ML24109A096: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot insert)
 
(StriderTol Bot change)
 
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                                      USAR-14 SECTION 14          PLANT SAFETY ANALYSIS                                                    Revision 40 Page 1 of 239
{{#Wiki_filter:}}
 
TABLE OF CONTENTS Section                                                                                                          Page
 
14.1                    Summary Description .............................................................................. 8
 
14.1.1                  General Safety Design Basis ...................................................................... 9 14.1.2                  Operational Design Basis ........................................................................... 9 14.1.3                  Primary System Integrity Design Basis ....................................................... 9 14.1.4                  Plant Stability Design Basis ........................................................................ 9 14.1.5                  Design Basis for Accidents ....................................................................... 10
 
14.2                    Fuel Cladding Integrity Safety Limit ..................................................... 10
 
14.3                    Operating Limits ..................................................................................... 11
 
14.3.1                  MCPR Calculational Procedure ................................................................ 11 14.3.1.1                General Assumptions and Models ............................................................ 11 14.3.1.2                Calculation of Operating Limit MCPR for Core Reload ............................. 12 14.3.2                  Calculation of MAPLHGR for Core Reload ............................................... 14 14.3.3                  Calculation of LHGR for Core Reload....................................................... 14 14.3.4                  Power to Flow Operating Map .................................................................. 15
 
14.4                    Transient Events Analyzed for Core Reload ........................................ 16
 
14.4.1                  Generator Load Rejection Without Bypass ............................................... 19 14.4.1.1                Starting Conditions and Assumptions ....................................................... 19 14.4.1.2                Event Description ..................................................................................... 19 14.4.1.3                Acceptance Criteria .................................................................................. 19 14.4.1.4                Main Physics Parameters ......................................................................... 20 14.4.1.5                Event Results ........................................................................................... 20 14.4.2                  Loss of Feedwater Heating ....................................................................... 20 14.4.2.1                Starting Conditions and Assumptions ....................................................... 20 14.4.2.2                Event Description ..................................................................................... 20 14.4.2.3                Acceptance Criteria .................................................................................. 21 14.4.2.4                Main Physics Parameters ......................................................................... 21 14.4.2.5                Event Results ........................................................................................... 21
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                                      USAR-14 SECTION 14          PLANT SAFETY ANALYSIS                                                      Revision 40 Page 2 of 239
 
TABLE OF CONTENTS (CONT'D)
Section                                                                                                          Page 14.4.3                  Rod Withdrawal Error ............................................................................... 21 14.4.3.1                Starting Conditions and Assumptions ....................................................... 21 14.4.3.2                Event Description ..................................................................................... 22 14.4.3.3                Acceptance Criteria .................................................................................. 22 14.4.3.4                Main Physics Parameters ......................................................................... 23 14.4.3.5                Event Results ........................................................................................... 23 14.4.4                  Feedwater Controller Failure - Maximum Demand ................................... 23 14.4.4.1                Starting Conditions and Assumptions ....................................................... 23 14.4.4.2                Event Description ..................................................................................... 24 14.4.4.3                Acceptance Criteria .................................................................................. 24 14.4.4.4                Main Physics Parameters ......................................................................... 24 14.4.4.5                Event Results ........................................................................................... 24 14.4.5                  Turbine Trip Without Bypass .................................................................... 25 14.4.5.1                Starting Conditions and Assumptions ....................................................... 25 14.4.5.2                Event Description ..................................................................................... 25 14.4.5.3                Acceptance Criteria .................................................................................. 25 14.4.5.4                Main Physics Parameters ......................................................................... 25 14.4.5.5                Event Results ........................................................................................... 25 Table 14.4-1            Transients and Accidents for Reload Evaluation ...................................... 26
 
14.5                    Special Events ........................................................................................ 27
 
14.5.1                  Vessel Pressure ASME Code Compliance Model .................................... 27 14.5.2                  Standby Liquid Control System Shutdown Margin ................................... 28 14.5.3                  Stuck Rod Cold Shutdown Margin ............................................................ 28
 
14.6                    Plant Stability Analysis .......................................................................... 29
 
14.7                    Accident Evaluation Methodology ........................................................ 30
 
14.7.1                  Control Rod Drop Accident Evaluation ..................................................... 31 14.7.1.1                Sequence of Events ................................................................................. 32 14.7.1.2                Model Parameters Sensitivities ................................................................ 33 14.7.1.3                Analytical Methods ................................................................................... 33 14.7.1.4                Effect of Fuel Densification ....................................................................... 35 14.7.1.5                Results ..................................................................................................... 35
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                                      USAR-14 SECTION 14          PLANT SAFETY ANALYSIS                                                      Revision 40 Page 3 of 239
 
TABLE OF CONTENTS (CONT'D)
Section                                                                                                          Page 14.7.1.6                Radiological Consequences ..................................................................... 35 14.7.1.6.1              Introduction ............................................................................................... 37 14.7.1.6.2              Source Term ............................................................................................. 37 14.7.1.6.3              Mitigation .................................................................................................. 38 14.7.1.6.4              Transport .................................................................................................. 38 14.7.1.6.5              Results ..................................................................................................... 39 14.7.2                  Loss-of-Coolant Accident ......................................................................... 39 14.7.2.1                Description of Design Basis LOCA ........................................................... 40 14.7.2.2                ECCS-LOCA Analysis Assumptions ......................................................... 41 14.7.2.3                Effects of Unavailability of Offsite Power .................................................. 41 14.7.2.4                Loss of Coolant Accident Analysis ........................................................... 42 14.7.2.5                Effect of Fuel Densification ....................................................................... 43 14.7.2.6                Analytical Methods ................................................................................... 43 14.7.2.6.1              Blowdown Analysis ................................................................................... 43 14.7.2.6.2              Refill/Reflood Analysis .............................................................................. 44 14.7.2.6.3              Heatup Analysis ....................................................................................... 44 14.7.2.6.4              Break Spectrum Analysis ......................................................................... 45 14.7.2.6.4.1            Limiting Single Failure .............................................................................. 47 14.7.2.6.4.2            Recirculation Line Breaks ......................................................................... 47 14.7.2.6.4.3            Non-Recirculation Line Breaks ................................................................. 49 14.7.2.6.4.4            Long Term Cooling ................................................................................... 49 14.7.2.7                Summary of ECCS-LOCA Results ........................................................... 49 14.7.2.8                Emergency Core Cooling System Performance ....................................... 50 14.7.2.8.1              ECC System Descriptions ........................................................................ 50 14.7.2.8.1.1            Emergency Diesel Generators ................................................................. 50 14.7.2.8.1.2            High Pressure Coolant Injection (HPCI) System ...................................... 51 14.7.2.8.1.3            Core Spray (CS) System .......................................................................... 51 14.7.2.8.1.4            Low Pressure Coolant Injection (LPCI) System ........................................ 52 14.7.2.8.1.5            Automatic Depressurization System (ADS) .............................................. 53 14.7.2.8.2              Single Failure Considerations ................................................................... 54 14.7.2.8.3              ECCS Equipment Performance ................................................................ 55 14.7.2.8.4              Evaluation of Parameter Variations .......................................................... 58 14.7.2.8.5              Reduced Power Considerations ............................................................... 59
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                                        USAR-14 SECTION 14            PLANT SAFETY ANALYSIS                                                      Revision 40 Page 4 of 239
 
TABLE OF CONTENTS (CONT'D)
Section                                                                                                              Page 14.7.2.8.6                Long Term Core Cooling Performance ..................................................... 60 14.7.2.9                  Radiological Consequences ..................................................................... 62 14.7.2.9.1                Introduction ............................................................................................... 62 14.7.2.9.2                Source Term ............................................................................................. 63 14.7.2.9.3                Mitigation .................................................................................................. 64 14.7.2.9.4                Transport .................................................................................................. 65 14.7.2.9.5                Results ..................................................................................................... 67 14.7.3                    Main Steam Line Break Accident Analysis ............................................... 67 14.7.3.1                  Reactor Primary System Transient Effects and Mass and Energy Releases .................................................................................................. 68 14.7.3.2                  Radiological Consequences ..................................................................... 69 14.7.3.2.1                Introduction ............................................................................................... 69 14.7.3.2.2                Source Term ............................................................................................. 70 14.7.3.2.3                Mitigation .................................................................................................. 71 14.7.3.2.4                Transport .................................................................................................. 71 14.7.3.2.5                Results ..................................................................................................... 71 14.7.4                    Fuel Loading Error Accident ..................................................................... 72 14.7.5                    One Recirculation Pump Seizure Accident Analysis ................................ 72 14.7.6                    Refueling Accident Analysis ..................................................................... 73 14.7.6.1                  Identification of Causes ............................................................................ 73 14.7.6.2                  Effect of Fuel Densification ....................................................................... 74 14.7.6.3                  Radiological Consequences ..................................................................... 74 14.7.6.3.1                Introduction ............................................................................................... 74 14.7.6.3.2                Source Term ............................................................................................. 75 14.7.6.3.3                Mitigation .................................................................................................. 76 14.7.6.3.4                Transport .................................................................................................. 76 14.7.6.3.5                Results ..................................................................................................... 76 14.7.7                    Accident Atmospheric Dispersion Coefficients ......................................... 77 14.7.7.1                  Meteorological Data .................................................................................. 77 14.7.7.2                  Control Room Atmospheric Dispersion Coefficients ................................. 77 14.7.7.3                  Offsite Atmospheric Dispersion Coefficients ............................................. 78 14.7.8                    Core Source Term Inventory .................................................................... 78
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                            USAR-14 SECTION 14        PLANT SAFETY ANALYSIS                                            Revision 40 Page 5 of 239
 
TABLE OF CONTENTS (CONT'D)
Section                                                                                            Page Table 14.7-2a        CRDA Radiological Consequences Analysis Inputs and Assumptions ............................................................................................. 79 Table 14.7-2b        CRDA Dose Consequences (Rem TEDE) ................................................ 81 Table 14.7-3          Appendix K Assumption ........................................................................... 82 Table 14.7-4          Maximum Break Areas ............................................................................. 83 Table 14.7-5          Initial Conditions for LOCA Analyses ........................................................ 84 Table 14.7-6          Fuel Parameters utilized in ECCS-LOCA Analysis ................................... 85 Table 14.7-7          High Pressure Coolant Injection System Parameters ............................... 86 Table 14.7-8          Core Spray System Parameters ............................................................... 87 Table 14.7-9          Low Pressure Coolant Injection System Parameters ............................... 88 Table 14.7-10        Automatic Depressurization System Parameters            ..................................... 90 Table 14.7-11        Single Failures and Available Systems        ..................................................... 91 Table 14.7-12        ECCS Injection Timing Parameters Used in ECCS            -LOCA Analysis ......... 92 Table 14.7-13        LOCA Radiological Consequences Analysis Inputs and Assumptions ............................................................................................. 93 Table 14.7-14        LOCA Dose Consequences (Rem TEDE)  Reference 153 and 154                    ......... 96 Table 14.7-15        Mass and Energy Release for Main Steamline Break Outside Containment - 1880 MWt Power (See Note 1)            ......................................... 97 Table 14.7-16        Mass and Energy Release for Main Steamline Break Outside Containment - Hot Standby (965 psia) ..................................................... 99 Table 14.7-17        Mass and Energy Release for Main Steamline Break Outside Containment - Hot Standby (1158 psia) ................................................. 102 Table 14.7-18        Mass Release from MSLBA        - Hot Standby ............................................. 105 Table 14.7-19        MSLBA Radiological Consequences Analysis Inputs and Assumptions ........................................................................................... 106 Table 14.7-20        MSLBA Dose Consequences (Rem TEDE)                ............................................ 107 Table 14.7-21        FHA Radiological Consequences Analysis Inputs and Assumptions
                      ............................................................................................................... 108 Table 14.7-22        FHA Dose Consequences (Rem TEDE) Reference 175 ........................ 109 Table 14.7-23        Atmospheric Dispersion Factors (X/Q) for Accident Analysis 3 (sec/m  ) .................................................................................................. 110 Table 14.7-24a        Core Inventory for GE14 Fuel at 35 GWD/MT Exposure @ T = 0 Hours in Ci/MWt (Reference 167)........................................................... 112 Table 14.7-24b        Core Inventory for ATRIUM 10XM Fuel @ T = 0 Hours in Ci/MWt (Reference 172) ..................................................................................... 113
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                                      USAR-14 SECTION 14          PLANT SAFETY ANALYSIS                                                    Revision 40 Page 6 of 239
 
TABLE OF CONTENTS (CONT'D)
Section                                                                                                          Page Table 14.7-24c          Core Inventory for GE14 Fuel at 37 GWD/MT Exposure @ T = 0 Hours in Ci/MWt (Reference 173)........................................................... 114 Table 14.7-24d          Core Inventory for ATRIUM 10XM Fuel at 41 GWD/MT Exposure
                        @ T = 0 Hours in Ci/MWt ....................................................................... 115 Table 14.7-24e          Core Inventory for ATRIUM 11 Fuel at 41 GWD/MT Exposure  @ T
                        = 0 Hours in Ci/MWt ............................................................................... 116
 
14.8                    Anticipated Transients Without Scram (ATWS) ................................. 117
 
14.8.1                  General ................................................................................................... 117 14.8.1.1                Final ATWS Rule .................................................................................... 117 14.8.1.2                Programmatic Issues .............................................................................. 118 14.8.2                  Evaluation of Events ............................................................................... 119 14.8.2.1                ATWS Licensing Basis Analysis ............................................................. 120 14.8.2.2                ATWS With Depressurization Analysis ................................................... 121 14.8.2.3                ATSW With Core Instability (ATWSI) Analysis ....................................... 122
 
14.9                    Deleted .................................................................................................. 124
 
14.10                    Other Analyses ..................................................................................... 124
 
14.10.1                  Adequate Core Cooling for Transients With a Single Failure ................. 124
 
14.11                    References ............................................................................................ 125
 
FIGURES                  ............................................................................................................... 139
 
Figure 14.7-1            Deleted ................................................................................................... 140 Figure 14.7-1a          Deleted ................................................................................................... 141 Figure 14.7-1b          Deleted ................................................................................................... 142 Figure 14.7-2            Deleted ................................................................................................... 143 Figure 14.7-3            Deleted ................................................................................................... 144 Figure 14.7-4            Deleted ................................................................................................... 145 Figure 14.7-5            Deleted ................................................................................................... 146 Figure 14.7-6            Deleted ................................................................................................... 147 Figure 14.7-7            Deleted ................................................................................................... 148 Figure 14.7-8            CS Flow Delivery Assumed for Break Spectrum .................................... 149
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                              USAR-14 SECTION 14        PLANT SAFETY ANALYSIS                                                Revision 40 Page 7 of 239
 
TABLE OF CONTENTS (CONT'D)
Section                                                                                                Page Figure 14.7-9          LPCI Flow Delivery Assumed for Break Spectrum ................................. 150 Figure 14.7-10        ADS Actuation Logic ............................................................................... 151 Figure 14.7-11        CS Initiation Logic................................................................................... 152 Figure 14.7-12        LPCI Initiation Logic ................................................................................ 153 Figure 14.7-13        Main Steam Line Break Accident, Break Location            .................................. 154 Figure 14.7-14        Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model ............... 155
 
Section 14A            ............................................................................................................... 156
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 8 of 239
 
14.1 Summary Description
 
The purpose of this section is to evaluate the ability of the plant to operate without undue risk to the safety of the public.
 
The analytical objective of this evaluation is to demonstrate that plant systems essential to safety are capable of performing their functions during transients or postulated accidents, concurrent with postulated equipment failures.
 
These transients and the limiting accident parameters are generally re-verified for applicability every core reload. Where applicable, the initial conditions, analytical methods, and results presented herein are for the current reload cycle.
 
As required by NRC Generic Letter 88-20 (Reference 84), an Individual Plant Examination (IPE) Report for Monticello was prepared and submitted to the NRC in February of 1992 (Reference 72), with additional information provided in February of 1993 (Reference 73).
By {{letter dated|date=May 26, 1994|text=letter dated May 26, 1994}} (Reference  77), the NRC transmitted the NRC Staff Evaluation of the Monticello IPE. The IPE is a full scope probabilistic risk assessment consisting of Level 1 and Level 2 analyses. The two analyses were used to determine an estimate of the probability and type of releases which could potentially result from a severe accident. The IPE report provides valuable insights concerning the safety significance of various postulated accidents and failures.
 
Generic Letter 88-20, Supplement No. 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, dated June 28, 1991 (Reference 85),
requested licensees to complete an IPEEE. Th  e purpose of the IPEEE is to develop appreciation of severe accident behavior, (2) understand the most likely severe accident sequences that occur under full power conditions, (3) gain a qualitative understanding of the overall likelihood of core damage and radioactive material release, and (4) to identify potential plant enhancements to reduce the overall likelihood of core damage and radioactive material releases. By letters dated December 20, 1991(Reference  86);
January 5, 1995 (Reference  87); March 1, 1995 (Reference 88); and November 20, 1995 (Reference 89); Monticello responded to Generic Letter 88-20, Supplement 4.
 
By {{letter dated|date=March 1, 1995|text=letter dated March 1, 1995}} (Reference 88), Monticello forwarded the report documenting the results of the Monticello Individual Plant Examination of External Events (IPEEE) as requested by Generic Letter 88  -20. This report addressed internal fires, high winds, floods and other credible events. By {{letter dated|date=November 20, 1995|text=letter dated November 20, 1995}} (Reference 89), Monticello submitted revised information concerning the evaluation of internal fires as well as the seismic event evaluation.
 
The IPEEE evaluation of seismic, internal fires, high winds, floods and other credible events provides valuable insights concerning the safety significance of various postulated accidents and failures. The NRC review of information submittals related to IPEEE has determined that no vulnerabilities associated with aspects of external events were identified and the staff considers these issues resolved for Monticello (Reference 119).
 
MNGP conducted an evaluation to identify the risk implications due to Extended Power Uprate (EPU) operation at 2004 MWt at MNGP. Risk impacts due to internal and external events were evaluated. The results indicate that the risk impact is acceptable. The risk assessment report and the associated NRC reviews are contained in References 130, 131, 132, 133, and 134.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 9 of 239
 
14.1.1  General Safety Design Basis
 
Limits on plant operation are established to ensure that the plant can be safely operated and not pose any undue risk to the health and safety of the public. This is accomplished by demonstrating that radioactive release from the plant for normal operation, transients, and postulated accidents meets applicable regulations in which conservative limits are documented.
 
14.1.2  Operational Design Basis
 
The objective for normal operation and transient events is to maintain nucleate boiling and thus avoid a transition to film boiling. Operating limits are specified to maintain adequate margin to the onset of the boiling transition. The figure of merit utilized for plant operation is the Critical Power Ratio (CPR). This is defined as the ratio of the critical power (bundle power at which some point within the assembly experiences onset of boiling transition) to the operating bundle power. The critical power is determined at the same mass flux, inlet temperature, and pressure which exists at the specified reactor condition. Thermal margin is stated in terms of the value of the Minimum Critical Power Ratio, MCPR, which corresponds to the most limiting fuel assembly in the core. To ensure that adequate margin is maintained, a design requirement based on a statistical analysis was selected as follows:
 
The limiting value of MCPR is to be established such that at least 99.9% of the fuel rods in the core are not expected to experience boiling transition during normal operation or AOOs (Reference 217).
 
Both the transient (safety) and normal operating thermal limits in terms of MCPR are derived from this basis. A discussion of these limits is given in Sections 14.2 and 14.3 (See Section 3.2 for a more detailed discussion.)
 
14.1.3  Primary System Integrity Design Basis
 
The ASME Boiler and Pressure Vessel Code and other codes and standards require that the pressure relief system prevent overpressurization of the primary system process barrier and the pressure vessel. The allowable pressure and prescribed evaluations are determined by these requirements. The analysis performed to demonstrate conformance to the requirements is documented in Section 14.5.
 
14.1.4  Plant Stability Design Basis
 
Three types of stability are considered in the design of boiling water reactors:  (1) reactor core (reactivity) stability; (2) channel hydrodynamic stability; and (3) total system stability. A stable system is analytically demonstrated if no inherent limit cycle or divergent oscillation develops within the system as a result of calculated step disturbances of any critical variable, such as steam flow, pressure, neutron flux, or recirculation flow. The criteria for evaluating reactor dynamic performance and stability are stated in terms of two compatible parameters. First is the decay ratio, x2/x0, which is the ratio of the magnitude of the second overshoot to the first overshoot resulting from a step perturbation. A plot of the decay ratio is a graphic representation of the physical responsiveness of the system which is readily evaluated in a time-domain analysis.
Second is the damping coefficient,  n the definition of which corresponds to the dominant pole pair closest to the imaginary axis in the s-plane for the system
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 10 of 239
 
closed-loop transfer function. As  n decreases, the closed-loop roots approach the imaginary axis and the response becomes increasingly oscillatory. This parameter also applies to the frequency-domain interpretation.
 
Detailed evaluations have been conducted to substantiate that the design of the Monticello Nuclear Generating Plant is adequate with respect to thermal hydraulic stability. Additional information concerning these evaluations is provided in Section 14.6.
 
14.1.5  Design Basis for Accidents
 
The effects of the various postulated accidents are investigated for a variety of plant conditions in Section 14.7. Accident limits are specified as follows:
: a. calculated radioactive material releases do not result in exposures exceeding the limits of 10CFR50.67;
: b. catastrophic failure of fuel cladding, including fragmentation of fuel cladding and excessive fuel enthalpy is not predicted;
: c. nuclear system or containment (when required) stresses in excess of those allowed for accidents by applicable codes will not result;
: d. dose received by Control Room operators will not exceed the limits of 10CFR50.67 or 10CFR50 Appendix A, GDC 19.
 
14.2 Fuel Cladding Integrity Safety Limit
 
The generation of the Safety Limit Minimum Critical Power Ratio (SLMCPR) requires a statistical analysis of the core near the limiting Minimum Critical Power Ratio (MCPR) condition. The statistical analysis is utilized to determine the MCPR corresponding to the transient design requirement given in Section 14.1. This MCPR established Fuel Cladding Integrity Safety Limit applies not only for core wide transients, but is also conservatively applied to the localized rod withdrawal error transient.
 
The statistical analysis utilizes a model of the BWR core which simulates the process computer function. This code produces a critical power ratio (CPR) map of the core based on inputs of power distribution and flow and on heat balance information.
 
Bounding cycle specific statistical analyses performed by Framatome provide conservative SLMCPRs for each operating cycle. The SLMCPRs for the current reactor fuel cycle are provided in Technical Specification 2.1.1.3 and the COLR.
 
In order to account for the differences in core flow and assembly power uncertainty during single loop operation, separate SLMCPRs are calculated for single and two loop operation.
(Reference 217)
 
See section 3.2.4.3 for discussion on calculation of the SLMCPR95/95 and SLMCPR99.9%.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 11 of 239
 
The ACE critical power correlation will be used for the ATRIUM 10XM fuel (Reference 214). The ACE critical power correlation for ATRIUM 11 fuel (Reference 1) is used for ATRIUM 11 fuel. These are NRC approved correlations based on Framatome full scale bundle test data.
 
As described in Reference 189, a 0.03 SLMCPR adder is required when evaluating events that initiate from above 42 MWt/Mlbm/hr in the EFW domain.
 
14.3 Operating Limits
 
14.3.1  MCPR Calculational Procedure
 
A reload specific Operating Limit Minimum Critical Power Ratio (OLMCPR) is established to ensure that the Fuel Cladding Integrity Safety Limit (i.e., the Safety Limit Minimum Critical Power Ratio, SLMCPR) is not exceeded for any moderate frequency transient. OLMCPR are established as a function of core power and as a function of core flow. Analyses of moderate frequency events are performed to establish OLMCPR for operation within the licensed Power - Flow operating map. OLMCPR are established by adding the maximum CPR for the most limiting moderate frequency event to the SLMCPR. A summary of the analyses performed in support of the current cycle can be found in Section 14A.
 
14.3.1.1  General Assumptions and Models
 
Safety Evaluation Methods
 
The transient, accident, and steady-state analysis methods used are consistent with the methods described in the NRC approved topical report for Monticello or an NRC approved topical report for the supplier of the analysis service (References 3, 20, 78, 106, 201, 202, 203, 217).
 
The fuel bundle critical power ratios are calculated using approved correlations specific to the fuel types used in the core.
 
A conservative, usually maximum, power condition, is assumed with thermally limited fuel conditions. The philosophy with respect to using the equipment performance components of the transient models for design and safety evaluations is to consider conservative performance of key components. Circuitry delays in the reactor protection system as well as other key equipment circuit delays are conservatively assumed. CPR limits are provided for varying scram insertion times. The setpoints for the safety/relief valves both in the safety and relief function for pressure scram are assumed at their specified limits with added uncertainties. Other equipment performance such as relief and safety valve opening characteristics, recirculation pump drive train inertia, and main steam line isolation valve closing times are conservatively assumed.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 12 of 239
 
End of Hot Full Power Reactivity Considerations
 
End of Hot Full Power (EHFP) conditions for nuclear data are used (except where specific exposure dependent evaluations are performed) to provide a varying level of conservatism associated with core exposure aspects. The nuclear data which are re-evaluated for each reload analysis are the scram reactivity function, void reactivity function and Doppler reactivity function.
 
Scram reactivity is the worth of control rods as a function of time or position following the scram signal. The scram reactivity insertion is normally lowest at the End of Hot Full Power (all rods out condition) because there are no stubbed rods to insert negative reactivity more quickly than the remaining blades of the control rod bank.
 
The void reactivity coefficient is an important parameter, not only in transient analysis, but also in core stability. The core average void coefficient must be negative; however, it must not be so negative as to yield such a strong positive reactivity feedback during void collapse events that core and vessel limits are threatened. Conversely, events with void increase must produce sufficient negative feedback to maintain operation within safety limits.
 
The presence of U-238 and, ultimately, Pu-240 contributes to yield a strong negative Doppler coefficient. This coefficient provides instantaneous negative reactivity feedback to any fuel temperature rise, either gross or local. The magnitude of the Doppler coefficient is not dependent on gadolinium position or concentration in any bundle because gadolinium has very little effect on the resonance group flux or on U-238 content of the core.
 
14.3.1.2  Calculation of Operating Limit MCPR for Core Reload
 
The Operating Limit Minimum Critical Power Ratio (OLMCPR) at full power, and off-rated power and flow conditions, is determined by analyzing the most limiting events and calculating a conservative margin which would prevent 99.9% of the fuel from entering into the transition boiling flow regime. The severity of event and the impact on the OLMCPR is primarily a function of the following factors:
 
cycle operating plan; including fuel characteristics, reload size, cycle length, setpoints, and operational flexibility,
 
operating power and flow conditions,
 
core depletion, and
 
measured cycle specific Control Rod Drive (CRD) scram times.
 
The factors listed above cover a range of operating conditions, therefore the OLMCPR may also vary with changes in operating conditions. The cycle specific analyses attempt to provide plant operational flexibility while maintaining the required margins to operating limits in order to ensure safety.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 13 of 239
 
The dependence of the OLMCPR on the core power and flow conditions is determined through the use of the methodology outlined in References 20 and 201 and the Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement Program (ARTS) (Reference 62).
 
The OLMCPR may also vary with the ability of the CRDs to insert within a specified time and mitigate the consequences of a transient or accident.
 
Reload dependent plant initial conditions, parameters, and Operating Limit Minimum Critical Power Ratio (OLMCPR) results for the limiting core wide transients are given in the current cycles reload licensing analysis. The results of the current cycles analysis are summarized in Section 14A. Densification power spiking is not considered in establishing the OLMCPR.
 
Reduced Flow Considerations - ARTS
 
Flow dependent CPR limits are necessary to assure that the Safety Limit Minimum CPR (SLMCPR) is not exceeded during flow runout events. The design basis flow runout event is a slow flow/power increase event which stabilizes at a new core power corresponding to the maximum possible core flow or may be terminated by a scram. Flow runout events are analyzed along a constant xenon flow control line assuming an equilibrium plant heat balance at each flow condition.
 
The results of this analysis are shown in the Flow Dependent MCPR(F) Limits Figure in the current cycle core operating limits report (Reference 125). In the figure, the flow dependent MCPR limit is referred to as MCPR(F).
 
Reduced Power Considerations - ARTS
 
For power levels above the power level P-Bypass, the point where the reactor scram signals from turbine stop valve closure and turbine control valve fast closure are bypassed, a boundary transient severity trend [CPR = f(P)] was established. Even with the transient severity increase included as a result of assuming constant core flow, large margins still exist between the required thermal limits and expected operating plant performance at lower power levels. Accordingly, above P-Bypass, bounding power dependent trend functions have been developed.
 
A conservative set of CPR limits were also established for operation below P-Bypass.
To maximize operating flexibility CPR limits are provided for both high and low flows.
Therefore, below P-Bypass, both high and low core flow sets of CPR operating limits are provided. No thermal monitoring is required below 25% power. For MNGP the average bundle power density at 25 percent  of rated power is 1.0 MWt. The design limit for not monitoring thermal limits is an average bundle power < 1.2 MWt. This supports no thermal limit monitoring is required below 25 percent power (Reference 134 and 159).
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 14 of 239
 
Coastdown Considerations
 
Once the plant reaches an EHFP condition it may shutdown for refueling or it may be placed in a coastdown mode of operation. In this type of operation the control rods are typically held in the all-rods-out position and the plant is allowed to coastdown to a lower percent of rated power while maintaining flow within the allowable areas of the Power-Flow operating map.
 
Transient analyses are performed to bound the Power  -Flow operating map at all cycle exposures including coastdown operation.
 
Refer to Section 3 for further discussion of operating limit thermal margins.
 
14.3.2  Calculation of MAPLHGR for Core Reload
 
Another Technical Specification limitation on plant operation is the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR). MAPLHGR limits originate from and are associated with LOCA analyses (see Section  14.7.2).
 
For Framatome ATRIUM 10XM fuel, the MAPLHGR ECCS        -LOCA limit is determined by applying the EXEM BWR  -2000 and AURORA-B Evaluation Model for the analysis of the limiting LOCA event (Reference 218 and 223). For ATRIUM 11 fuel, the MAPLHGR ECCS-LOCA limit is determined by applying the AURORA    -B LOCA Evaluation Model for the analysis of the limiting LOCA event (Reference 3, 4, and 221). Power- and flow-dependent MAPLHGR multipliers are not required. Operation with only one recirculation loop (single-loop operation) requires that a MAPLHGR multiplier be applied to the two-loop operation MAPLHGR limit. Calculations confirm that the LOCA acceptance criteria in the Code of Federal Regulations (10 CFR 50.46) are met for operation at or below these MAPLHGR limits.
 
14.3.3  Calculation of LHGR for Core Reload
 
An additional Technical Specification limitation on plant operation is the Linear Heat Generation Rate (LHGR). The current cycle core operating limits report (Reference 125) provides the LHGR standard limits and includes the flow and power dependent multipliers.
 
Framatome established power- and flow-dependent LHGR multipliers. LHGR multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during moderate frequency events initiated from off-rated conditions. LHGR multipliers are less than or equal to 1.0. An exposure dependent LHGR limit is established for each fuel design type (i.e. the same LHGR design limit is applicable to all ATRIUM 10XM fuel). The LHGR limit applicable when the core is operating at reduced core power is determined by multiplying the fuel design LHGR limit by the power dependent LHGR multiplier. Similarly, the LHGR limit applicable when the core is operating at reduced core flow is determined by multiplying the fuel design LHGR limit by the flow dependent LHGR multiplier. When the core is operating at reduced core power and reduced core flow, the LHGR limit for these conditions is the lower of the LHGR limit applicable for the reduced core power and the LHGR limit applicable for the reduced core flow.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 15 of 239
 
In order to limit the number of statistical runs because both AURORA-B and RODEX4 are statistical, conservative measures are used to ensure that RODEX4 calculation inputs appropriately bound the results of the AURORA-B AOO analysis. See ANP-2924P (Reference 7) for further discussion of the conservative measures.
 
14.3.4 Power to Flow Operating Map
 
The standard power/flow map as described in Figure 3.2-1 defines the region of normal plant operations. This includes the region which was added to increase operational flexibility which is an expansion of the power/flow map as defined in the FSAR (Figure 3.2-3) and the Extended Load Line Limit region as defined in Reference 66. The Maximum Extended Load Line Limit Analysis (MELLLA) was performed by GE (Reference 71). The analysis expands the allowable operating domain to the MELLL rod line. Subsequent to MELLLA, an Increased Core Flow (ICF) analysis was performed (Reference 74) which further expanded the power/flow map to areas with 6 core flows larger than the rated value of 57.6 x 10 lb/hr.
 
The expansion of the power/flow map into the ICF region was originally accomplished under Reference 74. Extended Power Uprate operation at 2004 MWt into the ICF region of the power/flow map was evaluated and determined to be acceptable (References 134 and 160).
 
NRC approved a transition to Framatome ATRIUM 10XM fuel and Framatome safety analysis methods in Reference 220 with new safety analysis methods and ATRIUM 11 fuel approved in Reference 5.
 
The operating domain was also expanded to include the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) region. The expansion involved a comprehensive safety analysis (Reference 182) that was approved by the NRC (Reference 184). The scope of the safety analysis included generic evaluations in accordance with the MELLLA+
Licensing Topical Report (Reference 185) that apply to MNGP and certain plant-specific analysis including operation in the ICF region. All lines on the Power to Flow map, other than those associated with the MELLLA+ expansion, are unchanged by MELLLA+. The existing MELLLA boundary is used to establish the operating domain for core flows outside of the MELLLA+ and ICF regions. Single Loop Operation (SLO), is allowed in the MELLLA region only. With the transition to Framatome fuel and methodologies, a power-flow map region designated as the Extended Flow Window (EFW) was created.
The boundaries of the EFW region are exactly the same as for MELLLA+, so prior non-fuel-dependent analyses for MELLLA+ remain applicable. See USAR section 3.2.6 for a description of the MELLLA+/EFW region of the power -flow map.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 16 of 239
 
14.4 Transient Events Analyzed for Core Reload
 
According to transient analysis performed for the initial licensing for Monticello, plant system disturbances caused by single operator error or a single equipment malfunction can be assigned to one of eight separate categories.
 
(1)  Nuclear system pressure increase - threatens to rupture the reactor coolant pressure boundary from internal pressure. Also a pressure increase collapses the voids in the moderator. This causes an insertion of positive reactivity which may result in exceeding the fuel cladding safety limits.
 
(2)  Reactor vessel water (moderator) temperature decreases - results in an insertion of positive reactivity as density increases. Positive reactivity insertions threaten the fuel cladding safety limits because of higher power.
 
(3)  Positive reactivity insertion - is possible from causes other than nuclear system pressure or moderator temperature changes. Such reactivity insertions threaten the fuel cladding safety limits because of higher power.
 
(4)  Reactor vessel coolant inventory decrease - threatens the fuel as the coolant becomes unable to maintain nucleate boiling.
 
(5)  Reactor core coolant flow decrease - threatens the fuel cladding safety limits as the coolant becomes unable to maintain nucleate boiling.
 
(6)  Reactor core coolant flow increase - reduces the void content of the moderator, resulting in a positive reactivity insertion. The resulting high power may exceed fuel cladding safety limits.
 
(7)  Core coolant temperature increase - could exceed fuel cladding safety limits.
 
(8)  Excess of coolant inventory - could result in damage resulting from excessive carry-over.
 
In order to address all of the credible transient events in these eight categories, the initial operating license for Monticello was based on the analysis of 16 FSAR events, each assignable to one of the above categories. In this manner, the most severe transient events relative to LHGR, CPR, and Reactor Coolant System pressure were identified. The relative and absolute severity of the consequences of the events are generally cycle specific. Most of the events result in fairly mild plant disturbances. Only a few events are severe enough to be potentially limiting. Although the most limiting event differs from reload-to-reload, experience shows that the most limiting transient comes from the same selected group of transient events.
 
The original FSAR transient analysis was migrated to the GE methodology for transient analysis as referenced in Core Operating Limits Report (COLR). The adoption of the GE methodology resulted in some refinement of event classification, reload transient analysis and the addition of stability transients. As approved by the NRC in Amendment 210 (Reference 5), transient analysis is now performed using Framatome methods. The core operating limits are developed using NRC approved methodology referenced in the COLR.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 17 of 239
 
The need to analyze various Anticipated Operational Occurrences (AOOs) as part of a reload safety analysis has been generically defined.
 
