|
|
| (4 intermediate revisions by the same user not shown) |
| Line 2: |
Line 2: |
| | number = ML16054A417 | | | number = ML16054A417 |
| | issue date = 01/26/2016 | | | issue date = 01/26/2016 |
| | title = Monticello - Revision 33 to the Updated Final Safety Analysis Report, Section 3, Reactor | | | title = 3 to the Updated Final Safety Analysis Report, Section 3, Reactor |
| | author name = | | | author name = |
| | author affiliation = Northern States Power Co, Xcel Energy | | | author affiliation = Northern States Power Co, Xcel Energy |
| Line 17: |
Line 17: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:SECTION 3 | | {{#Wiki_filter:}} |
| | |
| MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-03.01 SECTION 3 REACTOR Revision 31 Page 1 of 4 3.1 General Summary This section includes descriptions of the mechanical, thermal-hydraulic and nuclear characteristics of the current fuel load of the reactor. In addition, a functional description of the reactor control systems and reactor vessel internal components is given. The first of the following subsections (Section 3.2) presents a summary of design performance data for the core. In particular, the thermal-hydraulic characteristics are compared to the design criteria to demonstrate compliance. The thermal-hydraulic operating limits are described in Section 3.2.4. Performance data are presented in Section 3.2.5 for the various modes of operation. The safety limit and the boundaries on the Monticello operating range are discussed in Section 3.2.6. The results of transient analyses show a high degree of effectiveness of the protection system in preventing approach to levels of safety concern. Section 3.2.7 describes core performance capability. Section 3.3 describes the nuclear aspects of the fuel. Included in this section are analyses for the fuel cycle, reactivity control, and control rod worths. Also included are discussions of the reactivity coefficients and spatial xenon characteristics of the core. Section 3.4 describes the mechanical aspects of the fuel material (uranium dioxide), the fuel cladding, the fuel rods, and the arrangement of fuel rods in bundles. Section 3.5 describes the mechanical aspects of the movable control rods. These are provided to control reactivity. The control rod drive hydraulic system is designed so that sufficient energy is available to force the control rods into the core under conditions associated with abnormal operational transients and accidents. Control rod insertion speed is sufficient to prevent fuel damage as a result of any abnormal operational transient. Section 3.6 describes both the arrangement of the supporting structure for the core and reactor vessel internal components which are provided to properly distribute the coolant delivered to the reactor vessel. The reactor vessel internals are designed to allow the control rods and core standby cooling systems to perform their safety functions during abnormal operation transients and accidents. Detailed analytical comparisons of the thermal-hydraulic behavior for normal operation with typical and end-of-cycle power distributions respectively are provided. Data for development of the safety limit, the maximum safety system settings, and the limiting conditions for operation are included in the Technical Specifications. Certain AREVA safety analysis methods have been approved for use in Monticello Technical Specification Amendment 188. However, those methods are not invoked in the analysis-of-record until AREVA fuel is loaded in the core. Until that time, GEH (General Electric-Hitachi) safety analysis methods support core operation. Section 1.0 of the current Monticello COLR (Core Operating Limits Report) states whether GEH or AREVA methods support the current operating cycle. The physical description of the core, principal core design features and performance parameters for operation at 2004 MWt are presented in Table 3.1-1. The Design Basis Limits for Fission Product Barriers are located in Table 3.1-2 below.01101248 01496278 MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-03.01 Revision 31 Page 2 of 4 Table 3.1-1 Monticello Unit 1 Core Components Design (Page 1 of 2) Fuel Assembly Number of fuel assemblies: 484 Movable Control Rods Number 121 Shape Cruciform Pitch 12.0 inches Stroke 144 inches Width 9.75 inches Control Length 143 inches Control Material B4C granules in stainless steel tubes and sheath, and hafnium1 Number of Control Material 84 (original design, D-120 and D-140A), Tubes per Rod 72 (D-160 and D-230), or 56 (Marathon)
