ML16054A428

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Monticello - Revision 33 to the Updated Final Safety Analysis Report, Section 14, Plant Safety Analysis
ML16054A428
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 01/26/2016
From:
Northern States Power Co, Xcel Energy
To:
Office of Nuclear Reactor Regulation
Shared Package
ML16054A376 List:
References
L-MT-16-004
Download: ML16054A428 (177)
v • d • e


Text

SECTION 14

SECTION 1414.1

14.1.1 14.1.2

14.1.3

14.1.4

14.1.5 SECTION 1414.2

SECTION 1414.314.3.1

14.3.2

14.3.3

14.3.4

SECTION 1414.4

14.4.1

14.4.2

14.4.3

14.4.4

14.4.5

SECTION 1414.5

14.5.1

14.5.2 14.5.3

SECTION 1414.6

14.6.1 14.6.2 14.6.314.6.414.6.5

SECTION 1414.7

14.7.1

14.7.2

14.7.3

14.7.4

14.7.5

14.7.6

14.7.7

14.7.8

SECTION 1414.814.8.1

14.8.2

SECTION 1414.1014.10.1

SECTION 1414.11

SECTION 14

SECTION 14AUPDATED SAFETY ANALYSIS REPORT

?l 0 Xcel Energy* 2015 Xcol Enorgy Nuclear Analysis and Design Page 1 of 19 July 2015 Section Revision: 0 2 19 Revision: 0 3 19 Revision: 0 4 sections. ARTS Average Po'vV&r Range Monitor, Rod Block Monitor, and Technical Specification # Gard-e I OffiCial Core Monitoring System fo*r Monticello MCPR Minimum Critical Po\'VGr Ratio OPRM Oscillation Power Range Monitor Revision: 0 5 Option A Scram Time representative of the Technical Specification requirements Revision: 0 0 I wlo Revision: 0 7 19 The final core loading pattern analyzed in this report \VaS transmitted to GNF in correspondence 3). described in Cycle GE14*P1 < < 133781 26 2-17GZ 141n {41781 27 14" {43381 26 ,_ JYS001 04 ,.,.JYY541 Number Of Bundles 0 44 56 24 24 6 Initial Avg. 13 11 l1 3.91 3. 3. 3, 3.89 3.89 Revision: 0 The minimum shutdovvn margin (SDM) reported bek>w is based on moderator temperature of exposure of11.849 GWDJM'TU corresponds to tlile minimum previous cycle exposure Operations Manual 8.03.04 .. 05 requires that the minimum torus water temperature is greater calculated SDM results bel\wen 68°F and 6S"F is insignificant as compared to the uncertainties was found to be sufficient. A conservative depletion strategy was utilized in the evaluation of standby liquid control shutdown margin. The resuHs presented are cornsorvatively based on an end of tho prevtous reactor, from a full po\A/er and minimum control rod inventory to a sutrcritical condrtion at any time in the cycle under the most reactive free state by the injection of 660 ppm boron.

Revision: 0 9 19 This soction identifies the transient and accident analyses performed as part of the current cycle the limitations and conditions. These cycle-specific Himitations and are met for Monticello another issue were evaluated. These events are listed below.

1) 2) 3) 4) 7) Event Primary System Pressure Increase Generator load Reiection with Bypass Failure Turt:line Trip with Bypass Failure Main Steam Isolation Valve CloStxe (One I All Vatves) Turtine Trip with Bypass Failure w/o Position Scram 1 Main Steam Isolation Valve Closure Scram Loss of Condenser Vacuum Reactor Vessel Water Temperature Decrease Inadvertent HPCI Actuatron wrth L8 Turbine Tnp Posittve Reactivity Insertion Rod WlthdrawaJ Error Reactor Vessel Coolant Inventory Decrease Core Coolant Flow Decrease Trip ot One Recirculation Pump or Two Rearcu!ation Pumps Recirct.Jation Pump Seizure. Core Coolant Flow Increase-Slow Reci'ctJat!on Control Failure -Increase (MCPRF) Slow Reci'CIAation Control Failure -Increase (MAPlHGR,)" Fast RecirculaNon Control Failure -Increase Stanup of an Idle Recuculation LOop Fuel Loading Errors Misplaced Boodte-Acaclent 1 Performed ror ASME Vessel Overpressure Compliance. Revision: 0 Current ./ ./ ./ ./ ./ (ilSLO) ./ ./