The limiting events that are within the typical core reload evaluation scope are discussed in Reference 106 and in Section 3.1.5 of Reference 184 for the MELLLA+ operating domain.
The MELLLA+ evaluation included a review of Anticipated Operational Occurrences (AOO) transients and reported the results in Chapter 9 of the SAR (Reference 182). The result of this evaluation is that most transient analyses are either unaffected by the MELLLA+
operating domain extension or are bounded by other analyses. The AOOs analyzed in the SAR for the MELLLA+ domain extension include the following:
 
Generator Load Rejection Without Bypass (LRNBP)
 
Turbine Trip with Bypass (TTWBP)
 
Turbine Trip Without Bypass (TTNBP)
 
Feedwater Controller Failure (Maximum Demand) (FWCF)
 
Loss of Feedwater Heater (LFWH)
 
These AOOs were evaluated at 2004 MWt and two flows: the increased core flow (ICF) limit of 105 percent and the MELLLA+ reduced core flow limit of 80 percent. The comparisons show that for all cases, the ICF conditions are more limiting, indicating no impact for MELLLA+ operation on delta-CPR.
 
The Loss of Feedwater Flow event (LOFW) is not a reload transient but was evaluated in the MELLLA+ Safety Analysis Report in accordance with generic licensing requirements for power uprates. The results demonstrated that the RCIC system is capable of maintaining the water level inside the shroud above the top of active fuel during the LOFW transient.  (Reference 182)
 
The Control Rod Withdrawal Error from Subcritical or Low Power Startup was generically dispositioned in the MELLLA+ Safety Analysis Report and is discussed in USAR sections 7.3.4.3 and 14.4.3. (Reference 182)
 
Descriptions of certain limiting events are given below. The analytical results of the most limiting transient in each of the above types of events is provided in the reload licensing analysis. Input parameters and plant initial conditions used in the transient analysis for the current reload are listed in Section 14A.
 
Reference  6 describes transient analyses performed using Framatome    methodology.
Framatome  transient event analysis is for similar events and with a similar approach as GE methodology, but with different computer models and codes and with analysis inputs specific to those models and codes.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 18 of 239
 
Reference 6 includes two major types of results. One is a Disposition of Events (DOE).
The DOE reviewed each applicable section of the USAR to determine the impact of Framatome fuel on that section. The DOE determined whether no re-analysis was required, whether cycle-specific analysis was required, or whether initial-licensing-only analysis was required. More specifically, the DOE concluded one of the following:
: 1)  No further analysis required. This classification may result from one of the following:
* The consequences of the event have been previously shown to be bounded by consequences of a different event and the introduction of a new fuel design and new analysis methods does not change that conclusion.
* The consequences of the event are benign, i.e., the event causes no significant change in margins to the operating limits.
* The event is not affected by the introduction of a new fuel design, transition to new analysis methods and/or the current analysis of record remains applicable.
: 2)  Address event each following reload. The consequences of the event are potentially limiting and need to be addressed each reload.
: 3)  Address event for initial licensing analysis. This classification may result from one of the following:
* The analysis is performed using conservative bounding assumptions and inputs such that the initial licensing analysis results will remain applicable for following reloads of the same fuel design.
* Results from the initial licensing analysis will be used to quantitatively demonstrate that the results remain applicable for following reloads of the same fuel design because the consequences are benign or bounded by those of another event.
 
The second type of results described in Reference 6 are descriptions and conclusions of initial licensing analysis (item 3) above) and cycle-specific results for a representative core (item 2) above). The process of using a representative core for licensing fuel transitions has precedent. The precedent recognizes that a representative core design is adequate for the purposes of: (1) demonstrating the core design meets the applicability requirements of the new analysis methods, (2) demonstrating that the results can meet the proposed safety limits, and (3) demonstrating either existing Technical Specifications do not need to be revised for the fuel transition or that needed revisions are identified. The representative core design for these analyses assures the actual core design meets all these objectives.
When the first actual reload of Framatome fuel was operated, the applicable cycle-specific licensing analyses was updated for the then-current core loading and the Core Operating Limits Report (COLR) for cycles including Framatome fuel includes applicable limits to ensure safe core operation.
 
See Table 14.4-1 for the transients that are analyzed in the current cycle analysis. This section classifies each transient by disposition type.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 19 of 239
 
14.4.1  Generator Load Rejection Without Bypass
 
Fast closure of the turbine control valves is 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 rapidly as possible to prevent overspeed of the turbine generator rotor. The closing causes a sudden reduction of steam flow which results in a nuclear system pressure increase. The reactor is scrammed by the fast closure of the turbine control valve. The analysis is performed with a statistical approach.
 
14.4.1.1  Starting Conditions and Assumptions
 
The following plant operating conditions and assumptions form the principal basis for which reactor behavior is analyzed during a load rejection.
 
(1)  The reactor and turbine generator are initially operating at full power when the load rejection occurs.
 
(2)  All of the plant control systems continue normal operation
 
(3)  Auxiliary power is continuously supplied at rated frequency.
 
(4)  The reactor is operating in the manual flow control mode when load rejection occurs.
 
(5)  The turbine bypass valve system is failed in the closed position.
 
14.4.1.2  Event Description
 
Complete loss of the generator load produces the following sequence of events:
 
(1)  The power/load unbalance device steps 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.
 
(2)  Reactor scram is initiated upon sensing control valve fast closure.
 
(3)  If the pressure rises to the pressure relief set point, part or all of the relief valves open, discharging steam to the suppression pool.
 
14.4.1.3  Acceptance Criteria
 
The acceptance criteria for this transient are based on General Design Criteria (GDC) 10 and 26 for fuel design limits and GDC 15 with respect to reactor coolant pressure limits. This means the CPR for the transient is greater than the safety limit and the pressure in the RCS is less than 110% of the design pressure.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 20 of 239
 
14.4.1.4  Main Physics Parameters
 
The core behavior of interest is the pressure increase  which causes the collapse of steam voids with the corresponding increase in neutron flux level. The increase in power is curtailed by the Doppler feedback and reactor scram. Thus, the main physics parameters of interest are the void coefficient, Doppler coefficient and scram worth.
 
14.4.1.5  Event Results
 
Results of the analysis for this transient for the current cycle are shown in Section 14A.
 
14.4.2  Loss of Feedwater Heating
 
A loss of feedwater heating transient can occur as a result of a loss of extraction steam to a feedwater heater or from inadvertent actuation of high pressure coolant injection which delivers relatively cool water to the reactor through the feedwater sparger. Loss of feedwater heating results in a core power increase due to the increase in core inlet subcooling. If the neutron power exceeds the reactor trip setpoint, a scram occurs; otherwise the system settles to a steady state higher power condition until the operator intervenes. The analysis is performed with a statistical approach.
 
14.4.2.1  Starting Conditions and Assumptions
 
The following plant operating conditions and assumptions form the principal basis for which reactor behavior is analyzed during the loss of feedwater heating transient:
 
(1)  The plant is operating at full power.
 
(2)  The plant is operating in the manual flow control mode.
 
14.4.2.2  Event Description
 
Feedwater heating can be lost in at least two ways:
 
(1)  Steam extraction line to heater is closed.
 
(2)  Feedwater is bypassed around heater.
 
(3)  Inadvertent actuation of high pressure coolant injection.
 
The first case produces a gradual cooling of the feedwater. In the second case the feedwater bypasses the heater and no heating of the feedwater is generated. In the third case cool water is injected in the reactor through the feedwater sparger. In any of these cases the reactor vessel receives cooler feedwater. The maximum number of feedwater heaters which can be tripped or bypassed by a single event represents the most severe transient for analysis considerations. This event of an instantaneous loss of the feedwater heating capability of the plant causes an increase in core inlet subcooling. This increases core power due to the negative void reactivity coefficient.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 21 of 239
 
In any case power would increase at a very moderate rate. If power exceeded the normal full power flow control line, the operator would be expected to insert control rods to return the power and flow to their normal range. If this were not done the neutron flux could exceed the scram set point where a scram would occur.
 
If the reduction in feedwater heating is due to inadvertent HPCI injection, the feedwater temperature changes only in a single loop of feedwater. This asymmetric temperature distribution carries through to the core. The degree of mixing between the cold side (with HPCI injection) and the hot side (without HPCI injection) is conservatively determined via computational fluid dynamics.
 
14.4.2.3  Acceptance Criteria
 
The acceptance criteria for this transient are based on GDC 10, 15 and 26. The relevant criteria is the maintenance of the fuel cladding integrity by ensuring that the CPR remains above the safety limit.
 
14.4.2.4  Main Physics Parameters
 
The core behavior of interest is the increase in inlet subcooling and the subsequent reduction in core voids which result in an increase in core power. The increase in power is curtailed by the Doppler feedback and in some cases by a reactor scram.
Thus, the main physics parameters of interest are the void coefficient, Doppler coefficient and scram worth if a scram occurs.
 
14.4.2.5  Event Results
 
Results of the analysis for this transient for the current cycle are shown in Section 14A.
 
14.4.3  Rod Withdrawal Error
 
The current Rod Block Monitor (RBM) system for Monticello with power dependent setpoints was analyzed for the rod withdrawal error (RWE) using a statistical analysis approach.
 
14.4.3.1  Starting Conditions and Assumptions
 
The reactor operator has followed procedures and up to the point of the withdrawal error is in the normal mode of operation (i.e., the control rod pattern, flow set points, etc., are all within normal operating limits).
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 22 of 239
 
14.4.3.2  Event Description
 
For a RWE, it is assumed that the reactor is in a normal mode of operation and the operator makes a procedural error resulting in an uncontrolled withdrawal of the maximum worth control rod. The positive reactivity insertion causes the average core power to increase. More importantly, the local power in the vicinity of the withdrawn control rod will increase and could potentially cause cladding damage due to either overheating which may accompany the occurrence of boiling transition, by exceeding the 1% plastic strain limit imposed on the cladding, or due to fuel centerline melt, which are the assumed transient failure thresholds.
 
The control rod withdrawal is terminated either by the rod being fully withdrawn or by the RBM. The feedback from the voids and fuel temperature will limit the power increase and following termination of the control rod withdrawal a new equilibrium power level will be reached unless a reactor trip setpoint is reached.
 
Approximate Elapsed Time from Start of Rod Motion                                  Event 0                                  (1)  Event begins, operator selects the control rod, acknowledges any alarms and withdraws the rod at the maximum rod speed.
5 seconds                        (2)  Core average power and local power increase.
30 seconds                        (3) Event ends - rod block by RBM
 
Identification of Operator Actions:
 
Under most normal operating conditions, no operator action will be required since the transient which will occur will be very mild. If the peak linear power design limits are exceeded, the official core monitor will display the abnormal condition, and the operator will take appropriate action to rectify the situation.
 
If the rod withdrawal error is severe enough, the Rod Block Monitor (RBM) system will sound alarms, at which time the operator must acknowledge the alarm and take corrective action. Even for extremely severe conditions (i.e., for highly abnormal control rod patterns, operating conditions, and assuming that the operator ignores all alarms and warnings, and continues to withdraw the control rod), the RBM system will block further withdrawal of the control rod before exceeding either the OLMCPR or the 1% plastic strain or fuel centerline melt limits.
 
14.4.3.3  Acceptance Criteria
 
The acceptance criteria for this transient are based on GDC 10, 20, and 25. The fuel design criteria are met when the CPR for the transient is greater than the safety limit and when the uniform cladding strain does not exceed 1%.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 23 of 239
 
14.4.3.4  Main Physics Parameters
 
The core behavior of interest is the reactivity addition by a single rod with the corresponding increase in local power. The feedback from the voids and fuel temperature will limit the power increase and following termination of the control rod withdrawal a new equilibrium power level will be reached unless a reactor trip setpoint is reached.
 
In order to limit the number of statistical runs, conservative measures are used to ensure that RODEX4 calculation inputs appropriately bound the results of the AURORA-B AOO analysis.
 
14.4.3.5  Event Results
 
Results of the analysis for this transient for the current cycle are shown in Section 14A.
 
14.4.4  Feedwater Controller Failure - Maximum Demand
 
The feedwater controller is assumed to fail in such a manner as to cause an increase in feedwater flow and thus increasing the core coolant inventory and decreasing the coolant temperature. The most severe event is a feedwater controller failure during maximum flow demand in manual flow control mode. The influx of excess feedwater flow results in an increase in core subcooling which reduces the void fraction and thus induces an increase in reactor power. The excess feedwater flow also results in a rise in the reactor water level which eventually leads to high water level trip setpoint being exceeded causing a turbine trip, feedwater pump trip and a subsequent reactor scram due to turbine stop valve closure. The transient is mitigated by opening the turbine bypass valves and the safety/relief valves. The analysis is performed with a statistical approach.
 
14.4.4.1  Starting Conditions and Assumptions
 
The following plant operating conditions and assumptions form the principal basis for which reactor behavior is analyzed during a feedwater controller failure.
: a. Feedwater controller fails during maximum flow demand.
: b. Maximum feedwater pump run-out is assumed.
: c. The reactor is operating in a manual flow control mode which provides for the most severe transient.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 24 of 239
 
14.4.4.2  Event Description
 
A feedwater controller failure during maximum demand produces the following sequence of events:
: a. The reactor vessel receives an excess of feedwater flow.
: b. The excess flow results in an increase in core subcooling, which results in a rise in core power and reactor vessel water level.
: c. The rise in the reactor vessel water level eventually leads to high water level turbine trip, feedwater pump trip and reactor scram due to turbine stop valve closure.
: d. The transient is mitigated by opening the turbine bypass valves and the safety/relief valves.
 
14.4.4.3  Acceptance Criteria
 
The acceptance criteria for this transient are based on GDC 10, 15 and 26. The relevant criterion is the maintenance of the fuel cladding integrity by ensuring that the CPR remains above the safety limit.
 
14.4.4.4  Main Physics Parameters
 
The core behavior of interest is the increase in core subcooling due to the increased feedwater flow which causes a decrease in the core voids which results in an increase in core power. The increase in core power is curtailed by the Doppler feedback and the transient results in a reactor scram. Thus, the main physics parameters of interest are the void coefficient, Doppler coefficient and scram worth.
 
14.4.4.5  Event Results
 
The influx of excess feedwater flow results in an increase in core subcooling which reduces the void fraction and thus induces an increase in reactor power. The excess feedwater flow also results in a rise in the reactor water level which eventually leads to high water level; main turbine and feedwater trip and turbine bypass valves are actuated. Reactor scram trip is actuated from main turbine stop valve position switches. Relief valves open as steamline pressures reach relief valve setpoints.
 
Results of the analysis for the current cycle are shown in Section 14A.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 25 of 239
 
14.4.5  Turbine Trip Without Bypass
 
This transient is similar to the generator load rejection without bypass in that it results in a nuclear system pressure increase. The transient is initiated from a high power level without turbine bypass valves opening following closure of the turbine stop valves. The stop valve closure results in a scram and the primary system relief valves open to limit the pressure increase. For the case of bypass valves opening, the transient is less severe. The analysis is performed with a statistical approach.
 
14.4.5.1  Starting Conditions and Assumptions
 
The following plant operating conditions and assumptions form the principal basis for which reactor behavior is analyzed during turbine trip without bypass.
 
(1)  The reactor and turbine generator are initially operating at full power.
 
(2)  All of the plant control systems continue normal operation.
 
(3)  Auxiliary power is continuously supplied at rated frequency.
 
(4)  The reactor is operating in the manual flow control mode.
 
(5)  The turbine bypass valve system is failed in the closed position.
 
14.4.5.2  Event Description
 
Turbine trip without bypass produces the following sequence of events:
 
(1)  The turbine trip initiates a reactor scram on stop valve closure.
 
(2)  If the pressure rises to the pressure relief set point, part or all of the relief valves open, discharging steam to the suppression pool.
 
14.4.5.3  Acceptance Criteria
 
The acceptance criteria for this transient are based on GDC 10 and 26 for fuel design limits and GDC 15 with respect to reactor coolant pressure limits. This means the CPR for the transient is greater than the safety limit and the pressure in the RCS is less than 110% of the design pressure.
 
14.4.5.4  Main Physics Parameters
 
The core behavior of interest is the pressure increase which causes the collapse of steam voids with the corresponding increase in neutron flux level. The increase in power is curtailed by the Doppler feedback and reactor scram. Thus, the main physics parameters of interest are the void coefficient, Doppler coefficient and scram worth.
 
14.4.5.5  Event Results
 
Results of the analysis for this transient for the current cycle are shown in Section 14A.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                                  USAR-14 SECTION 14          PLANT SAFETY ANALYSIS                                                  Revision 40 Page 26 of 239
 
Table 14.4-1        Transients and Accidents for Reload Evaluation
 
Current                    No Further Event Type                                                                Cycle          Initial      Analysis Analysis2      Reload1      Required3
: 1)  Primary System Pressure Increase Generator Load Rejection with Bypass Failure Turbine Trip with Bypass Failure (TTNB)
Main Steam Isolation Valve Closure (One / All Valves)
Turbine Trip with Bypass Failure without Position Scram Turbine Trip with Bypass (TTWB) and Degraded Scram Speed Main Steam Isolation Valve Closure without Position Scram Loss of Condenser Vacuum Pressure Regulator Failure - Full Close (Downscale)
Loss of Auxiliary Power - All Grids Loss of Stator Cooling GE SIL 502 (Revision 1) Single Turbine Control Valve Slow Closure
: 2)  Reactor Vessel Water Temperature Decrease Feedwater Controller Failure - Maximum Demand Loss of Feedwater Heating (95.3 oF)
Inadvertent High Pressure Coolant Injection Actuation with L8 (HPCI/L8) Turbine Trip
: 3)  Positive Reactivity Insertion Rod Withdrawal Error Control Rod Drop Accident
: 4)  Reactor Vessel Coolant Inventory Decrease Pressure Regulator Failure - Full Open Inadvertent Opening of Safety/Relief Valve Loss of Feedwater Flow Loss of Auxiliary Power Transformers
: 5)  Core Coolant Flow Decrease Recirculation Flow Control Failure - Decrease Trip of One Recirculation Pump Trip of Two Recirculation Pumps Recirculation Pump Seizure                                          (in SLO)
: 6)  Core Coolant Flow Increase Slow Recirculation Control Failure - Increase (MCPRF, LHGRFACF) 4     
 
Fast Recirculation Control Failure - Increase Startup of an Idle Recirculation Loop
: 7)  Fuel Loading Errors Misoriented Bundle Accident Mislocated Bundle Accident                                             
 
1Disposition of Events for Initial ATRIUM 11 Reload can be found in Reference 6 2Analysis results of events for current cycle are found in Chapter 14A.
3Disposition of events that require no further analysis can be found in Reference 6 4MCPRF and LHGRFACF are flow-dependent MCPR and LHGRFAC.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 27 of 239
 
14.5 Special Events
 
Special events are those items that need to be analyzed to meet a licensing requirement as part of a reload safety evaluation but do not fit into the abnormal transient or accident categories. The special events to be analyzed are:
 
(1)  Overpressure Protection - MSIV and TSV Closure
 
(2)  Standby Liquid Control System Shutdown Margin
 
(3)  Stuck Rod Cold Shutdown Margin
 
In this section, each event is described and the acceptance criteria is given.
 
14.5.1  Vessel Pressure ASME Code Compliance Model
 
The pressure relief system was designed to prevent excessive overpressurization of the primary system process barrier and the pressure vessel and thereby preclude an uncontrolled release of fission products.
 
The vessel overpressure protection system was designed to satisfy the requirements of Section III, Nuclear Vessels, of the ASME Code, 1965 edition (Reference 90). The ASME Code, Section III, for Class I vessels permits pressure transients up to 10% over design pressure, and requires that the nominal setpoint of at least one safety or relief valve be not greater than the vessel design pressure and the setpoint of any additional required valves be not greater than 105% of the vessel design pressure. Section III of the code allows credit to be taken for the scram protection system as a pressure protection device when determining the required safety valve capacities for nuclear vessels. The code required the reactor pressure vessel to be designed to accommodate the normal operating loads and transient startup/shutdown and test cyclic loads expected during the 40-year life of the plant.
 
The two potentially limiting events for the ASME overpressure are the main steam isolation valve (MSIV) closure and the turbine stop valve (TSV) closure. Either event can be initiated by various plant conditions or by various operation actions. Normally, as the MSIVs close, a reactor scram is initiated by position switches which sense closure.
Similarly, as the TSVs close, a reactor scram is initiated by position switches which sense closure, although the TSV scram is bypassed at approximately 25% power based on the first stage turbine pressure. The TSV event from derated conditions (i.e., when the TSV scram is bypassed) is non-limiting compared to the event initiated from high power conditions. As the system isolates, pressure rises in the vessel until the safety/relief valves open to mitigate the event. Sensitivity evaluations indicate that peak pressure in the steam dome of the reactor vessel occurs when the ATWS-RPT (Anticipated Transient Without Scram - Recirculation Pump Trip) trips the recirculation pumps and peak pressure in the lower plenum is higher with the ATWS-RPT disabled, so the event is run both ways to evaluate both steam dome and lower plenum pressure.
The event assumes initiating conditions of 1040 psia, 102% power, failure of the position switch scram, and failure of the turbine bypass valves. The evaluation assumed that only five of the eight valves are operable and that they open at 1145 psig (Reference 228), which is a little over 103% of the valves stamped setpoint of 1109 psig. The analysis is performed with a statistical approach. See USAR Section 14.3.1.1 and Reference 20 for further details.
 
The acceptance criteria for this transient are based on GDC 10, 15 and 26. The main criteria is to demonstrate compliance with the ASME Code by showing that the pressure
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 28 of 239
 
in the reactor vessel remains below 110% of the design value and attached piping systems remain below 120% of design values.
 
The reactor vessel pressure limit is 1375 psig (110% of design). The piping attached to the bottom head has a limit of 1363.2 psig (120% of design). The recirculation pump discharge piping has a limit of 1497.6 psig (120% of design). ATWS RPT is not credited for the evaluation of reactor recirculation pump discharge piping because peak piping pressure occurs without ATWS-RPT. The recirculation pump suction piping has a limit of 1377.6 psig (120% of design). The steam piping  to the outboard MSIV has a limit of 1332 psig (120% of design). Compliance with a steam dome limit of 1332 psig ensures that pressures in the reactor vessel and attached piping remain below their respective limits.
 
The core behavior of interest is the initial pressure surge caused by the valve closure which in turn collapses voids causing a neutron flux spike. The reactor is scrammed due to high neutron flux or high reactor pressure, whichever occurs first. The Doppler feedback also contributes to limiting the power spike. Results of this event for the current cycle are provided in Section 14A.
 
14.5.2  Standby Liquid Control System Shutdown Margin
 
The design objective of the Standby Liquid Control System is to provide the capability of bringing the reactor to a sub-critical condition at any time in the cycle during the most reactive xenon-free state with all the control rods in the full-out condition.
 
To meet this objective, the Standby Liquid Control System is designed to inject a quantity of boron which produces an equivalent concentration of at least 660 ppm of natural boron in the reactor core in less than 125 minutes.
 
The requirements of this system are primarily dependent on the reactor power level along with the reactivity effects of voids and temperature between the hot full power and cold xenon-free condition. The calculations show that the Standby Liquid Control System has sufficient shutdown margin for the current cycle. These results are provided in Section 14A.
 
14.5.3  Stuck Rod Cold Shutdown Margin
 
Shutdown margin is the amount by which the reactor is subcritical with the most reactive control rod in its fully withdrawn position and all other rods fully inserted. The shutdown margin is calculated throughout the cycle for the most reactive core condition which is at the most reactive temperature of  68&deg;F and xenon free. Advanced fuel designs (e.g.,
ATRIUM 10XM) can be more reactive at temperatures greater than 68&deg;F for some exposures.
 
The shutdown margin is a limiting condition for operation (LCO) as specified in the Monticello Technical Specifications. The shutdown margin is required to be verified within limits prior to each in vessel fuel movement during fuel loading sequence and once within 4 hours after criticality following fuel movement within the reactor pressure vessel or control rod placement.
 
A three dimensional Boiling Water Reactor Simulator code was utilized to calculate the stuck rod cold shutdown margin. The calculations show that with the high worth rod out during the current cycle, the core has sufficient shutdown margin. The stuck rod cold shutdown margin results for the current cycle are provided in Section 14A.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 29 of 239
 
14.6 Plant Stability Analysis
 
The stability licensing basis for U.S. nuclear power plants is set forth in GDC-12. NRC Generic Letter 94-02 (Reference 91) requested licensees take actions to ensure compliance with GDC-12. GDC-12 requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. In response to NRC Bulletin 88-07 (Reference 92), the BWR Owners Group, in conjunction with General Electric, implemented a program to develop long-term solutions to the stability issue.
 
The long term stability solution is Best Estimate Enhanced Option III (BEO-III) as described in the approved topical report ANP-10344 (Reference 78), which is based on the Owners Group solution created in response to NRC bulletin 88-07 (Reference 92).
 
Per Reference 78 the hardware implementation, implementation of the channel instability exclusion trip, and specific method of providing automatic stability backup in the event of OPRM failure are plant specific and must be NRC-approved. The hardware implementation at Monticello uses the previously installed Power Range Neutron Monitoring System (PRNMS), which was approved for use at Monticello in License Amendment 159. (Reference 149)
 
The BEO-III solution combines closely spaced LPRM detectors into cells to effectively detect corewide or regional (local) modes of reactor instability. These cells are termed Oscillation Power Range Monitor (OPRM) cells and are configured to provide local area coverage with multiple channels. BEO-III uses the PRNMS hardware to combine the LPRM signals and to evaluate the cell signals with instability detection algorithms. The Period Based Detection Algorithm (PBDA) is the only algorithm credited in the Option III licensing basis. Two defenseindepth algorithms, referred to as the Amplitude Based Algorithm (ABA) and the Growth Rate Based Algorithm (GRBA), offer a high degree of assurance that fuel failure will not occur as a consequence of stability related oscillations.
BEO-III provides SLMCPR protection by generating a reactor scram if a reactor instability, which exceeds the specified trip setpoint, is detected. Settings are established per Reference 78 each cycle and are input into the OPRM software. These are settings i  n the OPRM software and no firmware change is required to perform these setting changes.
While the OPRM settings are calculated each cycle, the settings are not highly dependent on the neutronic fuel type. The BEO-III OPRM trip is armed only when plant operation is within the BEO-III OPRM tripenabled region. The BEO-III OPRM tripenabled region is generically defined as the region on the power/flow map with power  25% of RTP and recirculation drive flow < 60% of rated drive flow.
 
As part of the BEO-III analysis, it is confirmed that single channel instabilities would not result in violation of the SLMCPR. As a result, no automatic scram is required to preclude single channel instabilities. In addition to the automatic protective features, BEO-III also uses Backup Stability Regions I and II as described in Reference 78. As described in L-MT-15-065 (Reference 88), the actions to be taken in the event that the OPRMs are declared inoperable are those methods listed in OG 02-0119-260, Backup Stability Protection (BSP) for Inoperable Option III Solution (Reference 93). The BSP evolved from the stability interim corrective actions (ICAs), which restrict plant operation in the high power, low core flow region of the power/flow operating map. The ICAs provide guidance which reduces the likelihood of an instability event by limiting the period of operation in regions of the power and flow map most susceptible to thermal hydraulic instability. The ICAs also specify operator actions, which are capable of detecting conditions consistent with the onset of oscillations, and additional actions, which mitigate the consequences of oscillations consistent with degraded thermal hydraulic stability performance of the core.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 30 of 239
 
14.7 Accident Evaluation Methodology
 
As stated in Section 14.3, abnormal operating transients are evaluated to determine a plant normal operating MCPR limit. In addition to these analyses, evaluations of less frequent postulated events are made to ensure an even greater depth of safety. Accidents are events which have a projected frequency of occurrence of less than once in every one hundred years for every operating BWR. The broad spectrum of postulated accidents is covered by six categories of design basis events. These events are the control rod drop, loss of coolant accident, main steam line break, fuel assembly loading accidents, recirculation pump seizure, and refueling accident. A description of each of these events follows.
 
The design basis accident radiological consequences analyses were performed using the Alternative Source Term methodology provided in Regulatory Guide 1.183 (Reference 129). Regulatory limits for dose consequences are specified in 10CFR50.67.
Control room operator accident dose limits are also specified in GDC19 of 10CFR50 Appendix A. Offsite (EAB and LPZ) accident dose limits in 10CFR50.67 are supplemented by individual accident dose limits (adjusted for accident probability) specified in RG 1.183.
Control room operator and LPZ (Low Population Zone) doses are calculated for the duration of the accident. EAB (Exclusion Area Boundary) doses are calculated as the worst 2-hour dose for the accident period.
 
The radiological consequences of LOCA inside containment, MSLBA, ILBA, CRDA and FHA are bounded by the evaluation in the MELLLA domain and need not be reevaluated for the MELLLA+/EFW domain. The radiological results for all accidents remain below the applicable regulatory limits for the plant for operation in the MELLLA+/EFW domain (Reference 182).
 
Atmospheric dispersion coefficients (X/Q) were calculated based on site meteorological data from 1998-2002. These coefficients are shown in USAR Table 14.7-23 and in the individual accident sections.
 
The post-operation radiation sources in the core are primarily the result of accumulated fission products. Two separate forms of post-operation source data are normally applied.
The first of these is the core gamma-ray source, which is typically used in shielding calculations for the core and for individual fuel bundles. This source term is defined in terms of MeV/sec per Watt of reactor thermal power  (or equivalent) at various times after shutdown.
 
The second set of post-operation source data consists primarily of nuclide activity inventories for fission products in the fuel. These data are needed for post-accident evaluations, which are performed in compliance with regulatory guidance that applies different release and transport assumptions to different fission products. The core fission product inventories for these evaluations are based on an assumed fuel irradiation time, which develops equilibrium activities in the fuel (typically 3 years). Most radiologically significant fission products reach equilibrium within a 60-day period. The radionuclide inventories are determined in terms of Curies per megawatt of reactor thermal power at various times after shutdown. See Section  2.9.1 of NEDC-33322P (Reference 160).
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 31 of 239
 
The core source term for radiological accident analysis was developed using ORIGEN, Isotope Generation and Depletion Code Matrix Exponential Method. The core isotope inventory was generated from the ORIGEN source term with the exception of Co-58 and Co-60 which were obtained from the BWR default source term values from Table 1.4.3.2-3 of NUREG/CR-6604 (Reference 166). Core inventories are specified in sub-tables of Table 14.7-24 for various core loadings. The inventory used in each event is described in the respective USAR section.
 
Core inventory was developed assuming a power level of 2004 MWt, increased by 2% to account for power measurement uncertainties.
 
14.7.1  Control Rod Drop Accident Evaluation
 
The accidents that result in releases of radioactive material from the fuel with the reactor primary system, primary containment, and secondary containment initially intact are the results of various failures of the control rod drive system. Examples of such failures are collet finger failures in one control rod drive mechanism, a control rod drive system pressure regulator malfunction, and a control rod drive mechanism ball check valve failure. None of the single failures associated with the control rods or the control rod system result in a greater release of radioactive material from the fuel than the release that results when a single control rod drops out of the core after being disconnected from its drive and after the drive has been retracted to the fully withdrawn position. Thus, this control rod drop accident is established as the design basis accident for the category of accidents resulting in radioactive material release from the fuel with all other barriers initially intact.
 
There are many ways of inserting reactivity into a boiling water reactor; however, most of them result in a relatively slow rate of reactivity insertion and therefore pose no threat to the system. It is possible, however, that a rapid removal of a high worth control rod could result in a potentially significant excursion; therefore, the accident which has been chosen to encompass the consequences of a reactivity excursion is the Control Rod Drop Accident (CRDA).
 
The dropping of the rod results in a high local reactivity in a small region of the core and for large, loosely coupled cores, significant shifts in the spatial power generation during the course of the excursion.
 
The key reactivity feedback mechanism affecting the shutdown of the initial prompt power burst is the Doppler coefficient. Final shutdown is achieved by scramming all but the dropped rod.
 
Analysis is performed each reload to confirm that the total radioactivity released during a CRDA due to fuel failure is bounded by the radiological analysis of record and that the core remains in a coolable geometry. The two modes of fuel failure evaluated  are pellet-clad mechanical interaction (PCMI) and fuel temperature. Core coolablity is evaluated by ensuring that fuel enthalpy remains below 230 cal/g and that the mechanical energy of the event does not result in molten fuel contacting the coolant or rod burst. The radiological consequences are evaluating by combining the number of failed rods with appropriate fission gas release models to determine that the overall source term is bounded by the analysis of record. With the application of the AURORA  -B CRDA method (Reference 8) along with a hydrogen multiplier (Reference 7), the Monticello analysis of the CRDA complies with Regulatory Guide 1.236 (Reference      9).
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 32 of 239
 
14.7.1.1  Sequence of Events
 
The sequence of events and approximate time of occurrence for this postulated accident are described below.
Approximate Event                                                Elapsed Time (a)  Reactor is at a control rod pattern corresponding to maximum incremental rod worth.
 
(b)  Rod worth minimizer or operators are functioning within constraints of banked position withdrawal sequence (BPWS) (References 17 and 127). The control rod that will result in the maximum incremental reactivity worth addition at any time in core life under any operating condition while employing the BPWS becomes decoupled from the control rod drive.
 
(c)  Decoupled control rod sticks in the fully inserted position.
 
(d)  Operator selects and withdraws the drive of the decoupled rod along with the other control rods assigned to the banked-position group such that the proper core geometry for the maximum incremental rod worth exists.
 
(e)  Control rod becomes unstuck and drops at the maximum velocity determined from experimental data (3.11 fps).  (Reference 18).
 
(f)  Reactor goes on a positive period; initial power          1 sec burst is terminated by the Doppler reactivity feedback.
 
(g)  APRM Neutron Flux - High signal scrams reactor (conservative; in startup mode, APRM Neutron Flux-High (Setdown) scram would be operative in addition to the IRM).
 
(h)  Scram terminates accident                                5 sec
 
To limit the worth of the rod which could be dropped in a bank position withdrawal sequences (BPWS) plant, the rod worth minimizer system (RWM) is used below 10%
power to enforce the rod withdrawal sequence. The  RWM is programmed to follow the BPWS, which are generally defined in References 17 and 127.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 33 of 239
 
14.7.1.2  Model Parameters Sensitivities
 
Although there are many input parameters to the CRDA analysis, the resultant peak fuel enthalpy was most sensitive to the following input parameters:
 
(1)  Dropped control rod worth,
 
(2)  Doppler reactivity,
 
(3)  Effective delayed neutron fraction,
 
(4)  Fuel rod local peaking factor
 
(5)  Fuel cycle design
 
14.7.1.3  Analytical Methods
 
The AREVA methodology for CRDA analysis involves the selection of conservative initial conditions and candidate rods which would result in the most severe consequences for the CRDA. Following the selection of the candidate control rods, the rod drops are simulated with the AURORA-B code system. The results obtained from the simulation are then evaluated against the acceptance criteria to assess rod failure and subsequent fission product inventory release. The methodology includes generic components, plant specific components, and cycle specific components.
 
In the startup range the initial conditions that must be determined are the initial coolant properties, the initial core power, the initial core flow, and the inoperable control rods. The determination of the initial core flow and the inoperable control rods is determined on plant specific bases. Although the actual static rod worth varies little with moderator temperature, the lower temperature for a given rod worth results in higher fuel enthalpy. Sensitivity studies were performed with respect to the power level. Relative to the reference case, the magnitude of the event decreased as the initial power was increased. Sensitivity studies were performed with respect to the coolant flow rate and determined that total enthalpy rise increases as the initial core flow increases. Since startup occurs with at least one recirc pump at minimum speed, core flow would be greater than 10%, so 10% is used as conservative initial condition.
Inoperable rods locations are defined consistent with those allowed by plant technical specifications in such a manner to maximize the worth of the candidate rods. The rod selection process for the startup range CRDA is based on the static worth of the control rods and a broad range at which criticality is assumed to occur.
 