| |
| Tube Dimensions 0.188 inch OD x 0.025 inch wall (original design, D-120, D-140A and D-160) 0.220 inch OD x 0.020 inch wall (D-230) 0.298 inch OD x 0.024 inch wall (Marathon) Core Equivalent Core Diameter 149.0 inches Circumscribed Core Diameter 160.1 inches Core Lattice Pitch 12 inches THERMAL-HYDRAULIC2 General Operating Conditions Typical Power Distribution Design Thermal Output 2004 MWt Reactor Dome Pressure 1025 psia Steam Flow Rate 8.389 x 106lb/h Core Flow Rate 57.6 x 106lb/h Fraction of Power Appearing as Heat Flux 0.965 Power Density 48.3 KW/liter Maximum UO2 Temperature 2750°F Volumetric Average Fuel Temperature (typical) 925°F Power to Flow Ratio at <50 MWt/Mlbm/hr 100% Power and 99 % Core Flow
| |
| : 1. For Hybrid I Control Rods (D-160) see Figure 3.5-2a; for D-230 Control Blade see Figure 3.5-1a. 2. This data is for typical 100% power and flow condition and does not reflect data for reduced core flows in the MELLLA+ operating domain. 01101248 01101248 01457570 01457570 MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-03.01 Revision 31 Page 3 of 4 Table 3.1-1 Monticello Unit 1 Core Components Design (Page 2 of 2) Inlet Enthalpy 523.7 Btu/lb Core Average Exit Void Fraction 0.67 - 0.74 (varies with cycle exposure) Reactor Average Exit Quality 0.149 NUCLEAR DESIGN DATA Average Fuel Enrichment Bundle Type3 Wt% U-235 Cycle Loaded *DNAB392-16GZ 3.92 25 *DNAB375-16GZ 3.75 25 *DNAB392-16GZ 3.92 25 *DNAB373-16GZ 3.73 26 *DNAB391-16GZ 3.91 26 *DNAB391-15GZ 3.91 26 *DNAB391-12GZ 3.91 26 *DNAB372-17GZ 3.72 27 *DNAB386-16GZ 3.86 27 *DNAB386-16GZ 3.86 27 *DNAB389-11GZ 3.89 27 *DNAB387-16GZ 3.87 28 *DNAB389-11GZ 3.89 28 *DNAB384-15GZ 3.84 28 *DNAB374-16GZ 3.74 28 Excursion Parameters Prompt Neutron Lifetime 36 microseconds Effective Delayed Neutron Fraction 5.29 x 10-3
| |
| * Includes 6 inches of natural UO2 at bottom and 12 inches of natural UO2 at top of the fuel column. 3. The same bundle type identifier below does not necessarily indicate that the fuel pin enrichment layout or Gadolinia Pin enrichment and layout are the same. 01101248 01486387 01486387 MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-03.01 Revision 31 Page 4 of 4 Table 3.1-2 Design Basis Limits For Fission Product Barriers Monticello Boundary Design Bases Parameter Limit Reference Fuel Cladding MCPR 99.9% of fuel rods are prevented from experiencing transition boiling during abnormal operational transient USAR 3.2.4.3 Linear Heat Generation Rate GE14: 13.4 kW/ft during normal ops Core Operating Limits Report (use current revision) Fuel Enthalpy Not more than 280 cal/gram deposited during a reactivity accident. USAR 3.3.3.4 Cladding Strain Not more than 1% plastic strain during an anticipated operational occurrence USAR 3.4.3.2 Fuel Burnup - peak pellet GE14: 70 GWD/MTU Core Operating Limits Report (use current revision) Clad Temperature* Not more than 2200 °F during a Loss of Coolant Accident 10 CFR 50.46 Clad Oxidation* Not more than 17% local oxidation during a Loss of Coolant Accident. 10 CFR 50.46 RCS Boundary Pressure* 1332 psig steam dome pressure. (Higher pressures allowed at lower elevations. See TS.) TS 2.1.2 Stresses* ASME or ANSI B31.1 Code compliance for normal, upset, faulted conditions, etc., as appropriate for the accident. 10CFR50.55a Heat-up/Cool-down* Applicable ASME Code stress limits. 10CFR50.55a Primary Containment Pressure 62 psig maximum, 56 psig design. (ASME Code rating) USAR 5.2.1.1 *Limits controlled by 10CFR50.46, 10CFR50.55a and/or specific technical specifications require NRC approval to exceed or alter and would not be subject to evaluation under 10CFR50.59(c)(2)(vii). Notes on selection of Monticello DBLFPBs: 1. Fuel centerline melt is not a DBLFPB for BWRs because it is permitted in certain transients provided the cladding does not exceed 1% plastic strain as a result. See GNF Report NEDE-24011-P- 2. The 10CFR 50.46 limit of 1% overall clad oxidation in a LOCA is related to hydrogen generation and does not pertain to cladding integrity. 3. Reference MNGP Calc 01-SECTION 33.23.2.1 3.2.2
| |
| | |
| 3.2.3
| |
| | |
| 3.2.4
| |
| | |
| 3.2.5
| |
| | |
| 3.2.6
| |
| | |
| 3.2.7
| |
| | |
| SECTION 33.33.3.13.3.2
| |
| | |
| 3.3.3
| |
| | |
| 3.3.4
| |
