consequences for the DBA LOCA. The RADTRAD program is a radiological consequence analysis code used to estimate post-accident doses at plant offsite locations and in the control room.
15.6.5.5.4 Key Plant Assumptions The key input assumptions are presented in Table 15.6-7.
15.6.5.5.5 Results The reanalysis of the LOCA accident event using AST methodology is documented in calculation QDC-0000-N-1481 (Reference 61). The radiological consequences are presented in Table 15.6-8a.
Section 15.6.5.5.6 describes the radiological a ssessment performed prior to the adoption of an updated accident source term in accordanc e with 10 CFR 50.67, "Accident Source Term."
15.6.5.5.6 Historical Information The discussion in this subsection represents evaluations performed for initial licensing.
Analysis of LOCA conditions using the SAFER-GESTR
[12] codes results in lower values; hence, the following discussion of dose calc ulations are considered conservative. The original UFSAR licensing basis prior to Extended Power Uprate utilized the TID 14844 methodology, which establishes source term based on rated core thermal power. The impact of Extended Power Uprate and the use of SVEA-96 Optima2 fuel on the radiological consequences is discussed at the end of relevant sections.
[15.6-31]
QUAD CITIES - UFSAR Revision 11, October 2011 15.6-18 Offsite Dose Rates
The offsite radiological effects of a design bas is LOCA for the initial core (7x7 arrays) based on the preceding barrier performance analysis and 99% SBGTS efficiency are shown in Table 15.6-6. These doses are far below the guideline radiation doses listed in 10 CFR 100.
The original UFSAR licensing basis prior to Extended Power Uprate utilized the TID 14844 methodology, which establishes source term bas ed on rated core thermal power. This table was valid before the Extended Power Uprate, re gardless the change of fuel design from 7x7 to 8x8 to 9x9 to 10x10 fuel assemblies. It is retained for historical information.
[15.6-32]
The consequences of the postulated desi gn basis LOCA were calculated by the AEC
[27] using the conservative assumptions presented in the Commission's Safety Guide 3 entitled "Assumptions Used for Evaluating the Pote ntial Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Re actors". For the AEC analysis, the primary containment was assumed to leak at a constant rate of 1.3% of the containment volume per day at accident conditions for the duration of the accident without consideration of the effects of post accident decreases in pressure. The leakage was assumed to pass directly to the SBGTS (assuming 90% halogen removal effi ciency) without mixing with the reactor building atmosphere and then to the 310-foot chimney. The results of the AEC analysis
indicate 2-hour doses at the site boundary of 6 rem for the whole body, and 150 rem to the thyroid. [15.6-33]
An evaluation of the design basis LOCA was based on the primary containment maximum
allowable accident leak rate of 1.0%/day at 48 psig. The analysis showed that with this leak rate and a standby gas treatment system filter efficiency of 90% for halogens and 95% for particulates, and assuming the fission product release fractions stated in TID 14844, the
maximum total whole body passing cloud dose is about 5 rem, and the maximum total
thyroid dose is about 120 rem at the site bound ary over an exposure duration of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The resultant doses that would occur for the dura tion of the accident at the low population distance of 3 miles are lower than those stated due to the variability of meteorological conditions that would be expected to occur ov er a 30-day period. Thus, the doses reported are the maximum that would be expected in the unlikely event of a design basis LOCA.
These doses are also based on the assumption of no holdup in the secondary containment
resulting in a direct release of fission produ cts from the primary containment through the filters and stack to the environs. Therefore, the specified primary containment leak rate and filter efficiency are conservative and pr ovide margin between expected offsite doses and 10 CFR 100 guidelines.
[15.6-34]
The Extended Power Uprate increases the 2 ho ur thyroid dose from 120 rem to 152 rem and the whole body dose from 5 rem to 6 rem.
An evaluation of the post-LOCA offsite doses for SVEA-96 Optima2 fuel shows that the existing EPU doses are either bounded, maintained, or increased slightly
[44]. The SVEA-96 Optima2 doses are shown in Table 15.6-8a.
Thus, there is adequate margin in the design of the reactor and containment to limit the
consequences of large postulated accident s and still adequately protect the public.
[15.6-35]
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-18a Site Dose Rates
The radiological effects in the reactor buildin g a week after a design basis LOCA would be greater than 500 R/hr, restricting access to sho rt durations for life saving purposes only.
The turbine building would be accessible on a limited basis depending on location.
[28] [15.6-36]
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-19 Control Room Dose Rates
Subsequent to the offsite and site dose analyses , a control room dose analysis was performed in accordance with the guidance of NUREG 0737
[29] Item III.D.3.4 to determine compliance with the radiological requirements of General Design Criterion (GDC) 19 and Standard Review Plan (SRP) 6.4.
[30] The LOCA was considered in the analysis to be the radiological design basis accident (DBA). The results of this analysis are considered conservative. Several natural mechanisms will reduce or delay the radioactivity prior to release to the environment. Credit was taken only for iodine plateout on surfaces of the steam lines and condenser and radioactive decay prior to release.
[15.6-37]
Methodology
The guidelines given in SRP 6.4
[30] and Regulatory Guide 1.3
[31] have been used with the exceptions of the chi/Q for the control room and plateout of iodines during transportation within pipes. Realistically, the components of the main steam lines and the turbine-condenser complex
would remain intact following a design basis LOCA.
Therefore, plateout of iodines on surfaces of the main steam lines and the turbine-condenser complex would be expected.
Figure 15.6-7 shows the radiological control room model used for activity released through the SBGTS and through the MSIVs. The total control room 30-day integrated dose would be equal to the sum of the two dose models. The input paramete rs used to develop the activity levels in the control room are shown on Table 15.6-7.
[15.6-38]
Assumptions and Bases
[31] has been used to determine activity levels in the containment following a design basis LOCA. Activity releases are based on a containment leakage rate of 1% per day.
Table 15.6-7 lists the assumptions and parameters used in the analysis and dose point locations.
The majority of the containment leakage will be collected in the reactor building and exhausted to the atmosphere through the SBGTS as an elevated release from the main stack. Any SBGTS bypass leakage has been quantified by assuming t hat all MSIVs leak 11.5 scfh per main steam line when tested at 25 psig (Technical Specificat ion limit for all MSIV leakage paths is less than or equal to 46 scfh). The leak-rate was correcte d to the containment design pressure using the laminar flow extrapolation factors of ORNL NSIC-5
[41]. [15.6-39]
Leakage past the isolation valves could be rele ased through the outboard MSIV stems into the steam tunnel, or continue down the steam lines to the stop valves and into the turbine-condenser complex. The steam tunnel is exhausted by the SBGT S filtration system, thus eliminating it as a bypass pathway. The MSIV leakage travels down the steam piping to the turbine-condenser complex where it is released as a ground level re lease at a rate of 1% of the turbine-condenser volume per day. This leak rate is consistent with the assumptions used for the control rod drop accident in SRP 15.4.9.
[32] This assumption is conservative since the volumetric leakage from the condenser at 1% per day is greater than the MSIV leakage from the drywe ll. The MSIV leakage passes through three different volumes which prov ide holdup and plateout. The first volume consists of the steam lines between the inboard and outboard isolation valves, the second volume
consists of the steam lines between the outboard isolation valves and the turbine QUAD CITIES - UFSAR Revision 9, October 2007 15.6-20 stop valves, and the third volume includes the steam lines after the turbine stop valves and the turbine-condenser complex. The leakage pa th was conservatively treated as a single volume with a volume of 1.7 x 10 5 ft 3 and a surface area of 6.5 x 10 5 ft 2. The iodine removal rates were calculated for elemental and particula te iodines using a deposition velocity of 0.012 cm/s. The removal of organic iodine throug h plateout is not credited. Elemental and particulate iodine decontamination factors of over 100 can be calculated for the small travel distances and large travel times down the steam lines, refer to NUREG/CR-009 Section 5.1.2[33].
The MSIVs will leak to the turbine building which would be exhausted by the heating, ventilating, and air conditioning (HVAC) system if it was operational. Additional plateout on ductwork, fans, and unit coolers would further minimize the iodine releases. If the HVAC system was not operational, then any bypass flow would tend to collect in the building and be subject to additional decay and plateou t which are not credited in the analysis.
The activity which enters the main control room may be the result of bypass leakage, the
SBGTS exhaust in the outside air, or both, depend ing on wind direction. It is possible for the control room intake to be expose d to activity from both sources at the same time. Because the SBGTS exhaust is elevated, the concentrations from this source at the intake will be less than those due to bypass leakage. It is conservative ly assumed that the activity concentration at the intake is due to concurrent bypass leakage and chimney releases for the duration of the event.
Infiltration Analysis
During emergency operation, the control room ventilation system supplies 1800-2200 scfm of outdoor air to maintain the control room at 1/8 in w.g. positive overpr essure with respect to the adjacent areas. Intentionally admitting outdoor air into the habitable zone prevents infiltration through the habitable zone boundary by assuring that air is exfiltrating from the zone at a fairly significant velocity (a velocity through the habitable zone boundary penetrations of approximately 1400 ft/min is require d to develop a backpressure of 1/8 in w.g.).
Prior to start of the AFU (during the first 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 50 minutes of the accident), 1000 cfm is
assumed to infiltrate through the zone boundary based on SRP 6.4
[30] which recommends a base infiltration rate equal to 50% of the makeup rate required to pressurize the zone to 1/8 in w.g. [15.6-40]
The infiltration of unfiltered air into the control room habitable area is assumed to enter
through three different pathways s ubsequent to emergency isolation:
[15.6-41]
- 1. Through the habitable zone boundary;
- 2. Through the system components located outside of the habitable zone; and
- 3. Through backflow which could occur when the control room doors are opened.
In accordance with SRP 6.4
[30], the infiltration through the zone boundary is assumed to be zero when the system is pressurized.
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-21 Infiltration through the components locate d outside the habitable zone occurs through joints and seams in the ductwork, damper shafts, joints and penetrations in the air
handling units and the dampers that isolat e the habitable zone from the nonhabitable areas. Figure 6.4-1 identifies the system components which are located outside of the
habitable zone that are under a negative pr essure. The inleakage through the damper shafts and blades was based on damper specif ication requirements and vendor test data.
The Train A ductwork and air handling unit was assumed to leak at the maximum adjusted
leakage rate provided in ANSI N-509 of, 0.2 cfm/ft 2 at 10 in H
{2}O. The Train A ductwork leak rate was used in the analysis because it was more conservative than the Train B ductwork leak rate of 0.005 cfm/ft 2 at 10 in H 2 O gauge.
The opening and closing of boundary doors can induce infiltration to the habitable zone.
Standard Review Plan 6.4
[30] recommends that backflow through the control room doors be assumed to be 10 cfm.
The infiltration analysis resulted in a total unfiltered infiltration rate of 260 cfm. A
breakdown of the infiltration through the different leakage paths is as follows:
Leakage Path Infiltration (cfm) Through ductwork 150.1 Through equipment housings 32 Through isolation dampers 45.9 Through damper shafts 21.3 Through zone doors (backflow) 10 Through habitable zone boundary 0 Total infiltration 259.3 Atmospheric Dispersion Factor (chi/Q) The following discussion is an explanation of the reasons for the use of the Halitsky chi/Q methodology and a value of K
{C} = 2, instead of the Murphy methodology
[34] which SRP 6.4
[30] suggests as an interim position.
[15.6-42] Historically, the preliminary work on building wake chi/Qs was based on a series of wind tunnel tests by J. Halitsky, et al.
[35]. In 1974, K. Murphy and K. Campe of the NRC published their paper based on a survey of existing data. This chi/Q methodology, which presented equations without derivation or justification, was adopted as the interim methodology in SRP 6.4 in 1975. Since then, a series of actual building wake chi/Q measurements have been conducted at Rancho Seco
[36] and several other papers have been published documenting the results of additional wind tunnel tests.
QUAD CITIES - UFSAR Revision 4, April 1997 15.6-22 Murphy[34] suggested the following equation for the calculation of chi/Q: chi/Q = K{C}/AU where
K{C} = K + 2
K = 3/(s/d) 1.4 A = Cross-sectional area of the building
U = Wind speed
This formulation was derived from the Halitsky
[35] data in Figure 37 from Murphy's paper.
