RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 4.0, Reactor
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4.1-5 Reformatted October 2016 TABLE 4.1-1 REACTOR DESIGN PARAMETERS Thermal and Hydraulic Design Parameters Current RTP and Engineered Safeguards Design Rating
- 1. Reactor Core Heat Output, MWt 2900 2. Reactor Core Heat Output, 10 6 Btu/hr 9898 3. Heat Generated in Fuel, %
97.4 4. System Pressure, Nominal, psia 2250 5. System Pressure, Minimum Steady
-State, psia 2200 6. Minimum Departure from Nucleate Boiling Ratio for Design Transients (Design Limit) 1.23 (Typical Flow Channel) 1.23 (Thimble Flow Channel) 6a. DNB Correlation Used WRB-2 COOLANT FLOW
- 7. Total Thermal Design Flow Rate (TDF), 10 6 lbm/hr 104.0 7a. Minimum Measured Flow Rate, 10 6 lbm/hr 106.1 8. Effective Flow Rate for Heat Transfer, 10 6 lb/hr 94.74 9. Effective Flow Area for Heat Transfer, ft 2 44.0 10. Average Velocity Along Fuel Rods, ft/sec 13.7 11. Average Mass Velocity, 10 6 lbm/hr-ft 2 2.15 01-113 0 1-113 01-113 4.1-6 Reformatted October 2016 TABLE 4.1-1 (Continued)
REACTOR DESIGN PARAMETERS Thermal and Hydraulic Design Parameters Current RTP and Engineered Safeguards Design Rating COOLANT TEMPERATURE, F (BASED ON TDF)
- 12. Nominal Inlet 552.9 13. Average Rise in Vessel 69.0 14. Average Rise in core 74.8 15. Average in core 592.8 16. Average in Vessel 587.4 HEAT TRANSFER
- 17. Active Heat Transfer, Surface Area, ft 2 46,780 18. Average Heat Flux, BTU/hr
-ft 2 206,100 19. Maximum Heat Flux for Normal Operation, Btu/hr
-ft 2 504,900 20. Average Thermal Output, kW/ft 5.69 21. Maximum Thermal Output for Normal Operation, kW/ft 13.9 (3) 22. Peak Linear Power for Determination of Protection Setpoints, kW/ft <22.6 (1) 23. Heat Flux Hot Channel Factor, F Q 2.45 (2) 24. Peak Fuel Central Temperature at Maximum Thermal Output for Maximum Overpower Trip Point, F <4700 4.1-7 Reformatted October 2016 TABLE 4.1-1 (Continued)
REACTOR DESIGN PARAMETERS Core Mechanical Design Paramete rs VANTAGE+ Fuel Assembly FUEL ASSEMBLIES
- 25. Design RCC Canless
- 26. Number of Fuel Assemblies 157 27. UO 2 Rods per Assembly 264 28. Rod Pitch, in.
0.496 29. Overall Dimensions, in.
8.426 x 8.426
- 30. Fuel Weight (as UO 2), lb 174,894 31. Clad Weight, lb 35,244 32. Number of Grids per Assembly 2 - Type R (Inconel End) 6 - Zircaloy (Internal) 1 - Protective
- 33. Number of IFMs per Assembly 3 34. Loading Technique multi-region nonuniform
-checkered FUEL RODS 35. Number 41,448 36. Outside Diameter, in. 0.360 37. Diametral Gap, in.
0.0062 38. Clad Thickness, in.
0.0225 39. Clad Material ZIRLO/Optimized ZIRLO TM 9 6-043 0 1-113 RN 11-013 4.1-8 Reformatted October 2016 TABLE 4.1-1 (Continued)
REACTOR DESIGN PARAMETERS Core Mechanical Design Parameters VANTAGE+ Fuel Assembly FUEL PELLETS
- 40. Material UO 2 Sintered 41. Density (% of Theoretical) 95 42. Diameters, in. OD/ID 0.3088/-(Enriched) 0.3088/0.155 (Unenriched/Midenriched)
- 43. Length, in.
0.370 (Enriched) 0.458 (Unenriched/Midenriched)
ROD CLUSTER CONTROL ASSEMBLIES
- 44. Neutron Absorber Ag-In-Cd 45. Cladding Material Type 304 SS-Cold Worked
- 46. Clad Thickness, in.
0.0185 47. Number of Clusters 48 48. Number of Absorber Rods per Cluster 24 CORE STRUCTURE
- 49. Core Barrel, I.D./O.D., in.
133.85/137.875
- 50. Thermal Shield, I.D./O.D., in.
Neutron Pad Design 0 1-113 0 1-113 0 1-113 98-01 98-01 4.1-9 Reformatted October 2016 TABLE 4.1-1 (Continued)
REACTOR DESIGN PARAMETERS Core Mechanical Design Parameters VANTAGE+ Fuel Assembly STRUCTURE CHARACTERISTICS
- 51. Core Diameter, in. (Equivalent) 119.7 52. Core Height, in. (Active Fue l) 144 REFLECTOR THICKNESS AND COMPOSITION
- 53. Top - Water plus Steel, in.
10 54. Bottom - Water plus Steel, in.
10 55. Side - Water plus Steel, in.
15 56. H 2O/U Molecular Ratio core, Lattice (Cold) 2.73
(1) See Section 4.3.2.2.6
(2) This is the value of F Q for normal operation (3) This limit associated with an F Q = 2.45 0 1-113 95-022 4.1-10 Reformatted October 2016 TABLE 4.1-2 ANALYTIC TECHNIQUES IN CORE DESIGN Analysis Technique Computer Code Section Referenced Mechanical Design of Core Internals Loads, Deflections, and Stress Analysis Static and dynamic modeling Blowdown code, FORCE, finite element structural analysis code, and others 3.7.2.1 3.9.1 3.9.3 Fuel Rod Design Fuel Performance Characteristics (temperature, internal pressure, clad stress, etc.)
Semi-empirical thermal model of fuel rod with consideration of fuel density changes, heat transfer, fission gas release, etc.
Westinghouse fuel rod design model 4.2.1.3.1 4.3.3.1 4.4.2.2 4.4.3.4.2 Nuclear Design
- 1. Cross Sections and Group Constants Microscopic data; Macroscopic constants for homogenized core regions PHOENIX , NEXUS, PARAGON 4.3.3.2 4.3.3.2 Group constants for control rods with self
-shielding PHOENIX, NEXUS, PARAGON 4.3.3.2 2. X-Y-Z Power Distributions, Fuel Depletion, Critical Boro n Concentrations, x
-y Xenon Distributions, Reactivity Coefficients 3-D, Advanced Nodal Theory ANC 4.3.3.3 RN 15-022 RN 15-022 4.1-11 Reformatted October 2016 TABLE 4.1-2 (Continued)
ANALYTIC TECHNIQUES IN CORE DESIGN Analysis Technique Computer Code Section Referenced
- 3. Axial Power Distributions, Control Rod Worths, and Axial Xenon Distribution 1-D, 2-Group Diffusion Theory 3D, Diffusion Theory APOLLO NEXUS, PARAGON, ANC 4.3.3.3 4.3.3.3 4. Fuel Rod Power Integral Transport Theory LASER 4.3.3.1 Effective Resonance Temperature Monte Carlo Weighting Function REPAD Thermal-Hydraulic Design
- 1. Steady-State Subchannel analysis of local fluid conditions in rod bundles, including inertial and crossflow resistance terms, solution progresses from core-wide to hot assembly to hot channel THINC-IV 4.4.3.4.1 2. Transient Departure from Nucleate Boiling Analysis Subchannel analysis of local fluid conditions in rod bundles during transients by including accumulation terms in conservation equations THINC-I (THINC-III) 4.4.3.4.1 0 1-113 0 1-113 RN 15-022 9 5-022 4.1-12 Reformatted Per October 2016 TABLE 4.1-3 DESIGN LOADING CONDITIONS FOR REACTOR CORE COMPONENTS
- 1. Fuel Assembly Weight
- 2. Fuel Assembly Spring Forces
- 3. Internals Weight
- 4. Control Rod Trip (equivalent static load)
- 5. Differential Pressure
- 6. Spring Preloads
- 7. Coolant Flow Forces (stati c) 8. Temperature Gradients
- 9. Differences in Thermal Expansion
- a. Due to temperature differences
- b. Due to expansion of different materials
- 10. Interference Between Components
- 11. Vibration (mechanically or hydraulically induced)
- 12. One or More Loops Out of Service
- 13. All Operational Transients Listed in Table 5.2
-2 14. Pump Overspeed
- 15. Seismic Loads (operating basis earthquake and safe shutdown earthquake)
- 16. Blowdown Forces (due to cold and hot leg break)
4.2-1 Reformatted February 2018 4.2 MECHANICAL DESIGN The plant conditions for design are divided into 4 categories in accordance with their anticipated frequency of occurrence and risk to the public: Condition I - Normal Operation; Condition II - Incidents of Moderate Frequency; Condition III - Infrequent Incidents; Condition IV - Limiting Faults.
The reactor is designed so that its components meet the following performance and safety criteria:
- 1. The mechanical design of the reactor core components and their physical arrangement, together with corrective actions of the reactor control, protection and emergency cooling systems (when applicable) assure that:
- a. Fuel damage (1) is not expected during Condition I and Condition II events. It is not possible, however, to preclude a very small number of rod failures. These are within the capability of the plant cleanup system and are consistent with plant design bases.
- b. The reactor can be brought to a safe state following a Condition III event with only a small fraction of fuel rods damaged
[1] although sufficient fuel damage might occur to preclude resumption of operation without considerable outage time. c. The reactor can be brought to a safe state and the core can be kept subcritical with acceptable heat transfer geometry following transients arising from Condition IV events.
- 2. The fuel assemblies are designed to accommodate expected conditions for handling during assembly inspection, refueling operations and shipping loads.
- 3. The fuel assemblies are designed to accept control rod insertions in order to provide the required reactivity control for power operations and reactivity shutdown conditions.
- 4. All fuel assemblies have provisions for the insertion of incore instrumentation necessary for plant operation.
(1) Fuel damage as used here is defined as penetration of the fission product barrier (i.e., the fuel rod clad).
4.2-2 Reformatted February 2018
- 5. The reactor internals, in conjunction with the fuel assemblies, direct reactor coolant through the core to achieve acceptable flow distribution and to restrict bypass flow so that the heat transfer performance requirements can be met for all modes of operation. In addition, the internals provide core support and distribute coolant flow to the pressure vessel head so that the temperature differences between the vessel flange and head do not result in leakage from the flange during the Condition I and II modes of operation. Required inservice inspection can be carried out as the internals are removable and provide access to the inside of the pressure vessel.
4.2.1 FUEL 4.2.1.1 Design Bases The fuel rod and fuel assembly design bases are established to satisfy the general performance and safety criteria presented in Section 4.2 and specific criteria noted below. 4.2.1.1.1 Fuel Rods The integrity of the fuel rods is ensured by designing to prevent excessive fuel temperatures, excessive internal rod gas pressures due to fission gas releases, and excessive cladding stresses and strains. This is achieved by designing the fuel rods so that the following conservative design bases are satisfied during Condition I and Condition II events over the fuel lifetime:
- 1. Fuel Pellet Temperatures
- The center temperature of the hottest pellet is to be below the melting temperature of the UO 2 (melting point of 5080 F [1] unirradiated and decreasing by 58F per 10,000 MWD/MTU). While a limited amount of center melting can be tolerated, the design conservatively precludes center melting. A calculated fuel centerline temperature of 4700F has been selected as an overpower limit to assure no fuel melting. This provides sufficient margin for uncertainties as described in Sections 4.4.1.2 and 4.4.2.10.1.
- 2. Internal Gas Pressure
- The internal pressure of the lead rod in the reactor will be limited to a value below that which could cause: 1) the diametral gap to increase due to outward cladding creep during steady state operation, and 2) extensive DNB propagation to occur. Reference [20] shows that the DNB propagation criteria is satisfied.
- 3. Clad Stress
- The clad stresses are less than the clad yield stress, with due consideration for temperature and irradiation effects. While the clad has some capability for accommodating plastic strain, the yield stress has been accepted as a conservative design basis.
RN 95-022 96-043 RN 95-022 4.2-3 Reformatted February 2018 An alternative criterion has been developed to evaluate cladding stress in Reference [23]. Per the revised stress criterion, maximum cladding stress intensities excluding pellet clad interaction (PCI) induced stress will be evaluated using ASME pressure vessel guidelines. Cladding corrosion is accounted for as a loss of load carrying material. Stresses are combined to calculate a maximum stress intensity which is then compared to criteria based on the ASME code (Addendum 1
-A of Reference [23]).
- 4. Clad Tensile Strain
- The clad tensile strain is less than 1%. This limit is consistent with proven practice.
- 5. Strain Fatigue
- The cumulative strain fatigue cycles are less than the design strain fatigue life.
This basis is consistent with proven practice.
The preceding fuel rod design bases and other supplementary fuel design criteria/limits are given in Section 2 of Reference [23], Section 2 of Reference [27], Section A of Reference [7], and Appendix B of Reference [27]. The detailed fuel rod design establishes such parameters as pellet size and density, clad
-pellet diametral gap, gas plenum size, and helium pre
-pressure. The design also considers effects such as fuel density changes, fission gas release, clad creep, and other physical properties which vary with burnup.
An extensive irradiation testing and fuel surveillance operational experience program has been, and continues to be, conducted to verify the adequacy of the fuel performance and design bases. This program is discussed in Section 4.2.1.3.3. Fuel surveillance and testing results, as they become available, are used to improve fuel rod design and manufacturing processes and assure that the design bases and safety criteria are satisfied.
4.2.1.1.2 Fuel Assembly Structure Structural integrity of the fuel assemblies is assured by setting limits on stresses and deformations due to various loads and by determining that the assemblies do not interfere with the functioning of other components. Three (3) types of loads are considered
. 1. Nonoperational loads such as those due to shipping and handling.
- 2. Normal and abnormal loads which are defined for Conditions I and II. 3. Abnormal loads which are defined for Conditions III and IV.
These criteria are applied to the design and evaluation of the top and bottom nozzles, the guide thimbles, grids and thimble joints.
RN 95-022 RN 95-022 14-036 RN 14-005 4.2-4 Reformatted February 2018 The design bases for evaluating the structural integrity of the fuel assemblies are:
- 1. Nonoperational (e.g., shipping)
- 4 g axial loading and 6 g lateral loading with dimensional stability.
- 2. For the normal operating and upset conditions, the fuel assembly component structural design criteria are classified into 2 material categories, namely austenitic steels and zirconium alloys. The stress categories and strength theory presented in the ASME Boiler and Pressure Vessel Code,Section III, are used as a general guide. The maximum shear
-theory (Tresca criterion) for combined stresses is used to determine the stress intensities for the austenitic steel components. The stress intensity is defined as the numerically largest difference between the various principal stresses in a 3 dimensional field. The allowable stress intensity value for austenitic steels, such as nickel
-chromium-iron alloys, is given by the lowest of the following:
- a. 1/3 of the specified minimum tensile strength or 2/3 of the specified minimum yield strength at room temperature;
- b. 1/3 of the tensile strength or 90% of the yield strength at temperature but not to exceed 2/3 of the specified minimum yield strength at room temperature.
The stress limits for the austenitic steel components are given below. All stress nomenclature is per the ASME Boiler and Pressure Vessel Code,Section III. Stress Intensity Limits Categories Limit General Primary Membrane Stress Intensity Sm Local Primary Membrane Stress Intensity 1.5 Sm Primary Membrane plus Bending Stress Intensity 1.5 Sm Total Primary plus Secondary Stress Intensity 3.0 Sm The zirconium alloy structural components which consist of guide thimble and fuel tubes are in turn subdivided into 2 categories because of material differences and functional requirements. The fuel tube design criteria are covered separately in Section 4.2.1.1.1. The maximum shear theory is used to evaluate the guide thimble design. For conservative purposes, the zirconium alloy unirradiated properties are used to define the stress limits.
- 3. Abnormal loads during Conditions III or IV
- worst cases represented by combined seismic and blowdown loads.
- a. Deflections or failures of components cannot interfere with the reactor shutdown or emergency cooling of the fuel rods.
RN 96-043 RN 95-022 96-043 09-022 4.2-5 Reformatted February 2018
- b. The fuel assembly structural component stresses under faulted conditions are evaluated using primarily the methods outlined in Appendix F of the ASME Boiler and Pressure Vessel Code,Section III. Since the current analytical methods utilize elastic analysis, the stress allowables are defined as the smaller value of 2.4 Sm or 0.70 Su for primary membrane and 3.6 Sm or 1.05 Su for primary membrane plus primary bending. For the austenitic steel fuel assembly components, the stress intensity is defined in accordance with the rules described in the previous section for normal operating conditions. For the zirconium alloy components, the stress intensity limits are set at 2/3 of the material yield strength, Sy, at reactor operating temperature. This results in stress limits being the smaller of 1.6 Sy or 0.70 Su for primary membrane and 2.4 Sy or 1.05 Su for primary membrane plus bending. For conservative purposes, the zirconium alloy unirradiated properties are used to define the stress limits.
The grid component strength criteria are based on experimental tests. For grids, the limit is established based on the 95% confidence level on the true mean of the test data (experimental collapse load) at operating temperature.
4.2.1.2 Design Description Fuel assembly and fuel rod design data are given in Table 4.1
-1. NRC approval of the VANTAGE+ design was given in Reference [29].
Two hundred sixty four (264) fuel rods, or variations of fuel rods and filler rods (with respect to fuel assembly reconstitution, Reference [30]), 24 guide thimble tubes and 1 instrumentation thimble tube are arranged within a supporting structure to form a fuel assembly. The instrumentation thimble is located in the center position and provides a channel for insertion of an incore neutron detector, if the fuel assembly is located in an instrumented core position.
The guide thimbles provide channels for insertion of either a rod cluster control assembly, a neutron source assembly, a burnable absorber assembly or a plugging device (if used), depending on the position of the particular fuel assembly in the core. Figure 4.2
-1 shows a cross
-section of the fuel assembly array, and Figure 4.2-2 shows fuel assembly full length view. The fuel rods are loaded into the fuel assembly structure so that there is clearance between the fuel rod ends and the top and bottom nozzles.
Since the VANTAGE+ fuel is intended to replace either the Westinghouse LOPAR, Optimized, or VANTAGE 5 fuel designs, the VANTAGE+ exterior assembly envelope is equivalent to those of previous Westinghouse fuel designs. Also, the VANTAGE+ fuel assembly is designed to be mechanically and hydraulically compatible with the LOPAR, Optimized and VANTAGE 5 designs in full or transition cores, and the same functional requirements and design criteria as previously established for the Westinghouse VANTAGE 5 fuel assembly remain valid for the VANTAGE+ fuel assembly, Reference [29]. The VANTAGE+ fuel assembly design is provided in Table 4.1
-1. Figure 4.2
-2 compares the VANTAGE 5 and VANTAGE+ fuel assembly designs.
RN 95-022 96-043 RN 96-043 09-022 RN 95-022 RN 95-022 96-043 RN 95-022 96-043 RN 96-043 4.2-6 Reformatted February 2018 Each fuel assembly is installed vertically in the reactor vessel and stands upright on the lower core plate, which is fitted with alignment pins to locate and orient the assembly. After all fuel assemblies are set in place, the upper support structure is installed. Alignment pins, built into the upper core plate, engage and locate the upper ends of the fuel assemblies.
The upper core plate then bears downward against the fuel assembly top nozzle via the holddown springs to hold the fuel assemblies in place.
4.2.1.2.1 Fuel Rods The fuel rods consist of uranium dioxide ceramic pellets contained in slightly cold worked zirconium alloy tubing which is plugged and seal welded at the ends to encapsulate the fuel. A schematic of the fuel rods is shown in Figure 4.2
-3. The fuel pellets are right circular cylinders consisting of uranium dioxide powder which has been compacted by cold pressing and then sintered to the required density. The fuel rods also contain, annular Axial Blanket pellets, and an Integral Fuel Burnable Absorber (IFBA) coating on some of the enriched fuel pellets.
The bottom end plug has an internal grip feature to facilitate rod loading and provides appropriate lead
-in for the removable top nozzle reconstitution feature.
A chamfered pellet design is used with the objective of improving manufacturability while maintaining or improving performance (e.g., improved pellet chip resistance during manufacturing/handling).
The axial blankets are a nominal 6 inches of fuel pellets at each end of the fuel rod pellet stack. VANTAGE+ fuel uses unenriched annular axial blanket pellets. VANTAGE+ fuel with Performance+ features uses mid
-enriched annular axial blanket pellets. Axial blankets reduce neutron leakage and improve fuel utilization. The axial blankets utilize chamfered pellets physically different than the pellets, which help to prevent accidental mixing during manufacturing. The physical difference includes a longer pellet length than the pellet (see Table 4.1
-1). The IFBA coated fuel pellets are identical to the enriched uranium dioxide pellets except for the addition of a thin boride coating on the pellet cylindrical surface. Coated pellets occupy the central portion of the fuel column. The number and pattern of IFBA rods within an assembly may vary depending on specific application. The ends of the enriched coated pellets and enriched uncoated pellets are dished to allow for greater axial expansion at the pellet centerline and void volume for fission gas release. An evaluation and test program for the IFBA design features are given in Section 2.5 in Reference [24]. RN 95-022 96-043 RN 95-022 96-043 01-113 RN 95-022 96-043 01-113 4.2-7 Reformatted February 2018 To avoid overstressing of the clad or seal welds, void volume and clearances are provided within the rods to accommodate fission gases released from the fuel, differential thermal expansion between the clad and the fuel, and fuel density changes during burnup. Shifting of the fuel within the clad during handling or shipping prior to core loading is prevented by a stainless steel helical spring which bears on top of the fuel. At assembly the pellets are stacked in the clad to the required fuel height, the spring is then inserted into the top end of the fuel tube and the end plugs pressed into the ends of the tube and welded. All fuel rods are internally pressurized with helium during the welding process in order to minimize compressive clad stresses and creep due to coolant operating pressures.
The fuel rods are presently being pre
-pressurized and designed so that the internal gas pressure design basis (Section 4.2.1.1.1) is satisfied and clad flattening will not occur during the fuel core life as determined by the methods in Reference
[3]. 4.2.1.2.2 Fuel Assembly Structure The fuel assembly structure consists of a bottom nozzle, top nozzle, guide thimbles, and grids, as shown in Figure 4.2
-2. 4.2.1.2.2.1 Bottom Nozzle The bottom nozzle is a box
-like structure which serves as a bottom structural element of the fuel assembly and directs the coolant flow distribution to the assembly. The VANTAGE+ with Performance + features includes a 2 piece composite bottom nozzle design incorporating a highly machined stainless steel adaptor plate welded to a low cobalt investment casting as shown in Figure 4.2
-2. The plate itself acts to prevent a downward ejection of the fuel rods from their fuel assembly. The bottom nozzle is fastened to the fuel assembly guide tubes by weld
-locked screws which penetrate through the nozzle and mate with an inside fitting in each guide tube.
The debris filter bottom nozzle (DFBN) is used to reduce the possibility of fuel rod damage due to debris-induced fretting. The relatively large flow holes in a conventional bottom nozzle are replaced with a new pattern of smaller flow holes for the DFBN. The holes are sized to minimize passage of debris particles large enough to cause damage while providing sufficient flow area, comparable pressure drop, and continued structural integrity of the nozzle. Tests to measure pressure drop and demonstrate structural integrity have been performed to verify that the DFBN is totally compatible with the current design. Coolant flow through the fuel assembly is directed from the plenum in the bottom nozzle upward through the penetrations in the plate to the channels between the fuel rods. The penetrations in the plate are positioned between the rows of the fuel rods. RN 95-022 96-043 RN 96-043 RN 95-022 RN 95-022 4.2-8 Reformatted February 2018 Axial loads (holddown) imposed on the fuel assembly and the weight of the fuel assembly are transmitted through the bottom nozzle to the lower core plate. Indexing and positioning of the fuel assembly is controlled by alignment holes in 2 diagonally opposite bearing plates which mate with locating pins in the lower core plate. Any lateral loads on the fuel assembly are transmitted to the lower core plate through the locating pins.
4.2.1.2.2.2 Top Nozzle The top nozzle assembly functions as the upper structural element of the fuel assembly in addition to providing a partial protective housing for the rod cluster control assembly or other components. It consists of an adapter plate, enclosure, top plate, and pads. The integral welded assembly has holddown springs mounted on the assembly as shown in Figure 4.2
-2. The springs and bolts are made of Inconel
-718 and Inconel
-600, respectively, whereas other components are made of Type 304 stainless steel.
The square adapter plate is provided with round penetrations and semicircular ended slots to permit the flow of coolant upward through the top nozzle. Other round holes are provided to accept sleeves which are welded to the adapter plate and mechanically attached to the thimble tubes.
The ligaments in the plate cover the tops of the fuel rods and prevent their upward ejection from the fuel assembly. The enclosure is a box
-like structure which sets the distance between the adapter plate and the top plate. The top plate has a large square hole in the center to permit access for the control rods and the control rod spiders. Holddown springs are mounted on the top plate and are fastened in place by bolts and clamps located at 2 diagonally opposite corners. On the other 2 corners integral pads are positioned which contain alignment holes for locating the upper end of the fuel assembly.
To remove the top nozzle, a tool is first inserted through a lock tube and expanded radially to engage the bottom edge of the tube. An axial force is then exerted on the tool which overrides local lock tube deformations and withdraws the lock tube from the insert. After the lock tubes have been withdrawn, the nozzle is removed by raising it off the upper slotted ends of the nozzle inserts which deflect inwardly under the axial lift load. With the top nozzle removed, direct access is provided for fuel rod examinations or replacement. Reconstitution is completed by the remounting of the nozzle and the insertion of lock tubes. Additional details of this design feature, the design bases and evaluation of the reconstitutable top nozzle are given in Section 2.3.2 in Reference
[24]. 4.2.1.2.2.3 Guide and Instrument Thimbles The guide thimbles are structural members which also provide channels for the neutron absorber rods, burnable absorber rods, thimble plugs, or neutron source assemblies. Each one is fabricated from zirconium alloy tubing having 2 different diameters. The larger diameter at the top provides a relatively large annular area to permit rapid insertion of the control rods during a reactor trip as well as to accommodate the flow of coolant during normal operation. Four (4) holes are provided on the thimble tube above the dashpot to reduce the rod drop time. The lower portion of each guide thimble has a RN 95-022 RN 95-022 96-043 RN 95-022 96-043 4.2-9 Reformatted February 2018 reduced diameter to produce a dashpot action near the end of the control rod travel during normal operation and to accommodate the outflow of water from the dashpot during a reactor trip. The dashpot is closed at the bottom by means of an end plug which is provided with a small flow port to avoid fluid stagnation in the dashpot volume during normal operation. The top end of the guide thimble is fastened to a tabular insert by 3 expansion swages. The insert engages into the top nozzle and is secured into position by a lock tube.
The lower end of the guide thimble is fitted with an end plug which is then fastened into the bottom nozzle by a weld locked screw.
Each grid is fastened to the guide thimble assemblies to create an integrated structure. Westinghouse has chosen one 4 lobe bulge at each thimble and instrumentation tube location for the fastening technique as depicted in Figures 4.2
-4 and 4.2-5 for all but the top and bottom grids.
An expanding tool is inserted into the inner diameter of the thimble tube to the elevation of the sleeves that have been welded to the middle grid assemblies. The 4 lobed tool forces the thimble and sleeve outward to a predetermined diameter, thus joining the 2 components.
For the VANTAGE+ assembly, the top grid to nozzle attachment is shown in Figure 4.2-6. The thimbles are fastened to the top nozzle inserts by expanding the members as shown in Figures 4.2
-4 and 4.2-5. The inserts then engage the top nozzle and are secured into position by the insertion of lock tubes.
The bottom grid assembly is not mechanically fastened as described above, but rather is joined to the assembly as shown in Figure 4.2
-7. The stainless steel insert is spotwelded to the bottom grid and later captured between the guide thimble end plug and the bottom nozzle by means of a stainless steel thimble screw.
The central instrumentation thimble of each fuel assembly is not attached to either the top or bottom nozzles, but the thimble is constrained by its seating in counterbores of diameter does not vary, and incore neutron This thimble is expanded at the top and mid grids in the same manner as the previously discussed expansion of the guide thimbles to the grids.
The guide thimble tube ID provides an adequate nominal diametral clearance for the control rods. The thimble tube ID also provides sufficient diametral clearance for burnable absorber rods, source rods, and dually compatible thimble plugs (see Figures 4.2-20 and 4.2
-2). The instrumentation tube also has a sufficient diametral clearance for the flux thimble to traverse the tube without binding.
