ML21286A314
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BFN-26 3.2-1 3.2 FUEL MECHANICAL DESIGN The fuel assembly is comprised of the fuel bundle, channel, and channel fastener.
The fuel bundle is comprised of fuel rods, water rods (water channels), spacers, upper and lower tie plates, springs, and fittings.
Fuel licensing acceptance criteria for General Electric (GE) fuel designs are specified in GESTAR II (General Electric Standard Application For Reactor Fuel)
(Reference 1). Amendment 22 of GESTAR II established an approved set of licensing acceptance criteria for which fuel design compliance constitutes USNRC acceptance and approval without specific USNRC review. Current GE fuel designs that have received specific USNRC review and approval, or that have been shown to meet the approved fuel licensing acceptance criteria are documented in Reference 2 (General Electric Fuel Bundle Designs). GE designs documented in Reference 2 and approved for use in Browns Ferry reload cores include the GE13 and GE14 fuel product lines. The initial core 7x7 and 8x8 designs, the reload core unpressurized 8x8R design, and the pressurized P8x8R, BP8x8R, GE9B, GE11 designs that are currently in spent fuel storage will not be reinserted in any future reload cores.
Fuel licensing acceptance criteria for AREVA fuel designs are specified in ANF-89-98(P)(A) Revision 1 and Supplement 1. Generic Mechanical Design Criteria for BWR Designs (Reference 4). This document contains an approved set of licensing acceptance criteria for which fuel design compliance constitutes USNRC acceptance and approval without specific USNRC review.
For GE fuel, the generic information contained in GESTAR II is supplemented by plant cycle-unique information and analytical results. This cycle-unique information includes a list of the fuel to be loaded in the core and safety analysis results. This information is documented in a separate plant-unique cycle-dependent submittal for each reload. The format for this Supplemental Reload Licensing Report (SRLR) and a description of the transient and accident methods used are given in the country-specific supplement to GESTAR II (Reference 3).
For AREVA fuel, a reload-specific fuel mechanical design report is prepared to document compliance with Reference 4 Generic Design Criteria. The Generic Design Criteria lists the approved methodology documents. The fuel mechanical design report is referenced in the applicable cycle-specific Safety Analysis Report.
3.2.1 Power Generation Objective The objective of the nuclear fuel is to provide a high integrity assembly of fissionable material which can be arranged in a critical array. The assembly must be capable of efficiently transferring the generated fission heat to the circulating coolant water while maintaining structural integrity and containing the fission products.
BFN-26 3.2-2 3.2.2 Power Generation Design Basis The nuclear fuel shall be designed to assure (in conjunction with the core nuclear characteristics, the core thermal and hydraulic characteristics, the plant equipment characteristics, and the capability of the nuclear instrumentation and reactor protection system) that fuel damage limits will not be exceeded during either planned operation or abnormal operational transients caused by any single equipment malfunction or single operator error.
Limits are established to assure fuel operation remains within design bases. Three types of thermal limits are used. The first is the Minimum Critical Power Ratio Operating Limit. This limit is generated to protect against the phenomena of dryout, where the liquid film on the fuel cladding surface is boiled/stripped away, thereby creating conditions for a rapid rise in cladding temperature due to the elimination of effective boiling heat transfer. The second limit is the Linear heat Generation Rate.
This limit protects the fuel rod thermal/mechanical integrity during normal operation, as well as during anticipated operational occurrences. The third limit is the Maximum Average Planar Linear Heat Generation Rate. This limit protects the fuel from exceeding fuel performance requirements identified by Title 10 Code of Federal Regulations Part 50.46.
Power and Flow dependent multipliers are applied to all three types of thermal limits to protect operation at off-rated conditions. The basic limit types, along with the power/flow multipliers are generated for all fuel types, regardless of fuel vendor or unit.
Power and flow dependent limits (or multipliers) are applied to both the MCPR and thermal-mechanical limits to account for off-rated conditions. The thermal-mechanical limits are protected with off-rated corrections applied directly to the LHGR limits.