[34] The Halitsky data were from wind tunnel test s on a model of the EBR-II rounded (PWR type) containment and the validity of the data was limited to 0.5 < s/d <3. The origin and reason for
the +2 in K + 2 is not known. All other formulat ions use K only, and for the situation where K is less than 1, the use of K + 2 imposes an unrealistic limit on the chi/Q.
For Quad Cities, the building complex is composed of square-edged buildings and not a
round-topped cylindrical containment as was us ed in the Halitsky experiments. For an HVAC intake located near the south wall of the control room at elevation 633 feet 0 inches, the intake
will be subject to a building wake caused by a co mbination of the reactor building and the turbine building for any bypass leakage escaping from the turbine building. There will not be any reactor building bypass leakage because the building is kept at a negative pressure by the SBGTS which exhausts to the main chimney.
Because the Murphy methodology could not be applied, a survey of the literature was undertaken. It was found that the Halitsky wind tunnel test data
[35] conservatively overestimated K
{C} values "by factors of up to possibly 10." Given this conservatism, it was felt that the use of a reasonable K
{C} value from the Halitsky data on square-edged buildings should be acceptable. A review of Figure 5.27 from the Halitsky data
[35] resulted in K
{C} values in the 0.5 to 2 range. A value of K
{C} = 2 was chosen to determine a chi/Q for the control room. A building cross-sectional area of 1,550 m
^2^ was conservatively used. This corresponds to a projected area of one reactor building above grade. The use of a 1,550 m 2 area is very conservative because both of the reactor buildings are adjacent to each other and the combined projected area could be larger than the value used. Information from ot her sources, as indicated in the following, has also shown that this should be a conservative value.
In a paper by D. H. Walker,[37] control room chi/Qs were experimentally determined for floating power plants in wind tunnel tests. Different in take and exhaust combinations were considered.
Using the data for intake 6 and stack A exhaust,[37]chi/Q values of 1.77 x 10
-5 and 2.24 x 10
-5 were found after adjusting the wind speed from the 1.5 m/s to 1 m/s. These values are approximately two orders of magnitude lower then the conserva tively calculated value for Quad Cities.
In a wind tunnel test by P. N. Hatcher,[38] a model industrial complex was used to test dispersions due to a wake. Data obtained from these tests show that K C has a value less than 1, and decreases as the test points are moved closer to the structure. In a study to determine optimum stack heights, R. N. Meroney and B. T. Yang
[39] show that for short stacks (6/5 of building height), K C reaches a value of approximately 0.2 and decreases QUAD CITIES - UFSAR Revision 9, October 2007 15.6-23 closer to the building. They concluded that the Halitsky methodology was "overly conservative." These recent experimental tests show that K C = 2 used to determine the chi/Q for Quad Cities is a conservative estimate by at least a factor of 2 and possibly by a factor of 10 or more.
Field tests were made on the Rancho Seco facility,[36] and chi/Q values were obtained. The data indicates that the use of K C = 2 is conservative.
It was concluded that sufficient data and field tests exist to give a reasonable assurance that the chosen chi/Q is a conservative one, over and above the conservatism implied by using the fifth percentile wind speed and wind direction fa ctors. Based on the preceding discussion, the following equation is used in the calculation of chi/Q values.
chi/Q = 2/AU
Mechanisms for Reducing Iodine Releases
The following four mechanisms could result in significant quantities of iodine being removed before they would be released to the envi ronment. However, credit for the plateout mechanisms is the only credit taken in the calc ulation of radiological consequences.
- 1. Drywell sprays, suppression pool-to-ai r partitioning and condensation effects - The drywell sprays would reduce the iodine sou rce term if actuated. Even without the spray system, condensation would occur in the drywell and suppression chamber.
The iodines in the air and suppression pool would be expected to reach equilibrium due to this phenomenon. Because the iodines have a preference to remain in
solution due to the equilibrium partition factor of over 400 established by the
physical conditions in the containment, the iodines available for release by air leakage would be reduced significantly. The NSAC-14, Workshop on Iodine Releases in Reactor Accidents has indicate d that the iodine release assumption may be overly conservative. Most of the iodine may be released as cesium iodide instead of elemental iodine. The cesium iodide has a much higher solubility and tendency
to plateout than elemental iodine.
- 2. Plateout - The plateout removal constant used in this analysis is based on the lowest deposition veloci ty quoted in NUREG/CR-009.
[33] The NOAA's Technical/Memorandum, "Rancho-Seco Bu ilding Wake Effects on Atmospheric Diffusion,"[36] indicates that the deposition velociti es could be higher by a factor of 4, which would tend to increase the plateout.
- 3. Removal through valves and leakag e holes - Because the bypass leakage paths would be through minute holes in valves and valve seats, the leakage would be
subjected to filtration effects. Larger particulates could tend to plug the leak paths.[40]
- 4. Condensate within pipes - Condensati on would occur within the pipes when the pipes cool down to ambient temperature.
This could result in removal of iodines and particulates from the gas phase.
QUAD CITIES - UFSAR Revision 9, October 2007 15.6-24 Results The 30-day control room doses using the inputs in Table 15.6-7 with the control filter unit
initiated at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 50 minutes are shown in Table 15.6-8.
[43] [15.6-43]
The extended power uprate increases the calcul ated control room thyroid dose by 30%, and the whole body and beta doses by 20%. The post-EPU doses are shown in Table 15.6-8a and are within the regulatory limits.
An evaluation of the post-LOCA control room doses for Optima2 fuel shows that the existing EPU doses are either bounded , maintained, or increased slightly
[44]. The Optima2 doses are shown in Table 15.6-8a.
[END OF HISTORICAL INFORMATION]
QUAD CITIES - UFSAR Revision 10, October 2009 15.6-25 15.6.6 References
- 1. Letter from W.A. Paulsen (NRC) to L.
Del George (CECo), January 4, 1982,
Subject:
Dresden 2 SEP Topic XV-15, Inadvertent Opening of a BWR Safety/Relief Valve.
- 2. Letter from D. Crutchfield (NRC) to W.
Counsil (Northeast Nuclear Energy Co.), October 28, 1981,
Subject:
SEP Topic XV-15 for Millstone 1 (Docket 50-245).
- 3. Letter from D. Crutchfield (NRC) to I.
Finfrock (Jersey Centra l Power & Light Co.), December 4, 1981,
Subject:
SEP Topic XV-15 for Oyster Creek (Docket 58-219).
- 4. Moody, F. J., "Maximum Flowrate of a Single Component. Two Phase Mixture." Journal of Heat Transfer, Trans. ASME Series C, Vol. 87, p 134.
- 5. Wilson, J.F., et al., "The Velocity of Ri sing Steam in a Bubbling Two Phase Mixture." ANS Transactions, Vol. 5, No. 1, p 151, 1962.
- 6. Moody, F.J. "Liquid-Vapor Action in a Vessel During Blowdown,"APED-5177, June 1966.
- 7. Moody, F.J., "Maximum Two Phase Vessel Blowdown from Pipes, ASME Paper No.
65-WA/HT-1.
- 8. Singer, I. A., Frijjolo, J. A. and Smith, M. E. "The Prediction of the Rise of a Hot Cloud from Field Experiments," Journal of the Ai r Pollution Control Association, November 1964.
- 9. Watson, L.C., et. al., "Iodine Containment by Dousing in NPD-II," AECL 1130, October 1960.
- 10. Diffey, H.R., et. al., "Iodine Cleanup in a Steam Suppression System," International Symposium on Fission Product Release and Transport Under Accident Conditions, April 1956.
- 11. Deleted.
- 12. Deleted QUAD CITIES - UFSAR Revision 7, January 2003 15.6-26 13. Collins, D.A., et al., International Symposium on Fission Product Release and Transport Under Accident Conditions, Oak Ri dge Tennessee Paper 59, April 1965.
- 14. Parker, et al., Volume II, "Fission Pr oduct Release," SIFTOR Draft, Chapter 18.
- 15. Collins, R.D., and Hillary, International Symposium on Fission Product Release and Transport Under Accident Conditions, Oa k Ridge, Tennessee Paper 44, April 1965.
- 16. Diffey, et al., International Symposium on Fission Product Release and Transport Under Accident Conditions, Oak Ridge, Tennessee Paper 41, April 1965.
- 17. Miller, et al., International Symposium on Fission Product Release and Transport Under Accident Conditions, Oak Ridge, Tennessee Paper 12, April 1965.
- 18. Allen, T.L., and Keefer, R. M., "The Form ation of Hypoidus Acid and Hydrated Iodine Cation by the Hydrolysis of Iodine," JACS 77, No. 11, June 1955.
- 19. Watson, Bancroft, and Hoelke, "Iodine Containment by Dousing in NPD-II," AECL-1130, 1960.
- 20. Maccary, R.L., et al., "Leakage Characteri stics of Steel Containment Vessels and the Analysis of Leakage Rate Determinations." TID-20483, May 1964.
- 21. Keiholts, G.W., and Barton, C.J., "Behavior of Iodine in Reactor Containment Systems," ORNL-NSIC-4, p 64, February 1965.
- 22. Adams and Browning, International Sy mposium on Fission Product Release and Transport Under Accident Conditions, Oak Ridge, Tennessee, Paper 46, April 1965.
- 23. Collins, R.D., and Hillary, International Symposium on Fission Product Release and Transport Under Accident Conditions, Oak Ridge, Tennessee, Paper 44, April 1965.
- 24. Collins, and Eggleton, "Behavior of Iodi ne in Reactor Containment Systems," ORNL-NSIC-4, p 65 February 1965.
- 25. Adams and Browning, "Behavior of Iodine in Reactor Containment Systems," ORNL-NSIC-4, p 65 February 1965.
- 26. Collins, D.A., et al., International Symposium on Fission Product Release and Transport Under Accident Conditions, Oak Ri dge, Tennessee Paper 45, April 1965.
QUAD CITIES - UFSAR Revision 4, April 1997 15.6-27 27. U.S. Atomic Energy Commission, Division of Reactor Licensing, Safety Evaluation for Quad Cities Station, Units 1 and 2, Docket Nos. 50-254 and 50-265, August 25, 1971.
- 28. Brtis, and Lanti, "Post-Accident Radiation Levels," Sargent & Lundy Engineers, Chicago, IL Quad Cities Station Project Number 5954-00 December 1979.
- 29. U.S. Nuclear Regulatory Commission, "Clari fication of TMI Action Plan Requirements," NUREG 0737, November 1980.
- 30. U. S. Nuclear Regulatory Commission, "Stand ard Review Plan 6.4 for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 6.4, "Habitability Systems", NUREG 0800 Rev. 1, December 1978.
- 31. U. S. Nuclear Regulatory Commission Re gulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Conse quences of a Loss-of-Coolant Accident for Boiling Water Reactors," Rev. 2, June 1974.
- 32. U. S. Nuclear Regulatory Commission Stand ard Review Plan for the Review of Safety Analysis Reports for Nuclea r Power Plants, Section 15.4.9 "Spectrum of Rod Drop Accidents" (BWR), NUREG-0800 Rev. 1, December 1978.
- 33. "Technological Basis for Models of Sp ray Washout of Airborne Contaminants in Containment Vessels," NUREG/CR-009, Posta, A.K., Sherry, R.R., Tam, P.S., October 1978.
- 34. Murphy K.G. and Campe, K.M. Nuclear Po wer Plant Control Room Ventilation System Design for Meeting General Design Criter ion 19, 13th AEC Air Cleaning Conference.
- 35. Slade, D. H. ed., Meteorology and Atomic Energy, TID, 24190 (1968).
- 36. Start, G.E., Cate, J.H., Dickenson, C.R., Ri cks, N.R., Ackerman, G.H., and Sagendorf, J.F., "Rancho-Seco Building Wake Effects on Atmospheric Diffusion," NOAA Technical Memorandum, ERL ARL-69, 1977.
- 37. Walker, D.H., Nassano, R.N., Capo, M.A., 1976, Control Room Ventilation Intake Selection for the Floating Nuclear Power Plant 14th ERDA Air Cleaning Conference.