RN 95-022 RN 95-022 96-043 RN 95-022 96-043 RN 96-043 RN 95-022 96-043 RN 95-022 96-043 RN 95-022 96-043 4.2-10 Reformatted February 2018 The IFM grids employ a single bulge connection to the sleeve and thimble just as is used in the intermediate grids.
4.2.1.2.2.4 Grid Assemblies The fuel rods, as shown in Figure 4.2
-2, are supported laterally at intervals along their length by grid assemblies which maintain the lateral spacing between the rods throughout the design life of the assembly. Each fuel rod is afforded lateral support within each grid by the combination of support dimples and springs. The grid assembly consists of individual slotte-crate" arrangement to join the straps permanently at their points of intersection. The straps contain spring fingers, support dimples and mixing vanes.
The grid material is zirconium alloy (mid
-grids) and Inconel 718 (end grids) for VANTAGE+ assemblies. Inconel 718 was chosen for its corrosion resistance assembly and high strength properties. Zirconium Alloy was chosen for the VANTAGE+ assembly mid-grids due to its corrosion resistance, low neutron cross
-section, and adequate strength properties. The magnitude of the grid restraining force on the fuel rod is set high enough to minimize possible fretting, without overstressing the cladding at the points of contact between the grids and fuel rods. The grid assemblies also allow axial thermal expansion of the fuel rods without imposing restraint sufficient to develop buckling or distortion of the fuel rods.
For VANTAGE+ fuel with Performance + features, 4 types of grid assemblies are used. The top and bottom Inconel (non
-mixing vane) grids of the VANTAGE+ fuel assemblies are used. Six (6) intermediate (mixing vane) structural grids are used. They are made of Zirconium Alloy material. The third grid type for the VANTAGE+ assembly is an Intermediate Flow Mixer, also made of Zirconium Alloy. These Intermediate Flow Mixer (IFM) grids shown in Figure 4.2
-2 are located in the 3 uppermost spans between the Zirconium Alloy mixing vane structural grids and incorporate a similar mixing vane array. Their prime function is mid
-span flow mixing in the hottest fuel assembly spans.
Each IFM grid cell contains 4 dimples which are designed to prevent mid
-span channel closure in the spans containing IFMs and fuel rod contact with the mixing vanes. This simplified cell arrangement allows short grid cells so that the IFM grid can accomplish its flow mixing objective with minimal pressure drop.
The VANTAGE+ IFM grids are not intended to be structural members. The outer strap configuration is designed to preclude grid hang
-up and damage during fuel handling. Additionally, the grid envelope is smaller which further minimizes the potential for damage and reduces calculated forces during seismic/LOCA events.
A fourth type of grid used in VANTAGE+ fuel with Performance + features is a Robust Protective Grid (RPG). The grid is an Inconel 718 grid that is mounted below the bottom grid (on top of the bottom nozzle) and acts to strain out foreign materials before they can flow up the fuel assembly, be trapped by the other grids and cause damage to the fuel rods.
RN 95-022 96-043 RN 95-022 RN 96-043 11-013 RN 95-022 09-022 11-013 RN 95-022 96-043 11-013 RN 95-022 96-043 11-013 RN 95-022 RN 95-022 4.2-11 Reformatted February 2018 4.2.1.3 Design Evaluation 4.2.1.3.1 Fuel Rods The fuel rods are designed to assure the design bases are satisfied for Condition I and II events. This assures that the fuel performance and safety criteria (Section 4.2) are satisfied.
4.2.1.3.1.1 Materials
- Fuel Cladding The desired fuel rod clad is a material which has a superior combination of neutron economy (low absorption cross section), high strength (to resist deformation due to differential pressures and mechanical interaction between fuel and clad), high corrosion resistance (to coolant, fuel and fission products), and high reliability. ZIRLO
/Optimized ZIRLO TM has this desired combination of clad properties. As shown in Reference
[4], there is considerable PWR operating experience on the capability of Zircaloy and ZIRLO as a clad material. Optimized ZIRLO TM has the same chemical composition as ZIRLO with the exception of a lower tin content for improved corrosion resistance , Reference [16].
Clad hydriding has not been a significant cause of clad perforation since current controls on levels of fuel contained moisture were instituted , Reference [4]. Metallographic examination s of irradiated commercial fuel rods have shown occurrences of fuel/clad chemical interaction. Reaction layers of < 1 mil in thickness have been observed between fuel and clad at limited points around the circumference. Westinghouse metallographic data indicates that this interface layer remains very thin even at high burnup. Thus, there is no indication of propagation of the layer and eventual clad penetration.
Stress corrosion cracking is another postulated phenomenon related to fuel/clad chemical interaction. Reaction tests have shown that in the presence of high clad tensile stresses, large concentrations of iodine can chemically attack the Zirconium-based alloy tubing and can lead to eventual clad cracking. Westinghouse has no i n reactor evidence that this mechanism is operative in commercial fuel.
4.2.1.3.1.2 Materials
- Fuel Pellets Sintered, high density uranium dioxide fuel reacts only slightly with the clad, at core operating temperatures and pressures. In the event of clad defects, the high resistance of uranium dioxide to attack by water protects against fuel deterioration although limited fuel erosion can occur. As has been shown by operating experience and extensive experimental work, the thermal design parameters conservatively account for changes in the thermal performance of the fuel elements due to pellet fracture which may occur during power operation. The consequences of defects in the clad are greatly reduced by the ability of uranium dioxide to retain fission products including those which are gaseous or highly volatile. Observations from several operating Westinghouse PWRs RN 11-013 RN 95-022 11-013 RN 95-022 96-043 11-013 14-036 4.2-12 Reformatted February 2018 (References
[2] and [4]) have shown that fuel pellets can densify under irradiation to a density higher than the manufactured values. Fuel densification and subsequent incomplete settling of the fuel pellets results in local and distributed gaps in the fuel rods. Fuel densification has been minimized by improvements in the fuel manufacturing process and by specifying a nominal initial fuel density of 95%.
The effects of fuel densification have been taken into account in the nuclear and thermal hydraulic design of the reactor described in Sections 4.3 and 4.4, respectively.
4.2.1.3.1.3 Materials
- Strength Considerations One of the most important limiting factors in fuel element duty is the mechanical interaction of fuel and clad. This fuel/clad interaction produces cyclic stresses and strains in the clad, and these in turn consume clad fatigue life. The reduction of fuel/clad interaction is therefore a principal goal of design. In order to achieve this goal and to enhance the cyclic operational capability of the fuel rod, the technology for using pre- Initially the gap between the fuel and clad is sufficient to prevent hard contact between the two. However, during power operation a gradual compressive creep of the clad onto the fuel pellet occurs due to the external pressure exerted on the rod by the coolant. Clad compressive creep eventually results in the hard fuel/clad contact. During this period of fuel/clad contact, changes in power level could result in significant changes in clad stresses and strains. By using pre
-pressurized fuel rods to partially offset the effect of the coolant external pressure, the rate of clad creep toward the surface of the fuel is reduced. Fuel rod pre
-pressurization delays the time at which substantial fuel/clad interaction and hard contact occur and hence significantly reduces the number and extent of cyclic stresses and strains experienced by the clad both before and after fuel/clad contact. These factors result in an increase in the fatigue life margin of the clad and lead to greater clad reliability. If gaps should form in the fuel stacks, clad flattening will be prevented by the rod pre
-pressurization so that the flattening time will be greater than the fuel core life.
4.2.1.3.1.4 Steady-State Performance Evaluation In the calculation of the steady
-state performance of a nuclear fuel rod, the following interacting factors must be considered:
- 1. Clad creep and elastic deflection, 2. Pellet density changes, thermal expansion, gas release, and thermal properties as a function of temperature and fuel burnup, 3. Internal pressure as a function of fission gas release, rod geometry, and temperature distribution.
RN 14-036 4.2-13 Reformatted February 2018 These effects are evaluated using an overall fuel rod design model, per References [5, 6, 7, 17, 20, and 28] which include appropriate models for time dependent fuel densification. With these interacting factors considered, the model determines the fuel rod performance characteristics for a given rod geometry, power history, and axial power shape. In particular, internal gas pressure, fuel and clad temperatures, and clad deflections are calculated. The fuel rod is divided lengthwise into several sections and radially into a number of annular zones. Fuel density changes, clad stresses, strains and deformations, and fission gas releases are calculated separately for each segment. The effects are integrated to obtain the internal rod pressure.
The initial rod internal pressure is selected to delay fuel/clad mechanical interaction and to avoid the potential for flattened rod formation. It is limited, however, by the rod internal pressure design basis given in Section 4.2.1.1.1.
The gap conductance between the pellet surface and the clad inner diameter is calculated as a function of the composition, temperature, and pressure of the gas mixture, and the gap size or contact pressure between clad and pellet. After computing the fuel temperature for each pellet annular zone, the fractional fission gas release is assessed using an empirical model derived from experimental data in References [5, 17, and 28]. The total amount of gas released is based on the average fractional release within each axial and radial zone and the gas generation rate which in turn is a function of burnup. Finally, the gas released is summed over all zones and the pressure is calculated.
The model shows good agreement in fit for a variety of published and proprietary data on fission gas release, fuel temperatures and clad deflections, References [5, 17, and 28]. Included in this spectrum are variations in power, time, fuel density, and geometry. The in-pile fuel temperature measurement comparisons used are shown in References [5, 17, and 28]
. Typical fuel clad inner diameter and the fuel pellet outer diameter as a function of exposure are presented in Figure 4.2
-8. The cycle to cycle changes in the pellet outer diameter represent the effects of power changes as the fuel is moved into different positions as a result of refueling.
The gap size at any time is merely the difference between clad inner diameter and pellet outer diameter. Total clad
-pellet surface contact occurs near the end of Cycle
- 2. The figure represents hot fuel dimensions for a fuel rod operating at the power level shown in Figure 4.2
-9. Figure 4.2
-9 illustrates representative fuel rod internal gas pressure and linear power for the lead burnup rod versus irradiation time. In addition, it outlines the typical operating range of internal gas pressures which is applicable to the total fuel rod population within a region.
The "best estimate" fission gas release model was used in determining the internal gas pressures as a function of irradiation time. The plenum height of the fuel rod has been designed to ensure that the maximum internal pressure of the fuel rod will not exceed the design pressure of the reactor coolant.
RN 95-022 14-036 RN 95-022 14-036 RN 95-022 14-036 4.2-14 Reformatted February 2018 The clad stresses at a constant local fuel rod power are low. Compressive stresses are created by the pressure differential between the coolant pressure and the rod internal gas pressure. Because of the pre-pressurization with helium, the volume average effective stresses are less than 13,500 psi at the pressurization level used in this fuel rod design. Stresses due to the temperature gradient are not included in this average effective stress because thermal stresses are, in general, negative at the clad inside diameter and positive at the clad outside diameter and their contribution to the clad volume average stress is small. Furthermore, the thermal stress decreases with time during steady
-state operation due to stress relaxation. The stress due to pressure differential is highest in the minimum power rod at the beginning of life (due to low internal gas pressure). The thermal stress is highest in the maximum power rod (due to steep temperature gradient). Tensile stresses could be created once the clad has come in contact with the pellet. These stresses would be induced by the fuel pellet swelling during irradiation. As shown in Figure 4.2
-8, there is very limited clad pushout after pellet-clad contact. Fuel swelling can result in small clad strains (< 1%) for expected discharge burnups but the associated clad stresses are very low because of clad creep (thermal and irradiation
-induced creep). Furthermore, the 1% strain criterion is extremely conservative for fuel
-swelling driven clad strain because the strain rate associated with solid fission products swelling is very slow (-5 x 10-7 hr-1). In-pile experiments in Reference [8] have shown that Zircaloy tubing exhibits "superplasticity" at slow strain rates during neutron irradiation. Uniform clad strains of greater than 10% have been achieved under these conditions with no sign of plastic instability.
4.2.1.3.1.5 Transient Evaluation Method Pellet thermal expansion due to power increases is considered the only mechanism by which significant stresses and strains can be imposed on the clad. Power increases in commercial reactors can result from fuel shuffling (e.g., Region 3 positioned near the center of the core for Cycle 2 operation after operating near the periphery during Cycle 1), reactor power escalation following extended reduced power operation, and control rod movement. In the mechanical design model, lead rods are depleted using best estimate power histories as determined by core physics calculations. During the depletion, the amount of diametral gap closure is evaluated based upon the pellet expansion-cracking model, clad creep model, and fuel swelling model. At various times during depletion the rod power is assumed to increase locally up to the burnup dependent attainable power density as determined by core physics calculations. Historically, the radial, tangential, and axial clad stresses resulting from the power increase are combined into a volume average effective clad stress.
The Von Mises criterion is used to evaluate whether the clad yield stress has been exceeded. This criterion states that an isotropic material in multi
-axial stress will begin to yield plastically when the effective stress exceeds the yield stress as determined by a uniaxial tensile test. The yield stress is correlated for irradiated cladding since fuel/clad interaction occurs at high burnup. Furthermore, the effective stress is increased by an allowance which accounts for stress concentrations in the clad adjacent to radial cracks RN 95-022 RN 14-005 4.2-15 Reformatted February 2018 in the pellet prior to the comparison with the yield stress. This allowance was evaluated using a 2-dimensional (r,) finite element model.
According to the revised stress criterion documented in Reference
[23], the transient clad stress limit was designed to protect the cladding during pellet clad interaction (PCI).
This is one of the four criteria which were imposed to protect the cladding from PCI during Condition I and II operation.
The other three remaining criteria are transient strain less than 1%, no centerline fuel melt, and cladding total strain less than 1%.
The later three criteria are sufficient to protect the cladding from PCI.
The transient stress criterion is redundant and does not represent industry practice and the following describes the criterion for fuel rod clad stress per Addendum 1A of Reference
[23]: Maximum cladding stress intensities excluding PC I induced stress will be evaluated using ASME pressure vessel guidelines.
Cladding corrosion is accounted for as a loss of load carrying material.
Stresses are combined to calculate a maximum stress intensity which is then compared to criteria based on the ASME code.
Slow transient power increases can result in large clad strains because of clad creep and stress, therefore, a criterion on allowable clad positive strain is necessary. Based upon high strain rate burst and tensile test data on irradiated tubing, 1% strain was determined to be the lower limit on irradiated clad ductility and thus adopted as a design criterion.
In addition to the mechanical design models and design criteria, Westinghouse relies on performance data accumulated through transient power test programs in experimental and commercial reactors, and through normal operation in commercial reactors.
It is recognized that a possible limitation to the satisfactory behavior of the fuel rods in a reactor which is subjected to daily load follow is the failure of the clad by low cycle strain fatigue. During their normal residence time in reactor, the fuel rods may be subjected to 1,000 or more cycles with typical changes in power level from 50 to 100% of their steady-state values.
The assessment of the fatigue life of the fuel rod clad is subjected to a considerable uncertainty due to the difficulty of evaluating the strain range which results from the cyclic interaction of the fuel pellets and clad. This difficulty arises from such highly unpredictable phenomena as pellet cracking, fragmentation, and relocation. Nevertheless, since early 1968, Westinghouse has been investigating this particular phenomenon both analytically and experimentally. Strain fatigue tests on irradiated and nonirradiated hydrided Zircaloy
-4 claddings were performed which permitted a definition of a conservative fatigue life limit and recommendation of a methodology to treat the strain fatigue evaluation of the Westinghouse reference fuel rod designs.
However, Westinghouse is convinced that the final proof of the adequacy of a given fuel rod design to meet the load follow requirements can only come from incore experiments performed on actual reactors. The Westinghouse experience in load follow operation dates back to early 1970 with the load follow operation of the Saxton reactor.
RN 95-022 RN 14-005 RN 14-005 4.2-16 Reformatted February 2018 Successful load follow operation has been performed on reactor A (300 load follow cycles) and reactor B (150 load follow cycles). In both cases, there was no significant coolant activity increase that could be associated with the load follow mode of operation.
The following paragraphs present briefly the Westinghouse analytical approach to strain fatigue. A comprehensive review of the available strain
-fatigue models was conducted by Westinghouse as early as 1968. This included the Langer
- Reference [9], the Yao-Munse model, and the Manson
-Halford model.
Upon completion of this review and using the results of the Westinghouse experimental programs discussed below, it was concluded that the approach defined by Langer-modified in order to conservatively bound the results of the Westinghouse testing program. The Langer
- Where: S a = 1/2 E t = pseudo-stress amplitude which causes failure in N f cycles (lb/in
- 2) t = total strain range (in/in)
- 2) N f = number of cycles to failure RA = reduction in area at fracture in a uniaxial tensile test (%)
S e = endurance limit (lb/in
- 2) Both RA and S e are empirical constants which depend on the type of material, the temperature and irradiation. The Westinghouse testing program was subdivided in the following subprograms:
- 1. A rotating bend fatigue experiment on unirradiated Zircaloy
-4 specimens at room temperature and at 725F. Both hydrided and nonhydrided Zircaloy
-4 cladding were tested.
RN 14-036 4.2-17 Reformatted February 2018
- 2. A biaxial fatigue experiment in gas autoclave on unirradiated Zircaloy
-4 cladding both hydrided and nonhydrided.
- 3. A fatigue test program on irradiated cladding from the CVTR and Yankee Core V conducted at Battelle Memorial Institute.
The results of these test programs provided information on different cladding conditions including the effect of irradiation, hydrogen level, and temperature.
The Westinghouse design equations followed the concept for the fatigue design criterion according to the ASME Boiler and Pressure Vessel Code,Section III, namely:
- 1. The calculated pseudo
-stress amplitude (S a) has to be multiplied by a factor of 2 in order to obtain the allowable number of cycles (N f). 2. The allowable cycles for a given S a is 5% of N f, or a safety factor of 20 on cycles.
The lesser of the 2 allowable number of cycles is selected. The cumulative fatigue life fraction is then computed as:
Where: n k = number of diurnal cycles of mode k.
The potential effects of operation with waterlogged fuel are discussed in Section 4.4.3.6. Waterlogging is not considered to be a concern during operational transients.
4.2.1.3.1.6 Rod Bowing Reference [10] presents the model used for evaluation of fuel rod bowing.
RN 96-043 4.2-18 Reformatted February 2018 4.2.1.3.2 Fuel Assembly Structure 4.2.1.3.2.1 Stresses and Deflections The fuel assembly component stress levels are limited by the design. For example, stresses in the fuel rod due to thermal expansion and cladding irradiation growth are limited by the relative motion of the rod as it slips over the grid spring and dimple surfaces.
Clearances between the fuel rod ends and nozzles are provided so that cladding irradiation growth will not result in rod end interferences. Stresses in the fuel assembly caused by tripping of the rod cluster control assembly have little influence on fatigue because of the small number of events during the life of an assembly. Assembly components and prototype fuel assemblies made from production parts have been subjected to structural tests to verify that the design bases requirements are met.
The fuel assembly response resulting from the most limiting pipe break was analyzed using time
-history numerical techniques. Since the resulting vessel motion induces primarily lateral loads on the reactor core, a finite element model similar to the seism ic model described in Reference [31] was used to assess the fuel assembly deflections and impact forces.
The reactor core finite element model which simulates the fuel assembly interaction during lateral excitation consists of fuel assemblies arranged in a planer array with inter assembly gaps. For the Virgil C. Summer Nuclear Station the fuel assembly rows 15, 13, 11, 9, 7, and 3 were analyzed.
The fuel assemblies and the reactor baffle support s are represented by single beam elements
. Figure 4.2
-28 shows this single beam schematic for row 15, the maximum number of assemblies across the core diameter. The time-history motion for the upper and lower core plates and the barrel at the upper core plate elevation are simultaneously applied to the simulated reactor core model as illustrated in Figure 4.2-28. The 3 time
-history motions were obtained from the tim e-history analysis of the reactor vessel and internals finite element model.
An evaluation of fuel assembly structural integrity consider ed the lateral effects of a LOCA accident with an upflow conversion determined that the seismic analysis of record remains applicable to the VC Summer Upflow conversion. The grid impact load s resulting from seismic and LOCA accidents were combined by SRSS (Square root sum of square) method
. The safe shutdown earthquake and LOCA analyses indicated tha t the flow mixers will share some grid load among the structural grids. The analysis results confirmed that the calculated grid loads are less than the allowable limit except 3 fuel assembly (FA) row. According to Reference [32], the core coolable geometry is maintained when the crushed grid s occur at 3 FA row.
RN 95-022 96-043 RN 96-043 09-022 RN 09-022 RN 95-022 09-022 4.2-19 Reformatted February 2018 Because the Virgil C. Summer Nuclear Station is not bounded by the LOCA/Seismic loads considered in Reference
[25] and Reference [24], additional analyses have been performed to demonstrate fuel assembly structural integrity.
The evaluation of the VANTAGE 5 fuel assembly, in accordance with NRC requirements as given in SRP 4.2, Appendix A, shows that the VANTAGE 5 fuel is structurally acceptable for an all VANTAGE 5 core; the grids will not buckle due to combined impact forces of a seismic/LOCA event except in the 3 fuel assembly row. Since the 3 fuel assembly row is located at a low power ed region, the core coolable geometry is maintained based on Reference [32].
The stresses in the fuel assembly components resulting from seismic and LOCA loads are well within acceptable limits ensuring that the reactor can be safely shutdown under the combined faulted condition loads.
Since the VANTAGE+ assembly is structurally identical to that of VANTAGE 5, the evaluations discussed above are also applicable to VANTAGE+.
The fuel assembly design loads for shipping have been established at 6 g lateral and 4 g axial. Accelerometers are permanently placed into the shipping cask to monitor and detect fuel assembly accelerations that would exceed the criteria. Past history and experience has indicated that loads which exceed the allowable limits rarely occur. Exceeding the limits requires reinspection of the fuel assembly for damage. Tests on various fuel assembly components such as the grid assembly, sleeves, inserts and structure joints have been performed to assure that the shipping design limits do not result in impairment of fuel assembly function.
4.2.1.3.2.2 Dimensional Stability The VANTAGE 5 Mechanical Test Program description and results are given in Appendix A of Reference [24] and are considered to be applicable to VANTAGE+ as the 2 assemblies are essentially identical from a structural standpoint.
The coolant flow channels are established and maintained by the structure composed of grids and guide thimbles. The lateral spacing between fuel rods is provided an d controlled by the support dimples of adjacent grid cells. Contact of the fuel rods on the dimples is maintained through the clamping force of the grid springs. Lateral motion of the fuel rods is opposed by the spring force and the internal moments generated between the spring and the support dimples.
RN 95-022 96-043 RN 95-022 09-022 RN 95-022 4.2-20 Reformatted February 2018 No interference with control rod insertion into thimble tubes will occur during a postulated loss of coolant accident transient due to fuel rod swelling, thermal expansion, or bowing. In the early phase of the transient following the coolant break, the high axial loads, which potentially could be generated by the difference in thermal expansion between fuel clad and thimbles, are relieved by slippage of the fuel rods through the grids. The relatively low drag force restraint on the fuel rods will induce only minor thermal bowing, which is insufficient to close the fuel rod
-to-thimble tube gap. This rod-to-grid slip mechanism occurs simultaneously with control rod drop. Subsequent to the control rod insertion the transient temperature increase of the fuel rod clad can result in swelling sufficient to contact the thimbles.
4.2.1.3.2.3 Vibration and Wear Fuel rod vibrations are flow induced. The effect of the vibration on the fuel assembly and individual fuel rods is minimal. The cyclic stress range associated with deflections of such small magnitude is insignificant and has no effect on the structural integrity of the fuel rod.
The reaction force on the grid supports due to rod vibration is small and is much less than the spring preload. Firm contact is maintained. No significant wear of the clad or grid supports is expected during the life of the fuel assembly.
Clad fretting and fuel rod vibration has been experimentally investigated as discussed in Reference [24] for VANTAGE 5 fuel and is applicable directly to VANTAGE+ fuel design. 4.2.1.3.3 Operational Experience A discussion of fuel operating experience is given in Reference
[4]. 4.2.1.4 Tests and Inspections 4.2.1.4.1 Quality Assurance Program The Quality Assurance Program Plan of the Westinghouse Commercial Nuclear Fuel Division is summarized in Reference [13].
The program provides for control over all activities affecting product quality, commencing with design and development, and continuing through procurement, materials handling, fabrication, testing and inspection, storage, and transportation. The program also provides for the indoctrination and training of personnel and for the auditing of activities affecting product quality through a formal auditing program.
Westinghouse drawings and product, process, and material specifications identify the inspection to be performed.
RN 95-022 4.2-21 Reformatted February 2018 4.2.1.4.2 Quality Control Quality control philosophy is generally based on the following inspections being performed to a 95% confidence that at least 95% of the product meets specification, unless otherwise noted.
- 1. Fuel System Components and Parts The characteristics inspected depends upon the component parts and includes dimensional, visual, check audits of test reports, material certification and nondestructive examination such as X
-Ray and ultrasonic.
All material used in this core is verified and released by Quality control.
- 2. Pellets Inspection is performed for dimensional characteristics such as diameter, density, length and squareness of ends. Additional visual inspections are performed for cracks, chips and surface conditions according to approved standards.
Density is determined in terms of weight per unit length and is plotted on zo ne charts used in controlling the process. Chemical analyses are taken on a specified sample basis through pellet production.
- 3. Rod Inspection Fuel rod, control rodlet, burnable poison and source rod inspection consists of the following nondestructive examination techniques and methods, as applicable.
- a. Leak Testing Each rod is tested using a calibrated mass spectrometer with helium being the detectable gas.
- b. Enclosure Welds Rod welds are inspected by ultrasonic test or X
-ray in accordance with a qualified technique and Westinghouse specifications.
- c. Dimensional All rods are dimensionally inspected prior to final release.
The requirements include such items as length, camber, and visual appearance.
4.2-22 Reformatted February 2018
- d. Plenum Dimensions 100% of the fuel rods are inspected by gamma scanning or other approved methods to ensure proper plenum dimensions.
- e. Pellet-to-Pellet Gaps 100% of the fuel rods are inspected by gamma scanning or other methods to ensure that no significant gaps exist between pellets. f. Pellet Enrichment 100% of the fuel rods are active gamma scanned to verify enrichment control prior to acceptance for assembly loading.
- g. Traceability Traceability of rods and associated rod components is established by Quality Control. 4. Assemblies Each fuel, control rod, burnable absorber and source rod assembly is inspected for drawing and/or specification requirements. Other incore control component inspection and specification requirements are given in Section 4.2.3.4.
- 5. Other Inspections The following inspections are performed as part of the routine inspection operation:
- a. Tool and gage inspection and control including standardization to primary and/or secondary working standards. Tool inspection is performed at prescribed intervals on all serialized tools.
Complete records are kept of calibration and conditions of tools.
- b. Audits are performed of inspection activities and records to assure that prescribed methods are followed and that records are correct and properly maintained.
- c. Surveillance inspection where appropriate, and audits of outside contractors are performed to ensure conformance with specified requirements.
RN 14-036 4.2-23 Reformatted February 2018
- 6. Process Control To prevent the possibility of mixing enrichments during fuel manufacture and assembly, strict enrichment segregation and other process controls are exercised.
The UO 2 powder is kept in sealed containers or is processed in a closed system. The containers are either fully identified both by descriptive tagging and preselected color coding or, for the closed system, the material is monitored by a computer data management information system. For the sealed container system, a Westinghouse identification tag completely describing the contents is affixed to the containers before transfer to powder storage. Isotopic content is confirmed by analysis. Powder withdrawal from storage can be made by only one authorized group, which directs the powder to the correct pellet production line. All pellet production lines are physically separated from each other and pellets of only a single nominal enrichment and density are produced in a given production line.
Finished pellets are placed on trays and transferred to segregated storage racks within the confines of the pelleting area. Samples from each pellet lot are tested for physical and chemical properties including isotopic content and impurity levels prior to acceptance by Quality Control. Physical barriers prevent mixing of pellets of different nominal designs and enrichment in this storage area.
Unused powder and substandard pellets are returned to storage for disposition.
Loading of pellets into the clad is performed in isolated production lines and again only one density and enrichment is loaded on a line at a time.
A serialized traceability code is placed on each fuel tube which identifies the enrichment. The end plugs are inserted and then inert welded to seal the tube. The fuel tube remains coded and traceability identified until just prior to installation in the fuel assembly.