In order to determine the most limiting rod drop, each rod within a rod pull group is evaluated as if it is the first rod pulled in the group. This results in maximizing the worth of the control rod. As the control rod density in the core decreases and nuclear heating begins, the individual control rod worth also decreases. Following the establishment of the static rod worth values, the rods are identified for which detailed transient analyses are to be performed. Rod selection is based on rod worth and the exposure-dependent minimum failure threshold, see Appendix A of Reference 8 for details.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 34 of 239
 
The standard AURORA-B code system is supplemented with additional capabilities to model cold conditions, additional data transfer, and support for peak rod heat structures. The CRDA scenario is a relatively localized event around the target rod.
The individual fuel assemblies in an array around the target rod are explicitly modeled as individual channels. Both an average fuel rod and peak power fuel rod are evaluated for the individual channels. This detailed target array is then surrounded by buffer rings of hydraulically grouped assemblies. Within the buffer rings the fuel assemblies are assigned in groups with similar geometric, hydraulic, thermal power, and fuel thermal properties. The remainder of the fuel assemblies in the core are assigned to groups with similar geometric, hydraulic, thermal power, and fuel neutronic properties. For the CRDA analysis, a core only model is used in the AURORA-B system.
 
Figures of merit include prompt fuel enthalpy rise compared against the pellet clad mechanical interaction failure threshold, total enthalpy compared against the high temperature cladding failure threshold, and MCPR such that rods enter transition boiling for at-power evaluation. Additionally, peak system pressure, released fission product inventory, and core coolability are assessed.
 
The acceptance criteria make a distinction between prompt and total enthalpy rise. If acceptable results can be obtained using total enthalpy, then total enthalpy rise may be used for the evaluation. However, if there is a significant difference between the prompt and total enthalpy, then both the prompt and total enthalpy may be used for event evaluations. The prompt enthalpy is the radial average fuel enthalpy rise at the time corresponding to one pulse width after the peak of the prompt pulse. Pressure or clad hydrogen content dependent failure thresholds are taken from RG 1.236 (Reference 9) and the standard RODEX4 hydrogen correlation is adjusted to meet the requirements of RG 1.236 is used to determine clad hydrogen content. This adjustment to the standard RODEX4 hydrogen correlation was approved by the NRC in Amendment 210 (Reference 5).
 
The number of unique rods failed due to pellet clad mechanical interaction and/or high temperature is totaled for the cycle specific number of failed rods. No fuel rods are allowed to melt because even incipient melting could result in coolability failures.
 
CRDA in the power range was evaluated using rods in boiling transition as the figure of merit; this evaluation determined that the number of failed rods from at-power conditions is bounded by the number of failed rods from the startup range (Reference 8, section 7.7). Additionally, the vessel pressure impact was evaluated and determined that there is little impact on pressure regardless of whether the event is initiated from startup or at-power conditions, so ASME Service Limit C stresses would not be exceeded.
 
Core coolability is maintained by keeping peak radial average fuel enthalpy 230 cal/g, peak fuel temperature below the fuel melt temperature, and pulse width 25 ms.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 35 of 239
 
14.7.1.4  Effect of Fuel Densification
 
Localized power spikes due to axial gaps in the fuel column would result in a proportional increase in the calculated peak fuel enthalpy.
 
Framatome incorporates fuel densification and swelling effects into its RODEX computer codes, as described in References 219, 220, and 221.
 
14.7.1.5  Results
 
The postulated CRDA event is one in which a high worth control blade is stuck in the fully inserted position and is decoupled from its control rod drive. Sometime after, the control rod drive of this blade is withdrawn and the control blade subsequently drops at the maximum speed and creates a localized power excursion.
 
Cycle specific analysis demonstrates that core coolability is maintained, the number of failed rods less than the number of failed rods assumed in the USAR (850 8x8 equivalent rods - see section 14.7.1.2.), and the released fission product inventory is bounded by the dose analysis.
 
14.7.1.6  Radiological Consequences
 
The Control Rod Drop Accident radiological consequences were analyzed using Alternative Source Term methodology as provided in Regulatory Guide 1.183 (Reference 129). The accident parameters and assumptions used in the analysis (References 13 and 136-140) are summarized below and in USAR Table 14.7-2a, and are in accordance with the guidance provided in RG 1.183.
 
The total radioactive material release is confirmed to be bounded by the dose analysis of record inputs. Since fuel rod failures had been identified, revised release fractions or total release fraction (TOTR) are determined using the Licensing Basis Release Fractions (LBRF) from RG 1.183 (Reference 129) conservatively applied as the steady state release fractions (SSRF) with the transient fission gas release (TFGR) as described in RG 1.236 (Reference 9). A ratio of the new TOTR to the LBRF used in the original licensing basis is then generated following the method provided in ANP-10333P-A (Reference 8).
 
The transient release terms, from Reference RG 1.236, expressed as a fraction are:
 
Peak Pellet BU < 50 GWd/MTU:
                                      =  0.26  13  0 100 Peak Pellet BU  50 GWd/MTU:
                                        = 0.26    5  0 100
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 36 of 239
 
The total fuel rod release fraction TOTR is dependent on the nuclide group and the enthalpy dependent TFGR average over the 24 nodes of fuel for a full-length fuel rod.
(If the failure were in a different length fuel rod, the number of axial nodes would be adjusted accordingly.)
 
Burnup < 50:
 
[0.26  13]
24
                                  =          100       
 
Burnup  50:
 
[0.26    5]
24
                                    =        100       
 
Where,
 
H fuel enthalpy increase (cal/g)
 
SSRF is the steady state release fraction
 
Group                        GMUL    Applied to Stable long-lived isotopes    1.0    Kr-85 Cs-134 and Cs-137            1.414  Alkali Metals Short lived radioactive      0.333  Iodines, nobles except Kr-85, isotopes                              halogens
 
This release is then used to determine the release ratio for each rod for each main isotope or class of isotope. The maximum release ratio is then used to scale up the total number of failed rods calculated by AURORA-B. As long as this total number of failed rods is less than the amount assumed in the dose analysis of record, the dose analysis of record remains bounding. Due to the high number of rods assumed to fail in the dose analysis of record (approximately 14 bundles, see USAR 14.7.1.6.2) compared to a typical scaled failure number of  1 bundle, the dose analysis of record typically remains bounding but is checked each cycle. Alternatively, the specific inventory release could be directly used in a revised dose analysis.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 37 of 239
 
14.7.1.6.1 Introduction
 
The postulated CRDA involves the rapid removal (drop) of the highest worth control rod resulting in a reactivity excursion. The CRDA reactivity excursion is terminated by the APRM high flux scram or by the IRMs during startup if the APRMs are not operable. Activity released from damaged fuel is transported to the main condenser and then released to the environment. The release is assumed to terminate after 24 hours.
 
Two cases were performed to model possible pathways for the main condenser release. The SJAE release case is the limiting case and models the release from the main condenser through the steam jet air ejectors (SJAEs) to the offgas stack with the offgas storage system bypassed. The isolated condenser release case assumes that the mechanical vacuum pump (MVP) is operating at the beginning of the accident and models the release as leakage from the isolated main condenser following MVP trip on high radiation in the main steam lines.
 
14.7.1.6.2 Source Term
 
The core inventory used for the CRDA analysis source term was calculated assuming operation at 2044 MWt (2004 MWt increased by 2% to account for power measurement uncertainties) and operation at the total average burnup expected for a 24-month fuel cycle. See USAR Section 14.7.8 for further discussion of the inventory development.
 
The core inventory available at accident time T=0 for release is shown in USAR Tables 14.7-24b, 14.7-24d, and 14.7-24e for ATRIUM 10XM and ATRIUM 11 fuel, in any combination; that may be in the core. GE14 fuel is not analyzed because all GE fuel is removed from the core.
 
The source term consists of releases from melted fuel and the gap activity from fuel pins with cladding damage. Fuel cladding damage is assumed to occur in 850 rods, with 9 of those experiencing fuel melt. This represents cladding damage in 2.9% of the core (484 fuel assemblies with 60 rods per equivalent 8x8 assembly). A radial peaking factor of 1.7 is assumed.
 
For the fuel with cladding damage, 10% of the rod inventory of noble gases and iodines are in the fuel gap and are released into the reactor coolant. For the melted fuel, 100% of the noble gases and 50% of the iodines are released to the reactor coolant. The iodine species released to the reactor coolant are assumed to be 95%
aerosol (CsI), 4.85% elemental, and 0.15% organic. Although not specified in RG 1.183, alkali metals (Cs and Rb) are assumed to be released with a release fraction of 0.12 for fuel with cladding damage and 0.25 for melted fuel. As discussed in Reference 8, this source term is bounding over the total source term calculated in the cycle-specific analysis. The activity is released into the reactor coolant at time zero of the accident and is assumed to mix instantaneously in the reactor coolant within the reactor vessel.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 38 of 239
 
Of the activity released from the reactor coolant within the pressure vessel, 100% of the noble gases, 10% of the iodine, and 1% of the remaining radionuclides are assumed to instantaneously reach the turbine and condensers. Of the activity that reaches the main condenser, 100% of the noble gases, 10% of the iodines, and 1%
of the remaining nuclides are available for release to the environment. The iodine species released from the main condenser to the environment are assumed to be 97% elemental and 3% organic.
 
14.7.1.6.3 Mitigation
 
The CRDA reactivity excursion is terminated by the APRM high flux scram or by the IRMs during startup if the APRMs are not operable.
 
For the activity released to the reactor coolant, no credit is assumed for partitioning in the reactor vessel or for removal by the steam separators.
 
For the SJAE release case, no credit for main condenser isolation is assumed and the release is modeled through operating SJAEs to the offgas stack elevated release point. Condenser air inleakage is assumed to occur at the maximum rate that can be processed by the SJAEs, resulting in minimum holdup time for the release. The offgas storage system is bypassed.
 
For the isolated condenser release case, the mechanical vacuum pump is operating and isolates on a high radiation signal from the main steam line radiation monitors.
An isolation time of 10 seconds is assumed, including MVP suction valve closure time and instrument response time for radiation detection and isolation initiation.
Following MVP isolation, the condenser is assumed to leak at 1% per day to the Turbine Building. No credit for dilution or holdup in the Turbine Building is assumed.
 
Finally, a hybrid case is run to account for the brief time that both the MVP and SJAE are in service during a plant startup. This combines the release characteristics of the MVP and SJAE cases described previously.
 
CR ventilation is assumed to remain in the normal operating mode throughout the event and no credit for emergency mode filtration or isolation is assumed.
 
No credit is taken for operator action.
 
14.7.1.6.4 Transport
 
The activity released from the reactor coolant is assumed to be instantaneously transported to the main condenser.
 
For the SJAE release case, the SJAEs are conservatively assumed to be operating at their maximum capacity with 360.5 cfm of condenser air inleakage. The SJAEs discharge to the recombiners and then through the air ejector holdup line to the offgas stack for an elevated release. A 17-minute holdup time is provided for the release, based on condenser air inleakage (SJAE flow from recombiners) and the holdup line volume. The SJAEs continue to operate at their maximum capacity until the release is terminated after 24 hours.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 39 of 239
 
For the isolated condenser release case, the MVP is assumed to be initially operating at its maximum flow rate of 2,300 cfm. The MVP discharges through the steam packing holdup line to the offgas stack for an elevated release. The MVP is isolated within 10 seconds of the accident release by a high radiation signal from the Main Steam Isolation Radiation Monitors. Following MVP isolation, the isolated condenser is assumed to leak at a rate of 1% per day providing a ground level release from the Turbine Building vent until the release is terminated after 24 hours.
 
The release characteristics of the SJAE and MVP cases above are used to analyze the hybrid case with both running.
 
CR ventilation remains in the normal mode throughout the accident, with 7,440 cfm of CR air intake assumed, representing the maximum normal CR air intake rate (i.e., no intake blanking plates installed and no recirculation of intake). An additional 1,000 cfm of unfiltered inleakage is assumed. CR dose studies were performed at several lower air intake and unfiltered inleakage flow rates, verifying that the maximum flow rates of 7,440 cfm and 1,000 cfm are limiting.
 
Control Room and offsite atmospheric dispersion coefficients (X/Q) are shown in USAR Table 14.7-2a.
 
14.7.1.6.5 Results
 
Control Room operator and offsite accident doses are shown in USAR Table 14.7-2b.
 
14.7.2  Loss-of-Coolant Accident
 
Accidents that could result in release of radioactive material directly into the primary containment are the result of postulated nuclear system pipe breaks inside the drywell.
All possibilities for pipe break sizes and locations have been investigated including the severance of small pipe lines, the main steam lines upstream and downstream of the flow restrictors, and the recirculation loop pipelines. Historically, the most severe nuclear system effects and the greatest release of radioactive material to the primary containment result from a complete circumferential break of one of the recirculation loop pipelines. This accident is established as the design basis loss of coolant accident (LOCA).
 
The LOCA is analyzed at EPU and EFW operating conditions in conjunction with the emergency core cooling system (ECCS) performance evaluation in accordance with 10 CFR 50.46 and Appendix K to 10 CFR 50.  (See Section 6.2 for further discussion of ECCS design and performance.)  This evaluation is hereinafter referred to as the ECCS-LOCA analysis. A complete spectrum of postulated break sizes and locations is considered in the evaluation of ECCS performance. The objective of the ECCS-LOCA analysis is to demonstrate conformance with the ECCS acceptance criteria of 10 CFR 50.46 for the most limiting break size, break location and single failure combination for the plant. The required documentation for demonstrating that this objective is met is given in References 10 and 208.
 
The Evaluation Model used for the break spectrum analysis of ATRIUM 10XM fuel is the EXEM BWR-2000 LOCA analysis methodology described in Reference 218. The Evaluation Model used for the break spectrum analysis of ATRIUM 11 fuel is AURORA-B LOCA, described in Reference 3.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 40 of 239
 
14.7.2.1  Description of Design Basis LOCA
 
The LOCA is described in the Code of Federal Regulations 10 CFR 50.46 as a hypothetical accident that results in a loss of reactor coolant from breaks in reactor coolant pressure boundary piping up to and including a break equivalent in size to a double-ended rupture of the largest pipe in the reactor coolant system. There is not a specifically identified cause that results in the pipe break. However, for the purpose of identifying a design basis accident, the pipe break is postulated to occur inside the primary containment before the first isolation valve.
 
For a boiling water reactor (BWR), a LOCA may occur over a wide spectrum of break locations and sizes. Responses to the break vary significantly over the break spectrum. The largest possible break is a double-ended rupture of a recirculation pipe; however, this is not necessarily the most severe challenge to the emergency core cooling system (ECCS). A double-ended rupture of a main steam line causes the most rapid primary system depressurization, but because of other phenomena, steam line breaks are seldom limiting with respect to the event acceptance criteria (10 CFR 50.46). In addition to break location dependence, different break sizes in the same pipe produce quite different event responses, and the largest break area is not necessarily the most severe challenge to the event acceptance criteria. Because of these complexities, an analysis covering the full range of break sizes and locations is performed to identify the limiting break characteristics. This analysis is typically known as a break spectrum analysis.
 
Regardless of the initiating break characteristics, the event response is conveniently separated into three phases: the blowdown phase, the refill phase, and the reflood phase. The relative duration of each phase is strongly dependent upon the break size and location. The last two phases are often combined and discussed together.
 
During the blowdown phase of a LOCA, there is a net loss of coolant inventory, an increase in fuel cladding temperature due to core flow degradation, and for the larger breaks, the core becomes fully or partially uncovered. There is a rapid decrease in pressure during the blowdown phase. During the early phase of the depressurization, the exiting coolant provides core cooling. Later in the blowdown, core cooling is provided by lower plenum flashing as the system continues to depressurize and the injection of ECCS flows. The blowdown phase is defined to end when the system reaches the pressure corresponding to the rated Low Pressure Core Spray (LPCS) flow.
 
In the refill phase of a LOCA, the ECCS is functioning and there is a net increase of coolant inventory. During this phase the core sprays provide core cooling and, along with low-pressure and high-pressure coolant injection (LPCI and HPCI), supply liquid to refill the lower portion of the reactor vessel. In general, the core heat transfer to the coolant is less than the fuel decay heat rate and the fuel cladding temperature continues to increase during the refill phase.
 
In the reflood phase, the coolant inventory has increased to the point where the mixture level reenters the core region. During the core reflood phase, cooling is provided above the mixture level by entrained reflood liquid and below the mixture level by pool boiling. Sufficient coolant eventually reaches the core hot node and the fuel cladding temperature decreases.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 41 of 239
 
Indication of flooding to 2/3 core height for a DBA LOCA is expected by about 300 seconds after the initiation of the accident. At this point operator actions can be initiated for the transition to long term core and containment cooling described in Section 14.7.2.8.6 below (References 162 and 164).
 
14.7.2.2  ECCS-LOCA Analysis Assumptions
 
ECCS-LOCA licensing analyses are required to incorporate several limiting assumptions. This is to ensure that the ECCS design is capable of mitigating all postulated LOCA event scenarios. The required assumptions are:
: a. A break occurs in any steam or liquid line which forms part of the primary reactor coolant pressure boundary. (10 CFR 50, Appendix K)
: b. Coincident with the LOCA, offsite power may become unavailable.
Consequently, the limiting condition, either availability or unavailability of offsite power, must be evaluated. (10 CFR 50, Appendix A, General Design Criteria 35)
: c. A single component within the ECCS network fails coincident with the LOCA.
(10 CFR 50, Appendix K)
: d. The Reactor Core Isolation Cooling (RCIC) system is unavailable.
 
14.7.2.3  Effects of Unavailability of Offsite Power
 
The primary effect of the assumption that offsite power becomes unavailable coincident with the LOCA is an increase in the time delay for injection by the low pressure ECC systems. This occurs because the ECC systems must then wait for the emergency power supplied by the diesel generators. This unavailability of offsite power assumption is also implemented in other aspects of the ECCS-LOCA analysis as discussed below.
 
The unavailability of offsite power causes a trip of the reactor recirculation pumps at the beginning of the event. This causes both pumps to begin coasting down. A time constant of 5 secs (the minimum value derived from the rotational inertia time constant of the recirculation pump and motor-generator unit) is assumed for the coastdown. For the broken recirculation loop, the pump coastdown time constant is immaterial because the flow coastdown is dominated by the break flow dynamics (i.e., the break causes a rapid flow reversal in the broken loop which effectively results in an extremely rapid coastdown of the flow).
 
The feedwater pumps are also assumed to trip at the beginning of the event. The feedwater pumps are conservatively assumed to linearly coastdown from the initial value to zero in 5 secs.
 
Since the Reactor Protection System (RPS) is fail safe, the unavailability of offsite power will initiate a scram at the beginning of the event. Instead of crediting this beginning of event scram, the scram is conservatively initiated on low-low water level.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 42 of 239
 
14.7.2.4  Loss of Coolant Accident Analysis
 
Framatome Loss-of-Coolant Accident (LOCA) analysis, models, evaluation, and results are for a full core of ATRIUM 10XM or ATRIUM 11 fuel. The basis for applicability of PCT results from full cores of ATRIUM 10XM fuel and fuel cores of ATRIUM 11 fuel for a mixed (transition) core is provided in Reference 7.
Thermal-hydraulic characteristics of the ATRIUM 11 and ATRIUM 10XM fuel designs are similar as presented in Reference 7. Therefore, the core response during a LOCA will not be significantly different for a full core of ATRIUM 10XM fuel or a mixed core of ATRIUM 10XM and ATRIUM 11 fuel. In addition, since about 95% of the reactor system volume is outside the core region, slight changes in core volume and fluid energy due to fuel design differences will produce an insignificant change in total system volume and energy. Therefore, the existing ATRIUM 10XM LOCA analysis and resulting licensing PCT and MAPLHGR limits remain applicable for ATRIUM 10XM fuel in transition cores and current Framatome analysis and resulting licensing PCT and MAPLHGR limits are applicable for ATRIUM 11 fuel in transition cores and beyond.
 
Plant specific ECCS analyses provide peak cladding temperature (PCT) and maximum local metal-water reaction (MWR) values and establish MAPLHGR limits for each fuel design. For the limiting single failure and limiting break, calculations are performed to determine the PCT and MWR values over the expected exposure lifetime of the fuel when operating at the MAPLHGR limit. The limiting break is determined by evaluating a spectrum of potential break locations, sizes, and single failures.
 
The limiting single failure of ECCS equipment is that failure which results in the minimum margin to the PCT criterion. From a list of potentially limiting ECCS single failures, potentially limiting failures are analyzed and the worst single failure for each fuel type is identified.
 
Evaluations and analyses to establish the location of the limiting break are performed.
Analyses are performed for breaks on the suction and discharge sides of the recirculation pump. Non-recirculation line breaks are also evaluated but are generally non-limiting. The determination of the limiting location is based on minimum margin to the PCT criterion calculated for consistent fuel exposure conditions at each of the break locations. The MWR criterion is typically not challenged if the PCT limit is met and is normally reported for the highest PCT case. Analyses to establish the characteristics of the limiting break are performed. Hypothetical split and guillotine piping system breaks are evaluated up to and including those with a break area equal to the cross-sectional area of the largest pipe in the recirculation system piping. As with the location spectrum, the determination of the limiting break size is based on the minimum margin to the PCT criterion.
 
The condition of the fuel during the LOCA analysis is conservatively based on exposure conditions which assure that the highest value of fuel stored energy is used.
The condition of the fuel is based on fuel conditions associated with planar average exposure. The ATRIUM 10XM Appendix K LOCA methodology is referred to as the EXEM BWR-2000 Evaluation Model (Reference 218). The ATRIUM 11 Appendix K LOCA methodology is the AURORA      -B LOCA Evaluation Model (Reference 3). A summary of the Monticello initial condition is provided in Table 14.7-5.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 43 of 239
 
14.7.2.5  Effect of Fuel Densification
 
Power spiking due to in-reactor fuel densification has not been explicitly considered in LOCA calculations. Framatome incorporates fuel densification and swelling effects into its RODEX computer codes, as described in References 219, 220, and 221.
 
14.7.2.6  Analytical Methods
 
The Evaluation Model used for the ATRIUM 10XM break spectrum analysis is the EXEM BWR-2000 LOCA analysis methodology described in Reference 218. The EXEM BWR-2000 methodology employs three major computer codes to evaluate the system and fuel response during all phases of a LOCA. These are the RELAX, HUXY, and RODEX2 computer codes. RELAX is used to calculate the system and hot channel response during the blowdown, refill and reflood phases of the LOCA.
The HUXY code is used to perform heatup calculations for the entire LOCA, and calculates the PCT and local clad oxidation at the axial plane of interest. RODEX2 is used to determine fuel parameters (such as stored energy) for input to the other LOCA codes. The code interfaces for the LOCA methodology are illustrated in Figure 14.7-14 of Reference 218.
 
A complete ATRIUM 10XM analysis for a given break size starts with the specification of fuel parameters using RODEX2 (Reference  219). RODEX2 is used to determine the initial stored energy for both the blowdown analysis (RELAX hot channel) and the heatup analysis (HUXY). This is accomplished by ensuring that the initial stored energy in RELAX and HUXY is the same or higher than that calculated by RODEX2 for the power, exposure, and fuel des  ign being considered.
 
ATRIUM 11 fuel is analyzed using the AURORA  -B LOCA Evaluation Model described in Reference 3. RODEX4 (Reference 221) is used to determine the initial stored energy for both blowdown and heatu p analysis and provide fuel properties.
 
14.7.2.6.1  Blowdown Analysis
 
The RELAX code (Reference 218) is used to calculate the system thermal-hydraulic response during the blowdown phase of the LOCA  for ATRIUM 10XM. For the system blowdown analysis, the core is represented by an average core channel. The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and with decay heating as required by Appendix K of 10 CFR 50. The nodalization in the ATRIUM 10XM analysis is consistent with that used in the topical report submitted to the NRC (Reference 218).
 
The RELAX blowdown analysis is performed from the time of the break initiation through the end of blowdown (EOB). The system blowdown calculation provides the upper and lower plenum transient boundary conditions for the hot channel analysis.
 
Following the system blowdown calculation, another RELAX analysis is performed to analyze the maximum p ower assembly (hot channel) of the core. The RELAX hot channel blowdown calculation determines hot channel fuel, cladding, and coolant temperatures during the blowdown phase of the LOCA.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 44 of 239
 
The hot channel analysis is performed using the system blowdown results to supply the core power and the system boundary conditions at the core inlet and exit. The initial average fuel rod temperature at the limiting plane of the hot channel is conservative relative to the average fuel rod temperature calculated by RODEX2 for operation of the ATRIUM 10XM assembly at the MAPLHGR limit. The heat transfer coefficients and fluid conditions at the limiting plane of the RELAX hot channel calculation are used as input to the HUXY heatup analysis.
 
For ATRIUM 11 fuel, S-RELAP5 is used to calculate the thermal-hydraulic response during all phases of the LOCA. The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and decay heat as required by Appendix K of 10 CFR 50. The clad swelling and rupture models from NUREG-0630 have been incorporated into S-RELAP5. The nodalization in the ATRIUM 11 analysis is consistent with that used in the topical report submitted to the NRC (Reference 3), which is incorporated into the USAR by reference.
 
14.7.2.6.2  Refill/Reflood Analysis
 
The RELAX code is also used to compute the system and hot channel hydraulic response during the refill/reflood phase of the LOCA for ATRIUM 10XM. The RELAX system and RELAX hot channel analyses continue beyond the end of blowdown to analyze system and hot channel responses during the refill and reflood phases.
 
The refill phase is the period when the lower plenum is filling due to ECCS injection. The reflood phase is the period when some portions of the core and hot assembly are being cooled with ECCS water entering from the lower plenum. The purpose of the RELAX calculations beyond blowdown is to determine the time when the liquid flow via upward entrainment from the bottom of the core becomes high enough at the hot node in the hot assembly to end the temperature increase of the fuel rod cladding. This event time is called the time of hot node reflood. The time when the core bypass mixture level rises to the elevation of the hot node in the hot assembly is also determined.
 
RELAX provides a prediction of fluid inventory during the ECCS injection period.
Allowing for countercurrent flow through the core and bypass, RELAX determines the refill rate of the lower plenum due to ECCS water and the subsequent reflood times for the core, hot assembly, and the core bypass. The RELAX calculations provide HUXY with the time of hot node reflood and the time when the liquid has risen in the bypass to the height of the axial plane of interest (time of bypass reflood).
 
For ATRIUM 11 fuel, S-RELAP5 is used to calculate the thermal-hydraulic response during all phases of the LOCA.
 
14.7.2.6.3  Heatup Analysis
 
The HUXY code (Reference 222) is used to perform heatup calculations for ATRIUM 10XM for the entire LOCA transient and provides PCT and local clad oxidation at the axial plane of interest. The heat generated by metal-water reaction (MWR) is included in the HUXY analysis. HUXY is used to calculate the thermal response of each fuel rod in one axial plane of the hot channel assembly. These
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 45 of 239
 
calculations consider thermal-mechanical interactions within the fuel rod. The clad swelling and rupture models from NUREG-0630 have been incorporated into HUXY (Reference 224). The HUXY code complies with the 10 CFR 50 Appendix K criteria for LOCA Evaluation Models.
 
HUXY uses the EOB time and the times of core bypass reflood and core reflood at the axial plane of interest from the RELAX analysis. Until the EOB, HUXY uses RELAX hot channel heat transfer coefficients, fluid temperatures, fluid qualities, and power. Throughout the calculations, decay power is determined based on the ANS 1971 decay heat curve plus 20% as described in Reference 218. After the EOB and prior to the time of hot node reflood, HUXY uses Appendix K spray heat transfer coefficients for the fuel rods, water channel and fuel channel.
Experimental data for AREVA 10X10 fuel which supports the use of the convective heat transfer coefficients listed in Appendix K is documented in Reference 224.
After the time of hot node reflood, Appendix K reflood heat transfer coefficients are used in the HUXY analysis. The principal results of a HUXY heatup analysis are the PCT and the percent local oxidation of the fuel cladding, often called the
            %MWR.
 
For ATRIUM 11 fuel, S-RELAP5 is used to calculate the thermal -hydraulic response during all phases of the LOCA. The principle results of the S-RELAP5 analysis are PCT and maximum local oxidation (%MWR).
 
14.7.2.6.4  Break Spectrum Analysis
 
The results of a loss-of-coolant accident (LOCA) break spectrum analysis at extended power uprate (EPU) conditions for Monticello have been determined using the EXEM BWR-2000 and AURORA-B LOCA methods. The purpose of the break spectrum analysis is to identify the parameters that result in the highest calculated peak cladding temperature (PCT) during a postulated LOCA. The LOCA parameters determined include the following:
 
a)  Break location
 
b)  Break type (double-ended guillotine (DEG) or split)
 
c)  Break size
 
d)  Limiting emergency core cooling system (ECCS) single failure
 
e)  Axial power shape (top- or mid-peaked)
 
The analyses are performed with LOCA Evaluation Models developed by Framatome and approved for reactor licensing analyses by the U.S. Nuclear Regulatory Commission (NRC). See Section 14.7.2.7 for additional details. The calculations are performed in conformance with 10 CFR 50 Appendix K requirements and satisfy the event acceptance criteria identified in 10 CFR 50.46.
 
The break spectrum analyses were performed for a core composed entirely of ATRIUM'    10XM fuel at beginning-of-life (BOL) conditions. Calculations assumed an initial core power of 102% of 2004 MWt, providing a licensing basis power of 2044.08 MWt. The 2.0% increase reflects the maximum uncertainty in monitoring reactor power, as per NRC requirements. The limiting assembly in the core was
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 46 of 239
 
assumed to be at a maximum average planar linear heat generation rate (MAPLHGR) limit of 12.5 kW/ft. Other initial conditions used in the analyses included core flow, recirculation loop flow, steam flow, steam dome pressure, core inlet enthalpy, and maximum hot assembly MCPR, as well as ATRIUM 10XM fuel assembly parameters (Reference 208).
 
The results of the ATRIUM 10XM LOCA break spectrum analysis performed in support of the fuel transition at Monticello are presented in Reference 208. Break spectrum calculations were performed at several core flow and core power statepoints to ensure that limiting results were determined. Analyzed power levels were 2% higher than actual statepoint values to ensure that 10 CFR 50 Appendix K requirements were satisfied.
 
The break spectrum analyses were performed for a core composed entirely of ATRIUM'    11 fuel at beginning-of-life (BOL) conditions. Calculations assumed an initial core power of 102% of 2004 MWt, providing a licensing basis power of 2044.1 MWt. The 2.0% increase reflects the maximum uncertainty in monitoring reactor power per Appendix K requirements. The limiting assembly in the core was assumed to be at a maximum average planar linear heat generation rate (MAPLHGR) limit of 10 kW/ft. Other initial conditions used in the analyses included core flow, recirculation loop flow, steam flow, steam dome pressure, core inlet enthalpy, and maximum hot assembly MCPR, as well as ATRIUM 11 fuel assembly parameters (Reference 10).
 
The results of the ATRIUM 11 LOCA break spectrum analysis performed in support of the fuel transition at Monticello are presented in Reference 10. Break spectrum calculations were performed at several core flow and core power statepoints to ensure that limiting results were determined. Analyzed power levels were 2% higher than actual statepoint values to ensure that 10 CFR 50 Appendix K requirements were satisfied.
 
The objective of break spectrum analyses is to ensure that the limiting break location, break type, break size, and ECCS single failure are identified. The LOCA response scenario varies considerably over the spectrum of break locations.
Potential break locations are separated into two groups: recirculation line breaks and non-recirculation line breaks. The basis for the break locations and potentially limiting single failures is described in the following sections.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 47 of 239
 
14.7.2.6.4.1  Limiting Single Failure
 
Regulatory requirements specify that the LOCA analysis be performed assuming that all offsite power supplies are lost instantaneously and that only safety grade systems and components are available. In addition, regulatory requirements also specify that the most limiting single failure of ECCS equipment must be assumed in the LOCA analysis. The term "most limiting" refers to the ECCS equipment failure that produces the greatest challenge to event acceptance criteria. The limiting single failure can be a common power supply, an injection valve, a system pump, or system initiation logic. The most limiting single failure may vary with break size and location. The potential limiting single failures are shown below:
 
a)  DC power (SF-BATT)
 
b)  Diesel generator (SF-DGEN)
 
c)  LPCI injection valve (SF-LPCI)
 
d)  High-pressure coolant injection system (SF-HPCI)
 
e)  ADS valve (SF-ADS)
 
All three ADS valves are assumed operable in the analyses that do not consider SF-ADS. The single failures and the available ECCS for each failure assumed in these analyses are consistent with Table 14.7-11. Other potential failures are not specifically considered because they result in as much or more ECCS capacity.
 
A comparison of the systems remaining for each of the assumed failures shows that the diesel generator failure (SF-DGEN) always has more available ECCS capacity than the battery failure (SF-BATT). Therefore, PCT results obtained for SF-BATT bound those obtained for SF-DGEN. Similarly, a comparison of the systems remaining for HPCI failure (SF-HPCI) shows that it also always provides more available ECCS capacity than the SF-BATT. Therefore, PCT results obtained for SF-BATT bound those obtained for SF-HPCI. As a result, results reported in References 10 and 208 are those considering SF-BATT, SF-LPCI, and SF-ADS.
 
14.7.2.6.4.2  Recirculation Line Breaks
 
The response during a recirculation line LOCA is dependent on break size. The rate of reactor vessel depressurization decreases as the break size decreases.
The high pressure ECCS and ADS will assist in reducing the reactor vessel pressure to the pressure where the LPCI and LPCS flows start. For large breaks, rated LPCS and LPCI flow is generally reached before or shortly after the time when the ADS valves open so the ADS system is not required to mitigate the LOCA. ADS operation is an important emergency system for small breaks where it assists in depressurizing the reactor system faster, and thereby reduces the time required to reach rated LPCS and LPCI flow.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 48 of 239
 
The two largest flow resistances in the recirculation piping are the recirculation pump and the jet pump nozzle. For breaks in the discharge piping (PD), there is a major flow resistance in both flow paths from the reactor vessel to the break.
For breaks in the suction piping (PS), the major flow resistances are in the same flow path from the vessel to the break. As a result, pump suction side breaks experience a more rapid blowdown, which tends to make the event more severe. For suction side breaks, the recirculation discharge isolation valve on the broken loop closes which allows the LPCI flow to fill the discharge piping and supply flow to the lower plenum and core. For discharge side breaks with 2 break areas  0.4 ft, the LPCI Selection Logic directs all available LPCI flow to 2 the intact loop. No LPCI flow is credited for breaks < 0.4 ft  in the EXEM BWR-2000 methods (Reference 218). In the AURORA-B LOCA methodology, LPCI flow into the broken loop can be credited below the minimum detectable break size (References 3 & 5). Both suction and discharge recirculation pipe breaks are considered in the break spectrum analysis.
 
Two break types (geometries) are considered for the recirculation line break.
The two types are the double-ended guillotine (DEG) break and the split break.
 
For a DEG break, the piping is assumed to be completely severed resulting in two independent flow paths to the containment. The DEG break is modeled by setting the break area (at both ends of the pipe) equal to the full pipe cross-sectional area and varying the discharge coefficient between 1.0 and 0.4.
The range of discharge coefficients is used to cover uncertainty in the actual geometry at the break. Discharge coefficients below 0.4 are unrealistic and not considered in the EXEM BWR-2000 or AURORA-B LOCA methods. The most limiting DEG break is determined by varying the discharge coefficient. The labeling convention for guillotine breaks is to list the discharge coefficient before DEG. For example, a guillotine break with a discharge coefficient of 0.8 on the suction side of the recirculation pump would be labeled as 0.8DEGPS.
 
A split type break is assumed to be a longitudinal opening or hole in the piping that results in a single break flow path to the containment. Appendix K of 10 CFR 50 defines the cross-sectional area of the piping as the maximum split break area required for analysis. The labeling convention for split breaks is to list the flow area using the letter P instead of a period. For example, a split 2 break with a flow area of 3.5 ft on the suction side of the recirculation pump would be labeled as 3P5FT2PS. These labeling conventions for double-ended guillotine and split breaks are typically used in figures such as those in Section 7.0 of Reference 208. Break types, break sizes and single failures are analyzed for both suction and discharge recirculation line breaks.
 