- 38. Hatcher, R.N., Meroney, R.N., Peterka, J.A., and Kothari, K. "Dispersion in the Wake of a Model Industrial Complex," NUREG 0373 1978.
- 39. Meroney, R.N. and Yang, B.T. "Wind Tunnel Study on Gaseous Mixing Due to Various Stack Heights and Injection Rates Above an Isolated Structure," FDDL Report CER 71-72RNM-BTY16, Colorado State University 1971.
- 40. Morewitz, H.A., Johnson, R.P., Nelson, C.T., Vaughn, E.V., Guderjahn, C.A., Hillard, R.K., McCormack, J.D., and Posta, A.K., "Attenuatio n of Airborne Debris from Liquid-Metal Fast Breeder Reactor Accident," Nuclea r Technology, Volume 46, December 1979.
- 41. ORNL NSIC-5, "U.S. Containment Technolo gies", Oak Ridge National Laboratory and Bechtel Corp., August 1965.
QUAD CITIES - UFSAR Revision 14, October 2017 15.6-28 42. "LOCA Break Spectrum Analysis for Quad Cities Units 1 and 2," EMF-96-184(P), December 1996.
- 43. QDC-9400-M-0348, Revision 1, Assessment of Control Room Habitability with Increased SGTS Filter Efficiency, April 28, 1997.
- 44. Calculation QDC-000-N-1020, "Impact of Extended Power Uprate on Site Boundary and Control Room Doses from LOCA and Non-LOCA Events," Revision 001A, Attachment E, "Evaluation for Impact of Westinghouse Optima2 Fuel on Existing UFSAR Design Basis Accidents." 45. U. S. Atomic Energy Commission, Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," 1962.
- 46. NRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
- 47. Letter from M. Banerjee (U. S. NRC) to C. Cr ane (Exelon Corporation), Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments Re: Adoption of Alternative Source Term Methodology," dated September 11, 2006 [SER correction letter: D. Collins (U. S. NRC) to C. Crane (Exelon Corporation), "Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Correction of Safety Evaluation for Amendment Dated September 11, 2006," dated September 28, 2006].
- 48. NUREG-0800 (Standard Review Plan), Section 15.6.4, "Radiological Consequences of Main Steam Line Failure Outside Containment (BWR)."
- 49. NRC Regulatory Guide (Safety Guide) 1.5, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors," March
1971. 50. U.S. Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.
- 51. U.S. Federal Guidance Report No. 12, "Externa l Exposure to Radionuclides in Air, Water, and Soil," 1993.
- 52. QDC-0000-N-1266, "Re-analysis Main Steam Li ne Break (MSLB) Accident Using Alternative Source Term."
- 53. NRC Regulatory Guide 1.49, "Power Levels of Nuclear Power Plants," December 1973.
- 54. ORNL/TM-7175, "A Users' Manual for the ORIGEN2 Computer Code," A. G. Croff, July 1980.
- 55. Deleted.
- 56. Atmospheric Dispersion Code System for Evaluating Accidental Radioactivity Releases from Nuclear Power Stations; PAVAN, Version 2; Oak Ridge National Laboratory; U.S. Nuclear Regulatory Commission; December 1997.
QUAD CITIES - UFSAR Revision 14, October 2017 15.6-29 57. NRC Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," February 1983.
- 58. Atmospheric Relative Concentrations in Building Wakes; NUREG/CR-6331, PNNL-10521, Rev. 1; prepared by J. V. Ramsdell, Jr., C. A. Simmons, Pacific Northwest National Laboratory; prepared for U.S. Nuclear Regulatory Commission; May 1997 (Errata, July 1997).
- 59. NRC Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," June 2003.
- 60. NUREG/CR-6604, "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," April 1998 and Supplement 1, June 1999.
- 61. QDC-0000-N-1481, Revision 2, "Quad Cities Units 1 & 2 Post-LOCA EAB, LPZ, and CR Dose - AST Analysis."
- 62. NUREG/CR-0200, Revision 6, Volume 2, Section F7, "ORIGEN-S: SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms," ORNL, March 2000.
(Sheet 1 of 1) Revision 9, October 2007 QUAD CITIES UFSAR
Table 15.6-1
EFFECT OF MAIN STEAM ISOLATION VALVE CLOSURE TIME (Historical Information)
Steam Line Isolation Valve Closure Time (Seconds) Net Mass of Water and Steam Loss from Pressure Vessel (Pounds) (Includes 0.5 detection time) With Feedwater Without Feedwater 3.5 seconds 10,000 20,000 10.5 seconds 60,000 85,000
(Sheet 1 of 2)
Revision 7, January 2003 QUAD CITIES - UFSAR Table 15.6-2*
RADIOLOGICAL EFFECTS OF THE STEAM LINE BREAK ACCIDENT (INITIAL CORE ANALYSIS) (Original analysis, retained for historical purpose)
First 2 - Hour Dose Total Dose VS-1 MS-1 N-1 N-5 U-1 U-5 VS-1 MS-1 N-1 N-5 U-1 U-5 Distance (Miles)
Passing Cloud Whole Body Dose (rem) 1/4 1.1 x 10
-3 1.1 x 10-3 1.1 x 10-3 2.2 x 10-4 5.8 x 10-4 1.4 x 10-4 1 7.3 x 10
-4 6.8 x 10-4 4.6 x 10-4 8.8 x 10-5 1.1 x 10-4 2.9 x 10-5 3 3.9 x 10
-4 3.1 x 10-4 1.2 x 10-4 2.4 x 10-5 1.6 x 10-5 5.0 x 10-6 Not 5 2.4 x 10
-4 1.8 x 10-4 5.1 x 10-5 1.1 x 10-5 5.7 x 10-6 2.0 x 10-6 A pplicable 10 1.1 x 10
-4 6.7 x 10-5 1.2 x 10-5 3 x 10-6 1.2 x 10-6 5.3 x 10-7 Lifetime Thyroid Dose (rem) 1/4 2.3 x 10
-4 4.0 x 10-1 5.2 x 10-1 8.8 x 10-2 2.4 x 10-1 5.9 x 10-2 1 3.8 x 10
-3 2.6 x 10-1 1.2 x 10-1 3.6 x 10-2 2.3 x 10-2 6.0 x 10-3 3 2.5 x 10
-2 9.6 x 10-2 1.9 x 10-2 5.7 x 10-3 3.1 x 10-3 8.1 x 10-4 Not 5 3.6 x 10
-2 5.1 x 10-2 7.5 x 10-3 2.3 x 10-3 1.2 x 10-3 3.1 x 10-4 A pplicable 10 3.3 x 10
-2 1.9 x 10-2 2.2 x 10-3 6.5 x 10-4 3.3 x 10-4 8.7 x 10-5 Fallout Whole Body Dose (rem) 1/4 1.7 x 10
-7 5.2 x 10-4 1.1 x 10-3 9.0 x 10-4 1.0 x 10-3 1.3 x 10-3 5.7 x 10
-7 1.7 x 10-3 3.6 x 10-3 3.0 x 10-3 3.3 x 10-3 4.1 x 10-3 1 2.8 x 10
-6 3.4 x 10-4 2.5 x 10-4 3.7 x 10-4 9.7 x 10-5 1.3 x 10-4 9.1 x 10
-6 1.1 x 10-3 8.4 x 10-4 1.2 x 10-3 3.2 x 10-4 4.2 x 10-4 3 1.9 x 10
-5 1.3 x 10-4 3.9 x 10-5 5.9 x 10-5 1.3 x 10-5 1.7 x 10-5 6.1 x 10
-5 4.1 x 10-4 1.3 x 10-4 2.0 x 10-4 4.3 x 10-5 5.7 x 10-5 QUAD CITIES - UFSAR Table 15.6-2 RADIOLOGICAL EFFECTS OF THE STEAM LINE BREAK ACCIDENT (INITIAL CORE ANALYSIS) (Cont'd)
(Sheet 2 of 2)
Revision 4, April 1997 Fallout Whole Body Dose (rem) 5 2.6 x 10
-5 6.6 x 10-5 1.5 x 10-5 2.4 x 10-5 5.1 x 10-6 6.7 x 10-6 8.7 x 10
-5 2.2 x 10-4 5.1 x 10-4 7.8 x 10-5 1.7 x 10-5 2.2 x 10-5 10 2.4 x 10
-5 2.6 x 10-5 4.4 x 10-6 6.7 x 10-6 1.4 x 10-6 1.8 x 10-6 7.9 x 10
-5 8.3 x 10-5 1.5 x 10-5 2.2 x 10-5 4.7 x 10-6 6.1 x 10-6
- The Technical Specification limitations on the specific activity of the primary coolant ensure that the two hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100. Therefore, the radiological dose consequences are not in creased by fuel type or design changes.
Meteorology VS-2 Very stable 2 mph MS-2 Moderately stable 2 mph N-2 Neutral 3 mph N-10 Neutral 10 mph U-2 Unstable 2 mph U-10 Unstable 10 mph
(Sheet 1 of 1)
Revision 9, October 2007 QUAD CITIES - UFSAR Table 15.6-2a RADIOLOGICAL EFFECTS OF THE STEAM LINE BREAK ACCIDENT (Alternative Source Term)
Case 1 (normal equilibrium - 0.2
µC/g) Case 2 (iodine spike - 4.0
µC/g) Regulatory Limits: Control Room: 5.0; EAB & LPZ: 2.5 Control Room 5.0; EAB & LPZ: 25 Location Dose (Rem TEDE)
EAB 0.167 3.32 LPZ 0.0168 0.335 CR 0.189 3.79 (Sheet 1 of 1)
Revision 9, October 2007 QUAD CITIES - UFSAR
Table 15.6-3*
LOSS-OF-COOLANT ACCIDENT PRIMARY CONTAINMENT AIRBORNE FISSION PRODUCT INVENTORY (INITIAL CORE)
(Historical Information)
Time After Accident Noble Gases (curies) Halogens (curies) 1 minute 1.6 x 10 6 4.3 x 10 4 30 minutes 5.2 x 10 5 2.7 x 10 4 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 4.7 x 10 5 2.5 x 10 4 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.5 x 10 5 2.4 x 10 4 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 3.3 x 10 5 1.3 x 10 4 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> 3.2 x 10 5 1.2 x 10 4 1 day 2.7 x 10 5 1.0 x 10 4 2 days 2.5 x 10 5 9.3 x 10 3 10 days 7.5 x 10 4 2.4 x 10 3 25 days 1.2 x 10 4 4.7 x 10 2
- 1. The values in this table were determined to be unchanged due to the introduction of ATRIUM-9B fuel since the source terms are based on the TID 14844.
- 2. This table was valid prior to Extended Power Uprate. It is retained for historical purpose.
(Sheet 1 of 1)
Revision 9 October 2007 QUAD CITIES - UFSAR
Table 15.6-4*
LOSS-OF-COOLANT ACCIDENT REACTOR BUILDING AIRBORNE FISSION PRODUCT INVENTORY (INITIAL CORE)
(Historical Information)
Time After Accident Noble Gases (curies) Halogens (curies) 1 minute 8.8 x 10 0 2.4 x 10-1 30 minutes 4.8 x 10 1 2.5 x 10 0 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 8.9 x 10 1 4.8 x 10 0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.6 x 10 2 8.6 x 10 0 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 6.0 x 10 2 2.4 x 10 1 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> 6.6 x 10 2 2.6 x 10 1 1 day 7.4 x 10 2 2.8 x 10 1 2 days 8.3 x 10 2 3.1 x 10 1 10 days 2.0 x 10 2 6.5 x 10 0 25 days 2.5 x 10 1 9.5 x 10-1
- 1. The values in this table were determined to be unchanged due to the introduction of ATRIUM-9B fuel since the source terms are based on the TID 14844.
- 2. This table was valid prior to Extended Power Uprate. It is retained for historical purpose.