At the time of installation into an assembly, the traceability codes are removed and a matrix is generated to identify each rod in its position within a given assembly. After the fuel rods are installed, an inspector verifies that all fuel rods in an assembly carry the correct identification character describing the fuel enrichment and density for the core region being fabricated. The top nozzle is inscribed with a permanent identification number providing traceability to the fuel contained in the assembly. Similar traceability is provided for burnable absorber, source rods and control rodlets as required.
4.2-24 Reformatted February 2018 4.2.1.4.3 Onsite Inspection Onsite inspection programs for fuel, control rods, and internals are based on the NSSS led procedures. In the event reloads or other components are supplied procedures.
Loaded fuel containers, when received onsite, are externally inspected to ensure that labels and markings are intact and seals are unbroken. After the containers are opened, the shock indicators attached to the suspended internals are inspected to determine if movement during transit exceeded design limitations.
Following removal of the fuel assembly from the container in accordance with detailed procedures from the fuel fabricator the polyethylene wrapper is then removed and a visual inspection of the entire bundle is performed.
Control rod assemblies are shipped in fuel assemblies and are inspected.
The control rod assembly is withdrawn a few inches from the fuel assembly using the fuel handling tool to ensure free and unrestricted movement. The exposed section is then visibly inspected for mechanical integrity, replaced in the fuel assembly and stored with the fuel assembly. Reactor internals are visually inspected and manually checked for tightness upon receipt at the site. Clearance measurements performed during assembly and procedures, also serve to verify the mechanical integrity of the internals.
Surveillance of fuel and reactor performance is routinely conducted by operating personnel. Coolant activity and chemistry are followed to permit early detection of any fuel clad defects. Visual fuel inspection in accordance with the EPRI
-Fuel Reliability Guidelines Reference
[18] is conducted during refueling. This type of inspection is capable of identifying gross anomalies such as broken fuel rods or torn grid straps. Additional fuel inspections are dependent on the results of the operational monitoring and visual inspection.
RN 15-022 RN 15-022 4.2-25 Reformatted February 2018 4.2.2 REACTOR VESSEL INTERNALS 4.2.2.1 Design Bases The design bases for the mechanical design of the reactor vessel internals components are as follows:
- 1. The reactor internals in conjunction with the fuel assemblies shall direct reactor coolant through the core to achieve acceptable flow distribution and to restrict bypass flow so that the heat transfer performance requirements are met for all modes of operation. In addition, required cooling for the pressure vessel head shall be provided so that the temperature differences between the vessel flange and head do not result in leakage from the flange during reactor operation.
- 2. In addition to neutron shielding provided by the reactor coolant, the reactor internals are designed to limit the exposure of the pressure vessel in order to maintain the required ductility of the material for all modes of operation.
- 3. Provisions shall be made for installing incore instrumentation useful for the plant operation and vessel material test specimens required for a pressure vessel irradiation surveillance program.
- 4. The core internals are designed to withstand mechanical loads arising from operating basis earthquake, safe shutdown earthquake and pipe ruptures and meet the requirement of item 5 below.
- 5. The reactor shall have mechanical provisions which are sufficient to adequately support the core and internals and to assure that the core is intact with acceptable heat transfer geometry following transients arising from abnormal operating conditions.
- 6. Following the design basis accident, the plant shall be capable of being shutdown and cooled in an orderly fashion so that fuel cladding temperature is kept within specified limits. This implies that the deformation of certain critical reactor internals must be kept sufficiently small to allow core cooling.
The functional limitations for the core structures during the design basis accident are shown in Table 4.2
-1. To ensure no column loading of rod cluster control guide tubes, the upper core plate deflection is limited to not exceed the value shown in Table 4.2
-1. Details of the dynamic analyses, input forcing functions, and response loadings are presented in Section 3.9.
4.2-26 Reformatted February 2018 4.2.2.2 Description and Drawings The reactor vessel internals are described as follows:
The components of the reactor internals are divided into 3 parts consisting of the lower core support structure (including the entire core barrel and neutron shield pad assembly), the upper core support structure and the incore instrumentation support structure. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between fuel assemblies and control rod drive mechanisms, direct coolant flow past the fuel elements, direct coolant flow to the pressure vessel head, provide gamma and neutron shielding, and guides for the incore instrumentation. The coolant flows from the vessel inlet nozzles down the annulus between the core barrel and the vessel wall and then into a plenum at the bottom of the vessel. It then reverses and flows up through the core support and through the lower core plate. The lower core plate is sized to provide the desired inlet flow distribution to the core. After passing through the core, the coolant enters the region of the up per support structure and then flows radially to the core barrel outlet nozzles and directly through the vessel outlet nozzles. A small portion of the coolant flows between the baffle plates and the core barrel to provide additional cooling of the barrel. Beginning with Cycle 19, the direction of coolant flow between the baffle plates and core barrel was change d from a downflow configuration to an upflow configuration to reduce baffle jetting - related fuel failures. Similarly, a small amount of the entering flow is directed into the vessel head plenum and exits through the vessel outlet nozzles.
All the major material for the reactor internals is Type 304 stainless steel. Parts not fabricated from Type 304 stainless steel include bolts and dowel pins which are fabricated from Type 316 stainless steel and radial support key bolts which are fabricated of Inconel
-750. These materials are listed in Table 5.2
-12. There are no other materials used in the reactor internals or core support structures which are not otherwise included in ASME Code,Section III, Appendix I. The discussions provided in Sections 5.2.3 and 5.2.5 are applicable to the welding of reactor internals and core support components.
The discussions provided in Sections 5.2.5.1, 5.2.5.2, 5.2.5.3, 5.2.5.4, 5.2.5.5, and 5.2.5.6 are applicable to the degree of conformance of reactor internals and core support structures with Regulatory Guide 1.44.
The discussion provided in Section 5.2.5.7 addresses Regulatory Guide 1.31 with respect to reactor internals and core support structures.
Appendix 3A addresses Regulatory Guides 1.66 and 1.71 with respect to reactor internals and core support structures.
RN 09-022 4.2-27 Reformatted February 2018 Austenitic stainless steel is used for the majority of reactor internals structures, and this material is not subject to brittle fracture. The core hold down spring, however, is made of Type 403 stainless steel. Significant crack growth has been found to be impossible for this component considering the stress state and possible flaw size. The core hold down spring is the only stainless steel material in the reactor core support structure with a yield strength greater than 90,000 psi and is acceptable based upon Code Case 1337. All reactor internals are removable from the vessel for the purpose of their inspection as well as the inspection of the vessel internal surface.
4.2.2.2.1 Lower Core Support Structure The major containment and support member of the reactor internals is the lower core support structure, shown in Figure 4.2
-10. This support structure assembly consists of the core barrel, the core baffle, the lower core plate and support columns, the neutron shield pads, and the core support which is welded to the core barrel. All the major material for this structure is Type 304 stainless steel. The lower core support structure is supported at its upper flange from a ledge in the reactor vessel head flange and it s lower end is restrained in its transverse movement by a radial support system attached to the vessel wall. Within the core barrel are in axial baffle and a lower core plate, both of which are attached to the core barrel wall and form the enclosure periphery of the assembled core. The lower core support structure and principally the core barrel serve to provide passageways and control for the coolant flow.
The lower core plate is positioned at the bottom level of the core below the baffle plates and provides support and orientation for the fuel assemblies.
The lower core plate is a member through which the necessary flow distribution holes for each fuel assembly are machined. Fuel assembly locating pins (2 for each assembly) are also inserted into this plate. Columns are placed between this plate and the core support of the core barrel in order to provide stiffness and to transmit the core load to the core support. Adequate coolant distribution is obtained through the use of the lower core plate and core support.
The neutron shield pad assembly consists of 4 pads that are bolted and pinned to the outside of the core barrel. These pads are constructed of Type 304 stainless steel and are approximately 48 inches wide by 148 inches long by 2.8 inches thick. The pads are located azimuthally to provide the required degree of vessel protection. Specimen guides in which material surveillance samples can be inserted and irradiated during reactor operation are attached to the pads. The samples are held in the guides by a preloaded spring device at the top and bottom to prevent sample movement. Additional details of the neutron shield pads and irradiation specimen holders are given in Reference [14].
4.2-28 Reformatted February 2018 Vertically downward loads from weight, fuel assembly preload, control rod dynamic loading, hydraulic loads and earthquake acceleration are carried by the lower core plate partially into the lower core plate support flange on the core barrel shell and partially through the lower support columns to the core support and thence through the core barrel shell to the core barrel flange supported by the vessel head flange. Transverse loads from earthquake acceleration, coolant cross flow, and vibration are carried by the core barrel shell and distributed between the lower radial support to the vessel wall, and to the vessel flange. Transverse loads of the fuel assemblies are transmitted to the core barrel shell by direct connection of the lower core plate to the barrel wall and by upper core plate alignment pins which are welded onto the core barrel.
The main radial support system of the lower end of the core barrel is accomplished by "key" and "keyway" joints to the reactor vessel wall. At equally spaced points around the circumference, an Inconel clevis block is welded to the vessel inner diameter. Another Inconel insert block is bolted to each of these blocks and has a "keyway" geometry. Opposite each of these is a "key" which is attached to the internals. At assembly, as the internals are lowered into the vessel, the keys engage the keyways in the axial direction. With this design, the internals are provided with a support at the furthest extremity, and may be viewed as a beam fixed at the top and simply supported at the bottom.
Radial and axial expansions of the core barrel are accommodated but transverse movement of the core barrel is restricted by this design. With this system, cyclic stresses in the internal structures are within the ASME Section III limits. In the event of an abnormal downward vertical displacement of the internals following a hypothetical failure, energy absorbing devices limit the displacement after contacting the vessel bottom head. The load is then transferred through the energy absorbing devices of the internals to the vessel.
The energy absorbers, cylindrical in shape, are contoured on their bottom surface to the reactor vessel bottom head geometry. Assuming a downward vertical displacement the potential energy of the system is absorbed mostly by the strain energy of the energy absorbing devices.
4.2.2.2.2 Upper Core Support Assembly The upper core support assembly, shown in Figures 4.2
-11 and 4.2
-12 consists of the upper support plate assembly and the upper core plate between which are contained support columns and guide tube assemblies. The support columns establish the spacing between the upper support plate assembly and the upper core plate and are fastened at top and bottom to these plates. The support columns transmit the mechanical loadings between the 2 plates and serve the supplementary function of supporting thermocouple guide tubes. The guide tube assemblies, sheath and guide the control rod drive shafts and control rods. They are fastened to the top support plate and are restrained by pins in the upper core plate for proper orientation and support. Additional guidance for the control rod drive shafts is provided by the upper guide tube which is attached to the upper support plate and guide tube.
4.2-29 Reformatted February 2018 The upper core support assembly is positioned in its proper orientation with respect to the lower support structure by flat
-sided pins pressed into the core barrel which in turn engage in slots in the upper core plate. At an elevation in the core barrel where the upper core plate is positioned, the flat
-sided pins are located at angular positions of 90 from each other. Four (4) slots are milled into the core plate at the same positions. As the upper support structure is lowered into the main internals, the slots in the plate engage the flat
-sided pins in the axial direction. Lateral displacement of the plate and of the upper support assembly is restricted by this design. Fuel assembly locating pins protrude from the bottom of the upper core plate and engage the fuel assemblies as the upper assembly is lowered into place. Proper alignment of the lower core support structure, the upper core support assembly, the fuel assemblies and control rods are thereby assured by this system of locating pins and guidance arrangement. The upper core support assembly is restrained from any axial movements by a large circumferential spring which rests between the upper barrel flange and the upper core support assembly and is compressed by the reactor vessel head flange.
Vertical loads from weight, earthquake acceleration, hydraulic loads and fuel assembly preload are transmitted through the upper core plate via the support columns to the top support plate assembly and then the reactor vessel head. Transverse loads from coolant cross flow, earthquake acceleration, and possible vibrations are distributed by the support columns to the top support plate and upper core plate. The top support plate is particularly stiff to minimize deflection.
The following information describes repairs which were made to the original Reactor Vessel Head.
Although these components are no longer installed at VCS, this information is being retained for historical purposes.
During RF-20 and RF-21, repairs were performed to mitigate Primary Water Stress Corrosion Cracking (PWSCC) in the CRDM nozzle J
-groove welds in the Reactor Vessel Upper Head enclosure (per ECR
-50846). The J-groove weld repairs follow WCAP
-15987-P, Rev. 2-P-A , Basis for the Embed dwhich has been approved by the NRC on April 30, 2014
, under Relief Request RR 05, This repair required the permanent removal of the Part
-Length (P/L) Mechanism drive rod located in penetration #19 during RF-20. Additional P/L CRDM drive rods were cut in penetrations #1 and 18 in RF
-21 to facilitate the thermal sleeve degradation inspections. To retain the flow characteristics through the upper internals, flow restrictors were mechanically clamped into the Upper Internals assembly guide tubes which sit under RV Head penetration #1, 18 and 19.
4.2.2.2.3 Incore Instrumentation Support Structures The incore instrumentation support structures consist of an upper system to convey and support thermocouples penetrating the vessel through the head and a lower system to convey and support flux thimbles penetrating the vessel through the bottom (Figure 7.7-9 shows the basic flux
-mapping system).
RN 12-042 14-024 15-023 RN 16-003 4.2-30 Reformatted February 2018 The upper system utilizes the reactor vessel head penetrations. Instrumentation port columns are slip
-connected to inline columns that are in turn fastened to the upper support plate. These port columns protrude through the head penetrations. The thermocouples are carried through these port columns and the upper support plate at positions above their readout locations. The thermocouple conduits are supported from the columns of the upper core support system. The thermocouple conduits are sealed stainless steel tubes.
In addition to the upper incore instrumentation, there are reactor vessel bottom port columns which carry the retractable, cold worked stainless steel flux thimbles that are pushed upward into the reactor core. Conduits extend from the bottom of the reactor vessel down through the concrete shield area and up to a thimble seal line. The minimum bend radii are about 144 inches and the trailing ends of the thimbles (at the seal line) are extracted approximately 15 feet during refueling of the reactor in order to avoid interference within the core. The thimbles are closed at the leading ends and serve as the pressure barrier between the reactor pressurized water and the containment atmosphere.
Mechanical seals between the retractable thimbles and conduits are provided at the seal line. During normal operation, the retractable thimbles are stationary and move only during refueling or for maintenance, at which time a space of approximately 15 feet above the seal line is cleared for the retraction operation.
The incore instrumentation support structure is designed for adequate support of instrumentation during reactor operation and is designed to resist damage or distortion under the conditions imposed by handling during the refueling sequence. These are the only conditions which affect the incore instrumentation support structure. Reactor vessel surveillance specimen capsules are covered in Section 5.4.3.6.
4.2.2.3 Design Loading Conditions The design loading conditions that provide the basis for the design of the reactor internals are:
- 1. Fuel Assembly Weight
- 2. Fuel Assembly Spring Forces
- 3. Internals Weight
- 4. Control Rod Trip (equivalent static load)
- 5. Differential Pressure
- 6. Spring Preloads
4.2-31 Reformatted February 2018
- 7. Coolant Flow Forces (static)
- 8. Temperature Gradients
- 9. Differences in thermal expansion a. Due to temperature differences
- b. Due to expansion of different materials.
- 10. Interference between components
- 11. Vibration (mechanically or hydraulically induced)
- 12. One (1) or more loops out of service
- 13. All operational transients listed in Table 5.2
-2 14. Pump overspeed
- 15. Seismic loads (operating basis earthquake and safe shutdown earthquake)
- 16. Blowdown forces (due to cold and hot leg break)
The main objective of the design analysis is to satisfy allowable stress limits, to assure an adequate design margin, and to establish deformation limits which are concerned primarily with the functioning of the components. The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the materials. Both low and high cycle fatigue stresses are considered when the allowable amplitude of oscillation is established.
Dynamic analysis on the reactor internals are provided in Section 3.9. As part of the evaluation of design loading conditions, extensive testing and inspections are performed from the initial selection of raw materials up to and including component installation and plant operation. Among these tests and inspections are those performed during component fabrication, plant construction, startup and checkout, and during plant operation.
4.2-32 Reformatted February 2018 4.2.2.4 Design Loading Categories The combination of design loadings fit into either the normal, upset, emergency or faulted conditions as defined in the ASME Code,Section III. Loads and deflections imposed on components due to shock and vibration are determined analytically and experimentally in both scaled models and operating reactors. The cyclic stresses due to these dynamic loads and deflections are combined with the stresses imposed by loads from component weights, hydraulic forces and thermal gradients for the determination of the total stresses of the internals.
The reactor internals are designed to withstand stresses originating from various operating conditions as summarized in Table 5.2
-2. The scope of the stress analysis problem is very large requiring many different techniques and methods, both static and dynamic. The analysis performed depends on the mode of operation under consideration.
For normal operating conditions, downward vertical deflection of the lower core support plate is negligib le. For the loss of coolant accident plus the safe shutdown earthquake condition, the deflection criteria of critical internal structures are the limiting values given in Table 4.2-1. The corresponding no loss of function limits are included in Table 4.2
-1 for comparison purposes with the allowed criteria.
The criteria for the core drop accident is based upon analyses which have to determine the total downward displacement of the internal structures following a hypothesized core drop resulting from loss of the normal core barrel supports. The initial clearance between the secondary core support structures and the reactor vessel lower head in the hot condition is approximately 1/2 inch. An additional displacement of approximately 3/4 inch would occur due to strain of the energy absorbing devices of the secondary core support; thus the total drop distance is about 1
-1/4 inches which is insufficient to permit the tips of the rod cluster control assembly to come out of the guide thimble in the fuel assemblie
- s. Specifically, the secondary core support is a device which is never expected to be used, except during a hypothetical accident of the core support (core barrel, barrel flange, etc.). There are 4 supports in each reactor. This device limits the fall of the core and absorbs the energy of the fall which otherwise would be imparted to the vessel. The energy of the fall is calculated assuming a complete and instantaneous failure of the primary core support and is absorbed during the plastic deformation of the controlled volume of stainless steel, loaded in tension.
The maximum deformation of this austenitic stainless piece is limited to approximately 15%, after which a positive stop is provided to ensure support.
4.2-33 Reformatted February 2018 For additional information on design loading categories see Section 3.9. 4.2.2.5 Design Criteria Basis For normal operating conditionsSection III of the ASME Nuclear Power Plant Components Code is used as a basis for evaluating acceptability of calculated stresses. Both static and alternating stress intensities are considered. It should be noted that the allowable stresses in Section III of the ASME Code are based on unirradiated material properties. In view of the fact that irradiation increases the strength of the Type 304 stainless steel used for the internals, although decreasing its elongation, it is considered that use of the allowable stresses in Section III is appropriate and conservative for irradiated internal structures.
The allowable stress limits during the design basis accident used for the core support structures are based on the January 1971 draft of the ASME Code for Core Support Structures, Subsection NG, and the Criteria for Faulted Conditions.
4.2.3 REACTIVITY CONTROL SYSTEM 4.2.3.1 Design Bases Bases for temperature, stress on structural members, and material compatibility are imposed on the design of the reactivity control components.
4.2.3.1.1 Design Stresses The reactivity control system is designed to withstand stresses originating from various operating conditions as summarized in Table 5.2
-2. 1. Allowable Stresses For normal operating conditionsSection III of the ASME Boiler and Pressure Vessel Code is used. All pressure boundary components are analyzed as Class 1 components under Article NB
-3000. 2. Dynamic Analysis The cyclic stresses due to dynamic loads and deflections are combined with the stresses imposed by loads from component weights, hydraulic forces and thermal gradients for the determination of the total stresses of the reactivity control system.
4.2.3.1.2 Material Compatibility Materials are selected for compatibility in a PWR environment, for adequate mechanical properties at room and operating temperature, for resistance to adverse property changes in a radioactive environment, and for compatibility with interfacing components.
4.2-34 Reformatted February 2018 4.2.3.1.3 Reactivity Control Components The reactivity control components are subdivided into 2 categories:
- 1. Permanent devices used to control or monitor the core and, 2. Optional devices used to control or monitor the core.
The permanent type components are the rod cluster control assemblies (contain absorber rods), control rod drive mechanisms and neutron source assemblies.
The optional type components are the IFBA, burnable absorber assembly, and thimble plug assembly. Although the thimble plug assembly does not directly contribute to the reactivity control of the reactor, it is presented as a reactivity control system component in this document because it may be used to restrict bypass flow through those thimbles not occupied by absorber, source or burnable absorber rods. The design bases for each of the mentioned components, except the IFBA rods, are in the following paragraphs. The IFBA rods are discussed in the Fuel Section 4.2.1. 4.2.3.1.3.1 Absorber Rods The following are considered design conditions under Article NB
-3000 of the ASME Boiler and Pressure Vessel Code,Section III. 1. The external pressure equal to the reactor coolant system operating pressure.
- 2. The wear allowance equivalent to 1,000 reactor trips.
- 3. Bending of the rod due to a misalignment in the guide tube.
- 4. Forces imposed on the rods during rod drop.
- 5. Loads caused by accelerations imposed by the control rod drive mechanism.
- 6. Radiation exposure for maximum core life.
The control rod which is cold rolled Type 304 stainless steel is the only noncode material used in the control rod assembly. The stress intensity limit Sm for this material is defined at 2/3 of the 0.2% offset yield stress.
The absorber material temperature shall not exceed its melting temperature of 1470 F for silver-indium-cadmium alloy absorber material[15]. 4.2.3.1.3.2 Burnable Absorber Rods The burnable absorber rods may be of the design containing borosilicate glass or the Wet Annular Burnable Absorber (WABA) design containing Al 2 O 3-B 4C absorber material.
4.2-35 Reformatted February 2018 The burnable absorber rod clad (304 SS for the borosilicate design and Zircaloy 4 for the WABA design) is designed using the requirements of a Class 1 component under Article NB
-3000 of the ASME Boiler and Pressure Vessel Code,Section III, 1973 for Conditions I and II as a guide. For abnormal loads during Conditions III and IV, code stresses are not considered limiting. Failures of the burnable absorber rods during these conditions must not interfere with reactor shutdown or cooling of the fuel rods.
The burnable poison absorber material is nonstructural. The structural elements of the burnable absorber rod are designed to maintain the absorber geometry even if the absorber material is fractured. The rods are designed so that the borosilicate absorber material is below its softening temperature (1492 F [1] for reference 12.5 w/o boron rods), and the Al 2 O 3-B 4C material is below 1200F during normal operation or overpower transients.
4.2.3.1.3.3 Neutron Source Rods The neutron source rods are designed to withstand the following:
- 1. The external pressure equal to the reactor coolant system operating pressure and
- 2. An internal pressure equal to the pressure generated by released gases over the source rod life.
4.2.3.1.3.4 Thimble Plug Assembly (if used)
The thimble plug assemblies satisfy the following:
- 1. Accommodate the differential thermal expansion between the fuel assembly and the core internals.
- 2. Maintain positive contact with the fuel assembly and the core internals.
- 3. Be inserted into the fuel assembly by a force not exceeding 40 pounds. 4.2.3.1.4 Control Rod Drive Mechanisms The control rod drive mechanisms (CRDMs) pressure housings are Class 1 components designed to meet the stress requirements for normal operating conditions of Section III of the ASME Boiler and Pressure Vessel Code. Both static and alternating stress intensities are considered. The stresses originating from the required design transients are included in the analysis.
[1] Borosilicate glass is accepted for use in burnable absorber rods if the softening temperature is 1510 +/-18F. The softening temperature is defined in ASTM C 338.
RN 95-022 4.2-36 Reformatted February 2018 A dynamic seismic analysis is required on the CRDMs when a seismic disturbance has been postulated to confirm the ability of the pressure housing to meet ASME Code,Section III allowable stresses and to confirm its ability to trip when subjected to the seismic disturbance.
4.2.3.1.4.1 Control Rod Drive Mechanism Operation Requirements The basic operational requirements for the CRDMs are:
- 1. 5/8 inch step, 2. 144 inch travel, 3. 360 pound maximum load (includes drive rod weight), 4. Step in or out at 5 to 45 inches/minute (8 to 72 steps/minute), 5. Electrical power interruption shall initiate release of drive rod assembly, 6. Trip delay time of less than 150 milliseconds
- Free fall of drive rod assembly shall begin less than 150 milliseconds after power interruption no matter what holding or stepping action is being executed with any load and coolant temperature of 100 F to 550 F. 7. 40 year design life with normal refurbishment.
4.2.3.2 Design Description Reactivity control is provided by IFBA pins, burnable absorber rods, and a soluble chemical neutron absorber (boric acid). The boric acid concentration is varied to control long term reactivity changes such as:
- 1. Fuel depletion and fission product buildup.
- 2. Cold to hot, zero power reactivity change.
- 3. Reactivity change produced by intermediate term fission products such as xenon and samarium. 4. Burnable absorber depletion.
Chemical and volume control is covered in Section 9.3.4.
RN 99-134 4.2-37 Reformatted February 2018 The rod cluster control assemblies provide reactivity control for:
- 1. Shutdown. 2. Reactivity changes due to coolant temperature changes in the power range. 3. Reactivity changes associated with the power coefficient or reactivity.
- 4. Reactivity changes due to void formation.
The neutron source assemblies provide a means of monitoring the core during periods of low neutron activity.
The most effective reactivity control components are the rod cluster control assemblies and their CRDMs which are the only kinetic parts in the reactor. Figure 4.2
-13 identifies the full length rod cluster control and CRDM assembly, in addition to the arrangement of these components in the reactor relative to the interfacing fuel assembly and guide tubes. In the following paragraphs, each reactivity control component is described in detail. The guidance system for the control rod cluster is provided by the guide tube as shown in Figure 4.2
-13. The guide tube provides 2 sections of guidance: 1) In the lower section a continuous guidance system provides support immediately above the core. The system protects the rod against excessive deformation and wear due to hydrauli c loading. 2) The region above the continuous section provides support and guidance at uniformly spaced intervals.
The envelope of support is determined by the pattern of the control rod cluster as shown in Figure 4.2
-14. The guide tube assures alignment and support of the control rods, spider body, and drive rod while maintaining trip times at or below required limits.
4.2.3.2.1 Reactivity Control Components 4.2.3.2.1.1 Rod Cluster Control Assembly The rod cluster control assemblies are divided into 2 categories: control and shutdown. The control groups compensate for reactivity changes due to variations in operating conditions of the reactor; i.e., power and temperature variations. Two (2) criteria have been employed for selection of the control group. First the total reactivity worth must be adequate to meet the nuclear requirements of the reactor. Second, in view of the fact that these rods may be partially inserted at power operation, the total power peaking factor should be low enough to ensure that the power capability is met. The control and shutdown group provides adequate shutdown margin which is defined as the amount by which the core would be subcritical at hot shutdown if all rod cluster control assemblies are tripped assuming that the highest worth assembly remains fully withdrawn and assuming no changes in xenon or boron concentration.
4.2-38 Reformatted February 2018 A rod cluster control assembly comprises a group of individual neutron absorber rods fastened at the top end to a common spider assembly, as illustrated in Figure 4.2
-14. The absorber material used in the control rods is silver
-indium-cadmium alloy which is essentially "black" to thermal neutrons and has sufficient additional resonance absorption to significantly increase its worth. The alloy is in the form of extruded rods which are sealed in stainless steel tubes to prevent the rods from coming in direct contact with the coolant. In construction, the silver
-indium-cadmium rods are inserted into cold-worked stainless steel tubing which is then sealed at the bottom and the top by welded end plugs as shown in Figure 4.2
-15. Sufficient diametral and end clearance is provided to accommodate relative thermal expansions. A thin layer of chrome electroplate is applied to the stainless steel cladding. The bottom plugs are made bullet
-nosed to reduce the hydraulic drag during reactor trip and to aid smooth entry into the dashpot section of the fuel assembly guide thimbles. The upper plug is threaded for assembly to the spider and has a reduced end section to make the joint more flexible.
The material used in the absorber rod end plugs is Type 308 stainless steel. The design stresses used for the Type 308 material are the same as those defined in the ASME Code,Section III, for Type 304 stainless steel. At room temperature the yield and ultimate stresses per ASTM
-580 are exactly the same for the 2 alloys. In view of the similarity of the alloy composition, the temperature dependence of strength for the 2 materials is also assumed to be the same.
The allowable stresses used as a function of temperature are listed in Table 1-1.2 of Section III of the ASME Boiler and Pressure Vessel Code. The fatigue strength for the Type 308 material is based on the S
-N curve for austenitic stainless steels in Figu re 1-9.2 of Section III. There are no other applications of stressed wrought Type 308 stainless steel in the control rod assembly.