| | |
| SECTION 33.4
| |
| | |
| 3.4.1
| |
| | |
| 3.4.2
| |
| | |
| 3.4.3
| |
| | |
| 3.4.4
| |
| | |
| SECTION 33.53.5.1 3.5.2
| |
| | |
| 3.5.3
| |
| | |
| 3.5.4
| |
| | |
| 3.5.5
| |
| | |
| SECTION 33.63.6.1
| |
| | |
| 3.6.2
| |
| | |
| 3.6.3
| |
| | |
| 3.6.4
| |
| | |
| SECTION 33.7
| |
| | |
| SECTION 3
| |
| | |
| SECTION 3
| |
| | |
| MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-03.01 SECTION 3 REACTOR Revision 31 Page 1 of 4 3.1 General Summary This section includes descriptions of the mechanical, thermal-hydraulic and nuclear characteristics of the current fuel load of the reactor. In addition, a functional description of the reactor control systems and reactor vessel internal components is given. The first of the following subsections (Section 3.2) presents a summary of design performance data for the core. In particular, the thermal-hydraulic characteristics are compared to the design criteria to demonstrate compliance. The thermal-hydraulic operating limits are described in Section 3.2.4. Performance data are presented in Section 3.2.5 for the various modes of operation. The safety limit and the boundaries on the Monticello operating range are discussed in Section 3.2.6. The results of transient analyses show a high degree of effectiveness of the protection system in preventing approach to levels of safety concern. Section 3.2.7 describes core performance capability. Section 3.3 describes the nuclear aspects of the fuel. Included in this section are analyses for the fuel cycle, reactivity control, and control rod worths. Also included are discussions of the reactivity coefficients and spatial xenon characteristics of the core. Section 3.4 describes the mechanical aspects of the fuel material (uranium dioxide), the fuel cladding, the fuel rods, and the arrangement of fuel rods in bundles. Section 3.5 describes the mechanical aspects of the movable control rods. These are provided to control reactivity. The control rod drive hydraulic system is designed so that sufficient energy is available to force the control rods into the core under conditions associated with abnormal operational transients and accidents. Control rod insertion speed is sufficient to prevent fuel damage as a result of any abnormal operational transient. Section 3.6 describes both the arrangement of the supporting structure for the core and reactor vessel internal components which are provided to properly distribute the coolant delivered to the reactor vessel. The reactor vessel internals are designed to allow the control rods and core standby cooling systems to perform their safety functions during abnormal operation transients and accidents. Detailed analytical comparisons of the thermal-hydraulic behavior for normal operation with typical and end-of-cycle power distributions respectively are provided. Data for development of the safety limit, the maximum safety system settings, and the limiting conditions for operation are included in the Technical Specifications. Certain AREVA safety analysis methods have been approved for use in Monticello Technical Specification Amendment 188. However, those methods are not invoked in the analysis-of-record until AREVA fuel is loaded in the core. Until that time, GEH (General Electric-Hitachi) safety analysis methods support core operation. Section 1.0 of the current Monticello COLR (Core Operating Limits Report) states whether GEH or AREVA methods support the current operating cycle. The physical description of the core, principal core design features and performance parameters for operation at 2004 MWt are presented in Table 3.1-1. The Design Basis Limits for Fission Product Barriers are located in Table 3.1-2 below.01101248 01496278 MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-03.01 Revision 31 Page 2 of 4 Table 3.1-1 Monticello Unit 1 Core Components Design (Page 1 of 2) Fuel Assembly Number of fuel assemblies: 484 Movable Control Rods Number 121 Shape Cruciform Pitch 12.0 inches Stroke 144 inches Width 9.75 inches Control Length 143 inches Control Material B4C granules in stainless steel tubes and sheath, and hafnium1 Number of Control Material 84 (original design, D-120 and D-140A), Tubes per Rod 72 (D-160 and D-230), or 56 (Marathon)
| |