Revision: 0 Reactor full power initial conditions that apply to the limiting transient analysis are summarized transiont analyses are fisted in Table 4.3. Parameter Value Value Increased Core Low COfe Fklw Flow Rated Thermal Power 2004 46.1 Analysis Power I Core Flow Analysis Reactor Pressure Time in Cycle Number of*SIRVs for Analysis 5 in-service in-service two Pslg low Vessel Water Level Scram 1 %open Low low Vessel Water 1 High Pressure Pslg High Reactor Vessel Water Leve11 Psig Revision: 0 125 10915 85 85 *55 1162 1170 ror be times. Option 1 be be Nollimiting be 2 to 3 Revision: 0 OLMCPR for moasured scram insertion times. 3 5.2 177 1.57 1.62 core monitoring system uses more detailed lirrits for each fuel bundle lattice.

Revision: 0 15 111 2.1.2 specification states that the pressure measured in the reactor steam dome shall not exceed 1332 Psig. The pressure safety limit of 1332 Psig as measured in tho vessel steam space was derived from the design pressures of the reactor pressure vessel, steam space piping, and water space piping. The pressure safety limit was chosen as the lowor of the pressure 1332 initial conditions:

  • 100% (2004 1 105% (60.5 x1061010.0 1170.0 1320 Technical SpeciflCation limit of 1332 Psig. The calculated maximum steam line pressure of 1314 1332 1344 1375 Monticello uses TRACG for the MSIV cmure-Witoout Position Scram analysis. This ana!ysis is rt11 at 100% 1 (1010 psig) and uncertainty are applied as a pressure adder to the result of the event to obtain the final reported Revision: 0 rod thermal*mechanical design and licensing basis, with resped to both steady state operations, and transient and accident events. Thermal overpower limits arc defined to evaluate the potential for fuel centerline metting. Mechanical overpower limits are deftned to evaluate the potential for fuel cladding overstrain. Feedwater ControUer Faaure event mechanical design and lioensing basis criteria for the plant.

Revision: 0 To provide Monticello Nuclear Generating Plant -with operating improvemonts, expanded operating domain an:alyses were performed in Reference 1 for maximum extended load line limit reload analyses in Reference 1.

Revision: 0 A reload DSS.CO evaluation has beon performxt in accordance Volith tho licensing rnGthodology 0) trip linear segment. povver intercopt WasP.. TRIP Was.p.sREAK RDF-Recirculation Drive F'low Revision: 0 SECTION 14

SECTION 1414.1

14.1.1 14.1.2

14.1.3

14.1.4

14.1.5 SECTION 1414.2

SECTION 1414.314.3.1

14.3.2

14.3.3

14.3.4

SECTION 1414.4

14.4.1

14.4.2

14.4.3

14.4.4

14.4.5

SECTION 1414.5

14.5.1

14.5.2 14.5.3

SECTION 1414.6

14.6.1 14.6.2 14.6.314.6.414.6.5

SECTION 1414.7

14.7.1

14.7.2

14.7.3

14.7.4

14.7.5

14.7.6

14.7.7

14.7.8

SECTION 1414.814.8.1

14.8.2

SECTION 1414.1014.10.1

SECTION 1414.11

SECTION 14

SECTION 14AUPDATED SAFETY ANALYSIS REPORT

?l 0 Xcel Energy* 2015 Xcol Enorgy Nuclear Analysis and Design Page 1 of 19 July 2015 Section Revision: 0 2 19 Revision: 0 3 19 Revision: 0 4 sections. ARTS Average Po'vV&r Range Monitor, Rod Block Monitor, and Technical Specification # Gard-e I OffiCial Core Monitoring System fo*r Monticello MCPR Minimum Critical Po\'VGr Ratio OPRM Oscillation Power Range Monitor Revision: 0 5 Option A Scram Time representative of the Technical Specification requirements Revision: 0 0 I wlo Revision: 0 7 19 The final core loading pattern analyzed in this report \VaS transmitted to GNF in correspondence 3). described in Cycle GE14*P1 < < 133781 26 2-17GZ 141n {41781 27 14" {43381 26 ,_ JYS001 04 ,.,.JYY541 Number Of Bundles 0 44 56 24 24 6 Initial Avg. 13 11 l1 3.91 3. 3. 3, 3.89 3.89 Revision: 0 The minimum shutdovvn margin (SDM) reported bek>w is based on moderator temperature of exposure of11.849 GWDJM'TU corresponds to tlile minimum previous cycle exposure Operations Manual 8.03.04 .. 05 requires that the minimum torus water temperature is greater calculated SDM results bel\wen 68°F and 6S"F is insignificant as compared to the uncertainties was found to be sufficient. A conservative depletion strategy was utilized in the evaluation of standby liquid control shutdown margin. The resuHs presented are cornsorvatively based on an end of tho prevtous reactor, from a full po\A/er and minimum control rod inventory to a sutrcritical condrtion at any time in the cycle under the most reactive free state by the injection of 660 ppm boron.