The spider assembly is in the form of a central hub with radial vanes containing cylindrical fingers from which the absorber rods are suspended. Handling detents and detents for connection to the drive rod assembly are machined into the upper end of the hub. A coil spring inside the spider body absorbs the impact energy at the end of a trip insertion. The radial vanes are joined to the hub by tack welds and brazing while the fingers are joined to the vanes by brazing alone. A centerpost which holds the spring and its retainer is threaded into the hub within the skirt and welded to prevent loosening in service. All components of the spider assembly are made from Types 304 and 308 stainless steel except for the retainer which is of 17
-4 PH material and the springs which are Inconel
-718 alloy or oil tempered carbon steel where the springs do not contact the coolant. The absorber rods are fastened securely to the spider to assure trouble free service. The rods are first threaded into the spider fingers and then pinned to maintain joint RN 96-043 4.2-39 Reformatted February 2018 tightness, after which the pins are welded in place. The end plug below the pin position is designed with a reduced section to permit flexing of the rods to correct for small operating or assembly misalignments.
The overall length is such that when the assembly is withdrawn through its full travel the tips of the absorber rods remain engaged in the guide thimbles so that alignment between rods and thimbles is always maintained. Since the rods are long and slender, they are relatively free to conform to any small misalignments with the guide thimble.
4.2.3.2.1.2 Burnable Absorber Assembly Each burnable absorber assembly consists of borosilicate or wet annular burnable absorber rods attached to a hold down assembly. Conceptual burnable absorber assemblies (containing borosilicate) are shown in Figure 4.2-17. WABA rods may be used in place of the borosilicate rods.
The borosilicate absorber rods consist of borosilicate glass tubes contained within Type 304 stainless steel tubular cladding which is plugged and seal welded at the ends to encapsulate the glass. The glass is also supported along the length of its inside diameter by a thin wall tubular inner liner. The top end of the liner is open to permit the diffused helium to pass into the void volume and the liner overhangs the glass. The liner has an outward flange at the bottom end to maintain the position of the liner with the glass. A typical borosilicate burnable absorber rod is shown in longitudinal and transverse cross
-sections in Figure 4.2
-18. A WABA rod (Figure 4.2
-18a) consists of annular pellets of alumina
-boron carbide (Al 2 O 3-B 4C) burnable absorber material contained within 2 concentric Zircaloy tubes. These Zircaloy tubes, which form the inner and outer clad for the WABA rod, are plugged and welded at each end to encapsulate the annular stack of absorber material. The assembled rod is then internally pre
-pressurized to 650 psig and seal welded. The absorber stack lengths are positioned axially within the WABA rods by the use of Zircaloy bottom
-end spacers. An annular plenum is provided within the rod to accommodate the helium gas released from absorber material depletion during irradiation. The reactor coolant flows inside the inner tube and outside the outer tube of the annular rod. Further design details are given in Section 3.0 of Reference
[26]. The burnable absorber rods are statically suspended and positioned in selected guide thimbles within the fuel assemblies. The absorber rods in each assembly are attached together at the top end of the rods to a hold down assembly by a flat, perforated retaining plate which fits within the fuel assembly top nozzle and rests on the adaptor plate. The absorber rod assembly is held down and restrained against vertical motion through a spring pack which is attached to the plate and is compressed by the upper core plate when the reactor upper internals assembly is lowered into the reactor. This arrangement ensures that the absorber rods cannot be ejected from the core by flow forces. Each rod is permanently attached to the base plate by a nut which is locked into place.
4.2-40 Reformatted February 2018 The borosilicate rod clad is slightly cold worked Type 304 stainless steel, and the WABA rod clad is Zircaloy
- 4. All other structural materials are Types 304 or 308 stainless steel except for the springs which are Inconel
-718. The absorber rods and IFBAs provide sufficient boron content to meet the criteria discussed in Section 4.3.1.
4.2.3.2.1.3 Neutron Source Assembly The purpose of the neutron source assembly is to provide a base neutron level to ensure that the detectors are operational and responding to core multiplication neutrons. Since there is very little neutron activity during loading, refueling, shutdown, and approach to criticality, a neutron source may be placed in the reactor to provide a positive neutron count of at least 2 counts per second on the source range detectors attributable to core neutrons. The detectors, called source range detectors, are used primarily when the core is subcritical and during special subcritical modes of operations.
The source assembly also permits detection of changes in the core multiplication factor during core loading refueling, and approach to criticality. This can be done since the multiplication factor is related to an inverse function of the detector count rate. Therefore a change in the multiplication factor can be detected during addition of fuel assemblies while loading the core, a change in control rod positions, and changes in boron concentration.
Both primary and secondary neutron source rods are used. The primary source rod, containing a radioactive material, spontaneously emits neutrons during initial core loading and reactor startup. After the primary source rod decays beyond the desired neutron flux level, neutrons are then supplied by the secondary source rod. The secondary source rod contains a stable material, which must be activated by neutron bombardment during reactor operation. The activation results in the subsequent release of neutrons.
This becomes a source of neutrons during periods of low neutron flux, such as during refueling and the subsequent startups.
The reactor core may employ 2 secondary source assemblies. Each secondary source assembly contains a symmetrical grouping of 4 secondary source rods. Locations not filled with a source or burnable absorber rod contain a thimble plug. Conceptual source assemblies are shown in Figures 4.2
-19 and 4.2
-20. Neutron source assemblies are employed at opposite sides of the core. The assemblies are inserted into the rod cluster control guide thimbles in fuel assemblies at selected unrodded locations.
The source assemblies contain a hold down assembly identical to that of the burnable absorber assembly.
The primary and secondary source rods both utilize slightly coldworked 304 SS cladding material. The secondary source rods contain 500 grams of stacked antimony
-beryllium pellets and the rod is internally pre
-pressurized to 650 psig. The primary source rods contain capsules of Californium source material and alumina spacer rods to position the RN 99-141 RN 95-022 99-141 4.2-41 Reformatted February 2018 source material within the cladding. The rods in each assembly are permanently fastened at the top end to a hold down assembly, which is identical to that of the burnable absorber assemblies.
The other structural members are fabricated from Type 304 and 308 stainless steel except for the springs exposed to the reactor coolant. They are wound from an age hardened nickel base alloy for corrosion resistance and high strength.
4.2.3.2.1.4 Thimble Plug Assembly If it is desired to further limit bypass flow through the rod cluster control guide thimbles in fuel assemblies which do not contain either control rods, source rods, or burnable absorber rods, the fuel assemblies at those locations may be fitted with thimble plug assemblies.
The thimble plug assemblies as shown in Figure 4.2
-21 consist of a flat base plate with short rods suspended from the bottom surface and a spring pack assembly. The 24 short rods, called thimble plugs, project into the upper ends of the guide thimbles to reduce the bypass flow. Similar short rods are also used on the source assemblies and burnable absorber assemblies to plug the ends of all vacant fuel assembly guide thimbles. At installation in the core, the thimble plug assemblies interface with both the upper core plate and with the fuel assembly top nozzles by resting on the adaptor plate. The spring pack is compressed by the upper core plate when the upper internals assembly is lowered into place. Each thimble plug is permanently attached to the base plate by a nut which is locked to the threaded end of the plug.
All components in the thimble plug assembly, except for the springs, are fabricated from Type 304 and 308 stainless steel. The springs are wound from an age hardened nickel base alloy for corrosion resistance and high strength. 4.2.3.2.2 Control Rod Drive Mechanism All parts exposed to reactor coolant are fabricated of metals which resist the corrosive action of the primary coolant. Three (3) types of metals are used exclusively: stainless steels, nickel
-chrome-iron alloys and cobalt based alloys. In the case of stainless steels, only austenitic and martensitic stainless steels are used.
The discussions (Sections 5.2.5, 5.2.5.1, 5.2.5.2, 5.2.5.3, 5.2.5.4, 5.2.5.5, and 5.2.5.6) concerning the processes, inspections and tests on austenitic stainless steel components to assure freedom from increased susceptibility to intergranular corrosion caused by sensitization, and the discussions (Sections 5.2.5.5 and 5.2.5.7) on the control of welding of austenitic stainless steels, especially control of delta ferrite, are applicable to the austenitic stainless steel pressure retaining components of the control rod drive mechanisms.
Wherever magnetic flux is induced on parts exposed to the main coolant, 400 series stainless steel is used. Cobalt based alloys are used for the pins and latch tips.
4.2-42 Reformatted February 2018 Nickel-chrome-iron alloy is used for the springs of both latch assemblies and Type 304 stainless steel is used for all pressure containing parts as listed in Table 5.2-8. Hard chrome plating provides wear surfaces on the sliding parts and prevents galling between mating parts.
Position indicator assemblies slide over the control rod drive mechanism rod travel housings. Each assembly detects the drive rod position by means of 42 discrete coils that magnetically sense the entry and presence of the rod drive line through its center line over the normal length of the drive rod travel.
Control rod drive mechanisms are located on the dome of the reactor vessel.
They are coupled to rod control clusters which have absorber material over the entire length of the control rods. The full length control rod drive mechanism is shown in Figure 4.2
-22 and schematically in Figure 4.2
-23. The primary function of the control rod drive mechanism is to insert or withdraw a rod cluster control assembly within the core to control average core temperature and to shutdown the reactor.
The control rod drive mechanism is a magnetically operated jack. A magnetic jack is an arrangement of 3 electromagnets which are energized in a controlled sequence by a power cycler to insert or withdraw a rod cluster control assembly in the reactor core in discrete steps. Rapid insertion of the rod cluster control assembly occurs where electrical power is interrupted.
The control rod drive mechanism consists of 4 separate subassemblies.
They are the pressure vessel, coil stack assembly, latch assembly, and the drive rod assembly.
- 1. The pressure vessel includes a latch housing and a rod travel housing which are connected by a threaded, seal welded, maintenance joint which facilitates replacement of the latch assembly. The closure at the top of the rod travel housing is an integral forged end for pressure integrity.
The latch housing is the lower portion of the vessel and contains the latch assembly. The rod travel housing is the upper portion of the vessel and provides space for the drive rod during its upward movement as the control rods are withdrawn from the core.
- 2. The coil stack assembly includes the coil housings, an electrical conduit and connector, and 3 operating coils; 1) the stationary gripper coil, 2) the moveable gripper coil, and 3) the lift coil.
The coil stack assembly is a separate unit which is installed on the drive mechanism by sliding it over the outside of the latch housing. It rests on the base of the latch housing without mechanical attachment.
RN 16-003 RN 16-003 4.2-43 Reformatted February 2018 Energizing the operating coils causes movement of the pole pieces and latches in the latch assembly.
- 3. The latch assembly includes the guide tube, stationary pole pieces, moveable pole pieces, and the 2 sets of latches; 1) the moveable gripper latches and 2) the stationary gripper latches.
The latches engage grooves in the drive rod assembly. The moveable gripper latches are moved up or down in 5/8 inch steps by the lift pole to raise or lower the drive rod. The stationary gripper latches hold the drive rod assembly while the moveable gripper latches are repositioned for the next 5/8 inch step.
- 4. The drive rod assembly includes a flexible coupling, a drive rod, a disconnect button, a disconnect rod, and a locking button.
The drive rod has 5/8 inch grooves which receive the latches during holding or moving of the drive rod. The flexible coupling is attached to the drive rod and provides the means for coupling to the rod cluster control assembly.
The disconnect button, disconnect rod, and locking button provide positive locking of the coupling to the rod cluster control assembly and permits remote disconnection of the drive rod.
The control rod drive mechanism is a trip design. Tripping can occur during any part of the power cycler sequencing if electrical power to the coils is interrupted.
The control rod drive latch mechanism housing is full penetration welded on an adaptor on top of the reactor vessel and is coupled to the rod cluster control assembly directly below. The mechanism is capable of raising or lowering a 360 pound load, (which includes the drive rod weight) at a rate of 45 inches/minute. Withdrawal of the rod cluster control assembly is accomplished by magnetic forces while insertion is by gravity.
The mechanism internals are designed to operate in 650F reactor coolant. The pressure vessel is designed to contain reactor coolant at 650F and 2500 psia. The 3 operating coils are designed to operate at 392F with forced air cooling required to maintain that temperature.
The full length control rod drive mechanism, shown schematically in Figure 4.2
-23, withdraws and inserts a rod cluster control assembly, as shaped electrical pulses are received by the operating coils. An ON or OFF sequence, repeated by silicon controlled rectifiers in the power programmer, causes either withdrawal or insertion of the control rod. Position of the control rod is measured by 42 discrete coils mounted on the position indicator assembly surrounding the rod travel housing. Each coil magnetically RN 16-003 4.2-44 Reformatted February 2018 senses the entry and presence of the top of the ferromagnetic drive rod assembly as it moves through the coil center line.
During plant operation the stationary gripper coil of the drive mechanism holds the rod cluster control assembly in a static position until a stepping sequence is initiated at which time the moveable gripper coil and lift coil is energized sequentially.
- 1. Rod Cluster Control Assembly Withdrawal The rod cluster control assembly is withdrawn by repetition of the following sequence of events (refer to Figure 4.2
-23): a. Movable Gripper Coil (B) ON The latch locking plunger raises and swings the movable gripper latches into the drive rod assembly groove.
A 0.055 inch axial clearance exists between the latch teeth and the drive rod.
Note: The 0.055 inch referred to is a nominal value for a new CRDM at room temperature. This will change with wear and temperature.
- b. Stationary Gripper Coil (A) OFF The force of gravity, acting upon the drive rod assembly and attached control rod, causes the stationary gripper latches and plunger to move downward 0.055 inch until the load of the drive rod assembly and attached control rod is transferred to the movable gripper latches. The plunger continues to move downward and swings the stationary gripper latches out of the drive rod assembly groove.
- c. Lift Coil (C) ON The 5/8 inch gap between the movable gripper pole and the lift pole closes, and the drive rod assembly raises 1 step length (5/8 inch).
- d. Stationary Gripper Coil (A) ON The plunger raises and closes the gap below the stationary gripper pole. The 3 links, pinned to the plunger, swing the stationary gripper latches into a drive rod assembly groove.
The latches contact the drive rod assembly and lift it (and the attached control rod) 0.055 inch. The 0.055 inch vertical drive rod assembly movement transfers the drive rod assembly load from the movable gripper latches to the stationary gripper latches.
RN 16-003 RN 16-003 RN 16-003 RN 16-003 4.2-45 Reformatted February 2018
- e. Movable Gripper Coil (B) OFF The latch locking plunger separates from the movable gripper pole under the force of a spring and gravity. Three (3) links, pinned to the plunger, swing the 3 movable gripper latches out of the drive rod assembly groove.
- f. Lift Coil (C) OFF The gap between the movable gripper pole and lift pole opens.
The movable gripper latches drop 0.625 inch to a position adjacent to a drive rod assembly groove. g. Repeat Step a The sequence described above (Items a through f) is termed as 1 step or 1 cycle. The rod cluster control assembly moves 5/8 inch for each step or cycle. The sequence is repeated at a rate of up to 72 steps per minute and the drive rod assembly (which has a 5/8 inch groove pitch) is raised 72 grooves per minute. The rod cluster control assembly is thus withdrawn at a rate up to 45 inches per minute.
- 2. Rod Cluster Control Assembly Insertion The sequence for rod cluster control assembly insertion is similar to that for control rod withdrawal, except the timing of lift coil (C) ON and OFF is changed to permit lowering the control assembly.
- a. Lift Coil (C) ON The 0.625 inch gap between the movable gripper and lift pole closes. The movable gripper latches are raised to a position adjacent to a drive rod assembly groove.
- b. Movable Gripper Coil (B) ON The latch locking plunger raises and swings the movable gripper latches into a drive rod assembly groove. A 0.055 inch axial clearance exists between the latch teeth and the drive rod assembly.
- c. Stationary Gripper Coil (A) OFF The force of gravity, acting upon the drive rod assembly and attached rod cluster control assembly, causes the stationary gripper latches and plunger to move downward 0.055 inch until the load of the drive rod assembly and attached rod cluster control assembly is transferred to the movable gripper latches. The plunger continues to move downward and swings the stationary gripper latches out of the drive rod assembly groove.
RN 16-003 RN 16-003 RN 16-003 RN 16-003 4.2-46 Reformatted February 2018
- d. Lift Coil (C) OFF The force of gravity and spring force separate the movable gripper pole from the lift pole and the drive rod assembly and attached rod cluster control drop down 5/8 inch.
- e. Stationary Griper (A) ON The plunger raises and closes the gap below the stationary gripper pole. The 3 links, pinned to the plunger, swing the 3 stationary gripper latches into a drive rod assembly groove. The latches contact the drive rod assembly and lift it (and the attached control rod) 1/16 inch. The 1/16 inch vertical drive rod assembly movement transfers the drive rod assembly load from the movable gripper latches to the stationary gripper latches. f. Movable Gripper Coil (B) OFF The latch locking plunger separates from the movable gripper pole under the force of a spring and gravity. Three (3) links, pinned to the plunger, swing the 3 movable gripper latches out of the drive rod assembly groove.
- g. Repeat Step a The sequence is repeated, as for rod cluster control assembly withdrawal, up to 72 times per minute which gives an insertion rate of 45 inches per minute.
- 3. Holding and Tripping of the Control Rods During most of the plant operating time, the control rod drive mechanisms hold the rod cluster control assemblies withdrawn from the core in a static position. In the holding mode, only 1 coil, the stationary gripper coil (A), is energized on each mechanism. The drive rod assembly and attached rod cluster control assemblies hang suspended from the 3 latches.
If power to the stationary gripper coil is cut off, the combined weight of the drive rod assembly and the rod cluster control assembly plus the stationary gripper return spring is sufficient to move latches out of the drive rod assembly groove. The control rod falls by gravity into the core. The trip occurs as the magnetic field, holding the stationary gripper plunger half against the stationary gripper pole, collapses and the stationary gripper plunger half is forced down by the stationary gripper return spring and the weight acting upon the latches. After the rod cluster control assembly is released by the mechanism, it falls freely until the control rods enter the dashpot section of the thimble tubes in the fuel assembly.
4.2-47 Reformatted February 2018 4.2.3.3 Design Evaluation 4.2.3.3.1 Reactivity Control Components The components are analyzed for loads corresponding to normal, upset, emergency and faulted conditions. The analysis performed depends on the mode of operation under consideration.
The scope of the analysis requires many different techniques and methods, both static and dynamic.
Some of the loads that are considered on each component where applicable are as follows: 1. Control Rod Trip (equivalent static load)
- 2. Differential Pressure
- 3. Spring Preloads
- 4. Coolant Flow Forces (static)
- 5. Temperature Gradients
- 6. Differences in thermal expansion
- a. Due to temperature differences
- b. Due to expansion of different materials
- 7. Interference between components
- 8. Vibration (mechanically or hydraulically induced)
- 9. All operational transients listed in Table 5.2
-2 10. Pump Overspeed
- 11. Seismic Loads (operating basis earthquake and safe shutdown earthquake) 12. Blowdown Forces (due to cold and hot leg break)
The main objectives of the analysis are to satisfy allowable stress limits, assure an adequate design margin, and establish deformation limits which are concerned primarily with the functioning of the components. The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also to limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics 4.2-48 Reformatted February 2018 of the materials. Standard methods of strength of materials are used to establish the stresses and deflections of these components. The dynamic behavior of the reactivity control components has been studied using experimental test data and experience from operating reactors
. The design of reactivity component rods provides a sufficient cold void volume within the burnable absorber and source rods to limit the internal pressures to a value which satisfies the criteria in Section 4.2.3.1. The void volume for the helium in th e borosilicate glass burnable absorber rods is obtained through the use of glass in tubular form which provides a central void along the length of the rods. For the WABA rods, an annular void volume is provided between the 2 tubes at the top, and along the length of each WABA rod (Figure 4.2-18a). Helium gas is not released by the neutron absorber rod material, thus the absorber rod only sustains an external pressure during operating conditions. The internal pressure of source rods continues to increase from ambient until end of life at which time the internal pressure never exceeds that allowed by the criteria in Section 4.2.3.1. Except for WABA rods, the stress analysis of reactivity component rods assumes 100% gas release to the rod void volume, considers the initial pressure within the rod, and assumes the pressure external to the component rod is zero. Stress analysis for the WABA rods assumed a maximum 30% gas released, consistent with Reference [26].
Based on available data for properties of the borosilicate glass and on nuclear and thermal calculations for these burnable absorber rods, gross swelling or cracking of the glass tubing is not expected during operation. Some minor creep of the glass at the hot spot on the inner surface of the tube could occur but would continue only until the glass came in contact with the inner liner. The wall thickness of the inner liner is sized to provide adequate support in the event of slumping and to collapse locally before rupture of the exterior cladding if unexpected large volume changes due to swelling or cracking should occur. The top of the inner liner is open to allow communication to the central void by the helium which diffuses out of the glass.
An evaluation of the WABA rod design is given in Referen ce [26]. Sufficient diametral and end clearances have been provided in the neutron absorber, burnable absorber and source rods to accommodate the relative thermal expansions between the enclosed material and the surrounding clad and end plugs. There is n o bending or warping induced in the rods although the clearance offered by the guide thimble would permit a postulated warpage to occur without restraint on the rods. Bending, therefore, is not considered in the analysis of the rods. The radial and axial temperature profiles have been determined by considering gap conductance, thermal expansion, and neutron and/or gamma heating of the contained material as well as gamma heating of the clad. The maximum neutron absorber material temperature was found to be less than 850F which occurs axially at only the highest flux region. The maximum borosilicate glass temperature was calculated to be about 1200F and takes place following the initial rise to power. The glass temperature then decreases rapidly for the following reasons: 1) reduction in power generation due to B 10 depletion; 2) 4.2-49 Reformatted February 2018 better gap conductance as the helium produced diffuses to the gap; and 3) external gap reduction due to borosilicate glass creep. Rod, guide thimble, and dashpot flow analysis performed indicates that the flow is sufficient to prevent coolant boiling and maintain clad temperatures at which the clad material has adequate strength to resist coolant operating pressures and rod internal pressures.
Temperatures for thimbles at the bottom of the fuel assemblies range from approximately 530F to 553F. Mid-assembly temperatures reach a high of about 593F while the maximum temperatures at the top of the assemblies are about 641 F. Analysis on the rod cluster control spider indicates the spider is structurally adequate to withstand the various operating loads including the higher loads which occur during the drive mechanism stepping action and rod drop. Experimental verification of the spider structural capability has been completed (see Section 1.5).
The materials selected are considered to be the best available from the standpoint of resistance to irradiation damage and compatibility to the reactor environment. The materials selected partially dictate the reactor environment (e.g., control in the coolant). The current design type reactivity controls have been in service for many years with no unanticipated degradation of construction materials.
With regard to the materials of construction exhibiting satisfactory resistance to adverse property changes in a radioactive environment, it should be noted that work on breeder reactors in current design, similar materials are being applied. At high fluences the austenitic materials increase in strength with a corresponding decreased ductility (as measured by tensile tests) but energy absorption (as measured by impact tests) remain quite high. Corrosion of the materials exposed to the coolant is quite low and proper control of and O 2 in the coolant will prevent the occurrence of stress corrosion. All of the austenitic stainless steel base materials used are processed and fabricated to preclude sensitization. Although the control rod spiders are fabricated by furnace brazing, the procedure used requires that the pieces be rapidly cooled so that the time-at-temperature is minimized. The time that is spent by the control rod spiders in the sensitization range, 800
- 1500F, is not more than 0.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, as a maximum, during fabrication to preclude sensitization. The 17
-4 PH parts are all aged at the highest standard aging temperature of 1100F to avoid stress corrosion problems exhibited by aging at lower temperatures.
Analysis of the rod cluster control assemblies show that if the drive mechanism housing ruptures, the rod cluster control assembly will be ejected from the core by the pressure differential of the operating pressure and ambient pressure across the drive rod assembly. The ejection is also predicated on the failure of the drive mechanism to retain the drive rod/rod cluster control assembly position. It should be pointed out that a drive mechanism housing rupture will cause the ejection of only 1 rod cluster control assembly with the other assemblies remaining in the core. Analysis also showed that pressure drop in excess of 4000 psi must occur across a two
-fingered vane to break the vane/spider body joint causing ejection of 2 neutron absorber rods from the core. Since RN 95-022 4.2-50 Reformatted February 2018 the greatest pressure drop in the system is only 2250 psi, a pressure drop in excess of 4000 psi is incredible. Thus, the ejection of the neutron absorber rods is not possible.
Ejection of a burnable absorber or thimble plug assembly is conceivable based on the postulation that the holddown bar fails and that the base plate and burnable absorber rods are severely deformed. In the unlikely event that failure of the holddown bar occurs, the upward displacement of the burnable absorber assembly only permits the base plate to contact the upper core plate. Since this displacement is small, the major portion o f the borosilicate glass tubing remains positioned within the core. In the case of the thimble plug assembly, the thimble plugs will partially remain in the fuel assembly guide thimbles thus maintaining a majority of the desired flow impedance. Further displacement or complete ejection would necessitate the square base and burnable absorber rods be forced, thus plastically deformed, to fit up through a smaller diameter hole. It is expected that this condition requires a substantially higher force or pressure drop than that of the holddown bar failure.
Experience with control rods, burnable absorber rods, and source rods are discussed in Reference [4]. The mechanical design of the reactivity control components provides for the protection of the active elements to prevent the loss of control capability and functional failure of critical components. The components have been reviewed for potential failure and consequences of a functional failure of critical parts. The results of the review are summarized below.
4.2.3.3.1.1 Rod Cluster Control Assembly
- 1. The basic absorbing material is sealed from contact with the primary coolant, the fuel assembly and guidance surfaces by a high quality stainless steel clad. Potential loss of absorber mass or reduction in reactivity control material due to mechanical or chemical erosion or wear is therefore reliably prevented.
- 2. A breach of the cladding for any postulated reason does not result in serious consequences. The silver
-indium-cadmium absorber material is relatively inert and would still remain remote from high coolant velocity regions. Rapid loss of material resulting in significant loss of reactivity control material would not occur.
- 3. The individually clad absorber rods are doubly secured to the retaining spider vane by a threaded joint and a welded lock pin.
This joint has been qualified by functional testing and actual service in operating plants.
It should also be noted that in several instances of control rod jamming caused by foreign particles, the individual rods at the site of the jam have borne the full capacity of the control rod drive mechanism and higher impact loads to dislodge the jam without failure.
The conclusion to be drawn from this experience is that this joint is extremely insensitive to potential mechanical damage.
A failure of the RN 95-022 4.2-51 Reformatted February 2018 joint would result in the insertion of the individual rod into the core resulting in reduced reactivity. 4. The spider finger braze joint by which the individual rods are fastened to the vanes has also experienced the service described above and been subjected to the same jam freeing procedures also without failure. A failure of this joint would also result in insertion of the individual rod into the core.
- 5. The radial vanes are attached to the spider body, again by a brazed joint. The joints are designed to a theoretical strength in excess of that of the components joined. It is a feature of the design that the guidance of the rod cluster control is accomplished by the inner fingers of these vanes. They are therefore the most susceptible to mechanical damage. Since these vanes carry 2 rods, failure of the vane-to-hub joint such as the isolated incidents at Connecticut
-Yankee does not prevent the free insertion of the rod pair.
Neither does such a failure interfere with the continuous free operation of the drive line, also as experienced at Connecticut
-Yankee. Failure of the vane
-to-hub joint of a single rod vane could potentially result in failure of the separated vane and rod to insert. This could occur only at withdrawal elevations where the spider is above the continuous guidance section of the guide tube (in the upper internals).
A rotation of the disconnected vane could cause it to hang on one of the guide cards in the intermediate guide tube. Such an occurrence would be evident from the failure of the rod cluster control to insert below a certain elevation but with free motion above this point. This possibility is considered extremely remote because the single rod vanes are subjected to only vertical loads and very light lateral reactions from the rods. The consequences of such a failure are not considered critical since only 1 drive line of the reactivity control system would be involved. This condition is readily observed and can be corrected at shutdown.
- 6. The spider hub being of single unit cylindrical construction is very rugged and of extremely low potential for damage. It is difficult to postulate any condition to cause failure. Should some unforeseen event cause fracture of the hub above the vanes, the lower portion with the vanes and rods attached would insert by gravity into the core causing a reactivity decrease. The rod could then not be removed by the drive line. Fracture below the vanes cannot be postulated since all loads, including scram impact, are taken above the vane elevation.