References 10 and 208 provide a description and results summary for breaks in the recirculation line.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 49 of 239
 
14.7.2.6.4.3    Non-Recirculation Line Breaks
 
In addition to breaks in the recirculation line, breaks in other reactor coolant system piping must be considered in the LOCA break spectrum analysis.
Although the recirculation line large breaks result in the largest coolant inventory loss, they do not necessarily result in the most severe challenge to event acceptance criteria. The double-ended rupture of a main steam line is expected to result in the fastest depressurization of the reactor vessel. Special consideration is required when the postulated break occurs in ECCS piping.
Although ECCS piping breaks are small relative to a recirculation pipe DEG break, the potential to disable an ECCS system increases their severity.
 
Non-recirculation line breaks outside of the containment are inherently less challenging to fuel limits than breaks inside the containment. For breaks outside containment, isolation or check valve closure will terminate break flow prior to the loss of significant liquid inventory and the core will remain covered. If high-pressure coolant inventory makeup cannot be reestablished, ADS actuation may become necessary. Although analyses of breaks outside containment may be required to address non-fuel related regulatory requirements, these breaks are not limiting relative to fuel acceptance criteria such as PCT. For the Monticello break spectrum analysis, breaks in the main steam lines, feedwater lines, HPCI lines, LPCS lines, LPCI lines, RWCU lines, shutdown cooling lines, and instrument lines were all considered and found to be non-limiting.
 
14.7.2.6.4.4    Long Term Cooling
 
Framatome continues to credit the GEH evaluation of long term core cooling capability as described in 14.7.2.8.6.
 
14.7.2.7  Summary of ECCS-LOCA Results
 
The most limiting Peak Cladding Temperature (PCT) for ATRIUM 10XM fuel during Two Loop Operation (TLO) is 2102 degrees F in the break spectrum resulting in a final temperature of 22125 degrees F in the MAPLHGR analysis and occurs for a 0.25 ft split break in the recirculation discharge line with a top-peaked axial power shape and a failure of the battery bus (SF-BATT). For Single Loop Operation (SLO),
the most limiting PCT is 1862 degrees F and occurs for a double ended guillotine break in the recirculation suction line with a top-peaked axial power shape and a failure of LPCI injection (SF-LPCI). These are Appendix K results and are licensing PCT values. A MAPLHGR multiplier of 0.70 is established for SLO since LOCA is more severe when initiated during SLO. ECCS system flow rates, initiation times, and pressure ranges for the AREVA analysis are consistent with values shown in Tables 14.7-7 through 14.7-12 and Figures 14.7-8 through 14.7-12, as documented in Reference 208. Note that cycle-specific MAPLHGR limit determinations for various fuel designs can result in lowering the MAPLHGR limit be low the 12.5 kw/ft bounding value used in the ATRIUM 10XM ECCS-LOCA analysis (Reference 208).
 
Taking into account 50.46 notifications from Framatome, the current estimated licensing PCT for ATRIUM 10XM is 2144&deg;F (Reference 19).
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 50 of 239
 
The most limiting Peak Cladding Temperature (PCT) and maximum local oxidation for ATRIUM 11 fuel during Two Loop Operation (TLO) is 2120&deg;F in the MAPLHGR analysis and occurs for a 1.0 double ended guillotine break in the recirculation suction line with a top-peaked axial power shape and a failure of LPCI injection (SF-LPCI).
For Single Loop Operation (SLO), the most limiting break spectrum PCT is 1968&deg; F and occurs for a double ended guillotine break in the recirculation suction line with a top-peaked axial power shape and a failure of LPCI injection (SF-LPCI). These are Appendix K results and are licensing PCT values. A MAPLHGR multiplier of 0.80 is established for SLO since LOCA is more severe when initiated during SLO. ECCS system flow rates, initiation times, and pressure ranges for the AREVA analysis are consistent with values shown in Table 14.7-7 through Table 14.7-12 and Figures 14.7-8 through 14.7-12, as documented in Reference 10. Taking into account 50.46 notifications from Framatome, the current estimated licensing PCT for ATRIUM 11 is 2120&deg;F (Reference 19).
 
14.7.2.8  Emergency Core Cooling System Performance
 
14.7.2.8.1 ECC System Descriptions
 
The ECCS network consists of a High Pressure Coolant Injection (HPCI) System, the Core Spray (CS) System, the Low Pressure Coolant Injection (LPCI) mode of the Residual Heat Removal (RHR) System, and the Automatic Depressurization System (ADS).
 
Monticello is also equipped with a Reactor Core Isolation Cooling (RCIC) System, which is an alternative source of make-up water for the reactor. It is designed to provide adequate makeup to the reactor during normal plant shutdowns and transient events which lead to a loss of feedwater flow. The RCIC System is not part of the ECCS network.
 
14.7.2.8.1.1 Emergency Diesel Generators
 
The emergency diesel generators (EDGs) provide an alternative source of AC power in the event that the multiple redundant offsite power supplies are lost. They provide power to the emergency busses and must achieve rated operating conditions in a few seconds. The EDGs are designed to startup, achieve full speed, and be loaded within 10 secs. In order to accommodate future improvements to the EDGs, which may result in slowing the rapid start response, the EDG startup time to rated speed and voltage assumed in the ECCS-LOCA analysis was 15 secs.
 
The design basis, system description and performance analysis for the EDGs can be found in Section 8.4.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 51 of 239
 
14.7.2.8.1.2 High Pressure Coolant Injection (HPCI) System
 
The HPCI System is designed to provide rated flow over a vessel pressure range of 1120 to 150 psig. The HPCI System is capable of delivering (per Table 14.7-7) a minimum of 2700  gpm of coolant to the vessel. The system is initiated on either low-low reactor water level (Level 2) or high drywell pressure. The HPCI System is powered by reactor steam, and its control requirements and moto r operated valves needed for startup and operation are supplied by DC power. Consequently, it is independent of the emergency diesel generators. The HPCI System is normally aligned to take suction from the condensate storage tank and will automatically transfer to the suppression pool as an inventory supply, if necessary.
 
It should be noted that HPCI does not have a significant effect on the overall ECCS performance for large breaks. Large breaks depressurize the vessel before the steam-powered HPCI Syst    em has sufficient time to startup and inject a significant amount of coolant into the reactor vessel. The actual core cooling contribution of the HPCI System for most recirculation line breaks is also small. This is because HPCI injects coolant through the feedwater sparger into the downcomer region of the vessel. Injection at this location allows the coolant flow to be diverted from the core region out the postulated recirculation line break.
 
For small recirculation line breaks, the Monticello ECCS-LOCA analyses (References 10 and 208) takes no credit for HPCI operation because the limiting failure (battery) prevents the HPCI System from initiating.
 
The key HPCI system analysis parameters are shown in    Table 14.7-7. ECCS performance concerns no longer have a significant bearing on the se parameters, since little credit was received for the HPCI System in the Monticello ECCS-LOCA analyses. Consequently, several HPCI System performance assumptions were relaxed in the Monticello ECCS -LOCA analysis. These relaxations included reducing the HPCI flow rate from 3000 to 2700 gpm, increasing the overall system startup time from 30 to 45 secs, and lowering the low-low water level initiation point to the instrument tap elevation. By relaxing the ECCS requirements and constraints imposed on the HPCI System, the plant has the flexibility to make modifications to improve the system performance and/or reliability of the HPCI System in a manner similar to that for the Emergency Diesel Generators.
 
14.7.2.8.1.3 Core Spray (CS) System
 
The Core Spray (CS) system is designed to restore and maintain the coolant in the reactor vessel in combination with other emergency core cooling systems such that the core is adequately cooled to preclude fuel damage. See Section 6.2.2 for a complete discussion of the Core Spray Syst em design basis, system description, and performance evaluation.
 
The key CS parameters used in the ECCS    -LOCA analyses (References 10 and 208) are provided in Table 14.7-8. Some of these parameters have been modified and represent a relaxation in comparison to both the system design and the parameters assumed in the original ECCS-LOCA analysis. Analyses show that the acceptance criteria of the ECCS-LOCA analysis are still met when the CS loop flow rate assumption is as shown in Table 14.7-8 to account for EDG frequency uncertainty, the impact on pump discharge from NPSHr, and postulated maximum wear ring degradation (Reference 193). The ECCS-LOCA values reflect the CS
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 52 of 239
 
flow rates which are assumed to actually inject inside the core shroud. The limiting CS flow delivery curve (CS flow inside the core shroud versus reactor vessel-to-torus differential pressure) is shown in Figure 14.7-8. This delivery curve represents a quadratic fit obtained from the assumed delivery flow rate of 2700 gpm at 130 psid, pump shutoff head of 279 psid and a delivery flow of 3700 gpm at 0 psid (Reference 10 and 208).
 
The margins between the 2835 gpm Technical Specification flow requirement and the 2672 gpm and 2700 gpm ECCS-LOCA analysis and supplemental analysis values, respectively, are intended to account for expected pressure boundary leakage between the Core Spray loop and the core shroud. The areas of expected leakage result in flow being diverted from the Core Spray piping to the downcomer region, outside the shroud. This flow is lost out a recirculation line break.
Examples of expected leakage are the Core Spray T    -box vent hole (7.9 gpm),
leakage at holes machined for the T-box clamp fixture (25.3 gpm), and leakage associated with P5 and P6 reactor vessel welds. In addition, excess Core Spray flow is used to offset a small shortfall in the total amount of LPCI flow (50 gpm)
(References 111, 169, 170 and 171) (See Section 14.7.2.8.4).
 
For existing crack indications observed in the piping between the vessel and the core shroud, the worst case leakage has been determined with conservative assumptions for through wall conditions and crack growth over time. The predicted leakage and the ECCS-LOCA analyses flow requirements are totaled and the resulting flow rate is verified to be within the rated capability of the associated Core Spray pump (References 10, 111, 169, 170, 171, 193 and 208).
 
The CS flow rates described above are required to meet the requirements of 10 CFR 50.46 prior to the point in time where long term core cooling is credited.
See discussion in USAR section 14.7.2.8.6 for long term core cooling requirements.
 
14.7.2.8.1.4 Low Pressure Coolant Injection (LPCI) System
 
The LPCI is an operating mode of the multiple-purpose Residual Heat Removal (RHR) System. A portion of the LPCI initiation logic, known as the LPCI Loop Selection Logic, selects which of the two recirculation loops will receive the LPCI flow. This selection is based on comparing the pressure in the two loops to determine which one is broken. LPCI flow is then directed to the unbroken recirculation loop. The LPCI Loop Selection Logic is designed to correctly select 2 the unbroken loop for break areas of 0.4 ft in the recirculation line.
(Reference 143, 144, 145)  For smaller recirculation line breaks or non-recirculation line breaks, the LPCI Loop Selection Logic directs the LPCI flow to a predetermined default recirculation loop.
 
Another portion of the LPCI Loop Selection logic includes signaling the closure of the isolation valve on the discharge side of one of the reactor recirculation pumps.
The closure of this valve directs the LPCI flow upward through the recirculation piping and into the vessel through the jet pump drive lines. The flow can then pass directly through the jet pumps into the lower plenum region of the vessel, ensuring an efficient inventory delivery to the lower plenum.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 53 of 239
 
The RHR System is divided into two loops consisting of separate piping, pumps, and valves. Each RHR loop has two RHR pumps, and each loop is supplied by separate power sources under both normal and emergency power conditions.
Flow from both loops is directed to a single injection point, located in one of the two recirculation loops, by a normally open intertie line. When operating in the LPCI mode, each RHR pump takes suction from the suppression pool. As with the CS System, the LPCI function is designed to inject coolant when the reactor pressure is relatively low. Injection at pressures above the design value of the system is prevented by a low pressure permissive on the LPCI injection valves (the pressure permissive is intended to prevent overpressurization of the LPCI piping network).
The CS and LPCI Systems provide Monticello with two completely independent and diverse sources of low pressure coolant makeup flow. Consistent with this philosophy, the LPCI System is initiated by the same redundant LOCA signals as those used to initiate the CS System.
 
The key LPCI parameters used in the ECCS-LOCA analyses (Reference 10, 193 and 208) are provided in Table 14.7-9. Some of these parameters have been modified and represent a relaxation in comparison to both the system design and the parameters assumed in the original ECCS-LOCA analysis .
 
The LPCI flow rate entering the vessel is dependent upon the number of pumps which are providing flow through the injection line. The two-pump flow rate was assumed to be as shown in Table 14.7-9 which reflect adjustments made to account for EDG frequency uncertainty, the impact on pump discharge head from NPSHr and postulated maximum wear ring degradation in excess of 10%, still satisfy the acceptance criteria of the ECCS-LOCA analysis (Reference 193). The ECCS-LOCA and supplemental analyses values reflect the LPCI flow rates which are injected into the recirculation loop with a 20 psid differential pressure between the reactor vessel and torus (Reference 193).  (Note:  See Section 14.7.2.8.4 for further discussion on expected system performance.)  The difference between the pumped flow and that which reaches the vessel lower plenum is due to leakage from joints on the jet pump assemblies. Consequently, a conservative leakage allowance is taken to account for this effect (see section 14.7.2.8.4). The limiting LPCI flow delivery curves (LPCI flow into the core versus vessel pressure) for two-pump, three-pump and four-pump operation are based on the supplemental analysis (Reference 193) and are shown in Figure 14.7-9. These delivery curves were obtained by using a quadratic fit from the delivery flow rates determined by the supplemental analysis (Reference 193) for two, three, and four pump operation. These flow rates are listed on Table 14.7-9.
 
The LPCI mode of RHR is credited until long term core cooling as defined in Section 14.7.2.8.6 is satisfied. Once the core is recovered above top of active fuel (TAF) or reflooded to an indicated level of 2/3 core height, RHR is placed into a containment cooling mode of operation. Acceptable methods for containment cooling include use of containment spray, suppression pool cooling or LPCI injection cooling (Reference 155).
 
14.7.2.8.1.5 Automatic Depressurization System (ADS)
 
As indicated in Section 6.2.5, the ADS uses three of the safety relief valves (SRVs) to depressurize the reactor. The pertinent ADS parameters used in the ECCS-LOCA analyses (References 10 and 208) are provided in Table 14.7-10 and ADS initiation logic is shown in Figure 14.7-10.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 54 of 239
 
14.7.2.8.2 Single Failure Considerations
 
Regulatory requirements specify that the LOCA analysis be performed assuming that all offsite power supplies are lost instantaneously and that only safety grade systems and components are available. In addition, regulatory requirements also specify that the most limiting single failure of ECCS equipment must be assumed in the LOCA analysis. The term "most limiting" refers to the ECCS equipment failure that produces the greatest challenge to event acceptance criteria. The limiting single failure can be a common power supply, an injection valve, a system pump, or system initiation logic. The most limiting single failure may vary with break size and location.
The potential limiting single failures are shown below:
: a. DC power (SF-BATT)
: b. Diesel generator (SF-DGEN)
: c. LPCI injection valve (SF-LPCI)
: d. High-pressure coolant injection system (SF-HPCI)
: e. ADS valve (SF-ADS)
 
All three ADS valves are assumed operable in the analyses that do not consider SF-ADS. The single failures and the available ECCS for each failure assumed in these Framatome analyses are consistent with Table 14.7-11. Other potential failures are not specifically considered because they result in as much or more ECCS capacity.
 
A comparison of the systems remaining for each of the assumed failures shows that the diesel generator failure (SF-DGEN) always has more available ECCS capacity than the battery failure (SF-BATT). Therefore, PCT results obtained for SF-BATT bound those obtained for SF-DGEN. Similarly, a comparison of the systems remaining for HPCI failure (SF-HPCI) shows that it also always provides more available ECCS capacity than the SF-BATT. Therefore, PCT results obtained for SF-BATT bound those obtained for SF-HPCI. As a result, results reported in References 10 and 208 are those considering SF-BATT, SF-LPCI, and SF-ADS.
 
The single-failure evaluation and the list of available systems shown in Table 14.7-11 was in part the basis for Required Actions and Completion Times in TS 3.5.1, Emergency Core Cooling System, to allow a 72-hour completion time to restore a low-pressure ECCS subsystem to operable status after discovery of two low-pressure ECCS subsystem inoperable (Reference 151 and 152).
 
The single-failure evaluation showing the remaining ECCS following an assumed failure and the effects of a single failure or operator error that causes any manually controlled, electrically operated valve in the ECCS to move to a position that could adversely affect the ECCS are presented in Reference  40.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 55 of 239
 
14.7.2.8.3 ECCS Equipment Performance
 
The ECCS-LOCA analyses were performed using a relaxed set of ECCS injection timing (and other) parameters. The relaxed parameters were selected as a result of a mutual agreement between the fuel vendor and NSP prior to the start of the final ECCS-LOCA analysis calculations.
 
The effectiveness of the ECCS in mitigating the consequences of a LOCA depends upon the performance characteristics of the ECCS. These characteristics can be grouped into two broad categories:
 
ECCS injection time, and
 
inventory delivery.
 
ECCS injection time is defined as the time that elapses between the occurrence of a LOCA and the time that the ECCS flow enters the vessel. The injection time is controlled by a variety of parameters that are dependent not only on the equipment performance characteristics, but also on the reactor response to the LOCA (e.g., water level and vessel pressure), which is a function of the break size and location being considered. Inventory delivery is defined as the rate of ECCS flow being delivered to the vessel. Inventory delivery is controlled by both the ECCS pump and piping characteristics as well as the vessel response to a LOCA.
 
The ECCS-LOCA analysis is sensitive to changes in both the time of ECCS injection and the rate of inventory delivery. However, due to the relaxations incorporated into the analysis, variations in individual parameters can be tolerated without invalidating the calculated ECCS performance response. The criteria for determining acceptability is that the time of ECCS injection be no greater, and the inventory delivery no less, than that used in the licensing calculations. It is important to note that other non-LOCA effects associated with modifications to the ECCS performance characteristics must also be considered in order to determine the acceptability of a variation in any of these parameters. In order to determine the acceptability of individual parameter variations, the logic which controls ECCS injection and the parameters which control injection and inventory delivery must first be understood.
 
Figures 14.7-11 and 14.7-12 schematically show the ECCS initia  tion time logic diagrams for the CS and LPCI Systems. These logic diagrams are composed of two types of parameters: initiation signals and equipment performance parameters.
 
The initiation signals are plant permissives and signals which control ECCS injection.
Typical initiation signals are high drywell pressure signals, low water level signals, injection valve pressure permissives, and the pump shutoff head. The initiation signals are represented as rectangles in Figures 14.7-11 and 14.7-12. The times at which the initiation signals occur depend on the vessel blowdown and break flow rates which are functions of the break location and size.
 
The equipment parameters represent the time required for ECCS equipment to accomplish an action. Typical equipment parameters are diesel generator start time, delay time for loading equipment on the emergency busses, and valve stroke times.
Equipment parameters are represented as ellipses in Figures  14.7-11 and 14.7-12.
The individual ECCS equipment p  arameters used for all breaks in the Monticello ECCS-LOCA analyses (References 10 and 208) are listed in Table 14.7-12.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 56 of 239
 
The time required to complete a path is a combination of the initiation signal times and the equipment parameter times. The longest path (longest time) in a logic diagram determines the time of ECCS injection. As shown in Figures 14.7-11 and 14.7-12, several paths are represented, each of which must be completed before ECCS can inject:
 
Path 1:    The ECCS pumps must be at rated speed. For this to occur, the diesel generator must be started and powering the emergency busses and the pumps must be loaded on the emergency busses and allowed time to achieve rated flow.
Path 2:    The injection valves must be opened. In order to open the valves, two conditions must first be satisfied:  (a) Power must be established at the valves (i.e., the diesel generators must be started and powering the emergency busses), and (b) the reactor pressure must be less than the injection valve pressure permissive.
Once both of these conditions are satisfied, the injection valves must have time to stroke open.
Path 3:    The vessel pressure must be reduced to below the ECCS pump shutoff head. Coolant cannot be pumped into the vessel until the difference between reactor pressure and suppression pool pressure (source of ECCS coolant) is less than the pump shutoff head.
Path 4:    For the LPCI system (Figure 14.7-12), a fourth logic path is included. This involves closing the recirculation discharge valve in the unbroken loop. Closure of this valve directs the LPCI flow upward through the jet pump drive lines and into the jet pumps, thereby preventing the flow from being lost out the break. In order to close the recirculation discharge valve, power must be established at the valve (i.e., the diesel generators must be started and powering the emergency busses). Once this condition is satisfied the discharge valve must have time to stroke closed.
From Figures 14.7-11 and 14.7-12, equations can be developed to determine the time duration for the completion of each path. These equations are listed below. The variables are shown in Figures 14.7-11 and 14.7-12 and are defined below.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 57 of 239
 
CS Logic Paths (See Figure 14.7-11)
Path 1:    THDWS + THDW + TDGS + TDG  +  TCSPR Path 2a:    THDWS + THDW + TDGS + TDG + TCSPV + TCSIV Path 2b:    TCSPP + TCSIV Path 3:    TCSPH LPCI Logic Paths (See Figure 14.7-12)
Path 1:    THDWS + THDW + TDGS + TDG + TCIPR Path 2a:    THDWS + THDW + TDGS + TDG + TCIPV + TCIIV Path 2b:    TCIPP + TCIIV Path 3:    TCIPH Path 4:    THDWS + THDW + TDGS + TDG + TPDV + TDV Initiation Signals:
THDWS =  Time delay between time at which signal setpoint is reached (i.e.
High Drywell Pressure) and initiating signal transmitted.
THDW =    Time to reach high drywell pressure signal after LOCA initiation.
TCSPP =  Time for reactor pressure to drop to the pressure permissive of the CS injection valve.
TCSPH =  Time for reactor pressure to drop below the CS pump shutoff head.
TCIPP =  Time for reactor pressure to drop to the pressure permissive of the LPCI injection valve.
TCIPH =  Time for reactor pressure to drop below the LPCI pump shutoff head.
Equipment Parameters:
TDGS  =  Time delay from Diesel (DG) start signal until DG begins its start sequence.
TDG  =  Diesel generator (DG) startup time TCSPR =  Time for CS pump to achieve rated speed once the DG has started (includes any sequencing delays for pump breaker closure and time for pump to reach rated speed).
TCSPV =  Time to receive power at the CS injection valve once the DG has started (sequence delay time).
TCSIV =  Time to fully stroke open the CS injection valve once it has power.
TCIPR =  Time for the LPCI pump to achieve rated speed once the DG has started (includes any sequencing delays for pump breaker closure and time for pump to reach rated speed).
TCIPV =  Time to receive power at the LPCI injection valve once the DG has started (sequence delay time).
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 58 of 239
 
TCIIV =  Time to fully stroke open the LPCI injection valve once it has power.
(Note:  Rated LPCI flow was assumed to begin with the LPCI injection valve greater than 50% open for Monticello.)
TPDV  =  Time to receive power at the recirculation discharge valve once the DG has started (sequence delay time).
TDV  =    Time to fully stroke the recirculation discharge valve closed once it has power.
 
See the Plant Parameters Document (PPD) (References 11 and 228) for inputs to the ECCS-LOCA analysis.
 
14.7.2.8.4 Evaluation of Parameter Variations
 
Significant margin exists between the equipment performance parameters used in the ECCS-LOCA analyses (References 10 and 208) and actual plant equipment performance. The application of this margin is extremely flexible. The key is that any set of initiation signal times and equipment parameters may be acceptable, provided that no increase occurs in the time of ECCS injection. For example, the actual CS injection valve stroke time may exceed its analytical value of 20 secs by 2 secs if the actual D/G startup time is less than its analytical value by two or more seconds. This would lead to no increase in the time of ECCS injection. This is a simplified example, but it illustrates the flexibility in applying the available ECCS margin.
 
The following methodology should be applied in order to determine if an equipment parameter variation, which may affect the time of ECCS injection, will be acceptable in terms of the ECCS-LOCA analysis:
: 1)    Identify the postulated LOCA events that could be affected by the deviation.
: 2)    Calculate the CS and LPCI injection times using the logic diagrams (Figures 14.7-11 and 14.7-12) and the equations shown in Section 14.7.2.8.3 for the affected cases. The injection time will be the longest time of all the paths for that system.
 
a)  Use the ECCS-LOCA analysis results to evaluate the initiation signal times.
 
b)  Use plant data for the equipment parameters (the equipment parameters shown in Table 14.7-12 are the values used in the ECCS  -LOCA analysis).
: 3)    Verify that the calculated injection time for the actual plant hardware is less than the analytical injection time for the ECCS-LOCA analysis.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 59 of 239
 
The other performance characteristic which may effect the ECCS performance response is the ECCS inventory delivery. The flow delivery curves for the CS and LPCI systems used in the ECCS-LOCA analysis are provided in Figures 14.7-8 and 14.7-9. Deviations in these parameters or potential plant modifications can be evaluated using these flow delivery curves and the information in Sections 14.7.2.8.1.3 and 14.7.2.8.1.4.
 
Trading off small amounts of core spray flow for LPCI has two benefits from a calculational point of view. First, by injecting additional core spray water into the upper plenum, a pool of water will form more rapidly over the top of the fuel bundles.
Second, the small reduction in LPCI injection to the lower plenum delays the time at which the lower plenum becomes subcooled. This delays Counter Current Flow Limiting breakdown at the bottom of the core (side-entry orifice) which allows liquid to be held up in the fuel bundles for a longer period of time. Both of these effects improve the heat removal capability from the fuel and are expected to result in a small reduction in the calculated PCT.
 
The core reflooding time is based on the combined flow of all the ECCS Systems and will not be significantly affected by small tradeoffs in the amount of water delivered by LPCI versus core spray.
 
Thus, it is concluded that excess core spray flow can be used to offset a small shortfall in the total amount of LPCI flow (flow delivered inside shroud plus leakage).
As such, the actual flow requirement for each core spray pump has been established to be 2835 gpm at 130 psid by Technical Specifications. This increase in core spray pump performance over the ECCS-LOCA analysis flow rate of 2672 gpm at 130 psid offsets the 50 gpm leakage from LPCI plus assumptions for CS leakage. The two-pump flow rate requirement for the LPCI pumps does not have to take leakage into account and is identical to the flow rate injected inside the core shroud, 7740 gpm.
 
14.7.2.8.5 Reduced Power Considerations
 
ATRIUM fuel does not require any power/flow reduction in MAPLHGR limits (References 10, 125 and 207).
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 60 of 239
 
14.7.2.8.6 Long Term Core Cooling Performance
 
The NRC acceptance criteria for ECCS performance is contained in 10CFR50.46(b).
Criterion, (b)(5), states:
 
After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
 
The requirements for long term core cooling are met by having sufficient water injection to cover the core from any ECCS system or by providing one core spray pump injecting 3020 gpm of water to the core spray sparger nozzles with the core reflooded to 2/3 core height. The break area created by large recirculation line breaks preclude flooding above 2/3 core height unless the drywell is filled. Such breaks also result in full de-pressurization of the reactor (References 10, 164, and 208).
 
NEDO-20566A, General Electric Company Analytical Model for Loss      -of-Coolant Analysis in Accordance with 10CFR50, Appendix K (Reference 23) presents information concerning compliance with this long-term cooling criterion that is generic to all GE BWRs.
 
Long-term cooling considerations for Monticello include:
 
Recirculation Line Breaks. When the core refloods following the postulated LOCA, the fuel rods will return quickly to saturation temperature over their entire length. For large pipe breaks, the heat flux in the core will eventually be inadequate to maintain a two-phase water flow over the entire length of the core since the static water level inside the core shroud is approximately that of the jet pump suctions. So long as one core spray loop is available, the upper third of the core will remain wetted by the core spray water and there will be no further perforation or metal-water reaction. Table 14.7-11 summarizes the ECCS systems available for all limiting break locations and limiting ECCS single failures. The core spray break event listed in that table is not a long-term cooling concern since the core spray vessel penetrations are located well above the top of the active fuel and the core would remain covered for this event.
 
Recirculation Line Break with LPCI Injection Into Recirculation Piping. Even if a core spray loop is not available long-term, with axial power peaking at mid-plane or lower, the upper region of the core will be cooled by convection to the steam generated in the still-covered region and cladding temperatures will not reach values resulting in further perforation, significant additional oxidation, or significant additional metal-water reaction. Fuel management strategies resulting in axial power peaking above mid-plane require operation of at least one core spray pump to assure adequate core cooling. At least one LPCI System is available except for a recirculation line break with failure of the LPCI injection valve. In this case, two core spray loops will be available.
 
Pipe Breaks Other Than in the Recirculation System. The reactor vessel refloods for all pipe breaks other than the recirculation system, and the fuel cladding quickly cools to saturation temperature. No further perforation or metal-water reaction will result.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 61 of 239
 
During the review of a deviation request pertaining to Monticello plant Emergency Operating Procedures, the NRC evaluated the above considerations relative to the adequacy of core flooding to 2/3 core height for long-term cooling. Their conclusions are discussed in a December 10, 1998 Safety Evaluation Report (Reference 109).
 
Following a large recirculation line break LOCA, the long-term water level in the core will be restored to the top of the jet pumps (approximately two-thirds core height). For design and licensing basis evaluations, one core spray system is assumed available to maintain 2/3 core height and provide adequate long-term cooling to the uncovered upper third of the core. Operation of at least one core spray system is required to maintain adequate long-term core cooling for breaks in which the vessel water level cannot be restored above the minimum steam cooling reactor water level (MSCRWL). For these breaks, the core is quickly reflooded with a two -phase mixture and the fuel rods are cooled to saturation temperature.
 
Adequate long-term core cooling is provided when the cladding temperatures are low enough to prevent further fuel rod perforations, significant additional cladding oxidation, and significant additional metal-water reaction (hydrogen generation). Two different cladding temperature criteria were used during development of the Emergency Procedure Guidelines: (1) Peak Cladding Temperature (PCT) less than 1800&deg;F, the onset of significant metal-water reaction, and (2) PCT less than 1500&deg;F, the onset of significant fuel rod perforations. The acceptance criteria in 10CFR50.46 require that the PCT not exceed 2200&deg;F and that the loca  l cladding oxidation not exceed 17% of the initial clad thickness. For reasonable time duration to recover from the event and submerge the core, a long  -term PCT of less than about 1500&deg;F is required to ensure that both licensing basis criteria are met. Without core spray, the long term cladding temperature may exceed 1500&deg;F with a limiting top peaked core axial power shape.
 
In order to provide adequate long-term core cooling, the core spray system flow must be at the design rated system flow of 3020 gpm  delivered to the core spray sparger nozzles with a reactor to containment differential pressure of 0 psid.  (Break sizes large enough to preclude covering the core will depressurize the reactor). The required flow to the core is based on the original core spray sparger design and testing and relies on achieving the design flow rate through the spray nozzles.
Leakage through sparger cracks, spray piping repairs and pump minimum flow lines requires a pump flow of 3388 gpm at 0 psid to insure that the original core spray sparger design flow is delivered to the nozzles.
 
Adequate core cooling will have been restored when the indicated RPV water level reaches 2/3 core height (top of jet pumps) on the fuel zone level instruments. Once the indicated level stabilizes at 2/3 core height, the operators may begin taking actions to align the ECCS in the post-LOCA long-term cooling configuration. The operators will not be able to determine when the channel fill actually occurs since there is a time lag between core reflooding and the existence of indicated 2/3 core height. The fuel zone instruments (the jet pump level) will lag somewhat because the core is flooding with a highly voided mixture. Initially, the spillover through the jet pump will be a two-phase mixture. Because the level instrumentation senses the collapsed level (only the liquid fraction), the level indication will show as something below 2/3 core height. It will take some additional time for the void fraction in the lower plenum and jet pump to drop and the level indication to show a stable 2/3 core height.
The SAFER code results which support the ECCS-LOCA analysis include plots for jet pump level (two phase), break flow and break flow quality. Based on those plots, between a 15-60 second lag exists between the core being reflooded (channels filled)
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 62 of 239
 
and the fuel zone instrument showing a stable 2/3 core height. A time of 300 seconds for large break is the expected time to achieve stable level indication which allows for initiation of operator actions to throttle ECCS pump flow to long term (>600 seconds) flow rates and initiate containment cooling. Throttling of RHR and CS pumps prior to exceeding 600 seconds is required to meet safety analyses assumptions for core cooling described here, for containment heat removal, and for pump reliability associated with NPSH concerns (References 155, 157, 162, 164, 176, and 192).
 
The NRC reviewed Monticello's assessment for meeting long-term core cooling requirements under Extended Power Uprate (EPU) conditions to 2004 MWt and concluded it was acceptable (Reference 134).
 
14.7.2.9  Radiological Consequences
 
The Loss-of-Coolant Accident radiological consequences were analyzed using Alternative Source Term methodology as provided in Regulatory Guide 1.183 (Reference 129). The accident parameters and assumptions used in the analysis (References 136-140 and 153-154 are summarized below and in USAR Table 14.7-13, and are in accordance with the guidance provided in RG  1.183.
 
14.7.2.9.1 Introduction
 
Even though fuel failures are not predicted for the LOCA sequence of events, the radiological consequences analysis assumes significant fuel damage in accordance with the guidance of RG 1.183.
 
The activity is released from the damaged fuel to the primary containment and then transported to the environment through three pathways. Inhalation doses are calculated separately for each pathway.
 
Primary-to-Secondary Containment Leakage Pathway:  The primary containment is assumed to leak at the Technical Specification limit of 1.2% of containment air weight per day (La). A portion of this leakage is assumed to bypass secondary containment (SCB leakage) and is released via the MSIV/SCB Pathway; the rest is assumed to leak to the secondary containment.
 
MSIV/SCB Leakage Pathway:  The Main Steam Pathway, consisting of the main steam isolation valves (MSIVs) and main steam line drain valves, is assumed to leak at the Technical Specification limit of 200 scfh through the main steam lines and drains to the main condenser. SCB leakage is transported through drain lines to the main condenser. The combined leakage is released from the main condenser to the Turbine Building.
 
ECCS Leakage Pathway:  ECCS systems circulating outside primary containment are assumed to leak through system valve packing, pump seals, or flanged connections to the secondary containment.
 
A secondary containment positive pressure period (PPP) of 5 minutes is assumed at the beginning of the accident until the Standby Gas Treatment System (SGTS) can draw down secondary containment to a negative pressure with respect to the environment. During the PPP, releases to secondary containment are assumed to go directly to the environment. After the PPP, releases to secondary containment
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 63 of 239
 
are processed by the SGTS to the offgas stack. Justification for 5 minute assumption is provided in Reference 15.
 
External shine dose from confined sources to Control Room operators is calculated and added to the Control Room inhalation dose for the total Control Room operator dose.
 
RG 1.183 also directs that design leakage from ECCS systems interfacing with systems with direct release to the environment be considered. This pathway, including leakage to the Condensate Storage Tanks (CSTs) and the condensate service system, was assessed as insignificant compared to the other release pathways and is not included in the analysis.
 
14.7.2.9.2 Source Term
 
The core inventory used for the LOCA analysis source term was calculated assuming operation at 2044 MWt (2004 MWt increased by 2% to account for power measurement uncertainties) and operation at the total average burnup expected for a 24-month fuel cycle. See USAR Section 14.7.8 for further discussion of the inventory development.
 
The core inventory available at accident time T=0 for release is shown in USAR Tables 14.7-24a, 14.7-24b, and 14.7-24c for the three different fuels/burnups, in any combination; that may be in the core. The source terms in Tables 14.7-24d and 14.7-24e are not used for the LOCA at this time, so the LOCA radiological consequences evaluation limits the core average exposure to 37 GWD/MTU until the LOCA radiological consequences analysis is updated with Tables    14.7-24d and 14.7-24e.
 
Fission products from the damaged fuel are assumed to be released to the primary containment in two phases. The gap release phase is initiated 2 minutes after the start of the accident and lasts for one-half hour. During this phase, activity is released from the fuel rod gap with a release fraction of 0.05 of the total rod activity for noble gases, halogens, and alkali metals. A linear release of the gap activity over the phase duration is assumed.
 
The early in-vessel phase begins immediately following the end of the gap release phase and lasts for 1.5 hours. The release is from the damaged fuel pellets, with release fractions of 0.95 for noble gases, 0.25 for halogens, and 0.20 for alkali metals. Release fractions for the remaining nuclides are shown in USAR Table 14.7 -13. A linear release of the activity over the phase duration is assumed.
 