Revision: 0 9 19 This soction identifies the transient and accident analyses performed as part of the current cycle the limitations and conditions. These cycle-specific Himitations and are met for Monticello another issue were evaluated. These events are listed below.

1) 2) 3) 4) 7) Event Primary System Pressure Increase Generator load Reiection with Bypass Failure Turt:line Trip with Bypass Failure Main Steam Isolation Valve CloStxe (One I All Vatves) Turtine Trip with Bypass Failure w/o Position Scram 1 Main Steam Isolation Valve Closure Scram Loss of Condenser Vacuum Reactor Vessel Water Temperature Decrease Inadvertent HPCI Actuatron wrth L8 Turbine Tnp Posittve Reactivity Insertion Rod WlthdrawaJ Error Reactor Vessel Coolant Inventory Decrease Core Coolant Flow Decrease Trip ot One Recirculation Pump or Two Rearcu!ation Pumps Recirct.Jation Pump Seizure. Core Coolant Flow Increase-Slow Reci'ctJat!on Control Failure -Increase (MCPRF) Slow Reci'CIAation Control Failure -Increase (MAPlHGR,)" Fast RecirculaNon Control Failure -Increase Stanup of an Idle Recuculation LOop Fuel Loading Errors Misplaced Boodte-Acaclent 1 Performed ror ASME Vessel Overpressure Compliance. Revision: 0 Current ./ ./ ./ ./ ./ (ilSLO) ./ ./

Revision: 0 Reactor full power initial conditions that apply to the limiting transient analysis are summarized transiont analyses are fisted in Table 4.3. Parameter Value Value Increased Core Low COfe Fklw Flow Rated Thermal Power 2004 46.1 Analysis Power I Core Flow Analysis Reactor Pressure Time in Cycle Number of*SIRVs for Analysis 5 in-service in-service two Pslg low Vessel Water Level Scram 1 %open Low low Vessel Water 1 High Pressure Pslg High Reactor Vessel Water Leve11 Psig Revision: 0 125 10915 85 85 *55 1162 1170 ror be times. Option 1 be be Nollimiting be 2 to 3 Revision: 0 OLMCPR for moasured scram insertion times. 3 5.2 177 1.57 1.62 core monitoring system uses more detailed lirrits for each fuel bundle lattice.

Revision: 0 15 111 2.1.2 specification states that the pressure measured in the reactor steam dome shall not exceed 1332 Psig. The pressure safety limit of 1332 Psig as measured in tho vessel steam space was derived from the design pressures of the reactor pressure vessel, steam space piping, and water space piping. The pressure safety limit was chosen as the lowor of the pressure 1332 initial conditions:

  • 100% (2004 1 105% (60.5 x1061010.0 1170.0 1320 Technical SpeciflCation limit of 1332 Psig. The calculated maximum steam line pressure of 1314 1332 1344 1375 Monticello uses TRACG for the MSIV cmure-Witoout Position Scram analysis. This ana!ysis is rt11 at 100% 1 (1010 psig) and uncertainty are applied as a pressure adder to the result of the event to obtain the final reported Revision: 0 rod thermal*mechanical design and licensing basis, with resped to both steady state operations, and transient and accident events. Thermal overpower limits arc defined to evaluate the potential for fuel centerline metting. Mechanical overpower limits are deftned to evaluate the potential for fuel cladding overstrain. Feedwater ControUer Faaure event mechanical design and lioensing basis criteria for the plant.

Revision: 0 To provide Monticello Nuclear Generating Plant -with operating improvemonts, expanded operating domain an:alyses were performed in Reference 1 for maximum extended load line limit reload analyses in Reference 1.

Revision: 0 A reload DSS.CO evaluation has beon performxt in accordance Volith tho licensing rnGthodology 0) trip linear segment. povver intercopt WasP.. TRIP Was.p.sREAK RDF-Recirculation Drive F'low Revision: 0

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