| Tube Dimensions 0.188 inch OD x 0.025 inch wall (original design, D-120, D-140A and D-160) 0.220 inch OD x 0.020 inch wall (D-230) 0.298 inch OD x 0.024 inch wall (Marathon) Core Equivalent Core Diameter 149.0 inches Circumscribed Core Diameter 160.1 inches Core Lattice Pitch 12 inches THERMAL-HYDRAULIC2 General Operating Conditions Typical Power Distribution Design Thermal Output 2004 MWt Reactor Dome Pressure 1025 psia Steam Flow Rate 8.389 x 106lb/h Core Flow Rate 57.6 x 106lb/h Fraction of Power Appearing as Heat Flux 0.965 Power Density 48.3 KW/liter Maximum UO2 Temperature 2750°F Volumetric Average Fuel Temperature (typical) 925°F Power to Flow Ratio at <50 MWt/Mlbm/hr 100% Power and 99 % Core Flow
| |
| : 1. For Hybrid I Control Rods (D-160) see Figure 3.5-2a; for D-230 Control Blade see Figure 3.5-1a. 2. This data is for typical 100% power and flow condition and does not reflect data for reduced core flows in the MELLLA+ operating domain. 01101248 01101248 01457570 01457570 MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-03.01 Revision 31 Page 3 of 4 Table 3.1-1 Monticello Unit 1 Core Components Design (Page 2 of 2) Inlet Enthalpy 523.7 Btu/lb Core Average Exit Void Fraction 0.67 - 0.74 (varies with cycle exposure) Reactor Average Exit Quality 0.149 NUCLEAR DESIGN DATA Average Fuel Enrichment Bundle Type3 Wt% U-235 Cycle Loaded *DNAB392-16GZ 3.92 25 *DNAB375-16GZ 3.75 25 *DNAB392-16GZ 3.92 25 *DNAB373-16GZ 3.73 26 *DNAB391-16GZ 3.91 26 *DNAB391-15GZ 3.91 26 *DNAB391-12GZ 3.91 26 *DNAB372-17GZ 3.72 27 *DNAB386-16GZ 3.86 27 *DNAB386-16GZ 3.86 27 *DNAB389-11GZ 3.89 27 *DNAB387-16GZ 3.87 28 *DNAB389-11GZ 3.89 28 *DNAB384-15GZ 3.84 28 *DNAB374-16GZ 3.74 28 Excursion Parameters Prompt Neutron Lifetime 36 microseconds Effective Delayed Neutron Fraction 5.29 x 10-3
| |
| * Includes 6 inches of natural UO2 at bottom and 12 inches of natural UO2 at top of the fuel column. 3. The same bundle type identifier below does not necessarily indicate that the fuel pin enrichment layout or Gadolinia Pin enrichment and layout are the same. 01101248 01486387 01486387 MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-03.01 Revision 31 Page 4 of 4 Table 3.1-2 Design Basis Limits For Fission Product Barriers Monticello Boundary Design Bases Parameter Limit Reference Fuel Cladding MCPR 99.9% of fuel rods are prevented from experiencing transition boiling during abnormal operational transient USAR 3.2.4.3 Linear Heat Generation Rate GE14: 13.4 kW/ft during normal ops Core Operating Limits Report (use current revision) Fuel Enthalpy Not more than 280 cal/gram deposited during a reactivity accident. USAR 3.3.3.4 Cladding Strain Not more than 1% plastic strain during an anticipated operational occurrence USAR 3.4.3.2 Fuel Burnup - peak pellet GE14: 70 GWD/MTU Core Operating Limits Report (use current revision) Clad Temperature* Not more than 2200 °F during a Loss of Coolant Accident 10 CFR 50.46 Clad Oxidation* Not more than 17% local oxidation during a Loss of Coolant Accident. 10 CFR 50.46 RCS Boundary Pressure* 1332 psig steam dome pressure. (Higher pressures allowed at lower elevations. See TS.) TS 2.1.2 Stresses* ASME or ANSI B31.1 Code compliance for normal, upset, faulted conditions, etc., as appropriate for the accident. 10CFR50.55a Heat-up/Cool-down* Applicable ASME Code stress limits. 10CFR50.55a Primary Containment Pressure 62 psig maximum, 56 psig design. (ASME Code rating) USAR 5.2.1.1 *Limits controlled by 10CFR50.46, 10CFR50.55a and/or specific technical specifications require NRC approval to exceed or alter and would not be subject to evaluation under 10CFR50.59(c)(2)(vii). Notes on selection of Monticello DBLFPBs: 1. Fuel centerline melt is not a DBLFPB for BWRs because it is permitted in certain transients provided the cladding does not exceed 1% plastic strain as a result. See GNF Report NEDE-24011-P- 2. The 10CFR 50.46 limit of 1% overall clad oxidation in a LOCA is related to hydrogen generation and does not pertain to cladding integrity. 3. Reference MNGP Calc 01-SECTION 33.23.2.1 3.2.2
| |
| | |
| 3.2.3
| |
| | |
| 3.2.4
| |
| | |
| 3.2.5
| |
| | |
| 3.2.6
| |
| | |
| 3.2.7
| |
| | |
| SECTION 33.33.3.13.3.2
| |
| | |
| 3.3.3
| |
| | |
| 3.3.4
| |
| | |
| SECTION 33.4
| |
| | |
| 3.4.1
| |
| | |
| 3.4.2
| |
| | |
| 3.4.3
| |
| | |
| 3.4.4
| |
| | |
| SECTION 33.53.5.1 3.5.2
| |
| | |
| 3.5.3
| |
| | |
| 3.5.4
| |
| | |
| 3.5.5
| |
| | |
| SECTION 33.63.6.1
| |
| | |
| 3.6.2
| |
| | |
| 3.6.3
| |
| | |
| 3.6.4
| |
| | |
| SECTION 33.7
| |
| | |
| SECTION 3}}
| |