The release from the fuel to the primary containment is terminated at the end of the early in-vessel phase, with total release fractions from the fuel of 1.0 for noble gases, 0.30 for halogens, and 0.25 for alkali metals.
 
The suppression pool pH is maintained greater than 7 (basic) post -accident by injection of sodium pentaborate from the Standby Liquid Control System, resulting in primary containment radioiodine composition of 95% cesium iodide (C  sI) as an aerosol, 4.85% elemental iodine, and 0.15% organic iodide.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 64 of 239
 
The source term for the MSIV/SCB Leakage pathway consists of the activity released to the primary containment, as decreased by natural deposition within the drywell. At two hours post-accident the drywell airspace is assumed to mix with the torus airspace, thus diluting the primary containment activity source. Credit is also taken for the reduction in primary containment activity by Primary-to-Secondary Containment leakage.
 
The source term for the Primary-to-Secondary Containment Leakage pathway consists of the activity released to the primary containment, as decreased by natural deposition within the drywell. No credit is taken for the torus airspace or for reduction in primary containment activity by releases through other pathways.
 
The source term for the ECCS Leakage pathway consists of the total activity released from the fuel except for the noble gases. The activity is assumed to instantaneously mix in the suppression pool at the time of release from the core, then is recirculated by ECCS systems and released through system valve packing, pump seals, flanged connections, etc. 90% of the radioiodines and all of the radionuclides other than iodine are assumed to be retained in the liquid phase, resulting in a release consisting of 10% of the radioiodines in the leaked fluid. The 10% flash fraction is based on suppression pool liquid temperature maintained at less than 212&deg;F. The radioiodine released from the ECCS leakage is assumed to be 97%
elemental and 3% organic. No credit is taken for reduction in primary containment activity by releases through other pathways.
 
The source term for external shine dose to Control Room operators is the activity confined in the reactor building airspace, the activity in the airborne cloud external to the Control Room, activity deposited on the SGTS and EFT filters, and activity contained in ECCS piping recirculating reactor coolant inside the reactor building.
 
14.7.2.9.3 Mitigation
 
Natural deposition of the particulate (aerosol) activity within the drywell is credited using the Powers 10th Percentile Natural Deposition Model, reducing the amount of activity released to the MSIV/SCB and Primary-to-Secondary Leakage pathways.
 
Sodium pentaborate is assumed to be injected by the Standby Liquid Control System in sufficient quantity to maintain the suppression pool pH greater than 7, preventing any significant conversion of particulate radioiodine to elemental radioiodine and resulting in greater removal of radioiodine species prior to release to the environment. The injection is assumed to be completed within two hours post-accident.
 
No credit is assumed for suppression pool scrubbing, drywell or torus spray operation, or holdup/removal using drywell HVAC.
 
MSIV/SCB Leakage pathway:
For MSIV leakage, natural deposition of radioactive particulates is credited in the main steam lines and associated drains (Reference 110). Natural deposition is also credited for the secondary containment bypass leakage through the steam line drains. Further deposition and holdup for the combined leakage occurs in the main condenser prior to release to the environment via the Turbine Building vent.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 65 of 239
 
Primary-to-Secondary Containment Leakage:
Prior to secondary containment drawdown by the Standby Gas Treatment System, this activity is released directly to the environment as a ground-level release. After secondary containment drawdown, this leakage is collected in the secondary containment and released to the environment through the SGTS to the offgas stack for a filtered elevated release. The SGTS filter efficiency is assumed at 85% for the adsorber section, which removes elemental and organic iodines, and 98% for the particulate filters.
 
ECCS leakage:
Prior to secondary containment drawdown by the SGTS, this activity is released directly to the environment as a ground-level release. After the secondary containment drawdown, this leakage is collected in the secondary containment and released to the environment through the SGTS to the offgas stack for a filtered elevated release.
 
The Control Room ventilation system emergency mode (EFT) is initiated by the LOCA signals (drywell high pressure or reactor vessel low level) prior to the accident release initiation at two minutes post-accident. The EFT filter efficiency is assumed at 98% for both the adsorber section and the particulate filters.
 
14.7.2.9.4 Transport
 
MSIV/SCB Leakage Pathway:
 
The total core source term is released directly into the drywell airspace from the reactor vessel. At 2 hours post-accident, the primary containment airspace is increased to include both the drywell airspace and the torus airspace resulting in lower airborne concentration.
 
MSIV leakage at the Technical Specification limit of 200 scfh is initially assumed to transport through two of the four main steam lines at 100 scfh each with the further assumption that the inboard MSIV has failed open on one of the two lines. This assumed failure limits the piping surface area credited for natural deposition. Natural deposition of radioactive particulates is credited for the piping between the inboard and outboard MSIV on one steam line (the one of shortest distance) and in the drain lines from two of the main steam lines to the main condenser (in the two shortest drain line paths). Since a single failure of an inboard MSIV in one steam line is assumed, natural deposition is not credited between the MSIVs in this line.
 
Secondary containment bypass (SCB) leakage is initially assumed at 35.2 scfh from within the drywell through drain lines to the main condenser. Natural deposition of SCB leakage in the shortest drain line path to the main condenser is assumed.
 
The MSIV/SCB leakage rates decrease over time due to post-accident primary containment depressurization. The leakage is assumed to be 100% (200 scfh and 35.2 scfh) for the first 24 hours, at 66% for the next 66 hours, and at 50% for the remainder of the accident, based on the post-accident pressure/temperature profile.
 
Deposition and holdup of the combined MSIV/SCB leakage is assumed in the main condenser. A ground level release from the main condenser via the Turbine Building vent is assumed for a duration of 30 days post-accident.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 66 of 239
 
Primary-to-Secondary Containment Leakage Pathway:
 
The total core source term is released directly into the drywell airspace from the reactor vessel. The primary containment is assumed to leak to the secondary containment at the Technical Specification limit of 1.2% containment air weight per day, excluding the SCB leakage (Reference 115).
 
The leakage rate decreases over time due to post-accident primary containment depressurization. The leakage is assumed to be 100% (1.2% excluding SCB leakage) for the first 24 hours, at 66% for the next 66 hours, and at 50% for the remainder of the accident, based on the post-accident pressure/temperature profile.
 
At accident onset (T=0), the secondary containment is assumed to pressurize prior to drawdown by the SGTS. This positive pressure period (PPP) lasts for five minutes (Reference 15). During the PPP, all activity in the secondary containment is assumed to be released directly to the environment as a ground-level release.
Release duration is 3 minutes since there is no activity release to the drywell for the first two minutes of the accident.
 
Following the PPP, the release from this pathway is through the SGTS to the offgas stack for a filtered elevated release. No credit is taken for holdup or dilution in the secondary containment.
 
ECCS Leakage Pathway:
 
The total core source term, with the exception of noble gases, is assumed to release directly into the suppression pool liquid volume. The release from the core is assumed to occur over a 2 hour period and, as released, instantaneously and homogeneously mix in the suppression pool volume. No credit for reactor vessel or recirculation system piping volumes is assumed to further dilute the activity in the suppression pool.
 
The suppression pool liquid is recirculated by ECCS systems and released through system valve packing, pump seals, flanged connections, etc., to the secondary containment. The leakage is assumed to flash and release 10% of its radioiodine activity, based on a suppression pool temperature of less than 212&deg;F. Although different systems will operate for different durations over the course of the accident, a total combined leakage rate is conservatively assumed to remain constant for the entire accident duration. The leakage is based on a design rate of 1.31 gpm, which is then doubled in accordance with RG 1.183 for a total rate of 2.62 gpm.
 
During the PPP, all activity in the secondary containment is assumed to be released directly to the environment as a ground-level release. Release duration is 3 minutes since there is no activity release to the drywell for the first two minutes of the accident.
 
Following the PPP, the release from this pathway is through the SGTS to the offgas stack for a filtered elevated release. No credit is taken for holdup or dilution in the secondary containment.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 67 of 239
 
Control Room ventilation is assumed in the emergency mode throughout the accident release period, with 900 cfm of EFT filtered air intake assumed. An additional 500 cfm of unfiltered inleakage is assumed. Control Room dose studies were performed at several lower air intake and unfiltered inleakage flow rates, verifying that the flow rates given above are limiting.
 
For the elevated release from the offgas stack, fumigation (an atmospheric condition resulting in increased ground-level exposure to accident releases) is assumed for one-half hour. In accordance with the guidance of RG 1.183, the fumigation period is assumed during the worst 2-hour period for EAB exposure (1.7 hours to 2.2 hours).
For consistency, the same timing is assumed for the LPZ and Control Room dose assessments.
 
Control Room and offsite atmospheric dispersion coefficients are shown in USAR Table 14.7-13 (References 153 and 154).
 
14.7.2.9.5 Results
 
Control Room operator and offsite accident dose s are shown in USAR Table 14.7-14.
 
14.7.3  Main Steam Line Break Accident Analysis
 
Accidents that result in the release of radioactive materials outside the secondary containment are the results of postulated breaches in the nuclear system process barrier. The design basis accident is a complete severance of one main steam line outside the secondary containment. Table 14.7-13 shows the break location. The analysis of the accident is described in three parts as follows:
: a. Primary System Transient Effects
 
This includes analysis of the changes in primary system parameters pertinent to fuel performance and the determination of fuel damage.
: b. Radioactive Material Release
 
This includes determination of the quantity and type of radioactive material released through the pipe break and to the environs.
: c. Radiological Consequences
 
This portion determines the dose effects of the accident to offsite persons.
 
The analysis of the main steam line break accident depends on the operating thermal-hydraulic parameters of the overall reactor (such as pressure) and overall factors affecting the consequences (such as primary coolant activity). The most limiting main steam line break radiological consequences are associated with a steam line break outside containment. Consequences of the main steam line break accident are independent of fuel design, since the radioactive release is dependent on primary coolant activity and not on fuel design parameters.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 68 of 239
 
14.7.3.1  Reactor Primary System Transient Effects and Mass and Energy Releases
 
The mass and energy release for the Main Steam line Break outside containment (Reference 121) was calculated using the same SAFER/GESTR-LOCA model used in the Rerate (1880 MWt) ECCS-LOCA analysis for Monticello (Reference 21). There was no change in the mass and energy release for the hot standby main steam line break related to the increased power level associated with Extended Power Uprate (EPU) operation since at EPU conditions there is no increase in pressure and enthalpy in the reactor and the break locations (References 134 and 160).
 
The mass and energy release for the steam line break is largely determined by the amount of liquid discharged through the break. Following the break, the vessel rapidly depressurizes because the steam generation from the decay power cannot make up the steam loss through the break. The rapid d  epressurization causes the water in the vessel to flash and swell up to the steam lines, resulting in a steam-water mixture flowing out the break. This mixture flow continues until the MSIVs close. The core remains adequately cooled throughout the accident and no fuel damage will occur.
 
The steam line break flow is determined by the reactor pressure and the steam line flow restrictor area. The initial core power determines the amount of steam generation during this period, which in turn determines the depressurization rate and resulting level swell. A higher initial core power level results in a higher steam generation rate. The combination of the unchanged break flow and higher steam generation rate results in a lower vessel depressurization rate and delays the level swell. Because the MSIV closure time is constant, the delayed level swell results in less steam-water mixture being released out the break. The mass and energy release time histories for operation at 1880 MWt is shown in Table 14.7-15. Based on the preceding discussion, the total mass and energy release listed on Table 14.7-15 bounds the total mass and energy release that would occur for a break at rated EPU power level of 2004 MWt. An initial reactor pressure of 1025 psia was assumed. The MSIVs are fully closed in 10.5 secs. The Homogeneous Equilibrium Model (HEM) with a break flow multiplier of 1.2 was used to calculate the break flows for this case.
No frictional losses were assumed in the break flow calculations. As described in Reference 24, the HEM is generally accepted as providing the best fit to experimental data for saturated upstream conditions. Subcooled break flow is often under predicted by the HEM without adjustment. Comparison of the blowdown data    over the range of test conditions (saturated and subcooled) shows that virtually all of the test data fall within +/-20% of the HEM prediction. As shown in Table 14.7-15, the break flow remains saturated throughout the event. Therefore, a multiplier of 1.2 on the break flow provides a sufficient degree of conservatism for the mass release calculations.
 
The case shown in Table 14.7-16 assumes an initial reactor pressure of 965 psia, equal to the turbine inlet pressure. This is the reactor pressure expected in hot standby conditions where the steam from the reactor is being directed to the condenser. The case shown in Table 14.7-17 non-mechanistically assumes an initial reactor pressure of 1158 psia (SRV opening pressure with 3% tolerance) and was used to provide a bounding release for the radiological calculations. By minimizing the steam generation from the core, hot standby conditions maximize the level swell in the vessel, thus maximizing the mass and energy release from the break. Since hot standby conditions are not affected by power, the mass and energy release rates are valid for all licensed power levels. The total integrated mass releases shown in Tables 14.7-16 and 14.7-17 are directly calculated by the SAFER code; slight round
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 69 of 239
 
off errors were introduced when calculating the liquid and steam releases for the radiological analysis. No frictional losses were assumed in the break flow calculations. The Moody Slip break flow model was used to calculate the break flows for these cases.
 
The decrease in steam pressure at the turbine inlet initiates closure of the main steam line isolation valves after the break occurs (see Primary Containment and Reactor Vessel Isolation Control System Section 7.6.3). Also, main steam line isolation valve closure signals are generated as the differential pressures across the main steam line flow restrictors increase above isolation setpoints. The instruments sensing flow restrictor differential pressures generate isolation signals within about 600 milliseconds after the break occurs (Reference 168).
 
A reactor scram is initiated as the main steam line isolation valves begin to close (see Reactor Protection System Section 7.6.1). In addition to the scram initiated from main steam line isolation valve closure, voids generated in the moderator during depressurization contribute significant negative reactivity to the core even before the scram is complete. Because the main steam line flow restrictors are sized for the main steam line break accident, reactor vessel water level remain s above the top of the fuel throughout the transient.
 
14.7.3.2  Radiological Consequences
 
The main steam line break accident radiological consequences were analyzed using Alternative Source Term methodology as provided in Regulatory Guide 1.183 (Reference 129). The accident parameters and assumptions used in the analysis (References 16 and 136-140) are summarized below and in USAR Table 14.7-19, and are in accordance with the guidance provided in RG 1.183.
 
14.7.3.2.1 Introduction
 
The postulated accident involves a guillotine break of one of the four main steam lines outside the containment, resulting in mass loss from both ends of the break.
There is no fuel damage as a consequence of this event; therefore, the only activity released to the environment is that associated with the steam and liquid discharged from the break. Initially, only steam will issue from the broken end of the steam line.
Subsequently, rapid depressurization due to the break causes the reactor pressure vessel water level to rise, resulting in a steam-water mixture flowing from the break (blowdown) until the main steam isolation valves are closed.
 
It is assumed that the accident occurs at hot standby conditions. At these conditions, steam generation from the decay heat in the core is very low and cannot make up the steam loss through the break. The results are high rate of vessel depressurization and rapid rising of water level to the main steam line inlet. In addition to hot standby conditions, the Appendix K break flow model was assumed in order to maximize the two-phase break flow rate. Both of these assumptions yielded the maximum coolant mass releases through the break.
 
Hot standby (66.8 MWt) power, steam flow, and feedwater flow rate were used in the actual SAFER calculations to generate coolant mass releases. Two cases are studied: the first case assumes reactor pressure initially is at the safety relief valve opening setpoint plus 3%, 1158 psia. The second case assumes the initial reactor pressure at the pressure regulator setpoint, 965 psia. As shown in Table 14.7-18, the
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 70 of 239
 
total integrated mass leaving the reactor pressure vessel through the break is 86152 lb. in the first case, of which 71574 lb. is liquid. In the second case it is 78617 lb., of which 66223 lb. is liquid.
 
For the radiological consequences analysis, the mass of coolant released is the amount of mass in the steam line and connecting lines at the time of the break plus the amount passing through the MSIVs prior to closure, as analyzed above. The mass released from the break is taken from USAR Table 14.7-18 (Hot Standby Case
: 1) and then scaled upward by approximately 6% for added conservatism.
 
14.7.3.2.2 Source Term
 
The AST analysis for the Main Steam Line Break Accident (MSLBA) is performed at hot standby. Hot standby power level is assumed to be 66.8 MWt. This power level provides a more conservative impact on off site dose than a full power break (Reference 16). MELLLA+/EFW does not impact the AST analysis for MSLBA (Reference 182).
 
There is no fuel damage as a consequence of this accident; therefore, the only activity released to the environment is that associated with the steam and liquid discharged from the break, consisting of radioiodines and noble gases
 
Two cases were performed based on the allowable limits for reactor coolant iodine activity in MNGP Technical Specifications. The equilibrium case assumes an activity concentration of 0.2 Ci/g dose-equivalent I-131 in the released coolant. The pre-accident iodine spike case assumes an activity concentration of 2  Ci/g dose-equivalent I-131. The radioiodine species released from the coolant are assumed to be 95% aerosol, 4.85% elemental, and 0.15% organic.
 
A portion of the released coolant exists as steam prior to blowdown, and as such does not contain the same iodine concentration per unit mass as the steam generated through blowdown. Therefore, it is necessary to separate the initial steam mass from the total mass released and assign a certain percentage (2% carryover is assumed) of the fission product activity contained in this portion of steam by an equivalent mass of primary coolant. See USAR Table 14.7-18 for equivalent mass.
This equivalent mass does not apply to noble gases, which are released 100% from both the steam and liquid coolant.
 
An off-gas release rate of 300,000 Ci/sec after 30 minutes of decay is used to calculate the undecayed noble gas emission rate for the coolant release. This value exceeds the Technical Specification allowable limit for gross gamma activity by approximately 15%. The activity is assumed to consist of a standard isotopic fraction based on measurement data.
 
Alkali metals (Cs and Rb) were evaluated and indicated that the dose due to alkali metals in the released coolant was determined to be negligible (Reference 136).
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 71 of 239
 
14.7.3.2.3 Mitigation
 
The only mitigative action credited for the MSLBA is the termination of the release upon the automatic closure of the MSIVs. A closure time of 10.5 seconds is assumed, including valve closure time and instrument response time for break detection and valve closure initiation.
 
Control Room ventilation is assumed to remain in the normal operating mode throughout the event and no credit for emergency mode filtration or isolation is assumed. No credit is taken for operator action.
 
14.7.3.2.4 Transport
 
Noble gases are assumed to enter the steam phase instantaneously. The total mass of coolant released, prior to MSIV closure, is the amount in the steam line and connecting lines at the time of the break plus the amount that passes through the valves prior to closure. The radioactivity in the released coolant is assumed to be released instantaneously to the atmosphere as a ground level release from the Turbine Building vent. No credit for plateout, holdup, or dilution within the Turbine Building is assumed.
 
CR ventilation remains in the normal mode throughout the accident, with 7,440 cfm of CR air intake assumed, representing the maximum normal CR air intake rate (i.e.,
no intake blanking plates installed and no recirculation of intake). An additional 1,000 cfm of unfiltered inleakage is assumed. CR dose studies were performed at several lower air intake and unfiltered inleakage flow rates, verifying that the maximum flow rates of 7,440 cfm and 1,000 cfm are limiting.
 
Control Room and offsite atmospheric dispersion coefficients are shown in USAR Table 14.7-19.
 
14.7.3.2.5 Results
 
Control Room operator and offsite accident doses are shown in USAR Table 14.7-20.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 72 of 239
 
14.7.4  Fuel Loading Error Accident
 
A loading error in the core configuration is considered to be either an error in orientation (i.e., misoriented - rotated 90&deg; or 180&deg;) or location (misplaced) of one or more of the bundles.
 
Proper orientation of fuel assemblies in the reactor core is readily verified by visual observation and assured by verification procedures during core loading. Five separate visual indications of proper fuel assembly orientation exist:
 
(1)  The channel leaf spring assemblies, including the spring and guard used to maintain clearances between channels, are located at one corner of each fuel assembly adjacent to the center of the control rod.
 
(2)  The identification boss on the fuel assembly handle points toward the adjacent control rod.
 
(3)  The channel spacing buttons are adjacent to the control rod passage area.
 
(4)  The assembly identification numbers which are located on the fuel assembly handles are all readable from the direction of the center of the cell.
 
(5)  There is cell-to-cell replication.
 
Because of plant administrative procedures during fuel loading and the aforementioned bundle mechanical design features, the probability of a significant fuel loading error (based on the probability assessments given in Reference 42) is much less than once in a plant lifetime. Additionally, it requires multiple operator errors. Thus, the fuel loading error is classified as an accident, not a transient, so application of LHGR limits is not appropriate.
 
Improper loading and operation of a fuel assembly is evaluated relative to GDC 13 as it relates to instrumentation and monitoring and 10CFR50.67 as it relates to offsite consequences. The misloaded bundle accident is evaluated on a cycle-by-cycle basis.
The acceptance criteria used for the reload licensing analysis is that the MCPR in the core must be greater than the safety limit MCPR with a misloaded bundle present.
analyses show that this criteria is met, as well as that no fuel rods approach centerline melt or 1% strain (Reference 6). See USAR  Section  14A for the current cycle Misplaced and Misoriented Fuel Loading Error Accident results.
 
14.7.5  One Recirculation Pump Seizure Accident Analysis
 
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.
 
The pump seizure event is a very mild accident in relation to other accidents such as the LOCA. This is easily verified by consideration of the two events. In both accidents, the recirculation driving loop flow is lost extremely rapidly - in the case of the seizure, stoppage of the pump occurs; for the LOCA, the severance of the line has a similar, but more rapid and severe influence. Following a pump seizure event, flow continues, water level is maintained, the core remains submerged, and this provides a continuous core
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 73 of 239
 
cooling mechanism. However, for the LOCA, complete flow stoppage occurs and the water level decreases due to loss of coolant resulting in uncovery of the reactor core and subsequent overheating of the fuel rod cladding. In addition, for the pump seizure accident, reactor pressure does not significantly decrease, whereas complete depressurization occurs for the LOCA. Clearly, the increased temperature of the cladding and reduced reactor pressure for the LOCA both combine to yield a much more severe stress and potential for cladding perforation for the LOCA than for the pump seizure. Therefore, it can be concluded that the potential effects of the hypothetical pump seizure accident during two-loop operation are very conservatively bounded by the effects of a LOCA and specific analyses of the pump seizure accident are not required.
 
During single loop operation, however, seizure of the operating pump may be a limiting event. This event is evaluated on a cycle-dependent basis for Monticello against the acceptance criteria for plant transients. Acceptance criteria for transients are based on avoiding transition boiling and maintaining the fuel within thermal and mechanical limits.
This analysis is performed assuming single loop operation. The result of this event is provided in Section 14A.
 
14.7.6  Refueling Accident Analysis
 
14.7.6.1  Identification of Causes
 
Accidents that result in the release of radioactive materials directly to secondary containment can occur when the drywell is open. A survey of the various conditions that could exist when the drywell is open reveals that the greatest potential for the release of radioactive material occurs when the drywell head and reactor vessel head have been removed. In this case, radioactive material released as a result of fuel failure is available for transport directly to secondary containment.
 
Various mechanisms for fuel failure under this condition have been investigated.
With the current fuel design, refueling interlocks that impose restrictions on the movement of refueling equipment and control rods prevent an inadvertent criticality during refueling operations. Administrative procedures are also utilized to prevent an inadvertent criticality during refueling operations. In addition, the reactor protection system can initiate a reactor scram in time to prevent fuel damage for errors or malfunction occurring during planned criticality tests with the reactor vessel head off.
It is concluded that the only accident that could result in the release of significant quantities of fission products to the containment during this mode of operation is one resulting from the accidental dropping of a fuel bundle onto the top of the core or onto the fuel bundles in the spent fuel pool.
 
Analysis has demonstrated that the accident over the core is more limiting than the accident over the spent fuel pool, provided the spent fuel pool has sufficient depth of water (Reference 136).
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 74 of 239
 
14.7.6.2  Effect of Fuel Densification
 
This event occurs under non-operating conditions for the fuel. The key assumption of this postulated occurrence is the inadvertent mechanical damage to the fuel rod cladding as a consequence of the fuel bundle being dropped on the core while in the cold condition.
 
Fuel densification considerations do not enter into or affect the accident results.
 
14.7.6.3  Radiological Consequences
 
The Fuel Handling Accident radiological consequences were analyzed using Alternative Source Term methodology as provided in Regulatory Guide 1.183 (Reference 129). The accident parameters and assumptions used in the analysis (References 136-140 and 175) are summarized below and in USAR Table 14.7-21, and are in accordance with the guidance provided i n RG 1.183.
 
14.7.6.3.1 Introduction
 
The limiting fuel-handling accident assumes that the drywell head and the reactor vessel head are removed, and a fuel bundle was accidentally dropped on the core.
The drop height into the fuel pool will be less than that into the core. Therefore, a fuel bundle dropped on top of the core results in more damaged rods.
 
If the depth of water above the damaged fuel is 23 feet or greater, the decontamination factors for the elemental and organic iodine species are 500 and 1, respectively, giving an overall effective decontamination factor of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water). For the accident over the core in the reactor cavity, the water depth is much greater than 23 feet so a decontamination factor of 200 is assumed. Although there is less than 23 feet above damaged fuel in the spent fuel pool or the reactor flange area and therefore a lower decontamination factor, the reactor cavity accident remains bounding due to the greater amount of fuel damage in the cavity.
 
The number of rods assumed to fail in a fuel-handling accident is dependent both on the fuel design and the design of fuel handling equipment. The mast and any attached sections are assumed to be 650 lbf, the maximum drop height of 34 feet, and failure is assumed when the clad reaches in 1% strain in compression; further conservative assumptions are described in section S.2.2.3.5 of the US supplement of Reference 106. The fuel handling accident analysis assumes a drop over the core results in 125 failed rods in an 8x8 bundle that has 60 rods per bundle, totaling 2.083 failed bundles or 0.43% of the core.
 
While legacy fuel types 7x7, 8x8, 9x9, 10x10 (all types prior to GE14) fuel remain in the pool, all these legacy fuel types have cooled for several years and review of the half-lives of the most important isotopes in Tables 14.7-24a, b, and c shows that this decay time significantly reduces the source term. Therefore, even due to the heavier 7x7 bundles (Reference 106) and the various weights of other legacy bundle types (References 76 and 106) that result in a varied number of rod failures, the overall source term from failed fuel is bounded by the assumption of 2.083 recently irradiated bundle failures.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 75 of 239
 
The refueling accident analysis involving GE14 fuel was evaluated in References 122 and 123. It was concluded in Reference 123 that failure of 2.083 bundles bounds the amount of failed GE14 fuel.
 
Any drop onto ATRIUM 10XM fuel would have an equivalent 1.833 failed bundles (Reference 175). Drops onto ATRIUM 11 fuel would have an equivalent 1.788 failed bundles (Reference 175). Therefore, the total number of failed bundles for any current or recent fuel type is bounded by the 2.083 failed bundles used in the dose analysis.
 
14.7.6.3.2 Source Term
 
The core inventory used for the FHA analysis source term was calculated assuming operation at 2044 MWt (2004 MWt increased by 2% to account for power measurement uncertainties) and operation at the total average burnup expected for a 24-month fuel cycle. See USAR Section 14.7.8 for further discussion of the inventory development.
 
The core inventory available at accident time T=0 for release is shown in USAR Tables 14.7-24c, 14.7-24d, and 14.7-24e for the three different fuels/burnups, in. any combination; that may be in the core.
 
The source term for this event is the gap activity in the 125 fuel rods assumed damaged as a result of the drop in the reactor cavity. This number of fuel pins equals approximately 0.43% of the total number of fuel rods in the reactor core (125 equivalent 8x8 fuel rods from a total of 484 core bundles with 60 rods per bundle).
The total fuel rod gap activity available for release from the reactor core is based on the core inventory noted above, with a 24 hour decay period following reactor shutdown. The fraction of radionuclides in the fuel gap assumed available for release is shown in USAR Table 14.7-21.
 
Alkali metals (Cs, Rb) are released from the gap but are not included in the analysis source term since all particulate radionuclides are assumed to be retained in the water and no airborne alkali metals are produced as daughter products during the 2-hour event.
 
Of the gap activity released from the damaged fuel rods, 100% of the noble gases and a fraction of the iodines are assumed available for release. The chemical form of the radioiodine released from the fuel is 95% aerosol (CsI), 4.85% elemental and 0.15% organic. Due to the possibility of low pH in the pool, CsI is assumed to instantaneously disassociate, with the iodine re-evolving in elemental form. This results in 99.85% elemental and 0.15% organic iodine.
 
An overall pool Decontamination Factor (DF) of 200 is assumed for radioiodine releases. A DF of 0 (no retention) is assumed for noble gases, and an infinite DF (complete retention) is assumed for particulates.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 76 of 239
 
14.7.6.3.3 Mitigation
 
The primary mitigation mechanisms for the fuel-handling accident are radioactive decay and decontamination of releases by water in the pool above the damaged fuel.
 
A 24 hour decay period is assumed prior to fuel movement. Decontamination of the radioiodine gap activity as it rises (bubbles) to the surface through the pool water above the dropped assembly in the reactor vessel is credited. For a DF of 200 the minimum required water depth over the damaged fuel is 23 ft., which is exceeded in the reactor cavity by normal refueling water level requirements. Technical Specification limits on spent fuel pool water level ensure the fuel assembly drop over the reactor core remains bounding over the fuel assembly drop in the spent fuel pool or over the reactor flange.
 
No other mitigating actions are assumed. No credit for radiation monitor detection of the release or subsequent isolation of secondary containment and initiation of the SBGT system are assumed. Control Room ventilation is assumed to remain in the normal operating mode throughout the event and no credit for emergency mode filtration or isolation is assumed. No credit is taken for operator action.
 
14.7.6.3.4 Transport
 
The release of the gap activity from the damaged fuel rods is assumed to occur instantaneously. Radioactivity that escapes from the pool is assumed released to the environment from the Reactor Building Vent linearly over a period of 2 hours. No credit is assumed for mixing or dilution in the secondary containment.
 
The release is a ground-level release from the Reactor Building Vent. The RB Vent provides a bounding and representative release point regardless of whether the RB ventilation system or SBGT are operating.
 
CR ventilation remains in the normal mode throughout the accident, with 7,440 cfm of CR air intake assumed, representing the maximum normal CR air intake rate (i.e., no intake blanking plates installed and no recirculation of intake). An additional 1,000 cfm of unfiltered inleakage is assumed. CR dose studies were performed at several lower air intake and unfiltered inleakage flow rates, verifying that the maximum flow rates of 7,440 cfm and 1,000 cfm are limiting.
 
Control Room and offsite atmospheric dispersion coefficients are shown in USAR Table 14.7-21.
 
14.7.6.3.5 Results
 
Control Room operator and offsite accident doses are shown in USAR Table 14.7-22.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 77 of 239
 
14.7.7  Accident Atmospheric Dispersion Coefficients
 
Atmospheric dispersion factors (X/Q) provide values that represent the relative dispersion occurring between a source release location and a receptor location. The relative dispersion can then be used to determine the expected atmospheric radionuclide concentration at some defined distance from the source for a known quantity of released effluent.
 
14.7.7.1  Meteorological Data
 
Site meteorological data from the years 1998-2002 were used to calculate accident atmospheric dispersion factors. The site meteorological data collection system is described in USAR Section 2.3. The five years of data provide a representative long-term trend.
 
14.7.7.2  Control Room Atmospheric Dispersion Coefficients
 
Four release (source) points to the environment were modeled:
          - Closest Reactor Building (RB) wall to the CR (ground level release)
          - Reactor Building Vent (ground level release)
          - Turbine Building Vent (ground level release)
          - Offgas Stack (elevated release)
 
Two receptor locations with the potential for introducing outside air into the Control Room were modeled:
          - Control Room outside air intake
          - Administration Building (Admin Bldg) outside air intake
 
Consistent with the guidance of Regulatory Guide 1.194 (Reference 140), the meteorological data collection location closest to the release point was utilized, i.e.,
data collected at the 100 m height were used for the calculation of elevated releases and data collected at 43 m and 10 m were  used for the calculation of ground level releases.
 
The calculated atmospheric dispersion coefficients (X/Q) for the two CR receptor locations are shown in USAR Table 14.7-23. In the radiological DBA analyses, the bounding (larger) source-receptor X/Q values were selected as input to the dose calculations. These bounding values were used for all outside air sources to the Control Room, including the CR ventilation normal and emergency mode air intake and CR unfiltered inleakage.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 78 of 239
 
14.7.7.3  Offsite Atmospheric Dispersion Coefficients
 
The atmospheric dispersion coefficients for offsite receptors were calculated using the guidance of Regulatory Guide 1.194 (Reference 141) and Regulatory Guide 1.145 (Reference 142).
 
Two source (release) points to the environment were modeled:
          - Ground level release
          - Elevated release (offgas stack)
 
The ground level release was modeled as a bounding general release from the Reactor or Turbine Buildings.
 
Two receptor locations were modeled:
          - EAB (Exclusion Area Boundary)
          - LPZ (Low Population Zone).
 
The calculated atmospheric dispersion coefficients (X/Q) for the EAB and LPZ are shown in USAR Table 14.7-23.
 