(Sheet 1 of 1)
Revision 9, October 2007 QUAD CITIES - UFSAR
Table 15.6-5*
LOSS-OF-COOLANT ACCIDENT FISSION PRODUCT RELEASE RATE FROM CHIMNEY (INITIAL CORE)
(Historical Information)
Time After Accident Noble Gases (curies/sec) Halogens (curies/sec) 1 minute 1.0 x 10
-4 2.8 x 10-8 30 minutes 5.6 x 10
-4 2.9 x 10-7 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.0 x 10
-3 5.5 x 10-7 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.9 x 10
-3 1.0 x 10-6 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 7.0 x 10
-3 2.8 x 10-6 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> 7.7 x 10
-3 3.0 x 10-6 1 day 8.6 x 10
-3 3.2 x 10-6 2 days 9.7 x 10
-3 3.6 x 10-6 10 days 2.4 x 10
-3 7.5 x 10-7 25 days 2.9 x 10
-4 1.1 x 10-7
- 1. The values in this table were determined to be unchanged due to the introduction of ATRIUM-9B fuel since the source terms are based on the TID 14844.
- 2. This table was valid prior to Extended Power Uprate. It is retained for historical purpose.
Sheet 1 of 1 Revision 14, October 2017 QUAD CITIES - UFSAR
Table 15.6-5a
ISOTOPIC CORE INVENTORY (Ci/MWt) (Alternative Source Term)
Isotope Ci/MW t Isotope Ci/MW t Isotope Ci/MW t CO-58 .1529E+03 RU-103 .4310E+05 CS-136 .1953E+04 CO-60 .1830E+03 RU-105 .3077E+05 CS-137 .5073E+04 KR-85 .4609E+03 RU-106 .1890E+05 BA-139 .4973E+05 KR-85M .7427E+04 RH-105 .2901E+05 BA-140 .4807E+05 KR-87 .1436E+05 SB-127 .2974E+04 LA-140 .5172E+05 KR-88 .2022E+05 SB-129 .8819E+04 LA-141 .4542E+05 RB-86 .6465E+02 TE-127 .2957E+04 LA-142 .4376E+05 SR-89 .2715E+05 TE-127M .3979E+03 CE-141 .4542E+05 SR-90 .3747E+04 TE-129 .8687E+04 CE-143 .4244E+05 SR-91 .3382E+05 TE-129M .1290E+04 CE-144 .3780E+05 SR-92 .3647E+05 TE-131M .3945E+04 PR-143 .4111E+05 Y-90 .3846E+04 TE-132 .3846E+05 ND-147 .1814E+05 Y-91 .3481E+05 I-131 .2702E+05 NP-239 .5404E+06 Y-92 .3647E+05 I-132 .3912E+05 PU-238 .2105E+03 Y-93 .4178E+05 I-133 .5537E+05 PU-239 .1247E+02 ZR-95 .4609E+05 I-134 .6101E+05 PU-240 .1257E+02 ZR-97 .4575E+05 I-135 .5172E+05 PU-241 .7493E+04 NB-95 .4642E+05 XE-133 .5305E+05 AM-241 .1326E+02 MO-99 .5106E+05 XE-135 .2195E+05 CM-242 .2606E+04 TC-99M .4476E+05 CS-134 .7990E+04 CM-244 .3349E+03
Sheet 1 of 2 Revision 14, October 2017 QUAD CITIES - UFSAR Table 15.6-5b POST-LOCA REACTOR BUILDING ISOTOPIC INVENTORY Post-LOCA Reactor Building Isotopic Inventory (Ci) Total Containment + ESF Leakage Activity Isotope 0.667 hr 2.0 hr 4.0 hrs 8.0 hrs 16 hrs 24 hrs (Ci) Co-58 9.423E-03 6.373E-01 1.267E+00 9.365E-01 3.895E-01 1.583E-01 3.398E+00 Co-60 1.128E-02 7.633E-01 1.519E+00 1.124E+00 4.691E-01 1.912E-01 4.078E+00 Kr-85 3.672E+01 9.538E+02 3.373E+03 6.830E+03 1.038E+04 1.175E+04 3.332E+04 Kr-85m 5.337E+02 1.128E+04 2.927E+04 3.192 E+04 1.407E+04 4.620E+03 9.169E+04 Kr-87 7.954E+02 9.990E+03 1.187E+04 2.718E+03 5.275E+01 7.627E-01 2.543E+04 Kr-88 1.369E+03 2.568E+04 5.574E+04 4.252E+04 9.172E+03 1.474E+03 1.360E+05 Rb-86 3.441E+00 3.116E+01 5.448E+01 3.958E+01 1.628E+01 6.554E+00 1.515E+02 Sr-89 1.339E+01 9.050E+02 1.799E+03 1.329E+03 5.518E+02 2.239E+02 4.821E+03 Sr-90 1.848E+00 1.250E+02 2.488E+02 1.842E+02 7.686E+01 3.133E+01 6.681E+02 Sr-91 1.589E+01 9.753E+02 1.677E+03 9.274E+02 2.159E+02 4.909E+01 3.861E+03 Sr-92 1.517E+01 7.297E+02 8.706E+02 2.317E+02 1.249E+01 6.581E-01 1.860E+03 Y-90 2.152E-02 2.582E+00 9.929E+00 1.482E+01 1.207E+01 7.124E+00 4.655E+01 Y-91 1.720E-01 1.181E+01 2.416E+01 1.874E+01 8.273E+00 3.469E+00 6.663E+01 Y-92 5.195E-01 1.540E+02 5.395E+02 3.886E+02 6.000E+01 6.570E+00 1.149E+03 Y-93 1.968E-01 1.215E+01 2.108E+01 1.186E+01 2.858E+00 6.729E-01 4.883E+01 Zr-95 2.272E-01 1.537E+01 3.055E+01 2.258E+01 9.386E+00 3.812E+00 8.192E+01 Zr-97 2.195E-01 1.406E+01 2.578E+01 1.620E+01 4.869E+00 1.430E+00 6.256E+01 Nb-95 2.289E-01 1.549E+01 3.082E+01 2.282E+01 9.519E+00 3.880E+00 8.276E+01 Mo-99 3.126E+00 2.086E+02 4.064E+02 2.885E+02 1.107E+02 4.148E+01 1.059E+03 Tc-99m 2.760E+00 1.865E+02 3.695E+02 2.686E+02 1.073E+02 4.153E+01 9.762E+02 Ru-103 2.656E+00 1.795E+02 3.567E+02 2.633E+02 1.092E+02 4.426E+01 9.556E+02 Ru-105 1.709E+00 9.393E+01 1.368E+02 5.423E+01 6.490E+00 7.588E-01 2.939E+02 Ru-106 1.165E+00 7.882E+01 1.568E+02 1.161E+02 4.840E+01 1.972E+01 4.210E+02 Rh-105 1.789E+00 1.206E+02 2.369E+02 1.678E+02 6.171E+01 2.172E+01 6.106E+02 Sb-127 3.649E+00 2.444E+02 4.791E+02 3.442E+02 1.353E+02 5.193E+01 1.259E+03 Sb-129 9.770E+00 5.338E+02 7.706E+02 3.003E+02 3.471E+01 3.920E+00 1.653E+03 Te-127 3.638E+00 2.450E+02 4.841E+02 3.525E+02 1.424E+02 5.617E+01 1.284E+03 Te-127m 4.906E-01 3.320E+01 6.608E+01 4.893E+01 2.043E+01 8.330E+00 1.775E+02 Te-129 1.014E+01 5.967E+02 9.461E+02 4.295E+02 1.009E+02 2.853E+01 2.112E+03 Te-129m 1.591E+00 1.077E+02 2.142E+02 1.584E+02 6.574E+01 2.662E+01 5.742E+02
Sheet 2 of 2 Revision 14, October 2017 QUAD CITIES - UFSAR Table 15.6-5b (Continued)
POST-LOCA REACTOR BUILDING ISOTOPIC INVENTORY
Post-LOCA Reactor Building Isotopic Inventory (Ci) Total Isotope Containment + ESF Leakage Activity 0.667 hr 2.0 hr 4.0 hrs 8.0 hrs 16 hrs 24 hrs (Ci) Te-131m 4.790E+00 3.143E+02 5.971E+02 4.030E+02 1.398E+02 4.736E+01 1.506E+03 Te-132 4.714E+01 3.152E+03 6.163E+03 4.403E+03 1.712E+03 6.499E+02 1.613E+04 I-131 1.470E+03 1.503E+04 2.688E+04 2.056E+04 1.031E+04 6.023E+03 8.028E+04 I-132 1.880E+03 1.625E+04 1.980E+04 7.772E+03 2.102E+03 7.826E+02 4.859E+04 I-133 2.953E+03 2.902E+04 4.888E+04 3.319E+04 1.311E+04 6.040E+03 1.332E+05 I-134 1.964E+03 7.030E+03 2.604E+03 8.548E+01 7.894E-02 8.499E-05 1.168E+04 I-135 2.630E+03 2.349E+04 3.430E+04 1.749E+04 3.899E+03 1.013E+03 8.282E+04 Xe-133 4.225E+03 1.093E+05 3.832E+05 7.597E+05 1.105E+06 1.197E+06 3.558E+06 Xe-135 1.815E+03 4.639E+04 1.481E+05 2.247E+05 1.853E+05 1.135E+05 7.199E+05 Cs-134 4.257E+02 3.862E+03 6.774E+03 4.951E+03 2.061E+03 8.399E+02 1.891E+04 Cs-136 1.039E+02 9.400E+02 1.641E+03 1.189E+03 4.867E+02 1.949E+02 4.556E+03 Cs-137 2.703E+02 2.453E+03 4.301E+03 3.144E+03 1.310E+03 5.338E+02 1.201E+04 Ba-139 1.754E+01 6.070E+02 4.418E+02 4.375E+01 3.267E-01 2.384E-03 1.110E+03 Ba-140 2.367E+01 1.597E+03 3.163E+03 2.321E+03 9.509E+02 3.807E+02 8.436E+03 La-140 3.014E-01 4.353E+01 1.824E+02 2.807E+02 2.259E+02 1.301E+02 8.629E+02 La-141 1.992E-01 1.065E+01 1.490E+01 5.446E+00 5.542E-01 5.510E-02 3.180E+01 La-142 1.599E-01 5.942E+00 4.811E+00 5.896E-01 6.744E-03 7.353E-05 1.151E+01 Ce-141 5.600E-01 3.787E+01 7.526E+01 5.554E+01 2.302E+01 9.318E+00 2.016E+02 Ce-143 5.160E-01 3.395E+01 6.478E+01 4.409E+01 1.555E+01 5.359E+00 1.642E+02 Ce-144 4.660E-01 3.153E+01 6.273E+01 4.642E+01 1.935E+01 7.882E+00 1.684E+02 Pr-143 2.028E-01 1.376E+01 2.754E+01 2.060E+01 8.737E+00 3.600E+00 7.444E+01 Nd-147 8.931E-02 6.022E+00 1.192E+01 8.732E+00 3.568E+00 1.424E+00 3.175E+01 Np-239 6.609E+00 4.399E+02 8.542E+02 6.021E+02 2.278E+02 8.416E+01 2.215E+03 Pu-238 2.595E-03 1.756E-01 3.495E-01 2.587E-01 1.080E-01 4.401E-02 9.383E-01 Pu-239 1.538E-04 1.041E-02 2.071E-02 1.534E-02 6.408E-03 2.614E-03 5.564E-02 Pu-240 1.550E-04 1.049E-02 2.087E-02 1.545E-02 6.446E-03 2.628E-03 5.603E-02 Pu-241 9.238E-02 6.251E+00 1.244E+01 9.208E+00 3.842E+00 1.566E+00 3.340E+01 Am-241 6.540E-05 4.426E-03 8.813E-03 6.531E-03 2.731E-03 1.115E-03 2.368E-02 Cm-242 1.285E-02 8.693E-01 1.729E+00 1.279E+00 5.330E-01 2.170E-01 4.641E+00 Cm-244 1.652E-03 1.118E-01 2.224E-01 1.646E-01 6.869E-02 2.800E-02 5.971E-01 (Sheet 1 of 2)
Revision 8, October 2005 QUAD CITIES - UFSAR
Table 15.6-6*
OFFSITE RADIOLOGICAL EFFECTS OF THE LOSS-OF-COOLANT ACCIDENT (INITIAL CORE) (This table is retained for historical purposes)