- 7. The rod cluster control rods are provided a clear channel for insertion by the guide thimbles of the fuel assemblies. All fuel rod failures are protected against by providing this physical barrier between the fuel rod and the intended insertion channel. Distortion of the fuel rods by bending cannot apply sufficient force to 4.2-52 Reformatted February 2018 damage or significantly distort the guide thimble. Fuel rod distortion by swelling, though precluded by design, would be terminated by fracture before contact with the guide thimble occurs. If such were not the case, it would be expected that a force reaction at the point of contact would cause a slight deflection of the guide thimble. The radius of curvature of the deflected shape of the guide thimbles would be sufficiently large to have a negligible influence on rod cluster control insertion.
4.2.3.3.1.2 Burnable Absorber Assemblies The burnable absorber assemblies are static temporary reactivity control elements. The axial position is assured by the holddown assembly which bears against the upper core plate. Their lateral position is maintained by the guide thimbles of the fuel assemblies.
The individual rods are shouldered against the underside of the retainer plate and securely fastened at the top by a threaded nut which is then locked in place. The square dimension of the retainer plate is larger than the diameter of the flow holes through the core plate. Failure of the holddown bar or spring pack therefore does not result in ejection of the burnable absorber rods from the core.
The only incident that could potentially result in ejection of the burnable absorber rods is a multiple fracture of the retainer plate. This is not considered credible because of the light loads borne by this component. During normal operation the loads borne by the plate are approximately 5 pounds/rod or a total of 100 pounds distributed at the points of attachment. Even a multiple fracture of the retainer plate would result in jamming of the plate segments against the upper core plate, again preventing ejection. Excessive reactivity increase due to burnable absorber rod ejection is therefore prevented.
Burnable absorber rods are clad with either stainless steel or Zircaloy
- 4. The burnable absorber is either a borosilicate glass tube which is maintained in position by a central hollow stainless steel tube or Al 2 O 3-B 4 C annular pellets contained within 2 concentric Zircaloy tubes. Burnable absorber rods are placed in static assemblies and are not subjected to motion which might damage the rods. Further, the guide thimble tubes of the fuel assembly afford additional protection from damage.
During the accumulated thousands of years of burnable absorber rodlet operating experience, only 1 instance of penetration of the stainless steel burnable absorber cladding has been observed. The consequences of clad breach are also small. It is anticipated that upon clad breach, the B 4C or borosilicate glass would be leached by the coolant water and that localized power peaking of a few percent would occur; no design criteria would be violated. Additional information on the consequences of postulated WABA rod failures is presented in Reference
[26]. RN 95-022 4.2-53 Reformatted February 2018 4.2.3.3.1.3 Drive Rod Assemblies All postulated failures of the drive rod assemblies either by fracture or uncoupling lead to a decrease in reactivity. If the drive rod assembly fractures at any elevation, that portion remaining coupled falls with, and is guided by the rod cluster control assembly. This always results in reactivity decrease for control rods.
4.2.3.3.2 Control Rod Drive Mechanism 4.2.3.3.2.1 Material Selection All pressure containing materials comply with Section III of the ASME Boiler and Pressure Vessel Code, and are fabricated from austenitic (Type 304) stainless steel.
Magnetic pole pieces are fabricated from Type 410 stainless steel. All nonmagnetic parts, except pins and springs, are fabricated from Type 304 stainless steel. Haynes 25 is used to fabricate link pins. Springs are made from nickel
-chrome-iron alloy. Latch arm tips are clad with Cobalt-6 to provide improved wearability. Hard chrome plate and Cobalt-6 are used selectively for bearing and wear surfaces. At the start of the development program, a survey was made to determine whether a material better than Type 410 stainless steel was available for the magnetic pole pieces. Ideal material requirements are as follows:
- 1. High magnetic saturation value.
- 2. High permeability.
- 3. Low coercive force.
- 4. High resistivity.
- 5. High curie temperature.
- 6. Corrosion resistance.
- 7. High impact strength.
- 8. Nonoriented.
- 9. High machinability.
- 10. Low susceptibility to radiation damage.
After a comprehensive material study was made it was decided that the Type 410 stainless steel was satisfactory for this application. RN 16-003 RN 16-003 4.2-54 Reformatted February 2018 The cast coil housings require a magnetic material. Both low carbon cast steel and ductile iron have been successfully tested for this application.
The choice, made on the basis of cost, indicates that ductile iron will be specified on the control rod drive mechanism. The finished housings are zinc plated or flame sprayed to provide corrosion resistance.
Coils are wound on bobbins of molded Dow Corning 302 material, with double glass insulated copper wire. Coils are then vacuum impregnated with silicon varnish. A wrapping of mica sheet is secured to the coil outside diameter. The result is a well insulated coil capable of sustained operation at 200 centigrade.
The drive rod assembly utilizes a Type 410 stainless steel drive rod. The coupling is machined from Type 403 stainless steel. Other parts are Type 304 stainless steel with the exception of the springs which are nickel
-chrome-iron alloy and the locking button which is Haynes 25.
4.2.3.3.2.2 Radiation Damage As required by the equipment specification, the control rod drive mechanisms are designed to meet a radiation requirement of 10 rads/hour.
Materials have been selected to meet this requirement. The above radiation level which amounts to 1.753 x 10 6 rads in 20 years will not limit control rod drive mechanism life. Control rod drive mechanisms at Yankee Rowe which have been in operation since 1960 have not experienced problems due to radiation.
4.2.3.3.2.3 Positioning Requirements The control rod drive mechanism has a step length of 5/8 inch which determines its positioning capabilities. Positioning requirements are determined by reactor phy sics. 4.2.3.3.2.4 The ability of the pressure housing components to perform throughout the design lifetime as defined in the equipment specification is confirmed by the stress analysis report required by the ASME Boiler and Pressure Vessel Code,Section III. Internal components subjected to wear will withstand a minimum of 3,000,000 steps without refurbishment as confirmed by life tests. Latch assembly inspection is recommended after 2.5 x 10 6 steps have been accumulated on a single control rod drive mechanism.
4.2-55 Reformatted February 2018 4.2.3.3.2.5 Results of Dimensional and Tolerance Analysis With respect to the control rod drive mechanism system as a whole, critical clearances are present in the following areas:
- 1. Latch assembly (Diametral clearances).
- 2. Latch arm-drive rod clearances.
- 3. Coil stack assembly
-thermal clearances.
- 4. Coil fit in coil housing.
The following write
-up defines clearances that are designed to provide reliable operation in the control rod drive mechanism in these 4 critical areas. These clearances have been proven by life tests and actual field performance at operating plants.
4.2.3.3.2.6 Latch Assembly
- Thermal Clearances The magnetic jack has several clearances where parts made of Type 410 stainless steel fit over parts made from Type 304 stainless steel.
Differential thermal expansion is therefore important. Minimum clearances of these parts at 68F is 0.011 inch. At the maximum design temperature of 650F minimum clearance is 0.0045 inch and at the maximum expected operating temperatures of 550F is 0.0057 inch.
4.2.3.3.2.7 Latch Arm
- Drive Rod Clearances The control rod drive mechanism incorporates a load transfer action. The movable or stationary gripper latch are not under load during engagement, as previously explained, due to load transfer action.
Figure 4.2
-26 shows latch clearance variation with the drive rod as a result of minimum and maximum temperatures. Figure 4.2
-27 shows clearance variations over the design temperature range.
4.2.3.3.2.8 Coil Stack Assembly
- Thermal Clearances The assembly clearance of the coil stack assembly over the latch housing was selected so that the assembly could be removed under all anticipated conditions of thermal expansion.
At 70F the inside diameter of the coil stack is 7.308/7.298 inches. The outside diameter of the latch housing is 7.260/7.270 inches.
4.2-56 Reformatted February 2018 Thermal expansion of the mechanism due to operating temperature of the control rod drive mechanism results in the minimum inside diameter of the coil stack being 7.310 inches at 222F and the maximum latch housing diameter being 7.302 inches at 532 F. Under the extreme tolerance conditions listed above, it is necessary to allow time for a 70F coil housing to heat during a replacement operation.
Four (4) coil stack assemblies were removed from 4 hot control rod drive mechanisms mounted on 11.035 inch centers on a 550F test loop, allowed to cool, and then replaced without incident as a test to prove the preceding.
4.2.3.3.2.9 Coil Fit in Coil Housing Control rod drive mechanism and coil housing clearances are selected so that coil heat up results in a close to tight fit. This is done to facilitate thermal transfer and coil cooling in a hot control rod drive mechanism.
4.2.3.3.2.10 Protection from Pipe Rupture and Missiles The control rod drive mechanisms are protected from postulated pipe ruptures and missiles as discussed in Sections 3.6 and 3.5, respectively.
4.2.3.4 Tests, Verification and Inspections 4.2.3.4.1 Reactivity Control Components Tests and inspections are performed on each reactivity control component to verify the mechanical characteristics. In the case of the rod cluster control assembly, prototype testing has been conducted and both manufacturing test/inspections and functional testing at the plant site are performed.
During the component manufacturing phase, the following requirements apply to the reactivity control components to assure the proper functioning during reactor operation:
- 1. All materials are procured to specifications to attain the desired standard of quality.
- 2. A spider from each braze lot is proof tested by applying a 5000 pound load to the spider body, so that approximately 310 pounds is applied to each vane. This proof load provides a bending moment at the spider body approximately equivalent to 1.4 times the load caused by the acceleration imposed by the control rod drive mechanism.
- 3. All clad/end plug welds are checked for integrity by visual inspection, X
-ray, and are helium leak checked. All the seal welds in the neutron absorber rods, burnable absorber rods and source rods are checked in this manner.
RN 95-022 4.2-57 Reformatted February 2018
- 4. To assure proper fit with the fuel assembly, the rod cluster control, burnable absorber and source assemblies are installed in the fuel assembly without restriction or binding in the dry condition. Also a straightness of 0.01 in/ft is required on the entire inserted length of each rod assembly.
The rod cluster control assemblies are functionally tested, following core loading but prior to criticality to demonstrate reliable operation of the assemblies.
In order to demonstrate continuous free movement of the rod cluster control assemblies, and to ensure acceptable core power distributions during operations, partial movement checks are performed on the rod cluster control assemblies as required by the Technical Specifications. In addition, periodic drop tests of the rod cluster control assemblies are performed at each refueling shutdown to demonstrate continued ability to meet trip time requirements. During these tests the acceptable drop time of each assembly is not greater than 2.7 seconds, at full flow and operating temperatures, from the beginning of motion to dashpot entry.
If a rod cluster control assembly cannot be moved by its mechanism, adjustments in the boron concentration ensure that adequate shutdown margin would be achieved following a trip. Thus, inability to move 1 rod cluster control assembly can be tolerated. More than 1 inoperable rod cluster control assembly could be tolerated, but would impose additional demands on the plant operator. Therefore, the number of inoperable rod cluster control assemblies has been limited to 1.
4.2.3.4.2 Control Rod Drive Mechanisms Quality assurance procedures during production of control rod drive mechanisms include material selection, process control, mechanism component tests and inspections during production and hydrotests.
After all manufacturing procedures had been developed, several prototype control rod drive mechanisms and drive rod assemblies were life tested with the entire drive line under environmental conditions of temperature, pressure and flow. All acceptance tes ts confirm the 3 x 10 6 step unrefurbished life capability of the control rod drive mechanism and drive rod assembly.
These tests include verification that the trip time achieved by the control rod drive mechanisms meet the original design requirement of 2.7 seconds from start of rod cluster control assembly motion to dashpot entry. This trip time requirement will be confirmed for each control rod drive mechanism prior to initial reactor operation and at periodic intervals after initial reactor operation. In addition, a Technical Specification has been set to ensure that the trip time requirement is met.
It is expected that all control rod drive mechanisms will meet specified operating requirements for the duration of plant life with normal refurbishment. However, a Technical Specification pertaining to an inoperable rod cluster control assembly has been set. 98-01 98-01 98-01 4.2-58 Reformatted February 2018 To confirm the mechanical adequacy of the fuel assembly, the control rod drive mechanism, and rod cluster control assembly, functional test programs have been conducted on a full scale 12 foot control rod. The 12 foot prototype assembly was tested under simulated conditions of reactor temperature, pressure, and flow for approximately 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />. The prototype mechanism accumulated about 3,000,000 steps and 600 trips. At the end of the test the control rod drive mechanism was still operating satisfactorily. A correlation was developed to predict the amplitude of flow excited vibration of individual fuel rods and fuel assemblies. Inspection of the drive line components did not reveal significant fretting.
There are no significant differences between the prototype control rod drive mechanisms and the replaced production units. Design materials, tolerances and fabrication techniques (Section 4.2.3.3.2) are the same, or as noted
. Actual experience in many operating Westinghouse plants, indicates excellent performance of these control rod drive mechanisms.
All units are production tested prior to shipment to confirm ability of the control rod drive mechanism to meet design specification operational requirements.
Each control rod drive mechanism undergoes a production test as listed below:
Test Acceptance Criteria Cold (ambient) hydrostatic ASME Section III Confirm step length and load transfer (stationary gripper to movable gripper or movable gripper to stationary gripper) Step Length 5/8 0.015 inch axial movement Load Transfer 0.055 0.007 inch nominal axial movement Cold (ambient) performance Test at design load
- 5 full travel excursions Operating Speed 45 inches/minute Trip Delay Free fall of drive rod to begin within 150 MS 4.2.3.5 Instrumentation Applications Instrumentation for determining reactor coolant average temperature (Tavg) is provided to create demand signals for moving groups of rod cluster control assemblies to provide load follow (determined as a function of first stage turbine pressure) during normal operation and to counteract operational transients. The hot and cold leg resistance temperature detectors (RTDs) are described in Section 7.2 in the reactor coolant bypass loops. The location of the RTDs in each loop is shown on the flow diagrams in Chapter 5.0. The reactor control system which controls the reactor coolant average temperature by regulation of control rod bank position is described in Section 7.3. RN 16-003 RN 16-003 4.2-59 Reformatted February 2018 Rod position indication instrumentation is provided to sense the actual position of each control rod so that the actual position of the individual rod may be displayed to the operator. Signals are also supplied by this system as input to the rod deviation comparator. The rod position indication system is described in Chapter 7.0.
The reactor makeup control system whose functions are to permit adjustment of the reactor coolant boron concentration for reactivity control (as well as to maintain the desired operating fluid inventory in the volume control tank), consists of a group of instruments arranged to provide a manually preselected makeup composition that is borated or diluted as required to the charging pump suction header or the volume control tank. This system, as well as other systems including boron sampling provisions that are part of the chemical and volume control system, are described in Section 9.3.
Monitoring of the neutron flux for various phases of reactor power operation as well as of core loading, shutdown, startup, and refueling is by means of the nuclear instrumentation system. The monitoring functions, readout and indication characteristics for the following means of monitoring reactivity during these phases are included in the discussion on safety
-related display instrumentation in Section 7.5:
- 1. Nuclear Instrumentation System
- 2. Temperature Indicators
- a. T average (Measured)
- b. T (Measured)
- c. Auctioneered T average
- d. T reference
- 3. Demand Position of Rod Cluster Control Assembly Group
- 4. Actual Rod Position Indicator.
4.
2.4 REFERENCES
- 1. Christensen, J. A., Alio, R. J., and Biancheria, A., "Melting Point of Irradiated UO 2," WCAP-6065, February 1965. 2. Eggleston, F. T., "Safety
-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries, Spring, 1976," WCAP
-8768, June 1976. 3. Eng, G. H., George, R. A., Lee, and V. C., "Revised Clad Flattening Model," WCAP-8377 (Proprietary) and WCAP
-8381 (Non-Proprietary), July 1974.
4.2-60 Reformatted February 2018
- 4. Slagle, W. H., "Operational Experience with Westinghouse Cores Through December 31, 1994," WCAP
-8183, Revision 23, January 1996.
- 5. Miller, J. V.
(Ed.), "Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations," WCAP
-8720 (Proprietary) and WCAP
-8785 (non-Proprietary), October 1976. 6. Hellman, J. M., (Ed.), "Fuel Densification Experimental Results and Model for Reactor Application," WCAP
-8218-P-A (Proprietary) and WCAP
-8219-A (Non-Proprietary), March 1975.
- 7. ZIRLO and Optimized ZIRLO-12610-P-A & CENPD-404-P-A Addendum 2
-A (Proprietary) and WCAP-14342-A & CENPD-404-NP-A Addendum 2
-A (Non-Proprietary), October 2013. 8. Wood, D. S., "High Deformation Creep Behavior of 0.6 Inch Diameter Zirconium Alloy Tubes Under Irradiation," ASTM
-STP-551, American Society for Testing and Materials, 1973.
- 9. . and Longer, B. F., "Fatigue Design for Zircaloy Components," Nuclear Science and Engineering, 20, 1
-12, 1964. 10. Skaritka, J. (Ed.), "Fuel Rod Bow Evaluations," WCAP
-8691, Revision 1 (Proprietary), July 1979. 11. Gesinski, L., Chiang, D., and Nakazato, S., "Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP
-8236 (Proprietary), December, 1973 and WCAP
-8288 (Non-Proprietary), January 1974.
- 12. DeMario, E. E., "Hydraulic Flow Test of the 17 x 17 Fuel Assembly," WCAP
-8278 (Proprietary) and WCAP
-8279 (Non-Proprietary), February 1974.
- 13. Dollard, W. J., "Nuclear Fuel Division Quality Assurance Program Plan," WCAP
-7800, Revision 7
-A, December 1988.
- 14. Kraus, S., "Neutron Shielding Pads," WCAP
-7870, May 1972.
- 15. Cohen, J., "Development and Properties of Silver Base Alloys as Control Rod Materials for Pressurized Water Reactors," WAPD
-214, December 1959.
- 16. TM-12610-P-A & CENPD-404-P-A Addendum 1-A (Proprietary) and WCAP
-14342-A & CENPD-404-NP-A Addendum 1-A (Non-Proprietary), June 2006.
RN 95-022 96-043 RN 95-022 RN 14-036 RN 14-036 4.2-61 Reformatted February 2018
- 17. -15063-P-A Revision 1, with Errata (Proprietary) and WCAP-15064-NP-A Revision 1, with Errata (Non
-Proprietary), July 2000.
- 18. 004, Revision 2, July 2012.
- 19. Deleted 20. Risher, D. H., et al., "Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP
-8963 (Proprietary) and WC AP-8964 (Non-Proprietary), June 1977. 21. Deleted 22. Deleted 23. Davidson, S. L. (Ed.), et al., "Extended Burnup Evaluation of Westinghouse Fuel," WCAP-10125-P-A (Proprietary), December 1985 WCAP-10125-P-A, Addendum 1
-A, Revision 1
-A, May 2005.
- 24. Davidson, S. L. (Ed.), "Reference Core Report - VANTAGE 5 Fuel Assembly," WCAP-10444-P-A, September 1985. 25. Davidson, S. L., et al., (Ed.), "Verification Testing and Analysis of the 17 x 17 Optimized Fuel Assembly," WCAP
-9401-P-A, August 1981. 26. Skaritka, J., "Westinghouse Wet Annular Burnable Absorber Evaluation Report," WCAP-10021-P-A, Revision 1, October 1983. 27. Davidson, S.L., and Nuhfer, D.L., "VANTAGE+ Fuel Assembly Reference Core Report," WCAP
-12610 (Proprietary), June 1990.
- 28. Weiner, R.A., et al., "Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," WCAP
-10851-P-A (Proprietary) and WCAP
-11873-A (Non Proprietary), August 1988.
- 29. Thadani, A.C., "Acceptance For Referencing of Topical Report", WCAP
-12610 "VANTAGE+ Fuel Assembly Reference Core Report," (TAC No. 77258), July 1, 1991. 30. Slagle, W.H. (EMethodology," WCAP
-13060-P, September 1991.
- 31. Davidson, S. L. (Ed), "Reference Core Report
- VANTAGE 5 Fuel Assembly
," WCAP-10444-P-A , September 1985.
RN 09-022 RN 95-022 RN 14-005 RN 14-036 RN 15-022 4.2-62 Reformatted February 2018
- 32. Gergos, B. W., WCAP-16980, "Reactor Internals Upflow Conversion Program Engineering Report V. C. Summer Nuclear Plant
," December 2008.
- 33. Bamford, W. H., et Basis for the Embedded Flaw Process for Repair -15987-P Revision 2
-P-A (Proprietary), December 2003.
- 34. Letter from Chief Robert J. Pascarelli (NRC) to Mr. Thomas D. Summer Nuclear Station, Unit 1 Alternative Request Weld Repair for Reactor RN 15-023 RN 09-022 4.2-63 Reformatted February 2018 TABLE 4.2-1 MAXIMUM DEFLECTIONS ALLOWED FOR REACTOR INTERNAL SUPPORT STRUCTURES Component Allowable Deflections (in.)
No-Loss-of Function Deflections (in.)
Upper Barrel radial inward 4.1 8.2 radial outward 1.0 1.0 Upper Package (1) 0.10 0.15 Rod Cluster Guide Tubes 1.00 1.75 (1) The vertical motion of the upper core plate relative to the upper support plate shall not cause buckling of the guide tubes.
RN 01-113
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c::>:3-c:::>c::=::::>L£3\I--c::=::::>c::=::::>S-F c:::::..-F-l-c:::>c::......., GUIDE lutE-l-F-l 13 l-I=>b SV'POIl PlaT(SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Upper Core Support Assembly Figure 4.2-11 SUPPORT COLUMNS DEEP BEAM WELDMENTS GUIDE TUBEASSEMBLY LOCATIONS IHEAD AND VESSEL ALIGNMENT PIN LOCATION SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Plan View of Upper Core Support Structure Figure 4.2-12 SUPPORT COLUMNS DEEP BEAM WELDMENTS GUIDE TUBEASSEMBLY LOCATIONS IHEAD AND VESSEL ALIGNMENT PIN LOCATION SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Plan View of Upper Core Support Structure Figure 4.2-12 RCC ASSEM8L Y WITHDRAWN CHOM DRIVE ROO ASSEMBLY FULL RCC ASSEMBL Y GU I DE TUBE ASSEMBL Y FUEL ASSEMBLY SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STATION Full length Rod Cluster Control and Drive Rod Assembly with Interfacing II'Components I Figure 4.2-13 RCC ASSEM8L Y WITHDRAWN CHOM DRIVE ROO ASSEMBLY FULL RCC ASSEMBL Y GU I DE TUBE ASSEMBL Y FUEL ASSEMBLY SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STATION Full length Rod Cluster Control and Drive Rod Assembly with Interfacing II'Components I Figure 4.2-13 S Pi DE R 0.361 DIA 161.0 0.385 D IA MAX LENGTH ABSORBERSILVER 15"1.I ND I UM 5'1, CADMI UM SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION full Length Rod Cluster Control Assembly Outline Figure 4.2-14 S Pi DE R 0.361 DIA 161.0 0.385 D IA MAX LENGTH ABSORBERSILVER 15"1.I ND I UM 5'1, CADMI UM SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION full Length Rod Cluster Control Assembly Outline Figure 4.2-14 0.381 OIA.HOM'1 1l12.00 ABOSRBER MATERIAL.1'J-...
I::: I 151.73 I I t-SOUTH CAROLINA ElECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STA TION Full length Absorber Rod Figure 4.2-15 0.381 OIA.HOM'1 1l12.00 ABOSRBER MATERIAL.1'J-...
I::: I 151.73 I I t-SOUTH CAROLINA ElECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STA TION Full length Absorber Rod Figure 4.2-15 STA INLESS STL 151.60 142.0 REF.r PO I SON LENGTH 2.,&::.I-r.......-------_.-I-=---r-l-----fA.1\I I r I.(ti1itiIiJ I"" IlIl!III!77/"7_?r--f-------....----'" I I///-L---'---"'t:I I-J-/,J;:.'--SPRINGS J---II--" BURNABLE POISON ROO TH IMBLE PLUG AMENDMENT6 AUGUST,1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Borosilicate Absorber Assembly Figure 4.2-17 STA INLESS STL 151.60 142.0 REF.r PO I SON LENGTH 2.,&::.I-r.......-------_.-I-=---r-l-----fA.1\I I r I.(ti1itiIiJ I"" IlIl!III!77/"7_?r--f-------....----'" I I///-L---'---"'t:I I-J-/,J;:.'--SPRINGS J---II--" BURNABLE POISON ROO TH IMBLE PLUG AMENDMENT6 AUGUST,1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Borosilicate Absorber Assembly Figure 4.2-17 0 z:<r 0-..00'" 0 r" t1">r--I.--00 l.L.W 0"'"<.!J.....z t.Jww"'"....J o cr:<o UJ<Na:l::t'cr:-0 (/)a:l<l: o o 1..'"<i-<z: oe;.;>w (/)-'-wll'l:w'..L.AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Borosilicate Absorber Rod Cross Section Figure 4.2-180 z:<r 0-..00'" 0 r" t1">r--I.--00 l.L.W 0"'"<.!J.....z t.J w w"'"....J o cr:<o UJ<Na:l::t'cr:-0 (/)a:l<l: o o 1..'"<i-<z: oe;.;>w (/)-'-wll'l:w'..L.AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Borosilicate Absorber Rod Cross Section Figure 4.2-18 150.0 TYPICAL FLOW'ATH f*UPPER END PLUG ZlRCALOY lUBES....._HOlDDOWN DEVICE FLOW PATH 1.0 TYPICAl.ABSORBER LENGTH AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Wet Annular Burnable Absorber Rod Figure 4.2*18a 150.0 TYPICAL FLOW'ATH f*UPPER END PLUG ZlRCALOY lUBES....._HOlDDOWN DEVICE FLOW PATH 1.0 TYPICAl.ABSORBER LENGTH AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Wet Annular Burnable Absorber Rod Figure 4.2*18a
""'150.4 1.5 NOM.\-CALIFORNIUM
-,-
142.0 REF PO'SON LENGTH------------lI1lOO1 NOTE ALL OIMENSIONS ARE IN INCHES 0.381 o I A.JBl OIA PRIMARY SCOURCE BURNABLE POISON BOROS III CA TE GLASS TUBE SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Primary Source Assembly Figure 4.2-19""'150.4 1.5 NOM.\-CALIFORNIUM
-,-
142.0 REF PO'SON LENGTH------------lI1lOO1 NOTE ALL OIMENSIONS ARE IN INCHES 0.381 o I A.JBl OIA PRIMARY SCOURCE BURNABLE POISON BOROS III CA TE GLASS TUBE SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Primary Source Assembly Figure 4.2-19 ANTIMONY BERYLLIUM 88.0 SECONOARY SOURCE SECONOARY SOURCE , 0.381 OIA STAINLESS STU L THIMBLE*PLUG*FITS ONLY LOPAR ASSEMBLY THIMBLE TUBES.TO FIT INTO VANTAGE 5 THIMBLE TUBES, MUST USE OUALLY COMPATIBLE PLUG DESIGN WITH 0.424 INCH DIAMETER.NOTE ALL OIMENSIONS ARE IN INCHES AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Secondary Source Assembly Figure 4.2-20 ANTIMONY BERYLLIUM 88.0 SECONOARY SOURCE SECONOARY SOURCE , 0.381 OIA STAINLESS STU L THIMBLE*PLUG*FITS ONLY LOPAR ASSEMBLY THIMBLE TUBES.TO FIT INTO VANTAGE 5 THIMBLE TUBES, MUST USE OUALLY COMPATIBLE PLUG DESIGN WITH 0.424 INCH DIAMETER.NOTE ALL OIMENSIONS ARE IN INCHES AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Secondary Source Assembly Figure 4.2-20 0.424" OIA.(BOTH LOPAR ANO VANTAGE 5 ASSY.)-----.,E:::7-
--D c:(]---[0.434" OIA.(lOPAR ASSY ONLY)
AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Thimble Plug Assembly Figure 4.2-21 0.424" OIA.(BOTH LOPAR ANO VANTAGE 5 ASSY.)-----.,E:::7-
--D c:(]---[0.434" OIA.(lOPAR ASSY ONLY)
AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Thimble Plug Assembly Figure 4.2-21
RN 16-003 BEFORE LOAD TRANSFER AFTER LOAD TRANSFER M6 LIFT COIL ON 16.258 16.203 A AT 650°A B CD LIFT COIL OFF 15.679 0.038 LI FT COl L ON 16.387 16.291 0.01656 (SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Nominal Latch Clearance at Minimum and Maximum Temperature Figure 4.2-26 LIFT COIL OFF 15.625 15.578 A B 1I FT COl L OFF LI FT COl L ON LI FT COl L OFF II FT COl L ON 56 AT 70°A B (0 15.6 1 W 15.625 0.01516.26516.250 0.015 AT 650°A B C0 15.725 15.679 16.375 16.387 0.068 B A B BEFORE LOAD TRANSFER AFTER LOAD TRANSFER M6 LIFT COIL ON 16.258 16.203 A AT 650°A B CD LIFT COIL OFF 15.679 0.038 LI FT COl L ON 16.387 16.291 0.01656 (SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Nominal Latch Clearance at Minimum and Maximum Temperature Figure 4.2-26 LIFT COIL OFF 15.625 15.578 A B 1I FT COl L OFF LI FT COl L ON LI FT COl L OFF II FT COl L ON 56 AT 70°A B (0 15.6 1 W 15.625 0.01516.26516.250 0.015 AT 650°A B C0 15.725 15.679 16.375 16.387 0.068 B A B 0.080 700 600------500--------300 1100 TEMPERATURE (OF)LIFT COIL OFF l MG CLEARANCE AFTER'LIFT COIL ON f LOAD TRANSFER 200 100 ,,-//--/----.A'..----/-\L lIFT COIL 0'}SG CLEARANCE BEFORE LIFT COIL OFF LOAD TRANSFER (/)0.070 UJ:x: c:...>z: 0.060 UJ:>0 C 0::: 0.050w:>0::: 0.01l0 Cl 0:z: e((/)0.030 UJ:x: c:...>l-e(0.020-J UJ c:...>z: e(0.010 0::: e(UJ-J c:...>0.000 0 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Control Rod Drive Mechanism Latch Clearance Thermal Effect Figure 4.2-27 0.080 700 600------500--------300 1100 TEMPERATURE (OF)LIFT COIL OFF l MG CLEARANCE AFTER'LIFT COIL ON f LOAD TRANSFER 200 100 ,,-//--/----.A'..----/-\L lIFT COIL 0'}SG CLEARANCE BEFORE LIFT COIL OFF LOAD TRANSFER (/)0.070 UJ:x: c:...>z: 0.060 UJ:>0 C 0::: 0.050w:>0::: 0.01l0 Cl 0:z: e((/)0.030 UJ:x: c:...>l-e(0.020-J UJ c:...>z: e(0.010 0::: e(UJ-J c:...>0.000 0 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Control Rod Drive Mechanism Latch Clearance Thermal Effect Figure 4.2-27 LEGEND: tVV\IMPACT SPRING ELEMENT-n-VI SCOUS DAMP I NG ELEMENT-lGAP ELEMENTMASS ELEMENT......4-F 3 (t)UPPER CORE PlATE FUEL ASSEMBLY..1-tJ-f1 IIII II LI II F 2{t)LOWER CO.RE PLATE F 3 (t)......BARREL AT UPPER CORE PLATE£LEVATI ON<Vl-0:x:I e r-I (\f"'l Vl>>e:x:l m Z:x:I>>em f"'lf"'l r--I m:x:l:x:If"'l Vl QO-I">>>>:::!Vl Of"'l z9 Vl ('\:::r ro 3 OJ r+n':x:I f"'lro 0"0roro II> oQ.OJ ro r+-o'::J o....:x:I ro OJ ('\r+o"T1\C c:roIV N 00 tVV\IMPACT SPRING ELEMENT-n-VI SCOUS DAMP I NG ELEMENT LEGEND:-lGAP ELEMENTMASS ELEMENT F 2 (t) F 3 (t}....BARREL AT UPPER CORE PLATE ELEVATI ON VI<VI-0 ('\:x:I c:::r ro 3 r-I OJ t""It""I z..>>('\VI:x:I:x:I Co"T1 t""I ro\C 0"Q c:roro II>roIV o.....cm N Q.OJ t""It""I ro z.r--l 00-0 m:x:l::J>>n 0....:x:IQO:x:I ro OJ>>>>('\-lVl.....Ot""l 0z9 FUEL ASSEMBLY UPPER CORE PlATE
4.3-53 Reformatted Per Amendment 02-01 TABLE 4.3-3 REACTIVITY REQUIREMENTS FOR ROD CLUSTER CONTROL ASSEMBLIES (Typical Reload Cycle)
Reactivity Effects, Percent Beginning of Life End of Life
- 1. Control Requirements Fuel temperature (Doppler), % 1.11 1.06 Moderator temperature (1), % 0.06 1.06 Redistribution, % 0.85 1.00 Rod Insertion Allowance, % 0.50 0.50 2. Total Control, % 2.52 3.62 3. Estimated Rod Cluster Control Assembly Worth (52 Rods)
- a. All assemblies inserted, % 10.18 8.40 b. All but one (highest worth) assemblies inserted, % 6.43 7.61 4. Estimated Rod Cluster Control Assembly credit with 10 % adjustment to accommodate uncertainties (3b - 10 percent), % 5.79 6.85 5. Shutdown Margin Available (4-2), % 3.27 3.23 (2) (1) Includes void effects.