3.2.3 Safety Design Bases In meeting the power generation objectives, the nuclear fuel cladding shall be utilized as the initial barrier to the release of fission products. The fission product retention capability of the nuclear fuel shall be substantial during normal modes of reactor operation so that significant amounts of radioactivity are not released from the reactor fuel barrier.
For GE fuel, the detailed fuel thermal-mechanical design bases and limits are provided in GESTAR II. For AREVA fuel, the thermal-mechanical design bases and limits are provided in the AREVA Generic Design Criteria document (Reference 4).
The design bases address each of the fuel system damage, failure, and coolability criteria identified in the Standard Review Plan (NUREG-0800).
BFN-26 3.2-3 3.2.4 Description A core cell is defined as a control rod and the four fuel assemblies which immediately surround it. Each core cell is associated with a 4-lobed fuel support piece. Around the outer edge of the core, certain fuel assemblies are not immediately adjacent to a control rod and are supported by individual peripheral fuel support pieces.
A description of the fuel assembly and various fuel assembly components is provided in the following sub-sections.
3.2.4.1 Fuel Assembly The fuel assembly consists of a fuel bundle and a channel which surrounds it. The fuel bundle contains fuel rods and water rods (or water channel) which are spaced and supported in a square array by upper and lower tieplates, as well as fuel rod spacers. The lower tieplate has a nosepiece which has the function of supporting the fuel assembly in the reactor. The upper tieplate has a handle for transferring the fuel bundle from one location to another. The identifying fuel assembly serial number is engraved on the top of the handle. No two assemblies bear the same serial number. A boss projects from one side of the handle to aid in ensuring proper orientation of the assembly in the core. Finger springs located between the lower tieplate and channel are utilized to control the bypass flow through that flow path.
3.2.4.1.1 Fuel Rods Three types of fuel rods are used in a GE fuel bundle; tie rods, standard rods, and (in some designs) part length rods. The tie rods in each fuel bundle have lower end plugs which thread into the lower tieplate and threaded upper end plugs which extend through the upper tieplate. A nut and locking tab are installed on the upper end plug to hold the fuel bundle together. The tie rods support the weight of the assembly during fuel handling operations. All of the standard rods are full length rods. The part length rods are approximately 2/3 of the length of standard fuel rods.
AREVA fuel assemblies contain two basic rod types: standard rods and part length fuel rods (PLFRs). Tie rods are not necessary because the structural tie between the lower and upper tieplate is provided by the water channel. The PLFRs lengths vary by fuel design, relative to standard fuel rods.
During operation, the GE assembly is supported by the lower tieplate. The end plugs of the standard rods have shanks which fit into bosses in the tieplates. An expansion spring is located over the upper end plug shank of each rod in the bundle to keep the rods seated in the lower tieplate.
BFN-26 3.2-4 For the AREVA design, the lower ends of the fuel rods rest on top of the lower tieplate grid. The lower ends of the fuel rods are laterally restrained by an additional spacer grid located just above the lower tieplate. No expansion springs are necessary on each fuel rod because a single, large reaction spring is used on the water channel to hold the upper tieplate in the latched position.
Each fuel rod contains high density ceramic uranium dioxide fuel pellets stacked within Zircaloy cladding. The fuel rod is evacuated, backfilled with helium, and sealed with end plugs welded into each end. U-235 enrichments may vary from fuel rod to fuel rod within a bundle to reduce local peak-to-average fuel rod power ratios.
Selected fuel rods within each bundle may include small amounts of Gadolinia as a burnable poison.
Adequate free volume is provided within each fuel rod in the form of a pellet-to-cladding gap and a plenum region. A plenum spring, or retainer, is provided in the plenum space to minimize the movement of the column of fuel pellets inside the fuel rod during shipping and handling. For GE fuel product lines through GE13, a hydrogen getter has historically been provided in the plenum space as assurance against chemical attack from inadvertent admission of moisture or hydrogenous impurities into the fuel rod during manufacture. With enhanced hydrogen controls in place, the optional feature of a reactive getter has been removed for the current GE13 and GE 14 product lines. Likewise, hydrogen and moisture controls eliminate the need for a getter in the AREVA fuel rod design.