14.7.8 Core Source Term Inventory
 
The core inventory used for the accident analysis source terms involving fuel damage was calculated assuming operation at 2004 MWt, with 2% added for power measurement uncertainties. The Monticello core source term parameters are as follows (References 27, 174, 163, 172, and 173):
* GE14 (LOCA only), ATRIUM 10XM, or ATRIUM 11 (not analyzed for LOCA) fuel is used
* Maximum bundle average enrichment is 4.6 wt% for GE14 and ATRIUM 10XM fuel and 4.3% for ATRIUM 11 fuel
* GE14 fuel is analyzed at 35 GWD/MTU and 37 GWD/MTU for LOCA only
* ATRIUM 10XM fuel is analyzed at 37 GWD/MTU for all events and 41 GWD/MTU for CRDA and FHA
* ATRIUM 11 fuel is analyzed at 41 GWD/MTU for CRDA and FHA
* Maximum batch average discharge bundle exposure is 58 GWD/MTU
* Maximum initial bundle uranium mass is 182 kg (GE14), 179 kg (ATRIUM 10XM),
or 178.6 kg (ATRIUM 11)
* Maximum bundle average power is 5.75 MWt (at 102% of 2004 MWt) for GE14 fuel and 4.223 MWth for ATRIUM 10XM and ATRIUM 11 fuel
 
The core inventory available for accident release at time T=0 is shown in the subtables of Table 14.7-24
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 79 of 239
 
Table 14.7-2a  CRDA Radiological Consequences Analysis Inputs and Assumptions (Page 1 of 2)
 
Core Power (for establishing isotopic inventory)          2044 MWt (2004 MWt plus 2%)
Core Inventory at Accident Time T=0                      USAR Tables 14.7-24 b, d, and e Radial Peaking Factor                                    1.7 Total Fuel Rods Damaged (8x8 equivalent rods)            850 Cladding Damaged Rods                                841 Melted Rods                                              9 Activity Released to Coolant (cladding damaged rods):
Halogens                                            10%
Noble Gases                                          10%
Alkali Metals                                        12%
Activity Released to Coolant (melted rods):
Halogens                                              50%
Noble Gas                                            100%
Alkali Metals                                        25%
Percentage of Released Activity Reaching Condenser:
Halogens                                              10%
Noble Gases                                          100%
All Other Nuclides                                      1%
Percentage of Condenser Activity Available for Release to Environment:
Halogens                                              10%
Noble Gases                                          100%
All Other Nuclides                                      1%
 
Release Duration                                          24 hours Control Room Airspace (Free Volume)                      27,000 ft3 EFT System Operation                                      Not credited CR Outside Air Intake Rate (Normal Mode)                  7,440 cfm CR Envelope Unfiltered Inleakage Rate                    1,000 cfm Control Room Breathing Rate                              3.5E-04 m3/sec Control Room Occupancy Rate                              1.0
 
Offsite Breathing Rate:
0-8 hours                                            3.5E-04 m3/sec 8-24 hours                                          1.8E-04 m3/sec
 
SJAE and MVP Release Case
 
Steam Jet Air Ejector Flow Rate to Offgas Stack          360.5 scfm SJAE Release Holdup Time                                  17 minutes
 
Mechanical Vacuum Pump Flow Rate                          2,300 scfm Main Steam Line Radiation Monitor Setpoint                9 R/hr MVP Isolation Time                                        10 seconds MVP Release Holdup Time                                  0.38 minutes
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 80 of 239
 
Table 14.7-2a  CRDA Radiological Consequences Analysis Inputs and Assumptions
 
(Page 2 of 2)
 
Control Room X/Q, Elevated Release From Offgas Stack:
0-0.5 hr (fumigation)                                3.59E-4 sec/m3 0.5-2 hrs                                            4.06E-06 sec/m3 2-8 hrs                                              5.75E-07 sec/m3 8-24 hrs                                              2.24E-07 sec/m3 EAB X/Q, Elevated Release From Offgas Stack:
0-0.5 hr (fumigation)                                1.11E-4 sec/m3 0.5-2 hrs (used for accident duration)                4.22E-6 sec/m3
 
LPZ X/Q, Elevated Release From Offgas Stack:
0-0.5 hr (fumigation)                                3.86E-5 sec/m3 0.5-2 hrs                                            3.79E-6 sec/m3 2-8 hrs                                              2.14E-6 sec/m3 8-24 hrs                                              1.61E-6 sec/m3 Isolated Condenser Release Case
 
Mechanical Vacuum Pump Flow Rate                          2,300 scfm Main Steam Line Radiation Monitor Setpoint                9 R/hr MVP Isolation Time                                        10 seconds MVP Release Holdup Time                                    0.38 minutes Main Condenser Leak Rate (following MVP isolation)        1% per day
 
Control Room X/Q (Pre-MVP Trip), MVP Elevated Release From Offgas Stack (fumigation for 10 sec)                  3.59E-4 sec/m3 Offsite X/Q (Pre-MVP Trip), MVP Elevated Release From Offgas Stack:
EAB (fumigation for 10 seconds)                      1.11E-4 sec/m3 LPZ (fumigation for 10 seconds)                      3.86E-5 sec/m3
 
Control Room X/Q (Post-MVP Trip), Ground Level Release from Turbine Building Vent:
0-2 hrs                                              2.58E-3 sec/m3 2-8 hrs                                              1.85E-3 sec/m3 8-24 hrs                                              7.37E-4 sec/m3 EAB X/Q (Post-MVP Trip), Ground Level Release:
0-2 hrs                                              7.86E-4 sec/m3 2-8 hrs                                              5.08E-4 sec/m3 8-24 hrs                                              4.08E-4 sec/m3 LPZ X/Q (Post-MVP Trip), Ground Level Release:
0-2 hrs                                              1.53E-4 sec/m3 2-8 hrs                                              8.83E-5 sec/m3 8-24 hrs                                              6.71E-5 sec/m3
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 81 of 239
 
Table 14.7-2b    CRDA Dose Consequences (Rem TEDE)
 
Receptor                            Dose        AM 188        Regulatory Limit*
Dose SJAE and MVP Release Case:
Control Room Operator          1.88        1.89          5.0 EAB (2-hour)                    2.10        2.01          6.3 LPZ                            0.95        0.92          6.3
 
Isolated Condenser Release Case (MVP Operation with 10 Second Isolation):
Control Room Operator          0.61                      5.0 EAB (2-hour)                    0.21                      6.3 LPZ                            0.09                      6.3
*10CFR50.67 and RG 1.183
 
Note:  See References 13, 134 and 199 for additional information.
Am 188 dose is the dose most recently approved by the NRC.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 82 of 239
 
Table 14.7-3    Appendix K Assumption
 
Parameter                  ATRIUM 10XM                    ATRIUM 11
 
Decay Heat                    1971 ANS + 20%                1971 ANS + 20%
(Reference 100)              (Reference 100)
 
Transient boiling temperature  Transition boiling allowed    Transition boiling allowed during blowdown only until    during blowdown only until cladding superheat exceeds    cladding superheat exceeds 300&deg;F                        300&deg;F
 
Break flow                    Moody Slip Flow Model with    Moody Slip Flow Model with discharge coefficients of 1.0, discharge coefficients 0.8, 0.6, and 0.4            ranging from 0.4 to 1.0
 
Metal-Water Reaction          Baker-Just                    Baker-Just
 
Core Power                    2044 MWt (102% of 2004)      2044 MWt (102% of 2004)
 
Peak Average Planar Linear    12.5 X 1.02 kW/ft            10.0 X 1.02 kW/ft Heat Generation
 
Bypass Leakage Coefficients    Nominal Values                Nominal Values
 
Initial Operating Minimum      1.40                          1.41 Critical Power Ratio
 
ECCS Water Temperature        90&deg;F                          90&deg;F
 
ECCS System Performance        Per Plant Parameters          See tables 14.7-7, 14.7-8, and Single Failure Evaluation  Document (Reference 225)      14.7-9, 14.7-10, 14.7-11, and Inputs                                                      14.7-12
 
ECCS Available                Systems remaining after      Systems remaining after worst single failure          worst single failure
 
Stored Energy                  Conservatively calculated    Conservatively calculated using RODEX2                  using RODEX4 incorporated into S-RELAP5
 
Fuel Rod Internal Pressure    Conservatively calculated    Conservatively calculated using RODEX2                  using RODEX4 incorporated into S-RELAP5
 
Additional Information        207, 208, 225, and 226        3 and 10 References
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 83 of 239
 
Table 14.7-4    Maximum Break Areas
 
Break Location                Break Area (ft2)
 
Recirculation Suction Line            3.679
 
Core Spray Line                      Dispositioned
 
Feedwater Line                        Dispositioned
 
Steam Line (Inside Containment)      Dispositioned
 
Steam Line (Outside Containment)      Dispositioned
 
See References 3, 10, 207, and 208 for additional information.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                          USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                              Revision 40 Page 84 of 239
 
Table 14.7-5  Initial Conditions for LOCA Analyses
 
Steam Dome Power          Core Flow        Pressure (MWt)      (% of 57.6 Mlb/hr)    (psia)      Fuel Analyzed
 
2044      Minimum allowed      1038.7      ATRIUM 10XM
 
2044      Minimum allowed      1038.7      ATRIUM 10XM
 
1693.4      Minimum allowed      1007.5      ATRIUM 10XM
 
2044            105%              1048        ATRIUM 11
 
2044            80%              1048        ATRIUM 11
 
1886.8          70%              1035        ATRIUM 11
 
1349.1          52.5%            994          ATRIUM 11
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 85 of 239
 
Table 14.7-6  Fuel Parameters utilized in ECCS-LOCA Analysis
 
Fuel Parameter                            Value
 
MAPLHGR (kW/ft)                          12.5 x 1.02 (ATRIUM 10XM)
 
10.0 X 1.02 (ATRIUM 11)
 
Worse Case Pellet Exposure for                      NA ECCS Evaluation (MWd/MTU)
 
Initial Operating MCPR                      1.40 (ATRIUM 10XM)
 
1.41 (ATRIUM 11)
 
Number of Fuel Rods per Bundle              91 (ATRIUM 10XM)
 
112 (ATRIUM 11)
 
Axial Peaking Factor                Mid peak and top peak are analyzed
 
Note:  For additional information see References 10, 207, 208, 225, and 226.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 86 of 239
 
Table 14.7-7    High Pressure Coolant Injection System Parameters
 
Analysis Variable                                Units                      Value
: a. Operating pressure range
 
Maximum                                psid                      1120 (vessel to drywell)
 
Minimum                                psid                      150 (vessel to drywell)
: b. Minimum flow over the entire            gpm                        2700 pressure range above
: c. Initiating Signals
 
Low-low water level (Level 2)          inches                    422.1 (inches above vessel zero)
 
or
 
High drywell pressure                  psig                      3.0
: d. Maximum allowed delay time              sec                        45 from initiating signal to rated flow available and injection valve wide open
 
Note:  See References 10 and 225 for additional information.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 87 of 239
 
Table 14.7-8    Core Spray System Parameters
 
Variable                                  Units                    Analysis Value
: a. Maximum vessel pressure at which          psid                      279 pumps can inject flow (vessel to drywell)
: b. Minimum Flow into Reactor Core for        gpm                      2700 one CS loop at Vessel Pressure psid                      130
: c. Run-out flow at 0 psid (vessel to        gpm                      3700 drywell) for one CS pump
: d. Initiating Signals Low-low water level (Level 2) (inches    inches                    422.1 above vessel zero) and low vessel pressure                                  psig                      350 or High drywell pressure                    psig                      3
 
or Low-low water level (Level 2) sustained  minute                    > 20 for a time period of
: e. Timer setting for bypassing low reactor  minute                    < 20 pressure permissive in CS pump start                                (analytical limit) logic
: f. Maximum allowable delay time from        sec                      38 initiating signal to pump at rated speed and capable of rated flow Total system delay time from initiating signal until the system is ready to inject.
: g. CS injection valve Pressure at which CS injection valve      psig                      350 may be open CS injection valve stroke time            sec                      15
 
Note:  See References 10, 179, 193, 208, and 225 for additional information.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 88 of 239
 
Table 14.7-9    Low Pressure Coolant Injection System Parameters
 
(Page 1 of 2)
 
Variable                                Units                    Analysis Value
: a. Maximum vessel pressure at which        psid                      226 - 2 pump pumps can inject flow (vessel to drywell)                                                          222 - 4 pump
: b. Minimum Pump Flow into Reactor Core Vessel pressure below which listed      psid                      20 flow rates are quoted (vessel to drywell) 2 LPCI pumps operating                  gpm                      7740
 
3 LPCI pumps operating                  gpm
 
4 LPCI pumps operating                  gpm                      14184
: c. Run-out flow at 0 psid (vessel to drywell) 2 LPCI pumps operating                  gpm                      8038
 
3 LPCI pumps operating                  gpm
 
4 LPCI pumps operating                  gpm                      14733
: d. Initiating Signals Low-low water level (Level 2)          inches                    422.1 (inches above vessel zero) and Low vessel pressure                    psig                      350
 
or High drywell pressure                  psig                      3
 
Low-low water level (Level 2)          minute                    > 20 sustained for a time period of
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 89 of 239
 
Table 14.7-9  Low Pressure Coolant Injection System Parameters
 
(Page 2 of 2)
 
Variable                                Units                      Analysis Value
: e. Timer setting for bypassing low reactor  minute                    <20 pressure permissive in LPCI pump                                    (analytical limit) start logic
: f. Total system delay time from initiating  sec                        53.2 signal until the system is ready to inject
: g. LPCI injection valve Pressure at which LPCI injection valve  psig                      350 may be open LPCI injection valve stroke time        sec                        35(1)
: h. Recirculation discharge valve stroke    sec                        35 time
: i. Minimum recirculation break size        ft2                        0.4 assumed to be correctly detected by loop selection logic
 
(1) Rated LPCI flow to reactor vessel was assumed to occur at time LPCI injection value is greater than 50% open. In the ECCS-LOCA analysis, rated LPCI flow was assumed to occur at 35.0 seconds.
 
Note:  For additional information see References 10, 207, 208, 228, and 229.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 90 of 239
 
Table 14.7-10    Automatic Depressurization System Parameters
 
Variable                              Units                    Analysis Value
: a. Number of ADS valves
 
Total number of relief valves with ADS valves                    3 function
 
Number of ADS relief valves used in    valves                    3 analysis
: b. Minimum ADS flow rate
 
Minimum flow rate for one valve open  lb/hr                    829,000 (2 valves) at below listed pressure                                        800,000 (1 valve)
 
Vessel Pressure at which flow capacity psig                      1080 is quoted (vessel to suppression pool)
: c. Initiating Signals
 
Low-low water level (Level 2)          inches                    422.1 (inches above vessel zero) and Signal that at least 1 LPCI (pump      psig                      NA discharge pressure) or 1 LPCS pump is running (pump          psig                      NA discharge pressure) and ADS timer delay                        sec                      138
: d. Valve pressure setpoints
 
Vessel pressure below which ADS        psig                      NA valves close
 
Vessel pressure above which ADS        psig                      NA valves reopen
 
Note:  For additional information see References 10, 207, 208, 228, and 229.
Withheld Security-Related Information
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 91 of 239
 
Table 14.7-11  Single Failures and Available Systems
 
Note 1  Systems unavailable for each single failure are the same as the recirc suction break cases.
 
Note 2  Systems unavailable for each single failure are the same as the recirc suction break cases less the ECCS system in which the break occurs
 
Note:  See References 10 and 208 for additional information.
 
Withheld Security-Related Information MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 92 of 239
 
Table 14.7-12    ECCS Injection Timing Parameters Used in ECCS    -LOCA Analysis
 
Variable            Parameter                Value THDWS(1)            Delay Time to            0.0 sec to process high drywell pressure signal THDW(1)            Delay Time to Reach      0.0 sec High DW Press Signal After LOCA Initiation TDGS(1)            Delay Time to process    0.0 sec start sequence
 
TDG                D/G Start Time            15.0 sec
 
TCSPR              CS Pump Start Time        23.0 sec TCSPV              CS IV Sequencing            3.2 sec TCSIV              CS IV Stroke Time        15.0 sec
 
TCIPR              LPCI Pump Start Time      18.0 sec TCIPV              LPCI IV Sequencing          3.2 sec (2)
TCIIV              LPCI IV Stroke Time      69.0 sec
 
TPDV                Discharge Valve          3.2 sec Sequencing
 
TDV                Discharge Valve Stroke    35.0 sec Time
 
(1)  Delay times THDWS, THDW and TDGS are assumed to be included in the 15 sec, TDG, maximum delay time from EDG start signal until bus is at rated voltage.  (i.e., Both of these delay times were assumed to be 0.0 sec.)
 
(2)  Rated LPCI flow to reactor vessel was assumed to occur at time LPCI injection valve is greater than 50% open. In the analysis, rated LPCI flow was assumed to occur at 35.0 seconds.
 
Note:  See References 10 and 208 for additional information.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 93 of 239
 
Table 14.7-13  LOCA Radiological Consequences Analysis Inputs and Assumptions
 
(Page 1 of 3)
Core Power                                          2044 MWt (2004 MWt plus 2%)
Core Inventory at Accident Time T=0                  USAR Tables 14.7-24a, b, and c Release Onset                                        T= 2 minutes Gap Release Duration                                0.5 hours Gap Release Fractions:
Noble Gases                                    0.05 Halogens                                        0.05 Alkali Metals                                  0.05 Early In-Vessel Release Duration                    1.5 hours Early In-Vessel Release Fractions:
Noble Gases                                    0.95 Halogens                                        0.25 Alkali Metals                                  0.20 Tellurium Metals                                0.05 Ba, Sr                                          0.02 Noble Metals                                    0.0025 Cerium Group                                    0.0005 Lanthanides                                    0.0002 Standby Liquid Control Injection:
Completed By                                    T= 2 hours Final suppression pool pH                      Greater than 7 Drywell Natural Deposition                          Powers 10th Percentile Positive Pressure Period (PPP)                      5 minutes (starts at T=0)
Standby Gas Treatment System (SGTS):
Effective Filter Efficiency (Adsorber)          85%
Effective Filter Efficiency (Particulate)      98%
 
Primary-to-Secondary Containment Leakage Pathway Primary to Secondary Containment Leakage Rate (includes SCB Leakage, excludes MSIV leakage):
0-24 hours                                      1.2% per day by weight (La) 24-90 hours                                    66% of La 90 hrs - 30 days                                50% of La Release Point:
During PPP (3 minute release)                  Directly to environment After PPP (secondary containment negative)      Offgas stack via SGTS
 
ECCS Leakage Pathway ECCS Leakage Rate to Secondary Containment Design rate                                    1.31 gpm Analysis rate (design rate doubled)            2.62 gpm ECCS Leakage Radioiodine Flash Fraction              10%
Release Point:
During PPP (3 minute release)                  Directly to environment After PPP (secondary containment negative)      Offgas stack via SGTS
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 94 of 239
 
Table 14.7-13    LOCA Radiological Consequences Analysis Inputs and Assumptions
 
(Page 2 of 3)
MSIV/SCB Leakage Pathway MSIV Leakage Rate:
0-24 hours                                    200 scfh 24-90 hours                                    66% of 200 scfh 90 hrs - 30 days                              50% of 200 scfh SCB Leakage Rate:
0-24 hours                                    35.2 scfh 24-90 hours                                    66% of 35.2 scfh 90 hrs - 30 days                              50% of 35.2 scfh Main Condenser Leak Rate                            1% per day by weight Main Steam Line Radioiodine Deposition (Aerosol, Elemental, Organic)                      Well-Mixed Flow Model per RG 1.183 and AEB 98-03 Main Condenser Radioiodine Deposition (Aerosol and Elemental):
0-24 hours                                    98.62%
24-72 hours                                    99.09%
72 hrs - 30 days                              99.31%
Release Point                                      Turbine Building Vent
 
Control Room Airspace (Free Volume)                27,000 ft3 EFT System Operation:
Emergency Mode operating                      Prior to release onset at T= 2 min Filter Efficiency (Adsorber)                  98%
Filter Efficiency (Particulate)                98%
EFT Flow Rate                                  900 cfm Unfiltered Inleakage to EFT envelope          500 cfm Control Room Breathing Rate                        3.5E-04 m3/sec Control Room Occupancy Rate:
0-24 hours                                    1.0 1-4 days                                      0.6 4-30 days                                    0.4 Control Room X/Q, Ground Level Release from Turbine Building Vent (MSIV/SCB pathway):
0-2 hrs                                        2.58E-03 sec/m3 2-8 hrs                                        1.85E-03 sec/m3 8-24 hrs                                      7.37E-04 sec/m3 1-4 days                                        4.90E-04 sec/m3 4-30 days                                      3.84E-04 sec/m3 Control Room X/Q, Ground Level Release during PPP (Prim-Sec Cntmt and ECCS Leakage pathways):
0-2 hrs (Rx Bldg nearest wall)                1.43E-02 sec/m3
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 95 of 239
 
Table 14.7-13    LOCA Radiological Consequences Analysis Inputs and Assumptions
 
(Page 3 of 3)
Control Room X/Q, Elevated Release From Offgas Stack post-PPP (Prim-Sec Cntmt and ECCS Leakage pathways):
0-1.7 hrs                                      4.06E-06 sec/m3 1.7-2.2 hrs (fumigation)                      3.59E-04 sec/m3 2.2-8 hrs                                      5.75E-07 sec/m3 8-24 hrs                                      2.24E-07 sec/m3 1-4 days                                        2.90E-08 sec/m3 4-30 days                                      1.54E-09 sec/m3
 
Offsite Breathing Rate:
0-8 hours                                      3.5E-04 m3/sec 8-24 hours                                    1.8E-04 m3/sec 1-30 days                                      2.3E-04 m3/sec EAB X/Q, Ground Level Release (MSIV/SCB release; Prim-Sec Cntmt and ECCS Leakage release during PPP):
0-2 hours (used for accident duration)        7.86E-04 sec/m3 EAB X/Q, Elevated Release From Offgas Stack post-PPP (Prim-Sec Cntmt/ECCS release):
0-1.7 hrs                                      4.22E-06 sec/m3 1.7-2.2 hrs (fumigation)                      1.11E-04 sec/m3 2.2 hrs-30 days                                4.22E-06 sec/m3 LPZ X/Q, Ground Level Release (MSIV/SCB release; Prim-Sec Cntmt and ECCS Leakage release during PPP):
0-2 hrs                                        1.53E-04 sec/m3 2-8 hrs                                        8.83E-05 sec/m3 8-24 hrs                                      6.71E-05 sec/m3 1-4 days                                        3.70E-05 sec/m3 4-30 days                                      1.57E-05 sec/m3 LPZ X/Q, Elevated Release From Offgas Stack post-PPP (Prim-Sec Cntmt/ECCS release):
0-1.7 hrs                                      3.79E-06 sec/m3 1.7-2.2 hrs (fumigation)                      3.86E-05 sec/m3 2.2-8 hrs                                      2.14E-06 sec/m3 8-24 hrs                                      1.61E-06 sec/m3 1-4 days                                        8.64E-07 sec/m3 4-30 days                                      3.54E-07 sec/m3
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 96 of 239
 
Table 14.7-14    LOCA Dose Consequences (Rem TEDE)
Reference 153 and 154
 
Receptor                            Dose        AM 188        Regulatory Limit*
Dose Control Room Operator Internal (Inhalation) Dose      3.06        3.06 External (Shine) Dose          0.77        0.77 Total Dose                      3.83        3.83          5.0
 
EAB (2-hour)                    1.47        1.47          25 LPZ                            2.00        1.99          25
*10CFR50.67 and RG 1.183
 
Note:  See References 134, 153, 154 and 199 for additional information.
Am 188 dose is the dose most recently approved by the NRC.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 97 of 239
 
Table 14.7-15    Mass and Energy Release for Main Steamline    Break Outside Containment - 1880 MWt Power (See Note 1)
(Page 1 of 2)
 
Break    Integrated  Integrated Time,    Pressure,  Total Break  Break    Enthalpy,  Break    Break Enthalpy, Seconds    psia    Flow, lb/sec  Quality  BTU/lbm    Flow, lb      BTU
 
0.00      1025        4140      1.0000    1192      8.28      9.869E+03 0.25      1005        4058      1.0000    1193      1032      1.23 1E+06 0.50      987.3      3983      1.0000    1193      2037      2.430E+06 0.75      970.7      3914      1.0000    1194      3024      3.608E+06 1.00      955.0      3849      1.0000    1195      3994      4.767E+06 1.25      938.8      3782      1.0000    1195      4948      5.906E+06 1.50      923.5      3719      1.0000    1196      5886      7.027E+06 1.75      910.5      3665      1.0000    1196      6808      8.130E+06 2.00      898.4      3615      1.0000    1196      7718      9.219E+06 2.25      885.8      3564      1.0000    1197      8616      1.029E+07 2.50      873.3      3512      1.0000    1197      9500      1. 135E+07 2.75      863.3      3471      1.0000    1198    10370      1.240E+07 3.00      858.6      3452      1.0000    1198    11240      1.343E+07 3.25      853.7      3432      1.0000    1198    12100      1.446E+07 3.50      848.9      3412      1.0000    1198    12950      1.549E+07 3.75      844.0      3392      1.0000    1198    13800      1.651E+07 4.00      839.1      3372      1.0000    1198    14650      1.752E+07 4.25      834.1      3352      1.0000    1198    15490      1.853E+07 4.50      831.7      5611      0.2478    684.6    16630      1.949E+07 4.75      831.5      6122      0.1846    641.4    18110      2.047E+07 5.00      831.4      6199      0.1764    635.8    19650      2.145E+07 5.25      831.1      6123      0.1842    641.1    21200      2.243E+07 5.50      830.6      5972      0.1996    651.5    22710      2.341E+07 5.75      829.9      5835      0.2169    663.3    24180      2.438E+07 6.00      828.8      5704      0.2332    674.3    25630      2.535E+07 6.25      827.5      5591      0.2470    683.5    27040      2.630E+07 6.50      826.0      5515      0.2570    690.2    28420      2.726E+07 6.75      824.4      5464      0.2636    694.5    29800      2.821E+07 7.00      822.7      5429      0.2675    697.0    31160      2.915E+07 7.25      820.8      5405      0.2694    698.1    32510      3.010E+07 7.50      818.8      5391      0.2697    698.0    33860      3.104E+07 7.75      817.0      4898      0.2779    703.4    35150      3.194E+07 8.00      815.3      4338      0.2984    717.3    36300      3.276E+07 8.25      813.7      3779      0.3280    737.5    37320      3.350E+07 8.50      812.2      3224      0.3677    764.6    38190      3.416E+07 8.75      810.9      2683      0.4217    801.5    38930      3.473E+07
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 98 of 239
 
Table 14.7-15    Mass and Energy Release for Main Steamline    Break Outside Containment - 1880 MWt Power (Page 2 of 2)
 
Break  Integrated    Integrated Time,    Pressure,  Total Break  Break    Enthalpy,  Break    Break Enthalpy, Seconds    psia    Flow, lb/sec  Quality  BTU/lbm    Flow, lb      BTU 9.00      809.6      2157      0.4992    854.8    39540      3.523E+07 9.25      808.5      1652      0.6196    937.5    40010      3.565E+07 9.50      807.5      1174      0.8272    1080    40360      3.601E+07 9.75      806.7      812.1      1.000      1199    40600      3.629E+07 10.00      806.4      541.2      1.000      1199    40770      3.649E+07 10.25      806.8      270.7      1.000      1199    40870      3.661E+07 10.50      807.8      0.00      1.000      1199    40910      3.665E+07
 
Note 1:  Total mass and energy releases at 1800 MWt bound those at 2004 MWt.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 99 of 239
 
Table 14.7-16    Mass and Energy Release for Main Steamline Break Outside Containment - Hot Standby (965 psia)
(Page 1 of 3)
 
Break  Integrated    Integrated Time,  Pressure,  Total Break  Break    Enthalpy,  Break    Break Enthalpy, Seconds    psia    Flow, lb/sec  Quality  BTU/lbm    Flow, lb      BTU 0.0      965.0        0        1.000    1194        0        0.000E+00 0.1      957.8      3316        1.000    1194      333.1      3.978E+05 0.2      949.9      3288        1.000    1195      663.2      7.922E+05 0.3      946.7      3276        1.000    1195      991.4      1. 184E+06 0.4      943.7      3265        1.000    1195      1318      1.575E+06 0.5      940.6      3254        1.000    1195      1644      1.965E+06 0.6      937.5      3243        1.000    1195      1969      2.353E+06 0.7      934.5      3232        1.000    1195      2293      2.740E+06 0.8      931.6      3222        1.000    1195      2616      3.125E+06 0.9      928.6      3211        1.000    1195      2937      3.510E+06 1.0      925.7      3200        1.000    1196      3258      3.893E+06 1.1      922.7      3190        1.000    1196      3577      4.275E+06 1.2      919.8      3180        1.000    1196      3896      4.656E+06 1.3      916.9      3169        1.000    1196      4213      5.036E+06 1.4      914.4      6579      0.3291    748.5    4685      5.466E+06 1.5      912.5      8318      0.1959    659.5    5444      5.991E+06 1.6      911.0      9179      0.1476    627.0    6325      6.556E+06 1.7      909.6      9707      0.1224    610.0    7272      7.140E+06 1.8      908.4      10030      0.1070    599.6    8260      7.738E+06 1.9      907.2      10250    0.09674    592.5    9275      8.342E+06 2.0      906.1      10420    0.08947    587.5    10310      8.952E+06 2.1      905.0      10550    0.08420    583.8    11360      9.566E+06 2.2      903.9      10640    0.08026    581.0    12420      1.018E+07 2.3      902.9      10700    0.07731    578.9    13490      1.080E+07 2.4      901.9      10750    0.07507    577.3    14560      1.142E+07 2.5      900.9      10790    0.07340    576.0    15640      1.204E+07 2.6      899.9      10810    0.07215    575.0    16720      1.267E+07 2.7      898.9      10830    0.07128    574.3    17800      1.329E+07 2.8      897.9      10840    0.07069    573.7    18880      1.391E+07 2.9      896.9      10840    0.07033    573.3    19960      1.453E+07 3.0      896.0      10830    0.07017    573.1    21050      1.515E+07 3.1      895.0      10830    0.07017    573.0    22130      1.577E+07 3.2      894.0      10820    0.07032    572.9    23210      1.639E+07 3.3      893.0      10800    0.07058    572.9    24290      1.701E+07 3.4      892.0      10790    0.07095    573.0    25370      1.763E+07 3.5      891.0      10770    0.07141    573.2    26450      1.825E+07 3.6      890.0      10750    0.07194    573.4    27530      1.886E+07 3.7      889.0      10730    0.07254    573.7    28600      1.948E+07 3.8      888.0      10700    0.07320    573.9    29670      2.010E+07 3.9      887.0      10680    0.07392    574.3    30740      2.071E+07
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 100 of 239
 
Table 14.7-16    Mass and Energy Release for Main Steamline Break      Outside Containment - Hot Standby (965 psia)
(Page 2 of 3)
 
Break  Integrated  Integrated Time,  Pressure,  Total Break  Break    Enthalpy,  Break    Break Enthalpy, Seconds    psia    Flow, lb/sec  Quality  BTU/lbm    Flow, lb      BTU 4.0      886.0      10650    0.07469    574.6    31810      2.132E+07 4.1      885.0      10620    0.07550    575.0    32870      2.193E+07 4.2      884.0      10600    0.07634    575.4    33930      2.254E+07 4.3      882.9      10570    0.07724    575.9    34990      2.315E+07 4.4      881.9      10540    0.07817    576.4    36050      2.376E+07 4.5      880.8      10510    0.07913    576.9    37100      2.437E+07 4.6      879.8      10470    0.08010    577.3    38150      2.497E+07 4.7      878.7      10440    0.08112    577.9    39190      2.558E+07 4.8      877.7      10410    0.08217    578.4    40240      2.618E+07 4.9      876.6      10370    0.08323    579.0    41280      2.678E+07 5.0      875.5      10340    0.08432    579.6    42310      2.738E+07 5.1      874.4      10300    0.08541    580.1    43340      2.798E+07 5.2      873.3      10270    0.08653    580.7    44370      2.858E+07 5.3      872.2      10230    0.08769    581.3    45400      2.917E+07 5.4      871.1      10190    0.08884    581.9    46420      2.977E+07 5.5      869.9      10160    0.09007    582.6    47430      3.036E+07 5.6      868.8      10120    0.09123    583.2    48450      3.095E+07 5.7      867.7      10080    0.09248    583.9    49460      3.154E+07 5.8      866.6      10040    0.09374    584.6    50460      3.213E+07 5.9      865.4      10000    0.09498    585.2    51470      3.271E+07 6.0      864.3      9960      0.09629    586.0    52460      3.330E+07 6.1      863.1      9918      0.09763    586.7    53460      3.388E+07 6.2      861.9      9875      0.09900    587.5    54450      3.446E+07 6.3      860.7      9834      0.1004    588.2    55430      3.504E+07 6.4      859.5      9794      0.1019    589.0    56410      3.562E+07 6.5      858.3      9753      0.1034    589.9    57390      3.619E+07 6.6      857.0      9711      0.1049    590.8    58360      3.677E+07 6.7      855.8      9667      0.1066    591.7    59330      3.734E+07 6.8      854.6      9622      0.1082    592.6    60300      3.791E+07 6.9      853.3      9577      0.1099    593.6    61260      3.848E+07 7.0      852.1      9532      0.1116    594.6    62210      3.905E+07 7.1      850.8      9486      0.1134    595.6    63160      3.961E+07 7.2      849.5      9437      0.1152    596.6    64110      4.018E+07 7.3      848.2      9390      0.1170    597.7    65050      4.074E+07 7.4      846.9      9340      0.1189    598.8    65990      4.130E+07 7.5      845.6      9292      0.1207    599.8    66920      4.186E+07 7.6      844.3      8955      0.1229    601.1    67830      4.241E+07 7.7      843.0      8580      0.1258    602.9    68710      4.293E+07 7.8      841.8      8199      0.1293    605.1    69550      4.344E+07
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 101 of 239
 
Table 14.7-16    Mass and Energy Release for Main Steamline Break      Outside Containment - Hot Standby (965 psia)
(Page 3 of 3)
 
Break  Integrated    Integrated Time,  Pressure,  Total Break    Break  Enthalpy,  Break    Break Enthalpy, Seconds    psia    Flow, lb/sec  Quality  BTU/lbm    Flow, lb      BTU 8.0      839.5      7418      0.1383    610.9    71110      4.439E+07 8.1      838.3      7024      0.1436    614.4    71830      4.483E+07 8.2      837.3      6627      0.1496    618.3    72510      4.525E+07 8.3      836.2      6245      0.1564    622.7    73160      4.565E+07 8.4      835.2      5867      0.1636    627.6    73760      4.603E+07 8.5      834.2      5488      0.1717    633.0    74330      4.639E+07 8.6      833.2      5109      0.1807    639.0    74860      4.672E+07 8.7      832.3      4728      0.1910    645.9    75350      4.704E+07 8.8      831.4      4351      0.2027    653.7    75800      4.733E+07 8.9      830.6      3990      0.2158    662.6    76220      4.761E+07 9.0      829.8      3624      0.2315    673.2    76600      4.786E+07 9.1      829.0      3257      0.2498    685.6    76940      4.810E+07 9.2      828.2      2906      0.2720    700.7    77250      4.831E+07 9.3      827.5      2555      0.2981    718.5    77530      4.850E+07 9.4      826.9      2217      0.3308    740.8    77760      4.868E+07 9.5      826.2      1886      0.3719    768.9    77970      4.883E+07 9.6      825.7      1564      0.4265    806.1    78140      4.897E+07 9.7      825.2      1254      0.5010    857.1    78280      4.908E+07 9.8      824.7      961.3      0.6099    931.7    78390      4.918E+07 9.9      824.3      682.8      0.7912    1056      78470      4.926E+07 10.0      823.9      475.2      1.0000    1199      78530      4.933E+07 10.1      823.7      380.2      1.0000    1199      78570      4.938E+07 10.2      823.5      285.4      1.0000    1199      78600      4.942E+07 10.3      823.5      190.5      1.0000    1199      78630      4.945E+07 10.4      823.5      95.75      1.0000    1199      78640      4.946E+07 10.5      823.6      0.9482      1.0000    1199      78650      4.947E+07 10.52    823.7        0.0      1.0000    1199      78650      4.947E+07 7.9      840.6      7809      0.1335    607.8    70350      4.393E+07
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 102 of 239
 
Table 14.7-17    Mass and Energy Release for Main Steamline Break      Outside Containment - Hot Standby (1158 psia)
(Page 1 of 3)
 
Break  Integrated    Integrated Time,    Pressure,  Total Break  Break    Enthalpy,  Break    Break Enthalpy, Seconds    psia    Flow, lb/sec  Quality  BTU/lbm    Flow, lb      BTU 0.0      1158        0.0      1.0000    1187      0.0      0.000E+00 0.1      1148        4010      1.0000    1187      403.1      4.784E+05 0.2      1138        3973      1.0000    1187      802.2      9.522E+05 0.3      1133        3956      1.0000    1188      1199      1.423E+06 0.4      1128        3937      1.0000    1188      1593      1.892E+06 0.5      1122        3915      1.0000    1188      1986      2.358E+06 0.6      1117        3895      1.0000    1188      2376      2.822E+06 0.7      1111        3874      1.0000    1189      2765      3.284E+06 0.8      1106        3854      1.0000    1189      3151      3.743E+06 0.9      1100        3834      1.0000    1189      3535      4.200E+06 1.0      1095        3815      1.0000    1189      3918      4.655E+06 1.1      1091        6348      0.4638    849.9      4367      5.131E+06 1.2      1088        8617      0.2492    713.8      5135      5.715E+06 1.3      1086        9696      0.1818    670.8      6058      6.350E+06 1.4      1084      10310      0.1478    649.0      7061      7.011E+06 1.5      1082      10730      0.1274    635.9      8115      7.687E+06 1.6      1081      11010      0.1139    627.1      9203      8.374E+06 1.7      1079      11190      0.1046    620.9    10310      9.067E+06 1.8      1078      11340      0.0977    616.4    11440      9.764E+06 1.9      1076      11450      0.0927    613.0    12580      1.046E+07 2.0      1075      11530      0.0889    610.4    13730      1.117E+07 2.1      1073      11590      0.0860    608.4    14890      1.187E+07 2.2      1072      11630      0.0838    606.8    16050      1.258E+07 2.3      1070      11660      0.0822    605.6    17210      1.328E+07 2.4      1069      11680      0.0810    604.6    18380      1.399E+07 2.5      1068      11690      0.0802    603.9    19550      1.470E+07 2.6      1066      11700      0.0797    603.4    20720      1.540E+07 2.7      1065      11690      0.0795    603.1    21890      1.611 E+07 2.8      1064      11690      0.0794    602.9    23060      1.681E+07 2.9      1062      11680      0.0796    602.8    24220      1.752E+07 3.0      1061      11660      0.0799    602.8    25390      1.822E+07 3.1      1059      11640      0.0803    602.9    26560      1.892E+07 3.2      1058      11620      0.0808    603.0    27720      1.962E+07 3.3      1057      11600      0.0815    603.3    28880      2.032E+07 3.4      1055      11570      0.0822    603.5    30040      2.102E+07 3.5      1054      11540      0.0830    603.9    31190      2.172E+07 3.6      1053      11510      0.0839    604.2    32350      2.242E+07 3.7      1051      11480      0.0848    604.7    33500      2.311E+07 3.8      1050      11450      0.0858    605.1    34640      2.380E+07 3.9      1048      11410      0.0869    605.6    35790      2.450E+07 4.0      1047      11380      0.0880    606.1    36920      2.519E+07 4.1      1045      11340      0.0891    606.6    38060      2.588E+07 4.2      1044      11300      0.0903    607.2    39190      2.656E+07
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 103 of 239
 