(2) The design basis minimum shutdown is 1.77%.
4.3-54 Reformatted Per RN 03-017 TABLE 4.3-4
SUMMARY
OF SPENT FUEL RACK CRITICALITY BIASES AND STATISTICAL UNCERTAINTIES REGION 1 REGION 2 CALCULATIONAL & METHODOLOGY BIASES Methodology (Benchmark) Bias +0.0009 +0.0009 Pool Water Temperature Variation +0.0016 +0.0021 Axial Burnup Distribution N/A +0.0062 TOTAL Bias
+0.0025 +0.0092 TOLERANCES & UNCERTAINTIES UO 2 Enrichment Tolerance +0.0016 +0.0032 UO 2 Density Tolerance +0.0022 +0.0030 Storage Cell ID Tolerance N/A +0.0011 1 Boral Width Tolerance +0.0006 +0.0008 Boral Minimum B10 Content +0.0019 +0.0028 Depletion Uncertainty N/A +0.0146 Calculational Uncertainty (95/95) +0.0016 +0.0012
Methodology Bias Uncertainty
+0.0011 +0.0011 TOTAL Uncertainty (statistical)
+0.0086 +0.0157 TOTAL OF BIASES & UNCERTAINTIES +0.0111 +0.0249
1 As the box I.D. and cell pitch are interre lated a change in one of these parameters will necessarily the other param eter. It is assumed that both the cell pitch and box I.D. are manufactured at their minimum.
2 This assumes the maximum possible change in the water gap, predicated on the box I.D. and cell pitch being manufactured at their greatest tolerance in opposition to each other (i.e. maximum box I.D. and minimum cell pitch).
RN 03-017 4.3-55 Reformatted Per Amendment 02-01 TABLE 4.3-5 AXIAL STABILITY INDEX PRESSURIZER WATER REACTOR CORE WITH A 12 FOOT HEIGHT Stability Index (hr
-1) Burnup (MWD/T) F Z C B (ppm) Exp Calc 1550 1.34 1065 -0.041 -0.032 7700 1.27 700 -0.014 -0.006 Difference +0.027 +0.026 02-01 4.3-56 Reformatted Per Amendment 02-01 TABLE 4.3-6 TYPICAL NEUTRON FLUX LEVELS (n/cm 2 - sec) AT FULL POWER
E s1.0 Mev 5.53 Kev E 1.0 Mev 0.625 ev E 5.53 Kev E 0.625 ev (nv)0 Core Center
6.73 x 10 13 1.18 x 10 14 8.92 x 10 13 3.14 x 10 13 Core Outer Radius
at Midheight
3.39 x 10 13 6.03 x 10 13 4.85 x 10 13 9.03 x 10 12 Core Top, on Axis
1.60 x 10 13 2.54 x 10 13 2.20 x 10 13 1.71 x 10 13 Core Bottom, on Axis
2.48 x 10 13 4.13 x 10 13 3.67 x 10 13 1.53 x 10 13 Pressure Vessel Inner Wall, Azimuthal Peak, Core Midheight 2.90 x 10 10 6.03 x 10 10 6.32 x 10 10 8.78 x 10 10 02-01 4.3-57 Reformatted Per Amendment 02-01 TABLE 4.3-7 COMPARISON OF MEASURED AND CALCULATED DOPPLER DEFECTS Plant Fuel Type Core Burnup (MWD/MTU) Measured (pcm) (1) Calculated (pcm) 1 Air - Filled 1800 1700 1710 2 Air - Filled 7700 1300 1440 3 Air and Helium - Filled 8460 1200 1210 (1) pcm = 10 5 x ln k 1/k 2 02-01 4.3-58 Reformatted Per Amendment 02-01 TABLE 4.3-8 SAXTON CORE II ISOTOPICS ROD MY, AXIAL ZONE 6 Atom Ratio Measured (1) 2 Precision (%)
LEOPARD Calculation U-234/U 4.65 x 10
-5 29 4.60 x 10-5 U-235/U 5.74 x 10
-3 0.9 5.73 x 10-3 U-236/U 3.55 x 10
-4 5.6 3.74 x 10-4 U-238/U 0.993816 0.01 0.99385 Pu-238/Pu 1.32 x 10
-3 2.3 1.222 x 10
-3 Pu-239/Pu 0.73971 0.03 0.74497 Pu-240/Pu 0.19302 0.2 0.19102 Pu-241/Pu 6.014 x 10
-2 0.3 5.74 x 10-2 Pu-242/Pu 5.81 x 10-3 0.9 5.38 x 10-3 Pu/U (2) 5.938 x 10
-2 0.7 5.970 x 10
-2 Np-237/U-238 1.14 x 10-4 15 0.86 x 10-4 Am-241/Pu-239 1.23 x 10-2 15 1.08 x 10-2 Cm-242/Pu-239 1.05 x 10
-4 10 1.11 x 10-4 Cm-244/Pu-239 1.09 x 10
-4 20 0.98 x 10-4 (1) Reported in Reference [31]
(2) Weight ratio
02-01 02-01 4.3-59 Reformatted Per Amendment 02-01 TABLE 4.3-9 CRITICAL BORON CONCENT RATIONS, HZP, BOL Plant Type Measured Calculated 2-Loop, 121 Assemblies 10-foot Core
1583 1589 2-Loop, 121 Assemblies
12-foot Core
1625 1624 2-Loop, 121 Assemblies
12-foot Core
1517 1517 2-Loop, 157 Assemblies
12-foot Core 1169 1161 4.3-60 Reformatted Per Amendment 02-01 TABLE 4.3-10 COMPARISON OF MEASURED AND CALCULATED ROD WORTH 2-Loop Plant, 121 Assemblies, 10-foot Core Measured (pcm) Calculated (pcm) Group B Group A Shutdown Group
1885 1530 3050 1893 1649 2917 ESADA Critical (1), 0.69" Pitch, 2 w/o Pu0 2 8% Pu-240 9 Control Rods 6.21" Rod Separation
2.07" Rod Separation
1.38" Rod Separation 2250 4220 4100 2250 4160 4019 (1) Reported in Reference [32]
02-01 4.3-61 Reformatted Per Amendment 02-01 TABLE 4.3-11 COMPARISON OF MEASURED AND CALCULATED MODERATOR COEFFICIENTS AT HZP, BOL Plant Type/ Control Bank Configuration Measured iso (1) (pcm/ F) Calculated iso (pcm/ F) 3-Loop, 157 Assemblies, 12-foot Core
D at 160 steps -0.50 -0.50 D in, C at 190 steps -3.01 -2.75 D in, C at 28 steps -7.67 -7.02
B, C, and D in
-5.16 -4.45 2-Loop, 121 Assemblies, 12-foot Core
D at 180 steps +0.85 +1.02 D in, C at 180 steps -2.40 -1.90 C and D in, B at 165 steps -4.40 -5.58
B, C, and D in, A at 174 steps -8.70 -8.12
(1) Isothermal coefficients, which incl ude the Doppler effect in the fuel.
F T/k k ln 10 1 2 5 in 02-01
- p II..L"a,*DC.***.-**11 D*-
11/1/-----
---
1/1/------
..------------
1/1/1---------------------------------...----------------------------
0---
1/1/1------
---
1/1/1-
D no au..-un AMENDMENTS AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Typical Reload Core Fuel Loading Arrangement Figure 4.3-1*p II..L" a ,*DC.***.-**11 D*-
11/1/-----
---
1/1/------
..------------
1/1/1---------------------------------...----------------------------
0---
1/1/1------
---
1/1/1-
D no au..-un AMENDMENTS AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Typical Reload Core Fuel Loading Arrangement Figure 4.3-1 9 0 8-It-;::)..,-!i: (-8;::)-!i: O'l-::0.::: O'l::0.::: C.?6-12 LI.J enLI.J 0....0 0 5---....en 0-16 en c::: LI.J"-I:::z::: LI.J c;J It//;::)-20 u..:::z::: Pu239 u..u..7 0 0 z: z: 3 0 0-21t-I l-I-(,.)S;::)/c en 0 2 z: 0:::-28 0 Q.(,.)I
.....
-32."".."".--
---0-36 0 8 12 16 20 21t 283236 itO BURNUP, (GWD/MUT)SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Production and Consumption of Higher Isotopes Figure 4.3-2 9 0 8-It-;::)..,-!i: (-8;::)-!i: O'l-::0.::: O'l::0.::: C.?6-12 LI.J enLI.J 0....0 0 5---....en 0-16 en c::: LI.J"-I:::z::: LI.J c;J It//;::)-20 u..:::z::: Pu239 u..u..7 0 0 z: z: 3 0 0-21t-I l-I-(,.)S;::)/c en 0 2 z: 0:::-28 0 Q.(,.)I
.....
-32."".."".--
---0-36 0 8 12 16 20 21t 283236 itO BURNUP, (GWD/MUT)SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Production and Consumption of Higher Isotopes Figure 4.3-2 18000 15000'200D 1000 1000 3000 o c+----r----r-----r----,..---..,---....-iIl!
100 uoo 2000 I t i I I CYCU BIJRNUP lMWD/M'TlJ)
SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Boron Concentration Versus Burnup For Transition and Equilibrium Cores Figure 4.3-3 AMENDMENT 96-02 JULy 1996 fsar433.n:(359.jr14325}
JRl432S@jrl4325.
Wed Jul310:.$8:54 EDT 1996 18000 15000'200D 1000 1000 3000 o c+----r----r-----r----,..---..,---....-iIl!
100 uoo 2000 I t i I I CYCU BIJRNUP lMWD/M'TlJ)
SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Boron Concentration Versus Burnup For Transition and Equilibrium Cores Figure 4.3-3 AMENDMENT 96-02 JULy 1996 fsar433.n:(359.jr14325}
JRl432S@jrl4325.
Wed Jul310:.$8:54 EDT 1996 10 IF'8A ASSEIIBL T 121 InA lSSDeLT LECE!I): o FUEL ROD O GUIDE TUBE OR INSTRUMENT AT 1011 TUBE'.'InA ROD i.:<"'_1.DJ(335.jd432S)
JRl432S@jrl432S.FriJun28I1,39"'EDTI996 104 JF'IA NOTE: ALL FIGURES ARE lOP VIEI SOUTH CAROLINA ELECTRIC S.GAS CO.VIRGIL C.SUMMER NUCLEAR ST AlION Integral Fuel Burnable Absorber Rod Arrongement Within on Assembly (Sheet 1 of 2)figure 4.3-4 AMENDMENT 96-02 JULy 1996 10 IF'8A ASSEIIBL T 121 InA lSSDeLT LECE!I): o FUEL ROD O GUIDE TUBE OR INSTRUMENT AT 1011 TUBE'.'InA ROD i.:<"'_1.DJ(335.jd432S)
JRl432S@jrl432S.FriJun28I1,39"'EDTI996 104 JF'IA NOTE: ALL FIGURES ARE lOP VIEI SOUTH CAROLINA ELECTRIC S.GAS CO.VIRGIL C.SUMMER NUCLEAR ST AlION Integral Fuel Burnable Absorber Rod Arrongement Within on Assembly (Sheet 1 of 2)figure 4.3-4 AMENDMENT 96-02 JULy 1996
,:iII**.IeIIII e*I*1 J r-II*1I e:***\II:II:*I lIII1:*e', Ie: I:.!;I::.:III1:II.IIII , eo I i i;e, ,: III!, I:**;I I.,II I., I*I i I I 1 I!, j I.I I I I: , ,:: 16 InA ASSD8LY ,::: I.I i I i.!;!: I II III I I rJf*: e.:I°°e I!*III.1 j I I i I'..0!i.iI, e;ff..'I.!I Ie,IIII I: I!i..;i..;I**e'l;.: I,I i:!it i ei cit I!I..I:III I Ie., I: '.: I ,.1:., Ie!.: I Ie i I I I ItII***1 lei*.i°e:.1 iif-II I I:*I I I!I1III I I I..InA ASSDeLY LEGEND: o FUEL ItOO[J GUIDE TUIE OR ItlSTRUlEMUTION YUlE=e: IFBA ROD-fsr434_2.m(336.jr14325)
JR14325@jd432S.FriJun2811:43:59EDT1996 32 1F8A ASSEMBLY'14 I FlA*ASSEWLT MOTE: ALL F'ICURES ARE TOP VIEW SOUTH CAROLINA ELECTRICGAS CO.VIRGIL C.SUMMER NUCLEAR STATION Integral Fuel Burnable Absorber Rod Arrongement Within on Assembly<Sheet 2 of 2)Figure4*.3-4 AMENDMENT 96-02 JULY 1996 ,:iII**.IeIIII e*I*1 J r-II*1I e:***\II:II:*I lIII1:*e', Ie: I:.!;I::.:III1:II.IIII , eo I i i;e, ,: III!, I:**;I I.,II I., I*I i I I 1 I!, j I.I I I I: , ,:: 16 InA ASSD8LY ,::: I.I i I i.!;!: I II III I I rJf*: e.: I°°e I!*III.1 j I I i I'..0!i.iI, e;ff..'I.!I Ie,IIII I: I!i..;i..;I**e'l;.: I,I i:!it i ei cit I!I..I:III I Ie., I: '.: I ,.1:., Ie!.: I Ie i I I I ItII***1 lei*.i°e:.1 iif-II I I:*I I I!I1III I I I..InA ASSDeLY LEGEND: o FUEL ItOO[J GUIDE TUIE OR ItlSTRUlEMUTION YUlE=e: IFBA ROD-fsr434_2.m(336.jr14325)
JR14325@jd432S.FriJun2811:43:59EDT1996 32 1F8A ASSEMBLY'14 I FlA*ASSEWLT MOTE: ALL F'ICURES ARE TOP VIEW SOUTH CAROLINA ELECTRICGAS CO.VIRGIL C.SUMMER NUCLEAR STATION Integral Fuel Burnable Absorber Rod Arrongement Within on Assembly<Sheet 2 of 2)Figure4*.3-4 AMENDMENT 96-02 JULY 1996 II*IIMLKMa,*DC.t.I a***7*iii'**111 11 D'iii*....10..10 10*IS 10 10 10 10 10 10....10 10............................10 10....*101010 10 10 10*as 10..10....AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Typical Burnable Absorber Loading Pattern Figure 43*5 II*II M L KM a,*DC.t.I a***7*iii'**111 11 D'iii*....10..10 10*IS 10 10 10 10 10 10....10 10............................10 10....*10 10 10 10 10 10*as 10..10....AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Typical Burnable Absorber Loading Pattern Figure 43*5 I-1.031 1.2111 1.12'1.011 1.111 1.254 0**1.0.3'3-1.21.1.0521.2381.144 1.2'11.1811.057 0.3341.1211.232 1.138 1.281 1.171 1.232 0.801 1.011 1.1.3 1.210 1.181 1.252 1.014 G 1.111 1.21.1.1'731.2531.133 0**'7'1.254 1.1'71 1.238 1.01'7 0**'0 CALCULATED to*1.428 0.81'1.081 0.813 0**'0.313 0.331 I AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of life.Unrodded Core.Hot full Power.No Xenon (Transition Core)figure 4.3-6 I-1.031 1.2111 1.12'1.011 1.111 1.254 0**1.0.3'3-1.21.1.052 1.238 1.144 1.2'1 1.181 1.057 0.334 1.121 1.232 1.138 1.281 1.171 1.232 0.801 1.011 1.1.3 1.210 1.181 1.252 1.014 G 1.111 1.21.1.1'73 1.253 1.133 0**'7'1.254 1.1'71 1.238 1.01'7 0**'0 CALCULATED to*1.428 0.81'1.081 0.813 0**'0.313 0.331 I AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of life.Unrodded Core.Hot full Power.No Xenon (Transition Core)figure 4.3-6 I-1.031 1.227 1.133 LOll 1.1111 1.250 0.1" 0."00-1.227 1.OS1 1.2.3 1.1.1.211 1.1112 1.0411 0.3lIO 1.133 1.2410 1.1.3 1.213 1.1117 1.222 0.1112 1.011 1.141 1.212 1.1110 1.2.'1.011 0.31'1.1.1.212 1.181 1.2.1 1.1a 0.02 1.250 1.1" 1.22.1.013 0.413 CALQlLATID..1.413 0.1" 1.080 0**,1 0.311 0.*00 0.3" I AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of life, Unrodded Core, Hot Full Power, Equilibrium Xenon (Transition Core)Figure 4.3-7 I-1.031 1.227 1.133 LOll 1.1111 1.250 0.1" 0."00-1.227 1.OS1 1.2.3 1.1.1.211 1.1112 1.0411 0.3lIO 1.133 1.2410 1.1.3 1.213 1.1117 1.222 0.1112 1.011 1.141 1.212 1.1110 1.2.'1.011 0.31'1.1.1.212 1.181 1.2.1 1.1a 0.02 1.250 1.1" 1.22.1.013 0.413 CALQlLATID..1.413 0.1" 1.080 0**,1 0.311 0.*00 0.3" I AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of life, Unrodded Core, Hot Full Power, Equilibrium Xenon (Transition Core)Figure 4.3-7 I-1.037 1.227 1.130 1.081 1.11'1.241 0.*1.0.3131.227 100M 1.231 1.143 1.214 1.182 1.048 0.337 1.130 1.233 1.111 1.2.0 1.172 1.231 0.*,4 1.081 1.142 1.2'1 1.'8'1.2.1 1.021 0.311 1.'11 1.217'.174 1.280 1.13'0.417 1.241 1.1.'1.231 1.023 0.41'CALQlLAT!D r...*1.430 0.11.1.01'7 0.817 0.382 0.313 0.340 I AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of Life, Group D at HFP Insertion limit, Hot Full Power, Equilibrium Xenon (Transition Core)Figure 4.3-8 I-1.037 1.227 1.130 1.081 1.11'1.241 0.*1.0.3131.227 100M 1.231 1.143 1.214 1.182 1.048 0.337 1.130 1.233 1.111 1.2.0 1.172 1.231 0.*,4 1.081 1.142 1.2'1 1.'8'1.2.1 1.021 0.311 1.'11 1.217'.174 1.280 1.13'0.417 1.241 1.1.'1.231 1.023 0.41'CALQlLAT!D r...*1.430 0.11.1.01'7 0.817 0.382 0.313 0.340 I AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of Life, Group D at HFP Insertion limit, Hot Full Power, Equilibrium Xenon (Transition Core)Figure 4.3-8 I-1.041 1.213 1.103 1.047 1.131 1.231 0."1O*.ae r--1.253 l.OU 1.271 1.101 1.214 1.103 0.174 0.344 1.103 1.273 1.124 1.321 1.14'1.23.0**24 1.047 1.101 1.324 1.1111.3081.013 0.3" 1.1.1.211 1.141 1.308 1.110 0.117 1.2311.1081.231 1.01..0.11'CALCULATED to*1.3110*.,1 0.1'0 o.al 0.31'O.<iOII 0.347 I AMENDMENT6 AUGUST,1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Middle of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon (Transition Core)Figure 4.3-9 I-1.041 1.213 1.103 1.047 1.131 1.231 0."1O*.ae r--1.253 l.OU 1.271 1.101 1.214 1.103 0.174 0.344 1.103 1.273 1.124 1.321 1.14'1.23.0**24 1.047 1.101 1.324 1.111 1.308 1.013 0.3" 1.1.1.211 1.141 1.308 1.110 0.117 1.231 1.108 1.231 1.01..0.11'CALCULATED to*1.3110*.,1 0.1'0 o.al 0.31'O.<iOII 0.347 I AMENDMENT6 AUGUST,1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Middle of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon (Transition Core)Figure 4.3-9 I-1.(M1 1.212 1.017 1.03.1.11'1.230 0.'41 0.4"--.1.212 1.011 1.224 1.01.1.2....1.112 1.031 0.41'1.017 1.223 1.010 1.210 1.122 1.227 0**11 1.03'1.01.1.2411.1121.247 1.011 0.412 1.11'1.241 1.122 1.24'7 1.014 0.117 1.230 1.114 1.221 1.011 0.11'7 CALCULATED t.,.1.338 0.'41 1.031 0....0.4U[J AlSe-LV POWER..0.411 0.420 I AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near End of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon (Transition Core)Figure 4.3*10 I-1.(M1 1.212 1.017 1.03.1.11'1.230 0.'41 0.4"--.1.212 1.011 1.224 1.01.1.2....1.112 1.031 0.41'1.017 1.223 1.010 1.210 1.122 1.227 0**11 1.03'1.01.1.241 1.112 1.247 1.011 0.412 1.11'1.241 1.122 1.24'7 1.014 0.117 1.230 1.114 1.221 1.011 0.11'7 CALCULATED t.,.1.338 0.'41 1.031 0....0.4U[J AlSe-LV POWER..0.411 0.420 I AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near End of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon (Transition Core)Figure 4.3*10 0.811 1.131 1.019 1.293 1.196 1.384 0.855 0.304 1.131 BBBBBEJ 0.285 1.019 88BBB88B8B 0.318 1.I%BB8BB 1.384 888B 0.855 81 0.5610.3 17 1 CALCULATEDP""=15560.3040.282 soum CAROLINA ELEClRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of Life, Unrodded Core, Hot Full Power, No Xenon (Equilibrium Core)Figure 4.3-11 AMENDMENT 96-02 JULY 1996 0.811 1.131 1.019 1.293 1.196 1.384 0.855 0.304 1.131 BBBBBEJ 0.285 1.019 88BBB88B8B 0.318 1.I%BB8BB 1.384 888B 0.855 81 0.5610.3 17 1 CALCULATEDP""=15560.3040.282 soum CAROLINA ELEClRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of Life, Unrodded Core, Hot Full Power, No Xenon (Equilibrium Core)Figure 4.3-11 AMENDMENT 96-02 JULY 1996 0.837 1.155 1.034 1.296 1.191 1372 0315 1155 BBBBBB 0.293B8BBB 0570 1.296 BBBBB 0.324 1191BBBEJB L372BBBB 0.859 81 0567 11 0.3 23 1 CALCULATED F""=1530 0.315 0.291 SOum: CAROLINA ELEC1RlC&GAS CO.VIRGil..C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon (Equilibrium Core)Figure 4.3-12 AMENDMENT 96-02 JULY 1996 0.837 1.155 1.034 1.296 1.191 1372 0315 1155 BBBBBB 0.293B8BBB 0570 1.296 BBBBB 0.324 1191BBBEJB L372BBBB 0.859 81 0567 11 0.3 23 1 CALCULATED F""=1530 0.315 0.291 SOum: CAROLINA ELEC1RlC&GAS CO.VIRGil..C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon (Equilibrium Core)Figure 4.3-12 AMENDMENT 96-02 JULY 1996 0.837 1.157 1.0331.2981.1931.369 0.836 0.310 1.157 BBBBBB1033 BBBBB 0573 1298 BBBBB 0.327 1.193 BBBI 0.764 18 1.369BBB8 0.836 Ell 0570 II 0.3 27 1 CALCULAlED F AH=1.484 0.310 0.289 soum: CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of Life, Group D at HFP Insertion Limit, Hot Full Power, Equilibrium Xenon (Equilibrium Core)Figure 4.3-13 AMENDMENT 96-02 JULY 1996 0.837 1.157 1.0331.2981.1931.369 0.836 0.310 1.157 BBBBBB1033 BBBBB 0573 1298 BBBBB 0.327 1.193 BBBI 0.764 18 1.369BBB8 0.836 Ell 0570 II 0.3 27 1 CALCULAlED F AH=1.484 0.310 0.289 soum: CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Beginning of Life, Group D at HFP Insertion Limit, Hot Full Power, Equilibrium Xenon (Equilibrium Core)Figure 4.3-13 AMENDMENT 96-02 JULY 1996 0.933 1.288 1.087 1.355 1.149 1.326 0.843 0.346 1.288 BB8BBB 0.318 1.087 BBBB8 0580 1.355 888BB 0.341 U49BBBBEJ 1.326B8BB 0.843 BI 0.5790.341 I CALCULATlIDF",,=
1.439 0.346 0.316 soum CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Middle of llie, Unrodded Core, Hot Full Power, Equilibrium Xenon (Equilibrium Core)Figure 4.3-14 AMENDMENT 96-02 JULY 1996 0.933 1.288 1.087 1.355 1.149 1.326 0.843 0.346 1.288 BB8BBB 0.318 1.087 BBBB8 0580 1.355 888BB 0.341 U49BBBBEJ 1.326B8BB 0.843 BI 0.5790.341 I CALCULATlIDF",,=
1.439 0.346 0.316 soum CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near Middle of llie, Unrodded Core, Hot Full Power, Equilibrium Xenon (Equilibrium Core)Figure 4.3-14 AMENDMENT 96-02 JULY 1996 L 0.955 1.269 1.063 1.277 1.107 1.312 0.919 0.448 1.269 BBBBBB 0400 1.063 BBBBB 0.653 1.277 BBBBB 0.403 l.lWBBBBEJ 1.312BBBEJ 0.919 BI 0.652 1 0.403 1 CALCULATEDF",,=
1.373 0.448 0.400 SOurn: CAROLINA ELEC1RIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near End of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon (Equilibrium Core)Figure 4.3-15 AMENDMENT 96-02 JULY 1996 L 0.955 1.269 1.063 1.277 1.107 1.312 0.919 0.448 1.269 BBBBBB 0400 1.063 BBBBB 0.653 1.277 BBBBB 0.403 l.lWBBBBEJ 1.312BBBEJ 0.919 BI 0.652 1 0.403 1 CALCULATEDF",,=