First 2 - Hour Dose Total Dose Distance (Miles) VS-1 MS-1 N-1 N-5 U-1 U-5 VS-1 MS-1 N-1 N-5 U-1 U-5 Passing Cloud Whole Body Dose (rem) 1/4 1.2 x 10
-5 1.2 x 10-5 1.2 x 10-5 2.0 x 10-6 1.7 x 10-5 2.5 x 10-6 3.6 x 10
-4 3.6 x 10-4 3.7 x 10-4 6.3 x 10-5 5.3 x 10-4 7.9 x 10-5 1 6.8 x 10
-6 6.8 x 10-6 8.2 x 10-6 1.2 x 10-6 6.2 x 10-6 9.8 x 10-7 2.1 x 10
-4 2.1 x 10-4 2.6 x 10-4 3.8 x 10-5 1.9 x 10-4 3.1 x 10-5 3 3.0 x 10
-6 3.2 x 10-6 2.9 x 10-6 4.8 x 10-7 1.2 x 10-6 2.4 x 10-7 9.5 x 10
-5 1.0 x 10-4 9.1 x 10-5 1.5 x 10-5 3.7 x 10-5 7.5 x 10-6 5 1.9 x 10
-6 2.1 x 10-6 1.4 x 10-6 2.5 x 10-7 4.7 x 10-7 1.1 x 10-7 6.0 x 10
-5 6.6 x 10-5 4.3 x 10-5 7.9 x 10-6 1.5 x 10-5 3.5 x 10-6 10 9.5 x 10
-7 1.0 x 10-6 3.9 x 10-7 9.4 x 10-8 1.2 x 10-7 3.9 x 10-8 3.0 x 10
-5 3.2 x 10-5 1.2 x 10-5 3.0 x 10-6 3.0 x 10-6 1.2 x 10-6 Lifetime Thyroid Dose (rem) 1/4 a a 8.0 x 10
-10a 3.2 x 10
-6 2.0 x 10-7 a a 3.3 x 10-8a 1.3 x 10
-4 8.2 x 10-6 1 a 9.6 x 10
-9 1.8 x 10-6 1.6 x 10-7 1.6 x 10-6 2.2 x 10-7 a 4.0 x 10
-7 7.4 x 10-5 6.6 x 10-6 6.8 x 10-5 9.2 x 10-6 3 a 2.3 x 10
-7 6.4 x 10-7 1.1 x 10-7 2.8 x 10-7 4.8 x 10-8 a 9.4 x 10
-6 2.6 x 10-5 4.5 x 10-6 1.2 x 10-5 2.0 x 10-6 5 a 3.5 x 10
-7 3.1 x 10-7 5.8 x 10-8 1.3 x 10-7 2.3 x 10-8 a 1.4 x 10
-5 1.3 x 10-5 2.4 x 10-6 5.3 x 10-6 9.4 x 10-7 10 a 3.5 x 10
-7 1.1 x 10-7 2.3 x 10-8 4.4 x 10-8 8.2 x 10-9 3.9 x 10
-9 1.4 x 10-5 4.6 x 10-6 9.4 x 10-7 1.8 x 10-6 3.4 x 10-7 First 2 - Hour Dose Total Dose (Sheet 2 of 2)
Revision 8, October 2005 QUAD CITIES - UFSAR
Table 15.6-6* (Cont'd)
Distance(Miles) VS-1 MS-1 N-1 N-5 U-1 U-5 VS-1 MS-1 N-1 N-5 U-1 U-5 Whole Body Fallout Dose (rem) 1/4 a a a a 5.5 x 10
-9 1.7 x 10-9 a a a a 5.3 x 10
-7 1.7 x 10-7 1 a a 1.4 x 10
-9 6.1 x 10-10 2.9 x 10-9 1.9 x 10-9 a 4.0 x 10
-10 1.3 x 10-7 5.9 x 10-8 2.7 x 10-7 1.9 x 10-7 3 a a 4.9 x 10
-10 4.2 x 10-10 5.0 x 10-10 4.2 x 10-10 a 9.5 x 10
-9 4.7 x 10-8 4.0 x 10-8 4.8 x 10-8 3.1 x 10-8 5 a 1.5 x 10
-10 2.4 x 10-10 2.3 x 10-10 2.2 x 10-10 2.0 x 10-10 a 1.5 x 10
-8 2.3 x 10-8 2.2 x 10-8 2.1 x 10-8 1.9 x 10-8 10 a 1.5 x 10
-10 a a a a a 1.5 x 10
-8 8.3 x 10-9 8.3 x 10-9 7.3 x 10-9 6.8 x 10-9 (a) denotes less than 10 Wind Speed Meterology (m/sec)
- 1. The values in this table were determined VS-1 Very Stable 1 to be unchanged due to the introduction of MS-1 Moderately Stable 1 ATRIUM-9B fuel since the source terms are N-1 Neutral 5 based on the TID 14844. N-5 Neutral 1 U-1 Unstable 1 U-5 Unstable 5
- 2. This table is valid prior to Extended Power Uprate. It is retained for historical purpose.
(Sheet 1 of 4) Revision 14, October 2017 QUAD CITIES - UFSAR Table 15.6-7 LOSS-OF-COOLANT ACCIDENT INPUT PARAMETERS (Alternative Source Term)
Design Basis Assumption I. Data and Assumptions Used to Estimate Radioactive Source from Postulated Accidents.
A. Power level, MWt 3016* B. Fuel Burnup Optima2 & ATRIUM 10XM: Peak Fuel Rod Burnup, MWD/MTU Optima2 & ATRIUM 10XM: Core Average Fuel Burnup, MWD/MTU 62,000 39,000 C. Fission Product Release Fractions
Gap Early Release In-Vessel Group Phase Phase Total Noble Gases 0.05 0.95 1.0 Halogens 0.05 0.25 0.3 Alkali Metals 0.05 0.20 0.25 Tellurium Metals 0.00 0.05 0.05 Ba, Sr 0.00 0.02 0.02 Noble Metals 0.00 0.0025 0.0025 Cerium Group 0.00 0.0005 0.0005 Lanthanides 0.00 0.0002 0.0002 (NRC Regulatory Guide 1.183, Table 1)
D. Iodine chemical forms Organic Elemental Aerosol (CsI) 0.15%
4.85% 95% E. Fission product release timing
Phase Onset Duration Gap Release 2 min 0.5 hr Early In-vessel Release 0.5 hr 1.5 hr (NRC Regulatory Guide 1.183, Table 4)
(Sheet 2 of 4) Revision 14, October 2017 QUAD CITIES - UFSAR Table 15.6-7 (continued)
LOSS-OF-COOLANT ACCIDENT INPUT PARAMETERS (Alternative Source Term)
Design Basis Assumption II. Data and Assumptions Used to Estimate Activity Released A. Primary containment leak rate (total), %/day 3.00 B. Total MSIV leakage (at 48 psig) 150 scfh (0 to 30 days);
60 scfh max per line C. Volume of primary containment (drywell plus suppression chamber free air volume), cu. ft.
2.69 x 10 5 D. Drywell surface area, sq. ft 32,430 E. Volume of suppression pool water, cu. ft 110,000 F. Primary containment leak rate which goes to secondary, %/day 2.187 G. Primary containment leak rate which goes through MSIVs, %/day 0.813 H. SGTS adsorption and filtration efficiencies, %
Organic Iodide Elemental Iodine Particulate Aerosols 80 80 98 I. Secondary containment leak rate, %vol/day 0
(Sheet 3 of 4) Revision 9, October 2007 QUAD CITIES - UFSAR Table 15.6-7 (continued)
LOSS-OF-COOLANT ACCIDENT INPUT PARAMETERS (Alternative Source Term)
Design Basis Assumption J. Main Steam Line Deposition Two-node treatment, each well mixed, is used for each steam line in which flow occurs. The first node is from the reactor vessel to the inboard MSIV. The second node is from the inboard MSIV to the outboard MSIV.
Gravitational settling applied to aerosols on horizontal pipe projected ar eas. For Elemental Iodine Deposition, a DF of 2 or elemental removal efficiency of 50% is used per AEB 98-03, Appendix B. No credit is taken for holdup or plate-out in the main steam lines beyond the outboard MSIVs. No credit is taken for holdup and plate-out in the main condenser.
III. Data for Control Room A. Volume of control room habitable zone, cu. ft.
1.84 x 10 5 B. Volume of control room proper, cu. ft.
5.83 x 10 4 C. Control room intake flow, scfm 2000 +/- 10% D. Control room intake charcoal adsorption efficiencies, % Organic Iodide Elemental Iodine Particulate Aerosols 99 99 99 E. Control room intake, isolation (following LOCA), min 0 F. Control room filter unit start (following LOCA), min 40 G. Unfiltered inleakage (0 to 40 min), scfm 60,000 H. Unfiltered inleakage (40 min to 720 hrs), scfm 400 (Sheet 4 of 4)
Revision 9, October 2007
QUAD CITIES - UFSAR Table 15.6-7 (continued)
LOSS-OF-COOLANT ACCIDENT INPUT PARAMETERS (Alternative Source Term)
Design Basis Assumption I. Control room cleanup recirculation flowrate, scfm 0 J. Occupancy factors 0 to 1 day 1 to 4 days 4 to 30 days
1.0 0.6 0.4 K. CR Operator Breathing Rates, m 3/sec 3.5E-04
- The reactor power level after Extended Po wer Uprate is 2957 MWt. The radiological consequences are evaluated at 3016 MWt to include the 2% instrument error.
QUAD CITIES - UFSAR (Sheet 1 of 1)
Revision 9, October 2007 Table 15.6-8*
LOSS-OF-COOLANT ACCIDENT CONTROL ROOM RADIOLOGICAL EFFECTS (rem) [Pre-uprate]
(Historical Information)
Thyroid Wholebody Beta MSIV Leakage Activity inside control room 7.58 1.36x 10
-2 0.55 Plume shine __ 2.03 x 10 - Direct shine -- 5.70 x 10 - Stack Release Activity inside control room 15.17 2.25 x 10
-1 8.16 Plume shine --
1.66 x 10-2 --
Total Control Room Doses 22.8 0.314 8.71 SRP 6.4 Guidelines 30 5 30
- The values in this table were determined to be unchanged due to the introduction of ATRIUM-9B fuel since the source terms
are based on the TID 14844.
- This table is retained for historical purposes.
QUAD CITIES - UFSAR (Sheet 1 of 1)
Revision 14, October 2017 Table 15.6-8a Loss-of-Coolant Accident EAB, LPZ and Control Room Doses (Alternative Source Term) (**)
Control Room 30 days 4.07* 5.0 EAB 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (max) 8.85 25 LPZ 30 days 2.45 25 Notes:
- The doses here include the external cloud shine, control room filter shine, and inhalation
doses from radioactivity drawn into the control room.
- The radiological consequences are based on the Westinghouse Optima2 core inventory described in Section 15.6.5.5.1. These co nsequences are bounding for the AREVA ATRIUM 10XM fuel design.