Table 14.7-17    Mass and Energy Release for Main Steamline Break    Outside Containment - Hot Standby (1158 psia)
(Page 2 of 3)
 
Break    Integrated  Integrated Time,    Pressure,  Total Break  Break    Enthalpy,  Break  Break Enthalpy, Seconds    psia    Flow, lb/sec  Quality  BTU/lbm    Flow, lb      BTU 4.3      1042        11260      0.0915    607.8    40320      2.725E+07 4.4      1041        11230      0.0927    608.4    41450      2.793E+07 4.5      1039        11180      0.0940    609.0    42570      2.861E+07 4.6      1038        11140      0.0953    609.6    43680      2.929E+07 4.7      1036        11100      0.0967    610.3    44790      2.997E+07 4.8      1035        11060      0.0980    611.0    45900      3.065E+07 4.9      1033        11020      0.0994    611.7    47010      3.132E+07 5.0      1032        10980      0.1008    612.4    48110      3.200E+07 5.1      1030        10930      0.1022    613.1    49200      3.267E+07 5.2      1029        10890      0.1037    613.8    50290      3.334E+07 5.3      1027        10850      0.1052    614.6    51380      3.401E+07 5.4      1026        10810      0.1067    615.3    52460      3.467E+07 5.5      1024        10770      0.1082    616.1    53540      3.534E+07 5.6      1022        10720      0.1097    616.9    54620      3.600E+07 5.7      1021        10680      0.1113    617.7    55690      3.666E+07 5.8      1019        10640      0.1129    618.5    56750      3.732E+07 5.9      1017        10590      0.1145    619.3    57810      3.797E+07 6.0      1016        10550      0.1162    620.2    58870      3.863E+07 6.1      1014        10500      0.1178    621.0    59920      3.928E+07 6.2      1012        10450      0.1196    622.0    60970      3.993E+07 6.3      1011        10400      0.1214    623.0    62010      4.058E+07 6.4      1009        10350      0.1234    624.0    63050      4.123E+07 6.5      1007        10300      0.1254    625.1    64080      4.187E+07 6.6      1006        10240      0.1274    626.2    65110      4.252E+07 6.7      1004        10180      0.1296    627.3    66130      4.316E+07 6.8      1002        10130      0.1317    628.5    67150      4.379E+07 6.9      1000        10070      0.1339    629.7    68160      4.443E+07 7.0      998.7      10010      0.1362    630.9    69160      4.506E+07 7.1      996.9      9949      0.1385    632.2    70160      4.569E+07 7.2      995.1      9887      0.1408    633.5    71150      4.632E+07 7.3      993.3      9824      0.1432    634.8    72140      4.695E+07 7.4      991.5      9760      0.1456    636.2    73110      4.757E+07 7.5      989.6      9696      0.1480    637.5    74090      4.819E+07 7.6      987.8      9332      0.1509    639.2    75040      4.880E+07 7.7      986.0      8935      0.1547    641.4    75950      4.938E+07 7.8      984.3      8531      0.1593    644.2    76830      4.994E+07 7.9      982.6      8122      0.1646    647.5    77660      5.048E+07 8.0      9810        7709      0.1707    651.3    78450      5.099E+07 8.1      979.4      7294      0.1776    655.5    79200      5.148E+07 8.2      977.9      6877      0.1852    660.4    79910      5.195E+07 8.3      976.4      6458      0.1937    665.7    80570      5.239E+07 8.4      975.0      6044      0.2032    671.8    81200      5.281E+07
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 104 of 239
 
Table 14.7-17    Mass and Energy Release for Main Steamline Break      Outside Containment - Hot Standby (1158 psia)
(Page 3 of 3)
 
Break  Integrated    Integrated Time,  Pressure,  Total Break    Break  Enthalpy,  Break    Break Enthalpy, Seconds    psia    Flow, lb/sec  Quality  BTU/lbm    Flow, lb      BTU 8.5      973.5      5644      0.2137    678.5    81780      5.320E+07 8.6      972.2      5245      0.2253    686.0    82330      5.357E+07 8.7      970.9      4846      0.2383    694.3    82830      5.392E+07 8.8      969.7      4450      0.2529    703.8    83290      5.425E+07 8.9      968.5      4071      0.2697    714.7    83720      5.455E+07 9.0      967.3      3685      0.2895    727.6    84110      5.483E+07 9.1      966.2      3309      0.3129    742.8    84460      5.508E+07 9.2      965.2      2938      0.3407    761.0    84770      5.532E+07 9.3      964.2      2576      0.3747    783.2    85040      5.553E+07 9.4      963.3      2226      0.4156    810.1    85280      5.572E+07 9.5      962.4      1884      0.4685    844.8    85490      5.589E+07 9.6      961.7      1559      0.5354    888.8    85660      5.604E+07 9.7      961.0      1240      0.6339    953.5    85800      5.617E+07 9.8      960.3      939.4      0.7782    1048      85910      5.627E+07 9.9      959.7      667.5      1.0000    1194      85990      5.636E+07 10.0      959.3      556.1      1.0000    1194      86050      5.644E+07 10.1      959.0      445.0      1.0000    1194      86100      5.650E+07 10.2      958.8      333.9      1.0000    1194      86140      5.654E+07 10.3      958.7      223.0      1.0000    1194      86160      5.658E+07 10.4      958.7      112.0      1.0000    1194      86180      5.660E+07 10.5      958.9      1.110      1.0000    1194      86190      5.660E+07 10.5      959.0        0.0      1.0000    1194      86190      5.660E+07
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 105 of 239
 
Table 14.7-18    Mass Release from MSLBA - Hot Standby
 
Case 1        Case 2 Power level before accident (MWt)                                66.8          66.8
 
Initial reactor pressure (psia)                                1158            965
 
Total mass released through break (lbm)                      86,152        78,617
 
Total steam released through break (lbm)                      14,578        12,394
 
Total liquid released through break (lbm)                    71,574        66,223
 
Time for water level to cover steamline (sec)                    1.04          1.32
 
Initial steam released before steamline is covered (lbm)        4030          4243
 
Equivalent liquid released from break (lbm)                  82,203        74,459
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 106 of 239
 
Table 14.7-19  MSLBA Radiological Consequences Analysis Inputs and Assumptions
 
Power Level                                                66.8 MWt (Hot Standby)
Release Duration (MSIV Closure Time)                      10.5 seconds
 
Total Mass Release                                        91,834 lbm Liquid Mass Release                                  76,295 lbm Total Steam Mass Release                              15,540 lbm Initial Steam Mass Release (2% iodine carryover)      4,296 lbm Equivalent Mass Release (for Iodines)                      87,625 lbm
 
Iodine Concentration:
Equilibrium Case                                      0.2 Ci/g Dose Equivalent I-131 Pre-Accident Iodine Spike Case                        2.0 Ci/g Dose Equivalent I-131 Noble Gas Offgas Release Rate                              300,000 Ci/sec @30 min delay Noble Gas Offgas Release Fraction:
Kr-83m                                                9.36E-03 Kr-85m                                                1.64E-02 Kr-85                                                6.40E-05 Kr-87                                                5.11E-02 Kr-88                                                5.24E-02 Kr-89                                                2.18E-01 Xe-131m                                              5.23E-05 Xe-133m                                              7.82E-04 Xe-133                                                2.19E-02 Xe-135m                                              6.41E-02 Xe-135                                                5.92E-02 Xe-137                                                2.88E-01 Xe-138                                                2.18E-01
 
Normal Reactor Coolant Concentration (Ci/cc):
1-131                                                4.06E-03 1-132                                                1.78E-02 1-133                                                1.50E-02 1-134                                                3.83E-02 1-135                                                1.35E-02
 
Control Room Airspace (Free Volume)                        27,000 ft3 EFT System Operation                                      Not credited CR Outside Air Intake Rate (Normal Mode)                  7,440 cfm CR Envelope Unfiltered Inleakage Rate                      1,000 cfm Control Room X/Q, Ground Level Release from Turbine Building Vent (0-2 hr)                            2.58E-03 sec/m3 Control Room Breathing Rate                                3.5E-04 m3/sec Control Room Occupancy Rate                                1.0
 
Offsite X/Q, Ground Level Release EAB (0-2 hr)                                          7.86E-04 sec/m3 LPZ (0-2 hr)                                          1.53E-04 sec/m3 Offsite Breathing Rate                                    3.5E-04 m3/sec
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 107 of 239
 
Table 14.7-20  MSLBA Dose Consequences (Rem TEDE)
 
Receptor                              Dose                      Regulatory Limit*
 
Equilibrium Case Control Room Operator            0.33                                  5.0 EAB (2-hour)                    0.11                                  2.5 LPZ                              0.02                                  2.5
 
Pre-Accident Iodine Spike Case Control Room Operator            3.25                                  5.0 EAB (2-hour)                    1.05                                  25 LPZ                              0.20                                  25
 
*10CFR50.67 and RG 1.183
 
Note:  See Reference 134 for additional information.
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 108 of 239
 
Table 14.7-21    FHA Radiological Consequences Analysis Inputs and Assumptio  ns
 
Core Power (for establishing isotopic inventory)        2044 MWt (2004 MWt plus 2%)
Limiting Accident Location                              Reactor cavity Fuel Damage for Limiting Accident                        125 rods of equivalent 8x8 fuel Radial Peaking factor                                    1.7 Decay time (time since reactor shutdown)                24 hours Core inventory (Ci/MWt) at T=0 (reactor shutdown)        USAR Tables 14.7-24b, d, and e Percent of Activity Released from Damaged Rods:
I-131                                                  8 Kr-85                                                10 Other Noble Gases                                      5 Other Halogens                                        5 Water depth over damaged fuel                            >23 feet Overall Iodine Decontamination Factor (DF)              200 Release Duration                                        2 hours Secondary Containment, SBGT System Operation            Not credited Control Room Airspace (Free Volume)                      27,000 ft3 EFT System Operation                                    Not credited CR Outside Air Intake Rate (Normal Mode)                7,440 cfm CR Envelope Unfiltered Inleakage Rate                    1,000 cfm Control Room X/Q, Ground Level Release from Reactor Building Vent (0-2 hr)                          2.48E-03 sec/m3 Control Room Breathing Rate                              3.5E-04 m3/sec Control Room Occupancy Rate                              1.0
 
Offsite X/Q, Ground Level Release:
EAB (0-2 hr)                                        7.86E-04 sec/m3 LPZ (0-2 hr)                                        1.53E-04 sec/m3 Offsite Breathing Rate                                  3.5E-04 m3/sec MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 109 of 239
 
Table 14.7-22  FHA Dose Consequences (Rem TEDE) Reference 175
 
Receptor                        Dose        AM 188 Dose      Regulatory Limit*
Control Room Operator          4.53        4.67            5.0 EAB (2-hour)              1.76        1.74            6.3 LPZ                        0.34        0.34            6.3
*10CFR50.67 and RG 1.183
 
Note:  See References 134, 175, and 199 for additional information.
AM 188 dose is the dose most recently approved by the NRC.
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 110 of 239
 
Table 14.7-23    Atmospheric Dispersion Factors (X/Q) for Accident Analysis  (sec/m3)
 
(Page 1 of 2)
 
Control Room                                  Control Room              Admin Bldg Intake                    Intake
 
Elevated Release - Offgas Stack:
Fumigation                                3.37E-04                  3.59E-04*
0-2 hrs                                  3.77E-06                  4.06E-06*
2-8 hrs                                  5.74E-07                  5.75E-07*
8-24 hrs                                  2.24E-07*                  2.17E-07 1-4 days                                  2.90E-08*                  2.60E-08 4-30 days                                1.54E-09*                  1.24E-09 Ground Level Release - Turbine Building Vent:
0-2 hrs                                  2.51E-03                  2.58E-03*
2-8 hrs                                  1.73E-03                  1.85E-03*
8-24 hrs                                  6.86E-04                  7.37E-04*
1-4 days                                  4.70E-04                  4.90E-04*
4-30 days                                3.52E-04                  3.84E-04*
Ground Level Release - Reactor Building Vent:
0-2 hrs                                  2.48E-03*                  2.47E-03 2-8 hrs                                  1.81E-03*                  1.76E-03 8-24 hrs                                  6.58E-04*                  6.31E-04 1-4 days                                  4.67E-04*                  4.57E-04 4-30 days                                3.49E-04*                  3.41E-04 Ground Level Release - Reactor Building Nearest Wall to CR Intake (used for LOCA Positive Pressure Period):
0-2 hrs                                  1.00E-02                  1.43E-02*
2-8 hrs                                  7.09E-03                  9.69E-03*
8-24 hrs                                  2.75E-03                  3.82E-03*
1-4 days                                  1.90E-03                  2.65E-03*
4-30 days                                1.42E-03                  1.98E-03*
 
*Bounding receptor for use in radiological consequences analyses.
 
Offsite - Exclusion Area Boundary (EAB)
Elevated Release - Offgas Stack:
Fumigation                                1.11E-04 0-2 hrs                                  4.22E-06 2-8 hrs                                  2.23E-06 8-24 hrs                                  1.67E-06 1-4 days                                  7.88E-07 4-30 days                                3.11E-07 Ground Level Release:
0-2 hrs                                  7.86E-04 2-8 hrs                                  5.08E-04 8-24 hrs                                  4.08E-04 1-4 days                                  2.54E-04 4-30 days                                1.29E-04
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                      Revision 40 Page 111 of 239 Table 14.7-23    Atmospheric Dispersion Factors (X/Q) for Accident Analysis (sec/m3)
(Page 2 of 2)
 
Offsite - Low Population Zone (LPZ)
Elevated Release - Offgas Stack:
Fumigation                                3.86E-05 0-2 hrs                                  3.79E-06 2-8 hrs                                  2.14E-06 8-24 hrs                                  1.61E-06 1-4 days                                  8.64E-07 4-30 days                                3.54E-07 Ground Level Release 0-2 hrs                                    1.53E-04 2-8 hrs                                    8.83E-05 8-24 hrs                                  6.71E-05 1-4 days                                  3.70E-05 4-30 days                                  1.57E-05
 
Note:  See References 177 and 178 for additional information.
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                          USAR-14 SECTION 14    PLANT SAFETY ANALYSIS                              Revision 40 Page 112 of 239
 
Table 14.7-24a Core Inventory for GE14 Fuel at 35 GWD/MT Exposure @ T = 0 Hours in Ci/MWt (Reference 167)
 
Nuclide    Activity        Nuclide      Activity        Nuclide    Activity Co-58      1.379E+02        Ru-103      4.049E+04        Cs-136      1.863E+03 Co-60      1.329E+02        Ru-105      2.708E+04        Cs-137      3.470E+03 Kr-85      3.327E+02        Ru-106      1.409E+04        Ba-139      4.965E+04 Kr-85m    7.383E+03        Rh-105      2.461E+04        Ba-140      4.774E+04 Kr-87      1.424E+04        Sb-127      2.795E+03        La-140      4.915E+04 Kr-88      2.005E+04        Sb-129      8.518E+03        La-141      4.530E+04 Rb-86      6.346E+01        Te-127      2.838E+03        La-142      4.388E+04 Sr-89      2.684E+04        Te-127m      3.703E+02        Ce-141      4.534E+04 Sr-90      2.637E+03        Te-129      8.381E+03        Ce-143      4.228E+04 Sr-91      3.365E+04        Te-129m      1.243E+03        Ce-144      3.682E+04 Sr-92      3.621E+04        Te-131m      3.842E+03        Pr-143      4.134E+04 Y-90      2.805E+03        Te-132      3.817E+04        Nd-147      1.807E+04 Y-91      3.439E+04        I-131        2.677E+04        Np-239      5.223E+05 Y-92      3.636E+04        I-132        3.896E+04        Pu-238      9.040E+01 Y-93      4.177E+04        I-133        5.513E+04        Pu-239      1.086E+01 Zr-95      4.851E+04        I-134        6.087E+04        Pu-240      1.408E+01 Zr-97      4.993E+04        I-135        5.174E+04        Pu-241      4.092E+03 Nb-95      4.869E+04        Xe-133      5.478E+04        Am-241      4.610E+00 Mo-99      5.124E+04        Xe-135      2.532E+04        Cm-242      1.085E+03 Tc-99m    4.537E+04        Cs-134      5.346E+03        Cm-244      5.238E+01 NOTE:  Data are from Reference 163 with the exception of Co-58 and Co-60 which were obtained from the BWR default source term values from Table 1.4.3.2-3 of Reference 166.
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                            USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                Revision 40 Page 113 of 239
 
Table 14.7-24b  Core Inventory for ATRIUM 10XM Fuel @ T = 0 Hours in Ci/MWt (Reference 172)
 
Nuclide      Activity      Nuclide      Activity        Nuclide    Activity Co-58        1.38E+02      Ru-103      4.08E+04        Cs-136      1.52E+03 Co-60        1.33E+02      Ru-105      2.70E+04        Cs-137      4.99E+03 Kr-85        4.96E+02      Ru-106      1.50E+04        Ba-139      4.94E+04 Kr-85m      7.33E+03      Rh-105      2.54E+04        Ba-140      4.79E+04 Kr-87        1.45E+04      Sb-127      2.35E+03        La-140      4.99E+04 Kr-88        1.95E+04      Sb-129      7.30E+03        La-141      4.48E+04 Rb-86        4.51E+01      Te-127      2.32E+03        La-142      4.34E+04 Sr-89        2.70E+04      Te-127m      3.96E+02        Ce-141      4.53E+04 Sr-90        3.98E+03      Te-129      6.83E+03        Ce-143      4.24E+04 Sr-91        3.40E+04      Te-129m      1.32E+03        Ce-144      3.95E+04 Sr-92        3.62E+04      Te-131m      5.05E+03        Pr-143      4.14E+04 Y-90        4.10E+03      Te-132      3.81E+04        Nd-147      1.80E+04 Y-91        3.51E+04      I-131        2.67E+04        Np-239      4.56E+05 Y-92        3.66E+04      I-132        3.90E+04        Pu-238      8.43E+01 Y-93        4.10E+04      I-133        5.53E+04        Pu-239      1.15E+01 Zr-95        4.65E+04      I-134        6.24E+04        Pu-240      1.87E+01 Zr-97        4.62E+04      I-135        5.26E+04        Pu-241      4.08E+03 Nb-95        4.68E+04      Xe-133      5.42E+04        Am-241      6.16E+00 Mo-99        5.03E+04      Xe-135      2.44E+04        Cm-242      1.31E+03 Tc-99m      4.45E+04      Cs-134      4.95E+03        Cm-244      4.13E+01 NOTE:  Data are from Reference 172 with the exception of Co-58 and Co-60 which were obtained from the BWR default source term values from Table 1.4.3.2-3 of Reference 166.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                          USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                              Revision 40 Page 114 of 239
 
Table 14.7-24c Core Inventory for GE14 Fuel at 37 GWD/MT Exposure @ T = 0 Hours in Ci/MWt (Reference 173)
 
Nuclide    Activity        Nuclide      Activity      Nuclide    Activity Co-58      1.38E+02        Ru-103      4.08E+04      Cs-136      1.52E+03 Co-60      1.33E+02        Ru-105      2.70E+04      Cs-137      4.99E+03 Kr-85      4.96E+02        Ru-106      1.50E+04      Ba-139      4.94E+04 Kr-85m      7.33E+03        Rh-105      2.54E+04      Ba-140      4.79E+04 Kr-87      1.45E+04        Sb-127      2.35E+03      La-140      4.99E+04 Kr-88      1.95E+04        Sb-129      7.30E+03      La-141      4.48E+04 Rb-86      4.51E+01        Te-127      2.32E+03      La-142      4.34E+04 Sr-89      2.70E+04        Te-127m      3.96E+02      Ce-141      4.53E+04 Sr-90      3.98E+03        Te-129      6.83E+03      Ce-143      4.24E+04 Sr-91      3.40E+04        Te-129m      1.32E+03      Ce-144      3.95E+04 Sr-92      3.62E+04        Te-131m      5.05E+03      Pr-143      4.14E+04 Y-90        4.10E+03        Te-132      3.81E+04      Nd-147      1.80E+04 Y-91        3.51E+04        I-131        2.67E+04      Np-239      4.56E+05 Y-92        3.66E+04        I-132        3.90E+04      Pu-238      8.43E+01 Y-93        4.10E+04        I-133        5.53E+04      Pu-239      1.15E+01 Zr-95      4.65E+04        I-134        6.24E+04      Pu-240      1.87E+01 Zr-97      4.62E+04        I-135        5.26E+04      Pu-241      4.08E+03 Nb-95      4.68E+04        Xe-133      5.42E+04      Am-241      6.16E+00 Mo-99      5.03E+04        Xe-135      2.44E+04      Cm-242      1.31E+03 Tc-99m      4.45E+04        Cs-134      4.95E+03      Cm-244      4.13E+01 NOTE:  Data are from Reference 173 with the exception of Co-58 and Co-60 which were obtained from the BWR default source term values from Table 1.4.3.2-3 of Reference 166.
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                          USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                              Revision 40 Page 115 of 239
 
Table 14.7-24d  Core Inventory for ATRIUM 10XM Fuel at 41 GWD/MT Exposure
                              @ T = 0 Hours in Ci/MWt
 
Isotope    Activity        Isotope      Activity      Isotope    Activity Co-58      1.38E+02        Ru-103      4.20E+04      Cs-136      1.71E+03 Co-60      1.33E+02        Ru-105      2.87E+04      Cs-137      5.48E+03 Kr-85      5.30E+02        Ru-106      1.65E+04      Ba-139      4.91E+04 Kr-85m      7.03E+03        Rh-105      2.69E+04      Ba-140      4.75E+04 Kr-87      1.39E+04        Sb-127      2.42E+03      La-140      5.04E+04 Kr-88      1.86E+04        Sb-129      7.50E+03      La-141      4.44E+04 Rb-86      5.13E+01        Te-127      2.40E+03      La-142      4.29E+04 Sr-89      2.57E+04        Te-127m      4.13E+02      Ce-141      4.48E+04 Sr-90      4.29E+03        Te-129      7.03E+03      Ce-143      4.17E+04 Sr-91      3.26E+04        Te-129m      1.35E+03      Ce-144      3.90E+04 Sr-92      3.49E+04        Te-131m      5.17E+03      Pr-143      4.05E+04 Y-90        4.47E+03        Te-132      3.83E+04      Nd-147      1.79E+04 Y-91        3.36E+04        I-131        2.68E+04      Np-239      4.79E+05 Y-92        3.53E+04        I-132        3.92E+04      Pu-238      1.07E+02 Y-93        3.98E+04        I-133        5.52E+04      Pu-239      1.13E+01 Zr-95      4.79E+04        I-134        6.21E+04      Pu-240      2.08E+01 Zr-97      4.94E+04        I-135        5.26E+04      Pu-241      4.36E+03 Nb-95      4.81E+04        Xe-133      5.35E+04      Am-241      6.99E+00 Mo-99      5.01E+04        Xe-135      1.49E+04      Cm-242      1.67E+03 Tc-99m      4.44E+04        Cs-134      5.86E+03      Cm-244      6.73E+01
 
NOTE:  Data is from Reference 27 with the exception of Co-58 and Co-60 which were obtained from the BWR default source term values from Table 1.4.3.2-3 of Reference 166.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                          USAR-14 SECTION 14    PLANT SAFETY ANALYSIS                              Revision 40 Page 116 of 239
 
Table 14.7-24e  Core Inventory for ATRIUM 11 Fuel at 41 GWD/MT Exposure
                              @ T = 0 Hours in Ci/MWt
 
Isotope    Activity        Isotope      Activity      Isotope    Activity Co-58      1.38E+02        Ru-103      4.31E+04      Cs-136      1.74E+03 Co-60      1.33E+02        Ru-105      3.01E+04      Cs-137      5.47E+03 Kr-85      5.20E+02        Ru-106      1.75E+04      Ba-139      4.87E+04 Kr-85m      6.77E+03        Rh-105      2.80E+04      Ba-140      4.71E+04 Kr-87      1.33E+04        Sb-127      2.49E+03      La-140      5.01E+04 Kr-88      1.78E+04        Sb-129      7.68E+03      La-141      4.40E+04 Rb-86      5.28E+01        Te-127      2.46E+03      La-142      4.24E+04 Sr-89      2.47E+04        Te-127m      4.24E+02      Ce-141      4.45E+04 Sr-90      4.20E+03        Te-129      7.20E+03      Ce-143      4.11E+04 Sr-91      3.15E+04        Te-129m      1.38E+03      Ce-144      3.85E+04 Sr-92      3.38E+04        Te-131m      5.27E+03      Pr-143      4.00E+04 Y-90        4.39E+03        Te-132      3.84E+04      Nd-147      1.78E+04 Y-91        3.25E+04        I-131        2.69E+04      Np-239      5.07E+05 Y-92        3.42E+04        I-132        3.93E+04      Pu-238      1.11E+02 Y-93        3.89E+04        I-133        5.51E+04      Pu-239      1.10E+01 Zr-95      4.73E+04        I-134        6.18E+04      Pu-240      2.15E+01 Zr-97      4.91E+04        I-135        5.25E+04      Pu-241      4.40E+03 Nb-95      4.75E+04        Xe-133      5.34E+04      Am-241      6.82E+00 Mo-99      5.00E+04        Xe-135      1.38E+04      Cm-242      1.78E+03 Tc-99m      4.43E+04        Cs-134      6.07E+03      Cm-244      8.04E+01
 
NOTE:  Data is from Reference 27 with the exception of Co-58 and Co-60 which were obtained from the BWR default source term values from Table 1.4.3.2-3 of Reference 166.
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 117 of 239
 
14.8 Anticipated Transients Without Scram (ATWS)
 
14.8.1 General
 
ATWS was not considered in the original design or licensing basis of the Monticello plant and was not addressed in the Final Safety Analysis Report (FSAR).
 
Anticipated Transients Without Scram (ATWS) events were first identified by the Atomic Energy Commission as a safety issue shortly before a Provisional Operating License was issued to Monticello. In 1969, a consultant for the Advisory Committee on Reactor Safeguards (ACRS) pointed out that a common mode failure in the reactor protection system could prevent an automatic scram of the reactor following a plant transient.
 
In 1973, the AEC staff published WASH-1270 (Reference 101), a technical report on ATWS for water cooled power reactors, which established their position on ATWS.
Subsequently, the NSSS vendors developed methods for analyzing ATWS events.
 
At the Monticello Atomic Safety and Licensing Board (ASLB) Full Term Operating License hearings in May, 1975, information related to ATWS issues was presented. In an effort to close the ATWS issue and obtain a Full Term Operating License, Northern States Power Company agreed to install a Recirculation Pump Trip (RPT) System and an Alternate Rod Injection (ARI) System. The NRC reviewed and approved the proposed ATWS modifications in a letter and safety evaluation dated February 23, 1977 (Reference 102).
 
14.8.1.1  Final ATWS Rule
 
For the industry as a whole, the concerns related to ATWS required approximately 15 years to reach final resolution. The ATWS issue was resolved with a Rule issued by the Nuclear Regulatory Commission in 1984. The Final ATWS Rule, 10CFR50.62, was prescriptive in nature. The Rule directed that a number of modifications be made based on reactor type. Completion of these modifications was deemed by the Commission to provide the required level of plant protection for ATWS events.
 
For Boiling Water Reactors, the Final ATWS Rule required:
: a. An Alternate Rod Injection (ARI) system, diverse from the reactor protection system, to vent the scram air header automatically under ATWS conditions.
: b. A Recirculation Pump Trip (RPT) system to trip the reactor recirculation pumps automatically under ATWS conditions.
: c. A Standby Liquid Control System (SLCS) with the capability of inserting negative reactivity equivalent to 86 gpm of 13 weight percent of natural sodium pentaborate decahydrate solution into a 251-inch inside diameter reactor vessel.
 
Clarification of design features and quality assurance requirements for these modifications was provided in additional guidance issued by the NRC staff in 1985.
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 118 of 239
 
Based on agreements reached with the NRC staff to resolve Full Term Operating License open items, the ARI and RPT systems at Monticello were installed prior to the Final ATWS Rule. The NRC adopted a different position concerning ARI diversity requirements with the Final ATWS Rule, however, which the Monticello installation did not fully meet. The NRC later concluded that further modifications were not required at Monticello due to backfit considerations (References 68, 69, and 70).
 
The capacity of the SLCS at Monticello was upgraded by increasing the concentration of Boron-10 in the SLCS tank in accordance with the Final ATWS Rule.
SLCS related changes implemented at Monticello were reviewed and approved by the NRC (References 117 and 118).
 
Satisfaction of the Final ATWS Rule was confirmed by the NRC staff and the Monticello Technical Specifications were amended to include limiting conditions for operation and surveillance requirements for the required ATWS mitigation features.
 
Refer to Section 7.6.2 for a description of the RPT and ARI system. Refer to Section 6.6.1 for a description of SLCS compliance with the Final ATWS Rule.
 
14.8.1.2  Programmatic Issues
 
In 1983, both scram breakers failed to automatically open at Unit 1 of the Salem Nuclear Power Plant when an automatic reactor trip signal was received. This was considered to be an actual ATWS event.
 
A detailed NRC investigation of this event resulted in the issue of NRC Generic Letter 83-28 (Reference 103). This letter required licensees to make a number of programmatic improvements in reactor protection system reliability and general management. Improvements were specified in the following areas.
: a. Post-Trip Review
: b. Equipment Classification and Vendor Interface
: c. Post-Maintenance Testing
: d. Reactor Trip System Reliability Improvements
 
A number of improvements to satisfy the requirements of the General Letter were made at Monticello and found acceptable by the NRC staff.
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 119 of 239
 
14.8.2 Evaluation of Events
 
MNGP meets the ATWS requirements defined in 10 CFR 50.62 because:
* An Alternate Rod Injection (ARI) system is installed.
* Standby Liquid Control (SLC) system's automatic boron injection capability is equivalent to the control provided by 86 gpm of 13 wt% sodium pentaborate decahydrate solution.
* Reactor recirculation pump control logic automatically trips pumps (RPT) under conditions indicative of an ATWS event.
 
In addition, ATWS event analyses were performed to validate compliance with the ATWS acceptance criteria below. Three ATWS analyses were performed:  ATWS licensing basis analysis, ATWS with depressurization analysis, and ATWS with core instability (ATWSI) analysis. These analyses take credit for SLC and RPT but not ARI and ensure that the following ATWS acceptance criteria were met (References 7, 108, 134, 160, 182, 184, 188, 199, and 204):
 
Event mitigation is consistent with emergency procedure guidelines/severe accident guidelines (EPGs/SAGs)
* The potential for thermal-hydraulic instability is mitigated
* The peak vessel bottom pressure is less than the ASME Service Level C limit of 1500 psig;
* The peak clad temperature is within the 10 CFR 50.46 limit of 2200&deg;F
* The peak suppression pool temperature is less than the design limit
* The peak containment pressure is less than the containment design pressure
* Sufficient margin is available in the setpoint for the SLC system pump discharge relief valve such that SLC system operability is not affected by a postulated ATWS event (See USAR section 6.6.1.4 for further discussion)
 
The emergency operating procedures (EOPs) follow EPGs for mitigation of an ATWS event. Inputs, assumptions, and mitigation sequences used in the analyses are described in References 108 and 188, for EPU and MELLLA+, respectively. The NRC approved the MNGP ATWS mitigation strategy, event analysis, and the generic disposition of peak cladding temperature (PCT) and local cladding oxidation for EPU and MELLLA+, as applicable, with issuance of License Amendments 176 (Reference 134) and 180 (Reference 184), respectively.
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 120 of 239
 
14.8.2.1  ATWS Licensing Basis Analysis
 
The limiting events required to be evaluated for the licensing basis ATWS analysis are (Reference 188 for MELLLA+):
* Main steam isolation valve closure (MSIVC)
* Pressure regulator failure open (PRFO)
* Loss of offsite power (LOOP) (containment analysis only)
 
Using GNF methods, the reactor transient analysis of these events was performed using the approved ODYN methodology documented in Reference 190. ODYN calculates the peak vessel pressure used in non-vessel evaluations such as SBLC relief valve simmer margin (see section 6.6.1.4) and input to containment analysis.
Using Framatome methods, the reactor transient analysis was performed using the AURORA-B method documented in Reference 20 for peak vessel pressure compliance with ASME service level C limits. The STEMP model was used for the suppression pool heatup analysis. STEMP is not used for NPSH analysis as the containment pressure is non-conservative. (See USAR section 5.2.3.3 for evaluation of ECCS pump NPSH during ATWS event). As described in References    7, 206, and 207 and approved by the NRC in References 189, 199, and 5 respectively, the containment analyses are unaffected by the introduction of ATRIUM 10XM or ATRIUM 11 fuel and the containment analyses in Reference 108 and 188 remain the analyses of record. RHR and RHRSW pumps are assumed to operate in suppression pool cooling mode for these events. A loss of offsite power (LOOP) reduces the number of pumps available and thus the RHR heat exchanger effectiveness while in this mode.
 
The key operator actions credited in the licensing basis ATWS analysis, which are consistent with the EOPs, include:
* Manual FW flow reduction at 90 seconds following the start of MSIV closure.
The FW flow reduction from 100% to 0% rated flow occurs in 15 seconds.
* Water level control at top of active fuel (TAF) plus 5 feet due to limitations of ODYN code. (ODYN limitation and 5 foot adder only applicable for GNF analysis)
* Initiation of SLC system boron injection at 120 seconds following the high pressure ATWS RPT signal.
* Initiation of RHR suppression pool cooling at 600 seconds into the ATWS event.
 
The containment analysis is fuel independent per References 7, 206, and 207. The limiting peak suppression pool temperature occurs when the event starts from 80%
core flow (i.e., MELLLA+/EFW minimum flow) and results in peak suppression pool temperature of 197&deg;F, which is below the limit of 281&deg;F (Reference 188). The limiting peak containment pressure is also limiting from 80% core flow and results in a peak containment pressure of 13.6 psig, which is below the limit of 56.0 psig (Reference 188).
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 121 of 239
 
The overpressure analysis is performed each cycle.  (Reference 6) Representative analysis was presented and subsequently approved by the NRC as part of License Amendment 210. The limiting peak vessel pressure occurs when the event starts from 80% core flow (i.e., MELLLA+/EFW) and occurs in the vessel lower head. The representative analysis resulted in a peak pressure of 1462 psig (Reference 6).
 
Coolable core geometry is assured by meeting the 2200&#xba;F PCT and the 17% local cladding oxidation acceptance criteria of 10 CFR 50.46 (Reference 157). Previous ATWS analyses used to support generic assessments of ATWS have demonstrated that there is significant margin to the acceptance criteria of 10 CFR 50.46. The calculated PCTs for ATWS events have been consistently less than 1500&#xba;F. If the fuel temperature remains below 1600&#xba;F, cladding oxidation is insignificant compared to the acceptance criteria. This criteria is therefore met with no further analysis. The local fuel conditions are not changed with operation at increased power levels assumed for the current ATWS event analyses because the hot bundle operation is still constrained by the same operating thermal limits. Because the average channel power increases with EPU, the fraction of the flow passing through the hot channel increases. The increased flow keeps the peak cladding temperature and local oxidation from increasing with EPU. The peak clad temperature was not calculated during the transition to Framatome fuel because the results are bounded by the LOCA analysis, which includes a longer core uncovery phase.
 