1.373 0.448 0.400 SOurn: CAROLINA ELEC1RIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Power Density Distribution Near End of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon (Equilibrium Core)Figure 4.3-15 AMENDMENT 96-02 JULY 1996 1.2.1 1.2410 l.i**I.HI 1.2'" ,.*, I.U'1**2 1.101 1**a I.U'1**'I.UI l.na 1.211 1.212 1.211.1.22'1.'" 1.200 t***t.21t t....'.lI..t.nl t.170 1.117'.210 t.11I t**'t.1OI 1.2tO t.'80 1.211'.2.'t.,.,.lt2 t.1I2 ,.4102'.11't....'**t t.llt ,.." t.lla 1.12.t.21.t.2"'.lMO t....t.1IO'**21'.aI t.lIt t.182t."t." t."1**17 t...t.'"'.2" I.HI I.'" 1." 1**2.t.I2'I.C'" 1.17.1." t.COI 1.'" 1.1n t....1."'.610 t..." 1.:an 1.2M ,.11IO t....1**t.t**2.t**21 I**" 1**" I.'" 1....t....1.171 l.al2 1.2'70 1.2" I.In 1.214 1.172 1...n 1**'1.1" 1**U 1.11't.lla I.C" 1.'" 1.102 I.COI l.na 1._1.2" 1.11'1.17'1.211 I.U'1.<<Ie 1._1." I.C" 1....1.1" t."22 1.102 t**'1.C02 1.111 1.112 t.27't.....1.171 1.111 ,...,, 1...t1....,..*,.C11 t.4102 1.11'1.120 I.HI 1."1 I.NT I.HI 1.DO l.co.-1.10<1 1.11I 1.ft2 t**, 1.IIC I.C20 1.100 1.280 I.COI 1.....1.110 1.2" l.n1 1.112 1.271 1.110 ,.*,1 1."" 1.101 I**" 1.110 ,....t**', 1....1._1....1.270 1.102 1.2" 1.121 ,..., 1."'" 1**,1 1.*>>'.CI2 t.cn l.cn t.CII t.NT 1.107 I.UI 1.211 1....1." 1.107 I.COI 1.M2 I.U2 1." t....1.17.I.C.l.nl t.....t.C12 I.H2 1.180 1.217 1.2M 1.....1.21" 1.2M I.M2 1.27..1....1**n l.cal 1.180 1._1.271 1.182 1.'" 1.1'1 1....1.172 ,....1.172'.110 t.1I2 l.co.-1."'.1" 1.20'1.2" 1."" 1.,lIl 1..., l.lMO 1.220 1.12'1.2171.117 1....1.121 t.1a I.HI 1.2111 1.101 t.20" 1.,.a 1.2" 1.111 1.t.I" 1.221 1.2.2 1.H4 I.Ha 1.21'1.2" 1.111 1.210 1.211 1.27'1.2..1.212 1.2.'t.112 AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Rodwise Power Distribution in a Typical Assembly (G-10)Near BOL, HFP, Equilibrium Xenon, Unrodded Core Figure 4.3-16 1.2.1 1.2410 l.i**I.HI 1.2'" ,.*, I.U'1**2 1.101 1**a I.U'1**'I.UI l.na 1.211 1.212 1.211.1.22'1.'" 1.200 t***t.21t t....'.lI..t.nl t.170 1.117'.210 t.11I t**'t.1OI 1.2tO t.'80 1.211'.2.'t.,.,.lt2 t.1I2 ,.4102'.11't....'**t t.llt ,.." t.lla 1.12.t.21.t.2"'.lMO t....t.1IO'**21'.aI t.lIt t.182 t." t." t."1**17 t...t.'"'.2" I.HI I.'" 1." 1**2.t.I2'I.C'" 1.17.1." t.COI 1.'" 1.1n t....1."'.610 t..." 1.:an 1.2M ,.11IO t....1**t.t**2.t**21 I**" 1**" I.'" 1....t....1.171 l.al2 1.2'70 1.2" I.In 1.214 1.172 1...n 1**'1.1" 1**U 1.11't.lla I.C" 1.'" 1.102 I.COI l.na 1._1.2" 1.11'1.17'1.211 I.U'1.<<Ie 1._1." I.C" 1....1.1" t."22 1.102 t**'1.C02 1.111 1.112 t.27't.....1.171 1.111 ,...,, 1...t 1....,..*,.C11 t.4102 1.11'1.120 I.HI 1."1 I.NT I.HI 1.DO l.co.-1.10<1 1.11I 1.ft2 t**, 1.IIC I.C20 1.100 1.280 I.COI 1.....1.110 1.2" l.n1 1.112 1.271 1.110 ,.*,1 1."" 1.101 I**" 1.110 ,....t**', 1....1._1....1.270 1.102 1.2" 1.121 ,..., 1."'" 1**,1 1.*>>'.CI2 t.cn l.cn t.CII t.NT 1.107 I.UI 1.211 1....1." 1.107 I.COI 1.M2 I.U2 1." t....1.17.I.C.l.nl t.....t.C12 I.H2 1.180 1.217 1.2M 1.....1.21" 1.2M I.M2 1.27..1....1**n l.cal 1.180 1._1.271 1.182 1.'" 1.1'1 1....1.172 ,....1.172'.110 t.1I2 l.co.-1."'.1" 1.20'1.2" 1."" 1.,lIl 1..., l.lMO 1.220 1.12'1.217 1.117 1....1.121 t.1a I.HI 1.2111 1.101 t.20" 1.,.a 1.2" 1.111 1.t.I" 1.221 1.2.2 1.H4 I.Ha 1.21'1.2" 1.111 1.210 1.211 1.27'1.2..1.212 1.2.'t.112 AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Rodwise Power Distribution in a Typical Assembly (G-10)Near BOL, HFP, Equilibrium Xenon, Unrodded Core Figure 4.3-16
.t.ta..'1.'1.t.tM'1.'1.2 t.tI.'1.'1"'1.'110'I.'"'1.'1" t.'" 1.'10 t.'111 1.tH 1.tn 1.'111 1.tl1'1.1.'1.111 t.1'" t..., 1.tl1 t.U" t.JII 1.IM'1.112 1.1_1.21'1.1.1.111'1.'1.'1.'1"'1.'1.0 1.,n 1.'.t.n.1.'110 t.a**t.HI'I.....'I.....t.HI t.HI t.an 1." t.JOoI 1.'1"'1.'110'1.'1" 1.'1" 1.'.I."'I." t.nl 1.n.t.n.1.an t.Jl2 1."'I." 1.IU 1.1M.1.'In'I.'.t.U1 t.1a 1.11I'I." t.1n t.1SI 1....1."'I." 1.....'1.101'1.'11 1.101 t.1eO 1.1.I.'" 1.'1" 1.Ut1'I....t.1eO'1**2'I.*'t.IM t.1SI t.1Ot'1.101'1.1110 t.'.I.'" t.U'1." t.no'I." 1***1.HI'1**'1 t**'1**'1.170'1.100 t.HI 1.UI 1.1'1 1.1S2 t.'M t.ta 1.U't.:an'.nl t.J"'1.111 t**H t."'1.1'0'I."'1.100 t**1.111.211 1.U'1.'.'t.tH t.Ut t.HI 1.n.t**2'I." 1.100 t.IO'1.IM 1.11I'1.1"'1.101'1.'1'" 1.tn 1.a.1'1**1 1.n.t.IU t.aH 1.114'I."'I."'I."'1.100 t**1.11I t.nl t.JU t.t**.t.tl2 1.Jt.'I.*"'1.1412'I.**1.1" 1**"'1.1" t**'t....1.21.t.1OO 1.211'1**10 t.211 1.1"'1.'1.1.'112 t**,0'1.1'2'1.110 1**'t.all t.1IM'I.'" 1.eo.t.1Ot'1.112 t.101 t.'.1.100 t.....t.no 1.all t**1 1.no 1.an 1....1....t.HI'1.10'1 1.111 1.1Oa'I.*'t**" t:'10'1.'20 1.tlt t.220'.171 t.2n'I....t." t.n.'.n.'1**'1 t**1.100 1.211'1.101'1.'1"'1.'102 1.'.'t.tn'1.112'1.1.'t.8a'1.:11I 1.HI 1.HI 1.an 1.110'1.101 t.102'1.1" t.on'1.'111'1.'1**'1.'1" 1.101'1.1'1't.Ut'1.102 t.UI'1.1'10 t.a.'1.111'1.'1"'I.'.'I.'.'t.ttaO 1.011'1.01'1 1.'101'1.'121 t.'" t.tlt t.,n'1.'1" t.'" t.,n t.tlO t.,.,'I.I.'" t.'10 1.'" t.tI.AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Rodwise Power Distribution in a Typical Assembly (G-10)Near EOL, HFP., Equilibrium Xenon, Unrodded Core Figure 4.3-17.t.ta..'1.'1.t.tM'1.'1.2 t.tI.'1.'1"'1.'110'I.'"'1.'1" t.'" 1.'10 t.'111 1.tH 1.tn 1.'111 1.tl1'1.1.'1.111 t.1'" t..., 1.tl1 t.U" t.JII 1.IM'1.112 1.1_1.21'1.1.1.111'1.'1.'1.'1"'1.'1.0 1.,n 1.'.t.n.1.'110 t.a**t.HI'I.....'I.....t.HI t.HI t.an 1." t.JOoI 1.'1"'1.'110'1.'1" 1.'1" 1.'.I."'I." t.nl 1.n.t.n.1.an t.Jl2 1."'I." 1.IU 1.1M.1.'In'I.'.t.U1 t.1a 1.11I'I." t.1n t.1SI 1....1."'I." 1.....'1.101'1.'11 1.101 t.1eO 1.1.I.'" 1.'1" 1.Ut1'I....t.1eO'1**2'I.*'t.IM t.1SI t.1Ot'1.101'1.1110 t.'.I.'" t.U'1." t.no'I." 1***1.HI'1**'1 t**'1**'1.170'1.100 t.HI 1.UI 1.1'1 1.1S2 t.'M t.ta 1.U't.:an'.nl t.J"'1.111 t**H t."'1.1'0'I."'1.100 t**1.111.211 1.U'1.'.'t.tH t.Ut t.HI 1.n.t**2'I." 1.100 t.IO'1.IM 1.11I'1.1"'1.101'1.'1'" 1.tn 1.a.1'1**1 1.n.t.IU t.aH 1.114'I."'I."'I."'1.100 t**1.11I t.nl t.JU t.t**.t.tl2 1.Jt.'I.*"'1.1412'I.**1.1" 1**"'1.1" t**'t....1.21.t.1OO 1.211'1**10 t.211 1.1"'1.'1.1.'112 t**,0'1.1'2'1.110 1**'t.all t.1IM'I.'" 1.eo.t.1Ot'1.112 t.101 t.'.1.100 t.....t.no 1.all t**1 1.no 1.an 1....1....t.HI'1.10'1 1.111 1.1Oa'I.*'t**" t:'10'1.'20 1.tlt t.220'.171 t.2n'I....t." t.n.'.n.'1**'1 t**1.100 1.211'1.101'1.'1"'1.'102 1.'.'t.tn'1.112'1.1.'t.8a'1.:11I 1.HI 1.HI 1.an 1.110'1.101 t.102'1.1" t.on'1.'111'1.'1**'1.'1" 1.101'1.1'1't.Ut'1.102 t.UI'1.1'10 t.a.'1.111'1.'1"'I.'.'I.'.'t.ttaO 1.011'1.01'1 1.'101'1.'121 t.'" t.tlt t.,n'1.'1" t.'" t.,n t.tlO t.,.,'I.I.'" t.'10 1.'" t.tI.AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Rodwise Power Distribution in a Typical Assembly (G-10)Near EOL, HFP., Equilibrium Xenon, Unrodded Core Figure 4.3-17 1.50 TRANSIEO\J AO=4.66 US, I E Cl.70..J I Q.lSO 0.25 100 90 80 70 dO so o
'0 20 30 40 PERCE"'T Of ACTIVC CORE HEIGHT FROM BOTTOM SOUTH CAROLINA ELECTRICGAS CO.VIRGIL C.SUMMER NUCLEAR STATION Typicol Axiol Power Shopes Occurring ot Beginning of Life Figure 4.3-18 AMENDMENT 96-02 JULY 1996 isar431Sm(360.jr1432S)JRl432S@ir14325.WodJul311:03:1SEDT1996 1.50 TRANSIEO\J AO=4.66 US, I E Cl.70..J I Q.lSO 0.25 100 90 80 70 dO so o
'0 20 30 40 PERCE"'T Of ACTIVC CORE HEIGHT FROM BOTTOM SOUTH CAROLINA ELECTRICGAS CO.VIRGIL C.SUMMER NUCLEAR STATION Typicol Axiol Power Shopes Occurring ot Beginning of Life Figure 4.3-18 AMENDMENT 96-02 JULY 1996 isar431Sm(360.jr1432S)JRl432S@ir14325.WodJul311:03:1SEDT1996 1.50-y---------
__1.25 1 VANTAGE5/VANTAGE+./AO=-4.11fD.7S;:j I O.so 0.25 100 90 80 70 ISO&0 40 30 10 o-1.-----"""1-----------------------.....1
.20 f>£RCENT OF ACTIVE CORE H£IGHT FROM BO'TTOM SOUTH CAROLINA ELECTRIC 8<GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Typicol Axiol Power Shopes Occurring ot Middle of Life Figure 4.3-19 AMENDMENT 96-02 JULY 1996 fsar.4319.m(361Jr14325)
JR14325@jrl432S.
WedJul311:06:12 EDT 19%1.50-y---------
__1.25 1 VANTAGE5/VANTAGE+./AO=-4.11fD.7S;:j I O.so 0.25 100 90 80 70 ISO&0 40 30 10 o-1.-----"""1-----------------------.....1
.20 f>£RCENT OF ACTIVE CORE H£IGHT FROM BO'TTOM SOUTH CAROLINA ELECTRIC 8<GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Typicol Axiol Power Shopes Occurring ot Middle of Life Figure 4.3-19 AMENDMENT 96-02 JULY 1996 fsar.4319.m(361Jr14325)
JR14325@jrl432S.
WedJul311:06:12 EDT 19%
1..25 0..25 1.,"'....tttOl"l/'I&0--3.61 f E 0.76 i I cue 100 90 70 eo 150 30 20 10 o o__"""__-""I__...J 80 PERCENT OF ACifVE CORE HEIGHT FROM BOTTOW: SOUTH CAROLINA ELECTRIC 8<GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Typical Axial Power Shapes Occurring at End of Life Figure 4.3*20 AMENDMENT 96-02 JULY 1996 r..,.3211.0l(358.jd4325)
JR14325@jd4325.
WedJul310,;'<,09 EDT 1996 1..25 0..25 1.,"'....tttOl"l/'I&0--3.61 f E 0.76 i I cue 100 90 70 eo 150 30 20 10 o o__"""__-""I__...J 80 PERCENT OF ACifVE CORE HEIGHT FROM BOTTOW: SOUTH CAROLINA ELECTRIC 8<GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Typical Axial Power Shapes Occurring at End of Life Figure 4.3*20 AMENDMENT 96-02 JULY 1996 r..,.3211.0l(358.jd4325)
JR14325@jd4325.
WedJul310,;'<,09 EDT 1996 FREQUENCY OF GAPS PER ROD-F 9 AXIAL FREQUENCY DISTRIBUTION OF GAPS-F.)SIZE FREQUENCY DISTRIBUTION OF GAPS-F k SPIKE DUE TO SINGLE GAPS-S g\1-----1--------
7DRAW COMPUTER Y CODE MTG FROM DENSIFICATION MODEL (MAX.GAP SIZE AT GIVEN HEIGHT)MODE I PROBABILITIES OF EXCEEDING GIVEN SPIKE SIZE FOR EACH AXIAL LOCATION CONVOLUTION POWER SP IKE FACTOR-S(Z)MODE 2EXPECTED VALUES SEEN BY INCORE DETECTOR ROD CENSUS AMENDMENT6 AUGUST.1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGILC.SUMMER NUCLEAR STATION Flowchart for Determining Spike Model Figure 4.3-22 FREQUENCY OF GAPS PER ROD-F 9 AXIAL FREQUENCY DISTRIBUTION OF GAPS-F.)SIZE FREQUENCY DISTRIBUTION OF GAPS-F k SPIKE DUE TO SINGLE GAPS-S g\1-----1--------
7DRAW COMPUTER Y CODE MTG FROM DENSIFICATION MODEL (MAX.GAP SIZE AT GIVEN HEIGHT)MODE I PROBABILITIES OF EXCEEDING GIVEN SPIKE SIZE FOR EACH AXIAL LOCATION CONVOLUTION POWER SP IKE FACTOR-S(Z)MODE 2EXPECTED VALUES SEEN BY INCORE DETECTOR ROD CENSUS AMENDMENT6 AUGUST.1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGILC.SUMMER NUCLEAR STATION Flowchart for Determining Spike Model Figure 4.3-22 12.0 10.0 CI-8.0<[<.::l UJ....J c.::l z: en 0 f-UJ 6.0:=>0 UJ::w:: CI-V)f-z: UJ u 1t.0 a:::: UJ CI-2.0 o o 1.0 2.0 GAP SIZE (INCHES)3,0 ILO AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STATION Predicted Power Spike Due to Single Non-Flattened Gap in the Adjacent Fuel Figure 4.3-23 12.0 10.0 CI-8.0<[<.::l UJ....J c.::l z: en 0 f-UJ 6.0:=>0 UJ::w:: CI-V)f-z: UJ u 1t.0 a:::: UJ CI-2.0 o o 1.0 2.0 GAP SIZE (INCHES)3,0 ILO AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STATION Predicted Power Spike Due to Single Non-Flattened Gap in the Adjacent Fuel Figure 4.3-23 0 U"l:::r=t-O>">-I-(I)0 z: UJ N 0 u cr:: I-UJ::::E: 0 UJ 8-c;l_en....l UJ"":1:::z: l-t.)z: z: z:00 f-en 0 Q.....J<<0><1.0<<SOUTH CAROLINA ElECTRiC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION N o (z)s o 8 o N o AMENDMENT6 AUGUST, 1990 Power Spike Factor as a Function of Axial Position Figure 4.3-240 U"l:::r=t-O>">-I-(I)0 z: UJ N 0 u cr:: I-UJ::::E: 0 UJ 8-c;l_en....l UJ"":1:::z: l-t.)z: z: z:00 f-en 0 Q.....J<<0><1.0<<SOUTH CAROLINA ElECTRiC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION N o (z)s o 8 o N o AMENDMENT6 AUGUST, 1990 Power Spike Factor as a Function of Axial Position Figure 4.3-24 I i t o I*COAl HIJGHT.."*to 12 AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Maximum FQ x Power Versus Axial Height During Normal Operation Figure 4.3-25 I i t o I*COAl HIJGHT.."*to 12 AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Maximum FQ x Power Versus Axial Height During Normal Operation Figure 4.3-25 25r-----r-----r-----r------
20 t-----+------I------I------I 0-o 0...-)D 15...-I 0 l!t 0 i I W 10-.-I t-30 o AXIAL FLUX DIFFERENCE (I)30 AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Peak Power Ouring Control Rod Malfunction Overpower Transients Figure 4.3-26 25r-----r-----r-----r------
20 t-----+------I------I------I 0-o 0...-)D 15...-I 0 l!t 0 i I W 10-.-I t-30 o AXIAL FLUX DIFFERENCE (I)30 AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Peak Power Ouring Control Rod Malfunction Overpower Transients Figure 4.3-26 25.----.---------,.----..,.------...
...30 o AXIAL flUX DlmRENCE (I)30 60 AMENDMENT6 AUGUST.1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Peak Power During Boration/Dilution Overpower Transients Figure 4.3-2]25.----.---------,.----..,.------...
...30 o AXIAL flUX DlmRENCE (I)30 60 AMENDMENT6 AUGUST.1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Peak Power During Boration/Dilution Overpower Transients Figure 4.3-2]
0.759 0.792 4.3/1.217 1.224 0.67 0.774 1.255 0.800 1.249 3.41-0.5)1.229 1.229 1.225 I.2-0.37-0.9 1.109 L 077 I.107 1.092-0.2'/I*tt'1c, 1.202 1.223 I.1.256-2.7", 2.7'1: 0.523 1.217 1.221 1.217 0.548 1.203 1.233 1.210 4.6%I.I X 1.01-0.6'1 1.229 1.229 I.189 I.-3.3%-0.70 1.217CALCULATED 1.21 IMEASURED-0.5%DIFFERENCE
- 2.07 LOCATED AT M-8 SOUTH AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Comparison Between Calculated and Measured Relative Fuel Assembly Power Distribution Figure 4.3-28 0.759 0.792 4.3/1.217 1.224 0.67 0.774 1.255 0.800 1.249 3.41-0.5)1.229 1.229 1.225 I.2-0.37-0.9 1.109 L 077 I.107 1.092-0.2'/I*tt'1c, 1.202 1.223 I.1.256-2.7", 2.7'1: 0.523 1.217 1.221 1.217 0.548 1.203 1.233 1.210 4.6%I.I X 1.01-0.6'1 1.229 1.229 I.189 I.-3.3%-0.70 1.217CALCULATED 1.21 IMEASURED-0.5%DIFFERENCE
- 2.07 LOCATED AT M-8 SOUTH AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Comparison Between Calculated and Measured Relative Fuel Assembly Power Distribution Figure 4.3-28 o N o 8 Q..0 I-0 lJ.J I-0:: 0 lJ.J (/)0>z: z>00 z: f'-1;'Z lJ.J l.C)x l-N oCt 0 0 0 z: 0:: lJ.J"'" 00....J:3t z:00 oCt CO CL CO 0 ,.....a..oCt::::E lJ.J (.!)lJ.J<l: 0:: 0 0:: 0 lJ.J 0 I-lJ.J U I-c.o:::t::>-z:<l: Cl<l:-1>'Z....J N=:;,1>'Z u.J lJ.J::::E u M 0:: 0"....":r::: 0 0:: 0<l: 0 U w..<l: U<u.J Q:: 0 0 LO U u...0 I-z: u.J 0 U Q:::::r u.J 0..
lVIXV AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Comparison of Calculated and Measured Axial Shape Figure 4.3-29 o N o 8 Q..0 I-0 lJ.J I-0:: 0 lJ.J (/)0>z: z>0 0 z: f'-1;'Z lJ.J l.C)x l-N oCt 0 0 0 z: 0:: lJ.J"'" 00....J:3t z: 0 0 oCt CO CL CO 0 ,.....a..oCt::::E lJ.J (.!)lJ.J<l: 0:: 0 0:: 0 lJ.J 0 I-lJ.J U I-c.o:::t::>-z:<l: Cl<l:-1>'Z....J N=:;,1>'Z u.J lJ.J::::E u M 0:: 0"....":r::: 0 0:: 0<l: 0 U w..<l: U<u.J Q:: 0 0 LO U u...0 I-z: u.J 0 U Q:::::r u.J 0..
lVIXV AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Comparison of Calculated and Measured Axial Shape Figure 4.3-29 o o*o**o x REACTORI0 REACTOR\l REACTOR2*REACTOR 5*REACTOR 3 F Q 3.0 2.5*ii*x*o*o*2.0 8.xO.,.x 0**Q: o 9**x 1.5***o 00-50-ij5-40-35-30-25-20-15-10-505 10 152025 30 INCORE AXIAL OFFSET (%)AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Measured Values of FQ for Full Power Rod Configurations Figure 4.3-30 o o*o**o x REACTORI0 REACTOR\l REACTOR2*REACTOR 5*REACTOR 3 F Q 3.0 2.5*ii*x*o*o*2.0 8.xO.,.x 0**Q: o 9**x 1.5***o 00-50-ij5-40-35-30-25-20-15-10-505 10 152025 30 INCORE AXIAL OFFSET (%)AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Measured Values of FQ for Full Power Rod Configurations Figure 4.3-30
-1..,..-----------------------.
-1.2 It!-1A I EOL-1.'80L I-1.1-2 i-2.2-2.A-2.'+-..,..-..,..-.,...-...,--.,...-.,.--..,..-.,....-,....-,....----'
100 100 700 100 100'1000 1100 120013001400 1500 ,.00 IFFICTIVI PUIE1.TEMPERATURE Teff<<F)AMENDMENT 6 AUGUST, 1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STATION Doppler Temperature Coefficient at BOl and EOl Versus T eft for a Typical Reload Core Figure 4.3-31-1..,..-----------------------.
-1.2 It!-1A I EOL-1.'80L I-1.1-2 i-2.2-2.A-2.'+-..,..-..,..-.,...-...,--.,...-.,.--..,..-.,....-,....-,....----'
100 100 700 100 100'1000 1100 1200 1300 1400 1500 ,.00 IFFICTIVI PUIE1.TEMPERATURE Teff<<F)AMENDMENT 6 AUGUST, 1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STATION Doppler Temperature Coefficient at BOl and EOl Versus T eft for a Typical Reload Core Figure 4.3-31
-
100 10 10 20 o...,..-----------------------. t------,----,...---"'T"---...,..-----I I-.EOL i BOL I-to I i i-'2 POWER LEVEL o.RCENT OF FULL POWER)AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Doppler Only Power Coefficient Versus Power Level at SOL and EOL for a Typical Reload Core Figure 4.3 100 10 10 20 o...,..-----------------------. t------,----,...---"'T"---...,..-----I I-.EOL i BOL I-to I i i-'2 POWER LEVEL o.RCENT OF FULL POWER)AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Doppler Only Power Coefficient Versus Power Level at SOL and EOL for a Typical Reload Core Figure 4.3-32 O-.t'-------------------.-.
I I-..00 I i i-100 EOL BOL-1200-t-----r----..,..-----.----.,...----f o 20.-0 10 10 POWER LEVEL CPERCENT OF FULL POWER)100 AMENDMENTS AUGUST.1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Doppler Only Power Defect Versus Percent Power SOL and EOl for a Typical Reload Core Figure 4.3-33 O-.t'-------------------.-.
I I-..00 I i i-100 EOL BOL-1200-t-----r----..,..-----.----.,...----f o 20.-0 10 10 POWER LEVEL CPERCENT OF FULL POWER)100 AMENDMENTS AUGUST.1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Doppler Only Power Defect Versus Percent Power SOL and EOl for a Typical Reload Core Figure 4.3-33 10,.....--...,..----,----,..-----..,..---..,....---
5-i-...:lit loll-&oJ-I::0&oJ 000 PPU J 1700 PPM c i 1400 PPU!1100 PPM-5-10 IIlURATOR ttlF£RATlII: (or)AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STATION Moderator Temperature Coefficient BOl, No Rods, for a Typical Reload Core Figure 4.3-34 10,.....--...,..----,----,..-----..,..---..,....---
5-i-...:lit loll-&oJ-I::0&oJ 000 PPU J 1700 PPM c i 1400 PPU!1100 PPM-5-10 IIlURATOR ttlF£RATlII: (or)AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STATION Moderator Temperature Coefficient BOl, No Rods, for a Typical Reload Core Figure 4.3-34 5 0..........