QUAD CITIES - UFSAR (Sheet 1 of 1)
Revision 9, October 2007 Table 15.6-9 Atmospheric Dispersion Factors for LOCA (Alternative Source Term)
Receptor Time Interval Ground Level
(/Q) (s/m 3) Elevated (/Q) (s/m 3) EAB 0 - 0.5 hr
1.57E-4 0.5 - 2 hr
6.38E-6 0 - 2 hr 1.36E-3 ---- LPZ 0 - 0.5 hr
3.01E-5 0.5 - 2 hr
2.05E-5 0 - 2 hr 1.04E-4 ---- 2 - 8 hr 4.14E-5 8.76E-6 8 - 24 hr 2.62E-5 5.73E-6 24 - 96 hr 9.96E-6 2.28E-6 96 - 720 hr 2.52E-6 6.07E-7 Control Room 0 - 2 hr 1.02E-3 5.84E-6 2 - 8 hr 8.23E-4 2.68E-6 8 - 24 hr 3.55E-4 1.81E-6 24 - 96 hr 2.32E-4 7.77E-7 96 - 720 hr 1.38E-4 2.30E-7
(Sheet 1 of 1)
Revision 7, January 2003 QUAD CITIES - UFSAR
Table 15.7-1
REACTOR BUILDING AIRBORNE FISSION PRODUCT INVENTORY (Original analysis, retained for historical purposes)
Time After Accident Noble Gases (curies/s) Halogens (curies/s) 1 minute 5.6 x 10 3 5.1 x 10 2 30 minutes 5.4 x 10 3 4.7 x 10 2 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 5.2 x 10 3 4.5 x 10 2 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.8 x 10 3 4.0 x 10 2 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2.7 x 10 3 2.3 x 10 2 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> 2.3 x 10 3 2.1 x 10 2 1 day 1.5 x 10 3 1.8 x 10 2 2 days 4.5 x 10 2 1.2 x 10 2 10 days 5.3 x 10
-2 1.6 x 10 1 25 days 2.6 x 10
-9 5.0 x 10-1 (Sheet 1 of 1)
Revision 7, January 2003 QUAD CITIES - UFSAR
Table 15.7-2
REFUELING ACCIDENT FISSION PRODUCT RELEASE RATE FROM CHIMNEY (Original analysis, retained for historical purposes)
Time After Accident Noble Gases (curies) Halogens (curies) 1 minute 6.5 x 10
-2 5.9 x 10-5 30 minutes 6.2 x 10
-2 5.5 x 10-5 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 6.0 x 10
-2 5.2 x 10-5 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 5.6 x 10
-2 4.6 x 10-5 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 3.2 x 10
-2 2.7 x 10-5 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> 2.7 x 10
-2 2.5 x 10-5 1 day 1.7 x 10
-2 2.0 x 10-5 2 days 5.3 x 10
-3 1.4 x 10-5 10 days 6.2 x 10
-7 1.8 x 10-6 25 days 0 5.8 x 10
-8 (Sheet 1 of 2)
Revision 7, January 2003 QUAD CITIES - UFSAR
Table 15.7-3 RADIOLOGICAL EFFECTS OF THE REFUELING ACCIDENT (Original analysis, retained for historical purposes)
First 2-Hour Dose Total Dose VS-2 MS-2 N-2 N-10 U-2 U-10 VS-2 MS-2 N-2 N-10 U-2 U-10 Distance (Miles)
Passing Cloud Whole Body Dose (rem) 1/4 8.3 x 10
-4 8.2 x 10-4 8.3 x 10-4 1.4 x 10-4 1.2 x 10-3 1.8 x 10-4 4.0 x 10
-3 4.0 x 10-3 4.1 x 10-3 7.0 x 10-4 5.9 x 10-3 8.8 x 10-4 1 4.8 x 10
-4 4.8 x 10-4 5.9 x 10-4 8.7 x 10-5 4.4 x 10-4 7.0 x 10-5 2.4 x 10
-4 2.4 x 10-3 2.9 x 10-3 4.3 x 10-4 2.2 x 10-3 3.4 x 10-4 3 2.2 x 10
-4 2.3 x 10-4 2.1 x 10-4 3.4 x 10-5 8.4 x 10-5 1.7 x 10-5 1.1 x 10
-3 1.1 x 10-3 1.1 x 10-3 1.7 x 10-4 4.1 x 10-4 8.3 x 10-5 5 1.4 x 10
-4 1.5 x 10-4 9.7 x 10-5 1.8 x 10-5 3.4 x 10-5 8.0 x 10-6 6.7 x 10
-4 7.3 x 10-4 4.7 x 10-4 8.8 x 10-5 1.6 x 10-4 3.9 x 10-5 10 6.8 x 10
-5 7.2 x 10-5 2.8 x 10-5 6.8 x 10-6 8.6 x 10-6 2.8 x 10-6 3.3 x 10
-4 3.5 x 10-4 1.4 x 10-4 3.3 x 10-5 4.2 x 10-5 1.4 x 10-5 Lifetime Thyroid Dose (rem) 1/4 *
- 1.5 x 10
-7* 6.1 x 10
-4 3.8 x 10-5
- 2.8 x 10
-10 1.0 x 10-6 2.5 x 10-10 4.1 x 10-3 2.6 x 10-4 1
- 1.9 x 10
-6 3.4 x 10-4 3.1 x 10-5 3.2 x 10-4 4.3 x 10-5
- 1.3 x 10
-5 2.3 x 10-3 2.1 x 10-4 2.1 x 10-3 2.9 x 10-4 3
- 4.4 x 10
-5 1.2 x 10-4 2.1 x 10-5 5.5 x 10-5 9.3 x 10-6
- 2.9 x 10
-4 8.2 x 10-4 1.4 x 10-4 3.7 x 10-4 5.2 x 10-5 5
- 6.7 x 10
-5 5.9 x 10-5 1.1 x 10-5 2.4 x 10-5 4.4 x 10-6
- 4.5 x 10
-4 4.0 x 10-4 7.5 x 10-5 1.6 x 10-4 3.0 x 10-5 10 1.8 x 10
-8 6.7 x 10-5 2.2 x 10-5 4.4 x 10-6 8.5 x 10-6 1.6 x 10-6 1.2 x 10
-7 4.5 x 10-4 1.5 x 10-4 3.3 x 10-6 5.7 x 10-5 1.1 x 10-5 QUAD CITIES - UFSAR
Table 15.7-3 RADIOLOGICAL EFFECTS OF THE REFUELING ACCIDENT (Original analysis, retained for historical purposes)
(Sheet 2 of 2)
Revision 7, January 2003 First 2-Hour Dose Total Dose
VS-2 MS-2 N-2 N-10 U-2 U-10 VS-2 MS-2 N-2 N-10 U-2 U-10 Whole Body Fallout Dose (rem) 1/4 * * *
- 6.0 x 10
-7 1.9 x 10-7 *
- 1.2 x 10
-9* 1.0 x 10
-5 3.5 x 10-6 1
- 4.6 x 10
-10 1.5 x 10-7 6.6 x 10-8 3.1 x 10-7 2.1 x 10-7
- 8.5 x 10
-9 2.8 x 10-6 1.2 x 10-6 5.8 x 10-6 3.9 x 10-6 3
- 1.1 x 10
-8 5.3 x 10-8 4.5 x 10-8 5.3 x 10-8 4.5 x 10-8
- 2.0 x 10
-7 9.9 x 10-7 8.5 x 10-7 1.0 x 10-7 8.5 x 10-7 5
- 1.7 x 10
-8 2.6 x 10-8 2.4 x 10-8 2.4 x 10-8 2.1 x 10-8
- 3.1 x 10
-7 4.8 x 10-7 4.5 x 10-7 4.5 x 10-7 4.0 x 10-7 10
- 1.7 x 10
-8 9.3 x 10-9 9.4 x 10-9 8.2 x 10-9 7.7 x 10-9
- 3.1 x 10
-7 1.7 x 10-7 1.8 x 10-7 1.5 x 10-7 1.5 x 10-7 Meteorology Wind Speed (mph) VS-2 Very stable 2 MS-2 Moderately stable 2 N-2 Neutral 2 N-10 Neutral 10 U-2 Unstable 2 U-10 Unstable 10
- Denotes less than 10
-10.
(Sheet 1 of 1)
Revision 7, January 2003 QUAD CITIES - UFSAR Table 15.7-4 RADIOLOGICAL EFFECTS OF THE REFUELING ACCIDENT IN THE SPENT FUEL POOL OR CONTAINMENT Pre-Uprate Conditions
Summary of Calculated Doses in the Control Room, Low Po pulation Zone, and Exclusion Area Boundary following a postulated Fuel Handling Accident with the Spent Fuel Pool Water at a minimum level of 19 feet.
Location Thyroid Dose (Rem) Whole Body Dose (Rem)
Beta Dose (Rem) Control Room Calculated Value 7.66 1.20E-2 0.462 SRP 6.4 Limit 30 5 30 Low Population Zone Calculated Value 0.687 3.80E-2 9.60E-2 SRP 15.7.4 Limit 75 6 NA Exclusion Area Boundary Calculated Value 9.92 0.358 0.900 SRP 15.7.4 Limit 75 6 NA (This table is retained for historical purposes)
(Sheet 1 of 1)
Revision 9, October 2007 QUAD CITIES - UFSAR Table 15.7-4a Fuel Handling Accident in the Spent Fuel Pool or in Containment EAB, LPZ, and Control Room Doses Following EPU (Historical Information)
Location Organ Non-Optima2 Dose (Rem)
Optima2 Dose (Rem)
- Regulatory Dose Limit (Rem) EAB Thyroid 12.6 12.6 750 Whole Body 0.42 0.44 6.25 LPZ Thyroid 0.87 0.87 75 Whole Body 0.045 0.047 6.25 Control Room Thyroid 9.73 9.68 30 Whole Body 0.014 0.014 5 Beta 0.55 0.55 30
- Optima2 dose per QDC-0000-N-1020, Revision 001A (Reference 17 of Section 15.7).
(Sheet 1 of 1)
Revision 14, October 2017 QUAD CITIES - UFSAR
Table 15.7-4b Fuel Handling Accident in the Spent Fuel Pool or in Containment EAB, LPZ, and Control Room Doses (Alternative Source Term)
Location Case 1 (23 feet water coverage)
Case 2 (19 feet water coverage)
Limits CR 5.0; EAB&LPZ 6.3 EAB 3.59 4.90 LPZ 0.275 0.374 CR 1.14 1.68
Notes: 1 - The radiological consequences are based on the Westinghouse Optima2 core inventory described in Section 15.7.2.5.1. Thes e consequences are bounding for the AREVA ATRIUM 10XM design.
(Sheet 1 of 1)
Revision 9, October 2007 QUAD CITES - UFSAR Table 15.7-5 RADIOLOGICAL EFFECTS OF AN OFF-GAS SYSTEM COMPONENT FAILURE (This assessment has not been update d using Alternative Source Term)
Component Failed Primary Activity Released Percent Released Resultant Exposure for 100,000 mCi/s (mR) [Ref. 14]
Resultant Exposure for 350,000 mCi/s (mR) [Ref.15]
SRP 1 Limits (mR) 10 CFR 100 Limits (mR)
First Carbon Bed Iodine 1 17.4 60.9 500 300,000 12 Carbon Beds Noble Gas 10 1.77 6.20 500 25,000 Prefilter Particulates 1 48.0 168.0 500 25,000 Holdup Pipe Particulates 20 53.0 185.5 500 25,000
- 1. Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, Postulated Radioactive Releases Due to Waste Gas System Leak or Failure.
QUAD CITIES - UFSAR 15.8-1 Revision 14, October 2017 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM The AREVA methodology is only applicable to Quad Cities Unit 1.
This section covers the events, which result in an anticipated transient without scram (ATWS). Anticipated transient without scram ev ents are beyond design basis accidents.
Anticipated transients without scram are those low probability events in which an anticipated transient occurs and is not follo wed by an automatic reactor shutdown (scram) when required. The failure of the reactor to scram quickly during these transients can lead to unacceptable reactor coolant system pressure s and to fuel damage. Mitigation of the lack of scram must involve insertion of nega tive reactivity into the reactor, thereby terminating the long-term aspects of the ev ent. For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), AREVA demonstrated the PRFO remains the limiting ATWS event (Reference 9) and will confirm this on a cycle specific basis.
The occurrence of a common-mode failure, whic h completely disables the reactor scram function, is a very low probability event. Theref ore, no significant risk to public safety is presented by the combination of an infrequent event and a common-mode failure, which
prevents scram. Thus, attention is focuse d on those transient situations, which have a relatively high expected freque ncy of occurrence at a power condition at which serious plant disturbance might result.
In support of the transition to SVEA-96 Optim a2 fuel, Westinghouse demonstrated that a Pressure Regulator Failure Open to maximum steam demand (PRFO) is the limiting ATWS event. For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), AREVA demonstrated the PRFO remains the limiting AT WS event (Reference 9) and will confirm this on a cycle specific basis.