3.2.4.1.2 Water Rods or Water Channel For the GE fuel designs, water rods are hollow Zircaloy tubes with several holes around the circumference near each end to allow coolant to flow through. One water rod in each bundle axially positions the spacers. The AREVA design instead uses one larger internal water channel that has a square cross-section in contrast to round water rods. The water channel displaces 3x3 array of fuel rods in the fuel assembly interior.
3.2.4.1.3 Fuel Spacer The primary function of the spacer is to provide lateral support and spacing of the fuel rods, with consideration of thermal-hydraulic performance, fretting wear, strength, neutron economy, and producibility.
3.2.4.1.4 Finger Springs Finger Springs are employed to control the bypass flow through the channel-to-lower tieplate flow path.
BFN-26 3.2-5 3.2.4.1.5 Debris Filter Lower Tie Plate Fuel assemblies may include a debris filter as part of the lower tie plate products designed to reduce the probability of foreign material entering the fuel during normal operation. Both GE and AREVA debris filter designs utilize non-line-of-sight coolant flow paths to maximize the ability to stop and retain foreign material. The debris filter does not play a part in the structural performance of the fuel assembly.
Designs are explicitly tested to determine hydraulic performance impacts.
3.2.4.1.6 Channels The BWR Zircaloy fuel channel performs the following functions:
(1)
Forms the fuel bundle flow path outer periphery for bundle coolant flow.
(2)
Provides surfaces for control rod guidance in the reactor core.
(3)
Provides structural stiffness to the fuel bundle during lateral loadings applied from fuel rods through the fuel spacers.
(4)
Minimizes, in conjunction with the finger springs and bundle lower tieplate, coolant bypass flow at the channel/lower tieplate interface.
(5)
Transmits fuel assembly seismic loadings to the top guide and fuel support of the core internal structures.
(6)
Provides a heat sink during loss-of-coolant accident (LOCA).
(7)
Provides a stagnation envelope for incore fuel sipping.
The channel is open at the bottom and makes a sliding seal fit on the lower tieplate surface. The upper end of the fuel assemblies in a four-bundle cell are positioned in the corners of the cell against the top guide beams by the channel fastener springs.
At the top of the channel, two diagonally opposite corners have welded tabs supporting the weight of the channel on the threaded raised posts of the upper tieplate. One of these raised posts has a threaded hole. The channel is attached to the fuel bundle using the threaded channel fastener assembly, which also includes the fuel assembly positioning spring. Channel-to-channel spacing is assured by the fuel bundle spacer buttons located on the upper portion of the channel adjacent to the control rod passage area.
3.2.5 Safety Evaluation The GE thermal-mechanical evaluations performed for the fuel are described in GESTAR II. AREVA thermal-mechanical evaluations are described in the Generic Design Criteria (Reference 4). Areas evaluated include:
(1)
Fuel System Damage -- stress/strain, fatigue, fretting wear, oxidation, hydriding, corrosion, dimensional changes, internal gas pressure, and hydraulic loads.
BFN-26 3.2-6 (2)
Fuel Rod Failure -- hydriding, cladding collapse, fretting wear, overheating of cladding, overheating of pellets, excessive fuel enthalpy, pellet-cladding interaction, bursting, and mechanical fracturing.
(3)
Fuel Coolability -- cladding embrittlement, violent expulsion of fuel, generalized cladding melting, fuel rod ballooning, and structural deformation.
3.2.5.1 Evaluation Methods The GE methods used in performing thermal-mechanical evaluations for the fuel are described in GESTAR II. These evaluations are performed primarily using the NRC-approved GESTR-MECHANICAL fuel rod thermal-mechanical performance model.