..I-............1\....5""""""\1\:500 PPM , ,0 PPM , ," ,""""-20-25 o 100 200 JOO 400 IIIlERA TOIl 1'tIFtRA ME (or)500 600 AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C SUMMER NUCLEAR STATION Moderator Temperature Coefficient EOl for a Typical Reload Core Figure 4.3-35 5 0..........
..I-............1\....5""""""\1\:500 PPM , ,0 PPM , ," ,""""-20-25 o 100 200 JOO 400 IIIlERA TOIl 1'tIFtRA ME (or)500 600 AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C SUMMER NUCLEAR STATION Moderator Temperature Coefficient EOl for a Typical Reload Core Figure 4.3-35
- --,---...,.--....,r----.....---...,..---
AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C SUMMER NUClEAR STATION Moderator Temperature Coefficient as a Function of Boron Concentration BOl, No Rods, for a Typical Reload Core Figure 4.3-36**--,---...,.--....,r----.....---...,..---
AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C SUMMER NUClEAR STATION Moderator Temperature Coefficient as a Function of Boron Concentration BOl, No Rods, for a Typical Reload Core Figure 4.3-36 0...,..-----------------------_
li:-10 I 5/VA;'1TAGE+
Co,)!-20 , ,..'" C.', Ic: I-30-40
()3000 6000 9000 12000 1S000 1BOOO f=4337.m(362.jr14325)JRl4325@jrl4325.WedJul311,w,JSEDTI996 C'fC'J BURN'JP CMWO!MTUl SOUTH CAROLINA ELECTRIC S.GAS CO.VIRGIL C.SUMMER NUCLEAR ST AllaN Hot F'ullPower Temperature Coefficient at Critical Boron Concentration versus Burnup F'igure 4.3-37 AMENDMENT 96-02 JULY 1996 0...,..-----------------------_
li:-10 I 5/VA;'1TAGE+
Co,)!-20 , ,..'" C.', Ic: I-30-40
()3000 6000 9000 12000 1S000 1BOOO f=4337.m(362.jr14325)JRl4325@jrl4325.WedJul311,w,JSEDTI996 C'fC'J BURN'JP CMWO!MTUl SOUTH CAROLINA ELECTRIC S.GAS CO.VIRGIL C.SUMMER NUCLEAR ST AllaN Hot F'ullPower Temperature Coefficient at Critical Boron Concentration versus Burnup F'igure 4.3-37 AMENDMENT 96-02 JULY 1996
-10.,---------------------..
__"",..------------"""1 BOl-tIS I I-20 I..21 I EOl-30!e-31 o 20 10 10 100 POWER LEVEL PERCENT OF FULL POWER)*AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGllC.SUMMER NUCLEAR STATION Total Power Coefficient Versus Percent Power for BOl and EOl on a Typical Reload Core Figure 4.3-38-10.,---------------------..
__"",..------------"""1 BOl-tIS I I-20 I..21 I EOl-30!e-31 o 20 10 10 100 POWER LEVEL PERCENT OF FULL POWER)*AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGllC.SUMMER NUCLEAR STATION Total Power Coefficient Versus Percent Power for BOl and EOl on a Typical Reload Core Figure 4.3-38
-100 I-1000 I SOL I-'100-2000-2100 EOL-3000-t-----.,...----...----_
.....--_------..
o 20*eo 10 POWER LIVEl..(PERCENT OF FULL POWER)100 AMENDMENT 6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Total Power Defect BOl, EOl, for a Typical Reload Core Figure 4.3-39
-100 I-1000 I SOL I-'100-2000-2100 EOL-3000-t-----.,...----...----_
.....--_------..
o 20*eo 10 POWER LIVEl..(PERCENT OF FULL POWER)100 AMENDMENT 6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUCLEAR STATION Total Power Defect BOl, EOl, for a Typical Reload Core Figure 4.3-39 23456789 10 II 12 13 14 15 270 0'"/"'"\"\....A...J r-....:..;18/,"\/B'"\/, ,B...I c-...J'-.../MV"""\(0G)/'\/'"\/" 0'-...J/,/,/\/'"\,D...I ,-C..,J\..c..J\..0...1V"""\I(B).C0/'"\/",/'"\'-.A...J0"A...J A---..-..,;;...l iCc)/",/",/, ,B...I\.B...I'-c./&"---'o.:...l V'"\/"'"\r--....A..,J
\....0..J A B C D E F G J K L M N p R FUNCTION CONTROL BANK D CONTROL BANK C CONTROL BANK B CONTROL BANK A SHUTDOWN BANK S6 SHUTDOWN BANK SA NUMBER OF CLUSTERS 8 8 8 8 8 8 AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Rod Cluster Control Assembly Pattern Figure 4.3*40 23456789 10 II 12 13 14 15 270 0'"/"'"\"\....A...J r-....:..;18/,"\/B'"\/, ,B...I c-...J'-.../MV"""\(0G)/'\/'"\/" 0'-...J/,/,/\/'"\,D...I ,-C..,J\..c..J\..0...1V"""\I(B).C0/'"\/",/'"\'-.A...J0"A...J A---..-..,;;...l iCc)/",/",/, ,B...I\.B...I'-c./&"---'o.:...l V'"\/"'"\r--....A..,J
\....0..J A B C D E F G J K L M N p R FUNCTION CONTROL BANK D CONTROL BANK C CONTROL BANK B CONTROL BANK A SHUTDOWN BANK S6 SHUTDOWN BANK SA NUMBER OF CLUSTERS 8 8 8 8 8 8 AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Rod Cluster Control Assembly Pattern Figure 4.3*40 I I i 8 II: 20 10 o-rc.......---r-----,r-----.,..
......--..,....-...l..---I o 10 tOO tlO 200 210 STEPS wrTHDRAWN AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Accidental Simultaneous Withdrawal of Two Control Banks EOL, HZP Banks D and B Moving in the Same Plane Figure 4.3-41 I I i 8 II: 20 10 o-rc.......---r-----,r-----.,..
......--..,....-...l..---I o 10 tOO tlO 200 210 STEPS wrTHDRAWN AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Accidental Simultaneous Withdrawal of Two Control Banks EOL, HZP Banks D and B Moving in the Same Plane Figure 4.3-41 U I U u I i 1.1 I Iu I I u I I 1.4 I I u I I JaSKIlt1r LZ I L1 I I.U U U U U!'H JIR:II.,QSE raax:s)AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Rod Position Versus Time on Reactor Trip Figure 4.3-42 U I U u I i 1.1 I Iu I I u I I 1.4 I I u I I JaSKIlt1r LZ I L1 I I.U U U U U!'H JIR:II.,QSE raax:s)AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Rod Position Versus Time on Reactor Trip Figure 4.3-42 1J u u 1.4 L2 u u.,....----------------_
u 1.7 u u 102 U 1.1 I I I AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STATION Normalized RCCA Reactivity Worth Versus Percent Insertion Figure 4.3-43 1J u u 1.4 L2 u u.,....----------------_
u 1.7 u u 102 U 1.1 I I I AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGil C.SUMMER NUClEAR STATION Normalized RCCA Reactivity Worth Versus Percent Insertion Figure 4.3-43
,:/-'I ACCEPTABLE,.'/" , 1..1'"..I'".I II'I"..I'" 1/NOT ACCEPTABLE I IT II I./II I V 1./1..1'" 5 45 40 35 30 25 20 15 10-:::>:?E Ci 3: (!J-a.:::>z a::::>en w (!J a:<<:::r: ()en o--'enw en en<<o 2.0 2.5 3.0 3.5 4.0 4.5 5.0 MAXIMUM NOMINAL U-235 ENRICHMENT (WIG)Notes 1.Fuel assemblies with enrichments less than 2.0 W/O must meet the burn-up requirements of 2.0 W/O assemblies.
2.Use of the following polynomial fit is acceptable where E=Enrichment W/O: Assembly Discharge burmlp=0.1246 E 3-1.91 E 2+20.9205 E-30.2482 RN 03-017 December 2003 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Required Fuel Assembly Burn-up as a Function of Initial Enrichment to permit Storage in Region 2 Racks Figure 4.3-44 ,:/-'I ACCEPTABLE,.'/" , 1..1'"..I'".I II'I"..I'" 1/NOT ACCEPTABLE I IT II I./II I V 1./1..1'" 5 45 40 35 30 25 20 15 10-:::>:?E Ci 3: (!J-a.:::>z a::::>en w (!J a:<<:::r: ()en o--'enw en en<<o 2.0 2.5 3.0 3.5 4.0 4.5 5.0 MAXIMUM NOMINAL U-235 ENRICHMENT (WIG)Notes 1.Fuel assemblies with enrichments less than 2.0 W/O must meet the burn-up requirements of 2.0 W/O assemblies.
2.Use of the following polynomial fit is acceptable where E=Enrichment W/O: Assembly Discharge burmlp=0.1246 E 3-1.91 E 2+20.9205 E-30.2482 RN 03-017 December 2003 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Required Fuel Assembly Burn-up as a Function of Initial Enrichment to permit Storage in Region 2 Racks Figure 4.3-44
Figure 4.3-45 Deleted per RN 03-017, December 2003 RN 03-017 AMENDMENT6 AUGUST, 1990 o 00 N o ('t)I-:::t-00 00,I>-'I--::>-'0-::x::: CO_<<I-U)X L.lJ o:z:_0-;;;-00:::::>a::: 0 L.lJ::x::: Q..-L.lJ0:::::>u o lVIXV SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Axial Offset Versus Time PWR Core with a 12-ft.Height and 121 Assemblies Figure 4.3-46 AMENDMENT6 AUGUST, 1990 o 00 N o ('t)I-:::t-00 00,I>-'I--::>-'0-::x::: CO_<<I-U)X L.lJ o:z:_0-;;;-00:::::>a::: 0 L.lJ::x::: Q..-L.lJ0:::::>u o lVIXV SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Axial Offset Versus Time PWR Core with a 12-ft.Height and 121 Assemblies Figure 4.3-46
- t-(,0I 0 W (,0 U , U 0
- : 1.0 ,!.O;N:!.O J', a::.fJj;;:t-::t:;.N;;:t-I:::t" UJ:z:<.DU U"'-..Cl:: 00!.L.0 0 , 1.0" ('t")...J x\.,<31: w<<=>N Cl:::z: ('t")0->-t-OO 3:...J N-Cl:: co UJ<>:;;:t-l-t-;)N u..en<enCl:::::::>0::I:..-" (,0.,...." N<00******;;:t-***0 0 00 (,0;;:t-N 0 N::::I"\0 00 0 N::::I" (,0---I I I I I%(n
-l 1111 INVMaVnO AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION XV Xenon Test Thermocouple Response Quadrant Tilt Difference Versus Time Figure 4.3-47;;:t-(,0I 0 W (,0 U , U 0::: 1.0 ,!.O;N:!.O J', a::.fJj;;:t-::t:;.N;;:t-I:::t" UJ:z:<.DU U"'-..Cl:: 00!.L.0 0 , 1.0" ('t")...J x\.,<31: w<<=>N Cl:::z: ('t")0->-t-OO 3:...J N-Cl:: co UJ<>:;;:t-l-t-;)N u..en<enCl:::::::>0::I:..-" (,0.,...." N<00******;;:t-***0 0 00 (,0;;:t-N 0 N::::I"\0 00 0 N::::I" (,0---I I I I I%(n
-l 1111 INVMaVnO AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION XV Xenon Test Thermocouple Response Quadrant Tilt Difference Versus Time Figure 4.3-47 (wJd)
M31ddOa lVM831NI000000 0 0000 0 000 0 0000:::t N 0 00 (J:):::t N 00000NNN-----00 (J:):::t N 0Iiiiii I I i IIi 0 0 o C1>o 00 0 a::: I.LJ (J:)0---'....J z:::>a 0I-!.C'l<<0 en z I-w:z: a..a:z::::::-, I.Ll a I-(.)u<I: 0 a::: en:::r I.LJ::.:::z: 0-:f w z a..I-a ex:>::::: z-0 w I-...J U:::::<I: a-....J IX:z: IX=::>0 I-a w u:z: IX a.....J ('t)a a x<<:(u ex:>w u 0<J 0 N o o(d%/wJd)1M3 1 J I M31ddOa o o (J:)I o o L{')I o o:::r I o o ('t)I o o N I o o-I o AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Calculated and Measured Doppler Defect and Coefficients at BOL, Two-Loop Plant, 121 Assemblies, 12-ft.Core Figure 4.3-48 (wJd)
M31ddOa lVM831NI 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0:::t N 0 00 (J:):::t N 0 0 0 0 0 N N N-----00 (J:):::t N 0Iiiiii I I i IIi 0 0 o C1>o 00 0 a::: I.LJ (J:)0---'....J z:::>a 0I-!.C'l<<0 en z I-w:z: a..a:z::::::-, I.Ll a I-(.)u<I: 0 a::: en:::r I.LJ::.:::z: 0-:f w z a..I-a ex:>::::: z-0 w I-...J U:::::<I: a-....J IX:z: IX=::>0 I-a w u:z: IX a.....J ('t)a a x<<:(u ex:>w u 0<J 0 N o o(d%/wJd)1M3 1 J I M31ddOa o o (J:)I o o L{')I o o:::r I o o ('t)I o o N I o o-I o AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Calculated and Measured Doppler Defect and Coefficients at BOL, Two-Loop Plant, 121 Assemblies, 12-ft.Core Figure 4.3-48 8:::r 8 0 00 0 0 0 to 0 0
0:::r 0 0 0 N 0:::;:)0 Q..:::;:)8 z: CII::: 00:::;:)QQ 0 0 0 to 0 8:::r 8 0 N 0 0 AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGllC SUMMER NUCLEAR STATION Comparison of Calculated and Measured Boron Concentration for Two-loop Plant, 121 Assemblies, 12-ft.Core Figure 4.3-49 8:::r 8 0 00 0 0 0 to 0 0
0:::r 0 0 0 N 0:::;:)0 Q..:::;:)8 z: CII::: 00:::;:)QQ 0 0 0 to 0 8:::r 8 0 N 0 0 AMENDMENT6 AUGUST,1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGllC SUMMER NUCLEAR STATION Comparison of Calculated and Measured Boron Concentration for Two-loop Plant, 121 Assemblies, 12-ft.Core Figure 4.3-49 8 (\')Hdd8 00 a a N I'--a a::t-LD::::>a f-a::::e: LD-::t-o.Q..a::::>a z: to Cl::: (\')::::>00 o aN 8 to o AMENDMENT6 AUGUST, 1990 SOUTH CAROliNA ELECTRIC&GAS CO.VIRGIL C SUMMER NUCLEAR STATION Comparison of Calculated and Measured C B in Two-loop Plant, 121 Assemblies, 12*ft.Core Figure 4.3*50 8 (\')Hdd8 00 a a N I'--a a::t-LD::::>a f-a::::e: LD-::t-o.Q..a::::>a z: to Cl::: (\')::::>00 o aN 8 to o AMENDMENT6 AUGUST, 1990 SOUTH CAROliNA ELECTRIC&GAS CO.VIRGIL C SUMMER NUCLEAR STATION Comparison of Calculated and Measured C B in Two-loop Plant, 121 Assemblies, 12*ft.Core Figure 4.3*50 88 o o o en 8 00 o o CD 8 LO 8:::T 8 (Y)o o N 8 o 0 0 0:::T 0 0
0 (Y)0 0 0 N 0 0 0§0 8 en 8::>00_0 00 0 f'-..0-::>0 z: 0 oc:: 0::>CD 00 0 0 0 LO 0 0
0:::T 0 0 0 (Y)0 0 0 N 0 0 0 o AMENDMENT6 AUGUST,1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Comparison of Calculated and Measured C B in Three-Loop Plant, 157 Assemblies, 12*ft.Core Figure 4.3-51 8 8 o o o en 8 00 o o CD 8 LO 8:::T 8 (Y)o o N 8 o 0 0 0:::T 0 0 0 (Y)0 0 0 N 0 0 0§0 8 en 8::>00_0 00 0 f'-..0-::>0 z: 0 oc:: 0::>CD 00 0 0 0 LO 0 0 0:::T 0 0 0 (Y)0 0 0 N 0 0 0 o AMENDMENT6 AUGUST,1990 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Comparison of Calculated and Measured C B in Three-Loop Plant, 157 Assemblies, 12*ft.Core Figure 4.3-51
0.07 r'--------------------------
- 3200 96.0 96.5 96.5 97.9 97.0 95.0 97.0 PERCENT TH EO R ET I CAL DENSITY 2800.A 2qOO o 6.1200 1600 2000 TEMPERATURE (Oc)LEGEND: PERCENT THEORETICAL DEN SITY o GYLLANDER 96.ij 0 HOWARD&GULVIN*ASAMOTO 95.0*LUCKS&DEEM o GODFREY 93.ij<>DAN I EL ET AL 6.STORA 92.2.FEITH*A.D.BUSH 9ij.ij 0 VOGT ET AL*ASAMOTO 91.0 6.N1 SH IJ IMA V KRUGER 93.7" WHEELER&AI NSCOUGH SEE REFERENCES FOR THERMAL-HYDRAULIC SECTION 800 qOO*.*
A\J\J v*6.6.o 0.01 0.06 I-0 0 I E ()--..0.05..--I--::.:::.O.Oq:>--I-0::::)c:z;0 o 0.03...J<0::: Ul:t: I-0.02<V1-0;:l:lc ,...:3:V1;:l:l Co 3:,...3:-m Z;:l:l>>cm ("'1("'1 ,...-i m;:l:l;:l:l("'l V1 120-i>>_V1 0("'1 z9 ("'I o-i.,:r., l1)l1).,:4.3 l1)OJ....("'1 o 0 VlQ..on-i!:t.:r<l1);:+: 0,<., 0.....n" Cl1)0:J OJ'$.:n IQ C., l1).e:o:.:..\AI 0.07 3200 96.0 96.5 96.5 97.9 97.0 95.0 97.0 PERCENT TH EO R ET I CAL DENSITY 2800 2qOO o 1200 1600 2000 TEMPERATURE (OC)LEGEND: PERCENT THEORETICAL DEN SITY o GYLLANDER 96.ij 0 HOWARD&GULVIN*ASAMOTO 95.0*LUCKS&DEEM o GODFREY 93.ij<>DAN I EL ET AL 6.STORA 92.2.FEITH*A.D.BUSH 9ij.ij 0 VOGT ET AL*ASAMOTO 91.0 6.N 1 SH IJ IMA\l KRUGER 93.7" WHEELER&AI NSCOUGH SEE REFERENCES FOR THERMAL-HYDRAULIC SECTION 800 qOO o 0.01 0.060 0 I E ()--..0.05 I--::.:::O.Oq:>-I-0::::)c:z;0 0.03 0...J<0::: Ul:c I-0.02<V1-0::l:lc ,...:3:V1::l:l Co 3:,...3:z m>>::l:lm Z,...cm ("'1("'1 ,...-i>>("'1::l:l12O>>>>-iV1 0("'1 z9 ("'I o-i.,:r iD III n.,.-+3 III OJ Q..-'-+("'1 o 0 VlQ.. o n-i!:t.:r<III_.., III 0.-+.....n" C 0'":J 0 VI OJ_*.-+.-+OJ'$.
r\WWT I..-rT" I 1\\\*-Iol'I C\-"lI......-VI**Iol'I JI-**J
- ..1ol'I-..0\*..,.,-..-\.">.a:: ""'**.8"C: III.........-'"-o*JI"'Cl.:j-.,....1ol'I_C:C:
1".C\.-C.....-""'..*-.*
- ...*...:**\*'\....r\c.-.-c c.c c.....ei C f"'l.C C N.C c.C.c c-*-8*-C 0\*C-N t: c I CIQ a::*::: c.....i=-=-e....)<c:i:=...J a.r..l-e.c loW\C:.*cu-l-e-!on a::*u c c...,....u c-..c-*..., C a:: Q".C C M...,*a:: C c..-\C C-C"-N w-*::::: 11:1&U cr-c-*c AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Measured Versus Predicted Critical Heat Flux*WRB-2 Correlation Figure 4.4-6 r\WWT I..-rT" I 1\\\*-Iol'I C\-"lI......-VI**Iol'I JI-**J
- ..1ol'I-..0\*..,.,-..-\.">.a:: ""'**.8"C: III.........-'"-o*JI"'Cl.:j-.,....1ol'I_C:C:
1".C\.-C.....-""'..*-.*
- ...*...:**\*'\....r\c.-.-c c.c c.....ei C f"'l.C C N.C c.C.c c-*-8*-C 0\*C-N t: c I CIQ a::*::: c.....i=-=-e....)<c:i:=...J a.r..l-e.c loW\C:.*cu-l-e-!on a::*u c c...,....u c-..c-*..., C a:: Q".C C M...,*a:: C c..-\C C-C"-N w-*::::: 11:1&U cr-c-*c AMENDMENT6 AUGUST, 1990 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Measured Versus Predicted Critical Heat Flux*WRB-2 Correlation Figure 4.4-6
.._.0.12 0.10-0.08'-LEGENO: o 26" SPACING I-PHASE o 26" SPACING 2-PHASE o o I 1.0 I 2.0 I 3.0II IJ.O 5.0 G De Re='T (10-5)I I 7.0 8.0 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION TOC versus Reynolds Number for 26" Grid Spacing Figure 4.4-8
.._.0.12 0.10-0.08'-LEGENO: o 26" SPACING I-PHASE o 26" SPACING 2-PHASE o o I 1.0 I 2.0 I 3.0II IJ.O 5.0 G De Re='T (10-5)I I 7.0 8.0 SOUTH CAROLINA ElECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION TOC versus Reynolds Number for 26" Grid Spacing Figure 4.4-8 I ct-I.1631.000-KEY: r-!:J.h/!:J.h G/G Q B wr-;:l.'968r::lFOR RADIAL POWER DISTRIBUTION NEAR BEGINNING OF LIFE.HOT FULL POWER.EQUILIBRIUM XENON N CALCULATED F!:J.H=1.33 1.00 II.142 QQQQQ G DDDDDD DDDD DDD r-::lDQr::lQQQQ B I I I 1.094 1.001 I 1.204 1.000 I.109 1.001 I 1.209 1.001 1.179 1.002 I 1.050 1.001 0.803 0.998 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Radial Flow and Enthalpy.Distribution at 4 ft.Elevation Figure 4.4-9I ct-I.1631.000-KEY: r-!:J.h/!:J.h G/G Q B wr-;:l.'968r::lFOR RADIAL POWER DISTRIBUTION NEAR BEGINNING OF LIFE.HOT FULL POWER.EQUILIBRIUM XENON N CALCULATED F!:J.H=1.33 1.00 II.142 QQQQQ G DDDDDD DDDD DDD r-::lDQr::lQQQQ B I I I 1.094 1.001 I 1.204 1.000 I.109 1.001 I 1.209 1.001 1.179 1.002 I 1.050 1.001 0.803 0.998 SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Radial Flow and Enthalpy.Distribution at 4 ft.Elevation Figure 4.4-9 CL I CL-1.164 0.991-I-KE Y: f-GIG FOR RADIAL POWER DISTRIBUTION NEAR BEGINNING OF LIFE.HOT FULL POWER.EQUILIBRIUI1 XENON CALCULA TED F=1.33 r::l 0Q G QL.:J r::lQH L:J Qo 8 DDDDDD DDDD DDDQ G r-::lL:Jt:j QR L:J Q EJ I I I I 1.092 0.997 I 1.206 0.987 I.109 0.99, 1.211 0.986'.182 0.988 1.050 0.998 0.800 1.019 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Radial Flow and Enthalpy Distribution at 8 ft.Elevation Figure 4.4*10 CL I CL-1.164 0.991-I-KE Y: f-GIG FOR RADIAL POWER DISTRIBUTION NEAR BEGINNING OF LIFE.HOT FULL POWER.EQUILIBRIUI1 XENON CALCULA TED F=1.33 r::l 0Q G QL.:J r::lQH L:J Qo 8 DDDDDD DDDD DDDQ G r-::lL:Jt:j QR L:J Q EJ I I I I 1.092 0.997 I 1.206 0.987 I.109 0.99, 1.211 0.986'.182 0.988 1.050 0.998 0.800 1.019 SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Radial Flow and Enthalpy Distribution at 8 ft.Elevation Figure 4.4*10 KEY: I-t::.h 16': GIG QFOR RADIAL POWER DISTRIBUTION NEAR BEGINNING OF LIFE.HOT FULL POWER.EQUILIBRIUM XENON CALCULATED=1.33Qr::lI:.:lI-QQQDDDDDD DDDD DDDI.146 0.9921::1.I I III.i oS o 990 I 1.209 0.968 I.I 10 0.994 I.OSO 0.998 1.093 0.99S I.214 0.988 0.797 1.01 I 8.20:+0.90t 1...--+_-'
SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Radial Flow and Enthalpy Distribution at 12 ft.Elevation Core Exit Figure 4.4-11 KEY: I-t::.h 16': GIG QFOR RADIAL POWER DISTRIBUTION NEAR BEGINNING OF LIFE.HOT FULL POWER.EQUILIBRIUM XENON CALCULATED=1.33Qr::lI:.:lI-QQQDDDDDD DDDD DDDI.146 0.9921::1.I I III.i oS o 990 I 1.209 0.968 I.I 10 0.994 I.OSO 0.998 1.093 0.99S I.214 0.988 0.797 1.01 I 8.20:+0.90t 1...--+_-'
SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Normalized Radial Flow and Enthalpy Distribution at 12 ft.Elevation Core Exit Figure 4.4-11 z: o(.)<t Q:: u..o o>-REGION I NO BUBBLE DETACHMENT LOCAL BO III NG".-,a....J<<I:z: w:z: o>-a..0::..J:::><<I<<I-(I):z: W en..J"'"<<:5 is CO W REGION II BUBBLE DETACHMENT W 0:::z::::>0<<I-0::<<W 0:: a..:::>W:::<:1-0::
W<<:::>I-en I-<<o en 0::-..J W::::><Q..C>':::>:::<:_C>'W..J WI-BULK BOILING VOID FRACTION PREDICTED FROM THERMODYNAMIC QUALITY WITH NO SLI P o THERMODYNAMIC QUALITY, X=H-H SAT I Hg-H SAT SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Void Fraction versus Thermodynamic Quality H-HsAT/Hg-HsAT Figure 4.4-12 z: o(.)<t Q:: u..o o>-REGION I NO BUBBLE DETACHMENT LOCAL BO III NG".-,a....J<<I:z: w:z: o>-a..0::..J:::><<I<<I-(I):z: W en..J"'"<<:5 is CO W REGION II BUBBLE DETACHMENT W 0:::z::::>0<<I-0::<<W 0:: a..:::>W:::<:1-0::
W<<:::>I-en I-<<o en 0::-..J W::::><Q..C>':::>:::<:_C>'W..J WI-BULK BOILING VOID FRACTION PREDICTED FROM THERMODYNAMIC QUALITY WITH NO SLI P o THERMODYNAMIC QUALITY, X=H-H SAT I Hg-H SAT SOUTH CAROLINA elECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Void Fraction versus Thermodynamic Quality H-HsAT/Hg-HsAT Figure 4.4-12 234 5 6789 10 II 270 0 12 13 14 15 T 0 TO 0 0 TO TO TO TO TO 0 0 0 T 0 T T T 0TT 0 T 0 T 0 0 TO TO TO T T T 0 T 0 0 T 0 TO 0 T 0TT TO TO T T 0 0 0 TO TO 0 T T T T 0 0 TO TO 0 0 T TO TO T T T 0 T 0 TO TO 0 T T T DT-THERMOCOUPLE (51)o-MOVABLE IHCORE DETECTOR (50)A B c o E F G H J K L M N p R SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Distribution of Incore Instrumentation Figure 4.4-19 2 3 4 5 6789 10 II 270 0 12 13 14 15 T 0 TO 0 0 TO TO TO TO TO 0 0 0 T 0 T T T 0TT 0 T 0 T 0 0 TO TO TO T T T 0 T 0 0 T 0 TO 0 T 0TT TO TO T T 0 0 0 TO TO 0 T T T T 0 0 TO TO 0 0 T TO TO T T T 0 T 0 TO TO 0 T T T DT-THERMOCOUPLE (51)o-MOVABLE IHCORE DETECTOR (50)A B c o E F G H J K L M N p R SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Distribution of Incore Instrumentation Figure 4.4-19