Westinghouse performed a plant-unique ATWS analysis as part of the fuel licensing
analysis supporting the introduction of SVEA-96 Optima2 fuel. The analysis was
performed for two equilibrium cores (SVEA-96 Optima2 and GE14) using the same cycle design specifications at EPU conditions and fo r a transition core with 2/3 GE14 fuel and 1/3 SVEA-96 Optima2 fuel. The results of the analysis are reported in Reference 6.
Additional calculations for the long-term ATWS were also performed and reported in Reference 6. Long-term effects refer to the containment response and the general plant response with emphasis on the heat transfer to the suppression pool. The time delay from the symptom to initiating operator action credited for the SVEA-96 Optima2 fuel transition are equal to or conservatively longer than th e EPU licensing analysis of long-term ATWS.
The results confirm that for the limiting pr essure regulator failure open to maximum demand (PRFO) all acceptance criteria are me t after the introduction of SVEA-96 Optima2 fuel. Westinghouse has performed the limiting ATWS analysis for PRFO using the bounding ATWS inputs for Quad Cities. Westinghouse has analyzed the limiting ATWS event for the
plant changes including acoustic side branc h piping modification and change to 45 a/o enriched B-10 and 40 gpm SLC pump flow rate while considering applicable core design configurations. Westinghouse confirms the peak vessel bottom pressure for PRFO on a cycle-specific basis.Westinghouse ATWS analysis for Quad Cities has addressed all ATWS acceptance criteria. Following sp ecific criteria are satisfied:
QUAD CITIES - UFSAR 15.8-1a Revision 12, October 2013
- Peak vessel bottom pressure - The Westi nghouse calculation in Reference 6 considers the effect of core design changes and has established the licensing
basis for the reload of Westinghouse fuel.
- Peak cladding temperature (PCT) - West inghouse has confirmed in Reference 6 that the PCT for the limiting ATWS event is bounded by the LOCA analysis for
the Westinghouse fuel.
- Peak cladding oxidation - Westinghouse has confirmed in Reference 6 that the peak cladding oxidation for the limitin g ATWS event is bounded by the LOCA analysis for the Westinghouse fuel.
- Peak suppression pool temperature - The Westinghouse calculation in Reference 6 is cycle independent and confirms the peak suppression pool temperature is less than the acceptance criterion.
- Peak containment pressure - The Westinghouse calculation in Reference 6 is cycle independent and confirms the peak containment pressure is less than the design limit.
The PRFO would be the most severe postulat ed event from virtually all aspects when accompanied by a lack of scram. Other sign ificant ATWS events, which are postulated to occur are described in the following subsection
- s. Other transients such as closure of all main steam isolation valves (MSIVs), inadvertent opening of a relief or safety relief valve, and feedwater failure to maximum demand are less severe and are bounded by the PRFO
event described in Section 15.8.6. Section 15.
8.6 describes the limiting PRFO event in the Westinghouse analysis of Reference 6. Westinghouse has evaluated the impact of
installation of the Adjustable Speed Drives (ASDs) for the ATWS analysis (Reference 8), and has determined that there is no advers e impact due to this plant modification.
The following ATWS events have also been analyzed and have been demonstrated to be
bounded by the PRFO.
A. Closure of main steam isolation valves
B. Loss of A.C. Power
C. Loss of normal feedwater flow
D. Turbine generator trip
E. Loss of condenser vacuum QUAD CITIES - UFSAR 15.8-1b Revision 14, October 2017 The descriptions of the above ATWS events are provided in Sections 15.8.1 through 15.8.5 and are based upon analyses
[1,2], which utilized setpoints and intial conditions that differ from those currently in effect for Quad Cities.
The differences include a shorter rod insertion time than specified in the current design of the ATWS mitigation system (described in Section 7.8). The conclusions of these analyses
are not completely applicable to the current plant design or fuel cycle. Specific details contained in the descriptions and associated fi gures should be used only to understand the analysis and its conclusions. These specific details should not be used as sources of current
design information.
[15.8-3] For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), AREVA confirmed that the core PCT and cladding oxidatio n for the limiting ATWS are bound by the applicable LOCA analysis results. Also, the suppression pool temperature and containment pressure meet the acceptance cri teria and/or design limits as presented in Reference 10. For AREVA reload cores (starti ng with Quad Cities Unit 1 Cycle 25), AREVA confirms on a cycle-specific basis t hat the peak vessel pressure meets the 1500 psig ASME acceptance criterion.
15.8.1 Closure of Main Steam Line Isolation Valves
See the introduction to Section 15.8 for inform ation regarding use of details from this analysis description which may not be a pplicable to the current plant design.
15.8.1.1 Identification of Causes
Closure of all MSIVs is caused by any of a number of plant conditions such as low-low reactor water level, main steam line high fl ow, main steam low pressure, or main steam tunnel high temperature.
[15.8-4]
QUAD CITIES - UFSAR Revision 14, October 2017 15.8-2 15.8.1.2 Sequence of Events and Systems Operations
This transient would be initiated with the cl osure of the MSIVs. Closure of the MSIVs would produce an immediate increase in rea ctor pressure, which would result in a reduction in moderator voids and a rapid incre ase in reactor power. In the absence of normal scram, the fuel temperature would rise and the negative Doppler reactivity would limit the power. The opening of relief valves would tend to curtail increase in reactor pressure and power. The reactor pressure would reach the ATWS mitigation system setpoint of 1250 psig for Unit 2 and 1200 psig fo r Unit 1 (analytical limit) about five seconds after the start of the event. The ATWS mitigation system would initiate trip of the recirculation system pumps (RPT) and would in itiate alternate rod insertion (ARI). The RPT and ARI would introduce nega tive reactivity into the core. The reactor pressure would peak in about 12 seconds and then decrease to a value just above the relief valve setpoints.
[15.8-5]
15.8.1.3 Core and System Performance
A. Reactor Shutdown by RPT and ARI
The projected vessel pressure as a fun ction of time for this event is shown in Figure 15.8-1. In this case, the reactor pr essure would rise to the ATWS setpoint which would trip the recirculation pumps.
This would cause a rapid reduction of core flow and a corresponding increase in core moderator voids which would reduce core power. The re sulting neutron flux behavior is shown in Figure 15.8-
header which would result in insertion of the control rods. This transient would result in a peak reactor pressure of 1476 psig, which would satisfy the ATWS analysis guideline overpressure limit of 1500 psig
[2].
B. Reactor Shutdown by RPT and SBLC (No ARI)
In the event that control rod insertion (via ARI) is unavailable for shutdown of the reactor, the standby liquid control (SBLC) system would be used as an
alternative method of achieving reactor shutdown. The SBLC system would be actuated manually, and would inject an aqueous solution of sodium pentaborate
into the reactor vessel. When the boro n concentration in the reactor coolant reaches approximately 600 ppm, sufficient ne gative reactivity would be available to bring the core to hot shutdown. Th e SBLC pumps would continue injection until sufficient boron is in the core to achieve shutdown.
[15.8-6]
15.8.1.3.1 Reactor Water Level Response
The projected reactor water leve l response to the event with utilization of ARI is shown in Figure 15.8-3. The reactor water level woul d remain at near-normal level from event initiation until hot shutdown. Thereafter, ad equate water inventory would be maintained either by the feedwater system or by the hi gh-pressure coolant injection (HPCI) system.
[15.8-7]
QUAD CITIES - UFSAR 15.8-3 15.8.1.3.2 Containment and Suppression Pool Response
A. Reactor Shutdown by RPT and ARI
After the MSIVs are closed, the reacto r power would be dissipated by the relief valve discharge of steam into the suppression pool. The steam discharged into
the suppression pool would heat the suppre ssion pool. The reactor would be shut down by ARI and steam would continue to flow into the pool due to decay heat.
The operator would place the residual heat removal (RHR) system in the suppression pool cooling mode. The suppression pool temperature would
continue to increase until the decay heat input decreases below the heat removal capacity of the RHR heat exchangers.
B. Reactor Shutdown by RPT and SBLC (No ARI)
Achievement of reactor shutdown usin g SBLC rather than ARI would result in a peak containment pressure less than the design pressure. Containment pressures and suppression pool temperatur es will be higher using SBLC rather than ARI due to the greater amount of relief valve steam flow entering the
suppression pool.
15.8.1.3.3 Long Term Response
For ATWS considerations, the reactor conditio n of concern would be hot shutdown rather than cold shutdown, because the key factor wo uld be stopping thermal power generation during the event. The power generated prio r to reaching hot shutdown has the most significant potential impact on the plant. Co nsequently, the time required to achieve hot shutdown would be the important parameter fo r ATWS. After hot shutdown is achieved, further action would be required to bring the reactor to cold shutdown conditions.
[15.8-8]
15.8.1.3.4 Operator Actions
In case of an apparent ATWS, certain manual a ctions would be required to be performed by the operator if automatic features do not fun ction as designed. Possible operator actions would include trip of the recirculation pump, manual initiation of ARI, and actuation of
SBLC. [15.8-9]
Certain alarms and indications would be provided to the operator to support performance of the required manual actions within the time limits. Annunciator windows would alarm when the reactor water level or reactor pressure would reach the ATWS setpoints. At the beginning of the ATWS event, the recirculati on pumps would be signalled to trip and the ARI would be automatically initiated, and th e operator would be informed that an ATWS has occurred. The operator would then have su fficient time to perform the required actions.
Operator actions that would be required in the event of an ATWS are set forth in plant procedures. Plant procedures sp ecify that upon receipt of an automatic scram signal, if the reactor has not achieved shutdown using the control rods and rea ctor power is above a specified point, the operator is to actuate ARI. This action would insert the control rods.
Manual RPT would be achieved by tripping the recirculation pumps.
QUAD CITIES - UFSAR 15.8-4 Control room annunciators would inform the op erator of ATWS trouble, ATWS channel A or B manual push button armed, and ATWS channel A or B tripped conditions.
Manual initiation of ARI and RPT is descri bed in greater detail in Section 7.8.3.
In the event that control rod insertion is unavailable for shutting the reactor down, the SBLC system would be manually actuated to inject an aqueous solution of sodium
pentaborate into the reactor vessel.
[15.8-10]
Operator actions would involve actuation of th e residual heat removal system to cool the suppression pool. Operator actions are al so required to bring the reactor from hot shutdown to cold shutdown.
15.8.1.4 Barrier Response
During the MSIV closure transient without scra m the reactor fuel would experience a rapid power spike. Since heat removal through the fu el surface would follow the relatively slow dynamics of the fuel, a significant rise in fuel enthalpy would be encountered.
The analysis of the event shows the amount of cladding oxidation (<1% by volume) would be far less than the 17% guideline (per NEDE-25026
^[1]^), and peak fuel enthalpy would be less than the Regulatory Guide 1.77 limit of 280 cal/
- g. Few fuel rod perforations would be experienced.
15.8.1.5 Radiological Consequences
The radiological consequences would be minimal due to the small (if any) number of fuel
rod perforations.
15.8.2 Loss of Normal AC Power
See the introduction to Section 15.8 for inform ation regarding use of details from this analysis description which may not be a pplicable to the current plant design.
15.8.2.1 Identification of Causes
The loss of normal ac power would generally be caused by large grid disturbances which in turn would de-energize buses that supply power to auxiliary equipment such as the recirculation pumps, condensate pump s, and circulating water pumps.
[15.8-11]
QUAD CITIES - UFSAR 15.8-5 15.8.2.2 Sequence of Events and Systems Operations
When auxiliary power is lost, the circulati ng water pumps, feedwater pumps, and recirculation pumps would begin coasting down immediately. The reduction in core flow would begin to reduce the rea ctor power. A turbine-generato r trip would occur at the start of the event due to a general grid disturbance and would contribute to the pressurization of the reactor. The safety/relief valves wo uld open momentarily which would limit the pressure rise in the vessel. The peak vessel pressure experienced in this event would be less than in the MSIV closure event. The sho rt-term response would be much less severe than the MSIV ATWS event for the following reasons:
A. The recirculation pumps would trip at ti me zero which would result in lower core flow rather than tripping when the reactor pressure reaches the ATWS mitigation system setpoint.
B. The feedwater pumps would trip at time zero which would result in reduced core inlet subcooling, and hence in a lower reactor power.