The GESTR-MECHANICAL fuel rod performance model performs best estimate coupled thermal and mechanical analyses of a fuel rod experiencing a variable operating history. The model explicitly addresses the effects of:
- Fuel and cladding thermal expansion
- Fuel and cladding creep and plasticity
- Cladding irradiation growth
- Cladding irradiation hardening and thermal annealing of that irradiation hardening
- Fuel irradiation swelling
- Fuel irradiation-induced densification
- Fuel cracking and relocation
- Fuel hot pressing
- Fission gas generation and exposure-enhanced fission gas release including fission product helium release
- Differential axial expansion of the fuel and cladding reflecting axial slip or lockup of the fuel pellets with the cladding
- Fuel phase change volumetric expansion upon melting The GESTR-MECHANICAL material properties and component models represent the latest experimental information available.
The fuel rod cladding stress analyses are performed using a Monte Carlo statistical method in conjunction with distortion energy theory. Fuel cladding plasticity analyses are also performed when required by the loading conditions.
AREVA methods used in performing the thermal-mechanical evaluations for the fuel are described in the Generic Design Criteria (Reference 4) and by the approved topical reports referenced thereof. These analyses are performed primarily using the NRC-approved RODEX2A fuel rod thermal-mechanical performance code. This code combines best estimate and conservative models, coupled with a specific input methodology to produce conservative results. The code contains a number of different models to address the following phenomena:
BFN-26 3.2-7
- Fuel densification
- Fuel gaseous and solid swelling
- A physically-based fission gas release model coupled to the swelling model
- Columnar grain growth
- Instantaneous plastic deformation of the fuel
- Fuel cracking, crack volume closure and creep deformation
- Fuel pore migration
- Cladding anisotropic creep deformation, irradiation growth
- Cladding corrosion and hydriding
- Thermal-hydraulic conditions of the fuel rod sub-channel The code performs interactive calculations on a time incremental basis with conditions updated at each calculated increment. Cladding strain, fuel and cladding temperatures, fission gas release, rod internal pressure are calculated as well as time and burnup-dependent fuel and cladding properties. In addition, the code is used to establish initial conditions for ramping and accident analyses.
Starting with the ATRIUM-10X fuel design, the NRC has approved use of the newer RODEX4 method at BFN. RODEX4 supports thermal mechanical analysis and development of the LHGR limit. RODEX4 is considered a best estimate method, utilizing a Monte-Carlo analysis process. The older ROCEX2A methodology is still retained to provide conservative input to transient and LOCA analyses.
3.2.5.2 Evaluation Results The thermal-mechanical evaluations described above have been completed by GE for all fuel designs included in Reference 2 (General Electric Fuel Bundle Designs).
The evaluations demonstrate that these fuel designs meet the required thermal-mechanical licensing criteria documented in GESTAR II.
Similarly, AREVA thermal-mechanical evaluations are described in the applicable fuel mechanical design report prepared for each cycle. These evaluations document compliance of the fuel design to the NRC approved Generic Design Criteria topical report.
3.2.6 Inspection and Testing The GE fuel quality assurance program is described in GESTAR II. The AREVA fuel quality assurance program is described in FQM U.S. Version, Framatome ANP Fuel Sector Quality Manual. The program covers the quality control areas associated with the manufacture and inspection of new fuel for the areas of:
(1)
Material and component procurement.
BFN-26 3.2-8 (2)
Fabrication and assembly of components and systems.
(3)
Inspection and testing.
(4)
Cleaning, packaging, and shipping.
GE and AREVA also have active programs of interim and post-irradiation surveillance of both lead use assemblies and developmental BWR fuel. The GE program and the inspection techniques used are described in GESTAR II. The AREVA program is described in the referenced reports contained in the applicable fuel mechanical design report.
3.2.7 References
- 1. General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A, (See Appendix N for applicable revision).
- 2. General Electric Fuel Bundle Designs, NEDE-31152P, Rev. 7, June 2000.
- 3. General Electric Standard Application for Reactor Fuel (Supplement for United States), NEDE-24011-P-A-US, (See Appendix N for applicable revision).
- 4. Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A)
Revision 1 and Supplement 1, Advanced Nuclear Fuels Corp., May 1995.