Only in the ODYN MELLLA+ region analysis, which is no longer credited for peak vessel pressure, all SRVs must be in service to comply with the ASME service level C pressure limit. Framatome analysis in both the EFW and MELLLA regions demonstrates compliance with the ASME service level C pressure limit with one SRV out of service (Reference 12). The limitation to have all SRVs in service while operating above the MELLLA line (i.e., in the MELLLA+/EFW region) is retained in order to preserve the ODYN analysis results that were used for other evaluations.
 
The results of the licensing basis ODYN ATWS analysis meet the ATWS acceptance criteria. Therefore, the Monticello response to an ATWS event when initiated in any operating domain is acceptable.
 
14.8.2.2  ATWS With Depressurization Analysis
 
Monticello Emergency Operating Procedures (EOPs) require depressurization during an ATWS event when the suppression pool temperature reaches the heat capacity temperature limit (HCTL). A best estimate ATWS analysis, using TRACG04 methodology with input data from TGBLA06/PANAC11, was therefore performed as required by Reference 191 because hot shutdown was not achieved prior to reaching the HCTL based on the licensing basis ODYN calculation. The TRACG04 ATWS analysis was performed for the ATWS event initiated in the MELLLA+ operating domain with depressurization explicitly modeled (Reference 188).
 
TRACG04 is not the license basis calculation but was used to perform a complete assessment of possible conditions. ODYN cannot model depressurization. In the licensing basis ODYN ATWS analysis, ADS operation is inhibited and the vessel cycles on SRV setpoints until the reactor is shutdown. These limitations were factored into NRC approval for application of ODYN to ATWS.
Withheld Security-Related Information
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 122 of 239
 
The limiting event evaluated for the TRACG04 ATWS analysis is the MSIVC. The MSIVC and PRFO event behavior are essentially the same in the long-term as both events result in reactor isolation. Therefore, the MSIVC response is representative of both events for the long term simulation. One SRV was assumed out of service.
 
The key operator actions credited in the best estimate TRACG04 ATWS analysis, which are consistent with the plant specific EOPs, include:
* Water level control using the designated water level control strategy. Two different water level control strategies were investigated. Reactor level was controlled at either TAF or approximately TAF minus two feet.
* Initiation of SLC system boron injection at 120 seconds following the high pressure ATWS RPT signal.
* Following depressurization, operators maintain reactor vessel pressure between 20 and 50 psig by closing/re-opening ADS SRVs.
* Termination of all ECCS injection, except RCIC and SLC system, prior to depressurization.
 
The best-estimate TRACG04 calculations demonstrate that, depending on initial conditions, the HCTL may or may not be reached and emergency depressurization may not be required. The HCTL is a function of the reactor operating pressure and the suppression pool water level. For this reason, the best-estimate analysis was performed for bounding assumptions of HCTL of 150&deg;F to 175&deg;F. For the low HCTL value, depressurization is required, but not for the high level. The results of the analysis are presented in section 9.3.1.2 of Reference 182. For all cases analyzed, the ATWS acceptance criteria were satisfied. The containment analysis associated with the ATWS with depressurization is unaffected by the transition to Framatome fuel types.
 
14.8.2.3  ATSW With Core Instability (ATWSI) Analysis
 
The generic core instability evaluations continue to apply for the MELLLA+/EFW operating domain (Reference 185). However, a plant-specific ATWS instability calculation was performed, as required by Reference 191, to demonstrate that Monticello Emergency Operating Procedure (EOP) actions, including boron injection and water level control strategy (flow runback to uncover the feedwater spargers),
effectively mitigate an ATWS event with large power oscillations in the MELLLA+/EFW operating domain. The MNGP analysis is summarized in References 188 and 204. A detailed discussion of ATWS core instability and MELLLA+/EFW operation is included in Section 9.3.3 of Reference 182 and Reference 204. Limitations and requirements identified in the NRC review of Withheld Security-Related Information Withheld Security-Related Information
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 123 of 239
 
Reference 182 are addressed in Reference 184. TRACG04 and RAMONA5-FA calculations indicate that all applicable ATWS criteria are satisfied for ATWSI.
 
To support the transition to advanced methods, a peak clad temperature analysis was performed using Framatome code RAMONA5-FA (Reference 204).
Reference 14 describes the methods used in RAMONA5      -FA for ATWS-I analysis. As described in Reference  207 section 7.2.2 and Reference  7 section 7.3, the analysis for peak suppress ion pool temperature and peak containment pressure remains the TRACG04 analysis.
 
The limiting ATWS instability event was initiated from 102% current licensed thermal power and 80% rated core flow (MELLLA+). One SRV is assumed out of service.
Without oper ator action, the limiting event for peak suppression pool temperature and peak containment pressure is the turbine trip with bypass (TTWB). The TTWB isolates the feedwater heaters and increases core inlet subcooling. The event analysis included the use o f nominal inputs.
 
Reference  204 describes the fuel-specific ATWS -I analysis. The PCT analysis used balance of plant cond itions (e.g. initial power at natural circulation) from the TRACG04 analysis as boundary conditions (Reference      213 ). Peak clad temperature was analyzed for potenti ally limiting exposure statepoints. Analysis assumptions did not include use of margin to the LHGR limit.
 
To ensure that the analysis results bound future cycles, conservative measures were taken as described in Reference  204. Peak clad temperature for the limiting peak hot excess reactivity exposure point was  1100&deg;C (2012&deg;F), which is well below the limit of 2200&deg;F.
 
The two recirculation pump trip case was modeled with no operator action to lower water level, which resulted in an event that is bounded by the turbine trip case.
 
The key operator actions credited in the TTWB ATWS instability analysis include (Section 3.2 of Reference 188 and section 3.2.6 of reference 230:
 
ATWS mitigation features (i.e., prompt manual FW flow runback and early boron injection) are adequate to mitigate the ATWSI oscillations, and are still effective in the MELLLA+ domain. The calculations indicate that the ATWS acceptance criteria are satisfied even in the presence of unstable power oscillations when the MNGP specific timing for operator actions is used.
Withheld Security-Related Information MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 124 of 239
 
14.9 Deleted
 
14.10  Other Analyses
 
14.10.1 Adequate Core Cooling for Transients With a Single Failure
 
NUREG-0737, Task Item II.K.3.44 required licensees to demonstrate that the reactor core remains covered, or provide analysis to show that no significant fuel damage results from uncovering the core, for anticipated transients combined with the worst single failure, assuming proper operator actions. The General Electric BWR Owners Group responded to this item with a generic report applicable to BWR-2 through BWR-6 plants on December 19, 1980 (Refer  ence 53).
 
The BWR Owners Group report identified a loss-of-feedwater event as the worst anticipated transient, and loss of a high pressure inventory makeup or heat removal system as the worst single failure. The analyses showed that the reactor core remains covered for the combination of these worst-case conditions, without operator action to manually initiate the emergency core cooling system or other inventory makeup systems.
 
Item II.K.3.44 also included transients which result in a stuck open relief valve, in combination with the worst single failure, as a situation requiring analysis. Under these conditions, the analyses in the BWR Owners Group report showed that the reactor core remains covered with proper operator actions.
 
Northern States Power Company endorsed the BWR Owners Group report in reference to the Monticello Nuclear Generating Plant in a letter to the Director of Nuclear Reactor Regulation on November 12, 1981 (Reference 54). For Extended Power Uprate (EPU),
Probabilistic Risk Assessment (PRA) calculations were performed to evaluate a Loss of Feedwater Event at EPU conditions assuming a Stuck Open Relief Valve and using the RCIC System as the high-pressure injection source. The result of these evaluations demonstrated that adequate core cooling and containment integrity are maintained throughout the mitigation sequence (Reference 160).
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 125 of 239
 
14.11  References
: 1. ANP-10335P-A, Revision 0, "ACE/ATRIUM 11 Critical Power Correlation,"
May 2018.
: 2. Deleted.
: 3. ANP-10332P-A, Revision 0, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Loss of Coolant Accident Scenarios," March 2019.
: 4. XN-NF-82-07(P)(A) Revision 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," November 1982.
: 5. Letter from US NRC to C. Domingos "Monticello Nuclear Generating Plant -
Issuance of Amendment No. 210 RE: Revised Methodologies for Determining the Core Operating Limits (EPID L-2021-LLA-0144)," January 13, 2023, Ascension number ML22332A539.
: 6. ANP-4039P, Monticello Cycle 32 Reload Safety Analysis, March 2023.
: 7. ANP-3924, Revision 0, "Applicability of Framatome BWR Methods to Monticello with ATRIUM 11 Fuel.
: 8. ANP-10333P-A, Revision 0, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA).
: 9. Regulatory Guide 1.236, Revision 0, "Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents."
: 10. ANP-3934, Revision 0, "Monticello LOCA Analysis for ATRIUM 11 fuel."
: 11. ANP-3985, Revision 0, "Monticello Cycle 32 Plant Parameters Document."
: 12. ANP-3925, Revision 0, "Monticello ATRIUM 11 Transient Demonstration."
: 13. Monticello calculation number 04-040, Revision 4, "MNGP AST - CRDA Radiological Consequence Analysis."
: 14. ANP-10346P-A, Revision 0, "ATWS-I Analysis Methodology for BWRs Using RAMONA5-FA."
: 15. NSPM letter L-MT-09-048 (T J O'Connor) to NRC, "Monticello Extended Power Uprate: Response to NRC Containment and Ventilation Review Branch (SCVB)
Request for Additional Information (RAI) dated March 19, 2009, and March 26, 2009 (TAC No. MD9990)", dated July 13, 2009
: 16. Monticello calculation number 04-039, Revision 0B, "MNGP AST - MSLBA Radiological Consequences."
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 126 of 239
: 17. General Electric report, NEDO-21231, Banked Position Withdrawal Sequence, C J Paone, January 1977.
: 18. General Electric report, NEDO-10527, Rod Drop Accident Analysis for Large Boiling Water Reactors, C J Paone and J A Woolley, March 1972.
: 19. L-MT-22-046, "2022 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.46".
: 20. ANP-10300P-A, Revision 1, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Transient and Accident Scenarios.
: 21. General Electric report, NEDC-32514P, Revision 1, Monticello SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis, dated October 1997.
: 22. Deleted.
: 23. Deleted.
: 24. General Electric report, NEDC-23785-P, The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident; Volume III, SAFER/GESTR Application Methodology, Revision 1, October 1984.
: 25. Deleted.
: 26. Deleted.
: 27. 20-013 Rev. 0, Source Term for ATRIUM 11 Fuel.
: 28. Deleted.
: 29. Deleted.
: 30. Deleted.
: 31. Deleted.
: 32. Deleted.
: 33. Deleted.
: 34. Deleted.
: 35. Deleted.
: 36. Deleted.
: 37. Deleted.
: 38. Deleted.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 127 of 239
: 39. Deleted.
: 40. NSP (L O Mayer) letter to the NRC, Transmittal of ECCS Analysis, dated July 9, 1975.
: 41. Deleted.
: 42. General Electric (R E Engel) letter to the NRC (D G Eisenhut), Fuel Assembly Loading Error, November 30, 1977.
: 43. Deleted.
: 44. Deleted.
: 45. Deleted.
: 46. Deleted.
: 47. Deleted.
: 48. Deleted.
: 49. Deleted.
: 50. Deleted.
: 51. Deleted.
: 52. Deleted.
: 53. BWR Owners Group (D B Waters) letter to the NRC (D G Eisenhut), BWR Owners Group Evaluation of NUREG-0737 Requirements, dated December 29, 1980.
: 54. NSP (L O Mayer) letter to the NRC, Endorsement of BWR Owners Group Report on NUREG-0737, Item II.K.3.44, November 12, 1981.
: 55. Deleted.
: 56. Deleted.
: 57. Deleted.
: 58. Deleted.
: 59. Deleted.
: 60. Deleted.
: 61. Deleted.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 128 of 239
: 62. General Electric report, NEDC-30492-P, Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Monticello Nuclear Generating Plant, April 1984.
: 63. Deleted.
: 64. Deleted
: 65. Deleted
: 66. General Electric report, NEDC-30515, GE BWR Extended Load Line Limit Analysis for Monticello Nuclear Generating Plant, Cycle 11, March 1984.
: 67. Deleted.
: 68. NRC (J J Stefano) letter to NSP (T M Parker), Alternate Rod Injection System Instrument Diversity Requirement (TAC No. 73022), dated June 7, 1989.
: 69. NSP (T M Parker) letter to the NRC, Alternate Rod Injection System Instrument Diversity Requirement (TAC No. 73022), dated August 31, 1989.
: 70. NRC (W O Long) letter to NSP (T M Parker), Implementation of Alternate Rod Injection System (ARI) Diversity Requirements in 10CFR 50.62 (ATWS Rule) -
Monticello Nuclear Generating Plant (TAC 73022), dated March 29, 1991.
: 71. General Electric report, NEDC-31849P, Maximum Extended Load Line Limit Analysis for Monticello Nuclear Generating Plant Cycle 15, June 1992.
: 72. NSP (T M Parker) letter to the NRC, Submittal of Monticello Individual Plant Examination (IPE) Report, dated February 27, 1992.
: 73. NSP (T M Parker) letter to the NRC, Response to Request for Additional Information Concerning the Monticello Individual Plant Examination (IPE)
Submittal - Generic Letter 88-20 (TAC No. M74435), dated February 15, 1993.
: 74. General Electric report, NEDC-31778-P, Safety Review for Monticello Nuclear Generating Plant Increased Core Flow Operation Throughout Cycle 14, H X Hoang, December 1989.
: 75. Deleted.
: 76. Siemens report EMF-94-055(P), Monticello 9x9-1X Qualification Fuel Assemblies Safety Analysis Report, July 1994.
: 77. NRC (B A Wetzel) letter to NSP (R O Anderson), Staff Evaluation of the Monticello Nuclear Generating Plant Individual Plant Examination (IPE) - Internal Events Submittal, dated May 26, 1994.
: 78. ANP-10344P-A Revision 0, "Framatome Best-estimate Enhanced Option III Methodology," March 2021.
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                  USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 129 of 239
: 79. Deleted.
: 80. Deleted.
: 81. Deleted.
: 82. Deleted.
: 83. Deleted.
: 84. NRC (D Crutchfield) Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities - 10CFR50.54(f), dated November 23, 1988.
: 85. NRC (J G Partlow) Generic Letter 88-20, Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities -
10CFR50.54(f), dated June 28, 1991.
: 86. NSP (T M Parker) letter to the NRC, Response to Generic Letter 88-20, Supplement 4, Individual Plant Examination of External Events for Severe Accident Vulnerabilities, dated December 20, 1991.
: 87. NSP (W J Hill) letter to the NRC, Supplemental Response to Generic Letter 88-20, Supplement 4, Individual Plant Examination of External Events for Severe Accident Vulnerabilities, dated January 5, 1995.
: 88. NSP (W J Hill) letter to the NRC, Submittal of Monticello Individual Plant Examination of External Events (IPEEE) Report, dated March 1, 1995.
: 89. NSP (W J Hill) letter to the NRC, Submittal of Monticello Individual Plant Examination of External Events (IPEEE) Report, Revision 1; Seismic Analysis, Revision 0 and Internal Fires Analysis, Revision 1 (TAC M83644), dated November 20, 1995.
: 90. ASME Boiler & Pressure Vessel Code, Section III, Class I, 1965 Edition.
: 91. NRC (R P Zimmerman) Generic Letter 94-02, Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors, dated July 11, 1994.
: 92. NRC (C E Rossi) Bulletin 88-07, Power Oscillations in Boiling Water Reactors (BWRs), dated June 15, 1988.
: 93. OG 02-0119-260, "Backup Stability Protection (BSP) for Inoperable Option III Solution, July 17, 2002".
: 94. Deleted.
: 95. Deleted.
: 96. Deleted.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 130 of 239
: 97. Deleted.
: 98. Deleted.
: 99. American National Standards Institute ANSI/ANS 5.1 - 1979, Decay Heat in Light Water Reactors.
 
100. American Nuclear Standards Proposed Standard, ANS 5.1, Decay Energy Release Rates Following Shutdown for Uranium-Fueled Thermal Reactors (Proposed), October 1971.
 
101. WASH-1270, Technical Report on Anticipated Transients Without Scram for Water Cooled Power Reactors, September 1973.
 
102. NRC (K R Goller) letter to NSP (L O Mayer), Safety Evaluation of Proposed Installation of RPT as a Short Term ATWS Solution, dated February 23, 1977.
 
103. NRC (D G Eisenhut) Generic Letter 83-28, Required Actions Based on Generic Implications of Salem ATWS Events, dated July 8, 1983.
 
104. Deleted.
 
105. Deleted.
 
106. Global Nuclear Fuels - Americas Report, NEDE-24011-P-A-27, General Electric Standard Application For Reactor Fuel, Revision 27, August 2018 and U.S.
Supplement, NEDE-24011-P-A-27-US, August 2018.
 
107. Deleted.
 
108. Monticello calculation number 11-183, Revision 0, "Task Report T0902 -
Anticipated Transients Without Scram."
 
109. NRC (C F Lyon) letter to NSP (R O Anderson), Request for Deviation from Emergency Procedure Guidelines, Revision 4, NEDO-31331, March 1987 (TAC No. MA0168), dated December 10, 1998.
 
110. General Electric report NEDC-31858P BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, Revision 2, September 1993.
 
111. NSP (W J Hill) letter to the NRC, Supplement to License Amendment Request Titled - Revised Core Spray Pump Flow and Other Editorial Corrections, dated August 25, 1994.
 
112. Deleted.
 
113. Deleted.
 
114. Deleted.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 131 of 239
 
115. NRC (T Kim) letter to NSP (R O Anderson, Issuance of Amendment Re: Main Steam Isolation Valve and 10 CFR Part 50, Appendix J, Leak Test Requirement (TAC No. M93332), dated April 3, 1996.
 
116. Deleted.
 
117. NRC (D C Dilanni) letter to NSP (D M Musolf), Amendment No. 56 to Facility Operating License No. DPR-22: Standby Liquid Control, dated December 11, 1987.
 
118. NRC (J J Stefano) letter to NSP (D M Musolf), Amendment No. 57 to Facility Operating License No. DPR-22: Increased Sodium Pentaborate Boron-10 Enrichment, dated September 23, 1988.
 
119. NRC (C F Lyon) letter to NSP (M F Hammer), Review of Monticello Individual Plant Examination of External Events (IPEEE) Submittal (TAC No. M83644),
dated April 14, 2000.
 
120. Deleted.
 
121. General Electric (P T Tran) letter to NSP (S J Hammer), Revised Mass and Energy Release Rates for Power Rerate HELB Analysis (Task 27), GLN-96-086, dated December 10, 1996.
 
122. GE Hitachi Nuclear Energy Report NEDC-32868P, Revision 5, GE14 Compliance With Amendment 22 of NEDE-24011-P-A (GESTAR II), May 2013.
 
123. GE14 Fuel Design Cycle-Independent Analysis for Monticello Nuclear Generating Plant, GE-NE-0000-0013-9576P, GE Nuclear Energy (Proprietary), March 2003
 
124. Deleted.
 
125. Current Cycle Monticello Nuclear Generating Plant Core Operating Limits Report.
 
126. Deleted.
 
127. General Electric Report NEDO-33091-A, Improved BPWS Control Rod Insertion Process, J. Tuttle, July, 2004.
 
128. Deleted.
 
129. NRC Regulatory Guide 1.183, Alternative Radiological Source Terms For Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.
 
130. NSPM letter L-MT-08-052 (T J O'Connor) to NRC, " License Amendment Request:
Extended Power Uprate (TAC MD9990)", dated November 5, 2008.
 
131. NSPM letter L-MT-09-002 (T J O'Connor) to NRC, "Response to NRC Probabilistic Risk Assessment (PRA) Branch Requests For Additional Information (RAls) dated December 5, 2008 (TAC No. MD9990) ", dated February 9, 2009.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 132 of 239
 
132. NSPM letter L-MT-09-029 (T J O'Connor) NRC, "Monticello Extended Power Uprate: Response to NRC Probabilistic Risk Assessment (PRA) Licensing Branch Requests For Additional Information (RAl) dated April 29, 2009 (TAC No.
MD9990)", dated May 29, 2009.
 
133. NSPM letter L-MT-12-114 (M A Schimmel) to NRC, "Monticello Extended Power Uprate (EPU): Supplement for Gap Analysis Updates (TAC MD9990)", dated January 21, 2013.
 
134. Letter from NRC (T A Beltz) to NSPM (K D Fili), "Monticello Nuclear Generating Plant - Issuance of Amendment No. 176 to Renewed Facility Operating License Regarding Extended Power Uprate (TAC No. MD9990)", dated December 9, 2013.
 
135. Deleted.
 
136. NRC (P S Tam) letter to NMC (J T Conway), Monticello Nuclear Generating Plant
            - Issuance Of Amendment Re: Full-Scope Implementation Of The Alternative Source Term Methodology (TAC NO MC8971), dated December 7, 2006.
 
137. NRC (P S Tam) letter to NMC (J T Conway), Monticello Nuclear Generating Plant (MNGP) - Correction Of Safety Evaluation Associated With Alternative Source Term Amendment (TAC NO MC8971), dated April 17, 2007.
 
138. NMC (J T Conway) letter to the NRC, License Amendment Request - Full Scope Application of an Alternative Source Term, dated September 15, 2005.
 
139. NMC (J T Conway) letter to the NRC, Full Scope Alternate Source Term -
Supplemental Information, dated April 13, 2006.
 
140. NMC (J T Conway) letter to the NRC, Response to Request for Additional Information Related to License Amendment Request for Full Scope Alternative Source Term, dated August 22, 2006.
 
141. NRC Regulatory Guide 1.194, Atmospheric Relative Concentrations For Control Room Radiological Habitability Assessments At Nuclear Power Plants, June 2003.
 
142. NRC Regulatory Guide 1.145, Atmospheric Dispersion Models For Potential Accident Consequence Assessments At Nuclear Power Plants, November 1982.
 
143. General Electric Report, GE-NE-000-0052-3313-P-RO, Monticello Nuclear Generating Plant SAFER/GESTR ECCS-LOCA Analysis - LPCI Loop Selective Detachable Break Area. dated September 2006.
 
144. NMC (J T Conway) letter to the NRC,  License Amendment Request:  Revision to the allowable Value and Channel Calibration Surveillance interval for the Recirculation Riser Pressure - High Function, dated September 25, 2007.
 
145. NRC (P S Tam) letter to NSP-M (T J OConnor) Monticello Nuclear Generating Plant - Issuance of Amendment regarding Recirculation Riser Differential Pressure (TAC NO MD6864), dated April 7, 2009.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 133 of 239
 
146. Deleted.
 
147. Deleted.
 
148. Deleted.
 
149. NRC (P Tam) letter to NSPM, Issuance of Amendment Regarding the Power Range Neutron Monitoring System (TAC NO. MD8064), dated January 30, 2009.
 
150. Deleted.
 
151. NMC (J T Conway) letter to the NRC, License Amendment Request: Revision to Required Actions for Specification 3.5.1 Emergency Core Cooling System, dated June 26, 2008.
 
152. NRC (K D Feintuch) letter to NSPM (T J OConnor) Monticello Nuclear Generating Plant - Issuance of Amendment Regarding Completion Time to Restore a Low-Pressure Emergency Core Cooling Subsystem to Operable Status (TAC NO MD9170), dated July 10, 2009.
 
153. Monticello calculation number 04-038, Revision 3, "MNGP AST-LOCA Radiological Consequence Analysis".
 
154. Monticello calculation number 08-027, Revision 2, "MNGP EPU-CR and TSC Direct Dose".
 
155. GE Hitachi EPU Project Task Report GE-NE-0000-0060-9229-TR-R3, "Task T0400: Containment System Response", February 2011 (Monticello calculation number 11-173).
 
156. Deleted.
 
157. GE Hitachi EPU Project Task Report GE-NE-0000-0060-9286-TR-R2, Revision 2, "Task T0407: ECCS-LOCA SAFER/GESTR", October 2011 (Located in Monticello calculation number 11-180).
 
158. Deleted.
 
159. NSPM letter L-MT-09-049 (T J O'Connor) to NRC, "Monticello Extended Power Uprate: Response to NRC Reactor Systems Review Branch and Nuclear Code and Performance Review Branch Request for Additional Information (RAI) dated March 23, 2009 and Nuclear Code and Performance Review Branch Request for Additional Information dated April 27, 2009 (TAC No. MD9990)", dated July 23, 2009.
 
160. GE Hitachi Report NEDC-33322-P, Revision 3, "Safety Analysis Report for Monticello Constant Pressure Power Uprate", October 2008.
 
161. Deleted.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 134 of 239
 
162. NSPM letter L-MT-12-082 (M A Schimmel) to NRC, "Monticello Extended Power Uprate and Maximum Extended Load Line Limit Analysis Plus License Amendment Requests: Supplement to Address SECY 11-0014 Use of Containment Accident Pressure (TAC Nos. MD9990 and ME3145)", dated September 28, 2012.
 
163. GE Hitachi EPU Project Task Report GE-NE-0000-0064-6767-TR-R0, Revision 0, "Task T0802: Core Source Term", July 2007 (Monticello calculation number 11-245).
 
164. GE Hitachi Report 0000-0152-0735-R0, "Monticello Long Term Cooling",
September 18, 2012.
 
165. Deleted.
 
166. NUREG/CR6604, "RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation", April 1998, Supplement 1, June 8, 1999, and Supplement 2, October 2002.
 
167. Deleted.
 
168. GE Letter GLN-95-032 (P T Tran) to NSP (S J Hammer), "Emergency Core Cooling Parameters for use in SAFER/GESTR Power Rerate Analysis - Task 7.5",
dated September 12, 1995.
 
169. NRC (P S Tam) letter to NSPM (T J O'Connor), "Monticello Nuclear Generating Plant - Issuance of Amendment RE: Minimum Core Spray Pump Flow Rate (TAC No. ME5441)", dated January 11, 2012 (Amendment No. 167).
 
170. NRC (T J Kim) letter to NSP (R O Anderson), "Monticello Nuclear Generating Plant
            - Issuance of Amendment RE: Revised Core Spray Pump Flow (TAC No.
M85838)", dated July 12, 1995 (Amendment No. 93).
 
171. Monticello calculation number 96-085, "Structural Integrity of Core Spray Line Inside the Reactor".
 
172. 13-059 Rev 0, "MNGP EPU - Core inventory with AREVA Atrium 10-XM Fuel".
 
173. 16-090 Rev 0, "MNGP EPU - Core Inventory with GE14 Fuel with 37 GWD/MTU Exposure".
 
174. Deleted.
 
175. Monticello calculation number 04-041, Revision 4, "MNGP AST - FHA Radiological Consequence Analysis".
 
176. Monticello calculation number 12-040, Revision 0, "Containment Accident Pressure Assessment of ECCS Pump NPSHa Requirements for Extended Power Uprate (EPU)".
 
177. Monticello calculation number 04-036, Revision 1, "MNGP AST - Offsite Post-Accident Dispersal Analysis".
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 135 of 239
 
178. Monticello calculation number 04-037, Revision 3, "MNGP AST - CR/TSC Post-Accident Atmospheric Dispersal Analysis".
 
179. Monticello calculation number 03-036, Revision 2, "Instrument Setpoint Calculation Reactor Low Pressure Permissive Bypass Timer".
 
180. Deleted.
 
181. Deleted.
 
182. GE Hitachi Nuclear Energy Report NEDC-33435P, "Safety Analysis Report for Monticello Maximum Extended Load Line Limit Analysis Plus," Revision 1, December 2009.
 
183. Deleted.
 
184. Letter from NRC (T A Beltz) to NSPM (K D Fili), "Monticello Nuclear Generating Plant - Issuance of Amendment No. 180 to Renewed Facility Operating License Regarding Maximum Extended Load Line Limit Analysis Plus (TAC No. ME3145),"
dated March 28, 2014 (ADAMS Accession No. ML13317A866).
 
185. GE Nuclear Energy Report NEDC-33006P-A, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, Revision 3, dated June 2009 (ADAMS Accession No. ML091800530)
 
186. Deleted
 
187. Deleted.
 
188. Monticello calculation 11-343, Rev. 0, "Task T0902 M+ Anticipated Transients Without Scram."
 
189. Letter from NRC (R. Kuntz) to NSPM (P. Gardner), "Monticello Nuclear Generating Plant- Issuance of Amendment RE: Extended Flow Window (CAC NO. MF5002),"
February 23, 2017.
 
190. GE Nuclear Energy Report NEDC3242P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," February 1999.
 
191. Letter from NRC (H K Hieh) to General Electric (R E Brown), "Final Safety Evaluation for GE-Hitachi Nuclear Energy Americas, LLC (GHNE) Topical Report (TR) NEDC-3006P, "Maximum Extended Load Line Limit Analysis Plus" (TAC No.
MB6157)," September 17, 2007.
 
192. GE Hitachi MELLLA+ Project Task Report 0000-0096-6889-TR-R1, Revision 1, "Task T0407: ECCS-LOCA SAFER/GESTR", October 2011 (located in Monticello calculation number 11-180).
 
193. Monticello calculation 13-055, Rev. 2, "Core Spray and LPCI Flow Delivery to Reactor Vessel for Safety Analysis."
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 136 of 239
 
194. ANP-3119 Rev 1, "Mechanical Design Report for Monticello ATRIUM'    10XM Fuel Assemblies," October 2016.
 
195. ANF-89-98 Rev 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," May 1995.
 
196. ANP-3519 Rev 0, "Fuel Rod Thermal-Mechanical Design for Monticello ATRIUM'    10XM Fuel Assemblies, Cycle 29," November 2016 .
 
197. EMF-93-177 Rev 1, "Mechanical Design for BWR Fuel Channels, Framatome ANP Inc.," August 2005.
 
198. ANP-3092P Rev 1, "Monticello Thermal-Hydraulic Design Report for ATRIUM' 10XM Fuel Assemblies," June 2016.
 
199. T. Beltz to P. Gardner, Monticello Nuclear Generating Plant - Issuance of Amendment to Transition to Areva Atrium 10XM Fuel and Areva Safety Analysis Methods (TAC No. MF2479) (Am 188)"," June 5, 2015.
 
200. Deleted.
 
201. ANP-2637, Revision 9, "Boiling Water Reactor Licensing Methodology Compendium", April 2021.
 
202. ANP-3224 Rev 2, "Applicability of AREVA NP BWR Methods to Monticello,"
June 2013.
 
203. ANP-3135 Rev 0, "Applicability of AREVA BWR Methods to Extended Flow Window for Monticello," April 2014.
 
204. ANP-3933, Revision 0, "Monticello ATWS-I Evaluation for ATRIUM 11 Fuel" and FS1-0065675, Revision 1.0, Clarifications for Monticello ATWS-I Evaluation for ATRIUM 11 Fuel, combined in Monticello calculation 22-007.
 
205. Deleted
 
206. ANP-3213 Rev 1, "Monticello Fuel Transition Cycle 28 Reload Licensing Analysis (EPU/MELLLA)," June 2013. (Monticello calculation number 16-015)
 
207. ANP-3295 Rev 3, "Monticello Licensing Analysis for EFW (EPU/MELLLA+),"
February 2016. (Monticello calculation number 16-018)
 
208. ANP-3719P Rev 0, "Monticello LOCA Break Spectrum for EPU/EFW with ATRIUM 10XM Fuel and Revised Modeling Parameters," November 2018
 
209. Deleted.
 
210. 13-059 Rev 0, "MNGP EPU - Core inventory with AREVA Atrium 10-XM Fuel".
 
211. Deleted.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                            USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                Revision 40 Page 137 of 239
 
212. Deleted.
 
213. NSPM letter L-MT-15-081, P Gardner to US NRC, "License Amendment Request for AREVA Extended Flow Window Supplement to Provide Revised Analysis of Anticipated Transient Without Scram Instability (TAC No. MF5002)", December 8, 2016
 
214. ANP-10298P-A Revision 1, ACE/ATRIUM 10XM Critical Power Correlation, March 2014
 
215. Deleted.
 
216. Deleted.
 
217. ANP-10307P-A Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, June 2011
 
218. EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.
 
219. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal - Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
 
220. Letter from NRC to P. A. Gardner (Xcel Energy), "Monticello Nuclear Generating Plant - Issuance of Amendment to Transition to AREVA ATRIUM 10XM Fuel and AREVA Safety Analysis Methods (TAC NO. MF2479)", June 5, 2015 (Assession Number ML15072A141)
 
221. BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.
 
222. XN-CC-33(A) Revision 1, HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual, Exxon Nuclear Company, November 1975.
 
223. Deleted.
 
224. EMF-2292(P)(A) Revision 0, ATRIUM'  -10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
 
225. ANP-3666P Revision 0, "Monticello Cycle 30 Plant Parameters Document,"
June 2018
 
226. ANP-3720P Revision 0, "Monticello LOCA MAPLHGR Limits for EPU/EFW with ATRIUM 10XM Fuel and Revised Modeling Parameters," January 2019.
 
227. Deleted.
 
228. Monticello calculation 05-030, Rev. 2, "Determination of SRV Self Actuation Setpoint to be Used in Reactor Vessel Overpressure Analysis."
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 138 of 239
 
229. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990.
 
230. R. Kuntz to P Gardner, "Monticello Nuclear Generating Plant- Issuance of Amendment RE: Extended Flow Window (CAC NO. MF5002)."
 
231. Deleted.
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                            USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                Revision 40 Page 139 of 239
 
FIGURES
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 140 of 239
 
Figure 14.7-1    Deleted
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 141 of 239
 
Figure 14.7-1a  Deleted
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 142 of 239
 
Figure 14.7-1b  Deleted
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 143 of 239
 
Figure 14.7-2    Deleted
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 144 of 239
 
Figure 14.7-3    Deleted
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 145 of 239
 
Figure 14.7-4    Deleted
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 146 of 239
 
Figure 14.7-5    Deleted
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 147 of 239
 
Figure 14.7-6    Deleted
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 148 of 239
 
Figure 14.7-7    Deleted
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 149 of 239
 
Figure 14.7-8  CS Flow Delivery Assumed for Break Spectrum
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 150 of 239
 
Figure 14.7-9  LPCI Flow Delivery Assumed for Break Spectrum
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 151 of 239
 
Figure 14.7-10  ADS Actuation Logic
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 152 of 239
 
Figure 14.7-11  CS Initiation Logic
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 153 of 239
 
Figure 14.7-12  LPCI Initiation Logic
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                    Revision 40 Page 154 of 239
 
Figure 14.7-13  Main Steam Line Break Accident, Break Location
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                              USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 155 of 239
 
Figure 14.7-14 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                    USAR-14 SECTION 14    PLANT SAFETY ANALYSIS                          Revision 40 Page 156 of 239
 
UPDATED SAFETY ANALYSIS REPORT Section 14A
 
TD-XCL-NADOctober 2023-MN-054, Revision 0
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 157 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 158 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 159 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 160 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 161 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 162 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 163 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 164 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 165 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 166 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 167 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 168 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 169 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 170 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 171 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 172 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 173 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 174 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 175 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 176 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 177 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 178 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 179 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 180 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 181 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 182 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 183 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 184 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 185 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 186 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 187 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 188 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 189 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 190 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 191 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 192 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 193 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 194 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 195 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 196 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 197 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 198 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 199 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 200 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 201 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 202 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 203 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 204 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 205 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 206 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 207 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 208 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 209 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 210 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 211 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 212 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 213 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 214 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 215 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 216 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 217 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 218 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 219 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 220 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 221 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 222 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 223 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 224 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 225 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 226 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 227 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 228 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 229 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 230 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 231 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 232 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 233 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 234 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 235 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 236 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 237 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 238 of 239
 
MONTICELLO UPDATED SAFETY ANALYSIS REPORT                                USAR-14 SECTION 14      PLANT SAFETY ANALYSIS                                  Revision 40 Page 239 of 239}}

Latest revision as of 18:37, 4 October 2024

1 to the Updated Safety Analysis Report, Chapter 14, Plant Safety Analysis
ML24109A096
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 04/17/2024
From:
Northern States Power Company, Minnesota, Xcel Energy
To:
Office of Nuclear Reactor Regulation
Shared Package
ML24109A128 List:
References
L-MT-24-002
Download: ML24109A096 (1)
v • d • e


Text

Retrieved from "https://kanterella.com/w/index.php?title=ML24109A096&oldid=3677809"