15.8.2.3 Core and System Performance
A. Reactor Shutdown by RPT and ARI
The reactor would achieve hot shut down by utilizing ARI. The peak fuel enthalpy reached would be less than 280 cal/g.
Combined flow of the reactor core isolation cooling (RCIC) and HPCI system s would restore reactor water level to the normal range. Cold shutdown woul d be reached by performing the normal manual actions.
B. Reactor Shutdown by RPT and SBLC (No ARI)
In the event that insertion of the control rods via ARI is not achievable, the SBLC system would be utilized as an alte rnative method of effecting neutronic power shutdown. The vessel water level wo uld be restored by HPCI and RCIC.
Containment pressures and suppression p ool temperatures will be higher using SBLC rather than ARI due to the greater amount of steam flow entering the
suppression pool.
The operator actions associated with this event would be similar to those described in Section 15.8.1.3.4.
15.8.2.4 Barrier Performance
As in the MSIV closure event, cladding oxid ation would be less than 1% by volume. Peak fuel enthalpy would be less than 280 cal/g. Very few if any fuel rod perforations would be experienced.
QUAD CITIES - UFSAR 15.8-6 15.8.2.5 Radiological Consequences Radiological consequences would be minimal due to the small (if any) number of fuel rod perforations.
15.8.3 Loss of Normal Feedwater Flow
See the introduction to Section 15.8 for inform ation regarding use of details from this analysis description which may not be a pplicable to the current plant design.
15.8.3.1 Identification of Causes
Inadvertent trip of all feedwater pumps or water level controller failure (zero demand) would be a potential cause for loss of all norm al feedwater flow to the vessel. Loss of auxiliary power would also be a potential cause of this event.
[15.8-12]
15.8.3.2 Sequence of Events and Systems Operations
The short-term effects of this event would be less severe than those of the MSIV closure event.
Reactor core flow would be re duced when the feedwater flow reduction occurs, which would drop power gradually until a low water level scram is initiated. Gradual vessel water inventory reduction would occur until vessel isol ation is initiated. The RCIC/HPCI systems would be initiated automatically to mainta in proper water level until the event is terminated. When the reactor water level reac hes the low-low level, the ATWS logic would initiate ARI and RPT.
15.8.3.3 Core and System Performance
A. Reactor Shutdown by RPT and ARI
Hot shutdown would be achieved by using ARI. The recirculation pumps would also trip. The peak temperature reached in the pool would be slightly less than the temperature for the MSIV closure event. The peak vessel pressure
experienced in this event would be less than in the MSIV closure event.
B. Reactor Shutdown by RPT and SBLC (No ARI)
In the event that insertion of the control rods via ARI is not achievable, the SBLC system would be utilized as an alte rnative method of effecting neutronic power shutdown.
The cold shutdown condition would be achieved by normal manual actions similar to those performed during the MSIV closure event. Without the ARI QUAD CITIES - UFSAR 15.8-7 function, the plant long term response to the transient would be similar to the loss of normal ac power transient (SBLC without ARI).
The operator actions associated with this event would be similar to those described in Section 15.8.1.3.4.
15.8.3.4 Barrier Performance
This event would result in cladding oxidat ion of less than 1% by volume. Peak fuel enthalpy would be less than 280 cal/g. Very few if any fuel rod perforations would be experienced.
15.8.3.5 Radiological Consequences
Radiological consequences would be minimal due to the small (if any) number of fuel rod perforations.
15.8.4 Turbine-Generator Trip/Load Rejection
See the introduction to Section 15.8 for inform ation regarding use of details from this analysis description which may not be a pplicable to the current plant design.
15.8.4.1 Identification of Causes
Loss of generator electrical load would initiate fast closure of the turbine control valves to provide overspeed protection for the unit. A variety of equipment protection signals would lead to trip of the turbine stop valves directly.
Both the turbine control valve fast closure and turbine stop valve trip would have a simila r effect on the reactor. Normally, a scram would be initiated almost simultaneously with th e start of the control valve fast closure or with the stop valves starting to close. Howe ver, the scram is postulated not to occur.
[15.8-13]
15.8.4.2 Sequence of Events and Systems Operations
The fast closure of the valves would cause an abrupt reactor pressure rise which would be limited to well below design pressures by th e action of the bypass and the safety/relief valves.
When the dome pressure reaches the ATWS setpoint, a recirculation pump trip and alternate rod insertion would be initiated.
QUAD CITIES - UFSAR 15.8-8 15.8.4.3 Core and System Performance
A. Reactor Shutdown by RPT and ARI
The neutron flux, vessel pressure, and fuel transient peaks experienced in this event would be less than those in the MSIV closure event.
The long-term response of the plant is not analyzed, as it would be similar to the MSIV closure ATWS event. However, because of the availability of turbine bypass to the condenser, the steam flow into the suppression pool would be
considerably less than the MSIV closure ev ent. Long-term heat removal would be provided by the steam bypass to main condenser. Reactor coolant inventory would be maintained using the feedwater system.
B. Reactor Shutdown by RPT and SBLC (No ARI)
In the event that insertion of the control rods via ARI is not achievable, the SBLC system would be utilized as an alte rnative method of effecting neutronic power shutdown. The peak suppression pool temperature will be less than the temperature reached in the MSIV closure event (SBLC without ARI).
The operator actions associated with this event would be similar to those described in Section 15.8.1.3.4.
15.8.4.4 Barrier Performance
This event would result in cladding oxidatio n of less than 1% by volume. Peak fuel rod enthalpy would be less than 280 cal/g. Very few if any fuel rod perforations would be experienced.
15.8.4.5 Radiological Consequences
Radiological consequences are minimal due to the small (if any) number of fuel rod
perforations.
15.8.5 Loss of Condenser Vacuum
See the introduction to Section 15.8 for inform ation regarding use of details from this analysis description which may not be a pplicable to the current plant design.
QUAD CITIES - UFSAR 15.8-9 15.8.5.1 Identification of Causes
The reduction or loss of vacuum in the main condenser can be caused by loss of circulating
water pumps or ineffectual operation of the vacuum support equipment. The long-term results of the event would be similar to a co ndenser isolation unless enough vacuum can be maintained to preserve bypass flow. Preser ving the bypass flow to the condenser would permit decay heat removal through the conden ser instead of relying upon the suppression pool and the shutdown cooling systems.
[15.8-14]
15.8.5.2 Sequence of Events and Systems Operations
Loss of condenser vacuum would trip the turbine stop valves closed (which would normally scram the reactor). If the event is severe en ough the steam bypass valves would be closed.
These actions would occur normally over a pe riod of several minutes or at worst, 20-30 seconds. The initial sequence of events would be the same as a turbine-generator trip since all systems would function in the same way.
15.8.5.3 Core and System Performance
A. Reactor Shutdown by RPT and ARI
The loss of condenser vacuum event woul d result in short-term peak values that would be less severe than in the MSIV cl osure event. All ATWS logic would be rapidly activated by the high pressure tr ansient. The longer term nature of this event (assuming vacuum continues to deteri orate) would be converted to a nearly normal condenser isolation by action of the ARI which would achieve hot shutdown. The long-term response woul d be similar to the response for the MSIV closure event. The peak vessel pr essure experienced in this event would be less than in the MSIV closure event.
B. Reactor Shutdown by RPT and SBLC (No ARI)
In the event that insertion of the control rods via ARI is not achievable, the SBLC system would be utilized as an alte rnative method of effecting neutronic power shutdown.
The operator actions associated with this event would be similar to those described in Section 15.8.1.3.4
15.8.5.4 Barrier Performance
This event would result in cladding oxidatio n of less than 1% by volume. Peak fuel rod enthalpy would be less than 280 cal/g. Very few if any fuel rod perforations would be experienced.
QUAD CITIES - UFSAR 15.8-10 Revision 14, October 2017 15.8.5.5 Radiological Consequences The radiological consequences would be minima l due to the small (if any) number of fuel rod perforations.
15.8.6 Pressure Regulator Failure - Open to Maximum Demand
See the introduction to Section 15.8 for informatio n regarding the use of details from the analysis description for applicability to the current plant design.
15.8.6.1 Identification of Causes
Pressure regulator failure to the maximum demand is caused by the malfunction of the normal pressure regulator. The feedback pressure regula tor responds only to the low demand signal and does not intervene when the demand signal is high.
15.8.6.2 Sequence of Events and Systems Operations
This transient would be initiated by the failure of the pressure regulator, which generates a maximum demand signal for the turbine-generator. This signal yields the opening of all the
turbine bypass valves and forces the turbine control valves to the fully open position. This causes
the reactor vessel to depressuirze and to void the core, which in turn further reduces the reactor power and the vessel pressure. The MSIVs start to close once the low steamline pressure is detected at about 13 seconds. The reactor pre ssure would reach the ATWS mitigation system setpoint of 1200 psig for Unit 1 and 1250 psig for Unit 2 approximately 21 seconds after the start of the event (Reference 4). The ATWS mitigation system would initiate a recirculation pump trip (RPT) and would initiate alternate rod inserti on (ARI). The RPT and ARI would introduce negative reactivity into the core. The reactor pressure would peak in about 27 seconds (Reference 4). The standby liquid control (SBLC) system can be initiated if ARI is unavailable. The reactor water level is reduced and maintained at an el evation consistent with Emergency Operation procedures (EOP). The water level is restored to the normal level once the hot shutdown boron weight (HSBW) is injected into the vessel.
15.8.6.3 Core and System Performance
In the event that control rod insertion (via ARI) is unavailable for shutdown of the reactor, the SBLC system would be used as an alternative me thod of achieving shutdown. According to the Westinghouse analysis presented in Reference 6, the transient would result in a peak vessel pressure less than the overpressure limit of 1500 psig. Westinghouse checks the limiting ATWS pressure every reload. The peak suppression pool temperature is calculated to be less than the bounding post-accident suppression pool temperat ure limit of 202 degrees F. Therefore, for Westinghouse reload cores (Unit 2) the ATWS a cceptance criteria continue to be satisfied for SVEA-96 Optima2 fuel.
For AREVA reload cores (starting with Quad Cities Unit 1 Cycle 25), cycle-specific evaluations of ATWS peak vessel pressure are performed to conf irm the ATWS peak pressure criteria are met.
The evaluations use an ATWS mitigation syst em setpoint of 1200 psig (Reference 9).
15.8.6.4 Barrier Response This event would result in cladding oxidation of less than 1% by volume. Peak fuel rod enthalpy would be less than 280 cal/g. Very few (if any) fuel rod perforations would be experienced.
QUAD CITIES - UFSAR 15.8-11 Revision 14, October 2017 15.8.6.5 Radiological Consequences The radiological consequences would be minimal due to the small (if any) number of fuel
rod perforations.
15.8.7 References
- 1. "Studies of ATWS for Dresden 2, 3 and Quad Cities 1, 2 Nuclear Power Stations", General Electric Co., NEDE-25026, December 1976,
- 2. "Main Steam Isolation Valves Closure Event With ATWS/RPT and ARI for Dresden 2, 3 and Quad Cities 1, 2 Nuclear Generating Plants," General Electric Company, NSE 0880, August 1980.
- 3. Deleted.
- 4. "Dresden and Quad Cities Extended Power Uprate, Task T0902: Anticipated Transient Without Scram, " GE-NE-A22-00103-11-01, Revision 2, February 2002.
- 5. Deleted. 6. OPTIMA2-TR026QC-ATWS, Revision 1, "ATWS Analysis for the Introduction of SVEA-96 Optima2 Fuel at Quad Cities 1 & 2," April 2007.
- 7. GE-NE-0000-0050-6728-01, Revision 1, "Quad Cities Acoustical Side Branch Modification Evaluation of Current GE Tasks," March 2006.
- 8. NF-BEX-08-133, Revision 1, "Evaluation of the Planned Implementation of Adjustable Speed Drives in Quad Cities Units 1 and 2," February 2009.
- 9. ANP-3565P, Revision 0, "Quad Cities Unit 1 Cycle 25 Reload Safety Analysis," AREVA Inc., February 2017.
- 10. ANP-3338P Revision 1, "Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Exte nded Power Uprate," AREVA Inc., August 2015.